Text

Initial Exam 2011-301 Draft Simulator Scenarios
	ML110980768
Person / Time
Site:	Browns Ferry
Issue date:	04/04/2011
From:	; NRC/RGN-II/DRS/OLB
To:	; Tennessee Valley Authority
References
	50-259/11-301, 50-260/11-301, 50-296/11-301
	Download: ML110980768 (337)
	v • d • e

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

Examiners: Operators: ATC:

BOP:

Initial IC 190 / Unit 3 Reactor Power 83% / RHRSW Pump B2 is tagged out of service / APRM 3 Conditions: is bypassed for Surveillance Testing Turnover: Alternate Bus Duct Cooling Fans per 3-01-47 Section 6.1 1. 1[2]. Raise Reactor Power to 90% with Reactor Recirculation.

Event Event No. Maif. No. Type* Event Description N-BOP 1 Bus Duct Cooling Fan rotation 3-01-47 Section 6.1 1 .1 [2]

N..SRO R-ATC 2 Raise Reactor Power with Recirc R-SRO rdOla CRD Pump 3A Trip 1-BOP 4 og05a HWC Malfunction TS.SRO C-ATC 5 thl2a Recirc Pump 3A High Vibration CSRO hpOl C-BOP 6 HPCI Inadvertent Initiation TS..SRO hpO8 HPCI Steam Leak Fail to isolate / Loss of 480 V RMOV Bd 3A/ ED 7 hpO9 M-ALL onTemps th23 8 fwl2 C Startup Level Control Valve Failure 9 ad03b C 1 ADS Valve fails to operate

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

3-A Page 1 of 42 Console Operator Instructions A. Scenario File Summary

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

rhrsw pump B2 clearance ior ypobkrrhrswpb2 fail tcoil ior zlohs23 1 9a[ 1] off

aprm 3 bypassed for 3-sr-3.3.1.1.16

crd a pump trip imfrd0la(el 0)

hpci Initiation imfhpol (e5 0)

recirc pump a vibration high imfthl2a (elO 0)

hwc malfunction imfogo5a (e15 60) 99 iorxa5553a[10] (e15 0) alarm on trg 16nrc20110440 trg 16 = mmfog05a 100 360 99 Trigger nrc20110440 zdihs0440a[ 1] .eq. 1

HPCI Steam Leak/major (have to manually modify fj02 to close) mrffp02 (e20 0) close imfhp09 imfhp08 (e20 0) 8 600 4 trg2l nrc2011732 trg2l =imfedl2a ior ypovfcv733 (e20 0) fail_now imf fw 12 imf ad03b

if crew anticipates ED, may have to raise severity Trigger nrc2Ol 1732 zdihs732{1].eq.1

3-A Page 2 of 42 Console Operator Instructions Scenario 3-A DESCRIPTION/ACTION Simulator Setup manual Reset to IC 190 Simulator Setup Load Batch Bat nrc2OllaRl Simulator Setup manual Place APRM 3 in Bypass Simulator Setup Clearance out RHRSW Pump B2 and manual CRD Pump A Simulator Setup Verify Batch file loaded RCP required (83% 90% w/Recirc flow)

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

3-A Page 3 of 42 Scenario Summary:

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

When the NRC is satisfied with the reactivity manipulation, CRD Pump 3A will trip. ATC will perform 3-A0I-85-3 actions to start the Standby CRD Pump.

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

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

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

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

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

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

The Emergency classification is 3.1-S.

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

All Control Rods are inserted.

Emergency Depressurization complete.

Reactor Level is restored and maintained.

3-A Page 4 of 42 Simulator Event Guide:

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

SRO Directs BOP to rotate Bus Duct Cooling Fans.

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

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

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

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

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

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

Diiver Iriver Pre start walk down complete Inlet Daniper is Fully Open, Water has been drained from fan housing, B Fan is not rotating..

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

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

3-A Page 5 of 42 Simulator Event Guide:

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

Directs Power increase using Recirc Flow, per 3-GOI-l00-12.

[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, JAW 3-01-68, Section 6.2

[1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following; Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM),

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

AND/OR

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

3-HS-96-16A(1 6B).

[2] WhEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, TIIEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required:

RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32 c Whex sisflç4 with ReactMtyManij1ation, CRDPurnj 9tii1 DRWEi V1i&hrcfed by lead ex1çr, Tngger1 U) Pump Tnp

3-A Page 6 of 42 Simulator Event Guide:

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

SRO Announces entry into 3-AOI-85-3, CR1) System Failure.

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-l 1, in MAN at minimum setting.

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

CRD PUMP 3B, using 3-HS-85-2A

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

CRD CLG WTR HDR 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 CRD SYSTEM FLOW CONTROL, 3-FIC-85-11, and PLACE in AUTO or BALANCE.

f Dispaehed to CRD Pump 3% pump is extremely hot to touch.

CRDfp3B oil levels hi

- d eadyi% start, çnditipns normal start ORD 3 ieportbxeker tipped oii over current, Electrical Maint called.

When ready, JWC Malfunctionl.

Pnr Drtver ppnLead exanu4er duecIo, initiate Tngger 15 forIC1functio

3-A Page 7 of 42 Simulator Event Guide:

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

B. Dispatches personnel.

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

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

SRO Announces entry into 3-AOI-66-l, Off Gas 112 High.

DriYCr Whendispatehedto Panel report, 112 injection rates above (high) setpomt cannQt adjust BOP 3-AOI-66-1, Off Gas H2 High

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

[2.1] At HYDROGEN WATER CHEMISTRY CONTROL PANEL, 3-LPNL-925-0589, VERIFY that H2 and 02 injection rates are normal at Operator Interface Unit (OIU). (H2 injection rate should match the setpoint on the OIU. The 02 injection rate should match the setpoint on the ORJ,.

which should be half of the H2 injection rate during normal steady state conditions.)

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

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

[Panel 3-9-53] or

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

[Panel 3-9-5] or

. 3-HS-4-39 HWC SI{UTDOWN SWITCH [3-LPNL-925-0588].

Dri& 1?nrer f dije&ed to pcrfónn H CShutdow1oai1y inform Con&Rora tht scaffold is1ii e BOP Shutdown HWC System using either 3-HS-4-40A at panel 9-53 or 3-HS-4-40B at panel 9-5 SRO [4] IF hydrogen concentration is? 4%, THEN REFER TO TRM 3.7.2 itkc brIW?Yi 1Ier slowly

3-A Page 8 of 42 Simulator Event Guide:

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

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

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

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

On MAIN CONDENSERS (MN COND) ICS display:

Offgas pretreatment, post treatment, and stack radiation

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

NRC NR No mdøatioii ofFuel Failu iists, step S should be NA BOP Report H2 Concentration lowering slowly.

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

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

SRO REFER TO TRM 3.7.2 Condition A: With the concentration of hydrogen > 4% by volume Required Action A. 1: Restore the concentration to within the limit Completion Time: 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months c When ready, Recire Pump 3A.High Vibration..

Dnver Driver Upon Lead examiner direction, imtiate Tngger 10 for Recirc Pump 3A High Vibration

3-A Page 9 of 42 Simulator Event Guide:

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

BOP/ATC A. CHECKS temperatures for RECIRC PMP MTR 3A13B WINDiNG AND BRG TEMP recorder, 3-TR-68-71 on Panel 3-9-2 1 are below:

Pump motor bearing temperatures (< 190°F)

Pump motor winding temperatures (<255°F)

Pump Seal Cavity temperatures (< 180°F)

Pump cooling water from Seal Cooling temperature (< 140°F)

Pump motor cooling water from bearing temperature (< 140°F)

B. CI[ECKS for a rise in Drywell equip sump pumpout rate, due to seal leakage.

C. DISPATCHES personnel to 3-LPNL-925-0712, (Vibration Mon. System) on EL 565 (S-Ri 7), to REPORT the Vibration Data for Pump A and any other alarm indications, to the Unit Operator. The person shall advise the Unit Operator of any changes in the vibration values.

D. IF alarm seals in, THEN ADJUST pump speed slightly to try reset the alarm.

E. IF unable to reset alarm, THEN CONSULT with Unit Supervisor, and with his concurrence, SHUTDOWN the Recirc pump and REFER TO 3-AOI-68-1A or 3-AOI-68-1B.

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

t5Ij Pr( aaj ln points lind point o5fYi 12.5mi1s. After speed is1owere4,vibratioreading Iowered1ightiy, point 59D is 12 uls, ATC Lowers Pump Speed in an attempt to reset high vibration alarm.

Driver Dnv IF Speed isTowered a secona time, ibratioa&gs lowered aiarn ndpomt 59D is 10 inilsTllEN Delete thl2a SRO Determine whether to remove RR Pump 3A.

ATC Records seal parameters hourly for RR Pump 3A.

upjggçpji

3-A Page 10 of 42 Simulator Event Guide:

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

ATC 7.2 Stopping a Recirc Pump (Mode I) & Single Loop Operation CAUTIONS

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

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

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

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

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

Recirc Drive 3A RAISE SLOW, 3-HS-96-15A RAISE MEDRJM, 3-HS-96-l5B LOWER SLOW, 3-HS-96-17A LOWER MEDIUM, 3-HS-96-17B LOWER FAST, 3-HS-96-17C tri Driver If Reacior Engmeer s contacted, inform crew to follow Urgent Load Reduction RCP NRC When ready, BPCI Inadvertent Initiation.

Dnvei Drive UoiI&dçx&u& dueetion, qteTng&3 for IGflnijatLon

3-A Page 11 of 42 Simulator Event Guide:

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

ATC [6] To shutdown Recirc Drive 3A:

PERFORM the following: (Otherwise N/A)

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

[6.2] VERIFY Recirc Drive shuts down.

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

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

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

NRb nit atiou nvei Driver ruttio

3-A Page 12 of 42 Simulator Event Guide:

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

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

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

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

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

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

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

dyer Driver Acknowledge Notifications and directions.

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

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

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

[1.2] OBSERVE stable steam flow and Feedwater flow.

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

VERIFY green backlight for push-button illuminates.

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

[1 .5] CITECK_Reactor water level_stable.

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

NRC NRC When Ready, Mjc>rHPCI Steam Leaki Driver Driyçr Upon Lead examin direction, initiate Tngge 20 fqr IWCI Steam Leak

3-A Page 13of42 Simulator Event Guide:

Event 6 Component: HPCI Inadvertent Initiation BOP Reports Suppression Chamber Water Level Abnormal, greater than (-) 1.

SRO Enters EOI-2.

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

BOP Checks ECCS systems for sources of water.

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

Crew Directs AUO to valve locally to isolate.

Driver Drivei When dispatched, wait 3 niinutes and report ready to isblate at breaker. When dIrected by opertor,GO T** Component Ovenide, TUEN System 73, TUEN FCV-73-30 Fail Now.

SRO Directs pump down of Torus per App 18.

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

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

Driver 1ri When dispatcbe fait iunutes and report hieçL ip lopally to pump to2us BOP Aligns to pump down torus in Control Room, per Appendix 18.

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

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

1) ESTABLISH communications with Radwaste.

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

BOP Directs AUO to Start R}IR Drain Pump.

p. DflY WendedtartEBRPrainPump,IRFRHO9orRH1O and RH1 1A.or WhenReady, Nicr lPCLStea Lea Driver Dri tlpon Lead eminr direction, initiate nggr 2Q for HPCI Steam Leak:

3-A Page 14of42 Simulator Event Guide:

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

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

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

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

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

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

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

Attempts to isolate HPCI Steam Supply Valves.

Reports HPCI fails to isolate.

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

(Loop 1RHR and Loop 1 Core Spray unavailable.)

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

Crew Dispatches personnel to transfer RPS A to alternate.

Driver brver When reqsted, wait 4xmriites and place BPS A on mte IRF RPO4 and Rl?03

3-A Page 15 of 42 Simulator Event Guide:

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

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

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

ATC/BOP 3-EOI Appendix 8F

1. VERIFY PCIS Reset.

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

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

b. PLACE 3-HS-64-3A, REFUEL ZONE FANS ANT) 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 INBD ISOL DMPR

3-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR

3-FCO-64-1O, REFUEL ZONE EXH INBD ISOL DMPR V

Y V Dnver Priveti frequestedwait3 m eaiid repor ppep4!Ep9napIete, enterbat appOS.

3-A Page 16 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak SRO EOI-3 Secondary Containment (Temperature)

Monitor and Control Secondary Containment Temperature.

Operate available ventilation, per Appendix 8F.

Is Any Area Temp Above Max Normal? YES -

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

. Be operated by EOIs OR

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

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

CS#1 CS#1 Enters EOI-l RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe.

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

CS#2 Crew Monitors for Max Safe Temperatures, reports when two areas are above MAX Safe (HPCI Room greater than 270°F and RHR System II Pump Room greater than 215°F)

SRO EOI-3 Secondary Containment (Level)

Monitor and Control Secondary Containment Water Levels.

Is Any Floor Drain Sump Above 66 inches? NO Is_Any Area Water Level_Above_2_inches? NO -

3-A Page 17of42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak SRO EOI-3 Secondary Containment (Radiation) f Monitor and Control Secondary Containment Radiation Levels.

Is Any Area Radiation Level Max Normal?

1

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

. Be operated by EOIs OR

. Suppress a Fire

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

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

3-A Page 18 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak CS#1 SRO Enters EOI- I, RPV Control and directs Reactor Scram.

CS#1 ATC Scrams the Reactor and places the Mode Switch in Shutdown.

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

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

CS#2 CS#2 IF Emergency Depressurization is required, THEN exit RC/P and enter C2 Emergency Depressurization.

Answers YES; when two area temperatures have reached MAX Safe.

IF RPV water level cannot be determined? NO-Is any MSRV Cycling? NO -

IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? - NO IF Drywell Control air becomes unavailable? NO IF Boron injection is required? NO SRO Directs a Pressure Band. Should begin to lower Reactor Pressure with bypass valves, not to exceed 1000 cooldown; until SRO decides that ED is anticipated.

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

When directed to Anticipate ED, Opens all bypass valves.

3-A Page 19 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak SRO Reactor Level Monitor and Control Reactor Level Verify as required PCIS isolations group (1,2 and 3), ECCS and RCIC, Directs group 2 and 3 verified.

ATC/BOP Verifies Group 2 and 3 isolation.

SRO IF It has not been determined that the reactor will remain subcritical? NO IF RPV water level cannot be determined? NO -

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

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

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

3-A Page 20 of 42 Simulator Event Guide:

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

1. IF It is desired to use a reactor feed pump that is in operation, THEN CONTINUE at step 12 to control the operating pump.

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

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

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

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

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

VERIFY OPEN the following valves:

3-FCV-3-75, HP HTR 3A1 FW OUTLET ISOL VLV 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VLV 3-FCV-3-77, HP HTR 3C1 FW OUTLET ISOL VLV DEPRESS 3-HS-3-124A(150A)(175A), RFPT 3A(3B)(3C) TRIP RESET, and Verify that the turbine trii is RESET.

3-A Page 21 of42 Simulator Event Guide:

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

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

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

11. VERIFY OPEN 3-FCV-3-l9(12)(5), RFP 3A(3B)(3C) DISCHARGE VALVE.

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

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

Individual 3-SIC-46-8(9)(l0), RFPT 3A(3B)(3C) SPEED CONTROL in MANUAL, OR

3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 3-SIC-46-8(9)(l0), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO.

13. ADJUST RFPT speed as necessary to control injection using the methods of step 12.

14. WhEN RPV level is approximately equal to desired level AND automatic level control is desired, THEN PLACE 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)_SPEED CONTROL in AUTO.

3-A Page 22 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak CS#2 SRO Enters 3-C-2, Emergency Depressurization.

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

CS#2 ATC/BOP Opens 6 ADS Valves.

Reports 1 ADS Valve failed to Open.

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

ATC/BOP Opens 1 additional MSRV.

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

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

SRO Emergency Plan Classification is 3.1-S.

3-A Page 23 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC Appendix 6A Injection with Condensate

1. VERIFY CLOSED the following Feedwater heater return valves:

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

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

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

2. VERIFY CLOSED the following RFP discharge valves:

. 3-FCV-3-19, 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

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

4. VERIFY OPEN the following heater outlet valves:

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

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

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

5. VERIFY OPEN the following heater isolation valves:

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

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

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

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

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

3-FCV-3-77, HP HTR 3Cl FW OUTLET ISOL VLV

6. VERIFY OPEN the following REP suction valves:

3-FCV-2-83, RFP 3A SUCTION VALVE

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

. 3-FCV-2-108, RFP 3C SUCTION VALVE

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

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

ATC Reports failure of Start Up Level controller.

3-A Page 24 of 42 Simulator Event Guide:

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

1. VERIFY OPEN the following valves:

. 3-FCV-75-30, CORE SPRAY PUMP 3B SUPPR POOL SUCT VLV

. 3-FCV-75-39, CORE SPRAY PUMP 3D SUPPR POOL SUCT VLV

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

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

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

4. WhEN RPV pressure is below 450 psig, THEN ThROTTLE 3-FCV-75-53, CORE SPRAY SYS II fNBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

5. MONITOR Core Spray Pump NPSH using Attachment 1.

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

3-A Page 25 of 42 Simulator Event Guide:

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

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

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

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

4. VERIFY CLOSED the following valves:

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

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

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

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

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

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

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, PER 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, PER HX 3D RHRSW OUTLET VLV.

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

3-A Page 26 of 42 Simulator Event Guide:

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

EOI-2 (Drywell Temperature)

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

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

SRO Verifi H202 Analyzers placed in service, Appendix 19.

BOP Places H202 analyzers in service, LAW Appendix 19.

SRO EOI-2 Primary Containment (Pressure)

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

(Appendix 12)

Can Primary Containment pressure be maintained below 2.4 psig? YES-SRO EOI-2 Suppression Pool (Temperature)

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

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

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

BOPIATC Start RHR Loop 2 in Suppression Pool Cooling, if not being used for level control, LAW Appendix 17A

3-A Page 27 of 42 Simulator Event Guide:

Event 7 Major: HPCI Steam Leak ATC/BOP Initiates Suppression Pool Cooling per Appendix 1 7A IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary, by PLACING 3-HS-74-155A(B), LPCI SYS 1(11)

OUTBD INJ VLV BYPASS SEL in BYPASS.

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

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

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

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

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV

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

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

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

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

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

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

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

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

3-A Page 28 of 42 Simulator Event Guide:

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

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

(Touch screen actions unavailable in the simulator)

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

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

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

3-A Page 29 of 42 Critical Tasks Two-CS#1 With reactor at power and with a primary system discharging into the secondary containment, manually scram the reactor before any area exceeds the maximum safe operating level.

1. Safety Significance:

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

2. Cues:

Procedural compliance Secondary containment area temperature, level, and radiation indication.

Field reports.

3. Measured by:

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

OR Observation With a primary system discharging into secondary containment, US transitions to EOP- 1 and RO initiates scram upon report that a maximum safe condition has been reached.

4. Feedback:

Control rod positions Reactor power decrease CS#2-With a primary system discharging into the secondary containment, when two or more areas are greater than their maximum safe operating values for the same parameter, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

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

2. Cues:

Procedural compliance Secondary containment area temperatures, level, and radiation indications 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

3-A Page 30 of 42 Scenario Tasks EVENT TASK NUMBER K/A RO SRO 1 Rotate Bus Duct Cooling Fans RO U-047-NO-27 400000A4.01 3.1 3.0 2 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 3 CRD Pump Trip RO U-085-AL-07 201001A2.O1 3.2 3.3 SRO S-085-AB-03 4 Hydrogen Water Chemistry Malfunction RO U-066-AL-10 271000A1.13 3.2 3.7 SRO S-066-AB-01 5 Reactor Recirculation Pump High Vibrations RO U-068-AL-1 1 202001A4.05 3.3 3.3 6 HPCI Inadvertent Start RO U-073-NO-05 206000A2.17 3.9 4.3 7 HPCI Steam Leak RO U-073-AL-06 295032EA2.03 3.8 4.0 SRO S-000-AB-03 SRO S-000-EM-12 SRO T-000-EM-16

3-A Page 31 of 42 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-A 7 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List(1-3) 1 EOI Contingencies used: List (0-3) 60 Run Time (minutes) 2 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

3-A Page 32 of 42 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

RHRSW Pump B2 is out of service and tagged out.

APRM 3 is bypassed for IMD Surveillance testing.

Operations/Maintenance for the Shift:

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

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

Units 1 and 2 are at 90% power.

Unusual Conditions/Problem Areas:

None

FTI Fr 9N 1 0I M

0 mm CL UCA) r CD C?)

(31 C

-h NJ

aT:[

m

3-A Page 37 of 42 Airborne Effluents TR 3.7.2 TR 31 PLANT SYSTEMS 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%

byvotume, APPUCAGIUTY: During main condenser offgas treatment system operation NOTE.

TRM ICO 3.0.3 is not applicable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. With the concentration A,1 Restore the 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of hydrogen >4% by concentration to within volume.. the limit.

3-A Page 38 of 42 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS> AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injectionlspray subsystem and the Automatic Depressuilzation System (ADS) function of six safety/rellef valves shall be OPERABLE..

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

ACTIONS LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS Al Restore low pressure 7 days injectianfspray subsystem ECCS irjectionIspray inoperable, subsystem(s) to OPERABLE status.

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

(continued)

3-A Page 39 of 42 ECCS Operating 3.51 ACTIONS (continued)

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

B.2 Be tn MODE 4. 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months (continued)

3-A Page 40 of 42 ACTIONS (continued)

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

AND Cl Restore HPCI System to 14 days OPERABLE status.

D. HPCI System inoperable. Di Restore HPC1 System to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND OR Condition A entered.

D,2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injectionispray subsystem to OPERABLE status.

E. One ADS valve E. 1 Restore ADS valve to 14 days inoperable. OPERABLE status.

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

AND QR Condition A entered. F.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injectionlspray subsystem to OPERABLE status.

(continued)

3-A Page 41 of 42 ACTIONS (continued)

CONDiTION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves Gi Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable AND OR G2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion 150 psig.

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

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

OR HPCI System and one or more ADS valves inoperabIe

3-A Page 42 of 42 SECONDARY CONTAINMENT TEMPERATURE Description a 1 z

a 0

C 1 I I I r

m 1

3.1-S I I I TABLE I US I 0

An unisolable Pnrnary System leak rs dschargrng rnto Secondary Contamment m

AND Any area temperature exceeds the Maximum Sate Operating Temperature limit listed in Table 3.1.

OPERATING CONDrnON:

Modelor2or3 -<

31-G I I I TABLE US I An unisotable Primary System teak is discharging into Secondary Containment AND Any area temperature exceeds the Maximum Safe Operating Temperature limit listed in Table 3.1 AND Any indication of potential or significant fl.iel cladding failure exists. Refer to Table 3,l43/3.2G with RC$ Barrier intact inside Primary Containment m z

OPERAI1NG CONDITION Model or2or3

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

Examiners: Operators: ATC:

BOP:

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

Turnover:

Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 per 3-SR-3.6.1.3.5 Section 7.3. Raise Reactor Power to 95%.

Event Event No. Maif. No. Type* Event Description N-BOP 1 Stroke time 2 PCIVs. The second valve will fail open.

TS-SRO R-ATC 2 Raise Reactor Power with Recirc R-SRO 3 thl8d VFD Cooling Water Pump failure C-BOP Steam Packing Exhauster Trip / STBY Exhauster Starts but discharge 4 g C-SRO damper fails to open.

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

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

3-B Page 1 of 57 Console Operator Instructions A. Scenario File Summary

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

raw cooling water pump a clearance ior zlohs247a[1j off

surveillance 3.6.1.5 section 7.3 ior zlohs43 1 4a[ 1] (e3 5) off ior zlohs43 1 4a[2] (e3 5) on ior zlofcv43l4[1] (e3 5) off ior zlofcv43l4[2] (e3 5) on ior zloil64lb6{1] (e3 5) off trg3 nrc2Ol 14314 Trigger nrc20114314 zdihs43 14a(3).eq. 1

wide range pressure bypassed 3-207

vfd cooling pump failure ior zlohs682b2a[ 1] on ior zlohs682b2a[2] off mrfthl8d trip ior zdihs682bla[1] (el 0) off trg 2 nrc2Ol lbvfd trg2=batnrc2ollbl Trigger nrc2Ol lbvfd zdihs682b2a(3) .eq. 1 Batch nrc2Ollbl mrfthl8d close dor zlohs682b2a[ 1]

dor zlohs682b2a[2]

3-B Page 2 of 57

RBCCW pump trip imf sw02a (e5 0) ior zlohs7O48a[1] off ior zlohs7O48a[2] on ior xa554c19 alarm off trg 6 nrc2Ol 17048 trg 6 = bat nrc2Oi 1b2 Trigger nrc20117048 zdihs7o48a[1].eq. 1 Batch nrc2Ollb2 dor zlohs7O48a[i]

dor zlohs7O48a[2]

Steam packing blower trip ior ypomtrspea (eli 0) fail_control_power ior ypovfcv6635 (eli 0) fail_power_now ior zlohs6635a{i] on trg l0nrc20llspe trg 10 bat nrc2Ol ispe Trigger nrc2Ol ispe zdihs6635a{3].eq. 1 Batch nrc2Ollspe dor ypovfcv6635 dor zlohs6635a[1]

RHR A leak imfpcl4 (e15 0)10 iorxa554c[17] (e15 30) alarm_on ior xa554c[24] alarm_off ior xa554c[30] alarm_off ior xa554c{3 1] alarm_off

Major imfth33a (e20 0) .8 15 imftc02 (e20 0) 0 trg 25 nrc2Ol ldwspray2 ior zdihs7475a[2] auto imfrp07 imfth33b (e25 0) .5 180 Trigger nrc2Olldwspray2 zdihs7474a(3).eq. 1

3-B Page 3 of 57 Console Operator Instructions Scenario 3-B DESCRIPTION/ACTION Simulator Setup manual Reset to IC 191 Simulator Setup Load Batch Bat nrc2Ol lb Simulator Setup Place Green covers on Reactor manual Pressure indications two places.

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

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

Marked up Copy of 3-SR-3.6.1.3.5, for section 7.3 performance.

3-B Page 4 of 57 Scenario Summary:

BOP will perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 with 3-FCV-43-14 failing open. SRO will determine Technical Specification 3.6.1.3 Condition A required.

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

Once evaluators satisfied with Reactivity Manipulations, the VFD Cooling Water Pump for the B Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water, preventing a VFD and Reactor Recirc Pump trip.

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

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

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

A LOCA will occur, RPS will fail to de-energize, the crew will scram the Reactor by arming and depressing ARI, and enter EOI-l and EOI-2. All rods will insert on ARI, level control will be on feedwater; and pressure control will be on SRVs. The bypass valves fail closed during the scram.

The LOCA will cause increasing DW Pressure and Temperature; the crew will take action lAW EOI-2. When the crew attempts to spray the Drywell, the Drywell Spray valves will fail to open.

Unable to spray the drywell; the crew will need establish limits for DW pressure and temperature for anticipating ED and ED.

The Emergency classification is 2.1-A.

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

All Control Rods are inserted.

Emergency Depressurization is complete Reactor Level is restored and maintained.

3-B Page 5 of 57 Simulator Event Guide:

Event 1 Normal: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 JAW 3-SR-3.6.1.3.5 Section 7.3 SRO Directs BOPto perform 3-SR-3.6.1.3.5, Section 7.3.

BOP Performs 3-SR-3.6.l.3.5, Section 7.3.

7.3 Sampling and Water Quality System Valve Closure Timing

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

[2] On 3-LPNL-925-0009B (RB 621, near steps to Precoat Tank), PLACE the following switches to the OPEN position:

REACTOR RECIRC SAMPLE 1NBD ISOL VLV, 3-HS-043-0013B

REACTOR RECIRC OUTBD ISOLATION VLV, 3-HS-043-0014B

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

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

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

Normal Measured Maximum 0.6-1.6 5.0

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

NA [5] IF the stroke time measured in Step 7.3 [4] is less than or equal to the maximum stroke time value but outside the normal range, THEN PERFORM the following:

(Otherwise N/A this section.)

[5.1] OPEN the 3-FCV-43-13 using RX RECIRC SAMPLE 1NBD ISOLATION VLV, 3-HS-43-13A.

[5.2] CLOSE and TIME 3-FCV-43-13 using RX RECIRC SAMPLE INBD ISOLATION VLV, 3-HS-43-13A and RECORD the stroke time on Attachment 3.

[5.3] VERIFY the restroke time is less than or equal to the maximum closure time.

3-B Page 6 of 57 Simulator Event Guide:

Event 1 Normal: Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 JAW 3-SR-3.6.1.3.5 Section 7.3 BOP [6] RETURN 3-FCV-43-13, to the initial position recorded in Step 7.3[1], using RX RECTRC SAMPLE INBD ISOLATION VLV, 3-HS-43-13A.

