ES-401, REV 10 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 · ES-401, REV 70 T2G7 PWR EXAMINATION...

ES-401 PWR Examination Outline Form ES-401-2

Facility: OCONEE Date of Exam: DECEMBER 2016

—— RD K/ACateqy Pnts SRO-Only_Points

Tier GroupK K K K K K A A A A G* A2 G* Total12 34j6 1234 Total

1. 1 333 33 3 18 3 3 6Emergency &

Abnormal 2. N/A 2 1 N/A ..j_ 9 2 2 4

PlantEvolutions Tier Totals 5 4 5 5 4 4 27 5 5 10

1 3133132221232 3 28 3 2 52. I

Plant 2 101111111111 1 10 0 1 2Systems

Tier Totals 4 3 4 [ 4 3 3 3 [ 3 4 3 4 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7Categories

3 3 2 2 2 2 1 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RDand SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the “Tier Totals” ineach K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A isreplaced by a K/A from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. Thefinal point total for each group and tier may deviate by ±1 from that specified in the table based on NRCrevisions. The final RD exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions thatdo not apply at the facility should be deleted with justification; operationally important, site-specificsystems/evolutions that are not included on the outline should be added. Refer to Section D.1 .b of ES-401for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in thegroup before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall beselected. Use the RD and SRO ratings for the RD and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics

must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicableK/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics’ importanceratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enterthe group and tier totals for each category in the table above; if fuel handling equipment is sampled in acategory other than Category A2 or G* on the SRD-only exam, enter it on the left side of Column A2 forTier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RD and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs,and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

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ES-301 Administrative Topics Outline Form ES-301-1

Facility: Oconee Date of Examination: 12/05/2016 Examination Level: RO X SRO Operating Test Number: 1

Administrative Topic (see Note)

Type Code*

Describe activity to be performed

Conduct of Operations [KA:G2.1.45 (4.3/4.3)] (20 min)

D, R ADM-110, Diverse Verification of Reactor Power (RO & SRO)

Conduct of Operations [KA: G2.1.20 (4.6/4.6)] (20 min)

M, R ADM-111, Perform Manual RCS Leakage Calculation (RO Only)

Equipment Control [KA: G2.2.14 (3.9/4.3)] (30 min)

D, R ADM-205, Determine LTOP Requirements (RO Only)

Radiological Control [KA: G2.3.12 (3.2/3.7)] (20 min)

M, R ADM-305, Determine Posting and Access Requirements of LPI Room Based on Plan View (RO & SRO)

Emergency Plan N/A

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

* Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (≤ 3 for ROs; ≤ 4 for SROs & RO retakes) (N)ew or (M)odified from bank (≥ 1) (P)revious 2 exams (≤ 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1

Facility: Oconee Date of Examination: 12/05/2016

Examination Level: RO SRO X Operating Test Number: 1

Administrative Topic

(see Note)

Type

Code*

Describe activity to be performed


D, R ADM-110, Diverse Verification of Reactor Power (RO & SRO)


D, R ADM-S108, Determine if SRO License Requirements are Met (SRO Only)

Equipment Control [KA: G2.2.14 (3.9/4.3)] (30 min)

N, R ADM-S204, Determine LTOP Requirements and Any Required Actions (SRO Only)

Radiological Control [KA: G2.3.12 (3.2/3.7)] (20 min)

M, R ADM-305, Determine Posting and Access Requirements of LPI Room Based on Plan View (RO & SRO)

Emergency Plan [KA: G2.4.38 (2.4/4.4)] (30 min)

D, R

ADM-S405, Determine Emergency Classification and Protective Action Recommendations (Complete Emergency Notification Form) (SRO Only)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

* Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (≤ 3 for ROs; ≤ 4 for SROs & RO retakes) (N)ew or (M)odified from bank (≥ 1) (P)revious 2 exams (≤ 1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

Facility: ___Oconee____________________ Date of Examination: 12/05/2016

Exam Level: RO X SRO-I SRO-U Operating Test No.: ______1_______

Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U

System / JPM Title Type Code* Safety Function

a. RO-101b, Align Emergency Boration During an ATWS Unit 1 EOP Rule 1 [KA: BW/E13 EA1.1 (3.4/3.2)] (10 min)

A, M, S 1

b. RO-P403, Initiate HPI Forced Cooling EOP Rule 4 (Initiate HPI Forced Cooling) [KA: EPE 074 EA1.08 (4.2/4.2)] (10 min)

D, L, S 4P

c. RO-501a, ES Channels 5 and 6 Recovery Unit 1 EOP Encl. 5.41 (ES Recovery) [KA: SYS103 A4.04 (3.5*/3.5*)] (10 min)

A, D, EN, L, S 5

d. RO-702, Adjust Radiation Monitor Setpoints OP/1-2/A/1104/018 (GWD System) Encl. 4.9 & 4.10 PT/0/A/230/001, (Radiation Monitor Check) [KA: SYS073 A4.01 (3.9/3.9)] (25 min)

D, S 7

e. RO-S404a, Establish EFDW Flow Through Startup Valves EOP Encl. 5.27 (Alternate Methods for Controlling EFDW Flow) [KA: APE054 AA2.04 (4.2/4.3)] (15 min)

D, A, L, S 4S

f. RO-805a, Perform Required Actions For a Turbine Building Flood AP/1/A/1700/010 (Turbine Building Flood) [KA: APE BW/A07 AA1.3 (3.3/3.5)] (15 min)

D, A, L, S 8

g. RO-604, Perform a Manual Start of Keowee Hydro Unit 1 OP/0/A/1106/019 Encl. 4.5 (KHU-1 Manual Startup) [KA: SYS062 A4.07 (3.1*/3.1*)] (15 min)

D, S 6

h. RO-204, Align letdown with 1HP-14 failed in Bleed AP/1/A/170/002 (Excessive RCS Leakage) [KA: SYS002 A2.01 (4.3/4.4)] (15 min)

D, S 2

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. AO-804, AO Actions for Control Room Evacuation AP/1/A/1700/050 Encl. 5.6 (AP/EOP AO Actions for Control Room Evacuation) [KA: APE068 AA1.07 (4.1/4.2)] (15 min)

N, R, E, L 8

j. AO-701, Restoration of ICS Auto Power AP/1/A/1700/023 Encl. 5.2 (Restoration of ICS Auto Power) [KA: BW/A02 AA1.3 (3.4/3.6)] (15 min)

D, A, E 7

k. AO-S403, AO Actions for Supply of Water to SSF AP/0/A/1700/046 Encl. 5.9 (Supply of Water to SSF) [KA: GEN 2.1.20 (4.6/4.6)] (30 min)

N, E, L 4S

* All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

* Type Codes Criteria for RO / SRO-I / SRO-U

A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator

4-6 / 4-6 / 2-3 ≤ 9 / ≤ 8 / ≤ 4 ≥ 1 / ≥ 1 / ≥ 1 ≥1 / ≥1 / ≥ 1 (control room system) ≥ 1 / ≥ 1 / ≥ 1 ≥ 2 / ≥ 2 / ≥ 1 ≤ 3 / ≤ 3 / ≤ 2 (randomly selected) ≥ 1 / ≥ 1 / ≥ 1


Facility: __Oconee_____________________ Date of Examination: _12/05/2016__

Exam Level: RO SRO-I X SRO-U Operating Test No.: ______1_______




A, M, S 1


D, L, S 4P


A, D, EN, L, S 5

d. RO-702, Adjust Radiation Monitor Setpoints OP/1-2/A/1104/018 (GWD System) Encl. 4.9 & 4.10 PT/0/A/230/001, (Radiation Monitor Check) [KA: SYS073 A4.01 (3.9/3.9)] (25 min)

D, S 7

e. RO-S404a, Establish EFDW Flow Through Startup Valves EOP Encl. 5.27 (Alternate Methods for Controlling EFDW Flow) [KA: APE054 AA2.04 (4.2/4.3)] (15 min)

D, A, L, S 4S

f. RO-805a, Perform Required Actions For a Turbine Building Flood AP/1/A/1700/010 (Turbine Building Flood) [KA: APE BW/A07 AA1.3 (3.3/3.5)] (15 min)

D, A, L, S 8

g. RO-604, Perform a Manual Start of Keowee Hydro Unit 1 OP/0/A/1106/019 Encl. 4.5 (KHU-1 Manual Startup) [KA: SYS062 A4.07 (3.1*/3.1*)] (15 min)

D, S 6

h. N/A



N, R, E, L 8


D, A, E 7

k. AO-S403, AO Actions for Supply of Water to SSF AP/0/A/1700/046 Encl. 5.9 (Supply of Water to SSF) [KA: GEN 2.1.20 (4.6/4.6)] (30 min)

N, E, L 4S






Facility: __Oconee_____________________ Date of Examination: _12/05/2016__

Exam Level: RO SRO-I SRO-U X Operating Test No.: ______1______




A, M, S 1


D, L, S 4P


A, D, EN, L, S 5

d. N/A

e. N/A

f. N/A

g. N/A

h. N/A



N, R, E, L 8


D, A, E 7

k. N/A





FOR REVIEW ONLY - DO NOT DISTRIBUTEILT 16-2 ONS SRO NRC Examination QUESTION 1 C1

Given the following Unit 1 conditions:

Reactor power = 100%

RCS pressure = 2200 psig lowering

1RC-66 (PORV) indicates partially open

1RC-4 will NOT close from the control room 1) __(1)__ will be entered and will dispatch an operator to open a breaker to fail

1RC-66 closed. 2) The supply breaker for 1RC-66 __(2)__ located on 1DIB.

Which ONE of the following completes the statement above? A. 1. AP/02 (Excessive RCS Leakage)

2. is

B. 1. AP/02 (Excessive RCS Leakage)

2. is NOT

C. 1. AP/44 (Abnormal Pressurizer Pressure Control)

2. is

D. 1. AP/44 (Abnormal Pressurizer Pressure Control)

2. is NOT

APE008 2.4.34 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)APE008 GENERIC

Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5 / 45.13)

Thursday, September 29, 2016 Page 1 of 279


General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because there is no direction in AP/2 to open the breaker for 1RC-66. It is plausible because 1) you meet entry conditions for AP/2, 2) AP/44 directs entry into AP/2 and AP/2 Encl 5.9 does give direction to close 1RC-4 if leakage through 1RC-66 exceeds 1 gpm.

2nd part is correct. AP/44, Step 4.3 RNO directs opening the breaker for 1RC-66 on 1DIB. The PORV will fail closed (unless mechanically stuck) when power is removed.

Answer B Discussion

Incorrect:1st part is incorrect because there is no direction in AP/2 to open the breaker for 1RC-66. It is plausible because 1) you meet entry conditions for AP/2, 2) AP/44 directs entry into AP/2 and AP/2 Encl 5.9 does give direction to close 1RC-4 if leakage through 1RC-66 exceeds 1 gpm.

2nd part is incorrect because 1DIB bkr # 24 is the power supply to 1RC-66.

Answer C Discussion

CORRECT:1st part is correct. AP/44 entry conditions are met. Step 4.3 RNO dispatches an operator to open the breaker for 1RC-66 on 1DIB.

2nd part is correct. AP/44, Step 4.3 RNO directs opening the breaker for 1RC-66 on 1DIB. The PORV will fail closed (unless mechanically stuck) when power is removed.

Answer D Discussion

Incorrect:1st part is correct. AP/44 entry conditions are met. Step 4.3 RNO dispatches an operator to open the breaker for 1RC-66 on 1DIB.

2nd part is incorrect because 1DIB bkr # 24 is the power supply to 1RC-66.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT47 Q2

Student References ProvidedDevelopment References

ILT47 Q2AP/44AP/2EAP-AP44 Obj: 4

Remarks/Status

Basis for meeting the KA

This question matches the KA by requiring knowledge of operator tasks performed outside the CR during an emergency and its operational effects.

Basis for Hi Cog

Basis for SRO only

APE008 2.4.34 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)APE008 GENERIC

Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5 / 45.13)



Given the following Unit 1 conditions: Time = 0800:

SCM = 0°F

A Small Break LOCA is in progress

EOP Immediate Manual Actions are complete

EOP Enclosure 5.1 (ES Actuation) is in progress Time = 0804:

SCM = 0°F

ALL RCPs are operating

EOP Rule 2 (Loss of SCM) is initiated

ES Channel 1 failed to go to manual

The ES ODD Voter is in OVERRIDE

ES Channel 3 was manually initiated

RCS WR Pressure is 500 psig stable 1) In accordance with EOP Rule 2, RCPs should be __(1)__ by the operator

at Time = 0804. 2) In accordance with EOP Enclosure 5.1, prior to stopping the 1A LPI pump __(2)__. Which ONE of the following completes the statements above? A. 1. stopped

2. ES Channel 3 must be RESET

B. 1. stopped

2. NO other actions are required

C. 1. left running

2. ES Channel 3 must be RESET

D. 1. left running

2. NO other actions are required

EPE009 EA1.01 - Small Break LOCAAbility to operate and monitor the following as they apply to a small break LOCA: (CFR 41.7 / 45.5 / 45.6)

RCS pressure and temperature .....................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because part of the criteria to trip RCPs on a loss of SCM is that it has been less than 2 minutes since losing it (you are over 2 minutes). It is plausible because if time since losing SCM was less than 2 minutes, it would be correct.

2nd part is correct. Encl 5.1, Step 6, is an IAAT step: IAAT all exist: 1) Voter associated with ES channel in OVERRIDE (it is), 2) An ES channel is manually actuated (Ch 3 has) and 3) Components on that channel require manipulation (LPI Pumps need to be secured), THEN depress RESET on the required channel (3).

Answer B Discussion

1st part is incorrect because part of the criteria to trip RCPs on a loss of SCM is that it has been less than 2 minutes since losing it (you are over 2 minutes). It is plausible because if time since losing SCM was less than 2 minutes, it would be correct.

2nd part is incorrect because per Encl 5.1, Step 6, is an IAAT step: IAAT all exist: 1) Voter associated with ES channel in OVERRIDE (it is), 2) An ES channel is manually actuated (Ch 3 has) and 3) Components on that channel require manipulation (LPI Pumps need to be secured), THEN depress RESET on the required channel (3). It is plausible because ES channel was initiated manually therefore it would seem logical that you could just turn the equipment off without resetting anything.

Answer C Discussion

1st part is correct. Part of the criteria for stopping all RCPs is that it has been less than 2 minutes since losing SCM. You do not meet that criteria at time = 0804.

2nd part is correct. Per Encl 5.1, Step 6, is an IAAT step: IAAT all exist: 1) Voter associated with ES channel in OVERRIDE (it is), 2) An ES channel is manually actuated (Ch 3 has) and 3) Components on that channel require manipulation (LPI Pumps need to be secured), THEN depress RESET on the required channel (3).

Answer D Discussion

1st part is correct. Part of the criteria for stopping all RCPs is that it has been less than 2 minutes since losing SCM. You do not meet that criteria at time = 0804.

2nd part is incorrect because per Encl 5.1, Step 6, is an IAAT step: IAAT all exist: 1) Voter associated with ES channel in OVERRIDE (it is), 2) An ES channel is manually actuated (Ch 3 has) and 3) Components on that channel require manipulation (LPI Pumps need to be secured), THEN depress RESET on the required channel (3). It is plausible because ES channel was initiated manually therefore it would seem logical that you could just turn the equipment off without resetting anything.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT43 Q3


ILT43 Q3EP/1/A/1800/001 Rev 1Rule 2 EP/1/A/1800/001 Enclosure 5.1IC-ES Obj: 21


This question matches the KA by requiring the ability to operate plant components during a SB LOCA based on RCS pressure.

Basis for Hi Cog

Basis for SRO only

EPE009 EA1.01 - Small Break LOCAAbility to operate and monitor the following as they apply to a small break LOCA: (CFR 41.7 / 45.5 / 45.6)



Remarks/Status

RCS pressure and temperature .....................................


FOR REVIEW ONLY - DO NOT DISTRIBUTEILT 16-2 ONS SRO NRC Examination QUESTION 3 B3


Reactor power = 100% Time = 0440:

RCS pressure = 1120 psig stable

Reactor Building pressure peaked at 4.6 psig and is now 2.8 psig slowly lowering

EOP Enclosure 5.1 (ES Actuation) initiated Time = 0514:

RCS pressure = 1100 psig lowering Time = 0515:

RCS pressure = 178 psig lowering

Reactor Building pressure = 8.8 psig rising

EOP Enclosure 5.1 complete with outstanding IAATs At Time = 0515, 1A and 1B LPI pumps are _______. Which ONE of the following completes the statement above? A. operating from the initial ES actuation

B. off and must be restarted

C. operating after automatically re-starting when RCS pressure lowered to the ES setpoint of 500 psig

D. operating after automatically re-starting when RCS pressure lowered to the ES setpoint of 550 psig

EPE011 2.4.21 - Large Break LOCAEPE011 GENERIC

Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12)



General Discussion

Answer A Discussion

Incorrect:Incorrect because Enclosure 5.1 will direct securing the LPI pumps. It is plausible because the LPI pumps will have automatically started with ES and if no operator action were taken in Encl 5.1, it would be correct.

Answer B Discussion

CORRECT:The LPI pumps will start when RB pressure exceeds 3 psig (Time = 0440). Encl 5.1 will secure 1A & 1B LPI pumps due to running at shutoff head. When RCS pressure is < LPIP shutoff head, Encl 5.1 directs restarting LPIPs.

Answer C Discussion

Incorrect:Incorrect because after the LPIPs were secured per Encl 5.1, they will not start with an additional ES signal. In this case also, 500 psig is not the actual sepoint but the TS required setpoint is to be > 500 psig. It is plausible to think that with another ES signal, the pumps would restart.

Answer D Discussion

Incorrect:Incorrect because after the LPIPs were secured per Encl 5.1, they will not start with an additional ES signal. It is plausible to think that with another ES signal (ES 3 & 4 at 550 psig), the pumps would restart.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT39 Q4


ILT39 Q4IC ES Obj: 12Encl 5.1

Remarks/Status


Matches KA by requiring the ability to determine the status of safety functions (LPIPs) based on parameters provided and knowledge of equipment logic.

Basis for Hi Cog

Basis for SRO only

EPE011 2.4.21 - Large Break LOCAEPE011 GENERIC

Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12)


FOR REVIEW ONLY - DO NOT DISTRIBUTEILT 16-2 ONS SRO NRC Examination QUESTION 4 D4


Core Thermal Power = 100%

A Station Blackout occurs

AP/0/A/1700/025 (Standby Shutdown Facility Emergency Operating Procedure) has been initiated

Time = 0610:

1XSF is being powered from OXSF 1) In accordance with station Time Critical Actions, the latest time that SSF RCMU

flow can be established to Unit 1 RCP seals without potentially damaging/degrading the seals is __(1)__.

2) At time = 0610, 1HP-20 (RCP Seal Return) __(2)__ be operated from Unit 1 Control

Room. Which ONE of the following completes the statements above? A. 1. 0614

2. can

B. 1. 0614

2. can NOT

C. 1. 0620

2. can

D. 1. 0620

2. can NOT

APE015/017 AK2.07 - Reactor Coolant Pump (RCP) MalfunctionsKnowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 / 45.7)

RCP seals ......................................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because the TCA for establishing RCMU is 20 minutes of when CC and Seal Injection are lost to the RCP seals. This would be 0620. It is plausible because there is a TCA associated with the SSF which is 14 minutes (establishing SSF ASW to the SGs if all feedwater is lost).

2nd part is incorrect because 1HP-20 can not be operated from the control room once 1XSF power has been transferred to OSXF. It is plausible because if it were 1HP-21 (the other seal return isolation valve), it would be correct.

Answer B Discussion

Incorrect:1st part is incorrect because the TCA for establishing RCMU is 20 minutes of when CC and Seal Injection are lost to the RCP seals. This would be 0620. It is plausible because there is a TCA associated with the SSF which is 14 minutes (establishing SSF ASW to the SGs if all feedwater is lost).

2nd part is correct. 1HP-20 can not be operated from the control room once 1XSF power has been transferred to OSXF.

Answer C Discussion

Incorrect:1st part is correct. The TCA for establishing RCMU is 20 minutes of when CC and Seal Injection are lost to the RCP seals. This would be 0620.

2nd part is incorrect because 1HP-20 can not be operated from the control room once 1XSF power has been transferred to OSXF. It is plausible because if it were 1HP-21 (the other seal return isolation valve), it would be correct.

Answer D Discussion

CORRECT:1st part is correct. The TCA for establishing RCMU is 20 minutes of when CC and Seal Injection are lost to the RCP seals. This would be 0620.

2nd part is correct because 1HP-20 can not be operated from the control room once 1XSF power has been transferred to OSXF.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT43 Q5


ILT43 Q5EAP-SSF Obj: R29AP/0/A/1700/025 Rev. 062

Remarks/Status


This question matches the KA by requiring knowledge of relationship between losing RCP (blackout) including RCP seals and establishing RCP seals via the SSF.

Basis for Hi Cog

Basis for SRO only

APE015/017 AK2.07 - Reactor Coolant Pump (RCP) MalfunctionsKnowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 / 45.7)

RCP seals ......................................................



1) When re-establishing seal injection flow to a Unit 1 RCP, the major concern with thermal

shock to the RCP seals is seal failure resulting in __(1)__. 2) The MINIMUM condition that would require entry into AP/1/A/1700/014 (Loss Of Normal

HPI Makeup and/or RCP Seal Injection) is loss of seal injection to __(2)__ RCP(s).

Which ONE of the following completes the statements above? A. 1. excessive RCS leakage

2. ANY

B. 1. excessive RCS leakage

2. ALL

C. 1. RCP shaft binding

2. ANY

D. 1. RCP shaft binding

2. ALL.

APE022 AK1.01 - Loss of Reactor Coolant MakeupKnowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: (CFR 41.8 / 41.10 / 45.3)

Consequences of thermal shock to RCP seals . . . . . . . . . . . . . . . . . . . .



General Discussion

Answer A Discussion

1st part is correct.

2nd part is incorrect because per AP/14 entry condions for AP/14 is a complete a loss of seal injection to ALL RCPs. It is plausible in that a loss of seal injection to a single RCP is some type of RCMU malfunction.

Answer B Discussion

1st part is correct.

2nd part is correct. Per AP/14, the individual RCP throttle valve are adjusted locally to limit temperature increase when re-establishing seal flow.

Answer C Discussion

1st part is incorrect because the concern with thermal shock of the RCP seals is increasing RCS leakage. It particular, exceeding the capacity of the RCMU pump. It is plausible because damaging the seals will result in some type of binding / contact with the seal surfaces.

2nd part is incorrect because per AP/14 entry condions for AP/14 is a complete a loss of seal injection to ALL RCPs. It is plausible in that a loss of seal injection to a single RCP is some type of RCMU malfunction.

Answer D Discussion

1st part is incorrect because the concern with thermal shock of the RCP seals is increasing RCS leakage. It particular, exceeding the capacity of the RCMU pump. It is plausible because damaging the seals will result in some type of binding / contact with the seal surfaces.

2nd part is correct. Per AP/14, the individual RCP throttle valve are adjusted locally to limit temperature increase when re-establishing seal flow.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT46 Q5


ILT46 Q5AP/14EAP AP14 Obj: 3

Remarks/Status

9/13/16 - Received new K/A (APE022 AK1.01) from Dan Bacon following review of 10 preview questions. Could not write a discriminating question to the original K/A.


Question requires knowledge of the operational implications of re-establishing seal injection flow to the RCPs including thermal shock and possible damage..

Basis for Hi Cog

Basis for SRO only

APE022 AK1.01 - Loss of Reactor Coolant MakeupKnowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: (CFR 41.8 / 41.10 / 45.3)

Consequences of thermal shock to RCP seals . . . . . . . . . . . . . . . . . . . .




RCS cooldown in progress

LPI aligned in the Series Mode 1) The reason Series Mode was developed for Unit 2 was to provide __(1)__. 2) A loss of the __(2)__ LPI Pumps would result in a total loss of Decay Heat

Removal. Which ONE of the following completes the statements above? A. 1. a backup to the Switchover mode of LPI

2. 2A and 2C

B. 1. a backup to the Switchover mode of LPI

2. 2B and 2C

C. 1. additional cooling capacity during 2/0 pump ops

2. 2A and 2C

D. 1. additional cooling capacity during 2/0 pump ops

2. 2B and 2C

APE025 AA1.03 - Loss of Residual Heat Removal System (RHRS)Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: (CFR 41.7 / 45.5 / 45.6)

LPI pumps ......................................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because this was not the reason for the development of the "Series" lineup. It is plausible because it is the reason that the High Pressure Mode was added to Unit 2 LPI system.

2nd part is correct. Since the 2A and 2C LPI pumps are used in Series Mode, a loss of both pumps would result in a loss of decay heat removal.

Answer B Discussion

Incorrect:1st part is incorrect because this was not the reason for the development of the "Series" lineup. It is plausible because it is the reason that the High Pressure Mode was added to Unit 2 LPI system.

2nd part is incorrect because the 2A and 2C LPI pumps are used in Series Mode. It is plausible because the 2C LPI pump is used in series mode and the flowpath for the LPI fluid goes through the 2B LPI cooler first making it plausible that the 2B LPI pump would be used.

Answer C Discussion

Correct:1st part is correct. During cooldowns in the 2/0 RCP configuration on Units 1 and 2, it was noticed that, due to the heat added by the second RCP, it was impossible to cooldown at or near the procedural cooldown rate. A need for additional cooling capacity was thus determined. A way that was devised to provide the additional cooling needed was the LPI Series Mode. The flowpath for the LPI Series Mode is from the Decay Heat DropLine through LP-68, through the "B" LPI Cooler, through LP-73, LP-74 and LP-75 to the suction of the "A" and "C" LPI Pumps. The LPI Pump discharge is to the "A" LPI Cooler and then to the vessel through both headers via LPI cross-over.

2nd part is correct. Since the 2A and 2C LPI pumps are used in Series Mode, a loss of both pumps would result in a loss of decay heat removal.

Answer D Discussion

1st part is correct. During cooldowns in the 2/0 RCP configuration on Units 1 and 2, it was noticed that, due to the heat added by the second RCP, it was impossible to cooldown at or near the procedural cooldown rate. A need for additional cooling capacity was thus determined. A way that was devised to provide the additional cooling needed was the LPI Series Mode. The flowpath for the LPI Series Mode is from the Decay Heat DropLine through LP-68, through the "B" LPI Cooler, through LP-73, LP-74 and LP-75 to the suction of the "A" and "C" LPI Pumps. The LPI Pump discharge is to the "A" LPI Cooler and then to the vessel through both headers via LPI cross-over.

2nd part is incorrect because the 2A and 2C LPI pumps are used in Series Mode. It is plausible because the 2C LPI pump is used in series mode and the flowpath for the LPI fluid goes through the 2B LPI cooler first making it plausible that the 2B LPI pump would be used.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT46 Q6


ILT46 Q6PNS-LPI Obj: 9, 10, 17


Requires knowledge of the relationship between a Loss of DHR and LPI pump availability when in the Series Mode of DHR alignment with LPI.

Basis for Hi Cog

Basis for SRO only

APE025 AA1.03 - Loss of Residual Heat Removal System (RHRS)Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: (CFR 41.7 / 45.5 / 45.6)

LPI pumps ......................................................



Remarks/Status



Given the following Unit 1 conditions: Initial conditions:

Reactor is tripped

The SGTR tab is in progress due to a 120 gpm primary to secondary leak Current conditions:

1RC-1 (Pzr Spray) has failed closed

Auxiliary Spray is being considered to reduce Core SCM In accordance with the SGTR tab… 1) __(1)__ temperature will be used when determining if Auxiliary Spray is allowed. 2) If the suction source for Auxiliary Spray is too cold, it will not be used due to

Potential __(2)__. Which ONE of the following completes the statements above? A. 1. LDST

2. thermal shock of the Pzr Spray Nozzle

B. 1. LDST

2. subcooling of the Pzr

C. 1. BWST

2. thermal shock of the Pzr Spray Nozzle

D. 1. BWST

2. subcooling of the Pzr

APE027 AK3.03 - Pressurizer Pressure Control System (PZR PCS) MalfunctionKnowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: (CFR 41.5,41.10 / 45.6 / 45.13)

Actions contained in EOP for PZR PCS malfunction ...................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because the SGTR tab directs using BWST temperature to determine PZR spray nozzle delta T. It is plausible because HPI suction is normally aligned to the LDST.

2nd part is correct. The DT between the suction and Pzr temperature is limited to 410 degrees to prevent thermal shock of the Pzr Spray nozzles.

Answer B Discussion

Incorrect:1st part is incorrect because the SGTR tab directs using BWST temperature to determine PZR spray nozzle delta T. It is plausible because HPI suction is normally aligned to the LDST.

2nd part is incorrect because the DT between the suction and Pzr temperature is limited to 410 degrees to prevent thermal shock of the Pzr Spray nozzles. It is plausible because the Pzr will because putting ~ 110 degree water in the Pzr will probably cause it to become subcooled however, the concern (as stated in the SGTR tab) is thermal shock of the Pzr Spray Nozzle.

Answer C Discussion

CORRECT:1st part is correct. The SGTR tab directs using BWST temperature to determine PZR spray nozzle delta T.

2nd part is correct. The DT between the suction and Pzr temperature is limited to 410 degrees to prevent thermal shock of the Pzr Spray nozzles.

Answer D Discussion

Incorrect:1st part is correct. The SGTR tab directs using BWST temperature to determine PZR spray nozzle delta T.

2nd part is incorrect because the DT between the suction and Pzr temperature is limited to 410 degrees to prevent thermal shock of the Pzr Spray nozzles. It is plausible because the Pzr will because putting ~ 110 degree water in the Pzr will probably cause it to become subcooled however, the concern (as stated in the SGTR tab) is thermal shock of the Pzr Spray Nozzle.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

MODIFIED

Question Source

ILT43 Q9


ILT43 Q9EP/1/A/1800/001 (SGTR) Rev. 0EAP-SGTR Obj R6

Remarks/Status

DISCUSSED WITH CHIEF 7/27: OK WITH FIRST QUESTION MATCHING THE KA.


This question matches the KA by requiring knowledge of actions contained in the EOP for a Pzr pressure control malfunction.

Basis for Hi Cog

Basis for SRO only

APE027 AK3.03 - Pressurizer Pressure Control System (PZR PCS) MalfunctionKnowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: (CFR 41.5,41.10 / 45.6 / 45.13)

Actions contained in EOP for PZR PCS malfunction ...................





Feedwater and Diamond in MANUAL Current conditions:

BOTH Main Feedwater pumps trip

Reactor power = 35% lowering

RCS pressure = 2378 psig

Rule 1 (ATWS/Unanticipated Nuclear Power Production) in progress 1) Negative reactivity was FIRST added from __(1)__. 2) The 3RC-66 (PORV) is __(2)__. Which ONE of the following completes the statements above? A. 1. Moderator temperature coefficient

2. open

B. 1. Moderator temperature coefficient

2. closed

C. 1. Control Rods

2. open

D. 1. Control Rods

2. closed

EPE029 EK1.01 - Anticipated Transient Without Scram (ATWS)Knowledge of the operational implications of the following concepts as they apply to the ATWS: (CFR 41.8 / 41.10 / 45.3)

Reactor nucleonics and thermo-hydraulics behavior ...................



General Discussion

Answer A Discussion

Incorrect:1st part is correct. The RCS will start heating up immediately which will add negative reactivity.

2nd part is incorrect because RCS pressure is below the PORV setpoint. It is plausible because if RCS pressure were > 2450 psig, it would be correct. (PORV opens at 2450 psig and closes at 2400 psig)

Answer B Discussion

CORRECT:1st part is correct. The RCS will start heating up immediately which will add negative reactivity.

2nd part is correct. RCS pressure is below the PORV setpoint. (PORV opens at 2450 psig and closes at 2400 psig).

Answer C Discussion

Incorrect:1st part is incorrect because RCS temperature starts increasing before the operator starts inserting control rods per Rule 1. It is plausible because this would be the correct answer for normal power decreases done in accordance with plant procedures.

2nd part is incorrect because RCS pressure is below the PORV setpoint. It is plausible because if RCS pressure were > 2450 psig, it would be correct. (PORV opens at 2450 psig and closes at 2400 psig)

Answer D Discussion

Incorrect:1st part is incorrect because RCS temperature starts increasing before the operator starts inserting control rods per Rule 1. It is plausible because this would be the correct answer for normal power decreases done in accordance with plant procedures.

2nd part is correct. RCS pressure is below the PORV setpoint. (PORV opens at 2450 psig and closes at 2400 psig).

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

2009B Q8


2009B Q8PNS-PZR Obj: 15EOP Rule 1

Remarks/Status


Question requires knowledge of the reactor theory (nucleonics) and the operational impact of Rx theory (Moderator Temperature coefficient) during an ATWS.

Basis for Hi Cog

Basis for SRO only

EPE029 EK1.01 - Anticipated Transient Without Scram (ATWS)Knowledge of the operational implications of the following concepts as they apply to the ATWS: (CFR 41.8 / 41.10 / 45.3)

Reactor nucleonics and thermo-hydraulics behavior ...................





TDEFDW pump is OOS Time = 0401:

Both Main FDW pumps trip

Loss of IA to 1FDW-316 occurs Time = 0403:

1A SG level = 27 inches XSUR rising

1B SG level = 36 inches XSUR rising 1) At Time = 0403, __(1)__ SG has indications of a tube leak. 2) At Time = 0403, and in accordance with Rule 7 (SG Feed Control), the MAXIMUM

EFDW flow allowed to each SG is __(2)__ gpm. Which ONE of the following completes the statements above? A. 1. 1A

2. 600

B. 1. 1A

2. 1000

C. 1. 1B

2. 600

D. 1. 1B

2. 1000

EPE038 EA2.03 - Steam Generator Tube Rupture (SGTR)Ability to determine or interpret the following as they apply to a SGTR : (CFR 43.5 / 45.13)

Which S/G is ruptured ............................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because being at 27" XSUR increasing is not indication of a primary to secondary leak. It is plausible because the 1A SG level is increasing above the Main FDW setpoint of 25 inches. If MFW were in operation, it could be correct.

2nd part is correct. With the TD EFDWP OOS, the header flow will be limited to the maximum flow allowed from the MD EFDWP which is 600 gpm.

Answer B Discussion

Incorrect:1st part is incorrect because being at 27" XSUR increasing is not indication of a primary to secondary leak. It is plausible because the 1A SG level is increasing above the Main FDW setpoint of 25 inches. If MFW were in operation, it could be correct.

2nd part is incorrect because with the TD EFDWP OOS, the header flow will be limited to the maximum flow allowed from the MD EFDWP which is 600 gpm. It is plausible because the limit on the header is 1000 gpm and would be correct if the TD EFDW pump were also in operation.

Answer C Discussion

CORRECT:1st part is correct because being at 36" XSUR increasing is an indication that level is above level setpoint. The loss of IA to 1FDW-316 could be an indication for this valve to fail open however, there is N2 backup which will still allow operation.

2nd part is correct. With the TD EFDWP OOS, the header flow will be limited to the maximum flow allowed from the MD EFDWP which is 600 gpm.

Answer D Discussion

Incorrect:1st part is correct because being at 36" XSUR increasing is an indication that level is above level setpoint. The loss of IA to 1FDW-316 could be an indication for this valve to fail open however, there is N2 backup which will still allow operation.

2nd part is incorrect because with the TD EFDWP OOS, the header flow will be limited to the maximum flow allowed from the MD EFDWP which is 600 gpm. It is plausible because the limit on the header is 1000 gpm and would be correct if the TD EFDW pump were also in operation.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT45 Q10


ILT45 Q10CF-EF Obj: 51EAP-SGTR R2EOP Rule 7


This question matches the KA by requiring an evaluation of control indications to determine which SG is ruptured.

Basis for Hi Cog

Basis for SRO only

EPE038 EA2.03 - Steam Generator Tube Rupture (SGTR)Ability to determine or interpret the following as they apply to a SGTR : (CFR 43.5 / 45.13)

Which S/G is ruptured ............................................



Remarks/Status




Reactor power = 100% Current conditions:

See reference 1) The event causing the indications on the attached P/T display is a __(1)__. 2) The Reactor __(2)__ automatically trip by actuating the “Low RCS Pressure Trip”

setpoint first. Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. Steam line break

2. did

B. 1. Steam line break

2. did NOT

C. 1. LOCA

2. did

D. 1. LOCA

2. did NOT

APE040 AA2.03 - Steam Line RuptureAbility to determine and interpret the following as they apply to the Steam Line Rupture: (CFR: 43.5 / 45.13)

Difference between steam line rupture and LOCA .....................



General Discussion

Answer A Discussion

Incorrect:1st part is correct. The attachment is a PT display showing indications during a steam line rupture. In particular, the Tc, Th trends cooling off as well as the 1B SG pressure at 139 psig.

2nd part is incorrect because the RPS Low RCS Pressure trip was not the first signal to trip the reactor. The Variable Low Pressure Trip setpoint (the diagonal white line) was reached before the RCS Low Pressure Trip setpoint (the bottom horizontal while line). Tc (blue) and Th (red) lines track with RCS pressure. As RCS pressure decreases, the lines will track together and the Th line will hit a white line first. This is the Variable Low Pressure Trip signal. Also, it is very possible that the low steam pressure has already tripped both MFW pumps which has, in turn sent a trip signal to RPS as well. It is plausible because the Tc dot on the PT display has clearly crossed the horizontal white line.

Answer B Discussion

CORRECT:1st part is correct. The attachment is a PT display showing indications during a steam line rupture. In particular, the Tc, Th trends cooling off as well as the 1B SG pressure at 139 psig.

2nd part is correct because the RPS Low RCS Pressure trip was not the first signal to trip the reactor. The Variable Low Pressure Trip setpoint (the diagonal white line) was reached before the RCS Low Pressure Trip setpoint (the bottom horizontal while line). Tc (blue) and Th (red) lines track with RCS pressure. As RCS pressure decreases, the lines will track together and the Th line will hit a white line first. This is the Variable Low Pressure Trip signal. Also, it is very possible that the low steam pressure has already tripped both MFW pumps which has, in turn sent a trip signal to RPS as well.

Answer C Discussion

Incorrect:1st part is incorrect because the attachment is a PT display showing indications during a steam line rupture. It is plausible because the display does show some indications of a SBLOCA (RCS pressure decreasing rapidly, Pzr level decreasing).

2nd part is incorrect because the RPS Low RCS Pressure trip was not the first signal to trip the reactor. The Variable Low Pressure Trip setpoint (the diagonal white line) was reached before the RCS Low Pressure Trip setpoint (the bottom horizontal while line). Tc (blue) and Th (red) lines track with RCS pressure. As RCS pressure decreases, the lines will track together and the Th line will hit a white line first. This is the Variable Low Pressure Trip signal. Also, it is very possible that the low steam pressure has already tripped both MFW pumps which has, in turn sent a trip signal to RPS as well. It is plausible because the Tc dot on the PT display has clearly crossed the horizontal white line.

Answer D Discussion

Incorrect:1st part is incorrect because the attachment is a PT display showing indications during a steam line rupture. It is plausible because the display does show some indications of a SBLOCA (RCS pressure decreasing rapidly, Pzr level decreasing).

2nd part is correct because the RPS Low RCS Pressure trip was not the first signal to trip the reactor. The Variable Low Pressure Trip setpoint (the diagonal white line) was reached before the RCS Low Pressure Trip setpoint (the bottom horizontal while line). Tc (blue) and Th (red) lines track with RCS pressure. As RCS pressure decreases, the lines will track together and the Th line will hit a white line first. This is the Variable Low Pressure Trip signal. Also, it is very possible that the low steam pressure has already tripped both MFW pumps which has, in turn sent a trip signal to RPS as well.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


This question matches the KA by requiring knowledge of different indications associated with a steam line break vs a LOCA.

Basis for Hi Cog

Basis for SRO only



Student References Provided

Q10 PT Display

Development References

Q10 PT DisplaySF10, Obj R10OMP 1-18

Remarks/Status

APE040 AA2.03 - Steam Line RuptureAbility to determine and interpret the following as they apply to the Steam Line Rupture: (CFR: 43.5 / 45.13)

Difference between steam line rupture and LOCA .....................





FDW Loop Masters in HAND

OAC computer point readings

o FDW Flow A = 4.8 MPPH lowering

o FDW Flow B = 5.8 MPPH rising

o Generator output = 890 MWe stable

RB pressure = 1.1 psig slowly rising Time = 0801:

Reactor trip 1) Plant conditions at Time = 0800 indicate a __(1)__ inside containment.

2) When the Reactor trip occurs, the FDW Loop Masters will __(2)__. Which ONE of the following completes the statements above? A. 1. 1B Main Steam leak

2. swap to AUTOMATIC

B. 1. 1B Main Steam leak

2. remain in HAND

C. 1. 1B Main Feedwater line leak

2. swap to AUTOMATIC

D. 1. 1B Main Feedwater line leak

2. remain in HAND

APE054 AA2.08 - Loss of Main Feedwater (MFW)Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): (CFR: 43.5 / 45.13)

Steam flow-feed trend recorder ....................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect. Plausible because if MWe were lowering, it would be correct. Oconee does not have direct measurment of steam flow. The most direct indication of steam flow at Oconee is megawatts.

2nd part is incorrect because FDW Loop Masters will remain in HAND. It is plausible because if the Main FDW valve controllers, Startup FDW valve controllers or the MFW Pump Speed controllers were in HAND, it would be correct.

Answer B Discussion

Incorrect:1st part is incorrect. Plausible because if MWe were lowering, it would be correct. Oconee does not have direct measurment of steam flow. The most direct indication of steam flow at Oconee is megawatts.

2nd part is correct. FDW Loop Masters will remain in HAND when the reactor trips.

Answer C Discussion

Incorrect:1st part is correct. A break in the feedwater line to the 1B SG would cause an initial increase in FDW flow B and decrease in FDW flow A due to robbing flow. Generator output would not initially change due to the delayed effects of changing steam flow.

2nd part is incorrect because FDW Loop Masters will remain in HAND. It is plausible because if the Main FDW control valve controllers, Startup FDW valve controllers or the MFW Pump Speed controllers were in HAND, it would be correct.

Answer D Discussion

CORRECT:1st part is correct. A break in the feedwater line to the 1B SG would cause an initial increase in FDW flow B and decrease in FDW flow A due to robbing flow. Generator output would not initially change due to the delayed effects of changing steam flow.

2nd part is correct. FDW Loop Masters will remain in HAND when the reactor trips.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


ICS-03 Obj: 7STG-SG

Remarks/Status

Need New KA. ONS does not have a Steam-Feed flow recorder.

DISCUSSED WITH CHIEF 7/27: CAN USE FEED FLOW INDICATION AND STEAM FLOW INDICATION AND ASK WHAT EVENT CAUSED THE INDICATION.


This question matches the KA by requiring the ability to determine that a feedwater line break exist (loss of FDW to a SG) using Feed and Steam flow indications.

Basis for Hi Cog

Basis for SRO only

APE054 AA2.08 - Loss of Main Feedwater (MFW)Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): (CFR: 43.5 / 45.13)

Steam flow-feed trend recorder ....................................


FOR REVIEW ONLY - DO NOT DISTRIBUTEILT 16-2 ONS SRO NRC Examination QUESTION 12 A12

Given the following plant conditions: Initial conditions:

All three units Reactor power = 100% Current conditions:

All Unit’s 4160v Main Feeder Busses are de-energized

Unit 1, 2, and 3 EOP Blackout tabs in progress 1) Unit 1’s Essential Inverters KI, KX and KU are de-energized to extend the life of

batteries __(1)__.

2) These batteries are sized to carry the continuous emergency load for a period of __(2)__.

Which ONE of the following completes the statements above? A. 1. CA and CB

2. 1 hour

B. 1. CA and CB

2. 2 hours

C. 1. PA and PB

2. 1 hour

D. 1. PA and PB

2. 2 hours

EPE055 EK1.01 - Loss of Offsite and Onsite Power (Station Blackout)Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : (CFR 41.8 / 41.10 / 45.3)

Effect of battery discharge rates on capacity ..........................



General Discussion

Answer A Discussion

CORRECT:1st part is correct. KI, KU & KX are powered from batteries CA or CB.

2nd part is correct. Each battery is sized to carry the continuous emergency load for a period of one hour.

Answer B Discussion

Incorrect:1st part is correct. KI, KU & KX are powered from batteries CA or CB.

2nd part is incorrect because each battery is sized to carry the continuous emergency load for a period of one hour. It is plausible because when battery cell voltages are less than required by TS (on PSW batteries), a surveillance is required to be performed within 2 hours.

Answer C Discussion

Incorrect:1st part is incorrect because KI, KU & KX are powered from batteries CA or CB. It is plausible because if the question was about de-energizing Essential inverter KOAC it would be correct.

2nd part is correct. Each battery is sized to carry the continuous emergency load for a period of one hour.

Answer D Discussion

Incorrect:1st part is incorrect because KI, KU & KX are powered from batteries CA or CB. It is plausible because if the question was about de-energizing Essential inverter KOAC it would be correct.

2nd part is incorrect because each battery is sized to carry the continuous emergency load for a period of one hour. It is plausible because when battery cell voltages are less than required by TS (on PSW batteries), a surveillance is required to be performed within 2 hours.

Cognitive Level

Memory

Job Level

RO

QuestionType

MODIFIED

Question Source

2009 Q48


2009 Q48EOP BO Tab R1EOP Encl 5.32 R0EAP-SBO Obj: 7EL-DCDTS 3.7.10a

Remarks/Status


This question matches the KA by requiring knowledge of the discharge capacity of station batteries during a blackout.

Basis for Hi Cog

Basis for SRO only

EPE055 EK1.01 - Loss of Offsite and Onsite Power (Station Blackout)Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : (CFR 41.8 / 41.10 / 45.3)

Effect of battery discharge rates on capacity ..........................




Reactor power = 90% lowering

1RC-4 (PORV Block Valve) is closed due to 1RC-66 (PORV) leaking past the seat

Reactor trips

CT-1 Lockout occurs

TD EFDWP fails to start Time = 0803:

ALL 4160 VAC Switchgear (1TC, 1TD, 1TE) is de-energized

The OATC is directed to open 1RC-4

RCS pressure = 2300 psig rising 1) When the OATC rotates the switch for 1RC-4 (PORV Block Valve) to OPEN, the

valve will __(1)__.

2) At Time = 0803, Rule 4 (Initiation of HPI Forced Cooling) __(2)__ be performed. Which ONE of the following completes the statements above? A. 1. remain closed

2. will

B. 1. remain closed

2. will NOT

C. 1. open

2. will

D. 1. open

2. will NOT

APE056 AA1.33 - Loss of Offsite PowerAbility to operate and / or monitor the following as they apply to the Loss of Offsite Power: (CFR 41.7 / 45.5 / 45.6)

PORV block valve control switch ...................................



General Discussion

Answer A Discussion

CORRECT:1st part is correct. The PORV Block Valve is powered from 1XSF1/LC 1X8/1TC and will be de-energized (failed as is).

2nd part is correct. Even without power available to the HPIPs, Rule 4 will be performed to ensure some actions are performed. There are steps in Rule 4 that address not having any HPIPs available.

Answer B Discussion

Incorrect:1st part is correct. The PORV Block Valve is powered from 1XSF1/LC 1X8/1TC and will be de-energized (failed as is).

2nd part is incorrect because Rule 4 will be performed to ensure some actions are performed. It is plausible because Rule 4 would typically start all HPIPs and open the PORV. It would seem logical that since you do not have any HPIPs, that performing Rule 4 would not be required.

Answer C Discussion

Incorrect:1st part is incorrect because the PORV Block Valve is powered from 1XSF1/LC 1X8/1TC and will be de-energized (failed as is). It is plausible because if it were the PORV, it would be correct (its DC powered from 1DIB).

2nd part is correct. Even without power available to the HPIPs, Rule 4 will be performed to ensure some actions are performed. There are steps in Rule 4 that address not having any HPIPs available.

Answer D Discussion

Incorrect;1st part is incorrect because the PORV Block Valve is powered from 1XSF1/LC 1X8/1TC and will be de-energized (failed as is). It is plausible because if it were the PORV, it would be correct (its DC powered from 1DIB).

2nd part is incorrect because Rule 4 will be performed to ensure some actions are performed. It is plausible because Rule 4 would typically start all HPIPs and open the PORV. It would seem logical that since you do not have any HPIPs, that performing Rule 4 would not be required.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


EL-EPD Obj: 17Vital Powrer DWGPNS-Pzr Obj: 17Rule 3 & 4

Remarks/Status


The question matches the KA by requiring knowledge of the effect of a loss of offsite power on PORV block valve operation.

Basis for Hi Cog

Basis for SRO only

APE056 AA1.33 - Loss of Offsite PowerAbility to operate and / or monitor the following as they apply to the Loss of Offsite Power: (CFR 41.7 / 45.5 / 45.6)

PORV block valve control switch ...................................




A loss of both MFDW pumps occurs from 100% power

Rule 3 (Loss of Main or Emergency FDW) is in progress

1FDW-315 and 1FDW-316 are maintaining SG levels at setpoint Current conditions:

1KVIB is de-energized Assuming no additional operator actions, which ONE of the following will be directed by the EOP and why? A. Take manual control of 1FDW-315 due to the loss of power to the auto level

control circuitry

B. Take manual control of 1FDW-316 due to the loss of power to the auto level control circuitry

C. Feed the 1A SG through 1FDW-35 (1A STARTUP FDW CONTROL) since 1FDW-315 will fail open

D. Feed the 1B SG through 1FDW-44 (1B STARTUP FDW CONTROL) since 1FDW-316 will fail open

APE057 AK3.01 - Loss of Vital AC Electrical Instrument BusKnowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: (CFR 41.5,41.10 / 45.6 / 45.13)

Actions contained in EOP for loss of vital ac electrical instrument bus ...



General Discussion

Answer A Discussion

CORRECT:When KVIB is lost, the 1FDW-315 controller automatically swaps to its alternate source and the valve would fail to the open position due to loss of power to the auto level control circuitry and it does not auto swap to its alternate. Rule 3 will direct attempting to control the valve with the moore controller in manual and in this case it would be successful. KVIB is the normal supply to the 1FDW-315 Moore controller and the 1A SG XSUR Primary level train.

Answer B Discussion

Incorrect:KVIB does not power 1FDW-316. It is plausible because this would be correct for a loss of KVIC.

Answer C Discussion

Incorrect:Power to 1FDW-315 will automatically swap to KVIA. It is plausible since this would be correct under the assumption that the Moore controller for FDW-315 did not auto swap to its alternate source of power and therefore could not be used to control valve position. If 1FDW-315 did not work in auto or manual, Rule 3 would direct initiating Enclosure 5.27 which would direct EFW flow to the SG through 1FDW-35.

Answer D Discussion

Incorrect:KVIB does not power 1FDW-316. It is plausible because this would be correct under the assumption that the Moore controller for FDW-316 did not auto swap to its alternate source of power and therefore could not be used to control valve position. If 1FDW-316 did not work in auto or manual, Rule 3 would direct initiating Enclosure 5.27 which would direct EFW flow to the SG through 1FDW-44.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT47 Q13


ILT47 Q13CF-EF Obj: 35Rule 3

Remarks/Status


This question matches the KA by requiring knowledge of how RO actions in Rule 3 (EOP) will be dictated by a loss of vital AC power (KVIB).

Basis for Hi Cog

Basis for SRO only

APE057 AK3.01 - Loss of Vital AC Electrical Instrument BusKnowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: (CFR 41.5,41.10 / 45.6 / 45.13)

Actions contained in EOP for loss of vital ac electrical instrument bus ...




Reactor power = 100% Which ONE of the following would result in the Auxiliary Instrument Air system being required to maintain 2HP-5 OPEN assuming NO operator actions are taken? A. Unit 2 blackout occurs

B. Primary IA compressor fails

C. 2 inch IA header rupture in Unit 2 East Penetration room

D. 230KV Red Bus lockout occurs in the 230KV switchyard

APE065 AK3.04 - Loss of Instrument AirKnowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: (CFR 41.5,41.10 / 45.6 / 45.13)

Cross-over to backup air supplies ...................................



General Discussion

Answer A Discussion

Incorrect:Plausible since these conditions would result in losing all but 1 of the IA supply systems. Additionally plausible since this would mean that no 4160V power is available to Unit 2 and all unit AC powered equipment would lose power.

Answer B Discussion

Incorrect:Plausible since the Primary IA compressor is the normal supply to the IA header however the Backup IA compressors can maintain IA header pressure on a loss of the Primary IA compressor under normal conditions.

Answer C Discussion

CORRECT:The existing Instrument Air system, including the Primary Instrument Air Compressor, can handle a 1.5-inch diameter or smaller break.This means that if the break is > 1.5 inches, the system is not capable of maintaining the air pressure high enough to operate components.The AIA system has spring loaded check valves located at major junctions of the IA and AIA systems and at each of the individual components served. These will close if the IA system begins to depressurize

Answer D Discussion

Incorrect:Plausible since the normal supply to 2HP-5 is provided by the Primary IA system and the normal power supply to the Primary IA compressor is from the 230KV switchyard red buss and would therefore be unavailable to provide power to the Primary IA compressor following a red buss lockout. Additional plausibility is from having to assess if the backup IA compressors are going to be available following the red buss lockout however 4160V power would not be lost therefore the Backup IA compressors would still be available.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

2009B Q15


2009B Q15SSS-IA Obj: 23, 24

Remarks/Status


This question requires knowledge of the reason for providing Auxiliary Instrument Air supply to selected components. Knowledge of the fact that the AIA system is meant to provide air to selected components in the event of a line break of > 1.5 inches in the normal IA supply to the valve is required to arrive at the correct choice.

Basis for Hi Cog

Basis for SRO only

APE065 AK3.04 - Loss of Instrument AirKnowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: (CFR 41.5,41.10 / 45.6 / 45.13)

Cross-over to backup air supplies ...................................




AP/1/A/1700/034 (Degraded Grid) in progress

Generator output = 850 MWe and 450 MVARs lagging

Generator Hydrogen Pressure = 60 psig

Generator Output Voltage = 18.2 KV 1) The Generator output __(1)__ within the limits of the Generator Capability Curve. 2) If the generator exceeds the Underfrequency Maximum Allowable Time given in

AP/34 (Degraded Grid), the Main Turbine __(2)__ automatically trip. Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. is NOT

2. will

B. 1. is NOT

2. will NOT

C. 1. is

2. will

D. 1. is

2. will NOT

APE077 AK2.03 - Generator Voltage and Electric Grid DisturbancesKnowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)

Sensors, detectors, indicators......................................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because the generator is operating within the limits of the Generator Capability Curve. It is plausible because it would be correct if power factor were leading or if Gen H2 pressure were lower.

2nd part is correct. The main turbine will automatically trip if the underfrequency limits are exceeded.

Answer B Discussion

Incorrect:1st part is incorrect because the generator is operating within the limits of the Generator Capability Curve. It is plausible because it would be correct if power factor were leading or if Gen H2 pressure were lower. 2nd part is incorrect because the main turbine will automatically trip if the underfrequency limits are exceeded. It is plausible because AP 34 is worded such that you would think that this is a manual action. Step 4.19/20: IAAT the Maximum Allowable Time for a given frequency is reached AND the turbine is on line, THEN perform steps 4.21-4.23-- - - -Trip the reactor if > 50% or open PCB 20 and 21 if < 50%. The frequencies/times given in the AP are the turbine trip setpoints.

Answer C Discussion

CORRECT:1st part is correct. Since MVARS are positive, power factor is lagging and using the upper portion of the Gen Capacity Curve, this value is acceptable.

2nd part is correct. The main turbine will automatically trip if the underfrequency limits are exceeded.

Answer D Discussion

Incorrect:1st part is correct. Since MVARS are positive, power factor is lagging and using the upper portion of the Gen Capacity Curve, this value is acceptable. 2nd part is incorrect because the main turbine will automatically trip if the underfrequency limits are exceeded. It is plausible because AP 34 is worded such that you would think that this is a manual action. Step 4.19/20: IAAT the Maximum Allowable Time for a given frequency is reached AND the turbine is on line, THEN perform steps 4.21-4.23-- - - -Trip the reactor if > 50% or open PCB 20 and 21 if < 50%. The frequencies/times given in the AP are the turbine trip setpoints.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT44 Q16


AP/34 Gen Capacity Curve


ILT44 Q16STG-EHC Obj: 12AP/34EAP AP34


This question matches the KA by requiring the ability to use plant indications and the Generator Capacity Curve to determine if Generator output is acceptable during a grid disturbance. Also required the ability to utilize frequency indicators and predict plant response based on those indications.

Basis for Hi Cog

Basis for SRO only

APE077 AK2.03 - Generator Voltage and Electric Grid DisturbancesKnowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 / 45.8)



Remarks/Status

Sensors, detectors, indicators......................................................




Small Break LOCA has occurred

Rule 2 (Loss of SCM) in progress

The MD EFDW pumps will be used to feed both SGs using Startup FDW Control Valves to the LOSCM Setpoint

The following plant parameters exist:

o SG pressure = 725 psig

o RB temperature = 157°F

o RB pressure = 2.6 psig

Current conditions:

SG level = 275 inches XSUR 1) In accordance with Rule 7 (SG Feed Control), the initial feed rate to be established

to each SG is __(1)__.

2) Currently, SG level __(2)__ in the correct SG level band. Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. 0.15 x 106 lbm/hr

2. is

B. 1. 0.15 x 106 lbm/hr

2. is NOT

C. 1. 0.22 x 106 lbm/hr

2. is

D. 1. 0.22 x 106 lbm/hr

2. is NOT

BWE02 2.1.7 - Vital System Status VerificationBWE02 GENERIC

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)



General Discussion

Answer A Discussion

CORRECT:1st part is correct. With EFDWPs providing flow, Rule 7 directs 0.15 million lbm/hr as an initial feed rate.

2nd part is correct. With the parameters given, the correct SG level band would be 272 to 287 inches XSUR.

Answer B Discussion

Incorrect:1st part is correct. With EFDWPs providing flow, Rule 7 directs 0.15 million lbm/hr as an initial feed rate.

2nd part is incorrect because with the parameters given, the correct SG level band would be 272-287 inches. It is plausible because if the applicant used temperature only (disregarded RB pressure), they would get 280 to 295 inches.

Answer C Discussion

Incorrect:1st part is incorrect because with EFDWPs providing flow, Rule 7 directs 0.15 million lbm/hr as an initial feed rate. It is plausible because if MFW were providing flow, it would be correct.

2nd part is correct. With the parameters given, the correct SG level band would be 272 to 287 inches XSUR.

Answer D Discussion

Incorrect:1st part is incorrect because with EFDWPs providing flow, Rule 7 directs 0.15 million lbm/hr as an initial feed rate. It is plausible because if MFW were providing flow, it would be correct.

2nd part is incorrect because with the parameters given, the correct SG level band would be 272-287 inches. It is plausible because if the applicant used temperature only (disregarded RB pressure), they would get 280 to 295 inches.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


Rule 7, Table 2 & 5


Rule 2 & 7EAP-LOSCM Obj: 19

Remarks/Status


This question matches the KA by requiring the ability to evaluate plant parameters and determine the appropriate SG Levels (Vital System Verification).

Basis for Hi Cog

Basis for SRO only

BWE02 2.1.7 - Vital System Status VerificationBWE02 GENERIC

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)




A loss of ALL sources of Steam Generator feed has occurred

HPI Forced Cooling in progress

RCS pressure = 2210 psig slowly lowering

Pzr Level = 380 inches rising

Core SCM = 56°F rising In accordance with Rule 6 (HPI), HPI flow __(1)__ be throttled because __(2)__. Which ONE of the following completes the statement above? A. 1. may NOT

2. RCS pressure is lowering

B. 1. may NOT

2. CETCs are rising

C. 1. may

2. Pzr Level is rising

D. 1. may

2. CETCs are lowering

BWE04 EK2.2 - Inadequate Heat TransferKnowledge of the interrelations between the (Inadequate HeatTransfer) and the following:(CFR: 41.7 / 45.7)

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the properoperation of these systems to the operation of the facility.



General Discussion

Answer A Discussion

Incorrect:RCS pressure is not criteria used to determine if throttling HPI is allowed. It is plausible if slowly decreasing pressure is used as basis for Core SCM increasing.

Answer B Discussion

Incorrect:CETCs are decreasing. It is plausible if the correlation between increasing Core SCM and slowly decreasing pressure is not recognized as indication that CETC temperatures are decreasing.

Answer C Discussion

Incorrect:Pzr level is not part of the throttling criteria for HPI when in HPI FC. It is plausible because if HPI FC was not in progress, it would be correct.

Answer D Discussion

CORRECT:Criteria for throttling HPI during HPI cooling is based on Core SCM >0 and CETC decreasing. Core SCM increasing with RCS pressure slowly decreasing indicates that CETC temperatures are decreasing.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT46 Q16


ILT46 Q16EOP Rule 6EAP-HPI CD Obj: R3

Remarks/Status


This question matches the KA by requiring knowledge of criteria for throttling HPI during HPI Forced Cooling based on the status of core cooling provided by HPI (CETC trend and Core SCM status)

Basis for Hi Cog

Basis for SRO only

BWE04 EK2.2 - Inadequate Heat TransferKnowledge of the interrelations between the (Inadequate HeatTransfer) and the following:(CFR: 41.7 / 45.7)

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the properoperation of these systems to the operation of the facility.




Reactor power = 70%

Control Rod Group 7 Rod 3 drops into the core

ICS runback actuates

Turbine Control Valves do NOT move in the closed direction

Time = 0812:

Reactor power = 58%

1) During the runback, the automatic runback rate __(1)__ be adjusted.

2) At Time = 0812, the control rods __(2)__ respond to an automatic “withdraw” signal

to correct the Tave error. Which ONE of the following completes the statements above? A. 1. can

2. will

B. 1. can

2. will NOT

C. 1. can NOT

2. will

D. 1. can NOT

2. will NOT

APE003 AK1.02 - Dropped Control RodKnowledge of the operational implications of the following concepts as they apply to Dropped Control Rod: (CFR 41.8 / 41.10 / 45.3)

Effects of turbine-reactor power mismatch on rod control ..............



General Discussion

Answer A Discussion

CORRECT:1st part is correct. The runback signal is to 55% at 1%/minute. This signal can be increased with the thumbwheel on the load control panel.

2nd part is correct. With a dropped control rod, there is a control rod withdraw inhibit signal that is in place until power is < 60%. Because power is < 60% at 0812, the control rods will respond to a withdraw signal. The combination of the dropped control rod and the stuck turbine control valve will make Tave less than setpoint which will call for the withdraw signal.

Answer B Discussion

Incorrect:1st part is correct. The runback signal is to 55% at 1%/minute. This signal can be increased with the thumbwheel on the load control panel.

2nd part is incorrect because with a dropped control rod, there is a control rod withdraw inhibit signal that is in place until power is < 60%. Because power is < 60% at 0812, the control rods will respond to a withdraw signal. The combination of the dropped control rod and the stuck turbine control valve will make Tave less than setpoint which will call for the withdraw signal. It is plausible because if power were > 60%, it would be correct.

Answer C Discussion

Incorrect:1st part is incorrect because the runback can be increased with the thumbwheel on the load control panel. It is plausible because for any other ICS runback, it would be correct.

2nd part is correct. With a dropped control rod, there is a control rod withdraw inhibit signal that is in place until power is < 60%. Because power is < 60% at 0812, the control rods will respond to a withdraw signal. The combination of the dropped control rod and the stuck turbine control valve will make Tave less than setpoint which will call for the withdraw signal.

Answer D Discussion

Incorrect:1st part is incorrect because the runback can be increased with the thumbwheel on the load control panel. It is plausible because for any other ICS runback, it would be correct.

2nd part is incorrect because with a dropped control rod, there is a control rod withdraw inhibit signal that is in place until power is < 60%. Because power is < 60% at 0812, the control rods will respond to a withdraw signal. The combination of the dropped control rod and the stuck turbine control valve will make Tave less than setpoint which will call for the withdraw signal. It is plausible because if power were > 60%, it would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


ICS-01 Obj: 4IC-CRI Obj: 15


This question matches the KA by requiring knowledge of how, after a dropped CR that a turbine reactor power mismatch (Turbine fails to runback) affects CR motion.

Basis for Hi Cog

Basis for SRO only

APE003 AK1.02 - Dropped Control RodKnowledge of the operational implications of the following concepts as they apply to Dropped Control Rod: (CFR 41.8 / 41.10 / 45.3)

Effects of turbine-reactor power mismatch on rod control ..............



Remarks/Status





Control Rod Group 7 Position = 94%

1A Main FDW pump trips Current conditions:

Reactor power = 70%

Control Rod Group 7 Position = 86%

o Group 7 Rod 3 = 94%

o Group 7 Rod 1 = 94% 1) If NO operator action is taken, the plant __(1)__ automatically run back to the

Asymmetric Rod Load Limit setpoint. 2) In accordance with OMP 1-18 (Implementation Standard During Abnormal And

Emergency Events), a manual Reactor trip __(2)__ required. Which ONE of the following completes the statements above? A. 1. will

2. is

B. 1. will

2. is NOT

C. 1. will NOT

2. is

D. 1. will NOT

2. is NOT

APE005 AK2.02 - Inoperable/Stuck Control RodKnowledge of the interrelations between the Inoperable / Stuck Control Rod and the following: (CFR 41.7 / 45.7)

Breakers, relays, disconnects, and control room switches ........................................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because the automatic runback will not occur unless at least one CR is > 6.5% from the group average AND is at 0% or Group IN LIMIT. The > 6.5% criteria is met, but the 0% or group in-limit is not. It is plausible because both stuck Control Rods are > 6.5% out from the group average.

2nd part is correct. Per OMP 1-18, if two CRs are misaligned > 6.5% from the group average, you are directed to trip the Reactor.

Answer B Discussion

Incorrect:1st part is incorrect because the automatic runback will not occur unless at least one CR is > 6.5% from the group average AND is at 0% or Group IN LIMIT. The > 6.5% criteria is met, but the 0% or group in-limit is not. It is plausible because both stuck Control Rods are > 6.5% out from the group average.

2nd part is incorrect because per OMP 1-18, if two CRs are misaligned > 6.5% from the group average, you are directed to trip the Reactor. It is plausible because the OMP 1-18 Manual Reactor Trip criteria is two or more CRs drop into the core OR two or more CRs are misaligned > 9 inches.

Answer C Discussion

CORRECT:1st part is correct. The automatic runback will not occur unless at least one CR is > 6.5% from the group average AND is at 0% or Group IN LIMIT. The > 6.5% criteria is met, but the 0% or group in-limit is not.

2nd part is correct. Per OMP 1-18, if two CRs are misaligned > 6.5% from the group average, you are directed to trip the Reactor.

Answer D Discussion

Incorrect:1st part is correct. The automatic runback will not occur unless at least one CR is > 6.5% from the group average AND is at 0% or Group IN LIMIT. The > 6.5% criteria is met, but the 0% or group in-limit is not.

2nd part is incorrect because per OMP 1-18, if two CRs are misaligned > 6.5% from the group average, you are directed to trip the Reactor. It is plausible because the OMP 1-18 Manual Reactor Trip criteria is two or more CRs drop into the core OR two or more CRs are misaligned > 9 inches.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


OMP 1-18AP/1EAP AP1 Obj: 4


This question matches the KA by requiring the knowledge of 1) when a runback/relay is activated due to a stuck control rod and 2) when the reactor is required to be tripped ( Trip PB / CR Switch) based on stuck control rod position.

Basis for Hi Cog

Basis for SRO only

APE005 AK2.02 - Inoperable/Stuck Control RodKnowledge of the interrelations between the Inoperable / Stuck Control Rod and the following: (CFR 41.7 / 45.7)

Breakers, relays, disconnects, and control room switches ........................................................



Remarks/Status

Ask chief to make sure this is an ok KA match.

DISCUSSED WITH CHIEF 7/27: OK WITH KA MATCH.





1RC-66 (PORV) fails open

1RC-4 (PORV Block Valve) will NOT close Current conditions:

LOCA CD tab is in progress


RCS temperature = 597°F

RB pressure = 2.9 psig rising

Pzr level = 180 inches rising 1) Indicated Pzr level is rising due to __(1)__. 2) In accordance with Rule 6 (HPI) __(2)__ be throttled. Which ONE of the following completes the statements above? A. 1. bubble formation in the reactor vessel

2. can

B. 1. bubble formation in the reactor vessel

2. can NOT

C. 1. rise in containment temperature which causes Pzr level detector reference leg to “flash” to steam

2. can

D. 1. rise in containment temperature which causes Pzr level detector reference leg to “flash” to steam

2. can NOT

APE028 AK3.03 - Pressurizer (PZR) Level Control MalfunctionKnowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: (CFR 41.5,41.10 / 45.6 / 45.13)

False indication of PZR level when PORV or spray valve is open and RCS saturated .........................................



General Discussion

Answer A Discussion

Incorrect:1st part is correct. The given parameters indicate a saturated RCS. As a bubble forms in the reactor vessel, Pzr level will increase due to the displaced water out of the vessel.

2nd part is incorrect because Core SCM must be > 0 degrees to throttle HPI when not in HPI FC. It is plausible because Pzr level is increasing which is part of the throttle criteria.

Answer B Discussion

CORRECT:1st part is correct. The given parameters indicate a saturated RCS. As a bubble forms in the reactor vessel, Pzr level will increase due to the displaced water out of the vessel.

2nd part is correct. Core SCM must be > 0 degrees to throttle HPI when not in HPI FC. Based on the parameters given, the RCS is saturated.

Answer C Discussion

Incorrect:1st part is incorrect because with RCS pressure = 1485 psig, containment temperature would have to raise Pzr reference leg tempertaure to ~ 596 degrees to flash to steam. You are no where near that temperature at 3.4 psig containment pressure. It is plausible because when RCS pressure decreases rapidly, reference leg flashing could occur as RCS pressure drops below Tsat for the reference leg.

2nd part is incorrect because Core SCM must be > 0 degrees to throttle HPI when not in HPI FC. It is plausible because Pzr level is increasing which is part of the throttle criteria.

Answer D Discussion

Incorrect:1st part is incorrect because with RCS pressure = 1485 psig, containment temperature would have to raise Pzr reference leg tempertaure to ~ 596 degrees to flash to steam. You are no where near that temperature at 3.4 psig containment pressure. It is plausible because when RCS pressure decreases rapidly, reference leg flashing could occur as RCS pressure drops below Tsat for the reference leg.

2nd part is correct. Core SCM must be > 0 degrees to throttle HPI when not in HPI FC. Based on the parameters given, the RCS is saturated.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


Rule 6EAP-LCD Obj: R9

Remarks/Status


Question matches the KA by requiring knowledge of the reason that Pzr level will increase when the RCS is at saturated conditions.

Basis for Hi Cog

Basis for SRO only

APE028 AK3.03 - Pressurizer (PZR) Level Control MalfunctionKnowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: (CFR 41.5,41.10 / 45.6 / 45.13)

False indication of PZR level when PORV or spray valve is open and RCS saturated .........................................




Core reload is in progress

All Source Range NIs are operable

NI-1 and NI-3 are designated for refueling Current conditions:

NI-3 loses power 1) NI-3 is powered from __(1)__.

2) In accordance with OP/2/A/1502/007 (Operations Defueling /Refueling

Responsibilities), core alterations __(2)__.

Which ONE of the following completes the statements above? A. 1. 2KVIB

2. can continue with ONLY ONE designated NI

B. 1. 2KVIB

2. can NOT continue until 2 designated NIs are available

C. 1. 2KVIC

2. can continue with ONLY ONE designated NI

D. 1. 2KVIC

2. can NOT continue until 2 designated NIs are available

APE032 2.4.9 - Loss of Source Range Nuclear InstrumentationAPE032 GENERIC

Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)



General Discussion

Answer A Discussion

Incorrect:1st part is correct. NI-3 is powered from KVIB. Source Range NI (NI-1,2,3,4) power supplies are not in order (KVIA,,C,B,D). Power Range NI (NI-5,6,7,8) power supplies are in order (KVIA,C,B,D).

2nd part is incorrect because in the L&P of OP/1/A/1502/007: If the SR NI previously designated at the start of refueling operations becomes inoperable, core alterations will be suspended until an acceptable NI combination is returned to service. It is plausible because you are only required to have 2 Nis operable for core alterations.

Answer B Discussion

CORRECT:1st part is correct. NI-3 is powered from KVIB. Source Range NI (NI-1,2,3,4) power supplies are not in order (KVIA,,C,B,D). Power Range NI (NI-5,6,7,8) power supplies are in order (KVIA,C,B,D).

2nd part is correct. In the L&P of OP/1/A/1502/007: If the SR NI previously designated at the start of refueling operations becomes inoperable, core alterations will be suspended until an acceptable NI combination is returned to service. It is plausible because you are only required to have 2 Nis operable for core alterations.

Answer C Discussion

Incorrect:1st part is incorrect because NI-3 is powered from KVIB. It is plausible because if the power supplies went in order with the Nis, it would be correct.Source Range NI (NI-1,2,3,4) power supplies are not in order (KVIA,,C,B,D). Power Range NI (NI-5,6,7,8) power supplies are in order (KVIA,C,B,D).

2nd part is incorrect because in the L&P of OP/1/A/1502/007: If the SR NI previously designated at the start of refueling operations becomes inoperable, core alterations will be suspended until an acceptable NI combination is returned to service. It is plausible because you are only required to have 2 Nis operable for core alterations.

Answer D Discussion

Incorrect:1st part is incorrect because NI-3 is powered from KVIB. It is plausible because if the power supplies went in order with the NIs, it would be correct.Source Range NI (NI-1,2,3,4) power supplies are not in order (KVIA,,C,B,D). Power Range NI (NI-5,6,7,8) power supplies are in order (KVIA,C,B,D).

2nd part is correct. In the L&P of OP/1/A/1502/007: If the SR NI previously designated at the start of refueling operations becomes inoperable, core alterations will be suspended until an acceptable NI combination is returned to service. It is plausible because you are only required to have 2 Nis operable for core alterations..

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


TS 3.9.2OP/1/A/1502/007


This question matches the KA by requiring knowledge of how a loss of NIs will affect core reload evolution.

Basis for Hi Cog

Basis for SRO only



FH-FHS Obj: 17Vital IC Drawing

Remarks/Status

DISCUSSED WITH CHIEF 7/27: SUGGESTED FUEL HANDLING AND LOSE SR DETECTOR OR AUDIBLE COUNTER.

APE032 2.4.9 - Loss of Source Range Nuclear InstrumentationAPE032 GENERIC

Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)




Reactor trips from 100% Current conditions:

ALL TBVs are in AUTO and closed

Main Steam pressure cycling between 1050 psig and 1070 psig 1) __(1)__ would cause the TBVs to fail closed while in AUTO. 2) In accordance with the EOP, Main Steam pressure will be controlled using __(2)__. Which ONE of the following completes the statements above? A. 1. Condenser vacuum = 6 inches hg

2. TBVs in MANUAL

B. 1. Condenser vacuum = 6 inches hg

2. Atmospheric Dump Valves (ADVs)

C. 1. Selected Turbine Header Pressure failed low

2. TBVs in MANUAL

D. 1. Selected Turbine Header Pressure failed low

2. Atmospheric Dump Valves (ADVs)

APE051 AK3.01 - Loss of Condenser VacuumKnowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum: (CFR 41.5,41.10 / 45.6 / 45.13)

Loss of steam dump capability upon loss of condenser vacuum .........................................................



General Discussion

Answer A Discussion

Incorrect:1st part is correct. With condenser vacuum at 6" Hg, the TBVs will fail closed.

2nd part is incorrect because the TBVs will also fail closed if in MANUAL (from the control room). It is plausible because 1) if vacuum were > 7"Hg, it would be correct and 2) if controlling from the ASP, it would be correct (there is no guidance in the EOP for control of TBVs from the ASP).

Answer B Discussion

CORRECT:1st part is correct. With condenser vacuum at 6" Hg, the TBVs will fail closed.

2nd part is correct. The TBVs will also fail closed if in MANUAL (from the control room). The EOP directs using the TBVs or ADVs. With the TBVs not operable, the ADVs will be used.

Answer C Discussion

Incorrect:1st part is incorrect because after the reactor has tripped, the TBVs are controlled by SG outlet pressure. It is plausible because at power with the turbine in AUTO, Turbine Header Pressure is the pressure used to control the TBV's.

2nd part is correct. Per the SA tab, if TBVs cannot control SG presure at setpoint, then manually control pressure using either TBVs or ADVs. TBVs will be attempted first.

Answer D Discussion

Incorrect:1st part is incorrect because after the reactor has tripped, the TBVs are controlled by SG outlet pressure. It is plausible because at power with the turbine in AUTO, Turbine Header Pressure is the pressure used to control the TBV's.

2nd part is could be correct. Per the SA tab, if TBVs cannot control SG presure at setpoint, then manually control pressure using either TBVs or ADVs. If the TBVs did not work in manual, this would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

2009B Q21


2009B Q21ICS-02 Obj 4EOP SA tab

Remarks/Status


This question matches the KA by requiring knowledge of the reason normal steam pressure control is unavailable during a loss of condenser vacuum.

Basis for Hi Cog

Basis for SRO only

APE051 AK3.01 - Loss of Condenser VacuumKnowledge of the reasons for the following responses as they apply to the Loss of Condenser Vacuum: (CFR 41.5,41.10 / 45.6 / 45.13)

Loss of steam dump capability upon loss of condenser vacuum .........................................................




Reactor in MODE 5 Which ONE of the following describes ALL automatic actions that will occur as a resultof a HIGH alarm on 1RIA-4 (Reactor Building Hatch Monitor)? A. Statalarm ONLY

B. Statalarm AND a Local Alarm ONLY

C. Statalarm AND the RB Evacuation alarm ONLY

D. Statalarm, Local Alarm, AND the RB Evacuation alarm

APE061 AA2.04 - Area Radiation Monitoring (ARM) System AlarmsAbility to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: (CFR: 43.5 / 45.13)

Whether an alarm channel is functioning properly .....................



General Discussion

Answer A Discussion

Incorrect: The H IGH alarm will actuate a Statalarm but will also result in a localalarm and the RB Evacuation alarm.

Answer B Discussion

Incorrect: The H IGH alarm will actuate a Statalarm and a local alarm but will alsoresult in the RB Evacuation alarm.

Answer C Discussion

Incorrect: The H IGH alarm will actuate a Statalarm and the RB Evacuation alarm but will also actuate a local alarm.

Answer D Discussion

CORRECT: the HIGH alarm will result in a statalarm, a local audible horn will sound, and the RB evacuation alarm will be activated.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT40 Q23


ILT40 Q23RAD RIA Obj: 8

Remarks/Status


This question matches the KA by requiring the ability to interpret the response of RIA-4 to determine if the alarm channel is functioning properly (i.e. does the monitor actuate the RB evacuation alarm and since it does, is it actuated by the ALERT alarm or the HIGH alarm).

Basis for Hi Cog

Basis for SRO only

APE061 AA2.04 - Area Radiation Monitoring (ARM) System AlarmsAbility to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: (CFR: 43.5 / 45.13)

Whether an alarm channel is functioning properly .....................




There is an active fire in the cable room

The fire brigade has been dispatched

You are checking your Self-Contained Breathing Apparatus (SCBA) for use

o Cylinder pressure gage = 4500 psig

o Control module gage = 4405 psig Time = 0830:

You are currently combating the fire

The cylinder pressure of your SCBA has lowered to 250 psig 1) At Time = 0800, the SCBA that you have checked __(1)__ acceptable for use.

2) At Time = 0830, the low pressure alarm on your current SCBA __(2)__ be in alarm

due to low pressure. Which ONE of the following completes the statements above? A. 1. is

2. will

B. 1. is

2. will NOT

C. 1. is NOT

2. will

D. 1. is NOT

2. will NOT

APE067 AA1.01 - Plant Fire On SiteAbility to operate and / or monitor the following as they apply to the Plant Fire on Site: (CFR 41.7 / 45.5 / 45.6)

Respirator air pack ...............................................



General Discussion

Answer A Discussion

1st part is correct. The gages have to be within 5% (225 psig) or 100 psig. They are.

2nd part is correct. The alarm comes in when cylinder pressure lowers to 1125 psig.

Answer B Discussion

1st part is correct. The gages have to be within 5% (225 psig) or 100 psig. They are.

2nd part is incorrect because the alarm comes in when cylinder pressure lowers to 1125 psig. It is plausible because as stateted in the lesson, the alarm will remain in alarm until cylinder pressure is 200 psig or less. 200 psig seem like a more logical pressure for a low alarm to come in than 1125 psig.

Answer C Discussion

1st part is incorrect because the gages have to be within 5% (225 psig) or 100 psig. It is plausible because if pressure were 6 psig lower, it would be correct.

2nd part is correct. The alarm comes in when cylinder pressure lowers to 1125 psig.

Answer D Discussion

1st part is incorrect because the gages have to be within 5% (225 psig) or 100 psig. It is plausible because if pressure were 6 psig lower, it would be correct.

2nd part is incorrect because the alarm comes in when cylinder pressure lowers to 1125 psig. It is plausible because as stateted in the lesson, the alarm will remain in alarm until cylinder pressure is 200 psig or less. 200 psig seem like a more logical pressure for a low alarm to come in than 1125 psig.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


Fire Brigade Training Scott Air Pack

Remarks/Status

Request new KA. Can not write a discriminating question at the RO level for this KA.

DISCUSSED WITH CHIEF 7/27: OK WITH LOD FOR THIS QUESTION.


This question matches the KA by requiring the ability to monitor the operation of a respirator air pack (SCBA).

Basis for Hi Cog

Basis for SRO only

APE067 AA1.01 - Plant Fire On SiteAbility to operate and / or monitor the following as they apply to the Plant Fire on Site: (CFR 41.7 / 45.5 / 45.6)

Respirator air pack ...............................................




