ML041890346

Draft Section C Operating

ML041890346
Person / Time
Site:	Oyster Creek
Issue date:	04/19/2004
From:	Hackenberg J AmerGen Energy Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-219/04-301 50-219/04-301
Download: ML041890346 (55)

Text

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. NRC #1 Op Test No. ILT2004 Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor at 100% power with the A CRD Pump out of service. The crew Summary will begin by removing the EPR from service. The running steam seal exhaust blower trips. The crew will start the other exhaust blower. A reference leg leak will develop in a RPV Level instrument. The crew will take manual control of RPV level and transfer to the alternate signal. A loss of power to VMCC 1A2 will result in the crew restoring RPS and resetting the half scram. The CyFeedwater Pump trips requiring the crew to reduce power to maintain reactor level. The only available CRD pump trips, which will require the crew to scram the reactor. A RWCU leak will occur in the Reactor Building requiring entry into Secondary Containment Control EOP. A RWCU valve will fail preventing the isolation of the leak. Emergency Depressurization will be required to mitigate the primary leak into the Reactor Building.

Event Malfunction Event Event No. No. Type* Description SRO 1 N BOP Remove EPR from service for maintenance MAL MSS005A C SRO Steam seal exhaust blower trips, start other blower 3 MAL-NSSO11C I RO Level instrument reference leg leak develops, take manual control of 2%, 300s RPV level, swap instrument signals, return to auto control SRO 4 MAL-EDSOO4A C BOP Loss of Power to VMCC 1A2, restore RPS (Tech Spec)

SRO 5 MAL-CFWOO6C R BOP CFeedwater Pump Trip leads to Power Reduction to Control Level RO SRO 6 BKR CRD001 C RO Only available CRD pump trips - Results in Plant Scram

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SRO 7 MAL RCUl3 3% M Reactor Water Clean-up Leak into the Reactor Building 600s BOP [HELB]

SRO 8 VLV RCUOOl , 6 C BOP Reactor Water Clean-up Isolation Valve Failure, V-14-1, 14 & 61 fail to VLV RCU004,6 auto close. V-14-61 can be manually closed.

VLV RCUOl 1, 5

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

Operator Actions ES-D-2 b Op Test No.: lLT2004 Scenario No.: NRC #1 Event No.: 1 Page 1 of 8 Event

Description:

Remove EPR from service for maintenance Cause: Erratic EPR response Automatic Actions: None Effects: None Time Position Amlicant's Actions Or Behavior SRO Direct transfer from EPR to MPR IAW Operating Procedure 315.4, Transferring Pressure Regulators BOP IAW 315.4, step 3.3; 0 Slowly lower MPR setpoint by placing MPR Control Switch to lower (7%)

position for approximately one-second periods until MPR relay position indicator moves toward the EPR setting.

0 Adjust EPR control switch so that EPR pressure setpoint is 6-7 psig higher than the pressure at which it had been operating.

c

Operator Actions ES-D-2 Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 2 Page 2 o f 8 Event

Description:

Steam seal exhaust blower trips, start other blower Cause: Motor overload causes trip Automatic Actions: none Effects: Operator action required to start other blower Time Position Applicant's Actions Or Behavior BOP Recognize condition by reporting alarms; Q-8-C: EXHAUSTER TRIP 0

IAW RAP Q-8-c confirms:

loss of exhauster 0 checks gland steam pressure on panel 7F.

SRO Directs start of other gland steam exhauster BOP Places other gland steam exhauster in service, may refer to procedure 325 0 Starts Exhauster Blower #2 0 Closes V-7-38 0 Opens V-7-39 to maintain Gland Steam Vacuum between 15 and 17.5 inches vacuum BOP Monitors and reports gland steam pressure

Group Heading -

T U R B I N E V A C / S E A L S 0 c oLx/1-1 OUc/l-2 Reference Drawings :

GU 3B-612-19-017 GB 157B6350, Sh. 186A, 1868 I

ZONFIRMATORY ACTIONS :

%eck gland steam pressure on Panel 7P.

WIOMATIC ACTIONS : -

q0NE W A L CORRECTIVE ACTIONS:

Subject Procedure No.

Page 1 of 1 B O P 2000-RAP-3024.03 Q c A l a r m Response Procedures Revision NO: 118 I

MANUAL CORRECTIVE ACTIONS u

1.0 START standby gland steam exhauster [ I auster can be started co CE a pfant shutdown in accordance with Pro 203 [ I r System #7 Isolation Valve [ J ommand Center (OCC) [ I h for the Turbine Operating I 1 uum and offgas fl

[ I cticm of possible stem leak on

[ I 3.0 DETERMINE cause of overload on tripped exhauster blower [ I and remtbreaker 4.0 CORRECT cause of overlo [ I L

roup Heading .

T U R B I N E V A C / ' S E A L S Q c SETPOINTS: ACTUATING DEVICES:

3verload trip of either gland steam 3 3 - 0 amps OLX/ 1 - 1

>lower. OLX/1-2 Reference Drawings:

GU 33-611-17-017 GE 157B6350, Sh. 186A, 186B 30NFIRMATORY ACTIONS:

3heck gland steam pressure on Panel 7F.

4UTOMATIC ACTIONS:

NONE MANUAL CORRECTIVE ACTIONS:

Start standby gland steam exhauster. -termhe cause of overload on tripped exhauster blower. <-Correctcause of overload and reset breaker.

Subj ect Procedure No.

Page 1 of 1 B O P -

2 0 00 -RAP 3024 0 3 Q c Alarm Response Procedures Re-wision No: 118 I I

OYSTER CREEK GENERATING Number AmerGem STATION PROCEDURE 325 An ExelonlBnbsh Energy Company Title Revision No.

u Air Extraction and Off Gas System 54 3.2 Precautions and Limitations 3.2.1 The Main Turbine Steam Seal System must be operating when the turbine is hot and condenser vacuum is present.

3.3 Instructions for Placinq the Steam Seals Exhauster in Service 3.3.1 Place the control switch for the EXHAUSTER BLOWER l(2) on Panel 7F to START and verify that the EXHAUSTER BLOWER ON indicator itluminates.

3.3.2 NOTE With the Mechanical Vacuum Pump in service and Condenser Vacuum Breaker V-2-44 CLOSED,the GLAND STEAM HEADER VACUUM Gage (EPT-5) should read greater than 0 inches of vacuum.

Gradually throttle OPEN the EXHAUSTER VALVE I: V-7-38 (EXHAUSTER VALVE 2: V-7-39) until the vacuum in the GLAND STEAM HEADER increases by 5 to 6 inches of water, above the initial vacuum reading on Panel 7F (EPT-5).

3.3.3 NOTE As steam is admitted to the STEAM SEAL HEADER, the vacuum in the GLAND STEAM HEADER (Panel 7F: EPT-5) will decrease.

7.0

OYSTER CREEK GENERATING Number AmerGen, STATION PROCEDURE 325 1

An ExelanlBritish Energy Company

- Title Air Extraction and Off Gas System R g s i o n No.

Commence warming the STEAM SEAL HEADER in accordance with Procedure 318.

3.3.4 CAUTION Throttling of the Exhauster Valve to increase GLAND STEAM HEADER VACUUM (EPT-5) must be performed in 2 to 4 inches of water increments, with a 1 minute delay between each step increase, to prevent an EXHAUSTER BLOWER trip.

WHENthe GLAND STEAM HEADER VACUUM, as read on Panel 7F (EPT-5), decreases below 5 inches of water, THEN perform the following steps:

3.3.4.1 Throttle OPEN the EXHAUSTER VALVE I: V-7-38 (EXHAUSTER VALVE 2: V-7-39) until a vacuum of I O to 12 inches of water is established in the GLAND STEAM HEADER (Panel 7F: EPT-5).

3.3.4.2Maintain the GLAND STEAM HEADER VACUUM (EPT-5) at 10 to 12 inches of water for a period of 15 minutes from the time that the STEAM SEAL HEADER PRESSURE is established at 6 to 6.5 psig, as read on Panel 7F (EPT-4).

3.3.4.3Throttle OPEN the EXHAUSTER VALVE 1: V-7-38 (EXHAUSTER VALVE 2: V-7-39) until a vacuum of 15 to 17.5 inches of water is established in the GLAND STEAM HEADER (Panel 7F: EPT-5).

3.3.5 Adjust the EXHAUSTER VALVE 1: V-7-38 (EXHAUSTER VALVE 2: V 39), as necessary, to maintain GLAND STEAM HEADER VACUUM at 15 to 17.5 inches of water as indicated on Panel 7F (EPT-5).