[7] RECORD the initial position of RX RECIRC SAMPLE OUTBD ISOLATION VLV,_3-FCV-43-14._OPEN / CLOSED (Circle one)

[8] OPEN or VERIFY OPEN 3-FCV-43-14, using 3-HS-43-14A.

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

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

Normal Measured Maximum 0.4-1.4 5.0 Report Failure of 3-FCV-43-14 to stroke close.

SRO Dispatches personnel to investigate.

Refer to Technical Specification 3.6.1.3.

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

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

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

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

Completion Time: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for main steam line and Once per 31 days for isolation devices outside_primary_containment

3-B Page 7 of 57 Simulator Event Guide:

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

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

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

RAISE power using control rods or core flow changes.

REFER TO 3-SR-3.3.5(A) and 3-01-68.

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

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

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

3-HS-96-1 5A(i 5B).

AND/OR

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

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

[2] WhEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required:

RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-1-15-96-32 W poi çgWatexmp i1 WnJg

3-B Page 8 of 57 Simulator Event Guide:

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

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

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

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

SRO Concurs with start of Standby VFD Pump.

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

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

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

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

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

Crew Verifies Standby pump started by pulling up ICS displays.

BOP Dispatches personnel to VFD.

Tá4niinutes after dispatched, TUE çporttrippe1 VFD Ppmp is hot to touch, interna bkr c1osed 480 o1tb ifripped(48O V SD3A5D Dtt UIa dIiW&btioi, Ti aWauer fnp

3-B Page 9 of 57 Simulator Event Guide:

Event 4 Component: SPE Packing Exhauster Trip BOP Responds to Alarm 7A-12, Steam Packing Exhauster Vacuum Low.

7A-l2, Steam Packing Exhauster Vacuum Low Automatic Action: Alternate SPE fan starts and discharge damper opens, and the running fans trips.

A. CHECKS the following:

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

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

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

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

rrey.t?,

pit river npathjts dorrI&x.

N w,

Driver pt d çxa it Thgger I

3-B Page 10 of57 Simulator Event Guide:

Event 5 Component: RHR A Leak BOP/ATC Respond to Alarm 4C-l7 RHR LOOP I PUMP ROOM FLOOD LEVEL HIGH, A. DISPATCH personnel to visually check the RHR pump room.

B. IF alarm is valid, THEN PERFORM the following

VERIFY the floor drain sump pumps running.

VERIFY the floor drains for proper drainage.

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

ENTER 3-EOI-3 FLOWCHART.

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

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

C. REFER TO Tech Spec Section 3.6.2.6.

BOP/ATC Dispatches personnel to RHR Loop 1 area.

SRO Evaluates Tech Spec 3.6.2.6 and Enters EOI-3.

Ae esIshd portl&k3 oinm6n. ii1In iRü Pumps A and C ppearJeakwas caused by mntnaeewo+/-itfhe area, When thee 1Qses74 ana74-L2 report leak has stopped. Cannot acceWany manual valvep dueto amount ofwater spray,. Ifon1 one qftheRflR SupIesionPoø1 Suction Vaies is clod epo tbatleak has &s1owed. lii addfflo i about 8 inobeJnThi quad t1 shnber BOP/ATC Respond to Alarm 4C-3, SUPPR CHMB RM FLOOD LEVEL HIGH A. DISPATCH personnel to VISUALLY CHECK the suppression chamber room.

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

CHECK the floor drain sump pumps running.

CHECK the floor drains for proper drainage.

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

ENTER 3-EOI-3 FLOWCHART.

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

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

3-B Page 11 of57 Simulator Event Guide:

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

Monitor and Control Secondary CNTMT Water Levels.

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

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

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

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

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

iriver Drer After 744 an4 7442 are isolated, REPORT sump pumps are oper ngnorma1y, in reaof alarm. DELETE overrides on alarms, xa554e[3j alarm on and br xa554cj17] alarm on SRO EOI-3 (Temperature)

Monitor and Control Secondary Containment Temperatures.

Operate all available ventilation. (Appendix 8F)

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

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

SRO EOI-3 (Radiation)

Monitor and Control Secondary CNTMT Radiation Levels.

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

Driver Upon Lead exam nerdirction initiate TriggerS for Loss of RBCCW.

3-B Page 12 of 57 Simulator Event Guide:

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

Required Action A. 1: Restore low pressure ECCS injectionlspray subsystem to Operable status.

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

Required Action B. 1: Restore one RHR suppression pooi cooling subsystem to Operable status.

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

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

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

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

Completion Time: 7 Days TS 3.6.2.6: No Entry required Dner Up:nJ xgerdirection,. iia gger rJp 4C

3-B Page 13 of 57 Simulator Event Guide:

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

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

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

B. VERIFY RBCCW pumps A and B in service.

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

D. DISPATCH personnel to check the following:

RBCCW surge tank level locally.

RBCCW pumps for proper operation.

E. REFER TO 3-AOI-70-1, for RBCCW System failure and 3-01-70, for starting spare pump.

SRO Enters 3 -AOI 1.

ATC Closes 3-FCV-70-48 and report the sectionalizing valve failed to close automatically BOP Dispatch Personnel to investigate RBCCW Pump 3A trip Dnvç pwe When dispatched, repoxt qCWPump 3çis tripped f Thçre Io smell rniajig adq1iapg op the jaket ATC 3-AOI-70-l 4.1 Immediate Actions

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

SECURE RWCU Pumps.

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

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

3-B Page 14 of 57 Simulator Event Guide:

Event 6 Component: Loss of RBCCW 4.2 Subsequent Actions

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

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

Step 1 and 2 are NA

[3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (Otherwise N/A):

[3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions.

[3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s).

Dive Thivb Waen dispatched, reppr RCCWPump A tppeç1 free, Th is also siiç pf SRO [4] IF unable to restart a tripped pump, THEN PLACE Spare RBCCW Pump in service. REFER TO 3-01-70. Direct Unit 1 to place Spare RBCCW Pump in service Driye Drivqr called tojace pareBCeW Ptimp iservicwait 10 mnutes (J13F SW02).

giVit SRO [5] IF RBCCW flow was restored to two pump operation by placing the Spare RBCCW pump in service in the preceding step, THEN PERFORM the following:

[5.1] REOPEN RBCCW SECTIONALIZING VLV, 3-HS-70-48A.

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

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

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

NR wi oiWcn

3-B Page 15 of 57 Simulator Event Guide:

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

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

ATC Manually scrams the reactor.

CS#2 ATC Reports RPS failed to de-energize and initiates one channel of ART.

CS#2 ATC Verifies all Rods In, on ART Initiation.

SRO Enters EOI-1 and EOI-2.

SRO EOI-1 (Reactor Pressure)

Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? YES, but action Not Required IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO -

IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? NO-IF RPV water level cannot be determined? NO -

ATC/BOP Report failure of Bypass Valves to control Reactor Pressure Is any MSRV Cycling? YES Direct Manually open MSRVs until RPV Pressure drops to the pressure at which all turbine bypass valves are open. (Appendix 11 A) iF Steam cooling is required? NO-IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?- NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO -

IF Boron injection is required? NO-SRO Directs a Pressure Band with SRVs, JAW Appendix 1 1A.

Should begin to lower Reactor Pressure, not to exceed 1000 cooldown.

Control Reactor Pressure in assigned band, JAW Appendix 11 A.

3-B Page 16 of 57 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Pressure Control JAW Appendix hA, RPV Pressure Control SRVs IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTW CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.

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

3. OPEN MSRVs, using the following sequence, to control RPV pressure as directed bySRO:

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

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

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

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

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

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

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

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

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

j. 3-PCV-1-41 MN STM LINED RELIEF VALVE

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

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

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

3-B Page 17 of 57 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Pressure Control lAW Appendixi 1A RPV Pressure Control SRVs NA 4. IF Drywell Control Air header supplied from CAD System A, shows indications of being depressurized, as determined by Appendix 8G, THEN OPEN MSRVs supplied by CAD System B; using the following sequence to control RPV pressure; as directed by SRO:

NA 5. IF Drywell Control Air header supplied from CAD System B, shows indications of being depressurized, as determined by Appendix 8G, THEN OPEN MSRVs supplied by CAD System A; using the following sequence to control RPV pressure; as directed by SRO:

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

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

SRO EOI- 1 (Reactor Level)

Monitor and Control Reactor Water Level.

Directs Verification of PCIS isolations.

ATC/BOP Verifies PCIS isolations.

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

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

3-B Page 18 of 57 Simulator Event Guide:

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

1. IF It is desired to use a reactor feed pump that is in operation, THEN CONTINUE at step 12 to control the operating pump.

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

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

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

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

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

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

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

7. VERIFY OPEN the following valves:

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

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

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

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

3-B Page 19 of57 Simulator Event Guide:

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

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

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

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

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

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

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

3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 3-SIC-46-8(9)(l0), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO.

13. ADJUST RFPT speed as necessary to control injection, using the methods of step 12.

14. WHEN RPV level is approximately equal to desired level AND automatic level control is desired, THEN PLACE 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)_SPEED CONTROL in AUTO.

3-B Page 20 of 57 Simulator Event Guide:

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

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

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

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

BOP Places H202 analyzers in service, JAW Appendix 19.

SRO EOI-2 (Primary Containment Pressure)

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

SRO Directs venting of Primary Containment, per Appendix 12.

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

EOI-2 (Suppression Pool Temperature)

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

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

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

3-B Page 21 of57 Simulator Event Guide:

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

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

Can Suppression Pool Level Be Maintained above (-) 6 inches? - YES Can Suppression Pool Level Be Maintained below (-) 1 inch? - YES BOP Places H202 analyzers in service, JAW Appendix 19.

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

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

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

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

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

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

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

a. PLACE 3-11-76-110, 112/02 ANALYZER DW/SUPPR CHBR SELECT in DRYWELL position.

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

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

3-B Page 22 of 57 Simulator Event Guide:

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

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

a. TOUCH 3-MON-76-1 10 display screen.

b. DEPRESS Go To Panel PROCESS VALUES soft key.

c. DEPRESS Go To Panel MA1NT MENU soft key.

d. DEPRESS LOG ON soft key.

e. ENTER password 1915 on soft keypad.

f. DEPRESS ENT soft key on keypad.

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

h. VERIFY soft key reads STANDBY MODE OFF.

i. DEPRESS Go To Panel PROCESS VALUES soft key.

j. DEPRESS Go To Panel MAIN soft key.

k. VERIFY STANDBY MODE is NOT displayed.

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

7. VERIFY red LOW FLOW indicating light extinguished at 3-MON-76-1 10, H2/O2 ANALYZER (Panel 3-9-55).

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

3-B Page 23 of 57 Simulator Event Guide:

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

1. VERIFY at least one SGTS train in service.

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

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

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

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

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

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

7. CONTINUE in this procedure at:

Step 8 to vent the Suppression Chamber through 3-FCV-84-19, OR Step 9 to vent the Suppression Chamber through 3-FCV-84-20.

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

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

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

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

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

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

f. CONTINUE in this procedure at step 12.

3-B Page 24 of 57 Simulator Event Guide:

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

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

a. VERIFY OPEN 3-FCV-64-14l, DRYWELL DP COMP BYPASS VALVE (Panel 3-9-3).

b. PLACE keylock switch 3-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 3-9-54).

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

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

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

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

g. CONTINUE in this procedure at step 12.

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

Stable flow as indicated on controller, AND 3-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:

iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 iiCils ANT) 0-SI-4.8.B.1.a.1 release fraction of 1.

3-B Page 25 of 57 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC Place Suppression Pool Cooling in service JAW Appendix 17A IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary; by PLACING 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. PLACE RFIR SYSTEM II in Suppression Pool Cooling as follows:

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

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

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

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

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

d. IF Directed by SRO, THEN PLACE 3-XS-74-130, RHR. SYS II LPCI 2/3 CORE HEIGHT OVRD 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 INED INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-66, RHR SYS II

LPCI OUTBD INJECT VALVE.

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

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

i. TIIROTTLE 3-FCV-74-73, RHR SYS II 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-74-30, RHR SYSTEM II MIN FLOW VALVE.

k. MONITOR RHR Pump NPSH using Attachment 1.

3-B Page 26 of 57 Simulator Event Guide:

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

Operate all available Drywell Cooling.

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

SRO EOI-2 Primary Containment Pressure Before Suppression Chamber Pressure rises to 12 psig, Continue Initiate Suppression Chamber Sprays, Using only pumps not required to assure adequate core cooling by continuous injection. (Appendix 1 7C)

SRO Directs Operator to initiate Suppression Chamber Sprays, LAW Appendix 17C.

ATC/BOP Initiates Suppression Chamber Sprays, JAW Appendix 1 7C.

3-B Page 27 of 57 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP 1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

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

Step 3 and 4 are NA.

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR

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

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

f. VERIFY OPEN 3-FCV-74-71, RIlE. SYS II SUPPR CHBRIPOOL ISOL VLV.

3-B Page 28 of 57 Simulator Event Guide:

Event 7: Main Steam Line Leak inside Containment

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

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

i. VERIFY CLOSED 3-FCV-74-30, 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-23-46, RHR HX 3B RHRSW OUTLET VLV

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

3-B Page 29 of 57 Simulator Event Guide:

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

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

ATC Trips Reactor Recirculation Pumps.

BOP Places all Drywell Blowers in Off.

SRO Initiate DW Sprays, using only pumps not required to assure adequate core cooling; by continuous injection. (Appendix 17B)

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

SRO EOI-2 (Primary Containment Pressure)

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

ATC Trips Reactor Recirculation Pumps.

BOP Places all Drywell Blowers in Off.

SRO Initiate DW Sprays; using only pumps not required to assure adequate core cooling; by continuous injection. (Appendix 17B)

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

3-B Page 30 of 57 Simulator Event Guide:

Event 7 Major: Main Steam Line Leak inside Containment ATC/BOP Initiate DW Sprays, lAW Appendix 17B.

IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:

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

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

2. VERIFY Recirc Pumps and Drywell Blowers shutdown.

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

6. INITIATE Drywell Sprays using RHR Loop 1(11) as follows:

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

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

c. IF EIThER of the following exists:

LPCI Initiation signal is NOT present, OR

Directed by SRO, THEN PLACE keylock switch 1-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

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

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

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

g. OPEN the following valves:

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

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

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

3-B Page3l of57 Simulator Event Guide:

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

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

Emergency RPV Depressurization is required.

CS#1 SRO Enters EOI-C2.

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

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

CS#1 ATC/BOP Open 6 ADS Valves SRO Can 6 ADS Valves be opened YES -

SRO EOI-1 Level SRO Restore and Maintain RPV Water Level between +2 to 51 inches with one or more of the following injection sources. Condensate Appendix 6A, Core Spray Appendix 6D or 6E, LPCI Appendix 6C ATC/BOP Restore and maintain level +2 to +51 inches JAW Appendix 6A, 6D, 6E, or 6C SRO Emergency Plan Classification 2.1 -A

3-B Page 32 of 57 Simulator Event Guide:

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

1. VERIFY CLOSED the following feedwater heater return valves:

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

l-FCV-3-.72, HP HTR 1B1 LONG CYCLE TO CNDR

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

2. VERIFY CLOSED the following RFP discharge valves:

l-FCV-3-19, RFP 1A DISCHARGE VALVE

1-FCV-3-12, RFP lB DISCHARGE VALVE

1-FCV-3-5, RFP 1C DISCHARGE VALVE.

3. VERIFY OPEN the following drain cooler inlet valves:

l-FCV-2-72, DRAIN COOLER lA5 CNDS INLET ISOL VLV

l-FCV-2-84, DRAIN COOLER lBS CNDS INLET ISOL VLV

1-FCV-2-96, DRAIN COOLER 1C5 CNDS INLET ISOL VLV

4. VERIFY OPEN the following heater outlet valves:

l-FCV-2-l24, LP HEATER 1A3 CNDS OUTL ISOL VLV

l-FCV-2-125, LP HEATER 1B3 CNDS OUTL ISOL VLV

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

5. VERIFY OPEN the following heater isolation valves:

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

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

l-FCV-3-24, HP HTR lC2 FW INLET ISOL VALVE

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

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

l-FCV-3-77, HP HTR lCl FW OUTLET ISOL VALVE

6. VERIFY OPEN the following REP suction valves:

1-FCV-2-83, REP 1A SUCTION VALVE

l-FCV-2-95, REP lB SUCTION VALVE

l-FCV-2-108, REP 1C SUCTION VALVE.

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

9. ADJUST l-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 1 5).

10. VERIFY RFW flow to RPV.

3-B Page 33 of 57 Simulator Event Guide:

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

1. VERIFY OPEN the following valves:

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

1-FCV-75-1 1, CORE SPRAY PUMP 1C SUPPR POOL SUCT VLV

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

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

3. VERIFY CS Pump 1A andlor 1C running.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 1-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.

Core Spray System II Appendix 6E

1. VERIFY OPEN the following valves:

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

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

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

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

3. VERIFY CS Pump lB andJor 1D running.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 1-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.

3-B Page 34 of 57 Simulator Event Guide:

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

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

2. VERIFY OPEN the following valves:

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

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

3. VERIFY CLOSED the following valves:

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

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

1-FCV-74-71, RHR SYS II SUPPR CHBRJPOOL ISOL VLV

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

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

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

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

6. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 1-FCV-68-3,

______ RECIRC PUMP 1A DISCHARGE VALVE.

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

8. MONITOR RHR Pump NPSH using Attachment 1.

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

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

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

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

3-B Page 35 of 57 Critical Tasks Two CS#1-When Drywell Pressure cannot be maintained below the PSP limit, US determines that Emergency Depressurization is required and RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure

3. Measured by:

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

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

4. Feedback:

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

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Temperature

3. Measured by:

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

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

4. Feedback:

RPV pressure decreasing SRV open status indications

3-B Page 36 of 57 Critical Tasks Two CS#2- With a reactor scram required and the reactor not shutdown, take action to reduce power by initiating ART to cause control rod insertion.

1. Safety Significance:

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

Correct reactivity control.

2. Cues:

Reactor power indication.

Procedural compliance.

3. Measured by:

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

4. Feedback:

Reactor power trend.

Rod status indication.

3-B Page 37 of 57 Scenario Tasks EVENT TASK NuMBER KIA RQ SEQ 1 Stroke Time Containment Isolation Valves RO U-064-SU-08 223002A2.08 2.7 3.1 SRO S-000-AD-8 1 2 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 3 VFD Cooling Water Pump Failure RO U-068-AL-33 202001A2.22 3.1 3.2 SRO S-068-AB-O1 4 Steam Packing Exhauster Trip RO U-47C-AL-02 271000A1.O1 3.3 3.2 SRO S-047-AB-03 5 RHR Loop 1 Leak RO U-77A-AL-06 203000A4.02 4.1 4.1 SRO S-000-EM-09 6 Loss of RBCCW RO U-070-AL-03 206000A2.17 3.9 4.3 SRO S-070-AB-O1 7 Drywell LOCA RO U-000-EM-05 295028EA2.O1 4.0 4.1 SRO S-000-EM-04 SRO S-000-EM-05 SRO T-000-EM-15

3-B Page 38 of57 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-B 8 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 60 Run Time (minutes) 2 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:

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

3-PI-3-207 Bypassed for surveillance.

Operations/Maintenance for the Shift:

Perform Stroke Time Test on 3-FCV-43-13 and 3-FCV-43-14 per 3-SR-3.6.1.3.5 Section 7.3.

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

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

Unusual Conditions/Problem Areas:

None

Fr 0

N I

ftl lH M mm

IT rn

16 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PC.**iVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation

.fl Instrumentation ACTIONS

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

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

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

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1 Primary Containment when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME k

NOTE--- Al isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated AND

--- automatic valve, closed manual valve, blind 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured.

MSIV leakage not within limits.

AND (continued)

ACTIONS CONDITION REQUIFED ACTION COMPLETION TIME A. (continued) A2 -------------NOTE----

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

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

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETiON TIME B. --NOTE--------- B. I Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated

automatic valve, closed manual valve, or blind One or more penetration flange.

flow paths with two PCIVs inoperable except due to MSIV leakage not within limits.

C. ------------NOTE---------- C.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and de-activated

automatic valve, closed AND manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for flow paths with one PCIV EPCVs inoperable. AND C.2 ----------NOTE-----------

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

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

(continued)

ACTIONS (continued)

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

leakage not within limits.

E. Required Action and El Be in MODE 3.. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, B, C, AND orDnotmetin MODE I

2. or 3 E.2 Be n MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Required Action and F.1 Initiate action to suspend Immediately associated Completion operations with a Time of Condition A, B C, potential for draining the or D not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during PR MODE 4 or 5.

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

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

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS Operating LCO 3.5.1 Each ECCS injectionlspray subsystem and the Automatic Depressurization System (ADS) function of six safetylrellef valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant iniection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig, ACTIONS LCO 3,0.4,b is not applicable to IWO 1.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A. I Restore low pressure 7 days lnjectionlsp ray subsystem ECCS lnjectionlspray Inoperable. subsystem(s) to OPERABLE status.

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

(continued)

ECCS Operating 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B.2 Be tn MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

ACTIONS (continued)

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

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

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

AND OR Condition A entered D.2 Restore [ow pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

E. One ADS valve El Restore ADS valve to 1.4 days inoperable. OPERABLE status.

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

AND Condition A entered. P.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injection/spray subsystem to OPERABLE status.

(continued)

ACTIONS (continued)

CONDITION REQUiRED ACTION COMPLETION TIME

.G Two or more ADS valves G. 1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND OR G2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion 150 PS.

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

H. Two or more low pressure Hi Enter LCO 303. Immediately ECCS injectionfspray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperabIe

3.6 CONTAINMENT SYSTEMS 3,6.2.3 Residual Heat Removal (RHR> Suppression Pool Cooling LCO 3,6,2,3 Four RKR suppression pool cooling subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression Al Restore the RHR 30 days pool cooling subsystem suppression pool cooling inoperable. subsystem to OPERABLE status, B. Two RI-fR suppression 8.1 Restore one RHR 7 days pool cooling subsystems suppression pool cooling inoperable. subsystem to OPERABLE status.

C. Three or more RHR C.1 Restore required RHR B hours suppression pool cooling suppression pool cooling subsystems inoperable, subsystems to OPERABLE status.

(continued)

ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and Dl Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met, AND D.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

RHR Suppression Pool Spray 362.4 36 CONTAINMENT SYSTEMS 362.4 Residual Heat Removal (RHR) Suppression Pool Spray LCO 3,62.4 Four RIIR suppression pool spray subsystems shall be OPERABLE APPLICABILITY: MODES 1,2, and 3.

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

B. Two RHR suppression 81 Restore one RHR 7 days pool spray subsystems suppression pool spray inoperable. subsystem to OPERABLE status.

C. Three or more RHR C.1 Restore required RI-fR B hours suppression pool spray suppression pool spray subsystems inoperable, subsystems to OPERABLE status, D. Required Action and Dl Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not mAt AND

RHR [JrywelI Spray 3.62.5 3.6 CONTAfNMENT SYSTEMS 3.6.2.5 Residual Heat Removal (RHR) Drywell Spray LCO 362.5 Four RHR drywell spray subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

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

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

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

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

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

3.6 CONTAINMENT SYSTEMS 3.6.2.6 Drywell4o-Suppression Chamber Differential Pressure LCO 3.6.2.6 The drywell pressure shall be maintained 1.1 psid above the pressure of the suppression chamber.

This differential may be decreased to < 1.1 psid for a maximum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during required operability testing of the HPCI system, the RCIC system or the suppression chamber4odrywelI vacuum breakers.

APPUCABILITY: MODE 1 durIng the time period:

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

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell-to-suppression Al Restore differential 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months chamber differential pressure to within limit.

pressure not within limit.

8, Required Action and 8.1 Reduce THERMAL 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion POWER to 15% RTP.

Time not met.

I I I I I I I a C

z 2.1.AI I ITABLEI Drywell pressure at or above 2.45 psig AND I I I I I

Indication of Primary System leakage Into m Primary Containment. Refer to Table 21-A.

OPERATING CONDITION:

Model or 2 or 3 2iS ICURVEI I I 2.2-S I I I I Suppression Chamber pressure can NOT be Dfeli or Suppression Chamber maintained in the safe area of Curve 21-S. hydrogen concentration at or above 4%

in AND DrweII or Suppression Chamber oxygen concentration at or above 5%.

OPERATING CONDITION: OPERATING CONDITION:

Mode 1 or 2 or 3 Mode I or 2 Or 3 -< -

2.1GI 22-GI I I I Suppression Chamber pressure can NOT be DyeIl or Suppression Chamber maintained below 55 pslg, hydrogen concentration at or above 8%

AND Dryweli or Suppression Chamber oxygen concentration at or above 5%. in in in OPERATING CONDITION: OPERATING CONDITION:

Mode 1 or 2 or 3 Mode I OT 2013

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

Examiners: Operators: ATC:

BOP:

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

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

Event Event No. Maif. No. Type* Event Description N-BOP 1 8.13 Automatic Start Test of RFPT 3A Oil Pumps, 3-01-3 N-SRO R-ATC 2 Perform Control Rod Pattern adjust JAW RCP R-SRO 3 C-ATC Final(4 Control Rod manipulated continues to move 3 notches

)

th rd04r3823 TS-SRO beyond intended position C-BOP 4 rcO9 RCIC Room high temp / Fail to Isolate TS-SRO 5 fwO5b Loss of FW Heating C-ALL 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate Feedwater Line Break in Turbine Bldg / Drywell leak 6 fwl8 M-ALL Div 1 ECCS fails to initiate 7 csO4a I Core Spray Logic Power Failure 8 edl2b C 480V RMOV Board 3B Supply Breaker Trip

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

3-C Page 1 of 56 Console Operator Instructions Scenario File Summary File: batch and trigger files for scenario 3-C Batch nrc2Ollc

hpci tagout bat nrc2Ol lhpcito Batch nrc2Ol lhpcito ior zdihs732 close ior zdihs733a close ior zdihs738la close ior zlohs7347a[1] off ior ypovfcv732 (none 30) fail_now ior ypovfcv733 (none 30) fail_now ior ypovfcv73 81 (none 30) fail_now

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

CR Drift imfrdO4r3823 (el 0)

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

Loss of Feedwater Heating imffw05b (elO 0)100 300 75 ior ypovfcv052l fail_power_now ior zlohs052la{2] on trg 11 nrc20110521 trg 11 =batnrc20llcl Trigger nrc2Ol 10521 zdihs052l a[1].eq. 1 Batch nrc2Ollcl dor ypovfcv052l dor zlohs052la[2]

3-C Page 2 of 56

Major imffwl8 (e20 0) 50300 imfth2l (e25 30) .1 360 imf csO4a imfedl2b (e20 300) ior xa553c[27] alarm off ior xa553c[14] alarm off ior zloil756la[1] off ior zloil756lb[1] off trg2l nrc20117525 trg2l =batnrc20llc2 Trigger nrc2Ol 17525 zdihs7525a[3] .eq. 1 Batch nrc2Ollc2 dmf cs04a Console Operator Instructions Scenario 3-C DESCRIPTION/ACTION Simulator Setup manual Reset to 1C193 Simulator Setup Load Batch Bat nrc2Ol ic Simulator Setup manual Clearance out HPCI Simulator Setup Clearance out Stator Water Cooling manual Pump_3B Simulator Setup Verify batch file loaded RCP required (86% Power with Control Rod Pattern Adjust) Provide marked up copy of 3-GOI-100-12 and RCP for Urgent Load Reduction.

3-C Page 3 of 56 Scenario Summary:

The Plant is operating at 86% Reactor Power.

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

The SRO may declare the Control Rod Inoperable, Technical Specification 3.1.3 condition C.

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

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

A tube leak on High Pressure Feedwater Heater B2 results in isolation of Extraction Steam to the heater.

The crew will respond in accordance with 3-A0I-6-1A or 1C. The ATC will lower reactor power by 5%. The Operators refer to 3-AOI-6-1A or 1C and determine that all automatic actions failed to occur and the Operators isolate the Heater B2.

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

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

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

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

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

Level will be restored with Low Pressure ECCS.

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

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

3-C Page 4 of 56 Simulator Event Guide:

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

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

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

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

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

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

[4] TEST EBOP 3A3 as follows:

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

[4.2] VERIFY the following:

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

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

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

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

[4.4.1] VERIFY the following:

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

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

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

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

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

3-C Page 5 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP SRO Notify ODS of Power Increase Direct Power Increase after Control Rod Pattern Adjustment per 3-GOI-l00-12 section 5.0 step 21 5.0 INSTRUCTION STEPS

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

RAISE power using control rods or core flow changes.

REFER TO 3-SR-3.3.5(A) and 3-01-68.

MONITOR Core thermal limits using Illustration 1, ICS, and/or 0-11-248 ATC Raise Power with Control Rods per 3-01-85, section 6.6 ATC 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

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

CRD POWER, 3-HS-85-46 in ON.

