Text

Initial Submittal of the Scenarios for the Perry Initial Examinaton - March 2002
	ML021190681
Person / Time
Site:	Perry
Issue date:	03/05/2002
From:	Dante Johnson; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	50-440/02301, ES-D-1
	Download: ML021190681 (82)
	v • d • e

INITIAL SUBMITTAL OF THE SCENARIOS FOR THE PERRY INITIAL EXAMINATION - MARCH 2002

Appendix D Scenario Outline Form ES-D-1 Facility:

Perry Scenario No.: 1 Examiners:

Op-Test No.: 2002-01 Operators:

Initial Conditions: A Reactor startup is in progress following a brief outage. Reactor power is being held at 70%

power per SCC request. A xenon transient is in progress. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech. Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The MFP is in secured status to support recirc valve actuator work. SRV F041 E is weeping and causing the Suppression Pool to slowly heatup. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented.

Turnover: 1. BOP operator place RHR A in suppression pool cooling mode and lower Suppression Pool temperature to 75 F. ESW A and ECC A are in operation.

2. Maintain 70% power.

Target Critical Tasks: Emergency Depressurization, Restore RPV water level Event Malf. No.

Event Event No.

Type*

Description 1

N (BOP)

Startup RHR A in Suppression Pool cooling mode 2

MV06:

C (BOP)

RHR Min Flow valve (F064A) fails due to mechanical binding after RHR flow is 1E1 2-F064A established (TS 3.5.1.C, 3.6.1.7.B, 3.6.2.3.B) 3 CN02:

I (BOP)

CRDH flow controller failure in Auto mode 1C11 R0600 100%

RD12R1447 C (RO)

Single control rod drift inward (14-47) 4 Various C (RO)

Reactor Feed Pump A bearing failure R (RO)

Lower reactor power to 63% using recirc flow N (RO)

Remove RFPT A from service ZL1N27R0425A I (RO)

RFP A manual speed control dial pot failure (as is) 5 FW02 - 50%

M (All)

Feedwater System Pipe Break inside Drywell / Reactor Scram CP01:

C (BOP)

HPCS Pump shaft breaks 1E22C0001 MV04:

1E51F0013 C (BOP)

RCIC Injection Valve (F013) failure to Auto open RC03 C (BOP)

RCIC Turbine mechanical trip latch failure 6

TH02C C (All)

Recirc Bottom, Head D razir pipe break (5 minute, ramp) i;--

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75%

ZD1B21S34B C (BOP)

ADS B Inhibit Switch failure in Normal position 7

M RPV emergency depressurization / Inject with lbw pressure ECCS to maintain adequate core cooling

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 1 Objectives:

The BOP operator places RHR A in Suppression Pool Cooling mode to lower Suppression Pool temperature to 75 F. When RHR is initiated, the RHR minimum flow valve will remain open after flow is established. This will make RHR inoperable and the Tech. Specs. will be consulted.

Immediately after Tech Specs are referenced, the CRDH Flow Controller will fail in the Auto mode causing a single control rod to drift inward. The crew will recognize the drifting control rod and enter ONI-C51, Unplanned Change in Reactor Power or Reactivity. When the crew identifies the failed CRDH Flow Controller, the controller will be placed in Manual and CRDH parameters will be returned to normal allowing the drifted control rod to settle at notch 00.

After a plan is implemented to recover the drifted control rod, RFP A will experience sustained high bearing vibration requiring reactor power to be lowered to 63% using recirc flow to allow the RFP to be removed from service.

After the power reduction, as RFP A is being removed from service, the manual speed control dial potentiometer will fail requiring a trip of the RFP to remove it from service. When RFP A is tripped, a feedwater pipe breaks in the Drywell resulting in a reactor scram.

Following the scram, the HPCS pump will be unavailable due to a shaft break.

When RCIC is initiated for level control, the RCIC injection valve will fail to open automatically requiring operator action to manually open the injection valve. After injection is established with RCIC, the RCIC Turbine mechanical trip latch will fail making RCIC unavailable for injection, A small Recirc pipe break in the Drywell will develop and slowly increase in severity resulting in rising drywell temperature and pressure and lowering RPV water level. As reactor level continues to lower, the crew will place alternate injection systems in service, emergency depressurize the RPV, and maintain adequate core cooling with low pressure ECCS systems in accordance with PEI-B1I3, RPV Control (Non-ATWS). However, when ADS is inhibited, the ADS Inhibit Switch B will fail which may require the crew to take action to prevent an unintended ADS blowdown.

Discussion of Safety Significance for scenario I The BOP operator will place RHR in Suppression Pool Cooling. When RHR flow is initiated, the RHR Min flow valve stays open after flow is established. The BOP operator must note that the min flow valve has failed to close and inform the SRO because with the minimum flow valve failed open, RHR A is inoperable.

Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 1 The CRDH Flow Controller will fail open in the Auto mode causing a single control rod to drift inward. The RO must recognize that a control rod is drifting and determine which rod is drifting. The drifting control rod is safety significant because it directly effects core reactivity and core power distribution. Operators will be required to determine the cause of the drifting control rod (failed CRDH flow controller) and begin actions to recover the control rod.

After a plan is implemented to recover the drifted control rod, RFP A will experience sustained high bearing vibration. This is safety significant because it will require the operators to lower reactor power in a controlled manner and remove the RFP from service. Removing the RFP from service is safety significant because the operation directly affects RPV water level control. As the RFP is being removed from service, the manual speed control dial pot will fail 'as is' requiring a trip of the RFP, directly leading to a Feedwater System pipe break in the Drywell and a reactor scram.

Following the scram, the HPCS pump will be unavailable due to a shaft break.

When RCIC is initiated for level control, the RCIC injection valve will fail to open automatically. The RCIC injection valve failure is safety significant because RCIC is the only normal high-pressure injection system available. Operator action will be required to manually open the injection valve allowing RCIC injection. After manual flow control is established with RCIC, the RCIC Turbine mechanical trip latch will fail making RCIC unavailable for injection.

A small Recirc pipe break in the Drywell will develop and slowly increase in severity resulting in rising Drywell temperature and pressure and lowering RPV water level. The Recirc pipe break is safety significant because it directly contributes to the loss of coolant inventory requiring the crew to place alternate injection systems in service and prepare for RPV emergency depressurization to allow injection with low pressure ECCS systems. As RPV level continues to lower and ADS is inhibited, ADS Inhibit Switch B will fail. Failure of the capability to inhibit ADS is safety significant because it places the crew in a condition not anticipated by PEI-B 13, RPV Control (Non-A TWS), requiring the crew to take action to prevent unintended RPV emergency depressurization.

S............................"....

Final - Revision 0 Page 2 of 2

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 1 Page 1 of 1 Event

Description:

Place RHR A in Suppression Pool Cooling Time Position Applicant's Actions or Behavior Direct BOP to place RHR A in Suppression Pool Cooling in accordance with SOI-E12, Section 4.4 BOP Places RHR A in Suppression Pool Cooling.

Notify HP that a Suppression Pool evolution will be conducted Verifies ESW in operation Verifies ECC System in operation Places RHR A OUT OF SERVICE switch in INOP Expected Alarm - H13-P601-20 (A6), RHR A OUT OF SERVICE Verifies RHR A HX's Outlet Valve, 1 E12-F03A is open Close RHR A HX's Bypass Valve, 1 E1 2-F048A Throttle closed RHR A HX's Outlet Valve, 1 El 2-FO03A for 18 to 20 seconds Place RHR Pump A control switch to start Observes rising amps on RHR Pump A Expected Alarm - H 13-P601-19 (F9), ADS A PERMISSIVE LPCS/RHR A RUN Take RHR A Test Valve to Supr Pool, 1 El 2-F024A to open Recognizes that RHR Pump A Minimum Flow Valve 1 E12-F064A Power Loss Lamp illuminated I

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-I-NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 2 Page 1 of 2 Event

Description:

RHR Min Flow Valve stays open after RHR flow established Time Position Applicant's Actions or Behavior BOP Receives report F064A, RHR min flow valve, is closed Notifies SRO that RHR A min flow valve has failed closed SRO Acknowledge report of failed closed RHR A minimum flow valve Note: P&L # 9 of SOI-E12 requires RHR A be declared INOP SRO Directs one of the following:

Shutdown RHR A per SOl-E12, Section 6.3 and rack out RHR Pump A breaker or Shutdown RHR A per SOI-E12, Section 6.3 and place in secured status per SOI-E12, Section 6.5 or Provide increased monitoring whenever RHR Pump A is in operation BOP If directed, perform increased monitoring while RHR Pump A is in operation BOP If directed, shutdown RHR A per SOI-E12, Section 6.3 Take RHR A Test Valve to Supr Pool, 1 El 2-F024A to Close and verify valve fully closed Fully open RHR A HX's Outlet Valve, 1E12-F03A Take RHR Pump A control switch to Stop BOP If directed, direct NLO to rack out RHR Pump A breaker or place RHR A in secured status per SOI-E12, Section 6.5 NUREG-1021, Revision 8, Supplement 1 40 of 40

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Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 2 Page 2 of 2 Event

Description:

RHR Min Flow Valve stays open after RHR flow established Time Position Applicant's Actions or Behavior SRO/BOP Requests Responsible System Engineer and Maintaince assistance in the Control Room to support troubleshooting SRO References Tech Specs LCO 3.5.1, Enters Condition C LCO 3.6.1.7, Enters Condition B LCO 3.6.2.3, Enters Condition B SRO Notify Operations Management of RHR A inoperability t

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NUREG-1021, Revision 8, Supplement 1 40 of 40 I

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Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 3 Page 1 of 2 Event

Description:

CRDH Flow Controller failure in Auto mode concurrent with single Control Rod drift inward (14-47)

Time Position Applicant's Actions or Behavior RO Recognize and report Alarm H13-P680-5 (D10), ROD DRIFT Identify Control Rod 14-47 is drifting Report Control Rod 14-47 drifting inward Determine Control Rod 14-47 is fully inserted BOP Support RO with ARI review SRO Enter ONI-C51, Unplanned Change in Reactor Power or Reactivity Direct RO/BOP to evaluate subsequent actions Notify Reactor Engineering to determine if any power distribution limits have been exceeded SRO Notify Operations Management and NRC Resident of ONI-C51 entry due to drifting Control Rod 14-47 RO/BOP Direct NLO to investigate HCU 14-47 for drifted Control Rod BOP/RO Determines CRD Flow Controller output is failed high, indicated by:

Maximum CRD Drive Water Pressure Maximum CRD Cooling Water Pressure CRD Flow Controller output meter indicates maximum BOP Informs SRO CRD Flow Controller output is failed high NUREG-1 021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 3 Page 2 of 2 Event

Description:

CRDH Flow Controller failure in Auto mode concurrent with single Control Rod drift inward (14-47)

Time Position Applicant's Actions or Behavior SRO/RO/BOP NLO in Containment reports no abnormal indications at HCU for Control Rod 14-47 SRO Directs BOP to place CRD Flow Controller in Manual and restore CRDH parameters to normal BOP Restores CRDH parameters to normal RO Reports Alarm H13-P680-5 (D10), ROD DRIFT has cleared SRO References FTI-B02 for recovery actions Consults with Reactor Engineering to develop a recovery plan Notifies Operations Management of cause of Control Rod 14-47 drift and intended recovery plan Requests Special Maneuver Sheet from Reactor Engineering for recovery of Control Rod 14-47 I ___

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1-NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 4 Page 1 of 1 Event

Description:

Feedwater Pump A Bearing Failure / Manual trip of RFPT A Time Position Applicant's Actions or Behavior RO/BOP Respond to alarm - Review ARI 1 H1 3-P680-3 (C6) RFP A VIB Monitor RFPT A vibration level Dispatch an NLO to RFPT A Contact Responsible System Engineer (RSE)

RO/BOP/SRO NLO at RFPT A reports squealing noise from RFP end RO/BOP Reports requirement to trip RFPT A when vibration level has exceeded 4 mils for greater than 3 minutes SRO References ONI-N27 or SOI-N27 (63% power or 23.1 Kgpm limit)

Direct RO to reduce power with Recirc Flow Direct RFPT A removed from service when < 63% power or

< 23.1Kgpm flow RFP B RO/BOP Lowers power with Recirc Flow Using SOI-C34, transfer RFPT A from Master to Manual Flow Control and from Manual Flow Control to Manual Speed Control Dial Reports RFPT A Manual Speed Control Dial failure SRO Acknowledges RFPT A Manual Speed Control Dial failure and directs tripping of RFPT A (May not enter ONI-N27 due to RFP A not feeding RPV)

RO/BOP Trips RFPT A as directed NUREG-1 021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 5 Page 1 of 2 Event

Description:

Feedwater System Pipe Break Inside Drywell / Reactor Scram / HPCS shaft break / RCIC Injection Valve (F01 3) failure to Auto open / RCIC Turbine mechanical trip latch failure.

