Appendix D Scenario Outline Form ES-D-1 · 2012. 8. 31. · Scenario Event Description NRC Scenario...

Appendix D Scenario Outline Form ES-D-1

CPNPP NRC 2012 Sim Scenario #1 Rev e.docx

Facility: CPNPP 1 & 2 Scenario No.: 1 Op Test No.: June 2012 NRC

Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 908 ppm (by sample).

Turnover: Maintain steady-state power conditions.

Critical Tasks: Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-708, Feedwater Flow Instrument Malfunction.

Reduce Reactor Power to Less Than 100% Prior to Exiting ABN-304, Main Condenser and Circulating Water System Malfunction.

Initiate Emergency Boration Due to Loss of Digital Rod Position Indication Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

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 No. Malf. No. Event Type* Event Description

1 +10 min

RX01E

I (BOP, SRO) Steam Generator (1-03) Feed Flow Instrument (FT-530) Fails High.

2 +25 min

NI05E I (RO, SRO) TS (SRO)

Power Range Nuclear Instrument Channel (N-43) Fails High.

3 +35 min

CW02A R (RO) N (BOP, SRO)

Circulating Water Pump (1-01) Trip.

4 +45 min

RX09A I (RO, SRO) TS (SRO)

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

5 +55 min

RC19C M (RO, BOP, SRO) Small Break Loss of Coolant Accident Inside Containment at 1750 GPM on 300 second ramp.

6 +58 min

RD12C I (RO) Loss of Digital Rod Position Indication.

7 +58 min

CC02D C (BOP)

Component Cooling Water Pump (1-02) SI Sequencer Start Failure.

8 +75 min

SW01A C (BOP) Station Service Water Pump (1-01) Trip Upon Containment Isolation Phase A RESET.

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

Actual Target Quantitative Attributes 8 Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 4 Critical tasks (2-3)

Scenario Event Description NRC Scenario #1


SCENARIO SUMMARY NRC #1 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. The first event it is a high failure of Steam Generator (1-03) Feed Flow Instrument, FT-530. Operator actions are per ABN-708, Feedwater Flow Instrument Malfunction, Section 2.0. The crew must manually control Steam Generator level, transfer to an Alternate Channel, and restore Steam Generator (SG) Feedwater Flow Control to AUTO. The next event is high failure of Power Range Nuclear Instrument Channel (N-43). Crew actions are per ABN-703, Power Range Instrumentation Malfunction, Section 2.0, and include placing Rod Control in MANUAL, bypassing the failed Channel, restoring TAVE to program, and verifying Permissive Interlocks match current power level. The SRO will refer to Technical Specifications. When Technical Specifications have been reference, Circulating Water Pump (1-01) will trip. The crew enters ABN-304, Main Condenser and Circulating Water System Malfunction, Section 2.0, Circulating Water Pump Trip, and initiates a 50 MWe Turbine Load reduction when Reactor Power rises above 100%. Rod Control is returned to AUTO during this evolution. When plant conditions are stable, Main Turbine First Stage Pressure Transmitter (PT-505) will fail low. The crew enters ABN-709, Turbine 1st Stage Pressure Instrument Malfunction, Section 4.0, places Rod Control in MANUAL, disables the Steam Dump System, and transfers to Turbine Impulse Pressure Channel PT-506. The SRO will refer to Technical Specifications. When Technical Specifications have been referenced, a Small Break Loss of Coolant Accident will commence on a 300 second ramp inside Containment. The crew will observe lowering Pressurizer pressure and level and manually initiate a Reactor Trip and Safety Injection. EOP-0.0A, Reactor Trip or Safety Injection, is entered and actions implemented until a loss of primary coolant is diagnosed and then a transition into EOP-1.0A, Loss of Primary or Secondary Coolant, is made. An Emergency Boration due to loss of Digital Rod Position Indication must be initiated after Step 4 of EOP-0.0A. The scenario includes a manual Train B Component Cooling Water (CCW) Pump start following failure of the Safety Injection Sequencer and a Service Water Pump 1-01 trip that requires stopping the Train A Emergency Diesel Generator (EDG) when Containment Isolation Phase A is RESET while in EOP-1.0A. Additionally, the Reactor Coolant Pumps (RCPs) must be tripped when a loss of subcooling occurs in either EOP-0.0A or EOP-1.0A. This scenario is terminated when a Reactor Coolant System cooldown is commenced per EOS-1.2A, Post LOCA Cooldown and Depressurization. Risk Significance:

Failure of risk important system prior to trip: SG Feed Flow Instrument Failure

Risk significant core damage sequence: Small Break LOCA

Risk significant operator actions: Restore SG Feedwater Flow Control

Manually Initiate Emergency Boration

Manually Start Train B CCW Pump

Trip RCPs Due to Loss of Subcooling



BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP

Initialize to IC #18 and NRC #1 Scenario File.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE

INITIATING PARAMETER

SETUP RD12C Loss of Digital Rod Position Indication OFF Reactor Trip

CC02D CCW Pump (1-02) SI Sequencer Start Failure - SI Sequencer

OVRDE CCW Pump (1-02) Override OFF {DICCHS4519} STP K0

COND Delete CCW Pump (1-02) Override {DICCHS4519A}

DOR CCW to START

1 RX01E SG (1-03) Feed Flow Instrument (FT-530) Failure 5E6 lbm/hr K1

2 NI05E PR Nuclear Instrument Channel (N-43) Failure 200% K2

3 CW02A Circulating Water Pump (1-01) Trip OFF K3

4 RX09A Main Turbine 1st Stage Pressure (PT-505) Failure 0 PSIG K4

5 RC19C Small Break LOCA Inside Containment 1750 GPM K5 (300 sec. ramp)

6 RD12C Loss of Digital Rod Position Indication OFF Reactor Trip

7 CC02D CCW Pump (1-02) SI Sequencer Start Failure - SI Sequencer

7 OVRDE CCW Pump (1-02) Override OFF {DICCHS4519} STP K0

7 COND Delete CCW Pump (1-02) Override {DICCHS4519A}

DOR CCW to START

8 COND SW01A Station Service Water Pump (1-01) Trip TRIP Phase A RESET



Booth Operator: INITIALIZE to IC #18 and NRC #1 Scenario File. ENSURE all Simulator Annunciator Alarms are ACTIVE. ENSURE all Control Board Tags are removed. ENSURE Operator Aid Tags reflect current boron conditions. ENSURE Rod Bank Update (RBU) is performed. ENSURE Turbine Load Rate set at 10 MWe/minute. ENSURE 60/90 buttons DEPRESSED on ASD. ENSURE ASD speakers are ON at half volume. ENSURE Reactivity Briefing Sheet printout provided with Turnover. ENSURE procedures in progress are on SRO desk: - COPY of IPO-003A, Power Operations, Section 5.5, Operating at Constant Turbine Load. ENSURE Control Rods are in AUTO with Bank D at 215 steps.

Control Room Annunciators in Alarm: PCIP-1.1 – SR TRN A RX TRIP BLK PCIP-1.2 – IR TRN A RX TRIP BLK PCIP-1.4 – CNDSR AVAIL STM DMP ARMED C-9 PCIP-1.6 – RX ≥ 10% PWR P-10 PCIP-2.1 – SR TRN B RX TRIP BLK PCIP-2.2 – IR TRN B RX TRIP BLK PCIP-2.5 – SR RX TRIP BLK PERM P-6 PCIP-3.2 – PR TRN A LO SETPT RX TRIP BLK PCIP-4.2 – PR TRN B LO SETPT RX TRIP BLK

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 1 Page 6 of 36 Event Description: Steam Generator Feed Flow Transmitter Failure

Time Position Applicant’s Actions or Behavior


Booth Operator: When directed, EXECUTE Event 1 (Key 1). - RX01E, SG 1-03 Feed Flow Transmitter (FT-530) fails high.

Indications Available: 8A-3.8 – SG 3 STM & FW FLO MISMATCH 1-FI-530A, SG 3 FW FLO indication fails high

+30 sec BOP REFER to Annunciator Alarm Procedures.

BOP RECOGNIZE SG 1-03 Feed Flow Transmitter (FT-530) failed high.

Examiner Note: A Feedwater Flow Channel failing HIGH will cause feedwater flow to lower by closing the associated Feedwater Control Valve. Without operator action, the Reactor will trip at 38% Steam Generator level.

US DIRECT implementation of ABN-708, Feedwater Flow Instrument Malfunction, Section 2.0.

BOP DETERMINE controlling feed flow channel has failed. [Step 2.3.1 - YES]

CRITICAL TASK

STATEMENT Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-708, Feedwater Flow Instrument Malfunction.

CRITICAL

TASK BOP PLACE 1-FK-530, SG 3 FW FLO CTRL in MANUAL and CONTROL level. [Step 2.3.2 - YES]

BOP VERIFY automatic SG 1-03 level control – DESIRED: [Step 2.3.3 - YES]

VERIFY alternate SG 1-03 level control channel responding normally. [Step 2.3.3.a - YES]

VERIFY automatic SG 1-03 level control is desired. [Step 2.3.3.b - YES]

BOP SELECT an alternate channel. [Step 2.3.4 - YES]

PLACE 1-FS-530C, SG 3 FW FLO CHAN SELECT to FY-531B. [Step 2.3.4 - YES]

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 1 Page 7 of 36 Event Description: Steam Generator Feed Flow Transmitter Failure



BOP VERIFY SG 1-03 ready for AUTO Level Control: [Step 2.3.5 - YES]

VERIFY Feedwater and Steam Flows matched. [Step 2.3.5.a - YES]

VERIFY Steam Generator level stable at program. [Step 2.3.5.b - YES]

BOP PLACE 1-FK-530, SG 3 FW FLO CTRL in AUTO and MONITOR operation. [Step 2.3.6 - YES]

US INITIATE a Condition Report per STA-421. [Step 2.3.7 - YES]

+10 min US INITIATE repairs per STA-606. [Step 2.3.8 - YES]

When Feedwater Control is restored, or at Lead Evaluator’s discretion, PROCEED to Event 2.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 2 Page 8 of 36 Event Description: Power Range Nuclear Instrument Failure



Booth Operator: When directed, EXECUTE Event 2 (Key 2). - NI05E, Power Range Nuclear Instrument (N-43) detector fails high.

