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Text

Initial Exam 2013-301 Draft Simulator Scenarios
	ML13214A325
Person / Time
Site:	Browns Ferry
Issue date:	08/01/2013
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	Download: ML13214A325 (271)
	v • d • e

{{#Wiki_filter:Facility: Browns Ferry NPP Scenario No.: NRC 1- Op-Test No.: 1306 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: Reactor Power is 70%. HPCI is tagged out. Condensate Booster Pump 3A has been returned to service. Turnover: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3. Commence a power increase in accordance with Reactivity Control Plan to 90%. Event Maif. No. Event Event Description No. Type* N-BOP I Place 3A RFPT in service from 600 RPM lAW 3-01-3 N-SRO 2 - Power increase with Control Rods and flow to 90% R-SRO 3 RDO1A CRD Pump Trip C-ATC 4 sev file Control Rod 3 0-23 High CRD Temperature TS-SRO C-BOP 5 ed07b Loss of 480V Unit Board 3B IB OP Off Gas Recombiner Catalyst failure, with Hydrogen 6 ogO5aJogOl TS-SRO explosion Loss of Feedwater / RCIC Steam Leak / ED on High 7 Batch M-ALL Radiation 8 RHO 1AJC C 3A and 3C RHR Pumps trip 9 IOR I ADS SRV 1-34, Acoustic Monitor indication failure

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

NRC Scenario 1 ility Browns Ferry NPP Scenario No NRC 1- Op-Test No Jjj Examiners:____________________ Operators: SRO:_____________________ ATC:_______________ BOP:_______________ Initial Conditions: Reactor Power is 70%. HPCI is tagged out. Condensate Booster Pump 3A has been returned to service. Turnover: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3. Commence a power increase in accordance with Reactivity Control Plan to 90%. Event Maif. No. Event Type* Event Description No. 1 - Place 3A RFPT in service from 600 RPM lAW 3-01-3 N-SRO R-ATC 2 Power increase with Control Rods and flow to 90% 3 01A C Pump Trip 4 sev file Control Rod 3 0-23 High CR1) Temperature TS.SRO C-BOP 5 ed07b Loss of 480V Unit Board 3B 1-BOP 6 Off Gas Recombiner Catalyst failure, with Hydrogen explosion ogosa/ogOl TS-SRO 7 Batch M-ALL Loss of Feedwater / RCIC Steam Leak / ED on High Radiation 8 RHO1AJC C 3A and 3C RHR Pumps trip 9 IOR I ADS SRV 1-34, Acoustic Monitor indication failure

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

NRC Scenario 1 Critical Tasks Three

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 without delay.

1. Safety Significance:

Places the primary system in the lowest possible energy state, rejects heat to the suppression pooi 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. When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and prior to exceeding the PSP limit.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays lAW with EOI Appendix I 7B

AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment 2

NRC Scenario I Critical Tasks Three

When MAIN STEAM LINE RADIATION HIGH-HIGH is received and SLC injection is NOT required in accordance with RC/Q of EOI-l, MSIVs are closed. SRO will direct MSIV closure without delay and the RO will close MSIVs.

1. Safety Significance:

Isolate potential discharge path to environment.

2. Cues:

Procedural compliance. Reactor Pressure trend. Pressure Control on SRVs

3. Measured by:

When ARP-9-3A window 27 annunciator is received MSIVs are closed if alarm is verified valid and SLC injection lAW EOI-1 RC/Q is NOT required.

4. Feedback:

Reactor Pressure trend. MSIVs indicate closed. 3

NRC Scenario 1 Events

1. BOP places RFPT 3A in service from 600 rpm lAW 3-01-3, Reactor Feedwater System section 5.7

2. ATC will commence to raise power with control rods to a load line of 105 to 106, by withdrawing four control rods from 00 to 24.

3. CRDH pump 3A trips while the ATC is withdrawing the fourth control rod. ATC will perform 3-A0I-85-3 actions to start the Standby CR]) Pump and restore CR])

parameters.

4. Crew will respond to a loss of 480V Unit Board 3B, the BOP will restore from a trip of Bus Duct Cooling 3A Fan and take action JAW with ARPs, start standby Bus Duct Cooling Fan 3B.

5. Once Control Rod 3 0-23 is withdrawn to position 24, this control rod will experience a high CRD mechanism temperature. The ATC will respond JAW ARPs and 3-01-85, the operator will perform CRD flushing for a control rod at a position other than

48. The SRO will declare control rod 30-23 slow lAW Technical Specification 3.1.4-1 Note 1.

6. The Hydrogen Water Injection system will malfunction resulting in high hydrogen concentration in Off Gas The BOP Operator will respond lAW with ARPs and 3-AOI-66-1 an Off Gas System explosion will occur once off gas hydrogen concentration exceeds 5%. The SRO will evaluate TRM 3.7.2 and enter Condition A.

7. On the SCRAM a loss of Unit Board 3C will occur with a trip of Condensate Booster Pump 3A, this will result in only one Condensate Booster Pump in service, eventually Condensate Booster Pump 3B will trip resulting in a loss of feedwater. RCIC will auto start and slowly start to restore Reactor Level. Shortly after the scram a LOCA will occur requiring Suppression Chamber and Drywell Sprays. Reactor Level will begin to lower. An unisolable RCIC steam leak will develop and Emergency Depressurization will be required due to High Secondary Containment Radiation Levels

8. RHR Pumps on Division 1 will trip and be unavailable for Containment Cooling. The crew will use Division 2 RHR Pumps for Containment Cooling functions.

9. When the Emergency Depressurization is required the Acoustic Monitor on ADS SRV 1-34 will fail, requiring the crew to open an additional SRV.

10. Once Emergency Depressurization is complete the crew will restore Reactor Level with Core Spray Loops 1 or 2, and the Condensate Pumps.

4

NRC Scenario 1 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner Drywell Sprays initiated and secured Emergency Depressurization complete Reactor Level is restored and maintained SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 1 8 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) I Major Transients: List (1-2) 3 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 65 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) 5

NRC Scenario 1 Scenario Tasks TASK NTJMBER KIA RO SRO Place a RFPT in Service RO U-003-NO-04 259001A4.01 3.6 3.5 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-NO-31 2.2.2 4.6 4.1 CRD Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 CRD Mechanism Temperature High RO U-085-AL-08 201003A2.1O 3.0 3.4 SRO S-085-AB-03 Loss of 480V Umt Board 3B ROU-57A-AL-15 262001A2.04 3.8 4.2 SRO S-57B-NO-07 Off Gas Explosion RO U-066-AB-O1 271000 A2.06 3.5 3.9 SRO S-066-AB-01 Fuel Failure RO U-090-AL-07 272000A2.01 3.7 4.1 SRO S-000-EM-02 RCIC Steam Leak RO U-000-EM-10 295033EA2.01 3.8 3.9 SRO S-000-EM-10 LOCA RO U-000-EM-0l 295009AA2.01 4.2 4.2 SRO S-000-EM-01 6

NRC Scenario 1 Secondary Containment High Radiation RO U-090-AL-04 295033EA2.O1 3.8 3.9 SRO S-000-EM-15 Procedures Used/Referenced: Procedure Number Procedure Title 3-01-3 Reactor Feedwater System 3-G0I-lOO-12 Power Maneuvering 3-01-85 Control Rod Drive System 3-A0I-85-3 CRD System Failure 3-TI-393 Evaluation of CRD Temperature Alanns Technical Specifications 3-ARP-9-7A Panel 9-7 3-XA-55-7A 3-ARP-9-8A Panel 9-8 3-XA-55-8A 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-8C Panel 9-8 3-XA-55-8C 3-ARP-9-53 Panel 9-53 3-XA-55-53 3-A0I-66-l Off Gas H2 High 3-A0I-iOO-i Reactor Scram 3-E0I-i RPV Control 3-E0I-2 Primary Containment Control 3-E0I-3 Secondary Containment Control 3-E0I Appendix-5C Injection System Lineup RCIC

             .           Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 3-E0I Appendix-8F Isolation 3-E0I Appendix-17C        RHR System Operation Suppression Chamber Sprays 3-EOI Appendix-i 7B       RHR System Operation Drywell Sprays 3-ARP-9-3A                Panel 9-3 3-XA-55-3A 3-EOI Appendix-i iA       Alternate RPV Pressure Control Systems MSRVs 3-EOI Appendix-6A         Injection Subsystems Lineup Condensate 3-EOI Appendix-6D         Injection Subystems Lineup Core Spray System I 3-EOI Appendix-6E         Injection Subystems Lineup Core Spray System II 3-EOI-2-C-2               Emergency RPV Depressurization EPIP- 1                   Emergency Classification Procedure EPIP-4                     Site Area Emergency 7

NRC Scenario 1 Console Operator Instructions A. Scenario File Summary Batch File

HPCI tagout bat nrchpcito
bus duct cooling fan trip imf ed07b (el 0)
Rod 3 0-23 high temperature sev rdkxleak(63) -1.01 sev yua85795 240 trg 9 = sev yua85795 345 trg 10 = sev rdkxleak(63) -.9 trg 11 = sev rdkxleak(63) -.96 trg 12 = sev rdkxleak(63) .768
hwc malfunction imfog05a (e3 0) 85 1200 100 imfogOl (e5 0)
CRD A Pumptrip imfrd0la (e7 0)
Major ior zaoxi0 134 0 Acoustic monitor failure ADS SRV 1-34 imfed08c (e15 0) Loss of Unit Board 3C imffw02a (e15 0) Condensate Booster Pump 2A Trip imffw02b (e15 1:45) Condensate Booster Pump 2B Trip trg e20 MODESW imfrc09 (e20 12:20) 25 120 RCIC Steam Leak imfth2l (e20 5:00) .2 LOCA imfpcl6a(e20 5:10) Vacuum Breaker failure imfpcl6b (e20 5:10) imfth23 (e20 0)10 1000 Fuel Failure imf rc 10 RCIC Steam Valves fail to Auto Close ior zdihs7l2a open RCIC Steam Valve fails to close ior zdihs7l3a open RCIC Steam Valve fails to close trg e30 nrcRHR3C imfrh0lc (e30 :30) RHR 3C Trip imf rhO 1 a RHR 3A Trip 8

NRC Scenario 1 Trigger Files MODESW ZDIHS465(4) .NE. 1 nrcRHR3C 3C RHR Pump to start

zdihs74l6a[3] .eq. 1 Scenario 1 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 201 Simulator Setup close HPCI Valves and open drain valves manual so steam line pressure decays to zero Simulator Setup Load Batch bat nrcl3O6-l Simulator Setup Tag HPCI, Shift Bus Duct Fans to B on manual and A in Standby, swap lens covers on Bus_duct Simulator Setup Verify file loaded RCP required (70% 90% with control rods and flow) and RCP for Urgent Load Reduction

Provide marked up copy of 3-GOI-100-12 9

NRC Scenario 1 ( Simulator Event Guide: Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3 5.7 Placing the Second and Third RFP/RFPT In Service [2] IF RFP/RFPT is NOT warmed, reset and rolling, THEN PERFORM the following: (Otherwise N/A) [3] VERIFY RFP 3A MIN FLOW VALVE, 3-HS-3-20, in OPEN position.

CHECK OPEN M1N FLOW VALVE, 3-FCV-3-20.

[4] SLOWLY RAISE speed of RFPT using RFPT 3A SPEED CONT RAISE/LOWER, 3-HS-46-8A, to establish flow and maintain level in vessel. [5] WHEN RFPT discharge pressure is within 250 psig of reactor pressure, ThEN VERIFY OPEN RFP 3A DISCHARGE VALVE, 3-FCV-3-19. [6] SLOWLY RAISE RFPT speed using RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A, to slowly raise RFP discharge pressure and flow on the following indications (Panel 3-9-6):

RFP Discharge Pressure RFP 3A, 3-PI-3-16A

RFP Discharge Flow RFP 3A, 3-FI-3-20

[7] WHEN sufficient flow is established to maintain RFP 3A MIN FLOW VALVE, 3-FCV-3-20, in CLOSED position (approximately 2 x 106 lbm/hr), THEN PLACE RFP 3A M1N FLOW VALVE, 3-HS-3-20, in AUTO. [8] OBSERVE lowering in speed and discharge flows of other operating RFPs. 10

NRC Scenario 1 Simulator Event Guide: Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3 [9] IF transferring RFPT from MANUAL GOVERNOR to individual RFPT Speed Control PDS, THEN PERFORM the following: (Otherwise N/A) [9.1] PULL RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A, to FEEDWATER CONTROL position. [9.2] VERIFY amber light at switch extinguished above RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A. [9.3] PERFORM the following on RFPT 3A SPEED CONTROL(PDS), 3-SIC-46-8 (Panel 3-9-5): [9.3.1] SELECT Column 3. [9.3.2] VERIFY PDS in MANUAL. [10] IF transferring control of RFPT from individual RFPT Speed Control PDS to AUTO control using REACTOR WATER LEVEL CONTROL PDS, 3-LIC-46-5, THEN PERFORM the following: (Otherwise N/A) [10.1] VERIFY REACTOR WATER LEVEL CONTROL (PDS), 3-LIC-46-5 is functioning properly and ready to control second or third RFP. [10.2] SLOWLY RAISE RFP discharge flow and pressure by raising RFP speed. [10.3] WHEN RFP speed is approximately equal to operating RFP(s) speed, THEN PERFORM the following on RFPT 3A SPEED CONTROL (PDS), 3-SIC-46-8: [10.3.1] PLACE PDS in AUTO. [10.3.21 VERIFY Colunin 3 selected. 11

NRC Scenario 1 Simulator Event Guide: Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3 [11] WHEN RFP in automatic mode on REACTOR WATER LEVEL CONTROL, (PDS) 3-LIC-46-5, THEN CLOSE the following valves:

RFPT 3A LP STOP VLV ABOVE SEAT DR 3-FCV-6-120
RFPT 3A LP STOP VLV BELOW SEAT DR, 3-FCV-6-121
RFPT 3A HP STOP VLV ABOVE SEAT DR 3-FCV-6-122
RFPT 3A NP STOP VLV BELOW SEAT DR. 3-FCV-6-123
RFPT 3A FIRST STAGE DRAiN VLV, 3-FCV-6-124
RFPT A HP STEAM SHUTOFF ABOVE SEAT DRAIN, 3-FCV-006-0 153 (local control)
RFPT ALP STEAM SHUTOFF ABOVE SEAT DRAIN, 3-FCV-006-0 154 (local control)

[12] VERIFY CLOSED the following valves on first RFP started in Section 5.5:

RFPT 3B(3C) LP STOP VLV ABOVE SEAT DR, 3-FCV-6-125(130)
RFPT 3B(3C) LP STOP VLV BELOW SEAT DR, 3-FCV-6-126(13 1)
RFPT B(C) LP STEAM SHUTOFF ABOVE SEAT DR.

3-FCV-006-0156(01 58) (local control) [13] VERIFY both RFPT Main Oil Pumps running. [14] IF desired to stop Turning Gear for in service RFPT, THEN PLACE

appropriate handswitch in STOP and RETURN to AUTO:

RFPT 3A TURNiNG GEAR MOTOR. 3-HS-3-1O1A

[15] REFER TO Section 6.0.

                            . CONTROL and MONITOR RFW system operation.

12

NRC Scenario 1 Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase. Direct Power increase using control rods LAW 3-GOT-i 00-12. [21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:

RAISE power using control rods or core flow changes. REFER TO 3-SR-3.3.5(A) and 3-01-68.
MONITOR Core thermal limits using ICS, and/or 0-TI-248 ATC Raise Power with Control Rods lAW 3-01-85, section 6.6. Control Rods:

22-31 from 0 to 24, 30-39 from 0 to 24, 38-31 from 0 to 24 and 30-23 from 0 to 24 NOTES Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches. When continuously withdrawing a control rod to a position other than position 48, the CR1) t Notch Override Switch is held in the Override position and then the C Control Switch is held in the Rod Out Notch position.

Both switches should be released when the control rod reaches two notches prior to its intended position.

6.6.1 Initial Conditions Prior to Withdrawing Control Rods [2] VERIFY the following prior to control rod movement:

                                        . CRD POWER, 3-HS-85-46 in ON.
                                        . Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing.

6.6.2 Actions Required During and Following Control Rod Withdrawal [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]. 13

NRC Scenario 1 ( Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods [6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REiNITIALIZE the RBM: [6.1] PLACE CRD POWER, 3-HS-85-46 in the OFF position to deselect the Control Rod. [6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position. 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 LATCI{ED 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. Driver At NRC direction, trip CRD Pump 3A trigger 7 14

NRC Scenario 1 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 LATCI{ED into correct ROD GROUP when the Rod Worth Minimizer is enforcing. [4] VERIFY Control Rod is being withdrawn to a position greater than three notches. [5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: [5.1] PLACE AND HOLD CR1) NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE. [5.2] PLACE AN]) HOLD CR1) 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 CR1) CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE. 15

NRC Scenario 1 ( Simulator Event Guide: Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued) [5.5] WREN 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.

o 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. [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) 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 s desired, THEN: I. [1.1] PLACE CRD POWER, 3-HS-85-46, in OFF. [1.2] PLACE CR0 POWER, 3-HS-85-46, in ON. 16

NRC Scenario 1 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, CR1) System Failure. 4.1 Immediate Actions [11 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 CRD Pump using one of the following:

CR1) PUMP 3B, using 3-HS-85-2A

[1.3] ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, to establish the following conditions:

CRD CLG WTR HDR DP, 3-PDI-85-18A, approximately 20 psid
CR1) SYSTEM FLOW CONTROL, 3-FIC-85-1 1, between 40 and t

65 gpm. [1.4] BALANCE CR1) SYSTEM FLOW CONTROL, 3-FIC-85-1 1, and PLACE in AUTO or BALANCE. If Dispatched to CRD Pump 3A, pump is extremely hot to touch. CRD Pump 3B oil levels in band, pump ready for start, conditions normal after the start.

CR1) 3A report breaker tripped on over current, Electrical Maint called.

iseto 345 ishing Driver )0 degrees II1ILLL I I LU IUW UL III I LiI.J. .JIi*.. UI iw i liii iuugi monitoring CR]) Temps insert trigger 12 to return simulator to normal otherwise the last sev entered will stay in, resetting does not clear. 17

NRC Scenario 1 Simulator Event Guide: Event 4 Component: Control Rod 3 0-23 High Temperature ATC Responds to annunciator 9-5A window 17: CONTROL ROD DRIVE UNIT TEMP HIGH A. CHECK high temp of CRD on recorder 3-TR-85-7A, & 3-TR-85-7B (Panel 3-9-47) or on ICS. B. IF alarm is VALID, THEN PERFORM the following, as directed by the Unit Supervisor.

CHECK cooling water pressure and flow normal on Panel 3-9-5.
DISPATCH personnel to check for HCU scram discharge valve leaking as indicated by elevated discharge piping temperatures for associated CRD.
PERFORM 3-TI-393 for control rods with high temperatures or failed thermocouples.
REFER TO OPDP-4, 3-01-85, 3-AOI-85-3.
FLUSH CRD to unblock restricted cooling water flow. REFER TO 3-01-85.
DECLARE the control rod, which is in alarm, SLOW as directed by 3-TI-393 per Tech Spec Table 3.1.4-1 Note 1.
RAISE CRD Flow, as directed by Unit Supervisor, if required to keep the drives cool per CR1) Pump Operation At Elevated Flow section of 3-01-85.

Driver When called report CR1) Temperature from ICS, report no elevated pipmg temperature lx

NRC Scenario 1 I Simulator Event Guide: Event 4 Component: Control Rod 3 0-23 High Temperature ATC Flush of Control Rod 3 0-23 lAW 3-01-85 Section 8.29 CAUTION This section may only be used at the direction of the CRD System engineer to flush a CRD that has the characteristics of the cooling water flow being restricted by debris within the CRD. This flushing section should not be performed repeatedly (Greater than one time per week)on the same CRD to prevent thermal cycling on the CR1). GESIL 173. [1] IF flushing rods at Position 48, THEN: NOTE The following steps (performed only on rods at positions other than 48) perform an insert flush at reduced drive water DP to prevent the CRD from being inserted. [2] IF flushing rods at positions other than 48, THEN: [2.1] LOWER the CRD DRIVE WTR HDR DP, 3-PDI-85-17A, to <75 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A. [2.2] VERIFY CR]) POWER, 3-HS-85-46, in ON. [2.3] SELECT the control rod to be flushed by depressing the appropriate CR1) ROD SELECT pushbutton, 3-XS-85-40. [2.4] IF while performing this section Rod Motion is observed, THEN [2.4.1] IMMEDIATELY RELEASE CR]) CONTROL SWITCH 3-HS-85-48 and NOTIFY System Engineer. [2.4.2] RAISE the CRD DRIVE WTR HDR DP, 3-PDI-85-17A to between 250 psid and 270 psid, using CR]) DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A. [2.4.3] STOP further flushing until authorized by the Unit Supervisor. NOTE It could take up to two minutes for CRD DRIVE WTR HDR DP to stabilize. [2.5] CHECK CR1) DRIVE WTR JIDR DP, 3-PDI-85-17A, is stable and <75 psid. [2.6] PLACE AND HOLD CR1) CONTROL SWITCH 3-HS-85-48, in ROD IN for 30 seconds. ier 19

NRC Scenario 1 ,i Simulator Event Guide: Event 4 Component: Control Rod 30-23 High Temperature [2.7] ADJUST the CR1) DRIVE WTR HDR DP, 3-PDI-85-17A to between 250 psid and 270 psid, using CR1) DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A. ATC [2.8] DESELECT the control rod after flushing by PLACING CR1) POWER, 3-HS-85-46, in OFF, THEN PLACE CRD POWER, 3-HS-85-46, in ON. Evaluate 3-TI-393 and Tech Spec 3.1.4 EVALUATION OF CRD TEMPERATURE ALARMS SCOPE This procedure is utilized by Operations personnel to evaluate CRD high temperature alarms as directed by 3-ARP-9-5A. FREQUENCY This instruction will be performed for the following conditions: As required to evaluate new CRD high temperature alarms, 4.0 PRECAUTIONS AND LIMITATIONS 4.1 GE STE 173 discusses the potential effects on scram times for CRDs whose operating temperature is above 350 °F. Scram times can be increased by 0.150 to 0.500 seconds for a CR1) operating between 350 °F and 525 °F. cRDs with tempeatures above 350 °F are declared slow m accordance with Techmcal Specification Section-3 1 4ndieadded to the population ofrods scráni time tested per 3-SR-3. 1.4.1. SRO 7.1 Evaluation of New Alarms 7.1.1 Record the CRD number on Appendix B, Evaluation of New CRD Temperature Alarms. Record the CRD temperature or indicate that the alarm is due to an open thermocouple input (upscale trip of 3-TA-85-7 and unknown temperature data displayed on ICS). 7.1.2 Determine if the scram outlet valve is leaking by comparing the scram outlet piping temperature for the affected CRD to the scram outlet piping temperatures of adjacent HCUs. 7.1.4 If the CR1) temperature is greater than 350 °F and it has been determined that the scram outlet valve is not leaking, perform Section 8.28 of 3-01-85 to flush the CRD as directed by the System Engineer. If the CRD temperature returns to normal after flushing, document in the Comments field of Appendix B. 7.1.5 If the CRD temperature remains greater than 350 °F, 7.1.5.1 Declare the affected rod SLOW perNote 1 of Tech Spec Table 3.1.4-1, and contact Reactor Engineering to update the scram time data base. 20

NRC Scenario 1 Simulator Event Guide: Event 4 Component: Control Rod 30-23 High Temperature SRO Evaluate 3-TI-393 and Tech Spec 3J.4 3i REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 13 OPERABLE control rods shall be slow, in accordance with Table 3.1 A-I; and

b. No more than 2 OPERABLE control rods that are slow shall occupy adjacent locations.

APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQU1RED ACTION COMPLETION TIME A. Requirements of the LCD A.1 Be in MODE 3. 12 hours not met. Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times

                                          -----------     ---NOTES----

1. OPERABLE control rods with scram times not within the limits of this Table are considered slow.

2. Enter applicable Conditions and Required Actions of LCO 3.1.3, Control Rod OPERABILITY, for control rods with scram times> 7 seconds to notch position 0&

These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered slow. Driver At direction of NRC initiate trigger 1 for loss of 480V Unit BD 3B 21

NRC Scenario I Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, C rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 Announce loss of 480V Unit Board 3B SRO Prioritize annunciators 9-7A Window 31, Highest Priority BOP 9-7A Window 31 GEN BUS DUCT FAN FAILURE A. VERIFY Main Bus Cooling Fans, 3-HS-262-1A or 1-HS-262-2A, indicates running on Panel 3-9-8 AND START GEN BUS DUCT HX FAN A using 3-HS-262-1A, on panel 3-9-8 to start the standby fan. B. IF no Fans are operating and the Generator is tied to the grid and loaded to greater than the self cooled bus rating of 16,500 amps THEN, IMMEDIATELY iNSERT a manual reactor scram, AN]) TRIP the Main Generator. C. IF while executing this procedure, the Bus Duct Temperature is at or above the Temperature Excursion limit of 120°C, THEN IMMEDIATELY INSERT a manual reactor and TRIP the Main Generator. D. DISPATCH personnel as necessary to check the following:

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

2. Monitor Bus Duct temperature by available means including using a portable temperature monitor device locally at the 14 in-service thermostats. REFER to Window 32, Figure 1.

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

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

E. VERIFY the system is operating in accordance with 3-01-47. When sent to investigate inform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (51U) 22

NRC Scenario 1 I Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, C rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8A Window 4 and 35 Window 4 EXCITATION SYSTEM ABNORMAL A. ChECK the following displays on ICS to identify the cause: AVR ALARM CODES (AVRALM) AVR FAULT CODES (AVRFLT) B. ADJUST, as required, VOLTAGE REGULATOR LOWERIRAISE ADJUST, 3-HS-57-26, to maintain the following:

GENERATOR VOLTS, 3-EI-57-39, between 20,900V and 23,100V.
GENERATOR MVARS, 3-El-S 7-51 Window 35 COMMUNICATIONS ROOM COMMON ALARM A. Unit 3 ICS display:

{ 1. SYSTEM MIMiCS

2. ANNUNCIATOR MIMICS

3. ANNUNCIATOR 8A

4. COMM ROOM COMMON ALARM

5. Determine alarm point in alarm by EOR point number with RED ALM for point QUAL.

When sent to investigate inform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (51U) 23

NRC Scenario 1 / Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, C rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8B Window 17 Window 17 480V UNITBD 3B UV OR XFR Automatic action A. Undervoltage Trip (loss of following):

EHC hyd fluid system Pump 3B Stator cooling water Pump 3B
Motor suction Pump RFPT 3B 3B1 Main Oil Pump
Cond vac pump 3B
Lose one of two 480V AC power supplies to Generator Voltage Regulator.

Voltage Regulator remains in service.

COMPLETE listing on reference drawings.

B. Auto Transfer:

Alternate breaker closes
No loss of equipment A. VERIFY automatic action has occurred.

B. INSPECT 480V Unit Bd 3B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. C. REFER TO 0-OI-57B to re-energize board. D. REFER TO appropriate 01 for recovery or realignment of equipment. When sent to investigate inform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (51U). Unit Board 3B failed to transfer 24

NRC Scenario 1 ( Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8C Window 22, 23, and 30 Window 22 480V TB VENT BD 3B UV OR XFR A. VERIFY alarm by checking associated annunciator, TURBNE BLDG VENTILATION ABNORMAL, (3-XA-55-3D, Window 4) in alarm. D B. DISPATCH personnel to TB VENT Bd 3B to check equipment, board status, and abnormal conditions.

Turbine spaces supply fans 3A and 3B
Turbine room supply fans 3C, 3D, 3E
Turbine room exhaust fans 3E, 3F, 3G, 3H, 33
Electrical space supply fan 3A, 3B, and exhaust fan C. REFER TO 0-OI-57B to re-energize or transfer the board.

D. REFER TO appropriate 01 for recovery or realignment of equipment. Window 23 480V TURB MOV BD 3B UV OR XFR A. VERIFY alarm by checking light indication in Control Room for the following equipment:

Steam packing exhauster blower 3B
RFPT injection water pump 3B
RFPT oil tanks vapor extractor
RFPT 3A 3A2 Main Oil Pump
RFPT 3C 3C2 Main Oil Pump B. NOTIFY Radwaste of loss of the following:
Station sump pump 3A
Turb Bldg equipment and floor drain sump pump 3B
Turb Bldg cnds pump pit equipment and floor drain sump pump 3B C. CHECK board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.

D. REFER TO 0-OI-57B to re-energize or transfer the board. E. REFER TO appropriate 01 for recovery or realignment of equipment.

NRC Scenario 1 Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8C Window 22, 23, and 30 Window 30 480V TURB MOV BD 3C UV OR XFR A. VERIFY alarm by checking light indication in Control Room for the following equipment:

112 main seal oil pump
H2 seal oil vacuum pump
H2 recirculating seal oil pump
Steam packing exhauster blower 3A B. CHECK MOV board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.

