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Initial Exam 2014-301 Draft Simulator Scenarios
	ML14154A365
Person / Time
Site:	Browns Ferry
Issue date:	06/02/2014
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
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ML14155A059	List: ML14154A324; ML14154A337; ML14154A346; ML14154A365; ML14154A369; ML14154A373; ML14154A377; ML14154A383; ML14154A390; ML14154A393; ML14154A396; ML14154A399;
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{{#Wiki_filter:______ C Facility: Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.: 1404 Examiners:___________________ Operators: SRO:___________________ ATC:_____________ BOP:______________ Initial Conditions: 80% power, RFPT 3B and A3 RHRSW Pumps are tagged out. Turnover: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Raise power to 85% with flow and hold for RFPT 3B repairs. Event Maif. No. Event Type* Event Description No. Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and N-BOP 1 30B, Refuel damper 64-9 fails in mid position when Refuel TS-SRO Fans are in Off and is Open when Refuel Fans operating R-ATC 2 Commence power increase with flow to 85% R-SRO C-ATC 3 edlOb C-BOP Loss of 480V SD BD 3B TS-SRO C-BOP 4 Batch File Stator Water Cooling Pump trip C-SRO 5 fw30a RFPT 3A Governor fails low tc 1 Ob IATC EHC Pressure Transducer failure, with MSIV closure failure 6 I-SRO 7 Batch File M-ALL ATWS 8 Batch File M-ALL LOCA Loss of RPV Water Level 9 hpO7 C Loss of HPCI 120 VAC Power Supply

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

Critical Tasks Six With 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 unless all required injection systems are Terminated and Prevented.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS A/B I14HIBITED annunciator status. With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BuT) and inserting control rods. I. Safety Significance: Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.

2. Cues:

Procedural compliance. Suppression Pool temperature.

3. Measured by:

Observation If operating lAW EOI- 1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool. AND RO places SLC A / B Pump control switch in ON, when directed by US. AND Control Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:

Reactor Power trend. Control Rod indications. SLC tank level.

Critical Tasks Six-During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection (except for CRD, SLC and RCIC) from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US. I. Safety Significance: Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation No ECCS injection prior to being less than the MARFP. AND Observation Feedwater terminated and prevented until less than the MARFP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms. Injection system flow rates into RPV. With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

Maintaining adequate core cooling and preclude possibility of large power excursions.

2. Cues:

Procedural compliance. RPV pressure indication.

3. Measured by:

Observation Injection not commenced until less than MARFP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF.

4. Feedback:

RPV level trend. RPV pressure trend. Injection system flow rate into RPV.

Critical Tasks Six - When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation RO initiates Drywell Sprays

4. Feedback:

Diywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays JAW with EOI Appendix 1 7B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment

Critical Tasks Six After RPV water level drops to (-) 50 inches, when RPV level cannot be restored and maintained above (-) 180, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. RPV level indication.

3. Measured by:

At least 6 SRVs are opened when RPV level cannot be restored and maintained above -180.

4. Feedback:

RPV pressure trend. Suppression Pool temperature trend. SRV open status indication.

Events

1. BOP operator will alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B. Refuel damper 64-9 fails in mid position when Refuel Fans are in off and is open when Refuel Fans operating. Tech Spec 3.6.4.2 Condition A, required action A.1 and A.2.

2. ATC commences power increase 85% using recirculation flow.

3. The Crew will respond to a loss of 480V SD BD 3B, this will cause a loss of RPS B, loss of 480V RMOV BDs 3B and 3C. The Inboard MS1V A will have inadvertently closed. The crew will need to lower power to meet the main steam line flow guidance JAW 3 -AOI-3 -1. The crew will need to restore power to 480V SD BD 3B, reset RPS, reset PCIS and restore systems. The SRO will also have to enter the following AOIs; 3-AOI-1-3, 3-AOI-70-1, and 3-AOI-99-1. SRO will refer to the TRM and detennine Technical Surveillance Requirement 3.4. Li to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours. SRO will refer to Tech Spec 3.6.1.3 for failed closed MSIV and enter condition A. SRO will refer to Tech Spec 3.4.5 and determine Condition B is required for inoperable containment atmospheric monitoring equipment.

4. The running Stator Water Cooling Pump will trip and the standby pump will fail to AUTO start.

The BOP operator will be required to start the standby Stator Water Cooling pump to restore system flow and prevent an automatic Turbine Trip/Reactor scram.

5. RFPT 3A flow controller will slowly fail low, RFPT 3A speed will continue to decrease until the ATC or Crew notices. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to restore RFPT 3A speed in manual. SRO should direct entry into 3-AOI-3-1.

6. The EHC pressure transducer will slowly fail causing reactor pressure to slowly lower, the crew will enter 3-AOI-47-2. With RPV pressure lowering and the mode switch in RUN the MS1Vs will fail to automatically close. The ATC will scram the reactor and verify MSIVs closed.

7. An ATWS will occur on the scram and the power supply to HPCI will fail, leaving RCIC as the only source of high pressure makeup besides SLC and CRD. The crew will insert control rods manually, and maintain reactor level.

8. With RCIC, CRD and SLC as the only source of high pressure makeup as the LOCA degrades RPV Level will continue to lower. The SRO will determine Emergency Depressurization is required to restore RPV Level. The crew will ED and restore RPV Level with available systems.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Control Rods are being inserted Emergency Depressurization complete Reactor Level is restored

SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 9 Total Malfunctions liserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 6 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) Scenario Tasks TASK NUMBER KJA RO SR() Alternate Reactor and Refuel Zone Fans RO U-30A-NO-02 288000 A4.01 3.1 2.9 Raise Power with Recirc Flow RO U-000-NO-06 202002 A4.07 3.3 3.3 SRO S-000-AD-3 1

Scenario Tasks TASK NUMBER EQ EQ Loss of 480V SD BD 3B RO U-57B-AL-06 262001 A2.04 3.8 4.2 SRO S-57B-AL-09 Reactor Feed Pump Turbine Governor Failure ROU-003-AL-09 259002 A4.0l 3.8 3.6 SRO S-003-AB-01 Stator Water Cooling Pump Trip RO U-35A-AL-02 245000 A4.03 2.7 2.8 SRO S-070-AB-01 EHC Pressure Transducer Failure RO U-047-AB-02 241000 A2.03 4.1 4.2 SRO S-047-AB-02 LOCA/Low Level ED RO U-003-AL-24 295031 EA2.04 4.6 4.8 RO U-000-EM-01 SRO S-000-EM-14 SRO S-000-EM-15 SRO 5-000-EM-Ol ATWS RO U-000-EM-03 295015 AA2.01 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18

NRC Scenario 4 ( acility: 7 Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.: j44 Fximiiiers: Operators: SRO:_ ATC:_ BOP: Initial Conditions: 80% power, RFPT 3B and A3 RHRSW Pumps are tagged out. Turnover: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Raise power to 85% with flow and hold for RFPT 3B repairs. Event Maif. No. Event Type* Event Description No. N-BOP Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 1 30B, Refuel damper 64-9 fails in mid position when Refuel Fans TS-SRO are in Off and is Open when Refuel Fans operating R-ATC 2 Commence power increase with flow to 85% R-SRO C-ATC 3 edlOb C-BOP Loss of 480V SD BD 3B TS-SRO 4 Batch File Stator Water Cooling Pump trip 5 fw30a RFPT 3A Governor fails low 6 tcl Ob EHC Pressure Transducer failure, with MSIV closure failure 7 Batch File M-ALL ATWS 8 Batch File M-ALL LOCA Loss of RPV Water Level 9 hpO7 C Loss of HPCI 120 VAC Power Supply

       *    (N)ormal,   (R)eactivity, (I)nstrument,     (C)omponent,    (M)ajor 1

NRC Scenario 4 Critical Tasks Six With 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 unless all required injection systems are Terminated and Prevented.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS A/B INHIBITED annunciator status. With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BIIT) and inserting control rods.

1. Safety Significance:

Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.

2. Cues:

Procedural compliance. Suppression Pool temperature.

3. Measured by:

Observation If operating lAW EOI-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool. AND RO places SLC A / B Pump control switch in ON, when directed by US. AND Control Rod insertion commenced in accordance EO1 Appendixes.

4. Feedback:

Reactor Power trend. Control Rod indications. SLC tank level. 2

NRC Scenario 4 Critical Tasks Six During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection (except for CRD, SLC and RCIC) from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation No ECCS injection prior to being less than the MARFP. AND Observation Feedwater terminated and prevented until less than the MARFP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms. Injection system flow rates into RPV. With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

Maintaining adequate core cooling and preclude possibility of large power excursions.

2. Cues:

Procedural compliance. RPV pressure indication.

3. Measured by:

Observation Injection not commenced until less than MARFP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF.

4. Feedback:

RPV level trend. RPV pressure trend. Injection system flow rate into RPV. 3

NRC Scenario 4 Critical Tasks Six When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment 4

NRC Scenario 4 Critical Tasks Six After RPV water level drops to (-) 50 inches, when RPV level cannot be restored and maintained above (-) 180, RU initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. RPV level indication.

3. Measured by:

At least 6 SRVs are opened when RPV level cannot be restored and maintained above -180.

4. Feedback:

RPV pressure trend. Suppression Pool temperature trend. SRV open status indication. 5

NRC Scenario 4 Events

1. BOP operator will alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Refuel damper 64-9 fails in mid position when Refuel Fans are in off and is open when Refuel Fans operating. Tech Spec 3.6.4.2 Condition A, required action A. 1 and A.2.

2. ATC commences power increase 85% using recirculation flow.

3. The Crew will respond to a loss of 480V SD BD 3B, this will cause a loss of RPS B, loss of 480V RMOV BDs 3B and 3C. The Inboard MSIV A will have inadvertently closed. The crew will need to lower power to meet the main steam line flow guidance JAW 3-AOI-3-1. The crew will need to restore power to 480V SD BD 3B, reset RPS, reset PCIS and restore systems. The SRO will also have to enter the following AOIs; 3-AOI-1-3, 3-AOI-70-1, and 3-AOI-99-1. SRO will refer to the TRM and determine Technical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours. SRO will refer to Tech Spec 3.6.1.3 for failed closed MSIV and enter condition A. SRO will refer to Tech Spec 3.4.5 and determine Condition B is required for inoperable containment atmospheric monitoring equipment.

4. The running Stator Water Cooling Pump will trip and the standby pump will fail to AUTO start.

The BOP operator will be required to start the standby Stator Water Cooling pump to restore system flow and prevent an automatic Turbine Trip/Reactor scram.

5. RFPT 3A flow controller will slowly fail low, RFPT 3A speed will continue to decrease until the ATC or Crew notices. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to restore RFPT 3A speed in manual. SRO should direct entry into 3-AOI-3 -1.

6. The EHC pressure transducer will slowly fail causing reactor pressure to slowly lower, the crew will enter 3-AOI-47-2. With RPV pressure lowering and the mode switch in RUN the MSIVs will fail to automatically close. The ATC will scram the reactor and verify MSIVs closed.

7. An ATWS will occur on the scram and the power supply to HPCI will fail, leaving RCIC as the only source of high pressure makeup besides SLC and CR0. The crew will insert control rods manually, and maintain reactor level.

8. With RCIC, CR0 and SLC as the only source of high pressure makeup as the LOCA degrades RPV Level will continue to lower. The SRO will determine Emergency Depressurization is required to restore RPV Level. The crew will ED and restore RPV Level with available systems.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Control Rods are being inserted Emergency Depressurization complete Reactor Level is restored 6

NRC Scenario 4 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 9 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List(1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 6 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) 7

NRC Scenario 4 Scenario Tasks TASK NUMBER iQ RQ Alternate Reactor and Refuel Zone Fans RO U-30A-NO-02 288000A4.01 3.1 2.9 Raise Power with Recirc Flow RO U-000-NO-06 202002A4.07 3.3 3.3 SRO S-000-AD-3 1 Loss of 480V SD BD 3B RO U-57B-AL-06 26200 1A2.04 3.8 4.2 SRO S-57B-AL-09 Reactor Feed Pump Turbine Governor Failure RO U-003-AL-09 259002A4.01 3.8 3.6 SRO 5-003-AB-Ol Stator Water Cooling Pump Trip RO U-35A-AL-02 245000A4.03 2.7 2.8 SRO 5-070-AB-Ol EHC Pressure Transducer Failure RO U-047-AB-02 241000A2.03 4.1 4.2 SRO S-047-AB-02 LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-Oi ATWS RO U-000-EM-03 295015AA2.01 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18 8

NRC Scenario 4 Procedures Used/Referenced: Procedure Number Procedure Titie 3-01-3 OA Refuel Zone Ventilation System 3-0I-30B Reactor Zone Ventilation System 3 -GOl- 100-12 Power Maneuvering 3-01-68 Reactor Recirculation System 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-8C Panel 9-8 3-XA-55-8C 3-ARP-9-4C Panel 9-4 3-XA-55-4C 3-A0I-99-1 Loss of Power to One RPS Bus 3 -AOl- 1-3 Main Steam Isolation Valve Closure at Power 3-A0I-70-1 Loss of Reactor Building Closed Cooling Water 3-01-99 Reactor Protection System 3-A0I-64-2D Group 6 Ventilation System Isolation Technical Specifications Technical Requirements Manual 3-ARP-9-7A Panel 9-7 3-XA-55-7A 3-ARP-9-8A Panel 9-8 3-XA-55-8A 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3 -A0I-3 -1 Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-ARP-9-7B Panel 9-7 3-XA-55-7B 3-A0I-47-2 Turbine EHC Control System Malfunctions 3 -AOl- 100-1 Reactor Scram 3-EOI-1 RPV Control 3-C-5 Level/Power Control 3-EOI Appendix-iD Insert Control Rods Using Reactor Manual Control System 3-E0I Appendix-iF Manual Scram 3-E0I Appendix-2 Defeating ART Logic Trips 3-EOI Appendix-i 1A Alternate Pressure Control Systems MSRVs 3-EOI Appendix-5C Injection System Lineup RCIC 3-EOI-2 Primary Containment Control 3-EOI Appendix-17A R}IR System Operation Suppression Pool Cooling 3-EOI Appendix-i 7C RHR System Operation Suppression Chamber Sprays 3-EOI Appendix-i 7B RHR System Operation Drywell Sprays 3-AOI-85-3 CRD System Failure 3-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 3 -EOI Appendix-4 Prevention of Injection 3-C-2 Emergency RPV Depressurization 3-EOI Appendix-6B Injection Subsystems Lineup RFIR System I LPCI Mode 3-EOI Appendix-6C Injection Subsystems Lineup RHR System II LPCI Mode EPIP-1 Emergency Classification 9

NRC Scenario 4 Batch File

  #RFPT 3B and EECW Pump A3 tagout ior ypobkrrhrswpa3 fail_ccoil ior zlohs2385a[1] off ior ypomtreopb3 fail_control_power ior ypobkrmopbi fail_ccoil br ypobkrmopb2 fail_ocoil ior ypomtrtgmb fail_control_power ior ypovfcv03 12 fail_power_now ior ypovfcv0295 fail_power_now ior ypovfcv0l 129 fail_power_now ior ypovfcv0l 133 fail_power_now
  #Refuel Zone Damper 64-9 fails in Mid position ior zlohs649[2] on
  #Loss of 480V SD BD B imfedlOb (el 0) ior zdihs0l 14a[1] (el 0) close mrfrp09 (e3 0) reset trg e3 bat restorerpsb ior zdixs577l[1] (e4 0) normal
  #B stator water pump trip irfego2 (e5 0) off ior ypobkrscwpa (e5 0) fail_ccoil ior zdihs3535a[2] (e5 0) stop ior zlohs3535a[1] (e5 0) off
   #rft governor drift imffw30a (elO 0) 0 2500 72 trg eli nrcrfptA trgell dmffw30a
   #B EHC Pressure transducer failure ior zdihs0ll6[1] (e14 0) select ior zdihs472O4[1j (e14 0) null ior zlohs0l 16[1] off ior z1ohs47204[1] on imftcl0b (e14 0) 82 3000 79 imf ms06b imf ms06c imf ms06d imfms06e imf ms06f imf mso6g imf ms06h 10

NRC Scenario 4

  #major trg e 18 nrcmodesw bat nrcstickquad Imfth22 (e18 1:00)

Imfth2l (e18 10:00) 1 15:00 imfhp07 (e18 0) trg e23 bat appOif trg e24 = bat app02 trg e25 bat nrcstickquad-1 mrfrdO6 (e26 0) close mrfrd06 (e27 0) open trg e28 = bat nrcatws95 trg e 18 = bat nrcmsiv Trigger Files nrcrfptA zdihs468a[41 .ne. 1 Scenario 4 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 203 Simulator Setup Load Batch bat nrcl4O4-4 Simulator Setup manual Tag RFPT 3B and EECW Pump A3 Simulator Setup Verify file loaded. Log in to EHC System to ensure when operators try to access they are able to. RCP required (80% 85% with flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12 11

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B SRO Direct Refuel and Reactor Zone Fans alternated BOP 6.1 Alternating Refueling Zone Supply and Exhaust Fans 3-OI-30A [1] NOTIFY Unit 1 and Unit 2 Operators that the Refuel Zone fans are being alternated. [2] VERIFY the Refueling Zone supply and exhaust fans are operating. REFER TO Section 5.1. [3] REVIEW precautions and limitations in Section 3.0. NOTES

1) The preferred method to start the alternate Refueling Zone supply and exhaust fans is to use the common control Switch, 3-HS-64-3A, on Panel 3-9-25.

2) Refueling Zone supply and exhaust dampers, 3-FCO-064-0005,0006,0009, and 0010 will open or close automatically as necessary when fans are stopped and started.

3) Refueling Zone supply and exhaust fans are alternated every six weeks.

[4] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in OFF. [5] CHECK that the two red lights A(B) extinguish and the two green lights A(B) illuminate above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A. NOTE BOP If any damper does not meet the requirements of step 6.1 [6] IMMEDIATELY notify the Unit supervisor to evaluate SCIV damper operability (refer to TRM appendix A). If any listed damper indicates not full closed, it should be considered inoperable for its SCIV function, and the required actions of Tech Spec LCO 3.6.4.2 entered for all units. [6] CHECK the red (open) damper position indication lights extinguish and the green (closed) lights illuminate above the following control switches:

REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
REFUEL ZONE SPLY 1NBD ISOL DMPR, 3-HS-64-6
REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9
REFUEL ZONE EXH 1NBD ISOL DMPR, 3-HS-64-10

[7] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B) to start alternate fans. BOP Report Failure of REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9 12

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B BOP [8] ChECK that the two green lights A(B) extinguish and the two red lights A(B) illuminate above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A. [9] CHECK the red (open) damper position indication lights illuminate and green (closed) lights extinguish above the following control switches:

REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6
REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9
REFUEL ZONE EXH INBD ISOL DMPR, 3-HS-64-l0 NOTE A five minute time delay should be observed following Refuel Zone Supply and Exhaust Fan SLOW Start. The time delay allows the discharge dampers to fully open after SLOW start.

[10] IF Refueling Zone Supply and Exhaust Fan FAST speed operation is necessary, THEN: PERFORM the following: [10.1] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in FAST 3A (FAST 3B). [10.2] CHECK that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A. [11] CHECK the following conditions:

SUPPLY FANS FILTER DIFF PRESS Indicator, 3-PDI-064-0022, indicates less than 0.6 inches H2O at the Reactor Building/Refuel Floor Supply fan intake room at El 565.

REFUELING ZONE STATIC PRESS INTLK, 1-PDS-064-006 lA/C, on refuel floor Panel 25-220 indicates between (negative) -0.25 inches to -0.40 inches. BOP Contacts AUO for the above information Driver When contacted wait 4 minutes and report 3-PDI-064-0022, indicates 0.4 inches H20 and that 1-PDS-064-006 lA/C, indicates -0.3 inches. 13

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B SRO Direct Refuel and Reactor Zone Fans alternated BOP 6.1 Alternating Reactor Zone Supply and Exhaust Fans 3-OI-30B [1] VERIFY the Reactor Zone supply and exhaust fans are operating. REFER TO Section 5.1. [2] REVIEW all Precautions and Limitations in Section 3.0. NOTES

1) The preferred method to start the standby Reactor Zone supply and exhaust fans is to use the common control switch (3-HS-64-1 1A) on Panel 3-9-25.

2) Reactor Zone supply and exhaust dampers, 3-FCO-064-0013, 0014, 0042, and 0043 will open or close automatically as necessary when fans are stopped and started.

3) The Steam Vault Exhaust Booster Fan should normally be in service whenever the Unit is operating with Reactor Building Ventilation in service and fans in fast speed. Operation of the Steam Vault Exhaust Booster Fan with Reactor Zone Exhaust fans out of service is an ALARA concern due to backflow into the Reactor Building lower level ventilation ductwork. However, the Steam Vault Exhaust Booster fan may remain in service with Reactor Zone Exhaust fans out of service to cool the steam tunnel for short durations such as alternating fans, cycling reactor zone dampers, or RPS power transfers.

[3] IF Reactor Zone Ventilation is to remain Out of Service for an extended period (3 hours) and it is desired to leave the Steam Vault Exhaust Booster Fan in service, THEN (Otherwise N/A): Step is NA [4] IF required, THEN SHUT DOWN Steam Vault Exhaust Booster Fan. REFER TO Section 7.4. (Otherwise N/A). Step is NA [5] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in OFF. [6] VERIFY dampers close and fans stop as indicated by illuminated green lights above the following switches:

REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
REACTOR ZONE EXH INBD ISOL DMPR, 3-HS-64-42
REACTOR ZONE EXH OUTBD ISOL DMPR. 3-HS-64-43
REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1A

[7] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in SLOW A (SLOW B) to start alternate fans. 14

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans LAW 3-OI-30A and 30B BOP [8] VERIFY dampers open and fans start as indicated by illuminated red lights above the following switches:

REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
REACTOR ZONE EXH INBD ISOL DMPR, 3-HS-64-42
REACTOR ZONE EXH OUTBD ISOL DMPR, 3-HS-64-43
REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1A

[9] IF fast speed Reactor Zone Supply and Exhaust Fan operation is required, five minutes should be allowed after slow start for the discharge dampers to FULLY OPEN, THEN [9.1] PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11A, in FAST A (FAST B). [9.2] VERIFY that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A. [10] VERIFY the following conditions: [10.1] VERIFY REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-064-0002, on 3-LPNL-925-0213, located at R17-P El 639, indicates between -0.25 inches and -0.40 inches H2O. [10.2] IF REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-64-2, is not between -0.25 inches and -0.40 inches H20, THEN REFER TO 3-AOl-3 OB- 1, Reactor Building Ventilation Failure. [11] IF required, THEN START Steam Vault Exhaust Booster Fan. REFER TO Section 5.4. NOT Required BOP Contacts AUO for the above information Driver When contacted wait 4 minutes and report 3-PDIC-064-0002, indicates -0.35 inches H20. 15

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) LCO 3.6.4.2 Each SCIV shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs). SRO ACTIONS NOTES

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

Condition A. One or more penetration flow paths with one SCIV inoperable. Required Action A. 1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. Completion Time: 8 hours AND Required Action A.2 Verify the affected penetration flow path is isolated. Completion Time: Once per 31 days Call Work Control for a clearance on the REFUEL ZONE SPLY OUTBD ISOL DMPR, SRO 3 -FCV-64-5 TRM Appendix A Power Operated Secondary Containment Isolation Valves 3-DMP-64-9 REFUELiNG ZONE EXH DUCT OUTBD ISOL DMPR <10 SEC 16

NRC Scenario 4 Simulator Event Guide: Event 2 Reactivity: Power increase with Recirc Flow SRO Notifies ODS of power increase. Directs Power increase using Recirc Flow, per 3 -GOl- 100-12. [20] IF desired to raise power with only two(2) Reactor feedpumps in service, THEN RAISE Reactor power, as desired, maintaining each Reactor feedpump less than 5050 RPM. ATC Raise Power w/Recirc, lAW 3-01-68, Section 6.2 D. Individual pump speeds should be mismatched by 60 RPM during dual pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short time periods for testing or maintenance). [11 IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;

Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM), 3-HS-96-15A(15B).

AND/OR

Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM), 3-HS-96-16A(16B).

[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A &3B using the following push buttons as required: RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-HS-96-32 NRC At RR Pump Speeds of 1260rpm and 1200 rpm, power will be 85% and RFPT RPMs will be just below 5025 Driver When directed by NRC, Trigger 1 Loss of 480V SD BD 3B, If crew attempts to close alternate supply breaker or is going to close alternate supply breaker delete ED I OB in order to allow the crew to energized the Board Driver Wait 2 minutes and report license class 1404 was in the field, a trainee accidently tripped the normal feeder breaker. No problems indicated on Board. 17

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B Crew Responds to numerous alarms, diagnoses a loss of 480V SD BD 3B and 480V RMOV Bds 3Band3C Responds to the following alarms; 8B-30, 8C-17, 24, 29, 31, 4C-12 and 3D -32. SRO Enters 3-AOI-99-1, 3-AOI-1-3 and 3-A0I-70-l. BOP Alarm 8B-30: 480V SHUTDOWN BD 3B UV OR XFR A. CHECK for indication of 480V Shutdown Bd 3B loss: 1

RWCU Pump 3B shutdown
Fuel pool cooling Pump 3B shutdown
480V Shutdown Bd 3B voltage (3-EI-57-30)

B. IF 480V Shutdown Bd 3B is lost, THEN MANUALLY TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 3-9-8. C. IF manual transfer is accomplished, THEN REFER TO O-OI-57B, 3-01-99, and appropriate Ols for recovery or realignment of equipment. D. IF manual transfer is NOT accomplished, THEN REFER TO Tech Spec Section 3.8.1. Dispatches personnel to Breaker, may attempt to energize 480V SD BD 3B Driver If crew attempts to close alternate supply breaker or is going to close alternate supply breaker delete ED I OB in order to allow the crew to energized the Board Driver DO NOT Call until requested to investigate. Wait 2 minutes and report license class 1404 was in the field, a trainee accidently tripped the normal feeder breaker. No problems indicated on Board. If the crew directs you to restore 480V SD BD 3B to Normal supply trigger 4, ior zdihs577 1 [1] normal, if directed to restore Board on alternate supply change normal to ALT (alternate) 18

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B = BOP Alarm 8C-24: 480V REACTOR MOV BD 3B OR 3E UV OR XFR A. CHECK light indications for loss of any 480V equipment. B. CHECK 480V Rx MDV Bd 3B & 3E for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. C. IF Normal or Alternate feeder breaker tripped, THEN MANUALLY DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device. D. IF undervoltage or transfer has occurred:

1. REFER TO TS Section 3.8.7.

2. RESET possible half-scram. REFER TO 3-01-99.

BOP Alarm 8C-29: 1&C BUS B VOLTAGE ABNORMAL A. VERIFY the Alarm by checking:

Loss of instrument power and remote position indication to Core Spray Div II and R}IR Div II (Panel 3-9-3)
RWCU Filter Demin 3B Isolation
Reactor Zone/Refuel Zone Ventilation Isolation Verifies I&C Bus B Auto transferred to alternate feeder Alarm 8C-29: 480V REACTOR MOV BD 3C UV OR XFR A. VERIFY automatic action.

B. CHECK light indications for loss of 480V equipment. C. CHECK board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. D. IF Normal or Alternate feeder breaker tripped, THEN MANUALLY DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device. E. REFER TO O-OI-57B to re-energize or transfer the board. Driver When requested to restore steam tunnel booster fan wait two minutes and mrf PC 14 start 19

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B ATC Alarm 4C-12: RBCCW PUMP DISCH. HDR PRESS LOW A. VERIFY 3-FCV-70-48 CLOSiNG/CLOSED. B. VERIFY RBCCW pumps A and B in service. C. VERIFY RBCCW surge tank low level alarm is reset. E. REFER TO 3-AOI-70-1 for RBCCW System failure and 3-01-70 for starting spare pump. When 480V RMOV BD 3B is restored should VERIFY 3-FCV-70-48 CLOSING Report Alarm 3D-32: Reactor Zone Differential Pressure Low BOP EOI-3 Entry Condition. SRO Enters EOI-3 Secondary Containment Control When requested to restore RPS B, if requested to place on alternate trigger 3, mrf rpO9 reset Driver and bat restorerpsb, if requested to restore to normal then bat rpsreset and mrf rpO9 reset. if place back on normal ensure to reset alternate supply circuit protectors mrf rpO3 reset. 20

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B ATC Announces Power, Pressure and Level stable on Board loss Crew 3-AOI-99-1, Loss of Power to One RPS Bus 4.1 Immediate Actions [11 STOP all testing with potential RPS half-scrams or PCIS logic isolation signals. NOTES

3) Loss of RPS will isolate 3-RM-90-256, Drywell Air Monitor, and TS LCO 3.4.5 Condition B should be entered.

4.2 Subsequent Actions [1] VERIFY automatic actions occur. [2] ATTEMPT to determine cause of loss of RPS Bus using indicating lights inside RPS Circuit Protector cabinets. [3] NOTIFY Chemistry RWCU is isolated and no longer in-service and a sampling LCO per TRM 3.4.1 is to be entered. [4] NOTIFY Electrical Maintenance to correct cause. [5] RESTORE power to RPS Bus A(B) using alternate power supply. REFER TO 3-01-99 section for Immediate Restoration of Power to RPS Bus A(B) Using Alternate Power Supply. [5.1] DISPATCH operator to Aux. Instrument Room to reset ATU GROSS FAILURES. [6] WHEN system restoration is desired, THEN RESTORE systems to normal. REFER TO 3-01-99 section for Restoration to Normal Following RPS Bus Power Loss. When requested to restore RPS B, if requested to place on alternate trigger 3, mrfrpO9 reset Driver and bat restorerpsb, if requested to restore to normal then bat rpsreset and mrf rpO9 reset. If place back on normal ensure to reset alternate supply circuit protectors mrf rpO3 reset. 21

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B ATC/BOP Reports MSIV A Inboard Valve shut on loss of board. SRO Enters 3-AO1-1-3, Main Steam Isolation Valve Closure at Power ATC 4.2 Subsequent Actions [1] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s). [2] LOWER reactor power with recirc flow and insert control rods as directed by the Reactor Engineer/Unit Supervisor as necessary to ensure that rated steam line flow 3.54 x 106 lb/hr is not exceeded as indicated on Main Steam Line Flow Indicators. REFER TO 3-GOI-100-12 or 3-GOI-100-12A for the power reduction. [6] IF Drywell control air pressure is normal, THEN INITIATE trouble-shooting of the MSIV. (Otherwise N/A) Step is NA [7] EVALUATE Technical Specification 3.6.1.3, Primary Containment Isolation Valves. SRO Directs ATC to lower power to less than 3.54 x 106 lb/hr on Main Steam Line Flow Indicators. Directs recirc pump speeds matched when outside of the 1200 to 1300 rpm band. BOP Places MSIV A Inboard Valve handswitch in the close position ATC Lowers power as directed by SRO 22

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-AO1-70-1, Loss of Reactor Building Closed Cooling Water ATC 4.1 Immediate Actions [1] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A):

SECURE RWCU Pumps.
VERIFY RBCCW SECTIONALIZING VLV, 3-FCV-70-48 CLOSED.

Verifies RWCU Tripped, cannot verify sectionalizing valve at this time NO Power 4.2 Subsequent Actions [1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AND core flow is above 60%,THEN: [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A). One RBCCW Pump is in service with sectionalizing valve open due to loss of power [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A): [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). ATC When power is restored to 480V SD BD 3B RBCCW Pump will auto start 23

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-AOI-70-1, Loss of Reactor Building Closed Cooling Water ATC 4.1 Immediate Actions [1] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A): SECURE RWCU Pumps.

VERIFY RBCCW SECTIONALIZ1NG VLV, 3-FCV-70-48 CLOSED.

Verifies RWCU Tripped, cannot verify sectionalizing valve at this time NO Power 4.2 Subsequent Actions [1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AND core flow is above 60%,THEN: [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A). One RJ3CCW Pump is in service with sectionalizing valve open due to loss of power [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A): [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). 24

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System AfC/B0P 8.3 Restoration to Normal Following RPS Bus Power Loss [1] OBTAIN Unit Supervisor/SROs permission to restore to normal. [2] MOMENTARILY PLACE SCRAM RESET, 3-HS-99-5A/S5, as follows: [2.1] RESET FIRST. (Group 2/3) [2.2] RESET SECOND. (Group 1/4) [2.3] NORMAL [3] CHECK the following conditions: A. All eight SCRAM SOLENOID GROUP A/B LOGIC RESET lights illuminated. B. The following four lights ILLUMiNATED:

SYSTEM A BACKUP SCRAM VALVE, 3-IL-99-5A/AB
SYSTEM B BACKUP SCRAM VALVE, 3-IL-99-5A/CD C. Scram Discharge Volume vent and drain valves indicate OPEN.