Indications Available: 6D-1.3 – 1 OF 4 HI SETPT PR FLUX HI 6D-3.3 – 1 OF 4 PR FLUX RATE HI 6D-3.4 – PR CHAN DEV HI 6D-2.14 – OP HI FLUX ROD STOP C-2 1-NI-43B, PR POWER CHAN III indication fails high

+30 sec RO REFER to Annunciator Alarm Procedures.

RO RECOGNIZE Power Range Nuclear Instrument N-43 detector failure.

Examiner Note: If a Power Range Channel fails HIGH while the Rod Control System is an AUTO, Control Rods will be rapidly inserted.

US DIRECT implementation of ABN-703, Power Range Instrumentation Malfunction, Section 2.0.

RO VERIFY rapid Control Rod insertion – NOT REQUIRED. [Step 2.3.1 - YES]

VERIFY Reactor and Turbine Power – MATCHED. [Step 2.3.1.a - YES]

VERIFY TAVE less than 3°F above TREF. [Step 2.3.1.a - YES]

PLACE Rod Control in MANUAL. [Step 2.3.1.b - YES]

RO VERIFY Reactor Power < 75% rated thermal power. [Step 2.3.2 - NO]

US INITIATE actions to comply with Technical Specification SR 3.2.4.2. [Step 2.3.2 RNO - YES]

PERFORM at Channel N-43 Drawers: [Step 2.3.3 - YES]

RO/BOP At Detector Current Comparator Drawer, SELECT Rod Stop Bypass Switch to N-43. [Step 2.3.3.a - YES]

RO/BOP At Comparator and Rate Drawer, SELECT Comparator Channel Defeat Switch to N-43. [Step 2.3.3.b - YES]




RO/BOP At Detector Current Comparator Drawer, SELECT Upper Section Switch to N-43. [Step 2.3.3.c - YES]

RO/BOP At Detector Current Comparator Drawer, SELECT Lower Section Switch to N-43. [Step 2.3.3.d - YES]

RO/BOP At Detector Current Comparator Drawer, SELECT Power Mismatch Bypass Switch to N-43. [Step 2.3.3.e - YES]

RO/BOP At the Power Range A Drawer, SELECT Rate Mode Switch momentarily to RESET for N-43. [Step 2.3.3.f - YES]

RO/BOP PLACE 1/1-JS-411E, N16 PWR CHAN DEFEAT Switch to LOOP 3. [Step 2.3.3.g - YES]

RO/BOP PLACE 1/1-TS-412T, TAVE CHAN DEFEAT Switch to LOOP 3. [Step 2.3.3.g - YES]

RO/BOP PLACE 1/1-TS-411E, 1-TR-411 Channel Select to an OPERABLE channel. [Step 2.3.4 - YES]

RO RESTORE TAVE to within 1ºF of TREF. [Step 2.3.5 - YES]

US/RO Within one (1) hour, VERIFY Permissives P-7, P-8, P-9, and P-10 MATCH current power level: [Step 2.3.6 - YES]

RX & TURB ≤ 10% PWR P-7 (PCIP – 3.5) – DARK. [Step 2.3.6.a - YES]




RX ≤ 48% PWR 3-LOOP FLO PERM P-8 (PCIP – 4.5) – DARK. [Step 2.3.6.a - YES]

RX ≤ 50% PWR TURB TRIP PERM P-9 (PCIP – 1.7) – DARK. [Step 2.3.6.a - YES]

RX ≥ 10% PWR P-10 (PCIP – 1.6) – LIT. [Step 2.3.6.a - YES]

US/RO RECORD verification in Unit Log. [Step 2.3.6.b - NO]

US/RO CHECK Quadrant Power Tilt Ratio within limits: [Step 2.3.7 - YES]

CHECK Power Range Channels– ONE OR MORE INOPERABLE [Step 2.3.7.a - YES]

CHECK Reactor Power – GREATER THAN 50%. [Step 2.3.7.b - YES]

REFER to TS 3.2.4, Table 3.3.1-1, Items 2, 3 (ACTIONS D and E) and TR 13.2.33. [Step 2.3.7.c - YES]

Examiner Note: The next two (2) steps are only performed following I&C maintenance.

VERIFY I&C locates problem area AND completes appropriate repair within 72 hours of instrument malfunction. [Step 2.3.8 - NO]

RECORD data in Unit Log. [Step 2.3.8.a - NO]

VERIFY interlocks MATCH current power level: [Step 2.3.9 - DONE]

RX & TURB ≤ 10% PWR P-7 (PCIP – 3.5) – DARK

RX ≤ 48% PWR 3-LOOP FLO PERM P-8 (PCIP – 4.5) – DARK

RX ≤ 50% PWR TURB TRIP PERM P-9 (PCIP – 1.7) – DARK

RX ≥ 10% PWR P-10 (PCIP – 1.6) – LIT




US EVALUATE Technical Specifications. [Step 2.3.10 - YES]

LCO 3.3.1.D, Reactor Trip System Instrumentation.

CONDITION D - One Power Range Neutron Flux-High channel inoperable.

ACTION D.1.1 - Perform SR 3.2.4.2 within 12 hours from discovery of THERMAL POWER> 75% RTP, AND

ACTION D.1.2 - Place channel in trip within 72 hours, OR ACTION D.2 - Be in MODE 3 within 78 hours.

LCO 3.3.1.E, Reactor Trip System Instrumentation.

CONDITION E - One channel inoperable. ACTION E.1 - Place channel in trip within 72 hours, OR ACTION E.2 - Be in MODE 3 within 78 hours.

LCO 3.3.1.S, Reactor Trip System Instrumentation.

CONDITION S - One or more required channel(s) inoperable. ACTION S.1 - Verify interlock is in the required state for existing unit

conditions within 1 hour, OR ACTION S.2 - Be in MODE 3 within 7 hours.

LCO 3.3.1.T, Reactor Trip System Instrumentation.

CONDITION T - One or more required channel(s) inoperable. ACTION T.1 - Verify interlock is in the required state for existing unit

conditions within 1 hour. ACTION T.2 - Be in MODE 2 within 7 hours.



When the Technical Specification actions are addressed, or at Lead Evaluator’s discretion, PROCEED to Event 3.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 3 Page 12 of 36 Event Description: Circulating Water Pump Trip



Booth Operator: When directed, EXECUTE Event 3 (Key 3). - CW02A, Circulating Water Pump 1-01 trip.

Indications Available: 9A-11.8 – ANY CWP OVRLOAD / TRIP 1-II-2800, CWP 1 motor current at 0 amps 1-PI-2804, CWP 1 DISCH PRESS low

+10 sec BOP REFER to Annunciator Alarm Procedures.

BOP DETERMINE Circulating Water Pump 1-01 tripped.

US DIRECT performance of ABN-304, Main Condenser and Circulating Water System Malfunction, Section 2.0.

BOP VERIFY Circulating Water Pumps 1-02, 1-03, and 1-04 – OPERATING. [Step 2.3.1 - YES]

RO/BOP VERIFY Reactor Power less than or equal to 100% [Step 2.3.2 - NO]

CRITICAL TASK

STATEMENT Reduce Reactor Power to Less Than 100% Prior to Exiting ABN-304, Main Condenser and Circulating Water System Malfunction.

CRITICAL

TASK BOP INITIATE a 50 MW Turbine Load reduction. [Step 2.3.2 RNO - YES]




Examiner Note: The following steps are considered skill of the craft.

US DIRECT manual INITIATION of 50 MWe Turbine Runback.

RO PLACE 1/1-RBSS, Control Rod Bank Select Switch in AUTO.

BOP PERFORM the following to INITIATE 50 MWe Turbine Runback:

DEPRESS “50 MWe” Manual Runback button.

CLICK on “0/1” button.

CLICK on “EXECUTE” then VERIFY Runback in progress.

Examiner Note: The crew may initially execute a 50 MWe power reduction (Runback) for a loss of secondary efficiency. Rod Control will be placed in AUTO and boration may or may not be performed.

US/BOP On the TG Control Display, VERIFY Main Condenser Vacuum maintained – GREATER THAN 26.5″ HG. [Step 2.3.3 - NO]

On the Back Pressure Limit Display ENSURE Turbine NOT operating for more than 5 minutes in the NOT PERMISSIBLE region by lowering Turbine Load. [Step 2.3.3 RNO - YES]

US/BOP VERIFY Main Condenser Vacuum maintained – GREATER THAN 24.5″ HG AND Reactor Power LESS THAN 100%. [Step 2.3.4 - YES]

BOP ENSURE at least two Circulating Water Pumps – OPERATING per SOP-310A. [Step 2.3.5 - YES]

BOP VERIFY PCIP Window 1.4 – CNDSR AVAIL STM DMP ARMED C-9 – LIT. [Step 2.3.6 - YES]




BOP VERIFY at least three Circulating Water Pumps – OPERATING. [Step 2.3.7 - YES]

US/BOP VERIFY Auxiliary Condenser Vacuum – GREATER THAN 21″ HG AND STABLE. [Step 2.3.8 - YES]

BOP CHECK Chemistry Sampling Requirements: [Step 2.3.9 - YES]

VERIFY Steam Generator Atmospheric Relief Valves – REMAINED CLOSED AND Turbine Driven Auxiliary Feedwater Pump – REMAINED STOPPED. [Step 2.3.9.a - YES]

VERIFY Reactor Power Change – < 15% WITHIN ONE HOUR. [Step 2.3.9.b - YES]

Examiner Note: SOP-310A is addressed but NOT performed due to extensive field actions.

BOP

ENSURE Main Condenser Outlet Valves are adjusted as required to maintain CWP discharge pressure within limits of SOP-310A and CWP running currents of 150-160 amps. [Step 2.3.10 - YES]

US INITIATE repairs per STA-606. [Step 2.3.11 - YES]

+10 min US INITIATE a Condition Report per STA-421. [Step 2.3.12 - YES]

When the Turbine Runback is complete and power is stabilized, or at Lead Examiner discretion, PROCEED to Event 4.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 4 Page 15 of 36 Event Description: Main Turbine 1st Stage Pressure Transmitter Failure



Booth Operator: When directed, EXECUTE Event 4 (Key 4). - RX09A, Main Turbine 1st Stage Pressure Transmitter (PT-505) fails low.

