ML102080297

2010-07 Final Operating Test

ML102080297
Person / Time
Site:	Comanche Peak
Issue date:	07/15/2010
From:	Kelly Clayton NRC Region 4
To:	Luminant Generation Co
References
50-445/10-302, 50-446/10-302 50-445/10-302, 50-446/10-302
Download: ML102080297 (70)

Text

Appendix D Scenario Outline Form ES-D-1 CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Facility:

CPNPP 1 & 2 Scenario No.:

1 Op Test No.:

July 2010 NRC Examiners:

Operators:

Initial Conditions:

100% power MOL - RCS Boron is 910 ppm by Chemistry sample.

Train A Emergency Diesel Generator is OOS for governor repair.

Turnover:

Maintain steady-state 100% power conditions.

Critical Tasks:

Trip Reactor Coolant Pumps Upon Loss of Subcooling.

Manually Initiate Containment Isolation Phase A Upon Failure to Automatically Actuate.

Maintain Core Cooling During a Loss of Cold Leg Recirculation.

Event No.

Malf. No.

Event Type*

Event Description 1

+5 min RP06A I (RO, SRO)

TS (SRO)

Loop 1 N-16 Channel I (1-JI-411A/B) Fails High.

2

+10 min RX18 I (BOP, SRO)

Feed Header Pressure Transmitter (PT-508) Fails High.

3

+15 min CH03 C (BOP, SRO)

Neutron Detector Well Fan (FN-09) Motor Overcurrent.

4

+25 min RX05B I (RO, SRO)

TS (SRO)

Pressurizer Level Channel (LT-460) Fails Low.

5

+30 min RC17A M (RO, BOP, SRO) Large Break Loss of Coolant Accident (60 second ramp).

6

+35 min RP09A RP09B C (BOP)

Phase A Containment Isolation Trains A and B Fail to Automatically Actuate.

7

+35 min RH01B C (BOP)

Residual Heat Removal Pump (1-02) Trip Upon SI Sequencer Start.

8

+50 min RHR15 C (RO)

Containment Sump to Residual Heat Removal Pump (1-01) Suction Isolation Valve (1-8811A) Will Not Open.

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

Scenario Event Description NRC Scenario #1 CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc SCENARIO

SUMMARY

NRC #1 The crew will assume the watch and maintain steady-state conditions per IPO-003A, Power Operations. Train A Emergency Diesel Generator is out of service for governor repair.

The first event is a high failure of Loop 1 N-16 Channel I. Operator actions are per ABN-704, Tc/N-16 Instrumentation Malfunction, and include placing Rod Control in Manual and identifying the failed channel. The SRO will refer to Technical Specifications.

Once Technical Specifications are addressed, a Feedwater Header Pressure Transmitter fails high causing Main Feed Pump speed to decrease resulting in a drop in Feedwater flow. Operator actions are per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 5.0, and require manual Feedwater Header pressure control until repairs are made.

When Feedwater Header pressure is stable, the running Neutron Detector Well Fan will trip. Actions are per ALM-0031A, 1-ALB-3A, Window 2.1 - CNTMT FN MASTER TRIP. The BOP will start the standby Neutron Detector Well Fan per SOP-801A, Containment Ventilation System, and monitor for proper operation.

The next event is a Pressurizer Level instrument failure. Actions are per ABN-706, Pressurizer Level Instrumentation Malfunction. The RO will take manual control of Pressurizer level or Charging flow to maintain Pressurizer Level on program. When the failed instrument is identified, an alternate controlling channel is selected and Charging flow and Pressurizer Level control will be returned to Automatic.

Letdown flow will be restored per the Job Aid. The SRO will refer to Technical Specifications.

When Letdown flow is restored, a Large Break Loss of Coolant Accident develops inside Containment resulting in generation of signals for Safety Injection, Containment Isolation Phase A and Phase B.

Train B Residual Heat Removal Pump will trip upon pump start and the Train A Containment Sump to RHR Pump Suction Isolation Valve will NOT open. Reactor Coolant Pumps must be manually tripped due to a loss of subcooling and Containment Isolation Phase A must be manually initiated.

The crew enters EOP-0.0A, Reactor Trip or Safety Injection and at Step 14, transitions to EOP-1.0A, Loss of Reactor or Secondary Coolant. While in EOP-1.0A, the crew should recognize and transition to ECA-1.1A, Loss of Emergency Coolant Recirculation, at Step 11. When in ECA-1.1A, the crew will secure Containment Spray Pumps as required per Containment pressure conditions.

The scenario is terminated when Containment Spray flow is reduced in ECA-1.1A or when the conditions of FRZ-0.1A, Response to High Containment Pressure, have been met.

Risk Significance:

Risk important components out of service:

Train A Emergency Diesel Generator Risk significant core damage sequence:

LBLOCA with Loss of Coolant Recirculation Risk significant operator actions:

Manually Trip Reactor Coolant Pumps Manually Initiate Containment Isolation Manually Secure Containment Spray Pumps

Scenario Event Description NRC Scenario #1 CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP Initialize to IC #18 and Event File for NRC Scenario #1.

EVENT TYPE MALF #

DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP Train A EDG OOS PULLOUT RHR15 CNTMT Sump to RHRP 1-01 SUC ISOL VLV Breaker 1-8811A rackout RACKOUT Conditional on Rx Trip RP09A Containment Isolation Train A actuation failure K0 RP09B Containment Isolation Train B actuation failure K0 1

RP06A Loop 1 N-16 Channel I failure 150 K1 2

RX18 Feed Header Press (PT-508) transmitter failure 1500 K2 3

CH03 Neutron Detector Well Fan Motor overcurrent TRIP K3 3

AN3A 02 CB03 1-ALB-3A Window 2.1 Alarm OFF TRIP K9 4

RX05B Pressurizer Level Channel (LT-460) fails low 0%

K4 5

RC17A Large Break Loss of Coolant Accident 5,000 gpm K5 (60 second ramp) 6 RP09A Containment Isolation Train A actuation failure K0 6

RP09B Containment Isolation Train B actuation failure K0 7

RH01B RHR Pump (1-02) trip TRIP K0 upon SI 8

RHR15 CNTMT Sump to RHRP 1-01 SUC ISOL VLV Breaker 1-8811A rackout RACKOUT Conditional on Rx Trip

Scenario Event Description NRC Scenario #1 CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: INITIALIZE to IC #18 and NRC Scenario #1 SETUP file.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE Control Board Tags are hung:

- Red tag Train A Emergency Diesel Generator 1-01 & Breaker DG1 BKR 1EG1.

ENSURE Operator Aid Tags reflect current boron conditions.

ENSURE Control Rods are in AUTO with Bank D at 215 steps.

ENSURE Rod Bank Update (RBU) is performed.

ENSURE Turbine Load Rate set at 10 MWe/minute.

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.

Significant 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 10B-1.8 - DG 1 DISABLED

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

1 Page 5

of 22 Event

Description:

Loop 1 N-16 Channel I Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: When directed, EXECUTE Event 1.

- RP06A, Loop 1 N-16 Channel I (1-JI-411A/B) fails high.

Indications Available:

5C-1.5 - ANY N 16 DEV HI/LO 5C-2.5 - 1 OF 4 OT N 16 HI 5C-3.5 - ANY TAVE DEV HI/LO 5C-2.6 - 1 OF 4 OP N 16 HI 6D-1.10 - ANY TAVE TREF DEV 6D-2.13 - 1 OF 4 OP N 16 ROD STOP & TURB RUNBACK 6D-3.14 - 1 OF 4 OT N 16 ROD STOP & TURB RUNBACK 6D-1.7 - ANY ROD BANK AT LO LIMIT (May come in depending on time rods are in AUTO)

+30 secs RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE Loop 1 N-16 Channel failure at JI-411A/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 Switch in MANUAL.

RO PLACE 1-TS-412T, Tave CHAN DEFEAT in LOOP 1 position.

BOP VERIFY Steam Dump System is NOT actuated and NOT armed.

RO RESTORE TAVE to within 1ºF of TREF.

RO SELECT LOOP 1 on 1/1-JS-411E, N16 Power Channel Defeat.

RO ENSURE a valid N16 channel supplying recorder on 1/1-TS-411E, 1-TR-411 CHAN SELECT.

BOP VERIFY Steam Dump System is NOT armed by OBSERVING PCIP-3.4 alarm not LIT.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

1 Page 6

of 22 Event

Description:

Loop 1 N-16 Channel I Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc

+5 min US EVALUATE Technical Specifications.

LCO 3.3.1.E, Reactor Trip System Instrumentation CONDITION E - One channel INOPERABLE.

ACTION E.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in MODE 3 within 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />.

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 #

1 Event #

2 Page 7

of 22 Event

Description:

Feed Header Pressure Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: When directed, EXECUTE Event 2.

- RX18, Feed Header Pressure (PT-508) transmitter fails high.

Indications Available:

8A-1.8 - SG 1 STM & FW FLO MISMATCH 8A-2.8 - SG 2 STM & FW FLO MISMATCH 8A-3.8 - SG 3 STM & FW FLO MISMATCH 8A-4.8 - SG 4 STM & FW FLO MISMATCH 9A-3.2 - HDP 1 DISCH PRESS HI 9A-7.2 - HDP 2 DISCH PRESS HI 1-PI-508, Feed Header Pressure Indication pegged high

+1 min BOP RESPOND to Annunciator Alarm Procedures.

BOP RECOGNIZE Feed Header Pressure 1-PT-508 transmitter failure.

US DIRECT performance of ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 5.0.

BOP PLACE 1-SK-509A, FWPT Master Speed Controller in MANUAL.

BOP ADJUST 1-SK-509A, FWPT Master Speed Controller to maintain 80 psig to 170 psig between FWP discharge pressure and steam line pressure.

+5 min US INITIATE a Work Request per STA-606.

When manual control of feedwater is attained, or at Lead Examiner discretion, PROCEED to Event 3.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

3 Page 8

of 22 Event

Description:

Neutron Detector Well Fan Motor Overcurrent Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: When directed, EXECUTE Event 3.

- CH03, Neutron Detector Well Fan (1-09) overcurrent trip.

Indications Available:

3A-2.1 - CNTMT FN MASTER TRIP 11A-1.5 - NEUT DET WELL CH WTR RET FLO LO (on Panel CV-01) 1-HS-5435, NEUT DET WELL FN CLR FN 9 & DMPR green DAMPER & FAN and white TRIP lights illuminated

+1 min BOP RESPOND to Annunciator Alarm Procedures.

BOP RECOGNIZE Neutron Detector Well Fan 1-09 tripped.

US DIRECT performance of ALM-0031A, 1-ALB-3A, Window 2.1 - CNTMT FN MASTER TRIP.

Examiner Note: BOP may start fan per assumed operator knowledge prior to referencing the ALM.

BOP If NO fans are in service, START Neutron Detector Well Fan 10 per SOP-801A, Containment Ventilation System.

