Text

December 2002-301 Exam Final SRO Written Exam with Answers
	ML030210512
Person / Time
Site:	Browns Ferry
Issue date:	08/16/2003
From:	Ernstes M; Operator Licensing and Human Performance Branch
To:	Scalice J; Tennessee Valley Authority
Shared Package
ML032280265	List: ML030210512;
References
	50-259/02-301, 50-260/02-301, 50-296/02-301 50-259/02-301, 50-260/02-301, 50-296/02-301
	Download: ML030210512 (206)
	v • d • e

Final Submittal BROWNS FERRY EXAM 2002-301 50-259, 50-260, & 50-296 DECEMBER 13,16-19, 2002 Senior Operator Written Examination

QUESTIONS REPORT for SRO 2002-301 LXRBank

1. 201001A2.01 001 While operating at 25% power the Unit Operator reports the following:

- CRD Pump A Breaker Disagreement White Light.

- Motor Trip Out Annunciator and Horn Received.

- 1B CRD Pump in Standby.

Which ONE of the following describes the IMMEDIATE actions to be taken for the above conditions?

A. Start CRD pump 1B and open the CRD PUMP DISH TO UNIT 2. Adjust the CRD SYS FLOW CONTROLLER tapeset to adjust CRD cooling water header differential pressure to 20 psid and CRD system flow to approximately 60 gpm.

B. Place CRD SYS FLOW CONTROLLER in MAN at MAXIMUM setting. Start CRD pump lB. Open CRD PUMP DISCH TO UNIT 2. When CRD cooling water header differential pressure reaches 20 psid, and CRD system flow reaches between 45 and 75 gpm, balance CRD SYS FLOW CONTROLLER and place in AUTO.

C. Start CRD pump 1B. Place CRD SYS FLOW CONTROLLER in MAN at MINIMUM setting. Open CRD PUMP DISCH TO UNIT 2. Adjust CRD SYS FLOW CONTROLLER to establish 20 psid cooling water header differential pressure.

Balance CRD SYS FLOW CONTROLLER and place in AUTO.

DO Place CRD SYS FLOW CONTROLLER in MAN at MINIMUM setting. Start CRD pump 18. Open CRD PUMP DISCH TO UNIT 2. Adjust CRD SYS FLOW CONTROLLER to establish 20 psid CRD cooling water header differential pressure and CRD system flow between 40 and 65 gpm. Balance CRD SYS FLOW CONTROLLER and place in AUTO.

RO Tier: T2GI SRO Tier: T2G2 Keyword: CRD SYSTEM Cog Level: MEM 3.2/3.3 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:45 AM 1

QUESTIONS REPORT for SRO 2002-301 LXRBank

2. 201006K3.01 001 The following conditions exist on Unit 3:

Control Rod 22-35 at position 18 (group limit 00-12)

Core Power level is above the Low Power Setpoint The RWM program has NOT been initialized after being unbypassed Which ONE of the following is the reason that control rod 22-35 cannot be moved?

A. Withdraw Block is in effect.

B. Insert Block is in effect.

C. Select Block is in effect.

D. Withdraw Error has occured.

References:

OPL171.024 Rev. 10 pg 13-16 A, B and D are incorrect since a Select Block is in effect due to RWM Bypass Switch in NORMAL and the RWM program has not been initiated.

C. Correct answer.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: RWM Cog Level: C/A 3.2/3.5 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:45 AM 2

QUESTIONS REPORT for SRO 2002-301 LXRBank

3. 202002G2.2.3 002 Which ONE of the following choices correctly describes the response of the Unit 2 and Unit 3 reactor recirculation pump (RRP) speed control (Speed Feedback is enabled) to an increase in core differential pressure?

A. On Unit 2 the RRP speed must be manually adjusted by the operator but Unit 3 will automatically reposition the scoop tube to bring speed back to the setpoint if

-generator speed has changed by more than-5-rpm.

B. Both Unit 2 and Unit 3 must be manually adjusted by the operator to bring speed back to the setpoint.

C. Unit 2 will automatically reposition the scoop tube to bring speed back to the setpoint but on Unit 3 the RRP speed must be manually adjusted by the operator if generator speed has changed by more than 5 rpm.

Df Both Unit 2 RRP and Unit 3 will automatically reposition the scoop tube to bring speed back to the setpoint if generator speed has changed by more than 5 rpm.

References:

OPL171.007 Rev.20 pg 44 A, B and C are incorrect since both units recirc pumps will change speed automatically if generator speed has changed by more than 5 rpm.

D. Correct answer.

RO Tier: T2G I SRO Tier: T2G1 Keyword: RECIRC SYSTEM Cog Level: MEM 3.1/3.3 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:45 AM 3

QUESTIONS REPORT for SRO 2002-301LXRBank

4. 203000A4.01 001 During a level transient on Unit 2 the following events occurred:

- RPV water level decreased to -125 inches during the transient

- ADS actuated

- RHR Pump 2A and 2B started and injected to the reactor vessel

- RPV water level is now +25 inches and increasing

- -No operator actions have been taken Which ONE of the following statements describes the RHR system response if RHR Pump 2A control switch is placed to the STOP position?

A' RHR Pump 2A will stop and the amber auto-start lockout light will light.

B. RHR Pump 2A will stop and the amber auto-start lockout light will extinguish.

C. No change; RHR Pump 2A will continue to run until the LOCA initiation signal is reset.

D. RHR Pump 2A will stop and then restart when the switch is released. The amber auto-start lockout light will not change indication.

References:

OPL171.044 Rev. 10 pg 61 Enabling Objective #13 2-01-74 Rev. 0107 pg 8 A. Correct answer.

B. Incorrect since the RHR system is designed to allow a pump to be secured and auto-initiation lock-out.

C. Incorrect since the amber light is the auto-init. lockout indication and will not extinguish until the LOCA signal is reset.

D. Incorrect since both sentences are incorrect.

RO Tier: T2G1 SRO Tier: T2GI Keyword: RHR Cog Level: C/A 4.3/4.1 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:45 AM 4

QUESTIONS REPORT for SRO 2002-301LXRBank

5. 203000K3.02 001 Gross fuel failure is suspected on Unit 3. The crew is in 3-EOI Appendix 18 Suppression Pool water Inventory Removal and Makeup and have just closed 3-FCV-74-63;RHR RADWASTE SYS FLUSH VALVE. Suppression Pool level is -2.5 inches and steady.

Which ONE of the.following are the appropriate actions?

Af Exit 3-EOI Appendix 18 - Suppression Pool water Inventory Removal and Makeup since Suppression Pool water level is within acceptable limits.

B. Open 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE and direct Suppression Pool water to Radwaste ONLY.

C. Open 3-FCV-74-62, RHR MAIN CNDR FLUSH VALVE and direct Suppression Pool water to the Main Condenser ONLY.

D. Open 3-FCV-74-6Z RHR MAIN CNDR FLUSH VALVE and direct Suppression Pool water to the Main Condenser or open 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE and direct Suppression Pool water to Radwaste.

References:

3-EQI Appendix 18 WHEN Suppression Pool level can be maintained between -1 in. and -5.5 in. THEN EXIT this procedure.

RO Tier: T201 SRO Tier: T2GI Keyword: SUPPRESSION CHAMBER Cog Level: C/A 3.5/3.5 Source: B Exam: BF02301 Test: C Mise: TCK Tuesday, January 21, 2003 07:30:46 AM 5

QUESTIONS REPORT for SRO 2002-301LXRBank

6. 204000K5.05 001 Which ONE of the following sets of signals will cause the RWCU Blowdown Valve (FCV 69-15) to close automatically?

A. Low Reactor Water Level +2", Standby Liquid Control initiation.

B! High downstream pressure 140 psig, low upstream pressure 5 psig.

C. High RWCU Pump Rm temp 140'F, high temp on outlet of NRHX 140 0 F.

D. High flow 250 gpm, high differential pressure across valve 25 psid.

References:

OPL171.013 Rev.12 pg 22-24 A,C and D Incorrect since these signals do not close the blowdown valve.

B. Correct answer.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: RWCU SYSTEM Cog Level: MEM 2.6/2.6 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 6

QUESTIONS REPORT for SRO 2002-301 LXRBank

7. 205000A4.05 001 Unit 2 is in a refueling outage with Loop II of RHR in shutdown cooling. The RHR SYSTEM II MIN FLOW INHIBIT switch is in the INHIBIT position. The Unit Operator then places the RHR Loop IIMinimum Flow Valve (2-FCV-74-30) Control Switch to the OPEN position.

Which ONE of the following describes the effect on the Minimum Flow Valve?

A. Valve would not open.

BY Valve would open then immediately go back closed.

C. Valve would open regardless of RHR flow and remain open.

D. Valve would open only if RHR flow was less than min flow closing setpoint.

References:

OPLI171.044 Rev. 10 pg 33 and 34 2-01-74 Rev. 107 pg 73 Enabling Objective OPLI71.044 #10 A. Incorrect since valve would open and immediately close.

B. Correct answer.

C. Incorrect since valve would not remain open.

D. Incorrect since valve would open regardless of min flow signal.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: RHR SYSTEM Cog Level: MEM 3.2/3.2 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 7

QUESTIONS REPORT for SRO 2002-301 LXRBank

8. 206000A3.05 001 HPCI is operating in the pressure control mode (suction from the CST and return to the CST through FCV 73-35 and 36) when reactor water level lowers to -50".

Which ONE of the following describes HPCI response?

A. HPCI will be unaffected and continue to operate in the pressure control mode.

B. FCV 73-44 (inboard injection valve) opens; FCVs 73-35 and 36 remain open; HPCI does not inject to the reactor.

C. FCV 73-44 (inboard injection valve) opens: FCVs 73-35 and 36 close; HPCI injects to the reactor.

D. FCV 73-44 (inboard injection valve) opens; FCV 73-35 closes; FCV 73-36 remains open; HPCI injects to the reactor.

Reference:

OPLI171.042 Rev. 16 pg 42 A. Incorrect since HPCI has received an initiation signal from low water level. Setpoint is -45".

B. Incorrect since FCV's 73-35 and 36 receive a closed signal if they were open.

C. Correct answer.

D. Incorrect since both FCV's 73-35 and 36 receive a closed signal.

RO Tier: T2G1 SRO Tier: T2G1 Keyword: HPCI Cog Level: C/A 4.3/4.3 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 8

QUESTIONS REPORT for SRO 2002-301 LXRBank

9. 209001K5.05 001 A PMT is required on the vent system of Core Spray System I. The WO requires the Loop I Core Spray vent valves to be opened to verify the solenoid operated valve

--works as expected after replacement of the electrical solenoid.

Which ONE of the following describes the effect of opening the vent valves on the Core Spray system?i , ,-.

A. Core Spray System I will be inoperable as a result of venting however System II will not be affected.

B. Both Core Spray system will be inoperable as a result of venting.

C. Core Spray System operability will not be affected as long as CS & S is the ONLY source aligned to the Core Spray system.

D' Core Spray System operability will not be affected as long as discharge pressure meets the requirements of the TRM.

References:

Tech Spec 3.5.1, ECCS-Operating 2-SR-3.5.1.1 (CS 1) Rev. 1 pg 4 and 6 A. Incorrect since the grace period for the surveillance hasn't expired.

B. Incorrect since the grace period for the surveillance is 7.75 days and not 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. Incorrect since by the time that the 7.75 days expire then HPCI will be OPERABLE and a 7 day LCO entered.

D. Correct answer since HPCI will be OPERABLE by the time the 7.75 days has expired.

RO Tier: T2GI SRO Tier: T2GI Keyword: CORE SPRAY Cog Level: C/A 2.5/2.5 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 9

QUESTIONS REPORT for SRO 2002-301 LXRBank

10. 211000K1.01 001 Which ONE of the following describes the relationship between the SLC System and the Core Spray System?

Af The SLC sparger provides a sensing point for the Core Spray Break Detection logic.

B. The SLC sparger provides a sensing point for the Core Spray flow indication.

C. The Core Spray System is totally independent of the SLC System.

D. The same Shutdown Board powers the 2B SLC Pump and the 2B Core Spray Pump.

References:

OPI171.045 Rev.11 pg 13 OPL1 71.039 Rev.13 pg 14,26 and 27 Enabling Objective OPLI 71.039 #4 A. Correct answer.

B. Incorrect since the sparger has no input to Core Spray flow.

C. Incorrect since Core Spray does interact with SLC through the sparger.

D. Incorrect since 2B Core Spray is powered from SD BD "C" and 2B SLC pump is powered from SD BD "B".

RO Tier: T2GI SRO Tier: T2G I Keyword: SLC Cog Level: MEM 3.0/3.3 Source: N "Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 10

QUESTIONS REPORT for SRO 2002-301LXRBank

11. 211000K6.03 001 Which ONE of the following describes the power supply and interlocks of the SLC pumps?

A. One pump is powered from 250V RMOV Board A and one from 480V Shutdown Board B. The pumps are electrically interlocked so that both pumps run, if available.

B.! One pump is powered from 480V Shutdown Board A and one from 480V Shutdown Board B. The pumps are electrically interlocked so that only one pump will run at a time.

C. One pump is powered from 250V RMOV Board A and one from 480V Shutdown Board B. The pumps are electrically interlocked so that only one pump will run at a time.

D. One pump is powered from 480V Shutdown Board A and one from 480V Shutdown Board B. The pumps are electrically interlocked so that both pumps run, if available.

250 VDC is control power for the valves.

Two 100% capacity, triplex, positive displacement piston pumps powered from 480V Shutdown Bds A and B respectively are installed in parallel. The pumps are electrically interlocked so that only one pump can be run at a time to prevent overpressurization of the system.. This is accomplished by B-finger contacts in the start circuit of the running pump, opening contacts in the start circuit of the idle pump.

RO Tier: T2G1 SRO Tier: T2G1 Keyword: SBLC Cog Level: MEM 3.2/3.3 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:46 AM 11

QUESTIONS REPORT for SRO 2002-301 LXRBank

12. 212000A1.08 001 Unit 3 scrammed due to a spurious Group 1 isolation. The Mode Switch is in Shutdown and all rods are inserted. Reactor water level has been restored to the normal operating band. The Unit Supervisor has ordered the Reactor Operator to reset the scram.

Which ONE of the following describes the status of the Backup Scram Valves when the

-Reactor Operator moves the "Reset" switch to the right?

Both Backup Scram Valves should be...

A. energized and OPEN.

Bf de-energized and CLOSED.

C. energized and CLOSED.

D. de-energized and OPEN.

References:

OPL171.028 Rev.13 pg 22 A,C and D are incorrect since the Backup Scram Valves should be de-energized and CLOSED.

B. Correct answer.

RO Tier: T2GI SRO Tier: T2G1 Keyword: RPS Cog Level: MEM 3.4/3.4 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:46 AM 12

QUESTIONS REPORT for SRO 2002-301 LXRBank

13. 214000K4.01 001 I

Which ONE of the following statements is describes the operation of the Rod Position Information System (RPIS)?

A! If both of the S52 and SOO normal full-in reed switches are closed the full core display will be backlit green and display 00.

B. The S48 full-out digital display reed switches also supply rod position input signals to the "CONTROL ROD OVERTRAVEL" alarm.

C. On an uncoupled control rod, the full core display will show position 49 and no red backlight if the rod is withdrawn to the overtravel position.

D. When a CRD is driven beyond the full-in position the S51 over-travel reed switch will be actuated. The full-core digital display for that rod will display 00 and be backlit green.

References:

OPL171.029 Rev. 9 pg 19 and 20.

A. Correct answer.

B. Incorrect since the S50 switch provided indication for Rod Overtravel.

C. Incorrect since there is no position indication for an uncoupled control rod.

D. Incorrect since overtravel beyond full-in is --.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: RPIS SYSTEM Cog Level: MEM 3.0/3.1 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:47 AM 13

QUESTIONS REPORT for SRO 2002-301LXRBank

14. 215004K4.01 001 A reactor startup is in progress on Unit 2 with the following conditions:

Mode Switch is in START/HOT STBY IRM A is on range 2 with all other IRM's on range 3 The SRM's are partially withdrawn SRM count rate ranges between 80 and 90 cps The Reactor Operator attempts to withdraw control rod 24-33 but it will not move.

Which ONE of the following is the reason why the rod cannot be withdrawn?

A. SRM Downscale rod block.

Bf SRM Detector Wrong Position rod block.

C. SRM Hi rod block.

D. SRM Inop rod block.

Re .ferences: OPL1 71.019 Rev.6 pg 21 and 22 Enabling Objective OPLI171.019 #8 A. Incorrect since the SRM downscale rod block is <5 cps.

B. Correct answer.

C. Incorrect since the SRM Hi rod block is 6.8 X 104 D. Incorrect since SRM's are not INOP.

RO Tier: T2Gl SRO Tier: T2G1 Keyword: SRM Cog Level: C/A 3.7/3.7 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 14

QUESTIONS REPORT for SRO 2002-301 LXRBank

15. 215005K3.01 001 Which ONE of the following Mode Switch position and Nuclear Instrumentation signal combinations will cause ONLY a REACTOR CHANNEL "A"AUTO SCRAM?

A.' RUN; 2/4 Voter Al in TEST.

B. STARTUP; 2/4 Voter B2 in TEST.

C. RUN; IRM"G" Upscale.

D. STARTUP; Channel 2 OPRM PBA Trip and Channel 4 OPRM PBA Trip.

References:

Tech Specs 3.3.1.1-1 pg 3.3-7 and 3.3-8 OPL171.148 Rev.7 pg 24-56 A. Correct answer.

B. Only required in Mode 1.

C. Incorrect since IRM Hi does not generate trip with Mode Switch in Run.

D. Only required in Mode 1.

RO Tier: T2G1 SRO Tier: T20I Keyword: APRM Cog Level: MEM 4.0/4.0 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 15

QUESTIONS REPORT for SRO 2002-301 LXRBank

16. 216000A2.14 001 During a startup the operators begin to raise recirculation pump flow.

Which ONE of the following describes the effect on Panel 9-5 RPV level indicators when raising recirc flow from 50% to 65%?

A/ Emergency range indicated level will trend downward.

B. Narrow range indicated level will trend upward.

C. Emergency range indicated level will trend upward.

D. Narrow range indicated level will trend downward.

References:

RO Tier: T2G1 SRO Tier: T2GI Keyword: LEVEL INSTRUMENTS Cog Level: C/A 2.9/2.9 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:47 AM 16

QUESTIONS REPORT for SRO 2002-301 LXRBank

17. 217000K2.02 001 Unit 2 is operating at 100% RTP when the 250VDC Reactor MOV Board B Logic Bus A de-energizes. An operator has been sent to investigate and reports that the feed breaker has failed.

Which ONE of the following describes the operation of HPCI and RCIC if reactor water level decreases to -45"under these conditions?

A. HPCI and RCIC will both automatically initiate but will not auto isolate if needed.

Bf HPCI will automatically initiate but will not auto isolate if needed and RCIC will not automatically initiate.

C. Both HPCI and RCIC will not initiate automatically but may be operated manually.

D. HPCI will not automatically initiate and RCIC will automatically initiate but will not auto isolate if needed.

References:

2-ARP-9-3F pg 4 2-ARP-9-3C pg 2 A. Incorrect since RCIC initiation logic will not work.

B. Correct answer.

C. Incorrect since HPCI will still initiate automaticall*y.

D. Incorrect since HPCI will automatically initiate an d RCIC will NOT automatically initiate.

NOTE: RCIC and HPCI recieve an initiation signal vvhen RWL reaches -45".

RO Tier: T2G1 SRO Tier : T2GI Keyword: RCIC SYSTEM Cog Leve 1: C/A 2.8/2.9 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 17

QUESTIONS REPORT for SRO 2002-301LXRBank

18. 218000K6.06 001 Various electrical malfunctions have occurred on Unit 2. Existing conditions are as noted:

- 480V S/D Bd 2A deenergized

- 480V RMOV Bd 2C deenergized

- 250V-RMOVBd 2B deenergized

- No boards have been transferred Which ONE of the following identifies the systems that are still available?

A. ADS, HPCI, RCIC B. CS Loop I, RHR I, RCIC C. RHR Loop I, ADS, HPCI W. CS Loop II, RHR Loop I, HPCI RO Tier: T2G1 SRO Tier: T2G1 Keyword: 480V DISTRIBUTION Cog Level: C/A 3.4/3.6 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:47 AM 18

QUESTIONS REPORT for SRO 2002-301 LXRBank

19. 223001A2.11 001 Units 2 and 3 are operating at 100% RTP when a loss of Offsite power occurs. This condition is expected to exist for the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Which ONE of the following predicts the response of suppression pool level over the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ?

A/ Suppression pool water level will rise due to operation of SRV's, HPCI and/or RCIC.

B. Suppression pool water level will be controlled within the normal operating band due to the RHR drain pumps being used to control level.

C. Suppression pool water level will lower, makeup can be provided by opening the minimum flow valves on RCIC and HPCI from the CST's.

D. Suppression pool water level will remain stable except for heating by the use of SRV's, HPCI and/or RCIC causing indicated level to read high.

References:

A. Correct answer since MSIV's close on loss of off site power and all 3 systems are needed to control reactor pressure and reactor water level.

B. Incorrect since the drain pumps do not have a power supply.

C. Incorrect since suppression pool level will increase due to operation of SRV's and HPCI and/or RCIC.

D. Incorrect since suppression pool level will increase due to operation of SRV's and HPCI and/or RCIC.

RO Tier: T2GI SRO Tier: T2G1 Keyword: SUPPRESSION CHAMBER Cog Level: C/A 3.6/3.8 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 19

QUESTIONS REPORT for SRO 2002-301LXRBank

20. 223002K3.16 001 Unit 3 is in a refueling outage with Shutdown Cooling in operation on RHR Sys If. A spurious Group IIisolation is initiated by the Instrument Techs while performing a surveillance. All isolations occurred as designed.

Which ONE of the following describes the actions to take to allow re-opening 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VLV?

AY Isolation signal has been reset AND either Shutdown Cooling Suction Valve is fully closed.

B. RHR SYS I SD CLG INBD INJECT ISOL RESET pushbutton is depressed followed by the group II isolation signal being reset.

C. Either Shutdown Cooling Suction Valve fully closed followed by the RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton being depressed.

D. RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton is depressed followed by either Shutdown Cooling Suction Valve being fully closed.