Time Position Applicant's Actions or Behavior RO Inform SRO/BOP of reactor scram Performs ONI-C71-1, Reactor Scram, immediate actions RO/BOP Inform SRO of lowering water level RO/BOP Report that RPV water level is less than Level 3 RO/BOP Manually initiate RCIC SRO Enters PEI-B13, RPV Control (Non-ATWS) due to RPV < L3

- Verifies reactor is scrammed

- Confirms Reactor Mode Switch is in SHUTDOWN

- Start Hydrogen Analyzers

- Verifies reactor shutdown under all condition without boron

- Verifies SRMs and IRMs inserted

- Directs pressure control 800 to 1000 psig using BPVs Enters PEI-T23, Containment Control

- Operate all available Drywell Cooling

- Restore Drywell Cooling (PEI-SPI 2.1)

RO/BOP (May) report indications of Feedwater line break NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 6 Page 1 of 1 Event

Description:

Recirc Bottom Head Drain pipe break / ADS B Inhibit Switch failure in Normal position Time Position Applicant's Actions or Behavior SRO Determine RPV level cannot be maintained > 0 inches Direct maximum injection with CRD per PEI-SPI 4.1 Direct start of available low pressure ECCS pumps SRO Determines Containment Spray required at 2.25 psig.

Containment Spray unavailable if RHR A was secured BOP/RO Lineup and inject CRD per PEI-SPI 4.1 When directed, start available low pressure ECCS pumps BOP/RO Report MSIVs closed when RPV level reaches Li (16.5")

SRO When RPV level reaches Li (16.5")

- Executes PEI-M51/56, Hydrogen Control

- Direct Inhibit ADS

- Directs pressure control 700 to 900 psig using SRVs BOP/RO When directed, inhibit ADS Identify and report failure of ADS B Inhibit switch to SRO Report Alarm H13-P601-19 (D10), ADS LOGIC B TIME DELAY LOGIC TIMER RUNNING SRO Acknowledge failure of ADS B Inhibit switch Direct delay of ADS actuation in accordance with SOI-B21, Section 7.4 or direct RO and BOP to prepare for an automatic ADS actuation NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 5 Page 2 of 2 Event

Description:

Feedwater System Pipe Break Inside Drywell / Reactor Scram / HPCS shaft break / RCIC Injection Valve (F01 3) failure to Auto open / RCIC Turbine mechanical trip latch failure.

Time Position Applicant's Actions or Behavior SRO RPV Level Control

- Direct restore and maintain RPV level between 185 and 215 inches

- Direct level control with RCIC and HPCS

- Direct RO/BOP to isolate Feedwater line break (if recognized)

RO/BOP Report failure of RCIC Injection Valve to open Manually opens RCIC Injection Valve RO/BOP Determine RCIC has tripped Direct NLO to investigate RCIC trip Report RCIC trip to SRO Determine HPCS shaft is broken RO/BOP Report HPCS failure to SRO SRO Determine RPV level cannot be maintained between 185 and 215 inches Direct restore and maintain RPV level between 0 and 215 inches using SLC Demin Water Alternate Injection per PEI-SPI 4.5 BOP/RO When directed, line up and inject SLC per PEI-SPI 4.5 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 7 Page 1 of 1 Event

Description:

RPV emergency depressurization / Inject with low pressure ECCS to maintain adequate core cooling.

Time Position Applicant's Actions or Behavior RO/BOP Maintain RPV pressure as directed RO/BOP Delay ADS actuation in accordance with SOl-B21 as directed SRO When RPV level reaches 0 inches, confirm injection subsystem lined up with pump running When RPV level reaches - 25 inches, PEI-B1 3 Emergency Depressurization is entered and executed concurrently with PEI-B13, RPV Control (Non-ATWS), exit Pressure Control Leg SRO Directs BOP/RO actions per PEI-B13 Emergency Depressurization

- Confirm Reactor shutdown under all conditions without boron

- Determines Drywell pressure > 1.68 psig and low press ECCS required for adequate core cooling.

- Verifies eight or more SRVs are not open

- Verifies Suppression Pool level is > 5.25 feet

- Direct all ADS valves opened to rapidly depressurize the RPV BOP/RO When directed, open all ADS valves or verify automatic ADS actuation occurs when expected and all ADS valves open SRO Confirms all ADS valves are open Direct maximum injection flow with all available systems BOP/RO Maximize injection flow with available injection system until adequate core cooling is assured NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Page 1 of 1 Event

Description:

Scenario Termination Criteria Time Position Applicant's Actions or Behavior

1. All Control Rods are fully inserted

2. The reactor has been emergency depressurized

3. Low pressure ECCS Systems are being used to restore RPV level to 185 to 215 inches

4. Containment and Drywell parameters are being restored per PEI-T23, Containment Control i

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Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 6 Page 1 of 1 Event

Description:

Critical Task #1 Time Position Applicant's Actions or Behavior

This critical task does not apply if ADS actuation occurs automatically Critical Task #1 - When RPV water level cannot be maintained greater than zero inches, initiate Emergency Depressurization

1. Safety Significance:

- Maintain adequate core cooling

2. Cues:

- Procedural compliance

- Level lowering without adequate high pressure injection available

3. Measured by:

- Observation - at least 5 SRVs open

4. Feedback:

- Reactor pressure trend

- Suppression Pool temperature trend I.

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NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 1 Event No.: 7 Page 1 of 1 Event

Description:

Critical Task #2 Time Position Applicant's Actions or Behavior Critical Task #2 - When RPV water level cannot be maintained greater than zero inches and RPV depressurization is in progress, increase and control injection into the RPV to restore and maintain RPV level greater than zero inches.

1. Safety Significance:

- Establish adequate core cooling

2. Cues:

- RPV pressure trend

- Procedural compliance

3. Measured by:

- RPV injection is established and level is above the TAF and rising

4. Feedback:

- Lowering Suppression Pool level

- Lack of Hydrogen generation

- RPV level and pressure indications NUREG-1021, Revision 8, Supplement 1 40 of 40

Facility:

Perry Scenario No.: 2 Examiners:

Op-Test No.: 2002-01 Operators:

Initial Conditions: The plant is operating at 100% power. A xenon transient is in progress. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech.

Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented. HPCS testing is scheduled to support flow rate testing.

Turnover: 1. Place HPCS in full flow test mode to the suppression pool. HPCS ESW and HPCS Pump Room Cooler are in operation. 2. Maintain 100% power.

Target Critical Tasks: Initiate action to shutdown the reactor, Inhibit ADS, Terminate and Prevent injection into the RPV, Emergency Depressurization, Restore RPV water level.

Event Malf. No.

Event Event No.

Type*

Description 1

N (BOP)

Start HPCS in full flow test mode to the Suppression Pool 2

CP03:

C (BOP)

HPCS Pump flow degradation (2 minute ramp) (TS 3.5.1. B and C) 1 E22C0001 100%

3 AD01N C (BOP)

ADS/SRV B21-F047H cycling (TS 3.5.1.E, F and H /TS 3.0.3)

R (RO)

Lower reactor power to 90% using recirc flow 4

CN03:

I (RO)

Reactor Feed Pump Controller B oscillations (1 minute ramp) 1C34R060B 20%

5 ED061 C (BOP)

Loss of 480Vac Bus F-I-E TC05 C (ALL)

Turbine Control EHC leak I Main Turbine trip and reactor scram 10%

6 RD15 C (RO)

Failure of RPS and ARI to automatically shutdown the reactor M (All)

ATWS SL01A C (BOP)

SLC Squib Valve failures, C41-FO04A and C41-FO04B SLIB01B_

7 CBOl:

C (RO)

All RFBPs trip 1N27C0001A M (All)

Loss of all Feedwater capability 1N27C0001B 1N27Co001C CB01:

1 N27C0001D 8

M (All)

RPV emergency depressurization / Inject with low pressure ECCS to maintain adequate core cooling RV04:

C (BOP)

ADS/SRV B21-F041 F failure closed 1B21 F0041F "kiNi)ormal, kr*)eacivity, kl)nstrument, (L)omponent, (M)ajor Final - Revision 1

'k Appendix D Scenario Outline Form ES-D-1

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2002 Perry NRC Examination Scenario Objectives Safety Significance Discussions Scenario 2 Objectives:

The BOP operator will place HPCS in the full flow test mode to support flowrate testing. After HPCS flow is stabilized for the test, HPCS Pump flow will be degraded requiring the HPCS System to be declared inoperable.

After Tech. Specs. have been consulted, an ADS/SRV will open and cycle due to shorted switch contacts; the crew will enter ONI-B21-1, SRV Inadvertent' Opening/Stuck Open, evacuate the Containment, reduce reactor power to 90%

using recirc flow, and de-energize the SRV solenoids by removing control power fuses. After the SRV closes, Tech Specs must again be consulted and the loss of HPCS and ADS will require a Tech. Spec. 3.0.3 entry.

The crew will respond to a RPV level transient due to Feedwater Flow Controller A oscillations that require entry into ONI-C34, Feedwater Flow Control Malfunction, and manual control of feedwater flow.

After conditions have stabilized, a loss of 480 Volt Bus F-1-E will result in a loss of the running EHC Pump B, TBCC Pump C, and CVCW Chiller A (in addition to other loads). Following auto start of standby EHC Pump A, the EHC System will develop a leak which will slowly increase in severity until the reactor is manually scrammed and the Main Turbine is tripped or the Main Turbine trips automatically.

When the reactor scrams, the control rods will fail to fully insert due to blockage in the scram discharge volume. PEI-B13 (RPV Control-ATWS) is entered and executed to stabilize the plant. SLC squib valves will fail to fire.

After PEI ATWS actions are underway and RPV level has reached L2, (130" above TAF), Feedwater System capability is lost and the RCIC system will be unable to provide adequate makeup. Therefore, the crew will emergency depressurize the RPV to allow for low pressure ECCS injection. Two ADS/SRVs will fail to operate during the emergency depressurization.

Discussion of Safety Significance for.

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The BOP operator will place HPCS in -the full flow test mode After HPCS flow is stabilized for the test, the BOP operator must note that HPCS Pump flow is degraded. This is safety significant because the HPCS System is inoperable and unavailable for core cooling.

