Text

Initial Exam 2012-301 Final Simulator Scenarios
	ML12318A362
Person / Time
Site:	Browns Ferry
Issue date:	11/13/2012
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/12-301, 50-260/12-301, 50-296/12-301 50-259/12-301, 50-260/12-301, 50-296/12-301
	Download: ML12318A362 (186)
	v • d • e

Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC - I Op-Test No.: 1205 Examiners:________________ Operators: SRO:_________________

ATC:______________

BOP:_______________

Initial Conditions: 95% power. DG 3A is OOS. Tech Spec 3.8.1 Condition B has been entered and offsite power availability was verified 5 minutes ago.

Turnover: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 per 3-SR-3.6.1.3.5 Section 7.6 and Section 7.7. Raise Power to 100% with Recirc Flow.

Event Malt. No. Event Event Description No. Type*

N-BOP I N/A 3-SR-3.6.1.3.5, perform stroke time testing on 2 PCIVs.

2 N/A Raise Power to 100% with flow.

3 cuO4 RWCULeakwithfailuretoAutoisolate.

TSSRO C-BOP Bus Duct Cooling Fan A Trip with failure of standby fan 4 eg 13 a C-SRO to auto start.

C-ATC . .

5 th03a RR Pump A Trip with power oscillations.

TS-SRO l-BOP Level 2 instrument failure (58D) causes HPCI to Auto 6 th30d TS-SRO initiate.

7 hpO8 M-ALL HPCI Steam Leak without Isolation.

8 tcO2 C No bypass valves with ATWS.

9 edl2a c Loss of 480V RMOV BD 3A.

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

Appendix D Scenario Outline Form ES-D-1 Events

1. BOP will perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-.14.

2. ATC raises power to 100% with Recirc Flow.

3. The crew will respond to RWCU alarms indicating a leak and RWCU valve 3-FCV I will fail to automatically isolate. The ATC will isolate RWCU and take actions lAW 2-AOI-64-2A. The SRO will enter EOl-3 on High Secondary Containment Temperatures, evaluate Tech Spec 3.6.1.3, and determine Condition A must be entered. Also, TRM Technical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. BOP will respond to Bus Duct Cooling 3A Fan trip and take action lAW with ARPs, start standby Bus Duct Cooling Fan 3B.

5. Reactor Recirculation 3A Pump will trip, crew will respond lAW 3-AOl-68-1A. The ATC will close 3A RR Pump Discharge Valve. Small power oscillations will develop. The ATC will insert control rods to dampen oscillations and exit region 2. The SRO will evaluate Technical Specification 3.4.1; Condition A is required.

6. Level transmitter 58D will fail to less than -45 inches. This failure will result in a HPCI auto initiation. The crew will respond lAW ARPs. Crew will verify that level is in normal band and secure HPCI. The SRO will evaluate Technical Specification 3.3.4.2 Condition A and B, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition A and B, and 3.8.1 Condition D and J.

7. Shortly after the HPCI initiation a steam leak will develop in the HPCI Room, HPCI will fail to automatically and manually isolate. When attempting to manually isolate the HPCI steam valve, 3-FCV-73-2, the 3A 480V RMOV Board will be lost due to an electrical fault.

8. The crew will enter EOl-3 and scram the Reactor. A small ATWS will occur on the scram; power, level and pressure will be controlled lAW EOl-.1. When the second MAX safe temperature is reached the crew will Emergency Depressurize.

9. Turbine Bypass Valves will not be available on the scram with an ATWS of 20 rods.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

All but 6 Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained.

Appendix 0 Scenario Outline Form ES-D-1 Critical Tasks Four CT#1 -With a primary system discharging into the secondary containment, before any area exceeds the maximum safe operating level, manually scram the reactor.

1. Safety Significance:

Places the primary system at a lower energy state and reduces driving head and flow of system discharging into the secondary containment.

2. Cues:

Procedural compliance,.

Secondary containment area temperatures, level, and radiation indication.

Field reports.

3. Measured by:

Observation - US enters EOI-1 at RC-1 and RO manually inserts a scram.

4. Feedback:

Rod insertion.

Reactor power decreasing.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks CT#2-With a primary system discharging into the secondary containment, when two or more areas are greater than their maximum safe operating values for the same parameter, RO initiates Emergency Depressurization as directed by US. BEN Operations Management expectation is that this action will be taken within 5 minutes.

1. Safety Significance:

Places the primary system in the lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces driving head and flow of system discharging into the secondary containment.

2. Cues:

Procedural compliance.

Secondary containment area temperatures, level, and radiation indication.

Field reports.

3. Measured by:

Observation US transitions to C-2 and RO opens at least 6 SRVs when two or more areas are greater than their maximum safe operating values for the same parameter.

4. Feedback:

RPV pressure trend.

SRV status indications.

CT#3 During an ATVVS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection 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 RHR, CS, C&FW, or HPCI injection prior to being less than MARFP (190 psig).

AND Observation Feedwater terminated and prevented until less than the MARFP (190 psig).

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms.

Injection system flow rates into RPV.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks CT#4 With RPV pressure <MARFP (190 psig), 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 (190 psig), 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.

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: NRC 1 8 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) 3 EOIs used: List (1-3) 2 EOl Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix 0 Scenario Outline Form ES-D-1 Scenario Tasks EVENT TASK NUMBER K/A RD SRO I Stroke Time Containment Isolation Valves RD U-064-SU-08 223002A2.08 2.7 3.1 SRQ S-000-AD-81 2 Raise Power with Recirc Flow RD U-068-NO-17 SROS-000-NO-138 2.1.23 4.3 4.4 3 RWCU Leak with Failure to Auto Isolate RO U-069-AL-1O 223002A2.03 3.0 3.3 SRO S-000-EM-12 4 Bus Duct Cooling Fan Trip RO U-047-AL-13 245000A2.05 3.6 3.8 5 Reactor Recirculation Pump Trip RD U-068-AB-1 202001A2.03 3.6 3.7 SRO S-068-AB-1 6 Level 2 Instrument Failure RD U-073-NO-5 216000A3.01 3.4 3.4 RD U-071-NO-5 SRO S-000-AD-27 7 HPCI Steam Leak RD U-073-.AL-06 295032EA2.03 3.8 4.0 SRO S-000-AB-03 SRO S-000-EM-12 SRO 5-000-EM-I 5

1 Page 8 of 44 Procedures Used/Referenced:

Procedure Number Procedure Procedure Title Revision 3-SR-3.6.l .3.5 Primary Containment Isolation Valve Operability Test Rev 26 3-GOI-100-12 Power Maneuvering Rev 37 3-01-68 Reactor Recirculation System Rev 81 3-ARP-9-3D, W17 RWCU Leak Detection Temperature High Rev 28 3-AOl-64-2A Group 3 RWCU Isolation Rev 9 TS 3.6.1.3 Primary Containment Isolation Valves Amd 212 TRM 3.4.1 Coolant Chemistry Rev 21 3-EOI-3 Secondary Containment Control Rev I I 3-ARP-9-7A, W31 Generator Bus Duct Fan Failure Rev 23 Trip/Core Flow Decrease OPRMs 3-AOI-68-IA Rev 6 Operable 3-ARP-9-3F, W29 Reactor Water Level Low Low HPCI/RCIC Initiation Rev 28 Anticipated Transient Without Scram Recirculation TS 3.3.4.2 Amd 213 Pump Trip (ATWS-RPT) Instrumentation Emergency Core Cooling System (ECCS) Amd 213 TS 3.3.5.1 Instrumentation Reactor Core Isolation Cooling (RCIC) System Amd 213 TS 3.3.5.2 Instrumentation TS 3.8.1 AC Sources Operating

- Amd 244 TS 3.5.1 ECCS Operating

- Amd 244 TS 3.5.3 RCIC System Amd 244 3-ARP-9-3F, W1O HPCI Leak Detection Temperature High Rev 28 3-EOI-1 RPV Control Rev 8 3-AOl-i 00-1 Reactor Scram Rev 55 3-EOI-App-3A SLC Injection Rev 1 3-EOI-3-C-5 Level/Power Control Rev 9 3-EOl-App-4 Prevention of Injection Rev 5 3-EOl-3-C-2 Emergency RPV Depressurization Rev 8 3-EOl-App-6A Injection Subsystems Lineup Condensate Rev 2 Injection Subsystems Lineup RHR System II LPCI Rev 3 3-EOl-App-6C 3-EOI-App-2 Defeating ARI Logic Trips Rev 4

1 Page 9 of 44 Procedures Used/Referenced:

Procedure Number Procedure Procedure Title Revision 3-EOl-App-1 F Manual Scram Rev 2 Insert Control Rods using Reactor Manual Control Rev 2 3-EOl-App-1 D System 3-EOl-2 Primary Containment Control Rev 8 3-EOl-App-1 7A RHR System Operation Suppression Pool Cooling Rev 6 EPIP-1 Emergency Classification Rev 47 EPIP-4 Site Area Emergency Rev 32

1 Page 10 of 44 Simulator Instructor 1C205 Bat nrcllO8-1O Bat nrcstick2O Bat nrcunstickl4

3A DG tagged out imf rd06r2635 dmf rd06r3435 br ypobkrl 838 fail_ccoil imf rd06r3035 dmf rd06r3423 mrf dgola open imf rd06r3435 dmf rd06r2631 ior zlo3hs2l l3ea9a[1] off imf rd06r2235 dmf rd06r3431 imf rd06r2623 dmf rd06r2639

RWCU seal leak no auto iso imf rd06r3423 dmf rd06r3439 imfcu06 imfrdO6r263l dmfrdO6r3O27 imf cuO4 (el 0)100 300 50 imf rd06r3431 dmf rd06r3427 imf rd06r2639 dmf rd06r2243

bus duct cooling fan trip imf rd06r3439 dmf rd06r2643 imf egl3a (e5 0) imf rd06r2227 dmf rd06r3043 imf rd06r2627 dmf rd06r3443

Recirc pump A trip, with pwr oscillations imf rd06r3027 dmf rdO6rl 843 imfth03a(elO 0) 1mfrd06r3427 dmfrdO6rl8l9 imfcr02a(elO 30) 10120 imfrdO6r2243 imfcr02b (elO 30) 10120 imfrdO6r2643 imf rd06r3043

HPCI Initiate due to failed lnstr imf rd06r3443 imf th30d (el 5 0) 35 60 84 imf rdO6rl 843 imfhpOl (e160) imfrdO6rlBl9

ATWSImajor HPCI Leak (have to manually modify fpO2 to close) mrffp02 (e20 0) close imfhpo9 imfhpo8(e200)88204 trg2l nrc2011732 trg 21 = imfedl2a ior ypovfcv733 (e20 0) fail_now bat nrcstick20 imftc02 (e20 0) 0 trg 26 = bat appOif trg27=batappo2 trg28batappo8ae trg 29 = bat nrcunstickl4

if crew anticipates ED, may have to raise severity hpOB -25%

1 Page 11 of 44 NRC Scenario I DESCRIPTIONIACTION Simulator Setup manual Reset to IC 205 Simulator Setup Load Batch bat nrcl 108-10 Simulator Setup manual Clearance DG 3A Simulator Setup Verify file loaded Procedures needed:

RCP required (95% 100% with flow)

RCP for Urgent Load Reduction

Marked up copy of 3-GOI-100-12

Copy of 3-SR-36.1 .3.5 P&Ls and Section 7.6 and 77

1 Page 12 of 44 Simulator Event Guide:

Event 1 Normal: Perform stroke time testing on 2 PCIVs (3-FCV-43-13 and 3-FCV-43-14)

SRO Directs BOP to perform 3-SR-3.6.1.3.5, Section 7.6.

BOP Performs 3-SR-3.6.1.3.5, Section 7.6.

3-SR-3.6.1 .3.5, Primary Containment Isolation Valve Operability Test, Section 7.6.

NOTES

1) Valves 3-FCV-43-13 and 3-FCV-43-14 are normally closed.

2) The following section is performed on Panel 3-9-3 unless otherwise noted.

7.6 3-FCV-43-13 Valve Stroke Timing

[1] RECORD the initial position of RX RECIRC SAMPLE INBD ISOLATION VLV,_3-FCV-43-13._OPEN_I_CLOSED_(Circle_one)

[2] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE REACTOR RECIRC SAMPLE INBD ISOL VLV, 3-HS-043-0013B OPEN position.

Driver When called 3-HS-043-001 38 is in the OPEN position.

[3] VERIFY OPEN 3-FCV-43-13 using RX RECIRC SAMPLE INBD ISOLATION VLV, 3-HS-43-13A.

[4] CLOSE and TIME 3-FCV-43-13, using RX RECIRC SAMPLE INBD ISOLATION VLV, 3-HS-43-13A, and RECORD the closure time below.

3-FCV-43-13 Closure Time (Seconds)

Normal Measured Maximum 0.6-1.6 5.0

[5] CHECK 3-FCV-43-1 3 closure time is less than or equal to the maximum closure time.

NA [6] IF the time recorded in step 7.6[4] is more than the maximum value listed, THEN_(Otherwise_NIA this_section.)

NA [7] IF the stroke time measured in step 7.6[4] is less than or equal to the maximum stroke time but outside the normal range, THEN (Otherwise NA this section)

BOP [8] RETURN 3-FCV-43-13, to the initial position recorded in Step 7.6[1], using RX RECIRC SAMPLE_INBD_ISOLATION VLV,_3-HS-43-13A.

[9] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE REACTOR RECIRC SAMPLE INBD ISOL VLV, 3-HS-043-0013B to the CLOSE position.

Driver When called 3-HS-043-0013B is in the CLOSE position.

1 Page 13 of 44 Simulator Event Guide:

Event 1 Normal: Perform stroke time testing on 2 PCIVs SRO Directs BOP to perform 3-SR-3.6.1 .3.5, Section 7.7.

BOP Performs 3-SR-3.6.1.3.5, Section 7.7.

7.7 3-FCV-43-14 Valve Stroke Timing

[1] RECORD the initial position of RX RECIRC SAMPLE OUTBD ISOLATION VLV,_3-FCV-43-14._OPEN_I_CLOSED_(Circle_one)

[2] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE REACTOR RECIRC OUTBD ISOLATION VLV, 3-HS-043-0014B to the OPEN position.

Driver When called 3-HS-043-0014B is in the OPEN position.

[3] VERIFY OPEN 3-FCV-43-14 using RX RECIRC SAMPLE OUTBD_ISOLATION_VLV,_3-HS-43-14A.

[4] CLOSE and TIME 3-FCV-43-14, using RX RECIRC SAMPLE OUTBD ISOLATION VLV, 3-HS-43-14A, and RECORD the closure time below.

3-FCV-43-14 Closure Time (Seconds)

Normal Measured Maximum 0.4-1.4 5.0

[5] CHECK 3-FCV-43-14 closure time is less than or equal to the maximum closure time.

NA [6] IF the time recorded in step 7.7[4] is more than the maximum value listed, THEN_(Otherwise_N/A this_section.)

NA [7] IF the stroke time measured in step 7.7[4] is less than or equal to the maximum stroke time but outside the normal range, THEN (Otherwise NA this section)

BOP [8] RETURN 3-FCV-43-14, to the initial position recorded in Step 7.7[1], using RX_RECIRC_SAMPLE_OUTBD_ISOLATION VLV,_3-HS-43-14A.

[9] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE REACTOR RECIRC OUTBD ISOLATION VLV, 3-HS-043-0014B to the CLOSE position.

Driver When called 3-HS-043-0014B is in the CLOSE position.

BOP Informs SRO that 3-FCV-43-14 and 3-FCV-43-14 have passed the SR Driver IF crew contacts the System Engineer or Duty Engineer because the operator mis-timed the valve stroke, THEN, as the System Engineer, inform the crew that a second valve stroke is authorized and they should document this as required by the SR.

1 Page 14 of 44 Simulator Event Guide:

Event 2 Reactivity: Raise Power to 100% with Recirc Flow SRO Notifies ODS of power increase.

Directs Power increase using Recirc Flow, per 3-GOt-I 00-12.

[21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:

RAISE power using control rods or core flow changes.

REFER_TO_3-SR-3.3.5(A)_and_3-01-68.

ATC Raise Power w/Recirc, lAW 3-01-68, Section 6.2

[1] 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-96-1 5A(1 5B).

ANDIOR

Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM),

3-HS-96-1 6A(1 6B).

[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-96-31 RAISE MEDIUM, 3-HS-96-32 Driver When directed by NRC, insert TRIGGER I for RWCU Leak with failure to Auto isolate

1 Page 15of44 Simulator Event Guide:

Event 3 Component: RWCU leak with failure to auto isolate Report alarm RWCU LEAK DETECTION TEMP HIGH (3-9-3D Window 17)

ATC A. IF this alarm is received in conjunction with RWCU ISOL LOGIC CHANNEL A TEMP HIGH [3-XA-55-5B, window 32] and RWCU ISOL LOGIC CHANNEL B TEMP HIGH [3-XA-55-5B, window 33], THEN EXIT this procedure and GO TO 3-ARP-9-5B. Otherwise, CONTINUE in this procedure.

Report alarms RWCU ISOL LOGIC CHANNEL A TEMP HIGH (3-9-5B Window 32) and RWCU ISOL LOGIC CHANNEL B TEMP HIGH (3-9-5B Window 33)

RWCU ISOL LOGIC CHANNEL A TEMP HIGH (3-9-5B Window 32)

A. VERIFY alarm by checking:

1. ATUs on Panel 3-9-83 and 3-9-85.

3. Area temperature indications on LEAK DETECTION SYSTEM TEMPERATURE,_3-TI-69-29,_on_Panel_3-9-21.

B. IF leak is suspected, THEN MANUALLY ISOLATE RWCU or if RWCU automatically isolates, REFER TO 3-AOI-64-2A.

C. IF TIS-69-835A(C) indicates greater than 131°F, THEN ENTER 3-EOI-3.

ATC Reports RWCU Valve 69-1 failed to isolate ATC Closes 69-1 to stop RWCU Leak SRO Directs Penetration Isolated or concurs with the closure of 69-1 SRO Enter EOI-3 and 3-AOI-64-2A ATC 3-AOI-64-2A Group 3 RWCU Isolation 4.1 Immediate Actions

[1] VERIFY automatic actions occur.

[2] PERFORM any automatic actions which failed to occur.

Driver Acknowledge Notifications, when dispatched to ATUs report high temperatures in RWCU_HX_room_and temperature_lowering.

SRO Contact work management and radiation protection.

1 Page 16 of 44 Simulator Event Guide:

Event 3 Component: RWCU leak with failure to auto isolate BOP 3-AOI-64-2A Group 3 RWCU Isolation 4.2 Subsequent Actions

[1] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s).

[21 CHECK the following to confirm high area temperature condition exists:

LEAK DETECTION SYSTEM TEMPERATURE, 3-TI-69-29 (Panel 3-9-2 1)

ATUs in Auxiliary Instrument Room

[3] IF isolation is caused by high area temperature, THEN DETERMINE if a line break exists by:

RWCU ARMs 3-RI-90-9A, 13A, and 14A

Visual Observation

Rx Zone Exhaust Rad Monitors 3-RE-90-142A, 142B, 143A, and 1 43B

[4] PERFORM necessary Heat Balance adjustments. REFER TO 3-01-69.

[51 CHECK the following monitors for a rise in activity:

AREA RADIATION, 3-RR-90-1, Points 9, 13, and 14 (Panel 3-9-2)

AIR PARTICULATE MONITOR CONSOLE, 3-MON-90-50, 3-RM-90-55 and 57 (Panel 3-9-2)

RB, TB, and Refuel Zone Exhaust Rad on CHEMISTRY CAM, MON ITOR CONTROLLER, 0-MON-90-36 1 (Panel 1-9-2)

[6] IF it has been determined that leakage is the cause of the isolation, THEN NOTIFY RADCON of RWCU status.

[7] NOTIFY Chemistry that RWCU has been removed from service for the following evaluations:

The need to begin sampling Reactor Water

The need to remove the Durability Monitor from service

[8] IF the isolation cannot be reset, THEN

[9] EVALUATE Technical Requirements Manual Section 3.4.1, Coolant Chemistry, for limiting condWons for ooeration.

Driver When directed by NRC, insert trigger 5 for Bus Duct Cooling Fan Trip

1 Page 17 of 44 Simulator Event Guide:

Event 3 Component: RWCU leak with failure to auto isolate SRO Evaluate Technical Specification 3.6.1.3 and determine Condition A required and TRM 3.4.1. Notifies Chemistry that continuous monitoring is no longer available and to commence sampling perTRM Surveillance 3.4.1.1. Appendix R compensatory_Measure_A_is_required 3.6 CONTAINMENT SYSTEMS 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 REQUIRED ACTION COMPLETION TIME A. NOTE A.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by use of at least main steam line Penetration flow paths one closed and de-activated automatic With two PCIVs valve, closed manual valve, blind AND or check valve One or more penetration with flow through the valve secured. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main flow paths with one PCIV steam line inoperable except due to MSIV leakage not within A.2 NOTE Once per 31 days i mi t s. for isolation devices Isolation devices in high radiation areas may be outside primary verified by use of containment administrative means.

Verify the affected penetration flow path is Prior to entering MODE 2 isolated. or 3 from MODE 4, if Primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside

. primary containment Mn,,u,,I3, FiaoPot,ectionRepoa1 PtJNT, NON T3UIT() , 1/2/2 PAGE 137 o0O4 Vol. 1 I TITI.E: AppendixR Safe Shutdown Proprana SECTION: 4 REV: 11 II SECTION III REQUED SAFE SEUTDOWN EQUIPMENT - UNIT 3 EQUIPMENT DESCRIPTION UNIT(S) APPENDIXR COMPENSATORY AREA, ZONES APTECTED PUNCTION MEASURES AppR SYSTEM 960-REACTOR WATER CLEANUP 3.FCV4)69-3JJ1 KWCU 1N131) SOUL [SUN VLV CLOSL FROM MUlL A 12 3 FCV 060 5002 RWCU OUTBD SUCT ISLN CLOSE PROM MCR A 11,12,13,14,15,16.21,22, VLV 2-3, 2-4, 2-5, 2-6, 3.1 3.2. 3-4,4, 2, 6,7,1.9, 10,11, 13, 14, 15,17, 11, 19, 20. 21,22, 23, 24,25-1. 23.2, II 25-3,26

1 Page 18 of 44 Simulator Event Guide:

Event 3 Component: RWCU leak with failure to auto isolate Enters EOl-3 on High Secondary Containment Temperature.

Secondary Containment Temperature Monitor and Control Secondary Containment Temperature.

Operate available ventilation, per Appendix 8F.

Answers YES to: Is Any Area Temp Above Max Normal?

Isolate all systems that are discharging into the area Verifies RWCU Isolated Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels.

Answers NO to: Is Any Area Radiation Level above Max Normal?

Secondary Containment Level Monitor and Control Secondary Containment Water Levels.

Answers NO to: Is Any Floor Drain Sump Above 66 inches?

AND Answers NO to: Is Any Area Water Level Above 2 inches?

Driver When directed by NRC, insert trigger 5 for Bus Duct Cooling Fan Trip

1 Page 19 of 44 Simulator Event Guide:

Event 4 Component: Bus Duct Cooling Fan A Trip with failure of standby fan to auto start

= BOP Responds to alarm GEN BUS DUCT FAN FAILURE 7A-31 A. VERIFY Main Bus Coohng Fans, 3-HS-262-IA or 1-HS-262-2A, indicates running on Panel 3-9-8 AND START GEN BUS DUCT HX FAN A(B) using 3-HS-262-1A(2A), on panel 3-9-8 to start the standby fan.

B. IF no Fans are operating and the Generator is tied to the grid and loaded to greater than the self cooled bus rating of 16,500 amps THEN, IMMEDIATELY INSERT a manual reactor scram, AND TRIP the Main Generator.

C. IF while executing this procedure, the Bus Duct Temperature is at or above the Temperature Excursion limit of 120°C, THEN IMMEDIATELY INSERT a manual reactor scram, AND TRIP the Main Generator.

D. DISPATCH personnel as necessary to check the following:

1. Main Bus Cooling Fan on elevation 586 to check fan condition.

2. Monitor Bus Duct temperature by available means including using a portable temperature monitor device locally at the 14 in-service thermostats.

REFER to Window 32, Figure 1.

3. 480V Unit Board 3A on elevation 586 to check breaker 5C closed.

4. 480V Unit Board 3B on elevation 604 to check breaker 5C closed.

E. VERIFY the system is operating in accordance with 3-01-47.

BOP Start Standby Bus Duct Cooling Fan B and dispatches personnel SRO Concurs with start or directs start of Bus Duct Cooling Fan B BOP Dispatch personnel to breaker and bus duct cooling fans Driver Breaker for bus duct cooling fan A is tripped, no abnormal indications apparent, if asked to reset breaker, breaker trips again, no problems noted at fans SRO Contact work management and maintenance manager.

Driver When directed by NRC, insert TRIGGER 10 for RR pump A trip

1 Page 20 of 44 Simulator Event Guide:

Event 5 Component: RR Pump A Trip with power oscillations ATC Respond to numerous alarm and Report Trip of RR Pump A SRO Enter 3-AOI-68-IA Recirc Pump Trip/Core Flow Decrease OPRMs Operable ATC [1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A),

[2] IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve.

ATC Closes_3A_Recirc_Pump_Discharge_Valve ATC [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.

[4] RAISE core flow to greater than 45%. REFER TO 3-01-68.

[5] INSERT control rods to exit regions if not already 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 3-01-68.

Report_in_Region_2_of Power to_Flow_Map SRO Directs Load Line reduction to <95%

ROD ROD NUMBER FROM TO NUMBER FROM TO 22-23 08 00 14-39 48 00 22-39 08 00 46-39 48 00 38-23 08 00 46-23 48 00 38-39 08 00 14-23 48 00 30-3 1 24 00 06-3 1 48 00 14-31 08 00 30-55 48 00 30-47 08 00 54-31 48 00 46-3 1 08 00 30-07 48 00 30-15 08 00 06-39 48 00 22-47 48 00 54-39 48 00 38-47 48 00 54-23 48 00 38-15 48 00 06-23 48 00 22-15 48 00 NOTES:

I For all rod moves to the full out position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 3-01-85.

2 Second-party verification by a second qualified member of the plant staff (i.e., RE, STA or UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal power 10%. A Peer Checker may initial when second party is NOT required.

3 Record the rod number and any problems encountered, as applicable.

Driver When directed by NRC, insert trigger 15 for Level 2 instrument failure (580). When alarm 3F-29 (RXWTR LVL LOW LOW HPCI/RCIC INIT) comes in, insert trigger 16 cause HPCI to initiate

I Page 21 of 44 Simulator Event Guide:

Event 5 Component: RR Pump A Trip with power oscillations ATC Insert Rods per Emergency shove sheets until <95% Load Line Driver When First Control rod is inserted delete Power Oscillations crO2a and crO2b SRO Evaluate Tech Spec for Single Loop Operation TS 3.4.1 Condition A Condition A Requirements of the LCO not met.

Required Action Al Satisfy the requirements of the LCO Completion Time 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months MODE Change not permitted until setpoint changes complete.

Driver When directed by NRC, insert TRIGGER 15 for Level 2 instrument failure (58D).

When alarm 3F-29 (RX WTR LVL LOW LOW HPCIIRCIC INIT) comes in, insert TRIGGER_16_cause_HPCI_to_initiate SRO Contact work management and maintenance manager.

I Page 22 of 44 Simulator Event Guide:

Event 6 Instrument: Level 2 instrument failure (58D) causes HPCI to Auto initiate Driver When directed by NRC, insert trigger 15 for Level 2 instrument failure (58D).

When alarm 3F-29 (RX WTR LVL LOW LOW HPCI/RCIC INIT) comes in, insert trigger 16 cause HPCI to initiate BOP Report alarm 3F-29 RX WTR LVL LOW LOW HPCI/RCIC INIT ATC/BOP A. CHECK RPV water level using multiple indications.

Report indicated water level on B instrument is less than -45 inches but other indicators are normal.

ATC/BOP Trips HPCI and locks out Aux Oil Pump using 3-HS-73-47A Driver If dispatched to investigate failure if Ll-3-58BB, wait 2 minutes and report, still investigating the cause, but it appears to be a problem with the Level Transmitter LT-3-58D.

