Facility: Browns Ferry NPP Scenario No.: 3-24 Op-Test No ... · • Inject boron (prior to...

Appendix D Scenario Outline Form ES-D-1

Facility: Browns Ferry NPP Scenario No.: 3-24 Op-Test No.: ILT 1006

Examiners:

Operators:

SRO:

ATC:

BOP:

Initial Conditions:

Unit 3 is at 5 to 6% power with startup is in progress per 3-GOI-100-1A. IRM E is out of service. Maintenance has identified that EHC Pump A motor is operating hotter than normal. Unit 1 and 2 are at 100% power. A Severe Thunder Storm Warning has been issued for Limestone County for the next 4 hours.

Turnover: Complete EHC Auto Pump Start Test and Weekly Pump Alternation per 3-OI-47A. At step

67 to raise reactor power with rods to 8%. Continue unit startup to Mode 1.

Event No. Malf. No.

Event Type* Event Description

1 N-BOP N-SRO

Weekly EHC Pump Test

2 R-ATC R-SRO

Raise power with rods to 8%

3 rd05 TS-SRO C-ATC

Uncoupled Control Rod

4 ms01 C-BOP C-SRO

Steam Seal Regulator failure

5 hp08 TS-SRO C-BOP

HPCI Steam leak w/ auto isolation failure

6 rfpt c C-ATC C-SRO

RFPT C trips

7 vacuum loss M-All Loss of vacuum causes a loss of feedwater and pressure control

8 loca M-All Loss of High pressure injection with a LOCA, ATWS, and TAF

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


Critical Tasks

1. ADS inhibited before ADS actuates

2. With a reactor scram required and the reactor not shutdown: • Inserts control rods, to prevent exceeding the primary

containment design limits • Inject boron (prior to exceeding 110°F SP temperature)

3. During an ATWS with emergency depressurization required,

injection has been stopped and prevented, with the exception of RCIC, SLC, and CRD, into the RPV until reactor pressure is below Minimum Alternate RPV Flooding Pressure(MARFP)

4. Emergency depressurizes when RPV level cannot be restored and

maintained above -180". Emergency depressurization is initiated before level reaches -200"

5. When RPV pressure is below MARFP, injection reestablished to maintain RPV level above -180”

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With the unit at 5 to 6% power, the BOP operator will complete the weekly EHC Pump Test per 3-OI-47A section 6.2. Upon completion the ATC will commence power increase. The second rod to 48 will fail the coupling check. When actions are taken to couple the rod per 3-AOI-85-2, the rod will couple and power ascension may continue. SRO should refer to Tech Spec 3.1.3 for uncoupled rod. Follow up question after scenario for Tech Spec call if rod failed to couple. Tech Spec 3.1.3 Condition C. The Steam Seal Regulator will fail causing a loss of sealing steam. Off gas flow will increase and vacuum will slowly degrade. The BOP operator can restore seal steam by taking manual control of the steam seal regulator bypass valve. Off gas flow will slowly return to normal and vacuum will quickly recover. HPCI steam leak will develop and HPCI will fail to auto isolate. The BOP operator will isolate HPCI and stop the steam leak. EOI-3 will be entered on HPCI room temperature. With HPCI Inoperable, the SRO will evaluate Tech Spec 3.5.1 and enter condition C and Tech Spec 3.6.1.3 Condition B. With HPCI Inoperable, mode change can not be made. After failure of HPCI, RFPT C will trip. The ATC will take control of RFPT B and raise discharge pressure in order to restore level to the normal operating band or he will need to insert a manual reactor scram if unable to recover level. RFPT B is idling at 600 RPM at the start of the scenario; speed will need to be raised to about 3400 RPM for injection to the reactor. Once the level is stable on RFPT B or if the reactor is scrammed, a lighting strike at the CCW Capacitor Banks will trip all running Circ Water Pumps and create a loss of condenser vacuum. The vacuum loss will trip the RFPT and isolate bypass valves. The crew will attempt to scram the reactor but an electrical ATWS will exist and no rod motion will occur. Pressure control will be on SRVs and rods can be inserted manually per appendix 1D. When RCIC is started it will trip and will be unavailable. Level will lower to TAF, the SRO will determine that Reactor Level can not be restored and maintained above -180 inches. Once this determination is made the crew will transition to ED. Injection sources will be terminated and prevented and 6 ADS valves will be opened. Once Reactor Pressure lowers to the MARFP pressure of 190 psig, injection can commence with low pressure systems to restore Reactor level to the prescribed level band.

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VIII. Console Operator Instructions A. Scenario File Summary 1. File: batch and trigger files for scenario 3-24 Batch NRC201024 #steam seal failure imf ms01 (e1 0) imf mc04 (e1 0) 100 #uncoupled control rod imf rd05r2659 #rfpt c trip ior zdihs03176[1] (e10 0) trip #hpci steam leak ior xa554c[10] alarm_off imf hp09 imf hp08 (e5 0) 10 120 #loss of vacuum imf mc02a (e30 0) imf mc02b (e30 0) imf mc02c (e30 0) imf rp07 (e30 0) imf rp14a (e30 0) imf rp14b (e30 0) ior zdihs661a[3] (e30 0) open ior zlohs661a[2] (e30 0) off imf hp04 ior zdihs7316a[2] auto ior zdihs719a[1] (e29 0) trip trg e29 nrc2010718open imf th22 (e30 30) 100 imf th21 (e30 150) 0.8 600 imf rc08 #rps fuses out and in trg 20 = bat nrc201024rpsout trg 21 = bat nrc201024rpsin

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Batch NRC201024rpsout mrf rp05a (none 60) out mrf rp05c (none 120) out mrf rp05e (none 180) out mrf rp05g (none 240) out mrf rp05b (none 300) out mrf rp05d (none 360) out mrf rp05f (none 420) out mrf rp05h (none 480) out Batch NRC201024rpsout mrf rp05a (none 0) in mrf rp05c (none 30) in mrf rp05e (none 60) in mrf rp05g (none 90) in mrf rp05b (none 120) in mrf rp05d (none 150) in mrf rp05f (none 180) in mrf rp05h (none 210) in RCIC Start Trigger zdihs718a[2].ne.1 IX.

Console Operators Instructions

Scenario 3-24

DESCRIPTION/ACTION

Simulator Setup manual Reset to IC 183 Simulator Setup Load Batch Bat nrc201024 Delete the ATWS from trigger 30 Simulator Setup manual Bypass IRM E Simulator Setup Verify Batch file loaded