Examiner Note: The following steps are from SOP-801A, Containment Ventilation System.

BOP DETERMINE Prerequisites in Section 2.2 of SOP-801A are met.

BOP PLACE 1-HS-5440, NEUT DET WELL FN CLR, FN 10 and DMPR in START.

VERIFY Fan 10 Suction Damper OPEN and Fan 9 Suction Damper CLOSED.

BOP VERIFY 1-HS-6079, NEUT DET WELL FN CLR 10 CH WTR RET VLV automatically OPENS on Panel CV-01.

BOP ENSURE 1-HS-6084, CH WTR SPLY ISOL VLV ORC is OPEN.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

3 Page 9

of 22 Event

Description:

Neutron Detector Well Fan Motor Overcurrent Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Examiner Note: The following steps are from Window 2.1 - CNTMT FN MASTER TRIP.

BOP PLACE 1-HS-5435, NEUT DET WELL FN CLR FN 9 & DMPR in PULLOUT/STOP.

Booth Operator: When contacted, WAIT one minute then REPORT Fan 9 Breaker appears to have tripped due to overload.

US DISPATCH an operator to Fan 9 Breaker to determine cause of trip.

US When conditions permit, PERFORM a Containment Entry per STA-620 to determine cause of fan failure.

+5 min US INITIATE a work request per STA-606.

When Neutron Detector Well cooling is restored, 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 10 of 22 Event

Description:

Pressurizer Level Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: When directed, EXECUTE Event 4.

- RX05B, Pressurizer Level Transmitter (LT-460) fails low.

Indications Available:

5B-1.4 - PRZR HTR GRP C CTRL TRBL 5B-3.6 - PRZR LVL LO 6A-3.8 - CVCS HELB PS-5385A 6A-4.8 - CVCS HELB PS-5385 1-LI-460A, PRZR LVL CHAN II indication failed low

+30 secs RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE Letdown isolated and all PRZR Heaters deenergized and DETERMINE Pressurizer Level Channel (LT-460) malfunction.

RO REPORT Pressurizer Level Channel I (LT-460) failed low.

US DIRECT performance of ABN-706, Pressurizer Level Instrumentation Malfunction, Section 2.0.

RO PLACE PZR Level Control or Charging Flow in MANUAL to maintain level on program using one of the following controllers:

1-LK-459, PRZR LVL CTRL 1-FK-121, CCP CHRG FLO CTRL RO TRANSFER 1/1-LS-459D, PZR Level Control Channel Select to an OPERABLE channel.

RO TRANSFER 1/1-LS-459E, 1/1-LR-459 PZR Level Select to an OPERABLE channel.

RO RESTORE Letdown per Job Aid.

OPEN or VERIFY OPEN both Letdown Isolation Valves.

ENSURE 1-PK-131, LTDN HX OUT PRESS CTRL in MANUAL and 30%

(75 gpm) or 50% (120 gpm) DEMAND.

ENSURE 1-TK-130, LTDN HX OUT TEMP CTRL in MANUAL and 50%

DEMAND.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

4 Page 11 of 22 Event

Description:

Pressurizer Level Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc ADJUST Charging to desired flow and MAINTAIN Seal Injection flow between 6 and 13 gpm.

OPEN the desired Orifice Isolation Valves.

ADJUST 1-PK-131, LTDN HX OUT PRESS CTRL to ~310 psig on 1-PI-131, LTDN HX OUT PRESS then PLACE in AUTO.

ADJUST 1-TK-130, LTDN HX OUT TEMP CTRL to obtain ~95ºF on 1-TI-130, LTDN HX OUT TEMP, then place in AUTOMATIC.

RO RESTORE PZR Control Heater Group C.

Examiner Note: AUTO Pressurizer Level Control will not be restored until level is manually returned to program.

RO RESTORE PZR Level Control or Charging Flow Control to AUTO as desired.

RO VERIFY instruments on common instrument line - NORMAL.

DETERMINE PT-0456 and PT-0458 readings are NORMAL.

+10 min US EVALUATE Technical Specifications.

LCO 3.3.1.M, Reactor Trip System Instrumentation.

CONDITION M - One channel inoperable.

ACTION M.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

When Technical Specifications are addressed, or at Lead Examiner 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 12 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Booth Operator: When directed, EXECUTE Events 5, 6, 7, and 8.

- RC17A, Loss of Coolant Accident @ 5,000 gpm on 60 second ramp.

- RP09A, Containment Isolation Train A actuation failure.

- RP09B, Containment Isolation Train B actuation failure.

- RH01B, Residual Heat Removal Pump (1-02) trip.

- RH15, Containment Sump to RHR Pump (1-01) Suction Isolation Valve Breaker 1-8811A trip.

Indications Available:

Numerous Reactor Trip and Safety Injection related alarms

+10 secs RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE PRZR pressure decreasing.

RO/BOP INITIATE a Reactor Trip and Safety Injection.

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

RO VERIFY Reactor Trip:

DETERMINE Reactor trip breakers - OPEN.

DETERMINE Neutron flux - DECREASING.

RO DETERMINE all Control Rod Position Rod Bottom Lights - ON.

BOP VERIFY Turbine Trip:

DETERMINE all HP Turbine Stop Valves - CLOSED.

BOP VERIFY Power to AC Safeguards Buses:

DETERMINE both AC Safeguards Buses - ENERGIZED.

RO DETERMINE both Trains of SI actuated.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 13 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Examiner Note: EOP-0.0A, Attachment 2 steps performed by BOP are identified later in the scenario.

US/BOP INITIATE Proper Safeguards Equipment Operation Per Attachment 2.

RO VERIFY AFW Alignment:

DETERMINE both MDAFW Pumps - RUNNING.

DETERMINE Turbine Driven AFW Pump - RUNNING.

DETERMINE AFW total flow - GREATER THAN 460 GPM.

DETERMINE AFW valve alignment - PROPER ALIGNMENT.

RO VERIFY Containment Spray Not Required:

VERIFY 1-ALB-2B Window 1-8, CS ACT NOT illuminated.

VERIFY 1-ALB-2B Window 4-11, CNTMT ISOL PHASE B ACT NOT illuminated.

VERIFY Containment pressure < 18.0 PSIG.

VERIFY Containment Spray Heat Exchanger Outlet Valves - CLOSED.

VERIFY all Containment Spray Pumps - RUNNING.

Examiner Note: Operators may manually isolate the Main Steam Lines in anticipation of Containment HI-2 pressure (6.2 psig) prior to automatic isolation. If not performed here the isolation will occur later as Containment pressure rises.

RO DETERMINE Main Steam Lines Should Be Isolated:

VERIFY Main Steam Isolation complete:

DETERMINE Main Steam Isolation Valves - CLOSED.

DETERMINE before MSIV Drippot Isolation Valves - CLOSED.

RO CHECK RCS Temperature -

DETERMINE RCS AVERAGE TEMPERATURE less than 557°F.

RO STOP dumping steam.

RO REDUCE total AFW flow to minimize the cooldown:

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 14 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc MAINTAIN a minimum of 460 gpm UNTIL narrow range level greater than 50% in at least one SG.

VERIFY Turbine Driven AFW Pump - STOPPED.

RO CHECK PRZR Valve Status:

DETERMINE PRZR Safeties - CLOSED.

DETERMINE PRZR Spray Valves - CLOSED.

DETERMINE PORVs - CLOSED.

DETERMINE power to both PORV Block Valves - AVAILABLE.

DETERMINE both PORV Block Valves - OPEN.

US/RO CHECK If RCPs Should Be Stopped:

DETERMINE all ECCS Pumps - RUNNING.

DETERMINE RCS subcooling - LESS THAN 25ºF (55 ºF adverse).

CRITICAL TASK STATEMENT Manually Trip Reactor Coolant Pumps due to Loss of Subcooling Prior to Exiting EOP-0.0A.

CRITICAL TASK RO DETERMINE RCS subcooling less than 25ºF (55ºF adverse) and STOP all RCPs.

RO/BOP CHECK if Any Steam Generator Is Faulted:

DETERMINE pressure in all Steam Generators - NORMAL.

RO/BOP CHECK if any Steam Generator is Ruptured:

DETERMINE radiation levels in all Steam Generators - NORMAL.

RO/BOP CHECK if RCS is intact:

DETERMINE Containment pressure, radiation level and sump levels increasing.

+10 min US TRANSITION to EOP-1.0A, Loss of Reactor or Secondary Coolant, Step 1.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 15 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc Examiner Note: These steps are performed by the BOP as required per EOP-0.0A, Attachment

2. EOP-1.0A steps are identified later in the scenario.

BOP VERIFY SSW Alignment:

DETERMINE both SSW Pumps - RUNNING.

VERIFY Train B Diesel Generator Cooler SSW return flow.

BOP VERIFY Safety Injection Pumps - RUNNING.

DETERMINE both Safety Injection Pumps - RUNNING.

CRITICAL TASK STATEMENT Manually Initiate Containment Isolation Phase A due to Failure to Automatically Actuate Prior to Exiting EOP-0.0A.

CRITICAL TASK BOP Manually INITIATE both Trains of Containment Isolation Phase A.

PLACE 1/1-CIPAA1 CNTMT ISOL - PHASE A CONT VENT ISOL Switch in ACT position.

BOP VERIFY Containment Isolation Phase A.

BOP VERIFY Containment Ventilation Isolation.

BOP VERIFY both CCW Pumps - RUNNING.

DETERMINE both CCW Pumps - RUNNING.

BOP VERIFY both RHR Pumps - RUNNING.

DETERMINE Train A RHR Pump - RUNNING.

DETERMINE Train B RHR Pump - TRIPPED.

BOP VERIFY Proper CVCS Alignment:

DETERMINE both CCPs - RUNNING.

VERIFY Letdown Relief Valve isolation:

DETERMINE Letdown Orifice Isolation Valves - CLOSED.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 16 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc DETERMINE Letdown Isolation Valves - CLOSED.

BOP VERIFY ECCS flow:

VERIFY CCP SI flow indicated.

VERIFY RCS pressure < 1800 PSIG.

VERIFY SIP discharge flow indicators.

DETERMINE RCS pressure < 425 PSIG.

RHR to Cold Leg Injection Flow Indicators - CHECK FOR FLOW DETERMINE Train A RHR flow - INDICATED.

BOP VERIFY Feedwater Isolation Complete:

VERIFY Feedwater Isolation Valves - CLOSED.

VERIFY Feedwater Isolation Bypass Valves - CLOSED.

VERIFY Feedwater Bypass Control Valves - CLOSED.

VERIFY Feedwater Control Valves - CLOSED.

BOP VERIFY Train B Diesel Generator - RUNNING.

BOP VERIFY Monitor Lights For SI Load Shedding illuminated.

BOP VERIFY Proper SI alignment per MLB light indication.

BOP VERIFY Components Properly Aligned per Table 1.