Final - Revision 0 Page 1 of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussions Scenario 2 Next, an ADS/SRV will open and cycle due to shorted switch contacts; the crew will enter ONI-B21-1, evacuate the Containment, reduce reactor power, and deenergize the SRV solenoids by removing control power fuses. This is safety significant because the cycling SRV directly affects core reactivity and results in a rising Suppression Pool temperature. When SRV fuses are pulled, the crew must determine that the ADS/SRV is inoperable and unavailable. This is safety significant because it further degrades the status of the high-pressure ECCS systems, thus requiring entry into Tech. Spec. 3.0.3.

The crew will then respond to a RPV level transient due to Feedwater Flow Controller A oscillations. This is significant because it requires manual control of RPV water level to terminate the level transient.

After conditions have stabilized, a loss of 480 Volt Bus F-1-E will result in a loss of running EHC Pump B, TBCC Pump C, and CVCW Chiller A (in addition to other loads). This is safety significant because it requires the crew to evaluate plant status and enter the appropriate ONIs for the 480-Volt bus failure and TBCC Pump trip. Following the auto start of the standby EHC Pump A, a turbine control EHC leak will develop. This is safety significant because the crew must recognize the EHC leak, manually scram the reactor, and trip the Main Turbine before the Main Turbine trips automatically.

When the reactor is scrammed, the control rods fail to fully insert due to blockage in the scram discharge volume. The failure of control rods to insert is safety significant because it will require the operators to take PEI A 71/S actions to shutdown the reactor, control reactor level, and control reactor pressure.

During A TWS actions, the SL C squib valves will fail to fire. This failure is safety significant because it will lengthen the time required to achieve reactor shutdown.

Following the loss of the Feedwater System, the crew must determine that the RCIC System will be unable to provide adequate makeup, thereby challenging the ability to maintain adequate cQre cooling. This is safetysqinificant because emergency depressurization will be requirodtQ,1 ta, bhsh controlledijection with '

low pressure ECCS to assure adequate core cooling. TWo.AS/$RVs will fail to operate during the emergency depressurization. Failure 'of the ADS SRVs to open is safety significant because the crew must recognize the failure and open additional SRVs.

Final - Revision 0 Page 2 of 2

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 1 Page 1 of 1 Event

Description:

Place HPCS in Full Flow Test Time Position Applicant's Actions or Behavior SRO Direct BOP to place HPCS in Full Flow Test per SOI-E22A, Section 75 BOP Places HPCS in Full Flow Test

- Places HPCS OUT OF SERVICE Switch in INOP Expected Alarm H13-P601-16 (D4), HPCS OUT OF SERVICE

- Notify HP that a Suppression Pool evolution will be conducted

- Verifies HPCS ESW loop in operation

- Verifies HPCS Pump Room Cooling in operation

- Notifies SRO HPCS suction shift to the Supp Pool is required

- Performs SOI-E22, Section 5.2 as SRO directs Note: CST Suction automatically closes when Supp Pool Suction Valve opens. No expected alarms

- Verifies HPCS Supp Pool Suction Valve, 1 E22-FOl 5 is open

- Verifies HPCS CST Suction Valve, 1 E22-FOO1 is closed

- Place HPCS Pump control switch to start:

Observes rising pump discharge pressure Observes rising pump amps Expected Alarm, H13-P601-16 (A5), HPCS PUMP START SIGNAL RECEIVED Observes HPCS Min Flow Valve opens

- Hold HPCS Test Valve to Supp Pool, 1 E22-F023 in open:

Flow approximately 6900 gpm on E22-R603 with Test Valve fully open E22-R616, Pump Amps, approximately 320 amps E22-R601, Discharge Press, approximately 300 psig NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 2 Page 1 of 2 Event

Description:

HPCS Pump flow degradation Time Position Applicant's Actions or Behavior BOP Responds to unexpected alarm H13-P601-16 (H5), HPCS WATER LEG PUMP DISCH PRESS LO Determines HPCS Pump flow is degrading Notify SRO that HPCS Pump flow is degraded, as indicated by discharge press slowly going to 0 psig / flow going to 0 gpm SRO Acknowledge report of degraded HPCS Pump flow Directs BOP to shutdown HPCS per SOI-E22A, Section 7.5 SRO/BOP/RO Requests Maintaince and Responsible System Engineer (RSE) assistance in the Control Room to support troubleshooting SRO/BOP/RO NLO at HPCS Pump reports HPCS Pump is extremely noisy SRO References Tech Specs

- LCO 3.5.1, Enters Condition B

- Notify Operations Management of HPCS inoperability Note: Must declare HPCS inoperable per Tech Specs prior to receiving following recommendation from RSE SRO/BOP/RO NLO at HPCS Pump recommends placing HPCS in secured status SRO Directs BOP to place HPCS in Secured Status in accordance with SOI-E22A, Section 6.2 Note: May confer with Shift Manager prior to directing HPCS be placed in secured status NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 2 Page 2 of 2 Event

Description:

HPCS Pump flow degradation Time Position Applicant's Actions or Behavior BOP Inform SRO a Procedure Deviation is required because flow cannot be lowered to 500 to 600 gpm. (flow indicates 0 gpm)

SRO Authorizes a Procedure Deviation as required Note: May consult with Shift Manager prior to authorizing BOP Shutdown HPCS per SOI-E22A, Section 7.5

- Hold HPCS Test Valve to Supp Pool, 1 E22-F023 to Close and verify valve fully closed:

Alarm H13-P601-16 (H5), HPCS WATER LEG PUMP DISCH PRESS LO clears when 1E22-F023 is closed

- Take HPCS Pump control switch to Stop:

Alarm, H13-P601-16 (A5), HPCS PUMP START SIGNAL RECEIVED clears Places HPCS in Secured Status using SOI-E22A, Section 6.2 as directed.

Note: Placing HPCS in shutdown instead of secured status will not affect the scenario outcome.

NUREG-1021, Revision 8, Supplement 1 40 of 40 I

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Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 3 Page 1 of 2 Event

Description:

ADS/SRV 1 B21-F047H cycling Time Position Applicant's Actions or Behavior BOP/RO Reports unexpected alarms and consults ARIs:

H13-P601-19: (A7), SRV OPEN H13-P601-19: (B7), SRV OPEN SIGNAL RECEIVED RO Monitors RPV level, pressure, and power BOP Informs SRO/RO that ADS/SRV 1B21-F047H is cycling SRO Enters ONI-B21-1, SRV Inadvertent Opening/Stuck Open

- Directs RO/BOP to initiate evacuation of Containment

- Directs RO to reduce reactor power using recirc flow to < 90%

Note: Must reduce power to < 90% prior to attempting To close the SRV

- Directs BOP to attempt to close the SRV by placing both of its control switches from AUTO to OFF

- Directs BOP to de-energize the SRV solenoids by removing the applicable control power fuses

- Coordinates with RO/BOP to complete applicable Supplemental Actions RO/BOP Notifies SCC, Chem and HP of intent to lower reactor power (may occur after power reduction has begun)

SRO Provides SRO oversight for power reduction NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 3 Page 2 of 2 Event

Description:

ADS/SRV 1 B21-F047H cycling Time Position Applicant's Actions or Behavior Decreases reactor power to 90% using Recirc Loop Flow Control RO Maintain Recirc loop flows matched within 10%

BOP Attempt to close SRV by placing both control switches to off:

Expected Alarm, H13-P601-19: (A7), SRV DIV 1/2 SWITCH IN OFF Informs SRO/RO that SRV is still cycling open SRO/BOP Refer to ONI-B21-1 Attachment 1 and Attachment 2 to determine SRV fuses that must be pulled BOP Attempts to close SRV by removing its control power fuses inside panel H13-P628 and in H13-P631 Informs SRO/RO that SRV is closed Expect Alarms: H13-P601-19 (G9) (G11), ADS OUT OF SERVICE SRO References Tech Specs for a single, inoperable ADS SRV

- LCO 3.5.1, Enters Condition E, F, and H

- Enters LCO 3.0.3, due to ADS Valve and HPCS inoperable SRO Notifies OPS Management and NRC Resident of ONI entry and reason for entry, and of the various LCO entries and required TS 3.0.3 shutdown. Requests RSE and I&C assistance in the Control Room to support troubleshooting Review 101-3 for power reduction Begin preparations for required plant (TS 3.0.3) shutdown NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 4 Page 1 of 2 Event

Description:

Reactor Feed Pump Controller B oscillations Time Position Applicant's Actions or Behavior RO Reports Feed Flow / Steam Flow mismatch Reports reactor water level oscillations Reports RX LEVEL HI/LO L7/L4 alarm (H13-P680-3 (A9))

BOP Supports RO by consulting ARI-H13-P680-3 (A9)

SRO Acknowledges receipt of unexpected alarm Enters ONI-C34 due to malfunction of feedwater level control

_~SROO__

- Directs RO to transfer control of both RFPTs to the Manual Speed Control Dial and maintain reactor water level 192 to 200 inches

- Directs RO to place RFP A & B FLOW CONTROL for both RFPTs to Manual

- Coordinates with RO/BOP to complete applicable Supplemental Actions RO/BOP Requests Responsible System Engineer and I&C assistance in the Control Room to support troubleshooting SRO Notifies OPS Management and NRC Resident of ONI-C34 entry and reason for entry SRO Evaluates feedwater level control options NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 4 Page 2 of 2 Event

Description:

Reactor Feed Pump Controller B oscillations Time Position Applicant's Actions or Behavior SRO Directs RO to transfer RFPT A (B) from the Manual Speed Control Dial to the Startup Level Control per SOI-C34, Section 4.6

- Provides SRO oversight during feedwater level control shift

One RFPT will be on the SULC and the other RFPT will be base loaded SRO Directs BOP to monitor reactor power and reactor pressure during the feedwater level control shift BOP Monitors reactor power and reactor pressure RO Transfers RFPT A (B) from the Manual Speed Control Dial to the Startup Level Control

- Verifies RFPT B (A) is being controlled by RFPT B (A)

Manual Speed Control Dial

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1 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 5 Page 1 of 2 Event

Description:

Loss of 480Vac Bus F-1 -E. Turbine Control EHC leak / Main Turbine trip and reactor scram.

Time Position Applicant's Actions or Behavior SRO/BOP/RO Recognize and report Alarms H1 3-P870-1 (E4), 480 VOLT BUS UNDERVOLTAGE and H13-P870-1 (E6), BUS F-1-E BREAKER TRIP SRO Enters ONI-R23-2, Loss of Non-Essential 480 Volt Bus Refers to Plant Data Book for list of affected loads (PDB-H001 7):

CVCW deenergized TBCC Pump C deenergized Note: TBCC Pump breaker remains closed. No Alarm received when BUS F-1-E deenergized.

RO/BOP Walk down panels and assess plant status:

Acknowledge and report alarm H13-P870-9 (G2), EHC STBY PUMP START-HEADER PRESSURE LOW; refer to ARI Determines EHC Pump B tripped; EHC Pump A auto started Note: EHC Pump B breaker remained closed when Bus F-1-E deenergized SRO Enters and executes ONI-P44, Loss of Turbine Building Closed Cooling RO/BOP Acknowledge alarm H13-P870-9 (F2), EHC SYSTEM RESERVOIR HI/LO; refer to ARI and inform SRO SRO Directs reactor scram or enters and direct actions of 101-14, Fast Unload and Trip of Turbine NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 5 Page 2 of 2 Event

Description:

Loss of 480Vac Bus F-1-E. Turbine Control EHC leak / Main Turbine trip and reactor scram.