3.4 Instructions for Placinq the Standbv Steam Seals Exhauster In Service (325) 8.0

Operator Actions ES-D-2 OpTestNo.: ILT200 Event No.: 3 Page 3of 8 Event

Description:

nstrume-s Downscale, (Tech Spec)

Cause: Instrument failure causes downscale response Automatic Actions: /J/;a Effects: Operator action required to determine Tech Spec implication Time Position Applicants Actions Or Behavior Recognize condition by any one of the following plant parameters that are changing:

0 Gemac A and C level increasing 0 Yarway level, and B Gemac decreasing Feed flow decrease 0 Megawatts thermal decrease 0 Drywell pressure increasing Determines level transmitter malfunction/reference leg leak. Refers to ABN-17 and procedure 317, section 11.7 SRO 0 Directs manual feedwater control to regain reactor level 0 Directs swap to B GEMAC instrument for automatic level control 0 Directs feedwater control returned to AUTO 0 Evaluate TS sections 3.7, Auxiliary Electrical Power and determines that the EDG may be inoperable and the plant may remain in operation for 7 days BbB Attempts to regain RPV level by performing the following operator actions:

I 0 Places master feedwater controller in Manual and increases feedwater flow by increasing demand signal.

Swaps to alternate Level Select signal by performing the following:

0 Place LEVEL TRANSMITTER SELECT to the B Gemac 0 Select the S display on Master feed controller 0 Match S display readout to the P display readout 0 When S equals P, places Master feed controller to AUTO 0 Maintains(returns) level in(to) normal band or as directed

Operator Actions ES-D-2 L Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 4 Page 4 o f 8 Event

Description:

Loss of Power to VMCC 1A2, restore RPS (Tech Specs)

Cause: Breaker malfunction causes trip Automatic Actions: RPS System 1 half scram Effects: Operator action required to transfer RPS power supply and reset the half scram Time Position Applicants Actions Or Behavior BOP Recognize condition by reporting alarms; 0 9XF-3-a: PROT SYS PNL 1 PWR LOST 9XF-1-c: VLDP-1 PWO TRANSFER Diagnoses the loss of power to VMCC 1A2 based on the VLDP-1 transfer and/or other indications/alarms SRO Directs execution of ABN-50, Loss of VMCC 1A2 BOP Executes ABN-50 RO 0 Restores RPS IAW section 3.0 0 Confirm VMCC-1B2 breaker, C4L is closed 0 Confirm disconnect switch SW-733-169 (Lower Cable Spreading Room) is OFF and the Kirk Key removed 0 Confirm the Kirk Key inserted and disconnect switch, SW-733-170, is ON.

0 Confirm closed EPA breaker #5 0 Confirm closed EPA breaker #6 PLACE the POWER SELECT switch in the TRANS position 0 When power is restored to PSP-1, then RESET; Half scram, Main steam isolation, APRM lights on Panel 3R, APRM flow converters in Panels 3R and 5R, and associated annunciators 0 Confirms VLDP transfer 0 Declares V-14-33, 35 INOP 0 Declares C Battery INOP - Monitors volts 0 Reviews Attachment 50-3 for loads Follow-up Actions 0 Monitors 1-8 sump (312.9, 351.1,2) 0 Initiatestroubleshooting (Notifies WWM)

SRO 0 Monitors C Battery Room temperature (328.1) 0 Evaluate TS sections 3.7, Auxiliary Electrical Power and determines that the plant must be in cold shut down in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> due to loss of power 0 Evaluate TS section 3.3.D.4,5 for Reactor Coolant System Leakage and recognizes that UlLR monitoring capability must be determined

V I T A L P O W E R A C P W R L O S T 9 X F a I

P R O T S Y S P N L 1 P W R L O S T IES: SETPOINTS : ACTIJATING DEVICES :

Loss of power to Protection System None Relay K1 Panel #1 (PsP-1).

Reference Drawings:

GU 33-611-17-022 BR 3013, Sh. 1 GU 3C-733-11-010 GE 9133911 CONFIRMATORY ACTIONS:

Verify the following l o s s of power ala= and iindications:

Loss of Power Alarms:

Enqravinq Location RPS MG SET 1 TRIP G-2-c SCRAM CONTACTOR OPEN G-1-c Loss of Power Indications:

Indication Location SCRAM SOLENOIDS lights exthguished 6R & 4F TRANS OUTPUT and GEN OUTPUT lighks extinguished 6R REACTOR PROTECTIONS MG SET-1 v o l e indicates zero 6R HALF SCRAM 4F IRMs 11 through 14 fail downscale 4F APRMs 1 through 4 fail downnscale 4F LPRMs A and C fail downscaZe 4F ROD BLOCK 4F Subject Procedure No.

Page 1 of 2 E L E C T R I C A L 2 0 0 0 -RAP-3 024.- 02 9 X F a A l a r m Response Procedures Reviision EJQ: 81 I I (PANELXF)

~

Group Heading V I T A ' L P O W E R A C P W R L O S T 9 X F a P R O T S Y S P N L 1 P W R L O S T CONFIRMATORY ACTIONS (continued):

If power is 1ost.to both Protection System Panels 1 and 2, the following ~ 5 1 1occur:

FULL SCRAM 4F ALL RECIRC PUMPS TRIP 3F REACTOR isolation 11F PRIMARY CONTAINMENT isolation 11F SECONDARY CONTAINMENT isolation TURBINE-GENERATOR trip 7F ROPS function disabled 14XR AUTOMATIC ACTIONS:

None MANUAL CORRECTIVE ACTIONS:

Refer to Technical Specification 3.7 for limitations on continued plant operation.

Refer to the following Procedures for instructions on manual corrective amions:

2000-OPS-3024.10e, "Electrical Distribution: Reactor Protection System-Diagnostic and Restoration Actions".

408.12, "Operation of Reactor Protection System Panel 1-1 and Transfo-r PS-lt'.

Refer to drawing GU 3C-733-11-010 for PSP-1 Panel Schedule Subject Procedure No.

Page 2 of 2 E L E C T R I C A L 2000-RAP-3024.02 9 X F a Alarm Response procedures Revision No: 81

( PANELXF)

Group Heading V I T A L P O W E R A C X F E R S 9 X F c V L D P - 1 P W R X F E R CAUSES : SETPOINTS: ACTUATING DEVICES :

Automatic transfer switch for 460 Volt Dropout @ 70% v. Relays lV, 2V, 3V Vital AC Lighting Distribution Panel transferred to alternate source. Pickup 0 9 0 % v. Reference Drawingsr JC 19643 BR 3013 SH, 1 GU 33-611-17-022 CONFIRMATORY ACTIONS:

Verify operation of the automatic transfer switch.

AUTOMATIC ACTIONS:

460 volt supply power for the Vital AC Lighting Distribution Panel transfers f m m Vital MCC 1A2 to Vital MCC 1B2.

MANUAL CORRECTIVE ACTIONS:

Zheck for a fault or breaker trip on 460 Volt VMCC 1A2. Restore power to 460 Valt m C C 1A2. Refer to Procedure 339 to reset the transfer switch.

Refer to Procedure 2000-OPS-3024.10b, Electrical Distribution - 4 6 0 VAC Diagnagtic 2nd Restoration Actions.

subject Procedure No.

E L E C T R I C A L 2000-RAP-3024 -02 Page 1 of 1 9 X B c Alarm Response Procedures Revision No: 81 I I

Operator Actions ES-D-2

,P

• --/,

Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 5 Page 5 o f 8 Event

Description:

'C' Feedwater Pump Trips leads to Power Reduction to Control Level Cause: Motor malfunction causes overload trip Automatic Actions: Pump trip alarms Effects: Reactor power, steam flow and feed flow decrease. Operator action reduces required feedwater flow Time Position Applicant's Actions Or Behavior BOP Recognize condition by reporting alarms; J-1-d: FEED PUMP TRIP A J-24: FEED PUMP OL A IAW RAPS; confirm automatic actions and indications including Feed pump amps, discharge pressure, flow, etc.

SRO ,Direct reduction of Reactor power, using recirculation flow, to within the capacity of the remaining feed pumps (approximately 70% power) IAW ABN-17, Feedwater System Abnormal Conditions RO IAW ABN-17:

0 Reduce recirculation flow by dialing down on the Master Recirc Controller as required to reduce Reactor power and prevent scram on RPV low level and keep feed flow within the capacity of the remaining feed pumps.

0 Monitor Reactor parameters BOP Monitor Feedwater pumps and flow BOP Direct Equipment Operator to investigate feed pump and its breaker

G ro up Headin g FEED PUMPS 1A J-2-d FEED PUM OL P 2

A CAUSES: SETPOINTS: ACTUATING DEVICES:

Motor overload. Relay 74 Reference Drawings:

GE 223R0173 Sh 7 GU 3E-611-17-01 I CONFIRMATORY ACTIONS:

Check feedpump motor amps.

