ML111010606

Initial Exam 2011-301 Final Simulator Scenarios

ML111010606
Person / Time
Site:	Browns Ferry
Issue date:	04/04/2011
From:	NRC/RGN-II/DRS/OLB
To:	Tennessee Valley Authority
References
50-259/11-301, 50-260/11-301, 50-296/11-301
Download: ML111010606 (283)

Text

Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: A Op-Test No.: ILT 1102 FINAL SRO:

Examiners: Operators: ATC:

BOP:

Initial 1C190 / Unit 3 Reactor Power 83% / RHRSW Pump B2 is tagged out of service I APRM 3 Conditions: is bypassed for Surveillance Testing Turnover: Alternate Bus Duct Cooling Fans per 3-01-47 Section 6.11.1 [2]. Raise Reactor Power to 90% with Reactor Recirculation.

Event Event No. Maif. No. Type* Event Description N-BOP 1 Bus Duct Cooling Fan rotation 3-01-47 Section 6.11.1 [2]

NSRO R-ATC 2 Raise Reactor Power with Recirc RSRO C-ATC rd0la CRD Pump 3A Trip C-SRO 1-BOP 4 og05a HWC Malfunction TS-SRO C-ATC 5 thl2a Recirc Pump 3A High Vibration CSRO hpOl C-BOP 6 HPCI Inadvertent Initiation TS-SRO hpO8 HPCI Steam Leak Fail to isolate / Loss of 480 V RMOV Bd 3A1 ED 7 M-ALL hpO9 onTemps 8 ad03b C 1 ADS Valve fails to operate 9 fwl2 C Startup Level Control Valve Failure

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

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

1. Safety Significance:

Scram reduces to decay heat energy that the RPV may be discharging into the secondary containment.

2. Cues:

Procedural compliance.

Secondary containment area temperature, level, and radiation indication.

Field reports.

3. Measured by:

Observation With a primary system discharging into secondary containment, a reactor scram is initiated before a maximum safe condition is reached.

OR Observation With a primary system discharging into secondary containment, US transitions to EOP-l and RO initiates scram upon report that a maximum safe condition has been reached.

4. Feedback:

Control rod positions.

Reactor power decrease.

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.

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.

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

With the unit at 83% power, the BOP operator will rotate Bus Duct Cooling Fans JAW 3-01-47 section 6.11 .1 [2]. Upon completion the ATC will commence power increase with flow.

When the NRC is satisfied with the reactivity manipulation, CRD Pump 3A will trip. ATC will perform 3-AOI-85-3 actions to start the Standby CR1) Pump.

Once the Standby CRD Pump is started and CRD parameters are restored, the Hydrogen Water Injection system will malfunction resulting in high hydrogen concentration in Off Gas. The crew will respond JAW with ARPs and 3-AOI-66-1 and shutdown the Hydrogen Water Chemistry System. The SRO will address TRM 3.7.2 and Enter Condition A.

After shutdown of the HWC System, high vibration alarms on Reactor Recirculation Pump 3A will have the crew respond JAW the ARPs. The ARPs will direct the operators to adjust RR Pump 3A speed in an attempt to lower vibrations on RR Pump 3A. Once speed is adjusted, high vibration alarm will clear and vibrations will lower.

After the RR Pump 3A vibrations is addressed, HPCI will inadvertently initiate. The crew will verify the initiation is inadvertent and trip and lockout HPCI. The SRO will address Technical Specification 3.5.1 and Enter Condition C.

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 HPCI steam valve 73-2 the 3A RMOV Board will be lost due to an electrical fault.

The crew will enter EOI-3 and scram the Reactor. All rods will insert on the scram and level and pressure will be controlled JAW EOI-1. The crew should lower reactor pressure. As the second MAX safe temperature is approached, the crew should anticipate Emergency Depressurization and when the second MAX safe temperature is reached the crew will Emergency Depressurize.

During ED one ADS valve will fail and the operator will open an additional SRV. After ED, the startup level controller will fail. The crew will control level with Core Spray Loop 2 and place R}JR Loop 2 in Suppression Pool Cooling.

The Emergency classification is 3.1-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

All Control Rods are inserted.

Emergency Depressurization complete.

Reactor Level is restored and maintained.

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-A 7 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 4 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 60 Validation Time (minutes) 2 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No) Scenario Tasks

Appendix D Scenario Outline Form ES-D-1 EVENT TASK NUMBER K/A RO SRO 1 Rotate Bus Duct Cooling Fans RO U-047-NO-27 400000A4.01 3.1 3.0 2 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 3 CRD Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 4 Hydrogen Water Chemistry Malfunction RO U-066-AL-10 271000A1 .13 3.2 3.7 SRO S-066-AB-01 5 Reactor Recirculation Pump High Vibrations RO U-068-AL-1 1 20200 1A4.05 3.3 3.3 6 HPCI Inadvertent Start RO U-073-NO-05 206000A2.17 3.9 4.3 7 HPCI Steam Leak RO U-073-AL-06 295032EA2.03 3.8 4.0 SRO S-000-AB-03 SRO S-000-EM-12 SRO T-000-EM-15

3-A Page 6 of 43 Procedures Used/Referenced:

Procedure Number ] Procedure Title Procedure Revision 3-01-47 Turbine-Generator System Revision 91 3-GOT-i 00-12 Power Maneuvering Revision 35 3-01-68 Reactor Recirculation System Revision 80 3-AOI-85-3 CRD System Failure Revision 10 3-ARP-9-53 Alann Response Procedure Panel 3-9-53 Revision 24 3-AOT-66-1 Off Gas Hydrogen High Revision 6 TRM 3.7.2 Airborne Effluents Revision 0 3-ARP-9-4A Alarm Response Procedure Panel 3-9-4A Revision 39 TS 3.5.1 ECCS Operating Amendment 244 3-01-3 Reactor Feedwater System Revision 82 3-EOT-2 Primary Containment Control Flowchart Revision 7 3-E0I-APPENDIX-18 Suppression Pool Water Inventory Removal and Makeup Revision 2 3-ARP-9-3F Alarm Response Procedure Panel 3-9-3F Revision 28 3-EOI-3 Secondary Containment Control Flowchart Revision 9 Restoring Refuel Zone and Reactor Zone Ventilation 3E01APPENTMX8F Revision 2 Following_Group_6_Isolation 3-E0I-1 RPV Control Flowchart Revision 8 3-E0I-3-C-2 Emergency RPV Depressurization Flowchart Revision 8 3-EOI-APPENDIX-5A Injection Systems Lineup Condensate/Feedwater Revision 5 3-E0I-APPENDIX-6A Injection Subsystems Lineup Condensate Revision 2 3-EOI-APPENDIX-6B Injection Subsystems Lineup RHR System I LPCI Mode Revision 3 3-E0I-APPENDIX-6C Injection Subsystems Lineup RHR System II LPCI Mode Revision 3 3-EOT-APPENIMX-6D Injection Subsystems Lineup Core Spray System I Revision 3 3-EOI-APPENDJX-6E Injection Subsystems Lineup Core Spray System II Revision 3 Emergency Classification Procedure Event Classification EPTP-1

. Revision 46 Matrix EPIP-4 Site Area Emergency Revision 32 3-E0I-APPENDLX-i 9 H2/O2 Analyzer Operation Revision 0 3-EOT-APPENDJX-1 7A RFLR System Operation Suppression Pool Cooling Revision 5 3-A0I-100-1 Reactor Scram Revision 53

3-A Page 7 of 43 Console Operator Instructions A. Scenario File Summary

1. File: batch and trigger files for scenario 3-A Batch nrc2OllaRl

1. rhrsw pump B2 clearance ior ypobkrrhrswpb2 fail_tcoil ior zlohs23l9a[1] off

1. aprm 3 bypassed for 3 -sr-3 .3.1.1 .16

1. crd a pump trip imfrd0la(el 0)

1. hpci Initiation imfhpol (e5 0)

1. recirc pump a vibration high imfthl2a (elO 0)

1. hwc malfunction imfog05a (e15 60) 99 ior xa5553a[10] (e15 0) alarm on trg 16nrc20110440 trg 16 = mmfogo5a 100 36099 Trigger nrc2Ol 10440 zdihs0440a[ 1] .eq. 1

1. HPCI Steam Leak/major (have to manually modify fpO2 to close) mrff,02 (e20 0) close imf hpO9 imfhpo8 (e20 0) 8 600 4 trg2l nrc2011732 trg2l =imfedl2a ior ypovfcv733 (e20 0) fail_now imf fw 12 imfad03b

1. if crew anticipates ED, may have to raise severity Trigger nrc20 11732 zdihs732[1].eq.1

3-A Page 8 of 43 Console Operator Instructions Scenario 3-A DESCRIPTION/ACTION Simulator Setup manual Reset to IC 190 Simulator Setup Load Batch Bat nrc2Ol 1 aRl Simulator Setup manual Place APRM 3 in Bypass Simulator Setup manual Clearance out RHRSW Pump B2 Simulator Setup Verify Batch file loaded RCP required (83% 90% w/Recirc flow)

- Provide marked up copy of 3-GOI-100-12 and RCP for Urgent Load Reduction.

3-A Page 9 of 43 Simulator Event Guide:

Event 1 Normal: Bus Duct Cooling Fan rotation, 3-01-47, Section 6.11.1 [2]

SRO Directs BOP to rotate Bus Duct Cooling Fans.

BOP Rotate Bus Duct Cooling Fans, lAW 3-01-47, Section 6.11.1 [2]

[2] PERFORM the following to SWAP from Bus Duct Cooling Fan A to Fan B:

[2.1] VERIFY U-3 GEN BUS DUCT HTX B INLET VANE DMPR, 3-DMP-262-0057, is fully OPEN.

[2.2] DRAIN water from 3B bus duct fan housing as follows:

[2.2.1] Simultaneously OPEN GEN MAIN BUS COOLING FAN B DRAIN VALVE, 3-DRV-262-0002, and OBSERVE GEN MAIN BUS COOLING FAN B DRAIN SIGHT GLASS, 3-LG-262-0002, for water.

[2.2.2] WHEN GEN MAIN BUS COOLING FAN B DRAIN SIGHT GLASS, 3-LG-262-0002, no longer indicates water flow, THEN CLOSE GEN MAIN BUS COOLING FAN B DRAINVALVE, 3-DRV-262-0002.

DRIVER Pre start walk down complete Inlet Danper is Fully Open, Water has been drained from faniiousing, B Fan is not rotating.

BOP [2.3] On Panel 9-7, MOMENTARILY PLACE GEN BUS DUCT HX FAN A, 3-HS-262-0001A, in STOP.

[2.4] On Panel 9-7, MOMENTARILY PLACE GEN BUS DUCT HX FAN B, 3-HS-262-0002A, in START.

3-A Page 10 of43 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Flow SRO Notifies ODS of power increase.

Directs Power increase using Recirc Flow, per 3-GOT-i 00-12.

[21 j 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 andlor 3B speed with Recirc Individual Control, THEN PERFORM the following;

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

3-HS-96-1 5A(i5B).

AND/OR

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

3-HS-96-l 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 NC When atsfied wtbRctivityMampü1ation CRtJuip Tri E)RIVER When dirted byJead ei mner,Tgger C1) Pump

3-A Page 11 of43 Simulator Event Guide:

Event 3 Component: CRD Pump 3A Trip ATC Reports Trip of CR1) Pump 3A.

SRO Announces entry into 3-AOI-85-3, CRD System Failure.

4.1 Immediate Actions

[1] IF operating CRD PUMP has failed AND the standby CRD Pump is available, THEN PERFORM the following at Panel 3-9-5:

[1.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, in MAN at minimum setting.

[1.2] START associated standby CR1) Pump using one of the following:

• CRD PUMP 3B, using 3-HS-85-2A

[1.3] ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-11, to establish the following conditions:

• CRD CLG WTR HUR DP, 3-PDI-85-1 8A, approximately 20 psid.

• CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, between 40 and 65 gpm.

[1.4] BALANCE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, and PLACE in AUTO or BALANCE.

riuv IfDipafcbed to CJW Pi,mip 3A,j,umps extreme1y hot to tonh.

çum 3B a eve1sinbdpmpreadyfor st conditios ane sLrt RD 3A -xeportbreaker trijed on oyer ctirrent Electrical Maint e41ed NRC When ready, HWC Malfunction DRIVE1 Upon Lead examiner direction, in tiat Trigger 15 for HWC Malfunction;

3-A Page 12 of43 Simulator Event Guide:

Event 4 Instrument: HWC Malfunction BOP Respond to Off Gas Panel Alarms 9-53-10, 3, and 13 53-10, H2 Water Chemistry Abnormal A. Checks H2 concentration on H2 analyzer on 3-9-53.

B. Dispatches personnel.

53-3 and 13, High Offgas % H2 Train A and B A. CHECK H2 concentration on OFF-GAS HYDROGEN ANALYZER, at 3-H2R-66-96 (CH2), on Panel 3-9-53 to verify H2 concentration..

B. IF alarm is valid, THEN REFER TO 3-AOI-66-1.

SRO Announces entry into 3-AOI-66-1, Off Gas 1-12 High.

DRIVER jat1ie toPanI report, H2injtioii rates above (high) setpoint cannot adjust BOP 3-AOI-66-1, Off Gas H2 High

[2] IF HWC System injection is in service, THEN PERFORM the following

[2.1] At HYDROGEN WATER CHEMISTRY CONTROL PANEL, 3-LPNL-925-0589, VERIFY that H2 and 02 injection rates are normal at Operator Interface Unit (OIU). (H2 injection rate should match the setpoint on the OIU. The 02 injection rate should match the setpoint on the 01111, which should be half of the H2 injection rate during normal steady state conditions.)

[2.2] IF H2 and 02 injection rates do NOT meet the above conditions, THEN NOTIFY the Unit Supervisor and INITIATE a HWC System shut down using either:

• 3-HS-4-40A H2 WATER CHEMISTRY CONTROL

[Panel 3-9-53] or

• 3-HS-4-40B H2 WATER CHEMISTRY CONTROL

[Panel 3-9-5] or

. 3-HS-4-39 HWC SHUTDOWN SWITCH [3-LPNL-925-0588].

DRIVER If directed to perform HWC ShutdowniQeally, inform Control Room that scaffold is in the way áannot ac ess switth. ONCE HWC isshutdown and 112 conç...tionj e 4%

THEN delete ilüre MF OG5A BOP Shutdown HWC System using either 3-HS-4-40A at panel 9-53 or 3-HS-4-40B at panel 9-5 SRO [4] IF hydrogen concentration is 4%, THEN REFER TO TRM 3.7.2 OnceHWOs Sbutdowu,I2Conc trauoaillbegm to Iowçr 1pc

3-A Page 13 of43 Simulator Event Guide:

Event 4 Instrument: HWC Malfunction SRO 3-AOI-66-l, Off Gas H2 High SRO NOTE Fuel failure is indicated by, but NOT limited to, rising activity on the following:

• OFF-GAS PRETREATMENT RADIATION recorder, 3-RR-90-157 (Panel 3-9-2)

MAiN STEAM LINE RADIATION recorder, 3-RR-90-135 (Panel 3-9-2)

• OFFGAS POST-TREATMENT RADIATION recorder, 3-RR-90-265

• On MAIN CONDENSERS (MN COND) ICS display:

Offgas pretreatment, post treatment, and stack radiation

[5] IF high hydrogen concentration is a result of possible fuel failure, ThEN REDUCE core flow to 50 60 % (otherwise N/A).

IRC mdieaton oFi Fe Exists, sja3 should be NA BOP Report H2 Concentration lowering slowly.

SRO [7] WhEN any of the following conditions exist, THEN INITIATE actions to reduce hydrogen concentration within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

• Hydrogen Analyzer on Panel 3-9-53 indicates? 4% hydrogen.

SRO REFER TO TRM 3.7.2 Condition A: With the concentration of hydrogen> 4% by volume Required Action A. 1: Restore the concentration to within the limit Completion Time: 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Whenxáiy,Reefrc Pump 3A High Vibration DRWER Upon Lead z ii kection,initiateThggerlQ for Recirc Pump 3A High Vibration;

3-A Page 14 of43 Simulator Event Guide:

Event 5 Component: Recirc Pump 3A High Vibration ATC Responds to alarm, RECIRC PUMP MTR A VIBRATION HIGH.

BOP/ATC A. CHECKS temperatures for RECIRC PMP MTR 3A13B WINDING AND BRG TEMP recorder, 3-TR-68-71 on Panel 3-9-21 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. CHECKS for a rise in Drywell equip sump pumpout rate, due to seal leakage.

C. DISPATCHES personnel to 3-LPNL-925-0712, (Vibration Mon. System) on EL 565 (S-Ri 7), to REPORT the Vibration Data for Pump A and any other alarm indications, to the Unit Operator. 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 3-AOI-68-1A or 3-AOI-68-1B.

F. IF pump operation continues, TREN RECORD pump 3A seal parameters hourly on Attachment 1, Page 22 of this ARP.

DRWER: W1aendispatcbód, rpor all vibration points are elevated and point 3-XI-068-0059D is at Aftespee41oiyeed ybjaio 4ng1eiç41ightly, pojQt59lis 12 mils.

fspowedteiin 2GRvi inta1i3 de1ç th ndinfonha91Yis 10 tuils ATC Lowers Pump Speed in an attempt to reset high vibration alarm.

DRIVER IF Speed is lowereda second time, vibration readings and pointS9D islO nils. THEN Delete th12 a:

SRO Determine whether to remove RR Pump 3A.

ATC Records seal parameters hourly for RR Pump 3A.

NRc.

IttS7EI jjçjLeid ammnejdigtion, ml mat 1Ie5 for IJtiitmdfbn

3-A Page 15 of43 Simulator Event Guide:

Event 5 Component: Contingent if SRO removes RR Pump 3A SRO Directs RR Pump 3A Shutdown, JAW 3-01-68, Section 7.2.

NOTE: Tripping of theReactorRecinpup under these conditions is an undesirable i action ATC 7.2 Stopping a Recirc Pump (Mode 1) & Single Loop Operation CAUTIONS

1) Prior to stopping a Recirc Pump, all attempts should be made to evaluate where the plant conditions will end up, when a Recirc Pump is removed from service. If practical, the control rod line should always be below 95.2% before stopping a Recirc Pump. At BFN, deliberate entry into Regions 1, 2, or 3 is NOT permitted.

2) Per Technical Specifications, the reactor CAN BE operated indefinitely with one Recirc loop out of service, provided the requirements of T.S. 3.4.1 are implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of entering single loop operations.

ATC [1] IF stopping of the 3A Recirc Pump is immediately required, THEN PERFORM the following: (Otherwise N/A)

ATC [4] REDUCE reactor power by a combination of control rod insertions and core flow changes, as recommended by the Reactor Engineer/Unit Supervisor, to maintain operating recirc pump flow less than 46,600 gpm. REFER TO 3-G0I-100-12, 3-GOI-100-12A, and 3-SR-3.1.3.5(A).

ATC [5] WifEN desired to control Recirc Pumps 3A andlor 3B speed in preparation for shutting down a recirc drive, THEN ADJUST Recirc Pump speed 3A and/or 3B using the following push buttons as required:

Recirc Drive 3A RAISE SLOW, 3-HS-96-15A RAISE MEDIUM, 3-HS-96-15B LOWER SLOW, 3-HS-96-17A LOWER MEDIUM, 3-HS-96-17B LOWER FAST, 3-HS-96-17C DRMR If Reaciy Ener is conctedinfon crew to fcl1w UrghLd I&duction RCP NRC Inadvertent Jnitiation DRIVER Upon Lead examiner direction, initiate Tngger 5 for HPCI Initiation

3-A Page 16 of43 Simulator Event Guide:

Event 5 Component: Contingent if SRO removes RR Pump 3A NOTE**ripp gof the Rcaetor Recirepump under these conditions is an. undesirable NRC SRO Directs RR Pump 3A Shutdown, JAW 3-01-68, Section 7.2.

ATC [6] To shutdown Recirc Drive 3A:

PERFORM the following: (Otherwise N/A)

[6.1] FIRMLY DEPRESS RECIRC PUMP 3A SHUTDOWN, 3-HS-96-19.

[6.2] VERIFY Recirc Drive shuts down.

[6.3] VERIFY DRIVE RUNNING, 3-IL-96-41 is extinguished.

ATC [8] WHEN RECIRC LOOP A DIFF PRESS LOW 3-PDA-68-65 ALARMS, CLOSE, RECIRC PUMP 3A DISCHARGE VALVE, 3-HS-68-3A.

[10] WhEN conditions allow, THEN MAINTAIN operating jet pump loop flow greater than 41 x 106 lbmlhr (3-FI-68-46 or 3-FI-68-48).

NR When ready, HP CI Inadvertent Inftiation DRIVER Upon Lead examiner d$rççtioimtiate Trigger 5for HPC tb

3-A Page 17 of43 Simulator Event Guide:

Event 6 Component: HPCI Inadvertent Initiation BOP Recognizes and responds to an inadvertent HPCI initiation and reports it to the SRO.

Verifies by multiple indications that the initiation signal is not valid and reports it to the SRO.

SRO Directs BOP to trip HPCI and place the Aux Oil Pump in Pull-to-Lock.

BOP Trips HPCI and places the Aux Oil Pump in Pull-to-Lock (after turbine stops).

ATC Reports power / level! pressure stable after HPCI secured.

Reports FWLC system transferred from 3-element control to single-element control.

SRO Refer to Technical Specification 3.5.1 Condition C: HPCI System Inoperable Required Action C. 1: Verify by administrative means RCIC System is Operable C.2: Restore HPCI System to Operable status Completion Time C. 1: Immediately C.2: 14 Days Directs Instrument Mechanics to investigate the HPCI initiation logic.

DRIVER AcknówledgeNotffications and directions.

ATC Places FWLC system back in 3-element control per 3-01-3.

[1] IF desired to transfer level control from Single Element to Three Element, THEN PERFORM the following: (Otherwise N/A)

[1.1] VERIFY conditions in Note 2 are met for placing level control in Three Element.

[1.2] OBSERVE stable steam flow and Feedwater flow.

[1.3] DEPRESS THREE ELEMENT push-button, 3-HS-46-6/3.

. VERIFY green backlight for push-button illuminates.

[1.4] VERIFY extinguished green backlight for SINGLE ELEMENT push button, 3-HS-46-6/1.

[1.5] CHECK Reactor water level_stable.

Reports to US that FWLC placed back in 3-element control.

NRC WheaEeady, MajorHPCI Steam Leak Ptiorto starting HPCI steam leak O2tDLOSE,T inateThgger 2O fO

- HCJ SeL

3-A Page 18 of43 Simulator Event Guide:

Event 6 Component: HPCI Inadvertent Initiation NRC NOTE SuppressionPocil Level should not reach this pomt BOP Reports Suppression Chamber Water Level Abnormal, greater than (-) 1 SRO Enters EOI-2.

Monitor and Control Suppression Pool Level between -l inch and -6inch, (Appendix 1 8).

BOP Checks ECCS systems for sources of water.

Reports HPCI minimum flow 73-30 open, attempts close valve. (Valve will NOT remain closed with initiation signal in.)

Crew Directs AUO to valve locally to isolate.

biiiER When iera13 minutes and eoiready soltatb eiW1indiieoted by

óperator,GO TO Component Override, TUEN System 73, TilENFOV-73-30 Fail Now.

SRO Directs pump down of Torus per App 18.

SRO Can Suppression Pool Level Be Maintained Above -6 inches? - YES Can Suppression Pool Level Be Maintained Below -1 inches? - YES BOP/ATC Appendix 18 BOP/ATC IF Directed by SRO, THEN REMOVE water from Suppression Pool as follows:

DISPATCH personnel to perform the following (Unit 3 RB, El 519 ft, Torus Area):

ER, i hted BOP Aligns to pump down torus in Control Room, per Appendix 18.

b. IF Main Condenser is desired drain path, THEN OPEN 3-FCV-74-62, RHR MAIN CNDR FLUSH VALVE.

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

1) ESTABLISH communications with Radwaste.

2) OPEN 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE.

BOP Directs AUO to Start RHR Drain Pump.

DRIVER 4çtDrain Jimp, IRF]XOr RH10 and RHI 1A or]3 NRC When Ready, Major HPCI Steam Leak j)P Prior to sa II CI steam leak inodi

. ppçQSE THENimkate Trigger 20 fox

3-A Page 19 of43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak Crew Recognize rising HPCI Room Temperatures and Radiation Levels.

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

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

C. CHECK following on Panel 3-9-11 and NOTIFY RADCON if rising radiation levels are observed:

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

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

ATC/BOP VERIFIES HPCI STEAM LINE INBD ISOL VLV, 3-FCV-73-2 AN])

HPCI STEAM LINE OUTBD ISOL VLV, 3-FCV-73-3 CLOSE.

Attempts to isolate HPCI Steam Supply Valves.

Reports HPCI fails to isolate.

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

(Loop 1RHR 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.

DRJVER Wh reqiisd, wait 4 minutes and place RP A on alternateJRF RP4 and RPO3 Crew PA announcement to evacuate the HPCI quad or Reactor Building

3-A Page 20 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak SRO Enters EOI-3 on Secondary Containment (Area Radiation or Temperature).

SRO IF Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation 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/br, directs Operator to perform Appendix 8F.

DIVER frequesidvait S minutes utrepoiiApp&idix E conp1ete, enter bat ppO8e CT#i Enters EOI-1 RPV Control an directs P.eactor Scram before a** ternperatitre exceeds.

MAX Safe.

CT#2 Stopsatqp sign When ietuiiwc. ormorerças repveMix Safe Then ErnIjenc3r Dpre stiaiion is rejijd

3-A Page 21 of43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak CT#1 Enters EOI-i RPY Control and directs Reactor Scram before any temperature exceeds MAX Saf t!J2 Stcá a )tcp nWjn tem ijiyo orore areas are o/e Max Sz Then. Eniergpnc3r e ssunj1on içqirçd SRO EOI-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 EOIs 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 RFIR 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 -

DkIyER IF ED is Xitipated beeady to raise cI Steai teak JMF HPO$ d15 and take out the ramp to ensure we get greater than 215 degrees;

3-A Page 22 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak 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 DRIVER TEED is AxEi1pated be ready to raise IT,PC] Steam Leak IME HPt8 to 1 5 and take oufthe pensureye ggiter t1j4egees,

3-A Page 23 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak

, TI Enter&EOi4, RPV Control! and d octs eactor Scram 1

çi#i & pliiie Mitb Shutd SRO Reactor Pressure Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig ?- NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate.

Should Answer YES; during Scenario and direct Bypass Valves opened.

CT#Z THEN I defii qpessuiiation.

wesY$ wi1euLtw area peratur&s have çac1e 3XSafe(SEE PAGE:

IF RPV water level cannot be determined? NO -

Is any MSRV Cycling? NO -

IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? - NO IF Drywell Control air becomes unavailable? NO -

IF Boron injection is required? NO SRO Directs a Pressure Band. Should begin to lower Reactor Pressure with bypass valves, not to exceed 1000 cooldown; until SRO decides that ED is anticipated.

ATC/BOP Controls Reactor Pressure as directed with Bypass Valves.

When directed to Anticipate ED, Opens all bypass valves.

3-A Page 24 of 43 Simulator Event Guide:

Event 7 Major: HPCJ Steam Leak SRO 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? NO IF RPV water level cannot be determined? NO -

IF PC water level cannot maintained below 105 feet? - NO Restores and Maintains RPV Water Level between +2 and +51 inches, with one of the following injection sources:

Directs a Level Band of (+) 2 to (+) 51 inches with Feedwater, Appendix 5A.

ATC Maintains the prescribed level band, per Appendix 5A.

3-A Page 25 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC Maintains the prescribed level band, JAW 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.

2. VERIFY Condensate system in service, supplying suction to RFPs.

3. VERIFY OPEN 3-FCV-1-125(133)(141), RFPT 3A(3B)(3C) HP STEAM SUPPLY VALVE.

4. DEPRESS 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER, and VERIFY amber light is illuminated.

5. VERIFY a Main Oil Pump is running for RFPT to be started.

6. VERIFY that the green light is illuminated and the red light is extinguished above the following on Panel 3-9-5 3-HS-3-208A, RX WTR LVL CH A HI RFPT/MT TRIP RESET 3-HS-3-208B, RX WTR LVL CH B HI RFPT/MT TRIP RESET.