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

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

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

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

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

STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].

3-C Page 6 of 56 Simulator Event Guide:

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

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

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

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

6.6.3 Control Rod Notch Withdrawal

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

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

CRD ROD SELECT pushbutton is brightly ILLUMINATED

White light on the Full Core Display ILLUMINATED

Rod Out Pemiit light ILLUMINATED

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

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

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

[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:

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

3-C Page 7of56 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP ATC 6.6.3 Control Rod Notch Withdrawal (continued)

[6.2] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.

[6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes.

[6.4] IF Control Rod Coupling Integrity Check fails, THEN REFER TO 3-AOI-85-2.

6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 3-XS-85-40.

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

CRD ROD SELECT pushbutton is brightly ILLUMINATED

White light on the Full Core Display ILLUMINATED

Rod Out Permit light ILLUMiNATED

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[5.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

3-C Page 8 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-lOO-12 and in accordance with the RCP ATC 6.6.4 Continuous Rod Withdrawal (continued)

[5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47 and CRD CONTROL SWITCH, 3-HS-85-48.

[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.

[5.5] WHEN control rod settles into the intended notch, THEN CHECK the following.

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does NOT alarm.

[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

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

[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position with the control rod at position 48.

3-C Page 9 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP ATC 6.6.4 Continuous Rod Withdrawal (continued)

[6.4] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

[6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2

[7] IF continuously withdrawing the control rod to position 48 and the control rod coupling integrity check will be performed after the CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[7.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

[7.3] WHEN position 48 is reached, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

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

3-C Page 10 of 56 Simulator Event Guide:

Event 2 Reactivity: Raise Reactor Power after Completion of Control Rod Pattern Adjustment per 3-GOI-100-12 and in accordance with the RCP ATC 6.6.4 Continuous Rod Withdrawal (continued)

[7.6] CHECK control rod coupled by observing the following:

Four rod display digital readout AND full core display digital readout AND background light will remain illuminated.

CONTROL ROD OVERTRAVEL annunciator (3-XA-55-5A, Window 14) does NOT alarm.

[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[7.8] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2.

6.6.5 Return to Normal After Completion of Control Rod Withdrawal

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

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

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

liiiver itriver When ATC withdraws the Final (4) rod (38-23) Insert trigger 1, Rod will continue to move3 Notches beyond mtendedposition When Control Rod 38.23 reaches position 14 de1etemwctionidO4r3823 liEom the mdfntion menu.

3-C Page 11 of 56 Simulator Event Guide:

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

)

intended position iiriver When ATC withdraws the Final (4)r6d (38-23)Rodwiil continue toiiióve3Notches beyond intended position. When ControLRod 38-23 reaches position 14 delete malfunction rd04r3823 from the nia1fimction*menu ATC Reports CONTROL ROD DRIFT alarm and Control Rod 38-23 has drifted out 3 notches from intended position SRO Directs ATC to respond per ARP and 3-AOI-85-6 and/or 3-AOI-85-7 ATC 3-ARP-9-5A window 28 CONTROL ROD DRIFT A. DETERMINE which rod is drifting from Full Core Display.

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

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

2. RESET the annunciator.

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

E. REFER TO Tech Spec Section 3.1.3, 3.10.8.

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

Then resets the annunciator Responds per 3-AOI-85-6 and/or 3-AOI-85-7 Monitors Full Core Display for a second Control Rod Drift as per Immediate Actions of 3-AOI-85-6 ATC 3-AOI-85-6 Control Rod Drift 4.1 Immediate Actions

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

3-C Page 12 of 56 Simulator Event Guide:

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

[1] IF a Control Rod is moving from its intended position without operator actions, THEN SELECT the drifting control rod and INSERT to the FULL IN (00) position.

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

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

[4] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 3-AOI-l00-1.

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

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

[7] EVALUATE Tech Spec 3.1.3.

[8] INITIATE Service RequestlWork Order.

3-C Page 13of56 Simulator Event Guide:

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

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

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

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

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

[10] NOTIFY System Engineering to PERFORM 0-TI-20, Control Rod Drive System Testing and Troubleshooting to determine problem with faulty control rod.

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

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

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

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

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

3-C Page 14 of 56 Simulator Event Guide:

Event 3 Component: Final(4tI) Control Rod manipulated continues to move 3 notches beyond intended position j5jj AsReactorEngineer infowi that Core Thermal Limits the current Control Rod pattern will be evaluated.

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

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

. Evaluation of impact on thermal limits and PCIOMOR restraints.

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

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

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

Directs associated HCU removed from service per 3-01-85 DiE Drrifei IT contacted, as Reactor Engmee inform that all conditions listed above wilIbe ealuated If contacted, as Work Control inform that you will get workmg om a Work Order/SR.

If contacted, as System EngIneering inform that you will perforth0TI-2O.

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

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

[1] STOP all intentional control rod movement.

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

. Reactor Engineer (RE),

. Shift Technical Advisor (STA),

Unit Supervisor

. Shift Manager (SM)

. Operations Superintendent. [INPO SOER 84-002]

3-C Page 15 of 56 Simulator Event Guide:

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

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

[3.1] INSERT the mispositioned rod to 00.

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

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

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

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

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

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

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

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

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

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

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

EVALUATE the possible consequences

DOCUMENT in Reactor Engineer log.

3-C Page 16 of 56 Simulator Event Guide:

Event 3 Component: Final(4t) Control Rod manipulated continues to move 3 notches beyond intended position SRO Directs ATC to stop all intentional Control Rod Movement Informs all positions listed in step 2 of Subsequent Actions of Mispositioned Control Rod Directs ATC to Insert Mispositioned Control Rod to 00 Enters 3-GOl- 100-12, Power Maneuvering Initiates Service Request and Notifies Reactor Engineer to evaluate the possible consequences and document in the Reactor Engineering Log Drivei Diiver The SRO will direct the associated IICU removed from service if 3-A0I-85-6 is entered.:

Acknowledge orderto removel CUfromserviee Verify what stepiin 3-01-85 will be us to isolate the RCU4ajt 20 nes, to bgiççun11at9

iqpessnre ala,rpid r cremçye4 roai servçc bri ;yJ&ii: :ygers r jT1, e Ui qp i1t

3-C Page 17 of 56 Simulator Event Guide:

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

Driver Drrrer Acknowledge allposations informed rn step 2 of Subsequent Actions

. f, cóntcted,as Work Coi rolinform that you will get workhig onaWorkOrder)Servic ReQuest Ifçoxtaçjactoinfonnflqten<fltions lj$ç yer Divei Whe ect tFiigger for ibJ steam lea hado auto

3-C Page 18 of 56 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate Driver Driver When directed by NRC insert Trigger S for RCIC steam leak with failure to auto isolate BOP Respond to Annunciator RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 3-9-11. (Alarm on Panel 3-9-11 will automatically reset if radiation level lowers below setpoint.)

C. NOTIFY RADCON.

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

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

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

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

. RCIC STEAM LINE OUTBD ISOLATION VLV, 3-FCV-71-3 A. CHECK RCIC temperature switches on LEAK DETECTION SYSTEM TEMPERATURE indicator, 3-TI-69-29 on Panel 3-9-2 1.

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

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

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

E. DISPATCH personnel to investigate.

3-C Page 19of56 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate BOP Reports rising temperature in RCIC, reports RCIC failed to isolate and isolates RCIC Steam Line SRO Enter EOI-3 on Secondary Containment Area Radiation 31 && &tijdii SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. Then verify isolation of Reactor Zone or Refuel Zone and verify SGTS initiates If above 72 mr/hr direct Operator to verify isolation of ventilation system and SGTS initiated ATC/BOP Verifies Reactor Zone and Refuel Zone Ventilation Systems isolated and SGTS initiated SRO If Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels are below 72 mr/hr Then Restart Reactor Zone and Refuel Zone Ventilation per Appendix 8F If ventilation isolated and below 72 mr/hr directs Operator to perform Appendix 8F CS#3 SRO Enters EOI-3 on High Secondary Containment Temperature Secondary Containment Temperature Monitor and Control Secondary Containment Temperature Operate available ventilation per Appendix 8F Is Any Area Temp Above Max Normal YES -

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

Be operated by EOIs OR

Suppress a Fire CS#3 BOP Isolates RCIC Steam Lines and reports Temperatures and Radiation Levels lowering

3-C Page 20 of 56 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate SRO Enters EOI-3 on High Secondary Containment Temperature (continued)

Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels Is Any Area Radiation Level Max Normal NO -

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

Be operated by EOIs OR

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

Enters EOI-3 on High Secondary Containment Temperature (continued)

Secondary Containment Level Monitor and Control Secondary Containment Water Levels Is Any Floor Drain Sump Above 66 inches NO -

AND Is_Any Area Water Level Above 2_inches_- NO SRO Evaluates Technical Specification 3.5.3 Condition B and 3.6.1.3 Condition B. The SRO will determine RCIC Isolation Valves inoperable and RCIC System inoperable. With HPCI already Inoperable, plant shutdown is required to Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and less than 150 psig Steam Dome Pressure within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

SRO Notifies Operations management that Plant Shutdown is required ATC/BOP 3-EOI Appendix 8F

1. VERIFY PCIS Reset.

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

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

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

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

3-C Page 21 of 56 Simulator Event Guide:

Event 4 Component: RCIC Room high temp / Fail to Isolate ATC/BOP d. VERIFY OPEN the following dampers:

3-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR

3-FCO-64-.6, REFUEL ZONE SPLY INBD ISOL DMPR

3-FCO-64-9, REFUEL ZONE EXH OUTBD ISOL DMPR

3-FCO-64-1O, REFUEL ZONE EXH 11JBD 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 AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.

d. VERIFY OPEN the following dampers:

3-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR

3-FCO.-64-14, REACTOR ZONE SPLY INBD ISOL DMPR

3-FCO-64-42, REACTOR ZONE EXH INBD ISOL DMPR 3-FCO-64-43, REACTOR ZONE EXH OUTBD ISOL DMPR.

h D4 5 ka ipj

3-C Page 22 of 56 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate Driven Driver When directed byNRC insert Trigger 10 foiLos ofFeedwater Heating and 3-PCV-5-2 1, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate::

ATC/BOP Announces BYPASS VALVE TO CONDENSER NOT CLOSED and refers to 3-ARP-9-6A, window 18.

A. CHECK heater high or low level or moisture separator high or low level alarm window illuminated on Panel 3-9-6 or 3-9-7 to identify which bypass valve is opening.

B. CHECK ICS to determine which bypass valve is open.

C. DISPATCH personnel to check which valves light is extinguished on junction box 34-2 1, Col T-13 J-L1NE, elevation 565.

Dnver Iiivei Acknowledge dispatch, wait 1-2 mmutes and report 3-LCV-6-22B light is out on junction box34-2L ATC/BOP Announces HEATER B2 LEVEL HIGH and refers to 3-ARP-9-6A window 9.

A. CHECK the following indications:

Condensate flow recorder 2-29, Panel 3-9-6. Rising flow is a possible indication of a tube leak.

Heater B2 shell pressure, 3-PI-5-22 and drain cooler B5 flow, 3-FI-6-34, Panel 3-9-6. High or rising shell pressure or drain cooler flow is possible indication of a tube leak.

B. CHECK drain valve 3-FCV-6-95 open.

C. CHECK level on ICS screen, FEEDWATER HEATER LEVEL (FWHL).

IF the 3B2 heater indicates HIGH (Yellow), THEN VERIFY proper operation of the Drain and Dump Valves.

DISPATCH personnel to local Panel 3-LPNL-925-562C to VERIFY and MANUALLY control the level.

D. IF a valid HIGH HIGH level is received, THEN GO TO 3-AOI-6-1A or 3-AOI-6-1C.

ATC/BOP Checks condensate flow recorder, Heater B2 shell pressure and Drain Cooler B5 flow for indications of a tube leak Checks drain valve 3-FCV-6-95 open Checks 3B2 Heater level on ICS and dispatches personnel to verify and manually control level

3-C Page 23 of 56 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate ATC/BOP Announces B 1 and B2 High Pressure Heater Extraction Isolation SRO Directs crew to enter 3-AOI-6-1A or 3-AOI-6-1C ATC/BOP 3-AOI-6-1A High Pressure Feedwater Heater String/Extraction Steam Isolation 4.1 Immediate Actions

[1] REDUCE Core Thermal Power to? 5% below initial power level to maintain thermal margin.

4.2 Subsequent Actions

[1] REFER TO 3-01-6 for turbine/heater load restrictions.

[2] REQUEST Reactor Engineer EVALUATE and ADJUST thermal limits, as required.

[3] ADJUST reactor power and flow as directed by Reactor Engineer/Unit Supervisor to stay within required thermal and feedwater temperature limits. REFER TO 3-GOI-100-12 or 3-GOI-100-12A for the power reduction.

[4] ISOLATE heater drain flow from the feedwater heater string that isolated by closing the appropriate FEEDWATER HEATER A-2(B-2) or (C-2) DRAIN TO HTR A-3(B-3) or (C- 3), 3-FCV-6-94(95) or (96).

[5] IF a tube leak is indicated, THEN PERFORM manual actions of Attachment 1 for affected heaters.

[6] VERIFY automatic actions occur. REFER TO Attachment 1.

[7] MONITOR TURB THRUST BEARING TEMPERATURE, 3-TR-47-23, for rises in metal temperature and possible active/passive plate reversal.

[8] DETERMTNE cause which required heater isolation and PERFORM necessary corrective action.

3-C Page 24 of 56 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate ATC/BOP 3-AOI-6-1A High Pressure Feedwater Heater String/Extraction Steam Isolation (continued) 4.2 Subsequent Actions (continued)

[9] WHEN the condition which required heater isolation is no longer required, THEN RESTORE affected heater. REFER TO 3-01-6.

ATC Lower Reactor Power greater than 5% below initial power level using Recirc Pump flow adjustments BOP Refers to 3-01-6 for turbine/heater load restrictions Contacts Reactor Engineer to evaluate and adjust Thermal Limits, if needed Isolates heater drain flow B2 Heater Drain to B3 Heater by shutting 3-FCV-6-95 SRO Directs isolating FW to B HP heater string based on indications of tube leak by performing manual actions of Attachment 1 and verifying automatic actions occur 3-AOI-6-1A Attachment 1 B 1 or B2 The following valves must be manually closed:

3-FCV-3-3l, HP HTR 3B2 FW iNLET ISOL VALVE 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VALVE The following valves AUTO Isolate 3-FCV-5-9, HP HEATER 3B1 EXTR ISOL VLV 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV 3-FCV-6-74, MOISTURE SEP LC RES Bl ISOL VLV 3-FCV-6-l72, MOISTURE SEP LC RES B2 ISOL VLV Directs power reduction to < 79% power (Mid-power runback) per 3-01-6, Illustration 1 3-01-6 Illustration 1 HEATERS OUT (Tube and Shell Side) **

One HP string 920 MWe (79%)

One LP string 920 MWe (79%)

One HP and LP string 920 MWe (79%)

Enters 3-G0I-l00-12, Power Maneuvering Notifies Rx Eng. And ODS of Feedwater Heater isolation and power reduction

3-C Page 25 of 56 Simulator Event Guide:

Event 5 Component: Loss of Feedwater Heating and 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV Fail to isolate BOP Closes the following Feedwater Valves Manually 3-FCV-3-31, HP HTR 3B2 FW iNLET ISOL VALVE 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VALVE Verifies the following valves close automatically 3-FCV-5-9, HP HEATER 3B1 EXTR ISOL VLV 3-FCV-5-21, HP HEATER 3B2 EXTR ISOL VLV 3-FCV-6-74, MOISTURE SEP LC RES Bi ISOL VLV 3-FCV-6-172, MOISTURE SEP LC RES B2 ISOL VLV Takes action to manually shut 3-FCV-5-21 upon determining the valve did not automatically close and reports to SRO Recognizes HTR level lowers as a result of isolating the Condensate side of 3B HP HTR string (i.e. tube leak) and reports to crew eii ner jose4,vqpfyTnggerU g act1v w cqn&I to ei$ adjst thennal iim1ts c1moie4g ordejand1ien creyjs gpjipoqç ATC Lower Reactor Power to <920 MWe/<79% power by lowering recirc flow.

SRO Direct ATC to insert the first group of control rods on the Emergency Shove Sheet per Reactor Engineer recommendation.

ATC Inserts the first group of rods on the Emergency Shove Sheet using a peer check as directed by Rx Engineer & Unit Supervisor pe n edwatc n+/-Ni ih jea9totØI ee4iOQse gg9r2

&eaI

3-C Page 26 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate 1

ATC Responds to alarms RECTOR FEED PUMPS A, B, AND C ABNORMAL, RFWCS ABNORMAL and REACTOR WATER LEVEL ABNORMAL ATC 3-ARP-9-5A Reactor Water Level Abnormal A. VERIFY Reactor water level hi/low using multiple indications including Average Narrow Range Level on 3-XR-3-53 recorder, 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-LI-3-253 on Panel 3-9-5.

B. IF alarm is valid, THEN REFER TO 3-AOI-3-1 or 3-01-3.

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

ATC Monitors Reactor Water Level and Reports trend, recommends Manual Reactor Scram Determines Feedwater Leak in the Turbine Building due to both Feedwater Line Flows lowering to 0 and Reactor Feed Pump Flows Increasing with a Lowering Reactor Water Level SRO Directs a Manual Reactor Scram inserted Directs Reactor Feed Pumps to be tripped, Reactor Feed Pump Discharge Valves shut, and Condensate Booster Pumps then Condensate Pumps secured (Isolate and stop leak)

ATC Inserts Manual Reactor Scram Trips Reactor Feed Pumps and shuts Reactor Feed Pump Discharge Valves Secures Condensate Booster Pumps then Condensate Pumps

3-C Page 27 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate henqctoWaterLecreLreaches 410 fo ;12Oinches ineTngger 25 DzyweIlJ

çaI4 SRO Enters EOI-1 on Low Reactor Water Level RC/Q Monitor and Control Reactor Power Directs Exit of EOI-1 RC/Q Leg after ATC reports All Rods In on Scram Report RCIP Monitor and Control RPV Pressure Answers No to is any MSRV cycling Directs BOP to maintain RPV Pressure 800-1000 psig using Bypass Valves RCIL Monitor and Control RPV Water Level Verify as Required

PCIS Isolations (Groups 1, 2 and 3)

ECCS

RCIC Recognizes loss of all High Pressure Injection sources with exception of CRD and SLC. Directs maximizing CRD flow to the Vessel per Appendix 5B Answers No to can water level be Restored and Maintained above +2 inches Maintain RPV Water Level above -162 inches

3-C Page 28 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate Drr Diver etpr Water ieál CS#4 Enters EOI-1 on Low Reactor Water Level (cont)

Directs ADS inhibited when RPV Water Level drops below -120 inches Augments RPV Water Level Control with SLC per Appendix 7B Answers No to can RPV Water Level be maintained above -162 inches Exits RC/L and enters C- 1, Alternate Level Control ATC Appendix 5B

1. IF Maximum injection flow is NOT required, THEN VERIFY CRD aligned as follows:

a. VERIFY at least one CRD pump in service and aligned to Unit 3 CRD system.

b. ADJUST 3-FIC-85-11, CRD SYSTEM FLOW CONTROL, as necessary to obtain flow rate of 65 to 85 gpm.

c. THROTTLE 3-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV, to maintain 250 to 350 psid drive water header pressure differential.

d. EXIT this procedure.

2. IF BOTH of the following exist:

CRD is NOT required for rod insertion, AND Maximum injection flow is required, THEN LINE UP ALL available CRD pumps to the RPV as follows:

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

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

3-C Page 29 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate jjç When,RaetoiWater Level eaebps-11Oto -jO mnçhesinsertTrigger 25 Drywd iea1 ATC Appendix 5B (cont)

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

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

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

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

d. ADJUST 3-FIC-85-11, CRD SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WHILE maintaining 3-PI-85-13A, CRD ACCUM CHG WTR HDR PRESS, above 1450 psig, if possible.

ATC Appendix 7B

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

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

11. CHECK SLC injection by observing the following:

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

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

. SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (3-XA-55-5B, Window 20).

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

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

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

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

13. IF SLC tank level drops to 0%,

THEN STOP SLC pumps.

15. MONITOR and CONTROL SLC System as necessary to maintain injection.

3-C Page 30 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak/

Div 1 ECCS fails to initiate eTiggr 25frfwell ryyçll es 5

& 11jeovi1I1d Rear teriii9wefttI BOP Approximately 5 minutes after Feedwater Leak inserted recognizes loss of 480v RMOV Board B. Announces loss of Division II ECCS systems N1O NC op flIECfIII stilhnject ie to outoanjecpoiivtveopeii vit1io powet) an lx board injection valve stilhiiavingpower Wi1l be>miable to throttle flow; when Loqp II LPCLisiio iorgerrequird, puxps niustbe securd, LoopIt Corç Spy otfinctioi, SRO Enters C-i, Alternate Level Control CS#4 Verifies ADS Inhibited Directs lineup of Injection Systems Irrespective of Pump NPSH and Vortex limits (LPCI and CS) per Appendix 6B and 6D Answers Yes to can 2 or more CNDS, LPCI or CS Injection Subsystems be aligned with pumps running CS#i When RPV Water Level drops to -162 inches, Then continues Answers Yes to is any CNDS, LPCI or CS Injection Subsystem aligned with at least one pump running Before RPV Water Level drops to -180 inches continue Answers Yes to are pumps running that can restore and maintain RPV Water Level above -180 inches after Emergency Depressurization Emergency RPV Depressurization is Required Enters C-2 Directs maximizing RPV Injection from all available sources irrespective of pump CS#2 NPSH and Vortex Limits Answers Yes to can RPV Water Level be restored and maintained above -180 inches Exits C-i and enters EOI-1, RPV Control at step RC/L-1

3-C Page 31 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg I Drywell leak!

Div 1 ECCS fails to initiate CS#4 BOP/ATC Inhibits ADS Lines up LPCI and CS Loop I pumps for Injection per Appendix 6B and 6D After Emergency Depressurization commenced, verifies RPV Injection is maximized from CS#2 all available sources irrespective of pump NP SF1 and Vortex Limits BOPIATC Appendix 6B, Loop I LPCI

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

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

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

4. VERIFY CLOSED the following valves:

3-FCV-74-61, RHR SYS I DW SPRAY 1NBD VLV

3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

3-FCV-74-57, RHR SYS I SUPPR CHBR/POOL ISOL VLV

3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

3-C Page 32 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate BOP/ATC Appendix 6B, Loop I LPCI (cont)

5. VERIFY RHR Pump 3A and/or 3C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RHR SYS I LPCI INBD INJECT VALVE.

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

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

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

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

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV

3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV BOP/ATC Appendix 6D, Loop I Core Spray

1. VERIFY OPEN the following valves:

. 3-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCT VLV

. 3-FCV-75-11, CORE SPRAY PUMP 3C SUPPR POOL SUCT VLV

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

3-C Page 33 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg! Drywell leak/

Div 1 ECCS fails to initiate BOP/ATC Appendix 6D, Loop I Core Spray (cont)

2. VERIFY CLOSED 3-FCV-75-22, CORE SPRAY SYS I TEST VALVE.

3 VERIFY CS Pump 3A and/or 3C RUNNING.

4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 3- FCV-75-25, CORE SPRAY SYS I INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.

SRO Enters C-2, Emergency RPV Depressurization Answers Yes to will the Reactor remain subcritical without Boron under all conditions Answers Yes to is Drywell Pressure above 2.4 psig Does not prevent Injection from any Core Spray or LPCI pumps because they are all needed to assure adequate core cooling Answers Yes to is Suppression Pool Level above 5.5 feet CS#1 Directs opening of all ADS Valves Answers Yes to can 6 ADS Valves be opened Maintains 6 ADS Valves open until RPV cold shutdown Interlocks are clear BOP/ATC Reports Suppression Pool Level in Feet when directed by SRO CS#l Opens 6 ADS valves and verifies open when directed CS#2 When RPV Pressure is low enough for Injection of LPCI and Core Spray, operator should verify available systems are injecting. At this time operator should notice Core Spray Loop I Injection Valve not open and take action to manually open the valve.

When adequate core cooling is assured begins to throttle flow to prevent overfihling CS#2 RPV. Must secure pumps on Loop II LPCI to stop injection.

1o wheJfl

3-C Page 34 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO Enters EOI-2 on High Drywell Pressure DWIT Monitor and control Drywell temperature below I 60F using available Drywell cooling Answers No to can Drywell Temperature be maintained below 1 60F Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor (this will already be complete at this time)

Before Drywell Temperature rises to 280F continue Answers Yes to is Suppression Pool Level below 18 Feet Answers Yes to are Drywell Temperatures and Pressures within the safe area of curve 5 Directs Shutdown of Recirc Pumps and Drywell Blowers (should leave Drywell Blowers running due to being unable to spray because adequate core cooling is not assured)

Does not initiate Drywell Sprays Because Adequate Core Cooling is not assured at this time

3-C Page 35 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO Enters EOI-2 on High Drywell Pressure (cont)

PC/P Monitor and control Primary Containment pressure below 2.4 psig Answers No to can Primary Containment Pressure be maintained below 2.4 psig Before Suppression Chamber Pressure rises to 12 psig Initiate Suppression Chamber Sprays using only those pumps not required for Adequate Core Cooling (Does not initiate Suppression Chamber Sprays because Adequate Core Cooling is not assured at this time)

PC/H Monitor and Control Drywell and Suppression Chamber Hydrogen at or below 2.4% and Oxygen at or below 3.3% using the Nitrogen Makeup System SP/T Monitor and Control Suppression Pool Temperature below 95F using available Suppression Pool Cooling Answers Yes to can Suppression Pool Temperature be maintained below 95F (Once Emergency Depressurization has commenced Suppression Pool Temperature will exceed 95F, this step should be re-addressed once Adequate Core Cooling is assured)

3-C Page 36 of 56 Simulator Event Guide:

Event 6 Major: Feedwater Line Break in Turbine Bldg / Drywell leak!

Div 1 ECCS fails to initiate SRO spit.

Monitor and Control Suppression Pool Level between -1 and -6 inches Answers Yes to can Suppression Pool Level be maintained above -6 inches Answers Yes to can Suppression Pool Level be maintained below -1 inches SRO Enters EOI-3 on High Secondary Containment Temperature Secondary Containment Temperature Monitor and Control Secondary Containment Temperature Operate available ventilation per Appendix 8F Answers Yes to Is Any Area Temp Above Max Normal Isolate all systems that are discharging into the area except systems required to:

. Be operated by EOIs OR

Suppress a Fire Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels Answers No to Is Any Area Radiation Level Max Normal Secondary Containment Level Monitor and Control Secondary Containment Water Levels Answers No to Is Any Floor Drain Sump Above 66 inches AND Answers No to Is Any Area Water Level Above 2 inches uooe1 tham1e SRO The Emergency Classification is 1.1-Si

3-C Page 37 of 56 Critical Tasks Three CS#1-With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, initiate Emergency Depressurization before RPV level lowers to -180 inches.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Water level trend.

3. Measured by:

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

4. Feedback:

RPV pressure trend.

SRV status indications.

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

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

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

3. Measured by:

Operator manually starts or 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.

3-C Page 38 of 56 Critical Tasks Three CS#3-With a primary system discharging into the secondary containment, take action to manually isolate the break.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Area temperature indication.

3. Measured by:

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

4. Feedback:

Valve position indication

3-C Page 39 of 56 Scenario Tasks EVENT TASK NUMBER K/A RO $1Q 1 Automatic Start Test of RFPT 3A Oil Pumps RO U-003-NO-30 259001K4.06 2.5 2.6 2 Control Rod Pattern Adjustment RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 3 Control Rod Mispositioned or Drift RO U-085-AB-07 295014AA1.03 3.5 3.5 SRO S-085-AB-07 4 RCIC Steam Leak RO U-071-AL-19 295032EA1.05 3.7 3.9 SRO S-000-EM-12 5 Loss of Feedwater Heating ROU-006-AB-01 2.1.43 4.1 4.3 SRO S-006-AB-01 6 Feedwater Line Break RO U-000-EM-18 29503 1EA2.04 4.6 4.8 SRO S-000-EM- 19 SRO T-000-EM-15

3-C Page 40 of 56 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-C 7 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 3 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 60 Run Time (minutes) 4 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

RHRSW is tagged out for Preventive Maintenance.

Stator Water Cooling Pump 3B is tagged out.

Operations/Maintenance for the Shift:

BOP Operator perform 3-01-3 Section 8.13 Automatic Start Test of RFPT 3A EBOP 3A3 Oil Pump Once completed perform Control Rod Pattern adjustment in accordance with the Reactivity Control Plan Units 1 and 2 are at 100% power.