Turbine Building Flood (TBF) tab initiated

Main and Emergency Feedwater have been lost 1) In accordance with the TBF tab, RCS decay heat will be removed using __(1)__. 2) This heat removal method is chosen because __(2)__. Which ONE of the following completes the statements above? A. 1. HPI Forced Cooling

2. “Raw” lake water will damage the SGs

B. 1. HPI Forced Cooling

2. SSF-ASW suction source is CCW and ALL CCW pumps will be secured

C. 1. SSF-ASW

2. of anticipation of losing Condensate Booster Pumps

D. 1. SSF-ASW

2. of anticipation of losing LPSW pumps

BWA07 AA1.3 - FloodingAbility to operate and / or monitor the following as they apply tothe (Flooding ): (CFR: 41.7 / 45.5 / 45.6)

Desired operating results during abnormal and emergency situations.



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because per a note prior to step 19 in the TBF tab, it states that SSF-ASW is the preferred "long term" source of cooling over HPI FC. It is plausible because if you meet the criteria to perform Rule 4 (HPI FC) in the TBF tab, you are directed to perform it.

2nd part is plausible (if the first part was correct) because it is a valid reason for using HPI FC before SSF-ASW.

Answer B Discussion

Incorrect:1st part is incorrect because per a note prior to step 19 in the TBF tab, it states that SSF-ASW is the preferred "long term" source of cooling over HPI FC. It is plausible because if you meet the criteria to perform Rule 4 (HPI FC) in the TBF tab, you are directed to perform it.

2nd part is plausible (if the first part was correct) because CCW is the suction source for SSF-ASW and CCW is the most likely source of the flooding. There are actions taken to minimize CCW water that can get to the Turbine Building since it is the most probable source of flooding however these action do not isolate the SSF ASWP from its suction source in the CCW inlet piping.

Answer C Discussion

Incorrect:1st part is correct. Per a note prior to step 19 in the TBF tab, it states that SSF-ASW is the preferred "long term" source of cooling over HPI FC.

2nd part is incorrect because SSF -ASW is preferred because of the potential loss of LPSW. It is plausible since CBP feed is normally how heat removal occurs if both Main and Emergency FDW have been lost and the CBP's are located in the Turbine Building Basement and therefore will be impacted by the flooding.

Answer D Discussion

CORRECT:1st part is correct. Per a note prior to step 19 in the TBF tab, it states that SSF-ASW is the preferred "long term" source of cooling over HPI FC.

2nd part is correct. SSF-ASW is the preferred source due to the potential for losing LPSW.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT45 Q26


ILT45 Q26EAP-TBF R5EOP-TBF

Remarks/Status


This question matches the KA by requiring knowledge of the desired lineup (operating results) during an emergency situation.

Basis for Hi Cog

Basis for SRO only

BWA07 AA1.3 - FloodingAbility to operate and / or monitor the following as they apply tothe (Flooding ): (CFR: 41.7 / 45.5 / 45.6)

Desired operating results during abnormal and emergency situations.



Ask chief if this is ok for operating results.

DISCUSSED WITH CHIEF 7/27: CHIEF OK WITH KA MATCH.




A LOCA has occurred

The LOCA CD tab is in progress

Core SCM = 0°F

1SA-02/E-3 (HP LOOP “B” INJECTION FLOW HIGH) is in alarm

1B HPI Header flow = 480 gpm

1) In accordance with Rule 6 (HPI), 1HP-27 (1B HP INJECTION) __(1)__ be throttled. 2) If HPI flow to the “B” Injection Header were through 1HP-409 (1HP-27 Bypass),

1SA-02/E-3 __(2)__ detect a high flow condition. Which ONE of the following completes the statements above? A. 1. must

2. would

B. 1. must

2. would NOT

C. 1. is NOT allowed to

2. would

D. 1. is NOT allowed to

2. would NOT

BWE08 EK1.3 - LOCA CooldownKnowledge of the operational implications of the following concepts asthey apply to the (LOCA Cooldown)(CFR: 41.8 / 41.10 / 45.3)

Annunciators and conditions indicating signals, and remedial actions associated with the ((LOCACooldown)



General Discussion

Answer A Discussion

Incorrect:1st part is correct. The alarm setpoint is 475 gpm which is the maximum flow allowed per Rule 6, so flow in the B HPI header must be throttled.

2nd part is incorrect because the alarm is generated from FT-8A which is located upstream of 1HP-27 so it would not register flow through the crossover header. It is plausible because it is still Loop B injection flow.

Answer B Discussion

CORRECT:1st part is correct. The alarm setpoint is 475 gpm which is the maximum flow allowed per Rule 6, so flow in the B HPI header must be throttled.

2nd part is correct. The alarm is generated from FT-8A which is located upstream of 1HP-27 so it would not register flow through the crossover header.

Answer C Discussion

Incorrect:1st part is incorrect because per Rule 6, flow in the header with one HPIP operating is limited to 475 gpm (the alarm setpoint) so 1HP-27 must be throttled. It is plausible because there is criteria in Rule 6 to allow throttling HPI BUT ONLY IF SCM is > 0 degrees. The applicant has to understand the difference in (applicability of) the two criteria in that preventing the HPIP from running out is the higher priority.

2nd part is incorrect because the alarm is generated from FT-8A which is located upstream of 1HP-27 so it would not register flow through the crossover header. It is plausible because it is still Loop B injection flow.

Answer D Discussion

Incorrect:1st part is incorrect because per Rule 6, flow in the header with one HPIP operating is limited to 475 gpm (the alarm setpoint) so 1HP-27 must be throttled. It is plausible because there is criteria in Rule 6 to allow throttling HPI BUT ONLY IF SCM is > 0 degrees. The applicant has to understand the difference in (applicability of) the two criteria in that preventing the HPIP from running out is the higher priority.

2nd part is correct. The alarm is generated from FT-8A which is located upstream of 1HP-27 so it would not register flow through the crossover header.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


Rule 6PNS-HPI Obj: 21SA-2/E-3LOCA CD tabRule 2


This question matches the KA by requiring knowledge of actions associated with annunciators while in the LOCA CD tab of the EOP.

Basis for Hi Cog

Basis for SRO only

BWE08 EK1.3 - LOCA CooldownKnowledge of the operational implications of the following concepts asthey apply to the (LOCA Cooldown)(CFR: 41.8 / 41.10 / 45.3)



Remarks/Status

Annunciators and conditions indicating signals, and remedial actions associated with the ((LOCACooldown)




Reactor power = 65%

1LPSW-6 (UNIT 1 RCP COOLERS SUPPLY) fails closed Current conditions:

AP/1/A/1700/016 (Abnormal RCP Operation) in progress

RCP Temperatures: 1A1 1A2 1B1 1B2 Radial Bearing 182ºF 227ºF 188ºF 185ºF Temp Seal Return 169ºF 174ºF 227ºF 187ºF Temp Which ONE of the following is required per AP/16? A. Manually trip the Reactor and stop RCP 1A2 ONLY

B. Manually trip the Reactor and stop RCP 1B1 ONLY

C. Stop RCP 1A2 ONLY and verify FDW re-ratios properly

D. Stop RCP 1B1 ONLY and verify FDW re-ratios properly

SYS003 A1.02 - Reactor Coolant Pump System (RCPS)Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: (CFR: 41.5 / 45.5)

RCP pump and motor bearing temperatures ..........................



General Discussion

Answer A Discussion

Incorrect because the reactor does not have to be tripped prior to securing 1A2 RCP. It is plausible because if reactor power were > 70%, it would be correct.

Answer B Discussion

Incorrect because only 1A2 RCP is required to be secured and the reactor is not required to be tripped. It is plausible because if 1B1 Radial bearing temperature were 227 degrees, tripping the pump would be correct.

Answer C Discussion

CORRECT. AP/16 directs that if any RCP meets immediate trip criteria (Enclosure 5.1) and less than 3 RCPs will be remain operating, then manually trip the Rx and immediately stop the affected RCPs only. Immediate trip criteria for Radial Bearing temp of 225 is exceeded for 1A2 and Seal Return temp limit of 260 is not exceeded for 1B1.

Answer D Discussion

Incorrect because you do not meet the trip criteria for the 1B1 RCP. It is plausible because if its Radial Bearing Temp were 227 degrees, it could be correct (if RCP 1A2 did not meet the trip criteria).

Cognitive Level

Memory

Job Level

RO

QuestionType

MODIFIED

Question Source

ILT39 Q28


ILT39 Q28AP/16EAP AP16 Obj: 4

Remarks/Status


This question matches the KA by requiring the ability to monitor RCP motor parameters and determine how many pumps exceed temperature limits of AP/16. The limits of AP/16also must be known by the student.

Basis for Hi Cog

Basis for SRO only

SYS003 A1.02 - Reactor Coolant Pump System (RCPS)Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: (CFR: 41.5 / 45.5)

RCP pump and motor bearing temperatures ..........................





1B1 RCP upper seal completely fails

1) Seal return flow will __(1)__.

2) Differential pressure (P) across the middle seal will be approximately __(2)__. Which ONE of the following completes the statements above? A. 1. rise

2. 700 psid

B. 1. rise

2. 1050 psid

C. 1. lower

2. 700 psid

D. 1. lower

2. 1050 psid

SYS003 A3.03 - Reactor Coolant Pump System (RCPS)Ability to monitor automatic operation of the RCPS, including: (CFR: 41.7 / 45.5)

Seal D/P ........................................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because seal return flow will decrease. It is plausible because if either the middle or lower seal failed, it would be correct.

2nd part is incorrect because the DP should be ~ 1050 psid. It is plausible because if no seals had failed, it would be correct.

Answer B Discussion

Incorrect:1st part is incorrect because seal return flow will decrease. It is plausible because if either the middle or lower seal failed, it would be correct.

2nd part is correct. With one seal failed, the 2100 psid across the entire seal package is broken down in 2 steps or 1050 psid across each seal.

Answer C Discussion

Incorrect:1st part is correct. If the upper seal fails then the leakoff from the middle seal will go to seal leakage instead of seal return and seal return flow will decrease.

2nd part is incorrect because the DP should be ~ 1050 psid. It is plausible because if no seals had failed, it would be correct.

Answer D Discussion

CORRECT:1st part is correct. If the upper seal fails then the leakoff from the middle seal will go to seal leakage instead of seal return and seal return flow will decrease.

2nd part is correct. With one seal failed, the 2100 psid across the entire seal package is broken down in 2 steps or 1050 psid across each seal.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


PNS-CPS Obj: 8

Remarks/Status


This question matches the KA by requiring knowledge of the normal operation of RCP seals to determine indictions of a seal failure.

Basis for Hi Cog

Basis for SRO only

SYS003 A3.03 - Reactor Coolant Pump System (RCPS)Ability to monitor automatic operation of the RCPS, including: (CFR: 41.7 / 45.5)

Seal D/P ........................................................




Time = 0300


Reactor trip

CT-3 lockout occurs Current conditions:

MFBs re-energized

6900V power still unavailable

HPI system leak downstream of 3HP-31 occurs

3A1 RCP SI flow = 3.9 gpm slowly lowering

3A2 RCP SI flow = 3.7 gpm slowly lowering

3B1 RCP SI flow = 3.5 gpm slowly lowering

3B2 RCP SI flow = 3.4 gpm slowly lowering

Seal Inlet Header Flow = 40 gpm stable Which ONE of the following describes the status of the following RCP support system valve(s) two minutes later? A. ONLY 3HP-21 has closed

B. ONLY 3HP-21 AND 3HP-31 have closed

C. ONLY 3HP-21 and ALL individual seal return valves have closed

D. 3HP-21, 3HP-31, and ALL individual seal return valves have closed

SYS004 K3.04 - Chemical and Volume Control SystemKnowledge of the effect that a loss or malfunction of the CVCS will have on the following: (CFR: 41.7/45/6)

RCPS ..........................................................



General Discussion

Answer A Discussion

Incorrect:Incorrect because 3HP-21 is not the only seal valve that has closed. Each individual RCP seal return valve is also closed. All 4 of them being closed is what causes 3HP-21 to close. It is plausible because if 1HP-21 closes, the individual RCP seal return valves on Unit 1 do not close.

Answer B Discussion

Incorrect:Incorrect because 3HP-21 is not the only seal valve that has closed. Each individual RCP seal return valve is also closed. All 4 of them being closed is what causes 3HP-21 to close. 3HP-31will close if SI flow to each pump is < 4 gpm for > 1 minute. It is plausible because if 1HP-21 closes, the individual RCP seal return valves on Unit 1 do not close.

Answer C Discussion

Incorrect:Incorrect because 3HP-31 will close after 1 minute with individual seal flows < 4 gpm. Plausible since all of these valves would be closed and it is plausible to believe 3HP-31 would still be open since it would be true for Unit 1.

Answer D Discussion

CORRECT:With seal injection to each RCP < 4 gpm with the RCP off, each RCP seal return valve will close. When all 4 valves close, 3HP-21 will close. When seal flow is < 4 gpm for > 1 minute, 3HP-31 closes.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

2010A Q28


2010A Q28PNS-HPI Obj: 6

Remarks/Status


This question matches the KA by requiring knowledge of the effects that a seal injection malfunction will have on RCPs (seal return).

Basis for Hi Cog

Basis for SRO only

SYS004 K3.04 - Chemical and Volume Control SystemKnowledge of the effect that a loss or malfunction of the CVCS will have on the following: (CFR: 41.7/45/6)

RCPS ..........................................................





1A and 1B Letdown Coolers in service Current conditions:

1A Letdown Cooler is removed from service by the RO in the control room 1) CC flow to the 1B Letdown Cooler will __(1)__. 2) If letdown temperature rises, control rods will __(2)__ as a result of the

reactivity change. Which ONE of the following completes the statements above? A. 1. stay the same

2. insert

B. 1. stay the same

2. withdraw

C. 1. rise

2. insert

D. 1. rise

2. withdraw

SYS004 K5.15 - Chemical and Volume Control SystemKnowledge of the operational implications of the following concepts as they apply to the CVCS: (CFR: 41.5/45.7)

Boron and control rod reactivity effects as they relate to MTC ..........



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because HPI and CC are isolated when a letdown cooler is removed from service, the CC flow to the in service cooler will increase. CC is a closed loop system. It is plausible because if only HPI were isolated from the cooler, it would be correct.

2nd part is incorrect because increasing Letdown temperature will cause the demin to release Boron to the RCS causing a negative reactivity addition and control rods withdrawing. It is plausible because it is a common misconception that increasing IX inlet temperature will cause IX outlet boron concentration to decrease which would cause CRs to insert.

Answer B Discussion

Incorrect:1st part is incorrect because HPI and CC are isolated when a letdown cooler is removed from service, the CC flow to the in service cooler will increase. CC is a closed loop system. It is plausible because if only HPI were isolated from the cooler, it would be correct.

2nd part is correct. Increasing Letdown temperature will cause the demin to release Boron to the RCS causing a negative reactivity addition and control rods withdrawing.

Answer C Discussion

Incorrect:1st part is correct. Because HPI and CC are isolated when a letdown cooler is removed from service, the CC flow to the in service cooler will increase. CC is a closed loop system.

2nd part is incorrect because increasing Letdown temperature will cause the demin to release Boron to the RCS causing a negative reactivity addition and control rods withdrawing. It is plausible because it is a common misconception that increasing IX inlet temperature will cause IX outlet boron concentration to decrease which would cause CR to insert.

Answer D Discussion

CORRECT:1st part is correct. Because HPI and CC are isolated when a letdown cooler is removed from service, the CC flow to the in service cooler will increase. CC is a closed loop system.

2nd part is correct. Increasing Letdown temperature will cause the demin to release Boron to the RCS causing a negative reactivity addition and control rods withdrawing.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

IT45 Q7


ILT45 Q7PNS-CC Obj: 5, 10

Remarks/Status


This question matches the KA by requiring knowledge of how changing letdown alignment can effect reactivity and how that change in reactivity will effect plant components (control rods) due to the change in moderator temperature (moderator temperature coefficient).

Basis for Hi Cog

Basis for SRO only

SYS004 K5.15 - Chemical and Volume Control SystemKnowledge of the operational implications of the following concepts as they apply to the CVCS: (CFR: 41.5/45.7)

Boron and control rod reactivity effects as they relate to MTC ..........




RCS cooldown in progress

2B LPI cooler isolated due to cooler leak 1) The LPI Decay Heat Removal mode that will be used for the INITIAL transition to

LPI cooling is __(1)__. 2) The HIGHER RCS pressure (psig) that will allow aligning LPI in the Normal Decay

Heat Removal mode is __(2)__. Which ONE of the following completes the statements above? A. 1. Switchover

2. 115

B. 1. Switchover

2. 220

C. 1. High Pressure

2. 115

D. 1. High Pressure

2. 220

SYS005 K6.03 - Residual Heat Removal System (RHRS)Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7)

RHR heat exchanger ..............................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because with the 2B LPI cooler out of service, switchover mode will not be used. It is plausible because this would be correct if the 2A cooler were not available instead of the 2B cooler.

2nd part is correct. RCS pressure has to be reduced to < 125 psig to align to normal DHR mode.

Answer B Discussion

Incorrect:1st part is incorrect because with the 2B LPI cooler out of service, switchover mode will not be used. It is plausible because this would be correct if the 2A cooler were not available instead of the 2B cooler.

2nd part is incorrect because RCS pressure has to be reduced < 125 psig to align for normal DHR mode. It is plausible since it would be correct if asked on Unit 3.

Answer C Discussion

CORRECT:1st part is correct. High pressure mode will be used with the 2B LPI cooler out of service,

2nd part is correct. RCS pressure has to be reduced < 125 psig to align to normal DHR mode.

Answer D Discussion

Incorrect:1st part is correct. High pressure mode will be used with the 2B LPI cooler out of service,

2nd part is incorrect because RCS pressure has to be reduced < 125 psig to align for normal DHR mode. It is plausible since it would be correct if asked on Unit 3..

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT44 Q30


ILT44 Q30PNS-LPI Obj 13, 35

Remarks/Status


This question matches the KA by requiring knowledge of the effect that a loss of one of the DHR coolers will have on available DHR alignments.

Basis for Hi Cog

Basis for SRO only

SYS005 K6.03 - Residual Heat Removal System (RHRS)Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: (CFR: 41.7 / 45.7)

RHR heat exchanger ..............................................




Reactor power = 20% stable

Small Break LOCA occurs Time = 1210:

RCS pressure = 410 psig slowly lowering

RB pressure = 2.7 psig slowly rising Which ONE of the following contains ONLY valves that have received an Engineered Safeguards signal to OPEN at Time = 1210? A. 1HP-24 and 1BS-1

B. 1HP-24 and 1LP-17

C. 1LP-18 and 1BS-1

D. 1LP-18 and 1LPSW-15

SYS006 A3.06 - Emergency Core Cooling System (ECCS)Ability to monitor automatic operation of the ECCS, including: (CFR: 41.7 / 45.5)

Valve lineups ....................................................



General Discussion

Answer A Discussion

Incorrect:Incorrect because 1BS-1 will not have received an ES signal to open yet. 1HP-24 receives a signal to open when ES channel 1 actuates. 1BS-1 receives a signal to open when ES channel 7 actuates (10 psig in containment). It is plausible because ES channels 1-4 have actuated.

Answer B Discussion

CORRECT:With RCS pressure below 550 psig ES 1-4 have actuated. Since RB pressure is below 3 psig, ES 5-8 have not actuated. 1HP-24 opens on ES-1 and 1LP-17 opens on ES-3. Therefore they have both received a signal to open.

Answer C Discussion

Incorrect:Incorrect because 1BS-1 will not have received an ES signal to open yet. 1LP-18 receives a signal to open when ES channel 4 actuates. 1BS-1 receives a signal to open when ES channel 7 actuates (10 psig in containment). It is plausible because ES channels 1-4 have actuated.

Answer D Discussion

Incorrect:Incorrect because 1LPSW-15 will not have received an ES signal to open yet. 1LP-18 receives a signal to open when ES channel 4 actuates. 1LPSW-15 receives a signal to open when ES channel 5 or 6 actuates (3 psig in containment). It is plausible because ES channels 1-4 have actuated.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT40 Q32


ILT40 Q32IC-ES R18 Obj: 4

Remarks/Status


This question matches the KA by requiring the ability to monitor proper valve alignment based on auto actuation of ECCS systems.

Basis for Hi Cog

Basis for SRO only

SYS006 A3.06 - Emergency Core Cooling System (ECCS)Ability to monitor automatic operation of the ECCS, including: (CFR: 41.7 / 45.5)

Valve lineups ....................................................




OP/1/A/1103/002 (Filling and Venting RCS) Enclosure 4.14 (Establishing Pzr

Steam Bubble And RCS Final Vent) in progress

Quench Tank level = 82 inches

Quench Tank pressure = 0.5 psig

The Pressurizer is vented to the Quench Tank for 30 minutes Which ONE of the following describes QT parameters that would indicate Pzr Steam Bubble Formation is complete?

QT level (inches) QT pressure (psig) A. 82.1 0.6

B. 84.1 0.6

C. 82.1 2.5

D. 84.1 2.5

SYS007 2.2.44 - Pressurizer Relief Tank/Quench Tank System (PRTS)SYS007 GENERIC

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)



General Discussion

Answer A Discussion

Incorrect:Incorrect because when the Pzr has steam as opposed to N2 being vented, the steam will condense in the QT which will cause level in the QT to increase but have little effect on QT pressure. It is plausible if you have the misconception that as the steam condenses it will not contribute to QT level.

Answer B Discussion

CORRECT: Per OP/1103/002, Pzr steam bubble formation is complete (i.e., all the N2 gas is vented out of the Pzr) when a change (rise) in QT pressure of less than 0.2 psig occurs and QT level increases by 2 inches. Since the Pzr vent is underwater in the QT, when N2 is being vented it will rise to the surface and cause a corresponding increase in QT pressure, therefore minimal pressure response is a sign that all of the N2 has been vented. Additionally, as water is vented it is condensed under the water level of the QT, therefore minimal QT pressure change in conjunction with increasing QT level is indicative of all N2 being out of Pzr.

Answer C Discussion

Incorrect:Incorrect because when the Pzr has steam as opposed to N2 being vented, the steam will condense in the QT which will cause level in the QT to increase but have little effect on QT pressure. It is plausible if you have the misconception that the steam intruduction into the QT will increase pressure.

Answer D Discussion

Incorrect:Incorrect because when the Pzr has steam as opposed to N2 being vented, the steam will condense in the QT which will cause level in the QT to increase but have little effect on QT pressure. It is plausible if you have the misconception that the steam intruduction into the QT will increase pressure.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT45 Q34


ILT45 Q34PNS-PZR Obj. 23OP/1/A/1103/002, Encl. 4.10 pg 6 and 8

Remarks/Status


This question matches the KA by requiring the ability to interpret control room indications to determine the status of Pzr bubble formation.

Basis for Hi Cog

Basis for SRO only

SYS007 2.2.44 - Pressurizer Relief Tank/Quench Tank System (PRTS)SYS007 GENERIC






1A CC Pump switch in ON

1B CC Pump switch in AUTO

A Steam line break occurs

RB pressure = 3.5 psig rising

1) 1CC-7 and 8 (Component Cooling Return Penetration Block Valves) will close as a result of ES Channels __(1)__ actuating.

2) When ES has been reset and 1CC-7 and 8 are opened, __(2)__ CC Pump(s) will

automatically start. Which ONE of the following completes the statements above? A. 1. 1 and 2

2. one

B. 1. 1 and 2

2. both

C. 1. 5 and 6

2. one

D. 1. 5 and 6

2. both

SYS008 A3.08 - Component Cooling Water System (CCWS)Ability to monitor automatic operation of the CCWS, including: (CFR: 41.7 / 45.5)

Automatic actions associated with the CCWS that occur as a result of a safety injection signal ................................



General Discussion

Answer A Discussion

Incorrect:1st part is incorrect because CC 7 & 8 close upon receiving ES Channel 5 & 6 actuation. It is plausible because with RB pressure = 3.5 psig, ES1-6 will actuate which will close 1CC-7 & 8.

2nd part is incorrect because both CC pumps will start. It is plausible because with the switches in different positions, it would be logical to think that either 1) there is a time delay with the pump that is in AUTO that has to time out before starting or 2) the switch positioned to ON would have to be taken to OFF then back to AUTO or ON to start it (this is common reset procedure).

Answer B Discussion

Incorrect:1st part is incorrect because CC 7 & 8 close upon receiving ES Channel 5 & 6 actuation. It is plausible because with RB pressure = 3.5 psig, ES1-6 will actuate which will close 1CC-7 & 8.

2nd part is correct. Both CC pumps will start. The switch for CC Pump 1A is already in ON so it will start. CC Pump 1B will recognize CC flow < 575 gpm as soon as CC-7 & 8 are open (this is an automatic start signal).

Answer C Discussion

Incorrect:1st part is correct. CC 7 & 8 close upon receiving ES Channel 5 & 6 actuation.

2nd part is incorrect because both CC pumps will start. It is plausible because with the switches in different positions, it would be logical to think that either 1) there is a time delay with the pump that is in AUTO that has to time out before starting or 2) the switch positioned to ON would have to be taken to OFF then back to AUTO or ON to start it (this is common reset procedure).

Answer D Discussion

CORRECT:1st part is correct. CC 7 & 8 close upon receiving ES Channel 5 & 6 actuation.

2nd part is correct. Both CC pumps will start. The switch for CC Pump 1A is already in ON so it will start. CC Pump 1B will recognize CC flow < 575 gpm as soon as CC-7 & 8 are open (this is an automatic start signal).

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


PNS-CC Obj: 15IC-ES

Remarks/Status

9/13/16 - Discussed with Dan Bacon following his review of the 10 preview questions. Added "automatically" in the second part per Dan.

Ask chief about KA match. The CC system at ONS does not have interlocks associated with a safety injection (ES 1/2). It does have it with ES 5/6 (RB Press = 3 psig).


This question matches the KA by requiring knowledge of automatic actions that occur in the CC system upon receiving an ES actuation.

Basis for Hi Cog

Basis for SRO only

SYS008 A3.08 - Component Cooling Water System (CCWS)Ability to monitor automatic operation of the CCWS, including: (CFR: 41.7 / 45.5)

Automatic actions associated with the CCWS that occur as a result of a safety injection signal ................................



DISCUSSED WITH CHIEF 7/27: MAY BE OK IF WE CANT FIND ANYTHING THAT CHANGES CC FLOW TO ANYTHING WITH ES 1 & 2.





1RC-66 (PORV) is leaking past its seat


Quench tank pressure = 5 psig Because steam being throttled as it passes through the PORV is a constant __(1)__ process, the PORV tailpipe temperature will be approximately __(2)__. Which ONE of the following completes the statement above? A. 1. enthalpy

2. 228°F

B. 1. enthalpy

2. 648°F

C. 1. entropy

2. 228°F

D. 1. entropy

2. 648°F

SYS010 K5.02 - Pressurizer Pressure Control System (PZR PCS)Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: (CFR: 41.5 / 45.7)

Constant enthalpy expansion through a valve ..........................



General Discussion

Answer A Discussion

1st part is correct. The throttling process is an isenthalpic one.

2nd part is correct. At 2165 psig (2180 psia), the enthalpy is 1122 Btu/lbm. 1122 Btu/lbm at 5 psig (20 psia) constitutes a wet vapor. That being said, the temperature of the steam will be the saturation temperature for 20 psia or 228 degrees.

Answer B Discussion

1st part is correct. The throttling process is an isenthalpic one.

2nd part is incorrect because at 2165 psig (2180 psia), the enthalpy is 1122 Btu/lbm. 1122 Btu/lbm at 5 psig (20 psia) constitutes a wet vapor. That being said, the temperature of the steam will be the saturation temperature for 20 psia or 228 degrees. It is plausible because the saturization temperaure of the steam in the Pzr at 2265 psig is 648 degrees.

Answer C Discussion

1st part is incorrect because the throttling process is an isenthalpic one. It is plausible because other ideal processes in the thermal cycle (turbine and pumps) are isentropic processes.

2nd part is correct. At 2165 psig (2180 psia), the enthalpy is 1122 Btu/lbm. 1122 Btu/lbm at 5 psig (20 psia) constitutes a wet vapor. That being said, the temperature of the steam will be the saturation temperature for 20 psia or 228 degrees.

Answer D Discussion

1st part is incorrect because the throttling process is an isenthalpic one. It is plausible because other ideal processes in the thermal cycle (turbine and pumps) are isentropic processes.

2nd part is incorrect because at 2165 psig (2180 psia), the enthalpy is 1122 Btu/lbm. 1122 Btu/lbm at 5 psig (20 psia) constitutes a wet vapor. That being said, the temperature of the steam will be the saturation temperature for 20 psia or 228 degrees. It is plausible because the saturization temperaure of the steam in the Pzr at 2265 psig is 648 degrees.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


PNS-PZR Obj: 28

Remarks/Status


The question matches the KA by requiring knowledge of the operational implications (tailpiece temperature) as a result of constant enthalpy expansion through a valve (PORV).

Basis for Hi Cog

Basis for SRO only

SYS010 K5.02 - Pressurizer Pressure Control System (PZR PCS)Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: (CFR: 41.5 / 45.7)

Constant enthalpy expansion through a valve ..........................





1RC-1 (Pzr Spray) fails OPEN Time = 0405:


RCS pressure = 2115 psig rising Time = 0415:

RCS pressure = 2155 psig rising

Pzr temperature = 640ºF rising 1) At Time = 0405, the requirements of LCO 3.4.1 (RCS Pressure, Temperature, and

Flow Departure from Nucleate Boiling (DNB) Limits) __(1)__ met. 2) At Time = 0415, Pzr Heater Bank 2 (Groups B and D) are __(2)__. Which ONE of the following completes the statements above? A. 1. are NOT

2. de-energized

B. 1. are NOT

2. energized

C. 1. are

2. de-energized

D. 1. are

2. energized

SYS010 K6.02 - Pressurizer Pressure Control System (PZR PCS)Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: (CFR: 41.7 / 45.7)

PZR ...........................................................



General Discussion

Answer A Discussion

1st part is correct. The requirements of LCO 3.4.1 are not met because RCS pressure is below the COLR value of 2125 psig.

2nd part is incorrect because Pzr Heater Bank 2 Groups B and D will be energized due to the Pzr Saturation recovery Circuit. The Pzr is subcooled and Pzr Psat is greater than 20 psig below actual RCS pressure. Saturation pressure for 640 = 2060 psia = 2045 psig. It is plausible because RCS pressure is in the normal operating window and above the Bank #2 cutoff at 2150 psig.

Answer B Discussion

1st part is correct. The requirements of LCO 3.4.1 are not met because RCS pressure is below the COLR value of 2125 psig.

2nd part is correct. Pzr Heater Bank 2 Groups B and D will be energized due to the Pzr Saturation recovery Circuit. The Pzr is subcooled and Pzr Psat is greater than 20 psig below actual RCS pressure. Saturation pressure for 640 = 2060 psia = 2045 psig.

Answer C Discussion

1st part is incorrect because the requirements of LCO 3.4.1 are not met because RCS pressure is below the COLR value of 2125 psig. It is plausible because if pressure were 10 psig higher, it would be correct.

2nd part is incorrect because Pzr Heater Bank 2 Groups B and D will be energized due to the Pzr Saturation recovery Circuit. The Pzr is subcooled and Pzr Psat is greater than 20 psig below actual RCS pressure. Saturation pressure for 640 = 2060 psia = 2045 psig. It is plausible because RCS pressure is in the normal operating window and above the Bank #2 cutoff at 2150 psig.

Answer D Discussion

1st part is incorrect because the requirements of LCO 3.4.1 are not met because RCS pressure is below the COLR value of 2125 psig. It is plausible because if pressure were 10 psig higher, it would be correct.

2nd part is correct. Pzr Heater Bank 2 Groups B and D will be energized due to the Pzr Saturation recovery Circuit. The Pzr is subcooled and Pzr Psat is greater than 20 psig below actual RCS pressure. Saturation pressure for 640 = 2060 psia = 2045 psig.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT41 Q36


ILT41 Q36ADM-ITS R8PNS-PZR Obj: 12, 13TS 3.4.1Unit 1 COLR

Remarks/Status


This question matches the KA by requiring knowledge of how a Pzr Press Control malfunction affects the Pzr equipment.

Basis for Hi Cog

Basis for SRO only

SYS010 K6.02 - Pressurizer Pressure Control System (PZR PCS)Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: (CFR: 41.7 / 45.7)

PZR ...........................................................





BOTH Main Feedwater Pumps trip Current conditions:

Reactor power = 57% slowly lowering

1HP-24 & 1HP-25 will NOT open 1) In order to inject Boron into the RCS, Rule 1 (ATWS) will direct __(1)__. 2) The direction given to the operator opening the CRD breaker is to __(2)__

Arc Flash PPE. Which ONE of the following completes the statements above? A. 1. using CBAST

2. wear

B. 1. using CBAST

2. NOT wear

C. 1. using the BWST via the LPI system

2. wear

D. 1. using the BWST via the LPI system

2. NOT wear

SYS012 A2.06 - Reactor Protection System (RPS)Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Failure of RPS signal to trip the reactor .............................



General Discussion

Answer A Discussion

1st part is incorrect because Rule 1 will direct using a piggyback lineup to inject boron into the RCS. It is plausible because if the event was initiated during Mode 3, it would be correct.

2nd part is incorrect because Rule 1 directs the operator NOT to wear Flash PPE when opening CRD Breakers. It is plausible because the normal expectation is to wear Arc Flash PPE when operating a 600V breaker. Without the specific direction NOT to wear the PPE the outside operator may take unnecessary time to don this PPE.

Answer B Discussion

1st part is incorrect because Rule 1 will direct using a piggyback lineup to inject boron into the RCS. It is plausible because if the event was initiated during Mode 3, it would be correct.

Second part is correct. Rule 1directs the operator NOT to wear Arc Flash PPE when opening CRD Breakers.

Answer C Discussion

1st part is correct. Rule 1 will direct using a piggyback lineup to inject boron into the RCS.

2nd part is incorrect because Rule 1 directs the operator NOT to wear Flash PPE when opening CRD Breakers. It is plausible because the normal expectation is to wear Arc Flash PPE when operating a 600V breaker. Without the specific direction NOT to wear the PPE the outside operator may take unnecessary time to don this PPE.

Answer D Discussion

1st part is correct. Rule 1 will direct using a piggyback lineup to inject boron into the RCS.

Second part is correct. Rule 1directs the operator NOT to wear Arc Flash PPE when opening CRD Breakers.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT39 Q67


ILT39 Q67EAP-UNPP R8Rule 1

Remarks/Status


This question matches the KA by requiring knowledge of actions taken to mitigate a failure of the RPS to trip the reactor.

Basis for Hi Cog

Basis for SRO only

SYS012 A2.06 - Reactor Protection System (RPS)Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Failure of RPS signal to trip the reactor .............................




Small Break LOCA occurs

Reactor Building pressure = 4.5 psig rising

RCS pressure = 1500 psig lowering Current conditions:

Reactor Building pressure = 2.5 psig stable


ES reset is desired 1) Depress RESET for ES Channels __(1)__ to allow the HPI pumps ES logic

to be reset. 2) Based on current conditions, the Diverse HPI “Bistable Tripped” light __(2)__

be illuminated. Which ONE of the following completes the statements above? A. 1. 1 and 2

2. will

B. 1. 1 and 2

2. will NOT

C. 1. 3 and 4

2. will

D. 1. 3 and 4

2. will NOT

SYS013 A4.02 - Engineered Safety Features Actuation System (ESFAS)Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Reset of ESFAS channels .........................................



General Discussion

Answer A Discussion

1st part is correct. HPIPs are on ES channel 1 & 2.

2nd part is incorrect because the Diverse HPI "Bistable Tripped" light is off when the bistable automatically clears as RCS pressure increases above 1600 psig. It is plausible because ES logic will have to be manually reset.

Answer B Discussion

1st part is correct. HPIPs are on ES channel 1 & 2.

2nd part is correct. The Diverse HPI "Bistable Tripped" light is off because the bistable automatically clears when RCS pressure is above 1600 psig.

Answer C Discussion

1st part is incorrect because HPIPs are on ES channel 1 & 2. It is plausible because if it were LPI pumps, it would be correct.

2nd part is incorrect because the Diverse HPI "Bistable Tripped" light is off when the bistable automatically clears as RCS pressure increases above 1600 psig. It is plausible because ES logic will have to be manually reset.

Answer D Discussion

1st part is incorrect because HPIPs are on ES channel 1 & 2. It is plausible because if it were LPI pumps, it would be correct.

2nd part is correct. The Diverse HPI "Bistable Tripped" light is off because the bistable automatically clears when RCS pressure is above 1600 psig.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT43 Q55


ILT43 Q55EP/1/A/1800/001 Encl. 5.41 Rev. 39IC-ES Obj: 15

Remarks/Status


This question matches the KA by requiring knowledge of the process for resetting ES Channels.

Basis for Hi Cog

Basis for SRO only

SYS013 A4.02 - Engineered Safety Features Actuation System (ESFAS)Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Reset of ESFAS channels .........................................





1KVIA Panelboard de-energized Current conditions:

MSLB inside the Reactor Building occurs

Lowest RCS pressure = 1137 psig

Reactor Building pressure peaked at 32 psig Which ONE of the following describes ALL ES Actuation Logic Channels that have actuated? A. 1, 3, 5, 7

B. 2, 4, 6, 8

C. 1, 5, 7 ONLY

D. 2, 6, 8 ONLY

SYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS)Knowledge of bus power supplies to the following: (CFR: 41.7)

ESFAS/safeguards equipment control ...............................



General Discussion

Answer A Discussion

Incorrect because KVIA provides power to odd digital channels. With KVIA de-energized, the Odd channels cannot actuate. It is plausible because if it were KVIB, it would be correct.

Answer B Discussion

Correct. KVIA provides power to the odd digital channels. With KVIA de-energized, the Odd channels cannot actuate. Since RB pressure has exceeded 10 psig, all channels 8 channels would receive an actuation signal however only the odd channels have power and therefore they are all that can actuate.

Answer C Discussion

Incorrect because ES Channels 2,4,6 & 8 would actuate. It is plausible because RCS pressure has reached the Low RCS pressure setpoint for HPI injection but has not reached the LPI injection setpoint of 550 psig. The misconception that HPI and LPI only actuate from RCS pressure rather than from either RCS pressure OR RB pressure would lead to believing that channels 2 and 4 had not yet received an actuation signal. Additionally, the power supply's to the Actuation Logic channels is split based on odd and even channels. Channel 1&2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.

Answer D Discussion

Incorrect because ES Channels 2,4,6 & 8 will actuate. It is plausible since RCS pressure has reached the Low RCS pressure setpoint for HPI injection but has not reached the LPI injection setpoint of 550 psig. The misconception that HPI and LPI only actuate from RCS pressure rather than from either RCS pressure OR RB pressure would lead to believing that channels 2 and 4 had not yet received an actuation signal. Additionally, the power supply's to the Actuation Logic channels is split based on odd and even channels. Channel 1&2 RCS pressure setpoint has already been reached therefore under the misconception that ES channels 3 and 4 (LPI) only actuate on low RCS pressure this is plausible.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT44 Q65


ILT44 Q65IC-ES Obj: 2

Remarks/Status


This queston matches the KA by requiring knowledge of ES powers supplies for the equipment that controls actuation of the components.

Basis for Hi Cog

Basis for SRO only

SYS013 K2.01 - Engineered Safety Features Actuation System (ESFAS)Knowledge of bus power supplies to the following: (CFR: 41.7)

ESFAS/safeguards equipment control ...............................





1A and 1C RBCUs operating in HIGH speed Time = 0401:

ES Channels 1 - 8 actuate Time = 0405:

1A and 1B RBCUs in LOW speed

1C RBCU in HIGH speed 1) Continued operation in this alignment could damage the 1C RBCU fan __(1)__ due

to high humidity in the RB. 2) In accordance with EOP Enclosure 5.1 (ES Actuation) the 1C RBCU will __(2)__. Which ONE of the following completes the statements above? A. 1. motor

2. immediately be switched to LOW speed

B. 1. motor

2. ONLY be switched to LOW speed after directed by the CRS

C. 1. blades

2. immediately be switched to LOW speed

D. 1. blades

2. ONLY be switched to LOW speed after directed by the CRS

SYS022 K3.01 - Containment Cooling System (CCS)Knowledge of the effect that a loss or malfunction of the CCS will have on the following: (CFR: 41.7 / 45.6)

Containment equipment subject to damage by high or low temperature, humidity, and pressure .................................



General Discussion

Answer A Discussion

1st part is correct. The high humidity air in containment is denser and will put more strain on the RBCU motor.

2nd part is incorrect because per Encl 5.1, if any ES Channel 5 & 6 components are not in their ES position, the RO is to notify the SRO to evaluate the component and initiate action to place in ES position if desired. It is plausible because normally it is expected to start equipment that failed to initiate automatically.

Answer B Discussion

1st part is correct. The high humidity air in containment is denser and will put more strain on the RBCU motor.

2nd part is correct. Per Encl 5.1, if any ES Channel 5 & 6 components are not in their ES position, the RO is to notify the SRO to evaluate the component and initiate action to place in ES position if desired.

Answer C Discussion

1st part is incorrect because the high humidity air in containment is denser and will put more strain on the RBCU motor. It is plausible to think that the high humidity would cause impingement on the fan blades in the same manner that high moisture steam can imping the turbine blading.

2nd part is incorrect because per Encl 5.1, if any ES Channel 5 & 6 components are not in their ES position, the RO is to notify the SRO to evaluate the component and initiate action to place in ES position if desired. It is plausible because normally it is expected to start equipment that failed to initiate automatically.

Answer D Discussion

1st part is incorrect because the high humidity air in containment is denser and will put more strain on the RBCU motor. It is plausible to think that the high humidity would cause impingement on the fan blades in the same manner that high moisture steam can imping the turbine blading.

2nd part is correct. Per Encl 5.1, if any ES Channel 5 & 6 components are not in their ES position, the RO is to notify the SRO to evaluate the component and initiate action to place in ES position if desired.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT42 Q40


ILT42 Q40PNS-RBC Obj: 6EOP Encl. 5.1 Rev 1

Remarks/Status


This question matches the KA by requiring knowledge of the potential damage to equipment if one RBCU remaining in high speed after ES actuation (due to high humidity) and the procedural guidance on how to correct the issue.

Basis for Hi Cog

Basis for SRO only

SYS022 K3.01 - Containment Cooling System (CCS)Knowledge of the effect that a loss or malfunction of the CCS will have on the following: (CFR: 41.7 / 45.6)

Containment equipment subject to damage by high or low temperature, humidity, and pressure .................................




A Large Break LOCA occurs Time = 0800:

EOP Enclosure 5.12 (ECCS Suction Swap to RBES) is in progress

BWST level = 8.5 ft lowering Time = 0900:

BWST level = 5.5 ft lowering In accordance with Encl 5.12….. 1) 1LP-19 and 1LP-20 will INITIALLY be opened at time __(1)__. 2) If 1LP-19 fails to open when INITIALLY attempted, __(2)__ will be secured.

Which ONE of the following completes the statements above? A. 1. 0800

2. 1A RBS Pump

B. 1. 0800

2. 1A LPI Pump

C. 1. 0900

2. 1A RBS Pump

D. 1. 0900

2. 1A LPI Pump

SYS026 A2.07 - Containment Spray System (CSS)Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit



General Discussion

Answer A Discussion

1st part is correct. In Enclosure 5.12, WHEN BWST level is less than or equal to 9 ft, and RB level is rising, it directs you to simultaneously open 1LP-19 & 20.

2nd part is correct. Per Encl 5.12, IF 1LP-19 fails to open, it directs you to stop the 1A RPS Pump.

Answer B Discussion

1st part is correct. In Enclosure 5.12, WHEN BWST level is less than or equal to 9 ft, and RB level is rising, it directs you to simultaneously open 1LP-19 & 20.

2nd part is incorrect because per Encl 5.12, IF 1LP-19 fails to open, it directs you to stop the 1A RPS Pump. It is plausible because if BWST lever were < 6 ft, it would be correct.

Answer C Discussion

1st part is incorrect because in Enclosure 5.12, WHEN BWST level is less than or equal to 9 ft, and RB level is rising, it directs you to simultaneously open 1LP-19 & 20. It is plausible because when BWST level is < 6 ft, it has you verify that 1LP-19 & 20 are open, then it has you close the suction valves to the VCT.

2nd part is correct. Per Encl 5.12, IF 1LP-19 fails to open, it directs you to stop the 1A RPS Pump.

Answer D Discussion

1st part is incorrect because in Enclosure 5.12, WHEN BWST level is less than or equal to 9 ft, and RB level is rising, it directs you to simultaneously open 1LP-19 & 20. It is plausible because when BWST level is < 6 ft, it has you verify that 1LP-19 & 20 are open, then it has you close the suction valves to the VCT.

2nd part is incorrect because per Encl 5.12, IF 1LP-19 fails to open, it directs you to stop the 1A RPS Pump. It is plausible because if BWST lever were < 6 ft, it would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


Encl 5.12PNS-RBS Obj: 10


This question matches the KA by requiring knowledge of procedure steps taken when the suction of a Building Spray pump fails to open.

Basis for Hi Cog

Basis for SRO only

SYS026 A2.07 - Containment Spray System (CSS)Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit



Remarks/Status

Send as one of 10 questions to NRC.




Unit 1 startup in progress

Reactor power = 15%

The Main Generator is being paralleled to the grid per OP/1/A/1106/001 (Turbine Generator)

Turbine Master is in MANUAL When PCB-20 or PCB 21 (Generator Breaker) is closed….. 1) the Turbine Bypass Valves will be controlled by __(1)__ error.

2) you are directed to establish __(2)__. Which ONE of the following completes the statements above? A. 1. SG Outlet Header Pressure

2. approximately 35 MWe

B. 1. SG Outlet Header Pressure

2. a minimum of 90 MWe

C. 1. Turbine Header Pressure

2. approximately 35 MWe

D. 1. Turbine Header Pressure

2. a minimum of 90 MWe

SYS039 2.2.44 - Main and Reheat Steam System (MRSS)SYS039 GENERIC




General Discussion

Answer A Discussion

1st part is correct. When the Generator Breakers are closed, the Turbine Master is in MANUAL. After paralleling is complete, the Turbine Master is placed in AUTO at which time THP error is used to control TBVs. Then the turbine is "Loaded".

2nd part is correct. When the Generator Output breaker is closed, OP/1106/001 directs you to establish ~ 35 Mwe.

Answer B Discussion

1st part is correct. When the Generator Breakers are closed, the Turbine Master is in MANUAL. After paralleling is complete, the Turbine Master is placed in AUTO at which time THP error is used to control TBVs. Then the turbine is "Loaded".

2nd part is incorrect because when the Generator Output breaker is closed, OP/1106/001 directs you to establish ~ 35 Mwe. It is plausible because when the Generator is "Loaded" (Load PB is depressed and the Generator Load Status Flag is TRUE), it would be at least 90 Mwe.

Answer C Discussion

1st part is incorrect because until the Turbine Master is placed in AUTO, SG Outlet pressure error is used as input to the TBVs. It is plausible because it is a common misunderstanding that the error signal switches to THP when the Turbine is Loaded and that the Turbine is loaded when you parallel it with the grid and pick up load.

2nd part is correct. When the Generator Output breaker is closed, OP/1106/001 directs you to establish ~ 35 Mwe.

Answer D Discussion

1st part is incorrect because until the Turbine Master is placed in AUTO, SG Outlet pressure error is used as input to the TBVs. It is plausible because it is a common misunderstanding that the error signal switches to THP when the Turbine is Loaded and that the Turbine is loaded when you parallel it with the grid and pick up load.

2nd part is incorrect because when the Generator Output breaker is closed, OP/1106/001 directs you to establish ~ 35 Mwe. It is plausible because when the Generator is "Loaded" (Load PB is depressed and the Generator Load Status Flag is TRUE), it would be at least 90 Mwe.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


OP/1/A/1106/001 Turbine StartupICS-02 Obj: 4


This question matches the KA by requiring knowledge of 1) how operator actions (paralleling the generator) affect plant conditions and 2) how control room indications (Mwe) are used for turbine loading.

Basis for Hi Cog

Basis for SRO only

SYS039 2.2.44 - Main and Reheat Steam System (MRSS)SYS039 GENERIC




Remarks/Status




Reactor power = 80% stable

Tc Controller is in HAND

1B1 RCP trips

Crew performs Plant Transient Response

Crew enters AP/01 (Unit Runback)

Tc = +1.2°F and becoming more positive The operator will have to manually re-ratio feedwater such that feed to the __(1)__ SGwill rise because the RC Flow Ratio circuit ___(2)___. Which ONE of the following completes the statement above? A. 1. 1A

2. has failed

B. 1. 1A

2. is blocked when the Delta Tc controller is in HAND

C. 1. 1B

2. has failed

D. 1. 1B

2. is blocked when the Delta Tc controller is in HAND

SYS059 A4.10 - Main Feedwater (MFW) SystemAbility to manually operate and monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

ICS ............................................................



General Discussion

Answer A Discussion

1st part is correct. DTc = Tc (a loop) - Tc(b loop). If the re-ratio does not occur when the 1B1 RCP trips, the B SG has too much feed/steam flow which cools off the B loop. This will be seen as a + DTc.

2nd part is correct. The RC Flow Ratio circuit taps in down stream of the DTc controller so the controller being in HAND will not stop the RC Flow Ratio circuit from providing input to the summer. If this does not happen when RC flow changes, the RC Flow Ratio circuit has failed.

Answer B Discussion

1st part is correct. DTc = Tc (a loop) - Tc(b loop). If the re-ratio does not occur when the 1B1 RCP trips, the B SG has too much feed/steam flow which cools off the B loop. This will be seen as a + DTc.

2nd part is incorrect because its output taps in down stream of the DTc controller. It is plausible because it, as well as the DTc controller provides input to the re-ratio summer. It would be logical to think that its input would be controlled by the DTc controller.

Answer C Discussion

1st part is incorrect. DTc = Tc (a loop) - Tc(b loop). If the re-ratio does not occur when the 1B1 RCP trips, the B SG has too much feed/steam flow which cools off the B loop. This will be seen as a + DTc. It is plausible because it a common misconception to think that a + DTc means that the A SG needs less feedwater.

2nd part is correct. The RC Flow Ratio circuit taps in down stream of the DTc controller so the controller being in HAND will not stop the RC Flow Ratio circuit from providing input to the summer. If this does not happen when RC flow changes, the RC Flow Ratio circuit has failed.

Answer D Discussion

1st part is incorrect. DTc = Tc (a loop) - Tc(b loop). If the re-ratio does not occur when the 1B1 RCP trips, the B SG has too much feed/steam flow which cools off the B loop. This will be seen as a + DTc. It is plausible because it a common misconception to think that a + DTc means that the A SG needs less feedwater.

2nd part is incorrect because its output taps in down stream of the DTc controller. It is plausible because it, as well as the DTc controller provides input to the re-ratio summer. It would be logical to think that its input would be controlled by the DTc controller.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT47 Q24


ILT47 Q24AP/1ICS-03 Obj: 4ICS Feedwater Power Point

Remarks/Status


Question matches the KA by requiring knowledge of how to manipulate ICS to control feedwater.

Basis for Hi Cog

Basis for SRO only

SYS059 A4.10 - Main Feedwater (MFW) SystemAbility to manually operate and monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

ICS ............................................................





Both Main FDW pumps trip

Current conditions:

Enclosure 5.9 (Extended EFDW Operation) is in progress

UST level = 4 ft lowering

Preparations are being made to swap EFDW suction to the Hotwell 1) In accordance with Enclosure 5.9, when UST level = 2.9 ft, the MD EFDWPs will be

__(1)__. 2) When taking suction from the Hotwell, in accordance with Rule 7 (SG Feed

Control), each MD EFDWP is limited to a MAXIMUM flow rate of __(2)__. Which ONE of the following completes the statements above? A. 1. secured

2. 440 gpm

B. 1. secured

2. 600 gpm

C. 1. operating

2. 440 gpm

D. 1. operating

2. 600 gpm

SYS061 K1.05 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge of the physical connections and/or cause-effect relationships between the AFW and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Condensate system ...............................................



General Discussion

Answer A Discussion

1st part is incorrect because suction sources are not swapped until the UST level lowers to below 1 ft. It is plausible because vacuum is broken when UST level lowers to below 3 ft in preparation to swap suction sources.

2nd part is correct. When on the Hotwell, EFW flow is limited to 440 gpm.

Answer B Discussion

1st part is incorrect because suction sources are not swapped until the UST level lowers to below 1 ft. It is plausible because vacuum is broken when UST level lowers to below 3 ft in preparation to swap suction sources.

2nd part is incorrect because when on the Hotwell, EFW flow is limited to 440 gpm. It is plausible because if still on the UST, it would be correct.

Answer C Discussion

1st part is correct. Suction sources are not swapped until the UST level lowers to below 1 ft. At that time, the MD EFDW pumps are stopped while suctions are swapped.

2nd part is correct. When on the Hotwell, EFW flow is limited to 440 gpm.

Answer D Discussion

1st part is correct. Suction sources are not swapped until the UST level lowers to below 1 ft. At that time, the MD EFDW pumps are stopped while suctions are swapped.

2nd part is incorrect because when on the Hotwell, EFW flow is limited to 440 gpm. It is plausible because if still on the UST, it would be correct.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


Enclosure 5.9Rule 7CF-EF Obj: 6

Remarks/Status


This question matches the KA by requiring knowledge of the relationship between the EFW system and the condensate system.

Basis for Hi Cog

Basis for SRO only

SYS061 K1.05 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge of the physical connections and/or cause-effect relationships between the AFW and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Condensate system ...............................................




Reactor power = 100%.

Loss of all feedwater occurs

CBP feed is being established 1) In accordance with Rule 3 (Loss of Main or Emergency Feedwater), once CBP feed

is established, RCS temperature __(1)__. 2) When manipulating the Feedwater Startup Control Valves, “Two Handed” valve

operation __(2)__ allowed. Which ONE of the following completes the statements above? A. 1. will be stabilized at its current value

2. is

B. 1. will be stabilized at its current value

2. is NOT

C. 1. will lower to Tsat for the SG pressure

2. is

D. 1. will lower to Tsat for the SG pressure

2. is NOT

SYS061 K3.01 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge of the effect that a loss or malfunction of the AFW will have on the following: (CFR: 41.7 / 45.6)

RCS ...........................................................



General Discussion

Answer A Discussion

1st part is correct. Per Rule 3, Step 14 RNO: 1) Lower SG pressure in available SGs to ~ 500 psig. 2) control FDW flow to stabilize RCS P/T by throttling ..

2nd part is correct. Per OMP 1-2, Rules of Practice, there is a list of components where non-procedure driven simultaneous component manipulation is allowed. ICS Startup Control Valves are on that list.

Answer B Discussion

1st part is correct. Per Rule 3, Step 14 RNO: 1) Lower SG pressure in available SGs to ~ 500 psig. 2) control FDW flow to stabilize RCS P/T by throttling ..

2nd part is incorrect because per OMP 1-2, Rules of Practice, there is a list of components where non-procedure driven simultaneous component manipulation is allowed. ICS Startup Control Valves are on that list. It is plausible because this is a short list and most of the valves in the control room are not on this list.

Answer C Discussion

1st part is incorrect because per Rule 3, Step 14 RNO: 1) Lower SG pressure in available SGs to ~ 500 psig. 2) control FDW flow to stabilize RCS P/T by throttling .. . Its is plausible because if a SG level is established, it would be correct.

2nd part is correct. Per OMP 1-2, Rules of Practice, there is a list of components where non-procedure driven simultaneous component manipulation is allowed. ICS Startup Control Valves are on that list.

Answer D Discussion

1st part is incorrect because per Rule 3, Step 14 RNO: 1) Lower SG pressure in available SGs to ~ 500 psig. 2) control FDW flow to stabilize RCS P/T by throttling .. . Its is plausible because if a SG level is established, it would be correct.

2nd part is incorrect because per OMP 1-2, Rules of Practice, there is a list of components where non-procedure driven simultaneous component manipulation is allowed. ICS Startup Control Valves are on that list. It is plausible because this is a short list and most of the valves in the control room are not on this list.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


OMP 1-2, Rules of PracticeRule 3EAP-LOHT Obj: R6, R7


This question matches the KA by requiring knowledge of how a loss of AFW will affect the RCS (temperatue / cooldown) when CBP feed is established.

Basis for Hi Cog

Basis for SRO only

SYS061 K3.01 - Auxiliary / Emergency Feedwater (AFW) SystemKnowledge of the effect that a loss or malfunction of the AFW will have on the following: (CFR: 41.7 / 45.6)

RCS ...........................................................



Remarks/Status





1KI Breaker #1 is inadvertently opened

1SA-2/B-11 (ICS AUTO POWER FAILURE) actuates 1) Pzr level will __(1)__ as a result of the power failure. 2) 1HP-120 can now be operated in __(2)__. Which ONE of the following completes the statements above? A. 1. stay the same

2. manual ONLY

B. 1. stay the same

2. manual OR automatic

C. 1. lower

2. manual ONLY

D. 1. lower

2. manual OR automatic

SYS062 A1.03 - AC Electrical Distribution SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5)

Effect on instrumentation and controls of switching power supplies ..................................................



General Discussion

Answer A Discussion

1st part is correct. 1HP-120 will remain in AUTO upon a loss of auto power and maintain Pzr level at setpoint.

2nd part is incorrect because 1HP-120 can be controlled in HAND or AUTO. It is plausible because most of the ICS controls are not operable in automatic when ICS AUTO POWER is lost.

Answer B Discussion

1st part is correct. 1HP-120 will remain in AUTO upon a loss of auto power and maintain Pzr level at setpoint.

2nd part is correct. 1HP-120 can be controlled in HAND or AUTO.

Answer C Discussion

1st part is incorrect because 1HP-120 will remain in AUTO upon a loss of auto power and maintain Pzr level at setpoint. It is plausible because if 1HP-120 were to fail closed upon a loss of power, it would be correct.

2nd part is incorrect because 1HP-120 can be controlled in HAND or AUTO. It is plausible because most of the ICS controls are not operable in automatic when ICS AUTO POWER is lost.

Answer D Discussion

1st part is incorrect because 1HP-120 will remain in AUTO upon a loss of auto power and maintain Pzr level at setpoint. It is plausible because if 1HP-120 were to fail closed upon a loss of power, it would be correct.

2nd part is correct. 1HP-120 can be controlled in HAND or AUTO.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT41 Q45


ILT41 Q45ICS-05 Obj: 5AP/23

Remarks/Status


This question matches the KA by requiring knowledge of how 1HP-120 will operate when its AC power supply is swapped. 1HP-120 is important to ensure that the Pzr level is kept within the limits of TS.

Basis for Hi Cog

Basis for SRO only

SYS062 A1.03 - AC Electrical Distribution SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: (CFR: 41.5 / 45.5)

Effect on instrumentation and controls of switching power supplies ..................................................





ACB-4 Closed

Large Break LOCA occurs coincident with a loss of offsite power Time = 1205:

KHU-2 Emergency Lockout occurs Which ONE of the following describes how ECCS systems are being powered at Time = 1206? ASSUME NO OPERATOR ACTIONS A. KHU-1 through the Overhead power path

B. KHU-1 through the underground power path

C. CT-5 powered from Lee combustion turbine

D. CT-5 powered from Central Switchyard

SYS062 K4.03 - AC Electrical Distribution SystemKnowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Interlocks between automatic bus transfer and breakers ................



General Discussion

Answer A Discussion

Correct. ACB-4 closed indicates that KHU-2 is aligned to the underground power path. Following the LOCA/LOOP, MFB would be energized by KHU-2 through the underground power path. With a subsequent loss of KHU-2, retransfer to startup logic would transfer power to the overhead power path which would be supplied by KHU-1.

Answer B Discussion

Incorrect (see A explanation). Plausible for the following reasons.If candidate does not know which ACB is for the overhead power path and which one is for the underground.If candidate does not understand Power Switching Logic it is plausible to believe that PSL would align KHU-1 to the underground if KHU-2 is lost since there are conditions which result in the overhead KHU auto aligning to the underground.

Answer C Discussion

Incorrect (see A explanation). Plausible since this would be correct if Lee were energizing the Standby Buses prior to the LOCA/LOOP.

Answer D Discussion

Incorrect (see A explanation). Plausible since this would be correct if Central were energizing the Standby Buses prior to the LOCA/LOOP.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT46 Q49


ILT46 Q49EL-PSL Obj: 11

Remarks/Status


This question matches the KA by requiring knowledge of electrical interlocks that determine bus transfer logic operation.

Basis for Hi Cog

Basis for SRO only

SYS062 K4.03 - AC Electrical Distribution SystemKnowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Interlocks between automatic bus transfer and breakers ................



Given the following plant conditions: Time = 1230

All three units operating at 50% power

1TC is de-energized

KHU-1 is NOT running

KHU-1 is aligned to the underground Time = 1245:

PCB-8 and PCB-9 trip open and lock out

ES channel 1 and 2 Keowee Emergency Start Signal has been received ASSUME NO OPERATOR ACTION At Time = 1250, KHU-1 control power (via Bus 1DA) is __(2)__. Which ONE of the following completes the statement above? A. being supplied by Battery #1 ONLY

B. being supplied by Battery Charger #1 via Load Center 1X

C. NOT available due to 1TC being de-energized

D. NOT available due to PCB 8 and 9 being open

SYS063 K4.02 - DC Electrical Distribution SystemKnowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Breaker interlocks, permissives, bypasses and cross-ties. ...............



General Discussion

Answer A Discussion

Correct. Because power to 1X is not available, the only source of control power is the battery.

Answer B Discussion

Incorrect because Battery 1 is the only source of control power. It is plausible because it would be true if the unit was not connected to the underground feeder and a Switchyard Isolation had occurred.

Answer C Discussion

Incorrect because Battery 1 is supplying power. It is plausible because the 1TC is a source of power to the Battery charger.

Answer D Discussion

Incorrect because Battery 1 is supplying power. It is plausible because the Switchyard via PCB-8 and 9 is a source of power to the Battery charger.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT43 Q49


ILT43 Q49EL-KHG Obj: 18EL-DCD Figure 9

Remarks/Status


This question matches the KA by requiring knowledge of the control power (DC) interlocks / crossties of the KHU system

Basis for Hi Cog

Basis for SRO only

SYS063 K4.02 - DC Electrical Distribution SystemKnowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Breaker interlocks, permissives, bypasses and cross-ties. ...............




Reactor in MODE 5

RB Purge is in progress

Reactor Building Airborne activity is rising 1SA-8/D-9 (RM Reactor BLDG Purge Disch RAD Inhibit) will actuate __(1)__ the “Switchover” from 1RIA-45 to 1RIA-46 and will close __(2)__. Which ONE of the following completes the statement above? A. 1. prior to

2. 1PR-2 through 1PR-5 ONLY

B. 1. prior to

2. 1PR-1 through 1PR-6

C. 1. after

2. 1PR-2 through 1PR-5 ONLY

D. 1. after

2. 1PR-1 through 1PR-6

SYS073 K1.01 - Process Radiation Monitoring (PRM) SystemKnowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Those systems served by PRMs ....................................



General Discussion

Answer A Discussion

1st part correct. Both RIA-45 and RIA-46 will cause the isolation however, with no equipment failures, RIA-45 will cause the isolation before the switchover to RIA-46 occurs.

2nd part is correct. RIA-45 will only cause PR-2 through PR-5 to isolate.

Answer B Discussion

1st part correct. Both RIA-45 and RIA-46 will cause the isolation however, with no equipment failures, RIA-45 will cause the isolation before the switchover to RIA-46 occurs.

2nd part is incorrect because 1PR-1 and 1PR-6 do not close when the alarm is received. It is plausible because 1PR-1 through 1PR-6 do isolate on an ES signal.

Answer C Discussion

1st part is incorrect because the isolation will occur will occur before the switchover occurs. It is plausible because RIA-46 does perform the same function as RIA-45 and the switchover will occur as levels continue to rise. RIA-46 causing the isolation is a backup to RIA-45 in case it fails to isolate RB Purge.

2nd part is correct. RIA-45 will only cause PR-2 through PR-5 to isolate.

Answer D Discussion

1st part is incorrect because the isolation will occur will occur before the switchover occurs. It is plausible because RIA-46 does perform the same function as RIA-45 and the switchover will occur as levels continue to rise. RIA-46 causing the isolation is a backup to RIA-45 in case it fails to isolate RB Purge.

2nd part is incorrect because 1PR-1 and 1PR-6 do not close when the alarm is received. It is plausible because 1PR-1 through 1PR-6 do isolateon an ES signal

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT47 Q51


ILT47 Q51RAD RIA Obj: 81SA8 D9

Remarks/Status


This question matches the KA by requiring knowledge of the interactions between PRMs and their associated systems.

Basis for Hi Cog

Basis for SRO only

SYS073 K1.01 - Process Radiation Monitoring (PRM) SystemKnowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Those systems served by PRMs ....................................




Reactor power = 100% Which ONE of the following describes an effect of a loss of LPSW? ASSUME NO OPERATOR ACTIONS A. The HPI pump motors would overheat

B. The Primary IA Compressor would overheat

C. In a subsequent LOHT, the TDEFDW pump will overheat

D. In a subsequent LOCA, RCS cooling would be reduced during the recirculation phase

SYS076 K1.19 - Service Water System (SWS)Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

SWS emergency heat loads ........................................



General Discussion

Answer A Discussion

Incorrect because while the HPIPs are cooled by LPSW (which makes it plausible), HPSW will automatically provide cooling to the HPI pumps.

Answer B Discussion

Incorrect because LPSW does not provide cooling to the Primary IA compressor. It is plausible because LPSW does provide cooling for many loads in the Turbine Building such as the Backup IA compressors. The Primary IA Compressor is cooled by HPSW.

Answer C Discussion

Incorrect because LPSW does not provide cooling to the TD EFDWPs. It is plausible because it would be correct for the MDEFDW pumps.

Answer D Discussion

Correct. Although RB Spray is not directly cooled by LPSW it would be affected in this case due to the lack of cooling of the RBES contents by the LPI Coolers during the recirc phase.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT41 Q51


ILT41 Q51SSS-LPW R16

Remarks/Status


This question matches the KA by requiring knowledge of the effect that a loss of LPSW would have on heat loads (LPI HX during a LOCA).

Basis for Hi Cog

Basis for SRO only

SYS076 K1.19 - Service Water System (SWS)Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

SWS emergency heat loads ........................................



Which ONE of the following states all of the switchgear that can supply power to the B LPSW pump? A. 1TD only

B. 2TC only

C. 1TC or 2TC

D. 1TD or 2TD

SYS076 K2.01 - Service Water System (SWS)Knowledge of bus power supplies to the following: (CFR: 41.7)

Service water ....................................................



General Discussion

Answer A Discussion

Incorrect because it can be powered from 2TD as well. It is plausible because it can be powered from 1TD.

Answer B Discussion

Incorrect because the B LPSW pump can be powered from 1TD or 2TD. It is plausible because it would be correct for the C LPSW pump.

Answer C Discussion

Incorrect because the B LPSW pump can be powered from 1TD or 2TD. It is plausible because they both supply power to the LPSW pumps; A and C respectively.

Answer D Discussion

Correct. The B LPSW pump can be supplied power from 1TD or 2TD.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT41 Q50


ILT41 Q50SSS-LPW Obj: 8

Remarks/Status


This question matches the KA by requiring knowledge of the power supply for one of the LPSW pumps.

Basis for Hi Cog

Basis for SRO only

SYS076 K2.01 - Service Water System (SWS)Knowledge of bus power supplies to the following: (CFR: 41.7)

Service water ....................................................



Which ONE of the following is the power supply for Backup (Worthington) Instrument Air Compressor “B”? A. 1XF

B. 1XP

C. 3X1

D. 3X3

SYS078 K2.02 - Instrument Air System (IAS)Knowledge of bus power supplies to the following: (CFR: 41.7)

Emergency air compressor ........................................



General Discussion

Answer A Discussion

Correct. 1XF is the power supply to the "B" Backup IA Compressor.

Answer B Discussion

Incorrect because 1XF is the power supply to the "B" Backup IA Compressor. It is plausible because 1XP is the power supply to the Unit 1 AIA Compressor.

Answer C Discussion

Incorrect because 1XF is the power supply to the "B" Backup IA Compressor. It is plausible because 3X1 is the power supply to one of the Sulair Air Compressors.

Answer D Discussion

Incorrect because 1XF is the power supply to the "B" Backup IA Compressor. It is plausible because 3X1 is the power supply to one of the Sulair Air Compressors.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


SSS-IA Obj: 2

Remarks/Status


This question matches the KA by requiring knowledge of the power supply for the AIA compressors.

Basis for Hi Cog

Basis for SRO only

SYS078 K2.02 - Instrument Air System (IAS)Knowledge of bus power supplies to the following: (CFR: 41.7)

Emergency air compressor ........................................




Mode 1

You are dispatched to swap the operating Main Steam Line Cooling Fans for the East Penetration Room from Fan ‘A’ to Fan ‘B’

1) The control panel for the fans is located __(1)__ the East Penetration Room. 2) If the local breaker for Fan ‘B’ trips after starting, Fan ‘A’ __(2)__. Which ONE of the following completes the statements above? A. 1. inside

2. will start immediately due to Fan ‘B’ breaker position

B. 1. inside

2. will start after a time delay due to low flow in the ductwork

C. 1. outside

2. will start immediately due to Fan ‘B’ breaker position

D. 1. outside

2. will start after a time delay due to low flow in the ductwork

SYS103 2.1.30 - Containment SystemSYS103 GENERIC

Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)



General Discussion

Answer A Discussion

1st part is correct. The fan control panel for the East Penetration Room is located inside the East Pen Room.

2nd part is incorrect because the fan will start after a low flow conditions exists in the ductwork for 30 seconds. It is plausible because either condition will cause 1SA9/D4, MS 1A PEN. PRIMARY COOL FAN FAIL to alarm. The first manual action is to verify that the standby fan is operating.

Answer B Discussion

1st part is correct. The fan control panel for the East Penetration Room is located inside the East Pen Room.

2nd part is correct. Fan A will start after a low flow conditions exists in the ductwork for 30 seconds.

Answer C Discussion

1st part is incorrect because the fan control panel for the East Penetration Room is located inside the East Pen Room. It is plausible because if it were the West Penetration Room Fans, it would be correct.

2nd part is incorrect because the fan will start after a low flow conditions exists in the ductwork for 30 seconds. It is plausible because either condition will cause 1SA9/D4, MS 1A PEN. PRIMARY COOL FAN FAIL to alarm. The first manual action is to verify that the standby fan is operating.

Answer D Discussion

1st part is incorrect because the fan control panel for the East Penetration Room is located inside the East Pen Room. It is plausible because if it were the West Penetration Room Fans, it would be correct.

2nd part is correct. Fan A will start after a low flow conditions exists in the ductwork for 30 seconds.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


1SA9/D4STG-MS Obj: 2

Remarks/Status


This question matches the KA by requiring knowledge of the location of containment related control panels and how they operate.

Basis for Hi Cog

Basis for SRO only

SYS103 2.1.30 - Containment SystemSYS103 GENERIC

Ability to locate and operate components, including local controls. (CFR: 41.7 / 45.7)





Main Steam Line Break occurs

ES Channel 2 actuates at the designed setpoint

ES Channel 1 fails to actuate Which ONE of the following describes a valve that has received a signal to CLOSE? A. 1GWD-12

B. 1LWD-1

C. 1HP-21

D. 1LPSW-6

SYS103 K4.06 - Containment SystemKnowledge of containment system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Containment isolation system ......................................



General Discussion

Answer A Discussion

Incorrect 1GWD-12 is on ES Channel 1 so it would NOT have received a close signal. It is plausible because if ES Channel 1 had actuated, it would be correct. Also, it is plausible to think that the Even numbered valves would be associated with the even ES channel (some are, but not this one).

Answer B Discussion

Incorrect because 1LWD-1 is associated with ES Channel 1 so it did not receive a signal to close. It is plausible because if ES Channel 1 had actuated, it would be correct. Some of the valves assoaciated with the Odd ES channel have odd numbers and some of them have even numbers.

Answer C Discussion

CORRECT: 1HP-21 is on ES channel 2. With RCS pressure below the ES channel 1 actuation setpoint for RCS pressure (1600 psig) ES 2 will have actuated and sent a close signal to 1HP-21 for non essential containment isolation.

Answer D Discussion

Incorrect because1LPSW-6 is not associated with ES Channel 2. It is plausible because if RB pressure had reached 3 psig, it will have actuated from ES Channel 5 and ES Channel 6 (its on both).

Cognitive Level

Comprehension

Job Level

RO

QuestionType

MODIFIED

Question Source

ILT44 Q38


ILT44 Q38IC-ES R14, R18

Remarks/Status


This question matches the KA by requiring knowledge of containment isolation system.

Basis for Hi Cog

Basis for SRO only

SYS103 K4.06 - Containment SystemKnowledge of containment system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Containment isolation system ......................................





Main Turbine Trip occurs 1) If CRD Breakers A & C remain CLOSED, the Control Rod Drive Mechanisms

__(1)__ remain energized. 2) The RCS pressure setpoint that would actuate the Diverse Scram System (DSS) to

de-energize the CRDMs is __(2)__. Which ONE of the following completes the statements above? A. 1. will NOT

2. 2345 psig

B. 1. will NOT

2. 2450 psig

C. 1. will

2. 2345 psig

D. 1. will

2. 2450 psig

SYS001 K6.03 - Control Rod Drive SystemKnowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Reactor trip breakers, including controls .............................



General Discussion

Answer A Discussion

1st part is incorrect because if CRD Breakers A & C fail to open, the CRDM will remain energized. It is plausible because if it were CRD Breakers A & D or A & B, it would be correct.

2nd part is incorrect because the DSS signal is actuated at 2450 psig. It is plausible because 2345 psig is the RPS high pressure trip signal.

Answer B Discussion

1st part is incorrect because if CRD Breakers A & C fail to open, the CRDM will remain energized. It is plausible because if it were CRD Breakers A & D or A & B, it would be correct.

2nd part is correct. The DSS signal is actuated at 2450 psig.

Answer C Discussion

1st part is correct. If CRD breakers A & C failed to open, CRDMs would still be energized.

2nd part is incorrect because the DSS signal is actuated at 2450 psig. It is plausible because 2345 psig is the RPS high pressure trip signal.

Answer D Discussion

1st part is correct. If CRD breakers A & C failed to open, CRDMs would still be energized.

2nd part is correct. The DSS signal is actuated at 2450 psig.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


IC-RPS Obj: 6

Remarks/Status


This question matches the KA by requiring knowledge of how a failure of reactor trip breakers would effect the CRDMs and how the RPS would compensate for the failure.

Basis for Hi Cog

Basis for SRO only

SYS001 K6.03 - Control Rod Drive SystemKnowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7)

Reactor trip breakers, including controls .............................





RBNS level as shown and has been rising

Time = 0820:

RBNS is being pumped per OP/1-2/A/1104/007 (Liquid Waste Disposal System)

1SA-09/A-6 (REACTOR BUILDING NORMAL SUMP LEVEL HIGH/LOW) is in alarm

RBNS level = 1.0 inch 1) At time = 0800, 1SA-9/A-6 __(1)__ be in alarm.

2) At time = 0820, the RBNS pump __(2)__ shut down automatically due to RBNS

level. Which ONE of the following completes the statements above? A. 1. will

2. will

B. 1. will

2. will NOT

C. 1. will NOT

2. will

D. 1. will NOT

2. will NOT

7.8

RBNSLEVELIN H2O

10

5

0

15

20

25

30

15GAL/IN

SYS002 A3.01 - Reactor Coolant System (RCS)Ability to monitor automatic operation of the RCS, including: (CFR: 41.7 / 45.5)

Reactor coolant leak detection system ..............................



General Discussion

Answer A Discussion

1st part is incorrect because the alarm does not actuate until RBNS raises to 8". It is plausible because if level were .2" higher, it would be correct.

2nd part is correct. When pumping down the RBNS, as sump level lowers to 1", the pump will automatically shut down.

Answer B Discussion

1st part is incorrect because the alarm does not actuate until RBNS raises to 8". It is plausible because if level were .2" higher, it would be correct.

2nd part is incorrect because when pumping down the RBNS, as sump level lowers to 1", the pump will automatically shut down. It is plausible because in some applications (LDST) have an interlock function that occurs below the low level alarm (Lo-Lo alarm = 55" and interlock is 40").

Answer C Discussion

1st part is correct. The alarm will actuate when RBNS raises to 8".

2nd part is correct. When pumping down the RBNS, as sump level lowers to 1", the pump will automatically shut down.

Answer D Discussion

1st part is correct. The alarm will actuate when RBNS raises to 8".

2nd part is incorrect because when pumping down the RBNS, as sump level lowers to 1", the pump will automatically shut down. It is plausible because in some applications (LDST) have an interlock function that occurs below the low level alarm (Lo-Lo alarm = 55" and interlock is 40").

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


SA9 A6

Remarks/Status


This question matches the KA by requiring knowledge of the automatic operation (alarms) associated with RCS leakage into the RBNS.

Basis for Hi Cog

Basis for SRO only

SYS002 A3.01 - Reactor Coolant System (RCS)Ability to monitor automatic operation of the RCS, including: (CFR: 41.7 / 45.5)

Reactor coolant leak detection system ..............................




Reactor power = 60%

0% light for Group 4 Rod 3 is lit

AP/2/A/1700/001 (Unit Runback) is entered Time = 0802:

Reactor power = 60%

1) The “GROUP IN LIMIT” light for Control Rod Group 4 will be __(1)__.

2) At Time = 0802, the operator will__(2)__.

Which ONE of the following completes the statements above?

A. 1. OFF

2. place the Diamond AND FDW Loop Masters in MANUAL, then reduce power because the automatic runback has failed to occur

B. 1. OFF

2. leave ICS in AUTO because reactor power is below the runback setpoint for a dropped control rod

C. 1. ON

2. place the Diamond AND FDW Loop Masters in MANUAL, then reduce power because the automatic runback has failed to occur

D. 1. ON

2. leave ICS in AUTO because reactor power is below the runback setpoint for a dropped control rod

SYS014 A4.02 - Rod Position Indication System (RPIS)Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Control rod mode-select switch ....................................



General Discussion

Answer A Discussion

1st part is incorrect because if any control rod in the group is on the bottom, the Group In Limit light will be on. It is plausible because it would be logical to think that ALL control rods in the group would have to be at the in limit for the group in limit light to be lit.

2nd part is correct. The dropped rod runback setpoint is 55% power. Since ICS is not running back, AP/1 directs reducing power to < 55%. This will be done by taking the Diamond and FDW Loop Masters to HAND and reducing power.

Answer B Discussion

1st part is incorrect because if any control rod in the group is on the bottom, the Group In Limit light will be on. It is plausible because it would be logical to think that ALL control rods in the group would have to be at the in limit for the group in limit light to be lit.

2nd part is incorrect because the dropped rod runback setpoint is 55% power. Since ICS is not running back, AP/1 directs reducing power to < 55%. This will be done by taking the Diamond and FDW Loop Masters to HAND and reducing power. It is plausible because if it were an ICS runback due to a loss of RCP or loss of MFW Pump, it would be correct.

Answer C Discussion

1st part is correct. If any control rod in the group is on the bottom, the Group In Limit light will be on.

2nd part is correct. The dropped rod runback setpoint is 55% power. Since ICS is not running back, AP/1 directs reducing power to < 55%. This will be done by taking the Diamond and FDW Loop Masters to HAND and reducing power.

Answer D Discussion

1st part is correct. If any control rod in the group is on the bottom, the Group In Limit light will be on.

2nd part is incorrect because the dropped rod runback setpoint is 55% power. Since ICS is not running back, AP/1 directs reducing power to < 55%. This will be done by taking the Diamond and FDW Loop Masters to HAND and reducing power. It is plausible because if it were an ICS runback due to a loss of RCP or loss of MFW Pump, it would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


AP/2/A/1700/001IC-CRI Obj: 7Control Board UB1

Remarks/Status

DISCUSSED WITH CHIEF 7/27: LOOK AT TAKING DIAMOND TO MANUAL DUE TO INDICATION MALFUNCTION.


This question matches the KA by requiring the ability to determine when to place Diamond Panel to MANUAL based on given plant parameters.

Basis for Hi Cog

Basis for SRO only

SYS014 A4.02 - Rod Position Indication System (RPIS)Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

Control rod mode-select switch ....................................




Reactor startup in progress after refueling

Core Thermal Power Best (TPB) = 65%

NI Power:

o NI-5 = 66%

o NI-6 = 65%

o NI-7 = 68%

o NI-8 = 66% 1) Input to OAC CTPB is provided by __(1)__ ONLY.

2) The Reactor power increase __(2)__ have to be stopped to allow NI adjustment.

Which ONE of the following completes the statements above? A. 1. Primary Power

2. does

B. 1. Primary Power

2. does NOT

C. 1. Secondary Power

2. does

D. 1. Secondary Power

2. does NOT

SYS015 K5.04 - Nuclear Instrumentation System (NIS)Knowledge of the operational implications of the following concepts as they apply to the NIS: (CFR: 41.5 / 45.7)

Factors affecting accuracy and reliability of calorimetric calibrations ....



General Discussion

Answer A Discussion

1st part is incorrect because when power is > 50% power, input to the TPB OAC display is from Secondary Power ONLY. It is plausible because if power were < 15%, it would be correct.

2nd part is incorrect because per OP 1 A 1102 004, an NI would have to be > 9% conservative before the power increase would have to be stopped to re calibrate the NI. It is plausible because if it were 3% conservative when > 90%, it would be correct.

Answer B Discussion

1st part is incorrect because when power is > 50% power, input to the TPB OAC display is from Secondary Power ONLY. It is plausible because if power were < 15%, it would be correct.

2nd part is correct. Per OP 1 A 1102 004, an NI would have to be > 9% conservative before the power increase would have to be stopped to re calibrate the NI.

Answer C Discussion

1st part is correct. When power is > 50% power, input to the TPB OAC display is from Secondary Power ONLY.

2nd part is incorrect because per OP 1 A 1102 004, an NI would have to be > 9% conservative before the power increase would have to be stopped to re calibrate the NI. It is plausible because if it were 3% conservative when > 90%, it would be correct.

Answer D Discussion

1st part is correct. When power is > 50% power, input to the TPB OAC display is from Secondary Power ONLY.