7. VERIFY OPEN the following valves:

3-FCV-3-75, HP HTR 3A1 FW OUTLET ISOL VLV 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VLV 3-FCV-3-77, HP HTR 3C1 FW OUTLET ISOL VLV

8. DEPRESS 3-HS-3-124A(150A)(175A), RFPT 3A(3B)(3C) TRIP RESET, and Verify that the turbine trip is RESET.

3-A Page 26 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC Maintains the prescribed level band, JAW Appendix 5A.

9. VERIFY OPEN 3-FSV-3-20(13)(6), RFP 3A(3B)(3C) M1N FLOW VALVE.

10. PLACE 3-HS-46-1 12A(138A)(163A), RFPT 3A(3B)(3C) START/LOCAL ENABLE, in START, AND VERIFY RFPT speed increases to approximately 600 rpm.

11. VERIFY OPEN 3-FCV-3-19(12)(5), RFP 3A(3B)(3C) DISCHARGE VALVE.

12. SLOWLY ADJUST RFPT speed UNTIL feedwater flow to the RPV is indicated, using ANY of the following methods on Panel 3-9-5:

• Individual 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

• Individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL in MANUAL, OR 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) 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 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)_SPEED CONTROL in AUTO.

3-A Page 27 of 43 Simulator Event Guide:

Event 8 Component: 1 ADS Valve Fails to Operate CT#2; Enters 3-C2, Emergency Depressunzation.

Will the Reactor Remain Subcritical Without Boron Under All Conditions ?- YES Is Drywell Pressure Above 2.4 psig? - NO Is Suppression Pool Level Above 5.5 feet? - YES Directs All ADS Valves Open.

CT#2 Opens 6Ai Valve&

Reports 1 ADS Valve failed to Open.

SRO Can 6 ADS Valves Be Opened? - NO Directs Opening of Additional MSRVs, as necessary, to establish 6 MSRVs Open.

ATC/BOP Opens 1 additional MSRV.

SRO Are At Least 4 MSRVs Open? - YES SRO Directs Reactor Level Restored to (+) 2 to (+) 51 inches with Condensate (Appendix 6A) or Core Spray (Appendix 6D, 6E) or LPCI (Appendix 6B, 6C)

ATC/BOP Restores Reactor Level to prescribed level band, reports Startup Level Controller failure and restores level with Core Spray Loop 2 or RHR Loop 2.

SRO Emergency Plan Classification is 3.1-S.

3-A Page 28 of 43 Simulator Event Guide:

Event 9 Component: Startup Level Control Valve Failure ATC Appendix 6A Injection with Condensate

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

3. VERIFY OPEN the following drain cooler inlet valves:

• 3-FCV-2-72, DRAiN COOLER 3A5 CNDS INLET ISOL VLV

• 3-FCV-2-84, DRAIN COOLER 3B5 CNDS INLET ISOL VLV

. 3-FCV-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-FCV-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-FCV-.3-24, HP HTR 3C2 FW INLET ISOL VLV

• 3-FCV-3.-75, HP HTR 3Al FW OUTLET ISOL VLV

• 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VLV

.__3-FCV-.3-.77, HP HTR 3C1_FW OUTLET ISOL VLV

6. VERIFY OPEN the following RFP suction valves:

• 3-FCV-2-83, REP 3A SUCTION VALVE

• 3-FCV-2-95, REP 3B SUCTION VALVE

• 3-FCV-2-108, RFP 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, RFW START-UP LEVEL CONTROL, to control injection.

ATC Reports failure of Start Up Level controller.

3-A Page 29 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC/BOP Appendix 6E Injection with Core Spray Loop 2

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. VERLFY 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 3FCV-75-53, CORE SPRAY SYS II 1NBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

Restores Level (+) 2 to (+) 51 inches.

3-A Page 30 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC/BOP Appendix 6C Injection with RHR Loop 2 LPCI Mode

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-FeV-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 CHBRJPOOL 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, R}{R SYS II LPCI OUTBD INJECT VALVE, as necessary to control injection.

9. MONITOR RHE. 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.

Restores Level (+) 2 to (+) 51 inches.

3-A Page 31 of43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak SRO Continues to evaluate Suppression Pool Level and other legs of EOI-2.

EOI-2 (Drywell Temperature)

SRO Monitor and Control DW Temp Below 160°F, using available DW Cooling.

Can Drywell Temp Be Maintained Below 160°F? YES -

SRO Verify H202 Analyzers placed in service, Appendix 19.

BOP Places H202 analyzers in service, lAW Appendix 19.

SRO EOI-2 Primary Containment (Pressure)

Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary. (Appendix 12)

Can Primary Containment pressure be maintained below 2.4 psig? YES -

SRO EOI-2 Suppression Pool (Temperature)

Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary. (Appendix 1 7A)

Can Suppression Pool Temperature Be Maintained Below 95°F? NO -

Operate all available suppression pool cooling using only R}IR Pumps not required to assure adequate core cooling by continuous injection (Appendix 1 7A)

Start RHR Loop 2 in Suppression Pool Cooling, if not being used for level control, JAW BOP/ATC Appendix 17A Tenninate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

All Control Rods are inserted.

Emergency Depressurization complete.

Reactor Level is restored and maintained.

3-A Page 32 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC/BOP Initiates Suppression Pool Cooling per Appendix 17A 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 R}IR SYSTEM 1(11) in Suppression Pool Cooling as follows:

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

b. VERIFY R}IRSW 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, RUR FLX 3B RHRSW OUTLET VLV

• 3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV

• 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV

d. IF Directed by SRO, THEN PLACE 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT 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 1NBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

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

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

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

All Control Rods are inserted.

Emergency Depressurization complete.

Reactor Level is restored and maintained.

3-A Page 33 of 43 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak BOP Places H202 analyzers in service, JAW Appendix 19.

5. IF H2/02 Analyzer is in STANDBY at 3-MON-76-l 10 (Panel 3-9-55), THEN PLACE H2/02 Analyzer in service at as follows:

( Touch screen actions unavailable in the simulator)

6. VERIFY 112/02 ANALYZER SAMPLE PUMP ninring using 3-XI-76-l 10 (Panel 3-9-55).

7. VERIFY red LOW FLOW indicating light extinguished at 3-MON-76-1 10, H2/02 ANALYZER (Panel 3-9-55).

8. WhEN 112/02 Analyzer has been aligned and sampling for 10 minutes or greater, THEN OBTAIN H2 and 02 readings from 3-XR-76-1 10112/02 CONCENTRATION recorder (Panel 3-9-54).

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

All Control Rods are inserted.

Emergency Depressurization complete.

Reactor Level is restored and maintained.

3-A Page 34 of 43 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

RHRSW Pump B2 is out of service and tagged out.

APRM 3 is bypassed for IMD Surveillance testing.

Operations/Maintenance for the Shift:

Rotate Bus Duct Cooling Fans JAW 3-01-47 Section 6.11.1 [2j.

Once completed raise power with flow to 90% JAW 3-GOI-100-12 section 5.0 step 21 and the Reactivity Control Plan.

Units 1 and 2 are at 90% power.

Unusual Conditions/Problem Areas:

None

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3-A Page 38 of 43 AirbDme Efflirt TR 312 TR 17 PLNJT SYSTEMS TR 31.2 irborn Effiunts LCO 312 Whenever the SJAE s i serice the concentration of hydrogen in Lti wikni ur L[it uribirr [idI i lirii1U lu 4%

by volume.

PP1lCA5LiTY: During ma condaneroffgasfreathient stern operation

-NOTE TRM ICO 3.0.3 is not applicable.

ACTIONS CODTICN REQUIRED ACTIO$ COMPLETION TIME A With the concentration kl Restcre the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> o hyciogen >4% ty concentraicn to itNn oIunie. the lirnil

3-A Page 39 of 43 35 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.51 ECCS - Operating LCO 35l Each ECCS injectionispray subsystem and the Automatic Depressurization System (ADS) function of six safetylrelief valves shall be OPERABLE.

APPLICABILiTY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

-NOTE-LCO 3.04b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS 4.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem(s) to OPERABLE status.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inOperable.

(continued)

3-A Page 40 of 43 ECCS Operating 3.5i ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B. I Be ri MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

Bi Be n MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

3-A Page 41 of43 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C... HPCI System inoperable. C...1 Verify by administrative Immediately means RCIC System is OPERABLE.

AND C.2 Restore HPCI System to 14 days OPERABLE status.

D. HFCI System inoperable. 0.1 Restore HPCI System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

AND OR Condition A entered.

D.2 Restore [ow pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem to OPERABLE status.

E. One ADS valve E. I Restore ADS valve to 14 days inoperable.. OPERABLE status.

F. One ADS valve F.I Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status, AND Condition A entered. F.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem to OPERABLE status.

(continued)

3-A Page 42 of 43 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves G:l Be in MODE 3. 12 flours inoperable AND OR G2 Reduce reactor steam 36 flours Required Achon and dome pressure to associated Completion 150 psig.

Time of Condition C, D, E, or F not met H.. Two or more Low pressure Hi Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

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Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: B Op-Test No.: ILT 1102 FINAL SRO:

Examiners: Operators: ATC:

BOP:

Initial 1C191/ Unit 3 Reactor Power 90%. RCW Pump 3A tagged. 3-PI-3-207 Bypassed for Conditions: surveillance.

Turnover:

Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 per 3-SR-3.6.i.3.5 Section 7.6 and 7.7. Raise Reactor Power to 95%.

Event Event No. Maif. No. Type* Event Description N-BOP i Stroke time 2 PCIVs. The second valve will fail stroke time.

TS-SRO R-ATC 2 Raise Reactor Power with Recirc RSRO 3 thi8d VFD Cooling Water Pump 3-B-i failure C-BOP Steam Packing Exhauster Trip / STBY Exhauster Starts but discharge 4 trg 11 C-SRO damper fails to open.

5 pcl4 Leak on RHR Loop 1 Minimum Flow Line C-ATC Loss of RBCCW 3A Pump trip with Sectionalizing Valve 3-70-48 6 sw02a C-SRO failure to close 7 th33a M-ALL Drywell Leak with Emergency Depressurization on Drywell Temps 8 tcO2 C Bypass Valves Fail Closed 9 trg25 C RHR Loop I and II Drywell Sprays Fail 10 adO3 C 10 SRVs Fail Closed

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

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#1-When Drywell Pressure cannot be maintained below the PSP limit, US determines that Emergency Depressurization is required and RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure

3. Measured by:

Observation US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Drywell pressure exceeds the PSP limit.

AND Observation RO opens at least 6 SRVs during performance of Emergency Depressurization actions.

4. Feedback:

RPV pressure decreasing SRV open status indications OR CT#1-When Drywell Temperature cannot be maintained below the Drywell Temperature limit of 280°F, US determines that Emergency Depressurization is required and RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Temperature

3. Measured by:

Observation US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Drywell Temperature exceeds the limit of 280°F.

AND Observation RO opens at least 6 SRVs during performance of Emergency Depressurization actions or if six SRVs cannot be opened takes additional actions to depressurize the Reactor.

4. Feedback:

RPV pressure decreasing SRV open status indications

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#2- With a reactor scram required and the reactor not shutdown, take action to reduce power by initiating ARI to cause control rod insertion.

1. Safety Significance:

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

Correct reactivity control.

2. Cues:

Reactor power indication.

Procedural compliance.

3. Measured by:

Observation ARI pushbuttons armed and depressed to cause control rod insertion.

4. Feedback:

Reactor power trend.

Rod status indication.

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

BOP will perform Stroke Time Test on 3-FCV-43.13 and 3-FCV-43-14 with 3-FSV-43-14 failing the stroke time test. SRO will determine Technical Specification 3.6.1.3 Condition A required.

Then, the ATC will raise power with Reactor Recirculation flow to 95%.

Once evaluators satisfied with Reactivity Manipulations, 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.

Steam Packing Exhauster will trip and the STBY Exhauster will Start but the discharge damper will fail to open. The BOP will open the Steam Packing Exhauster discharge damper and restore Steam Packing Exhauster operation JAW with ARPs.

A leak will develop on RHR Loop 1 common minimum flow line, field reports will indicate the leak can be isolated by closing RHR A and C Pump suction valves. Once suction valves are closed SRO will determine Technical Specification 3.5.1 Condition A is required, TS 3.6.2.3 Condition B, 3.6.2.4 Condition B, and 3.6.2.5 Condition B all 7 Days.

After RHR Loop 1 is isolated an RBCCW Pump will trip and the sectionalizing valve will fail to close automatically. Operators will take actions JAW 3-AOI-70-1 and trip RWCU Pumps and close the sectionalizing valve for RBCCW.

A LOCA will occur, RPS will fail to de-energize, the crew will scram the Reactor by arming and depressing ARI, and enter EOI-1 and EOI-2. All rods will insert on ART, level control will be on feedwater and pressure control will be on SRVs(only three SRVs are available. The bypass valves fail closed during the scram. The LOCA will cause increasing DW Pressure and Temperature; the crew will take action JAW EOI-2. When the crew attempts to spray the Drywell, the Drywell Spray valves will fail to open. Unable to spray the drywell the crew will need to establish limits for DW pressure and temperature for anticipating ED and ED.

The Emergency classification is 2.1-A Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

All Control Rods are inserted.

Emergency Depressurization is complete Reactor Level is restored and maintained.

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-B 9 Total Malfunctions Inserted: List (4-8)

4. Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 90 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 RO SRO 1 Stroke Time Containment Isolation Valves RO U-064-SU-08 223002A2.08 2.7 3.1 SRO S-000-AD-8 1 2 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 3 VFD Cooling Water Pump Failure RO U-068-AL-33 202001A2.22 3.1 3.2 SRO S-068-AB-O1 4 Steam Packing Exhauster Trip RO U-47C-AL-02 271000A1.O1 3.3 3.2 SRO S-047-AB-03 5 RHR Loop 1 Leak RO U-77A-.AL-06 203000A4.02 4.1 4.1 SRO S-000-EM-09 6 Loss of RBCCW RO U-070-AL-03 206000A2.17 3.9 4.3 SRO S-070-AB-O1 7 Drywell LOCA RO U-000-EM-05 295028EA2.O1 4.0 4.1 SRO S-000-EM-04 SRO S-000-EM-05 SRO T-000-EM-15

3-B Page 7 of 65 Procedures Used/Referenced:

Procedure Number ] Procedure Title Procedure Revision 3-SR-3 .6.1.3.5 Primary Containment Isolation Valve Operability Test Revision 24 TS 3.6.1.3 Primary Containment Isolation Valves Amendment 212 3-GOI-100-12 Power Maneuvering Revision 35 3-01-68 Reactor Recirculation System Revision 80 3-ARP-9-4B Alarm Response Procedure Panel 3-9-4B Revision 42 3-ARP-9-7A Alarm Response Procedure Panel 3-9-7A Revision 22 3-ARP-9-4C Alarm Response Procedure Panel 3-9-4C Revision 33 3-ARP-9-3B Alarm Response Procedure Panel 3-9-3B Revision 18 TS 3.6.2.6 Drywell-to-Suppression Chamber Differential Pressure Amendment 212 3-EOI-3 Secondary Containment Control Flowchart Revision 9 TS 3.5.1 ECCS Operating

- Amendment 244 TS 3.6.2.3 Residual Heat Removal Suppression Pool Cooling Amendment 230 TS 3.6.2.4 Residual Heat Removal Suppression Pool Spray Amendment 212 TS 3.6.2.5 Residual Heat Removal Drywell Spray Amendment 212 3-A0I-70-1 Loss of Reactor Building Closed Cooling Water Revision 16 3-EOI-1 RPV Control Flowchart Revision 8 3-E0I-2 Primary Containment Control Flowchart Revision 7 3-EOI-APPENDIX-1 1A Alternate RPV Pressure Control Systems MSRVs Revision 2 3-EOI-APPENDIX-5A Injection Systems Lineup Condensate/Feedwater Revision 5 3-EOI-APPENDIX-1 9 H2/O2 Analyzer Operation Revision 0 3-EOI-APPENDIX-12 Primary Containment Venting Revision 3 3-EOI-APPENDIX-1 7A RHR System Operation Suppression Pool Cooling Revision 5 3-EOI-APPENDIX-1 7C RHR System Operation Suppression Chamber Sprays Revision 6 3-EOI-APPENIMX-17B R}IR System Operation Drywell Sprays Revision 5 3-EOI-3-C-2 Emergency RPV Depressurization Flowchart Revision 8

3-B Page 8 of 65 Procedures Used/Referenced Continued:

_Procedure Number ] Procedure Title Procedure Revision_1 3-EOI-APPENDIX-6A Injection Subsystems Lineup Condensate Revision 2 3-EOI-APPEMDIX-6D Injection Subsystems Lineup Core Spray System I Revision 3 3-EOI-APPENDD(-6E Injection Subsystems Lineup Core Spray System II Revision 3 3-EOI-APPENDIX-6C Injection Subsystems Lineup RHR System II LPCI Mode Revision 3 Emergency Classification Procedure Event Classification EPIP-l Revision 46 Matrix EPIP-3 Alert Revision 33 3-AOl-i 00-1 Reactor Scram Revision 53

3-B Page 9 of 65 Console Operator Instructions A. Scenario File Summary

1. File: batch and trigger files for scenario 3-B Batch nrc2Ollb

1. raw cooling water pump a clearance ior zlohs247a[l] off

1. surveillance 3.6.1.5 section 7.7 ior zlohs43 1 4a[2] (e3 0) on ior zlofcv43 14[2] (e3 0) on ior zloil64lb6[l] (e3 0) off

1. wide range pressure bypassed 3-207

1. vfd cooling pump failure ior zlohs682b2a[l] on ior zlohs682b2a[2] off mrfthl 8d trip ior zdihs682bla{1] (el 0) off trg 2 nrc2Ol lbvfd trg 2 = bat nrc2Ol ibi Trigger nrc2Ol lbvfd zdihs682b2a(3) .eq. 1 Batch nrc2Ollbl mrfthl8d close dor zlohs682b2a[1]

dor zlohs682b2a[2]

1. RBCCW pump trip imf sw02a (e5 0) ior zlohs7048a[1] off ior zlohs7O48a{2] on trg 6 nrc20117048 trg 6 = bat nrc2Ol 1b2 Trigger nrc20117048 zdihs7048a[ 1] .eq. 1 Batch nrc2Ollb2 dor zlohs7O48a{l]

dor zlohs7048a[2]

3-B Page 10of65

1. Steam packing blower trip ior ypomtrspea (eli 0) fail_controlpower ior ypovfcv6635 (eli 0) failpower_now ior zlohsó635a[l] on trg lOnrc2Ollspe trgi0=batnrc2oilspe Trigger nrc2Ol ispe zdihs6635a[3].eq. 1 Batch nrc2Ollspe dor ypovfcv6635 dor zlohs6635a[l]

1. RHR A leak imfpcl4 (el5 0)10 ior xa554c[17j (e15 30) alarm_on ior xa554c[24] alarm off ior xa554c[30] alarm off ior xa554c[3 1] alarm off

1. Major imfth33a (e20 0) .8 15 imftc02 (e20 0) 0 trg 25 nrc20l ldwspray2 ior zdihs7475a[2] auto imfth33b (e25 0) .5 180 imfrp07 imf ad03a imf ad03b imf ad03c imf ad03d imf ad03e imf ad03f ior xa553e[iOj (e30 0) alarm_on ior zdihs0l23[1] close/auto ior zdihs0l30[i] close/auto ior zdihs0l3la[i] close/auto ior zdihs0i42[lJ close/auto ior zdihs0l55a{2] auto ior zdihs0i56a[2] auto ior zdihs0i58a[2] auto ior zdihs0l59a[2] normal Trigger nrc2Ol ldwspray2 zdihs7474a(3).eq. 1

3-B Page 11 of65 Console Operator Instructions Scenario 3-B DESCRIPTION/ACTION Simulator Setup manual Reset to IC 191 Simulator Setup Load Batch Bat nrc2Ol lb Simulator Setup Place Green covers on Reactor manual Pressure indications two places.

Verify 3-PI-3-207 bypassed Simulator Setup manual Clearance out RCW Pump 3A Simulator Setup Verify Batch file loaded, clear VFD alarms RCP required (90% 95% wLRecirc flow)

- Provide marked up copy of 3-GOI-100-12 and RCP for Urgent Load Reduction.

Marked up Copy of 3-SR-3.6.1.3.5, for section 7.6 and 7.7 performance.

3-B Page 12 of65 Simulator Event Guide:

Event 1 Normal: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 JAW 3-SR-3.6.1.3.5 Section 7.6 and 7.7 SRO Directs BOP to perform 3-SR-3.6.1.3.5, Section 7.6.

BOP Performs 3-SR-3.6.l.3.5, Section 7.6.

7.6 3-FCV-43-13 Valve Stroke Timing

[1] RECORD the initial position of RX RECIRC SAMPLE INBD ISOLATION VLV, 3-FCV-43-l3. OPEN / CLOSED (Circle one)

[2] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE REACTOR RECWC SAMPLE 1NBD ISOL VLV, 3-HS-043-0013B OPEN position.

Pivq

[3] VERIFY OPEN 3-FCV-43-13 using RX REC1RC 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] VERIFY 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 N/A this section.)

[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)

3-B Page 13 of65 Simulator Event Guide:

Event 1 Normal: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 lAW 3-SR-3.6.l.3.5 Section 7.6 and 7.7 BOP [8] RETURN 3-FCV-43-13, to the initial position recorded in Step 7.6[l], 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.

Driv& Whca11ed 3-H Ø39is the CLOSIpbsitgn 7.7 3-FCV-43-14 Valve Stroke Timing

[1] RECORD the initial position of RX RECIRC SAMPLE OUTBD ISOLATION VLV, 3-FCV-43-13. OPEN / CLOSED (Circle one)

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

Piiver

[3] VERIFY OPEN 3-FCV-43-14 using RX RECIRC SAMPLE OUTBD ISOLATION VLV, 3-HS-43-14A.

W1enValv3-FC-43-l4 isopenjnsertTrIgger3 udprepare to delete the 3 overrides 04 trigger 3. When3-FCV-43-l4is closed wait<a111iiw3m of 5pon in.letelh ovenides sothat 43-14 exceeds the maxinium stroke time

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

SRO [6] IF the time recorded in step 7.7[4] is more than the maximum value listed, THEN_DECLARE the valve INOPERABLE BOP Report Failure of 3-FCV-43-14 to stroke close within the Maximum allowed time.

3-B Page 14 of 65 Simulator Event Guide:

Event 1 Normal: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 JAW 3-SR-3.6.l.3.5 Section 7.6 and 7.7 SRO Dispatches personnel to investigate.

Refer to Technical Specification 3.6.1.3.

Condition A: NOTE Only applicable to penetration flow paths with two PCIVs.

One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits.

Required Action A. 1: Isolate the affected penetration flow path by use of at least one closed and dc-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

Required Action A.2 : Verif the affected penetration flow path is isolated.

Completion Time: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for main steam line and Once per 31 days for isolation devices outside primary containment

3-B Page 15 of 65 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Flow SRO Notifies ODS of power increase.

Direct Power increase using Recirc Flow, per 3-GOI-100-12.

[211 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

[lj 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-l 5A(l 5B).

AND/OR

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

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

[2] WhEN desired to control Recirc Pumps 3A and/or 3B speed with the RECII{C 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 WaterPmjFaiire, ioomatermi

3-B Page 16 of65 Simulator Event Guide:

Event 3 Component: VFD Cooling Water Pump 3-B-i Failure ATC Reports the following annunciators 4B-12, 28 and 32 RECIRC DRIVE 3B COOLANT FLOW LOW, REC1RC DRIVE 3B DRIVE ALARM and RECIRC DRIVE 3B PROCESS ALARM.

ATC Reports the 3-B-i VFD Cooling Water Pump for the B Recirc Pump, has tripped.

ATC Reports Standby Recirc Drive Cooling Water Pump3-B-2, failed to auto start.

ATC RECIRC DRVIE 3B COOLANT FLOW LOW STARTS RECIRC DRIVE cooling water pump and DISPATCHES personnel to the RECC 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.

Crew Verifies Standby pump started by pulling up ICS displays.

BOP Dispatches personnel to VFD.

Wait 4 minutes after dispatched, THEN report tripped VFD Pump is hot to touch, internal bc1osed, 48Ovoitb tripped (480 V SDBD 3A5D).

R1ER Upon Lead examiner direction, initiate Trigger II for Steam Packing Exhauster trip

3-B Page 17 of65 Simulator Event Guide:

Event 4 Component: SPE Packing Exhauster A Trip

• BOP Responds to Alarm 7A-12, Steam Packing Exhauster Vacuum Low.

7A-12, Steam Packing Exhauster Vacuum Low Automatic Action: Alternate SPE fan starts and discharge damper opens, and the running fans trips.

A. CHECKS the following:

1. Alternate STEAM PACKING EXHR BLOWER 3B, 3-HS-66-50A started.

2. 3B DISCHARGE VLV, 3-HS-66-34A opens.

BOP Determines that Alternate Blower started, but discharge damper fails to open.

Opens 3B DISCHARGE VLV, 3-HS-66-34A to restore SPE Vacuum.

NYIEJSPBB B)owçrindieatioawi}J baye Red and Greeiiit In qMerr Red 11ht cnly ndication L1ie crew wou1&hav to stop the A SP. lAW 3-I47Ø J& infes ud ortijbvidiis prb1eins tSPpr Breaker Wi I,

3-B Page 18 of 65 Simulator Event Guide:

Event 5 Component: RHR A Leak BOP/ATC Respond to Alarm 4C-17 RHR LOOP I PUMP ROOM FLOOD LEVEL HIGH, A. DISPATCH personnel to visually check the RHR pump room.

B. IF alarm is valid, THEN PERFORM the following

• VERIFY the floor drain sump pumps running.

• VERIFY the floor drains for proper drainage.

• IF possible, THEN DETERMINE the source of the leak and the leak rate.

• ENTER 3-EOI-3 FLOWCHART.

BOP/ATC Respond to Alarm 3B-26, DRYWELL TO SUPPR CHAMBER DIFF PRESS ABNORMAL A. CHECK alarm by checking Drywell to Suppression Chamber DP.

B. REFER TO 1-AOI-64-l.

C. REFER TO Tech Spec Section 3.6.2.6.

BOP/ATC Dispatches personnel to RHR Loop 1 area.

SRO Evaluates Tech Spec 3.6.2.6 and Enters EOI-3.

DRiVER 3 minuteflerthspathed, report leak is on thecommon minimum flow line for RHR Pumps A and C, th6 leak is betwçen the pumps and the Mm Flow Valve; appears leak was ausedby maintenance work in the area. Wbex the crew closes 74-1 and 74-i2report leak has stoppe4 an change PC14 toO. Cmnot aqçes any manual valves due to amount of vatersray. IfdnLyone of the R.HR Supwessionoo1 Suction Valves is <1osea, report that leak has iot slowed. In add&ton, reportwater 1eveisabout 8 inc s4iqiad ai1ere BOP/ATC Respond to Alarm 4C-3, SUPPR CHMB RM FLOOD LEVEL HIGH A. DISPATCH personnel to VISUALLY CHECK the suppression chamber room.

B. IF alarm is valid, THEN PERFORM the following:

• CHECK the floor drain sump pumps running.

• CHECK the floor drains for proper drainage.

• IF possible, THEN DETERMINE the source of the leak and the leak rate.

• ENTER 3-EOI-3 FLOWCHART.

SRO When leak source is reported, directs BOP to close 74-1 and 74-12, RHR Pump 3A and 3C Suppression Pool Suction Valves.

BOP Closes 74-1 and 74-12, RHR Pump 3A and 3C Suppression Pool Suction Valves.

DJUWER nd3CWait2O minutes then go tomponent vthdesan

3-B Page 19 of65 Simulator Event Guide:

Event 5 Component: RHR A Leak SRO EOI-3 (Secondary Containment Water Level)

Monitor and Control Secondary CNTMT Water Levels.

Answers Yes to: Is Any Area Water Level Above 2 inches?

Answers Yes to: Is Any Floor Drain Sump Water Level Above 66 inches?

Restores and Maintains floor drain sump levels and area water levels, using all available sump pumps.

When source of leak is determined and isolated, Answers Yes to: Can all floor drain and area water levels be restored and maintained?

BOP/ATC Contacts Radwaste to determine status of sump Pumps.

