ES-401 PWR Examination Outline FORM ES-401-2 · 28,49 003 Reactor Coolant Pump 04 0 2 Effects of...

ES-401 PWR Examination Outline FORM ES-401-2

Facility Name: Date of Exam:

K 1

K 2

K 3

K 4

K 5

K 6

A 1

A 2

A 3

A 4

G *

Total Total

1 3 2 4 3 3 3 18 6

2 1 1 2 2 2 1 9 4

Tier Totals 4 3 6 5 5 4 27 10

1 3 2 2 3 2 3 2 4 3 2 2 28 5

2 1 1 1 2 1 1 1 0 0 1 1 10 0 2 3

Tier Totals 4 3 3 5 3 4 3 4 3 3 3 38 8

1 2 3 4

1 2 2 2

Note: 1.

Note: 2.

Note: 3.

Note: 4.

Note: 5.

Note: 6.

Note: 7.*

Note: 8.

Note: 9.

ES-401, 21 of 33

2

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

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

Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the “Tier Totals” in each K/A category shall not be less than two).

4

3 2

4 0

4

2. Plant Systems

3. Generic Knowledge and Abilities Categories

1

710

1

3 2

32

1. Emergency &

Abnormal Plant

Evolutions 8

N/A N/A

Tier GroupSRO-Only PointsRO K/A Category Points

A 2 G *

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

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.

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

Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

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 applicable K/As.

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

3 2

5 3

2

ES-401 2 Form ES-401-2

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

Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)

Q# E/APE # / Name / Safety FunctionK1

K2

K3

A1

A2

G K/A Topic(s) IR #

1 000007 Reactor Trip - Stabilization - Recovery / 102

Shutdown margin 3.4 1

2 000008 Pressurizer Vapor Space Accident / 308

PRT level pressure and temperature 3.8 1

3 000009 Small Break LOCA / 314

Actions to be taken if PTS limits are violated 3.8 1

4 000011 Large Break LOCA / 303

Consequences of managing LOCA with loss of CCW 3.7 1

5000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4

02.22

Knowledge of limiting conditions for operations and safety limits. 4.0 1

000022 Loss of Rx Coolant Makeup / 2 0

6 000025 Loss of RHR System / 401

RHR heat exchangers 2.9 1

7 000026 Loss of Component Cooling Water / 806

Control of flow rates to components cooled by the CCWS 2.9 1

8000027 Pressurizer Pressure Control System Malfunction / 3

03

Controllers and positioners 2.6 1

9 000029 ATWS / 101

Reactor nucleonics and thermo-hydraulics behavior 2.8 1

000038 Steam Gen. Tube Rupture / 3 0

10000040 Steam Line Rupture - Excessive Heat Transfer / 4

06

Containment temperature and pressure considerations 3.4

WE12 Uncontrolled Depressurization of all Steam Generators / 4

000054 (CE/E06) Loss of Main Feedwater / 4 0

11 000055 Station Blackout / 602

Actions contained in EOP for loss of offsite and onsite power 4.3 1

12 000056 Loss of Off-site Power / 617

Operational status of PZR backup heaters 3.4 1

000057 Loss of Vital AC Inst. Bus / 6 0

13 000058 Loss of DC Power / 603

Vital and battery bus components 3.1 1

14 000062 Loss of Nuclear Svc Water / 402

The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS

3.6 1

15 000065 Loss of Instrument Air / 804.11

Knowledge of abnormal condition procedures. 4.0 1

W/E04 LOCA Outside Containment / 3 0

16 W/E11 Loss of Emergency Coolant Recirc. / 404.20

Knowledge of the operational implications of EOP warnings, cautions, and notes.

3.8 1

17BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4

03

Annunciators and conditions indicating signals, and remedial actions associated with the Loss of Secondary Heat Sink

3.9 1

18000077 Generator Voltage and ElectricGrid Disturbances / 6

01

Reactor and turbine trip criteria 3.9 1

K/A Category Totals: 3 2 4 3 3 3 18

ES-401, 22 of 33

Group Point Total:

1

ES-401 3 Form ES-401-2


Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)


K2

K3

A1

A2

G K/A Topic(s) IR #

19 000001 Continuous Rod Withdrawal / 1 03Proper actions to be taken if automatic safety functions have not taken place

4.5 1

20 000003 Dropped Control Rod / 1 03 Relationship of reactivity and reactor power to rod movement 3.5 1

000005 Inoperable/Stuck Control Rod / 1 0

21 000024 Emergency Boration / 1 02 Actions contained in EOP for emergency boration 4.2 1

000028 Pressurizer Level Malfunction / 2 0

000032 Loss of Source Range NI / 7 0

000033 Loss of Intermediate Range NI / 7 0

000036 Fuel Handling Accident / 8 0

22 000037 Steam Generator Tube Leak / 3 01 Maximum controlled depressurization rate for affected S/G 3.7 1

000051 Loss of Condenser Vacuum / 4 0

000059 Accidental Liquid RadWaste Rel. / 9 0

000060 Accidental Gaseous Radwaste Rel. / 9 0

000061 ARM System Alarms / 7 0

000067 Plant Fire On-site / 8 0

000068 Control Room Evac. / 8 0

23 000069 Loss of CTMT Integrity / 5 03 Personnel access hatch and emergency access hatch 2.8

W/E14 High Containment Pressure / 5

24 000074 Inad. Core Cooling / 4 06 RCPS 3.6

W/E06 Degraded Core Cooling / 4

W/E07 Saturated Core Cooling / 4

25 000076 High Reactor Coolant Activity / 9 06 Actions contained in EOP for high reactor coolant activity 3.2 1

W/E01 Rediagnosis / 3

26 W/E02 SI Termination / 3 02Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments

3.5

27 W/E13 Steam Generator Over-pressure / 404.35

Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

3.8 1

W/E15 Containment Flooding / 5 0

W/E16 High Containment Radiation / 9 0

W/E03 LOCA Cooldown - Depress. / 4 0

W/E09 Natural Circulation Operations / 4

W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4

W/E08 RCS Overcooling - PTS / 4 0


ES-401, 23 of 33

Group Point Total:

1

0

1

1

ES-401 4 Form ES-401-2


Plant Systems - Tier 2/Group 1 (RO)

Q# System # / NameK1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

28,49 003 Reactor Coolant Pump 0402

Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow; RCP pump and motor bearing temperatures

3.2; 2.9

2

50 004 Chemical and Volume Control15

PZR pressure and temperature 3.5 1

30 005 Residual Heat Removal11

RWST 3.5 1

55,31 006 Emergency Core Cooling16

13

Interlocks between RHR valves and RCS; Inadvertent SIS actuation

3.2; 3.9

2

32 007 Pressurizer Relief/Quench Tank09

Relationships between PZR level and changing levels of the PRT and bleed holdup tank

2.5 1

52,33 008 Component Cooling Water04

02

RCS, in order to determine source(s) of RCS leakage into the CCWS; CCW pump, including emergency backup

3.3; 3 2

34 010 Pressurizer Pressure Control01

Pressure detection systems 2.7 1

51,35 012 Reactor Protection02

01.32

Loss of instrument power; Ability to explain and apply system limits and precautions.