Indications Available: 6D-1.10 – AVE TAVE TREF DEV PCIP-2.4 – LO TURB PWR ROD WITHDRW BLK C-5 1-PI-505 – Turbine Impulse Pressure Channel I indication fails low 1-TI-412A – Ave TAVE TREF Deviation indication to maximum

+30 secs RO/BOP RESPOND to Annunciator Alarm Procedures.

RO/BOP RECOGNIZE Control Rods INSERTING due to Turbine Impulse Pressure Instrument failure.

RO/BOP REPORT PT-505, Turbine Impulse Pressure Channel I has failed low.

US

DIRECT implementation of ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 4.0.

RO PLACE 1/1-RBSS Control Rod Bank Select Switch in MANUAL. [Step 4.3.1 - YES]

BOP VERIFY Steam Dumps - CLOSED WITH NO OPEN DEMAND. [Step 4.3.2 - NO]

If Steam Dump operation NOT required, PLACE at least one Steam Dump Interlock Select Switch – OFF. [Step 4.3.2 RNO - YES]

43/1-SDA, STM DMP INTRLK SELECT.

43/1-SDB, STM DMP INTRLK SELECT.




BOP RESTORE Steam Dump availability, PLACE Steam Dumps in STEAM PRESSURE Mode per Attachment 7. [Step 4.3.3 - YES]

Examiner Note: The following five (5) steps are from ABN-709, Attachment 7, Transferring Steam Dumps and are performed using the Control Board Job Aid.

BOP ENSURE 1-PK-507, STM DMP PRESS CTRL is in MANUAL. [Step 1 - YES]

BOP MATCH 1-PK-507, STM DUMP PRESS CTRL demand to current Steam Dump Valve position. [Step 2 - YES]

BOP VERIFY 1-PCIP, Window 1.4 – CNDSR AVAIL STM DMP ARMED C-9 is ON. [Step 3 - YES]

BOP PLACE 43/1-SD, STM DMP MODE SELECT in STM PRESS. [Step 4 - YES]

BOP ENSURE both STM DMP INTLK SELECT switches are ON. [Step 5 - YES]

BOP If desired to control Steam Dumps in AUTO, then PERFORM the following: [Step 6 - YES]

VERIFY 1-PI-507, MS HDR PRESS indicates current MSL pressure [Step 6.A - YES]

ENSURE 1-PK-507, STM DMP PRESS CTRL set to control at 1092 psig for "no load" conditions (Pot setting 6.86). [Step 6.B - YES]

PLACE 1-PK-507, STM DMP PRESS CTRL in AUTO [Step 6.C - YES]

RO TRANSFER 1-PS-505Z, TURB IMP PRESS CHAN SELECT to PT-506. [Step 4.3.4 - YES]




Examiner Note: When PT-506 is selected, the major event should be commenced. The Technical Specification call associated with the failed 1st Stage Pressure Transmitter can be performed when the scenario is completed. This will prevent delays associated with restoring TAVE to within 1ºF of TREF.

RO ENSURE TAVE within 1ºF of TREF. [Step 4.3.5 - NO]

RO PLACE 1/1-RBSS, CONTROL ROD BANK SELECT Switch in AUTO. [Step 4.3.6 - YES]

US/RO CHECK Reactor Power in – MODE 1. [Step 4.3.7 - YES]

US/BOP CHECK Turbine Power – GREATER THEN 10% POWER. [Step 4.3.8 - YES]

US Within 1 hour, VERIFY PCIP Window 4.6 – TURB ≤ 10% PWR P-13, IN PROPER STATE for existing plant conditions (DARK). [Step 4.3.9 - YES]


LCO 3.3.1.T, Reactor Trip System Instrumentation (Function 18.f).

CONDITION T - One or more required channels inoperable. ACTION T.1 - Verify interlock is in required state for existing unit

conditions within one (1) hour, OR ACTION T.2 - Be in MODE 2 within 7 hours.

US VERIFY PCIP Window 1.3 – AMSAC BLK TURB < 40% PWR C-20 – IN PROPER STATE (DARK) for actual Turbine power. [Step 4.3.10 - YES]

If AMSAC actuation blocked and Turbine power >40%, ENSURE Automatic Actions of ALB-9B, Window 3.7 – AMSAC ACT TURB TRIP as necessary.






When the PT-506 is placed in service, or at Lead Evaluator’s discretion, PROCEED to Events 5, 6, 7, and 8.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 1 Event # 5, 6, 7, & 8 Page 19 of 36 Event Description: Small Break Loss of Coolant Accident / Loss of Digital Rod Position Indication / Component Cooling

Water Pump Start Failure / Station Service Water Trip Time Position Applicant’s Actions or Behavior


Booth Operator: When directed, EXECUTE Events 5, 6, 7, (Key 5) and 8. - RC19C, Small Break LOCA at 1750 GPM on 300 second ramp. - RD12C, Loss of Digital Rod Position Indication upon Reactor trip. - CC02D, CCW Pump (1-02) fails to start on SI Sequencer. - SW01A, Service Water Pump (1-01) trip upon Containment Isolation Phase A RESET.

Indications Available: 2A-2.8 – ANY CNTMT SMP PMP RUN 2B-4.12 – CNTMT FN CLR 1 & 2 CNDS FILL RATE HI 2B-3.12 – CNTMT FN CLR 3 & 4 CNDS FILL RATE HI 5B-3.4 – PRZR 1 OF 4 PRESS LO 5B-4.4 – PRZR 1 OF 4 SI PRESS LO 5B-3.6 – PRZR LVL LO 5C-1.2 – PRZR LVL DEV LO 5C-3.3 – PRZR PRESS LO BACKUP HTRS ON 6A-3.4 – CHRG FLO HI / LO

+10 sec RO/BOP RECOGNIZE Pressurizer level and pressure – LOWERING.

US DIRECT performance of EOP-0.0A, Reactor Trip or Safety Injection.

Examiner Note: The following steps are from EOP-0.0A, Reactor Trip or Safety Injection.

RO VERIFY Reactor Trip: [Step 1 - YES]

VERIFY Reactor Trip Breakers – OPEN. [Step 1.a - YES]

VERIFY Neutron flux – DECREASING. [Step 1.a - YES]

VERIFY all Control Rod Position Rod Bottom Lights – ON. [Step 1.b - NO]

BOP VERIFY Turbine Trip: [Step 2 - YES]

VERIFY all HP Turbine Stop Valves – CLOSED. [Step 2 - YES]

BOP VERIFY Power to AC Safeguards Buses: [Step 3 - YES]

VERIFY AC Safeguards Buses – AT LEAST ONE ENERGIZED. [Step 3.a - YES]

VERIFY both AC Safeguards Buses – ENERGIZED. [Step 3.b - YES]




RO CHECK SI status: [Step 4 - YES]

RO CHECK if SI is actuated. [Step 4.a - NO]

RO CHECK if SI is required. [Step 4.a RNO - YES]

VERIFY Steam Line Pressure < 610 PSIG. [Step 4.a RNO - NO]

VERIFY Pressurizer Pressure < 1820 PSIG. [Step 4.a RNO - YES]

VERIFY Containment Pressure > 3.0 PSIG. [Step 4.a RNO - NO]

RO PLACE 1/1-SIA2, SI MAN ACT Switch to ACT position at CB-07 and DETERMINE SI has actuated. [Step 4.a RNO - YES]

RO VERIFY Both Trains SI Actuated: [Step 4.b - YES]

SI Actuated blue status light – ON NOT FLASHING.

Examiner Note: Emergency boration is performed per the Foldout Page of EOP-0.0A using steps outlined in ABN-107, Emergency Boration, Attachment 1.

CRITICAL TASK

STATEMENT Initiate Emergency Boration Due to Loss of Digital Rod Position Indication Prior to Exiting EOP-0.0, Reactor Trip or Safety Injection.

CRITICAL

TASK RO INITIATE Emergency Boration of 3600 gallons of Boric Acid.

ENSURE a Charging Pump is RUNNING. [Step 1 - YES]

START a Boric Acid Transfer Pump: [Step 2 - YES]

PLACE 1/1-APBA1, BA XFER PMP 1 in START.

PLACE 1/1-8104, EMER BORATE VLV in OPEN. [Step 3 - YES]

VERIFY flow on 1-FI-183A, EMER BORATE FLO. [Step 4 - YES]

VERIFY flow on 1-FI-121A, CHRG FLO. [Step 5 - YES]

Examiner Note: EOP-0.0A, Attachment 2 steps performed by BOP are identified later in the scenario. The RCPs may be tripped if subcooling is observed to be < 25ºF.




US/BOP INITIATE Proper Safeguards Equipment Operation Per Attachment 2. [Step 5 - YES]

RO VERIFY AFW Alignment: [Step 6 - YES]

VERIFY both MDAFW Pumps – RUNNING. [Step 6.a - YES]

PLACE TDAFW Pump in PULLOUT per Foldout Page. [Step 6.b - YES]

VERIFY AFW total flow – GREATER THAN 460 GPM. [Step 6.c - YES]

VERIFY AFW valve alignment - PROPER ALIGNMENT. [Step 6.d - YES]

RO VERIFY Containment Spray NOT Required: [Step 7 - YES]

VERIFY 1-ALB-2B, Window 1-8, CS ACT – NOT ILLUMINATED. [Step 7.a - YES]

VERIFY 1-ALB-2B, Window 4-11, CNTMT ISOL PHASE B ACT – NOT ILLUMINATED. [Step 7.a - YES]

VERIFY Containment pressure – LESS THAN 18.0 PSIG. [Step 7.a - YES]

VERIFY Containment Spray Heat Exchanger Outlet Valves – CLOSED. [Step 7.b - YES]

VERIFY Containment Spray Pumps – RUNNING. [Step 7.c - YES]

RO CHECK if Main Steam lines should be ISOLATED: [Step 8 - YES]

VERIFY Containment pressure – GREATER THAN 6.0 PSIG. [Step 8.a - NO]




VERIFY Steam Line pressure – LESS THAN 610 PSIG. [Step 8.a - NO]

GO to Step 9. [Step 8.a RNO - YES]

RO CHECK RCS Temperature: [Step 9 - YES]

VERIFY RCS Average Temperature – STABLE AT OR TRENDING TO 557°F. [Step 9 - YES]

RO CHECK PRZR Valve Status: [Step 10 - YES]

VERIFY PRZR Safeties – CLOSED. [Step 10.a - YES]

VERIFY Normal PRZR Spray Valves – CLOSED. [Step 10.b - YES]

VERIFY PORVs – CLOSED. [Step 10.c - YES]

VERIFY Power to at least 1 Block Valve – AVAILABLE. [Step 10.d - YES]

VERIFY Block Valves – AT LEAST ONE OPEN. [Step 10.e - YES]

CRITICAL TASK STATEMENT


CRITICAL

TASK RO VERIFY RCS subcooling less than 25ºF (55ºF FOR ADVERSE

CONTAINMENT) and STOP all RCPs. VERIFY RCS subcooling – LESS THAN 25°F. [Step 11.a - YES] VERIFY ECCS Pumps (CCP or SI) – AT LEAST ONE RUNNING.