AUTOMATIC ACTIONS:

Eireaker trips at 1040 amps.

MANUAL CORRECTIVE ACTIONS:

1. Check pump bearing and motor temperatures.

?. If overload persists, reduce reactor power as necessary and remove pump from service.

Page 1 of 1 BOP 2000-RAP-3024.03 t J-2-d Alarm Response Procedures Revision No: 121 Panel J

Operator Actions ES-D-2 L- Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 6 Page 6of 8 Event

Description:

'B' CRD Pump trips - Results in Plant Scram Cause: Breaker malfunction causes CRD pump trip Automatic Actions: Pump trip alarms with subsequent low Charging Water pressure alarms Effects: Requires operator action to scram reactor on Accumulator LeveVPressure Rod Block illuminated Time Position Applicant's Actions Or Behavior RO Recognize condition by reporting alarms; BOP H-2-C:PUMP B OL H-7-C:CHARG WTR PRESS LO IAW RAP H-2-c confirms:

0 Check CRD system flow 0 Check position of CRD pump minimum flow valve 0 Recognizes loss of only available CRD pump a a D y c, 0 Diagnoses the loss of CRD charging pressure at panel 4F SRO 0 Per RAP H-7-c, directs manual scram when Accumulator LeveVPress Rod Block illuminates (indication of 2"daccumulator trouble alarm) x 0 May enter ABN-01 and direct recirculation flow reduced to 8.5 x 1O4 gpm RO 0 Scrams the reactor when Accumulator LeveVPressure Rod Block illuminates I"

and enters ABN-1 0 Depresses both manual scram push buttons 0 Places mode switch in SHUTDOWN 0 verifies reactor shutdown, rods fully inserted to 00, 02 or 04 verifies power decrease inserts SRMs and IRMs

- NO

Operator Actions ES-D-2 Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 7 Page 7 o f 8 Event

Description:

Reactor Water Clean-up (RWCU) Leak into the Reactor Building Cause: RWCU pipe leak Automatic Actions: RWCU auto isolation Effects: Operator action required Time Position Amlicant's Actions Or Behavior RO Recognize condition by observing indications or reporting alarms:

BOP D-1-d/D-2-d: RWCU HELB 0 Increase in unidentified leak rate 0 Increase in containment parameters IAW RAPS; confirm automatic actions and indications including RWCU system status, area temperatures, area radiation levels SRO Enter and execute EOP 3200.1 1, Secondary Containment Control 0 Direct the RWCU system isolation be verified RO Recognize that RWCU is not fully isolated.

BOP 0 Attempt to isolate the RWCU system and report the failure of the isolation valves. See Event 8 for details.

c

Group Heading CLEANUP SYSTEM D-1-d 0

CAUSES:

RWCU HELB I i SETPOINTS: ACTUATING DEVICES:

High ambient temperature in Cleanup Pump 160°F Increasing TSH-215-006 area, indicating possible RWCU HELB or downstream of isolation valves. TSH-215-007 K

Reference Drawings:

GU 3E-611-17-006 GE 237E566 Sh. 1 Check temperature indicator on Panel 1OR for affected area(s)/cmponent(s).

Check Alarm D-8-d.

AUTOMATIC ACTIONS:

Isolationof Cleanup System Valves V-16-1, V-16-2, V-I 6-14, and V-16-61 if alarm D-2-d is received concurrently.

NOTE: The closure of the isolation valves will cause the operating Cleanup Pump@)to trip.

MANUAL CORRECTIVE ACTIONS:

Ifhigh temperature confirmed without system isolation, immediately isolate the CU System by closing V-16-1, V-16-2, V-16-14, and V-16-61.

Ifthe system is manually isolated and V-16-2 is open, then Bypass the V-16-2 torque switch by removing the bypass plug from BP14 and inserting into BP13 (EOP BYPASS PLUG PANEL behind Panel 3F), then close V-16-2.

B Enter Procedure EMG-3200-11 Secondary Containment Control.

B Check radiation level indicationfor CU equipment on Panel 2R.

D Check Cleanup System area for source of high temperature (steam leak, fire) if conditions permit.

Subject Procedure No.

Page 1 of 1 NSSS 2000-RAP-3024.01 D-1 -d Alarm Response-Procedures Revision No: 121 (Panel D/19)

Group Heading CLEANUP SYSTEM D-2-d RWCU HELB High ambient temperature in Cleanup Pump 160°F Increasing TSH-215-008 area, indicating possible RWCU HELB or downstream of isolation valves. TSH-215-009 Reference Drawings:

GU 3E-611-17-006 GE 237E566 Sh. 1

! CONFIRMATORYACTIONS:

Check temperature indicator on Panel 1OR for affected area(s)/component(s).

Check Alarm D-8-d.

AUTOMATIC ACTIONS:

Isolation of Cleanup System Valves V-16-1, V-16-2, V-16-14, and V-16-61 if alarm D-1-d is received concurrently.

NOTE: The closure of the isolation valves will cause the operating Cleanup Pump(s) to trip.

MANUAL CORRECTIVE ACTIONS:

Ifhigh temperature confirmed without system isolation, immediately isolate the CU System by closing V-16-1, V-16-2, V-16-14, and V-16-61.

If the system is manually isolated and V-16-2 is open, then Bypass the V-16-2 torque switch by removing the bypass plug from BP14 and inserting into BP13 (EOP BYPASS PLUG PANEL behind Panel 3F), and then close V-16-2.

Enter Procedure EMG-3200-11Secondary Containment Control.

Check radiation level indication for CU equipment on Panel 2R.

Check Cleanup System area for source of high temperature (steam leak, fire) if conditions permit.

Isolate or shutdown equipment as necessary to control leak.

Subject Procedure No. ,

Page 1 of 1 NSSS 2000-RAP-3024.01 D-2-d Alarm !?espc)!?se ..

Procedures Revision No: 121 I I (Panel D/20)

Operator Actions ES-D-2

--- Op Test No.: ILT2004 Scenario No.: NRC #1 Event No.: 8 Page 8of 8 Event

Description:

Reactor Water Clean-up Isolation Valve Failure Cause: Breaker malfunction prevents auto valve closure Automatic Actions: none Effects: Incomplete RWCU system isolation. Operator action required mitigate unisolable leak Time Position Applicants Actions Or Behavior RO Identify failure of RWCU system to fully isolate. Attempts to close V-14-1, 14 &

BOP 61. Able to close V-14-61. Unable to close V-14-1 & 14.

SRO Determine that a primary system is discharging into the secondary containment.

Before exceeding Max Safe temperatures, enterheenter EOP 3200.01 A, RPV Control - No ATWS.

BOP Record and/or report area temperature and radiation indicati CT SRO Directs Emergency Depressurization IAW EOP 3 Depressurization - No ATWS

---/,

0 Direct bypassing Reactor Overfill Protection System (ROPS) 0 Direct manually opening all EMRVs RO Bypass ROPS CT BOP Opens all EMRVs RO Control reactor level during the depressurization TERMINATION CRITERIA: Once ED is performed and reactor is depressurizing, or at the discretion of the lead evaluator, the scenario may be terminated POST SCENARIO EMERGENCY CLASSIFICATION: ALERT: primary containment isolation required and isolation valves malfunction causing unisolated release path or confirmed leak-rate exceeds 50 gpm from reactor coolant system. EAL: H-lc or H-2

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. NRC #4 Op Test No. ILT2004 Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor startup in progress at 5-10% power with mode switch in RUN.

Summary Control rods will be moved to raise power. The crew will swap the Service Water Pumps. Running CRD pump trips, start standby pump. Reactor Level Instrument RE02A Fails Downscale causing the Core Spray to start but EDG #2 does not start and idle. Core Spray will be manually secured. APRM 4 will then fail upscale requiring the crew to evaluate Tech Specs, bypass the APRM, and reset the half scram. The running RBCCW pump trips requiring the standby pump to be started. An RPV steam leak will result in increase in Drywell temperature and pressure. Drywell pressure will increase requiring Drywell Sprays using the Containment Spray system. The drywell spray valve fails to automatically realign and must operated manually to permit sprays to function.