5yij fter74-1 a6474-i2 are isoated REPO1t sump pumps arc opçr iia11y, Iarm. LEI qverride ona1aim iotxa554cj1 7] 1amaon SRO EOI-3 (Temperature)

Monitor and Control Secondary Containment Temperatures.

Operate all available ventilation. (Appendix 8F)

Defeat isolation interlocks, as necessary. (Appendix 8E)

Answers NO to: Is Any Area Temperature Above Max Normal?

SRO EOI-3 (Radiation)

Monitor and Control Secondary CNTMT Radiation Levels.

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

DRIEJ UpoiiLe1 exa t&ii. mtitTrigger 1

i of RBCO

3-B Page 20 of 65 Simulator Event Guide:

Event 5 Component: RHR A Leak SRO Refer to Technical Specification 3.5.1, 3.6.2.3, 3.6.2.4, 3.6.2.5, and 3.6.2.6 TS 3.5.1 Condition A: One low pressure ECCS injection/spray subsystem inoperable.

Required Action A. 1: Restore low pressure ECCS injection/spray subsystem to Operable status.

Completion Time: 7 Days TS 3.6.2.3 Condition B: Two RHR suppression pooi cooling subsystems inoperable.

Required Action B. 1: Restore one R}TR suppression pool cooling subsystem to Operable status.

Completion Time: 7 Days TS 3.6.2.4 Condition B: Two RHR suppression pool spray subsystems inoperable.

Required Action B. 1: Restore one RHR suppression pool spray subsystem to Operable status.

Completion Time: 7 Days TS 3.6.2.5 Condition B: Two RHR drywell spray subsystems inoperable.

Required Action B. 1: Restore one RHR drywell spray subsystem to Operable status.

Completion Time: 7 Days TS 3.6.2.6: No Entry required M at

3-B Page 21 of65 Simulator Event Guide:

Event 6 Component: Loss of RBCCW Pump 3A Responds to alarm 4C-12, RBCCW PUMP DISCH. HDR PRESS LOW BOP/ATC Report Trip of RBCCW Pump 3A.

BOP/ATC Automatic Action: Closes 3-FCV-70-48, non-essential ioop, closed cooling water sectionalizing MOV.

A. VERIFY 3-FCV-70-48 CLOSiNG/CLOSED.

B. VERIFY RBCCW pumps A and B in service.

C. VERIFY RBCCW surge tank low level alarm is reset.

D. DISPATCH personnel to check the following:

• RBCCW surge tank level locally.

• RBCCW pumps for proper operation.

E. REFER TO 3-AOI-70-l, for RBCCW System failure and 3-01-70, for starting spare pump.

SRO Enters 3-AOI-70-l.

ATC Closes 3-FCV-70-48 and report the sectionalizing valve failed to close automatically BOP Dispatch Personnel to investigate RBCCW Pump 3A trip DRIVER When dispathed. rep&tRJ3CCW Pump 3A lreaker istripped free, Ther isaIsQ. sriieI of urnt irg ançt earringontbe brea1er ATC 3-AOI-70-1 4.1 Immediate Actions

[1] IF RBCCW Pump(s) has tripped, THEN Perform the following

. SECURE RWCU Pumps.

• VERIFY RBCCW SECTIONALIZING VLV, 3-FCV-70-48 CLOSED.

ATC Secures RWCU Pumps and Closes 3-FCV-70-48.

3-B Page 22 of 65 Simulator Event Guide:

Event 6 Component: Loss of RBCCW Pump 3A 4.2 Subsequent Actions

[1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AND core flow is above 60%,THEN: (Otherwise N/A):

[2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (Otherwise N/A).

Step 1 and 2 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] IFno damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s).

D1MI &diai4hed. ¶COWi&np Araicer i iisiiói Jgdjhaiing SRO [4] IF unable to restart a tripped pump, THEN PLACE Spare RBCCW Pump in service. REFER TO 3-01-70. Direct Unit 1 to place Spare RBCCW Pump in service r-When calletho place spareRBCCW 1umpm service, wait (F,SW E1{

jrvce 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 SECTIONALIZ1NG VLV, 3-HS-70-48A.

[5.2] RESTORE the RWCU system to operation. (REFER TO 3-01-69)

Directs ATC or BOP to Open Sectionalizing Valve and Restore RWCU.

ATC Opens Sectionalizing Valve, 3-FCV-70-48.

dj$fain pe ptpme DR1VEI upon Lead4x r4çppox, tiø7pggQO fge,Lni

3-B Page 23 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment Crew Recognize rising Drywell Pressure and Temperature.

SRO Directs a Reactor Scram, prior to 2.4 psig in the Drywell.

ATC Manually scrams the reactor.

T VerffjesU KoThitiion!

SRO Enters EOI-1 and EOI-2.

SRO EOI- 1 (Reactor Pressure)

Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? YES, but action Not Required IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO-IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? NO IF RPV water level cannot be determined? NO

3-B Page 24 of 65 Simulator Event Guide:

Event 8 Component: Bypass Valves Fail Closed ATC/BOP Report failure of Bypass Valves to control Reactor Pressure Is any MSRV Cycling? YES Direct Manually open MSRVs until RPV Pressure drops to the pressure at which all turbine bypass valves are open. (Appendix 11 A)

IF Steam cooling is required? NO-IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?- NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO-IF Boron injection is required? NO-SRO Directs a Pressure Band with SRVs, lAW Appendix 1 1A.

Should begin to lower Reactor Pressure, not to exceed 100° cooldown.

Control Reactor Pressure in assigned band, JAW Appendix 1 1A.

3-B Page 25 of 65 Simulator Event Guide:

Event 8 Component: Bypass Valves Fail Closed ATCIBOP Pressure Control JAW Appendixi 1A, RPV Pressure Control SRVs 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 bySRO:

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

b. 3-PCV-1-180 MN STM LINED RELIEF VALVE.

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

d. 3-PCV-1-31 MN STM LINE C RELIEF VALVE desowør1

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

f. 3-PCV-i -42 MN STM LINE D RELIEF VALVE dOS nvork

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

h. 3-PCV-i-19 MN STM LINE B RELIEF VALVE. dsnotvpr1

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

j. 3-PCV-i-41 MN STM LINED RELIEF VALVE do

k. 3-PCV-1-22 MN STM LINE B RELIEF VALVE &esnot1o

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

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

3-B Page 26 of 65 Simulator Event Guide:

Event 8 Component: Bypass Valves Fail Closed EOI-l RPV Pressure Augment RPV Pressure control, as necessary; with one or more of the following depressurization systems: HPCI Appendix 11C, RCIC Appendix 11B, RFPTs SRO on minimum flow Appendix 1 lF, Main Steam System Drains Appendix 1 1D, Steam Seals Appendix 11 G, SJAEs Appendix 11 G, Off Gas Preheater Appendix 11 G, RWCU Appendix liE.

ATC/BOP Augments RPV Pressure Control, if directed by SRO.

SRO EOI- 1 (Reactor Level)

Monitor and Control Reactor Water Level.

Directs Verification of PCIS isolations.

ATC/BOP Verifies PCIS isolations.

SRO Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with one or more of the following injection sources. (Condensate and Feedwater, Appendix 5A)

ATC Maintains the prescribed level band, lAW Appendix 5A.

3-B Page 27 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC Maintains the prescribed level band lAW 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.

2. VERIFY Condensate system in service, supplying suction to REPs.

3. VERIFY OPEN 3-FCV-l-125(133)(141), RFPT 3A(3B)(3C) HP STEAM SUPPLY VALVE.

4. DEPRESS 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER, and VERIFY amber light is illuminated.

5. VERIFY a Main Oil Pump is running for RFPT to be started.

6. VERIFY that the green light is illuminated and the red light is extinguished above the following on Panel 3-9-5

• 3-HS-3-208A, RX WTR LVL CH A HI RFPT/MT TRIP RESET

• 3-HS-3-208B, RX WTR LVL CH B HI RFPT/MT TRIP RESET.

7. VERiFY OPEN the following valves:

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

• 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VLV

• 3-FCV-3-77, HP HTR 3C1 FW OUTLET ISOL VLV

8. DEPRESS 3-HS-3-124A(150A)(175A), RFPT 3A(3B)(3C) TRIP RESET, and VERIFY that the turbine trip is RESET.

3-B Page 28 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC Maintains the prescribed level band, JAW Appendix 5A.

9. VERIFY OPEN 3-FSV-3-20(13)(6), RFP 3A(3B)(3C) MIN FLOW VALVE.

10. PLACE 3-HS-46-l 12A(138A)(163A), RFPT 3A(3B)(3C) START/LOCAL ENABLE, in START, AND VERIFY RFPT speed increases to approximately 600 rpm.

11. VERIFY OPEN 3-FCV-3-19(12)(5), REP 3A(3B)(3C) DISCHARGE VALVE.

12. SLOWLY ADJUST RFPT speed UNTIL feedwater flow to the RPV is indicated, using ANY of the following methods on Panel 3-9-5:

• Individual 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

• Individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL in MANUAL, OR

• 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) 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 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)_SPEED CONTROL in AUTO.

3-B Page 29 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment SRO Enters EOI-2 all legs, EOI-2 (Drywell Temperature)

SRO Monitor and Control DW Temp Below 160°F, using available DW Cooling.

Can Drywell Temp Be Maintained Below 160°F? NO -

SRO Directs H202 Analyzers placed in service, JAW Appendix 19.

BOP Places H202 analyzers in service, lAW Appendix 19.

SRO EOI-2 (Primary Containment Pressure)

Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary. (Appendix 12)

SRO Directs venting of Primary Containment, per Appendix 12.

Can PC Pressure Be Maintained Below 2.4 psig? NO-Vents Primary Containment, JAW Appendix 12.

EOI-2 (Suppression Pool Temperature)

Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool_Cooling As Necessary._(Appendix_17A)

Can Suppression Pool Temperature Be Maintained Below 95°F? NO -

ATC Places Suppression Pool Cooling in service, lAW Appendix 1 7A.

3-B Page 30 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment SRO EOI-2 (Suppression Pool Level)

Monitor and Control Suppression Pool Level between (-) 1 inch and (-) 6 inches. (Appendix 18)

Can Suppression Pool Level Be Maintained above (-) 6 inches? - YES Can Suppression Pool Level Be Maintained below (-) 1 inch? - YES

3-B Page 31 of65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment BOP Places H202 analyzers in service, JAW Appendix 19.

1. IF A Group 6 PCIS signal exists, THEN PLACE 3-HS-76-69, H2/02 ANALYZER ISOLATION BYPASS switch in BYPASS (Panel 3-9-54).

2. DEPRESS 3-HS-76-91, H2/02 ANALYZER ISOLATION RESET.

3. IF H2/02 Analyzer is to sample the Suppression Chamber, THEN ALIGN Analyzer as follows (Panel 3-9-54):

a. PLACE 3-HS-76-1 10, H2/02 ANALYZER DW/SUPPR CHBR SELECT in SUPPR CHBR position.

b. VERIFY SUPPR CHBR SMPL VLVS 3-FSV-76-55/56 OPEN using 3-IL-76-49-1.

c. VERIFY OPEN SMPL RTN VLVS 3-FSV-76-57/58 using 3-IL-76-49-3.

4. IF H2/02 Analyzer is to sample the Drywell, THEN ALIGN Analyzer as follows (Panel 3-9-54):

a. PLACE 3-HS-76-1 10, H2/02 ANALYZER DW/SUPPR CHBR SELECT in DRYWELL position.

b. VERIFY OPEN DRYWELL SMPL VLVS 3-FSV-76-49/50 using 3-IL-76-49-2.

c. VERIFY OPEN SMPL RTN VLVS 3-FSV-76-57/58 using 3-JL-76-49-3.

3-B Page 32 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment BOP Places H202 analyzers in service, JAW Appendix 19.

5. IF H2/02 Analyzer was in service prior to sample path isolation (Panel 3-9-55), THEN RETURN H2/02 Analyzer to service as follows:

a. TOUCH 3-MON 110 display screen if required to restore display.

b. DEPRESS flashing FLOW / 0/P RESET soft key in upper right quarter of the MAiN (2 GAS MONITORiNG) screen.

6. IF H2/02 Analyzer is in STANDBY at 3-MON-76-l 10 (Panel 3-9-55), THEN PLACE H2/O2 Analyzer in service at as follows:

a. TOUCH 3-MON-76-l 10 display screen.

b. DEPRESS Go To Panel PROCESS VALUES soft key.

c. DEPRESS Go To Panel MAINI MENU soft key.

d. DEPRESS LOG ON soft key.

e. ENTER password 1915 on soft keypad.

f. DEPRESS ENT soft key on keypad.

g. DEPRESS STANDBY MODE ON soft key to enable sample pump operation.

h. VERIFY soft key reads STANDBY MODE OFF.

i. DEPRESS Go To Panel PROCESS VALUES soft key.

j. DEPRESS Go To Panel MAIN soft key.

k. VERIFY STANDBY MODE is NOT displayed.

7. VERIFY H2/02 ANALYZER SAMPLE PUMP running using 3-XI-76-1 10 (Panel 3-9-55).

8. VERIFY red LOW FLOW indicating light extinguished at 3-MON 110, H2/O2 ANALYZER (Panel 3-9-55).

9. WHEN H2/O2 Analyzer has been aligned and sampling for 10 minutes or greater, THEN OBTAIN H2 and 02 readings from 3-XR-76-1 10 H2/O2 CONCENTRATION recorder (Panel 3-9-54).

3-B Page 33 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment BOP Vents Primary Contaitunent JAW Appendix 12 VERIFY at least one SGTS train in service.

2. VERIFY CLOSED the following valves (Panel 3-9-3 or Panel 3-9-54):

• 3-FCV-64-31, DRYWELL INBOARD ISOLATION VLV,

. 3-FCV-64-29, DRYWELL VENT INBD ISOL VALVE,

. 3-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV,

• 3-FCV-64-32, SUPPR CHBR VENT JNBD ISOL VALVE.

Steps 3, 4, 5 and 6 are If! Then steps that do not apply.

7. CONTINUE in this procedure at:

Step 8 to vent the Suppression Chamber through 3-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 3-FCV-84-20.

8. VENT the Suppression Chamber using 3-FIC-84-19, PATH B VENT FLOW CONT, as follows:

a. PLACE keylock switch 3-HS-84-35, DW/SUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 3-9-54).

b. VERIFY OPEN 3-FCV-64-32, SUPPR CHBR VENT 1NBD ISOL VALVE (Panel 3-9-54).

c. PLACE 3-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfiri (Panel 3-9-55).

d. PLACE keylock switch 3-HS-84-19, 3-FCV-84-19 CONTROL, in OPEN (Panel 3-9-55).

e. VERIFY 3-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfim.

f. CONTINUE in this procedure at step 12.

3-B Page 34 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment BOP Vents Primary Containment lAW Appendix 12

9. VENT the Suppression Chamber using 3-FIC-84-20, PATH A VENT FLOW CONT, as follows:

a. VERIFY OPEN 3-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 3-9-3).

b. PLACE keylock switch 3-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 3-9-5 4).

c. VERIFY OPEN 3-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV (Panel 3-9-54).

d. VERIFY 3-FIC-.84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 3-9-55).

e. PLACE keylock switch 3-HS-84-20, 3-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 3-9-55).

f. VERIFY 3-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.

g. CONTINUE in this procedure at step 12.

12. ADJUST 3-FIC-84-19, PATH B VENT FLOW CONT, or 3-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:

Stable flow as indicated on controller, AND 3-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 jiCi/s AND 0-SI-4.8.B.1.a.l release fraction of 1.

RIVER Acknowledges Notification.

3-B Page 35 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC Place Suppression Pool Cooling in service lAW Appendix 1 7A 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-74-155B, LPCI SYS II OUTBD 1NJ VLV BYPASS SEL in BYPASS.

2. PLACE RHR SYSTEM II 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-23-46, RHR HX 3B RHRSW OUTLET VLV

• 3-FCV-23-52, RHR HX 3D RI-IRSW OUTLET VLV

d. IF Directed by SRO, THEN PLACE 3-XS-74-l 30, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD in MANTJAL OVERRIDE.

e. IF LPCI INITIATION Signal exists, ThEN MOMENTARILY PLACE 3-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT in SELECT.

f. IF 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

g. OPEN 3-FCV-74-71, RHR SYS II SUPPR CHBRIPOOL ISOL VLV.

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

i. THROTTLE 3-FCV-74-73, RHR SYS II SIJPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-FI-74-64, RHR SYS II 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-74-30, RHR SYSTEM II MIN FLOW VALVE.

k. MONITOR RHR Pump NPSH using Attachment 1.

3-B Page 36 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment SRO Can Drywell Temp Be Maintained Below 160°F? NO -

Operate all available Drywell Cooling.

Before DW Temperature rises to 200°F, Continue EOI-1 RPV Control and SCRAM the Reactor Before DW Temperature rises to 280°F, Continue Stops at STOP sign.

SRO EOI-2 Primary Containment Pressure Before Suppression Chamber Pressure rises to 12 psig, Continue Initiate Suppression Chamber Sprays, Using only pumps not required to assure adequate core cooling by continuous injection. (Appendix 17C)

3-B Page 37 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment SRO Directs Operator to initiate Suppression Chamber Sprays, lAW Appendix 17C.

ATC/BOP Initiates Suppression Chamber Sprays, JAW Appendix 17C.

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

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

Step 3 and 4 are NA.

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRII)E.

c. MOMENTARILY PLACE 3-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 3-FCV-74-67, RHR SYS II INED INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS II OUTBD INJECT VALVE.

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

f. VERIFY OPEN 3-FCV-74-71, RHR SYS II SUPPR CHBRJPOOL ISOL VLV.

3-B Page 38 of65 Simulator Event Guide:

Event 7: Main Steam Line Leak inside Containment ATC/BOP g. OPEN 3-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE.

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

i. VERIFY CLOSED 3-FCV-74-30, R}IR. SYSTEM II M1N FLOW VALVE.

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

k. MONITOR RHR Pump NPSH using Attachment 2.

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

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

• 3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV

• 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

3-B Page 39 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment SRO EOI-2 (Drywell Temperature)

Before DW Temperature rises to 280°F, Continue Is Suppression Pool level below 18 feet? YES Are DW Temperature and Pressure within the safe area of curve 5? YES Direct Operators to shutdown Recirc Pumps and Drywell Blowers.

ATC Trips Reactor Recirculation Pumps.

BOP Places all Drywell Blowers in Off.

SRO Initiate DW Sprays, using only pumps not required to assure adequate core cooling; by continuous injection. (Appendix 1 7B)

ATC/BOP Initiate DW Sprays, JAW Appendix 17B.

SRO EOI-2 (Primary Containment Pressure)

When Suppression Chamber Pressure exceeds 12 psig, THEN Continue Is Suppression Pool level below 18 feet YES Are DW Temperature and Pressure within the safe area of curve 5 YES Directs Operators to shutdown Recirc Pumps and Drywell Blowers.

ATC Trips Reactor Recirculation Pumps.

BOP Places all Drywell Blowers in Off.

SRO Initiate DW Sprays; using only pumps not required to assure adequate core cooling; by continuous injection. (Appendix 1 7B)

ATC/BOP Initiate DW Sprays, JAW Appendix 17B.

3-B Page 40 of 65 Simulator Event Guide:

Event 9 Component: RHR Loop I and II Drywell Sprays Fail ATC/BOP Initiate DW Sprays, lAW Appendix 17B.

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 1-HS-74-155A, LPCI SYS I OUTBD JNJ VLV BYPASS SEL in BYPASS.

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

2. VERIFY Recirc Pumps and Drywell Blowers shutdown.

3. IF Directed by SRO to spray the Drywell using RHR System II, THEN CONTINUE in this procedure at Step 6 using RHR Loop II.

6. INITIATE Drywell Sprays using R}IR Loop 1(11) as follows:

a. BEFORE drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 9.

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

c. IF EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, THEN PLACE keylock switch l-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

d. MOMENTARILY PLACE 1-.XS-74-29, RHR SYS II CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

e. IF l-FCV-74-67, RHR SYS II LPCI 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED l-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

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

g. OPEN the following valves:

• l-FCV-74-74, RT{R SYS II DW SPRAY OUTBD VLV.

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

ATC/BOP Reports Failure of Drywell Spray Valve on RHR Loop II.

3-B Page 41 of65 Simulator Event Guide:

Event 9 Component: RHR Loop I and II Drywell Sprays Fail SRO When Loop 2 Drywell Sprays fails direct DW Sprays using Standby Coolant ATC/BOP Initiate DW Sprays, lAW Appendix 17B. with Standby Coolant

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

8. INITIATE Drywell Spray 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 3-XS-74-122, RHR SYS I LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

b. MOMENTARILY PLACE 3-XS-74-121, RHR SYS I CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

c. IF 3-FCV-74-53, RHR SYS I LPCI 1ISTBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52, RHR SYS I LPCI OUTBD iNJECT VALVE.

d. 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-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

• 3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV.

e. VERIFY RHR Pumps 3A and 3C are NOT running.

f. PLACE 3-BKR-074-0 100, RHR HTX A-C DISCH XTIE (TO U-2) VLV FCV-74-100 (M010-171) to ON (480V RMOV Board 3B, Compartment 19A).

Driver Rçport 3-EKR-074-0100 Tnps free and cannot be closeçl Mamtenance contacted flV,çr If required mcreç severity ofDrywefl Steamakto ensure Crew EDs ATC/BOP Reports Failure of Drywell Sprays using Standby Coolant.

3-B Page 42 of 65 Simulator Event Guide:

Event 10 Component: 10 SRVs Fail Closed SRO EOI-2 (Drywell Temperature)

CT#1 Can Drywell Temperature be Maintained below 280°F? NO -

Emergency RPV Depressurization is required.

CT#1 Enters EOI-C2.

Will the Reactor remain subcritical without Boron under all conditions? YES Is Drywell Pressure Above 2.4 psig? YES Prevent Injection from only those CS and LPCI Pumps; not required to assure adequate core cooling. (Appendix 4)

Is Suppression Pool level above 5.5 feet? YES Direct ATC/BOP to Open all ADS Valves.

CT#1 Open 6 ADS Valves SRO Can 6 ADS Valves be opened NO -

Open additional MSRVs as necessary to establish 6 MSRVs open Are at least 4 MSRVs open NO (dependent upon whether crew opens additional MSRVs from the back-up control panel)

DR1VE1 ftsth back-up janefio AJ5SiSR1s wai1miuts and msert et,faipr ad3a,c,eandf

3-B Page 43 of 65 Simulator Event Guide:

Event 10 Component: 10 SRVs Fail Closed Is RPV Press 70 PSI or more above Suppression Chamber Pressure YES APOLY tPRtSUkK.E fl4& TO L 1 ovtw crn siwm o oooi LOWGSYSThMS OPROSA1OYS1LM X*JNSLi I HPC TTSY 44O LNOUP. LtVL1Jt TLSTMO lie F? 4MTLO MAN MSYSTMbRMIS 10

$EAMSTALS 110 110

s 0A0LH1A0F 10 1{ACVINT

-pci i0COPANS I Li RWJ 0 OR010O IlL 04Ji0LYAN iOuNO Ol& 0:

• AiN LC?LO i Li?)

• IWL RLACTO wti RL% SClOl0L wpIOiJr eoFI?wrnRAu. ThN0

?SLLNOM$

SRO Directs additional Depressurization Methods from the chart above or directs ADS Valves be opened from Back Up Control Panel SRO EOI-1 Level SRO Restore and Maintain RPV Water Level between +2 to 51 inches with one or more of the following injection sources. Condensate Appendix 6A, Core Spray Appendix 6D or 6E, LPCI Appendix 6C ATC/BOP Restore and maintain level +2 to +51 inches lAW Appendix 6A, 6D, 6E, or 6C SRO Emergency Plan Classification 2.1 -A

3-B Page 44 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Restore and maintain level +2 to +51 inches JAW Appendix 6A, 6D, 6E, or 6C Condensate Appendix 6A

1. VERIFY CLOSED the following feedwater heater return valves:

• 1-FCV-3-71, HP HTR 1A1 LONG CYCLE TO CNDR

• 1-FCV-3-72, HP HTR 1B1 LONG CYCLE TO CNI)R

• l-FCV-3-73, HP HTR 1C1 LONG CYCLE TO CNDR

2. VERIFY CLOSED the following RFP discharge valves:

• 1-FCV-3-19, RFP 1A DISCHARGE VALVE

• l-FCV-3-12, RFP lB DISCHARGE VALVE

• l-FCV-3-5, REP 1C DISCHARGE VALVE.

3. VERIFY OPEN the following drain cooler inlet valves:

• 1-FCV-2-72, DRAIN COOLER 1A5 CNI)S INLET ISOL VLV

• l-FCV-2-84, DRAiN COOLER 1B5 CNI)S 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 lA3 CNDS OUTL ISOL VLV

• l-FCV-2-125, LP HEATER lB3 CNDS OUTL ISOL VLV

• 1-FCV-2-126, LP HEATER 1C3 CNDS OUTL ISOL VLV.

5. VERIFY OPEN the following heater isolation valves:

• l-FCV-3-38, HP HTR 1A2 FW INLET ISOL VALVE

• l-FCV-3-31, HP HTR 1B2 FW INLET ISOL VALVE

• l-FCV-3-24, HP HTR 1C2 FW iNLET ISOL VALVE

• 1-FCV-3-75, HP HTR 1A1 FW OUTLET ISOL VALVE

• l-FCV-3-76, HP HTR 1B1 FW OUTLET ISOL VALVE

• l-FCV-3-77, HP HTR 1C1 FW OUTLET ISOL VALVE

6. VERIFY OPEN the following REP suction valves:

• l-FCV-2-83, REP 1A SUCTION VALVE

• l-FCV-2-95, REP lB SUCTION VALVE

• l-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 RFW flow to RPV.

3-B Page 45 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Restore and maintain level +2 to +51 inches JAW Appendix 6A, 6D, 6E, or 6C Core Spray System I Appendix 6D

1. VERIFY OPEN the following valves:

• l-FCV-75-2, CORE SPRAY PUMP 1A SUPPR POOL SUCT VLV

• 1-FCV-75..ll, CORE SPRAY PUMP 1C SUPPR POOL SUCTVLV

• l-FCV-75-23, CORE SPRAY SYS I OUTBD INJECT VALVE.

2. VERIFY CLOSED 1-FCV-75-22, CORE SPRAY SYS I TEST VALVE.

3. VERIFY CS Pump 1A and/or 1C running.

4. WHEN RPV pressure is below 450 psig, THEN ThROTTLE 1-FCV-75-25, CORE SPRAY SYS I 1NBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

Core Spray System II Appendix 6E

1. VERIFY OPEN the following valves:

• l-FCV-75-30, CORE SPRAY PUMP lB SUPPR POOL SUCT VLV

• 1-FCV-75..39, CORE SPRAY PUMP 1D SUPPR POOL SUCT VLV

• 1-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.

2. VERIFY CLOSED 1-FCV-75-50, CORE SPRAY SYS II TEST VALVE

3. VERIFY CS Pump lB and/or 1D running.

4. WHEN RPV pressure is below 450 psig, THEN ThROTTLE 1-.FCV-75-53, CORE SPRAY SYS II ]NBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

3-B Page 46 of 65 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Restore and maintain level +2 to +51 inches LAW Appendix 6A, 6D, 6E, or 6C LPCI Appendix 6C

1. IF Adequate core cooling is assured AND It becomes necessary to bypass LPCI Injection Valve auto open signal to control injection, THEN PLACE 1 -HS-74.-1 55B, LPCI SYS II OUTBD 1NJ VLV BYPASS SEL in BYPASS.

2. VERIFY OPEN the following valves:

• 1-FCV-74-24, RHR PUMP lB SUPPR POOL SUCT VLV.

• l-FCV-74-35, RHR PUMP 1D SUPPR POOL SUCT VLV.

3. VERIFY CLOSED the following valves:

• l-FCV-74-75, RI{R SYS II DW SPRAY INBD VLV

• 1-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV

• 1-FCV-74-71, RHR SYS II SUPPR CHBR/POOL ISOL VLV

• 1-FCV-74-72, RI{R SYS II SUPPR CHBR SPRAY VALVE

• 1-FCV-74-73, RHR SYS II SIJPPR POOL CLG/TEST VLV.

4. VERIFY RHR Pump lB and/or 1D running.

5. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 1-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.

6. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 1-FCV-68-3, RECII{C PUMP 1A DISCHARGE VALVE.

7. ThROTTLE 1-FCV-74-66, RHR SYS II LPCI OUTBD iNJECT VALVE, as necessary to control injection.

8. MONITOR RHR Pump NPSH using Attachment 1.

9. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers_discharging to_the_RPV.

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

• l-FCV-23-46, RHR HX lB RHRSW OUTLET VLV

• l-FCV-23-52, RHR HX 1D RHRSW OUTLET VLV.

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

All Control Rods are inserted.

Emergency Depressurization is complete Reactor Level is restored and maintained

SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:

RCW Pump 3A is out of service and tagged out.

3-PI-3-207 Bypassed for surveillance.

Operations/Maintenance 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.

Once completed raise power with flow to 95% JAW 3-GOI-100-12 section 5.0 step 21 and the Reactivity Control Plan.

Units 1 is in a forced outage and Unit 2 is at 100% power.

Unusual Conditions/Problem Areas:

None

w*

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P L

F, H a

n

.1 N

I

m 3.6 CONTAINMENT SYSTEMS 361.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.61.3 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, uPrimary Containment Isolation instrumentation ,

ACTIONS NOTES.----

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

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

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

4. Enter applicable Conditions and Required Actions of LCO 3.6.11, Primary Containment when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME

---

NOTE------ Al Isolate the affected 4 hOurs except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCJVs. and de-activated AND automatic valve, closed manual valve, blind B hours for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured MSIV leakage not within limits.

AND (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A. 2 ------------NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is for isolation isolated. devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was do-mailed while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (rrrifinr ii1

ACTI ONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTE- B.1 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic valve, closed manual valve, or blind One or more penetration flange.

flow paths with two PC [Vs inoperable except due to MSIV leakage not within limits.

C. NOTE---------- C.i Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVS) with only one PC IV. and de-activated automatic valve, closed AND manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for flow paths with one PCIV EFCVs inoperable. AND C.2 NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means.

Verity the affected Once per 31 days penetration flow path is isolated.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration D I Restore leakage rate to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flow paths with MSIV within limit.

leakage not within limits.

E. Required Action and E. I Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, C, AND orDnot metin MODEl, 2, or 3 E.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F. Required Action and F. I Initiate action to suspend Immediately associated Completion operations with a Time of Condition A, B, C, potential for draining the or D not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during PR MODE 4 or 5.

F.2 NOTE Only applicable for inoperable RHR Shutdown Cooling Valves.

Initiate action to restore Immediately valve(s) to OPERABLE status.

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.51 ECCS Operating LCO 3,5.1 Each ECCS injectionlspray subsystem and the Automatic Depressurization System (ADS) (unction of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS LCO 3.O.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A. 1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable. subsystem(s) to OPERABLE status.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

(continued)

ECCS Operahng 351 ACTIONS continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.i Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Comptetion Time of Condition A not AND met B2 Be n MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. HPCI System inoperable. C. I Verify by administrative Immediately means RCIC System is OPERABLE AND Cl Restore HPCI System to 14 days OPERABLE statu&

D. HPCI System inoperable.. Di Restore HPCI System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

AND OR Condition A entered.

Dl Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injectionlspray subsystem to OPERABLE status.

E. One ADS valve El Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One ADS valve F. 1 Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

AND Condition A entered Fl Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS i*njectionlspray subsystem to OPERABLE status.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COW PLETION TIME C. Two or more ADS valves Ci Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND OR G2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action -i dome pressure to associated Completion 150 psig.

Time of Condition C, D, E, or F not met.

H. Two or more low pressure Hi Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO 3.62.3 Four RHR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression Al Restore The RHR 30 days pool cooling subsystem suppression pool cooling inoperable, subsystem to OPERABLE status.

B. Two RI-fR suppressIon 8.1 Restore one RHR 7 days pool cooling subsystems suppression pool cooling inoperable. subsystem to OPERABLE status.

C. Three or more RHR C.1 Restore required RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> suppression pool cooling suppression pool cooling subsystems inoperable, subsystems to OPERABLE status.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and Dl Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met, AND 0.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

RHR Suppression Pool Spray 3.62A

&6 CONTAINMENT SYSTEMS 362,4 Residual Heat Removal (RHR) Suppression Pool Spray LCO 3,62A Four RHR suppression pool spray subsystems shall be OPERABLE, APPUCABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression Al Restore the RHR 30 days pool spray subsystem suppression pool spray inoperable, subsystem to OPERABLE status.

B. Two RHR suppression 61 Restore one RHR 7 days pool spray subsystems suppression pool spray inoperable, subsystem to OPERABLE status.

C. Three or more RHR Cl Restore required RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> suppression pool spray suppression pool spray subsystems mnoperable. subsystems to OPERABLE status.

D. Required Action and Dl Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not mAt AND

RHR Drywall Spray 3.6,2.5 3.6 CONTAINMENT SYSTEMS 3.6.2,5 Resklual Heat Removal (RHR) Drywall Spray LCO 3.6.2.5 Four RHR drywell spray subsystems shaH be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETiON TIME A. One RHR drywall spray Al Restore the RHR drywall 30 days subsystem inoperable, spray subsystem to OPERABLE status.

B. Two RHR drywall spray 8.1 Restore one RHR drywall 7 days subsystems inoperable, spray subsystem to OPERABLE status.

C. Three or more RHR Cl Restore required RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> drywall spray subsystems drywell spray subsystems inoperable, to OPERABLE status.

D. Required Action and Dl Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

36 CONTAINMENT SYSTEMS 3.&2.6 Drywell-toSuppression Chamber Differential Pressure LCO 3.62.6 The dTywell pressure shall be maintained 11 psid above the pressure of the suppression chamber.

This differential may be decreased to < 1.1 pskt for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during required operability testing of the HPCI system, the RCIC system or the suppression chamber4o-drywell vacuum breakers.

APPLICABILITY: MODE I during the time period:

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is> 15% RTP following startup, to

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DrywelI-tosuppression Al Restore differential B hours chamber differential pressure to within limit.

pressure not within limit.

B. Required Action and 81 Reduce THERMAL 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion POWER to 15% RTP, Time not met.

I I I I I I I I C

z C

0 C

m m

z

-I 2,iAI ITABLEI I I I I Drywell pressure at or above 2,45 psig AND Indication of Primary System leakage into Primary Coitainment. Refer to Table 2.1-A.

OPERATING CONDITION:

Model cr2 cr3 2i-S ICURVEI I I 2.24 1 I I Suppression Chamber pressure can NOT be Oryweli or Suppression Chamber maintained in the safe area of Curve 2.1-S. hydrogen concentration at or above 4%

m AND Dpjwell or Suppression Chamber oxygen concentration at or above 5%,

OPERATING CONDITION: OPERATING CONDITION:

Modelor2or3 Mode lor2or3 - -< -

2141 I I I 22-GI I 1 Suppression Chamber pressure can NOT be Drywell or Suppression Chamber maintained below 55 psig. hydrogen concentration at or above 6%

AND DrwetI or Suppression Chamber oxygen concentration at or above 5%. rn m

rn OPERATING CONDITION: OPERATING CONDITION:

Mode I cr2 cr3 Model cr2 cr3

Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: C Op-Test No.: ILT 1102 FINAL SRO:

Examiners: Operators: ATC:

BOP:

Initial IC 192/ Unit 3 Reactor Power 86% / HPCI tagged out for PMs. Stator Water Cooling Conditions: Pump 3B tagged out.

Turnover: BOP Operator Perform 3-01-3 Section 8.13 Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump. Perform Control Rod Pattern adjust JAW RCP.

Event Event No. Maif. No. Type* Event Description 1 8.13 Automatic Start Test of RFPT 3A Oil Pumps, 3-01-3 R-ATC 2 Perform Control Rod Pattern adjust lAW RCP R-SRO C-ATC Final(4th) Control Rod (3 8-23) manipulated continues to move 3 3 rdO4r3 823 TS-SRO notches beyond intended position 4 rcO9 RCIC Room high temp / Fail to Isolate Loss of FW Heating 5 fwO5b C-ALL 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate Feedwater Line Break in Turbine Bldg / Drywell leak 6 fwl8 M-ALL Div 1 ECCS fails to initiate 7 edl2b C 480V RMOV Board 3B Supply Breaker Trip 8 csO4a I Loop I Core Spray Logic Power Failure

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

Appendix B Scenario Outline Form ES-D-1 Critical Tasks Four CT#1-With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, initiate Emergency Depressurization before RPV level lowers to -

180 inches.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Water level trend.

3. Measured by:

Observation At least 6 SRVs must be opened before RPV level lowers to -180 inches.

4. Feedback:

RPV pressure trend.

SRV status indications.

CT#2-With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

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

3. Measured by:

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

4. Feedback:

Reactor water level trend.

Reactor pressure trend.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks Four CT#3-With a primary system discharging into the secondary containment, take action to manually isolate the break.

1. Safety Significance:

Isolating high energy sources can preclude failure of secondary containment and subsequent radiation release to the public.

2. Cues:

Procedural compliance.

Area temperature indication.

3. Measured by:

With the reactor at pressure and a primary system discharging into the secondary containment, operator takes action to manually isolate the break.

4. Feedback:

Valve position indication CT#4-To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B INHIBITED annunciator status.

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

The Plant is operating at 86% Reactor Power.

The BOP Operator will perform Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump, 3-01-3 Section 8.13 The ATC will adjust the Control Rod Pattern JAW RCP. When the 4 th control rod is withdrawn, it will continue to move 3 notches beyond its intended positions. The ATC will completely insert the Control Rod JAW 3-AOI-85-6 or 3-AOI-85-7. Accumulator must be declared mop if charging water is isolated. The SRO may declare the Control Rod Inoperable Technical Specification 3.1.3 condition C.

A RCIC Steam Line Break will result in high Room temperature with a failure of RCIC to Isolate.

The BOP will isolate RCIC. The SRO will determine RCIC Isolation Valves inoperable and RCIC System inoperable. With HPCI already Inoperable, plant shutdown is required. Technical Specification 3.5.3 Condition B and 3.6.1.3 Condition A.

A tube leak on High Pressure Feedwater Heater B2 results in isolation of Extraction Steam to the heater. The crew will respond in accordance with 3 -AOI 1 A or 1 C. The ATC will lower reactor power by 5%. The Operators refer to 3-AOI 1 A or 1 C and determine that all automatic actions failed to occur and the Operators isolate the Heater B2.

A Feedwater line break will occur in the Turbine Building. The Loss of Feedwater Flow 3-AOI-3-1 should be entered and a manual Scram inserted. EOI- 1 will be entered on Reactor Level.

EOI-2 will be entered on High Drywell Pressure / Temperature. Actions of EOI-2 will be directed.

SRO will enter C-i on lowering Reactor Level. CRD should be maximized and SLC should be initiated as Reactor Level continues to lower.

Reactor level will decrease to TAF and an Emergency Depressurization will be initiated per C-2.

Div 1 ECCS will fail to auto initiate and will have to be manually initiated.

Level will be restored with Low Pressure ECCS.

The Emergency Classification is 1.1-Si Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

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

Appendix B Scenario Outline Form ES-B-i SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-C 7 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 3 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 70 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix D Scenario Outline Form ES-D-1 Scenario Tasks EVENT TASK NUMBER KJA RO SRO 1 Automatic Start Test of RFPT 3A Oil Pumps RO U-003-NO-30 25900 1K4.06 2.5 2.6 2 Control Rod Pattern Adjustment RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 3 Control Rod Mispositioned or Drift RO U-085-AB-07 295014AA1.03 3.5 3.5 SRO S-085-AB-07 4 RCIC Steam Leak RO U-071-AL-19 295032EA1.05 3.7 3.9 SRO S-000-EM-12 5 Loss of Feedwater Heating RO U-006-AB-01 2.1.43 4.1 4.3 SRO 5-006-AB-Ol 6 Feedwater Line Break RO U-000-EM-18 29503 1EA2.04 4.6 4.8 SRO S-000-EM-19 SRO T-000-EM-15

3-C Page 7 of 58 Procedures Used/Referenced:

Procedure Number Procedure Title Procedure Revision 3-01-3 Reactor Feedwater System Revision 82 3-GOI-1 00-12 Power Maneuvering Revision 35 3-01-85 Control Rod Drive System Revision 70 3-ARP-9-5A Alarm Response Procedure Panel 3-9-5A Revision 41 3-A0I-85-6 Rod Drift Out Revision 9 3-AOI-85-7 Mispositioned Control Rod Revision 5 TS 3.1.3 Control Rod Operability Amendment 212 3-ARP-9-3A Alarm Response Procedure Panel 3-9-3A Revision 43 3-ARP-9-3D Alarm Response Procedure Panel 3-9-3D Revision 28 3-EOI-3 Secondary Containment Control Flowchart Revision 9 Restoring Refuel Zone and Reactor Zone Ventilation Fans 3.EOPAPPENDD(8F Revision 2 Following Group 6 Isolation TS 3.5.3 RCIC System Amendment 244 TS 3.6.1.3 Primary Containment Isolation Valves Amendment 212 3-ARP-9-6A Alarm Response Procedure Panel 3-9-6A Revision 20 High Pressure Feedwater Heater String/Extraction Steam 3-AOI-6-1A . Revision 18 Isolation High and Low Pressure Feedwater Heater String/Extraction 3-AOI-6-1C . Revision 15 Steam_Isolation 3-01-6 Feedwater Heating and Misc Drains System Revision 67 3-ARP-9-5A Alarm Response Procedure Panel 3-9-5A Revision 41 3-ARP-9-6C Alarm Response Procedure Panel 3-9-6C Revision 21 3-EOI-1 RPV Control Flowchart Revision 8 3-EOI-APPENDIX-5B Injection System Lineup CRD Revision 1 3-EOI-APPENDIX-7B Alternate RPV Injection System Lineup SLC System Revision 2 3-EOI-3-C-l Alternate Level Control Flowchart Revision 9 3-EOI-3-C-2 Emergency RPV Depressurization Flowchart Revision 8 3-EOI-APPENDJX-6B Injection Subsystems Lineup RHR. System I LPCI Mode Revision 3

3-C Page 8 of 58 Procedures Used/Referenced Continued:

Procedure Number } Procedure Title Procedure Revision 3-EOI-APPENDIX-6D Injection Subsystems Lineup Core Spray System I Revision 3 3-EOI-2 Primary Containment Control Flowchart Revision 7 Emergency Classification Procedure Event Classification EPll-l Revision 46 Matrix EPJP-4 Site Area Emergency Revision 32 3-AOI-100-l Reactor Scram Revision 53

3-C Page 9 of 58 Console Operator Instructions Scenario File Summary File: batch and trigger files for scenario 3-C Batch nrc2Ollc

1. hpci tagout bat nrc2Ol lhpcito Batch nrc2Ol lhpcito ior zdihs732 close ior zdihs733a close ior zdihs738la close ior zlohs7347a[1] off ior ypovfcv732 (none 30) fail_now ior ypovfcv733 (none 30) fail_now ior ypovfcv738l (none 30) fail_now

1. stator water pump b tagout ior zlohs3536a[1] off ior zlohs3536a[2J off

1. CR Drift imfrdO4r3823 (el 0)

1. RCIC leak fail to isolate imfrc09 (e5 0)100 120 imf rc 10

1. Loss of Feedwater Heating imffw05b (elO 0)10030075 ior ypovfcv052l fail_power_now ior zlohs052la{2j on trg 11 nrc20110521 trg 11 =batnrc20llcl Trigger nrc2Ol 10521 zdihs052 1 a[ 1] .eq. 1 Batch nrc2Ollcl dor ypovfcv052l dor zlohsO52la[2]

3-C Page 10 of 58

1. Major imffwl8 (e20 0) 50 300 imfth2l (e25 30) .1 360 imf cs04a imfedl2b (e20 300) ior xa553c[27j alarm off ior xa553c[14] alarm off ior zloil756la[l] off ior zloil756lb[l] off trg2l nrc20117525 trg2l =batnrc20llc2 Trigger nrc2Ol 17525 zdihs7525a[3] .eq. 1 Batch nrc2Ollc2 dmf cs04a

3-C Page 11 of58 Console Operator Instructions Scenario 3-C DESCRIPTION/ACTION Simulator Setup manual Reset to IC 192 Simulator Setup Load Batch Bat nrc2Ol ic Simulator Setup manual Clearance out HPCI Simulator Setup Clearance out Stator Water Cooling manual Pump3B Simulator Setup Verify batch file loaded RCP required (86% Power with Control Rod Pattern Adjust) Provide marked up copy of 3-GOI-100-12 and RCP for Urgent Load Reduction.

3-C Page 12 of58 Simulator Event Guide:

Event 1 Normal: Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump per 3-01-3 Section 8.13 SRO Direct BOP to perform Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump per 3-01-3 Section 8.13 BOP 8.13 Automatic Start Test of RFPT 3A Oil Pumps

[1] OBTAIN Unit Supervisor approval to perform this test.

[2] VERIFY the following switches in Normal after START or STOP:

• RFPT 3A 3A1 MAiN OIL PUMP, 3-HS-3-103A

• RFPT 3A 3A2 MAiN OIL PUMP, 3-HS-3-250A

[3] VERIFY RFPT 3A EBOP 3A3, 3-HS-3-102A, in AUTO.

[4] TEST EBOP 3A3 as follows:

[4.1] DEPRESS and HOLD 3A3 EBOP TEST push-button, 3-HS-3-1 05A.

[4.2] VERIFY the following:

• Red (running) light and amber (auto start) light at push-button illuminated.

• RFPT OIL PUMP AUTO START annunciation, 3-XA-55-6B Window 29, in alarm.

[4.3] RELEASE 3A3 EBOP TEST push-button, 3-HS-3-105A.

[4.4] PLACE RFPT 3A EBOP 3A3 switch, 3-HS-3-102A, in START (return to AUTO).

[4.4.1] VERIFY the following:

• Amber (auto start) light extinguished at 3A3 EBOP TEST push-button, 3-FIS-3-105A.

• RFPT OIL PUMP AUTO START annunciation, 3-XA-55-6B Window 29, will reset.

[4.5] PLACE RFPT 3A EBOP 3A3, 3-HS-3-102A, in STOP (return to AUTO).

• CHECK Red light extinguished at 3A3 EBOP TEST push-button.

BOP Perform 3-01-3 section 8.13 steps 1-4 to Test Automatic Start of RFPT 3A EBOP 3A3 Oil Pump

3-C Page 13 of58 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP SRO Notifr ODS of Power Increase Direct Power Increase after Control Rod Pattern Adjustment per 3-GOI-100-12 section 5.0 step 21 5.0 INSTRUCTION STEPS

[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.

• MONITOR Core thermal limits using Illustration 1, ICS, and/or 0-TI-248 ATC Raise Power with Control Rods per 3-01-85, section 6.6. Control Rods to be withdrawn:

22-23, 22-39, 38-39and38-23 start at 00 and goto 10..

ATC 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

[2] VERIFY the following prior to control rod movement:

• CRD POWER, 3-HS-85-46 in ON.

• Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

• Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.

• Nuclear Instrumentation responds as control rods move through the core (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize automatic RBM Rod Block as follows:

• STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[61.

3-C Page 14 of 58 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP ATC 6.6.2 Actions Required During and Following Control Rod Withdrawal (contd)

[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:

[6.1] PLACE CRD POWER, 3-HS-85-46 in the OFF position to deselect the Control Rod.

[6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position.

6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 3-XS-85-40.

[2] OBSERVE the following for the selected control rod:

• CRD ROD SELECT pushbutton is brightly ILLUMINATED

• White light on the Full Core Display ILLUMINATED

• Rod Out Permit light ILLUMiNATED

[4] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE.

[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.

3-C Page 15 of58 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP ATC 6.6.5 Return to Normal After Completion of Control Rod Withdrawal

[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:

[1.1] PLACE CRD POWER, 3-HS-85-46, in OFF.

[1.2] PLACE CRD POWER, 3-HS-85-46, in ON.

bWI WhenATC withdraws hc Final (4) rod. (38-23)Jnsert triggerI,Rod will contixzue td pov3 o&hes beyond intndedposition. After Control Rod 8-23 reachespotion 14 da1etenia1ncionidO43823 from the malfunctioarnenw Vt od to go a1east 1

notches past iuteiided position of 1O, eg position l6.:

3-C Page 16 of58 Simulator Event Guide:

Event 3 Component: th Final(4 Control Rod manipulated continues to move 3 notches beyond

)

intended position DRiVE When ATC withdraws the Final (4th) rod (3823) insert trigger 1, Rod will continue to move 3 Notebes beyond intended position. After Control Rod 38-23 reaches position 14 delete malfimction rd04r3823 from the mii1inction menu. Waxit rod to gç at least 3 notches past ii4ended pos1tion of 10, egposition 16.

ATC Reports CONTROL ROD DRIFT alarm and Control Rod 38-23 has drifted out 3 notches from intended position SRO Directs ATC to respond per ARP and 3-AOI-85-6 and/or 3-AOI-85-7 ATC 3-ARP-9-5A window 28 CONTROL ROD DRIFT A. DETERMINE which rod is drifting from Full Core Display.

B. IF no control rod motion is observed, THEN RESET rod drift as follows:

1. PLACE ROD DRIFT ALARM TEST switch, 3-HS-85-3A-S7, in RESET and RELEASE.

2. RESET the annunciator.

C. IF rod drifting in, THEN REFER TO 3-AOI-85-5 and 3-AOI-85-7 D. IF rod drifting out, THEN REFER TO 3-AOI-85-6 and 3-AOI-85-7.

E. REFER TO Tech Spec Section 3.1.3, 3.10.8.

ATC Resets the CONTROL ROD DRIFT alarm when rod motion has stopped by placing the ROD DRIFT ALARM TEST switch, 3-HS-85-3A-S7, in RESET and RELEASE.

Then resets the annunciator Responds per 3-AOI-85-6 and/or 3-AOI-85-7 Monitors Full Core Display for a second Control Rod Drift as per Immediate Actions of 3-AOI-85-6 j NOTE: If crew identifies Control Rod 38-23 as a driltthisis the coirectAOI. if the crew identifiçs ControIRod 3823 as Mispositioned then referto Page 19 ATC 3-AOI-85-6 Control Rod Drift 4.1 Immediate Actions

[1] IF multiple control rod drifts are identified, THEN MANUALLY SCRAM the reactor and enter 3-AOI-100-1.

3-C Page 17 of58 Simulator Event Guide:

Event 3 Component: Final(4t) Control Rod manipulated continues to move 3 notches beyond intended position ATC 3-AOI-85-6 Control Rod Drift (continued) 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 TN (00) position.

[2] IF control rod drive does NOT respond to iNSERT signal, THEN

[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 3-AOI-100-1.

[5] IF the control rod will not latch into position 00 and continues to demonstrate occurrences of inadvertent withdrawal, THEN

[6] IF the control rod is latched into position 00, THEN REMOVE associated HCU from service per 3-01-85.

[7] EVALUATE Tech Spec 3.1.3.

[8] INITIATE Service Request/Work Order.

3-C Page 18 of58 Simulator Event Guide:

Event 3 Component: Final(4th) Control Rod manipulated continues to move 3 notches beyond intended position ATC 3-AOI-85-6 Control Rod Drift (continued) 4.2 Subsequent Actions(continued)

[9] NOTIFY Reactor Engineer to perform the following for current condition:

• EVALUATE condition of core to assure no resultant fuel damage has occurred.

• EVALUATION of impact on thermal limits and PCIOMOR restraints. (N/A if scram was initiated.)

• DETERMINE if other control rods need to be repositioned in order to safely restore core symmetry to prevent local fuel damage. (N/A if scram was initiated.)

[10] NOTIFY System Engineering to PERFORM O-TI-.20, Control Rod Drive System Testing and Troubleshooting to determine problem with faulty control rod.

[11] IF a manual scram was not inserted and Reactor Startup or Shutdown is not in progress, THEN

[12] WHEN control rod fault has been corrected, THEN

[13] NOTIFY Reactor Engineer to EVALUATE impact on preconditioning envelope, prior to returning to normal power operation.

ATC Selects Control Rod 3 8-23 and inserts to position 00 Notifies the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

Removes the associated HCU from service per 3-01-85

3-C Page 19 of 58 Simulator Event Guide:

Event 3 Component: th Final(4 Control Rod manipulated continues to move 3 notches beyond

)

intended position AsJ{eactor Engineer inform that Core ThermalE sanjI Preconditioning imfts fç the current Cono1 Rod pattern will bç evlu$e.

SRO Evaluates Tech Spec 3.1.3 Condition C Initiates Work Order/Service Request Notifies Reactor Engineer to perform the following for current condition:

• Evaluation of condition of core to assure no resultant fuel damage has occurred.

• Evaluation of impact on thermal limits and PCIOMOR restraints.

• Determination if other control rods need to be repositioned in order to safely restore core symmetry to prevent local fuel damage.

Notifies System Engineering to perform 0-TI-20, Control Rod Drive System Testing and Troubleshooting to determine problem with faulty control rod.

Enters 3-GOI-100-12, Power Maneuvering, for the power change that occurred.

Directs associated HCU removed from service per 3-01-85 i)RPER If contacted, as Reactor Engineer infonnthat all conditions listed above will bó eva1uated

- If contacted, as Work Control inform that you. will get workmg on a Work Order/SR If contacted, as System Engineering inform that you will perform 0-TI-20.

SRO The SRO may direct entry into 3-AOI-85-7, Mispositioned Control Rod, if so the following procedure will be used.

ATC 3-AOI-85-7 Mispositioned Control Rod 4.1 Immediate Actions None 4.2 Subsequent Actions

[ 1] STOP all intentional control rod movement.

[2] IF Control Rod is determined to be mispositioned, THEN NOTIFY the following:

• Reactor Engineer (RE),

• Shift Technical Advisor (STA),

• Unit Supervisor

• Shift Manager (SM)

• Operations Superintendent. [1ISTPO SOER 84-002]

3-C Page 20 of 58 Simulator Event Guide:

Event 3 Component: Final(4t) Control Rod manipulated continues to move 3 notches beyond intended position ATC 4.2 Subsequent Actions (continued)

[3] IF the Control Rod is > 2 notches from the intended position, THEN PERFORM the following: (Otherwise N/A)

[3.1] INSERT the mispositioned rod to 00.

[3.2] IF a Reactor Startup or Shutdown is not in progress, THEN (Otherwise N/A)

[4] IF the Control Rod is less than or equal to 2 notches from the intended position, THEN (Otherwise N/A)

[5] CHECK the following radiation recorders for a rise in activity to determine if any fuel damage occurred:

• MAIN STEAM LINE RADIATION, 3-RR-90-l35 (Panel 3-9-2)

• OFFGAS RADIATION, 3-RR-90-266, on Panel 3-9-2.

• OFFGAS RADIATION, 3-RR-90-160 (Panel 3-9-2)

• OFFGAS PRETREATMENT RADIATION, 3-RR 157 (Panel 3-9-2)

[6] IF there is any evidence of fuel damage, THEN

[7] INTIATE a Service Request/PER for Control Rod error or mispositioned Control Rod.

[8] IF possible, THEN DETERMINE how long the rod has been mispositioned

[9] NOTIFY Reactor Engineer to perform the following when time permits:

• EVALUATE the possible consequences

• DOCUMENT in Reactor Engineer log.

3-C Page 21 of58 Simulator Event Guide:

Event 3 Component: Final(4th) Control Rod manipulated continues to move 3 notches beyond intended position SRO Directs ATC to stop all intentional Control Rod Movement Informs all positions listed in step 2 of Subsequent Actions of Mispositioned Control Rod Directs ATC to Insert Mispositioned Control Rod to 00 Enters 3-GOI-100-12, Power Maneuvering Initiates Service Request and Notifies Reactor Engineer to evaluate the possible consequences and document in the Reactor Engineering Log bthvi The SkçiwiTRirect tfie assQciatedHCUr&novdfronseryice iA3 ii enter AcknowJedge order to remove HCUfromservice, ez wbat stepsin3-OJ-85 will betsed to isgiate theICtL Wait 20 minutes then insert mallimcti9nr4O8 t4gçumii1atoj ervioq DRJVEl: ty4RCiifiggrRcste withai a*isoie SRO Evaluate Tech Spec 3. 1.3 Condition C One or more control rods inoperable for reasons other than Condition A or B Required Action C. 1 Fully Insert inoperable control rod Completion Time 3 Hours AND Required Action C.2 Disarm the associated CRD Completion Time 4 Hours

3-C Page 22 of 58 Simulator Event Guide:

Event 3 Component: Final(4tl) Control Rod manipulated continues to move 3 notches beyond intended position ATC Stops all intentional control rod movement When directed inserts Control Rod to Position 00 Evaluates Radiation Recorders to determine if Fuel Damage Exists and determines how long rod has been mispositioned.