References:

3-01-74 Rev.52 pf 12 Enabling Objective OPLI171.044 Rev.10 #B10 A. Incorrect since RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton must be depressed after either of the listed conditions clears.

B. Incorrect since RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton must be depressed AFTER the condition clears.

C. Correct answer.

D. Incorrect since RHR SYS II SD CLG INBD INJECT ISOL RESET pushbutton must be depressed AFTER the condition clears.

RO Tier: T2GI SRO Tier: T2G1 Keyword: RHR SYSTEM Cog Level: C/A 3.2/3.3 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 20

QUESTIONS REPORT for SRO 2002-301 LXRBank

21. 226001K1.09 001 Unit 3 is at 90% RTP when a LOCA occurs. The following conditions are present in the Containment:

Drywell Pressure 12.5 psig Drywell Temperature 260°F Suppression Pool Level 16 ft Suppression Pool Temperature 150 0 F The Unit Supervisor has ordered Drywell Sprays to be initiated per EOI-2, Primary ContainmentControl.

Which ONE of the following describes the affect on Containment when Drywell Sprays are initiated? (Assume Suppression Chamber sprays have been initiated)

A. A large rapid reduction in Drywell pressure followed by the opening of the Reactor Building to Suppression Chamber vacuum breakers followed by the opening of the Suppression Chamber to Drywell vacuum breakers.

B. A slow reduction in Drywell pressure followed by the opening of the Reactor Building to Suppression Chamber vacuum breakers followed by the opening of the Suppression Chamber to Drywell vacuum breakers.

C. A slow reduction in Drywell pressure followed by the opening of the Suppression Chamber to Drywell vacuum breakers.

D' A large rapid reduction in Drywell pressure followed by the opening of the Suppression Chamber to Drywell vacuum breakers.

References:

OPL171.044 Rev.10 pg 59 A. Incorrect since the Suppression Chamber to Drywell vacuum breakers open first.

B. Incorrect since the pressure reduction is rapid due to mainly steam in the Drywell.

C. Incorrect since the pressure reduction is rapid due to mainly steam in the Drywell.

D. Correct answer.

RO Tier: T2G2 SRO Tier: T2GI Keyword: SUPPRESSION CHAMBER Cog Level: C/A 3.0/3.1 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:47 AM 21

QUESTIONS REPORT for SRO 2002-301 LXRBank

22. 233000K1.02 001 Which ONE of the following is CORRECT regarding RHR Supplemental Fuel Cooling?

A. The RHR pumps are preferred for use in this mode over the RHR drain pu B. -RHR Drain Pump B cannot be used to provide flow.

C' Should only be used when required to maintain Fuel Pool temperature bel 125 0 F.

D. RHR pump suction is taken from the fuel pool cooling pump discharge line

References:

OPI171.052 page 25 2-01-74 Rev. 107 pg 94 Enabling Objective #6 A. Incorrect since the drain pumps are preferred for use over the RHR pumps.

B. Incorrect since RHR Drain Pump B can be used for this function.

C. Correct answer.

D. Incorrect since the suction is taken from the Skimmer Surge Tank outlet.

Changed the correct answer to a totally different answer.

RO Tier: T2G3 SRO Tier: T2G3 Keyword: FUEL POOL COOLING Cog Level: MEM 2.9/3.0 Source: M Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 22

QUESTIONS REPORT for SRO 2002-301 LXRBank

23. 239001K5.08 001 DC power has been lost to a MSIV solenoid valve.

Which ONE of the following describes the effect on the MSIV?

A. The valve will close if open.

B! The valve-will remain open if open.

C. The valve cannot be opened if closed.

D. The slow closure capability of the valve is lost.

References:

OPL171.009 Rev.8 pg 26 NOTE: Modified the stem slightly and reordered answers.

A. Incorrect since the AC and DC solenoids must de-energize to close the valve.

B. Correct answer.

C. Incorrect since only ONE of the solenoid valves must be energized to operate the valve.

D. Incorrect since the solenoid valves do not affect the testing circuit.

RO Tier: T2G2 SRO Tier: T2G3 Keyword: MAIN STEAM Cog Level: MEM 2.6/2.7 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 23

QUESTIONS REPORT for SRO 2002-301 LXRBank

24. 239002K5.04 001 The following plant conditions exist:

Reactor Power 100% RTP Reactor Pressure 1000 psig Safety Relief Valve (SRV) 1-4 has lifted and failed to reseat.

Which ONE of the following SRV tailpipe temperatures would you expect to see on the SRV that failed to close? (Steam Tables attached)

A. 212 0 F Bf 290°F C. 345°F D. 545°F JUSTIFICATION A. Incorrect since this is saturation temperature for steam at tailpipe pressure (atmospheric).

B. Correct answer. This is a throttling process and is therefore isenthalpic.

C. 340'F would be incorrectly determined if the candidate considered the process to be isenthalpic to the saturation line, then followed the constant superheat line to atmospheric pressure.

D. Incorrect since this is saturation temperature for reactor pressure.

RO Tier: T2GI SRO Tier: T2GI Keyword: RELIEF VALVE Cog Level: C/A 3.3/3.5 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 24

QUESTIONS REPORT for SRO 2002-301 LXRBank

25. 241000A4.11 001 Which ONE of the following is the controlling parameter that is illuminated on the Turbine Control Panel during a turbine roll to 1800 rpm?

A. Valve position.

B. Pressure.

Cf Speed.

D. Load.

References:

OPLI71.228 Rev. 0 pg Enabling Objective OPLI171.228 #9 A, B and D are incorrect since SPEED is the controlling parameter until the turbine reaches "AT SET SPEED".

C. Correct answer.

RO Tier: T2G1 SRO Tier: T2G1 Keyword: EHC SYSTEM Cog Level: C/A 3.1/3.1 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 25

QUESTIONS REPORT for SRO 2002-301 LXRBank

26. 245000K5.03 001 The main turbine shell is being warmed in accordance with GOI-1 00-1 A, Unit Startup and Power Operation and 01-47, Turbine Generator System.

Which one of the following is the correct turbine valve configuration?

CONTROL VALVES STOP VALVES INTRCPT STOPS INTRCPT CONTROL A. Full open 1,3 & 4 closed Full closed Full open

(# 2 BP open)

Bf Full open 1,3 &4 closed Full closed Full closed

(#2 BP open)

C. Full closed Full closed Full open Full open D. Full open Full open Full open Full closed Taskno: U-047-NO-02 RO Tier: T2G2 SRO Tier: T2G2 Keyword: TURBINE CONTROLS Cog Level: C/A 2.6/2.6 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:48 AM 26

QUESTIONS REPORT for SRO 2002-301LXRBank

27. 259001A1.01 001 Unit 2 is at 100% RTP. A heater tube leak activates alarm 2-LA-6-4, HEATERA2 LEVEL HIGH. The Operator checks the ICS screen and verifies a valid HIGH HIGH (Red) level. Heater level continues to rise.

Which ONE of the following describes the required Operator action and the response of the plant?

A.' The Operator should be directed to reduce Core Thermal Power and verify 2A2 heater high level dump valve to the main condenser OPENS.

B. The Operator should be directed to hold power constant and verify the 2A2 high level dump valve to the heater drain cooler OPENS.

C. The Operator should be directed to reduce Core Thermal Power and verify HP Heater 2A1 extraction isolation valve is OPEN.

D. The Operator should be directed to hold power constant and verify the drain inlet flow from the 2A2 heater to the 2A1 heater is isolated.

References:

2-ARP-9-6A Rev. 16 pg 10 Note: Modified from a question on the last exam.

A. Correct answer.

B. Incorrect since core thermal power should be lowered.

C. Incorrect since the drain for the 2A2 heater to the 2A1 heater should be open.

D. Incorrect since core thermal power should be lowered.

RO Tier: T2G1 SRO Tier: T2G2 Keyword: FEEDWATER HEATERS Cog Level: C/A 3.3/3.3 Source: M Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 27

QUESTIONS REPORT for SRO 2002-301 LXRBank

28. 259002G2.2.22 001 Unit 2 experienced a Group I isolation with a failure to scram. Rx steam dome pressure increased to 1320 psig which caused a break in the Recirc Suction piping.

The following conditions exist at this time:

Rx Power 8% RTP Reactor Water Level -165 inches .

Reactor Pressure 185 psig Core Flow 5% of rated Drywell Pressure 13 psig Which ONE of the following describes the Safety Limit violated and the corrective action?

A. Reactor steam dome pressure; insert all insertable control rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Core Flow vs. Thermal Power; insert all insertable control rods and restore Thermal Power to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

C.f Reactor Vessel Water Level; restore level above -162" and insert all insertable control rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

D. Drywell Pressure; restore to within limits and be in Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

References:

Tech Spec Section 2.0, Safety Limits Tech Spec Bases pg B 2.0-6 A. Incorrect since Reactor steam dome safety limit was not exceeded (limit is 1325 psig)

B. Incorrect since the safety limit is > 25% RTP and core flow < 10% rated.

C. Correct answer.

D. Incorrect since Drywell Pressure is not a safety limit.

RO Tier: SRO Tier: T2G I Keyword: SAFETY LIMIT Cog Level: C/A 3.4/4.1 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:48 AM 28

QUESTIONS REPORT for SRO 2002-301 LXRBank

29. 261000K2.03 001 The SGT "A" master control switch (HS-65-18A on Unit 1) is in the pull-to-lock position.

Which ONE of the following conditions would still cause the SGT "A" to start even though the control switch is in the pull-to-lock position?

A. Unit 3 SGT "A" start pushbutton is depressed.

B. Unit 2 drywell pressure rises to 2.5 psig and continues to rise.

Cf The local (SGT Building) SGT "A" start pushbutton is depressed.

D. SGT TRAIN "A" INBD ISOL TEST SIG keylock switch (HS-65-48A) is placed in the TEST position.

RO Tier: SRO Tier: T2G1 Keyword: SBGT Cog Level: C/A 2.3/2.5 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:48 AM 29

QUESTIONS REPORT for SRO 2002-301 LXRBank

30. 262001K4.06 001 Unit 2 is operating at 100% RTP.

-A combination of errors cause an inadvertent Group 1 and Group 4 isolation.

-A loss of I&C 2A also occurs. Panel 9-9 cabinet 2 does not transfer.

-Reactor Water Level is currently at 22".

-The causing event for the group isolations is quickly corrected, however I&C

-cannot be restored.

Which ONE of the following lists the systems that can be utilized immediately to restore reactor water level?

A. RCIC and CRD only.

B' HPCI, RCIC and CRD.

C. Core Spray, HPCI, and RCIC.

D. Reactor Feed Pumps, RCIC and CRD.

References:

2-AOI-57-5A, Rev. 37 pg 2 and 3 Bank question -,Revised answers slightly and reworded last portion of stem.

A. Incorrect since HPCI is also available for injection since the Group 4 isolation is able to be reset even with a loss of I&C A.

B. Correct answer.

C. Incorrect since Core Spray cannot be used with reactor at normal operating pressure.

D. Incorrect since Reactor Feedwater Pumps are not available due to Group 1 isolation not being able to be reset until I&C A is restored.

RO Tier: T2G2 SRO Tier: T2GI Keyword: AC DISTRIBUTION Cog Level: C/A 3.6/3.9 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 30

QUESTIONS REPORT for SRO 2002-301LXRBank

31. 262002K6.02 001 Unit 2 UPS Distribution Bus Battery Board 2 Panel 11 has just de-energized.

Which ONE of the following describes the effect this has on the equipment that is supplied by Panel 9-9 Cabinet 6?

A. The equipment is de-energized until power to Panel 9-9 Cabinet 6 is manually transferred to Batt Bd 3 Panel 11.

B. The equipment is de-energized until power is restored to Batt Bd 2 Panel 11.

Cf The equipment remains energized due to power supply to Panel 9-9 Cabinet 6 auto transfers to Batt Bd 3 Panel 11.

D. The equipment remains energized due to MMG power supply automatically transferring to its 250VDC supply.

References:

OPL171.102 Rev.4 pg 14 and 15 Enabling Objective OPI171.102 #2a and 2b A. Incorrect since the equipment remains energized due to auto transfer.

B. Incorrect since the equipment remains energized due to auto transfer.

C. Correct answer.

d. Incorrect since the MMG set power supply does not affect the power to Panel 9-9.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: 480V DISTRIBUTION Cog Level: MEM 2.8/3.1 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 31

QUESTIONS REPORT for SRO 2002-301 LXRBank

32. 263000K1.04 001 The Unit 2 Unit Operator receives alarm BATBD 2 BKR TRIPOUT/FUSE BLOWN OR GROUND.

Which ONE of the following describes where the Field Operator would be sent to check for a ground?

A. Battery Board Room No. 2, 250V Charger 2A panel.

B. 250V DC Distribution Panel SBA.

C. Battery Board Room No. 2, Panel 1.

D. 4KV Shutdown Bd 250V DC Distribution Panel SD-3EB.

References:

2-ARP-9-8C Page 8 Tile #7 0-01-57D Rev.62 Pg 42, 46 and 47.

C. Correct answer.

A, B and D. Plausible distractors.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: GROUND DETECTION Cog Level: MEM 2.6/2.9 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 32

QUESTIONS REPORT for SRO 2002-301 LXRBank

33. 264000A1.03 001 Diesel Generator 3A is synchronized to 4KV Shut Down Board 3A. The instrumentation readings for the diesel generator are as follows:

voltage: 4160 VAC frequency = 59.8

.current = 340 amps vars = 1600 K-vars watts = 2585 KW oil temp = 145 0 F Which ONE of the following actions are required ifthe diesel is expected to be operated for an extended period? (Supply 01-82 illustration #11)

A. The operator must take the voltage regulator control switch to raise to reduce field current.

B. The operator must take the voltage regulator control switch to lower to reduce field current.

C' The operator must take the governor control switch to lower to reduce stator amps.

D. The operator must take the governor control switch to raise to reduce stator amps.

References:

01-82 OPL171 .038 Rev. 9, page 31 RO Tier: T2G1 SRO Tier: T2G1 Keyword: DIESEL GENERATOR Cog Level: C/A 2.8/2.9 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:49 AM 33

TITLE: STANDBY DIESEL GENERATOR SYSTEM UNIT 0 0-01-82 ILLUSTRATION 1 REV 0081 (Page 1 of 1)

DG KW vs KVAR LOADING 3201 0 11 1 1 11 1 1 1 1 -

300(

0 280,

P----------------------

2600

---P----- -- - - - - - - - - - - -

2400 2200 2800 KW -75O PERCENT OF CONTINUOUS RATING 2000 1800 KW 1600 1400 REPRE - -NT LIN E------------------------------- -------------------

1200 ------------ ------------------------------------

1000 2000 PECN -7OF CONTINUOUSRATIN -W

--- +/-:-- -- -

DIAGGONAL

- - - - - - ---- -T--

8oo OR OUTDOINGIIII 600 400 200 0

0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 KVAR (LAGGING OR OUTGOING)

Page 97 of 99

QUESTIONS REPORT for SRO 2002-301 LXRBank

34. 268000A4.01 001 Given the following information:

Unit 2 has been at 100% RTP for 3 weeks.

Current 2-FQ-77-6 Reading at 0800 63624.3 Previous Days 2-FQ-77-6 Reading at 0800 63125.4 Previous Days Leakrate .34 gpm Current calculated Leakrate - 3.12 gpm Which ONE of the following describes the status of the LEAKAGE limits?

A. No limits are being exceeded.

B/ Increase in unidentified LEAKAGE limit is being exceeded.

C. Unidentified LEAKAGE limit is being exceeded.

D. Increase in unidentifed LEAKAGE and unidentified LEAKAGE limit are both being exceeded.

References:

2-SR-2 Rev.29 pg 20 A. Incorrect since increase in LEAKAGE limit is being exceeded at 3.12 gpm. Limit is

< 2 gpm. If use decimal point when subtracting readings then this would be the answer that the student would get. Procedure says to ignore decimal point.

B. Correct answer. Increase in LEAKAGE is at 3.12 gpm and limit is < 2 gpm.

C. Incorrect since the increase in LEAKAGE is the only limit not met.

D. Incorrect since the increase in LEAKAGE is the only limit not met.

RO Tier: T2G3 SRO Tier: T2G3 Keyword: LEAKAGE LIMITS Cog Level: C/A 3.4/3.6 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 34

QUESTIONS REPORT for SRO 2002-301LXRBank

35. 271000K3.02 001 Which ONE of the following describes the effect on offsite release rates if the Off-Gas System Glycol pumps fail?

Af Offsite Release Rates will INCREASE due to the Charcoal Adsorbers becoming less efficient.

B. Offsite Release Rates will DECREASE due to better H2 0 2 Recombination.

C. Offsite Release Rates will INCREASE due to the Off-Gas Condenser becoming le' ss efficient.

D. Offsite Release Rates will DECREASE due to the Charcoal Adsorbers becoming more efficient.

References:

OPL171.030 Rev. 13 Pg 29 and 31 A. Correct answer. The glycol cools the Cooler Condenser which is used to remove moisture from the gases entering the Charcoal Adsorbers. Water is a poison to the adsorbers so if the gases contain more moisture then the adsorbers are less efficient.

B. Incorrect since the glycol system has no affect on the Recombiners.

C. Incorrect since the Condensate System supplies cooling to the Off-Gas condenser.

D. Incorrect since the Charcoal Adsorbers become less efficient.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: OFF-GAS SYSTEM Cog Level: C/A 3.3/3.9 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 35

QUESTIONS REPORT for SRO 2002-301 LXRBank

36. 286000A3.01 001 The following conditions currently exist on Unit 2:

- A fire at one station service transformer has actuated the water spray system.

- Fire header pressure has been 115 psig for 35 seconds after the spray system actuated.

- All system controls are in a normal lineup.

Based on these conditions, the diesel fire pump....

A. and all three electric fire pumps are operating.

B. and two of the three electric fire pumps are operating.

C. is in standby and all three electric fire pumps are operating.

D. and two electric fire pumps are in standby; the selected electric fire pump is operating.

References:

OPL171.049 Rev. 12 pg 43 Enabling Objective (HLT) 5 0-01-26 Rev. 55 pg 10 A. Incorrect since the diesel fire pump doesn't start until 45 seconds after pressure is below 120#.

B. Incorrect since the diesel fire pump doesn't start until 45 seconds after pressure is below 120#.

C. Correct answer.

D. Incorrect since all of the electric fire pumps should be running.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: FIRE PROTECTION Cog Level: C/A 3.4/3.4 Source: M Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 36

QUESTIONS REPORT for SRO 2002-301 LXRBank

37. 290001A3.01 001 Which ONE of the following conditions will cause the Reactor Bldg ventilation trip and isolate?

A. A scram which results in reactor water level reaching +14 inches.

B. Drywell pressure reaches 2.3 psig before the Drywell can be vented.

C' Reactor Bldg static pressure reaches +.6 inches of water due to high windE D. Reactor Zone exhaust duct radiation level reaches 62 mR/hr due to a stear

References:

OPI171.016 Rev.12 pg 62 and 63 A. Incorrect since the isolation setpoint for RWL is +11.2".

B. Incorrect since the isolation setpoint for Drywell High pressure is +2.45 psig.

C. Correct answer.

D. Incorrect since the isolation setpoint for exhaust duct hi rads is 72 mR/hr.

RO Tier: T2G2 SRO Tier: T2GI Keyword: SECONDARY CONTAINMEN Cog Level: MEM 3.9/4.0 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 37

QUESTIONS REPORT for SRO 2002-301LXRBank

38. 290002K4.03 001 Which ONE of the following describes the design and purpose of the orificing in the lower section of the reactor core?

A. All orifices are the same size to ensure all bundles have the same flow.

B. The interior bundles have more orifices to ensure equalized core flow at high power levels.

C. Center portions of the core have smaller orifices to ensure the neutron thermalization is equalized across the core.

Df The outer portions of the core have smaller orifices to ensure adequate cooling in the interior fuel bundles at high power levels.

References:

OPL1 71.002 Rev.5 pg 24-26 Enabling Objective OPL-1711.002 #2 A. Incorrect since all orifices are not the same size.

B. Incorrect since more orifices are not provided but larger orifices are provided.

C. Incorrect since center orifices are larger than outer orifices.

D. Correct answer.

RO Tier: T2G3 SRO Tier: T2G3 Keyword: VESSEL INTERNALS Cog Level: MEM 3.2/3.3 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:49 AM 38

QUESTIONS REPORT for SRO 2002-301 LXRBank

39. 290003G2.1.12 001 Unit 2 is at 100% power. Two Unit 1 &2 Control Room Air Conditioning Subsystems are inoperable. Actions to restore one subsystem to OPERABLE status are in progress.

Which ONE of the following states the actions/limitations imposed by Tech Specs?

A. Unit 2 must be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and MODE 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

B." Place an alternate method of cooling in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and restore one subsystem to OPERABLE status within 7 days.

C. Restore one subsystem to OPERABLE status within 30 days.

D. Place an alternate method of cooling in service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and restore one subsystem to OPERABLE status within 30 days.

References:

Tech Spec 3.7.4 Condition B Revised answers from Bank question to prevent having more than one correct answer since you are in Condition A until you return BOTH systems to OPERABLE status.

A. Incorrect since this is the action to take if Condition B completion time cannot be met.

B. Correct answer.

C. Incorrect since at least ONE subsystem must be restored to OPERABLE status within 7 days along with other actions.

D. Incorrect since one subsystem must be returned to operable status within 7 days.

RO Tier: SRO Tier: T2G2 Keyword: TECH SPECS Cog Level: C/A 2.9/4.0 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:50 AM 39

QUESTIONS REPORT for SRO 2002-301LXRBank

40. 295001AA2.01 001 Unit 3 is in the process of starting up A 100% rod line has been established Currently raising recirc flow A fault in the 3B Recirc MG set occurs causing a trip of the MG set.

The operator notes the following conditions after the pump trip:

- MWE = 560

- MWT = 1745

- Core Flow = 32%

- OPRM's are INOPERABLE Using the attached Illustration 1 from 3-01-68 determine which ONE of the following describes the appropriate action to take?

A. Region 2 has been entered, scram the reactor immediately.

B.! Region 1 has been entered, scram the reactor immediately.