SRO 4 MAL-NSS007H I BOP Reactor Level Instrument Fails Downscale, Core Spray starts but EDG MAL-DGN003A does not idle 5 I MAL-NIS020D Ro I APRM 4 Fails Upscale (Tech Spec) 6 MAL-RBC001A C BOP Running RBCCW pump trips MAL-NSSO17A SRO 7 1%, 300s M RO Steam leak develops in the Drywell leads to spraying DW 2%, 1800s BOP SRO 8 VLV CNS008, C BOP Containment Spray Valve fails to realign automatically when sprays opt 6 are required

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

Operator Actions ES-D-2 k d Op Test No.: ILT2004 Scenario No.: NRC #4 Event No.: 1 Page 1 of 9 Event

Description:

Pull rods to raise power Cause: Complete power ascension Automatic Actions: None Effects: None Time Position Applicant's Actions Or Behavior SRO Review rod withdrawal sequence from turnover Finished Group 7-2, Pull 20, Step 3 Next action, complete Group 7-2, Pull 20, Step 4 and then complete Group 7-3 Step 21 (all 4 rods from 8 to 12)

Direct resumption of power ascension IAW 201 RO 0 Verify rod selected is correct 0 Begins to pull rods in sequence IAW Rod Withdrawal Sequence For each rod selected:

0 verifies rod 0 selects rod 0 notch withdraws rod to position 12 0 verifies correct notch 0 initials completed action BOP 0 Assist in verifying correct rod 0 Second check on rod movements

Operator Actions ES-D-2 i/ Op Test No.: lLT2004 Scenario No.: NRC #4 Event No.: 2 Page 2of 9 Event

Description:

Swap Service water pumps Cause: Equipment rotation Automatic Actions: none Effects: none Time Position Applicant's Actions Or Behavior SRO Direct that Service water pumps be wapped IAW Op rating Procedure 322, Service Water System BOP IAW 322, section 5.0; Dispatch Equipment Operator to the Intake and establish communication with the control room 0 Start idle pump by placing its control switch to START at 5F/6F Stop originally running pump by placing its control switch to STOP at 5F/6F 0 Verify discharge check valve closure by observing NO reverse rotation of the now idle pump 0 OPEN operating pump continuous vent valve and confirm flow from it

Operator Actions ES-D-2

'- Op Test No.: ILT2004 Scenario No.: NRC #4 Event No.: 3 Page 3of 9 Event

Description:

Running CRD pump trips (Tech Specs)

Cause: Breaker problem causes pump trip Automatic Actions: none Effects: Operator action required to start standby pump Time Position Applicant's Actions Or Behavior RO Recognize condition by reporting alarms; H-l-c: PUMP A OL H-7-C:CHARG WTR PRESS LO IAW RAP H-1-c confirms:

0 the running CRD pump has tripped 0 availability of standby CRD pump SRO 0 Direct start of standby CRD pump 0 Direct reference to 3024.08, Control Rod Hydraulics - Diagnostic and Restoration Actions 0 Notify Work Management to troubleshoot and repair the pump 0 Evaluate TS sections 3.4., Emergency Cooling and determines that the CRD pump may be inoperable and the plant may remain in operation for 7 days Starts standby CRD pump.

Monitor CRD parameters and valve positions Refers to 3024.08, Control Rod Hydraulics - Diagnostic and Restoration Actions to determine cause of equipment problem

Operator Actions ES-D-2 L, Op Test No.: ILT2004 Scenario No.: NRC #4 Event No.: 4 Page 4of 9 Event

Description:

Reactor Level Instrument RE02 Fails Downscale, Core Spray starts but EDG does not idle (Tech Spec)

Cause: Instrument failure Automatic Actions: Core Spray starts but EDG does not idle Effects: Requires operator action to secure Core Spray system and determine Tech Spec Time Position Applicant's Actions Or Behavior BOP Recognize condition by reporting alarms; 0 B-l-e: SYSTEM 1 AUTOSTART B-l-f: SYSTEM 2 AUTOSTART IAW RAP B-1-e & 1 -f confirms:

0 both Core Spray systems are running 0 verifies #2 EDG has idle started and that #1 EDG failed to start and idle.

0 Using multiple indications verifies that a valid lo-lo signal does not exist L

Based on alarms and indications, reports that both Core Spray systems started due to RE02 failure, but that #1 EDG did not idle.

SRO 0 Confirms that a valid 10-10 signal does not exist 0 Requests Work Management assistance and/or may direct the I&C technician to investigate the problem 0 Evaluate TS 3.7.C.2, Auxiliary Electrical Power, and enters a 7 day LCO Directs URO/BOP to secure core spray in accordance with Procedure 308 BOP Secures Core Spray IAW 308 section 5.0:

Depresses OVERRIDE push buttons and then depresses ACTUATED push buttons to reset Core Spray Logic 0 Confirm the parallel isolation valves are closed Secures running booster pumps and then main pumps in each system 0 Verify system is in standby readiness

Group Heading C O R E S P R A Y 1 i B e S Y S T E M 1 A U T O S T A R T

~

SETPOINTS: ACTUATING DEVICES:

Low low reactor water level 90" above TAF RE02AY5 REG 2BY5

- OR - REO 2CY5 RE02DY5 High drywell pressure Drywell press. (Panel 18R & 19R Relay 2.9 psig Modules P.S. RV46 A, B, C, D Reference Drawings :

NU 506036003 GU 33-611-17-004 Sh. 1 CONFIRMATORY ACTIONS:

Verify low low reactor water l e v e l o r high drywell pressure.

Confirm System 1 pumps and d i e s e l generators running.

UJTOMATIC ACTIONS:

starts Core Spray pumps and diesel generators.

W A L CORRECTIVE ACTIONS:

If drywell pressure exceeds 3 psig, enter Procedures EMG-3200.02, Primary lontainment Control and EMG-3200.01A, RPV Control - No ATWS or EMG-3200.01B, RPV Jontrol with ATWS.

rhis alarm indicates that a parameter has exceeded or has the potential to exceed an Zmergency Action Level (EAL). Enter Procedure EPIP-OC-.Ol, "Classification of 3rnergency Conditions11. EAL - EPV Level subject Procedure No. I I

i Page 1 of 1 N S S S 2QOO-RAP-3O24-01 i B e Alarm Response Procedures (PanelB/21)

~

Group Heading C O R E S P R A Y 2 B f S Y S T E M 2 A U T O S T A R T SETPOINTS: ACTUATING DEVICES:

Low low reactor water level 90" above TAF REO 2AY5 REO 2BY5

- OR - REO 2CY5 REO2DY5 High drywell pressure Drywell press. (Panel 18R & 19R Relay 2.9 psig Modules )

P.S. RV46 A, B, C, D Reference Drawings:

Nu 506036003 GU 3E-611-17-004Sh. 1 CONFIRMATORY ACTIONS:

Verify low low reactor water level or high drywell pressure.

Confirm core spray system pumps and diesel generators running.

AUTOMATIC ACTIONS:

Starts Core Spray pumps and diesel generators.

MANUAL, CORRECTIVE ACTIONS:

If drywell pressure exceeds 3 psig, enter Procedures EMG-3200.02, Primary Containment Control and EMG-3200.01A, RPV Control - No ATWS or EMG-3200.01B, RPV Control with ATWS.

This alarm indicates that a parameter has exceeded or has the potential to exceed an Emergency Action Level (EAL). Enter Procedure EPIP-OC-.O1, "Classification of Emergency Conditions". EAL - EPV Level Subject Procedure No. I Page 1 of 1 N S S S 2000-RAP-3024.01 B f Alarm Response Procedures Revision No: 127

( PanelB/32)

Operator Actions ES-D-2 i/

Op Test No.: lLT2004 Scenario No.: NRC #4 Event No.: 5 Page 5of 9 Event

Description:

APRM 4 Fails Upscale (Tech Spec)

Cause: Instrument failure causes upscale response Automatic Actions: none Effects: Requires operator action to bypass APRM and reset the half scram Time Position Applicant's Actions Or Behavior RO Recognize condition by reporting alarms; G-1-c: SCRAM CONTACTOR OPEN G-1-f: APRM HI-HI/INOP G-3-f: APRM HI G-l-d: CHANNEL I IAW Response to Alarm Procedures (RAPS); confirm automatic action and indications including RPS system 1 scram lights out on 4F and APRM 4 indications on 4F.

Based on alarms and indications, reports RPS system 1 half scram due to APRM 4 failing upscale.

SRO 0 Refers to Procedure 403, LPRM-APRM System Operations Requests Work Management assistance and/or may direct the I&C technician to investigate the problem 0 Evaluate TS 3.1, Protective Instrumentation, to ensure that it permits the APRM to be bypassed 0 Directs APRM 4 to be bypassed and the half scram to be reset RO Bypass APRM input IAW 403 section 5.3.3:

0 Check Section 5.4 and Tech Spec Section 3.1 to determine if channel may bebypassed 0 Bypasses APRM 4 by placing the joystick in bypass BOP 0 Verify on 3R/5R that selected APRM indicates it is bypassed 0 Update Attachment 403-2 as determined by US 0 Reset the half scram

Group Heading REACTOR/RPS G-l-c SCRAM CONTACTOR ACTUATING DEVICES:

Any of the Reactor Protection System automatic or Scram relay tripped. Relays:

manual scram relays tripped.