C. REFER TO 0-OI-57B to re-energize or transfer the board. D. REFER TO appropriate 01 for recovery or realignment of equipment. SRO Will call outside US and determine if board can be re-energized

           .       When sent to investigate inform crew that 480V TURB MOV BD 3C transferred. DO NOT Driver              .

re-energize 480V Unit BD At NRC direction insert trigger 3 for Off Gas Malfunction when hydrogen is above 5% and Driver at NRC_direction_initiate trigger_5_for Hydrogen Explosion 26

NRC Scenario 1 / Simulator Event Guide: Event 6 Component: OFF-GAS H2 RECOMB1NER CATALYST FAILURE BOP Responds to alarm the following alarms: HIGH OFFGAS % H2 TRAiN A (9-53, Window 3) HIGH OFFGAS % H2 TRAiN B (9-53, Window 13) 112 WATER CHEMISTRY ABNORMAL, (9-53, window 10) BOP Reports a rise in hydrogen concentration on OFF GAS HYDROGEN ANALYZERs (CH 1-Analyzer 3A, CH 2-Analyzer 3B) recorder, 3-XR-66-103/1 and 2, Panel 9-53. SRO Enters 3-AOI-66-1, Off-Gas H2 High. 4.1 Immediate Actions None 4.2 Subsequent Actions [1] PLACE both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK! BYPASS control switches, 3-HS-066-1007 and 1008, on OFFGAS SAMPLE PANEL, 3-LPNL-925-0588, in BYPASS to assure continuous availability of hydrogen monitoring. [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 and the 02 injection rate should match the setpoint on the OIU which should be half of the H2 injection rate during normal steady state conditions.) [2.2] IF 112 and 02 injection rates do NOT meet the above conditions, THEN [2.3] IF oxygen concentration as indicated on OFFGAS TRAIN A or TRAiN B SAMPLE 02 iNDICATOR, 3-021-066-105 1 or 1052, on 3-LPNL-925-0588 indicates less than 5% oxygen AND an automatic HWC shutdown has NOT occurred, THEN INITIATE a HWC System shut down using either, 3-HS-4-40A 112 WATER CHEMISTRY CONTROL [Panel 3-9-53], 3-HS-4-40B 112 WATER CHEMISTRY CONTROL [Panel 3-9-5] or 3-HS-4-39 HWC SHUTDOWN SWITCH [3-LPNL-925-0588]. When directed wait 2 minutes and report that both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK / BYPASS control switches, 3-HS-066-1007 and 1008 are in bypass. Driver Report that H2 and 02 injection rates are normal. When called to report 02 concentration for step 2.3 above report a concentration of 4.5%. 27

NRC Scenario 1 Simulator Event Guide: Event 6 Component: OFF-GAS H2 RECOMBINER CATALYST FAILURE [3] VERIFY proper operation of in service SJAE. (Steam jet may have failed to isolate on low supply steam pressure.) [3.1] IF a failure of the in service SJAE is indicated AND hydrogen concentration is less than 4%, THEN PLACE standby SJAE in service. REFER TO 3-01-66 (otherwise N/A) [4] IF hydrogen concentration is 4%, THEN REFER TO TRM 3.7.2 SRO TR 3.7.2 Airborne Effluents LCO 3.7.2 Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to 4% by volume. APPLICABILITY: During main condenser offgas treatment system operation Condition A: With the concentration of hydrogen >4% by volume. Required Action A. 1: Restore the concentration to within the limit. Completion Time: 48 hours When directed wait 2 mmutes and report that both OFFGAS TRAIN A(B) AUTO Driver CHANNEL CHECK / BYPASS control switches, 3-HS-066-1007 and 1008 are in bypass Report that 112 and 02 injection rates are normal 28

NRC Scenario 1 Simulator Event Guide: Event 6 Component: OFF-GAS H2 RECOMBINER CATALYST FAILURE NOTE Fuel failure is indicated by, but NOT limited to, rising activity on the following:

OFF-GAS PRETREAT MENT 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).

[6] WHEN any of the following conditions exist OBTAIN US approval, THEN SCRAM the Reactor (REFER TO 3-AOI-100-1):

Hydrogen ignition/explosion as indicated by rising temperature, and/or pressure, and/or flow in the Off-Gas System which may cause some or all of the following annunciators to alarm:

OFF-GAS HOLDUP VOLUME PRESS HIGH, (3-XA-55-7A, Window 1) OFF-GAS HOLDUP VOLUME TEMP HIGH, (3-XA-55-7A, Window 2)

OG CHARCOAL BED VESSEL TEMP NIGH, (3-XA-55-53, Window 9)
HOLDUP LINE INLET FLOW HIGH, (3-XA-55-53, Window 14)
CHARCOAL BED VAULT TEMP HIGH, (3-XA-55-53, Window 19)
CHARCOAL BED GAS REHTR OUTL DEW PT TEMP HIGH, (3-XA-55-53, Window 27)
CHARCOAL TRAiN INLET/OUTLET PRESS HIGH, (3-XA-55-53, Window 28)
Charcoal combustion in Adsorber beds as indicated by rising temperature and/or radiation in the Off-Gas System which may cause some or all of the following windows to alarm:
CHARCOAL BED VESSEL TEMP HIGH, (3-XA-55-53, Window 9)

                              . CHARCOAL BED VAULT TEMP HIGH, (3-XA-55-53, Window 19)

OG POST TRTMT RADIATION HIGH, (3-XA-55-4C, Window 33)
Unit Supervisor direction The Following alarms will come in when Off Gas Explosion occurs: Panel 7A window NRC 1,4,9; Panel 4C window 33,34,35; Panel 3A window 6,13; Panel 53 window 4,14,8,9,18,27,28 SRO Directs_Reactor_SCRAM_and_enters_3-AOI-lOO-1 Driver Just before scram or as scram is directed msert trigger 15 for loss of feedwater 29

NRC Scenario 1 Simulator Event Guide: Event 6 Component: OFF-GAS H2 RECOMBINER CATALYST FAILURE BOP Responds to annunciator 9-7A window 1, OFF-GAS HOLDUP VOLUME PRESS HIGH A. CHECK off-gas flow recorder, 3-FR-66-20 on Panel 3-9-8 and pressure to off-gas preheater 3-PI-66-71 on Panel 3-9-53. B. CHECK for indications of H2 ignition: 1.0FF GAS HOLDUP VOLUME TEMP HIGH alarm, 3-XA-55-7A, Window 2. 2.0FF GAS SAMPLE FLOW ABNORMAL alarm, 3-XA-55-3A, Window 33.

3. Various abnormal indications or alarms NOT limited to, but may include one or more of the following on Panel 3-9-53:

a. GAS RHTR OUTLET DEW POINT HIGH TEMP alarm, 3-XA-55-53, Window 27.

b. Upscale reading on Recorder 3-TRS-66-108.

c. Adsorber Vault (3-TRS-66-120) or Adsorber Vessel (3-TRS-66-1 15) high temperature.

d. Lowering inlet recombiner temperature 3-TI-66-75A(B).

C. IF indication of H2 ignition is present, THEN VERIFY the following valves are closed:

3-FCV-66-14, 18 SJAE discharge valves.

3-FCV-66-11, 15- SJAE Air inlet valves.
3-PCV-1-151,153,166,167 SJAE steam pressure control valves.

3-FCV-1-155,156,172,173 SJAE steam isolation valves.

3-FSV-1-150,152 SJAE inter-condenser drain valves, AND REFER TO 3-AOI-66-1.

SRO Directs Reactor SCRAM and enters 3-AOl- 100-1 Dnver Just before scram or as scram is directed msert trIgger 15 for loss of feedwater 30

NRC Scenario 1 Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater SRO Directs Reactor SCRAM and enters 3-AOI-100-1 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5A/S3B, on Panel 3-9-5. [2] II? scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A) [3] Refuel Mode One Rod Permissive Light check [3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) [4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in SHUTDOWN. [5] REPORT the following status to the US:

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

[3] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit. [3.1] DOWNRANGE IRMs as necessary to follow power as it lowers. [4] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5. [5] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, as follows: Crew Reports loss of Unit Board 3C ATC Report Loss of all Condensate Booster Pumps 31

NRC Scenario 1 Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater SRO Enters EOI- 1 on Reactor Level SRO EOI- 1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO

IF Emergency Depressurization is or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO-Is any MSRV Cycling? No IF Steam cooling is required? NO-( IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO

IF Drywell Control air becomes unavailable? NO-IF Boron injection is required? NO-SRO Directs a Pressure Band with MS1Vs SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs_Verification_of PCIS_isolations. ATC/BOP Verifies PCIS isolations. Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with SRO the following injection source. (RCIC, App SC) ATC Initiates RCIC lAW App 5C 32

NRC Scenario 1 I Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater ATC/BOP Maintain Directed Level Band with RCIC, Appendix SC.

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

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

5. OPEN the following valves:

                               . 3-FCV-71-39, RCIC PUMP INJECTION VALVE
                               . 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE
                               . 3-FCV-71-25, RCIC LUBE OIL COOLING WTRVLV.

6. PLACE 3-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.

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

8. CHECK proper RCIC operation by observing the following:

(

a. RCIC Turbine speed accelerates above 2100 rpm.

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

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

d. 3-FCV-71-34, RCIC PUMP MTh4 FLOW VALVE, closes as flow rises above 120 gpm.

9. IF BOTH of the following exist? NO

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

33

NRC Scenario 1 ( Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater Crew Respond to_Turbine_Building Radiation High A. DETERMINE area with high radiation level on Panel 3-9-il. Report Numerous Turbine Building areas B. IF the TSC is NOT manned, THEN USE public address system to evacuate area where high airborne conditions exist. Respond to MAIN STEAM LiNE RADIATION HIGH A. CHECK following radiation recorders:

1. MAiN STEAM L]NE RADIATION, 3-RR-90-135_on Panel 3-9-2.

Respond to RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 3-9-11. RX BLDG 565 Area 3-RI-90-21A D. IF the TSC is NOT manned and a VALID radiological condition exists., THEN USE public address system to evacuate area where high radiological conditions exist I. ENTER 3-EOI-3 Flowchart. Enters EOI-3 on Secondary Containment Radiation SRO SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO If Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels SRO are below 72 mr/hr Then Restart Reactor Zone and Refuel Zone Ventilation. Directs Restart of Reactor Zone and Refuel Zone Exhaust Ventilation (App 8F) SRO Directs_Defeat_of isolation_interlocks_(App_8E) Calls for Appendix 8E and when complete Restarts Reactor Zone and Refuel Zone Exhaust BOP/ATC Ventilation Is Any Area Radiation Level 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?       NO, NO System Discharging Before any area radiation rises to Max Safe (table 4) Continue and enter EOI- 1 (EOI- 1_has_afready_been_entered_after Reactor_Scram)

Crew Monitors for Max Safe Radiation and reports ( DRiVER If Called for Appendix 8E, Wait 4 minutes, enter bat appO8e and report complete 34

NRC Scenario I ( Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater ATC/BOP Appendix 8F RESTORING REFUEL ZONE AND REACTOR ZONE VENTILATION

FANS FOLLOWING GROUP 6 ISOLATION VERIFY PCIS Reset.

2. PLACE Refuel Zone Ventilation in service as follows (Panel 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 1NBD ISOL DMPR
3-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR
3-FCO-64-lO, REFUEL ZONE EXH INBD ISOL DMPR.

3. PLACE Reactor Zone Ventilation in service as follows (Panel 3-9-25):

a. VERIFY 3-HS-64-l 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.

b. PLACE 3-HS-64-l 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 1A, REACTOR ZONE FANS ANT) DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

3-FCO-64-l3, REACTOR ZONE SPLY OUTBD ISOL DMPR
3-FCO-64-14, REACTOR ZONE SPLY INED ISOL DMPR
3-FCO-64-42, REACTOR ZONE EXH INBD ISOL DMPR
3-FCO-64-43, REACTOR ZONE EXH OUTBD ISOL DMPR.

35

NRC Scenario 1 f Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater Crew Report nsmg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 1 60F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES ( Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) 36

NRC Scenario 1 Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 37

NRC Scenario 1 Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 38

NRC Scenario 1 Simulator Event Guide: Event 7 Major: SCRAM and Loss of Feedwater Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (14% on control room indicators continue, does not continue Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WI-lEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -l in. and -6 in. (APPX 18) Can suppr p1 lvl be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES 39

NRC Scenario 1 Simulator Event Guide: Event 8 Component: RHR Pumps 3A and 3C Trip ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

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

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

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

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

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

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

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

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

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

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

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

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

ATC/BOP When RI{R Pumps 3A and 3C are started they will trip, operator will report and align Loop 2 of RHR 40

________ _______ NRC Scenario 1 ( Simulator Event Guide: Event 8 Component: RHR Pumps 3A and 3C Trip ATC/BOP 3-EOI APPENDIX-17C, R[1R System Operation Suppression Chamber Sprays

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

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

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

k. MONITOR RHR Pump NPSH using Attachment 2.

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

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

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

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

6. WHEN EITHER of the following exists:

Before Suppression Pool pressure drops below 0 psig, OR
Directed by SRO to stop Suppression Chamber Sprays, THEN STOP Suppression Chamber Sprays as follows:

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

b. VERIFY CLOSED 3-FCV-74-100, RHR SYS I U-2 DISCH XTIE

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

d. STOP RHR Pumps 3A and 3C (3B and 3D).

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

41

NRC Scenario 1 Simulator Event Guide: Event 8 Component: RHR Pumps 3A and 3C Trip = ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

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

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

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

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

6. INITIATE Drywell Sprays as follows:

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

b. ll EITHER of the following exists:

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

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

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

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

f. OPEN the following valves:

3-FCV-74-60(74), RIIR SYS 1(11) DW SPRAY OUTBD VLV
3-FCV-74-61(75), RUR SYS 1(11) DW SPRAY INBD VLV.

ATC/BOP When RHR Pumps 3A and 3C are started they will trip, operator will report and align Loop 2 of RHR 42

NRC Scenario 1 ( Simulator Event Guide: Event 8 Component: R}IR Pumps 3A and 3C Trip ATC/BOP 3-EO1 APPENDIX-17B, RHR System Operation Drywell Sprays

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

h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System 1(11) RHR Pump in service.

i. MONITOR RHR Pump NPSH using Attachment 2.

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

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

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

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

7. WHEN EITHER of the following exists:

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

a. VERIFY CLOSED the following valves:

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

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

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

d. STOP RHR Pumps 3A and 3C (3B and 3D).

43

NRC Scenario 1 7 Simulator Event Guide: Event 9 Major: Fuel Failure ATC/BOP Report MAIN STEAM LINE RADIATION HIGH-HIGH A. VERIFY alarm on 3-RM-90-136 thru 137 on Panel 3-9-10. B. CONFIRM main steam line radiation level on recorder 3-RR 135, Panel 3-9-2. C. IF alarm is VALID and scram has NOT occurred, THEN PERFORM the following: IF core flow is above 60%, THEN

1. LOWER core flow to between 50-60%.

2. MANUALLY SCRAM the Reactor.

3. REFER TO 3-AOI-100-1.

D. IF SLC injection per RC/Q of EOI-1 is NOT required, THEN VERIFY the MSIVs closed. SRO Directs MSIVs Closed and transitions pressure control to Appendix 1 1A ATC/BOP Close MS1Vs and commences pressure control with Appendix 1 1A

44

_________ ___________ NRC Scenario 1 Simulator Event Guide: Event 9 Major: Fuel Failure ATC/BOP Commence pressure control with Appendix 1 1A, Alternate RPV Pressure Control Systems MSRVs

1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CA]) TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.

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

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

a. 1 3-PCV-1-179 MN STM LiNE A RELIEF VALVE.

b. 2 3-PCV-1-180 MN STM LiNE D RELIEF VALVE.

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

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

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

f. 6 3-PCV-1-42 MN STM LINE D RELIEF VALVE.

g. 7 3-PCV-1-30 MN STM LINE C RELIEF VALVE.

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

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

j. 10 3-PCV-l-41 MN STM LINED RELIEF VALVE.

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

1. 123-PCV-l-18 MN STM LINE B RELIEF VALVE.

m. 13 3-PCV-l-34 MN STM LINE C RELIEF VALVE.

45

NRC Scenario 1 [ Simulator Event Guide: Event 9 Major: Fuel Failure ATC/BOP Report RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH AND RCIC STEAM LiNE LEAK DETECTION TEMP HIGH RCIC STEAM LINE LEAK DETECTION TEMP HIGH A. CHECK following instrumentation:

1. RCIC temperature elements on LEAK DETECTION SYSTEM TEMPERATURE Recorder, 1-TR-69-29 (Points 9-12) on 1-9-21.

2. CS / RCIC ROOM EL. 519 RX BLDG, 1-RI-90-26A, on Panel 1-9-11.

3. RX REFUEL ZONE EXHAUST RADIATION, 1-RR-90-144, on Panel 1-9-2 B. IF RCIC is NOT in service AND 1-FI-71-1A(B), RCIC STEAM FLOW indicates flow, THEN ISOLATE RCIC AND verify temperatures lowering.

C. IF high temperature is confirmed, ThEN ENTER 1-EOI-3 Flowchart. ATC/BOP Reports rising temperature in RCIC Area, If temperature continues to rise it will cause a RCIC Isolation @ 165°F in Torus area or RCIC Pump Room temp. SRO Re-Enters EOI-3 46

NRC Scenario 1 ( Simulator Event Guide: Event 9 Major: Fuel Failure RE Enters EOI-3 on Secondary Containment Temperature SRO SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO 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. SRO Directs Restart of Reactor Zone and Refuel Zone Exhaust Ventilation (App 8F) Directs_Defeat_of isolation_interlocks_(App_8E) BOP/ATC Calls for Appendix 8E and when complete Restarts Reactor Zone and Refuel Zone Exhaust Ventilation SRO Monitors and Controls Secondary Containment Temps Operate Available ventilation (APPX 8F) Defeat Isolation interlocks if necessary (APPX 8E) Is any Area Temp Above MAX Normal, Yes RCIC Room Isolate all systems that are discharging into the area except systems required to:

Be operated by EOIs OR
Suppress a Fire ATC/BOP Attempt to Isolate RCIC, RCIC fails to isolate SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before ANY Area Temp Rises to MAX Safe Continue, Completed When Temps in 2 or more areas are above MAX Safe Then Continue, Does Not Continue in the Temperature leg, NO area at or above MAX Safe Temperature Enters EOI-3 on Secondary Containment Radiation SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before any area radiation rises to Max Safe (table 4) Continue When Radiation Levels in 2 or more Areas are Above MAX Safe Then Continue, Report Two areas are above MAX SAFE Radiation Levels 3-RI-90-26A CS/RCIC Room ATC/BOP and_3-RI-90-29A_Supp_Pool Area_519_Elev.

SRO Continues to Emergency RPV Depressurization 3-C-2 47

NRC Scenario 1 Simulator Event Guide: ( Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure SRO Emergency RPV Depressurization 3 -C-2 RPV Water Level CANNOT be determined, NO Containment Water Level CANNOT be Maintained Below 44 Feet, NO DW Control Air Becomes Unavailable, NO Will the Reactor remain subcritical without boron under all conditions, YES Is DW pressure above 2.4 psig, YES Prevent Injection from ONLY those CS and LPCI Pumps NOT required to assure adequate core cooling (APPX 4), NOT required currently no injection systems aligned for injection Is Suppression Pool Level above 5.5 feet, YES Directs all ADS Valves Open ATC/BOP Opens six ADS Valves, reports indication on that ADS Valve 1-34 failed to open SRO Can 6 ADS Valves be Opened, NO Open Additional MSRVs as necessary to establish 6 MSRVs open ATC/BOP Open an additional MSRV, reports 3 MSRVs open 48

NRC Scenario 1 Simulator Event Guide: Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure SRO EOI-1 RPV Control SRO Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with the following injection source. Condensate APPX 6A, Core Spray APPX 6D, 6E SRO Can RPV Water Level be mamtained above 2 inches, Yes ATC/BOP Align injection sources directed by the SRO and Restores Level to 2 to 51 inches Injection Subsystems Lineup 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, RFP 3A DISCHARGE VALVE 3-FCV-3-12, RFP 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-3-3 1, 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, lIP 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, RFP 3A SUCTION VALVE 3-FCV-2-95, RFP 3B SUCTION VALVE 3-FCV-2-108, RFP 3C SUCTION VALVE. 49

NRC Scenario 1 Simulator Event Guide: Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure ATCIBOP Injection Subsystems Lineup Condensate

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).

11. VERIFY RFW flow to RPV.

ATC/BOP Injection Subsystems Lineup Core Spray System I

1. VERIFY OPEN the following valves:

3-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCT VLV 3-FCV-75-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 TBROTTLE 3-FCV-75-25, CORE SPRAY SYS I INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core_Spray Pump NPSH using Attachment_1.

50

NRC Scenario 1 Simulator Event Guide: Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure ATC/BOP Injection Subsystems Lineup Core Spray System II

1. VERIFY OPEN the following valves:

3-FCV-75-30, CORE SPRAY PUMP 3B SUPPR POOL SUCT VLV 3-FCV-75-39, CORE SPRAY PUMP 3D SUPPR POOL SUCT VLV 3-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.

2. VERIFY CLOSED 3-FCV-75-50, CORE SPRAY SYS II TEST VALVE.

3. VERIFY CS Pump 3B and/or 3D RUNNING.

4. WHEN RPV pressure is below 450 psig, TI{EN THROTTLE 3-FCV-75-53, CORE SPRAY SYS II INBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

SRO Emergency Classification EPIP-l 3.2-S An unisolable Primary System leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2. TABLE 3.2 MAXIMUM SAFE OPERATING AREA RADIATION LIMITS AREA RAD MONITOR MAX SAFE VALUE MRIHR UNITI UNIT2 UNIT3 RHR West Room 90-25A 1000 1000 1000 RHR East Room 90-28A 1000 1000 1000 HPCI Room 90-24A 1000 1000 1000 CSIRCIC Room 90-26A 1000 1000 1000 Core Spray Room 90-27A 1000 1000 1000 Suppr Pool Area 90-29A 1000 1000 1000 CRD-HCU West Area 90-20A 1000 1000 1000 CRD-HCU EastArea 90-21A 1000 1000 1000 TIP Drive Area 90-23A 1000 1000 1000 North RWCU System Area 90-13A 1000 1000 1000 South RWCU System Area 90-14A 1000 1000 1000 RWCU System Area 909A 1000 1000 1000 MG Set Area 90-4A 1000 1000 1000 Fuel Pool Area 90-lA 1000 1000 1000 Serqice FlrArea 90-2A 1000 1000 1000 New Fuel Storage 90-3A 1000 NA NA 51

NRC Scenario 1 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs: HPCI is Tagged Out Operations/Maintenance for the Shift: Unit 3 is at 70% Power. Condensate Booster Pump 3A has been placed back in service. RFPT 3A is warmed and operating at 600 rpm, place RFPT 3A in service in accordance with 3-01-3 Reactor Feedwater System section 5.7. An RPHP is effect for placing the RFPT in service, all data and signatures are recorded on Appendix A. Then commence power ascension with Control Rods and Flow in accordance with the reactivity control plan. Units 1 and 2 are at 100% power Unusual Conditions/Problem Areas: None 52

Facility: Browns Ferry NPP Scenario No.: NRC -2 Op-Test No.: 1306 Examiners:___________________ Operators: SRO:___________________ ATC:______________ BOP:______________ Initial Conditions: 4.5% power. 2-GOI-100-1A Section 5.4 Step 64 and 66 Turnover: Continue to pull rods to 8% power, verify IRM/APRM overlap at 5% and continue to 8% power and hold for Mode Change. Event Maif. No. lvent Type* Event Description No. R-ATC 1 Raise power with Control Rods R-SRO N-BOP Start SBGT Fan C and align to Reactor Zone Ventilation 2 TS-SRO lAW 0-01-65 section 5.2 C-ATC 3 imfrd06 Control Rod 50-11 and 18-59 difficult to withdraw C-SRO imf 1-BOP 4 Condenser Hotwell Level Automatic Makeup controller failure fw20a I-SRO C-BOP Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto 5 imfed07a C-SRO start 6 imfth3lb TS-SRO Pressure Transmitter PT-3-22BB, fails high C-ATC 7 ior RFPT C Trip C-SRO 8 tb23 M-ALL ATWS, Fuel Failure, LOCA RWCU Leak, failure of auto isolation only, high Secondary 9 cu06alb C Containment Radiation, ED not required RHR Loop 1 Outboard Injection Bypass switch fails, cannot 10 ior C terminate and prevent RHR Loop 1

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

NRC Scenario 2 Critical Tasks Six With a reactor scram required and the reactor not shutdown, initiate action to reduce power by inserting control rods.

1. Safety Significance:

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

2. Cues:

Procedural compliance CRD Pump B operating

3. Measured by:

Observation Control Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:

Reactor Power trend. Control Rod indications. 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. 9

NRC Scenario 2 Critical Tasks Six 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 Based on RPV pressure operate available system(s) to maintain RPV water level above T.A.F. (-162 inches).

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance. HPCI system aligned for injection Pressure below low pressure ECCS system(s) shutoff head 320 psig Pressure below Condensate Booster Pump shutoff head 425 psig

3. Measured by:

Operator manually aligns, starts initiates the available injection system and injects into the RPV to maintain water level above -162 inches.

4. Feedback:

Reactor water level trend. Reactor pressure trend. 3

NRC Scenario 2 Critical Tasks Six When MAiN STEAM LINE RADIATION HIGH-HIGH is received and SLC injection is NOT required in accordance with RC/Q of EOI- 1, MSIVs are closed.

1. Safety Significance:

Isolate potential discharge path to environment.

2. Cues:

Procedural compliance. Reactor Pressure trend.

3. Measured by:

When ARP-9-3A window 27 annunciator is received MSIVs are closed if alarm is verified valid and SLC injection JAW EOI-l RC/Q is NOT required.

4. Feedback:

Reactor Pressure trend. MSIVs indicate closed. During an ATWS, when pressure lowers below the shutoff head of operating injection systems, Terminate and Prevent RPV injection as required by EOI-C-5 to prevent uncontrolled reactor vessel injection.

1. Safety Significance:

Prevention of power spikes and fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation No Core Spray injection.

AND Observation Feedwater discharge valves indicate closed.

AND Observation RHR Pumps 2A and 2C are secured.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms. Injection system flow rates into RPV. 4

NRC Scenario 2 EVENTS

1. ATC Continues Power ascension with control rods to 8%. IRM/APRM overlap required to be verified at 5% power, prior to continued ascension to 8%. IRM B will be failed as is and during power ascension the ATC will identify that IRM B is not responding to increasing power.

2. BOP staxts SBGT Fan C and aligns to Reactor Bldg lAW 0-01-65 section 5.2. The relative humidity heater will fail to start and the SRO will evaluate Technical Specification 3.6.4.3 and determine Condition A is entered.

3. Control Rod 50-11 will not withdraw with normal methods from position 00, the ATC will take action lAW 2-01-85 for control difficult to withdraw. The ATC will double clutch control rod 50-11, the control rod will withdraw. Control Rod 18-59 will not withdraw with normal methods from position 12, the ATC will take action JAW 2-01-85 for control difficult to withdraw. The ATC will raise drive water pressure and the control rod will withdraw.

4. Condenser Hotwell Level Automatic Makeup controller will fail closed. The BOP operator will responds to alarms on Panel 9-6A-5, 6, 7 OR notice Hotwell Level on 2-LR-2-9 trending down or CST FLOW to Hotwell at 0 on 2-FT-2-48. The operator will take manual control and begin to restore hotwell level.

5. A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip with failure of standby pump to auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.

6. Pressure Transmitter PT-3-22BB, fails high. This will cause a Half Scram and a high Reactor Pressure annunciator. The SRO will evaluate Tech Specs 3.3.1.1 and determine condition A is entered.

7. RFPT 2C Trips, ATC increases the speed of RFPT 2B to maintain level in the normal level band.

SRO will enter 2-AOI-3 -1. While restoring Reactor Level if the restoration is too rapid, power oscillations will occur and the SRO may direct a Reactor Scram at this point.

8. After Reactor Level is restored a fuel failure will occur. Rising radiation levels will occur throughout the Unit, the SRO will determine when to scram. Eventually the Main Steam Line High-High will alarm requiring a Reactor Scram and MSIVs to be isolated if SLC is not required JAW the RC/Q leg of EOI- 1. An ATWS will result on the scram and the SRO will evaluate conditions.

9. Shortly after the scram RWCU will develop a small leak, with the fuel failure two areas in Secondary Containment will be at or above max safe radiation levels. The SRO will enter EOI-3 and evaluate the need to Emergency Depressurize. The crew will isolate RWCU and with no primary system discharging to Secondary Containment ED will not be required.

10. The LCPI System 1 Outboard Injection Valve Bypass switch will fail, RHR Loop 1 injection flow will not be able to terminated and prevented or controlled. RHR Pumps 2A and 2C will have to be tripped to prevent an uncontrolled injection to the reactor when Reactor Pressure lowers to below the shutoff head of the RHR System.