RCP required (Start up) – Provide marked up copy of 3-GOI-100-1A

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X. Scenario Summary: With the unit at 5 to 6% power, the BOP operator will complete the weekly EHC Pump Test per 3-OI-47A section 6.2. Upon completion the ATC will commence power increase. The second rod to 48 will fail the coupling check. When actions are taken to couple the rod per 3-AOI-85-2, the rod will couple and power ascension may continue. SRO should refer to Tech Spec 3.1.3 for uncoupled rod. Follow up question after scenario for Tech Spec call if rod failed to couple. Tech Spec 3.1.3 Condition C. The Steam Seal Regulator will fail causing a loss of sealing steam. Off gas flow will increase and vacuum will slowly degrade. The BOP operator can restore seal steam by taking manual control of the steam seal regulator bypass valve. Off gas flow will slowly return to normal and vacuum will quickly recover. HPCI steam leak will develop and HPCI will fail to auto isolate. The BOP operator will isolate HPCI and stop the steam leak. EOI-3 will be entered on HPCI room temperature. With HPCI Inoperable, the SRO will evaluate Tech Spec 3.5.1 and enter condition C and Tech Spec 3.6.1.3 Condition B. With HPCI Inoperable, mode change can not be made. After failure of HPCI, RFPT C will trip. The ATC will take control of RFPT B and raise discharge pressure in order to restore level to the normal operating band or he will need to insert a manual reactor scram if unable to recover level. RFPT B is idling at 600 RPM at the start of the scenario; speed will need to be raised to about 3400 RPM for injection to the reactor. Once the level is stable on RFPT B or if the reactor is scrammed, a lighting strike at the CCW Capacitor Banks will trip all running Circ Water Pumps and create a loss of condenser vacuum. The vacuum loss will trip the RFPT and isolate bypass valves. The crew will attempt to scram the reactor but an electrical ATWS will exist and no rod motion will occur. Pressure control will be on SRVs and rods can be inserted manually per appendix 1D. When RCIC is started it will trip and will be unavailable. Level will lower to TAF, the SRO will determine that Reactor Level can not be restored and maintained above -180 inches. Once this determination is made the crew will transition to ED. Injection sources will be terminated and prevented and 6 ADS valves will be opened. Once Reactor Pressure lowers to the MARFP pressure of 190 psig, injection can commence with low pressure systems to restore Reactor level to the prescribed level band.

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

Control Rod insertion in progress Emergency Depressurization Reactor Water Level restored after ED

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Event 1: EHC Pump Test

SRO Direct BOP to perform EHC Auto Pump Start Test & Weekly Pump Alternation

BOP Perform EHC Auto Pump Start Test & Weekly Pump Alternation

[1] VERIFY the EHC System is in service. REFER TO Section 5.1. [2] REVIEW Precautions and Limitations listed in Section 3.0. [3] DEPRESS the EHC PUMP 3A(3B) TEST, 3-HS-47-4A(5A), and CHECK the following actions occur: • EHC Hydraulic Fluid Pump 3A(3B) starts. • Annunciator STANDBY EHC PUMP RUNNING, 3-XA-47-108 (3-XA-55-7B Window 8), ANNUNCIATES. • Red light above test switch is ILLUMINATED (PS-47-1B(2B)) (positive indication of pump discharge pressure). [4] CHECK the started EHC HYD PUMP A(B) DISCH PRESS, 3-PI-047-0001(0002), indicates between 1550 psig and 1750 psig, locally at the EHC skid.

Driver Driver Pre start walk down complete Pump ready for start Local Pressure is 1600 psig Volts are 0 Volts

BOP [5] IF the started EHC pump discharge pressure is NOT between 1550 psig and 1750 psig, THEN [6] CHECK locally on Junction Box 3-JBOX-047-10166 that the following voltmeters indicate approximately 0 volts: • TURBINE EHC HDR PRESS 3-PS-47-63A TRIP IND, 3-EI-047-0063A • TURBINE EHC HDR PRESS 3-PS-47-63B TRIP IND, 3-EI-047-0063B • TURBINE EHC HDR PRESS 3-PS-47-63C TRIP IND, 3-EI-047-0063C [7] IF alternating operating EHC pumps, THEN STOP EHC pump 3B(3A) using EHC HYD FLUID PUMP 3B(3A), 3-HS-47-2A(1A). [8] IF NOT alternating EHC pumps, THEN

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XII. Simulator Event Guide: Event 1: EHC Pump Test

BOP [9] IF the pump stopped in Step 6.2[7] or 6.2[8] fails to stop or remain stopped, THEN [10] CHECK EHC HEADER PRESSURE, 3-PI-47-7, indicates between 1550 psig and 1650 psig. [11] IF EHC header pressure is NOT between 1550 psig and 1650 psig, THEN [12] IF pumps were alternated, THEN RESET any disagreement flags by placing the operating EHC pump handswitch, 3-HS-47-1A(2A), to START.

[13] CHECK EHC PMP 3A(3B) MTR AMPS, 3-EI-47-1(2), indicates less than 140 amps for the operating pump. [14] VERIFY EHC Pump 3A(3B), 3A(3B) EHC PMP SUCT FLT, 3-FLT-047-0229(0230) and 3A(3B) EHC PMP AUX FLT, 3-FLT-047-0217(0218), indicate CLEAN, locally at the EHC skid. [15] VERIFY MANUAL FLOW CONTROL VALVE, 3-SHV-047-0588, is set at 2.0 turns clockwise.

Driver Driver Filter indicates Clean, Flow control valve is 2 turns open Volts are 0 Volts

[16] CHECK locally on Junction Box 3-JBOX-047-10166 that the following voltmeters indicate approximately 0 volts: • TURBINE EHC HDR PRESS 3-PS-47-63A TRIP IND, 3-EI-047-0063A • TURBINE EHC HDR PRESS 3-PS-47-63B TRIP IND, 3-EI-047-0063B • TURBINE EHC HDR PRESS 3-PS-47-63C TRIP IND, 3-EI-047-0063C

Driver Driver Post Start Walk down complete, No Leaks Pump running normally

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XII. Simulator Event Guide: Event 2: Unit Startup

SRO 3-GOI-100-1A Section 5.0 Instruction Steps [67] CONTINUE to withdraw control rods to raise Reactor power to approximately 8%. (REFER TO 3-OI-85 and 3-SR-3.1.3.5(A)) Directs ATC to continue power ascension with rods

ATC 3-OI-85 6.6 Control Rod Withdrawal 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 Supervisor’s 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.

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ATC 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

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing. [4] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE. [5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes. [6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows: [6.1] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE. [6.2] CHECK control rod coupled by observing the following: • Four rod display digital readout and the full core display digital readout and background light remain illuminated. • CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm. [6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes. [6.4] IF Control Rod Coupling Integrity Check fails, THEN REFER TO 3-AOI-85-2.

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ATC 6.6.4 Continuous Rod 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

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the 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, 3-HS-85-47, in NOTCH OVERRRIDE. [5.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH. [5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47 and CRD CONTROL SWITCH, 3-HS-85-48. [5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE. [5.5] WHEN control rod settles into the intended notch, THEN CHECK the following. • Four rod display digital readout and the full core display digital readout and background light remain illuminated. • CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does NOT alarm. [5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

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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. [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.

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ATC 6.6.4 Continuous Rod Withdrawal (continued) [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. [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.

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SRO 3-GOI-100-1A Section 5.0 Instruction Steps [68] VERIFY all operable APRM downscale alarms are reset and no rod blocks exist.