3.6; 3.8

2

36013 Engineered Safety Features Actuation

01

Fuel 4.4 1

37 022 Containment Cooling03

Automatic containment isolation 3.6 1

025 Ice Condenser 0

29,38 026 Containment Spray09

06

Prevention of path for escape of radioactivity from containment to the outside (interlock on RWST isolation after swapover); Containment spray pump cooling

3.7; 2.7

2

39 039 Main and Reheat Steam05

Bases for RCS cooldown limits 2.7 1

40 059 Main Feedwater04

Feeding a dry S/G 2.9 1

41 061 Auxiliary/Emergency Feedwater02

Pumps 2.6 1

42 062 AC Electrical Distribution01

All breakers (including available switchyard) 3.3 1

43 063 DC Electrical Distribution01

Major DC loads 2.9 1

44,53 064 Emergency Diesel Generator08

04.34

Fuel oil storage tanks; Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

3.2; 4.2

2

45 073 Process Radiation Monitoring02

Detector failure 2.7 1

46 076 Service Water07

ESF loads 3.7 1

47,54 078 Instrument Air02

01

Service air; Air pressure2.7; 3.1

2

48 103 Containment01

Containment isolation 3.9 1

0

K/A Category Totals: 3 2 2 3 2 3 2 4 3 2 2 28

ES-401, Page 24 of 33

Group Point Total:

ES-401 5 Form ES-401-2


Plant Systems - Tier 2/Group 2 (RO)


K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

56 001 Control Rod Drive01

One-line diagram of power supply to M/G sets 3.5 1

57 002 Reactor Coolant05

Detection of RCS leakage 3.8 1

58 011 Pressurizer Level Control12

Criteria and purpose of PZR level program 2.7 1

014 Rod Position Indication 0

59 015 Nuclear Instrumentation02

Discriminator/compensation circuits 2.6 1

60 016 Non-nuclear Instrumentation04

MFW System 2.6 1

017 In-core Temperature Monitor 0

027 Containment Iodine Removal 0

028 Hydrogen Recombiner and Purge Control

0

029 Containment Purge 0

61 033 Spent Fuel Pool Cooling04.31

Knowledge of annunciator alarms, indications, or response procedures.

4.2 1

034 Fuel Handling Equipment 0

62 035 Steam Generator12

RPS 3.7 1

63 041 Steam Dump/Turbine Bypass Control02

Steam pressure 3.1 1

64 045 Main Turbine Generator06

Turbine stop valves 2.8 1

055 Condenser Air Removal 0

056 Condensate 0

068 Liquid Radwaste 0

071 Waste Gas Disposal 0

072 Area Radiation Monitoring 0

075 Circulating Water 0

079 Station Air 0

65 086 Fire Protection01

Adequate supply of water for FPS 3.1 1


ES-401, Page 25 of 33

Group Point Total:

ES-401 2 Form ES-401-2


Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)


K2

K3

A1

A2

G K/A Topic(s) IR #

000007 Reactor Trip - Stabilization - Recovery / 1 0

000008 Pressurizer Vapor Space Accident / 3 0

76 000009 Small Break LOCA / 302

Possible leak paths 3.8 1

000011 Large Break LOCA / 3 0

000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4

0

000022 Loss of Rx Coolant Makeup / 2 0

000025 Loss of RHR System / 4 0

000026 Loss of Component Cooling Water / 8 0

000027 Pressurizer Pressure Control System Malfunction / 3

0

000029 ATWS / 1 0

78 000038 Steam Gen. Tube Rupture / 309

Existence of natural circulation, using plant parameters 4.2 1

79000040 Steam Line Rupture - Excessive Heat Transfer / 4

01.07

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.7

WE12 Uncontrolled Depressurization of all Steam Generators / 4

80 000054 (CE/E06) Loss of Main Feedwater / 403

Conditions and reasons for AFW pump startup 4.2 1

000055 Station Blackout / 6 0

000056 Loss of Off-site Power / 6 0

81 000057 Loss of Vital AC Inst. Bus / 602.22

Knowledge of limiting conditions for operations and safety limits. 4.7 1

000058 Loss of DC Power / 6 0

000062 Loss of Nuclear Svc Water / 4 0

000065 Loss of Instrument Air / 8 0

77 W/E04 LOCA Outside Containment / 301

Facility conditions and selection of appropriate procedures during abnormal and emergency operations

4.3 1

W/E11 Loss of Emergency Coolant Recirc. / 4 0

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4

0

000077 Generator Voltage and ElectricGrid Disturbances / 6

0


ES-401, 22 of 33

Group Point Total:

1

ES-401 3 Form ES-401-2


Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)


K2

K3

A1

A2

G K/A Topic(s) IR #

000001 Continuous Rod Withdrawal / 1 0

000003 Dropped Control Rod / 1 0

000005 Inoperable/Stuck Control Rod / 1 0

000024 Emergency Boration / 1 0

83 000028 Pressurizer Level Malfunction / 2 08 PZR level as a function of power level 3.5 1

000032 Loss of Source Range NI / 7 0

000033 Loss of Intermediate Range NI / 7 0

000036 Fuel Handling Accident / 8 0

000037 Steam Generator Tube Leak / 3 0

000051 Loss of Condenser Vacuum / 4 0

000059 Accidental Liquid RadWaste Rel. / 9 0

000060 Accidental Gaseous Radwaste Rel. / 9 0

000061 ARM System Alarms / 7 0

85 000067 Plant Fire On-site / 8 15 Requirements for establishing a fire watch 3.9 1