2nd part is correct. Per OP 1 A 1102 004, an NI would have to be > 9% conservative before the power increase would have to be stopped to re calibrate the NI.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


OP 0 A 1102 004IC-NI Obj: 17THF-CTP

Remarks/Status

9/13/16 - Discussed with Dan Bacon following his review of the 10 preview questions. He agreed that giving a power level and asking whether Primary or Secondary power is providing input to CTPB would be acceptable and match the K/A. Power level affects the accuracy of the primary and secondary calculations.


This question matches the KA because it requires knowledge of how power affects the accuracy of the CTP best calculation/display.

Basis for Hi Cog

Basis for SRO only

SYS015 K5.04 - Nuclear Instrumentation System (NIS)Knowledge of the operational implications of the following concepts as they apply to the NIS: (CFR: 41.5 / 45.7)

Factors affecting accuracy and reliability of calorimetric calibrations ....




Reactor power = 9%

ALL air is lost to 3MS-31 (Turbine Bypass Valve) When the air is lost to 3MS-31, RCS temperature will initially __(1)__ before eventually stabilizing at the temperature equivalent to the TBV __(2)__. Which ONE of the following completes the statement above? A. 1. lower

2. setpoint

B. 1. lower

2. setpoint plus 50 psig

C. 1. rise

2. setpoint

D. 1. rise

2. setpoint plus 50 psig

SYS041 K3.02 - Steam Dump System (SDS)/Turbine Bypass ControlKnowledge of the effect that a loss or malfunction of the SDS will have on the following: (CFR: 41.7 / 45.6)

RCS ...........................................................



General Discussion

Answer A Discussion

1st part is incorrect because the TBV will fail close upon a loss of air. When that happens, RCS temperature will initially increase due to removing less heat from the primary. When SG pressure increases, the other TBV will open to maintain SG pressure. It is plausible because if the TBV failed open, it could be correct.

2nd part is correct. At 9% power, the turbine is not loaded yet (turbine LOAD status flag is false) so the 50 psig bias is not applied to the TBV setpoint.

Answer B Discussion

1st part is incorrect because the TBV will fail close upon a loss of air. When that happens, RCS temperature will initially increase due to removing less heat from the primary. When SG pressure increases, the other TBV will open to maintain SG pressure. It is plausible because if the TBV failed open, it could be correct.

2nd part is incorrect because at 9% power, the turbine is not loaded yet (turbine LOAD status flag is false) so the 50 psig bias is not applied to the TBV setpoint. It is plausible because if the turbine were loaded (LOAD status flag true), it would be correct.

Answer C Discussion

1st part is correct. The TBV will fail close upon a loss of air. When that happens, RCS temperature will initially increase due to removing less heat from the primary. When SG pressure increases, the other TBV will open to maintain SG pressure.

2nd part is correct. At 9% power, the turbine is not loaded yet (turbine LOAD status flag is false) so the 50 psig bias is not applied to the TBV setpoint.

Answer D Discussion

1st part is correct. The TBV will fail close upon a loss of air. When that happens, RCS temperature will initially increase due to removing less heat from the primary. When SG pressure increases, the other TBV will open to maintain SG pressure.

2nd part is incorrect because at 9% power, the turbine is not loaded yet (turbine LOAD status flag is false) so the 50 psig bias is not applied to the TBV setpoint. It is plausible because if the turbine were loaded (LOAD status flag true), it would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


ICS-02 Obj: 4STG-MS

Remarks/Status


This question matches the KA by requiring knowledge of how a malfunction of the TBVs affect RCS temperature.

Basis for Hi Cog

Basis for SRO only

SYS041 K3.02 - Steam Dump System (SDS)/Turbine Bypass ControlKnowledge of the effect that a loss or malfunction of the SDS will have on the following: (CFR: 41.7 / 45.6)

RCS ...........................................................




Reactor in MODE 5

RCS Loops dropped

RB Purge in progress Time = 1205:

1RIA-49 (Reactor Building High Gas) in HIGH alarm

RB Purge is still in operation 1) The Containment Evacuation alarm __(1)__ AUTOMATICALLY actuate as a result

of the 1RIA-49 HIGH alarm. 2) RB Purge __(2)__ fail to isolate. Which ONE of the following completes the statements above? A. 1. will

2. did

B. 1. will

2. did NOT

C. 1. will NOT

2. did

D. 1. will NOT

2. did NOT

SYS029 2.4.46 - Containment Purge System (CPS)SYS029 GENERIC

Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12)



General Discussion

Answer A Discussion

1st part is correct. RIA-49 in high alarm will actuate the RB evacuation alarm.

2nd part is incorrect because the RB Purge system does not isolate from an RIA-49 alarm. It is plausible because 1) if it were RIA-45, it would be correct and 2) RIA is RB Gas alarm so it would seem logical that the RB Purge system would isolate to prevent the release of the gas to the atmosphere.

Answer B Discussion

1st part is correct. RIA-49 in high alarm will actuate the RB evacuation alarm.

2nd part is correct. The RB Purge system does not isolate from an RIA-49 alarm.

Answer C Discussion

1st part is incorrect because the containment evaluation alarm will actuate upon an RIA-49 high alarm. It is plausible since some other RIA's in the RB do NOT actuate the Containment Evacuation alarm (Ex. RIA-3, 57, 58)

2nd part is incorrect because the RB Purge system does not isolate from an RIA-49 alarm. It is plausible because 1) if it were RIA-45, it would be correct and 2) RIA is RB Gas alarm so it would seem logical that the RB Purge system would isolate to prevent the release of the gas to the atmosphere.

Answer D Discussion

1st part is incorrect because the containment evaluation alarm will actuate upon an RIA-49 high alarm. It is plausible since some other RIA's in the RB do NOT actuate the Containment Evacuation alarm (Ex. RIA-3, 57, 58)

2nd part is correct. The RB Purge system does not isolate from an RIA-49 alarm.

Cognitive Level

Memory

Job Level

RO

QuestionType

MODIFIED

Question Source

ILT48 Q55


ILT48 Q55PNS-RIA Obj: 8PNS-RBP Obj: 7

Remarks/Status


This question matches the KA by requiring knowledge of correct plant status when specific alarms are in.

Basis for Hi Cog

Basis for SRO only

SYS029 2.4.46 - Containment Purge System (CPS)SYS029 GENERIC




Given the following plant conditions:

Unit 1&2 Spent Fuel Pool level = -0.2 feet, lowering at ~ 0.1 ft per minute

AP/1-2/A/1700/035 (Loss of SFP Cooling and/or Level) initiated 1) The low SFP level interlock trip of the SF Cooling pumps is minus (-) __(1)__ feet.

2) Per AP/35, Emergency Dose limits __(2)__ in effect. Which ONE of the following completes the statements above? A. 1. 2.5

2. are

B. 1. 2.5

2. are NOT

C. 1. 3.5

2. are

D. 1. 3.5

2. are NOT

SYS033 K4.01 - Spent Fuel Pool Cooling System (SFPCS)Knowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Maintenance of spent fuel level ....................................



General Discussion

Answer A Discussion

1st part is correct. The SF Cooling pump auto trip on low SFP level setpoint is -2.5 feet.

2nd part is correct. AP directs Emergency Dose Limits are in effect if SFP leakage is excessive. A note prior to Step 4.13 defines "Excessive Leakage" as SFP leak rate > 500 gpm (~ 0.1 ft every 3 minutes).

Answer B Discussion

1st part is correct. The SF Cooling pump auto trip on low SFP level setpoint is -2.5 feet.

2nd part is incorrect because the AP directs Emergency Dose Limits are in effect if SFP leakage is excessive. A note prior to Step 4.13 defines "Excessive Leakage" as SFP leak rate > 500 gpm (~ 0.1 ft every 3 minutes). It is plausible because if leakage were smaller that 0.1 gpm every 3 minutes, it would be correct.

Answer C Discussion

1st part is incorrect because because the SF Cooling pump auto trip on low SFP level setpoint is -2.5 feet. It is plausible because this is the level of the pump suction lines below the normal pool level..

2nd part is correct. AP directs Emergency Dose Limits are in effect if SFP leakage is excessive. A note prior to Step 4.13 defines "Excessive Leakage" as SFP leak rate > 500 gpm (~ 0.1 ft every 3 minutes).

Answer D Discussion

1st part is incorrect because because the SF Cooling pump auto trip on low SFP level setpoint is -2.5 feet. It is plausible because this is the level of the pump suction lines below the normal pool level..

2nd part is incorrect because the AP directs Emergency Dose Limits are in effect if SFP leakage is excessive. A note prior to Step 4.13 defines "Excessive Leakage" as SFP leak rate > 500 gpm (~ 0.1 ft every 3 minutes). It is plausible because if leakage were smaller that 0.1 gpm every 3 minutes, it would be correct.

Cognitive Level

Memory

Job Level

RO

QuestionType

MODIFIED

Question Source

ILT40 Q61


ILT40 Q61FH-SFC Obj: 5AP/35

Remarks/Status


This question matches the KA by requiring knowledge of how the SFP is maintained including when SFP pumps trip.

Basis for Hi Cog

Basis for SRO only

SYS033 K4.01 - Spent Fuel Pool Cooling System (SFPCS)Knowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

Maintenance of spent fuel level ....................................




Reactor power = 50%

AP/2/A/1700/031 (Primary to Secondary Leakage) is in progress due to a tube leak on 2A SG

1) 2RIA- __(1)__ will convert radiation level into a leak rate in GPD or GPM.

2) If the leak rate from 2A SG is determined to be 180 gpd, entry into TS 3.4.13 (RCS

Operational Leakage) __(2)__ required. Which ONE of the following completes the statements above? A. 1. 16

2. is

B. 1. 16

2. is NOT

C. 1. 59

2. is

D. 1. 59

2. is NOT

SYS072 A1.01 - Area Radiation Monitoring (ARM) SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: (CFR: 41.5 / 45.5)

Radiation levels .................................................



General Discussion

Answer A Discussion

1st part is incorrect because RIA-16 reads out in count rate based off of radiation level. It is plausible because it does detect if a primary to secondary leak exists.

2nd part is correct. A leak rate of 180 gpd exceeds the allowable leak rate fo 150 gpd per TS 3.4.13.

Answer B Discussion

1st part is incorrect because RIA-16 reads out in count rate based off of radiation level. It is plausible because it does detect if a primary to secondary leak exists.

2nd part is incorrect because a leak rate of 180 gpd exceeds the allowable leak rate fo 150 gpd per TS 3.4.13. It is plausible because if the leak rate were < 150 gpd, it would be correct.

Answer C Discussion

1st part is correct. RIA-59 reads out in GPM or GPD.

2nd part is correct. A leak rate of 180 gpd exceeds the allowable leak rate fo 150 gpd per TS 3.4.13.

Answer D Discussion

1st part is correct. RIA-59 reads out in GPM or GPD.

2nd part is incorrect because a leak rate of 180 gpd exceeds the allowable leak rate fo 150 gpd per TS 3.4.13. It is plausible because if the leak rate were < 150 gpd, it would be correct.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


RAD-RIA Obj: 8TS 3.4.13

Remarks/Status


This question matches the KA by requiring the ability to determine if a given primary to secondary leak rate as determined by an ARM exceeds TS limits.

Basis for Hi Cog

Basis for SRO only

SYS072 A1.01 - Area Radiation Monitoring (ARM) SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including: (CFR: 41.5 / 45.5)

Radiation levels .................................................




Reactor power = 50%

CCW pumps 1A, 1B and 1C are operating

1C CCW pump trips

Condenser vacuum = 27.5 inches Hg lowering

AP/1/A/1700/027 (Loss of Condenser Vacuum) is entered Time = 0810:

Reactor tripped

Condenser vacuum is continuing to lower

Turbine header pressure (THP) = 935 psig At Time = 0810…. 1) The reactor tripped due to the __(1)__ tripping due to low condenser vacuum.

2) In accordance with the Subsequent Actions tab, TBVs will be throttled __(2)__ to

maintain THP at the desired setpoint.

Which ONE of the following completes the statements above? A. 1. Main Turbine

2. open

B. 1. Main Turbine

2. closed

C. 1. Main Feedwater Pumps

2. open

D. 1. Main Feedwater Pumps

2. closed

SYS075 A2.03 - Circulating Water SystemAbility to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Safety features and relationship between condenser vacuum, turbine trip, and steam dump ................................



General Discussion

Answer A Discussion

1st part is correct. The main turbine will automatically trip if main condenser vacuum decreases to 21.75 Hg". This will in turn, trip the reactor.

2nd part is incorrect because the SA tab directs you to maintain THP at the desired setpoint. The desired setpoint is the THP setoint (885 psig) plus 125 psig bias after the trip = 1010 psig. To get there, the TBVs will have to be throttled closed. It is plausible because to maintian THP at setpoint (885 psig), the TBVs would have to be throttled open.

Answer B Discussion

1st part is correct. The main turbine will automatically trip if main condenser vacuum decreases to 21.75 Hg". This will in turn, trip the reactor.

2nd part is correct. The SA tab directs you to maintain THP at the desired setpoint. The desired setpoint is the THP setoint (885 psig) plus 125 psig bias after the trip = 1010 psig. To get there, the TBVs will have to be throttled closed.

Answer C Discussion

1st part is incorrect because the main turbine will automatically trip if main condenser vacuum decreases to 21.75 Hg". It is plausible because if power were below 27.75% power, it would be correct because, while the turbine would still trip, it would not cause a reactor trip. When the FDW pumps tripped on low vacuum at 19 Hg", it would trip the reactor.

2nd part is incorrect because the SA tab directs you to maintain THP at the desired setpoint. The desired setpoint is the THP setoint (885 psig) plus 125 psig bias after the trip = 1010 psig. To get there, the TBVs will have to be throttled closed. It is plausible because to maintian THP at setpoint (885 psig), the TBVs would have to be throttled open.

Answer D Discussion

1st part is incorrect because the main turbine will automatically trip if main condenser vacuum decreases to 21.75 Hg". It is plausible because if power were below 27.75% power, it would be correct because, while the turbine would still trip, it would not cause a reactor trip. When the FDW pumps tripped on low vacuum at 19 Hg", it would trip the reactor.

2nd part is correct. The SA tab directs you to maintain THP at the desired setpoint. The desired setpoint is the THP setoint (885 psig) plus 125 psig bias after the trip = 1010 psig. To get there, the TBVs will have to be throttled closed.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


IC-RPS Obj: 6SA Tab


This question matches the KA by requiring knowledge of the safety features associated with the turbine trip from low condenser vacuum (resulting from a reduction of Circulating Water) and the actions required (SA tab actions).

Basis for Hi Cog

Basis for SRO only

SYS075 A2.03 - Circulating Water SystemAbility to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Safety features and relationship between condenser vacuum, turbine trip, and steam dump ................................



Remarks/Status




All units at 100% power

Instrument air pressure = 89 psig lowering

AP/2/A/1700/022 (Loss of Instrument Air) in progress

An operator has been dispatched to perform Encl. 5.4 (Emergency Start of the Diesel Air Compressor)

1) The Diesel Air Compressor __(1)__ have automatically started.

2) Encl. 5.4 will direct opening SA-143 (SA to IA Controller Bypass) __(2)__.

Which ONE of the following completes the statements above? A. 1. will

2. ONLY if SA-141 (SA to IA Controller) will not maintain IA header pressure

B. 1. will

2. regardless of IA header pressure

C. 1. will NOT

2. ONLY if SA-141 (SA to IA Controller) will not maintain IA header pressure

D. 1. will NOT

2. regardless of IA header pressure

SYS079 K1.01 - Station Air System (SAS)Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

IAS ............................................................



General Discussion

Answer A Discussion

1st part is correct. The Diesel Air Compressor will automatically start when IA pressure drops to < 90 psig.

2nd part is incorrect because Enclosure 5.4 directs opening the valve. It is plausible because it is the bypass for a valve that will open at 85 psig which hasn't been reached yet.

Answer B Discussion

1st part is correct. The Diesel Air Compressor will automatically start when IA pressure drops to < 90 psig.

2nd part is correct. Enclosure 5.4 directs opening the valve. It is not dependent on IA pressure.

Answer C Discussion

1st part is incorrect because the Diesel Air Compressor will automatically start when IA pressure drops to < 90 psig. It is plausible because 1) the AIA compressors have setpoint of 88 psig IA pressure and 2) the name of the enclosure implies that they have not started yet.

2nd part is incorrect because Enclosure 5.4 directs opening the valve. It is plausible because it is the bypass for a valve that will open at 85 psig which hasn't been reached yet.

Answer D Discussion

1st part is incorrect because the Diesel Air Compressor will automatically start when IA pressure drops to < 90 psig. It is plausible because 1) the AIA compressors have setpoint of 88 psig IA pressure and 2) the name of the enclosure implies that they have not started yet.

2nd part is correct. Enclosure 5.4 directs opening the valve. It is not dependent on IA pressure.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


SSS-IA Obj: 41AP/22

Remarks/Status


This question matches the KA by requiring knowledge of the physical connection and actions directed to connect the systems following a loss of IA pressure.

Basis for Hi Cog

Basis for SRO only

SYS079 K1.01 - Station Air System (SAS)Knowledge of the physical connections and/or cause-effect relationships between the SAS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

IAS ............................................................





See reference 1) HPI pump suction __(1)__ currently aligned to the BWST.

2) The Pzr __(2)__ saturated. Which ONE of the following completes the statements above? ASSUME NO OPERATOR ACTIONS REFERENCE PROVIDED A. 1. is

2. is

B. 1. is

2. is NOT

C. 1. is NOT

2. is

D. 1. is NOT

2. is NOT

GEN2.1 2.1.19 - GENERIC - Conduct of OperationsConduct of Operations

Ability to use plant computers to evaluate system or component status. (CFR: 41.10 / 45.12)



General Discussion

Answer A Discussion

1st part is correct. At LDST level = 40", HPIP suction will automatically line up to the BWST (HP-24 & 25 open).

2nd part is correct. This can be calculated using Pzr temperature (631 degrees) an RCS pressure (1910 psig) or you could observe the yellow "P" and see that it is on the blue "saturation" line.

Answer B Discussion

1st part is correct. At LDST level = 40", HPIP suction will automatically line up to the BWST (HP-24 & 25 open).

2nd part is incorrect because the Pzr is saturated. This can be calculated using Pzr temperature (631 degrees) an RCS pressure (1910 psig) or you could observe the yellow "P" and see that it is on the blue "saturation" line. It is plausible because if you calculate using parameters, there are numerous errors that you could make (psia vs psig, get on the incorrect line in the steam tables…).

Answer C Discussion

1st part is incorrect because at LDST level = 40", HPIP suction will automatically line up to the BWST (HP-24 & 25 open). It is plausible because if level were 2 inches higher, it would be correct.

2nd part is correct. This can be calculated using Pzr temperature (631 degrees) an RCS pressure (1910 psig) or you could observe the yellow "P" and see that it is on the blue "saturation" line.

Answer D Discussion

1st part is incorrect because at LDST level = 40", HPIP suction will automatically line up to the BWST (HP-24 & 25 open). It is plausible because if level were 2 inches higher, it would be correct.

2nd part is incorrect because the Pzr is saturated. This can be calculated using Pzr temperature (631 degrees) an RCS pressure (1910 psig) or you could observe the yellow "P" and see that it is on the blue "saturation" line. It is plausible because if you calculate using parameters, there are numerous errors that you could make (psia vs psig, get on the incorrect line in the steam tables…).

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


PT Display


PT Display for LOHTSF-10 Obj: R4PNS-HPI

Remarks/Status

9/13/16 - rewrote first part of question and discussed with Dan Bacon following his review of the 10 preview questions. He was agreeable to making LDST level < 40 inches and ask whether HPI suction is from the BWST or not.


This question matches the KA by requiring the ability to use the plant computer to determine plant status.

Basis for Hi Cog

Basis for SRO only


Ability to use plant computers to evaluate system or component status. (CFR: 41.10 / 45.12)




AP/1/A/1700/029 (Rapid Unit Shutdown) is in progress

Enclosure 5.1 (Support Actions During Rapid Unit Shutdown) is being performed

Turbine/Generator shutdown is required

Load = 350 MWe (~39% Core Thermal Power) In accordance with Encl 5.1…. 1) Main Feeder Busses 1 & 2 NORMAL FDR breakers will be __(1)__. 2) 1D1 and 1D2 Heater Drain Pumps __(2)__ allowed to be secured. Which ONE of the following completes the statement above? A. 1. closed

2. are

B. 1. closed

2. are NOT

C. 1. open

2. are

D. 1. open

2. are NOT .


Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)



General Discussion

Answer A Discussion

1st part is incorrect because when you are below 80%, you are directed to transfer power (Close E1 and verify N1 opens). There is a WHEN step at 65% and you have to be below 65% to be 350 Mwe. Therefore, to be at 350MWe, the previous steps have to be complete (power has to be transferred). It is plausible because if power were > 80%, it would be correct.

2nd part is correct. The D HDPs are not secured until load is < 400 Mwe.

Answer B Discussion

1st part is incorrect because when you are below 80%, you are directed to transfer power (Close E1 and verify N1 opens). There is a WHEN step at 65% and you have to be below 65% to be 350 Mwe. Therefore, to be at 350MWe, the previous steps have to be complete (power has to be transferred). It is plausible because if power were > 80%, it would be correct.

2nd part is incorrect because the D HDPs are not secured until load is < 400 Mwe. It is plausible because 1) if it were the E HDPs, it would be correct or 2) if you were > 400 Mwe, you would be correct.

Answer C Discussion

1st part is correct. When you are below 80%, you are directed to transfer power (Close E1 and verify N1 opens). There is a WHEN step at 65% and you have to be below 65% to be 350 Mwe. Therefore, to be at 350MWe, the previous steps have to be complete (power has to be transferred).

2nd part is correct. The D HDPs are not allowed to be secured until load is < 400 Mwe.

Answer D Discussion

1st part is correct. When you are below 80%, you are directed to transfer power (Close E1 and verify N1 opens). There is a WHEN step at 65% and you have to be below 65% to be 350 Mwe. Therefore, to be at 350MWe, the previous steps have to be complete (power has to be transferred).

2nd part is incorrect because the D HDPs are not secured until load is < 400 Mwe. It is plausible because 1) if it were the E HDPs, it would be correct or 2) if you were > 400 Mwe, you would be correct.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

NEW

Question Source


AP/29EAP-AP29 Obj: 4

Remarks/Status


This question matches the KA by requiring knowledge of how to perform integrated plant procedures.

Basis for Hi Cog

Basis for SRO only


Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)





Reactor trips due to a Large Break LOCA Current conditions:

The OATC performs Immediate Manual Actions (IMAs)

The BOP performs a symptom check

Rule 2 (Loss of SCM) is performed due to Core SCM = 0°F

When RCP switches are taken to TRIP, 1A1 RCP remains running 1) Assuming that the OATC and BOP start their actions at the same time, the __(1)__

will be the operator performing Rule 2.

2) When RCP 1A1 does NOT trip, the RNO column in Rule 2 directs the operator to de-energize __(2)__.

Which ONE of the following completes the statements above? A. 1. BOP

2. 1TA only

B. 1. BOP

2. 1TA and 1TB

C. 1. OATC

2. 1TA only

D. 1. OATC

2. 1TA and 1TB


Ability to manage the control room crew during plant transients. (CFR: 41.10 / 43.5 / 45.12 / 45.13)



General Discussion

Answer A Discussion

1st part is correct. SCM is checked during the performance of a symptom check which will be performed first by the BOP while the OATC is performing IMAs.

2nd part is incorrect beause per Rule 2, if ALL RCPs are not stopped, the RNO directs you to place 1TA and 1TB AUTO/MAN switches in MAN and open 1TA and 1TB 6.9 KV FDRs. It is plausible because RCP 1A1 is powered from 1TA.

Answer B Discussion

1st part is correct. SCM is checked during the performance of a symptom check which will be performed first by the BOP while the OATC is performing IMAs.

2nd part is correct beause per Rule 2, if ALL RCPs are not stopped, the RNO directs you to place 1TA and 1TB AUTO/MAN switches in MAN and open 1TA and 1TB 6.9 KV FDRs.

Answer C Discussion

1st part is incorrect becaue SCM is checked during the performance of a symptom check which will be performed first by the BOP while the OATC is performing IMAs. It is plausible because if conditions were satisfied for the performance of Rule 1 (ATWS), the OATC would be performing it.

2nd part is incorrect beause per Rule 2, if ALL RCPs are not stopped, the RNO directs you to place 1TA and 1TB AUTO/MAN switches in MAN and open 1TA and 1TB 6.9 KV FDRs. It is plausible because RCP 1A1 is powered from 1TA.

Answer D Discussion

1st part is incorrect becaue SCM is checked during the performance of a symptom check which will be performed first by the BOP while the OATC is performing IMAs. It is plausible because if conditions were satisfied for the performance of Rule 1 (ATWS), the OATC would be performing it.

2nd part is correct beause per Rule 2, if ALL RCPs are not stopped, the RNO directs you to place 1TA and 1TB AUTO/MAN switches in MAN and open 1TA and 1TB 6.9 KV FDRs.

Cognitive Level

Memory

Job Level

RO

QuestionType

NEW

Question Source


OMP1-18Rule 2IMAsEAP-IMA Obj: R3


This question matches the KA by requiring the ability to determine which operator is performing which actions upon a LB LOCA with a LOSCM.

Basis for Hi Cog

Basis for SRO only


Ability to manage the control room crew during plant transients. (CFR: 41.10 / 43.5 / 45.12 / 45.13)



Remarks/Status

This is an SRO KA. Request New KA.

DISCUSSED WITH CHIEF 7/27: RECOMMEND ASKING ABOUT WHO DOES WHAT ON IMAs VS SYMPTOM CHECK.




Reactor power is being reduced from 100% to 88% to perform surveillance testing

OP/2/A/1102/004 (Operation at Power) Enclosure 4.2 (Power Reduction) is in progress

1) The SOC __(1)__ required to be notified. 2) The “E” Heater Drain Pumps __(2)__ required to be secured. Which ONE of the following completes the statements above? A. 1. is

2. are

B. 1. is

2. are NOT

C. 1. is NOT

2. are

D. 1. is NOT

2. are NOT

GEN2.2 2.2.17 - GENERIC - Equipment ControlEquipment Control

Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. (CFR: 41.10 / 43.5 / 45.13)



General Discussion

Answer A Discussion

1st part is correct. When using OP/2/A/1102/004, you are required to notify the SOC anytime that power is reduced.

2nd part is incorrect because the E HDP are not secured until going below 80% power however it is plausible since the E HDP's are not started until power is > 90%. That means without specific knowledge of the procedure requirements it would natural to assume they are secured when going below 90%..

Answer B Discussion

1st part is correct. When using OP/2/A/1102/004, you are required to notify the SOC anytime that power is reduced.

2nd part is correct because the E HDP are not secured until going below 80% power

Answer C Discussion

1st part is incorrect because the SOC is required to be notified if a power reduction is to occur. It is plausible because there are several thresholds for notifications during a power reduction (2nd chem @ 6%, primary chemistry @ 15%, maneuvering plan @ 10%) . There is not a threshold for notifying the SOC however.

2nd part is incorrect because the E HDP are not secured until going below 80% power however it is plausible since the E HDP's are not started until power is > 90%. That means without specific knowledge of the procedure requirements it would natural to assume they are secured when going below 90%.

Answer D Discussion

1st part is incorrect because the SOC is required to be notified if a power reduction is to occur. It is plausible because there are several thresholds for notifications during a power reduction (2nd chem @ 6%, primary chemistry @ 15%, maneuvering plan @ 10%) . There is not a threshold for notifying the SOC however.

2nd part is correct because the E HDP are not secured until going below 80% power

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT47 Q70


ILT47 Q70OP 1 A 1102 004 Encl 4.2CP 12 Obj : 5


This question matches the KA by requiring knowledge of process of coordinating with the transmission system operator (SOC) during maintenance activities.

Basis for Hi Cog

Basis for SRO only


Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. (CFR: 41.10 / 43.5 / 45.13)



Remarks/Status





Group 8 movement is required for imbalance control 1) The MINIMUM actions required prior to moving Group 8 rods is to __(1)__. 2) Group 7 rods __(2)__ respond to neutron error while group 8 is selected. Which ONE of the following completes the statements above? A. 1. place Diamond panel Group Select switch to Group 8 ONLY

2. will

B. 1. place Diamond panel Group Select switch to Group 8 ONLY

2. will NOT

C. 1. select SEQUENCE OVERRIDE and place Diamond panel Group Select switch to Group 8

2. will

D. 1. select SEQUENCE OVERRIDE and place Diamond panel Group Select switch to Group 8

2. will NOT


Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 / 45.2)



General Discussion

Answer A Discussion

1st part is correct. The only action required is selecting Group 8 on the Group Select switch and then the joystick will manually control the position of Group 8.

2nd part is correct. When the Group Select is selected to Group 8, Group 7 remains able to respond to neuron error commands for movement.

Answer B Discussion

1st part is correct. The only action required is selecting Group 8 on the Group Select switch and then the joystick will manually control the position of Group 8.

2nd part is correct. When the Group Select is selected to Group 8, Group 7 remains able to respond to neuron error commands for movement. It is plausible because it is correct when the Diamond is placed in Manual and/or sequence override.

Answer C Discussion

1st part is incorrect because the only action required is selecting Group 8 on the Group Select switch and then the joystick will manually control the position of Group 8. It is plausible because it would be correct if moving any group other then Group 8 while at 100%.

2nd part is correct. When the Group Select is selected to Group 8, Group 7 remains able to respond to neuron error commands for movement.

Answer D Discussion

1st part is incorrect because the only action required is selecting Group 8 on the Group Select switch and then the joystick will manually control the position of Group 8. It is plausible because it would be correct if moving any group other then Group 8 while at 100%.

2nd part is correct. When the Group Select is selected to Group 8, Group 7 remains able to respond to neuron error commands for movement. It is plausible because it is correct when the Diamond is placed in Manual and/or sequence override.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

2009B Q69


2009B Q69IC-CRI Obj: 18

Remarks/Status


The question requires demonstrating the ability to manipulate Group 8 Control Rods and the knowledge of how the CRD system will respond to error signals while group 8 is being manipulated.

Basis for Hi Cog

Basis for SRO only


Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 / 45.2)




A loss of power has occurred Time = 0801:

Unit auxiliaries are being supplied from CT-3 via the 230 KV switchyard

Subsequent Actions tab in progress 1) Subsequent Actions directs restarting __(1)__. 2) The __(2)__ RCP will provide the best Pressurizer Spray. Which ONE of the following completes the statements above? A. 1. one RCP ONLY

2. 3A1

B. 1. one RCP ONLY

2. 3B1

C. 1. one RCP per loop

2. 3A1

D. 1. one RCP per loop

2. 3B1


(multi-unit license) Knowledge of the design, procedural, and operational differences between units. (CFR: 41.5 / 41.6 / 41.7 / 41.10 / 45.12)



General Discussion

Answer A Discussion

1st part is incorrect because the SA tab directs starting one RCP per loop. It is plausible because there are times when the EOP directs to have only 1 RCP operating.

2nd part is incorrect because RCP 3B1 will provide the best pressurizer spray. It is plausible because if it were Unit 1, RCP 1A1 would be started.

Answer B Discussion

1st part is incorrect because the SA tab directs starting one RCP per loop. It is plausible because there are times when the EOP directs to have only 1 RCP operating.

2nd part is correct. RCP 3B1 will provide the best pressurizer spray.

Answer C Discussion

1st part is correct. The SA tab directs starting one RCP per loop.

2nd part is incorrect because RCP 3B1 will provide the best pressurizer spray. It is plausible because if it were Unit 1, RCP 1A1 would be started.

Answer D Discussion

1st part is correct. The SA tab directs starting one RCP per loop.

2nd part is correct. RCP 3B1 will provide the best pressurizer spray.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT44 Q17


ILT44 Q17EAP-SA Obj R39Unit 3 SA

Remarks/Status

Ask chief about overlap with 16-1 Q36. I think were ok. 16-1 asks about taking RCPs off due to stuck spray valve on Unit 1. This Q asks about starting RCPs to initiate spray (Unit 3).

DISCUSSED WITH CHIEF 7/27: OK WITH THIS QUESTION NOT OVERLAPPING WITH 16-1 Q36.


This question matches the KA by requiring knowledge of procedural and design differences between Unit 1 and Unit 3 relative to guidance provided in the Subsequent Actions tab. The Pzr spray line on Unit 1 comes off the 1A1 loop while on Unit 3, it comes off of the 3B1 loop.

Basis for Hi Cog

Basis for SRO only






3A GWD gas tank release in progress

Release is at 2/3 Station Limit 1) 1RIA-45 High and Alert setpoints will be set at __(1)__ the normal 1/3 Station Limit

as listed in PT/0/A/230/001 (Radiation Monitor Check). 2) If 1RIA-45 High alarm setpoint is reached, the 3A GWD gas tank release __(2)__. Which ONE of the following completes the statements above? A. 1. the same as

2. will automatically terminate

B. 1. the same as

2. must be manually terminated

C. 1. half

2. will automatically terminate

D. 1. half

2. must be manually terminated

GEN2.3 2.3.11 - GENERIC - Radiation ControlRadiation Control

Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)



General Discussion

Answer A Discussion

1st part is incorrect because per PT/0/A/230/001 (Radiation Monitor Check) the setpoint on the non-releasing unit is set at half the value in the PT. It is plausible because it is on the non-releasing unit.

2nd part is incorrect because per OP/3/A/1104/018 (GWD System) if RIA-45 High alarm actuates on a non-releasing unit, the other unit must be notified to manually terminate the release. RIA-37/38 are the process monitors that are interlocked to terminate the release. It is plausible because if RIA-37/38 were in alarm, it would be correct.

Answer B Discussion

1st part is incorrect because per PT/0/A/230/001 (Radiation Monitor Check) the setpoint on the non-releasing unit is set at half the value in the PT. It is plausible because it is on the non-releasing unit.

2nd part is correct. Per OP/3/A/1104/018 (GWD System) if RIA-45 High alarm actuates on a non-releasing unit, the other unit must be notified to manually terminate the release. RIA-37/38 are the process monitors that are interlocked to terminate the release.

Answer C Discussion

1st part is correct. Per PT/0/A/230/001 (Radiation Monitor Check) the setpoint on the non-releasing unit is set at half the value in the PT.

2nd part is incorrect because per OP/3/A/1104/018 (GWD System) if RIA-45 High alarm actuates on a non-releasing unit, the other unit must be notified to manually terminate the release. RIA-37/38 are the process monitors that are interlocked to terminate the release. It is plausible because if RIA-37/38 were in alarm, it would be correct.

Answer D Discussion

1st part is correct. Per PT/0/A/230/001 (Radiation Monitor Check) the setpoint on the non-releasing unit is set at half the value in the PT.

2nd part is correct. Per OP/3/A/1104/018 (GWD System) if RIA-45 High alarm actuates on a non-releasing unit, the other unit must be notified to manually terminate the release. RIA-37/38 are the process monitors that are interlocked to terminate the release.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT39 Q70


ILT39 Q70WE-GWD Obj: 7PT/0/A/0230/001 ENCL 13.21

Remarks/Status


This question matches the KA by requiring knowledge of the process for releasing at 2/3 the station limit.

Basis for Hi Cog

Basis for SRO only


Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)



In accordance with PD-RP-ALL-0001 (Radiation Worker Responsibilities), which ONE of the following states…. 1) the MAXIMUM annual Whole Body Dose Limit (REM) allowed by the NRC? 2) the MAXIMUM lifetime Planned Special Exposure limit (REM) allowed by the NRC? A. 1. 5

2. 25

B. 1. 5

2. 50

C. 1. 2

2. 25

D. 1. 2

2. 50


Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)



General Discussion

Answer A Discussion

1st part is correct. The maximum annual dose limit allowed by the NRC is 5 REM.

2nd part is correct. The maximum lifetime planned special exposure limit allowed by theNRC is 5 times the annual limit (25 REM).

Answer B Discussion

1st part is correct. The maximum annual dose limit allowed by the NRC is 5 REM.

2nd part is incorrect because the maximum lifetime planned special exposure limit allowed by the NRC is 5 times the annual limit (25 REM). It is plausible because 50 REM is the annual limit allowed by the NRC for SDE to skin and extremities and CDE to any tissue or organ except lens of the eye.

Answer C Discussion

1st part is incorrect because the maximum annual dose limit allowed by the NRC is 5 REM. It is plausible because 2 REM is the annual dose limit allowed by Duke Energy.

2nd part is correct. The maximum lifetime planned special exposure limit allowed by theNRC is 5 times the annual limit (25 REM).

Answer D Discussion

1st part is incorrect because the maximum annual dose limit allowed by the NRC is 5 REM. It is plausible because 2 REM is the annual dose limit allowed by Duke Energy.

2nd part is incorrect because the maximum lifetime planned special exposure limit allowed by the NRC is 5 times the annual limit (25 REM). It is plausible because 50 REM is the annual limit allowed by the NRC for SDE to skin and extremities and CDE to any tissue or organ except lens of the eye.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT48 Q72


ILT48 Q72RAD-RPP Obj: 5PD-RP-ALL-0001

Remarks/Status


This question matches the KA by requiring knowledge of emergency worker radiation exposure limits and when they apply.

Basis for Hi Cog

Basis for SRO only


Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)





AP/1/A/1700/050 (Challenging Plant Fire) initiated due to a fire in the Control Room

1) Per AP/50, the Unit 1 Reactor Operators will relocate to the __(1)__. 2) A method used in RP/0/A/1000/029 (Fire Brigade Response) to dispatch the full fire

brigade is __(2)__. Which ONE of the following completes the statements above? A. 1. Auxiliary Shutdown Panel

2. using the plant paging system

B. 1. Auxiliary Shutdown Panel

2. having Security dispatch fire brigade

C. 1. Standby Shutdown Facility

2. using the plant paging system

D. 1. Standby Shutdown Facility

2. having Security dispatch fire brigade

GEN2.4 2.4.39 - GENERIC - Emergency Procedures / PlanEmergency Procedures / Plan

Knowledge of RO responsibilities in emergency plan implementation. (CFR: 41.10 / 45.11)



General Discussion

Answer A Discussion

1st part is incorrect because for a control room evacuation due to fire, AP/8 directs you to AP/50 which directs you to evacuate to the SSF. It is plausible because if the control room evacuation was not due to a fire, AP/8 would direct you to the ASP.

2nd part is correct. Per RP/0/A/1000/029, Encl 4.1, a method to dispatch the full fire brigade is to use the plant PA system. This is significant since all fire brigade members do not have radios and pagers therefore the plant page is used to ensure all members are notified.

Answer B Discussion

1st part is incorrect because for a control room evacuation due to fire, AP/8 directs you to AP/50 which directs you to evacuate to the SSF. It is plausible because if the control room evacuation was not due to a fire, AP/8 would direct you to the ASP.

2nd part is incorrect because security does not dispatch the fire brigade. It is plausible because per RP/0/A/1000/029, Encl 4.1, Security is used to dispatch the MERT to a medical emergency to respond along with the fire brigade.

Answer C Discussion

1st part is correct. For a control room evacuation due to fire, AP/8 directs you to AP/50 which directs you to evacuate to the SSF.

2nd part is correct. Per RP/0/A/1000/029, Encl 4.1, a method to dispatch the full fire brigade is to use the plant PA system. This is significant since all fire brigade members do not have radios and pagers therefore the plant page is used to ensure all members are notified.

Answer D Discussion

1st part is correct. For a control room evacuation due to fire, AP/8 directs you to AP/50 which directs you to evacuate to the SSF.

2nd part is incorrect because security does not dispatch the fire brigade. It is plausible because per RP/0/A/1000/029, Encl 4.1, Security is used to dispatch the MERT to a medical emergency to respond along with the fire brigade.

Cognitive Level

Memory

Job Level

RO

QuestionType

BANK

Question Source

ILT39 Q74


ILT39 Q74RP/1000/029IC-FDS Obj: 6AP/50

Remarks/Status


This question matches the KA by requiring knowledge of RO actions (going to the SSF) during emergency plan implementation.

Basis for Hi Cog

Basis for SRO only


Knowledge of RO responsibilities in emergency plan implementation. (CFR: 41.10 / 45.11)



Unit 1 plant conditions:


1DIB inverter input breaker trips Based on the above plant conditions, which ONE of the following represents the conditions of the statalarms on 1SA-1 for ES Actuation Logic channels one and two? A.

B.

C.

D.





General Discussion

Answer A Discussion

Correct. If the Digital ES channels lose power, they fail in the untripped state. 1DIB feeds KVIB. Since there is no auto backup to the vital panel boards and since 1KVIB feeds the EVEN digital channels, a loss of power to the 1KVIB panel board would not result in actuation of any of the digital ES channels.

Answer B Discussion

Incorrect because neither light will be lit (see A). It is plausible because KVIA panel board provides power to the odd digital ES channels and KVIB feeds power to the even ES digital channels therefore reversing which panel board feeds which channels is an easy misconception. RPS channels trip when they lose power and ES channels fail untripped when they lose power therefore reversing which ones trip and which ones do not trip is also a plausible misconception.

Answer C Discussion

Incorrect because neither light will be lit (see A). It is plausible since KVIB panel board (fed from 1DIB) feeds the even ES channels therefore this would be correct under the misconception that the digital ES channels actuate when they lose power (like the RPS channels do).

Answer D Discussion

Incorrect: That both ES 1 and 2 are affected is plausible since KVIA feeds one halfof the ES digital channels and KVIB feeds the other half therefore it is plausible tohave a misconception about which specific channels are fed from whichpanel board. Specifically, since there are 4 ES functions (Lo RCS pressure, Lo LoRCS, Hi RB Pressure and HI HI RB pressure) and there are 4 Vital panel boardsfeeding the ES channels it would be plausible to believe that a single panel boardfeeds each ES function. Since RPS channels trip when they lose power, it isplausible to believe that the ES channels affected would trip.

Cognitive Level

Comprehension

Job Level

RO

QuestionType

BANK

Question Source

ILT40 Q13


ILT40 Q13IC-ES Obj: 2, 10, 20

Remarks/Status


This question matches the KA by requiring the ability of the applicant to verify that the alarms received are consistent with the described plant conditions.

Basis for Hi Cog

Basis for SRO only







Pzr Safety Valve fails open

Reactor trips

Loss of all Feedwater

PREV CANC F1=SPDS STATU F2=PRINT PATH F3=PRINT INDE F4=VIEW PATH F5=VIEW INDET F6=LOGIC

SUBCRIT ICC HEAT SINK RCS INTEG CONTAIN RCS INVEN

GREEN YELLOW ORANGE WHITEORANGE GREEN

Time = 0615:

LOSCM tab in progress

TD EFWP supplying both SGs

RCS Pressure = 1685 psig

CETCs = 613°F lowering 1) At time = 0600, the highest priority Critical Safety Function indicated on the (SPDS)

Displayed above is __(1)__. 2) At time = 0615, in accordance with the LOSCM Tab, the __(2)__ Tab will be

initiated. Which ONE of the following completes the statements above? A. 1. Heat Sink

2. LOHT

B. 1. Heat Sink

2. LOCA CD

C. 1. RCS Integrity

2. LOHT

D. 1. RCS Integrity

2. LOCA CD

APE008 AA2.28 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: (CFR: 43.5 / 45.13)

Safety parameter display system indications ..........................



General Discussion

Answer A Discussion

1st part is correct. When two safety functions are the same color, the order of priority is left to right so the Heat Sink is the highest priority safety function.

2nd part is incorrect because with the the TD EFDW pump in operation CETCs lowering, the LOSCM tab directs you to GO TO the LOCA CD tab. It is plausible because if if CETCs are increasing, it directs you to GO TO the LOHT tab.

Answer B Discussion

1st part is correct. When two safety functions are the same color, the order of priority is left to right so the Heat Sink is the highest priority safety function.

2nd part is correct. TD EFDW pump in operation CETCs lowering, the LOSCM tab directs you to GO TO the LOCA CD tab.

Answer C Discussion

1st part is incorrect because when two safety functions are the same color, the order of priority is left to right so the Heat Sink is the highest priority safety function. It is plausible because with the RCS integrity issue (Safety failed open) leading to the LOSCM and the LOSCM being a higher priority than the LOHT, the RCS Integrity could seem correct.

2nd part is incorrect because with the the TD EFDW pump in operation CETCs lowering, the LOSCM tab directs you to GO TO the LOCA CD tab. It is plausible because if if CETCs are increasing, it directs you to GO TO the LOHT tab.

Answer D Discussion

1st part is incorrect because when two safety functions are the same color, the order of priority is left to right so the Heat Sink is the highest priority safety function. It is plausible because with the RCS integrity issue (Safety failed open) leading to the LOSCM and the LOSCM being a higher priority than the LOHT, the RCS Integrity could seem correct.

2nd part is correct. TD EFDW pump in operation CETCs lowering, the LOSCM tab directs you to GO TO the LOCA CD tab.


The question matches the KA by requiring knowledge of the SPDS indications during a Pzr Vapor Space Accident (failed open Safety Valve).

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2E. Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal,abnormal, or emergency) and then 2) selecting a procedure or section of a procedureto mitigate, recover, or with which to proceed. One area of SRO level knowledge (withrespect to selecting a procedure) is knowledge of the content of the procedure versusknowledge of the procedure’s overall mitigative strategy or purpose.The applicant’s knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) byensuring that the additional knowledge of the procedure’s content is required tocorrectly answer the written test item, for example: Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures. Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.



Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


LOSCM TabCAE SF-12 Obj: 1.5

Remarks/Status

APE008 AA2.28 - Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: (CFR: 43.5 / 45.13)

Safety parameter display system indications ..........................





1SA-16/A-6 (RC PUMP MOTOR 1A1 OIL POT LOW LEVEL) in alarm

MTR UPPER TH BRG TEMP 1 = 165°F and rising at 5°F per minute

AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation) is in progress

The Shift Manager has directed reducing power immediately to remove 1A1 RCP from service

1) The power reduction will be performed using __(1)__. 2) If 1A1 RCP Thrust Bearing temperature continues to rise at the same rate, RCP

Immediate trip criteria will first be met at __(2)__.

Which ONE of the following completes the statements above? A. 1. AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation)

2. 0805

B. 1. AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation)

2. 0812

C. 1. AP/1/A/1700/029 (Rapid Unit Shutdown)

2. 0812

D. 1. AP/1/A/1700/029 (Rapid Unit Shutdown)

2. 0805

APE015/017 AA2.02 - Reactor Coolant Pump (RCP) MalfunctionsAbility to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 43.5 / 45.13)

Abnormalities in RCP air vent flow paths and/or oil cooling system . . . . .



General Discussion

Answer A Discussion

1st part is correct. If a rapid shutdown is required, AP/16 directs you to Enclosure 5.2 of AP/16 for the power reduction.

2nd part is correct. The RCP emergency trip criteria in AP/16 Enclosure 5.1 for the RCP Thrust Bearing is 190 degrees. 165 to 190 degrees at 5 degrees per minute will take 5 minutes which will be at 0805.

Answer B Discussion

1st part is correct. If a rapid shutdown is required, AP/16 directs you to Enclosure 5.2 of AP/16 for the power reduction.

2nd part is incorrect because the RCP emergency trip criteria in AP/16 Enclosure 5.1 for the RCP Thrust Bearing is 190 degrees. 165 to 190 degrees at 5 degrees per minute will take 5 minutes which will be at 0805. It is plausible because if it were a RCP Radial Bearing temperature (225 degrees), it would be correct.

Answer C Discussion

1st part is incorrect because If a rapid shutdown is required, AP/16 directs you to Enclosure 5.2 of AP/16 for the power reduction. It is plausible because not all Aps have directions contained in their bodies for a rapid power reduction (AP/2) or any time not it an AP that a rapid power reduction is required.

2nd part is correct. The RCP emergency trip criteria in AP/16 Enclosure 5.1 for the RCP Thrust Bearing is 190 degrees. 165 to 190 degrees at 5 degrees per minute will take 5 minutes which will be at 0805.

Answer D Discussion

1st part is incorrect because If a rapid shutdown is required, AP/16 directs you to Enclosure 5.2 of AP/16 for the power reduction. It is plausible because not all APs have directions contained in their bodies for a rapid power reduction (AP/2) or any time not it an AP that a rapid power reduction is required.

2nd part is incorrect because the RCP emergency trip criteria in AP/16 Enclosure 5.1 for the RCP Thrust Bearing is 190 degrees. 165 to 190 degrees at 5 degrees per minute will take 5 minutes which will be at 0805. It is plausible because if it were a RCP Radial Bearing temperature (225 degrees), it would be correct.


This question matches the KA by requiring the ability to determine actions that apply to the plant based on abnormalities in the RCP oil cooling system (low level).

Basis for Hi Cog

Basis for SRO only




Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


AP/16EP-AP Obj: 4

Remarks/Status

Check with Chief to make sure that "oil cooling system" can refer to the oil system (as written) or does it have to be the system that cools the oil system?

DISCUSSED WITH CHIEF 7/27: CHIEF OK KA MATCH AND SRO CRITERIA.

APE015/017 AA2.02 - Reactor Coolant Pump (RCP) MalfunctionsAbility to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 43.5 / 45.13)

Abnormalities in RCP air vent flow paths and/or oil cooling system . . . . .




Reactor in MODE 6

Fuel Transfer Canal full

SF-1 and SF-2 are open Current conditions:

Fuel Transfer Canal level slowly lowering

AP/26 (Loss of Decay Heat Removal) in progress Which one of the following describes the action that would be performed FIRST, in accordance with AP/26, and why? A. Secure ALL LPI Pumps in preparation for closing 1SF-1 and 1SF-2

B. Secure ALL LPI Pumps to determine if leak is on discharge of LPI Pumps

C. Secure SF Cooling pump used for Refueling Cooling Mode to determine if leak is on discharge of SF Cooling Pump

D. Secure SF Cooling pump used for Refueling Cooling Mode in preparation for closing 1SF-1 and 1SF-2

APE025 2.1.20 - Loss of Residual Heat Removal System (RHRS)APE025 GENERIC

Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)



General Discussion

Answer A Discussion

1st part is correct. LPI pumps are secured early in the procedure

2nd part is incorrect because this is not the reason for securing LPI pumps. It is plausible because the current SFC alignment in the refueling mode provides for taking a suction off the fuel transfer canal via the decay heat drop line and discharging to the SFP. That alignment must be secured prior to closing SF-1 and 2 to prevent pumping FTC to SFP however it is the B SF pump being used in this alignment and not the LPI pumps.

Answer B Discussion

Correct. Once you make your way to the correct section of AP/26 , it will direct securing all LPI pumps in an effort to determine the location and isolability of the leak. If securing the pumps do not change the leak rate then they will be restarted.

Answer C Discussion

Incorrect because the LPI pumps will be secured first in the procedure to see if that is the source of the leak. It is plausible since the B SFC Pump is being used in the Refueling Mode alignment and securing the pump and monitoring leak rate could help determine if the source of the leak is on the discharge of SFC pump. Since the Fuel Transfer Canal is full, securing the pump is plausible.Additionally, this action is actually directed by AP/26 although it is a later action after transferring to the condition specific section of the AP. It is the LPI pumps that are initially secured.

Answer D Discussion

Incorrect because the LPI pumps will be secured first in the procedure to see if that is the source of the leak. It is plausible since the SFC pumps are secured later in AP/26 prior to closing SF-1 and 2 to prevent pumping FTC to SFP.


This question matches the KA by requiring knowledge of actions taken in AP/26 based on a decreasing fuel transfer canal and the ability to interpret the results of those actions todetermine location and isolability of Fuel Transfer Canal leak.

Basis for Hi Cog

Basis for SRO only

Securing the LPI pumps is in the body of the procedure and securing SF Cooling pumps is in section 4D. The actions are in different sections of the procedure although the question itself is not asking which sections are applicable.




Cognitive Level

Comprehension

Job Level

SRO

QuestionType

BANK

Question Source

ILT41 Q78


ILT41 Q78EAP-AP26 Obj: 4AP/26

Remarks/Status

APE025 2.1.20 - Loss of Residual Heat Removal System (RHRS)APE025 GENERIC





Loss of ALL Main and Emergency Feedwater

OATC is performing Rule 3

LOHT tab is in progress Time = 0804:

SG pressure is being reduced to establish Condensate Booster Pump (CBP) feed

1A SG level = 12 inches XSUR

1B SG level = 11 inches XSUR Time = 0810:

CBP feed is lost

Rule 4 is complete

1A HPIP is the only HPI pump available 1) At Time = 0804, when establishing CBP feed, SG pressure should be lowered to

500 psig __(1)__.

2) At Time = 0810, the CRS will __(2)__. Which ONE of the following completes the statements above? A. 1. at the maximum rate to establish feed to the SGs

2. GO TO the HPI CD tab

B. 1. at the maximum rate to establish feed to the SGs

2. remain in the LOHT tab

C. 1. slowly to prevent overcooling

2. GO TO the HPI CD tab

D. 1. slowly to prevent overcooling

2. remain in the LOHT tab

APE054 2.4.20 - Loss of Main Feedwater (MFW)APE054 GENERIC

Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13)



General Discussion

Answer A Discussion

1st part is incorrect because as per a caution statement prior to depressurizing SGs: Until SGs are dry, lower SG pressure slowly to prevent overcooling (12" XSUR is considered dry). It is plausible because the RCS is losing its heat sink so it would seem logical to establish feed as soon as possible. Also, if XSUR were < 12", it would be correct.

2nd part is correct. With degraded HPI, the transition to HPI CD will still occur after Reactor Vessel Head and RCS Loop Vents are opened.

Answer B Discussion

1st part is incorrect because as per a caution statement prior to depressurizing SGs: Until SGs are dry, lower SG pressure slowly to prevent overcooling (12" XSUR is considered dry). It is plausible because the RCS is losing its heat sink so it would seem logical to establish feed as soon as possible. Also, if XSUR were < 12", it would be correct.

2nd part is incorrect because the transition to the HPI CD tab will occur after Reactor Vessel Head and RCS Loop Vents are opened. It is plausible because if NO HPIPs were operating (also degraded HPI), it would be correct. Rule 4 (HPI CD) is performed with degraded HPI even if no HPIPs are operating.

Answer C Discussion

1st part is correct. As per a caution statement prior to depressurizing SGs: Until SGs are dry, lower SG pressure slowly to prevent overcooling (12" XSUR is considered dry).

2nd part is correct. With degraded HPI, the transition to HPI CD will still occur after Reactor Vessel Head and RCS Loop Vents are opened.

Answer D Discussion

1st part is correct. As per a caution statement prior to depressurizing SGs: Until SGs are dry, lower SG pressure slowly to prevent overcooling (12" XSUR is considered dry).

2nd part is incorrect because the transition to the HPI CD tab will occur after Reactor Vessel Head and RCS Loop Vents are opened. It is plausible because if NO HPIPs were operating (also degraded HPI), it would be correct. Rule 4 (HPI CD) is performed with degraded HPI even if no HPIPs are operating.


This question matches the KA by requiring knowledge of the operational implications associated with depressurizing the SGs with level in them (over cooling) due to a loss of MFW (and EFW).

Basis for Hi Cog

Basis for SRO only




Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


Rule 3LOHT tabEAP-LOHT Obj: R1

Remarks/Status

9/13/16 - Changed 2nd question back to LOHT tab or HPI CD tab per discussion with Dan Bacon following his review of the 10 preview questions. KDS - 9/13

APE054 2.4.20 - Loss of Main Feedwater (MFW)APE054 GENERIC

Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13)



Given the following plant conditions: Date = 12/02

Unit 1 = Mode 5

Unit 2 = Mode 1

Unit 3 = Mode 1

Battery Charger 2CA is out of service Date = 12/03

Time = 0800 o Battery Charger 1CA breakers trip open

Date = 12/04

Time = 1200

o Battery Chargers 1CA and 2CA are out of service

o Commenced Unit 2 and 3 shutdown

Time = 1500

o Units 2 and 3 in Mode 3 1) Battery Charger 1CA (when available) is a power source for Unit __(1)__. 2) In accordance with AD-LS-ALL-0006 (Notification / Reportability Evaluation),

the latest time on 12/04 that an ENS notification is required is __(2)__ Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. 2

2. 1600

B. 1. 2

2. 1900

C. 1. 3

2. 1600

D. 1. 3

2. 1900

APE058 2.4.30 - Loss of DC PowerAPE058 GENERIC

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)



General Discussion

Answer A Discussion

1st part is incorrect because Battery Charger 1CA is counted as a power source for Units 1 and 3. It is plausible because 2CA is counted as a power supply for Unit 1.

2nd part is correct. Per AD-LS-ALL-0006, Table 1, for a plant shutdown required by Tech Specs, ENS notification is required to be within 4 hours of "initiation" of S/D required by TS.

Answer B Discussion

1st part is incorrect because Battery Charger 1CA is counted as a power source for Units 1 and 3. It is plausible because 2CA is counted as a power supply for Unit 1.

2nd part is incorrect because per AD-LS-ALL-0006, Table 1, for a plant shutdown required by Tech Specs, ENS notification is required to be within 4 hours of "initiation" of S/D required by TS. It is plausible because if the requirement was within 4 hours of the time that the unit is shutdown, it would be correct.

Answer C Discussion

1st part is correct. Battery Charger 1CA is counted as a power source for Units 1 and 3.

2nd part is correct. Per AD-LS-ALL-0006, Table 1, for a plant shutdown required by Tech Specs, ENS notification is required to be within 4 hours of "initiation" of S/D required by TS.

Answer D Discussion

1st part is correct. Battery Charger 1CA is counted as a power source for Units 1 and 3.

2nd part is incorrect because per AD-LS-ALL-0006, Table 1, for a plant shutdown required by Tech Specs, ENS notification is required to be within 4 hours of "initiation" of S/D required by TS. It is plausible because if the requirement was within 4 hours of the time that the unit is shutdown, it would be correct.

Cognitive Level

Memory

Job Level

SRO

QuestionType

NEW

Question Source


AD-LS-ALL-0006


AD-LS-ALL-0006TS 3.8.3EL-DCD Obj: 7


This question matches the KA by requiring the ability to determine the reportability requirements due to consequences from a loss of DC power.

Basis for Hi Cog

Basis for SRO only

Per ES-401 Attachment 2 (Clarification Guidance for SRO-only Questions), requires knowledge of the NRC reporting requirements under Conditions and limitations in the facility license (10CFR55.43(b)(1).

APE058 2.4.30 - Loss of DC PowerAPE058 GENERIC

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11)



Remarks/Status



BWE05 EA2.2 - Excessive Heat TransferAbility to determine and interpret the following as they apply tothe (Excessive Heat Transfer)(CFR: 43.5 / 45.13)Ability to determine and interpret the following as they apply tothe (Excessive Heat Transfer)(CFR: 43.5 / 45.13)

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.




Steam line break in 1A Steam Generator has occurred from 100%

Rule 5 (Main Steam Line Break) is complete

EHT tab in progress Time = 0810:

Steam line break occurs in 1B Steam Generator

Rule 5 re-performed 1) At Time = 0810, the __(1)__ tab will be used to mitigate the event. 2) If neither SG can be “trickle fed”, the __(2)__ tab will be directed by the CRS. Which ONE of the following completes the statements above? A. 1. EHT

2. FCD

B. 1. EHT

2. HPI CD

C. 1. LOHT

2. FCD

D. 1. LOHT

2. HPI CD



General Discussion

Answer A Discussion

1st part is correct. Per the EHT tab, Parallel Action (Yellow) page: If indications of excessive heat transfer in another SG after EHT tab initiated, RETURN TO beginning of EHT tab.

2nd part is incorrect because with no SGs available for heat transfer, Rule 4 will be performed and transfer to HPI CD is directed in the EHT tab. It is plausible because if trickle feed were available to one SG, it would be correct.

Answer B Discussion

1st part is correct. Per the EHT tab, Parallel Action (Yellow) page: If indications of excessive heat transfer in another SG after EHT tab initiated, RETURN TO beginning of EHT tab.

2nd part is correct. With no SGs available for heat transfer, Rule 4 will be performed and transfer to HPI CD is directed in the EHT tab.

Answer C Discussion

1st part is incorrect because per the EHT tab, Parallel Action (Yellow) page: If indications of excessive heat transfer in another SG after EHT tab initiated, RETURN TO beginning of EHT tab. It is plausible because with NO SGs available for heat transfer, the LOHT tab would seem applicable because it is a higher priority tab.

2nd part is incorrect because with no SGs available for heat transfer, Rule 4 will be performed and transfer to HPI CD is directed in the EHT tab. It is plausible because if trickle feed were available to one SG, it would be correct.

Answer D Discussion

1st part is incorrect because per the EHT tab, Parallel Action (Yellow) page: If indications of excessive heat transfer in another SG after EHT tab initiated, RETURN TO beginning of EHT tab. It is plausible because with NO SGs available for heat transfer, the LOHT tab would seem applicable because it is a higher priority tab.

2nd part is correct. With no SGs available for heat transfer, Rule 4 will be performed and transfer to HPI CD is directed in the EHT tab.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

BANK

Question Source

2009B Q93


The question matches the KA by requiring knowledge of appropriate procedures used for an EHT event.

Basis for Hi Cog

Basis for SRO only





2009B Q93EAP-EHT Obj: 5EOP-EHT

Remarks/Status

BWE05 EA2.2 - Excessive Heat TransferAbility to determine and interpret the following as they apply tothe (Excessive Heat Transfer)(CFR: 43.5 / 45.13)Ability to determine and interpret the following as they apply tothe (Excessive Heat Transfer)(CFR: 43.5 / 45.13)

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.




Reactor power = 6%

Startup in progress Time = 0801:

Control rods begin moving in the OUT direction

Reactor power = 13% rising

OATC depresses the Reactor Trip pushbutton

1RC-66 (PORV) opens at setpoint Time = 0802

Rule 1 (ATWS) in progress 1) In accordance with Rule 1, an operator is dispatched to open the CRD breakers

__(1)__ aligning HPI injection from the BWST. 2) The Emergency Plan classification (EAL) for this event is __(2)__.

Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. prior to

2. Alert

B. 1. prior to

2. Site Area Emergency

C. 1. after

2. Alert

D. 1. after


APE001 2.4.41 - Continuous Rod WithdrawalAPE001 GENERIC

Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11)



General Discussion

Answer A Discussion

1st part is incorrect because Rule 1 directs aligning the BWST (opening 1HP-24 & 25) prior to dispatching an operator to open CRD breakers. It is plausible because if 1HP-24 & 25 fail to open, an operator is dispatched to open CRD breakers prior to lining up the BWST (via piggyback).

1st part is incorrect because a SAE is the proper classification. Failure of an automatic reactor trip meets the criteria for an ALERT. If DSS AND the manual trip from the control room also fail to trip the reactor, it becomes a SAE. It is plausible because if the manual trip or DSS trip was successful, it would be correct.

Answer B Discussion

1st part is incorrect because Rule 1 directs aligning the BWST (opening 1HP-24 & 25) prior to dispatching an operator to open CRD breakers. It is plausible because if 1HP-24 & 25 fail to open, an operator is dispatched to open CRD breakers prior to lining up the BWST (via piggyback).

1st part is correct. Failure of an automatic reactor trip meets the criteria for an ALERT. Because DSS AND the manual trip from the control room also fail to trip the reactor, it becomes a SAE.

Answer C Discussion

1st part is correct. Rule 1 directs opening 1HP-24 & 25 prior to dispatching an operator to open CRD breakers.

1st part is incorrect because a SAE is the proper classification. Failure of an automatic reactor trip meets the criteria for an ALERT. If DSS AND the manual trip from the control room also fail to trip the reactor, it becomes a SAE. It is plausible because if the manual trip or DSS trip was successful, it would be correct.

Answer D Discussion

1st part is correct. Rule 1 directs opening 1HP-24 & 25 prior to dispatching an operator to open CRD breakers.

1st part is correct. Failure of an automatic reactor trip meets the criteria for an ALERT. Because DSS AND the manual trip from the control room also fail to trip the reactor, it becomes a SAE.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT46 Q.9


This question matches the KA by requiring the ability to analyze plant conditions to make an accurate emergency classification.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2F. Procedures and limitations involved in initial core loading, alterations in coreconfiguration, control rod programming, and determination of various internal andexternal effects on core reactivity. [10 CFR 55.43(b)(6)]Some examples of SRO exam items for this topic include:• Evaluating core conditions and emergency classifications based on core conditions.• Administrative requirements associated with low power physics testing processes.• Administrative requirements associated with refueling activities, such as approvalsrequired to amend core loading sheets or administrative controls of potential dilutionpaths and/or activities.• Administrative controls associated with the installation of neutron sources.• Knowledge of TS bases for reactivity controls.




RP 0 A 1000 001


RP/0/A/1000/001Rule 1EAP SEP Obj: R12

Remarks/Status

9/13/16 - New K/A (APE001 2.4.41) received from Dan Bacon following 10 preview question review. Could not write a discriminating SRO ONLY question to the original K/A.

APE001 2.4.41 - Continuous Rod WithdrawalAPE001 GENERIC

Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11)





Pressurizer (Pzr) Level 2 selected

SASS is in AUTO Current conditions:

The diaphragm in the DP cell for Pzr level 2 detector ruptures 1) As Pzr level deviates from the other Pzr level detectors, SASS will select Pzr Level

Detector __(1)__. 2) The bases for the limit on Pzr level in TS 3.4.9 (Pressurizer) is to __(2)__. Which ONE of the following complete the statements above? A. 1. 1

2. prevent exceeding the RCS pressure Safety Limit due to a subsequent Pzr insurge

B. 1. 1

2. ensure adequate Pzr heater capacity to return the Pzr to saturated conditions following all anticipated transients

C. 1. 3

2. prevent exceeding the RCS pressure Safety Limit due to a subsequent Pzr insurge

D. 1. 3

2. ensure adequate Pzr heater capacity to return the Pzr to saturated conditions following all anticipated transients

APE028 AA2.01 - Pressurizer (PZR) Level Control MalfunctionAbility to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: (CFR: 43.5 / 45.13)

PZR level indicators and alarms ....................................



General Discussion

Answer A Discussion

1st part is incorrect because SASS will select the detector in the other TRAIN (#3 in this case). It is plausible because level indication 1 & 2 are on the same train.

2nd part is correct. TSB 3.4.9 states: "If the level limits were exceeded prior to a transient that creates a large pressurizer insurge volume, the maximum RCS pressure could exceed the design Safety Limit (SL) of 2750 psig."

Answer B Discussion

1st part is incorrect because SASS will select the detector in the other TRAIN (#3 in this case). It is plausible because level indication 1 & 2 are on the same train.

2nd part is incorrect because TSB 3.4.9 states: "If the level limits were exceeded prior to a transient that creates a large pressurizer insurge volume, the maximum RCS pressure could exceed the design Safety Limit (SL) of 2750 psig." It is plausible because the Pzr does become subcooled on many anticipated transients and at some point two much water would prevent the heaters from restoring the Pzr to saturated conditions.

Answer C Discussion

1st part is correct. SASS will select the detector in the other TRAIN (#3 in this case).

2nd part is correct. TSB 3.4.9 states: "If the level limits were exceeded prior to a transient that creates a large pressurizer insurge volume, the maximum RCS pressure could exceed the design Safety Limit (SL) of 2750 psig."

Answer D Discussion

1st part is correct. SASS will select the detector in the other TRAIN (#3 in this case).

2nd part is incorrect because TSB 3.4.9 states: "If the level limits were exceeded prior to a transient that creates a large pressurizer insurge volume, the maximum RCS pressure could exceed the design Safety Limit (SL) of 2750 psig." It is plausible because the Pzr does become subcooled on many anticipated transients and at some point two much water would prevent the heaters from restoring the Pzr to saturated conditions.

Cognitive Level

Memory

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT41 Q83


This question matches the KA by requiring the ability to determine how a Pzr level indication malfunction will effect Pzr level level control.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]Some examples of SRO exam items for this topic include: Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1). Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4). Knowledge of TS bases that are required to analyze TS required actions and terminology. Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).




ILT41 Q83PNS-Pzr Obj: 26TSB 3.4.9

Remarks/Status

APE028 AA2.01 - Pressurizer (PZR) Level Control MalfunctionAbility to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: (CFR: 43.5 / 45.13)

PZR level indicators and alarms ....................................




Reactor trip from 100% power

Excessive Heat Transfer on 1A SG

Rule 5 (Main Steam Line Break) initiated

The CRS transfers to the EHT tab Time = 0801:

ALL SCMs = 0°F

RCPs secured in accordance with Rule 2 (Loss of SCM)

CRS transfers to the LOSCM tab

RCS temperature = 490°F At Time = 0801... 1) Criteria __(1)__ met to perform Rule 8, Pressurized Thermal Shock. 2) In accordance with the LOSCM tab, the CRS will __(2)__ to address plant

conditions. Which ONE of the following completes the statements above? A. 1. is

2. remain in the LOSCM tab

B. 1. is

2. transfer back to the EHT tab

C. 1. is NOT

2. remain in the LOSCM tab

D. 1. is NOT

2. transfer back to the EHT tab

BWE03 EA2.1 - Inadequate Subcooling MarginAbility to determine and interpret the following as they apply tothe (Inadequate Subcooling Margin)(CFR: 43.5 / 45.13)

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.



General Discussion

Answer A Discussion

1st part is correct. Criteria (HPI injected through an open or throttled open 1HP-26, 27, 409, 410 with All RCPs off) has been met so Rule 8 is required to be performed.

2nd part is incorrect because step 3 in the LOSCM tab will "Verify EHT tab has been performed" (it has not ). The RNO is to GO TO the EHT tab. It is plausible because the LOSCM tab is a higher priority and you just transferred from the EHT tab.

Answer B Discussion

1st part is correct. Criteria (HPI injected through an open or throttled open 1HP-26, 27, 409, 410 with All RCPs off) has been met so Rule 8 is required to be performed.

2nd part is correct. Step 3 in the LOSCM tab will "Verify EHT tab has been performed" (it has not ). The RNO is to GO TO the EHT tab.

Answer C Discussion

1st part is incorrect because criteria (HPI injected through an open or throttled open 1HP-26, 27, 409, 410 with All RCPs off) has been met so Rule 8 is required to be performed. It is plausible because the other criteria (cooldown below 400 degrees Tc at > 100 degrees/hr has occurred) has not been satisfied.

2nd part is incorrect because step 3 in the LOSCM tab will "Verify EHT tab has been performed" (it has not ). The RNO is to GO TO the EHT tab. It is plausible because the LOSCM tab is a higher priority and you just transferred from the EHT tab.

Answer D Discussion

1st part is incorrect because criteria (HPI injected through an open or throttled open 1HP-26, 27, 409, 410 with All RCPs off) has been met so Rule 8 is required to be performed. It is plausible because the other criteria (cooldown below 400 degrees Tc at > 100 degrees/hr has occurred) has not been satisfied.

2nd part is correct. Step 3 in the LOSCM tab will "Verify EHT tab has been performed" (it has not ). The RNO is to GO TO the EHT tab.


This question matches the KA by requiring ability to select the appropriate procedures based on plant parameters.

Basis for Hi Cog

Basis for SRO only




Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


LOSCM tabRule 5EAP-LOSCM Obj: 2Rule 8

Remarks/Status

9/13/16 - Discussed with Dan Bacon following his review of the 10 preview questions. Added Rule 2 securing RCPs to time 0801 and changed first part to Rule 8 conditions per his suggestion. He agreed second part was OK as is.

BWE03 EA2.1 - Inadequate Subcooling MarginAbility to determine and interpret the following as they apply tothe (Inadequate Subcooling Margin)(CFR: 43.5 / 45.13)

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.




Reactor power = 100% Time = 0405:

Core SCM = 0°F

Rule 2 (Loss of SCM) is in progress

1HP-24 fails closed

RCS pressure = 1700 psig lowering 1) In accordance with Rule 2, when the step to open 1HP-26 and 1HP-27 is reached,

there will be __(1)__ HPI pumps injecting. 2) The MINIMUM number of HPI trains required to mitigate this event is __(2)__. Which ONE of the following completes the statements above? A. 1. two

2. two

B. 1. two

2. one

C. 1. three

2. two

D. 1. three

2. one

BWE13 2.1.20 - EOP RulesBWE13 GENERIC




General Discussion

Answer A Discussion

1st part is correct. With 1HP-24 failing to open, you are limited to 2 HPI pumps running per Rule 6. IAW Rule 2, 1) you will be directed to start an HPIP in the other header or 2) if ES has occurred, you will be directed to secure the 1B HPIP prior to opening 1HP26 & 27.

2nd part is correct. IAW TS bases 3.5.5.2 two HPI trains are required in this event because reactor power was greater than 75% when the event occurs.

Answer B Discussion

1st part is correct. With 1HP-24 failing to open, you are limited to 2 HPI pumps running per Rule 6. IAW Rule 2, 1) you will be directed to start an HPIP in the other header or 2) if ES has occurred, you will be directed to secure the 1B HPIP prior to opening 1HP26 & 27.

2nd part is incorrect because IAW TS bases 3.5.5.2 two HPI trains are required in this event because reactor power was greater than 75% when the event occurs. It is plausible because if power were < 75%, it could be correct.

Answer C Discussion

1st part is incorrect because with 1HP-24 failing to open, you are limited to 2 HPI pumps running per Rule 6. IAW Rule 2, 1) you will be directed to start an HPIP in the other header or 2) if ES has occurred, you will be directed to secure the 1B HPIP prior to opening 1HP26 & 27. It is plausible because SCM is 0 degrees which typically means that you must start all available HPIPs. In this case however, your available HPIPs is limited to 2 because of the suction valve.

2nd part is correct. IAW TS bases 3.5.5.2 two HPI trains are required in this event because reactor power was greater than 75% when the event occurs.

Answer D Discussion

1st part is incorrect because with 1HP-24 failing to open, you are limited to 2 HPI pumps running per Rule 6. IAW Rule 2, 1) you will be directed to start an HPIP in the other header or 2) if ES has occurred, you will be directed to secure the 1B HPIP prior to opening 1HP26 & 27. It is plausible because SCM is 0 degrees which typically means that you must start all available HPIPs. In this case however, your available HPIPs is limited to 2 because of the suction valve.

2nd part is incorrect because IAW TS bases 3.5.5.2 two HPI trains are required in this event because reactor power was greater than 75% when the event occurs. It is plausible because if power were < 75%, it could be correct.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT42 Q84


This question requires knowledge about HPI system limits and how to execute them while performing LOSCM procedure steps.

Basis for Hi Cog

Basis for SRO only





ILT42 Q84Rule 2Rule 6TS B3.5.2EAP-LOSCM Obj: 5

Remarks/Status

BWE13 2.1.20 - EOP RulesBWE13 GENERIC




Unit 1 plant conditions Time = 0800:


Chemistry results indicate DEI = 7 µCi/ml Time = 0830:

RCS leak = 140 gpm rising

Reactor tripped Time = 0900:


DEI = 287 µCi/ml

1RIA-57 = 285 R/hr

1RIA-58 = 140 R/hr 1) The EAL classification at 0800 is __(1)__. 2) The EAL classification at 0900 is __(2)__.

Based on the above plant conditions, which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. Alert


B. 1. Alert

2. General Emergency

C. 1. Unusual Event


D. 1. Unusual Event

2. General Emergency

SYS004 A2.19 - Chemical and Volume Control SystemAbility to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/ 43/5 / 45/3 / 45/5)

High secondary and primary concentrations of chloride, fluoride, sodium and solids ........................................



General Discussion

Answer A Discussion

1st part is incorrect because in RP/1000/001, Encl 4.2,System Malfunctions under the Unusual Event Column, #5 (Fuel Clad Degradation) it states that a DEI of > 5 microcuries/ml classifies as an Unusual Event. It is plausible because in Encl 4.3, Abnormal Rad Levels/Radiological Effluent. 1. A, it states that a PR sample reading of > 6.62 microcuries/ml Xe 133 for > 15 minutes would be classified as and ALERT.

2nd part is correct. Per Encl 4.1 Fission Product Barrier Matrix: With 1RIA-57 = 285 R/Hr at 0.5 hr after the reactor tripped, There are 5 pts for RCS Barrier and 5 pts for Fuel Clad Barrier. 10 points = SAE.

Answer B Discussion

1st part is incorrect because in RP/1000/001, Encl 4.2,System Malfunctions under the Unusual Event Column, #5 (Fuel Clad Degradation) it states that a DEI of > 5 microcuries/ml classifies as an Unusual Event. It is plausible because in Encl 4.3, Abnormal Rad Levels/Radiological Effluent. 1. A, it states that a PR sample reading of > 6.62 microcuries/ml Xe 133 for > 15 minutes would be classified as and ALERT.

2nd part is incorrect because per Encl 4.1 Fission Product Barrier Matrix: With 1RIA-57 = 285 R/Hr at 0.5 hr after the reactor tripped, There are 5 pts for RCS Barrier and 5 pts for Fuel Clad Barrier. 10 points = SAE. It is plausible because under the Containment Barrier column, if the RIA-57 dose rate were present 2 or more hours after the trip, it would be 1 point which would put the total at 11 points which is a GE.

Answer C Discussion

1st part is correct. Per RP/1000/001, Encl 4.2,System Malfunctions under the Unusual Event Column, #5 (Fuel Clad Degradation) it states that a DEI of > 5 microcuries/ml classifies as an Unusual Event.

2nd part is correct. Per Encl 4.1 Fission Product Barrier Matrix: With 1RIA-57 = 285 R/Hr at 0.5 hr after the reactor tripped, There are 5 pts for RCS Barrier and 5 pts for Fuel Clad Barrier. 10 points = SAE.

Answer D Discussion

1st part is correct. Per RP/1000/001, Encl 4.2,System Malfunctions under the Unusual Event Column, #5 (Fuel Clad Degradation) it states that a DEI of > 5 microcuries/ml classifies as an Unusual Event.

2nd part is incorrect because per Encl 4.1 Fission Product Barrier Matrix: With 1RIA-57 = 285 R/Hr at 0.5 hr after the reactor tripped, There are 5 pts for RCS Barrier and 5 pts for Fuel Clad Barrier. 10 points = SAE. It is plausible because under the Containment Barrier column, if the RIA-57 dose rate were present 2 or more hours after the trip, it would be 1 point which would put the total at 11 points which is a GE.


This question matches the KA by requiring knowledge of the procedure (OP and SLC) requirements when primary chemistry is not within limits.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2F. Procedures and limitations involved in initial core loading, alterations in coreconfiguration, control rod programming, and determination of various internal andexternal effects on core reactivity. [10 CFR 55.43(b)(6)]Some examples of SRO exam items for this topic include:• Evaluating core conditions and emergency classifications based on core conditions.• Administrative requirements associated with low power physics testing processes.• Administrative requirements associated with refueling activities, such as approvalsrequired to amend core loading sheets or administrative controls of potential dilutionpaths and/or activities.• Administrative controls associated with the installation of neutron sources.• Knowledge of TS bases for reactivity controls.



Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


RP/0/A/1000/001


RP/0/A/1000/001EAP SEP Obj: R12

Remarks/Status

Need new KA. Can not write a descriminating question at the SRO level for this KA.

DISCUSSED WITH CHIEF 7/27: ANY TS ON CHEMISTRY WILL MEET THIS KA.

SYS004 A2.19 - Chemical and Volume Control SystemAbility to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/ 43/5 / 45/3 / 45/5)

High secondary and primary concentrations of chloride, fluoride, sodium and solids ........................................




An RO is performing PT/1/A/600/001 (Periodic Instrument Surveillance) and is reviewing the chart of control room indicated BWST level below:

1) At Time = 1330, in accordance with PT/1/A/0600/001, the BWST __(1)__ operable. 2) In accordance with TS Bases 3.5.4, Borated Water Storage Tank, the MINIMUM

__(2)__ limits of the BWST ensure the solution in the RB Emergency sump following a LOCA is within a specified pH range.

Which ONE of the following completes the statements above? A. 1. is

2. Level

B. 1. is

2. boron concentration

C. 1. is NOT

2. Level

D. 1. is NOT

2. boron concentration

SYS006 2.4.47 - Emergency Core Cooling System (ECCS)SYS006 GENERIC

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 / 45.12)



General Discussion

Answer A Discussion

1st part is correct. Per PT/600/001 the BWST level limit is 47 feet. This is the error corrected value.

2nd part is incorrect because the minimum BWST level does NOT ensure the pH in the RBES. It is plausible because ensuring pH is the basis for limiting the maximum level in the BWST.

Answer B Discussion

1st part is correct. Per PT/600/001 the BWST level limit is 47 feet. This is the error corrected value. It is plausible because it the question was asking for the TS limit (46 feet), it would be correct.

2nd part is correct. The minimum and maximum concentration limits both ensure that the long term solution in the sump following a LOCA is within a specified pH range that will allow the BS system to minimize the evolution of iodine and the effect of chloride and caustic stress corrosion cracking on the mechanical systems and components.

Answer C Discussion

1st part is incorrect because per PT/600/001 the BWST level limit is 47 feet. This is the error corrected value. It is plausible because it the question was asking for the TS limit (46 feet), it would be correct.

2nd part is incorrect because the minimum BWST level does NOT ensure the pH in the RBES. It is plausible because ensuring pH is the basis for limiting the maximum level in the BWST.

Answer D Discussion

1st part is incorrect because per PT/600/001 the BWST level limit is 47 feet. This is the error corrected value. It is plausible because it the question was asking for the TS limit (46 feet), it would be correct.

2nd part is correct. The minimum and maximum concentration limits both ensure that the long term solution in the sump following a LOCA is within a specified pH range that will allow the BS system to minimize the evolution of iodine and the effect of chloride and caustic stress corrosion cracking on the mechanical systems and components.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT43 Q89


This question matches the KA by requiring the ability to monitor a control room trend and determine if TS requirements are met.