[Step 11.b - YES]

STOP all RCPs. [Step 11.c - YES]

RO/BOP CHECK if Any Steam Generator Is Faulted: [Step 12 - YES] VERIFY any Steam Generator pressure – DECREASING IN AN

UNCONTROLLED MANNER. [Step 12.a - NO]

VERIFY any Steam Generator pressure – COMPLETELY DEPRESSURIZED. [Step 12.a - NO]


RO/BOP CHECK if Steam Generator Tubes Are NOT Ruptured: [Step 13 - YES]




VERIFY Condenser Off Gas radiation – NORMAL.

VERIFY Main Steam Line radiation – NORMAL.

VERIFY SG Blowdown Sample Radiation Monitor – NORMAL.

VERIFY levels in all Steam Generators – NORMAL.

RO/BOP CHECK if RCS is Intact: [Step 14 - YES]

VERIFY Containment pressure – LESS THAN 1.3 PSIG. [Step 14 - NO]

VERIFY Containment recirculation sump levels – NORMAL. [Step 14 - YES]

VERIFY Containment radiation levels – NORMAL. [Step 14 - NO] GO to EOP-1.0A, Loss of Reactor or Secondary, Step 1.

[Step 14 RNO - YES]

US TRANSITION to EOP-1.0A, Loss of Reactor or Secondary Coolant, Step 1.

Examiner Note: EOP-1.0A, Loss of Reactor or Secondary Coolant steps begin here.

US CHECK If RCPs Should Be Stopped: [Step 1 - YES] VERIFY RCS subcooling less than 25ºF (55ºF FOR ADVERSE

CONTAINMENT). [Step 1.a - YES]

VERIFY ECCS Pumps (CCP or SI) – AT LEAST ONE RUNNING. [Step 1.b - YES]






CRITICAL

TASK RO VERIFY RCS subcooling less than 25ºF (55ºF FOR ADVERSE

CONTAINMENT) and STOP all RCPs.

STOP all RCPs. [Step 1.c - YES]

RO/BOP CHECK if Any Steam Generator Is Faulted: [Step 2 - YES] VERIFY any Steam Generator pressure – DECREASING IN AN

UNCONTROLLED MANNER. [Step 2.a - NO] VERIFY any Steam Generator pressure – COMPLETELY

DEPRESSURIZED. [Step 2.a - NO] GO to Step 3. [Step 2.a RNO - YES]

US CHECK Intact Steam Generator Levels: [Step 3 - YES]

VERIFY Narrow range level – GREATER THAN 43% (50% FOR ADVERSE CONTAINMENT). [Step 3.a - NO]

MAINTAIN total AFW flow greater than 460 GPM until narrow range level GREATER THAN 43% (50% FOR ADVERSE CONTAINMENT). [Step 3.a RNO - YES]

US CHECK Secondary Radiation NORMAL: [Step 4 - YES]

VERIFY Condenser off gas radiation – NORMAL.

VERIFY Main Steam Line radiation – NORMAL.

VERIFY SG Blowdown Sample Radiation Monitor – NORMAL.

VERIFY levels in all Steam Generators – NORMAL.




US CHECK PRZR PORVs and Block Valves: [Step 5 - YES]

VERIFY power to Block Valves – AVAILABLE. [Step 5.a - YES]

VERIFY PORVs – CLOSED. [Step 5.b - YES]

VERIFY Block Valves – AT LEAST ONE OPEN. [Step 5.c - YES]

US/RO CHECK if ECCS Flow Should Be Reduced: [Step 6 - YES]

VERIFY Secondary heat sink conditions – SATISFIED. [Step 6.a - YES] VERIFY total AFW flow to Intact SGs – GREATER THAN 460 GPM. VERIFY intact SG NR level – GREATER THAN 43% (50% FOR

ADVERSE CONTAINMENT)

VERIFY RCS subcooling – GREATER THAN 25°F (55°F FOR ADVERSE CONTAINMENT). [Step 6.b - NO]

GO to Step 7 and OBSERVE CAUTIONS prior to Step 7. [Step 6.b RNO - YES]

RO/BOP RESET ESF Actuation Signals. [Step 7 - YES]

RO/BOP PLACE both EDG EMERG STOP/START handswitches in START. [Step 7.a - YES]

RO/BOP RESET SI. [Step 7.b - YES]

RO/BOP RESET SI Sequencers. [Step 7.c - YES]




Examiner Note: SSW pump 1-01 will trip when Train B Containment Isolation Phase A is reset.

RO/BOP RESET Containment Isolation Phase A and Phase B. [Step 7.d - YES]

RO/BOP RESET Containment Spray Signal. [Step 7.e - YES]

BOP Respond to alarm 1-ALB-01, 1.8, SSWP 1/2 OVRLOAD/TRIP.

BOP DETERMINE SSW Pump 1-01 Trip.

Examiner Note: The following is an initial operator action per ABN-501, section 2.3.

Examiner Note: Diamond steps (◊) are Initial Operator Actions.

◊ BOP ◊ PLACE Train A EDG Emergency Start/Stop handswitch in PULL-OUT. [Step 1 – YES]

Examiner Note: EOP-1.0A, Loss of Reactor or Secondary Coolant steps continue here.

US CHECK If RHR Pumps Should Be Stopped: [Step 8 - YES]

RO/BOP VERIFY RCS pressure – GREATER THAN 325 PSIG (425 PSIG FOR ADVERSE CONTAINMENT). [Step 8.a.1) - YES]

RO/BOP VERIFY RCS pressure – STABLE OR INCREASING. [Step 8.b.2) - YES]

RO/BOP VERIFY RHR Pumps – RUNNING WITH SUCTION ALIGNED TO RWST. [Step 8.b - YES]

RO/BOP STOP RHR Pumps and PLACE in standby. [Step 8.c - YES]

RO/BOP RESET RHR Auto Switchover. [Step 8.d - YES]




US CHECK RCS and SG Pressures: [Step 9 - YES]

RO/BOP VERIFY RCS pressure – STABLE OR DECREASING. [Step 9 - YES]

RO/BOP VERIFY all SG pressures – STABLE OR INCREASING. [Step 9 - YES]

US CHECK If Diesel Generators Should Be Stopped: [Step 10 - YES]

RO/BOP VERIFY AC Safeguards Buses ENERGIZED by Offsite Power. [Step 10.a - YES]

PLACE Train B EDG EMERG STOP/START handswitch in STOP. [Step 10.b - YES]

US INITIATE Evaluation of Plant Status. [Step 11 - YES]

RO/BOP VERIFY Cold Leg Recirculation capability: [Step 11.a - YES] VERIFY Train A RHR Pump – AVAILABLE. [Step 11.a.1) - YES] VERIFY CCW to Train A RHR Pump – AVAILABLE.

[Step 11.a.1) - YES] VERIFY 1/1-8811A, CNTMT SMP TO RHRP 1 SUCT ISOL VLV –

AVAILABLE. [Step 11.a.1) - YES]

VERIFY Train B RHR Pump – AVAILABLE. [Step 11.a.1) - YES] VERIFY CCW to Train B RHR Pump – AVAILABLE.

[Step 11.a.1) - YES] VERIFY 1/1-8811B, CNTMT SMP TO RHRP 2 SUCT ISOL VLV –

AVAILABLE. [Step 11.a.1) - YES] VERIFY 1/1-8804A, RHRP 1 TO CCP SUCT VLV – AVAILABLE.

[Step 11.a.2) - YES] VERIFY 1/1-8804B, RHRP 2 TO SIP SUCT VLV – AVAILABLE.

[Step 11.a.2) - YES]




RO/BOP CHECK Auxiliary Building and Safeguards Building radiation – NORMAL: [Step 11.b - YES] CHECK PC-11 monitors – NORMAL OR Notify Radiation Protection

to take local Radiation Surveys. [Step 11.b - YES]

US NOTIFY Chemistry to obtain RCS samples to assist in determining extent of the accident. [Step 11.c - YES]

US CONTACT Plant Staff to EVALUATE plant equipment. [Step 11.d - YES]

US CHECK if RCS Cooldown and Depressurization Is Required: [Step 12 - YES]

RO/BOP VERIFY RCS pressure – GREATER THAN 325 PSIG (425 PSIG FOR ADVERSE CONTAINMENT). [Step 12.a - YES]

US GO to EOS-1.2A, Post LOCA Cooldown and Depressurization, Step 1. [Step 12.b - YES]

Examiner Note: EOS-1.2A, Post LOCA Cooldown and Depressurization, steps begin here. Steps in [brackets] are from the associated EOS-1.2A Attachments.

RO/BOP [1.D] VERIFY Diesel Generators – NOT RUNNING. [Step 1 - YES]

RO/BOP [1.D] VERIFY SI – RESET. [Step 2 - YES]

RO/BOP [1.D] VERIFY SI Sequencers – RESET. [Step 3 - YES]




RO/BOP [1.D] VERIFY Containment Isolation Phase A and Phase B – RESET. [Step 4 - YES]

RO/BOP [1.D] VERIFY Containment Spray Signal – RESET. [Step 5 - YES]

Booth Operator: If required, EXECUTE remote functions EDR73, 1EB3/11D BKR to CLOSE and EDR74, 1EB4/11D BKR to CLOSE to RESET Instrument Air Compressors 1-01 & 1-02.