000068 Control Room Evac. / 8 0

000069 Loss of CTMT Integrity / 5

W/E14 High Containment Pressure / 5

000074 Inad. Core Cooling / 4

84 W/E06 Degraded Core Cooling / 4 02Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments

4.1

W/E07 Saturated Core Cooling / 4

000076 High Reactor Coolant Activity / 9 0

W/E01 Rediagnosis / 3

82 W/E02 SI Termination / 3 01Facility conditions and selection of appropriate procedures during abnormal and emergency operations

4.2

W/E13 Steam Generator Over-pressure / 4 0

W/E15 Containment Flooding / 5 0

W/E16 High Containment Radiation / 9 0

W/E03 LOCA Cooldown - Depress. / 4 0

W/E09 Natural Circulation Operations / 4

W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4

W/E08 RCS Overcooling - PTS / 4 0


ES-401, 23 of 33

Group Point Total:

1

0

0

1

ES-401 4 Form ES-401-2


Plant Systems - Tier 2/Group 1 (SRO)


K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

86 003 Reactor Coolant Pump05

Effects of VCT pressure on RCP seal leakoff flows 2.8 1

004 Chemical and Volume Control 0

87 005 Residual Heat Removal02.25

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

4.2 1

006 Emergency Core Cooling 0

007 Pressurizer Relief/Quench Tank 0

88 008 Component Cooling Water09

Results of excessive exit temperature from the letdown cooler, including the temperature effects on ion-exchange resins

2.8 1

010 Pressurizer Pressure Control 0

012 Reactor Protection 0

013 Engineered Safety Features Actuation

0

022 Containment Cooling 0

025 Ice Condenser 0

89 026 Containment Spray05

Failure of chemical addition tanks to inject 4.1 1

039 Main and Reheat Steam 0

059 Main Feedwater 0

90 061 Auxiliary/Emergency Feedwater04.18

Knowledge of the specific bases for EOPs. 4.0 1

062 AC Electrical Distribution 0

063 DC Electrical Distribution 0

064 Emergency Diesel Generator 0

073 Process Radiation Monitoring 0

076 Service Water 0

078 Instrument Air 0

103 Containment 0

0


ES-401, Page 24 of 33

Group Point Total:

ES-401 5 Form ES-401-2


Plant Systems - Tier 2/Group 2 (SRO)


K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

001 Control Rod Drive 0

002 Reactor Coolant 0

011 Pressurizer Level Control 0

014 Rod Position Indication 0

91 015 Nuclear Instrumentation01

Power supply loss or erratic operation 3.9 1

016 Non-nuclear Instrumentation 0

017 In-core Temperature Monitor 0

027 Containment Iodine Removal 0

028 Hydrogen Recombiner and Purge Control

0

029 Containment Purge 0

033 Spent Fuel Pool Cooling 0

034 Fuel Handling Equipment 0

92 035 Steam Generator01

Faulted or ruptured S/Gs 4.6 1

041 Steam Dump/Turbine Bypass Control 0

045 Main Turbine Generator 0

055 Condenser Air Removal 0

056 Condensate 0

068 Liquid Radwaste 0

071 Waste Gas Disposal 0

072 Area Radiation Monitoring 0

075 Circulating Water 0

079 Station Air 0

93 086 Fire Protection04.11

Knowledge of abnormal condition procedures. 4.2 1


ES-401, Page 25 of 33

Group Point Total:

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3

Facility Name: Date of Exam:

Q# IR # IR #

66 2.1. 26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). 3.4 1

67 2.1. 03 Knowledge of shift or short-term relief turnover practices. 3.7 1

68 2.1. 04 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, “no-solo” operation, maintenance of active license status, 10CFR55, etc. 3.3 1

94 2.1. 34 Knowledge of primary and secondary plant chemistry limits. 3.5 1

2.1.

2.1.

Subtotal 3 1

69 2.2. 39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. 3.9 1

70 2.2. 37 Ability to determine operability and/or availability of safety related equipment. 3.6 1

95 2.2. 42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. 4.6 1

96 2.2. 20 Knowledge of the process for managing troubleshooting activities. 3.8 1

2.2.

2.2.

Subtotal 2 2

71 2.3. 05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. 2.9 1

72 2.3. 14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. 3.4 1

73 2.3. 12Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

3.2 1

97 2.3. 11 Ability to control radiation releases. 4.3 1

98 2.3. 15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. 3.1 1

2.3.

Subtotal 3 2

74 2.4. 06 Knowledge of EOP mitigation strategies. 3.7 1

75 2.4. 32 Knowledge of operator response to loss of all annunciators. 3.6 1

99 2.4. 26 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage. 3.6 1

100 2.4. 16Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

4.4 1

2.4.

2.4.

Subtotal 2 210 7

ES-401, Page 26 of 33

RO SRO-Only

Tier 3 Point Total

3. Radiation Control

4. Emergency Procedures / Plan

2. Equipment Control

Category Topic

1. Conduct of Operations

K/A #

ES-401 Record of Rejected K/As Form ES-401-4

Tier / Group

Randomly Selected K/A

Reason for Rejection

1/1 011 EA2.03 Question 4

Could not write a question with regard to existence of natural circulation during a Large Break LOCA as there is no natural circulation in the RCS with a large break LOCA. Replaced K/A 011 EA2.09 with K/A 011 EA2.03.

1/2 003 AK1.03 Question 20

Could not write an operationally valid question for an Inoperable/Stuck Rod Xenon transient as CPNPP does not allow extended operation with an inoperable or stuck control rod. Replaced K/A 005 AK1.03 with 003 AK1.03.

1/2 W/E 13 G.2.4.35 Question 27

Could not write a question with regard to immediate operator actions without reference to procedures during a steam generator over-pressure event as there are no immediate operator actions which would be performed without procedure reference when addressing steam generator over-pressure. Steam generator over-pressure is a YELLOW path FRG at CPNPP. Replaced K/A W/E 13 G.2.4.49 with W/E 13 G.2.4.35.

2/1 003 K5.04 Question 28

Could not write an operationally valid question on the operational implications of the effects of RCP shutdown on TAVE as CPNPP trips the Reactor at all times in Modes 1 and 2 when an RCP trips, thus the effects on TAVE are only evaluated with respect to a reactor trip where temperature control is what is monitored versus loop variations. Replaced K/A 003 K5.03 with 003 K5.04.