1K21A, 1K51, 1K52, 2K21A, 2K51,2K52 Reference Drawings:

GE 237E566 Sheet 2,3,6, & 7 GU 3E-611-17-009 Sh. 1 CONFIRMATORY ACTIONS:

If necessary, check scram relay panels to determine which d a y has tripped. Verify proper RPS MG set operation at Panel 6RnR.

AUTOMATIC ACTIONS:

Reactor scram with coincident Channel I and Channd II trips.

MANUAL CORRECTIVE ACTIONS:

Check other Reactor Protection System alarms at Panel 5FBF to determine cause of trip. For reactor scram verify automatic actions and perform followup actions as defined in Procedure 2000-ABN-3200.01, "Reactor Scram". If RPS bus voltage is lost, refer to Procedure 2000-OPS-3024,1Oe, "RPS Restoration - Diagnostic 8 Restoration".

half scram signal is present and the condition has cleared, '

E THEN reset the half scram upder direction of Unit Supervisor

/

/'

Subject /

Procedure No.

Page 1 of 1 NSSS 2000-RAP-3024.01 Alarm Response 1 G-1-C Procedures rnvision No. 127 (Panel G/12)

Group Heading REACTOR NEUTRON MONITORS G-1-d SETPOlNTS: ACTUATING DEVICES:

These trips are inputs to RPS Channel I:

NOTE: Below trips are operable only in STARTUP or REFUEL modes.

IRM HI HI Level greater than 118% on or hop: 125% scale or 38% on 40% scale or channel inoperative.

NOTE: Below trips are operable only in RUN mode.

IRM HI HI Level greater than 118% on or 125% scale or 38% on 40%

Inop/APRM scale or channel Downscale: inoperative coincident with APRM channel less than 2%.

NOTE: Below trip is operable in all modes of operation.

APRM HI HI Level greater than 3r INOP: (.95x Flow) + 60%

(For flow greater than 7.68 x lo4gpm'

(.98x flow) + 37.7%

I Reference Drawings:

(For less than 7.68 x 1O4 gpm) GE 237E455 Sh. 1A with a maximum value of 118% or GU 3E-611-17-009 Sh. 1 channel inoperative.

I

ON FIRMAT0RY ACT10NS :

4UTOMATIC ACTIONS:

3eactor scram coincident with Channel I I trip.

Subject Procedure No.

Page 1 of 2 NSSS 2000-RAP-3024.01 G-1-d Alarm Response Procedures Revision No. 127 (Panel GI18)

Group Heading REACTOR NEUTRON MONITORS G-1-d CHANNEL I

~

MANUAL CORRECTIVE ACTIONS:

For half scram signal, check IRMs and APRMs to determine cause of trip. Correct cause as follows and reset half scram.

IRM HI HI Check IRM recorders at Panel 4F and IRM cabinets at Panels 3R and 5R.

or lnop For high alarm, adjust range switch in affected unit one position higher to maintain level between high and low trips. For inoperative unit, check for loss of high voltage power to the detector unit, selector switch in other than operate, or the module disconnected. One high or inoperative channel in each protection system may be bypassed in accordance with procedure 402.4, IRM Bypass Operation.

IRM HI HI Determine which APRM has caused alarm by noting Dnscl or h o p light or on Panel 4F. Check for downscale LPRM inputs to affected channel Inop/APRM at Panels 3R and 5R. Verify failed LPRMs with TIP trace and bypass Downscale: affected units in accordance with Procedure 403. Re-adjust affected APRM as required.

A+RM HI HI Determine cause of trip by checking APRM recorders at Panel 4F and or h o p APRM cabinets at Panels 3R and 5R. For high APRM, reduce reactor power by inserting control rods. For inoperative unit, check the affected APRM cabinet for improper mode switch position, Status LED on the FCTR Card lit solid red, module removed, or more than three LPRM inputs bypassed. Refer to Procedure 403.

For full scram condition, verify automatic actions and perfom followup actions as defined in Procedure 2000-ABN-3200.0tIReactor Scram.

Subject Procedure No.

Page 2 of 2 NSSS 2000-RAP-3024.01 G-t -d Alarm Response Procedures Revision No. 127 I I

~\

\

(Panel G/19)

Group Heading REACTOR NEUTRON MONITORS I G-I-f HI-HI/INOP I CAUSES: SETPOINTS: ACTUATING DEVICES:

Core power exceeding predetermined level for the (.95 x Flow) + 60% RJ19A and RJ19B existing recirculation flow condition as described (For flow greater than

(.95 x Flow) +- 60°/0 7.68 x lo4 gpm)

(For flow greater than 7.68 x l o 4 gpm) (.98 x flow) + 37.7%

(-98 X flow) + 37.7% (For less than (For less than 7.68 x 1O4 gpm) 7.68 x l o 4 gpm)

Maximum value of 118% (Maximum Setpoint of or module inoperable, indicating mode switch on 118% power) or Reference Drawings:

APRM drawer not in operate position, module Module Inoperable.

removed, or more than three LPRM inputs bypassed. GE 237E566 Sh. 1A & 1B These are trip signal inF)utsto Reactor Protection GE 706E812 Sh. 19 & 22 System Channel I. GU 3E-611-17-009 Sh. 1 CONFIRMATORY ACTIONS:

For half scram signal, determine cause of trip by checking APRM r - --

recorders at Panel 4F and APRM cabinets at Panels 3R and 5 R . /'

\

4UTOMATIC ACTIONS:

?eactor scram coincident with Channel II trip.

MANUAL CORRECTIVE ACTIONS:

=or high APRM, reduce reactor power by inserting control rods in accordance with the rod sequence. For noperative unit, check the affected APRM cabinet for improper mode sviiitch position, module removed Status LE@

i n the FCTR Card is list solid red, or more than three LPRM inputs bypassed. Refer to Procedure 403. For full .

jcram condition, verify automatic actions and perform followup actions as defined in Procedure 2000-ABN-3200.01, Reactor Scram.

f failure of APRM channels results in conditions less conservative than those permitted by Technical W - ~

Specifications, shutdown the reactor. If all APRM indication is lost, manually scram the reactor per Procedure 2000-ABN-3200.01, Reactor Scram. Use IRMs and SRMs to monitor reactor power.

Subject Procedure No.

Page 1 of 1 NSSS 2000-RAP-3024.01 G-I-f Alarm Response Procedures Revision No. 127 (Panel G/35)

Group Heading REACTOR NEUTRON MONITORS G-3-f CAUSES: SETPOINTS: ACTUATING DEVICES:

Core power has exceeded the predeterminedlevel (0.90x 1o-6w) RJlSA, B, C, D for the existing recirculation flow condition. + 53.1 with a maximum value of 108%

Reference Drawings:

G E 706E812 Sh. 19,22, 26 & 29 GU 3E-613-17-009 Sh. 1 CONFIRMATORY ACTIONS:

w Verify high APRM at Panel 4F. control rods were not moved, check recirculation flow, feedwater flow and temperature, reactor presyfre, and reactor water level to determine causL,-

/f AUTOMATIC ACTIONS:

Rod withdrawal block. 1

\

MANUAL CORRECTIVE ACTIONS:

derify alarm transfer point has been reset at affected unit. Check APRM channel for LPRM failurk6a rod block setpoint is exceeded by 4%, power by inserting contmt-fods. Refer a0 Procedure 403 before sypassing an APRM Subject Procedure No.

Page 1 of 1 NSSS 2000-RAP-3024.01 G-3-f Alarm Response Procedures Revision No. 127 (Panel G/37)

Operator Actions ES-D-2 Op Test No.: ILT2004 Scenario No.: NRC #4 Event No.: 6 Page 6of 9 Event

Description:

Running RBCCW pump trips Cause: Breaker electrical problem causes pump trip Automatic Actions: none Effects: Operator action required to start standby pump to prevent reactor scram Time Position Applicant's Actions Or Behavior BOP Recognize condition by reporting alarms; C-4-C:PUMP 1-2 TRIP 0

IAW RAP C-4-c confirms:

0 pump status and system pressure on 13R 0 the running RBCCW pump has tripped 0 availability of standby RBCCW pump SRO 0 Direct start of st Direct Failure Response and repair the pump BOP CT 0 Starts standby RBCCW pump.

0 Monitor RBCCW and Service Water parameters 0 Refers to ABN-19, RBCCW Failure Response and 2000-OPS-3024.21, RBCCW System Diagnostic and Restoration procedure to determine follow-up actions

Group Heading R B C C W c c 4

CAUSES :

Overload trip of Reactor Building Closed Cooling Water Pump 1-2.