Unusual Conditions/Problem Areas:

None

r C,

S

rr 1%,

I N

N m.

1 z

Li i:

rn

m 3I m

3 3

Cl) rn C,

0 C,

0 3

I m

I 3

16 CONTAINMENT SYSTEMS 16.1 3 Primary Containment Isolation Valves (PCIVs)

LCO 1613 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 13.6.1, Primary Containment Isolation Instrumentation ACTIONS NOTES t Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

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

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

4. Enter applicable Conditions and Required Actions of LCO 16.1.1w Primary Containment7 when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME Al isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs and dc-activated AND automatic valve, closed manual valve, bEind B hours for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to vaLve secured.

MSIV leakage not within lirnits AND (continued):

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 .-----------NOTE Isolation devices in high radiation areas may be verified by use of administrative means.

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

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B ----------NOTE-------- B I Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated

- automatic valve, closed nianual valve, or blind One or more penetration flange flow paths with two PCIVs inoperable except due to MSIV leakage not within limits.

C. NOTE------ Ci Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and de-activated

automatic valve, closed AND manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for flow paths with one PCIV EECVs inoperable. AND Cl Isolation devices in high radiation areas may be verified by use of administrative means Verify the affected Once per 31 days penetration flow path is isolated.

(continued)

ACTIONS (continued)

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

leakage not within limits.

E. Required Action and E. I Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, B, C, AND or D not met in MODE 1, 2, or 3 E2 Be n MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Required Action and P.1 Initiate action to suspend Immediately associated Completion operations with a Time of Condition A, B, C, potential for draining the or D not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during PR MODE 4 or5.

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

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

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.3 RCIC System LCO 3,5.3 The RCIC System shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure> 150 psig.

ACTIONS NOTE LCO 3.0.4.b is not applicable to RCIC.

CONDITION REQUIRED ACTION COMPLETION TiME A. RCIC System inoperable. A.1 VerifSi by administrative Immediately means High Pressure Coolant Injection System s OPERABLE.

AND A.2 Restore RC1C System to 14 days OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months dome pressure to 150 psig.

3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS NOTE Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control --NOTE rod stuck. Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, tControI Rod Block lnstrumentation, if required, to allow continued operation.

A..1 Verify stuck control rod Immediately separation criteria are met.

AND A.2 Disarm the associated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months control rod drive (CRD).

AND (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Perform SR 3.1.3.2 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from SR 3.1.3.3 for each discovery of withdrawn OPERABLE Condition A control rod, concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A.4 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Two or more withdrawn BA Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stuck.

C. One or more control rods C.1 -------------NOTE inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B. LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod, AND C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. NOTE Di Restore compliance with 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when BPWS.

THERMAL POWER

>1O%RTP. .QE D.2 Restore control rod to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Two or more inoperable OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and El Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, C, or D not met.

OR Nine or more control rods inoperable.

BROWNS FERRY I EMERGENCY CLASSIFICATION PROCEDURE j EVENT CLASSIFICATION MATRIX EPIP-1 Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity with irradiated fuel assemblies expected to Pool with irradiated fuel assemblies expected to z remain covered by water. remain covered by water.

C m

OPERATING CONDITION: OPERATING CONDITION Mode5 ALL 1.1-All INOTEI 1.l.A21 I I Uncontrolled water level decrease in Reactor Uncontrolled water level decrease in Spent Fuel Cavity expected to result in irradiated fuel Storage Pool expected to result in irradiated fuel assemblies being uncovered, assemblies being uncovered.

0 OPERATING CONDITION: OPERATING CONDITION:

Mode 5 ALL 1.1-SI I I NOTE I I 1l-S2 I Reactor water level can NOT be maintained Reactor water level can NOT be determined.

above -162 inches. (TAF) m m

m C) m OPERATING CONDITION: OPERATING CONDITION:

ALL Mode I or 2 or 3 -<

1.1-GI I I l.l-G2 I I NOTE I TABLE p US Reactor water level can NOT be restored and Reactor water level can NOT be determined maintained above -180 inches. AND Either of the following exists:

. The reactor will remain subcritical without boron under all conditions, and m

> Less than 4 MSRVs can be opened, or 2 Reactor pressure can NOT be restored and maintained above Suppression Chamber pressure by at least r

+ UNIT19opsi m

. UNIT280 psi

+ UNIT37opsi

. It has NOT been determined that the reactor will remain subcritical without boron under all z conditions and unable to restore and maintain C)

MARFP in Table 1.1-G2, OPERATING CONDITION: OPERATING CONDITION:

Mode lor2or3 Model or2or3

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

Initial Conditions: IC 193/ Unit 3 Reactor Power 4%! Condensate Pump 3A tagged Turnover: Aligning Charcoal Filters for Parallel Flow JAW 3-01-66 Section 5.1 1. Raise Power with Control Rods for Mode Change Event Event No. Maif. No. Type* Event Description 1 Aligning Charcoal Filters for Parallel Flow 5.11 2 Raise Power with Control Rods for Mode Change J

C-ATC 3 thO3b Reactor Recirc Pump B Trip TS..SRO TS-SRO 4 trg CS inadvertent initiation CBOP 5 msOl Steam Seal Regulator failure C-ATC 6 fw3Oc Feedwater Pump Governor drifts up C..SRO 7 pci 4 M-ALL Torus Leak / ATWS 3-FCV-74-57 fails to open (If repair team called for, open valve after 8 C ED started) 9 trg 20 C 3-FCV-73-30 Fails to Open

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

3-D Page 1 of 54 Console Operator Instructions Scenario File Summary File: batch and trigger files for scenario 3-D Batch nrc2OlldRl

cp pump 3a clearance ior ypobkrcndpa fail_power

Recirc Pump B trip imfth03b (el 0)

cs Initiation ior zdihs755a[4] (e5 0) start ior ypobkrcspmpa (e6 0) fail_power

steam seal failure imfms0l (elO 0) imfmc04 (elO 0)100

FWLC fail imffw30c (e15 0)100 3000 54 trg 7 nrc2011fptc trg 7 = dmffw30c Trigger nrc2Ol lfptc zdihs46 1 Oa[4] .ne. 1

SP LEAK ATWS/major bat atws75 imfpcl4 (e20 0)100 300 75 ior ypovfcv7330 (e20 0) fail_now trg2l =batatws-1 trg 22 = bat appOif trg 23 = bat appo2 ior zdihs7457a[2] auto bat nrcstick20 trg 24 = bat nrcunstickl4 trg25 =batsdv

3-D Page 2 of 54 Batch nrcstick2O imfrd06r30l 5 imf rdO6r3 023 imfrd06r303 1 imfrd06rl 851 imfrdO6rl439 imfrd06rl43 1 imfrdO6r34l5 imfrdO6r38l5 imf rd06r42 15 imfrdO6r463l imf rd06r543 9 imf rdO6r3 027 imfrd06r263 1 imfrd06r26 15 imf rd06r2239 imf rdO6r3 839 imfrd06rl4l 5 imfrd06r30l 5 imfrdO6r46lS imf rd06r2223 Batch nrcunstickl4 dmfrdO6r3435 dmf rdO6r3 423 dmfrd06r263 1 dmfrd06r343 1 dmf rd06r263 9 dmfrdO6r3439 dmf rdO6r3 027 dmfrdO6r3427 dmf rd06r2243 dmf rdO6r2 643 dmf rdO6r3 043 dmf rdO6r3 443 dmf rdO6rl 843 dmfrd06rl 819

3-D Page 3 of 54 Console Operator Instructions Scenario 3-D DESCRIPTION/ACTION Simulator Setup manual Reset to IC 193 Simulator Setup Load Batch Bat nrc2OlldRl Simulator Setup manual Clearance out Condensate pump 3A Simulator Setup Verify Batch file loaded RCP required (Raise Power from 4% to 8% with Control Rods for Mode Change) Provide marked up copy of 3-GOI-100-1A and RCP

3-D Page 4 of 54 Scenario Summary:

The Plant is operating at 4% Reactor Power.

The BOP Operator will Align Charcoal Filters for Parallel Flow lAW 3-01-66 section 5.11.

The ATC will withdraw control rods in order to raise power to 8% for a mode change from 2 to 1.

Once the NRC is satisfied with the reactivity manipulation, Reactor Recirculation Pump B will trip. The SRO will direct entry to 3-AOI-68-1A; the ATC will close RR Pump B discharge valve.

The SRO will evaluate Technical Specification 3.4.1, Condition A is required.

Core Spray Pump 3A inadvertently initiates, BOP Operator verifies initiation is inadvertent and with SRO concuence stops Core Spray Pump 3A JAW with ARPs. The SRO will evaluate Technical Specification 3.5.1, Condition A is required.

The Steam Seal regulator will fail; the BOP Operator will take action lAW with the ARPs and restore steam seal pressure with the bypass valve.

The operating feedwater pump controller will fail and level will slowly rise until the ATC or Crew notices the Reactor Level change. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to maintain Reactor Level control in manual. SRO should direct entry into 3-AOI-3-1.

An unisolable Torus leak will commence. Suppression Pool level will start to lower and continue to lower. The SRO will enter E0I-3 on flood alarms and eventually EOI-2 on Suppression Pool Level.

The SRO will determine that Suppression Pool level cannot be maintained above 11.5 feet and enter EOI-1 to scram the reactor and then to Emergency Depressurize.

An ATWS will exist on the scram, the crew will work through EOI- 1 and C-5 to insert control rods, maintain reactor level and pressure. The SRO will transition to C-2 to Emergency Depressurize.

Attempts to add water to the suppression pool will be unsuccessful with the failure of 3-FCV-73-30 and 3 -FCV-74-57.

The Emergency Classification is 1.2-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

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

3-D Page 5 of 54 Simulator Event Guide:

Event 1 Normal: Aligning Charcoal Filters for Parallel Flow JAW 3-01-66 Section 5.11 SRO Direct BOP to align Charcoal Filters for parallel flow.

BOP Align Charcoal Filters for Parallel Flow lAW 3-01-66 section 5.11.

5.11 Aligning Charcoal Filters for Parallel Flow:

[1] PLACE the OFFGAS TREATMENT SELECT handswitch, 3-XS-66-1 13, in TREAT.

[2] OPEN the CHARCOAL ADSORBER TRAIN 2 INLET VALVE, using 3-HS-66-1 17.

[3] OPEN the CHARCOAL ADSORBER TRAIN 1 DISCH VALVE, using 3-HS-66-1 18.

[4] CLOSE the CHARCOAL ADSORBER TRAINS SERIES VLV, using 3-ES-66-1 16.

[5] CHECK dewpoint temperature on OFFGAS MOIST SEP REHEATER TEMPERATURE recorder, 3-TRS-66-108, indicates 45°F or less (Red Pen).

[6] IF the Off-Gas System is intended to be operated with charcoal beds in parallel with the charcoal beds on another (shut down) unit, THEN

3-D Page 6 of 54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase Direct Power increase using Control Rods per 3-GOI-100-1A, section 5.4 5.4 Withdrawal of Control Rods while in Mode 2

[67] CONTINUE to withdraw control rods to raise Reactor power to approximately 8%. (REFER TO 3-01-85 and 3-SR-3. 1 .3.5(A))

ATC Raise Power with Control Rods per 3-01-85, section 6.6 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

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

CRD POWER, 3-HS-85-46 in ON.

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

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

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

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

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

STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].

[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to RE1NITIALIZE the RBM:

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

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

I

3-D Page 7of54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.3 Control Rod Notch Withdrawal

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

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

CRD ROD SELECT pushbutton is brightly ILLUMiNATED

White light on the Full Core Display ILLUMiNATED

Rod Out Permit light ILLUMiNATED

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

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

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

[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:

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

[6.2] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.

[6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes.

[6.4] IF Control Rod Coupling Integrity Check fails, THEN REFER TO 3-AOI-85-2.

3-D Page 8of54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 3-XS-85-40.

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

CRD ROD SELECT pushbutton is brightly ILLUMiNATED

White light on the Full Core Display ILLUMINATED

Rod Out Permit light ILLUMINATED

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[5.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

[5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47 and CRD CONTROL SWITCH, 3-HS-85-48.

[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.

3-D Page 9 of 54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[5.5] WHEN control rod settles into the intended notch, THEN CHECK the following.

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does NOT alarm.

[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

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

[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

3-D Page 10 of 54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2

[7] IF continuously withdrawing the control rod to position 48 and the control rod coupling integrity check will be performed after the CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.

[7.2] PLACE AND HOLD CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH.

[7.3] WHEN position 48 is reached, THEN RELEASE CRD NOTCH OVERRIDE, 3-HS-85-47, and CRD CONTROL SWITCH, 3-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

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

[7.6] CHECK control rod coupled by observing the following:

Four rod display digital readout AND full core display digital readout AND background light will remain illuminated.

CONTROL ROD OVERTRAVEL annunciator (3-XA-55-5A, Window 14) does NOT alarm.

3-D Page 11 of 54 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC [7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[7.8] IF control rod coupling integrity check fails, THEN REFER TO 3-AOI-85-2.

6.6.5 Return to Normal After Completion of Control Rod Withdrawal

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

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

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

Drivei Drivd WhenNRC directs 3 insert Trigger 1 for RcactbtRecic Pump 313 tri

3-D Page 12 of 54 Simulator Event Guide:

Event 3: Reactor Recirc Pump 3B Trip Driver Driver When NRC directs, insert Trigger .1 forReactorRecirç trip.

ATC Respond to numerous alarms and Report Trip of Reactor Recirc Pump 3B SRO Enter 3-AOI-68-1A Recirc Pump Trip/Core Flow Decrease OPRMs Operable ATC 4.2 Subsequent Actions

[1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A),

[1.1] SCRAM the Reactor.

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

Closes 3B Recirc Pump Discharge Valve ATC [3] IF Region I or II of the Power to Flow Map is entered, THEN Steps 3 through 8 are N/A SRO [9] NOTIFY Reactor Engineer to PERFORM the following:

Tech Specs 3.4.1

3-SR-3.4.1(SLO), Reactor Recirculation System Single Loop Operation

O-TI-248, Core Flow Determination in Single Loop Operation

3-D Page 13 of 54 Simulator Event Guide:

Event 3: Reactor Recirc Pump 3B Trip SRO Evaluate Tech Spec for Single Loop Operation TS 3.4.1 Condition A Condition A Requirements of the LCO not met.

Required Action A. 1 Satisfy the requirements of the LCO Completion Time 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months MODE Change not permitted until setpoint changes complete.

ATC [10] [NER/C] WHEN the Recirc Pump discharge valve has been closed for at least five minutes (to prevent reverse rotation of the pump) [GE SIL-517], THEN (N/A if Recirc Pump was isolated in Step 4.2[8])

OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibrium.

Opens Recirc Pump 3B discharge valve BOP [11] REFER TO the following ICS screens to help determine the cause of recirc pump trip/core flow lowering. VFDPMPB and VFDBAL

[12] CHECK parameters associated with Recirc Drive and Recirc Pump/Motor 3B on ICS and 3-TR-68-7l to determine cause of trip.

Dispatch personnel [13] PERFORM visual inspection of tripped Reactor Recirc Drive.

Dispatch personnel [14] PERFORM visual inspection of Reactor Recirc Pump Drive relay boards for relay targets.

Dnf The AsR&rEngmeer n1ie reqies reponp ups cus s 3 Dnve Dnv& When NRC direts, insertIigger5 for Corpray Pump 3Aadvertent starl

3-D Page 14 of 54 Simulator Event Guide:

Event 4: Core Spray Pump 3A Inadvertent Initiation Driver Driver When NRCdirçts insert TriggerS for core Spray Pump 3A inadvertent start Delete Trigger oeffide*immediatlyafier pump starts to allow.operator tosecurç pump Whn operator places Core Spray3Ato off insert.trjgger 6..

BOP Report inadvertent start of Core Spray Pump 3A and alarm CORE SPRAY SYS I PUMP A START BOP A. VERIFY auto start signals by multiple indications.

B. VERIFY Pump 3A operation by motor amps, discharge pressure, and flow on Panel 3-9-3.

B. IF pump is NOT needed, THEN STOP Pump before 5 mm time limit at minimum flow expires.

D. WHEN the auto start signal is reset and Core Spray is NOT required for Core Cooling, THEN E. RETURN system to standby readiness.

BOP Report drywell pressure and reactor level normal and stops Core Spray Pump 3A nyçr Lir n.operat rq pump thsgg BOP When pump is stopped reports loss of indication on pump (no lights)

BOP Dispatches personnel to breakers P4Y oei,k SRO Evaluate Technical Specification 3.5.1 Condition A One low pressure ECCS injection/spray subsystem inoperable.

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

Completion Time 7 Days Driver Dnver When NRC direct insert Trigger 10 for Ste Seal Regulator Failure

3-D Page 15of54 Simulator Event Guide:

Event 5: Steam Seal Regulator Failure ii 5i :i:. detdüp BOP Respond to Annunciator STEAM TO STEAM SEAL REG PRESS LOW A. CHECK steam seal header pressure, 3-PI-1-148, Panel 3-9-7.

B. VERIFY proper valve alignment on Panel 3-9-7.

C. IF pressure is low, THEN OPEN steam seal bypass valve 3-FCV-1-145.

D. DISPATCH personnel to check 3-PIC-1-147 (El 617 Turb Bldg).

E. CHECK condenser vacuum on 3-P/TR-2-2 (Panel 3-9-6) and turbine vibration on 3-XR-47-15 (Panel 3-9-7) normal.

BOP Responds to Annunciators STEAM PACKING EXHAUSTER VACUUM LOW OG HOLDUP LINE INLET FLOW HIGH BOP Recommends opening steam seal bypass valve 3-FCV- 1 -145 to restore steam pressure SRO Concurs with actions to restore steam seal pressure BOP Dispatches personnel and checks condenser vacuum p

3rer Jes&i1 ad I411iaS fail&1 Wb irir indatioj oic euté&s Lcac SRO Evaluate entry to 3-AOI-47-3 Loss of Condenser Vacuum BOP Once steam seal pressure is restored resets annunciators and verifies condenser vacuum is improving.

If 1&Cs privei w

3-b Page 16of54 Simulator Event Guide:

Event 6: Feedwater Pump Governor Drifts Up nyei iS hii&lC setTdggi5 GoyerWiwe. W1i bperatorfrikes the PE o

4 vcmor to manual themt1funetiou is autoiuatióally dqet4 thre1ore JoI?erafo; piWs the Gover ontrolicnob back ou c14 mneomustbe mul serted an1 deleted wbeji the qperato returns th Govenpr control kndb bacç doitoftree t1e operator to controflevel manually.

ATC Report Rising Reactor Water Level and RFPT is not responding.

SRO Direct manual control of operating RFPT and Enter 3-AOI-3-1.

4.2 Subsequent Actions

[1] VERIFY applicable automatic actions.

[16] IF Feedwater Control System has failed, THEN PERFORM the following:

[16.1] PLACE individual RFPT Speed Control Raise/Lower switches in MANUAL GOVERNOR (depressed position with amber light illuminated).

[16.2] ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain level.

[20] IF level continues to rise, THEN TRIP a RFP, as necessary.

[22] IF RFPs are in manual control, THEN LOWER speed of operating RFPs.

[23] EXPECT a possible Reactor power rise due to a rise in moderation.

[24] IF unit remains on-line, THEN PERFORM the following:

RETURN Reactor water level to normal operating level of 33 (normal range).

REQUEST Nuclear Engineer check core limits.

ATC Take MANUAL GOVERNOR control of RFPT and maintain Reactor Water Level Manually in the Normal Level Band. Operator may attempt to control RFPT with PDS.

PDS will not respond.

I3

3-D Page 17 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS i% Wii WbcUk ATC/BOP Respond to alarm multiple Pump Room Flood Level alarms and SIJPPR CHAMBER WATER LEVEL ABNORMAL ATC/BOP Report lowering suppression pool water level A. CHECK level using multiple indications.

B. IF level is low, THEN DISPATCH personnel to check for leaks.

C. IF level is high, THEN D. REFER TO 3-01-74, Sections 8.2, 8.3, and 8.4.

E. REFER TO Tech Spec Section 3.6.2.2.

F. IF level is above -1 or below -6.25, THEN ENTER 3-E0I-2 Flowchart.

tvq Qnye When. 4jpatcbetwait c minutes and report water 1e,e1 is4incbes and nsing iijiç Fputheast Qad. tsjg i41 fpitl apçt dçtein 9f1c1 SRO Enter EOI-2 on Low Suppression Pool Level Monitor and Control Suppression Pool Level Between -l inch and -6 inches (Appendix 18)

Answers No to Can Suppression Pool Level Be Maintained Above -6 inches Answers Yes to Can Suppression Pool Level Be Maintained Below -l inches CS#5 SRO Sets a Value for HPCI to place in Pull to Lock prior to 12.75 feet CS#5 ATC/BOP Places HPCI in Pull to Lock before Suppression Level lowers to 12.75 feet

3-D Page 18of54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Directs Appendix 18 BOP Appendix 18

6. IF Directed by SRO to add water to suppression pooi, THEN MAKEUP water to Suppression Pool as follows:

a. VERIFY OPEN 3-FCV-73-40, HPCI CST SUCTION VALVE.

b. OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE

c. IF HPCI is NOT available for Suppression Pool makeup, THEN MAKEUP water to Suppression Pool using RCIC as follows:

1) VERIFY OPEN 3-FCV-71-19, RCIC CST SUCTION VALVE.

2) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.

BOP Attempts to makeup water to the Suppression Pool using HPCI; 3-FCV-73-30 has lost power. Utilizes RCIC to makeup water to the Suppression Pool and dispatches personnel to investigate 3-FCV-73-30.

]

i 5

vr 3-FCV-73-30 pwcrJails when the Torus 1ekis inserted, crew will dispatd personnel to investigate owJedgJnyetigition and pzovidno further iIifbnnationi CS#2 SRO Detennines a trigger value for inserting a Reactor Scram on lowering Suppression Pool Water Level and enters EOI- 1, Scrams Reactor before Suppression Pool level reaches 11.5 feet.

SRO Determines that Emergency Makeup to the Suppression Pool using Standby Coolant is required and directs BOP to line up Standby Coolant to the Suppression Pool per Appendix 18.

3-D Page 19 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS BOP Appendix 18

5. IF Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN CONTINUE in this procedure at Step 9.

9. IF Directed by SRO to Emergency Makeup to the Suppression Pool using Standby Coolant Supply, THEN MAKEUP water to the Suppression Pool as follows:

a. VERIFY CLOSED the following valves:

3..FCV..74-61, RHR SYS I DW SPRAY INBD VALVE

3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VALVE

3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

3-FCV-23-46, RHR HX 3B RHRSW OUTLET VALVE

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

c. PLACE 3-BKR-074-O100, RHR HTX A-C DISCH XTIE (TO U-2) VLV FCV-74-lOO (MOlO-171) to ON (480V RMOV Board 3B, Compartment 19A).

d. START RHRSW Pumps Bi and B2.

e. NOTIFY Unit 1 Operator to VERIFY CLOSED 1-FCV-23-46, RHR HEAT EXCHANGER B COOL WATER OUTLET VLV

3-D Page 20 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS Driver en personnel dipateh d to close 3-BKR-074-0100, wait 1 minutes then close breaker and report delete ovemde for break control power When requested 1-FCV-2346 is closed: When requestd toopen 2.FCV43-57 insert i!te function swO9 open BOP f. NOTIFY Unit 2 Operator to perform the following

1) VERIFY CLOSED 2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2) OPEN 2-FCV-23-57, STANDBY COOLANT VLV FROM RHRSW.

g. INJECT Standby Coolant into the Suppression Pool as follows:

1) CLOSE 3-FCV-74-52, RHR SYS I LPCI OUTBD fNJECT VLV.

2) OPEN 3-FCV-74-lO0, RHR SYS I U-2 DISCH XTIE.

3) OPEN 3-FCV-74-57, RHR SYS I SUPPR CHMBR,POOL ISOL VLV.

4) THROTTLE OPEN 3-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV to control injection.

BOP Determines 3-FCV-74-57 will not open and is unable to Emergency Makeup to the Suppression Pool, dispatches personnel to determine cause of valve failure.

ic& cwledge spat dprpide frfoiiunfiLek hasopened Ml v41 Onqe all vye ened dele oyerrid ihs74i7a[2jJ auto and nfonm crey abqyaIv would iot çp di 4 cóntcts w hppioØ has bifixed SRO Enters EOI-3 on Flood Alarms

3-D Page 21 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS Enters EOI-3 on Flood Alarms SRO EOI-3 Secondary Containment Temp Monitor and Control Secondary CNTMT Temp Answers No to Is Any Area Temp Above Max Normal EOI-3 Secondary Containment Radiation Monitor and Control Secondary CNTMT Radiation Levels Answers No to Is Any Area Radiation Level Above Max Normal EOI-3 Secondary Containment Level Monitor and Control Secondary CNTMT Water Level Answers Yes to Is Any Floor Drain Sump Above 66 inches Answers Yes to Is Any Area Water Level Above 2 inches Restore and Maintain Water Levels using all available sump pumps Answers No to Can All Water Levels be Restore and Maintained Below Isolate all systems that are discharging into the area except systems required to:

Be operated by EOIs OR

Suppress a Fire Answers No to Will Emergency Depressurization Reduce Discharge Into Secondary Containment.

Enters EOI-1 at pre-determined trigger value and directs Reactor Scram based on EOI-2 step SRO SP/L-7.

ATC Inserts Reactor Scram, Initiates One Channel of ARI and reports rods out

3-D Page 22 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Enters EOI-1 from EOI-2 step SP/L-7 Verify Reactor Scram EOI-1 RC/P Monitor and Control RPV pressure Exits RC/P and enters C-2, Emergency RPV Depressurization, based on Override step RC/P-4.

EOI-1 RC/L Monitor and Control RPV Water Level Verify as Required:

PCIS Isolations (Groups 1,2 and 3)

ECCS

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

Crew may determine Reactor Subcritical and exit RC/Q, as long as NQ Boron has been injected, at any point during execution. If this is done Crew would enter AOl-i 00-1, Reactor Scram, based on override RC/Q-2.

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

Verify Reactor Mode Switch is in Shutdown Initiate second channel of ARI Verify Recirc Pump Runback (Pump speed 480rpm or less)

Answers No to is Reactor Power above 5% or Unknown (The Following steps N/A if RC/Q exited)

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

Initiates SLC per Appendix 3A

3-D Page 23 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS SRO EOI-1 RCIQ (cont)

Inhibit ADS Verify RWCU System Isolation Answers Yes to is SLC injecting into the RPV Stops at step RC/Q-1 8 until SLC has injected into the RPV to a tank level of 43%, then exits RC/Q and enters AOl- 100-1 Trips the SEC pump when SLC tank level drops to 0%

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

Verifies Recirc Pump at 480 rpm or less.

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

CS#4 BOP/ATC Appendix 3A UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A13B, control switch in START PUMP 3A or START PUMP 3B position.

2. CHECK SLC System for injection by observing the following:

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

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

SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 20).

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

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

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

3-D Page 24 of 54 Simulator Event Guide:

Event 7 Major: Torus LeakIATWS BOP/ATC Appendix 3A (continued)

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

4. VERIFY RWCU isolation by observing the following:

RWCU Pumps 3A and 3B tripped

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

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

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

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

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

SRO Enters C-5 from EOI-1 step RCIL-3 Override Step C5-l, states that IF Emergency Depressurization is required, THEN continue at step C5-l 9, however, if the SRO has not determined that ED is required at this time then he will continue at step C5-2 (below) lithibit ADS Answers Yes to is any Main Steam Line Open Bypass the following Isolation Interlocks:

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

RB Ventilation Low RPV Water Level (APPX 8E)

Crosstie CAD to DW Control Air, if necessary (APPX 8G) (Step N/A) hdt6esdgo

3-D Page 25 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS SRO Enters C-5 from EOI-1 step RC/L-3 (continued)

Answers No to is Reactor Power Above 5% or Unknown Establishes Reactor Water Level Band between -180 and +51 inches utilizing available injection sources listed on step CS-iS.

Upon detennination that Emergency Depressurization is required continues at step CS#112 C5-i9 and enters C-2 by direction of EOI-2 step SP/L-6 and from EOI-1 step RC/P 4 and directs Crew to Stop and Prevent all Injection Sources to the RPV Except from RCIC, CRD and SLC per step C5-20, in accordance with Appendix 4.

BOP/ATC Inhibits ADS (if not already done per Appendix 3A)

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

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

When directed performs Appendix 4 to Stop and Prevent all Injection Sources to the RPV Except from RCIC, CRD and SLC CS#1 BOP/ATC Appendix 4

1. PREVENT injection from HPCI by performing the following:

a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURB1NE TRIP push-button.

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

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

3-D Page 26 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS CS#1 BOPIATC Appendix 4 (continued)

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

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

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

OR

b. BEFORE RPV pressure drops below 450 psig,

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

AND

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

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

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

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

OR

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.