2/1 004 K4.09 Question 29

Could not develop operationally valid distracters for the effect of a loss or malfunction on the spray/heater combination in the pressurizer to assure uniform boron concentration. Replaced K/A 004 K6.01 with 026 K4.09.


Tier / Group



2/1 007 A4.09 Question 32

Identified that the Recognition of a Leaking PORV or Safety Valve via PRT indications was an overlap to Question 2 which was PRZR Vapor Space Accident monitoring of the PRT pressure and temperature. Replaced K/A 007 A4.10 with 007 A4.09.

2/1 061 K6.02 Question 41

Questions 41 and 52 both sampled system 061 in the K1 category. Replaced K/A 061 K1.02 with K/A 061 K6.02 to preventing over sampling of the K1 category for auxiliary/emergency feedwater. K6 was selected to provide balance in outline.

2/2 033 G.2.4.31 Question 61

Could not write a question with regard to immediate operator actions performed without reference to procedures for the Spent Fuel Pool Cooling System as there are no immediate operator actions which would be performed without procedure reference when addressing Spent Fuel Pool Cooling system malfunctions. This is addressed in an AOP at CPNPP. Replaced K/A 033 G.2.4.49 with 033 G.2.4.31.

2/2 086 K4.01 Question 65

Could not write a question as the loss or malfunction of the Fire Protection system would not affect, fire, smoke and heat detectors. Replaced K/A 086 K6.04 with 086 K4.01.

1/1 W/E 04 EA2.01 Question 77

Could not write an operationally valid question on calculation of expected values of flow in the RCS loop with RCP secured. CPNPP trips the Reactor at all times in Modes 1 and 2 when an RCP trips, thus the resulting loop flow is not considered procedurally as forced flow exist in the remaining three loops. Replaced K/A 015 AA2.07 with W/E 04 EA2.01.


Tier / Group



1/1 054 AA2.03 Question 80

Could not write an SRO only question with regard to immediate operator actions during a loss of main feedwater event. The immediate operator actions in the AOP for a Main Feedwater pump trip are actions the ROs perform from memory at CPNPP so this is not an SRO only topic for CPNPP. Additionally, other areas which could make the question an SRO level are lacking in loss of main feedwater as nearly the entire system is non safety-related and procedure selection is at the RO knowledge level. Replaced K/A 054 G.2.4.49 with K/A 054 AA2.03.

1/1 057 G.2.2.22 Question 81

Could not write a question as system limits and precautions do not apply to Station Blackout. Replaced K/A 055 G.2.1.32 with 057 G.2.2.22.

1/2 005 AA2.03 Question 82

Generic K/A 2.1.43 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 003 G.2.1.43 with 005 AA2.03.

1/2 074 EA2.02 Question 84

Could not write an SRO level question on loss of condenser vacuum as this is a non-safety system without Technical Specifications. Replaced K/A 051 G.2.4.46 with 074 EA2.02.

2/1 061 G.2.4.18 Question 90

Generic K/A 2.4.23 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 061 G.2.4.23 with 061 G.2.4.18.

2/2 035 A2.01 Question 92

Could not write a Technical Specification question on Spent Fuel Pool Cooling System as there are no Technical Specifications for this system. A similar topic was sampled on Question 61. Replaced K/A 033 G.2.2.25 with K/A 035 A2.01.


Tier / Group



2/2 086 G.2.4.11 Question 93

Generic K/A 2.4.25 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 086 G.2.4.25 with 086 G.2.4.11.

3 G.2.1.34 Question 94

Generic K/A 2.1.15 has no 10CFR43 tie in NUREG 1122. Replaced K/A G.2.1.15 with K/A G.2.1.34

1/1 077 AK3.01 Question 18

Could not develop an operationally valid question including distracters for the effect of generator voltage and electric grid disturbances on the turbine/generator control. Replaced K/A 077 AK2.07 with 077 AK3.01.

2/2 015 K6.02 Question 59

Could not develop an operationally valid question on disagreement between channels without discussing annunciators or alarms. Replaced K/A 015 A3.04 with 015 K6.02.

3 G.2.3.15 Question 98

Could not develop an SRO level question on knowledge of radiation exposure limits. Replaced K/A G.2.3.04 with G.2.3.15.

2/1 010 K6.01 Question 34

Could not develop an operationally valid question on the ability to monitor the Pressurizer Pressure Control System that did not conflict with the operational exam. Replaced K/A 010 A3.02 with 010 K6.01.

2/1 008 K1.04 Question 52

Auxiliary Feedwater System was over sampled. Replaced K/A 061 K1.04 with 008 K1.04.

2/2 016 K3.04 Question 60

Auxiliary Feedwater System was over sampled. Replaced K/A 016 K3.06 with 016 K3.04.

Administrative Topics Outline

CPNPP NRC 2014 ES-301-1 RO Admin JPM Outline.docx

Facility: CPNPP Units 1 and 2 Date of Examination: 06/09/14

Examination Level RO Operating Test Number: NRC

Administrative Topic (see Note)

Type Code* Describe Activity to be Performed

Conduct of Operations (RA1) N, R

2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (3.9).

JPM: Determine Minimum and Maximum RHR Flow Rate (RO1404).

Conduct of Operations (RA2) M, R

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (4.3)

JPM: Perform a Shutdown Margin Calculation (RO1010D).

Equipment Control (RA3) M, R

2.2.12 Knowledge of surveillance procedures (3.7).

JPM: Perform a Manual Quadrant Power Tilt Ratio Calculation (RO1803D).

Radiation Control (RA4) M, R

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. (3.2).

JPM: Determine Radiation Doses during System Alignment (RWT029D).

Emergency Plan — —

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

*Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom

(D)irect from bank (≤ 3 for ROs; ≤ for 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (≥ 1)

(P)revious 2 exams (≤ 1; randomly selected)

Administrative Topics Outline Task Summary

CPNPP NRC 2014 ES-301-1 RO Admin JPM Outline.docx

RA1 The applicant will determine minimum and maximum Residual Heat Removal

(RHR) System flow per ABN-104, Residual Heat Removal System Malfunction, Attachment 3, Minimum RHR Flow Guideline for Decay Heat Removal, Attachment 4, RHR Maximum Flow Limit, and Attachment 16, Actual Versus Indicated Reactor Vessel Level. Critical steps include calculating the minimum RHR flow rate given time after shutdown and maximum RHR flow rate given recalculated RCS level. This is a new JPM.