SETPOINTS:

Breaker tripped ACTUATING DEVICES:

30T Relay Overload trip setpoint is 300 amps Reference Drawings:

BR E1126 GU 33-611-17-005Sh. 1 CONFIRMATORY ACTIONS:

Check pump status on Panel 13R and system pressure at Panel 13R on PI-IA18.

AUTOMATIC ACTIONS:

NONE MANUAL CORRECTIVE ACTIONS:

1. Start any available RBCCW pump.

2. RPV Temperature is >212OF IF RBCCW to the drywell is not restored within one minute, THEN perform the following:

a) -

IF the reactor is in the Startup or Run modes, THEN scram the reactor in accordance with Proc. 2000-ABN-3200.01, Reactor Scram.

b) Trip all recirc pumps, and refer to Proc. 2000-ABN-3200.02, Recirculation Pump Trip.

C) Confirm that the suction and main discharge valves in at least 1 recirc loop are open.

d) Refer to Proc. 2000-ABN-3200.19, RBCCW Failure Response, and 2000-OPS-3024.21, RBCCW System Diagnostic and Restoration Actions.

3. RPV Temperature is I212OF Refer to Proc. 2000-ABN-3200.19, RBCCW Failure Response, and 2000-OPS-3024.21,RBCCW System Diagnostic and Restoration Actions.

Subj ect Procedure No.

Page 1 of 1 N S S S 2000-RAP-3024.01 c c Alarm Response Procedures Revision No: 126 (PANEL C)

Operator Actions ES-D-2 11 Op Test No.: lLT2004 Scenario No.: NRC #4 Event No.: 7 Page 7of 9 Event

Description:

Steam leak develops in the Drywell Cause: Main Steam line breaks Automatic Actions: none Effects: Operator action required to vent drywell and scram reactor prior to 3 psig Time Position Amlicants Actions Or Behavior Recognize condition by reporting; RO 0 Unidentified leak-rate change Bop 0 Containment pressure and temperature change C-3-f; DW PRESS HVLO 0 Reference RAP C-3-f, DW PRESS HVLO SRO 0 Direct venting of Containment IAW procedure 312.1 1, Section 4.3, Nitrogen System and Containment Atmosphere Control.

Direct monitoring of Containment and investigate potential in-leakage paths using 2000-OPS-3024.09, Drywell Cooling System Diagnostic procedure BOP Vent Containment IAW procedure 312.11 ii 0 Vent the drywell via the Torus by opening Torus vent valves V-28-47 and V-28-18 on panel 11F.

OR 0 Vent the drywell via the drywell by opening drywell vent valves V-23-21 and V-23-22 on panel 12XR.

RO Identify the cause of the high drywell pressure condition as directed.

BOP SRO Before drywell pressure reaches 3.0 psig, directs manual scram IAW ABN-1 scram pushbuttons on 4F before Drywell pressure to SHUTDOWN Report that all control have inserted on the scram Verifies power decrease 0 Inserts SRMs and IRMs

3roup Heading T O R U S / D R Y W E L L C f USES : SETPOINTS: ACTUATING DEVICES:

Iigh or low pressure in the Drywell, or High 1.4 psig Recorder 12XR-6

!ailed transmitter loop. LOW 1.0 psig (PT-51)

Reference Drawings:

SN 15081.02-ETLD-007 GE 112C2827 GU 33-611-17-005Sh. 1

ONFIRMATORYACTIONS:

erify abnormal pressure at Panel 4F, PI-622-1407. If digital indicator is lashing, most probable cause is a failure of PT-51. If this is the case, isregard indication on recorder 12RX-6.

UTOMATIC ACTIONS:

ONE ANUX CORRECTIVE ACTIONS:

i/,

etermine the cause of the hi/low Drywell pressure.\ Add Nitrogen or vent in (//

ccordance with Procedure 312.11, Nitrogen System and Containment Atmosphere ontrol as necessary to adjust pressure within normal operating band (greater than

.I# and less than 1 . 3 # ) . If Primary Containment requires venting, the potential xists for airborne activity to be higher than normal. Consideration should be given o vent through the Standby Gas Treatment System in accordance with Procedure 3309 - .-

tandby Gas Treatment System. Stack and Reactor Building radiation monitors shall e monitored whenever the PrimarylContainmentis vented.

f frequent or persistent high pressure alarms are occurring, check containment emperature, Torus water level A, Drywell Cooler operation (including RBCCW flow;;a amperature) and potential in-leakage paths to determine the cause of the high ressure condition. Refer to Procedure 2000-OPS-3024.09, Drywell Cooling System -

iagnostic & Restoration Actions, or 2000-OPS-3024.21, Reactor Building Closed

>ding Water System - Diagnostic and Restoration Actions.

E frequent or persistent low pressure alarms are occurring, check for proper

?eration of the Nitrogen Makeup System, Drywell Cooling System, and check mtainment valve lineup to determine the cause of the low pressure condition. Refer 3 Procedure 2000-OPS-3024.06,Containment Ventilation System - Diagnostic &

?storation Actions.

m: Any leak resulting in an increase in Torus level shall be calculated and added to the Unidentified Leakage Rate, in accordance with Procedure 106.

&ject Procedure No.

Page 1 of 1 N S S S 2000-RAP-3024.01 c f Alarm Response Procedures Revision No: 126 I

(PANEL C)

Operator Actions ES-Dd ii Op Test No.: ILT2004 Scenario No.: NRC #4 Event No.: 8 Page 8of 9 Event

Description:

Containment Spray Valve fails to realign automatically when sprays are required Cause: Breaker malfunction prevents valve movement Automatic Actions: none Effects: System configuration does not automatically realign. Operator action required to manually open valve Time Position Applicants Actions Or Behavior SRO Enter and execute EOP 3200.02, Primary Containment Control when Drywell Pressure exceeds 3 psig.

Direct lineup of Containment Spray in the Drywell Spray Mode per Support CT Procedure 29, Initiation of the Containment Spray System for Drywell Sprays BOP IAW SP-29; Lineup Containment Spray in the Drywell Spray mode 0 Report failure of V-21-11, DW Spray Discharge Valve, to open CT Direct Equipment operator to manually open V-21-11 Spray Drywell when V-21-11 is open and conditions for spraying Drywell are met Must initiate containment spray before 281 e -FO V e s s U W h a d w k RO Control Reactor Level and Pressure as Directed TERMINATION CRITERIA: Once Drywell spray has been initiated and Drywell pressure is being controlled between 4 - 12 psig, or at the discretion of the lead evaluator, the scenario may be terminated POST SCENARIO EMERGENCY CLASSIFICATION: Declares an ALERT due to torus pressure r 12 psig EAL:

Procedure EMG-3200.02 Support Proc. 29 Rev. 16 Attachment G Page 1 of 4 SmPORT PROCEDURE 29 INITIATION OF THE CONTAINMENT SFRAY SYSTEM FOR DRYG?XLL SPRAYS 1.0 PREREOUISITES Manual initiation of Drywell Sprays has been directed by the Emergency Operating Procedures.

2.0 PREPARATION 2.1 Select the Containment Spray S y s w to be used by confirming either SYSTEM 1 MODE SELECT or SYSTEM 2 MODE SELECT switch in DW SPRAY position (Panel 1 F / 2 F ) ) ,

2.2 Verify that the system TORUS CLG DISCHARGE valve closes and DW SPRAY DISCHARGE valve opens (-el 1F/2F).

3.0 PROCEDURE Containment Spray suction strainer plugging may occur due to debris in the Primary Containment and result in a loss of Containment Spray System Flow.

Diesel Generator overload will result if a containment Spray pump and ESW pump are started with a Diesel Generator load of greater than 2 8 6 0 XW.

IF Bus 1C or ID are being suppnied by an Emergency Diesel Generator, THEN verify that adequate load margin is available so as NOT to exceed EDG load limitt when starting Containment Spray and ESW pumps.

OVER

( 3 2 0 0 0 2/ 9 ) 337-1

. -..i Procedure ELPG-3200.02 Support P r o c - 29

. . . Rev. 16

.i . Attachment 6

.. . . Page 2 of 4 3.2 -

WHEN directed to initiate Drywell sprays, THEN complete the following:

-1. Confirm all Reactor Recirculation Pumps tripped.

2. Confirm the Drywell Recirc Fans tripped (Panel 11R).

3-3 CAUTION NPSH problems will develop on all operating pumps if more than 4 Containment Spray/Core Spray Main pumps are operated at the same time.

IF 4 Containment Spray/Core Spray Main pumps are in operation, THEN do not start additional Containment Spray pumps until Core Spray Main pumps can be secured.