Time Position Applicant's Actions or Behavior Rfl As directed, decreases reactor power using Recirc Flow Control Maintain Recirc loop flows matched within 10%

SRO/BOP Monitor EHC pressure and report lowering pressure trend SRO Direct RO to perform either a fast reactor shutdown or to manually scram the reactor prior to trip of the Main Turbine RO Complete reducing core flow to 58 x 10' Ibm/hr and then arm and depress RPS Manual Scram Pushbuttons or arm and depress RPS Manual Scram Pushbuttons prior to automatic Main Turbine trip

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NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 6 Page 1 of 3 Event

Description:

ATWS with SLC squib valve failures Time Position Applicant's Actions or Behavior RO Recognizes incomplete scram and APRMs not downscale and informs SRO/BOP SRO Directs RO/BOP actions per PEI-B13, RPV Control (Non-ATVVS)

- Arms and depresses all RPS Manual Scram PBs

- Places the Reactor Mode Switch in SHUTDOWN

- Starts Hydrogen Analyzers

- Verifies ARI Initiated RO/BOP Executes PEI-B13, RPV Control (Non-ATWS) actions per SRO direction SRO Determines reactor is still NOT shutdown under all conditions without boron SRO Exits PEI-B1 3, RPV Control (Non-ATWS) and enters PEI-B1 3, RPV Control (ATWS)

- Directs initiation of Standby Liquid Control and ADS inhibited

- Directs RO to runback Recirc FCVs to minimum position, then trip Recirc Pumps

- Directs RO to use PEI-SPI 1.3 to insert Control Rods

- Directs BOP/RO to terminate/ prevent injection of inside the shroud systems using PEI-SPI 5.1 & 5.2 (DGs will auto start)

- Directs RO/BOP to line up at least two outside the shroud injection systems using PEI-SPI 6.1 & 6.2 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 6 Page 2 of 3 Event

Description:

ATWS with SLC squib valve failures Time Position Applicant's Actions or Behavior SRO (cont.)

PEI-B13, RPV Control (ATWS) actions:

- Directs actions to maintain MSIVs open per PEI-SPI 2.3 and 2.8

- Directs RPV level stabilized in a band of - 25 to +100"

- Directs pressure band of 800 to 1000 psig Note: RFPTs may trip on L8 following scram. When L8 is reset, water level should be stabilized above L2

" HPCS may be in secured status

" Crew should be able to maintain RPV level > -25 inches until loss of feedwater capability. RCIC will auto start at L2 (+130")

Crew should maintain RPV level > Level 1 (+16.5 inches) to maintain MSIVs open RO Runback Recirc FCVs to minimum position, trips Recirc Pumps Insert Control Rods using PEI-SPI 1.3, Manual Insertion BOP Starts Hydrogen Analyzers Initiates SLC, reports squib valve failures to SRO/RO Inhibits ADS Aligns for outside the shroud injection using PEI-SPI 6.1 & 6.2 Bypasses MSIV LI and IA Isolations per PEI-SPI 2.3 and 2.8 Performs terminate/prevent actions per PEI-SPI 5.1 & 5.2 Maintains reactor pressure band directed by SRO NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 6 Page 3 of 3 Event

Description:

ATWS with SLC squib valve failures Time Position Applicant's Actions or Behavior SRO Directs Alternate Boron Injection per PEI-SPI 1.8 BOP Coordinates Alternate Boron Injection per PEI-SPI 1.8 BOP/RO Maintain level band directed by SRO using Condensate and Feedwater, CRD, and RCIC SRO/BOP/RO Monitor for power > 4% and RPV level > 0" and Supr Pool Temp

> 11 0°F and any SRV open or Drywell pressure > 1.68 psig SRO If ALL above monitored conditions are met, direct terminate and prevent all injection into the RPV except boron and CRD BOP/RO As directed, terminate and prevent injection per PEI-SPI 5.1, 5.2, and 5.3 Depress RCIC Turbine Remote Trip pushbutton SRO When power < 4% or RPV level drops to 0" or all SRVs remain closed and Drywell pressure < 1.68 psig, direct RPV level band between -25" and the level to which RPV was lowered BOP/RO Maintain level band directed by SRO using Condensate and Feedwater, CRD, and RCIC NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 7 Page 1 of 1 Event

Description:

Loss of Feedwater capability Time Position Applicant's Actions or Behavior RO/BOP Report loss of all RFBPs and all Reactor Feed Pumps RO/BOP Monitor and trend reactor water level SRO If not already running, directs RCIC initiation if permitted BOP/RO As directed, start and maximize injection with RCIC BOP/RO Notifies SRO of RPV level approach to -25" SRO Determines RPV level cannot be maintained > -25" PEI-B13, Emergency Depressurization is entered and executed concurrently with PEI-B13, RPV Control (ATWS)

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NUREG-1021, Revision 8, Supplement 1 40 of 40 I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 8 Page 1 of 2 Event

Description:

RPV emergency depressurization / Inject with low pressure ECCS to maintain adequate core cooling. Two ADS/SRVs fail to open Time Position Applicant's Actions or Behavior SRO Directs BOP/RO actions per PEI-B1 3, Emergency Depressurization

- Directs terminate and prevent all injection except boron and CRD

Must hold here until all injection into the RPV has been terminated (except boron and CRD)

BOP/RO As directed, terminate and prevent injection per PEI-SPI 5.3 Depress RCIC Turbine Remote Trip pushbutton SRO Verifies eight or more SRVs are not open Verifies Suppression Pool level is > 5.25 feet Directs all ADS valves opened to rapidly depressurize the RPV BOP/RO When directed, open all ADS valves Determine two ADS/SRVs have failed to open and report to SRO SRO Confirms all ADS valves are NOT open Directs additional SRVs be opened to obtain 8 SRVs open BOP/RO Monitor and trend reactor pressure

Must hold here until RPV pressure is less than MARFP NUREG-1021, Revision 8, Supplement 1 40 of 40 I

I _______________

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2001-01 Scenario No.: 2 Event No.: 8 Page 2 of 2 Event

Description:

RPV emergency depressurization / Inject with low pressure ECCS to maintain adequate core cooling. Two ADS/SRVs fail to open.

Time Position Applicant's Actions or Behavior As soon as RPV pressure is < MARFP, direct injection with SRO outside the shroud systems to restore and maintain RPV level > -25"

LPCI B unavailable for injection outside the shroud BOP/RO Injects into the RPV to systems to restore and maintain RPV level using outside the shroud injection systems as directed

Must hold here until the reactor is shutdown under all conditions without boron

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1-NUREG-1021, Revision 8, Supplement 1 40 of 40 I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Page 1 of 1 Event

Description:

Scenario Termination Criteria Time Position Applicant's Actions or Behavior

1. Control Rods are being inserted

2. The reactor has been emergency depressurized

3. Low pressure ECCS Systems are being used to restore RPV level above -25 inches I

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4 8, Supplement 1 40 of 40 i

i NUREG-1021, Revision

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 6 Page 1 of 1 Event

Description:

Critical Task #1 Time Position Applicant's Actions or Behavior Critical Task #1 - With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron and/or inserting control rods

1. Safety Significance:

- Shutting down the reactor can preclude failure of Containment or equipment necessary for the safe shutdown of the plant

2. Cues:

- Procedural compliance

3. Measured by:

- SLC Pump control switches taken to ON and control rod insertion before the end of the scenario

4. Feedback:

- Reactor power trend 4

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1-NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 6 Page 1 of 1 Event

Description:

Critical Task #2 Time Position Applicant's Actions or Behavior Critical Task #2 - With a reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS

1. Safety Significance:

- Precludes core damage due to an uncontrolled reactivity addition

2. Cues:

- Procedural compliance

3. Measured by:

- ADS logic inhibited prior to an automatic initiation of the ADS System unless all required injection systems are terminated and prevented

4. Feedback:

- RPV pressure and level trends

- ADS "Out of Service" annunciator status I

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NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 8 Page 1 of 1 Event

Description:

Critical Task #3 Time Position Applicant's Actions or Behavior Critical Task #3 - When RPV water level cannot be maintained

> -25" and the reactor is at pressure, initiate Emergency Depressurization

1. Safety Significance:

- Maintain adequate core cooling

2. Cues:

- Procedural compliance

- Level lowering without adequate high pressure injection available 3 Measured by:

- Observation - at least 5 SRVs open prior to re-establishing injection after terminate and prevent actions are completed

4. Feedback:

- Reactor pressure trend

- Suppression Pool temperature trend r

t NUREG-1021, Revision 8, Supplement 1 40 of 40 I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 8 Page 1 of 1 Event

Description:

Critical Task #4 Time Position Applicant's Actions or Behavior Critical Task #4 - During an ATWS with Emergency Depressurization required, terminate and prevent injection, with the exception of SLC and CRD, into the RPV until reactor pressure is below MARFP

1. Safety Significance:

- Prevention of fuel damage due to uncontrolled feeding

2. Cues:

- Procedural compliance

3. Measured by:

- Observation - no injection into the RPV except for SLC and CRD prior to reaching the MARFP that causes a reactor short period alarm or power increase to APRM upscale alarms

4. Feedback:

- Reactor power trend, power spikes, reactor short period alarms NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 2 Event No.: 8 Page 1 of 1 Event

Description:

Critical Task #5 Time Position Applicant's Actions or Behavior Critical Task #5 - When RPV water level cannot be maintained during an ATWS, when RPV pressure is below the MARFP, slowly increase and control injection into the RPV to restore and maintain RPV level above the TAF

1. Safety Significance:

- Establish adequate core cooling

2. Cues:

- RPV pressure trend

- Procedural compliance

3. Measured by:

- RPV level is established and controlled above the TAF

4. Feedback:

- Lack of power excursion

- Lack of Hydrogen generation

- RPV level and pressure indications I.

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4 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Scenario Outline Form ES-D-1 Facility:

Perry Scenario No.: 3 Examiners:

Op-Test No.: 2002-01 Operators:

Initial Conditions: Reactor startup is in progress with the plant at 5% of rated power. RHR B is in secured status for preventive maintenance on the pump breaker. The OPRMs are functional but are inoperable per Tech. Spec.

3.3.1.3. Required Action A.3 has been implemented.

Turnover: Plant startup continues: withdraw control rods to 10% power, transfer the Reactor Mode Switch to RUN, and continue power ascension. All required MODE 1 change paperwork has been reviewed and approved.

Target Critical Tasks: Emergency Depressurization, RPV Flooding to restore and maintain adequate core cooling Event Malf. No.

Event Event No.

Type*

Description 1

R (RO)

Increase reactor power to 10% using control rods RD01:R11043 C (RO)

Control rod 10-43 stuck at position 8 8%

2 NM02H I (RO)

IRM H failure upscale (bypass failed IRM) 100%

(TS 3.3.1.1 and OR 6.2.3) 3 N (RO)

Verify NI overlap / Transfer Reactor Mode Switch to RUN I Withdraw IRMs 4

RD17A C (BOP)

CRDH Pump A trip due to loss of lube oil.

50%

N (BOP)

Perform CRD Pump trip recovery RD05R5443 Accumulator fault HCU 54-43 (TS 3.1.5) (1 minute time delay) 5 CP02:

C (BOP)

Service Water Pump 'B' trip due to shaft seizure (start standby Service Water OP41CO01B Pump) 6 bat C (All)

Seismic Event (OBE) or/seismic_2 RPO1A C (All)

RPS A' EPA Breaker Trip (loss of RPS Bus 'A') (30 second time delay)

TH02A / TH02B M (All)

Recirc Loop pipe rupture (reactor scram on high Drywell pressure) 100%

(TH02A - 6 minute time delay & 5 minute ramp)

(TH02B - 8 minute time delay & 5 minute ramp)

MV08:

C (BOP)

NCC Drywell Isolation Valve P43-F215 failure when valve becomes fully closed OP43F0215 RD01 R4219 C (RO)

Control rod 42-19 stuck at position 12 (during scram) 7 bat I (All)

Loss of all RPV level indication msllossleve12 v.-,,

J

M (All)

Emergency Depressurization I RPV Flooding to restore and maintain adequate core cooling

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 3 Objectives:

The crew will continue the startup. Prior to placing the Reactor Mode Switch in RUN, control rod 10-43 will not withdraw using normal drive water pressure. This will require the crew to take action per ONI-C1 1, Inability to Move Control Rods, and SOI-Cll (RCIS) to free the stuck control rod.