5

NRC Scenario 2 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner. Control Rods are being inserted Reactor Level is maintained RWCU has been isolated SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2 10 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 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 80 Validation Time (minutes) 6 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) 6

NRC Scenario 2 Scenario Tasks TASK NUMBER KJA RO SRO Manual Initiation of SBGT Fan C RO U-065-NO-02 261000A4.07 3.1 3.2 SRO S-000-A1)-27 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to Withdraw RO U-085-AB-07 201002A2.02 3.2 3.3 SRO S-085-AB-07 Hotwell Level Automatic Makeup controller failure RO U-002-AL-O1 201002A2.02 3.2 3.3 SRO S-002-AL-Oi EHC Pump Trip RO U-47A-AL-07 256000A2.06 3.2 3.2 SRO S-57B-NO-07 RFPT C Trip ROU-003-AB-O1 259001A2.O1 3.7 3.7 SRO S-003-AB-O1 RWCU Leak RU U-069-AL-09 295033EA1.05 3.9 4.0 SRO 5-000-EM-b 7

NRC Scenario 2 Scenario Tasks TASK NUMBER RQ $RQ Fuel Failure RO U-099-AL-05 272000A2.O1 3.7 4.1 SRO S-066-AB-02 ATWS RO U-000-EM-03 295015AA2.O1 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18 LOCA RO U-000-EM-O1 29503 1EA2.04 4.6 4.8 RO U-000-EM-05 RO U-000-EM-17 SRO 5-000-EM-Ol SRO S-000-EM-05 8

NRC Scenario 2 Procedures Used/Referenced: Procedure Number Procedure Title 0-01-65 Standby Gas Treatment System 2-GOl- 100-1 A Power Maneuvering 2-01-85 Control Rod Drive System 2-ARP-9-6A Panel 9-6 2-XA-55-6A 2-01-2 Condensate System 2-ARP-9-7A Panel 9-7 2-XA-55-7A 2-ARP-9-7B Panel 9-7 2-X.A-55-7B 2-ARP-9-8B Panel 9-7 2-XA-55-8B 2-ARP-9-8C Panel 9-7 2-XA-55-8C 2-ARP-9-5B Panel 9-5 2-XA-55-5B 2-ARP-9-4A Panel 9-4 2-XA-55-4A Technical Specifications 2-ARP-9-6C Panel 9-6 2-XA-55-6C 2-A0I-3 -1 Loss of Reactor Feedwater or Reactor Water Level 2-ARP-9-5A Panel 9-5 2-XA-55-5A 2-ARP-9-3A Panel 9-3 2-XA-55-3A 2-E0I-3 Secondary Containment Control 2-E0I-1 RPV Control 2-A0I-100-1 Reactor Scram 2-E0I-C-5 LevellPower Control 2-E0I Appendix-2 Defeating ARI Logic Trips 2-E0I Appendix-iF Manual Scram 2-E0I Appendix-iD Insert Control Rods Using Reactor Manual Control System 2-E0I Appendix-i iA Alternate RPV Pressure Control MSRVs 2-E0I-2 Primary Containment Control 2-E0I Appendix-6A Injection Subsystems Lineup Condensate 2-E0I Appendix-SD Injection System Lineup HPCI 2-E0I Appendix-SC Injection System Lineup RCIC 2-E0I Appendix-i 7C RHR System Operation Suppression Chamber Sprays 2-EOI Appendix-17B R}IR System Operation Drywell Sprays EPIP- 1 Emergency Classification Procedure 9

NRC Scenario 2 Simulator Instructor IC 104

Batch File NRC/l3O6nrc-2 imfcu06a imf cu06b trg e5 NRC/ehc trg e5= bat NRC/ehcpumptrip-1 trg elO NRC/modesw imfcu04 (elO 2:00) 70 180 0 RWCU Leak imfth2l (elO 6:00) .6 LOCA ior zdihs74l 55a normal imfrd06r50l 1 bat atws80 br ypomtrsbgthtrrh fail_control_power ior ypobkrrhfpmb fail_ccoil imfnm05b 21 imfrdO6rl859 Preference File NRC/l3O6nrc-2 pfk0l tog pfk 02 ann silence pflc 03 bat NRC/l3O6nrc-2 pfk 04 imf fw2O 0 Condenser hotwell level auto makeup failure pfk 05 imfed07a 480V unit bd 2A loss pfk 06 imfth3lb Pressure Transmitter PT-3-22BB fails high pfk07 pfk 08 ior zdihsO3 176 trip RFPT 2C Trip pfk 09 imfth23 15 180 0 Fuel Failure pfk 10 bat sdv pfk 11 bat app0lf p1k 12 bat app02 pfksl p1k s2 p1k s3 bat app08ae p1k s4 mrfrd06 close p1k s5 mrfrd06 open 10

NRC Scenario 2 Scenario 2 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 104 restorepref NRC/i 3O6nrc-2 Simulator Setup Load Batch F3 NRC/i3O6nrc-2 Simulator Setup Verify file loaded Verify Rod Worth Minimizer Working Verify IRM B on range 8, when the batch file is loaded range all IRMs upscale and then back to mask IRM B Failure, Dial 2-LC-2-3 Hotwell High Level Dumpback to an indication on 2-FT-2-47 of Manual about 151. Set 2-LC-2-3 at 30 in manual and 2-LC-2-6 at 30.5 in auto Swap lens covers on Bus Duct Cooling Fans Advance chart recorders in fast Insert pref key f4 as soon as the crew assumes the Manual shift._imf fw2O_0 RCP required (4% -8%), Marked up copy of 2-GOI-100-1A Unit Startup ii

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 SRO Direct Power increase using Control Rods per 2-GOI-100-1A, Section 5.4 5.4 Withdrawal of Control Rods while in Mode 2 [64] VERIFY IRMJAPRM overlap by operator visual observation before exceeding 5% power. [66] CONTINUE to withdraw control rods to raise Reactor power to approximately_8% per 2-01-85_and 2-SR-3.1.3.5(A). ATC Raise Power with Control Rods JAW 2-01-85, Section 6.6 Group 40 18-03 and 02-19 from 00 to 12 Group 41 = 10-11, 10-51, 50-51, 50-11 from 00 to 12 Group 42 02-43, 18-59, 42-59, 58-43, 58-19, 42-03, 18-03, 02-19 from 12 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 LATCI{ED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel inRPV).

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

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

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

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

12

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 [6] IF Control Rod movement was stopped to keep from exceeding a RBM Setpoint or ATC was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM: [6.1] PLACE the CRD Power, 2HS-85-46 to the OFF position to deselect the control Rod. [6.2] PLACE the CRD Power, 2-HS-85-46 to the ON position. [6.3] II? desired, THEN CONTINUE to withdraw Control Rods and PERFORM applicable section for Control Rod withdraw. 6.6.3 Control Rod Notch Withdrawal [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40. [2] OBSERVE the following for selected control rod:

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

[3] VERIFY ROD WORTH MINIMIZER operable and LATCHED in to correct ROD GROUP when Rod Worth Minimizer is enforcing. [4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [5] OBSERVE control rod settles into desired position A1[I) ROD SETTLE light extinguishes. 13

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [6] II? control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows: [6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [6.2] CUECK control rod coupled by observing the following:

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

14) does not alarm.

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

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

2) When in areas of high notch worth, single notch withdrawal should be used instead of continuous rod withdrawal. Information concerning high notch worth is identified by Reactor Engineering in Control Rod Coupling Integrity Check, 2-SR-3 .1.3 .5A.

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

Both switches should be released when the control rod reaches two notches prior to its intended position. (Example: If a control rod is to be withdrawn from position 00 to position 12, the CRD Notch Override Switch and the CRD Control Switch would be used to move the control rod until reaching position 08, then both switches would be released.)
If the rod settles in a notch prior to the intended position, the CR1) Control Switch should be used to withdraw the rod to the intended position. (using the above example; If the control rod settles at a notch prior to the intended position of 12, the CRD Control Switch would be used to withdraw the control rod to position 12.)

14

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40. [21 OBSERVE the following for selected control rod:

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

[3] VERIFY ROD WORTH MINIIvIIZER operable and LATCHED into correct ROD GROUP when 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 OVERRIDE. [5.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and 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. [5.5] WHEN control rod settles into the intended notch, THEN CHECK the following:

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

14) does not alarm.

[5.6] CHECK control rod settles at intended position and ROD SETTLE light extinguishes. ATC During power ascension IRM B fails to respond to continuous steady counts, ATC reports to Unit Supervisor SRO Directs Startup to continue, have all needed IRM instruments. Need 6 of 8 IRMs. 15

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 = ATC NOTE When continuously withdrawing a control rod to position 48, the control rod coupling integrity check can be performed by one of the two following methods:

1) Coupling integrity check while maintaining the CR1) Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position. If this method is selected, perform Step 6.6.4[6] and N/A Step 6.6.4[7].

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

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A) [6.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE. [6.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CR1) Control Switch in the Rod Out Notch position with the control rod at position 48. [6.4] CHECK control rod coupled by observing the following:

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

14) does not alarm.

[6.5] RELEASE both CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48. [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes. [6.7] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2. 16

NRC Scenario 2 Simulator Event Guide: Event 1 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [7] IF continuously withdrawing the control rod to position 48 control rod coupling integrity check is to be performed after the CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-8548 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A): [7.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE. [7.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [7.3] WHEN position 48 is reached, THEN RELEASE CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48. [7.4] VERIFY control rod settles into position 48. [7.5] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE. [7.6] CHECK control rod coupled by observing the following:

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

14) does not alanu.

[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes. [7.8] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2. 6.6.5 Return to Normal after Completion of Control Rod Withdrawal [1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN: [1.1] PLACE CR1) POWER, 2-HS-85-46, in OFF. [1.2] PLACE CR1) POWER, 2-HS-85-46, in ON. 17

NRC Scenario 2 Simulator Event Guide: Event 2 Normal: Start SBGT Fan C and align to Reactor Bldg LAW 0-01-65 section 5.2 SRO Directs Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 section 5.2 BOP Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 5.2 Standby Gas Treatment System Manual Initiation [1] VERIFY the following requirements are satisfied:

SGT Train A(B)(C) in standby readiness.
Main Stack Radiation Monitoring in Service.

[2] REVIEW the Precautions and Limitations in Section 3.0. [3] VERIFY suction path is aligned to SGT System as follows: [3.2] IF alignment to Reactor Zone Ventilation suction path is desired, THEN VERIFY OPEN the following dampers for the desired unit(s) to be aligned.

REACTOR ZONE EXH TO SGTS dampers, 2-HS-64-40 and 2-HS-64-41 on Panel 2-9-25

[4] START SGT FAN C as follows: [4.2] IF starting SGT FAN C from Panel 2-9-25, THEN PLACE SGTS FAN C, 0-HS-65-69A12 in START. [5] CHECK SGT TRAiN C INLET DAMPER as follows: [5.3] IF SGT FAN C was started, THEN CHECK OPEN SGTS TRAIN C INLET DAMPER, 0-HS-65-51A indicates OPEN on Panel 2-9-25. [6] CHECK SGT TRAIN C RH CONTROL HTR as follows: [6.2] IF SGT FAN C was started, THEN CHECK ENERGIZED SGTS TRAIN C RH CONTROL HTR, 0-HS-65-60 on Panel 2-9-25. [7] RECORD start time and filter bank differential pressure for SGT Train as follows: [7.2] IF SGT FAN C was started, THEN RECORD start time and FILTER BANK DIFFERENTIAL PRESSURE, 0-PDI-65-53 on Panel 2-9-25, in the Narrative Log. [8] DISPATCH Operator to the Standby Gas Treatment building as soon as time allows to check for abnormal conditions (i.e. belt tightness, rubbing or vibration noises). [9] MONITOR Standby Gas Treatment Train operation. REFER TO Section 6.0. BOP Reports failure of RH Heater 18

NRC Scenario 2 Simulator Event Guide: Event 2 Normal: Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 NRC NRC .nform the SRO that the relative humidity heater failed to energize,

                                      . .e. miner will noti the booth driver to call the SRO (as U0) and inform
         /Driver 111111 01 we pro ulern.

2-ARP-9-3B, Window 5 - SGT TRAiN C SWITCHES MISALIGNED A. CHECK each hand switch in normal operating position in accordance with 0 65, Attachment 2. B. If possible, CLEAR initiating signal. Otherwise REFER TO Tech Spec 3.6.4.3. C. NOTIFY UNIT SUPERVISORJSRO and Unit 1 and Unit 3. SRO SRO Evaluate Technical Specification 3.6.4.3 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3, During operations with a potential for draining the reactor vessel (OPDRV5). Condition A One SGT subsystem inoperable Required Action A. 1 Restore SGT subsystem to OPERABLE status Completion Time 7 Days NOTE: This LCO applies to ALL 3 UNITs 19

NRC Scenario 2 Simulator Event Guide: Event 3 Component: Control Rod 50-hand 18-59 difficult to withdraw 8.15 Control Rod Difficult to Withdraw ATC [1] VERIFY the control rod will not notch out. Refer to Section 6.6. [2] REVIEW all Precautions and Limitations in Section 3.0 [3] IF RWM is enforcing, ThEN VERIFY RWM is operable and LATCHED in to the correct ROD GROUP. NOTES

1) Steps 8.15[4] through 8.15[6] should be used when the control rod is at Position 00 while Step 8. 15 [7] should be used when the control rod is at OR between Positions 02 and 46.

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

3) Step 8.1 5[4] may be repeated as necessary until it is determined that this method will not free the control rod.

[4] IF the control rod problem is not believed to be air in the hydraulic system, THEN PERFORM the following to double clutch the control rod at Position 00: [4.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in EMERG ROD IN, for several seconds. [4.2] CHECK the control rod full in indication (double green dashes) on the Full Core Display for the associated control rod. [4.3] SIMULTANEOUSLY PLACE CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRIDE AND CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH. [4.4] WHEN EITHER of the following occur:

Control rod begins to move, OR
It is determined the rod will not move, THEN RELEASE 2-HS-85-47 AND 2-HS-85-48.

[4.5] IF the control rod successfully notches out, THEN PROCEED TO Section 6.6 and WITHDRAW the control rod to the appropriate position. ATC Will Double clutch Control Rod 50-11 from position 00 Driver when ATC goes to double clutch 50-11, delete stuck rod 50-11 20

NRC Scenario 2 Simulator Event Guide: Event 3 Component: Control Rod 50-hand 18-59 difficult to withdraw [7] IF the control rod is at or between Positions 02 and 46, THEN PERFORM the following to withdraw the control rod using elevated drive water pressure: [7.1] RAISE the CRD DRIVE WTR }ll)R DP, 2-PDI-85-17A, to 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. ATC [7.2] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48. [7.3] IF the control rod successfully notches out, THEN LOWER CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL [7.4] IF the control rod still fails to NOTCH OUT, THEN RAISE CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to 350 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. [7.5] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48. [7.6] LOWER CRD DRIVE WTR IIDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A. [7.7] IF the control rod still fails to NOTCH OUT using elevated CRD DRIVE WTR FIDR DP, THEN CONTACT Reactor Engineer and NOTIFY Unit Supervisor for further instructions. ATC Will raise drive water pressure to 350 psid to successfully move Control Rod 1 8-59 from position 12 Driver when ATC raises drive water pressure to 350 psid, delete stuck rod 1 8-59 21

NRC Scenario 2 Simulator Event Guide: Event 4 Instrument: Condenser Hotwell Level Automatic Makeup controller failure This malfunction will be in from the beginning of the scenario, hotwell level starts at 30 NRC inches and this alarm is at 24 inches. May notice before alarm and take action at that time Responds to Alarms on Panel 9-6A-5, 6, 7 OR notices Hotwell Level on 2-LR-2-9 trending BOP down or CST FLOW to Hotwell at 0 on 2-FT-2-48 6A-5, HOTWELL A LEVEL ABNORMAL 6A-6, HOTWELL B LEVEL ABNORMAL 6A-7, HOTWELL C LEVEL ABNORMAL A. VERIFY abnormal level on 2-LIC-2-3, Panel 2-9-6.

1. IF level is high, THEN RAISE hotwell reject.

2. IF level is low, THEN RAISE makeup.

B. ChECK condenser vacuum normal. C. CHECK:

1. Hotwell level recorder 2-LR-2-9C and 2-LIC-2-6.

2. Hotwell makeup flow 2-FR-2-48, Panel 2-9-6.

3. Hotwell dump flow 2-FR-2-47, Panel 2-9-6.

4. Conductivity rising recorder 2-CR-43-1A, Panel 2-9-6; recorder

2-CR-43-1 1AJ12A, Panel 2-9-4.

a. IF conductivity is rising, THEN REFER TO 2-AOI-2-1.

5. Bypass valves 2-LCV-2-4 and -7 closed.

6. Locally verify level in Hotwell level sightglass 2-LG-2-261.

D. IF cause is malfunction of automatic level control, THEN REFER TO 2-01-2, Section 8.3, Manual HW Level Control. BOP Refers to 2-01-2 Section 8.3 8.3 Manual Hotwell Level Control [3] IF desired to raise Hotwell Level, THEN PERFORM the following, as required: [3.1] THROTTLE OPEN HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LCV-2-6, using 2-LC-2-6 in MAN to maintain desired level. [3.2] OPEN CONDENSATE MAKEUP BYPASS VLV, 2-LCV-2-7, using 2-HS-2-7A, if necessary. [3.3] VERIFY CLOSED CONDENSATE DUMPBACK BYPASS VLV, 2-LCV-2-4, using 2-HS-2-4A. [3.4] THROTTLE CLOSED HOTWELL NIGH LEVEL DUMPBACK CONTROL, 2-LC-2-3, as necessary to raise Hotwell level. 22

NRC Scenario 2 Simulator Event Guide: Event 4 Instrument: Condenser Hotwell Level Automatic Makeup controller failure BOP Refers to 2-01-2 Section 8.3 8.3 Manual Hotwell Level Control [5] II? desired to maintain Hotwell Level, THEN: [5.1] PLACE HOTWELL IilGH LEVEL DUMPBACK CONTROL, 2-LC-2-3, in MAN and ESTABLISH approximately equal to or greater than 60 x 103 ibm/hr flow from the Hotweil to the CST as indicated on 2-FR-2-47. [5.2] PLACE HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, in MAN and ESTABLISH the required amount of flow from the CST to the Hotweil as necessary to stabilize desired HOTWELL LEVEL. [5.3] IF auto control of HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, is desired, THEN PLACE HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, in AUTO with controller set at desired HOTWELL LEVEL. 23

NRC Scenario 2 Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to mvestigate wait 3 mmutes and mform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alarms; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-l and 7B-15. 7A-22 GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action of this ARP 2-XA-55-9-8A Window 1 alarms THEN,

1. VERIFY all available Stator Cooling Water Pumps running.

2. Attempt to RESET alarm

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

Generator NOT on line, verifies Stator water cooling pump running 8B-16 480V UMT BD 2A UV OR XFR A. VERIFY automatic action has occurred. B. INSPECT 480V Unit Bd A for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. C. REFER TO O-OI-57B to re-energize board.

                   *D. REFER TO appropriate 01 for recovery or realignment of equipment.

Reports trip_of 480V_Unit Bd 2A,_dispatches_operators 8C-3 480V RX BLDG VENT BD 2A UV OR XFR A. VERIFY automatic action has occurred. B. CHECK or START refuel floor and reactor zone exhaust fans 2A or 2B. C. CHECK board for abnormal condition: relay targets, smoke, burned paint, breaker position, etc. D. REFER TO O-OI-57B to re-energize or transfer board. E. REFER TO appropriate 01 for recovery or realignment of equipment. 24

NRC Scenario 2 Simulator Event Guide: Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay. BOP Responds to the following alarms; 7A-22, 8C-3, 8C-lO, 8C-15, 8C-16, 8C-25, 8B-16, 7B-l and 7B-15. 8C-lO 480V RX BLDG VENT BD 2B UV OR XFR A. VERIFY automatic action has occurred. B. CHECK or START refuel floor and reactor zone fans 2A or 2B. C. CHECK 480V Reactor Bldg Vent Bd 2B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. D. REFER TO O-OI-57B to re-energize board. E. REFER TO appropriate 01 for recovery or realignment of equipment. 8C-15 480V TB VENT BD 2A UV OR XFR A. VERIFY alarm by checking the following:

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

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

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

C. REFER TO O-OI-57B to re-energize or transfer the board. D. REFER TO appropriate 01 for recovery or realignment of equipment. If called to verif if the following boards transferred report that these boards have auto

transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 25

NRC Scenario 2 Guide: Simulator Event rd 2A om po ne nt : Lo ss of 480V Unit Boa Event 5 C , when sent to en ce ke y F5 for imf edO7a has am re qu es te d by th e NRC in se rt pr ef er al Su pp ly B re ak er for Unit Board 2A When the Nor m Dnver minutes and inform investigate wait 3 ercurrent relay. -25, 8B-16, actuation of the ov -2 2, 8C -3 , 8C -1 O , 8C-15, 8C-16, 8C llowing alarms; 7A BOP Responds to the fo 7B-1 and 7B-15. OR XFR MOV BD 2A UV 8C-16 480V TURB uipment: lig ht in di ca tio n to the following eq checking ion valves. A. VERIFY alarm by ,2 C 1 ,2 C2 ) extraction isolat B 1 ,2 B2

RFW heaters (2 ain Oil Pump.
RFPT 2B 2B2 M ain Oil Pump.
RFPT 2C 2C1 M rned co nd iti on s: re la y targets, smoke, bu al spected for abnorm B. CHECK board in ion, etc.

paint, breaker posit OCESS A A L or V FD BAL and verify PR screen VFD C. REFER TO ICS Fa n Power status is O K. M In te rn al FI X ALAR r the board. 0- O F -5 7B to re -energize or transfe D. REFER TO uipment. fo r re co ve ry or realignment of eq opriate 01 E. REFER TO appr OR XFR 0V CN D S D EMIN BD 2 UV 8C-2 5 48 in Bd 2 to ch in g pe rs on ne l to 480V Cnds Dem at ic transfer by disp A. VERIFY automat wing: check for the follo . e lights illuminated ate disconnect 2A closed.

Power availabl op en an d al te rn ct switch 1A
Normal disconne on s such as breaker tri ps.

no rm al co nd iti

Any ab te Operator.

NOTIFY Radwas aste: B. sta tu s of th e fo llowing from Radw TAIN ailable, THEN OB C. IF power NOT av er at io n. s precoat op Condensate demin er at ion. wash transfer op

Condensate back .

er pit floor drains

Backwash receiv ctions.

re st or at io n or transfer instru 7B for po w er have auto REFER TO O-OI-5 fe rr ed re po rt that these boards D. ards tra ns X BLDG lle d to ve rif y if the following bo X B LD G V EN T BD 2A, 480V R If ca 480V R CDS fe rr ed to th ei r alternate supply. 0V TU R B M O V BD 2A, and 480V tra ns 2A, 48 0V TB VENT BD VENT BD 2B, 48

Dnye1 DEMIN BD 2 26

NRC Scenario 2 Simulator Event Guide: t Board 2A Event 5 Component: Loss of 480V Uni t ference key F5 for imf ed07a, when sen to Driver When requested by the NRC insert pre Uni t Boa rd 2A has an the Normal Supply Breaker for investigate wait 3 minutes and inform actuation of the overcurrent relay. 8B-16, 2, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, BOP Responds to the following alarms; 7A-2 7B-1 and 7B-15. LOW 7B-l EHC HYD FLUID HDR PRESS 2-HS-47-2A(1A) running. A. VERIFY Standby EHC PUMP 2B(2A), 1650 indicator, 2-PI-47-7 between 1550 and B. CHECK EHC HEADER PRESSURE psig. pump unit. C. DISPATCH personnel to inspect EHC VERIFY turbine trips at or below 1100 D. IF EHC Hydraulic system fails, THEN psig. ED 7B-15 STANDBY EHC PUMP FAIL A. On Panel 2-9-7: . ADER PRESSURE indicator, 2-PI-47-7

1. VERIFY alarm by checking EHC HE

( A 7-2A and/or EHC PUMP 2A, 2-HS-47-1

2. VERIFY EHC PUMP 2B, 2-HS-4 running.

R CURRENT 2-EI-47-2 and/or EHC

3. CHECK EHC PUMP 2B PUMP MT 1.

PUMP 2A PUMP MTR CURRENT 2-EI NOTE and illuminate at 1500 psig rising. Lights extinguish at 1300 psig lowering EHC 2A TEST pushbutton 2-HS-47-4A and

4. CHECK lights above EHC PUMP A.

PUMP 2B TEST pushbutton 2-HS-47-5 to check for abnormal conditions. B. DISPATCH personnel to pumping unit VERIFY turbine trips at or below 1100 C. IF EHC Hydraulic System fails, THEN psig. pressure returns to normal Start s Stan dby EHC Pump 2B and verifies EHC BOP in aux key F6, and when asked to investigate When directed by NRC insert preference Driver _failed high instrument room report 2-PIS-3-22BB_is 27

NRC Scenario 2 Simulator Event Guide: Event 6 Instrument: Pressure Transmitter PT-3-22BB, fails high SRO 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. Table 3.3.1.1-1 (page2of3) Reaotr Protecon System Irsthimertafon APPUCABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION 0.1

3. ReactorVessel Steam Dome 12 G SR 3.3.1.1.1 IOO psIg Pressure High(d)

SR 3.3.1.1.8 SR 3.3.1.1.10 SR 32,1.1.14 SRO Condition A: One or more required channels inoperable Required Action A. 1: Place channel in trip. Completion Time: 12 hours OR Required Action A.2: NOTE Not applicable for Functions 2.a, 2.b, 2.c, 2.d or 2.f. Place associated trip system in trip. Completion Time: 12 hours r When directed by NRC insert preference key F8 for a trip of RFPT 2C. After RFPT 2C is tripped wait 3 minutes after dispatched and report a scaffold crew removing scaffold from the area believes they may have tripped RFPT. 29

NRC Scenario 2 Simulator Event Guide: Event 7 Component: RFPT 2C Trip ATC Responds to a trip of RFPT 2C, reports Reactor level lowering, Reactor power lowering and reactor pressure stable. Responds to the following alarms 6C-15, 6C-23, 6C-29, 6C-32 and 5A-8 6C-15 RFPT C ABNORMAL A. CHECK other RFP alarms on Panel 2-9-6 to determine problem area. B. REFER TO appropriate alarm response procedure. 6C-23 RFPT TRIP CIRCUIT ABNORMAL A. VERIFY alarm and RFPT trip by checking Panel 2-9-6, RFPT speed, governor valve position and discharge flow. B. VERIFY reactor power is within the capacity of operating RFPs. C. IF BKR TRIPOUT PNL 2-9-9 DC DIST (2-XA-55-8C, alarm window 20) is illuminated, THEN Step is NA D. IF RFP is tripped, THEN REFER TO 2-01-3, Section 8.1 or 2-AOI-3-1. 6C-29 RFPT TRIPPED A. VERIFY reactor power is within the capacity of operating RFPs. B. CHECK core limits. C. WHEN RFPT coasts down to zero speed, unless RFPT is rolling on minimum flow, THEN VERIFY turning gear motor starts and engages. D. REFER TO_2-AOI-3-1_or 2-01-3,_Section_8.1. 6C-32 RFP DISCH FLOW LOW A. VERIFY reactor Feedpump flow on Panel 2-9-6. B. REFER TO 2-01-3, Section 8.7. 30

NRC Scenario 2 Simulator Event Guide: Event 7 Component: RFPT 2C Trip ATC Responds to a trip of RFPT 2C, reports Reactor level lowering, Reactor power lowering and reactor pressure stable. Responds to the following alanns 6C-15, 6C-23, 6C-29, 6C-32 and 5A-8 5A-8 REACTOR WATER LEVEL ABNORMAL A. VALIDATE Reactor water level hi/low using multiple indications including Average Narrow Range Level on 2-XR-3-53 recorder, 2-LI-3-53, 2-LI-3-60, 2-3 -206 and 2-LI-3-253 on Panel 2-9-5. B. IF alarm is valid, THEN REFER TO 2-AOI-3-1_or 2-01-3. SRO Directs entry to 2-AOI-3 -1, Loss of Reactor Feedwater or Reactor Water Level ATC 2-AOI-3-1 5.0 LOW REACTOR WATER LEVEL OR LOSS OF FEEDWATER [7] IF RFPs in manual control, THEN RAISE speed of operating RFPs. [9] IF RFPT has tripped and will not be required to maintain level, THEN REFER TO 2-01-3 and SHUT DOWN tripped RFPT. [11] IF unit remains on-line, THEN RETURN Reactor water level to 33 (normal range). ATC Takes control of RFPT 2B and raises RFPT speed and discharge pressure until RFPT 2B injects and raises and maintains RPV Level. NRC Power will drop to about 3% as level lowers, depending on how aggressive level is restored power may peak and cause and auto scram at 14% reactor power, or the SRO may direct a scram when he sees power oscillating depending on the actions of the ATC operator on level control. Driver When directed by NRC or if the crew scrams the reactor msert preference key F9 for fuel failure When Reactor SCRAM insert preference key Fl 0 for BAT SDV 31

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA BOP Respond to the following radiation alarms 3A-29, 3A-5, 3A-22 and 3A-7 Alarm 3A-29, TURBiNE BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 2-9-11. (Alarm on Panel 2-9-1 1 will automatically reset if radiation level lowers below setpoint.) B. IF the TSC is NOT manned, THEN USE public address system to evacuate area where high airborne conditions exist. D. NOTIFY RAD PRO. Alarm 3A-5, OG PRETREATMENT RADIATION HIGH A. VERIFY high radiation on following:

1. OFFGAS RADIATION recorder, 2-RR-90-266 on Panel 2-9-2.

2. OG PRETREATMENT RAP MON RTMR, 2-RM-90-157 on Panel 2-9-10.

3. OFFGAS RAD MON RTMR, 2-RM-90-160 on Panel 2-9-10.

B. CHECK off-gas flow normal. C. CHECK following radiation recorders and associated radiation monitors:

1. MAIN STEAM LINE RADIATION, 2-RR-90-135 on Panel 2-9-2.

2. OFFGAS POST-TREATMENT RADIATION, 2-RR-90-265 on Panel 2-9-2.

3. STACK GAS/CONT RM RADIATION, 0-RR-90-147 on Panel 1-9-2.

D. NOTIFY RAD PRO. E. NOTIFY Chemistry to perform radiochemical analysis to determine source 32

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA BOP Respond to the following radiation alarms 3A-29, 3A-5, 3A-22 and 3A-7 Alarm 3A-22, RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 2-9-11. (Alarm on Panel 2-9-11 will automatically reset if radiation level lowers below setpoint.) B. IF Dry Cask storage activities are in progress, THEN NOTIFY CASK Supervisor. C. IF alarm is from the HPCI Room while Flow testing is performed, NOTIFY personnel at the }IPCI Quad to validate conditions. D. NOTIFY RAD PRO. E. IF the TSC is NOT manned and a VALID radiological condition exists, THEN USE public address system to evacuate area where high radiological conditions exist. J. For all radiation indicators except FUEL STORAGE POOL radiation indicator, 2-RI-90-30, ENTER 2-EOI-3 Flowchart. BOP Reports RWCU South Area 90-14a and RWCU 621 Elev 90-9a in alarm and each are EOI 3 entry conditions SRO Announce Entry to EOI-3 BOP Alarm 3A-7, MAiN STEAM LINE RADIATION HIGH A. CHECK following radiation recorders:

1. MAIN STEAM LINE RADIATION, 2-RR-90-135 on Panel 2-9-2.

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

3. STACK GAS/CONT RM RADIATION, O-RR-90-147 on Panel 1-9-2.

B. NOTIFY RAD PRO. C. NOTIFY Chemistry to perform radiochemical analysis of primary coolant. 33

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA SRO Enters EOI-3 on Secondary Containment Radiation SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO 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. NO SRO WHEN Any Area Radiation Level Above Max Normal? - YES, proceeds 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?         NO, NO System Discharging Before any area radiation rises to Max Safe (table 4) Continue and enter EOI-1 (EOI- 1 has already been entered after Reactor Scram)

SRO If Reactor not SCRAMMED, Direct Scram Crew Monitors for Max Safe Radiation and reports 34

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA BOP Respond to the following radiation alarm 3A-27, MAIN STEAM LINE RADIATION HIGH-HIGH A. VERIFY the alarm on 2-RM-90-136 and 2-RM-90-137 on Panel 2-9-10. B. CONFIRM main steam line radiation level on recorder 2-RR 135, Panel 2-9-2. C. IF alarm is valid and Reactor Scram has not occurred, THEN PERFORM the following:

1. IF core flow is above 60% ThEN LOWER core flow to between 5 0-60%..

2. MANUALLY SCRAM the Reactor.

3. REFER to 2-AOI-100-1.

D. Li? SLC injection per RC/Q of EOI-1 is NOT required, THEN VERIFY the MSIVs closed. E. NOTIFY RAD PRO. SRO If have NOT directed SCRAM, directs SCRAM and entry to 2-AOl- 100-1, Reactor Scram ATC 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/53A and 2-HS-99-5A/S3B, on Panel 2-9-5. [2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA [3] REFUEL MODE ONE ROD PERMISSIVE light check: [3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-X1-85-46, is not illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) Step is NA [4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN. [5] REPORT the following status to the US:

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

( Dnver When Reactor SCRAM insert preference key FlO for BAT SDV 35

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATC 4.2 Subsequent Actions [1] ANNOUNCE Reactor SCRAM over PA system. [2] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following: [2.1] INITIATE ARI by Arming and Depressing BOTH of the following:

ARI Manual Initiate, 2-HS-68-1 19A
ARJ Manual Initiate, 2-HS-68-1 19B

[2.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed. [2.3] REPORT ATWS Actions Complete and power level. [3] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit. [3.11 DOWNRANGE IRMs as necessary to follow power as it lowers. [5] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, as follows: [5.2] IF required to maintain reactor water level, THEN TRIP reactor feed pumps as necessary to prevent exceeding High Reactor Water Level Trip setpoint. [5.5] IF Reactor Water Level exceeds +51 inches, THEN VERIFY TRIPPED the following turbines:

HPCI
RCIC

[5.6] IF Reactor Water Level exceeds +55 inches, THEN VERIFY TRIPPED the following turbines:

Main Turbine
Reactor Feed Pump Turbines SRO Enters EOI- 1 on RPV Water Level or Reactor Power 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 - 36

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA SRO EOI- 1 (Reactor Pressure) IF Emergency Depressurization is or has been required THEN exit RC/P and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO-Is any MSRV Cycling? 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 MSIV Closed and a Reactor Pressure Band with MSRVs (APPX) 1 1A SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs Verification of PCIS isolations. ATC/BOP Verifies PCIS isolations. SRO IF It has NOT been determined that the reactor will remain subcritical without boron under all condition THEN EXIT RC/L and ENTER C5, Level / Power Control SRO Enters C-5, Level/Power Control Inhibit ADS ATC/BOP lithibits ADS SRO Is any main steam line open No - Is reactor power above 5% or unknown No - Maintain RPV water level between -180 inches and +51 inches with the following injection sources: RCIC - APPX SC SRO Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with the following injection source. (RCIC, App 5C) ( Initiates RCIC lAW App SC ATC 37

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C.