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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 17 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc 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-2398A SG 2 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2399A SG 3 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2400A SG 4 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 CV-01 X-HS-6188 PRI PLT SPLY FN 18 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6195 PRI PLT SPLY FN 19 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6202 PRI PLT SPLY FN 20 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6209 PRI PLT SPLY FN 21 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6216 PRI PLT SPLY FN 22 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6223 PRI PLT SPLY FN 23 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6230 PRI PLT SPLY FN 24 & INTK DMPR 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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 18 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc 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 CV-03 X-HS-5733 SFP EXH FN 34 STOPPED/DEENERGIZED CV-03 X-HS-5727 SFP EXH FN 35 STOPPED/DEENERGIZED CV-03 X-HS-5729 SFP EXH FN 36 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.

Examiner Note: EOP-1.0A, Loss of Reactor or Secondary Coolant, steps begin here. Steps for FRZ-0.1A, Response to High Containment Pressure, are identified later in the scenario.

US CHECK if RCPs Should Be Stopped:

DETERMINE all RCPs STOPPED.

US CHECK if Any Steam Generator Is Faulted:

DETERMINE pressure in all Steam Generators - NORMAL.

US CHECK Intact Steam Generator Levels:

DETERMINE Narrow range levels - GREATER THAN 50%.

CONTROL AFW flow to maintain NR level between 50% and 60%.

US CHECK Secondary Radiation NORMAL:

DETERMINE no Steam Generator tubes ruptured.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 19 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc US CHECK PRZR PORVs and Block Valves:

DETERMINE power to both PORV Block Valves - AVAILABLE.

DETERMINE PORVs - CLOSED.

DETERMINE both PORV Block Valves - OPEN.

US/RO DETERMINE ECCS Flow Should NOT Be Reduced:

VERIFY Secondary heat sink:

DETERMINE total AFW flow to intact SGs > 460 GPM.

DETERMINE Narrow range level in all SGs > 50%.

DETERMINE RCS subcooling < 25ºF (55ºF adverse).

RO/BOP RESET ESF Actuation Signals.

RO/BOP PLACE Train B Diesel Generator EMERG STOP/START Handswitch in START.

RO/BOP RESET SI.

RO/BOP RESET SI Sequencers.

RO/BOP RESET Containment Isolation Phase A and Phase B.

RO/BOP RESET Containment Spray Signal.

US CHECK If RHR Pumps Should Be Stopped.

DETERMINE RCS pressure > 325 PSIG (425 PSIG adverse).

STOP RHR Pump 1-01and PLACE in standby.

RESET RHR Auto Switchover.

US CHECK RCS and SG Pressures.

DETERMINE RCS pressure STABLE or DECREASING, AND

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 20 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc DETERMINE all SG pressures are STABLE or INCREASING.

RO/BOP DETERMINE AC Safeguards Buses ENERGIZED by Offsite Power.

PLACE Train B DG EMERG STOP/START Handswitch in STOP.

Booth Operator: When contacted, WAIT five minutes then REPORT Train B RHR Pump has Phase B 50/51 overcurrent relay flags dropped and the motor has an acrid odor.

Booth Operator: If contacted to investigate breaker status for 1/1-8811A, WAIT three minutes and REPORT breaker is tripped free and will not reset. If dispatched to locally open 1/1-8811A, REPORT that RP will not allow entry into area due to high radiation levels.

US INITIATE Evaluation of Plant Status:

DETERMINE Cold Leg Recirculation Capability - NOT AVAILABLE.

DETERMINE 1-8811A, Containment Sump to Train A RHR Pump Suction Isolation Valve will NOT open.

DETERMINE Train B RHR Pump - TRIPPED.

+20 min US TRANSITION to ECA-1.1A, Loss of Emergency Coolant Recirculation, Step 1.

Examiner Note: ECA-1.1A, Loss of Emergency Coolant Recirculation, steps begin here.

US CHECK If Emergency Coolant Recirculation Equipment - AVAILABLE PER ATTACHMENT 2.

ATTEMPT to restore at least one train.

US VERIFY Train B EDG - STOPPED.

RO/BOP VERIFY SI reset.

RO/BOP VERIFY SI Sequencers reset.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 21 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc RO/BOP VERIFY Containment Isolation Phase A and Phase B reset.

RO/BOP VERIFY Containment Spray Signal reset.

RO/BOP RESET RHR Auto Switchover.

US NOTIFY Plant Staff to DETERMINE if Containment Fan Coolers should be started.

US CHECK RWST Level - GREATER THAN RWST EMPTY.

US/RO DETERMINE Containment Spray Requirements:

DETERMINE Containment Spray Pump suction - ALIGNED TO RWST.

DETERMINE Containment pressure - less than 18 PSIG AND LOWERING.

DETERMINE zero (0) Containment Spray Pumps REQUIRED from Table 1.

CRITICAL TASK STATEMENT Determine Cold Leg Recirculation Capability Does NOT Exist and Maintain Core Cooling Prior to Refueling Water Storage Tank Level Reaching 0%.

CRITICAL TASK BOP STOP all Containment Spray Pumps.

Examiner Note: FRZ-0.1A, Response to High Containment Pressure, steps begin here.

US ENTER FRZ-0.1A, Response to High Containment Pressure, due to an ORANGE Path.

US DETERMINE Containment Pressure NOT GREATER THAN 50 PSIG and alignment was NOT verified in EOP-0.0A.

RO/BOP VERIFY Containment Isolation Phase A - APPROPRIATE MLB LIGHT INDICATION.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

1 Event #

5, 6, 7, & 8 Page 22 of 22 Event

Description:

Large Break LOCA / Phase A Containment Isolation Failure / Loss of Emergency Coolant Recirculation Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #1 Rev c.doc RO/BOP VERIFY Containment Ventilation Isolation - APPROPRIATE MLB LIGHT INDICATION.

RO/BOP CHECK If Containment Spray Is Required:

DETERMINE Containment pressure was >18.0 PSIG.

DETERMINE Containment Spray AND Phase B Actuation - INITIATED.

DETERMINE all RCPs - STOPPED.

DETERMINE ECA-1.1A, Loss of Emergency Coolant Recirculation, is in effect.

OPERATE Containment Spray per ECA-1.1A, Loss of Emergency Coolant Recirculation.

RO/BOP VERIFY Main Steam Isolation Valves - CLOSED.

US DETERMINE Feed Flow Should NOT Be Isolated To Any SG.

+30 min US RETURN To Procedure And Step In Effect.

When Containment Spray flow is secured or the actions of FRZ are complete, TERMINATE the scenario.

Appendix D Scenario Outline Form ES-D-1 CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Facility:

CPNPP 1 & 2 Scenario No.:

2 Op Test No.:

July 2010 NRC Examiners:

Operators:

Initial Conditions:

72% power MOL - RCS Boron is 916 ppm by Chemistry sample.

Train A Emergency Diesel Generator is OOS for governor repair.

Turnover:

Maintaining 72% power per Load Controller direction. Rod Control in AUTO.

Critical Tasks:

Emergency Borate Required for Two Stuck Control Rods.

Perform Actions to Identify and Isolate Faulted Steam Generator.

Perform Actions to Initiate Feed Isolation to Faulted Steam Generator.

Event No.

Malf. No.

Event Type*

Event Description 1

+10 min RX09A I (RO, BOP, SRO)

TS (SRO)

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

2

+20 min CC02A CC03A C (BOP, SRO)

TS (SRO)

Train A Component Cooling Water Pump (1-01) Trip.

Train B Component Cooling Water Pump (1-02) Auto Start Failure.

3

+30 min RX08A I (RO, SRO)

TS (SRO)

Pressurizer Pressure Channel (PT-455) Fails Low.

4

+40 min RX04A I (BOP, SRO)

TS (SRO)

Steam Generator (1-01) Level Channel (LT-551) Fails High.

5

+45 min MS03A M (RO, BOP, SRO) Steam Generator (1-01) Steam Line Break Outside Containment Before Main Steam Isolation Valve (300 second ramp).

6

+45 min RD04K6 RD04K8 C (RO)

Two Control Rods Fail to Insert Upon Reactor Trip.

Emergency Boration Required.

7

+55 min SI04D C (BOP)

Safety Injection Pump (1-02) Fails to Start.

8

+55 min FW38 A/B/C/D C (BOP)

Feed Line Isolation Valves (HV-2134 to HV-2137) Fail to Close.

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

Scenario Event Description NRC Scenario #2 CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc SCENARIO

SUMMARY

NRC #2 The crew will assume the watch at 72% power with no scheduled activities per IPO-003A, Power Operations. The Grid Controller has requested that power remain at this level due to transmission line overload until further notice. Train A Emergency Diesel Generator is out of service for governor repair.

The first 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 plant conditions are stable, the Train A Component Cooling Water (CCW) Pump will trip and the Train B CCW Pump will fail to automatically start. The crew will respond per ABN-502, Component Cooling Water System Malfunctions, Section 2.0, and manually start Train B CCW Pump and perform equipment adjustments as required by procedure. The SRO will refer to Technical Specifications.

When ABN-502 actions are complete, a Pressurizer Pressure Channel will fail low. Response is per ABN-705, Pressurizer Pressure Malfunction, Section 2.0, to ensure Pressurizer Heaters are controlled and Power Operated Relief Valves remain closed. The SRO will refer to Technical Specifications.

Once systems are stable, a Steam Generator Level Transmitter fails high. ABN-710, Steam Generator Level Instrumentation Malfunction is referenced and the BOP takes Manual control of the Feedwater Control Valve to prevent a Unit trip on low Steam Generator level. Once identified, an Alternate Channel is selected and Automatic control restored. The SRO will refer to Technical Specifications.

When Technical Specifications have been addressed, a Main Steam Line Break Outside Containment before the Main Steam Isolation Valve will ramp in over five minutes on Steam Generator 1-01. With lowering Pressurizer pressure and Reactor Coolant System temperature, the Unit Supervisor will direct a Reactor and Turbine Trip.

The crew will enter EOP-0.0A, Reactor Trip or Safety Injection, and then transition to EOP-2.0A, Faulted Steam Generator Isolation, at Step 12. While performing the actions of EOP-0.0A, the RO will be required to manually initiate an Emergency Boration due to two stuck Control Rods and the BOP will start a Safety Injection Pump and manually close the Feed Line Isolation Valves while in Attachment 2.

Once the faulted Steam Generator is isolated, the Unit Supervisor will transition to EOS-1.1A, Safety Injection Termination. The scenario is terminated after EOS-1.1A, Safety Injection Termination, is entered and the actions to secure Safety Injection flow are performed.

Risk Significance:

Risk important components out of service:

Train A Emergency Diesel Generator Failure of risk important system prior to trip:

Train A Component Cooling Water System Risk significant core damage sequence:

Main Steam Line Break Outside Containment Risk significant operator actions:

Emergency Borate Due to Two Stuck Rods Isolate Faulted Steam Generator Isolate Feedwater to Faulted SG

Scenario Event Description NRC Scenario #2 CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP Initialize to IC #50 and Event File for NRC Scenario #2. Rods in AUTO.