After the control rod is moving normally, IRM H will fail upscale resulting in a RPS half-scram. The startup will be placed on hold while Tech Specs are referenced, the IRM is bypassed, and the half-scram is reset.

The startup continues and after the Reactor Mode Switch has been placed in RUN, the running CRDH Pump will trip due to a loss of lube oil, requiring the standby CRDH Pump to be started. While performing CRD pump trip recovery, an HCU accumulator fault alarm will be received requiring the crew to monitor for additional accumulator fault alarms.

After the standby CRDH Pump is started, Service Water Pump B will trip.

ONI-P41, Loss of Service Water, will be entered requiring manual start of a standby Service Water Pump.

Immediately after a standby Service Water Pump is started, a seismic event occurs which causes an RPS EPA breaker to trip and a Recirc Loop pipe break in the Drywell, resulting in rising Drywell pressure and temperature. NCC Drywell Isolation Valve P43-F215 will fail when it is fully closed further degrading the crew's ability to control Drywell temperature.

The reactor will be manually scrammed or will automatically scram; however, one control rod will fail to insert.

Following the scram, rising Drywell temperature will result in a loss of all level indication. RPV Flooding and Emergency Depressurization is performed and low pressure injection systems are used to maintain adequate core cooling.

Discussion of Safety Significance for scenario 3:

As the startup continues, prior to placing the Reactor Mode Switch in RUN, one control rod will not withdraw using normal drive water pressure requiring the crew to take action to get the control rod to move. This is safety significant because the actions directly affect core reactivity.

After the control rod is unstuck, an IRM will fail upscale. This is safety significant because it will result in a RPS half-scram, require the crew to determine the IRM is inoperable, bypass the IRM, and reset the half-scram.

Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 3 As the startup continues and the Reactor Mode Switch has been placed in RUN, the running CRDH Pump will trip that requires the standby CRDH Pump to be started. An accumulator fault alarm will also occur. This is safety significant because Tech. Specs. would require a manual reactor scram after 20 minutes if an additional HCU accumulator fault alarm were to be received.

After the standby CRDH Pump is started, a Service Water Pump will trip. This will require a manual start of a standby Service Water Pump to avoid high temperatures on components and systems cooled by the Service Water System.

After a standby Service Water Pump has been started, a seismic event occurs which causes a RPS EPA breaker to trip and a Recirc pipe break in the Drywell.

This is safety significant because the reactor will be manually scrammed or will automatically scram following the seismic event. One control rod will fail to insert following the scram. The failure of the control rod to insert is safety significant because it will require the crew to make a timely assessment of control rod positions to determine if A TWS actions are necessary.

Following the scram, conditions in the Drywell will cause a loss of all level indication. This is safety significant because the crew must enter RPV Flooding and emergency depressurize the RPV in order to allow low-pressure injection systems to be used to maintain adequate core cooling.

Final - Revision 0 Page 2 of 2

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 1 Page 1 of 1 Event

Description:

Power ascension continues. Control Rod 10-43 stuck at position 8.

Time Position Applicant's Actions or Behavior SRO Conduct reactivity brief Provides SRO oversight for power ascension RO Withdraw control rods as directed in accordance with the approved rod withdrawal sequence RO/BOP Recognize Control Rod 10-43 will not move and inform SRO SRO Acknowledge alarm and report of immovable Control Rod 10-43 Enter and execute SOl-Cl 1 (RCIS), Section 7.9.2:

- Directs RO to raise CRDH drive water differential pressure in 50 psid increments until control rod motion is achieved RO Raise CRDH drive water d/p.-as directed, attempts Control Rod motion Reports Control Rod movement to SRO SRO Directs RO to return CRDH drive water d/p to normal band NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 2 Page 1 of 1 Event

Description:

IRM H failure upscale / Bypass failed IRM.

Time Position Applicant's Actions or Behavior RO/BOP Recognizes and reports IRM H failure upscale Recognizes and report RPS half-scram Recognizes and reports control rod withdrawal block Closely monitors remaining IRMs BOP Assists RO by consulting ARIs:

- ARI-H13-P680-6 (E3), IRM UPSCALE TRIP/INOP

- ARI-H13-P680-6 (C2), ROD BLOCK IRM UPSCALE

- ARI-H13-P680-5 (B9), % SCRAM B/D

-ARI-H13-P680-5 (B7), RPS NEUTRON MON TRIP

- ARI-H13-P680-5 (El0), ROD WITHDRAWAL BLOCK SRO Acknowledge report of IRM failure, /2 scram, and rod block Suspends Control Rod withdrawal Directs IRM back panel indications checked SRO/BOP Requests I&C and Responsible System Engineer assistance in the Control Room to support troubleshooting Consults Tech Spec 3.3.1.1 and ORM 6.2.3 Direct RO to bypass IRM H and reset 1/2 scram Notifies Operations Management of IRM failure and actions taken RO Bypass IRM H per SOI-C51 (IRM), Section 7.1 Reset half-scram B/D per SOI-C71, Section 7.3 Observes IRM, Rod Block, and RPS alarms clear NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 3 Page 1 of 3 Event

Description:

Verify NI overlap / Transfer Reactor Mode Switch to RUN / Withdraw IRMs.

Time Position Applicant's Actions or Behavior SRO Direct RO to continue Control Rod withdrawal Provides SRO oversight for power ascension RO As directed, continue Control Rod withdrawal in accordance with approved Startup Pullsheets SRO Prepare for transfer of the Reactor Mode Switch to RUN:

- Verify Main Steam pressure > 807 psig, and the MSL ISOL MAIN STEAM LINE PRESSURE LOW annunciator has cleared

- Verify Condenser pressure is less than 21.5 inches HgA and the MSL ISOL MAIN CONDENSER VACUUM LOW annunciator has cleared

- Verify APRM ROD BLOCK DOWNSCALE annunciator cleared Consults ARI-H13-P601-19 (A7)

- Verify reactor water level is in the range of 192" to 200" II I _______________

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NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 3 Page 2 of 3 Event

Description:

Verify NI overlap / Transfer Reactor Mode Switch to RUN / Withdraw IRMs.

{TimeJ Position Applicant's Actions or Behavior SRO/RO Verify IRM/APRM overlap per 101-1, Section 4.9.3.f SRO When IRM/APRM overlap has been verified, direct transfer of the Reactor Mode Switch to the RUN position RO Transfers Reactor Mode Switch to the RUN position SRO Directs transfer of all IRM/APRM Recorder Select Switches to APRM RO Transfers all IRM/APRM Recorder Select Switches to APRM SRO Directs verification that APRM rod block setpoints have transferred to the flow biased setpoints RO/BOP Verifies APRM rod block setpoints transferred to the flow biased setpoints SRO Verify that the MSIV CLOSURE SCRAM BYP annunciator H13-P680-5 (A6) cleared Direct withdrawal of all IRM detectors to the full out position and placement of IRMs on Range 3

+

+/-

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 3 Page 3 of 3 Event

Description:

Verify NI overlap / Transfer Reactor Mode Switch to RUN / Withdraw IRMs.

Time Position Applicant's Actions or Behavior RO Withdraw all IRM detectors to the full out position and place on Range 3 per SOI-C51 (IRM), Section 5.1 SRO As necessary, direct RO to continue to withdraw Control Rods per FTI-B02 to maintain Bypass Valve 1 approximately 75% open RO As directed, re-commences withdraw of Control Rods per FTI-B02 I

4 I.

4 4

.4 I.

.4 4

.4 NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 4 Page 1 of 2 Event

Description:

CRDH Pump A trip due to loss of lube oil. Perform CRD Pump trip recovery. Accumulator fault HCU 54-43.

Time Position Applicant's Actions or Behavior RO/BOP Responds to, reports, and references ARIs for the following alarms:

H13-P601-22 (C3), CRD SYS CHARGING WATER PRESSURE LOW H13-P601-22 (D2), CRD PUMP AUTO TRIP H13-P601-22 (F2), CRD PUMP A TRIP OIL PRESS LOW H1 3-P877-1 (G3), Bus XH1 I BREAKER TRIP Recognizes and reports CRDH Pump A trip/ Aux Oil Pump running RO/BOP Dispatches NLOs to CRDH Pump A and breaker to investigate RO/BOP/SRO Receives report from NLO of a broken oil line on CRDH Pump A BOP Recognizes Aux Oil Pump running and secures Aux Oil Pump BOP/RO/SRO Requests Maintenance and Responsible System Engineer assistance in the Control Room to support troubleshooting RO Responds to, reports, and references ARIs for alarm H13-P601-22 (Al) CRD MECHANISM TEMP HIGH SRO Acknowledges CRDH Pump trip and receipt of unexpected alarms Enters ONI-C1-1, Inability to Move Control Rods

- Directs plant parameters maintained as steady as possible

- Directs CRD Pump trip recovery per SOI-C1 1 (CRDH)

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 4 Page 2 of 2 Event

Description:

CRDH Pump A trip due to loss of lube oil. Perform CRD Pump trip recovery. Accumulator fault HCU 54-43.

Time Position Applicant's Actions or Behavior SRO (cont.)

- Directs RO to monitor for HCU accumulator fault alarms RO Observes and reports accumulator fault on HCU 54-43 SRO Acknowledges accumulator fault on HCU 54-43 Directs RO to monitor for a second HCU accumulator fault BOP Performs CRD Pump trip recovery per SOl-Cl 1 (CRDH)

- Take CRD Aux Oil Pump B, 1C1 1-CO02B to Start

- Place CRD Hydraulics Flow Control, 1 C11-R600, in Manual

- Using CRD Hydraulics Flow Control, 1 C11-R600, fully close Flow Control Valve, 1 C11 -FO02B

- Take CRD PUMP B, 1 C11-C01 B, to Start and observes:

increasing amps for CRD Pump B

- CHARGING WATER LOW PRESSURE alarm clears

- Slowly throttle open CRD Flow Control Valve until flow is returned to the pre-transient setting on CRD Hydraulics Flow Control, 1C11-R600

- Coordinate with NLO to complete CRD Pump trip recovery per SOI-Cl1 (CRDH)

SRO Notifies OPS Management and NRC Resident of ONI entry and reason for entry May direct CRDH Pump A and Aux Oil Pump breakers racked out NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 5 Page 1 of 2 Event

Description:

Service Water Pump B trip due to shaft seizure. / Manual start of standby Service Water Pump.

Time Position Applicant's Actions or Behavior BOP Responds to, reports, and references ARIs for alarms:

H13-P970-1 (B8), SW PUMP DISCH HEADER PRESSURE LOW H13-P877-1 (G3), BUS XH11 BREAKER TRIP Recognizes and reports Service Water Pump B trip BOP/RO Dispatches NLOs to Service Water Pump B and pump breaker BOP/RO/SRO Requests Maintenance and Responsible System Engineer assistance in the Control Room to support troubleshooting SRO Acknowledges report of Service Water Pump B trip Enters ONI-P41, Loss of Service Water

- Directs BOP to start the standby Service Water Pump per SOI-P40/41, Section 5.1 Notifies OPS Management and NRC Resident of ONI entry and reason for entry Directs BOP to complete shutdown of the Service Water Pump B per SOI-P40/41, Section 6.2 I.