[4] At Panel 3-9-4, RESET PCIS trip logic as follows: [4.1] MOMENTARILY PLACE PCIS DIV I RESET, 3-HS-64-16A-S32, to left and right RESET positions. [4.2] CHECK the following red lights ILLUMINATED:

MSIV GROUP Al, 3-IL-64-A1
MSIV GROUP Bi, 3-IL-64-B1

[4.3] MOMENTARILY PLACE PCIS DIV II RESET, 3-HS-64-16A-S33, to left and right RESET positions. [4.4] CHECK the following red lights ILLUMINATED:

MSIV GROUP A2, 3-IL-64-A2
MSIV GROUP B2, 3-IL-64-B2

[6] RESTORE Reactor and Refuel Zone Ventilation to normal operation. REFER TO 3-AO1-64-2D, Group 6 Ventilation System Isolation. BOP 3-AOI-64-2D, Group 6 Ventilation System Isolation 25

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System BOP 8.3 Restoration to Normal Following RPS Bus Power Loss [7] RESTORE Standby Gas Treatment System to standby readiness. REFER TO 0-01-65. BOP [8] At Panel 3-9-3, PLACE PSC head tank pumps in service as follows:

PLACE SUPPR POOL DRAIN INBD ISOL VALVE, 3-HS-75-57A, in AUTO After OPEN.
PLACE SUPPR POOL DRAIN OUTBD ISOL VALVE, 3-HS-75-58A, in AUTO After OPEN.

[10] At Panel 3-9-3, RESTORE Drywell DP Compressor to automatic operation as follows: [10.1] DEPRESS DRYWELL DP CPRSR SUCT VLV RESET pushbutton, 3-HS-64-139A. [10.2] DEPRESS DRYWELL DP CPRSR DISCH VLV RESET pushbutton, 3-HS 140A. [10.3] VERIFY OPEN DW TO SGT INBD ISOL VALVE using 3-HS-64-31. [10.4] VERIFY OPEN SUPPR CHBR SGT INBD ISOL VALVE using 3-HS-64-34. BOP [11] At Panel 3-9-4, RESTORE Drywell Floor and Equipment Drain Systems to normal operation as follows: [11.11 NOTIFY Radwaste Operator that Drywell Equipment and Floor Drain Sump isolation valves are being reopened. [11.2] PLACE DRYWELL EQPT DR 1NBD ISOL VLV, 3-HS-77-15A, in AUTO After OPEN. [11.3] PLACE DRYWELL EQPT DR OUTBD ISOL VLV, 3-HS-77-15B, in AUTO After OPEN. [11.4] PLACE DRYWELL FLOOR DR INBD ISOL VLV, 3-HS-77-2A, in AUTO After OPEN. [11.5] PLACE DRYWELL FLOOR DR OUTBD ISOL VLV, 3-HS-77-2B, in AUTO After OPEN. Driver when directed by NRC trigger 5 for Stator water pump trip 26

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System BOP .3 Restoration to Normal Following RPS Bus Power Loss [12] IF DW Radiation Monitor CAM, 3-RM-90-256 was secured due to a preplanned transfer, THEN (otherwise N/A) Step is NA [13] IF DW Radiation Monitor CAM, 3-RM-90-256, isolated due to loss of RPS, THEN MOMENTARILY DEPRESS the following RESET pushbuttons on Panel 3-9-2.

DW RAD MON UPPER INBD SUPPLY ISV RESET, 3-HS-90-254A-A (opens FCV-90-254A)
DW RAD MON LOWER INBD SUPPLY ISV RESET, 3-HS-90-254B-A (opens FCV-90-254B)
DW RAD MON OUTBD RETURN ISV RESET, 3-HS-90-257A-A (opens FCV 90-257A
DW RAD MON OUTBD SUPPLY ISV RESET, 3-HS-90-255A (opens FCV 255
DW RAD MON 1NBD RETURN ISV RESET, 3-HS-90-257B-A (opens FCV 257B)

[14] At Panel 3-9-54, PLACE H2/02 Analyzer in service per 3-01-76. [15] At Panel 3-9-55, VERIFY DRYWELL OR SUPPRESSION CHAMBER EXHAUST TO SGTS, 3-FIC-84-20, in AUTO with setpoint at 100 scfm. [19] At Panels 3-9-10 and 3-9-1 1, RESTORE Radiation Monitoring System as follows: [19.1] DEPRESS applicable RESET pushbuttons. [19.2] RESTORE Radiation Monitoring System to normal. REFER TO 3-01-90. [20] RESTORE Main Steam System to normal. REFER TO 3-01-1. [22] At Panel 3-9-13, DEPRESS TIP ISOLATION RESET pushbutton, 3-HS-94-7D-S2. Depresses Fault rest pushbuttons on each VFD on Panel 9-4 in able to clear Recirc Drive ATC/BOP Alarms on Panel 9-4A and 4B. 27

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B BOP 3-AOI-64-2D, Group 6 Ventilation System Isolation [1] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s). [2] VERIFY Group 6 isolation valves penetrating Primary Containment are closed. UTILIZE Panel 3-9-3 mimic or Containment Isolation Status System on Panel 3-9-4. [3] IF Refuel Zone Isolation is due to high radiation, as indicated on 3-RM-90-140 and/or 3-RM-90-141, Panel 3-9-10, and/or associated recorder on Panel 3-9-2, THEN. (Otherwise N/A) Step is NA [7] ChECK the following to confirm condition:

REACTOR & REFUEL ZONE EXHAUST RADIATION, 3-RR-90-144
RX & REFUEL ZONE EXH CH A RAD MON RTMR, 3-RM-90-140/142
RX & REFUEL ZONE EXH CH B RAI) MON RTMR, 3-RM-90-141/143

[13] WHEN initiating signal has been corrected AN]) necessary repairs are made, THEN [13.1] VERIFY PCIS RESET: RESET PCIS DIV I RESET, 3-HS-64-16A-S32.

RESET PCIS DIV II RESET, 3-HS-64-16A-S33.

[13.2] RESET Reactor/Refuel isolation logic, as required:

PLACE REFUEL ZONE FANS AND DMPRS, 3-HS-64-3A, in OFF.
PLACE REACTOR ZONE FANS AND DMPRS, 3-HS-64-1 1A, in OFF.

[13.3] START Reactor/Refuel zone ventilation, as required:

PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-1 1A, in SLOW A (SLOW B).
PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B).

[13.4] For the fans started, VERIFY that the dampers open and fans start as indicated by illuminated red lights above the following switches:

The two green lights A(B) above REACTOR ZONE FANS AND DAMPERS Switch 3-HS-64-l 1A, extinguish and the two red lights A(B) illuminate.
The two green lights A(B) above REFUEL ZONE FANS AND DAMPERS Switch 3-HS-64-3A, extinguish and the two red lights A(B) illuminate.

Driver when directed by NRC trigger 5 for Stator water pump trip 28

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Tech Spec Actions from loss of 480V SD BD 3B Evaluate TRM 3.4.1 TSR 3.4.1.1 Monitor reactor coolant conductivity. Continuously OR 4 hours when the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or 5 OR 8 hours when the continuous conductivity monitor is inoperable_and_the_reactor_is_in_MODE_4_or_5 Informs Chemistry have lost Continuous reactor coolant conductivity monitoring SRO Evaluate Tech Spec 3.4.5 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell floor drain sump monitoring system; and

b. One channel of either primary containment atmospheric particulate or atmospheric gaseous monitoring system.

APPLICABILITY: MODES 1,2, and 3. Condition B: Required primary containment atmospheric monitoring system inoperable. Required Action B. 1: Analyze grab samples of primary containment atmosphere. Completion Time: Once per 12 hours Required Action B.2: Restore required primary containment atmospheric monitoring system to OPERABLE status. Completion Time: 30 days Driver when directed by NRC trigger 5 for Stator water pump trip 29

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Tech Spec Actions from loss of 480V SD BD 3B SRO Evaluate Technical Specification 3.6.1.3. 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation. Condition A: NOTE Only applicable to penetration flow paths with two PCIVs. One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits Required Action A. 1: Iso late the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. Completion Time: 8 hours for main steam lines -.. Required Action A.2: NOTE Isolation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated. Completion Time: Once per 31 days for isolation devices outside primary containment Driver when directed by NRC trigger 5 for Stator water pump trip 30

NRC Scenario 4 Simulator Event Guide: Event 4 Component: Stator Water Cooling Pump trip BOP Responds to alanns 7A-22 and 8A-1 BOP Announces trip of Stator Water Cooling Pump 3B 7A-22, GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action of this ARP 3 -XA-5 5-9-8A Window 1 alarms THEN,

1. VERIFY all available Stator Cooling Water Pumps running.

2. Attempt to RESET alarm

3. IF alarm fails to reset, AND reactor power is above turbine bypass valve capability THEN SCRAM the Reactor B. VERIFY a stator cooling water pump is running and CHECK stator temperature recorder, 3-TR-57-59, Panel 3-9-8.

C. CHECK alarm by dispatching personnel to check the Stator Coolant Control Cabinet. 8A-1, TURBINE TRIP TIMER INITIATED A. CHECK Stator Cooling Water Flow and Temperature and Generator Stator temperatures using ICS. B. VERIFY all available Stator Cooling Water Pumps running. NOTE The full capacity of the Turbine Bypass valves with all nine valves open is 25% reactor power. To determine the capacity of the bypass valves, subtract 3% for each out of service bypass valve from the 25%. (Example, one bypass valve out of service, [25% 3% 22%], therefore, the capacity of the bypass valves with one bypass valve out of service is 22%.) C. IF all of the following conditions exist:

Alarm fails to reset,
Low Stator Cooling Water flow OR High Generator or Stator Cooling temperatures are observed on ICS,
Reactor Power is above turbine bypass valve capability, THEN, SCRAM the reactor. (Otherwise N/A)

BOP Starts Stator Water Cooling Pump 3B Driver When dispatched wait two minutes and report pump is extremely hot to touch, at breaker breaker_is tripped no_other indications Driver When directed by NRC to insert RFPT 3A governor failure, veril start value is between 71 and 72. If not modif start value to the current value and ensure fmal severity is set to zero and ensure ramp time remains unchanged and then insert trigger 10 mffw3Oa (elO 0) 0 1300 72, When operator takes manual control of RFPT 3A ensure trigger 11 goes active to allow the operator to control. Be prepared to insert the next event if crew decides to scram, see next page driver instructions. the 31

NRC Scenario 4 Simulator Event Guide: Event 5 Instrument: RFPT 3A Governor slowly fails low ATC Notices lowering speed on RFPT 3A or rising speed on RFPT 3C, or responds alarm 5A-8. 5A-8, REACTOR WATER LEVEL ABNORMAL A. VERIFY Reactor water level hi/low using multiple indications including Average Narrow Range Level on 3-XR-3-53 recorder, 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-LI-3-253 on Panel 3-9-5. B. IF alarm is valid, THEN REFER TO 3-AOI-3-1 or 3-01-3. ATC Report Reactor level less than 27 inches and lowering, reports RFPT 3A flow has lowered. Takes manual control of RFPT 3A to attempt to control RPV Level SRO Directs entry 3-AOI-3 -1, Loss Of Reactor Feedwater or Reactor Water Level High/Low ATC [1] VERIFY applicable automatic actions. [2] IF level OR Feedwater flow is lowering due to loss of Condensate, Condensate Booster, or Feedwater Pump(s), THEN REDUCE Recirc flow as required to avoid scram on low level. [4] IF Feedwater Control System has failed, THEN [4.1] PLACE individual RFPT Speed Control Raise/Lower switches in Manual Governor (depressed position with amber light illuminated). [4.2] ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain level. [24] IF unit remains on-line, THEN PERFORM the following:

RETURN Reactor water level to normal operating level of 33(normal range).
REQUEST Nuclear Engineer check core limits.

Driver when directed by NRC or if the crew decides to scram, verify start value is between 79 and

80. If not modify start value to the current value and ensure final severity is set to 82 and ensure ramp time remains unchanged and then insert trigger 14.

imftclOb (e14 0) 82 1600 79 32

NRC Scenario 4 Simulator Event Guide: Event 6 Instrument: EHC Pressure Transducer Failure Responds to alarm 7B-6, EHC/TSI SYSTEM A. On EHC Workstation computer on Panel 3-9-7, Alarm Summary screen, ATC/BOP ATTEMPT to RESET alarm input. B. IF necessary, THEN REQUEST assistance from Site Engineering. ATC Recognizes lowering Reactor Pressure and generator megawatts. SRO Directs entry into 3-AOI-47-2. 3-AOI-47-2 Turbine EHC Control System Malfunctions 4.1 Immediate Actions [1] IF Reactor Pressure lowers to or below 900 psig, THEN MANUALLY SCRAM the_Reactor and_CLOSE the_MSIVs. 4.2 Subsequent Actions [3] IF a Group 1 isolation has occurred, THEN PLACE EHC PUMP 3A and 3B, 3-HS-47-1A and 3-HS-47-2A, to PULL TO LOCK. Places EHC Pumps 3A and 3B in Pull to Lock ) BOP Directs manual scram, closing of the MSIVs, and entry into 3-AOI-100-l. SRO ATC Manually scrams the reactor. 33

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS ATC 3-AOI-100-1, Reactor Scram 4.1 Immediate Actions [I] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5AIS3A and 3-HS-99-5AIS3B, on Panel 3-9-5. [2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (N/A) [3] Refuel Mode One Rod Permissive Light check [3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (N/A) [4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-Sl, in SHUTDOWN. I ) [5] REPORT the following status to the US:

Reactor Scram
Mode Switch is in Shutdown
All rods in or rods out
Reactor Water Level and trend (recovering or lowering).
Reactor pressure and trend
MSIV position (Open or Closed)
Power level 4.2 Subsequent Actions

[2] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following: [2.1] INITIATE ARI by Arming and Depressing BOTH of the following:

ART Manual Initiate, 3-HS-68-1 19A
ARI Manual Initiate, 3-HS-68-1 19B

[2.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 3-9-4. [2.3] REPORT ATWS Actions Complete and power level. [3] DRiVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit. [3.1] DOWNRANGE IRMs_as_necessary_to_follow_power_as_it lowers. 34

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS SRO Enters EOI- 1 on Reactor Level SRO EOI- 1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO-IF Emergency Depressurization is or has been required THEN exit RC/P and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO - Is any MSRV Cycling? YES, but MSIVs closed IF Steam cooling is required? NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO IF Boron injection is required? NO SRO Directs a Pressure Band with SRVs lAW APPX hA SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs Verification of PCIS isolations. ATCIBOP Verifies PCIS isolations. SRO W It has NOT been determined that the reactor will remain subcritical without boron under all conditions THEN Exit RC/L and Enter C5, Level/Power control. SRO Exits RC/L and Enters 3-C-5, Level/Power Control 35

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS ATCIBOP Commence pressure control with Appendix 1 1A, Alternate RPV Pressure Control Systems MSRVs

1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G.

CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.

2. IF Suppression Pool level is at or below 5.5 ft, THEN CLOSE MSRVs and CONTROL RPV pressure using other options.

3. OPEN MSRVs using the following sequence to control RPV pressure as directed by SRO:

a. I 3-PCV-l-179 MN STM LINE A RELIEF VALVE.

b. 23-PCV-l-180 MN STM LINE D RELIEF VALVE.

c. 3 3-PCV-1-4 MN STM LINE A RELIEF VALVE.

d. 4 3-PCV-1-3 1 MN STM LINE C RELIEF VALVE.

e. 53-PCV-l-23 MN STM LINE B RELIEF VALVE.

f. 63-PCV-l-42 MN STM LINE D RELIEF VALVE.

g. 73-PCV-1-30 MN STM LINE C RELIEF VALVE.

h. 8 3-PCV-l-19 MN STM LINE B RELIEF VALVE.

i. 9 3-PCV-1-5 MN STM LINE A RELIEF VALVE.

j. 10 3-PCV-1-41 MN STM LINE D RELIEF VALVE.

k. 11 3-PCV-1-22 MN STM LINE B RELIEF VALVE.

1. 12 3-PCV-1-18 MN STM LINE B RELIEF VALVE.

m. 133-PCV-1-34 MN STM LINE C RELIEF VALVE.

36

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS SRO EOI- 1 (Power) Monitor and Control Reactor Power Verify Reactor Mode Switch in shutdown Yes Initiate ARI completed Will tripping Recirc Pumps cause trip of main turbine, RFP, HPCI or RCIC No Is reactor power above 5% or unknown No - SLC Leg When periodic APRM oscillations greater than 25% peak to peak persist continue OR Before Suppression Pool temperature rises to 110°F continue Direct SLC injection (APPX 3A) Inhibit ADS Verify RWCU system isolation completed earlier Insert Control Rods Leg Reset AR! and defeat ARI logic trip (APPX 2) Insert Control Rods using any of the following methods: APPX-1A Deenergize scram solenoids No APPX-1B Vent the scram air header No APPX-1C Scram individual control rods No APPX-1D Drive Control Rods Yes APPX-1E Vent control rod over piston No - APPX-1F Reset scram/RE-SCRAM Yes APPX-1G Raise CR1) cooling water dp No - 37

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS ATC Inserting Control Rods Calls for 3-EOI Appendix-2 and the field portion of 3-EOI Appendix-iF 3-EOI Appendix-iF

2. WHEN RPS Logic has been defeated, THEN RESET Reactor Scram.

3. VERIFY OPEN Scram Discharge Volume vent and drain valves.

4. DRAIN SDV UNTIL the following annunciators clear:

WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 1)
EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 29).

5. DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER ISOL.

6. WHEN CR1) Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.

3-EOI Appendix-iD

1. VERIFY at least one CR1) pump in service.

2. IF Reactor Scram or ART CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SOy.

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:

a. SELECT control rod.

b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.

c._REPEAT_Steps_5.a and_5.b_for_each_control rod to_be_inserted. Driver When called for Appendix 2 wait 2 minutes and trigger 24, Appendix-iF wait 3 minutes and trigger 23, when SCRAM is reset trigger 25 to unstick rods. Before the crew scrams or the insert trigger 28 for bat nrcatws95 If requested to close 85-586 trigger 26 to close and trigger 27 to open 38

NRC Scenario 4 Simulator Event Guide: Event 9 Component: Loss of FIPCI 120 VAC Power Supply SRO Enters C-5, LeveliPower Control Inhibit ADS ATC/BOP Inhibits ADS SRO Is any main steam line open No-Is reactor power above 5% or unknown No - Maintain RPV water level between -180 inches and +51 inches with the following injection sources: CRD - APPX 5B, RCIC - APPX 5C, SLC - APPX 7B SRO Directs a Level Band maintained by RCIC ATC Initiate RCIC lAW Appendix-5C and maintains level in directed band, if possible BOP Reports Loss of HPCI 120 VAC Power Supply, HPIC NOT available for Level Control Crew Calls for investigation and repair in order to use NPCI for Level control 39

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C.

1. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRJP/THROT VALVE RESET.

2. VERIFY 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 600 gpm.

5. OPEN the following valves:

                            . 3-FCV-71-39, RCIC PUMP iNJECTION VALVE
                            . 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE
                            . 3-FCV-71-25, RCIC LUBE OIL COOLING WTR VLV.

6. PLACE 3-HS-71-3 IA, RCIC VACUUM PUMP, handswitch in START.

7. OPEN 3-FCV-71-8, RCIC TURBINE STEAM SUPPLY VLV, to start RCIC Turbine.

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at 600 gpm.

c. 3-FCV-71-40, RCIC Testable Check Vlv, opens by observing 3-ZI-71-40A, DISC POSITION, red light illuminated.

d. 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE, closes as flow rises above 120 gpm.

9. IF BOTH of the following exist? NO

10. ADJUST 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller as necessary to control injection.

40

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-l), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Stops the first time through when the LOCA worsens will continue at that time Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 41

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue SRO Enters EOI-2 on High Drywell Pressure SPIT MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, directs RHR Pumps in Pool Cooling Enters EOI-2 on High DryweJi Pressure SPIL MOMTOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18) Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES 42

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/BOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

1. IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock AS NECESSARY:

PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.
PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. VERIFY RIIRSW pump supplying desired RHR Heat Exchanger(s).

c. THROTTLE the following in-service R1-IRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RFIR HX 3D RHRSW OUTLET VLV.

d. IF Directed by SRO, THEN PLACE 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRi) in MANUAL OVERRIDE.

e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.

f. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

g. OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBRJPOOL ISOL VLV.

h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.

CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage. ATC/BOP Aligns directed RHR Pumps in Pool Coolmg 43

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATCIBOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

i. THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-FI-74-50(64), RHR SYS 1(11) FLOW:

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE

k. MONITOR RUR Pump NPSH using Attachment 1.

1. NOTIFY Chemistry that RHRSW is aligned to in-service RI{R Heat Exchangers.

m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional_RHR and RHRSW pumps_in_service_using_Steps_2.b through_2.1.

ATCIBOP . . Aligns directed RHR Pumps in Pool Coolmg 44

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF Adequate core cooling is assured, OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD NJ VLV BYPASS SEL in BYPASS.

3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, TFIEN CONTINUE in this procedure at Step 7.

4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.

5. IMTIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 3-XS-74-121(129), RJ{R SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 3-FCV-74-53(67), RHR SYS 1(11) 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired R}IR System 1(11) pump(s) for Suppression Chamber Spray.

f. VERIFY OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV.

g. OPEN 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.

ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 45

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA ATC/BOP 3-EOI APPENIMX-17C, RHR System Operation Suppression Chamber Sprays

h. IF RHR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k.

i. VERIFY CLOSED 3-FCV-74-7(30), RFIR SYSTEM 1(11) M1N FLOW VALVE.

j. RAISE system flow by placing the second RHR System 1(11) pump in service as necessary.

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RHRSW pump supplying desired RFIR Heat Exchanger(s).

m. THROTTLE the following in-service RFIRSW outlet valves to obtain between 1350 and 4500 gpm flow:

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
3-FCV-23-46, RHR lix 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR HX 3D RFIRSW OUTLET VLV.

n. NOTIFY Chemistry that RE{RSW is aligned to in-service R}IR Heat Exchangers.

AIC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 46

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

1. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

2. IF Adequate core cooling is assured, OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD INJ VLV BYPASS SEL in BYPASS.

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.

5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.

6. INITIATE Drywell Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RER SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI II4BD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD iNJECT VALVE.

e. VERIFY OPERATING the desired System 1(11) RHR pump(s) for Drywell Spray.

f. OPEN the following valves:

3-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV
3-FCV-74-61(75), RI{R SYS 1(11) DW SPRAY INBD VLV.

ATC/BOP . Aligns directed RHR Pumps in Drywell Sprays 47

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

g. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) M1N FLOW VALVE.

h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System II RHR Pump in service.

i. MONITOR RFIR Pump NPSH using Attachment 2.

j. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

k. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
3-FCV-23-46, RNR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

1. NOTIFY Chemistry that R}IRSW is aligned to in-service RHR Heat Exchangers.

) 7. WHEN EITHER of the following exists:

Before drywell pressure drops below 0 psig, OR
Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:

a. VERIFY CLOSED the following valves:

3-FCV-74-100, RHR SYS I U-2 DISCH XTIE
3-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV
3-FCV-74-61(75), RHR SYS 1(11) DW SPRAY INBD VLV.

b. VERIFY OPEN 3-FCV-74-7(30), RHR SYSTEM 1(11) M1N FLOW VALVE.

c. IF RHR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.

d. STOP RHR Pumps.

ATC/BOP Aligns directed RUE. Pumps in Drywell Sprays 48

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA SRO C5 Level/Power Control SRO As Level continues to lower with RCIC injection, directs use of SLC APPX-7B ATC Initiates SLC lAW APPX 7B

2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step 10 to inject SLC Boron Tank to RPV.

10. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A13B, control switch in START PUMP 3A or START PUMP 3B (Panel 3-9-5).

11. CHECK SLC injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch.

                                   . Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,
                                   . SLC SQUII3 VALVE CONTINUITY LOST Annunciator in alarm (3-XA-55-5B, Window 20).
                                   . 3-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

System flow, as indicated by 3-IL-63-1 I, SLC FLOW, red light illuminated,

                                   . SLC INJECTION FLOW TO REACTOR Annunciator in alarm (3-XA-55-5B, Window 14).

12. IF Proper system operation CANNOT be verified, THEN RETURN TO Step 10 and START other SLC pump.

As RPV Level continues to lower, CAN RPV water level be restored and maintained above SRO

                     -180 inches No-Are at least 2 MSRVs open No   -

Emergency Depressurization is Required 3-C-2 and 3-C-S Will the reactor remain subcritical without boron under all conditions NO When all injection into the RPV is stopped and prevented except from RCIC, CRD, and SLC_per_CS,_Level/Power control_Step_C5-22 Stop and Prevent ALL injection into RPV Except from RCIC, CRD, and SLC (APPX 4) 49

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA BOP/ATC Stop and Prevent ALL injection into RPV Except from RCIC, CRD, and SLC (APPX 4)

3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.

4. PREVENT injection from LPCI SYSTEM I by performing the following:

a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 3-HS-74-155A, LPCI SYS I OUTBD 1NJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 3-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE.

5. PREVENT injection from LPCI SYSTEM II by performing the following:

a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 3-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

6. PREVENT injection from CONDENSATE and FEED WATER by performing the following:

a. IF Immediate injection termination from a reactor feedwater pump is required, THEN PERFORM step 6.d for the desired pump.

c. CLOSE the following valves BEFORE RPV pressure drops below 450 psig:

3-FCV-3-19, RFP 3A DISCHARGE VALVE

3-FCV-3-12, RFP 3B DISCHARGE VALVE
3-FCV-3-5, RFP 3C DISCHARGE VALVE
3-LCV-3-53, RFW START-UP LEVEL CONTROL 50

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA SRO C2 Emergency Depressurization and C5 LevellPower Control Is suppression pool level above 5.5 feet Yes Open all ADS Valves BOP/ATC Opens all 6 ADS Valves SRO Can at least two MSRVs be opened per C2 Emergency RPV Depressurization Yes - When RPV pressure is below MSCP Table 1A 190 psig Start and Slowly raise RPV injection with the following injection sources to restore and maintain RPV water level above -180 inches Directs injection with LPCI APPX 6B and 6C to restore RPV Level to directed band BOP/ATC Injects with LPCI JAW APPX 6B and/or 6C to restore RPV water level SRO Emergency Classification EPIP-1 1.1-Si Reactor water level can NOT be maintained above -162 inches. (TAF) OR 1.2-S Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical. 51

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA BOP/ATC Injects with LPCI JAW APPX 6B to restore RPV water level I. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN 3-FCV-74-.1, RHR PUMP 3A SUPPR POOL SUCT VLV.

3. VERIFY OPEN 3-FCV-74-12, RFIR PUMP 3C SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

3-FCV-74-61, RHR SYS I DW SPRAY INBD VLV
3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV
3-FCV-74-57, RHR SYS I SUPPR CHBRJPOOL ISOL VLV
3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

                              *  . 3-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 3A andlor 3C running.

6. WI{EN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RHR SYS I LPCI INBD iNJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 3-FCV-68-79, RECIRC PUMP 3B DISCHARGE VALVE.

8. THROTTLE 3-FCV-74-52, RHR SYS I LPCI OUTBD iNJECT VALVE, as necessary to control injection.

9. MOMTOR RI{R Pump NPSH using Attachment 1.

10. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers discharging to the RPV.

11. THROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:

                               *   . 3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
                               *   . 3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV.

52

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOCA BOP/ATC Injects with LPCI JAW APPX 6C to restore RPV water level

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 3-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN 3-FCV-74-24, RHR PUMP 3B SUPPR POOL SUCT VLV.

3. VERIFY OPEN 3-FCV-74-35, RHR PUMP 3D SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

3-FCV-74-75, RHR SYS II DW SPRAY INBD VLV
3-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV

                              *  . 3-FCV-74-71, RHR SYS II SUPPR CHBR/POOL ISOL VLV

3-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE

                              *  . 3-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 3B and/or 3D running.

6. WI-lEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, RI-JR SYS II LPCI 1NBD INJECT VALVE.

7. IF RPV pressure is below 230 psig, TI{EN VERIFY CLOSED 3-FCV-68-3, RECIRC PUMP 3A DISCHARGE VALVE.

8. THROTTLE 3-FCV-74-66, RHR SYS II LPCI OUTBD iNJECT VALVE, as necessary to control injection.

9. MONITOR RHR Pump NPSH using Attachment 1.

10. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers discharging to the RPV.

11. THROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:

                               *   . 3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
                               *   . 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

53

NRC Scenario 4 SHIFT TURNOVER SHEET Equipment Out of ServiceILCOs: RFPT 3B and EECW Pump A3 Operations/Maintenance for the Shift: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B. Commence a power increase to 85% in accordance with the RCP Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas: The following Control Rods are identified as SLOW: 30-19, 34-23, 14-51, 02-19, 46-51, and 06-43. 54

NRC Scenario 5 aci1ity: Browns Ferry NPP Scenario No.: NRC 5 Op-Test No.: Examiners:_____________________ Operators: SRO:______________________ ATC:_______________ BOP:_______________ Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation. Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing. Event Maif. No. Event Type* Event Description No. N-BOP 1 Return LPRM 8-49B to Operate JAW 2-OI-92B N-SRO R-ATC 2 Commence power decrease with flow to 90% R-SRO C-BOP 3 edl8a Lossofl&CBusA TS-SRO R-ATC 4 adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the standby 5 thl8a C-SRO pump to auto start C-ATC LOCA Recirculation Pump 2A Inboard and Outboard seal 6 R-ATC thlO/1 la failure TS-SRO Two Level instruments fail high tripping Feedwater and HPCI / 7 Batch File M-ALL LOCA / ED on Reactor Level 8 edl0a C Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not Auto 9 Batch I open 10 rcO8 C RCIC Steam Valve fails to Auto open

      *  (N)ormal,    (R)eactivity, (I)nstrument,     (C)omponent,    (M)ajor 1

4*

NRC Scenario 5 Critical Tasks Three With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates at least one low pressure ECCS system and injects into the RPV to restore water level above -162 inches.

4. Feedback:

Reactor water level trend. Reactor pressure trend. With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to -180 inches.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance. Water level trend.

3. Measured by:

Observation At least 6 SRVs opened

4. Feedback:

RPV pressure trend. SRV status indications.

NRC Scenario 5 Critical Tasks Three To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS AIB iNHIBITED annunciator status.

NRC Scenario 5 Events

1. BOP operator will return LPRM 8-49B to Operate lAW 2-OI-92B.

2. ATC lowers power to 90% using recirculation flow.

3. The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high H2 will result in Off Gas, the SRO will evaluate TRM 3.7.2 and enter Condition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.

4. During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to I&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.

5. The VFD Cooling Water Pump for the A Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water preventing a VFD and Reactor Recirc Pump trip.

6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump lAW with 2-A0I-68-iA. ATC will insert control rods to exit Region 2 of the power to flow map. SRO will determine Technical Specification 3.4.1 Condition A, is applicable again with 24 hours to establish single loop conditions.

7. Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.

8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but no throttle capability with exist. RHR Loop 1 will not be available for Containment cooling operation.

9. With Division 2 Accident logic bypassed RHR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RHR Loop 2 will be available for Containment Cooling functions until required for injection.

A

NRC Scenario 5 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOl entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) c

NRC Scenario 5 Scenario Tasks TASK NUMBER RQ $R.Q Restore an LPRM from Bypass RO U-92B-NO-05 215005 A4.04 3.2 3.2 Lower Power with Recirc Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 262001 A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-001-AB-01 239002 A2.03 4.1 4.2 SRO 5-001-AB-Ol VFD Cooling Water Pump Failure RO U-068-AL-19 202001 A2.22 3.1 3.2 SRO 5-068-AB-Ol RR Pump Seal Failure RO U-068-AL-09 202001 A2.10 3.5 3.9 SRO 5-068-AB-Ol Loss of 480V SD BD 2A RO U-57B-AL-06 262001 A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 295031 EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-15 SRO 5-000-EM-Ol

NRC Scenario 5 ( aci1ity: Browns Ferry NPP Scenario No.: NRC 5 Op-Test No.: 1404 Examiners: Operators: SRO:_____ ATC:___ BOP: Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation. Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing. Event Maif. No. Event Type* Event Description No. N-BOP 1 Return LPRM 8-49B to Operate lAW 2-OI-92B N-SRO R-ATC 2 Commence power decrease with flow to 90% R-SRO C-BOP 3 edl8a Loss of I&C Bus A TS-SRO R-ATC 4 adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the standby 5 thl8a C-SRO pump to auto start C-ATC LOCA Recirculation Pump 2A Inboard and Outboard seal 6 R-ATC thlO/1 ia failure TS-SRO Two Level instruments fail high tripping Feedwater and HPCI / 7 Batch File M-ALL LOCA / ED on Reactor Level 8 edl0a C Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not Auto 9 Batch I open 10 rcO8 C RCIC Steam Valve fails to Auto open

       *   (N)ormal,    (R)eactivity, (I)nstrument,     (C)omponent,   (M)ajor 1

NRC Scenario 5 Critical Tasks Three With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates at least one low pressure ECCS system and injects into the RPV to restore water level above -162 inches.