EVENT TYPE MALF #

DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP Train A EDG OOS PULLOUT RD04K6 Control Rod fails to insert 12 Steps K0 RD04K8 Control Rod fails to insert 228 Steps K0 COND Feed Line Isolation Valves close with handswitch DIFWHS 2134.Value=0 DMF FW 38A DIFWHS 2135.Value=0 DMF FW 38B DIFWHS 2136.Value=0 DMF FW 38C DIFWHS 2137.Value=0 DMF FW 38D CLOSE K0 SI04D Safety Injection Pump 1-02 auto start failure K0 1

RX09A Turbine 1st Stage PT-505 failure 0%

K1 2

CC02A Train A CCW Pump 1-01 trip TRIP K2 2

CC03A Train B CCW Pump 1-02 auto start failure START FAILURE K2 3

RX08A Pressurizer Pressure Channel PT-455 failure 1700 psig K3 4

RX04A SG 1-01 Level Transmitter LT-551 fails high 100%

K4 5

MS03A Steam Generator 1-01 Main Steam Line Break Outside Containment before MSIV 9.5 ft2 K5 (300 sec ramp) 6 RD04K6 Control Rod fails to insert 12 Steps K0 6

RD04K8 Control Rod fails to insert 228 Steps K0 7

SI04D Safety Injection Pump 1-02 auto start failure K0 8

FW38 A/B/C/D Feed Line Isolation Valves fail to close OPEN K0

Scenario Event Description NRC Scenario #2 CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: INITIALIZE to IC #50 and NRC Scenario #2 SETUP file.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

ENSURE Control Board Tags are hung:

- Red tag Train A Emergency Diesel Generator 1-01& Breaker DG1 BKR 1EG1.

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 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 at 179 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 10B-1.8 - DG 1 DISABLED

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

1 Page 5

of 23 Event

Description:

Main Turbine Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: When directed, EXECUTE Event 1.

- RX09A, Main Turbine 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 DETERMINE Control Rods INSERTING in AUTO and PLACE 1/1-RBSS Control Rod Bank Select Switch in MANUAL.

BOP DETERMINE Steam Dumps - CLOSED with 100% DEMAND.

OBSERVE 1-UI-500, STM DMP DEMAND indicates 100% DEMAND.

BOP PLACE at least one (1) Steam Dump Interlock Switch in OFF.

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

BOP ENSURE 1-PK-507, STM DMP PRESS CTRL is in MANUAL.

BOP MATCH 1-PK-507, STM DUMP PRESS CTRL demand to current Steam Dump Valve position.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

1 Page 6

of 23 Event

Description:

Main Turbine Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc BOP VERIFY window PCIP-1.4, CNDSR AVAIL STM DMP ARMED C-9 is ON.

BOP PLACE 43/1-SD, STM DMP MODE SELECT in STM PRESS.

BOP ENSURE both STM DMP INTLK SELECT switches are ON.

US DIRECT transfer of 1-PS-505Z, Turbine Impulse Pressure Channel Select to PT-506.

RO PLACE PT-506, Turbine Impulse Pressure Channel II in service.

Examiner Note: Crew should hold reactivity brief to establish plan for restoring rods to pre-event position (179 steps on Control Bank D).

RO ENSURE TAVE within 1ºF of TREF then PLACE 1/1-RBSS Control Rod Bank Select Switch in AUTO.

Examiner Note: Performing the next step meets the REQUIRED ACTION for Technical Specification LCO 3.3.1, Table 3.3.1-1. A Procedure Enhancement will be submitted for ABN-709 specifically identifying Technical Specifications should be referenced to ensure compliance with the LCO.

US Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, VERIFY PCIP Window 4.6, TURB 10% PWR P-13, in proper state for existing plant conditions (DARK).

US EVALUATE Technical Specifications.

LCO 3.3.1.T, Reactor Trip System Instrumentation.

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.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

1 Page 7

of 23 Event

Description:

Main Turbine Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc US VERIFY PCIP Window 1.3, AMSAC BLK TURB < 40% PWR C-20 (LIT).

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

+10 min US INITIATE a Work Request per STA-606.

When the Turbine impulse pressure transmitter actions are addressed, or at Lead Evaluators discretion, PROCEED to Event 2.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

2 Page 8

of 23 Event

Description:

Train A CCW Pump Trip / Train B CCW Pump Start Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: When directed, EXECUTE Event 2.

- CC02A, Train A CCW Pump (1-01) trip.

- CC03A, Train B CCW Pump (1-02) start failure.

Indications Available:

3C-1.6 - CSP 1 & 3 SEAL CLR CCW RET FLO LO 3C-1.8 - SFP HX 1 CCW RET FLO LO 3C-1.14 - RCDT HX CCW RET FLO LO 3C-1.16 - SEAL WTR HX CCW RET FLO LO 3C-2.3 - CCWP 1 / 2 OVLD TRIP 3C-2.6 - CSP 2 & 4 SEAL CLR CCW RET FLO LO 3C-2.12 - ANY RCP MTR CLR CCW RET FLO LO 3C-3.3 - CCW TRAIN B SFGD LOOP PRESS LO 3C-3.11 - ANY RCP THBR CLR CCW RET FLO LO 3C-3.12 - ANY RCP UP BRG L/O CLR CCW RET FLO LO 3C-4.7 - RHRP 2 SEAL CLR CCW RET FLO LO 3C-4.12 - ANY RCP LOW BRG L/O CLR CCW RET FLO LO 3C-4.13 - XS LTDN HX CCW RET FLO LO 4A-1.7 - SFTY CH WTR TR A/B TRBL/TRIP 01-1.10 - SSW TO CCW TRN A HX P LO 01-2.10 - SSW TO CCW TRN B HX P LO

+1 min BOP RESPOND to Annunciator Alarm Procedures.

BOP RECOGNIZE Train A CCW Pump trip with failure of Train B CCW Pump to start.

US DIRECT implementation of ABN-502, Component Cooling Water System Malfunctions, Section 2.0.

BOP DETERMINE Train B CCW Pump did NOT auto start and START CCW Pump 1-02.

BOP VERIFY Train B Station Service Water Pump running.

RO/BOP VERIFY Train B Safety Chiller Recirc Pump 1-06 is running.

BOP VERIFY CCW heat exchanger outlet flow < 17,500 gpm per heat exchanger.

1-FI-4536A, CCW Heat Exchanger #1 Outlet Flow 1-FI-4537A, CCW Heat Exchanger #2 Outlet Flow

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

2 Page 9

of 23 Event

Description:

Train A CCW Pump Trip / Train B CCW Pump Start Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc BOP STOP equipment on Train A as necessary for plant conditions.

BOP VERIFY CCW Heat Exchanger outlet temperature did NOT exceed 122ºF with pump running.

US EVALUATE Technical Specifications.

LCO 3.7.7.A, Component Cooling Water System.

CONDITION A - One CCW train inoperable.

ACTION A.1 - Restore CCW train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

+10 min US INITIATE a Work Request per STA-606.

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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

3 Page 10 of 23 Event

Description:

Pressurizer Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: When directed, EXECUTE Event 3.

- RX08A, Pressurizer Pressure Channel (PT-455) fails low.

Indications Available:

5B-3.4 - PRZR 1 OF 4 PRESS LO 5B-4.4 - PRZR 1 OF 4 SI PRESS LO 5C-3.3 - PRZR PRESS LO BACKUP HTRS ON 5C-3.6 - CORE CLG MICRO PROC TRN A SYS FAIL

+1 min RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE PRZR pressure rising with PRZR heaters ON.

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

Examiner Note: The next three (3) steps are Initial Operator Actions.

RO VERIFY PORV closed.

RO PLACE 1-PK-455A, PRZR Master Pressure Control in MANUAL.

RO ADJUST 1-PK-455A for current RCS pressure.

RO TRANSFER to an alternate controlling channel, 1/1-PS-455F, PRZR Press Control Channel Select.

RO PLACE 1-PK-455A in AUTO.

RO VERIFY automatic control restoring Pressurizer pressure to 2235 psig.

RO ENSURE a valid channel selected to recorder 1/1-PS-455G, 1-PR-455 PRZR Pressure Select.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

3 Page 11 of 23 Event

Description:

Pressurizer Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Examiner Note: Performing the next step meets the REQUIRED ACTION for Technical Specification LCO 3.3.2, Table 3.3.2-1. A Procedure Enhancement will be submitted for ABN-705 specifically identifying Technical Specifications should be referenced to ensure compliance with the LCO.

US Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, VERIFY PCIP window 2.6, PRZR PRESS SI BLK PERM P-11 in required state for current pressure (DARK).

US/RO VERIFY other instruments on common instrument line - NORMAL.

DETERMINE LT-459, LT-459F, and PT-455F readings are NORMAL.

+10 min US EVALUATE Technical Specifications.

LCO 3.3.1.E, Reactor Trip System Instrumentation.

• CONDITION E - One channel inoperable.

• ACTION E.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LCO 3.3.1.M, Reactor Trip System Instrumentation.

• CONDITION M - One channel inoperable.

• ACTION M.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LCO 3.3.2.L, ESFAS Instrumentation.

CONDITION L - One channel inoperable.

ACTION L.1-Verify interlock in required state for existing condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

LCO 3.3.2.D, ESFAS Instrumentation.

• CONDITION D - One channel inoperable.

• ACTION D.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

When Technical Specifications are addressed, or at Lead Examiner discretion, PROCEED to Event 4.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

4 Page 12 of 23 Event

Description:

Steam Generator Level Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: When directed, EXECUTE Event 4.

- RX04A, SG 1-01 Level Transmitter (LT-551) fails high.

Indications Available:

8A-1.12 - SG 1 LVL DEV 8A-1.8 - SG 1 STM & FW FLO MISMATCH 9A-3.2 - HDP 1 DISCH PRESS HI 9A-7.2 - HDP 2 DISCH PRESS HI 1-LI-551, SG 1 LVL (NR) CHAN I indication fails high 1-FK-510, SG 1 FW FLO CTRL VLV starts to close

+30 sec BOP REFER to Annunciator Alarm Procedures.

BOP RECOGNIZE Steam Generator 1-01 Level Transmitter (LT-551) failed high.

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

BOP DETERMINE controlling level channel has failed.

BOP PLACE 1-FK-510, SG 1 FW FLO CTRL in MANUAL and OPEN valve to CONTROL level.

BOP VERIFY instruments on common instrument line - NORMAL.

DETERMINE LT-501 reading is NORMAL.

RO VERIFY all HI-HI level bistable Windows on TSLB-3 for SG 1 DARK.

DETERMINE Windows 2.2, 3.2, and 4.2 are DARK.

BOP VERIFY automatic SG 1-01 level control AVAILABLE and DESIRED:

DETERMINE Alternate Level Control Channel responding normally.