C. Region 2 has been entered, and must be exited within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

D. Region 1 has been entered, insert control rods to less than a 95.2% rod line.

Task no: U-068-AB-01 RO Tier: SRO Tier: T1G2 Keyword: PWR/FLOW MAP Cog Level: C/A 3.5/3.8 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:50 AM 40

QUESTIONS REPORT for SRO 2002-301 LXRBank

41. 295002AA1.05 001 The following plant conditions exist on Unit 2:

- Reactor mode switch: STARTUP/HOT STANDBY

- Main turbine: Shell warming

- Feedwater lineup: RFP A maintaining level in single element Which ONE of the following statements describes-the expected sequence of actions as a condensate system leak causes condenser vacuum to decrease from 27 inches Hg Vacuum to atmospheric pressure?

A. The RFP turbine trips, then later, the turbine bypass valves close, followed by a reactor scram on low condenser vacuum.

B. The RFP turbine trips and the main turbine bypass valves close at the same time, then later, the Main Turbine trips.

Co The Main Turbine trips, then later, the RFP turbine trips and the main turbine bypass valves close at the same time.

D. The Main Turbine trips and the reactor scrams in response to the turbine trip, then later, the RFP turbine trips and Main Turbine bypass valves close at the same time.

JUSTIFICATION

a. There is no reactor scram on low main condenser vacuum.

b. A true statement at 7" Hg Vac; however, this is preceded by a main turbine trip at 21" Hg Vac.

c. Correct answer.

d. The reactor won't trip on a turbine trip below 30% RTP.

RO Tier: TIG2 SROTier: T1G2 Keyword: MAIN TURBINE Cog Level: C/A 3.2/3.2 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:50 AM 41

QUESTIONS REPORT for SRO 2002-301 LXRBank

42. 295003AK1.02 001 Unit 2 is at 100% power and has a special test in progress with the C DIG tied to 4KV SD Bd C as the sole source. The following occurs:

MSIVs go closed due to high steam tunnel temperature.

All rods do not insert.

Reactor pressure is 800 psig.

Reactor power is 2.5%.

Reactor level is -45".

If reactor water level decreased to -122 inches, which ONE of the following describes the effect this would have on the RBCCW system?

(Assume no operator actions)

A! Both pumps trip, 2A will auto restart in 40 seconds.

B. RBCCW pump 2B will trip, pump 2A not effected.

C. Both pumps trip and auto restart in 40 seconds.

D. No effect on the system.

References:

OPL171.072 Rev. 8 pg 7 & 15 Enabling Objective OPLI171.072 #4 Bank Question - Comment: 480V load shed will occur due to C diesel, only 2A pump auto restarts.

A. Correct answer since the D/G is the sole power supply to the Shutdown Board and water level reaches 122".

B. Incorrect since both pumps trip and the 2A restarts after 40 seconds.

C. Incorrect since the 2B pump restarts automatically only if the 2A pump fails to start.

D. Incorrect since the RBCCW pumps trip due to D/G tied to the Shutdown Board and water level reaches -122".

RO Tier: TIG2 SROTier: TIGI Keyword: LOAD SHED Cog Level: C/A 3.1/3.2 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:51 AM 42

QUESTIONS REPORT for SRO 2002-301 LXRBank

43. 295003G2.1.12 001 Unit 2 is at 100% RTP. Core Spray Pump 2B has just been tagged for maintenance when 250V DC Reactor MOV Board 2B trips.

Which ONE of the following describes the Tech Spec required actions for Unit 2?

A! Enter LCO 3.0.3 immediately.

B. Restore the 250V DC Reactor MOV Board 2B OR Core Spray Pump 2B to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

C. Restore Core Spray Pump 2B AND 250V DC Reactor MOV Board 2B to OPERABLE status within 7 days.

D. Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND be in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

References:

Tech Spec 3.5.1 Refer to OPDP-8. The 250V Board is a support system for Core Spray, and there is a LOSF since the 250V Board is Division I logic. Therefore both Core Spray systems are inoperable, requiring entering LCO 3.0.3 per 3.5.1.H. If 250V Rx MOV Board 2B is placed on its alternate supply the board is still considered inoperable per BASES B 3.8.7.

A. Correct answer since two low pressure ECCS spray subsystems are INOPERABLE.

LCO 3.0.3 is entered per 3.5.1.G.

B. Incorrect since this assumes that HPCI is also INOPERABLE.

C. Incorrect since the 7 day clock is for 1 low pressure spray subsystem being INOPERABLE.

D. Incorrect since this action is required only if the required actions or completion time for Condition A is not met.

RO Tier: SRO Tier: TIGI Keyword: CORE SPRAY Cog Level: C/A 2.9/4.0 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:51 AM 43

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

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

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

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

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 in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued)

BFN-UNIT 2 3.5-1 Amendment No. 253- 269 March 12, 2001

ECCS - Operating 3.5.1 ACTIONS (continued)

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

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

D. HPCI System inoperable. D.1 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 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.

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 OR 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 injection/spray subsystem to OPERABLE status.

(continued)

BFN-UNIT 2 3.5-2 Amendment No. 243-, 269 March 12, 2001

ECCS - Operating 3.5.1 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 G.2 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 H.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

BFN-UNIT 2 3.5-3 Amendment No. 23, 269 March 12, 2001

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the injection valve.

SR 3.5.1.2 --------------------- NOTE ------------

Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) low pressure permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each ECCS injection/spray subsystem 31 days manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.1.3 Verify ADS air supply header pressure is _Ž81 31 days psig.

SR 3.5.1.4 Verify the LPCI cross tie valve is closed and 31 days power is removed from the valve operator.

(continued)

BFN-UNIT 2 3.5-4 Amendment No. 253

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.5 ------------

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

1. Only required to be performed when in MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

2. Not required to be performed if performed within the previous 31 days.

Verify each recirculation pump discharge Once prior to valve cycles through one complete cycle of entering MODE 2 full travel. from MODE 3 or 4

SR 3.5.1.6 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the Inservice corresponding to the specified pressure. Testing Program SYSTEM HEAD CORRESPONDING TO A VESSEL TO TORUS NO. OF DIFFERENTIAL SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray Ž6250 gpm 2 105 psid INDICATED NO. OF SYSTEM SYSTEM FLOW RATE PUMPS PRESSURE LPCI > 12.000 gpm 2 > 250 psig LPCI > 9,000 gpm 1 >_125 psig (continued)

BFN-UNIT 2 3.5-5 Amendment No. 253

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.7 ------------------ NOTE ------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 1040 and 92 days

_>950 psig, the HPCI pump can develop a flow rate > 5000 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.8 ------------------ NOTE ------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure _<165 psig, the 24 months HPCI pump can develop a flow rate > 5000 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.9 --------------------- NOTE--- ------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

(continued)

BFN-UNIT 2 3.5-6 Amendment No. 255 November 30, 1998

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.1.10 -------------------------- NOTE ------------

Valve actuation may be excluded.

Verify the ADS actuates on an actual or 24 months simulated automatic initiation signal.

SR 3.5.1.11 -------------------------- NOTE ------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify each ADS valve opens when manually 24 months actuated, SR 3.5.1.12 Verify automatic transfer of the power supply 24 months from the normal source to the alternate source for each LPCI subsystem inboard injection valve and each recirculation pump discharge valve.

BFN-UNIT 2 3.5-7 Amendment No. 255 November 30, 1998

QUESTIONS REPORT for SRO 2002-301LXRBank

44. 295004AA1.03 001 Unit 2 was operating at 100% power when a reactor scram occurs. The following plant conditions exist:

Main turbine is tripped.

Position indication for DC powered RCIC valves is out.

CORE SPRAY SYS I LOGIC POWER FAILURE annunciator is lit.

Which ONE of the following is the likely cause of this event?

A. Loss of 250 VDC RMOV Bd "A".

BY Loss of 250 VDC RMOV Bd "B".

C. Loss of 250 VDC RMOV Bd "C".

D. Loss of 250 VDC Turb Bldg Dist. Bd 2.

References:

2-ARP-9-8C #11 0-01-57D Rev.

This is a bank question. B is the correct answer. Not verified yet.

RO Tier: T1G2 SRO Tier: TIG2 Keyword: 250 VDC Cog Level: MEM 3.21/3.5 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:52 AM 44

QUESTIONS REPORT for SRO 2002-301 LXRBank

45. 295005AA2.02 001 Unit 3 is at 24% RTP during a startup from a refueling outage. The Control Room Operator reports that the vibrations for the turbine are at 12 mils and continuing to increase slowly.

Which ONE of the following describes the actions you should take as Control Room Supervisor?

A. Order the Control Room Operator to Trip the Turbine immediately and verify a reactor scram.

B. Order the Control Room Operator to monitor the Turbine vibrations and to Trip the Turbine ifthe vibrations continue for greater than 15 minutes.

C. Enter EOI-1 since a scram should have occured due to Turbine Trip from High vibrations.

Do Order the Control Room Operator to Trip the Turbine immediately. Verify Generator output breaker trips.

References:

OPI171.147 Rev. 3 pg 10, 19, 22 A. Incorrect since a reactor scram should not occur from the turbine trip due to Rx power <30%.

B. Incorrect since the turbine must be tripped immediately if vibrations reach 12 mils.

C. Incorrect since an entry condition does not exist at this time.

D. Correct answer.

RO Tier: SRO Tier: TIG2 Keyword: TURBINE CONTROLS Cog Level: C/A 2.4/2.7 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 45

QUESTIONS REPORT for SRO 2002-301 LXRBank

46. 295006AA2.03 001 Unit 3 is operating at 100% RTP when the unit scrams. The following conditions exist after the scram:

All rods are in Reactor Pressure 510 psig lowering slowly Reactor Water Level -155" steady Drywell Pressure 11 psig All systems operating as designed Which ONE of the following lists ALL the systems that may be used and are capable of injecting and maintaining vessel level greater than -162"?

Af Feedwater, HPCI, RCIC, CRD and SLC.

B. Feedwater, HPCI, CRD and Core Spray.

C. Feedwater, HPCI, RCIC, and CRD.

D. Feedwater, HPCI, RCIC, CRD, SLC, Core Spray and RHR.

References:

EOI-1 Rev 5 A. Correct answer.

B. Incorrect since Core Spray is not able to inject at this pressure.

C. Incorrect since SLC is also able to be used to maintain level >-162".

D. Incorrect since Core Spray and RHR cannot inject at this time.

RO Tier: SROTier: TIG1 Keyword: LEVEL CONTROL Cog Level: C/A 4.0/4.2 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 46

QUESTIONS REPORT for SRO 2002-301 LXRBank

47. 295007AK1.01 001 Unit 2 has scrammed with the following conditions present:

All rods inserted.

Reactor pressure 475 psig Reactor water level +53" 4KV Shutdown Board "A" de-energized MSIVs open Which ONE of the following lists the systems capable of injecting at this time?

A. HPCI, RCIC, 2A CRD Pump.

B. Reactor Feedwater Pumps, 1B CRD Pump, SLC.

Cf SLC, Reactor Feedwater Pumps, 2A CRD Pump.

D. Core Spray, RHR, HPCI.

References:

OPL171.026 Rev.11 pg 25 OPLI171.040 Rev.18 pg 27 OPL1 71.042 Rev.16 pg 41 OPLI171.044 Rev.10 pg 26 OPI171.045 Rev.11 pg 15 A. Incorrect since HPCI and RCIC are isolated due to reactor high water level.

B. Incorrect since 1 B CRD Pp does not have power.

C. Correct answer.

D. Incorrect since Core Spray and RHR injection permissive is 450# and HPCI is isolated due to high reactor water level.

ROTier: TIG1 SROTier: TIGI Keyword: REACTOR LEVEL Cog Level: C/A 2.9/3.2 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 47

QUESTIONS REPORT for SRO 2002-301 LXRBank

48. 295008AK1.02 001 Unit 2 scrammed from high Drywell pressure with the following conditions present:

Reactor pressure 920 psig Reactor water level + 53 inches and increasing Drywell pressure 3.8 psig Feedwater pumps running HPCI injecting 5000 gpm RCIC in standby Which ONE of the following describes the action to be taken and the reason for the action?

A. Trip HPCI only to prevent moisture carryover into the steam lines.

B. Take manual control of HPCI and reduce flow to prevent reaching feedwater high level trip setpoint.

C' Trip HPCI and Feedwater pumps to prevent overflowing the main steam lines while pressurized.

D. Trip HPCI and Feedwater pumps to prevent violating MCPR and LHGR during a feedwater controller minimum demand failure.

References:

OPL171.003 Rev.15 pg 26 and 27 Enabling Objective OPLI171.003 B#7 A. Incorrect since feedwater pumps should also be tripped since they are approaching the high level trip setpoint. Also, RCIC should be tripped since it exceeded the trip setpoint of 51 inches.

B. Incorrect since HPCI should be tripped due to exceeding high level trip setpoint.

C. Correct answer.

D. Incorrect since the reason for tripping the equipment is to prevent exceeding MCPR and LHGR limits during a feedwater controller failure to maximum demand.

RO Tier: T1G2 SRO Tier: TIG2 Keyword: MSIV Cog Level: C/A 2.8/2.8 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 48

QUESTIONS REPORT for SRO 2002-301 LXRBank

49. 295009AK2.01 001 Unit 2 is making preparations to perform a startup after a maintenance outage.

Reactor vessel level is being maintained at +33" with Loop I RHR in Shutdown Cooling.

2A Recirc Pump is operating at minimum speed.

Which ONE of the following instruments provide the most accurate level indication under these conditions?

A/ LI 3-53 (0 to +60) on the 9-5 panel.

B. LI 3-58A (-155 to +60) on the 9-5 panel.

C. LI 3-208B (0 to +60) on the 9-3 panel.

D. LI 3-52 (-268 to +32) on the 9-3 panel.

References:

OPL171.003 Rev. 15 pg 19-21 A. Correct since instrument is pressure compensated.

B. Incorrect since instrument is calibrated at normal operating temperatures and pressures.

C. Incorrect since this instrument is calibrated under hot conditions.

D. Incorrect since level band does not reach +33 inches.

ROTier: TIGI SROTier: TIGI Keyword: LEVEL INSTRUMENTS Cog Level: C/A 3.9/4.0 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 49

QUESTIONS REPORT for SRO 2002-301 LXRBank

50. 295010AA1.01 001 During a loss of offsite power, which ONE of the following conditions would prevent the Drywell Air Cooler fans from operating?

A. Drywell pressure at 2.25 psig with reactor pressure at 435 psig.

B. Reactor water level at -110" with reactor pressure at 425 psig.

C. Reactor water level at +1" with reactor pressure at 475 psig.

Df Drywell pressure at 2.55 psig with reactor pressure at 440 psig.

References:

OPLI7I.016 Rev.12 pg 70 OP-1 71.045 Rev. 11 pg 12 A. Incorrect since Drywell pressure is below 2.45 psig.

B. Incorrect since Reactor water level is greater than -122".

C. Incorrect since Reactor water level is greater than -122" and pressure is grea iter than 450 psig.

D. Correct answer since Drywell pressure is greater than 2.45 psig and Reactor pressure is less than 450 psig.

ROTier: TIGI SROTier: TIGI Keyword: DRYWELL COOLING Cog Level: MEM 3.8/4.0 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 50

QUESTIONS REPORT for SRO 2002-301 LXRBank

51. 295012AK2.01 001 The Unit 2 Reactor Operator notices that the Drywell Temperature is increasing slowly as reactor power is increased. He verifies the normal Drywell Cooling Units are in operation on Panel 2-9-25.

Which ONE of the following indicates the normal lineup of the Drywell Cooler Fans and the Drywell temperature that should be maintained?

A. 3 of 5 cooling units in each train should be operating and attempting to maintain Drywell temperature less than or equal to 135 0 F.

B. 4 of 5 cooling units in each train should be operating and attempting to maintain Drywell temperature less than or equal to 150 0 F.

C. All the Drywell cooling units should be in operation and attempting to maintain Drywell temperature less than or equal to 135 0 F.

D. 4 of 5 cooling units in each train should be operating and attempting to maintain Drywell temperature less than or equal to 150 0 F.

References:

2-01-64 Rev.74 pg 13 A. Incorrect since 4 of 5 coolers should be operating in each train trying to maintain temp less than 135 0 F.

B. Incorrect since 4 of 5 coolers should be operating in each train trying to maintain temp less than 135 0 F.

C.Correct answer.

D. Incorrect since the coolers should be trying to maintain temp less than 135 0 F.

RO Tier: TIG2 SRO Tier: TIG2 Keyword: DRYWELL COOLING Cog Level: MEM 3.4/3.5 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:52 AM 51

QUESTIONS REPORT for SRO 2002-301LXRBank

52. 295013AK3.02 001 Which ONE of the following describes why the reactor must be shutdown immediately if the Suppression Pool temperature reaches >11 0°F?

A. To ensure that the design pressure of 56 psig is not reached during a Design Basis Accident.

BY To ensure that the pool is not heated beyond design limits by the steam generated if the reactor is not shutdown.

C. To maintain HPCI and RCIC OPERABLE since they exhaust into the suppression pool.

D. This ensures that the non-condensibles will remain in the suppression pool air space following a Design Basis Accident.

References:

Tech Spec Bases 3.6.2.1, Suppression Pool Average Temperature pg B 3.6-57.

A. Incorrect since the design pressure that is being protected is 62 psig.

B. Correct answer.

C. Incorrect since pool temperature does not affect the operation of HPCI and RCIC.

D. Incorrect since the temperature limit also ensures that complete steam condensation occurs.

RO Tier: T1G2 SROTier: TIGI Keyword: SUPPRESSION CHAMBER Cog Level: MEM 3.6/3.8 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 52

QUESTIONS REPORT for SRO 2002-301LXRBank

53. 295014AA1.03 001 A startup is in progress with the RWM bypassed. A shift turnover has just been completed when the on coming peer check notices that two rods in RWM Group 6 (16-48) are atfposition 16 and the operator is pulling rods in Group 16.

Which ONE of the following describes the proper action to take?

A. Insert a manual scram.

B. Verify no indications of fuel damage and continue withdrawal of rods.

C. Stop rod withdrawal and notify the Shift Manager, Shift Technical Advisor, Operations Superintendent, and Reactor Engineer.

D. With the concurrence of the Reactor Engineer and Shift Manager withdraw the control rods to their required position and continue the startup.

References:

2-AOI-85-7, Section 4.2.1-4.2.3, rev. 14 A. Incorrect since this is not a required action.

B. Incorrect since must recover mispositioned control rods prior to normal rod withdrawal.

C. Correct answer. Operator must determine that control rod is mispositioned otherwise he doesn't notify the Operations Superintendent.

D. Incorrect since must notify other individuals prior to withdrawing mispositioned control rods to their correct position.

Note: Reworded stem slightly and reordered answers.

ROTier: TIGI SROTier: TIGI Keyword: REACTIVITY ADDITION Cog Level: MEM 3.5/3.5 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 53

QUESTIONS REPORT for SRO 2002-301 LXRBank

54. 295014AA2.03 001 Which ONE of the following is a symptom of an inadvertent HPCI injection with the Unit at 100% RTP?

A. APRM flow comparator alarm.

B. Decrease in reactor pressure.

C. Recirc pump runback.

Df Positive reactor period.

RO Tier: SRO Tier: TIGt Keyword: HPCI Cog Level: C/A 4.0/4.3 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:53 AM 54

QUESTIONS REPORT for SRO 2002-301 LXRBank

55. 295015AK2.09 001 Unit 3 has received a Scram signal and all of the control rods fully inserted except one rod is still at position 48.

Per 3-A01-100-1, ReactorScram which ONE of the following actions would detect this condition?

A. Verifying the "REFUEL MODE ONE ROD PERMISSIVE" light is lit with the Mode Switch in Shutdown.

B! Verifying the "REFUEL MODE ONE ROD PERMISSIVE" light is out with the Mode Switch in Refuel.

C. Pausing in START/HOT STBY for 5 seconds when moving the Mode Switch to Refuel.

D. Move the Mode Switch to Shutdown and back to Refuel to look for the "REFUEL MODE ONE ROD PERMISSIVE" light to be lit.

References:

3-AOI-100-1 Rev.29 pg 2 A. Incorrect since the ONE ROD PERMISSIVE light should only light with all rods in and the Mode Switch in Refuel.

B. Correct answer.

C. Incorrect since this action is taken if the scram is due to a loss of RPS.

D. Incorrect since there is no direction to move the Mode Switch back to Refuel once it is in Shutdown.

ROTier: TIG1 SROTier: TIGI Keyword: REACTOR SCRAM Cog Level: MEM 3.5/3.6 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 55

QUESTIONS REPORT for SRO 2002-301 LXRBank

56. 295016AA1.08 002 The control room has been abandoned.

All MSRV transfer switches at panel 25-32 have been placed in EMERGENCY.

All MSRV control switches at panel 25-32 have been checked in CLOSE.

Which ONE of the following statements below describes the operation of these MSRVs?

A. The associated ADS valves will open upon receipt of an ADS initiation signal.

B. Any associated ADS valve will open only when its control switch is placed in OPEN.

C' The associated ADS valves will open if their respective pressure relief setpoints are exceeded.

D. The associated ADS valves will open if their respective control switches on panel 9-3 are placed in OPEN.

References:

OPL1 71.009 Rev.8 pg 22 Enabling Objective OPL1 71.009 #3 A. Incorrect since automatic operation of ADS is prevented with transfer switches in EMERGENCY.

B. Incorrect since valves will open when the pressure setpoint is reached.

C. Correct answer.

D. Incorrect since function from the 9-3 Panel is prevented with transfer switches in EMERGENCY.

RO Tier: TIG2 SROTier: TIGI Keyword: ADS Cog Level: MEM 4.0/4.0 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 56

QUESTIONS REPORT for SRO 2002-301 LXRBank

57. 295016G2.2.3 001 The following conditions exist at Browns Ferry Nuclear Station:

- Both Unit 2 and 3 are in Hot Shutdown

- U1/2 4 KV Shutdown Board "A" supplied by Diesel Generator as sole source

- U3 4 KV Shutdown Board "3ED" supplied by Diesel Generator as sole source

- All other electrical equipment aligned normal Which ONE of the following describes the effects on Unit 1/2 and U3 480 Volt Load Shed as a result of water level lowering to -130" on Unit 3 only (U2 water level remains normal)?