3-D Page 27 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS cs#i BOP/ATC Appendix 4 (continued)

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

Using 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL AND individual 3-SIC 8(9)(l0), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO, OR

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

SPEED CONTROL in MANUAL, OR

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

SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR.

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

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

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

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

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

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

3-HS-3-125A, RFPT 3A TRIP

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

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

CS#2 SRO Determines Emergency Depressurization is required and enters C-2 Answers No to will the reactor remain subcritical under all conditions. Waits until he receives the report that Appendix 4 is complete.

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

The Reactor will remain Subcritical without Boron under all conditions OR

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

OR

The Reactor is Subcritical and No Boron has been injected into the RPV

3-D Page 28 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS SRO Determines Emergency Depressurization is required and enters C-2 (continued)

Stops execution of execution of C-2 until Shutdown Cooling RPV Pressure Interlocks are clear Maintain RPV in Cold Shutdown per Appendix 1 7D BOP/ATC Reports when Appendix 4 is complete Reports Suppression Pool Level in Feet when Directed CS#2 Opens and Verifies Open ALL ADS Valves when directed SRO Upon commencement of Emergency Depressurization Continues in C-5 at step C5-21 Answers Yes to are at least 2 MSRVs open per C-2, Emergency RPV Depressurization Stops until RPV Pressure is below MARFP (l9opsig with 6 MSRVs open)

CS#3 Then continues Directs crew to Start and Slowly raise RPV Injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSH limits and Suppression Pool level per Appendix 6A or per Appendix 6B, 6C CS#3 BOP/ATC Appendix 6A

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

3-D Page 29 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS CS#3 BOP/ATC Appendix 6A (continued)

5. VERIFY OPEN the following heater isolation valves:

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

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

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

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

3-FCV-3-76, HP HTR 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

7. VERIFY at least one condensate pump running.

8. VERIFY at least one condensate booster pump running.

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

10. VERIFY RFW flow to RPV.

CS#3 BOP/ATC 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, THEN PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

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

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

3-D Page 30 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak!ATWS CS#3 BOP/ATC Appendix 6B (continued)

4. VERIFY CLOSED the following valves:

3-FCV-74-61, RHR SYS I DW SPRAY TNBD VLV

3-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV

3-FCV-.74-57, RHR SYS I SUPPR CHBR/POOL ISOL VLV

3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

5. VERIFY RHR Pump 3A and/or 3C running.

6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RHR SYS I LPCI INBD INJECT VALVE.

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

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

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

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

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV

3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV BOP/ATC Appendix 6C

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

3-D Page 31 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS BOP/ATC Appendix 6C (continued)

4. VERIFY CLOSED the following valves:

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

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

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

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

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

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

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

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

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

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

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

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

3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV BOP/ATC Starts and Slowly raises RPV Injection to Restore and Maintain RPV Water Level above

-180 inches irrespective of pump NPSH limits and Suppression Pool level per Appendix 6A or per Appendix 6B, 6C

3-D Page 32 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CS#4 SRO EOI-1 RC/Q steps RC/Q-20 and RC/Q-21 Reset ART Defeat ARI Logic Trips if necessary (APPX 2) (This step is N/A, however, crew may choose to perform this step)

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

1) Reset Scram Defeat RPS Logic Trips if necessary

2) Drain SDV

3) Recharge Accumulators

4) Initiate Reactor Scram Appendix 1D: Manual Control Rod Insertion Method

1) Drive Control Rods. Bypass RWM if necessary BOP/ATC Dispatch personnel to perform Appendix 2(N/A) and outside portions of Appendix iF.

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

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

3-D Page 33 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CS#4 ATC Appendix iF

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

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

4. DRAIN SDV UNTIL the following annunciators clear:

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

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

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

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

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

ALL control rods are fully inserted, OR

NO inward movement of control rods is observed, OR

SRO directs otherwise.

ggpjiop, ai 3Wn f1iinsert mrf b6çfosj Tien popjj I When directed tore-open 3-FCV-85-586 wait 3 mmutesthen msemr1rdO open Then rçport completion

3-D Page 34 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS CS#4 ATC Appendix 1D

1. VERIFY at least one CRD pump in service.

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

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

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

a. SELECT control rod.

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

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

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

ATC Continue performance of Appendix 1 F and 1 D until all rods inserted OR Until EOI-1 RC/Q is exited due to Reactor determined to be Subcritical at which point continue to insert rods per 3-AOI-100-.l and 3-01-85

3-D Page 35 of 54 Simulator Event Guide:

Event 7 Major: Tonis Leak/ATWS SRO Executes all legs of EOI-2 concurrently (SP/L leg has been previously addressed)

EOI-2 DW/T Monitor and control Drywell Temperature below 1 60F using available Drywell Cooling Answers Yes to can Drywell Temperature be maintained below 1 60F EOI-2 PC/P Monitor and control Primary Containment pressure below 2.4 psig using the vent system (APPX 12) as necessary Answers Yes to can Primary Containment pressure be maintained below 2.4 psig EOI-2 PC/fl Monitor and control Drywell and Suppression Chamber

Hydrogen at or below 2.4%

AND

Oxygen at or below 3.3%

Using the Nitrogen Makeup System (APPX 14A)

EOI-2 SP/T Monitor and control Suppression Pool temperature below 95F using available Suppression Pool Cooling (APPX 1 7A) as necessary Answers No to can Suppression Pool temperature be maintained below 95F (This is assuming Emergency Depressurization is complete and Reactor Water Level has been restored, if Emergency Depressurization has not been conducted yet, the answer will be Yes. If Reactor Water Level has not been restored yet, after Emergency Depressurization, this is not a priority.)

Directs Line up of all available Suppression Pool Cooling using only RHR pumps not required to assure adequate core cooling by continuous injection (APPX 1 7A) (After Emergency Depressurization complete and Reactor Water level restored)

BOP Performs Appendix 17A to place Suppression Pool cooling in service after Emergency Depressurization and restoration of Reactor Water level.

3-D Page 36 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS BOP Appendix 17A

1. If Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, Then BYPASS LPCI injection valve auto open signal as necessary by PLACiNG 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD INJ VLV BYPASS SEL in BYPASS.

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

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

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

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

3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV

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

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

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

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

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

f. If 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, Then VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

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

3-D Page 37 of 54 Simulator Event Guide:

Event 7 Major: Torus Leak/ATWS

= BOP Appendix 17A (cont)

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

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

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

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

k. MONITOR RHR Pump NPSH using Attachment 1.

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

m. If Additional Suppression Pool Cooling flow is necessary, Then PLACE additional RHR and RHRSW pumps in service using Steps 2.b through 2.1.

3-D Page 38 of 54 Critical Tasks Five CS#1-During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

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

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

4. Feedback:

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

Injection system flow rates into RPV.

CS#2-When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of Containment.

2. Cues:

Procedural compliance.

Suppression Pool level trend.

3. Measured by:

Observation US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet.

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

4. Feedback:

RPV pressure trend.

Suppression Pool temperature trend.

SRV status indication.

3-D Page 39 of 54 Critical Tasks Five CS#3-With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

RPV pressure indication.

3. Measured by:

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

4. Feedback:

RPV level trend.

RPV pressure trend.

Injection system flow rate into RPV.

CS#4-With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BITT) and inserting control rods.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Suppression Pool temperature.

3. Measured by:

Observation If operating JAW EOI-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool.

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

AND Control Rod insertion commenced in accordance EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

3-D Page 40 of 54 Critical Tasks Five CS#5-When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.

1. Safety Significance:

Prevent failure of Primary Containment from pressurization of the Suppression Chamber.

2. Cues:

Procedural compliance.

Suppression Pool Level indication

3. Measured by:

Observation HPCI Auxiliary Pump placed in Pull to Lock

4. Feedback:

HPCI does not Auto initiate No RPM indication on HPCI

3-D Page 41 of 54 Scenario Tasks EVENT TASK NUMBER KIA RO SRO 1 Align Charcoal Filters RO U-066-NO-22 271000A4.09 3.3 3.2 2 Raise Power with Control Rods RO U-085-NO-06 SRO S-000-AD-31 2.2.2 4.6 4.1 3 Reactor Recirc Pump Trip RO U-068-AB-O1 202001A2.03 3.6 3.7 SRO S-068-AB-O1 4 Core Spray Inadvertent Initiation RO U-075-NO-O1 209001A3.02 3.8 3.7 5 Steam Seal Regulator Failure RO U-OO1-AL-O1 2450001(6.01 2.8 2.9 SRO S-047-AB-03 6 Reactor Feed Pump Turbine Governor Failure RO U-003-AL-09 259002A4.01 3.8 3.6 SRO S-003-AB-O1 7 Torus Leak/ATWS RO U-000-EM-14 295030EA2.01 4.1 4.2 RO U-000-EM-17 RO U-000-EM-83 SRO S-000-EM-07 SRO S-000-EM-15 SRO S-000-EM- 18

3-D Page 42 of 54 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3-D 7 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List(l-3) 2 EOI Contingencies used: List (0-3) 90 Run Time (minutes) 5 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

Condensate Pump 3A tagged Out of Service.

Operations/Maintenance for the Shift:

Align Charcoal Filters for Parallel Flow lAW 3-01-66 Section 5.11.

Once completed Raise Power with Control Rods for Mode Change JAW 3-GOJ-100-1A, section 5.4 step

[67] and the Reactivity Control Plan Units 1 and 2 are at 100% power.

Unusual Conditions/Problem Areas:

None

Fn m

F rn t%)

I N LU

.1.

ii.

m.

.0

ni I

f k iri IIPi

i1, I

Lc, o r

a L

I tiliF hLwjtH ¶9#Eff 1 L

C, S

U H fi JJ P

Ui

3.4 REACTOR COOLANT SYSTEM (RCS) 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 ICO is applicable:

I

a. LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation ilnilts 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 (RIS) lnstrumentatlon Function 2,b (Average Power Range Monitors Flow Biased Simulated Thermal Power High), Allowable Value of Table 3,3.1,1.4 is reset for single loop operation, APPLICABILITY: MODES I and 2.

ACTIONS CONDITION REOUIREO ACTION COMPLETION TIME A. Requirements of the LCO Al Satisfy the requirements 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months not met. of the LCO.

5. Required Action and 5.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met OR No recirculation loops in operation,

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 35.1 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.

APPUCABIL1TY: MODEl, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig, ACTIONS LCO 3.0A.b is not applicable to HPCI, CONDITION REQUIRED ACTION COMPLETION TIME A, One low pressure ECCS A. I Restore low pressure 7 days Injection/spray subsystem ECCS lnjectton!spray inoperable, subsystem(s) to OPERABLE status.

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

(continued)

ECCS Operating 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and 81 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AND met.

B2 Be n MODE 4 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

ACTIONS (continued)

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

AND C2 Restore HPCI System to 14 days OPERABLE status D. HPCI System inoperabIe Dl Restore HPCI System to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status AND OR Condition A entered.

D2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injectionispray subsystem to OPERABLE status.

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

F. One ADS valve Fl Restore ADS valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status AND Condition A entered F2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ECCS injectionispray subsystem to OPERABLE status.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves Gi Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperabIe AND OR G.2 Reduce reactorsteam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and dome pressure to associated Completion 150 psig.

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

OR HPCI System and one or more ADS valves inoperable.

SCRAM FAILURE REACTOR COOLANT ACTMTY a

rceaOlor coolant activity exce rn dose C equiv&erttl-431 (Thc*wlcal Spe froaon LtniIs z as deiernned b chemistry sample.

I m

OPERATING CONDiTION m ALL z

-1 1,2A I I NOTE I I t3A 1 I I I Faihre of RPS automatic scram functians to bring Reactor coolant activity exceeds 300 I1Cm dose the reactor subcriticat equivalent lodine-ISI as determined hy chemistry AND saib.

Manual scram or ARt automatic or manual) was m successfuL OPERATING CONDrnON:

OPERATING CONDrnON: Model cr2 or 3 Mode lor2 i2S 1 INOTE I I I I I I Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical, m

m rn OPERATING CONDITION:

Model -<

1,2G ICURVEI I I OS I I I Failure of automatic scram, manual scram, and ARL Reactor power is ove 3%

AND Either of the fclicwing conditions exists:

Suppression Pool temp exceeds HCTL. in RefertoCurvet24,

Reactor water level can NOT be restored in and maintained ar or aloint -UI) inches, in z

OPERATING CONDITION:

MarIe Ior2

Appendix D Scenario Outline Form ES-D-1 rFacility: Browns Ferry NPP Scenario No.: E Op-Test No.: lET 1102 SRO:

Examiners: Operators: ATC:

BOP:

Initial IC 105/ Unit 2 Reactor Power 20%! Start Bus 1 A Out of Service! Shutdown Board C Conditions: Alternate Feeder Breaker is out of service for PMs. Severe Thunderstorm Warning in affect.

Turnover: Raise Power with Control Rods to commence Main Turbine Roll. Roll the Main Turbine lAW 2-01-47 section 5.4. Severe Thunderstorm Warning in affect.

Event Event No. Maif. No. Type* Event Description R-ATC 1 Raise Power with Control Rods R-SRO 2 rdO5rl 859 Uncoupled Control Rod N-BOP 3 Commence Main Turbine Roll N-SRO 4 trg 1 Turbine Speed Control Unit Failure

-I C-ATC RFPT Seal Injection Pump Trip C.SRO C-BOP 6 dgolc DG C Automatic start failure TS..SRO 7 th3O M-ALL Level 8 Instrument Failures Loss of High Pressure Injection! ATWS 8 th2l C LOCA 9 cs02b C Core Spray Loop 2 Inboard Injection Valve Auto Open failure

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

2-E Page 1 of 67 Console Operator Instructions A. Scenario File Summary

1. File: batch and trigger files for scenario 2-E Batch NRC/110203 Bat atws 90 Trgei NRC/iOOrpm = zdihs4777e.eq.i Trgei = bat NRC/i 10203-1 Trge2 NRC/turbtrip = zdihs4767d.eq. 1 Trge2 = bat NRC/i 10203-2 Trge3 NRC/rfpseal = zdihso369a[3].eq.1 Trge3 = bat NRC/i 10203-4 Imfdgolc Trge4 NRC/dgstart = zdiohs82oala[4].eq.1 Trg e4 = dmf ed09c Imfth30e (e5 0)100 60 Imfth30f(e5 120) 100 60 Imfth30g (e5 60) 100 60 Imfth30h (e5 120) 100 30 Imfth22 (e5 180) 100 Imfth2l (e5 420) .25 240 0 Imf cs02b br zlohsO369a[2] off br zlohsO369a[i] on Batch NRC/110203-i br zlohs4777h[1] off br zlohs4777e[1] on br zdihs4777{h] (none 30) select Batch NRC/i 10203-2 Dor zlohs4777e[1]

Dor zdihs4777h

2-E Page 2 of 67 Batch NRC/110203-3 br xa556c15 alarm on br xa556c1 8 alarm on br zdihs0368a[1] stop br zlohsO368a[1] off Batch NRC/i 10203-4 Dor xa556c15 Dor xa556c18 Dor zlohsO369a[2]

Dor zlohsO369a[1j Pref NRC/i10203 F3 imfrdO5ri859 F4 bat NRC/i 10203 F5 bat NRC/i 10203-3 F6 imfed09c F7 mrfrd06 open F8 bat appOif F9 bat app02 FiO mrfrd06 close Fli bat sdv Console Operator Instructions Scenario 2-E DESCRIPTION/ACTION Simulator Setup manual Reset to IC i 05 Simulator Setup Load File RestorePref NRC/il 0203 Simulator Setup Tag out Start Bus lA, SD BD C manual Alternate Feeder Breaker 1624 Tagged Out Simulator Setup F3 and F4 Verify file loaded RCP required (Raise Power with Rods to obtainS to 6 turbine bypass valves open)

RCP for Urgent Load Reduction Provide marked up copy of 2-GOI-100-1A Step 18

2-E Page 3 of 67 With Unit 2 at 20% power, the ATC will withdraw control rods to obtain 5 bypass valves open and the sixth bypass valve 25% to 50% open; in order to commence Main Turbine roll.

When withdrawing control rods, Control Rod 18-59 will fail the coupling check, the ATC will respond lAW 2-A0I-85-2 and re-couple Control Rod 18-59. The SRO will refer to Technical Specification 3.1.3; with the control rod coupled, no entry will be required. Follow up question after the scenario for Tech Spec action if control rod failed to couple.

Once the required bypass position is achieved, the BOP will commence Main Turbine roll JAW 2-01-47 section 5.4. Once Main Turbine roll is commenced, the Turbine Speed Control Unit will fail and turbine speed will continue to increase. The BOP will be required to trip the Main Turbine.

The operating RFPT Seal Injection Pump will trip, with a failure of the standby seal injection pump to start. The ATC will respond JAW the ARPs and start the standby seal injection pump to restore seal water to the operating RFPTs.

Maintenance work in the area of Shutdown Board C will cause the Normal Supply Breaker to trip. Diesel Generator C will fail to automatically start and tie to the shutdown board. The BOP will respond and start DG C. When DG C is started, it will automatically tie to the board. The SRO will evaluate Technical Specifications and determine TS 3.8.1 Condition B is entered.

Since one required offsite source is also out of service for SD BD C, Condition G is also entered; and Shutdown Board C is declared Inoperable. The SRO will then evaluate Technical Specification 3.8.7 and Condition A is entered.

The 208A D Level instruments will begin to fail high, tripping the operating RFPTs and not allowing HPCI or RCIC to inject. The crew will insert a scram on lowering Reactor Level. An ATWS will exist and the crew will enter 2-EOI-l, RPV Control and 2-C-5, Level/Power Control to take required actions.

2-E0I-2, Primary Containment Control will be entered when a small LOCA develops; eventually the SRO will determine that Reactor Level cannot be restored and maintained above -

180 inches and will transition to 2-C-2, Emergency RPV Depressurization. Core Spray Loop 2 inboard injection valve will fail to auto open but can be manually opened.

The Crew will insert all control rods and restore level to +2 to +51 inches.

The Emergency Classification is 1.2-S or 1.1-S 1 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

Control are inserted or being inserted Emergency Depressurization Complete Reactor Level restored and maintained above TAF

2-E Page 4 of 67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods SRO Direct Power Increase Direct Power increase JAW Reactivity Control Plan

[18] WHEN 5 to 6 turbine bypass valves are open (being careful NOT to exceed 25% Reactor power)

ATC Raise Power with Control Rods per 2-01-8 5, section 6.6 6.6.1 Initial Conditions Prior to Withdrawing Control Rods

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

CRD POWER, 2-HS-85-46 in ON.

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

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

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

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

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

STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 2-9-5 and PERFORM Step 6.6.2[6].

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

[6.1] PLACE CRD POWER, 2-HS-85-46 in the OFF position to deselect the Control Rod.

[6.2] PLACE CRD POWER, 2-HS-85-46, in the ON position.

2-E Page 5 of 67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods ATC 6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

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

CRD ROD SELECT pushbutton is brightly ILLUMiNATED

White light on the Full Core Display ILLUMINATED

Rod Out Permit light ILLUMiNATED

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH, and RELEASE.

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

[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:

[6.1] PLACE CRD CONTROL SWITCH, 2-115-85-48, in ROD OUT NOTCH, and RELEASE.

[6.2] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does not alarm.

[6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes.

[6.4] IF Control Rod Coupling Integrity Check fails, THEN REFER TO 2-AOI-85-2.

2-E Page 6 of 67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 2-XS-85-.40.

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

CRD ROD SELECT pushbutton is brightly ILLUMiNATED

White light on the Full Core Display ILLUMINATED

Rod Out Permit light ILLUMINATED

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[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 CR1) 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 CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

2-E Page 7 of 67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[5.5] WhEN control rod settles into the intended notch, THEN CHECK the following.

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does NOT alarm.

[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[6.3] MAINTAIN the CR1) Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48.

II

2-E Page 8of67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods ATC [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 2-AOI-85-2

[7] IF continuously withdrawing the control rod to position 48 and the control rod coupling integrity check will be performed after the CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[7.2] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[7.3] WhEN position 48 is reached, THEN RELEASE CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

[7.5] PLACE CRU CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[7.6] CHECK control rod coupled by observing the following:

Four rod display digital readout AND full core display digital readout AND background light will remain illuminated.

CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window 14) does NOT alarm.

2-E Page 9 of 67 Simulator Event Guide:

Event 1 Reactivity: Raise Power with Control Rods ATC [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 CRD POWER, 2-HS-85-46, in OFF.

[1.2] PLACE CRD POWER, 2-HS-85-46, in ON.

2-E Page 100f67 Simulator Event Guide:

Event 2 Component: Uncoupled Control Rod ATC Respond to Annunciator 5A-l4 and 28 CONTROL ROD OVERTRAVEL and CONTROL ROD DRIFT CONTROL ROD OVERTRAVEL A. VALIDATE alarm as follows:

1. Full Core Display will have no digital readout.

2. Background light extinguished.

3. Rod DRIFT light on.

B. IF alarm valid, THEN REFER TO 2-AOI-85-2.

C. NOTIFY Reactor Engineer.

D. REFER TO Tech Spec 3.1.3, 3.10.8.5, 3.3.2.1, and TRM TABLE 3.3.5-1.

Reports Control Rod 18-59 is Uncoupled or over traveled SRO Enter 2-AOI-85-2 Uncoupled Control Rod ATC 2-AOI-85-2 4.1 Immediate Actions

[1] STOP all control rod withdrawal.

SRO 4.2 Subsequent Actions

[1] NOTIFY Reactor Engineer to evaluate the suspect uncoupled control rod for its impact on core thermal limits and rod worth.

[2] ADJUST the rod pattern as directed by the Reactor Engineer throughout the performance of this procedure.

Driver T)ri góuncotpIedeontrcilrod,concur wftkcoupling tteifpei2-AOI-85-2

2-E Page 11 of 67 Simulator Event Guide:

Event 2 Component: Uncoupled Control Rod SRO [3] IF the control rod drive is at position 48 AND with Reactor Engineer concurrence, THEN PERFORM the following:

Direct ATC to attempt to couple Control Rod 18-59 PERFORM the following:

ATC [3.11 NOTCH INSERT the control rod drive to position 46 to attempt to couple the control rod.

[3.2] RESET associated annunciators.

[3.3] NOTCH WITHDRAW the control rod drive to position 48.

[3.4] PERFORM a coupling check.

[3.5] IF coupling integrity check fails, THEN CONTINUOUSLY INSERT control rod drive to position 00 to attempt to latch control rod with control rod drive mechanism.

[3.5.1] RESET associated annunciators.

[3.5.2] NOTCH WITHDRAW control rod to position 48.

[3.5.3] PERFORM a coupling check.

Report Control Rod 18-59 is Coupled SRO Makes notifications that Control Rod 1 8-59 is Coupled SRO Exits 2-AOI-85-2 and directs power ascension to continue qi SWfQ4Jeci

,j

2-E Page 12 of 67 Simulator Event Guide:

Event 3 Normal: Main Turbine Roll SRO Direct Turbine Roll

[18] WHEN 5 to 6 turbine bypass valves are open (being careful NOT to exceed 25% Reactor power), THEN:

[18.1] ROLL Turbine-Generator. REFER TO 2-01-47.

BOP Roll the Main Turbine JAW 2-01-47 section 5.4 starting at step 11 BOP [11] VERIFY OPEN the following valves on Panel 2-9-7:

STEAM LINES TO HP TURBINE DR VLV, 2-HS-6-109A

STOP VALVE I BEFORE SEAT DR VLV, 2-HS-6-100A

STOP VALVE 2 BEFORE SEAT DR VLV, 2-HS-6-1O1A

STOP VALVE 3 BEFORE SEAT DR VLV, 2-HS-6-102A

STOP VALVE 4 BEFORE SEAT DR VLV, 2-HS-6-103A

CONTROL VALVE I BEFORE SEAT DR VLV, 2-HS-6-104A

CONTROL VALVE 2 BEFORE SEAT DR VLV, 2-HS-6-105A

CONTROL VALVE 3 BEFORE SEAT DR VLV, 2-HS-6-106A

CONTROL VALVE 4 BEFORE SEAT DR VLV, 2-HS-6-107A

LP STEAM LiNES TO RFPTS DRAIN VALVES, 2-HS-6-1 1 1A

LP STEAM LINES TO RFPTS DRAiN VALVES, 2-HS-6-1 12A

2-E Page 13 of 67 Simulator Event Guide:

Event 3 Normal: Main Turbine Roll BOP [12] START EMERGENCY BEARING OIL PUMP using 2-HS.47-8A.

[13] OBSERVE pump operates for at least two minutes, THEN STOP EMERGENCY BEARING OIL PUMP using 2-HS-47-8A.

[14] PERFORM EHC Control System Lamp Test. REFER TO Section 6.2.

[15] VERIFY MOTOR SUCTION PUMP, 2-HS-47-12A, is in service.

[16] IF any Turbine lift pump motor has been disabled, N/A.

[17] VERIFY the following on EHC TURBINE CONTROL panel:

EHC SETPOINT, 2-PI-47-1 62, indicates approximately 955 psig in Reactor Pressure Control, if available.

Either REACTOR PRESSURE CONTROL, 2-HS-47-204, or HEADER PRESSURE CONTROL, 2-HS-1-16, is ILLUMINATED.

BPV DEMAND, 2-ZI-47-l 30, indicates zero.

TURBiNE TRIPS NORMAL green light, 2-IL-47-87, is illuminated.

VACUUM TRIP BYPASSED, 2-IL-47-72, is extinguished.

BPV VACUUM INHIBIT, 2-IL-47-73, is extinguished.

ALL VALVES CLOSED pushbutton backlight, 2-HS-47-77D, is illuminated.

CV POSITION LIMIT, 2-XI-47-157, set at approximately 66%.

LOAD SET, 2-XI-47-75, set at approximately zero.

2-E Page 14 of 67 Simulator Event Guide:

Event 3 Normal: Main Turbine Roll BOP [18] CHECK all first out annunciators are RESET.

[19] STATION an Operator on the turbine floor and PERFORM the following:

[19. 1] CHECK the following oil pressures at the front standard:

TURB SHAFT OIL PUMP SUCTION PRESS, 2-PI-47-53, indicates greater than 15 psig.

TURB BRG OIL HEADER PRESS, 2-PI-07-56, indicates greater than 15 psig.

[19.2] MONITOR for rubbing, vibration, or unusual noises during turbine roll, AND CHECK for adequate bearing oil flow by comparison to other bearing discharge weirs.

Dr Prive 1 re fer no abnormal oippI&r

[20] IF Turbine seals have been in service with the Turbine Turning Gear secured AND unit is to be returned to operation, THEN (Otherwise N/A)

[21] VERIFY the following initial conditions are satisfied.

Turbine shaft eccentricity is less than 0.5 mils or has been less than two mils for the last one hour as indicated on TURB GEN ECC/SPEED/VALVE P0 SN, 2-XR-47-1 6, Local Vibration Monitoring from portable analyzer, or from Engineering Judgment from Observation of Trend Recorder at EHC Work Station on Panel 9-31, Auxiliary Instrument Room.

EXTERNAL VALVE CHEST UPPER INNER SURFACE temperature, 2-XR-47-20, Point 1, located on Panel 2-9-8, indicates greater than 485°F.

TURB FIRST STAGE BOWL INNER SURFACE TEMP, 2-XR-47-20, Point 5, indicates greater than 25 0°F.

Reactor is at rated pressure with 4.5 (four and a half) to 6 (six) turbine bypass valves open.

Condenser vacuum is greater than 25 Hg vacuum.

Turbine oil temperature is being controlled by TURBiNE OIL TEMPERATURE CONT, 2-TIC-24-75.

IF oil temperature is lower than 2-TIC-24-75 setpoint, THEN (Otherwise N/A)

2-E Page 15of67 Simulator Event Guide:

Event 3 Normal: Main Turbine Roll BOP [22] SHIJT DOWN from chest warming as follows.

[22.1] DEPRESS MSV-2 Pilot Position LOWER pushbutton, 2-HS-47-78B, UNTIL MSV-2 PILOT POSITION, 2-ZI-47-78, indicates zero.

[22.2] MOMENTARILY DEPRESS Main Stop Valve Position Demand OFF pushbutton, 2-HS-47-78A.

[23] PERFORM the following:

[23.1] At TURBINE OIL TEMPERATURE CONT, 2-TIC-24-75:

[23.1 .1] VERIFY controller in auto.

[23.1.2] ADJUST setpoint to 100°F using and keys.

[23.1.3] DEPRESS SET/ENT key to lock in new setpoint at 100°F.

[24] NOTIFY Radiation Protection an RPHP is in effect for the impending action to roll the main turbine. RECORD time Radiation Protection notified in NOMS Narrative Log.

[24.1] VERIFY appropriate data and signatures recorded on Appendix A per Appendix A instructions.