RA2 The applicant will perform a Shutdown Margin Calculation per OPT-301,

Shutdown Margin Calculation, Attachment 10.1.1, Manual Generation of OPT-301-9, Shutdown Margin. Critical tasks include identifying individual parameters associated with the calculation and determining the Shutdown Margin. This is a modified bank JPM.

RA3 The applicant will perform a manual Quadrant Power Tilt Ratio calculation per

OPT-302, Calculating Power Tilt Ratio, and determine whether Acceptance Criteria are met. The critical steps include recording data, accurately performing calculations and applying Acceptance Criteria. This is a modified bank JPM.

RA4 The applicant will determine radiation doses during a system alignment per

STA-657, ALARA Job Planning/Debriefing. The critical steps include calculating the dose received in a radiation field with and without protective shielding. This is a modified bank JPM.

Administrative Topics Outline

CPNPP NRC 2014 ES-301-1 SRO Admin JPM Outline.docx


Examination Level SRO Operating Test Number: NRC

Administrative Topic (see Note)

Type Code* Describe Activity to be Performed

Conduct of Operations (SA1) N, R

2.1.1 Knowledge of conduct of operations requirements. (4.2)

JPM: Determine Technical Specification and Event Reportability (SO1005D).

Conduct of Operations (SA2) M, R

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (4.4)

JPM: Perform a Shutdown Margin Calculation (SO1002A).

Equipment Control (SA3) M, R

2.2.12 Knowledge of surveillance procedures (4.1). JPM: Perform a Manual Quadrant Power Tilt Ratio

Calculation and Evaluate Technical Specifications (SO1202C).

Radiation Control (SA4) D, R

2.3.6 Ability to approve release permits. (3.8)

JPM: Review a Gaseous Waste Release Permit (SO1039C).

Emergency Plan (SA5) N, R

2.4.41 Knowledge of the emergency action level thresholds and classifications. (4.6)

JPM: Classify an Emergency Plan Event

(SO1136G).

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

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

Administrative Topics Outline Task Summary

CPNPP NRC 2014 ES-301-1 SRO Admin JPM Outline.docx

SA1 The applicant will identify impacted Technical Specification Limiting Conditions

for Operations and determine Event Reportability per STA-501, Non-Routine Reporting for the Station Electrical System. The critical steps include identifying the Technical Specification Limiting Condition for Operation, Required Action, Completion Time and determining the Oral and Written Reporting Requirements. This is a new JPM.

SA2 The applicant will perform a Shutdown Margin Calculation per OPT-301,

Shutdown Margin Calculation, Attachment 10.1.1, Manual Generation of OPT-301-9, Shutdown Margin. Critical tasks include reviewing individual parameters, verifying adequate Shutdown Margin, and identifying any required action when the Acceptance Criteria is not met. This is a modified bank JPM.

SA3 The applicant will perform a manual Quadrant Power Tilt Ratio calculation per

OPT-302, Calculating Power Tilt Ratio, and determine whether Acceptance Criteria are met. The critical steps include recording data, accurately performing calculations, applying Acceptance Criteria, and identifying any impacted Technical Specification CONDITION, REQUIRED ACTION, and COMPLETION TIME. This is a modified bank JPM.

SA4 The applicant will review a Gaseous Waste Release Permit per STA-603, Control

of Station Radioactive Effluents. The critical steps include identifying any errors and actions required prior to approving the release. This is a bank JPM.

SA5 The applicant will classify an Emergency Plan event per EPP-201, Assessment

of Emergency Action Levels, Emergency Classification, and Plan Activation. Critical steps include determining the Event Category and Event Classification using the Hot, Common, and Cold Emergency Action Level Classification Charts. This is a new JPM.

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

Page 1 of 3 CPNPP NRC 2014 ES-301-2 RO & SRO JPM Outline.docx


Exam Level: RO SRO(I) SRO (U) Operating Test No.: NRC

Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

S-1 001 – Control Rod Drive System (RO1026A) (RO Only)

Respond to Reactor Startup Continuous Rod Insertion

A, L, M, S 1

S-2 004 – Chemical and Volume Control System (RO1309A) Lower Letdown Flow and Adjust Charging Flow

N, S

2

S-3 010 – Pressurizer Pressure Control System (RO1222)

Respond to a Pressurizer Spray Valve Failure

A, D, S

3

S-4 005 – Residual Heat Removal System (RO1507A) Transfer RHR & SI Pumps to Hot Leg Recirculation

A, D, EN, L, S 4P

S-5 061 – Auxiliary Feedwater System (RO3516A) Respond to Inadvertent Start of TDAFW Pump

A, EN, N, S 4S

S-6 064 – Emergency Diesel Generator System (RO4215B) Restore Safeguards Bus 1EA1 to Offsite Power

A, D, S 6

S-7 016 – Non-Nuclear Instrumentation System (RO1833)

Respond to Feedwater Flow Instrument Failure

D, S

7

S-8 086 – Fire Protection System (RO4406C) Respond to a Fire in the Control Room

M, S 8

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

P-1 045 – Main Turbine Generator System (RO4217D)

Align a Main Generator Vent and Purge

D, E 4S

P-2 103 – Containment System (RO2119A) Perform Containment Phase A Local Isolation

D, E, R 5

P-3 071 – Waste Gas Disposal System (RO4006) Terminate Release of Radioactive Gas

N, R 9



@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 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 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank ≤ 9 / ≤ 8 / ≤ 4 (E)mergency or abnormal in-plant ≥ 1 / ≥ 1 / ≥ 1 (EN)gineered safety feature - / - / ≥ 1 (control room system) (L)ow Power / Shutdown ≥ 1 / ≥ 1 / ≥ 1 (N)ew or (M)odified from bank including 1(A) ≥ 2 / ≥ 2 / ≥ 1 (P)revious 2 exams ≤ 3 / ≤ 3 / ≤ 2 (randomly selected) (R)CA ≥ 1 / ≥ 1 / ≥ 1 (S)imulator

NRC JPM Examination Summary Description

S-1 The applicant will perform actions for a continuous rod insertion during a Reactor Startup.