I.

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 5 Page 2 of 2 Event

Description:

Service Water Pump B Trip due to shaft seizure. / Manual start of standby Service Water Pump.

Time Position Applicant's Actions or Behavior BOP Start standby Service Water Pump per SOI-P40/41 Section 5.1, Shifting Service Water Pumps

- Take SW Pump Discharge Valve control switch to Open and press the Stop button when the blue light comes on

- Take SW Pump control switch to START

- When SW Pump amps stabilize, take Discharge Valve control switch to Open

- Throttle NCC HX SW Bypass Valve, P41-F400, as necessary to maintain discharge pressure of all operating SW Pumps at 55-60 psig

- Notify Chemistry to place the Service Water Chlorination System in operation per SOI-P48 BOP Complete shutdown of Service Water Pump B per SOI-P40/41, Section 6.2:

- Take SW Pump Discharge valve control switch to Close and press the Stop button when the blue light comes on

- Take Discharge Valve control switch to Close (Note: Section 6.2 may not be performed to facilitate troubleshooting of Service Water Pump B trip)

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 6 Page 1 of 4 Event

Description:

Seismic Event leading to steam leak in Drywell and reactor scram. Control Rod 42-19 stuck at position 12 (during scram).

Time Position Applicant's Actions or Behavior RO/BOP Responds to, reports, and references ARIs for alarms:

H13-P680-8 (B3), SEISMIC ALARM P969 H13-P680-8 (C3), SEISMIC MOMITOR TROUBLE Recognizes half scram and Div 1 BOP isolation and informs SRO Determines EPA Breaker trip and informs SRO RO/BOP Recognizes rising Drywell pressure and informs SRO SRO Acknowledges Seismic Monitor alarms, report of 11/2 scram, Div 1 BOP isolation, and rising Drywell pressure RO/BOP Dispatches NLO to investigate cause of 1/2 scram and Div 1 BOP isolation Observes and reports CVCW A trip (due to BOP isolation)

Observes and reports RWCU isolated Observes and reports 1/2 MSIV isolation, and Outboard MSIVs have lost position indication ROIBOP/SRO Receive multiple reports of earthquake from plant personnel RO/BOP Monitor and trend rising Drywell pressure SRO Enters ONI-D51, Earthquake:

- Verifies earthquake is greater than OBE by verifying amber and red lights on panel H13-P969

- Directs all Emergency Service Water Pumps started NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 6 Page 2 of 4 Event

Description:

Seismic Event leading to steam leak in Drywell and reactor scram. Control Rod 42-19 stuck at position 12 (during scram).

Time Position Applicant's Actions or Behavior SRO (Cont)

- Directs all Plant Underdrain Pumps stopped

- Directs plant personnel to begin walkdown of plant areas

- Directs RO and BOP to check major plant variables Enters ONI-C71-2, Loss of RPS Bus BOP/RO Check and monitor major plant variables BOP Checks panel H1 3-P969, informs SRO amber and red lights are on Starts all ESW Pumps, as directed by the SRO Contacts NLOs to secure Plant Underdrain Pumps SRO Due to rising Drywell pressure orders Rx scram, enters ONI-C71-1

If not manually scrammed, reactor automatically scrams at 1.68 psig (a LOCA signal also occurs at 1.68 psig)

RO If directed, arms and depresses RPS Manual Scram Pushbuttons prior to 1.68 psig Drywell pressure Recognizes incomplete scram, determines one Control Rod not inserted and informs SRO, report APRMs are downscale SRO Enters PEI-B13, RPV Control (Non-ATWS) due to RPV < L3 and Drywell pressure > 1.68 psig and enters PEI-T23, Containment Control due to Drywell pressure > 1.68 psig NUREG-1021, Revision 8, Supplement 1 40 of 40 I ___

I _______________

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 6 Page 3 of 4 Event

Description:

Seismic Event leading to steam leak in Drywell and reactor scram. Control Rod 42-19 stuck at position 12 (during scram).

Time Position Applicant's Actions or Behavior SRO Directs PEI-B13, RPV Control (Non-ATWS)

- Verifies reactor is scrammed

- Confirms Reactor Mode Switch is in SHUTDOWN

- Start Hydrogen Analyzers

- Verifies reactor shutdown under all condition without boron

- Verifies SRMs and IRMs inserted

- Directs pressure control 800 to 1000 psig SRO RPV Level Control

- Restores and maintains RPV level between 185 and 215 inches

"* Feedwater - available

"* CRD - available

"* RCIC - available

"* HPCS - available RPV Pressure Control

- Confirms no SRVs are cycling

- Stabilizes RPV pressure 800 to 1000 psig using Bypass Valves

- Override low pressure ECCS Pumps per PEI-SPI 5.2

Condensate and Feedwater will restore RPV level to 185 to 215" RO/BOP Executes PEI-B13, RPV Control (Non-ATWS) actions per SRO direction NUREG-1021, Revision 8, Supplement 1 40 of 40 I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 6 Page 4 of 4 Event

Description:

Seismic Event leading to steam leak in Drywell and reactor scram. Control Rod 42-19 stuck at position 12 (during scram).

Time Position Applicant's Actions or Behavior SRO Directs actions per PEI-T23, Containment Control, when Drywell pressure reaches 1.68 psig Drywell Temperature Control

- Operates all available DW cooling

- Restores NCC to the DW Drywell & Containment Pressure Control

- Maintains Containment pressure below PSP Containment Temperature Control

- Operates all available Containment cooling

- Restores CVCW System

- Maintains Containment average temperature less than 185°F Suppression Pool Level Control

- Restores and maintains SP level between 17.8 and 18.5 ft Suppression Pool Temperature Control

- Maintains both SP average temperature and RPV pressure below HCL RO/BOP Executes PEI-T23 actions per SRO direction Reports P43-F215, NCC Containment Return Inboard Isolation Valve will NOT open (P43-F215 is failed closed)

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 7 Page 1 of 1 Event Description Loss of all RPV level indication. Emergency Depressurization and Injection systems to restore and maintain adequate core cooling.

Time Position Applicant's Actions or Behavior BOP/RO Recognizes loss of all level indication and informs the SRO SRO Determines that RPV level cannot be determined and transitions to PEI-B1 3, RPV Flooding

- Verifies all Control Rods inserted

- Verifies Suppression Pool level greater than 5.25 feet

- Directs eight (8) ADS SRVs to be opened RO/BOP Open eight (8) ADS SRVs, when directed SRO Directs actions to isolate the reactor vessel RO/BOP Closes MSIV's, MSL drains and RCIC steam valves, when directed SRO Directs RO/BOP to inject to establish RPV pressure 60 psig greater than Containment pressure and at least five (5) SRVs open RO/BOP Starts designated systems and injects to vessel when directed:

RFBPs (PEI-SPI 2.7) or HPCS (PEI-SPI 2.4) are available and low pressure ECCS Systems are available BOP/RO Maintain RPV pressure 60 psig greater than Containment pressure and at least five (5) SRVs open NUREG-1021, Revision 8, Supplement I 40of40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Page 1 of 1 Event

Description:

Scenario Termination Criteria Time Position Applicant's Actions or Behavior

1. Control Rods are fully inserted

2. The reactor has been emergency depressurized

3. Low pressure ECCS Systems are being used to establish RPV pressure at least 60 psig above Containment pressure

4. Containment and Drywell parameters are being restored per PEI-T23, Containment Control 4

4 4

NUREG-1021, Revision 8, Supplement 1 40 of 40 i ________________

i __________________________________________________________________________

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 7 Page 1 of 1 Event

Description:

Critical Task #1 Time Position Applicant's Actions or Behavior Critical Task #1 - When RPV water level cannot be determined and the reactor is at pressure, initiate Emergency Depressurization

1. Safety Significance:

- Maintain adequate core cooling

2. Cues:

- Procedural compliance

- Loss of all water level indication

3. Measured by:

- Observation - at least 5 SRVs open prior to re-establishing injection after terminate and prevent actions are completed

4. Feedback:

- Reactor pressure trend

- Suppression Pool temperature trend

+

1 I.

I I.

I L

I I

4 I

+

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 3 Event No.: 7 Page 1 of 1 Event

Description:

Critical Task #2 Time Position Applicant's Actions or Behavior Critical Task #2 - When RPV water level cannot be determined with RPV pressure below the MRFP, increase and control injection into the RPV to restore and maintain RPV pressure above the MRFP

1. Safety Significance:

- Establish adequate core cooling

2. Cues:

- RPV pressure trend

- Procedural compliance

3. Measured by:

- RPV pressure is established and controlled above the MRFP

4. Feedback:

- Lack of water level indication

- Lack of Hydrogen generation

- RPV pressure indications 4

4.

4 4.

4 1-NUREG-1021, Revision 8, Supplement 1 40 of 40

I Perry NRC Exam LC 00-01 Scenario 3 Simulator Setup and Cues

1.

Simulator Setup

a.

Initial Conditions

1)

2)

3)

4)

5)

6)

7)

8)

9)

10)

11)

12)

13)

14) 15)

Reset to IC93 Using Simulator Startup Pullsheet, withdraw all rods in step 47 Unbypass IRM Place the Off-Gas Adsorbers in service per SOI-N64/62, Section 4.6 Complete the Main Turbine warmup (pressurize the steam chest) per SOI-N32/39/41/51, Section 4.4, steps 24-27 Snap IC setup for future use. Record IC # 88 Update 101-1 to Section 4.9, Step 1 and sign-off all applicable Mode Change steps Update Simulator Startup Pullsheet, step 48, gang 40 at notch 08 Place copy of OPRM PLCO (Form 7158) in ALCO/PLCO Tracking Book Perform annunciator test Load Batch File LNCO001-3 Load Event Trigger File CRDDRIVEP Load Override File SEISMIC_2 Load Malfunction File LOSSLEVEL2 Verify no Triggers went active

b.

Special Procedures

1)

SOI-C 11(RC&IS) Attachment 11

2.

Batch File LNCOOO1-3

a.

IMFRDOIRI043

b.

IMF RD01R4219

c.

IMF MV08:OP43F0215

d.

IMF NM02H

e.

IMF RD I7A

f.

IMF RD05R5443

g.

IMF CP02: OP41COOOB

h.

IMF RP01A

i.

IMF TH02A

j.

IMF TH02B Active Control rod 10-43 stuck at position 08 Active Control rod 54-43 stuck at position 12 Active NCC Drywell Isolation Valve P43-F215 failure when fully closed TRG I IRM H failure upscale 100% severity TRG 2 CRD Pump A loss of lube oil 50% severity TRG 2 HCU 54-43 accumulator fault 1 minute time delay TRG 3 Service Water Pump 'B' shaft seizure TRG 4 RPS 'A' EPA Breaker trip 30 second time delay TRG 4 Recirc Loop 'A' pipe break 100% severity, 6 minute time delay, 5 minute ramp TRG 4 Recirc Loop 'B' pipe break 100% severity, 8 minute time delay, 5 minute ramp I

2

3.

Event Trigger File

a.

CRDDRIVEP RDXMlC11NO008.GT. 295

4.

Override File

a.

SEISMIC_2

see attached printout of file

5.

Malfunction File

a.

LOSSLEVEL2

see attached printout of file

6.

Commands

a.

TRG 4 = ROR SEISMIC_2

b.

TRG 5 = RMF LOSSLEVEL2

c.

TRG 6 CRDDRIVEP

d.