BOP SELECT an alternate channel.

PLACE 1-LS-519C, SG 1 LVL CHAN SELECT to 1-LY-519 position.

BOP VERIFY Steam Generator 1-01 ready for AUTO Level Control:

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

4 Page 13 of 23 Event

Description:

Steam Generator Level Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc DETERMINE Feedwater and Steam Flows matched.

VERIFY Steam Generator level stable at program.

BOP PLACE 1-FK-510, SG 1 FW FLO CTRL in AUTO and MONITOR operation.

US INITIATE repairs per STA-606.

+10 min US EVALUATE Technical Specifications.

LCO 3.3.1.E, Reactor Trip System Instrumentation.

• CONDITION E - One channel inoperable.

• ACTION E.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LCO 3.3.2.D, ESFAS Instrumentation.

• CONDITION D - One channel inoperable.

• ACTION D.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

LCO 3.3.2.I, ESFAS Instrumentation.

• CONDITION I - One channel inoperable.

• ACTION I.1 - Place channel in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

When Feedwater Control is restored, or at Lead Evaluators discretion, PROCEED to Events 5, 6, 7, and 8.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 14 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc Booth Operator: When directed, EXECUTE Events 5, 6, 7, and 8.

- MS03A, Steam Generator (1-01) Steam Line Break outside Containment.

- RD04K6 @ 12 steps, Control Rod fails to insert upon Reactor trip.

- RD04K8 @ 228 steps, Control Rod fails to insert upon Reactor trip.

- SI04D, Safety Injection Pump (1-02) fails to auto start.

- FW38A/B/C/D, Feed Line Isolation Valves (HV-2134 to HV-2137) fail to close.

Indications Available:

6C-3.7 - MSL PRESS LO SI ACT Numerous plant trip alarms

+30 secs RO/BOP RECOGNIZE lowering RCS temperature and pressure.

RO/BOP DETERMINE Reactor Trip required and manually TRIP Reactor.

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

RO VERIFY Reactor Trip:

DETERMINE Reactor trip breakers - OPEN.

DETERMINE Neutron flux - DECREASING.

RO DETERMINE all Control Rod Position Rod Bottom Lights - NOT LIT (two (2) stuck rods).

BOP VERIFY Turbine Trip:

DETERMINE all HP Turbine Stop Valves - CLOSED.

BOP VERIFY Power to AC Safeguards Buses:

DETERMINE both AC Safeguards Buses - ENERGIZED.

RO DETERMINE both Trains of Safety Injection actuated.

CRITICAL TASK STATEMENT Initiate Emergency Boration with Two or More Stuck Control Rods Prior to Exiting EOP-0.0.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 15 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc CRITICAL TASK RO INITIATE Emergency Boration of 3600 gallons of Boric Acid.

ENSURE a Charging Pump is RUNNING.

START either Boric Acid Transfer Pump.

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

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

PLACE 1/1-8104, EMER BORATE VLV in OPEN.

VERIFY flow on 1-FI-183A, EMER BORATE FLO.

VERIFY flow on 1-FI-121A, CHRG FLO.

Examiner Note: EOP-0.0A, Attachment 2, steps performed by BOP are identified later in the scenario. Ensure CRITICAL TASK listed is performed during Attachment 2.

US/BOP INITIATE Proper Safeguards Equipment Operation Per Attachment 2.

RO VERIFY AFW Alignment:

DETERMINE both MDAFW Pumps - RUNNING.

PLACE Turbine Driven AFW Pump in PULL-OUT per Foldout Page.

CONTROL AFW Flow as follows:

CONTROL AFW flow as necessary to maintain narrow range level >

43% in any SG or total AFW flow > 460 gpm per Foldout Page.

STOP AFW flow to Faulted SG 1-01 per Foldout Page.

MAINTAIN proper AFW valve alignment.

RO VERIFY Containment Spray Not Required:

VERIFY 1-ALB-2B Window 1-8, CS ACT NOT illuminated.

VERIFY 1-ALB-2B Window 4-11, CNTMT ISOL PHASE B ACT NOT illuminated.

VERIFY Containment pressure remained < 18.0 PSIG.

VERIFY Containment Spray Heat Exchanger Outlet Valves - CLOSED.

VERIFY all Containment Spray Pumps - RUNNING.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 16 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc RO DETERMINE Main Steam Lines Should be Isolated and PERFORM the following:

VERIFY Main Steam Isolation Complete:

DETERMINE Main Steam Isolation Valves - CLOSED.

DETERMINE Before MSIV Drip Pot Isolation Valves - CLOSED.

RO CHECK RCS Temperature -

DETERMINE RCS Average Temperature less than 557ºF.

RO VERIFY NOT dumping steam.

RO REDUCE total AFW flow to minimize the cooldown:

MAINTAIN a minimum of 460 gpm UNTIL narrow range level greater than 43% in at least one SG.

If necessary, STOP Turbine Driven AFW Pump.

RO CHECK PRZR Valve Status:

VERIFY PRZR Safeties - CLOSED.

VERIFY Normal PRZR Spray Valves - CLOSED.

VERIFY PORVs - CLOSED.

VERIFY Power to at least one Block Valve - AVAILABLE.

VERIFY Block Valves - AT LEAST ONE OPEN.

RO CHECK if RCPs Should Be Stopped:

DETERMINE all ECCS Pumps - RUNNING.

DETERMINE RCS subcooling - GREATER THAN 25ºF.

Continue RUNNING Reactor Coolant Pumps.

RO CHECK If Any SG Is Faulted:

DETERMINE SG 1-01 completely DEPRESSURIZED.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 17 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc US TRANSITION to EOP 2.0A, Faulted Steam Generator Isolation, Step 1.

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

2. EOP-2.0A steps are identified later in the scenario.

BOP VERIFY SSW Alignment:

VERIFY both SSW Pumps - RUNNING.

VERIFY Train B Diesel Generator Cooler SSW return flow.

BOP VERIFY Safety Injection Pumps - RUNNING.

DETERMINE Safety Injection Pump 1-02 failed to start and MANUALLY START Safety Injection Pump 1-02.

BOP VERIFY Containment Isolation Phase A.

BOP VERIFY Containment Ventilation Isolation.

BOP VERIFY CCW Pump 1 RUNNING.

BOP VERIFY both RHR Pumps - RUNNING.

BOP VERIFY Proper CVCS Alignment:

DETERMINE both CCPs - RUNNING.

VERIFY Letdown Relief Valve isolation:

DETERMINE Letdown Orifice Isolation Valves - CLOSED.

DETERMINE Letdown Isolation Valves - CLOSED.

BOP VERIFY ECCS flow:

VERIFY CCP SI flow indicated.

VERIFY RCS pressure < 1700 PSIG (1800 PSIG adverse).

VERIFY SIP discharge flow indicators.

DETERMINE RCS pressure > 325 PSIG (425 PSIG adverse).

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 18 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc CRITICAL TASK STATEMENT Isolate Feedwater Flow to a Faulted Steam Generator Prior to Exiting From EOP-2.0.

CRITICAL TASK BOP DETERMINE Feedwater Isolation Valves NOT closed and manually CLOSE Valves to all Steam Generators.

HV-2134 - CLOSED.

HV-2135 - CLOSED.

HV-2136 - CLOSED.

HV-2137 - CLOSED.

BOP VERIFY Feedwater Isolation Complete:

VERIFY Feedwater Isolation Valves - CLOSED.

VERIFY Feedwater Isolation Bypass Valves - CLOSED.

VERIFY Feedwater Bypass Control Valves - CLOSED.

VERIFY Feedwater Control Valves - CLOSED.

BOP VERIFY Train B Diesel Generator - RUNNING.

BOP VERIFY Monitor Lights For SI Load Shedding illuminated.

BOP VERIFY Proper SI alignment per MLB light indication.

BOP VERIFY Components Properly Aligned per Table 1.

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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 19 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc 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-2398A SG 2 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2399A SG 3 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2400A SG 4 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 CV-01 X-HS-6188 PRI PLT SPLY FN 18 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6195 PRI PLT SPLY FN 19 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6202 PRI PLT SPLY FN 20 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6209 PRI PLT SPLY FN 21 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6216 PRI PLT SPLY FN 22 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6223 PRI PLT SPLY FN 23 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6230 PRI PLT SPLY FN 24 & INTK DMPR 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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 20 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc 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 CV-03 X-HS-5733 SFP EXH FN 34 STOPPED/DEENERGIZED CV-03 X-HS-5727 SFP EXH FN 35 STOPPED/DEENERGIZED CV-03 X-HS-5729 SFP EXH FN 36 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.

Examiner Note: EOP-2.0A, Faulted Steam Generator Isolation steps begin here.

+15 min US/RO CHECK Main Steam line Isolation Valves - CLOSED.

US/RO CHECK at Least One Steam Generator Pressure - STABLE OR INCREASING.

US/RO IDENTIFY Faulted Steam Generator 1-01.

CRITICAL TASK STATEMENT Perform Actions to Identify and Isolate Faulted Steam Generator Prior to exiting EOP-2.0A.

CRITICAL TASK RO/BOP ISOLATE Faulted Steam Generator 1-01.

ISOLATE Main Feed Line to Steam Generator 1-01.

ISOLATE AFW flow to Steam Generator 1-01.

ISOLATE Blowdown and Sample Lines to Steam Generator 1-01.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 21 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc ENSURE 1-HS-2452-2, AFWPT STM SPLY VLV MSL 1 in PULL OUT.

ENSURE Steam Generator 1-01 Atmospheric Valve - CLOSED.

ENSURE Main Steam Line Drip Pot Isolation Valve - CLOSED.

RO CHECK CST Level - GREATER THAN 10%.

Examiner Note: EOP-2.0A, Attachment 2 actions are performed outside of the Control Room.

US/BOP VERIFY Faulted Steam Generator 1-01 Break Outside Containment.

DIRECT performance of EOP-2.0A, Attachment 2.

US/RO CHECK Secondary Radiation:

REQUEST periodic activity samples of all Steam Generators.

CHECK available Secondary Radiation Monitors - NORMAL.

US/RO CHECK if ECCS Flow to Should Be Reduced:

VERIFY Secondary heat sink:

DETERMINE Total AFW Flow to intact SGs > 460 GPM.

DETERMINE Narrow Range Level in SGs 1-02, 1-03, & 1-04 > 43%.

VERIFY RCS subcooling > 25ºF.

VERIFY RCS pressure - STABLE OR INCREASING.

VERIFY PRZR level > 13%.

US DETERMINE ECCS flow should be reduced and TRANSITION to EOS-1.1A, Safety Injection Termination, Step 1.

Examiner Note: EOS-1.1A, Safety Injection Termination, steps begin here.

Examiner Note: The following six (6) steps are performed per EOS-1.1A, Attachment 1.D.