Basis for Hi Cog

Basis for SRO only





ILT43 Q89TSB 3.5.4 PT/1/A/0600/001ADM-TSSP Obj: 4

Remarks/Status

SYS006 2.4.47 - Emergency Core Cooling System (ECCS)SYS006 GENERIC

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 / 45.12)




LOHT tab is in progress

CBP feed is providing feed to both SGs

The DC breaker to the TD EFDW pump Auxiliary Oil Pump tripped

An Auxiliary Operator has been dispatched to manually start the TD EFDW pump

Current conditions:

The TD EFDW pump is supplying feed to both SGs

Heat transfer has been verified 1) When the TD EFDW pump Auxiliary Oil pump failed, __(1)__ failed to open. 2) Based on current conditions, the LOHT tab directs you to GO TO the __(2)__ tab. Which ONE of the following completes the statements above? A. 1. 1MS-93 (TD EFDWP Steam Supply Trip Valve)

2. Subsequent Actions

B. 1. 1MS-93 (TD EFDWP Steam Supply Trip Valve)

2. Forced Cooldown

C. 1. 1MS-95 (TD EFWP Governor Valve)

2. Subsequent Actions

D. 1. 1MS-95 (TD EFWP Governor Valve)

2. Forced Cooldown

SYS061 A2.03 - Auxiliary / Emergency Feedwater (AFW) SystemAbility to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Loss of dc power ................................................



General Discussion

Answer A Discussion

1st part is incorrect because 1MS-95 is opened by hydraulic oil pressure. It is plasuible because 1MS-93 also receives an open signal when the TD EFDW pump starts. It will open because it is air operated.

2nd part is correct. When the LOHT tab is complete, it directs you to the SA tab of the EOP.

Answer B Discussion

1st part is incorrect because 1MS-95 is opened by hydraulic oil pressure. It is plasuible because 1MS-93 also receives an open signal when the TD EFDW pump starts. It will open because it is air operated.

2nd part is incorrect because when the LOHT tab is complete, it directs you to the SA tab of the EOP. It is plausible because if either SG were isolated, it would be correct. CBP feed was added to make this more plausible.

Answer C Discussion

1st part is correct. 1MS-95 is opened by hydraulic oil pressure. When the AOP failed to start, there was no hydraulic pressure to open the valve.

2nd part is correct. When the LOHT tab is complete, it directs you to the SA tab of the EOP.

Answer D Discussion

1st part is correct. 1MS-95 is opened by hydraulic oil pressure. When the AOP failed to start, there was no hydraulic pressure to open the valve.

2nd part is incorrect because when the LOHT tab is complete, it directs you to the SA tab of the EOP. It is plausible because if either SG were isolated, it would be correct. CBP feed was added to make this more plausible.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


This question matches the KA by requiring the ability to predict the impact of a loss of DC power to the EFW system and how procedures are used to address this issue.

Basis for Hi Cog

Basis for SRO only





LOHT tabCF-EF Obj: 18

Remarks/Status

KA nomenclature in the NRC provided sample plan for 061 A2.03 doesn't match KA Catalog:A2.03 is "Loss of DC power" in the catalog.In the sample plan it states "Flow rates expected from various combinations of AFW pump discharge valves" which is A2.08 in the KA Catalog.Need to discuss with Chief.

DISCUSSED WITH CHIEF 7/27: LOSS OF DC POWER IS CORRECT K/A.

SYS061 A2.03 - Auxiliary / Emergency Feedwater (AFW) SystemAbility to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Loss of dc power ................................................




Unit 1 = 100% power

Unit 2 = 100% power

Load center 1X8 is to be powered from LC 2X11 to allow maintenance to be performed

1) In order to prevent powering 1X8 and 2X8 from LC 2X11 at the same time, __(1)__.

2) When in this lineup, entry into TS 3.8.8 (Electrical Power Systems, Distribution Systems – Operating) __(2)__ required for Unit 1.

Which ONE of the following completes the statements above? A. 1. a kirk-key interlock prevents closing the individual breakers to 1X8 and 2X8

at the same time

2. is

B. 1. a kirk-key interlock prevents closing the individual breakers to 1X8 and 2X8 at the same time

2. is NOT

C. 1. there is ONLY one breaker in LC 2X11 which is moved to either the 1X8 or 2X8 position to supply that load

2. is

D. 1. there is ONLY one breaker in LC 2X11 which is moved to either the 1X8 or 2X8 position to supply that load

2. is NOT

SYS062 A2.06 - AC Electrical Distribution SystemAbility to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Keeping the safeguards buses electrically separate ....................



General Discussion

Answer A Discussion

1st part is incorrect because there is only one breaker in the load center. It is plausible because this is the method used in other similar applications (SSF power supplies, Vital AC power supplies).

2nd part is correct. TS 3.8.8 is applicable because the ES Power String that includes 1X8 is inoperable when powered from Unit 2.

Answer B Discussion

1st part is incorrect because there is only one breaker in the load center. It is plausible because this is the method used in other similar applications (SSF power supplies, Vital AC power supplies).

2nd part is incorrect because TS 3.8.8 is applicable due to the ES Power String that includes 1X8 being inoperable when powered from Unit 2. It is plausible because the ES string is still receiving power and separate from the other Unit 1 power strings.

Answer C Discussion

1st part is correct. There is only one breaker in the load center. It has to be moved to the designated location in the breaker cabinet and then racked in to provide power.

2nd part is correct. TS 3.8.8 is applicable because the ES Power String that includes 1X8 is inoperable when powered from Unit 2.

Answer D Discussion

1st part is correct. There is only one breaker in the load center. It has to be moved to the designated location in the breaker cabinet and then racked in to provide power.

2nd part is incorrect because TS 3.8.8 is applicable due to the ES Power String that includes 1X8 being inoperable when powered from Unit 2. It is plausible because the ES string is still receiving power and separate from the other Unit 1 power strings.

Cognitive Level

Memory

Job Level

SRO

QuestionType

NEW

Question Source


TS 3.8.8TSB 3.8.8


This question matches the KA by requiring knowledge of how safeguards busses are kept electrically separate and how procedures (TS) are used to limit the time that one units ES string can be powered from a different unit.

Basis for Hi Cog

Basis for SRO only




EL-EPD Obj: 23/24/25OP/0/A/1107/007

Remarks/Status

9/13/16 - Discussed the kirk key interlock and EL-EPD drawing (page 151 of 165) with Dan Bacon following his review of the 10 preview questions. Determined the drawing matches the lesson plan and procedure. Question OK.

SYS062 A2.06 - AC Electrical Distribution SystemAbility to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Keeping the safeguards buses electrically separate ....................




Reactor shutdown is in progress in accordance with OP/1/A/1102/004 (Operation at Power)

Reactor power = 28% 1) The loss of panelboard __(1)__ would result in a loss of power to the Main Turbine

Electro Hydraulic Control. 2) When the above failure occurs, the SRO will __(2)__. Which ONE of the following completes the statements above? A. 1. 1DIA

2. stay in OP/1/A/1102/004

B. 1. 1DIA

2. enter the EOP

C. 1. 1DP 2. stay in OP/1/A/1102/004

D. 1. 1DP

2. enter the EOP

SYS063 2.4.2 - DC Electrical Distribution SystemSYS063 GENERIC

Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8)



General Discussion

Answer A Discussion

1st part is correct. Panel board 1DIA is the power supply to the EHC controls.

2nd part is incorrect because with reactor = 28% power DECREASING, the reactor will trip which will require EOP entry. It is plausible because if power were increasing, it could be correct. The reactor trip based on a main turbine trip does not arm until 29.75% power when power is increasing and does no bypass when decreasing power until 27.75% power. Since power is decreasing, the reactor trip is still armed.

Answer B Discussion

1st part is correct. Panel board 1DIA is the power supply to the EHC controls.

2nd part is correct because with reactor = 28% power DECREASING, the reactor will trip which will require EOP entry. The reactor trip based on a main turbine trip does not arm until 29.75% power when power is increasing and does no bypass when decreasing power until 27.75% power. Since power is decreasing, the reactor trip is still armed.

Answer C Discussion

1st part is incorrect because panel board 1DIA is the power supply to the EHC controls. It is plausible because1DP is the power supply to the Main Turbine Emergency Oil pumps.

2nd part is incorrect because with reactor = 28% power DECREASING, the reactor will trip which will require EOP entry. It is plausible because if power were increasing, it could be correct. The reactor trip based on a main turbine trip does not arm until 29.75% power when power is increasing and does no bypass when decreasing power until 27.75% power. Since power is decreasing, the reactor trip is still armed.

Answer D Discussion

1st part is incorrect because panel board 1DIA is the power supply to the EHC controls. It is plausible because1DP is the power supply to the Main Turbine Emergency Oil pumps.

2nd part is correct because with reactor = 28% power DECREASING, the reactor will trip which will require EOP entry. The reactor trip based on a main turbine trip does not arm until 29.75% power when power is increasing and does no bypass when decreasing power until 27.75% power. Since power is decreasing, the reactor trip is still armed.


This question matches the KA by requiring knowledge of how a loss of DC power to certain loads will result in EOP entry.

Basis for Hi Cog

Basis for SRO only




Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


IC-RPS Obj: 6EL-DCD Obj: 19STG-EHC Obj: 25EL-VPC

Remarks/Status

DISCUSSED WITH CHIEF 7/27: ASK ANY DC LOAD THAT WOULD CAUSE TURBINE TRIP AT POWER OR REACTOR TRIP.

SYS063 2.4.2 - DC Electrical Distribution SystemSYS063 GENERIC

Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8)





Pressurizer (Pzr) Level 3 selected

SASS in MANUAL

ICCM Train "1B" experiences a total loss of power 1) Due to the loss of power, 1HP-120 will __(1)__. 2) If power can NOT be restored, TS 3.3.8 (Post Accident Monitoring Instrumentation)

__(2)__ require a shutdown within 12 hours. Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. close

2. does

B. 1. close

2. does NOT

C. 1. open

2. does

D. 1. open

2. does NOT

SYS011 2.2.22 - Pressurizer Level Control System (PZR LCS)SYS011 GENERIC

Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2)



General Discussion

Answer A Discussion

1st part is incorrect. ICCM Train "A" feeds Pzr level 1 and 2. Train B feeds Pzr level 3. A loss of power to 1B ICCM Train will cause Pzr Level 3 to fail low and cause 1HP-120 to fully open. It is plausible because if power were lost to 1HP-120, it could be correct. Air operated valves failure position on a loss of power typically occur because air is vented off. Numerous valves in the HPI system (5, 6 & 21) fail closed if air pressure to the valve goes away.

2nd part is incorrect because TS 3.3.8 will not require a shutdown. It is plausible because if the applicant goes directly to the Table in TS 3.3.8 for Pzr level, it states that 2 channels are required. If 2 channels are not available, it directs you to condition H. Condition H requires a shutdown to M-3 within 12 hours if the condition cannot be corrected (2 trains are required and Level 3 (ICCM Tr B) is one of those trains).

Answer B Discussion

1st part is incorrect. ICCM Train "A" feeds Pzr level 1 and 2. Train B feeds Pzr level 3. A loss of power to 1B ICCM Train will cause Pzr Level 3 to fail low and cause 1HP-120 to fully open. It is plausible because if power were lost to 1HP-120, it could be correct. Air operated valves failure position on a loss of power typically occur because air is vented off. Numerous valves in the HPI system (5, 6 & 21) fail closed if air pressure to the valve goes away.

2nd part is correct. TS 3.3.8 Condition A applies which allows 30 days to restore to operable status. Unless both channels are inoperable, you do not reference the table 3.3.8-1.

Answer C Discussion

1st part is correct because Pzr level 3 fails low, causing 1HP-120 to open in response.

2nd part is incorrect because TS 3.3.8 will not require a shutdown. It is plausible because if the applicant goes directly to the Table in TS 3.3.8 for Pzr level, it states that 2 channels are required. If 2 channels are not available, it directs you to condition H. Condition H requires a shutdown to M-3 within 12 hours if the condition cannot be corrected (2 trains are required and Level 3 (ICCM Tr B) is one of those trains).

Answer D Discussion

1st part is correct because Pzr level 3 fails low, causing 1HP-120 to open in response

2nd part is correct. TS 3.3.8 Condition A applies which allows 30 days to restore to operable status. Unless both channels are inoperable, you do not reference the table 3.3.8-1.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT47 Q91


This question matches the KA by requiring knowledge of the impact of Pzr level indication failing low on system operation.

Basis for Hi Cog

Basis for SRO only





TS 3.3.8


ILT47 Q91PNS-PZR Obj: 21TS 3.3.8TS 3.3.8 B

Remarks/Status

SYS011 2.2.22 - Pressurizer Level Control System (PZR LCS)SYS011 GENERIC

Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 / 43.2 / 45.2)





Electrical malfunction results in loss of ALL Control Rod position indications 1) In accordance with Tech Spec 3.1.7 (Position Indicator Channels), the

MAXIMUM time allowed to declare all Control Rods inoperable is __(1)__. 2) Assuming indications are NOT restored, Tech Spec 3.1.4 (Control Rod Group

alignment limits) __(2)__ require reducing Reactor Power to a MAXIMUM of 60% RTP within 2 hours of declaring all Control Rods inoperable.

Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. immediately

2. does

B. 1. immediately

2. does NOT

C. 1. one hour

2. does

D. 1. one hour

2. does NOT

SYS016 A2.02 - Non-Nuclear Instrumentation System (NNIS)Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Loss of power supply .............................................



General Discussion

Answer A Discussion

1st part is correct. TS 3.1.7 required declaring all rods that have no position indication operable inoperable immediately.

2nd part is correct. This puts you in TS 3.1.4 where both Condition A and Condition C apply. While Condition C requires Mode 3 within 12 hours, the Condition A requirement to be below 60% RTP within 2 hours would also still be applicable.

Answer B Discussion

1st part is correct. TS 3.1.7 required declaring all rods that have no position indication operable inoperable immediately.

2nd part is incorrect because this puts you in TS 3.1.4 where both Condition A and Condition C apply. While Condition C requires Mode 3 within 12 hours, the Condition A requirement to be below 60% RTP within 2 hours would also still be applicable. It is plausible since that requirement stems from Condition A. Condition A applies with one Control Rod inoperable. It is a common misconception when applying Tech specs that you are only in the specific CONDITION that matches plant status. i.e. I would not be inCondition A since it is for one inoperable rod because I have more than one inoperable rod. That misconception would lead the candidate to believe that only the requirements of Condition C apply. Further plausibility comes from the fact that Oconee does have an outlier SLC (16.8.3) that does work as the above "misconception" states. There is a condition for "a single battery" being inoperable and the condition only applies when there is one and ONLY one battery inoperable.

Answer C Discussion

1st part is incorrect because TS 3.1.7 required declaring all rods that have no position indication operable inoperable immediately. It is is plausible because there are many Tech Specs that allow 1 hr as a competition time and the 1 hr CT would mean that thiswould still be a required memory item.

2nd part is correct. This puts you in TS 3.1.4 where both Condition A and Condition C apply. While Condition C requires Mode 3 within 12 hours, the Condition A requirement to be below 60% RTP within 2 hours would also still be applicable.

Answer D Discussion

1st part is incorrect because TS 3.1.7 required declaring all rods that have no position indication operable inoperable immediately. It is is plausible because there are many Tech Specs that allow 1 hr as a competition time and the 1 hr CT would mean that thiswould still be a required memory item.

2nd part is incorrect because this puts you in TS 3.1.4 where both Condition A and Condition C apply. While Condition C requires Mode 3 within 12 hours, the Condition A requirement to be below 60% RTP within 2 hours would also still be applicable. It is plausible since that requirement stems from Condition A. Condition A applies with one Control Rod inoperable. It is a common misconception when applying Tech specs that you are only in the specific CONDITION that matches plant status. i.e. I would not be inCondition A since it is for one inoperable rod because I have more than one inoperable rod. That misconception would lead the candidate to believe that only the requirements of Condition C apply. Further plausibility comes from the fact that Oconee does have an outlier SLC (16.8.3) that does work as the above "misconception" states. There is a condition for "a single battery" being inoperable and the condition only applies when there is one and ONLY one battery inoperable..


This question matches the KA by requiring knowledge of procedures to mitigate consequences of loss of power to RPIS. Predicting the impact is understanding that all Control Rods must be delcared inoperable and procedure use is met by applying the requirments of TS 3.1.4.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]Some examples of SRO exam items for this topic include: Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1). Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4). Knowledge of TS bases that are required to analyze TS required actions and



Cognitive Level

Comprehension

Job Level

SRO

QuestionType

BANK

Question Source

ILT48 Q92


TS 3.1.4


ILT48 Q92ADM-TSSP Obj: 5TS 3.1.4TS 3.1.7

Remarks/Status

terminology. Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).

SYS016 A2.02 - Non-Nuclear Instrumentation System (NNIS)Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Loss of power supply .............................................




ICC tab in progress

Depressurization of both SGs is in progress 1) The MINIMUM CETC which would require entry into the OSAG is __(1)__. 2) In the ICC tab, the RB Aux fans are started to __(2)__ the containment

atmosphere. Which ONE of the following completes the statements above? A. 1. 700°F

2. mix

B. 1. 700°F

2. cool

C. 1. 1200°F

2. mix

D. 1. 1200°F

2. cool

SYS017 2.2.38 - In-Core Temperature Monitor (ITM) SystemSYS017 GENERIC

Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)



General Discussion

Answer A Discussion

1st part is incorrect because entry into the OSAG is not directed to the TSC unless CETCs are > 1200 degrees. It is plausible because with CETCs temperatures between 700 and 1200 degrees a RCP is started in the ICC tab even though there is no subcooling.

2nd part is correct. The RB Aux fans are started to improve mixing and to obtain a better analyzer indication of the hydrogen concentration in containment.

Answer B Discussion

1st part is incorrect because entry into the OSAG is not directed to the TSC unless CETCs are > 1200 degrees. It is plausible because with CETCs temperatures between 700 and 1200 degrees a RCP is started in the ICC tab even though there is no subcooling.

2nd part is incorrect because the RB Aux fans are started to improve mixing and to obtain a better analyzer indication of the hydrogen concentration in containment. It is plausible because the Aux fans do normally provide RB cooling (but that is not why they are started in this case).

Answer C Discussion

1st part is correct. Entry into the OSAG is directed to the TSC when CETCs are > 1200 degrees.

2nd part is correct. The RB Aux fans are started to improve mixing and to obtain a better analyzer indication of the hydrogen concentration in containment.

Answer D Discussion

1st part is correct. Entry into the OSAG is directed to the TSC when CETCs are > 1200 degrees.

2nd part is incorrect because the RB Aux fans are started to improve mixing and to obtain a better analyzer indication of the hydrogen concentration in containment. It is plausible because the Aux fans do normally provide RB cooling (but that is not why they are started in this case).


This question matches the KA by requiring the ability to evaluate plant performance and make operational judgments based on CETC's.

Basis for Hi Cog

Basis for SRO only




Cognitive Level

Memory

Job Level

SRO

QuestionType

BANK

Question Source

ILT39 Q92


ILT39 Q92EAP-ICC R11ICC tab

Remarks/Status

Ask Chief if minimum CETC for OSAG entry meets KA.

DISCUSSED WITH CHIEF 7/27: OK WITH OSAG MEETING KA.

SYS017 2.2.38 - In-Core Temperature Monitor (ITM) SystemSYS017 GENERIC

Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)





Main T/G Journal Bearing #7 vibration is indicated below:

Which ONE of the following describes: 1) The EARLIEST time that the Main Turbine must be manually tripped in accordance

with Limits and Precautions of OP/1/A/1106/001 (Turbine Generator)? 2) The procedure that would be used to take the Main Turbine off line without tripping

the Reactor? A. 1. 1216

2. AP/1/A/1700/029 (Rapid Unit Shutdown)

B. 1. 1216

2. OP/1/A/1106/001 (Turbine Generator)

C. 1. 1224

2. AP/1/A/1700/029 (Rapid Unit Shutdown)

D. 1. 1224

2. OP/1/A/1106/001 (Turbine Generator)


Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12)



General Discussion

Answer A Discussion

1st part is incorrect because 1224 represents 12 mils which is the setpoint requiring immediate tripping of the Main Turbine if at rated speed. It is plausible because if vibrations were at 10 mils for 15 minutes, it would be correct.

2nd part is correct. AP/29 would have to be used to prevent a reactor trip as it contains guidance to simply trip the turbine once below 18% CTP. Using OP/1106/01 would not allow enough time to get the MT off line prior to reaching 12 mils. Also, unless AP/29 is used, Rx power cannot get low enough to take the MT off line without tripping the Rx in the time available.

Answer B Discussion

1st part is incorrect because 1224 represents 12 mils which is the setpoint requiring immediate tripping of the Main Turbine if at rated speed. It is plausible because if vibrations were at 10 mils for 15 minutes, it would be correct.

2nd part is incorrect because AP/29 would have to be used to prevent a reactor trip as it contains guidance to simply trip the turbine once below 18% CTP. Using OP/1106/01 would not allow enough time to get the MT off line prior to reaching 12 mils. Also, unless AP/29 is used, Rx power cannot get low enough to take the MT off line without tripping the Rx in the time available. It is plausible because it is how the Main Turbine is taken off line in situations that are less time critical. Additionally, it would be plausible to believe that AP/29 would direct taking the MT off line using this procedure instead of just directing theoperator to trip the MT.

Answer C Discussion

1st part is correct. 1224 represents 12 mils which is the setpoint requiring immediate tripping of the Main Turbine if at rated speed.

2nd part is correct. AP/29 would have to be used to prevent a reactor trip as it contains guidance to simply trip the turbine once below 18% CTP. Using OP/1106/01 would not allow enough time to get the MT off line prior to reaching 12 mils. Also, unless AP/29 is used, Rx power cannot get low enough to take the MT off line without tripping the Rx in the time available.

Answer D Discussion

1st part is correct. 1224 represents 12 mils which is the setpoint requiring immediate tripping of the Main Turbine if at rated speed.

2nd part is incorrect because AP/29 would have to be used to prevent a reactor trip as it contains guidance to simply trip the turbine once below 18% CTP. Using OP/1106/01 would not allow enough time to get the MT off line prior to reaching 12 mils. Also, unless AP/29 is used, Rx power cannot get low enough to take the MT off line without tripping the Rx in the time available. It is plausible because it is how the Main Turbine is taken off line in situations that are less time critical. Additionally, it would be plausible to believe that AP/29 would direct taking the MT off line using this procedure instead of just directing theoperator to trip the MT.


This question matches the KA by requiring the ability to apply the limits on Main Turbine Vibration.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2E. Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal,abnormal, or emergency) and then 2) selecting a procedure or section of a procedureto mitigate, recover, or with which to proceed. One area of SRO level knowledge (withrespect to selecting a procedure) is knowledge of the content of the procedure versusknowledge of the procedure’s overall mitigative strategy or purpose.The applicant’s knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) byensuring that the additional knowledge of the procedure’s content is required tocorrectly answer the written test item, for example: Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. Knowledge of diagnostic steps and decision points in the emergency operating



Cognitive Level

Comprehension

Job Level

SRO

QuestionType

BANK

Question Source

ILT44 Q95


ILT44 Q95STG-MT Obj: 20AP/29OP 1 A 1106 001

Remarks/Status

procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures. Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.


Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12)




Refueling in progress

FTC level = 22 feet stable

No water additions are being made to the system

2A LPI train is operable and in service Current conditions:

Refueling SRO desires stopping the 2A LPI pump to aid in inserting a fuel assembly

2A LPI pump has been in continuous operation for the previous 24 hours Which ONE of the following describes whether the 2A LPI pump may be stopped in accordance with OP/2/A/1502/007 (Operations Defueling/Refueling Responsibilities) AND the bases for this action? A. 2A LPI pump may be stopped

FTC provides adequate backup decay heat removal

B. 2A LPI pump may be stopped

Spent Fuel Cooling system provides adequate backup decay heat removal

C. 2A LPI pump may NOT be stopped

FTC does NOT provide adequate backup decay heat removal

D. 2A LPI pump may NOT be stopped

Spent Fuel Cooling does NOT provide adequate backup decay heat removal


Knowledge of refueling administrative requirements. (CFR: 41.10 / 43.5 / 45.13)



General Discussion

Answer A Discussion

Correct. TS 3.9.4 (Refueling Ops- DHR and Coolant Circulation (High Water Level) is in effect as water level is ≥ 21.34 ft. This condition requires only 1 DHR loop to be operable and in service since the water can provide adequate backup decay heat removal. TS and Refueling procedures limits & precautions allow SRO to grant permission for the operating loop to be secured for up to 1 hour every 8 hours with adequatelevel.

Answer B Discussion

Incorrect. 1st part is correct.

2nd part is incorrect because this is not the bases for allowing the pump to be secured for up to one hour. It is plausible since Spent Fuel Cooling (SFC) helps to provide decay heat removal.

Answer C Discussion

1st part is incorrect because the 2A LPI pump can be stopped. It is plausible if TS 3.9.5 criteria are misapplied to this situation. TS3.9.5 (DHR and Coolant Circulation LowWater Level) requires 2 operable DHR loops with one loop in service (no time is allowed for a pump to be secured).

2nd part is incorrect because the FTC does provide adequate backup decay heat removal. It is plausible in that it would be true if level was below the refueling level of 21.34 ft.

Answer D Discussion

1st part is incorrect because the 2A LPI pump can be stopped. It is plausible if TS 3.9.5 criteria are misapplied to this situation. TS3.9.5 (DHR and Coolant Circulation LowWater Level) requires 2 operable DHR loops with one loop in service (no time is allowed for a pump to be secured).

2nd part is incorrect because it is not the bases used for the criteria to secure an LPI pump. It is plausible because the Spent Fuel Cooling (SFC) helps to provide decay heat removal.

Cognitive Level

Memory

Job Level

SRO

QuestionType

BANK

Question Source

ILT47 Q94


This question matches the KA by requiring knowledge of administrative requirements (Tech Spec) for securing the only operating LPI pump during fuel handling.

Basis for Hi Cog

Basis for SRO only

Requires knowledge of procedures and limitations involved in core alterations (43.6), TS and Bases (43.2). Requires knowledge of the bases of TS 3.9.4.





ILT47 Q941502/07TS 3.9.4TSB 3.9.4ADM-TSSP Obj: 4

Remarks/Status


Knowledge of refueling administrative requirements. (CFR: 41.10 / 43.5 / 45.13)



In accordance with AD-EG-ALL-1132 (Preparation and Control of Design Change Engineering Changes):

1) An on-line temporary design change is required to have a plan that specifies removal of the change within __(1)__ from installation.

2) The Operational Control Group (Operations) __(2)__ responsible for maintaining a log of installed changes.

Which ONE of the following completes the statements above? A. 1. thirty days

2. is

B. 1. thirty days

2. is NOT

C. 1. one year

2. is

D. 1. one year

2. is NOT


Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13)



General Discussion

Answer A Discussion

1st part is incorrect because the time specified is 1 year. It is plausible because 30 days is an often used surveillance time.

2nd part is correct. OCG (operations) is responsible for maintaining the log.

Answer B Discussion

1st part is incorrect because the time specified is 1 year. It is plausible because 30 days is an often used surveillance time.

2nd part is incorrect because the OCG maintains the log. Plausible since the process is "owned" by engineering.

Answer C Discussion

1st part is correct. Per AD-EG-ALL-1132, a plan for removal within one year must be in place before temporary design change will be installed.

2nd part is correct. The operational control group (OPS in this case) is required to maintain a log of installed changes.

Answer D Discussion

1st part is correct. Per AD-EG-ALL-1132, a plan for removal within one year must be in place before temporary design change will be installed.

2nd part is incorrect because the OCG maintains the log. Plausible since the process is "owned" by engineering.

Cognitive Level

Memory

Job Level

SRO

QuestionType

BANK

Question Source

ILT 16-1 Q95


ILT16-1 Q95ADM-SD Obj: 13, 14 AD-EG-ALL-1132

Remarks/Status


This question matches the KA by requiring knowledge of the process for controlling Temporary Design Changes.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2C. Facility licensee procedures required to obtain authority for design and operatingchanges in the facility. [10 CFR 55.43(b)(3)]Some examples of SRO exam items for this topic include: 10 CFR 50.59 screening and evaluation processes. Administrative processes for temporary modifications. Administrative processes for disabling annunciators. Administrative processes for the installation of temporary instrumentation. Processes for changing the plant or plant procedures.


Knowledge of the process for controlling temporary design changes. (CFR: 41.10 / 43.3 / 45.13)




A main steam line break occurred inside containment

The EHT tab was performed

The crew transferred to the Forced Cooldown (FCD) Tab Time = 0830:

The decision has been made to perform a natural circulation cooldown Time = 1500:

RCS temperature = 240°F

RCS pressure = 250 psig 1) At this point in the cooldown, the FCD tab directs using the __(1)__ to complete

the RCS cooldown. 2) Transition to OP/2/A/1102/010 (Controlling Procedure for Unit Shutdown) is done

__(2)__. Which ONE of the following completes the statements above? A. 1. Normal Decay Heat Removal Mode

2. ONLY after the LPI alignment in (1) above is made

B. 1. Normal Decay Heat Removal Mode

2. to perform the alignment directed in (1) above

C. 1. LPI Series Mode

2. ONLY after the LPI alignment in (1) above is made

D. 1. LPI Series Mode

2. to perform the alignment directed in (1) above





General Discussion

Answer A Discussion

1st part is incorrect because you will use Series Mode. It is plausible because if it were Unit 3, it would be correct.

2nd part is correct. As directed in the FCD tab, the transition is made to 1102/010 after LPI DHR is established and temperature is 150 to 180 degrees.

Answer B Discussion

1st part is incorrect because you will use Series Mode. It is plausible because if it were Unit 3, it would be correct.

2nd part is incorrect but plausible since this procedure does contain directions required to align LPI cooling and under other conditions the OP is what is used to perform alignment to LPI cooling.

Answer C Discussion

1st part is correct because you are directed by the Unit 2 FCD tab to line up for Series mode.

2nd part is correct. As directed in the FCD tab, the transition is made to 1102/010 after LPI DHR is established and temperature is 150 to 180 degrees.

Answer D Discussion

1st part is correct because you are directed by the Unit 2 FCD tab to line up for Series mode.

2nd part is incorrect but plausible since this procedure does contain directions required to align LPI cooling and under other conditions the OP iswhat is used to perform alignment to LPI cooling.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT47 Q85


This question matches the KA by requiring knowledge of the differences in Units procedural actions when perform a natural circulation cooldown.

Basis for Hi Cog

Basis for SRO only





ILT47 Q85FCD tab U2FCD tab U3EAP FCD Obj: R8

Remarks/Status






Shift is preparing for ONE (1) GWR release of the 1A Gaseous Waste Disposal (GWD) tank at the 1/3 Station Limit

The release will be through the P/A/C filter

1A GWD tank holdup time is 41 days 1) The Shift Manager (SM) __(1)__ the MINIMUM level of authority for approval of the

above release in accordance with OP/1-2/A/1104/018 (GWD System). 2) The SLC basis for limiting the Curie content of the GWD Tanks limits Whole Body

exposure of an individual at the nearest exclusion boundary to < __(2)__. Which ONE of the following completes the statements above? A. 1. is

2. 0.5 Rem in the event of a GWD tank rupture

B. 1. is

2. 100 mrem during a planned GWD tank release

C. 1. is NOT

2. 0.5 Rem in the event of a GWD tank rupture

D. 1. is NOT

2. 100 mrem during a planned GWD tank release


Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10)



General Discussion

Answer A Discussion

1st part is incorrect because OSM approval not required. Plausible because if the hold up time was < 30 days or the PAC filter was not used, it would be correct.

2nd part is correct.

Answer B Discussion

1st part is incorrect because OSM approval not required. Plausible because if the hold up time was < 30 days or the PAC filter was not used, it would be correct.

2nd part is incorrect. Plausible because yearly limit at site boundary for unmonitored personnel is 100 mrem.

Answer C Discussion

Correct.1st part is correct. With > 30 day holdup time, only SRO approval is required.

2nd part is correct: Basis for limit in GWD tanks based on exposure at site boundary of .5 REM.

Answer D Discussion

1st part is correct. With > 30 day holdup time, only SRO approval is required.

2nd part is incorrect. Plausible because yearly limit at site boundary for unmonitored personnel is 100 mrem.

Cognitive Level

Comprehension

Job Level

SRO

QuestionType

BANK

Question Source

ILT46 Q98


ILT46 Q98SLC 16.11.13OP 1-2 A 1104 018WE GWD Obj: 7


KA matched because the test item evaluates the SROs understanding of conditions that will determine approval authority of a gaseous radioactive release

Basis for Hi Cog

Basis for SRO only

NUREG 1021 ES 401 Attachment 2D. Radiation hazards that may arise during normal and abnormal situations, includingmaintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]Some examples of SRO exam items for this topic include:• Process for gaseous/liquid release approvals, i.e., release permits.• Analysis and interpretation of radiation and activity readings as they pertain toselection of administrative, normal, abnormal, and emergency procedures.Analysis and interpretation of coolant activity, including comparison to emergencyplan criteria and/or regulatory limits.


Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10)



Remarks/Status

9/13/16 - Received new K/A (Gen2.3 2.3.6) from Dan Bacon following 10 preview question review. Could not write a discriminating SRO ONLY question to the original K/A.




Reactor in MODE 5

A and B LPSW pumps failed due to a seismic event

RCS temperature 131°F slowly rising Time = 1205:

ALERT declared in accordance with RP/0/A/1000/001 (Emergency Classification)

Time = 1430:

Plant conditions require escalating the Emergency Plan classification to an SITE AREA EMERGENCY

1) __(1)__ is a condition that will require cross connecting with Unit 3's LPSW system? 2) __(2)__ has the responsibility to upgrade the classification at TIME = 1430 in

accordance with the Emergency Coordinator Procedures. Which ONE of the following completes the statements above? A. 1. Loss of C LPSW Pump

2. TSC

B. 1. Loss of C LPSW Pump

2. SM

C. 1. RCS temperature = 205°F slowly increasing

2. TSC

D. 1. RCS temperature = 205°F slowly increasing

2. SM


Knowledge of crew roles and responsibilities during EOP usage. (CFR: 41.10 / 45.12)



General Discussion

Answer A Discussion

1st part is correct. AP/24 states that if NO unit 1 & 2 LPSW pumps are available and Unit 3 LPSW system is available..cross connect.

2nd part is correct. When the ALERT is declared at 1205, the TSC is required to be manned by 1305. At this point, the TSC would make the upgrade at 1430.

Answer B Discussion

1st part is correct. AP/24 states that if NO unit 1 & 2 LPSW pumps are available and Unit 3 LPSW system is available..cross connect.

2nd part is incorrect because when the ALERT is declared at 1205, the TSC is required to be manned by 1305 so the TSC would make the upgrade at 1430. It is plausible because if it were a NOUE at 1205, it could be correct.

Answer C Discussion

1st part is incorrect because AP/24 states that if NO unit 1 & 2 LPSW pumps are available and Unit 3 LPSW system is available..cross connect. It is plausible because going above 200 degrees would result in an increase in the EPLAN classification to an ALERT so to avoid escalation, cross connecting to Unit 3 LPSW would seem to be the correct preventative measure.

2nd part is correct. When the ALERT is declared at 1205, the TSC is required to be manned by 1305. At this point, the TSC would make the upgrade at 1430.

Answer D Discussion

1st part is incorrect because AP/24 states that if NO unit 1 & 2 LPSW pumps are available and Unit 3 LPSW system is available..cross connect. It is plausible because going above 200 degrees would result in an increase in the EPLAN classification to an ALERT so to avoid escalation, cross connecting to Unit 3 LPSW would seem to be the correct preventative measure.

2nd part is incorrect because when the ALERT is declared at 1205, the TSC is required to be manned by 1305 so the TSC would make the upgrade at 1430. It is plausible because if it were a NOUE at 1205, it could be correct.

Cognitive Level

Memory

Job Level

SRO

QuestionType

MODIFIED

Question Source

ILT40 Q78


Requires knowledge of the lines of authority required during Emergency Planimplementation. Since RP/1000/02 is only required to be performed during implementation of the emergency plan and this question requires knowledge of who hasthe authority to change the Eplan classification. This is tied to a loss of DHR becausethere are other events that can change the answer to the second part of the question

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]Some examples of SRO exam items for this topic include: Evaluating core conditions and emergency classifications based on core conditions. Administrative requirements associated with low power physics testing processes. Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities. Administrative controls associated with the installation of neutron sources. Knowledge of TS bases for reactivity controls.




ILT40 Q78EAP-SEP Obj: R3RP/1000/001RP/1000/002AP/24

Remarks/Status


Knowledge of crew roles and responsibilities during EOP usage. (CFR: 41.10 / 45.12)



Given the following site conditions:

Keowee reservoir elevation is 806 feet

A Site Area Emergency has been declared due to imminent failure of the Keowee Hydro Dam

1) Anytime that Protective Action Recommendations are required, State and County

agencies are required to be notified within a MAXIMUM of __(1)__

2) In accordance with RP/0/A/1000/002 (Control Room Emergency Coordinator Procedure), Protective Action Recommendations __(2)__ required for this event.

Which ONE of the following completes the statements above? REFERENCE PROVIDED A. 1. 15 minutes

2. are

B. 1. 15 minutes

2. are NOT

C. 1. 1 hour

2. are

D. 1. 1 hour

2. are NOT


Knowledge of emergency plan protective action recommendations. (CFR: 41.10 / 41.12 / 43.5 / 45.11)



General Discussion

Answer A Discussion

1st part is correct. Per RP/0/A/1000/002, State and County agencies shall be notified within 15 minutes of E-plan declaration, Classification upgrades, and Protective Action Recommendations.

2nd part is correct. For Immanent dam failure, RP/0/A/100/002, Condition A response actions are required. Per Condition A actions, you are required to provide Protective Action Recommendations to Oconee and Pickens County.

Answer B Discussion

1st part is correct. Per RP/0/A/1000/002, State and County agencies shall be notified within 15 minutes of E-plan declaration, Classification upgrades, and Protective Action Recommendations.

2nd part is incorrect because for Immanent dam failure, RP/0/A/100/002, Condition A response actions are required. Per Condition A actions, you are required to provide Protective Action Recommendations to Oconee and Pickens County. It is plausible because if it were "Potential Dam Failure", Condition B actions would be required which do NOT require Protective Action Recommendations.

Answer C Discussion

1st part is incorrect because per RP/0/A/1000/002, State and County agencies shall be notified within 15 minutes of E-plan declaration, Classification upgrades, and Protective Action Recommendations. It is plausible because if the question was asking about the time requirement associated with the NRC, it would be correct.

2nd part is correct. For Immanent dam failure, RP/0/A/100/002, Condition A response actions are required. Per Condition A actions, you are required to provide Protective Action Recommendations to Oconee and Pickens County.

Answer D Discussion

1st part is incorrect because per RP/0/A/1000/002, State and County agencies shall be notified within 15 minutes of E-plan declaration, Classification upgrades, and Protective Action Recommendations. It is plausible because if the question was asking about the time requirement associated with the NRC, it would be correct.

2nd part is incorrect because for Immanent dam failure, RP/0/A/100/002, Condition A response actions are required. Per Condition A actions, you are required to provide Protective Action Recommendations to Oconee and Pickens County. It is plausible because if it were "Potential Dam Failure", Condition B actions would be required which do NOT require Protective Action Recommendations.


This question matches the KA by requiring knowledge of PAR requirements.

Basis for Hi Cog

Basis for SRO only

NUREG 1021, ES 401, Attachment 2F. Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]Some examples of SRO exam items for this topic include: Evaluating core conditions and emergency classifications based on core conditions. Administrative requirements associated with low power physics testing processes. Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities. Administrative controls associated with the installation of neutron sources. Knowledge of TS bases for reactivity controls.



Cognitive Level

Comprehension

Job Level

SRO

QuestionType

NEW

Question Source


RP.0.A.1000/002


RP/0/A/1000/001RP/0/A/1000/002EAP-SEP Obj: R12

Remarks/Status

DISCUSSED WITH CHIEF ON 7/27: GIVE CONDITIONS IN A SAE AND ASK IF PAR REQUIRED (YES/NO) AND TIME TO MAKE PAR (15 MIN).


Knowledge of emergency plan protective action recommendations. (CFR: 41.10 / 41.12 / 43.5 / 45.11)


Appendix D Scenario Outline Form ES D-1 ILT16-2 NRC Exam

Facility: Oconee Scenario No.: 1 Op-Test No.: 1 Examiners: ________________________ Operators: ________________________ SRO

________________________ ________________________ OATC

________________________ ________________________ BOP Initial Conditions:

• Reactor Power = 75% Turnover:

• SASS is in Manual for calibration • AMSAC/DSS is bypassed for calibration • PT/0/A/0610/017 is in progress to perform functional verification of SL breakers

Event No.

Malfunction No. Event Type* Event

Description

0a Override AMSAC/DSS Bypassed

0b Override SASS in Manual

0c Override Standby CC Pump Auto Start

1 N: BOP, SRO Functional Verification of SL Breakers

2 MPS290 C: BOP, SRO 1A CC Pump Trips & Standby CC Pump Fails to Auto Start

3 Override I: BOP, SRO 1NI-5 Power Failure

4 Override C: OATC, SRO (TS) 1A HPI Pump Trips & 1B HPI Pump Fails to Auto Start

5 MPI040 MPI070 I: OATC, SRO Loop 1A RC Flow Fails Low

6 MPS031 MPS031D R: OATC, SRO (TS) 80 gpm RCS Leak / Manual Power Reduction

7 MPS031D MPS150 Override

M: ALL Small Break LOCA

• 1C HPI Pump Fails to Start on ES • 1HP-24 Fails Closed

* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Required Operator Actions Form ES-D-2 ILT16-2 NRC Exam

Page 2 of 58

SCENARIO 1 EVENT SUMMARY

Event 1: When the crew takes the shift, the SRO will direct the BOP to perform PT/0/A/0610/017 Enclosure 13.11 (Functional Verification Of SL Breaker(s). The SRO will enter TS 3.8.1 Condition D by procedure and then the BOP will perform a functional check of SL1 and SL2 breakers.

Event 2: The operating Component Cooling Pump (1A CC Pump) will trip and the Standby CC Pump (1B CC Pump) will fail to auto start. The BOP will reference an Alarm Response Guide which will direct manually starting the Standby CC Pump. Since there will be no CC flow until the Standby CC Pump is started, Letdown temperature will begin to increase and 1HP-5 will automatically close on high Letdown temperature (135°F). The SRO will then enter AP/1/A/1700/032 (Loss of Letdown) to restore Letdown.

Event 3: 1NI-5 will lose DC power which will result in 1NI-5 reading 0% power. The BOP will reference an Alarm Response Guide which will direct the crew to bypass the 1A RPS channel in accordance with OP/1/A/1105/014 (Control Room Instrumentation Operation and Information).

Event 4: The operating High Pressure Injection Pump (1A HPI Pump) will trip and the Standby HPI Pump (1B HPI Pump) will fail to auto start. The SRO will enter AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal Injection) to start the Standby (1B) HPI Pump and re-establish normal HPI Makeup and RCP Seal Injection. The SRO will enter TS 3.5.2 Condition A for the inoperable HPI Pump.

Event 5: 1A Loop RCS Flow instrument will fail low which will cause ICS to re-ratio Feedwater in an attempt to restore the primary to secondary heat balance. Once alarms are received, the crew will perform Plant Transient Response to stabilize the plant. The OATC will re-ratio Feedwater to restore the heat balance. Once the plant is stable, the SRO will enter AP/1/A/1700/028 (ICS Instrument Failures) and ensure the appropriate ICS stations are in manual and BOP will perform an instrument surveillance for the failed instrumentation.

Event 6: The RCS will develop an unidentified leak of approximately 80 gpm. Once alarms are received, the SRO will enter AP/1/A/1700/002 (Excessive RCS Leakage). Since the RCS leak will be greater than Letdown Storage Tank makeup capability from 1A Bleed Holdup Tank, the crew will initiate AP/1/A/1700/029 (Rapid Unit Shutdown) and manually shutdown the unit. The SRO will enter TS 3.4.13 Conditions A & B.

Event 7: Once Reactor power has been decreased > 10% and auxiliary power has been transferred to the Startup Transformer in Event 6, the RCS leak will propagate into a Small Break LOCA. The 1C HPI Pump will fail to automatically start and 1HP-24 will fail closed when ES actuates. The SRO will Transfer to the LOSCM tab of the EOP. One RO will perform Rule 2 (Loss of SCM) and the other RO will perform EOP Enclosure 5.1 (ES Actuation).


Page 3 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 1 Page 1 of 2 Event Description: Functional Verification of SL Breakers (N: BOP, SRO) Time Position Applicant’s Actions or Behavior

SRO

BOP

PT/0/A/0610/017 Crew response: SRO directs the BOP to perform PT/0/A/0610/017 Enclosure 13.11 (Functional Verification Of SL Breakers) PT/0/A/0610/017 Encl 13.11 (Functional Verification of SL Breakers) rev 29

2.1 Perform the following: • IF Unit 1 in MODE 1, 2, 3 or 4, enter TS 3.8.1 Condition ‘D’ • IF Unit 1 in MODE 5, 6 or during movement of recently irradiated fuel

assemblies, AND Underground Power Path being credited as an emergency power source, enter TS 3.8.2 Condition ‘B’ (N/A)

• IF Unit 2 in MODE 1, 2, 3 or 4, enter TS 3.8.1 Condition ‘D’ • IF Unit 2 in MODE 5, 6 or during movement of recently irradiated fuel


• IF Unit 3 in MODE 1, 2, 3 or 4, enter TS 3.8.1 Condition ‘D’ • IF Unit 3 in MODE 5, 6 or during movement of recently irradiated fuel


2.2 IF required, perform functional check of SL1 CT5 STBY BUS 1 FEEDER as follows: 2.2.1 Verify SL1 & SL2 TRIP INTERLOCK DEFEAT CH1 switch in

“CENTRAL” position 2.2.2 Verify CT5 BUS 1 AUTO/MAN transfer switch in “MAN” 2.2.3 Close SL1 CT5 STBY BUS 1 FEEDER 2.2.4 Verify ≈ 4160V on STANDBY BUS 1 VOLTS (2AB3) 2.2.5 Open SL1 CT5 STBY BUS 1 FEEDER

This event is complete when the SRO has exited TS 3.8.1 for Unit 1, or as directed by the Lead Examiner.


Page 4 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 1 Page 2 of 2 Event Description: Functional Verification of SL Breakers (N: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

PT/0/A/0610/017 Crew response: PT/0/A/0610/017 Encl 13.11 (Functional Verification of SL Breakers)

2.3 IF required, perform functional check of SL2 CT5 STBY BUS 2 FEEDER as follows: 2.3.1 Verify SL1 & SL2 TRIP INTERLOCK DEFEAT CH2 switch in

"CENTRAL" position 2.3.2 Verify CT5 BUS 2 AUTO/MAN transfer switch in "MAN" 2.3.3 Close SL2 CT5 STBY BUS 2 FEEDER 2.3.4 Verify ≈ 4160V on STANDBY BUS 2 VOLTS (2AB3) 2.3.5 Open SL2 CT5 STBY BUS 2 FEEDER

2.4 Perform the following: • IF entered on Unit 1, evaluate exiting TS 3.8.1 Condition 'D' • IF entered on Unit 1, evaluate exiting TS 3.8.2 Condition 'B' (N/A) • IF entered on Unit 2, evaluate exiting TS 3.8.1 Condition 'D' • IF entered on Unit 2, evaluate exiting TS 3.8.2 Condition 'B' (N/A) • IF entered on Unit 3, evaluate exiting TS 3.8.1 Condition 'D' • IF entered on Unit 3, evaluate exiting TS 3.8.2 Condition 'B' (N/A)

This event is complete when the SRO has exited TS 3.8.1 for Unit 1, or as directed by the Lead Examiner.


Page 5 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 2 Page 1 of 3 Event Description: 1A CC Pump Trips & Standby CC Pump Fails to Auto Start (C: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

Plant response:

• 1SA-9/B-1 (CC CRD RETURN FLOW LOW) • 1SA-9/C-1 (CC COMP COOLING RETURN FLOW LOW) • 1SA-2/C-1 (LETDOWN TEMPERATURE HIGH) • 1HP-5 (Letdown Isolation) will close due to high letdown temperature • CC Total Flow Low • Component Cooling Pressure Low

Crew response: Refer to ARG 1SA-9/B-1 OR 1SA-9/C-1

ARG 1SA-09/B-1 ARG 1SA-09/B-1 3.1 IF ES 5 or 6 has actuated, (N/A) 3.2 IF 1CC-7 or 1CC-8 are closed, (N/A) 3.3 IF 1SA-09/C-1 is in alarm AND the Standby CC Pump did NOT start,

perform the following: 3.3.1 Verify CC Surge Tank level > 12” 3.3.2 Start Standby CC Pump

3.4 IF NO CC Pumps are operating, Go To AP/20 (Loss of Component Cooling)

ARG 1SA-09/C-1 ARG 1SA-09/C-1 3.1 IF ES 5 or 6 has actuated, (N/A) 3.2 IF 1CC-7 or 1CC-8 are closed, (N/A) 3.3 IF Standby CC Pump did NOT start, perform the following:

3.3.1 Verify CC Surge Tank level > 12” 3.3.2 Start Standby CC Pump

3.4 IF NO CC Pumps are operating, Go To AP/20 (Loss of Component Cooling)

Examiner Note: Once the 1B CC pump is started, the SRO will refer to

AP/32 (Loss of Letdown).

This event is complete when the Standby HPI Pump is placed back in Auto, or as directed by the Lead Examiner.


Page 6 of 58


SRO/BOP

AP/1/A/1700/032 Crew response: AP/1/A/1700/032 (Loss of Letdown) rev 07

4.1 Verify a total loss of letdown exists

4.2 Place 1HP-120 in HAND and reduce demand to zero

4.3 Position the standby HPI pump switch to OFF

CAUTION: RCP individual seal return valves will close if seal injection is < 22 gpm with CC flow < 575 gpm.

4.4 Throttle 1HP-31 to establish 12 - 15 gpm SEAL INLET HDR FLOW

NOTE: The running HPIP may operate below 65 gpm for up to 4 hours. HPIP time of operation below minimum flow is cumulative.

4.5 Verify HPI pump flow ≥ 65 gpm (30 gpm Recirc + ___ SI +___ MU)

RNO: Log beginning time for HPI pump flow below minimum

4.6 Initiate makeup to the LDST as required. (Using EOP Enclosure 5.5 or OP/1/A/1103/004 for batch additions)

4.7 Notify the OSM to reference OMP 1-14, Emergency Plan, and notify the STA

4.8 Verify 1HP-5 closed

4.9 Dispatch an operator to 1HP-5 to establish communication with the CR

NOTE

• TS 3.4.9 applies when PZR level > 260" (corrected value for 285"). • Conditions where it is known that letdown CANNOT be restored do not

require waiting until 260" to begin a rapid shutdown.

4.10 IAAT either of the following exist: • PZR level > 260 inches AND letdown CANNOT be established

• Plant conditions exist such that letdown will NOT be restored THEN initiate unit shutdown per AP/29 (Rapid Unit Shutdown)

4.11 IAAT PZR level ≥ 375 inches, THEN trip Rx

4.12 Determine the cause of loss of letdown: Actual LD Temperature high: GO TO Step 4.29



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BOP

AP/1/A/1700/032 Crew response:

4.29 Notify FIN24 to initiate repairs on failed equipment 4.30 IAAT letdown can be re-established, THEN perform Steps 4.30 - 4.44 4.31 Place CC System in operation 4.32 Close 1HP-6 4.33 Close 1HP-7 4.34 Open 1HP-1, 1HP-2, 1HP-3, and 1HP-4 4.35 Verify letdown temperature < 135°F

RNO: 1. Open 1HP-13 2. Close 1HP-8, 1HP-9 & 11 3. Verify NO deborating IXs in service 4. Select LETDOWN HI TEMP INTLK BYP switch to BYPASS

4.36 Open 1HP-5 4.37 Throttle open 1HP-7 to establish ≈ 20 gpm 4.38 WHEN letdown temperature < 130°F, THEN place LETDOWN HI TEMP

INTLK BYP switch in NORMAL 4.39 Open 1HP-6 4.40 Adjust 1HP-7 to control desired letdown flow 4.41 Re-establish normal makeup through 1HP-120 4.42 Re-establish normal RCP seal injection flow 4.43 Position the standby HPI pump switch to AUTO

Examiner Note: This concludes the event. It is not required to put Purification IX in service at step 4.44

4.44 Verify any purification IX in service

RNO: IF purification IX operation is desired, THEN initiate OP/1/A/1103/004 B to establish desired IX operation

4.45 EXIT this procedure

Tech Spec 3.4.1 (RCS Pressure, Temperature, and Flow DNB Limits) requires RCS pressure to stay > 2125 psig when in MODE 1 Steady State. During Letdown flow perturbations in this event it is possible that RCS pressure goes below 2125 psig, If that occurs the SRO will enter TS 3.4.1 Condition A (One or more RCS DNB parameters not within limits) which has a 2 hour completion time to restore parameter to within limits.



Page 8 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 3 Page 1 of 2 Event Description: 1NI-5 Power Failure (I: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

BOP

Plant response:

• 1SA-05/A-5 (1A RPS TROUBLE) • 1SA-05/A-6 (1NI-5 PWR FAIL) • 1SA-4/C-1 (QUADRANT POWER TILT) will alarm after couple minutes • OAC alarm 1A RPS NI FLUX DEV • OAC alarm 1NI-5 +15V POWER SUP • OAC alarm 1NI-5 -15V POWER SUP • OAC alarm NI FLUX #1 NI-5 MISMATCH

1SA-05/A-6 Crew response: The BOP will refer to Statalarm 1SA-5/A-6 (1NI-5 PWR FAIL)

ARG 1SA-05/A-6 (1NI-5 PWR FAIL) rev 19

3.1 IF all Wide Range AND Power Range Nuclear Instrument channels fail at power, AND the Reactor has NOT tripped, manually trip the Reactor

3.2 Refer to TS 3.3.1

3.3 IF the other three RPS channels are in service (NOT bypassed), bypass the channel per OP/1/A/1105/014 (Control Room Instrumentation Operation and Information)

3.4 Initiate Work Request for I&E to investigate cause and restore power

OP/1/A/1105/014 OP/1/A/1105/014 Encl 4.7 (Removal and Restoration of RPS Channels) rev 42

2.1 Verify one of the following: 2.1.1 A procedure requires RPS Channel to be placed in Trip or Bypass 2.1.2 Equipment failure requires RPS Channel to be placed in Trip or

Bypass 2.2 Identify affected RPS Channel 1A

(1A, 1B, 1C, 1D)

This event is complete when the 1A RPS Channel is placed in Manual Bypass, or as directed by the Lead Examiner.


Page 9 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 3 Page 2 of 2 Event Description: 1NI-5 Power Failure (I: BOP, SRO) Time Position Applicant’s Actions or Behavior

OP/1/A/1105/014 Crew response: OP/1/A/1105/014 Encl 4.7 (Removal and Restoration of RPS Channels)

3.1 IF affected RPS channel is NOT required per TS 3.3.1, perform one of the following: 3.1.1 IF Manual Bypass of affected RPS channel is desired, perform the

following: A. Obtain Key #314 B. Declare affected RPS Channel inoperable C. Place affected RPS Channel in MANUAL BYPASS keyswitch

in “BYP” (Cab. 2, 4, 6, or 8)

Examiner Note: Statalarm 1SA-05/A-1 will actuate when the channel is placed in BYP

3.1.2 IF Manual Trip of affected RPS channel is desired, perform the following: (NOT desired to trip channel per the NOTE)

3.2 IF affected RPS channel is required per TS 3.3.1, perform the following: (Channel is NOT required per TS 3.3.1)

3.3 IF RPS Channel removed from service due to equipment failure, perform the following: • Initiate Work Request • IF required per OMP 1-14 (Notifications), perform appropriate

notifications

Examiner Note: The SRO may announce TS 3.3.1 for tracking purposes.

This event is complete when the 1A RPS Channel is placed in Manual Bypass, or as directed by the Lead Examiner.

NOTE: Placing RPS channel in Manual Bypass is preferred to minimize risk of Reactor trip.


Page 10 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 4 Page 1 of 4 Event Description: 1A HPI Pump Trips & 1B HPI Pump Fails to Auto Start (C: OATC, SRO) (TS) Time Position Applicant’s Actions or Behavior

SRO

SRO/OATC

Plant response:

• 1SA-2/B-2 (HP RCP Seal Injection Flow High/Low) • 1SA-2/C-2 (HP Injection Pump Disch. Header Pressure High/Low) • RC Makeup Flow ≈ 0 gpm • RCP SI flow ≈ 0 gpm • 1A HPI Pump amps low = 0 amps • PZR level will begin to decrease and LDST level will begin to increase

Crew response: The SRO will enter AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal Injection)

AP/1/A/1700/014 Crew response: AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal Inj) rev 18

3.1 IAAT RCP seal injection flow is lost, AND Component Cooling is lost, THEN perform the following: A. Trip the Rx B. Stop all RCPs C. Initiate AP/25 (SSF EOP)

3.2 IAAT loss of suction to operating HPI pumps is indicated: • Motor amps low or cycling • Discharge pressure low or cycling • Abnormal LDST level trend

THEN GO TO Step 3.3 RNO: GO TO Step 4.7

4.7 Announce AP entry using PA System

4.8 Verify any HPI pump operating

RNO: 1. Close 1HP-5 2. Place 1HP-120 in HAND and closed 3. Place 1HP-31 in HAND and closed 4. Attempt to start the Standby HPIP (1B HPIP starts) 5. IF standby HPI pump started,

THEN GO TO Step 4.111

This event is complete when Seal Inlet Header Flow is ≈ 32 gpm and 1HP-31 is in Auto (Step 4.124), or as directed by the Lead Examiner.


Page 11 of 58


AP/1/A/1700/014 Crew response: AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal Inj)

4.111 Place 1HP-31 in HAND

4.112 Slowly open 1HP-31 until ≈ 8 gpm/RCP is achieved

4.113 Re-establish normal makeup through 1HP-120

4.114 Ensure proper operation of the Component Cooling System

4.115 Reduce 1HP-7 demand to 0%

4.116 Close 1HP-6

4.117 Open the following: 1HP-1 1HP-2 1HP-3 1HP-4

Booth Note: Crew may contact the WCC to direct AO to rack out the 1A HPIP breaker (wait 10 minutes and then Use Quick Strike to remove fuses and report that the 1A HPI pump has been tagged out).

4.118 Open 1HP-5

4.119 Throttle open 1HP-7 for ≈ 20 gpm letdown flow

4.120 Open 1HP-6

4.121 Adjust 1HP-7 for desired letdown flow

4.122 Open the following:1HP-228, 1HP-226,1HP-232, 1HP-230

4.123 Open 1HP-21

4.124 IAAT SEAL INLET HDR FLOW ≈ 32 gpm, THEN place 1HP-31 in AUTO

4.125 Monitor RCP seal parameters

4.126 Maintain RCP seal injection flows as required

4.127 Log thermal cycle of 1A HPI header

4.128 WHEN conditions permit, THEN EXIT this procedure Examiner Note: Crew may enter AP/16 (Abnormal RCP Operation) as a

result of high seal return temperatures. Steps are on the next page.



Page 12 of 58


AP/1/A/1700/016 Crew response: AP/1/A/1700/016 (Abnormal RCP Operation) rev 34

4.1 IAAT any RCP meets immediate trip criteria… (does not) RNO: GO TO Step 4.12

NOTE If affected RCP has a seal failure, and immediate trip criteria not met, then continue to Section 4A (Seal Failure) to ensure Seal Failure guidance steps are read. Section 4A (Seal Failure) contains steps to quickly secure affected RCP should the need arise.

4.12 IAAT either of the following apply: __ Any RCP approaching immediate trip criteria of Encl 5.1 __ There is an immediate need to stop a RCP at this time THEN perform Steps 4.13 - 4.15

RNO: GO TO Step 4.16

4.16 Announce AP entry using the PA system

4.17 Notify OSM to request evaluation by RCP Component Engineer

4.18 IAAT the failure is identified, THEN GO TO the applicable section per the following table:

AP/16 (Abnormal RCP Operation) Section 4D 1. IAAT any RCP meets immediate trip criteria… (does not)

RNO: GO TO Step 12

12. Monitor RCP parameters for abnormalities (Turn on Code “RCP”).

13. Open 1HP-20 and 1HP-21

14. Open1HP-228, 1HP-226, 1HP-232, and 1HP-230

15. Verify either of the following conditions apply..… (not met) RNO: GO TO Step 17



Page 13 of 58


AP/1/A/1700/016 Crew response: AP/1/A/1700/016 (Abnormal RCP Operation) rev 34

17. Verify RCP seal return low flow alarms off

RNO: Request that RCP Component Engineer provide the following: • Immediate evaluation • Additional monitoring requirements

Examiner Cue: If candidate attempts to monitor the Loose part Monitor, indicate that the noise is normal.

Examiner Note: Due to sequence of events, SRO may not review the TS during the scenario. Follow-up questions may be required to ensure knowledge of this competency.

TS 3.5.2 HIGH PRESSURE INJECTION Condition A (72 hours) Restore HPI pump to OPERABLE status



Page 14 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 5 Page 1 of 3 Event Description: Loop 1A RC Flow Fails Low (I: OATC, SRO) Time Position Applicant’s Actions or Behavior

OATC/BOP

Plant response:

• 1SA-02/A-3 (RC Loop A Flow Low) • 1SA-02/A-5 (RC Total Flow Low) • 1SA-02/A-11 (ICS Runback) • 1SA-02/A-12 (ICS Tracking) • 1SA-05/B-5 (1B RPS Trouble)

Crew response:

• When the Statalarms are received, the candidates should utilize the “Plant Transient Response” (PTR) process to stabilize the plant, which should include: o Placing ICS to HAND (Feedwater Masters and Diamond) o Inserting Control Rods as needed to control RCS pressure (Performed

by the BOP)

o Decreasing or re-ratio Feedwater to control Reactor power and delta Tcold (Performed by the OATC)

SRO may direct the OATC to perform the actions of the ARG for 1SA-02/A-3 Statalarm, but probably will move directly from PTR to AP/28.

ARG for 1SA-02/A-3 3.1. Ensure reactor power is reduced below the flux to flow minus imbalance

trip ratio

3.2. Ensure Feedwater demand re-ratios properly

Examiner Note: This failure will cause FDW flow to re-ratio. If the crew immediately recognizes it is an instrument failure and not an actual loss of flow, they may choose to adjust FDW and therefore prevent high delta Tc from being established causing high Quadrant Power Tilt values. The SRO should direct a band for delta Tc of 0 ± 2°F as he establishes bands for parameters being manually controlled.

SRO directs performance of AP/1/A/1700/028, ICS Instrument Failures (see next page)

This event is complete when Step 5 of Section 4E is complete, or as directed by the Lead Examiner.


Page 15 of 58


AP/1/A/1700/028 Crew response: AP/1/A/1700/028 (ICS Instrument Failures) rev 20

4.1. Provide control bands as required

4.2. Initiate notification of the following: • OSM to reference the following:

o OMP 1-14 (Notifications) o Emergency Plan

• STA

4.3. Verify a power transient ≥ 5% has occurred

RNO: GO TO Step 4.5

Examiner Note: If power change was ≥ 5%, then step 4.4 will be performed. It depends on the speed of crew response to the failure.

4.4. Notify Rx Engineering and discuss the need for a maneuvering plan

Booth Cue: We will develop a maneuvering plan. 4.5. Use the following, as necessary, to determine the applicable section

from table in Step 4.6: • OAC alarm video • OAC display points • Control Board indications • SPOC assistance, as needed

4.6. GO TO the applicable section per the following table:

√ Section Failure 4A RCS Temperature

4B Turbine Header Pressure

4C Controlling NI

4D Feedwater Loop Flow

4E RCS Flow



Page 16 of 58


AP/1/A/1700/028 Plant response: AP/1/A/1700/028 Section 4E (RCS Flow)

NOTE The following will occur when an ICS RCS flow loop signal fails: • ICS RUNBACK • Controlling Tave swaps to RCS loop with higher flow • Delta Tc station re-ratios loop feedwater flows

1. Ensure the following in HAND: • 1A FDW MASTER • 1B FDW MASTER

2. Ensure DIAMOND in MANUAL

3. Notify SPOC to perform the following: • Select a valid RCS flow input to ICS per AM/1/A/0326/020 (Control of

Unit 1 Star Module Signal Selection Function) • Investigate and repair the failed RCS flow instrumentation

4. PERFORM an instrumentation surveillance using applicable table in Encl 5.2 (ICS Instrument Surveillances) for the failed instrument

The RO will refer to Table 3 and determine that the surveillances can NOT be met as written with the failed instrument and therefore the SRO would ensure that a surveillance evaluation is initiated. (Surveillance is required in Mode 1, Steady State Operation. Steady State is defined as operation within a 4% (e.g. 88% - 92% RTP) power band for ≥ 4 hours)

5. Verify instrumentation surveillance in Encl 5.2 (ICS Instrument Surveillances) was performed satisfactorily as written

Inform the SRO that the surveillance cannot be met as written

Determine that a Surveillance Evaluation should be initiated

Booth Cue: If crew requests Unit 2 to perform the surveillance evaluation, respond that "Unit 2 will perform the surveillance evaluation".

6. WHEN notified by SPOC that a valid RCS flow input has been restored to ICS, THEN GO TO OP/1/A/1102/004 A Encl (Placing ICS Stations To Auto)

Examiner Note: ICS will remain in Manual for the remainder of the scenario.



Page 17 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 6 Page 1 of 11 Event Description: 80 GPM RCS Leak (Requires Manual Power Reduction) (R: OATC, SRO) (TS) Time Position Applicant’s Actions or Behavior

SRO

SRO

Plant response:

• OAC RB Normal Sump Temp HI HI • 1SA-9/A-6 (RB NORMAL SUMP HIGH/LOW) • 1SA-8/B-9 (RM Process Monitor Radiation HIGH) • 1SA-8/E-9 (RM Reactor Building Normal Sump Isolate) • PZR and LDST level deceasing • RC makeup flow increasing • RB normal sump level increasing

Crew response:

• The SRO may refer to TS 3.4.13 (RCS Operational Leakage) and determine that Condition A, Reduce leakage to within limits within 4 hours and Condition B, Be in MODE 3 in 12 hours are in effect. This is for an unidentified leak > 1 gpm.

TS 3.4.13 RCS OPERATIONAL LEAKAGE Condition A (4 hours) Reduce leakage to within limits

Condition B (12 hours) Be in MODE 3 AND (36 hours) Be in MODE 5

• The SRO will enter AP/1/A/1700/002 (Excessive RCS Leakage) (next page)

This event is complete when Reactor power is decreased > 10% and auxiliaries have been transferred, or as directed by the Lead Examiner.


Page 18 of 58


AP/1/A/1700/002 AP/1/A/1700/002 (Excessive RCS Leakage) rev 15

3.1 Verify HPI operating

3.2 IAAT RC makeup flow is > 100 gpm, AND Pzr level is decreasing, THEN close 1HP-5

3.3 IAAT all the following exist: (does not apply) • HPI flow is > NORMAL MAKEUP CAPABILITY (≈ 160 gpm) with

letdown isolated • Pzr level decreasing • SG Tube Leakage NOT indicated • LPI DHR NOT in service

THEN perform the following: A. Ensure Rx is tripped B. Initiate Unit 1 EOP

4.1 Initiate Pzr and LDST level makeup using Unit 1 EOP Encl 5.5, as necessary (page 47)

Booth Cue: If requested by the crew to close 1CS-48, wait two minutes and then use Manual Valves to close 1CS-48. Then call back to report that 1CS-48 is closed.


4.3 IAAT LPI DHR in service, AND RCS leakage > LDST makeup capability (≈ 50 gpm) THEN GO TO AP/26

4.4 Initiate the following notifications:

__ OSM to reference the following: • RP/1000/001 (Emergency Classification) • OMP 1-14 (Notifications) • Encl 5.9 (Oversight Guidelines)

__ STA and RP



Page 19 of 58

- Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 6 Page 3 of 11 Event Description: 80 GPM RCS Leak (Requires Manual Power Reduction) (R: OATC, SRO) (TS) Time Position Applicant’s Actions or Behavior

AP/1/A/1700/002 Crew response: AP/1/A/1700/002 (Excessive RCS Leakage)

4.5 Monitor the following trends to determine leak area (AB or RB) and trend for degradation: • T6 AP02 • T6 WASTE • RIAs

Examiner/Booth Note: AP/1/A/1700/018 entry conditions will also be met due to RB RIA alarms. If the crew asks, Unit 2 will perform AP/18 (Abnormal Release of Radioactivity) actions.

4.6 Verify specific leak location is identified

RNO: Notify WCC SRO to initiate Encl 5.2 (Primary Leak Check) and of the leak area (AB or RB), if known

Note: Crew should determine that the leak is in the Reactor Building due to RB RIAs increasing, RBNS rate increasing, and NO RCP seal failure indications.

4.7 Initiate Encl 5.1 (Leak Rate Determination) (See page 24 for actions of Encl 5.1)

4.8 WHEN leak area/failure is identified, THEN GO TO applicable step that best fits leak area/failure

√ Area/ Failure

Symptoms Step

Rx Bldg ↑ RB RIAs

↑ RBNS rate

NO RCP seal failure indications

4.53



Page 20 of 58



4.53 IAAT in MODE 1 AND leak is > LDST makeup capability from 1A BHUT, THEN initiate a shutdown using AP/29 (Rapid Unit Shutdown)

Examiner Note: The 80 gpm leak will eventually be greater than LDST makeup capability from 1A BHUT so the SRO should initiate AP/29.

4.54 IAAT leak rate is ≥ 10 gpm, THEN discontinue pumping RBNS

4.55 IAAT either of the following conditions exist: • RCS pressure ≤ 50 psig and RCS leakage ≥ 10 gpm • RCS pressure ≥ 50 psig and RCS leakage ≥ 1 gpm THEN perform Steps 4.56 - 4.59

4.56 Verify the RB is occupied (It is not) RNO: GO TO Step 4.58

4. 58 Verify LPI DHR in service (It is not) RNO: GO TO Step 4.60

4.60 Verify RB pressure ≥ 3 psig

RNO: Maximize RB Cooling by performing the following: • Ensure all available RBCUs operating in HIGH • Open 1LPSW-18 • Open 1LPSW-21 • Open 1LPSW-24



Page 21 of 58



4.61 IAAT RB is accessible to locate the leak, THEN GO TO Step 4.62

RNO: GO TO Step 77

4.77 Verify 1HP-5 is closed

RNO: GO TO Step 4.82 (next page)

4.78 Place standby CC pump switch in OFF

NOTE Statalarm 1SA-9C-1 (CC COMP COOLING RETURN FLOW LOW) will alarm when the letdown coolers are isolated

4.79 Close the following:

__ 1CC-1/1HP-1

__ 1CC-2/1HP-2

4.80 Verify leak is isolated (it will not be) RNO: 1. Perform the following to shutdown and depressurize the RCS

A. Initiate shutdown by one of the following as necessary: __ AP/29 (Rapid Unit Shutdown) (page 25) __ OP/1/A/1102/004 (Operation at Power)

B. Initiate OP/1/A/1102/010 (Controlling Procedure For Unit Shutdown) to continue shutdown and depressurization

2. WHEN conditions permit, THEN EXIT this procedure

Booth Cue: If the crew elects to use the OP to shutdown the unit, call as the SM and inform the crew that a more rapid shutdown is desired.



Page 22 of 58



4.82 Isolate 1A Letdown Cooler by closing the following: __ 1CC-1/1HP-1 __ 1HP-3

4.83 Verify leak is isolated

RNO: 1. Restore 1A Letdown Cooler by performing the following: A. Open 1HP-3 B. Open 1CC-1/1HP-1 2. GO TO Step 4.85

4.85 Isolate 1B Letdown Cooler by closing the following: __ 1CC-2/1HP-2 __ 1HP-4

4.83 Verify leak is isolated

RNO: 1. Restore 1B Letdown Cooler by performing the following: A. Open 1HP-4 B. Open 1CC-2/1HP-2 2. GO TO Step 4.88



Page 23 of 58



NOTE • It is desirable to isolate the LD line (to check for a leak), if the leak in the

RB is so large that a RB entry (to locate leak) will not be made.

• The following steps to isolate letdown are performed at station management discretion dependent on the rate of leak. Isolating letdown will result in a complicated shutdown. Guidance is provided to restore letdown if leak is not isolated.

4.88 IAAT Station Management desires to isolate LD to determine if leak is downstream of 1HP-3 and 1HP-4, THEN GO TO Step 4.89

Booth Cue: If contacted as Station Management to determine if it is desired to isolate Letdown, state that “It is not desired to isolate Letdown to determine if the leak is downstream of 1HP-3 and 1HP-4”.

RNO: 1. Perform the following to shutdown and depressurize the RCS: A. Initiate shutdown by one of the following, as necessary: __ AP/29 (Rapid Unit Shutdown) (page 25) __ OP/1/A/1102/004 (Operation At Power) B. Initiate OP/1/A/1102/010 (Controlling Procedure For Unit

Shutdown) to continue shutdown and depressurization 2. WHEN conditions permit, THEN EXIT this procedure

Booth Cue: If the crew elects to use the OP to shutdown the unit, call as the SM and inform the crew that a more rapid shutdown is desired.



Page 24 of 58


AP/1/A/1700/002 Crew response: AP/1/A/1700/002 Enclosure 5.1 (Leak Rate Determination)

1. Stabilize RCS Temperature

2. Notify WCC to secure all primary draining/RB washdown evolutions if applicable

NOTE Depending on leak location, leakage may NOT be detected by all the formulas. One or more of the following methods may be necessary to determine RCS leak rate.

3. Calculate leak rate using the following, as required:

Examiner Note: There are several other methods to calculate leakage rate. While one of the two below will most likely be used, depending on plant conditions they may not be the only correct methods available.

Method #1: Calculation of RCS Volume Loss:

Leak Rate = ______ + ______ - ______ - ______ = ______ MU SI LD TSR

Where: MU = makeup Flow SI = Seal Inlet Hdr Flow LD = Letdown Flow TSR = Total Seal Return Flow

Method #2: LDST Level Change:

Leak Rate = (LDST level change) x (31 gal/inch) + BTP Flowrate (gpm) (minutes)

Leak Rate = ( inches) x 31 gal/inch + ______ gpm = _________ gpm _____minutes

4. Notify SM and SRO of calculated leak rate



Page 25 of 58


AP/1/A/1700/029 Crew response: AP/1/A/1700/029 (Rapid Unit Shutdown) rev 13

4.1 Initiate Encl 5.1 (Support Actions During Rapid Unit Shutdown) (page 27) 4.2 Announce AP entry using the PA system.

4.3 IAAT both of the following apply: __ It is desired to stop power decrease __ CTP > 18%

THEN perform Steps 4.4 – 4.7

RNO: GO TO Step 4.8

4.4 Verify ICS in AUTO

RNO: 1. Stop manual power reduction 2. GO TO Step 4.6

4.6 Initiate OP/1/A/1102/004 (Operation at Power) power reduction enclosure

4.7 WHEN conditions permit, THEN perform one of the following: __ Depress MAXIMUM RUNBACK __ GO TO appropriate operating procedure for continued operation to

resume power reduction

4.8 Verify ICS in AUTO (ICS is NOT in Auto) RNO: 1. Initiate manual power reduction to desired power level

2. GO TO Step 4.10

Note: OATC reduces power by first reducing feedwater and then inserting control rods as necessary.

4.10 Verify both Main FDW pumps running:


NOTE • 1B Main FDW Pump is the preferred pump to be shutdown first. • To lower 1B Main FDW Pump suction flow, bias is adjusted counter-

clockwise. • To lower 1A Main FDW Pump suction flow, bias is adjusted clockwise.

4.11 Adjust bias for first Main FDW pump desired to be shutdown (1B) until its suction flow is ≈ 1 X 106 lbm/hr less than remaining Main FDW pump suction flow



Page 26 of 58



4.12 WHEN core thermal power is < 65% FP, THEN continue

4.13 IAAT both Main FDW pumps running, AND both of the following exist: __ 1B Main FDW Pump is first pump to be shut down __ Any of the following alarms actuate and remain in alarm:

• FWP B FLOW MINIMUM (1SA-16/A-3) • FWP B FLOW BELOW MIN (1SA-16/A-4)

THEN trip 1B Main FDW Pump

4.14 IAAT both Main FDW pumps running, AND both of the following exists: __ 1A Main FDW pump is the first pump to be shut down __ Any of the following alarms actuate and remain in alarm:

• FWP A FLOW MINIMUM (1SA-16/A-1) • FWP A FLOW BELOW MIN (1SA-16/A-2)

THEN trip 1A Main FDW Pump 4.15 Verify Turbine-Generator shutdown is required

4.16 Start the TURBINE TURNING GEAR OIL PUMP

4.17 Start 1A through 1E TURBINE BRNG OIL LIFT PUMPS

4.18 Start the TURBINE MOTOR SUCTION PUMP

4.19 IAAT both of the following apply: __ ICS in automatic __ NI power is ≤ 18%

THEN deselect MAXIMUM RUNBACK (does NOT apply) 4.20 Verify Turbine-Generator shutdown is required (it is required) 4.21 WHEN NI power ≤18% THEN depress turbine TRIP pushbutton



Page 27 of 58


AP/1/A/1700/029 Crew response: AP/1/A/1700/029 Enclosure 5.1

1. Notify WCC SRO to initiate Encl 5.2 (WCC SRO Support During Rapid Unit Shutdown)

2. Start the following pumps: • 1A FDWP SEAL INJECTION PUMP • 1A FDWP AUXILIARY OIL PUMP • 1B FDWP AUXILIARY OIL PUMP • 1B FDWP SEAL INJECTION PUMP

3. WHEN CTP is ≤ 80%, THEN continue

4. Stop 1E1 HTR DRN PUMP

5. Place 1HD-254 switch to OPEN



8. Verify Turbine-Generator shutdown is required (It is required) 9. Place the following transfer switches to MAN

• 1TA AUTO/MAN • 1TB AUTO/MAN

10. Close 1TA SU 6.9 KV FDR

11. Verify 1TA NORMAL 6.9 KV FDR opens

12. Close 1TB SU 6.9 KV FDR

13. Verify 1TB NORMAL 6.9 KV FDR opens

14 Place the following transfer switches to MAN • MFB1 AUTO/MAN • MFB2 AUTO/MAN

15. Close E11 MFB1 STARTUP FDR

16. Verify N11 MFB1 NORMAL FDR opens





Page 28 of 58

Op-Test No.: ILT16-2 Scenario No.: 1 Event No.: 7 Page 1 of 10 Event Description: Small Break LOCA (M: ALL) Time Position Applicant’s Actions or Behavior

Plant response:

• 1SA-2/D-3 (RC PRESS HI/LOW) • RCS pressure and PZR level decreasing • ES Channels 1-6 actuate • RCS subcooling margin will indicate 0°F shortly after the Rx trips • Reactor Building Emergency Sump level increasing

Crew response: The SRO will direct the OATC to perform IMAs.

3.1 Depress REACTOR TRIP pushbutton 3.2 Verify reactor power < 5% FP and decreasing 3.3 Depress the turbine TRIP pushbutton 3.4 Verify all turbine stop valves closed 3.5 Verify RCP seal injection available

The SRO will direct the BOP to perform a Symptoms Check

Power Range NIs NOT < 5% Power Range NIs NOT decreasing

Rule 1, ATWS/Unanticipated Nuclear Power Production

Any SCM < 0°F Rule 2, Loss Of SCM

Loss of Main and Emergency FDW (including unsuccessful manual initiation of EFDW)

Rule 3, Loss of Main or Emerg FDW Rule 4, Initiation of HPI Forced Cooling (Inability to feed SGs and > 2300 psig, NDT limit reached, or PZR level > 375")

Uncontrolled Main steam line(s) pressure decrease

Rule 5, Main Steam Line Break

CSAE Offgas alarms Process monitor alarms (RIA-40, 59,60), Area monitor alarms (RIA-16/17)

None (SGTR Tab is entered when identified SG Tube Leakage > 25 gpm)

SRO will transfer from the Subsequent Actions Tab to the LOSCM tab (page 29) from the Parallel Actions Page (page 56) to direct crew activities Once the RCS saturates, one of the ROs will perform Rule 2 (page 33) The RO not performing Rule 2 will begin performing Enclosure 5.1 due to ES actuation (page 38)

This event is complete when the SRO transfers to the LOCA CD tab, or as directed by the Lead Examiner.


Page 29 of 58


LOSCM Tab Crew Response: LOSCM tab rev 01

1. Ensure Rule 2 (Loss of SCM) is in progress or complete

2. Verify LOSCM caused by excessive heat transfer

RNO: GO TO Step 4

4. IAAT either exists: LPI FLOW TRAIN A plus LPI FLOW TRAIN B ≥ 3400 gpm Only one LPI header in operation with header flow ≥ 2900 gpm

THEN GO TO LOCA CD tab

5. Verify SSF activated per AP/25 with SSF RC Makeup required

RNO: GO TO Step 7

7. Verify all exist: __ NO RCPs operating __ HPI flow in both HPI headers __ Adequate total HPI flow per Figure 1 (Total Required HPI Flow)



Page 30 of 58


LOSCM Tab Crew Response: LOSCM tab (continued)

8. GO TO Step 104

104. Open 1AS-40 while closing 1MS-47

105. Verify HPI forced cooling in progress

RNO: Close 1RC-4

106. Close 1GWD-17, 1HP-1, 1HP-2, and 1RC-3

107. Verify either: • Core superheated • Rx vessel head level at 0″

RNO: GO TO Step 109

109. IAAT BWST level is ≤ 19', THEN initiate Encl 5.12 (ECCS Suction Swap to RBES)

CAUTION If TDEFDWP is being used for SG feed, reducing SG pressure below ≈ 250 psig can result in reduced pumping capability

110. Maintain SG pressure < RCS pressure utilizing either: __ TBVs __ ADVs

111. Verify any SG available for feeding/steaming

112. Initiate Encl 5.16 (SG Tube-to-Shell ∆T Control)



Page 31 of 58


LOSCM Tab Crew response: LOSCM tab (continued)

113. Verify indications of SGTR exist

RNO: GO TO Step 116

116. Verify HPI forced cooling in progress

RNO: GO TO Step 118

118. Verify CETCs trend decreasing

119. Verify primary to secondary heat transfer is excessive

RNO: GO TO Step 121

121. Verify indications of SGTR ≥ 25 gpm

RNO: GO TO Step 123

123. Verify required RCS makeup flow within normal makeup capability

RNO: GO TO LOCA CD tab (page 32)



Page 32 of 58


LOCA Cooldown Tab Crew Response: LOCA Cooldown Tab rev 0

1. IAAT BWST level is ≤ 19’, THEN initiate Encl 5.12 (ECCS Suction Swap to RBES)

2. Verify ES actuated

3. GO TO Step 7

7. Perform the following: • Ensure all RBCUs in low speed • Open 1LPSW-18 • Open 1LPSW-21 • Open 1LPSW-24

8. Initiate Encl 5.35 (Containment Isolation)

9. Start all RB Aux fans

10. IAAT either of the following exists: • LPI FLOW TRAIN A plus LPI FLOW TRAIN B ≥ 3400gpm • Only one LPI header in operation with header flow ≥ 2900 gpm

THEN GO TO Step 11

RNO: GO TO Step 43

43 Initiate Encl 5.36 (Equipment Alignment For Plant Shutdown)

44. IAAT all the following exist: __ All SCMs > 0°F __ RCS pressure > LPI shutoff head __ Required HPI within normal makeup capability THEN GO TO Step 45

RNO: GO TO Step 48

45. Verify primary to secondary heat transfer exists



Page 33 of 58


CT-1

CT-3

Rule 2 Crew Response: Rule 2 (Loss of SCM)

1. IAAT all the following exist: • Any SCM ≤ 0°F • Rx power ≤ 1% • ≤ 2 minutes elapsed since loss of SCM

THEN perform steps 2 & 3

2. Stop all RCPs 3. Notify CRS of RCP status

4. Verify Blackout exists

RNO: GO TO Step 6

6. Open 1HP-24 & 25

7. Start all available HPI Pumps

Examiner Note: The 1C HPI pump will fail to start on ES signal but will start manually.