The Control Rods will initially be withdrawn in accordance with IPO-002A, Plant Startup from Hot Standby, Step 5.4.1. The alternate path requires operator action per ABN-712, Rod Control System Malfunction, Section 2.0, Abnormal Control Rod Response in MODE 1 or 2 and/or EOP-0.0A, Reactor Trip or Safety Injection, Step 1. This is a modified bank JPM under Control Rod Drive System – Reactivity Control Safety Function. This is a PRA significant action. (K/A 001.A2.11 - IR 4.4 / 4.7)

S-2 The applicant will lower Letdown flow and adjust Charging flow in accordance with

SOP-103A, Chemical and Volume Control System, Sections 5.2.4, Lowering/Securing Letdown Flow and 5.2.2, Raising/Lowering Charging with a CCP in Operation. This is a new JPM under the Chemical and Volume Control System – Reactor Coolant System Inventory Control Safety Function. (K/A 004.A4.19 - IR 3.1 / 2.8)

S-3 The applicant will perform the actions for a Pressurizer Spray Valve failing open in

accordance with ABN-705, Pressurizer Pressure Malfunction. The alternate path includes actions to trip the Reactor and Reactor Coolant Pump when the Pressurizer Spray Valve fails to close. This is a bank JPM under the Pressurizer Pressure Control System – Reactor Pressure Control Safety Function. This is a PRA significant action. (K/A 010.A2.02 - IR 3.9 / 3.9)

S-4 The applicant will transfer Residual Heat Removal Pumps and Safety Injection Pumps to

Hot Leg Recirculation per EOS-1.4A, Transfer to Hot Leg Recirculation. The alternate path occurs when the Safety Injection to Hot Legs 2 & 3 Injection Isolation Valve fails to open at Step 2d. This is a bank JPM under the Residual Heat Removal System – Primary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A 011.EA1.11 - IR 4.2 / 4.2)



S-5 The applicant will respond to an inadvertent opening of the steam admission valve to the Turbine Driven Auxiliary Feedwater Pump in accordance with ALM-0082A, 1-ALB-8B, Window 4.5 – TD AFWP STM SPLY VLV LEAKING HV-2452-1/2 and then by ABN-305, Auxiliary Feedwater Malfunction, Section 6.0, Inadvertent Turbine Driven AFW Pump Start (Steam Supply Valve Fails Open). The alternate path occurs when the steam supply valve fails to close. This is a new JPM under the Main Feedwater System – Secondary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A 061.A2.07 – IR 3.4 / 3.5)

S-6 The applicant will restore Offsite Power to 6.9 kV Safeguards Bus 1EA1 from

Transformer XST1 in accordance with SOP-609A, Diesel Generator System, Section 5.7, Transferring From DG Supplying Alone to Normal or Alternate Supply. The alternate path occurs when a lowering frequency requires separating the Emergency Diesel Generator from the grid. This is a bank JPM under the Emergency Diesel Generator System – Electrical Safety Function. (K/A 064.A4.07 - IR 3.4 / 3.4)

S-7 The applicant will respond to a Feed Flow Instrument failure in accordance with

ALM-0081A, 1-ALB-8A, Window 2.8 – SG 2 STM FLO & FW FLO MISMATCH and ABN-708, Feed Flow Instrument Malfunction. This is a bank JPM under the Non-Nuclear Instrumentation System – Instrumentation Safety Function. (K/A 059.A2.11 - IR 3.0 / 3.3)

S-8 The applicant will perform the actions for a Control Room fire in accordance with

ABN-803A, Respond to a Fire in the Control Room or Cable Spreading Room, Attachment 1, Reactor Operator Actions to Achieve Hot Shutdown. This is a modified bank JPM under the Fire Protection System – Plant Service Systems Safety Function. (K/A 068.AK3.12 - IR 4.1 / 4.5)

P-1 The applicant will align a Main Generator Vent and Purge per ABN-601, Response to a

138/345 KV System Malfunction, Attachment 11, Main Generator Hydrogen Vent and Argon Purge. This is a bank JPM under the Main Turbine Generator System – Secondary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A G 2.4.35 - IR 3.8 / 4.0)

P-2 The applicant will perform local actions to isolate Containment Phase A Isolation Valves

in accordance with EOP-0.0A, Reactor Trip or Safety Injection, Attachment 4, Phase A Isolation. This is a bank JPM under the Containment System – Containment Integrity Safety Function. (K/A 069.AA1.01 - IR 3.5 / 3.7)

P-3 The applicant will perform actions to terminate a radioactive gas release in accordance

with RWS-201, Gaseous Waste Processing System, Section 5.4.5, Gas Decay Tank X-01 Discharge to the Ventilation System, Step 5.4.5.N. This is a new JPM under the Waste Gas Disposal System – Radioactivity Release Safety Function. (K/A 060.AA2.06 - IR 3.6 / 3.8)

Appendix D Scenario Outline Form ES-D-1

CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Facility: CPNPP 1 & 2 Scenario No.: 1 Op Test No.: June 2014 NRC

Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample).

Turnover: Maintain steady-state power conditions. Tornado Warning from the National Weather Service. Pressurizer Steam Space Sample is in progress by Chemistry.

Critical Tasks: Emergency Stop Train B Diesel Generator within 15 minutes of Breaker Failure in accordance with ECA-0.0A, Loss of All AC Power. (Event 6)

Isolate Reactor Coolant System Leakage Paths in accordance with ECA-0.0A, Loss of All AC Power prior to initiating the Steam Generator depressurization. (Event 7)

Initiate an Operator Induced Cooldown in accordance with ECA-0.0A, Loss of All AC Power. (Event 8)

Event No. Malf. No. Event Type* Event Description

1 +10 min

RP05D I (RO, SRO) TS (SRO)

Reactor Coolant System Loop (1-04) Narrow Range Cold Leg Temperature Instrument (TI-441A) Fails High.

2 +20 min

RP03J I (BOP, SRO) TS (SRO)

Main Steam Line (1-04) Pressure Transmitter (PT-544) Channel I Fails Low.

3 +30 min

RX05A I (RO, SRO) TS (SRO)

Pressurizer Level Channel (LT-459A) Fails Low.

4 +45 min

FW14B TC09I

R (RO) N (BOP, SRO) TS (SRO)

Heater Drain Pump (1-02) Trip. Automatic Turbine Runback Failure.

5 +50 min

ED01

M (RO, BOP, SRO)

Loss of All AC Power Due to Loss of Offsite Power.

6 +50 min

EG06A EG16B

C (BOP) Emergency Diesel Generator (1-01) Air Start Failure. Emergency Diesel Generator (1-02) Output Breaker Failure.