[69] VERIFY the following: • Hotwell Pressure is below -24" Hg. • CONDENSER A, B OR C VACUUM LOW annunciator, (3-XA-55-7B, window 17) is reset on Panel 3-9-7. [70] VERIFY all operable MSIVs are open on Panel 3-9-3. [71] IF primary containment purge and/or Primary Containment Ventilation is in service, THEN PLACE the following switches in the BYPASS position (Panel 3-9-3): • PC PURGE DIV I RUN MODE BYPASS, 3-HS-64-24. • PC PURGE DIV II RUN MODE BYPASS, 3-HS-64-25. [72] IF Recirculation System is in Single Loop Operation, THEN (Otherwise N/A) VERIFY that 3-SR-3.4.1(SLO) is completed to satisfy Tech Specs and SR-3.0.4. Initials Date Time

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XII. Simulator Event Guide: Event 3: Uncoupled Control Rod

ATC Respond to Annunciator CONTROL ROD OVERTRAVEL and CONTROL ROD DRIFT A. VALIDATE alarm as follows: 1. Full Core Display will have no digital readout. 2. Background light extinguished. 3. Rod DRIFT light on. B. IF alarm valid, THEN REFER TO 2-AOI-85-2. C. NOTIFY Reactor Engineer. D. REFER TO Tech Spec 3.1.3, 3.10.8.5, 3.3.2.1, and TRM TABLE 3.3.5-1. Reports Control Rod 26-59 is Uncoupled or over traveled

SRO Enter 2-AOI-85-2 Uncoupled Control Rod

ATC 3-AOI-85-2 4.1 Immediate Actions [1] STOP all control rod withdrawal.

SRO 4.2 Subsequent Actions [1] NOTIFY Reactor Engineer to evaluate the suspect uncoupled control rod for its impact on core thermal limits and rod worth. [2] ADJUST the rod pattern as directed by the Reactor Engineer throughout the performance of this procedure.

Driver Driver Acknowledge uncoupled control rod, concur with coupling attempt per 3-AOI-85-2

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XII. Simulator Event Guide: Event 3: Uncoupled Control Rod

SRO [3] IF the control rod drive is at position 48 AND with Reactor Engineer concurrence, THEN PERFORM the following: Direct ATC to attempt to couple Control Rod 26-59

ATC

PERFORM the following: [3.1] NOTCH INSERT the control rod drive to position 46 to attempt to couple the control rod. [3.2] RESET associated annunciators. [3.3] NOTCH WITHDRAW the control rod drive to position 48. [3.4] PERFORM a coupling check. [3.5] IF coupling integrity check fails, THEN CONTINUOUSLY INSERT control rod drive to position 00 to attempt to latch control rod with control rod drive mechanism. [3.5.1] RESET associated annunciators. [3.5.2] NOTCH WITHDRAW control rod to position 48. [3.5.3] PERFORM a coupling check. Report Control Rod 26-59 is Coupled

Driver Driver When control rod 26-59 is inserted to position 46 delete the failure rd05r2659

SRO Makes notifications that Control Rod 26-59 is Coupled

SRO Exits 3-AOI-85-2 and directs power ascension to continue

NRC NRC At completion of scenario if SRO does not address Tech Spec 3.1.3 Follow up for Tech Spec call if rod failed to couple TS 3.1.3 Condition C, C.1 Fully insert inoperable control rod in 3 hours AND disarm the associated CRD within 4 hours

Driver Driver Upon Lead examiner direction, initiate Trigger 1 for Steam Seal Regulator Failure

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XII. Simulator Event Guide: Event 4: Steam Seal Regulator Failure

BOP Respond to Annunciator STEAM TO STEAM SEAL REG PRESS LOW A. CHECK steam seal header pressure, 3-PI-1-148, 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-1-145. D. DISPATCH personnel to check 3-PIC-1-147 (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 Reports Condenser Vacuum stable or slowly degrading

Driver Driver 3 minutes after dispatch report No noticeable problems at 3-PIC-1-147, once steam seal pressure is restored delete mc04 (e1 0) 100 condenser air in leakage

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.

SRO Direct Power Ascension to continue and contacts personnel to investigate Steam Seal Regulator

Driver Driver If I&C is dispatched acknowledge communication

Driver Driver Upon Lead examiner direction, initiate Trigger 5 for HPCI Steam Leak

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XII. Simulator Event Guide: Event 5: HPCI Steam Leak

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

BOP

Respond to annunciator HPCI LEAK DETECTION TEMP HIGH Continued rise will cause the following valves to isolate (at Steamline Space Temperature of 165°F Torus Area or 185°F HPCI Pump Room): HPCI STEAM LINE INBD ISOL VALVE, 3-FCV-73-2 HPCI STEAM LINE OUTBD ISOL VALVE, 3-FCV-73-3 HPCI STEAM LINE WARM-UP VALVE, 3-FCV-73-81. 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. D. DISPATCH personnel to investigate for leaks consistent with ALARA considerations in HPCI Turbine Area (El 519) and HPCI Steam Supply Area (El 550).

BOP Reports rising temperature in HPCI, reports HPCI failed to isolate and isolates HPCI Steam Line

SRO Enter EOI-3 on Secondary Containment Area Radiation

Driver Driver If dispatched to HPCI area report after 5 minutes that can not access area at this time.

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SRO EOI-3 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

BOP Isolates HPCI Steam Lines and reports Temperatures and Radiation Levels lowering

SRO EOI-3 Secondary Containment Radiation

Monitor and Control Secondary Containment Radiation Levels

Answers Yes to Is Any Area Radiation Level Max Normal

Isolate all systems that are discharging into the area except systems required to: • Be operated by EOIs OR • Suppress a Fire

SRO Ensures no systems are still discharging to Secondary Containment, remains in EOI-3 until entry conditions are cleared.

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SRO EOI-3 Secondary Containment Level

Monitor and Control Secondary Containment Water Levels

Answers No to Is Any Floor Drain Sump Above 66 inches and Answer No to Is Any Area Water Level Above 2 inches

SRO Directs Power Ascension stopped

Evaluates Tech Spec 3.5.1 and 3.6.1.3

Tech Spec 3.5.1 enter Condition C, C.1 Verify by administrative means RCIC is OPERABLE Immediately and C.2 Restore HPCI System to operable status in 14 Days. Tech Spec 3.6.1.3 and Enter Condition A and B, B Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve or blind flange 1 Hour.

Notifies Operations management that Mode Change can not be completed

Driver Driver Upon Lead examiner direction, initiate Trigger 10 for RFPT C Trip

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XII. Simulator Event Guide: Event 6: RFPT C Trip

ATC Report RFPT C Trip

SRO Enters 3-AOI-3-1 Loss of Reactor Feedwater

ATC Actions for Low Reactor Water Level or Loss of Feedwater [10] IF RFPs are in manual control, THEN RAISE speed of operating RFPs. [14] IF unit remains on-line, THEN RETURN Reactor water level to normal operating level of 33” (normal range).

SRO Direct RFPT B used to restore level or directs a Reactor Scram if level cannot be restored

ATC Raises speed of RFPT B to restore Reactor Level

BOP Dispatches personnel to RFPT C area

SRO Establishes a Reactor Level at which a scram will be inserted, automatic scram is plus 2 inches.