A. Unit 2 Both Divisions 480 Volt Load Shed occurs Unit 3 Only Division II480 Volt Load Shed occurs B. Unit 2 No 480 Volt Load Shed occurs Unit 3 Both Division 480 Volt Load Shed occurs C. Unit 2 No 480 Volt Load Shed occurs Unit 3 No 480 Volt Load Shed occurs D. Unit 2 Only Division 1480 Volt Load Shed occurs Unit 3 No 480 Volt Load Shed occurs

References:

OPL171-072 Rev. 8 A,B and D are incorrect because, the accident signal is generated from unit 3 which does not initiate 480 volt load shed on unit 2. However would trip the DG output breakers on unit priority retrip but no 480 load shed received. Unit 3 480 load shed is divisionalized and senses which DG is supplying the 480 volt MCCs. 3ED is the only DG on U3 that is not capable of supplying the 480 volt system, thus has no input into load shed on unit 3 even though the accident signal is from that unit.

C. Correct answer.

RO Tier: SRO Tier: TIGI Keyword: CONTROL ROOM Cog Level: MEM 3.1/3.3 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 57

QUESTIONS REPORT for SRO 2002-301 LXRBank

58. 295017AA2.04 001 A release of radioactivity is in progress. The following radiological conditions are present:

Stack Gas Rad Monitors RM-90-147 & 148 rising radiation levels Reactor Bldg/Refuel Zone Rad Monitors reading normal levels Turbine Bldg Exhaust Rad Monitor RM 90-250 rising radiation levels Drywell Rad Monitors rising radiation levels Torus Rad Monitors reading normal levels Which ONE of the following describes the probable source of the release?

A. Main Steam leakage outside the Primary Containment.

B. Reactor Water Cleanup leakage outside the Primary Containment.

C' Fuel clad failure release thru the Offgas System.

D. Major airborne activity in the Radwaste Bldg.

References:

OPI171.067 Rev.10 pg 19 - 22 A. Incorrect since the steam release is only through Primary Containment. Secondary Containment is still in tact.

B. Incorrect since the RWCU leak is through Primary Containment.

C. Correct answer. Turbine Bldg Vents take a suction from the Offgas Areas and the offgas also goes to the main stack. These are the only areas with increased rad levels other than the Drywell.

D. Incorrect since the Radwaste Bldg Ventilation has its own exhaust stack.

RO Tier: SROTier: TIGI Keyword: OFF SITE RELEASE Cog Level: C/A 3.6/4.3 Source: M Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 58

QUESTIONS REPORT for SRO 2002-301LXRBank

59. 295018AK3.03 001 Unit 2 is operating at 100% RTP. Alarm "RECIRC PUMP A COOLING WATER FLOW LOW" is received at 8:07 am on 10/22/02. It is confirmed that RBCCW Seal Cooling is lost to the 2A Recirc Pump but CRD seal purge is still in operation.

Which ONE of the following describes the actions that should be taken and the reason for that action?

A/ Monitor seal temperatures and no further action is required; the Recirc Pump can be operated indefinitely under these conditions.

B. Restore RBCCW seal cooling by 8:14 am or Trip the 2A Recirc Pump; Recirc seal temperatures will exceed 200 0 F after 7 minutes.

C. Trip the 2A Recirc Pump immediately; Recirc seal temperatures will exceed 200°F in a short period of time.

D. Reduce 2A Recirc Pump speed to minimum by 8:14 am; the Recirc Pump can be operated indefinitely at minimum speed under these conditions.

References:

2-01-68 Rev 91 pg 11 2-ARP-9-4A Rev. 18 pg 37 A. Correct answer.

B. Incorrect since the 7 minute time frame is when both CRD and RBCCW are lost to the Recirc Pump seals.

C. Incorrect since the Recirc Pump only needs to be tripped if seal cavity temperatures exceed 2000 F.

D. Incorrect since the speed of the Recirc Pump doesn't need to be reduced.

RO Tier: TIG2 SROTier: TIG2 Keyword: RECIRC SYSTEM Cog Level: C/A 3.1/3.3 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:53 AM 59

QUESTIONS REPORT for SRO 2002-301LXRBank

60. 295020AA1.03 001 Unit 2 is holding load at 24% RTP after starting up from a refueling outage. Drywell inerting is in progress per 2-01-76, Containment Inerting System. A Scram occurs from Scram Air Header Low Pressure with the following conditions present:

- Oxygen concentration at 7% by volume and decreasing.

- Leak has been isolated.

- Mode Switch is in Shutdown.

- All rods are inserted.

- No entry conditions have been met for the EOI's.

Concerning the Drywell, which ONE of the following describes the status of inerting the containment?

A. Drywell inerting has been isolated due to a containment isolation when reactor water level decreased to +0" on the scram.

B. Drywell inerting has been isolated due to the Mode Switch being taken out of Run on the scram.

C. Drywell inerting is still in progress since there has not been an isolation signal processed for this event.

D. Drywell inerting is still in progress but will isolate when the PC PURGE DIV I AND II RUN MODE BYPASS switches are taken to NORMAL.

References:

OPL171.032 Rev.10 pg 14-17 Enabling Objective OPLI171.032 #4 2-01-76 Rev.46 pg 10 A. Incorrect since reactor water level did not reach 0". Stem says no EOI's have been entered.

B. Incorrect since valves do not close when Mode Switch is taken out of Run.

C. Correct answer.

D. Incorrect since Mode Switch is no longer in Run.

RO Tier: TIG2 SRO Tier: TIG2 Keyword: CONTAINMENT Cog Level: C/A 2.9/3.1 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 60

QUESTIONS REPORT for SRO 2002-301 LXRBank

61. 295021AA2.04 001 The following conditions exist on Unit 3:

- Reactor-Water Level -- +460 --...

Reactor Coolant Temp. 185 0 F Reactor Pressure 0 psig Rx Vessel Head bolted and torqued RHR loop 1 in shutdown cooling Both Recirculation Pumps off Which ONE of the following describes the operational consequence of the Inboard Shutdown Cooling Valve (FCV 74-48) failing closed?

A. Transition boiling will occur immediately causing fuel cladding damage.

B. Damage could occur to the running RHR pump since it no longer has a suction path.

C. Coolant temperature may rise unmonitored and change the reactor condition from MODE 4 to MODE 3.

D. Since the valve is a failed PCIV then the Outboard Shutdown Cooling Valve (FCV 74-47) must be closed and deactivated.

References:

3-AOI-74-1 Rev. 25 pg 1 and 4 Tech Specs 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

A. Incorrect since transition boiling may not occur for a long period of time.

B. Incorrect since the RHR pump will trip if it loses its suction path.

C. Correct answer.

D. Incorrect since the outboard isolation valve does not need to be closed. The failed valve can be used as the isolation valve.

RO Tier: SRO Tier: T1G2 Keyword: SHUTDOWN COOLING Cog Level: C/A 3.2/3.3 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 61

QUESTIONS REPORT for SRO 2002-301 LXRBank

62. 295022G2.1.20 001 The following conditions exist on Unit 2 when the 2A CRD Pump trips on overcurrent:

Mode Switch position START/HOT STBY Reactor Pressure 800 psig Reactor Temperature 485 0 F 1B CRD Pump Out-of-Service Plant heatup in progress Which ONE of the following should the Unit Supervisor direct the Reactor Operator to perform ?

A. Scram the reactor if 2 or more scram accumulator alarms are received AND charging water pressure cannot be restored to 940 psig within 20 minutes.

B. Monitor CRD temps and attempt to return 1B CRD Pp to service. Scram the reactor if CRD temps exceed 350 0 F.

C' Manually Scram the reactor and place the Mode Switch in Shutdown immediately.

D. Attempt to restart the 2A CRD Pump. Ifthe pump fails to start then commence a normal shutdown.

References:

2-AOI-85-3 Rev.20 pg 3 and 4 A. Incorrect since these actions are taken if Rx pressure is >900 psig.

B. Incorrect since you do not wait to return a CRD Pp to service.

C. Correct answer.

D. Incorrect since you do not perform a normal shutdown and you do not try to restart a pump on overcurrent.

RO Tier: SRO Tier: TIG2 Keyword: CRD SYSTEM Cog Level: C/A 4.3/4.2 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 62

QUESTIONS REPORT for SRO 2002-301LXRBank

63. 295023AK3.02 001 Interlocks or limit switches on the refueling equipment are provided for specific protective functions.

Which ONE of the following describes these protection devices?

A. The Refueling Interlocks are not-required during fuel handling as long as there is a second qualified individual performing the functions of the interlocks.

B. Jumpering a refueling interlock should not cause the refuel bridge operator any concern as long as a TACF tag is clearly visible at the controls.

C. Switches and interlocks act as a backup protection rather than principle means for stopping travel of the refueling equipment.

D. Fuel handlers may rely on limits and interlocks to terminate refuel equipment travel, as long as they are within their surveillance frequency.

References:

O-GOI-100-3A Rev. 29 pg 14 A. Incorrect because Tech Specs do not allow for a second qualified individual to take the place of the refueling interlocks.

B. Incorrect since jumpering interlocks is a concern.

C. Correct answer.

D. Incorrect since the operators should not rely on the interlocks.

RO Tier: T1G3 SROTier: TIGI Keyword: REFUELING Cog Level: MEM 3.4/3.8 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 63

QUESTIONS REPORT for SRO 2002-301 LXRBank 64i 295024EK1.01 001 Given the following conditions:

SSuppression Chamber pressure 53.0 psig

- Drywell temperature 350°F

- RPV pressure 425 psig Which ONE of the following is the reason why the Drywell or the Suppression Chamber is vented under these conditions irrespective of offsite release rates?

A. Pressure capability of the containment will be reached if Suppression Chamber pressure reaches 55 psig.

B. The maximum containment pressure that the vent valves can be opened and closed to reject decay heat will be reached at 55 psig.

C. The maximum containment pressure that the MSRV's can be opened and remain open will be reached at 55 psig.

D. Chugging is prevented ifthe containment is vented prior to reaching 55 psig.

References:

OPLI171.203 Rev. 5 pg 29 and 36 Enabling Objective OPL1 71.203 #8 A. Incorrect since the pressure capability of the containment is approx. 100 psig.

B. Correct answer.

C. Incorrect since the pressure limit for the MSRV's is 65 psig.

D. Incorrect since chugging depends on the amount of non condensibles in the containment.

ROTier: TIGI SROTier: TIG1 Keyword: CONTAINMENT Cog Level: MEM 4.1/4.2 Source: N Exam: BF02301 Test: C Mise: TCK Tuesday, January 21, 2003 07:30:54 AM 64

QUESTIONS REPORT for SRO 2002-301 LXRBank

65. 295025EA2.03 001 The following conditions exist on Unit 3:

Reactor scrammed due to stuck open relief valve.

Reactor pressure 500 psig Suppression Pool level 15 FT Suppression Pool temperature 200'F increasing l°F/min Which ONE of the following actions should the Unit Supervisor direct the RO to perform?

(Refer to 3-EOI-1 and 2)

A. Emergency Depressurize the reactor.

B. Lower reactor pressure to 300 psig without exceeding 100°F cooldown rate.

C. Exit pressure control and enter steam cooling.

D" Lower reactor pressure to 300 psig irrespective of cooldown rate.

References:

3-EOI-1 Rev.5 A. Incorrect since procedure directs lowering reactor pressure to within curve 3 limits irrespective of cooldown rates.

B. Incorrect since procedure directs lowering reactor pressure to within curve 3 limits irrespective of cooldown rates.

C. Incorrect since steam cooling is not required.

D. Correct answer per Curve 3 and override RC/P-7.

RO Tier: SRO Tier: TIGI Keyword: HEAT CAPACITY Cog Level: C/A 3.9/4.1 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 65

QUESTIONS REPORT for SRO 2002-301 LXRBank

66. 295025EK1.06 001 The Unit 2 Mode Switch is in the S/U position with the Unit at normal operating pressure and temperature following a short maintenance outage. The RVLIS system is out-of-service at this time. The Wide Range level instruments (+60 to -155") are reading approximately +34" at this time.

Which ONE of the following describes the accuracy of the instruments under these conditions?

A?' The instruments are showing accurate level indication due to being calibrated for normal operating pressure and temperature.

B. The instruments are NOT showing accurate level indication due to being calibrated for cold shutdown conditions.

C. The instruments are showing accurate level indication because they are within the level range of the instruments.

D. The instruments are NOT showing accurate level indication because the RVLIS system is not providing flow to the reference leg fill lines.

Reference:

OPLI171.003 Rev. 15 pg 20 A. Correct answer.

B. Incorrect since the instruments are calibrated for hot conditions.

C. Incorrect since being within the indicated range doesn't mean that the instrument is accurate.

D. Incorrect since RVLIS has no affect on how the instrument reads. It does have an affect on the indication when a rapid depressurization occurs.

RO Tier: T1GI SRO Tier: TIGI Keyword: LEVEL INSTRUMENTS Cog Level: C/A 3.9/4.0 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:54 AM 66

QUESTIONS REPORT for SRO 2002-301 LXRBank

67. 295026G2.1.12 001 Unit 2 is conducting a plant startup per 2-GOI-100-1A, "Unit Startup and Power Operation", and surveillance 2-SR-3.5.1.8, "HPCI Main and Booster Pump Set

-Developed Head and Flow Rate-Test at 150 psig" is in progress. The following conditions exist:

Suppression Pool water temperature is 98 0F and rising.

Both loops of Suppression Pool cooling are in service.

The Unit Supervisor is implementing the actions of EOI-2,"Primary Containment Control".

In accordance with Unit 2 Technical Specifications, which ONE of the following actions is required once Suppression Pool temperature exceeds 105 0 F?

A. Place the Reactor Mode Switch in SHUTDOWN.

BV Suspend all testing that adds heat to the Suppression Pool.

C. Reduce THERMAL POWER to range 7 on IRM's within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Depressurize the reactor vessel to less than 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

References:

Tech Specs 3.6.2.1, Suppression Pool Average Temperature A. Incorrect since the Mode Switch doesn't need to be placed in Shutdown until average Suppression Pool temperature exceeds 1 10'F.

B. Correct answer.

C. Incorrect since this is done if cannot meet Required Action and associated Completion time of Condition A.

D. Incorrect since this is performed if average Suppression Pool temperature exceeds 120 0 F.

RO Tier: SRO Tier: TIGI Keyword: SUPPRESSION CHAMBER Cog Level: MEM 2.8/3.1 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:54 AM 67

QUESTIONS REPORT for SRO 2002-301LXRBank

68. 295028EK3.01 001 Which ONE of the following describes the reason why Emergency Depressurization is required if Drywell Temperature cannot be maintained below 280°F?

A. At this temperature all of the RPV level instruments are affected such that there is no reliable level indication and RPV flooding is required.

BY Primary Containment has reached the structural design limit and actions are required to minimize further release of energy from the RPV.

C. Ensures the increase in Drywell pressure does not result in exceeding the Heat Capacity Temperature Limit.

D. Above 280°F containment failure is emminent which would cause the release rates at the site boundary to reach 10 CFR 100 limits.

References:

OPL1 71.203 Rev.5 pg 26 A. Incorrect since these conditions do not make all of the level instruments unreliable.

B. Correct answer.

C. Incorrect since the containment is threatened and not the EQ equipment.

D. Incorrect since containment failure is not emminent and it would not cause the 10 CFR limits to be exceeded.

RO Tier: T1G2 SROTier: TIG2 Keyword: EOI INSTRUCTIONS Cog Level: MEM 3.6/3.9 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:55 AM 68

QUESTIONS REPORT for SRO 2002-301 LXRBank

69. 295029EKI.01 001 Unit 2 is operating at 100% RTP. The Suppression Pool water level is required to be maintained at > -6.25 inches and < -1.0 inches.

Which ONE of the following is available to protect the containment against overpressurization if Suppression Pool water level is allowed to go above the maximum level? .

A. Drywell Cooling.

B. Reactor Building-to-Suppression Chamber Vacuum Breakers.

C. Drywell Spray system.

D. Residual Heat Removal (RHR) Suppression Pool Cooling System.

References:

Tech Spec Bases Section B 3.6.2.2 pg B 3.6-66 A. Incorrect since these valves protect the Drywell from negative pressure upon inadvertent operation of the Drywell Spray system.

B. Incorrect since these valves protect the Suppression Chamber from negative pressure upon inadvertent operation of the Suppression Pool Spray system.

C. Correct answer.

D. Incorrect since this system is needed to maintain the containment within design temperature limitations.

RO Tier: TIG2 SROTier: TIG2 Keyword: PRIMARY CONTAINMENT Cog Level: C/A 3.4/3.7 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:55 AM 69

QUESTIONS REPORT for SRO 2002-301LXRBank

70. 295030EA1.02 002 The following conditions exist on Unit 2:

Reactor Water Level -60" and steady Reactor Pressure 920 psig Suppression Chamber Level 9 ft Drywell Pressure 1.3 psig

- Suppression Pool Temperature 98 0F RCIC Pump Room Temperature 150 0 F CST Suction not available Operator reports leak on suction header of Torus.

The Reactor Operator informs the Unit Supervisor that the RCIC turbine has tripped.

Which ONE of the following is the most likely cause of the turbine trip?

A. High RCIC Pump room temperature.

B. High RCIC exhaust pressure.

C. Low suction pressure.

D. High Reactor Water Level.

References:

OPI171.040 Rev.18 pg 29 and 30 Enabling Objective OPL1 71.040 #5 A. Incorrect since the High RCIC Room Temperature isolation is at 160 0 F.

B. Incorrect since high exhaust pressure can't happen if Drywell pressure is low.

C. Correct answer based on low Drywell pressure and low torus level.

D. Incorrect since the High Reactor Water Level trip is +51".

RO Tier: T1G2 SROTier: TIGI Keyword: RCIC SYSTEM Cog Level: C/A 3.4/3.5 Source: N Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:55 AM 70

QUESTIONS REPORT for SRO 2002-301 LXRBank

71. 295030EK2.01 001 Unit 3 EOI-2, "Primary Containment Control", has the operators perform the following action if suppression pool water level CANNOT be maintained above 12.75 feet.

Secure HPCI irrespective of adequate core cooling.

Which ONE of the following HPCI system responses will this action prevent?

A.' Overpressurization of the primary containment.

B. Loss of back pressure on the exhaust line.

C. HPCI exhaust check valve chatter.

D. Unstable HPCI operation.

References:

OPI171.203 Rev 5 pg 50 & 51 Enabling Objective OPL1 71.203 #7 A. Correct answer.

B. Incorrect since the exhaust line will still have the backpressure from the torus airspace.

C. Incorrect since water level in the torus doesn't affect the HPCI exhaust check valve.

D. Incorrect since torus water level doesn't affect HPCI operation.

Note: Reordered answers.

RO Tier: T1G2 SROTier: TiGI Keyword: HPCI Cog Level: MEM 3.8/3.9 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:55 AM 71

QUESTIONS REPORT for SRO 2002-301LXRBank

72. 29503 1EA2.04 001 The following conditions exist on Unit 2 after a scram from 100% power:

Reactor Power 6%

SLC INJECTING Reactor Water level UNKNOWN Suppression pool temp 105 degrees F Suppression Pool Pressure 3.0 psig Drywell Pressure 3.0 psig Six ADS valves OPEN RPV injection Stopped and Prevented per Appx. 4 Which ONE of the following indicates the reactor pressure at which injection to the vessel should be reestablished?

(Supply copy of EOI-1 and C-4)

A. When reactor pressure is below 65 psig.

B. When reactor pressure is below 68 psig.

C. When reactor pressure is below 180 psig.

D. When reactor pressure is below 220 psig.

References:

2-EOI-1 Rev. 8 C-4 Rev. 6 A. Incorrect since this is the pressure that needs to be above Suppression Chamber if the reactor will stay shutdown under all conditions.

B. Incorrect since this pressure added to Suppression Chamber pressure is the pressure required to be maintained if the reactor will stay shutdown under all conditions.

C. Correct answer with 6 or more MSRV's open.

D. Incorrect since this is the pressure with only 5 MSRV's open.

RO Tier: SROTier: TIGI Keyword: RPV FLOODING Cog Level: C/A 4.6/4.8 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:55 AM 72

QUESTIONS REPORT for SRO 2002-301 LXRBank

73. 29503 1EK3.05 001 A loss of all high pressure injection systems has resulted in RPV level lowering to TAF.

An emergency RPV depressurization has been directed.

Which ONE of the following states the reason that a minimum of 4 MSRVs must be opened?

A7' Ensures that sufficient steam flow will exist to remove decay heat at low enough pressure for the lowest head ECCS pump to make up for steam flow.

B. Ensures that at the worst case in core life, the APLHGR thermal limit will not be exceeded and inhibit adequate radiant heat transfer.

C. Ensures that the reactor will be depressurized to below ECCS shut off head before the RPV level reaches two thirds core height.

D. Prevents exceeding 1% plastic strain on the hottest fuel pin in the core allowing fuel cladding failure to release radioactive fission products.

References:

OPLI 71.205 Rev. 4 pg 29 A. Correct answer.

B,C and D are incorrect since 4 relief valves open do not affect these conditions.

ROTier: TIGI SROTier: TIGI Keyword: EMERG DEPRESS Cog Level: MEM 4.2/4.3 Source: B Exam: BF02301 Test: C Mise:

Tuesday, January 21, 2003 07:30:55 AM 73

QUESTIONS REPORT for SRO 2002-301 LXRBank

74. 295032EK3.03 001 Which ONE of the following is the basis for the Main Steam Line (MSL) Tunnel high temperature isolation?

A. Protect the integrity of the secondary containment and ensure the continued operability of safe shutdown equipment.

B. Prevent exceeding the Environmental Qualification temperature limits on the MSIV control air solenoids.

C' Minimize radioactive releases to the environment and limit the inventory loss from the reactor under all accident conditions.

D. Limit the escape of radioactivity from the MSL Tunnel to the Reactor Building HVAC system.

PCIS purpose BSEP BANK LOI-CLS-LP-012A*017001 RO Tier: TIG3 SRO Tier: TIG2 Keyword: MAIN STEAM Cog Level: MEM EK3.03 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:55 AM 74

QUESTIONS REPORT for SRO 2002-301 LXRBank

75. 295033EK2.01 002 Unit 2 is in a Refueling outage with work in progress on the Turbine Floor. When the High Pressure Turbine casing is removed the radiation levels increase significantly.

Which ONE of the following describes the indications available to the Control Room Operator due to the increased radiation levels and the actions required to be taken?

A. Turbine Bldg ventilation trips and isolates. The Control Room Operator announces evacuation of turbine floor and contacts RADCON.