Qiivi cntactes Wttr if ioiwwrkiig on Or&r/Svice Request.

If coace4 &Engm foi &illi eiiate all on1Itions hsl,o DRIVER Wlie ct1iyNRC msertTngg5 forRCjqsteam 1ak with failure t<to i1te

3-C Page 23 of 58 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate (Leak downstream of Isolation valves)

DRIVER WhiireciIWNkC Se rgger fpr RC anek with failie toso1e BOP Respond to Annunciator RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 3-9-11. (Alarm on Panel 3-9-11 will automatically reset if radiation level lowers below setpoint.)

C. NOTIFY RADCON.

D. IF the TSC is NOT manned and a VALID radiological condition exists., THEN USE public address system to evacuate area where high airborne conditions exist.

BOP Determine RCIC Area Radiation Monitor is in Alarm and report, Evacuate affected area and notify radiation protection.

BOP Respond to annunciator RCIC STEAM LINE LEAK DETECTION TEMP HIGH If temperature continues to rise it will cause isolation of the following valves at steam line space temperature of 165°F Torus Area or 165°F RCIC Pump Room.

• RCIC STEAM LINE INBD ISOLATION VLV, 3-FCV-71-2

• RCIC STEAM LINE OUTBD ISOLATION VLV, 3-FCV-71-3 A. CHECK RCIC temperature switches on LEAK DETECTION SYSTEM TEMPERATURE indicator, 3-TI-69-29 on Panel 3-9-21.

B. IF RCIC is NOT in service AND 3-FI-71-1A(B), RCIC STEAM FLOW indicates flow, THEN ISOLATE RCIC and VERIFY temperatures lowering.

C. IF high temperature is confirmed, THEN ENTER 3-EOI-3 Flowchart.

D. CHECK CS/RCIC ROOM El 519 RX BLDG radiation indicator, 3-RI-90-26A on Panel 3-9-11 and NOTIFY RADCON if rising radiation levels are observed.

E. DISPATCH personnel to investigate.

3-C Page 24 of 58 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate (Leak downstream of Isolation valves)

BOP Reports rising temperature in RCIC, reports RCIC failed to isolate and isolates RCIC Steam Line SRO Enter EOI-3 on Secondary Containment Area Radiation Dk1VE1 SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. Then verify isolation of Reactor Zone or Refuel Zone and verify SGTS initiates If above 72 mr/hr direct Operator to verify isolation of ventilation system and SGTS initiated ATC/BOP Verifies Reactor Zone and Refuel Zone Ventilation Systems isolated and SGTS initiated SRO If Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels are below 72 mr/hr Then Restart Reactor Zone and Refuel Zone Ventilation per Appendix 8F If ventilation isolated and below 72 mr/hr directs Operator to perform Appendix 8F SRO Enters EOI-3 on High Secondary Containment Temperature Secondary Containment Temperature Monitor and Control Secondary Containment Temperature Operate available ventilation per Appendix 8F Is Any Area Temp Above Max Normal YES -

To1te all ystem that arc dischargng nto the area CT#3 to

• B opeedbBOI OR

• Suppresai?ie Cf#3 ttenLLines and ptTeiiiperat&e na jdtzon Lyenn SRO Evaluates Technical Specification 3.6.1.3 Condition B Condition B One or more penetration flow paths with two PCIVs inoperable except due to MS1V leakage not within limits.

Required Action B. 1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

Completion Time 1 Hour

3-C Page 25 of 58 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate (Leak downstream of Isolation valves)

SRO Evaluates Technical Specification 3.5.3 Condition A Condition A RCIC System Operable Required Action A. 1 Verif by administrative means that HPCI is operable Completion Time Immediately AND Required Action A.2 Restore RCIC system to operable status Completion Time 14 Days Evaluate Technical Specification 3.5.3 Condition B Condition B Required Action and associated completion time not met Required Action B.1 Be in Mode 3 Completion Time 12 Hours AND Required Action B .2 Reduce Reactor Steam Dome Pressure to < or equal to 150 PSIG Completion Time 36 Hours SRO Enters EOI-3 on High Secondary Containment Temperature (continued)

Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels Is Any Area Radiation Level Max Normal NO -

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

• Be operated by EOIs OR

• Suppress a Fire Ensures no systems are still discharging to Secondary Containment, remains in EOI-3 until entry conditions are cleared.

SRO Enters EOI-3 on High Secondary Containment Temperature (continued)

Secondary Containment Level Monitor and Control Secondary Containment Water Levels Is Any Floor Drain Sump Above 66 inches NO -

AND Is Any Area Water Level Above 2 inches NO -

3-C Page 26 of 58 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate (Leak downstream of Isolation valves)

ATC/BOP 3-EOI Appendix 8F

1. VERIFY PCIS Reset.

2. PLACE Refuel Zone Ventilation in service as follows (Panel 3-9-25):

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

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

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

d. VERIFY OPEN the following dampers:

• 3-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR

• 3-FCO-64-6, REFUEL ZONE SPLY INED ISOL DMPR

• 3-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR

• 3-FCO-64-1O, REFUEL ZONE EXH INBD ISOL DMPR.

3. PLACE Reactor Zone Ventilation in service as follows (Panel 3-9-25):

a. VERIFY 3-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.

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

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

d. VERIFY OPEN the following dampers:

• 3-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR

• 3-FCO-64-14, REACTOR ZONE SPLY INBD ISOL DMPR

• 3-FCO-64-42, REACTOR ZONE EXH INBD ISOL DMPR

• 3-FCO-64-43, REACTOR ZONE EXH OUTBD ISOL DMPR.

PYi: WI en directed by NRC insert Trigger 10 for LssofFeedaterHeafingCV-5-21,

çRIç)LyJdy, 2JJ1

3-C Page 27 of 58 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate PRJER Wien ctg 10 for L3FGV-Iil NEAT JSQJ YY Fail tqisate1 ATC/BOP Announces BYPASS VALVE TO CONDENSER NOT CLOSED and refers to 3-ARP-9-6A, window 18.

A. CHECK heater high or low level or moisture separator high or low level alarm window illuminated on Panel 3-9-6 or 3-9-7 to identify which bypass valve is opening.

B. CHECK ICS to determine which bypass valve is open.

C. DISPATCH personnel to check which valves light is extinguished on junction box 34-21, Col T-13 J-LINE, elevation 565.

Aelow1edgedispatch wajt 1-2 miutes lights put onji4iction j,ox 34-21 ATC/BOP Announces HEATER B2 LEVEL HIGH and refers to 3-ARP-9-6A window 9.

A. CHECK the following indications:

• Condensate flow recorder 2-29, Panel 3-9-6. Rising flow is a possible indication of a tube leak.

• Heater B2 shell pressure, 3-PI-5-22 and drain cooler B5 flow, 3-FI-6-34, Panel 3-9-6. High or rising shell pressure or drain cooler flow is possible indication of a tube leak.

B. CHECK drain valve 3-FCV-6-95 open.

C. CHECK level on ICS screen, FEEDWATER HEATER LEVEL (FWHL).

• IF the 3B2 heater indicates HIGH (Yellow), THEN VERIFY proper operation of the Drain and Dump Valves.

• DISPATCH personnel to local Panel 3-LPNL-925-562C to VERIFY and MANUALLY control the level.

D. IF a valid HIGH HIGH level is received, THEN GO TO 3-AOI-6-1A or 3-AOI-6-1C.

ATC/BOP Checks condensate flow recorder, Heater B2 shell pressure and Drain Cooler B5 flow for indications of a tube leak Checks drain valve 3-FCV-6-95 open Checks 3B2 Heater level on ICS and dispatches personnel to verify and manually control level

$cno4edge or4ta ye axdinantiafly eontroUevd on B2at W minutes andrort nab1ctotakemanual contxoIpfB2Heater.

3-C Page 28 of 58 Simulator Event Guide:

EventS Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate ATC/BOP Announces B 1 and B2 High Pressure Heater Extraction Isolation SRO Directs crew to enter 3-AOI-6-1A or 3-AOI-6-1C ATC/BOP 3-AOI-6-1A High Pressure Feedwater Heater String/Extraction Steam Isolation 4.1 Immediate Actions

[1] REDUCE Core Thermal Power to ? 5% below initial power level to maintain thermal margin.

4.2 Subsequent Actions

[1] REFER TO 3-01-6 for turbine/heater load restrictions.

[2] REQUEST Reactor Engineer EVALUATE and ADJUST thermal limits, as required.

[3] ADJUST reactor power and flow as directed by Reactor Engineer/Unit Supervisor to stay within required thermal and feedwater temperature limits. REFER TO 3 -GOl- 100-12 or 3-G0I-100-12A for the power reduction.

[4] ISOLATE heater drain flow from the feedwater heater string that isolated by closing the appropriate FEEDWATER HEATER A-2(B-2) or (C-2) DRAIN TO HTR A-3(B-3) or (C- 3), 3-FCV-6-94(95) or (96).

[5] IF a tube leak is indicated, THEN PERFORM manual actions of Attachment 1 for affected heaters.

[6] VERIFY automatic actions occur. REFER TO Attachment 1.

[7] MONITOR TURB THRUST BEARING TEMPERATURE, 3-TR-47-23, for rises in metal temperature and possible active/passive plate reversal.

[8] DETERMINE cause which required heater isolation and PERFORM necessary corrective action.

3-C Page 29 of 58 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate ATC/BOP 3-AOI-6-1A High Pressure Feedwater Heater String/Extraction Steam Isolation (continued) 4.2 Subsequent Actions (continued)

[9] WHEN the condition which required heater isolation is no longer required, THEN RESTORE affected heater. REFER TO 3-01-6.

ATC Lower Reactor Power greater than 5% below initial power level using Recirc Pump flow adjustments BOP Refers to 3-01-6 for turbine/heater load restrictions Contacts Reactor Engineer to evaluate and adjust Thermal Limits, if needed Isolates heater drain flow B2 Heater Drain to B3 Heater by shutting 3-FCV-6-95 SRO Directs isolating FW to B HP heater string based on indications of tube leak by performing manual actions of Attachment 1 and verifying automatic actions occur 3-AOI-6-1A Attachment 1 Bi or B2 The following valves must be manually closed:

3-FCV-3-31, HP HTR 3B2 FW INLET ISOL VALVE 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VALVE The following valves AUTO Isolate 3-FCV-5-9, HP HEATER 3Bl EXTR ISOL VLV 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV 3-FCV-6-74, MOISTURE SEP LC RES Bi ISOL VLV 3-FCV-6-172, MOISTURE SEP LC RES B2 ISOL VLV Directs power reduction to 920 MWe (79%) power (Power Reduction with RCP flow or Control Rods) per 3-01-6, Illustration 1 3-01-6 Illustration 1 HEATERS OUT (Tube and Shell Side) **

One HP string 920 MWe (79%)

One LP string 920 MWe (79%)

One HP and LP string 920 MWe (79%)

Enters 3-GOI-l00-12, Power Maneuvering Notifies Rx Eng. And ODS of Feedwater Heater isolation and power reduction

3-C Page 30 of 58 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate BOP 3-AOI-6-1A Attachment 1 Closes the following Feedwater Valves Manually 3-FCV-3-31, HP HTR 3B2 FW INLET ISOL VALVE 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VALVE Verifies the following valves close automatically 3-FCV-5-9, HP HEATER 3B1 EXTR ISOL VLV 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV 3-FCV-6-74, MOISTURE SEP LC RES Bi ISOL VLV 3-FCV-6-172, MOISTURE SEP LC RES B2 ISOL VLV Takes action to manually shut 3-FCV-5-21 upon determining the valve did not automatically close and reports to SRO Recognizes HTR level lowers as a result of isolating the Condensate side of 3B HP HTR string (i.e. tube leak) and reports to crew DI tfte.

t& EjCd irc ew to 1low leie OftirgeD pa4reducan-OJ ATC Lower Reactor Power to <920 MWe/<79% power by lowering recirc flow.

SRO Direct ATC to insert the first group of control rods on the Emergency Shove Sheet per Reactor Engineer recommendation.

ATC Inserts the first group of rods on the Emergency Shove Sheet using a peer check as directed by Rx Engineer & Unit Supervisor D% e gg2ØaLine&&in Turlme ig eterI reaches -f 10 to -tirn suiedTngjer 5JçyWe1 1ea1

3-C Page 31 of58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate RWE Whc eedwaçer Line i4nç BI4 Wh],qWater Lev4re cbes1fq-12O iikhesinsert4ggçr 5rywell 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 A. VERIFY Reactor water level hi/low using multiple indications including Average Narrow Range Level on 3-XR-3-53 recorder, 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-LI-3-253 on Panel 3-9-5.

B. IF alarm is valid, THEN REFER TO 3-AOI-3-1 or 3-01-3.

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 Leak in the Turbine Building due to both Feedwater Line Flows lowering to 0 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

3-C Page 32 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate

)RPER When Rctor Water cy! ciIs-ILQicL2P icles insert Trigger 25 Dye1J.

leak SRO Enters EOI-1 on Low Reactor Water Level RCIQ Monitor and Control Reactor Power Directs Exit of EOI-1 RC/Q Leg after ATC reports All Rods In on Scram Report RC/P Monitor and Control RPV Pressure Answers No to is any MSRV cycling Directs BOP to maintain RPV Pressure 800-1000 psig using Bypass Valves RCJL Monitor and Control RPV Water Level Verify as Required

• PCIS Isolations (Groups 1, 2 and 3)

• ECCS

• RCIC Recognizes loss of all High Pressure Injection sources with exception of CRD and SLC. Directs maximizing CRD flow to the Vessel per Appendix 5B Answers No to can water level be Restored and Maintained above +2 inches Maintain RPV Water Level above -162 inches

3-C Page 33 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate J\YB eactW?er e1reach -110 to -12b mthini& fie25 Drdll eak Enters EOI-1 on Low Reactor Water Level (cont)

T#4 Directs i1ibiteiben %l7atereve1 drops bqw42icbe Augments RPV Water Level Control with SLC per Appendix 7B Answers No to can RPV Water Level be maintained above -162 inches Exits RC/L and enters C-i, Alternate Level Control

3-C Page 34 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate ATC Appendix 5B

1. IF Maximum injection flow is NOT required, THEN VERIFY CRD aligned as follows:

a. VERIFY at least one CRD pump in service and aligned to Unit 3 CRD system.

b. ADJUST 3-FIC-85-l 1, CRD SYSTEM FLOW CONTROL, as necessary to obtain flow rate of 65 to 85 gpm.

c. THROTTLE 3-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV, to maintain 250 to 350 psid drive water header pressure differential.

d. EXIT this procedure.

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-l 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-C Page 35 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate jpj Yh tot Wt Ixcac1 tp 420 inches insertTiigger25j)ywj1 ATC Appendix 7B

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

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

11. CHECK SLC injection by observing the following:

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

• Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,

• SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (3-XA-55-5B, Window 20).

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

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

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

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

13. IF SLC tank level drops to 0%,

THEN STOP SLC pumps.

15. MONITOR and CONTROL SLC System as necessary to maintain injection.

3-C Page 36 of 58 Simulator Event Guide:

Event 7 Component: 480V RMOV Board 3B Supply Breaker Trip When Trigger 25, Drywell Leak, is inserted Dryweli Pressure will begin oise an Reactor Water Level vill bego lower at a faster rate.

BOP Approximately 5 minutes after Feedwater Leak inserted recognizes loss of 480v RMOV Board B. Announces loss of Division II ECCS systems Lk Loop II t1(jedlve on(w1tiu> i5iUId inboard ixijectim vthe stiWhaviiig power. Willbe unabfe o throttle flow; whenlO9p II LPCfls no ereqedpwnps i eqired,jgpJçCore Spray isot

!funetloL SRO Enters C-i, Alternate Level Control Directs lineup of Injection Systems Irrespective of Pump NPSH and Vortex limits (LPCI and CS) per Appendix 6B and 6D Answers Yes to can 2 or more CNDS, LPCI or CS Injection Subsystems be aligned with pumps running When RPV Water Level drops to -162 inches, Then continues Answers Yes to is any CNDS, LPCI or CS Injection Subsystem aligned with at least one pump running Before RPV Water Level drops to -180 inches continue CT#1 A;nwp es to a& piimjs run atabxstore and mamtaiRPV I above 48 Dprsuiization EmergenqyEPV Dçpressunzton is1çqired Enters C-2 Directs iimizing RPV Injection from all available sources irrespective of pump NP and Vortex Limits CT#2 Answers Yes to can RPV Water Level be restored and maintained above -180 inches Exits C-i and enters EOI-1, RPV Control at step RC/L-i BOP/ATC

• Tnhi T#4 Lme upTP [op I pumps for opAjpendix 6i46D r*

C After uergency reswizafIon pinmenced veiflesJPV Tneçtoiized 4jj

3-C Page 37 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate 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 SUCT VLV.

4. VERIFY CLOSED the following valves:

• 3-FCV-74-61, RHR SYS I DW SPRAY INBD VLV

• 3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

• 3-FCV-74-57, RHR SYS I SUPPR CHBR!POOL ISOL VLV

• 3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

• 3-FCV-74-59, RHR SYS I SUPPR POOL CLG!TEST VLV

5. VERIFY RHR Pump 3A and/or 3C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RHR SYS I LPCI TNBD 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-C Page 38 of 58 Simulator Event Guide:

Event 8 Component: Loop I Core Spray Logic Power Failure BOP/ATC Appendix 6D, Loop I Core Spray

1. VERIFY OPEN the following valves:

• 3-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCT VLV

• 3-FCV-75-l 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.

3-C Page 39 of 58 Simulator Event Guide:

Event 8 Component: Loop I Core Spray Logic Power Failure 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 YIV wo ciii 64DS Valye be j,eue4 Maintains 6 A pniIPV co tdow lnterkcl are clear BOP/ATC Reports Suppression Pool Level in Feet when directed by SRO Ope,sØ 4DS v rtfie hçtç rr 4fl,r CT#Z When RPV Pressure is low enough fi Injection ofLPCI and Core Spray, operator should vçdfy available systems are injecting. At this time perator shoühuotice Core Spra oopIJujeciiou Valve not opeiiaudtakç çti,utp manually open e vaIve When adequate core cooling is assured begins to throttle flow to prevent overfilling RPV. Must secure pumps on Loop II LPCI to stop injection.

RC iii bpar&irijection/dIve s lLhavingpøwer. Wilibe imabieW throttle flow, wben Loop U QiJqug rçquiredpimps must ccd tetio1

3-C Page 40 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Answers No to can Drywell Temperature be maintained below 1 60F Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI-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-C Page 41 of58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO Enters EOI-2 on High Dryweli Pressure (cont)

Pc/P Monitor and control Primary Containment pressure below 2.4 psig 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 (Does not initiate Suppression Chamber Sprays because Adequate Core Cooling is not assured at this time)

Pc/fl 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 SP/T Monitor and Control Suppression Pool Temperature below 95F using available Suppression Pool Cooling Answers Yes to can Suppression Pool Temperature be maintained below 95F (Once Emergency Depressurization has commenced Suppression Pool Temperature will exceed 95F, this step should be re-addressed once Adequate Core Cooling is assured)

3-C Page 42 of 58 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO swL Monitor and Control Suppression Pool Level between -l and -6 inches Answers Yes to can Suppression Pool Level be maintained above -6 inches Answers Yes to can Suppression Pool Level be maintained below -l inches SRO Enters EOI-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 except systems required to:

• Be operated by EOIs OR

• Suppress a Fire Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels Answers No to Is Any Area Radiation Level 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 Secondary Water Levççiiditions may 1ek IqlQt isçJated tjniey manner SRO The Emergency Classification is 1.1-Si Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

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

3-C Page 43 of 58 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs:

HPCI is tagged out for Preventive Maintenance.

Stator Water Cooling Pump 3B is tagged out.

Operations/Maintenance for the Shift:

BOP Operator perform 3-01-3 Section 8.13 Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump Once completed perform Control Rod Pattern adjustment in accordance with the Reactivity Control Plan Units 1 and 2 are at 100% power.

Unusual Conditions/Problem Areas:

None

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3-C Page 49 of 58 3.6 CONTAINMENT SYSTEMS 361.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.61.3 Each PCIV. except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES i 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation.

ACTIONS NOTES--- --

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

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

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

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, Primary Containment when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

3-C Page 50 of 58 CONDITION REQUIRED ACTION COMPLETION TIME A. -----

NOTE------ k 1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only apphcable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs and de-activated AND

---

automatic valve, dosed manual valve, blind 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured MSIV leakage not within limits.

AND (continued)

3-C Page 51 of58 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A (continued) A2 -NOTE Isolation devices in high radiation areas may be verified by use of administrative means -

Verily the affected Once per 31 days penetration flow path is for isolation isolated, devices outside primary containment AND Prior to entering MODE 2 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 (rrntini urfl

3-C Page 52 of 58 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B - NOTE --.- 5.1 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated

---

automatic valve, closed manual valve, or blind One or more penetration flange.

flow paths with two PCIVs inoperable except due to MSIV leakage not within limits.

C. NOTE C,1 Isolate the affected: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EECVs) with only one PC IV. and de-activated

---

automatic valve, closed AND manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for flow paths with one PCIV EECVs inoperable. AND C.2 --------------NOTE----

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

(continued)

3-C Page 53 of 58 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration 0.1 Restore leakage rate to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flow paths with MSIV within limit.

leakage not within limits.

E. Required Action and E. I Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, C, AND or 0 not met in MODE 1, 2, or 3 El Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F. Required Action and F. 1 Initiate action to suspend Immediately associated Completion operations with a Time of Condition A, B, C, potential for draining the or 0 not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during MODE 4 or 5.

F.2 NOTE Only applicable for inoperable RHR Shutdown Cooling Valves.

Initiate action to restore Immediately valve(s) to OPERABLE status.

3-C Page 54 of 58 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.

ACTIONS NOTE LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System inoperable. A.1 Verify by administrative Immediately means High Pressure Coolant Injection System is OPERABLE.

AND A.2 Restore RCIC System to 14 days OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND 8.2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> dome pressure to 150 psig.

3-C Page 55 of 58 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES I and 2.

ACTIONS

• NOTE Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control NOTE rod stuck. Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, ControI Rod Block if required, to allow continued operation.

A.1 Verify stuck control rod Immediately separation criteria are met.

AND A.2 Disarm the associated 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> control rod drive (CRD).

AND (continued>

3-C Page 56 of 58 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Perform SR 3.1.3.2 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from SR 3.1.3.3 for each discovery of withdrawn OPERABLE Condition A control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A.4 Perform SR 3.1.1.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> control rods stuck.

C. One or more control rods CA -----

NOTE inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control rod.

AND 0.2 Disarm the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> CRD, (continued)

3-C Page 57 of 58 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. NOTE --- D.1 Restore compliance with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Not applicable when BPWS.

THERMAL POWER

>1O%RTP. QE D.2 Restore control rod to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Two or more inoperable OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, C, or D not met.

OR Nine or more control rods inoperable.

3-C Page 58 of 58 BROWNS FERRY I EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 ti-Ui I NOTE I 11-1.12 I I I I Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity with irradiated fuel assemblies expected to Pool with irradiated fuel assemblies expected to Z remain covered by water. remain covered by water.

C r

m OPERATING CONDITION: OPERATING CONDITION Mode 5 ALL 1.1-Al I INOTEI I l.1-A21 I I Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity expected to result in irradiated fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered, assemblies being uncovered.

OPERATING CONDITION: OPERATING CONDITION:

Mode 5 ALL 1.1-SI I INOTEI I II-S2p I I I Reactor water level can NOT be maintained Reactor water level can NOT be determined.

above -162 inches, (TAF) m m

m

J Q

m OPERATING CONDITION: OPERATING CONDITION:

ALL Model or2or3 -<

l.1-G1 I I I I 1.l-G2 I I NOTE 1 TABLE I US Reactor water level can NOT be restored and Reactor water level can NOT be determined maintained above -180 inches. AND Either of the following exists:

. The reactor will remain subcritical without boron under all conditions, and m Less than 4 MSRVs can be opened, or Z

> Reactor pressure can NOT be restored and maintained above Suppression Chamber pressure by at least

• • UNIT19opsi rn

• • UNIT28opsi

+ UNIT37opsi

. It has NOT been determined that the reactor will remain subcritical without boron under all z conditions and unable to restore and maintain C)

MAREP in Table 1.1-G2. -C OPERATING CONDITION: OPERATING CONDITION:

Modelor2or3 Model or2or3

Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: D Op-Test No.: ILT 1102 FINAL SRO:

Examiners: Operators: ATC:

BOP:

Initial 1C193 / Unit 3 Reactor Power 4% / Condensate Pump 3A tagged Conditions:

Turnover: Aligning Charcoal Filters for Parallel Flow lAW 3-01-66 Section 5.11. Raise Power with Control Rods for Mode Change Event Event No. Maif. No. Type* Event Description 1 Aligning Charcoal Filters for Parallel Flow 5.11 2 - Raise Power with Control Rods for Mode Change R-SRO C-ATC 3 thO3b Reactor Recirc Pump 3B Trip TS-SRO TS-SRO CS Pump 3A inadvertent initiation with loss of minimum flow 4 trg5 C-BOP protection 5 msO 1 Steam Seal Regulator failure 6 fw3Oc Feedwater Pump 3C Governor drifts up 7 pcl4 M-ALL Torus Leak / ATWS 8 trg 20 C 3-FCV-73-30 Fails to Open

-FCV-74-57 fails to open (If repair team called for, open valve after 9 C ED started)

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

Appendix D Scenario Outline Form ES-fl-i Critical Tasks Five CT#i-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.

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.

CT#2-When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of Containment.

2. Cues:

Procedural compliance.

Suppression Pool level trend.

3. Measured by:

Observation US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet.

AND Observation RO opens at least 6 SRVs during performance of Emergency Depressurization actions.

4. Feedback:

RPV pressure trend.

Suppression Pool temperature trend.

SRV status indication.

Appendix P Scenario Outline Form ES-D-1 Critical Tasks Five CT#3-With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

RPV pressure indication.

3. Measured by:

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

4. Feedback:

RPV level trend.

RPV pressure trend.

Injection system flow rate into RPV.

CT#4-With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BITT) and inserting control rods.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Suppression Pool temperature.

3. Measured by:

Observation If operating lAW EOI-l and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool.

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

AND Control Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

Appendix D Scenario Outline Form ES-fl-i Critical Tasks Five CT#5-When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.

1. Safety Significance:

Prevent failure of Primary Containment from pressurization of the Suppression Chamber.

2. Cues:

Procedural compliance.

Suppression Pool Level indication

3. Measured by:

Observation HPCI Auxiliary Pump placed in Pull to Lock

4. Feedback:

HPCI does not Auto initiate No RPM indication on HPCI

Appendix P Scenario Outline Form ES-P-i Scenario Summary:

The Plant is operating at 4% Reactor Power.

The BOP Operator will Aligning Charcoal Filters for Parallel Flow JAW 3-01-66 section 5.11.

The ATC will withdraw control rods in order to raise power to 8% for a mode change from 2 to 1.

Once the NRC is satisfied with the reactivity manipulation, Reactor Recirculation Pump B will trip.

The SRO will direct entry to 3-A0I-68-1A, the ATC will close RR Pump B discharge valve.

The SRO will evaluate Technical Specification 3.4.1 Condition A is required.

Core Spray Pump 3A inadvertently initiates with loss of minimum flow protection. BOP Operator verifies initiation is inadvertent and with SRO concurrence stop Core Spray Pump 3A JAW with ARPs. The SRO will evaluate Technical Specification 3.5.1 Condition A is required.

The Steam Seal regulator will fail, the BOP Operator will take action JAW with the ARPs and restore steam seal pressure with the bypass valve.

The operating feedwater pump controller will fail, level will slowly rise until the ATC or Crew notices the Reactor Level change. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to maintain Reactor Level control in manual.

SRO should direct entry into 3-AOI-3-l.