BOP

[1.D] PLACE Train B Diesel EMER START/STOP Handswitch in START.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 22 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc BOP

[1.D] RESET SI.

BOP

[1.D] RESET SI Sequencers.

BOP

[1.D] RESET Containment Isolation Phase A and B.

BOP

[1.D] RESET Containment Spray Signal.

BOP/RO

[1.D] ESTABLISH Instrument Air and Nitrogen to Containment.

RO STOP one CCP and PLACE in Standby.

US/RO CHECK RCS Pressure - STABLE OR INCREASING.

Examiner Note: The following two (2) steps are performed per EOS-1.1A, Attachment 1.J.

RO

[1.J] ISOLATE CCP Injection Line Flow Path:

VERIFY CCP - SUCTION ALIGNED TO RWST.

ALIGN CCP Miniflow Valves:

OPEN 1/1-8110 and 1/1-8111, CCP Miniflow Valves.

CLOSE 1/1-8511A and 1/1-8511B, CCP Alternate Miniflow Isolation Valves.

PLACE Charging Flow Control Valve in MANUAL and 35% demand.

CLOSE 1/1-8801A and 1/1-8801B, CCP Injection Line Isolation Valves.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

2 Event #

5, 6, 7, & 8 Page 23 of 23 Event

Description:

Steam Line Break Outside Containment Before MSIV / Two Stuck Control Rods / Safety Injection Pump Start Failure / Feed Line Isolation Valve Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #2 Rev d.doc RO

[1.J] ESTABLISH Charging Flow Path:

OPEN 1/1-8105 and 1/1-8106, Charging Line Isolation Valves.

ADJUST Charging Flow Control Valve to establish Charging flow.

ADJUST RCP seal flow to maintain between 6 gpm and 13 gpm.

When EOS-1.1A, Safety Injection Termination, Attachment 1.J is complete, TERMINATE the scenario.

CP 2010-07 Retake Exam (Scenarios Only, No JPMs)

Shift turnover sheets for both scenarios follow this cover sheet.

CPNPP July 2010 NRC Sim Scenario #1 Shift Relief Checklist R.doc UNIT:

1 UNIT SUPERVISOR RELIEF CHECKLIST ((NNRRCC SScceennaarriioo ##11))

PART I TO BE PREPARED BY THE OFF-GOING UNIT SUPERVISOR.

1.0 SHIFT ACTIVITIES:

1.1 Activities Completed This Shift:

None 1.2 Activities In-Progress:

None 1.3 Planned Activities:

None 2.0 PLANT AND EQUIPMENT STATUS:

2.1 Technical Specification Related Equipment Summary:

Train A Emergency Diesel Generator 1-01 out-of-service for Governor repair.

Estimated Return-to-Service time is four (4) hours.

2.2 Non-Technical Specification Equipment Summary:

3.0 GENERAL INFORMATION:

Maintain steady state conditions per IPO-003A, Power Operation.

Diluted 45 gallons three (3) times last shift.

4.0 END OF SHIFT REVIEW:

LOGS - RO/BOP X

LOGS-PEO X

CLOSED eLCOARs ARCHIVED X

OPTS COMPLETED X

DAILY ACTIVITIES LIST X

LCOARs REVIEWED X

PART II TO BE COMPLETED BY THE ON-COMING UNIT SUPERVISOR.

1.0 CRITICAL PARAMETERS:

MODE:

1 REACTOR POWER:

100%

MWE:

1264 RCS TAVE:

585 °F CONTROL ROD POSITION 215 ON BANK D

Cb:

910 ppm RCS PRESS:

2235 psig X

Protected Train - Train A X

Unit 2 is in Mode 1 @ 100% power X

Risk Assessment - GREEN X

BAT CB = 7447 ppm

CPNPP July 2010 NRC Sim Scenario #2 Shift Relief Checklist R.docc UNIT:

1 UNIT SUPERVISOR RELIEF CHECKLIST ((NNRRCC SScceennaarriioo ##22))

PART I TO BE PREPARED BY THE OFF-GOING UNIT SUPERVISOR.

1.0 SHIFT ACTIVITIES:

1.1 Activities Completed This Shift:

None 1.2 Activities In-Progress:

None 1.3 Planned Activities:

None 2.0 PLANT AND EQUIPMENT STATUS:

2.1 Technical Specification Related Equipment Summary:

Train A Emergency Diesel Generator 1-01 out-of-service for Governor repair.

Estimated Return-to-Service time is four (4) hours.

2.2 Non-Technical Specification Equipment Summary:

3.0 GENERAL INFORMATION:

Maintaining 875 MWe in accordance with Load Controller direction per IPO-003A, Power Operation.

Diluted 35 gallons three (3) times last shift.

4.0 END OF SHIFT REVIEW:

LOGS - RO/BOP X

LOGS-PEO X

CLOSED eLCOARs ARCHIVED X

OPTS COMPLETED X

DAILY ACTIVITIES LIST X

LCOARs REVIEWED X

PART II TO BE COMPLETED BY THE ON-COMING UNIT SUPERVISOR.

1.0 CRITICAL PARAMETERS:

MODE:

1 REACTOR POWER:

72%

MWE:

875 RCS TAVE:

577 °F CONTROL ROD POSITION 179 ON BANK D

Cb:

916 ppm RCS PRESS:

2235 psig X

Protected Train - Train A X

Unit 2 is in Mode 1 @ 100% power X

Risk Assessment - GREEN X

BAT CB = 7447 ppm

Appendix D Scenario Outline Form ES-D-1 CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Facility:

CPNPP 1 & 2 Scenario No.: 3 (Spare)

Op Test No.:

July 2010 NRC Examiners:

Operators:

Initial Conditions:

~1X10-8 amps BOL - RCS Boron is 1545 ppm by Chemistry sample.

Steam Dump System in service for RCS Temperature Control.

Turnover:

Raise Power to 2% in preparation for plant startup to 100% power.

Critical Tasks:

Restore Feedwater Flow to any Affected Steam Generator.

Determine Inadvertent Safety Injection & Secure Charging Prior to Pressurizer Overfill.

Determine Loss of Coolant Accident in Progress and Reinitiate Safety Injection.

Event No.

Malf. No.

Event Type*

Event Description 1

+20 min R (RO)

N (BOP, SRO)

Raise Reactor power to 2%.

2

+30 min FW24A C (BOP)

TS (SRO)

Motor Driven Auxiliary Feedwater Pump (1-01) Trip.

3

+35 min RP17D TS (SRO)

Containment Pressure Transmitter (PT-937) Fails High.

4

+40 min MS13B I (BOP, SRO)

Atmospheric Relief Valve (1-02) Fails Open due to Steam Pressure Transmitter (PT-2326) Failure.

5

+45 min RP14A M (RO, BOP, SRO) Spurious Train A Safety Injection Actuation Signal.

6

+45 min CV01B CV01E C (RO)

Centrifugal Charging Pump (1-01) Trip.

Centrifugal Charging Pump (1-02) SI Sequencer Start Failure.

7

+60 min RC17C C (RO)

Loss of Coolant Accident at 1700 gpm Following Isolation of High Head Injection.

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

Scenario Event Description NRC Scenario #3 CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc SCENARIO

SUMMARY

NRC #3 The crew will assume the watch with a Plant Startup in progress and will continue raising power to approximately 2% per IPO-002A, Plant Startup from Hot Standby.

When conditions are stable, Motor Driven Auxiliary Feedwater Pump 1-01 will trip. The crew will refer to ABN-305, Auxiliary Feedwater System Malfunction, Section 3.0, and determine that Steam Generator levels are slowly decreasing and start the Turbine Driven Auxiliary Feedwater Pump. The SRO will refer to Technical Specifications.

When the Steam Generator levels are stable, a Containment Pressure Transmitter will fail high. Crew response will be per ALM-0022A, 1-ALB-2B, Window 3.10, CNTMT 1 OF 4 PRESS HI-3, and include verifying that only one channel is affected. The SRO will refer to Technical Specifications.

When Technical Specifications have been referenced, a Steam Generator Atmospheric Relief Valve (ARV) fails open due to a Main Steam Pressure Transmitter failure. This event is recognized by a Reactor power increase and the ARV Controller indicating 100% demand. The BOP will place the affected Controller in Manual and close the ARV. ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 2.0, will be referenced.

The major event begins with a spurious Train A Safety Injection Actuation Signal. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection, and perform immediate actions including actuation of both Trains of Safety Injection.

When it is determined that a spurious Safety Injection actuation has occurred, the crew will transition to EOS-1.1A, Safety Injection Termination. During the event the Train A Centrifugal Charging Pump will trip and the Train B Centrifugal Charging Pump will fail to auto start on the Safety Injection Sequencer.

When the high head injection alignment is secured in EOS-1.1A, a Small Break Loss of Coolant Accident will occur. This will require re-initiation of Safety Injection flow per the Foldout Page Criteria of EOS-1.1A.

Event termination will occur when the crew has reinitiated Safety Injection and transitioned to EOP-1.0A, Loss of Reactor or Secondary Coolant.

Risk Significance:

Failure of risk important system prior to trip:

Loss of MDAFW Pump SG Atmospheric Relief Valve Failure Risk significant core damage sequence:

Small Break LOCA Following SI Termination Risk significant operator actions:

Initiate Charging Flow upon Safety Injection Secure Charging Prior To Pressurizer Overfill Reinitiate Safety Injection Flow

Scenario Event Description NRC Scenario #3 CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc BOOTH OPERATOR INSTRUCTIONS for SIMULATOR SETUP Initialize to IC #8 and Event File for NRC Scenario #3.

EVENT TYPE MALF #

DESCRIPTION DEMAND VALUE INITIATING PARAMETER SETUP CV01E Centrifugal Charging Pump (1-02) start failure K0 1

N/A Raise Reactor power to 2%

2 FW24A Train A Auxiliary Feedwater Pump (1-01) trip TRIP K2 3

RP17D Containment pressure channel (PT-937) failure 60 psig K3 4

MS13B PT-2326 failure fails ARV 1-02 open 1300 psig K4 5

RP14A Spurious Train A Safety Injection actuation K5 6

CV01B Centrifugal Charging Pump (1-01) trip (NOTE 1)

TRIP K6 6

CV01E Centrifugal Charging Pump (1-02) start failure START FAILURE K0 NOTE 1: Initiate K6 after EOP-0.0A Immediate Actions are complete.

7 RC17C Loss of Coolant Accident (NOTE 2) 1700 gpm K7 NOTE 2: Initiate K7 after RHR Pumps are stopped at Step 13 of EOS-1.1A.

Scenario Event Description NRC Scenario #3 CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: INITIALIZE to IC #8 and NRC Scenario #3 SETUP file.

ENSURE all Simulator Annunciator Alarms are ACTIVE.

VERIFY all Control Board Tags are removed.

ENSURE Control Rods are in MANUAL with Control Rod Bank C @ 215 steps and Bank D @ 100 steps.