4. Feedback:

Reactor water level trend. Reactor pressure trend. With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to -180 inches.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance. Water level trend.

3. Measured by:

Observation At least 6 SRVs opened

4. Feedback:

RPV pressure trend. SRV status indications. 7

NRC Scenario 5 Critical Tasks Three To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS A/B iNHIBITED annunciator status. 2

NRC Scenario 5 Events

1. BOP operator will return LPRM 8-49B to Operate lAW 2-OI-92B.

2. ATC lowers power to 90% using recirculation flow.

3. The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high 112 will result in Off Gas, the SRO will evaluate TRM 3.7.2 and enter Condition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.

4. During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to I&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i -l actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.

5. The VFD Cooling Water Pump for the A Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water preventing a VFD and Reactor Recirc Pump trip.

6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump lAW with 2-A0I-68-1A. ATC will insert control rods to exit Region 2 of the power to flow map. SRO will determine Technical Specification 3.4.1 Condition A, is applicable again with 24 hours to establish single loop conditions.

7. Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.

8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. R}IR Loop 1 may be used for injection but no throttle capability with exist. RHR Loop 1 will not be available for Containment cooling operation.

9. With Division 2 Accident logic bypassed RHR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RHR Loop 2 will be available for Containment Cooling functions until required for injection.

A

NRC Scenario 5 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) c

NRC Scenario 5 Scenario Tasks TASK NUMBER SRO Restore an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 Lower Power with Recirc Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 262001A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-OO1-AB-O1 239002A2.03 4.1 4.2 SRO 5-001-AB-Ol VFD Cooling Water Pump Failure RO U-068-AL-19 202001A2.22 3.1 3.2 SRO 5-068-AB-Ol RR Pump Seal Failure RO U-068-AL-09 203000A4.02 4.1 4.1 SRO 5-068-AB-Ol Loss of 480V SD BD 2A RO U-57B-AL-06 226001A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO 5-000-EM-iS SRO S-000-EM-01

NRC Scenario 5 Procedures Used/Referenced: Procedure Number Procedure Title 2-0I-92B Average Power Range Monitoring 2-G0I-100-12 Power Maneuvering 2-01-68 Reactor Recirculation System 2-A0I-57-5A Loss of I&C Bus A 2-ARP-9-8C Panel 9-8 2-XA-55-8C 2-ARP-9-7A Panel 9-7 2-XA-55-7A 2-ARP-9-6C Panel 9-6 2-XA-55-6C 2-ARP-9-7C Panel 9-7 2-XA-55-7C 2-ARP-9-3C Panel 9-3 2-XA-55-3C 2-ARP-9-3D Panel 9-3 2-XA-55-3D 2-ARP-9-53 Panel 9-53 2-XA-55-53 2-E0I-3 Secondary Containment Control 2-A0I-64-2D Group 6 Ventilation System Isolation

             .           Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 2-E0I Appendix-8F Isolation 2-01-66                   Off-Gas System 2-A0I-66-l                0ff-Gas H2 High Technical Specifications Technical Requirements Manual 2-AOl-i -1                Relief Valve Stuck open 2-01-74                   Residual Heat Removal System 2-E0I-2                   Primary Containment Control 2-E0I Appendix-i 8        Suppression Pool Water Inventory Removal and Makeup 2-ARP-9-4A                Panel 9-4 2-XA-55-4A 2-A01-68-iA               Recirc Pump Trip/Core Flow Decrease OPRMs Operable 2-AOI-lOO-1               Reactor Scram 2-E0I-l                   RPV Control 2-EOI Appendix-8B         Reopening MSIVs / Bypass Valve Operation 2-EOI-i-C-i               Alternate Level Control 2-E0I Appendix-6A         Injection Subsystems Lineup Condensate 2-EOI Appendix-i 7C       RHR System Operation Suppression Chamber Sprays 2-EOI Appendix-i 7B       RHR System Operation Drywell Sprays 2-EOI-3-C-2               Emergency RPV Depressurization 2-EOI Appendix-5B         Injection System Lineup CRD 2-EOI Appendix-7B        Alternate RPV Injection System Lineup SLC System 2-E01 Appendix-6B        Injection Subsystems Lineup RHR System I LPCI Mode 2-EOI Appendix-6C         Injection Subsystems Lineup RFLR System II LPCI Mode 2-E0I Appendix-6E         Injection Subsystems Lineup Core Spray System II EPIP- 1                   Emergency Classification 7

NRC Scenario 5 Console Operator Instructions A. Scenario File Summary Batch File NRC/l3O6nrc-5 ior zloOil2l ld2Ob[1] off ior zloOil2l ld2Ob[2J off ior zloOhs2l lOd2Oa[l] off Tag DG D ior zloOhs2l lOd2Oa[2] off ior zloOhs2l lOd2Oa[2] off mrfdg0ld open ior zdihs7O8a null ior zlohs7o8a[lj off Tag RBCCW 2B ior zlohs708a[2] off ior zlohs708a[3] off ior zlohs682a2a[1j on ior zlohs682a2a[2] off A VFD Cooling Pump Trip mrfthl8b trip trg 1 NRC/avfd trg 1= bat NRC/130605-1 A VFD Cooling Pump Trip imfth30f (e5 0)100 imfth30h (e5 60) 100 45 55 Level 8 instrument failures imfrc08 RCIC steam supply valve failure imfthl0a(e3 0)100 imfthl la (e3 60) 100 90 0 RR 2A Pump seal mrf cs09b inhibit mrfrhl5 inhibit Div 2 accident logic bypassed ior zloil75S6a off ior zloil74 I 54a off mrfedl3 open momentary loss of I&C Bus A Batch File NRC/l3O6nrc-5-1 imfth2l (none 330) .6 600 .1 LOCA imfedl0a (none 370) Loss of 480V SD BD 2A 2

NRC Scenario 5 Preference File NRC/l3O6nrc-5 pfk 01 tog pflc 02 ann silence pfk 03 mrf swO2 align align spare RBCCW Pump pfk 04 bat NRC/l3O6nrc-5 pfk 05 imfedl8a Loss of I&C Bus A pfk 06 ior zdihs682ala[1] off VFD A Cooling Pump trip pflc 07 imfadOic 10 ADS SRV Leak by pfk08trg!e3 RR Pump A Seal Failure pfko9trg! e5 Loss of Feedwater pfk 10 bat NRC/i3O6nrc-5-1 LOCA and Loss of 480V SD BD 2A pfk 11 mrfad0lc out pflc 12 ior zdixs0l22 Back up control panel switch to emergency pfk si pfk s2 pfk s3 pflc s4 ior zdihs0l22c open Back up control panel srv 1-22 pfk s5 ior zdihs0l22c close Back up control panel srv 1-22 pfk s6 bat appl8rhra pflc s7 bat appl8rhrb pfk s8 mrfedi3 close Scenario 5 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 28 manual Bypass LPRM 8-49B restorepref NRC/i 3O6nrc-5 mrf swO2 align RBCCW wait one F3 minute and turn off RBCCW Pump 2B Simulator Setup Load Batch F4 bat NRC/l3O6nrc-5 Simulator Setup manual Tag DG D and RBCCW Pump 2B Simulator Setup Verify file loaded, Clear alarms for Reactor Recirc RCP required (100% -90% with flow) and RCP for Urgent Load Reduction a

NRC Scenario 5 .. Simulator Event Guide: Event 1 Normal: Return LPRM 8-49B to Operate from a Bypassed Condition lAW 2-OI-92B Driver At NRC direction call the control room as Reactor Engineer and request LPRM 08-49B be returned to service SRO Directs BOP to return LPRM 8-49B to Operate JAW 2-OI-92B BOP Return LPRM 8-49B to Operate JAW 2-OI-92B 6.4 Returning an LPRM to Operate From a Bypassed Condition [1] REVIEW all precautions and limitations. REFER TO Section 3.0. [2] REFERENCE Illustration 4 to find the APRM/LPRM Channel associated with the desired LPRM to be returned to normal. [3] At Panel 2-9-14, DEPRESS any softkey to illuminate the display on the desired APRM!LPRM channel chassis. [4] DEPRESS the ETC softkey until BYPASS SELECTIONS illuminates on the bottom row of the display. [5] DEPRESS BYPASS SELECTIONS softkey, enter the password, and DEPRESS ENT. [6] SELECT the desired LPRM to be returned to service by using the left or right arrows on the softkey board until the inverse video illuminates the correct LPRM. [7] DEPRESS the OPERATE softkey. [8] CHECK the BYP/HV OFF is replaced by OPERATE below the selected LPRM. [9] DEPRESS EXIT softkey to return display to the desired bargraph. [10] VERIFY, as a result of returning this LPRM to operate, that any alarms received on Panel 2-9-5 or on the APRM/LPRM channel are reset. in

NRC Scenario 5 Simulator Event Guide: Event 2 Reactivity: Power decrease with Recirc Flow SRO Notifies ODS of power decrease. Directs Power decrease using Recirc Flow, lAW 2-GOI-100-12. [1] REVIEW all Precautions and Limitations listed in Section 3.0. [2] VERIFY Prerequisite listed in Section 4.0 is satisfied. [3] NOTIFY Operations Duty Specialist (ODS) and Chattanooga Load Coordinator of impending power reduction. [4] NOTIFY Radiation Protection of purpose for power reduction, the target power level (see above note), and RECORD time Radiation Protection notified in NOMS Narrative Log. [6] IF power is being reduced (less than 10%) for any of the following reasons:

Weekly Control Rod Exercise
Main Turbine Valve Testing
Ultimate heat Sink temperature> 92.5°F

[6.1] REDUCE Recirculation flow. REFER TO 2-01-68. [6.2] MAINTAIN Reactor thermal power within the limits shown on ICS and 0-Tl-248, Station Reactor Engineer, as appropriate. [10] PERFORM the following while reducing Reactor power: [10.1] WHEN Reactor power is at approximately 90%, THEN REFER TO 2-01-3 and START a REP Injection Water Pump. ATC Lower Power w/Recirc, JAW 2-01-68, Section 6.2 Driver When directed by NRC, insert preference key F5 imfedl8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close NRC Two additional power decreases in Scenario, can continue when ready. will have to mismatch speeds at 1300 rpm. 11

NRC Scenario 5 Simulator Event Guide: Event 2 Reactivity: Power decrease with Recirc Flow ATC Lower Power w/Recirc, lAW 2-01-68, Section 6.2 D. Individual pump speeds should be mismatched by --60 RPM during dual pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short periods for testing or maintenance). [1] IF desired to control Recirc Pumps 2A andlor 2B speed with Recirc Individual Control, THEN PERFORM the following:

RAISE Recirc Pump 2A using RAISE SLOW (MEDIUM), 2-HS-96-15A(15B).

(Otherwise N/A)

LOWER Recirc Pump 2A using SLOW(MEDIUM)(FAST),

2-HS 1 7A( 1 7B)( 1 7C). (Otherwise N/A). AND/OR

RAISE Recirc Pump 2B using RAISE SLOW (MEDIUM), 2-HS-96-16A(16B).

(Otherwise N/A)

LOWER Recirc Pump 2B using SLOW(MEDIUM)(FAST),

2-HS-96-18A(18B)(18C). (Otherwise N/A). [2] WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump Speed 2A & 2B using the following pushbuttons as required. RAISE SLOW, 2-HS-96-3 1 RAISE MEDIUM, 2-HS-96-32 LOWER SLOW, 2-HS-96-33 LOWER MEDIUM, 2-HS-96-34 LOWER FAST, 2-HS-96-35 Driver When directed by NRC, insert preference key F5 imf edi 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close. When dispatched wait two minutes and report Failure of 9-9 Throwover Switch, switch tripped to alternate. NRC Two additional power decrease in Scenario, can continue when ready 17

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are_8C-21,_6C-12,_3C-25,_7C-22_and_3D-3,_19,_and_32. ATC Announces Power, Level, and Pressure are stable BOP Alarm 8C-21, I&C BUS A VOLTAGE ABNORMAL A. VERIFY alarm by checking the following:

Loss of instrument power and remote position indication to Core Spray Div I and RHR Div I (Panel 9-3).
RWCU Filter Demin A isolation.
Reactor Building/Refuel Zone Ventilation isolation.

B. NOTIFY Unit 3 Unit Supervisor. C. REFER TO 2-AOI-57-5A and 0-GOI-300-2. SRO Aimounce entry to 2-AOI-57-5A, Loss of I&C Bus A. ATC Alarm 6C-12, RFPT GOVERNOR POWER FAILURE OR GOV ABNORMAL A. VERIFY RFPT/RFPs continue to control Reactor Water Level. B. IF a RFPTIRFP has tripped, THEN VERIFY other RFPTs in Automatic operation raise or lower output flow to maintain reactor water level. C. DISPATCH personnel to UNIT 2 Auxiliary Instrument Room to PERFORM the following at Panels 2-9-48,49,50:

CHECK Power Supply lights illuminated.
CHECK display screens for Governor abnormal conditions.

Announced RPV Level stable, dispatches personnel BOP Alarm 3C-25, MAIN STEAM RELIEF VALVE OPEN A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 2-TR-l-1, on Panel 2-9-47 and SRV Tailpipe Flow Monitor on Panel 2-9-3 for raised temperature and flow indications. B. REFER TO 2-AOl-I-I. BOP Announce Main Steam Relief Valve Open alarm cleared, but have indication on acoustic monitor of SRV partially open or leaking by. ADS SRV 1-22 3C-25 alarms on a loss of I&C Bus A, when the bus re-energizes ADS SRV will show NOTE acoustic monitoring indication of leaking by. BOP operator should report to SRO and SRO enter 2-AOl-i-i. These events will occur under event four.

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32. BOP Alarm 7C-22, DRYWELL/SUPPR CHAMBER H202 ANALYZER FAILURE A. CHECK Panel 2-9-54 and 2-9-5 5 for abnormal indicating lights such as low flow, H2 or 02 downscale, pump off, etc. B. IF sample pump is NOT running, THEN ATTEMPT to start pump using 2-HS-76-l 10/S5. C. IF sample pump will NOT start OR H2/02 analyzer malfunction, THEN PLACE H2/02 Analyzer in Service per 2-01-76 section 5.4. D. REFER TO TRM 3.3.11 and TRM Section 3.6.2. BOP Resets H2/02 ANALYZER ISOLATION RESET, 2-HS-76-9l Resets Alarm on 2-MON-76-l 10, touch screen. BOP Alarm 3D-19, DRYWELL LEAK DETECTION RADIATION DNSC A. DETERMINE cause of alarm by performing the following:

1. CHECK AIR PARTICULATE MONITOR CONTROLLER, 2-MON-90-50 on Panel 2-9-2 for condition bringing in alarm

2. DISPATCH personnel to determine which alarm is annunciating using the HELP button (REFER TO 2-01-90 for complete annunciator list).

E. REFER TO Tech Specs 3.4.4, 3.4.5, and TRM 3.3.10 for CAM LCO requirements and IMPLEMENT appropriate TS/TRM actions as required. F. WHEN conditions permit, THEN RESET alarm per 2-01-90, Section 6.5. BOP Determines DW Radiation Monitor Cam isolated, resets the following to restore to operation. UPPER INBD SUPPLY ISOL VALVE RESET, 2-HS-90-254A-A LOWER 1NBD SUPPLY ISOL VALVE RESET, 2-HS-90-254B-A OUTBD RETURN ISOL VALVE RESET, 2-HS-90-257A-A OUTBD SUPPLY ISOL VALVE RESET, 2-HS-90-255A 1NBD RETURN ISOL VALVE RESET, 2-HS-90-257B-A 14

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32. BOP Alarms 3D-3, RX BLDG VENIILATION ABNORMAL A. IF PCIS group 6 isolation exists, THEN REFER TO 2-AOI-64-2d. B. NOTIFY Unit Supervisors, Unit 1 and Unit 3. C. VERIFY standby fans start. D. DISPATCH personnel to check Bldg zP (PDIC 64-2, El 639, Rx Bldg.) E. IF zP is at or above -0.17 in. H20 THEN ENTER 2-EOI-3 Flowchart, 2-XA-55-3D, window 32. BOP Alarms 3D-32, REACTOR ZONE DIFFERENTIAL PRESSURE LOW D. IF alarm is valid, THEN INFORM Unit Supervisor of 2-EOI-3 entry condition. E. REQUEST personnel to check fans locally for any apparent problems. F. REFER TO 2-OI-30B and PLACE standby fan in service to restore normal differential pressure. SRO Enters 2-EOI-3, Secondary Containment Control and 2-AOI-64-2D, Group 6 Ventilation System Isolation Directs Reactor and Refuel Zone Ventilation returned to service by either 2-EOI Appendix-8F, Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation or 2-AOI-64-2D The above procedures for restoring ventilation are basically the same will describe NOTE Appendix-8F below. The only action in EOI-3 is to restore ventilation. 15

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A ATC/BOP Appendix 8F Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation VERIFY PCIS Reset.

2. PLACE Refuel Zone Ventilation in service as follows (Panel 2-9-25):

a. VERIFY 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch is in OFF.

b. PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch to SLOW A (SLOW B).

c. CHECK two SPLY/EXH A(B) green lights above 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch extinguish and two SPLYJEXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

2-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR
2-FCO-64-6, REFUEL ZONE SPLY INBD ISOL DMPR
2-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR
2-FCO-64-1O, REFUEL ZONE EXH INBD ISOL DMPR.

N

3. PLACE Reactor Zone Ventilation in service as follows (Panel 2-9-2 5):

a. VERIFY 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.

b. PLACE 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch in SLOW A (SLOW B).

c. CHECK two SPLY/EXH A(B) green lights above 2-HS-64-l 1A, REACTOR ZONE FANS AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

2-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR
2-FCO-64-14, REACTOR ZONE SPLY INBD ISOL DMPR
2-FCO-64-42, REACTOR ZONE EXH INBD ISOL DMPR
2-FCO-64-43, REACTOR ZONE EXH OUTBD ISOL DMPR.

5. IF Reactor Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).

6. IF Refuel Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).

16

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of 1&C Bus A SRO Enters 2-AOI-57-5A 4.2 Subsequent Actions [1] VERIFY Automatic Actions have occurred. [2] IF a Reactor Scram occurs, THEN PERFORM 2-AOl- 100-I concurrently with this procedure. [3] VERIFY a flow path for Condensate System, or STOP the condensate pumps/booster pumps. REFER TO 2-01-2. [4] START Standby Gas Train(s) and CHECK Reactor Building pressure at or below 0.25 H20 vacuum (PDIC 64-1, Panel 25-215; PDIC 64-2, Panel 25-213). REFER TO 0-01-65, Section Standby Gas Treatment System Manual Initiation. [5] VERIFY SJAE B in service to maintain condenser vacuum. REFER TO 2-01-66. [6] IF Auto Transfer of Panel 2-9-9, Cabinet 2, failed THEN (otherwise N/A) [7] WHEN Reactor water level is normal, THEN RESET PCIS Group 6 inboard ) isolation and RETURN the affected systems to service or standby readiness. REFER TO 2-AOl- 100-1, if a Reactor Scram occurred, otherwise REFER TO 2-AOI-64-2D. SRO Directs restoration of Reactor Building DP, should restore Ventilation lAW Appendix-8F or 2-AOI-64-2D._May_call_Unit_1_to_start_Standby_Gas_Fans SRO Directs restoration of SJAE, lAW 2-01-66 hard card BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card 17

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card [1] VERIFY RESET Off-Gas isolation using 2-HS-90-155, OG OUTLET/DRAIN ISOLATION VLVS. NOTE With power back to I&C Bus A, once RO resets 2-HS-90-155, can place SJAE A back in service or can transfer to SJAE B. All steps are listed below for either. [2] VERIFY OPEN the following valves:

2-HS-66-1 1(15), SJAE 2A(2B) INLET VALVE.
2-HS-1-155A(156A), STEAM TO SJAE 2A(2B).

[3] VERIFY in AUTO/OPEN 2-HS-66-14(18), SJAE 2A(2B) OG OUTLET VALVE. [4] PLACE 2-HS-1-150(152), SJAE 2A(2B) PRESS CONTROLLER, in CLOSE and then in OPEN. [5] VERIFY OPEN the following valves (red light illuminated):

2-PCV-1-151/166 (153/167), STEAM TO SJAE 2A(2B) STAGES 1,2, AND 3.
2-FCV-1-150(152), SJAE 2A(2B) INTMD CONDENSER DRAiN.

[6] MONITOR hotwell pressure as indicated on recorder 2-XR-2-2, HOTWELL TEMP AND PRESS, on Panel 2-9-6. [7] FOR the SJAE not being placed in service, VERIFY CLOSED the following valves:

2-HS-66-18(14), SJAE 2B(2A) OG OUTLET VALVE.
2-HS-1-152(150), SJAE 2B(2A) PRESSURE CONTROLLER.
2-HS-1-156A(155A) STEAM TO SJAE 2B(2A)

BOP Acknowledge Panel 2-9-53 Alarms, Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % H2 TRAIN A, and HIGH OFFGAS % H2 TRAIN BB 1R

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of 1&C Bus A BOP Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % 112 TRAIN A, and HIGH OFFGAS % H2 TRAIN BB A. CHECK Off-gas Hydrogen Analyzer, 2-H2R-66-96 (CH 1) on Panel 2-9-53 to verif 142 concentration. B. IF alarm is valid, THEN REFER TO 2-AO1-66-1. SRO Enters 2-AOI-66-1, Off-Gas 112 High BOP/ATC [1] PLACE both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK I BYPASS control switches, 2-HS-066-1007 and 1008, on OFFGAS SAMPLE PANEL, 2-LPNL-925-0588, in BYPASS to assure continuous availability of hydrogen monitoring. [2] IF HWC System injection is in service, THEN (otherwise N/A) [3] VERIFY proper operation of in service SJAE. [4] IF hydrogen concentration is greater than or equal to 4%, THEN REFER TO TRM 3.7.2. j, [10] MOMTOR the following parameters at Control Room Panel 9-53 and 9-8:

RECOMBINER 2AI2B TEMPERATURE, 2-TRS-66-77, for abnormal trend; I.

either rising or lowering.

OFF GAS HYDROGEN ANALYZER, 2-H2R-66-96, for hydrogen concentration.

NOTE H2 concentration will rise to 8 to 12% and return to a normal value of less than 1% 10

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A SRO Tech Specs for Loss of I&C Bus A For Drywell CAM 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell floor drain sump monitoring system; and

b. One channel of either primary contaimnent atmospheric particulate or atmospheric gaseous monitoring system.

APPLICABILITY: MODES 1,2, and 3. Condition B: Required primary containment atmospheric monitoring system inoperable. Required Action B. I: Analyze grab samples of primary contaimnent atmosphere. Completion Time: Once per 12 hours Required Action B.2: Restore required primary containment atmospheric monitoring system to OPERABLE status. Completion Time: 30 days For High H2 TR 3.7.2 Airborne Effluents LCO 3.7.2 Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to < 4% by volume. APPLICABILITY: During main condenser offgas treatment system operation Condition A: With the concentration of hydrogen >4% by volume. Required Action A. 1: Restore the concentration to within the limit. Completion Time: 48 hours Below is event 4 which started with the failure of I&C Bus A, depending on SRO priorities NOTE may have addressed SRV first and then restoration from Bus loss. 20

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A SRO Tech Specs for Loss of I&C Bus A For H202 Monitor TR 3.3.11 Hydrogen Monitoring Instrumentation LCO 3.3.11 The primary containment hydrogen analyzer shall be OPERABLE APPLICABILITY: MODE 1 during the time period

a. From 24 hours after THERMAL POWER is> 15% RTP following startup, to

b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Condition A: Primary containment hydrogen analyzer inoperable. Required Action A. 1: Restore primary containment analyzer to OPERABLE status. Completion Time: 7 days TR 3.6.2 Oxygen Concentration Monitor LCO 3.6.2 The Primary Containment oxygen concentration monitor shall be OPERABLE. APPLICABILITY: MODE 1 during the time period

a. From 24 hours after THERMAL POWER is> 15% RTP following startup, to

b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Condition A: Primary containment oxygen concentration monitor inoperable. Required Action A. 1: Begin alternate sampling and analyze results. Completion Time: Immediately AND Once per 7 days thereafter. Below is event 4 which started with the failure of I&C Bus A, depending on SRO priorities NOTE may have addressed SRV first and then restoration from Bus loss. 21

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking Event 4 started with the failure of I&C Bus A, depending on SRO priorities may have NOTE addressed SRV first and then restoration from Bus loss. SRO Enters 2-AOl-i-i BOP 4.1 Immediate Action [1] IDENTIFY stuck open relief valve by OBSERVING the following:

SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3, OR
MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-1-1 on Panel 2-9-47.

ATC [2] IF relief valve transient occurred while operating above 90% power, THEN REDUCE reactor power to 90% RTP with recirc flow. BOP [31 WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch several times as required:

CLOSE to OPEN to CLOSE positions

[4] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (N/A) . 4.2 Subsequent Action J 4.2.2 Attempt to close valve from Panel 9-3: [1] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the OFF position. [2] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the ON position. [3] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (N/A) [4] PLACE MSRV AUTO ACTUATION LOGIC 1NFIIBIT, 2-XS-1-202 in INHIBIT: [5] IF relief valve closes, THEN OPEN breaker or PULL fuses as necessary using Attachment 1 (Unit 2 SRV Solenoid Power Breaker/Fuse Table). [6] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT 2-XS-1-202, in AUTO. [7] IF the SRV valve did not close, THEN PERFORM the appropriate section from table below. RELIEF STEP Switch Breaker Fuse VALVE Number Location Location Location SRV 1-22 Step 4.2.3[2] Panel 25-32 Multiple Panel 25-32

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking Actions for SRV 1-22, wait two minutes and for taking control at panel 25-32 Preference key Fl 2, for cycling SRV preference key shift F4 open, shift F5 close to 10, shift F4 open,

shift F5 close to 10.

Dnver . Contact control room and determme if valve closed. When told to remove power preference key Fl 1. When back to normal at panel 25-32 delete override for annunciator xa553e10. When told to power back up srv 1-22 mrf adO Ic in. Driver [2] IF 2-PCV-1-22 is NOT closed, THEN PERFORM the following: [2.1] On Panel 2-25-32 PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR,, 2-XS-l-22 in EMERG position. [2.2] IF the SRV does NOT close, THEN PERFORM the following while OBSERVING the indications for the 2-PCV-1-22 on the Acoustic Monitor: CYCLE the MAIN STM LINE B RELIEF VALVE, 2-HS-1-22C to the following positions several times. CLOSE/AUTO to OPEN to CLOSE/AUTO [2.3] IF the SRV does NOT close, THEN PERFORM the following: A. VERIFY the MAIN STM LINE B RELIEF VALVE, 2-HS-1-22C, in the CLOSE/AUTO position. B. PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-1-22 in NORM position. r)river [2.4] IF the SRV does NOT close, THEN REMOVE the power from 2-PCV-1-22 by performing one of the following: A. OPEN the following breakers (Preferred method) [2.5] IF the valve does NOT close, THEN CLOSE the breakers or REINSTALL fuses removed in Step 4.2.3 [2.4]. BOP [2.6] CONTINUE at Step 4.2.4. 23

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 lealdng BOP [2.6] CONTINUE at Step 4.2.4. 4.2.4 Other Actions and Documentation [1] NOTIFY Reactor Engineering of current conditions. [2] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s). [3] REFER TO Technical Specifications Sections 3.5.1 and 3.4.3 for Automatic Depressurization System and relief valve operability requirements. [4] IMTIATE suppression pooi cooling as necessary to maintain suppression pool temperature less than 95°F. [5] IF the relief valve can NOT be closed AND suppression pool temperature CANNOT be maintained less than or equal to 95°F, THEN PLACE the reactor in Mode 4 in accordance with 2-GOI-100-12A. [6] DOCUMENT actions taken and INITIATE Work Order (WO) for the valve. SRO Directs Suppression Pool Cooling JAW 2-01-74

1) BOP Initiates Pool Cooling as directed SRO Refers to Tech Specs 3.5.1 ECCS Operating LCO 3.5.1 Each ECCS injectionlspray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (F1PCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure < 150 psig. Condition E: One ADS valve inoperable. Required Action E. 1: Restore ADS valve to OPERABLE status. Completion Time: 14 days. BOP/ATC Inform SRO when Suppression Pool Level meets EOI-2 entry requirements SRO Enter EOI-2 on Suppression Pool Level NOTE One RHR Pump will almost maintain pool temperature depending on reactor power, Do NOT expect pool temperature to exceed 95°F. 24

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed 8.5 Initiation of Loop 1(11) Suppression Pool Cooling CAUTION PSA concerns with RHR in Suppression Pool Cooling Mode with a LOCA and a LOSP identify that severe water hammer may occur during the pump restart. Therefore, the following guidelines should be used to try and maintain the system below the PSA Risk Assessment goals:

RHR in suppression pool cooling should be minimized.
Two Loops of RHR in suppression pool cooling should be minimized.
Use two pumps per loop, if needed, to minimize total time spent in suppression pool cooling.
Suppression pool cooling run times are tracked in 2-SR-2 to ensure risk assessment goals are not exceeded.

NOTES

1) Suppression Pool Cooling is required to be initiated whenever necessary to maintain suppression pool temperature less than 95°F or when directed by other procedures.

[1] VERIFY RHR Loop 1(11) is in Standby Readiness. REFER TO Section 4.0 [2] REVIEW the precautions and limitations in Section 3.0. [3] NOTIFY other units of placing Loop 1(11) of RHR in suppression pool cooling, the subsequent start of common equipment (i.e., RHRSW pumps) and associated alarms are to be expected. [4] NOTIFY Radiation Protection for impending action to initiate Suppression Pool Cooling. RECORD name and time of Radiation Protection representative notified in NOMS narrative log [5] IF possible, THEN BEFORE placing RHRSW in service, NOTIFY Chemistry that RHRSW sampling is to be initiated (RHRSW sampling requirements). [6] VERIFY at least one RHRSW Pump is operating on each EECW Header. NOTE One RHR Pump will almost maintain pooi temperature constant depending on reactor power, Do NOT expect pool_temperature to exceed 95°Fr 25

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed [7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows: [7.1] START an RHRSW Pump to supply RHR Heat Exchanger A(C). [7.2] ESTABLISH RHRSW flow by performing one the following: [7.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, (RHRSW Pump A(C) and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23. OR [7.2.2] THROTTLE OPEN RHR HX 2A(2C) R}IRSW OUTLET VLV, 2-FCV-23-34(40), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm R1{RSW flow as indicated on 2-FI-23 -3 6(42), RFII{ IITX 2A(2C) RHRSW FLOW. [7.3] VERIFY CLOSED RHR SYS I LPCI INBD iNJECT VALVE, 2-FCV-74-53. [7.4] II? NO RHR PUMP (1A OR 1C) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RHR. SYS I SUPPR POOL CLG/TEST VALVE, 2-FCV-74-59. [7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-5 8. [7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60. [7.7] VERIFY OPEN R}IR SYS I SUPPR C1{BR/POOL ISOL VLV, 2-FCV-74-57. [7.8] IF desired to operate without the Drywell DP Compressor, THEN: [7.9] START RHR PUMP A(C) using 2-HS-74-5A(16A). [7.10] THROTTLE RHR SYS I SUPPR POOL CLG/TEST VLV, 2-FCV-74-59, to maintain RHR flow within limits, as indicated on RHR SYS I CTMT SPRAY FLOW, 2-FI-74-56: RHR Pumps in 1 2 Operation Loop Flow 7,000 to <13,000 gpm & 10,000 gpm & Blue Blue light Iiqht illuminated illuminated 26

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed [7.11] IF desired to raise Suppression Pool Cooling flow and only one Loop I pump is in service, ThEN PLACE the second Loop I RHR Pump and Heat Exchanger in service by REPERFORMING Step 8.5 [7] for the second pump. [8] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:

Amber breaker spring charged light on,
Closing spring target indicates charged.