1. VERIFY RESET and OPEN 2-FCV-71-9, RCIC TURB TRIP/THROT VALVE RESET.

2. 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 MIN FLOW VALVE
                             . 2-FCV-71-25, RCIC LUBE OIL COOLiNG WTRVLV.

6. PLACE 2-HS-71-3 1A, 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-71-40A, DISC POSITION, red light illuminated.

d. 2-FCV-71-34, RCIC PUMP MN 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.

38

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATC/BOP Commence pressure control with Appendix 1 1A, Alternate RPV Pressure Control Systems MSRVs

1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CA]) 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 by SRO:

a. 1 2-PCV-l-179 MN STM LINE A RELIEF VALVE.

b. 2 2-PCV-1-180 MN STM LINED RELIEF VALVE.

c. 3 2-PCV-1-4 MN STM LINE A RELIEF VALVE.

d. 4 2-PCV-1-31 MN STM LINE C RELIEF VALVE.

e. 5 2-PCV-1-23 MN STM LINE B RELIEF VALVE.

f. 6 2-PCV-1-42 MN STM LINE D RELIEF VALVE.

g. 7 2-PCV-1-30 MN STM LINE C RELIEF VALVE.

h. 8 2-PCV-l-19 MN STM LINE B RELIEF VALVE.

i. 9 2-PCV-1-5 MN STM LINE A RELIEF VALVE.

                          . 10 2-PCV-1-41 MN STM LINE D RELIEF VALVE.

k. 11 2-PCV-1-22 MN STM LINE B RELIEF VALVE.

1. 12 2-PCV-1-18 MN STM LINE B RELIEF VALVE.

m. 13 2-PCV-1-34 MN STM LINE C RELIEF VALVE.

NOTE Due to low power scenario Reactor pressure should be between 700 psig and 800 psig and 39

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA SRO EOI- 1 (Power) Monitor and Control Reactor Power Verif Reactor Mode Switch in shutdown Yes Initiate ARI completed Will tripping Recirc Pumps cause trip of main turbine, RFP, HPCI or RCIC No Is reactor power above 5% or unknown No - SLC Leg When periodic APRM oscillations greater than 25% peak to peak persist NO OR Before Suppression Pool temperature rises to 110°F NO-Insert Control Rods Leg Reset ARI and defeat ARI logic trip (APPX 2) Insert Control Rods using any of the following methods: APPX- 1 A Deenergize scram solenoids No APPX- 1 B Vent the scram air header No APPX-1C Scram individual control rods No APPX-1D Drive Control Rods Yes APPX- 1 E Vent control rod over piston No

APPX-1F Reset scram/RE-SCRAM Yes APPX- 1 G Raise CRD cooling water dp No - 40

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATC Inserting Control Rods Calls for 2-EOI Appendix-2 and the field portion of 2-EOI Appendix-iF 2-EOI Appendix-iF

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

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

4. DRAIN SDV UNTIL the following annunciators clear:

WEST CRD DISCH VOL WTR LVL NIGH 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 ISOL.

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

2-EOI Appendix-iD

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

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

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

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

a. SELECT control rod.

b. PLACE CR1) NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.

c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.

Driver When called for Appendix 2 wait 3 mmutes and preference key Fl 2, Appendix- iF wait 5 minutes and preference key Fl 1, If requested to close 85-5 86 preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 41

NRC Scenario 2 ( Simulator Event Guide: Event 9 Component: RWCU Leak, failure of Auto Isolation Reports alarm 3D-17, RWCU LEAK DETECTION TEMP HIGH ATC/BOP RE Enters EOI-3 on Secondary Containment Temperature SRO SRO Monitors and Controls Secondary Containment Temps Operate Available ventilation (APPX 8F) Defeat Isolation interlocks if necessary (APPX 8E) Is any Area Temp Above MAX Normal, Yes RWCU Room Isolate all systems that are discharging into the area except systems required to:

                         . Be operated by EOIs OR

Suppress a Fire Reports High Temperature RWCU Outboard Isolation Valve Area 2-TS-69-29F BOP Reports RWCU Radiation Monitors 90-14A and 90-13A at MAX Safe SRO Before ANY Area Temp Rises to MAX Safe Continue, Completed When Temps in 2 or more areas are above MAX Safe Then Continue, Does Not Continue in the Temperature leg, NO area at or above MAX Safe Temperature Enters EOI-3 on Secondary Containment Radiation SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before any area radiation rises to Max Safe (table 4) Continue When Radiation Levels in 2 or more Areas are Above MAX Safe Then Continue, ATC Reports RWCU failed to Auto Isolate, RWCU has been manually isolated with the Hand Switches Will Emergency Depressurization Reduce Discharge Into Secondary Containment? NO SRO Does NOT direct Emergency Depressurization, NO systems Discharging into Secondary Containment.

42

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 160F using available Drywell cooling Can Drywell Temperature be maintained below 160F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI-1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) 43

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 44

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18) Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES 45

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

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

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

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

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

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR
Directed by SRO, THEN PLACE keylock switch 2-XS 122(130), RHR SYS 1(11)

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

c. MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-53(67), RHR SYS 1(11) 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT VALVE.

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

f. VERIFY OPEN 2-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBRIPOOL ISOL VLV.

g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHER SPRAY VALVE.

ATCIBOP Aligns directed RHR Pumps in Suppression Chamber Sprays 46

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA ATCIBOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

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

i. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MN FLOW VALVE.

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

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RHRSW pump supplying desired R}{R Heat Exchanger(s).

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

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

n. NOTIFY Chemistry that RHRSW is aligned to in-service RRR Heat Exchangers.

ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 47

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA = ATC/BOP 2-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

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

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

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

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

6. INITIATE Drywell Sprays as follows:

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

b. iF EITHER of the following exists:

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

c. MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF 2-FCV-74-53(67), RIJR SYS 1(11) LPCI 114BD II4JECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

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

f. OPEN the following valves:

2-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBD VLV
2-FCV-74-61(75), RHR SYS 1(11) DW SPRAY INBD VLV.

ATC/BOP Aligns directed RHR Pumps in Drywell Sprays 48

NRC Scenario 2 Simulator Event Guide: Event 8 Major: Fuel Failure, ATWS, LOCA = ATCIBOP 2-EOI APPENDLX-17B, RHR System Operation Drywell Sprays

g. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MN FLOW VALVE.

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

i. MONITOR RHR Pump NPSH using Attachment 2.

j. VERIFY RI{RSW pump supplying desired RHR Heat Exchanger(s).

k. THROTTLE the following in-service RI{RSW outlet valves to obtain between 1350 and 4500 gpm R}{RSW flow:

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

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

7. WHEN EITHER of the following exists:

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

a. VERIFY CLOSED the following valves:

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

b. VERIFY OPEN 2-FCV-74-7(30), RHR SYSTEM 1(11) MN FLOW VALVE.

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

d. STOP RHR Pumps.

ATC/BOP Aligns directed RHR Pumps in Drywell Sprays 49

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails SRO C-5, LevelfPower Control Maintain RPV water level between -180 inches and +51 inches with the following injection SRO sources: may direct use of HPCI APPX 5D

BOP IF directed initiates HPCI JAW App 5D SRO Can RPV water level be restored and maintained above -180 inches YES-Direct additional injection sources to maintain RPV Level above -180 inches Condensate APPX 6A LPCI APPX 6B, 6C Directs Core Spray Terminated and Prevented BOP/ATC Operates directed systems to maintain RPV Level above -180 inches SRO Emergency Classification EPIP- 1 1.2-S Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical. When Reactor Pressure lowers to less than 450 psig, ECCS injection systems RHR and Core Spray will initiate. Core Spray should be terminated and prevented and RHR should NOTE be used to control RPV Level, RHR Loop 1 cannot be throttled and the pumps will have to be tripped. If RFPT discharge valves are not closed then condensate will inject uncontrollably. 50

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails BOP IF directed initiates HPCI lAW App 5D

4. VERIFY 2-IL-73-18B, HPCI TURBiNE TRIP RX LVL NIGH 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 5300 gpm.

7. PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in START.

8. PLACE 2-HS-73-1OA, NPCI STEAM PACKING EXHAUSTER, handswitch in START.

9. OPEN the following valves:

2-FCV-73-30, HPCI PUMP IvIlN FLOW VALVE
2-FCV-73-44, HPCI PUMP INJECTION VALVE.

10. OPEN 2-FCV-73-16, HPCI TURBiNE STEAM SUPPLY VLV, to start NPCI 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 5300 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.

51

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails BOP IF directed to inject with Condensate JAW App 6A

1. VERIFY CLOSED the following feedwater heater return valves:

2-FCV-3-71, HP HTR 2A1 LONG CYCLE TO CNDR 2-FCV-3-72, HP HTR2B1 LONG CYCLE TO CNDR

2-FCV-3-73, HP HTR 2C1 LONG CYCLE TO CNI)R

2. VERIFY CLOSED the following RFP discharge valves:

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

3. VERIFY OPEN the following drain cooler inlet valves:

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

4. VERIFY OPEN the following heater outlet valves:

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

5. VERIFY OPEN the following heater isolation valves:

2-FCV-3-38, HP HTR 2A2 FW iNLET ISOL VLV
2-FCV-3-3 1, lIP HTR 2B2 FW INLET ISOL VLV
2-FCV-3-24, HP HTR 2C2 FW INLET ISOL VLV
2-FCV-3-75, HP HTR 2A1 FW OUTLET ISOL VLV
2-FCV-3-76, HP HTR 2B1 FW OUTLET ISOL VLV
2-FCV-3-77, HP HTR 2C1 FW OUTLET ISOL VLV

6. VERIFY OPEN the following RFP suction valves:

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

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

9. ADJUST 2-LIC-3-53, RFW START-UP LEVEL CONTROL, to control 10.

VERIFY RFW flow to RPV. 52

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails BOP IF directed to inject with RHR Loop 1 lAW App 6B

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 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

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

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

4. VERIFY CLOSED the following valves:

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

5. VERIFY RHR Pump 2A andJor 2C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-53, RHR SYS I LPCI 1NBD INJECT VALVE.

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

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

Report 2-FCV-74-52 is NOT able to be throttled, Trips RHR Pumps 2A and 2C to prevent BOP/ATC uncontrolled injection 53

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails BOP IF directed to inject with RHR Loop 1 lAW App 6B

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 2-HS-74-155B, LPCI SYS II OUTBD lThTJ VLV BYPASS SEL in BYPASS.

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

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

4. VERIFY CLOSED the following valves:

2-FCV.-74-75, RRR SYS II DW SPRAY 1NBD VLV
2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
2-FCV-74-71, RHR SYS II SUPPR CHBR/POOL ISOL VLV
2-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE
2-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 2B and/or 2D running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.

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

8. THROTTLE 2-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE, as necessary to control injection.

54

NRC Scenario 2 Simulator Event Guide: Event 10 Component: RHR Loop 1 Outboard Injection Bypass switch fails BOP Core Spray Terminated and Prevented JAW App 4 NOTE Following receipt of a CORE SPRAY automatic initiation signal, it is NOT necessary to wait until a pump starts before performing step 3.

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

55

NRC Scenario 2 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs: None Operations/Maintenance for the Shift: 4.5% power. 2-GOT-lOU-lA Section 5.4 Step 64 and 66. Perform IRMJAPRM overlap before exceeding 5% power. RFPT 2B is warmed lAW 2-01-3 Section 5.6. Continue to pull rods to 8% lAW RCP. When Engineering is ready in the field need to start SBGT Fan C with alignment to Reactor Zone Ventilation lAW 0-01-65 section 5.2. Engineering will call. Units 1 and 3 are 100% Power Unusual Conditions/Problem Areas: Severe Thunderstorm Warnings are in affect for the entire area for the next 2 hours. 56

Facility: Browns Ferry NPP Scenario No.: NRC 3 Op-Test No.: 1306 Examiners: Operators: SRO: ATC: BOP: Initial Conditions: 95% power, CRD Pump 3A, EHC Pump 3A and B3 RHRSW Pump are tagged out. Turnover: Remove DG 3B from Parallel operation in accordance with 3-01-82. Raise power to 100% with flow. Event Maif. No. Event Type* Event Description No. R-ATC 1 Commence power increase with flow to 100% R-SRO N-BOP 2 IOR Remove DG 3B from Parallel operation lAW 3-01-82 TS-SRO 3 sw03m D3 EECW Pump Trip 4 RD22 CRD Flow Controller 3-FCV-85-1 lA, fails high C-ATC 5 Batch File RFPT 3C Trips, Reactor Recirc fails to runback C-SRO

                   . C-BOP       Loss of 4KV Unit BD 3C, Loss of Circ Water Pump 3C 6         Batch File                         .

C-SRO and the discharge valve fails to close automatically 7 Batch File M-ALL Station Blackout 8 RCO4 I RCIC Controller fails in Auto, Manual available 9 imfth2l M-ALL LOCA

      *  -  (N)ormal,  (R)eactivity, (I)nstrument,   (C)omponent,       (M)ajor cz4 V

NRC Scenario 3 Critical Tasks Three

With a loss of all Off Site Power and NO 480V SD BD energized, energizes 4KV SD BD 3EC from 4KVSDBD3ED.

1. Safety Significance:

Provides Power for ECCS Systems

2. Cues:

Procedural compliance 480V Shutdown Board 3B Energized, along with 480V RMOV Boards 3B, 3C, 3D, and 3E

3. Measured by:

Observation RO crossties to 4 KV SD BD 3ED to 3EC AND Observation 4 KV SD BD 3EC indicates energized

4. Feedback:

Power to 4 KV SD BD 3EC Powerto48OV SD BD 3B Prior to the Drywell Pressure accident signal actuating Operator has placed the following switches in Test/Inhibit, at Panel 2-9-3: ECCS SYS I HI DW PRESS Test/Inhibit, 2-HS-75-59 AND ECCS SYS II HI DW PRESS Test/Inhibit, 2-HS-75-60.

1. Safety Significance:

Prevent CAS initiation Load Shed due to high Drywell Pressure and lowering Reactor Pressure to maintain power to 4KV SD BD 3EC.

2. Cues:

Procedural compliance. No Load Shed when Drywell Pressure exceeds 2.45 psig and Reactor Pressure lowers to 450 psig

3. Measured by:

Observation 2-HS-75-59 and 60 in Test/Inhibit.

Observation No Load Shed, 4KV SD BD 3EC remains energized

4. Feedback:

Powerto4KV SDBD 3EC Power to 480V SD BD 3B 2

NRC Scenario 3 Critical Tasks Three

Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

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

AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment 3

NRC Scenario 3 Events

1. ATC commences power increase 100% using recirculation flow.

2. BOP remove DG 3B from Parallel operation lAW 3-01-82 section 8.1 step 17. A low lube oil pressure condition will occur on DG 3B, requiring an Emergency Shutdown of the DG 3B. SRO will evaluate TS 3.8.1 and Enter Condition B.

3. D3 EECW pump will trip, with B3 EECW pump tagged, the crew will respond lAW the ARP and manually start the Dl EECW pump. The SRO will refer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Dl EECW pump has been aligned the SRO will determine TS 3.7.1 Condition A now applies.

4. The ATC will respond to the CRD flow element failing high causing 3-FIC-85-1 1 CRD flow control valve to close. The ATC will take action to take manual control of the CR1) flow controller and restore CR1) system parameters. Depending on how long the crew takes to restore CR1) system parameters Tech Spec actions may be required if CRD temperatures increase high enough.

5. RFPT 3C will trip, with the RFPT trip and when level reaches 27 inches a Reactor Recirc runback should have occurred. The runback will fail to occur, RFPT 3A and 3B speed will increase to maintain level but much higher than the prescribed procedure limit of 5050 rpm. The ATC will have to manually initiate the runback or lower power manually.

6. The crew will respond to a Loss of 4KV Unit Board 3C. The 3C CCW Pump will trip, the Discharge valve will fail to close and the crew will close the discharge valve, to prevent a loss of vacuum. If the crew fails to close the discharge valve vacuum will quickly degrade requiring a SCRAM.

7. On the SCRAM or just before a Loss of Offsite Power will occur, due to equipment failures Unit 3 will enter a Station Blackout. No 4KV Shutdown Boards will be energized. The crew will take action lAW 0-AOI-5 7-lA and restore 4 KV SD BD B. RCIC and HPCI will be available for Level and Pressure control.

8. The RCIC flow controller will fail in automatic, the crew will take manual control to maintain Reactor level.

9. After the SCRAM a LOCA will develop requiring the crew to maintain RPV Level and Containment parameters JAW EOI-1 and EOI-2.

4

NRC Scenario 3 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Control Rods are inserted Reactor Level is restored and maintained SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3 9 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) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) 5

NRC Scenario 3 Scenario Tasks TASK NUMBER RO SRO Shutdown DG 3B RO U-082-NO-05 264000 A4.04 3.7 3.7 SRO S-082-NO-02 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 CRD Controller Failure RO U-085-AB-03 201001 A2.07 3.2 3.1 SRO S-085-AB-03 Loss Unit Board 3C/CCW Pump Trip RO U-047-AB-03 295002 AA1.07 3.1 2.9 SRO S-047-AB-03 EECW Pump Trip RO U-067-NO-12 400000 A2.01 3.3 3.4 SRO S-000-AD-27 RFPT Trip / Failure of Core Flow Runback ROU-003-AL-16 259001 A2.01 3.7 3.7 RO U-068-AL-20 SRO S-003-AB-01 Loss of Offsite Power/Station Blackout RO U-57A-AB-01 295003 AA1.03 4.4 4.4 SRO S-57A-AB-01 LOCA RO U-000-EM-01 295031 EA2.01 4.6 4.6 RO U-000-EM-05 SRO S-000-EM-01 SRO S-000-EM-05 6

NRC Scenario 3 Procedures Used/Referenced: Procedure Number Procedure Title 3-01-82 Standby Diesel Generator System 3-ARP-9-23B Panel 9-23 3-XA-55-23B 3-GOl- 100-12 Power Maneuvering 3-01-68 Reactor Recirculation System Technical Specifications 3-ARP-9-20A Panel 9-20 3-XA-55-20A 0-01-67 Emergency Equipment Cooling Water System 3-A0I-85-3 CRD System Failure 3-01-85 Control Rod Drive System 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3-ARP-9-6A Panel 9-6 3-XA-55-6C 3-A0I-3-1 Loss of Reactor Feedwater or Reactor Water Level High/Low 3-01-3 Reactor Feedwater System 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-A0I-47-3 Loss of Condenser Vacuum 3-A0I-100-1 Reactor Scram 0-A0I-57-1A Loss of Offsite Power (161 and 500 KV)/Station Blackout 3-A0I-30B-1 Reactor Building Ventilation Failure 3-A0I-78-1 Fuel Pool Cleanup System Failure 3-E0I-1 RPV Control 3-E0I Appendix-i 1A Alternate Pressure Control Systems MSRVs 3-E0I Appendix-5D Injection System Lineup HPCI 3-E0I Appendix-5C Injection System Lineup RCIC 3 -E0I Appendix-i 7C RHR System Operation Suppression Chamber Sprays 3-E0I Appendix-i 7B RHR System Operation Drywell Sprays 3 -E0I Appendix-i 7A RHR System Operation Suppression Pool Cooling 3-E0I Appendix-il C Alternate RPV Pressure Control Systems HPCI Test Mode 3-E0I-2 Primary Containment Control EPIP-3 Alert EPIP-4 Site Area Emergency EPIP- 1 Emergency Classification 7

NRC Scenario 3 Console Operator Instructions A. Scenario File Summary Batch File

  #tagout CRD 3A, EHC 3A, and B3 EECW ior ypobkrpumpa fail_ecoil ior zlohs85la[1] off ior zlohs47la[1] off ior zdihs47la[1] ptl ior ypobkrrhrswpb3 fail_ecoil ior zdihs2388a[2j stop ior zlohs2388a[1] off
  #DG 3B Loss of Lube Oil ior zlo3il82l3b (e30 0) on ior 0xa5523b[2] (e30 0) alarm on ior 0xa5523b[4] (e30 0) alarm on
  #EECW D3 Trip imfsw03m (ci 0) mrf swO5 (e2 0) close
  #CRD flow controller fails high imf rd22 (e4 0)100
   #RFPT 3C Trip! RR Fails to runback ior zdihsO3l76[1] (e8 0) trip imf fw26e (e8 0) 60
   #Circ water 3C pump trip and discharge valve fails to close, Loss of Unit BD 3C imfedO8c (e12 0) ior ypovfcv2729 fail_power_now ior zlohs2729a[21 on trg e13 nrcccwpumpC trg e13= bat nrcccwpumpC
   #Major imf edO 1 (e 18 0)               Loss of Off Site Power imfed09a (e18 0)                 Loss of 4KV SD BD 3EA imf dgo6c                       DG 3C fails to start trg e20 MODESW imfth2l (e20 1200) 1            LOCA imfth33d (e20 1200) 3.0 300 .5 imf rcO4 (e20 0)10              RCIC Flow controller Failure trg e23 = bat eecw trg e24 = bat eecw-1 trg e25 = bat rpsreset 8

NRC Scenario 3 trge26=batca Trigger Files nrcccwpumpC zdihs2729a[1] .eq. 1 MODESW ZDIHS465(4) .NE. 1 Scenario 3 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 202 Verify CR1) Pump 3B in service, EHC manual Pump 3B in service and EHC Pump 3A in Pull-to-Lock Simulator Setup Load Batch bat nrcl3O6-3 Simulator Setup Tag CRD 3A, EHC 3A and EECW Pump manual B3

Simulator Setup Verify file loaded RCP required (95% 100% with flow) and RCP for Urgent Load Reduction

Provide marked up copy of 3-GOI-100-12 Provide marked up copy of 3-01-82, section 8.1 up to step 17 Provide marked up Illustration 2 9

NRC Scenario 3 Simulator Event Guide: Event 1 Reactivity: Power increase with Recire Flow SRO Notifies ODS of power increase. Directs Power increase using Recirc Flow, per 3-GUI-i 00-12. [21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:

RAISE power using control rods or core flow changes.

REFER TO 3-SR-3.3.5(A) and 3-01-68. ATC Raise Power w/Recirc, lAW 3-01-68, Section 6.2 D. Individual pump speeds should be mismatched by 60 RPM during dual pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short time periods for testing or maintenance). [1] IF desired to control Recire Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;

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

AND/OR

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

[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required: RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-HS-96-32 iriver When directed byNRC;Triggørl to tripEECWPuthpD3, 10

NRC Scenario 3 Simulator Event Guide: Event 2 Normal/Component: Remove DG 3B from Parallel operation lAW 3-01-82 SRO Directs BOP to remove DG 3B from parallel operation JAW 3-01-82 section 8.1 BOP 8.1 Parallel with System Operation at Panel 9-23 [17] WHEN Parallel With System operation is no longer desired, ThEN UNLOAD the Diesel Generator as follows: CAUTION When unloading the Diesel Generator, failure to slowly approach the 300 kWI250 kVAR limit may result in a reverse power trip of the Diesel Generator output breaker. [17.1] USE the associated Diesel Generators governor control switch and voltage regulator control switch to reduce generator load to approximately 300 kW and 250 kVAR. DC 38 GOVERNOR CONTROL 3-HS-82-3B13A DG 38 VOLT REGULATOR CONT 3-HS-82-38/2A 3B 3-9-23 DG 3B KILOWATTS 3-JI-82-38/A DG 38 KILOVARS 3-VAR-82-38/A [17.2] PLACE the associated Diesel Generator breaker control switch in TRIP. DG3B BKR 1842 3-HS-21 1-3EA/9A 3-9-23 [17.3] PULL and PLACE the associated Diesel Generator mode selector switch in SINGLE UNIT. DG_3B_Mode_Select__3-HS-82-3B/5A__3-9-23 [17.4] RELEASE the Diesel Generator mode selector switch and OBSERVE the SINGLE UNIT light illuminated [17.5] RECORD the time/date unloaded on Illustration 2. [17.6] DISPATCH personnel to visually inspect the Diesel Generator output breaker to verify the closing springs are fully charged. Both the amber light and mechanical flag should be checked to indicate a charged spring. [18] if operation of the Diesel Generator is no longer required, THEN REFER TO Section_7.0_and_SHUT_DOWN the Diesel_Generator. DRIVER Prior to the operator shuttmg down the DG insert trigger 30, to cause a low lube oil condition and report significant oil leak from a damaged oil line. Unable to obtain oil pressure reading, cannot access panel. Delete override on amber light after Emergency Shutdown depressed. 11

NRC Scenario 3 Simulator Event Guide: Event 2 Normal/Component: Remove DG 3B from Parallel operation JAW 3-01-82 Responds to annunciators DIESEL GEN 3B TROUBLE and DIESEL GEN 3B LUBE OIL BOP ABNORMAL DIESEL GEN 3B TROUBLE A. CHECK panel 9-23 alarm panel for other alarms. B. DISPATCH personnel to diesel generator room to check alarm. DIESEL GEN 3B LUBE OIL ABNORMAL A. CHECK panel 9-23 to see if Low Low Lube Oil pressure light is illuminated. B. DISPATCH personnel to diesel generator room to check:

1. Oil pressure on local gauges for lube oil engine, lube oil filter inlet, and turbocharger compressor lube oil.

2. Scavenging, main, and piston cooling pumps running.

3. Oil level visible on dipstick.

4. Oil filters and strainers.

5. Any leakage.

C. SH1JT DOWN the diesel generator by EMERGENCY SHUTDOWN Pushbutton, if necessary. 7.4 Emergency Shutdown at Panel 9-23 [1] DEPRESS the associated Diesel Generator Emergency SHUTDOWN (STOP) push-button. DG 3B EMERGENCY SHUTDOWN 3-HS-82-3B/4 3-9-23 BOP [2] VERIFY OPEN DG 3B Output Bkr 1842. [3] DISPATCH personnel to locally VERIFY Diesel Generator stops. [6] INITIATE corrective action to return the Diesel Generator to an operable status. [7] RECORD time/date_stopped_on Illustration 2. Dnver Report DG is Stoppmg 12

NRC Scenario 3 Simulator Event Guide: Event 2 NormallComponent: Remove DG 3B from Parallel operation lAW 3-01-82 SRO Evaluate Tech Spec 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating

LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1 E AC Electrical Power Distribution System;

b. Unit 3 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 1 and 2 DG(s) capable of supplying the Unit 1 and 2 4.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems Operating.

APPLICABILITY: MODES 1,2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 6.2 Evaluate availability of 1 hour both temporaiy diesel generators (TDGs). AND AND Once per 12 hours therealter 8.3 Declare required 4 hours from feature(s), supported by discovery of the inoperable Unit 3 DG, Condition B inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable, redundant required feature(s) B. One required Unit 3 DG 8.1 Verify power availability 1 hour inoperable, from the otIsite transmission network, Once per 8 hours thereafter AND With B3 EECW Pump tagged out a 4 hour clock on EECW Pump C3 start with DO SRO inoperable. 13

NRC Scenario 3 Simulator Event Guide: Event 2 NormallComponent: Remove DG 3B from Parallel operation JAW 3-01-82 SRO Evaluate Tech Spec 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME AND B.4.1 Determine OPERABLE 24 hours Unit 3 DG(s) are not inoperable due to common cause failure. OR 6.4.2 Perform SR 3.8.1.1 for 24 hours OPERABLE Unit 3 DG(s). AND B. (continued) 6.5 Restore Unit 3 DG to 7 days from OPERABLE status. discovery of unavailability of TDG(s) AND 24 hours from discovery of Condition B entry 6 days concurrent with unavailability of TDG(s) AND 14 days 14

NRC Scenario 3 Simulator Event Guide: Event 3 Component: EECW pump D3 trip BOP Respond to Motor Trip Out annunciator. Report Trip of D3 EECW Pump, No EECW Pumps on the South Header C. The EECW System is aligned as follows:

1. At least one RHRSW pump, assigned to the EECW System, should be running on each header to maintain the header charged at all times. If no pumps are running on a header and header pressure lowers to 0 psig, the header shall be declared inoperable and appropriate actions taken, as required by Technical Specifications.