Driver Driver 4 minutes after dispatched report workers in area accidentally tripped RFPT C

ATC Reports that level is recovering when RFPT B is feeding the vessel

Report Power level and SRM alarms as level is restored

ATC Control speed of RFPT B to maintain Reactor Level

Driver Driver If the crew inserts a scram or level reaches level 3 then initiate trigger 30 for lighting strike at CCW Capacitor Banks, Call control room as security and report lighting strike at CCW Capacitor Banks

Driver Driver Upon Lead examiner direction, initiate Trigger 30 for lighting strike at CCW Capacitor Banks, Call control room as security and report lighting strike at CCW Capacitor Banks

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XII. Simulator Event Guide: Event 7: Loss of Condenser Vacuum / ATWS

Crew Responds to trip of operating Circ Water Pumps and Degrading Vacuum

SRO Enters 3-AOI-47-3 Loss of Vacuum

SRO [6] CHECK CCW pumps for proper operation. If available, START additional CCW PUMPS. [7] VERIFY CLOSED CONDENSER VAC BREAKERS 1A AND 1B, 3-HS-66-1A, Panel 9-8. Directs BOP to start CCW Pump A and verify vacuum breakers closed

BOP Attempts start of CCW Pump A and reports trip

SRO Directs Reactor Scram

ATC Manually scrams Reactor

ATC Reports no control rods movement, initiates one channel of ARI

ATC Places mode switch in shutdown and reports power above 5%

SRO Enters EOI-1 RPV Control, Power Directs second channel of ARI

Answers Yes to is Power Above 5%

Directs ATC to Trip Recirc Pumps

ATC Initiates second channel of ARI and trips Recirc Pumps and reports power

CT#2 SRO Directs Appendix 1A, 1D and 2

ATC Contacts AUO to perform field actions of appendix 1A, 1D and 2

Driver Driver When Notified to perform appendix 1A and 2, wait 2 minutes initiate trigger 20 and verify batch file for removing fuses is active. If requested to open/close 3-shv-085-0586 wait 3 minutes to close or open and use remote function rd06.

Driver Driver When requested to install fuses initiate trigger 21 and verify batch file goes active

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XII. Simulator Event Guide: Event 7: ATWS

ATC Appendix 1D to manually insert rods

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 (RB NE, El 565 ft). 3. VERIFY REACTOR MODE SWITCH in SHUTDOWN. 4. BYPASS Rod Worth Minimizer. 5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:

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

CT#2 ATC Inserts Control Rods

SRO Stops at Stop sign Before Suppression Pool Temperature Rises to 110°F Continue

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SRO EOI-1 Pressure Monitor and Control RPV Pressure Answer No to is any MSRV cycling Directs RPV pressure band of 800 to 1000 psig, initially on bypass valves then on

MSRVs

BOP Maintains RPV pressure band of 800 to 1000 psig

SRO EOI-1 Level Monitor and Control RPV Water Level Directs Verification of PCIS isolations

ATC/BOP

Verifies PCIS isolations

SRO Exits RC/L and Enters C5 Level Power Control

CT#1 SRO Directs ADS Inhibited

CT#1 BOP/ATC Inhibits ADS

SRO Answers Yes to Is Any Main Steam Line Open

Bypass the following isolation interlocks • MSIV Low Low Low RPV Water Level (Appendix 8A) • RB Ventilation Low RPV Water Level (Appendix 8E)

ATC/BOP Contacts AUO to perform Appendix 8A and 8E

Answer Yes to Is Reactor Power Above 5%

Driver Driver If requested to perform Appendix 8A and 8E wait 5 minutes insert Bat app08ae and report field actions for appendix 8A are complete and appendix 8E is complete

ATC/BOP Complete Appendix 8A 3. NOTIFY Unit Operator to verify closed the following valves (Unit 3 Control Room, Panel 9-3): 3-FCV-43-13, RX RECIRC SAMPLE INBD ISOLATION VLV 3-FCV-43-14, RX RECIRC SAMPLE OUTBD ISOLATION VLV

SRO Directs Stop and Prevent All Injection Into RPV Except From RCIC, CRD and SLC (Appendix 4)

3-24 Rev 0 Page 26 of 53


3-24 Rev 0 Page 27 of 53

BOP/ATC Terminate and Prevent per Appendix 4

1. PREVENT injection to RPV from the following systems in any order as required: a. HPCI 1) IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURBINE TRIP push-button. 2) 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. c. CORE SPRAY NOTE: After receipt of an automatic initiation signal, it is NOT necessary to wait until the pump actually starts before performing the next step. Following an initiation signal, PLACE ALL Core Spray pump control switches in STOP.

BOP/ATC

d. LPCI SYSTEM I PREVENT injection by EITHER of the following methods: • Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP. OR • BEFORE RPV pressure drops below 450 psig, 1) PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL, in BYPASS. AND 2) VERIFY CLOSED 3-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE. e. LPCI SYSTEM II PREVENT injection by EITHER of the following methods: • Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP. OR • 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.

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XII. Simulator Event Guide: Event 7: EOI-3

BOP/ATC Terminate and Prevent per Appendix 4

f. CONDENSATE and FEEDWATER 1) 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 PDS, in MANUAL AND individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL PDS in AUTO, OR • Using individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL PDS in MANUAL, OR • Using individual 3-HS-46-8A(9A)(10A), RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR. 2) 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. 3) 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.

SRO When RPV Water Level Drops Below -50 inches Then Continue Directs a RPV Level Band of -50 to -100 inches using RCIC and CRD

ATC/BOP

Reports RCIC failed to auto start and attempts to start RCIC Report failure of RCIC and dispatch personnel

Driver Driver When dispatched to RCIC wait 5 minutes and report Turbine is tripped and tripped linkage can not be reset

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XII. Simulator Event Guide: Event 8: Loss of High Pressure Injection

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Crew Monitor Reactor Water Level

CT#2 SRO Before Suppression Pool Temperature rises to 110°F directs Boron Injection

CT#2 ATC Initiates SLC per Appendix 3A

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-63-11, 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). SRO Direct Verification of RWCU Isolation

ATC/BOP Verifies RWCU isolated

SRO If RPV Water Level Cannot Be Restored And Maintained Above -180 inches Then Continue at Step C5-19

CT#4 Prior to -180 inches transitions to Emergency Depressurization

CT#3 Directs Stop and Prevent All Injection Into RPV Except From RCIC, CRD and SLC (Appendix 4)

CT#3 BOP/ATC

Confirms Appendix 4 Injection Systems are stopped and prevented

SRO Answers Yes to Is Suppression Pool Level Above 5.5 feet

CT#4 Directs All ADS Valves Open

CT#4 ATC/BOP

Opens 6 ADS Valves

SRO Answers Yes to Can 6 ADS Valves Be Opened

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XII. Simulator Event Guide: Event 8: Loss of High Pressure Injection

CS#5 SRO When Reactor Pressure is below 190 psig (MARFP Pressure) Then Continue

Direct Start and Slowly Raise RPV Injection with the following injection sources to restore and maintain RPV water level above -180 inches Condensate and Feedwater, CRD, LPCI

CT#5 ATC/BOP Restore Level using Condensate and/or LPCI to the prescribed level band

SRO Answers Yes to Can RPV Water Level Be Restored And Maintained Above -180 inches

Answers Yes to Was Water Level Lowered For Step C5-12

Directs the same level band that was in use prior to ED

ATC/BOP Restore Level using Condensate and/or LPCI to the prescribed level band

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XII. Simulator Event Guide: Event 8: EOI-2

SRO Enters EOI-2 Primary Containment Control

EOI-2 Drywell Temperature

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

Answers No to Can Drywell Temp Be Maintained Below 160°F

Operate All Available Drywell Cooling

Stops at Stop sign Before Drywell Temperature Rises to 280°F Then Continue

SRO EOI-2 Primary Containment Pressure

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

Directs venting of Primary Containment per Appendix 12

Answers No to Can PC Pressure Be Maintained Below 2.4 psig

Evaluates stop sign Before Suppression Chamber Pressure Rises To 12 psig Continue

Initiates Suppression Chamber Sprays Using Only Pumps Not Required to Assure Adequate Core Cooling By Continuous Injection (Appendix 17C)