[10] PLACE RHR Pump and Heat Exchanger B(D) in service as follows: [10.2] ESTABLISH RHRSW flow by one of the following methods: [10.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23. OR [10.2.2] THROTTLE OPEN RHR HX 2B(2D) RHRSW OUTLET VLV, 2-FCV-23-46(52), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 2-FI-23-48(54), RHR HX 2B(2D) RHRSW FLOW. [10.3] VERIFY CLOSED RUR SYS II LPCI 1NBD INJECT VALVE, 2-FCV-74-67. [10.4] IF NO RHR PUMP (lB or 1D) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73. [10.5] VERIFY CLOSED RHR SYS II SUPPR CHBR SPRAY VALVE, 2-FCV-74-72. [10.6] VERIFY CLOSED RHR SYS II DW SPRAY OUTBD VLV, 2-FCV-74-74. [10.7] VERIFY OPEN RHR SYS II SUPPR CRBR/POOL ISOL VLV, 2-FCV-74-7 1. 27

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP initiates Pool Cooling as directed [10.8] IF desired to operate without the Drywell DP Compressor, THEN: [10.9] START RUR PUMP 2B(2D) using 2-HS-74-28A(39A). [10.10] THROTTLE RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73, to maintain RHR flow within limits, as indicated on RHR SYS II CTMT FLOW, 2-FI-74-70. RI-IR Pumps in 1 2 Operation Loop Flow 7,000to <13,000 gpm& 10,000 gpm & Blue Blue light light illuminated illuminated [11] iF desired to RAISE Suppression Pool Cooling flow and only one Loop II pump is in service, THEN PLACE the second Loop II RHR Pump AND Heat Exchanger in service. REPERFORM Step 8.5[l0] for the second pump. [12] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:

Amber breaker spring charged light on,
Closing spring target indicates charged.

SRO Tech Spec 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be.? -6.25 inches with and -7.25 inches without differential pressure control and -1.0 inches. APPLICABILITY: MODES 1,2, and 3. Condition A: Suppression pool water level not within limits. Required Action A. 1: Restore suppression pool water level to within limits. Completion Time: 2 hours. Note AS the SRV remains open adding inventory to suppression pool, pool level spec will be appropriate. 28

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking SRO Enter EOI-2 on Suppression Pool Level SRO PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue SPIT MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue PC/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, does not continue , DW/T Monitor and control Drywell temperature below I 60F using available Drywell cooling Can Drywell Temperature be maintained below 160F, YES SP/L MOMTOR and CONTROL suppr p1 lvl between -1 in. and -6 in. (APPX 18) Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES SRO Direct Appendix 18, Suppression Pool Water Inventory Removal And Makeup BOP Calls for Operator to perform field action of Appendix 18 29

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Calls for Operator to perform field action of Appendix 18

3. IF Directed by SRO, THEN REMOVE water from Suppression Pool as follows:

a. DISPATCH personnel to perform the following (Unit 2 RB, El 519 ft, Torus Area):

1) VERIFY OPEN 2-SHV-074-0786A(B), RHR DR PUMP 2A(2B)

DISCH TO MN CNDR!RW SOy.

2) OPEN the following valves:

2-SHV-074-0564A(B), RHR DR PMP 2A(2B) SEAL WATER SUPPLY SOV 2-SHV-074-0529A(B), RHE. DRAiN PUMP A(B) SHUTOFF VLV.

3) UNLOCK and OPEN 2-SHV-074-0765A(B), RHR DR PUMP 2A(2B)

DISCH SOy.

4) NOTIFY Unit Operator that RHR Drain Pump 2A(2B) is lined up to remove water from Suppression Pool.

5) REMAIN at torus area UNTIL Unit 2 Operator directs starting of RHR Drain Pump 2A(2B).

b. IF Main Condenser is desired drain path, THEN OPEN 2-FCV-74-62, RHR MAIN CNDR FLUSH VALVE.

c. IF Radwaste is desired drain path, THEN PERFORM the following:

1) ESTABLISH communications with Radwaste.

2) OPEN 2-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE

d. NOTIFY personnel in Unit 2 RB, El 519 fi, Torus Area to start RHR Drain Pump 2A(2B).

e. THROTTLE 2-FCV-74-l08, RHR DR PUMP 2A/B DISCH HDR VALVE, as necessary.

Driver When dispatched to remove water from the suppression pool, wait 10 minutes and call and report aligned step 4 above, when the operator calls you back to start the RNR Drain Pump Shift F6 for bat appl 8rhra and Shift F7 for bat appi 8rhrb 30

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Appendix 18

4. WFIEN Suppression Pool level reaches -5.5 in., THEN SECURE RHR Drain System as follows:

a. DISPATCH personnel to STOP the Drain System as follows (Unit 2 RB, El 519 ft, Torus Area):

1) STOP RHR Drain Pump 2A(2B).

2) CLOSE the following valves:

2-SHV-074-0564A(B), RHR. DR PMP 2A(2B) SEAL WATER SUPPLY SOV
2-SHV-074-0529A(B), RIIR DRAiN PUMP A(B) SHUTOFF VLV.

3) CLOSE and LOCK 2-SHV-074-0765A(B), R}IR DR PUMP 2A(2B)

DISCH SOy.

b. CLOSE 2-FCV-74-108, RHR DR PUMP 2A!B DISCH HDR VALVE.

c. VERIFY CLOSED 2-FCV-74-62, RI-JR MAIN CNDR FLUSH VALVE.

d. VERIFY CLOSED 2-FCV-74-63, RFLR RADWASTE SYS FLUSH VALVE.

Driver When directed by NRC for VFD Cooling Pump trip, Preference Key F6 31

NRC Scenario 5 Simulator Event Guide: Event 5 Component: VFD Cooling Water Pump 2A trip Driver When directed by NRC for VFD Cooling Pump trip, Preference Key F6 ATC Respond to the following alarms, 4A-12, 4A-28 and 4A-32 ATC Report Trip of Recirc Drive 2A Cooling Pump 2A1, and failure of standby pump to start Alarm 4A-12, RECIRC DRiVE 2A COOLANT FLOW LOW Automatic Action Standby RECIRC DRIVE cooling water pump will auto start. A. VERIFY RECIRC DRIVE cooling water pump running. B. DISPATCH personnel to the RECIRC DRIVE to check the operation of the Recirc Drive cooling water system. Alarm 4A.28, RECIRC DRIVE 2A PROCESS ALARM A. IF 2-XA-55-4B Window 28 is also in alarm, THEN (N/A) B. Refer to ICS screen VFDAAL and determine cause of alarm Alarm 4A-32, RECJRC DRIVE 2A DRIVE ALARM A. REFER TO ICS Group Display GD @VFDADA and DETERMINE cause of alarm. B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system. ATC Start Standby Recirc Drive 2A Cooling Pump 2A2, dispatches personnel to investigate Wait 4 minutes after dispatched, THEN report tripped VFD Pump 2A1 is hot to the

touch, internal bkr closed, 480 volt bkr tripped (480 V SD BD 2A 5C).

Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 32

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A lnboard and Outboard seal failure Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 ATC Respond to alarm 4A-25, RECIRC PUMP A NO. 1 SEAL LEAKAGE ABN A. DETERMINE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 2-9-4 or ICS.

Plugging of No. 1 RO No. 2 seal cavity pressure indicator drops toward zero.
Plugging of No. 2 RO No. 2 seal pressure approaches no. 1 seal pressure.
Failure of No. 1 seal No. 2 seal pressure is greater than 50% of the pressure of No. 1.
Failure of No. 2 seal no. 2 seal pressure is less than 50% of the No. 1 seal.

B. RECORD pump seal parameters hourly on Attachment 1, ATC Report of failure of number 1 seal or inner seal Respond to alarm 4A-18, RECIRC PUMP A NO.2 SEAL LEAKAGE HiGH A. COMPARE No. 2 cavity pressure indicator (2-PI-68-63A) to No. 1 cavity pressure indicator (2-PI-68-64A). No. 2 seal degradation is indicated if the pressure at No. 2 seal is less than 50% of the pressure at No. 1 seal. ATC .) Reports the second seal is failed both pressure indicators trending toward zero psig. C. IF dual seal failure is indicated, THEN

1. SHUTDOWN Recirc Pump 2A by depressing RECIRC DRiVE 2A SHUTDOWN, 2-HS-96-19.

2. VERIFY TRIPPED, RECIRC DRiVE 2A NORMAL FEEDER, 2-HS-57-17.

3. VERIFY TRIPPED, RECIRC DRIVE 2A ALTERNATE FEEDER, 2-HS-57-15.

4. CLOSE Recirculation Pump 2A suction valve.

5. CLOSE Recirculation Pump 2A discharge valve.

6. REFER TO 2-AOl-68-1A or 2-AOl-68-1B AND 2-01-68.

7. DISPATCH_personnel_to_secure_Recirculation_Pump_2A_seal_water.

ATC Trips RR Pump 2A and closes suction and discharge valves Reports rising Drywell Pressure, reports DW Pressure stable once valves are closed SRO Enters 2-A0I-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 33

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure SRO Enters 2-AOI-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 4.2 Subsequent Actions [1] II? both Recirc Pumps are tripped in modes 1 or 2, ThEN (N/A), [2] IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve. [3] IF Region I or II of the Power to Flow Map is entered, THEN IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. Refer to O-TI-464, Reactivity Control Plan Development and Implementation. [4] RAISE core flow to greater than 45%. REFER TO 2-01-68. [5] INSERT control rods to exit regions if not afready exited. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation. [6] MAINTAIN operating Recirc pump flow less than 46,600 gpm. Refer to 2-01-68. [7] WHEN plant conditions allow, THEN, MAINTAIN operating jet pump loop flow greater than_41_x_106_ibm/hr_(2-FI-68-46_or 2-FI-68-48). SRO Direct inserting control rods JAW Urgent Load Reduction and Rod Shove Sheets ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Driver When dispatched to isolate seal water waitS minutes and then mrf rdO3 close and report closed 34

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Inserts all of the following Control Rods to lower rod line to < 95%: Control Rods 22-31, 30-39, 38-31, 30-23 from 08 to 00 Control Rods 22-3 9, 3 8-39, 3 8-23, 22-23 from 16 to 00 Control Rod 30-31 from 22 to 00 Control Rods 14-31, 30-47, 46-3 1, 30-15 from 48 to 00 ATC Raise Speed of RR Pump B until core flow is 46 to 50% and ensure RR Pump B drive flow is below 46,600 pm Report Exit from Region II of Power to Flow Map SRO Tech Spec 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single ioop operation limits specified in the COLR;

b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR),

single loop operation limits specified in the COLR; C. LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power High), Allowable Value of Table 3.3.1.1-1 is reset for single ioop operation; APPLICABILITY: MODES 1 and 2. Condition A: Requirements of the LCO not met. Required Action A. 1: Satisfy the requirements of the LCO. Completion Time: 24 hours. Driver When directed by NRC, Preference Key F9, Level Instruments Fail high. When mode switch is out of run or NOT in run Preference Key Fl 0 35

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI ATC Report Trip of Main Turbine and RFPTs and Reactor Scram ATC 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AIS3A and 2-HS-99-5AJS3B, on Panel 2-9-5. [2] if scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA [3] REFUEL MODE ONE ROD PERMISSIVE light check: [3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S 1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-X1-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, is not illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) Step is NA [4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S 1, in SHUTDOWN. [5] REPORT the following status to the US: F)

Reactor Scram
Mode Switch is in Shutdown
All rods in or rods out
Reactor Level and trend (recovering or lowering).
Reactor pressure and trend
MSJV position (Open or Closed)
Power level

[1] ANNOUNCE Reactor SCRAM over PA system. [3] DRiVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit. [3.1] DOWNRANGE IRMs as necessary to follow power as it lowers. [5] MOMTOR and CONTROL Reactor Water Level between +2 and +51 , or as directed by US, as follows: ATC/BOP Open RCIC Steam Supply Valve to start RCIC for Level Control, RCIC has received an Auto Start signal but the Steam Supply Valve failed to Open. Driver When mode switch is in out of run Preference Key FlO 36

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI SRO Enters EO1-1 on RPV Water Level SRO EOI-1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO IF Emergency Depressurization is or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO-Is any MSRV Cycling? No IF Steam cooling is required? NO

( IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO IF Boron injection is required? NO Stabilize RPV pressure below 1073 psig with the main turbine bypass valves (APPX 8B) SRO Direct a pressure band, may direct a cooldown lAW Appendix 8B ATC/BOP Maintain directed pressure with Bypass Valves lAW Appendix 8B, Reopening MSIVs I Bypass Valve Operation 37

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI ATC/BOP Maintain directed pressure with Bypass Valves JAW Appendix 8B, Reopening MS1Vs I Bypass Valve Operation

1. IF pressure control with bypass valves is desired and MSIVs are open, TFIEN proceed to step 10.

10. Verify Condenser Vacuum is greater than 7

11. IF manual opening of Bypass Valves is desired, THEN perform the following step:

a. Depress the Bypass Valve Opening Jack Raise Pushbutton, 2-HS 130B to slowly open the Bypass Valves.

b. Adjust BPV Positio0n as necessary by using the raise, 2-HS-47-130B and Lower 2-HS-47-130A pushbuttons to maintain desired cooldown rate.

12. IF EHC Auto Cooldown is desired, THEN perform the following steps:

a. Verify EHC is in Pressure Control using 2-HS-47-204

b. Verify Bypass Valve Demand is set at ZERO

c. On the EHC Work Station on Panel 2-9-7:

1) Select Main Menu from the toolbar at bottom of the screen.

2) Select Log In on Display Screen and Enter OPS for name and OPS for password.

3) Select Auto Cooldown from list of function on the screen.

d. On the Auto Cooldown Display Screen

1) Check the following are displayed.

Turbine Tripped or All Valves Closed indicates reset
RX Press Ctrl indicates reset

2) Select the block above the FINAL PRESSURE TARGET

3) Enter the desired pressure using the display screen or keyboard

4) Select OK

5) Depress the START button

6) When Are You Sure You Want to Initiate Auto Cooldown? appears, Select YES

7) Check the following:

EHC PRESSURE SETPOINT, 2-PI-47-162, is lowering
EHC AUTO COOLDOWN displays IN PROCESS 38

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs_Verification_of PCIS_isolations. ATC/BOP Verifies PCIS isolations. SRO IF It has NOT been determined that the reactor will remain subcritical without boron under all condition THEN EXIT RC/L NO - RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO - Restore and Maintain RPV water level between +2 inches and +51 inches with RCIC (APPX 5C) ATC/BOP RCIC failed to auto start, Opens RCIC Steam Supply Valve and verifies RCIC operation.

1. VERIFY 2-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO withsetpointat600gpm.

7. OPEN 2-FCV-71-8, RCIC TURBINE STEAM SUPPLY VLV, to start RCIC Turbine.

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at 600 gpm.

c. 2-FCV-71-40, RCIC Testable Check Vlv, opens by observing 2-ZI-71-40A, DISC POSiTION, red light illuminated.

d. 2-FCV-71-34, RCIC PUMP M1N FLOW VALVE, closes as flow rises above 120 gpm.

39

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Report rising Drywell Pressure and Temperature ATC/BOP Report loss of 480V SD BD 2A and 480V RMOV BD 2A ATC/BOP Dispatch personnel to investigate loss of Board SRO Re-Enter EOI-2 on High DW Pressure and Temperature ATC/BOP IF RHR Loop 1 was in Pool Cooling for leaking SRV, then operators report that RHR Loop 1 remains in Pool cooling. NOTE RHR Loop 1 has lost power to almost all valves but NO valves reposition on board loss SRO EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 160F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOl- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) Driver When dispatched for Board loss, wait 4 minutes and report overcurrent trip of supply breaker on 480V SD BD 2A. If requested to energize 480V RMOV BD 2A from alternate supply, wait 3 minutes and report that unable to restore power to Board 40

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 17C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue 41

NRC Scenario 5 Event Guide: EventS Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywefi Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr pi cooling (APPX 17A), Pool Temp below 95° WFIEN suppr pl temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18) Can suppr p1 ivl be maintained above -6 in., YES Can suppr p1 lvi be maintained below -1 in., YES SRO Direct Suppression Chamber Sprays and Drywell Sprays on RHR Loop II ONLY 42

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF Adequate core cooling is assured, OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLAC ING 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN CONTINUE in this procedure at Step 7.

4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.

5. INITIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR Directed by SRO, THEN PLACE keylock switch 2-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-67, RHR SYS II INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired RHR System II pump(s) for Suppression Chamber Spray.

f. VERIFY OPEN 2-FCV-74-71, RHR SYS II SUPPR CHBRIPOOL ISOL VLV.

g. OPEN 2-FCV-74-72, RFIR SYS II SUPPR CHBR SPRAY VALVE.

ATC/BOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 43

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

h. IF RFIR System II is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k.

i. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II MUST FLOW VALVE.

j. RAISE system flow by placing the second RHR System II pump in service as necessary.

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY R}IRSW pump supplying desired RHR Heat Exchanger(s).

m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:

2-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
2-FCV-23-52, RUR HX 3D RHRSW OUTLET VLV.

n. NOTIFY Chemistry that RHRSW is aligned to in-service RI{R Heat Exchangers.

ATC/BOP Aligns RI-JR Loop II Pumps in Suppression Chamber Sprays 44

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17B, RHR System Operation Drywell Sprays I. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

2. IF Adequate core cooling is assured, OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.

5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.

6. INITIATE Drywell Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 2-XS-74-130, RI{R SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II LPCI OUTBD iNJECT VALVE.

e. VERIFY OPERATING the desired System II RHR pump(s) for Drywell Spray.

f. OPEN the following valves:

2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
2-FCV-74-75, RHR SYS II DW SPRAY INBD VLV.

ATC/BOP Aligns RHR Loop II Pumps in Drywell Sprays 45

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP 2-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

g. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II MIN FLOW VALVE.

h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System II RHR Pump in service.

i. MONITOR RHR Pump NPSH using Attachment 2.

j. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

k. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

2-FCV-23-46, RHR FIX 3B RHRSW OUTLET VLV
2-FCV-23-52, RHR. HX 3D RHRSW OUTLET VLV.

1. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

7. WHEN EITHER of the following exists:

Before drywell pressure drops below 0 psig, OR
Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:

a. VERIFY CLOSED the following valves:

2-FCV-74-100, RHR SYS I U-2 DISCH XTLE
2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
2-FCV-74-75, RHR. SYS II DW SPRAY 1NBD VLV.

b. VERIFY OPEN 2-FCV-74-30, RHR SYSTEM II MN FLOW VALVE.

c. IF RHR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.

d. STOP RHR Pumps.

ATC/BOP Aligns RHR Loop II Pumps in Drywell Sprays 46

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Report lowermg RPV water level unable to maintain with RCIC SRO EOI-1 Reactor Level RPV water level drops below -120 inches OR The ADS timer has initiated NO IF RPV water level CANNOT be restored and maintained between +2 and +51 inches, THEN Restore and maintain RPV water level above -162 inches Augment RPV water level control as necessary with any of the following SRO Directs additional level control systems: SLC (boron tank) APPX-7B CRD APPX-5B ATC/BOP Place SLC and an additional CR1) Pump in service JAW Appendix 7B and 5B SRO EOI- 1 Reactor Level Can RPV water level be restored and maintained above -162 inches NO SRO Announces entry to EOI-C- 1 Alternate Level Control RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO - IF RPV water level can be restored and maintained above -162 inches NO Inhibit ADS ATCIBOP Inhibits ADS SRO Restore and maintain RPV water level above -162 inches using any of the following: Condensate APPX 6A LPCI System I APPX 6B LPCI System II APPX 6C Core Spray System II APPX 6E SRO Directs 2 or more of the above systems lined up for injection ATCIBOP Aligns the directed systems for Injection 47

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Aligns CRD and SLC JAW Appendix SB and 7B ATC CRD Appendix 5B

2. IF BOTH of the following exist:

CRD is NOT required for rod insertion, AND

                                . Maximum injection flow is required, THEN LINE UP ALL available CRD pumps to the RPV as follows:

a. IF CRD Pump 2A is available, THEN VERIFY RIJNMNG CRD Pump 2A.

b. IF CRD Pump lB is available, TFIEN PERFORM the following:

1) NOTIFY Unit 1 Operator to verify closed 1-FCV-85-8, CRD PUMP B DISCHARGE VALVE (Unit 1, Panel 9-5).

2) START CRD Pump lB.

3) OPEN 2-FCV-85-8, CRD PUMP lB DISCH TO U2.

c. OPEN the following valves to increase CRIJ flow to the RPV:

2-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV
2-PCV-85-27, CRD CLG WATER PRESS CONTROL VLV
2-FCV-85-50, CRD EXH RTN LiNE SHUTOFF VALVE.

d. ADJUST 2-FIC-85-1 1, CRD SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WHILE maintaining 2-Pl-85-13A, CRD ACCUM CHG WTR HDR PRESS, above 1450 psig, if possible.

e. IF Additional flow is necessary to prevent or mitigate core damage, THEN DISPATCH personnel to fully open the following valves as required:

2-THV-085-0527, PUMP DISCH THROTTLING (RB NE, el 565)
2-BYV-085-0551, PUMP TEST BYPASS (RB NE, el 565).

Driver When called as unit one operator FCV-85-8, CRD PUMP B DISCHARGE VALV is E closed 48

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Aligns CRD and SLC lAW Appendix 5B and 7B ATC SLC Appendix 7B

2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step 10 to inject SLC Boron Tank to RPV.

10. UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B (Panel 9-5).

11. CHECK SLC injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch.

                             . Squib valves fire, as indicated by SQUIB VALVE A and B CONTiNUITY blue lights extinguished,

SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (2-XA-55-5B, Window 20).

                             . 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.
                             . System flow, as indicated by 2-IL-63-1 1, SLC FLOW, red light illuminated,
                             . SLC INJECTION FLOW TO REACTOR Annunciator in alarm (2-XA                                   5B, Window 14).

12. IF Proper system operation CANNOT be verified, THEN RETURN to Step 10 and START other SLC pump.

49

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC Aligns Condensate JAW Appendix 6A

1. VERIFY CLOSED the following feedwater heater return valves:

                                . 2-FCV-3-71, HP HTR 2A1 LONG CYCLE TO CNDR
                                . 2-FCV-3-72, HP HTR 2B1 LONG CYCLE TO CNDR
                               . 2-FCV-3-73, HP HTR 2C1 LONG CYCLE TO CNDR.

2. VERIFY CLOSED the following RFP discharge valves:

                               . 2-FCV-3-19, RFP 2A DISCHARGE VALVE
                               . 2-FCV-3-12, RFP 2B DISCHARGE VALVE
                               . 2-FCV-3-5, RFP 2C DISCHARGE VALVE.

3. VERIFY OPEN the following drain cooler inlet valves:

                               . 2-FCV-2-72, DRAIN COOLER 2A5 CNDS INLET ISOL VLV
                               . 2-FCV-2-84, DRAIN COOLER 2B5 CNDS INLET ISOL VLV
                               . 2-FCV-2-96, DRAIN COOLER 2C5 CNDS INLET ISOL VLV.

4. VERIFY OPEN the following heater outlet valves:

                              . 2-FCV-2-124, LP HEATER 2A3 CNDS OUTL ISOL VLV
                              .*   2-FCV-2-125, LP HEATER 2B3 CNDS OUTL ISOL VLV 2-FCV-2-126, LP HEATER 2C3 CNDS OUTL ISOL VLV.

5. VERIFY OPEN the following heater isolation valves:

                              . 2-FCV-3-38, NP HTR 2A2 FW INLET ISOL VLV
                              . 2-FCV-3-31, HP FITR 2B2 FW INLET ISOL VLV
                              . 2-FCV-3-24, HP HTR 2C2 FW INLET ISOL VLV
                              . 2-FCV-3-75, HP HTR 2A1 FW OUTLET ISOL VLV
                              . 2-FCV-3-76, HP HTR 2B1 FW OUTLET ISOL VLV
                              . 2-FCV-3-77, HP HTR 2C1 FW OUTLET ISOL VLV.

6. VERIFY OPEN the following REP suction valves:

                              . 2-FCV-2-83, REP 2A SUCTION VALVE
                              . 2-FCV-2-95, RFP 2B SUCTION VALVE
                              . 2-FCV-2-108, RFP 2C SUCTION VALVE.

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

9. ADJUST 2-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 2-9-5).

10. VERIFY RFW flow to REV.

50

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA SRO EOI-C-1 Alternate Level Control SRO Can 2 or more Condensate, LPCI or Core Spray injection subsystems be lined up YES-When RPV Water level drops to -162 inches Proceeds at TAF or -1 62 inches Is any Condensate, LPCI or Core Spray injection subsystems lined up for injection with at least one pump running YES-Is any RPV injection source lined up with at least one pump running YES BEFORE RPV water level drops to -180 inches CONTINUE Continues - Emergency Depressurization is required Inject into the RPV with any available sources SRO Enters EOI-C-2 Emergency Depressurization Will the reactor remain subcritical without boron under all conditions YES

Is DW pressure above 2.4 psig YES Prevent injection from only those Core Spray and LPCI pumps not required NO Is suppression pooi level above 5.5 feet YES Open all ADS Valves Directs ADS valves open ATC/BOP Opens all 6 ADS valves, reports all ADS valves open When pressure is below the shutoff head of the available injection systems direct injection SRO to restore level to +2 to +51 inches ATC/BOP Injects with available systems to restore level SRO Emergency Classification EPIP-1 1.1-SI Reactor water level can NOT be maintained above -162 inches. (TAF) 51

NRC Scenario 5 Simulator Event Guide: Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, lAW BOP Appendix 6B, 6C and 6E Although most valve power is lost for RHR Loop I, injection is still available, the pumps have power, the Outboard Injection Valve does not have power but is normally open and the only valve with power is the Inboard Injection Valve which can be opened. NOTE If RHR Loop I is used the only to control injection is to turn pumps on and off. In addition if it was aligned for Pool Cooling those valves will still be open, so the injection pressure to the vessel will be much lower. BOP Appendix 6B

1. IF Adequate core cooling is assured, ANT) It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-155A, LPCI SYS I OUTBD 1NJ VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN 2-FCV-74-l, RHR PUMP 2A SUPPR POOL SUCT VLV.

3. VERIFY OPEN 2-FCV-74-12, RHR PUMP 2C SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

                                . 2-FCV-74-61, RHR SYS I DW SPRAY LNBD VLV
                                . 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV
                                    . 2-FCV-74-57, RHR SYS I SUPPR CHBR/POOL ISOL VLV
                                   . 2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE
                                   . 2-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 2A and/or 2C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-53, RHR SYS I LPCI INBD iNJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-68-79, RECIRC PUMP 2B DISCHARGE VALVE.

8. THROTTLE 2-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE, as necessary to control injection.

Can inject but cannot throttle 2-FCV-74-52 and will have to open 2-FCV-74-53 with the BOP handswitch 52

NRC Scenario 5 Simulator Event Guide: (* Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop BOP I, JAW Appendix 6B, 6C and 6E BOP Appendix 6C

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-7 4- 55B, LPCI SYS II OUTBD Th4J VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN 2-FCV-74-24, RHR PUMP 2B SUPPR POOL SUCT VLV.

3. VERIFY OPEN 2-FCV-74-35, RHR PUMP 2D SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

                                 . 2-FCV-74-75, RHR SYS II DW SPRAY INBD VLV

2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
2-FCV-74-71, RHR SYS II SUPPR CFII3R/POOL ISOL VLV
2-FCV-74-72, RRR SYS II SUPPR CHBR SPRAY VALV E
2-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 2B andlor 2D running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV 67, RHR SYS II LPCI 111BD INJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-.68-3, RECIRC PUMP 2B DISCHARGE VALVE.

8. THROTTLE 2-FCV74-66, RI{R SYS II LPCI OUTBD INJECT VALVE, as necessary to control injection.

BOP Will have to open 2-FCV-74-67 with the handswitch 53

NRC Scenario 5 Simulator Event Guide: Event 9 Component: RFIR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, JAW BOP Appendix 6B, 6C and 6E BOP Appendix 6E

1. VERIFY OPEN the following valves:

                              . 2-FCV-75-30, CORE SPRAY PUMP 2B SUPPR POOL SUCT VLV
                              . 2-FCV-75-39, CORE SPRAY PUMP 2D SUPPR POOL SUCT VLV
                              . 2-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.

2. VERIFY CLOSED 2-FCV-75-50, CORE SPRAY SYS II TEST VALVE.

3. VERIFY CS Pump 2B and/or 2D running.

4. WHEN RPV pressure is below 450 psig, THEN ThROTTLE 2-FCV-75-53, CORE SPRAY SYS II INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

Coordinate RPV Level Control to restore and maintain Level +2 to +51 inches. BOP/ATC Condensate and Core Spray will restore and maintain level. When RPV pressure is low enough Condensate System will maintain directed level band. BOP Will have to open 2-FCV-75-53 with the handswitch 54

NRC Scenario 5 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation. Temporary DGs are NOT provided. Operations/Maintenance for the Shift: Lower Power with flow to 91% for Main Turbine Valve Testing. Unit 1 and 3 are at 100% Power Unusual Conditions/Problem Areas: Severe Thunderstorms are forecast for today, currently no watches or warnings are in effect. 55

Facility: Browns Ferry NPP Scenario No.: NRC 6- Op-Test No.: Examiners:__________________ Operators: SRO:_________ ATC: BOP: Initial Conditions: 1.3% power, operating in 2-GOI-100-1A Section 5.4 steps 63.3 and 65. Turnover: Warm RFPT B lAW 2-01-3, section 5.6 and then Continue to pull rods for Mode Change. Event Maif. No. Event Event Description No. Type* N-BOP 1 Warm RFPT B lAW 2-01-3, section 5.6 N-SRO R-ATC 2 Raise power with Control Rods R-SRO C-ATC Control Rod will difficult to withdraw, control rod at position 3 rd05r3435 C-SRO other than 00, and then control rod stuck TS-SRO C-BOP Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto 4 ed07a C-SRO start 1-BOP 5 rcOl One level instrument fails and RCIC inadvertently starts TS-SRO C-ATC 6 th03b Reactor Recirculation Pump 2B trip TS-SRO 7 batch M-ALL SSI Fire 25-1 8 adOim C SRV fails open 9 hpO3 I HPCI Flow controller will not operate in Auto

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

Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed Path A Vent Flow Controller, 2-FIC-84-20, in MANUAL and 0 SCFM, at Panel 2-9-55.

1. Safety Significance:

Maintaining adequate RHR Pump NPSH.

2. Cues:

Procedural compliance. Containment Pressure indication.

3. Measured by:

Observation 2-FIC-84-20 in manual and set at 0 SCFM. Observation 2-FCV-84-20 closed.

4. Feedback:

Containment Pressure trend. No flow through A vent path. Within 10 minutes of recorded time in SSI an Operator has initiated a controlled 100°F per hour cooldown rate using HPCI and relief valves as required.

1. Safety Significance:

Prevent Drywell Temperature from exceeding design basis temperature.

2. Cues:

Procedural compliance. Reactor Pressure indication.

3. Measured by:

Observation HPCI in Pressure Control Mode. Observation SRVs opened to lower pressure.

4. Feedback:

Reactor Pressure trend.

Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed the following switches in Test/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Test/Inhibit, 2-HS-75-59 ANT) ECCS SYS II HI DW PRESS Test/Inhibit, 2-HS-75-60.

1. Safety Significance:

Prevent CAS initiation due to actual high Drywell Pressure, and minimize the number of subsequent additional actions (to secure/realign both credited and non-credited pumps).

2. Cues:

Procedural compliance. No AUTO initiation of ECCS when Drywell Pressure exceeds 2.45 psig.

3. Measured by:

Observation 2-HS-75-59 and 60 in Test/Inhibit. Observation No AUTO initiation on high drywell pressure.

4. Feedback:

ECCS Pumps green lights ON and Red Lights Off.

EVENTS

1. BOP Operator warms RFPT B lAW 2-01-3 Feedwater System, section 5.6.

2. ATC Continues Power ascension with control rods.

3. During power ascension Control Rod 34-3 5 will fail to withdraw. The crew will respond lAW 2-01-85. Once Drive water pressure is at 350 psig or greater the control rod will triple notch to position 14 which is one notch beyond the banked position of 12. The Unit Supervisor should enter 2-AOI-85-7 for a mispositioned control rod. All attempts to insert the control rod to the correct position will fail. The control rod will be declared stuck and the SRO will enter Tech Specs and determine TS 3.1.3 condition A.

4. A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip due to loss of power and the standby pump will not auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.

5. Level transmitter 58D will fail to less than -45 inches. This failure will result in a RCIC inadvertent initiation. The BOP Operator will respond JAW ARPs. BOP Operator will verify that level is in normal band and secure RCIC. The SRO will evaluate Technical Specification 3.5.3 Condition A, 3.3.4.2 Condition A, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition A and B, and 3.8.1 Condition D.