DRIVER If contacted as Unit 1 operator, you did not secure the D3 EECW Pump SRO Direct operator to place Dl RHRSW Pump in service to the South Header BOP 8.4 Operation of RIIRSW Pump Dl (for EECW in place of P3) CAUTION Only one RHRSW pump in a given RHRSW pump room may be counted toward meeting Technical Specification 3.7.2 requirements for EECW pump operability. NOTES

1) RHRSW Pump Dl may be aligned for service by this section when:

It is used to meet the minimum number of Tech. Spec. operable pumps; or
At the discretion of the Unit Supervisor, it is needed to replace another pumps operation; or
At the discretion of the Unit Supervisor, it is needed to assist in supplying header flow/pressure demand.

2) If used to meet EECW requirements, RHRSW pump Dl must be aligned to EECW, the pump started, and should remain running. RHRSW Pump Dl does NOT have the same auto start signals as RHRSW Pump D3.

3) Prior to aligning Dl RHRSW Pump to EECW, Technical Requirements Manual 3.5.2 must be reviewed to ensure Standby Coolant requirements are met.

4) When RHRSW Pump Dl is aligned for EECW, its RHRSW function required by the Safe Shutdown Program (Appendix R) is inoperable. Appendix R program equipment operability requirements of FPR-Volume 1 shall be addressed.

5) The RHRSW pump control switches and amp meters are located on Panel 9-3, Unit 1,2, and3.

15

NRC Scenario 3 Simulator Event Guide: Event 3 Component: EECW pump C3 trip 8.4 Operation of RIIRSW Pump Dl (for EECW in place of P3) [1] To line up RHRSW Pump Dl for EECW System operation, PERFORM the following: [1.1] VERIFY EECW System is in prestartup/standby readiness alignment in accordance with Section 4.0. [1.2] REVIEW all precautions and limitations in Section 3.0. [1.3] VERIFY RHRSW Pump Dl is in standby readiness in accordance with 0-01-23. [1.4] VERIFY RRRSW Pump Dl upper and lower motor bearing oil level is in the normal operating range. [1.5] UNLOCK and CLOSE RHRSW PMP Dl & D2 CROSSTIE, 0-23-563 at RHRSW D Room. [1.6] OPEN RHRSW PMP Dl CROSSTIE TO EECW, 0-FCV-67-48 using one of the following:

RHRSW PUMP Dl SUPPLY TO EECW, 0-HS-67-48A13 on Unit 3

[1.7] REQUEST a caution order be issued to tag RHRSW Pump Dl and its associated crosstie valves to infonu Operations personnel that it is aligned for EECW system operation and that the Dl pump should remain running to be operable for EECW. [2] To start RI{RSW (EECW) Pump Dl, PERFORM the following: [2.1] START RHRSW Pump Dl using one of the following:

RHRSW PUMP Dl, 0-HS-23-23A/3 on Unit 3

[2.2] VERIFY RHRSW Pump Dl running current is less than 53 amps using one the following:

RHRSW PUMP Dl AMPS, 0-EI-23-23/3 on Unit 3

[2.3] VERIFY locally, RI{R SERVICE WATER PUMP Cl breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged. [2.4] VERIFY RFLRSW Pump Cl upper and lower motor bearing oil level is in the normal operating range. Dnver If dispatched to check D3 EECW pump breaker, report breaker tripped on overload and breaker smells burnt but no visible smoke or flames (4kv SD BD D) 16

NRC Scenario 3 Simulator Event Guide: Event 3 Component: EECW pump C3 trip 8.4 Operation of RHRSW Pump Dl (for EECW in place of D3) [2.5] NOTIFY Chemistry of running R}IRSW (EECW) pump(s). [2.6] VERIFY a caution order has been issued to tag RHRSW Pump Dl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Dl pump should remain running to be operable for EECW. Driver When chemistry contacted, acknowledge report When contacted as Work Control for Caution Order, acknowledge direction and inform will begin working on a Caution Order When c ched as intake AUO to check Oil Levels and close 0-23-563 valve wait 2 minutes *d insert trigger 2 (mrf swOS close), then report oil levels are normal and the 0-23-56 vaLve is closed breaker charging spring recharged for the Dl EECW pump, wait ber breaker spring charged light is on and closing spring target 1aIu1aL uaj.cu i,i.ijx 4kv SD BD 3EB). When contacted as Intake AUO for second Oil Level check, report Oil Levels are normal Evaluate Technical Specification 3.7.2 before the Dl EECW Pump is aligned 3.7.2 Emergency Equipment Cooling Water (EECW) System and Ultimate Heat Sink (UHS) SRO LCO 3.7.2 The EECW System with three pumps and IJHS shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3. Condition A: Two required EECW pumps inoperable. (D3 and B3) Required Action A.l: Be in Mode 3 in 12 hrs, Mode 4 in 36 hrs Completion Time: 7 days In addition due to DG 3B Inoperable, 4 hour from DG 3B inoperability C3 EECW Pump SRO will have to be declared inoperable per Tech Spec 3.8.1 Condition B.3 and at that time Tech Spec_3.7.2_Condition A will_be_required_to_be entered. 17

NRC Scenario 3 Simulator Event Guide: Event 3 Component: EECW pump C3 trip Evaluate Technical Specification 3.7.1 after the Dl EECW Pump is aligned 3.7.1 Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS) LCO 3.7.1 NOTE The number of required RHRSW pumps may be reduced by one for each fueled unit that has been in MODE 4 or 5 for [1 24 hours. SRO Four RHRSW subsystems and UHS shall be OPERABLE with the number of OPERABLE pumps as listed below:

1. 1 unit fueled four OPERABLE RHRSW pumps.

2. 2 units fueled six OPERABLE RHRSW pumps.

3. 3 units fueled eight OPERABLE RHRSW pumps.

APPLICABILITY: MODES 1,2, and 3. Condition A: One required RHRSW pump inoperable Required Action A. 1: Only applicable for the 2 units fueled condition. (NA) Required Action A.2: Restore required RHRSW pump to OPERABLE status. Completion Time: 30 days In addition due to DG 3B Inoperable, 4 hour from DG 3B inoperability C3 EECW Pump SRO will have to be declared inoperable per Tech Spec 3.8.1 Condition B.3 and at that time Tech Spec_3.7.2 Condition A will be_required to be entered. 18

NRC Scenario 3 Simulator Event Guide: Event 4 Component: CRD flow element fails high causing 3-FIC-85-1 1 CRD flow control valve to close. Driver When directed by the NRC trigger 4 (imf rd22 100) to fail the CRD flow element ATC Report Alarm 5A-10 CR1) ACCUM CHG WTR HDR PRESS HIGH A. VERIFY pressure high on CR1) ACCUM CHG WTR FIDR 3-PI-85-13A, B. CHECK 3-FCV-85-1 1A (B) in service. C. IF in-service controller has failed, THEN REFER TO 3-01-85. D. IF pressure is still greater than 1510 psig after verifing proper controller operation, THEN THROTTLE PUMP DISCH THROTTLING, 3-THV-085-0527, to maintain between 1475 and 1500 psig. ATC Report CR]) controller has failed in Automatic, takes manual control and restores CRD Parameters SRO Directs ATC to take manual control of 3-FIC-85-1 1 and restore CRD parameters The crew may use OPDP-1 guidance listed below, or 3-01-85 Section 8.32 to take manual control of 3-FIC-85-1 1. OPDP-1 Conduct of Operations Exammer 3.5 Manual Control of Automatic Systems Note A. If an automatic control or an automatic action is confirmed to have ma1firnctinne1 fk prompt actions to place that control in manual or to accomplish the desired function. (e.g. Establishment of manual level control following automatic FCV failure to control level or manual start of an EDG that failed to auto start.) 8.32 AUTOMATIC/MANUAL Operation of 3-FIC-85-1 1 [1] REVIEW all Precautions and Limitations in Section 3.6. [2] IF transferring 3-FIC-85-1 1 from AUTO to MANUAL, THEN: [2.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1 in BAL. [2.2] BALANCE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, by turning ATC Manual Control Pot inside Control Selector Wheel until red deviation pointer is in the Green Band. [2.3] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, in MAN. [2.4] ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, manual potentiometer to establish the desired system flow. Refer to Section 5.1 or Section 6.11. 19

NRC Scenario 3 Simulator Event Guide: Event 5 Instrument: RFPT C Flow instrument fails as is on a trip of RFPT C, Reactor Recirculation System will fail to runback. I)river When directedbytheNRC trigger 8 to fripRFPT C iorzdihs03l76[1] trip andto fail FW flow element imf fw26e 60 ATC Reports a Trip of RFPT 3 C, and the following annunciators Panel 6C window 15, 23, and 29 RFPT C ABNORMAL, RFPT TRIP CIRCUIT ABNORMAL, RFPT TRIPPED RFPT C ABNORMAL A. CHECK other RFP alarms on Panel 3-9-6 to determine problem area. B. REFER TO appropriate alarm response procedure. RFPT TRIP CIRCUIT ABNORMAL A. VERIFY alarm and RFPT trip by checking Panel 3-9-6, RFPT speed, governor valve position and discharge flow. B. VERIFY reactor power is within the capacity of operating RFPs. D. IF REP is tripped, THEN REFER TO 3-01-3, Section 8.1 or 3-AOI-3-1. RFPT TRIPPED A. VERIFY reactor power is within the capacity of operating RFPs. B. CHECK core limits. C. WHEN RFPT coasts down to zero speed, unless RFPT is rolling on minimum flow, THEN VERIFY turning gear motor starts and engages. D. REFER TO 3-AOI-3-l or 3-01-3, Section 8.1. SRO Directs Entry to 3-AOI 1 Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-AOI-3 -1 Loss Of Reactor Feedwater or Reactor Water Level High/Low AUTOMATIC ACTIONS ATC Recirc Pumps receive run back signal to 75% speed at 27 (normal range) if the discharge flow of a REP is less than 889,000 lb/hr 19% (rated flow). [1] VERIFY applicable automatic actions. [2] IF level OR Feedwater flow is lowering due to loss of Condensate, Condensate Booster, or Feedwater Pump(s), THEN REDUCE Recirc flow as required to avoid scram on low level. 20

NRC Scenario 3 Simulator Event Guide: Event 5 Instrument: RFPT C Flow instrument fails as is on a trip of RFPT C, Reactor Recirculation System will fail to runback. ATC Reports annunciator REACTOR WATER LEVEL ABNORMAL and level is below 27 inches and slowly lowering RFPTs 3B and 3A speed is increasing to recover level. REACTOR WATER LEVEL ABNORMAL A. VERIFY Reactor water level hillow 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-l or 3-01-3. ATC Reports Power remained stable, pressure remained stable and level is recovering For operating Feed Pumps, monitor and maintain the following parameters within ranges described below. RFPT Speed: 5050 rpm maximum (3-9-6). ATC_reports_RFPT_3A and_3B_speed at greater than_5400_RPM SRO If ATC has NOT initiated a power reduction Directs a Core flow runback ATC Initiates a Mid Power Core Flow Runback, to lower RFPT speeds to less than 5050 RPM When called to report reason for KH I JU trip, wait minutes and report laborers in area, Driver they believe the tripped the RFPT. Delete override ior zdihs03 1 76[ 1] trip At direction of NRC trigger 12 Loss of 4kv Unit Board 3C, when called wait 4 minutes and report cause of Unit Board trip is actuation of Unit Board Feeder 86-XXX Lockout Driver Relay. When operator closes Circ Water Pump Discharge Valve ensure trigger 13 goes active 21

NRC Scenario 3 Simulator Event Guide: Event 6 Instrument: Loss of 4KV Unit BD 3 C, Loss of Circ Water Pump 3 C and the discharge valve fails to close automatically Crew Respond to 8B-14 4KV UNIT BD 3C UNDERVOLTAGE A. VERIFY Unit in stable condition by checking:

Condensate Pump 3C
Condensate Booster Pump 3C

                              *RCWPump3C

CCW Pump 3C

                              *CRDPump3A B.       IF undervoltage has occurred, THEN

1. CLEAR disagreement lights on breakers.

2. REDUCE load as necessary to maintain stable operating conditions.

3. Condenser discharge may need to be throttled for two CCW pump operation.

REFER TO 3-01-27. C. CHECK Unit Bd 3C for abnormal conditions: relay targets, smoke, burned paint, etc. D. REFER TO O-OI-57A to re-energize board. Report loss of 4KV Unit Board 3C ATC Reports loss of Condensate Pump 3C and Condensate Booster Pump 3C, Power Pressure Level stable BOP Will respond to a loss of Circ Water Pump 3C, report failure of discharge valve to close and close discharge valve ATC/BOP If discharge Valve is NOT closed, report degrading condenser Vacuum SRO Enters 3-AOI-47-3 Loss of Condenser Vacuum IF Operator fails to close CCW Pump 3C Discharge Valve Vacuum will quickly degrade causing a Turbine Trip and Reactor SCRAM. If the Turbine Trips insert trigger 18 loss of DRIVER Offsite Power. If the operator closes the discharge valve insert trigger 18 at the direction of the NRC 22

NRC Scenario 3 Simulator Event Guide: Event 6 Instrument: Loss of 4KV Unit BD 3C, Loss of Circ Water Pump 3C and the discharge valve fails to close automatically BOP/ATC 3-AOI-47-3 Loss of Condenser Vacuum [2] IF unable to maintain hotwell pressure below -25 inches Hg as indicated on 3-XR-2-2, with Reactor power less_than_30%,_THEN_TRIP the_main_turbine. [4] REDUCE reactor power in an attempt to maintain condenser vacuum. [5] VERIFY automatic actions. [6] CHECK CCW pumps for proper operation. If available, START additional CCW PUMPS. [7] VERIFY CLOSED CONDENSER VAC BREAKERS 1A AND lB. 3-HS-66-1A, Panel 9-8. [8] CHECK OFF-GAS FLOW TO 6-HOUR HOLDUP VOLUME, 3-FR-66-20, Panel 9-8, between 20 and 180 scfm. [9] VERIFY OPEN, 3-FCV-66-28, OFF-GAS SYSTEM ISOLATION VALVE. [10] IF SJAE 3A is in service , THEN VERIFY the following:

3-PCV-00l-0151 and 3-PCV-00l-0166 OPEN using STEAM TO SJAE 3A STAGES 1,2, AND 3, 3-ZI-l-151/166 on Panel 3-9-7.
SJAE 3A INTMD CONDENSER DRAIN, 3-ZI-l-150, on Panel 3-9-7, indicates OPEN.
3-FCV-066-001 1 OPEN using SJAE 3A INLET VALVE, 3-HS-66-1 1 on Panel 3-9-8.
Main Steam supply pressure at SJAE 3A STAGE I & II STEAM PRESS, 3-PI-001-0150, on 3-LPNL-925-0 105, is being maintained between 190 and 225 psig.

(TB EL 586 T12-C) IF Operator fails to close CCW Pump 3C Discharge Valve Vacuum will quickly degrade causing a Turbine Trip and Reactor SCRAM. If the Turbine Trips insert trigger 18 loss of DRIVER Offsite Power. If the operator closes the discharge valve insert trigger 18 at the direction of the NRC 23

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power lAW O-AOI-57-1A 4.1 Immediate Actions [1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators. [2] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators. [3] IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes). BOP Report only DG 3D is carry 4KV SD BD D, dispatches personnel to determine problems with DGs and SD Boards. Crew Verifies RHRSW Pump Dl Operating. IF NOT operating starts RHRSW Pump Dl within 8 minutes. Once Dl is started Operators have one hour to start an additional RI{RSW Pump not using the same strainer. SRO Declares a Station Blackout (SBO) is defined as a loss of 161 and 500kV systems and a failure of the two diesel generators which supply normal power to the two 480V Shutdown Boards on a unit. Currently NO 480V Shutdown Boards are energized. When sent to investigate 4KV SD BD 3EA has an actuation of overcurrent relay 51. DG 3A is operating normally. Unknown why DG 3C will not start contacting maintenance. If called Driver as Unit 1 or 2 report all 4 DGs on Unit 1 and 2 have started and are tied to thefr respective board. 24

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A Crew 4.2 Subsequent Actions [1] IF ANY EOI entry condition is met, THEN REFER to the appropriate E0I(s). [2] VERIFY automatic actions and PERFORM any that failed to occur. NOTES

1) If a Unit is in a Station Blackout condition, performance of this instruction will also require implementation of 1 (2)(3)-AOI-3 OB- 1, Reactor Building Ventilation Failure, on the Unit in Station Blackout.

2) EECW supply valves to the Control Air Compressors and RBCCW are air operated. If initial air pressure is low, air compressors may trip on high temperature, until cooling water flow is established.

3) At US discretion, the O-FCV-67-53 valve can be placed in the open position with hand switch. The valve will automatically come open once EECW pressure is above setpoint. REFER to 01-67 for valve operation.

4) The North header supply to Unit 1 RBCCW, the North header supply to Unit 2 RBCCW and the South header supply to Unit 3 RBCCW are normally isolated with a manual valve; therefore no flow will occur when either 1-FCV-67-50, 2-FCV-67-50 or 3-FCV-67-5 1 opens.

BOP/ATC 3-A0I-3 OB- 1, Reactor Building Ventilation Failure [2] IF this procedure is entered due to Station Blackout (O-A0I-57-1A) THEN GO TO Step 4.2[14]. NOTE Steps 4.2[14] ,4.2[15j and 4.2[16} may be entered directly from Step 4.2[2] if this procedure is entered due to Station Blackout (O-A0I-5 7-lA) [14] IF reactor building pressure CANNOT be maintained more negative than -0.25 inch 1120, THEN START Standby Gas Treatment. REFER TO 0-01-65. [15] IF the unit is in Mode 4 or Mode 5 THEN, Step 15 and 16 NA Unit not in Mode 4 or 5 BOP/ATC Verifies SGT Fans started 25

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power lAW O-AOI-57-1A [ Crew 4.2 Subsequent Actions [3] REFER to 1(2)(3)-AOI-78-1, FPC System Failure for a complete Loss of AC POWER, as necessaiy. NOT NECESSARY [4] WHEN EECW header pressure is restored above the reset pressure setpoint (psig) for the valves listed below, THEN Common Unit 1 Unit 2 Unit 3 0-FCV-67-53 106 FCV-67-50 - 90 91 92 FCV-67-51 - 107 109 113 RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel l-LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. Refer to the EECW to the RCW Crossties for Control Air & RBCCW section of 0-01-67. [5] START Control Air Compressors A, D and G as required and MONITOR system pressure. Refer to 0-AOI-3 2-1. [5.1] IF an air compressor trips on high temperature, THEN (Otherwise N/A) NOTII?Y Unit Supervisor for instructions. [6] REFER to 3-A0I-32-2, Loss of Control Air, as necessary [7] PLACE RPS MG Sets A and B in service. Refer to 3-01-99. Crew Calls for Control Air and EECW reset, calls for RPS Reset once 4KV SD BD 3EB energized or can request now but should inform Outside US power is not available for reset of RPS. Driver 3 mmutes after called for EECW trigger 23 bat eecw and trigger 24 bat eecw-1 and 4 mmutes_for Control Air trigger 26 bat ca Cannot reset RPS until operator has energized 4KV SD BD 3EB, when 4KV SD BD B is iver energized and 480 V SD BD 3B is energized wait 3minutes and trigger 25 bat rpsreset 26

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A NOTES

1) Station Blackout (SBO) is defmed as a loss of 161 and 500kV systems and a failure of the two diesel generators which supply normal power to the two 480 V Shutdown Boards on a unit. Exiting the SBO can occur through Cross-connect capabilities as long as it does not place the Non-SBO unit in jeopardy. Analysis takes credit for only one unit being in an SBO Event.

2) This section is to be performed if at any time during the loss of 161 and 500 kV Offsite power, the required Diesel Generators (for the Units 480 V Shutdown Bds) become inoperable thereby placing the unit in a SBO event. All times start with the recognition of an SBO Event, except for the time since shutdown.

3) The purpose of the alternate curves are to replace the normal curves (especially the Crew PSP curve) which would force an Emergency Depressurization (thus losing RCIC level control) before the end of the 4 hour coping period of the SBO analysis. Cooldown must be commenced as soon as possible at near maximum allowable rates to ensure that reactor pressure on the SBO unit is below 235 psig before 155 minutes have elapsed.

4) To support one unit in a LOOP/LOCA and two units in a LOOP, 6 RHR pumps, 2 Core Spray pumps, 6 RHRSW pumps and 2 EECW pumps are required long term (greater than 10 minutes). The units in the LOOP each require 2 RHR pumps and 2 RHRSW pumps in suppression pool cooling for long term cooling requirements until shutdown cooling can be placed in service where only 1 RHR pump and 1 RHRSW pump per non-accident unit is required. DG load management will ensure the 2 hour de-rated DG limit is not exceeded by manually removing non-required loads.

27

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A [8] IF the Unit(s) are under a Station Blackout THEN PERFORM the following: PROCEED TO Step 4.2[9] [8.1] ESTABLISH Level control with RCIC per the EOIs. [8.2] ESTABLISH Pressure Control with SRVs per the EOIs. [8.3] PRIOR to Reactor pressure decreasing below 450 psig, at panel 9-3, PLACE the following switches to TEST/INHIBIT: [8.3.1] ECCS SYS I HI DW PRESS TEST/INHIBIT, HS-75-59 [8.3.2] ECCS SYS I HI DW PRESS TEST/iNHIBIT, HS-75-60 [8.4] As soon as possible but within 60 minutes of the SBO event, INITIATE a cooldown at less than 90°F per hour in accordance with the EOIs until reactor pressure is between 150 and 230 psig. CREW [8.5] CONTROL Reactor Pressure between 150 and 230 psig using SRVs. NOTES

1) The following step will allow the use of SBO modified PSP and HCTL curves for EOI usage. The curves are only valid if both 155 minutes have elapsed since the time of reactor shutdown, AND the reactor pressure vessel is maintained below 235 psig.

2) If one curve is substituted, then BOTH curves are to be used.

3) The purpose of the alternate curves are to replace the normal curves (especially the PSP curve) which would force an Emergency Depressurization (thus losing RCIC level control) before the end of the 4 hour coping period of the SBO analysis.

[8.6] IF the EOI PSP Curve 6, or HCTL Curve 3, is close to being exceeded, AND the following conditions apply, (Otherwise N/A), Step is NA Directs Level control in RCIC and pressure control with SRVs, commences a cooldown SRO with SRVs. Direct ECCS Inhibit switches placed in l 1 nhibit Test Initiate RCIC for Level Control and Commence Cooldown with SRVs as directed. Inhibit BOP/ATC ECCS High Drywell Pressure signal. When RCIC started reports failure of Flow controller in Auto, takes manual control and BOP/ATC controls level with RCIC. 28

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power lAW 0-AOI-57-1A [9] START the Diesel Driven Fire Pump. Refer to 0-01-26. [11] IF containment isolation is required, THEN VERIFY the following containment isolation valves closed UNLESS they are required to be open by EOIs (RG 1.155): FCV-1-56 MN STM LINE OUTBD DRAiN ISOL FCV-69-2 RWCU OUTBD SUCT ISOLATION FCV-71-3 RCIC OUTBD SUCT ISOLATION FCV-71-18 RCIC SUPPR POOL OUTBD SUCT VALVE Crew FCV-73-3 HPCI STEAM LiNE OUTBD ISOL VALVE FCV-73-26 HPCI SUPPR POOL 1NBD SUCTION VLV FCV-73-30 HPCI MAIN PUMP MINIMUM FLOW VLV FCV-74-47 RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV NOTES

1) The UNIT SUPERVISOR should prioritize board energization to ensure common HVAC equipment powered from 480V boards is returned to service within 1 hour as directed by Attachment 5.

Request start of Diesel Fire Pump. Verifies Containment isolation status, RCIC valves will BOP/ATC be open When requested to start Diesel Fire Pump wait one minute and start diesel fire pump iif iver f04d start 29

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power lAW O-AOI-57-1A [12] VERIFY the following boards are energized. IF NOT, THEN REFER to Attachment_1_to_restore_affected_busses_while_continuing_with_this_instruction. Unit 1 Unit 2 Unit 3 4KV Shutdown Boards A, C B, D 3EA, 3EB, 3EC, 3ED 480V Shutdown Boards 1A, lB 2A, 2B 3A, 3B 480V DSL Aux Boards A B 3EA, 3EB 48OVRMOVBoards 1A,1B 2A,2B 3A,3B 480V Control Bay Vent A B Boards 480V HVAC Board B [13] VERIFY the following LPCI MG Sets in service to their respective Reactor MOV boards.

LPCI MG Sets 3D and 3E Crew Proceed to Attachment 1 to determine how to energize 4KV SD BD 3EB NOTE To ensure adequate cooling, required 480V loads should be re-energized within 1 hour of the event.

[19] VERIFY the following 4kV Shutdown Boards AUTO/LOCKOUT RESET switches in MANUAL:

U-3 4kV Shutdown Boards 3-43-21 1-3EA, 3-43-21 1-3EB, 3-43-21 1-3EC, 3-43-2 1 1-3ED.

[20] MAINTAIN diesel generator loading within the limits of Attachment 6. Crew NOTES

1) The following methods for reactor depressurization are listed in order of preference, but plant conditions may warrant other methods or a different order of preference.

HPCI system operating in CST to CST recirc.
RCIC system operating in CST to CST recirc.
Cycle SRVs.

[27] COMMENCE Reactor depressurization as soon as conditions permit. Cooldown is to_be_limited to_90°F/hr or_less_unless_otherwise_specified_by the_EOIs. SRO Direct HPCI place in pressure control if possible Place HPCI in Pressure control, if possible. If level is low can recover level with HPCI and ATC/BOP once level is high enough can reset start signal and place HCPI in pressure control mode until Drywell pressure exceeds 2.45 psig. 30

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A 1.0 RESTORATION OF ELECTRICAL BUSSES [1] DETERMINE restoration sequence of electrical busses and RE-ENERGIZE busses as necessary. Refer to Table 1. Table 1 Conditions Perform ENERGIZING 4Kv Shutdown Board 3EA using 4Kv Shutdown Board 3EB OR ENERGIZING 4Kv Shutdown Board 3EC using 4Kv Shutdown Board 3ED. Attachment 10 Driver If called for Temporary Diesels, They are not available Determines that Attachment 10 is the required attachment to restore power to a Shutdown C rew Board. Energize 4KV SD BD 3EC from DG 3D Attachment 10 ENERGIZING 4KV SD BD 3EA or 3EC during Station Blackout NOTES

1) This attachment is used to energize 4kV Shutdown Board 3EA using Shutdown Board 3EB or to energize 4Kv Shutdown Board 3EC using 4Kv Shutdown Board 3ED.

2) The use of 3EB-3EA and 3ED -3EC cross-ties are required to mitigate Unit 3 Station Blackout (SBO) scenarios (i.e. only 3B or 3D Diesel Generators available).

Crew 1.0 BOTH 4KV SHUTDOWN BOARDS 3EA AND 3EC ARE DE-ENERGIZED [1] IF both 4Kv Shutdown Boards 3EA and 3EC are de-energized, ThEN PERFORM the following: [1.1] DETERMINE which 4Kv Shutdown Board is energized (3EB or 3ED). [1.2] RE-ENERGIZE the desired shutdown board (3EA or 3EC) using the available_board_(3EB_or_3ED)_using_Step_1 .0[2]_or_STEP_1 .0[3]_as_applicable. SRO Emergency Plan classification 5.1-Al or 5.1-S 5.1 -Al Loss of voltage to ANY THREE unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND Only ONE source of power available to the remaining board. OPERATING CONDITION: Mode 1 or 2 or 3 5.1 -S Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes. OPERATING CONDITION: Mode 1 or 2 or 3 31

NRC Scenario 3 , Simulator Event Guide: Event 7 Component: Station Blackout Crew Attachment 10 ENERGIZING 4KV SD BD 3EA or 3EC during Station Blackout [3] IF desired to re-energize 4KV Shutdown Board 3EC using cross-tie from 4Kv Shutdown Board 3ED, THEN PERFORM the following: [3.1] VERIFY Diesel Generator 3D is supplying 4Kv Shutdown Board 3ED. [3.2] VERIFY 4Kv Shutdown Board 3EC is de-energized. [3.3] VERIFY OPEN 4Kv Bus Tie Board breaker 1632 (3-IL-210-1/6B). [3.4] VERIFY 4KV SD BD 3EC AUTO/LOCKOUT RESET switch, 3-43 -21 1-3EC, is tripped to MANUAL. [3.5] VERIFY 4KV SD BD 3ED AUTO/LOCKOUT RESET switch, 3-43 -21 1-3ED, is tripped to MANUAL. [3.6] PLACE in ON synchronizing switch 4KV SD BD 3ED BKR 1628 SYNC, 3-25-2 1 1-3ED/1A. [3.7] CLOSE 4KV SD BD 3ED ALT FDRBKR 1628, 3-HS-211-3ED/1A. [3.8] PLACE in OFF synchronizing switch 4KV SD BD 3ED BKR 1628 SYNC, 3-25-2 1 1-3ED/1A. [3.9] PLACE in ON synchronizing switch 4KV SD BD 3EC BKR 1626 SYNC, 3-25-2 1 1-3EC/3A. [3.10] CLOSE 4KV SD BD 3EC ALT FDR BKR 1626, 3-HS-21 1-3EC/3A. [3.11] PLACE in OFF synchronizing switch 4KV SD BD 3EC BKR 1626 SYNC, 3-25-2 1 1-3EC/3A. [3.12] VERIFY 4Kv_Shutdown Board 3EC_is energized. When 4KV SD BD 3EC is energized verifies 480V SD BD 3B energized and the following rew 480V RMOV BDs, 3C, 3D and 3E energize 32

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Attachment 9 Restoration of EECW Pumps NOTES

1) EECW Pumps may be restored by using one or both of the methods listed in this attachment.