8. GO TO step 13

13. Open 1HP-26 & 27

14. Verify at least two HPI pumps are operating using two diverse indications

15. IAAT ≥ 2 HPI pumps operating and HPI flow in any header is in Unacceptable Region of Fig. 1, THEN perform Steps 16-21

RNO: GO TO Step 17

17. IAAT flow limits are exceeded THEN perform Steps 18 - 20

RNO: GO TO Step 21

18. Place Diverse HPI in BYPASS

19. Perform both: • Place ES CH 1 in MANUAL • Place ES CH 2 in MANUAL

20. Throttle HPI to maximize flow ≤ flow limit



Page 34 of 58


Rule 2 Crew Response: Rule 2 ( Loss of SCM) (Continued)

21. Notify CRS of HPI status

22. Verify RCS pressure >550 psig

23. IAAT either exists: • LPI FLOW TRAIN A plus LPI FLOW TRAIN B ≥ 3400 gpm • Only one LPI header in operation with header flow ≥ 2900 gpm THEN GO TO Step 24

RNO: GO TO Step 35

35. IAAT TBVs are unavailable, THEN A. Dispatch two operators to perform Encl 5.24 (Operation of ADVs) B. Notify CRS the ADVs are being aligned for use

36. Select OFF for both Digital Channels on AFIS HEADER A

37. Select OFF for both Digital Channels on AFIS HEADER B

38. Verify any EFDW pump operating

RNO: Place 1FDW 315 and 1FDW-316 in MANUAL and close

39. Start MD EFDW pumps on all intact SGs: • 1A MD EFDWP (1A MD EFDWP will not start) • 1B MD EFDWP

RNO: Start 1 TD EFDW PUMP


41. Verify both SGs intact

42. Establish 300 gpm EFDW flow to each SG

43. Verify both MD EFDWPs operating (the 1A MD EFDWP did not start) RNO: 1. IF 1TD EFDW PUMP is operating, OR NO Main FDW pumps

operating, GO TO Step 45

2. GO TO Step 47

45. Trip both Main FDW pumps



Page 35 of 58


Rule 2 Crew Response: Rule 2 ( Loss of SCM) (Continued)

46. Place FDW block valve switches in CLOSE: • 1FDW-33 • 1FDW-31 • 1FDW-42 • 1FDW-40

47. Utilize Rule 7 (SG Feed Control) to feed all intact SGs to the appropriate SG Level Control Point using available feed sources; EFDW/Main FDW

48. IAAT SG Level Control Point is reached, THEN maintain SG Level Control Point by feeding and steaming as necessary

49. Notify CRS of SG feed status

CAUTION If 1 TD EFDW PUMP is being used for SG feed and Unit 1 is supplying the Auxiliary Steam header, reducing SG pressure below ≈ 250 psig can result in reduced pumping capability.

50. IAAT SG pressure is > RCS pressure, THEN reduce SG pressure <

RCS pressure using either: • TBVs • Dispatch two operators to perform Encl 5.24 (operation of the ADVs)

51. Verify any Main FDW pump operating

RNO: GO TO Step 58

58. Ensure Rule 3 (Loss of Main or Emergency FDW) is in progress or complete (page 36)

59. WHEN directed by CRS, THEN EXIT



Page 36 of 58


Rule 3 Crew Response: Rule 3 (Loss of Main or Emergency Feedwater)

1. Verify loss of MFDW and /or EFDW was due to any of the following: __ Turbine Building Flooding __ Actions taken to increase SG level due to Turbine Building Flooding

RNO: GO TO Step 3

3. IAAT NO SGs can be fed with FDW (Main/CBP/Emergency/PSW), AND any of the following exist: • RCS pressure reaches 2300 psig or NDT limit • PZR level reaches 375” (340” acc)

THEN PERFORM Rule 4 (HPI Forced Cooling)

4. Start operable EFDW pumps, as required, to feed all intact SGs


6. GO TO Step 38

38. IAAT an EFDW valve CANNOT control in AUTO, OR manual operation of EFDW valve is desired to control flow/level, THEN perform Steps 39 - 43

RNO: GO TO Step 44

44. Verify any SCM ≤ 0°F

RNO: IF overcooling OR exceeding limits in Rule 7, THEN throttle EFDW as necessary

45. IAAT Unit 1 EFDW is in operation, THEN initiate Encl 5.9 (Extended EFDW Operation) (page 37)




Page 37 of 58


EOP Encl 5.9 Crew Response: EOP Encl 5.9 (Extended EFDW Operation)

1. Monitor EFDW parameters on EFW graphic display

2. IAAT UST level is < 4', THEN GO TO Step 120

3. IAAT feeding both SGs with one MD EFDWP is desired, THEN perform steps 4 - 7

RNO: GO TO Step 8 8. Perform as required to maintain UST level > 7.5'

• Makeup with demin water • Place CST pumps in AUTO

9. IAAT all exist: __ Rapid cooldown NOT in progress __ MD EFDWP operating for each available SG __ EFDW flow in each header < 600 gpm

THEN place 1 TD EFDW PUMP switch in PULL TO LOCK

10. Verify 1 TD EFDW PUMP operating

RNO: GO TO Step 12

11. Start TD EFDWP BEARING Oil Cooling Pump

NOTE • Loss of the condensate system for ≥ 25 minutes results in cooling down to

LPI using the ADVs. If NO HWPs are operating, continuing this enclosure to restore the condensate system is a priority unless the CR SRO deems EOP activities higher priority. The 25 minute criterion is satisfied when a HWP is started and 1C-10 is 10% open.

• If the condensate system is operating, the remaining guidance establishes FDW recirc, monitors and maintains UST, and transfers EFDW suction to the hotwell if required.

12. Notify CR SRO to set priority based on the NOTE above and EOP activities Note: The SRO should determine that continuing in Encl 5.9 is not a

priority at this time and direct the RO from the LOSCM Tab of the EOP. (page 29)



Page 38 of 58

EOP Enclosure 5.1 (ES Actuation)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. __ Determine all ES channels that should have actuated based on RCS pressure and RB pressure:

2. __ Verify all ES channels associated with actuation setpoints have actuated.

NOTE Voter OVERRIDE extinguishes the TRIPPED light on the associated channels that have auto actuated. Pressing TRIP on channels previously actuated will reposition components that may have been throttled or secured by this Enclosure.

__ Depress TRIP on affected ES logic channels that have NOT previously been actuated.

3. __ IAAT additional ES actuation setpoints are exceeded, THEN perform Steps 1 - 2.

4. __ Place Diverse HPI in BYPASS. __ Place Diverse HPI in OVERRIDE.

5. Perform both: __ Place ES CH 1 in MANUAL. __ Place ES CH 2 in MANUAL.

NOTE • Voter OVERRIDE affects all channels of the

affected ODD and/or EVEN channels. • In OVERRIDE, all components on the

affected ODD and/or EVEN channels can be manually operated from the component switch.

1. __ IF ES CH 1 fails to go to MANUAL, THEN place ODD voter in OVERRIDE.

2. __ IF ES CH 2 fails to go to MANUAL, THEN place EVEN voter in OVERRIDE.

Actuation Setpoint

(psig)

Associated ES Channel

1600 (RCS) 1 & 2 550 (RCS) 3 & 4 3 (RB) 1, 2, 3, 4, 5, & 6 10 (RB) 7 & 8


Page 39 of 58


6. __ IAAT all exist:

__ Voter associated with ES channel is in OVERRIDE

__ An ES channel is manually actuated __Components on that channel require

manipulation THEN depress RESET on the required channel.

7. __ Verify Rule 2 in progress or complete. __ GOTO Step 74.

8. __ Verify any RCP operating. __ GOTO Step 10.

9. Open: __ 1HP-20 __ 1HP-21

10. __ IAAT any RCP is operating, AND ES Channels 5 and 6 actuate, THEN perform Steps 11 - 15.

__ GOTO Step 16.

11. Perform all: __ Place ES CH 5 in MANUAL. __ Place ES CH 6 in MANUAL.






12. __ Verify any RCP is operating __ GO TO Step 16

13. Open: __ 1CC-7 __ 1CC-8 __ 1LPSW-15 __ 1LPSW-6

14. __ Ensure only one CC pump operating.

15. __ Ensure Standby CC pump in AUTO.


Page 40 of 58


16. __ IAAT ES Channels 3 & 4 are actuated,

THEN GO TO Step 17. __ GO TO Step 54.

17. __ Place Diverse LPI in BYPASS. __ Place Diverse LPI in OVERRIDE.







CAUTION

LPI pump damage may occur if operated in excess of 30 minutes against a shutoff head. {6}

19. __ IAAT any LPI pump is operating against a shutoff head, THEN at the CR SRO's discretion, stop affected LPI pumps. {6, 22}

20. __ IAAT RCS pressure is < LPI pump shutoff head, THEN perform Steps 21 - 22.

__ GOTO Step 23.

21. Perform the following: __ Open 1LP-17. __ Start 1A LPI PUMP.

1. __ Stop 1A LPI PUMP. 2. __ Close 1LP-17.

22. Perform the following: __ Open 1LP-18. __ Start 1B LPI PUMP.

1. __ Stop 1B LPI PUMP. 2. __ Close 1LP-18.


Page 41 of 58


23. __ IAAT 1A and 1B LPI PUMPs are

off / tripped, AND all exist: __ RCS pressure < LPI pump shutoff

head __ 1LP-19 closed __ 1LP-20 closed THEN perform Steps 24 - 25.

__ GO TO Step 26.

24. Open: __ 1LP-9 __ 1LP-10 __ 1LP-6 __ 1LP-7 __ 1LP-17 __ 1LP-18 __ 1LP-21 __ 1LP-22

25. __ Start 1C LPI PUMP.

26. __ IAAT 1A LPI PUMP fails while operating, AND 1B LPI PUMP is operating, THEN close 1LP-17.

27. __ IAAT 1B LPI PUMP fails while operating, AND 1A LPI PUMP is operating, THEN close 1LP-18.

28. Start: __ A OUTSIDE AIR BOOSTER FAN __ B OUTSIDE AIR BOOSTER FAN

CT- 2

29. Notify Unit 3 to start: __ 3A OUTSIDE AIR BOOSTER FAN __ 3B OUTSIDE AIR BOOSTER FAN


Page 42 of 58


30. Verify open:

__ 1CF-1 __ 1CF-2

__ IF CR SRO desires 1CF-1 and 1CF-2 open, THEN open: __ 1CF-1 __ 1CF-2

31. __ Verify 1HP-410 closed. 1. __ Place 1HP-120 in HAND. 2. __ Close 1HP-120.

32. __ Secure makeup to the LDST.

33. __ Verify all ES channel 1 - 4 components are in the ES position.

1. __ IF 1HP-3 fails to close, THEN close 1HP-1.


3. __ IF 1HP-20 fails to close, AND NO RCPs operating, THEN close:

__ 1HP-228 __ 1HP-226 __ 1HP-232 __ 1HP-230 4. __ Notify SRO to evaluate components NOT

in ES position and initiate action to place in ES position if desired.

34. __ Verify Unit 2 turbine tripped. __ GOTO Step 37.

35. __ Close 2LPSW-139.

36. __ Verify total LPSW flow to Unit 2 LPI coolers ≤ 6000 gpm.

__ Reduce LPSW to Unit 2 LPI coolers to obtain total LPSW flow ≤ 6000 gpm.

37. __ Close 1LPSW-139.

38. Place in FAIL OPEN: __ 1LPSW-251 FAIL SWITCH __ 1LPSW-252 FAIL SWITCH

39. __ Start all available LPSW pumps.


Page 43 of 58


40. Verify either:

__ Three LPSW pumps operating __ Two LPSW pumps operating when

Tech Specs only requires two operable

__ GOTO Step 42.

41. Open: __ 1LPSW-4 __ 1LPSW-5

__ IF both are closed: __ 1LPSW-4 __ 1LPSW-5 THEN notify SRO to initiate action to open at least one valve prior to BWST level ≤ 19'.

42. __ IAAT BWST level ≤ 19', THEN initiate Encl 5.12 (ECCS Suction Swap to RBES).

1. __ Display BWST level using OAC Turn-on Code "SHOWDIG O1P1600".

2. __ Notify crew of BWST level IAAT step.

43. __ Dispatch an operator to perform Encl 5.2 (Placing RB Hydrogen Analyzers In Service). (PS)

44. __ Select DECAY HEAT LOW FLOW ALARM SELECT switch to ON.

45. __ IAAT ES channels 5 & 6 have actuated, THEN perform Step 46.

__ GOTO Step 47.

NOTE RBCU transfer to low speed will NOT occur until 3 minute time delay is satisfied.

46. __ Verify all ES channel 5 & 6 components are in the ES position.

__ Notify SRO to evaluate components NOT in ES position and initiate action to place in ES position if desired.


Page 44 of 58


47. __ IAAT ES channels 7 & 8 have actuated,

THEN perform Steps 48 - 49. __ GOTO Step 50.









50. __ Notify U2 CR SRO that SSF is inoperable due to OTS1-1 open.

51. __ Ensure any turnover sheet compensatory measures for ES actuation are complete as necessary.

52. __ IAAT conditions causing ES actuation have cleared, THEN initiate Encl 5.41 (ES Recovery).

53. __ WHEN CR SRO approves, THEN EXIT.

••• END •••


Page 45 of 58


Unit Status

ES Channels 3 & 4 have NOT actuated.


CT-2


56. Verify open: __ 1CF-1 __ 1CF-2







3. __ IF 1HP-20 fails to close, AND NO RCPs operating, THEN close:

__ 1HP-228 __ 1HP-226 __ 1HP-232 __ 1HP-230 4. __ Notify SRO to evaluate components NOT

in ES position and initiate action to place in ES position if desired.


61. __ Close 2LPSW-139.



63. __ Close 1LPSW-139.


Page 46 of 58


64. Place in FAIL OPEN:

__ 1LPSW-251 FAIL SWITCH __ 1LPSW-252 FAIL SWITCH


66. Verify either: __ Three LPSW pumps operating __ Two LPSW pumps operating when


__ GOTO Step 68.

67. Open: __ 1LPSW-4 __ 1LPSW-5





69. __ Dispatch an operator to perform Encl 5.2 (Placing RB Hydrogen Analyzers In Service). (PS )





••• END •••


Page 47 of 58

Enclosure 5.5

Pzr and LDST Level Control


NOTE Maintaining Pzr level >100″ [180″ acc] will ensure Pzr heater bundles remain covered.

1. __ Utilize the following as necessary to maintain desired Pzr level: • 1A HPI Pump • 1B HPI Pump • 1HP-26 • 1HP-7 • 1HP-120 setpoint or valve demand • 1HP-5

__ IF 1HP-26 will NOT open, THEN throttle 1HP-410 to maintain desired Pzr level.

2. __ IAAT makeup to the LDST is desired, THEN makeup from 1A BHUT.

3. __ IAAT it is desired to secure makeup to LDST, THEN secure makeup from 1A BHUT.

4. __ IAAT it is desired to bleed letdown flow to 1A BHUT, THEN perform the following:

A. Open: 1CS-26 1CS-41

B. Position 1HP-14 to BLEED.

C. Notify SRO.

5. __ IAAT letdown bleed is NO longer desired, THEN position 1HP-14 to NORMAL.


Page 48 of 58

Enclosure 5.5 Pzr and LDST Level Control


6. __ IAAT 1C HPI PUMP is required,

THEN perform Steps 7 - 9.

__ GO TO Step 10.

7. Open: • 1HP-24 • 1HP-25

1. __ IF both BWST suction valves (1HP-24 and 1HP-25) are closed, THEN perform the following:

A. Start 1A LPI PUMP. B. Start 1B LPI PUMP. C. Open:

1LP-15 1LP-16 1LP-9 1LP-10 1LP-6 1LP-7

D. __ IF two LPI Pumps are running only to provide HPI pump suction, THEN secure one LPI pump.

E. __ Dispatch an operator to open 1HP-363 (Letdown Line To LPI Pump Suction Block) (A-1-119, U1 LPI Hatch Rm, N end).

F. GO TO Step 8.

2. __ IF only one BWST suction valve (1HP-24 or 1HP-25) is open, THEN perform the following:

A.__ IF three HPI pumps are operating, THEN secure 1B HPI PUMP.

B.__ IF < 2 HPI pumps are operating, THEN start HPI pumps to obtain two HPI pump operation, preferably in opposite headers.

C. GO TO Step 9.


Page 49 of 58



8. Start 1C HPI PUMP.

__ IF at least two HPI pumps are operating, THEN throttle 1HP-409 to maintain desired Pzr level.

9. Throttle the following as required to maintain desired Pzr level: __ 1HP-26 1HP-27

1. IF at least two HPI pumps are operating, AND 1HP-26 will NOT open, THEN throttle 1HP-410 to maintain desired Pzr level.

2. __ IF 1A HPI PUMP and 1B HPI PUMP are operating, AND 1HP-27 will NOT open, THEN throttle 1HP-409 to maintain desired Pzr level.


Page 50 of 58


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 10. __ IAAT LDST level CANNOT be

maintained, THEN perform Step 11.

__ GO TO Step 12.

11. Perform the following: • Open 1HP-24. • Open 1HP-25. • Close 1HP-16.


A. Start 1A LPI PUMP.

B. Start 1B LPI PUMP.

C. Open:

1LP-15

1LP-16

1LP-9

1LP-10

1LP-6

1LP-7



F. __ GO TO Step 13.

2. __ IF only one BWST suction valve (1HP-24 or 1HP-25) is open, AND three HPI pumps are operating, THEN secure 1B HPI PUMP.

NOTE Maintaining Pzr level > 100” [180” acc] will ensure Pzr heater bundles remain covered.

12.__ Operate Pzr heaters as required to maintain heater bundle integrity.


Page 51 of 58


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 13. __ IAAT additional makeup flow to LDST

is desired, AND 1A BLEED TRANSFER PUMP is operating, THEN dispatch an operator to close 1CS-48 (1A BHUT Recirc) (A-1-107, Unit 1 RC Bleed Transfer Pump Rm.).

14. __ IAAT two Letdown Filters are desired, THEN perform the following: Open 1HP-17. Open 1HP-18

15. IAAT all of the following exist: Letdown isolated LPSW available Letdown restoration desired THEN perform Steps 16 - 34. {41}

GO TO Step 35.

16. Open: __ 1CC-7 __ 1CC-8

1. __ Notify CR SRO that letdown CANNOT be restored due to inability to restart the CC system.

2. __ GO TO Step 35.

17. __ Ensure only one CC pump running.

18. __ Place the non-running CC pump in AUTO.

19. Verify both are open: __ 1HP-1 __ 1HP-2

1. __ IF 1HP-1 is closed due to 1HP-3 failing to close, THEN GO TO Step 21.


20. GO TO Step 23.

NOTE Verification of leakage requires visual observation of East Penetration Room.

21. Verify letdown line leak in East Penetration Room has occurred.

GO TO Step 23.

22. GO TO Step 35.


Page 52 of 58


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 23. __ Monitor for unexpected conditions

while restoring letdown.

24. Verify both letdown coolers to be placed in service.

1. __ IF 1A letdown cooler is to be placed in service, THEN open: 1HP-1 1HP-3

2. __ IF 1B letdown cooler is to be placed in service, THEN open: 1HP-2 1HP-4


25. Open: 1HP-1 1HP-2 1HP-3 1HP-4

26. __ Verify at least one letdown cooler is aligned.

Perform the following:

A. __ Notify CR SRO of problem.

B. __ GO TO Step 35. 27. __ Close 1HP-6.

28. __ Close 1HP-7.

29. __ Verify letdown temperature < 125°F. 1. __ Open 1HP-13.

2. Close: 1HP-8 1HP-9&11

3. __ IF any deborating IX is in service, THEN perform the following: A. Select 1HP-14 to NORMAL. B. Close 1HP-16.

4. __ Select LETDOWN HI TEMP INTLK BYP switch to BYPASS.


Page 53 of 58


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 30. Open 1HP-5.

31. Adjust 1HP-7 for ≈ 20 gpm letdown.

32. WHEN letdown temperature is < 125°F, THEN place LETDOWN HI TEMP INTLK BYP switch to NORMAL.

33. Open 1HP-6.

34. __ Adjust 1HP-7 to control desired letdown flow.

NOTE AP/32 (Loss of Letdown) provides direction to cool down the RCS to offset increasing pressurizer level.

35. __ IAAT it is determined that letdown is unavailable due to equipment failures or letdown system leakage, THEN notify CR SRO to initiate AP/32 (Loss of Letdown).

36. __ IAAT > 1 HPI pump is operating, AND additional HPI pumps are NO longer needed, THEN perform the following:

A. __ Obtain SRO concurrence to reduce running HPI pumps.

B. Secure the desired HPI pumps.

C. __ Place secured HPI pump switch in AUTO, if desired.

37. IAAT all the following conditions exist: Makeup from BWST NOT required LDST level > 55″ All control rods inserted Cooldown Plateau NOT being used THEN close: 1HP-24 1HP-25


Page 54 of 58


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 38. Verify 1CS-48 (1A BHUT Recirc) has

been closed to provide additional makeup flow to LDST.

__ GO TO Step 40.

39. __ WHEN 1CS-48 (1A BHUT Recirc) is NO longer needed to provide additional makeup flow to LDST, THEN perform the following:

A. __ Stop 1A BLEED TRANSFER PUMP.

B. __ Locally position 1CS-48 (1A BHUT Recirc) one turn open (A-1-107, Unit 1 RC Bleed Transfer Pump Rm.).

C. Close 1CS-46.

D. __ Start 1A BLEED TRANSFER PUMP.

E. __ Locally throttle 1CS-48 (1A BHUT Recirc) to obtain 90 - 110 psig discharge pressure.

F. __ Stop 1A BLEED TRANSFER PUMP.

40. Verify two Letdown Filters in service, AND only one Letdown filter is desired.

__ GO TO Step 42.

41. Perform one of the following: Place 1HP-17 switch to CLOSE. Place 1HP-18 switch to CLOSE.

42. __ WHEN directed by CR SRO, THEN EXIT this enclosure.

• • • END • • •


Page 55 of 58

Rule 6 HPI

HPI Pump Throttling Limits

• HPI must be throttled to prevent violating the RV-P/T limit.

• HPI pump operation must be limited to two HPIPs when only one BWST suction valve

(1HP-24 or 1HP-25) is open. • HPI must be throttled ≤ 475 gpm/pump (including seal injection for A header) when only

one HPI pump is operating in a header. • Total HPI flow must be throttled ≤ 950 gpm including seal injection when 1A and 1B HPI

pumps are operating with 1HP-409 open. • Total HPI flow must be throttled < 750 gpm when all the following exist:

- LPI suction is from the RBES

- piggyback is aligned

- either of the following exist:

• only one piggyback valve is open (1LP-15 or 1LP-16)

• only one LPI pump operating • HPI may be throttled under the following conditions:

HPI Forced Cooling in Progress:

HPI Forced Cooling NOT in Progress:

All the following conditions must exist: • Core SCM > 0

• CETCs decreasing

All the following conditions must exist: • All WR NIs ≤ 1%

• Core SCM > 0

• Pzr level increasing

• SRO concurrence required if throttling following emergency boration

HPI Pump Minimum Flow Limit

• Maintain ≥ 170 gpm indicated/pump. This is an instrument error adjusted value that

ensures a real value of ≥ 65 gpm/pump is maintained. HPI pump flow less than minimum is allowed for up to 4 hours.


Page 56 of 58

Subsequent Actions EP/1/A/1800/001

Parallel Actions Page 1 of 1

CONDITION ACTIONS

1. PR NIs ≥ 5% FP

OR

NIs NOT decreasing

GO TO UNPP tab.

UNPP

2. All 4160V SWGR de-energized

{13}

GO TO Blackout tab. BLACKOUT

3. Core SCM indicates superheat GO TO ICC tab. ICC

4. Any SCM = 0°F GO TO LOSCM tab. LOSCM

5. Both SGs intentionally isolated to stop excessive heat transfer

GO TO EHT tab.

LOHT 6. Loss of heat transfer (including loss of all Main and Emergency FDW)

GO TO LOHT tab.

7. Heat transfer is or has been excessive

GO TO EHT tab.

EHT

8. Indications of SGTR ≥ 25 gpm GO TO SGTR tab. SGTR

9. Turbine Building flooding NOT caused by rainfall event

GO TO TBF tab.

TBF

10. Inadvertent ES actuation occurred Initiate AP/1/A/1700/042 (Inadvertent ES Actuation).

ES

11. Valid ES actuation has occurred or should have occurred

Initiate Encl 5.1 (ES Actuation).

ES

12. Power lost to all 4160V SWGR and any 4160V SWGR re-energized

• Initiate AP/11 (Recovery from Loss of Power).

• IF Encl 5.1 (ES Actuation) has been initiated, THEN reinitiate Encl 5.1.

ROP

13. RCS leakage > 160 gpm with letdown isolated

Notify plant staff that Emergency Dose Limits are in affect using PA system.

EDL

14. Individual available to make notifications

• Announce plant conditions using PA system.

• Notify OSM to reference the Emergency Plan and AD-LS-ALL-0006 (Notification/Reportability Evaluation).

NOTIFY


Page 57 of 58

CRITICAL TASKS

CT-1 Secure all RCPs within two minutes of SCM ≤ 0°F per Rule 2 (BWOG CT-1)

CT-2 Start Outside Air Booster Fans within 30 minutes of initiation of LOCA (BWOG CT-27)

CT-3 Start 1C HPI Pump within 10 minutes of LOCA to provide flow in both headers to preclude quarter core cooling

SAFETY: Take a Minute UNIT 0 (OSM)

SSF Operable: No U2/U3: Yes

KHU's Operable: U1 - OH, U2 - UG LCTs Operable: 2 Fuel Handling: No

UNIT STATUS (CR SRO)

Unit 1 Simulator Other Units

Mode: 1 Unit 2 Unit 3

Reactor Power: 75% Mode: 1 Mode: 1

Gross MWE: 698 100% Power 100% Power

RCS Leakage: 0.01 gpm No WCAP Action

EFDW Backup: Yes EFDW Backup: Yes

RBNS Rate: 0.01 gpm Technical Specifications/SLC Items (CR SRO)

Component/Train OOS Date/Time

Restoration Required Date/Time

TS/SLC #

AMSAC/DSS Today/0300 7 Days SLC 16.7.2 SSF Today/0100 7 Days TS 3.10.1

Shift Turnover Items (CR SRO) Primary • Due to unanalyzed condition, the SSF should be considered INOPERABLE for Unit 1 if power

levels are reduced below 85%. Evaluations must be performed prior to declaring the SSF operable following a return to power (after going below 85%).

• 1RIA-3 and 5 removed from RB • SASS is in Manual for calibration Secondary • AMSAC/DSS bypassed for calibration • Unit 2 is supplying the AS header • PT/0/A/0610/017 (Operability Test of 4160V Breakers) Encl 13.11 (Functional Verification Of SL

Breaker(s) is in progress and complete up to Step 2.1 to perform functional verification of SL1 and SL2 breakers

• Keowee operability test was performed per PT/0/A/0620/009 (Keowee Hydro Operation) at 0700 today

• Unit 2 BOP will simulate performing CV during SL breaker testing • 1SSH-1, 1SSH-3, 1SD-2, 1SD-5, 1SD-140, 1SD-303, 1SD-355, 1SD-356 and 1SD-358 are

closed with power supply breakers open per the Startup Procedure for SSF Overcooling Event.

Reactivity Management (CR SRO)

RCS Boron 83 ppmB Gp 7 Rod Position: 77% Withdrawn

Batch additions as required for volume control.

Human Performance Emphasis (OSM) Procedure Use and Adherence



________________________ ________________________ OATC



• SASS is in Manual for calibration • AMSAC/DSS is bypassed for calibration • PT/1/A/0290/003 Encl 13.2 (Control Valve Movement At Power) in progress to test CV3 & CV4

Event No.


Description



1 N: BOP, SRO Turbine Control Valve Movement PT (CV3 & CV4)

2 Override C: BOP, SRO CCW Emergency Discharge Level Low

3 MPI121 I: BOP, SRO PZR Level 1 Fails Low

4 Override I: OATC, SRO (TS) Inadvertent ES Channel 3 Actuation

5 MPI281 I: OATC, SRO ΔTc Controller Failure

6 Override R: OATC, SRO (TS) One Dropped Control Rod (Group 1 Rod 6) Requiring Manual Power Decrease

7 MPI290 MPI300 MEL120

M: ALL ATWS (Loss of 1TA Switchgear)

• Turbine Fails to Trip • 1HP-26 Fails Closed



Page 2 of 43


Event 1: When the crew takes the shift, the BOP will perform PT/1/A/0290/003 Enclosure 13.2 (Control Valve Movement At Power) to test Turbine Control Valves CV3 & CV4. Once the test is complete, the Turbine Bypass Valves (TBVs) will be closed, returned to automatic, and the Standby EHC pump will be secured and placed in automatic.

Event 2: An OAC alarm will indicate that CCW Emergency Discharge Level is low. This will require the BOP to re-prime the Condenser Emergency Discharge Line in accordance with OP/1/A/1104/012 Enclosure 4.4 (Repriming Condenser Emergency Discharge Line).

Event 3: Pressurizer Level 1 will fail low which will cause 1HP-120 (RCS Volume Control) to open in an attempt to restore indicated Pressurizer level. The BOP will select Pressurizer level 2 or Pressurizer level 3 to restore normal RCS makeup flow.

Event 4: Engineered Safeguards (ES) Channel 3 will inadvertently actuate which will cause the 1A LPI Pump to start, 1LP-17 to open, and ‘C’ LPSW Pump to start. Once the crew determines that the ES actuation is not valid, the SRO will enter AP/1/A/1700/042 (Inadvertent ES Actuation) to restore the ES components to normal operation. The SRO will enter TS 3.3.7 and TS 3.7.7.

Event 5: The ΔTc Controller will fail causing 1A Feedwater flow to increase and 1B Feedwater flow to decrease. Once alarms are received, the crew will perform Plant Transient Response to stabilize the plant. Once the plant is stable, the SRO will enter AP/1/A/1700/028 (ICS Instrument Failures) and ensure the appropriate ICS stations are in manual.

Event 6: One Control Rod (Group 1 Rod 6) will fully insert. Once alarms are received, the crew will perform Plant Transient Response to stabilize the plant. The SRO will enter AP/1/A/1700/001 (Unit Runback). Since ICS is in manual (from event 5), the OATC will perform a manual power decrease to ≤ 55% power. The SRO will enter TS 3.1.4, TS 3.1.5, and TS 3.10.1.

Event 7: 1TA Switchgear will lockout and the Reactor will fail to trip automatically or manually (ATWS). The OATC will perform Rule 1 and the SRO will enter the UNPP tab of the EOP. The turbine will fail to trip using the Turbine Trip Pushbutton which will require the operator to lockout both EHC Pumps to prevent severe overcooling of the RCS. 1HP-26 will fail closed requiring 1HP-410 to be opened to allow full HPI flow from the BWST to borate the RCS and shutdown the Reactor.


Page 3 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 1 Page 1 of 3 Event Description: Turbine Control Valve Movement PT (N: BOP, SRO) Time Position Applicant’s Actions or Behavior

PT/1/A/0290/003 Crew response: SRO directs the BOP to perform PT/1/A/0290/003 Encl 13.2 to test CV3 & CV4 PT/1/A/0290/003 Encl 13.2 (Control Valve Movement At Power) rev 15

2.5 IF AT ANY TIME reactor/turbine trip OR significant transient occurs, ensure Turbine Bypass Valves to AUTOMATIC

2.6 IF CV3 and CV4 test required:

2.6.1 Select "Control Valve 3 & 4 Test"

2.6.2 Verify the following: ___ "Test Permissive" is ON for CV3 ___ "Test Permissive" is ON for CV4

2.6.3 Record CV3 and CV4 pretest positions: • CV3 pretest position:________ % Open • CV4 pretest position: ________% Open

2.6.4 Select "Initiate CV3 and CV4 Test" 2.6.5 IF any of the following conditions occur, select "Abort CV3 &

CV4 Test" NI POWER changes > 2% ICS Turbine Master trips to HAND Turbine vibration > 10 mils for > 5 seconds

2.6.6 IF "Test Failed" is "ON" AND CV3 is NOT fully closed, select "Abort CV3 & CV4 Test"

This event is complete when the Standby EHC pump switch is placed in AUTO (step 2.9.5), or as directed by the Lead Examiner.

NOTE: CV3 & CV4 testing is combined into single test with CV3 ramping closed and CV4 opening. When CV3 is fully tested, CV4 is disc dumped at < 6 % open to complete test.

CAUTION: If CV3 remains in the closed position with the Test Failed indication "ON", initiating Abort Test could result in a reactivity management event.


Page 4 of 43


PT/1/A/0290/003 Crew response: PT/1/A/0290/003 Encl 13.2 (Control Valve Movement At Power)

2.6.7 IF "Test Failed" is "ON" AND CV3 remained closed perform the following: A. Do NOT select "Abort CV3 & CV4 Test" B. Notify WCC & Engineering that the (FASV) for the Control

Valve under test is stuck in the energized state

C. Monitor Turbine Vibrations closely if in this abnormal state

2.6.8 Perform EITHER for CV3: A. Verify "Test Successful" indicated for CV3 B. IF "Test Successful" NOT indicated for CV3, verify CV3

moved towards closed position

2.6.9 Perform EITHER for CV4: A. Verify "Test Successful" indicated for CV4 B. IF "Test Successful" NOT indicated for CV4, verify CV4

moved towards closed position


NOTE: If a Control Valve remains closed after it has disc-dump, it may start going open at any time (i.e. 1 min, 5 min, 1 hour,...). When the Control Valve starts going back open it will open at its normal rate. No transients are expected during this scenario since the EHC Control System will simply continue with the test logic and return the Control Valve to its normal position at the normal controlled test rate.

NOTE: Control Valves which are not in their normal position could result in asymmetrical loading on the Turbine bearings.


Page 5 of 43


PT/1/A/0290/003 Crew response: PT/1/A/0290/003 Encl 13.2 (Control Valve Movement At Power)

2.6.10 Verify CV3 test indicator reset

2.6.11 Verify CV4 test indicator reset

2.6.12 Verify CV3 within ± 5.0% of pretest position

2.6.13 Verify CV4 within ± 5.0% of pretest position

2.6.14 Perform the following:

Verify acceptance criteria met

IF acceptance criteria NOT met, notify SRO

Examiner Note: Steps 2.7 and 2.8 are not applicable. 2.9 Perform the following:

2.9.1 Ensure all Turbine Bypass Valves CLOSED 2.9.2 Verify CV-4 returns to its expected pre-test position

(Reference step 2.4.1 or Encl 13.1 step 2.3.1)

Examiner Note: CV-4 should return to ≈ 4% open. 2.9.3 Ensure the following:

• 1A Turbine Bypass Valve AUTO • 1B Turbine Bypass Valve AUTO

2.9.4 Stop Standby EHC pump 2.9.5 Place Standby EHC pump control switch to "AUTO" 2.9.6 IF required, make the following plant page announcement: "Personnel can re-enter the Unit 1 Turbine Building"


NOTE: • CV3 and CV4 test indicator resets after ≈ 20 seconds

• Test indicators reset is when "Test Permissive" is ON and all other Test Indicators are OFF


Page 6 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 2 Page 1 of 2 Event Description: CCW Emergency Discharge Level Low (C: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

BOP

Plant response:

• OAC alarm CCW EMER DISCHARGE LEVEL LOW Crew response: The BOP will refer to OAC alarm response

OAC Alarm OAC alarm CCW EMER DISCHARGE LEVEL LOW

1. Reprime ECCW per OP/1/A/1104/012 (CCW System)

Booth Cue: If contacted as Unit 2 and/or Unit 3 about OAC alarm for

CCW EMER DISCHARGE LEVEL LOW, state that the alarm indicates LOW.

OP/1/A/1104/012 Encl 4.4 OP/1/A/1104/012 (CCW System) Encl 4.4 (Repriming Condenser Emergency Discharge Line) rev 82

Initial Conditions

1.1 Verify CCW System in service 1.2 Review Limits and Precautions

Procedure

2.1 Ensure closed CCW-8 (EMERGENCY CCW DISCHARGE TO TAILRACE)

2.2 Open 1DP-F5C (CCW-8 Bkr (Emerg CCW Disch To Tailrace))

2.3 Open 1CCW 1-6 (WATERBOX EMERG DISCH) 2.4 Throttle open CCW-9 (EMERGENCY CCW DISCHARGE TO INTAKE

CANAL) to provide vent path

This event is complete when Step 2.5 is complete (CCW-8 Breaker is closed), or as directed by the Lead Examiner.


Page 7 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 2 Page 2 of 2 Event Description: CCW Emergency Discharge Level Low (C: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

OP/1/A/1104/012 Crew response: OP/1/A/1104/012 (CCW System) Encl 4.4 (Repriming Condenser Emergency Discharge Line)

2.5 WHEN OAC indicates line is primed OR after 30 minutes:

2.5.1 Close CCW-9 (EMERGENCY CCW DISCHARGE TO INTAKE CANAL)

2.5.2 Close 1CCW 1-6 (WATERBOX EMERG DISCH) 2.5.3 IF OAC available, verify by OAC indication the following valves

are closed: • 1CCW-1 (CCW 1A1 Emer Disch) (CP O1D0299) • 1CCW-2 (CCW 1A2 Emer Disch) (CP O1D0301) • 1CCW-3 (CCW 1B1 Emer Disch) (CP O1D0303) • 1CCW-4 (CCW 1B2 Emer Disch) (CP O1D0305) • 1CCW-5 (CCW 1C1 Emer Disch) (CP O1D0307) • 1CCW-6 (CCW 1C2 Emer Disch) (CP O1D0309)

2.5.4 IF OAC NOT available, verify locally the following valves are closed: (T-3, catwalk) (N/A)

2.6 Close 1DP-F5C (CCW-8 Bkr (Emerg CCW Disch To Tailrace))

This event is complete when Step 2.5 is complete (CCW-8 Breaker is closed), or as directed by the Lead Examiner.

NOTE: Any units' O*D2673 point is acceptable to verify ECCW discharge piping full.


Page 8 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 3 Page 1 of 2 Event Description: PZR Level 1 Fails Low (I: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP

BOP

Plant response:

• 1SA-02/C-3 (RC Pressurizer Level High/Low) • 1SA-02/C-4 (RC Pressurizer Level Emerg High/Low) • PZR level 1 indicates 0 inches • 1HP-120 (RC Volume Control) throttles open • Makeup flow increases

ARG 1SA-02/C-3 Crew response: The BOP will refer to Statalarm 1SA-02/C-3 and/or 1SA-02/C-4 ARG 1SA-02/C-3 (RC Pressurizer Level High/Low) rev 33

3.1 Check alternate PZR level indications

3.2 Check for proper Makeup/Letdown flows and adjust to restore proper level

3.3 Refer to the following procedures as required: • AP/1/A/1700/002 (Excessive RCS Leakage) • AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal

Injection) • AP/1/A/1700/032 (Loss of Letdown)

3.4 Refer to Technical Specification 3.4.9

3.5 Refer to Technical Specification 3.3.8

3.6 Refer to OP/1/A/1105/014 (Control Room Instrumentation Operation And Information)

ARG 1SA-02/C-4 ARG 1SA-02/C-4 (RC Pressurizer Level Emerg High/Low) rev 33

3.1 Check alternate PZR level indications

3.2 Check for proper makeup/letdown flows and adjust to restore proper level

3.3 Refer to AP/1/A/1700/002 (Excessive RCS Leakage) and/or AP/1/A/1700/014 (Loss of Normal HPI Makeup and/or RCP Seal Injection) for additional instructions

3.4 IF PZR level decreases to less than 80", verify PZR heaters have deenergized, IF NOT, do so manually

This event is complete when PZR level 2 or PZR level 3 is selected, or as directed by the Lead Examiner.


Page 9 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 3 Page 2 of 2 Event Description: PZR Level 1 Fails Low (I: BOP, SRO) Time Position Applicant’s Actions or Behavior

OP/1/A/1105/014 Crew response: Examiner Note: The SRO may direct the BOP to select Pressurizer level 2

or level 3 (located on 1UB1) prior to referencing OP/1/A/1105/014 (Control Room Instrumentation Operation And Information).

Examiner Note: The SRO may direct the BOP to take manual control of 1HP-120 to control pressurizer level prior to referencing OP/1/A/1105/014 (Control Room Instrumentation Operation And Information).

OP/1/A/1105/014 Encl 4.11 (SASS Information) rev 42

3.2 SASS (Smart Automatic Signal Selector) Manual Operation

3.2.1 IF "MISMATCH" light is on and "TRIP ‘A’" or "TRIP ‘B’" light is on, a SASS trip has occurred A. Controlling signal will be signal selected from CR keyswitch

(for parameters in ICS Cabinet #8) B. Select valid signal as controlling signal by positioning CR

keyswitch or pushbutton for Pzr level to valid signal (for parameters in ICS Cabinet #8)

3.2.2 IF "MISMATCH" light is on, a mismatch has occurred: A. Controlling signal will be signal selected from CR keyswitch

(for parameters in ICS Cabinet #8) B. Select valid signal as controlling signal by positioning CR

keyswitch or pushbutton for Pzr level to valid signal (for parameters in ICS Cabinet #8)

3.2.3 Initiate a Work Request to repair faulty signal

This event is complete when PZR level 2 or PZR level 3 is selected, or as directed by the Lead Examiner.


Page 10 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 4 Page 1 of 5 Event Description: Inadvertent ES Channel 3 Actuation (I: OATC, SRO) (TS) Time Position Applicant’s Actions or Behavior

BOP

Plant response:

• 1SA-01/C-10 (ES 3 TRIP) • 1A LPI Pump starts • 1LP-17 (1A LPI INJECTION) opens • ‘C’ LPSW Pump starts

Crew response:

• The crew will determine that the ES actuation is not valid • The SRO may direct the BOP to refer to ARG 1SA-01/C-10 (ES 3 TRIP) • The SRO will enter AP/42 (Inadvertent ES Actuation) (page 11)

ARG 1SA-01/C-10

ARG 1SA-01/C-10 (ES 3 TRIP) rev 14

3.1 Determine if ES condition exists (RCS pressure ≤ 550 psig OR RB pressure ≥ 3.0 psig)

3.2 IF RCS pressure is ≤ 550 psig OR RB pressure is ≥ 3.0 psig, Go To EP/1/A/1800/001 (Emergency Operating Procedure)

3.3 IF ES condition does NOT exist, Initiate AP/1/A/1700/042 (Inadvertent ES Actuation)

3.4 Refer to OP/1/A/1105/014 (Control Room Instrumentation Operation And Information)

This event is complete when the SRO reaches Step 4.26, or as directed by the Lead Examiner.


Page 11 of 43


OATC

AP/1/A/1700/042 Crew response: AP/1/A/1700/042 (Inadvertent ES Actuation) rev 04

4.1 Verify any of the following have inadvertently actuated: (N/A) __ Diverse HPI __ ES Channel 1 __ ES Channel 2

RNO: GO TO Step 4.4

4.4 Verify any of the following have inadvertently actuated: (N/A) __ ES Channel 5 __ ES Channel 6

RNO: 1. IF ES Channel 1, ES Channel 2, or Diverse HPI have inadvertently actuated, AND it is desired to restore letdown, THEN initiate AP/42 Encl 5.2 (Letdown Restoration). (N/A)

2. GO TO Step 4.10

4.10 Close the following: __ 1HP-24 (already closed) __ 1HP-25 (already closed)

4.11 Ensure AP/42 Encl 5.1 (Required Operator Actions) is in progress (page 13)

4.12 Verify any of the following have inadvertently actuated: __ Diverse LPI (not actuated) __ ES Channel 3 __ ES Channel 4 (not actuated)

4.13 Verify Diverse LPI has inadvertently actuated (N/A) RNO: GO TO Step 4.15

4.15 Perform the following on all inadvertently actuated system(s): __ Ensure ES CH-3 is in MANUAL __ Ensure ES CH-4 is in MANUAL (N/A)


NOTE If personnel are available, should continue while Encl 5.1 (Required Operator Actions) is in progress.


Page 12 of 43


OATC

AP/1/A/1700/042 Crew response: AP/1/A/1700/042 (Inadvertent ES Actuation)

4.16 Verify LPI was aligned in decay heat removal mode prior to ES actuation

RNO: 1. Stop the following: __ 1A LPI PUMP __ 1B LPI PUMP (not operating) 2. Simultaneously close the following: __ 1LP-17 __ 1LP-18 (already closed)

4.17 Verify the Rx is critical

4.18 Verify ICS in Auto

4.19 Verify control rods are outside the desired control band


4.21 Verify any of the following have inadvertently actuated: __ ES Channel 1 (not actuated) __ Diverse HPI (not actuated)


4.24 Notify SPOC to investigate and repair the cause of the inadvertent ES actuation, as necessary

4.25 Initiate logging TS/SLC Entry/Exit, as applicable, in accordance with Encl 5.4 (TS/SLC Requirements)

4.26 WHEN all the following exist: __ Reason for inadvertent ES Channel or Diverse HPI/LPI actuation

has been resolved __ ES Channel or Diverse HPI/LPI reset is desired __ OSM concurs

THEN continue



Page 13 of 43


BOP

AP/1/A/1700/042 Encl 5.1 Crew response: AP/1/A/1700/042 Enclosure 5.1 (Required Operator Actions)

1. Initiate announcement of AP Entry using the PA system

2. Verify any of the following have inadvertently actuated:

RNO: GO TO Step 5

5. Verify any of the following have inadvertently actuated: __ ES Channel 7 __ ES Channel 8

RNO: GO TO Step 9

9. Perform the following: A. Open the following to restore RB RIAs:

__ 1PR-7 __ 1PR-8 __ 1PR-9 __ 1PR-10

Examiner Note: The SRO may elect to NOT stop the RB RIA sample pump to prevent inadvertently entering TS 3.4.15 since ES Channel 3 actuation does not isolated the RB RIAs.

B. From the ENABLE CONTROLS screen on the RIA View Node, perform the following: 1) Select OFF for RB RIA sample pump 2) Start the RB RIA sample pump

10. Verify any of the following have inadvertently actuated: __ Diverse HPI __ ES Channel 1

RNO: GO TO Step 12

12. EXIT this enclosure


NOTE If channels are bypassed or in override, 1SA-1/A-10 (ES 1 Trip) and 1SA-1/B-10 (ES 2 Trip) will be off even though the channel may have actuated.


Page 14 of 43


SRO

Tech Specs Crew response:

TS 3.3.7 ESPS AUTOMATIC ACTUATION OUTPUT LOGIC CHANNELS Condition A (1hour) Place associated component(s) in ES configuration

OR (1 hour) Declare the associated component(s) inoperable

TS 3.7.7 LOW PRESSURE SERVICE WATER SYSTEM (LPSW) Condition B (7 days) Restore the LPSW WPS to OPERABLE status If the RIA sample pump is turned off per AP/42 Encl 5.1 step 9, TS 3.4.15 Condition B will be entered while the sample pump is off. TS 3.4.15 RCS LEAKAGE DETECTION INSTRUMENTATION Condition B (Once per 24 hours) Analyze grab samples of the containment atmosphere

OR (Once per 24 hour) Perform SR 3.4.13.1

AND Restore required containment atmosphere radioactivity monitor to OPERABLE status



Page 15 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 5 Page 1 of 3 Event Description: ΔTc Controller Failure (I: OATC, SRO) Time Position Applicant’s Actions or Behavior

OATC

BOP

SRO

Plant response:

• FDW flow will ratio incorrectly based on the failure • “A” FDW flow will decrease causing “A” loop TC to increase • “B” FDW flow will increase causing “B” loop TC to decrease • This will cause actual ∆TC to increase (become more positive). Failure to

correctly adjust FDW flow will result in QPT alarms. • 1SA-02/B-5 (RC Cold Leg Diff. Temperature High) will actuate if actual ∆TC

increases to ± 5°F • 1SA-02/C-9 (MS STM GEN ‘B’ LEVEL High/Low) will actuate if 1B SG

Operating Range Level reaches ≥ 86%

Crew response: When the Statalarms are received, the crew should perform Plant Transient Response (PTR) to stabilize the plant • Diagnose the ∆TC failure by observing the ∆TC meter on 1UB1 • The OATC will place the Diamond and Feedwater Masters to MANUAL

and re-ratio feedwater using the Loop TC meters and/or OAC (RCS01) to return actual ∆TC to near zero

The OATC should: • Communicate to the CRS the initial alarm (if applicable) followed by reactor

power level and direction • Place the appropriate ICS stations in manual (Diamond and both FDW

Masters in this case) in manual if any of the following occur: o NI power increasing above the pre-transient power level o Failed instrument is diagnosed o Invalid input exists and the CRS directs the ICS be placed in manual

• Remain focused on reactor power level and FDW response during the transient

The BOP should: • Determine if a valid ICS runback exists and inform the CRS • Monitor plant response and verify operating limits NOT exceeded • If ICS is placed in Manual, remain focused on RCS pressure, SG outlet

pressure and RCS inventory The SRO should: • Refer to AP/28 (ICS Instrument Failures) (page 16) • Ensure FIN-24 (SPOC) is contacted to repair the failed instrument

This event is complete when the SRO reaches Step 4 of AP/28 Section 4F, or as directed by the Lead Examiner.


Page 16 of 43


OATC/SRO


4.1 Provide control bands as required (OMP 1-18 Attachment I) OMP 1-18 Attachment I: 1. Plant Conditions Stable or TPB ≤ Pre-transient Conditions

1.1 The following bands are to be established during manual control of plant conditions (as needed) but may be adjusted by the CRS if required 1.1.1 NI Power ± 1% not to exceed the pre-transient or allowable

power. If at the pre-transient or allowable level, band is NI Power – 1%

1.1.2 Current Tave ± 2°F 1.1.3 Current SG Outlet Pressure ± 10 PSIG (N/A) 1.1.4 Delta Tc 0oF ± 2°F

4.2 Initiate notification of the following: __ OSM to reference the following:

• OMP 1-14 (Notifications) • Emergency Plan

__ STA

4.3 Verify a power transient ≥ 5% has occurred

RNO: GO TO Step 4.5

4.4 Notify Rx Engineering and discuss the need for a maneuvering plan

4.5 Use the following, as necessary, to determine the applicable section from table in Step 4.6: • OAC alarm video • OAC display points • Control Board indications • SPOC assistance, as needed



Page 17 of 43


OATC/SRO

AP/1/A/1700/028 Crew response: AP/1/A/1700/028 (ICS Instrument Failures)

4.6 GO TO the applicable section per the following table:

√ Section Failure

4F Delta Tc AP/1/A/1700/028 Section 4F (Delta Tc Failure)

NOTE • This Section applies to Delta Tc controller failures. Tc input signal failures

are addressed in Section 4A

The following may occur when an ICS Delta Tc controller fails: • Delta Tc controller may re-ratio loop feedwater flows • Possible ICS RUNBACK

1. Ensure the following in HAND: __ 1A FDW MASTER __ 1B FDW MASTER __ DELTA Tc

CAUTION Total feedwater flow should be maintained constant while individual loop flows are adjusted to establish the desired ΔTc. Maintaining total FDW flow constant will prevent unwanted changes in reactor power.

2. Re-ratio feedwater flow, as required, to establish desired DELTA Tc while maintaining total feedwater flow constant

3. Notify SPOC to perform the following: __ Investigate and repair the failed Delta Tc controller

4. WHEN notified by SPOC that Delta Tc controller has been repaired, THEN GO TO OP/1/A/1102/004 A Encl (Placing ICS Stations To Auto)



Page 18 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 6 Page 1 of 7 Event Description: One Dropped Control Rod (R: OATC, SRO) (TS) Time Position Applicant’s Actions or Behavior

Plant Response:

• Group 1 Rod 6 drops into the core • Statalarm 1SA-2/A-10 (CRD GLOBAL SYSTEM FAULT) • Statalarm 1SA-2/B-10 (CRD ASYMMETRIC ROD POSITION ERROR) • Statalarm 1SA-2/D-9 (CRD OUT INHIBIT) • Statalarm 1SA-4/C-1 (QUADRANT POWER TILT) (in at ≈ 2 minutes) • Statalarm 1SA-5/A-5 (1A RPS TROUBLE) • Statalarm 1SA-5/D-5 (1D RPS TROUBLE)

Crew Response:

Crew should perform Plant Transient Response (PTR) and determine that no manual FDW adjustments are required and that a runback condition exists but due to ICS being in manual it is not occurring. • OATC reports to the SRO reactor power level and direction of movement. • The BOP reports expected AUTO Runback did not occur, and monitors RCS

pressure and inventory and inserts Control Rods as needed. • The OATC will adjust FDW and/or control rods as necessary to restore

reactor power to the desired control band.

SRO should enter AP/1/A/1700/001 (Unit Runback)

AP/1/A/1700/001 AP/1/A/1700/001 (Unit Runback) rev 15

4.1 GO TO the most limiting section per the following table:

√ Section Runback

4H Asymmetric Control Rod (1%/min to 55%power)

This event is complete when Rx Power has decreased to < 55%, or as directed by the Lead Examiner.


Page 19 of 43


AP/1/A/1700/001 Section 4H Crew response: AP/1/A/1700/001 Section 4H (Asymmetric Control Rod)

1. IAAT a more limiting runback occurs, THEN GO TO Subsequent Actions Step 4.1

2. IAAT more than one control rod is dropped or misaligned ≥ 6.5% (9″) from the group average, THEN trip the Rx

NOTE NIs should NOT be calibrated per guidelines contained in OP/1/A/1102/004 (Operation at Power) due to actual power re-distribution within the core as a result of a dropped/misaligned rod.

3. Verify Rx is critical

4. Verify power > 55% when the rod was dropped or misaligned

5. Verify Rx runback to 55% core thermal power in progress • CTPD set at 55% • ASYMETRIC RODS Runback Light lit • CTP Demand decreasing • Reactor power will decrease when the runback catches up with the

initial power decrease from the dropped rod

RNO: 1. Initiate power reduction to ≤ 55% core thermal power at ≥ 1%/min 2. IF control rods will not insert manually, THEN perform the following:

A. Trip reactor B. GO TO Unit 1 EOP

6. Initiate Encl 5.1 (Control of Plant Equipment During Shutdown) (page 23)

NOTE The following actions should be performed as quickly as possible due to the complexity of resetting RPS trip setpoints and Tech Spec time limits.

7. Notify SPOC to perform the following: __ Investigate cause of dropped or misaligned control rod. __ Prepare to reduce the following trip setpoints:

• RPS Flux/Flow-Imbalance • RPS High Flux



Page 20 of 43



8. Notify the OSM to ensure the requirements of the following Tech Specs are met: (page 22) __ TS 3.1.4 (Control Rod Group Alignment Limits) __ TS 3.1.5 (Safety Rod Position Limits) __ TS 3.2.3 (Quadrant Power Tilt)

Booth Cue: When SM is contacted, inform the team that the SM is occupied on Unit 3 and can NOT verify TS requirements at this time.

9. Notify OSM to make notifications as required per OMP 1-14 (Notifications).

10. Verify > 1% SDM with allowance for the inoperable control rod per PT/1/A/1103/015 (Reactivity Balance Calculation) within one hour.

Examiner Note: Shutdown Margin will be adequate 11. Reduce core thermal power ≤ the following limits, based on the number

of RCPs operating, within two hours:

RCPs Allowable Thermal Power (% FP) 3 45

4 60

NOTE The following ensures adequate margin in preparation for resetting RPS trip setpoints.

12. IAAT the power decrease is complete, AND any NI is > the following:

RCPs Maximum NI Power (% FP) 3 40

4 55 THEN reduce power until all NIs are ≤ the Maximum NI Power limit for the operating RCP combination per Encl 5.4 (Power Reduction).



Page 21 of 43



13. WHEN all NIs are ≤ the Maximum NI Power limit for the operating RCP combination, THEN notify SPOC to reduce RPS trip setpoints per AM/1/A/0315/017 (TXS RPS Channel A, B, C, And D Parameter Changes For Abnormal/Normal Operating Conditions.)

NOTE Due to the power decrease initiated in this AP, the current plant configuration must be compared to the normal plant configuration in OP/1/A/1102/004 (Operation at Power) power reduction enclosure. Equivalent steps performed by this AP should be signed off as intent met. Any steps NOT performed by this AP must be evaluated in preparation for power increase or continued shutdown.

14. Initiate OP/1/A/1102/004 (Operation at Power) power reduction enclosure.

NOTE The following will prevent a potential MODE change from MODE 2 to MODE 1 if unit power stabilizes ≤ 5%.

15. IAAT reactor power is ≤ 5%, THEN GO TO the following as necessary to shutdown the reactor prior to rod recovery. • OP/1/A/1102/10 (Controlling Procedure for Unit Shutdown) • OP/1/A/1102/004 (Operation at Power)

16. IAAT another runback has or should have occurred, THEN GO TO Subsequent Actions.

17. WHEN the control rod is repaired, THEN perform the following: A. Notify Duty Reactor Engineer for any necessary maneuvering limits. B. GO TO OP/1/A/1105/019 (Control Rod Drive System) to recover the

control rod.



Page 22 of 43


SRO


TS 3.1.4 CONTROL ROD GROUP ALIGNMENT LIMITS Condition A (1hour) Restore control rod alignment or verify SDM (2 hours) Reduce Thermal Power to ≤ 60% of allowable Thermal Power. (10 hours) RPS trip setpoints must be reduced based on flux and flux/flow imbalance to ≤ 65.5% of the allowable thermal power. TS 3.1.5 SAFETY ROD POSITION LIMITS Condition A (1 hour) Verify SDM and declare associated control rod INOPERABLE. TS 3.2.3 QUADRANT POWER TILT Condition A if QPT is greater than the steady state limit (+3.5) and less than or equal to the transient limit (+7.11) (10 hours) RPS trip setpoints must be reduced ≥ 2% RTP for each 1% of QPT greater than the steady state limit (24 hours) QPT restored to less than or equal to the steady state limit

Condition B “Possible” – QPT greater than the transient limit (+7.11) and less than or equal to the maximum limit (+16.55) due to misalignment of a Control Rod or an APSR (30 minutes) Reduce Thermal Power ≥ 2% RTP from Allowable Thermal Power for each 1% of QPT greater than the steady state limit. (2 hours) Restore QPT to less than or equal to the transient limit.

TS 3.10.1 STANDBY SHUTDOWN FACILITY (SSF) Conditions A-E (7 days) Restore to operable status

Once Reactor Power is reduced to below 85% the SSF must be declared inoperable and therefore Tech Spec 3.10.1 applies. Conditions A-E should be entered. Examiner Note: SDM will be verified using PT/1/A/1103/015, Encl 13.18

(Page 25) and Curve for Control Rod Position Setpoints, 1 Inoperable Rod, 4 Pump Flow, BOC to EOC from the COLR (Page 26).



Page 23 of 43


AP/1/A/1700/001 Enclosure 5.1 Crew response: AP/1/A/1700/001 Enclosure 5.1 (Control of Plant Equipment During Runback)

1. IAAT SRO determines all appropriate actions have been taken, AND the runback is complete, THEN EXIT this Enclosure

2. Notify the WCC SRO to initiate Enclosure 5.2 (WCC SRO Support During Unit Runback)

3. Start the following pumps: __ 1A FDWP SEAL INJECTION PUMP __ 1A FDWP AUXILIARY OIL PUMP __ 1B FDWP AUXILIARY OIL PUMP __ 1B FDWP SEAL INJECTION PUMP

4. WHEN CTP ≤ 80%, THEN stop the following pumps: {3} {4} __ 1E1 HTR DRN PUMP __ 1E2 HTR DRN PUMP

5. WHEN CTP ≤ 65%, THEN continue this Enclosure

6. Place the following in MANUAL and close: __ 1FDW-53 __ 1FDW-65

NOTE 1B FDWP is the preferred pump to shut down first.

7. Verify both Main FDWPs operating

8. Verify 1B FDWP to be shut down first

9. Adjust the FWP bias counter-clockwise to lower 1B FWP suction flow ≈ 1 x 106 lb/hr < 1A FWP suction flow

10. GO TO Step 12



Page 24 of 43


AP/1/A/1700/001 Enclosure 5.1 Crew response: AP/1/A/1700/001 Enclosure 5.1 (Control of Plant Equipment During Runback)

12. IAAT both Main FDW pumps running, AND both of the following exist: __ 1B Main FDW pump is first pump to be shut down __ Any of the following alarms occur:

• 1SA-16/A-3 (FWP B FLOW MINIMUM) • 1SA-16/A-4 (FWP B FLOW BELOW MIN)

THEN trip 1B Main FDW Pump

13. IAAT both Main FDW pumps running, AND both of the following exist: __ 1A Main FDW pump is first pump to be shut down __ Any of the following alarms occur:

• 1SA-16/A-1 (FWP A FLOW MINIMUM) • 1SA-16/A-2 (FWP A FLOW BELOW MIN)

THEN trip 1A Main FDW Pump

14. IAAT the operating FDWP suction flow < 1.5 x 106 lb/hr, THEN slowly throttle the associated recirc control valve to establish 2300 - 6000 gpm total Condensate flow: • 1FDW-53 • 1FDW-65

15. Maintain Pzr level between 220” - 250″ 16. IAAT load is ≤ 550 MWe, THEN perform Steps 17 - 18 17. Stop the following:

__ 1A MSRH DRN PUMP __ 1B MSRH DRN PUMP

18. Place the following in DUMP: __ 1HD-37 __ 1HD-52



Page 25 of 43

Enclosure 13.18 rev 75

PT/1/A/1103/015 Shutdown margin Calculation at Power

Page 1 of 1 Performed By: __________Date: ________ Time: __________ 1. Purpose

1.1 The purpose of this enclosure is to perform a shutdown margin calculation while at power.

2. Procedure

NOTE: Step 2.1 applies only to control rod Groups 1 through 6. Group 7 and the APSRs may be positioned as required.

____ 2.1 IF any rod groups are NOT at 100% wd (other than Group 7 and the APSRs) due to

CRD movement PT:

____ 2.1.1 Verify that only one group is NOT at 100% wd.

____ 2.1.2 Verify that the inserted group is ≥ 95% wd.

NOTE: For a dropped rod/stuck rod scenario, utilize the 1 inoperable rod graphs in the COLR. The rod position used should be the position of the controlling group (if the dropped/stuck rod is in the controlling group, the rod positions of the remaining rods in that group should be used, NOT the group average).

____ 2.2 Verify one of the following:

____ ____ 2.2.1 Available shutdown margin is > 1.00% ∆Κ/K. This is shown by verifying that the control rod position and power level are within the Acceptable Region or the Restricted Region on the appropriate curve for the number of RC Pumps and Inoperable rods in the COLR.

OR

____ ____ 2.2.2 Appropriate actions are taken per TS 3.1.4, 3.1.5, and 3.2.1. IV

IV


Page 26 of 43


Page 27 of 43

Op-Test No.: ILT16-2 Scenario No.: 2 Event No.: 7 Page 1 of 6 Event Description: ATWS (Loss of 1TA Switchgear) (M: ALL) Time Position Applicant’s Actions or Behavior

OATC

OATC

BOP

Plant Response:

• Statalarms 1SA-01/A-1, B-1, C-1, D-1 (RPS Channel A-D Trip) • 1SA-1/A-7, B-7, C-7, D-7 (RCP/Flux Trip) •

1TA lockout will occur. This will cause a loss of 6900V power to the 1A1 and 1B1 RCPs. RPS alarms will occur indicating that the Reactor should have tripped, but it will remain at power. The crew will attempt to manually trip the Reactor but it will not trip from the control room. Crew Response:

Recognize the Reactor should have tripped and attempt to manually trip the Reactor in accordance with OMP 1-18 Attachment A (< 3 RCPs operating with Reactor power > 2%). Since the Reactor will not trip from the control room, the OATC will initiate Rule 1 while performing Immediate Manual Actions (IMAs) of the EOP.

IMMEDIATE MANUAL ACTIONS 3.1 Depress REACTOR TRIP pushbutton [Reactor will NOT trip] 3.2 Verify reactor power < 5% FP and decreasing

RNO: GO TO Rule 1 (ATWS/Unanticipated Nuclear Power Production)

The BOP will perform a symptom check.










BOP will inform the SRO: • No symptoms to report except that Power Range NIs are > 5%, OATC is

performing Rule 1.

This event is complete when the crew transfers to Subsequent Actions, or as directed by the Lead Examiner.


Page 28 of 43


BOP/SRO

CT-1

UNPP Tab Crew response: Examiner Note: Rule 1 will direct the OATC to notify the CRS to go to the

UNPP tab UNPP Tab rev 0

1. Ensure Rule 1 (ATWS/Unanticipated Nuclear Power Production) is in progress or complete (page 32)

2. Verify Main FDW is operating and in AUTO

RNO: IF Main FDW is operating in MANUAL, THEN adjust Main FDW flow, as necessary to control RCS temperature

3. IAAT Main FDW is NOT operating, THEN: A. Trip the turbine-generator B. Start all available EFDW pumps C. Ensure Rule 3 (Loss of Main or Emergency FDW) is in progress or

complete

4. IAAT all power range NIs are < 5% FP, THEN perform Steps 5 - 6

RNO: GO TO Step 7

5. Depress turbine TRIP pushbutton [the turbine will NOT trip from P/B] 6. Verify all turbine stop valves closed

RNO: Place both EHC pumps in PULL TO LOCK

7. Verify any wide range NI > 1% FP

8. Open 1RC-4

9. Verify 1HP-5 open

10. Maximize letdown using 1HP-7 while maintaining letdown temperature < 120°F



Page 29 of 43


UNPP Tab Crew response: UNPP Tab

11. Verify Main FDW available

12. Adjust Main FDW flow as necessary to control RCS temperature

13. Verify overcooling in progress [Over cooling is NOT in progress] RNO: GO TO Step 16

16. Secure makeup to LDST

17. WHEN all wide range NIs are ≤ 1% FP, AND decreasing, THEN continue

18. Control RCS temperature as follows: __ Tave ≤ 555°F- Adjust SG pressure as necessary to stabilize RCS

temperature using either: • TBVs • Dispatch two operators to perform Encl 5.24 (Operation of the

ADVs) __ Tave > 555°F

• Utilize Rule 7 (SG Feed Control) to control SG feed rate as necessary to maintain cooldown rate within Tech Spec limits during the approach to the SG Level Control Point

19. Throttle HPI per Rule 6 (HPI) (page 41)

20. WHEN RCS pressure < 2300 psig, THEN continue

21. Verify PORV closed

22. Adjust letdown flow as desired

23. Verify RCP seal injection available

24. GO TO Subsequent Actions (page 30)



Page 30 of 43


SUBSEQUENT ACTIONS Tab Crew response: Subsequent Actions Tab rev 01

4.1 Verify all control rods in Groups 1 – 7 fully inserted

4.2 Verify Main FDW in operation

4.3 Verify either: __ Main FDW overfeeding causing excessive temperature decrease __ Main FDW underfeeding causing SG level decrease below setpoint

RNO: GO TO Step 4.5

4.5 IAAT Main FDW is operating, AND level in any SG is > 96% on the Operating Range, THEN perform Steps 4.6 - 4.8

RNO: GO TO Step 4.9

4.9 IAAT TBVs CANNOT control SG pressure at desired setpoint, AND TBVs NOT intentionally isolated, THEN manually control pressure in affected SGs using either: __ TBVs __ Dispatch two operators to perform Encl 5.24 (Operation of the ADVs)

4.10 Verify 1RIA-40 operable with CSAE OFF-GAS BLOWER operating

4.11 GO TO Step 4.14

4.14 Verify both are closed: __ 1MS-17 __ 1MS-26

4.15 Verify ES is required

RNO: 1. Initiate Encl 5.5 (Pzr and LDST Level Control) (Page 33) 2. GO TO Step 4.17

4.17 Open: __ PCB 20 __ PCB 21



Page 31 of 43


SUBSEQUENT ACTIONS Tab Crew response: Subsequent Actions Tab

4.18 Verify Generator Field Breaker open

4.19 Verify EXCITATION is OFF

4.20 Verify Aux Bldg and Turbine Bldg Instrument Air pressure ≥ 90 psig

4.21 Verify ICS/NNI power available

4.22 Verify all 4160V switchgear (1TC, 1TD & 1TE) energized

4.23 Verify both SGs > 550 psig

4.24 Verify Main FDW operating

4.25 Verify any RCP operating

4.26 Verify AP/0/A/1700/025 (SSF EOP) Encl (Unit 1 OATC Actions During Fire) in progress or complete

RNO: Ensure SGs approaching 25” – 35” [55” – 65” acc] S/U level

4.27 Place switches in CLOSE: __ 1FDW-31 __ 1FDW-40



Page 32 of 43


OATC

CT-2

RULE 1 Crew response: Rule 1

1. Verify any Power Range NI ≥ 5% FP

2. Initiate manual control rod insertion to the IN LIMIT

3. Verify Main FDW is feeding the SGs

4. Notify CRS to GO TO UNPP tab (Page 28)

5. Open: __ 1HP-24 __ 1HP-25

6. Ensure at least one operating: __ 1A HPI PUMP __ 1B HPI PUMP

7. Start 1C HPI PUMP

8. Open: __ 1HP-26 [1HP-26 will NOT open] __ 1HP-27

RNO: 1. IF 1HP-26 will NOT open, THEN open 1HP-410 2. IF at least two HPI pumps are operating,

AND 1HP-27 will NOT open THEN: A. Start the standby HPI pump B. Stop 1C HPI pump C. Open 1HP-409

9. Dispatch one operator without wearing Arc Flash PPE to open 600V CRD breakers: __ 1X9-5C (U-1 CRD Norm Fdr Bkr) (U1 Equipment Rm) __ 2X1-5B (U-1 CRD Alternate Fdr Bkr) (T-3/Dd-28)

Examiner Note: When the operator is dispatched to open CRD breakers, a 4 minute timer will be initiated to open the CRD breakers.

10. Verify only two HPI pumps operating

11. EXIT



Page 33 of 43











C. Notify SRO.



Page 34 of 43





__ GO TO Step 10.

7. Open: • 1HP-24 • 1HP-25



1LP-15 1LP-16 1LP-9 1LP-10 1LP-6 1LP-7



F. GO TO Step 8.




C. GO TO Step 9.


Page 35 of 43









Page 36 of 43




__ GO TO Step 12.





C. Open:

1LP-15

1LP-16

1LP-9

1LP-10

1LP-6

1LP-7








Page 37 of 43






GO TO Step 35.

16. Open: __ 1CC-7 __ 1CC-8








20. GO TO Step 23.



GO TO Step 23.

22. GO TO Step 35.


Page 38 of 43








25. Open: 1HP-1 1HP-2 1HP-3 1HP-4





28. __ Close 1HP-7.


2. Close: 1HP-8 1HP-9&11




Page 39 of 43





33. Open 1HP-6.










Page 40 of 43




__ GO TO Step 40.




C. Close 1CS-46.





__ GO TO Step 42.



• • • END • • •


Page 41 of 43

Rule 6

HPI

HPI Pump Throttling Limits • HPI must be throttled to prevent violating the RV-P/T limit. • HPI pump operation must be limited to two HPIPs when only one BWST suction valve

(1HP-24 or 1HP-25) is open. • HPI must be throttled ≤ 475 gpm/pump (including seal injection for A header) when only one

HPI pump is operating in a header. • Total HPI flow must be throttled ≤ 950 gpm including seal injection when 1A and 1B HPI pumps

are operating with 1HP-409 open. • Total HPI flow must be throttled < 750 gpm when all the following exist:

- LPI suction is from the RBES - piggyback is aligned - either of the following exist: • only one piggyback valve is open (1LP-15 or 1LP-16) • only one LPI pump operating • HPI may be throttled under the following conditions:

HPI Forced Cooling in Progress: HPI Forced Cooling NOT in Progress:

All the following conditions must exist:

• Core SCM > 0


All the following conditions must exist:

• All WR NIs ≤ 1%

• Core SCM > 0


• CRS concurrence required if throttling following emergency boration

HPI Pump Minimum Flow Limit • Maintain ≥ 170 gpm indicated/pump. This is an instrument error adjusted value that



Page 42 of 43

CRITICAL TASKS

CT-1 Lockout EHC Pumps to trip the Main Turbine prior to entry into PTS conditions (cooldown to < 400°F at > 100°F/hour)

CT-2 Prior to exiting the UNPP tab, take the Reactor subcritical (i.e. < 1% power on WR NIs)


SSF Operable: Yes KHU's Operable: U1 - OH, U2 - UG LCTs Operable: 2 Fuel Handling: No UNIT STATUS (CR SRO)










TS/SLC #

AMSAC/DSS 0300 7 Days SLC 16.7.2



• 1RIA-3 and 5 removed from RB • SASS is in Manual for calibration • Rx power ≈ 97% stable and ready to perform PT/1/A/0290/003 (Turbine Valve Movement) Encl

13.2 for CV3 & CV4 only. • PT/1/A/0290/003 Encl 13.2 (Control Valve Movement At Power) is in progress and complete up

to Step 2.5 to test CV3 & CV4 • Unit 1 will be maintaining 97% until Rx Engineering updates the maneuvering plan Secondary • AMSAC/DSS bypassed for calibration • Unit 2 is supplying the AS header • 1SSH-1, 1SSH-3, 1SD-2, 1SD-5, 1SD-140, 1SD-303, 1SD-355, 1SD-356 and 1SD-358 are








________________________ ________________________ OATC



• SASS is in Manual • AMSAC/DSS is bypassed • OP/1/A/1106/002B Encl 4.13 in progress to restore 1B Main FDW Pump from Handjack

Event No.


Description



1 N: BOP, SRO Restore 1B Main FDW Pump From Handjack

2 MPS090 C: OATC, SRO 1HP-120 Fails Closed

3 Override C: BOP, SRO 1D1 HDP Low Oil Level (Auto Power Decrease)

4 MCS004 I: OATC, SRO Controlling NR Tave Fails High

5 Override C: BOP, SRO (TS) 1A2 RCP High Oil Level

6 R: OATC, SRO Manual Power Decrease

7 MEL170 Override SRO (TS) CT-1 Lockout

8 MEL090 MEL180 MSS330

M: ALL `Switchyard Isolate

• KHU-2 Emergency Lockout • TD EFDW Pump fails to start in Auto



Page 2 of 56


Event 1: When the crew takes the shift, the SRO will direct the BOP to restore 1B Main FDW Pump from Handjack. Once proper operation of the FDW Pump Motor Gear Unit is verified, the 1B Main FDW Pump will be placed in automatic.

Event 2: 1HP-120 will fail closed which will cause RCS makeup to be lost. Pressurizer level will begin to decrease and Letdown Storage Tank level will begin to increase. Once the crew responds to the event and takes actions to stabilize RCS makeup and letdown, 1HP-120 will be repaired and the crew will place the valve back in automatic and restore normal operation.

Event 3: The 1D1 Heater Drain Pump (HDP) bearing temperatures will begin to increase due to low oil level. The OATC will perform a power decrease to ≤ 87% with ICS in automatic. The BOP will start 1A and 1B FDW Pump Seal Injection Pumps and Auxiliary Oil Pumps prior to securing the 1D1 HDP.

Event 4: Controlling Narrow Range Tave will fail high. FDW flow will increase and Control Rods will insert in an attempt to restore normal parameters. Once alarms are received, the crew will perform Plant Transient Response to stabilize the plant. The OATC will decrease Feedwater flow and the BOP will insert Control Rods to restore the heat balance. Once the plant is stable, the SRO will enter AP/1/A/1700/028 (ICS Instrument Failures) and ensure the appropriate ICS stations are in manual and the BOP will perform an instrument surveillance for the failed instrument.

Event 5: The crew will receive alarms indicating high oil level in the 1A2 RCP Lower Oil Pot. The SRO will enter AP/1/A/1700/016 (Abnormal RCP Operation). Once the 1A2 RCP is secured, the OATC will be required to re-ratio FDW in order to stabilize the plant.

Event 6 The OATC will perform a manual power decrease to ≤ 70% prior to the BOP securing the 1A2 RCP.

Event 7: The CT-1 Transformer (Startup Transformer) will lockout causing PCB-17 and PCB-18 to trip open. This will require the SRO to enter TS 3.8.1.

Event 8: A Switchyard Isolation will occur and seconds later the Reactor will trip causing a Blackout. The SRO will enter the Blackout Tab of the EOP. The Turbine Driven Emergency Feedwater Pump (TD EFDWP) will not automatically start which will require an operator to manually start the pump to restore heat removal from the core. Power will be restored in accordance with EOP Enclosure 5.38 (Restoration of Power) from a Keowee Hydro Unit.


Page 3 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 1 Page 1 of 2 Event Description: Restore 1B Main FDW Pump From Handjack (N: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP/SRO

OP/1/A/1106/002 B Encl 4.13 Crew response: SRO directs the BOP to restore 1B Main FDW Pump from Handjack per OP/1/A/1106/002 B Enclosure 4.13 OP/1/A/1106/002 B Encl 4.13 (Taking 1B FDWPT Off Handjack) rev 39

2.1 IF in Mode 1 OR Mode 2, perform the following:

2.1.1 WHILE enclosure is in progress monitor the following indications: • Appropriate ranged NIs • Neutron error • RCS Loop ΔT (curve for "Loop ΔT Vs Reactor Power" is in

PT/1/A/0600/001) • FDW Flow (curve for "Expected Feedwater Flow Per Header

Vs Reactor Power" is in OP/0/A/1108/001)

2.2 Remove "T/O SHEET" CR tag from 1B MAIN FDW PUMP (ICS) station

2.3 Run 1B MAIN FDW PUMP (ICS) station to "HSS" (high speed stop)

2.4 Perform the following: 2.4.1 Establish communication with Operator at 1B FDWPT 2.4.2 Run 1B MAIN FDW PUMP (ICS) to low speed stop 2.4.3 Run 1B MAIN FDW PUMP (ICS) to high speed stop

(≈ 1/8" from hard stop) 2.4.4 Verify Motor Gear Unit operated smoothly through entire

operation

Booth Cue: When asked, state that the MGU operated smoothly through the entire operation.

2.5 Turn 1B FDWP HANDJACK switch to "OFF"

This event is complete when the 1B Main FDW Pump is placed in Auto (Step 2.15), or as directed by the Lead Examiner.

NOTE: If both FDWPT(s) are in HAND changes in FDW valve ΔP can cause CTP to change

NOTE: Operator should locally verify Motor Gear Unit moves smoothly from low speed stop and back to high speed stop


Page 4 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 1 Page 2 of 2 Event Description: Restore 1B Main FDW Pump From Handjack (N: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP/SRO

OP/1/A/1106/002 B Encl 4.13 Crew response: OP/1/A/1106/002 B Encl 4.13

2.6 IF Unit 1 is in Mode 1 or 2 AND both FDWPT ICS stations are in "HAND" (N/A)

2.7

2.7 Decrease 1B MAIN FDW PUMP (ICS) until 1B FDWPT controlled by 1B MAIN FDW PUMP (ICS) station

2.8 Increase 1B FDWP Motor Speed Changer

2.9 Verify 1B FDWPT speed does NOT increase

2.10 Position 1B FDWP MOTOR SPEED CHANGER to 'FR' under "RAISE" until 1B FDWP MOTOR SPEED CHANGER is at "HSS"

2.11 After 1B FDWP MOTOR SPEED CHANGER reaches "HSS", hold 1B FDWP MOTOR SPEED CHANGER switch in 'FR' for 3 to 5 seconds to make all contacts

2.12 IF Unit 1 is in Mode 3: (N/A) 2.13 IF Unit 1 is in Mode 1 or 2 with 1A FDWPT shutdown: (N/A)

2.14 IF Unit 1 is in Mode 1 or 2 with 1A FDWPT operating but NOT in auto: (N/A)

2.15 IF Unit 1 is in Mode 1 or 2 with 1A FDWPT in auto: 2.15.1 Verify 1A MAIN FDW PUMP (ICS) in "AUTO" 2.15.2 Place 1B MAIN FDW PUMP (ICS) in "AUTO" 2.15.3 Verify ICS adjusts 1B FDWPT speed to balance suction flow

2.16 IF required, remove Turnover Sheet note for control of 1B FDWPT with Motor Speed Changer

This event is complete when the 1B Main FDW Pump is placed in Auto (Step 2.15), or as directed by the Lead Examiner.