Crew Determines that Level Transmitter LT-3-58D has failed therefore causing Level Indicator LI-3-58BB to indicate Low Driver If dispatched to investigate the cause of HPCI Auto-Initiation, acknowledge dispatch SRO Technical Specifications 3.3.4.2 Condition A Required Action A Completion Time 14 days 3.3.5.1 Condition A, B, and F for the following functions: 1 .a, 2.a, 3.a, 4.a, 5.a Required Action A Completion Time Immediately Required Action B.1 Completion Time 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action B.2 Completion Time 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Required Action B.3 Completion Time 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action F.1 Applies however Indefinite Completion time due to ADS Initiation Capability not lost in both Trip Systems Required Action F.2 Completion Time 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 3.3.5.2 Condition A and B for function 1 Required Action A Completion Time Immediately Required Action B.1 Applies however Indefinite Completion time due to RCIC Initiation Capability not lost Required_Action_B.2_Completion_Time_24_hours Diver When directed by NRC, insert trigger 20 for HPCI steam leak without isolation, Manually modify fpO2 to Close

1 Page 23 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A When trigger 20 is inserted a HPCI steam leak with a failure to auto-isolate will occur. Also, 3-FCV-73-3, HPCI outboard Isolation valve will fail on the same trigger

20. When the operator attempts to close 3-FCV-73-2, HPCI inboard Isolation NOTE NOTE valve, 480V RMOV Board 3A will be lost. Therefore, 3-FCV-73-2 will lose power and be de-energized in the not full close position. This will eventually require the crew to Emergency Depressurize on 2 Max Safe Secondary Containment Temperatures.

Recognize rising HPCI Room Temperatures and Radiation Levels.

3F-10 HPCI LEAK DETECTION TEMP HIGH A. CHECK HPCI temperature switches on LEAK DETECTION SYSTEM TEMPERATURE, 3-Tl-69-29 on Panel 3-9-21.

Crew B. IF high temperature is confirmed, THEN ENTER 3-EOl-3 Flowchart.

C. CHECK following on Panel 3-9-1 1 and NOTIFY RADCON if rising radiation levels are observed:

1. HPCI ROOM EL 519 RX BLDG radiation indicator, 3-Rl-90-24A.

2. RHR WEST ROOM EL 519 RX BLDG radiation indicator, 3-Rl-90-25A.

VERIFIES HPCI STEAM LINE INBD ISOL VLV, 3-FCV-73-2 AND ATCIBOP HPCI STEAM LINE OUTBD ISOL VLV, 3-FCV-73-3 CLOSE.

Attempts to isolate HPCI Steam Supply Valves.

Reports HPCI fails to isolate.

During attempts to isolate HPCI Steam Supply Valves, report a loss of 3A RMOV ATC/BOP Board._(Loop_1 RHR_and_Loop_1_Core_Spray_unavailable.)

Crew Contacts personnel to investigate loss of 3A RMOV Board.

Crew Dispatches personnel to transfer RPS A to alternate.

When requested, wait 4 minutes and place RPS A on alternate, in niO4 and Driver When requested to reset RPS ATU Gross Failures in rpO9 Crew PA announcement to evacuate the HPCI quad or Reactor Building SRO Contact work management and maintenance manager.

1 Page 24 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SRO Enters EOl-3 on Secondary Containment (Area Radiation or Temperature).

IF Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation SRO radiation levels are below 72 mr/hr, THEN Restart Reactor Zone and Refuel Zone Ventilation, per Appendix 8F. Defeat isolation interlocks if necessary, Appendix 8E.

If ventilation isolated and below 72 mr/hr, directs Operator to perform Appendix 8F.

Driver If requested, wait 3 minutes and report Appendix 8E complete, enter bat appO8e CT #1 Enters EOI-1 RPV Control and directs Reactor Scram before any temperature SRO exceeds MAX Safe.

CT #2 Stops at Stop sign When temperatures in two or more areas are above Max Safe, SRO Then Emergency Depressurization is required.

I Page 25 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT #1 SRO Enters EOI-1 RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe.

CT #2 SRO Stops at Stop sign When temperatures in two or more areas are above Max Safe, Then Emergency Depressurization is required.

SRO EOl-3 Secondary Containment (Temperature)

Monitor and Control Secondary Containment Temperature.

Is Any Area Temp Above Max Normal? YES -

Isolate all systems that are discharging into the area except systems required to:

. Be operated by EOls OR

. Suppress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment?

-YES Proceeds to the STOP sign Before any area temp rises to Max Safe (table 5)

Continue:

Crew Monitors for Max Safe Temperatures, reports when two areas are above MAX Safe (HPCI Room greater than 270°F and RHR System II Pump Room greater than 215°F)

SRO EOI-3 Secondary Containment (Level)

Monitor and Control Secondary Containment Water Levels.

Is Any Floor Drain Sump Above 66 inches? NO Is Any Area Water Level Above 2 inches? NO -

1 Page 26 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SRO EOI-3 Secondary Containment (Radiation)

Monitor and Control Secondary Containment Radiation Levels.

Is Any Area Radiation Level Max Normal? - YES Isolate all systems that are discharging into the area except systems required to:

. Be operated by EOIs OR

. Suppress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES -

Proceeds to the STOP sign Before any area radiation rises to Max Safe (table 4)

Continue

1 Page 27 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT #1 SRO Enters EOI-1 RPV Control and directs Reactor Scram before any temperature exceeds_MAX_Safe_based_on_EOl-3_step_SC/T-6.

CT #1 ATC Inserts Reactor Scram, Initiates One Channel of ARI and reports rods out SRO Enters EOI-1 from EOI-3 step SCIT-6 Verify Reactor Scram EOl-1 RCIP Monitor and Control RPV pressure Exits RC/P and enters C-2, Emergency RPV Depressurization, based on Override step RC/P-4.

EOI-1 RCIL Monitor and Control RPV Water Level Verify as Required:

PCIS Isolations (Groups 1,2 and 3)

ECCS

RCIC Exits RC/L and enters C-5, Level/Power Control, based on override RC/L-3 EOI-1 RCIQ Monitor and Control Reactor Power

Crew may determine Reactor Subcritical and exit RC/Q, as long as NO Boron has been injected, at any point during execution. If this is done Crew would enter AOl-I 00-1, Reactor Scram, based on override RC/Q-2.

(The following steps will be executed through AOl-100-1 if RCIQ exited)

Verify Reactor Mode Switch is in Shutdown

Initiate second channel of ARI

Verify Recirc Pump Runback (Pump speed 480rpm or less)

Answers No to is Reactor Power above 5% or Unknown (The Following steps NIA if RCIQ exited)

Before Suppression Pool Temperature rises to 11 OF, determines Boron Injection is Required.

Initiates SLC per Appendix 3A

1 Page 28 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SRO EOI-1 RCIQ (cont)

Inhibit ADS Verify RWCU System Isolation Answers Yes to is SLC injecting into the RPV Stops at step RCIQ-18 until SLC has injected into the RPV to a tank level of 43%, then exits RCIQ and enters AOl-I 00-1 Trips the SLC pump when SLC tank level drops to 0%

ATC Initiates Second Channel of ARI and reports no rod movement.

Verifies Recirc Pump at 480 rpm or less.

Reports Reactor Power less than 5% during Scram Report Should insert IRMs to determine if Reactor is subcritical BOP/ATC Verify and Report PCIS Isolations, ECCS and RCIC If directed, Initiate SLC per Appendix 3A, Inhibit ADS, and Verify RWCU System Isolation (These steps N/A if RC/Q exited and AOl-i 00-1 entered)

ATC/BOP Performs actions of 3-AOl-i 00-1 Scram Hardcards ATC Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS

[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5AIS3A and 3-HS 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 (Otherwise 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. (Otherwise_N/A)

1 Page 29 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A ATC [4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, 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 2.0 SUBSEQUENT ACTIONS:

[1] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise N/A):

[1.1] INITIATE ARI by Arming and Depressing BOTH of the following:

ARI Manual Initiate, 3-HS-68-1 I 9A

ARI Manual Initiate, 3-HS-68-1 1 9B

[1.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 3-9-4.

[1.3] REPORT_ATWS_Actions_Complete_and_power_level.

[2] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit.

[3] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5.

[4] MONITOR and CONTROL Reactor Water Level between +2 and +51 ,

or as directed by US, using RFP/RFPT.

[5] RETURN to body of procedure at step 4.2[5] AND CONTINUE with actions as required.

BOP Reactor Scram BOP Unit Operator Hard Card 1.0 SUBSEQUENT ACTIONS: PANELS 3-9-7 & 3-9-8

[I] At 50 MWe, or as directed by the Unit Supervisor, VERIFY TRIPPED the Main Turbine as follows:

[1.1] DEPRESS the TRIP pushbutton, 3-HS-47-67D on Panel 3-9-7.

[1.2] VERIFY OPEN GENERATOR PCB 234.

1 Page 30 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A BOP [1.3] RESET disagreement white light as follows: PLACE GENERATOR PCB 234, 3-HS-35-234, to NORMAL AFTER TRIP (NAT).

[1.4] VERIFY TRIPPED GENERATOR EXCITER FIELD BREAKER using 3-HS-57-24.

[1.5] PLACE GENERATOR EXCITER FIELD BREAKER, 3-HS-57-24 in the NORMAL AFTER TRIP (NAT) position.

[2] ANNOUNCE Reactor SCRAM over PA system.

2.0 SUBSEQUENT ACTIONS: PANELS 3-9-3

[1] MONITOR and CONTROL RPV pressure to keep below 1073 psig and stable, or as directed by US.

[1 .1] IF RPV pressure is lowering rapidly, THEN CLOSE the MSIVs.

(Otherwise N/A)

[1.2] IF MSRVs are cycling and bypass valves are available, THEN MANUALLY OPEN MSRVs on Panel 3-9-3 until RPV pressure is below 965 psig. (Otherwise N/A)

[1.3] IF MSRVs are cycling and bypass valves are NOT available, THEN MANUALLY OPEN MSRVs on Panel 3-9-3 until RPV pressure is controlled between 800 and 1000 psig. (Otherwise N/A)

[2] IF any PCIS isolation signal is received, THEN VERIFY PCIS isolations using any of the following: (Otherwise N/A)

Containment Isolation Status System on Panel 3-9-4

PCIS Mimic and individual control switch indications

ICS

3-01-64

[3] IF HPCI and/or RCIC are in service and injecting to the vessel, THEN MONITOR and CONTROL Reactor Water Level as necessary.

(Otherwise N/A)

BOP/ATC If directed, Initiate SLC per Appendix 3A, Inhibit ADS, and Verify RWCU System Isolation (These steps N/A if RCIQ exited and AOl-i 00-1 entered)

1 Page 31 of44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A BOP/ATC Appendix 3A

1. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A13B, control switch in START PUMP 3A or START PUMP 3B position.

2. CHECK SLC System for 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 on Panel 3-9-5 (3-XA-55-5B, Window 20).

3-Pl-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

System flow, as indicated by 3-IL-63-1 1, SLC FLOW, red light illuminated on Panel 3-9-5,

SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 3-9-5 (3-)(A-55-5B, Window 14).

3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.

4. VERIFY RWCU isolation by observing the following:

RWCU Pumps 3A and 3B tripped

3-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE closed

3-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed.

3-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

7. MONITOR 3-LI-63-IA, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

1 Page 32 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SRO Enters C-5 from EOI-1 step RCIL-3 Override Step C5-1, states that IF Emergency Depressurization is required, THEN continue at step C5-1 9, however, if the SRO has not determined that ED is required at this time then he will continue at step C5-2 (below)

Inhibit ADS Answers Yes to is any Main Steam Line Open Bypass the following Isolation Interlocks:

. MSIV Low Low Low RPV Water Level (APPX (8A)

. RB Ventilation Low RPV Water Level (APPX 8E)

Crosstie_CAD_to_DW_Control_Air,_if_necessary_(APPX_8G)_(Step_N!A)

Driver When requested for appendix 8A and 8E wait 4 minutes and insert trigger 28 for bat appo8ae and report complete SRO Answers No to is Reactor Power Above 5% or Unknown Establishes Reactor Water Level Band between -180 and +51 inches utilizing available injection sources listed on step C5-15.

Upon determination that Emergency Depressurization is required continues at step C5-19 and enters C-2 by direction of EOI-3 step SCIT-9 and from EOl-1 step RC/P CT#3 4 and directs Crew to Stop and Prevent all Injection Sources to the RPV Except from RCIC, CRD and SLC per step C5-20, in accordance with Appendix 4.

BOPIATC Inhibits ADS (if not already done per Appendix 3A)

If directed, dispatches personnel to perform Appendices 8A and 8E.

Maintains Reactor Water Level until directed to Stop and Prevent per Appendix 4.

When directed performs Appendix 4 to Stop and Prevent all Injection Sources to CT#3 the_RPV Except from_RCIC,_CRD_and_SLC Driver May need to adjust the HPCI Steam Leak to drive crew into 2 max safes

1 Page 33 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT#3 BOP/ATC Appendix 4

1. PREVENT injection from HPCI by performing the following:

a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURBINE TRIP push-button.

b. WHEN HPCI Turbine is at zero speed, THEN PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 3-HS-73-18A, HPCI TURBINE TRIP push-button.

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:

NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.

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 INJ 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:

NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

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

OR

1 Page 34 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT#3 BOP/ATC Appendix 4 (continued)

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 FEEDWATER 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.

b. LOWER RFPT 3A(3B)(3C) speed to minimum setting (approximately 600 rpm) using ANY of the following methods on Panel 3-9-5:

Using 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL AND individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO, OR

Using individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)

SPEED CONTROL in MANUAL, OR

Using individual 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C)

SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR.

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

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

3-FCV-3-12, REP 3B DISCHARGE VALVE

3-ECV-3-5, RFP 3C DISCHARGE VALVE

3-LCV-3-53, RFW START-UP LEVEL CONTROL

d. TRIP REPTs as necessary to prevent injection by DEPRESSING the following push-buttons:

3-HS-3-1 25A, RFPT 3A TRIP

3-HS-3-151A, RFPT 3B TRIP

3-HS-3-176A, RFPT 3C TRIP.

1 Page 35 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT #2 SRO Determines Emergency Depressurization is required and enters C-2 Answers No to will the reactor remain subcritical under all conditions. Waits until he receives the report that Appendix 4 is complete.

Answers Yes to is Suppression Pool Level above 5.5 ft Directs All ADS Valves opened Answers Yes to can Six ADS Valves be opened Stops execution of C-2 until:

The Reactor will remain Subcritical without Boron under all conditions OR

SLC has injected into the RPV to a tank level of 43%

OR

The Reactor is Subcritical and No Boron has been injected into the RPV Stops execution of execution of C-2 until Shutdown Cooling RPV Pressure Interlocks are clear Maintain RPV in Cold Shutdown per Appendix 17D CT#2 BOP/ATC Reports when Appendix 4 is complete Reports Suppression Pool Level in Feet when Directed Opens and Verifies Open ALL ADS Valves when directed SRO Upon commencement of Emergency Depressurization Continues in C-5 at step C5-21 Answers Yes to are at least 2 MSRVs open per C-2, Emergency RPV Depressurization Stops until RPV Pressure is below MARFP (l9Opsig with 6 MSRVs open)

Then continues Directs crew to Start and Slowly raise RPV Injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSH limits and CT#4 Suppression_Pool_level_per Appendix_6A_or_per Appendix_6C

1 Page 36 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT#4 BOP/ATC Appendix6A

1. VERIFY CLOSED the following Feedwater heater return valves:

3-FCV-3-71, HP HTR 3A1 LONG CYCLE TO CNDR

3-FCV-3-72, HP HTR 3B1 LONG CYCLE TO CNDR

3-FCV-3-73, HP HTR 3C1 LONG CYCLE TO CNDR

2. VERIFY CLOSED the following RFP discharge valves:

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

3-FCV-3-12, REP 3B DISCHARGE VALVE

3-ECV-3-5, REP 3C DISCHARGE VALVE

3. VERIFY OPEN the following drain cooler inlet valves:

3-ECV-2-72, DRAIN COOLER 3A5 CNDS INLET ISOL VLV

3-ECV-284, DRAIN COOLER 3B5 CNDS INLET ISOL VLV

3-ECV-2-96, DRAIN COOLER 3C5 CNDS INLET ISOL VLV

4. VERIFY OPEN the following heater outlet valves:

3-FCV-2-124, LP HEATER 3A3 CNDS OUTL ISOL VLV

3-FCV-2-125, LP HEATER 3B3 CNDS OUTL ISOL VLV

3-ECV-2-126, LP HEATER 3C3 CNDS OUTL ISOL VLV

5. VERIFY OPEN the following heater isolation valves:

3-FCV-3-38, HP HTR 3A2 FW INLET ISOL VLV

3-FCV-3-31, HP HTR 3B2 FW INLET ISOL VLV

3-ECV-3-24, HP HTR 3C2 EW INLET ISOL VLV

3-FCV-3-75, HP HTR 3A1 FW OUTLET ISOL VLV

3-ECV-3-76, HP HTR 3B1 EW OUTLET ISOL VLV

3-ECV-3-77, HP HTR 3C1 EW OUTLET ISOL VLV

6. VERIFY OPEN the following REP suction valves:

3-ECV-2-83, REP 3A SUCTION VALVE

3-FCV-2-95, REP 3B SUCTION VALVE

3-ECV-2-108, REP 3C SUCTION VALVE

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

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

10. VERIFY REW flow to RPV.

1 Page 37 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT#4 BOP/ATC 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, THEN PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ 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 SUCTVLV

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. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN 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.

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

1 Page 38 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A CT#4 BOP/ATC Starts and Slowly raises RPV Injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSH limits and Suppression Pool level per Appendix 6A or per Appendix 6C SRO EOl-1 RCIQ steps RCIQ-20 and RCIQ-21 Reset ARI Defeat ARI Logic Trips if necessary (APPX 2) (This step is N/A, however, crew may choose to perform this step)

Insert Control Rods by performing Appendix I F and 1 D Appendix I F: Scram Valves Opened but SDV is Full

1) Reset Scram Defeat RPS Logic Trips if necessary

2) Drain SDV

3) Recharge Accumulators

4) Initiate Reactor Scram Appendix ID: Manual Control Rod Insertion Method I)__Drive_Control_Rods._Bypass_RWM_if necessary BOP/ATC Dispatch personnel to perform Appendix 2(N/A) and outside portions of Appendix IF.

Dispatch personnel to close 3-FCV-85-586 (while awaiting completion of Appendix 1 F)

Drive Rods per Appendix 1 D while waiting for completion of Appendix 1 F

1 Page 39 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A ATC 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 CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.

7. CONTINUE to perform Steps 1 through 6 UNTIL ANY of the following exists:

ALL control rods are fully inserted, OR

NO inward movement of control rods is observed, OR

SRO directs otherwise.

Driver When directed to perform Appendix 2 and outside portions of Appendix I F wait 3 minutes. Insert Triggers 26, 27, and 29 then report completion.

If directed to close 3-FCV-85-586 wait 3 minutes then insert mn rdO6 close. Then report completion.

If/When directed to re-open 3-FCV-85-586 wait 3 minutes then insert mn rdO6 open. Then report completion.

I Page 40 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A

- ATC Appendix ID

1. VERIFY at least one CRD pump in service.

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

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.

6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SOV (RB_NE,_El_565 ft).

ATC Continue performance of Appendix IF and ID until all rods inserted OR Until EOI-I RC/Q is exited due to Reactor determined to be Subcritical at which point continue to insert rods per 3-AOl-I 00-I and 3-01-85

1 Page 41 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SRO Executes all legs of EOl-2 concurrently EOl-2 DWIT Monitor and control Drywell Temperature below 160F using available Drywell Cooling Answers Yes to can Drywell Temperature be maintained below 160F EOl-2 PCIP Monitor and control Primary Containment pressure below 2.4 psig using the vent system (APPX 12) as necessary Answers Yes to can Primary Containment pressure be maintained below 2.4 psig EOl-2 PCIH Monitor and control Drywell and Suppression Chamber

Hydrogen at or below 2.4%

AND

Oxygen at or below 3.3%

Using the Nitrogen Makeup System (APPX 14A)

EOl-2 SPIT Monitor and control Suppression Pool temperature below 95F using available Suppression Pool Cooling (APPX 17A) as necessary Answers No to can Suppression Pool temperature be maintained below 95F (This is assuming Emergency Depressurization is complete and Reactor Water Level has been restored, if Emergency Depressurization has not been conducted yet, the answer will be Yes. If Reactor Water Level has not been restored yet, after Emergency Depressurization, this is not a priority.)

Directs Line up of all available Suppression Pool Cooling using only RHR pumps not required to assure adequate core cooling by continuous injection (APPX 17A) (After Emergency Depressurization complete and Reactor Water level restored)

BOP Performs Appendix I 7A to place Suppression Pool cooling in service after Emerqencv Depressurization and restoration of Reactor Water level.

1 Page 42 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A SOP Appendix 17A

1. If Adequate core cooling is assured, OR Directed to cool the Suppression Pool 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.

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 RHRSW pump supplying desired RHR Heat Exchanger(s).

c. 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, 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 bySRO, Then PLACE 3-XS-74-1 22(1 30), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD 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 INBD 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 CHBR/POOL ISOL VLV.

1 Page 43 of 44 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak without Isolation, Loss of RMOV Board 3A

= BOP Appendix 17A (cont)

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

I. THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicatedon 3-Fl-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 RHR Pump NPSH using Attachment 1.

I. NOTIFY Chemistry that RHRSW 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.

SRO Emergency Plan Classification is 3.1-S.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

All but six Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained

SHIFT TURNOVER SHEET Equipment Out of ServiceiLCOs:

DG 3A is Out of Service. Tech Spec 3.8.1 Condition B has been entered and offsite power availability was verified 5 minutes ago.

OperationslMaintenance for the Shift:

Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 per 3-SR-.3.6.1.3.5 Section 7.6 and 7.7. Raise Power to 100% with Recirc Flow.

Unit I and 2 are 100% Power Unusual ConditionslProblem Areas:

None

Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 3

- Op-Test No.: 1205 Examiners:___________________ Operators: SRO:_______

ATC:______

BOP:

Initial Conditions: Unit 3 is at 100% power. Unit 1 and 2 are at 90% power. CRD Pump A is out of service.

Turnover: Complete Weekly EHC Pump Test per 3-Ol-47A section 6.2. Lower power to 90% with Recirc Flow.

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor FINAL

Appendix D Scenario Outline Form ES-D-1 Events

1. The BOP Operator will complete the weekly EHC Pumps test per 3-OI-47A section 6.2.

2. ATC will reduce Reactor Power to 90% RTP with Recirc flow as an immediate action of 3-AOl-I-I, Relief Valve Stuck Open. ADS SRV 1-41 will fail open.

3. ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 3-AOl-I-I actions to close SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition E.

4. After the NRC is satisfied with the power reduction, the VFD Cooling Water Pump for the B 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.

5. After VFD cooling water restored, 3ED 4KV Shutdown Board will lose power and the 3D Diesel Generator will fail to automatically tie to the Shutdown Board. The BOP will manually tie the Diesel to the board. SRO will refer to Tech Specs and determine TS 3.8.1 condition B and G, and TS 3.8.7.A.

6. High vibrations and low oil pressure alarms will come in on RFPT C, the RFPT will fail to trip and the ATC will have to trip in order to avoid extensive pump and equipment damage. The ATC will also have to lower power an additional 5% so the remaining RFPT5 are below their limit of 5050 RPM.

7. Once the plant is stable, the A Feedwater line will break in the Steam Tunnel, a scram will be inserted due to loss of feedwater and lowering level, EOI-1 will be entered. HPCI will be locked out to prevent feeding the leak. An ATWS will occur on the Scram; the SRO will enter C-5 and perform ATWS recovery actions to insert all control rods. Reactor level will decrease to TAF and Emergency Depressurization will be initiated per C-2.

8. MSIV D INBD valve will fail to isolate, crew will isolate Inboard MSIV D. D/G A will fail to automatically start, the crew will manually start D/G A Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

Emergency Depressurization complete Reactor Level is restored and maintained.

All Control Rods inserted.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks - Five CT#1 During power operations, with a stuck open Safety Relief Valve take actions to close the Safety Relief Valve OR Scram the Reactor prior to the suppression pool temperature reaching lICE.

I. Safety Significance:

Prevent a violation of the facility license condition (T.S. 3.6.2.1).

2. Cues:

Procedural compliance.

Suppression Pool temperature trend.

3. Measured by:

With rising Suppression Pool Temperatures, the REACTOR MODE SWITCH is placed in SHUTDOWN, OR The Safety Relief Valve is closed, prior to exceeding 1100 in the Suppression Pool.

4. Feedback:

Reactor Power trend.

Control Rod indication.

Suppression Pool temperature trend.

CT#2- 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 annunciator status

Appendix D Scenario Outline Form ES-D-1 Critical Tasks CT#3 -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 110°F) 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 EOl-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOl placekeeping action) before exceeding 110 degrees F in the Suppression Pool.

AND RO places SLC A I B Pump control switch in ON, when directed by US.

4. Feedback:

Reactor Power trend Control Rod indications SLC tank level CT#4 During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US (for these conditions MARFP is 190 psig)

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.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks CT#5 With RPV pressure <MARFP (190 psig), slowly increase and control injection into RPVto 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.