7 +55 min

WDR04 C (RO) Pressurizer Steam Space Sample Valves (1/1-4165A & 1/1-4176A) Fail to Auto Close. Manual Closure of 1-HV-4165A Required.

8 +75 min

MS13A I (BOP)

Atmospheric Relief Valve (1-01) Fails Closed due to Steam Pressure Instrument (PT-2325) Failure.

* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications

Actual Target Quantitative Attributes

9 Total malfunctions (5-8)

4 Malfunctions after EOP entry (1-2)

4 Abnormal events (2-4)

1 Major transients (1-2)

1 EOPs entered/requiring substantive actions (1-2)

1 EOP contingencies requiring substantive actions (0-2)

3 Critical tasks (2-3)



SCENARIO SUMMARY NRC 1 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. A Tornado Warning from the National Weather Service is in progress. ABN-907, Acts of Nature, Section 5.0, Severe Weather, has been performed and includes placing the High Flux at Shutdown switches in BLOCK and realigning Control Room Ventilation to the Emergency Recirculation Mode. A Pressurizer Steam Space sample is in progress. In the first event a Reactor Coolant System (RCS) Loop 4 TCOLD Instrument will fail high. The crew enters ABN-704, TC/N-16 Instrumentation Malfunction, Section 2.0, places Rod Control in MANUAL and defeats the affected channel. The SRO will refer to Technical Specifications. The next event is a Main Steam Line 1-04 Channel I Pressure Transmitter failing low. Operator actions are per ABN-709, Steam Line Pressure Instrument Malfunction, Section 2.0, and require taking MANUAL control of the Feedwater Flow Control Valve, transferring to an Alternate Channel, and restoring Feedwater Flow Control to AUTO. The SRO will refer to Technical Specifications. When plant conditions are stable a low failure of Pressurizer Level Channel, LT-459A, will occur. Operator actions are per ABN-706, Pressurizer Level Instrument Malfunction, Section 2.0. The crew must manually control either the Charging Flow Controller or the Pressurizer Level Controller, transfer to an Alternate Channel, and restore Reactor Coolant System (RCS) Letdown flow. The SRO will refer to Technical Specifications. The next event is a trip of a Heater Drain Pump with an automatic Turbine Runback failure. The crew responds per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 4.0. When it is determined that automatic plant response has not activated, Control Rods are placed in AUTO and a manual Turbine Runback will be initiated. The crew will stabilize load at 700 MWe. During this event, Control Rod position may drop below the Rod Insertion Limit (RIL) and when informed, the SRO will refer to Technical Specifications. Additionally, ABN-401, Main Turbine Malfunction, Section 8.0, must be entered to reset the Turbine Runback. The major event is a Loss of Offsite Power with an air start failure of Train A Diesel Generator and a breaker failure of Train B Diesel Generator resulting in a Total Loss of All AC Power. The crew enters either EOP-0.0A, Reactor Trip or Safety Injection and then exits to ECA-0.0A, Loss of All AC Power, or enters ECA-0.0A directly. While in ECA-0.0A, Reactor Coolant System leakage paths are isolated and a cooldown is initiated to 270°F. When the cooldown is commenced the crew will restore power to the Train B Safeguards Bus. The event is complicated by failure of a Pressurizer Steam Space Sample Valve to automatically close and an Atmospheric Relief Valve that must be manually opened in order to facilitate Reactor Coolant System cooldown. This scenario is terminated when power is restored to the Train B Safeguards Bus and transition to ECA-0.0A, Step 26, Stabilize Steam Generator Pressures is performed. Risk Significance:

Failure of risk important system prior to trip: Turbine Runback Failure

Risk significant core damage sequence: Loss of All AC Power

Risk significant operator actions: Stop Train B Diesel Generator

Isolate RCS Leakage Paths

Initiate RCS Cooldown

Restore Safeguards Bus Power






Turnover: Maintain steady-state power conditions.

Critical Tasks: Trip Reactor Coolant Pumps within 10 minutes upon a Loss of Subcooling per EOP-0.0A, Reactor Trip or Safety Injection or EOP-1.0A, Loss of Reactor or Secondary Coolant, Foldout Pages. (Event 5)

Initiate Cooldown of Reactor Coolant System Prior to Exiting EOS-1.2A, Post LOCA Cooldown and Depressurization. (Event 6)

Manually Initiate Train A and Train B Safety Injection Signal Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 7)


1 +5 min

RX12 C (BOP, SRO) Main Steam Header Pressure Transmitter (PT-507) Fails Low.

2 +15 min

RX08A

I (RO, SRO) TS (SRO)

Pressurizer Pressure Channel (PT-455) Fails High.

3 +25 min

RX04C I (BOP, SRO) TS (SRO)

Steam Generator (1-03) Level Transmitter (LT-553) Fails High.

4 +40 min

AFP 13_89

N (RO, SRO) TS (SRO)

Fire in Auxiliary Building Fire Area AC. Centrifugal Charging Pump (1-02) Manual Start Required.

5 +45 min

RX16A RX16B

C (RO) Power Operated Relief Valves (PCV-455A/456) Fail Open. Reactor Trip Required.

6 +50 min

RCR23 M (RO, BOP, SRO)

Power Operated Relief Valve Block Valve (1/1-8000A) Fails Open upon Breaker Trip.

7 +55 min

RP07A RP07B

C (RO)

Train A Safety Injection Fails to Automatically Actuate. Train B Safety Injection Fails to Automatically Actuate.

8 +55 min

RH01D C (BOP) Residual Heat Removal Pump (1-02) Safety Injection Sequencer Start Failure.