3.4 Start a Containment Spray Pump as follows:

3.4.1 Select a Containment Spray Pump to be started.

3.4.2 Place and hold the System Pump Start Permissive Keylock for the selected pump in the appropriate position (Panel 1 F / 2 F ) .

3.4.3 Start the selected Containment Spray Pump using its control switch (Panel 1 F / 2 F ) .

3.5 Start an associated ESW Pump using its control switch (Panel 1 F / 2 F ) .

3.6 CAUTION Operation of Containment Spray pumps with flow above the NPSH or vortex limits may result in equipment damage. When operating beyond any flow limits, periodic evaluations should be made to verify that continued operation beyond these limits is still required.

c 4 Monitor System parameters for expected performance.

L A (320002/9) E7-2

Procedure EMG-3200.02 Support Proc. 29 Rev. 16 Attachment G Page 3 of 4 3.7 NOTE Valves V-5-147, 166 and 167 240 not seal in when the control switch is taken to CLOSE. The control switch must be held in CLOSE until the valve indicates closed.

Confirm the following RBCCW Isolation valves closed (Panel I F / ~ F :

)

v-5-147 V-5-148 V-5-166 V-5-167 3.8 Diesel generator overload will result if a Containment Spray pump and ESW pump are started with a Diesel Generator Start additional Containment Spray and ESW Pumps in accordance 3.9 with Steps 3.3 through 3.6 as.directed by the LOS.

IF while performing the following steps, Containment Spray 1

L Pumps fail to trip, THEN place the respective syszem MODE SELECT switch in TORUS CLG position.

3.10 Maintain primary containment pressure in a band of 4 to 12 p s i g unless otherwise directed by the LOS as follows:

3.10.1 Secure Drywell. Sprays when Drywell pressure drops 1

to 4 psig.

3.10.2 WHEN Torus or Drywell pressure increases to 12 psig, THEN initiate Drywell sprays in accordance with i

Steps 3.3 through 3.6.

3.11 E any Core Spray Booster pump is running AND Torus Drywell pressure drops to 2 psig, THEN confirm termination of Dnryylell Sprays due to NPSH concerns.

I i,

OVER I

(320002/9) E?- 3

Procedure EMG-3200. 02 Support Proc. 2 9 R e v . 16 Attachment G Page 4 of 4 3.12 IF no Core Spray Booster pump is running Torus OR Drywell pressure drops below 1 psig, THEN confirm termination of Drywell Sprays to prevent deinertion of the Primary Containment.

(320002/9) E7-4

Group Heading C O R E S P R A Y 1 B e SETPOINTS: ACTUATING DEVICES:

High pressure differential across Core 0.3 2 0.3 psid D P I S RV30A Spray System 1 sparger nozzles due to Core Spray line break in the vessel annulus.

Reference Drawings:

GE 148F712 GE 885D781 GE 112C2845 Sh. 3 GU 33-611-17-004 Sh. 2 CONFIRMATORY ACTIONS:

Verify pressure differential at instrument rack RK04.

AUTOMATIC ACTIONS :

None MANUAL CORRECTIVE ACTIONS:

If instrument reading is greater than or equal to 1 psid, consider Core Spray System 1 inoperable. Verify operability of System 2.

Notify Licensed Operations Supervisor. Core APLHGR must be brought within 90% of limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Contact Core Engineering by referencing the Core Maneuvering Daily Instructions for guidance on rod movement and power changes.

Subject Procedure No.

Page 1 of 1 N S S S 2000-RAP-3024.01 I B e Alarm Response Procedures Revision No: 127 (PanelB/28)

=up Heading REACTOR NEUTRON MONITORS G-6-f ACTUATING DEVICES:

RJ18 A, B,C, D RJ19 A, 6, C, D Reference Drawings:

GE 706E812 Sh. 17, 18,20,21, 24,25,27,28,31,32,33,34, 35,36,37 and 38 GU 3E-611-17-009 Sh. 2 CONFIRMATORY ACTIONS:

Determine which detector is high at Panel 4F. Insure made switch on amplifier of LPRM is in OPERATE position.

If necessary, verify high flux by taking a TIP traoe of theaffected LPRM string as directed by the LOS.

AUTOMATIC ACTIONS:

Ifthe LPRM is an input to an APRM channel, there can b e auto action (Le., APRM Rod Block, and/or APRM High)

MANUAL CORRECTIVE ACTIONS:

Check that fuel limits are not being exceeded. Refer to Procedure 403 before bypassing an inoperative LPRM.

An alarm may be reset for a chamber that has momentarily exceeded the setpoint by depressing the reset button of the affected LPRM at Panels 3R and 5R. Cmtact Come Engineering for additional guidance as necessary.

Subject Procedure No.

Page 1 of 1 NSSS 2000-RAP-3024.04 I G-6-f Alarm Response Procedures Revision No: 127 I I (Panel G/40)

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. NRC #2 Op Test No. ILT2004 Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor at 99% power with the ACRD pump out of service. The crew Summary will begin by placing an alternate Reactor Building HVAC fan in service and secure the running fan. A Drywell recirculationfans trips and an alternate fan will be started. The RBHVAC ventilation radiation monitor will fail upscale, causing RBHVAC to trip, but the SGTS will fail to start. The crew will start the SGTS manually. A control rod drifts out and it will be restored to its programmed position. The running Service Water pump trips requiring the standby pump to be started. The rod drift will cause a small fuel failure. Power will be reduced to lower radiation levels. A leak in the Torus will require the Reactor to be scrammed and eventually this will lead to Emergency Depressurization. Five rods will fail to insert on the scram.

Event Malfunction Event Event No. No. Type* Description 1 SRO N BOP Swap Reactor Building HVAC Supply fans 2 MAL-PCN004D C SRO Drywell recirc fan trips 1 BOP I I SRO I 3 MAL-RMS005M I RBHVAC Rad Ventilation monitor Fails Upscale, RBHVAC trips, SGTS MAL-SCN005 BOP fails to start. (Tech Spec)

SRO 4 MAL- C RO Control Rod Drifts Out CRD005-2239 .B8B SRO 5 MAL-SWS001B C BOP Running Service Water pump trips MAL-RXS001, SRO 6 .00075,120s R RO Small Fuel Failure leads to Power Reduction to Lower Radiation

-B8p Levels MAL-CSS001A, SRO 7 8000.900s M RO TorusWaterLeak BOP MAL-CRD022 SRO 8 -1039, -421 1, C RO Five Rods Fail to Insert on the Scram (power > 2%)

-2635, -1 423,

-34.43

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

Operator Actions ES-D-2 u Op Test No.: ILT2004 Scenario No.: NRC #2 Event No.: 1 Page 1 of 8 Event

Description:

Swap Reactor Building HVAC Supply fans Cause: Equipment rotation Automatic Actions: none Effects: none Time Position Aeplicant's Actions Or Behavior SRO Direct that RBHVAC fans be swapped IAW Operating Procedure 329, Reactor Building Heating, Cooling and Ventilation System, section 8.3 BOP IAW 329, section 8.3; Dispatch Equipment Operator to Office Building roof to CONFIRM manual inlet dampers are OPEN for fan SF-1-13 and align SF-1-13 discharge dampers approximately equal to SF-1-12 dampers Start RBHVAC supply fan SF-1-13 by placing its control switch to ON at 11R Stop RBHVAC supply fan SF-1-12 by placing its control switch to OFF at 11R Verify RB dP is less than -0.25 in WG

Operator Actions ES-D-2 L/ Op Test No.: lLT2004 Scenario No.: NRC #2 EventNo.: 2 Page 2of 8 Event

Description:

Drywell recirc fan trips Cause: Motor malfunction Automatic Actions: None Effects: Operator action required to start alternate fan Time Position Amlicant's Actions Or Behavior BOP Recognize condition by reporting alarms; 0 L-4-a: RF 4 TRIP IAW RAP L-4-a confirms:

0 trip of RF-1-4 0 availability of RF-1-3.

SRO Directs start of Recirc Fan 1-3 BOP Places drywell Recirc Fan 1-3 in service RO Monitors and reports drywell parameters including pressure and temperature SRO Refers to procedure 2000-OPS-3024.09 - Drywell Cooling System - Diagnostic BOP and Restoration Actions to determine cause of equipment problem

~

Group Heading L a H & V D W R E C I R C F A N S R F 4 T R I P CAUSES : SETPOINTS: ACTUATING DEVICES:

Overload trip of Drywell Recirculating 39.5 amps OLX Relay Fan RF-1-4.

Reference Drawings:

GE 157B6350, Sh 96B BR 2011, Sh 1 GU 33-611-17-013 I

CONFIRMATORY ACTIONS:

Verify fan trip at Panel 11R.

AUTOMATIC ACTIONS:

Trip of RF-1-4.