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: NRC 3 8 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3)

I Major Transients: List (1-2) 3 EOls used: List (1-3) 2 EQI Contingencies used: List (0-3) 75 Validation Time (minutes) 5 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix 0 Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER K/A Q SRO

1. Alternate EHC Pumps ROU-47A-NO-04 241000A4.10 2.9 2.9

2. Lower Power with Recirc Flow RO U-068-NO-25 2.1.23 4.3 4.4 SRO 5-000-NO-I 38

3. ADS SRV Fails Open RO U-001-AB-1 239002A2.03 4.1 4.2 SRO S-000 1 -AB-1

4. VFD Cooling Water Pump Failure RO U-068-AL-33 202001A2.22 3.1 3.2 SRO S-068-AB-01

5. DIG Automatic Board tie failure RO U-082-NO-01 262001A4.01 3.4 3.7 SRO S-000-AD-27

6. RFPT C High Vibrations RO U-003-AL-11 259001A2.01 3.7 3.7 SRO S-003-AB-1

7. Feedwater Line Break RO U-000-EM-13 295031 EA2.04 4.6 4.8 SRO S-000-EM-14 SRO S-000-EM-15

3 Page 8 of 62 Procedures Used/Referenced:

Procedure Number Procedure Procedure Title Revision 3-Ol-47A

[ EHC System Relief Valve Stuck Open Rev 37 Rev 11 3-AOl-i-I 3-GOI-100-12 Power Maneuvering Rev 37 3-01-68 Reactor Recirculation System Rev 81 3-9-3C Window 25 MAIN STEAM RELIEF VALVE OPEN Rev 23 3-AOl-I-i Relief Valve Stuck Open Rev 1 1 3-01-74 Residual Heat Removal System Rev 100 3-9-4B Window 12 RECIRC DRIVE 3B COOLANT FLOW LOW Rev 42 3-9-4B Window 28 RECIRC DRIVE 3B PROCESS ALARM Rev 42 3-9-4B Window 32 RECIRC DRIVE 3B DRIVE ALARM Rev 42 3-9-6C Window 16 RFPT BRG OIL PRESS LOW Rev 12 3-9-6C Window 17 RFPT BRG OIL PRESS LOW Rev 12 Loss Of Reactor Feedwater or Reactor Water 3-AOI-3-1 Rev 9 Level High/Low 3-9-5A, Window 8 REACTOR WATER LEVEL ABNORMAL Rev 41 3-EOI-3 Secondary Containment Control Rev 1 1 3-AOl-i 00-1 Reactor Scram Rev 55 3-EOI-1 RPV Control Rev 8 3-EOI-Appendix-1 1A Alternate Pressure Control Systems MSRVs Rev 2 3-EOI-Appendix-5C Injection System Lineup RCIC Rev 3 3-EOI-3-C-2 Emergency RPV Depressurization Rev 8 3-EOI-2 Primary Containment Control Rev 8 EPIP-i Emergency Classification Rev 47

3 Page 9 of 62 Simulator Instructor 1C28 bat nrc2Ol 020:

crd A pump oos bat nrc2OlO2Oa:

or zlohs85la[1] off mmffw33g (e4 0)12 60

srv open mmffw33h (e4 0)1260 imfadolk(elO 0)70 mmffw33c (e4 0) 36 60 trg 11 nrc2OlO2Osrv mmffw33d (e4 0) 3660 trg 11 =dmfad0lk, mmffw33e (e4 0)1260 trg 12=mrfad0lkout mmf fw33f (e4 0) 12 60 mmffw33j (e4 0)1260

vfd cooling pump failure mmffw33k(e40)11 60 ior zlohs682b2a[1] on mmffw33l (e4 0)1160 ior zlohs682b2a[2] off mmffw33m(e40)11 60 mrfthl8d trip iorzdihs682bla[1] (el 0) off trg 2 nrc2Ol O2Ovfd trg 2 bat nrc2OlO2O2b bat nrc2OlO2O2b:

Loss of 4KV SD BD 3ED!DG 0 Fail to tie imf dgo3d mrfthl8d close ior zdi34321 I 3ed (e5 0) trip dor zlohs682b2a[1]

or zdi3hs2ll3ed8a (e5 0) trip dor zlohs682b2a[2]

rfpt low oil pressure! high vibration! oil pump trip imffw33g (e3 0)48 120 imffw33h (e3 0)4560 bat nrcmsivd imffw33c (e3 0)65 120 imffw33d (e3 0)7090 dmf mso6g imffw33e (e3 0)5245 imffw33f(e3 0)6770 imffw33j (e3 0)3830 imffw33k (e3 0)82 100 bat nrcsteamleak:

imffw33l (e3 0)6380 imffw33m (e3 0)4060 imfth35d 5360 1 ior zIohsO3l 54a[2] (e3 0) off ior xa556c[11] (e3 0) alarm_on ior xa556c[1 5] (e3 0) alarm_on ior xa556c[1 6] (e3 0) alarm_on ior xa556c[26] (e3 0) alarm_on trg 4 = bat nrc2OlO2Oa

Major FW leak in steam tunnel imffwl9 (e20 0)30300 imf mso6g imf dgola trg 14 nrc2ollmodesw trg 14 = bat nrcsteamleak trg 15 nrcmsivd trg 15 = bat nrcmsivd bat atws70 trg 25 = bat appOlf trg 26 = bat appo2 trg 27 = bat appo8ae trg 28 = bat atws-1 trg 29 = bat sdv imf s102 60 trg 30 nrcslc trg 30 = br an:xa555b14 alarm_on trg 6 nrcslcl trci 6 = br zlohs636al2 on

3 Page 10 of 62 S,mulator Instructor 1C28 DESCRI PTIONIACTION Simulator Setup manual Reset to IC 28 Simulator Setup Shift to 3B CRD pump and manual clearance out 3A CRD pump Simulator Setup Fault reset and clear alarm manual on_Recirc_Pump_3B Simulator Setup Load Batch bat nrc2OlO2O Simulator Setup manual Verify file loaded Procedures:

RCP required (100% 90% with flow)

RCP for Urgent Load Reduction

Provide marked up copy of 3-GOI-100-12

3 Page 11 of 62 Simulator Event Guide:

Event 1 Normal: Weekly EHC Pump Test per 3-Ol-47A section 6.2.

6.2 EHC Auto Pump Start Test & Weekly Pump Alternation NOTES

1) This section is performed from Panel 3-9-7 unless otherwise specified.

2) This test should be performed during weekly alternation of pumps.

3) This section describes the actions necessary to test (standby) EHC Pump 3A. Testing EHC Pump 3B is the same and the component numbers are enclosed in parenthesis.

4) Operations personnel should be present at the El-IC skid to observe proper system operation.

5) If EHC PUMP 3A(3B) TEST, 3-HS-47-4A(5A), is depressed for longer than 10 seconds, annunciator STANDBY EHC PUMP FAILED, 3-XA-47-111 (3-XA-55-7B,Window 15), will alarm.

BOP [1] VERIFY the EHC System is in service. REFER TO Section 5.1.

[2] REVIEW Precautions and Limitations listed in Section 3.0.

BOP

[3] DEPRESS the EHC PUMP 3A(3B) TEST, 3-HS-47-4A(5A),

BOP and CHECK the following actions occur:

. EHC Hydraulic Fluid Pump 3A(3B) starts.

. Annunciator STANDBY EHC PUMP RUNNING, 3-XA-47-1 08 (3-XA-55-7B Window 8), ANNUNCIATES.

. Red light above test switch is ILLUMINATED (PS-47-1 B(2B)) (positive indication of pump discharge pressure).

NOTE ALLOW both EHC pumps to operate for at least 30 seconds to allow the Standby pump to expel any air which may have accumulated in the pump casing.

3 Page 12 of 62 Simulator Event Guide:

Event I Normal: Weekly EHC Pump Test per 3-Ol-47A section 6.2.

[4] CHECK the started EHC HYD PUMP A(B) DISCH PRESS, BOP 3-PI-047-0001 (0002), indicates between 1550 psig and 1750 psig, locally at the EHC skid.

Report EHC HYD PUMP A(B) DISCH PRESS, DRIVER 3-PI-047-0001 (0002), indicates between 1650 psig, locally at the EHC skid.

[5] IF the started EHC pump discharge pressure is NOT between 1550 psig and 1750 psig, THEN BOP ADJUST the pressure compensator for the started EHC pump to adjust the pump discharge pressure. REFER TO Step 8.6[1}

NOTES

1) The voltmeters in Step 6.2[6] normally indicate approximately zero volts. When an EHC header pressure switch actuates, the associated voltmeter will indicate approximately mid-scale. The Unit Operator should be notified if a voltmeter indicates greater than 5 volts with the EHC System in service.

2) If two out of three EHC header pressure switches actuate, a turbine trip will occur.

[6] CHECK locally on Junction Box 3-JBOX-047-1 01 66 that the following voltmeters indicate approximately 0 volts:

TURBINE EHC HDR PRESS 3-PS-47-63A TRIP IND, 3-El-047-0063A

TURBINE EHC HDR PRESS 3-PS-47-63B TRIP IND, 3-El-047-0063B

TURBINE EHC HDR PRESS 3-PS-47-63C TRIP IND, 3-El-047-0063C BOP Direct AUO to check local voltmeters DRIVER Report local voltmeters indicate 0 volts.

[7] IF alternating operating EHC pumps, THEN BOP STOP EHC pump 3B(3A) using EHC HYD FLUID PUMP 3B(3A), 3-HS-47-2A(1A).

3 Page 13 of 62 Simulator Event Guide:

Event I Normal: Weekly EHC Pump Test per 3-Ol-47A section 6.2.

[8] IF NOT alternating EHC pumps, THEN STOP EHC pump 3A(3B) using EHC HYD FLUID PUMP 3A(3B), 3-HS-47-1 A(2A).

[9] IF the pump stopped in Step 6.2[7] or 6.2[8] fails to stop or remain stopped, THEN PERFORM the following:

[9.1] DEPRESS and RELEASE EHC PUMP 3A(3B) TEST, 3-HS-47-4A(5A), to exercise and attempt to correct the positioning of 3-FSV-47-4(5).

[9.2] RESET annunciator 3-XA-55-7B, Window 15.

[9.3] STOP the desired EHC pump using EHC HYD FLUID PUMP 3A(3B), 3-HS-47-IA(2A).

[9.4] IF the pump failed to stop, THEN REPEAT Steps 6.2[9.1] through 6.2[9.3], as necessary, up to 2 additional times.

[9.5] IF the pump failed to stop after performing the preceding step, THEN NOTIFY Site Engineering and/or Maintenance to assist in returning 3-FSV-47-4(5) to its normal position.

[9.6] WHEN 3-FSV-47-4(5) has been returned to its normal position, THEN REPEAT Steps 6.2[9.1] through 6.2[9.3], and CONTINUE at Step 6.2[10].

[10] CHECK EHC HEADER PRESSURE, 3-Pl-47-7, indicates BOP between 1550 psig and 1650 psig.

[11] IF the started EHC pump discharge pressure is NOT between 1550 psig and 1750 psig, THEN ADJUST the pressure compensator for the started EHC pump to adjust the_pump_discharge_pressure._REFER_TO_Step_8.6[1]

[12] IF pumps were alternated, THEN BOP RESET any disagreement flags by placing the operating EHC pump handswitch, 3-HS-47-IA(2A), to START.

3 Page 14 of 62 Simulator Event Guide:

Event 2 Reactivity: Lower Power with Flow to 90%.

Directs Power Reduction using Recirc Flow per 3-GOI-1 00-1 2:

[9] REDUCE reactor power by a combination of control rod insertions and SRO core flow changes, as recommended by Reactor Engineer.

REFER TO 3-SR-3.1.3.5(A) and 3-01-68. (N/A if entering 3-GOI-100-12 to recover from Recirc Pump Trip) 3-01-68 Precaution and Limitations 3.5.3 Dual Pump Operation 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).

Lowers Power w/Recirc using 3-01-68 section 6.2 and establishes 60 rpm ATC split 3-01-68 Section 6.2, Adjusting Recirc Flow

[1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following; (Otherwise N/A)

. Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM),

3-HS-96-15A(15B). (Otherwise N/A)

ATC

Lower Recirc Pump 3A using SLOW (MEDIUM) (FAST),

3-HS-96-1 7A(1 7B)(1 7C). (Otherwise N/A)

AND/OR

Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM),

3-HS-96-16A(16B). (Otherwise N/A)

Lower Recirc Pump 3B using SLOW (MEDIUM) (FAST),

3-HS-96-1 8A(1 8B)(1 8C). (Otherwise N/A)

3 Page 15 of 62 Simulator Event Guide:

Event 2 Reactivity: Lower Power with Flow to 90%.

[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:

ATC RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32 LOWER SLOW, 3-HS-96-33 LOWER MEDIUM, 3-1-15-96-34 LOWER FAST, 3-HS-96-35 NRC NRC When satisfied with Reactivity Manipulation, ADS SRV Fails Open requiring power to_be_lowered_to_less_than_90%

Driver Driver At lead examiner direction, insert TRIGGER 10 for failure of ADS SRV 1-41

3 Page 16 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

BOP Report alarm MAIN STEAM RELIEF VALVE OPEN (3-9-3C Window 25)

MAIN STEAM RELIEF VALVE OPEN (3-9-3C Window 25)

A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 3-TR-1-1, on Panel 3-9-47 and SRV Tailpipe Flow Monitor on Panel 3-9-3 for raised temperature and flow indications.

B. REFER TO 3-AOl-I -1.

C. IF alarm is due to sensor malfunction, THEN REFER TO 0-01-55 and OPDP-4.

SRO Enters 3-AOl-I-i Relief Valve Stuck Open 3-AOl-I -I Relief Valve Stuck Open NOTE Once a MSRV is operated, a time delay of 15 to 30 seconds can be expected before a response can be detected on 3-TR-1-1. ICS can be used to monitor the discharge tailpipe temperature, but the appropriate indications on 3-TR-1-1 must be confirmed.

BOP Identifies ADS SRV 1-41 open 4.1 Immediate Action

[1] IDENTIFY stuck open relief valve by BOP OBSERVING the following:

SRV TAILPIPE FLOW MONITOR, 3-FMT-1-4, on Panel 3-9-3, OR

MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 3-TR-1-1 on Panel 3-9-47.

[2] IF relief valve transient occurred while operating above 90%

power, THEN ATC REDUCE reactor power to 90% RTP with recirc flow.

3 Page 17 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

[3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; BOP 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. (Otherwise N/A)

NOTES

1) Once initial transient of SRV opening has stabilized (pressure regulator compensation) the Heat Balance will indicate bad data.

2) The SRV TAILPIPE FLOW MONITOR may seal-in an OPEN position indication.

3-AOl-I-I 4.2 Subsequent Action NOTES

1) Once initial transient of SRV opening has stabilized (pressure regulator compensation) the Heat Balance will indicate bad data.

2) The SRV TAILPIPE FLOW MONITOR may seal-in an OPEN position indication.

4.2.1 Action if a fire exists with SRV stuck open

[1] IF an SRV is open and a fire exists in ANY Appendix R fire area, THEN (Otherwise N/A):

INITIATE a manual scram before the Suppression Pool temperature_exceeds 95°F.

3 Page 18 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes w/ inhibit switch.

4.2.2 Attempt to close valve from Panel 9-3:

[1] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the OFF position.

BOP

[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.

(Otherwise N/A)

[4] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT, 3-XS-1-202 in INHIBIT:

CT#1 Observe and report when 3-XS-1-202 is placed in Inhibit, ADS SRV 1-41 BOP closes.

[5] IF relief valve closes, THEN OPEN breakers (250V RMOV BD 3A BRKR 9B1, 250V RMOV BD 3C BRKR 8A, and BB2 BRKR 710)or PULL fuses (3-FU1-001-0041A, 3-FU1 -001-0041 B, 3-FUI-001-0041 C, and 3-FUI-001 -0041 D) as necessary usingAttachment I (Unit 3 SRV Solenoid Power Breaker/Fuse Table).

DRIVER Insert TRIGGER 12 to pull fuses 2 minutes after direction DRIVER Operator Does NOT perform step 6 until Breaker opened or fuses pulled.

[6] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT 3-XS-1 -202, in BOP AUTO.

NOTES

1) Only the appropriate sections for the stuck open relief valve is required to be performed.

2) The ADS valves that have more than one power supply will AUTO TRANSFER on a loss of power, and are NORMAL SEEKING.

3) ADS Relief valves with hand-switches on Panel 25-32 are listed below and should be operated from that location first.

4) When opening breakers and pulling fuses, opening the breakers is the preferred method when time permits. However, the breakers with multiple locations will require opening each breaker to de-energize the control circuit. In this case, pulling the fuses from Panel 25-32 may be quicker than opening the breakers.

3 Page 19 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

[7] IF the SRV valve did not close, THEN PERFORM the appropriate section from table below.

Directs AUO to Remove Power from SRV 1-41 REMOVE the power from 3-PCV-1 -41 by performing one of the following:

A. OPEN the following breakers (Preferred method)

250V RMOV 3A, compartment 9B1

250V RMOV 3C, compartment 8A BOP

Battery Board 2, breaker 710 OR B. In Panel 3-25-32 PULL the following fuses as necessary

Fuse 3-FUI-001-0041A (BlockAA, F2)

Fuse 3-FUI-001-0041B (BlockAA, F7)

Fuse 3-FU1-001-0041C (BlockAA, F12)

Fuse 3-FUI-001-0041D (BlockAA, F15)

SRO Evaluate Tech Spec 3.5.1 CONDITION REQUIRED ACTION COMPLETION TIME E. One ADS valve E.1 Restore ADS Valve 14 Days Inoperable to OPERABLE status SRO Evaluate Appendix R Page 806 of 904 3-PCV-O01-0041 MAIN STEAM LINE D RELIEF VLV 3 ISOLATE AND OPEN AT PANEL 3-LPNL-925-0032 Compensatory Measures A Area/Zone Affected 16 3 OPEN FROM MCR Compensatory Measures A Area/Zone Affected 3-1, 12

3 Page 20 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

SRO May direct Suppression Pool Cooling placed in service lAW 3-01-74, Section 8.5 If Directed, places Suppression Pool Cooling in Service Loop I:

BOP 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 supp pool cooling.

Suppression pool cooling run times are tracked in 3-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 < 95°F, or when directed by other procedures.

2) All operations are performed at Panel 3-9-3 unless otherwise noted.

[1] VERIFY RHR Loop 1(11) in Standby Readiness. REFER TO Section 4.0.

[2] REVIEW the precautions and limitations in Section 3.0.

[3] NOTIFY the 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 the Narrative Log

[5] IF possible before placing RHRSW in service, THEN NOTIFY Chemistry that RHRSW sampling is to be initiated (RI-IRSW sampHng requirements).

3 Page 21 of62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

[6] VERIFY at least one RHRSW Pump is operating on each EECW Header NOTES

1) Step 8.5[7] initiates Suppression Pool Cooling for RHR Loop I.

2) Step 8.5[8] initiates Suppression Pool Cooling for RHR Loop II.

3) RHR Pump(s) may be operated with no RHRSW flow through the associated RHR Heat Exchanger(s), in support of maintenance or testing, provided the RHRSW side of the heat exchanger is pressurized and is approved by the Unit Supervisor.

[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).

CAUTIONS

1) DO NOT EXCEED 4500 gpm RHRSW flow through any single RHR Heat Exchanger.

2) When operating RHRSW through the heat exchangers, damage can occur to the RHRSW discharge valves for the RHR Heat Exchanger if operating at low flows and high differential pressures for long periods. In order to lower the differential pressure the valves experience, flow through the in service heat exchanger(s) should be established such that the total header flow is 4000 gpm. When operating RHRSW in split mode with other units, this is calculated by adding the individual flows from each of the in service RHR heat exchangers.

[7.2] ESTABLISH RHRSW flow by performing one the following:

[7.2.2] THROTTLE OPEN RHR HX 3A(3C) RHRSW OUTLET VLV, 3-FCV-23-34(40), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 3-Fl-23-36(42),

RHR HTX 3A(3C) RHRSW FLOW.

[7.3] VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 3-FCV-74-53.

[7.4] VERIFY CLOSED RHR SYS I SUPPR POOL CLGJTEST VLV, 3-FCV-74-59.

[7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 3-FCV-74-58.

[7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 3-FCV-74-60.

[7.7] VERIFY OPEN RHR SYS I SUPPR CHBR/POOL ISOL VLV, 3-FCV-74-57.

3 Page 22 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

[7.8] VERIFY OPEN RHR SYS I SUPPR CHBRIPOOL ISOL VLV, 3-FCV-74-57.

CAUTIONS

1) To prevent excessive vibration, RHR pumps should not be allowed to operate for more than 3 minutes at minimum flow.

2) Capacitor bank fuses are subject to clearing when the unit boards are being supplied from the 161kV source and large pumps are started. Unit Supervisors should evaluate placing the Capacitor Banks in Manual prior to starting RHR pumps per 0-Ol-57A.

3) When throttling RHR SYS I SUPPR POOL CLG/TEST VLV, 3-FCV-74 59, maintain Blue light illuminated in order to maintain LPCI operability.

NOTE RHR flow should be monitored while in operation on 3-Fl-74-50, RHR SYS I FLOW.

RHR flow should remain 10,000 gpm for 1-pump operation and is limited to <

1 3000 gpm, for two pump operation, due to the flow restricting orifice in the test return line.

[7.9] START RHR PUMP A(C) using 3-HS-74-5A(16A).

[7.10] THROTTLE RHR SYS I SUPPR POOL CLG/TEST VLV, 3-FCV-74-59, to maintain RHR flow within limits, as indicated on RHR SYS I CTMT SPRAY FLOW, 3-FI-74-56.

RHR Pumps in 1 2

[peration Loop Flow 7,000 to <13,000 gpni &

10000 gpm & Blue Blue light Light illuminated illuminated

[7.1 1] 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.

BOP If Directed places Suppression Pool Cooling in Service Loop 2

[8] PLACE RI-IR Pump and Heat Exchanger B(D) in service as follows:

[8.1] START_an_RHRSW_Pump to_supply_RHR_Heat_Exchanger_B(D).

CAUTIONS

1) DO NOT EXCEED 4500 gpm RHRSW flow through any single RHR Heat Exchanger.

2) When operating RHRSW through the heat exchangers, damage can occur to the RHRSW discharge valves for the RHR Heat Exchanger if operating at low flows and high differential pressures for long periods. In order to lower the differential pressure the valves experience, flow through the in service heat exchanger(s) should be established such that the total header flow is 4000 gpm. When operating RHRSW in split mode with other units, this is calculated by adding the individual flows from each of the in_service_RHR_heat_exchangers.

3 Page 23 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes w/ inhibit switch.

[8.2] THROTTLE OPEN RHR HX 3B(3D) RHRSW OUTLET VLV, 3-FCV-23-46(52), as required for cooling (Refer to caution 2 above.)

[8.3] IF required to maintain Total RHRSW Flow for RHRSW Pump B(D) between 4000 and 4500 gpm, THEN REQUEST another unit to establish RHRSW flow for the associated RHRSW Pump B(D) and maintain Total RHRSW flow between 4000 and 4500 gpm per 0 23. (Otherwise N/A)

[8.4] VERiFY CLOSED RHR SYS II LPCI INBD INJECT VALVE, 3-FCV-74-67.

[8.5] VERIFY CLOSED RHR SYS II SUPPR POOL CLG/TEST VLV, 3-FCV-74-73. (N/A if starting the second Loop II RHR Pump D(B))

[8.6] VERIFY CLOSED RHR SYS II SUPPR CHBR SPRAY VALVE, 3-FCV-74-72.

[8.7] VERIFY CLOSED RHR SYS II DW SPRAY OUTBD VLV, 3-FCV-74-74.

[8.8] VERIFY OPEN RHR SYS II SUPPR CHBRIPOOL ISOL VLV, 3-FCV-74-71.

CAUTIONS

1) To prevent excessive vibration, RHR pumps should not be allowed to operate for more than 3 minutes at minimum flow.

2) Capacitor bank fuses are subject to clearing when the unit boards are being supplied from the 161kV source and large pumps are started. Unit Supervisors should evaluate placing the Capacitor Banks in Manual prior to starting RHR pumps, as referenced in 0 57A.

3) When throttling RHR SYS II SUPPR POOL CLGITEST VLV, 3-FCV 73, maintain Blue light illuminated in order to maintain LPCI operability.

NOTE RHR Flow should be monitored while in operation on 3-FI-74-64, RHR SYS II FLOW. RHR Flow should remain 10,000 gpm for 1-pump operation and is limited to < 13000 gpm, for two pump operation, due to the flow restricting orifice in the test return line.

[8.9] START RHR PUMP B(D) using 3-HS-74-28A(39A).

[8.10] THROTTLE RHR SYS II SUPPR POOL CLGITEST VLV, 3-FCV-74-73, to maintain RHR flow within limits, as indicated on RHR SYS II CTMT SPRAY FLOW, 3-FI-74-70.

RHR Pumps in 1 2 Operation Loop Flow 7,000 to <13,000 gprn &

10,000 gpm & Blue Blue light light illuminated illuminated

3 Page 24 of 62 Simulator Event Guide:

Event 3 Component: ADS SRV 1-41 fails open, closes wI inhibit switch.

[8.11] 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 by REPERFORMING Step 8.5[8] for the second pump.

SRO Contacts maintenance to investigate cause of SRV failing open.

Driver Driver At NRC direction, insert TRIGGER I for trip of 2B1 VFD Cooling Pump NOTE NOTE Move BOP away from 9-4 for next event

3 Page 25 of 62 Simulator Event Guide:

Event 4 Component: VFD Cooling Water Pump failure.

NOTE NOTE Move BOP away from 9-4 for this event ATC Reports the following annunciators 9-4B-12, 28 and 32 RECIRC DRIVE 3B COOLANT FLOW LOW, RECIRC DRIVE 3B PROCESS ALARM, and RECIRC DRIVE 3B DRIVE ALARM ATC Reports the 2B2 VFD Cooling Water Pump for the B Recirc Pump, has tripped.

ATC Reports Standby Recirc Drive Cooling Water Pump 2B2, failed to auto start.

ATC RECIRC DRIVE 3B COOLANT FLOW LOW STARTS RECIRC DRIVE cooling water pump 2B2 and DISPATCHES personnel to the RECIRC DRIVE, to check the operation of the Recirc Drive cooling water system.

SRO Concurs with start of Standby VFD Pump.

BOP RECIRC DRIVE 3B DRIVE ALARM A. REFER TO ICS Group Display GD @VFDBDA and determine cause of alarm.

B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system.

C. IF the problem is conductivity in the cooling water system, THEN VERIFY demineralizer is in service.

D. IF a problem with power supplies is indicated, THEN VERIFY all the low voltage supply breakers are CLOSED/ON.

E. For all other alarms, or any problems encountered CONTACT system engineering.

SRO Contacts maintenance to investigate VFD cooling and contacts the RE to evaluate_thermal_limits_at_current_power_level Crew Verifies Standby pump started on VFD ICS display GD @VFDBDA.

BOP Dispatches personnel to VFD.

Wait 4 minutes after dispatched, THEN report tripped VFD Pump 2B1 is hot DRIVER to the touch, internal bkr closed, 480 volt bkr tripped (480 V SD BD 3A-5D).

3 Page 26 of 62 Simulator Event Guide:

Event 5 Component: Loss of 4KV Shutdown Board 3ED, 3D D/G fails to AUTO tie.

DRIVER Insert TRIGGER 5 (imf edO9d) and then TRIGGER 6 (dmf edO9d)to cause a loss of Shutdown Board 3ED, ensure EDO9D is deleted prior to operator start DG 3ED.

Recognizes Loss of Shutdown Board D and failure of DG 3ED automatically BOP start. Manually Starts DG 3ED and closes DG Output Breaker Reports Loss of Shutdown Board 3ED, failure of DG 3ED to start, and BOP manual start of DG 3ED to SRO.

SRO Dispatch AUO to D/G and call Electrical Maintenance, Shift Manager DRIVER When requested to investigate the shutdown board: Report that while Maintenance was moving a spare breaker in the area of the Shutdown Board 3ED, it bumped the racking tool shutter door for the Shutdown Board 3ED Normal breaker. This caused the Normal breaker to open.

As AUO, report are issues with the board (i.e. NO Lockouts).

3 Page 27 of 62 Simulator Event Guide:

Event 5 Component: Loss of 4KV Shutdown Board 3ED, 3D DIG fails to AUTO tie.

SRO Evaluates Tech Specs 3.8.1 (condition B and G) and 3.8.7 (condition A) 3.8.1 AC Sources Operating CONDITION REQUIRED ACTION COMPLETION TIME B. One required B.1 Verify power 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Unit 3 DG availability inoperable. from the offsite AND transmission network.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months AND thereafter B.2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 B.3.1 Determine 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit 3 DG(s) are not inoperable due to common cause failure.

OR B.3.2 Perlorm SR 3.8.1.1 24 Hours for OPERABLE Unit 3 DG(s).

AND B.4 Restore Unit 3 DG to 7 days OPERABLE status.

AND 14 days from discovery of failure to meet LCO hours

3 Page 28 of 62 Simulator Event Guide:

Event 5 Component: Loss of 4KV Shutdown Board 3ED, 3D D/G fails to AUTO tie.

3.8.1 AC SourcesOperating (CONTINUED)

CONDITION REQUIRED ACTION COMPLETION TIME NOTE Applicable when only one 4.16 kVshutdown board is affected.

G.1 Declare the affected Immediately C. One required offsite 4.16KV shutdown board circuit inoperable.

inoperable.

AND One Unit 3 DG inoperable.

3.8.7 Distribution Systems Operating CONDITION REQUIRED ACTION COMPLETION TIME NOTE Enter applicable Conditions and Required Actions of Condition B, C, D, and G when Condition A results in no power source to a required 480 volt board.

A. One Unit 3 4.16 kV A.1 Restore the Unit 3 5 days Shutdown Board 4.16 kV Shutdown Board inoperable, to OPERABLE status. AND 12 days from discovery of failure to meet LCO AND A.2 Declare associated Immediately diesel generator inoperable.

3 Page 29 of 62 Simulator Event Guide:

Event 5 Component: Loss of 4KV Shutdown Board 3ED, 3D D/G fails to AUTO tie.

Appendix R Mauual#: Fire Protection Report PLtT: BFN UNIT(S) 1/2/3 PAGE 888 of 904

[ Vol.1 I TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 11 APPENDIX R TESTING EQUIPMENT UNIT (S) FUNCTION REQUIREMENTS PROCEDURES Diesel 0 Start and load Verify diesel can be 3-SR-3.8.11(3D)

Generator 3D from Main Control started and loaded 3-SR-3.8.l.7(3D)

Room from Main Control Room DRIVER When directed by lead examiner insert trigger 3 for RFPT High Vibrations

3 Page 30 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

DRIVER When directed by lead examiner insert trigger 3 for RFPT High Vibrations ATC Respond to numerous RFPT C Alarms and indications on Panel 9-6.

ATC Report Vibrations increasing on RFPT C and the 3C oil pump has tripped ATC RFPT HP OIL PRESS LOW A. CHECK Feedwater flow, discharge pressure, RPMs, and valve position on affected Feedpump, Panel 3-9-6.