RO/BOP [1.D] ESTABLISH Instrument Air and Nitrogen to Containment. [Step 6 - YES]

ESTABLISH Instrument Air: [Step 6.a - YES] VERIFY Air Compressor – RUNNING. [Step 6.a.1) - YES] ESTABLISH Instrument Air to Containment: [Step 6.a.2) - YES]

ESTABLISH Nitrogen: [Step 6.b - YES]

VERIFY ACCUM 1●4 VENT CTRL, 1-HC-943 – CLOSED. [Step 6.b.1) - YES]

OPEN SI/PORV ACCCUM N2 ISOL VLV 1/1-8880. [Step 6.b.2) - YES]

BOP VERIFY all AC Buses – ENERGIZED BY OFFSITE POWER. [Step 7 - YES]

RO DEENERGIZE PRZR Heaters: [Step 8 - YES]

PLACE all PRZR heater switches in OFF position. [Step 8.a - YES]

CONSULT Plant Staff for a recommended minimum indicated PRZR water level that will ensure heaters are covered. [Step 8.b - YES]




US CHECK If RHR Pumps Should Be Stopped. [Step 9 - YES]

VERIFY RHR Pumps – ANY RUNNING WITH SUCTION ALIGNED TO RWST. [Step 9.a - NO]


US CHECK Intact SG Levels: [Step 10 - YES]

VERIFY narrow range level – GREATER THAN 43% (50% FOR ADVERSE CONTAINMENT): [Step 10.a - YES]

CONTROL AFW flow to maintain narrow range level between 43% (50% FOR ADVERSE CONTAINMENT) and 60%. [Step 10.b - YES]

+30 min US INITIATE RCS Cooldown to Cold Shutdown: [Step 11 - YES]

MAINTAIN cooldown rate in RCS Cold Legs – LESS THAN 100°F/HR. [Step 11.a - YES]

BLOCK Low Main Steam Pressure SI signal when Pressurizer pressure – LESS THAN 1960 PSIG. [Step 11.b - YES]

DUMP steam from intact Steam Generators via ARVs. [Step 11.c - YES]




US CHECK RCS subcooling – GREATER THAN 25°F (55°F FOR ADVERSE CONTAINMENT). [Step 12 - NO]

GO to Step 25. [Step 12 RNO - YES]

US VERIFY ECCS Flow Not Required: [Step 25 - YES]

CHECK RCS subcooling – GREATER THAN 25°F (55°F FOR ADVERSE CONTAINMENT). [Step 25.a - NO]

Manually START ECCS Pumps as necessary. [Step 25.a RNO - YES]


When a transition to Step 26 of EOS-1.2A is required, TERMINATE the scenario.




Examiner Note: These steps are performed by the BOP per EOP-0.0A, Attachment 2.

BOP VERIFY SSW Alignment: [Step 1 - YES]

VERIFY SSW Pumps – RUNNING. [Step 1.a - YES]

VERIFY Train A EDG Cooler SSW return flow. [Step 1.b - YES]

BOP VERIFY Safety Injection Pumps – RUNNING. [Step 2 - YES]

BOP VERIFY Containment Isolation Phase A – APPROPRIATE MLB LIGHT INDICATION (RED WINDOWS). [Step 3 - YES]

BOP VERIFY Containment Ventilation Isolation – APPROPRIATE MLB LIGHT INDICATION (GREEN WINDOWS). [Step 4 - YES]

BOP VERIFY CCW Pumps – RUNNING. [Step 5 - NO]

BOP DETERMINE CCW Pump 1-02 failed to start and MANUALLY START CCW Pump 1-02. [Step 5 RNO - YES]

BOP VERIFY RHR Pumps – RUNNING. [Step 6 - YES]

BOP VERIFY Proper CVCS Alignment: [Step 7 - YES]

VERIFY CCPs – RUNNING. [Step 7.a - YES]

VERIFY Letdown Relief Valve Isolation: [Step 7.b - YES]

Letdown Orifice Isolation Valves – CLOSED. [Step 7.b.1) - YES]

Letdown Isolation Valves 1/1-LCV-459 & 1/1-LCV-460 – CLOSED. [Step 7.b.2) - YES]




BOP VERIFY ECCS flow: [Step 8 - YES]

CCP SI flow indicator – CHECK FOR FLOW. [Step 8.a - YES]

RCS pressure – LESS THAN 1700 PSIG (1800 PSIG FOR ADVERSE CONTAINMENT). [Step 8.b - YES]

SIP discharge flow indicator – CHECK FOR FLOW. [Step 8.c - YES]

RCS pressure – LESS THAN 325 PSIG (425 PSIG FOR ADVERSE CONTAINMENT). [Step 8.d - NO]

GO to Step 9. [Step 8.d RNO) - YES]

BOP VERIFY Feedwater Isolation Complete: [Step 9 - YES]

Feedwater Isolation Valves – CLOSED.

Feedwater Isolation Bypass Valves – CLOSED.

Feedwater Bypass Control Valves – CLOSED.

Feedwater Control Valves – CLOSED.

BOP VERIFY Train B Diesel Generator – RUNNING. [Step 10 - YES]

BOP VERIFY Monitor Lights for SI Load Shedding on 1-MLB-9 and 1-MLB10 – LIT. [Step 11 - YES]

BOP VERIFY Proper SI alignment – PROPER MLB LIGHT INDICATION. [Step 12 - YES]

BOP INITIATE periodic monitoring of Spent Fuel Cooling. [Step 13 - YES]

Spent Fuel Pool temperature (T2900A, T2901A).

Spent Fuel Pool level (L4800A, L4801A, L4802A, L4803A).




BOP VERIFY Components on Table 1 are Properly Aligned. [Step 14 - YES]

Location Equipment Description Condition

CB-03 X-HS-5534 H2 PRG SPLY FN 4 STOPPED

CB-03 X-HS-5532 H2 PRG SPLY FN 3 STOPPED

CB-04 1/1-8716A RHRP 1 XTIE VLV OPEN

CB-04 1/1-8716B RHRP 2 XTIE VLV OPEN

CB-06 1/1-8153 XS LTDN ISOL VLV CLOSED

CB-06 1/1-8154 XS LTDN ISOL VLV CLOSED

CB-07 1/1-RTBAL RX TRIP BKR OPEN

CB-07 1/1-RTBBL RX TRIP BKR OPEN

CB-07 1/1-BBAL RX TRIP BYP BKR OPEN/DEENERGIZED

CB-07 1/1-BBBL RX TRIP BYP BKR OPEN/DEENERGIZED

CB-08 1-HS-2397A SG 1 BLDN HELB ISOL VLV CLOSED




CB-08 1-HS-2111C FWPT A TRIP TRIPPED

CB-08 1-HS-2112C FWPT B TRIP TRIPPED

CB-09 1-HS-2490 CNDS XFER PUMP STOPPED (MCC deenergized on SI)

CV-01 X-HS-6181 PRI PLT SPLY FN 17 & INTK DMPR

STOPPED/DEENERGIZED


STOPPED/DEENERGIZED





STOPPED/DEENERGIZED


STOPPED/DEENERGIZED


STOPPED/DEENERGIZED


STOPPED/DEENERGIZED


STOPPED/DEENERGIZED


STOPPED/DEENERGIZED

CV-01 X-HS-3631 UPS & DISTR RM A/C FN 1 & BSTR FN 42

STARTED

CV-01 X-HS-3632 UPS & DISTR RM A/C FN 2 & BSTR FN 43

STARTED

CV-01 1-HS-5600 ELEC AREA EXH FN 1 STOPPED/DEENERGIZED

CV-01 1-HS-5601 ELEC AREA EXH FN 2 STOPPED/DEENERGIZED

CV-01 1-HS-5602 MS & FW PIPE AREA EXH FN 3 & EXH DMPR

STOPPED/DEENERGIZED

CV-01 1-HS-5603 MS & FW PIPE AREA EXH FN 4 & EXH DMPR

STOPPED/DEENERGIZED

CV-01 1-HS-5618 MS & FW PIPE AREA SPLY FN 17

STOPPED/DEENERGIZED

CV-01 1-HS-5620 MS & FW PIPE AREA SPLY FN 18

STOPPED/DEENERGIZED

CV-03 X-HS-5855 CR EXH FN 1 STOPPED/DEENERGIZED

CV-03 X-HS-5856 CR EXH FN 2 STOPPED/DEENERGIZED

CV-03 X-HS-5731 SFP EXH FN 33 STOPPED/DEENERGIZED




Examiner Note: The next four (4) steps would be performed on Unit 2.

CB-03 2-HS-5538 AIR PRG EXH ISOL DMPR CLOSED

CB-03 2-HS-5539 AIR PRG EXH ISOL DMPR CLOSED

CB-03 2-HS-5537 AIR PRG SPLY ISOL DMPR CLOSED




CB-03 2-HS-5536 AIR PRG SPLY ISOL DMPR CLOSED

BOP NOTIFY Unit Supervisor attachment instructions complete AND to IMPLEMENT FRGs as required. [Step 14 - YES]

EOP-0.0A, Attachment 2 steps are now complete.



Facility: CPNPP 1 & 2 Scenario No.: 2 Op Test No.: June 2012 NRC

Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 908 ppm (by sample). Turnover: Maintain steady-state power conditions. Boric Acid Transfer Pump 1-01 is OOS.

Critical Tasks: Close PORV Block Valve to Maintain RCS Pressure Greater Than Low Pressure Reactor Trip Setpoint Following Entry Into ABN-705, Pressurizer Pressure Malfunction.

Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-710, Steam Generator Level Instrumentation Malfunction.

Initiate Emergency Boration due to Two (2) Stuck Control Rods Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

Trip Reactor Coolant Pumps Prior to Establishing Bleed and Feed per FRH-0.1A, Response to Loss of Secondary Heat Sink.

Establish Core Cooling via Reactor Coolant System Bleed and Feed per FRH-0.1A, Response to Loss of Secondary Heat Sink.

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

1 +10 min

RX08B RX16B

I (RO, SRO) TS (SRO)

Pressurizer Pressure Channel (PT-456) Fails High. Stuck Power Operated Relief Valve (PCV-456) at 50% Open.

2 +20 min

RX04B I (BOP, SRO) TS (SRO)

Steam Generator (1-02) Level Transmitter (LT-552) Fails High.

3 +30 min

RP06B I (RO, SRO) TS (SRO)

Loop 2 N-16 Channel (JI-421A/B) Fails Low.

4 +40 min

FW03A TC09C

R (RO) N (BOP, SRO)

Main Feedwater Pump A Trip. Automatic Turbine Runback Failure.