[25] SELECT proper turbine start up rate as follows:

[25.1] IF TURB FIRST STAGE BOWL INNER SURFACE TEMP, 2-XR-47-20, Point 5, located on Panel 2-9-8, indicates between 250°F and 350°F, THEN DEPRESS MEDIUM pushbutton, 2-HS-47-77B.

[25.2] IF TURB FIRST STAGE BOWL INNER SURFACE TEMP, 2-XR-47-20, Point 5, located on Panel 2-9-8, indicates greater than 350°F, THEN DEPRESS FAST pushbutton, 2-HS-47-77C.

[26] VERIFY all Turbine lift pump motors are operating.

[27] VERIFY MAiN TURBINE VIB TRIP BYPASS, 2-HS-47-26, in TRIP BYPASS, AND CHECK amber light illuminates above the handswitch.

2-E Page 16 of 67 Simulator Event Guide:

Event 4 Component: Turbine Speed Control Unit Failure BOP [28] On TURBINE SPEED SELECT Section, DEPRESS 100 RPM pushbutton, 2-HS-47-77E, AND OBSERVE the following:

Main Stop Valve, MSV-2, 2-ZI-l -78, begins to open.

When MSV-2 reaches full open, MSV-1, MSV-3, and MSV-4 open slowly.

Combined Intermediate Valves ISV/TV-i, ISVIIV-3 and ISV/1V-5 open slowly.

When IV-l, 3, and 5 are full open, IV-2, 1V-4, and IV-6 will open.

Control Valves, CV-l, CV-2, CV-3, and CV-4 throttle open and the turbine rolls off the turning gear.

The TURBINE ACCEL light, 2-IL-47-77A, illuminates.

CONTROLLING PARAMETER Section SPEED light, 2-IL-47-77, is illuminated.

TURB GEN ECC/SPEED/VALVE POSN, 2-XR-47-16, tracks turbine speed instead of eccentricity, (If Available). (Red Pen)

Digital speed indication rising on TURBINE SPEED, 2-SI-47-77.

[29] PLACE Generator Condition Monitor in service. Refer to 2-01-35.

[30] PLACE hydrogen purity analyzer from vent to normal operation. Refer to 2-01-3 5.

[311 IF while performing step 5.4[32]the turbine continues to accelerate, indicating a Control System Failure ThEN PERFORM the following:

[31.1] DEPRESS Turbine TRIP pushbutton, 2-HS-47-67D.

[31.2] VERIFY the Main Stop, Control and Combined Intermediate Valves CLOSE and Turbine speed lowers.

BOP Trips the Main Turbine Ic

2-E Page 17 of 67 Simulator Event Guide:

Event 4 Component: Turbine Speed Control Unit Failure SRO Enter 2-AOI-47-1 BOP 4.2 Subsequent Actions

[1] VERIFY Automatic Actions listed in Section 3.0 have occurred.

[2] PLACE TURNING GEAR OIL PUMP, 2-HS-47-1 1A, in START.

[3] PLACE MOTOR SUCTION PUMP, 2-HS-47-12A, in START.

[4] SET TURBiNE OIL TEMPEPATURE CONT, 2-TIC-24-75, to 85°F.

[5] OPEN the following drain valves by placing the following control switches to OPEN:

STOP VALVE A BEFORE SEAT DR VLV, 2-115-6-bOA.

STOP VALVE B BEFORE SEAT DRVLV, 2-HS-6-1O1A.

STOP VALVE C BEFORE SEAT DR VLV, 2-HS 1 02A.

STOP VALVE D BEFORE SEAT DR VLV, 2-HS-6-103A.

CONTROL VALVE A BEFORE SEAT DR VLV, 2-HS-6-104A.

CONTROL VALVE B BEFORE SEAT DR VLV, 2-HS-6-105A.

CONTROL VALVE C BEFORE SEAT DR VLV, 2-HS-6-106A.

CONTROL VALVE D BEFORE SEAT DR VLV, 2-HS-6-107A.

STEAM LINES TO HP TURBINE DR VLV, 2-HS-6-109A.

LP STEAM LINES TO RFPTS DRAIN VLVS, 2-HS-6-1 1 1A.

LP STEAM LINES TO RFPTS DRAIN VLVS, 2-HS-6-l 12A.

[6] WHEN Turbine speed lowers to 900 RPM, THEN START the bearing lift pumps by placing TURBINE TURNING GEAR MOTOR, 2-HS-47-1OA, to ON.

[7] VERIFY the exhaust hood sprays maintain exhaust hood temperature below 135°F.

Nq ic

2-E Page 18 of 67 Simulator Event Guide:

Event 5 Component: RFPT Seal Injection Pump Trip ATC Respond to Alarm 6C-18 6C-l 8 RFP Seal Injection Water Pressure Low NOTE If the alarm is the result of a common loss of injection water pressure, the spare injection water pump will auto start at 22 psid on injection water common header.

A. IF only one Feedpump abnormal alarm is illuminated, THEN B. IF one or more Feedpump abnormal alarm is illuminated and cannot be cleared by the above methods, ThEN

1. VERIFY spare injection water pump started on Panel 2-9-6.

2. DISPATCH personnel to check injection water pumps.

3. ALTERNATE if required injection water strainers in accordance with 2-01-3, if 2-PDIS-003 -0230 is reading 5 psid or greater.

ATC Start Spare RFPT Seal Injection Pump to return seal injection to service.

BOP Dispatch personnel ye Dijy Lt9 BeV -99 is tnppedEfre frequestedstrainerDPis2.Qpsid when standbypump started feeder breker Cei

2-E Page 19 of 67 Simulator Event Guide:

Event 6 Component: DG C Automatic Start Failure BOP Respond to alarms 8B-29, 30 and 33, 23C-28, 23B-19 and 20, 23A-20 8B-33 4KV SHUTDOWN BD C DEGRADED VOLTAGE A. VERIFY Diesel starts and ties to board, if required.

B. REFER TO 0-OI-57A, Tech Spec 3.8.7 and 3.8.8.

8B-29 480V SHUTDOWN BD 2A UV OR XFR B. IF 480V Shutdown Bd 2A is lost, THEN Manually TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 2-9-8.

C. IF manual transfer is accomplished, THEN REFER TO 0-OI-57B, 2-01-99, and appropriate Ols for recovery or realignment of equipment.

8B-30 480V SHUTDOWN BD 2B UV OR XFR B. IF 480V Shutdown Bd 2B is lost, THEN Manually TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 2-9-8.

C. IF manual transfer is accomplished, THEN REFER TO 0-OI-57B, 2-01-99, and appropriate Ols for recovery or realignment of equipment.

BOP Starts DG C, verifies DG Energizes Shutdown Board C Crew Dispatch personnel to investigate SRO Evaluate Technical Specification 3.8.1

2-E Page 20 of 67 Simulator Event Guide:

Event 6 Component: DG C Automatic Start Failure BOP Respond to alarms 8B-29, 30 and 33, 23C-28, 23B-19 and 20, 23A-20 23C-28 4160V SD BD C DEGRADED VOLTAGE Automatic Action A. Diesel generator starts.

Operator Action A. VERIFY unit in stable condition by checking:

1. Auto actions have occurred and diesel at rated frequency and voltage.

2. Diesel connects properly to assigned 41 60V Shutdown Bd.

3. Diesel reaches proper loading when supplying equipment fed from board.

BOP Starts DG C, verifies DG Energizes Shutdown Board C ftç D9 øs redc11 otJit aearea rççe4 iteaked71 8witha hand tru<1cwbile sfaging equipmntTeBreker trippe4 SRO Evaluate Technical Specification 3. 8.1

ØJI ggrg!

2-E Page 21 of67 Simulator Event Guide:

Event 6 Component: DG C Automatic Start Failure SRO Evaluate Technical Specification 3.8.1 Condition B One required Unit 1 and 2 DG inoperable Required Action B.1 Verify power availability from the offsite transmission network.

Required Action B.2 Declare required feature(s), supported by the inoperable Unit 1 and 2 DG, inoperable when the redundant required feature(s) are inoperable.

Required Action B.3.1 Determine OPERABLE Unit 1 and 2 DG(s) are not inoperable due to common cause failure.

OR B.3.2 Perform SR 3.8.1.1 for OPERABLE Unit 1 and 2 DG(s).

Required Action B.4 Restore Unit 1 and 2 DG to OPERABLE status.

Completion Time B. 1 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter B.2 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of condition B concurrent with inoperability of redundant required feature(s).

B.3.1 or B.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.4 7 Days and 14 days from discovery of failure to meet LCO Condition G Note: Applicable when only one 4.16 kv shutdown board is affected.

One required offsite circuit inoperable.

And One Unit 1 and 2 DG Inoperable.

Required Action G. 1 Declare the affected 4.16 kv shutdown board inoperable.

Completion Time Immediately SRO Evaluate Technical Specification 3.8.7 Condition A One Unit 1 and 2 4.16 kv Shutdown Board inoperable.

Required Action A. 1 Restore the Unit 1 and 2 4.16 kv Shutdown Board to OPERABLE status.

Required Action A.2 Declare the associated diesel generator inoperable.

Completion Time A. 1 5 days and 12 days from discovery of failure to meet LCO.

A.2 Immediately Lc Iw

2-E Page 22 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Tnj ection/ATWS ATC Recognizes Trip of RFPTs on Level 8 and lowering Reactor Level SRO Direct Reactor Scram and Enter EOI- 1 RPV Control on Scram ATC Scram the Reactor and place the mode switch in shutdown, report ATWS and initiate one channel of ART SRO EOI-1 Reactor Pressure Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig NO if Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate NO IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization NO IF RPV water level cannot be determined NO Is any MSRV Cycling 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 Direct a Pressure Band of 800 to 1000 psig ATC/BOP Maintain directed pressure band

2-E Page 23 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS SRO EOI-1 Reactor Level Monitor and Control Reactor Level Verify as required PCIS isolations group (1,2 and 3), ECCS and RCIC, Directs group 2 and 3 verified ATCIBOP Verifies Group 2 and 3 isolation SRO IF It has not been determined that the reactor will remain subcritical THEN Exit RC/L enter C5 Level I Power Control IF Emergency Depressurization is required NO -

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

...S#1 SRO Direct ADS Inhibited CS#1 ATCIBOP Inhibits ADS SRO Is any Main Steam Line Open YES -

Direct Bypass of isolation interlocks Appendix 8A and Appendix 8E Crew Call for Appendix 8A and 8E enp1for 4endix8A SE is cornpIp ATCIBOP Appendix 8A

3. Operator to verifies closed the following valves (Unit 2 Control Room, Panel 9-3):

2-FCV-43-13, RX RECIRC SAMPLE INBD ISOLATION VLV 2-FCV-43-14, RX RECIRC SAMPLE OUTBD ISOLATION VLV.

2-E Page 24 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ection!ATWS SRO C5 Level / Power Control Crosstie CAD to DW Control Air if necessary (Appendix 8G) NOT Necessary IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND An MSRV is open or cycling or drywell pressure is above 2.4 psig AND RPV water level is above -162 inches -NO SRO Is Reactor Power above 5% IF YES Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC (Appendix 4)

Direct Terminate and Prevent lAW Appendix 4 ATC/BOP Terminate and Prevent lAW Appendix 4 CS#2 BOP/ATC Appendix 4

1. PREVENT injection from HPCI by performing the following:

a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 2-HS-73-18A, HPCI TURBINE TRIP push-button.

b. WHEN HPCI Turbine is at zero speed, THEN PLACE 2-HS 47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 2-HS-73-18A, HPCI TURBiNE TRIP push-button.

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

2-E Page 25 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS CS#2 BOP/ATC Terminate and Prevent lAW Appendix 4 Appendix 4 (continued)

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

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

a. Following automatic pump start, PLACE R}IR SYSTEM I pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

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

AN))

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

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

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

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

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 2-HS74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-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.

2-E Page 26 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS CS#2 BOP/ATC Terminate and Prevent lAW Appendix 4 Appendix 4 (continued)

c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:

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

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

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

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

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

2-HS-3-125A, RFPT 3A TRIP

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

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

SRO Direct a Level Band

2-E Page 27 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Injection/ATWS CS#5 SRO IF RPV Water Level Cannot be Restored and Maintained Above -180 inches THEN continue at step C5-19 C5-l 9 Emergency Depressurization is required Enter C2 Emergency Depressurization Will the Reactor remain subcritical without Boron under all conditions NO CS#2 When all injection into the RPV is stopped and prevented except from RCIC, CRD, and SLC per CS THEN Continue Is Suppression Pool Level above 5.5 Feet YES Open all ADS Valves CS#5 ATC/BOP Opens 6 ADS Valves SRO Can 6 ADS Valves be Opened YES -

C5 Level / Power Control Are at least 2 MSRVs Open per C2 Emergency RPV Depressurization YES -

CS#3 When RPV Pressure is Below MARFP (190 psig) THEN Continue Start and slowly raise RPV injection with the following injection sources to restore and maintain RPV water level above -180 inches. Condensate Appendix 6A, LPCI Appendix 6B, 6C CS#3 ATC/BOP Once Reactor Pressure is below 190 psig commence injection lAW Appendix 6A, 6B and 6C.

2-E Page 28 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS SRO EOI-l Power Control Monitor and Control Reactor Power IF The reactor will remain sub subcritical without boron under all conditions NO IF The reactor is subcritical and No boron has been injected NO Verify Reactor Mode Switch in Shutdown Initiate ART ATC Initiates ART SRO Verify Recirc Runback ( pump speed 480 rpm)

ATC Verifies Recirc Runback SRO Is Power above 5% YES -

Trip Recirc Pumps ATC Trips Recirc Pumps CS#4 SRO Before Suppression Pool temperature rises to 110°F Continue Boron injection is required (If still critical with challenge to BITT)

CS#4 ATC/BOP Initiate SLC JAW Appendix 3A CS#4 SRO Direct API Reset Appendix 2 CS#4 Insert Control Rods Using one or more of the following methods:

Appendix 1F, Appendix 1D jirtver Dnv Whn cE& imijutes a&Iport apndi 2 complete arid>fieldactionlorappenthx Ieorp1ete8 bat appQ1fandF9batppO2 CS#4 ATC Insert Control Rods JAW Appendix 1D and 1F

2-E Page 29 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ection/ATWS CS#4 ATC Insert Control Rods JAW Appendix iF

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

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

4. DRAIN SDV UNTIL the following annunciators clear:

WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 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 CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ART.

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

ALL control rods are fully inserted, OR

NO inward movement of control rods is observed, OR

SRO directs otherwise.

Whea dispatch o,ex wM 2 mimitcs 4ppo 2HV-O85OSØ qpg wbei 1ced tçc1 afItes adrçpo jO6 c1ose74Qp

2-E Page 30 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS CS#4 ATC Insert Control Rods JAW Appendix 1D

1. VERIFY at least one CRD pump in service.

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

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

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

a. SELECT control rod.

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

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

6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER SOV (RB NE, El 565 +/-1).

ftedqa1cbed biosabiiiinutes nr osIV-Osq8 1se4ywiie4 s1ced4o opqrwait2nijnutes ortoyei. JW nfrdO6 close F7 thrfr4Op

2-E Page 31 of67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS CS#4 BOP/ATC Initiate SLC lAW Appendix 3A

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

2. CHECK SLC System for injection by observing the following:

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

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

. SLC SQU]B VALVE CONTINUITY LOST Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 20).

2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

System flow, as indicated by 2-IL-63-l 1, SLC FLOW, red light illuminated on Panel 2-9-5,

. SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 14).

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

4. VERIFY RWCU isolation by observing the following:

. RWCU Pumps 2A and 2B tripped

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

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

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

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

7. MONITOR 2-LI-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

2-E Page 32 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS CS#3 ATC/BOP Once Reactor Pressure is below 190 psig commence injection 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, REP 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 CNI)S INLET ISOL VLV 2-FCV-2-96, DRMN 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-3-31, HP 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 REP suction valves:

2-PCV-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 Restores Level to directed level band ATC/BOP Once Reactor Pressure is less than 450 psig report failure of Core Spray Loop 2 Inboard Injection Valve to Auto Open.

2-E Page 33 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS CS#3 ATC/130P Once Reactor Pressure is below 190 psig commence injection lAW 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, THEN PLACE 2-HS 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, R1{R SYS I SUPPR CHI3R/POOL ISOL VLV

2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE

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

5. VERIFY R}IR Pump 2A and/or 2C running.

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

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

8. THROTTLE 2-FCV-74-52, RHR SYS I 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:

2-FCV-23-34, RHR FIX 2A RHRSW OUTLET VLV 2-FCV-23-40, RHR lix 2C RHRSW OUTLET VLV.

Restores Level to directed level band ATC/BOP Once Reactor Pressure is less than 450 psig report failure of Core Spray Loop 2 Inboard Injection Valve to Auto Open.

2-E Page 34 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Injection!ATWS CS#3 ATC/BOP Once Reactor Pressure is below 190 psig commence injection JAW Appendix 6C.

1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 2-HS l55B, LPCI SYS II OUTBD INJ VLV BYPASS_SEL in BYPASS.

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

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

4. VERIFY CLOSED the following valves:

2-FCV-74-75, RHR. SYS II DW SPRAY ESIBD 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 andlor 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.

9. MONITOR RHR Pump NPSH using Attachment 1.

10. PLACE RHRSW pumps in service as soon as possible on ANY ERR Heat Exchangers_discharging to_the RPV.

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

2-FCV-23-46, ERR HX 2B RHRSW OUTLET VLV 2-FCV-23-52, RHE. HX 2D RHRSW OUTLET VLV.

Restores Level to directed level band ATC/BOP Once Reactor Pressure is less than 450 psig report failure of Core Spray Loop 2 Inboard Injection Valve to Auto Open.

2-E Page 35 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS SRO Enter EOI-2 Containment Control EOI-2 Drywell Temperature SRO Monitor and Control DW Temp Below 160°F using available DW Cooling Can Drywell Temp Be Maintained Below 160°F NO -

Operate all Available Drywell Cooling SRO Primary Containment Hydrogen If PCIS Group 6 isolation exists YES Then

1. Place analyzer isolation bypass keylock switches to bypass

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps BOP 1. Place analyzer isolation bypass keylock switches to bypass

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps SRO EOI-2 Suppression Pool Temperature Monitor and Control Suppression Pool Temperature Below 95°F Using Available Suppression Pool Cooling As Necessary (Appendix 1 7A)

Can Suppression Pool Temperature Be Maintained Below 95°F YES until Emergency Depressurization After Emergency Depressurization Operate all available Suppression pool cooling using only RHR Pumps not required to assure adequate core cooling by continuous injection Appendix 17A ATC/BOP Place an RHR System in Pool Cooling when directed JAW Appendix 17A SRO The Emergency Classification is 1.2-S or 1.1-Si

2-E Page 36 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS SRO EOI-2 Suppression Pool Level Monitor and Control Suppression Pool Level between -l inch and -6 inches (Appendix 18)

Can Suppression Pool Level be maintained above -6 inches YES Can Suppression Pool Level be maintained below -1 inch YES SRO EOI-2 Primary Containment Pressure Monitor and Control PC Pressure Below 2.4 psig Using the Vent System As Necessary (Appendix 12)

Direct Appendix 12 Vent SRO Can Primary Containment pressure be maintained below 2.4 psig NO BOP Vent Containment lAW Appendix 12 SRO Before Suppression Chamber Pressure rises to 12 psig Continue Initiate Suppression Chamber Sprays Using only pumps not required to assure adequate core cooling by continuous injection. Appendix 1 7C ATC/BOP Initiate Suppression Chamber Sprays JAW Appendix 17C SRO When Suppression Chamber Pressure exceeds 12 psig THEN Continue SRO The Emergency Classification is 1.2-S or 1.1-Si

2-E Page 37 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS BOP Vents Primary Containment lAW Appendix 12

1. VERIFY at least one SGTS train in service.

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

2-FCV-64-3 1, DRYWELL INBOARD ISOLATION VLV, 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE.

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

7. CONTINUE in this procedure at:

Step 8 to vent the Suppression Chamber through 2-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.

8. VENT the Suppression Chamber using 2-FIC-84-19, PATH B VENT FLOW CONT, as follows:

a. PLACE keylock switch 2-HS-84-35, DW/SUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).

c. PLACE 2-FIC-84-l9, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

d. PLACE keylock switch 2-HS-84-19, 2-FCV-84-19 CONTROL, in OPEN (Panel 2-9-55).

e. VERIFY 2-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.

f. CONTINUE in this procedure at step 12.

2-E Page 38 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS BOP Vents Primary Containment JAW Appendix 12

9. VENT the Suppression Chamber using 2-FIC-84-20, PATH A VENT FLOW CONT, as follows:

a. VERIFY OPEN 2-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 2-9-3).

b. PLACE keylock switch 2-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).

c. VERIFY OPEN 2-FCV-64-34, SUPPR CHBR INBOARI) ISOLATION VLV (Panel 2-9-54).

d. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).

e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-55).

f. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.

g. CONTINTJE in this procedure at step 12.

12. ADJUST 2-FIC-84-19, PATH B VENT FLOW CONT, or 2-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:

Stable flow as indicated on controller, AND 2-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:

iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 jtCi/s ANT) 0-SI-4.8.B. l.a. 1 release fraction of 1.

L

2-E Page 39 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Injection!ATWS ATC Place Suppression Pool Cooling in service lAW Appendix 17A IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock AS NECESSARY:

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

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

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

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

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

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

2-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2-FCV-23-40, RHR HX 2C RHRSW OUTLET VLV

2-FCV-23-52, RRR HX 2D RHRSW OUTLET VLV.

d. IF Directed by SRO, THEN PLACE 2-XS-74-122(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.

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

f. IF 2-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11)

LPCI OUTBD INJECT VALVE.

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

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

i. ThROTTLE 2-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 2-FI-74-50(64), RHR SYS 1(11) FLOW:

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

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

k. MONITOR RHR Pump NPSH using Attachment 1.

2-E Page 40 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS ATC/BOP Initiate Suppression Chamber Sprays JAW Appendix l7C 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.

Step 3 and 4 are NA

5. INITIATE Suppression Chamber Sprays as follows:

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

b. IF EITHER of the following exists:

LPCI Initiation signal is NOT present, OR

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

c. MOMENTARILY PLACE 3-XS-74-12l(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[-IR System 1(11) pump(s) for Suppression Chamber Spray.

f. VERIFY OPEN 2-FCV-74-57(71), RHR SYS 1(11) SJJPPR CHBR/POOL ISOL VLV.

2-E Page 41 of 67 Simulator Event Guide:

Event 7 Major: Loss of High Pressure InjectionlATWS ATC/BOP Initiate Suppression Chamber Sprays JAW Appendix 17C

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

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

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

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

k. MONITOR RHR Pump NPSH using Attachment 2.

1. VERIFY RHRSW pump supplying desired RHR 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 2A RHRSW OUTLET VLV

2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2-FCV-23-40, RHR HX 2C RHRSW OUTLET VLV

2-FCV-23-52, RHR HX 2D RHRSW OUTLET VLV.

2-E Page 42 of 67 Critical Tasks Five CS#1-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 Inhibited annunciator status.

CS#2-During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

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

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

4. Feedback:

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

Injection system flow rates into RPV.

2-E Page 43 of 67 Critical Tasks Five CS#3-With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF (-180) 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 or equal to

-180.

4. Feedback:

RPV level trend.

RPV pressure trend.

Injection system flow rate into RPV.

CS#4-With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BITT) and inserting control rods.

1. Safety Significance:

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

2. Cues:

Procedural compliance.

Suppression Pool temperature.

3. Measured by:

Observation If operating per EOI- 1 and C-5, US detennines that SLC is required (indicated by verbal direction or EOT placekeeping action) before exceeding 110 degrees in the Suppression Pool.

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

AND Control Rod insertion commenced in accordance with EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

2-E Page 44 of 67 Critical Tasks Five CS#5-After RPV water level drops to -50 inches, when RPV level cannot be restored and maintained above -180, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Maintaining adequate core cooling.

2. Cues:

Procedural compliance.

RPV level indication.

3. Measured by:

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

4. Feedback:

RPV pressure trend.

Suppression Pool temperature trend.

SRV open status indication.

2-E Page 45 of 67 Scenario Tasks EVENT TASK NUMBER KIA RO SRO 1 Main Turbine Roll RO U-047-NO-04 245000A1 .02 2.6 2.5 2 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 3 Uncoupled Control Rod RO U-085-AL-06 201003A2.02 3.7 3.8 SRO S-085-AB-02 4 Turbine Speed Control Unit Failure RO U-047-AB-02 245000A3.12 3.3 3.5 SRO 5-047-AB-Ol 5 RFPT Seal Injection Pump Trip RO U-003-AL-12 259001K6.10 2.5 2.5 6 DG C Auto Start Failure RO U-082-AL-07 264000A4.04 3.7 3.7 SRO S-000-AD-27 7 Loss of High Pressure InjectionlATWS RO U-000-EM-17 29503 1EA1.06 4.4 4.4 SRO S-000-EM-03 SRO S-000-EM-05 SRO S-000-EM- 18

2-E Page 46 of 67 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2-E 8 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) 90 Run Time (minutes) 5 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

Start Bus 1 A is out of service Unit 3 is performing 0-SR-3 .8.1 .A. 1 Verification of Offsite Power Availability to 4.16 KV Shutdown Boards every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Shutdown Board C Alternate Feeder Breaker 1624 is out of service for Breaker Swap.

Operations/Maintenance for the Shift:

Raise Reactor Power with Control Rods to meet step 18 of 2-GOI-100-1A section 5.5.

When 5 to 6 turbine bypass valves are open, complete Main Turbine Roll JAW 2-01-47 section 5.4 starting at step 11.

Units 1 and 3 are at 100% power.

Unusual Conditions/Problem Areas:

Severe Thunderstorm Warning in affect for the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

t3.

at (I

r m

r 0

m Cr, 0

z r

C,

U I 1 NI P4<

II

0 CI, 0

z

Z.

C) 0

-4 2

k

31 REACTIVITY CONTROL SYSTEMS

313 Control Rod OPERABILITY LCD 3.t3 Each control rod shall be OPERABLE APPLICABILITY: MODES I and 2.

ACTIONS

---NOTE---------------

Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control rod stucic Rod worth minimizer (RWM) may be bypassed as allowed by LCO 332.I, ControI Rod Block lnstrumentation, if required, to allow continued operation.

Al Verity stuck control rod Immediately separation criteria are met.

AND A.2 Disarm the associated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months control rod drive (CRD).

AND

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A3 Perform SR 3.1 32 and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from SR 31 3.3 for each discovery of withdrawn OPERABLE Condition A control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A4. Perform SR 311.1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

8. Two or more withdrawn 81 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stuck.

C. One or more control rods Cl -----NOTE------------.

inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B LCO 3321, if required, to allow insertion of inoperable control rod and continued operation Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND C2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

ACTIONS (continued)

CONDITION REQUiRED ACTION COMPLETION TIME D. ------NOTE------------ Di Restore compliance with 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Not applicable when BPWS, THERMAL POWER

>1O%RTP.

D..2 Restore control rod to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Two or more inoperable OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods.

E. Required Action and El Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time :Qf Condthon A, C, or D not meL OR Nine or more control rods inoperable.

3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO &8.1 The following AC electrical power sources shall be OPERA8LE:

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

b. Unit I and 2 diesel generators (DOS) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 3 DO(s) capable of supplying the Unit 34.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems Operating.

APPLICABILIrY: MODES 1,2, and 3.

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

CONDITION REQUIRED ACTION COMPLETION TIME A, One required offstte A.1 Verify power avaflability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months circuit inoperable, from the remaining OPERABLE offsite AND transmission network.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A2 Declare requIred 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from feature(s) with no otfsite discovery of no power available offsite power to inoperable when The one shutdown redundant required board concurrent feature(s) are inoperable, with inoperabftity of redundant required feature(s)

AND A3 Restore required offsite 7 days circuit to OPERABLE status. AND 14 days from discovery of failure to meet LCO B. One required Unit I and 2 .1 VeTify power availability 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months DO inoperable, from the offsite transmission network. AND Once per B hours thereafter AND (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B2 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from feature(s), supported by discovery of the inoperable Unit I and Condition B 2 DG, inoperable when concurrent with the redundant required inoperability of feature(s) are inoperable, redundant required feature(s)

AND 8,31 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Unit I and 2 DG(s) are not inoperable due to common cause failure.

OR 8.3.2 Perform SR 3.8.1.1 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE Unit I and 2 DG(s).

AND 8.4 Restore Unit I and 2 DG 7 days to OPERABLE status, AND 14 days from discovery of failure to meet LCO (continued)

ACTIONS (continuedi CONDITION REQUIRED ACTION COMPLETION TIME C. One division of 480 V Ci Restore required division 7 days load shed logic of 480 V load shed logic inoperable, to OPERABLE status.