B. IF oil pressure is lowering, THEN IMMEDIATELY START companion Main Oil Pump.

C. DISPATCH personnel to RFPT Room to check hydraulic oil pressure, oil tank level and oil system for leaks.

D. IF hydraulic pressure lowers to 75 psig as seen on the RFPT control panel, THEN MANUALLY TRIP the Feedpump turbine.

RFPT C ABNORMAL 3-9-6C-Window 15 REFER TO appropriate alarm response procedure.

RFPT BRG OIL PRESS LOW 3-9-6C-Window 16 A. IF the RFPT has tripped, THEN REFER TO 3-01-3, Section 8.1.

B. IF not running, THEN START Companion Main Oil Pump.

C. DISPATCH personnel to RFPT Room to check oil pressure, strainer DP, oil tank level and oil system for leaks.

3-01-3, RFW System Precautions and Limitations GG. For operating Feed Pumps, monitor and maintain the following parameters within ranges described below.

1. RFPT Hydraulic Pressure: RFPT Hydraulic Pressure:

2. Lube Oil Pressure to RFP Bearings: Lube Oil Pressure to RFP

3. Lube Oil Pressure to RFPT Bearings: Lube Oil Pressure to RFPT

4. Bearing lube oil from cooler: IIOEI. Bearing lube oil from cooler: 1 Computer Point Ids 24-56, 24-54, and 24-52).

5. Bearing lube oil to cooler: 180L1. Bearing lube oil to cooler: 18 Computer Point Ids TBDO25, TBDO32, and TBDO39).

6. Maximum Oil Temp Rise across the Turbine Bearings: 50EI.

7. Vertical Vibration at RFP Bearing Supports: 2 mils double amplitude.

8. RFPT Speed: 5050 rpm maximum (3-9-6).

3 Page 31 of62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

SRO Directs normal Shutdown of RFPT 3C and call work control to initiate work order RFPT VIB OR AXIAL POSITION HIGH-HIGH 3-9-6C-Window 17 A. CHECK RFPT/RFP vibration readings on 3-XR-3-177 on Panel 3-9-6 AND RFPT and REP Vibration display(RFPTV) on ICS.

B. DISPATCH personnel to Panel 3-LPNL-025-0673, Vibration Monitoring Panel, located outside of RFP Room 3A, EL 617, to PERFORM the following:

REPORT vibration data for affected RFPT/RFP.

REPORT all alarm/alert conditions on panel.

Advise the Unit Operator of any changes in vibration data.

C. IF a sustained vibration exceeding the DANGER setpoints (REFER TO setpoints on the next page) is confirmed on both pump inboard and outboard bearings or any turbine bearing, THEN REMOVE the RFPT from Service.

Driver Driver After dispatched wait 3 minutes and report an Oil Pressure of 65# and lowering, High Strainer DP and a small oil leak, not able to determine the source of the leak at this time. Oil Level is 5/8. If Pump has not been tripped report that you cannot enter the RFPT Room Driver Driver When RFPT is tripped initiate Trigger 4 to lower vibration readings (bat nrc2OlO2Oa).

SRO Directs trip of RFPT 3C RFPT TRIP CIRCUIT ABNORMAL A. VERIFY alarm and RFPT trip by checking Panel 3-9-6, RFPT speed, ATC governor valve position and discharge flow.

B. VERIFY reactor power is within the capacity of operating RFPs.

C. IF BKR TRI POUT PNL 3-9-9 DC DIST (3-XA-55-8C, alarm window 20) is illuminated, THEN CHECK for tripped breakers 105, 106, and 107 on Panel 3-9-9.

D. IF REP is tripped, THEN REFER TO 3-01-3, Section 8.1 or 3-AOI-3-1.

Crew . .

Plant Announcement Tripping 3C RFPT

3 Page 32 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

RFPT TRIPPED; RFPT VIBIAXIAL POSITION HIGH OR PWR FAILURE 3-9-6C Window 33 ATC A. CHECK RFPT/RFP vibration readings on 3-XR-3-177 on Panel 3-9-6 AND RFPT and REP Vibration display(REPTV) on ICS.

B. DISPATCH personnel to Panel 3-LPNL-025-0673, Vibration Monitoring Panel, located outside of REP Room 3A, El 617, to PERFORM the following:

CHECK Power Supplies on panel energized.

REPORT vibration data for affected REPT/REP.

REPORT all alarm/alert conditions on panel.

ADVISE Unit Operator of any changes in vibration data.

C. IF vibration is high, THEN PERFORM the following:

CHECK condenser vacuum on 3-PITR2-2.

CHECK oil and bearing temperatures on the computer printout.

D. VERIFY l&C 3B power (Panel 3-9-9, BKR 329) NOT tripped.

E. REQUEST assistance from Site Engineering.

F. ADJUST load on pump, as necessary.

Driver Driver Vibration Report: If RFPT is not tripped, report high vibrations numerous alarm and alert conditions and a rumble in the RFPT C Room. If RFPT has been tripped report vibration readings lowering and numerous alarm and alert conditions were in. Acknowledge communications with other organizations ATC Recommends Trip of RFPT C or Trips RFPT C SRO Direct Trip of REPT C and enter Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-AOl-3-1 NOTE NOTE The SRO may chose to runback Reactor Power with Recirc flow prior to tripping the RFPT, OR upon tripping the RFPT, the Recirc Pumps receive a run back signal to 75% speed at 27 (normal range) if the discharge flow of a REP is less than 889,000 lb/hr 19% (rated flow).

Driver Driver When RFPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 33 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

3-01-3 7.1 RFPIRFPT Shutdown CAUTIONS

1) FAILURE to monitor SJAE/OG CNDR CNDS FLOW, 3-Fl-2-42, on Panel 3-9-6 for proper flow (between 2 x 106 and 3 x 106 lbm/hr) may result in SJAE isolation.

2) Changes in Condensate System flow may require adjustment to SPE CNDS BYPASS, 3-FCV-002-0190.

3) When isolating the Reactor Feedwater Pump(s) for maintenance, the associated injection water should also be isolated to prevent high seal differential pressure and allow the RFW Pump shafts to rotate freely.

[1] REFER TO Section 3.0 and REVIEW Precautions and Limitations.

NOTE It may be necessary to switch to SINGLE ELEMENT mode from THREE ELEMENT mode earlier than recommended if Feedwater control becomes unstable.

[2] IF REP being removed from service is last operating RFP OR IF at any time Feedwater control becomes unstable, THEN DEPRESS SINGLE ELEMENT push-button, 3-HS-46-6/1. (Otherwise N/A)

VERIFY green backlight for push-button illuminated.

[3] VERIFY in AUTO, RFPT 3A(3B)(3C) TURNING GEAR MOTOR, 3-HS IO1A(127A)(152A).

NOTES

1) When selected, then Column 1 on individual RFPT Speed Control Panel Display Stations (PDS) displays actual pump speed and is NOT controlled in any mode.

2) When selected, then Column 2 on individual RFPT Speed Control PDS displays pump flow bias and is changed with Ramp Up/Ramp Down push buttons with controller in AUTO.

3) When selected, then Column 3 on individual RFPT Speed Control PDS displays RFPT speed demand and is changed with Ramp Up/Ramp Down push-buttons with controller in MANUAL.

4) Illustration 2 has additional information on RFPT Speed Control PDSs.

Driver Driver When RFPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 34 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

[4] LOWER speed of RFPT/RFP being removed from service by performing either one of the following:

ElF using individual RFPT Manual Governor switch, THEN GO TO Step 7.1[5].

LJIF using individual RFPT Speed Control PDS in MANUAL, THEN GO TO Step 7.1[6].

[5] LOWER speed of RFPT using individual RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER switch 3-HS-46-8A(9A)(1OA) as follows (Panel 3-9-5):

[5.1] DEPRESS RFPT Speed Control Raise/Lower switch to MANUAL GOVERNOR.

L:IVERIFY illuminated amber light at switch.

[5.2] SLOWLY LOWER RFPT speed by placing RFPT Speed Control switch in RAISE or LOWER positions as necessary.

[5.3] IF this is NOT the last operating REP, THEN OBSERVE rise in speed of any operating REPT in auto as REW Control System maintains Reactor water level.

[6] LOWER speed of RFPT using individual REPT 3A(3B)(3C) SPEED CONTROL PDS, 3-SIC-46-8(9)(1O) as follows (Panel 3-9-5):

[6.1] PLACE PDS in MANUAL and VERIFY Column 3 selected.

[6.2] SLOWLY LOWER RFPT speed using Ramp Up/Ramp Down push-buttons as necessary.

[6.3] IF this is NOT the last operating REP, THEN OBSERVE rise in speed of any operating REPT in auto as REW Control System maintains Reactor water level.

CAUTION REP Discharge Check Valve failure may be experienced while removing REP from service.

[7] IF at any time RFP Discharge Check Valve failure is experienced while removing REP from service, THEN PERFORM the following: (Otherwise N/A)

[7.1] DEPRESS RFP Discharge Testable Check valve push-button for apDroximatelv ten seconds (Panel 3-9-6).

Driver briver When REPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 35 of 62 Simulator Event Guide:

Event 6 ComDonent: RFPT C high vibrations with failure to trip.

[7.2] CHECK reverse flow through check valve has stopped.

[7.3] IF REP Discharge Testable Check Valve failure is still being experienced, THEN PERFORM one of the following:

[7.3.1] IMMEDIATELY RETURN REP to service.

[7.3.2] PERFORM the following:

A. VERIFY open REP Minimum Elow Valve.

B. CLOSE REP Discharge Valve.

C. TRIP REPT.

[8] CONTINUE to slowly lower REPT speed to minimum speed setting (approximately 600 rpm).

[9] IF REPT/RFP being removed from service is NOT the last operating REP, THEN GO TO Step 7.1[11]

[11] WHEN RFPT 3A(3B)(3C) is ready to be shutdown, THEN DEPRESS REPT 3A(3B)(3C) TRIP, 3-HS-3-125A(151A)(176A), to trip RFPT being removed from service.

NOTES

1) Check valve position indicator should NOT be relied upon for positive valve closure indication.

2) Step 7.1 [13] is performed only if REP Discharge Check Valve failure occurs.

[12] VERIFY CLOSED, REP 3A(3B)(3C) DISCL-I TESTABLE CHECK VLV, 3-FCV-3-94(93)(92), by one of the following:

DUD DDLDLDLD

[13] IF REP Discharge Check Valve failure is experienced, THEN PERFORM the following:

[13.1] DEPRESS RFP 3A(3B)(3C) DISCHARGE TESTABLE CK VLV push-button, 3-HS-3-94A(93A)(92A).

LI VERIFY Discharge Check Valve closed.

[13.2] IF REP Discharge Check Valve failure is still being experienced, THEN PERFORM one of the following:

[13.2.1] IMMEDIATELY RETURN REP to service.

[13.2.2] PERFORM the following:

A. VERIFY OPEN RFP 3A(3B)(3C) MIN ELOW VALVE, 3-ECV-3-20(1 3)(6).

B. CLOSE REP 3A(3B)(3C) DISCHARGE VALVE using 3-HS-3-1 9A(1 2A)(5A).

C. VERIFY REPT tripped.

[13.2.3] REQUEST assistance from Site Engineering and Maintenance.

Driver Driver When RFPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 36 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

NOTE Turning Gear motor will lockout if Turning Gear does NOT engage within five seconds of reaching zero speed. Lockout can be reset by placing control switch to OFF and pulling switch out (in OFF position).

[14] IF REP is NOT rolling on minimum flow AND RFPT coasts down to zero speed, THEN VERIFY Turning Gear motor starts and engages.

[15] CLOSE REP 3A(3B)(3C) DISCHARGE VALVE, 3-FCV-3-1 9(1 2)(5).

[16] PLACE RFP 3A(3B)(3C) MIN FLOW VALVE, 3.-HS-3-20(13)(6), in CLOSE.

[17] VERIFY Turning Gear engaged.

CAUTION When isolating the Reactor Feedwater Pump(s) for maintenance, the associated injection water should also be isolated to prevent high seal differential pressure and allow the RFW Pumo shafts to rotate freely.

[18] IF Unit Supervisor determines it necessary to close suction valve when isolating the Reactor Feedwater Pump for maintenance, THEN PERFORM the following for the Reactor Feedwater Pump to be isolated: (Otherwise N/A)

[18.1] CLOSE REP 3A(3B)(3C) SUCTION VALVE, 3-FCV 83(95)(1 08).

[18.2] CLOSE REP 3A(3B)(3C) INJECTION WATER SUPPLY SOV, 3-SHV-003-0593(0604)(061 6).

[19] IF the Reactor Feedwater Pump is NOT being isolated for maintenance and Unit Supervisor determines it necessary to close suction valve, THEN CLOSE REP 3A(3B)(3C) SUCTION VALVE, 3-FCV-2-83(95)(108)

(otherwise N/A).

[20] CLOSE the following applicable valve (Panel 3-9-6):

. U DUIUUI. LILILILIIILH

. U U U i U U D U U U U I. U U U U U [1 LI U U U U. U U LIII U

. UUUDUDUUEUUEUUUUUUUUUUUJiiUü

[21] CLOSE the following applicable valve (Panel 3-9-6):

. UU U U U U U U U U U U U U U U U .I.ü U U Dii U

. UUU U U UUUU UUUUU U U U U 1. U U Dii U Driver Driver When RFPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 37 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

[22] OPEN the following drain valves for RFPT being removed from service:

A. RFPT 3A(3B)(3C) LP STOP VLV ABOVE SEAT DR, 3-FCV 120(1 25)(1 30)

B. RFPT 3A(3B)(3C) LP STOP VLV BELOW SEAT DR, 3-FCV 121(126)(131)

C. RFPT 3A(3B)(3C) HP STOP VALVE ABOVE SEAT DR, 3-FCV 122(1 27)(1 32)

D. RFPT 3A(3B)(3C) HP STOP VLV BELOW SEAT DR, 3-FCV 123(1 28)(1 33)

E. RFPT A(B)(C) FIRST STAGE DRAIN VLV, 3-FCV 124(1 29)(1 34)

F. RFPT 3A(3B)(3C) HP STEAM SHUTOFF ABOVE SEAT DRAIN, 3-FCV-006-01 53(01 55)(01 57), Local Control G. RFPT A(B)(C) LP STEAM SHUTOFF ABOVE SEAT DRAIN, 3-FCV-006-01 54(01 56)(01 58), Local Control CAUTION DO NOT remove Seal Steam from RFPT until Reactor is de-pressurized.

NOTE The remainder of Section 7.1 is required to be performed only when RFPT/RFP being removed from service is last operating RFPT/RFP and is NOT required to maintain Condenser Vacuum.

[23] REFER TO 3-OI-47C and REMOVE Seal Steam from all three RFPTs.

NOTE Illustration 7 provides instruction for controlling Raw Cooling Water through RFP lube oil cooler.

[24] WHEN lube oil to RFP and RFPT bearings reaches 1 10°F AND seal steam has been removed, THEN PLACE the following switches in OFF:

. he following switches in OFF:ngs reaches

[25] VERIFY the following temperature control valves fully closed:

.. L LDLILLEDLi(EI586, T13-G)

. 13-G), EDEILILILLE (El 586, T13-F)

. 13-F), (El 586, T13-E)

Driver Driver When RFPT is tripped initiate trigger 4 to lower vibration readings (bat nrc2Ol 020a).

3 Page 38 of 62 Simulator Event Guide:

Event 6 Component: RFPT C high vibrations with failure to trip.

[26] REFER TO 3-01-24 and MANUALLY ISOLATE cooling water supply to lube oil cooler.

[27] SHUTDOWN RFPT/RFP Oil Pumps as follows:

[27.1] IF Shutting down RFPT/RFP 3A Oil Pumps, THEN PLACE the following switches in STOP and PULL TO LOCK:

ELiLiiLiLEELiLiLiLiLiLii Li Li Li Li Li Li Li Li Li [1. E Li E Lii Li

[27.2] IF Shutting down RFPT/RFP 3B Oil Pumps, THEN PLACE the following switches in STOP and PULL TO LOCK:

Li Li Li Li Li Li Li Li Li 11 Li Li Li Li Li I. ELi E Li Li Li Li D Li Li Li Li D Li Li Li Li Li ]. E Li DLII Li Li 11

[27.3] IF Shutting down RFPT/RFP 3C Oil Pumps, THEN PLACE the following switches in STOP and PULL TO LOCK:

Li Li Li Li Li Li LI Li Li Li Li Li Li Li Li [1. E Li ELi Li Li Li D Li Li Li Li Li Li Li 11 Li Li , E Li ED iLi Li Li

[28] PLACE the following in STOP:

Li Li Li Li Li Li Li Li Li Li Li Li Li Li ] Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li Li 11. E Li LiE Li Li LI

[29] IF directed by Unit Supervisor, THEN STOP RFPT OIL TANK VAPOR EXTRACTOR, using 3-HS-3-126A.

SRO Call Radwaste to lock out TB sumps and contact Chemistry to determine possible_oil_intrusion_into_Condensate_system SRO Contact RE to evaluate Thermal Limits at current power level DRIVER When directed by Lead Examiner Insert Trigger 20 Feedwater Line Break in Turbine Bldg

3 Page 39 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

DRIVER When directed by Lead Examiner Insert Trigger 20 Feedwater Line Break in Turbine Bldg ATC Responds to alarms RECTOR FEED PUMPS A, B, AND C ABNORMAL, RFWCS ABNORMAL and REACTOR WATER LEVEL ABNORMAL ATC 3-ARP-9-5A Reactor Water Level Abnormal 3-9-5A Window 8 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-AOl-3-1 or 3-01-3.

C. IF 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-LI-3-253 has failed or is invalid, THEN with SRO permission, BYPASS the affected level instrument. REFER TO 3-01-3, Section 8.2.

ATC Monitors Reactor Water Level and Reports trend, recommends Manual Reactor Scram Determines Feedwater Line A Leak in the Turbine Building on line due to high Feedwater Line A Flow and Reactor Feed Pump Flows Increasing with a Lowering Reactor Water Level.

SRO Directs a Manual Reactor Scram inserted Directs Reactor Feed Pumps to be tripped, Reactor Feed Pump Discharge Valves shut, and Condensate Booster Pumps then Condensate Pumps secured (Isolate and stop leak)

ATC Inserts Manual Reactor Scram Trips Reactor Feed Pumps and shuts Reactor Feed Pump Discharge Valves Secures Condensate Booster Pumps then Condensate Pumps DRIVER DRIVER Wh reactor is scrammed, insert TRIGGER 29 (bat sdv).

NOTE NOTE Not Wdswiil insert.

3 Page 40 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

Reactor Scram OATC Hard Card IMMEDIATE ACTIONS ATC [1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5NS3B, 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 (Otherwise 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-Xl-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 Overiravel, or Full In.(Otherwise N/A)

[4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, 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

3 Page 41 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

2.0 SUBSEQUENT ACTIONS:

ATC [1] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise N/A):

[1.1] INITIATE ARI by Arming and Depressing BOTH of the following:

ARI Manual Initiate, 3-HS-68-119A

ARI Manual Initiate, 3-HS-68-119B

[1.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 3-9-4.

[1.3] REPORT ATWS Actions Complete and power level.

[2] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit.

[3] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5.

[4] MONITOR and CONTROL Reactor Water Level between +2 and

+51, or as directed by US, using RFP/RFPT.

3 Page 42 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRO Enters EOl-1 on Low Reactor Water Level EOI-1 (Reactor Pressure)

SRO 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 SRO RPV with the Main Turbine Bypass Valves irrespective of cooldown rate ?-

NO IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? - NO 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.

SRO Direct a Pressure Band of 800 to 1000 psig, Appendix 1 1A.

ATC/BQP Maintain directed pressure band, lAW Appendix I IA.

EOI-1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems:

HPCI Appendix I IC RC(C Appendix IIB RFPTs on minimum flow Appendix 1 1 F SRO Main Steam System Drains Appendix 1 1 D Steam Seals Appendix hG SJAEs Appendix 11 G Off Gas Preheater Appendix I I G RWCU Appendix 11 E.

3 Page 43 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

ATC/BOP Pressure Control lAW Appendixi IA, RPV Pressure Control SRVs

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. 3-PCV-1-179 MN STM LINE A RELIEF VALVE

b. 3-PCV-1-180 MN STM LINE D RELIEF VALVE.

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

d. 3-PCV-1-31 MN STM LINE C RELIEF VALVE

e. 3-PCV-1-23 MN STM LINE B RELIEF VALVE

f. 3-PCV-1-42 MN STM LINE D RELIEF VALVE

g. 3-PCV-1-30 MN STM LINE C RELIEF VALVE

h. 3-PCV-1-19 MN STM LINE B RELIEF VALVE.

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

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

k. 3-PCV-1-22 MN STM LINE B RELIEF VALVE I. 3-PCV-1-18 MN STM LINE B RELIEF VALVE

m. 3-PCV-1-34 MN STM LINE C RELIEF VALVE

3 Page 44 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

Event 8 Component: MSIV D INBD valve fails to AUTO close Crew Recognizes High Pressure Coolant Injection is feeding the Feedwater leak.

SRO Directs trip and lockout of HPCI, isolates additional leakage source.

SRO EOI-1 (Reactor Level)

Monitor and Control Reactor Level.

Verify as required PCIS isolations group (1,2 and 3), ECCS and RCIC, Directs_group_2_and_3_verified.

ATC/BOP Verifies Group 2 and 3 isolation.

SRO IF it has not been determined that the reactor will remain subcritical, THEN Exit_RCIL;_ENTER_C5_Level_I_Power Control.

SRO C5 Level I Power Control If Emergency Depressurization is required? NO -

RPV Water level cannot be determined? NO The reactor will remain subcritical without Boron under all conditions? NO PC water level cannot be maintained below 105 feet OR Suppression Chamber pressure cannot be maintained below 55 psig? NO -

CT#2 SRO Directs ADS Inhibited.

CT#2 ATCIBOP Inhibits ADS.

During ATWS recover actions MSIV D INBD valve fails to NOTE NOTE automatically close on -122 RPV water level.

The crew will recognize the isolation failure and manually close the MSIV 0 Crew INBD valve on panel 9-3.

SRO Is any Main Steam Line Open?- NO ATCIBOP Closes MSIV D INBD valve and reports to SRO.

Crew Calls for Appendix 8A and 8E. Turbine Building AUO Driver When called for Appendix 8A and 8E, wait 6 minutes. Call back and report, Driver Field actions are complete for Appendix 8A and 8E. ENTER TRIGGER 27 (bat appo8ae)

IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND a MSRV is open or cycling OR drywell pressure is above 2.4 psig_AND_RPV water_level_is_above_-162_inches?__NO Is Reactor Power above 5% ?- NO

3 Page 45 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRO Is Power above 5% or unknown? NO Is Power above 5% or unknown SRO AND RPV Water Level above -50 inches?- NO SRO Directs a Level Band with RCIC, Appendix 5C and CRD, Appendix 5B.

BOP Control Reactor Water Level with RCIC lAW 3-EOl Appendix-5C

1. IF BOTH of the following exist:

AND

. . LDDLEDLiLJü DEÜDDLLDD necessary, THEN EXECUTE EOI Appendix 16A concurrentlywith this procedure.

2. IF BOTH of the following exist:

AND

. . DDDEDDLDLD LDDLDD interlocks, THEN PERFORM the following:

a. EXECUTE EOl Appendix 16K concurrently with this procedure.

b. RESET auto isolation logic using 3-XS-71-51A(B),

RCICAUTO-ISOL LOGIC A(B) RESET pushbuttons.

3. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRIP/THROT VLV.

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.

3 Page 46 of 62 Simulator Event Guide:

Event 7 Major: FW hne break in steam tunnel; ATWS.

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCJC flow to RPV stabilizes and is controlled automatically at 620 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:

RCIC Initiation signal is NOT present, AND

RCIC flow is below 60 gpm, THEN VERIFY OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.

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

ATC/ Appendix 5B BOP 1. IF Maximum injection flow is NOT required, THEN VERIFY CRD aligned as follows:

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 3A is available, THEN VERIFY RUNNING CRD Pump 3A or 3B.

b. IF CRD Pump 3B is available, THEN VERIFY RUNNING CRD Pump 3A or 3B.

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

3-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV

3-PCV-85-27, CRD CLG WATER PRESS CONTROL VLV

3-FCV-85-50, CRD EXH RTN LINE SHUTOFF VALVE.

d. ADJUST 3-FIC-85-1 1, CRD SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WHILE maintaining 3-PI-85-13A, CRD ACCUM CHG WTR HDR PRESS, above 1450 psig, if possible.

3 Page 47 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRQ EOI-1 (Power Control)

Monitor and Control Reactor Power.

Will the reactor will remain sub subcritical without boron under all conditions?

-NO If the reactor subcritical and No boron has been injected?- NO Verify Reactor Mode Switch in Shutdown.

Initiate ARI.

ATC Initiates ARI.

SRO Verify Recirc Runback ( pump speed 480 rpm).

ATC Verifies Recirc Runback.

SRO Is Power above 5%? YES Directs tripping Recirc Pumps ATC Trips Recirc Pumps.

CT#3 SRO Before Suppression Pool temperature rises to 1 10°F, continue:

Insert Control Rods Using one or more of the following methods:

. Appendix I F

. AppendixiD DRIVER WHEN directed to perform Appendix I F and Appendix 2, wait 4 minutes and insert TRIGGER 25 and TRIGGER 26 THEN report appendix 2 complete and field action for appendix I F complete.

WHEN the Scram has been reset THEN insert TRIGGER 28 to enter bat AWlS-I

3 Page 48 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

CT#3 ATC Inserts Control Rods, lAW Appendix ID and 1 F.

ATC Insert Control Rods, lAW Appendix I F.

1. VERIFY Reactor Scram and ARI reset.

a. IF ARI CANNOT be reset, THEN EXECUTE EOl Appendix 2 concurrently with Step 1 .b of this procedure.

b. IF Reactor Scram CANNOT be reset, THEN DISPATCH personnel to Unit 3 Auxiliary Instrument Room to defeat ALL RPS logic trips as follows:

1) REFER to Attachment 1 and OBTAIN four 3-ft banana jack jumpers from EOI Equipment Storage Box.

2) REFER to Attachment 2 and JUMPER the following relay terminals in Panel 9-15, Rear:

a) Relay 5A-KIOA (DQ) Terminal 2 to Relay 5A-K12E (ED) Terminal 4 (Bay 1).

b) Relay 5A-KIOC (AT) Terminal 2 to Relay 5A-KI2G (BH) Terminal 4 (Bay 3).

3) REFER to Attachment 3 and JUMPER the following relay terminals in Panel 9-17, Rear:

a) Relay 5A-KIOB (DQ) Terminal 2 to Relay 5A-KI2F (ED) Terminal 4 (Bay 1).

b) Relay 5A-K1OD (AT) Terminal 2 to Relay 5A-K12H (BH) Terminal 4 (Bay 3).

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 SHUTOFF.

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

7. CONTINUE to perform Steps 1 through 6, UNTIL ANY of the following exists:

ALL control rods are fully inserted, OR

NO inward movement of control rods is observed, OR

SRO directs otherwise.

DRIVER DRIVER When the scram is reset, inset TRIGGER 28 (bat atws-1) to remove the ATWS condition and allow rods to insert on subsequent scram.

NOTE NOTE It will take approximately 7 minutes for the scram discharge volume to drain after the scram is reset.

ATC After scram discharge volume is drained, inserts manual reactor scram. All rods insert.

3 Page 49 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS Appendix ID NOTE: This EOI Appendix may be executed concurrently with EOI Appendix IA or lB at SRO discretion when time and manpower permit.

I. VERIFY at least one CRD pump in service.

NOTE: Closing 3-SHV-085-0586, CHARGING WATER SOV, valve may reduce the effectiveness of EQI Appendix 1A or lB.

2. IF Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SOy (RB NE, El 565 ft).

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.

NOTE: A ladder may be required to perform the following step. REFER to Tools and Equipment, Attachment 1.

6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SOV (RB NE. El 565 ft).

3 Page 50 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

CT#3 BOP/ATC Initiate SLC lAW Appendix 3A

1. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B position.

2. CHECK SLC System for 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 on Panel 3-9-5 (3-XA-55-5B, Window 20).

. 3-Pl-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

. System flow, as indicated by 3-IL-63-1 1, SLC FLOW, red light illuminated on Panel 3-9-5.

. SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 14).

3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.

4. VERIFY RWCU isolation by observing the following:

. RWCU Pumps 2A and 2B tripped.

. 3-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE closed.

. 3-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed.

. 3-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

7. MONITOR 3-Ll-63-IA, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

3 Page 51 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRO Determines RPV Level Cannot be restored and maintained above -180 inches.

SRO Determined Emergency Depressurization is Required.

CT#4 SRO Direct Terminate and Prevent lAW Appendix 4.

ATC/BOP Terminate and Prevent lAW Appendix 4 CT#4 BOP/ATC Appendix 4

1. PREVENT injection from HPCI by performing the following:

a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURBINE TRIP push button.

b. WHEN HPCI Turbine is at zero speed, THEN PLACE 3-HS-73- 47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 3-HS-73-18A, HPCI TURBINE TRIP push-button.

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

3 Page 52 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

Appendix 4 (continued)

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

NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.

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 INJ 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 NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

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

3 Page 53 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

Appendix 4 (continued)

NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

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 FEEDWATER 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 500 psig:

3-FCV-3-.19, RFP 2A DISCHARGE VALVE

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

3-FCV-3-5, REP 2C DISCHARGE VALVE

3-LCV-3-53, RFW START-UP LEVEL CONTROL

d. TRIP REPTs as necessary to prevent injection by DEPRESSING the following push-buttons:

3-HS-3-1 25A, RFPT 3A TRIP

3-HS-3-151A, RFPT 3B TRIP

3-HS-3-176A, RFPT 3C TRIP.

3 Page 54 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRO Enters C-2, Emergency RPV Depressurization Answers Yes to will the Reactor remain subcritical without Boron under all conditions Answers Yes to is Drywell Pressure above 2.4 psig Does not prevent Injection from any Core Spray or LPCI pumps because they are all needed to assure adequate core cooling Answers Yes to is Suppression Pool Level above 5.5 feet Directs opening of all ADS Valves Answers Yes to can 6 ADS Valves be opened Maintains 6 ADS Valves open until RPV cold shutdown Interlocks are clear BOP/ATC Reports Suppression Pool Level in Feet when directed by SRO Opens 6 ADS valves and verifies open when directed CT#5 When RPV Pressure is low enough for Injection of LPCI and Core Spray, operator should verify available systems are injecting.

When adequate core cooling is assured begins to throttle flow to prevent overlilling RPV.

3 Page 55 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

= BOP/ATC Appendix 6B, Loop I LPCI

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-74-155A, 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, RHR PUMP 3C SUPPR POOL SUCTVLV.

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 CHBR/POOL ISOL VLV

3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

3-FCV-74-59, RHR SYS I SUPPR POOL CLGITEST VLV

5. VERIFY RHR Pump 3A and/or 3C running.

6. WHEN 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.

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

3 Page 56 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

BOP/ATC Appendix 6C, Loop II LPCI

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 INJ 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 CLGITEST VLV.

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

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.

7. IF RPV pressure is below 230 psig, THEN 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.

3 Page 57 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

BOP/ATC Appendix 6D, Loop I Core Spray

1. VERIFY OPEN the following valves:

3.-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCTVLV

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 INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

BOP/ATC Appendix 6E, Loop II Core Spray

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 II INBD INJECT VALVE, as necessary to control_injection_at_or_below 4000_gpm_per_pump.

SRO Contacts RE to determine if the Reactor will remain subcritical under all conditions (Note 1)

3 Page 58 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

SRO Enters EOI-2 on High Drywell Pressure DWIT Monitor and control Drywell temperature below 160F using available Drywell cooling Answers No to can Drywell Temperature be maintained below 160F Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOl-1 and Scram Reactor (this will already be complete at this time)

Before Drywell Temperature rises to 280F continue Answers Yes to is Suppression Pool Level below 18 Feet Answers Yes to are Drywell Temperatures and Pressures within the safe area of curve 5 Directs Shutdown of Recirc Pumps and Drywell Blowers (should leave Drywell Blowers running due to being unable to spray because adequate core cooling is not assured)

Does not initiate Drywell Sprays Because Adequate Core Cooling is not assured at this time

3 Page 59 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

ATC/BOP Vent Containment lAW Appendix 12

1. VERIFY at least one SGTS train in service.

2. VERIFY CLOSED the following valves (Panel 2-9-3 or Panel 2-9-54):

2-FCV-64-31, DRYWELL INBOARD ISOLATION VLV, 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE.

Steps 3, 4, 5 and 6 are If I Then steps that do not apply

7. CONTINUE in this procedure at:

Step 8 to vent the Suppression Chamber through 2-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.

8. VENT the Suppression Chamber using 2-FIC-84-1 9, PATH B VENT FLOW CONT, as follows:

a. PLACE keylock switch 2-HS-84-35, DWISUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).

c. PLACE 2-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

d. PLACE keylock switch 2-HS-84-19, 2-FCV-84-19 CONTROL, in OPEN (Panel 2-9-55).

e. VERIFY 2-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.

f._CONTINUE_in_this_procedure_at_step_12.

3 Page 60 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

BOP Vents Primary Containment lAW Appendix 12

9. VENT the Suppression Chamber using 2-FIC-84-20, PATH A VENT FLOW CONT, as follows:

a. VERIFY OPEN 2-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 2-9-3).

b. PLACE keylock switch 2-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

c. VERIFY OPEN 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV (Panel 2-9-54).

d. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-55).

f. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.

g. CONTINUE in this procedure at step 12.

Steps 10 and 11 are to Vent the Drywell and will not be used since the crew will be successful at venting through the Suppression Chamber

12. ADJUST 2-FIC-84-19, PATH B VENT FLOW CONT, or 2-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:

Stable flow as indicated on controller, AND 2-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:

iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 pCi/sAND 0-SI-4.8.B.1.a.1 release fraction of 1.

DRIVER Acknowledge Notification

3 Page 61 of 62 Simulator Event Guide:

Event 7 Major: FW line break in steam tunnel. ATWS.

During ATWS recover actions D/G A will fail to automatically start on 2.45#

Crew DW pressure or -122 RPV water Level. The crew will recognize the auto start failure and manually start D/G A at panel 9-23.

SRO The Emergency Classification is 1.1-SI or I.2S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

Emergency Depressurization complete Reactor Level is restored and maintained.

All Control Rods inserted.

SHIFT TURNOVER SHEET Unit 3 is at 100% power Units I and 2 are at 90% power Equipment Out of ServicelLCOs:

CRD Pump A is out of service.

OperationslMaintenance for the Shift:

Complete Weekly EHC Pump Test per 3-OI-47A section 6.2. Lower power to 90% with Recirc Flow for a Rod Pattern Adjustment Unusual ConditionslProblem Areas:

A flash flood watch is in effect for the next six hours.

Appendix 0 Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 5- Op-Test No.: 1205 Examiners:________________ Operators: SRO:_________________

ATC:______________

BOP:_______________

Initial Conditions: The unit is at approximately 90% power. RHR Pump 2A is tagged out for scheduled maintenance. Tech Spec 3.5.1 Condition A has been entered.

Turnover: Perform RFPT 2B Overspeed Trip Exerciser Test per 2-01-3 section 8.10 and then raise power to 100% with Recirc Flow.

Event Event No. Maif. No. Type* Event Description N-BOP RFPT 2B Overspeed Trip Exerciser Test per 2-01-3 Section I N/A N-SRO 8.10.

R-ATC 2 N/A Increase reactor power to 100% using recirc flow per 2-01-68 R...SRO 3 sw02b RBCCW pump B trips. 2-FCV-70-48 fails to AUTO close.

TS..SRO 4 ogo4a Loss of SJAE, swap to standby SJAE.

C-ATC 30-23 Control rod drifts out of the core, 2-A0I-85-6. Then 5 rdO7rl 835 TS-SRO sticks when driven in, 2-01-85 and 2-A0I-85-7.

6 batch file EHC pump trip. Standby pump fails to AUTO start.

EHC leak on pump casing leads to turbine trip and scram. A 7 tcO7 M-ALL Reactor Coolant Leak inside the Drywell, during the ATWS recovery actions, will require Containment venting and sprays.

8 s102 C-ALL SLC failure to inject 2C RHR pump will fail to start and RHR Loop II select logic will 9 zdixs74l29 C-ALL fail. Operator must use RHRSW w/ RHR loop I to spray the suppression chamber and drywell.

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor FINAL

Appendix D Scenario Outline Form ES-D-1 Events

1. BOP Operator will perform RFPT 2B Overspeed Trip Exerciser Test per 2-01-3 Section 8.10.

2. Crew will raise power to 100% using Recirc Flow.

3. RBCCW pump 2B will trip. The ATC Operator will secure RWCU. 2-FCV-70-48 will fail to AUTO close on low pressure and the ATC operator will manually close the valve with the control switch on Panel 2-9-6. ATC will contact Unit 1 and align the spare RBCCW pump to Unit 2 and re-open 2-FCV-70-48. 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. The crew will respond to a loss of the in-service SJAE (A). The BOP operator will shift SJAEs to B in service per 2-01-66.

5. The ATC will recognize and respond to control rod 30-23 drifting out of the core per 2-AC 1-85-5. The ATC will select and drive the rod. When rod is being driven IN, the rod will stick. The SRO will refer to Tech Specs and determine TS 3.1.3 condition A is appropriate for one stuck rod and depending on the timing the rod may not have rod position indication when it sticks and TS 3.1.3. Condition C will be appropriate.

6. The running EHC pump will trip and the standby pump will fail to AUTO start. The BOP will start the standby EHC pump. When the standby EHC pump starts a leak will develop on the pump casing.

7. The EHC leak will force the crew to manually Scram the reactor, trip the turbine, and lock out the running EHC pump. The Turbine Bypass Valves will be unavailable. An ATWS will occur on the Reactor Scram and the crew will perform ATWS recovery actions to insert all control rods with exception of the stuck rod 18-35. During the crews response to the ATWS, a Reactor Coolant leak inside the Drywell will require Suppression Chamber Sprays and Containment Sprays to be performed.

8. A failure of RHR Loop II Select Logic and a failure of RHR Pump 2C to start will cause the crew to use RHRSW via Loop I of RHR to spray the suppression chamber and drywell. SLC will fail to inject during performance of ATWS actions.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

All Control Rods are inserted Suppression Chamber /Drywell Sprays performed using RH.RSW

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#1- With a reactor scram required and the reactor not shutdown, initiate action to reduce power by inserting control rods before containment parameters require Emergency Depressurization (DW temperature 2800 F).

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 CRD Pump B operating

3. Measured by:

Observation - Control Rod insertion commenced in accordance EOI Appendixes IA, IF, 2, ID.

4. Feedback:

Reactor Power trend.

Control Rod indications.

CT#2-. When suppression chamber pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit(DSIL) curve using RHRSW Standby Coolant.

I. 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 17B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHRSW flow to containment

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: NRC 5 8 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)

I EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 2 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix D Scenario Outline Form ES-D-1 Scenario Tasks EVENT TASK NUMBER K/A RD SRO I RFPT 2B Overspeed Trip Exerciser Test 2.1.20 4.6 4.6 2 Raise reactor power using recirc flow RD U-068-NO-17 SROS-000-NO-138 2.1.23 4.3 4.4 3 Loss of RBCCW RD U-070-AL-03 206000A2.17 3.9 4.3 SRO S-070-AB-01 4 Loss of SJAE RD U-066-NO-7 295002AA2.01 2.9 3.1 SRO S-047-AB-3 5 Control Rod Drift RD U-085-AL-12 201003A2.03 3.4 3.7 SRO S-085-AB-5 6 EHC Fluid Leak RD U-47A-AL-05 241 000A2.20 2.5 2.6 SRO S-047-AB-02 7 Reactor Coolant leak and ATWS RD U-000-EM-5 295024EA2.04 3.9 3.9 SRO S-000-EM-3 SRO S-000-EM-.5 SRO S-000-EM-15 RD U-000--EM-1 SRO S-000-EM-1 SRO S-000-EM-2

5 Page 6 of 42 Procedures Used/Referenced:

Procedure Number Procedure Title Procedure Revision 2-01-3 Reactor Feedwater System Rev 139 2-G0l-100-12 Power Maneuvering Rev 41 2-01-68 Reactor Recirculation System Rev 139 2-ARP-9-4C Alarm Response Procedure Rev 30 2-AOl-70-1 Loss of Reactor Building Closed Cooling Water Rev 29 2-01-70 Reactor Building Closed Cooling Water System Rev 63 2-01-69 Reactor Water Cleanup System Rev 106 TS 3.4.1 Recirculation Loops Operating Amd 258 2-A0l-47-3 Loss of Condenser Vacuum Rev 19 2-ARP-9-53 Alarm Response Procedure Rev 36 2-ARP-9-5A Alarm Response Procedure Rev 48 2-AOl-85-6 Rod Drift Out Rev 19 TS 3.1.3 Control Rod Operability Amd 301 2-01-85 Control Rod Drive System Rev 130 2-ARP-9-7B Alarm Response Procedure Rev 29 2-AOl-laO-I Reactor Scram Rev 97 BFN-ODM-4.20 Strategies for Successful Transient Mitigation Rev 0 2-01-92 Source Range Monitors Rev 21 2-OI-92A Intermediate Range Monitors Rev 28 2-01-47 Turbine Generator System Rev 163 2-01-64 Primary Containment System Rev 116 2-EOl-1 RPV Control Flowchart Rev 12 2-EOl Appendix-3A SLC Injection Rev 5 2-E0l-2-C-5 Level Power Control Flowchart Rev I I 2-EOI Appendix-i 1A Alternate RPV Pressure Control Systems MSRVs Rev 4 RPV Low Low Low Level Rev 3 2-EOl Appendix-8A

. Bypassing Group 6 Low RPV Level and Rev 2 2-EOl Appendix-8E .

High Drywell Pressure Isolation Interlocks

5 Page 7 of 42 Procedures Used/Referenced:

Procedure Number Procedure Procedure Title Revision 2-EQI Appendix-5A Injection Systems Lineup Condensate/Feedwater Rev 9 2-EOl Appendix-2 Defeating ARI Logic Trips Rev 4 Insert Control Rods Using Reactor Manual Control 2-EOI Appendix-I D Rev 6 System Removal and Replacement of RPS Scram Solenoid 2-EOI Appendix-lA Rev 6 Fuses 2-EOl-2 Primary Containment Control Flowchart Rev 10 2-EOI Appendix-i 2 Primary Containment Venting Rev 4 RHR System Operation Suppression Chamber 2-EOl Appendix-I7C Rev I I Sprays 2-EOI Appendix-17B RHR System Operation Drywell Sprays Rev 10 2-EOI Appendix-6A Injection Subsystems Lineup Condensate Rev 4 EPIP-1 Emergency Classification Rev 47 EPIP-4 Site Area Emergency Rev 32

5 Page 8 of 42 Simulator Instructor 1C28 prefINRCIl 205-5 BatchINRCIl 205-5 pfk 01 tog bat atws80 pfk 02 ann silence bat rhra pfk 03 bat NRC/I 205-5 mf rhOlc pfk 04 imf sw02b ior zdihs74l6a null pfk 05 imf ogo4a imf rpO7 pfk 06 imf rd04r3023 imfrpl4b pfk07dmfrd04r3023 ior zdihs68ll9apl asis pfk 08 imf rd06r3023 trg elO NRC/modesw pfk 09 bat NRC/ehcpumptrip trg eIO= imfth22 100 pfk 10 imftc07 20 180 trg eIO= imfth2l 0.15 pfk 11 bat appalaout trg eIO= mmftc07 100 pfk 12 bat appOlain br zdihxs74l29 asis pfksl morypovfcv74loo norm bat NRC/7048ftc pfk s2 mrf swO9 open trg el NRC/HS7048 pfk s3 bat appo2 trg eI= bat NRC/7048-1 pfk s4 bat appOif trg e5 NRC/ehc pfk s5 bat appo8ae trg e5= bat NRC/ehcpumptrip-I pfk s6 bat sdv imf slO2 pfk s7 bat atws-1 Scenario 5 DESCRIPTIONIACTION Simulator Setup manual Reset to IC 1 10 Simulator Setup Load prefs restorepref NRC/I 205-5 Simulator Setup Load Batch F3 (load batch file)

Simulator Setup Verify file loaded Simulator Setup manual Clearance out RHR Pump 2A Procedures:

RCP required (90% 100% with flow)

RCP for Urgent Load Reduction

Provide marked up copy of 2-GOI-100-12

5 Page 9 of 42 Simulator Event Guide:

Event 1 Normal: RFPT 2B Overspeed Trip Exerciser Test SRO Directs BOP to perform RFPT 2B Overspeed Trip Exerciser, 2-01-3 section 8.10 8.10 Overspeed Trip Exerciser Test NOTES

1) The following steps will test the circuitry associated with the overspeed trip test and mechanical overspeed trip lockout. The test may be performed at speed or stopped; it should be performed prior to rolling the turbine off the turning gear and periodically with the applicable REP in operation. This test will normally be performed during scheduled reductions in power with reactor power less than 90%.

2) The success of this test depends on the mechanical overspeed trip bolt actuating as a result of oil pressure. The system was designed for this test to be performed with the RFPT at normal operating speed. The test may not function properly at less than 75% speed (4125 rpm) due to the trip mechanism being sluggish at speeds well below rated. If this is the case, no action is required except to reperform the test at a higher speed.

3) The following steps are performed at Panel 2-9-6.

BOP Performs RFPT 2B Overspeed Trip Exerciser Test, 2-01-3 section 8.10

[1] OBTAIN Unit Supervisor approval to perform this test.

[2] For REPT being tested, VERIFY Main Oil Pump running:

. RFPT 2B 2B1(2B2) MAIN OIL PMP

[3] PLACE RFPT (2B) OVERSPEED TEST TRIP LOCKOUT switch, 2-HS-3-(135A), in MECH.

[3.1] CHECK the following:

. Green (normal) light extinguished at switch.

. Amber (mechanical lockout) light to right of green light ill urn i nated.

[4] DEPRESS and HOLD RFPT (2B) OVERSPEED TEST pushbutton, 2-HS-3-(1 36).

[4.1] CHECK the following:

. White (trip) light illuminated at pushbutton.

__Green_(normal)_light_extinguished_at_pushbutton.

5 Page 10of42 Simulator Event Guide:

Event I Normal: RFPT 2B Overspeed Trip Exerciser Test BOP [5] RELEASE RFPT (2B) OVERSPEED TEST pushbutton.

[6] DEPRESS and HOLD RFPT (2B) OVERSPEED TEST RESET pushbutton, 2-HS-3-(132).

[6.1] CHECK the following:

. White (trip) light illuminated at 2-HS-3-(1 32) pushbutton.

. White (trip) light extinguished at 2-HS-3-(136) pushbutton.

[7] RELEASE RFPT (2B) OVERSPEED TEST RESET pushbutton, 2-HS (132).

[7.1] CHECK the following:

. White light extinguished at 2-HS-3-(1 32) pushbutton.

. Green (normal) light illuminated at 2-HS-3-(136) pushbutton.

NOTE Waiting 30 seconds before placing Overspeed Test Trip Lockout switch in normal position in Step 8.1 0[8] allows ample time for Overspeed test trip logic to reset.

[8] WHEN 30 seconds has elapsed, THEN PLACE (2B)RFPT OVERSPEED TEST TRIP LOCKOUT switch, 2-HS-3-(135A), in NORM.

[8.1] CHECK the following:

. Green (normal) light illuminated at switch.

. Amber (mechanical lockout) light to right of Green light is extinguished.

BOP [9] NOTIFY Unit Supervisor when test is complete.

BOP Informs SRO that RFPT 2B Overspeed Trip Exerciser Test Completed Satisfactorily

5 Page 11 of 42 Simulator Event Guide:

Event 2 Reactivity: Raise Power to 100% with Recirc Flow SRO Notifies ODS of power increase.

Directs Power increase using Recirc Flow, per 2-G0I-100-12.

[21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:

RAISE power using control rods or core flow changes.

REFER_TO_2-SR-3.3.5(A)_and_2-01-68.

2-01-68 Reactor Recirc Precaution and Limitation 3.5.3:

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).

SRO Personally oversee reactivity changes, lAW OPDP-1, Section 3.6.

ATC Raise Power w/Recirc, lAW 2-01-68, Section 6.2

[1] IF desired to control Recirc Pumps 2A and/or 2B speed with Recirc Individual Control, THEN PERFORM the following;

Raise Recirc Pump 2A using, RAISE SLOW (MEDIUM),

2-HS-96-1 5A(1 5B).

ANDIOR

Raise Recirc Pump 2B using, RAISE SLOW (MEDIUM),

2-HS-96-1 6A(1 6B).

[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 push buttons as required:

RAISE SLOW, 2-HS-96-31 RAISE MEDIUM, 2-HS-96-32 BOP Provides Peer Check for reactivity change Driver Wh&i directed by NRC, insert F4 for RBCCW pump 2B trip with 2-FCV-70-48 FTC

5 Page 12 of 42 Simulator Event Guide:

Event 3 Component: RBCCW pump B trips. 2-FCV-70-48 fails to AUTO close.

Driver When directed by NRC, insert F4for RBCCW pump 2B trip with 2-FCV-70-48 FTC Responds to alarm 4C-12, RBCCW PUMP DISCH. HDR PRESS LOW BOP/ATC Report Trip of RBCCW Pump 2B.

BOP/ATC Automatic Action: Closes 2-FCV-70-48, non-essential loop, closed cooling water sectionalizing MOV.

A. VERIFY 2-FCV-70-48 CLOSI NGICLOSED.

B. VERIFY RBCCW pumps A and B in service.

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 2-AOI-70-1 for RBCCW System failure and 2-01-70, for starting spare pump.

SRO Enters 2-AOI-70-1, Loss of Reactor Building Closed Cooling Water.

ATC Closes 2-FCV-70-48 and report the sectionalizing valve failed to close automatically SRO Contacts Maintenance Shift Manager to investigate failure of sectionalizing valve to close.

BOP Dispatch Personnel to investigate RBCCW Pump 2B trip ATC 2-AOI-70-1 Loss of Reactor Building Closed Cooling Water 41 Immediate Actions

[1] IF RBCCW Pump(s) has tripped, THEN Perform the following

SECURE RWCU Pumps.

VERIFY RBCCW SECTIONALIZING VLV, 2-FCV-70-48 CLOSED. (NO)

ATC Secures RWCU Pumps and Closes 2-FCV-70-48.

4.2 Subsequent Actions

[1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, SRO AND core flow is above 60%,THEN: (Otherwise N/A):

[2] IF any EOl entry condition is met, THEN ENTER appropriate EQI(s) (Otherwise N/A).

5 Page 13 of 42 Simulator Event Guide:

Event 3 Component: RBCCW pump B trips. 2-FCV-70-48 fails to AUTO close.

Steps I and 2 (on previous page) are NA

[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).

Driver When dispatched, report RBCCW Pump 2B breaker is tripped. There is also a smell of burnt wiring and charring on the breaker.

SRO [4] IF unable to restart a tripped pump, THEN PLACE Spare RBCCW Pump in service. REFER TO 2-01-70. Direct Unit I to place Spare RBCCW Pump in service Driver When called to place spare RBCCW Pump in service, wait 3 minutes (mrf swO2 align). THEN inform Unit 2 Operator that spare RBCCW Pump is in service.

SRO [5] IF RBCCW flow was restored to two pump operation by placing the Spare RBCCW pump in service in the preceding step, THEN PERFORM the following:

[5.1] REOPEN RBCCW SECTIONALIZING VLV, 2-HS-70-48A.

[5.2] RESTORE the RWCU system to operation. (REFER TO 2-01-69)

Directs ATC or BOP to Open Sectionalizing Valve and Restore RWCU.

NRC Note to NRC: ATC not expected to perform restoration of RWCU ATC Opens Sectionalizing Valve, 2-FCV-70-48.

SRO Determines RBCCW SECTIONALIZING VALAVE, 2-HS-48-70 is NOT in Appendix R

SRO References TR 3.4.ICOOLANT CHEMISTRY. With RWCU out of service determines the TSR3.4.1.1 is no longer met by continuous monitoring of reactor coolant conductivity.

SRO Calls Chemistry to commence sampling for reactor coolant conductivity

5 Page 14 of 42 Simulator Event Guide:

Event 3 Component: RBCCW pump B trips. 2-FCV-70-48 fails to AUTO close.

TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.1.1 -

Continuously Not required when there is no fuel in the reactor vessel.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when Monitor reactor coolant conductivity, the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or 5 OR 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the continuous conductivity monitor is inoperable and the reactor is in MODE 4 or 5 Driver When directed by NRC, insert F5 for loss of SJAE

5 Page 15of42 Simulator Event Guide:

Event 4 Component: Loss of SJAE, swap to standby SJAE.

Driver When directed by NRC, insert F5 for loss of SJAE SRO Enters AOI-47-3 Loss of Condenser Vacuum.

BOP Offgas Panel 9-53 Alarms:

Window 4, OG HOLDUP LINE INLET FLOW LOW:

Operator action:

VERIFY OPEN, FCV-66-28, off-gas system isolation valve.

VERIFY that SJAE auto isolation has NOT occurred.

Window 10, H2 WATER CHEMISTRY ABNORMAL:

Operator action:

None at this time Window 20, H2 WATER CHEMISTRY SHUTDOWN:

Operator action:

None at this time BOP Swaps to B SJAE lAW 2-AOl-47-3 Loss of Condenser Vacuum.

BOP 4.2 Subsequent Actions (continued)

[11] IF a failure of the in-service SJAE is indicated, THEN PLACE the standby SJAE in service as follows:

NOTES

1) This section may be used to return either SJAE to service following a shutdown or an isolation.

2) Potential causes of PCV valve closure are:

. Condensate pressure from SJAE A(B) less than 60 psig, 2-Pl-2-34(40),

Panel 25-105.

. SJAE 2A(28) CONDENSATE INLET VALVE closed at 2-HS-2-31A(36),

Panel 2-9-6.

. SJAE 2A(2B) CONDENSATE OUTLET VALVE closed at 2-HS-2-35A(41A),

Panel 2-9-6.

. STEAM TO SJAE A(B) STAGE I & II, 2-Pl-1-150(152), Panel 25-105 is less than 155 psig. (disabled for the SJAE selected by 2-HS-001-0375)

. Loss of l&C bus A(B), power is required to be restored to return the SJAE to service.

3) 2-HS-001-0375, SJAE TRAIN PERMISSIVE, should be placed in the position for the SJAE being placed in service. This switch will normally be in the position of the standby SJAE.

5 Page 16 of 42 Simulator Event Guide:

Event 4 Component: Loss of SJAE, swap to standby SJAE.

SRO Contact Maintenance Shift Manager to investigate SJAE failure BOP

[11.1] PLACE SJAE TRAIN PERMISSIVE 2-HS-001-0375 in the position for the SJAE being placed in service. This switch will normally be in the position of the Standby SJAE. (Panel 925-1 05 on junction box 8595) (N/A if Placing the standby SJAE in service)

[11.2] VERIFY off gas isolation is reset, using OG OUTLET/DRAI N ISOLATI ON VLVS, 2-HS-90-1 55, Panel 2-9-8.

[1 1 .3] VERIFY the following valves are OPEN:

. SJAE 2A(2B) INLET VALVE, 2-HS-66-11(15),

Panel 2-9-8

. STEAM TO SJAE 2A(2B), 2-.HS-1-155A(156A),

Panel 2-9-7 BOP [11.4] VERIFY SJAE 2A(2B) OG OUTLET VALVE, 2-HS-66-1 4(18), AUTO/OPEN (Panel 2-9-8)

[11.5] PLACE SJAE 2A(2B) PRESS CONTROLLER 2-HS-1-150(152) in CLOSE and then in OPEN at Panel 2-9-7.

[1 1 .6] VERIFY the following valves OPEN (red lights illuminated) at Panel 2-9-7.

. STEAM TO SJAE 2A(2B) STAGES I ,2, AND 3, 2-PCV-1-151/166 (153/167).

. SJAE 2A(2B) INTMD CONDENSER DRAIN 2-FCV-1 -150(152).

[11.7] MONITOR hotwell pressure as indicated on HOTWELL PRESS AND TEMP recorder, 2-XR-2-2 (Panel 2-9-6).