An unisolable Torus leak will commence. Suppression Pool level will start to lower and continue to lower. The SRO will enter EOI-3 on flood alarms and eventually EOI-2 on Suppression Pool Level.

The SRO will determine that Suppression Pool level cannot be maintained above 11.5 feet and enter E0I-l to scram the reactor and then to Emergency Depressurize.

An ATWS will exist on the scram, the crew will work through EOI-1 and C-S to insert control rods, maintain reactor level and pressure. The SRO will transition to C-2 to Emergency Depressurize.

Attempts to add water to the suppression pool will be unsuccessful with the failure of 3-FCV-73-30 and 3-FCV-74-57.

The Emergency Classification is 1.2-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

Appendix B Scenario Outline Form ES-B-i SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-D 7 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 4 EOIs used: List(1-3) 2 EOI Contingencies used: List (0-3) 90 Validation Time (minutes) 5 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix B Scenario Outline Form ES-B-i Scenario Tasks EVENT TASK NUMBER KIA RO SRO 1 Align Charcoal Filters RO U-066-NO-22 271000A4.09 3.3 3.2 2 Raise Power with Control Rods RO U-085-NO-06 SRO S-000-AD-31 2.2.2 4.6 4.1 3 Reactor Recirc Pump Trip RO U-068-AB-O1 20200 1A2.03 3.6 3.7 SRO S-068-AB-01 4 Core Spray Inadvertent Initiation RO U-075-NO-0l 209001A3.02 3.8 3.7 5 Steam Seal Regulator Failure RO U-001-AL-01 245000K6.01 2.8 2.9 SRO S-047-AB-03 6 Reactor Feed Pump Turbine Governor Failure RO U-003-AL-09 259002A4.01 3.8 3.6 SRO S-003-AB-01 7 Torus Leak/ATWS RO U-000-EM-14 295030EA2.01 4.1 4.2 RO U-000-EM- 17 RO U-000-EM-83 SRO S-000-EM-07 SRO 5-000-EM-iS SRO S-000-EM- 18

3-D Page 8 of 56 Procedures Used/Referenced:

Procedure Number Procedure Title Procedure Revision 3-01-66 Off Gas System Revision 59 3-G0I-100-1A Unit Startup Revision 91 3-01-85 Control Rod Drive System Revision 70 3-A0I 1 A Recirc Pump Trip/Core Flow Decrease OPRMs Operable Revision 6 TS 3.4.1 Recirculation Loops Operating Amendment 221 3-ARP-9-3C Alarm Response Procedure Panel 3-9-3C Revision 26 TS 3.5.1 ECCS Operating Amendment 244 3-ARP-9-6B Alarm Response Procedure Panel 3-9-6B Revision 11 3-ARP-9-7A Alarm Response Procedure Panel 3-9-7A Revision 22 3-A0I-47-3 Loss of Condenser Vacuum Revision 11 Loss of Reactor Feedwater or Reactor Water Level 3-A0I-3-1 .

Revision 9 High/Low 3-ARP-9-3B Alarm Response Procedure Panel 3-9-3B Revision 18 3-ARP-9-4C Alarm Response Procedure Panel 3-9-4C Revision 30 TS 3.6.2.2 Suppression Pool Water Level Amendment 212 3-EOI-2 Primary Containment Control Flowchart Revision 7 3-EOI-APPENDJX-1 8 Suppression Pool Water Inventory Removal and Makeup Revision 2 3-EOI-3 Secondary Containment Control Flowchart Revision 9 3-EOI-1 RPV Control Flowchart Revision 8 3-EOI-APPENDIX-3A SLC Injection Revision 1 3-EOI-3-C-5 Level-Power Control Flowchart Revision 9 3-EOI-APPENDIX-4 Prevention of Injection Revision 5 3-EOI-3-C-2 Emergency RPV Depressurization Flowchart Revision 8 3-EOI-APPENDIX-6A Injection Subsystems Lineup Condensate Revision 2 3-E0I-APPENDIX-6B Injection Subsystems Lineup RHR System I LPCI Mode Revision 3 3-EOI-APPENDIX-6C Injection Subsystems Lineup RHR System II LPCI Mode Revision 3

3-D Page 9 of 56 Procedures Used/Referenced Continued:

Procedure Number ] Procedure Title Procedure Revision 3-EOI-APPENDIX-1F Manual Scram Revision 2 3-EOI-APPENDIX-l D Insert Control Rods Using Reactor Manual Control System Revision 2 3-EOI-APPENDIX-2 Defeating ART Logic Trips Revision 4 Bypassing Group 1 RPV Low Low Low Level 3-EOI-APPENDIX-8A Revision 1 Isolation_Interlocks Bypassing Group 6 Low RPV Level and High Drywell 3E01-APPENIIIX8E Revision 1 Pressure Isolation Interlocks 3-AOl-i 00-1 Reactor Scram Revision 53 3-EOI-APPENDJX-1 7A RHR System Operation Suppression Pool Cooling Revision 5 Emergency Classification Procedure Event Classification EPIP-1 Revision 46 Matrix EPIP-4 Site Area Emergency Revision 32

3-D Page 10 of 56 Console Operator Instructions Scenario File Summary File: batch and trigger files for scenario 3-D Batch nrc2OlldRl

1. cp pump 3 a clearance ior ypobkrcndpa fail_power

1. Recirc Pump B trip imfth03b (el 0)

1. cs Initiation ior zdihs755a[4] (e5 0) start ior zdihs759a[2] (e5 0) close

1. steam seal failure imfms0l (elO 0) imf mcO4 (e 10 0)100

1. FWLC fail imffw30c (e15 0)100 3000 54 trg 7 nrc20llfptc trg 7 = dmffw30c Trigger nrc2O 1 lfptc zdihs46 1 Oa{4] .ne. 1

1. SP LEAK ATWS/major bat atws75 imfpcl4 (e20 0)100 300 75 ior ypovfcv733o (e20 0) fail_now trg2l =batatws-1 trg 22 bat appOif trg23 =batappo2 ior zdihs7457a[2] auto bat nrcstick20 trg 24 = bat nrcunstickl4 trg25=batsdv

3-D Page 11 of56 Batch nrcstick2O imfrdO6r3Ol5 imf rdO6r3 023 imfrd06r303 1 imfrdO6rl85l imf rdO6rl 439 imfrd06rl43 1 imf rd06r34 15 imfrdO6r38l5 imf rd06r42 15 imfrd06r463 1 imfrdO6r5439 imf rdO6r3 027 imfrd06r263 1 imf rd06r26 15 imf rd06r223 9 imf rdO6r3 839 imfrd06rl4l 5 imfrd06r30l 5 imfrdO6r46l5 imfrdO6r2223 Batch nrcunstickl4 dmf rd06r343 5 dmfrdO6r3423 dmfrd06r263 1 clmfrdO6r343 1 dmf rd06r263 9 dmfrdO6r3439 dmf rdO6r3 027 dmfrdO6r3427 dmfrdO6r2243 dmf rdO6r2 643 dmf rdO6r3 043 dmf rdO6r3 443 dmfrd06rl 843 dmfrdO6rl8l9

3-D Page 12 of56 Console Operator Instructions Scenario 3-D DESCRIPTION/ACTION Simulator Setup manual Reset to IC 193 Simulator Setup Load Batch Bat nrc2Ol ldRl Simulator Setup manual Clearance out Condensate pump 3A Simulator Setup Verify Batch file loaded RCP required (Raise Power from 4% to 8% with Control Rods for Mode Change) Provide marked up copy of 3-GOI-100-1A and RCP

3-D Page 13 of56 Simulator Event Guide:

Event 1 Normal: Aligning Charcoal Filters for Parallel Flow lAW 3-01-66 Section 5.11 SRO Direct BOP to align Charcoal Filters for parallel flow.

BOP Align Charcoal Filters for Parallel Flow JAW 3-01-66 section 5.11.

5.11 Aligning Charcoal Filters for Parallel Flow:

[1] PLACE the OFFGAS TREATMENT SELECT handswitch, 3-XS-66-1 13, in TREAT.

[2] OPEN the CHARCOAL ADSORBER TRAIN 2 INLET VALVE, using 3-HS-66-1 17.

[3] OPEN the CHARCOAL ADSORBER TRAIN 1 DISCH VALVE, using 3-HS-66-1 18.

[4] CLOSE the CHARCOAL ADSORBER TRAINS SERIES VLV, using 3-S.-66-116.

[5] CHECK dewpoint temperature on OFFGAS MOIST SEP REHEATER TEMPERATURE recorder, 3 -TRS 108, indicates 45°F or less (Red Pen).

[6] IF the Off-Gas System is intended to be operated with charcoal beds in parallel with the charcoal beds on another (shut down) unit, THEN NOT TypógrjhicaI errori do not requirstopping pr&cIure perforriiaiice These NPG-SPP errors should be noted, and corrected following performance ofthe 12 procedure Tins does not apply to changes m component identifiers, numerical units, values, Innits, work sequence or where the potential exists for impropei operation ofplant equipment

3-D Page 14 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase Direct Power increase using Control Rods per 3-GOl- 100-1 A, section 5.4 5.4 Withdrawal of Control Rods while in Mode 2

[67] CONTINUE to withdraw control rods to raise Reactor power to approximately 8%. (REFER TO 3-01-85 and 3-SR-3.1.3.5(A))

ATC Raise Power with Control Rods per 3-01-85, section 6.6. The following are the first 10 rods to be withdrawn b5, 26-34-59,38-35, 58,27, 34-O326-O3, &-27,06-47 an&14-55 ill rods start at 12 and go to 48 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

[2] VERIFY the following prior to control rod movement:

CRD POWER, 3-HS-85-46 in ON.

• Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

• Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.

• Nuclear Instrumentation responds as control rods move through the core (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize automatic RBM Rod Block as follows:

• STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].

[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REfNITIALIZE the RBM:

[6.1] PLACE CRD POWER, 3-HS-85-46 in the OFF position to deselect the Control Rod.

[6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position.

I

3-D Page 15 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 3-XS-85-40.

[2] OBSERVE the following for the selected control rod:

• CRD ROD SELECT pushbutton is brightly ILLUMiNATED

• White light on the Full Core Display ILLUMiNATED

• Rod Out Permit light ILLUMINATED

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

3-D Page 16 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[6.2] PLACE and HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

• Four rod display digital readout and the full core display digital readout and background light remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

3-D Page 17 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2

[7] IF continuously withdrawing the control rod to position 48 and the control rod coupling integrity check will be performed after the CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[7.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

[7.3] WHEN position 48 is reached, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

[7.5] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.

[7.6] CHECK control rod coupled by observing the following:

• Four rod display digital readout AND full core display digital readout AND background light will remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator (3-XA-55-5A, Window 14) does NOT alarm.

3-D Page 18 of56 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[7.8] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2.

6.6.5 Return to Normal After Completion of Control Rod Withdrawal

[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:

[1.1] PLACE CRD POWER, 3-HS-85-46, in OFF.

[1.2] PLACE CRD POWER, 3-HS-85-46, in ON.

DRJVER When NRC dIrects, insert Trigger 1 for Ieacfor RecirePnp 313 trip

3-D Page 19 of56 Simulator Event Guide:

Event 3: Reactor Recirc Pump 3B Trip PYER ATC Respond to numerous alarms and Report Trip of Reactor Recirc Pump 3B SRO Enter 3-AOI-68-1A Recirc Pump Trip/Core Flow Decrease OPRMs Operable ATC 4.2 Subsequent Actions

[1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A),

[1.1] SCRAM the Reactor.

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

Closes 3B Recirc Pump Discharge Valve ATC [3] IF Region I or II of the Power to Flow Map is entered, THEN Steps 3 through 8 are N/A SRO [9] NOTIFY Reactor Engineer to PERFORM the following:

• Tech Specs 3.4.1

• 3-SR-3.4.1(SLO), Reactor Recirculation System Single Loop Operation

• O-TI-248, Core Flow Determination in Single Loop Operation

3-D Page 20 of 56 Simulator Event Guide:

Event 3: Reactor Recirc Pump 3B Trip SRO Evaluate Tech Spec for Single Loop Operation TS 3.4.1 Condition A Condition A Requirements of the LCO not met.

Required Action A.1 Satisfy the requirements of the LCO Completion Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> MODE Change not permitted until setpoint changes complete.

ATC [10] [NER/C] WHEN the Recirc Pump discharge valve has been closed for at least five minutes (to prevent reverse rotation of the pump) [GE SIL-517], THEN (N/A if Recirc Pump was isolated in Step 4.2[8])

OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibrium.

Opens Recirc Pump 3B discharge valve BOP [11] REFER TO the following ICS screens to help determine the cause of recirc pump trip/core flow lowering. VFDPMPB and VFDBAL

[12] CHECK parameters associated with Recirc Drive and Recirc Pump/Motor 3B on ICS and 3-TR-68-7 1 to determine cause of trip.

Dispatch personnel [13] PERFORM visual inspection of tripped Reactor Recirc Drive.

Dispatch personnel [14] PERFORM visual inspection of Reactor Recirc Pump Drive relay boards for relay targets.

DRTVER AsR Eeeracknow1edeequest dieps.if er iksRE fr directions on completion ofrod withdrawal, direct to complete the rod that was in progress of being withdrawn and STOP Any field investigation for pump trip, reportno obvious causes: Pump reaker: 4KV Recire BD3B DR1VR Wen RC direcJnsrt igger$for Corpray Pimj3 nadvertut start.

3-D Page 21 of56 Simulator Event Guide:

Event 4: Core Spray Pump 3A Inadvertent Initiation DWEi WhnCdirfs,insert TiCgger for orprayPuiij3Ainadverteritstart:

Delete Pnhp Start override immediateIyaftepumpstarts{o allow operator to ccc wp.

BOP Report inadvertent start of Core Spray Pump 3A and alarm CORE SPRAY SYS I PUMP A START BOP A. VERIFY auto start signals by multiple indications.

B. VERIFY Pump 3A operation by motor amps, discharge pressure, and flow on Panel 3-9-3.

B. IF pump is NOT needed, THEN STOP Pump before 5 mm time limit at minimum flow expires.

D. WHEN the auto start signal is reset and Core Spray is NOT required for Core Cooling, THEN E. RETURN system to standby readiness.

BOP Report drywell pressure and reactor level normal and stops Core Spray Pump 3A BOP Dispatches personnel to investigate pump start DRIVR detetrnrewhy pump started I electrical maintenance will be coacte SRO Evaluate Technical Specification 3.5.1 Condition A One low pressure ECCS injection/spray subsystem inoperable.

Required Action A. 1 Restore low pressure ECCS injection/spray subsystem(s) to Operable status.

Completion Time 7 Days pmV hen NRC directs insert.Triggçr 10 for Steam Seal Reg4tor Fai1ure

3-D Page 22 of 56 Simulator Event Guide:

Event 5: Steam Seal Regulator Failure DRIy)3R i*c dir ts,ner(TriggerlO for Steaiii ea1RegojFailure BOP Respond to Annunciator STEAM TO STEAM SEAL REG PRESS LOW A. CHECK steam seal header pressure, 3-PI-l-148, Panel 3-9-7.

B. VERIFY proper valve alignment on Panel 3-9-7.

C. IF pressure is low, THEN OPEN steam seal bypass valve 3-FCV-l-145.

D. DISPATCH personnel to check 3-PIC-l-l47 (El 617 Turb Bldg).

E. CHECK condenser vacuum on 3-P/TR-2-2 (Panel 3-9-6) and turbine vibration on 3-XR-47-15 (Panel 3-9-7) normal.

BOP Responds to Annunciators STEAM PACKING EXHAUSTER VACUUM LOW OG HOLDUP LINE INLET FLOW HIGH BOP Recommends opening steam seal bypass valve 3 -FCV- 1-145 to restore steam pressure SRO Concurs with actions to restore steam seal pressure BOP Dispatches personnel and checks condenser vacuum DRIVER Repxts Condenseracuum 1ç or s1oIyç1egrading DIER J(periiel dis edep I-I-T4Thafaile4 iw no air pressu indication, çr44çte a1nction mcOcondens ir)çkage SRO Evaluate entry to 3-AOI-47-3 Loss of Condenser Vacuum BOP Once steam seal pressure is restored resets annunciators and verifies condenser vacuum is improving.

• i ai iiert TnggerI fo Fiedpr Punp 9avernqr<3Iure

3-D Page 23 of 56 Simulator Event Guide:

Event 6: Feedwater Pump 3C Governor Drifts Up

.P . ;V /

DRIYER When NRC 6irects,insert Trigger 15 for Feedwater Pump Governor Failure. Whe operator takes the RFPT Govçrnorto nianuaf the malfunction is automatically deleted, therefore,, IF the operatorpulls theGovernor control knob back out, the malfunction must be manually reinserted and dleted when the operator returns the

çycotroi knob ba to force the operator to rol level inanua1ly ATC Report Rising Reactor Water Level and RFPT is not responding.

SRO Direct manual control of operating RFPT and Enter 3-AOI-3.. 1.

4.2 Subsequent Actions

[1] VERIFY applicable automatic actions.

[16] IF Feedwater Control System has failed, THEN PERFORM the following:

[16.1] PLACE individual RFPT Speed Control Raise/Lower switches in MANUAL GOVERNOR (depressed position with amber light illuminated).

[16.2] ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain level.

[20] IF level continues to rise, THEN TRIP a RFP, as necessary.

[22] IF RFPs are in manual control, THEN LOWER speed of operating RFPs.

[23] EXPECT a possible Reactor power rise due to a rise in moderation.

[24] IF unit remains on-line, THEN PERFORM the following:

• RETURN Reactor water level to nonnal operating level of 33 (normal range).

• REQUEST Nuclear Engineer check core limits.

ATC Take MANUAL GOVERNOR control of RFPT and maintain Reactor Water Level Manually in the Normal Level Band. Operator may attempt to control RFPT with PDS.

PDS will not respond.

DRIER If a sram is msertTr at NRC &ctiou rnitia tngger 2Ofore Suppression Pool Leak

3-D Page 24 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS DRIVER If a scram is inserted or atNRC direction initiate trigger 20 for the Suppression Pool Leak ATC/BOP Respond to alarm multiple Pump Room Flood Level alarms and SUPPR CI1A1v1BER WATER LEVEL ABNORMAL ATC/BOP Report lowering suppression pool water level A. CHECK level using multiple indications.

B. IF level is low, THEN DISPATCH personnel to check for leaks.

C. IF level is high, THEN D. REFER TO 3-01-74, Sections 8.2, 8.3, and 8.4.

E. REFER TO Tech Spec Section 3.6.2.2.

F. IF level is above -1 or below -6.25, THEN ENTER 3-E0I-2 Flowchart.

D1VER Whi dispaciIwuE 6unuts tepøxtwr 1ev1 is 4 inches and rising n the oiitheast Quad, Waeris foingin from he Toiis Are, Cpnotdeteiine source of the 1ek.

SRO Enter EOI-2 on Low Suppression Pool Level Monitor and Control Suppression Pool Level Between -l inch and -6 inches (Appendix 18)

Answers No to Can Suppression Pool Level Be Maintained Above -6 inches Answers Yes to Can Suppression Pool Level Be Maintained Below -1 inches SRO CT#5 et a a eorJ1PC1to 1aee inPuto Lock pior to ifee ATC/BOP cn

3-D Page 25 of 56 Simulator Event Guide:

Event 8 Component: 3-FCV-73-30, HPCI PUMP MTN FLOW VALVE, fails to open SRO Directs Appendix 18 BOP Appendix 18

6. IF Directed by SRO to add water to suppression pool, THEN MAKEUP water to Suppression Pool as follows:

a. VERIFY OPEN 3-FCV-73-40, HPCI CST SUCTION VALVE.

b. OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE

c. IF HPCI is NOT available for Suppression Pool makeup, THEN MAKEUP water to Suppression Pool using RCIC as follows:

1) VERIFY OPEN 3-FCV-7l-19, RCIC CST SUCTION VALVE.

2) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.

BOP Attempts to makeup water to the Suppression Pool using HPCI; 3-FCV-73-30 has lost power. Utilizes RCIC to makeup water to the Suppression Pool and dispatches personnel to investigate 3-FCV-73-30.

DRIVE 3-FCV-73-3Q poweriailswhn the Torus leak is inserted, crew will dispatch personnel to investigate. Acknowledge investigation:and provide no fuftherinformation.

SRO Determines vthgger vlrefor inserting a Rjctor Scram. on lowering Supprsioa T# Pool Water Level cqteO-I, Sns aotSr before Srpprcssion.PooI ieyel bc is SRO Detemiines that Emergency Makeup to the Suppression Pool using Standby Coolant is required and directs BOP to line up Standby Coolant to the Suppression Pool per Appendix 18.

3-D Page 26 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS BOP Appendix 18

5. IF Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN CONTINUE in this procedure at Step 9.

9. IF Directed by SRO to Emergency Makeup to the Suppression Pool using Standby Coolant Supply, THEN MAKEUP water to the Suppression Pool as follows:

a. VERIFY CLOSED the following valves:

• 3-FCV-74-61, RHR SYS I DW SPRAY INBD VALVE

• 3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VALVE

• 3.FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

• 3-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VALVE

• 3-FCV-23-46, RHR HX 3B RHRSW OUTLET VALVE

b. VERIFY RHR Pumps 3A and 3C are NOT running.

c. PLACE 3-BKR-074-OlOO, RHR HTX A-C DISCH XTIE (TO U-2) VLV FCV-74-lOO (MO1O-171) to ON (480V RMOV Board 3B, Compartment 19A).

d. START RHRSW Pumps Bi and B2.

e. NOTIFY Unit 1 Operator to VERIFY CLOSED l-FCV-23-46, RHR HEAT EXCHANGER B COOL WATER OUTLET VLV

3-D Page 27 of 56 Simulator Event Guide:

Event 9 Component: 3-FCV-74-57, RHR SYS I SUPPR CHMBR/POOL ISOL VLV, fails to open DRIVER When personnel dispatched to close 3-BKR-074-OIOO, wait I niinutes.then close breaker and report, delete override for breaker control power. When requested 1-FCV-23-46is closed When requested toopen 2-FCV-23-57 insert remote function swO9 open BOP Appendix 18 (continued)

f. NOTIFY Unit 2 Operator to perform the following

1) VERIFY CLOSED 2.-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2) OPEN 2-FCV-23-57, STANDBY COOLANT VLV FROM RHRSW.

g. INJECT Standby Coolant into the Suppression Pool as follows:

1) CLOSE 3-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VLV.

2) OPEN 3-FCV-74-lOO, RHR SYS I U-2 DISCH XTIE.

3) OPEN 3-FCV-74-57, RHR SYS I SUPPR CHMBR/POOL ISOL VLV.

4) THROTTLE OPEN 3-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV to control injection.

BOP Determines 3-FCV-74-57 will not open and is unable to Emergency Makeup to the Suppression Pool, dispatches personnel to determine cause of valve failure.

Acknowledges-dispatch and provides no further infonnation until crew has opened al valves Once all ADS valves are opened delete ovemde zdzhs7457a[2] auto and inform crew that the valve would not open due to dirty contacts and the problem has been fixcd; SRO Enters EOI-3 on Flood Alarms

3-D Page 28 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS Enters EOI-3 on Flood Alarms SRO EOI-3 Secondary Containment Temp Monitor and Control Secondary CNTMT Temp Answers No to Is Any Area Temp Above Max Normal EOI-3 Secondary Containment Radiation Monitor and Control Secondary CNTMT Radiation Levels Answers No to Is Any Area Radiation Level Above Max Normal EOI-3 Secondary Containment Level Monitor and Control Secondary CNTMT Water Level Answers Yes to Is Any Floor Drain Sump Above 66 inches Answers Yes to Is Any Area Water Level Above 2 inches Restore and Maintain Water Levels using all available sump pumps Answers No to Can All Water Levels be Restore and Maintained Below Isolate all systems that are discharging into the area except systems required to:

• Be operated by EOIs OR

• Suppress a Fire Answers No to Will Emergency Depressurization Reduce Discharge Into Secondary Containment.

SRO Enters EOI-1 at pre-determined trigger value and directs Reactor Scram based on EOI-2 step SPIL-7.

Dfly After the first channel ofARI, initiate Tngger 25 for Eat SOy, further ATWS áctión are on page 41.

ATC Inserts Reactor Scram, Initiates One Channel of ARI and reports rods out

3-D Page 29 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Enters EOI-1 from EOI-2 step SP[L-7 Verify Reactor Scram EOI-1 RC/P Monitor and Control RPV pressure Exits RC/P and enters C-2, Emergency RPV Depressurization, based on Override step RC/P-4.

EOI-1 RC/L 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.-l RC/Q 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 AOI-100-1, Reactor Scram, based on override RC/Q-2.

(The following steps will be executed through AOI-lOO-1 if RC/Q exited)

Verify Reactor Mode Switch is in Shutdown Initiate second channel of ART Verify Recirc Pump Runback (Pump speed 480rpm or less)

Answers No to is Reactor Power above 5% or Unknown (The Following steps N/A if RC/Q exited)

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

Initiates SLC per Appendix 3A

3-D Page 30 of 56 Simulator Event Guide:

Event 7 Major: Torus LeakIATWS SRO EOI-1 RCIQ (cont)

Inhibit ADS Verify RWCU System Isolation Answers Yes to is SLC injecting into the RPV Stops at step RC/Q-l 8 until SLC has injected into the RPV to a tank level of 43%, then exits RC/Q and enters AOl-i 00-1 Trips the SLC pump when SLC tank level drops to 0%

ATC Initiates Second Channel of ART 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_ifReactor_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)

3-D Page 31 of56 Simulator Event Guide:

Event 7 Major: Torus LeakJATWS BOP/ATC c4

1. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A/3B, 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-PI-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 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-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

3-D Page 32 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Enters C-5 from EOI-1 step RC/L-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:

• MSW 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)

DRWER WI en requested for appendix 8A and 8E wait 4 minutes and insert bat app08ae and report PP!

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 rcquired continues at step 05-19 CT#1/2 and eaters C-2,y directionf EOI-2 step SP/L-6 and-from EOI-1 step RC/P4 and directs Crew toStop andPjvent all Jnjection Sourees tothe RPV Eeptfrom.çIC,CRDan<

sçpçrsteç5-2o, j BOP/ATC 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 the RPV Except from RCIC, CRD and SLC

3-D Page 33 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS BOP/ATC Appenlix4 CT#i

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 NJ 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

3-D Page 34 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS BOP/ATC Appendix 4 (eontinued

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 8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO, OR

• Using individual 3-SIC-46-8(9)(l 0), 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, RFP 3B DISCHARGE VALVE

• 3-FCV-3-5, RFP 3C DISCHARGE VALVE

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

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

• 3-HS-3-125A, RFPT 3A TRIP

• 3-HS-3-151A, RFPT 3B TRIP

• 3-HS-3-176A, RFPT 3C TRIP.

3-D Page 35 of56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS T#Z Determines Eegçuçy Depissrization is requeeutçrç-2 Answers No to will the reactor remain subcritical under all conditions. Waits until SRO 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 BOP/ATC Reports when Appendix 4 is complete Reports Suppression Pool Level in Feet when Directed CT#2 Opens an4 Verifiespen.ALL ADS Va1veswhen dire ted 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 until RP(Presure is keJoSvMA1P (f0psig with i4SRVs open)

CT#3- Then continued Directs, crew to Start and Slowly raise RPV injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSHlmnts and Sppression PóoHvel per AppeIIdi 6Aor per Appejidix 613. 6C

3-D Page 36 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CT#3 Appendix 6A

1. VERIFY CLOSED the following Feedwater heater return valves:

BOP/ATC

• 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, RFP 3A DISCHARGE VALVE

• 3-FCV-3-12, RFP 38 DISCHARGE VALVE

• 3-FCV-3-5, RFP 3C DISCHARGE VALVE

3. VERIFY OPEN the following drain cooler inlet valves:

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

• 3-FCV-2-84, DRAIN COOLER 3B5 CNDS INLET ISOL VLV

• 3-FCV-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-l25, LP HEATER 3B3 CNDS OUTL ISOL VLV

• 3-FCV-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-FCV-3-24, HP HTR 3C2 FW INLET ISOL VLV

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

• 3-FCV-3-76, HP HTR 3Bl FW OUTLET ISOL VLV

• 3-FCV-3-77, HP HTR 3C1 FW OUTLET ISOL VLV

6. VERIFY OPEN the following RFP suction valves:

• 3-FCV-2-83, RFP 3A SUCTION VALVE

• 3-FCV-2-95, RFP 3B SUCTION VALVE

• 3-FCV-2-108, RFP 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, RFW START-UP LEVEL CONTROL, to control injection (Panel 3-9-5).