* (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications











SCENARIO SUMMARY NRC 2 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. The first event is a low failure of Main Steam Header Pressure Transmitter (PT-507). The crew enters ABN-709, Steam Header Pressure Instrument Malfunction, Section 3.0, and places the Main Feedwater Pump Master Speed Control in MANUAL and restore required Steam Generator levels. The controller will remain in MANUAL and may require adjustment when the Steam Generator Level Transmitter fails later in the scenario. The next event is a high failure of a Pressurizer Pressure Channel. Operator actions are per ABN-705, Pressurizer Pressure Malfunction, Section 2.0, and require closing the Power Operated Relief Valve (PORV) and its associated Block Valve, placing the Pressurizer Master Pressure Controller in MANUAL, selecting an alternate controlling Channel, and restoring Pressurizer pressure to normal. The SRO will refer to Technical Specifications. When plant conditions are stable a high failure of a Steam Generator (1-03) Level Transmitter will occur. Crew actions are per ABN-710, Steam Generator Level Instrumentation Malfunction, Section 2.0, and include placing Steam Generator (SG) Level Control in MANUAL, stabilizing the plant, aligning an Alternate Channel, and transferring SG Level Control back to AUTO. The SRO will refer to Technical Specifications. When plant parameters are restored to normal, a fire alarm in Auxiliary Building Fire Area AC will be initiated. The crew enters ABN-805A, Response to Fire in the Auxiliary Building or the Fuel Building, Section 3.0. Actions include placing the Train B Centrifugal Charging Pump in service. The fire will continue to spread in the Auxiliary Building resulting in the inadvertent opening of both Power Operated Relief Valves (PORVs) as addressed in ABN-805A, Section 7.0, and will require a manual Reactor Trip. The crew will enter and perform actions of EOP-0.0A, Reactor Trip or Safety Injection. The Power Operated Relief Valves will fail to close and one PORV Block Valve breaker will trip before the valve closes resulting in a Small Break Loss of Coolant Accident. The crew will then transition from EOP-0.0A to EOP-1.0A, Loss of Reactor or Secondary Coolant, in preparation for an eventual cooldown and depressurization. The scenario includes a failure of both trains of Safety Injection to automatically actuate along with a Train B Residual Heat Removal Pump that fails to start upon initiation of the Safety Injection Sequencer. This scenario is terminated when a cooldown is commenced via the Steam Dump Valves in EOS-1.2A, Post LOCA Cooldown and Depressurization. Risk Significance:

Failure of risk important system prior to trip: Train A Centrifugal Charging Pump

Risk significant core damage sequence: Small Break LOCA

Risk significant operator actions: Manually Initiate Safety Injection

Trip Reactor Coolant Pumps

Manually Start Train B RHR Pump

Initiate RCS Cooldown






Turnover: Maintain steady-state power conditions.

Critical Tasks: Manually Trip Reactor Due to Reactor Protection System Failure Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 5)

Initiate Train A and/or Train B Containment Isolation Phase A due to Failure to Automatically Actuate Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 8)

Identify and Isolate Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture. (Event 6)

Initiate Cooldown of Reactor Coolant System Prior to Exiting EOP-3.0A, Steam Generator Tube Rupture. (Event 6)


1 +10 min

SG01A C (SRO) Steam Generator (1-01) Tube Leak at 50 gpd (0.0347 gpm).

2 +20 min

RX09A I (RO, BOP, SRO)TS (SRO)

Main Turbine 1st Stage Pressure Transmitter (PT-505) Fails High.

3 +30 min

SW01B C (BOP, SRO) TS (SRO)

Station Service Water Pump 1-02 Trip.

4 +50 min

FW16 R (RO) C (BOP, SRO)

Lowering Condenser Vacuum Requiring Power Reduction Followed by Total Loss of Condenser Vacuum.

5 +55 min

RP01 RP13A

I (BOP) Automatic Reactor Trip Failure. Manual Reactor Trip Failure from CB-07.

6 +65 min

SG01A M (RO, BOP, SRO) Steam Generator (1-01) Tube Rupture at 500 gpm (600 second ramp) upon Turbine Trip.

7 +70 min

OVRDE C (RO/BOP) Steam Generator (1-01) Blowdown Isolation Valve (1-HS-2397) Fails to Close on Safety Injection.

8 +70 min

RP09A RP09B

C (BOP)

Containment Isolation Phase A Train A and Train B Auto Actuation Failure.

*(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications











SCENARIO SUMMARY NRC 3 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. The first event is a Steam Generator #1 tube leak of 50 gpd. Crew actions are per ABN-106, High Secondary Activity, Section 2.0, and include verification of leakage rate to ensure proper procedural action. The SRO will refer to Technical Specifications. The next event is a Main Turbine 1st Stage Pressure Transmitter failure. The crew responds per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure and Feed Header Pressure Instrument Malfunction, Section 4.0. Several actions are required on the part of the RO and BOP to stabilize plant conditions. The SRO will refer to Technical Specifications. When conditions are stable, Station Service Water Pump 1-02 will trip. The crew will enter ABN-501, Station Service Water System Malfunction, Section 2.0. Initial operator actions include placing the Train B Emergency Diesel Generator in PULLOUT. The SRO will refer to Technical Specifications. When Pressurizer conditions are stable, a loss of Condenser vacuum will occur. The crew will respond per ABN-304, Main Condenser and Circulating Water System Malfunction, Section 3.0. Actions include lowering of Main Turbine load in 50 MWe increments in an attempt to maintain Condenser vacuum. Once a load reduction is performed, Condenser vacuum will continue to deteriorate until a Reactor Trip is required. Once it is determined that Condenser vacuum cannot be maintained, the Reactor must be manually tripped and EOP-0.0A, Reactor Trip or Safety Injection, entered. The automatic Reactor Trip function and one manual Reactor Trip switch are disabled to ensure appropriate crew actions and communication. The crew will transition from EOP-0.0A to EOS-0.1A, Reactor Trip Response and then return to EOP-0.0 when Foldout Page criteria for Pressurizer level or Reactor Coolant System (RCS) subcooling are not met. A Steam Generator Tube Rupture is diagnosed in EOP-0.0A and a transition to EOP-3.0A, Steam Generator Tube Rupture will be performed. The scenario is complicated by a Train A and B Containment Isolation Phase A failure and a Steam Generator Blowdown Valve that fails to automatically isolate. While in EOP-3.0A, the ruptured Steam Generator will be isolated and a RCS cooldown commenced using the Atmospheric Relief Valves. The scenario is terminated when steps to cooldown the Reactor Coolant System are reached in EOP-3.0A.

Risk Significance:

Failure of risk important system prior to trip: Steam Generator Tube Leak

Train B Emergency Diesel Generator

Risk significant core damage sequence: Automatic Reactor Trip Failure

Steam Generator Tube Rupture

Risk significant operator actions: Manually Initiate Reactor Trip

Isolate SG Blowdown Flow

Initiate Containment Phase A Isolation

Identify and Isolate Ruptured SG

ES-401 PWR Examination Outline FORM ES-401-2 · 28,49 003 Reactor Coolant Pump 04 0 2 Effects of...

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