2) Actions in this attachment should be continued until two EECW Pumps from different strainers are in service.

[ 1] SECURE any Diesel Generator prior to 8 minutes of operation without cooling water. [2) IF no EECW Pumps are in service supplying cooling water to the Diesel Generators, THEN PERFORM EITHER of the following to restore an EECW Pump electrically:

RE-ENERGIZE 4KV Shutdown Board C per Attachment 8, Energizing a Unit 1/2 4KV Shutdown Board using Another Unit 1/2 DG, THEN VERIFY EECW Pump B3 in service.
RE-ENERGIZE 4KV Shutdown Board D per Attachment 8, Energizing a Unit 1/2 4KV Shutdown Board using Another Unit 1/2 DG, THEN VERIFY EECW Pump D3 in service.

[3] SECURE any Diesel Generator prior to 1 hour of operation with only one EECW Pump supplying cooling water. Crew Verifies RHRSW Pump Dl Operating 33

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout Crew Attachment 9 Restoration of EECW Pumps 2.0 EECW PUMP CROSSTIE RESTORATION [1] CROSSTIE desired in-service Number One EECW Pump:

UNLOCK and OPEN R}{RSW PMP Al CROSSTIE TO EECW, O-SHV-067-0088 in RHRSW A Room.
OPEN RHRSW PMP Cl CROSSTIE TO EECW, O-FCV-067-0049 in U 1/2/3 MCR using one of the handswitches (if DG Aux Board A is energized), or ATC/BOP locally in RHRSW C Room using the handwheel.

[2] UNLOCK and CLOSE desired in-service Number One EECW Pump crosstie valve to Number Two EECW Pump:

RHRSW PMP Al & A2 CROSSTIE, O-SHV-23-504 in RHRSW A Room.
RHRSW PMJ Cl & C2 CROSSTIE, O-SHV-23-544 in RHRSW C Room.

Places second EECW Pump in operation from a different strainer, should start either Al or Crew If Cl is the RHRSW pump when called wait two minutes and mrf swO6 open, call and report RHRSW PMP Cl & C2 CROSSTTE, O-SHV-23 -544 CLOSED. Driver If Al is the RHRSW Pump when called wait two minutes and mrf swO7 align and report RHRSW PMP Al CROSSTIE TO EECW, O-SHV-067-0088 is OPEN and RIIRSW PMP Al & A2 CROSSTIE, O-SHV-23-504 is CLOSED 34

NRC Scenario 3 Simulator Event Guide: Event 7 Component: Station Blackout SRO Enters EOI- 1 on Reactor Level SRO EOI- 1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO - IF Emergency Depressurization is or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO - Is any MSRV Cycling? YES, but MS1Vs closed IF Steam cooling is required? NO-IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO-IF Drywell Control air becomes unavailable? NO-IF Boron injection is required? NO-SRO Directs a Pressure Band with SRVs LAW APPX 1 1A, and a controlled cooldown lAW Station Blackout. In addition may direct pressure control with HPCI (APPX 1 1C) SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs_Verification_of PCIS_isolations. ATCIBOP Verifies PCIS isolations. SRO Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with the following injection source. (RCIC, App 5C) may also direct at this time HPCI (App 5D), when the LOCA starts HPCI will be directed to maintain level (+) 2 to (+) 51 inches ATC Initiates RCIC lAW App 5C 35

NRC Scenario 3 Simulator Event Guide: Event 8 Instrument: RCIC Controller fails in Auto, Manual available ATCIBOP Maintain Directed Level Band with RCIC, Appendix 5C.

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

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

5. OPEN the following valves:

                            . 3-FCV-71-39, RCIC PUMP INJECTION VALVE
                            . 3-PCV-71-34, RCIC PUMP MIN FLOW VALVE
                            . 3-FCV-71-25, RCIC LUBE OIL COOLiNG WTR VLV.

6. PLACE 3-HS-71-3 1A, RCIC VACUUM PUMP, handswitch in START.

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

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

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

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

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

9. IF BOTH of the following exist? NO-

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

Reports level controller malfunction and takes manual control to inject with RCIC 36

NRC Scenario 3 Simulator Event Guide: Event 8 Instrument: RCIC Controller fails in Auto, Manual available ATC/BOP Maintain Directed Level Band with HPCI, Appendix 5D. CAUTION Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.

Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.

4. VERIFY 3-IL-73-18B, HPCI TURBINE TRIP RX LVL HIGH amber light extinguished.

5. VERIFY at least one SGTS train in operation.

6. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.

7. PLACE 3-HS-73-47A, HPCI AUXiLIARY OIL PUMP, handswitch in START.

8. PLACE 3-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, handswitch in START.

9. OPEN the following valves:

3-FCV-73-30, HPCI PUMP MIN FLOW VALVE
3-FCV-73-44, HPCI PUMP INJECTION VALVE.

10. OPEN 3-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. 3-FCV-73-45, HPCI TESTABLE CHECK VLV, opens by observing 3-ZI-73-45A, DISC POSITION, red light illuminated.

c. HPCI flow to RPV stabilizes and is controlled automatically at 5300 gpm.

d. 3-FCV-73-30, HPCI PUMP MN 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 3-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in AUTO.

14. ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

37

NRC Scenario 3 Simulator Event Guide: Event 7 Major: Station Blackout ATC/BOP Commence pressure control with Appendix 11 A, Alternate RPV Pressure Control Systems MSRVs

1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.

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

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

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

b. 23-PCV-1-180 MN STM LINED RELIEF VALVE.

c. 3 3-PCV-1-4 MN STM LiNE A RELiEF VALVE.

d. 4 3-PCV-1-31 MN STML1NE C RELIEF VALVE.

e. 53-PCV-1-23 MN STM LINE B RELIEF VALVE.

f. 6 3-PCV-1-42 MN STM LINE D RELIEF VALVE.

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

h. 83-PCV-1-19 MN STM LINE B RELIEF VALVE.

i. 9 3-PCV-i-S MN STM LiNE A RELIEF VALVE.

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

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

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

m. 13 3-PCV-1-34 MN STM LINE C RELIEF VALVE.

38

NRC Scenario 3 Simulator Event Guide: Event 7 Major: Station Blackout ATCIBOP Commence pressure control with Appendix 11 C, Alternate RPV Pressure Control Systems HPCI TEST MODE CAUTION

Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.
Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.

4. IF HPCI Turbine is operating, THEN ALIGN HPCI in test mode as follows:

a. OPEN 3-FCV-73-35, HPCI PUMP CST TEST VLV.

b. OPEN 3-FCV-73-36, HPCI/RCIC CST TEST VLV.

c. CLOSE 3-FCV-73-44, HPCI PUMP INJECTION VALVE.

d. CONTINUE in this procedure at Step 6.

5. IF UPCI is in standby readiness, THEN START HPCI as follows:

a. VERIFY at least one SGTS Train in operation.

b. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.

c. PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP handswitch, in START.

d. PLACE 3-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, in START.

e. OPEN the following valves:

3-FCV-73-36, HPCI/RCIC CST TEST VLV

3-FCV-73-35, HPCI PUMP CST TEST VLV
3-FCV-73-30, HPCI PUMP M1N FLOW VALVE.

f. OPEN 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, to start HPCI Turbine.

g. VERIFY HPCI Auxiliary Oil Pump starts and turbine accelerates above 2400 rpm.

39

NRC Scenario 3 Simulator Event Guide: Event 7 Major: Station Blackout ATC/BOP Commence pressure control with Appendix 1 1C, Alternate RPV Pressure Control Systems HPCI TEST MODE

6. VERIFY proper HPCI minimum flow valve operation as follows:

a. IF HPCI flow is above 1200 gpm, THEN VERIFY CLOSED 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

b IF HPCI flow is below 600 gpm, THEN VERIFY OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

7. THROTTLE 3-FCV-73..35, HPCI PUMP CST TEST VLV, to control HPCI pump discharge pressure at or below 1100 psig.

8. ADJUST 3-FIC-73-33, I-IPCI SYSTEM FLOW/CONTROL, controller to control RPV pressure.

9. IF HPCI injection to the RPV becomes necessary, THEN ALIGN HPCI to the RPV as follows:

a. OPEN 3-FCV-73-44, HPCI PUMP INJECTION VALVE.

b. THROTTLE 3-FCV-73-35, HPCI PUMP CST TEST VLV, to control injection 40

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA Crew Report rising Drywell Pressure, may enter EOI-2 earlier on Suppression Pool Temperature SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 1 60F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) 41

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 42

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When H2 is detected in PC (2.4% on control room indicators continue, does not continue SRO Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, directs RHR Pump 3D in Pool Cooling Enters EOI-2 on High Drywell Pressure SP/L MONITOR and CONTROL suppr p1 lvi between -l in. and -6 in. (APPX 18) Can suppr p1 ivl be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES SRO Emergency Plan classification 5.1-Al or 5.1-S 5.1 -Al Loss of voltage to ANY THREE unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND Only ONE source of power available to the remaining board. OPERATING CONDITION: Mode 1 or 2 or 3 5.1-s Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes. OPERATING CONDITION: Mode 1 or 2 or 3 43

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATCIBOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

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

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

2. PLACE RHR SYSTEM!! 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-23-52, RHR HX 3D RHRSW OUTLET VLV.

d. IF Directed by SRO, THEN PLACE 3-XS-74-130, RUR SYS II LPCI 2/3 CORE HEIGHT OVRJ) in MANUAL 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 1NBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

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

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

CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage. ATC/BOP ONLY aligns RHR Pump 3D m Pool Cooling, when power is available

                               .                               .

44

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATC/BOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

i. THROTTLE 3-FCV-74-73, RHR SYS 1(11) SUPPR 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-74-7(30), RI-JR SYSTEM 1(11) MIN FLOW VALVE

k. MONITOR RFIR Pump NPSH using Attachment 1.

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

m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional RHR and RIIRSW pumps_in_service_using_Steps_2.b through 2.1.

ATCIBOP ONLY aligns RHR Pump 3D in Pool Cooling, when power is available 45

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

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

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

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

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

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

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 INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS H 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 CHBRIPOOL ISOL VLV.

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

ATC/BOP ONLY aligns RHR Pump 3D in Chamber Sprays 46

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays

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, RHR SYSTEM II MIN 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. TIIROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:

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

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

47

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays BEFORE Diywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

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

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

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

6. INITIATE Drywell Sprays as follows:

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

b. IF EITHER of the following exists:

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

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 LPCI INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS II LPCI OUTBD iNJECT VALVE.

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

f. OPEN the following valves:

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

ONLY aligns R}[R Pump 3D in Drywell Sprays 48

NRC Scenario 3 Simulator Event Guide: Event 10 Major: LOCA ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Diywell Sprays

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

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

i. MONITOR RNR Pump NPSH using Attachment 2.

j. VERIFY RHRSW pump supplying desired R[IR Heat Exchanger(s).

k. 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 RHRSW OUTLET VLV.

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

7. WHEN EITHER of the following exists:

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

a. VERIFY CLOSED the following valves:

3-FCV-74-l00, RHR SYS I U-2 DISCH XTIE
3-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
3-FCV-74-75, RHR SYS II DW SPRAY INBD VLV.

b. VERIFY OPEN 3-FCV-74-30, RHR SYSTEM II MIN FLOW VALVE.

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

d. STOP RHR Pumps 3B and 3D.

49

NRC Scenario 3 SHIFT TURNOVER SHEIT Equipment Out of ServicelLCOs: CRD Pump 3A, EHC Pump 3A, and EECW Pump B3 Operations/Maintenance for the Shift: DG 3B is operating in parallel with Offsite power for the last hour, parallel with system operation is no longer desired. Remove DG 3B from parallel operation and shutdown DG 3B lAW 3-01-82 Commence a power increase to 100% Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas: None 50

Facility: Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.: 1306 Examiners:___________________ Operators: SRO:___________________ ATC:_____________ BOP:______________ Initial Conditions: 80% power, RFPT 3B and A3 RHRSW Pumps are tagged out. Turnover: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B. Raise power to 85% with flow and hold for RFPT 3B repairs. Event Maif. No. Event Type* Event Description No. N-BOP Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A 1 N-SRO and 30B R-ATC 2 Commence power increase with flow to 85% R-SRO C-ATC 3 edlOb C-BOP Loss of 480V SD BD 3B TS-SRO C-BOP 4 Batch File Stator Water Cooling Pump trip C-SRO fw30a RFPT 3A Governor fails low

         .                I-ATC 6       imf tel Oh                  EHC Pressure Transducer failure I-SRO 7        Batch File     M-ALL       ATWS 8         Batch File    M-ALL       LOOP/LOCA Loss of RPV Water Level 9           dgOla           C       3EA DG fails to auto start 10          dgOld           C       3ED DG fails to automatically tie to shutdown board
      *     (N)ormal,  (R)eactivity,  (I)nstrument,   (C)omponent,    (M)aj or

NRC Scenario 4 Critical Tasks Five

With reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS.

1. Safety Significance:

Precludes core damage due to an uncontrolled reactivity addition.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation unless all required injection systems are Terminated and Prevented.

4. Feedback:

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

1. Safety Significance:

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

2. Cues:

Procedural compliance. Suppression Pool temperature.

3. Measured by:

Observation If operating lAW EOI-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. 2

NRC Scenario 4 Critical Tasks Five

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

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation No ECCS injection prior to being less than the MARFP.

AND Observation Feedwater terminated and prevented until less than the MARFP.

4. Feedback:

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

1. Safety Significance:

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

2. Cues:

Procedural compliance. RPV pressure indication.

3. Measured by:

Observation Injection not commenced until less than MARFP, and injection controlled such

that power spikes are minimized, level restored and maintained greater than TAF.

4. Feedback:

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

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

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

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

AND Observation RO initiates Drywell Sprays

4. Feedback:

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

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation US directs Drywell Sprays JAW with EOI Appendix 17B

AND Observation RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment 4

NRC Scenario 4 Events

1. BOP operator will alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B.

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

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

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

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

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

6. An ATWS will occur on the scram and the power supply to HPCI will fail, leaving RCIC as the only source of high pressure makeup besides SLC and CR1). The crew will insert control rods manually, and maintain reactor level.

7. A LOOP will occur the 3D DG will fail to automatically tie to the shutdown board and will have to be manually tied and the 3A DG will fail to auto start but can be manually started. With RCIC, CR1) and SLC as the only source of high pressure makeup as the LOCA degrades RPV Level will continue to lower. When Reactor Water Level drops below -122 inches the crew will have to tie the 3D DG again due to Load Shed. The SRO will determine Emergency Depressurization is required to restore RPV Level. The crew will ED and restore RPV Level with available systems.

Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner: Control Rods are being inserted Emergency Depressurization complete Reactor Level is restored 5

NRC Scenario 4 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 9 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) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 5 Crew Critical Tasks: (2-5) YES Technical Specifications Exercised (Yes/No) 6

NRC Scenario 4 Scenario Tasks TASK NUMBER K/A Q $jQ Alternate Reactor and Refuel Zone Fans ROU-30A-NO-02 288000A4.01 3.1 2.9 Raise Power with Recirc Flow RO U-000-NO-06 202002A4.07 3.3 3.3 SRO S-000-AD-3 1 Loss of 480V SD BD 3B RO U-57B-AL-06 26200 1A2.04 3.8 4.2 SRO S-57B-AL-09 Reactor Feed Pump Turbine Governor Failure RO U-003-AL-09 259002A4.01 3.8 3.6 SRO S-003-AB-01 Stator Water Cooling Pump Trip RO U-35A-AL-02 245000A4.03 2.7 2.8 SRO 5-070-AB-Ol EHC Pressure Transducer Failure RO U-047-AB-02 241000A2.03 4.1 4.2 SRO S-047-AB-02 LOOP RO U-57A-AB-0l 295003AA1.03 4.4 4.4 RO U-082-AL-07 SRO S-57A-AB-01 7

NRC Scenario 4 Scenario Tasks TASK NUMBER K/A RO SRO LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 SRO S-000-EM-14 SRO S-00 0-EM- 15 SRO S-000-EM-01 ATWS RO U-000-EM-03 295015AA2.01 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18 8

NRC Scenario 4 Procedures Used/Referenced: Procedure Number Procedure Title 3-01-3 OA Refuel Zone Ventilation System 3-0I-30B Reactor Zone Ventilation System 3-GOl- 100-12 Power Maneuvering 3-01-68 Reactor Recirculation System 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-8C Panel 9-8 3-XA-55-8C 3-ARP-9-4C Panel 9-4 3-XA-55-4C 3-A0I-99-1 Loss of Power to One RPS Bus 3 -AOl- 1-3 Main Steam Isolation Valve Closure at Power 3-AOI-70-i Loss of Reactor Building Closed Cooling Water 3-01-99 Reactor Protection System 3-AOI-64-2D Group 6 Ventilation System Isolation Technical Specifications Technical Requirements Manual 3-ARP-9-7A Panel 9-7 3-XA-55-7A 3-ARP-9-8A Panel 9-8 3-XA-55-8A 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3-AOI-3 -1 Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-AR.P-9-7B Panel 9-7 3-XA-55-7B 3-AOI-47-2 Turbine EHC Control System Malfunctions 3-AOI-100-1 Reactor Scram 3-EOI-i RPV Control 3-C-5 LevellPower Control 3-EOI Appendix-iD Insert Control Rods Using Reactor Manual Control System 3-EOI Appendix-iF Manual Scram 3-EOI Appendix-2 Defeating ARJ Logic Trips 3-EOI Appendix-i iA Alternate Pressure Control Systems MSRVs 3-EOI Appendix-SC Injection System Lineup RCIC 3-EOI-2 Primary Containment Control 3-EOI Appendix-17A RHR System Operation Suppression Pool Cooling 3-E0I Appendix-i 7C RHR System Operation Suppression Chamber Sprays 3-EOI Appendix-i 7B RHR System Operation Drywell Sprays 0-AOI-57-1A Loss of Offsite Power (i6i and 500 KV)/Station Blackout 3-AOI-85-3 CRD System Failure 3-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 3-EOI Appendix-4 Prevention of Injection 3 -C-2 Emergency RPV Depressurization 3-EOI Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode 3-EOI Appendix-6C Injection Subsystems Lineup RHR System II LPCI Mode EPIP- i Emergency Classification 9

NRC Scenario 4 Console Operator Instructions A. Scenario File Summary Batch File

  #RFPT 3B and EECW Pump A3 tagout ior ypobkrrhrswpa3 fail_ccoil ior zlohs2385a[1] off ior ypomtreopb3 fail_controlpower ior ypobkrmopbi fail_ccoil ior ypobkrmopb2 fail_ocoil ior ypomtrtgmb fail_control_power ior ypovfcvo3 12 fail_power_now ior ypovfcvo295 fail_power_now ior ypovfcvol 129 fail_power_now ior ypovfcvol 133 fail_power_now
  #Loss of 480V SD BD B imfediOb (el 0) ior zdihs0l 14a[1J (el 0) close ior ypovfcv7o48 fail_power_now ior zlohs7048a[2] on trg e2 nrc7048 trg e2 = bat nrc7048-1 mrfrp09 (e3 0) reset trg e3 bat restorerpsb ior zdihs577l [1] (e4 0) normal
  #B stator water pump trip irf egO2 (e5 0) off ior ypobkrscwpa (e5 0) fail_ccoil ior zdihs3535a[2] (e5 0) stop ior zlohs3535a[1] (e5 0) off
   #rfpt governor drift imffw3Oa(elOO)0 130072 trg eli nrcrfptA trgell dmffw30a
   #B EHC Pressure transducer failure ior zdihs0il6[1] (e14 0) select ior zdihs472O4[ij (e14 0) null ior zlohs0i 16[1] off ior zlohs472O4[i] on imftci0b (e14 0) 82 1600 79 10

NRC Scenario 4

   #major trg e 18 nrcmodesw bat nrcstickquad ImfedOl (e18 8:00)                        LOOP Imfth22 (e18 1:00)                        LOCA Imfth2l (e18 10:00) 115:00                LOCA imfhp07 (e18 0)                           HPCI Fails imf dg0 1 a                               DG 3A will not auto start imf dg03d                                 DG 3B will not auto tie trg e19 = bat eecw trg e20 = bat eecw-1 trg e21 = bat rpsreset trg e22 = bat ca trg e23 = bat appOif trg e24 = bat app02 trg e25 = bat nrcstickquad-1 mrfrd06 (e26 0) close mrf rdO6 (e27 0) open trge28 =batnrcatws90 trg e30 = bat sdv Trigger Files nrc7048 zdihs7048a[1] .eq. 1 nrcriptA zdihs468a[41 .ne. 1 nrcmsivD zdihs0 1 52a[1] .eq. 1 Scenario 4 DESCRIPTION/ACTION Simulator Setup                           manual           ResettoiC 203 Simulator Setup                         Load Batch         bat nrc 1306-4 Simulator Setup                           manual           Tag RFPT 3B and EECW Pump A3 Simulator Setup                                            Verify file loaded. Log in to EHC System to ensure when operators try to access they_are_able_to.

RCP required (80% 85% with flow) and RCP for Urgent Load Reduction

Provide marked up copy of 3-GOI-100-12 11

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

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

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

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

[4] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in OFF. [5] CHECK that the two red lights A(B) extinguish and the two green lights A(B) illuminate above REFUEL ZONE FANS_AND DAMPERS_switch,_3-HS-64-3A. NOTE If any damper does not meet the requirements of step 6.1 [6] IMMEDIATELY notify the Unit supervisor to evaluate SCIV damper operability (refer to TRM appendix A). If any listed damper indicates not full closed, it should be considered inoperable for its SCIV function, and the required actions of Tech Spec LCO 3.6.4.2 entered for all units. [6] CHECK the red (open) damper position indication lights extinguish and the green (closed) lights illuminate above the following control switches:

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

[7] PLACE REFUEL ZONE FANS ANT) DAMPERS switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B) to start alternate fans. 12

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

REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
REFUEL ZONE SPLY 1NBD ISOL DMPR, 3-HS-64-6
REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9
REFUEL ZONE EXH INBD ISOL DMPR, 3-HS-64-l0 NOTE A five minute time delay should be observed following Refuel Zone Supply and Exhaust Fan SLOW Start. The time delay allows the discharge dampers to fully open after SLOW start.

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

SUPPLY FANS FILTER DIFF PRESS Indicator, 3-PDI-064-0022, indicates less than 0.6 inches H20 at the Reactor Building/Refuel Floor Supply fan intake room at El 565.
REFUELING ZONE STATIC PRESS INTLK, 1-PDS-064-006 lA/C, on refuel floor_Panel_25-220_indicates_between (negative)_-0.25_inches_to_-0.40_inches.

BOP Contacts AUO for the above information Driver When contacted wait 4 minutes and report 3-PDI-064-0022, mdicates 04 inches H20 and 13

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

1) The preferred method to start the standby Reactor Zone supply and exhaust fans is to use the common control switch (3-HS-64-1 1A) on Panel 3-925.

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

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

[3] IF Reactor Zone Ventilation is to remain Out of Service for an extended period (3 hours) and it is desired to leave the Steam Vault Exhaust Booster Fan in service, THEN (Otherwise N/A): Step is NA [4] IF required, THEN SHUT DOWN Steam Vault Exhaust Booster Fan. REFER TO Section 7.4. (Otherwise N/A). Step is NA [5] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-11A, in OFF. [6] VERIFY dampers close and fans stop as indicated by illuminated green lights above the following switches:

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

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

NRC Scenario 4 Simulator Event Guide: Event 1 Component: Alternate Refuel and Reactor Zone Fans lAW 3-O1-30A and 30B BOP [8] VERIFY dampers open and fans start as indicated by illuminated red lights above the following switches:

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

[9] IF fast speed Reactor Zone Supply and Exhaust Fan operation is required, five minutes should be allowed after slow start for the discharge dampers to FULLY OPEN, THEN [9.1] PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-1 1A, in FAST A (FAST B). [9.2] VERIFY that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A. [10] VERIFY the following conditions: [10.1] VERIFY REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-064-0002, on 3-LPNL-925-0213, located at R17-P El 639, indicates between -0.25 inches and -0.40 inches H2O. [10.2] IF REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-64-2, is not between -0.25 inches and -0.40 inches H2O, THEN REFER TO 3-AOl-3 OB- 1, Reactor Building Ventilation Failure. [11] IF required, THEN START Steam Vault Exhaust Booster Fan. REFER TO Section 5.4. NOT Required BOP Contacts AUO for the above information Dnver When contacted wait 4 mmutes and report 3-PDIC-064-0022, mdicates -0 35 mehes 1120 15

NRC Scenario 4 / Simulator Event Guide: Event 2 Reactivity: Power increase with Recirc Flow SRO Notifies ODS of power increase. Directs Power increase using Recirc Flow, per 3-GOT-i 00-12. [20] IF desired to raise power with only two(2) Reactor feedpumps in service, THEN RAISE Reactor power, as desired, maintaining each Reactor feedpump less than 5050 RPM. ATC Raise Power w/Recirc, JAW 3-01-68, Section 6.2 D. Individual pump speeds should be mismatched by -60 RPM during dual pump operation between 1200 and 1300 RPM to minimize hannonic vibration (this requirement may be waived for short time periods for testing or maintenance). [1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;

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

AND/OR

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

[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A &3B using the following push buttons as required: RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-HS-96-32 NRC At RR Pump Speeds of 1260rpm and 1200 zpm, power will be 85% and RFPT RPMs will be just below 5025 Driver When directed by NRC, Trigger 1 Loss of 480V SD BD 3B, If crew attempts to close alternate supply breaker or is going to close alternate supply breaker delete ED 1 OB in order to allow the crew to energized the Board Dnver Wait 2 minutes and report license class 1404 was m the field, a trainee accidently tripped the normal feeder breaker. No problems indicated on Board. 16

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

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

B. IF 480V Shutdown Bd 3B is lost, THEN MANUALLY TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 3-9-8. C. IF manual transfer is accomplished, THEN REFER TO O-OI-57B, 3-01-99, and appropriate Ols for recovery or realignment of equipment. D. IF manual transfer is NOT accomplished, THEN REFER TO Tech Spec Section 3.8.1. Dispatches personnel to Breaker, may attempt to energize 480V SD BD 3B Driver ernate supply Driver ecidently tripped directs you to 1*fdjitdt 17

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

1. REFER TO TS Section 3.8.7.

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

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

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

B. CHECK light indications for loss of 480V equipment. C. CHECK board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc. D. IF Normal or Alternate feeder breaker tripped, THEN MANUALLY DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device. E. REFER TO O-OI-57B to re-energize or transfer the board. 18

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B ATC Alarm 4C-12: RBCCW PUMP DISCH. FIIJR PRESS LOW A. VERIFY 3-FCV-70-48 CLOSING/CLOSED. B. VERIFY RBCCW pumps A and B in service. C. VERIFY RBCCW surge tank low level alarm is reset. E. REFER TO 3-AOI-70-1 for RBCCW System failure and 3-01-70 for starting spare pump. When 480V RMOV BD 3B is restored should VERIFY 3-FCV-70-48 CLOSING ATC Announces Power, Pressure and Level stable on Board loss Crew 3-AOI-99-1, Loss of Power to One RPS Bus 4.1 Immediate Actions [1] STOP all testing with potential RPS half-scrams or PCIS logic isolation signals. NOTES

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

4.2 Subsequent Actions [1] VERIFY automatic actions occur. [2] ATTEMPT to determine cause of loss of RPS Bus using indicating lights inside RPS Circuit Protector cabinets. [3] NOTIFY Chemistry RWCU is isolated and no longer in-service and a sampling LCO per TRIvI 3.4.1 is to be entered. [4] NOTIFY Electrical Maintenance to correct cause. [5] RESTORE power to RPS Bus A(B) using alternate power supply. REFER TO 3-01-99 section for Immediate Restoration of Power to RPS Bus A(B) Using Alternate Power Supply. [5.1] DISPATCH operator to Aux. Instrument Room to reset ATU GROSS FAILURES. [6] W1{EN system restoration is desired, THEN RESTORE systems to normal. REFER TO 3-01-99 section for Restoration to Normal Following RPS Bus Power Loss. When requested to restoreRPS B, if requested to place on alternate trigger 3, mrf rpO9 reset D nver and bat restorerpsb, if requested to restore to normal then bat ipsreset and mrfrp09 reset 19

NRC Scenario 4 Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B ATC/BOP Reports MSIV A Inboard Valve shut on loss of board. SRO Enters 3 -AOl- 1-3, Main Steam Isolation Valve Closure at Power ATC 4.2 Subsequent Actions [1] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s). [2] LOWER reactor power with recirc flow and insert control rods as directed by the Reactor Engineer/Unit Supervisor as necessary to ensure that rated steam line flow 3.54 x 106 lb/hr is not exceeded as indicated on Main Steam Line Flow Indicators. REFER TO 3-GOI-100-12 or 3-GOI-100-12A for the power reduction. [6] fl? Drywell control air pressure is normal, THEN INITIATE trouble-shooting of the MSIV. (Otherwise N/A) Step is NA [7] EVALUATE Technical Specification 3.6.1.3, Primary Contaimnent Isolation Valves. SRO Directs ATC to lower power to less than 3.54 x 106 lb/hr on Main Steam Line Flow Indicators ATC Lowers power as directed by SRO 20

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

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

Verifies RWCU Tripped, cannot verify sectionalizing valve at this time NO Power 4.2 Subsequent Actions { 1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AND core flow is above 60%,THEN: [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A). One RBCCW Pump is in service with sectionalizing valve open due to loss of power [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A): [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). When power is restored to 480V SD BD 3B RBCCW Pump will auto start, the sectionalizing valve should also auto close when board is energized. ATC reports RBCCW restored and RBCCW sectionalizing valve failed to close when ATC power restored. Undervoltage Will NOT trip supply bkr, but after 5.3 sec will trip all SD Bd 3B load

breakers except RBCCW Pump and Drywell Blower 21

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

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

Verifies RWCU Tripped, cannot verify sectionalizing valve at this time NO Power 4.2 Subsequent Actions [1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AND core flow is above 60%,THEN: [2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A). One RBCCW Pump is in service with sectionalizing valve open due to loss of power [3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A): [3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions. [3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s). When power is restored to 480V SD BD 3B RBCCW Pump will auto start, the sectionalizing valve should also auto close when board is energized. ATC reports RBCCW restored and RBCCW sectionalizing valve failed to close when ATC power restored. Undervoltage Will NOT trip supply bkr, but after 5.3 sec will trip all SD Bd 3B load

breakers except RBCCW Pump and Drywell Blower 22

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

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

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

MS1V GROUP Al, 3-IL-64-A1
MSIV GROUP Bi, 3-IL-64-B1

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

MSIV GROUP A2, 3-IL-64-A2
MS1V GROUP B2, 3-IL-64-B2

[6] RESTORE Reactor and Refuel Zone Ventilation to normal operation. REFER TO 3-AOI-64-2D,_Group_6_Ventilation_System Isolation. BOP 3-AOI-64-2D, Group 6 Ventilation System Isolation 23

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

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

[10] At Panel 3-9-3, RESTORE Drywell DP Compressor to automatic operation as follows: [10.1] DEPRESS DRYWELL DP CPRSR SUCT VLV RESET pushbutton, 3-HS 139A. [10.2] DEPRESS DRYWELL DP CPRSR DISCH VLV RESET pushbutton, 3-HS-64-140A. [10.3] VERIFY OPEN DW TO SGT 1NBD ISOL VALVE using 3-HS-64-31. [10.4] VERIFY OPEN SUPPR CHER SGT 1NBD ISOL VALVE using 3-HS-64-34. BOP [11] At Panel 3-9-4, RESTORE Drywell Floor and Equipment Drain Systems to normal operation as follows: [11.1] NOTIFY Radwaste Operator that Drywell Equipment and Floor Drain Sump isolation valves are being reopened. [11.2] PLACE DRYWELL EQPT DR 1NBD ISOL VLV, 3-HS-77-15A, in AUTO After OPEN. [11.3] PLACE DRYWELL EQPT DR OUTBD ISOL VLV, 3-HS-77-15B, in AUTO After OPEN. [11.4] PLACE DRYWELL FLOOR DR INBD ISOL VLV, 3-HS-77-2A, in AUTO After OPEN. [11.5] PLACE DRYWELL FLOOR DR OUTBD ISOL VLV, 3-HS-77-2B, in AUTO After OPEN. Driver when directed by NRC triggerS for Stator water pump tnp 24

NRC Scenario 4 / Simulator Event Guide: Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System BOP 8.3 Restoration to Normal Following RPS Bus Power Loss [12] IF DW Radiation Monitor CAM, 3-RM-90-256 was secured due to a preplanned transfer, THEN (otherwise N/A) Step is NA [13] IF DW Radiation Monitor CAM, 3-RM-90-256, isolated due to loss of RPS, THEN MOMENTARILY DEPRESS the following RESET pushbuttons on Panel 3-9-2.