SRO Should not initiate sprays or terminates sprays prior to ED

SRO EOI-2 Suppression Pool Level

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

Answers Yes to Can Suppression Pool Level Be Maintained Above -6 inches Answers Yes to Can Suppression Pool Level Be Maintained Below -1 inches

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XII. Simulator Event Guide:

Event 8: EOI-2

SRO EOI-2 Suppression Pool Temperature

Monitor and Control Suppression Pool Temperature Below 95°F Using Available Suppression Pool Cooling As Necessary (Appendix 17A)

Answers No to Can Suppression Pool Temperature Be Maintained Below 95°F

Operate all available Suppression Pool Cooling using only RHR Pumps not required to assure adequate core cooling by continuous injection (Appendix 17A)

SRO Should not initiate Pool Cooling or terminates pool cooling prior to ED

SRO EOI-2 Directs H2O2 Analyzers placed in service

BOP Places H2O2 analyzers in service: 1.Place Analyzer Isolation Bypass Switch to Bypass 2. Select DW or SUPPR CHMBR and pulls out select switch handle to start pumps.

SRO Emergency Plan Classification is 1.2-S

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Event 8: EOI-2

BOP Vents the suppression chamber per Appendix 12

1. 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 INBD ISOL VALVE.

NOTE: Venting may be accomplished using EITHER: 3-FIC-84-19, PATH B VENT FLOW CONT, OR 3-FIC-84-20, PATH A VENT FLOW CONT.

Driver Driver If requested to start a Standby Gas Train After 1 minute, Remote function PC01A or B or C

7. CONTINUE in this procedure at: Step 8 to vent the Suppression Chamber through 3-FCV-84-19, 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 INBD ISOL VALVE (Panel 3-9-54). c. PLACE 3-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (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 scfm. f. CONTINUE in this procedure at step 12.

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Event 8: EOI-2


1. 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 INBD ISOL VALVE.

NOTE: Venting may be accomplished using EITHER: 3-FIC-84-19, PATH B VENT FLOW CONT, OR 3-FIC-84-20, PATH A VENT FLOW CONT.

Driver Driver If requested to start a Standby Gas Train After 1 minute, Remote function PC01A or B or C

7. CONTINUE in this procedure at: Step 8 to vent the Suppression Chamber through 3-FCV-84-19, 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 INBD ISOL VALVE (Panel 3-9-54). c. PLACE 3-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (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 scfm. f. CONTINUE in this procedure at step 12.

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Event 8: EOI-2


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 3-9-54). 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: i. IF PRIMARY CONTAINMENT FLOODING per C-1, Alternate Level Control, is in progress, THEN MAINTAIN release rates below those specified in Attachment 2. ii. IF Severe Accident Management Guidelines are being executed, THEN MAINTAIN release rates below those specified by the TSC SAM Team. 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 µCi/s AND 0-SI-4.8.B.1.a.1 release fraction of 1.

BOP Call AUO to Monitor Release Rates

Driver Driver Acknowledge Notification

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Event 8: EOI-2

ATC/BOP Initiate Suppression Chamber Sprays per Appendix 17C

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6. 2. IF Adequate core cooling is assured, OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS I(II) OUTBD INJ VLV BYPASS SEL in BYPASS.

5. INITIATE Suppression Chamber Sprays as follows: a. VERIFY at least one RHRSW pump supplying each EECW header. b. IF ............. EITHER of the following exists: • LPCI Initiation signal is NOT present, OR • Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS I(II) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE. c. MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS I(II) CTMT SPRAY/CLG VLV SELECT, switch in SELECT. d. IF 3-FCV-74-53(67), RHR SYS I(II) INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS I(II) OUTBD INJECT VALVE. e. VERIFY OPERATING the desired RHR System I(II) pump(s) for Suppression Chamber Spray.

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Event 8: EOI-2

ATC/BOP

f. VERIFY OPEN 3-FCV-74-57(71), RHR SYS I(II) SUPPR CHBR/POOL ISOL VLV. g. OPEN 3-FCV-74-58(72), RHR SYS I(II) SUPPR CHBR SPRAY VALVE. h. IF RHR System I(II) is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k. i. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM I(II) MIN FLOW VALVE. j. RAISE system flow by placing the second RHR System I(II) pump in service as necessary. k. MONITOR RHR Pump NPSH using Attachment 2.

l. 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-23-34, 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.

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XIII. Crew Critical Tasks: 6. ADS inhibited before ADS actuates

7. With a reactor scram required and the reactor not shutdown:

• Inserts control rods, to prevent exceeding the primary containment design limits

• Inject boron (prior to exceeding 110°F SP temperature)

8. During an ATWS with emergency depressurization required, injection has been stopped and prevented, with the exception of RCIC, SLC, and CRD, into the RPV until reactor pressure is below Minimum Alternate RPV Flooding Pressure(MARFP)

9. Emergency depressurizes when RPV level cannot be restored

and maintained above -180". Emergency depressurization is initiated before level reaches -200"

10. When RPV pressure is below MARFP, injection reestablished to maintain RPV level above -180”

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XIV. Scenario Verification Data EVENT TASK NUMBER K/A RO SRO

1 EHC Pump Shift RO U-074-AL-16 241000 A4.10 2.9 2.9 2 Reactor Startup RO U-000-NO-03 201002A4.01 3.5 3.4 SRO S-000-NO-001 2.1.7 4.4 4.7 3 Uncoupled Control Rod RO U-085-AB-02 201003A2.02 3.7 3.8 SRO S-085-AB-02 4 Steam Seal Regulator Fail RO U-001-AL-01 239001 K3.06 2.6 2.7 5 HPCI Steam Leak

RO U-073-AL-06 295032EA2.03 3.8 4.0 SRO S-000-EM-10

6 RFPT Trip RO U-003-AL-16 259001A2.01 3.7 3.7 SRO S-003-AB-01 7 Loss of Condenser Vacuum RO U-047-AB-03 295002AA1.06 3.0 3.1 SRO S-047-AB-03 295037EA2.02 4.1 4.2 SRO S-000-EM-18

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XV. SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: HLTS-3-24 8

Total Malfunctions Inserted: List (4-8)

3

Malfunctions that occur after EOI entry: List (1-4)

4

Abnormal Events: List (1-3)

2

Major Transients: List (1-2)

3

EOI’s used: List (1-3)

2

EOI Contingencies used: List (0-3)

90

Run Time (minutes)

5

Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)


SHIFT TURNOVER SHEET

Equipment Out of Service/LCO’s:

IRM E is out of service and bypassed

__ _

Operations/Maintenance for the Shift:

Maintenance has identified that EHC Pump A motor is operating hotter than normal. The Shift

Manager in discussion with plant management has decided to remove EHC Pump A from

service but for it to remain in standby. Complete EHC Auto Pump Start Test & Weekly Pump

Alternation per 3-OI-47A and then continue plant startup, IAW 3-GOI-100-1A section 5.4,

mode change from Mode 2 to Mode 1.

Unit 1 and 2 are 100% power

Unusual Conditions/Problem Areas:

Severe Thunderstorm warning is affect for Limestone, Morgan, and Madison Counties for the

next two hours.