10. VERIFY RFW flow to RPV.

3-D Page 37 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CT#3 pp46fl BOP/ATC 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 SUCT VLV

4. VERIFY CLOSED the following valves:

• 3-FCV-74-61, RHR SYS I DW SPRAY INBD VLV

• 3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

• 3-FCV-74-57, RHR SYS I SUPPR CHBR/POOL ISOL VLV

• 3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

5. VERIFY RHR Pump 3A andJor 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-D Page 38 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS 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 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-FCV74-73, 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

3-D Page 39 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS 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 6B, 6C SRO EOI-1 RCIQ steps RC/Q-20 and RC/Q-21 Reset ART Defeat ART Logic Trips if necessary (APPX 2) (This step is N/A, however, crew may choose to perform this step)

CT#4 Insert Control Rods by performing Appendix IF and 1D Appendix iF: 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

1) Drive Control Rods. Bypass RWM if necessary BOP/ATC Dispatch personnel to perform Appendix 2(N/A) and outside portions of Appendix 1 F.

Dispatch personnel to close 3-FCV-85-5 86 (while awaiting completion of Appendix 1 F)

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

3-D Page 40 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CT#4 Appcmlix ATC

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.

DIUVE1 1 recie4 opefoxm Appendti2 an4 ouIide portioxis of Appndix fF ii 3 Wii&

mhfl4es. pert I ggers 21,22,23., and24 then pftcoiiWletion.

If directed.to close 3-FcV-85-586 wait 3 minutes then insert mrfrdO6 c1osc Then porom.pIçtipn WWhen direJted to reQpen 3FCy8-586 wa 3 mitestben iflsert mrfrdO6 Qp1p1etiQ1

3-D Page 41 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS ppçndiiU ATC 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

[N 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 1 F and 1 D until all rods inserted OR Until EOI-1 RC/Q is exited due to Reactor determined to be Subcritical at which point continue to insert rods per 3-AOI-lOO-1 and 3-01-85

3-D Page 42 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Executes all legs of EOI-2 concurrently (SP/L leg has been previously addressed)

EOI-2 DWIT Monitor and control Drywell Temperature below 1 60F using available Drywell Cooling Answers Yes to can Drywell Temperature be maintained below 1 60F EOI-2 PC/P 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 EOI-2 PC/H 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)

EOI-2 SP/T Monitor and control Suppression Pool temperature below 9SF using available Suppression Pool Cooling (APPX 1 7A) 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 1 7A) (After Emergency Depressurization complete and Reactor Water level restored)

BOP Performs Appendix 1 7A to place Suppression Pool cooling in service after Emergency Depressurization and restoration of Reactor Water level.

3-D Page 43 of 56 Simulator Event Guide:

Event 7 Major: Torus LeakJATWS BOP 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 NJ 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 by SRO, Then PLACE 3-XS-74-122(130), 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 CHBRJPOOL ISOL VLV.

3-D Page 44 of 56 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS

= 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 indicated on 3-FI-74-50(64), RHR SYS 1(11) FLOW:

• Between 7000 and 10000 gpm for one-pump operation.

OR

• At or below 13000 gpm for two-pump operation.

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

k. MONITOR RHR Pump NPSH using Attachment 1.

1. 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.

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

3-D Page 45 of 56 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:

Condensate Pump 3A tagged Out of Service.

Operations/Maintenance for the Shift:

Align Charcoal Filters for Parallel Flow lAW 3-01-66 Section 5.11.

Once completed Raise Power with Control Rods for Mode Change lAW 3-GOI-100-1A, section 5.4 step

[67] and the Reactivity Control Plan Units 1 and 2 are at 100% power.

Unusual Conditions/Problem Areas:

None

0 Ii

ru F

2

9N I 0

0 mm 11 CD 4.

00 C

cI

m m

C, I

CD C

ji QN J

3-D Page 51 of56 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall te 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 loop operation limits specified in the COLR:

b. ICO 3.2,2, MINIMUM CRITICAL POWER RATIO (MCPR),

single loop operation limits specified in the COIR;

c. LCO 3.3.1.1, Reactor Protection System (RPS) lnstrumentatIon, Function 2.b (Average Power Range Monitors Flow 6iased Simulated Thermal PowerS. High), Allowable Value of Table 3.3.1.1.1 is reset for single loop operation.

APPLICA5ILITY: MODES I and 2.

ACTIONS CONDITION REOUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satls.v the requirements 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> not met. of the LCO.

B. Required Action and 6.1 8e in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met OR No recirculation loops in operation.

3-D Page 52 of 56 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS LCO 3.OA.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injectionlspray inoperable, subsystem(s) to OPERABLE status.

OR One low pressure coolant injection (LPCI) pump in both LPC) subsystems inoperable.

(continued)

3-D Page 53 of 56 ECCS Operating 35i ACTIONS (continued)

CONDITION FEQUIRED ACTION COMPLETION TIME B. Required Action arid B.i Be in MODE 1 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met.

B2 Be in MODE 4 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

3-D Page 54 of 56 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. HPCI System inoperable. C.1 Verify by administrative immediately means RC IC System is OPERABLE AND C2 Restore HPCI System to 14 days OPERABLE status.

D. I-IPCI System inoperable. 0.1 Restore HPCI System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status AND OR Condition A entered.

02 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem to OPERABLE status.

E. One ADS valve El Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One ADS valve F. I Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

AND Condition A entered. F.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem to OPERABLE status.

(continued)

3-D Page 55 of 56 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME ci Two or more ADS valves cii Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperabla AND OR ci2 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action and dome pressure to associated Completion 150 psig Time of Condition C, D, E, or F not met.

EL Two or more low pressure ELI Enter LCO 102. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

3-D Page 56 of 56 SCRAM FAILURE REACTOR COOLANT ACTIVITY I

cim cwy 4 C

equivaIen 1-IS1 (Th nical Specificaon Lirnts z as delennined Sy chemrysame, C (4

C OPERATINS COND1TION m ALL 12..A I I NOTE I I t3A I I Failure of RPS automalic scram functions to bnng Reactor olant activity exceeds 300 iC1m dose the reactor subctitic equivalertt lodine4Sl as detennined y chemistty AND sample.

Manual scram or ARt (automaticor m.anual)was successful.

OPERATING CONDITION; OPERATING CONDITION; Mode 1 or2 or 3 Mode ler2 I 2S I I NOTE I I I I I Falure of automatic scram. manual scram, and ARI to bring the reactor subcritical.

m nl i

OPERATING CONDITION; Mode 1 12 I CURVE I I I Failure of automatic scram, manual scram, and US 1 I ARI. Reactor power is above 3% 0 AND z

Either of the following conditions evists:

Suppression Pool temp exceeds HCTL, m Refer to Curve t24.

Reactor water level can NOT be restored and maintained at or above 480 inches, OPERATING CONDITION:

Mode Ior2

Appendix D Scenario Outline Form ES-fl-i Facility: Browns Ferry NPP Scenario No.: F Op-Test No.: ILT 1102 FJNAI.

SRO:

Examiners: Operators: ATC:

BOP:

Initial IC 104/ Unit 2 Reactor Power 70%! EECW A3 Pump tagged Out! RFPT B Out of Service Conditions:

Turnover: Remove LPRM 8-49B from bypass JAW 2-OI-92B section 6.4, then raise power with Control Rods as directed by the RCP.

Event Event No. Maif. No. Type* Event Description N-BOP 1 Remove an LPRM from Bypass 8-49B N-SRO R-ATC 2 Raise Power with Control Rods R-SRO rd25 C-ATC RPIS Position Failure rod 14-3 5, will drift in when inserted to position rd07r1435 TS-SRO 46 C-BOP 4 sw03m D3 EECW Pump Trip TSSRO 5 ms05h - Outboard MSJV D Partial Closure TS-SRO I-ATC 6 fw26a!b Feedwater Flow Transmitters fail ISR 7 mcO4 M-ALL Degrading Vacuum, ATWS with out MSIVs 8 iaO2 C Loss of Drywell Control Air 9 rdOl C 2A CRD Pump Trip

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

Appendix B Scenario Outline Form ES-D-1 Critical Tasks Three CT#1-With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BuT) and inserting control rods.

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 per EOI- 1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping) before exceeding 1100 F in the Suppression Pool.

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

AND Control Rod insertion commenced in accordance with EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

CT#2-During an ATWS, when conditions are met to deliberately lower RPV level, Terminate and Prevent injection into the RPV from ECCS and Feedwater until conditions are met to reestablish injection.

1. Safety Significance:

Precludes loss of primary containment integrity and uncontrolled release of radioactivity into the environment.

2. Cues:

Procedural compliance.

3. Measured by:

Observation With Emergency Depressurization not required and >5% power, injection systems are terminated and prevented until:

• <5% power or < -162 with Suppression Pool Temp> 1100 F OR

• Level <(-) 50 inches with Suppression Pool Temp < 1100 F

4. Feedback:

Injection system flow rates into RPV Reactor Power lowering

Appendix D Scenario Outline Form ES-B-i Critical Tasks Three CT#3-With reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS.

1. Safety Significance:

Precludes core damage due to an uncontrolled reactivity addition.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation unless all required injection systems are Terminated and Prevented.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B INHIBITED annunciator status.

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

BOP will remove LPRM 8-49B from bypass JAW 2-OI-92B section 6.4.

ATC will raise Reactor Power with control rods as directed by the Reactivity Control Plan.

During power ascension Control Rod 14-3 5 will experience an RPIS position failure. The crew will respond JAW ARPs and 2-AOI-85-4. The ATC will insert Control Rod 14-3 5 one notch to establish position indication. After Control Rod 14-3 5 is inserted it will begin to drift in, the ATC will respond JAW 2-AOI-85-5 and insert the control rod to position 00.

EECW D3 Pump will trip and the standby EECW Pump B3 will fail to auto start, the BOP will respond JAW ARPs and start EECW Pump B3 to EECW flow to the north header.

The SRO will evaluate Technical Specification 3.7.2 and Condition A is entered.

Outboard MSIV D will drift closed, the crew will respond JAW 2-AOI-1-3. The ATC will lower Reactor Power to less than 66% and the BOP will fully close Outboard MSIV D. The SRO will evaluate Technical Specification 3.6.1.3 and Condition A is entered.

Feedwater Flow Transmitters will fail the crew will respond JAW ARPs and 2-AOI-3 -1, the ATC will report that Feedwater Level Control transferred to single element and will transfer to single element. Reactor Level will stabilize after the initial transient.

Vacuum will begin to degrade and the crew will respond JAW 2-AOI-47-3, the crew will insert a manual Reactor scram prior to the Main Turbine trip. An ATWS will exist and the crew will enter EOI- 1 and C-5.

After the scram and airline break will occur in the drywell causing MSJV closure and transition to SRVs for pressure control and RCIC for level control. Until the crew performs Appendix 8G, SRV operation will degrade due to the loss of air.

CRD Pump 2A will trip and the ATC will start CRD Pump lB in order to insert control rods.

The crew will maintain directed level and pressure bands, insert all control rods and enter EOI-2 and place RHR in Suppression Pool Cooling.

The Emergency Classification is 1.2-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

Controls Rods are being inserted Reactor Level is being maintained in directed level band

Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2-F 6 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 60 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

Appendix B Scenario Outline Form ES-B-i Scenario Tasks EVENT TASK NUMBER KIA RO SRO 1 Remove an LPRM from Bass RO U-92B-NO-05 215005A4.04 3.2 3.2 2 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 3 RPIS Position Failure ROU-085-AL-14 214000A2.01 3.1 3.3 SRO S-085-AB-04 4 EECW Pump Trip RO U-067-NO-12 400000A2.01 3.3 3.4 5 MSIV Partial Closure RO U-001-AB-02 239001A2.03 4.0 4.2 SRO S-001-AB-02 6 Feedwater Flow Transmitter Failure RO U-003-NO-12 259002A2.02 3.3 3.4 SRO S-003-AB-01 7 Vacuum Loss/ATWS RO U-000-EM-17 295037EA2.06 4.0 4.1 SRO S-000-EM-06 SRO S-000-EM- 18 SRO S-032-AB-02

2-F Page 7 of 61 Procedures Used/Referenced:

Procedure Number Procedure Title Procedure Revision 2-0I-92B Average Power Range Monitor Revision 38 2-GOT- 100-12 Power Maneuvering Revision 40 2-01-8 5 Control Rod Drive System Revision 125 2-ARP-9-5A Alarm Response Procedure Panel 2-9-5A Revision 46 2-A0I-85-4 Loss of RPIS Revision 20 TRM 3.3.5 Surveillance Instrumentation Revision 0 2-A0I-85-5 Rod Drift In Revision 19 2-ARP-9-20A Alarm Response Procedure Panel 2-9-20A Revision 24 2-ARP-9-23D Alarm Response Procedure Panel 2-9-23D Revision 12 0-01-67 Emergency Equipment Cooling Water System Revision 91 Emergency Equipment Cooling Water System and Ultimate TS 3 7 2

. Amendment 254 Heat Sink 2-ARP-9-5B Alarm Response Procedure Panel 2-9-SB Revision 25 2-AOl- 1-3 Main Steam Isolation Valve Closure at Power Revision 22 TS 3.6.1.3 Primary Containment Isolation Valves Amendment 253 2-ARP-9-6C Alarm Response Procedure Panel 2-9-6C Revision 19 Loss of Reactor Feedwater or Reactor Water Level 2-AOI-3-l . Revision 20 HighlLow 2-ARP-9-53 Alarm Response Procedure Panel 2-9-53 Revision 35 2-AOI-47-3 Loss of Condenser Vacuum Revision 18 2-E0I-1 RPV Control Flowchart Revision 12 2-E0I-APPENDIX-8G Crosstie CAD to Drywell Control Air Revision 4 2-EOI-APPENDIX-l 1A Alternate RPV Pressure Control Systems MSRVs Revision 4 2-E0I-2-C-5 Level-Power Control Flowchart Revision 11 2-.EOI-APPENDIX-4 Prevention of Injection Revision 10 2-E0I-APPENDIX-5C Injection System Lineup RCIC Revision 4 2-EOI-APPENDIX-5D Injection System Lineup HPCI Revision 6

2-F Page 8 of6l Procedures Used/Referenced Continued:

_Procedure Number Procedure Title Procedure Revision 2-EOI-APPENDIX-3A SLC Injection Revision 5 2-EOI-APPENDIX-2 Defeating ARI Logic Trips Revision 4 2-EOI-APPENDIX-1F Manual Scram Revision 5 2-EOI-APPENDIX- 1 D Insert Control Rods Using Reactor Manual Control System Revision 6 2-AOI-85-3 CRD System Failure Revision 23 2-EOI-2 Primary Containment Control Flowchart Revision 10 2-EOI-APPENDIX- 1 7A RHR System Operation Suppression Pool Cooling Revision 12 2-EOI-APPENDIX-12 Primary Containment Venting Revision 3 Emergency Classification Procedure Event Classification EPIP-l Revision 46 Matnx EP]P-4 Site Area Emergency Revision 32

2-F Page 9 of 61 Console Operator Instructions A. Scenario File Summary

1. File: batch and trigger files for scenario 2-F Batch NRC/110202 Imf sw07b Bat atws70 Imf rd0la (e2 120)

Trgel NRC/msivd = zdthsols2a[1].eq.1 Trgel = bat NRC/i 10202-3 br xa555b23 alarm off Trge3 NRC/singleelement = zdihs466a.eq. 1 Trge3 = bat NRC/i 10202-4 Imf ia02a(e2 15)100100 Imf ia02b (e2 60) 100 30 0 Batch NRC/1102-1 br zlohs466a off br zlohs466b on Imf fw26a (none 0) 0 bmffw26b (none 60) 100 30 0 Batch NRC/1102-2 bmfth27e br zlohs0 1 52a[2] on Imf ms05h br za0fi464 1.6 Batch NRC/1102-3 Dor zlohsOlS2a[2]

Dor zaofi464 Batch NRC/1102-4 Dor zlohs466a Dor zlohs466b

2-F Page 10 of6l Pref file F3 imfrd25 F4 imfrdO7rl43S F5dmf rdO7 1435 F6 imfsw03m F7 bat NRC/i 10202 F8 bat NRC/i 102-i F9 bat NRC/i 102-2 FlO dmfth27e Fli F 12 trg e2 modesw Shift fi imf mcO4 100 Shift f4 mrfrd06 open Shift f5 bat appOif Shift f6 bat app02 Shift f7 mrfrd06 close Shift f8 bat sdv

2-F Page 11 ofól Console Operator Instructions Scenario 2-F DESCRIPTION/ACTION Simulator Setup manual Reset to IC 104 Simulator Setup Load Batch RestorePref NRC/i 10202 Simulator Setup manual Tag Out EECW Pump A3 Simulator Setup manual F7 and F12 Simulator Setup Verify file loaded RCP required (70% 85% with control rods and flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12

2-F Page 12 of6l Simulator Event Guide:

Event 1 Normal: Remove LPRM 8-49B from bypass JAW 2-OI-92B section 6.4 SRO Directs LPRM 8-49B un-bypassed.

BOP Removes LPRM 8-49B from bypass lAW 2-OI-92B section 6.4.

6.4 Returning an LPRM to Operate From a Bypassed Condition

[1] REVIEW all precautions and limitations. REFER TO Section 3.0.

[2] REFERENCE Illustration 4 to find the APRM/LPRM Channel associated with the desired LPRM to be returned to normal.

[3] At Panel 2-9-14, DEPRESS any softkey to illuminate the display on the desired APRM/LPRM channel chassis.

[4] DEPRESS the ETC softkey until BYPASS SELECTIONS illuminates on the bottom row of the display.

[5] DEPRESS BYPASS SELECTIONS softkey, enter the password, and DEPRESS ENT.

[6] SELECT the desired LPRM to be returned to service by using the left or right arrows on the softkey board until the inverse video illuminates the correct LPRM.

[7] DEPRESS the OPERATE softkey.

[8] CHECK the BYP/HV OFF is replaced by OPERATE below the selected LPRM.

[9] DEPRESS EXIT softkey to return display to the desired bargraph.

[10] VERIFY, as a result of returning this LPRM to operate, that any alarms received on Panel 2-9-5 or on the APRM/LPRM channel are reset.

2-F Page 13 of 61 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase.

Direct Power increase using Recirc Flow per 2-G0I-l 00-12.

[20] IF desired to raise power with oniy two (2) Reactor feedpumps in service, THEN RAISE Reactor power, as desired, maintaining each Reactor feedpump less than 5850 RPM.

ATC Raise Power with Control Rods per 2-01-85, section 6.6. Control Rods 22-3 1, 30-39, 38-31 and 30-23 from 00 to 16, 30-31 from 00 to 48.

6.6.1 Initial Conditions Prior to Withdrawing Control Rods

[2] VERIFY the following prior to control rod movement:

• CRD POWER, 2-HS-85-46 in ON.

• Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

• Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.

• Nuclear Instrumentation responds as control rods move through the core. (This ensures control rod is following drive during Control Rod movement.)

[5] ATTEMPT to minimize automatic RBM Rod Block as follows:

• STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 2-9-5 and PERFORM Step 6.6.2[6].

[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:

[6.1] PLACE CRD POWER, 2-HS-85-46 in the OFF position to deselect the Control Rod.

[6.2] PLACE CRD POWER, 2-HS-85-46, in the ON position.

2-F Page 14 of 61 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[2] OBSERVE the following for the selected control rod:

• CRD ROD SELECT pushbutton is brightly ILLUMiNATED.

• White light on the Full Core Display ILLUMiNATED.

• Rod Out Permit light ILLUMINATED.

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH, and RELEASE.

[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.

2-F Page 15 of6l Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 2-XS-85-40.

[2] OBSERVE the following for the selected control rod:

• CRD ROD SELECT pushbutton is brightly iLLUMINATED.

• White light on the Full Core Display ILLUMINATED.

• Rod Out Permit light ILLUMINATED.

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRDE.

[5.2] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[5.3] WhEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 2-HS-85-47 and CRD CONTROL SWITCH, 2-HS-85-48.

[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

2-F Page 16 of6l Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[5.5] VI1EN control rod settles into the intended notch, THEN CHECK the following.

• Four rod display digital readout and the full core display digital readout and background light remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does NOT alarm.

[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position, with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

• Four rod display digital readout and the full core display digital readout and background light remain illuminated.

• CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48.

r DR1VEJ ne4z ytdrwajc 4

2-F Page 17 of6l Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 2-AOI-85-2.

ATC 6.6.5 Return to Normal After Completion of Control Rod Withdrawal

[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:

[1.1] PLACE CRD POWER, 2-HS-85-46, in OFF.

[1.2] PLACE CRU POWER, 2-HS-85-46, in ON.

13kflfE1 Jnrt rod 3O3l is bemgthd&h to 4

2-F Page 18 of6l Simulator Event Guide:

Event 3 Instrument: RPIS Position Failure Control Rod 14-35 ATC Report Control Rod Drift Alarm 5A-28, reports no control rods drifting.

Reports loss of position indication on Control Rod 14-35.

SRO Enter 2-AOI-85-4 Loss of RPIS.

ATC 4.1 Immediate Actions

[1] STOP all control rod movement.

SRO 4.2 Subsequent Actions NOTE Reference TRM 3.3.5, RPIS Indicated Channel Operability, for applicable 7 or 30 day LCO relating to an inoperable RPIS indication.

[1] IF control rod movement is required with a Total loss of RPIS, THEN MANUALLY SCRAM reactor.

[2] NOTIFY the Operations Superintendent and Reactor Engineer for actions to be taken in a timely manner.

SRO [9] IF unable to restore position indication for an individual control rod or rods, THEN NOTIFY Reactor Engineer and DETERMiNE additional corrective action. Control Rods may be moved to an Operable Position Indication as a means of position verification (REFER TO Tech Spec Bases SR 3.1.3.1). As a minimum, rod position will be verified, preferably with an independent position indication or other method.

DRtVER A idp.g ations,f çd ijk caTl&to reminendaction.

Rrdto co np1Iç inoy Woeq1jd 3kto 4.

SRO Direct ATC to insert Control Rod 14-35 one notch to attempt to establish position indication.

ATC Insert Control Rod 14-35 to position 46.

SRO Evaluate Technical Requirements Manual 3.3.5. Information LCO Condition A and Condition C from table 3.3.5-1 DRIVER. Weiieontrrc,d f4-S5 nseetp3siti&i>46 ij iiifr4O7445, when t rI sposi__aj ti

2-F Page 19 of 61 Simulator Event Guide:

Event 3 Instrument: RPIS Position Failure Control Rod 14-35 ATC Report Control Rod Drift Alarm 5A-28, reports Control Rod 14-3 5 drifting in.

SRO Enter 2-AOI-85-5 Rod Drift In.

ATC 4.1 Immediate Actions

[1] IF multiple rods are drifting into core, THEN MANUALLY SCRAM Reactor.

Refer to 2-AOI-100-1.

SRO 4.2 Subsequent Actions

[1] IF a Control Rod is moving from its intended position without operator actions, THEN INSERT the Control Rod to position 00 using CONTINUOUS iN.

[2] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

[3] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 2-AOI-100-1.

ATC Inserts Control Rod 14-35 to position 00.

[4] CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).

[5] ADJUST control rod pattern as directed by Reactor Engineer and CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).

Crew Dispatch AUO to check scram valves.

jrj As eaetc> engineer when ca1leihavecrew stop contrcro4 As UO aftr dispate repo cam va1re are nonn1i SRO Evaluate Tech Spec 3.1.3 Condition C One or more control rods inoperable for reasons other than Condition A or B Required Action C. 1 Fully Insert inoperable control rod Completion Time 3 Hours AND Required Action C.2 Disarm the associated CRD Completion Time 4 Hours Nic

4Wi

2-F Page 20 of 61 Simulator Event Guide:

Event 4 Component: EECW Pump D3 Trip BOP Respond to alarms 20A-21 and 23D-26.

23D-26 41 60V SD BD D MOTOR OL or TRIP Overload or trip out, on any one of the following:

CS pump 1D, 2D, RHR pump 1D, 2D, RHRSW pump D2, D3 A. CHECK control room for white light illuminated on effected equipment.

B. DISPATCH personnel to check:

1. Relays at associated electrical bd.

2. Equipment for abnormal conditions, relay targets, smell, burned paint, breaker.

20A EECW SOUTH HDR DG SECTION PRESS LOW B. CHECK Panel 2-9-3 for status of South header pump(s) breaker lights and pump motor amps normal.

C. NOTIFY UNIT SUPERVISOR, Unit 1 and Unit 3.

D. START standby pump for affected header. REFER TO 0-01-67.

8.11 Recovering from an EECW Pump Trip

[1] VERIFY < 25 minutes has elapsed since the EECW pump trip and header pressure

> 0 psig.

[3] IF the south header pump has tripped, THEN:

[3.1] START desired RRRSW Pump using one of the following:

• RHRSW PUMP D3(B3) EECW SOUTH HDR, 0-HS 94A/2(88A/2) on Unit 2.

[4] For the EECW(RHRSW) pump(s) started, PERFORM the following:

• VERIFY running current is less than 53 amps.

• VERIFY locally, Pump breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged.

• VERIFY Pump upper and lower motor bearing oil level is in the normal operating range.

• NOTIFY Chemistry of running RHRSW (EECW) pump(s).

BOP Start EECW Pump B3.

SRO Evaluate Technical Specification 3.7.2.

P When dtspatched rpoECWump 1)3 nothing pip, brclç

2-F Page 21 of6l Simulator Event Guide:

Event 4 Component: EECW Pump D3 Trip SRO Evaluate Technical Specification 3.7.2.

Condition A: One required EECW pump inoperable.

Required Action A. 1: Restore the required EECW pump to OPERABLE status.

Completion Time: 7 days NR When ready, MSfl?iriia1Iosure DRIVER Wken directed P9 bat NRC/I 10202-2,when alarm 5B-18 alarms F10 dm1 th27e.

2-F Page 22 of 61 Simulator Event Guide:

Event 5 Component: Outboard MSIV D Partial Closure ATC Respond to alarm 5B-18 MAiN STEAM LINE CH A FLOW HIGH.

5B-18 MAiN STEAM LINE CH A FLOW HIGH A. VERIFY alarm by checking main steam flow indicators.

B. IF alarm is valid on any steam line, THEN MANUALLY SCRAM Reactor and PLACE Rx Mode Sw. in Shutdown and CLOSE MSIVs.

cmis 41ni oyaiid)

C. IF any flow indicators are low, THEN CHECK all MSIVs open.

D. REFER TO 2-AOI-l-3.

E. REFER TO Tech Spec Table 3.3.6.1-1.

ATC Report Steam flow in D line is lower than A, B and C lines.

ATC/BOP Report Outboard MSIV D 1-52 indicates partially closed.

SRO Enter 2-AOI-1-3, MSIV Closure at Power.

4.1 Immediate Action None 4.2 Subsequent Action

[1] IF any EOI entry condition is met, THEN (Otherwise N/A):

ATC [2] LOWER reactor power with recirc flow and insert control rods as necessary, when directed by the Reactor Engineer/Unit Supervisor, to ensure that rated steam line flow (3.54 x 106 lbmlhr) is NOT exceeded; as indicated on Main Steam Line Flow Indicators.

ATC/BOP [3] IF an MSIV is partially closed, THEN:

[3.1] LOWER reactor power to 66%.

[3.2] PLACE the associated MSW control switch to CLOSE.

ATC Lower Power to 66%.

BOP PLACE the Outboard MSIV D 1-52 control switch to CLOSE.

2-F Page 23 of6l Simulator Event Guide:

Event 5 Component: MSIV Partial Closure SRO Evaluate Technical Specification 3.6.1.3.

Condition A: NOTE Only applicable to penetration flow paths with two PCIVs.

One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits Required Action A. 1: Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

Completion Time: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main steam lines Required Action A.2: NOTE Isolation devices in high radiation areas may be verified by use of administrative means.

Verif the affected penetration flow path is isolated.

Completion Time: Once per 31 days for isolation devices outside primary containment NRC When ready, Fedwater Flow Transr terFailur DIYR When directe418bat)4RC/11O2O2-i.