[11.8] For the SJAE not being placed in service,

. VERIFY CLOSED SJAE 2B(2A) OG OUTLET VALVE, 2-HS-66-18(14) (Panel 2-9-8).

. VERIFY CLOSED SJAE 2B(2A) PRESSURE CONTROLLER, 2-HS-1-152(150) (Panel 2-9-7)

[11.9] VERIFY SJAE TRAIN PERMISSIVE, 2-HS-001-0375, in the position for the SJAE selected for Standby operation SJAE A(SJAE B). (Panel 925-1 05 on junction box 8595)

5 Page 17 of 42 Simulator Event Guide:

Event 4 Component: Loss of SJAE, swap to standby SJAE.

Note NRC Candidate may use Hard Card from 2-01-66, Appendix B to place spare SJAE in service (procedure below)

BOP Standby SJAE System Lineup Hard Card 1.0 OPERATOR ACTIONS NOTES Radiation Protection should be notified prior to placing a SJAE in service. If time does not permit this due to plant conditions then notification should be made when possible. 2-HS-OO1-0375, SJAE TRAIN PERMISSIVE (located on 2-LPNL-925-0105, U2 TB, el 586) should normally be in the position of the standby SJAE. If problems are encountered while placing a SJAE in service and time permits, operate this switch as required during the performance of this section.

[1] VERIFY RESET 0ff-Gas isolation using 2-HS-90-155, OG OUTLET/DRAIN ISOLATION VLVS.

[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 AUTOIOPEN 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 (1 53/1 67), STEAM TO SJAE 2A(2B) STAGES 1,2, AND 3.

2-FCV-1-1 50(1 52), 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)

5 Page 18of42 Simulator Event Guide:

Event 5 Component: 30-23 Control rod drifts out of the core, then sticks when driven in.

Driver When directed by NRC, insert F6 for Control Rod 30-23 drift out ATC Announces and responds to alarm 9-5A window 28, Control Rod Drift A. DETERMINE which rod is drifting from Full Core Display.

ATC Identifies Control Rod 30-23 is drifting out of the core and announces SRO Directs performance of ARP and direction that if 2 rods are drifting, a Manual Reactor Scram will be required B. IF no control rod motion is observed, THEN RESET rod drift as follows.

1. PLACE ROD DRIFT ALARM TEST switch, 2-HS-85-3A-S7, in RESET and RELEASE.

2. RESET the annunciator.

C. IF rod drifting in, THEN REFER TO 2-AOl-85-5 and 2-AOI-85-7.

D. IF rod drifting out, THEN REFER TO 2-AOl-85-6 and 2-AOl-85-7 SRO Directs entry into 2-AOl-85-6, Rod Drift Out ATC Responds per 2-AOI-85-6 F. REFER TO Tech Spec 3.1.3, 3.10.8.

2-AOI-85-6, Rod Drift Out 4.0 OPERATOR ACTIONS CAUTION

[NRC/C] Operations outside of the allowable regions shown on the Recirculation System Operating Map could result in thermal-hydraulic power oscillations and subsequent fuel damage.

4.1 Immediate Actions

[1] IF multiple control rod drifts are identified, THEN MANUALLY SCRAM the reactor and enter 2-AOl-I 00-1.

BOP Identifies that Rod 30-23 is the only rod drifting and does not Manually Scram the Reactor SRO Dispatches AUO and SRO with heat gun to measure HCU temperatures.

Request Reactor Engineer to come to the control room.

Contact Maintenance Shift Manager for support troubleshooting control rod.

5 Page 19 of 42 Simulator Event Guide:

Event 5 Component: 30-23 Control rod drifts out of the core, then sticks when driven in.

SRO Contact System 85 Engineer.

ATC 4.2 Subsequent Actions

[1] IF a Control Rod is moving from its intended position without operator actions, THEN SELECT the drifting control rod and INSERT to the FULL IN (00) position.

ATC Selects Control Rod 30-23 and attempts to insert the control rod NA [2] IF control rod drive does NOT respond to INSERT signal, THEN PERFORM the following: (Otherwise N/A)

[2.1] REDUCE Total Core Flow, as indicated on TOTAL CORE FLOW/CORE PRESS DROP, 2-XR-68-50 on Panel 2-9-5, by approximately 10% to control possible power increase.

[2.2] IF drifting control rod is causing Reactor power to rapidly rise at a rate which can NOT be controlled by reducing recirculation flow, THEN MANUALLY SCRAM_the_Reactor.(Otherwise_N/A)

Driver When ATC selects rod 30-23 and gives an insert signal delete the rod drift out malfunction by pressing F7, then when rod gets to position 08, insert a stuck rod malfunction by pressing F8.

[3] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

[4] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 2-AOl-ba-i.

SRO/ATC Monitors for a second Control Rod Drift and contacts Reactor Engineer to evaluate Core Thermal Limits ATC Recognizes Control Rod 30-23 sticks at approximately position 08 and announces Stuck Control Rod SRO Directs ATC to perform actions for Control Rod Difficult to Insert in 2-01-85 ATC 2-01-85, Control Rod Drive System 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.

[2] 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.

5 Page 20 of 42 Simulator Event Guide:

Event 5 Component: 30-23 Control rod drifts out of the core, then sticks when driven in.

ATC [4] CHECK CRD SYSTEM FLOW is between 40 gpm and 65 gpm, indicated by 2-FIC-85-1 1.

[5] CHECK CRD DRIVE WTR HDR DP, 2-PDI-85-17A is between 250 psid and 270 psid.

[6] IF the CRD SYSTEM FLOW or CRD DRIVE WTR HDR DP had to be adjusted, THEN PROCEED TO Section 6.7.

ATC Checks all pressures and flows associated with the CRD system and determines that they_are_within_the_required_range NA [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 FLUSH the control rod by placing the CRD CONTROL SWITCH, 2-HS-85-48, in ROD IN, for several minutes OR until the control rod begins to insert.

[9] IF the control rod begins to insert normally, THEN PROCEED TO Section 6.7.

NA [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 PROCEED TO Section 8.8 to vent the HCU, and RETURN to Step 8.16[11].

[11] ATTEMPT to notch rod in using CRD CONTROL SWITCH, 2-HS-85-48.

[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.

ATC Contacts the Reactor Engineer to refer to Stuck Control Rod-Test to determine if there is a hydraulic problem or if the Rod is mechanically bound

5 Page 21 of 42 Simulator Event Guide:

Event 6 Component: EHC pump trip. Standby pump fails to AUTO start.

SRO Refers to Tech Spec 3.1.3 and determines Condition A applies Technical Specification 3.1.3 Condition A Required Action Al Completion Time Immediately Required Action A2 Completion Time 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Required Action A3 Completion Time 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Required Action A4 Completion Time 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months If the malfunction is inserted such that the rod sticks between reed switches, it will

NOT have indication and the SRO may declare the control rod INOP and enter NOTE Condition C and Condition E since the Condition C Required Actions cannot be completed.

Condition C Required Action Cl Completion Time 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Required Action C2 Completion Time 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Condition E Required Action El Completion Time 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Driver When directed by NRC: insert F9 for EHC pump trip, with standby pump fail to start

5 Page 22 of 42 Simulator Event Guide:

Event 6 Component: EHC pump trip. Standby pump fails to AUTO start.

Diver When directed by NRC, insert F9 for EHC pump trip, with standby pump fail to start BOP Announces Alarm 9-7B, Window 15, Standby EHC Pump Failed SRO Directs response per ARP A. On Panel 2-9-7:

1. VERIFY alarm by checking EHC HEADER PRESSURE indicator, 2-Pl-47-7.

2. VERIFY EHC PUMP 2B, 2-HS-47-2A and/or EHC PUMP 2A, 2-HS-47-IA running.

Driver 30 seconds after BOP starts EHC Pump 2B, insert FlO for unisolable EHC leak on a 180 second ramp BOP Determines that EHC Pump 2A is tripped and that EHC Pump 2B failed to start automatically. BOP starts 2B EHC Pump and verifies pressure returns to normal

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.

BOP NOTE On EHC Hydraulic System failure accumulator and check valve arrangement will provide approximately one minute bypass valve operation.

C. IF EHC Hydraulic System fails, THEN VERIFY turbine trips at or below 1 100 psig.

Driver When dispatched to the EHC pump skid, wait 3 minutes and report that oil is spraying_from_the_casing_of the_2B_EHC_Pump_and_it_cannot_be_isolated SRO Determines that EHC Fluid Pressure will soon be lost and a Manual Scram and Turbine trip must be inserted. May direct Core Flow Runback if time permits.

BOP Calls Radwaste to lockout Turbine Building sumps Driver Ensure trigger 10 goes active on the MODESWITCH

5 Page 23 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram SRO Directs ATC to insert manual Reactor scram and BOP to trip the turbine and lock out the EHC pumps ATC Performs actions of OATC Hard card Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS

[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AIS3A and 2-HS 5A/S3B, on Panel 2-9-5.

ATC Depresses Reactor Scram Pushbuttons and determines RPS did not de-energize NA [2] IF scram is due to a loss of RPS, THEN

[3] Refuel_Mode_One_Rod_Permissive_Light_check

[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1 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 ATC Places Reactor Mode switch in Shutdown and determines RPS still has not de energized. Performs Scram report and proceeds to subsequent actions of Hard Card 2.0 SUBSEQUENT ACTIONS:

[1] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise N/A):

[1 .1] INITIATE ARI by Arming and Depressing BOTH of the following:

CT#1

ARI Manual Initiate, 2-HS-68-119A

ARI Manual Initiate, 2-HS-68-119B

[1.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 2-9-4.

[1.3] REPORT_ATWS Actions_Complete_and_power_level.

ATC Initiates both channels of ARI and determines rod movement, Runs Recirc Pumps CT#1 back to minimum speed, Announces ATWS Actions complete and Reactor Power is approximately 80%.

Driver After ARI is initiated and some rods have inserted insert Shift F6 (bat sdv)

5 Page 24 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram

[2] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.

[3] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 2-9-5.

[4] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, using RFPIRFPT.

[5] RETURN to body of procedure at step 4.2[5] AND CONTINUE with actions as required.

ATC Drives in IRMs and SRMs as time permits, verifies SDV vents and drains are closed, and monitors and controls Reactor Water Level as directed by US BOP Performs actions of BOP Reactor Scram Hard Card Reactor Scram BOP Unit Operator Hard Card 1.0 SUBSEQUENT ACTIONS: PANELS 2-9-7 & 2-9-8 NOTES

1) The following steps are not required to be performed in order, but only as required to maintain stable conditions.

2) It is desired to trip the turbine prior to receiving the GEN REVERSE PWR FIRST RELAY OPERATION 2-EA-57-136 (2-XA-55-8A, Window 7) alarm to avoid motorizing the generator.

[1] At 50 MWe, or as directed by the Unit Supervisor, VERIFY TRIPPED the Main Turbine as follows:

[1.1] DEPRESS the TRIP pushbutton, 2-HS-47-67D on Panel 2-9-7.

[1.2] VERIFY OPEN Generator Output Breaker, by placing GENERATOR PCB 224, 2-HS-242-0224A, to TRIP.

[1.3] IMMEDIATELY PLACE VOLTAGE REGULATOR START/STOP SEL, 2-HS-57-24, to STOP and release.

[1.4] CHECK the following at 2-HS-57-24:

. GREEN light illuminated

. RED light extinguished

[2] ANNOUNCE Reactor SCRAM over PA system.

BOP Trips the Main turbine and locks out the EHC pumps, opens the generator breaker and places the voltage regulator in the STOP position, and announces Reactor Scram over the PA system

[1] MONITOR and CONTROL RPV pressure to keep below 1073 psig and stable.

[1.1] IF RPV pressure is lowering rapidly, THEN NA NA [1.2] IF MSRVs are cycling and bypass valves are available, THEN

5 Page 25 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram

[1.3] IF MSRVs are cycling and bypass valves are NOT available, THEN MANUALLY OPEN MSRVs on Panel 2-9-3 until RPV pressure is controlled between 800 and 1000 psig.

BOP Manually opens MSRVs on panel 9-3 to control Reactor Pressure 800-1000 psig until further direction is provided by the US

[2] IF any PCIS isolation signal is received, THEN VERIFY PCIS isolations using any of the following: (Otherwise N/A)

. Containment Isolation Status System on Panel 2-9-4

. PCIS Mimic and individual control switch indications

. CS

. 2-01-64

[3] IF HPCI and/or RCIC are in service and injecting to the vessel, THEN MONITOR and CONTROL Reactor Water Level as necessary. (Otherwise N/A)

BOP Verifies PCIS isolation signals received and reports condition of PCIS and HPCI/RCIC to US SRO Enters EOI-1 on Scram condition and Reactor Power above 5%

RCIQ Monitor and Control Reactor Power.

Verify Rx Mode Switch in shutdown Done -

Initiate ARI Done Verify RR pumps run back to 480 RPM or less Done If Rx Power >5%, trips RR pumps (SRO directs ATC to trip RR pumps)

Before SP temp rises to 1 1 OF, initiate SLC (SRO directs ATC, if necessary)

RCIP Monitor and Control RPV Pressure.

Answers NO to: Is any MSRV cycling? (BOP has already opened SRVs as necessary)

Directs BOP to maintain RPV Pressure 800 -1000 psig using Appendix hA.

RCIL Monitor and Control RPV Water Level.

Verify as Required:

. PCIS Isolations (Groups 1, 2 and 3)

. ECCS

. RCIC SRO exits RC/L and enters C-5, because the Reactor will not remain subcritical_without_Boron_under_all_conditions_(Note_1)

ATC Trips Recirc pumps, injects SLC by direction of SRO iaw with EOl-Appendix-3A, if necessary

5 Page 26 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram SRO Directs initiation of SLC if Reactor Power is Greater than 5% when step RCIQ-1 0 is reached of EOI-1 (this step should be NA based on validation results)

NRC SLC may be initiated depending what power is when step RCIQ-1 0 is reached.

Based on Scenario validation, Reactor Power should be less than 5% power when step RCIQ-1 0 is reached. BFN-ODM-4.20, Strategies for Successful Transient Mitigation, section 4.7.3, step 0 states, When EOI-1, Step RC/Q-1 0 is reached, IF reactor power is greater than APRM downscale, THEN INITIATE SLC.

ATC Initiate SLC lAW Appendix 3A, if necessary

1. UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B position.

2. CHECK SLC System for 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 on Panel 2-9-5 (2-XA-55-5B, Window 20).

. 2-Pl-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

. System flow, as indicated by 2-IL-63-11, SLC FLOW, red light illuminated on Panel 2-9-5,

. SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 14).

3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.

4. VERIFY RWCU isolation by observing the following:

. RWCU Pumps 2A and 2B tripped

. 2-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE closed

. 2-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed.

. 2-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

7. MONITOR 2-Ll-63-1A, SLC STORAGE TANK LEVEL, and CHECK that_level_is_dropping_approximately_1%_per_minute.

NOTE NOTE SLC will fail to initiate and will not be obvious to the operator, there is no action for the operator to take to correct the SLC failure to inject

5 Page 27 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Pressure Control lAW AppendixilA, RPV Pressure Control SRVs

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. 2-PCV-1-179 MN STM LINE A RELIEF VALVE

b. 2-PCV-1-180 MN STM LINE D RELIEF VALVE.

c. 2-PCV-1-4 MN STM LINE A RELIEF VALVE

d. 2-PCV-1-31 MN STM LINE C RELIEF VALVE

e. 2-PCV-1-23 MN STM LINE B RELIEF VALVE

f. 2-PCV-1-42 MN STM LINE D RELIEF VALVE

g. 2-PCV-1-30 MN STM LINE C RELIEF VALVE

h. 2-PCV-1-19 MN STM LINE B RELIEF VALVE.

i. 2-PCV-1-5 MN STM LINE A RELIEF VALVE.

j. 2-.PCV-1-41 MN STM LINE D RELIEF VALVE

k. 2-PCV-1-22 MN STM LINE B RELIEF VALVE I. 2-PCV-1-18 MN STM LINE B RELIEF VALVE

m. 2-PCV-1-34 MN STM LINE C RELIEF VALVE SRO Enters C-5, based on override RC/L-3 from EOI-1 Calls Maintenance Shift Manager for support on restoration of RHR C, CS Logic and SLC

5 Page 28 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram SRO IF It has not been determined that the reactor will remain subcritical, THEN Exit RC/L; enter C5 Level I Power Control.

IF Emergency Depressurization is required? NO -

IF RPV Water level cannot be determined? NO The Reactor will remain subcritical without Boron under all conditions - NO IF PC water level cannot be maintained below 105 feet OR Suppression Chamber pressure cannot be maintained below 55 psig? - NO SRO Directs ADS Inhibited.

ATC/BOP Inhibits ADS.

SRO Is any Main Steam Line Open? - YES Direct Bypass of isolation interlocks, Appendix 8A and Appendix 8E.

Crew Calls for Appendix 8A and 8E.

Driver When called for Appendix 8A and 8E, wait 6 minutes. Call back and report, Field actions are complete for Appendix 8A and 8E. ENTER Shift F5 (bat app08ae)

ATCIBOP Appendix 8A

3. Operator to verifies closed the following valves (Unit 2 Control Room, Panel 9-3):

2-FCV-43-13, RX RECIRC SAMPLE INBD ISOLATION VLV 2-FCV-43-14,_RX_RECIRC_SAMPLE_OUTBD_ISOLATION VLV.

SRO C5 Level I Power Control Crosstie CAD to DW Control Air if necessary (Appendix 8G). - NOT Necessary IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND An MSRV is open or cycling or drywell pressure is above 2.4 psig AND RPV water_level_is_above_(-)_162_inches?_-_NO SRO Is Reactor Power above 5%? NO Maintain RPV Water Level between -180 inches and +51 inches with Condensate and_Feedwater App_5A_and/or_HPCI_App_SD SRO Directs RPV Water level maintained +2 to +51 inches with Condensate and Feed iaw EOI App-5A

5 Page 29 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram ATC Performs actions necessary to maintain RPV Water Level +2 to +51 inches iaw EOI App-5A Appendix-5A

1. IF It is desired to use a reactor feed pump that is in operation, THEN CONTINUE at step 12 to control the operating pump.

12. SLOWLY ADJUST RFPT speed UNTIL feedwater flow to the RPV is indicated, using ANY of the following methods on Panel 2-9-5:

. Individual 2-HS-46-8A(9A)(1OA), RFPT 2A(2B)(2C) SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

. Individual 2-SIC-46-8(9)(1O), RFPT 2A(2B)(2C) SPEED CONTROL in MAN UAL, OR

. 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 2-SIC-46-8(9)(1O), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO.

13. ADJUST RFPT speed as necessary to control injection using the methods of step 12.

14. WHEN RPV level is approximately equal to desired level AND automatic level control is desired, THEN PLACE 2.-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 2-SIC-46-8(9)(1 0), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO.

CT#1 SRO Directs actions necessary to insert rods jaw EOI-1, steps RCIQ-20 and 21 SRO Directs ARI reset and ARI logic trips defeated jaw with EOI App-2,DEFEATING ARI LOGIC TRIPS, and EOI App-lA, REMOVAL AND REPLACEMENT OF RPS SCRAM SOLENOID FUSES. Directs ATC to drive control rods until scram can be reset and SDV have drained Driver When scram is reset insert Shift F7 (bat atws-1)

Driver When directed to perform Appendix 2 wait 3 minutes and insert Shift F3 (bat appo2) then report completion.

5 Page 30 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram ATC Appendix ID

1. VERIFY at least one CRD pump in service.

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

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 EM ERG 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.

6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER SOV (RB NE, El 565 ft).

Driver If directed to close 2-FCV-85-586, wait 3 minutes, then insert mrf rdO6 close and report completion.

lfiWhen directed to re-open 2-FCV-85-586, wait 3 minutes, then insert mrf rdO6 open and report completion.

ATC When report received that Appendix-2 is complete, reset the Reactor Scram and verify SDV vents and drains open to drain the SDV Driver When scram is reset insert Shift F6 (bat sdv) and Shift F7 (bat atws-1)

SRO When SDV alarms 9-4A windows I and 29 have cleared call the OSUS and direct Appendix-IA completed to de-energize RPS and attempt to insert rods Driver When called to complete Appendix-IA, wait 3 minutes and insert Fli (bat appOl out)

CT#1 ATC Determine that all rods inserted upon completion of Appendix-IA and announce to the SRO SRO Upon determination that all rods have been inserted, SRO exits C-5 and enters RC/L from EOI-1 based on override step C5-I and exits RC!Q based on step RC/Q-2 and enters AOl-i 00-I Driver When called to re-install fuses RPS fuses per Appendix IA, wait 1 minute and insert F12 (bat appOlin)

5 Page 31 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram NRC A leak inside the Drywell develops shortly after the Reactor is Scrammed, EOl-2 will be entered on Drywell pressure and all legs executed concurrently Executes all legs of EOl-2 concurrently EOl-2 DWIT Monitor and control Drywell Temperature below 160°F, using available Drywell Cooling.

Answers NO to: Can Drywell Temperature be maintained below 160°F?

PCIP SRO Monitor and control Primary Containment pressure below 2.4 psig using the Vent System (Appendix 12) as necessary Direct Appendix 12 Answers No to can Primary Containment Pressure be maintained below 2.4 psig Before Suppression Chamber Pressure rises to 12 psig Initiate Suppression Chamber Sprays using only those pumps not required for Adequate Core Cooling Directs Suppression Chamber Sprays EOI-2 PCIH Monitor and control Drywell and Suppression Chamber:

. Hydrogen at or below 2.4%

AND

. Oxygen at or below 3.3%

SRO Using the Nitrogen Makeup System (APPX 14A).

EOI-2 SPIT Monitor and control Suppression Pool temperature below 95°F, using available Suppression Pool Cooling (APPX 17A) as necessary.

SRO EOl-2 SPIL Monitor and control Suppression Pool Level between -1 inch and -6 inches.

Can Suppression pool level be maintained above -6 inches YES Can Suppression pool level be maintained below -1 inch YES IL

5 Page 32 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Vent Containment lAW Appendix 12 VERIFY at least one SGTS train in service.

2. VERIFY CLOSED the following valves (Panel 2-9-3 or Panel 2-9-54):

2-FCV-64-31, DRYWELL INBOARD ISOLATION VLV, 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE.

Steps 3, 4, 5 and 6 are If I Then steps that do not apply

7. CONTINUE in this procedure at:

Step 8 to vent the Suppression Chamber through 2-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.

8. VENT the Suppression Chamber using 2-FIC-84-19, PATH B VENT FLOW CONT, as follows:

a. PLACE keylock switch 2-HS-84-35, DWISUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).

c. PLACE 2-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

d. PLACE keylock switch 2-HS-84-1 9, 2-FCV-84-1 9 CONTROL, in OPEN (Panel 2-9-55).

e. VERIFY 2-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.

f. CONTINUE_in_this_procedure_at_step_12.

5 Page 33 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Vents Primary Containment lAW Appendix 12

9. VENT the Suppression Chamber using 2-FIC-84-20, PATH A VENT FLOW CONT, as follows:

a. VERIFY OPEN 2-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 2-9-3).

b. PLACE keylock switch 2-HS-84-36, SUPPR CHBRIDW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

c. VERIFY OPEN 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV (Panel 2-9-54).

d. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-55).

f. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.

g. CONTINUE in this procedure at step 12.

Steps 10 and 11 will not apply in this scenario

12. ADJUST 2-FIC-84-19, PATH B VENT FLOW CONT, or 2-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:

Stable flow as indicated on controller, AND 2-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:

iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 pCi/s AND 0-SI-4.8.B.1.a.1 release fraction of 1.

Driver Acknowledge Notification

5 Page 34 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Initiate Suppression Chamber Sprays per Appendix 17C

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 open interlock as necessary:

PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

BOP 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-74-1 22(1 30), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2.-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT VALVE.

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

f. VERIFY OPEN 2-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV.

g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.

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 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN 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.

BOP Determines that Select Logic on RHR Loop II is not functioning and therefore neither the Drywell nor the Suppression Chamber can be sprayed from RHR Loop II

5 Page 35 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Attempts to spray The Suppression Chamber from RHR Loop I with 2C RHR pump, however, the 2C RHR pump immediately trips after attempted start BOP Informs SRO that Containment Sprays are unavailable with either Loop of RHR and the containment must be sprayed with RHRSW using Loop I of RHR (Standby Coolant)

SRO Directs containment sprays using Loop I of RHR with RHRSW (Standby Coolant) iaw EOI App-i 7C for the Suppression Chamber and EOl App-i 7B for the Drywell BOP Initiate Suppression Chamber Sprays with Standby Coolant per Appendix 17C NOTE Step 7 is performed ONLY if directed from Step 3 to spray the Suppression Chamber using Standby Coolant Supply to RHR Loop I.

7. INITIATE Suppression Chamber Sprays on RHR Loop I using Standby Coolant Supply as follows:

a. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR

Directed by SRO, THEN PLACE keylock switch 2-XS-74-i22, RHR SYS I LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

b. MOMENTARILY PLACE 2-XS-74-i2i, RHR SYS I CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

c. IF 2-FCV-74-53, RHR SYS I INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52, RHR SYS I OUTBD INJECT VALVE.

d. VERIFY CLOSED the following valves:

2-FCV-74-61, RHR SYS I DW SPRAY INBD VLV

2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

2-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV

2-FCV-23-52, RHR HX 2D RHRSW OUTLET VLV.

e. VERIFY RHR Pumps 2A and 2C are NOT running.

f. START RHRSW Pumps Dl and D2.

5 Page 36 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP NOTE 2-BKR-074-0100, RHR SYS I U-i DISCH XTIE breaker compartment is maintained in the OPEN position as an Appendix R requirement.

g. NOTIFY Unit I Operator to perform the following:

1) VERIFY CLOSED i-FCV-23-52, RHR HEAT EXCHANGER D COOL WATER OUTLET VLV (Unit 1, Panel 1-9-3).

2) OPEN 1 -FCV-23-57, STANDBY COOLANT VALVE FROM RHRSW (Unit 1, Panel 1-9-3).

3) DISPATCH personnel to place 2-BKR-074-O100, RHR SYS I U-i DISCH XTIE in ON (480V RMOV BD lB Compartment 19A).

Driver As unit I Operator, when called, inform that I -FCV-23-52 is closed. When called to open 1-FCV-23-57, insert Shift F2 (mrf swO9 open) and inform that 1-FCV-23-57 is open Driver When called to place 2-BKR-74-100 in ON, wait 3 minutes and insert Shift Fl (mor ypovfcv74l 00 norm) then inform 2-BKR-74-1 00 is ON

h. NOTIFY Unit 3 Operator to VERIFY CLOSED 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV_(Unit_3,_Panel_3-9-3).

Driver As unit 3 Operator, when called, inform that 3-FCV-23-52 is closed

i. OPEN the following valves:

2-FCV-74-100, RHR SYS I U-i DISCH XTIE

2-FCV-74-57, RHR SYS I SUPPR CHBR/POOL ISOL VLV

2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VLV.

BOP Informs SRO that he is spraying the Suppression Chamber with Standby Coolant using RHR Loop I SRO When Suppression Chamber Pressure exceeds 12 psig, determines that Drywell CT#2 Sprays are required.

Directs Loop I of RHR to be placed in Drywell Sprays with Standby Coolant per EOl Appendix 17B.

5 Page 37 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram BOP Drywell Sprays per appendix 17B

1. IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, CT#2 THEN BYPASS LPCI injection valve open interlock as necessary:

PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. VERIFY Recirc Pumps and Drywell Blowers shutdown.

SRO Directs Recirc Pumps secured (if not already done) and directs Drywell Blowers secured ATC Secures both Reactor Recirc Pumps BOP Goes around back and secures the Drywell Blowers on panel 9-25

3. IF Directed by SRO to spray the Drywell using RHR System 1(11), THEN CONTINUE in this procedure at Step 6 using RHR Loop 1(11).

4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure At Step 8 using RHR Loop I OR At Step 9 using RHR Loop II.

BOP Continues in procedure at step 8 to spray the drywell using Standby Coolant with RHR Loop I

i. OPEN THE FOLLOWING VALVES:

2-FCV-74-100, RHR SYS I U-i DISCH XTIE CT#2 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

2-FCV-74-61, RHR SYS I DW SPRAY INBD VLV NRC All steps in Appendix 17B for using Standby Coolant on Loop I of RHR are the same up to step i.