B. Reactor Bldg ventilation trips and isolates. SGT starts automatically. The Control Room Operator announces evacuation of the turbine floor and contacts RADCON.

C. TURBINE BLDG AREA RADIATION HIGH Alarm sounds. The Control Room Operator announces evacuation of the turbine floor, contacts RADCON and monitors other alarms with inputs to this annunciator.

D. TURBINE BLDG AREA RADIATION HIGH Alarm sounds. The Control Room Operator notifies Unit Supervisor this is an expected alarm since the turbine casing is being removed.

References:

OPL171.034 Rev.8 pg 16 2-ARP-9-3A Rev.18 pg 31 Enabiling Objective OPLI1711.034 B5 A. Incorrect since Turbine Building vents do not trip.

B. Incorrect since Reactor Building vents do not trip.

C. Correct answer.

D. Incorrect since this is not an expected alarm. The ARP actions should be followed.

RO Tier: TIG2 SRO Tier: TIG2 Keyword: RAD MONITORS Cog Level: C/A 3.8/4.0 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:55 AM 75

QUESTIONS REPORT for SRO 2002-301 LXRBank

76. 295034EK1.01 001 Which ONE of the following describes why the Reactor Zone and Refueling Floor Exhaust Radiation - High allowable values are set at their current levels?

A." They provide timely detection of system process barrier leaks inside containment but are far enough above background levels to avoid spurious isolations.

B. They provide positive indication of system leaks but they are low enough to ensure proper instrument indications.

C. The values are set to ensure the isolation function is fast enough to prevent exceeding the 10 CFR 20 exposure limits at the site boundary.

D. The values are set such that trends are able to be determined before the isolations occur.

References:

Tech Spec Section 3.3 Bases pg B 3.3-251 A. Correct answer.

B,C and D. Incorrect per Bases statement.

RO Tier: T102 SRO Tier: TIG2 Keyword: SECONDARY CONTAINMEN Cog Level: MEM 3.8/4.1 Source: N Exam: BF02301 Test: C Mise: TCK Tuesday, January 21, 2003 07:30:56 AM 76

QUESTIONS REPORT for SRO 2002-301 LXRBank

77. 295036EK2.03 001 A relief valve is leaking on the Unit 2 RBCCW system which is causing the Reactor Building Equipment Drain Sump to fill up. The Reactor Building Equipment Drain Sump Pump has started.

Which ONE of the following identifies the first indication that Radwaste will see due to the increased leakage?

A. Chemical Waste Tank level will increase.

B. Floor Drain Collector Tank level will increase.

C' Waste Collector Tank level will increase.

D. Waste Surge Tank level will increase.

References:

OPL1 71.084 Rev.3 pg 17 A. Incorrect since the water from the Reactor Bldg Equipment Drain Sump goes to the Waste Collector Tank first.

B. Incorrect since the water from the Reactor Bldg Equipment Drain Sump goes to the Waste Collector Tank first.

C. Correct answer.

D. Incorrect since the water from the Reactor Bldg Equipment Drain Sump goes to the Waste Collector Tank first.

RO Tier: T1G3 SRO Tier: TIG2 Keyword: RADWASTE Cog Level: C/A 2.8/3.1 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 77

QUESTIONS REPORT for SRO 2002-301LXRBank

78. 295037EA1.04 001 The reactor has experienced an ATWS and you have been directed to initiate SLC injection. SLC Pump A was started at 0700 with the tank level at 65%. The following conditions exist at 0730 for SLC Pump A:

"* Red Light On

"* Squib Continuity Lights Off

Flow Light On

"* Alarm "SLC Injection Flow to Reactor"

"* Alarm "SLC Squib Valve Continuity Lost"

"* SLC Pressure 1200 psig

"* Reactor Pressure 1000 psig

"* Tank Level 55%

Which ONE of the following is the appropriate action to take?

A. Start SLC Pump B and continue running SLC Pump A.

B." Stop SLC Pump A and start SLC Pump B.

C. Initiate Alternate SLC Injection.

D. Continue running SLC Pump A.

References:

OPL171.039 Rev. 13 pg 17, 26 and 27 2-01-63 Rev. 26 pg 4 Enabling Objective OPL1 71.039 # 4 A. Incorrect since an interlock is installed to prevent running both pumps at the same time.

B. Correct answer since tank level should be down to 55% if SLC Pump A was operating properly.

C. Incorrect since B SLC Pump should be started first.

D. Incorrect since A SLC Pump is pumping at a degraded rate.

RO Tier: TIGI SROTier: TIGI Keyword: SLC Cog Level: C/A 4.5/4.5 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 78

QUESTIONS REPORT for SRO 2002-301LXRBank

79. 295038EK3.01 001 An accicent has happened on Unit 2 which causes radiation levels at the site boundary to reach 11 mRem/Hr. An ALERT has been declared by the Shift Manager.

Which ONE of the following describes why the Emergency Plan was implemented for this condition?

A. Ensures that all individuals are accounted for at the time of the accident.

B. Provides protective measures only for TVA employees and contractors located on the site at the time of the accident.

C. Ensures lines of communication are established between the site and the NRC.

Df Provides protective measures for TVA employees and the public.

References:

OPL171.075 Rev.17 pg 9 Enabling Objective OPL1 71.075 #B1 A. Incorrect since implementing the Emergency Plan does not ensure all people are accounted for.

B. Incorrect since it also provides protective measures for the public.

C. Incorrect since implementing the Emergency Plan does not mean that communication lines are open with the NRC.

D. Correct answer.

RO Tier: TIG2 SROTier: TIGI Keyword: EMERGENCY PLAN Cog Level: MEM 3.6/4.5 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 79

QUESTIONS REPORT for SRO 2002-301 LXRBank

80. 295038G2.3.4 001 Unit 2 has experienced a severe accident.

Which ONE of the following conditions would require EMERGENCY DEPRESSURIZATION per EOI-4, RadioactivityRelease Control?

A. Release rates above an ALERT due to fuel failure and a leak from the suppression pool.

B. Release rates exceeding an Unusual Event classification and there is an unisolable leak from the HPCI steam line.

C. Release rates approaching a General Emergency and there is an unisolable steam leak in the turbine building.

D. Release rates exceeding a General Emergency but no indications of a primary system discharging outside of containment.

References:

EOI-4, Radioactive Release Control, Rev.5 A. Incorrect since lowering reactor pressure would not slow down the leak.

B. Incorrect since release rate is below the alert level and not approaching the General Emergency level.

C. Correct answer.

D. Incorrect since nothing is being discharged outside of primary or secondary containment.

RO Tier: SRO Tier: TIGi Keyword: RADIATION CONTROL Cog Level: MEM 2.5/3.1 Source: B Exam: BF02301 Test: S Mise: TCK Tuesday, January 21,2003 07:30:56 AM 80

QUESTIONS REPORT for SRO 2002-301 LXRBank

81. 300000K4.03 001 The Raw Cooling Water regulating valve to the "A" Control Air Compressor has failed closed.

Which ONE of the following conditions would trip the "A" Control Air Compressor?

A/ Air discharge temperature reading 312 0 F.

B. Air discharge temperature reading 128 0 F.

C. Lube oil temperature reading 1750 F.

D. Seal Air Pressure reading 8 psig.

References:

OPL171.054 Rev.4 pg 13, 41 and 45 Learning Objective OPL1 71.054 #2 A. Correct answer.

B. Incorrect but it is correct for the G Compressor.

C. Incorrect since Lube Oil Hi Temp trip is 180°F.

D. Incorrect but it is correct for the G Compressor.

RO Tier: T2G2 SRO Tier: T2G2 Keyword: CONTROL AIR Cog Level: MEM2.8/2.8 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 81

QUESTIONS REPORT for SRO 2002-301 LXRBank

82. 500000EK2.07 001 An event has occurred on Unit 3 with the following conditions present:

-Drywell Pressure 30 psig Drywell Temperature 275 0 F SGT Systems "A" out of service, "B" and "C" fail to start SGT Inlet pressure - +0.5 psig CAD System Shutdown The Unit Supervisor has ordered the Drywell to be vented due to high H2 concentrations per Appendix 12.

Which ONE of the following describes the reason why the Drywell CANNOT be vented at this time?

A/vSGT System is NOT in operation.

B. SGT System inlet pressure is too high.

C. Drywell pressure is too high.

D. Drywell temperature is too high.

References:

3-01-83 Rev. 17 pg 5 OPLI171.032 Rev.10 pg 21 Enabling Objective OPLI171.032 #B.4 A. Correct answer.

B. Incorrect since the pressure for the SGT inlet is .79 psig.

C. Incorrect since High Drywell pressure prevents nitrogen purge.

D. Incorrect since Drywell Temp does not affect venting the containment.

RO Tier: TIG1 SRO Tier: TIGI Keyword: DRYWELL VENTING Cog Level: C/A 3.2/3.7 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 82

QUESTIONS REPORT for SRO 2002-301 LXRBank

83. 60000002.4.27 001 A fire rated door that is required to be OPERABLE is found to be propped open with work in progress on one side of the door. There is no fire detection equipment available to protect either side of the fire door. The door is located in a contamination zone.

Which ONE of the following is the MINIMUM action that must be taken to compensate for having the fire door propped open per BFNP FIRE PROTECTION PLAN?

A. Establish a roving hourly fire watch to monitor the both areas until the door is restored to an operable status.

B. If hot work is to be performed in either of the adjacent rooms, establish a continuous fire watch on either side of the open door.

C. No compensatory actions are required as long as the fire door is verified to be closed at least one time daily.

OW Establish a continuous/dedicated fire watch to monitor the impaired fire door area until the door is restored to an operable status.

References:

Fire Protection Plan Vol.1 Rev.17 pg 25 Modified so that the examinee must determine that the door is inoperable.

A. Incorrect since a continuous/dedicated fire watch is required until the door is restored to operable status.

B. Incorrect since a continuous/dedicated fire watch is required until the door is restored to operable status.

C. Incorrect since the fire door is inoperable and compensatory actions are required.

D. Correct answer.

RO Tier: SRO Tier: TIG2 Keyword: FIRE PROTECTION Cog Level: C/A 3.0/3.5 Source: M Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:56 AM 83

Manual #: Fire Protection Report PLANT: BFN UNIT: (s): 2/3 PAGE: 75 of 557 Vol. 1 TITLE: Fire Protection Plan REV: 0020 9.3/9.4 FIRE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATING AND SURVEILLANCE REQUIREMENTS (continued) 9.3.11.G FIRE-RATED ASSEMBLIES 9.4.11!G FIRE-RATED ASSEMBLIES

1. All fire barrier assemblies, including 1. Each of the required fire-rated walls, floor/ceilings, conduit wraps, assemblies and penetration sealing and other fire barriers; separating devices shall be verified OPERABLE:

fire areas or separating systems important to safe shutdown within a a. At least once per 18* months by fire area; and all sealing devices in performing a visual inspection of fire rated assembly penetrations, the exposed surfaces of each fire including fire doors shown in Table rated assembly",

9.3.1l.E, fire dampers shown in Table 9.3.1l.F, and fire-rated cable and b. At least once per 12 months by piping penetration seals, shall be performing a visual inspection of OPERABLE at all times. Fire barriers 20% of the fire damper and the are identified by compartmentation associated hardware. Dampers shall drawings 47W216-51 through 62. fNRC/C'] be selected, such that each damper will be inspected at least once per 5 years.

a. With one or more of the required * (Once per operating cycle for areas fire-rated assemblies and/or inside the Steam Tunnel and top of sealing devices inoperable, torus areas of Unit(s) 2 Tnd/or 3 Reactor Building(s).

a.1 Establish a continuous fire watch on one side of the ** (Includes walls, floors, ceilings, affected assembly if no fire fire wraps, structural fireproofing, detection (as listed in Table and penetration seals.)

9.3.11.A) is available to protect either side of the inoperable barrier.

or a.2 Establish a roving fire watch on one side of the affected assembly if fire detection (as listed in Table 9.3.1l.A) is OPERABLE to protect at least one side of the inoperable barrier. ENCO 880210002]

Manual #: Fire Protection Report PLANT: BFN UNIT: (s): 2/3 PAGE: 76 of 557 Vol. 1 TITLE: Fire Protection Plan REV: 0020 9.3/9.4 FIRE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATING AND SURVEILLANCE REQUIREMENTS (continued) 9.3.1l.G FIRE-RATED ASSEMBLIES (contd.) 9.4.1l.G FIRE-RATED ASSEMBLIES (contd.)

a.3 No compensatory measures required 2. Each of the required fire doors shall if fire detection (as listed in be verified OPERABLE by inspecting the Tables 9.3.1l.A) is OPERABLE to automatic hold-open, release, and protect both sides of the closing mechanisms and latches at inoperable barrier, least semiannually and by verifying:

a. The OPERABILITY of the fire door supervision system for each electrically supervised fire door by performing a CHANNEL FUNCTIONAL.

TEST at Least monthly.

b. That each locked-closed fire door is verified closed at least weekly.

c. That doors with automatic hold-open and release mechanisms are free of obstructions at least daily and perform a FUNCTIONAL TEST of these:

mechanisms at least once peri*

months.

d. That each unlocked normally closed fire door without electrical.

supervision is verified close at least daily.

9.3.11.H OPEN FLAMES, WELDING AND BURNING IN THE CABLE SPREADING ROOM Hot work activities during plant operation, shall only be permitted on a case by case basis, after a satisfactory evaluation by Fire Operations and Site Engineering (Mechanical). The evaluation will specifically address accessibility to the area by site fire brigade, accessibility of manual fire fighting equipment, ventilation, and exposure protection. In addition, a member of the fire brigade shall be present in the area and a continuous fire watch shall be provided during performance of any hot work activity.

Manual #: Fire Protection Report PLANT: BFN UNIT(i):2/3 PAGE 105 of 557 Vol. 1 TITLE: Fire Protection Plan REV 0020 TABLE 9.3.11.E FIRE RATED DOORS DOOR NO. LOCATION DETECTION PANEL DOOR NO. LOCATION DETECTION PANEL 30 P7-N 519 NONE 500 R14-P 593 0-LPNL-25-555 31 P7-R 519 NONE 501 R14-T 593 2-LPNL-25-546 34 RS-R 519 2-LPNL-25-545 506 R16-T 593 3-LPNL-25-546 35 RP-N 519 2-LPNL-25-545 510 R21-P 593 0-LPNL-25-555 36 R14-N 519 2-LPNL-25-545 514 R21-R 593 0-LPNL-25-555 37 R14-R 519 2-LPNL-25-545 531 P2-N 606 0-LPNLL25-556 40 R15-R 519 NONE 539 RP-S 593 NONE 41 R15-N 519 NONE 541 R14-S 593 0-LPNL-25-555 42 P7-T 541 NONE 600 P2-N 617 0-LPNL-25-557 43 R8-T 541 2-LPNL-25-545 630 P1-P 621 0-LPNL-25-557 44 R14-T 541 2-LPNL-25-545 631 RI-S 621 NONE 45 R15-T 541 NONE 632 R2-S 621 NONE 221 R7-N 565 1-25-286 632A RP-S 621 NONE

235 R8-M 565 2-LPNL-25-545 635 R6-T 621 1-25-303

237 R9-M-565 2-LPNL-25-545 637 R2-S 621 NONE 240 R8-T 565 2-LPNL-25-545 640 R8-T 621 2-LPNL-25 7 547 242 R14-T 565 2-LPNL-25-545 642 R13-P 621 O-LPNL-25-557 244 R14-N 565 2-LPNL-25-545 643 R13-S 621 0-LPNL-257557

248 R15-M 565 3-LPNL-25-545 644 R14-S 621 0-LPNL-25-557

250 R16-N 565 3-LPNL-25-545 644A R14-S 621 0-LPNL-25-557 298 R1-Q 583 0-LPNL-25-544 647 R15-T 621 2-LPNL-25t547 455 P2-N 593 O-LPNL-25-555 649 R13-S 621 0-LPNL-25-557 460 R3-N 593 0-LPNL-25-555 651 R16-T 621 3-LPNL-25-547 462 P4-N 593 0-LPNL-25-555 654 R21-P 621 0-LPNL-25-557 4 62A P4-N 593 0-LPNL-25-555 655 R21-R 621 0-LPNL-25-557 466 R9-N 593 0-LPNL-25-555 656 R20-R 621 0-LPNL-25-'557 4 6 6A R10-N 593 0-LPNL-25-555 658 R20-R 621 0-LPNL-25-557 468 RI0-N 593 0-LPNL-25-555 672 P8-T 639 2-LPNL-25-547 476 R18-N 593 0-LPNL-25-555 673 R14-T 639 2-LPNL-25-547 476A R19-N 593 0-LPNL-25-555 810 U3DGB 565 3-LPNL-25-543 479 R19-N 593 0-LPNL-25-555 811 U3DGB 565 3-LPNL-25-543 482 R20-N 593 0-LPNL-25-555 812 U3DGB 565 3-LPNL-25-543 485 RI-P 593 0-LPNL-25-555 824 U3DGB 583 3-LPNL-25-543 490 P6-T 593 1-25-287 825 U3DGB 583 3-LPNL-25-543 497 P8-T 593 2-LPNL-25-546 827 R21-P 593 0-LPNL-25-555

[NRC/C] Fire doors and associated fire detection panels [NRC NCO 890198001]

The following list of doors must be breached simultaneously in order to have a fire protection assembly out of service requiring compensatory measures to be taken:

Door(s) 30/35, 31/34, 36/41, 37/40, 42/43, 44/45.

NOTE: In accordance with SER-Appendix R Exemptions A02-881027-003, Section 3.1.2.1 fire doors of the main steam and feedwater piping tunnels, door nos. 220, 239 and 252 are exempt from surveillance, testing and the administrative control program.

In accordance with the Fire Hazard Analysis for Fire Areas 1 and 3, Fire Zone Doors 490 and 635 are not required within Fire Area 1 when Unit 2 and/or Unit 3 is the only operating unit. Fire Zone Doors 506 and 651 are not required within Fire Area 3 when Unit 2 and/or Unit 1 is the only operating unit.

Credit will be taken for detection capabilities within the Reactor Building due to lack of any combustibles within the air locks.

zu 0

7 2

Manual #: Fire Protection Report PLANT: BFN UNIT(s): 2/3 PAGE 107 of 557 Vol. 1 TITLE: Fire Protection Plan REV 0020 TABLE 9.3.11.F (continued)

FIRE RATED DAMPERS DAiPER ROOM COORDINATE Dfl44PER ROOM ROC'TIO iI 'COORDIHAT N1C. UNID HO. LOCAT i ON LOCATION NO. U81II 11O. LOCA-.TiO1 DELETED 145 0-31-2520 .ATT RI! I R,-3 144 631 I'--- SET ,!1 1 146 1-31-25% AYUINST RN 1 R2-S 147 1-91-2518 R4 -H 602 6':'-

R4 -N 149 -5 -2 1 ,BHR 1 143 1-31-2525 MG SET RN 1 60%2 15: 2-~31-2578 MG SET RH 2 R9-N! 151 2-31-2647 RI,'-I 602 6:3 HG SET RH 2 RIO-H 153 DELETED 152 2-31-2512 603 R~i1-P 154 E'ELLTED 155 "-31-2510 C07-1 BATT RH 6092 157 3-31-1303 4C" BBR £3? 3 9J3 DGE?

156 3,-31 -!0? 4?,", EPBR D'SE 3 U3 DGE 53K' 4K:. SDBR 32ED 3 D'S?

3-31-13:9 49,; SD'R BED! U3 D'3E DG3E 3 579, 3-31-1313 4??." SDER lEA U3 G 16,:> 4KV SDE:R 3E£ T83 CGE£ 161 3-31-1311 531 1 61 1 i 1 --

4[71 SDER 3 2C 163 2-:KPD-31-2653 IN- SET RI 2 DGB? 3, A&B 16-I BBR 1 Ri0-H 165 2-XFD-31-2655 .ATT R! 2 U3 DGB CE7 SERD RN B KRi 9-I 167 3-31-1314 .TAIRWAY U3 163 ,1-39-803:;

590 611 CB IA RI C-R 3-31-164? 4f)- SDBR 63, 3-31-257 4(7V SDBR 3B R21 -R 171 3-31--577 40'.,V SDER 3? R21-K 171 635 63.

173 1-31--2']0 U1 RB/250V R2-4 174 1-31 -2001 UI RB 250V E/B?

F/BE RH 63: RN US R? 250V' s/BE: R13-R 177 2-31-,y0, U- RE: "50V R20-H 171 2-31-2007 E/BE RII 63: RH R10)-E 132 1-31-2016 UI PREP NG 3 131 17-_. U' CABLE SPRD RHA 606 606

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Manual #: Fire Protection Report PLANT: BFN UNIT($): 2/3 PAGE 109 of 557 Vol. 1 0020 TITLE: Fire Protection Plan IIIREV 0020 TABLE 9.3.11.F(continued)

FIRE RATED DAMPERS E4NI PER ROOM COORDIIATE DAM1PER ROOY! COORDIHiAT NiO. *1:::D Ii,. LOCATION LOCATI ON l1O. T T11D 1CO. LOC'ATION LOtATICI

-64-6' ý4 U' R" BLDG Rg-P 3-64-,-5, U R EBLDG RI-U

'31 9.3 223 3-64-6(1 U3 R, BLDG R20-P -24 3-,4-6'3 3 R"" BLDG Ri,6-P S503 591 3-64-6C4 U3 RI BLDG RI6-F -349 '-64-6,.5 Ut RI BLDG Ri3-P 503 0-1 131 1-64-606 UI R- BLDG P9-P '3' 67 UT RI' ELDG F.,3-U Is - 6Ž1 1-64-6)3 Ut RŽ: BLDG P11-U 234 3-,64-693ý US RIK'BLDG FPi-U 611 6211 S-64-605 U3 RX BLDG RIO-P 136 3-64-6,6 U3 RI ELDG P16-P 6-1 141 1-64-69A U, R" BLDG Ri4-T 141 1-64-61, Ut RP LBLD.G R13-T 639 639 143 611 UT RF EBLDG Ri3-U 144 1-64-61 U2 &I BLDG RlK-s 639 639. v 145 1-64-613 U1 REBLDG R1I-, 146 C4-3-64-619 R5: EBLD' tj'3U 639 A '

1-64-699 US RI BLDG P1!-S 143 4-;.4-13 U R: EBLDG RP1-s 6 0 639 149 3-64-611 U3 RI BLDG P1g-S 15) 1-64-614 Ut RI ELDG PI1-0 639 621 153 3-64-614 U3 RI BLDG R19-QJ 55.- 64 U^ RI. BLDG P11-P 621 593 2-64-615 U3 R` 2LDG R!9-F, 268 16 -. RZ' BLDG3 Rl!2rP 593 C21 1-64-617 UT RI ELDG RI2-Q 060 170 2-64-613 U4 RI BLDG RIE-Q 021 611 RS-S .171 2-64-.610 UI RII BLDG P!0-S 211 1-64-619 U" R" BLDG 621 Eli 174 1-64-615 Ut£ R1 BLDG RIG-F 173 61I U2 RX BLDG R14-T 1621 621

'75 1-64-6C3 U- R` BLDG RIO-P 176 2- 64-6124 U' RI BLDG RiO-P 611 211

Manual #: Fire Protection Report PLANT: BFN UNIT(s):2/3 PAGE 110 of 557 Vol. 1 TITLE: Fire Protection Plan REV 0020 TABLE 9.3.11.F (continued)

FIRE RATED DAMPERS D?-2uBER RC'C'!I COORDT:[ATE DCX-IPER ROT' I C OR DI 1i i..T DhIFE UIIIE , '. LOCAT IcOY! LOC..MZ i HIO. rlHID 11C,. LOA io LOCAT I (iI 1-,64-611 Ut RI B:LD' RIG'-P G-4-66 U - RG EBLDG Ri 1-U Ro - Ij RD-P "27 9. 3i-4-516 U3 R' ELDG 3-,4-617 U3 R: ELt,3 RI 9

,621 5-64i-613' Uo RP ELDG' 3-64-619 UJ R:: BLDG RI5-T Rh1-S "283 .- 64-610 U3 FR:: BLDG -84 64-64 1 US RI: ELDG 3-,i-,325 [IS R:: BL* ,,D .5-64-616 US3 RI BLD3 R17-P 21.