Facility: Browns Ferry NPP Scenario No.: 3-22 Op-Test No.: ILT 1006

Examiners:

Operators:

SRO:

ATC:

BOP:

Initial Conditions:

Unit 3 is at 93% power. Unit 1 and 2 are at 100% power. RHR Pump A is out of service. RBCCW Pump A work is complete and ready to be returned to service. Spare RBCCW Pump is aligned to Unit 3.

Turnover: Rotate Raw Cooling Water Pumps C on and A off and then raise Reactor Power to 100%.

Event No. Malf. No.

Event Type* Event Description

1 N-BOP N-SRO

Rotate Raw Cooling Water Pumps C on and A off

2 R-ATC R-SRO

Raise Power with Recirc

3 pc014 C-SRO C-BOP

Suppression Pool leak on RHR A Loop, isolation available

4 ed16b C-BOP TS-SRO

B Shutdown Board Supply Breaker trips DG B fails to start

5 sw02b C-ATC C-SRO RBCCW Pump B trips

6 rd04 C-ATC TS-SRO

Rod 30-47 drifts out

7 th23 M-ALL Fuel Failure: scram on increasing Radiation and close MSIVs

8 rc09 M-ALL RCIC Steam leak no isolation ED on High Radiation

9 ad03c/f C-ATC C-SRO

Two ADS SRVs will fail to open for emergency depressurization

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


Critical Tasks 1 Emergency depressurization initiated within 5minutes with a primary

system discharging into the secondary containment and area radiation/water/temperature levels exceed maximum safe operating levels in two or more areas.

2 Maintains RPV water level above TAF (-162")

3 Manually isolate systems discharging into secondary containment before Max Safe Operating Value is exceeded. Isolates RHR Loop A

3-22 Rev 0 Page 3 of 55

With the unit at 93% power the BOP operator will rotate Raw Cooling Water Pumps C on and A off per 3-OI-24 section 6.1.1. Upon completion the ATC will commence power increase to 100%. When the NRC is satisfied with power maneuver, a suppression pool leak will develop on RHR A suction due to maintenance in the area. After field personnel are dispatched, field reports will indicate that the leak is isolable. The BOP operator will close RHR A Suppression Pool Suction Valve and the leak will stop. Depending on the timeliness of the crews actions EOI-2 may be entered on low Suppression Pool level. The SRO will evaluate EOI-2 entry conditions and Tech Specs. Once the plant is stable Shutdown Board B supply breaker will trip and DG B will fail to auto start to energize the board. In addition, RBCCW Pump B will trip shortly after the DG failure. The BOP operator will start DG B to restore Shutdown Board B safety related loads. The ATC will start RBCCW Pump A and restore RBCCW to a normal cooling lineup. The SRO will declare DG B Inoperable and refer to Tech Spec 3.8.1 for one DG out of service, condition B will be entered. In addition with RHR A out of service evaluate Tech Spec 3.6.2.3, 3.6.2.4 and 3.6.2.5 and enter condition B for each one. After the above transient Control Rod 30-47 will drift out. The ATC will select and insert Control Rod 30-47 per 3-AOI-85-6. The SRO will refer to Tech Spec 3.1.3 and enter Condition A. The rod drift and insertion will cause a fuel element defect. Plant radiation levels will begin to rise. The SRO will establish power reduction strategies and determine the point at which to scram the reactor. Shortly after the reactor is scrammed, a triple high radiation alarm on MSIVs will come in forcing the crew to isolate the MSIVs. The crew will transition level control to RCIC and pressure control to SRVs. A steam leak on RCIC will develop and with no isolation the crew will have to emergency depressurize due to secondary containment radiation levels. HPCI can be used for level control but the automatic operation of the controller will fail, Two ADS SRVs will fail to open requiring two additional SRVs opened. The crew will maintain level above TAF and after ED restore level to a band of 2 to 51 inches.

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VIII. Console Operator Instructions A. Scenario File Summary 1. File: batch and trigger files for scenario 3-22 Batch NRC201022 #spare rbccw pump aligned irf sw02 align #rhr a pump out of service ior ypobkrpmparh fail_power #rcic no isolation/ steam leak imf rc10 imf rc09 (e10 600) 25 180 ior zdihs712a[2] auto ior zdihs713a[2] auto trg e10 modesw #hpci controller failure in auto/ ads valve failure imf hp03 (e10 0) 0 imf ad03c imf ad03f #suppression pool leak imf pc14 (e1 0) 10 ior xa554c[3] alarm_off mor xa554c[3] (e1 270) alarm_on ior xa554c[17] (e1 30) alarm_on ior xa554c[24] alarm_off ior xa554c[30] alarm_off ior xa554c[31] alarm_off #loss of shutdown board b/ dg fails to auto start imf dg01b (e5 0) imf ed16b (e5 0) trg 6 nrc2010dgb trg 6 = dmf ed16b #rbccw trips imf sw02b (e8 0) #rod drift/fuel failure imf rd04r3047 (e15 0) imf rd23 imf th23 (e15 180) 20 1200

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DG B Start Trigger zdi3hs823b1a(4) .eq. 1 IX. Console Operators Instructions

Scenario 3-22

DESCRIPTION/ACTION

Simulator Setup

manual Reset to IC 28, Lower Power to 93%, Verify 60 RPM difference if between 1200 and 1300 RPM on Recirc Pumps

Simulator Setup Load Batch Bat nrc201022 Place RBCCW Pump 3A in STBY Simulator Setup manual Tag RHR A Pump Switch Simulator Setup Verify Batch file loaded

RCP required (w/Recirc flow) – Provide marked up copy of 3-GOI-100-12 RCP for Urgent Load Reduction

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X. Scenario Summary: With the unit at 93% power, the BOP operator will rotate Raw Cooling Water Pumps C on and A off per 3-OI-24 section 6.1.1. Upon completion, the ATC will commence power increase to 100%. When the NRC is satisfied with power maneuver, a suppression pool leak will develop on RHR A suction due to maintenance in the area. After field personnel are dispatched, field reports will indicate that the leak is isolable. The BOP operator will close RHR A Suppression Pool Suction Valve and the leak will stop. Depending on the timeliness of the crews actions EOI-2 may be entered on low Suppression Pool level. The SRO will evaluate EOI-2 entry conditions and Tech Specs. Once the plant is stable Shutdown Board B supply breaker will trip and DG B will fail to auto start to energize the board. In addition, RBCCW Pump B will trip shortly after the DG failure. The BOP operator will start DG B to restore Shutdown Board B safety related loads. The ATC will start RBCCW Pump A and restore RBCCW to a normal cooling lineup. The SRO will declare DG B Inoperable and refer to Tech Spec 3.8.1 for one DG out of service, condition B will be entered. After the above transient Control Rod 30-47 will drift out. The ATC will select and insert Control Rod 30-47 per 3-AOI-85-6. The SRO will refer to Tech Spec 3.1.3 and enter Condition C. The rod drift and insertion will cause a fuel element defect. Plant radiation levels will begin to rise. The SRO will establish power reduction strategies and determine the point at which to scram the reactor. Shortly after the reactor is scrammed, a high-high radiation alarm on MSIVs will come in forcing the crew to isolate the MSIVs. The crew will transition level control to RCIC and pressure control to SRVs. A steam leak on RCIC will develop and with no isolation the crew will have to emergency depressurize due to secondary containment radiation levels. HPCI can be used for level control but the automatic operation of the controller will fail. Two ADS SRVs will fail to open requiring two additional SRVs opened. The crew will maintain level above TAF and after ED restore level to a band of 2 to 51 inches.