5 +45 min

FW06A M (RO, BOP, SRO)

Main Feedwater Pump A Recirculation Valve (FV-2289) Fails Open. Loss of both Main Feedwater Pumps.

6 +45 min

RD04K2 RD04K8

C (RO)

Two Control Rods Fail to Insert Upon Reactor Trip. Emergency Boration Required.

7 +47 min

FW08B

C (BOP)

Motor Driven Auxiliary Feedwater Pump (1-02) Start Failure.

8 +47 min

OVRDE FW09B

C (BOP) Turbine Driven Auxiliary Feedwater Pump (TDAFW) Steam Valves Fail to Auto Open Followed by TDAFW Trip (60 second time delay).

9 +50 min

ED05H C (BOP) M (RO, BOP, SRO)

6900 Volt Train A Safeguards Bus 1EA1 86-1 Trip and Lockout Five minutes Post Trip. Loss of Feedwater Flow.

* (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) 2 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 5 Critical tasks (2-3)



SCENARIO SUMMARY NRC #2 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. Boric Acid Transfer Pump 1-01 is out-of-service. The first event it 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 Block Valve on a partially open Power Operated Relief Valve (PORV), 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. The next event is a high failure of a Steam Generator (1-02) Level Transmitter. 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 Technical Specifications are referenced, a low failure of TCOLD Temperature Instrument, TI-421A will occur. Operator actions are per ABN-704, TC/N-16 Instrumentation Malfunction, Section 2.0, and require stopping Control Rod motion and stabilizing Reactor Coolant System (RCS) temperature and Pressurizer level. The SRO will refer to Technical Specifications. When RCS temperature is restored to normal, a Main Feedwater Pump (MFW) will trip and an AUTO Turbine Runback will fail to initiate. Rod Control will be placed in AUTO and a Manual Turbine Runback to 900 MWe is commenced per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 2.0. The MFW Pump Recirculation Valve will fail open shortly after the Runback initiates and result in a lowering MFW Pump suction pressure and a trip of the second MFW Pump. When this occurs, the crew will manually trip the Reactor and enter EOP-0.0A, Reactor Trip or Safety Injection. Two Control Rods failed to insert on the Reactor Trip and an Emergency Boration is required. While performing the actions of EOP-0.0A, a Motor Driven Auxiliary Feedwater (AFW) Pump will fail to start and require placing the Turbine Driven AFW (TDAFW) Pump in service. Shortly after the TDAFW Pump is started, the pump will trip. The crew will transition from EOP-0.0A to EOS-0.1A, Reactor Trip Response. Five minutes after the Reactor trips, Train A Safeguards Bus 1EA1 will trip and lockout resulting in a Loss of Feedwater Flow. Entry into FRH-0.1A, Response to Loss of Secondary Heat Sink, will be performed with an immediate transition to Step 12 for RCS Bleed and Feed. The scenario success path includes tripping the Reactor Coolant Pumps and opening the Pressurizer and Reactor Head Vents. Risk Significance:

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

Risk significant core damage sequence: Loss of Feedwater Flow

Risk significant operator actions: Close PORV Block Valve

Restore Steam Generator 1-02 Level Control

Start TDAFW Pump

Trip Reactor Coolant Pumps

Initiate RCS Bleed and Feed



BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP

Initialize to IC #18 and NRC #2 Scenario File.

EVENT TYPE MALF # DESCRIPTION DEMAND VALUE

INITIATING PARAMETER

SETUP Boric Acid Transfer Pump 1-01 PULLOUT -

Boric Acid Transfer Pump 1-02 AUTO -

RD04K2 Stuck Control Bank B Control Rod K2 STUCK K0

RD04K8 Stuck Control Bank A Control Rod K8 STUCK K0

FW08B Motor Driven AFW Pump (1-02) Start Failure OFF K0

OVRDE TDAFW Pump Steam Valves Failure - K0

FW09B TDAFW Pump Trip TRIP 60 second TD Post Start

ED05H Train A Safeguards Bus 1EA1 Trip and Lockout TRIP 300 second TDPost Trip

NOTE: ENSURE keys for Pressurizer & Reactor Head Vent Valves are moved to top of Key Locker.

1 RX08B Pressurizer Pressure Channel (PT-456) Failure 100% K1

1 RX16B Stuck PORV (PCV-456) at 50% Open 50% K1

1 RCR24 PORV Block Valve 1/1-8000B Breaker OPEN K6

2 RX04B SG (1-02) Level Transmitter (LT-552) 100% K2

3 RP06B Loop 2 N-16 Channel (JI-421A/B) Failure 0% K3

4 FW03A Main Feedwater Pump (1-01) Trip TRIP K4

4 TC09C Automatic Turbine Runback Failure FAIL K4

5 FW06A MFW Pump (1-01) Recirculation Valve Failure OPEN K4 5 second delay

6 RD04K2 Stuck Control Bank B Control Rod K2 STUCK K0

6 RD04K8 Stuck Control Bank A Control Rod K8 STUCK K0

7 FW08B Motor Driven AFW Pump (1-02) Start Failure OFF K0



8 OVRDE DIFWHS24521

TDAFW Pump Steam Supply Valve 1-HS-2451-1 from SG 1-04 (NOTE 1)

CLOSED K0

8 OVRDE DIFWHS24522

TDAFW Pump Steam Supply Valve 1-HS-2451-2 from SG 1-01 (NOTE 1)

CLOSED K0

NOTE 1: 1-HS-2452-1/2, TDAFW Pump Valves will NOT AUTO open but will open when taken to START. This will prevent an AUTO start on 86-1 Trip and Lockout if TDAFW Pump was not previously started.

8 FW09B TDAFW Pump Trip TRIP 60 second TD Post Start

8 TDAFW Pump Trip {AOFW SI 2452A. Value > 1600} IMF FW09B f:1 d:60

9 ED05H Train A Safeguards Bus 1EA1 Trip and Lockout 86-1 300 second TDPost Trip

9 EDR74 IA Compressor 1-02 1EB4/11D FDR BKR CLOSED When directed

RCR24 PORV Block Valve 1/1-8000B Breaker (NOTE 2) ON K7

NOTE 2: Power is restored to PORV Block Valve 1/1-8000B when FRH-0.1A is entered.



Booth Operator: INITIALIZE to IC #18 and NRC #2 Scenario File. ENSURE all Simulator Annunciator Alarms are ACTIVE. ENSURE Boric Acid Transfer Pump 1-01 in PULLOUT and TAGGED. ENSURE Boric Acid Transfer Pump 1-02 in AUTO. ENSURE keys for PRZR & RX Head Vent Valves moved to top of Key Locker. ENSURE Operator Aid Tags reflect current boron conditions. ENSURE Rod Bank Update (RBU) is performed. ENSURE Turbine Load Rate set at 10 MWe/minute. ENSURE 60/90 buttons DEPRESSED on ASD. ENSURE ASD speakers are ON at half volume. ENSURE Reactivity Briefing Sheet printout provided with Turnover. ENSURE procedures in progress are on SRO desk: - COPY of IPO-003A, Power Operations, Section 5.5, Operating at Constant Turbine Load. ENSURE Control Rods are in AUTO with Bank D at 215 steps.

Control Room Annunciators in Alarm: PCIP-1.1 – SR TRN A RX TRIP BLK PCIP-1.2 – IR TRN A RX TRIP BLK PCIP-1.4 – CNDSR AVAIL STM DMP ARMED C-9 PCIP-1.6 – RX ≥ 10% PWR P-10 PCIP-2.1 – SR TRN B RX TRIP BLK PCIP-2.2 – IR TRN B RX TRIP BLK PCIP-2.5 – SR RX TRIP BLK PERM P-6 PCIP-3.2 – PR TRN A LO SETPT RX TRIP BLK PCIP-4.2 – PR TRN B LO SETPT RX TRIP BLK

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 1 Page 7 of 28 Event Description: Pressurizer Pressure Channel Failure / Power Operated Relief Valve Failure



Booth Operator: When directed, EXECUTE Event 1 (Key 1). - RX08B, Pressurizer Pressure Channel (PT-456) fails high. - RX16B, PORV (PCV-456) fails 50% open.

Indications Available: 5B-3.1 – PRZR PORV OUT TEMP HI 5B-4.1 – PRZR ANY SFTY RLF VLV OUT TEMP HI 5C-1.4 – PORV 455A / 456 NOT CLOSE 5C-2.1 – PRZR PRESS HI 5C-3.1 – PRZR 1 OF 4 PRESS HI 5C-3.3 – PRZR PRESS LO BACKUP HTRS ON 6A-3.4 – CHRG FLO HI / LO (~10 seconds later) 1-TI-463, PRZR PORV OUT TEMP indication rising 1-PI-456, PRZR PRESS CHAN II indication failed high 1/1-PCV-456, PRZR PORV indicates mid position

+1 min RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE PRZR pressure lowering.

US DIRECT performance of ABN-705, Pressurizer Pressure Malfunction, Section 2.0.


◊ RO ◊ VERIFY PORV – CLOSED. [Step 2.3.1 - NO]

PLACE 1/1-PCV-456, PRZR PORV in CLOSE. [Step 2.3.1 RNO - YES]


Close PORV Block Valve to Maintain Reactor Coolant System Pressure Greater Than Low Pressure Reactor Trip Setpoint Following Entry into ABN-705, Pressurizer Pressure Malfunction.




CRITICAL

TASK ◊ RO ◊ CLOSE PORV Block Valve. [Step 2.3.1 RNO - YES]

◊ RO ◊ PLACE 1/1-8000B, PRZR PORV BLK VLV in CLOSE. [Step 2.3.1 RNO - YES]

◊ RO ◊ PLACE 1-PK-455A, PRZR MASTER PRESS CTRL in MANUAL. [Step 2.3.2 - YES]

◊ RO ◊ ADJUST 1-PK-455A for current RCS pressure. [Step 2.3.3 - YES]

RO TRANSFER 1/1-PS-455F, PRZR PRESS CTRL CHAN SELECT to an Alternate Controlling Channel. [Step 2.3.4 - YES]

RO PLACE 1-PK-455A, PRZR MASTER PRESS CTRL in AUTO. [Step 2.3.5 - YES]

RO VERIFY automatic control restoring Pressurizer pressure to 2235 PSIG. [Step 2.3.6 - YES]

RO ENSURE valid channel to recorder 1/1-PS-455G, 1-PR-455 PRZR PRESS SELECT. [Step 2.3.7 - YES]

RO VERIFY 1/1-PCV-456, PRZR PORV failed in mid position. [Step 2.3.8 - YES]

RO MAINTAIN 1/1-8000B, PRZR PORV BLK VLV in CLOSE due to PORV in mid position. [Step 2.3.9 - YES]

US/RO Within one hour, VERIFY PCIP Window 2.6 - PRZR PRESS SI BLK PERM P-11 – DARK. [Step 2.3.10 - YES]

US/RO VERIFY other instruments on common instrument line - NORMAL. [Step 2.3.11 - YES]

VERIFY Loop 2 Instruments LT-460 and PT-458 responding normally per Attachment 1.