ENSURE Rod Bank Update (RBU) is performed.

PLACE Plant Computer, right hand RO and US Computer screens for MODE 2.

ENSURE Operator Aid Tags reflect current boron conditions.

ENSURE all PRZR Heaters energized.

ENSURE Reactivity Briefing Sheet printout provided with Turnover.

ENSURE procedures in progress are on SRO desk:

- COPY of IPO-002A, Plant Startup from Hot Standby, Section 5.4, Increasing Reactor Power to ~2% Following Reactor Startup.

Significant Control Room Annunciators in Alarm:

PCIP-1.1 - SR TRN A RX TRIP BLK PCIP-1.3 - AMSAC BLK TURB <40% PWR C-20 PCIP-1.4 - CNDNSR AVAIL STM DUMP ARMED C-9 PCIP-1.7 - RX 50% PWR TURB TRIP PERM P-9 PCIP-2.1 - SR TRN B RX TRIP BLK PCIP-2.4 - LO TURB PWR ROD WTHDRWL BLK C-5 PCIP-2.5 - SR RX TRIP BLK PERM P-6 PCIP-3.5 - RX & TURB 10% PWR P-7 PCIP-4.5 - RX 48% PWR 3-LOOP FLO PERM P-8 PCIP-4.6 - TURB 10% PWR P-13 6D-1.1 - SR HI VOLT FAIL 6D-1.3 - SR SHTDN FLUX ALM BLK 7B-1.6 - FW FLUSH VLV NOT CLOSE HV-2166 7B-4.8 - FWP A / B RECIRC VLV NOT CLOSED 8A-1.3 - FWPT B TRIP 8A-1.10 - 1 OF 4 TURB STOP VLV CLOSE Numerous 9A Feedwater alarms

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

1 Page 5

of 20 Event

Description:

Power Ascension to ~2%

Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: ENSURE Simulator in RUN when crew is ready to assume the watch.

+1 min US DIRECT performance of IPO-002A, Plant Startup from Hot Standby.

RO ESTABLISH a startup rate of ~ 0.5 DPM to increase Reactor power to ~ 2%.

RO Gradually REDUCE startup rate to ~0.2 DPM as the Intermediate Range channels approach 3x10-6 amps.

RO VERIFY Power Range channels begin to respond when the Intermediate Range Channels are between 3x10-6 amps and 5x10-6 amps.

BOP VERIFY Steam Dump operations maintain RCS temperature at 557ºF and Main Steam pressure at 1092 psig.

RO VERIFY annunciator alarm PCIP-3.6, TAVE LO LO P-12 is DARK.

RO MAINTAIN Reactor power between 2% and 3%.

BOP ADJUST Auxiliary Feedwater flow as necessary to maintain SG levels between 60% and 75%.

When power has been raised to 2 - 3%, or at Lead Examiner discretion, PROCEED to Event 2.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

2 Page 6

of 20 Event

Description:

Motor Driven Auxiliary Feedwater Pump Trip Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: When directed, EXECUTE Event 2.

- FW24A, Motor Driven Auxiliary Feedwater Pump 1-01 trip.

Indications Available:

8B-4.3 - MD AFWP 1 / 2 OVRLOAD / TRIP 1-PI-2453A, MD AFWP 1 DISCH PRESS indication reading 0 psig 1-FI-2456A, MD AFWP 1 DISCH FLO indication reading 0 gpm 1-HS-2450A, MD AFWP 1 white TRIP light lit

+1 min BOP RESPOND to Annunciator Alarm Procedure.

BOP RECOGNIZE Motor Driven Auxiliary Feedwater Pump 1-01 trip.

US DIRECT performance of ABN-305, Auxiliary Feedwater System Malfunction, Section 3.0.

BOP DETERMINE MDAFW Pump 1-02 running and PLACE MDAFW Pump 1-01 in STOP or PULLOUT.

BOP DETERMINE Steam Generator Levels trending down.

CRITICAL TASK STATEMENT Restore Feedwater Flow to Steam Generators 1-01 and 1-02 Prior to Reaching 0% Wide Range Steam Generator Level.

CRITICAL TASK BOP START the TDAFW Pump and FEED Steam Generators 1-01 and 1-02.

OPEN 1-HS-2452-1, AFWPT STM SPLY VLV MSL 4 from SG 1-04.

OPEN 1-HS-2452-2, AFWPT STM SPLY VLV MSL 1 from SG 1-01.

RO/BOP DISPATCH a Plant Equipment Operator to check breaker status of Auxiliary Feedwater Pump 1-01.

1EA1/5/BKR, 1APMD1, Auxiliary Feedwater Pump 1-01 BKR.

Booth Operator: When contacted about the condition of the MDAFW Pump breaker, REPORT an acrid smell and overcurrent flag on Phase B 50/51 relays.

BOP VERIFY MDAFW Pump suction pressure 10 psig.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

2 Page 7

of 20 Event

Description:

Motor Driven Auxiliary Feedwater Pump Trip Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc RO/BOP DISPATCH a Nuclear Equipment Operator to check Auxiliary Feedwater Pump 1-01.

Booth Operator: When contacted about the condition of MDAFW Pump 1-01, REPORT that the pump motor is hot.

+10 min US EVALUATE Technical Specifications.

LCO 3.7.5.B, Auxiliary Feedwater (AFW) System.

CONDITION B - One AFW train inoperable for reasons other than Condition A.

ACTION B.1 - Restore AFW train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

When Technical Specifications are addressed, or at Lead Examiner discretion, PROCEED to Event 3.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

3 Page 8

of 20 Event

Description:

Containment Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: When directed, EXECUTE Event 3.

- RP17D, Containment Pressure Transmitter (PT-937) fails high.

Indications Available:

2B-3.10 - CNTMT 1 OF 4 PRESS HI-3 1-PT-937, CNTMT PRESS (IR) CHAN I Containment pressure transmitter indication failed high

+30 secs RO RESPOND to Annunciator Alarm Procedures.

RO RECOGNIZE Containment pressure transmitter 1-PT-937, CNTMT PRESS (IR) CHAN I failed high.

US DIRECT performance of ALM-0022A, 1-ALB-2B, Window 3.10, CNTMT 1 OF 4 PRESS HI-3.

RO MONITOR Containment pressure.

1-PI-934, CNTMT PRESS (IR) CHAN IV 1-PI-935, CNTMT PRESS (IR) CHAN III 1-PI-936, CNTMT PRESS (IR) CHAN II 1-PI-937, CNTMT PRESS (IR) CHAN I RO DETERMINE 1-PI-937, CNTMT PRESS (IR) CHAN I is reading greater than or equal to 2.5 psig from the other channels.

RO REPORT to Unit Supervisor to REFER to Technical Specifications.

US EVALUATE Technical Specifications.

LCO 3.3.2.E, ESFAS Instrumentation.

CONDITION E - One Containment Pressure channel inoperable.

ACTION E.1 - Place channel in bypass within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

3 Page 9

of 20 Event

Description:

Containment Pressure Transmitter Failure Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc

+5 min US CORRECT the condition or INITIATE a Work Request per STA-606.

When Technical Specifications are addressed, or at Lead Examiner discretion, PROCEED to Event 4.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

4 Page 10 of 20 Event

Description:

Steam Pressure Control Channel Fails High Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: When directed, EXECUTE Event 4.

-MS13B, Steam Pressure Channel (PT-2326) fails high.

Indications Available:

1-PI-2326, MSL 2 PRESS pegged high 1-ZL-2326, SG 2 ATMOS RLF VLV read OPEN light LIT Y6845D Plant Computer alarm

+1 min BOP RESPOND to Dynamic Alarm Display (DAD) Alarm.

BOP RECOGNIZE Steam Generator 1-02 Steam Pressure Transmitter (PT-2326) failed high.

US DIRECT performance of ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure, and Feed Header Pressure Instrument Malfunction, Section 2.0.

BOP DETERMINE Steam Generator Atmospheric Relief Valve - OPEN.

US DIRECT closing of Steam Generator 1-02, Atmospheric Relief Valve.

BOP PLACE 1-PK-2326, SG 2 ATMOS RLF VLV CTRL in MANUAL and 0%

DEMAND to CLOSE Valve.

+5 min US NOTIFY Chemistry that a release has occurred.

When the Atmospheric Relief Valve is closed, or at Lead Examiner discretion, PROCEED to Events 5, 6, and 7.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 11 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: When directed, EXECUTE Event 5.

- RP14A, Spurious Train A Safety Injection Actuation Signal.

Indications Available:

6C-1.1 - MAN RX TRIP Numerous plant trip alarms

+30 secs RO/BOP RECOGNIZE Reactor Trip and Safety Injection.

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

RO VERIFY Reactor Trip:

DETERMINE Reactor Trip Breakers - OPEN.

DETERMINE Neutron flux - DECREASING.

RO DETERMINE all Control Rod Position Rod Bottom Lights - ON.

BOP VERIFY Turbine Trip:

DETERMINE all HP Turbine Stop Valves - CLOSED.

BOP VERIFY Power to AC Safeguards Buses:

DETERMINE both AC Safeguard Buses - ENERGIZED.

RO CHECK if SI is Actuated:

DETERMINE SI is actuated on Train A.

Manually ACTUATE SI to initiate dual train SI.

PLACE 1/1-SI2, SI MAN ACT Switch to ACT position.

RO VERIFY both Trains of Safety Injection actuated.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 12 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc Booth Operator: When EOP-0.0A, Step 4 has been verbally verified AND the Master Silence Alarm has timed out, EXECUTE Event 6.

- CV01B, Centrifugal Charging Pump (1-01) trip.

- CV01E, Centrifugal Charging Pump (1-02) fails to start.

RO RECOGNIZE CCP 1-01 trip.

START CCP 1-02.

VERIFY at least one CCP - RUNNING.

VERIFY Seal Injection Flow to each RCP -BETWEEN 6 GPM AND 13 GPM.

Examiner Note: EOP-0.0A, Attachment 2 steps performed by BOP are identified later in the scenario.

US/BOP INITIATE Proper Safeguards Equipment Operation Per Attachment 2.

RO VERIFY AFW Alignment:

DETERMINE MDAFW Pump 1 RUNNING.

DETERMINE Turbine Driven AFW Pump - RUNNING.

DETERMINE AFW total flow - GREATER THAN 460 GPM.

DETERMINE AFW valve alignment - PROPER ALIGNMENT.

RO VERIFY Containment Spray NOT Required:

VERIFY 1-ALB-2B Window 1-8, CS ACT NOT illuminated.

VERIFY 1-ALB-2B Window 4-11, CNTMT ISOL PHASE B ACT NOT illuminated.

DETERMINE Containment pressure remained < 18.0 PSIG.

VERIFY Containment Spray Heat Exchanger Outlet Valves - CLOSED.

VERIFY all Containment Spray Pumps - RUNNING.