0. One division of common Di Restore required divIsion 7 days accident signal logic of common accident inoperable. signal logic to OPERABLE status.

E. Two required offsite El Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from circuits inoperable, feature(s) inoperable discovery of when the redundant Condition E required feature(s) are concurrent with inoperable. inoperability of redundant required feature(s)

AND E.2 Restore one required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months offsite circuit to OPERABLE status.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

--NOTE--- --NOTE----

Only applicable when more Enter applicable Conditions and than one 416 kV shutdown Required Actions of LCO 3&7, board is affected, Distribution Systems

Operating, when Condition F is entered with no AC power source F. One required offsite to any 416 KV shutdown board.

circuit inoperable.

AND Fl Restore required offslte 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months circuit to OPERABLE One Unit 1 and 2 OG status, inoperable.

OR F.2 Restore Unit I and 2 DO 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to OPERABLE status,

NOTE--

Applicable when only one 416 kV shutdown board is affected,

0. One required offsite G.1 Declare the affected Immediately circuit inoperable. 4.16 kV shutdown board inoperable, AND One Unit I and 2 DG inoperable.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Two or more Unit 1 H. I Restore all but one Unit 1 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 2 DGs and 2 OG to OPERABLE inoperable, status.

Required Action and LI Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Associated Completion Time of ANQ Condition A, B, C, D, E, F, or H not met. L2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months J. One or mare required J.1 Enter LCO 3.0.3, Immediately offsite circuits and two or more Unit I and 2 DGs inoperable.

OR Two required ofisite circuits and one or more Unit 1 and 2 OGs inoperable.

OR Two divisions of 480 V load shed logic inoperable.

OR Two divisions of common accident signal logic inoperable, (continued>

ACTIONS (continued I CONDITION REQUIRED ACTION COMPLETION TIME K. One or more required K.1 Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months tram Unit 3 DGS feature(s) supported by discovery of inoperable, the inoperable UnIt 3 DC Condition K inoperable when the concurrent with redundant required inoperability of feature(s) are Inoperable, redundant required feature(s)

AND K.2 Declare affected SGT and 30 days CREVS subsystem(s) inoperable.

3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution Systems Operating LCO 3.8.7 The foflowing AC and DC electiical power distribution subsystems shall be OPERABLE:

a. Unit I and 2 4.16 KV Shutdown Boards;

b. Unit 2480 V Shutdown Boards; c, Unit 2480 V RMOV Boards 21k, 2B, 2D, and 2E;

d. Unit 1 and 2 DC Auxiliary Boards;

e. Unit DC Boards and 250 V DC RMOV Boards 2.4, 2B, and 2C; I, Unit I and 2 Shutdown Board DC Distribution Panels; and

g. Unit I and 3 AC and DC Boards needed to support equipment required to be OPERABLE by LCO 3.6.4.3, Standby Gas Treatment (SGT) System, and ICO 3.7.3, Control Room Emergency Ventilation (CREV) System.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Unit I and 2 4.16 kV ------NOTE--.----

Shutdown Board Enter applicable Conditions and inoperable. Required Actions of Condition B, C, 0, and G when Condition A results in no power source to a required 480 volt board.

Al Restore the Unit 1 and 2 5 days 4.16 kV Shutdown Board to OPERABLE status. AND 12 days from discovery of failure to meet LCO AND A2 Declare associated diesel Immediately generator inoperable.

(continued)

ACTIONS continuedI CONDITION REQUIRED ACTION COMPLETION TIME

8. One Unit 2 480 V Shutdown Board Enter Condition C when inoperable. Condition B results in no power source to 480 volt RMOV board OR 2Dor2E.

480 V RMOV Board 2A Bi Restore Board to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Inoperable. OPERABLE status.

AND 480 V RMOV Board 28 Inoperable, failure to meet LCO C. Unit 2 480 V RMOV Ci Declare the affected RI-fR Immediately Board 2D inoperable, subsystem inoperable.

OR Unit 2 480 V RMOV Board 2E inoperable.

IX One Unit 1 and 2 DG D.1 Restore Unit 1 and 2 DG 5 days Auxiliary Board Auxiliary Board to inoperable. OPERABLE status, AND 12 days from discovery of failure to meet

. LCO (continued)

ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME E. One Unit DC Board El Restore required Board or 7 days inoperable. ShUtdoWn Board DC Dlstrbution Panel to AND OR OPERABLE status, 12 days from One Unit 1 and 2 discovery of Shutdown Board DC failure to meet Distribution Panel LCO inoperable.

250 V DC RMOV Board 2A inoperable.

OR 250 V DC RMOV Board 28 inoperable.

OR 250 V DC RMOV Board 2C inoperable.

(continued)

ACTIONS Ccontinuecfl CONDON REQUI RED ACTION COMPLETION TIME F. Unit 1 and 2 416 kV Shutdown Board A and B Enter applicable conditions and inoperable, required actions of Condition B, C, 0, and G when Condition F OR results in no power source to a required 480 volt board.

UnIt 1 and 2 4.16 kV Shutdown Board C and D inoperable. Fl Restore one 4.16 kY B hours Shutdown Board to OPERABLE status. AND 12 days from discovery of failure to meet LCO G. One or more required G.l Declare the affected SGT Immediately Unit 1 or 3 AC or DC or CREV subsystem Boards inoperable, inoperable.

H. Required Action and 11,1 Be in MODE 3 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, B, D, &12 E orFnotmet.

11.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months L Two or more electrical LI Enter LCO 3.0.3. ImmedIately power distribution subsystems inoperable that result in a loss of function.

I I a I 1.3-U I I I I Reactor coolant activity exceeds 26 jicilgm dose C equivalent [.131 (Technical Specification Limits) z as determined by chemistry sample. C 0

OPERATING CONDITION m ALL z

-I 1 .2-A I I NOTE I I 1 .3-A I I I Failure of RPS automatic scram functions to bring Reactor coolant activity exceeds 300 iCi/gm dose the reactor subcritical equivalent lodine-131 as determined by chemistry AND sample.

r Manual scram or ARt (automatic or manual) was m successfuL OPERATING CONDITION:

OPERATiNG CONDITION: Mode I or 2 or 3 Mode I or2 t2-S j INOTEI I I I I Failure of automatic scram, manual scram, and cj>

ARI to bring the reactor subcritical.

tn rn m

a m

OPERATING CONDITION:

Model 1.2-G I CURVE I I I US I I I I Failure of automatic scram, manual scram, and ARI, Reactor power is above 3%

AND m Either of the following conditions exists:

. Suppression Pool temp exceeds HCTL, m Refer to Curve l2-G,

. Reactor water level can NOT be restored and maintained at or above .180 inches.

m z

OPERATING CONDITION:

Mode I or2

UNUSUAL EVENT ALERT SITE EMERGENCY GENERAL EMERGENCY N

U) c 2 C N

0 w 0 C

_J.

C

. v 0

U

.s

ØC N 0 r a.

41 oci. iZ 2C 2

0 VV V

VVC_

0<

N Dø 0<

N ft: o: N o: u.

0 0

00 0 0 C 4) a a C 0 (U (3 N4)3 NC V (3 0 v..--

V C V Z 0 0 Z 0

Z 0 0

,. I 0 N I

>s Z o

C) z

(, > 03 z

1) >

z 03 0 (3 0 C , ,C 0 0 0 N z z _o Z _0 V. <

hE DUo 0ft 0<aE

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

Examiners: Operators: ATC:

BOP:

Initial IC 104/ Unit 2 Reactor Power 70%! EECW A3 Pump tagged Out! RFPT B Out of Service Conditions:

Turnover: Remove LPRM 8-49B from bypass JAW 2-OJ-92B section 6.4, then raise power with Control Rods as directed by the RCP.

Event Event No. Maif. No. Type* Event Description N-BOP 1 Remove an LPRM from Bypass 8-49B N-SRO 2 Raise Power with Control Rods RPIS Position Failure rod 14-3 5 with drift in on insert rd r

1 35 C-BOP 4 sw03m EECW Pump Trip TS-SRO C-BOP 5 ms05h MSJV Partial Closure TS..SRO 6 fw26a!b Feedwater Flow Transmitter failures 7 mcO4 M-ALL Degrading Vacuum, ATWS with out MSJVs 8 iaO2 C Loss of Drywell Control Air 9 rdOl C CRD Pump Trip

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

2-F Page 1 of 55 Console Operator Instructions A. Scenario File Summary

1. File: batch and trigger files for scenario 2-F Batch NRC/110202 Imfsw07b Bat atws70 Imf rd0ia (e2 120)

Trge 1 NRC/msivd = zdihs0 1 52a[1] .eq. 1 Trgel = bat NRC/i 10202-3 br xa555b23 alarm off Trge3 NRC/singleelement = zdihs466a.eq. 1 Trge3 = bat NRC/i 10202-4 Imf ia02a(e2 15)100100 Imf ia02b (e2 60) 100 30 0 Batch NRC/1102-1 br zlohs466a off br zlohs466b on Imf fw26a (none 0) 0 Imffw26b (none 60) 100 30 0 Batch NRC/1102-2 Imf th27e br zlohs0l52a[2] on Imf ms05h br za0fi464 1.6 Batch NRC/1102-3 Dor zlohs0 1 52a[2j Dor zaofi464 Batch NRC/1102-4 Dor zlohs466a Dor zlohs466b

2-F Page 2 of 55 Pref file F3 imfrd25 F4 imfrdO7ri435 F5dmf rdO7 1435 F6 imfsw03m F7 bat NRC/i 10202 F8 bat NRC/1102-1 F9 bat NRC/1102-2 FlO dmfth27e Fl 1 F12 trg e2 modesw Shift fi imfmc04 100 Shift f4 nirfrd06 open Shift f5 bat appOif Shift f6 bat app02 Shift 17 mrf rdO6 close Shift f8 bat sdv Console Operator Instructions Scenario 2-F DESCRIPTION/ACTION Simulator Setup manual Reset to IC 104 Simulator Setup Load Batch RestorePref NRC/i 10202 Simulator Setup manual Tag Out EECW Pump A3 Simulator Setup manual F7 and F12 Simulator Setup Verify file loaded RCP required (70% 85% with control rods and flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12

2-F Page 3of55 BOP will remove LPRM 8-49B from bypass JAW 2-OJ-92B section 6.4.

ATC will raise Reactor Power with control rods as directed by the Reactivity Control Plan.

During power ascension Control Rod 14-3 5 will experience an RPIS position failure. The crew will respond JAW ARPs and 2-AOI-85-4. The ATC will insert Control Rod 14-3 5 one notch to establish position indication. After Control Rod 14-3 5 is inserted it will begin to drift in; the ATC will respond JAW 2-AOI-85-5 and insert the control rod to position 00.

EECW D3 Pump will trip and the standby EECW Pump B3 will fail to auto start, the BOP will respond JAW ARPs and start EECW Pump B3 to EECW flow to the north header.

The SRO will evaluate Technical Specification 3.7.2 and Condition A is entered.

Outboard MSIV D will drift closed; the crew will respond JAW 2-AOI-1-3. The ATC will lower Reactor Power to less than 66% and the BOP will fully close Outboard MSIV D. The SRO will evaluate Technical Specification 3.6.1.3 and Condition A is entered.

Feedwater Flow Transmitters will fail; the crew will respond JAW ARPs and 2-AOJ-3 -1; the ATC will report that Feedwater Level Control transferred to single element and will transfer to single element. Reactor Level will stabilize after the initial transient.

Vacuum will begin to degrade and the crew will respond JAW 2-AOI-47-3, the crew will insert a manual Reactor scram prior to the Main Turbine trip. An ATWS will exist and the crew will enter 2-EOI-1 and 2-C-S.

After the scram, an airline break will occur in the drywell causing MSIV closure and transition to SRVs for pressure control and RCIC for level control. Until the crew performs Appendix 8G, SRV operation will degrade due to the loss of air.

CRD Pump 2A will trip and the ATC will start CRD Pump lB in order to insert control rods.

The crew will maintain directed level and pressure bands, insert all control rods and enter EOI-2 and place RHR in Suppression Pool Cooling.

The Emergency Classification is 1.2-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.

Controls Rods are being inserted Reactor Level is being maintained in directed level band

2-F Page 4 of 55 Simulator Event Guide:

Event 1 Normal: Remove LPRM 8-49B from bypass lAW 2-OI-92B section 6.4 SRO Directs LPRM 8-49B un-bypassed.

BOP Removes LPRM 8-49B from bypass lAW 2-OI-92B section 6.4.

6.4 Returning an LPRM to Operate From a Bypassed Condition

[1] REVIEW all precautions and limitations. REFER TO Section 3.0.

[2] REFERENCE Illustration 4 to find the APRM/LPRM Channel associated with the desired LPRM to be returned to normal.

[3] At Panel 2-9-14, DEPRESS any sofikey to illuminate the display on the desired APRM/LPRM channel chassis.

[4] DEPRESS the ETC sofikey until BYPASS SELECTIONS illuminates on the bottom row of the display.

[5] DEPRESS BYPASS SELECTIONS softkey, enter the password, and DEPRESS ENT.

[6] SELECT the desired LPRM to be returned to service by using the left or right arrows on the softkey board until the inverse video illuminates the correct LPRM.

[7] DEPRESS the OPERATE softkey.

[8] CHECK the BYP/HV OFF is replaced by OPERATE below the selected LPRM.

[9] DEPRESS EXIT softkey to return display to the desired bargraph.

[10] VERIFY, as a result of returning this LPRM to operate, that any alarms received on Panel 2-9-5 or on the APRM/LPRM channel are reset.

2-F Page 5 of 55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase.

Direct 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 5850 RPM.

ATC Raise Power with Control Rods per 2-01-85, section 6.6.

6.6.1 Initial Conditions Prior to Withdrawing Control Rods

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

CRD POWER, 2-HS-85-46 in ON.

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

6.6.2 Actions Required During and Following Control Rod Withdrawal

[4] OBSERVE the following during control rod repositioning:

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

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

[5] ATTEMPT to minimize automatic RBM Rod Block as follows:

STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 2-9-5 and PERFORM Step 6.6.2[6].

[6] fi? Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a REM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:

[6.1] PLACE CRD POWER, 2-HS-85-46 in the OFF position to deselect the Control Rod.

[6.2] PLACE CRD POWER, 2-HS-85-46, in the ON position.

2-F Page 6 of 55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2XS-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 Pennit light ILLUMINATED.

[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH, and RELEASE.

[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.

[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:

[6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH, and RELEASE.

[6.2] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does NOT alarm.

[6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes.

[6.4] IF Control Rod Coupling Integrity Check fails, TI[EN REFER TO 2-AOI-85-2.

2-F Page 7of55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal

[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 2-XS.-85-40.

[2] OBSERVE the following for the selected control rod:

CRD ROD SELECT pushbutton is brightly ILLUMINATED.

White light on the Full Core Display ILLUMiNATED.

Rod Out Permit light ILLUMINATED.

[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.

[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.

[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)

[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[5.2] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[5.3] WhEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 2-HS-85-47 and CRD CONTROL SWITCH, 2-HS-85-48.

[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

2-F Page 8 of 55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal (Continued)

[5.5] WhEN control rod settles into the intended notch, THEN CHECK the following.

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does NOT alarm.

[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.

[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)

[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRJDE.

[6.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[6.3] MAINTAIN the CR1) Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position, with the control rod at position 48.

[6.4] CHECK control rod coupled by observing the following:

Four rod display digital readout and the full core display digital readout and background light remain illuminated.

CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does not alarm.

[6.5] RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48.

2-F Page 9 of 55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.

[6.7] IF control rod coupling integrity check fails, THEN REFER TO 2-AOI-85-2.

[7] IF continuously withdrawing the control rod to position 48; and the control rod coupling integrity check will be performed after the CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48 are to be released, THEN:

PERFORM control rod coupling integrity check as follows (otherwise N/A):

[7.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.

[7.2] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.

[7.3] WhEN position 48 is reached, THEN:

RELEASE CRD NOTCH OVERRiDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48.

[7.4] VERIFY control rod settles into position 48.

[7.5] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[7.6] CHECK control rod coupled by observing the following:

Four rod display digital readout AND full core display digital readout AND background light will remain illuminated.

CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window 14) does NOT alarm.

2-F Page 10of55 Simulator Event Guide:

Event 2 Reactivity: Raise Power with Control Rods ATC [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 CRD POWER, 2-HS-85-46, in OFF.

[1.2] PLACE CRD POWER, 2-HS-85-46, in ON.

V Ig_ J4c ie

2-F Page 11 of 55 Simulator Event Guide:

Event 3 Instrument: RPIS Position Failure Control Rod 14-3 5 ATC Report Control Rod Drift Alarm 5A-28, reports no control rods drifting.

Reports loss of position indication on Control Rod 14-35.

SRO Enter 2-AOI-85-4 Loss of RPIS.

ATC 4.1 Immediate Actions

[1] STOP all control rod movement.

SRO 4.2 Subsequent Actions NOTE Reference TRM 3.3.5, RPIS Indicated Channel Operability, for applicable 7 or 30 day LCO relating to an inoperable RPIS indication.

[1] IF control rod movement is required with a Total loss of RPIS, THEN MANUALLY SCRAM reactor.

[2] NOTIFY the Operations Superintendent and Reactor Engineer for actions to be taken in a timely manner.

SRO [9] IF unable to restore position indication for an individual control rod or rods, ThEN NOTIFY Reactor Engineer and DETERMINE additional corrective action. Control Rods may be moved to an Operable Position Indication as a means of position verification (REFER TO Tech Spec Bases SR 3.1.3.1). As a minimum, rod position will be verified, preferably with an independent position indication or other method.

- wi wj I 1. cv r F WW i iyçi Jvç onenatio QrQrecQnW1 SRO Direct ATC to insert Control Rod 14-35 one notch to attempt to establish position indication.

ATC Insert Control Rod 14-35 to position 46.

SRO Evaluate Technical Requirements Manual 3.3.5.

Information LCO for TRM 3.3.5 Condition A and C from table 3.3.5-1.

4K i

2-F Page 12of55 Simulator Event Guide:

Event 3 Instrument: RPIS Position Failure Control Rod 14-3 5 ATC Report Control Rod Drift Alarm 5A-28, reports Control Rod 14-35 drifting in.

SRO Enter 2-AOI-85-5 Rod Drift In.

ATC 4.1 Immediate Actions

[1] IF multiple rods are drifting into core, THEN MANUALLY SCRAM Reactor.

Refer to 2-AOI-100-1.

SRO 4.2 Subsequent Actions

[1] IF a Control Rod is moving from its intended position without operator actions, THEN INSERT the Control Rod to position 00 using CONTINUOUS IN.

[2] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

[3] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 2-AOI-100-l.

ATC Inserts Control Rod 14-35 to position 00.

[4] CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).

[5] ADJUST control rod pattern as directed by Reactor Engineer and CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).

ATC [6] IF CRD Cooling Water Header DP is excessive and causing the control rod drift, THEN:

[7] VERIFY scram pilot air header aligned to scram inlet and outlet valves.

Crew Dispatch AUO to check scram valves.

s eac sAUi eøxe air are cw:Ti

2-F Page 13of55 Simulator Event Guide:

Event 4 Component: EECW Pump D3 Trip BOP Respond to alarms 20A-21 and 23D-26.

23D-26 4160V SD BD D MOTOR OL or TRIP Overload or trip out, on any one of the following:

CS pump 1D, 2D, RHR pump 1D, 2D, RJIRSW pump D2, D3 A. CHECK control room for white light illuminated on effected equipment.

B. DISPATCH personnel to check:

1. Relays at associated electrical bd.

2. Equipment for abnormal conditions, relay targets, smell, burned paint, breaker.

20A-21EECW NORTH HDR DG SECTION PRESS LOW B. CHECK Panel 2-9-3 for status of North header pump(s) breaker lights and pump motor amps normal.

C. NOTIFY UNIT SUPERVISOR, Unit 1 and Unit 3.

D. START standby pump for affected header. REFER TO 0-01-67.

8.11 Recovering from an EECW Pump Trip

[1] VERIFY < 25 minutes has elapsed since the EECW pump trip and header pressure

> 0 psig.

[3] IF the south header pump has tripped, THEN:

[3.1] START desired RHRSW Pump using one of the following:

RHRSW PUMP D3(B3) EECW SOUTH HDR, 0-HS 94A!2(88A!2) on Unit 2.

[4] For the EECW(RHRSW) pump(s) started, PERFORM the following:

VERIFY running current is less than 53 amps.

VERIFY locally, Pump breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged.

VERIFY Pump upper and lower motor bearing oil level is in the normal operating range.

NOTIFY Chemistry of running RHRSW (EECW) pump(s).

BOP Start EECW Pump B3.

SRO Evaluate Technical Specification 3.7.2.

2-F Page 14 of 55 Simulator Event Guide:

Event 4 Component: EECW Pump D3 Trip SRO Evaluate Technical Specification 3.7.2.

Condition A: One required EECW pump inoperable.

Required Action A. 1: Restore the required EECW pump to OPERABLE status.

Completion Time: 7 days IE I

2-F Page 15of55 Simulator Event Guide:

Event 5 Component: MSIV Partial Closure ATC Respond to alarm 5B-1 8 MAIN STEAM LiNE CH A FLOW HIGH.

5B-l 8 MAiN STEAM LINE CH A FLOW HIGH A. VERIFY alarm by checking main steam flow indicators.

B. IF alarm is valid on any steam line, THEN MANUALLY SCRAM Reactor and PLACE Rx Mode Sw. in Shutdown and CLOSE MSIVs.

C. IF any flow indicators are low, THEN CHECK all MSIVs open.

D. REFER TO 2-AOI-l-3.

E. REFER TO Tech Spec Table 3.3.6.1-1.

ATC Report Steam flow in D line is lower than A, B and C lines.

ATC/BOP Report Outboard MSIV D 1 -52, indicates partially closed.

SRO Enter 2-AOl-i -3, MS1V Closure at Power.

4.1 Immediate Action None 4.2 Subsequent Action

[1] IF any EOI entry condition is met, THEN (Otherwise N/A):

ATC [2] LOWER reactor power with recirc flow and insert control rods as necessary, when directed by the Reactor Engineer/Unit Supervisor, to ensure that rated steam line flow (3.54 x 106 lbm!hr) is NOT exceeded; as indicated on Main Steam Line Flow Indicators.

ATC/BOP [3] IF an MSIV is partially closed, THEN:

[3.1] LOWER reactor power to 66%.

[3.2] PLACE the associated MSIV control switch to CLOSE.

ATC Lower Power to 66%.

BOP PLACE the Outboard MSIV D, 1 -52 control switch, to CLOSE.

2-F Page 16 of 55 Simulator Event Guide:

Event 5 Component: MSIV Partial Closure ATC Respond to alarm 5B-l 8 MAiN STEAM LINE CII A FLOW HIGH.

5B-18 MAIN STEAM LINE CH A FLOW HIGH A. VERIFY alarm by checking main steam flow indicators.

B. IF alarm is valid on any steam line, THEN MANUALLY SCRAM Reactor and PLACE Rx Mode Sw. in Shutdown and CLOSE MSIVs.

C. IF any flow indicators are low, THEN CHECK all MSIVs open.

D. REFER TO 2-AOI-1-3.

E. REFER TO Tech Spec Table 3.3.6.1-1.

ATC Report Steam flow in D line is lower than A, B and C lines.

ATC/BOP Report Outboard MSIV D 1-52 indicates partially closed.

SRO Enter 2-AOI-l-3, MSIV Closure at Power.

4.1 Immediate Action None 4.2 Subsequent Action

[1] IF any EOI entry condition is met, THEN (Otherwise N/A):

ATC [2] LOWER reactor power with recirc flow and insert control rods as necessary, when directed by the Reactor Engineer/Unit Supervisor, to ensure that rated steam line flow (3.54 x 106 lbmlhr) is NOT exceeded; as indicated on Main Steam Line Flow Indicators.

ATC/BOP [3] IF an MSIV is partially closed, THEN:

[3.1] LOWER reactor power to 66%.

[3.2] PLACE the associated MSIV control switch to CLOSE.

ATC Lower Power to 66%.

BOP PLACE the Outboard MSIV D 1-52 control switch to CLOSE.

2-F Page 17 of 55 Simulator Event Guide:

Event 5 Component: MSIV Partial Closure SRO Evaluate Technical Specification 3.6.1.3.

Condition A: NOTE Only applicable to penetration flow paths withtwo PCIVs.

One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits Required Action A. 1: Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

Completion Time: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main steam lines Required Action A.2: NOTE Isolation devices in high radiation areas may be verified by use of administrative means.

Verif the affected penetration flow path is isolated.

Completion Time: Once per 31 days for isolation devices outside primary containment

&i I [i1 3i

2-F Page 18of55 Simulator Event Guide:

Event 6 Instrument: Feedwater Flow Transmitter Failures Respond to alarm 6C-14 RFWCS INPUT FAILURE.

A. VERIFY RFWCS continues to maintain Reactor Water level.

B. IDENTIFY badlinvalid signal by checking Control Room instrumentation and/or ICS. REFER TO ATTACHMENT 1, on next page, for list of RFWCS ATC instrumentation. REFER TO ICS RX FW LVL CONTROL SYS display (FWLCS).

C. REQUEST assistance from Site Engineering.

D. BYPASS the bad/invalid signal with Unit Supervisor approval.

ATC Report Feedwater Flow signal has failed for FW Line A.

ATC Report FW Line B Feedwater Flow signal failing.

SRO Enter 2-AOI-3-l, Loss of Feedwater or Reactor Water Level High/Low.

4.1 Immediate Actions None 4.2 Subsequent Actions

[2] IF Feedwater Flow signal fails (FI-3-78A, FI-3-78B), THEN PERFORM the following:

A. With SROs permission, REFER TO 2-01-3 and BYPASS failed Feedwater Flow Instrument in Unit 1 &2 Computer Room; or Unit 2 Aux Instrument Room.

[2.1] IF both Feedwater Flow Instruments fail, THEN VERIFY level control transfers to SINGLE ELEMENT.

ATC Verifies Reactor Level control in single element, level control failed to transfer to single element; Operator depresses single element pushbutton to transfer.

[6] IF Reactor Water Level continues to rise, THEN TRIP RFP, as necessary.

[7] IF RFPs in automatic control, THEN VERIFY 2-LIC-46-5 lowers flow of operating RFPs.

ATC Verifies RFPTs maintain water level.

bI1 tI

2-F Page 19of55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP Respond to alarm 53-14 OG HOLDUP LINE iNLET FLOW HIGH.

ATC Report degrading condenser Vacuum.

SRO Enter 2-AOI-47-3, Loss of Condenser Vacuum.

4.1 Immediate Actions None 4.2 Subsequent Actions

[1] IF ANY EOI entry condition is met?_THEN:

[2] IF unable to maintain hotwell pressure below -25 inches Hg, as indicated on 2-XR-2-2, with Reactor power_less_than_30%,_THEN_TRIP the_main turbine.

[4] REDUCE reactor power in an attempt to maintain condenser vacuum.

SRO Determines a trigger value for Reactor Scram prior to Turbine Trip; at 25 inches.

ATC Insert Reactor Scram when directed; and place mode switch in shutdown. Report ATWS and initiate first channel of ARI.

I SRO Enter 2-EOI-, RPV Control.

SRO EOI-1 (Reactor Pressure)

Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? - NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate ?- NO IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? - NO IF RPV water level cannot be determined? - NO Is any MSRV Cycling? - YES IF Steam cooling is required? NO-IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?-NO

2-F Page 20 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO 2-EOI-1 (Reactor Pressure)

IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? YES.

THEN crosstie CAD to Drywell Control Air, Appendix 8G.

IF Boron injection is required? NO SRO Direct a Pressure Band of 800 to 1000 psig, Appendix 1 1A.

ATC/BOP Maintain directed pressure band, JAW Appendix 1 1A.

BOP Crosstie CAD to Drywell control air, JAW Appendix 8G.

SRO IF Main Steam Relief Valve Air Accumulator Low annunciator, (XA-55-3D-l 8) is in alarm, THEN: place each MSRV Control Switch in Close/Auto AND Place MSRV Auto Actuation Logic Inhibit XS-1 -202 to Inhibit.

ATC/BOP Places XS-l-202 to inhibit.

EOI- 1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems: HPCI Appendix 1 1C, RCIC Appendix 1 1B, RFPTs on SRO minimum flow Appendix 1 iF, Main Steam System Drains Appendix 1 1D, Steam Seals Appendix 11 G, SJAEs Appendix 11 G, Off Gas Preheater Appendix 11 G, RWCU Appendix 11 E.

2-F Page 21 of55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Pressure Control JAW Appendixl 1A, RPV Pressure Control SRVs

1. IF Drywell Control Air is NOT available, THEN:

EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.