Examiner Note: The next two (2) steps are only performed following I&C maintenance.

Within 72 hours, PERFORM the following: [Step 2.3.12 - YES]

- PLACE 1/1-RBSS, CONTROL ROD BANK SELECT in MANUAL. [Step 2.3.12.a - YES] - SELECT the failed channel on following switches: [Step 2.3.12.b - YES]

1-TS-412T, TAVE CHAN DEFEAT 1/1-JS-411E, N16 PWR CHAN DEFEAT

- ENSURE valid N16 channel is supplying recorder: [Step 2.3.12.c - YES] 1/1-TS-411E, 1-TR-411 CHAN SELECT

- CONTACT I&C to place bistable test switches for PT-456 in CLOSE. [Step 2.3.12.d - YES]

VERIFY appropriate alarms and trip status lights ON per Attachment 4. [Step 2.3.13 - YES]

OBSERVE TSLB-1, Window 2.7 – PRZR PRESS LO PB-456D is LIT.

OBSERVE TSLB-5, Window 2.2 – PRZR PRESS LO PB-456C is LIT.

OBSERVE TSLB-5, Window 2.3 – PRZR PRESS HI PB-456A is LIT.

OBSERVE TSLB-9, Window 2.3 – PRZR PRESS SI PERM PB-456B is LIT.




+10 min US EVALUATE Technical Specifications. [Step 2.3.14 - YES]

LCO 3.3.1.E, Reactor Trip System Instrumentation (Function 8.b).

CONDITION E - One channel inoperable. ACTION E.1 - Place channel in trip within 72 hours, OR ACTION E.2 - Be in MODE 3 within 78 hours.

LCO 3.3.1.M, Reactor Trip System Instrumentation (Function 8.a).

CONDITION M - One channel inoperable. ACTION M.1 - Place channel in trip within 72 hours, OR ACTION M.2 – Reduce THERMAL POWER to < P-7 within 78 hours.

LCO 3.3.2.D, ESFAS Instrumentation (Function 1.d).

CONDITION D - One channel inoperable. ACTION D.1 - Place channel in trip within 72 hours, OR ACTION D.2.1 - Be in MODE 3 within 78 hours, AND ACTION D.2.2 - Be in MODE 4 within 84 hours.

LCO 3.3.2.L, ESFAS Instrumentation (Function 8.b).

CONDITION L - One or more required channel(s) inoperable. ACTION L.1 - Verify interlock is in required state for existing unit

condition within one hour, OR ACTION L.2.1 - Be in MODE 3 within 7 hours, AND ACTION L.2.2 - Be in MODE 4 within 13 hours.

Booth Operator: When contacted to remove power to 1/1-8000B, PRZR PORV BLK VLV, EXECUTE remote function RCR24 (Key 6), 1/1-8000B to OFF.

LCO 3.4.11.B, Pressurizer Power Operated Relief Valve.

CONDITION B - One PORV inoperable and not capable of being manually cycled.

ACTION B.1 - Close associated block valve within 1 hour, AND ACTION B.2 - Remove power from associated block valve within 1

hour, AND ACTION B.3 - Restore PORV to OPERABLE status within 72 hours.




US INITIATE a work request per STA-606. [Step 2.3.15 - YES]


When Technical Specifications are addressed, or at Lead Examiner discretion, PROCEED to Event 2.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 2 Page 12 of 28 Event Description: Steam Generator Level Channel Failure



Booth Operator: When directed, EXECUTE Event 2 (Key 2). - RX04B, Steam Generator 1-02 Level Channel (LT-552) fails high.

Indications Available: 8A-2.8 – SG 2 STM & FW FLO MISMATCH 8A-2.12 – SG 2 LVL DEV 1-LI-552, SG 2 LVL (NR) CHAN II indication failed high

+30 sec BOP RESPOND to Annunciator Alarm Procedures.

BOP RECOGNIZE Steam Generator 1-02 level lowering.

Examiner Note: Steam Generator (SG) Level Channel II failing HIGH will cause the Feedwater Control Valve to CLOSE, thereby lowering SG level. Unit 1 LOW LEVEL REACTOR TRIP is at 38%.

US DIRECT performance of ABN-710, Steam Generator Level Instrumentation Malfunction, Section 2.0.

BOP RECOGNIZE Steam Generator 1-02 Level Channel (LT-552) failed high. [Step 2.3.1 - YES]

BOP VERIFY controlling level channel – FAILED. [Step 2.3.1 - YES]

CRITICAL TASK

STATEMENT Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-710, Steam Generator Level Instrumentation Malfunction.

CRITICAL

TASK BOP PLACE 1-FK-520, SG 2 FW FLO CTRL in MANUAL and CONTROL level. [Step 2.3.2 - YES]

BOP VERIFY instruments on common instrument line – NORMAL. [Step 2.3.3 - YES]

VERIFY Loop 2 Instrument LT-502 responding normally per Attachment 1.




BOP VERIFY all HI-HI level bistable windows on TSLB-3 for SG 1-02 – DARK. [Step 2.3.4 - YES]

OBSERVE TSLB-3, Window 1.3 – SG 2 LVL HI-HI LB-529A is DARK.



BOP VERIFY automatic SG level control available: [Step 2.3.5 - YES]

OBSERVE alternate level control channel 1-LI-529A indication NORMAL.[Step 2.3.5.a - YES]

DETERMINE automatic level control desired by Unit Supervisor. [Step 2.3.5.b - YES]

BOP SELECT Alternate Channel: [Step 2.3.6 - YES]

PLACE 1-LS-529C, SG 2 LVL CHAN SELECT to the LY-529 position.

BOP VERIFY affected SG conditions for auto level control: [Step 2.3.7 - YES]

OBSERVE Feedwater and Steam Flows – MATCHED.

OBSERVE Steam Generator Level – STABLE AT PROGRAM.




BOP PLACE 1-FK-520, SG 2 FW FLO CTRL in AUTO and MONITOR operation. [Step 2.3.8 - YES]

Examiner Note: The next two (2) steps are performed after I&C troubleshooting is completed.

WITHIN 72 hours, CONTACT I&C to PLACE bistable test switches for Level Channel II in CLOSE per Attachments 2 and 3. [Step 2.3.9 - YES]

VERIFY appropriate alarms and Trip Status Lights LIT per Attachment 3 and NOTE in the Unit Log. [Step 2.3.10 - YES]

OBSERVE TSLB-3, Window 1.3 – SG 2 LVL HI-HI LB-529A is LIT.

OBSERVE TSLB-5, Window 2.5 – SG 2 LVL LO-LO LB-552B is LIT.


LCO 3.3.1.E, Reactor Trip System Instrumentation (Function 14).

CONDITION E - One channel inoperable (Channel 2 LO-LO). ACTION E.1 - Place channel in trip within 72 hours, OR ACTION E.2 – Be in MODE 3 within 78 hours.

LCO 3.3.2.D, ESFAS Instrumentation (Functions 6.c).

CONDITION D - One channel inoperable. ACTION D.1 - Place channel in trip within 72 hours, OR ACTION D.2.1 - Be in MODE 3 within 78 hours, AND ACTION D.2.2 - Be in MODE 4 within 84 hours.

LCO 3.3.2.I, ESFAS Instrumentation (Function 5.b).

CONDITION I - One channel inoperable (Channel 2 HI-HI). ACTION I.1 - Place channel in trip within 72 hours, OR ACTION I.2 – Be in MODE 3 within 78 hours.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 3 Page 16 of 28 Event Description: Reactor Protection System Loop 2 N16 Channel Failure



Booth Operator: When directed, EXECUTE Event 3 (Key 3). - RP06B, Loop 2 N-16 Channel (1-JI-421A/B) fails low.

Indications Available: 5C-1.3 – PRZR LVL DEV HI 5C-1.5 – ANY N16 DEV HI/LO 5C-3.5 – ANY TAVE DEV HI/LO 6D-1.10 – ANY TAVE TREF DEV 6D-3.13 – TREF-AUCT LO TAVE MISMATCH 6D-4.13 – AUCT TAVE LO 1-JI-421A/B, RC LOOP 2 N16 PWR CHAN II indication failed low 1-TI-421B, RC LOOP 2 OT N16 SETPOINT CHAN II indication reads higher than the others

+30 sec RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE Loop 4 N-16 Channel failure at 1-JI-441A/B on CB-05 and/or CB-07.

US DIRECT performance of ABN-704, Tc / N16 Instrumentation Malfunction, Section 2.0.

RO PLACE 1/1-RBSS, CONTROL ROD BANK SELECT in MANUAL. [Step 2.3.1 - YES]

RO PLACE 1-TS-412T, TAVE CHAN DEFEAT in LOOP 2 position. [Step 2.3.2 - YES]

BOP VERIFY Steam Dump System – NOT ACTUATED AND NOT ARMED. [Step 2.3.3 - YES]




RO Using Control Rods, RESTORE TAVE to within 1ºF of TREF. [Step 2.3.4 - YES]

RO PLACE 1/1-JS-411E, N16 PWR CHAN DEFEAT in LOOP 2 position. [Step 2.3.5 - YES]

RO ENSURE valid N16 Channel supplying recorder on 1/1-TS-411E, 1-TR-411 CHAN SELECT. [Step 2.3.6 - YES]

BOP VERIFY Steam Dump System – NOT ARMED. [Step 2.3.7 - YES]

PCIP Window 3.4 – TURB LOAD REJ STM DMP ARMED C-7 is DARK.