3-64 -611 TU RI ELD3

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623 2 -- 1-u635 4%" SDEF XE F.-S 611 IU3 R:: ELD, 639 F2 ,-U 30I 6:1 US RI ELD3 R j-IT

'- -- 64-6,7 US RIF BLDG

QUESTIONS REPORT for SRO 2002-301LXRBank

84. G2.1.10 001 You are in the position of SM. There are several NRC personnel in the control area conducting initial reactor operator examinations. You feel these activities are interfering withldaily plant activities and you are concerned with the number of people in the control room.

Which ONE of the following actions should be taken to address your concerns per SPP 10.0, Plant Operations?

A. Notify the NRC personnel that they must leave the control room immediately.

B. Contact the NRC Senior Resident Inspector for resolution.

C. Contact the Plant Manager for resolution.

D. Nothing. NRC personnel shall be allowed to enter or be present in the control room for any activity.

SPP10.0, Plant Operations Rev.2 pg 9: Section 3.6 Control Room Activities A. Incorrect since the NRC have unfettered access to the control room.

B. Incorrect since the Plant Manager should be notified to resolve the issue per SPP 10.0.

C. Correct answer since the SM, US and the UO have the authority to restrict access to or remove personnel from the control room. NRC personnel shall be allowed to enter or be present in the control room. Ifthe SM has concern over the number or activities of NRC personnel, he shall contact the Plant Manager for resolution.

D. Incorrect since the Plant Manager should be contacted to resolve the situation per SPP 10.0.

RO Tier: SRO Tier: T3 Keyword: COMMAND AND CONTROL Cog Level: MEM 2.7/3.9 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:56 AM 84

QUESTIONS REPORT for SRO 2002-301 LXRBank

85. G2.1.12 001 The following plant conditions exist on U-2.

--Mode 5

- Reactor Mode Switch REFUEL

- Spent Fuel Pool Shuffle in progress The following plant conditions exist on U-3.

- Mode 4

- Reactor Mode Switch SHUTDOWN

- Preparations for startup underway Which ONE of the following describes the requirements for the SBGT System?

(Tech Spec reference 3.6.4.3 provided)

A. NOT required to be operable B. required to be operable for U-2.

C. required to be operable for U-3.

DO required to be operable for both units.

References:

Technical Specification 3.7.B A, B, C are incorrect because there are fuel movements in progress within secondary containment which is common to both units.

D. Correct answer.

RO Tier: SRO Tier: T3 Keyword: SBGT Cog Level: C/A 2.9/4.0 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:57 AM 85

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SGT subsystem A.1 Restore SGT subsystem 7 days inoperable, to 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 of Condition A not AND met in MODE 1, 2, or 3.

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

BFN-UNIT 2 3.6-51 Amendment No. 253

SGT System 3.6.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and ----- --------------- NOTE -------

associated Completion LCO 3.0.3 is not applicable.

Time of Condition A not met during movement of irradiated fuel assemblies C.1 Place two OPERABLE Immediately in the secondary SGT subsystems in containment, during operation.

CORE ALTERATIONS, or during OPDRVs. OR C.2.1 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.

AND C.2.2 Suspend CORE Immediately ALTERATIONS.

AND C.2.3 Initiate action to suspend Immediately OPDRVs.

D. Two or three SGT D.1 Enter LCO 3.0.3. Immediately subsystems inoperable in MODE 1, 2, or 3.

(continued)

BFN-UNIT 2 3.6-52 Amendment No. 253

SGT System 3.6.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two or three SGT E.1----------- NOTE-----

subsystems inoperable LCO 3.0.3 is not during movement of applicable.

irradiated fuel assemblies in the secondary containment, during Suspend movement of Immediately CORE ALTERATIONS, or irradiated fuel assemblies during OPDRVs. in secondary containment.

AND E.2 Suspend CORE Immediately ALTERATIONS.

AND E.3 Initiate action to suspend Immediately OPDRVs.

BFN-UNIT 2 3.6-53 Amendment No. 253

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each SGT subsystem for > 10 31 days continuous hours with heaters operating.

SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter Testing with the VFTP Program (VFTP).

SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 24 months actual or simulated initiation signal.

SR 3.6.4.3.4 Verify the SGT decay heat discharge 12 months dampers are in the correct position.

BFN-UNIT 2 3.6-54 Amendment No. 255 November 30, 1998

QUESTIONS REPORT for SRO 2002-301LXRBank

86. G2.1.28 001 Which ONE of the following describes the purpose and function of the TIP Shear Valves?

A. Provides an automatic emergency means to seal the TIP guide tube should the guide tube leak with the TIP probe extended and unable to be retracted. The shear valve cuts the cable and closes off the guide tube.

B./ Provides a manual emergency means to seal the TIP guide tube should the guide tube leak with the TIP probe extended and unable to be retracted. The shear valve cuts the cable and closes off the guide tube.

C. Provides the normal means to seal the TIP guide tube should the guide tube leak with the TIP probe extended and unable to be retracted. The shear valve is held closed by a fail-safe spring.

D. Provides a manual emergency means to seal the TIP indexer should the indexer leak nitrogen. The shear valve closes off the index mechanism for the leaking Tip guide tube.

References:

OPL1 71.023 Rev.4 pg 10 and 11 Enabling Objective OPL171.023 B2 A. Incorrect since the shear valve does not operate automatically.

B. Correct answer.

C. Incorrect since the shear valve is not a normal means to isolate the guide tube.

D. Incorrect since the shear valve does not isolate the index mechanism.

RO Tier: T3 SRO Tier: T3 Keyword: SYSTEM STATUS Cog Level: MEM 3.2/3.3 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:57 AM 86

QUESTIONS REPORT for SRO 2002-301 LXRBank

87. G2.1.3 001 The on-coming Unit 3 Board Unit Operator (BUO) has been on vacation for 7 days.

The BUO is preparing to assume shift at 0700 on 12/30/2002.

Which ONE of the following is the date, at a minimum, that the BUO must review back to concerning the Unit 3 Narrative Log?

A. 0700 on 12/29/2002.

B. 1500 on 12/27/2002.

C' 0700 on 12/25/2002.

D. 1500 on 12/23/2002.

References:

SSP-12.1, Section 3.12.2, page 64 C. Correct answer.

A, B and D are incorrect since the operator must only review the previous 5 days in the narrative log.

Note: Did not have a copy of the procedure to verify answer.

RO Tier: T3 SRO Tier: T3 Keyword: ADMIN Cog Level: C/A 3.0/3.4 Source: M Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:57 AM 87

QUESTIONS REPORT for SRO 2002-301 LXRBank

88. G2.2.12 001 The Unit Operator has just completed the required readings for his shift and documented them in 2-SR-2, Instrument Checks and Observations.

Which ONE of the following lists the individuals that are qualified to perform an independent review of the readings performed by the Unit Operator?

A. RO or SRO.

B. RO or STA.

C. Ops Manager or SRO.

D. STA or SRO.

References:

2-SR-2, Rev.29 pg 8 A. Incorrect since RO cannot perform independent review.

B. Incorrect since RO cannot perform independent review.

C. Incorrect since Ops Manager cannot perform independent review unless he is an SRO or qualified STA.

D. Correct answer.

RO Tier: T3 SRO Tier: T3 Keyword: SURVEILLANCE REQUIRE Cog Level: MEM 3.0/3.4 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:57 AM 88

QUESTIONS REPORT for SRO 2002-301 LXRBank

89. G2.2.14 001 Unit 3 is in a Refueling Outage with a 24 month Group 2 PCIS surveillance in progress.

The Test Director assigned to coordinate the surveillance is in the control room monitoring the test when the SGT system receives an actual Auto Start signal. The Test Director stops the surveillance until the conditions that caused SGT to auto start are corrected.

Which ONE of the following describes the actions the Test Director must perform to complete the surveillance?

A/ Re-verify the Initial conditions with Operations and ensure equipment performance will not be jeopardized by completing the remainder of the procedure.

B. Discard the surveillance test that was interupted and repeat the surveillance from the beginning.

C. Verify with Operations that they are able to support the remainder of the surveillance test and continue the surveillance by re-performing the last step that was completed and continuing until the test is complete.

D. The Test Director and Operations can review the procedure to verify all of the Acceptance Criteria is met. Ifthe Acceptance Criteria is met then the surveillance can be signed off as complete.

Reference:

SPP-8.1 Rev.2 Pg 9 OPLI71.078 Rev.11 Pg 14 Enabling Objective OPL1 71.078 #611 A. Correct answer.

B. Incorrect since the previous procedure is not discarded.

C. Incorrect since the Test Director and Operations must re-verify the initial conditions prior to restarting the test and there is no direction to repeat a step.

D. Incorrect since the surveillance cannot be signed off as complete until all steps are completed or N/A'd.

RO Tier: SRO Tier: T3 Keyword: TESTING Cog Level: MEM 2.1/3.0 Source: N Exam: BF02301 Test: S Mise: TCK Tuesday, January 21, 2003 07:30:57 AM 89

QUESTIONS REPORT for SRO 2002-301 LXRBank

90. G2.2.3 001 Assume that one of the 48V DC inverters that supplies 120V AC to the Control Room annunciators has failed.

Which ONE of the following describes the effects this will have?

Af Ifthe failure is on Unit 1 a buzzer will sound and a white light will illuminate.

B. Ifthe failure is on Unit 1,2 or 3 a buzzer will sound and a red light will illuminate.

C. Ifthe failure is on Unit 2 the power supply will auto swap to Battery Board 2.

D. Ifthe failure is on Unit 3 the power supply will auto swap to Battery Board 3.

References:

OPLI171.037 Rev.8 pg 16 Enabling Objective OPL1 71.037 #B8 Note: Taken from 2001 Exam.

A. Correct answer.

B. Incorrect since a red light will not illuminate on Unit I but it will illuminate on Units 2 and 3.

C. Incorrect since the power supply does not swap.

D. Incorrect since the power supply does not swap.

RO Tier: T3 SRO Tier: T3 Keyword: DC SYSTEMS Cog Level: MEM 3.1/3.3 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:57 AM 90

QUESTIONS REPORT for SRO 2002-301 LXRBank

91. G2.2.5 001 The System Engineer for the RHR System has performed an Evaluation for a Temporary Alteration to the system. The temporary alteration is not Quality Related.

The Systems Engineering Manager has approved the temporary alteration.

Which ONE of the following lists the individuals that are required to give approval prior to installing the temp alt?

(Supply copy of SPP)

A. Shift Manager, PORC, and Plant Manager.

B. Regulatory Assurance Manager and Plant Manager.

C. Shift Manager and Plant Manager.

D. Shift Manager, Regulatory Assurance Manager, and PORC.

References:

OPLI71.079 Rev.11 pg 9 A. Incorrect since PORC is not required since the temp alt is not quality related.

B. Incorrect since the Regulatory Assurance Manager is not a required signature.

C. Correct answer since temp alt is not Quality related.

D. Incorrect since Regulatory Assurance Manager and PORC are not required signatures.

RO Tier: SRO Tier: T3 Keyword: TEMP ALT Cog Level: C/A 1.6/2.7 Source: N Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:57 AM 91

Tennessee TITLE SPP-9.5 Valley Authority Rev. 4 Page 1 of 21 TVAN STANDARD TEMPORARY ALTERATIONS Quality Related 0 Yes E] No PROGRAMS AND PORC Required Z Yes [I No PROCESSES 10CFR50.59 Review E5 Yes 0 No Effective Date 6/29/2001 RESPONSIBLE PEER TEAM: Engineering & Technical Services Organization CONCURRENCES Doug F. Helms 6/13/01

Primary Sponsor Date Jon R. Rupert for JAB 6/25/01 Peer Team Mentor Date APPROVAL For Nuclear Assurance Sponsored SPPs N/A General Manager,NA Date Karl W. Singer 6/27/01

Senior Vice President,Nuclear Operations Date

Site-specific changes are approved by Site Sponsor and Site Vice President (see PCF)

TVA 40480 [05-2000J Page 1 of 1 ;SPP-2.1-1 [05-30-200 ]

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 2 of 21 REVISION LOG Revision Effective Pages Description of Revision Number Date Affected 0 12/19/97 All Initial issue. This SPP replaces SSP-12.4 (SQN and BFN) and (BFN, COC, SSP-12.04 (WBN).

WBN) 4-3-98 Yso (SQN) 4/2/98 1 07/31/98 2, 9,10, Changed 3.4.D to 3.4.C. Changed the reference of NEDP-3 to EAI 18 3.10 for WBN. Also changed "Chemistry" to "Environmental Section" on Form SPP-9.5-1. (Minor/editorial changes).

2 12-31-98 2,3,5,6, Added TPS Temporary Alteration Order (TAO) process to 9,12, 13, exclusions, Section 2.2.L. Reference CHPER970040.

19,22 Changed quarterly reviews to semi-annual reviews for BFN and SQN. Clarified paragraph 3.1.B.1.g. Remove step to provide copy of review sheet to plant manager on review form.

3 7/17/00 2, 4, 5, 7, These changes were made as a result of revisions to SPP-9.3 and 11, 13,22 9.4. The safety assessment was eliminated in SPP-9.4. SPP-9.3 was revised to add a technical evaluation and a determination that a change was safe. The wording in this SPP was revised to eliminate references to the safety assessment and to identify SPP-9.3 as the procedure to look at for guidance on performing a technical evaluation. Minor/editorial changes.

4 6/29/01 2, 4, 6-14, Revised to incorporate changes per BFPER 00-009052-000, 16,17,18, Corrective Action 3 and BFPER 00-009447-000, Corrective Action 20, 21 5. Added Maintenance Rule evaluation. Incorporated recommendations per Self Assessments SA-98-14 and SA-WBP 00-017-01.

TVANSTANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 3 of 21 TABLE OF CONTENTS Page 1 of 1 Section Title Page Revision Log ........................................................................................................................................

Table of Contents ................................................................................................................................

1.0 P UR P O S E ..............................................................................................................................

2 .0 S C OP E ...................................................................................................................................

2 .1 Ap p lic a b ility .............................................................................................................................

2 .2 Ex c lu sio ns ...............................................................................................................................

3.0 INSTRUCTIONS .....................................................................................................................

3.1 Initiation and Evaluation ..........................................................................................................

3.2 Recommendations and Approval .............................................................................................

3 .3 In sta llatio n ...............................................................................................................................

3.4 Return to Normal .....................................................................................................................

3 .5 R e v is io n s ................................................................................................................................

3 .6 R ev iew s ..................................................................................................................................

3 .7 Tra c king ..................................................................................................................................

4.0 RECORDS ..............................................................................................................................

4.1 QA Records ............................................................................................................................

4.2 Non-QA Records .....................................................................................................................

5.0 DEFINITIONS .........................................................................................................................

APPENDIXES Appendix A TACF Flowchart .................................................................................................. 15 Appendix B Requirements for Hanging Temporary Alteration Tags ........................................ 16 Appendix C Material Requirements for Temporary Alterations ................................................ 17 FORMS SPP 9.5-1 Temporary Alteration Control Form ........................................................................ 18 SPP 9.5-2 TACF Review Form ............................................................................................... 21

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 4 of 21 1.0 PURPOSE This SPP provides the requirements for controlling temporary alterations (TA) to the systems, structures, and components of TVA's Nuclear Plants in a manner which ensures operator awareness, conformance with design basis and operability requirements, and preservation of plant safety and reliability.

2.0 SCOPE 2.1 Applicability A. This SPP shall be used for quality and non quality-related systems, structures, and components and provides requirements for installing, controlling, returning to normal, and documenting TAs. The TA process should not be used to circumvent the permanent design change process and TAs are intended to be minor in scope, of short duration, and few in number.

B. TAs and Temporary Alteration Control Forms (TACFs) are quality-related when the affected equipment is designated as safety related or quality-related on the site Q-list.

2.2 Exclusions The following TAs are excluded from this SPP provided the programs which govern them require consideration of effect, limitations, any special actions, tests demonstrating return to normal, the need for a 10CFR50.59 review per SPP-9.4, and time frame which the TA may remain installed.

A. TAs positively identified and controlled in procedures or Work Orders (WO) which meet the following requirements.

1. Procedures or WO shall be approved.

2. Operations notification at procedure or WO initiation and completion unless the TA is controlled by an Operation Instruction.

3. Documentation of installation and Return to Normal (RTN) of the TA, or identification of the Work Initiating Document (WID) that performs installation and RTN of the TA is contained in the procedure or WO.

4. Tagging of lifted leads or tagging of other TAs when using Information Tags when the TA is to remain in place unattended beyond the end of shift unless controlled by a WID.

B. TAs performed by WIDs on equipment that has been removed from service for maintenance, repair, lay up, calibration, or troubleshooting and are RTN prior to equipment being returned to service.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 5 of 21 C. Temporary connection of sample monitoring equipment to sample system piping provided the connection does not circumvent any automatic isolation features.

Hoses to system vents and drains for the purpose of draining or filling and venting. Hoses to normal service connections such as service air or breathing air. These connections are excluded provided the flow path is not changed.

D. Temporary annunciator disabling/enabling performed in accordance with applicable procedures.

E. Breaches performed in accordance with applicable procedures.

F. Fire Protection Impairment Permits.

G. Scaffolding erection performed in accordance with applicable procedures.

H. Temporary shielding and catch containment installation performed in accordance with applicable procedures.

1. Furmanite repairs on non-quality-related components which do not affect component function.

J. TAs installed by and evaluated under WIDs for the purpose of troubleshooting in-service equipment and are removed at completion of troubleshooting.

K. Equipment outside configuration control as defined by the Design Change Process Program.

L. TPS Temporary Alteration Orders issued by the Power System Dispatcher in accordance with TPS-ESO-Operating Letter.

Any temporary alteration to a transmission system component that adversely affects the site's offsite power requirements, or that could challenge safety systems by creating transients that could decrease the margin of nuclear safety, or increase the potential for a unit trips are not exempt from this SPP.

In maintaining configuration control of the transmission system for maintenance activities, temporary alterations of transmission system components, such as circuit breakers, air compressors, transformers, relays, indicating devices, and communication equipment, such as microwave systems, power line carrier, and supervisory control and data acquisition (SCADA), may be necessary. Before a transmission component is placed under a TPS Temporary Alteration Order, the Power System Dispatcher must first make a predetermination of the impact on the transmission system and offsite power requirements delineated in TPS Operating Guides/Instructions for each nuclear site. TPS-ESO-Operating Letter, Temporary Alteration Order process, has been evaluated by a 10CFR50.59 review. All TPS Temporary Alteration Orders are coordinated with Plant Operations prior to being placed.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 6 of 21 3.0 INSTRUCTIONS NOTE The TACF process is delineated in Appendix A, TACF Flowchart.

3.1 Initiation and Evaluation A. Any individual who has the need to install a TA requests initiation of a TACF from the appropriate System Engineer or Design Engineer.

B. The System Engineer or Design Engineer shall perform the following:

1. Perform the "Initiation" and "Evaluation" sections of the TACF by completing the following. Use the Temporary Alteration Control Form, Form Number SPP-9.5-1.

NOTE The site TACF log book is maintained by the TACF Coordinator.

a. Obtain a TACF number from the site TACF Log Book. The numbering system shall be as follows:

Unit number - year - numerical sequence - system no. (e.g., 2 96-001-065).

b. Equipment altered.

c. Quality-related.

d. Mode restraint and mode number, if required.

e. Affected systems.

f. Description, need, location, and materials. For material requirements, see Appendix C.

g. Reason/justification.

h. Drawings (dwg) affected.

1. Prepare markup of affected Category 1 drawings and/or sketches reflecting the TA for drawing updates, if required, and attach with original TACF.

2. Prepare markup of secondary drawings and/or sketches needed to provide any fabrication/installation requirements, if required, and attach with original TACF.

3. List secondary drawings for Curator update, if required.

Drawings to which this applies are those that would be revised via a DCN for the same permanent change.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 7 of 21 Coordinates with plant organizations any procedure, instruction or document [including operating procedures, Preventative Maintenance Instructions (PM), Instrument Data Packages (IDP), Scaling Setpoint Documents (SSD), Master Equipment List (MEL)] revisions, if required.

j. Estimated RTN date and milestone.

k. System Engineer or Design Engineer, phone, section of System Engineer or Design Engineer, and date.

2. Obtain an environmental evaluation by the Environmental Section, if required. Sign and record date.

3. Complete effects, limitations, actions, system status required as a result of the TA. This should include an evaluation of effect on the Maintenance Rule, if applicable.