6. Reactor Recirculation Pump 2B will trip, the crew will respond lAW 2-A0I-68-1A. The SRO will enter Tech Specs and determine TS 3.4.1 condition A.

7. The crew will respond to a fire and enter 0-AOI-26-1 and SSI 25-1, Intake Pumping Station Pump El. 550, Cable Tunnel to Fire Door 440, RHRSW Room B, RHRSW Pump Room D.

The SRO will also enter EOI-1 and 2 and perform actions that do not conflict with the SSI guidance.

8. An SRV will fail open which causing an increased cooldown rate, the crew will respond lAW 2-AOl-i-i and take actions to close the SRV.

9. Shortly after entering the SSI the crew will commence a controlled cooldown lAW the SSI utilizing HPCI and SRVs, the HPCI flow controller will fail in Auto but will operate in manual.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Reactor Level is maintained Controlled Cooldown in progress SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 6 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No)

Scenario Tasks TASK NUMBER 1Q RQ Warm RFPT 2B lAW 2-01-3 RO U-003-N0-23 259001 A4.02 3.9 3.7 Raise Power with Control Rods RO U-085-N0-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to withdraw from a position other than 00 ROU-085-N0-19 201001 A4.04 3.1 3.1 SRO S-000-AD-3 1 Control Rod Mispositioned RO U-085-AB-07 201002 A2.02 3.2 3.3 SRO S-085-AB-07 Reactor Recirculation Pump Trip RO U-068-AB-1 20200 1A2.03 3.6 3.7 SRO S-068-AB-1 Loss of 480V Unit BD 2A RO U-57B-AL-06 226001 A2.04 3.8 4.2 SRO S-57B-N0-07 RCIC Inadvertent Start ROU-071-N0-5 217000 A2.01 3.8 3.7 SRO S-000-AD-27 SSI FIRE RO U-000-EM-85 600000 AA2.16 3.0 3.5 RO U-000-SS-30 RO U-000-N0-32 SRO S-000-EM-30 SRO S-000-SS-30 SRO S-000-SS-3 1

C

-3

NRC Scenario 6 Facility: Browns Ferry NPP Scenario No.: NRC 6- Op-Test No.: Examiners:_____________________ Operators: SRO:______________________ ATC:_______________ BOP:________________ Initial Conditions: 1.3% power, operating in 2-GOI-100-1A Section 5.4 steps 63.3 and 65. Turnover: Warm RFPT B JAW 2-01-3, section 5.6 and then Continue to pull rods for Mode Change. Event Maif. No. Event Type* Event Description No. N-BOP 1 Warm RFPT B lAW 2-01-3, section 5.6 N-SRO R-ATC 2 Raise power with Control Rods R-SRO C-ATC Control Rod will difficult to withdraw, control rod at position other 3 rd05r3435 C-SRO than 00, and then control rod stuck TS-SRO C-BOP 4 ed07a Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto start C-SRO 5 rcOl One level instrument fails and RCIC inadvertently starts C-ATC . 6 tho3b Reactor Recirculation Pump 2B trip TS-SRO 7 batch M-ALL SSI Fire 25-1 8 adOim C SRV fails open 9 hpO3 I HPCI Flow controller will not operate in Auto

      *   (N)ormal,   (R)eactivity, (I)nstrument,     (C)omponent,   (M)aj or

NRC Scenario 6 Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed Path A Vent Flow Controller, 2-FIC-84-20, in MANUAL and 0 SCFM, at Panel 2-9-55.

1. Safety Significance:

Maintaining adequate RI-IR Pump NPSH.

2. Cues:

Procedural compliance. Containment Pressure indication.

3. Measured by:

Observation 2-FIC-84-20 in manual and set at 0 SCFM. Observation 2-FCV-84-20 closed.

4. Feedback:

Containment Pressure trend. No flow through A vent path. Within 10 minutes of recorded time in SSI an Operator has initiated a controlled 100°F per hour cooldown rate using HPCI and relief valves as required.

1. Safety Significance:

Prevent Drywell Temperature from exceeding design basis temperature.

2. Cues:

Procedural compliance. Reactor Pressure indication.

3. Measured by:

Observation HPCI in Pressure Control Mode. Observation SRVs opened to lower pressure.

4. Feedback:

Reactor Pressure trend.

NRC Scenario 6 Critical Tasks Three Within 10 minutes of recorded time in SSI an Operator has placed the following switches in Test/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Testflnhibit, 2-HS-75-59 AND ECCS SYS II HI DW PRESS Test/Inhibit, 2-HS-75-60.

1. Safety Significance:

Prevent CAS initiation due to actual high Drywell Pressure, and minimize the number of subsequent additional actions (to secure/realign both credited and non-credited pumps).

2. Cues:

Procedural compliance. No AUTO initiation of ECCS when Drywell Pressure exceeds 2.45 psig.

3. Measured by:

Observation 2-HS-75-59 and 60 in Test/Inhibit. Observation No AUTO initiation on high drywell pressure.

4. Feedback:

ECCS Pumps green lights ON and Red Lights Off.

NRC Scenario 6 EVENTS

1. BOP Operator warms RFPT B lAW 2-01-3 Feedwater System, section 5.6.

2. ATC Continues Power ascension with control rods.

3. During power ascension Control Rod 34-3 5 will fail to withdraw. The crew will respond lAW 2 85. Once Drive water pressure is at 350 psig or greater the control rod will triple notch to position 14 which is one notch beyond the banked position of 12. The Unit Supervisor should enter 2-AOI-85-7 for a mispositioned control rod. All attempts to insert the control rod to the correct position will fail.

The control rod will be declared stuck and the SRO will enter Tech Specs and determine TS 3.1.3 condition A.

4. A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip due to loss of power and the standby pump will not auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.

5. Level transmitter 58D will fail to less than -45 inches. This failure will result in a RCIC inadvertent initiation. The BOP Operator will respond lAW ARPs. BOP Operator will verify that level is in normal band and secure RCIC. The SRO will evaluate Technical Specification 3.5.3 Condition A, 3.3.4.2 Condition A, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition A and B, and 3.8.1 Condition D.

6. Reactor Recirculation Pump 2B will trip, the crew will respond lAW 2-AOI-68-1A. The SRO will enter Tech Specs and determine TS 3.4.1 condition A.

7. The crew will respond to a fire and enter 0-AOI-26-1 and SSI 25-1, Intake Pumping Station Pump El. 550, Cable Tunnel to Fire Door 440, RHRSW Room B, RHRSW Pump Room D. The SRO will also enter EOI-l and 2 and perform actions that do not conflict with the SSI guidance.

8. An SRV will fail open which causing an increased cooldown rate, the crew will respond lAW 2-AOl-i-i and take actions to close the SRV.

9. Shortly after entering the SSI the crew will commence a controlled cooldown lAW the SSI utilizing HPCI and SRVs, the HPCI flow controller will fail in Auto but will operate in manual.

4

NRC Scenario 6 C Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Reactor Level is maintained Controlled Cooldown in progress SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 6 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) I Major Transients: List (1-2) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) c

NRC Scenario 6 Scenario Tasks TASK NUMBER RO SRO Warm RFPT 2B lAW 2-01-3 RO U-003-N0-23 259001 A4.02 3.9 3.7 Raise Power with Control Rods RO U-085-N0-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to withdraw from a position other than 00 ROU-085-N0-19 201001 A4.04 3.1 3.1 SRO S-000-AD-3 1 Control Rod Mispositioned RO U-085-AB-07 201002 A2.02 3.2 3.3 SRO S-085-AB-07 Reactor Recirculation Pump Trip RO U-068-AB-1 20200 1A2.03 3.6 3.7 SRO S-068-AB-1 Loss of 480V Unit BD 2A RO U-57B-AL-06 226001 A2.04 3.8 4.2 SRO S-57B-N0-07 RCIC Inadvertent Start ROU-071-N0-5 217000 A2.0l 3.8 3.7 SRO S-000-AD-27 SSI FIRE RO U-000-EM-85 600000 AA2.16 3.0 3.5 RO U-000-SS-30 RO U-000-N0-32 SRO S-000-EM-30 SRO S-000-SS-30 SRO S-000-SS-31 6

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NRC Scenario 6 Simulator Instructor IC 97 Batch File NRC/l4O4nrc-6 imfcu06b RWCU valve failure trg e5 NRC/ehc trg e5 bat NRC/ehcpumptrip-1 trg e10 NRC/ads 1-179 trg elO = dmfad0lm imfrd06r3435 stuck control rod imf rd26b triple notch Preference File NRC/l4O4nrc-6 p1k 01 tog p1k 02 arm silence p1k 03 bat NRC/l4O4nrc-6 p1k 04 imfth03b 2B Reactor Recirc Pump trip p1k 05 imfed07a 480V unit bd 2A loss p1k 06 imf rcO2 RCIC start p1k 07 imfth30d 28 level instrument fails low p1k 08 ior zdihs2388a start RHRSW Pump start B3 p1k 09 ior zdihs23 1 a start RHRSW Pump start Al p1k 10 ior zdihs238a start RHRSW Pump start Cl p1k 11 dor zdihs23la 1k 12 dor zdihs238a p p1ksl p1k s2 imfrdO6r3435 Stuck Control Rod p1k s3 bat ulu3scrarn p1k s4 bat NRC/1404-25-1 1k s5 bat NRC/1404-25-la p pfks6 1k s7 p p1k s8 ior z1ohs6749a2[1] off RHRSW Cl Supply valve p1k s9 ior z1ohs6749a2[2] off p1k slO mrf edl6a emerg_on p1k sli mrfedl6b emerg_on p1k si 2 mrf ed 1 6c emerg_on 8

NRC Scenario 6 Batch File NRC/1404-25-1 imfth30h 100 LT 3-208d failed high imfth3lg 100 PT 3-207 failed high imf th24d condensing pot 3-822 equalization imfad02g SRV 1-4 imf ad02k SRV 1-41 imfad02e SRV 1-31 imf ad02b SRV 1-19 imfadOim 35 SRV 1-179 open ior zdihs0 11 8a close/auto SRV 1-18 ior zdihs0 123 close/auto SRV 1-23 ior zdihs0l 1 80a close/auto SRV 1-180 ior zdihs0l 142 close/auto SRV 1-42 imfhpO3 10 HPCI flow controller imf swO3f RHRSW Pump B3 trip irf fpO4a start Fire Pump A start irf fpO4b irf fpO4c Batch File NRC/1404-25-la imf thO3a imf tho3b SSI Attachment action ior zlohs68aii[2] off ior zlohs685aii[1] off q

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NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B JAW 2-01-3 SRO Directs Warm RFPT 2B JAW 2-01-3 section 5.6 BOP Warm RFPT 2B JAW 2-01-3 section 5.6 5.6 Warming the Second and Third RFPIRFPT [1] REFER TO Section 3.0 and REVIEW Precautions and Limitations. [2] IF RFP/RFPT is warmed and rolling, THEN (Otherwise N/A): [2.1] DEPRESS RFPT Trip Reset.

RFPT 2B TRIP RESET, 2-HS-150A

[2.2] CHECK the following:

Blue light extinguished
HP and LP Stop Valves open NOTE Normal operating range for RFP lube oil to bearings is 110°F to 120°F. Illustration 7 has instruction for controlling Raw Cooling Water through REP lube oil cooler.

CAUTIONS

1) Do not raise RFP discharge pressure to greater than Reactor Pressure to prevent injection to the vessel.

2) Do not exceed 1100 RPM until lube oil temperature is at least 110°F.

[2.3] START RFPT from Panel 2-9-6 as follows:

PLACE RFPT 2B START/LOCAL ENABLE, 2-HS-46-138A, in START and OBSERVE RFPT accelerates to approximately 600 rpm on RFPT 2B SPEED, 2-SI-46-9A.

[2.4] IF lube oil to bearings is not at least 110°F, THEN (Otherwise N/A):

ADJUST RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, as necessary to RAISE RFPT speed to approximately 1 100 rpm.

BOP NOTE Low pressure steam supply valves to second and third RFPT placed in service are opened when directed by 2-01-47, following warming of Main Turbine. [3] IF REP turbine to be started is cold, THEN (Otherwise N/A): 11

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B JAW 2-01-3 BOP Warm RFPT 2B lAW 2-01-3 section 5.6 [3] IF RFP turbine to be started is cold, THEN (Otherwise N/A): [3.1] VERIFY the following valves are OPEN: 2-FCV-1-133, RFPT 2B HP STEAM SUPPLY VALVE using handswitch 2-HS-1-133A. [3.2] IF returning Feedpump to service after maintenance with unit at power AND Main Generator is on line, THEN (Otherwise N/A) [3.3] VERIFY OPEN the following valves to drain Condensate and warm RFPT steam supply lines: [3.3.2] RFPT 2B.

2-FCV-6-127, RFPT 2B HP STOP VLV ABOVE SEAT DR using handswitch 2-HS-6-127A.
2-FCV-6-128, RFPT 2B HP STOP VLV BELOW SEAT DR using handswitch 2-HS-6-128A.
2-FCV-6-129, RFPT 2B FIRST STAGE DRAIN VLV using handswitch 2-HS 129A.
2-FCV-006-O 155, RFPT 2B HP STEAM SHUTOFF ABOVE SEAT DRAIN using handswitch 2-HS-006-0155B (local control).
RFPT 2B LP STOP VLV ABOVE SEAT DR. using 2-HS 125A.
2-FCV-6-126, RFPT 2B LP STOP VLV BELOW SEAT DR using handswitch 2-HS-6-126A.
2-FCV-006-O 156, RFPT 2B LP STEAM SHUTOFF ABOVE SEAT DRAIN using handswitch 2-HS-006-0156B (local control).

          .        When contacted for local report valves are open 2-FCV-006-O 155 and 2-FCV-006-O 156 Driver

[3.4] VERIFY the following on Panel 2-9-5. BOP

RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A is depressed in MANUAL GOVERNOR position with amber light at switch illuminated 12

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B JAW 2-01-3 section 5.6 [3.5] VERIFY one of the following:

RFPT being started has been on Turning Gear or rolling on minimum flow for at least one hour. (Otherwise N/A)
With Unit Supervisor permission, RFPT is rolling on turning gear or minimum flow and no abnormal rubbing noises or vibration is observed.

(Otherwise N/A) [3.6] REFER TO 2-0I-47C and VERIFY Seal Steam being supplied to RFPT being placed in service. [3.7] VERIFY applicable switch in NORMAL and amber light extinguished (Panel 2-9-6):

RFP 2B NPSH TRIP BYPASS, 2-HS-2-122 CAUTION

[NER/C] Failure to have idle RFP discharge valves open during startup after first RFP has been placed in service may result in over-pressurization of piping between RFP discharge valve and discharge check valve due to thermal heating. [INPO SER 92-002] NOTE When returning Feedpump to service with Reactor at rated pressure, Step 5 .7[6] will provide the instructions to reopen Feedpump discharge valves. [3.8] VERIFY OPEN REP discharge valve for pump being placed in service:

RFP 2B DISCHARGE VALVE, 2-FCV-3-12

[3.9] OPEN applicable valve for pump being started (Panel 2-9-6):

RFP 2B SUCTION VALVE, 2-FCV-2-95, using 2-HS-2-95A NOTE B oi Blue light at turbine trip RESET pushbutton should now be illuminated. This indicates all trip signals are reset with control valves closed allowing trip to be reset.

[3.10] ChECK blue light at turbine trip RESET pushbutton illuminated. 13

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B lAW 2-01-3 section 5.6 [3.11] PERFORM RFPT Trip Reset: [3.11.2] DEPRESS RFPT 2B TRIP RESET, 2-HS-3-150A and CHECK the following:

Blue light extinguishes.
HP Stop Valves open.
LP Stop Valves open.

[3.12] REFER TO Section 8.11 and PERFORM Thrust Bearing Wear Detector Periodic Test. 8.11 Thrust Bearing Wear Detector Periodic Test [1] OBTAIN Unit Supervisor approval to perform test. [3] IF testing RFPT 2B Thrust Bearing Wear Detector, THEN: [3.1] VERIFY running:

RFPT 2B 2B1 MAIN OIL PMP
RFPT 2B 2B2 MAIN OIL PMP

[3.2] VERIFY the following on RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139:

Key inserted in switch.
Switch in NORM position.

BOP [3.3] PLACE in TEST RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139. [3.3.1] CHECK the following illuminated:

White light above RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139.
Both white lights above RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140.

[3.4] While maintaining 2-HS-3-139 in TEST position, REMOVE key. [3.5] INSERT key in RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140. [3.6] PLACE RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140 in ACTiVE TEST.

CHECK white test light (right) extinguishes.

14

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B JAW 2-01-3 BOP Warm RFPT 2B JAW 2-01-3 section 5.6 [3.7] PLACE RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140 in NORMAL.

CHECK white test light (right) illuminates.

[3.8] PLACE RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140 in iNACTIVE TEST.

CHECK white test light (left) extinguishes.

[3.9] PLACE RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140 in NORMAL.

CHECK white test light (left) illuminates.

[3.10] REMOVE key from RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS 140. [3.11] INSERT key in RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139. [3.12] PLACE RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139 in NORMAL. ) [3.13] CHECK the following extinguished.

White light above RFPT 2B THRUST BRG WEAR TEST, 2-HS-3-139.
Both white lights above RFPT 2B THRUST BRG WEAR TEST SELECT, 2-XS-3-140.

[3.13] REFER TO Section 8.10 and PERFORM Overspeed Trip Exerciser Periodic Test. 8.10 Overspeed Trip Exerciser Test [1] OBTAIN Unit Supervisor approval to perform this test. [2] For RFPT being tested, VERIFY Main Oil Pump running: BOP *RFPT2B2B1MAINOILPMP

RFPT 2B 2B2 MAiN OIL PMP

[3] PERFORM the following for the RFPT to be Tested. [3.2] PLACE RFPT 2B OVERSPEED TEST TRIP LOCKOUT, 2-HS-3-135A, in MECH and CHECK indications:

Green (normal) light extinguished at switch Amber (mechanical lockout) light to the right of the green light is illuminated 15

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B LAW 2-01-3 section 5.6 [4] PERFORM the following RFPT Overspeed Test. [4.2] DEPRESS and HOLD RFPT 2B OVERSPEED TEST 2-HS-3-136on Panel 2-9-6 and CHECK indications:

White (trip) light illuminated at pushbutton
Green (normal) light extinguished at pushbutton

[5] RELEASE RFPT OVERSPEED TEST pushbutton.

RFPT 2B OVERSPEED TEST pushbutton, 2-HS-3-136

[6] PERFORM the following RFPT Overspeed Test Reset. [6.2] DEPRESS and HOLD RFPT 2B OVERSPEED TEST RESET, 2-HS-3-132, and CHECK the following on Panel 2-9-6:

RFPT 2B OVERSPEED TEST RESET, 2-HS-3-132, white (trip) light illuminated.
RFPT 2B OVERSPEED TEST, 2-HS-3-136 white (trip) light extinguished

[7] RELEASE Overspeed Test Reset on Panel 2-9-6 and CHECK the following. [7.2] RFPT 2B

RFPT 2B OVERSPEED TEST RESET, 2-HS-3-132, white (trip) light extinguished.
RFPT 2B OVERSPEED TEST 2-HS-3-136 green (normal) light illuminated.

NOTE Waiting 30 seconds before placing Overspeed Test Trip Lockout switch in normal position BOP in Step 8.l0[8]. allows ample time for Overspeed test trip logic to reset. [8] PERFORM the following, when 30 seconds have elapsed for the RFPT to be Tested. [8.2] PLACE RFPT 2B OVERSPEED TEST TRIP LOCKOUT, 2-HS-3-135A, in NORM and CHECK indications:

Green (normal) light illuminated at switch
Amber (mechanical lockout) light to the right of the green light is extinguished 16

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B lAW 2-01-3 section 5.6 [3.14] REFER TO Section 8.12 and PERFORM HP and LP Stop Valve Test. 8.12 HP and LP Stop Valve Test [1] OBTAIN Unit Supervisor approval to perform test. [2] IF turbine is stopped OR on turning gear, THEN VERIFY turbine trip has been reset and stop valves open. [4] TEST RFPT 2B HP STOP VALVE as follows: [4.1] CHECK HP SV POSITION, 2-ZI-l-135 Red light illuminated. (located above RFPT 2B HP STOP VLV (2-FCV-l-135) TEST pushbutton, 2-HS-3-134). [4.2] CHECK HP CV POSITION, 2-ZI-1-132B Green light illuminated. (located above RFPT 2B HP STOP VLV (2-FCV-1-135) TEST pushbutton, 2-HS-3-134). [4.3] DEPRESS and HOLD RFPT 2B HP STOP VLV (2-FCV-1-135) TEST pushbutton, 2-HS 134. [4.4] CHECK the following three lights illuminate during valve travel:

HP SV POSITION, 2-ZI-l-135 Red light
HP SV POSITION, 2-ZI-1-135 Green light
HP CV POSITION, 2-ZI-1-132B Green light

[4.5] CHECK HP SV POSITION, 2-ZI-1-135 Red light extinguishes as valve reaches full closed. [4.6] RELEASE RFPT 2B HP STOP VLV (2-FCV-1-135) TEST pushbutton, 2-HS-3-134. [4.7] CHECK RFPT 2B HP STOP VALVE returns to open position using the following indications:

HP SV POSITION, 2-ZI-1-135 Red light illuminated.
HP SV POSITION, 2-ZI- 1-135 Green light extinguished.
HP CV POSITION, 2-ZI-1-132B Green light illuminated.

[7] TEST 2B RFPT LP Stop valves as follows: BOP [7.1] CHECK LP SV TEST POSN, 2-ZI-3-133 Red light is illuminated. (located above RFPT 2B LP STOP VLV (2-FCV-1-131) TEST pushbutton, 2-HS-3-133). 17

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B JAW 2-01-3 BOP Warm RFPT 2B JAW 2-01-3 section 5.6 [7.2] DEPRESS and HOLD RFPT 2B LP STOP VLV (2-FCV-1-13 1) TEST pushbutton, 2-HS 133. [7.3] CHECK the following lights illuminate during valve travel:

LP SV TEST POSN, 2-ZI-3-133 Red light
LP SV TEST POSN, 2-ZI-3-133 Green light

[7.4] CHECK the following light illuminates indicating the test position has been reached:

LP SV TEST POSN, 2-ZI-3-133 White light.

[7.5] RELEASE RFPT 2B LP STOP VLV (2-FCV-1-13 1) TEST pushbutton, 2-HS-3-133. [7.6] CHECK RFPT 2B LP STOP VALVE returns to open position using the following indication:

LP SV TEST POSN, 2-ZI-3-133 Red light illuminated.
LP SV TEST POSN, 2-ZI-3-133 Green light extinguished.
LP SV TEST POSN, 2-ZI-3-133 White light is extinguished.

[3.15] PLACE control switches for the following in AUTO and VERIFY OPEN RFP minimum flow valve for pump being placed in service:

RFP 2A MThl FLOW VALVE, 2-FCV-3-20
RFP 2B MIN FLOW VALVE, 2-FCV-3-13
RFP 2C M1N FLOW VALVE, 2-FCV-3-6 CAUTIONS

1) Do not allow RFPT speed to exceed 1100 rpm until RFPT manual trip has been satisfactorily tested.

2) Do not raise RFP discharge pressure to greater than Reactor pressure to prevent BOP injection to vessel during RFPT trip testing.

NOTE RFPT speed is raised to approximately 1100 rpm to obtain positive indication of discharge pressure. [3.16] START RFPT from Panel 2-9-6 as follows:

PLACE RFPT 2B START/LOCAL ENABLE, 2-HS-46-138A, in START and OBSERVE RFPT accelerates to approximately 600 rpm on RFPT 2B SPEED, 2-SI-46-9A.

18

NRC Scenario 6 Simulator Event Guide: Event 1 Normal: Warm RFPT 2B lAW 2-01-3 BOP Warm RFPT 2B JAW 2-01-3 section 5.6 [3.17] PERFORM the following locally at RFPT: A. CHECK no abnormal rubbing noises or vibration is observed. Driver Report NO abnormal rubbing noises or vibrations [3.18] CHECK RFPT Speed Control in MANUAL GOVERNOR.

USE RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, as necessary to RAISE RFPT speed to approximately 1100 rpm.

[3.20] PLACE RFPT Turning Gear Motor, in AUTO as follows:

RFPT 2B TURNING GEAR MOTOR, using 2-HS-3-127A.

[3.21] PERFORM RFPT Trip Test on Panel 2-9-6 as follows:

DEPRESS RFPT 2B TRIP, using 2-HS-3-151A, and VERIFY HP and LP Stop Valves close by using the position lights.

[3.22] CHECK Turning Gear automatically engages or RFP rolling on minimum flow. [3.23] PERFORM RFPT Trip Reset: BOP B. DEPRESS RFPT 2B TRIP RESET, 2-HS-3-150A and CHECK the following:

Blue light extinguishes.
HP Stop Valves open.
LP Stop Valves open.

CAUTION Do NOT raise RFP discharge pressure to greater than Reactor Pressure to prevent injection to vessel. NOTE Normal operating range for RFP lube oil to bearings is 110°F to 120°F. Illustration 7 has instructions to control Raw Cooling Water through RFP lube oil cooler. [3.24] START RFPT from Panel 2-9-6 as follows:

PLACE RFPT 2B START/LOCAL ENABLE, 2-HS-46-138A, in START and OBSERVE RFPT accelerates to approximately 600 rpm on RFPT 2B SPEED, 2-SI-46-9A.

19

NRC Scenario 6 Simulator Event Guide: Event 1 Nomial: Warm RFPT 2B JAW 2-01-3 BOP Warm RFPT 2B lAW 2-01-3 section 5.6 [3.25] IF the lube oil to the bearings is below 110°F, perform the following: ADJUST RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, as necessary to RAISE RFPT speed to approximately 1100 rpm. [3.261 WHEN lube oil to the bearings reaches 110°F, perform the following:

REDUCE RFPT speed using RFPT 2B SPEED CONT RAISE/LOWER switch, 2-HS-46-9A, until RFP 2B discharge pressure, 2-PI-3-9A on Panel 2-9-6 is less than Reactor pressure.

20

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 SRO Direct Power increase using Control Rods per 2-GO1-100-1A, Section 5.4 5.4 Withdrawal of Control Rods while in Mode 2 [64] VERIFY IRM/APRM overlap by operator visual observation before exceeding 5% power. [66] CONTINUE to withdraw control rods to raise Reactor power JAW 2-01-85 ATC Raise Power with Control Rods JAW 2-01-85, Section 6.6 Group 33 10-35, 26-51, 34-51, 50-35, 50-27, 34-1 1, 26-11, 10-27 from 08 to 12 Group 34 18-43, 42-43, 42-19, 18-19 from 08 to 12 Group 35 = 26-35, 34-35, 34-27, 26-27 from 08 to 12 Group 36 = 02-35, 26-59, 34-59, 58-35, 58-27, 34-03, 26-03, 02-27 from 00 to 12 6.6.1 Initial Conditions Prior to Withdrawing Control Rods [2] VERIFY the following prior to control rod movement:

CRD POWER, 3-HS-85-46 in ON.
Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).

6.6.2 Actions Required During and Following Control Rod Withdrawal [1] IF control rod fails to withdraw, THEN Refer to Section 8.15 for additional methods to reposition control rod. [2] IF control rod double notches, or withdraws past its correct/desired position, THEN Refer to Section 6.7 for inserting control rod to its correct/desired position. [3] IF at any time while driving a selected rod during the performance of this section, the Control Rod moves more than one notch from its intended position, THEN Refer to 2-AOI-85-7, MISPOSITIONED CONTROL ROD. [4] OBSERVE the following during control rod repositioning:

Control rod reed switch position indicators (four rod display) agree with indication on Full Core Display.
Nuclear Instrumentation responds as control rods move through the core (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize Automatic RBM Rod block as follows:

STOP Control Rod Withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM Displays on Panel 9-5 and perform step 6.6.2[6].

21

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 [61 IF Control Rod movement was stopped to keep from exceeding a RBM Setpoint or ATC was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REiNITIALIZE the RBM: [6.1] PLACE the CRD Power, 2-HS-85-46 to the OFF position to deselect the control Rod. [6.2] PLACE the CRD Power, 2-HS-85-46 to the ON position. [6.3] iF desired, THEN CONTINUE to withdraw Control Rods and PERFORM applicable section for Control Rod withdraw. 6.6.3 Control Rod Notch Withdrawal [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40. [2] OBSERVE the following for selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMiNATED.
White light on the Full Core Display ILLUMINATED
Rod Out Permit light ILLUMINATED.

[3] VERIFY ROD WORTH MINIMIZER operable and LATCHED in to correct ROD GROUP when Rod Worth Minimizer is enforcing. [4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [5] OBSERVE control rod settles into desired position AND ROD SETTLE light extinguishes. 22

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows: [6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [6.2] CHECK control rod coupled by observing the following:

Four rod display digital readout AND full core display digital readout AND background light remain illuminated.
CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[6.3] CHECK control rod settles into Position 48 and ROD SETTLE light extinguishes. [6.4] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2. 6.6.4 Continuous Rod Withdrawal NOTES

1) Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches.

2) When in areas of high notch worth, single notch withdrawal should be used instead of continuous rod withdrawal. Information concerning high notch worth is identified by Reactor Engineering in Control Rod Coupling Integrity Check, 2-SR-3.1.3 .5A.

3) When continuously withdrawing a control rod to a position other than position 48, the CRD Notch Override Switch is held in the Override position and then the CRD Control Switch is held in the Rod Out Notch position.

Both switches should be released when the control rod reaches two notches prior to its intended position. (Example: If a control rod is to be withdrawn from position 00 to position 12, the CRD Notch Override Switch and the CRD Control Switch would be used to move the control rod until reaching position 08, then both switches would be released.)
If the rod settles in a notch prior to the intended position, the CRD Control Switch should be used to withdraw the rod to the intended position. (using the above example; If the control rod settles at a notch prior to the intended position of 12, the CRD Control Switch would be used to withdraw the control rod to position 12.)

23

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40. [2] OBSERVE the following for selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED.
White light on the Full Core Display ILLUMINATED
Rod Out Permit light ILLUMiNATED.

[3] VERIFY ROD WORTH MINIMIZER operable and LATCHED in to correct ROD GROUP when Rod Worth Minimizer is enforcing. [41 VERIFY Control Rod is being withdrawn to a position greater than three notches. [5] IF withdrawing the control rod to a position other than 48, THEN PERFO RM the following: (Otherwise N/A) [5.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE. [5.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48. [5.4] IF control rod settles at notch before intended notch, THEN PLACE CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [5.5] WHEN control rod settles into the intended notch, THEN ChECK the following:

Four rod display digital readout and full core display digital readout and background light will remain illuminated.
CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[5.6] CHECK control rod settles at intended position and ROD SETTLE light extinguishes. ATC During power ascension IRM B fails to respond to continuous steady counts, ATC reports to Unit Supervisor SRO Directs Startup to continue, have all needed WM instruments. Need 6 of 8 IRMs. 24

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC NOTE When continuously withdrawing a control rod to position 48, the control rod coupling integrity check can be performed by one of the two following methods:

1) Coupling integrity check while maintaining the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position. If this method is selected, perform Step 6.6.4[6] and N/A Step 6.6.4[7].

2) Coupling integrity check after releasing the CRD Notch Override Switch and the CRD Control Switch. If this method is selected, perform Step 6.6.4[7] and N/A Step 6.6.4[6].

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A) [6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE. [6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position with the control rod at position 48. [6.4] CHECK control rod coupled by observing the following:

Four rod display digital readout and full core display digital readout and background light will remain illuminated.
CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[6.5] RELEASE both CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48. [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes. [6.7] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2. 25

NRC Scenario 6 Simulator Event Guide: Event 2 Reactivity: Raise power with Control Rods, Group 33, 34, 35 and 36 ATC [7] IF continuously withdrawing the control rod to position 48 control rod coupling integrity check is to be performed after the CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A): [7.1] PLACE and HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE. [7.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [7.3] WHEN position 48 is reached, THEN RELEASE CRD NOTCH OVERRIDE, 2-HS-.85-47, and CR1) CONTROL SWITCH, 2-HS-85-48. [7.4] VERIFY control rod settles into position 48. [7.5] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [7.6] CHECK control rod coupled by observing the following:

Four rod display digital readout and full core display digital readout and background light will remain illuminated.
CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window

14) does not alarm.