BOP Opens RHR SYS I DW SPRAY INBD VLV 2-FCV-74-61 and RHR SYS I DW SPRAY OUTBD VLV 2-FCV-74-60 and informs SRO that Drywell Sprays are in progress using Standby Coolant with RHR Loop I BOP Informs SRO that both Drywell and Suppression Chamber pressures are lowering SRO Directs BOP to secure Drywell and Suppression Chamber Sprays before 0 psig

5 Page 38 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram SRO Directs App 17A, RHR SYS OPERATION SUPPRESSION POOL COOLING BOP Initiates Suppression Pool Cooling per Appendix 17A BOP Suppression Pool Cooling per App 17A

1. IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:

. PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

. PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

NOTE: Step may have been performed as part of App 17B BOP 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 RHRSW pump supplying desired RHR Heat Exchanger(s)

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

. 2-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

. 2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

. 2-FCV-23-40, RHR HX 2C RHRSW OUTLET VLV

.__2-FCV-23-52,_RHR_HX 2D_RHRSW OUTLET VLV

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

e. IF LPCI Initiation signal exists, THEN... MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT

f. IF 2-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE is OPEN, THEN.. .VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

5 Page 39 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram

g. OPEN 2-FCV-57(71), RHR SYS 1(11) SIPPR CHBR/POOL 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.

i. THROTTLE OPEN 2-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLGITEST VLV, to maintain EITHER of the following as indicated on 2-FI-74-50(64), RHR SYS 1(11) FLOW:

Between 7000 and 10000 gpm for one-pump operation OR

At or below 13000 for two-pump operation

j. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE

k. MONITOR RHR Pump NPSH using Attachment 1.

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

e. IF Additional Suppression Pool Cooling flow is necessary, THEN... PLACE additional RHR and RHRSW pumps in service using Steps 2.b through 2.1

5 Page 40 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram SRO Directs All available Low Pressure ECCS Pumps secured when/if 2.45 psig in the

________ drywell with 450 psig Reactor Pressure is reached to prevent overfilling the RPV BOP/ATC Secures Core Spray Pumps and RHR Loop II pumps when/if 2.45 psig in the drywell with 450 psig Reactor Pressure is reached to prevent overfilling the RPV SRO Directs ATC to transition RPV Water Level control to EOl App-6A when/if RPV pressure drops below 500 psig ATC Appendix 6A

1. VERIFY CLOSED the following feedwater heater return valves:

. 1-FCV-3-71, HP HTR IAI LONG CYCLE TO CNDR

. 1-FCV-3-72, HP HTR IBI LONG CYCLE TO CNDR

. 1-FCV-3-73, HP HTR ICI LONG CYCLE TO CNDR.

2. VERIFY CLOSED the following REP discharge valves:

. 1-FCV-3-19, REP IA DISCHARGE VALVE

. 1-FCV-3-12, REP lB DISCHARGE VALVE

. 1-FCV-3-5, REP IC DISCHARGE VALVE.

3. VERIFY OPEN the following drain cooler inlet valves:

. 1-ECV-2-72, DRAIN COOLER 1A5 CNDS INLET ISOL VLV

. 1 -FCV-2-84, DRAIN COOLER 1 B5 CNDS INLET ISOL VLV

. 1-FCV-2-96, DRAIN COOLER 1C5 CNDS INLET ISOL VLV

4. VERIFY OPEN the following heater outlet valves:

. 1-FCV-2-124, LP HEATER 1A3 CNDS OUTL ISOL VLV a 1-ECV-2-125, LP HEATER 1B3 CNDS OUTL LSOL VLV

. 1-FCV-2-126, LP HEATER 1C3 CNDS OUTL ISOL VLV.

5. VERIFY OPEN the following heater isolation valves:

. 1-FCV-3-38, HP HTR 1A2 FW INLET ISOL VALVE

. 1-FCV-3-31, HP HTR 1B2 EW INLET ISOL VALVE a 1-FCV-3-24, HP HTR 1C2 FW INLET ISOL VALVE a 1-FCV-3-75, HP HTR IAI FW OUTLET ISOL VALVE

. 1-FCV-3-76, HP HTR IBI EWOUTLET ISOL VALVE

. 1-ECV-3-77, HP HTR 1C1 FWOUTLET ISOL VALVE.

6. VERIFY OPEN the following REP suction valves:

a 1-FCV-2-83, REP 1A SUCTION VALVE

. 1-FCV-2-95, REP lB SUCTION VALVE a 1-FCV-2-108, REP 1C SUCTION VALVE.

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

9. ADJUST 1-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 1-9-5).

10. VERIFY REW flow to RPV.

5 Page 41 of 42 Simulator Event Guide:

Event 7 Major: EHC leak on pump casing leads to turbine trip and scram ATC Transitions RPV Water Level control to EQI App-6A and verifies REP Discharge valves_are_closed_to_prevent_overfilling_RPV SRO REP Classification is a Site Are Emergency. EAL 1.2-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

All Control Rods are inserted Suppression Chamber /Drywell Sprays performed using RHRSW

SHIFT TURNOVER SHEET Thunderstorms are expected in the area for the next 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Equipment Out of ServicelLCOs:

RHR Pump 2A is tagged out for scheduled maintenance.

The following Tech Specs have been entered Tech Spec 3.5.1 Condition A.

Tech Spec 3.6.2.3 Condition A Tech Spec 3.6.2.4 Condition A Tech Spec 3.6.2.5 Condition A TRM 3.5.1 Condition A OperationslMaintenance for the Shift:

Perform RFPT 2B Overspeed Trip Exerciser Test per 2-01-3 section 8.10 and then raise power to 100% with Recirc Flow.

Unusual ConditionslProblem Areas:

None

Appendix 0 Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 9- Op-Test No.: 1205 Examiners:________________ Operators: SRO:_________________

ATC:______________

BOP:________________

Initial Conditions: Unit 2 is at 6% power. A shutdown is in progress in accordance with 2-GOl-100-12A, section 5.3.4 [8]. The Auxiliary Boilers are running and Steam Seals have been shifted to the Aux Boiler. A3 EECW pump is tagged for maintenance.

Turnover: Lower power to < 5% and transfer SJAE and Off-Gas Preheaters from nuclear steam to auxiliary. A Severe thunderstorm watch is in effect for the next 10 to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Event Event No. MaIf. No. Type* Event Description R-ATC I N/A Lower power to < 5% with control rods.

R-SRO C-ATC CR0 flow element fails high causing 2-FIC-85-1 I CR0 flow 2 rd22 C-SRO control valve to close.

N-BOP 3 N/A Place SJAE & OG Preheaters on Aux steam.

N-SRO C-BOP Trip of C3 EECW Operator manually starts and aligns the 4 swO3J TS-SRO Cl EECW pump per ARP.

C-BOP 5 bat PSC Head Tank Pump Failure.

C-SRO C-ATC 6 bat Recirc pump A high vibration, dual seal failure, trip, isolable.

TS-SRO 7 th33a M-ALL Steam line break in Drywell.

8 edl0a C-ALL Loss of 480V Shutdown Board 2A.

9 zdihs7475 C-ALL RHR loop II Drywell spray valve 74-75 fails to open.

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

Appendix D Scenario Outline Form ES-D-1 Events

1. Starting 6% power, the crew will lower power to < 5% by inserting control rods per 2-GOl-100-12A, RCP, and 2-01-85.

2. During the power reduction the crew will respond to the CRD flow element failing high causing 2-FIC-85-1 I CRD flow control valve to close. The ATC will take action to take manual control of the CRD flow controller and restore CRD system parameters.

3. Turbine sealing steam has already been swapped to aux boiler steam and the crew will swap the operating SJAE and 0ff-Gas Preheaters from nuclear steam to aux boiler steam per 2-G0I-100-12A and 2-01-66.

4. C3 EECW pump will trip, with A3 EECW pump tagged, the crew will respond per the ARP and manually start the Cl EECW pump. 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.

5. The crew will respond to a PSC head Tank low level and recognize that the neither PSC head pump is running. Attempts to start the first PSC head tank pump will fail and the second PSC head tank pump will start when the BOP manually starts it.

6. Reactor Recirculation Pump 2A high vibration will be received followed by an inner seal failure and then an outer seal failure. The crew will trip Recirc pump 2A and enter 2-AOl-68-IA. The operator will isolate the recirc loop. The SRO will refer to Tech Specs and determine TS 3.4.1 Condition A is applicable.

7. A steam line break in the Drywell will cause the crew to insert a manual scram. All rods will insert.

8. 480V Shutdown Board 2A will be lost and can be transferred to alternate power by the crew from panel 2-9-8 in the control room. Transferring the 480V S/D Board 2A will allow the crew to spray the drywell with RHR Loop I.

9. The drywell leak will cause Drywell pressure to increase and the crew will respond per EOl-1 and EOl-2 to spray the Suppression Chamber and Drywell. RHR Loop II Drywell sprays will be unavailable due to 2-FCV-74-75 failing to open.

Terminate the scenario when the following condition is satisfied or upon request of Lead Examiner.

Drywell Sprays in service with RHR Loop I

Appendix 0 Scenario Outline Form ES-D-1 Critical Tasks - One CT#1 -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 17B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment OR CT#1- 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 EOl Appendix 17B AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment

Appendix D Scenario Outline Form ES-D-j SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: NRC 9 7 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOl entry: List (1-4) 4 Abnormal Events: List (1-3)

I Major Transients: List (1-2) 2 EOls used: List (1-3) 0 EOl Contingencies used: List (0-3) 75 Validation Time (minutes)

I Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix D Scenario Outline Form ES-D-1 Scenario Tasks EVENT TASK NUMBER K/A RO SRO I Lower Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 2 Swap SJAEs and Preheaters from Main Steam to Auxiliary Boiler Steam RO U-066-NO-14 271000A1.01 3.3 3.2 3 CRD Flow Element Failure RO U-085-AB-03 201001A3.01 3.0 3.0 4 EECW Pump Trip RO U-067-NO-12 400000A2.01 3.3 3.4 5 Respond to PSC Head Tank Low Level RO U-075-AL-03 203000A4.03 3.4 3.4 6 Recirc pump high vibration/dual seal failure U-068-AL-11 202001A1.14 3.1 3.2 7 DryweilLOCA RO U-000-EM-05 295028EA2.01 4.0 4.1 SRO S-000-EM-04 SRO S-000-EM-05 SRO T-000-EM-15

9 Page 6 of 38 Procedures Used/Referenced:

Procedure Number Procedure Procedure Title Revision 2-G0l-100-12 Power Maneuvering Rev 41 2-01-68 Reactor Recirculation System Rev 139 2-01-85 Control Rod Drive System Rev 130 2-ARP-9-5A Alarm Response Procedure Rev 48 2-01-66 Off-Gas System Rev 104 0-01-67 EECW System Rev 94 2-ARP-9-20A Alarm Response Procedure Rev 25 TS 3.7.1 RHRSW System and UHS Amd 254 TS 3.7.2 EECW System and UI-IS Amd 254 2-ARP-9-3A Alarm Response Procedure Rev 44 2-ARP-9-4A Alarm Response Procedure Rev 37 Trip/Core Flow Decrease OPRMs 2-A0l-68-IA Rev 8 Operable Drywell Pressure and/or Temperature High, or 2 AOl 64 1

- R ev 24

- Excessive Leakage into Drywell TS 3.4.1 Recirculation Loops Operating Amd 258 2-ARP-9-5B Alarm Response Procedure Rev 26 2-ARP-9-3B Alarm Response Procedure Rev 20 2-AOl-i 00-1 Reactor Scram Rev 97 2-EOI-i RPV Control Flowchart Rev 12 2-EOl Appendix-5A Injection Systems Lineup Condensate/Feedwater Rev 9 2-E0I-2 Primary Containment Control Flowchart Rev 10 RHR System Operation Suppression Chamber Rev ii 2-EOI Appendix-17C Sprays 2-ARP-9-8B Alarm Response Procedure Rev 14 2-EOI Appendix-I 7B RHR System Operation Drywell Sprays Rev 10

9 Page 7 of 38 Simulator Instructor 1C90 prefINRCIl 205-9 pfk 01 tog pfk 02 ann silence pfk 03 bat NRC/i 205-9 pfk 04 imf rd22 100 pfko5mrfrdo5b pfk 06 mrf 0gb norm pfk 07 mrf og03a open pfk 08 mrf ogo3b open pfk 09 mn ogo4a aux pfk 10 mn og04b aux pfk 11 imf sw03j pfk 12 mrlsw06 close pfk si bat NRC/pscpump pfks2 bat RRPAVIBI pfk s3 imfth33a I pfk s4 dmf edi Oa batchINRCIl 205-9 mrfms0l on ior ypobkrrhrswpa3 fail_ccoil ior zIohs2385a[1] off ior zdihs2385a NASP trg e3 NRC/pscpump trg e3= dor xa553a26 trg e5 NRC/modesw imfth33b (e5 360) 156003 trg e5 imfedl0a trg e5= ior zdihs7475a asis trg e5= imf pcl6a imf hpO4 Scenario 9 DESCRI PTIONIACTION Simulator Setup manual Reset to IC 90 Simulator Setup Load Batch restorepref NRCII2O5-9 Simulator Setup manual F3 (load batch file)

Simulator Setup Verify file loaded Simulator Setup Clearance out A3 EECW Pump Simulator Setup manual Bring up RWM & turn CRD power off RCP required (10% < 5% with rods) Provide marked up copy of 2-GOI-100-12 Simulator Event Guide:

9 Page 8 of 38 Simulator Event Guide:

Event I Reactivity: Lower Power to < 5% with Control Rods SRO Direct Power decrease with rods to < 5% power per RCP and 2-01-85 ATC Lower Power with Control Rods per 2-01-85, section 6.7. Control Rods:

CAUTION Positioning control rods should be done with utmost diligence and care. Notch Inserting control rods provides the most deliberate controlled method of inserting control rods.

NOTES

1) The following steps are performed from Panel 2-9-5 unless noted otherwise.

2) If rod insertion to Position 00 is required and core thermal power is 10%,

entry into LCO 3.1.6 may be required.

6.7 Control Rod Insertion 6.7.1 Initial Requirements

[1] REVIEW Precautions and Limitations in Section 3.7 and 3.8.

[2] 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.)

[3] VERIFY the following prior to control rod movement:

CRD POWER, 2-HS-85-46 in ON.

When Rod Worth Minimizer is enforcing, the ROD WORTH MINIMIZER is operable and LATCHED in to the correct ROD GROUP.

[4] PERFORM the following to insert the control rod as appropriate.

Control Rod Notch Insertion per Section 6.7.2.

Control Rod Continuous Insertion per Section 6.7.3.

9 Page 9 of 38 Simulator Event Guide:

Event 1 Reactivity: Lower Power to <5% with Control Rods ATC 6.7.2 Notch Insertion of Control Rod

[1] VERIFY Section 6.7.1 has been performed.

[2] SELECT desired control rod by depressing appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[3] OBSERVE the following for selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED.

White light on Full Core Display ILLUMINATED

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, to ROD IN and RELEASE.

[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.

[6] IF control rod is to be notch inserted, THEN: PERFORM either of the following as desired: (Otherwise N/A)

[6.1] IF the control rod moved and settled back to the initial position, THEN ATTEMPT to re-insert control rod per step 6.7.2[4].

[6.2] Refer to Section 8.16 for additional methods to reposition control rod.

[7] IF control rod settles one notch past its intended position, THEN With Unit Supervisors permission return the control rod to the intended position per Section 6.6.

[8] IF the Control Rod moves more than one notch from its intended position, THEN: Refer to 2-AOI-85-7 MISPOSITIONED CONTROL ROD.

[9] WHEN control rod movement is no longer required AND deselecting control rods is desired, THEN:

[9.1] PLACE CRD POWER, 2-HS-85-46, in OFF.

[9.2] PLACE CRD POWER, 2-HS-85-46, in ON.

Rod sequence: FROM TO 10-11 12 00 50-11 12 00 50-51 12 00 10-51 12 00 02-19 12 00 18-03 12 00 42-03 12 00 58-19 12 00 58-43 12 00 42-59 12 00 18-59 12 00 02-43 12 00

9 Page 100f38 Simulator Event Guide:

Event I Reactivity: Lower Power to <5% with Control Rods ATC 6.7.3 Continuous Insertion of Control Rod

[1] VERIFY Section 6.7.1 has been performed.

[2] SELECT desired control rod by depressing appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[3] OBSERVE the following for selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED.

White light on Full Core Display ILLUMINATED

[4] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, to ROD IN.

[5] WHEN control rod notch reaches even rod notch position prior to desired final control rod notch position, THEN RELEASE CRD CONTROL SWITCH, 2-HS 48.

[6] OBSERVE the control rod settles into desired position AND ROD SETTLE light extinguishes.

[7] IF control rod settles one notch past its intended position, THEN With Unit Supervisors permission return the control rod to the intended position per Section 6.6.

[8] IF the Control Rod moves more than one notch from its intended position, THEN: Refer to 2-AOI-85-7 MISPOSITIONED CONTROL ROD.

[9] WHEN control rod movement is no longer required AND deselecting control rods is desired, THEN:

[9.1] PLACE CRD POWER, 2-HS-85-46, in OFF.

[9.2] PLACE CRD POWER,_2-HS-85-46, in ON.

ATC Down ranges IRMs per 2-GOI-100-l2Ato maintain on range between 25 and 75 as reactor power is lowered.

Driver When directed by the NRC and after several Rods have been inserted, insert F4 (imf rd22 100) to fail the CRD flow element NRC When CRD flow element is failed, ATC should take action to restore CR0 parameters_and_continue_driving_rods_to < 5%_power

9 Page 11 of 38 Simulator Event Guide:

Event 2 Component: CRD flow element fails high causing 2-FIC-85-1 I CRD flow control valve to close.

Reports Alarm 9-5A-window 10 CRD ACCUM CHG WTR HDR PRESS HIGH A. VERIFY pressure high on CRD ACCUM CHG WTR HDR 2-PI-85-13A, B. CHECK 2-FCV-85-1 IA (B) in service.

ATC C. IF in-service controller has failed, THEN REFER TO 2-01-85.

D. IF pressure is still greater than 1510 psig after verifying proper controller operation, THEN THROTTLE PUMP DISCH THROTTLING, 2-THV-085-0527, to_maintain_between_1475_and_1500_psig.

ATC Report CRD controller is not responding in Automatic.

The crew may use OPDP-1 guidance listed below, or 2-01-85 Section 8.33 to take manual control of 2-FIC-85-1 1.

OPDP-1 Conduct of Operations Examiner 3.5 Manual Control of Automatic Systems Note A. If an automatic control or an automatic action is confirmed to have malfunctioned, take prompt actions to place that control in manual or to accomplish the desired function. (e.g. Establishment of manual level control following automatic FCV failure to control level or manual start of an EDO that failed to auto start.)

Directs ATC to take MANUAL control of 2-FIC-85-1 I and adjust flow to normal SRO band of 40-65 gpm.

ATC Takes manual control and restores CRD Parameters.

2-01-85 Control Rod Drive System 8.33 AUTOMATICIMANUAL operation of 2-FIC-85-1 I

[1] REVIEW all Precautions and Limitations in Section 3.6.

[2] IF transferring 2-FIC-85-1 1 from AUTO to MANUAL THEN:

(Otherwise N/A)

[2.1] PLACE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 I in BAL.

[2.2] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, by turning Manual Control Pot inside Control Selector Wheel until red deviation pointer is in the Green Band.

[2.3] PLACE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, in MAN.

[2.4] ADJUST CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 I manual potentiometer to establish the desired_system_flow._Refer to_Section_5.1_or_6.10.

SRO Contacts Work Control to investigate 2-FIC-85-1 I failure.

Examiner Normal CRD parameters are:

Note Cooling Water DIP: 10-14 psid Drive Water D/P: 250-270 psid CRD System Flow: 40-65 gpm

9 Page 12 of 38 Simulator Event Guide:

Event 2 Component: CRD flow element fails high causing 2-FIC-85-1 I CRD flow control valve to close.

Examiner If ATCIUS is unable to diagnose that the CR0 controller can be shifted to manual Note and parameters can be restored the US may direct shifting CRD Flow Control Valves iaw 2-01-85 section 6.3 6.3 Shifting CRD Flow Control Valves

[1] VERIFY Control Rod Drive Hydraulic System in operation. Refer to Section 5.1.

[2] REVIEW all Precautions and Limitations in Section 3.6.

NOTE

1) Erratic operation of CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 I, may be observed during refueling/shutdown operations when larger APs exists due to low reactor pressure and CRD pressure.

2) As CRD Flow Control Valves are shifting, CRD System Flow Controller should be adjusted, as needed, to maintain a constant flow.

[3] PERFORM the following for Flow Control Valve being brought into service from Reactor Bldg El 565:

[3.1] OPEN FCV-85-1 1A(B) INLET SOV, 2-SHV-085- 0563(0561).

[3.2] OPEN FCV-85-1 IA(B) OUTLET SOy, 2-SHV-085- 0564(0562).

[3.3] VERIFY OPEN PCV BYPASS SOV TO FCV-85-1 IA(B), 2 317(318).

[3.4] CHECK OPEN PCV 85-11 SOV, 2-85-247.

[3.5] CHECK OPEN HDR ISOL TO FCV-85-1 IA & B, 2-85-313.

[3.6] CHECK FCV-85-1 IA THREE WAY ISOL valve handle in Horizontal position for 2-85-251.

[3.7] CHECK FCV-85-1 I B, THREE WAY ISOL valve handle in Horizontal position for 2-85-252.

9 Page 13 of 38 Simulator Event Guide:

Event 2 Component: CRD flow element fails high causing 2-FIC-85-1 1 CRD flow control valve to close.

Driver When/if directed to perform step [3] of 2-01-85, wait 3 minutes and report completion. If directed to perform entire section 6.3 of 2-01-85, then wait 3 minutes and insert mrf rdO5 b and report completion 6.3 Shifting CRD Flow Control Valves (continued)

[4] VERIFY CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, in BAL position on outer control selector wheel.

[4.1] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, by turning Manual Control Pot inside Control Selector Wheel UNTIL red deviation pointer is in Green Band.

[4.2] TURN CR0 SYSTEM FLOW CONTROL, 2-FIC-85-1 1, Control Selector from BAL position to MAN position.

[5] REDUCE CRD SYSTEM FLOW using 2-FIC-85-1 1, to approximately 40 gpm with Manual Control Pot on 2-FIC-85-1 1.

[6] PLACE CRD SYSTEM FLOW CV SELECTOR SW, 2-XS-85-1 1, on 2-LPNL-925-0018B, to select Flow Control Valve being brought into service, in VALVE A(VALVE B).

Driver Whenlif directed to perform step [6] of 2-01-85 or to perform entire section 6.3 of 2-01-85, wait 3 minutes and insert mit rdO5 b then report completion

[7] CHECK selected in-service valve opening and out-of-service valve closing.

[8] PERFORM the following for Flow Control Valve being removed from service:

[8.1] CLOSE FCV-85-1 1A(B) INLET SOV, 2-SHV-085-0563(0561).

[8.2] CLOSE FCV-85-1 1A(B) OUTLET SOV, 2-SHV-0850564(0562).

[9] ADJUST CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, to establish between 40 gpm and 65 gpm.

[10] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, by TURNING Flow Demand Thumb Wheel until Red Deviation Pointer is in Green band, AND PLACE in AUTO OR BALANCE.

[11] VERIFY CRD STABILIZING FLOW, 2-Fl-85-22, is approximately 6 gpm (locally on 2-LPNL-925-0018B).

[12] VERIFY CR0 DR WATER HDR FLOW, 2-FI-85-15A, is approximately 0 gpm.

6.3 Shifting CRD Flow Control Valves (continued)

[13] ESTABLISH the following by alternately adjusting tape setpoint of CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, AND throttled position of CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A:

. CR0 CLG WTR HDR DP, 2-PDI-85-18A between 10 psid and 20 psid.

. CRD DRIVE WTR HDR DP, 2-PDI-85-17A between 250 psid and 270 psid.

.__CR0_SYSTEM_FLOW._2-FIC-85-1 1_between_40_om_and_65_porn.

9 Page 14 of 38 Simulator Event Guide:

Event 2 Component: CRD flow element fails high causing 2-FIC-85-1 I CRD flow control valve to close.

NOTE PUMP DISCH THROTTLING valve, 2-85-527, has been set to supply 1500 psig charging water pressure and Unit Supervisor authorization is required prior to changing valve position.

[14] IF CRD ACCUM CHG WTR HDR PRESSURE, 2-PI-85-13A, is less than 1475 psig, OR greater than 1500 psig, THEN THROTTLE PUMP DISCH THROTTLING, 2-THV-085-0527, to maintain pressure within normal operating range of between 1475 psig and 1500 psig, as indicated on 2-PI-85-1 3A.

If/When ATC has swapped CRD FCVs he will realize that this has not corrected the ATC failed flow element problem SRO Consu Its Technical Specifications:

3t4 Control Rod Scram Times LCO 3.1.4

a. No more than 13 OPERABLE control rods shall be slow, in accordance with Table 3.1 .4-1; and

b. No more than 2 OPERABLE control rods that are slow shall occupy adjacent locations.

CONDITION REQUIRED ACTION COMPLETIO N TIME A. Requirements of the A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LCO not met.

9 Page 15of38 Simulator Event Guide:

Event 2 Component: CRD flow element fails high causing 2-FIC-85-1 I CRD flow control valve to close.

3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control C.1 ------ NOTE rods inoperable for RWM may be bypassed reasons other than as allowed by Condition A orB. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable control rod..

3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months AND C.2 Disarm the associated 4 h ours CRD.

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, C, or D not met.

OR Nine or more control rods inoperable.

9 Page 16 of 38 Simulator Event Guide:

Event 3 Normal: Place SJAE & OG Preheaters on Aux steam.

SRO Direct BOP to place SJAE & OG Preheaters on Aux Steam per 2-GOt-I 00-12A 2-GOI-1 00-1 2A

[8] WHEN Reactor power is less than 5% and PRIOR TO Turbine Bypass Valves closing, THEN REDUCE nuclear steam loads as follows:

[8.1] IF auxiliary steam is available, THEN TRANSFER SJAE and Offgas Preheaters from nuclear steam to auxiliary steam per 2-01-66. (Otherwise N/A)

BOP 2-01-66, 0ff-Gas System 8.12 Swapping SJAEs and Preheaters from MS to Aux Boiler Steam

[1] CHECK the following initial conditions are satisfied:

A. All Precautions and Limitations in Section 3.0 have been reviewed.

B. SJAE and/or preheaters are in service using main steam.

C. Swapping to Aux Boiler Steam has been directed by Unit Supervisor/SRO or 2-GOI-1 00-1 2A.

D. Auxiliary Boiler(s) in service per 0-01-12 and boiler pressure greater than or equal to 165 psig.

[2] VERIFY OPEN UNITS I & 2 SJAE STM valve, 0-12-662 (TB EL 565 north of Aux. Boiler A in overhead).

Driver When dispatched to verify open Units I & 2 Steam Valve, 0-12-662, wait 1 minute and report valve 0-12-662 is OPEN

[3] CHECK Auxiliary Steam Supply pressure at 2-Pl-1-150 and 152 on Panel 25-105, is between 170 and 250 psig.

briver Report Auxiliary Steam Pressures on 2-PI-1-150 and 2-PI-1-152 are approximately 190 psig BOP CAUTION The following step terminates steam flow to the SJAEs until the Auxiliary Boiler Steam supply valves are opened. Close coordination between personnel operating Auxiliary Boiler valves and Unit Operator is required.

[4] PLACE both of the following to CLOSE at Panel 2-9-7.

A. SJAE 2A PRESSURE CONTROLLER, 2-HS-1-150.

B. SJAE 2B PRESSURE CONTROLLER, 2-HS-1 -1 52.

[5] VERIFY in NORM, SJAE TRAIN PERMISSIVE, 2-HS-001-0375 (Panel 925-1 05 on junction box 8595).

Driver Insert F6 (mrf oglO norm) when contacted to place SJAE train permissive in NORM and report completion

9 Page 17 of 38 Simulator Event Guide:

Event 3 Normal: Place SJAE & OG Preheaters on Aux steam.