4. Identify post implementation and post RTN testing, if required.

5. Obtain a reactivity management evaluation by Reactor Engineering. If Reactor Engineering is not available, the reactivity management evaluation may be performed by the on-shift Senior Reactor Operator (SRO). Sign and record date.

6. Perform/obtain a technical evaluation per Section 3.1.4 and Appendix C of SPP-9.3, a IOCFR50.59 review per SPP-9.4, and reference necessary administrative controls in effects, limitations, actions, and system status. Obtain Design Engineering review or approval of the technical evaluation and 10CFR50.59 review, if not prepared by Design Engineering. Sign and record date.

a. TAs affecting the protected and vital area barriers or alter security or safeguards measures can not be initiated without coordinating with Security to determine proper compensatory measures.

b. TAs that could alter the calculated flood level from pipe breaks must be evaluated by Design Engineering (DE). Examples include the addition, removal, or modification of curbs; door seals or weather-stripping; blocking of vents in doors; pipe rerouting or mechanical jumpers in operable systems; wall or floor bores; covering or blocking of floor drains; or changes in pipe wall thickness.

c. To prevent contamination of breathing air, the breathing air system or service air system must not be cross connected (mechanical jumper) to any system in a way which could possible cause radioactive or chemical contamination to the breathing air supply.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 8 of 21

d. TAs that could result in material falling on or impacting safety related systems, structures, and components in a seismic event must be evaluated by DE.

7. Obtain DE concurrence with the TACF. Sign and record date.

8. Design Engineering to evaluate the TACF and consider if a FSAR update is required based on the temporary modification expected to be in place throughout the next required periodic FSAR update cycle or when no schedule for TACF return to normal has been established.

Sign and record date.

9. Record tag number, location, and description on TACF Tag List for tags to be installed, if required. Refer to Appendix B for correct usage.

C. The System Engineer evaluates and concurs with TACF initiation. Sign and record date.

3.2 Recommendations and Approval A. Normal TA Approval.

1. System Engineer or Design Engineer obtains review and concurrence from System Engineering Manager/Designee. Sign and record date.

2. System Engineer or Design Engineer obtains a Shift Manager (SM) or Unit Manager [licensed Senior Reactor Operator (SRO)], Plant Operating Review Committee (PORC) review (quality-related TAs only),

and Plant Manager or designee approval prior to implementation. Sign and record date for each.

3. System Engineer or Design Engineer sends a copy of the TACF including marked up drawings to the applicable data entry group coordinator for data entry and status update. Secondary drawings are entered into Curator but are not updated. Sign and record date.

4. System Engineer or Design Engineer forwards a copy of the TACF to the TACF Coordinator. Sign and record date.

5. System Engineer or Design Engineer forwards a copy of the TACF to the Training Section to ensure training impacts are addressed. Sign and record date.

6. System Engineer or Design Engineer transmits 10CFR50.59 to record storage per SPP-9.4. Sign and record date.

B. Canceling TACFs NOTE TACFs can only be canceled prior to installation.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 9 of 21

1. If a TACF is no longer needed, prior to installation, the requester voids the TACF number in the site TACF Log Book.

2. The requester provides the original TACF to the TACF Coordinator for canceling.

3.3 Installation NOTE Refer to Appendix C for material requirements.

A. Precautions and Limitations

1. Ensure lifted leads are suitably insulated and jumpers securely attached.

2. Verify the affected component is properly identified and in the correct circuit.

3. De-energize electrical circuits prior to the installation of jumpers or the removal of leads, whenever possible.

4. Ensure Category 1 drawings are updated and in the Control Room.

5. Ensure secondary drawings are listed and referenced in Curator.

6. Ensure materials (cables, jumpers, etc.) are equal to or better than the quality engineered into the host component or system.

7. If the original TACF is removed, a copy of the TACF should replace the original TACF while it is removed.

B. Concerning installation, the SM or designee is responsible for the following:

1. Review the TACF to ensure required reviews and documentation have been completed and that the TA can be performed without adverse effect on plant operation.

2. Authorize installation of the TA and maintain control of the approved TACF. Sign and record date.

C. The System Engineer shall perform the following:

1. Ensure the WID number is recorded.

2. Verify all post installation testing has been successfully completed, if required. Sign and record date.

3. Verify TA installation and TACF tags, if required, have been properly placed. Sign and record date.

4. Forward a copy of the TACF to the TACF Coordinator. Sign and record date.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 10 of 21

5. In urgent situations, when Category 1 drawing updates cannot support operability, the applicable drawings may be manually marked up (red lined) with SM concurrence. Requests for mark up must be kept to a minimum. Obtain SM concurrence to Red Line drawings. All drawing markups shall be performed in accordance with NEDP-3, "Drawing Control." Sign and record date.

6. Forward a copy of the TACF to Engineering Drawing Services (EDS) for Category 1 drawing updates, if required. Sign and record date.

7. Verify TA is marked on affected Category 1 drawings, if required. Sign and record date.

8. Verify procedure/instruction/document (including operating procedures, PMs, SSDs, IDPs, MEL) revisions required to support the TA have been made, if required. Sign and record date.

9. Obtain SM or designee signature designating notification of installation completion. Sign and record date.

3.4 Return to Normal A. Concerning RTN, the SM or designee is responsible for the following:

1. Reviewing the request for RTN of the TA and determining that RTN can be performed without adverse affect on plant operation.

2. Authorize RTN of the TA. Sign and record date.

B. The System Engineer shall perform the following:

1. Request RTN of the TA in accordance with an approved WID. Sign and record date.

2. Ensure the WID number is recorded.

3. Ensure all post RTN testing has been successfully completed, if required. Sign and record date.

4. Verify that equipment has been RTN and all TACF tags, if required, are removed. Sign and record date.

5. Forward a copy of the TACF to EDS for Category I drawings update, if required. Sign and record date.

6. Verify TA removed from affected Category 1 drawings, if required. Sign and record date.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 11 of 21

7. Forward a copy of the TACF to the applicable data entry group coordinator for Category 1 and secondary drawings status update. Sign and record date.

8. Verify procedure/instruction/document (including operating procedures, PMs, SSDs, IDPs, MEL) revisions required to support RTN of the TA have been made, if required. Sign and record date.

9. Obtain SM or designee signature designating notification of RTN completion. Sign and record date.

10. Forward original TACF to TACF Coordinator. Sign and record date.

11. Transmit a copy of the TACF to the Training Section to ensure training impacts are addressed. Sign and record date.

C. TACF RTN by Transferring TAs to Other Tracking Documents NOTE The intent of transferring TAs to other tracking documents is to relax the requirements for maintaining TACFs on systems/components that have appropriate documentation (WOs, DCNs, etc.) that can control the TA as long as the system/component is out-of service or placed in lay-up.

1. The System Engineer ensures the appropriate tracking document (WO, DCN, etc.) exists with instructions to RTN the TA.

2. The System Engineer enters "Transferred to (appropriate document)"

under the WID number in the RTN Documentation section.

3. The System Engineer removes TAOs associated with the TACF and replaces these with "Information Only" tags.

4. The SM or Designee removes MCPTs associated with the TACF.

5. The System Engineer processes the TACF in accordance with Sections 3.4.A, 3.4.B and 3.4.C and marks the appropriate signature lines "N/A".

3.5 Revisions A. Minor editorial changes

1. Made by the Reviser provided the intent and scope of the TA will not change, the change is documentation only, and the 10CFR50.59 review is not impacted.

2. Made by the reviser indicating the changes on the original TACF and initialing and dating the changes.

3. Concurred with by Systems Engineering Manager/Designee by initialing and dating the changes.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 12 of 21 B. Prior to installation

1. Reviser revises and obtains review and approval in accordance with original initiation, evaluation, and approval instructions.

C. Changes affecting hardware or impacted documentation

1. Reviser completes a new TACF for the revised TA and uses the next higher revision level (example: 1-96-002-067 Ri). Indicate in the "Reason/Justification" that "This TACF supersedes TACF" and list the original TACF number.

2. Reviser verifies that the TACF revision number appears on the Category 1 drawings and the drawings are updated, if required.

3. Reviser re-performs or revises the technical evaluation and the 10CFR50.59 review, if needed.

4. Reviser processes the revised TACF in accordance with the initiation, evaluation, approval, and installation sections of this SPP.

5. Reviser closes the original TACF by performing the following:

a. State on the Return to Normal WID, "This TACF has been superseded by TACF" and list the new TACF number.

b. Enter N/A on the remaining RTN signature lines.

c. Verify that procedures/instructions/documents (including operating procedures, PMs, SSDs, IDPs, MEL) have been changed to include the affects of the revision including the new revision level, if required.

3.6 Reviews A. Semi-Annual reviews as a minimum.

Document the review on the TACF Review Form, Form Number SPP-9.5-2.

NOTE TACFs on unit(s) in layup, when affected equipment is not needed to support the operating unit(s), may be reviewed annually.

1. Performed by the System Engineer to determine:

a. TACF is free of administrative errors.

b. TA is in good material condition.

C. TA needs to remain in effect.

"WVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 13 of 21

d. If TA is configured as described by the TACF.

e. If TA should be revised or converted to a plant modification.

f. If an extension of the RTN date is required and justification for the extension.

g. If the environmental evaluation is still accurate, if required.

h. TACF tags, if required, and barriers are in good condition.

i. TAO/MCPT are correctly installed and legible, if required.

j. Category 1 drawings are correctly marked, if required.

k. Procedures/instructions/documents are correctly revised, if required.

2. Performed by DE if 10CFR50.59 review is needed to confirm validity:

a. Review 10CFR50.59 to determine validity.

B. Other reviews Each site must ensure the following reviews are addressed.

1. Changes to the status of TAs will be reviewed by Operations during shift turnover.

2. TACFs will be reviewed by Operations prior to return from a scheduled outage to verify documentation and physical integrity.

3.7 Tracking A. The TACF Coordinator is responsible for the following:

1. Ensuring that TACFs are tracked to minimize duration and to ensure reviews are performed as required.

2. Site TACF Log Book.

3. Maintaining a status report which is periodically reviewed by Site Management personnel.

4.0 RECORDS 4.1 Quality Assurance (QA) Records Completed quality-related TACF including the form, Category 1 drawings and procedures/instructions/documents, TACF Review Form and 10 CFR 50.59 review.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 14 of 21 4.2 Non-QA Records Completed non quality-related TACFs.

5.0 DEFINITIONS Applicable Data Entry Group Coordinator - The individual who is responsible for updating Curator.

Main Control Point Tag (TWA 6265A) - A tag installed at the main point of control for a component with a TA to supply needed information to the operator or identify affect on operational characteristics.

Mechanical Jumper - Any device which has not been specifically designed for a particular use, such as a hose, pipe or spool piece, which is installed in a system on a temporary basis.

Mechanical jumpers are installed for various purposes including testing, maintenance, repair, modification, habitability, and hygiene.

TACF Coordinator - An individual assigned by the System Engineering Manager to track the activities associated with the installation and return to normal of Temporary Alterations as documented by TACFs.

Temporary Alteration - A Temporary Alteration is a temporary modification to systems and/or components which may include temporary bypass lines; temporary electrical jumpers, inhibits, and wire lifts; temporary connections to electrical, pneumatic, or hydraulic systems, instrumentation, or control devices; temporary instrument settings, pulled circuit cards, mechanical jumpers, installed/removed blank flanges, disabled relief or safety valves; or other plant configuration changes that do not conform with approved drawings or other design output documents.

Temporary Alteration Control Form - A Temporary Alteration Control Form is a form used to initiate, document review, and authorize the implementation and RTN of a TA. It is maintained by the SM after approval to document the status of the TA and is maintained as a record after the TA is returned to normal.

Temporary Alteration Order (TVA 6265) - Temporary Alteration Order is the tag which is used in the plant to identify the TA and is used only in conjunction with a TACF.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 15 of 21 APPENDIX A Page 1 of I TACF FLOWCHART

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 16 of 21 APPENDIX B Page 1 of 1 REQUIREMENTS FOR HANGING TEMPORARY ALTERATION TAGS A. TAO (TVA 6265) - Determine the number of tags required and their locations, and record this information on the TACF Tag List, Form number SPP-9.5-1. Place one TAO tag on the component affected by the TA. Place additional TAO tags on the components in other systems affected by the TA. Tags are not required inside the primary containment for ALARA and FME considerations. Do not place paper TAO tags in a C-zone. As an alternative, place the TAO tag at the C-zone entrance, and so indicate on the TACF and in the TACF Tag List.

B. Record requested information on TAO (TVA 6265), that is, Temporary Alteration Permit Number; Equipment and Description; Effect, Limitations, and/or Action; and Date Installed.

C. Jumper - Place one TAO tag on each jumper. Place a TAO tag on each end of jumpers where both ends are not visible.

D. Inhibit - Group all inhibits on a device under one TAO tag. For example, ifthree boots were used on a relay, use one TAO tag. Record the number of inhibits covered by the TAO tag on the TACF Tag List.

E. Wire Lifts - Use one TAO tag per terminal point, considering each side of a terminal block a point or identify, group, and tag all wires removed as a bundle. If all wires can not be bundled in one group, use additional TAO tags. Tape or cover each lifted wire with an insulator to prevent a personnel and equipment hazard.

F. Jumper/Inhibit(s) Removal - Use the "Return to Normal" portion of the TACF.

G. MCPT (TVA 6265A) - When the MCPT is used, place one tag on the main point of control of the system or component affected. Place duplicate tags as required to properly identify the temporary condition. Use the MCPT when the main point of control is the Main Control Room panel. Indicate all MCPTs on the TACF Tag List.

"WVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Rev. 4 PROCESSES Page 17 of 21 APPENDIX C Page I of I MATERIAL REQUIREMENTS FOR TEMPORARY ALTERATIONS A. TA material or equipment shall be of equal to or better than the quality level engineered in the host component or system. If material quality less than the host component or system is used, justification for use shall be attached to the TACF. The material shall be compatible for this intended use (e.g., size, terminal, type, insulation, pressure rating, material, piping construction, etc.).

B. Install jumpers in a conspicuous manner to be easily distinguished from permanent wiring.

C. Fabricate inhibits from bright colored heat-shrinkable polyvinyl chloride tubing shrunk on one end for holding purposes.

D. Mark printed circuit cards and other miniature devices too small to identify with tags, jumpers, or inhibits. Identify jumpers and inhibits conspicuously.

E. Mark temporary cables with yellow and red striped tape at intervals not to exceed five feet, and intervals not to exceed two inches in an enclosure.

F. Do not install temporary cables in permanent conduits or cable trays unless no other route is available.

G. If a temporary cable for quality-related equipment becomes a permanent plant feature through appropriate means (DCN, etc.), remove the temporary cable and install permanent cable.

H. Ensure temporary cables are sufficiently rated such that the temperature rating of the insulation will not be exceeded under full load conditions. Consider the route, separation criteria, adjacent equipment, and ambient conditions.

1. Route temporary cable such that interference with existing equipment and other temporary cables is minimized.

J. Limit the use of alligator clips to the following:

1. An urgent situation.

2. Use of ring tongue or spade type lug is not possible.

3. Installation of ring tongue or spade type lug presents an undue hazard.

K. Do not use jumpers with alligator clips to complete secondaries of current transformer (CT) circuits. Such circuits are normally found in protective relaying and metering circuits.

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Date 06-29-2001 PROCESSES Page 18 of 21 TEMPORARY ALTERATION CONTROL FORM TACF Number Page of INITIATION Equipment Altered Quality-Related Yes EW No El Mode Restraint Yes 0 No E] Mode Number*

Affected Systems Description, Need, Location, Materials Reason/Justification Drawings Affected* Drawings/Sketches Attached Procedures/Instructions/Documents Affected (including operating procedures, PMs, SSDs, IDPs, MEL)*

Estimated RTN Date and Milestone Initiated By Phone Section Date EVALUATION Environmental Evaluation by Environmental Section* Date Effects, Limitations, Actions, System Status Post Implementation Test(s)*

Post RTN Test(s)*

Reactivity Management Evaluation Date 10CFR50.59 Review and Technical Evaluation Complete Date Design Engineering Concurrence Date System Engineering Concurrence Date RECOMMENDATION AND APPROVAL System Engineering Manager/Designee Date SM or Unit Manager (Licensed SRO) Date PORC Chairperson (Quality-Related only)* Date Plant Manager/Designee Approval Date DISTRIBUTION Copy of TACF sent to Applicable Data Entry Group Coordinator Date Copy of TACF sent to TACF Coordinator Date Copy of TACF sent to Training Section Date 10CFR50.59 Review Transmitted to Record Storage Date

N/A if not required TVA 40573 [06-2001] Page 1 of 3 SPP-9.5-1 [06-29-2001]

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Date 06-29-2001 PROCESSES Page 19 of 21 TEMPORARY ALTERATION CONTROL FORM TACF TAG LIST TACF Number Page of TAG NUMBER LOCATION DESCRIPTION (Wire number, Terminal Board Number, Valve Number)

TVA 40573 [06-20011 Page 2 of 3 SPP-9.5-1 [06-29-2001]

TVANSTANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Date 06-29-2001 PROCESSES Page 20 of 21 TEMPORARY ALTERATION CONTROL FORM TACF Number Page of INSTALLATION DOCUMENTATION TACF Authorized (SM)# Date WID Post Installation Testing Complete* Date Installation Verified, TAOs & MCPTs Installed Date Copy of TACF sent to TACF Coordinator Date SM Concurrence to Red Line Drawings#* Date AFFECTED DRAWINGS AND PROCEDURES/INSTRUCTIONS/DOCUMENTS EDS Notified to Mark Drawings* Date TA Marked on Drawings by EDS* Date Procedures/Instructions/Documents (including operating procedures, PMs, SSDs, LDPs, MEL) Updated* Date TA Installation Completion Notification (SM)# Date RETURN TO NORMAL DOCUMENTATION Requested By Date WID TA RTN Authorized (SM)# Date Post RTN Testing Complete* Date TA RTN Verified, TAOs and MCPTs Removed Date REINSTATING DRAWINGS AND PROCEDURESIINSTRUCTIONS/DOCUMENTS EDS Notified to Mark Drawings* Date TA Removed from Drawings by EDS* Date Copy of TACF Sent to Applicable Data Entry Group Coordinator Date Procedures/Instructions/Documents (including operating procedures, PMs, SSDs, IDPs, MEL) Updated* Date TA RTN Completed Notification (SM)# Date DISTRIBUTION Original TACF Sent to TACF Coordinator Date _I Copy of TACF sent to Training Section Date

N/A if not required

SM or Designee TVA 40573 [06-2001] Page 3 of 3 SPP-9.5-1 [06-29-2001]

TVAN STANDARD SPP-9.5 PROGRAMS AND TEMPORARY ALTERATIONS Date 06-29-2001 PROCESSES Page 21 of 21 TACF REVIEW FORM

1. TACF Number

2. System Engineer Telephone Number

3. Section

4. TACF Initiation Date TACF Installation Date

5. Latest RTN Milestone/Date

6. TACP Status:

a. TACF is free of administrative errors YES El NO El

b. TA is in good material condition YESE NO El C. TA needs to remain in effect YES -' NO El

d. TA needs to be revised YES El NO

e. TA needs to be converted to a plant modification YES El NO El

f. Environmental evaluation is still accurate YES El NO ELE] N/A []

g. TA configuration is correct YES El NO El

h. TAO/MCPT correctly installed YES El NO El N/AEl

i. TAO/MCPT are legible YES El NO El N/AE]

j. Category I drawings are correctly marked YES El NO El N/AEl

k. Procedures/instructions/documents are correctly revised YES El NO El N/AEl I. Extension of TACF RTN date required YES El NO If YES, provide justification for extension in comment/corrective actions.

7. Comments/Corrective Actions

8. System Engineer Date

9. DE notified to review Technical Evaluation & YES D NOI-], If YES, THEN 10CFR50.59 review Technical Evaluation & 10CFR50.59 review has been reviewed and is valid Date Design Engineer

10. Current review attached to TACF Date System Engineer

11. Copy of review sent to System Engineering Manager/Designee Date System Engineer

12. Copy of review sent to TACF Coordinator Date System Engineer TVA 40574 [06-2001] Page 1 of 1 SPP-9.5-2 [06-29-20011

QUESTIONS REPORT for SRO 2002-301 LXRBank

92. G2.3.1 001 You are called at home and directed to go to the Hatch Facility to assist in the recovery efforts following a refueling accident. You are informed that you will require a TILD during the assist visit. -..

Which ONE of the following describes the dosimetry requirement for this emergency visit per SPP-5. 1, RADIOLOGICAL CONTROLS?

A. You must obtain your BFN dosimetry and wear it along with the dosimetry provided by Hatch. Following your return you must report to RADCON to obtain any required bioassay and update your exposure records.

B. You must inform RADCON of your intended visit and obtain your BFN dosimetry to wear with the dosimetry provided by Hatch. Upon your return you must present copies of your dose records from Hatch.

C. You must turn in your dosimetry and check out prior to leavi ng BFN, unless exempted by the Shift Manager or Operations Manager.

D. You must turn in your dosimetry and check out prior to leavi ng BFN, unless exempted by RADCON.

References:

SPP-5.1, Radiological Controls Rev.3 pg 9 A. Incorrect since a bioassay is not required and you must also turn in your BFN dosimetry.

B. Incorrect since you must turn in your BFN dosimetry.

C. Incorrect since you can only get exemption from RADCON.

D. Correct answer.

RO Tier: T3 SRO Tier: T3 Keyword: RADIATION CONTROL Cog Level: MEM 2.6/3.0 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:57 AM 92

QUESTIONS REPORT for SRO 2002-301 LXRBank

93. G2.3.4 001 Given the following exposure history data for an individual:

Lifetime Exposure: 19500 mrem TEDE (NRC form 4 on file)

Annual Exposure: 4600 mrem TEDE Quarterly Exposure: 600 mrem TEDE Age: 22 Sex: Male Which ONE of the following is the maximum additional whole body exposure the individual is allowed to receive in the current calender quarter under Federal (10 CFR

20) exposure limits? (Not a planned special exposure)

Af 400 mrem TEDE.

B. 600 mrem TEDE.

C. 800 mrem TEDE.

D. 1400 mrem TEDE.

References:

10 CFR 20 Subpart C 20.1201 A. Correct answer.

B, C and D. Incorrect since worker is limited to 5000 mRem annually.