[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes. [7.8] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2. 6.6.5 Return to Normal after Completion of Control Rod Withdrawal [1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN: [1.1] PLACE CR1) POWER, 2-HS-85-46, in OFF. [1.2] PLACE CR1) POWER, 2-HS-85-46, in ON. 26

NRC Scenario 6 Simulator Event Guide: Event 3 Component: Control Rod 34-35 difficult to withdraw 8.15 Control Rod Difficult to Withdraw ATC [1] VERIFY the control rod will not notch out. Refer to Section 6.6. [2] REVIEW all Precautions and Limitations in Section 3.0 [3] IF RWM is enforcing, THEN VERIFY RWM is operable and LATCI{ED in to the correct ROD GROUP. NOTES

1) Steps 8.15[4] through 8.15[6] should be used when the control rod is at Position 00 while Step 8.15 [7] should be used when the control rod is at OR between Positions 02 and 46.

2) Double clutching of a control rod at Position 00 will place the rod at the overtravel in stop, independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet fmgers from engaging the 00 notch.

3) Step 8.15[4] may be repeated as necessary until it is determined that this method will not free the control rod.

Wb [7] IF the control rod is at or between Positions 02 and 46, THEN PERFORM the (_ 1 following to withdraw the control rod using elevated drive water pressure: [7.1] RAISE the CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. [7.2] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48. [7.3] IF the control rod successfully notches out, THEN LOWER CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A, AND PROCEED TO Section 6.6. CAUTION To prevent a drive from double notching in a high rod worth region and to reduce exposure of drive seals and directional control valves to excessive pressures, the CRD DRIVE WTR HDR DP should be returned to between 250 psid and 270 psid as soon as possible. 27

NRC Scenario 6 Simulator Event Guide: Event 3 Component: Control Rod 34-35 difficult to withdraw [7.4] IF the control rod still fails to NOTCH OUT, THEN RAISE CRD DRIVE ATC WTR HDR DP, 2-PDI-85-17A, to 350 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. [7.5] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48. [7.6] LOWER CRD DRIVE WTR FIDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. [7.7] IF the control rod still fails to NOTCH OUT using elevated CRD DRIVE WTR HDR DP, THEN CONTACT Reactor Engineer and NOTIFY Unit Supervisor for further instructions. ATC Will raise drive water pressure to 350 psid to successfully move Control Rod 34-3 5 from position 8 Driver when ATC raises drive water pressure to 350 psid, delete stuck rod 34-35 NOTE When the control rod moves with drive water pressure at 350 psid the control will triple notch Driver After control rod triple notches RWM Alarm will come in when that alarm energizes insert Shift F2 to stick control rod 34-35 (rd06r3435) ATC Report Control Rod 34-3 5 triple notched to position 14 SRO Enter 2-AOI-85-7, Mispositioned Control Rod Mispositioned Control Rod 2-AOI-85-7 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions [1] STOP all intentional control rod movement. [2] IF Control Rod is determined to be mispositioned, ThEN NOTIFY the following:

Reactor Engineer (RE)
Shift Technical Advisor (STA)
Unit Supervisor
Shift Manager (SM)
Operations Superintendent.

28

NRC Scenario 6 Simulator Event Guide: Event 3 Component: Control Rod 34-35 difficult to withdraw [3] IF the Control Rod is> 2 notches from the intended position, THEN ATC PERFORM the following: (Otherwise N/A) CAUTION NRC/C] Operations outside of the allowable regions shown on the Recirc ulation System Operating Map could result in thermal-hydraulic power oscillations and subsequent fuel damage. Monitoring to be performed during a power decrease and require d actions are contained in 2-GOI-100-12. [NCO 940245010] NOTE

1) Rod moves to recover from mispositioned rod will NOT be any different from normal rod moves.

2) Power level at which recovery is performed and movement of other rods to support recovery are as allowed by 0-TI-248.

3) Rod movement during recovery is governed by 2-GOI-100-1A, 2-GOI

                                                                                                  -100-12, 2-GOI-100-12A and 2-01-85.

[4] IF the Control Rod is 2 notches from the intended position, ThEN PERFORM the following: [4.1] OBTAIN recommendation from the Reactor Engineer. [4.2] IF a Reactor Startup or Shutdown is not in progress, THEN VERIFY 2-GOl- 100-12, Power Maneuvering has been entered if a power change is anticipated. (Otherwise N/A) [4.3] IF required to allow rod movement to correct the rod position error, THEN REDUCE core thermal power as recommended by the Reactor Engineer/Reactivity Control Plan. (Otherwise N/A) [4.4] MOVE Control Rods with the Unit Supervisor permission and Shift Manager concurrence to recover from the rod positioning error as recommended by the Reactor Engineer/Reactivity Control Plan. [6] IF evidence of fuel damage, THEN REFER TO EPIP-1 for emergency classification. [7] INTIATE a Service Request/PER for Control Rod error or mispositioned Control Rod. Driver When called as RE concur with the Unit Supervisors recommendation SRO Direct ATC to insert control rod to position 12 ATC Attempts to insert Control Rod 34-3 5 to position 12, reports control rod failed to move. 29

NRC Scenario 6 Simulator Event Guide: Event 3 Component: Control Rod 34-35 difficult to INSERT 8.16 Control Rod Difficult to Insert [1] VERIFY the control rod will not notch in, in accordance with Section 6.7 or 8.19. [21 REVIEW all Precautions and Limitations in Section 3.0. [3] IF RWM is enforcing, ThEN VERIFY RWM operable and LATCHED in to the correct ROD GROUP. ATC [4] CHECK CRD SYSTEM FLOW is between 40 gpm and 65 gpm, indicated by 2-FIC-85-l1. [5] CHECK CRD DRIVE WTR FIDR DP, 2-PDI-85-17A is between 250 psid and 270 psid. [6] IF the CRD SYSTEM FLOW or CRD DRIVE WTR HIJR DP had to be adjusted, THEN PROCEED TO Section 6.7. [7] IF control rod motion is observed, but the CRD fails to notch-in with normal operating drive water pressure, ThEN: NA [8] IF the control rod problem is believed to be air in the hydraulic system, THEN [9] IF the control rod begins to insert nonnally, ThEN [10] IF the control rod still fails to notch in AND the control rod problem is believed to be air in the hydraulic system, THEN [12] IF the control rod still fails to notch in, ThEN: [12.1] NOTIFY the Unit Supervisor and Reactor Engineer to Refer to section Stuck Control Rod-Test to distinguish a Hydraulic Problem from Mechanical Binding, TI-20, and RETURN to Section 8.16. [12.2] REQUEST the Unit Supervisor and Reactor Engineer to evaluate the control rod operability. Refer to Tech Spec 3.1. SRO Contact Reactor Engineering and Evaluate Tech Spec 3.1 Driver Acknowledge RE communication will commence working on a course of action Driver When directed by the NRC insert preference key F5 for loss of 480V Unit BD 2A 30

NRC Scenario 6 Simulator Event Guide: Event 3 Component: Control Rod 34-3 5 difficult to INSERT SRO Contact Reactor Engineering arid Evaluate Tech Spec 3.1 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE. APPLICABILITY: MODES 1 and 2 Condition A: One withdrawn control rod stuck Required Action A. 1: Verify stuck rod separation criteria are met Completion Time: Immediately AND Required Action A.2: Disarm the associated control rod drive (CRD) Completion Time: 2 hours AND Required Action A.3: Perform SR 3.1.3.3 for each withdrawn OPERABLE control rod Completion Time: 24 hours from discovery of Condition A concurrent with Thermal power greater than the low power setpoint (LPSP) of the RWM AND Required Action A.4: Perform SR 3.1.1.1 Completion Time: 72 hours Driver Acknowledge RE communication will commence working on a course of action Driver When directed by the NRC insert preference key F5 for loss of 480V Unit BD 2A 31

NRC Scenario 6 Simulator Event Guide: Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alarms; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-l and 7B-15. 7A-22 GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action of this ARP 2-XA-55-9-8A Window 1 alarms THEN,

1. VERIFY all available Stator Cooling Water Pumps running.

2. Attempt to RESET alarm

3. IF alarm fails to reset, AND reactor power is above turbine bypass valve capability THEN SCRAM the Reactor B. VERIFY a stator cooling water pump is running and CHECK stator temperature recorder, 2-TR-57-59, Panel 2-9-8.

Generator NOT on line, verifies Stator water cooling pump running 8B-16 480V UNIT BD 2A UV OR XFR A. VERIFY automatic action has occurred. B. INSPECT 480V Unit Bd A for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. C. REFER TO O-OI-57B to re-energize board. D. REFER TO appropriate 01 for recovery or realignment of equipment. Reports trip of 480V Unit Bd 2A, dispatches operators 8C-3 480V RX BLDG VENT BD 2A UV OR XFR A. VERIFY automatic action has occurred. B. CHECK or START refuel floor and reactor zone exhaust fans 2A or 2B. C. CHECK board for abnormal condition: relay targets, smoke, burned paint, breaker position, etc. D. REFER TO O-OI-57B to re-energize or transfer board. E. REFER TO appropriate 01 for recovery or realignment of equipment. 32

NRC Scenario 6 Simulator Event Guide: Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alanns; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-l and 7B-15. 8C-1O 480V RX BLDG VENT BD 2B UV OR XFR A. VERIFY automatic action has occurred. B. CHECK or START refuel floor and reactor zone fans 2A or 2B. C. CHECK 480V Reactor Bldg Vent Bd 2B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. D. REFER TO O-OI-57B to re-energize board. E. REFER TO appropriate 01 for recovery or realignment of equipment. 8C-15 480V TB VENT BD 2A UV OR XFR A. VERIFY alarm by checking the following:

Associated annunciator, TURBiNE BLDG VENTILATION ABNORMAL, (2-XA-5 5-3D, Window 4) in alarm.
Power light on MTOT vapor extractor and EHC fluid heaters, Panel 2-9-7.

B. DISPATCH Personnel to 480V Turb Bldg Vent Bd 2A to CHECK equipment and board status for abnormal conditions.

Mechanical spaces supply fan 2A, 2B and exhaust fan.
Turb room supply fan 2A, 2B and exhaust fans 2A, 2B, 2C, 2D.
EHC fluid transfer and filtering pump.

C. REFER TO O-OI-57B to re-energize or transfer the board. D. REFER TO appropriate 01 for recovery or realignment of equipment. If called to verify if the following boards transferred report that these boards have auto transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Dnver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMTh4 BD 2 33

NRC Scenario 6 Simulator Event Guide: Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alarms; 7A-22, 8C-3, 8C-lO, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15. 8C-16 480V TURB MOV BD 2A UV ORXFR A. VERIFY alarm by checking light indication to the following equipment:

RFW heaters (2B1,2B2,2C1,2C2) extraction isolation valves.
RFPT 2B 2B2 Main Oil Pump.
RFPT 2C 2C1 Main Oil Pump.

B. CHECK board inspected for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. C. REFER TO ICS screen VFDAAL or VFDBAL and verify PROCESS ALARM Internal }{X Fan Power status is OK. D. REFER TO O-OI-57B to re-energize or transfer the board. E. REFER TO appropriate 01 for recovery or realignment of equipment. 8C-25 480V CNDS DEMIN BD 2 UV OR XFR A. VERIFY automatic transfer by dispatching personnel to 480V Cnds Demin Bd 2 to check for the following:

Power available lights illuminated.
Normal disconnect switch 1A open and alternate disconnect 2A closed.
Any abnormal conditions such as breaker trips.

B. NOTIFY Radwaste Operator. C. IF power NOT available, THEN OBTAIN status of the following from Radwaste:

Condensate demins precoat operation.
Condensate backwash transfer operation.
Backwash receiver pit floor drains.

D. REFER TO O-0I-57B for power restoration or transfer instructions. If called to verify if the following boards transferred report that these boards have auto transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 34

NRC Scenario 6 - Simulator Event Guide: Event 4 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alarms; 7A-22, 8C-3, 8C-lO, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15. 7B-l EHC HYD FLUID HDR PRESS LOW A. VERIFY Standby EHC PUMP 2B(2A), 2-HS-47-2A(1A) running. B. CHECK EHC HEADER PRESSURE indicator, 2-PI-47-7 between 1550 and 1650 psig. C. DISPATCH personnel to inspect EHC pump unit. D. IF EHC Hydraulic system fails, THEN VERIFY turbine trips at or below 1100 psig. 7B-15 STANDBY EHC PUMP FAILED A. On Panel 2-9-7:

1. VERIFY alarm by checking EHC HEADER PRESSURE indicator, 2-PI-47-7.

2. VERIFY EHC PUMP 2B, 2-HS-47-2A and/or EHC PUMP 2A, 2-HS-47-1A running.

3. CHECK EHC PUMP 2B PUMP MTR CURRENT 2-EI-47-2 and/or EHC PUMP 2A PUMP MTR CURRENT 2-EI-47-1.

NOTE Lights extinguish at 1300 psig lowering and illuminate at 1500 psig rising.

4. CHECK lights above EHC PUMP 2A TEST pushbutton 2-HS-47-4A and EHC PUMP 2B TEST pushbutton 2-HS-47-5A.

B. DISPATCH personnel to pumping unit to check for abnormal conditions. C. IF EHC Hydraulic System fails, ThEN VERIFY turbine trips at or below 1100 psig. BOP Starts Standby EHC Pump 2B and verifies EHC pressure returns to normal Driver When directed by NRC insert preference key F7 followed 5 seconds later by preference key F6, and when asked to investigate in aux instrument room report US 3-58B is failed at (-) 95 inches. 35

NRC Scenario 6 Simulator Event Guide: Event 5 Instrument: Level Transmitter LT-3-58D, fails low and RCIC starts BOP Responds to alarms 3F-29 BOP 3F-29 RX WTR LVL LOW LOW HPCI/RCIC INIT A. CHECK RPV water level using multiple indications. B. REFER TO the LOIs. Report RCIC has initiated If RCIC initiates on an invalid initiation signal, it is expected that the operato r report the initiation to the US, verify the initiation signal is not valid, and with concur rence from the US trip RCIC. BOP Trip RCIC Driver When sent to the breaker for the RCIC Minimum Flow Valve 250V DC RMOV BD 2B compartment 5D wait two minutes and when told to de-energize the 7 1-34 valve ior ypovfcv7 134 fail_now Crew Determines RCIC minimum flow valve being open is adding water o the suppre ssion pool SRO Directs RCIC minimum flow valve isolated BOP Coordinates with plant operator to de-energize 7 1-34 when the valve is closed SRO Evaluate Technical Specifications Technical Specification 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATW S-RPT) Instrumentation LCO 3.3.4.2 Two channels per trip system for each ATWS.-RPT instrumentation Function listed below shall be OPERABLE:

a. Reactor Vessel Water Level Low Low, Level 2; and

b. Reactor Steam Dome Pressure High.-

APPLICABILITY: MODE 1. Condition A: One or more channels inoperable Required Action A. 1: Restore channel to OPERABLE status OR Required Action A.2: Place channel in trip Completion Time: 14 days 36

NRC Scenario 6 Simulator Event Guide: Event 5 Instrument: Level Transmitter LT-3-58D, fails low and RCIC starts SRO Evaluate Technical Specifications 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LCO 3.3.5.1 The ECCS instrumentation for each Function in Table 3.3.5.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.5.1-1. APPUCABI.E CONDITIONS MODES REQUIRED REFERENCED FUNCTION OR OTHER CHANNELS FROM SPECIFIED PER REQUIRED CONDITIONS FUNCTION ACTION A I

3. High Pressure Cco(arrt Injeebon (HPCI) System a Reaor Vessel Water Level -

4 B Low Low, Level 2 (e) (2d), ( 3 d)

4. Aumabc Depressurization System (ADS) Trip System A a Reaor Vessel Water Level 2 F

Low Low Low, Level 1 (e) 2 ( d) ( 3 d) Condition A: One or more channels inoperable I Required Action A. 1: Enter the Condition referenced in Table 3.3.5.1-1 for the channel. Completion Time: Immediately Condition B: As required by Required Action A. 1 Required Action B.1: NA Required Action B.2: Declare High Pressure Coolant Injection System inoperable. Completion Time: 1 hour from discovery of a loss of HPCI initiation capability Required Action B.3: Place channel in trip Completion Time: 24 hours Condition F: As required by Required Action A. 1 Required Action F. 1: Declare ADS valves inoperable. Completion Time: 1 hour from discovery of a loss of ADS initiation capability in both trip systems Required Action F.2: Place channel in trip Completion Time: 96 hours from discovery of inoperable channel concurrent with HPCI or RCIC inoperable Driver When directed by Lead Evaluator insert F4 for RR Pump 2B trip 37

NRC Scenario 6 Simulator Event Guide: Event 5 Instrument: Level Transmitter LT-3-58D, fails low and RCIC starts SRO Evaluate Technical Specifications 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.2 The RCIC System instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE. APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure

                                                >150 psig.          -

CON DtTIONS REQUIRED REFERENCED FUNCTION CHANNELS PER FROM REQUIRED FUNCTION ACTION A.1 1 Rer Vess Water Level 4 B L Law, Level ( 2 a) Condition A: One or more channels inoperable. Required Action A. 1: Enter the Condition referenced in Table 3.3.5.2-1 Completion Time: Immediately Condition B: As required by Required Action A. 1 Required Action B. 1: Declare RCIC System inoperable. Completion Time: 1 hour from discovery of loss of RCIC initiation capability Required Action B.2: B.2 Place channel in trip. Completion Time: 24 hours 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE. APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure

                                               > 150 psig.

Condition A: RCIC System inoperable. Required Action A. 1: Verify by administrative means HPCI is OPERABLE. Completion Time: Immediately Required Action A.2: Restore RCIC System to OPERABLE status. Completion Time: 14 Days Driver When directed by Lead Evaluator insert F4 for RR Pump 2B trip 38

NRC Scenario 6 Simulator Event Guide: Event 6 Component: RR Pump 2B Trip ATC Responds to numerous alarms and report trip of RR 2B Pump SRO Enter 2-AOI 1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operab le [2] IF a single Recirc Pump tripped, ThEN CLOSE tripped Recirc Pump discha rge valve. ATC [3] IF Region I or II of the Power to Flow Map is entered, THEN IMMEDIATE LY take actions to INSERT control rods to less than 95.2% loadline. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation. ATC Range IRMs Down [9] NOTIFY Reactor Engineer to PERFORM the following: [10] WHEN the Recirc Pump discharge valve has been closed for at least five minute s (to prevent reverse rotation of the pump), THEN OPEN Recirc Pump discha rge valve as necessary to maintain Recirc Loop in thermal equilibrium. SRO Tech Spec 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation. APPLICABILITY: MODES 1 and 2. Condition A: Requirements of the LCO not met. Required Action A. 1: Satisfy the requirements of the LCO. Completion Time: 24 hours. 39

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire Driver When directed by NRC or if the crew scrams the reactor insert mrf fpO 1 alarm for the fire alarm. Commence Radio traffic between the Driver and the shift manager Report a fire reported at RHRSW Pump Rooms Insert Preference Key F8 start of RHRSW Pump B3 Two minutes after initial report, report as the Incident Commander that there was a fuel oil truck accident at the RHRSW Pump Rooms, RHRSW Pump Room B is fully involved and the fire appears to be spreading to RHRSW Pump Room D Insert Preference Key F9 start of RHRSW Pump Al Has incident Commander request off site Fire fighting support. Insert Preference Key FlO start of RHRSW Pump Cl Insert Preference Key Fl 1 and F12 after RHRSW Pump Cl start Two minutes later inform crew Fire is still active but silencing fire alarm As the Shift Manager call Unit 2 and have them enter O-SSI-25-l, after RHRSW Pump Cl starts. Enters 0-SSI-25-l Reactor is scrammed insert preference key shift F3 to scram Unit 1 and 3 and Shift F4 for bat file NRC/1404-25-1, after batch file 1404-25-1 is started go to remote function summary (RFS) and start the three fire pumps. 0-SSI-25-l 2.0 UNIT 2 CONTROL ROOM OPERATOR ACTIONS NOTE When 0-SSI-25-l is in use, other plant procedures such as Ols, AOIs, and EOIs will be used concurrently with this SSI as plant conditions warrant. In the case where direction conflicts between this SSI and another plant procedure, this SSI will take precedence. SRO TBD-2 [1] DIRECT Unit 3 Unit Supervisor to PERFORM Section 3.0 of 0-SSI-25-1 to Scram Unit 3, AND PROCEED TO cold shutdown. [21 DIRECT Unit 1 Unit Supervisor to PERFORM Section 4.0 of 0-SSI-25-1 to Scram Unit 1, AND PROCEED TO cold shutdown. 40

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire Crew NOTE The following instruments are those which have been credited for safe shutdown, and must be referenced when executing actions for this fire zone: 2-LI-3-58A and 2-PI-3-74A for reactor level and pressure. 2-XR-64-50 CH 2 (PT-64-50) for drywell pressure. 2-XR-64-50 CH 1 (TE-64-52C) for drywell temperature 2-FI-74-50, RHR SYS I Flow 2-FI-23-42, RHR I{X 2C RHRSW Flow TBD-81 2-L1-64-66 and 2-TI-64-161 for suppression pool level and temperature. (0 Mm) (0 Mm) [3] DIRECT Unit 2 Operator to perform the following: TBD-3 TBD-1 [3.1] VERIFY reactor Scram AND RECORD current time (SSI time of entry). Time SRO TBD-13 [4] DIRECT all operators to perform the following: (90 Mm) A. Operator 1 Section 1.0 of Attachment 1 (Places Fire Pump B local controls to EMERG at 4KV Shutdown Board) (90 MinJl2O Mm) B. Operator 2 Section 1.0 of Attachment 2 (Places Fire Pumps A & C local controls to EMERG at 4KV Shutdown Boards, trips Unit 2 RPT breakers) SRO NOTE Diesel Generator loading should be closely monitored and maintained within the limits of 0-01-82. Prompt action to secure non-Appendix R designated loads should be taken to prevent overloading a Diesel Generator/4KV Shutdown Board prior to an RHR pump start. Some 480V non-essential loads may require load shedding to keep Diesel Generator AI4KV Shutdown Board A within load limits. (0 Mm) [5] DIRECT Unit 2 Operator to perform the following: TBD-4 [5.1] VERIFY Main Steam Isolation Valves closed. [5.2] VERIFY Reactor Recirculation Pumps tripped. ATC/BOP Close MSIV and Trip Reactor Recirc Pumps 41

NRC Scenario 6 Simulator Event Guide: Event 8 Component: SRV 1-179 fails open BOP Respond to SRV Open Alarm 3C-25 and reports SRV 1-179 is partially open Alarm 3C-25, MAIN STEAM RELIEF VALVE OPEN A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 2-TR-1-1, on Panel 2-9-47 and SRV Tailpipe Flow Monitor on Panel 2-9-3 for raised temper ature and flow indications. B. REFER TO 2-AOl-i-i. SRO Enters 2-AOl-i-i BOP Relief Valve Stuck Open 2-AOl-i-i 4.1 Immediate Action [1] IDENTIFY stuck open relief valve by OBSERVING the following:

SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3, OR
MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-l-1 on Panel 2-9-47.

[3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch as require d up to three times:

CLOSE to OPEN to CLOSE positions

[4] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. BOP Reports SRV 1-179 is partially open cycles SRV 1-179 and report relief valve closed 42

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire SRO (0 mm) [6] DIRECT Unit 2 Operator to verify the following pumps are NOT operati ng, at Panel 2-9-3:

RHRSW PUMP B2, O-HS-23-19A12
RHRSW PUMP D2, O-HS-23-27A12 ATCIBOP Verifies RHRSW Pumps B2 and D2 are NOT operating SRO CAUTION TBD-87 To prevent loss of the HPCI systems, reactor pressure should NOT be allowe d to decrease below 150 psig until low pressure system injection is available.

NOTES L) HPCI system injection and relief valve operation will be used to maintain reactor water level and a controlled 100°F per hour cooldown rate until the proper alignments for LPCI injection can be perfonned.

2) The HPCI system flow controller should be left in the AUTO position if functioning properly.

3) Credited relief valves for this unit as listed on Illustration 1 are, 2-PC V-00l-0005, 2-PCV-00 1-0022, 2-PCV-00 1-0030, and 2-PCV-00 1-0034.

TBD-102 (10 Mill) [7] DIRECT Unit 2 Operator to initiate a controlled 100°F per hour cooldown rate using HPCI (pressure control mode) and relief valves. ATCJBOP Places HPCI in pressure control within 10 minutes of the scram and commences a controlled cooldown not to exceed 100°F per hour cooldown rate. Report FIPCI controller failed in automatic but is controlling in manual. SRO TBD-101 (10 Mm) [8] DIRECT Unit 2 Operator to place PATH A VENT FLOW CONT, 2-FIC-84-20, in MANTJAL and 0 SCFM, at Panel 2-9-55. ATC/BOP Places 2-FIC-84-20 in Manual and 0 SCFM within 10 minutes 43

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire TBD-89 (10 Mm) [9] DIRECT Unit 2 Operator to place the following switches in TEST/INHIB IT, at Panel 2-9-3: A. ECCS SYS I HI DW PRESS TEST/iNHIBIT, 2-HS-75-59. B. ECCS SYS II HI DW PRESS TEST/INHIBIT, 2-HS-75-60. ATC/BOP Place 2-HS-75-59 and 2-HS-75-60 in Testllnhibit within 10 minutes SRO CAUTIONS

1) RHRSW flow may be diverted to EECW by 0-FCV-067-0049 which may also be potentially affected by the fire and has the power removed at 480V Diesel Aux Board by Operator 3 during performance of Attachment 1.

2) To keep from overloading the Diesel Generator, the RHR pump should be started prior to starting the RI{RSW pump.

3) RHRSW Pumps Al and Cl are potentially affected by the fire and could spuriously start, if so the pumps can be secured from the Main Control Room to preven t dead heading. The Al pump is credited on Unit 3 for decay heat removal, when required, and can be restarted from MCR.

(10 Mm) [10] IF Cl RHRSW Pump has spuriously started, THEN STOP Cl RHRS

W PUMP, using 0-HS-23-8A12.

ATC/BOP Stops Cl RHRSW Pump after it had spuriously started SRO (10Mm) [11] IF B3 RHRSW Pump has spuriously started, THEN STOP B3 RHRSW PUMP, using 0-HS-23-88A/2. (10 Mm) [12] IF Al RHRSW Pump has spuriously started, THEN STOP Al RHRSW PUMP, using 0-HS-23-1A12. ATC/BOP Stops RHRSW Pumps B3 and Al after they spuriously started TBD-76 (10 mm) [13] DIRECT Unit 2 Operator to verify R}{RSW PUMP Cl SPLY TO EECW SRO FCV-67-49, 0-HS-67-49/A2, is closed. [14] DIRECT Unit 2 Operator to verify the following pumps are NOT operati ng at Panel 2-9-3:

RHRSW PUMP C2, 0-HS-23-12A/2.
RHRSW PUMP Al, 0-HS-23-1A12.

44

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire Driver After RHRSW B3 and Al are stopped call control room and report Attachment 1 Section 1 complete. Insert Preference Key Shift F5 Bat NRC/1404-25-la SRO NOTE When suppression pool cooling is required, then RHR Pump 2C and RIIRS W Pump C2 are the credited pumps for this unit. RHRSW Pump C2 can be operated from the Control Room as needed. TBD.-106 (90 mm) [15] WHEN RHRSW Pump C2 is required to be started for suppression pool coolin g, THEN DIRECT Unit 2 Operator to start the pump using RHRSW PUMP C2, handswitch 0-HS-23-(12A12), [16] VERIFY the following complete: A. Attachment 1 Section 1.0 (Placed Fire Pump B local controls to EMERG at 4KV Shutdown Board) B. Attachment 2 Section 1.0 (Placed Fire Pumps A & C local controls to EMERG at 4KV Shutdown Boards, tripped Unit 2 RPT breakers) Driver When contacted report Attachment 2 Section 1 complete. SRO [17] NOTIFY site Operations personnel of the following: A. Safe Shutdown procedure 0-SSI-25-l is in effect. TBD-10 B. Spurious equipment operation may occur. [18] NOTIFY Security Shift Supervisor of the following [18.1] When implementing safeguard measures NOT to lock down doors. [18.2] DISPATCH Security personnel to ensure the following Electrical Board Room doors to Unit 1(2,3) Rx Buildings are unlocked to allow Operator ingress/egress:

Door 634 (El. 621, Electrical Board Rm 1A)
Door 648 (El. 621, Electrical Board Rni 2A)
Door 657 (El. 621, Electrical Board Rm 3A)
Door 538 (El. 593, Electrical Board Rrn 1B)
Door 540 (El. 593, Electrical Board P.m 2B)
Door 513 (El. 593, Electrical Board Rm 3B).

45

NRC Scenario 6 Simulator Event Guide: Event 7 MAJOR: SSI Fire SRO TBD-18 (30 Mm) [19] DIRECT Unit 2 Operator to close 2-FCV-069-0002, using RWCU OUTB D SUCT ISOLATION VALVE, 2-HS-69-2A, at Panel 2-9-4. ATC/BOP Close 2-FCV-069-2 TBD-37 (60 Mm) [20] DIRECT Operator 1 to perform Section 2.0 of Attachment 1 to reset Battery Chargers I, 2A, and 3. NOTE Steps 2.0[21] and 2.0[22] may be performed concurrently. TBD-105 (90 Mm) [21] DIRECT Operator 1 to perform Section 3.0 of Attachment 1 to de-ene rgize and verify closed 0-FCV-067-0049. TBD-30 TBD-31 TBD-32 TBD-33 (240 Mm) [22] DIRECT Operator 2 to perform Section 2.0 of Attachment 2 to align and start Unit and 2 ventilation systems and electric board room AHUs. TBD-82 (540 Mm) [23] DIRECT Unit 2 Operator to verify closed 2-FCV-001-0055 using MN STM LINE DRAIN INBD ISOLATION VLV, 2-HS-1-55A, on Panel 2-9-3. TBD-1 00 [24] IF drywell temperature approaches 281 °F, THEN INITIATE shutdo wn cooling. When directed to perform Section 2 of attachment 1 wait 4 minutes and insert Preference Key Shift FlO, Shift Fl I and Shift F12 and report Attachment 1 section 2 complete Driver When directed to perform section 3 of attachment 1 wait 5 minutes and insert Preference Key Shift F8 and Shift F9 and report Attachment 1 section 3 complete When directed to perform section 2 of attachment 2 acknowledge direction 46

NRC Scenario 6 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs: None OperationsIMaintenance for the Shift: 1.3% power. 2-GOI-100-IA Section 5.4 Step 63.3 and 65. Warm RFPT 2B lAW 2-01-3 Section 5.6. Continue to pull rods in accordance with RCP. Units 1 and 3 are 100% Power Unusual Conditions/Problem Areas: Severe Thunderstorm Warnings are in affect for the entire area for the next 2 hours. 47

Appendix D Scenario Outline Form ES-B-i acility: Browns Ferry NEP Scenario No.: NRC 7 Op-Test No.: 1404 Examiners:_____________________ Operators: SRO:______________________ ATC:_______________ BOP:________________ Initial Conditions: Reactor Power is 23%, Unit startup is in progress lAW 3-GOI-lO0-1A. EECW A3 and Steam Packing Exhauster 3A are out of service. Turnover: Inert the Primary Containment in accordance with 3-01-76 starting at step 47. Commence a power increase with Control Rods to 30% in accordance with Reactivity Control Plan. The Spare RBCCW pump is in service and the 3B RBCCW Pump is ready to be returned to service. Event Malf. No. Event Type* Event Description No. N-BOP 1 Inert the Primary Containment in accordance with 3-01-76 N-SRO R-ATC 2 Power increase with Control Rods 30% R-SRO I-ATC 3 rd0la CRDPump3Atrip I-SRO C-ATC 4 sw02a RBCCW 3A Pump trip TS-SRO dg03b C-BOP Loss of EECW C3 Pump through loss of 4KV S/D Board 3EB TS-SRO C-BOP 6 ms0 1 Loss of Condenser Vacuum C-SRO 7 edO 1 M-ALL Loss of Offsite Power 8 dg0la C DG 3EA Fails to Auto start 9 th2l M-All LOCA 10 hpO4 C HPCI Steam Supply Valve fails to auto open.

     *  (N)ormal,   (R)eactivity, (I)nstrument,    (C)omponent,  (M)aj or

Appendix D Scenario Outline Form ES-D-1 Critical Tasks - Four RPV Level maintained above -162 inches, HPCI has been manually initiated.

1. Safety Significance:

Maintaining adequate core cooling

2. Cues:

RPV level indication

3. Measured by:

HPCI injecting at required flow rate

4. Feedback:

RPV level trend HPCI injection valve open With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to maintain or restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates available low pressure ECCS systems and injects into the RPV to maintain or restore water level above -162 inches.