MANUAL CORRECTIVE ACTIONS:

Place any available Drywell Recirculating Fan in service. Refer to Procedure 2000-OPS-3024.09, Drywell Cooling System - Diagnostic and Restoration Actions. If more than one Drywell Recirculating Fan has tripped, monitor Drywell pressure and temperature closely. During normal power operation, four recirculating fans are required to maintain Drywell temperature.

Subject I Procedure No.

B O P 2000-RAP-3024.03 L a Alarm Response procedures

Operator Actions ES-D-2

'u Op Test No.: lLT2004 Scenario No.: NRC #2 Event No.: 3 Page 3of 8 Event

Description:

RBHVAC Rad Ventilation monitor Fails Upscale, RBHVAC trips, SGTS fails to start (Tech Spec)

Cause: Instrument failure causes upscale response Automatic Actions: RBHVAC trips

-Effects: Operator action required to manually initiate SGTS Time Position Applicant's Actions Or Behavior RO Recognize condition by observing indications or reporting alarms; 1OF-l-f: VENT HI 0 Verify high radiation level on redundant indicators on Panel 2R IAW RAPS, confirm Reactor Building isolation and trip of RBHVAC and initiation of Standby Gas Treatment System [SGTS].

Verify that an actual ventilation high radiation condition does NOT exist SRO Recognize and report that the expected start of SGTS did NOT occur.

BOP RO SRO Direct SGTS be placed in service manually IAW procedure 330 0 Evaluate compliance with TS 3.5 0 Can remain in operation for 7 days if remaining system is operable.

0 Notify Work Management to troubleshoot and repair the instrument.

BOP IAW procedure 330, take the following actions when directed Confirm Standby Gas Select switch to SYS 1 on panel 11R 0 Place Exhaust Fan EF-1-8 to HAND on 11R 0 Verify EF-1-8 starts, and valves V-28-23, 24 & 26 open 0 After flow is established, verify V-28-24 closes and V-28-28 opens 0 Place V-28-48 control switch to CLOSE and verify GREEN close light LIT 0 Verify RBHVAC secured if directed by supervisor

Operator Actions ES-D-2 u Op Test No.: ILT2004 Scenario No.: NRC #2 Event No.: 4 Page 4of 8 Event

Description:

Control Rod 22-39 Drifts Out(/k%ht 9J"s/ .

Cause: Relay malfunction causes rod motion Automatic Actions: Rod Drift annunciator alarms Effects: Requires operator action to reduce reactor power Time Position Amlicant's Actions Or Behavior RO Recognize condition by observing indications or reporting alarms; 0 H-6-a: ROD DRIFT 0 Confirm only one rod drifting out IAW RAPS; confirm automatic actions and indications including control rod identification and direction of movement SRO Direct implementation of ABN-6, Abnormal Control Rod Motion Direct the rod to be selected and returned to its programmed position Monitor for indications of fuel failure 0 Notify the Reactor Engineers of abnormal Control Rod Motion b

CT RO Select rod and drive to its programmed position.

BOP IAW ABN-6; Notifies Reactor Engineering Request Reactor coolant sample from Chemistry Monitors Off-Gas and Main steam line radiation Reports observed indications to SRO

'1

Operator Actions ES-D-2 Op Test No.: ILT2004 Scenario No.: NRC #2 Event No.: 5 Page 5 o f 8 Event

Description:

Running Service Water pump trips Cause: Motor problem causes pump trip Automatic Actions: none Effects: Operator action required to start standby pump Time Position Atmlicant's Actions Or Behavior BOP Recognize condition by reporting alarms; K-1-f: SVC WATER PUMP TRIP 0

IAW RAP K-1-f confirms:

0 the running Service Water pump has tripped 0 availability of standby Service Water pump

'i, SRO 0 Direct start of standby Service Water pump' 0 Direct reference to ABN-18, Service Water Failure Notify Work Management to troubleshoot and repair the pump BOP 0 Starts standby Service Water pump.

0 Monitor RBCCW and Service Water parameters 0 Refers to ABN-18, Service Water Failure

Operator Actions ES-D-2 L,

- Op Test No.: ILT2004 Scenario No.: NRC #2 Event No.: 6 Page 6of 8 Event

Description:

Small Fuel Failure leads to Power Reduction to Lower Radiation Levels Cause: Caused by abnormal control rod motion Automatic Actions: none Effects: Main steam and Off-Gas radiation levels increase. Operator action required to lower power to mitigate failed fuel affects. Reactor power, steam flow and feed flow decrease.

Time Position Applicant's Actions Or Behavior BOP Recognize condition by reporting alarms; 1OF-1-k: AREA MON HI 1OF-1-c: OFF GAS HI-HI 1OF-2-C:OFF GAS HI 1OF-1-d: STACK EFFLUENT HI-HI 10F-2-d: STACK EFFLUENT HI IAW RAP 1OF-1-k confirms:

0 Verify radiation levels on panel 2R 0 Check Main Steam and Off Gas radiation monitors IAW RAPS;confirm radiation levels and trends indicating failed fuel.

SRO Direct implementation of ABN-26, Increase in Main Steam Line/Off Gas activity BOP Monitors Off-Gas and Main steam line radiation Request Off-Gas sample from Chemistry Request guidance from Reactor Engineering NOTE Reactor Engineering may prompt a power reduction in response to the crew's request for assistance.

SRO Direct reduction of radiation levels by lowering Reactor power, using recirculation flow IAW procedure 202.1, Power Operations RO IAW procedure 202.1 ;

0 Reduce reactor power with recirculationflow as required BOP Monitors Off-Gas and Main steam line radiation

Operator Actions ES-D-2

- Op Test No.: lLT2004 Scenario No.: NRC #2 Event No.: 7 Page 7of 8 Event

Description:

Torus Water Leak Cause: Torus piping failure Automatic Actions: none Effects: Torus level decrease. Increasing Secondary Containment Water Levels. Operator action is required to mitigate the Torus level decrease.

Position Applicants Actions Or Behavior BOP Recognize condition by observing indications or reporting alarms; 0 Torus Water level decrease 0 C-5-e: TORUS LEVEL HI/LO IAW SDRP or RAPS; confirm automatic action and indications including Torus water level at panel 11F and 16R, confirm Torus intact, direct inspection of Reactor building corner rooms.

SRO 0 Announce Entry into EOP 3200.02, Primary Containment Control due to low BOP Torus level RO 0 Announce Entry into EOP 3200.11, Secondary Containment Control due to L water levels in the Secondary Containment SRO Direct actions IAW EOP 3200.02, Primary Containment Control:

Direct adding water to the Torus using Fire Water per Support Procedure 37 0 When water level can not be maintained above 110 inches, then Enter EOP 3200.01 A, RPV Control - No ATWS at A and perform it concurrently 0 Recognize that not all rods fully inserted on the scram BOP Attempt to restore Torus level using the Core Spray System per Support Procedure 37 when directed.

Close V-20-91 Open V-20-83 Place CSS #1 Main pump in PTL Place breaker off for V-20-27 Open V-20-27 Spray Drywell when lined up and >12 psig RO IAW ABN-1:

e 0 Depress both manual scram pushbuttons on 4F before 110 inches Torus Level 0 Place Reactor mode switch to SHUTDOWN Report that not all control have inserted on the scram NOTE: most actions will be driven by the Primary Containment Control procedure before they are reached in Secondary Containment Control.

Operator Actions ES-D-2 L, Op Test No.: ILT2004 Scenario No.: NRC #2 Event No.: 8 Page 8of 8 Event

Description:

Five Rods Fail to Insert on the Scram Cause: CRD malfunction causes some rods not to insert Automatic Actions: none Effects: Operator action required to insert control rods Time Position Applicant's Actions Or Behavior RO 0 Report that not all rods fully inserted on the scram.

L Before water level drops to 110 inches, direct an Emergency Depressurization.

Perform the following actions for EOP for RPV Control -With ATWS when directed:

Power Control 0 Initiation of Alternate Rod Injection Bypass ROPS n Manually drive control rods, Close V-15-52 0 LeveVPower control 0 Bypass the following Initiations and Isolations 0 ADS 0 MSlV Low-Low Water Level Isolation 0 RBCCW Drywell Isolation 0 Use Support Procedure - 17 to control level as directed RO Initiate ARI when directed

CT 0 Open all EMRVs when directed RO IAW EOPs & ABN 01; BOP 0 Control Reactor level 0 Control Reactor pressure 0 Perform remaining scram actions TERMINATION CRITERIA: When the Emergency Depressurization is in progress, or at the discretion of the lead evaluator, the scenario may be terminated POST SCENARIO EMERGENCY CLASSIFICATION: ALERT: Scram signal received and Rx power remains

>2%

EAL: C-1 New EAL MA4