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Terminate the scenario when the following conditions are satisfied or upon

request of Lead Examiner. All Control Rods are inserted Emergency Depressurization Reactor Water Level restored after ED

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XII. Simulator Event Guide: Event 1: Alternate Raw Cooling Water Pumps

BOP 6.1.1 Alternating RCW Pumps [1] REVIEW all Precautions and Limitations in Section 3.0. [2] NOTIFY Chemistry prior to alternating a RCW pump. [3] For the pump to be placed in continuous operation, START the RCW PUMP using the appropriate hand-switch. (Refer to Step 3.2F for pump alteration.) • RCW PUMP 3A, 3-HS-24-7A • RCW PUMP 3B, 3-HS-24-10A • RCW PUMP 3C, 3-HS-24-13A • RCW PUMP 3D, 3-HS-24-16A • RCW PUMP 3E, 3-HS-24-232A

[4] For the pump to be removed from continuous operation, STOP the RCW PUMP using the appropriate hand-switch. (Refer to Step 3.2F for pump alteration.) • RCW PUMP 3A, 3-HS-24-7A • RCW PUMP 3B, 3-HS-24-10A • RCW PUMP 3C, 3-HS-24-13A • RCW PUMP 3D, 3-HS-24-16A • RCW PUMP 3E, 3-HS-24-232A

[5] For the pump to be placed in automatic operation, PLACE the applicable RCW PUMP hand-switch to the mid position and PULL to AUTO. (Refer to Step 3.2F for pump alteration.) • RCW PUMP 3A, 3-HS-24-7A • RCW PUMP 3B, 3-HS-24-10A • RCW PUMP 3C, 3-HS-24-13A • RCW PUMP 3D, 3-HS-24-16A • RCW PUMP 3E, 3-HS-24-232A [6] VERIFY RCW Header Pressure are within the following: • RCW header pressure >61 psig with only 3 pumps operating in parallel. • RCW header pressure >59 psig with only 4 pumps operating in parallel. • RCW header pressure >58 psig with only 5 pumps operating in parallel. • RCW header pressure >57 psig with 6 or more pumps operating in parallel.

BOP Starts Raw Cooling Water Pump C, Stops A and places D or E in Standby

Driver Driver If dispatched to Raw Cooling Water Pumps report Oil levels are normal pump ready for start, after start report no abnormal leaks, pump operating normally.

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XII. Simulator Event Guide: Event 2: Raise Power with Recirc flow

SRO Notify ODS of power increase

Direct Power increase using Recirc Flow per 3-GOI-100-12 [20] 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-OI-68. • MONITOR Core thermal limits using Illustration 1, ICS, and/or 0-TI-248

ATC Raises Power w/Recirc using 3-GOI-100-12 and 3-OI-68, section 6.2

[1] IF NOT in single loop operation, THEN ADJUST Recirc Pump speeds 3A using, RAISE SLOW (MEDIUM), 3-HS-96-15A(15B) AND/OR ADJUST Recirc Pump speed 3B using, RAISE SLOW (MEDIUM), 3-HS-96-16A(16B)

[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required: RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32

NRC When satisfied with Reactivity Manipulation; an additional reactivity manipulation will be required when Control Rod 30-47 drifts out and for the fuel failure.

SRO Notify ODS of power increase

Driver Driver When directed by lead examiner insert trigger 1, to start suppression pool leak

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XII. Simulator Event Guide: Event 3: RHR A Suppression Pool Leak

BOP/ATC

Respond to Annunciator 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 Annunciator DRYWELL TO SUPPR CHAMBER. DIFF PRESS ABNORMAL A. CHECK alarm by checking Drywell to Suppression Chamber DP. B. REFER TO 1-AOI-64-1. C. REFER TO Tech Spec Section 3.6.2.6.

BOP/ATC Dispatch personnel to RHR Loop 1 area

SRO Evaluates Tech Spec 3.6.2.6 and Enters EOI-3

Driver Driver 3 minutes after dispatched report leak is between 3-FCV-74-1 RHR Pump 3A Suppression Pool Suction Valve and RHR Pump 3A, appears leak was caused by maintenance work in the area. When the crew closes 74-1 report leak has slowed; some minor leakage or residual leakage remains. Change PC14 to .01 In addition report water level is about 3 to 5 inches in this quad and there is a few inches in the suppression chamber area.

BOP/ATC

Respond to Annunciator 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.

CT #3 SRO When report of leak source is reported direct BOP to close 74-1 RHR Pump 3A Suppression Pool Suction Valve

CT #3 BOP Close 74-1 RHR Pump 3A Suppression Pool Suction Valve

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XII. Simulator Event Guide: Event 3: RHR A Suppression Pool 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

Contact Radwaste to determine status of sump Pumps

Driver Driver Report sump pumps in area of alarm are operating normally, after 74-1 is isolated delete overrides on alarms xa554c[3] alarm_on and ior xa554c[17] alarm_on

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

Driver Driver Upon Lead examiner direction, initiate Trigger 5 for Shutdown Board B Supply Breaker Trip

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XII. Simulator Event Guide: Event 4: Shutdown Board B Supply Breaker Trip

Crew Respond to numerous Annunciators the key ones being 4KV SHUTDOWN BD 3EB DEGRADED VOLTAGE, A. VERIFY Diesel starts and ties to board, if required. B. REFER TO 0-OI-57A and Tech Spec Section 3.8.1. 4160V SD BD 3EB DEGRADED VOLTAGE A. VERIFY unit in stable condition by checking: 1. Auto actions has occurred and diesel at rated frequency and voltage. 2. Diesel connects properly to assigned 4160V Shutdown Bd. 3. Diesel reaches proper loading when supplying equipment feed from board.

BOP/SRO

Recognize loss of Shutdown Board 3EB with a failure of the 3B Diesel Generator to start

BOP Recommend or start DG 3B

SRO Concur with start of DG 3B or have BOP start DG 3B

BOP Report DG 3B started and energized Shutdown Board 3EB

Driver Driver Once crew energizes the bus report trip of supply breaker due to staging of equipment by contractor personnel, inadvertently bumped breaker with hand truck.

BOP Responds to remaining annunciators which will reset and clear once DG 3B is started.

BOP Reports other annunciators are clear and were 480 volt alarms due to loss of 4 KV Shutdown Board 3EB

ATC/BOP Determine what equipment tripped on the loss of Shutdown Board 3EB

Driver Driver While the SRO and BOP are addressing the loss of Shutdown Board B insert trigger 8 for trip of RBCCW Pump 3B, after the RBCCW Pump Trips delete the malfunction sw02.

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XII. Simulator Event Guide: Event 4: Shutdown Board B Supply Breaker Trip

SRO Evaluates Technical Specification 3.8.1 Condition B, B.1 Verify power availability from the offsite transmission network. 1 Hour AND Once per 8 hours thereafter AND B.2 Declare required feature(s), supported by the inoperable Unit 3 DG, inoperable when the required feature(s) are inoperable. 4 hours from the discovery of Condition B concurrent with inoperability of redundant required feature(s) AND B.3.1 Determine OPERABLE Unit 3 DG(s) are not inoperable due to common cause failure. 24 hours OR Perform SR 3.8.1.1 for OPERABLE DG(S). 24 hours AND B.4 Restore Unit 3 DG to OPERABLE status 7 days.