2. IF Suppression Pool level is at or below 5.5 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. 2-PCV-l-179 MN STM LINE A RELIEF VALVE

b. 2-PCV-l-180 MN STM LINED RELIEF VALVE.

c. 2-PCV-l-4 MN STM LINE A RELIEF VALVE

d. 2-PCV-l-31 MN STM LiNE C RELIEF VALVE

e. 2-PCV-l-23 MN STM LINE B RELIEF VALVE

f. 2-PCV-l-42 MN STM LINED RELIEF VALVE

g. 2-PCV-l-30 MN STM LINE C RELIEF VALVE

h. 2-PCV-1-19 MN STM LINE B RELIEF VALVE.

i. 2-PCV-l-5 MN STM LINE A RELIEF VALVE.

j. 2-PCV-l-41 MN STM LINED RELIEF VALVE

k. 2-PCV-l-22 MN STM LINE B RELIEF VALVE

1. 2-PCV-l-l8 MN STM LINE B RELIEF VALVE

m. 2-PCV-l-34 MN STM LINE C RELIEF VALVE

2-F Page 22 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Pressure Control lAW AppendixilA, RPV Pressure Control SRVs

3. IF Drywell Control Air header, supplied from CAD System A; shows indications of being depressurized, as determined by Appendix 8G, THEN:

OPEN MSRVs supplied by CAD System B, using the following sequence to control RPV pressure, as directed by SRO:

4. IF Drywell Control Air header, supplied from CAD System B; shows indications of being depressurized, as determined by Appendix 8G, THEN:

OPEN MSRVs supplied by CAD System A, using the following sequence to control RPV pressure, as directed by SRO:

6. IF BOTH Drywell Control Air headers are depressurized, THEN PERFORM the following as directed by EOI-l, RPV Control, RC/P Section:

PLACE each MSRV control switch in CLOSE/AUTO, and PLACE 2-XS-1-202, MSRV AUTO ACTUATION LOGIC INHIBIT, to INHIBIT.

AND

MINIMIZE MSRV cycling by using sustained openings for RPV depressurization.

EOI- 1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems: HPCI Appendix 11C, RCIC Appendix 1 1B, RFPTs on SRO minimum flow Appendix 1 iF, Main Steam System Drains Appendix liD, Steam Seals Appendix 11 G, SJAEs Appendix 11 G, Off Gas Preheater Appendix 11 G, RWCU Appendix liE.

ATC/BOP Augment RPV Pressure Control, if directed by SRO.

2-F Page 23 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP Crosstie CAD to Drywell control air, JAW Appendix 8G.

1. OPEN the following valves:

O-FCV-84-5, CAD SYSTEM A N2 SHUTOFF VALVE ( Panel 9-54)

O-FCV-84-16, CAD SYSTEM B N2 SHUTOFF VALVE (Panel 9-55).

2. VERIFY 0-PI-84-6, N2 VAPORIZER A OUTLET PRESSURE, and 0-PI-84-17, N2 VAPORIZER B OUTLET PRESSURE, indicate approximately 100 psig (Unit 1, Panel 9-54 and 9-55).

3. PLACE keylock switch 2-HS-84-48, CAD A CROSS TIE TO DW CONTROL AR, in OPEN (Unit 2, Panel 9-54).

4. CHECK OPEN 2-FSV-84-48, CAD A CROSS TIE TO DW CONTROL AIR, (Unit 2, Panel 9-54).

5. PLACE keylock switch 2-HS-84-49, CAD B CROSS TIE TO DW CONTROL AIR, in OPEN (Unit 2, Panel 9-55).

6. CHECK OPEN 2-FSV-84-49, CAD B CROSS TIE TO DW CONTROL AIR (Unit 2, Panel 9-55).

7. CHECK MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 2-PA-32-31, alarm cleared (2-XA-5 5-3D, Window 1 8).

8. IF MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 2-PA-32-31, annunciator is or remains in alarm (2-XA-55-3D, Window 18), THEN DETERMINE which Drywell Control Air header is depressurized as follows:

a. DISPATCH personnel to Unit 2, RB, El 565 fi, to MONITOR the following indications for low pressure:

b. MONITOR 0-FI-84-7(18), CAD LINE A(B) N2 FLOW, on Unit 1, Panel 1-9-54(55) for high flow.

c. MONITOR inboard MSIV indication status for valves drifting closed.

9. iF Drywell Control Air header supplied from CAD System A shows indications of being depressurized, THEN CLOSE the following valves:

0-FCV-84-5, CAD SYSTEM A N2 SHUTOFF VALVE (Panel 9-5 4)

2-FSV-84-48, CAD A CROSS TIE TO DW CONTROL AIR (Panel 9-54).

10. IF Drywell Control Air header supplied from CAD B shows indications of being depressurized, THEN CLOSE the following valves:

0-FCV-84-16, CAD SYSTEM B N2 SHUTOFF VALVE (Panel 9-55)

2-FSV-84-49, CAD B CROSS TIE TO DW CONTROL AIR (Panel 9-55).

iv i7H fl

2-F Page 24 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO EOI-l (Reactor Level)

Monitor and Control Reactor Level.

Verify as required PCIS isolations group (1,2 and 3), ECCS and RCIC, Directs group 2 and 3 verified.

ATC/BOP Verifies Group 2 and 3 isolation.

SRO IF it has not been determined that the reactor will remain subcritical, THEN Exit RC/L; ENTER C5 Level / Power Control.

Is Emergency Depressurization is required? NO -

RPV Water level cannot be determined? NO The reactor will remain subcritical without Boron under all conditions? NO-PC water level cannot be maintained below 105 feet OR Suppression Chamber pressure cannot be maintained below 55 psig? NO -

S#3 SRO Directs ADS Inhibited.

CS#3 ATC/BOP Inhibits ADS.

SRO Is any Main Steam Line Open?- NO

2-F Page 25 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO C5 Level / Power Control Crosstie CAl) to DW Control Air, if necessary (Appendix 8G).

IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND a MSRV is open or cycling OR drywell pressure is above 2.4 psig AND RPV water level is above -162 inches? NO Is Reactor Power above 5% ?- YES Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC (Appendix 4).

WhEN RPV Level drops below -50 inches; THEN Continue:

CS#2 Direct Terminate and Prevent lAW Appendix 4.

IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND a MSRV is open or cycling OR drywell pressure is above 2.4 psig AND RPV water level is above -162 inches IF YES?

Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC; irrespective of any consequent reactor power or reactor water level oscillations.

WHEN RPV Level drops below -50 inches and any of the following exist:

Power drops below 5% OR

. All MSRVs remain closed and DW pressure remains below 2.4 psig OR

Water level reaches -162 inches THEN Continue:

CS#2 Direct Terminate and Prevent, lAW Appendix 4.

ATC/BOP Terminate and Prevent lAW Appendix 4 CS#2 BOP/ATC Appendix 4

1. PREVENT injection from HPCI by performing the following:

a. IF HPCI Turbine is NOT at zero speed, ThEN PRESS and HOLD 2-HS-73-l8A, HPCI TURBINE TRIP push-button.

b. WHEN HPCI Turbine is at zero speed, THEN PLACE 2-HS 47A, HPCI AUXIIJARY OIL PUMP control switch in PULL TO LOCK and RELEASE 2-HS-73-18A, HPCI TURBiNE TRIP push-button.

3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.

2-F Page 26 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs BOP/ATC Terminate and Prevent JAW Appendix 4 CS#2 Appendix 4 (continued)

4. PREVENT injection from LPCI SYSTEM I by performing the following:

NOTE Injection maybe prevented by performing EITHER step 4.a or step 4.b.

a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 2-I-JS-74-155A, LPCI SYS I OUTBD NJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE.

5. PREVENT injection from LPCI SYSTEM II by performing the following:

NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV.

BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-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.

2-F Page 27 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs CS#2 BOP/ATC Terminate and Prevent JAW Appendix 4 Appendix 4 (continued)

c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:

2-FCV-3-19, RFP 2A DISCHARGE VALVE

2-FCV-3-12, RFP 2B DISCHARGE VALVE

2-FCV-3--5, REP 2C DISCHARGE VALVE

2-LCV-3-53, RFW START-UP LEVEL CONTROL

d. TRIP RFPTs as necessary to prevent injection by DEPRESSING the following push-buttons:

2-HS-3-l25A, RFPT 3A TRIP

2-HS-3-l5lA, RFPT 3B TRIP

2-HS-3-176A, RFPT 3C TRIP.

SRO WhEN RPV Level drops below -50 inches ThEN Continue:

OR WhEN RPV Level has dropped below -50 inches AND Power is below 5% OR Reactor Level reaches -162 inches, THEN Continue:

Directs a Level Band with RCIC and HPCI.

ATC/BOP Maintain Directed Level Band with RCIC, Appendix SC and HPCI, Appendix SD.

2-F Page 28 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C

3. VERIFY RESET and OPEN 2-FCV-71-9, RCIC TURB TRIP/THROT VALVE RESET.

4. VERIFY 2-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 600 gpm.

5. OPEN the following valves:

. 2-FCV-71-39, RCIC PUMP iNJECTION VALVE

. 2-FCV-71-34, RCIC PUMP MIN FLOW VALVE

. 2-FCV-71-25, RCIC LUBE OiL COOL]NG WTR VLV.

6. PLACE 2-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.

7. OPEN 2-FCV-71-.8, RCIC TURBINE STEAM SUPPLY VLV, to start RCIC Turbine.

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at 600 gpm.

c. 2-FCV-71-40, RCIC Testable Check Vlv, opens by observing 2-ZI 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.

2-F Page 29 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs ATC/BOP Maintain Directed Level Band with HPCI, Appendix 5D

4. VERIFY 2-IL-73-18B, HPCI TURBINE TRIP RX LVL HIGH, amber light extinguished.

5. VERIFY at least one SGTS train in operation.

6. VERIFY 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5,000 gpm.

7. PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in START.

8. PLACE 2-HS-73-1OA, I{PCI STEAM PACKiNG EXHAUSTER, handswitch in START.

9. OPEN the following valves:

2-FCV-73-30, HPCI PUMP M]N FLOW VALVE

2-FCV-73-44, HPCI PUMP INJECTION VALVE.

10. OPEN 2-FCV-73-16, HPCI TURBiNE STEAM SUPPLY VLV, to start HPCI Turbine.

11. CHECK proper HPCI operation by observing the following:

a. HPCI Turbine speed accelerates above 2400 rpm.

b. 2-FCV-73-45, RPCI Testable Check Vlv, opens by observing 2-ZI-73-45A, DISC POSITION, red light illuminated.

c. HPCI flow to RPV stabilizes and is controlled automatically at 5000 gpm.

d. 2-FCV-73-30, HPCI PUMP 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 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in AUTO.

14. ADJUST 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

2-F Page 30 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO EOI-l (Power Control)

Monitor and Control Reactor Power.

Will the reactor will remain sub subcritical without boron under all conditions? NO Is the reactor subcritical and No boron has been injected?- NO Verify Reactor Mode Switch in Shutdown.

Initiate ART.

ATC Initiates ART.

SRO Verify Recirc Runback ( pump speed 480 rpm).

ATC Verifies Recirc Runback.

SRO Is Power above 5%? YES -

Directs tripping Recirc Pumps.

ATC Trips Recirc Pumps.

CS#1 SRO Before Suppression Pool temperature rises to 110°F, continue:

CS#1 Boron injection is required.

CS#1 ATC/BOP Initiate SLC, JAW Appendix 3A.

SRO Directs ART Reset Appendix 2.

CS#1 Insert Control Rods Using one or more of the following methods:

Appendix iF

Appendix 1D iVo CS#1 ATC Inserts Control Rods, JAW Appendix 1D and iF.

2-F Page 31 of55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs CS#1 ATC Insert Control Rods, JAW 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 UNTTh the following annunciators clear:

WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 1)

EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 29).

5. DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER SHUTOFF.

6. WhEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ART.

7. CONTINUE to perform Steps 1 through 6, UNTIL ANY of the following exists:

ALL control rods are fully inserted, OR

NO inward movement of control rods is observed, OR

SRO directs otherwise.

DEed 8p58Wdéii. ftf:d6oppn)J

2-F Page 32 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs CS#1 ATC Insert Control Rods lAW Appendix 1D Reports Trip of CRD 2A and Start CRD Pump 1B, lAW 2-AOI-85-3

[1] IF operating CRD pump has failed AND standby CRD pump is available, THEN PERFORM the following at Panel 2-9-5:

[1.1] PLACE CR1) SYSTEM FLOW CONTROL, 2-FIC-85-1 1, in MAN at minimum setting.

[1.2] START associated standby CRD Pump using one of the following:

CRD PUMP lB. using 2-HS-85-2A.

[1.3] IF CRD Pump lB was started, THEN OPEN CR1) PUMP lB DISCH TO U2, using 2-HS-85-8A

[1.4] ADJUST CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, to establish the following conditions:

CRD CLG WTR HDR DP, 2-PDI-85-1 8A, approximately 20 psid.

CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, between 40 and 65 gpm.

[1.5] BALANCE CR1) SYSTEM FLOW CONTROL, 2-FIC-85-11, AND PLACE in AUTO or BALANCE.

1. VERIFY at least one CRD pump in service.

2. IF Reactor Scram or ART CANNOT be reset, THEN DISPATCH personnel to CLOSE 2-SHV-085-0586, CHARGiNG WATER SHUTOFF (RB NE, El 565).

3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4. BYPASS Rod Worth Minimizer.

5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:

a. SELECT control rod.

b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.

c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.

6. WhEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 2-SHV-085-0586, CHARGING WATER SHUTOFF

2-F Page 33 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs CS#1 BOP/ATC Initiate SLC JAW Appendix 3A

1. UNLOCK and PLACE 2-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B position.

2. ChECK SLC System for injection by observing the following:

Selected pump starts, as indicated by red light illuminated above pump control switch.

Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished.

SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 20).

. 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

System flow, as indicated by 2-IL-63-l 1, SLC FLOW, red light illuminated on Panel 2-9-5.

. SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 2-9-5 (2-XA-55-5B, Window 14).

3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.

4. VERIFY RWCU isolation by observing the following:

. RWCU Pumps 2A and 2B tripped.

. 2-FCV-69-1, RWCU INBD SUCT ISOLATION VALVE closed.

. 2-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed.

2-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.

5. VERIFY ADS inhibited.

6. MONITOR reactor power for downward trend.

7. MONITOR 2-LI-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.

2-F Page 34 of 55 Simulator Event Guide:

Event 7 Major: ATWS without MSIVs SRO ENTER 2-EOI-2, Primary Containment Control EOI-2 (Drywell Temperature)

SRO Monitor and Control DW Temp Below 160°F using available DW Cooling.

Can Drywell Temp Be Maintained Below 160°F? YES -

SRO EOI-2 (Primary Containment Hydrogen)

If PCIS Group 6 isolation exists? YES THEN DIRECTS:

1. Place analyzer isolation bypass keylock switches to bypass.

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.

BOP 1. Place analyzer isolation bypass keylock switches to bypass.

2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.

SRO EOI-2 (Suppression Pool Temperature)

Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary (Appendix 1 7A)

Can Suppression Pool Temperature Be Maintained Below 95°F? NO Operate all available Suppression pool cooling, using only RHR Pumps not required to assure adequate core cooling by continuous injection, Appendix 1 7A.

ATC/BOP Place an RHR System in Pool Cooling, when directed JAW Appendix l7A.

SRO Before Suppression Pool Temperature rises to 110°F Continue in EOJ-1 RPV Control Can Suppression Pool temperature and level be maintained within a safe area of curve 3? -

YES SRO The Emergency Classification is 1.2-S.

2-F Page 35 of 55 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Inj ectionlATWS SRO EOI-2 (Suppression Pool Level)

Monitor and Control Suppression Pool Level between -1 inch and -6 inches, (Appendix 18).

Can Suppression Pool Level be maintained above -6 inches? YES Can Suppression Pool Level be maintained below -1 inch? YES SRO EOI-2 (Primary Containment Pressure)

Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary, (Appendix 12)

SRO Can Primary Containment pressure be maintained below 2.4 psig? YES SRO The Emergency Classification is 1.2-S.

2-F Page 36 of 55 Simulator Event Guide:

Event 7 Major: Loss of High Pressure Injection!ATWS ATC Place Suppression Pool Cooling in service, JAW Appendix 1 7A.

IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock AS NECESSARY:

PLACE 2-HS-74-155A, LPCI SYS I OUTBD JNJ VLV BYPASS SEL in BYPASS.

PLACE 2-HS-74-155B, LPCI SYS II OUTBD TNJ VLV BYPASS SEL in BYPASS.

2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:

a. VERIFY at least one R1{RSW pump supplying each EECW header.

b. VERIFY RHRSW pump supplying desired RIW Heat Exchanger(s).

c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

2-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

2-FCV-23-46, RHR HX 2B RHRSW OUTLET VLV

2-FCV-23-40, RHR I{X 2C RHRSW OUTLET VLV

2-FCV-23-52, RHR HX 2D RRRSW OUTLET VLV.

d. IF Directed by SRO, THEN PLACE 2-XS-74-122(130), RHR SYS 1(11)

LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.

e. IF LPCI NITIATION Signal exists, THEN MOMENTARILY PLACE 3-XS-74-121(129), R}IR SYS 1(11) CTMT SPRAY!CLG VLV SELECT in SELECT.

f. IF 2-FCV-74-53(67), RHR SYS 1(11) LPCI INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSE]) 2-FCV-74-52(66), RHR SYS 1(11)

LPCI OUTBD INJECT VALVE.

g. OPEN 2-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV.

h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.

i. THROTTLE 2-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 2-FI-74-50(64), RHR SYS 1(11) FLOW:

Between 7000 and 10000 gpm for one-pump operation.

OR

At or below 13000 gpm for two-pump operation.

j. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MN FLOW VALVE.

k. MONITOR RHR Pump NPSH using Attachment 1.

2-F Page 37 of 55 Critical Tasks - Three CS#1 With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BuT) and inserting control rods.

1. Safety Significance:

Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.

2. Cues:

Procedural compliance.

Suppression Pool temperature.

3. Measured by:

Observation If operating per EOI-l and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping) before exceeding 1100 F in the Suppression Pool.

AND RO places SLC A / B Pump control switch in ON, when directed by US.

AND Control Rod insertion commenced in accordance with EOI Appendixes.

4. Feedback:

Reactor Power trend.

Control Rod indications.

SLC tank level.

CS#2 During an ATWS, when conditions are met to deliberately lower RPV level, Terminate and Prevent injection into the RPV; from ECCS and Feedwater; until conditions are met to reestablish injection.

1. Safety Significance:

Precludes loss of primary containment integrity and uncontrolled release of radioactivity into the environment.

2. Cues:

Procedural compliance.

3. Measured by:

Observation - With Emergency Depressurization not required and> 5% power, injection systems are terminated and prevented until:

<5% power or < (-)162 with Suppression Pool Temp> 1100 F OR

Level < (-) 50 inches with Suppression Pool Temp < 1100 F

4. Feedback:

Injection system flow rates into RPV Reactor Power lowering

2-F Page 38 of 55 Critical Tasks Three CS#3 With reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS.

1. Safety Significance:

Precludes core damage due to an uncontrolled reactivity addition.

2. Cues:

Procedural compliance.

3. Measured by:

ADS logic inhibited prior to an automatic initiation, unless all required injection systems are Terminated and Prevented.

4. Feedback:

RPV pressure trend.

RPV level trend.

ADS ADS LOGIC BUS A/B INHIBITED annunciator status.

2-F Page 39 of 55 Scenario Tasks EVENT TASK NUMBER K/A RO SRO 1 Remove an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 2 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 3 RPIS Position Failure ROU-085-AL-14 214000A2.0l 3.1 3.3 SRO S-085-AB-04 4 EECW Pump Trip RO U-067-NO-12 400000A2.O1 3.3 3.4 5 MSIV Partial Closure RO U-001-AB-02 239001A2.03 4.0 4.2 SRO S-0O1-AB-02 6 Feedwater Flow Transmitter Failure RO U-003-NO-12 259002A2.02 3.3 3.4 SRO 5-003-AB-Ol 7 Vacuum Loss/ATWS RO U-000-EM-17 295037EA2.06 4.0 4.1 SRO S-000-EM-06 SRO S-000-EM- 18 SRO S-032-AB-02

2-F Page 40 of 55 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2-F 6 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 60 Run Time (minutes) 3 Crew Critical Tasks: (2-5)

YES Technical Specifications Exercised (Yes/No)

SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:

EECW Pump A3 is out of service and tagged out.

RFPT B Out of Service Operations/Maintenance for the Shift:

Remove LPRM 8-49B from bypass JAW 2-OI-92B section 6.4.

Once completed adjust load line JAW RCP and 2-GOI-lOO-12 section 5.0 step 20 and continue power ascension as directed by the RCP.

Units 1 and 3 are at 100% power.

Unusual Conditions/Problem Areas:

None

I z

z rn:

-L 0

I

C)

I U 0 I;

111 r

I 01

\OI

TR 3.3 INSTRUMENTATION TR 3,3,5 Surveillance Instrumentation LCO 3.3.5 The surveillance instrumentation for each parameter in Table 3.3.5-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6-1 TRM LCO 3.0.4 is not applicable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required Al Enter the Condition Immediately channels inoperable, referenced in Table 3.3.5-i for the channel.

B. As required by Bi Restore required control 7 days Required Action Al room indication channel and referenced in to OPERABLE status.

Table 3.3.6-i.

C. As required by C.1 Restore one of the 7 days from Required Action Al required control room discovery of both and referenced in indication channels for redundant channels Table 3.3.5-I. each associated for one or more parameter to OPERABLE associated status. parameters not indicating in the ANI2 control room C.2 Restore required control 30 days room indication channels to OPERABLE status.

(continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

0. As required by 0.1 Monitor torus Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action Al temperature to observe and referenced in any unexplained Table 3,3.5-i. temperature increase which might be indicative of an open relief valve.

AND 0.2 Restore control room 30 days indication by either the Tailpipe Thermocouple Temperature or Acoustic Monitor to OPERABLE status for each relief valve.

D.3 When inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months more than 30 days, initiate a Problem Evaluation Report (PER),

(continued)

ACTIONS CONDON REQUIRED ACTION COMPLETION TIME E. As required by NOTE Required Action A.1 Required Actions E1.1 and and referenced in E.1.2 are not applicable when in Table 3.351. MODES 4 and 5, E.1.1 Restore required control 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months room indication channel to OPERABLE status.

El .2 Initiate the preplanned 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months alternate method of monitoring the parameter.

AND E.2 When inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months more than seven days, initiate a Problem Evaluation Report (PER).

(continued)

ACTIONS CONDON REQUIRED ACTION COMPLETION TIME F. Required Action and Fl Be in MODE 4. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months associated Completion Time of Condition B or D not met OR Required Action and associated Completion Time of Condition C not met for instruments 3,a or 3.b, G. Required Action and (3.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition C not met for Instruments 2.a, 2.b, 4.a, or4.b, H. Required Action and 1+1 Reduce THERMAL 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion POWER to 15% RTP, Time of Condition C not met for Instrument 5 channels.

TABLE 33.&-1 (page 1 of 2>

Surveillance Instrumentation PARAMETER AND APPliCABLE REQUIRED CONDITIONS TECHNICAL TPE INSTRUMENTS MODES OR CHANNELS REFERE1vCED SURVEILlANCE INDICATION OTHER FROM REQUIREMENTS AND RANGE SPECIFIED REQUIRED CONDITIONS ACTiON A. I 1, Suppeacn I B TSR 3,3.5,1 Recorder Dhamber Air TSR 3348 D-4r#F Teperatre (XR.84-52)

2. Control Rod Motion

a. Control Rod 12 1(b) C TSR 3lI.52 Indtor00-48 Position (a)

b. Neutron 12 1(O) C TSR 3353 $RM Indicators Monitorir (a) TSR 3.3&4 0I-1D cp IRM TSR 3L57 Indicators 0-125 TSR 335,8 LPRM Indicators TSR 33.5O 0-125

3. DryII Pressure Temperature ALarm a Dywell Pressure 1,2 1 C TSR 33.5J4 Alarm at 35 psig 1PS-84478) (<I) 1>. DywelI 1,23 I C TSR 33,&1O Alarm if temp.

Teneratwe and TSR 3.3.5i3 >281F and Pressure and pressure> iS Timer pgafter3O (T544-52A and rnnute delay P1S-84-SBA and lS-54-87A3 (d)

(onhinued)

(a) The channel of ContiI Rod Position instruments and the channel of Neutron Monitoring instruments are considered redundant to each other far the parameter of Control Rod Molion.

(b) The Control Rod Position channel consists of full core display position indicators or four-rod display position indicators capable of detennining position of aft OPERABLE control rods. Position indicators are considered to be capable of determining rod position when they display the rod position or the rod can be moved to a position where rod position is displayed.

(c) The Neutron Monitoring channel contains the following:

1, In MODE 2 with IRMs on Range 2 or below a minimum ot3 OPERABLE channels of SRMs.

2. In MODE 2 a minimum of B OPERABLE channels of IRMs,

3. in MODES I and 2,43 LPRM detector assemblies, each containing four fission chambers.

indMdual failed chambers can be bypassed to the extent that APRMs remain OPERABLE.

{d) The channel of Drywell Pressure and the channel of Dr,well Temperature and Pressure and Thier Instruments are considered redundant to each other for The parameter of Diywell Pressurefremperature Alarm.

37 PLANT SYSTEMS 3.72 Emergency Equipment CoolIng Water (EECW) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 The EECW System with three pumps and UHS shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required EECW A. I Restore the required 7 days pump inoperable. EECW pump to OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not AN me.

8.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Two or mare required EECW pumps inoperable.

a UKS inoperable.

3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor buildiagtosuppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, When assocIated instrumentation is required to be OPERABLE per LCO 3,3.6i, Primary Containment Isolation Instrumentatioit ACTIONS

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6,1.1, Primary Containment, when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE A.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated AND

- automatic valve, closed manual valve, blind 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main One or more penetration flange, or check valve steam line flow paths with one POW with flow through the inoperable except due to valve secured.

MSIV leakage not within limits.

AND (continued)

ACTIONS CONDITION REOIIIRED ACTION COMPLETION TIME A. (continued) A2 NOTE Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is for isolation isolated. devices outside primary containment AND Prior to entering MODE 2or3frorn MODE 4, if primary containment was de4nerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued)

ACTIONS (continued)

GONDITION REQUIRED ACTION COMPLETION TIME

8. NOTE 81 Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Only applicable to penetration flow path by penetration flow paths use of at least one dosed with two PCIVs. and de-activated automatic valve, closed manual valve, or blind One or more penetration flange.

flow paths with Iwo PCIVs inoperable except due to MSIV leakage not within limits.

C. NOTE Ci Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and de-activatad automatic valve, closed manual valve, or blind One or more penetration flange. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for flow paths with one PCIV EFCVS inoperable.

C2 NOTE Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected Once per 31 days penetration flow path is isolated.

(continued)

ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME D. One or more penetration Di Restore leakage rate to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months flow paths with MSIV within limit.

leakage not within limits.

E, Required Action and El Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, B, C, ANP.

or D not met in MODE 1, 2, or 3 E2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. Required Action and F.1 Initiate action to suspend Immediately associated Completion operations with a Time of Condition A, B, C, potential for draining the or D not met for PCIV(s) reactor vessel (OPDRVs).

required to be OPERABLE during QR MODE 4 or &

F.2 NOTE Only applicable for inoperable RHR Shutdown Cooling Valves.

Initiate action to restore Immediately valve(s) to OPERABLE status.

SCRAM FAILURE REACTOR COOLANT ACTMTY Irrr,tnn I1p!IrIfltiflfl I I I I 1.3-IJI I I Reactor coolant activity exceeds 26 pCi!gm dose .

equivalent 1-131 (Technical Specitication Limits) Z as determined by chemisby sample.

r OPERATING CONDITION m LL z

-4 1.2AI INOTEI I 1.3-Al Failure of iPS automatic scram functions to bflng Reactor coolant activity exceeds 300 iCi!gm dose the reactor subcritical equivalent Iodine-I 31 as determined by chemistry AND sample.

I-Manual scram or ARI (automatic or manual) was 111 successful, OPERATING CONDITION:

OPERATING CONDITION: Model or 2 or 3 Modelor2 1 2-S I I NOTE 1 I F Failure of automatic scram, manual scram, and AR1 to bring the reactor subcrItical, m

m In OPERATING CONDITION:

Model C 1.2-G ICURVEI I I US I Failure of automatic scram, manual scram, and ARL Reactor power is above 3%

AND Either of the follo1ng co*dftions edsts a Suppression Pool temp exceeds HCTL In Refer to Curve t2-G.

a Reactor water level can NOT be restored and maintained at or above -180 inches.

m z

OPERATING CONDITION:

Model or 2

ML110980768

Text

Navigation menu

Search