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

[14] At Panel 3-9-54, PLACE H2/02 Analyzer in service per 3-01-76. [15] At Panel 3-9-55, VERIFY DRYWELL OR SUPPRESSION CHAMBER EXHAUST TO SGTS, 3-FIC-84-20, in AUTO with setpoint at 100 scfin. [19] At Panels 3-9-10 and 3-9-11, RESTORE Radiation Monitoring System as follows: [19.1] DEPRESS applicable RESET pushbuttons. [19.2] RESTORE Radiation Monitoring System to normal. REFER TO 3-01-90. [20] RESTORE Main Steam System to normal. REFER TO 3-01-1. [22] At Panel 3-9-13, DEPRESS TIP ISOLATION RESET pushbutton, 3-HS-94-7D-S2. 25

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

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

[13] WHEN initiating signal has been corrected AND necessary repairs are made, THEN [13.1] VERIFY PCIS RESET:

RESET PCIS DiV I RESET, 3-HS-64-16A-S32.
RESET PCIS Dlvii RESET, 3-HS-64-16A-S33.

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

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

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

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

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

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

Driver when directed by NRC triggerS for Stator water pump trip 26

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

a. Drywell floor drain sump monitoring system; and

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

APPLICABILITY: MODES 1,2, and 3. Condition B: Required primary containment atmospheric monitoring system inoperable. Required Action B. 1: Analyze grab samples of primary containment atmosphere. Completion Time: Once per 12 hours Required Action B.2: Restore required primary containment atmospheric monitoring system to OPERABLE status. Completion Time: 30 days Driver when directed by NRC trigger 5 for Stator water pump trip 27

NRC Scenario 4 Simulator Event Guide Event 3 Component: Loss of 480V SD BD 3B SRO Tech Spec Actions from loss of 480V SD BD 3B SRO Evaluate Technical Specification 3.6.1.3. 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. APPLICABILITY: MODES 1,2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Contaiinnent Isolation Instrumentation. Condition A: NOTE Only applicable to penetration flow paths with two PCIVs. One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits Required Action A. 1: 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 hours for main steam lines Required Action A.2: NOTE Isolation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated. Completion Time: Once per 31 days for isolation devices outside primary containment Dnver when directed by NRC trigger 5 for Stator water pump trip 28

NRC Scenario 4 Simulator Event Guide: Event 4 Component: Stator Water Cooling Pump trip BOP Responds to alarms 7A-22 and 8A-l BOP Announces trip of Stator Water Cooling Pump 3B 7A-22, GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action ofthis ARP 3-XA-55-9-8A Window 1 alarms THEN, 1 VERifY all available Stator Cooling Water Pumps running.

2. Attempt to RESET alarm 3 IF alarm fails to reset, AND reactor power is above turbine bypass valve

capability THEN SCRAM the Reactor B. VERIFY a stator cooling water pump is running and CHECK stator temperature recorder, 3-TR-57-59, Panel 3-9-8. C. CHECK alarm by dispatching personnel to check the Stator Coolant Control Cabinet. 8A- 1 , TURBiNE TRIP TIMER iNITIATED A. CHECK Stator Cooling Water Flow and Temperature and Generator Stator temperatures using ICS. B. VERIFY all available Stator Cooling Water Pumps running. NOTE The full capacity ofthe Turbine Bypass valves with all nine valves open is 25% reactor power. To determine the capacity ofthe bypass valves, subtract 3% for each out of service bypass valve from the 25%. (Example, one bypass valve out of service, [25% 3% = 22%],

therefore, the capacity ofthe bypass valves with one bypass valve out ofservice is 22%.) C. IF all ofthe following conditions exist:

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

BOP Starts Stator Water Cooling Pump 3B Driver When report pump is extremely hot to touch, at breaker bk Driver verify start value is between 71 ure final seventy is set to zero ger 10 29

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

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

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

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

imftclOb (e14 0) 82 1600 79 30

NRC Scenario 4 Simulator Event Guide: Event 6 Instrument: EHC Pressure Transducer Failure Responds to alarm 7B-6, EHC/TSI SYSTEM A. On EHC Workstation computer on Panel 3-9-7, Alarm Summary screen, ATCIBOP ATTEMPT to RESET alarm input. B. IF necessary, ThEN REQUEST assistance from Site Engineering. ATC Recognizes lowering Reactor Pressure and generator megawatts. SRO Directs entry into 3-AOI-47-2. 3-AOI-47-2 Turbine EHC Control System Malfunctions 4.1 Immediate Actions [1] IF Reactor Pressure lowers to or below 900 psig, THEN MANUALLY SCRAM the Reactor and CLOSE the MS1Vs. 4.2 Subsequent Actions [3] IF a Group 1 isolation has occurred, THEN PLACE EHC PUMP 3A and 3B, 3-HS-47-1A and 3-HS-47-2A, to PULL TO LOCK. BOP Places EHC Pumps 3A and 3B in Pull to Lock SRO Directs manual scram, closing of the MS1Vs, and entry into 3-AOI-lO0-l. ATC Manually scrams the reactor. 31

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS ATC 3-AOI-100-l, Reactor Scram 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5A/S3B, on Panel 3-9-5. [2] LF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-Sl, in START & HOT STBY AND PAUSE for approximately 5 seconds (N/A) [3] Refuel Mode One Rod Permissive Light check [3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (N/A) [4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in SHUTDOWN. [5] REPORT the following status to the US:

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

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

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

[2.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 3-9-4. [2.3] REPORT ATWS Actions Complete and power level. [3] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit. [3.1] DOWNRANGE IRMs as_necessary to follow power_as it lowers. 32

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS SRO Enters EOI- 1 on Reactor Level SRO EOI- 1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO - IF Emergency Depressurization is or has been required THEN exit RC/P and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO - Is any MSRV Cycling? YES, but MSIVs closed IF Steam cooling is required? NO

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

IF Boron injection is required? NO SRO Directs a Pressure Band with SRVs lAW APPX 1 1A SRO EOI- 1 (Reactor Level) Monitor and Control Reactor Water Level. Directs_Verification_of PCIS_isolations. ATC/BOP Verifies PCIS isolations. SRO IF It has NOT been determined that the reactor will remain subcritical without boron under all conditions THEN Exit RCIL and Enter C5, Level/Power control. SRO Exits RCIL and Enters 3-C-5, Level/Power Control 33

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS ATCLBOP Commence pressure control with Appendix 1 1A, Alternate RPV Pressure Control Systems MSRVs IF Drywell Control Air is NOT available, ThEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRY WELL 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. 1 3-PCV-1-179 MN STM LINE A RELIEF VALVE.

b. 2 3-PCV-1-180 MN STM LINED RELIEF VALVE.

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

d. 4 3-PCV-1-31 MN STM LINE C RELIEF VALVE.

e. 5 3-PCV-1-23 MN STM LINE B RELIEF VALVE.

f. 6 3-PCV-1-42 MN STM LiNE D RELIEF VALVE.

g. 7 3-PCV-1-30 MN STM LINE C RELIEF VALVE.

h. 83-PCV-1-19 MN STM LINE B RELIEF VALVE.

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

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

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

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

m. 13 3-PCV-1-34 MN STM LiNE C RELIEF VALVE.

34

NRC Scenario 4 Simulator Event Guide: Event 7 Major: ATWS SRO EOI- 1 (Power) Monitor and Control Reactor Power Verify Reactor Mode Switch in shutdown Yes Initiate ARI completed Will tripping Recirc Pumps cause trip of main turbine, RFP, HPCI or RCIC No Is reactor power above 5% or unknown No - SLC Leg When periodic APRM oscillations greater than 25% peak to peak persist continue OR Before Suppression Pool temperature rises to 110°F continue

Direct SLC injection (APPX 3A) Irthibit ADS Verify RWCU system isolation completed earlier Insert Control Rods Leg Reset ARI and defeat ARI logic trip (APPX 2) Insert Control Rods using any of the following methods: APPX-1A Deenergize scram solenoids No APPX- 1 B Vent the scram air header No APPX- 1 C Scram individual control rods No APPX-1D Drive Control Rods Yes APPX-1E Vent control rod over piston No-APPX-1F Reset scram/RE-SCRAM Yes APPX- 1 G Raise CRD cooling water dp No - 35

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

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

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

4. DRAIN SDV UNTIL the following annunciators clear:

WEST CR1) 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.

3-EOI Appendix-iD

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

2. 1F Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SOV.

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

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

a. SELECT control rod.

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

c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.

Driver When called for Appendix 2 wait 2 mmutes and trigger 24, Appendix-iF wait 3 minutes and trigger 23, when SCRAM is reset trigger 25 to unstick rods Before the crew scrams or the LOOP insert trigger 28 for bat nrcatws90 and trigger 30 for bat sdv If requested to close 85-586 trigger 26 to close and trigger 27 to open 36

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS SRO Enters C-5, Level/Power Control Inhibit ADS ATC/BOP Inhibits ADS SRO Is any main steam line open No

Is reactor power above 5% or unknown No

Maintain RPV water level between -180 inches and +51 inches with the following injection sources: CRD APPX SB, RCIC APPX 5C, SLC APPX 7B

                         -                 -               -

SRO Directs a Level Band maintained by RCIC ATC Initiate RCIC JAW Appendix-SC and maintains level in directed band, if possible 37

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

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

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

5. OPEN the following valves:

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

6. PLACE 3-HS.71-3 1A, RCIC VACUUM PUMP, handswitch in START.

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

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

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

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

d. 3-FCV-71-34, RCIC PUMP M1N FLOW VALVE, closes as flow rises above 120 gpm.

9. IF BOTH of the following exist? NO

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

38

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig Continues

Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Stops the first time through when the LOCA worsens will continue at that time Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 39

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19) When 112 is detected in PC (2.4% on control room indicators continue, does not continue SRO Enters EOI-2 on High Drywell Pressure SPIT MOMTOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, directs R}IR Pumps in Pool Cooling Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18) Can suppr p1 lvi be maintained above -6 in., YES Can_suppr p1 lvi be maintained below -1_in., YES 40

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/BOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock AS NECESSARY:

PLACE 3-HS-74-155A, LPCI SYS I OUTBD 1NJ VLV BYPASS SEL in BYPASS.
PLACE 3-HS-74-155B, LPCI SYS II 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 gpmRHRSW flow:

3-FCV-23-34, RHR HX 3A RHR.SW OUTLET VLV
3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHR.SW 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), R}IR 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), RUR SYS 1(11) LPCI OUTBD INJECT VALVE.

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

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

CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and eciuipment damage. ATC/BOP Aligns directed RHR Pumps in Pool Cooling 41

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/130P 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling

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

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

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

k. MONITOR R}IR Pump NPSH using Attachment 1.

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

m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional RHR and RNRSW pumps in service using Steps 2.b through 2.1.

ATC/BOP Aligns directed RHR Pumps in Pool Cooling 42

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATCIBOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

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

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

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

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

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

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

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

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

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

f. VERIFY OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CI{13R/POOL ISOL VLV.

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

ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 43

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/BOP 3-EOI APPENDJX-17C, RHR System Operation Suppression Chamber Sprays

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

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

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

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RHRSW pump supplying desired RI-IR 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 RFIRSW OUTLET VLV
3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV
3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.

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

ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 44

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

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

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

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

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

6. INITIATE Drywell Sprays as follows:

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

b. IF EITHER of the following exists:

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

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

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

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

f. OPEN the following valves:

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

ATC/BOP . Aligns directed R}{R Pumps in Drywell Sprays 45

NRC Scenario 4 Simulator Event Guide: Event 7 Component: ATWS = ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays

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

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

i. MONITOR RIJR Pump NPSH using Attachment 2.

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

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

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

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

7. WHEN EITHER of the following exists:

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

a. VERIFY CLOSED the following valves:

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

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

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

d. STOP RHR Pumps.

ATC/BOP Aligns directed RHR Pumps in Drywell Sprays 46

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A 4.1 linmediate Actions [1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators. [2] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators. [3] IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes). BOP Report DG 3A failed to start and DG 3D failed to tie to 4KV SD BD 3ED. DG 3B and 3C have started and energized their respective Shutdown Board BOP Verifies 2 EECW Pumps not using same strainer are in service 47

NRC Scenario 4 Simulator Event Guide: Event 9 and 10 Component: DG 3A Fails to auto start and DG 3D fails to tie to Shutdown BD BOP Report DG 3A failed to start and DG 3D failed to tie to 4KV SD BD 3ED. Start DG 3A and verifies that the output breaker closes and energizes SD BD 3EA BOP Closes DG 3D Output Breaker 1836 to energize SD BD 3ED. NOTE Later in scenario when load shed logic activates BOP Operator will have to reclose Output Breaker 1836 to energize SD BD 3ED 48

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA Crew Responds to a Loss of Offsite Power lAW O-AOI-57-1A Crew 4.2 Subsequent Actions [1] IF ANY EOI entry condition is met, THEN REFER to the appropriate EOI(s). [2] VERIFY automatic actions and PERFORM any that failed to occur. NOTES

1) If a Unit is in a Station Blackout condition, performance of this instruction will also require implementation of 1 (2)(3)-A0I-3 OB- 1, Reactor Building Ventilation Failure, on the Unit in Station Blackout.

2) EECW supply valves to the Control Air Compressors and RBCCW are air operated. If initial air pressure is low, air compressors may trip on high temperature, until cooling water flow is established.

3) At US discretion, the O-FCV-67-53 valve can be placed in the open position with hand switch. The valve will automatically come open once EECW pressure is above setpoint. REFER to 01-67 for valve operation.

4) The North header supply to Unit 1 RBCCW, the North header supply to Unit 2 RBCCW and the South header supply to Unit 3 RBCCW are normally isolated with a manual valve; therefore no flow will occur when either 1-FCV-67-50, 2-FCV-67-50 or 3-FCV-67-5 1 opens.

SRO EOIs are already entered 49

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A Crew 4.2 Subsequent Actions [3] REFER to 1(2)(3)-AOI-78-l, FPC System Failure for a complete Loss of AC POWER, as necessary. NOT NECESSARY [4] WHEN EECW header pressure is restored above the reset pressure setpoint (psig) for the valves listed below, ThEN Common Unit 1 Unit 2 Unit 3 O-FCV-67-53 106 FCV-67-50 - 90 91 92 FCV-67-51 - 107 109 113 RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel 1-LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. Refer to the EECW to the RCW Crossties for Control Air & RBCCW section of 0-01-67. [5] START Control Air Compressors A, D and G as required and MONITOR system pressure. Refer to 0-A0I-32-1. [5.1] IF an air compressor trips on high temperature, THEN (Otherwise N/A) NOTIFY Unit Supervisor for instructions. [6] REFER to 3-A0I-32-2, Loss of Control Air, as necessary [7] PLACE RPS MG Sets A and B in service. Refer to 3-01-99. Calls for Control Air and EECW reset, calls for reset of RPS. 50

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A [9] START the Diesel Driven Fire Pump. Refer to 0-01-26. [11] IF containment isolation is required, THEN VERIFY the following containment isolation valves closed UNLESS they are required to be open by EOIs (RG 1.155): FCV-1-56 MN STM LINE OUTBD DRAiN ISOL FCV-69-2 RWCU OUTBD SUCT ISOLATION FCV-71-3 RCIC OUTBD SUCT ISOLATION FCV-7 1-18 RCIC SUPPR POOL OUTBD SUCT VALVE Crew FCV-73-3 HPCI STEAM LINE OUTBD ISOL VALVE FCV-73-26 HPCI SUPPR POOL INBD SUCTION VLV FCV-73-30 NPCI MAIN PUMP MINIMUM FLOW VLV FCV-74-47 RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV NOTES

1) The UNIT SUPERVISOR should prioritize board energization to ensure coimnon HVAC equipment powered from 480V boards is returned to service within 1 hour as directed by Attachment 5.

Request start of Diesel Fire Pump. Verifies Containment isolation status, RCIC valves will BOP/ATC be open When requested to start Diesel Fire Pump wait one mmute and start diesel fire pump irf iver f04d start [12] VERIFY the following boards are energized. IF NOT, THEN REFER to Attachment 1 to restore affected busses while continuing with this instruction. Unit 1 Unit 2 Unit 3 4KV Shutdown Boards A, C B, D 3EA, 3EB, 3EC, 3ED 480V Shutdown Boards 1A, lB 2A, 2B 3A, 3B 48OVDSLAuxBoards A B 3EA,3EB 48OVRMOVBoards 1A,1B 2A,2B 3A,3B 480V Control Bay Vent A B Boards 480V HVAC Board B Crew Once DG 3A and 3D are tied to their SD BDs, all boards should be energized 51

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA ATC On the LOOP the 3B CRD Pump 3B will have to placed in service lAW 3-AOI-85-3 4.1 Immediate Actions [1] IF operating CR1) PUMP has failed AND the standby CR1) Pump is available, THEN PERFORM the following at Panel 3-9-5: [1.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-11, in MAN at minimum setting. [1.2] START associated standby CRD Pump using one of the following:

CR1) PUMP 3B, using 3-HS-85-2A
CR1) Pump 3A, using 3-HS-85-1A

[1.3] ADJUST CR1) SYSTEM FLOW CONTROL, 3-FIC-85-1 1, to establish the following conditions:

CR1) CLG WTR H1)R DP, 3-PDI-85-18A, approximately 20 psid.
CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, between 40 and 65 gpm.

[1.4] BALANCE CR1) SYSTEM FLOW CONTROL, 3-FIC-85-11, and PLACE in AUTO or BALANCE. 52

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

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

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

11. CHECK SLC injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch.
Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,
SLC SQUIB VALVE CONTiNUITY LOST Annunciator in alarm (3-XA-5 5-SB, Window 20).
3-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.
System flow, as indicated by 3-IL 11, SLC FLOW, red light illuminated,
SLC INJECTION FLOW TO REACTOR Annunciator in alarm (3-XA-55-5B, Window 14).

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

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

                     -180 inches No -

Are at least 2 MSRVs open No - Emergency Depressurization is Required 3 -C-2 and 3 -C-S Will the reactor remain subcritical without boron under all conditions NO

When all injection into the RPV is stopped and prevented except from RCIC, CR1), and SLC_per_C5,_Level/Power_control_Step_C5-22 Stop and Prevent ALL injection into RPV Except from RCIC, CRD, and SLC (APPX 4) 53

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

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

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

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

OR

b. BEFORE RPV pressure drops below 450 psig,

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

AND

2) VERIFY CLOSED 3-FCV-74-52, RI-IR SYS I LPCI OUTBD iNJECT VALVE.

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

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

OR

b. BEFORE RPV pressure drops below 450 psig,

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

AND

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

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

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

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

3-FCV-3-19, RFP 3A DISCHARGE VALVE
3-FCV-3-12, REP 3B DISCHARGE VALVE
3-FCV-3-5, RFP 3C DISCHARGE VALVE
3-LCV-3-53, RFW START-UP LEVEL CONTROL 54

NRC Scenario 4 Simulator Event Guide: Event 8 Major: LOOP/LOCA SRO C2 Emergency Depressurization and C5 Level/Power Control Is suppression pooi level above 5.5 feet Yes Open all ADS Valves BOP/ATC Opens all 6 ADS Valves SRO Can at least two MSRVs be opened per C2 Emergency RPV Depressurization Yes - When RPV pressure is below MSCP Table 1A 190 psig Start and Slowly raise RPV injection with the following injection sources to restore and maintain RPV_water level above_-180_inches Directs injection with LPCI APPX 6B and 6C to restore RPV Level to directed band BOP/ATC Injects with LPCI lAW APPX 6B and/or 6C to restore RPV water level SRO Emergency Classification EPIP-1 1.1-Si Reactor water level can NOT be maintained above -162 inches. (TAP) OR 1.2-S Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical. 55

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

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN .PLACE 3-HS-74-155A, LPCI SYS I OUTBD 1NJ 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, RUR PUMP 3C SUPPR POOL SUCT VLV.

4. VERIFY CLOSED the following valves:

3-FCV-74-61, RHR SYS I DW SPRAY INBD VLV
3-PCV-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 1NBD INJECT VALVE.

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

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

9. MONITOR R}IR Pump NPSH using Attachment 1.

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

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

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV

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

56

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

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 3-HS-74-155B, LPCI SYS II OUTBD 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, RUR 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 CFIBR SPRAY VALVE
3-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV

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

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

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

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

9. MONITOR RHR Pump NPSH using Attachment 1.

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

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

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

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

57

NRC Scenario 4 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs: RFPT 3B and EECW Pump A3 Operations/Maintenance for the Shift: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Commence a power increase to 85% in accordance with the RCP Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas: None 58

Facility: Browns Ferry NPP Scenario No.: NRC 5 Op-Test No.: 1306 Examiners:__________________ Operators: SRO:___________________ ATC:______________ BOP:______________ Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation. Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing. Event Maif. No. Event Type* Event Description No. 1 Return LPRM 8-49B to Operate lAW 2-OI-92B R-ATC 2 Commence power decrease with flow to 90% RSRO 3 edl8a Loss of I&C Bus A R-ATC 4 ad0lc TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the 5 thl8 a C-SRO standby pump to auto start LOCA Recirculation Pump 2A Inboard and Outboard seal

6 thlO/1 la failure

                  .                Two Level instruments fail high tripping Feedwater and 7       Batch File     M-.ALL HPCI I LOCA / ED on Reactor Level 8          edl0a          C       Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not 9          Batch          I Auto open 10          rcO8          C       RCIC Steam Valve fails to Auto open
     *    (N)ormal,  (R)eactivity, (I)nstrument,     (C)omponent,   (M)ajor

NRC Scenario 5 Critical Tasks Three

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

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

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

3. Measured by:

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

4. Feedback:

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

1. Safety Significance:

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

2. Cues:

Procedural compliance. Water level trend.

3. Measured by:

Observation At least 6 SRVs opened

4. Feedback:

RPV pressure trend. SRV status indications. 2

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

1. Safety Significance:

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

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation.

4. Feedback:

RPV pressure trend. RPV level trend. ADS ADS LOGIC BUS A/B INHIBITED annunciator status. 3

NRC Scenario 5 Events

a. BOP operator will return LPRM 8-49B to Operate JAW 2-OI-92B.

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

c. The crew will respond to a(noment7joss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come 1nThe BOP operator will shift SJAEs to B or reset SJAE A and return to service JAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high H2 will result in Off Gas, the SRO will evaluate TRM 3.7.2 and enter Condition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.

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

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

f. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full

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

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

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

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

4

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

NRC Scenario 5 Scenario Tasks TASK NUMBER Q Q Restore an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 Lower Power with Recirc Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 26200 1A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-001-AB-Ol 239002A2.03 4.1 4.2 SRO S-001-AB-O1 VFD Cooling Water Pump Failure RO U-068-AL-19 202001A2.22 3.1 3.2 SRO S-068-AB-01 RR Pump Seal Failure RO U-068-AL-09 203000A4.02 4.1 4.1 SRO S-068-AB-01 Loss of 480V SD BD 2A RO U-57B-AL-06 226001A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-01 6

NRC Scenario 5 Procedures Used/Referenced: Procedure Number j Procedure Title 2-OI-92B Average Power Range Monitoring 2-GOT-i 00-12 Power Maneuvering 2-01-68 Reactor Recirculation System 2-AOI-57-5A Loss of I&C Bus A 2-ARP-9-8C Panel 9-8 2-XA-55-8C 2-ARP-9-7A Panel 9-7 2-XA-55-7A 2-ARP-9-6C Panel 9-6 2-XA-55-6C 2-ARP-9-7C Panel 9-7 2-XA-55-7C 2-ARP-9-3C Panel 9-3 2-XA-55-3C 2-ARP-9-3D Panel 9-3 2-XA-55-3D 2-ARP-9-53 Panel 9-53 2-XA-55-53 2-E0I-3 Secondary Containment Control 2-A0I-64-2D Group 6 Ventilation System Isolation

             .           Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 2-EOI Appendix-8F Isolation 2-01-66                   0ff-Gas System 2-A0I-66-1                Off-Gas H2 High Technical Specifications Technical Requirements Manual 2-Aol-i -1                Relief Valve Stuck open 2-01-74                   Residual Heat Removal System 2-E0I-2                   Primary Containment Control 2-E0I Appendix-18         Suppression Pool Water Inventory Removal and Makeup 2-ARP-9-4A                Panel 9-4 2-XA-55-4A 2-AOI-68-1A               Recirc Pump Trip/Core Flow Decrease OPRMs Operable 2-AOI-100-1               Reactor Scram 2-EOI-i                   RPV Control 2-EOI Appendix-8B         Reopening MSIVs / Bypass Valve Operation 2-EOI- i-C-i              Alternate Level Control 2-EOI Appendix-6A         Injection Subsystems Lineup Condensate 2-EOI Appendix-i 7C       RHR System Operation Suppression Chamber Sprays 2-EOI Appendix-i 7B       RHR System Operation Drywell Sprays 2-EOI-3-C-2               Emergency RPV Depressurization 2-EOI Appendix-5B         Injection System Lineup CRD 2-EOI Appendix-7B         Alternate RPV Injection System Lineup SLC System 2-EOI Appendix-6B        Injection Subsystems Lineup RHR System I LPCI Mode 2-EOI Appendix-6C         Injection Subsystems Lineup RHR System II LPCI Mode 2-EOI Appendix-6E        Injection Subsystems Lineup Core Spray System II EPIP- 1                   Emergency Classification 7

NRC Scenario 5 Console Operator Instructions A. Scenario File Summary Batch File NRC/l3O6nrc-5 ior zloOil2l ld2Ob[1j off ior zloOil2l ld2Ob[21 off ior zloOhs2l lOd2Oa[1j off Tag DG D ior zloOhs2l lOd2Oa[2] off ior zlo0hs2 11 Od2Oa[2] off mrfdg0ld open ior zdihs708a null ior zlohs7o8a[l] off Tag RBCCW 2B ior zlohs708a[2] off ior zlohs7o8a[3j off ior zlohs682a2a[1j on ior zlohs682a2a[2] off A VFD Cooling Pump Trip mrfthl8b trip trg 1 NRC/avfd trg 1= bat NRC/130605-l A VFD Cooling Pump Trip imfth30f (e5 0)100 imfth30h (e5 60) 100 45 55 Level 8 instrument failures imfrc08 RCIC steam supply valve failure imfthl0a(e3 0)100 imfthlla(e3 60)100900 RR2A Pump seal mrf cs09b inhibit mrfrhl5 inhibit Div 2 accident logic bypassed ior zloil7556a off ior z1oi174154a off mrf edi 3 open momentary loss of I&C Bus A Batch File NRC/l3O6nrc-5-1 imfth2l (none 330) .6 600 .1 LOCA imfedl0a (none 370) Loss of 480V SD BD 2A 8

NRC Scenario 5 Preference File NRC/l3O6nrc-5 pflcOi tog pflc 02 ann silence pflc 03 mrf swO2 align align spare RBCCW Pump pfk 04 bat NRC/l3O6nrc-5 pfk 05 imfedl8a Loss of I&C Bus A pfk 06 ior zdihs682ala[1] off VFD A Cooling Pump trip pflc 07 imfadOic 10 ADS SRV Leak by pfk08trg! e3 RR Pump A Seal Failure pflco9frg! e5 Loss of Feedwater pflc 10 bat NRC/i3O6nrc-5-l LOCA and Loss of 480V SD BD 2A pfk 11 mrfad0lc out pflc 12 ior xa553e10 alarm_on pfk si pfks2mmfad0lc 100 pflc s3 mmfad0lc 10 pfk s4 mmfad0ic 100 pfk s5 mmfad0lc 10 pfk s6 bat appl8rhra pflc s7 bat appl8rhrb pflc s8 mrfedl3 close Scenario 5 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 28 manual Bypass LPRM 8-49B restorepref NRC/I 3O6nrc-5 mrf swO2 align RBCCW wait one F3 minute and turn off RBCCW Pump 2B Simulator Setup Load Batch F4 bat NRC/i 3O6nrc-5 Simulator Setup manual Tag DG D and RBCCW Pump 2B Simulator Setup Verify file loaded, Clear alarms for Reactor Recirc RCP required (100% -90% with flow) and RCP for Urgent Load Reduction 9

NRC Scenario 5 / Simulator Event Guide: Event I Normal: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B SRO Directs BOP to return LPRM 8-49B to Operate JAW 2-OI-92B BOP Return LPRM 8-49B to Operate JAW 2-OI-92B 6.4 Returning an LPRM to Operate From a Bypassed Condition [I] 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 sofikey, 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 BYPIHV 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. 10

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

Main Turbine Valve Testing
Ultimate heat Sink temperature > 92.5°F

[6.1] REDUCE Recirculation flow. REFER TO 2-01-68. [6.2] MAINTAIN Reactor thermal power within the limits shown on ICS and 0-TI-248, Station Reactor Engineer, as appropriate. [10] PERFORM the following while reducing Reactor power: [10.1] WHEN Reactor power is at approximately 90%, THEN REFER TO 2-01-3 and_START a RFP_Injection Water Pump. ATC Lower Power w/Recirc, JAW 2-01-68, Section 6.2 Driver When directed by NRC, insert preference key F5 imf edi 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close NRC Two additional power decreases in Scenario, can continue when ready. will have to mismatch speeds at 1300 rpm. 11

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

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

(Otherwise N/A)

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

2-HS-96-17A(1 7B)(1 7C). (Otherwise N/A). AND/OR

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

(Otherwise N/A)

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

2-HS 1 8A( 1 8B)( 1 8C). (Otherwise N/A). [2] WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECJRC MASTER CONTROL, THEN ADJUST Recirc Pump Speed 2A & 2B using the following pushbuttons as required. RAISE SLOW, 2-HS-96-3 1 RAISE MEDIUM, 2-HS-96-32 LOWER SLOW, 2-HS-96-33 LOWER MEDIUM, 2-HS-96-34 LOWER FAST, 2-HS-96-35 Dnver When directed by NRC, insert preference key F5 imf edl 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close When dispatched wait two minutes and report Failure of 9-9 Throwover Switch, switch tripped to alternate. N Two additional power decrease in Scenario, can continue when ready 12

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

Loss of instrument power and remote position indication to Core Spray Div I and RHR Div I (Panel 9-3).