RO DETERMINE Main Steam Lines Should NOT Be Isolated:

VERIFY Containment pressure < 6 PSIG.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 13 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc VERIFY Main Steam Line pressure > 610 PSIG.

RO CHECK RCS Temperature:

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

RO CHECK PRZR Valve Status:

VERIFY PRZR Safeties - CLOSED.

VERIFY Normal PRZR Spray Valves - CLOSED.

VERIFY PORVs - CLOSED.

VERIFY Power to at least one Block Valve - AVAILABLE.

VERIFY Block Valves - AT LEAST ONE OPEN.

RO CHECK if RCPs Should Be Stopped:

DETERMINE both SI Pumps - RUNNING.

DETERMINE Train A CCP Pump 1 TRIPPED.

START Train B CCP Pump 1-02.

DETERMINE RCS subcooling - GREATER THAN 25ºF (55ºF adverse).

Continue RUNNING Reactor Coolant Pumps.

RO/BOP CHECK if Any Steam Generator Is Faulted:

DETERMINE pressure in all Steam Generators - NORMAL.

RO/BOP CHECK if any Steam Generator is Ruptured:

DETERMINE radiation levels in all Steam Generators - NORMAL.

RO/BOP CHECK if RCS is intact:

DETERMINE Containment pressure, radiation level and sump levels - NORMAL.

US/RO CHECK if ECCS Flow to Should Be Reduced:

VERIFY Secondary heat sink:

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 14 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc DETERMINE Total AFW Flow to intact SGs > 460 GPM.

DETERMINE Narrow Range Level in SGs > 43%.

VERIFY RCS subcooling > 25ºF.

VERIFY RCS pressure - STABLE OR INCREASING.

VERIFY PRZR level > 13%.

US DETERMINE ECCS flow should be reduced and TRANSITION to EOS-1.1A, Safety Injection Termination.

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

2. EOS-1.1A steps are identified later in the scenario.

BOP VERIFY SSW Alignment:

VERIFY both SSW Pumps - RUNNING.

VERIFY Diesel Generator Cooler SSW return flow.

BOP VERIFY both Safety Injection Pumps - RUNNING.

BOP VERIFY Containment Isolation Phase A.

BOP VERIFY Containment Ventilation Isolation.

BOP VERIFY both CCW Pumps - RUNNING.

BOP VERIFY both RHR Pumps - RUNNING.

BOP VERIFY Proper CVCS Alignment:

DETERMINE Train A CCP Pump - TRIPPED.

If secured, START Train B CCP Pump.

VERIFY CCP 1 RUNNING.

VERIFY Letdown Relief Valve isolation:

DETERMINE Letdown Orifice Isolation Valves - CLOSED.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 15 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc DETERMINE Letdown Isolation Valves - CLOSED.

BOP VERIFY ECCS flow:

DETERMINE Train B CCP SI flow indicated.

DETERMINE RCS pressure > 1700 PSIG (1800 PSIG adverse).

DETERMINE NO SI Pump flow indicated.

DETERMINE RCS pressure > 325 PSIG (425 PSIG adverse).

BOP VERIFY Feedwater Isolation Complete:

VERIFY Feedwater Isolation Valves - CLOSED.

VERIFY Feedwater Isolation Bypass Valves - CLOSED.

VERIFY Feedwater Bypass Control Valves - CLOSED.

VERIFY Feedwater Control Valves - CLOSED.

BOP VERIFY both Diesel Generators - RUNNING.

BOP VERIFY Monitor Lights For SI Load Shedding illuminated.

BOP VERIFY Proper SI alignment per MLB light indication.

BOP VERIFY Components Properly Aligned per Table 1.

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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 16 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc CB-08 1-HS-2397A SG 1 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2398A SG 2 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2399A SG 3 BLDN HELB ISOL VLV CLOSED CB-08 1-HS-2400A SG 4 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 CV-01 X-HS-6188 PRI PLT SPLY FN 18 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6195 PRI PLT SPLY FN 19 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6202 PRI PLT SPLY FN 20 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6209 PRI PLT SPLY FN 21 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6216 PRI PLT SPLY FN 22 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6223 PRI PLT SPLY FN 23 & INTK DMPR STOPPED/DEENERGIZED CV-01 X-HS-6230 PRI PLT SPLY FN 24 & INTK DMPR 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

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 17 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc CV-03 X-HS-5856 CR EXH FN 2 STOPPED/DEENERGIZED CV-03 X-HS-5731 SFP EXH FN 33 STOPPED/DEENERGIZED CV-03 X-HS-5733 SFP EXH FN 34 STOPPED/DEENERGIZED CV-03 X-HS-5727 SFP EXH FN 35 STOPPED/DEENERGIZED CV-03 X-HS-5729 SFP EXH FN 36 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.

Examiner Note: EOS-1.1A, Safety Injection Termination steps begin here.

Examiner Note: The following six (6) steps are performed per EOS-1.1A, Attachment 1.D.

BOP

[1.D] PLACE Train B Diesel EMER START/STOP Handswitch in START.

BOP

[1.D] RESET SI.

BOP

[1.D] RESET SI Sequencers.

BOP

[1.D] RESET Containment Isolation Phase A and B.

BOP

[1.D] RESET Containment Spray Signal.

BOP/RO

[1.D] ESTABLISH Instrument Air and Nitrogen to Containment.

RO DETERMINE CCP 1 RUNNING.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 18 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc US/RO CHECK RCS Pressure - STABLE OR INCREASING.

Examiner Note: The following two (2) steps are performed per EOS-1.1A, Attachment 1.J.

CRITICAL TASK STATEMENT Determine Inadvertent Safety Injection & Terminate ECCS flow prior to exiting EOS-1.1A.

CRITICAL TASK RO

[1.J] ISOLATE CCP Injection Line Flow Path:

VERIFY CCP - SUCTION ALIGNED TO RWST.

ALIGN CCP Miniflow Valves:

OPEN 1/1-8110 and 1/1-8111, CCP Miniflow Valves.

CLOSE 1/1-8511A and 1/1-8511B, CCP Alternate Miniflow Isolation Valves.

PLACE Charging Flow Control Valve in MANUAL and 35% demand.

CLOSE 1/1-8801A and 1/1-8801B, CCP Injection Line Isolation Valves.

RO

[1.J] ESTABLISH Charging Flow Path:

OPEN 1/1-8105 and 1/1-8106, Charging Line Isolation Valves.

ADJUST Charging Flow Control Valve to establish Charging flow.

ADJUST RCP seal flow to maintain between 6 gpm and 13 gpm.

RO CONTROL Charging Flow to maintain PRZR Level.

RO CHECK If SI Pumps Should Be Stopped:

CHECK RCS pressure:

DETERMINE RCS pressure - STABLE OR INCREASING.

DETERMINE RCS pressure > 1700 PSIG.

STOP both SI Pumps and PLACE in Standby.

RO CHECK If RHR Pumps Should Be Stopped:

DETERMINE RHR Pumps RUNNING with suction aligned to RWST.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 19 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc STOP both RHR Pumps and PLACE in Standby.

RESET RHR Auto Switchover.

Booth Operator: When RHR Pumps are stopped and RHR Auto Switchover is reset, EXECUTE Event 7.

- RC17C, Small Break Loss of Coolant Accident @ 1700 gpm.

+20 min US/RO RECOGNIZE PZR level and subcooling lowering and TRANSITION to EOP-1.0.A, Loss of Reactor or Secondary Coolant.

Examiner Note: Crew may choose to start ECCS Pumps, open High Head Injection Valves or initiate Safety Injection. Any of these actions meets the CPNPP EOP Guidelines.

CRITICAL TASK STATEMENT Manually Initiate Safety Injection Flow Prior to exiting EOS-1.1A.

CRITICAL TASK RO/BOP START ECCS Pumps and/or REALIGN High Head Injection Valves and/or manually INITIATE Safety Injection.

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

RO CHECK If RCPs Should Be Stopped:

DETERMINE ECCS Pumps - AT LEAST ONE RUNNING.

DETERMINE RCS subcooling < 25ºF (55ºF adverse).

STOP all RCPs.

RO DETERMINE That NO Steam Generator Is Faulted.

BOP CHECK Intact SG Levels:

DETERMINE Steam Generator narrow range levels > 43% (50%

adverse).

CONTROL AFW Flow to maintain level between 50% and 60%.

Appendix D Operator Action Form ES-D-2 Operating Test :

NRC Scenario #

3 Event #

5, 6, & 7 Page 20 of 20 Event

Description:

Spurious Train A Safety Injection Actuation Signal / Centrifugal Charging Pump Trip / Centrifugal Charging Pump Start Failure / Small Break Loss of Coolant Accident Time Position Applicants Actions or Behavior CPNPP July 2010 NRC Sim Scenario #3 Rev b.doc RO/BOP CHECK Secondary Radiation - NORMAL.

RO CHECK PRZR PORVs And Block Valves - AVAILABLE.

+40 min US DETERMINE ECCS Flow Should NOT Be Reduced.

When High Head Safety Injection flow is restored, TERMINATE the scenario.

ES-301 Transient and Event Checklist Form ES-301-5 CPNPP July 2010 NRC ES-301-5 Transient and Event Checklist R.doc Facility:

CPNPP 1 and 2 Date of Exam:

07/07/10 Operating Test No.:

NRC A

P P

L I

C A

N T

E V

E N

T T

Y P

E SCENARIOS CPNPP #1 CPNPP #2 CPNPP #3 (Spare)

T O

T A

L MINIMUM(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

SROI-1 RX 0

1 1

0 NOR 0

1 1

1 I/C 1,2,3, 4

1,3, 6

7 4

4 2

MAJ 5

5 2

2 2

1 TS 1,4 2

0 2

2 RO-1 RX 0

1 1

0 NOR 0

1 1

1 I/C 1,4, 8

1,2, 4, 7,8 8

4 4

2 MAJ 5

5 2

2 2

1 TS 0

0 2

2 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

ES-301 Competencies Checklist Form ES-301-6 CPNPP July 2010 NRC ES-301-6 Competencies Checklist Rev c.doc Facility:

CPNPP Date of Examination:

07/07/10 Operating Test No. NRC 1, 2, 3 Applicants SROI-1 RO-1 Competencies SCENARIO SCENARIO SCENARIO 1

2 1

2 Interpret/Diag-nose Events and Conditions 1,2,3, 4,5 1,3, 5,6 1,4,5, 8

1,2,4, 5,7,8 Comply With and Use Procedures (1)

ALL 1,3, 5,6 1,4,5, 8

1,2,4, 5,7,8 Operate Control Boards (2)

N/A 1,3, 5,6 1,4,5, 8

1,2,4, 5,7,8 Communicate and Interact ALL 1,2,3, 5,6 1,4,5, 6,7,8 1,2,4, 5,7,8 Demonstrate Supervisory Ability (3)

ALL N/A N/A N/A Comply With and Use Tech.

Specs. (3) 1,4 N/A N/A N/A Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.