NOTE: Changes in FDW valve ΔP will cause swings in CTP. Decreases in FDW valve ΔP will cause CTP to decrease

NOTE: • Motor Gear Unit control indicated by FDWPT speed and/or suction flow decreasing • Two successful decreases verifies control with Motor Gear Unit

NOTE: Changes in FDW valve ΔP will cause swings in CTP. Decreases in FDW valve ΔP will cause CTP to decrease


Page 5 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 2 Page 1 of 4 Event Description: 1HP-120 Fails Closed (C: OATC, SRO) Time Position Applicant’s Actions or Behavior

SRO

BOP

Plant Response:

• RCS makeup flow goes to ≈ 8 – 10 gpm (HPI Warming Flow) • PZR level begins to decrease • LDST level begins to increase • Valve position demand for 1HP-120 will increase to the 100% demand

value and valve position indication will indicate closed (green light) • 1SA-02/ B-1 (HP LETDOWN TANK LEVEL HIGH/LOW) will illuminate after

several minutes Crew Response: Examiner Note: If the 1HP-120 failure is recognized before the LDST

Statalarm actuates, the SRO will make a direct entry into AP/14 (next page). Otherwise the crew may perform the ARG and OP below and enter AP/14 when the SRO determines the entry conditions are met.

ARG 1SA-02/B-1

The crew may refer to ARG 1SA-02/B-1 (HP LETDOWN TANK LEVEL HIGH/LOW) and perform the required actions. rev 33

3.1 Instrument Failed: 3.1.1 Compare alternate channels to verify alarm validity:

• O1A1042 LDST LEVEL 1 • O1A1043 LDST LEVEL 2

3.2 Verify LDST pressure does not exceed LDST level/pressure operability requirement per OP/0/A/1108/001 (Curves and General Information).

3.3 IF High Level alarm is received: 3.3.1 Bleed as required by OP/1/A/1103/004 (Soluble Poison

Concentration Control).

This event is complete when Step 4.186 of AP/14 is complete, or as directed by the Lead Examiner.


Page 6 of 56


OATC/SRO

OP/1/A/1103/004 Crew Response:

Examiner Note: The crew may perform Enclosure 4.8 as necessary to reduce inventory. Enclosure 4.6 is performed to makeup/batch.

OP/1/A/1103/004 Enclosure 4.8 (Reducing RCS Inventory) rev 105

2.1 Verify HPI System operating

2.2 Ensure open 1CS-26 (LETDOWN TO RC BHUT)

2.3 Ensure open 1CS-41 (1A RC BHUT INLET)

2.4 Position 1HP-14 (LDST BYPASS) to “BLEED”

2.5 WHEN desired LDST level achieved, position 1HP-14 to “NORMAL” Examiner Note: The crew may initiate EOP Encl 5.5 for RCS inventory

control due LDST level increase (page 47)

AP/1/A/1700/014 AP/1/A/1700/014 (Loss of Normal Makeup and/or RCP Seal Injection) rev 18

3.1 IAAT RCP seal injection flow is lost, AND Component Cooling is lost, THEN perform the following: A. Trip the Rx B. Stop all RCPs C. Initiate AP/25 (SSF EOP)

3.2 IAAT loss of suction to operating HPI pumps is indicated: • Motor amps low or cycling • Discharge pressure low or cycling • Abnormal LDST level trend THEN GO TO Step 3.3

RNO: GO TO Step 4.7

4.7 Announce AP entry using PA System

4.8 Verify any HPI pump operating



Page 7 of 56


OATC/SRO

AP/1/A/1700/014 Crew Response: AP/1/A/1700/014

4.9 Verify RCP seal injection or HPI makeup line leak indicated by any of the following: • Report of line leak • Abnormal LDST level decrease • 1RIA-32 (AUX BLDG GAS) • 1RIA-45 (NORM VENT GAS) • RB RIAs in alarm • Abnormal RBNS level increase • Abnormal LAWT or HAWT level increase


4.11 Verify RCP seal injection flow exists to any RCP

4.12 Verify 1HP-120 operable in AUTO

RNO: 1. Attempt to operate 1HP-120 in HAND (will NOT operate) 2. IF 1HP-120 fails to operate, THEN GO TO Step 4.176

4.176 Perform the following as necessary to maintain Pzr level > 200” • Close 1HP-6 • Throttle 1HP-7 • Throttle 1HP-26

4.177 Place 1HP-120 in HAND and close

4.178 Notify SPOC to investigate and repair 1HP-120

4.179 WHEN 1HP-120 is repaired, THEN slowly re-establish flow through 1HP-120

Booth Cue: When directed by the Lead Examiner, FIRE TIMER 12 and

then notify the crew as FIN-24 that 1HP-120 has been repaired and ready to be placed in service.



Page 8 of 56


OATC/SRO

AP/1/A/1700/014 Crew Response:

4.180 Place 1HP-120 in AUTO

4.181 Close 1HP-26

4.182 Verify 1HP-122 (RC VOLUME CONTROL BYPASS) throttled


4.184 Verify 1HP-5 open

Examiner Note: 1HP-6 may be open or closed based on the actions taken in Step 4.176

4.185 Verify 1HP-6 open

RNO: 1. Throttle 1HP-7 for ≈ 20 gpm letdown flow 2. Open 1HP-6

4.186 Adjust 1HP-7 for desired letdown

4.187 WHEN conditions permit, THEN EXIT this procedure



Page 9 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 3 Page 1 of 5 Event Description: 1D1 HDP Low Oil Level (Auto Power Decrease) (C: BOP, SRO) Time Position Applicant’s Actions or Behavior

SRO

BOP/SRO

Plant response:

• OAC alarm HDP 1D1 Thrust Bearing Temp Hi • OAC alarm HDP 1D1 MTR Lower Bearing Temp Hi-Hi

Crew response: Refer to the OAC Alarm Response:

• Remove the pump from service per OP/1/A/1106/002 D (HDP Operation)

The SRO will enter AP/1/A/1700/029 to reduce Reactor power ≤ 87% to secure the 1D1 Heater Drain Pump (page 10)

Booth Cue: If an AO is sent to look at the HDP, report as an AO, that the lower oil sight glass of the 1D1 HDP is empty. The oil has been contained in the HDP sump and cleanup is in progress.

SRO should direct BOP to secure the 1D1 HDP once Reactor power has been reduced to ≤ 87%. FIN-24 (SPOC) should be notified to investigate and repair the oil leak.

OP/1/A/1106/002D The BOP should secure the 1D1 HDP per OP/1/A/1106/002 D Enclosure 4.4 (Shutdown of 1D1 HDP) rev 13

2.1 Verify CTP DEMAND ≤ 87%

3.1 Stop 1D1 HTR DRN PUMP

3.2 Verify 1HD-190 (1D1 Flash Tank Emer Level Control) opens by observing computer indication (O1D2092)

3.3 Verify 1HD-208 (1D1 HDP RECIRC) switch in "AUTO"

3.4 IF 1D1 HD pump breaker is no longer required, rack out 1TE-7 (1D1 Heater Drain Tank Pump Mtr)

Note: The SRO may determine that tagging the 1D1 HDP breaker is not

required at this time

Note: The SRO may refer to RP/0/A/1000/017 (Spill Response)

Booth Cue: If asked, Unit 2 will implement RP/0/A/1000/017 (Spill Response)

This event is complete when the 1D1 HDP has been secured, or when directed by the Lead Examiner.


Page 10 of 56


BOP/SRO

AP/1/A/1700/029 Crew response: AP/1/A/1700/029 (Rapid Unit Shutdown) rev 13

4.1 Initiate Encl 5.1 (Support Actions During Rapid Unit Shutdown) (page 12)


4.3 IAAT both of the following apply: __ It is desired to stop power decrease __ CTP > 18 % THEN perform Steps 4.4 - 4.7

RNO: GO TO Step 4.8


4.5 Deselect MAXIMUM RUNBACK

4.6 Initiate OP/1/A/1102/004 (Operation at Power) power reduction enclosure

4.7 WHEN conditions permit, THEN perform one of the following: __ Depress MAXIMUM RUNBACK to resume power reduction __ GO TO appropriate operating procedure for continued operation


NOTE Due to the power decrease initiated in this AP, the current plant configuration must be compared to the normal plant configuration in OP/1/A/1102/004 (Operation at Power) power reduction enclosure. Equivalent steps performed by this AP should be signed off as intent met. Any steps NOT performed by this AP must be evaluated in preparation for power increase or continued shutdown.


Page 11 of 56


OATC/SRO

SRO



4.9 Depress MAXIMUM RUNBACK

4.10 Verify both Main FDW pumps running

4.11 Adjust bias for first Main FDW pump desired to be shutdown until suction flow is ≈ 1 x 106 lbm/hr less than remaining Main FDW pump suction flow

4.12 WHEN core thermal power is < 65% FP, THEN continue

Examiner Note: If Reactor power is reduced below 85% during this event, TS 3.10.1 will apply.

TS 3.10.1 STANDBY SHUTDOWN FACILITY Conditions A-E are applicable due to Reactor power being < 85% which

makes the SSF inoperable Required Action: Restore SSF systems to OPERABLE status

(7 days)


NOTE • 1B Main FDW Pump is the preferred pump to be shutdown first • To lower 1B Main FDW Pump suction flow, bias is adjusted counter-

clockwise • To lower 1A Main FDW Pump suction flow, bias is adjusted clockwise


Page 12 of 56


BOP/SRO

AP/1/A/1700/029 Enclosure 5.1 Crew response: AP/29 Enclosure 5.1 (Support Actions During Rapid Unit Shutdown)

1. Notify WCC SRO to initiate Encl 5.2 (WCC SRO Support During Rapid Unit Shutdown)

2. Start the following pumps: __ 1A FDWP SEAL INJECTION PUMP __ 1A FDWP AUXILIARY OIL PUMP __ 1B FDWP AUXILIARY OIL PUMP __ 1B FDWP SEAL INJECTION PUMP

3. WHEN CTP is ≤ 80%, THEN continue





8. Verify Turbine-Generator shutdown is required

9. Place the following transfer switches to MAN: __ 1TA AUTO/MAN __ 1TB AUTO/MAN

10. Close 1TA SU 6.9 KV FDR

11. Verify 1TA NORMAL 6.9 KV FDR opens

12. Close 1TB SU 6.9 KV FDR

13. Verify 1TB NORMAL 6.9 KV FDR opens



Page 13 of 56


AP/1/A/1700/029 Enclosure 5.1 Crew response: AP/29 Enclosure 5.1 (Support Actions During Rapid Unit Shutdown)

14. Place the following transfer switches to MAN: __ MFB1 AUTO/MAN __ MFB2 AUTO/MAN





19. Notify CR SRO that Unit auxiliaries have been transferred

20. IAAT 1SSH-9 is NOT closed, AND CTP is ≤ 75%, THEN throttle 1SSH-9 to Maintain Steam Seal Header pressure 2.5 - 4.5 psig

21. WHEN CTP ≤ 65%, THEN place the following in MANUAL and close: __ 1FDW-53 __ 1FDW-65

22. IAAT load is ≤ 550 MWe, THEN perform Steps 23 - 24



Page 14 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 4 Page 1 of 4 Event Description: Controlling NR Tave Fails High (I: OATC, SRO) Time Position Applicant’s Actions or Behavior

OATC

CT-1

BOP

SRO

Plant response:

• Controlling NR Tave digital display reads ≈ 586°F • Controlling Tave Chessell display reads ≈ 586°F • 1SA-2/A-12 (ICS Tracking) • 1SA-2/C-11 (ICS Loss of OAC CTP Signal) • Control Rods will insert and FDW flow will increase • RCS pressure will decrease

Examiner Note: If Reactor power decreases below 85% during this event, the SRO will enter Tech Spec 3.10.1 Conditions A-E which all have a 7 day completion time (SSF OOS).

Crew response: When the Statalarms are received, the crew should perform Plant Transient Response (PTR) to stabilize the plant

The OATC should: • Communicate to the CRS the initial alarm (if applicable) followed by reactor

power level and direction • Place the appropriate ICS stations in manual (Diamond and both FDW

Masters in this case) in manual if any of the following occur: o NI power increasing above the pre-transient power level o Failed instrument is diagnosed o Invalid input exists and the CRS directs the ICS be placed in manual

• Remain focused on reactor power level and FDW response during the transient

Examiner Note: The OATC will have to reduce FDW in order to stabilize power below the pre-transient level.

The BOP should: • Determine if a valid ICS runback exists and inform the CRS • Monitor plant response and verify operating limits NOT exceeded • If ICS is placed in Manual, remain focused on RCS pressure, SG outlet

pressure and RCS inventory The SRO should: • Refer to AP/28 (ICS Instrument Failures) (page 15) • Ensure FIN-24 (SPOC) is contacted to repair the failed instrument

This event is complete when the SRO reaches Step 6 of Section 4A, or as directed by the Lead Examiner.


Page 15 of 56


OATC/SRO


4.1 Provide control bands as required (OMP 1-18 Attachment I) OMP 1-18 Attachment I: 1. Plant Conditions Stable or TPB ≤ Pre-transient Conditions

1.1 The following bands are to be established during manual control of plant conditions (as needed) but may be adjusted by the CRS if required 1.1.1 NI Power ± 1% not to exceed the pre-transient or allowable

power. If at the pre-transient or allowable level, band is NI Power – 1%

1.1.2 Current Tave ± 2°F 1.1.3 Current SG Outlet Pressure ± 10 PSIG (N/A) 1.1.4 Delta Tc 0oF ± 2°F

4.2 Initiate notification of the following: __ OSM to reference the following:

• OMP 1-14 (Notifications) • Emergency Plan

__ STA

4.3 Verify a power transient ≥ 5% has occurred

RNO: GO TO Step 4.5.

4.4 Notify Rx Engineering and discuss the need for a maneuvering plan

4.5 Use the following, as necessary, to determine the applicable section from table in Step 4.6: • OAC alarm video • OAC display points • Control Board indications • SPOC assistance, as needed



Page 16 of 56


OATC/SRO

AP/1/A/1700/028 Crew response: AP/1/A/1700/028 (ICS Instrument Failures)

4.6 GO TO the applicable section per the following table:

√ Section Failure

4A RCS Temperature

AP/1/A/1700/028 Section 4A AP/1/A/1700/028 Section 4A (RCS Temperature Failure)

NOTE • If Tave instrument circuit failed high, the following may have occurred

depending on initial ICS station status: • Unit to TRACK due to Rx Cross Limits • Control Rod insertion • Feedwater flow increase

• If Tave instrument circuit failed low, the following may have occurred depending on initial ICS station status: • Unit to TRACK due to Rx Cross Limits • Control Rod withdrawal • Feedwater flow decrease • Feedwater re-ratio

1. Ensure the following in HAND: __ 1A FDW MASTER __ 1B FDW MASTER

2. Ensure DIAMOND in MANUAL



Page 17 of 56


OATC/SRO

AP/1/A/1700/028 Section 4A Crew response:

3. Notify SPOC to perform the following: __ Select a valid RCS Tave and Delta Tc input to ICS per

AM/1/A/0326/020 (Control of Unit 1 Star Module Signal Selection Function).

__ Investigate and repair the failed RCS temperature instrumentation

4. PERFORM an instrumentation surveillance using applicable table in Encl 5.2 (ICS Instrument Surveillances) for the failed instrument

Examiner Note: This step will require the BOP to refer to Table 1 (RCS Temperature) of Enclosure 5.2 (next page) and determine if the surveillance for SR 3.4.1.2 can be met as it is written with the failed temperature instrument. Based on plant response to the failure the BOP will determine either: 1. The surveillance is not required to be met due to plant

conditions (not steady state) 2. The surveillance can be met as written 3. The surveillance can NOT be met as written This information will be reported to the SRO when asked in Step 5.

5. Verify instrumentation surveillance in Encl. 5.2 (ICS Instrument Surveillances) was performed satisfactorily as written

6. WHEN notified by SPOC that a valid RCS Tave and Delta Tc input have been restored to ICS, THEN GO TO OP/1/A/1102/004 A Encl. (Placing ICS Stations To Auto)



Page 18 of 56


Page 19 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 5 Page 1 of 10 Event Description: 1A2 RCP High Oil Level (C: BOP, SRO) Time Position Applicant’s Actions or Behavior

BOP/SRO

Plant response:

• OAC alarm RCP 1A2 MTR LOWER OIL POT LEVEL

Crew response: The BOP will refer to OAC alarm response guide for 1A2 RCP high oil level

1. If RCP operating and RCP was not started in the past 6 hours, GO TO AP/1/A/1700/016 (Abnormal Reactor Coolant Pump Operation)

AP/1/A/1700/016

AP/1/A/1700/016 (Abnormal RCP Operation) rev 34

4.1 IAAT any RCP meets immediate trip criteria of Encl 5.1 (RCP Immediate Trip Criteria), THEN perform Steps 4.2 - 4.11


Examiner Note: In step 4.12, the crew may conservatively decide to secure the RCP because the oil pot level is threatening to go off-scale high. If so, they will proceed to step 4.13 (page 23) and will not end up in Section 4C. Either procedure path is acceptable.

4.12 IAAT either of the following apply: __ Any RCP approaching immediate trip criteria of Encl 5.1 (RCP

Immediate Trip Criteria) __ There is an immediate need to stop a RCP at this time THEN perform Steps 4.13 - 4.15 (page 23)



4.17 Notify OSM to request evaluation by RCP Component Engineer

This event is complete when the 1A2 RCP has been secured and ΔTc is 0 ± 2°F, or as directed by the Lead Examiner.


Page 20 of 56


BOP/SRO


4.18 IAAT the failure is identified, THEN GO TO the applicable section per the following table:

√ Section Failure

4A Seal Failure

4B Abnormal Vibration

4C High or Low Oil Pot Level 4D Loss of Seal Return

4E Abnormal RCP Temperatures

AP/1/A/1700/016 Section 4C AP/1/A/1700/016 Section 4C (High or Low Oil Pot Level)

1. IAAT any RCP meets immediate trip criteria of Encl 5.1 (RCP Immediate Trip Criteria), THEN perform Steps 2 - 11

RNO: GO TO Step 12

12 Start trending RCP oil pot levels (Turn-on Code "unitpump"RCPT3, example 1A2RCPT3)

NOTE • RCP oil pot level indication range is +1.5 to -1.5 inches • RCP motor oil pot temperatures and RCP motor guide bearing temperatures

may be used to validate low oil pot level

13 IAAT oil pot level threatens to go off scale high or low for an operating RCP, THEN perform Steps 14 - 24

14 Verify MODE 1 or 2

15 Verify three RCPs will remain operating after affected RCP is tripped



Page 21 of 56


BOP/SRO

AP/1/A/1700/016 Section 4C Crew response: AP/1/A/1700/016 Section 4C

16. Verify Rx power is ≤ 70% as indicated on all NIs

RNO: 1. Direct an RO to initiate Encl 5.2 (Rapid Power Reduction) (page 29) 2. WHEN Rx power is ≤ 70% on all NIs,

THEN continue this procedure

17. Verify any SG on Low Level Limits

RNO: GO TO Step 20

20. Verify FDW Masters in Auto

RNO: 1. Stop the affected RCP 2. Manually adjust FDW masters to achieve desired ΔTc 3. GO TO Step 23

23. Initiate Encl 4.3 (Special Instructions for < 4 RCP Operation) of OP/1/A/1102/004 (Operation at Power) (page 26)

24. Make the following notifications: __ Notify OSM to make required notifications of OMP 1-14

(Notifications) __ Notify Rx Engineering and request a power maneuver plan, if

needed __ Notify SOC if load reduction was required __ Notify Chemistry to take RCS boron samples on a 1 hour frequency


CAUTION Total feedwater flow should be maintained constant to prevent changes in core reactivity


Page 22 of 56


BOP/SRO

AP/1/A/1700/016 Section 4C Crew response: AP/1/A/1700/016 Section 4C

25 IAAT an RB fire exists, THEN perform Steps 26 - 29

RNO: GO TO Step 30

30 IAAT either of the following conditions is met: __ a RCP with low oil level has been shut down for ≥ 3 hours

__ a RCP with high oil level has been shut down THEN close the associated RCP motor cooler inlet/outlet valve: __ 1LPSW-7&8 (1A1 RCP) __ 1LPSW-9&10 (1B1 RCP) __ 1LPSW-13&14 (1A2 RCP) __ 1LPSW-11&12 (1B2 RCP)

31 IAAT a RCP has been tripped due to exceeding Immediate Trip Criteria on a RCP motor, THEN contact RCP engineer prior to restart

32 IAAT both are met: __ There has been a failure of the DELTA Tc controller __ The DELTA Tc controller has been repaired THEN initiate OP/1/A/1102/004 A Encl (Placing ICS Stations To Auto)

33 WHEN conditions permit, THEN EXIT this procedure.



Page 23 of 56


BOP/SRO

AP/1/A/1700/016

Alternate path from Step 4.12

Crew response: Examiner Note: If the crew determines that It is desired to secure the

1A2 RCP in Step 4.12 due to oil levels going off-scale high, below lists the procedure path to secure the 1A2 RCP.

4.12 IAAT either of the following apply: __ Any RCP approaching immediate trip criteria of Encl 5.1 (RCP

Immediate Trip Criteria) __ There is an immediate need to stop a RCP at this time THEN perform Steps 4.13 - 4.15

4.13 Verify Rx Power > 70%

4.14 Initiate Encl 5.2 (Rapid Power Reduction) (page 29)

4.15 WHEN Rx Power is ≤ 70%, THEN GO TO Step 4.2

4.2 Verify MODE 1 or 2

4.3 Verify Rx power is ≤ 70% as indicated on all NIs

4.4 Verify three RCPs will remain operating after affected RCP is tripped

4.5 Verify any SG on Low Level Limits

RNO: GO TO Step 4.8

4.8 Verify FDW masters in Auto

RNO: 1. Stop the affected RCP 2. Manually adjust FDW masters to achieve desired ΔTc 3. GO TO Step 4.29


CAUTION Total feedwater flow should be maintained constant to prevent changes in core reactivity


Page 24 of 56


BOP/SRO

AP/1/A/1700/016


Crew response: AP/1/A/1700/016

4.29 IAAT any of the following indicate external RCP seal leakage: __ RB RIAs increasing or in alarm __ RCS Tave constant with LDST level decreasing more than normal __ Quench Tank level rate increasing __ RB Normal Sump rate increasing __ Visual confirmation THEN initiate AP/02 (Excessive RCS Leakage)

4.30 Initiate Encl 4.3 (Special Instructions for < 4 RCP Operation) of OP/1/A/1102/004 (Operation at Power)

4.31 IAAT either of the following conditions is met: __ a RCP has been shut down for ≥ 3 hours __ a RCP with high oil level has been shut down THEN close the associated RCP motor cooler inlet/outlet valve: __ 1LPSW-7&8 (1A1 RCP) __ 1LPSW-9&10 (1B1 RCP) __ 1LPSW-13&14 (1A2 RCP) __ 1LPSW-11&12 (1B2 RCP)

4.32 IAAT either of the following has exceeded 260°F including transient situations: __ O1A1253 - O1A1256 (RCP UPPER SEAL HOUSING TEMP) __ O1A1910 - O1A1913 (RCP SEAL RETURN TEMP) THEN closely monitor seal parameters for degradation until an Engineering evaluation is completed due to potential for seal ring and elastomer damage



Page 25 of 56


BOP/SRO

AP/1/A/1700/016


Crew response:

4.33 Verify 1RIA 57 or 1RIA 58 have increased


4.35 IAAT a RCP has been tripped due to exceeding Immediate Trip Criteria on a RCP motor, THEN contact RCP engineer prior to restart

4.36 IAAT both are met: __ There has been a failure of the DELTA Tc controller __ The DELTA Tc controller has been repaired THEN initiate OP/1/A/1102/004 A Encl (Placing ICS Stations To Auto)

4.37 Verify any RCP that was shut down had a high vibration alarm




NOTE Operating experience has shown that failure of RC Pump components located internal to the RCS can create loose debris which can lead to fuel clad failures. These type RC Pump failures may cause Loose Parts Monitor alarms immediately and increased RCS radioactivity later.


Page 26 of 56


BOP/SRO

OP/1/A/1102/004 Encl 4.3 Crew response: OP/1/A/1102/004 (Operations At Power) Encl 4.3 (Special Instructions For < 4 RCP Operations) Rev 148 2.1 IF conditions permit, log the current quadrant power tilt and the position of

the ΔTc controller prior to securing a RCP during power operations

2.2 Perform one of the following: (Continue) • Ensure TS 3.4.4 Condition A entered when fourth RCP secured • Verify the following:

__ O1E4021 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4022 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4023 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4024 (1A RPS Var Flux Trip Value) set at 79.75%

2.3 Using turn-on code T6 3RCP, digitally trend the following data at one minute intervals:

Point ID Description

O1P0889 CORE THERMAL POWER BEST O1P0877 INCORE IMBALANCE O1E3335 API GROUP AVE FOR GROUP 7 O1E3336 API GROUP AVE FOR GROUP 8 O1P0737 INCORE TILT QUADRANT W-X O1P0738 INCORE TILT QUADRANT X-Y O1P0739 INCORE TILT QUADRANT Y-Z O1P0740 INCORE TILT QUADRANT Z-W O1I0828 RC COLD LEG A1 TEMP O1I0829 RC COLD LEG A2 TEMP O1I0830 RC COLD LEG B1 TEMP O1I0831 RC COLD LEG B2 TEMP


NOTE: • Instructions for performing OAC trends are located in Working With Trends enclosure of OP/0/A/1103/020 A (Operator Aid Computer Use)

• Only the first 6 points will be displayed initially; press "Page Down" key to see second 6 points


Page 27 of 56


BOP/SRO

OP/1/A/1102/004 Encl 4.3 Crew response:

NOTE: The 100% Power Imbalance curves also apply for runs at reduced power.

2.4 Maintain Control Rod position and Power Imbalance within COLR limits

2.5 IF NI calibration NOT within requirements of Limit and Precaution Step 2.2.6, calibrate NIs to Thermal Power Best

2.6 Perform the following per AM/1/A/0315/017 (TXS RPS Channels A, B, C, And D Parameter Changes For Abnormal/Normal Operating Conditions):

2.6.1 Notify I&E to reset RPS Overpower High Trip Setpoint for three RCP Operation ________________________ _____ Person Notified Date

2.6.2 IF AT ANY TIME Quadrant Power Tilt problems exist, notify I&E to Adjust Flux/Imbalance/Flow trip setpoints as required to comply with TS 3.2.3 ________________________ _____ Person Notified Date

NOTE: The Maximum Allowed Power Setpoint (Pmax) is reduced when

operating for extended periods when only three RCPs operating as a conservative action.

2.6.3 IF expected to operate for an extended period of time with only 3 RCPs operating, notify I&E to adjust Flux/Imbalance /Flow trip setpoints for 3 RCP operation ________________________ _____ Person Notified Date

2.7 IF AT ANY TIME notified by I&E that RPS Overpower High Trip Setpoint adjusted for three RCP Operation, perform the following:

2.7.1 Verify the following: __ O1E4021 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4022 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4023 (1A RPS Var Flux Trip Value) set at 79.75% __ O1E4024 (1A RPS Var Flux Trip Value) set at 79.75%

2.7.2 Evaluate exiting TS 3.4.4 condition A



Page 28 of 56


BOP/SRO

OP/1/A/1102/004 Encl 4.3 Crew response: 2.8 Initiate review of PT/1/A/0600/001 (Periodic Instrument Surveillance) to

determine if any limits approached

NOTE: • Operations Management/Reactor Engineering Group should be consulted for value to use for high flux alarm setpoint.

• Instructions for Adjusting Alarm Setpoints On The NI Recorder are in OP/0/A/1108/001 (Curves And General Information).

2.9 Adjust high flux alarm setpoint per Operations Management/Reactor Engineering Group recommendations. (Alarm setpoint is adjusted on the NI Recorder)

NOTE: 'D' bleed pressure may NOT be high enough to run the FDWP turbines.

2.10 Maintain Auxiliary Steam available to the FDWP turbines.

2.11 IF 1SSH-9 (SSH DISCH CTRL BYPASS) is being used to control Steam Seal Header pressure, throttle 1SSH-9 as required to maintain desired SSH pressure

NOTE: RCS pressure decrease in the loop with two RCPs running is

expected. This may cause acceptance criteria of PT/1/A/0600/001 (Periodic Instrument Surveillance) NOT to be met.

2.12 Place note on CR turnover sheet indicating the following:

"Be aware of the effect of the indicated pressure on the margin to trip setpoint for the Reactor Protective System trips associated with RCS pressure"



Page 29 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 6 Page 1 of 1 Event Description: Manual Power Decrease (R: OATC, SRO) Time Position Applicant’s Actions or Behavior

OATC/SRO

AP/1/A/1700/016 Encl 5.2 Crew response: The OATC will perform AP/1/A/1700/016 Encl 5.2 (Rapid Power Reduction) to reduce Reactor power < 70% to secure the 1A2 RCP AP/1/A/1700/016 Enclosure 5.2 (Rapid Power Reduction)

1. Verify ICS in AUTO RNO: 1. Notify SRO to provide target power level, average power reduction

rate, and control bands, as required 2. Initiate manual power reduction to ≤ 70% 3. WHEN Rx Power ≤ 70% as indicated by all NIs,

THEN notify CR SRO 4. Stop the following pumps: __ 1E1 HTR DRN PUMP __ 1E2 HTR DRN PUMP 5. GO TO Step 7

7. Verify Rx Power was reduced ≥ 15% within a 1 hour period

8. Notify Primary Chemistry to perform Tech Spec SR 3.4.11.2 as required


This event is complete when Rx power is < 70% on all NIs, or as directed by the Lead Examiner.

NOTE • This enclosure should be performed by an RO • The step to verify ICS in AUTO means that the ICS is capable of

responding to a MAXIMUM RUNBACK signal


Page 30 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 7 Page 1 of 1 Event Description: CT-1 Lockout (SRO) (TS) Time Position Applicant’s Actions or Behavior

BOP

SRO

Plant response:

• 1SA-15/C-2 (EL SU SOURCE VOLT MONIT LOGIC UNDERVOLTAGE) • SA-3/B-4 (Oconee White Startup Transformer CT1 Tie PCB 17) • SA-3/C-4 (Startup Transformer CT1 PCB 18) • CT-1 transformer will be de-energized • PCB-17 and PCB-18 will trip open

Crew response: The BOP will refer to Statalarm 1SA-15/C-2

3.1 IF the start-up source was carrying the unit load AND the start-up breakers tripped, verify that power is supplied from the standby source

3.2 Refer to EP/1/A/1800/001 (Emergency Operating Procedure) or AP/1/A/1700/011 (Recovery from Loss of Power) if required

3.3 IF only one phase trips, perform the following: (N/A) 3.4 Refer to TS 3.8.1 3.5 Return to the normal (preferable) or start-up source of power as soon as

conditions permit Examiner Cue: If the crew trips the Reactor, Event 8 will auto actuate to

initiate a Switchyard Isolation. The SRO will refer to Tech Specs

TS 3.8.1 AC SOURCES – OPERATING Condition A (1 hour) Perform SR 3.8.1.3 AND (12 hours) Align the emergency startup bus to share another unit’s

startup transformer AND (36 hours) Restore transformer to OPERABLE status and normal

startup bus alignment OR (36 hours) Designate one unit sharing the startup transformer to be

shutdown

This event is complete when the SRO has determined Tech Specs, or as directed by the Lead Examiner.


Page 31 of 56

Op-Test No.: ILT16-2 Scenario No.: 3 Event No.: 8 Page 1 of 16 Event Description: Switchyard Isolate (M: ALL) Time Position Applicant’s Actions or Behavior

SRO

OATC

SRO

BOP

Plant response: • Switchyard Isolation occurs • RCS Pressure increases • Reactor will trip on RCS High Pressure (2345 psig) • When the Reactor trips, a loss of power will occur due to KHU-2 lockout and

CT-1 lockout Crew response: • Crew may manually trip the reactor prior to reaching the RCS High

Pressure setpoint • SRO will direct the OATC to perform IMAs and the BOP to perform a

symptom check

IMMEDIATE MANUAL ACTIONS The SRO will direct the OATC to perform IMAs

3.1 Depress REACTOR TRIP pushbutton 3.2 Verify reactor power < 5% FP and decreasing 3.3 Depress the turbine TRIP pushbutton 3.4 Verify all turbine stop valves closed 3.5 Verify RCP seal injection available

RNO: IF CC is unavailable, THEN immediately: A. Stop all RCPs B. Notify CR SRO to initiate AP/25

SYMPTOM CHECK The SRO will direct the BOP to perform a Symptoms Check.








CSAE Off-gas alarms Process monitor alarms (RIA-40, 59,60), Area monitor alarms (RIA-16/17)


This event is complete when the SRO transfers to the Subsequent Actions tab, or as directed by the Lead Examiner.


Page 32 of 56


Crew response: The SRO will direct the BOP to initiate AP/25

Examiner Note: When the BOP attempts to proceed to the SSF, inform him/her that a Unit 2 RO will perform AP/25 actions

BLACKOUT TAB

Blackout Tab rev 02

Examiner Note: When the Unit 2 CRS is notified to perform Unit 2 EOP Encl 5.42 (PSW Power and Pump Alignment), a booth instructor will enter the simulator to perform the enclosure. Performing this enclosure will have no effect on the outcome of the scenario.

1. Ensure Rule 3 (Loss of Main or Emergency FDW) is in progress or complete

2. Verify TDEFDWP feeding SGs

3. Direct RO to perform Encl 5.45 (PSW Feed and RCP Seals) to establish RCP seals (page 43)

4. Verify two ROs available to perform Control Room actions

5. Ensure Encl 5.45 (PSW Feed and RCP Seals) in progress or complete

6. Notify one RO to perform Encl 5.38 (Restoration of Power) (page 41)

7. Verify any: __ Any SG is being fed from Emergency FDW or PSW __ SSF or PSW is available to feed SGs

8. Verify both: __ Blackout exists on all three units __ PSW Power NOT available

RNO: GO TO Step 10



Page 33 of 56


BLACKOUT TAB Crew response:

10. Verify RCS temperature ≥ 550°F

RNO: 1. Feed and steam available SGs, as necessary, to stabilize RCS P/T using one of the following:

__ TBVs __ Dispatch two operators to perform Encl 5.24 (Operation of the

ADVs) to prepare to steam intact SGs 2. GO TO Step 13

11. Verify feeding SGs with TDEFDW pump 12. Feed available SGs as necessary to stabilize RCS P/T in bands

prescribed by Rule 7 (SG Feed Control)

13. IAAT NO SGs are being fed, AND any source of EFDW (Unit 1 or another unit) becomes available, THEN perform Steps 14 - 18

RNO: GO TO Step 19

19. IAAT the SSF or PSW is available to feed SGs, AND EFDW from any source is insufficient to maintain stable RCS P/T, THEN perform one of the following:

__ Notify SSF operator that feeding SGs with SSF ASW is required (SSF CR x-2766)

__ Initiate Encl 5.45 (PSW Feed and RCP Seals) and feed SGs with PSW


NOTE • Feeding SGs with EFDW is desired above HPI Forced Cooling. If a feed

source becomes available, step 13 should be performed prior to re-performing Rule 3

• 100 gpm could cause overcooling if adequate decay heat levels do NOT exist

NOTE • The EFW system operation is expected to last 2 - 4 hours during a three

unit blackout without PSW power • If battery power is lost and PSW power is not available, EFDW control will

be lost


Page 34 of 56


BLACKOUT TAB Crew response:

20. IAAT power is restored to any of the following: __ 1TC __ 1TD __ 1TE THEN GO TO Step 21

21. Ensure any SG is being fed or action is being taken per Step 13

22. Verify SSF activated

23. Communicate status of SG feed and seal injection to SSF operator using x-2766, radio, or plant page

24. Initiate AP/11 (Recovery from Loss of Power) (page 37)

25. GO TO Subsequent Actions tab (page 35)



Page 35 of 56


SUBSEQUENT ACTIONS TAB Crew response: Subsequent Actions Tab

4.1 Verify all control rods in Groups 1 – 7 fully inserted

4.2 Verify Main FDW in operation

4.3 Verify either: __ Main FDW overfeeding causing excessive temperature decrease __ Main FDW underfeeding causing SG level decrease below setpoint

RNO: GO TO Step 4.5

4.5 IAAT Main FDW is operating, AND level in any SG is > 96% on the Operating Range, THEN perform Steps 4.6 - 4.8

RNO: GO TO Step 4.9

4.9 IAAT TBVs CANNOT control SG pressure at desired setpoint, AND TBVs NOT intentionally isolated, THEN manually control pressure in affected SGs using either: __ TBVs __ Dispatch two operators to perform Encl 5.24 (Operation of the ADVs)

4.10 Verify 1RIA-40 operable with CSAE OFF-GAS BLOWER operating


4.14 Verify both are closed: __ 1MS-17 __ 1MS-26

4.15 Verify ES is required

RNO: 1. Initiate Encl 5.5 (Pzr and LDST Level Control) (page 47) 2. GO TO Step 4.17

4.17 Open: __ PCB 20 __ PCB 21



Page 36 of 56


SUBSEQUENT ACTIONS TAB Crew response: Subsequent Actions Tab

4.18 Verify Generator Field Breaker open

4.19 Verify EXCITATION is OFF

4.20 Verify Aux Bldg and Turbine Bldg Instrument Air pressure ≥ 90 psig

4.21 Verify ICS/NNI power available

4.22 Verify all 4160V switchgear (1TC, 1TD & 1TE) energized

4.23 Verify both SGs > 550 psig

4.24 Verify Main FDW operating

4.25 Verify any RCP operating

4.26 Verify AP/0/A/1700/025 (SSF EOP) Encl (Unit 1 OATC Actions During Fire) in progress or complete

RNO: Ensure SGs approaching 25” – 35” [55” – 65” acc] S/U level

4.27 Place switches in CLOSE: __ 1FDW-31 __ 1FDW-40



Page 37 of 56


AP/1/A/1700/011 Crew response: AP/1/A/1700/011 (Recovery From Loss of Power) rev 55

4.1 Announce AP entry using OMP 1-18 placard

4.2 IAAT all exist: __ 1KI energized __ Pzr level > 80″ [180″ acc] __ Pzr heaters are desired THEN ensure Pzr heaters in AUTO

4.3 Verify load shed of inverters was performed per Unit 1 EOP Encl (Load Shed of Inverters During SBO)

RNO: GO TO Step 4.9

4.9 Verify load shed has initiated as indicated by either of the following statalarms on: __ 1SA-15/D-4 (EL LOAD SHED CHNL A LOGIC INITIATE) __ 1SA-14/D-4 (EL LOAD SHED CHNL B LOGIC INITIATE)

4.10 Verify load shed is complete as indicated by LOAD SHED COMPLETE on any ES Channel (Channel 1 or 2)

4.11 Verify breakers closed: __ 1TC INCOMING FDR BUS 1 __ 1TC INCOMING FDR BUS 2 __ 1TD INCOMING FDR BUS 1 __ 1TD INCOMING FDR BUS 2 __ 1TE INCOMING FDR BUS 1 __ 1TE INCOMING FDR BUS 2

4.12 Verify 1SA-15/E-6 (EL SWYD ISOLATION CONFIRMED CHNL A LOGIC) is OFF

RNO: GO TO Step 4.15 4.15 Verify any Oconee unit receiving power form its normal source

(1T, 2T, 3T)



Page 38 of 56


AP/1/A/1700/011 Crew response: AP/1/A/1700/011 (Recovery From Loss of Power)

4.15 Verify any Oconee unit receiving power form its normal source (1T, 2T, 3T)


4.17 Verify load shed was initiated as indicated by either of the following statalarms on: __ 1SA-15/D-4 (EL LOAD SHED CHNL A LOGIC INITIATE) __ 1SA-14/D-4 (EL LOAD SHED CHNL B LOGIC INITIATE)

4.18 Verify ES has occurred


4.20 Simultaneously press RESET on both of the following pushbuttons to reset Main Feeder Bus Monitor Panel Load Shed Circuitry: • MFB UNDERVOLTAGE CHANNEL 1 RESET • MFB UNDERVOLTAGE CHANNEL 2 RESET

4.21 Verify load shed signal reset as indicated by both of the following statalarms off: __ 1SA-15/D-4 (EL LOAD SHED CHNL A LOGIC INITIATE) __ 1SA-14/D-4 (EL LOAD SHED CHNL B LOGIC INITIATE)

4.22 IAAT electrical loads are added, AND either MFB is energized by:

• CT-4 • CT-5 • Backcharged 1T THEN ensure transformer is within limits of the applicable enclosure:

√ Source Encl CT-4 5.1A

CT-5 5.1B

Transformer 1T 5.1C


CAUTION Normal plant loads can overload the Auxiliary Transformer, CT-4, or CT-5


Page 39 of 56


CT-2

RULE 3 Crew response: EOP Rule 3

1. Verify loss of MFDW and/or EFDW was due to any of the following: __ Turbine Building Flooding __ Actions taken to increase SG level due to Turbine Building Flooding

RNO: GO TO Step 3

3. IAAT NO SGs can be fed with FDW (Main/CBP/Emergency/PSW), AND any of the following exist:

__ RCS pressure reaches 2300 psig OR NDT limit __ Pzr level reaches 375” [340” acc] THEN PERFORM Rule 4 (Initiation of HPI Forced Cooling)



6. GO TO Step 38


RNO: GO TO Step 44

44. Verify any SCM ≤ 0°F

RNO: IF overcooling, OR exceeding limits in Rule 7 (SG Feed Control), THEN throttle EFDW, as necessary




CAUTION ATWS events may initially require throttling to prevent exceeding pump limits and additional throttling once the Rx is shutdown to prevent overcooling


Page 40 of 56


EOP Encl 5.9 Crew response: EOP Enclosure 5.9 (Extended EFDW Operation) rev 01

1. Monitor EFDW parameters on EFW graphic display

2. IAAT UST level is < 4', THEN GO TO Step 120

3. IAAT feeding both SGs with one MD EFDWP is desired, THEN perform Steps 4 – 7

RNO: GO TO Step 8

8. Perform as required to maintain UST level > 7.5' __ Makeup with demin water __ Place CST pumps in AUTO

9. IAAT all exist: __ Rapid cooldown NOT in progress __ MD EFDWP operating for each available SG __ EFDW flow in each header < 600 gpm THEN place 1 TD EFDW PUMP switch in PULL TO LOCK

10. Verify 1 TD EFDW PUMP operating

11. Start TD EFDWP BEARING OIL COOLING PUMP 12. Notify CR SRO to set priority based on the NOTE above and EOP

activities


NOTE • Loss of the condensate system for ≥ 25 minutes results in cooling down to

LPI using the ADVs. If NO HWPs are operating, continuing this enclosure to restore the condensate system is a priority unless the CR SRO deems EOP activities higher priority. The 25 minute criterion is satisfied when a HWP is started and 1C-10 is 10% open.



Page 41 of 56


EOP Encl 5.38 Crew response: EOP Enclosure 5.38 (Restoration of Power) rev 0

1. Verify power has been restored

RNO: GO TO Step 3

3. Place 1HP-31 in HAND and reduce demand to 0

4. Close 1HP-21

5. Verify any of the following energized: __ MFB1 __ MFB2

RNO: GO TO Step 8

8. Verify CT-1 indicates ≈ 4160 volts

RNO: GO TO Step 18

18. Verify both Standby Bus #1 and Standby Bus #2 are de-energized (0 volts)

19. Verify both Keowee units operating

RNO: 1. Emergency start Keowee units: __ KEOWEE EMER START CHANNEL A __ KEOWEE EMER START CHANNEL A 2. IF NO Keowee units are operating, THEN perform the following: A. Notify Keowee operator to restore a Keowee unit to operable

status B. GO TO Step 38

Examiner Note: Keowee Hydro Unit 1 will be operating



Page 42 of 56


CT-3

EOP Encl 5.38 Crew response: EOP Enclosure 5.38

20. Verify both Keowee units in Oconee Control (statalarms on): __ UNIT 1 OCONEE CONTROL (2SA-17/E-1) __ UNIT 2 OCONEE CONTROL (2SA-18/E-1)

RNO: Notify Keowee Operator to place both Keowee units Master Transfer switches to remote

21. Verify both Keowee units operating

RNO: 1. IF UNIT 1 EMER FDR ACB 3 is closed, AND Unit 1 Keowee is NOT operating, THEN open UNIT 1 EMER FDR ACB 3

2. IF UNIT 2 EMER FDR ACB 4 is closed, AND Unit 2 Keowee is NOT operating, THEN open UNIT 2 EMER FDR ACB 4

22. Ensure one of the following is closed for an operating Keowee unit:

√ Unit 1 √ Unit 2 UNIT 1

EMER FDR ACB 3

UNIT 2 EMER FDR ACB 4

23. Verify 4160 volt power has been restored to the MFB

24. GO TO Step 35

35. Verify any of the following energized: __ 1TC __ 1TD __ 1TE

36. Notify Unit 1 CRS of status of 4160V SWGR




Page 43 of 56


EOP Encl 5.45 Crew response: EOP Enclosure 5.45 (PSW Feed and RCP Seals) rev 03

1. Verify Unit 2 EOP Encl 5.42 (PSW Power and Pump Alignment) in

progress or complete

RNO: Notify Unit 2 to PERFORM Unit 2 EOP Encl 5.42 (PSW Power and Pump Alignment)

Examiner Note: Once an RO notifies Unit 2 to perform EOP Encl 5.42, a booth instructor will enter the simulator to perform the enclosure. This enclosure will have no effect on the outcome of the scenario.

2. WHEN the Unit 1 PSW 4KV POWER AVAILABLE light lit, THEN continue

3. Verify it is desired to power HPI from PSW

RNO: GO TO Step 9

4. Verify PSW SELECTED HPI PUMP selected to the 1A HPI pump

RNO: GO TO Step 6

6. Verify PSW SELECTED HPI PUMP selected to the 1B HPI pump

7. Perform the following: A. Place 1B HPI PUMP POWER TRANSFER to PSW B. Do NOT wait for the PSW power light, continue


NOTE While running this enclosure along with a loss of power to HPIPs, Rule 4 should not be performed until SG feed has been established from some source

NOTE There is a 40 second time delay in the swap from Normal to PSW power on HPIPs


Page 44 of 56



8. Place POWER TRANSFER switch to PSW on the following: • 1HP-24 • 1HP-26 • 1RC-159/1RC-160 • 1RC-157/1RC-158 • 1RC-155/1RC-156

9. Position the switch for 1PSW-6 to open

10. Notify the CRS that PSW is aligned and ready to supply SG feed and RCP Seals

11. IAAT it is desired to supply PSW SG feed, THEN GO TO Step 42

12. IAAT it is desired to supply RCP seals, THEN GO TO Step 57 (page 45) 13. IAAT Pzr level ≥ 85" (165" ACC), AND it is desired to power the Pzr

heaters from PSW, THEN dispatch an operator to perform EOP Encl 5.46 (Aligning PZR Heaters From PSW)

14. IAAT notified that PSW is NO longer required, THEN GO TO Step 16

15. WHEN directed by CRS, THEN EXIT this enclosure


NOTE If RCS head vents, loop vents, or PORV have been opened, ACC levels must be used


Page 45 of 56



57. IAAT notified that SSF Feed has been established or is capable of feeding SGs, THEN simultaneously close the PSW control valves: __ 1PSW-22 (1A SG) __ 1PSW-24 (1B SG)

58. IAAT PSW SG feed is aligned but NOT feeding, AND RCS pressure > 2300 psig, THEN PERFORM Step 43.

59. IAAT feeding SGs with PSW AND Tc 550 - 555°F, THEN throttle the

following valves, as required, to slowly raise SG levels to 240"- 260" XSUR (270" – 290" ACC) while maintaining Tc 550 - 555°F: __ 1PSW-22 (1A SG) __ 1PSW-24 (1B SG)

60. Perform the following:

__ Close 1HP-139 __ Open 1HP-24 __ Close 1HP-26

61. Ensure 1HP-120 in manual with demand at zero

62. Verify BWST available based on Control Room level indication



NOTE ES may have actuated requiring manual control of HPI


Page 46 of 56



63. Start an HPI Pump by positioning HPI PUMP START FROM PSW

POWER to START

64. Throttle 1HP-140 to obtain as close to 30 - 35 gpm RCP Seal flow as possible

65. Maintain 30 - 35 gpm Seal Inlet Hdr Flow by throttling 1HP-140 as

necessary

66. Close 1HP-5

67. Close 1HP-21

68. IAAT Pzr level ≥ 85" (165" ACC), AND it is desired to power the Pzr heaters from PSW, THEN dispatch an operator to perform EOP Encl 5.46 (Aligning Pzr Heaters from PSW)

69. Notify CRS PSW RCP seals have been established


NOTE • HPI pump ammeter will not respond when HPI pump is powered from PSW • When the SSF and PSW are both supplying seals it may not be possible to

get 30 – 35 gpm flow through 1HP-140

NOTE RCS pressure higher than normal RCS operating pressure could restrict flow

through 1HP-140



Page 47 of 56

Enclosure 5.5

Pzr and LDST Level Control










C. Notify SRO.



Page 48 of 56





__ GO TO Step 10.

7. Open: • 1HP-24 • 1HP-25



1LP-15 1LP-16 1LP-9 1LP-10 1LP-6 1LP-7



F. GO TO Step 8.




C. GO TO Step 9.


Page 49 of 56









Page 50 of 56




__ GO TO Step 12.





C. Open:

1LP-15

1LP-16

1LP-9

1LP-10

1LP-6

1LP-7








Page 51 of 56






GO TO Step 35.

16. Open: __ 1CC-7 __ 1CC-8








20. GO TO Step 23.



GO TO Step 23.

22. GO TO Step 35.


Page 52 of 56








25. Open: 1HP-1 1HP-2 1HP-3 1HP-4





28. __ Close 1HP-7.


2. Close: 1HP-8 1HP-9&11




Page 53 of 56





33. Open 1HP-6.










Page 54 of 56




__ GO TO Step 40.




C. Close 1CS-46.





__ GO TO Step 42.



• • • END • • •


Page 55 of 56

CRITICAL TASKS

CT-1 ICS must be taken to HAND and FDW flow adjusted to prevent tripping the reactor on

RPS Low Pressure (1810 psig) or RPS Variable Low Pressure. Adjustment will be dependent on how much time it takes to place ICS in HAND.

CT-2 Start the TD EFDWP to supply Feedwater flow to the SGs for heat removal to preclude initiation of HPI Forced Cooling.

CT-3 Restore power from Keowee Hydro Unit 1 prior to the completion of EOP Enclosure 5.38 (BWOG CT-3).

of 56


SSF Operable: Yes KHU's Operable: U1 - OH, U2 - UG LCTs Operable: 2 Fuel Handling: No UNIT STATUS (CR SRO)










TS/SLC #

AMSAC/DSS Today/0300 7 Days SLC 16.7.2



• 1RIA-3 and 5 removed from RB. • SASS is in Manual for calibration Secondary • AMSAC/DSS bypassed for calibration • Unit 2 is supplying the AS header • OP/1/A/1106/002B Encl 4.13 in progress to restore 1B Main FDW Pump from Handjack • 1SSH-1, 1SSH-3, 1SD-2, 1SD-5, 1SD-140, 1SD-303, 1SD-355, 1SD-356 and 1SD-358 are








________________________ ________________________ OATC


• Reactor Power = Critical below POAH Turnover:

• 1A CFT level high. Lower 1A CFT level per OP/1/A/1104/001 Encl 4.17 • Unit startup in progress. Increase Reactor power from below the POAH to ≈ 3.0% and place

ICS in Auto per OP/1/A/1102/001 Encl 4.7

Event No.


Description

0a Override AFIS Fails to Actuate

0b Override ES Channel 1 Fails to go to Manual

0c

1 N: BOP, SRO Lower 1A CFT Level

2 R: OATC, SRO Increase Power to ≈ 3.0% and Place ICS in Auto

3 Override C: BOP, SRO 1B FWPT Auxiliary Oil Pump Trip

4 MPS061

MPS061D C: BOP, SRO (TS) 1A Letdown Cooler Leak

5 SRO (TS) TD EFDWP Oil Sump Dry

6 Override C: OATC, SRO 1RC-1 (PZR Spray Valve) Fails Open

7 MCR022 MCR028 C: OATC, SRO Two Dropped Control Rod(s) Requiring a Reactor

Trip

8 MSS360

M: ALL

1A MSLB Inside Containment • AFIS Fails to Actuate • ES Channel 1 Fails to go to Manual



Page 2 of 75


Event 1: When the crew takes the shift, the SRO will direct the BOP will lower the level in the 1A Core Flood Tank using OP/1/A/1104/001 Enclosure 4.17.

Event 2: The OATC will increase power from below the POAH to ≈ 3.0% and place ICS in automatic.

Event 3: The Auxiliary Oil Pump for the 1B FWPT will trip causing the Emergency Oil Pump to start. Per the Alarm Response Guide (ARG), the BOP will attempt to start the Auxiliary Oil Pump which will fail. Shortly afterwards, the alarm for FWPT 1B Emergency Oil Pump Overload will alarm which will require the candidate to attempt to start the Auxiliary Oil Pump. Not being able to start the pump, the ARG will direct the BOP to stop the Turning Gear Motor to prevent damaging the Feedwater Pump.

Event 4: The 1A Letdown Cooler will develop a leak. The SRO will enter AP/1/A/1700/018 (Abnormal Release of Radioactivity) and AP/1/A/1700/002 (Excessive RCS Leakage). AP/1/A/1700/002 will isolate the 1A Letdown Cooler. The SRO will enter TS 3.4.13

Event 5: The Work Control Center SRO will contact the control room stating that an AO reports that the Turbine Driven Emergency FDW Pump oil sump indicates no oil on the dipstick. The SRO should direct the OATC to place the TD EFDW pump switch in Pull To Lock (PTL). The SRO will enter TS 3.7.5 for an inoperable TD EFDW Pump.

Event 6: The 1RC-1 (Pressurize Spray Valve) will fail open causing RCS pressure to unexpectedly decrease. The operator will be required to close 1RC-3 (Pressurizer Spray Block Valve) in order to stop the pressure decrease and prevent tripping the Reactor.

Event 7: Two control rods will drop into the core requiring a manual Reactor trip. Once the Reactor is manually tripped in accordance with OMP 1-18 (Implementation Standards During Abnormal and Emergency Events), Event 8 will automatically initiate.

Event 8: When the Reactor trips, a Main Steam Line Break on the 1A Steam Generator will occur inside containment. Engineered Safeguards (ES) Channels 1-6 will actuate on high Reactor Building pressure. The SRO will enter the Excessive Heat Transfer tab and one RO will initiate Rule 5 (MSLB) while the other RO initiates EOP Enclosure 5.1 (ES Actuation). The Automatic Feedwater Isolation System (AFIS) will fail to actuate which will require an operator to manually isolate Feedwater to the 1A SG to prevent overcooling the RCS and Reactor Core. ES Channel 1 will fail to go to manual which will require an operator to override the Odd Voters to take manual control of ES Channel 1 components.


Page 3 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 1 Page 1 of 1 Event Description: Lower 1A CFT Level (C: BOP, SRO)

Time Position Applicant's Actions or Behavior

SRO

BOP

OP/1/A/1104/001 Encl 4.17 Crew Response: OP/1/A/1104/001 Enclosure 4.17 (Lowering CFTs Level To MWHUT) rev 79

2.1 Open 1CF-38 (CF Drn to MWHUT Blk). (A-4-W Pent Rm)

2.2 IF required to decrease level in 1A CFT, perform the following:

2.2.1 Record 1A CFT level: _______ feet

2.2.2 IF Containment Operability is required, enter Technical Specification 3.6.3 Condition 'A'

2.2.3 Open 1CF-3 (1A CFT SAMPLE & DRAIN)

2.2.4 IF Containment Operability is required, establish constant communication with CR and personnel near 1CF-7. (A-4-W Pent Rm)

2.2.5 Begin monitoring 1A CFT level and pressure

2.2.6 Throttle 1CF-7 (CF to MWHUTs). (A-4-W Pent Rm)

2.2.7 IF AT ANY TIME ES actuation occurs, perform the following: • Notify personnel to close 1CF-7 (CF to MWHUTs). (A-4-W

Pent Rm) • Close 1CF-3 (1A CFT SAMPLE & DRAIN)

2.2.8 WHEN 1A CFT is at desired level, lock closed 1CF-7 (CF to MWHUTs). (A-4-W Pent Rm)

2.2.9 Close 1CF-3 (1A CFT SAMPLE & DRAIN)

2.2.10 IF entered, evaluate exiting Technical Specification 3.6.3 Condition 'A'

2.2.11 Verify 1A CFT level and pressure stable

2.3 IF required to decrease level in 1B CFT, perform the following: (N/A) 2.4 Close 1CF-38 (CFT Drn to MWHUT Blk) (A-4-W Pent Rm)

This event is complete when 1CF-38 is closed (Step 2.4), or when directed by the Lead Examiner.


Page 4 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 2 Page 1 of 2 Event Description: Increase Power to ≈ 3% and Place ICS in Auto (R: OATC, SRO)


SRO

OATC

Examiner Note: During the power increase, the Unit 1 CRS will assume the role of the dedicated Reactivity Management SRO.

OP/1/A/1102/001 Encl 4.7 Crew Response: • SRO directs the OATC to increase power to ≈ 3% using OP/1/A/1102/001

(Controlling Procedure for Unit Startup), Encl. 4.7, starting at step 3.36 OP/1/A/1102/001 (Controlling Procedure for Unit Startup) Encl. 4.7 Step 3.36 rev 310

NOTE: • Point Of Adding Heat (POAH) is normally achieved from 0.05 to 0.15% power on Wide Range Indications • When POAH is achieved: TBVs will begin to open, 1HP-120 will

begin to close, TAVE will increase, and SUR will decrease with negative Moderator Temperature Coefficient

• Wide Range indications are used since Source Range NIs saturate

3.36 Begin reactor power increase to 0.5 - 1.0 % at ≤ 0.5 DPM SUR

3.37 WHEN above POAH, begin reactor power increase to 2.5 - 3.5%

3.38 WHILE power increases, begin increasing 1HP-120 (RC VOLUME CONTROL) setpoint to establish 215" to 225" PZR Level

NOTE: • TAVE error is blocked when on Low Level Limit and TAVE is < setpoint • Core reactivity effects are minimized with Rx in automatic

3.39 WHEN at 2.5 - 3.5 % Power, perform the following: (R.M.) 3.39.1 Place REACTOR MASTER to "AUTO" 3.39.2 Place DIAMOND to "AUTO" 3.39.3 Ensure TURBINE MASTER Setpoint to 880 - 890 psig

3.40 Perform the following: • Ensure complete Enclosure "Prior To Entry Into MODE 1" of

PT/1/A/0630/001 (Mode Change Verification) • Review mechanical maneuvering rates and allowable ramp rates in

PT/0/A/1103/020 (Power Maneuvering Guidelines)

This event is complete when power is ≈ 3% and ICS is in Auto (Step 3.39), or when directed by the Lead Examiner.


Page 5 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 2 Page 2 of 2 Event Description: Increase Power to ≈ 3% and Place ICS in Auto (R: OATC, SRO)


OATC

OP/1/A/1102/001 Encl 4.7 Crew Response: 3.41 IF AT ANY TIME all the following:

• All operable T Cold indications > 550°F • SSF inoperable only due to low decay heat, Notify SM the SSF is Available for Unit 1 SM notified:______________________ Time/Date:_______________

3.42 Ensure acceptable point status for plant startup for MODE 1: • OAC Alarm Screen Review • OAC Point Processing Log

NOTE: OAC Points can be found on Turn-On Code FDW02

3.43 Set temporary alarms on the following:

• OAC Point O1E2129 (FDW LOOP A COMPOSITE VALVE DEMAND) temporary alarm set at 9.8%

• OAC Point O1E2130 (FDW LOOP B COMPOSITE VALVE DEMAND) temporary alarm set at 9.8%

• Note on Turnover sheet temporary alarms set on Composite Valve Demand

This event is complete when power is ≈ 3% and ICS is in Auto (Step 3.39), or when directed by the Lead Examiner.


Page 6 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 3 Page 1 of 1 Event Description: 1B FWPT Auxiliary Oil Pump Trip (C: BOP, SRO)


SRO/BOP

Plant Response:

• 1SA-8/B-7 FWPT “B” HYDRAULIC PRESSURE LOW • 1SA-8/C-6 FWPT “B” EMERGENCY OIL PUMP RUN in alarm • FWPT B Emergency Oil Pump starts

Crew Response: SRO will direct the BOP to refer to 1SA-08/C-6

1SA-08/C-3 1SA-08/C-6 FWPT “B” EMERGENCY OIL PUMP RUN rev 36

3.1 Verify that 1B FWPT emergency oil pump is running

3.2 IF applicable, verify 1B FWPT is still on turning gear [It will be] 3.3 Try to restart 1B FWPT auxiliary oil pump [It will fail to start] 3.4 IF restart fails, notify Maintenance for repairs Booth Cue: Approximately 2 minutes after firing timer 3,

Timer 12 will auto actuate which will trip the Emergency Oil Pump and cause 1SA-08/C-7, (FWPT “B” EMERGENCY OIL PUMP OVERLOAD) to alarm.

Booth Cue: If dispatched as an AO to investigate 1B FWPT, notify crew that all oil pumps are off and the FDWPT is on the turning gear. Report that the TG motor is much noisier than normal.

1SA-08/C-7

1SA-08/C-7 FWPT “B” EMERGENCY OIL PUMP OVERLOAD

3.1 IF available, start Auxiliary Oil Pump and stop Emergency Oil Pump

3.2 IF AOP is NOT available, stop Turning Gear Motor and emergency Bearing Oil Pumps

3.3 Notify Maintenance for repairs

3.4 As soon as EBOP OR Auxiliary Oil Pump becomes available, start pump and place Turbine on Turning Gear

This event is complete when Turning Gear Motor is secured, or as directed by the Lead Examiner.


Page 7 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 4 Page 1 of 8 Event Description: 1A Letdown Cooler Leak (C: BOP, SRO) (TS)


SRO/ BOP

Plant Response:

• 1SA-08/B-9 (PROCESS MONITOR RADIATION HIGH) • 1SA-09/D-1 (COMPONENT COOLING SURGE TANK HIGH/LOW)

Crew Response: The SRO will direct the BOP to refer to ARG 1SA-08/B-9

1SA-08/B-9

1SA-08/B-9 rev 36

3.1.1 Determine radiation monitors in alarm.

3.1.2 IF radiation monitoring data from PMC is NOT in service, refer to OP/1/A/1103/026, (Loss of Sorrento Radiation Monitor).

Examiner Note: Steps 3.2 through 3.11 are IF statements for which RIA is in alarm. In this case, the crew determines the radiation monitor alarming is 1RIA-50, so step 3.4 applies

3.4 IF any of the following RIAs have valid alarms, GO TO AP/18 (Abnormal Release of Radioactivity).

1RIA-50

AP/1/A/1700/018

AP/1/A/1700/018 Abnormal Release of Radioactivity rev 23 (Can be performed by Unit 2 if AP/2 has been entered)

4.1 Perform the following:

At the discretion of the CRS, make a PA announcement of the event including any necessary precautions to be observed.

Notify Shift Manager to reference the following: • RP/0/A/1000/001 (Emergency Classification). • NSD-202 (Reportability) • OMP 1-14 (Notifications)

4.2 GO TO appropriate sections for any monitors in High or Alert alarm:

Examiner Note: Crew should go to Section 4I for 1RIA-50.

This event is complete when 1A Letdown cooler has been isolated or as directed by the Lead Examiner.


Page 8 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 4 Page 2 of 8 Event Description: 1A Letdown Cooler Leak (C: BOP, SRO) (TS) Time Position Applicant's Actions or Behavior

SRO /BOP

AP/1/A/1700/018 Crew Response: AP/1/A/1700/018 Section 4I

1 Verify either of the following: • CC SURGE TANK increasing ≥ 5″/hour • CC SURGE TANK off-scale high

2. Initiate AP/2 (Excessive RCS Leakage) Examiner Note: Crew may enter AP/2 directly because the following entry

condition is met: “Reactor Coolant leakage into CC system at ≥ 5″/hour on CC Surge Tank (≈ 0.65 gpm) or CC Surge Tank level off-scale high”

AP/1/A/1700/002

AP1/A/1700/002 (Excessive RCS Leakage) rev 15

Immediate Actions 3.1 Verify HPI operating 3.2 IAAT RC makeup flow is > 100 gpm,

AND Pzr level is decreasing, THEN close 1HP-5

3.3 IAAT all the following exist: • RCS leakage > NORMAL MAKEUP CAPABILITY (≈160 gpm) with

letdown isolated • Pzr level decreasing • SG Tube Leakage NOT indicated • LPI DHR NOT providing core cooling THEN perform the following: A. Ensure Rx is tripped B. Initiate Unit 1 EOP



Page 9 of 75


SRO /BOP

AP/1/A/1700/002 Crew Response: AP/1/A/1700/002 Subsequent Actions

NOTE Other than a SGTR, 1HP-26 should NOT need be open with the Rx critical.

4.1 Initiate Pzr and LDST level makeup using Unit 1 EOP Encl 5.5 (Pzr and LDST Level Control), as necessary (Page 44)


NOTE “LPI DHR in service” means that prior to the event the unit was shutdown with the decay heat drop line aligned for decay heat removal operation.

4.3 IAAT LPI DHR in service, AND RCS leakage > LDST makeup capability, THEN GO TO AP/26 (Loss of Decay Heat Removal) [N/A]

4.4 Initiate the following notifications: __ OSM to reference the following:

• RP/0/A/1000/001 (Emergency Classification) • OMP 1-14 (Notifications) • Encl 5.9 (Oversight Guidelines)

__ STA __ RP

4.5 Monitor the following trends to determine leak area (AB or RB) and trend for degradation: • "T6 AP02" • "T6 WASTE" • RIAs

4.6 Verify specific leak location is identified

RNO: Notify WCC SRO to initiate Encl. 5.2 (Primary Leak Check) and of the leak area (AB or RB), if known



Page 10 of 75


SRO /BOP


4.7 Initiate Encl 5.1 (Leak Rate Determination) (Page 13)

Examiner Note: This will calculate RCS leakage based on CC surge tank level increase and is a very gross calculation since there is no OAC point for CC surge tank level and therefore control room gage must be used.

4.8 WHEN leak area/failure is identified, THEN GO TO applicable step that best fits leak area/failure:

4.16 Verify all of the following:

• CC Surge Tank level increasing at ≥ 5″/ hour (≅0.65 gpm) or level is off-scale high

• 1RIA-50 in alarm or increasing

NOTE • CC Surge Tank is hard piped to overflow to LAWT • Chemicals from the CC system will rapidly exhaust demineralizers used to

process LAWT water. Radwaste has limited storage capability and a Rx shutdown may be required if the leak CANNOT be isolated.

4.17 IAAT CC Surge Tank level is off-scale high, THEN notify Radwaste that the CC Surge Tank has overflowed to the LAWT

NOTE • Closing 1CC-7 and 1CC-8 will contain the RCS leak in the RB. Relief valves

on the CC system will lift inside the RB when 1CC-7 and 1CC-8 are closed. • 1CC-7 and 1CC-8 should be closed prior to letting RC to CC system leak fill

the LAWT above the Hi/Hi OAC alarm (72” on O1A1352(Low Activity Waste Tank Level)).

• Flashing of the CC system may be indicated by CRD flow low or fluctuating, CC flow low or fluctuating, and RB Sump Level increase.



Page 11 of 75


SRO /BOP


4.18 IAAT RCS leakage is flashing the CC system, OR threatens to overflow the LAWT, THEN perform the following: (No flashing indicated) A. Trip Rx B. Close 1HP-5 C. Close the following:

• 1CC-7 • 1CC-8 • 1CC-3 • 1CC-4 • 1CC-5 • 1CC-6

D. Initiate AP/32 (Loss of Letdown) E. GO TO Step 4.24

NOTE RCS to CC leakage may be indicated by one RCP cooler outlet temperature increasing more than the others (use historical temperature trend) near the time of increase on 1RIA-50.

4.19 Verify leakage indicated by change in RCP cooler outlet temperatures (Turn-on code "GD AP02") [No leakage indicated]


NOTE RCS leakage to CC in the letdown coolers may be indicated by a cooler outlet temperature increasing more than the other cooler. Due to CC system setup, letdown cooler CC outlet temperatures may be different. A historical OAC temperature trend may be required to determine if leakage exists and if actions taken are successful in leak isolation. If leaking cooler CANNOT be identified, the coolers will be isolated one at a time starting with the 1A Letdown Cooler.

4.24 Monitor letdown cooler outlet temperatures to determine which cooler is leaking (Turn-on code "GD AP02"): • O1A0065 (LETDOWN COOLER 1A CC OUTLET TEMP) • O1A0066 (LETDOWN COOLER 1B CC OUTLET TEMP)



Page 12 of 75

Op-Test No ILT16-2 Scenario No.: 4 Event No.: 4 Page 6 of 8 Event Description: 1A Letdown Cooler Leak (C: BOP, SRO) (TS) Time Position Applicant's Actions or Behavior

SRO /BOP


4.25 GO TO the appropriate step to isolate affected cooler

√ Letdown Cooler to be Isolated

GO TO Step

1A 4.26 1B 4.33

Unknown 4.26

4.26 Verify 1B Letdown Cooler is isolated

RNO: 1. Isolate the 1A Letdown Cooler by performing the following: A. Close 1CC-1/1HP-1 B. Close 1HP-3

2. GO TO Step 4.31

4.31 Verify the leak isolation was successful: • CC Surge Tank level stable if 1CC-7 and 1CC-8 open • Decrease in RCS leakage


4.44 Verify both the following are closed; • 1CC-7 • 1CC-8





Page 13 of 75


SRO /BOP

AP/1/A/1700/002 Crew Response: AP/1/A/1700/002 Encl 5.1 (Leak Rate Determination)

1. Stabilize RCS Temperature

2. Notify WCC to secure all primary draining/RB washdown evolutions if applicable

NOTE Depending on leak location, leakage may NOT be detected by all the formulas. One or more of the following methods may be necessary to determine RCS leak rate.

3. Calculate leak rate using the following, as required:

• Calculation of RCS Volume Loss: Leak Rate = ______ + ______ - ______ - ______ = ______

MU SI LD TSR Where: MU = makeup Flow

SI = Seal Inlet Hdr Flow LD = Letdown Flow TSR = Total Seal Return Flow

• LDST Level Change:

Leak Rate = (LDST level change) x (31 gal/inch) + BTP Flowrate (gpm) (minutes)

Leak Rate = ( inches) x 31 gal/inch + ______ gpm = ________ gpm ______minutes

• HAWT/LAWT Level Change: (Turn-on code “LWD”) Leak Rate = (Change in HAWT/LAWT lvl) x (18 gal/inch) =(____inches) x 18 gal/inch =______ gpm

(minutes)

• RBNS Level Change: Leak Rate = (Change in RBNS level) x (15 gal/inch) = ( inches) x 15 gal/inch = ____ gpm

(minutes) ____ minutes



Page 14 of 75


SRO /BOP

SRO


• RCS Leakage Calculation per PT/1/A/0600/010 (Reactor Coolant Leakage) • Calculation of RCS Volume Loss:

Leak Rate = Makeup Flow rate with stable level = ____gpm

NOTE Pressurizer 23.94 gal/inch Quench Tank: 34.94 gal/inch BWST: 7608 gal/foot SFP (Unit 1 & 2) 1512 gal/ 0.1 foot SFP/FTC (Unit 1&2) 2300 gal/inch Corer Flood Tank: 5 gal/ 0.01foot RB Normal Sump: 15 gal/inch LDST: 31.3 gal/inch CC Surge Tank Level: 7.8 gal/inch MWHUT: See OP/0.A/1108/001 (Curves and Gen Info)

• Tank Level Change: Leak Rate = (Change level) x (Tank volume/height) = ( ) x ( ) = ____gpm

(minutes) ____minutes

4. Notify OSM and SRO of calculated leak rate

3.4.13 RCS Operational Leakage Condition A (4 hours) Reduce leakage to within limits

Examiner Note: TS will apply until the 1A Letdown Cooler is isolated



Page 15 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 5 Page 1 of 1 Event Description: TD EFDWP oil sump dry (SRO) (TS)


SRO

Plant response: Simulator Operator call the Control Room as the WCC SRO and report that the Basement AO found the Unit 1 TD EFDWP oil sump with no oil indicating on the dipstick. Report that the WCC and FIN-24 are investigating the situation to determine the cause of the problem. No oil has been found leaking from any equipment.

Crew response:

• SRO should make the decision to place TD EFDWP in “Pull to Lock”

TS 3.3.14 EMERGENCY FEEDWATER (EFW) PUMP INITIATION CIRCUITRY Condition B.1 (Immediately) Declare the affected EFW pump inoperable. TS 3.7.5 EMERGENCY FEEDWATER (EFW) SYSTEM Condition B.1 (72 hours) Restore turbine driven EFW pump and EFW flow path to OPERABLE status.

• SRO refer to TS 3.3.14 Condition B • Declare the affected EFWP inoperable Immediately • SRO refer to TS 3.7.5 Condition B

Restore TD EFDWP within 72 hours Booth Cue: If asked, inform crew that the TD EFDWP was placed on the

oil purifier last shift. Examiner Note: The SRO may direct an RO to place the TD EFDW pump

in Pull-To-Lock per AD-OP-ALL-1000 Step 5.6.3.3.b 5.6.3 Common Tasks

3. Written procedures are not necessary for situations where: b. Prompt action is necessary to prevent the deterioration of

plant conditions or components to a possibly unsafe or unstable level. If time permits, approval from the SM/CRS shall be obtained.

Booth Cue: If crew does not place TD EFDWP in PTL, then at direction of

Lead Examiner, call as WCC SRO and direct the crew to place Unit 1 TDEFDWP switch in PTL per R&R.

Examiner Note: Ensure the Pzr is saturated prior to initiating Event 6.

This event is complete when the Tech Spec determination has been made or when directed by the Lead Examiner.


Page 16 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 6 Page 1 of 3 Event Description: Spray Valve Fails Open (C: OATC, SRO) (TS)


SRO/OATC

CT-1

Examiner Note: Ensure Pzr saturated prior to Event 6. AP/1/A/1700/044

Plant Response: • 1SA-18/A-1 ( Pressurizer Relief Valve Flow) in alarm • 1RC-66 indicates open • Acoustic monitor indicates 1RC-66 open • RCS pressure decreasing

Crew Response: • OATC should recognize that RCS pressure is below the Pzr Spray Valve

setpoint and close the Pzr Spray Block Valve (1RC-3). This is an Immediate Action from AP/44 Abnormal Pressurizer Pressure Control.

• SRO should enter AP/1/A/1700/044 AP/1/A/1700/044 Abnormal Pressurizer Pressure Control rev 04

Immediate Manual Actions 3.1 IAAT PORV is open, AND RC pressure is < setpoint (2400 psig (HIGH)

or 480 psig (LOW)), THEN close 1RC-4

3.2 IAAT RC pressure < 2155 psig, AND 1RC-1 indicates open, THEN select 1RC-1 to CLOSE

3.3 IAAT all the following conditions exist: __ RC pressure < 2155 psig __ RC pressure decreasing without a corresponding decrease in PZR

level THEN close 1RC-3

Subsequent Actions


4.2 GO TO the applicable step per the following table:

√ Failure Caused RCS Pressure

Step

Decrease 4.3 Increase 4.18

This event is complete when 1RC-3 is closed and RCS pressure is stable, or as directed by the Lead Examiner.


Page 17 of 75



SRO/OATC


4.3 Verify 1RC-4 is closed

RNO: IF PORV is open, AND 1RC-4 has failed to close, THEN perform the following:

A. Dispatch an operator to open 1DIB Panelboard breaker #24 B. Manually trip the reactor C. Initiate AP/02 (Excessive RCS Leakage)

4.4 Verify 1RC-3 is closed

NOTE 1RC-3 must NOT be allowed to be closed for ≥ 36 minutes at a time to avoid a thermal transient in piping between 1RC-3 and the PZR spray nozzle.

4.5 Position 1RC-3 as required to maintain RC pressure within desired band

4.6 GO TO Step 4.13

4.13 Verify PZR heaters maintaining RCS pressure within desired band

4.14 Notify SPOC to repair malfunctioning component



Page 18 of 75



SRO/OATC


4.15 Ensure requirements of following are met: __ TS 3.4.1 (RCS Pressure, Temperature, and Flow Departure from

Nucleate Boiling Limits) __ TS 3.4.9 (Pressurizer) __ TS 3.4.12 (Low Temperature Overpressure Protection System) __ SLC 16.5.1 (Reactor Coolant System Vents)

4.16 WHEN repairs complete, THEN place following components in desired position for current plant conditions as determined by CR SRO: __ 1RC-1 __ 1RC-3 __ 1RC-4 __ PZR heater bank #1 __ PZR heater bank #2 __ PZR heater bank #3 __ PZR heater bank #4

Examiner Note: If RCS pressure goes below 2125 psig, TS 3.4.1 will apply

TS 3.4.1, REACTOR COOLANT SYSTEM Condition A (2 hours) Restore RCS DNB parameter(s) to within limit.

COLR DNB Limit = 2125 psig



Page 19 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 7 Page 1 of 1 Event Description: Two Dropped Control Rod(s) Requiring a Reactor Trip (C: OATC, SRO)


SRO/OATC

CT-2

AP/1/A/1700/001 Plant Response: • Group 6 Rod 3 drops into the core • Group 6 Rod 6 drops into the core • Statalarm 1SA-2/A-10 (CRD GLOBAL TROUBLE) • Statalarm 1SA-2/B-10 (CRD ASYMMETRIC ROD POSITION ERROR) • Statalarm 1SA-2/D-9 (CRD OUT INHIBIT) • Statalarm 1SA-5/A-5 (1A RPS TROUBLE) • Statalarm 1SA-5/B-5 (1B RPS TROUBLE) • Statalarm 1SA-5/D-5 (1D RPS TROUBLE)

Crew Response: • Crew should recognize 2 dropped control rods and trip the reactor in

accordance with OMP 1-18

Examiner Note: Event 8 will auto actuate when the Reactor is tripped.

If the crew does not recognize the 2nd dropped control rod, they will enter AP/1/A/1700/001 (Unit Runback) AP/1/A/1700/001 (Unit Runback) rev15

NOTE • If more than one runback condition exists, ICS will respond by selecting the

fastest runback rate and the lowest load limit. The most limiting runback will be the one with the fastest runback rate and the shortest duration.

• It is possible for a FDWP to become unable to feed the SGs but not be tripped. In this case a signal would not be sent to RPS or the EFDWP start circuit.

4.1 GO TO the most limiting section per the following table:

√ 4H Asymmetric Control Rod (1% / min to 55% power)

AP/1A/1700/001 Section 4H (Asymmetric Control Rod)

1. IAAT a more limiting runback occurs, THEN GO TO Subsequent Actions Step 4.1.

2. IAAT more than one control rod is dropped or misaligned ≥ 6.5% (9″) from the group average, THEN trip the Rx.

This event is complete when the Reactor is manually tripped, or as directed by the Lead Examiner.


Page 20 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 8 Page 1 of 10 Event Description: 1A MSLB inside containment (M: All)


OATC

BOP

SRO

EOP Plant Response: • 1SA-2/D-3 (RC Press High/Low) • 1SA-02/A-9 (MS PRESS HIGH/LOW)

Crew Response: Examiner Note: Crew will be performing IMAs and Subsequent Actions

as a result of the manual Reactor trip due to two dropped control rods.

IMAs EOP Immediate Actions rev 01

3.1 Depress REACTOR TRIP pushbutton.

3.2 Verify reactor power < 5% FP and decreasing.

3.3 Depress the turbine TRIP pushbutton

3.4 Verify all turbine stop valves closed.

3.5 Verify RCP seal injection available.

SYMPTOM CHECK The BOP will verify the following:



Any SCM < 0°F Rule 2, Loss Of SCM Loss of Main and Emergency FDW (including unsuccessful manual initiation of EFDW)






BOP will perform Rule 5 (Main Steam Line Break) (page 27) SRO will review IMAs and transfer to the Subsequent Actions Tab.

This event is complete when the crew has transferred to the FCD tab, or as directed by the Lead Examiner.


Page 21 of 75

Op-Test No.: ILT16-2 Scenario No.: 4 Event No.: 8 Page 2 of 10 Event Description: 1A MSLB inside containment (M: All) Time Position Applicant's Actions or Behavior

SRO/OATC

Subsequent Actions Tab Crew Response: SRO will review the Subsequent Action Tab Parallel Action (Yellow) page (Page 72) and transfer to the Excessive Heat Transfer (EHT)Tab.

EHT Tab SRO will review the EHT Tab Parallel Action (Yellow) page (page 73) and determine that ES has actuated and direct the OATC to perform Encl. 5.1 ES Actuation (Page 52)

Excessive Heat Transfer (EHT) Tab rev 0

1. Verify any SG pressure < 550 psig. [1A SG should be < 550 psig at this point]

2. Ensure Rule 5 (Main Steam Line Break) in progress or complete.

3. Place the following in HAND and decrease demand to zero on all affected SGs:

1A SG 1B SG 1FDW-32 1FDW-41

1FDW-35 1FDW-44

4. Close the following on all affected SGs:


1MS-17 1MS-26

1MS-79 1MS-76

1MS-35 1MS-36

1MS-82 1MS-84

1FDW-368 1FDW-369

5. Verify level in both SGs < 96% O.R.

6. IAAT core SCM is > 0°F, THEN perform Steps 7 and 8

RNO: GO TO Step 9

7. Throttle HPI per Rule 6 (HPI)

8. Verify letdown in service

RNO: IF desired to restore letdown, THEN initiate Encl 5.5 (Pzr and LDST Level Control). (page 44)



Page 22 of 75



SRO/OATC

EHT Tab Crew Response:

9. Verify any SG has an intact secondary boundary (intact SG) [1B SG is intact]

NOTE If only one SG is intact and has been isolated for SGTR, the following steps will unisolate and use it for heat removal.