BOP [6] For the SJAE to be returned to service DEPRESS the open pushbutton for AUX STM TO SJAE A(B) 1st, 2nd & 3rd STG, 2-HS-12-3A(5A) UNTIL valve is fully open at JB 3524 El. 586 T6-C.

Insert F7 to lineup SJAE A OR insert F8 to lineup SJAE B when contacted to Driver Depress the open pushbutton for Aux Steam to SJAE that is to be returned to service and report completion BOP [7] CHECK Auxiliary Steam Supply pressure at 2-PI-1-150 and/or 152 on Panel 25-105, is between 175 and 250 psig.

Driver Report Auxiliary Steam Pressures on 2-PI-1 -1 50 and 2-PI-1-152 are approximately 190 psig BOP [8] PLACE both of the following to CLOSE at Panel 2-9-7.

A. STEAM TO SJAE 2A, 2-HS-1-155A.

B. STEAM to SJAE 2B, 2-HS-1-156A.

[9] SWAP steam to the preheaters by performing the following located in turbine bldg breezeway T-7 B LINE El 586:

[9.1] OPEN AUX STEAM TO OFF-GAS PREHEATER 2A, using 2-HS-012-0074B.

[9.2] OPEN AUX STEAM TO OFF-GAS PREHEATER 2B, using 2-HS-012-0075B.

[9.3] CLOSE STEAM TO OFF-GAS PREHEATER 2A, 2-HS-001-0176C.

[9.4] CLOSE STEAM TO OFF-GAS PREHEATER 2B, 2-HS-001-0176D.

Driver Insert F9 AND FlO to open Aux Steam to OG Preheaters 2A (2-HS-012-0074B) and 2B (2-HS-012-0075B) when contacted and report completion. Report Steam to OG Preheaters 2A (2-HS-00i -01 76C) and 2B (2-HS-001 -01 76D) are closed Driver At NRC direction, insert Eli (imf sw03j) to enter trip of C3 EECW Pump

9 Page 18 of 38 Simulator Event Guide:

Event 4 Component: Trip of C3 EECW Operator manually starts and aligns the Cl EECW pump per ARP.

Driver At NRC direction, insert Fl I (imf swO3j) to enter trip of C3 EECW Pump BOP Respond to alarm 20A-35.

20A-35 EECW SOUTH HDR DG SECTION PRESS LOW B. CHECK Panel 2-9-3 for status of North header pump(s) breaker lights and pump motor amps normal.

C. NOTIFY UNIT SUPERVISOR, Unit 1 and Unit 3.

D. START standby EECW Pump for affected header, if available.

H. IF pump failure is cause of alarm, THEN REFER TO Tech Spec 3.7.2.

If contacted, as Unit 3 Operator, inform that 4KV SD BD 3EB received a Motor Driver Overload or Trip alarm If contacted as Unit I operator, you did not secure the C3 EECW Pump 0-01-67, EECW System 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3)

CAUTION Only one RHRSW pump in a given RHRSW pump room may be counted toward meeting Technical Specification 3.7.2 requirements for EECW pump operability.

NOTES

1) RHRSW Pump CI may be aligned for service by this section when:

It is used to meet the minimum number of Tech. Spec. operable pumps; or

. At the discretion of the Unit Supervisor, it is needed to replace another pumps operation; or

At the discretion of the Unit Supervisor, it is needed to assist in supplying header flow/pressure demand.

2) If used to meet EECW requirements, RHRSW pump Cl must be aligned to EECW, the pump started, and should remain running. RHRSW Pump CI does NOT have the same auto start signals as RHRSW Pump C3.

3) The RHRSW pump control switches and amp meters are located at Control Room Panel 9-3, Unit 1, 2, and 3.

4) When RHRSW Pump Cl is aligned for EECW, its RI-IRSW function required by the Safe Shutdown Program (Appendix R) is inoperable. Appendix R program equipment operability requirements of FPR-Volume 1 shall be addressed.

[1] To line up RHRSW Pump Cl for EECW System operation, PERFORM the following:

[1 .1] VERIFY EECW System is in prestartup/standby readiness alignment in accordance with Section 4.0.

[1.2] REVIEW all precautions and limitations in Section 3.0.

[1.3] VERIFY RHRSW Pump Cl is in standby readiness in accordance with 0-01-23.

I

9 Page 19of38 Simulator Event Guide:

Event 4 Component: Trip of C3 EECW Operator manually starts and aligns the Cl EECW pump per ARP.

8.3 Operation of RHRSW Pump Cl (for EECW in place of C3) (contd)

[1.4] VERIFY RHRSW Pump Cl upper and lower motor bearing oil level is in the normal operating range.

[1.5] UNLOCK and CLOSE RHRSW PMP Cl & C2 CROSSTIE, 0-23-544 at RHRSW C Room.

[1.6] OPEN RHRSW PMP Cl CROSSTIE TO EECW, 0-FCV-67-49 using one of the following:

RHRSW PMP Cl CROSSTIE TO EECW, 0-l-IS-67-49A11 on Unit 1

RHRSW PUMP Cl SUPPLY TO EECW, 0-HS-67-49A/2 on Unit 2

RHRSW PUMP Cl SUPPLY TO EECW, 0-HS-67-49A13 on Unit 3

[1.7] REQUEST a caution order be issued to tag RFIRSW 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.

[2] To start RHRSW (EECW) Pump Cl, PERFORM the following:

[2.1] START RHRSW Pump Cl using one of the following:

RI-IRSW PUMP Cl, 0-HS-23-8N1 on Unit I

RHRSW PUMP Cl 0-HS-23-8A12 on Unit 2

RHRSW PUMP Cl, 0-HS-23-8N3 on Unit 3

[2.2] VERIFY RHRSW Pump Cl running current is less than 53 amps using one the following:

RHRSW PUMP Cl AMPS, 0-El-23-8/1 on Unit I

RHRSW PUMP Cl AMPS, 0-El-23-8/2 on Unit 2

RHRSW PUMP Cl AMPS, 0-El-23-8/3 on Unit 3

[2.3] VERIFY locally, RHR SERVICE WATER PUMP Cl breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged.

[2.4] VERIFY RHRSW Pump Cl upper and lower motor bearing oil level is in the normal operating range.

Driver If dispatched to check C3 EECW pump breaker, report breaker tripped on overload and breaker smells burnt but no visible smoke or flames (3EB 4kv SD BD) 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3) (contd)

[2.5] NOTIFY Chemistry of running RHRSW (EECW) pump(s).

[2.6] VERIFY a caution order has been 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 ooerable for EECW.

9 Page 20 of 38 Simulator Event Guide:

Event 4 Component: Trip of C3 EECW Operator manually starts and aligns the Cl EECW pump per ARP.

Driver When chemistry contacted, acknowledge report When contacted as Work Control for Caution Order, acknowledge direction and inform will begin working on a Caution Order When dispatched as intake AUOto check Oil Levels and close 0-23-544 valve wait 2 minutes and insert Fl 2 (mrf swOB close), then report oil levels are normal and the 0-23-544 valve is closed When contacted to check breaker charging spring recharged for the Cl EECW pump, wait 2 minutes and inform amber breaker spring charged light is on and closing spring target indicates charged.

When contacted as Intake AUO for second Oil Level check, report Oil Levels are normal SRO Evaluate Technical Specification 3.7.2 before the Cl EECW Pump is aligned 3.7.2 Emergency Equipment Cooling Water (EECW) System and Ultimate Heat Sink(UHS)

LCO 3.7.2 The EECW System with three pumps and UHS shall be OPERABLE.

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.l Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Two or more required EECW pumps inoperable.

OR UHS inoperable.

9 Page 21 of 38 Simulator Event Guide:

Event 4 Component: Trip of C3 EECW Operator manually starts and aligns the Cl EECW pump per ARP.

SRO Evaluate Technical Specification 3.7.1 after the Cl EECW Pump is aligned 3.7.1 Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS)

LCO 3.7.1 The number of required RHRSW pumps may be reduced by one for each fueled unit that has been in MODE 4 or 5 for L 24 hours.

Four RHRSW subsystems and UHS shall be OPERABLE with the number of OPERABLE pumps as listed below:

1. 1 unit fueled four OPERABLE RHRSW pumps.

2. 2 units fueled six OPERABLE RHRSW pumps.

3. 3 units fueled eight OPERABLE RHRSW pumps.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required RHRSW A.1 NOTES -

pump inoperable. 1. Only applicable for the 2 units fueled condition.

2. Only four RHRSW pumps powered from a separate 4 kV shutdown board are required to be OPERABLE if the other fueled unit has been in MODE 4 or 5 for> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Verify five RHRSW pumps powered from separate 4 kV shutdown boards are OPERABLE. IMMEDIATELY OR A.2 Restore required RHRSW 30 days pump to SRO Contacts the Work Control to get caution tag on Cl RHRSW pump to identify it is aligned to EECW.

9 Page 22 of 38 Simulator Event Guide:

Event 4 Component: Trip of C3 EECW Operator manually starts and aligns the Cl EECW pump per ARP.

SRO Contacts Maintenance to investigate the trip of C3 EECW pump.

9 Page 23 of 38 Simulator Event Guide:

Event 5 Component: PSC Head Tank Pump Failure.

Driver At NRC direction, insert Shift Fl (bat NRC/pscpump) to enter PSC head tank pump failure BOP Report alarm PSC HEAD TANK LEVEL LOW 2-9-3A Window 26 A. VERIFY both PSC Head Tank Pumps are running.

B. VERIFY power available to pumps.

C. CHECK PSC PUMP SUCTION INBD and OUTBD SQL VALVEs, 2-FCV-75-57 and 58, open.

BOP Reports PSC Head Tank Pumps failed to auto start SRQ Directs the BOP to attempt to start PSC head tank pumps per ARP BOP Starts PSC 2B Pump to restore PSC tank level.

Driver Approximately 30 seconds after PSC Pump started, delete alarm override -

xa553a[26]

Driver If dispatched to check PSC head tank, report no leaks noted and no abnormalities noted with the PSC pump 2A.

SRO Refers to TRM and confirms keepfill pressures are within specification:

TR 3.5.4 Maintenance of Filled Discharge Pipe LCO 3.5.4 The OPERABLE pressure indicators on the discharge of the RHR and CS pumps shall indicate not less than listed below.

P1-75-20 39 psig P1-75-48 39 psig P1-74-51 48 psig P1-74-65 35 psig Driver When requested by lead examiner, insert Shift F2 (bat RRPAVIB) for Recirc Pump A trip

9 Page 24 of 38 Simulator Event Guide:

Event 6 Component: Recirc pump A high vibration, dual seal failure, trip, isolable.

Driver When requested by lead examiner, insert Shift F2 (bat RRPAVIB) for Recirc Pump A trip ARP 2-9-4A Window 20 RECIRC PUMP MTR IA VIBRATION HIGH A. CHECK temperatures on RECIRC PMP MTR IA & I B WINDING AND BRG TEMP recorder, 1-TR-68-71 on Panel 1-9-21 are below:

Pump motor bearing temperatures (<190°F)

Pump motor winding temperatures (<255°F)

Pump Seal Cavity temperatures (<180°F)

Pump cooling water from Seal Cooling temperature (<140°F)

Pump motor cooling water from bearing temperature (<140°F)

B. CHECK for a rise in Drywell equip sump pumpout rate due to seal leakage.

C. DISPATCH personnel to 1-LPNL-925-0712 (Vibration Mon. System) on EL 565 and REPORT the Vibration Data for Pump A to the Unit Operator and any other alarm indications. The person shall advise the Unit Operator of any changes in the vibration values.

D. IF alarm seals in, THEN ADJUST pump speed slightly to try reset the alarm.

E. IF unable to reset alarm, THEN

CONSULT with Unit Supervisor, and with his concurrence,

SHUTDOWN the Recirc pump, and

REFER TO 1-AOl-68-IA or 1-AOI-68-1B.

ATC Reports failure of the #1 seal on Reactor Recirc Pump A

9 Page 25 of 38 Simulator Event Guide:

Event 6 Component: Recirc pump A high vibration, dual seal failure, trip, isolable.

RECIRC PUMP A NO. I SEAL LEAKAGE ABN, 2-9-4A Window 25:

A. DETERMINE initiating cause by comparing No. I and 2 seal cavity pressure indicators on Panel 2-9-4 or ICS.

Plugging of No. I 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. I 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.

NOTL

1) Possible indtions of lual sea failure i,clue:

a Window 18 on this pane Irn,in i ecnjinction with :hi window.

Ring dryv.eII pressure and/or tern,erature.

Increased Ieake into the dyeil surnp

. Inr.rsd .,ibn1inn nf Ih rcirc pump When AQU dispatched sent to investigate vibration, report 15 mils and Driver rising.

Enters:

2-AOI-68-IA, Recirc Pump TriplCore Flow Decrease OPRMs Operable, SRO 2-AOI-64-I, Drywell Pressure andlor Temperature High, or Excessive Leakage Into Drywell.

2-AOl-68-IA, Recirc Pump TriplCore Flow Decrease OPRMs Operable,

[1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A),

[1 .1] SCRAM the Reactor.

CAUTION

[NER/C] Failure to restart Reactor Recirculation pumps in a timely manner may result in exceeding the differential temperature limit for pump start and subsequently require plant depressurization to avoid exceeding pressure-tern perature limits for the reactor vessel.

9 Page 26 of 38 Simulator Event Guide:

Event 6 Component: Recirc pump A high vibration, dual seal failure, trip, isolable.

[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 (Otherwise N/A)

IMMEDIATELY take actions to INSERT control rods to less than 95.2%

loadline. Refer to O-Tl-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 already 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] [NER/C] WHEN plant conditions allow, THEN, (Otherwise N/A)

MAINTAIN operating jet pump loop flow greater than 41 x 106 Ibm/hr (2-FI 46 or 2-Fl-68-48).

SRO AOI-64-1 Directs BOP to Vent the Drywell Driver Whenlif requested to start a standby gas fan remote function pcola or b or c

9 Page 27 of 38 Simulator Event Guide:

Event 6 Component: Recirc pump A high vibration, dual seal failure, trip, isolable.

SRO Evaluates Tech Spec 3.4.1 and enters Condition A 3.4.1 Recirculation Loops Operating LCO 3.4.lTwo 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 resetforsingle loop operation; APPLICABILITY: MODES 1 and 2.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A. 1 Satisfy the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO not met. requirements of the LCO.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

OR No recirculation loops in operation.

BOP Contacts Control Bay AUO to perform vent log SRO Calls the Reactor Engineer to change the single loop MCPR limits.

Driver At NRC direction, insert Shift F3 (imfth33a 1) for steam leak inside containment, when contacted to perform vent log acknowledge

9 Page 28 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

Driver At NRC direction, insert Shift F3 (irnf th33a 1) for steam leak inside containment BOP/ATC Announces Drywell Pressure rising and responds to alarms 9-5B, Window 31 and 9-3B, Window 19 9-5B, Window 31, Drywell Pressure Abnormal A. VERIFY alarm using multiple indications.

B. IF RBCCW has been lost, THEN REFER TO 2-AOl-70-1.

C. REFER TO 2-AOI-64-1.

9-3B, Window 19, Drywell Norm Operating Pressure High A. CHECK drywell pressure and temperature for rise.

B. CHECK weather report for atmospheric pressure.

C. IF Drywell DP Compressor is running, THEN STOP compressor.

D. CHECK N2 makeup valves to Suppression Chamber and Drywell closed.

E. CHECK Drywell Control Air System Flow Elements 2-FIQ-032-00092 (Rx Bldg 565 Ri O-S) and 2-FIQ-032-0075 (Rx Bldg 565 R20-TO) < 1.7 SCFM.

F. IF pressure rise is due to normal startup, THEN REFER TO 2-01-64 for normal venting instructions.

G. IF Drywell pressure is high, THEN REFER TO 2-AOI-64-i.

BOP Confirms that a Drywell leak is present based on Drywell Temperature and Pressure rise SRO Directs entry into 2-AOl-64-i and pre-determines a Drywell Pressure value at which_the_ATC_shall_insert_a_manual_scram 2-AOl 1 4.2 Subsequent Actions

[1] IF any E0I entry condition is met, THEN ENTER appropriate EOI(s). (Otherwise N/A)

[2] IF Drywell Pressure is High, THEN PERFORM the following: (Otherwise N/A)

[2.1] CHECK Drywell pressure using multiple indications.

[2.2] IF Drywell pressure rising rate indicates Reactor Scram at 2.45 psi is imminent, THEN REDUCE Reactor power via Recirc flow to minimize the impact of a scram from high power. (Otherwise N/A)

[2.3] CHECK Drywell pressure using multiple indications.

[2.4] ALIGN and START additional Drywell coolers and fans as necessary. REFER TO 2-01-64.

[2.5] VENT Drywell as follows:

BOP Verifies Drywell venting is still in progress from Reactor Recirc Pump seal failure Driver After Modeswitch is placed in the Shutdown position, insert Shift F4 (drnfedlOa)

9 Page 29 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

SRO Directs ATC to insert a Reactor Scram prior to Drywell Pressure reaching 2.45 psig Inserts Manual Reactor Scram and performs immediate actions of 2-AOl-I 00-1 and ATC Hard Card actions Driver Ensure trigger 5 goes active on the MODESWITCH Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS

[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AIS3A and 2-HS 5AIS3B, on Panel 2-9-5.

Depresses Reactor Scram Pushbuttons and reports rod movement ATC NA [2] IF scram is due to a loss of RPS, THEN

[3] Refuel Mode One Rod Permissive Light check

[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL.

[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-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)

[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, 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 Places Reactor Mode switch in Shutdown, Performs Refuel Mode One Rod ATC Permissive Light Check and announce All Rods In on Scram report. Proceeds to subsequent actions of Hard Card Driver After Modeswitch is placed in the Shutdown position, insert Shift F4 (dmf edl0a)

9 Page 30 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

2.0 SUBSEQUENT ACTIONS:

NA [1] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise N/A):

[2] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.

[3] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 2-9-5.

[4] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, using RFP/RFPT.

[5] RETURN to body of procedure at step 4.2[5] AND CONTINUE with actions as required.

Drives in all lRMs and SRMs as time permits, verifies SDV Vents and Drains are ATC closed,_monitors_RPV water_level BOP Performs actions of BOP Reactor Scram Hard Card Reactor Scram BOP Unit Operator Hard Card 1.0 SUBSEQUENT ACTIONS: PANELS 2-9-7 & 2-9-8 NOTES

1) The following steps are not required to be performed in order, but only as required to maintain stable conditions.

2) It is desired to trip the turbine prior to receiving the GEN REVERSE PWR FIRST RELAY OPERATION 2-EA-57-136 (2-XA-55-8A, Window 7) alarm to avoid motorizing the generator.

[1] At 50 MWe, or as directed by the Unit Supervisor, VERIFY TRIPPED the Main Turbine as follows:

[1 .1] DEPRESS the TRIP pushbutton, 2-HS-47-67D on Panel 2-9-7.

[1.2] VERIFY OPEN Generator Output Breaker, by placing GENERATOR PCB 224, 2-HS-242-0224A, to TRIP.

[1.3] IMMEDIATELY PLACE VOLTAGE REGULATOR START/STOP SEL, 2-HS-57-24, to STOP and release.

[1 .4] CHECK the following at 2-HS-57-24:

. GREEN light illuminated

. RED light extinguished

[2] ANNOUNCE Reactor SCRAM over PA system.

Trips the Main turbine, opens the generator breaker and places the voltage BOP regulator in the STOP position, and announces Reactor Scram over the PA system Driver After Modeswitch is placed in the Shutdownposition, insert Shift F4 (dmf edl0a)

9 Page 31 of38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

[1] MONITOR and CONTROL RPV pressure to keep below 1073 psig and stable.

[1.1] IF RPV pressure is lowering rapidly, THEN NA

[1.2] IF MSRVs are cycling and bypass valves are available, THEN MANUALLY OPEN MSRVs on Panel 2-9-3 until RPV pressure is below 965 psig. (Otherwise N/A)

BOP Manually opens SRVs until RPV Pressure is below 965 psig then closes SRVs

[1.3] IF MSRVs are cycling and bypass valves are NOT available, THEN NA MANUALLY OPEN MSRVs on Panel 2-9-3 until RPV pressure is controlled between 800 and 1000 psig.

[2] IF any PCIS isolation signal is received, THEN VERIFY PCIS isolations using any of the following: (Otherwise N/A)

Containment Isolation Status System on Panel 2-9-4

. PCIS Mimic and individual control switch indications

. ICS

. 2-01-64

[3] IF HPCI and/or RCIC are in service and injecting to the vessel, THEN MONITOR and CONTROL Reactor Water Level as necessary. (Otherwise N/A)

Verifies PCIS isolation signals received and reports condition of PCIS and BOP HPCI/RCIC to US.

Enters EOI-1 on High Drywell Pressure RCIQ Monitor and Control Reactor Power.

Exits RC/Q based override step RC/Q-2 and enters 2-AOl-i 00-1 RCIP Monitor and Control RPV Pressure.

Answers NO to: Is any MSRV cycling? (BOP has already opened SRVs as necessary)

Directs BOP to maintain Rx Pressure with Turbine Bypass Valves.

SRO RCIL Monitor and Control RPV Water Level.

Verify as Required:

. PCIS Isolations (Groups 1, 2 and 3)

. ECCS

. RCIC Restore and maintain RPV water between +2 and +51 inches with one or more of the following inj. sources

. CNDS AND FW Appendix-5A Can_RPV water_level_be_restored_and_maintained_above_+2_inches-YES

9 Page 32 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

Directs ATC to maintain RPV water level +2 to +51 inches with condensate and SRO feedwater per Appendix-5A Performs actions necessary to maintain RPV Water Level +2 to +51 inches iaw EOI ATC App-5A Appendix-5A

1. IF It is desired to use a reactor feed pump that is in operation, THEN CONTINUE at step 12 to control the operating pump.

12. SLOWLY ADJUST RFPT speed UNTIL feedwater flow to the RPV is indicated, using ANY of the following methods on Panel 2-9-5:

. Individual 2-HS-46-8A(9A)(IOA), RFPT 2A(2B)(2C) SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

. Individual 2-SIC-46-8(9)(1O), RFPT 2A(2B)(2C) SPEED CONTROL in MAN UAL, OR

. 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 2-SIC-46-8(9)(1O), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO.

13. ADJUST RFPT speed as necessary to control injection using the methods of step 12.

14. WHEN RPV level is approximately equal to desired level AND automatic level control is desired, THEN PLACE 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 2-SIC-46-8(9)(1O), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO.

9 Page 33 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

EOl-1 and EOl-2 will be entered on Drywell pressure and all legs executed NRC concurrently Executes all legs of EOl-2 concurrently EOl-2 DWIT Monitor and control Drywell Temperature below 160°F, using available Drywell Cooling.

Answers NO to: Can Drywell Temperature be maintained below 160°F?

pCIp SRO Monitor and control Primary Containment pressure below 2.4 psig using the Vent System (Appendix 12) as necessary In Progress Answers No to can Primary Containment Pressure be maintained below 2.4 psig Before Suppression Chamber Pressure rises to 12 psig Initiate Suppression Chamber Sprays using only those pumps not required for Adequate Core Cooling Directs Suppression Chamber Sprays EOl-2 PC!H Monitor and control Drywell and Suppression Chamber:

. Hydrogen at or below 2.4%

AND SRO

Oxygen at or below 3.3%

Using the Nitrogen Makeup System (APPX 14A).

EOI-2 SPIT Monitor and control Suppression Pool temperature below 95°F, using available Suppression Pool Cooling (APPX 17A) as necessary.

SRO EOl-2 SPIL Monitor and control Suppression Pool Level between -1 inch and -6 inches.

Can Suppression pool level be maintained above -6 inches YES Can Suppression pool level be maintained below -1 inch YES SRO Directs Suppression Chamber Sprays with both loops of RHR per Appendix I 7C

9 Page 34 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

BOP Initiate Suppression Chamber Sprays per Appendix 17C

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 open interlock as necessary:

PLACE 2-HS-.74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

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-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT BOP VALVE.

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

f. VERIFY OPEN 2-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV.

g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.

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 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN 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.

Determines that Select Logic on RHR Loop II is not functioning and BOP therefore neither the Drywell nor the Suppression Chamber can be sprayed from RHR Loop II SRO Directs Suppression Chamber Sprays from RHR Loop I I

9 Page 35 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

Diagnoses that RHR Loop I containment spray valves have lost power and BOP determines that 480 Volt Shutdown Board 2A has been lost B0P Responds to loss of 480 Volt Shutdown Board per ARP 9-8B, Window 29 9-8B, Window 29, 480V Shutdown BD 2A UV or XFR A. CHECK for indication of 480V Shutdown Bd 2A loss:

. RWCU Pump 2A shutdown.

. Fuel pool cooling Pump 2A shutdown.

. 480V Shutdown Bd 2A voltage (2-El-57-30).

. RBCCW Pump 2A Shutdown.

. Half Scram.

B. IF 480V Shutdown Bd 2A is lost, THEN Manually TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 2-9-8.

C. IF manual transfer is accomplished, THEN REFER TO 0-0l-57B, 2-01-99, and appropriate Ols for recovery or realignment of equipment.

D. IF manual transfer is NOT accomplished, THEN REFER TO Tech Spec 3.8.7 and 3.8.8.

SRO Directs Transferring 480 Volt Shutdown Board 2A to alternate iaw the ARP Transfers 480 Volt Shutdown Board 2A to alternate by placing CS in ALTERNATE BOP position on panel 2-9-8 Commences Suppression Chamber Sprays iaw Appendix-i 7C (PROCEDURE BOP ABOVE)

SRO When Suppression Chamber Pressure exceeds 12 psig, determines that Drywell CT#1 Sprays are required.

Directs Loop I of RHR to be placed in Drywell Sprays per EOI Appendix 17B.

9 Page 36 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

Drywell Sprays per appendix 17B

1. IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BOP BYPASS LPCI injection valve open interlock as necessary:

CT#1

PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

. PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. VERIFY Recirc Pumps and Drywell Blowers shutdown.

Directs Recirc Pumps secured (if not already done) and directs Drywell Blowers SRO secured Secures both Reactor Recirc Pumps ATC Goes around back and secures the Drywell Blowers on panel 9-25 BOP

3. IF Directed by SRO to spray the Drywell using RHR System 1(11), THEN CONTINUE in this procedure at Step 6 using RHR Loop 1(11).

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-74-1 22(1 30), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-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:

2-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV

2-FCV-74-61(75), RHR SYS 1(11) DW SPRAY INBD VLV

g. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.

9 Page 37 of 38 Simulator Event Guide:

Event 7 Major: Steam line break in Drywell.

h. IF Additional Drywell Spray flow is necessary, THEN.. .PLACE the second System 1(11) 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:

BOP

2-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2-FCV-23-40, RHR HX 2C RHRSW OUTLET VLV

2-FCV-23-52, RHR HX 2D RHRSW OUTLET VLV I. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

Directs Drywell Sprays secured before 0 psig in the Drywell SRO

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-74-1 00(1 01), RHR SYS I U-1(SYS II U3) DISCH XTIE

2-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV

2-FCV-74-61(75), RHR SYS 1(11) DW SPRAY INBD VLV

b. VERIFY OPEN 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:

Drywell Sprays in service with RHR Loop I

SHIFT TURNOVER SHEET Unit 2 is at approximately 6% power Units I and 3 are operating at 100% power Equipment Out of ServicelLCOs:

A3 EECW pump is tagged for maintenance. Appendix R compensatory measure A is in effect for the appropriate fire zones associated with RHRSW Pump A3, 0-PMP-023-0085, on page 460 of the Fire Protection Report.

OperationslMaintenance for the Shift:

Unit 2 is at 6% power. A shutdown is in progress in accordance with 2-GOI-100-12A, section 5.3.4 [8].

Reactor Pressure is being controlled with Turbine Bypass Valves. RFPTs B and C are in service controlling Reactor Water Level with RFPT C in Auto. The Main Turbine is on the Turning Gear. The Auxiliary Boilers are in service and Steam Seals have been shifted to the Auxiliary Boiler.

Lower power to <5% and transfer SJAE and Off-Gas Preheaters from nuclear steam to auxiliary. All Surveillance Requirements for Modes 3 and 4 are complete.

Unusual ConditionslProblem Areas:

The National Weather Service has issued a Severe Thunderstorm Watch for the next 10 to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ML12318A362

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