SRO Evaluate Technical Specification 3.6.2.3, 3.6.2.4, 3.6.2.5 3.6.2.3 Condition B, B.1 Restore one RHR suppression pool cooling subsystem to OPERABLE status. 7 days.

3.6.2.4 Condition B, B.1 Restore one RHR suppression pool spray subsystem to OPERABLE status. 7 days.

3.6.2.5 Condition B, B.1 Restore one drywell spray subsystem to OPERABLE status. 7 days.

Driver Driver While the SRO and BOP are addressing the loss of Shutdown Board B insert trigger 8 for trip of RBCCW Pump 3B, after the RBCCW Pump Trips delete the malfunction sw02.

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XII. Simulator Event Guide: Event 5: RBCCW Pump 3B Trip

ATC Respond to RBCCW Pump Trip Annunciators, RBCCW PUMP DISCH. HDR PRESS LOW and RBCCW 3-FCV-70-48 CLOSED, report trip of RBCCW Pump 3B

ATC RBCCW PUMP DISCH. HDR PRESS LOW A. VERIFY 3-FCV-70-48 CLOSING/CLOSED. B. VERIFY RBCCW pumps A and B in service. C. VERIFY RBCCW surge tank low level alarm is reset. D. DISPATCH personnel to check the following: • RBCCW surge tank level locally. • RBCCW pumps for proper operation. E. REFER TO 3-AOI-70-1 for RBCCW System failure and 3-OI-70 for starting spare pump.

SRO Enter 3-AOI-70-1 Loss of Reactor Building Closed Cooling Water

ATC Immediate Actions [1] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A): • SECURE RWCU Pumps. • VERIFY RBCCW SECTIONALIZING VLV, 3-FCV-70-48 CLOSED.

ATC Secures RWCU Pumps and verifies 3-FCV-70-48 closes

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ATC Respond to RBCCW Pump Trip Annunciators, RBCCW PUMP DISCH. HDR PRESS LOW and RBCCW 3-FCV-70-48 CLOSED, report trip of RBCCW Pump 3B

SRO 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): [1.1] REDUCE core flow to between 50-60%. [1.2] MANUALLY SCRAM the Reactor and PLACE Mode Switch to SHUTDOWN. REFER TO 3-AOI-100-1. [1.3] SHUT DOWN both Recirc Pumps. • DEPRESS RECIRC DRIVE 3A SHUTDOWN, 3-HS-96-19. • DEPRESS RECIRC DRIVE 3B SHUTDOWN, 3-HS-96-20. [1.4] INITIATE a 90°F/HR cooldown rate. REFER TO 3-AOI-100-1. [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A).

ATC Subsequent Actions [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following: [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). [4] IF unable to restart a tripped pump, THEN: PLACE Spare RBCCW Pump in service. REFER TO 3-OI-70. [5] IF RBCCW flow was restored to two pump operation by placing the Spare RBCCW pump in service in the preceding step, THEN PERFORM the following: [5.1] REOPEN RBCCW SECTIONALIZING VLV, 3-HS-70-48A [5.2] RESTORE the RWCU system to operation. (REFER TO 3-OI-69)

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ATC Starts RBCCW Pump 3A or 3B, opens 3-FCV-70-48 and dispatches personnel to RBCCW Pumps / Breaker

Driver Driver If requested wait 2 minutes report surge tank level normal and stable in middle of band. If requested to verify position of RBCCW Pump 3A discharge valve wait 2 minutes and report valve is open. If requested, wait 2 minutes and report RBCCW Pump 3B breaker (480 volt SD BD 3B) is tripped if Pump was not started; if 3B RBCCW Pump was started, report nothing abnormal. If requested wait 2 minutes and report prior to a pump start nothing abnormal at pump or breaker for RBCCW 3B

SRO Concurs with start of RBCCW Pump 3A or 3B and prioritizes recovery plan for RWCU

SRO 3.4.1 Notifies Chemistry that continuous monitoring is no longer available and to commence sampling per TRM Surveillance 3.4.1.1, Monitor reactor coolant activity 4 hours when the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or 5

[6] IF RBCCW loss is partial, THEN CONTROL OR REDUCE Drywell and Recirc Pump Temperatures.(Otherwise N/A) REFER TO 3-GOI-100-12A [6.1] REDUCE Recirculation flow as necessary until 50-60% core flow is reached. [6.2] INSERT control rods per 3-SR-3.1.3.5(a) as directed by Reactor Engineer until below 66.7% rod line. [6.3] REDUCE Recirculation flow as necessary until minimum core flow is reached. [6.4] INSERT control rods per 3-SR-3.1.3.5(a) as directed by Reactor Engineer.

[7] IF RBCCW Surge Tank level is low, THEN: (Otherwise N/A):

[8] VERIFY all available Drywell Coolers in service. REFER TO 3-OI-64.

[9] IF RBCCW PUMP SUCTION HDR TEMP, 3-TIS-70-3 (Panel 3-9-4), cannot be maintained below 105°F, THEN: (Otherwise N/A):

[11] MONITOR Drywell Temperature on 3-XR-64-50 (Panel 3-9-3).

[12] MONITOR the following Recirc Pump and Motor 3A(3B) Winding and Bearing Temperatures on 3-TR-68-71 (Panel 3-9-21):

Driver Driver Upon Lead examiner direction, initiate Trigger 15 for Control Rod 30-47 drift out

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XII. Simulator Event Guide: Event 6: Control Rod 30-47 Drift Out

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ATC Report CONTROL ROD DRIFT on Control Rod 30-47 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. C. IF rod drifting in, THEN REFER TO 3-AOI-85-5. D. IF rod drifting out, THEN REFER TO 3-AOI-85-6. E. REFER TO Tech Spec Section 3.1.3, 3.10.8.

SRO Enter 3-AOI-85-6 Rod Drift Out

ATC 4.1 Immediate Actions [1] IF multiple control rod drifts are identified, THEN MANUALLY SCRAM the reactor and enter 3-AOI-100-1. Subsequent Actions [1] IF a Control Rod is moving from its intended position without operator actions, THEN SELECT the drifting control rod and INSERT to the FULL IN (00) position. [2] IF control rod drive does NOT respond to INSERT signal, THEN PERFORM the following:

SRO [3] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

Driver Driver Acknowledge Reactor Engineer communication, when control rod 30-47 is being driven in to 00 deletes the malfunction rd04r3047 and rd23

ATC

[4] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN [5] IF the control rod will not latch into position “00” and continues to demonstrate occurrences of inadvertent withdrawal, THEN INSERT and MAINTAIN control rod at position 00. (Otherwise N/A)

SRO [6] REMOVE associated HCU from service per 3-OI-85.

BOP/ATC Dispatch personnel to remove HCU 30-47 from service

Driver

Driver

When dispatched to remove HCU from service wait 4 minutes and report HCU 30-47 removed from service

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XII. Simulator Event Guide: Event 6: Control Rod 30-47 Drift Out

SRO [7] EVALUATE Tech Spec 3.1.3. [8] INITIATE Service Request/Work Order.

SRO Evaluate Technical Specification 3.1.3 Condition C, C.1 Fully insert inoperable control rod 3 hours AND C.2 Disarm the associated CRD 4 hours

[9] NOTIFY Reactor Engineer to perform the following for current condition: • EVALUATE condition of core to assure no resultant fuel dam

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