BOP VERIFY Steam Dump System – NOT BLOCKED. [Step 2.3.8 - YES]

Examiner Note: The next two (2) steps are performed after I&C troubleshooting is completed.

WITHIN 72 hours, CONTACT I&C to PLACE bistable test switches for LOOP 2 in CLOSE per Attachments 1 and 2. [Step 2.3.9 - YES]

VERIFY appropriate alarms and Trip Status Lights LIT per Attachment 3 and NOTE in the Unit Log. [Step 2.3.10 - YES]

VERIFY TSLB-5, Window 2.8 – RC LOOP 2 OT N16 is LIT.

VERIFY TSLB-9, Window 2.4 – OT N16 ROD STOP & TURB RUNBACK is LIT.


LCO 3.3.1.E, Reactor Trip System Instrumentation (Function 6 & 7).

CONDITION E - One channel inoperable (Channel 2 LO-LO). ACTION E.1 - Place channel in trip within 72 hours, OR ACTION E.2 – Be in MODE 3 within 78 hours.


Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 4 & 5 Page 19 of 28 Event Description: Main Feedwater Pump Trip / Automatic Turbine Runback Failure / MFWP Recirculation Valve Failure



Booth Operator: When directed, EXECUTE Event 4 (Key 4). - FW03A, Main Feedwater Pump A trip. - TC09C, Auto Turbine Runback failure.

Indications Available: 7B-1.12 – FWPT A TRIP 8A-1.8 / 2.8 / 3.8 / 4.8 – SG 1 / 2 / 3 / 4 STM & FW FLO MISMATCH 8A-1.12 / 2.12 / 3.12 / 4.12 – SG 1 / 2 / 3 / 4 LVL DEV (~10 seconds later) 6D-1.9 – ANY TURB RUNBACK EFFECTIVE (when Manual Runback initiated) 6D-1.10 – AVE TAVE-TREF DEV (when Manual Runback initiated) 1-HS-2111, FWPT A TRIP light LIT Steam Dump System Group 1 Valves OPEN Control Rods stepping IN

+30 sec RO/BOP RESPOND to Annunciator Alarm Procedures.

BOP RECOGNIZE trip of Main Feedwater Pump A with no Automatic Turbine Runback.

US DIRECT performance of ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 2.0.


◊ RO/BOP ◊ VERIFY automatic plant response. [Step 2.3.1 - NO]

◊ RO ◊ - VERIFY Control Rods in – AUTO.




◊ BOP ◊ - VERIFY Turbine Runback – IN PROGRESS.

◊ RO/BOP ◊ If Turbine Power is > ~700 MWe, PERFORM the following: [Step 2.3.1 RNO - YES].

◊ RO ◊ - PLACE 1/1-RBSS, CONTROL ROD BANK SELECT in AUTO. [Step 2.3.1.a RNO - YES]. ◊ BOP ◊ - ENSURE Turbine Runback to 700 MWE initiated. [Step 2.3.1.b RNO - YES]. DEPRESS “700 MWe” MANUAL RUNBACK button.

CLICK on “0/1” button.

CLICK on “Execute” then VERIFY Manual Runback in progress.

Examiner Note: 5 seconds following Main Feedwater Pump A Trip.

- FW06A, Main Feedwater Pump A Recirculation Valve fails open.

Examiner Note: When Main Feedwater Pump (MFWP) A Recirculation Valve fails open, Steam Generator level will lower and an automatic Reactor Trip will occur.

RO/BOP STABILIZE Reactor power using one or more of the following: [Step 2.3.2 - YES]

- Control Rods / Steam Dumps / Boration / Turbine Load.

BOP VERIFY Main Feedwater flow to Steam Generators. [Step 2.3.3 - YES]

- VERIFY one Main Feedwater Pump – RUNNING. [Step 2.3.3.a - NO] - VERIFY Main Feedwater – ALIGNED. [Step 2.3.3.a - NO] PERFORM the following: [Step 2.3.3.a RNO - YES]

If SG level is lowering in an uncontrolled manner, TRIP the Reactor and GO to EOP-0.0A. [Step 2.3.3.a.1) RNO - YES]




THROTTLE AFW Control Valves to 25%. [Step 2.3.3.a.2) RNO - YES]

START both Motor Driven AFW Pumps. [Step 2.3.3.a.3) RNO - YES]

ENSURE Feedwater System Bypass Valves – CLOSED. [Step 2.3.3.a.4) RNO - YES]

ENSURE Main Feedwater ALIGNED. [Step 2.3.3.a.5) RNO - NO]

+10 min ADJUST AFW Flow Control Valves to maintain SG narrow range level between 60% and 75%. [Step 2.3.3.a.6) RNO - YES]

When the Reactor is tripped, or at Lead Examiner discretion, PROCEED to Events 6, 7, 8, and 9.

Appendix D Operator Action Form ES-D-2 Operating Test : NRC Scenario # 2 Event # 6, 7, 8, & 9 Page 22 of 28 Event Description: Motor Driven Auxiliary Feedwater (AFW) Pump Start Failure / Two Stuck Control Rods / Turbine

Driven AFW Pump Steam Supply Valve Failure & Trip / Train A Safeguards Bus Trip and Lockout Time Position Applicant’s Actions or Behavior


Booth Operator: When directed, EXECUTE Events 6, 7, 8, and 9. - RD04K2 / RD04K8, Two (2) Stuck Control Rods. - FW08B, Motor Driven Auxiliary Feedwater Pump 1-02 start failure. - OVRDE, Turbine Driven Auxiliary Feedwater Pump Steam Supply Valves fail to AUTO open. - FW09B, Turbine Driven Auxiliary Feedwater Pump trip. - ED05H, Train A Bus 1EA1 86-1 trip and lockout (5 minutes post trip).

Indications Available: Numerous Reactor Trip alarms

US DIRECT performance of EOP-0.0A, Reactor Trip or Safety Injection.

Examiner Note: The following steps are from EOP-0.0A, Reactor Trip or Safety Injection.

RO VERIFY Reactor Trip: [Step 1 - YES]


VERIFY Neutron flux – DECREASING. [Step 1.a - YES]

VERIFY all Control Rod Position Rod Bottom Lights – ON. [Step 1.b - NO]

BOP VERIFY Turbine Trip: [Step 2 - YES]

VERIFY all HP Turbine Stop Valves – CLOSED. [Step 2 - YES]

BOP VERIFY Power to AC Safeguards Buses: [Step 3 - YES]

VERIFY AC Safeguards Buses – AT LEAST ONE ENERGIZED. [Step 3.a - YES]

VERIFY both AC Safeguards Buses – ENERGIZED. [Step 3.b - YES]

RO CHECK SI status: [Step 4 - YES]

CHECK if SI is actuated. [Step 4.a - NO]

CHECK if SI is required. [Step 4.a RNO - YES]

VERIFY Steam Line Pressure < 610 PSIG. [Step 4.a RNO - NO]

VERIFY Pressurizer Pressure < 1820 PSIG. [Step 4.a RNO - NO]

VERIFY Containment Pressure > 3.0 PSIG. [Step 4.a RNO - NO]




US IF SI is NOT required, GO to EOS-0.1A, Reactor Trip Response, Step 1. [Step 4.a RNO - YES]

Examiner Note: Emergency boration is performed per the Foldout Page of EOP-0.0A using steps outlined in ABN-107, Emergency Boration, Attachment 1.

CRITICAL TASK

STATEMENT Initiate Emergency Boration Due to Two (2) Stuck Control Rods Prior to Exiting EOP-0.0, Reactor Trip or Safety Injection.

CRITICAL

TASK RO INITIATE Emergency Boration of 3600 gallons of Boric Acid.

ENSURE a Charging Pump is RUNNING. [Step 1 - YES]

START a Boric Acid Transfer Pump: [Step 2 - YES]

PLACE 1/1-APBA1, BA XFER PMP 2 in START.

PLACE 1/1-8104, EMER BORATE VLV in OPEN. [Step 3 - YES]

VERIFY flow on 1-FI-183A, EMER BORATE FLO. [Step 4 - YES]

VERIFY flow on 1-FI-121A, CHRG FLO. [Step 5 - YES]

Examiner Note: The following steps are from EOS-0.1A, Reactor Trip Response.

Examiner Note: Depending on crew response, the RED PATH for Heat Sink will occur at some point early in EOS-0.1A. This condition occurred at Step 3 during validation.

RO CHECK RCS Temperature: [Step 1 - YES]

VERIFY RCS average temperature stable at or trending to 557ºF.




RO/BOP CHECK FW Status: [Step 2 - YES]


CHECK RCS average temperatures < 564ºF. [Step 2.b - NO]

CONTINUE with Step 3. When temperature is less than 564°F PERFORM Steps 2.c & 2.d. [Step 2.b RNO - YES]

BOP VERIFY total AFW flow to SGs – GREATER THAN 460 GPM. [Step 2.d - YES]

RO CHECK PRZR Level Control: [Step 3 - YES]

VERIFY PRZR Level – GREATER THAN 17%. [Step 3.a - YES]

VERIFY Charging – IN SERVICE. [Step 3.b - YES]

VERIFY Letdown – IN SERVICE. [Step 3.c - YES]

VERIFY PRZR Level – TRENDING TO 25%. [Step 3.d - YES]

RO CHECK PRZR Pressure Control: [Step 4 - YES]

VERIFY Pressurizer Pressure – GREATER THAN 1820 PSIG. [Step 4.a - YES]

VERIFY Pressurizer Pressure – STABLE AT OR TRENDING TO 2235 PSIG. [Step 4.b - YES]

RO CHECK Steam Generator Levels: [Step 5 - YES]

VERIFY Steam Generator Level – GREATER THAN 43%. [Step 5.a - NO]

MAINTAIN total feed flow greater than 460 GPM until narrow range level greater than 43% in at least one SG. [Step 5.a RNO - YES]

VERIFY Pressurizer Pressure – STABLE AT OR TRENDING TO 2235 PSIG. [Step 4.b - YES]




RO VERIFY All AC Buses – ENERGIZED BY OFFSITE POWER: [Step 6 - YES]

Examiner Note: FRH-0.1A,

Appendix D Scenario Outline Form ES-D-1 · 2012. 8. 31. · Scenario Event Description NRC Scenario...

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Transcript of Appendix D Scenario Outline Form ES-D-1 · 2012. 8. 31. · Scenario Event Description NRC Scenario...