TRG 6 = DMF RDO0R1043 2

3 Instructor Cues:

Event I None

Control Rod 10-43 malfunction will automatically delete itself when CRD drive water d/p is raised above 295 psig.

Event 2 Insert TRG 1 when directed during Control Rod withdrawal.

As I&C, report that a Work Order will have to be initiated in order to troubleshoot IRM H failure.

Event 3 None Event 4 Insert TRG 2 when directed.

As NLO / RSE, report that CRD Pump A has a lube oil leak.

As NLO, report that HCU 54-43 accumulator pressure is 1480 psig.

Event 5 Insert TRG 3 when directed.

As NLO, report that breaker XH 1203 for Service Water Pump B has tripped on overcurrent on all 3 phases.

Event 6 Insert TRG 4 when directed.

As various plant personnel, report that you have felt an earthquake within the plant.

As NLO, report that RPS A EPA Breaker 1C71-SO03A is tripped, reason unknown.

As NLO, report that disconnect EF1C07-NN for 1P43-F215 has a blown main line on Phase A.

Event 7 Insert TRG 5 when Drywell temperature reaches 212 'F.

3

Appendix D Scenario Outline Form ES-D-1 Facility:

Perry Scenario No.: 4 Examiners:

Op-Test No.: 2002-01 Operators:

Initial Conditions: The plant is operating at 100% power. RHR B is in secured status for preventive maintenance on the pump breaker. RHR B was declared inoperable five hours ago per Tech. Spec.3.5.1, Action A; 3.6.1.7, Action A; and 3.6.2.3, Action A. The OPRMs are functional but are inoperable per Tech. Spec. 3.3.1.3. Required Action A.3 has been implemented.

Turnover: 1. Shift NCC Pumps (start NCC Pump C and shutdown NCC Pump A).

Target Critical Tasks: Manually start RHR Pump A (failure to auto start), Emergency Depressurization Event Malf. No.

Event Event No.

Type*

Description 1

N (BOP)

Shift NCC Pumps (start NCC Pump C and shutdown NCC Pump A) 2 CN02:

I (BOP)

RFPT A Lube Oil Temp controller failure in Auto mode 1 P44R0450 0%

3 PT02:

I (RO)

Reactor Narrow Range Level Transmitter N004A Offset (3 minute ramp) 1C34N0004A (ORM 6.2.1.3) 17%

4 RF FW66 C (RO)

RFPT A spurious trip TH12B C (RO)

Reactor Recirculation FCV B runback failure (TS 3.4.1) 5 CP02:

C (BOP)

TBCC Pump A trip (start standby TBCC Pump) 1P44C0001A SW03 C (ALL)

TBCC System Process Piping Leakage (1 minute time delay and 3 minute ramp) 25%

Fast reactor shutdown required R (RO)

Decrease reactor power to 66% using recirc flow (58 Mlbs/hr) 6 TH28 M (All)

MSL Break in Drywell 1%

PCO1A DW/CNTMT Bypass Leakage (to be modified in Event #7) 0%

CB04:

C (BOP)

RHR Pump A fails to auto start on Drywell high pressure (required for 1 E12C0002A Containment Spray mode)

CB01:

C (BOP)

RHR PumpAtdips when-flow is aligned to cQntainrnent spray.

1E 12C0002A M (All)

RPV emergency depressurization to control Containment pressure PT01:

I (RO)

Reactor Level Transmitter N081 C failure downscale 1B21NO081C 5%

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Final - Revision 1

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 4 Objectives:

With the plant operating at 100% power, the BOP operator will shift NCC Pumps.

After NCC Pump C has been placed in service and NCC pump A has been shutdown, RFPT A Lube Oil Temp controller will fail closed in Auto requiring the crew to place the controller in Manual and restore lube oil temperature to normal.

After lube oil temperature is restored, the Narrow Range level transmitter for the in-service channel fails low (offset). The crew enters ONI-C34, Feedwater Flow Control Malfunction, to select an operable Narrow Range level channel and restore level to normal.

When conditions have stabilized, RFPT A will trip. During the transient, Reactor Recirculation FCV B will fail to automatically runback requiring Technical Specifications to be referenced for a loop flow mismatch.

After level and power are stabilized, TBCC Pump A will trip. ONI-P44, Loss of TBCC, will be entered requiring manual start of the standby TBCC Pump.

Immediately following start of the standby TBCC Pump, a TBCC System piping leak will be initiated and grow progressively worse until a fast reactor shutdown is required.

Following the scram, a MSL pipe break in the Drywell occurs resulting in an MSIV isolation and rising Containment pressure.

When signaled to start on high Drywell pressure, RHR Pump A will not automatically start and must be manually started. When aligned for containment spray, RHR Pump A breaker will trip. Eventually the RPV must be depressurized to control the Containment pressure rise. During emergency depressurization, Reactor Level Transmitter N081C will fail downscale.

Discussion of Safety Significance for scenario 4 After NCC Pump C has been placed in service and NCC Pump A has been shutdown, RFPT A Lube Oil Temp controller will fail closed. This is safety significant because failure to recognize and correct the failure would eventually result in RFP bearing damage and a loss of the RFP, thereby challenging RPV water level control.

Final - Revision 0 Page I of 2

2002 Perry NRC Examination Scenario Objectives Safety Significance Discussion Scenario 4 After lube oil temperature is restored, the Narrow range level transmitter for the in-service channel fails low (offset). This is safety significant because it will require manual operation of the Feedwater System to select an operable Narrow Range level channel and retum RPV level to normal.

When conditions have stabilized, RFPT A will trip and during the transient, Reactor Recirculation FCV B will fail to automatically runback. The RFPT A trip is safety significant because a reactor scram could result if the expected plant response is not verified. Failure of Reactor Recirculation FCV B to automatically runback is safety significant because Recirc Loop Flows will not be matched as required by Tech. Specs.

After level and power are stabilized, TBCC Pump A will trip and operator action will be required to manual start the standby TBCC Pump to allow for continued plant operation. Manual start of the standby TBCC Pump will also lead to a TBCC System piping leak that will grow progressively worse and require the crew to initiate a fast reactor shutdown.

Following the scram, a MSL pipe break in the Drywell occurs resulting in an MSIV isolation and rising Containment pressure. This is safety significant because PEI-B 13, RPV Control (Non-A TWS,) and PEI-T23, Containment Control, must be entered and executed to maintain key Containment and RPV parameters.

When signaled to start on high Drywell pressure, RHR Pump A will not automatically start. This is safety significant because RHR Pump A must be manually started to assure it will perform its LPCI design function if required.

The RHR Pump A trip is safety significant because the crew must determine that RHR flow will be unavailable for containment spray, requiring RPV emergency depressurization to control the Containment pressure rise. The Reactor Level Transmitter downscale failure is safety significant because the failure could complicate RPV level control during and following the emergency depressurization.

Final - Revision 0 Page 2 of 2

Perry Plant Initial License NRC examination Operating Examination Risk Significance Each candidate will be examined in a dynamic simulator setting on a minimum of 2 events identified in the Perry IPE.

The dynamic simulator evaluation contains the following Perry IPE events.

Transient with a Loss of Power Conversion System:

"* MSIV Isolation

"* Various High Pressure Injection Systems unavailable

"* RPV Emergency Depressurization required

"* RPV level control with the ECCS low pressure systems Intermediate LOCA Events:

"* Steam Leak in Drywell

"* Steam Break in Containment

"* Failure of Long-Term Containment Heat Removal with RHR ATWS:

"* Failure to Insert All Control Rods

"* Failure of RPV Level Control

"* Failure of Standby Liquid Control (SLC) System

"* RPV Emergency Depressurization required

"* RPV level control with the ECCS low pressure systems Additionally, two of four scenarios contain transient events that could lead to Perry IPE events if operator action is inappropriate or ineffective:

"* IORV transient

"* Trip of Service Water Pump SRO candidates will be required to determine Risk related to equipment availability as part of the Administrative Examination.

Confidential Page 1 11/15101

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 1 Page 1 of 1 Event

Description:

Shift NCC Pumps (start NCC Pump C and shutdown NCC Pump A)

Time Position Applicant's Actions or Behavior SRO Direct BOP to place NCC Pump C in service and secure NCC Pump A per SOI-P43 BOP Shift NCC Pumps per SOI-P43, Section 5.1

- Directs NLO to throttle NCC Pump C Discharge VIv, P43-F513C, To 10% open

- Take NCC Pump C control switch to START Observes P43-R352, Pump Amps Observes P43-R026C, Discharge Pressure

- Directs NLO to open NCC Pump C Discharge VIv, P43-F513C Observes P43-R352, change in pump amps Observes P43-R026C, change in discharge pressure

- Verify header pressure has stabilized between 94 and 123 psig

- Directs NLO to throttle NCC Pump A Discharge VIv closed until it Is 2% open

- Immediately take NCC Pump A control switch to STOP

- Directs NLO to open NCC Pump A Discharge VIv, P43-F513A

- Directs NLO to verify proper discharge check valve operation by confirming no reverse pump rotation BOP Verify NCC System parameters in accordance with SOI-P43 Section 5.0

1.

1

+

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 2 Page 1 of 1 Event

Description:

RFPT A Lube Oil Temperature Controller failure in Auto mode Time Position Applicant's Actions or Behavior BOP Respond to, report, and reference ARI for Alarm H1 3-P870-8 (A2),

RFPT A LUBE OIL CLR OUTLET TEMP HIGH BOP Determine RFPT LUBE OIL TEMP CONTROL A, 1P44-R450 has failed in Auto mode SRO Acknowledge alarm report and RFPT A Lube Oil Temperature Controller failure in the Auto mode BOP Place the RFPT LUBE OIL TEMP CONTROL A, 1 P44-R450, in Manual and increase cooling water flow Inform SRO that 1 P44-R450 has been placed in Manual and cooling water flow has been increased to RFPT Lube Oil Cooler SRO In accordance with Operations Expectations:

- Assign BOP operator as clear "owner" to closely monitor RFPT Lube Oil temperature

- Consider placing an Information Tag on RFPT A LUBE OIL TEMP CONTROL, 1 P44-R450 BOP Closely monitor RFPT A Lube Oil temperature BOP/RO Request I&C and Responsible System Engineer assistance to support troubleshooting NUREG-1021, Revision 8, Supplement 1 40 of 40 I

I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 3 Page 1 of 1 Event

Description:

Reactor Level Transmitter C34-NO04A Offset Time Position Applicant's Actions or Behavior "0OP/SRO Recognizes abnormal water level indication Refers to RPV Level Validation Screen on SPDS as necessary ROIBOP/I0 Requests I&C and Responsible System Engineer assistance in the Control Room to support troubleshooting BOP Supports RO by attending to alarms and consulting ARI-H1 3-P680-3 (B6) and other ARIs as required SRO Enters ONI-C34, Feedwater Flow Control Malfunction.