RWCU Filter Demin A isolation.

Reactor BuildingfRefhel Zone Ventilation isolation.

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

CHECK display screens for Governor abnormal conditions.

Announced_RPV_Level_stable,_dispatches_personnel BOP Alarm 3C-25, MAiN STEAM RELIEF VALVE OPEN A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 2-TR-1-1, on Panel 2-9-47 and SRV Tailpipe Flow Monitor on Panel 2-9-3 for raised temperature and flow indications. B. REFER TO 2-AOl-i-i. BOP Announce Main Steam Relief Valve Open alarm cleared, but have indication on acoustic monitor of SRV partially open or leaking by. ADS SRV 1-22 3C-25 alarms on a loss of I&C Bus A, when the bus re-energizes ADS SRV will show NOTE acoustic momtonng indication of leaking by BOP operator should report to SRO and SRO enter 2-AOl-i-i These events will occur under event four 13

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32. BOP Alarm 7C-22, DRYWELL/SUPPR CHAMBER H202 ANALYZER FAILURE A. ChECK Panel 2-9-54 and 2-9-55 for abnormal indicating lights such as low flow, F12 or 02 downscale, pump off, etc. B. IF sample pump is NOT running, THEN ATTEMPT to start pump using 2-HS-76-1 10/S5. C. IF sample pump will NOT start OR H2/02 analyzer malfunction, THEN PLACE H2/02 Analyzer in Service per 2-01-76 section 5.4. D. REFER TO TRM 3.3.11 and TRM Section 3.6.2. BOP Resets H2/02 ANALYZER ISOLATION RESET, 2-HS-76-91 Resets Alarm on 2-MON 110, touch screen. BOP Alarm 3D-19, DRYWELL LEAK DETECTION RADIATION DNSC A. DETERMINE cause of alarm by performing the following:

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

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

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

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32. BOP Alarms 3D-3, RX BLDG VENTILATION ABNORMAL A. IF PCIS group 6 isolation exists, ThEN REFER TO 2-AOI-64-2d. B. NOTIFY Unit Supervisors, Unit 1 and Unit 3. C. VERIFY standby fans start. D. DISPATCH personnel to check Bldg AP (PDIC 64-2, El 639, Rx Bldg.) E. IF AP is at or above -0.17 in. H2O ThEN ENTER 2-EOI-3 Flowchart, 2-XA-55-3D, window 32. BOP Alarms 3D-32, REACTOR ZONE DIFFERENTIAL PRESSURE LOW D. IF alarm is valid, THEN INFORM Unit Supervisor of 2-EOI-3 entry condition. E. REQUEST personnel to check fans locally for any apparent problems. F. REFER TO 2-OI-30B and PLACE standby fan in service to restore normal differential pressure. SRO Enters 2-EOI-3, Secondary Containment Control and 2-AOI-64-2D, Group 6 Ventilation System Isolation Directs Reactor and Refuel Zone Ventilation returned to service by either 2-EOI Appendix-8F, Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation or 2-AOI-64-2D The above procedures for restoring ventilation are basically the same will describe NOTE Appendix-8F below. The only action in EOI-3 is to restore ventilation. 15

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

6 Isolation VERIFY PCIS Reset.

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

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

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

c. CHECK two SPLYIEXH A(B) green lights above 2-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:

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

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

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

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

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

d. VERIFY OPEN the following dampers:

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

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

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

16

NRC Scenario 5 Simulator Event Guide: Event 3 Component: Loss of I&C Bus A SRO Enters 2-AOI-57-5A 4.2 Subsequent Actions [1] VERIFY Automatic Actions have occurred. [2] IF a Reactor Scram occurs, THEN PERFORM 2-AOl- 100-1 concurrently with this procedure. [31 VERIFY a flow path for Condensate System, or STOP the condensate pumps/booster pumps. REFER TO 2-01-2. [4] START Standby Gas Train(s) and CHECK Reactor Building pressure at or below 0.25 H2O vacuum (PDIC 64-1, Panel 25-215; PDIC 64-2, Panel 25-213). REFER TO 0-01-65, Section Standby Gas Treatment System Manual Initiation. [5] VERIFY SJAE B in service to maintain condenser vacuum. REFER TO 2-01-66. [6] IF Auto Transfer of Panel 2-9-9, Cabinet 2, failed THEN (otherwise N/A) [7] WHEN Reactor water level is normal, THEN RESET PCIS Group 6 inboard isolation and RETURN the affected systems to service or standby readiness. REFER TO 2-AOl- 100-1, if a Reactor Scram occurred, otherwise REFER TO 2-A0I-64-2D. SRO Directs restoration of Reactor Building DP, should restore Ventilation JAW Appendix-8F or 2-AOI-64-2D._May_call Unit_1_to_start_Standby_Gas_Fans SRO Directs restoration of SJAE, JAW 2-01-66 hard card BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card 17

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

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

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

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

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

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

BOP Acknowledge Panel 2-9-53 Alarms, Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % 112 TRAIN A, and HIGH OFFGAS % H2 TRAIN BB 18

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

RECOMB1NER 2A/2B TEMPERATURE, 2-TRS-66-77, for abnormal trend; either rising or lowering.
OFF GAS HYDROGEN ANALYZER, 2-H2R-66-96, for hydrogen concentration.

NOTE H2 concentration will nse to 8 to 12% and return to a normal value of less than 1% 19

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

a. Drywell floor drain sump monitoring system; and

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

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

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

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

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

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

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

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

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

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking Event 4 started with the failure of I&C Bus A, dependmg on SRO priontles may have NOTE addressed SRV first and then restoration from Bus loss SRO Enters 2-AOl-i-i BOP 4.1 Immediate Action [1] IDENTIFY stuck open relief valve by OBSERVING the following: SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3, OR

MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-l-i on Panel 2-9-47.

ATC [2] IF relief valve transient occurred while operating above 90% power, THEN REDUCE reactor power to 90% RTP with recirc flow. BOP [3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch several times as required:

CLOSE to OPEN to CLOSE positions

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

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking Actions for SRV 1-22, wait two minutes and for taking control at panel 25-32 Preference key F 12, for cycling SRV preference key shift F2 open, shift F3 close to 10, shift F4 open, shift F5 close to 10. Driver Contact control room and determine if valve closed. When told to remove power preference key Eli. When back to normal at panel 25-32 delete override for annunciator xa553e1 0. When told to power back up srv 1-22 mrf ad0lc in. Driver [2] IF 2-PCV-i-22 is NOT closed, THEN PERFORM the following: [2.11 On Panel 2-25-3 2 PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-1-22 in EMERG position. [2.3] IF the SRV does NOT close, THEN PERFORM the following: A. VERIFY the MAIN STM LINE B RELIEF VALVE, 2-HS-1-22C, in the CLOSE/AUTO position. B PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR,, 2-XS-l-22 mNORM position [2.4] IF the SRV does NOT close, THEN REMOVE the power from 2-PCV-l-22 by performing one of the following: A. OPEN the following breakers (Preferred method) F2.5] 11? the valve does NOT close, THEN CLOSE the breakers or REINSTALL fuses removed in Step 4.2.3 [2.4]. [2.6] CONTINUE at Step 4.2.4. 23

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP [2.6] CONTINUE at Step 4.2.4. 4.2A Other Actions and Documentation [1] NOTIFY Reactor Engineering of current conditions. [2] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s). [3] REFER TO Technical Specifications Sections 3.5.1 and 3.4.3 for Automatic Depressurization System and relief valve operability requirements. [4] INITIATE suppression pooi cooling as necessary to maintain suppression pooi temperature less than 95°F.

                 / [5]      IF the relief valve can NOT be closed AND suppression pooi temperature
                 \          CANNOT be maintained less than or equal to 95°F, ThEN PLACE the reactor in Mode 4 in accordance with 2-GOI-iOO- 12A.

[6] DOCUMENT actions taken and INITIATE Work Order (WO) for the valve. SRO Directs Suppression Pool Cooling lAW 2-01-74 BOP Initiates Pool Cooling as directed SRO Refers to Tech Specs 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 (I{PCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

Condition E: One ADS valve inoperable. Required Action E. 1: Restore ADS valve to OPERABLE status. Completion Time: 14 days. BOP/ATC Inform SRO when Suppression Pool Level meets EOI-2 entry requirements SRO Enter EOI-2 on Suppression Pool Level NOTE One RHR Pump will almost mamtam pool temperature dependmg on reactor power, Do

                    .rrm            .1                      .1 flcOV 24

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

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

NOTES

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

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

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Coolmg as directed [7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows: [7.1] START an RHRSW Pump to supply RHR Heat Exchanger A(C). [7.2] ESTABLISH RHRSW flow by performing one the following: [7.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, (RHRSW Pump A(C) and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23. OR [7.2.2] THROTTLE OPEN RHR fiX 2A(2C) RHRSW OUTLET VLV, 2-FCV-23 -3 4(40), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm R[{RSW flow as indicated on 2-FI-23 -36(42), RHR HTX 2A(2C) RHRSW FLOW. D [7.3] VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 2-FCV-74-53. [7.4] IF NO RHR PUMP (1A OR 1C) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RHR SYS I SUPPR POOL CLG/TEST VALVE, 2-FCV-74-59. [7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-5 8. [7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60. [7.7] VERIFY OPEN RHR SYS I SUPPR CHI3RIPOOL ISOL VLV, 2-FCV-74-57. [7.8] IF desired to operate without the Drywell DP Compressor, THEN: [7.9] START RHR PUMP A(C) using 2-HS-74-5A(16A). [7.10] THROTTLE RHR SYS I SUPPR POOL CLG/TEST VLV, 2-FCV-74-59, to maintain RHR flow within limits, as indicated on RHR SYS I CTMT SPRAY FLOW, 2-FI-74-56: RHR Pumps in 1 2 Operation Loop Flow 7,000 to <13,000 gpm & 10,000 gpm & Blue Blue light light illuminated illuminated 26

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

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

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

NRC Scenario 5 Simulator Event Guide: Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed [10.8] IF desired to operate without the Drywell DP Compressor, THEN: [10.9] START RHR PUMP 2B(2D) using 2-HS-74-28A(39A). [10.10] THROTTLE RI{R SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73, to maintain RHR flow within limits, as indicated on RHR SYS II CTMT FLOW, 2-FI-74-70. RHR Pumps in 1 2 Operation Loop Flow 7,000 to <130OO gpm & 10,000 gpm & Blue Blue Fight light illuminated illuminated [11] IF desired to RAISE Suppression Pool Cooling flow and only one Loop II pump is in service, THEN PLACE the second Loop II RHR Pump AND Heat Exchanger in service. REPERFORM Step 8.5[10] for the second pump. [12] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:

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

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

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

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

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

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

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

DISCH TO MN CNDR/RW SOy.

2) OPEN the following valves:

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

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

DISCH SOy.

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

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

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

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

1) ESTABLISH communications with Radwaste.

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

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

e. THROTTLE 2-FCV-74-108, RUR DR PUMP 2A/B DISCH FIUR VALVE, as necessary.

Driver When dispatched to remove water from the suppression pool, wait 10 mmutes and call and report aligned step 4 above, when the operator calls you back to start the RHR Drain Pump Shift F6 for bat appl8rhra and Shift F7 for bat appl8rhrb 30

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

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

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

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

2) CLOSE the following valves:

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

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

DISCH SOy.

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

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

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

Dnver When directed by NRC for VFD Coolmg Pump trip, Preference Key F6 31

NRC Scenario 5 Simulator Event Guide: Event 5 Component: VFD Cooling Water Pump 2A trip Driver When directed by NRC for VFD Coolmg Pump trip, Preference Key F6 ATC Respond to the following alarms, 4A-12, 4A-28 and 4A-32 ATC Report Trip of Recirc Drive 2A Cooling Pump 2A1, and failure of standby pump to start Alarm 4A-12, RECIRC DRIVE 2A COOLANT FLOW LOW Automatic Action Standby RECIRC DRIVE cooling water pump will auto start. A. VERIFY RECIRC DRIVE cooling water pump running. B. DISPATCH personnel to the RECIRC DRIVE to check the operation of the Recirc Drive cooling water system. Alarm 4A-28, RECIRC DRIVE 2A PROCESS ALARM A. IF 2.-XA-55-4B Window 28 is also in alarm, ThEN (N/A) B. Refer to ICS screen VFDAAL and determine cause of alarm Alarm 4A-32, RECIRC DRIVE 2A DRIVE ALARM A. REFER TO ICS Group Display GD @VFDADA and DETERMINE cause of alarm. B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system. ATC Start Standby Recirc Drive 2A Cooling Pump 2A2, dispatches personnel to investigate Wait 4 minutes after dispatched, THEN report tripped VFD Pump 2A1 is hot to the touch, internal bkr closed, 480 volt bkr tripped (480 V SD BD 2A 5C). 9]y When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 32

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure

Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 ATC Respond to alarm 4A-25, RECIRC PUMP A NO. 1 SEAL LEAKAGE ABN A. DETER1UNE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 2-9-4 or ICS.

Plugging of No. 1 RO No. 2 seal cavity pressure indicator drops toward zero.

Plugging of No. 2 RO No. 2 seal pressure approaches no. 1 seal pressure.

Failure of No. 1 seal No. 2 seal pressure is greater than 50% of the pressure of

No. 1.

Failure of No. 2 seal no. 2 seal pressure is less than 50% of the No. 1 seal.

B. RECORD pump seal parameters hourly on Attachment 1, ATC Report of failure of number 1 seal or inner seal Respond to alarm 4A-1 8, RECIRC PUMP A NO.2 SEAL LEAKAGE HIGH A. COMPARE No. 2 cavity pressure indicator (2-PI-68-63A) to No. 1 cavity pressure indicator (2-PI-68-64A). No. 2 seal degradation is indicated if the pressure at No. 2 seal is less than 50% of the pressure at No. 1 seal. ATC g Reports the second seal is failed both pressure indicators trending toward zero psig. C. IF dual seal failure is indicated, THEN

1. SHUTDOWN Recire Pump 2A by depressing RECIRC DRIVE 2A SHUTDOWN, 2-HS 19.

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

3. VERIFY TRIPPED, REC1RC DRIVE 2A ALTERNATE FEEDER, 2-HS-57-15.

4. CLOSE Recirculation Pump 2A suction valve.

5. CLOSE Recirculation Pump 2A discharge valve.

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

7. DISPATCH_personnel_to_secure_Recirculation_Pump_2A_seal_water.

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

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure

SRO Enters 2-AOI-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 4.2 Subsequent Actions [1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (N/A), [2] IF a single Recirc Pump tripped, ThEN CLOSE tripped Recirc Pump discharge valve. [3] IF Region I or II of the Power to Flow Map is entered, THEN IMISIEDIATELY take actions to iNSERT control rods to less than 95.2% loadline. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation. [4] RAISE core flow to greater than 45%. REFER TO 2-01-68. [5] INSERT control rods to exit regions if not afready exited. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation. [6] MAINTAIN operating Recirc pump flow less than 46,600 gpm. Refer to 2-01-68. [7] WHEN plant conditions allow, THEN, MAINTAIN operating jet pump loop flow greater than_41_x_106_lbmlhr_(2-FI-68-46_or_2-FI-68-48). SRO Direct inserting control rods lAW Urgent Load Reduction and Rod Shove Sheets ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Driver When dispatched to isolate seal water wait 5 minutes and then mrf rdO3 close and report closed 34

NRC Scenario 5 Simulator Event Guide: Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure

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

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

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

single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power High), Allowable Value of Table 3.3.1.1-1 is reset for

single loop operation; APPLICABILITY: MODES 1 and 2. Condition A: Requirements of the LCO not met. Required Action A. 1: Satisfy the requirements of the LCO. Completion Time: 24 hours. Driver When directed by NRC, Preference Key F9, Level Instruments Fail high When mode switch is out of run or NOT in run Preference Key Fl 0 35

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI ATC Report Trip of Main Turbine and RFPTs and Reactor Scram ATC 4.1 Immediate Actions [1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A153A and 2-HS-99-5A/S3B, on Panel 2-9-5. [2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA [3] REFUEL MODE ONE ROD PERMISSIVE light check: [3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL. [3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XJ-85-46. [3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, is not illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A) Step is NA [4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-Sl, in SHUTDOWN. [5] REPORT the following status to the US:

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

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

NRC Scenario 5 Simulator Event Guide: Event 7 Major: Loss of Feedwater and HPCI SRO Enters EOI- 1 on RPV Water Level SRO EOI-1 (Reactor Pressure) Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO

IF Emergency Depressurization is or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO - IF RPV water level cannot be determined? NO - Is any MSRV Cycling? No IF Steam cooling is required? NO

IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO

IF Drywell Control air becomes unavailable? NO

IF Boron injection is required? NO

Stabilize RPV pressure below 1073 psig with the main turbine bypass valves (APPX 8B) SRO Direct a pressure band, may direct a cooldown lAW Appendix 8B ATC/BOP Maintain directed pressure with Bypass Valves JAW Appendix 8B, Reopening MSJVs / Bypass Valve Operation 37

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

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

10. Verify Condenser Vacuum is greater than 7

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

a. Depress the Bypass Valve Opening Jack Raise Pushbutton, 2-HS 13 OB to slowly open the Bypass Valves.

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

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

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

b. Verify Bypass Valve Demand is set at ZERO

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

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

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

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

d. On the Auto Cooldown Display Screen

1) Check the following are displayed.

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

2) Select the block above the FiNAL PRESSURE TARGET

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

4) Select OK

5) Depress the START button

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

7) Check the following:

EHC PRESSURE SETPO1NT, 2-PI-47-162, is lowering
EHC AUTO COOLDOWN displays IN PROCESS 38

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

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

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

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

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

c. 2-FCV-71-40, RCIC Testable Check Vlv, opens by observing 2-ZI-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.

39

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Report rising Drywell Pressure and Temperature ATC/BOP Report loss of 480V SD BD 2A and 480V RMOV BD 2A ATC/BOP Dispatch personnel to investigate loss of Board SRO Re-Enter EOI-2 on High DW Pressure and Temperature ATC/BOP IF RHR Loop 1 was in Pool Cooling for leaking SRV, then operators report that RHR Loop 1 remains in Pool cooling. NOTE RHR Loop 1 has lost power to almost all valves but NO valves reposition on board loss SRO EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 160F using available Drywell cooling Can Drywell Temperature be maintained below 160F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7B) Driver When dispatched for Board loss, wait 4 minutes and report overcurrent trip of supply breaker on 480V SD BD 2A. If requested to energize 480V RMOV BD 2A from alternate supply, wait 3 minutes and report that unable to restore power to Board 40

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Dryweli Pressure PC/p Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1), PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using oniy those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue SRO Enters EOI-2 on High Drywell Pressure PC/H Verify 11202 analyzer in service (APP 19) When 112 is detected in PC (2.4% on control room indicators continue, does not continue 41

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvl between -1 in. and -6 in. (APPX 18) Can suppr p1 lvl be maintained above -6 in., YES Can_suppr p1_lvl_be maintained below -1_in.,_YES SRO Direct Suppression Chamber Sprays and Drywell Sprays on RHR Loop II ONLY 42

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays 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 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS SEL in BYPASS.

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

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

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

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

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

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

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

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

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

ATCIBOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 43

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

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 2-FCV-74-30, RHR SYSTEM II MIN 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:

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

n. NOTIFY Chemistry that RJ{RSW is aligned to in-service RHR Heat Exchangers.

ATC/BOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 44

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

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

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

3. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

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

6. INITIATE Drywell Sprays as follows:

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

b. IF EITHER of the following exists:

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

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

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

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

f. OPEN the following valves:

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

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

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

g. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II M1N FLOW VALVE.

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

i. MONITOR RUR Pump NPSH using Attachment 2.

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

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

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

I. NOTIFY Chemistry that Rl{RSW is aligned to in-service Rf1R Heat Exchangers.

7. WHEN EITHER of the following exists:

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

a. VERIFY CLOSED the following valves:

2-FCV-74-100, RRR SYS I U-2 DISCH XTIE
2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
2-FCV-74-75, RHR SYS II DW SPRAY II4BD VLV.

b. VERIFY OPEN 2-FCV-74-30, R}IR SYSTEM Ii MIN FLOW VALVE.

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

d. STOP R}IR Pumps.

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

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP Report lowermg RPV water level unable to maintam with RCIC SRO EOI- I Reactor Level RPV water level drops below -120 inches OR The ADS timer has initiated NO IF RPV water level CANNOT be restored and maintained between +2 and +51 inches, THEN Restore and maintain RPV water level above -162 inches Augment RPV water level control as necessary with any of the following SRO Directs additional level control systems: SLC (boron tank) APPX-7B CRD APPX-5B ATC/BOP Place SLC and an additional CR1) Pump in service lAW Appendix 7B and SB SRO EOI- 1 Reactor Level Can RPV water level be restored and maintained above -162 inches NO

SRO Announces entry to EOI-C- 1 Alternate Level Control RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO - IF RPV water level can be restored and maintained above -162 inches NO

Inhibit ADS ATC/BOP Inhibits ADS SRO Restore and maintain RPV water level above -162 inches using any of the following: Condensate APPX 6A LPCI System I APPX 6B LPCI System II APPX 6C Core Spray System II APPX 6E SRO Directs 2 or more of the above systems lined up for injection ATCIBOP Aligns the directed systems for Injection

47

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP Aligns CR1) and SLC JAW Appendix 5B and 7B ATC CRD Appendix 5B

2. IF BOTH of the following exist:

                                . CRD is NOT required for rod insertion, AND
                                . Maximum injection flow is required, THEN LINE UP ALL available CRD pumps to the RPV as follows:

a. IF CRD Pump 2A is available, THEN VERIFY RUNMNG CR1) Pump 2A.

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

1) NOTIFY Unit 1 Operator to verify closed 1-FCV-85-8, CR1) PUMP B DISCHARGE VALVE (Unit 1, Panel 9-5).

2) START CRD Pump lB.

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

c. OPEN the following valves to increase CR1) flow to the RPV:

                                . 2-PCV-85-23, CRD DRiVE WATER PRESS CONTROL VLV
                                . 2-PCV-85-27, CR1) CLG WATER PRESS CONTROL VLV
                                . 2-FCV-85-50, CR1) EXH RTN LINE SHUTOFF VALVE.

d. ADJUST 2-FIC-85-1 1, CR1) SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WFIILE maintaining 2-PI-85-13A, CR1) ACCUM CHG WTR HDR PRESS, above 1450 psig, if possible.

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

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

Driver When called as unitbne pertor FCV-85-8, CRDPUMP B DISCHAR.GEYALVEis closed 48

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

2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN_CONTINUE_at_Step_10 to_inject SLC Boron Tank to RPV.

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

11. CHECK SLC injection by observing the following:

                              . Selected pump starts, as indicated by red light illuminated above pump control switch.
                              . Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,

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

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

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

49

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

1. VERIFY CLOSED the following feedwater heater return valves:

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

2-FCV-3-73, HP HTR 2C1 LONG CYCLE TO CNDR.

2. VERIFY CLOSED the following RFP discharge valves:

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

3. VERIFY OPEN the following drain cooler inlet valves:

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

4. VERIFY OPEN the following heater outlet valves:

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

                             **   2-FCV-2-125, LP NEATER 2B3 CNDS OUTL ISOL VLV 2-FCV-2-l26, LP NEATER 2C3 CNDS OUTL ISOL VLV.

5. VERIFY OPEN the following heater isolation valves:

                              . 2-FCV-3-38, HP HTR 2A2 FW INLET ISOL VLV
                              . 2-FCV-3-3 1, HP HTR 2B2 FW INLET ISOL VLV

2-FCV-3-24, liP HTR 2C2 FW INLET ISOL VLV
2-FCV-3-75, HP HTR 2A1 FW OUTLET ISOL VLV
2-FCV-3-76, HP HTR 2Bl FW OUTLET ISOL VLV
2-FCV-3-77, HP HTR 2C1 FW OUTLET ISOL VLV.

6. VERIFY OPEN the following RFP suction valves:

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

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

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

10. VERIFY RFW flow to RPV.

50

NRC Scenario 5 Simulator Event Guide: Event 8 Component: Loss 480V SD BD 2A and LOCA SRO EOI-C-1 Alternate Level Control SRO Can 2 or more Condensate, LPCI or Core Spray injection subsystems be lined up YES - When RPV Water level drops to -162 inches Proceeds at TAF or -162 inches Is any Condensate, LPCI or Core Spray injection subsystems lined up for injection with at least one pump running YES - Is any RPV injection source lined up with at least one pump running YES

BEFORE RPV water level drops to -180 inches CONTINUE Continues - Emergency Depressurization is required Inject into the RPV with any available sources SRO Enters EOI-C-2 Emergency Depressurization Will the reactor remain subcritical without boron under all conditions YES Is DW pressure above 2.4 psig YES Prevent injection from only those Core Spray and LPCI pumps not required NO Is suppression pool level above 5.5 feet YES Open all ADS Valves Directs ADS valves open ATC/BOP Opens all 6 ADS valves, reports all ADS valves open When pressure is below the shutoff head of the available injection systems direct injection SRO to restore level to +2 to +51 inches ATC/BOP Injects with available systems to restore level SRO Emergency Classification EPIP-1 1.1-Si Reactor water level can NOT be maintained above -162 inches. (TAF) 51

NRC Scenario 5 Simulator Event Guide: Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, lAW BOP Appendix 6B, 6fl irni1 6E he pumps ope.an4 NOTE In addition i pressure to BOP Appendix 6B

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-l 55A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

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

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

4. VERIFY CLOSED the following valves:

                               . 2-FCV-74-61, RHR SYS I DW SPRAY 1NBD VLV
                               . 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV
                               . 2-FCV-74-57,         SYS I SUPPR CHBPOOL ISOL VLV
                                  .

2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

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

6. WREN RPV pressure is below 450 psig, TI{EN VERIFY OPEN 2-FCV-74-53, RHR SYS I LPCI INBD iNJECT VALVE.

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

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

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

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

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS_SEL in BYPASS.

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

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

4. VERIFY CLOSED the following valves:

                               . 2-FCV-74-75, RHR SYS II DW SPRAY INBD VLV
                               . 2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
                               . 2-FCV-74-71, RHR SYS II SUPPR CIIBRIPOOL ISOL VLV
                               . 2-FCV-74-72, RUR SYS II SUPPR CUBR SPRAY VALVE
                               . 2-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV

5. VERIFY RHR Pump 2B and/or 2D running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-67, RHR SYS II LPCI INBD iNJECT VALVE.

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

8. THROTTLE 2-FCV-74-66, RUR SYS II LPCI OUTBD INJECT VALVE, as necessary to control injection.

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

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

1. VERIFY OPEN the following valves:

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

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

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

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

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

NRC Scenario 5 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation. Temporary DGs are NOT provided. Operations/Maintenance for the Shift: Return LPRM 8-49B to Operate from a Bypassed Condition lAW 2-OI-92B. Lower Power with flow to 90% for Main Turbine Valve Testing. Unit 1 and 3 are at 100% Power Unusual Conditions/Problem Areas: Severe Thunderstorms are forecast for today, currently no watches or warnings are in effect. 55}}

ML13214A325: Difference between revisions

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ML13214A325: Difference between revisions

Revision as of 15:14, 4 November 2019

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