10. Open the following on all intact SGs:

√ 1A SG √ 1B SG 1FDW-372 1FDW-382

1FDW-368 1FDW-369

1MS-17 1MS-26

11. Start MDEFDWP associated with all intact SGs:

√ 1A SG √ 1B SG 1A MDEFDWP 1B MDEFDWP

12. Feed and steam all intact SGs to stabilize RCS P/T using either: • TBVs • Dispatch two operators to perform Encl 5.24 (Operation of the ADVs)

13. GO TO Step 32

32. Verify any: __ HPI has operated in the injection mode while NO RCPs were

operating __ A cooldown below 400°F at > 100°F/hr has occurred

RNO: GO TO Step 34

33. Initiate Rule 8 (Pressurized Thermal Shock (PTS))

34. Verify both closed: __ 1MS-24 __ 1MS-33



Page 23 of 75



SRO/OATC


35. Open 1AS-8

36. Close 1SSH-9

37. Perform notifications: __ Notify Chemistry to determine RCS boron concentration __ Notify Secondary Chemistry to check for indications of SGTR __ Notify RP to check for indications of SGTR

38 IAAT RCS boron is determined to be insufficient for adequate SDM THEN initiate Encl 5.11 (RCS Boration)

39. IAAT all exist: __ ES Bypass Permit satisfied __ All SCMs > 0°F __ RCS pressure controllable THEN perform Steps 40 - 41

RNO: GO TO Step 42

40. Bypass applicable ES: To Bypass HPI: __ Bypass HPI ES CH A,B,C To Bypass LPI: __ Bypass LPI ES CH A,B,C

41. Bypass applicable Diverse ES: To Bypass HPI: __ Bypass Diverse HPI To Bypass LPI: __ Bypass Diverse LPI

42. Verify any SG is dry.

NOTE • Minimizing SCM reduces tensile stress on the SG • PORV should be used if Pzr spray is not available • Procedure progression may continue when actions to minimize SCM are in

progress



Page 24 of 75



SRO/OATC


43. Maintain minimum SCM using the following methods as necessary: __ De-energize all Pzr heaters __ Use Pzr spray __ Throttle HPI to maintain Pzr level > 100″ [180″ acc] __ Use PORV

44. Verify any RCP operating

RNO: GO TO Step 46.

45. Maintain RCP NPSH • OAC • Encl 5.18 (P/T Curves)

46. Initiate Encl 5.16 (SG Tube-to-Shell ΔT Control)

NOTE RCP 1A1 provides the best Pzr spray

47. IAAT all exist: __ < one RCP operating in any loop __ All SCMs > 0°F __ RCP available in an idle loop THEN initiate Encl 5.6 (RCP Restart) to start one RCP in each idle loop

48. IAAT all exist: __ RBS actuated __ RB pressure < 10 psig __ 1RIA-57 NOT in alarm __ 1RIA-58 NOT in alarm THEN stop both RBS pumps.

49. IAAT Tcold approaches 470°F, AND all RCPs are operating, THEN ensure < four RCPs are operating

50. IAAT BWST level is ≤ 19', THEN initiate Encl 5.12 (ECCS Suction Swap to RBES)

51. Verify all SCMs > 0°F



Page 25 of 75



SRO/OATC


52. Verify indications of SGTR ≥ 25 gpm.

RNO: GO TO Step 54

54 Verify required RCS makeup flow within normal makeup capability

55. Verify either: __ Any SG isolated __ Any SG has an unisolable steam leak

56. GO TO FCD tab

Forced Cooldown Tab Rev 0

1. IAAT cooldown rate CANNOT be controlled within Tech Spec limits: • Tcold ≥ 270°F: ≤ 50°F / ½ hr • Tcold < 270°F: ≤ 25°F / ½ hr THEN GO TO EHT tab

2. Verify letdown in service

RNO: 1. Ensure CC System in operation 2. IF 1A Letdown Cooler available, THEN open the following:

__ 1HP-1 __ 1HP-3

3. IF 1B Letdown Cooler available, THEN open the following: __ 1HP-2 __ 1HP-4

4. Close the following: __ 1HP-6 __ 1HP-7

5. Open 1HP-5 6. Adjust 1HP-7 for ≈ 20 gpm letdown 7. Open 1HP-6 8. Adjust 1HP-7 to control desired letdown flow

3. Establish and maintain appropriate level per Rule 7 (SG Feed Control) and pressure in available intact SGs

4. IAAT Tcold approaches 470°F, THEN ensure < four RCPs operating



Page 26 of 75



SRO/OATC

Forced Cooldown Tab Crew Response:

5 IAAT Tcold approaches 300°F, THEN ensure < three RCPs operating

6. IAAT all the following exist: __ ES Bypass Permit satisfied __ All SCMs > 0°F __ RCS pressure controllable THEN perform Steps 7 - 8

7. Bypass applicable ES: To Bypass HPI: __ Bypass HPI ES CH A,B,C To Bypass LPI: __ Bypass LPI ES CH A,B,C

8. Bypass applicable Diverse ES: To Bypass HPI: __ Bypass Diverse HPI To Bypass LPI: __ Bypass Diverse LPI

9. IAAT any SG is < 700 psig, AND AFIS is NOT actuated on that SG, THEN select OFF on both Digital Channels 1&2 for that header:

√ A Header √ B Header

DIG CH 1 OFF DIG CH 1 OFF

DIG CH 2 OFF DIG CH 2 OFF

10. Stabilize RCS temperature

11. Close 1HP-26

12. Stop 1C HPI pump

13. Adjust 1HP-120 for desired setpoint



Page 27 of 75


BOP

CT-3

RULE 5 Crew Response: Rule 5 (Main Steam Line Break) rev 01

1. Perform on affected headers:

√ A Header √ B Header

On AFIS HEADER A, depress CH. 1 INIT. On AFIS HEADER B,

depress CH. 1 INIT.

On AFIS HEADER A, depress CH. 2 INIT. On AFIS HEADER B,

depress CH. 2 INIT.

Select OFF for 1A MD EFDWP. Select OFF for

1B MD EFDWP.

Trip both Main FDWPTs. Trip both Main

FDWPTs. Close 1FDW-315. Close 1FDW-316.

Place 1FDW-33 switch to CLOSE.




Close 1PSW-22. Close 1PSW-24. Close 1PSW-23. Close 1PSW-25.

2. Verify 1 TD EFDW PUMP operating.

RNO: 1. IF MD EFDWP for the intact SG is operating SG is operating, THEN GO TO Step 5. [IT WILL NOT BE OPERATING] 2. Start 1 TD EFDW PUMP

3. Verify 1 TD EFDW PUMP is feeding affected SGs [1FDW-315 is closed]

RNO: GO TO Step 5 5. Verify 1B SG is an affected SG

RNO: GO TO Step 7



Page 28 of 75


BOP

RULE 5 Crew Response:

7. WHEN overcooling is stopped, THEN adjust steaming of unaffected SG to maintain CETCs constant using either: __ TBVs __ Dispatch two operators to perform Encl 5.24 (Operation of the ADVs)

CAUTION Thermal shock conditions may develop if HPI is NOT throttled and RCS pressure NOT controlled.

8. WHEN all exist: __ Core SCM > 0°F __ Rx power ≤ 1% __ Pzr level increasing THEN continue

9. Verify ES HPI actuated

10. Place Diverse HPI in BYPASS

11. Perform both: __ Place ES CH 1 in MANUAL __ Place ES CH 2 in MANUAL

12. Perform the following to stabilize RCS P/T: __ Throttle HPI __ Reduce 1HP-120 setpoint to control at >100″ [180″ acc] __ Adjust steaming of unaffected SG as necessary to maintain CETCs

constant

13. WHEN CETCs have stabilized, THEN resume use of Tc for RCS temperature control

14. Ensure Rule 3 (Loss of Main or Emergency FDW) is in progress or complete (Page 29)

15. Ensure Rule 8 (Pressurized Thermal Shock (PTS)) is in progress or complete




Page 29 of 75


OATC/BOP

RULE 3 Rev 1 Crew Response: Rule 3 (Loss of Main of Emergency FDW)

1. Verify loss of MFDW and/or EFDW was due to any of the following: __ Turbine Building Flooding __ Actions taken to increase SG level due to Turbine Building Flooding

RNO: GO TO Step 3

3. IAAT NO SGs can be fed with FDW (Main/CBP/Emergency/PSW), AND any of the following exist: __ RCS pressure reaches 2300 psig OR NDT limit __ Pzr level reaches 375” [340” acc] THEN PERFORM Rule 4 (Initiation of HPI Forced Cooling)



6. GO TO Step 38


RNO: GO TO Step 44

44. Verify any SCM ≤ 0oF

RNO: IF overcooling, OR exceeding limits in Rule 7 (SG Feed Control), THEN throttle EFDW, as necessary




CAUTION ATWS events may initially require throttling to prevent exceeding pump limits and additional throttling once the Rx is shutdown to prevent overcooling


Page 30 of 75

Enclosure 5.9

Extended EFDW Operation Rev 1

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1. __ Monitor EFDW parameters on EFW

graphic display.

2. __ IAAT UST level is < 4', THEN GO TO Step 120.

3. __ IAAT feeding both SGs with one MD EFDWP is desired, THEN perform Steps 4 - 7.

__ GO TO Step 8.

4. Place EFDW control valve on SG with NO EFDW flow to MANUAL and closed:


5. Locally open:

__ 1FDW-313 (1A EFDW Line Disch To 1A S/G X-Conn) (T-1, 1' N of M-16, 18' up)

__ 1FDW-314 (1B EFDW Line Disch To 1B S/G X-Conn) (T-1, 3' S of M-24, 10' up)

6. Ensure a MD EFDWP is operating.

7. Throttle EFDW control valve on SG with NO EFDW flow to establish appropriate level per Rule 7 (SG Feed Control):


8. Perform as required to maintain UST level > 7.5': Makeup with demin water. Place CST pumps in AUTO.

9. IAAT all exist: Rapid cooldown NOT in progress __ MD EFDWP operating for each

available SG __ EFDW flow in each header

< 600 gpm THEN place 1 TD EFDW PUMP switch in PULL TO LOCK.


Page 31 of 75

Enclosure 5.9

Extended EFDW Operation

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 10. Verify 1 TD EFDW PUMP operating.

GO TO Step 12.

11. __ Start TD EFDWP BEARING OIL COOLING PUMP.

NOTE

• Loss of the condensate system for ≥ 25 minutes results in cooling down to LPI using the ADVs. If NO HWPs are operating, continuing this enclosure to restore the condensate system is a priority unless the CR SRO deems EOP activities higher priority. The 25 minute criterion is satisfied when a HWP is started and 1C-10 is 10% open.


12. __ Notify CR SRO to set priority based on the NOTE above and EOP activities.

13. __ IAAT it is determined that condensate flow CANNOT be restored within 25 minutes, THEN GO TO Step 90.

14. Verify any HWP operating. 1. Place all CBP control switches to OFF. 2. GO TO Step 20.

15. Verify any CBP operating. 1. __ IF AP/11 restarted a HWP, THEN GO TO Step 22.

2. GO TO Step 41.

16. __ Verify 1C COND BOOSTER PUMP operating.{12}

1. __ Ensure only one CBP is operating. 2. __ GO TO Step 18.

17. Stop: {12} 1A COND BOOSTER PUMP 1B COND BOOSTER PUMP

18. Ensure only one HWP is operating.

19. GO TO Step 44.


Page 32 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 20. Verify a loss of power event caused the

loss of the secondary system.

__ GO TO Step 24.

21. __ Ensure AP/11 (Recovery From Loss of Power) is in progress.

22. __ WHEN AP/11 (Recovery From Loss of Power) has restored 600v load centers, AND a HWP is operating, THEN dispatch an operator to start all CBP Aux Oil Pumps. (T-1/J-21)

23. __ WHEN notified that all CBP Aux Oil pumps are operating, THEN GO TO Step 41.

24. Place all HWP control switches to OFF.

25. Place all CBP control switches to OFF.

26. Place valve switches to close until valve travel is initiated: 1FDW-4 1FDW-9

__ Continue.

27. Start: __ 1A FDWP AUXILIARY OIL PUMP __ 1B FDWP AUXILIARY OIL PUMP

Start as necessary: __ 1A FDWP EMERGENCY BRNG OIL

PUMP __ 1B FDWP EMERGENCY BRNG OIL

PUMP

28. Verify both: __ FWPT A BRG LUBE OIL

PRESS > 4 psig FWPT B BRG LUBE OIL

PRESS > 4 psig

1.__ IF both FDW pumps have BRG LUBE OIL PRESS < 4 psig, THEN GO TO Step 90.

2. Perform for the FDW pump that has BRG LUBE OIL PRESS < 4 psig:

Close 1FDW-1 for 1A FDW pump.

Close 1FDW-6 for 1B FDW pump.

29. Place in MANUAL and close:

1FDW-53 1FDW-65


Page 33 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 30. __ Place 1C-10 FAIL SWITCH in

MANUAL.

31. Close 1C-10.

32. __ Make plant page to clear basement and third floor of non-essential personnel.

33. Start one HWP.

34. __ Verify < 25 minutes elapsed since loss of condensate.

1. __ Stop all HWPs.


35. __ Throttle 1C-10 controller 10% open to satisfy 25 minute system restart criteria.

36. WHEN FWP SUCT HDR PRESS (1VB3) is ≥ 100 psig, THEN open 1C-10.

37. __ Place 1C-10 FAIL SWITCH in FAIL OPEN.

38. __ Dispatch an operator to start all CBP Aux Oil Pumps. (T-1/J-21)

39. Maximize total recirc flow < 1200 gpm with one of the following: 1FDW-53 1FDW-65

40. __ WHEN five minutes have elapsed, AND notified that all CBP Aux Oil pumps are operating, THEN continue procedure.

41. Start a second HWP.

42. Start 1C COND BOOSTER PUMP. {12} Start one available CBP.

43. Stop one operating HWP.

44. __ Place control switch for one secured HWP in AUTO.

45. __ Place control switch for one secured CBP in AUTO.


Page 34 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 46. __ Perform the following:

__ Position HWP LOAD SHED DEFEAT switch to a running HWP.

__ Position CBP LOAD SHED DEFEAT switch to a running CBP.

47. Place in MANUAL: 1FDW-53 1FDW-65

48. Establish 2300 - 6000 gpm total recirc flow with one of the following: 1FDW-53 1FDW-65

49. __ IAAT UST level CANNOT be maintained > 8.5', THEN locally open 1C-899 (Cond Recirc To UST Riser Throttle) (T-1/J-23).

50. __ IAAT UST level increases > 11', THEN perform as required: Throttle demin water __ Locally throttle 1C-899 (Cond

Recirc To UST Riser Throttle) (T-1/J-23)

51. Verify closed: 1FDW-4 1FDW-9

GO TO Step 58.


Page 35 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 52. Position switches in CLOSE:

1FDW-33 1FDW-31 1FDW-42 1FDW-40

53. Ensure closed: 1FDW-33 1FDW-31 1FDW-42

1FDW-40

54. __ Locally open: 1FDW-5 (1A FDWP Discharge Bypass) (T-1/SE of D-24 12' up) 1FDW-10 (1B FDWP Discharge Bypass) (T-1/N of D-26 9' up)

55. __ WHEN FWP DISCH HDR PRESS (1VB3) is approximately equal to either of the following: • O1A1014 (FDWP 1A

DISCHARGE PRESS) • O1A1391 (FDWP 1B

DISCHARGE PRESS) THEN open: 1FDW-4 1FDW-9


Page 36 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 56. Locally close:

__ 1FDW-5 (1A FDWP Discharge Bypass) (T-1/SE of D-24 12' up)

__ 1FDW-10 (1B FDWP Discharge Bypass) (T-1/N of D-26 9' up)

NOTE Windmill protection may have required closure of FDW pump suction valve.

57. Verify open: __ 1FDW-1 1FDW-6

1. __ IF required, notify the WCC SRO to initiate investigation.

2. Note on Turnover sheet that FDW pump associated with closed valve is not available for use until problem resolved.

58. __ IAAT it is desired to re-establish Main FDW, THEN initiate Encl (Re-establishing Main FDW) of OP/1/A/1106/002 (Condensate And FDW System).

59. __ IAAT EFDW has been secured per Encl (Re-establishing Main FDW) of OP/1/A/1106/002 (Condensate And FDW System), THEN EXIT.


Page 37 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 60. __ WHEN UST level is < 4',

THEN dispatch two operators to perform Encl 5.24 (Operation of the ADVs) in preparation for loss of vacuum. (PS)

61. __ Verify power available to 1V-186 by using valve position indicating light.

Dispatch an operator to be in position at 1V-186 (Vacuum Breaker) (T-3, catwalk at 1C2 waterbox).

NOTE 1C-573 will be closed after vacuum is broken.

62. Dispatch an operator with a safety harness to 1C-573 (MD EFDWPs Suction From UST) (T-1, SW of E-24, 8' above floor) to: Unlock and remove chain from 1C-573.

Establish communication with Control Room.

63. __ WHEN UST level is < 3', THEN continue.

64. __ Open 1V-186. __ Notify operator to open 1V-186 (Main Condenser Vacuum Breaker) (T-3, catwalk at 1C2 waterbox).

65. Stop all main vacuum pumps.

66. Stop all CBPs.

67. Stop all HWPs.

68. Close: 1MS-47 1AS-40

Dispatch an operator to close: 1MS-49 (1A CSAE Steam Supply)

(T-3/F-26) __ 1MS-58 (1B CSAE Steam Supply)

(T-3/G-26) __ 1MS-67 (1C CSAE Steam Supply)

(T-3/H-26)


Page 38 of 75

Enclosure 5.9



NOTE • 1C-573 is open unless Step 75 has been completed. • While EFDW is secured, a transfer to LOHT is required only when directed by this enclosure

or Rule 4 (Initiation of HPI Forced Cooling) conditions are met.

69. __ IAAT UST level is < 1', AND 1C-573 (MD EFDWPs Suction From UST) is open, THEN perform Steps 70 - 71.

__ GO TO Step 72.

70. Perform the following: Stop 1A MD EFDWP. Stop 1B MD EFDWP.

71. Verify 1C-391 open. 1. Stop 1TD EFDW PUMP.

2. Close: __ 1FDW-315 __ 1FDW-316


A. __ Reduce MD EFDWP flow to < 440 gpm per pump.

B. __ Notify crew of MD EFDWP flow limit while aligned to hotwell.

NOTE Vacuum gage or computer can be used. Vacuum is broken when either start to flat line. Do NOT change scale on computer trend once started.

73. __ WHEN vacuum is broken, THEN continue.


Page 39 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 74. __ IAAT MD EFDWPs are operating,

OR available to operate, THEN PERFORM Steps 75 - 77.

__ GO TO Step 78.

75. Locally close 1C-573 (MD EFDWPs Suction From UST) (T-1, SW of E-24, 8' above floor).

1. __ IF 1TD EFDW PUMP is operating, OR operable, THEN GO TO Step 78.

2. __ IF NO EFDW pumps are operating, THEN:

A. Notify CR SRO that a LOHT exists from loss of EFDW suction source.

B. Notify CR SRO that Rule 3 will be performed to cross connect with alternate unit.

C. Consider all U1 EFDW pumps inoperable, AND GO TO Rule 3.

76. Verify MD EFDWPs were stopped due to UST level < 1'.

__ GO TO Step 78.


A. __ Restart all MD EFDWPs that were stopped due to UST level < 1'.

B. Resume feeding available SGs.


Page 40 of 75

Enclosure 5.9


ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 78. Verify 1 TD EFDW PUMP operating.

GO TO Step 82.

79. Dispatch operator to 1C-157 (TD EFDWP Suction From UST) to establish communication with CR (T-1/C-20).

80. WHEN operator in place at 1C-157, THEN continue.

81. Stop 1 TD EFDW PUMP.

82. Locally close 1C-157 (TD EFDWP Suction From UST) (T-1/C-20).

1. IF NO EFDW pumps are operating, THEN:




2. GO TO Step 84.

83. Open 1C-391. 1. Attempt to locally open 1C-391 (TD EFDWP Suction From Hotwell) (T-1/C-20).

2. IF 1C-391 CANNOT be opened, AND NO EFDW pumps are operating, THEN:





Page 41 of 75

Enclosure 5.9


84. IAAT 1 TD EFDW PUMP operation is

desired, AND all exist: Hotwell level is > 1″. Vacuum is broken.

1 TD EFDW PUMP successfully aligned to hotwell.

THEN:

A. Start 1 TD EFDW PUMP.

B. Feed available SGs as required.

85. Dispatch an operator to open: 1C-188 (Hotwell Emerg Makeup #1 Control Bypass) (T-1/W of E-24). {18}

1C-912 (UST Riser To HW Emerg Makeup #2 Auto Isol Bypass) (T-1/G-23)

86. Notify TSC to evaluate methods to maintain secondary inventory including strategies located in EM 5.1 (Engineering Emergency Response Plan) and EM 5.2 (Evaluation By Station Management in the TSC - Beyond Design Basis Mitigation Strategies).


Page 42 of 75

Enclosure 5.9



87. IAAT hotwell level is ≤ 1″, THEN:

A. Stop all EFDWPs.

B. Consider all U-1 EFDW pumps inoperable, AND GO TO Rule 3.

NOTE • This step provides general plant directions for the SRO and Management team. The user shall

continue after the notification has been made. • Swapping from TBVs to ADVs prevents overfilling the hotwell/condenser. • Securing steam seals limits the water (condensation) that reaches the oil systems. Vacuum

must be broken to secure steam seals. • Engineering will determine when to allow secondary system restart. • Beginning a cooldown assumes HPI is operating. If the SSF is supplying seals, then further

discussion with the Management team should be undertaken prior to cooldown.

88. Notify the CR SRO to direct the following as time and resources allow:

• Transfer steam control from TBVs to ADVs.

• Operate ADVs per U1 EOP Encl 5.24 (Operation of ADVs).

• Begin Unit cool down to LPI per OP/1/A/1102/010 (Controlling Procedure For Unit Shutdown) using the ADVs.

• Break vacuum per OP/1-2/A/1106/016 (Condenser Vacuum System).

• Secure Steam Seals per OP/1/A/1106/13 (Steam Seal System).

89. WHEN directed by CR SRO, THEN EXIT.


Page 43 of 75

Rule 6 Rev 1 HPI

HPI Pump Throttling Limits

• HPI must be throttled to prevent violating the RV-P/T limit.

• HPI pump operation must be limited to two HPIPs when only one BWST suction valve

(1HP-24 or 1HP-25) is open. • HPI must be throttled ≤ 475 gpm/pump (including seal injection for A header) when only One HPI pump is operating in a header.

• Total HPI flow must be throttled ≤ 950 gpm including seal injection when 1A and 1B HPI

pumps are operating with 1HP-409 open. • Total HPI flow must be throttled < 750 gpm when all the following exist:

- LPI suction is from the RBES

- piggyback is aligned

- either of the following exist:

• only one piggyback valve is open (1LP-15 or 1LP-16)

• only one LPI pump operating • HPI may be throttled under the following conditions:

HPI Forced Cooling in Progress:

HPI Forced Cooling NOT in Progress:

All the following conditions must exist: • Core SCM > 0


All the following conditions must exist: • All WR NIs ≤ 1%

• Core SCM > 0


• SRO concurrence required if throttling following emergency boration

HPI Pump Minimum Flow Limit

• Maintain ≥ 170 gpm indicated/pump. This is an instrument error adjusted value that



Page 44 of 75

Enclosure 5.5 Rev 1 Pzr and LDST Level Control



1. Utilize the following as necessary to maintain desired Pzr level: • 1A HPI Pump • 1B HPI Pump • 1HP-26 • 1HP-7 • 1HP-120 setpoint or valve demand • 1HP-5

IF 1HP-26 will NOT open, THEN throttle 1HP-410 to maintain desired Pzr level.

2. IAAT makeup to the LDST is desired, THEN makeup from 1A BHUT.

3. IAAT it is desired to secure makeup to LDST, THEN secure makeup from 1A BHUT.

4. IAAT it is desired to bleed letdown flow to 1A BHUT, THEN perform the following:



C. Notify SRO.

5. IAAT letdown bleed is NO longer desired, THEN position 1HP-14 to NORMAL.


Page 45 of 75





__ GO TO Step 10.

7. Open: • 1HP-24 • 1HP-25




C. Open:

1LP-15

1LP-16

1LP-9

1LP-10

1LP-6

1LP-7



F. GO TO Step 8.


A. __ IF three HPI pumps are operating, THEN secure 1B HPI PUMP.

B. __ IF < 2 HPI pumps are operating, THEN start HPI pumps to obtain two HPI pump operation, preferably in opposite headers.

C. GO TO Step 9.


Page 46 of 75









Page 47 of 75




__ GO TO Step 12.





C. Open:

1LP-15

1LP-16

1LP-9

1LP-10

1LP-6

1LP-7



F. GO TO Step 13.

2. IF only one BWST suction valve (1HP-24 or 1HP-25) is open, AND three HPI pumps are operating, THEN secure 1B HPI PUMP.




Page 48 of 75






GO TO Step 35.

16. Open: __ 1CC-7 __ 1CC-8








20. GO TO Step 23.



GO TO Step 23.

22. GO TO Step 35.


Page 49 of 75







3. GO TO Step 26.

25. Open: 1HP-1 1HP-2 1HP-3 1HP-4




B. __ GO TO Step 35.

27. __ Close 1HP-6.

28. __ Close 1HP-7.


2. Close: 1HP-8 1HP-9&11




Page 50 of 75





33. Open 1HP-6.






B. __ Secure the desired HPI pumps.




Page 51 of 75




__ GO TO Step 40.




C. Close 1CS-46.





__ GO TO Step 42.




Page 52 of 75

EOP Enclosure 5.1 (ES Actuation)Rev 1


1. __ Determine all ES channels that should have actuated based on RCS pressure and RB pressure:

2. __ Verify all ES channels associated with actuation setpoints have actuated.

NOTE Voter OVERRIDE extinguishes the TRIPPED light on the associated channels that have auto actuated. Pressing TRIP on channels previously actuated will reposition components that may have been throttled or secured by this Enclosure.

__ Depress TRIP on affected ES logic channels that have NOT previously been actuated.

3. __ IAAT additional ES actuation setpoints are exceeded, THEN perform Steps 1 - 2.

4. __ Place Diverse HPI in BYPASS. __ Place Diverse HPI in OVERRIDE.







Actuation Setpoint

(psig)

Associated ES Channel

1600 (RCS) 1 & 2 550(RCS) 3 & 4 3(RB) 1, 2, 3, 4, 5, & 6 10(RB) 7 & 8


Page 53 of 75


6. __ IAAT all exist:

__ Voter associated with ES channel is in OVERRIDE

__ An ES channel is manually actuated __Components on that channel require

manipulation THEN depress RESET on the required

channel.

7. __ Verify Rule 2 in progress or complete. __ GOTO Step 74.


9. Open: __ 1HP-20 __ 1HP-21

10. __ IAAT any RCP is operating, AND ES Channels 5 and 6 actuate, THEN perform Steps 11 - 15.

__ GOTO Step 16.







12. Verify any RCP is operating. __ GOTO Step 16.





Page 54 of 75


16. __ IAAT ES Channels 3 & 4 are actuated, THEN GO TO Step 17.

__ GO TO Step 54.








CAUTION


19. __ IAAT any LPI pump is operating against a shutoff head, THEN at the CR SRO's discretion, stop affected LPI pumps. {6, 22}

20. __ IAAT RCS pressure is < LPI pump shutoff head, THEN perform Steps 21 - 22.

__ GOTO Step 23.






Page 55 of 75


23. __IAAT 1A and 1B LPI PUMPs are

off / tripped, AND all exist: __ RCS pressure < LPI pump shutoff

head __ 1LP-19 closed __ 1LP-20 closed THEN perform Steps 24 - 25.

__ GO TO Step 26.








Page 56 of 75


30. Verify open:

__ 1CF-1 __ 1CF-2







3. __ IF 1HP-20 fails to close, AND NO RCPs operating, THEN close: __ 1HP-228 __ 1HP-226 __ 1HP-232 __ 1HP-230

4. __ Notify SRO to evaluate components NOT in ES position and initiate action to place in ES position if desired.


35. __ Close 2LPSW-139.



37. __ Close 1LPSW-139.




Page 57 of 75




__ GOTO Step 42.

41. Open: __ 1LPSW-4 __ 1LPSW-5








__ GOTO Step 47.





Page 58 of 75


47. __ IAAT ES channels 7 & 8 have actuated, THEN perform Steps 48 - 49.

__ GOTO Step 50.













••• END •••


Page 59 of 75


Unit Status ES Channels 3 & 4 have NOT actuated.













61. __ Close 2LPSW-139.



63. __ Close 1LPSW-139.


Page 60 of 75






__ GOTO Step 68.

67. Open: __ 1LPSW-4 __ 1LPSW-5










••• END •••


Page 61 of 75


74. Open: __ 1HP-24 __ 1HP-25

1. __ IF both BWST suction valves (1HP-24 and 1HP-25) are closed, THEN: A. __ Start 1A LPI PUMP. B. __ Start 1B LPI PUMP. C. Open:

__ 1LP-15 __ 1LP-16 __ 1LP-9 __ 1LP-10 __ 1LP-6 __ 1LP-7



F. __ GOTO Step 75. 2. __ IF only one BWST suction valve

(1HP-24 or 1HP-25) is open, THEN: A. __ IF three HPI pumps are operating,

THEN secure 1B HPI PUMP. B. __ IF< 2 HPI pumps are operating,

THEN start HPI pumps to obtain two HPI pump operation, preferably in opposite headers.

C. __ GO TO Step 76.


Page 62 of 75


75. __ Ensure at least two HPI pumps are operating.

76. Verify open: __ 1HP-26 __ 1HP-27

1. __ IF HPI has been intentionally throttled, THEN GOTO Step 77.

2. Open: __ 1HP-26 __ 1HP-27


Page 63 of 75


77. __ IAAT at least two HPI pumps are operating, AND HPI flow in any header that has NOT been intentionally throttled is in the Unacceptable Region of Figure 1, THEN open the following in the affected header:

Figure 1

Required HPI Flow Per Header

HPI Pump

Runout Region For 1

Pump In Header

(including seal

injection for A

header)

Unacceptable Region (excluding seal

injection)

1A Header 1B Header 1HP-410 1HP-409


Page 64 of 75



79. Open: __ 1HP-20 __ 1HP-21

80. __ IAAT ES Channels 5 and 6 is actuated, THEN perform Steps 81 - 85.

__ GOTO Step 86.

81. Perform the following for actuated channels: __ Place ES CH 5 in MANUAL. __ Place ES CH 6 in MANUAL.






82. Verify any RCP operating. __ GO TO Step 86.





Page 65 of 75


86. __ IAAT ES Channels 3 & 4 are actuated, THEN GO TO Step 87.

__ GO TO Step 124.








CAUTION


89. __ IAAT any LPI pump is operating against a shutoff head, THEN at the CR SRO's discretion, stop affected LPI pumps.{6, 22}


Page 66 of 75


90. __ IAAT RCS pressure is < LPI pump shutoff head, THEN perform Steps 91-92.

__ GOTO Step 93.





93. __ IAAT 1A and 1B LPI PUMPs are off / tripped, AND all exist: __ RCS pressure < LPI pump shutoff

head __ 1LP-19 closed __ 1LP-20 closed THEN perform Steps 94 -95.

__ GO TO Step 96.






Page 67 of 75














105. __ Close 2LPSW-139.



107. __ Close 1LPSW-139.


Page 68 of 75






__ GOTO Step 112.

111. Open: __ 1LPSW-4 __ 1LPSW-5








__ GOTO Step 117.





Page 69 of 75


117. __ IAAT ES channels 7 & 8 have actuated, THEN perform Step 118 - 119.

__ GOTO Step 120.














Page 70 of 75

Unit Status ES Channels 3 & 4 have NOT actuated.













131. __ Close 2LPSW-139.



133. __ Close 1LPSW-139.


Page 71 of 75






__ GOTO Step 138.

137. Open: __ 1LPSW-4 __ 1LPSW-5










••• END •••


Page 72 of 75

Subsequent Actions EP/1/A/1800/001


CONDITION ACTIONS

1. PR NIs ≥ 5% FP

OR

NIs NOT decreasing

GO TO UNPP tab.

UNPP

2. All 4160V SWGR de-energized

{13}

GO TO Blackout tab. BLACKOUT


4. Any SCM = 0°F GO TO LOSCM tab. LOSCM

5. Both SGs intentionally isolated to stop excessive heat transfer

GO TO EHT tab.

LOHT 6. Loss of heat transfer (including loss of all Main and Emergency FDW)

GO TO LOHT tab.

7. Heat transfer is or has been excessive

GO TO EHT tab.

EHT

8. Indications of SGTR ≥ 25 gpm GO TO SGTR tab. SGTR

9. Turbine Building flooding NOT caused by rainfall event

GO TO TBF tab.

TBF


ES



ES




ROP



EDL



• Notify OSM to reference the Emergency Plan and AD-LS-ALL-0006 (Notifications/Reportability Evaluation).

NOTIFY


Page 73 of 75

EHT EP/1/A/1800/001


CONDITION ACTIONS

1. PR NIs ≥ 5% FP

OR

NIs NOT decreasing

GO TO UNPP tab.

UNPP

2. All 4160V SWGR de-energized GO TO Blackout tab. BLACKOUT


4. Any SCM = 0°F AND HPI forced cooling NOT in progress

IF LOSCM tab has NOT been entered due to current EHT event THEN GO TO LOSCM tab.

LOSCM

5. Both SGs intentionally isolated to stop excessive heat transfer after EHT tab initiated

RETURN TO beginning of EHT tab.

LOHT 6. Loss of heat transfer AND at

least one SG NOT isolated

GO TO LOHT tab.

7. Indications of excessive heat transfer in another SG after EHT tab initiated

RETURN TO beginning of EHT tab.

EHT


ES



ES




ROP


OR SGTR .> 25 gpm


EDL



• Notify OSM to reference the Emergency Plan and AD-LS-ALL-0006 (Notification /Reportability Evaluation).

NOTIFY


Page 74 of 75

CRITICAL TASKS

CT-1 The PZR Spray Valve (1RC-1) must be isolated by closing 1RC-3 to stop the RCS pressure decrease to prevent a reactor trip.

CT-2 Trip the Reactor due to two dropped control rods (unanalyzed condition). CT-3 Manually isolate Feedwater to the 1A SG prior to entering PTS conditions (Rule 8) to prevent overcooling the RCS and Reactor Core. .


SSF Operable: Yes for Units 2 and 3

KHU's Operable: U1 - OH, U2 - UG

LCTs Operable: 2 Fuel Handling: No

UNIT STATUS (CR SRO)



Reactor Power: Below POAH Mode: 1 Mode: 1







TS/SLC #

SSF 2 days ago / 0400 5 days / 0400 TS 3.10.1 Cond A,B,C,D,E Shift Turnover Items (CR SRO) Primary • Due to unanalyzed condition, the SSF should be considered INOP for Unit 1 if power levels are

reduced below 85%. Evaluations must be performed prior to declaring the SSF operable following a return to power (after going below 85%).

• 1RIA-3 and 5 removed from RB. • After turnover, the BOP is to lower 1A CFT level to ≈ 13 ft per OP/1/A/1104/001, Encl. 4.17. Then

OATC is to continue in OP/1/A/1102/001 Encl 4.7 to increase Rx power. • Increase Reactor power to ≈ 3% per OP/1/A/1102/001 Encl. 4.7 beginning at step 3.36. Once Rx

power is ≈ 3% and step 3.39 is complete, hold further power increase and CRS return to oversight role until Rx Engineering updates the maneuvering plan.

• During the Reactor power increase, Unit 2 CRS will assume the oversight role for Unit 1. Unit 1 CRS will assume the role of the dedicated Reactivity Management SRO.

Secondary

• 1SSH-1, 1SSH-3, 1SD-2, 1SD-5, 1SD-140, 1SD-303, 1SD-355, 1SD-356 and 1SD-358 are closed with power supply breakers open per the Startup Procedure for SSF Overcooling Event.

• Enclosure 13.10 (Prior To Entry Into Mode 1) of PT/1/A/0630/001 (Mode Change Verification) is complete


RCS Boron 1778 ppmB Gp 7 Rod Position: 5% Withdrawn Batch additions as required per maneuvering plan

Human Performance Emphasis (SM) Procedure Use and Adherence

RO-101b a Page 1 of 14

REGION II JOB PERFORMANCE MEASURE

RO-101b

ALIGN EMERGENCY BORATION DURING AN ATWS

Alternate Path: Yes Alt Path Failure: HP-24/25 failed / A LPI pump failed / C HPI pump failed_____________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By:

Date:

Reviewed By: Date:

Approved By: Date:


REGION II

JOB PERFORMANCE MEASURE

Task Title : Align Emergency Boration During an ATWS Task Number : N/A Alternate Path: Yes Time Critical: No Validation Time: 10 minutes K/A Rating(s): System: BW/E13 K/A: EA 1.1 Rating: 3.4/3.2 Task Standard: Candidate aligns emergency boration during an ATWS per Rule 1 (ATWS)

References: EP/1/A/1800/001, Rule 1 (ATWS) (Rev. 0L)

Tools/Equipment/Procedures Needed: EP/1/A/1800/001, Rule 1 (ATWS) (Rev. 0L)

===================================================================================== Candidate: __________________________________ Time Start: _______________

NAME Time Finish: ______________ Performance Rating: SAT ______ UNSAT ______ Performance Time: _________ Examiner: ______________________________ ___________________________________/_________ NAME SIGNATURE DATE =====================================================================================

Comments


SIMULATOR OPERATOR JPM SETUP INSTRUCTIONS

1. RECALL SNAP 211

2. IMPORT FILES for RO-101b 3. Go to RUN to acknowledge alarms

4. RESET flags on HPIPs / LPIPs

5. Go to FREEZE

6. ENSURE clean copy of Rule 1 in place on Control Board

7. Go to RUN when directed by Lead Examiner


READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

INITIAL CONDITIONS

Unit 1 is at 100% power

The CRS has determined that the reactor needs to be tripped

INITIATING CUES

The CRS directs you to trip the reactor and perform Immediate Manual Actions (IMAs).


START TIME: _______

SEQ STEP

PROC STEP DESCRIPTION

1 1

IMAs

Depress REACTOR TRIP pushbutton.

STANDARD: Candidate performs EOP IMAs from memory (per OMP 1-18) and depresses the REACTOR TRIP pushbutton located on UB1.

Candidate continues to Step 2. COMMENTS:

___ SAT

___ UNSAT

2 2

IMAs

Verify reactor power < 5% FP and decreasing.

STANDARD: Candidate performs EOP IMAs from memory and verifies reactor power < 5% and decreasing and determines that reactor power is > 5% on all NIs.

Candidate performs Step 2 RNO.

COMMENTS:

___ SAT

___ UNSAT

3 2

RNO IMAs

GO TO Rule 1 (ATWS/Unanticipated Nuclear Power Production).

STANDARD: Candidate refers to Rule 1 Step 1. COMMENTS:

___ SAT

___ UNSAT


4

1

Verify any Power Range NI ≥ 5% FP. STANDARD: Candidate verifies any Power Range NI ≥ 5% FP and

determines that all Power Range NIs are > 5% FP.

Candidate continues to Step 2.

COMMENTS:

___ SAT ___ UNSAT

5 2

Initiate manual control rod insertion to the IN LIMIT. STANDARD: Candidate initiates a manual control rod insertion by

depressing the Manual pushbutton on the Diamond control panel and then locks the Control Rod Joystick in the INSERT (forward) position.


COMMENTS:

CRITICAL STEP

___ SAT ___ UNSAT

6 3

Verify Main FDW is feeding the SGs.

STANDARD: Candidate determines Main FDW is feeding the SGs.


COMMENTS:

___ SAT ___ UNSAT

7 4

Notify CRS to GO TO UNPP tab STANDARD: Candidate notifies the CRS to go to the UNPP tab of

the EOP.


COMMENTS:

___ SAT ___ UNSAT


8 5

Open the following: __ 1HP-24 __ 1HP-25 STANDARD: Candidate locates 1HP-24 and 1HP-25 on 1UB1 and

rotates the switches to the open position. Candidate determines 1HP-24 and 1HP-25 do not open by observing the green closed lights illuminated and the red open lights off.

Candidate continues to step 5 RNO.

COMMENTS:

___ SAT ___ UNSAT

9 5

RNO

ALTERNATE PATH IF both are closed:

__ 1HP-24 __ 1HP-25 THEN GO TO Step 32. STANDARD: Candidate determines 1HP-24 and 1HP-25 are both

closed by observing the green closed lights illuminated and the red open lights off.

Candidate continues to step 32.

COMMENTS:

___ SAT ___ UNSAT


10 32

Dispatch one operator without wearing Arc Flash PPE to open 600V CRD breakers: __ 1X9-5C (U-1 CRD Norm Fdr Bkr) (U1 Equipment Rm) __ 2X1-5B (U-1 CRD Alternate Fdr Bkr) (T-3/Dd-28)

STANDARD: Candidate dispatches one operator to open Unit 1 normal and alternate CRD feeder breakers without wearing Arc Flash PPE.


BOOTH CUE: FIRE TIMER 01 to trip normal and alternate CRD breakers after 4 minute delay.

COMMENTS:

___ SAT ___ UNSAT

11 33

Start 1A LPI Pump. STANDARD: Candidate rotates the Control Switch for 1A LPI Pump

to the start position. Candidate determines the 1A LPI Pump did not start by observing the red ON lights are off and/or 0 amps are indicated on the amp meter.

Candidate continues to Step 33 RNO.

COMMENTS:

___ SAT ___ UNSAT

12 33

RNO

1. *Start 1B LPI Pump. 2. IF NO LPI pumps are operating, THEN:

a. __Open 1LP-6 b. __Open 1LP-7 c. __Start 1C LPI pump

STANDARD: Candidate rotates the Control Switch for 1B LPI Pump

to the start position. Candidate determines the 1B LPI Pump starts by observing the red ON lights are illuminated and/or amps are indicated on the amp meter.

COMMENTS:

*CRITICAL STEP

___ SAT ___ UNSAT


13 34

Open:

__*1LP-15

__*1LP-16

__*1LP-9

__*1LP-10

__1LP-6

__1LP-7

STANDARD: Candidate rotates the control switch for 1LP-15 to the

open position and verifies the green closed light off and the red open light illuminated.

Candidate rotates the control switch for 1LP-16 to the open position and verifies the green closed light off and the red open light illuminated.





Examiner Note: The Critical step is to align sufficient valves to supply one header for HPI pump suction (minimum of 1LP-9, 1LP-10, and 1LP-15 OR 1LP-16).

COMMENTS:

*CRITICAL STEP

___ SAT ___ UNSAT


14 35

Dispatch an operator to open 1HP-363 (Letdown Line to LPI Pump Suction Block) (A-1-119, U1 LPI Hatch Area, 28’ W of North door).

STANDARD: Operator is dispatched to open 1HP-363.

BOOTH CUE: Use manual valves program to open 1HP-363.

COMMENTS:

___ SAT ___ UNSAT

15 36

Ensure at least one operating: __ 1A HPI PUMP __ 1B HPI PUMP

STANDARD: Candidate ensures either the 1A or 1B HPI pump is operating.


NOTE: The 1A HPI Pump is operating.

COMMENTS:

___ SAT ___ UNSAT

16 37

Start 1C HPI PUMP.

STANDARD: Candidate rotates the 1C HPI pump switch to the

CLOSE position and determines the pump did not start by observing the white open light illuminated and the red closed lights OFF and/or amps indicating 0 on the amp meter.

Candidate continues to Rule 1 Step 37 RNO.

COMMENTS:

___ SAT ___ UNSAT


17 37

RNO

1. __ Start standby HPI pump. 2. __ IF at least two HPI pumps are operating, THEN open 1HP-409.

STANDARD: Candidate rotates the 1B HPI pump switch to the CLOSE position and determines the pump started by observing the white open light OFF and the red closed lights illuminated and/or amps indicated on the amp meter.

Candidate determines two HPI pumps are operating and opens 1HP-409 by rotating the switch to the open position and verifying the green closed light off and the red open light illuminated.

Candidate continues to Rule 1 Step 38.

COMMENTS:

CRITICAL STEP

___ SAT ___ UNSAT

18 38

Open the following: __ *1HP-26 __ 1HP-27

STANDARD: Candidate rotates the control switch to the open position for 1HP-26 and verifies the green closed light off and the red open light illuminated. (critical)

Candidate determines that 1HP-27 is already open


COMMENTS:

*CRITICAL STEP

___ SAT ___ UNSAT


19 39

Verify only two HPI pumps operating. STANDARD: Candidate verifies two HPI pumps are operating by

verifying the red ON lights are illuminated for the 1A and 1B HPI pumps and/or verify HPI pump motor amps are indicated for the operating HPI pumps.

COMMENTS:

END OF TASK

___ SAT ___ UNSAT

TIME STOP: ________


CRITICAL STEP EXPLANATIONS

SEQ STEP #

Explanation

5 This step is required to initiate a manual control rod insertion and start adding negative

reactivity to the core.

12 This step is required to align emergency boration from the BWST to add negative reactivity to the core.

13 This step is required to align emergency boration from the BWST to add negative

reactivity to the core.

17 This step is required to ensure HPI flow in the "B" HPI header.

18 This step is required to ensure HPI flow in the "A" HPI header.

a

CANDIDATE CUE SHEET

(TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS

Unit 1 is at 100% power

The CRS has determined that the reactor needs to be tripped

INITIATING CUES

The CRS directs you to trip the reactor and perform Immediate Manual Actions (IMAs)

RO-204 h Page 1 of 11


RO-204

Align Letdown With 1HP-14 Failed in Bleed

Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Align letdown with 1HP-14 failed in "Bleed” Task Number : Alternate Path: No Time Critical: No Validation Time: 14 minutes K/A Rating(s): System: 002 K/A: A2.01 Rating: 4.3/4.4 Task Standard: Students align letdown with 1HP-14 failed in the ‘Bleed’ position IAW AP/1/A/1700/002 (Excessive RCS Leakage) References: AP/1/A/1700/002 (Excessive RCS Leakage) EOP Enclosure 5.5 (PZR and LDST Level Control) Tools/Equipment/Procedures Needed: AP/1/A/1700/002 (Excessive RCS Leakage) (Rev. 15) EOP Enclosure 5.5 (PZR and LDST Level Control) (0M, Rev. 01) ===============================================================================

Candidate: __________________________________ Time Start: _______________


Comments



1. RECALL SNAP 212 2. IMPORT RO-204 simulator files

3. ENSURE clean procedures in place for candidate

4. Go To RUN when directed by the Lead Examiner

================================================================================

Directions without a SNAP: 1. Initialize simulator to desired IC. (Example Recall IC-41 100% EOL)

2. Take any required action: (Example: Perform EOP Encl.5.1, stop RCPs,

open pump breakers, throttle EFDW flow etc.)

3. Activate Malfunctions, Build Timers and Overrides as required: (Example: Activate MPS400 @ 100% on AOR)

4. Acknowledge alarms and Freeze the Simulator at desired point 5. Save to a SNAP Number and password protect, as desired.


READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you. INITIAL CONDITIONS • Reactor Power is 100%

• AP/1/A/1700/002 is in progress due to 1HP-14 failing in the BLEED position

• Another RO is making up to the LDST per Encl 5.5 of the EOP INITIATING CUE The CRS directs you to continue with AP/1/A/1700/002 (Excessive RCS Leakage) beginning at step 4.155


START TIME: _______

SEQ STEP


1 4.155

Verify 1A LD Filter in service.

STANDARD: Student observes 1B Letdown (LD) Filter is in service and the 1A LD Filter is NOT in service using either the OAC or Control Board indications and goes to the RNO.

COMMENTS:

___ SAT

___ UNSAT

2 4.155 RNO

1. IF 1A LD Filter is OOS for maintenance, THEN restore 1A LD Filter per in progress procedure. (N/A)

2. Open 1HP-17

STANDARD: Student recognizes that 1A LD Filter is available.

*Student opens 1HP-17 and observes the red open light ON and the green closed light OFF.

Examiner Cue: If asked, notify candidate that 1A Letdown filter is NOT OOS for maintenance.

COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT


3 4.156

Close 1HP-6

STANDARD: Student closes 1HP-6 and observes the red open light OFF and the green closed light ON.

COMMENTS:

___ SAT

___ UNSAT

4 4.157

Adjust 1HP-7, as needed, to control:

• BLEED flow out of failed 1HP-14

• Pzr level

STANDARD: Student adjusts 1HP-7 as necessary to control PZR level.

COMMENTS:

___ SAT

___ UNSAT

5 4.158

Dispatch an operator to open 1HP-196 (Filter Diversion Inlet) (A-2-LDST Hatch Area)

STANDARD: Student dispatches an AO to open 1HP-196

BOOTH CUE: Open 1HP-196 with Manual Valves

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

NOTE TS 3.4.9 applies when indicated PZR Level > 260˝ (corrected value for 285˝).


6 4.159

Verify CC System in Operation.

STANDARD: Student observes:

• At least one CC pump in operation • CC return flow is > 575 gpm

COMMENTS:

___ SAT

___ UNSAT

7 4.160

Position the standby HPI pump switch to OFF.

STANDARD: Student places the standby (1B) HPIP switch to OFF

COMMENTS:

___ SAT

___ UNSAT

8 4.161

Initiate monitoring RCP parameters.

STANDARD: Student refers to OAC Turn-on Code RCP and control board indications to monitor RCP parameters.

Booth Cue: Notify the candidate that 1HP-196 is open. COMMENTS:

___ SAT

___ UNSAT


9 4.162

Throttle 1HP-31 to establish 12-15 gpm SEAL INLET HEADER FLOW.

STANDARD: Student throttles 1HP-31 to establish 12-15 gpm.

COMMENTS:

___ SAT

___ UNSAT

10 4.163

WHEN 1HP-196 (Filter Diversion Inlet) (A-2-LDST Hatch Area) is open, THEN close 1CS-26.

STANDARD: Student closes 1CS-26 and observes the red open light OFF and the green closed light ON.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

11 4.164

Close the following;

• 1CS-27 • 1CS-32 & 37

STANDARD: Student verifies 1CS-27, 1CS-32, and 1CS-37 are closed by observing the red open lights OFF and the green closed lights ON.

COMMENTS:

___ SAT

___ UNSAT


12 4.165

Open 1HP-6

STANDARD: Student opens 1HP-6 and observes the red open light ON and the green closed light OFF.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

13 4.166

Throttle 1HP-31 to establish ≈ 32 gpm SEAL INLET HDR FLOW.

STANDARD: Student throttles open 1HP-31 to establish ≈ 32 gpm SEAL INLET HDR FLOW.

COMMENTS:

___ SAT

___ UNSAT

14 4.167

Adjust 1HP-7 to establish desired letdown flow.

STANDARD: Student throttles open 1HP-7 to establish ≈ 70 - 80 gpm letdown flow.

COMMENTS:

___ SAT

___ UNSAT

15 4.168

Position the standby HPI pump switch to AUTO

STANDARD: Student places the standby (1B) HPIP switch to AUTO

EXAMINER CUE: Notify the candidate that another operator will continue with this procedure.

COMMENTS:

END TASK

CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

2 This step is required to align the alternate letdown flow path. 5 This step is required to align the alternate letdown flow path.

10 This step is required to isolate the flow path going to the BHUTs.

12 This step is required to return letdown flow to normal.

15 This step is required to protect the RCP seals by allowing the standby HPIP to automatically start if low seal injection flow (< 22 gpm) is detected.

CANDIDATE CUE SHEET


INITIAL CONDITIONS • Reactor Power is 100%

• AP/1/A/1700/002 is in progress due to 1HP-14 failing in the BLEED position

• Another RO is making up to the LDST per Encl 5.5 of the EOP INITIATING CUE The CRS directs you to continue with AP/1/A/1700/002 (Excessive RCS Leakage) beginning at step 4.155

RO-501a c Page 1 of 12


RO-501a

ES CHANNELS 5 AND 6 RECOVERY

Alternate Path: Yes Alt Path Failure: ______1A/1B CC Pumps, 1HP-5__________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By:

Date:

EP Review By: N/A Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : ES Channels 5 and 6 Recovery

Task Number : Alternate Path: Yes Time Critical: No Validation Time: 10 minutes K/A Rating(s): System: 103 K/A: A4.04 Rating: 3.5*/3.5* Task Standard: ES Channels 5 and 6 are returned to normal using EOP Enclosure 5.41 (ES Recovery) and Letdown isolated using AP/1/A/1700/020 (Loss of Component Cooling).

References: EOP Encl. 5.41 (ES Recovery)

AP/1/A/1700/020 (Loss of Component Cooling)

Tools/Equipment/Procedures Needed: EOP Encl. 5.41 (ES Recovery) (Rev. 0Q)

AP/1/A/1700/020 (Loss of Component Cooling) (Rev. 12)



Comments



1. RECALL SNAP 215 2. PLACE Letdown Temp Interlock Bypass to BYPASS to exit static check, then place back in

NORMAL

3. IMPORT RO-501a files

4. Go to RUN and acknowledge alarms

5. Go to FREEZE

6. ENSURE clean procedures (EOP Encl. 5.41 and AP/20) in place for candidate



READ TO OPERATOR


INITIAL CONDITIONS A MSLB has occurred

ES Channels 1 - 6 have actuated on High RB pressure

EOP Encl. 5.1 (ES Actuation) is complete

ALL ES Channels are being reset

EOP Encl. 5.41 (ES Recovery) is complete through step 40

INITIATING CUES

The CRS directs you to continue in EOP Encl. 5.41 (ES Recovery) beginning at step 41 to reset ES Channels 5 and 6.


START TIME: _______

SEQ STEP


1 41

Verify reset of any of the following is desired:

• ES Channel 5 • ES Channel 6

STANDARD: Determines, from Initiating Cue, that reset of ES Channels 5 and 6 is desired.

COMMENTS:

___ SAT

___ UNSAT

2 42

Verify Stat Alarms have cleared:

• 1SA-7/A-1 (1A1 ES TRIP) • 1SA-7/B-1 (1B1 ES TRIP) • 1SA-7/C-1 (1C1 ES TRIP) • 1SA-7/A-2 (1A2 ES TRIP) • 1SA-7/B-2 (1B2 ES TRIP) • 1SA-7/C-2 (1C2 ES TRIP)

STANDARD: Determine that all of the above statalarms on 1SA-7 are clear by observing that their respective window lights are NOT illuminated.

COMMENTS:

___ SAT

___ UNSAT

3 43

Perform the following to reset ES (1UB1):

• Depress RESET for CH 5. • Depress RESET for CH 6.

STANDARD: Candidate depresses the RESET pushbuttons for Channels 5 and 6 located on 1UB1

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


4 44

Verify:

• CH 5 TRIPPED light off • CH 6 TRIPPED light off

STANDARD: Verifies white tripped lights OFF for ES Channels 5 and 6

COMMENTS:

___ SAT

___ UNSAT

5 45

Open:

• 1CC-7 • 1CC-8

STANDARD: Candidate verifies 1CC-7 and 1CC-8 are open by observing RED open lights ON and GREEN closed lights OFF.

COMMENTS:

___ SAT

___ UNSAT

6 46

Ensure one CC pump is running in the ON position.

STANDARD: Rotates the switch for the 1B CC Pump to ON.

EXAMINER NOTE: The 1B CC Pump will trip when the switch is placed in the ON position.

COMMENTS:

___ SAT

___ UNSAT


7 46

RNO

ALTERNATE PATH

IF NO CC pumps are operating, THEN perform the following:

1. Initiate AP/20 (Loss of Component Cooling).

2. GO TO Step 48.

STANDARD: Candidate initiates AP/20 (Loss of Component Cooling) from either step 46 RNO or 1SA-9/B-1 or 1SA-9/C-1.

EXAMINER CUE: Once AP/20 is initiated, inform candidate the

CRS directs them to perform AP/20. COMMENTS:

___ SAT

___ UNSAT

8 AP/20

3.1

IAAT both of the following are lost:

• CC to RCPs • RCP seal injection THEN perform the following: A. Trip Rx B. Stop all RCPs C. Initiate AP/25 (SSF EOP)

STANDARD: Determine RCP seal injection is available and proceed to step 3.2.

COMMENTS:

___ SAT

___ UNSAT


9 3.2

NOTE If CRD stator cooling is lost, stator temperatures will reach 180°F in ≈ 4 minutes. CRD stator temperatures can be read on the OAC, turn-on code "1CRDTP".

IAAT ≥ two CRD stator temperatures ≥ 180°F,

THEN trip Rx.

STANDARD: Determine Rx is already tripped and proceed to step 4.1.

COMMENTS:

___ SAT

___ UNSAT

10 4.1

Verify at least one CC Pump operating.

STANDARD: Determine NO CC Pumps are operating. Proceed to RNO.

COMMENTS:

___ SAT

___ UNSAT

11 4.1

RNO

1. IF either of the following:

• CC Pumps have been manually secured due to CC system leakage

• AP/2 directed initiation of AP/20 THEN GO TO step 4.12.

2. GO TO step 4.3

STANDARD: Determine CC Pumps were NOT secured due to system leakage and AP/2 did NOT direct initiation of AP/20. GO TO step 4.3.

COMMENTS:

___ SAT

___ UNSAT


12 4.3

Open:

• 1CC-7 • 1CC-8

STANDARD: Candidate verifies 1CC-7 and 1CC-8 are open by observing red open lights ON and green closed lights OFF.

COMMENTS:

___ SAT

___ UNSAT

13 4.4

Verify > one CC Pump operating.

STANDARD: Determine NO CC Pumps are operating. Proceed to RNO.

COMMENTS:

___ SAT

___ UNSAT


14 4.4

RNO

1. IF CC Surge Tank level > 12",

THEN attempt to start a CC Pump (will NOT start) 2. IF unable to start any CC Pump,

THEN perform the following:

A. Close 1HP-5 (will NOT close) B. IF 1HP-5 fails to close

THEN perform the following:

• *Close 1HP-3 (critical) • *Close 1HP-4 (critical)

C. Notify WCC SRO to initiate actions to regain a CC Pump.

D. Initiate AP/32 (Loss of Letdown).

STANDARD: Verify CC Surge Tank level > 12" and attempt to start a CC Pump. Determine neither CC Pump will start.

Place 1HP-5 switch to close. Determine 1HP-5 will not close by observing the red open light ON and the green closed light OFF.

Place the switches for 1HP-3 and 4 to close. Determine 1HP-3 and 4 are closed by observing the green closed light ON and the red open light OFF.

Notify WCC SRO to initiate actions to regain a CC Pump.

Initiate AP/32. EXAMINERS CUE: Another operator will continue with this

procedure.

COMMENTS:

END TASK

*CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

3 This step is required to reset ES channels 5 & 6 so the associated components can be

re-aligned for normal operation.

14

This step is required to isolate letdown.

c

CANDIDATE CUE SHEET


INITIAL CONDITIONS A MSLB has occurred

ES Channels 1 - 6 have actuated on High RB pressure

EOP Encl. 5.1 (ES Actuation) is complete

ALL ES Channels are being reset

EOP Encl. 5.41 (ES Recovery) is complete through step 40

INITIATING CUES

The CRS directs you to continue in EOP Encl. 5.41 (ES Recovery) beginning at step 41 to reset ES Channels 5 and 6.

RO-604 g Page 1 of 11


RO-604

Perform a Manual Start of KHU 1


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Perform A Manual Start Of Keowee Hydro Unit 1

Task Number : Alternate Path: No Time Critical: No Validation Time: 15 minutes K/A Rating(s):

System: 062 K/A: A4.07 Rating: 3.1/3.1

Task Standard:

Keowee Hydro Unit 1 is manually started and synchronized to the 230 KV switchyard.

References:

OP/0/A/1106/019 (Keowee Hydro At Oconee)

Tools/Equipment/Procedures Needed:

OP/0/A/1106/019 (Keowee Hydro At Oconee), Encl 4.5 (KHU-1 Manual Startup) (Rev. 101)

===============================================================================



Comments



1. RECALL SNAP 216 2. IMPORT RO-604 simulator files

3. ENSURE KHU 1 in REMOTE

4. Go To RUN when directed by the Lead Examiner

================================================================================







READ TO OPERATOR


INITIAL CONDITIONS

Keowee Unit 1 is aligned to the Overhead.

Keowee Unit 1 is to be manually started and tied to the grid.

This is NOT an emergency.

OP/0/A/1106/19, Enclosure 4.5 (KHU-1 Manual Startup) is in progress

INITIATING CUE

The CRS directs you to continue with a manual start of Keowee Unit 1, and tie it to the grid with no load, per OP/0/A/1106/019, Enclosure 4.5 beginning at Step 2.1.


START TIME: _______

SEQ STEP


1 2.1

NOTE: AC HP Lift Pump starts and Generator Cooling Water valve opens when Unit 1 MASTER SELECTOR in MAN.

Place UNIT 1 MASTER SELECTOR in ”MAN"

STANDARD: Candidate positions UNIT 1 MASTER SELECTOR

switch to "MAN".

Continue to Step 2.2

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2 2.2

Place UNIT 1 SYNC 230 KV selector in "MAN".

STANDARD: Candidate positions UNIT 1 SYNC 230 KV selector

switch to "MAN".


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


3 2.3

IF this is NOT an emergency situation, notify Keowee operator to activate the Tailrace Rising Water Alarm Test button on CB-5.

STANDARD: Candidate determines this is not an emergency

situation and notifies Keowee Operator to activate the Tailrace Rising Water Alarm Test button on CB-5.

Booth cue: Keowee operator (Casey Holder) reports Tailrace Rising Water Alarm Test button on CB-5 is active.


COMMENTS:

___ SAT

___ UNSAT

4 2.4

Place AND hold UNIT 1 LOCAL MASTER switch to "START" position for > 10 seconds until KHU starts.

STANDARD: Candidate rotates the UNIT 1 LOCAL MASTER switch to the "START" position for >10 seconds and verifies that the Turb 1 Gate Position increases on 2AB3.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


5 2.5

WHEN TURBINE 1 GATE POSITION indicator settles out:

• Momentarily place EXCITER STOP/START to "START". • Verify EXCITER STOP/START Red START light ON,

Green STOP light OFF. STANDARD: Candidate observes that the TURBINE 1 GATE

POSITION has settled out and then: o *Momentarily places EXCITER STOP/START to

"START". o Observes EXCITER STOP/START Red START

light ON, Green STOP light OFF and GEN FREQ about 60 HZ and Keowee 1 Output volts about 13.8 KV.


COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT


6 2.6

IF required, synchronize KHU-1 to the grid:

2.6.1 Verify open KPF-9 (KHU 1 FDR)

2.6.2 Make the following adjustments concurrently as required:

• Using UNIT 1 AUTO VOLTAGE ADJUSTER, match KEOWEE 1 LINE VOLTS with KEOWEE 1 OUTPUT VOLTS when 13.8 KV SYNCHROSCOPE pointer is vertical.

• Adjust KHU-1 Generator speed with UNIT 1 SPEED CHANGER MOTOR until slow clockwise rotation of 13.8 KV SYNCHROSCOPE is established.

CAUTION: To prevent possible equipment damage, it is important to realize that 5° is NOT equivalent to 5 minutes.

2.6.3 WHEN 13.8 KV SYNCHROSCOPE pointer is ≈ 5° before reaching vertical, close ACB 1 KEOWEE 1 GENERATOR BKR.

STANDARD: 2.6.1 Candidate verifies KPF-9 is open by observing the red

closed light OFF and the white open light ON.

2.6.2 Candidate determines that KHU-1 is required to be synced to the grid and makes the following adjustments:

• UNIT 1 AUTO VOLTAGE ADJUSTER manipulated to match KEOWEE 1 LINE VOLTS with KEOWEE 1 OUTPUT VOLTS when 13.8 KV SYNCHROSCOPE pointer is vertical.

• KHU-1 Generator speed adjusted with UNIT 1 SPEED CHANGER MOTOR until slow clockwise rotation of 13.8 KV SYNCHROSCOPE is established.

2.6.3 ACB 1 KEOWEE 1 GENERATOR BKR switch is taken to close when the 13.8 KV SYNCHROSCOPE pointer is ≈ 5° before vertical, and candidate observes the red close light ON and the white open light OFF.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


7 2.7

CAUTION: KHU-1 load should NOT be reduced below indicated zero (0) MWs.

IF required, perform the following concurrently:

• Adjust load to zero (0) MWs with UNIT 1 SPEED CHANGER MOTOR.

• Adjust MVARS to zero (0) with UNIT 1 AUTO VOLTAGE ADJUSTER.

STANDARD: Candidate manipulates the UNIT 1 SPEED

CHANGER MOTOR as required to obtain (0) MWs and the UNIT 1 AUTO VOLTAGE ADJUSTER as required to obtain ≈ zero (0) MVARS.

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 This step is required to manually start KHU-1. 2 This step is required to manually start KHU-1. 4 This step is required to start KHU-1. 5 This step is required to excite the KHU-1 generator. 6 This step is required to synchronize KHU-1 to the grid.

CANDIDATE CUE SHEET


INITIAL CONDITIONS

Keowee Unit 1 is aligned to the Overhead.

Keowee Unit 1 is to be manually started and tied to the grid.

This is NOT an emergency.

OP/0/A/1106/19, Enclosure 4.5 (KHU-1 Manual Startup) is in progress

INITIATING CUE

The CRS directs you to continue with a manual start of Keowee Unit 1, and tie it to the grid with no load, per OP/0/A/1106/019, Enclosure 4.5 beginning at Step 2.1.

RO-702 - d Page 1 of 12


RO-702

Adjust Radiation Monitor Setpoints


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:

RO-702 - d Page 2 of 12

REGION II


Task Title : Adjust Radiation Monitor Setpoints Task Number : Alternate Path: No Time Critical: No Validation Time: 25 minutes K/A Rating(s):

System: 073 K/A: A4.01 Rating: 3.9 / 3.9 Task Standard: Adjust 1RIA-37 and 1RIA-38 for a GWD release per procedure.

References: OP/1-2/A/1104/018, GWD system PT/0/A/230/01, Radiation Monitor Check

Tools/Equipment/Procedures Needed: OP/1-2/A/1104/018, GWD system, Encl. 4.9 and 4.10 (Rev. 78) PT/0/A/230/01, Radiation Monitor Check, Encl. 13.6 (Rev. 172)

PT/1-2/A/0230/002, GWD Tank Release (Rev. 6)

===============================================================================



Comments

RO-702 - d Page 3 of 12


1. RECALL SNAP 217

2. ENSURE procedures in place for candidate

3. Select plant HMI screen to GWD system


================================================================================






RO-702 - d Page 4 of 12

READ TO OPERATOR


INITIAL CONDITIONS Release of “B” GWD tank is desired No other GWRs in progress OP/1-2/A/1104/018, Enclosure 4.9 (GWD Tank Release) is in progress and completed up to Step 3.4 INITIATING CUES SRO directs you to continue the release of the “B” GWD tank beginning at Step 3.4 of Enclosure 4.9

RO-702 - d Page 5 of 12

START TIME: _______

SEQ STEP


1 3.4

Perform the following on Enclosure 4.10 (GWD Tank Sample Request):

3.4.1 Record Initial GWD Tank Pressure AND GWD Tank Volume

3.4.2 Record Start Date AND Time

3.4.3 Record background readings for 1RIA-37 AND 1RIA-38

STANDARD: Determine B GWD tank pressure is 65.3 psig by looking on either the OAC on the chart on 1VB1 and record on Enclosure 4.10.

Determine GWD Tank Volume is ≈ 6000 Ft3 by using the curve in OP/1108/001 (General Curves and Information) and record on Enclosure 4.10.

Determine background readings for 1RIA-37 & 1RIA-38 by observing the RIA readings on the RIA View Node and record on Enclosure 4.10.

Note: The background readings should be as follows: • 1RIA-37 Background reading is: ≈ 367 or 3.67E2 cps • 1RIA-38 Background reading is: ≈ 11 or 1.1E1 cps


COMMENTS:

___ SAT

___ UNSAT

2 3.5

Record recommended 1RIA-37 and 38 Alert and High setpoints from Enclosure 4.10 (GWD Tank Sample Request):

• 1RIA-37 3.47 E5 cpm above background

• 1RIA-38 2.78 E2 cpm above background

STANDARD: Obtain this information from the sample request and record in the procedure.


COMMENTS:

___ SAT

___ UNSAT

RO-702 - d Page 6 of 12

3 3.6

IF 1RIA-37 is out-of-service OR GWD piping NOT purged, verify the following per Enclosure 4.10 (GWD Tank Sample Request):

• Independent Data Entry Checks completed

• Independent Sample agrees with initial sample

STANDARD: Determine step does not apply and N/A the step.


Examiner Cue: If asked, inform the candidate that the GWD piping is purged.

COMMENTS:

___ SAT

___ UNSAT

4 3.7

NOTE: If N2 was added to the most recently released GWD tank until 1RIA-37 indicated < 700 cpm, OR if the tanks’ radioactivity was < 2.1E-05µCi/ml when it was released, the GWD piping is considered “purged”.

IF GWD piping purged, calculate actual setpoints as follows:

3.7.1 IF 1RIA-37 Operable, calculate Alert and High setpoints as follows:

3.67E2 cpm + 3.47E5 = 3.47E5 cpm

1RIA-37' Add to background' 1RIA-37 background value from Step 3.5 Alert and High alarm setpoints

STANDARD: Calculate set points using formula above. (3.47E5 to 3.48E5)


COMMENTS:

CRITICAL STEP

___ SAT ___UNSAT


RO-702 - d Page 7 of 12

5 3.8

IF 1RIA-38 Operable, calculate Alert and High setpoints as follows:

1.1E1 cpm + 2.78E2 = 2.89E2 cpm

1RIA-38' Add to background' 1RIA-38 background value from Step 3.5 Alert and High alarm setpoints STANDARD: Calculate set points using formula above (2.89E2).


COMMENTS:

CRITICAL STEP

___ SAT ___UNSAT

RO-702 - d Page 8 of 12

6 3.9

Adjust 1RIA-37 setpoints for release as follows:

3.9.1 Perform one of the following:

A. IF all of the following: • Calculated setpoints are < 1E+07 CPM • 1RIA-37 operable • GWD pipping purged Set alarms as follows: • Set 1RIA-37 Alert setpoint at 3.47E5 cpm per

PT/0/A/0230/001 (Radiation Monitor Check). (from Step 3.7.1) • Set 1RIA-37 High setpoint at 3.47E5 cpm per

PT/0/A/0230/001 (Radiation Monitor Check). (from Step 3.7.1)

NOTE: If GWD piping NOT purged on most recent release, RP independently verifies release data and 1RIA-37 setpoint is set at zero to allow the tank to be released.

B. IF any of the following: (N/A) • Calculated setpoints are > 1E+07 CPM • 1RIA-37 out-of-service • GWD piping NOT purged

STANDARD: Determine that calculated setpoints are < 1E+07 CPM, 1RIA-37 is operable, and GWD piping is purged and set 1RIA-37 alert and high setpoints at 3.47E5 per PT/0/A/0230/001 (see Step 7).

(3.47E5 to 3.48E5)


COMMENTS:

___ SAT

___ UNSAT


RO-702 - d Page 9 of 12

7

PT/0/A/0230/001 Enclosure 13.6 (1RIA-37 and 1RIA-38 Setpoints)

1.1 Verify Unit 1&2 GWD Tank release planned per OP/1-2/A/1104/018 (GWD System)

NOTE: Each RIA procedure section may be performed in any sequence.

2.1 Set process monitor setpoints as required

3.1 Document 1RIA-37 Alert setpoint per OP/1-2/A/1104/018: 3.47E5 CPM

3.2 Insert 1RIA-37 Alert setpoint from Step 3.1

3.3 Document 1RIA-37 High setpoint per OP/1-2/A/1104/018: 3.47E5 CPM

3.4 Insert 1RIA-37 High setpoint from Step 3.3

STANDARD: Candidate documents the Alert and High setpoints for 1RIA-

37 per OP/1-2/A/1104/018 (3.47E5 to 3.48E5)

Candidate inserts the Alert and High setpoints into 1RIA-37 (3.47E5 to 3.48E5)

COMMENTS:

CRITICAL STEP

___ SAT ___UNSAT

RO-702 - d Page 10 of 12

8 3.10

Adjust 1RIA-38 setpoints for release as follows:

3.10.1 Perform one of the following:

A. IF calculated setpoints are < 1E+06 CPM AND 1RIA-38 operable, set alarms as follows: • Set 1RIA-38 Alert setpoint at 2.89E2 cpm per

PT/0/A/0230/001 (Radiation Monitor Check). (from Step 3.8) • Set 1RIA-38 High setpoint at 2.89E2 cpm per

PT/0/A/0230/001 (Radiation Monitor Check). (from Step 3.8)

NOTE: Due to type of radiation 1RIA-38 monitors, it is NOT sensitive enoughto perform an adequate N2 purge.

B. IF calculated setpoints are > 1E+06 CPM OR 1RIA-38 out of

service, perform the following: (N/A)

STANDARD: Candidate documents the Alert and High setpoints for 1RIA-38 per OP/1-2/A/1104/018 (2.89E2)

Candidate inserts the Alert and High setpoints into 1RIA-37 (2.89E2) per PT/0/A/0230/001 (per Step 9)

COMMENTS:

___ SAT

___ UNSAT

9

PT/0/A/0230/001 Enclosure 13.6 (1RIA-37 and 1RIA-38 Setpoints)

4.1 Document 1RIA-38 Alert setpoint per OP/1-2/A/1104/018: 2.89E2 CPM

3.2 Insert 1RIA-38 Alert setpoint from Step 4.1

3.3 Document 1RIA-38 High setpoint per OP/1-2/A/1104/018: 2.89E2 CPM

3.4 Insert 1RIA-38 High setpoint from Step 4.3

STANDARD: Candidate documents the Alert and High setpoints for 1RIA-

38 per OP/1-2/A/1104/018 (2.89E2)

Candidate inserts the Alert and High setpoints into 1RIA-38 (2.89E2)

COMMENTS:

END TASK

CRITICAL STEP

___ SAT ___UNSAT

TIME STOP: ________

RO-702 - d Page 11 of 12


SEQ STEP #

Explanation

4 Required to ensure proper setpoint is calculated. 5 Required to ensure proper setpoint is calculated. 7 Required to set RIA to prevent station release above limits. 9 Required to set RIA to prevent station release above limits.

CANDIDATE CUE SHEET


INITIAL CONDITIONS Release of “B” GWD tank is desired No other GWRs in progress OP/1-2/A/1104/018, Enclosure 4.9 (GWD Tank Release) is in progress and completed up to Step 3.4 INITIATING CUES SRO directs you to continue the release of the “B” GWD tank beginning at Step 3.4 of Enclosure 4.9

RO-805a f Page 1 of 18


RO-805a

Perform Required Actions For a Turbine Building Flood

Alternate Path: Yes Alt Path Description: ______1CCW-12 will not close_______________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Perform Required Actions For A Turbine Building Flood

Task Number : Alternate Path: Yes Time Critical: No Validation Time: 15 minutes K/A Rating(s):

System: APE BW/A07 K/A: AA1.3 Rating: 3.3/3.5 Task Standard: Perform required actions for a Turbine Building Flood in accordance with AP/10, Turbine Building Flood

References:

AP/1/A/1700/01010, Turbine Building Flood

Tools/Equipment/Procedures Needed: AP/1/A/1700/01010, Turbine Building Flood (Rev. 10)

===============================================================================



Comments



1. RECALL SNAP 218 2. IMPORT files for RO-805a

3. ENSURE clean procedure in place for candidate

4. Go to RUN when directed by the Lead Examiner

================================================================================







READ TO OPERATOR


INITIAL CONDITIONS

Unit 1 was operating at 100% power

Unit 2 is in cold shutdown

The 2A1 condenser inlet expansion joint has ruptured and Turbine Building Basement water level is rising

Unit 1 has been tripped and the EOP is being completed by other team members.

INITIATING CUE

The CRS directs you to initiate Unit 1 AP/10 (Turbine Building Flood).


START TIME: _______

SEQ STEP


1 4.1

Trip the Rx.

STANDARD: Depress the Rx trip pushbutton on 1UB1. Verify the RX is tripped by observing all control rods are inserted.

Continue to Step 4.2.

COMMENTS:

___ SAT

___ UNSAT

2 4.2

Place 1CCW-1-6 switch in PULL TO LOCK.

STANDARD: Candidate locates and places the control switch for 1CCW-1-6 in PULL TO LOCK.


COMMENTS:

___ SAT

___ UNSAT


3 4.3

NOTE

Tripping the CCW pumps will cause loss of condenser vacuum which will result in TBVs failing closed and trip of the MFDWPs.

Ensure all CCW pumps are shutdown.

STANDARD: Candidate rotates the 1A, 1B, and 1C CCW pump switches to the TRIP position and verifies the red close lights extinguished and the white open lights illuminated. (located on 1AB3)


Examiner Cue: If the candidate initiates AP/24, notify them that another operator will perform AP/24

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

4 4.4

Press both of the following on the TURB BLDG FLOOD EMER CLOSING ALL CCWP DISCH VALVES control switch:

• CLOSE 1

• CLOSE 2 STANDARD: "CLOSE 1" and "CLOSE 2" on the TURB BLDG

FLOOD EMER CLOSING ALL CCWP DISCH VALVES control switch are both depressed. (located on 1AB3)


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


5 4.5

Verify all the following closed.

__ 1CCW-10

__ 1CCW-11

__ 1CCW-12

__ 1CCW-13

STANDARD: The green CLOSED lights for 1CCW-10, 11, and 13 are verified illuminated and the red OPEN lights for the same valves are verified extinguished. (located on 1AB3)

Determine that 1CCW-12 is NOT closed by observing the red open light lit on 1AB3.

Continue to Step 4.5 RNO.

COMMENTS:

___ SAT

___ UNSAT


6 4.5

RNO

ALTERNATE PATH

NOTE

The control switches for the CCW pump discharge valves do not seal-in. The valves act as throttle valves unless the associated CCW pump switch is in the TRIP position.

Dispatch an operator to ensure all CCW pump discharge valves are closed (Equipment Rm):

• 1XS1-F2C (1CCW-10 Bkr 1A CCW Pump Disch)

• 1XS1-F3C (1CCW-13 Bkr 1D CCW Pump Disch)

• 1XS2-F2D (1CCW-11 Bkr 1B CCW Pump Disch)

• 1XS3-2E (1CCW-12 Bkr 1C CCW Pump Disch)

STANDARD: An operator is dispatched to the Equipment Room to ensure 1CCW-12 is closed.


Booth cue: After 2 minutes call as the AO and inform candidate that 1CCW-12 will NOT close.

COMMENTS:

___ SAT

___ UNSAT


7 4.6

NOTE

Field tasks that may affect multiple units should be coordinated with Unit 2 and Unit 3.

Dispatch operators to ensure all three flood doors are closed and in the SECURED position from the AB side (A-1, at Unit 1, 2, and 3 stairwell entrances to TB): {2}

• Unit 1 Flood Door



STANDARD: Dispatch an AO to ensure all three flood doors are closed and in the SECURED position from the AB side.


Booth cue: When called, inform candidate that an operator has been dispatched.

COMMENTS:

___ SAT

___ UNSAT


8 4.7

Dispatch an operator to position the waterbox discharge valve switches (T-3, East of condenser catwalk):

• 1CCW-20 (1A1) switch to HAND • 1CCW-20 (1A1) switch to CLOSE • 1CCW-21 (1A2) switch to HAND • 1CCW-21 (1A2) switch to CLOSE • 1CCW-22 (1B1) switch to HAND • 1CCW-22 (1B1) switch to CLOSE • 1CCW-23 (1B2) switch to HAND • 1CCW-23 (1B2) switch to CLOSE • 1CCW-24 (1C1) switch to HAND • 1CCW-24 (1C1) switch to CLOSE • 1CCW-25 (1C2) switch to HAND • 1CCW-25 (1C2) switch to CLOSE

STANDARD: Candidate should dispatch an operator to place the

above valves to HAND and CLOSE.


Booth Cue: When called, inform candidate that an operator has been dispatched.

COMMENTS:

___ SAT

___ UNSAT


9 4.8

Verify all the following closed.

__ 1CCW-10

__ 1CCW-11

__ 1CCW-12

__ 1CCW-13

STANDARD: The green CLOSED lights for 1CCW-10, 11, and 13 are verified illuminated and the red OPEN lights for the same valves are verified extinguished. (located on 1AB3)

Determine that 1CCW-12 is NOT closed by observing the red open light lit on 1AB3.

Continue to Step 4.8 RNO.

COMMENTS:

___ SAT

___ UNSAT


10 4.8

RNO

ALTERNATE PATH

NOTE

If the pipe cap cannot be removed by hand, then there is a strap wrench in the black tool box which is staged on the CCW Intake structure near catwalk.

Dispatch an operator to perform the following (ESV Trench, Intake):

• Remove pipe cap at 1CCW-26 (CCW Inlet High Point Vent).

• Open 1CCW-26 (CCW Inlet High Point Vent).

• Remove pipe cap at 1CCW-28 (CCW Inlet High Point Vent).

• Open 1CCW-28 (CCW Inlet High Point Vent).

STANDARD: Candidate should dispatch an operator to position the above valves as directed.



COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

11 4.9

Open 1V-186

STANDARD: Candidate rotates the switch for 1V-186 to the open position until the red open light is illuminated and the green closed light is extinguished.

COMMENTS:

___ SAT

___ UNSAT


12 4.10

Make a PA announcement of the event including restricting entry into TB basement due to electrical hazard.

STANDARD: PA announcement is made using the Control Room

phone.


COMMENTS:

___ SAT

___ UNSAT

13 4.11

Begin attempts to locate and isolate the source of the flooding.

STANDARD: Dispatch operators to locate and isolate the source of

the flooding.



Examiner Cue: Notify candidate that operators are attempting to isolate the source of the flooding.

COMMENTS:

___ SAT

___ UNSAT


14 4.12

Ensure all Unit 1 ESV pumps are shutdown:

• 1A ESV Pump

• 1C ESV Pump

• 1B ESV Pump

STANDARD: Stop ALL Unit 1 ESV pumps by rotating their pump switches to stop and verify that the red on light extinguishes and the green off light illuminates. (located on 1AB3)


COMMENTS:

___ SAT

___ UNSAT

15 4.13

Dispatch an operator to perform Encl 5.1 (Isolation of SSW to CCW Pumps).