4. Feedback:

Reactor water level trend. Reactor pressure trend.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Four When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays JAW with EOI Appendix 1 7B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RI-JR flow to containment

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Four To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS AJB INHIBITED annunciator status.

Appendix D Scenario Outline Form ES-D-1 Events

1. BOP will continue with Inerting Primary Containment lAW 3-01-76

2. ATC will commence to raise power with control rods to 30%.

3. CR0 Pump 3A will trip during control rod movements, the ATC will recover CR0 parameters lAW 3 -A0I-85 -3. Inserting the rod one notch will restore position indication.

4. The crew will respond to a trip of RBCCW Pump 3A lAW 3-A0I-70-1, The spare RBCCW pump is aligned and will be the only one running until the crew places RBCCW Pump 3B in service. The RBCCW sectionalizing valve will fail to auto close on the trip of RBCCW Pump 3A, the ATC will close the sectionalizing valve. The SRO will evaluate TRM 3.4.1 and take actions for failure to meet surveillance requirement TSR 3.4.1.1.

5. The crew will respond to a loss of 4KV Shutdown Board 3EB. This will result in a loss of the running EECW pump C3. The operator will take action to start EECW pump Cl. The SRO will refer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Cl EECW pump has been aligned the SRO will determine TS 3.7.1 Condition A now applies. In addition Technical Specification 3.8.7 Condition A with the board inoperable. TS 3.8.7 Condition A requires the associated diesel generator to be declared inoperable and enter TS 3.8.1 Condition B.

6. Condenser Vacuum will begin to degrade the SRO will initially enter 3-A0I-47-3 and then when the main turbine is tripped the SRO will enter 3-A0I-47-1. Condenser Vacuum will continue to degrade.

7. Prior to the SRO directing a Reactor Scram on vacuum a Loss of Offsite Power will occur. The crew will respond to the Reactor Scram JAW 3-A0I-l00-1 and 0-A0I-57-1A.

8. During the LOOP DG 3EA will fail to automatically start and will have to be manually started.

9. Sometime after the LOOP a LOCA will develop requiring the crew to utilize systems to maintain Reactor Level and Containment parameters.

10. The HPCI Steam Supply Valve, 3-FCV-73-16, will fail to OPEN on an automatic HPCI initiation signal.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Drywell Sprays initiated Reactor Level is maintained above TAF

Appendix D Scenario Outline Form ES-D-1 SCENARIO NUMBER: 7 10 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yeso)

Appendix P Scenario Outline Form ES-P-i Scenario Tasks TASK NUMBER EQ SEQ Inert Primary Containment RO U-076-NO-01 223001 A4.03 3.4 3.4 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-NO-31 2.2.2 4.6 4.1 Loss of RBCCW RO U-070-AL-03 400000A2.01 3.3 3.4 SRO S-070-AB-01 CRD Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 Loss of Condenser Vacuum RO U-047-AB-03 295002 AA1.05 3.2 3.2 SRO S-047-AB-03 LOOP RO U-57A-AB-01 295003 AA1.03 4.4 4.4 RO U-082-AL-07 SRO S-57A-AB-01 LOCA RO U-000-EM-01 RO U-000-EM-02 295024 EA1 .11 4.2 4.2 RO U-000-EM-05 295031 EA2.04 4.6 4.8 RO U-000-EM-3 1 RO U-000-EM-32 RO U-000-EM-80 SRO S-000-EM-01 SRO S-000-EM-02 SRO S-000-EM-05

C NRC Scenario 7 cility: a7 Browns Ferry NPP Scenario No.: NRC 7 Op-Test No.: Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Reactor Power is 23%, Unit startup is in progress JAW 3-GOI-l00-1A. EECW A3 and Steam Packing Exhauster 3A are out of service. Turnover: Inert the Primary Containment in accordance with 3-01-76 starting at step 47. Commence a power increase with Control Rods to 30% in accordance with Reactivity Control Plan. The Spare RBCCW pump is in service and the 3B RBCCW Pump is ready to be returned to service. Event Maif. No. Event Type* Event Description No. N-BOP 1 Inert the Primary Containment in accordance with 3-01-76 N-SRO R-ATC 2 Power increase with Control Rods 30% R-SRO C-ATC 3 rd0 1 a CRD Pump 3A trip C-SRO C-ATC 4 sw02a CCW 3A Pump trip. TSSRO 5 dg03b Loss of EECW C3 Pump through loss of 4KV S/D Board 3EB C-BOP 6 Loss of Condenser Vacuum. msOl C-SRO 7 edO 1 M-ALL Loss of Offsite Power 8 dg0la C DG 3EA Fails to Auto start 9 th2l M-ALL LOCA 10 hpO4 C HPCI Steam Supply Valve fails to auto open.

      *  (N)ormal,  (R)eactivity, (I)nstrument,    (C)omponent,   (M)aj or

NRC Scenario 7 Critical Tasks Four RPV Level maintained above -162 inches, HPCI has been manually initiated.

1. Safety Significance:

Maintaining adequate core cooling

2. Cues:

RPV level indication

3. Measured by:

HPCI injecting at required flow rate

4. Feedback:

RPV level trend HPCI injection valve open With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to maintain or restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. Pressure below low pressure ECCS system(s) shutoff head.

3. Measured by:

Operator manually starts initiates available low pressure ECCS systems and injects into the RPV to maintain or restore water level above -162 inches.

4. Feedback:

Reactor water level trend. Reactor pressure trend. n

NRC Scenario 7 Critical Tasks Four When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays JAW with EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment

NRC Scenario 7 Critical Tasks Four To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

Maintain adequate core cooling, prevent degradation of fission product barrier.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS A/B iNHIBITED annunciator status.

NRC Scenario 7 Events

1. BOP will continue with Inerting Primary Containment lAW 3-01-76

2. ATC will commence to raise power with control rods to 30%.

3. CRD Pump 3A will trip during control rod movements, the ATC will recover CRD parameters lAW 3-A0I-85-3. Inserting the rod one notch will restore position indication.

4. The crew will respond to a trip of RBCCW Pump 3A JAW 3-A0I-70-1.The spare RBCCW pump is aligned and will be the only one running until the crew places RBCCW Pump 3B in service. The RBCCW sectionalizing valve will fail to auto close on the trip of RBCCW Pump 3A, the ATC will close the sectionalizing valve. The SRO will evaluate TRM 3.4.1 and take actions for failure to meet surveillance requirement TSR 3.4.1.1.

5. The crew will respond to a loss of 4KV Shutdown Board 3EB. This will result in a loss of the running EECW pump C3. The operator will take action to start EECW pump Cl. The SRO will refer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Cl EECW pump has been aligned the SRO will determine TS 3.7.1 Condition A now applies. In addition Technical Specification 3.8.7 Condition A with the board inoperable. TS 3.8.7 Condition A requires the associated diesel generator to be declared inoperable and enter TS 3.8.1 Condition B.

6. Condenser Vacuum will begin to degrade the SRO will initially enter 3-AOI-47-3 and then when the main turbine is tripped the SRO will enter 3-A0I-47-1. Condenser Vacuum will continue to degrade.

7. Prior to the SRO directing a Reactor Scram on vacuum a Loss of Offsite Power will occur. The crew will respond to the Reactor Scram JAW 3-A0I-100-1 and 0-A0I-57-1A.

8. During the LOOP DG 3EA will fail to automatically start and will have to be manually started.

9. Sometime after the LOOP a LOCA will develop requiring the crew to utilize systems to maintain Reactor Level and Containment parameters.

10. The HPCI Steam Supply Valve, 3-FCV-73-16, will fail to OPEN on an automatic FIPCI initiation signal.

NRC Scenario 7 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Drywell Sprays initiated Reactor Level is restored and maintained above TAF SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 7 10 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No)

NRC Scenario 7 Scenario Tasks TASK NUMBER EQ SEQ Inert Primary Containment RO U-076-NO-01 223001 A4.03 3.4 3.4 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-NO-31 2.2.2 4.6 4.1 Loss of RECCW RO U-070-AL-03 400000A2.01 3.3 3.4 SRO S-070-AB-01 CRD Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 Loss of Condenser Vacuum RO U-047-AB-03 295002 AA1.05 3.2 3.2 SRO S-047-AB-03 LOOP RO U-57A-AB-01 295003 AA1.03 4.4 4.4 RO U-082-AL-07 SRO S-57A-AB-01 LOCA RO U-000-EM-01 RO U-000-EM-02 295024 EA1 .11 4.2 4.2 RO U-000-EM-05 295031 EA2.04 4.6 4.8 RO U-000-EM-31 RO U-000-EM-32 RO U-000-EM-80 SRO 5-000-EM-Ol SRO S-000-EM-02 SRO S-000-EM-O5

NRC Scenario 7 Procedures Used/Referenced: Procedure Number Procedure Title 3-01-76 Containment Inerting System 3-G0I-iOO-1A Unit Startup 3-01-85 Control Rod Drive System 3-A0I-85-3 CRD System Failure 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3-01-70 Reactor Building Closed Cooling Water 3-01-69 Reactor Water Cleanup 3-A0I-70-i Loss of Reactor Building Closed Cooling Water 3-ARP-9-4C Panel 9-4 3-XA-4C Technical Specifications 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-23B Panel 9-23 3-XA-55-23B 3-ARP-9-20A Panel 9-20 3-XA-55-20A 0-01-67 Emergency Equipment Cooling Water System 3-A0I-47-3 Loss of Condenser Vacuum 3-A0I-47-l Unplanned Turbine Trip Below 30 Percent Reactor Power (w/o Rx Scram) 3-ARP-9-53 Panel 9-53 3-XA-55-53 3-ARP-9-7B Panel 9-7 3-XA-55-7B 0-A0I-57-1A Loss of Offsite Power (161 and 500 KV)/Station Blackout 3-A0I-100-1 Reactor Scram 3-E01-1 RPV Control Flowchart 3-E01-2 Primary Containment Control 3-E01-3-C- I Alternate Level Control 3-E0I Appendix-hA Alternate RPV Pressure Control Systems MSRVs 3 -E0I-Appendix- 11 C Alternate RPV Pressure Control Systems HPCI Test Mode 3-E0I Appendix-17C RHR System Operation Suppression Chamber Sprays 3-EOI Appendix-i 7B RHR System Operation Drywell Sprays 3 -EOI Appendix-i 7A RHR System Operation Suppression Pool Cooling 3-EOI Appendix-SB Injection System Lineup CRD 3-EOI Appendix-SC Injection System Lineup RCIC 3-EOI-Appendix-SD Injection System Lineup HPCI 3-EOI-Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode 3-EOI-Appendix-6C Injection Subsystems Lineup RHR System II LPCJ Mode 3-EOI Appendix-6D Injection Subystems Lineup Core Spray System I 3-EOI Appendix-6E injection Subystems Lineup Core Spray System II 3-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 3-01-74 Residual Heat Removal System 3-EOI Appendix-8B Reopening MS1Vs / Bypass Valve Operation EPIP- 1 Emergency Classification

NRC Scenario 7 Batch File nrcl4O4-7.bat

Tagout ior ypobkrrhrswpa3 fail_ccoil ior zdihs23 85a[2] stop ior zlohs2385a[1j off imf sw03c ior ZDIHS665OA[1] null ior zlohs665Oa[2] off ior zlohs6650a[1] off
Drywell purge lineup irfpcO6 (el 0) start
CRD Pump Trip imf rd0 1 a (e2 0)
RBCCW Pump Trip imf sw02a (e5 0) ior ypovfcv7o48 failpower_now ior zlohs7O48a[2j on trg 6 nrc2Ol 17048 trg 6 bat nrcrbccw

#Condenser vacuum loss imfmc04 (e15 0)100 imfmc03 (e15 0)
#4kv 3eb loss imfed09b (elO 0) imfsw03j (elO 0:30) irfsw06 (e12 0) close irfsw07 (e13 0) aligned
#Major imfedOl (e20 0) imf dg0 I a imfdgo2b (e20 0:45) imf hpO4 imfedllc (e20 1:00) imfedlld(e20 1:00) trg e23 bat eecw trg e24 bat eecw- 1 trg e25 bat rpsreset trg e26 bat ca mrf swO9 (e27 0) open imfth2l (e20 10:00)4 10:00 mrf swO2 (e3 0 0) align

NRC Scenario 7 nrcrbccw.bat dor ypovfcv7O48 dor zlohs7O48a[2] Trigger Files nrc2Ol 17048 zdihs7O48a[ 1] .eq. 1 Scenario 7 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 188 Simulator_Setup Simulator Setup Load Batch bat nrcl4O4-7 Simulator_Setup Simulator Setup Verify file loaded RCP required 24% -30% with control rods Provide marked up copy of 3-GOI-100-1A at step 33 Provide marked up copy of 3-01-76 at step 47 of section 5.1 I fl

NRC Scenario 7 Simulator Event Guide: Event 1 Normal: Inert the Primary Containment in accordance with 3-01-76 SRO Directs BOP to continue with Inerting Primary Containment BOP Inerts Primary Containment lAW 3-01-76 [47] PERFORM the following to align the Drywell for purging: [47.1] OPEN DRYWELL VENT 1NBD ISOL VALVE, using 3-HS-64-29 (Panel 3-9-3). [47.2] OPEN DRYWELL VENT OUTBD ISOLATION VLV, using 3-HS-64-30 (Panel 3-9-3). [48] START CTMT PURGE FILTER FAN, using local pushbutton, 3-HS-64-13 1 (Reactor Bldg El 621) (N/A if Containment Purge Filter Fan is not used). [49] OBSERVE purge filter differential pressure is less than 6 H20, as indicated by the total DP across HEPA filter, 3-PDI 125, and charcoal filter, 3-PDI 126, (Reactor Bldg El 621). [50] OPEN DRYWELL ATM SUPPLY 1NBD ISOLATION VLV, using 3-HS-64-18 I) (Panel 3-9-3). [511 VERIFY OPEN, N2 TANK A LIQUID SHUTOFF VLV, 0-SHV-076-0504, at the N2 storage tank, to ensure pressure build circuit is in service to maintain 70 to 80 psig storage tank pressure. [52] VERIFY OPEN PR! CTMT N2 PURGE OUTBD ISOLATION VLV, using 3-HS-76-24 (Panel 3-9-3). [53] While throttling N2 TANK A PURGE VAPORIZER N2 iNLET SHUTOFF VLV, 0-SHV-076-0508 and/or N2 PURGE PCV BYPASS, 0-76-514 as necessary, VERIFY N2 PURGE SUPPLY PRESSURE, 0-PI-76-21B, is between 18 and 22 psig (0-LPNL-925-01 89). [54] CHECK DW/SUPP CHBR PURGE AIRJN2, 3-FI-76-25, stabilizes between 3000 and 4500 SCFM at Panel 3-9-3. Driver Step 48 Trigger 1 to start irfpco6 (el 0) start Containment Purge Filter Fan HS-76-131

NRC Scenario 7 Simulator Event Guide: Event 1 Normal: Inert the Primary Containment in accordance with 3-01-76 BOP Inerts Primary Containment JAW 3-01-76 [55] CONTINUOUSLY MONITOR frostlme on piping to the nitrogen purge vaporizer at the N2 storage tank. [56] THROTTLE OPEN, STEAM CONTROL VALVE BYPASS, 0-76-534, as necessary to prevent frostline from extending past the purge vaporizer. [57] MONITOR N2 TANK A LIQUID LEVEL, 0-LI-076-0003(0-LPNL-925-0 189), periodically. [58] IF N2 STORAGE TANK requires filling, THEN NA [59] MONITOR DW/SUPPR CHBR PURGE AIRJN2 TEMP, 3-TI-76-26, periodically during inerting to ensure?: 50°F (Panel 3-9-3). [60] IF unable to maintain nitrogen temperature?: 5 0°F, THEN CLOSE PRI CTMT N2 PURGE OUTBD ISOLATION VLV, using 3-HS-76-24, Panel 3-9-3, NA [61] VERIFY CLOSED, N2 PURGE PCV BYPASS, 0-76-5 14. NOTE 4 Since inerting or purging of the Drywell and the Suppression Chamber is NOT allowed to be performed at the same time, only one 02 analyzer will be affected by the purge. Nitrogen purging will continue until oxygen indication, 3-XR 110 (Panel 3-9-54), shows a value of < 3.5% oxygen. If possible, continue inerting until oxygen indication, 3-XR-76-110 (Panel 3-9-54), shows a value of< 2% oxygen. This will require 3 to 4 volume changes (approximately 700,000 SCF of nitrogen) and may take up to 4 hours to accomplish. [62] WHEN the Analyzer Oxygen indication 3-XR 110 (Panel 3-9-54) shows a value of< 3.5%, preferably < 2%, THEN STOP CTMT PURGE FILTER FAN, using local pushbutton 3-HS 131 (Reactor Bldg. El 621) [63] CLOSE the following valves (Panel 3-9-3):

DRYWELL VENT INBD ISOL VALVE, using 3-HS-64-29
DRYWELL VENT OUTBD ISOLATION VLV, using 3-HS-64-30

[64] WHEN Drywell pressure reaches 0.9 to 1.1 psig, THEN [64.1] CLOSE DRYWELL ATM SUPPLY INBD ISOLATION VLV, using 3-HS-64-18. [64.2] CLOSE PRI CTMT N2 PURGE OUTBD ISOLATION VLV, using 3-HS-76-24.

NRC Scenario 7 Simulator Event Guide: Event 1 Normal: Inert the Primaiy Containment in accordance with 3-01-76 BOP Inerts Primary Containment lAW 3-01-76 [65] CLOSE the following valves (at the N2 storage tank):

N2 TANK A LIQUID SHUTOFF VLV, 0-SHV-076-0504(to remove the pressure build coils from service)
N2 TANK A PURGE VAPORIZER N2 INLET SHUTOFF VLV, O-SHV-076-0508
N2 TANK A LIQUID WITHDRAWAL SHUTOFF VALVE, 0-SHV-076-05 13

[66] VERIFY key lock PC PURGE DIV I and (II) RUN MODE BYPASS, 3-HS-64-24 and 3-HS-64-25, in NORMAL. [67] VERIFY CLOSED, STEAM CONTROL VALVE BYPASS, 0-76-534 (at storage tank). [68] CLOSE PCV-12-66 INLET, 0-12-802, (Unit 1 Turbine Bldg, El 565, T-4 M-Line). [69] WHEN purge piping to nitrogen vaporizer reaches approximately ambient temperature, as indicated by frost-free piping (weather permitting), THEN CLOSE the following valves:

N2 PURGE PCV OUTLET, 0-76-516 D
N2 PURGE PCV INLET, 0-76-515

[70] VERIFY CLOSED, N2 PURGE PCV BYPASS, 0-76-514. Driver Do not proceed past step 69

NRC Scenario 7 Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase. Direct Power increase using control rods JAW 3-GOl- 100-1 A. [34] CONTINIJE control rod withdrawals in combination with core flow changes, as recommended by Reactor Engineer, until approximately 30% Reactor power. Raise Power with Control Rods JAW 3-01-85, section 6.6. Control Rods: 18-43, 42-43, 42-19, and 18-19 from 36 to 48 ATC 10-35, 26-5 1, 34-5 1, 50-35, 50-27, 34-11, 26-11, and 10-27 from 24 to 36 26-3 5,_34-3 5,_34-27,_and 26-27_from_12_to_24 NOTES Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches. When in areas of high notch worth, single notch withdrawal should be used instead of continuous rod withdrawal. Information concerning high notch worth is identified by Reactor Engineering in Control Rod Coupling Integrity Check, 3-SR-3.1.3 .5A. When continuously withdrawing a control rod, the CRD Notch Override Switch is held in the Override position and the CRD Control Switch is held in the Rod Out Notch position.

When the control rod reaches two notches below its intended position, both switches should be released.
If the rod settles in a notch below to the intended position, the CRD Control Switch should be used to withdraw the rod to the intended position.

6.6.1 Initial Conditions Prior to Withdrawing Control Rods [2] VERIFY the following prior to control rod movement:

CRD POWER, 3-HS-85-46 in ON.
Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing.

6.6.2 Actions Required During and Following Control Rod Withdrawal ATC [4] OBSERVE the following during control rod repositioning:

Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.
Nuclear Instrumentation responds as control rods move through the core. (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize automatic REM Rod Block as follows:

STOP Control Rod withdrawal (if possible) prior to reaching any REM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].

NRC Scenario 7 Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods [6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REiNITIALIZE the RBM: [6.1] PLACE CRD POWER, 3-HS-85-46 in the OFF position to deselect the Control Rod. [6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position. 6.6.3 Control Rod Notch Withdrawal [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 3-XS-85-40. [2] OBSERVE the following for the selected control rod:

                                    . CRD ROD SELECT pushbutton is brightly ILLUMINATED.
                                    . White light on the Full Core Display ILLUMINATED.

Rod Out Permit light ILLUMINATED.

ATC [3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing. [4] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE. [5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.

NRC Scenario 7 Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods 6.6.4 Continuous Rod Withdrawal [1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 3-XS.85-40. [2] OBSERVE the following for the selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED.
White light on the Full Core Display ILLUMiNATED.
Rod Out Permit light ILLUMINATED.

[3] VERIFY Rod Worth Minimizer operable and LATCI{ED into correct ROD GROUP when the Rod Worth Minimizer is enforcing. [4] VERIFY Control Rod is being withdrawn to a position greater than three notches. ATC [5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: [5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE. [5.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH. [5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47 and CRD CONTROL SWITCH, 3-HS-85-48. [5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.

NRC Scenario 7 Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods 6.6.4 Continuous Rod Withdrawal (Continued) [5.5] WHEN control rod settles into the intended notch, THEN CHECK the following.

Four rod display digital readout and the full core display digital readout and background light remain illuminated.
CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does NOT alarm.

ATC [5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes. [6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A) 6.6.5 Return to Normal After Completion of Control Rod Withdrawal [1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN: ATC [1.1] PLACE CRD POWER,, 3-HS-85-46, in OFF. [1.21 PLACE CRD POWER, 3-HS-85-46, in ON. NRC NRC When satisfied with Reactivity Manipulation, CR]) Pump Trip DRIVER When directed by lea examiner, Trigger 2 CR1) Pump Trip 17

NRC Scenario 7 Simulator Event Guide: Event 3 Component: Trip of CRD pump 3A ATC Reports trip of CRD pump 3A and start CRD pump 3B, JAW 3-AOI-85-3 SRO Announces entry into 3-AOI-85-3, CRD System Failure. 4.1 Immediate Actions [1] IF operating CRD PUMP has failed AND the standby CR1) Pump is available, THEN PERFORM the following at Panel 3-9-5: (Otherwise N/A) [1.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, in MAN at minimum setting. [1.2] START associated standby CR1) Pump using one of the following:

CR1) PUMP 3B, using 3-HS-85-.2A

[1.31 ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-l 1, to establish the following conditions: ATC

CR1) CLG WTR HDR DP, 3-PDI-85-1 8A, approximately 20 psid.
CRD SYSTEM FLOW CONTROL, 3-FIC-85-l 1, between 40 and 65 gpm.

[1.4] BALANCE CR1) SYSTEM FLOW CONTROL, 3-FIC-85-11, and PLACE in AUTO or BALANCE. [2] IF Reactor Pressure is less than 900 PSIG and either of the following conditions exists:

In-service CRD Pump tripped and neither CRD Pump can be started, OR
Charging Water Pressure can NOT be restored and maintained above 940 PSIG, THEN PERFORM the following: N/A 4.2 Subsequent Actions

[1] IF the operating CRD Pump has tripped AND backup CRD Pump is NOT available, THEN PERFORM the following at Panel 3-9-5: N/A [2] IF Reactor Pressure is greater than or equal to 900 PSIG AND

Charging water Pressure can NQJ be restored and maintained above 940 PSIG within 20 minutes, AND
Two or more Scram accumulators are mop with associated control rod NOT fully ATC inserted, THEN PERFORM the following: N/A

[3] IF operating CRD Pump has NOT tripped, THEN PERFORM the following: N/A [4] IF CRD System hydraulic pressure is restored, THEN RESTORE Control Rod Drive System alignment. REFER TO 3-01-85. (Otherwise N/A) [5] NOTIFY Reactor Engineer of CRD system failure. If Dispatched to CRD Pump 3A, pump is extremely hot to touch. CRD Pump 3B oil levels in band, pump ready for start, conditions normal after the Driver start. CR1) 3A report breaker tripped on over current, Electrical Maint called. 12

NRC Scenario 7 Simulator Event Guide: Event 4 Component: Trip of RBCCW pump 3A NRC When ready, Trip of RBCCW pump 3A. Driver Driver Upon Lead examiner direction, initiate TriggerS: Trip of RBBCW pump 3A Responds to alarm 4C-12, RBCCW PUMP DISCH. HDR PRESS LOW BOP/ATC Report Trip of RBCCW Pump 3A. Automatic Action: Closes 3-FCV-70-48, non-essential loop, closed cooling water sectionalizing MOV. A. VERIFY 3-FCV-70-48 CLOSING/CLOSED. B. VERIFY RBCCW pumps A and B in service. BOP/ATC C. VERIFY RBCCW surge tank low level alarm is reset. D. DISPATCH personnel to check the following:

RBCCW surge tank level locally.
RBCCW pumps for proper operation.

E. REFER TO 3-AOI-70-1, for RBCCW System failure and 3-01-70, for starting spare pump. SRO Enters 3-A01-70-l, Loss of Reactor Building Closed Cooling Water. ATC Closes 3-FCV-70-48 and report the sectionalizing valve failed to close automatically Driver When the HS for 3-FCV-70-48 is taken to close veri1 TRG 6 is active to allow the valve to close SRO Contacts Maintenance Shift Manager to investigate failure of sectionalizing valve to close. BOP Dispatch Personnel to investigate RBCCW Pump 3A trip ATC 3-AOI-70-1 Loss of Reactor Building Closed Cooling Water 4.1 Immediate Actions 11] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A):

SECURE RWCU Pumps.
VERIFY RBCCW SECTIONALIZ1NG VLV, 3-FCV-70-48 CLOSED.

ATC Secures RWCU Pumps and Closes 3-FCV-70-48. 10

NRC Scenario 7 Simulator Event Guide: Event 4 Component: Trip of RBCCW pump 3A SRO 4.2 Subsequent Actions [1] IF Reactor is at power AND Drywell Cooling carmot be immediately restored, AND core flow is above 60%,THEN: N/A [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s): N/A Determines 1 and 2 are N/A ATC or [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW BOP pump, THEN PERFORM the following: [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). Calls RB AUO to inspect Pump and Control Bay AOU to inspect Breaker 480V SD BD 3A-6B Notifies SRO of 3A RBBCW pump breaker damage Driver When dispatched, report RBCCW Pump 3A breaker is tripped. There is also a smell of burnt wiring and charring on the breaker. 3A pump appears fine other than being tripped. SRO [4] IF unable to restart a tripped pump, THEN: PLACE Spare RBCCW Pump in service. REFER TO 3-01-70. Directs ATC or BOP to place the 3B RBCCW in service lAW 3-01-70 Contacts Maintenance to investigate 3A Pump Breaker 480V SD BD 3A-6B

NRC Scenario 7 Simulator Event Guide: Event 4 Component: Trip of RBCCW pump 3A BOP 3-01-70 Section 8.12 Placing a Single RBCCW Pump and a Single Heat Exchanger In Service [4.2] IF desired to return RBCCW PUMP 38 to service, THEN PERFORM the following: [4.2.1] VENT RBCCW Pump 3B as follows A. OPEN PUMP B CASE VT, 3-70-615B. B. WHEN valve has been open for 2 minutes, THEN CLOSE PUMP B CASE VT, 3-70-615B. [4.2.2] THROTTLE OPEN PUMP B DISCH, 3-70-501B (2 to 4 turns). [4.2.3] START RBCCW PUMP 3B using 3-HS-70-8A, on Panel 3-9-4. [4.2.4] FULLY OPEN PUMP B DISCH, 3-70-50 lB and OBSERVE the RBCCW Pump B suction pressure is greater than 18 psig on 3-PI-70-7.

Directs RB AOU to perform steps 4.2.1 and 4.2.2
Starts 3B RBCCW pump
Directs RB AOU to perform steps 4.2.4 Driver When directed, wait 3 minutes and report Steps 4.2.1 and 4.2.2 are complete.

When directed wait 2 minutes and report Step 4.2.4 is complete and the RBCCW Pump B suction pressure is greater than 18 psig on 3-PI-70-7. SRO 3-A0I-70-1 Continued [5] IF RBCCW flow was restored to two pump operation, THEN PERFORM the following (Otherwise N/A): [5.1] REOPEN RBCCW SECTIONALIZING VLV, 3-HS-70-48A [5.2] RESTORE the RWCU system to operation. (REFER TO 3-01-69) When 3B RBCCW pumps is RTS, directs the BOP to Open Sectionalizing Valve and Restore RWCU BOP Opens Sectionalizing Valve, 3-FCV-70-48 SRO References TR 3.4.1COOLANT CHEMISTRY. With RWCU out of service determines the TSR3 .4.1.1 is no longer met by continuous monitoring of reactor coolant conductivity. Reactor water continuous conductivity monitoring is no longer available and Chemistry is NRC required to_sample per TRM_Surveillance 3.4.1.1 SRO Calls Chemistry to commence sampling for reactor coolant conductivity 21

NRC Scenario 7 Simulator Event Guide: Event 4 Component: Trip of RBCCW pump 3A TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.1 1 NOTE Continuously Not required when there is no fuel in the reactor vessel. 4 hours when Monitor reactor coolant conductivity, the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or 5 OR 8 hours when the continuous conductivity monitor is inoperable and the reactor is in MODE 4 or 5 Drive r When directed by NRC insert TRG 10 for Loss of EECW C3 Pump through loss of 4KV SID Board 3EB

NRC Scenario 7 Simulator Event Guide: Event 5 Component: Loss of EECW C3 Pump through loss of 4KV S Board 3EB Recognizes loss of Shutdown Board 3EB due to lockout Responds lAW 3-ARP-9-23B Automatic Action: A. 4kV Bus Tie Bd. 3EB overcurrent and bus differential relays TRIP and LOCK OUT source breakers. B. Loss of the following:

1. CS pump 3C

2. RHRpump3C

3. RHRSW pump C3

4. Possible loss of 480V SHTDN BD 3A.

Operator Action: BOP A. VERIFY automatic action have occurred. B. HAVE 4kV Shtdn Bd. 3EB inspected for abnormal condition; relay target(s), smoke, burned paint, breaker position, etc. C. REFER TO O-GOI-300-2 and O-OI-57A to re-energize the board. D. REFER TO Tech Spec Sections 3.5.1, 3.5.2, 3.7.2, 3.8.7 and 3.8.8. Verifies automatic actions occurred Dispatches AOU to investigate 4kV Shtdn Bd. 3EB Notifies SRO of lockout on 4KV SD BD 3EB and loss of EECW pump C3,RHR pump3C, and CS pump 3C

       .          When Dispatched to investigate 4KV SD BD 3EB, wait 2 minutes and report 86 lockout on Driver the normal feeder breaker with a 51 G flagm BOP          Responds to alarms 23B-26 and 20A-35 23B-26: 4160V SD BD 3EB MOTOROL OR TRIP Overload or tripout on any one of the following: CS pump 3C,RHR pump 3C, RHRSW pump C3 A. CHECK control room for white light illuminated on effected equipment.

BOP B. DISPATCH personnel to check:

1. Relays at associated electrical bd.

2. Equipment for abnormal conditions, relay targets, smell, burned paint, breaker position, etc.

2

NRC Scenario 7 Simulator Event Guide: Event 5 Component: Loss of EECW C3 Pump through loss of 4KV S Board 3EB 9-20A window 35 EECW SOUTH HDR DG SECTION PRESS LOW B. CHECK Panel 3-9-3 for status of north header pump(s) breaker lights and pump motor amps normal. BOP C. NOTIFY Unit Supervisor, Ui and U2. D. START standby pump for affected header. REFER TO 0-01-6 7. 0-01-67 Section 8.11 Recovering from an EECW Pump Trip [1] VERIFY <25 minutes has elapsed since the EECW pump trip and header pressure >0 psig. [2] IF the north header pump has tripped, THEN [2.1] START desired RHRSW Pump using: [2.2] IF an RHRSW pump has been aligned to EECW north header and is desired to be placed in service to EECW, THEN START desired RHRSW Pump using one of the following: RHRSW PUMP A1(Cl), 0-HS-23-1A13(8A13) on Unit 3 BOP [4] For the EECW(RHRSW) pump(s) started, PERFORM the following:

VERIFY running Al pump current is less than 61 amps OR VERIFY running Cl pump current is less than 61 amps.
VERIFY locally, Pump breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged.
VERIFY Pump upper and lower motor bearing oil level is in the normal operating range.
NOTIFY Chemistry of running RHRSW (EECW) pump(s).