- Directs RO to take manual control of RFPTs and maintain RPV level 192 to 200 inches (201" required for Level Program)

- After crew determines Rx Level Transmitter C34N004A has failed, directs RO to select NR Level Channel B

Crew can place RFPTs back on MLC when NR Level Channel B is selected and level is returned to normal

- Directs RO to shift both RFPTs from their Manual Speed Control Dial to the MLC per SOI-C34, Sections 4.5, 4.9, and 4.10 As directed, takes manual control of feedwater and maintains level; selects NR Level Channel B, and shifts both RFPTs from their Manual Speed Control Dial to the MLC SRO Notifies OPS Management and NRC Resident of ONI entry and reason for entry Refers to ORM 6.2.13, Action B NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 4 Page 1 of 2 Event

Description:

RFPT A trip. Reactor Recirculation FCV B fails to automatically runback Time Position Applicant's Actions or Behavior RO Observes the loss of RFPT A Informs SRO/BOP of RFPT A trip Verifies expected automatic plant response

- Starts or verifies auto start of MFP

- Verifies Reactor Recirculation Flow Control Valves will runback to 48% loop flow position Informs SRO/BOP of failure of FCV B to runback Takes immediate actions per ONI-N27

- If required, reduce reactor power by reducing recirculation flow and/or inserting Cram Rods in accordance with FTI-B02 to maintain steam flow and feed flow balanced and reactor water level within the normal operating range of 192 to 200 inches BOP Backs up RO by reviewing ARIs for annunciators received:

ARI-H13-P680-3 (A9), RX LEVEL HI/LO L7/L4 ARI-H13-P680-3 (D6), RFPT A TRIP ARI-H13-P680-4 (B4), RCIRC A FCV RUNBACK SRO Enters ONI-N27, Feedwater Pump Trip

- Verifies MFP auto started and shifted to the Master Level Controller

- Directs RO to maintain RPV level 192-200"

- Directs RFPT A shutdown using SOI-N27 Section 6.6, Shutdown to 1100 rpm and Section 6.7, Shutdown from 1100 rpm NUREG-1021, Revision 8, Supplement 1 40 of 40 I

I

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 4 Page 2 of 2 Event

Description:

RFPT A trip. Reactor Recirculation FCV B fails to automatically runback Time Position Applicant's Actions or Behavior SRO Enters ONI-C51, Unplanned Change in Reactor Power or Reactivity and references Section 4.2, Reactor Recirculation Flow Control Malfunction RO/BOP Complete immediate and supplemental actions as directed Shutdown RFPT A using SOI-N27 as directed BOP/SRO Contact Maintenance and Responsible System Engineer Directs NLO investigate RFPT A trip RO Informs SRO that there is a > 5% loop flow mismatch Determines Reactor Recirculation FCV B failed to automatically runback and informs SRO.

(Note: RO may manually runback Recirc FCV B)

SRO References Tech Specs for a Recirc loop flow mismatch

- LCO 3.4.1 (Recirc Loops Operating) - Enters Condition A

Reactor Engineering should be notified of the loop flow mismatch SRO Notifies OPS Management and NRC Resident of ONI entry and reason for entry and of entry into Tech Spec LCOs Reviews 101-3 for power decrease SRO/BOP/RO Notifies Chemistry, HP, and SCC of power change NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 5 Page 1 of 2 Event

Description:

TBCC Pump A trip / Manual start of standby TBCC Pump / TBCC System Process Piping Leakage / Fast reactor shutdown is required Time Position Applicant's Actions or Behavior BOP Recognizes and reports TBCC Pump A trip Responds to alarm. Consults ARI-H13-P870-1 (B6)

BOP/RO Dispatches NLOs to TBCC Pump A and pump breaker BOP/RO/SRO Requests Maintenance assistance in the Control Room to support troubleshooting SRO Acknowledges report TBCC Pump A trip Enters ONI-P44, Loss of Turbine Building Closed Cooling

- Directs BOP to start the standby TBCC pump per SOI-P44 Notifies OPS Management of ONI entry and reason for entry BOP Starts standby TBCC Pump per SOI-P44, Section 5.1.

- Take standby TBCC Pump control switch to start

May direct NLO to reset overcurrent trip BOP/RO Responds to, reports, and references ARIs for alarms:

- ARI-H13-P870-2 (H2), TBCC SURGE TANK LEVEL LOW

-ARI-H13-P870-2 (H4), TBCC PUMP SUCTION FLOW LOW May dispatch NLO to makeup to Surge Tank via the bypass valve BOP/RO Receives report bypass valve is open and surge tank level is continuing to decrease NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 5 Page 2 of 2 Event

Description:

TBCC Pump A trip / Manual start of standby TBCC Pump / TBCC System Process Piping Leakage I Fast reactor shutdown is required.

Time Position Applicant's Actions or Behavior BOP/RO Recognizes and reports loss of TBCC SRO Acknowledges report of loss of TBCC Re-enters ONI-P44, Loss of Turbine Building Closed Cooling

- Orders fast reactor shutdown, enters ONI-C71-1, Rx Scram

Will also enter PEI-B13, RPV Control (Non-ATWS) due to < L3 PEI-B13 RPV Control actions are described in Event six (6)

- Directs ONI-C71-1 Supplemental Actions

- Directs ONI-P44 Supplemental Actions as time permits

Any component or system served by the TBCC System that reaches its temperature limit shall be placed in the secured status per its applicable SOI SRO/BOP/RO May direct NLO to walkdown TBCC System SRO/BOP/RO If NLO was directed to perform system walkdown, receives report of water in Turbine Building at Elevation 605 RO/BOP Respond to, report, and reference ARIs for Alarms:

H13-P680-9 (DI), ISOPHASE BUS CLG TRBL H13-P680-7 (B131), GENERATOR TEMP P811 H13-P680-7 (D 9), H2 SEAL/STATOR CLG TRBL RO/BOP Carry out ONI-C71-1 actions as SRO directs RO/BOP Carry out ONI-P44 actions as SRO directs NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 6 Page 1 of 2 Event

Description:

Main Steam Line break in Drywell. RHR Pump A fails to auto start on high Drywell pressure (required for Containment Spray mode) / May be manually started.

Time Position I Applicant's Actions or Behavior SRO/RO/BOP Recognize rising Drywell Pressure condition

If not manually scrammed, Rx scrams on igh Drywell Pressure SRO Enters PEI-B13 RPV Control (Non-ATWS) due to < L3, high RPV press, and Drywell pressure > 1.68 psig.

Enters PEI-T23, Containment Control due to Drywell pressure

> 1.68 psig SRO Directs PEI-B13, RPV Control (Non-ATWS)

- Verifies reactor is scrammed

- Confirms Reactor Mode Switch is in SHUTDOWN

- Start Hydrogen Analyzers

- Verifies reactor shutdown under all condition without boron

- Verifies SRMs and IRMs inserted

- Directs pressure control 700 to 900 psig using SRVs RPV Pressure Control

- Stabilizes RPV pressure 800 to 1000 psig (Note: As Containment pressure rises, SRO may direct RPV rapidly depressurized to the Main Condenser using Main Turbine Bypass Valves)

I I

t I

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 6 Page 2 of 2 Event

Description:

Main Steam Line break in Drywell. RHR Pump A fails to auto start on high Drywell pressure (required for Containment Spray mode) / May be manually started.

Time Position Applicant's Actions or Behavior SRO (Cont.)

RPV Level Control

- Restores and maintains RPV level between 185 and 215 inches

CRD - available

RCIC - available

HPCS - available

Reactor Feedwater Booster Pumps and Motor Feed Pump unavailable for level control due to loss of TBCC RO/BOP Executes PEI-B1 3, RPV Control (Non-ATWS) actions per SRO direction RO/BOP Verifies automatic plant response is as expected Recognizes failure of RHR Pump A to auto start on high Drywell pressure and manually starts RHR Pump A RO/BOP When conditions allow, informs SRO of failure of RHR Pump A to auto start SRO Acknowledges failure of RHR Pump A to auto start SRO Directs actions per PEI-T23, Containment Control, when Drywell press reaches 1.68 psig (described in Event 7)

NUREG-1021, Revision 8, Supplement 1 40 of 40

[

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 7 Page 1 of 2 Event

Description:

RHR Pump A trips when flow is aligned to spray Containment / RPV emergency depressurized to control Containment pressure.

Time Position Applicant's Actions or Behavior SRO Directs actions per PEI-T23, Containment Control, when Drywell pressure reaches 1.68 psig Drywell and Containment Temperature Control

- Operates all available DW cooling

- Restores NCC to the DW

- Operates all available Containment cooling

- Restores CVCW System

- Attempts to maintain Cont. average temperature less than 1850F Drywell & Containment Pressure Control

- Attempts to maintain Containment pressure below PSP

- Directs RHR Loop A in the Containment Spray Mode when Containment pressure exceeds 2.25 psig Suppression Pool Level Control

- Restores and maintains SP level between 17.8 and 18.5 ft Suppression Pool Temperature Control

- Maintains both SP average temperature and RPV pressure below HCL RO/BOP Executes PEI-T23 actions per SRO direction Recognizes RHR Pump A trips when flow is aligned to Spray Containment and immediately informs SRO NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 7 Page 2 of 2 Event

Description:

RHR Pump A trips when flow is aligned to spray Containment / RPV emergency depressurized to control Containment pressure.

Time Position Applicant's Actions or Behavior SRO Prior to exceeding the Pressure Suppression Pressure (PSP),

exits PEI-B13, RPV Control (Non-ATWS), RPV Pressure Leg and enters PEI-B13, Emergency Depressurization SRO Executes PEI-B13, RPV Control (Non-ATWS), RPV Level Control Leg concurrently with PEI-B1 3, Emergency Depressurization SRO Directs RO/BOP actions per PEI-B1 3, Emergency Depressurization

- Confirms that the reactor is shutdown under all conditions without boron

- Verifies Drywell pressure is > 1.68 psig

- Verifies no low pressure ECCS are required for adequate core cooling

- Prevents injection from LPCS and LPCI

- Verifies eight or more SRVs are not open

- Verifies Suppression Pool level is > 5.25 ft

- Opens all ADS valves to rapidly depressurize the RPV

Crew should continue to restore and maintain RPV level 185-215" using available injection systems during Emergency Depressurization RO/BOP Executes PEI-B13, Emergency Depressurization actions per SRO direction NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Page 1 of 1 Event

Description:

Scenario Termination Criteria Time Position Applicant's Actions or Behavior

1. Control Rods are fully inserted

2. The reactor has been emergency depressurized

3. Feedwater or low pressure ECCS Systems are being used to restore RPV level to 185 to 215 inches

4. Containment and Drywell parameters are being restored per PEI-T23, Containment Control

4.

4-4-

4-

4.

4-4-

+

1-1-

4-4-

+

1-4-

.1-

+

NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No.: 4 Event No.: 7 Page 1 of 1 Event

Description:

Critical Task #1 Time Position Applicant's Actions or Behavior Critical Task #1 -With Containment pressure exceeding 2.25 psig, and prior to exceeding the Pressure Suppression Pressure, attempt to initiate Containment Spray

1. Safety Significance:

- Precludes an unrequired Emergency Depressurization

2. Cues:

- Containment pressure increase

- Procedural compliance

3. Measured by:

- Observation - With Containment pressure at least 2.25 psig, Containment Spray is manually initiated prior to exceeding the Pressure Suppression Pressure

4. Feedback:

"-"Containment Spray Start Signal Received" alarm status NUREG-1021, Revision 8, Supplement 1 40 of 40

Appendix D Operator Actions Form ES-D-2 Op-Test No.: 2002-01 Scenario No: 4 Event No.: 7 Page 1 of 1 Event

Description:

Critical Task #2 Time Position Applicant's Actions or Behavior Critical Task #2 - When Containment pressure cannot be maintained below the Pressure Suppression Pressure, initiate Emergency depressurization of the RPV prior to exceeding PSP

1. Safety Significance:

- Precludes degradation of a fission product barrier

2. Cues:

- Increasing Containment pressure

- Procedural compliance

3. Measured by:

- Observation - At least 5 SRVs must be open prior to exceeding the Pressure Suppression Pressure

4. Feedback:

- RPV pressure decreasing

- SRV status indications

4.

1-J.

4

+

4.

I1 I

.4-4

4.

4

+

4.

NUREG-1021, Revision 8, Supplement 1 40 of 40 I ___________

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ML021190681

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