2-F Page 24 of 61 Simulator Event Guide:

Event 6 Instrument: Feedwater Flow Transmitter Failures Respond to alarm 6C-14 RFWCS INPUT FAILURE.

A. VERIFY RFWCS continues to maintain Reactor Water level.

B. IDENTIFY bad/invalid signal by checking Control Room instrumentation and/or ICS. REFER TO ATTACHMENT 1, on next page, for list of RFWCS ATC instrumentation. REFER TO ICS RX FW LVL CONTROL SYS display (FWLCS).

C. REQUEST assistance from Site Engineering.

D. BYPASS the bad/invalid signal with Unit Supervisor approval.

ATC Report Feedwater Flow signal has failed LOW for FW Line A.

ATC Report FW Line B Feedwater Flow signal failing ifiGH.

SRO Enter 2-AOI-3-1, Loss of Feedwater or Reactor Water Level High/Low.

4.1 Immediate Actions None 4.2 Subsequent Actions

[2] IF Feedwater Flow signal fails (FI-3-78A, FI-3-78B), THEN PERFORM the following:

A. With SROs permission, REFER TO 2-01-3 and BYPASS failed Feedwater Flow Instrument in Unit l&2 Computer Room; or Unit 2 Aux Instrument Room.

[2.1] IF both Feedwater Flow Instruments fail, THEN VERIFY level control transfers to SiNGLE ELEMENT.

ATC Verifies Reactor Level control in single element, level control failed to transfer to single element; Operator depresses single element pushbutton to transfer.

[6] IF Reactor Water Level continues to rise, THEN TRIP RFP, as necessary.

[7] IF RFPs in automatic control, THEN VERIFY 2-LIC-46-5 lowers flow of operating RFPs.

ATC Verifies RFPTs maintain water level.

NRC When ready for Major Vacuum Leak DRIVBR UpçnLe

çqtioi Q4OO:

2-F Page 25 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP Respond to alarm 53-14 OG HOLDUP LINE INLET FLOW HIGH.

ATC Report degrading condenser Vacuum.

SRO Enter 2-AOI-47-3, Loss of Condenser Vacuum.

4.1 Immediate Actions None 4.2 Subsequent Actions

[1] IF ANY EOI entry condition is met?_THEN:

[2] IF unable to maintain hotwell pressure below -25 inches Hg, as indicated on 2-XR-2-2, with Reactor power less_than_30%,_THEN_TRIP the main turbine.

[4] REDUCE reactor power in an attempt to maintain condenser vacuum.

SRO Determines a trigger value for Reactor Scram prior to Turbine Trip; at 25 inches.

ATC Insert Reactor Scram when directed; and place mode switch in shutdown. Report ATWS and initiate first channel of ARI.

DRIVER Ri.ght after the scram eater<shift ES> Bat SDYi SRO Enter 2-EOI-1, RPV Control.

SRO EOI-1 (Reactor Pressure)

Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO -

IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate ?- NO IF Emergency Depressurization is 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

2-F Page 26 of6l Simulator Event Guide:

Event 8 Component: ATWS without MSIVs SRO 2-EOI-1 (Reactor Pressure) ll Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? YES.

THEN crosstie CAD to Drywell Control Air, Appendix 8G.

IF Boron injection is required? NO SRO Direct a Pressure Band of 800 to 1000 psig, Appendix 1 1A.

ATC/BOP Maintain directed pressure band, JAW Appendix 1 1A.

BOP Crosstie CAD to Drywell control air, JAW Appendix 8G.

SRO IF Main Steam Relief Valve Air Accumulator Low annunciator, (XA-5 5-3D- 18) is in alarm, THEN: place each MSRV Control Switch in Close/Auto AND Place MSRV Auto Actuation Logic Inhibit XS-1 -202 to Inhibit.

ATC/BOP Places XS-l-202 to inhibit.

EOI- 1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems: HPCI Appendix 1 1C, RCIC Appendix 11B, RFPTs on SRO minimum flow Appendix 1 iF, Main Steam System Drains Appendix 1 1D, Steam Seals Appendix 11 G, SJAEs Appendix 11 G, Off Gas Preheater Appendix 11 G, RWCU Appendix 1 1E.

WIJ

2-F Page 27 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Pressure Control JAW Appendix 1 1A, 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 fi, THEN:

CLOSE MSRVs_and CONTROL RPV pressure using other options.

3. OPEN MSRVs; using the following sequence to control RPV pressure, as directed bySRO:

a. 2-PCV-l-179 MN STM LINE A RELIEF VALVE

b. 2-PCV-1-180 MN STM LINED RELIEF VALVE.

c. 2-PCV-l-4 MN STM LINE A RELIEF VALVE

d. 2-PCV-l-31 MN STM LINE C RELIEF VALVE

e. 2-PCV-l-23 MN STM LINE B RELIEF VALVE

f. 2-PCV-1-42 MN STM LINED RELIEF VALVE

g. 2-PCV-1-30 MN STM LINE C RELIEF VALVE

h. 2-PCV-l-19 MN STM LINE B RELIEF VALVE.

i. 2-PCV-l-5 MN STM LINE A RELIEF VALVE.

j. 2-PCV-l-41 MN STM LINED RELIEF VALVE

k. 2-PCV-1-22 MN STM LINE B RELIEF VALVE

1. 2-PCV-.1-18 MN STM LINE B RELIEF VALVE

m. 2-PCV-1-34 MN STM LINE C RELIEF VALVE

2-F Page 28 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Pressure Control lAW Appendixl 1A, RPV Pressure Control SRVs

3. IF Drywell Control Air header, supplied from CAD System A; shows indications of being depressurized, as determined by Appendix 8G, THEN:

OPEN MSRVs supplied by CAD System B, using the following sequence to control_RPV pressure,_as_directed by_SRO:

4. IF Drywell Control Air header, supplied from CAD System B; shows indications of being depressurized, as determined by Appendix 8G, THEN:

OPEN MSRVs supplied by CAD System A, using the following sequence to control RPV pressure, as directed by SRO:

6. IF BOTH Drywell Control Air headers are depressurized, THEN PERFORM the following as directed by EOI-l, RPV Control, RC/P Section:

• PLACE each MSRV control switch in CLOSE/AUTO, and PLACE 2-XS-1-202, MSRV AUTO ACTUATION LOGIC INHIBIT, to INHIBIT.

AND

• MINIMIZE MSRV cycling by using sustained openings for RPV depressurization.

EOI- 1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems: HPCI Appendix 1 1C, RCIC Appendix 1 1B, RFPTs on SRO minimum flow Appendix 1 1F, Main Steam System Drains Appendix 1 lD, Steam Seals Appendix 1 1G, SJAEs Appendix 1 lG, Off Gas Preheater Appendix 1 1G, RWCU Appendix 1 1E.

ATC/BOP Augment RPV Pressure Control, if directed by SRO.

DRWER If Appendix 8G is performed, TUEN delete Instrument Air Leaks iaO2a and iaO2b

2-F Page 29 of6l Simulator Event Guide:

Event 8 Component: Loss of Drywell Control Air BOP Crosstie CAD to Drywell control air, lAW Appendix 8G.

1. OPEN the following valves:

• O-FCV-84-5, CAD SYSTEM A N2 SHUTOFF VALVE ( Panel 9-54)

• 0-FCV-84-16, CAD SYSTEM B N2 SHUTOFF VALVE (Panel 9-55).

2. VERIFY 0-PI-84-6, N2 VAPORIZER A OUTLET PRESSURE, and 0-PI-84-17, N2 VAPORIZER B OUTLET PRESSURE, indicate approximately 100 psig (Unit 1, Panel 9-54 and 9-55).

3. PLACE keylock switch 2-HS-84-48, CAD A CROSS TIE TO DW CONTROL AIR, in OPEN (Unit 2, Panel 9-54).

4. CHECK OPEN 2-FSV-84-48, CAD A CROSS TIE TO DW CONTROL AIR, (Unit 2, Panel 9-54).

5. PLACE keylock switch 2-HS-84-49, CAD B CROSS TIE TO DW CONTROL AIR, in OPEN (Unit 2, Panel 9-55).

6. CHECK OPEN 2-FSV-84-49, CAD B CROSS TIE TO DW CONTROL AIR (Unit 2, Panel 9-55).

7. CHECK MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 2-PA-32-31, alarm cleared (2-XA-55-3D, Window 18).

8. IF MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 2-PA-32-31, annunciator is or remains in alarm (2-XA-5 5-3D, Window 18), THEN DETERMINE which Drywell Control Air header is depressurized as follows:

a. DISPATCH personnel to Unit 2, RB, El 565 ft, to MONITOR the following indications for low pressure:

b. MONITOR 0-FI-84-7(18), CAD LINE A(B) N2 FLOW, on Unit 1, Panel 1-9-54(55) for high flow.

c. MONITOR inboard MSIV indication status for valves drifling closed.

9. IF Drywell Control Air header supplied from CAD System A shows indications of being depressurized, THEN CLOSE the following valves:

• 0-FCV-84-5, CAD SYSTEM A N2 SHUTOFF VALVE (Panel 9-54)

• 2-FSV-84-48, CAD A CROSS TIE TO DW CONTROL AIR (Panel 9-54).

10. IF Drywell Control Air header supplied from CAD B shows indications of being depressurized, THEN CLOSE the following valves:

• 0-FCV-84-16, CAD SYSTEM B N2 SHUTOFF VALVE (Panel 9-55)

• 2-FSV-84-49, CAD B CROSS TIE TO DW CONTROL AIR (Panel 9-55).

• I jwii OxIs

2-F Pag e 30 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO EOI-1 (Reactor Level)

Monitor and Control Reactor Level.

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

ATC/BOP Verifies Group 2 and 3 isolation.

or will remain subcritical, THEN Exit RC/L; SRO iF it has not been determined that the react ENTER C5 Level / Power Control.

NO Is Emergency Depressurization is required?

NO RPV Water level cannot be determined?

Boron under all conditions? NO The reactor will remain subcritical without w 105 feet OR Suppression Chamber pressure PC water level cannot be maintained belo cannot be maintained below 55 psig? NO-SRO cr# pçtsJithbtç4 ATC/BOP CT#3 hthi1fts ADS.

SRO Is any Main Steam Line Open?- NO

2-F Page 31 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO C5 Level / Power Control Crosstie CAD to DW Control Air, if necessary (Appendix 8G).

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% ?- YES Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC (Appendix 4).

WHEN RPV Level drops below -50 inches; THEN Continue:

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 IF YES?

Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC; irrespective

• of any consequent reactor power or reactor water level oscillations.

WhEN RPV Level drops below -50 inches and any of the following exist:

• Power drops below 5% OR

. All MSRVs remain closed and DW pressure remains below 2.4 psig OR

• Water level reaches -162 inches THEN Continue:

CTØ

2-F Page 32 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Terminate and Prevent JAW Appendix 4 Appendix 4

1. PREVENT injection from HPCI by performing the following:

BOP/ATC

a. ll HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 2-HS-73-18A, HPCI TURBINE TRIP push-button.

b. WhEN HPCI Turbine is at zero speed, THEN PLACE 2-HS 47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 2-HS-73-l 8A, 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 LPCJ 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 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-FCV-74-52, RHR SYS I LPCJ 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 RI{R SYSTEM II pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-FCV-74-66, R}IR 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.

2-F Page 33 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP/ATC c!wz Ten inae d1ve Apnd&

Appendix 4 (continued)

c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:

• 2-FCV-3-19, RFP 2A DISCHARGE VALVE

• 2-FCV-3-12, RFP 2B DISCHARGE VALVE

• 2-FCV-3-5, RFP 2C DISCHARGE VALVE

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

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

• 2-HS-3-125A, RFPT 3A TRIP

• 2-HS-3-151A, RFPT 3B TRIP

• 2-HS-3-176A, RFPT 3C TRIP.

SRO WHEN RPV Level drops below -50 inches THEN Continue:

OR WHEN RPV Level has dropped below -50 inches AND Power is below 5% OR Reactor Level reaches -162 inches, THEN Continue:

Directs a Level Band with RCIC and HPCI.

ATC/BOP Maintain Directed Level Band with RCIC, Appendix SC and HPCI, Appendix SD.

2-F Page 34 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Maintain Directed Level Band with RCIC, Appendix SC

3. VERIFY RESET and OPEN 2-FCV-71-9, RCIC TURJ3 TRIP/THROT VALVE RESET.

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

5. OPEN the following valves:

• 2-FCV-71-39, RCIC PUMP iNJECTION VALVE

• 2-FCV-71-34, RCIC PUMP M1N FLOW VALVE

• 2-FCV-71-25, RCIC LUBE OIL COOLING WTR VLV.

6. PLACE 2-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.

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

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

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

c. 2-FCV-7 1-40, RCIC Testable Check Vlv, opens by observing 2-ZI-7 1 -

40A, DISC POSITION, red light illuminated.

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

9. IF BOTH of the following exist? NO

10. ADJUST 2-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller as necessary to control injection.

2-F Page 35 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Maintain Directed Level Band with HPCI, Appendix 5D

4. VERIFY 2-IL-73-18B, HPCI TURBiNE TRIP RX LVL HIGH, amber light extinguished.

5. VERIFY at least one SGTS train in operation.

6. VERIFY 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5,000 gpm.

7. PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in START.

8. PLACE 2-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, handswitch in START.

9. OPEN the following valves:

• 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE

• 2-FCV-73-44, HPCI PUMP INJECTION VALVE.

10. OPEN 2-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, to start HPCI Turbine.

11. CHECK proper HPCI operation by observing the following:

a. HPCI Turbine speed accelerates above 2400 rpm.

b. 2-FCV-73-45, HPCI Testable Check Vlv, opens by observing 2-ZI-73-45A, DISC POSITION, red light illuminated.

c. HPCI flow to RPV stabilizes and is controlled automatically at 5000 gpm.

d. 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds 1200 gpm.

12. VERIFY HPCI Auxiliary Oil Pump stops and the shaft driven oil pump operates properly.

13. WhEN HPCI Auxiliary Oil Pump stops, THEN PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP,_handswitch in AUTO.

14. ADJUST 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

2-F Page 36 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO EOT-l (Power Control)

Monitor and Control Reactor Power.

Will the reactor will remain sub subcritical without boron under all conditions? NO Is the reactor subcritical and No boron has been injected?- NO Verif 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.

SRO fj jipj p&ij)%

ATC/BOP SRO Directs ART Reset Appendix 2.

C#I se(ont pds1Jsmg one or mjre of the fo1Iomjmthods ppeps1l pperdix oiiplet ndI1eldcjoi fçr appendix F 9ompletd aA _afjb2 4!ws-1 ATC iii Sdnfr6d1AW jj&ix 1DndI

2-F Page 37 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC CT#1 çoiodsJ AWenx 1

2. W1TEN P.PS 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 2-9-4, 2-XA-55-4A, Window 1)

• EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 29).

5. DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER SHUTOFF.

6. WhEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ART.

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.

pjy. WREN &sptthed to close Charging W4tr Shutofyif endpp HV-05-0586 cIosed (<ShfftInrrl06 c1os WREN asiced to opqn Chargg Waer Jiutoff, nes and reppIJ 085-0586 ol,en, (cShift>F4mrfrdO6 oeri,

2-F Page 38 of6l Simulator Event Guide:

Event 9 Component: 2A CRD Pump Trip Reports Trip of CR1) Pump 2A and Starts CRD Pump 1B, lAW 2-AOI-85-3

[1] IF operating CR1) pump has failed AND standby CR1) pump is available, THEN PERFORM the following at Panel 2-9-5:

[1.1] PLACE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, in MAN at minimum setting.

[1.2] START associated standby CR1) Pump using one of the following:

• CRD PUMP lB. using 2-HS-85-2A.

[1.3] IF CR1) Pump lB was started, THEN OPEN CR1) PUMP lB DISCH TO U2, using 2-HS-85-8A

[1.4] ADJUST CR1) SYSTEM FLOW CONTROL, 2-FIC-85-1 1, to establish the following conditions:

• CRD CLG WTR HDR DP, 2-PDI-85-18A, approximately 20 psid.

• CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, between 40 and 65 gpm.

[1.5] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, AND PLACE in AUTO or BALANCE.

2-F Page 39 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC

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

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

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 2-SHV-085-0586, CHARGING WATER SHUTOFF (RB NE, El 565 fi).

WUEN dpateW to 1ase C1rgin Water Shutoff,yait 2in4es a1 repqr 2 IV.o85-as86 closed (<SkftF7 ifrdO6 close)

WREN askedo opeChrging Water Shtoff wait 2 minutçs and rçport 2-SHV 085-0586 open. (<$hift>4 mrfrdfló open

2-F Page 40 of 61 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP/ATC CT#1

1. UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A!2B, 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 SQULB VALVE A and B CONTiNUITY blue lights extinguished.

• SLC SQUB VALVE CONTINUITY LOST Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 20).

• 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

• System flow, as indicated by 2-IL-63-1 1, SLC FLOW, red light illuminated 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-LI-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

2-F Page 41 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO ENTER 2-EOI-2, Primary Containment Control EOI-2 (Drywell Temperature)

SRO Monitor and Control DW Temp Below 160°F using available DW Cooling.

Can Drywell Temp Be Maintained Below 160°F? YES -

SRO EOI-2 (Primary Containment Hydrogen)

If PCIS Group 6 isolation exists? YES THEN DIRECTS:

1. Place analyzer isolation bypass keylock switches to bypass.

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.

BOP 1. Place analyzer isolation bypass keylock switches to bypass.

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.

SRO EOI-2 (Suppression Pool Temperature)

Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary (Appendix 17A)

Can Suppression Pool Temperature Be Maintained Below 95°F? NO Operate all available Suppression pool cooling, using only RHR Pumps not required to assure adequate core cooling by continuous injection, Appendix 1 7A.

ATC/BOP Place an RHR System in Pool Cooling, when directed JAW Appendix 1 7A.

SRO Before Suppression Pool Temperature rises to 110°F Continue in EOI-1 RPV Control Can Suppression Pool temperature and level be maintained within a safe area of curve 3? -

YES SRO The Emergency Classification is 1 .2-S.

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

Controls Rods are being inserted Reactor Level is being maintained in directed level band

2-F Page 42 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSJVs SRO EOI-2 (Suppression Pool Level)

Monitor and Control Suppression Pool Level between -1 inch and -6 inches, (Appendix 1 8).

Can Suppression Pool Level be maintained above -6 inches? YES Can Suppression Pool Level be maintained below -1 inch? YES SRO EOI-2 (Primary Containment Pressure)

Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary, (Appendix 12)

SRO Can Primary Containment pressure be maintained below 2.4 psig? YES SRO The Emergency Classification is 1.2-S.

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

Controls Rods are being inserted Reactor Level is being maintained in directed level band

2-F Page 43 of6l Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC Place Suppression Pool Cooling in service, JAW Appendix 1 7A.

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

• PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

• PLACE 2-HS-74-155B, LPCI SYS II OUTBD JNJ 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:

• 2-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

• 2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

• 2-FCV-23-40, RFIR 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-l22(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRI) in MANUAL OVERRIDE.

e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLACE 3-XS-74-121(l29), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.

f. IF 2-FCV-74-53(67), RHR SYS 1(11) LPCI 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11)

LPCI OUTBD INJECT VALVE.

g. OPEN 2-FCV-74-57(71), RHR SYS 1(11) SUPPR CBBR/POOL ISOL VLV.

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

i. THROTTLE 2-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST 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 gpm 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.

2-F Page 44 of 61 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:

EECW Pump A3 is out of service and tagged out.

RFPT B Out of Service Operations/Maintenance for the Shift:

Remove LPRM 8-49B from bypass lAW 2-OI-92B section 6.4.

Once completed adjust load line lAW RCP and 2-GOI-100-12 section 5.0 step 20 and continue power ascension as directed by the RCP.

Units 1 and 3 are at 100% power.

Unusual Conditions/Problem Areas:

None

Ic C

ci) g I>:

m 1m CD C

C, I

1 ft 01 rI C

2-F Page 48 of 61 TR 3.3 INSTRUMENTATION TR 3.3.5 Surveillance Instrumentation LGO3.3.5 The surveillance instrumentation for each parameter in Table 3.35-1 shall be OPERABLE.

APPLICA6ILITY: According to Table 3.3.6-1 NOTE TRM LCO 3.0.4 is not applicable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. I Enter the Condition Immediately channels inoperable, referenced in Table 3.3.5-1 for the channel.

B. As required by B.I Restore required control 7 days Required Action Al room indication channel and referenced in to OPERABLE status.

Table 3.3.5-1.

C. As required by Cl Restore one of the 7 days from Required Action Al required control room discovery of both and referenced in indication channels for redundant channels Table 3,35-1. each associated for one or more parameter to OPERABLE associated status, parameters not indicating in the NI2 control room C.2 Restore required control 30 days room indication channels to OPERABLE status.

(continued)

2-F Page 49 of 61 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. As required by 0.1 Monitor torus Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action Al temperature to obsere and referenced in any unexplained Table 135-1. temperature increase which might be indicative of an open relief valve.

AND D.2 Restore control room 30 days indication by either the Tailpipe Thermocouple Temperature or Acoustic Monitor to OPERABLE status for each relief valve.

AND 0.3 When inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> more than 30 days, initiate a Problem Evaluation Report (PER).

(continued)

2-F Page 50 of6l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by NOTE Required Action Al Required Actions Eli and and referenced in E12 are not applicable when in Table 3.3.5-1, MODES 4 and 5.

E.1.1 Restore required control 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> room indication channel to OPERABLE status.

R El .2 Initiate the preplanned 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> alternate method of monitoring the parameter.

AND E.2 When inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> more than seven days, initiate a Problem Evaluation Report (PER).

(continued)

2-F Page 51 of 61 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and Fl Be in MODE 4. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> associated Completion Time of Condition B or D not met.

OR Required Action and associated Completion Time of Condition C not met for Instruments Ia or 3.b.

S. Required Action and G.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C not met for Instruments 2.a, 2.b, 4a, or4,b.

H. Required Action and Hi Reduce THERMAL 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion POWER to 15% RTP, Time of Condition C not met for Instrument 5 channels.

2-F Page 52 of 61 TABLE 1351 (page 1 of 2)

Surveillance Instrumentation PARAMETER AND APPLJCALE REQUIRED CONDITIONS TECHNiCAL TYPE INSTRUMENTS MODES OR CHANNELS REFERENCED SURVEILLANCE INDtCATION OTHER FROM REQUIREMENTS AND RANGE SPECIFIED REDUIRED CONDITIONS ACTION A. I

1. SUPPMEsicn 123 1 8 TSR 3.a&1 Recorder ChamberAlr TSR 3.35.8 040F Temperature (XR4452)

2. Contrd Rod Motion

a. Contrd Rod 12 1(b) C TSR 3.3.52 indicatora 00-48 Pcoition (a)

b. Neuron 1.2 c) C TSR 3.3.5.3 $RM Indicalora Monitoring (a) TSR 3.3.5.4 0.1-10 cpe IRM TSR 3.3.5.7 Indicatora 0425 TSR 3.3.5.8 LPRM Indicatora TSR 3.3.5.0 0-125

3. Dr,well Prewref Temp ature Atam

a. Drywell Prure 1.2.3 1 C TSR 13.5.14 Alarm at 35 peig (P544478) (d)

I,. Drywell 1,23 I C TSR 3.3.5.10 Alarm if temp.

Tperature and TSR 3.3.5.13 > 281F and Preesure and preeaure> 2.5 Time pe.i after 30 (TS-8442A and nvnute delay P1844-58A and 134447A) (d)

(continued)

(a) The channel of Control Rod Position instruments and the channel of Neutron Monitoring instruments are considered redundant to each other for the parameter of Control Rod Motion.

(h) The Control Rod Position channel consists of full core display position indicators or four-rod display position indicators capable of determining position of all OPERABLE control rods. Position indicators are considered to be capable of determining rod position when they display the rod position or the rod can be moved to a position where rod position Is displayed.

(c) The Neutron Monitoring channel contains the following:

1. In MODE 2 with IRMs on Range 2 or below a minimum of 3 OPERABLE channels of SRMs..

2. In MODE 2 a minimum of OPERABLE channels of IRMs.

3. In MODES I and 2, 43 LPRM detector assemblies, each containing four fission chambers, lndMdual farled chambers can be bypassed to the extent that APRMS remain OPERABLE.

(d} The channel of Orywell Pressure and the channel of Drywell Temperature and Pressure and Timer instruments are considered redundant to each other for the parameter of Drywell Pressurerremnperature Alarm.

2-F Page 53 of6l 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

• NOTE Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control NOTE rod stuck. Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1. ControI Rod Block lnstrumeritation if required, to allow continued operation.

A.1 Verify stuck control rod Immediately separation criteria are met.

AND A.2 Disarm the associated 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> control rod drive (CRD).

AND (continued)

2-F Page 54 of6l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Perform SR 3.1 .3.2 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from SR 3.1.3.3 for each discovery of withdrawn OPERABLE Condition A control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A.4 Perform SR 3,tl.1, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Two or more withdrawn 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> control rods stuck.

C. One or more control rods C.1 NOTE inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control rod.

AND C.2 Disarm the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> CRD.

(continued)

2-F Page 55 of 61 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. NOTE ---- D1 Restore compliance with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Not applicable when BPWS.

THERMAL POWER

>1O%RTP.

D.2 Restore control rod to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Two or more inoperable OPERABLE status control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, C, or D not met.

OR Nine or more control rods mope rable,

2-F Page 56 of6l 31 PLANT SYSTEMS 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, APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required EECW A,I Restore the required 7 days pump inoperable. EECW pump to OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AN2 met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Two or mare required EECW pumps inoperable.

OR UHS inaperabIe

2-F Page 57 of 61 3.6 CONTAiNMENT SYSTEMS 3.61.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PC1V, except reactor buiIdingto-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, ?rimary Containment Isolation Instrumentation.

ACTIONS

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

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

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

4. Enter applicable Conditions and Required Actions of LCO 3,6.1.1, Primary Containment, when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE Al Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one dosed with two PCIVs. and de-activated AND

- automatic valve, closed manual valve, blind 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured.

MSIV leakage not within limits.

AND (continued)

2-F Page 58 of6l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A2 --NOTE isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is for isolation isolated. devices outside primary containment AND Prior to entering MODE 2or3from 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 (continued)

2-F Page 59 of6l ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 8 --NOTE 81 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic valve, closed manual valve, or blind One or more penetration flange.

flow paths with two PCIVs inoperable except due to MSIV leakage not within limits.

C. NOTE CI Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and da-activated

- automatic valve, closed manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for flow paths with one PCIV EECVs inoperable. N2 Cl NOTE Isolation devices in high radiation areas may be verified by use of administrative means Verify the affected Once per 31 days penetration flow path is isolated.

(continued)

2-F Page 60 of6l ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration Di Restore leakage rate to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flow paths with MSIV within limit.

leakage not within limits.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, 8, C, AN.Q.

or D not met in MODE 1, 2,or3. E.2 Be n MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> F. Required Action and Fi InitIate action to suspend Immediately associated Completion operations with a Time of Condition A, B, C, potential for draining the or D not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during MODE °r F.2 NOTE Only applicable for inoperable RI-IR Shutdown Cooling Valves.

Initiate action to restore Immediately valve(s) to OPERABLE status.

2-F Page 61 of6l SCRAM FAILURE REACTOR COOLANT ACTMTY flaaarin+ien I II I 1.3-U I I I I Reactor coolant actMty exceeds 26 pCilgm dose equivalent 1-131 (Technical Specification Limits) Z as determined by chemistiy sample.

r OPERATING CONDON in

-4 1.2-Al INOTEI 1.3-A I I I I Failure of RPS automatic scram functions to bnn Reactor coolant activity exceeds 300 iCi!gm dose the reactor subcritical equivalent Iodine-I 31 as detemiined by cheniislxy AND sample.

r Manual scram or ARt (automatic or manual) was m succesofti OPERATING CONDITION:

OPERA11NG CONDITION: Model or2or3 Model or 2 1.2-SI INOTEI I I I I Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical.

m m

OPERATING CONDfflON:

Model 1.2-G ICIJRVEI I I US I I I I Failure of automatic scram, manual scram, and ARL Reactor power is above 3% C)

AND Either of the following condthons exists:

. Suppression Pool temp exceeds HCTL. m Refer to Curve l.2-G.

. Reactor water level can NOT be restored and maintained at or above -180 inches. C) m z

OPERATING CONDITION:

Mode I or 2