If starting RHRSW Pump Al (for EECW in place of A3): 8.1 Operation of RHRSW Pump Al (for EECW in place of A3) [1] To line up RHRSW Pump Al for EECW System operation, PERFORM the following: [1.5] UNLOCK and CLOSE RHRSW PMP Al & A2 CROSSTIE Valve, 0 504 at RHRSW A Room. BOP [1.6] UNLOCK and OPEN RHRSW PMP Al CROSSTIE TO EECW, 0-SHV-067-0088 at RHRSW A Room. [1.7] REQUEST a caution order be issued to tag RHRSW Pump Al and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Al pump should remain running to be operable for EECW. To align Al or Cl RHRSW pump wait 3 minutes and insert: Al:TRG 13 or Cl: TRG 12 Driver I provide filed report for whichever pump you are told to align 24

NRC Scenario 7 Simulator Event Guide: Event 5 Component: Loss of EECW C3 Pump through loss of 4KV S/D Board 3EB If starting RHRSW Pump Cl (for EECW in place of C3): 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3) [[1] To line up RHRSW Pump Cl for EECW System operation, PERFORM the following: [1.5] UNLOCK and CLOSE RHRSW PMP Cl & C2 CROSSTIE, 0-23-544 at RHRSW C Room. BOP [1.6] OPEN RHRSW PMP Cl CROSSTIE TO EECW, 0-FCV-67-49 using RHRSW PUMP Cl SUPPLY TO EECW, 0-HS-67-49A/3 [1.7] REQUEST a caution order be issued to tag RHRSW Pump Cl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Cl pump should remain running to be operable for EECW To align Al or Cl RHRSW pump wait 3 minutes and insert: Al:TRG 13 or Cl: TRG 12 Dri V er provide filed report for whichever pump you are told to align BOP

                ] Calls for alignment of Al or Cl RHRSW pump than starts Al or Cl RHRSW pump 25

NRC Scenario 7 Simulator Event Guide: Event 5 Component: Loss of EECW C3 Pump through loss of 4KV SID Board 3EB Dispatch Maintenance to 4KV SD BD 3EB SRO Evaluate Tech Spec 3.8.1,3.8.7,3.5.1, and 3.7.2 SRO T.S. 3.8.1 Conditions B and G apply B. One required Unit 3 DG B.1 Verify power availability 1 hour inoperable, from the offsite transmission network. AND Once per 8 hours thereafter AND (continued) B. (continued) B.2 Evaluate availability of 1 hour both temporary diesel generators (TDGs). AND AND Once per 12 hours thereafter B.3 Declare required 4 hours from feature(s), supported by discovery of the inoperable Unit 3 DG, Condition B inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable. redundant required feature(s) AND 8.4.1 Determine OPERABLE 24 hours Unit 3 DG(s) are not inoperable due to common cause failure. OR B.4.2 Perform SR 3.81.1 for 24 hours OPERABLE Unit 3 DG(s). AND (continued)

NRC Scenario 7 Simulator Event Guide: Event 5 Component: Loss of EECW C3 Pump through loss of 4KV S/D Board 3EB B. (continued) B.5 Restore Unit 3 DG to 7 days from OPERABLE status. discovery of unavailability of TDG(s) AND 24 hours from discovery of Condition B entry 6days concurrent with SRO unavailability of TDG(s) AND 14 days AND 21 days from discovery of failure to meet LCO NOTE Applicable when only one 4.16 kV shutdown board is affected. SRO G. One required offsite G.1 Declare the affected Immediately circuit inoperable. 4.16 kV shutdown board inoperable. AND One Unit 3 DG inoperable. 27

NRC Scenario 7 Simulator Event Guide: gh loss of 4KV S/D Board 3EB Event 5 Component: Loss of EECW C3 Pump throu SRO T.S. 3.8.7 Condition A applies A. OneUnit34i6kV

                                                    ------------   NOTE Sruitclown Board            Enter applicable Conditions and inoperable.                 Required Actions of Condition B, C, D, and G when Condition A results in no power source to a required 480 volt board.

Al Restore tfle Unit 3 5 days 4.16 KV Shutdown Board to OPERABLE status. ANP 12 days from discovery of failure to meet LCO AND A.2 Declare associated diesel Immediately generator inoperable. SRO T.S. 3.5.1 Condition A applies A.1 Restore low pressure 7 days A. One low pressure ECCS injection/spray subsystem E CCS injection/spray inoperable, subsystem(s) to OPERABLE status. OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable. 28

NRC Scenario 7 Simulator Event Guide: Board 3EB Event 5 Component: Loss of EECW C3 Pump through loss of 4KV S/D SRO T.S. 3.7.2 Condition A applies A. One required EECW Al Restore the required 7 days pump inoperable EECW pump to OPERABLE status. Driver When directed by the NRC, TRG 15 Condenser Vacuum Loss 29

NRC Scenario 7 . Simulator Event Guide: Event 6 Component: Condenser Vacuum Loss BOP Respond to alarm 53-14 OG HOLDUP LINE INLET FLOW HIGH ATC Report degrading condenser Vacuum. SRO Enter 3-AOI-47-3, Loss of Condenser Vacuum. 4.1 Immediate Actions None 4.2 Subsequent Actions [1] IF ANY EOI entry condition is met, THEN: ENTER the appropriate EOI(s). [2] IF unable to maintain hotwell pressure below -25 inches Hg as indicated on 3-XR-2-2, with Reactor power less than 30%, THEN TRIP the main turbine. [3] IF condenser vacuum is lost, THEN OPEN the HOTWELL SAMPLE DR TO FL DR. 3-DRV-043-1019 (565@ T-16 D Line) and CON DEM1N SAMPLE TO CRW VLV, 3-DRV-043-1061, (565 @T-12 G Line) to establish flow through the sample lines. [4] REDUCE reactor power in an attempt to maintain condenser vacuum. [51 VERIFY automatic actions. SRO Directs Trip of the Main Turbine At NRC direction insert TRG 20 Major LOOP Or D river If the Reactor is scrammed, then in conjunction with the turbine Trip insert TRG2O Major LOOP BOP Trips Main Turbine Enter 3 -AOI 1, Unplanned Turbine Trip Below 30% Reactor Power (Without Reactor SRO Scram) 30

NRC Scenario 7 Simulator Event Guide: Event 6 Component: Condenser Vacuum Loss BOP 4.1 Immediate Actions None 4.2 Subsequent Actions [1] VERIFY Automatic Actions listed in Section 3.0 have occurred. [2] PLACE TURNING GEAR OIL PUMP, 3-HS-47-l 1A, to START. [3] PLACE MOTOR SUCTION PUMP, 3-HS-47-12A, to START. [4] SET TURBINE OIL TEMPERATURE CONT, 3-TIC-24-75, to85°F. [5] OPEN the following drain valves by placing the following control switches to OPEN:

STOP VALVE 1 BEFORE SEAT DRyLy, 3-HS-6-100A
STOP VALVE 2 BEFORE SEAT DR VLV, 3-HS-6-1O1A
STOP VALVE 3 BEFORE SEAT DRVLV, 3-HS-6-102A
STOP VALVE 4 BEFORE SEAT DR VLV, 3-HS-6-103A
CONTROL VALVE 1 BEFORE SEAT DR VLV, 3-HS-6-104A
CONTROL VALVE 2 BEFORE SEAT DR VLV, 3-HS-6-105A
CONTROL VALVE 3 BEFORE SEAT DR VLV, 3-HS-6-106A
CONTROL VALVE 4 BEFORE SEAT DR VLV, 3-HS-6-107A
STEAM LINES TO HP TURBINE DR VLV, 3-HS-6-109A
LP STEAM LINES TO RFPTS DRAIN VLVS, 3-HS-6-1 1 1A
LP STEAM LINES TO RFPTS DRAIN VLVS,3-HS-6-l 12A

[6] WhEN Turbine speed lowers to 900 RPM, THEN START the Bearing Lift Pumps by placing TURBINE TURNING GEAR MOTOR, 3-HS-47-1OA, to ON. [7] VERIFY the exhaust hood sprays maintain exhaust hood temperature below 135°F. At NRC direction insert TRG 20 Major LOOP Or D river If the Reactor is scrammed, then in conjunction with the turbine Trip insert TRG2O Major LOOP 31

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power Event 8 Component: DG3EA fails to start Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A 4.1 Immediate Actions [1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators. [2] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators. [3] IF two EECW Pumps (not using the same EECW strainer) are not in service supply ing Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes) (Otherwise N/A). [4] PERFORM the following to ensure at least one train of Diesel Generator Room Fans are energized:

VERIFY 480V DSL Aux Board A or B energized.
VERIFY 480V DSL Aux Board 3EA or 3EB energized.

fl BOP Performs actions per O-A01057-1A. Recognizes D/G 3EA failed to start. Starts D/G 3EA r and manually ties to S/D Board 3EA. 32

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power Crew 4.2 Subsequent Actions [1] IF ANY EOI entry condition is met, THEN REFER TO the appropriate EOI(s). (Otherwise N/A) [2] IF any Unit is under a Station Blackout THEN ONLY PERFORM Attach ment 12 for that Unit: N/A [3] VERIFY automatic actions and PERFORM any that failed to occur. [4] REFER TO 1(2)(3)-AOI-78-1, FPC System Failure for a complete Loss of AC POWER, as necessary. NOT NECESSARY [5] WHEN EECW header pressure is restored above the reset pressure setpoin t (psig) for the valves listed below, THEN Common Unit 1 Unit 2 Unit 3 0-FCV-67-53 106 FCV-67-50 - 90 91 92 FCV-67-51 - 107 109 113 RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel 1-j LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. REFER TO the EECW to the RCW Crossties for Control Air & RBCCW section of 0-01-67. [6] START Control Air Compressors A, D and G as required and MONITOR system pressure. REFER TO 0-AOI-32-l. [6.1] IF an air compressor trips on high temperature, THEN (Otherwise N/A) NOTIFY Unit Supervisor for instructions. [7] REFER TO 1(2)(3)-AOI-32-2, Loss of Control Air, as necessary [8] PLACE RPS MG Sets A and B in service. REFER TO 1(2,3)-O1-99. Crew Calls for Control Air and EECW reset, calls for RPS Reset energized Driver When called to restore Control Air wait 3 minutes and insert TRG 26 When called to reset EECW wait 3 minutes and insert TRG23 than insert TRG24 When called to reset RPS wait 4 minutes insert TRG25 33

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power SRO Enters EOI SRO Enter 3 -EOI- 1, RPV Control. EOI- 1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure is Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate ?- NO, Main turbine Bypass valves will be unavailable when MS1Vs go closed IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? NOT at this time IF RPV water level cannot be determined? - NO Is any MSRV Cycling? - YES IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? - NO IF Drywell Control air becomes unavailable? NO. THEN crosstie CAD to Drywell Control Air, Appendix 8G. IF Boron injection is required? - NO SRO Direct a Pressure Band of 800 to 1000 psig, Appendix hA or 11C ATC/BOP Maintain directed pressure band, JAW Appendix 11 A or 11 C 34

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power ATC/BOP APP 1 1A

3. OPEN MSRVs using the following sequence to control RPV pressure as directed by SRO: a. 1 3-PCV-l-179 MN STM LINE A RELIEF VALVE.

b. 23-PCV-1-180 MN STM LINED RELIEF VALVE.

c. 3 3-PCV-l-4 MN STM LINE A RELIEF VALVE.

d. 4 3-PCV-l-31 MN STM LINE C RELIEF VALVE.

e. 5 3-PCV-1-23 MN STM LINE B RELIEF VALVE.

f. 63-PCV-1-42 MN STM LINE D RELIEF VALVE.

g. 73-PCV-l-30 MN STM LINE C RELIEF VALVE.

h. 83-PCV-1-19 MN STM LINE B RELIEF VALVE.

i. 9 3-PCV-i-S MN STM LINE A RELIEF VALVE.

j. 10 3-PCV-I-41 MN STM LINE D RELIEF VALVE.

k. 11 3-PCV-l-22 MN STM LINE B RELIEF VALVE.

1. 123-PCV-1-18 MN STM LiNE B RELIEF VALVE.

m. 13 3-PCV-1-34 MN STM LINE C RELIEF VALVE.

ATC/BOP APP 11 C

5. IF HPCI is in standby readiness, THEN START HPCI as follows:

a. VERIFY at least one SGTS Train in operation.

b. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.

c. PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP handswitch, in START.

d. PLACE 3-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, in START.

e. OPEN the following valves:

3-FCV-73-36, HPCI/RCIC CST TEST VLV
3-FCV-73-35, HPCI PUMP CST TEST VLV
3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

f. OPEN 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, to start HPCI Turbine.

g. VERIFY HPCI Auxiliaiy Oil Pump starts and turbine accelerates above 2400 rpm.

6. VERIFY proper HPCI minimum flow valve operation as follows:

a. IF HPCI flow is above 1200 gpm, THEN VERIFY CLOSED 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

b IF HPCI flow is below 600 gpm, THEN VERIFY OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

7. THROTTLE 3-FCV-73-35, HPCI PUMP CST TEST VLV, to control HPCI pump discharge pressure at or below 1100 psig.

8. ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller to control RPV pressure.

9. IF HPCI injection to the RPV becomes necessary, THEN ALIGN HPCI to the RPV as follows:

a. OPEN 3-FCV-73-44, HPCI PUMP INJECTION VALVE.

b. THROTTLE 3-FCV-73-35, HPCI PUMP CST TEST VLV, to control injection.

c. GO TO EOI Appendix-5D.

35

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Level. Veri+/- as required PCIS isolations group (1,2 and 3), ECCS and RCIC, ATC/BOP Verifies required PCIS isolations have been received SRO IF it has not been determined that the reactor will remain subcritical- NO RPV Water level cannot be determined? NO PC water level cannot be maintained below 105 feet OR Suppression Chamber pressu re cannot be maintained below 55 psig? NO - SRO Restore and maintain RPV Water Level between +2 and +51 inches with one or more of the following injection sources SRO Directs Level band of +2 to +51 with RCIC (App 5C) ATCIBOP Aligns RCIC per APP 5C

3. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TR[P/THROT VLV J 4. VERIFY 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 620 gpm.

5. OPEN the following valves:

3-FCV-71-39, RCIC PUMP iNJECTION VALVE
3-FCV-71-34, RCIC PUMP MIN FLOW VALVE
3-FCV-71-25, RCIC LUBE OIL COOLING WTRVLV.

6. PLACE 3-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.

7. OPEN 3-FCV-71-8, RCIC TURBINE STEAM SUPPLY VLV, to start RCIC Turbine.

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at620 gpm.

c. 3-FCV-71-40, RCIC TESTABLE CHECK VLV, opens by observing 3-ZI 40A, DISC POSITION, red light illuminated.

d. 3-FCV-71-34, RCIC PUMP MN FLOW VALVE, closes as flow rises above 120 gpm.

SRO EOI- 1 (Reactor Power) Monitor and Control Reactor Power If Reactor will remain subcritical without Boron under all condition YES then exit RC/Q and enter AOl- 100-1 SRO Exits RC/Q and directs ATC to continue performing actions of 3 -AOl- 100-1 36

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power ATC 3-AOI-lOO-1 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5A/S3B, on Panel 3-9-5. [2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWETCH, 3-HS-99-5A-Sl, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A) [3] Refuel Mode One Rod Permissive Light check [3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-Sl, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSiVE light, 3-XI-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) [4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-Sl, in SHUTDOWN. [5] REPORT the following status to the US:

Reactor Scram
Mode Switch is in Shutdown
All rods in or rods out
Reactor Water Level and trend (recovering or lowering).
Reactor pressure and trend MSIV position (Open or Closed)
Power level 4.2 Subsequent Actions

[1] ANNOUNCE Reactor SCRAM over PA system [3] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit. [3.1] DOWNRANGE IRMs as necessary to follow power as it lowers. [4] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5 37

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power rATcoP IF bulk suppression pool reaches 95F notifies SRO of EOI-2 entry condit [ ion SRO When bulk suppression pool reaches 95F enters EOI-2 and directs suppression Pool Cooling using APP 1 7A ATC/BOP 3 -EOI APPENDIX-i 7A, RHR System Operation Suppression Pool Coolin g

1. IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core coolin g,

THEN BYPASS LPCI injection valve open interlock AS NECESSAR Y:

PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.
PLACE 3-HS-74-i55B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. PLACE R.HR SYSTEM 1(11) in Suppression Pool Cooling as follow s:

a. VERIFY at least one RHRSW pump supplying each EECW header

b. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpmRHRSW flow:

3-FCV-23-34, RHR FIX 3A RHRSW OUTLET VLV
3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

d. IF Directed by SRO, THEN PLACE 3-XS-74-122(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRI) in MANUAL OVERRIDE.

e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLAC E

3-XS-74-i2i(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELEC T in SELECT.

f. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI 1NBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

g. OPEN 3-FC V-74-57(71), RHR SYS 1(11) SUPPR CHBRJPOOL ISOL VLV.

h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operati ng.

CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage. j -, ATC/BOP Aligns directed R}IR Pumps in Pool Cooling 38

NRC Scenario 7 Simulator Event Guide: Event 7 Major: Loss of Offsite Power ATC/BOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

i. THROTTLE 3-FCV-74-59(73), RFIR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-FI-7 4-50(64), RHR SYS 1(11) FLOW:

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE

k. MONITOR RUR Pump NPSH using Attachment 1.

1. NOTIFY Chemistry that R}{RSW is aligned to in-service RHR Heat Exchangers.

m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional_RHR and RHRSW pumps_in_service_using_Steps_2

                                                                                                .b through_2.1.

ATC/BOP Aligns directed R1{R Pumps in Pool Cooling 3.6.2.1 Suppression Pool Average Temperature ( LCO 3.6.2.1 Suppression pool average temperature shall be:

a. 95°F when any OPERABLE intermediate range monitor (IRM) channel is> 70/125 divisions of full scale on Range 7 and no testing that adds heat to the suppression pool is being performed;

b. 105°F when any OPERABLE IRM channel is> 70/125 divisions of full scale on Range 7 and testing that adds heat to the suppression pool is being performed; and

c. < 11OL]F when all OPERABLE IRM channels are < 70/125 divisions of SRO full scale on Range 7.

APPLICABILITY: MODES 1,2, and 3. Condition A: Suppression pool average temperature >95°F but 1 10°F. AND Any OPERABLE IRM channel >70/125 divisions of full scale on range 7. AND Not performing testing that adds heat to the suppression pool. Required Action A.1: Verify suppression pooi average temperature 110°F. Completion Time: Once per hour AND Required Action A.2: Restore suppression pool average temperature to S95°F. Completion Time: 24 hours 39

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA Event 10 Component: HPCI Steam Supply Valve fails to auto open. SRO Renter EO1- 1 on Reactor Water Level EOI- 1 Reactor Level Restore and Maintain RPV water level between +2 inches and +51 inches with any of the following: CRD Appendix-SB HPCI - App 5D If RPV water level CANNOT be restored and maintained between +2 inches and

                                                                                                       +51 inches Then Restore and Maintain RPV water level above -162 inches. Augm ent RPV water level control as necessary with any of the following:

SLC (boron tank) Appendix-7B SRO Direct Appenix-5B CRD, HPCI App SD, and Appendix-7B SLC ATC/BOP Aligns CRD,HPCI, and SLC If water level reaches -45 and HPCI already not in operation: ATC/BOP HPCI 73-16 fails to auto open, diagnoses failure and opens 73-16. Reports failure of HPCI to automatically start.

SRO Can RPV water level be maintained above -162 inches NO -

Enter Cl, Alternate Level Control ATC/BOP Report rising Drywell Temperature and Pressure SRO Re-Enters EOI-2 on Drywell Temperature and Pressure 40

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP Aligns CR1) and SLC JAW Appendix 5B and 7B ATC CRD Appendix 5B

2. IF BOTH of the following exist:

                                   . CR1) is NOT required for rod insertion, AND
                                   . Maximum injection flow is required, THEN LINE UP ALL available CR1) pumps to the RPV as follow s:

a. iF CR1) Pump 3A is available, THEN VERIFY RUNNING CRD Pump 3A.

b. IF CRD Pump 3B is available, THEN VERIFY RUNNING CRD Pump 3B

c. OPEN the following valves to increase CR1) flow to the RPV:

3-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV

                                   . 3-PCV-85-27, CR1) CLG WATER PRESS CONTROL VLV

3-FCV-85-50, CRD EXH RTN LINE SHUTOFF VALVE.

d. ADJUST 3-FIC-85-l 1, CR1) SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WFIILE maintaining 3-PI-85-13A, CRD ACCU M CHG WTR HDR PRESS, above 1450 psig, if possible.

( e. IF Additional flow is necessary to prevent or mitigate core damag DISPATCH personnel to fully open the following valves as require e, THEN d:

                                  . 3-THV-085-0527, PUMP DISCH THROTTLING (RB NE, el 565)
                                  . 3-BYV-085-0551, PUMP TEST BYPASS (RB NE, el 565).

ATC HPCI APP SD [4) VERIFY at least one SGTS train in operation. [5] VERIFY 3-FJC-73-33, HPC1 SYSTEM FLOW/CONTROL, contro ller is in one of the following configurations, as desired:

in AUTO and set for 5300 gpm for rapid injection
in AUTO and set for approximately 2500 gpm for slower injection
in MANUAL with output at approximately 50% for slower injectio n.

[6] IF high reactor water level trip logic is actuated, THEN [6.1] DEPRESS HPCI TURBINE TRIP RX LVL HIGH RESET pushbu tton. [6.2] CHECK HPCI TURBINE TRIP LVL HIGH amber light has extinguished. [7] PLACE FIPCI AUXILIARY OIL PUMP handswitch in START. [8] PLACE HPCI STEAM PACKING EXHAUSTER handswitch in STAR T. [9] OPEN the following valves:

3-FCV-73-30, HPCI PUMP MEN FLOW VALVE
3-FCV-73-44, FIPCI PUMP INJECTION VALVE.

[10] OPEN 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, to start HPCI Turbine. 41

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA [11] CHECK proper HPCI operation by observing the following: A. HPCI Turbine speed accelerates. B. 3-FCV-73-45, HPCI TESTABLE CHECK VLV, opens by observing 3-ZI-7 3-45A, DISC POSITION, red light illuminated. C. HPCI flow to RPV stabilizes and is controlled automatically at the setpoin t. (N/A if controller in manual). D. 3-FCV-73-30, FIPCI PUMP MIN FLOW VALVE, closes as flow exceed s approximately 1200 gpm. [12] ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, contro ller as necessary to control injection. [13] VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly. [14] WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE HPCI AUXIL IARY OIL PUMP handswitch in AUTO. ATC SLC APP 7B

2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step 10 to inject SLC Boron Tank to RPV.

10. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A13B, control switch in START-A or START-B (Panel 9-5).

11. CHECK SLC injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch.
Squib valves fire, as indicated by SQUIB VALVE A and B CONT INUITY blue lights extinguished,
SLC SQUIB VALVE CONTiNUITY LOST Annunciator in alarm (3-XA-55-SB, Window 20).
3-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressu re.
System flow, as indicated by 3 -IL-63 -11, SLC FLOW, red light illuminated,
SLC iNJECTION FLOW TO REACTOR Annunciator in alarm (3-XA SB, Window 14).

12. IF Proper system operation CANNOT be verified, ThEN RETURN to Step 10 and START other SLC pump.

42

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA SRO RE- Enter EOI-2 on Drywell Temperature and Pressure SRO PC/H Verif 11202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue SPIT MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX i7A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, continues Operate all available suppression pool cooling using only RHR Pumps NOT required to assure adequate core cooling by continuous injection Appendix-17A PC/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig Continue Continues Initiate suppression chamber sprays using only those pumps not required to assure adequate core cooling by continuous injection Appendix-i 7C Directs Appendix-i 7C When suppression pressure exceeds 12 psig continues Is suppression pool level below 19 feet Yes Is Drywell temperature within the safe area of Curve 5 Yes Shutdown Recirculation pumps and Drywell Blowers Initiate Drywell sprays using only those pumps not required to assure adequa te core cooling by continuous injection Appendix-i 7B Directs Appendix-i 7B 43

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA SRO RE- Enter EOI-2 on Drywell Temperature and Pressure SRO DW/T Monitor and control Drywell temperature below 1 60F using available Drywe ll cooling When Drywell Temperature CANNOT be maintained below 1 60F, NO Continues Operate all available DW cooling Before Drywell temperature rises to 200°F Continue continues Before Drywell temperature rises to 280°F Continue continues Is suppression pool level below 19 feetYes Is Drywell temperature within the safe area of Curve 5 Yes Shutdown Recirculation pumps and Drywell Blowers Initiate Drywell sprays using oniy those pumps not required to assure adequa te core cooling by continuous injection Appendix-i 7B Directs Appendix- 17 spin MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, continues Operate all available suppression pool cooling using only RI-IR Pumps NOT required to assure adequate core cooling by continuous injection Appendix-i 7A SPIL MOMTOR and CONTROL suppr p1 lvl between -1 in. and -6 in. (APPX 18) Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES 44

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF Adequate core cooling is assured, OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD NJ VLV BYPASS SEL in BYPASS.

3. IF Directed by SRO to spray the Suppression Chamber using Standb y Coolant Supply, THEN CONTINUE in this procedure at Step 7.

4. IF Directed by SRO to spray the Suppression Chamber using Fire Protec tion, THEN CONTINUE in this procedure at Step 8.

5. INITIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 3-XS-74-l21(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 3-FCV-74-53(67), RHR SYS 1(11) 1NBD INJECT VALVE, is OPEN THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) OUTB D

INJECT VALVE.

e. VERIFY OPERATING the desired RI{R System 1(11) pump(s) for Suppression Chamber Spray.

f. VERIFY OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBRIPOOL ISOL VLV.

g. OPEN 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.

45

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP 3-EOI APPENDIX-i 7C, RHR System Operation Suppression Chamber Sprays

h. IF RHR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k.

i. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALV E.

j. RAISE system flow by placing the second R1{R System 1(11) pump in service as necessary.

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RFIRSW pump supplying desired RHR Heat Exchanger(s).

m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR lix 3D RHRSW OUTLET VLV.

n. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

6. WHEN EITHER of the following exists:

Before Suppression Pool pressure drops below 0 psig, OR
Directed by SRO to stop Suppression Chamber Sprays, THEN STOP Suppression Chamber Sprays as follows:

a. CLOSE 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALV E.

b. VERIFY CLOSED 3-FCV-74-100, RHR SYS I U-2 DISCH XTJE

c. IF RHR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.

d. STOP RHR Pump(s) 3A (3B and 3D).

e. CLOSE 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHER/POOL ISOL VLV.

46

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATCIBOP 3-EOI APPENDLX-17B, RHR System Operation Drywell Sprays

1. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

2. IF Adequate core cooling is assured, OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD INJ VLV BYPASS SEL in BYPASS.

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.

5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.

6. INITIATE Drywell Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RFIR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI 1NBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RUR SYS 1(11) LPCI OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired System 1(11) R}IR pump(s) for Drywell Spray.

f. OPEN the following valves:

3-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV
3-FCV-74-61(75), RHR SYS 1(11) DW SPRAY 1NBD VLV.

47

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP 3 -EOI APPENDIX-i 7B, RHR System Operation Drywell Sprays

g. VERIFY CLOSEI) 3-FCV-74-7(30), RHR SYSTEM 1(11) MEN FLOW VALVE.

h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System 1(11) RHR Pump in service.

i. MOMTOR RFIR Pump NPSH using Attachment 2.

j. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

k. THROTTLE the following in-service RFIRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV
3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

1. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

7. WHEN EITHER of the following exists:

Before drywell pressure drops below 0 psig, OR
Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:

a. VERIFY CLOSED the following valves:

3-FCV-74-l00, RHR SYS I U-2 DISCH XTIE
3-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV
3-FCV-74-61(75), RI-IR SYS 1(11) DW SPRAY INBD VLV.

b. VERIFY OPEN 3-FCV-74-7(30), RHR SYSTEM 1(11) MEN FLOW VALVE.

c. I-F RHR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.

d. STOP RHR Pump(s) 3A (3B and 3D).

48

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA SRO Enters 3-C-i, Alternate Level Control IF RPV water level can be restored and maintained above -162 inches Then Exit NO Inhibit ADS Directs ADS inhibited ATC/BOP Inhibits ADS Restore and Maintain RPV level above -162 inches using any of the following: SRO Condensate and Feedwater NO, CRD Yes, RCIC YES, FIPCI YES, Condensate NO; LPCI system i YES, LPCI System 2 NO, CS System 1 YES, CS System 2 YES NOTE-RHR3C and CS 3C are unavailable due to loss of 4k SDBD 3EB Can 2 or more CNDS, LPCI or CS injection subsystems be lined up for injection YES-Commence preparing as many of the following alternate injection subsystems as possible for injection Determines which if any of the alternate injection systems can be aligned. Is any Condensate, LPCI or Core Spray injection subsystem lined up for injection with at SRO least one pump running YES Maximize injection with alternate injection subsystems listed Is ANY RPV injection source lined up with at least one pump running YES SRO Inject into the RPV with ANY available source 49

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP Injection Subsystems Lineup Core Spray System II

1. VERIFY OPEN the following valves:

3-FCV-75-30, CORE SPRAY PUMP 3B SUPPR POOL SUCT VLV 3-FCV-75-39, CORE SPRAY PUMP 3D SUPPR POOL SUCT VLV 3-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.

2. VERIFY CLOSED 3-FCV-75-50, CORE SPRAY SYS II TEST VALVE.

3. VERIFY CS Pump 3B and/or 3D RUNNING.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 3-FCV-75-53, CORE SPRAY SYS 11 1NBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

ATC/BOP Injection Subsystems Lineup Core Spray System I

1. VERIFY OPEN the following valves:

3-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCT VLV 3-FCV-75-1 1, CORE SPRAY PUMP 3C SUPPR POOL SUCT VLV 3-FCV-75-23, CORE SPRAY SYS I OUTBD iNJECT VALVE.

2. VERIFY CLOSED 3-FCV-75-22, CORE SPRAY SYS I TEST VALVE.

3. VERIFY CS Pump 3A and/or 3C RUNNING.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 3-FCV-75-25, CORE SPRAY SYS I 1NBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

ONLY 3A pump is available. 50

NRC Scenario 7 Simulator Event Guide: Event 9 Major: LOCA ATC/BOP Injection Subsystems Lineup RHR System I LPCI Mode [1] IF Adequate core cooling is assured AND It becomes necessary to bypass LPCI Injection Valve auto open signal to control injection, THEN PLACE 3-HS-74-155A, LPCI SYS-I OUTBD INJ VLV BYPASS SEL, in BYPASS. [2] VERIFY OPEN 3-FCV-74-7, RHR SYSTEM I M1N FLOW VLV. [3] VERIFY OPEN the following valves:

3-FCV-74-1, RHR PUMP 3A SUPPR POOL SUCT VLV
3-FCV-74-12, RHR PUMP 3C SUPPR POOL SUCT VLV.

[4] VERIFY CLOSED the following valves:

3-FCV-74-61, RHR SYS I DW SPRAY 1NBD VLV
3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV
3-FCV-74-57, RHR SYS I SUPPR CHBRIPOOL ISOLVLV
3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE
3-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV

[5] VERIFY RFIR Pump 3A and / or 3C running. [6] WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RI{R SYS I LPCI 114BD INJECT VALVE. [7] IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 3-FCV-68-79, RECLRC PUMP 3B DISCHARGE VALVE. [8] THROTTLE 3-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE, as necessary to control injection. [9] MOMTOR RHR Pump NPSH using Attachment 1. [10] PLACE RHRSW pumps in service as soon as possible on ANY RRR Heat Exchangers discharging to the RPV. [11] THROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:

3-FCV-23-34, RRR HX 3A RUR.SW OUTLET VLV
3-FCV-23-40, RHR FIX 3C RHRSW OUTLET VLV.

[12] NOTIFY Chemistry that RHRSW is aligned to in-service RHR heat exchangers. 51

NRC Scenario 7 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs: EECW A3 and Steam Packing Exhauster 3A are out of service. Operations/Maintenance for the Shift: Reactor Power is 23%, Unit startup is in progress lAW 3-GOI-100-1A. Inert the Primary Containment in accordance with 3-01-76 starting at step 47. Commence a power increase with Control Rods to 30% in accordance with Reactivity Control Plan. The Spare RBCCW pump is in service and the 3B RBCCW is ready to be returned to service. Unusual ConditionsLProblem Areas: NONE 52}}

ML14154A365: Difference between revisions

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ML14154A365: Difference between revisions

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