STANDARD: Dispatch an operator to perform Encl 5.1 (Isolation of

SSW to CCW Pumps).



COMMENTS:

___ SAT

___ UNSAT


16 4.14

IAAT HPSW is NOT available,

THEN ensure Encl 5.31 (Temporarily Charging the HPSW System) of Unit 1 EOP has been initiated.

STANDARD: Determine that HPSW is available by observing the

HPSW Jockey pump operating and level in the EWST.


COMMENTS:

___ SAT

___ UNSAT

17 4.15

Coordinate with WCC/TSC to perform the following:

• Recover damaged equipment.

• Remove water from accessible sumps. STANDARD: Candidate should call the WCC and discuss the

following:

o Recover damaged equipment.

o Remove water from accessible sumps.


Booth Cue: When contacted, notify candidate that recovery efforts are in progress.

COMMENTS:

___ SAT

___ UNSAT


18 4.16

WHEN recovery efforts are in progress, THEN EXIT this procedure.

STANDARD: Candidate exits procedure.

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

3 This step is required to attempt to terminate the flooding and not contribute to the water

level in the TB basement. 4 This step is required to attempt to terminate the flooding and not contribute to the water

level in the TB basement.

10 This step is required to ensure CCW inlet piping is vented to break siphon.

CANDIDATE CUE SHEET


INITIAL CONDITIONS

Unit 1 was operating at 100% power

Unit 2 is in cold shutdown

The 2A1 condenser inlet expansion joint has ruptured and Turbine Building Basement water level is rising

Unit 1 has been tripped and the EOP is being completed by other team members

INITIATING CUE

The CRS directs you to initiate Unit 1 AP/10 (Turbine Building Flood)

RO-P403 - b

Page 1 of 14


RO-P403

Initiate HPI Forced Cooling

Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: Yes Time Critical Criteria: HPI Forced Cooling initiated within 5 minutes of criteria being met

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:

RO-P403 - b

Page 2 of 14


Task Title : Initiate HPI Forced Cooling Task Number : Alternate Path: No Time Critical: Yes Validation Time: 10 minutes K/A Rating(s):

System: EPE 074 K/A: EA1.08 Rating: 4.2/4.2 Task Standard: Perform Rule 4 (Initiate HPI Forced Cooling).

References: EOP Rule 3 (Loss of Main or Emergency FDW) EOP Rule 4 (Initiate HPI Forced Cooling) TCA #26, Initiate HPI Forced Cooling when required Tools/Equipment/Procedures Needed: EOP Rule 3 (Loss of Main or Emergency FDW) (Rev. 0L) EOP Rule 4 (Initiate HPI Forced Cooling) (Rev. 0L)

(Note: Below this line is used only for Initial NRC Exams) ===============================================================================



Comments

RO-P403 - b

Page 3 of 14


1. RECALL SNAP 214 2. IMPORT simulator files for RO-P403

3. ENSURE clean in-progress Rule 3 available for candidate

4. ENSURE clean Rule 4 in place on control board 5. Go to RUN when directed by lead examiner

================================================================================






RO-P403 - b

Page 4 of 14

READ TO OPERATOR DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

INITIAL CONDITIONS

• Unit 1 has tripped following a total loss of feedwater

• IMAs are complete

• The crew has been performing Rule 3 (Loss of Main or Emergency FDW) to regain heat transfer

• CBP feed could NOT be established and PSW SG feed is NOT available

• Efforts to restore steam generator heat transfer per Rule # 3 have NOT been successful

• You are at the Step 23 (WHEN) step in Rule 3 (Loss of Main or Emergency FDW)

INITIATING CUE

The CRS directs you to review outstanding IAAT Steps

This JPM is Time Critical

RO-P403 b Page 5 of 14

START TIME: _______

SEQ STEP


1

Rule 3

IAAT Step

3

The candidate determines that it is necessary to perform Rule 4 based on IAAT in Rule 3 or in accordance with OMP 1-18

STANDARD: Candidate announces the initiation of Rule 4.

Examiner Cue: If requested, provide concurrence (as CRS) for initiation of Rule 4.

Examiner Note: This starts the 5 minute “Time critical” time clock.

Time = ___________.

COMMENTS:

___ SAT

__ UNSAT

2

Rule 4

Step 1

Verify any HPI pump powered from 1TC, 1TD, or 1TE can be operated.

STANDARD: The candidate recognizes one HPI pump is in operation.

COMMENTS:

___ SAT

__ UNSAT

3 2

OPEN the following:

• 1HP-24

• 1HP-25

STANDARD: The candidate:

• Rotates 1HP-24 switch on 1UB1 to the OPEN position and observes the red open light come ON and the green closed light go OFF.

• Rotates 1HP-25 switch on 1UB1 to the OPEN position and observes the red open light come ON and the green closed light go OFF.

COMMENTS:

CRITICAL STEP

___ SAT

__ UNSAT


4 3

Start all available HPI pumps STANDARD: Candidate starts the 1B and 1C HPI pumps by rotating the 1B

HPI pump control switch to the START position and rotating the 1C HPI pump control switch to the CLOSE position located on 1UB1.

COMMENTS:

___ SAT __ UNSAT

5 4

OPEN the following:

• *1HP-26 • 1HP-27


• Rotates and holds 1HP-26 switch on 1UB1 to the OPEN position and observes the green “CLOSED light go OFF and the red “OPEN” light come ON.

• Locates 1HP-27 ('1B' HP Injection) on 1UB1 and verifies red 'OPEN' light is ON, and the green 'CLOSED' light is OFF.

COMMENTS:

*CRITICAL

STEP

___ SAT __ UNSAT

6 5

Open 1RC-4

STANDARD: The candidate locates 1RC-4 control switch on 1UB1 and verifies that the red "OPEN" indication is illuminated and the green "CLOSED" indication is extinguished.

Examiner Note: This valve will already be open

COMMENTS:

___ SAT __ UNSAT


7 6

VERIFY flow exists in any HPI header.

STANDARD: The candidate locates HPI Flow Train A and B flow meters on 1UB1 and flow is verified.

COMMENTS:

CRITICAL

STEP

___ SAT __ UNSAT

8 7


A. Place 1RC-66 SETPOINT SELECTOR to OPEN

B. Depress 1RC-66 OPEN PERMIT pushbutton


• *Rotates 1RC-66 SETPOINT SELECTOR switch on 1UB1 to the OPEN position

• *Depresses 1RC-66 OPEN PERMIT pushbutton on 1UB1

• Verifies PORV is open by verifying that the red "OPEN" indication is illuminated and the PORV Flow Statalarm (1SA18/A1) is in alarm.

EXAMINER NOTE: This stops the 5 minute “Time Critical” time clock.

Time = _____________.

COMMENTS:

*CRITICAL

STEP

___ SAT __ UNSAT

9 8

Verify at least two HPI pumps operating.

STANDARD: The candidate verifies that three HPI pumps are operating.

COMMENTS:

___ SAT ___ UNSAT


10 9

Verify flow in both HPI headers is in the acceptable region of Figure 1 (Required HPI Flow Per Header).

STANDARD: The candidate verifies flow in both HPI headers is in the acceptable region of Figure 1 below.

COMMENTS:

___ SAT __UNSAT

11 10

Verify flow exists in any HPI header STANDARD: The candidate locates HPI Flow Train A and B flow meters on

1UB1. Loop A and Loop B flow is verified.

Examiner Note: This flow has already been verified in step 7

COMMENTS:

___ SAT __ UNSAT


12 11


A. Place 1RC-66 SETPOINT SELECTOR to OPEN B. Depress 1RC-66 OPEN PERMIT pushbutton


• Verifies 1RC-66 SETPOINT SELECTOR switch on 1UB1 in the OPEN position

• Depresses 1RC-66 OPEN PERMIT pushbutton on 1UB1

• Verifies PORV is open by verifying that the red "OPEN" indication is illuminated and the PORV Flow Statalarm (1SA18/A1) is in alarm.

Examiner Note: This flow has already been verified in step 8

COMMENTS:

___ SAT ___ UNSAT

13 12

Verify > one RCP operating.

STANDARD: Candidate determines that two RCPs are operating.

COMMENTS:

___ SAT ___ UNSAT

14 13

NOTE:

1A1 RCP provides the best Pzr spray and is preferred to be left running in case recovery from HPI forced cooling is performed and a Pzr bubble drawn.

Stop all but one RCP.

STANDARD: The candidate stops ALL but one RCP by rotating their control switches to “OFF” position.

COMMENTS:

CRITICAL STEP

___ SAT ___ UNSAT


15 14

IAAT the following limits are exceeded

THEN throttle HPI to maximize flow ≤ flow limit.

STANDARD: The candidate verifies header flows less than the limits in the table above.

COMMENTS:

___ SAT ___ UNSAT

16 15

De-energize all PZR heaters.


• Rotates the PZR heater bank #1 switch on 1UB1 to the "OFF" position.

• Presses the OFF pushbutton controls for PZR heater banks 2, 3 and 4 on 1UB1

COMMENTS:

CRITICAL STEP

___ SAT ___ UNSAT


17 16

Close 1HP-5 (LETDOWN ISOLATION)


• Rotates the switch for 1HP-5 on 1UB1 to the closed position.

• Observes the red OPEN light go off and the green CLOSED light come on.

COMMENTS:

CRITICAL STEP

___ SAT ___ UNSAT

18 17

Close the following:

• __ TBVs • __ 1FDW-35 • __ 1FDW-44

STANDARD: The candidate places the TBVs in HAND and reduces demand to zero using the toggle switch OR if the Turbine Master is in manual, verifies the TBVs are closed by observing the green closed light ON and the red open light OFF.

The candidate places 1FDW-35 and 1FDW-44 to HAND and reduce demands to zero using the toggle switches.

COMMENTS:

___ SAT ___ UNSAT


19 18

IAAT all HPI is lost,

THEN:

A. Stop all RCPs B. Position 1RC-66 SETPOINT SELECTOR to HIGH

STANDARD: The candidate verifies HPI is available and operating and the IAAT step does not apply at this time.

COMMENTS:

___ SAT ___ UNSAT

20 19

WHEN directed by CRS,

THEN EXIT.

STANDARD: The candidate announces that Rule 4 is complete with outstanding IAATs and returns the Cue sheet to the examiner indicating he has completed his JPM.

COMMENTS:

END TASK

___ SAT ___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

3 Step is required provide proper system alignment for HPI forced cooling flow 5 Step required to allow flow in both HPI headers 7 Step is required to verify HPI flow available for forced cooling. 8 Step is required to open the PORV to initiate HPI forced cooling through the core (TCA

#26).

14 Step required to limit the heat input to the RCS

16 Step required to limit the heat input to the RCS

17 Step required to Reduce the amount of heat energy added to the RB Containment

RO-P403 - b

Page 14 of 14

CANDIDATE CUE SHEET


INITIAL CONDITIONS

• Unit 1 has tripped following a total loss of feedwater

• IMAs are complete

• The crew has been performing Rule 3 (Loss of Main or Emergency FDW) to regain heat transfer

• CBP feed could NOT be established and PSW SG feed is NOT available

• Efforts to restore steam generator heat transfer per Rule # 3 have NOT been successful

• You are at the Step 23 (WHEN) step in Rule 3 (Loss of Main or Emergency FDW)

INITIATING CUE

The CRS directs you to review outstanding IAAT Steps

This JPM is Time Critical

RO-S404a - e Page 1 of 15


RO-S404a

Establish EFDW Flow Through Startup Valves

Alternate Path: Yes Alt Path Description: ____1FDW-315 failed closed_________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Establish EFDW flow through Startup valves

Task Number : Alternate Path: Yes Time Critical: No Validation Time: 15 minutes K/A Rating(s): System: APE-054 K/A: AA2.04 Rating: 4.2/4.3 Task Standard: EFDW flow is established to the affected header through the startup valve.

References:

EOP Rule 3, (Loss of Main or Emergency FDW) EOP Rule 7, (SG Feed Control) EOP Enclosure 5.27, (Alternate Methods for Controlling EFDW Flow)

Tools/Equipment/Procedures Needed: EOP Rule 3, (Loss of Main or Emergency FDW) (0L, Rev. 01) EOP Rule 7, (SG Feed Control) (0L, Rev. 01) EOP Enclosure 5.27, (Alternate Methods for Controlling EFDW Flow) (0O, Rev. 01) ===============================================================================



Comments



1. RECALL SNAP 213

2. IMPORT files for RO-S404a 3. ENSURE clean procedure in place for candidate

4. WHEN directed by Lead Examiner, go to RUN

================================================================================







READ TO OPERATOR


INITIAL CONDITIONS

• The reactor has just tripped from 25% power

• Immediate Manual Actions are complete

INITIATING CUES

The CRS directs you to perform a Symptoms Check


START TIME: _______

SEQ STEP


1

Performs a Symptom Check

STANDARD: Performs Symptom Check and determines that there are no

symptoms to report but will perform Rule 3 due to a loss of Main Feedwater

EXAMINER CUE: CRS acknowledges performing Rule 3 due to a Loss of Main Feedwater.

COMMENTS:

___ SAT

___ UNSAT

2

Rule 3

Step 1

Verify loss of MFDW and/or EFDW was due to any of the following:

__ Turbine Building Flooding. __ actions taken to increase SG level due to Turbine Building

Flooding.

STANDARD: Observes that TB Flooding is NOT occurring by 2SA-18/A-11 (Turbine BSMT Water Emergency High Level) NOT illuminated. (OP/2/A/6102/018)

Candidate will perform RNO step and GO TO step 3.

COMMENTS:

___ SAT

___ UNSAT


3 3

IAAT NO SGs can be fed with FDW (Main/CBP/Emergency/PSW),

AND any of the following exist:

• RCS pressure reaches 2300 psig OR NDT limit • Pzr level reaches 375″ [340″ acc]

THEN PERFORM Rule 4 (Initiation of HPI Forced Cooling).

STANDARD: Candidate determines Rule 4 is not required.

Continue to Step 4.

COMMENTS:

___ SAT

___ UNSAT

4 4

Start operable EFDW pumps, as required, to feed all intact SGs.

STANDARD: Observes MD EFDWP & TD EFDWP running with switch red lights on and normal discharge pressure.

Continue to Step 5.

COMMENTS:

___ SAT ___ UNSAT

5 5

Verify any EFDW Pump is operating.

STANDARD: Checks MD EFDWP & TD EFDWP red lights are on and Pumps have normal discharge pressure.

Continue to Step 6.

COMMENTS:

___ SAT

___ UNSAT


6 6

GO TO step 38.

STANDARD: GO TO step 38.

COMMENTS:

___ SAT

___ UNSAT

7 38

ALTERNATE PATH

IAAT an EFDW valve CANNOT control in AUTO, OR manual operation of EFDW valve is desired to control flow/level, THEN perform Steps 39 - 43.

STANDARD: Determines that 1FDW-315 is NOT controlling properly (1A SG level is < 30” and decreasing). Concludes that 1FDW-315 has failed closed.

Continue to Step 39.

COMMENTS:

___ SAT

___ UNSAT

8 39

Place EFDW valve in MANUAL.

STANDARD: Place 1FDW-315 in MANUAL by depressing the A/M pushbutton on the Moore controller.


COMMENTS:

___ SAT

___ UNSAT


9 40

Control EFDW flow with EFDW valve in MANUAL.

STANDARD: Determine that 1FDW-315 will NOT control in MANUAL.

Perform RNO, GO TO Step 42.

Continue to step 42

COMMENTS:

___ SAT

___ UNSAT

10 42

Notify CRS that Encl 5.27 (Alternate Methods for Controlling EFDW Flow) is being initiated.

STANDARD: Removes Encl. 5.27 from EOP and initiates.

EXAMINER CUE: CRS acknowledges entry into Enclosure 5.27.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

11

Encl. 5.27 Step

1

Identify the failure: 1FDW-315 has failed CLOSED [GO TO Step 2]

STANDARD: Candidate determines the next procedural step from table in Step 1.Continue to Step 2.

COMMENTS:

___ SAT

___ UNSAT


12 2

Verify 1A MD EFDWP is operating

STANDARD: Candidate verifies 1A MD EFDWP is operating. Verify red light on and white light off and normal discharge pressure.

Continue to Step 3.

COMMENTS:

___ SAT

___ UNSAT

13 3

Stop 1A MD EFDWP

STANDARD: *Candidate places 1A MD EFDWP switch to OFF.

Verify red light off and white light illuminated.

Continue to Step 4.

COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

14 4

Verify 1B MD EFDWP is operating.

STANDARD: Candidate verifies 1B MD EFDWP is operating.

Verify red light on and white light off and normal discharge pressure.

Continue to Step 5.

COMMENTS:

___ SAT

___ UNSAT


15 5

Place 1 TD EFDW Pump in PULL TO LOCK

STANDARD: *Candidate places the U1 TD EFDW Pump in PULL to LOCK. Candidate verifies red light is OFF and green light is ON.

Continue to Step 6.

COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

16 6

Place 1FDW-35 in HAND and set demand to 0%

STANDARD: Candidate places 1FDW-35 in HAND and uses toggle switch to reduce demand to 0%.

Continue to Step 7.

COMMENTS:

CRITICAL

STEP

___ SAT

___ UNSAT

17 7

Close 1FDW-33

STANDARD: *Candidate closes 1FDW-33.

Candidate verifies red light is OFF and green light is ON.

Continue to Step 8.

COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT


18 8

Verify 1A MD EFDWP will be used.

STANDARD: Candidate determines that the 1A MD EFDWP will be used.

Continue to Step 9.

EXAMINER CUE: If candidate asks if 1A MD EFDWP will be used, state that the CRS directs using the 1A MD EFDWP.

COMMENTS:

___ SAT

___ UNSAT

19 9

Open 1FDW-374

STANDARD: *Candidate locates and opens 1FDW-374.

Candidate observes green closed light off and red open light on.


COMMENTS:

*CRITICAL

STEP

___ SAT

___ UNSAT

20 10

Verify the following:

• 1FDW-36 closed • 1FDW-38 open

STANDARD: Candidate determines 1FDW-36 is closed by observing the green closed light ON and red open light OFF on 1VB3 or uses OAC indication.

Candidate determines 1FDW-38 is open by observing the green closed light OFF and red open light lit ON 1VB3 or uses OAC indication.


COMMENTS:

___ SAT

___ UNSAT


21 11

Start 1A MD EFDWP

STANDARD: *Candidate places 1A MD EFDWP switch to RUN.

Candidate verifies pump start by observing white light is OFF and red light is ON and proper discharge pressure.


COMMENTS:

*CRITICAL

STEP

___ SAT

___ UNSAT

22 12

NOTE:

Flow from the TD EFDWP through a S/U control valve should be read on the FDW SU FLOW gauge.

Flow from a MD EFDWP through a S/U control valve should be read on the MDEFWP DISCH FLOW gauge.

100 gpm could cause overcooling if adequate decay heat levels do NOT exist.

Verify either of the following exists:

• HPI Forced Cooling is maintaining core cooling • CBP Feed providing SG feed

STANDARD: Candidate determines that neither condition is met and goes to the RNO.

Continue to Step 12 RNO.

COMMENTS:

___ SAT

___ UNSAT


23 RNO

12

IF any SG is being fed, THEN perform the following:

• *Throttle 1FDW-35 to establish a maximum of 100 gpm.

• *Throttle 1FDW-35 to obtain desired SG level per Rule 7 (SG Feed Control)

• Notify CR SRO of SG Feed Status

STANDARD: Candidate throttles 1FDW-35 to attain ~ 100 gpm flow initially on MD EFWP DISCH FLOW gauge or OAC EFW graphic, then throttles 1FDW-35 as necessary to establish ~ 25” S/U level (per Rule 7 table 4)

Rule 7 Table 4 (All SCMs > 0°F AND any RCP on) specifies the level to establish when using EFDWP is 30” (use MFDW setpoint if feeding via S/U CVs). The MFDW setpoint (i.e. when using the S/U CVs) is 25” S/U level since RCS temperature is > 500°F.

Examiner Note: The candidate does not need to establish this level to complete the JPM. 1FDW-35 must be open and SG level increasing.

Examiner Cue: Another operator will continue with this procedure.

COMMENTS:

END TASK

*CRITICAL

STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

10 This step is required to ensure the correct procedure is used.

13 This step is required to ensure the valves will operate.

15 This step is required to ensure the valves will operate.

16 This step is required to align the MD EFDWP to the S/U header.



21 This step is required to start the 1A MD EFDWP and provide a SG feed source.

23 This step is required to establish flow to the 1A SG.

CANDIDATE CUE SHEET


INITIAL CONDITIONS

• The reactor has just tripped from 25% power

• Immediate Manual Actions are complete

INITIATING CUES

The CRS directs you to perform a Symptoms Check

(AO-701) j Page 1 of 9


AO-701

RESTORATION OF ICS AUTO POWER

Administrative: No Alternate Path: Yes Alt Path Description: _________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Restoration of ICS Auto Power Task Number : N/A Alternate Path: Yes Time Critical: No Validation Time: 15 minutes K/A Rating(s):

System: BW/A02 K/A: AA1.1 Rating: 4.0/3.8 Task Standard: ICS AUTO power is restored per AP/3/A/1700/023 (Loss of ICS Power) References: AP/3/A/1700/023 (Loss of ICS Power) Enclosure 5.2 (Restoration of ICS AUTO Power) (Rev. 23) Tools/Equipment/Procedures Needed: AP/3/A/1700/023 (Loss of ICS Power) Enclosure 5.2 (Restoration of ICS AUTO Power) (Rev. 23)




Comments



None


READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you. INITIAL CONDITIONS Unit 3 operating at 100% power 3SA-2/B-11 (ICS AUTO POWER FAILURE) is actuated INITIATING CUE The Control Room SRO directs you to use AP/3/A/1700/023 (Loss of ICS Power) Enclosure 5.2 (Restoration of ICS AUTO Power) to restore ICS AUTO power on Unit 3.


START TIME: _______

SEQ STEP


1 1

Verify the following breakers closed (Unit 3 Cable Rm): __ 3KRA breaker #1 (100A 1P, ICS POWER PANELBOARD 3KI) __ 3KI breaker #1 (30A 1P, AUTO POWER (ICS/ACS))

STANDARD: Locate 3KRA panel board breaker #1 and verify it is closed.

Examiner Cue: Indicate to the candidate that 3KRA breaker #1 is CLOSED

Locate 3KI panel board breaker #1 and verify it is closed

Examiner Cue: Indicate to the candidate that 3KI breaker #1 is tripped OPEN.

Continue Step 1 RNO COMMENTS:

___ SAT

___ UNSAT

2 1

RNO

1. Reset and close the affected breakers (Unit 3 Cable Rm): __ 3KRA breaker #1 (100A 1P, ICS POWER PANELBOARD 3KI) __ *3KI breaker #1 (30A 1P, AUTO POWER (ICS/ACS))

2. IF either of the above breakers fails to remain closed, THEN notify CR SRO to have SPOC assist in resetting and closing any open breakers

3. WHEN both of the above breakers are closed, THEN continue

STANDARD: Resets and closes 3KI BREAKER #1 by taking the breaker to the open position and then closing it.

Examiner Cue: Indicate to the candidate that 3KI breaker #1 is CLOSED.

COMMENTS:

___ SAT

___ UNSAT


3 2

ALTERNATE PATH

Verify ICS AUTO power has been restored as indicated by 3SA-2/B-11 (ICS AUTO POWER FAILURE) off

STANDARD: Contacts Unit 3 Control Room to verify ICS AUTO power has been restored as indicated by 3SA-2/B-11 (ICS AUTO POWER FAILURE) off

Examiner Cue: Inform candidate that 3SA-2/B-11 (ICS AUTO POWER FAILURE) remains actuated.

COMMENTS:

___ SAT

___ UNSAT

4 2

RNO

IF ICS AUTO power has NOT been restored, THEN bypass 3KI inverter as follows (KI BYP SW cabinet, Unit 3 Equip Rm, North of col Q88) A. Position SW#1 OFF (left switch) B. Position SW#3 OFF (right switch) C. Position SW#2 ON (center switch)

STANDARD: Positions SW#1 to OFF (left switch).

Examiner Cue: Indicate to the candidate that SW#1 is OFF. Positions SW#3 to OFF (right switch)

Examiner Cue: Indicate to the candidate that SW#3 is OFF.

Positions SW#2 to ON (center switch Examiner Cue: Indicate to the candidate that SW#2 is ON. COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

NOTE The following step will require contacting Unit 3 Control Room


5 3

Notify Unit 3 CR that all applicable steps of this enclosure have been completed

STANDARD: Notify Unit 3 CR that all applicable steps of this enclosure have been completed.

Examiner Cue: Inform candidate that the Control Room has been notified.

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

4 Step is required in proper sequence to align power to the 3KI bus.

CANDIDATE CUE SHEET


INITIAL CONDITIONS Unit 3 operating at 100% power 3SA-2/B-11 (ICS AUTO POWER FAILURE) is actuated INITIATING CUE The Control Room SRO directs you to use AP/3/A/1700/023 (Loss of ICS Power) Enclosure 5.2 (Restoration of ICS AUTO Power) to restore ICS AUTO power on Unit 3.

AO-804 i Page 1 of 9


AO-804

AO Actions for Control Room Evacuation

Administrative: No Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : AO Actions for Control Room Evacuation Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 12 minutes K/A Rating(s):

System: APE068 K/A: AA1.07 Rating: (4.1/4.2) Task Standard: AO actions for Control Room Evacuation are completed IAW AP/1/A/1700/050 Encl. 5.6.

References: AP/1/A/1700/050

Tools/Equipment/Procedures Needed: AP/1/A/1700/050, Encl. 5.6 (Rev. 03)




Comments



None


READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you. INITIAL CONDITIONS A fire has occurred in Unit 1 and 2 Control Room AP/1/A/1700/050 (Challenging Plant Fire) has been initiated Unit 1 and 2 Control Rooms are being evacuated AP/1/A/1700/050, Encl. 5.6 (AP/EOP AO Actions for Control Room Evacuation) is in progress up to step 2. INITIATING CUE The CRS directs you to continue in AP/1/A/1700/050, Encl. 5.6 (AP/EOP AO Actions for Control Room Evacuation), beginning at step 2.


START TIME: _______

SEQ STEP


1 2

Locally close the following:

__ 1FDW-315 (1A S/G EFDW CONTROL) (A-4, E Pen Rm, 10' N of Qa-68 against RB Wall)

__ 1FDW-316 (1B S/G EFDW CONTROL) (A-4, W Pen Rm, 4' NE of W-70a, 4' up)

STANDARD: Candidate locates 1FDW-315 and 1FDW-316 and

simulates rotating the handwheel clockwise to a hard stop.

Examiner Cue: Inform candidate that each valve is closed

following handwheel rotation in the clockwise direction.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

NOTE Dress out is NOT required. RP will be notified at completion of this Enclosure.


2 3

Open the following breakers (A-4/ East Pen Rm):

__ 1XH-3A (Pzr Htr Group E Bkr) (already open) __ 1XH-4A (Pzr Htr Group H Bkr)

__ 1XI-3A (Pzr Htr Group F Bkr)

__ 1XI-4A (Pzr Htr Group I Bkr)

__ 1XJ-2A (Pzr Htr Group D Bkr)

__ 1XJ-3A (Pzr Htr Group G Bkr) (already open) __ 1XJ-4A (Pzr Htr Group J Bkr)

__ 1XK-1A (Pzr Htr Group A-K Bkr) STANDARD: Candidate locates and simulates opening the

following breakers by rotating the breaker handle counter clockwise to the OFF position: __ 1XH-3A (Pzr Htr Group E Bkr)

__ 1XH-4A (Pzr Htr Group H Bkr)

__ 1XI-3A (Pzr Htr Group F Bkr)

__ 1XI-4A (Pzr Htr Group I Bkr)

__ 1XJ-2A (Pzr Htr Group D Bkr)

__ 1XJ-3A (Pzr Htr Group G Bkr)

__ 1XJ-4A (Pzr Htr Group J Bkr)

__ 1XK-1A (Pzr Htr Group A-K Bkr) COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


3 4

Notify the CRS in the SSF Encl 5.6, (AP/EOP AO Actions for Control Room Evacuation) actions are completed.

STANDARD: Candidate notifies the CRS in the SSF that Encl. 5.6

actions are completed. Examiner Cue: Inform the candidate that the CRS has been

notified and another operator will continue with the procedure.

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 This step is required to secure EFDW to prevent challenging operation of the SSF. 2 This step is required to secure Pressurizer Heaters to prevent challenging operation of the SSF.

CANDIDATE CUE SHEET


INITIAL CONDITIONS A fire has occurred in Unit 1 and 2 Control Room AP/1/A/1700/050 (Challenging Plant Fire) has been initiated Unit 1 and 2 Control Rooms are being evacuated AP/1/A/1700/050, Encl. 5.6 (AP/EOP AO Actions for Control Room Evacuation) is in progress up to step 2. INITIATING CUE The CRS directs you to continue in AP/1/A/1700/050, Encl. 5.6 (AP/EOP AO Actions for Control Room Evacuation), beginning at step 2.

AO-S403 k Page 1 of 11


AO-S403

AO Actions for Supply of Water to SSF

Administrative: No Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : AO Actions for Supply of Water to SSF Task Number : Alternate Path: No Time Critical: No Validation Time: 25 minutes K/A Rating(s):

System: Gen K/A: 2.1.20 Rating: (4.6/4.6) Task Standard: Flow path from CCW Discharge back to CCW inlet is aligned in accordance with AP/0/A/1700/046, Encl. 5.9 (Supply of Water to SSF). References: AP/0/A/1700/046 (Extensive Damage Mitigation) (Rev. 10)

Tools/Equipment/Procedures Needed: AP/0/A/1700/046, Encl. 5.9 (Supply of Water to SSF) (Rev. 10)




Comments



None


READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you. INITIAL CONDITIONS Unit 2 refueling outage in progress with Unit 2 CCW Condenser inlet piping isolated. An explosion occurred that disabled the Unit 3 Control Room and operating crew. AP/0/A/1700/046 (Extensive Damage Mitigation) was initiated. Unit 3 is being maintained in hot shutdown from the SSF in accordance with AP/46. The dedicated Submersible Pump is inoperable and cannot be installed. Keowee Lake level is 798 feet. INITIATING CUE The Unit 2 CRS directs you to perform AP/0/A/1700/046, Encl. 5.9 (Supply of Water to SSF), beginning at step 7, to align the CCW Discharge back to the CCW inlet.


START TIME: _______

SEQ STEP


1 7

Locally open the following CCW crossover valves to align all units to Unit 2 and the SSF: __ *1CCW-40 (UNIT 1 CROSSOVER TIE) (T-1/K-22) __ *2CCW-41 (UNIT 2 CROSSOVER TIE) (T-1/J-40) __ *3CCW-42 (UNIT 3 CROSSOVER TIE) (T-1/K-44/45) __ 3CCW-94 (UNIT 3 CROSSOVER TIE) (T-1/K-50)

STANDARD: Candidate opens the following valves by rotating the handwheel in the counter clockwise direction to a hard stop; __ *1CCW-40 (UNIT 1 CROSSOVER TIE) (T-1/K-22) __ *2CCW-41 (UNIT 2 CROSSOVER TIE) (T-1/J-40) __ *3CCW-42 (UNIT 3 CROSSOVER TIE) (T-1/K-

44/45) __ 3CCW-94 (UNIT 3 CROSSOVER TIE) (T-1/K-50)

COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

2 8

Verify both of the following: __ Keowee Lake level > 791' __ Flow path from CCW Discharge back to CCW Inlet is desired

STANDARD: Candidate determines Keowee lake level is > 791 feet and the flow path from CCW Discharge back to CCW Inlet is desired from information contained on the cue sheet.

Examiner Cue: If asked, inform the candidate that Keowee lake

level is > 791 feet and the flow path from CCW Discharge back to CCW Inlet is desired.

COMMENTS:

___ SAT

___ UNSAT


3 9

Locally open the following Unit 1 valves: __ 1CCW-75 (CONDENSATE COOLERS CCW SUPPLY)

(T-1/F-25) __ 1CCW-76 (1A CONDENSATE COOLER CCW INLET)

(T-1/F-25) __ 1CCW-78 (1B CONDENSATE COOLER CCW INLET)

(T-1/F-25) __ 1CCW-77 (1A CONDENSATE COOLER CCW OUTLET)

(T-1/E-25) __ 1CCW-79 (1B CONDENSATE COOLER CCW OUTLET)

(T-1/E-25) __ 1CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL

BYPASS) (T-1/E-26)

STANDARD: The following valves are opened by rotating the handwheel in the counter clockwise to a hard stop, or verified open by the position indicator: __ 1CCW-75 (CONDENSATE COOLERS CCW

SUPPLY) (T-1/F-25)

__ 1CCW-76 (1A CONDENSATE COOLER CCW INLET) (T-1/F-25)

__ 1CCW-78 (1B CONDENSATE COOLER CCW INLET) (T-1/F-25)

__ 1CCW-77 (1A CONDENSATE COOLER CCW OUTLET) (T-1/E-25)

__ 1CCW-79 (1B CONDENSATE COOLER CCW OUTLET) (T-1/E-25)

__ 1CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL BYPASS) (T-1/E-26)

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


4 10





(T-1/D-30) __ 2CCW-79 (2B CONDENSATE COOLER CCW OUTLET)

(T-1/D-30, 4' W.) __ 2CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL

BYPASS) (T-1/ D-30, 4' W.)


SUPPLY) (T-1/F-30)



__ 2CCW-77 (2A CONDENSATE COOLER CCW OUTLET) (T-1/D-30)

__ 2CCW-79 (2B CONDENSATE COOLER CCW OUTLET) (T-1/D-30, 4' W.)

__ 2CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL BYPASS) (T-1/ D-30, 4' W.)

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


5 11





(T-1/D-44) __ 3CCW-79 (2B CONDENSATE COOLER CCW OUTLET)

(T-1/D-44) __ 3CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL

BYPASS) (T-1/ D-43/44)


SUPPLY) (T-1/F-44)



__ 3CCW-77 (2A CONDENSATE COOLER CCW OUTLET) (T-1/D-44)

__ 3CCW-79 (2B CONDENSATE COOLER CCW OUTLET) (T-1/D-44)

__ 3CCW-86 (CONDENSATE COOLERS CCW FLOW CONTROL BYPASS) (T-1/ D-43/44)

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


6 12

Notify TSC to continue efforts to supply water to the Unit 2 CCW Inlet piping for extended SSF operation

STANDARD: Candidate notifies the TSC to continue efforts to supply water to the Unit 2 CCW Inlet piping for extended SSF operation

Examiner Cue: Inform the candidate the TSC has been notified. COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 This step is required to align the flow path from the CCW Discharge back to the CCW Inlet. 3 This step is required to align the flow path from the CCW Discharge back to the CCW Inlet. 4 This step is required to align the flow path from the CCW Discharge back to the CCW Inlet. 5 This step is required to align the flow path from the CCW Discharge back to the CCW Inlet.

CANDIDATE CUE SHEET


INITIAL CONDITIONS Unit 2 refueling outage in progress with Unit 2 CCW Condenser inlet piping isolated. An explosion occurred that disabled the Unit 3 Control Room and operating crew. AP/0/A/1700/046 (Extensive Damage Mitigation) was initiated. Unit 3 is being maintained in hot shutdown from the SSF in accordance with AP/46. The dedicated Submersible Pump is inoperable and cannot be installed. Keowee Lake level is 798 feet. INITIATING CUE The Unit 2 CRS directs you to perform AP/0/A/1700/046, Encl. 5.9 (Supply of Water to SSF), beginning at step 7, to align the CCW Discharge back to the CCW inlet.

ADM-110 1.1B Page 1 of 12


ADM-110

Diverse Verification of Reactor Power

Administrative: Yes Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Diverse Verification of Reactor Power Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 15 minutes K/A Rating(s):

System: Gen K/A: 2.1.45 Rating: 4.3/4.3 Task Standard: Compare the average NI Power Range Rx power to the average Loop delta T power and determine that they are NOT within 2% of each other.

References: PT/1/A/0600/001 (Periodic Instrument Surveillance) (Rev. 337) PT/0/A/0600/001 A (Loss of Computer) Picture of front board indications Tools/Equipment/Procedures Needed: PT/1/A/0600/001 (Periodic Instrument Surveillance) (Rev. 337) Picture of front board indications




Comments



None


READ TO OPERATOR


INITIAL CONDITIONS

Unit 1 is at 80% RTP of Cycle 29 PT/1/A/0600/001 (Periodic Instrument Surveillance) Enclosure 13.1 (Mode 1&2) has just been initiated for your shift OAC is unavailable Transient Monitor is unavailable INITIATING CUE

The Time = 2000 requirements on page 1 of 29 of PT/1/A/0600/001 (Periodic Instrument Surveillance), Enclosure 13.1 (Mode 1&2) are being performed. The CRS directs you to perform Step 1 and determine if the procedural requirements are met that will allow signing Step 1 as complete. Document your answer below.


START TIME: _______

SEQ STEP


1 1 If Rx critical, perform the following: (R.M.)

2 1.1

Record Core Thermal Power Indication every 2 hours. (R.M.)

STANDARD: Candidate reviews the note prior to step 1.1 concerning priority of power Indications to use for surveillance and determines that A through E are not available due to the OAC and Transient Monitor being out of service.

Candidate refers to the attached graphic and averages the four NI power ranges to determine (%) Reactor Power in accordance with Note 1.1F as follows:

NI-5 = 80.1

NI-6 = 79.9

NI-7 = 79.8

NI-8 = 80.0

80.1 + 79.9 + 79.8 + 80.0 = 319.8 / 4 = 79.95

Candidate records 79.95 in the % Reactor Power (R.M.) column at 2000 and initials the Step 1.1 (R.M.) column at 2000.

COMMENTS:

___ SAT

___ UNSAT

3 1.2

Verify NO CTP OAC alarms OR CTP OAC alarm response in progress. (R.M.)

STANDARD: Candidate determines there are no CTP OAC alarms. The OAC is unavailable. Candidate initials the Step 1.2 (R.M.) column at 2000.

COMMENTS:

___ SAT

___ UNSAT


4 1.3 Verify diverse reactor power indications within + 2% per one of the following:

5 1.3.1

Compare the following: (refer to Note 1.1)

• OAC Calculated Thermal Power Best

OR

• OAC Calculated Thermal Power Secondary (if above ~ 25% power)

OR

• OAC Calculated Thermal Power Primary (if below ~ 25% power)

versus

• OAC Calculated Thermal Power Delta T

STANDARD: Candidate determines these indications are not available due to the OAC being out of service and proceeds to step 1.3.2.

COMMENTS:

___ SAT

___ UNSAT


6 1.3.2

Compare the following:

• Average of NI-5, NI-6, NI-7, and NI-8 (refer to Note 1.1.F)

OR

• PID A5081 ICS Calculated Thermal Power Best available from the Transient Monitor. See Note 1.1.E above.

versus

• Average two RC Loop ΔTs from RC Loop ΔT gauge and use Enclosure "Loop ΔT Vs Reactor Power" to determine percent power from ΔT

STANDARD: Candidate refers to the attached graphic and averages the four NI power ranges, or uses the value for reactor power recorded per step 1.1, since PID A5081 ICS Calculated Thermal Power Best is not available from the Transient Monitor.

NI-5 = 80.1

NI-6 = 79.9

NI-7 = 79.8

NI-8 = 80.0

80.1 + 79.9 + 79.8 + 80.0 = 319.8 / 4 = 79.95

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


7 1.3.2 con't.



OR


versus


STANDARD: Candidate refers to the attached graphic and averages the Loop A and Loop B ΔT indications:

Loop A = 37.2 Loop B = 37.4

37.2 + 37.4 = 74.6 / 2 = 37.3 °F COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

8 1.3.2 con't.



OR


versus


STANDARD: Candidate refers to PT/1/A/0600/001, Enclosure13.11 (Loop ΔT Vs Reactor Power), and using 37.3 °F determines that Reactor Power from delta T is 85% ( + 2%).

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


9 1.3.2 con't.



OR


versus


STANDARD: Candidate compares the average of NI-5, NI-6, NI-7, and NI-8 (79.95%) to the percent power from ΔT acquired from Enclosure 13.11 (Loop ΔT Vs Reactor Power) (85% + 2%), and determines the values are NOT within + 2%. Therefore step 1.3 cannot be signed off as complete.

EXAMINER NOTE: The candidate may stop prior to performing step 10 when they determine procedure step 1.3 cannot be signed off as complete. If so, the task will end here at step 9 and step 10 will not be performed.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

ADM-110 1.1B Page 10 of 12

10 1.4

Verify CTPD set < 100%. (R.M.)

STANDARD: Candidate refers to the attached graphic of the Load Control Panel (LCP) and determines CTPD set = 80%.

Candidate initials the Step 1.4 (R.M.) column at 2000.

EXAMINER NOTE: The candidate may stop prior to performing step 10 when they determine procedure step 1.3 cannot be signed off as complete. If so, the task will end at step 9 and step 10 will not be performed.

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________

ADM-110 1.1B Page 11 of 12


SEQ STEP #

Explanation

6 This step is required to determine % Rx Power from the average of NI-5, 6, 7, and 8. 7 This step is required to determine the correct ΔT to use. 8 This step is required to determine % Rx Power from ΔT. 9 This step is required to determine if the acceptance criteria for the step has been met.

CANDIDATE CUE SHEET


INITIAL CONDITIONS

Unit 1 is at 80% RTP of Cycle 29 PT/1/A/0600/001 (Periodic Instrument Surveillance) Enclosure 13.1 (Mode 1&2) has just been initiated for your shift OAC is unavailable Transient Monitor is unavailable INITIATING CUE

The Time = 2000 requirements on page 1 of 29 of PT/1/A/0600/001 (Periodic Instrument Surveillance), Enclosure 13.1 (Mode 1&2) are being performed. The CRS directs you to perform Step 1 and determine if the procedural requirements are met that will allow signing Step 1 as complete. Document your answer below.

(ADM-111) 1.2R Page 1 of 11


ADM-111

Perform Manual RCS Leakage Calculation


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:

(ADM-111) 1.2R Page 2 of 11

REGION II


Task Title : Perform Manual RCS Leakage Calculation Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 20 minutes K/A Rating(s):

System: Generic K/A: 2.1.20 Rating: 4.6/4.6 Task Standard: RCS Leakage is calculated per the attached key. References: PT/0/A/0600/001A (Loss Of Computer) (Rev 42) PT/1/A/0600/010 (Reactor Coolant Leakage) (Rev 95) Tools/Equipment/Procedures Needed: PT/1/A/0600/010 (Reactor Coolant Leakage) Encl. 13.3 (Manual RCS Leakage Calculation Data Sheet) (Rev 95)

(Note: Below this line is used only for Initial NRC Exams) =====================================================================================



Comments

(ADM-111) 1.2R Page 3 of 11


None

(ADM-111) 1.2R Page 4 of 11

READ TO OPERATOR

DIRECTIONS TO STUDENT I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you. INITIAL CONDITIONS Oconee Unit 1 OAC computer repairs are expected to be extended through turnover for hardware replacement. The Loss of Computer procedure PT/0/A/0600/001A (Loss of Computer) is in progress, however an RCS Leakage Calculation has not been performed during this shift. INITIATING CUES The CRS directs you to perform a manual RCS leakage calculation per PT/1/A/0600/010 (Reactor Coolant Leakage) Encl. 13.3 (Manual RCS Leakage Calculation Data Sheet) using the data collected below. Round calculations to three (3) decimal places.

Manual RCS Leakage Data

Parameter Initial Final Time 0015 0115 Pzr level 219.0 inches 220.7 inches Tave Indication 579.0°F 579.3°F Quench Tank Level 86.5 inches 86.5 inches LDST Level 76.7 inches 76.1 inches Power Range NI 100.0% 100.1% RCS NR Pressure 2150 psig 2150 psig Group 7 Control Rod Position 92.7% 92.7%

(ADM-111) 1.2R Page 5 of 11

START TIME: _______

SEQ STEP


1 2.1

Record Initial Data in "Table #1"

Value Initial Data Minus Final Data Equals Change

Pzr Lvl 219 Inches - inches = Inches (1)

RCS T 579 °F - °F = °F (2)

QT Lvl 86.5 inches - inches = Inches (3)

LDST Lvl 76.7 inches - inches = Inches (4)

Rx Pwr 100.0 % - % = %

RCS P 2150 psig - psig = psig

CR Pos 92.7 % - % = %

STANDARD: Student enters initial set of data into “Table #1” of Enclosure 13.3 (Manual Leakage Calculation Data Sheet) from the initiating cue.

COMMENTS:

___ SAT

___ UNSAT

2 2.2

Record time Initial Data recorded: 0015

STANDARD: Student enters 0015 (from initiating cue) COMMENTS:

___ SAT

___ UNSAT

NOTE: • If Tave is on-scale, use Tave. If Tave is off-scale, use Tcold

• If RCS NR Pressure is on-scale, use NR Pressure. If RCS NR Pressure is off-scale, use RCS WR Pressure

(ADM-111) 1.2R Page 6 of 11

3 2.3

After 1 hour, record Final Data in "Table #1"


Pzr Lvl 219 Inches - 220.7 inches = Inches (1)

RCS T 579 °F - 579.3 °F = °F (2)

QT Lvl 86.5 inches - 86.5 inches = Inches (3)

LDST Lvl 76.7 inches - 76.1 inches = Inches (4)

Rx Pwr 100.0 % - 100.1 % = %

RCS P 2150 psig - 2150 psig = psig

CR Pos 92.7 % - 92.7 % = %

STANDARD: Student enters final set of data into “Table #1” of Enclosure 13.3 (Manual Leakage Calculation Data Sheet) (from initiating cue)

COMMENTS:

___ SAT

___ UNSAT

4 2.4

Record time Final Data recorded: 0115

STANDARD: Student enters 0115 COMMENTS:

___ SAT

___ UNSAT

5 2.5

Record elapsed time: 60 minutes

STANDARD: Student enters 60 minutes COMMENTS:

___ SAT

___ UNSAT

(ADM-111) 1.2R Page 7 of 11

6 2.6

Calculate and record "Change" values in "Table # 1"


Pzr Lvl 219 Inches - 220.7 inches = -1.7 nches (1)

RCS T 579 °F - 579.3 °F = -0.3 °F (2)

QT Lvl 86.5inches - 86.5 inches = 0 Inches (3)

LDST Lvl 76.7inches - 76.1 inches = +0.6 inches (4)

Rx Pwr 100.0 % - 100.1 % = -0.1 %

RCS P 2150 psig - 2150 psig = 0 psig

CR Pos 92.7 % - 92.7 % = 0 %

STANDARD: Student calculates the change in data and fills in Table 1 change column

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

7 2.7

Calculate Corrected Pzr Level Change:

-1.7 inches – (6.831 inches/°F X -0.3 °F) = +.349 inches (1) PZR Level Change Level Change due to Temp (2) Corrected PZR Level Change

STANDARD: -1.7 inches – (6.831 inches/°F X -.3°F) = +.349 inches (+.339 to +.359) COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

NOTE: • Change = Initial - Final

• Negative sign (-) should be included with values as appropriate

NOTE: Conversion Factor (6.831 inches/°F) must be positive (+) value

(ADM-111) 1.2R Page 8 of 11

8 2.8

Convert Corrected Pzr Level Change to gallons:

+.349 inches X 14.364 gallons / inch = +5.013 gallons Corr PZR Lvl Change Conversion corr to 68°F PZR Vol Change

STANDARD: +.349 inches X 14.364 gallons / inch = +5.013 gallons (+.339 to + .359) (+4.869 to + 5.157) COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

9 2.9

Convert QT Level Change to gallons:

0 inches X 34.94 gallons / inch = 0 gallons (3) QT Level Change Conversion corr to 68°F QT Vol Change

STANDARD: Determine there is no change for the QT COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

10 2.10

Convert LDST Level Change to gallons:

+0.6 inches X 30.956 gallons / inch = +18.574 gallons (4) LDST Level Change Conversion corr to 68°F LDST Vol Change

STANDARD: +0.6 inches X 30.956 gallons / inch = +18.574 gallons (18.5 to 18.6) COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

(ADM-111) 1.2R Page 9 of 11

11 2.11

Calculate Total Volume Change:

5.013 gallons + 0 gallons + 18.574 gallons = 23.587 gallons PZR Vol Change QT Vol Change LDST Vol Change Total Change

STANDARD: 5.013 gal + 0 gal + 18.574 gal = 23.587 gallons (4.869-5.157) (18.5-18.6) (23.369-23.757) COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

12 2.12

Calculate "RCS Unidentified Leakage Rate":

23.587 gallons ÷ 60 minutes Minutes = 0.393 gpm

Total Change (5) Duration RCS Leakage Rate

STANDARD: 23.587 gallons / 60 minutes = 0.393 gpm (23.369-23.757) (0.389-0.396) COMMENTS:

*CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________

(ADM-111) 1.2R Page 10 of 11


SEQ STEP #

Explanation

6 Necessary data calculation to properly determine manual RCS leakage rate 7 Necessary data calculation to properly determine manual RCS leakage rate 8 Necessary data calculation to properly determine manual RCS leakage rate 9 Necessary data calculation to properly determine manual RCS leakage rate

10 Necessary data calculation to properly determine manual RCS leakage rate



CANDIDATE CUE SHEET


INITIAL CONDITIONS Oconee Unit 1 OAC computer repairs are expected to be extended through turnover for hardware replacement. The Loss of Computer procedure PT/0/A/0600/001A (Loss of Computer) is in progress, however an RCS Leakage Calculation has NOT been performed during this shift. INITIATING CUE The CRS directs you to perform a manual RCS leakage calculation per PT/1/A/0600/010 (Reactor Coolant Leakage) Encl. 13.3 (Manual RCS Leakage Calculation Data Sheet) using the data collected below. Round calculations to three (3) decimal places.

Manual RCS Leakage Data

Parameter Initial Final Time 0015 0115 Pzr level 219.0 inches 220.7 inches Tave Indication 579.0°F 579.3°F Quench Tank Level 86.5 inches 86.5 inches LDST Level 76.7 inches 76.1 inches Power Range NI 100.0% 100.1% RCS NR Pressure 2150 psig 2150 psig Group 7 Control Rod Position 92.7% 92.7%

(ADM-205) 2R Page 1 of 11

*Similar to SRO JPM ADM-S204 and must be performed on the same day


ADM-205

Determine LTOP Requirements


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Determine LTOP Requirements Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 30 minutes K/A Rating(s):

System: GENERIC K/A: 2.2.14 Rating: 3.9/4.3 Task Standard: Per OP/1/A/1104/049, Low Temperature Overpressure Protection, based on conditions provided determine that logic paths required by Enclosure 4.13 will satisfy LTOP requirements. References: OP/1/A/1104/049 Low Temperature Overpressure Protection (LTOP) OP/0/A/1108/001 Curves and General Information Tools/Equipment/Procedures Needed: Unit 1 Conditions For Determining LTOP Requirements (Last page of JPM) OP/1/A/1104/049 (Low Temperature Overpressure Protection) (Rev. 50) OP/0/A/1108/001 Curves and General Information (Rev. 111)

(Note: Below this line is used only for Initial NRC Exams) =============================================================================== Candidate: __________________________________ Time Start: _______________


Comments



None


READ TO OPERATOR


INITIAL CONDITIONS Unit 1 controlling procedure for unit shutdown and cooldown is in progress for entering a refueling outage Plant conditions are as described in the attachment provided CRS has just been notified that the 1HP-120 (RC VOLUME CONTROL) travel stop is NOT installed correctly INITIATING CUE In accordance with Limit and Precaution 2.12 of OP/1/A/1104/049 (Low Temperature Overpressure Protection), determine if LTOP requirements are still met. Document your determination as directed in the Limit and Precaution. Once you have completed the determination, sign and date in the blanks provided on Enclosure 4.13 (LTOP Requirements Logic Diagram) and return it to the examiner.


START TIME: _______

SEQ STEP


1

EXAMINER NOTE: The sequence of steps is NOT critical Determine if LTOP Train 1 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles or identifies the following: • RC-4 open • PORV Auto Operable

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2

Indicate the path that satisfies LTOP requirements for Train 1

STANDARD: Candidate should circle the plant/equipment configuration status and trace the logic path that satisfies Train 1 LTOP requirements as indicated on the attached key

COMMENTS:

___ SAT

___ UNSAT


3

Determine if upper portion LTOP Train 2 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles or identifies the following: • 1A & 1B HPIP’s racked out or in test, and tagged • HPI START FROM PSW POWER CR switch

tagged "Do Not Operate" • 1C HPIP bkr racked out or in test, and tagged • 1CF-1 Closed/handwheel tagged and bkr tagged

open • 1B CFT depressurized to < 373 psig

Examiner Note: Candidate may indicate additional equipment as

meeting LTOP requirements, but only the equipment listed above that satisfies the logic of Enclosure 4.13 is critical.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

4

Indicate the path that satisfies LTOP requirements for the upper portion of Train 2

STANDARD: Candidate should circle the plant/equipment configuration status and trace the logic path that satisfies the upper portion of Train 2 LTOP requirements as indicated on the attached key

COMMENTS:

___ SAT

___ UNSAT


5

Determine if lower portion LTOP Train 2 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles or identifies the following: • HPI NOT in operation and aligned via 1HP-120 • LTOP computer point O1L3153 not in alarm • 1SA-2/C-3, C-4 cards not pulled/not in alarm

“HIGH” • 1SA-18/A-3 card not pulled/not in alarm • 1SA-18/A-4 card not pulled/not in alarm • OAC operable • 1N-121 Tagged Closed • Pzr Htr Bank 4 Deactivated • RCS pressure within limits • Pzr level within limits

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

6

Indicate the path that satisfies LTOP requirements for the lower portion of Train 2

STANDARD: Candidate should circle the plant/equipment configuration status and trace the logic path that satisfies the lower portion of Train 2 LTOP requirements as indicated on the attached key

COMMENTS:

___ SAT

___ UNSAT

7

Indicate that the logic path for LTOP requirements is satisfied

STANDARD: Candidate indicates that the logic paths for LTOP requirements are being satisfied in accordance with the attached key

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


8

Fill in the “Performed By” and “Date/Time” blanks

STANDARD: Candidate completes the “Performed By” and “Date/Time” blanks

COMMENTS:

END TASK

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 Required to determine if Train 1 requirements are met 3 Required to determine if Train 2 requirements are met 5 Required to determine if Train 2 requirements are met 7 Required to determine if all LTOP requirements are met

CANDIDATE CUE SHEET


INITIAL CONDITIONS Unit 1 controlling procedure for unit shutdown and cooldown is in progress for entering a refueling outage Plant conditions are as described in the attachment provided CRS has just been notified that the 1HP-120 (RC VOLUME CONTROL) travel stop is NOT installed correctly INITIATING CUE In accordance with Limit and Precaution 2.12 of OP/1/A/1104/049 (Low Temperature Overpressure Protection), determine if LTOP requirements are still met. Document your determination as directed in the Limit and Precaution. Once you have completed the determination, sign and date in the blanks provided on Enclosure 4.13 (LTOP Requirements Logic Diagram) and return it to the examiner.

Unit 1 Conditions For Determining LTOP Requirements


Procedures in progress: OP/1/A/1102/010, Controlling Procedure for Unit Shutdown

• Encl. 4.10, S/D from Mode 3 to Mode 4 in progress RCS Parameters:

• Tc = 225°F slowly decreasing • RCS Pressure = 285 psig decreasing • Pressurizer Level = 218 inches decreasing slowly • LPI operation (Series Mode) in progress

Primary Systems/Components status:

• All HPI pumps are OFF • 1B and 1C HPI Pumps are racked out and tagged • 1A HPI Pump is racked in the TEST position and tagged • HPI START FROM PSW POWER CR switch tagged "Do Not Operate" • 1HP-26 (1A HP Injection) and 1HP-27 (1A HP Injection) are closed; hand wheels

tagged and breakers tagged open • 1HP-409 switch in CR has been tagged closed • 1HP-410 switch in CR is closed but not tagged • 1CF-1 is Closed / handwheel tagged and breaker tagged open • 1CF-2 is Closed / handwheel tagged and breaker not tagged open • 1A CFT pressure = 375 psig stable • 1B CFT pressure = 368 psig stable • Computer point O1L3153 is not in alarm • Computer points O1X2285 and O1X2256 are locked out • 1SA-2/C-3, 1SA-2/C-4, 1SA-18/A-3, and 1SA-18/A-4 cards are not pulled and not in

alarm • OAC is Operable • 1N-121 is closed and tagged • Pressurizer Heater Bank 4 is deactivated with breaker white tagged • No dedicated LTOP Operator assigned • An LTOP Vent path (≥ 3.6 square inches) is not established • 1RC-4 is open • PORV is operable with setpoint selected to LOW

ADM-305 3B Page 1 of 7


ADM-305

Determine Posting and Access Requirements of LPI Room Based on Plan View

Administrative: Yes Alternate Path: No Alt Path Failure: ____________________________________________________________ Time Critical: No Time Critical Criteria: ________________________________________________________

Prepared By:

Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Determine Posting and Access Requirements of LPI Room Based on Plan View Task Number : (N/A) Alternate Path: (N/A) Time Critical: No Validation Time: 15 minutes K/A Rating(s): System: GEN K/A: 2.3.12 Rating: 3.2/3.7 Task Standard: Determine that:

1. Area should be posted as a High Radiation Area 2. Continuous RP coverage is NOT required. 3. The total accumulated annual dose at completion of job is 1058 mrem.

References: PD-RP-ALL-0001 Radiation Worker Responsibilities (Rev 4) Survey Map of Room 61

Tools/Equipment/Procedures Needed: PD-RP-ALL-0001 Radiation Worker Responsibilities (Rev 4) Survey Map of Room 61



Comments



1. NONE


READ TO OPERATOR


INITIAL CONDITIONS You are a Reactor Operator. An event has occurred which resulted in a significant change in the radiological conditions in Room 61 (LPI and RBS Pumps).

Your total whole body dose this year is 986 mrem.

INITIATING CUE Using the survey map provided, determine the following for Room 61:

• The RP related area posting(s) you would expect to encounter at the entry to Room 61

• If Continuous RP coverage is required for you to perform venting of the 1A LPI Pump

• If venting the 1A LPI pump takes two (2) hours, determine your total accumulated annual dose when the job is complete and you have exited the room based on the following:

o 10 minutes spent at 1A LPI Pump initiating the vent o 5 minutes spent at 1A LPI Pump securing the vent o 1 hours 45 minutes spent in the LDWA


START TIME: _______

SEQ STEP


1

Using the survey map provided, determine the area posting requirements.

STANDARD: Candidate reviews the survey map and determines:

• Based on an area around LPI sump pumps being 930 mr/hr (> 100 mr/hr at 30 cm), the area should be posted as a High Radiation Area (HRA).

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2

Determine RP coverage required to vent the 1A LPI Pump

STANDARD: Candidate determines that continuous RP coverage is NOT required since the area is a HRA.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

3

Determine total accumulated annual dose after completion of job.

STANDARD: Candidate determines the dose as a result of the venting as follows:

.25 hours X 92 mr/hr = 23 mrem

1.75 hours X 28 mr/hr = 49 mrem

23 + 49 = 72 mrem

Determine the total accumulated annual dose following completion of the job.

986 + 72 = 1058 mrem

COMMENTS:

END OF TASK

CRITICAL STEP

___ SAT ___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 Determines General Area posting requirements for the room to be entered. 2 3

Determines RP continuous coverage requirements. Determines total accumulated annual dose.

3.B

CANDIDATE CUE SHEET


INITIAL CONDITIONS You are a Reactor Operator. An event has occurred which resulted in a significant change in the radiological conditions in Room 61 (LPI and RBS Pumps).

Your total whole body dose this year is 986 mrem.

INITIATING CUE Using the survey map provided, determine the following for Room 61:

• The RP related area posting(s) you would expect to encounter at the entry to Room 61

• If Continuous RP coverage is required for you to perform venting of the 1A LPI Pump

• If venting the 1A LPI pump takes two (2) hours, determine your total accumulated annual dose when the job is complete and you have exited the room based on the following:

o 10 minutes spent at 1A LPI Pump initiating the vent o 5 minutes spent at 1A LPI Pump securing the vent o 1 hours 45 minutes spent in the LDWA

ADM-S108 1.2S Page 1 of 8


ADM-S108

Determine if SRO License Requirements are met


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Determine if SRO License Requirements are met Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 15 minutes K/A Rating(s):

System: Gen K/A: 2.1.4 Rating: 3.3/3.8 Task Standard: Completes Form 512-1 and determines requirements of NSD 512 are met.

References: NSD 512 (Maintenance of RO/SRO NRC Licenses) (Rev. 07)

Tools/Equipment/Procedures Needed:

NSD 512 (Maintenance of RO/SRO NRC Licenses) (Rev. 07)

(Note: Below this line is used only for Initial NRC Exams)

===============================================================================



Comments



None


READ TO OPERATOR


INITIAL CONDITIONS You are an SRO. Today is 10/29/16. Your work history for October of this year is as follows:

10/05/16 Worked 10 hours as Unit 3 BOP (day shift). Took turnover at beginning of shift and gave turnover at the end.

10/06/16 Worked 12 hours as Unit 3 BOP (day shift). Took turnover at beginning of shift and gave turnover at end of shift.

10/07/16 Worked 12 hours as Unit 3 OATC (day shift). Took turnover at beginning of shift and gave turnover at end of shift.

10/12/16 Worked 12 hours as Unit 1 BOP (night shift). Took turnover at beginning of shift and gave turnover at end of shift.

10/13/16 Worked 12 hours (night shift). Worked 3 hours as the CR SRO and then 9 hours as WCC SRO. Took turnover at beginning and gave turnover at the end for his CR SRO position and his WCC SRO position .

10/14/16 Worked 12 hours as Unit 1 CR SRO (night shift). Took turnover at beginning of shift and gave turnover at end of shift.


INITIATING CUE The SM directs you to review your work history for October, complete Section 3 of form NSD 512-1 based on the above work history, and determine if you meet NSD 512 requirements to maintain an active SRO license for the following quarter.


START TIME: _______

SEQ STEP


1

Evaluate 10/05/16 work period.

STANDARD: Determines that requirement is not met; <12hrs worked in position.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2


STANDARD: Determines that requirement is met and adds this period to Form 512-1. Required position for 12 hrs. with Turnover at beginning and end of shift.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

3



COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

4



COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


5


STANDARD: Determines that requirement is not met since 9 of the hours were not at a required position (WCC SRO).

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

6



COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

7



COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

8

Compares credited time vs minimum requirements.

STANDARD: Determines that 5 12-hr shifts at a required position meets one of the SRO position requirements and that the NSD-512 requirement of at least 1 shift/qtr at an SRO position is also met; so the individual meets the requirements to maintain their SRO License active.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 Required to determine if minimum On Shift Experience requirements of NSD 512 have

been met. 2 Required to determine if minimum On Shift Experience requirements of NSD 512 have been

met. 3 Required to determine if minimum On Shift Experience requirements of NSD 512 have been





met. 8 Required to make the determination regarding the minimum license requirement.

CANDIDATE CUE SHEET


INITIAL CONDITIONS You are an SRO. Today is 10/29/16. Your work history for October of this year is as follows:

10/05/16 Worked 10 hours as Unit 3 BOP (day shift). Took turnover at beginning of shift and gave turnover at the end.

10/06/16 Worked 12 hours as Unit 3 BOP (day shift). Took turnover at beginning of shift and gave turnover at end of shift.


10/12/16 Worked 12 hours as Unit 1 BOP (night shift). Took turnover at beginning of shift and gave turnover at end of shift.

10/13/16 Worked 12 hours (night shift). Worked 3 hours as the CR SRO and then 9 hours as WCC SRO. Took turnover at beginning and gave turnover at the end for his CR SRO position and his WCC SRO position .

10/14/16 Worked 12 hours as Unit 1 CR SRO (night shift). Took turnover at beginning of shift and gave turnover at end of shift.


INITIATING CUE The SM directs you to review your work history for October, complete Section 3 of form NSD 512-1 based on the above work history, and determine if you meet NSD 512 requirements to maintain an active SRO license for the following quarter.

ADM-S204 2S Page 1 of 11

*Similar to RO JPM ADM-205 and must be performed on the same day


ADM-S204

Determine LTOP Requirements and Required Actions


Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Determine LTOP Requirements and Required Actions Task Number : N/A Alternate Path: No Time Critical: No Validation Time: 35 minutes K/A Rating(s):

System: GENERIC K/A: 2.2.14 Rating: 3.9/4.3 Task Standard: Based on conditions provided, determine that logic paths required by OP/1/A/1104/049, Low Temperature Overpressure Protection, Enclosure 4.13 are NOT satisfied and Tech Spec 3.4.12 Condition F must be entered. References: OP/1/A/1104/049 Low Temperature Overpressure Protection (LTOP) (Rev. 50) OP/0/A/1108/001 Curves and General Information (Rev. 111) Tools/Equipment/Procedures Needed: Unit 1 Conditions For Determining LTOP Requirements (Last page of JPM) OP/1/A/1104/049 (Low Temperature Overpressure Protection) (Rev. 50) OP/0/A/1108/001 Curves and General Information (Rev. 111) Tech Spec 3.4.12 (Low Temperature Overpressure Protection (LTOP) System)




Comments



None


READ TO OPERATOR


INITIAL CONDITIONS Unit 1 controlling procedure for unit shutdown and cooldown is in progress for entering a refueling outage Plant conditions are as described in the attachment provided CRS has just been notified that the 1HP-120 (RC VOLUME CONTROL) travel stop is NOT installed correctly INITIATING CUE In accordance with Limit and Precaution 2.12 of OP/1/A/1104/049 (Low Temperature Overpressure Protection), determine if LTOP requirements are still met. Document your determination as directed in the Limit and Precaution. Once you have completed the determination, sign and date in the blanks provided on Enclosure 4.13 (LTOP Requirements Logic Diagram) and list required actions, if any, below.


START TIME: _______

SEQ STEP


1

EXAMINER NOTE: The sequence of steps is NOT critical Determine if LTOP Train 1 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles the following: • RC-4 open • PORV Auto Operable

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2

Indicate the path that satisfies LTOP requirements for Train 1

STANDARD: Candidate should circle the plant/equipment configuration status and trace the logic path that satisfies Train 1 LTOP requirements as indicated on the attached key

COMMENTS:

___ SAT

___ UNSAT


3

Determine if upper portion LTOP Train 2 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles the following: • 1A & 1B HPIP’s racked out or in test, and tagged • HPI START FROM PSW POWER CR switch

tagged "Do Not Operate" • 1C HPIP bkr racked out or in test, and tagged • 1CF-1 Closed/handwheel tagged and bkr tagged

open • 1B CFT depressurized to < 373 psig

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

4

Indicate the path that satisfies LTOP requirements for the upper portion of Train 2

STANDARD: Candidate should circle the plant/equipment configuration status and trace the logic path that satisfies the upper portion of Train 2 LTOP requirements as indicated on the attached key

COMMENTS:

___ SAT

___ UNSAT


5

Determine if lower portion LTOP Train 2 requirements are met

STANDARD: Candidate Refers to OP/1/A/1104/049 (LTOP) Encl. 4.13 and attachment and circles the following: • HPI NOT in operation and aligned via 1HP-120 • LTOP computer point O1L3153 not in alarm • 1SA-2/C-3, C-4 cars not pulled/not in alarm “HIGH” • 1SA-18/A-3 card not pulled/not in alarm • 1SA-18/A-4 card not pulled/not in alarm • OAC operable • Pzr Htr Bank 4 Deactivated • RCS pressure within limits • Pzr level within limits

Candidate determines the lower portion of Train 2 is NOT satisfied due to 1N-121 being closed only (NOT tagged closed).

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

6

Indicate the path that satisfies LTOP requirements for the lower portion of Train 2

STANDARD: Candidate determines the lower portion of Train 2 is not satisfied due to 1N-121 being closed only (NOT tagged closed).

COMMENTS:

___ SAT

___ UNSAT

7

Indicate that the logic path for LTOP requirements is satisfied

STANDARD: Candidate determines the logic path for LTOP requirements is NOT satisfied due to 1N-121 being closed only (NOT tagged closed).

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


8

Fill in the “Performed By” and “Date/Time” blanks

STANDARD: Candidate completes the “Performed By” and “Date/Time” blanks

COMMENTS:

___ SAT

___ UNSAT

9

Determine required actions.

STANDARD: Candidate determines compensatory measures to monitor for initiation of an LTOP event (dedicated LTOP operator) must be established within 4 hours, in accordance with Tech Spec 3.4.12 Cond. F.

COMMENTS:

END TASK

CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 Required to determine if Train 1 requirements are met 3 Required to determine if Train 2 requirements are met 5 Required to determine if Train 2 requirements are met 7 Required to determine if all LTOP requirements are met 9 Required to determine required actions.

CANDIDATE CUE SHEET


INITIAL CONDITIONS Unit 1 controlling procedure for unit shutdown and cooldown is in progress for entering a refueling outage Plant conditions are as described in the attachment provided CRS has just been notified that the 1HP-120 (RC VOLUME CONTROL) travel stop is NOT installed correctly INITIATING CUE In accordance with Limit and Precaution 2.12 of OP/1/A/1104/049 (Low Temperature Overpressure Protection), determine if LTOP requirements are still met. Document your determination as directed in the Limit and Precaution. Once you have completed the determination, sign and date in the blanks provided on Enclosure 4.13 (LTOP Requirements Logic Diagram) and list required actions, if any, below.

Unit 1 Conditions For Determining LTOP Requirements


Procedures in progress: OP/1/A/1102/010, Controlling Procedure for Unit Shutdown

• Encl. 4.10, S/D from Mode 3 to Mode 4 in progress RCS Parameters:

• Tc = 225°F slowly decreasing • RCS Pressure = 285 psig decreasing • Pressurizer Level = 218 inches decreasing slowly • LPI operation (Series Mode) in progress

Primary Systems/Components status:

• All HPI pumps are OFF • 1B and 1C HPI Pumps are racked out and tagged • 1A HPI Pump is racked in the TEST position and tagged • HPI START FROM PSW POWER CR switch tagged "Do Not Operate" • 1HP-26 (1A HP Injection) and 1HP-27 (1A HP Injection) are closed; hand wheels

tagged and breakers tagged open • 1HP-409 switch in CR has been tagged closed • 1HP-410 switch in CR is closed but not tagged • 1CF-1 is Closed / handwheel tagged and breaker tagged open • 1CF-2 is Closed / handwheel tagged and breaker not tagged open • 1A CFT pressure = 375 psig stable • 1B CFT pressure = 368 psig stable • Computer point O1L3153 is not in alarm • Computer points O1X2285 and O1X2256 are locked out • 1SA-2/C-3, 1SA-2/C-4, 1SA-18/A-3, and 1SA-18/A-4 cards are not pulled and not in

alarm • OAC is Operable • 1N-121 is closed • Pressurizer Heater Bank 4 is deactivated with breaker white tagged • No dedicated LTOP Operator assigned • An LTOP Vent path (≥ 3.6 square inches) is not established • 1RC-4 is open • PORV is operable with setpoint selected to LOW



ADM-S405 Determine Emergency Classification and Protective Action

Recommendations

Administrative: Yes Alternate Path: No Alt Path Description: _________________________________________________________ Time Critical: Yes Time Critical Criteria: ________________________________________________________

Prepared By: Date:

EP Review By: Date:

Reviewed By: Date:

Approved By: Date:


REGION II


Task Title : Determine Emergency Classification and Protective Action Recommendations

Task Number : N/A Alternate Path: No Time Critical: Yes Validation Time: 30 minutes K/A Rating(s):

System: Gen K/A: 2.4.38 Rating: 2.4/4.4 Task Standard: Appropriate classification is determined and associated Emergency Notification Form is completed.

References:

RP/0/A/1000/01, Emergency Classification (Rev. 05) RP/0/A/1000/02, Control Room Emergency Coordinator Procedure (Rev. 11) RP/0/A/1000/015A, Offsite Communications From The Control Room BASIS Document (Volume “A”, Section “D” of the Emergency Plan) Tools/Equipment/Procedures Needed: RP/0/A/1000/01, Emergency Classification (Rev. 05) RP/0/A/1000/02, Control Room Emergency Coordinator Procedure (Rev. 11)




Comments



None


READ TO OPERATOR


INITIAL CONDITIONS

Time: 3 hours ago:

• The Oconee County Sheriff reported that a large group of “anti-nukes” had assembled at the Visitor Center

CURRENT CONDITIONS

• ALL three Unit’s at 100% power

• Oconee County Sheriff reports that an explosion has occurred at the Keowee Hydro Dam, however the dam shows only minor damage with no breach

• Keowee personnel report no visible damage to the KHUs

• Security Shift Supervisor notifies the SM that security is responding to armed adversaries in the vicinity of the Unit 1 Main Transformer

INITIATING CUE

You are to perform the required actions of the Emergency Coordinator by referring to RP/0/A/1000/01, Emergency Classification: 1. Determine Emergency Classification at present time. 2. Complete appropriate Emergency Notification Form (ENF) for the current conditions. Inform the examiner when you have made the classification. THIS IS A TIME CRITICAL JPM Note: Do not use Emergency Coordinator’s judgment as the basis for classifying the

event. When required, an operator will maintain the Emergency Coordinator’s Log and assume the duties of the Control Room Offsite Communicator.


START TIME: _______

SEQ STEP


1

Classify the Event

STANDARD: Refer to RP/0/A/1000/01 (Emergency Classification) Enclosure 4.6 (Fire/Explosions and Security Actions).

Classify the event as a “Site Area Emergency”(4.6.S.1) due to the following:

A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by Security Shift Supervision

STOP TIME #1: Time SAE Declared __________ (Actual time) (SAT is < Start Time + 15 minutes)

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

2 1.1

Commence the Off-Site Notification Form.

STANDARD: Go to RP/0/A/1000/002 (Control Room Emergency Coordinator Procedure) and initiate procedure by determining symptoms for entry exist and check Step 1.1

COMMENTS:

___ SAT

___ UNSAT

3 2.1

IF an EAL exists, Declare the appropriate Emergency Classification level.

Classification __SAE_____ (UE, ALERT, SAE, GE)

Time Declared: _________

STANDARD: Candidate documents SAE and time declared from step 1.

COMMENTS:

___ SAT

___ UNSAT


4 2.2

IF a Security event is in progress THEN GO TO Step 2.4

STANDARD: Determine Step 2.2 does apply and GO TO Step 2.4

COMMENTS:

___ SAT

___ UNSAT

5 2.4

Direct Control Room Offsite Communicator(s) to perform the following:

• Record Name _______________________

• REFER TO RP/0/A/1000/015 A (Offsite Communications From The Control Room), Immediate Actions steps 2.1 and 2.2 AND Enclosure 4.7 (Guidelines for Manually Transmitting a Message) in preparation for notifying offsite agencies.

STANDARD: Any name (real or imaginary) is acceptable. Continue to step 2.5 COMMENTS:

___ SAT

___ UNSAT

6 2.5

IF AT ANY TIME changing plant conditions require an emergency classification upgrade, ...

STANDARD: An upgrade is not expected. Candidate should circle the step

number and continue to step 2.6. COMMENTS:

___ SAT

___ UNSAT


7 2.6

NOTE: If more than one EAL of a classification level is met, use the EAL description of the most interest to offsite agencies.

Additional message sheets listing other information of interest to offsite agencies (e.g. transporting injured personnel) may be sent, if needed. For the case of more than one unit with different EAL entry conditions, the SM would declare the highest applicable EAL on the initial ENF to meet the time limit.

Obtain the applicable Emergency Notification Form (ENF) from the control room and complete as follows:

STANDARD: Initial Site Area Emergency form # 4.6.S.1 is selected and candidate continues to Step 2.6.1.

COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

8 2.6.1

IF pre-printed forms are NOT available , perform the following: A. Manually perform initial notification per RP/0/A/1000/015 A, enclosure for

Guidelines for Manually Completing an Initial Message.

B. GO TO Step 2.7.

STANDARD: Candidate determines pre-printed forms are available and this step does not apply.

Continues to step 2.6.2

COMMENTS:

___ SAT

___ UNSAT


9 2.6.2

Ensure EAL # and description as determined by RP/0/A/1000/001 matches Line 4.

STANDARD: Candidate ensures EAL # on line 4 is 4.6.S.1 and EAL description is Hostile Action within the Protected Area. Current plant conditions DO NOT threaten public safety.

Continues to step 2.6.3.

COMMENTS:

___ SAT

___ UNSAT

10 2.6.3

Enter message # - Beginning at one, message numbers will increment every message, regardless of whether it is an UPGRADE.

STANDARD: Candidate enters 1 as the message #.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT

11 2.6.4

Line 1 - Mark appropriate box "Drill", "Actual Declaration" or "Termination"

STANDARD: Candidate marks Drill in line 1.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


12 2.6.5

Line 4 - Military time and date of declaration (Refer to time in Step 2.1)

STANDARD: Candidate enters today's date and declaration time from step 2.1.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


13 2.6.6

NOTE: • An EMERGENCY RELEASE is defined as an unplanned, quantifiable, airborne radiological release to the environment attributable to the emergency event.

• Some examples of conditions where airborne releases would be

considered In-Progress are provided below. Not all bullets are required.

• This list is not intended to be all inclusive.

• The SM should then evaluate whether the release in-progress is considered attributable to the emergency event. If a radiological release was occurring prior to the emergency event and the radioactive release rate did NOT change because of the emergency event, then a release is NOT In-Progress and ENF line 5 would be marked “None”.

Release Source Potential Indications

SG Tube Leak • 1,2,3 RIA-40 in alarm or rising • Steam pressure maintained or rising in an isolated SG • Level maintained or rising in an isolated SG

Release from Unit Vent • 1,2,3 RIA-45 or 46 in alarm or rising • Notified by RP of significant release to the unit vent

Release from Containment • Containment pressure greater than 1 psig • Actual Containment breach is determined

Line 5 - Perform one of the following: A. IF there is NO emergency release occurring as a result of the declared

event as defined above and as determined by the Shift Manager, mark "NONE".

B. IF there is an ongoing emergency release in accordance with the definition

above as determined by the Shift Manager, mark "IS OCCURRING".

C. IF an emergency release as defined above and as determined by the Shift Manager has occurred previously during the declared emergency and is no longer occurring, mark "HAS OCCURRED".

STANDARD: Candidate determines NO release is occurring and marks NONE.


COMMENTS:

CRITICAL STEP

___ SAT

___ UNSAT


14 2.6.7

Line 6 - IF Imminent Failure exists for Jocassee OR Keowee Hydro dam/dike, ensure following: A. Check "OTHER".

B. Write: Move residents living downstream of the Keowee Hydro Project dams to higher ground. Prohibit traffic flow across bridges identified on your inundation maps until the danger has passed.

STANDARD: Candidate determines imminent failure does NOT exist for Jocassee or Keowee Hydro dam/dike.

Candidate may mark NONE or leave step blank. Either is acceptable.


COMMENTS:

___ SAT

___ UNSAT


15 2.6.8

NOTE: LINE 12 should be used to provide information important to offsite agencies. The following are examples of information which should be provided:

• Emergency Response Actions Underway (for example, site

evacuation) • Requests for offsite assistance (for example, ambulance or fire

support) • Facility Activation Status • Injured/contaminated individuals • Any reason causing/requiring a PAR change • Estimate of any surface contamination in the plant, onsite or

offsite

• If EAL description includes an "or". (example -Hostile action based EALs that include both a Hostile action "or" an airliner attack, provide more specific information in remarks section).

Line 12 - IF the SM has no remarks, write "None"

STANDARD: Candidate may include information related to the event or write None in line 12. Either is acceptable.


COMMENTS:

___ SAT

___ UNSAT


16 2.6.9

Line 13 - SM signature, CURRENT Time/Date (MUST SIGN)

STANDARD: Candidate signs and enters current time/date in line 13.

STOP TIME #2: Time for Notification __________ (Actual time) (SAT is < Stop Time #1 + 15 minutes)

COMMENTS:

END TASK

CRITICAL STEP

___ SAT

___ UNSAT

TIME STOP: ________



SEQ STEP #

Explanation

1 The candidate needs to be able to utilize the procedure and determine the conditions meet a

Site Area Emergency classification. This is a time critical step. The candidate needs to declare the SAE within 15 minutes of beginning the JPM. (The start of the JPM is the beginning of the assessment period)

7 The correct form that matches the EAL # is selected.

10 The correct message number is input on the ENF.

11 Drill is marked on the ENF.

12 Declaration time and date are input on the ENF.

13 None is marked on line 5 of the ENF to indicate no release in progress.

16 The candidate signs and enters time and date on line 13 of the ENF. This is a time critical step. The time entered must be within 15 minutes of the declaration time in step 1.

CANDIDATE CUE SHEET


INITIAL CONDITIONS

Time: 3 hours ago:

• The Oconee County Sheriff reported that a large group of “anti-nukes” had assembled at the Visitor Center

CURRENT CONDITIONS

• ALL three Unit’s at 100% power

• Oconee County Sheriff reports that an explosion has occurred at the Keowee Hydro Dam, however the dam shows only minor damage with no breach

• Keowee personnel report no visible damage to the KHUs

• Security Shift Supervisor notifies the SM that security is responding to armed adversaries in the vicinity of the Unit 1 Main Transformer

INITIATING CUE

You are to perform the required actions of the Emergency Coordinator by referring to RP/0/A/1000/01, Emergency Classification: 1. Determine Emergency Classification at present time. 2. Complete appropriate Emergency Notification Form (ENF) for the current conditions. Inform the examiner when you have made the classification. THIS IS A TIME CRITICAL JPM Note: Do not use Emergency Coordinator’s judgment as the basis for classifying the

event. When required, an operator will maintain the Emergency Coordinator’s Log and assume the duties of the Control Room Offsite Communicator.

ES-401, REV 10 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 · ES-401, REV 70 T2G7 PWR EXAMINATION...

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Transcript of ES-401, REV 10 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 · ES-401, REV 70 T2G7 PWR EXAMINATION...