Note: Changed annual dose in stem to 4600 instead of 4300. This changed the correct answer.

RO Tier: SRO Tier: T3 Keyword: RADIATION CONTROL Cog Level: C/A 2.5/3.1 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 93

QUESTIONS REPORT for SRO 2002-301 LXRBank

94. G2.3.9 001 Unit 2 startup is in progress with the following conditions existing:

Reactor Power 10% RTP Reactor Pressure 920 psig Mode Switch Position START/HOT STBY Containment is-being inerted.

Purge filter fan is in service.

Which ONE of the following describes the results of placing the Reactor Mode Switch to the RUN postion?

A. Initiates a Group 6 PCIS isolation unless Bypass switches are placed in BYPASS on panel 9-3.

B. Automatically closes all valves required for inerting with the purge filter fan unless Bypass switches are placed in BYPASS on panel 9-3.

C. Automatically closes all valves required for inerting with the purge filter fan unless the Drywell/Torus Bypass switch on panel 9-3 is taken to Torus position and any SGT fan is running.

D. Automatically closes the drywell and suppression chamber exhaust isolation valves unless the Drywell/Suppression Chamber Train A/B Vent keylock switches are positioned to DRYWELL.

References:

OPL171.032 Rev.10 pg 14 and 15 A. Incorrect since placing the Mode Switch in RUN will not initiate a Group 6 isolation.

B. Correct answer.

C. Incorrect since the Drywell/Torus bypass switch on the 9-3 panel is the incorrect switch.

D. Incorrect since placing the Drywell/Suppression Chamber Train A/B Vent keylock switches to the DRYWELL position does not prevent valve movement.

RO Tier: SRO Tier: T3 Keyword: CONTAINMENT Cog Level: C/A 2.5/3.4 Source: B Exam: BF02301 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 94

QUESTIONS REPORT for SRO 2002-301 LXRBank

95. G2.4.1 001 Which ONE of the following is NOT an immediate action of 2-AOl-1-1, Relief Valve Stuck Open?

Af Inhibit MSRV Auto Actuation Logic on Panel 2-9-3.

B. PLACE affected relief valve control switch from CLOSE to OPEN to CLOSE several times, and OBSERVE indications to see whether valve closes.

C. IDENTIFY stuck open relief valve by OBSERVING SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3.

D. IDENTIFY stuck open relief valve by OBSERV ING MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-1-1 on Panel 2-9-47.

4.1 Immediate actions:

4.1.1 IDENTIFY stuck open relief valve by OBSERVING following:

4.1.1.1 SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 293, or 4.1.1.2 MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2TRI 1 on Panel 2947.

4.1.2 PLACE affected relief valve control switch from CLOSE to OPEN to CLOSE several times, and OBSERVE indications to see whether valve closes.

4.2 Subsequent Action 4.2.1 IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s).

RO Tier: T3 SRO Tier: T3 Keyword: RELIEF VALVE Cog Level: C/A 4.3/4.6 Source: B Exam: BF02301 Test: C Misc:

Tuesday, January 21, 2003 07:30:58 AM 95

QUESTIONS REPORT for SRO 2002-301LXRBank

96. G2.4.10 001 Unit 2 is operating at 70% power, alarm window "Turbine Trip Timer Initiated" on the 2-9-8 Panel is illuminated.

Which ONE of the following describes the required operator action(s)?

A. Immediately scram the reactor.

B. Depress the Core Flow runback PB on 2-9-5, verify recirc runback to 50-60% core flow.

C. Start the standby Stator Cooling Water pump, ifalarm does not reset, scram the reactor.

D. Immediately trip the turbine.

References:

2-ARP-9-8A Rev.20 A. Incorrect since the first action is to verify all availble SCW in service.

B. Incorrect since power cannot be reduced to below 30 percent in time.

C. Correct answer.

D. Incorrect since the turbine trip is imminent without starting the standby SCW pump.

RO Tier: T3 SRO Tier: T3 Keyword: RPS Cog Level: MEM 3.0/3.1 Source: N Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 96

QUESTIONS REPORT for SRO 2002-301 LXRBank

97. G2.4.11 001 A diesel fuel oil spill has occurred in the Diesel Generator Building. The On Scene Coordinator requests that absorbent material be delivered to the scene.

Which ONE of the following contains a list of the locations of stored absorbent material?

A. MSDS for diesel fuel oil.

B. Emergency Plan Implementing Procedure.

C. Spill Prevention Control and Countermeasures Plan.

D. Browns Ferry Master Materials Index Data Base (MMIDB).

Reference:

OPI1 71.068 Rev. 4 pg 10 Enabling Objective #5 C is the only answer that provides a table for location of spill cleanup material therefore answers A,B and C are incorrect.

Note: Reordered answers.

RO Tier: T3 SRO Tier: T3 Keyword: ADMIN Cog Level: MEM 3.4/3.6 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 97

QUESTIONS REPORT for SRO 2002-301 LXRBank

98. G2.4.29 001 It is noted by the Unit 2 Supervisor that a Site Area Emergency classification level was exceeded but the present situation indicates that only an Alert classification level is being exceeded ...........

Which ONE (1) of the following is correct concerning the Radiological Emergency Plan?

A. the higher classification should be reported to CECC, if staffed, or ODS, if the CECC is not staffed, and the lower classification reported to NRC.

B! the higher classification shall be reported to NRC and the CECC if staffed, or the ODS, ifthe CECC is not staffed, but the higher classification should not be declared.

C. only the lower classification should be reported to NRC and the CECC if staffed, or the ODS, if the CECC is not staffed.

D. the higher classification shall be declared.

References:

EPIP-1 Rev.29 Pg 3 A. Incorrect since the higher classification should be reported to the NRC.

B. Correct answer.

C. Incorrect since the higher classification should be reported to the NRC.

D. Incorrect since the higher classification should not be declared but should be reported to everyone.

RO Tier: SRO Tier: T3 Keyword: EMERGENCY PLAN Cog Level: MEM 2.6/4.0 Source: B Exam: BF02301 Test: S Misc:

Tuesday, January 21, 2003 07:30:58 AM 98

EMERGENCY EPIP-1 CLASSIFICATION SECTION I CLASSIFICATION PROCEDURE INTRODUCTION INDEX EVENT CLASSIFICATION INDEX SECTION 1.0 REACTOR 1.1 WATER LEVEL 1.2 SCRAM FAILURE 1.3 REACTOR COOLANT ACTIVITY 1.4 MSL/OFFGAS RADIATION 1.5 LOSS OF DECAY HEAT REMOVAL SECTION 2.0 PRIMARY 2.1 PRIMARY CONTAINMENT PRESSURE CONTAINMENT 2.2 PRIMARY CONTAINMENT HYDROGEN 2.3 DRYWELL RADIATION 2.4 DRYWELL INTERNAL LEAKAGE 2.5 LOSS OF PRIMARY CONTAINMENT SECTION 3.0 SECONDARY 3.1 SECONDARY CONTAINMENT CONTAINMENT TEMPERATURE 3.2 SECONDARY CONTAINMENT RADIATION SECTION 4.0 RADIOACTIVITY 4.1 GASEOUS EFFLUENT RELEASES 4.2 MAIN STEAM LINE BREAK 4.3 LIQUID EFFLUENT SECTION 5.0 LOSS OF POWER 5.1 LOSS OF AC POWER 5.2 LOSS OF DC POWER SECTION 6.0 HAZARDS 6.1 RADIOLOGICAL 6.2 CONTROL ROOM EVACUATION 6.3 TURBINE FAILURE 6.4 FIRE/EXPLOSION 6.5 TOXIC GASES 6.6 FLAMMABLE GASES 6.7 SECURITY 6.8 VEHICLE CRASH SECTION 7.0 NATURAL EVENTS 7.1 EARTHQUAKE 7.2 TORNADO/HIGH WINDS 7.3 FLOOD SECTION 8.0 EMERGENCY 8.1 TECHNICAL SPECIFICATIONS DIRECTOR 8.2 LOSS OF COMMUNICATION JUDGEMENT 8.3 LOSS OF ASSESSMENT CAPABILITY 8.4 OTHER CLASSIFICATION PAGE 11 OF 207 REVISION 29 1 INDEX

EMERGENCY EPIP-1 CLASSIFICATION SECTION I CLASSIFICATION PROCEDURE INTRODUCTION INDEX I

THIS PAGE INTENTIONALLY BLANK REVISION 29 PAGE 12 OF 207

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR I

REACTOR 1.0 1.0OREACTOR PAGE 13 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

1.1-U1/1.1-A1 Applicable when the Reactor Head is removed and the Reactor Cavity is flooded.

1.1-si Applicable in Mode 5 when the Reactor Head is installed.

1.1-G2 The reactor will remain subcritical under all conditions without boron when:

All control rods are inserted to or beyond position 02

All control rods except one are inserted to or beyond position 00

Determined by reactor engineering CURVES/TABLES:

NUMBER OF OPEN MSRVs MAREP (PSIG) 6 or More1819 5.......0 4 290 ...

REVISION 30 PAGE 14 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR A LEVE DESCRIPTION DESCRIPTION 1.1-U1 1.l-U2 Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel Pool c with irradiated fuel assemblies expected to remain with irradiated fuel assemblies expected to remain covered by water. covered by water.

OPERATING CONDITION: OPERATING CONDITION:

- Mode 5 - All 1.1-Al K 1.1-A2 Uncontrolled water level decrease in Reactor Cavity Uncontrolled water level decrease in Spent Fuel expected to result in irradiated Fuel assemblies being Storage Pool expected to result in irradiated fuel uncovered, assemblies being uncovered.

OPERATING CONDITION: OPERATING CONDITION:

- Mode 5 - All 1.-S 1.1-2 Reactor water level CANNOT be maintained above Reactor water level CANNOT be determined.

-162 IN. (TAF)

OPERATING CONDITION: OPERATING CONDITION:

- All - Mode I - Mode 3

-Mode2 1.l-G1 1.1-G2 N I Reactor water level CANNOT be determninedM Reactor water level CANNOT be restored and AND e maintained above: EITHER ofthe following conditions exists:

UNIT UNIT 2 -190 IN.

Theractorwill mn ticalw/obomdroa cnditions

3 -185 IN. and

L=tmha4MSRVs nbeopaR adrpractsv CANNOT be w

resoed and nmatained at least 65 PSI above on Chamler

It has NOT been determined that the reactor will remain subcritical w/o boron under all conditions and unable to restoreandnainaaMARFP in Table 1.1-02.

OPERATING CONDITION:

- Mode I - Mode 3 OPERATING CONDITION:

-Mode2 -Model -Mode3

-Mode2 1.0 REACTOR PAGE 15 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

1.2 Subcritical is defined as Reactor power below the heating range and not trending upward.

CURVES/TABLES:

CURVE 1.2-G HEAT CAPACITY TEMP LIMIT 26 0 , .......

250,-li- - SAFE WHEN RX PRESS 0 BELOW 65 PSIG 2 *IS ý-o XY:,

240 R V r s 65 NS .23. . 5

0. .

a- 220,I Lu 210-rss 300 S200 RPV Press. 500 S190 09MRPV Press. 700 Cl 180 RPv rPress. 900 170, RPV Press. 11 _

11.5 12 12.5 13 13.5 14 14.5 15 15.5 16 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT)

ACTION REQUIRED IF ABOVE CURVE FOR EXISTING RX REVISION 30 PAGE 16 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION HI PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR A FIU 6 6 DESCRIPTION DESCRIPTION Reactor coolant activity exceeds 26 gCi/gm dose equivalent 1-131 (Technical Specification Limit) as determined by chemistry sample.

OPERATING CONDITION:

- ALL 1.2-A K 13 Failure of automatic scram functions to bring the Reactor coolant activity exceeds 300 gCi/gm dose Reactor subcritical equivalent Iodine-131 as determined by chemistry AND sample. t Manual scram or ARI was successful.

OPERATING CONDITION: OPERATING CONDITION:

- Mode l -Model -Mode 3

- Mode 2 - Mode 2 Failure of automatic scram, manual scram, and ARI to bring the Reactor subcritical.

OPERATING CONDITION:

- Mode l 1.2-G Failure of automatic scram, manual scram, and ARI.

Reactor power >3%

AND EITHER of the following conditions exists:

Suppression Pool temp exceeds HCTL.

Refer to Curve 1.2-G.

Reactor water level CANNOT be restored and maintained at or above:

UNIT2 -190 IN.

UNIT 3 -185 IN.

OPERATING CONDITION:

- Mode 1

- Mode 2 1.0 REACTOR PAGE 17 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION H CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE CURVES/TABLES:

CURVE 1.5-S HEAT CAPACITY TEMP LIMIT 260 250 SAFE WHEN RX PRESS IS BELOW 65 PSIG 240 RPV Press. 65 230 220 210 RPV Press. 300 200 RPV Press. 500 190 RPV Press. 700 180 RPV Press. 900 170 RPV Press. 1135 SAFE 160 150 11.5 12 12.5 13 13.5 14 14.5 15 15.5 16 16.5 17 17.5 18 18.5 19 SUPPR PL LVL (FT)

ACTION REQUIRED IFABOVE CURVE FOR EXISTING RX REVISION 30 PAGE 18 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 1.0 REACTOR I DESCRIPTION DESCRIPTION S1,4-U Valid MAIN STEAM LINE RADIATION HIGH-HIGH zA alarm, RA-90-135C OR Valid OG PRETREATMENT RADIATION HIGH alarm, RA-90-157A.

OPERATING CONDITION:

-Mode 1 -Mode 3

- Mode 2 1.SA Reactor moderator temperature CANNOT be maintained below 2120 F whenever Technical Specifications require Mode 4 conditions or during r operations in Mode 5.

M OPERATING CONDITION:

-Modee4

-Mode 5 1.5-S K>

Suppression Pool temperature, level and RPV pressure CANNOT be maintained in the safe area of Curve 1.5-S.

OPERATING CONDITION:

-Model -Mode 3

______________________________________ - Mode 2 to 1.0 REACTOR PAGE 19 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION H CLASSIFICATION 1.0 REACTOR EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 30 PAGE 20 OF 207 1.0 REACTOR

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT I

PRIMARY CONTAINMENT 2.0 2.0 PRIMARY PAGE 21 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

DRYWELL CAM ACTIVITY INCREASING

'IDRYWELL TEMPERATURE HIGH ALARM CHEMISTRY SAMPLE RADIONUCLIDE COMPARSION TO RX WATER REVISION 28 PAGE 22 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT I M DESCRIPTION DESCRIPTION z

ct r

2.1-A I Drywell pressure at or above 2.45 PSIG AND Indications of Primary System leakage into Primary Containment. Refer to Table 2.1 -A.

OPERATING CONDITION:

-Model -Mode 3

-Mode2 2.1-S H2.

Suppression Chamber Pressure CANNOT be Drywell or Suppression Chamber hydrogen maintained in the safe area of Curve 2.1 -S. concentration at or above 4%

AND Drywell or Suppression Chamber oxygen concentration at or above 5%.

OPERATING CONDITION: OPERATING CONDITION:

-Mode l -Mode 3 -Mode I -Mode 3

- Mode 2 -Mode2 Suppression Chamber Pressure CANNOT be Drywell or Suppression Chamber hydrogen maintained below 55 PSIG. concentration at or above 6%

AND Dxywell or Suppression Chamber oxygen concentration at or above 5%.

OPERATING CONDITION: OPERATING CONDITION:

-Mode l -Mode 3 -Mode l -Mode 3

- Mode 2 -Mode2 2.0 PRIMARY PAGE 23 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

TABLE 2.3-A/2.3-$2 DRYWELL RADIATION LEVELS WITH RCS BARRIER INTACT UNIT 2 UNIT 3 RAD MONITOR R/HR RAD MONITOR R/HR 2-RE-90-272A 345 3-RE-90-272A 106 2-RE-90-273A 164 3-RE-90-273A 164 TABLE 2.3/2.5-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT UNEXPLAINED LOSS OF PRESSURE EXCEEDING SI-4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIs/SAMGs REVISION 28 PAGE 24 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT I1YEa'AXUMREADIATIONg IDESCRIPTION iDESCRIPTION z

Ci) r H

2.3-A Drywell radiation levels at or above the values listed in Table 2.3-A/2.3-S2, with the RCS barrier intact.

OPERATING CONDITION:

Mode 1 -Mode 3 Mode 2 2.3-si 2.3-S2 CI]

Drywell radiation levels at or above the values listed Drywell radiation levels at or above 4880 R/HR wihteRS otitc.AND are in Table 2.3-A/2.3-S2, with the RCS barrier intact.

with the RCS baffler not intact. Either of the following exists: R

"*Indications of loss of Primary Containment. Refer to Table 2.3/2.5-U. 0

"*Primary Containment integrity CANNOT be maintained. M OPERATING CONDITION: OPERATING CONDITION:

-Model -Mode 3 -Mode -Mode 3

-Mode2 - Mode 2 2.3-G1 2.3-G2 K .

Drywell radiation levels at or above 19500 R/HR Drywell radiation levels at or above 4880 R/HR with M with the RCS barrier not intact. the RCS barrier not intact.

AND Either of the following exists:

Indications of loss of Primary Containment.

Refer to Table 2.3/2.5-U.

Primary Containment integrity CANNOT be maintained.

OPERATING CONDITION: OPERATING CONDITION:

-Mode I - Mode 3 -Mode l -Mode 3

-Mode2 - Mode 2 2.0 PRIMARY PAGE 25 OF 207 REVISION 28 CONTAINMENT

EPIP-1 EMERGENCY 2.0 PRIMARY SECTION II CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

TABLE 2.3/2.5-U INDICATIONS OF LOSS OF PRIMARY CONTAINMENT UNEXPLAINED LOSS OF PRESSURE EXCEEDING SI-4.7.A.2.a LIMITS INABILITY TO ISOLATE ANY LINE EXITING CONTAINMENT WHEN ISOLATION IS REQUIRED VENTING IRRESPECTIVE OF OFFSITE RELEASE RATES PER EOIs REVISION 28 PAGE 26 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPLP-1 CLASSIFICATION SECTION 11 2.0 PRIMARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT CLASSIFICATION Sr II OF PRIMARY P UEAAEVT CONTAINMENT DESCRIPTION DESCRIPTION l 2.4-U ] 2.5-U K*

z Drywell unidentified leakage exceeds 10 GPM Inability to maintain Primary Containment pressure Ct OR boundary. Refer to Table 2.3/2.5-U.

Drywell identified leakage exceeds 40 GPM.

r z

OPERATING CONDITION: OPERATING CONDITION:

-Model -Mode 3 -Model -Mode3

- Mode 2 - Mode 2 2.4-A Drywell unidentified leakage exceeds 50 GPM.

OPERATING CONDITION:

-Mode -Mode 3

- Mode 2

-10 2.0 PRIMARY PAGE 27 OF 207 REVISION 28 CONTAINMENT

EPIP-l EMERGENCY 2.0 PRIMARY SECTION H CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 28 OF 207 2.0 PRIMARY CONTAINMENT

EMERGENCY EPLP-l CLASSIFICATION SECTION H 3.0 SECONDARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT EMERGNCY PIP-SECONDARY CONTAINMENT 3.0 3.0 SECONDARY PAGE 29 OF 207 REVISION 29 CONTAINMENT

EPIP-1 EMERGENCY 3.0 SECONDARY SECTION HI CLASSIFICATION CONTAINMENT EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

APPLICABLE PANEL 9-21 0

AREA TEMPERATURE ELEMENTS MAX SAFE OPERATING VALUE F (UNLESS OTHERWISE NOTED) UNIT 2 UNIT 3 RHR A/C PUMP ROOM 74-95A 150 155 RHR B/D PUMP ROOM 74-95B 210 215 HPCI TURBINE AREA 73-55A 270 270 RCIC TURBINE AREA 71-41A 190 190 CS PUMP ROOM HIGH HUMIDITY OR TEMP (XA-55-3E-29) PANEL 9-3 TI-75-69B 140 150 HIGH RCIC STEAM SUPPLY AREA 71-41B, 41C, 41D 200 250 HPCI STEAM SUPPLY AREA 73-55B, 55C, 55D 240 240 RHR AC PUMP SUPPLY AREA 74-95H 240 240 RHR B/D PUMP SUPPLY AREA 74-95G 240 240 MAIN STEAM LINE LEAK DETECTION HIGH (XA-55-3D-24) PANEL 9-3 TIS-1-60A 315 315 RHR VALVE ROOM 74-95E 170 175 RWCU ISOL LOGIC CHANNEL A/B TEMP HIGH (XA-55-5B-32/33) PANEL 9-5 170 175 69-835A, B, C, D AUX INST ROOM RWCU OUTBD ISOL VLV AREA 69-29F 220 220 RWCU -X AREA 69-29G 220 220 RWCU HX EXH DUCT 69-29H 220 220 RWCU RECIRC PUMP A AREA 69-29D 215 215 RWCU RECIRC PUMP B AREA 69-29E 215 215 RHR A/C HX ROOM 74-95C 195 200 RHR B/D HX ROOM 74-95D 195 200 FPC HX AREA 74-95F 150 155 MAXIMUM.~.......RA

..... SAE...A.. AITINLM AREA RAD MONITOR MAX SAFE VALUE MR/HR RHR WEST ROOM 90-25A 1000 RHR EAST ROOM 90-28A 1000 HPCI ROOM 90-24A 1000 CS/RCIC ROOM 90-26A 1000 CORE SPRAY ROOM 90-27A 1000 SUPPR POOL AREA 90-29A 1000 CRD-HCU WEST AREA 90-20A 1000 CRD-HCU EAST AREA 90-21A 1000 TIP DRIVE AREA 90-23A 1000 NORTH RWCU SYSTEM AREA 90-13A 1000 SOUTH RWCU SYSTEM AREA 90-14A 1000 RWCU SYSTEM AREA 90-9A 1000 MG SET AREA 90-4A 1000 FUEL POOL AREA 90-IA 1000 SERVICE FLR AREA 90-2A 1000 NEW FUEL STORAGE 90-IA 1000 DRYWELL RADIATION UNIT 2 DRYWELL RADIATION UNIT 3 2-RE-90-272A I >345 k/HR 3-RE-90-272A I >106 R 2-RE-90-273A >164k/HR 3-RE-90-273A >164R/HR REACTOR COOLANT ACTIVITY > 300 ptCI/gm DOSE REACTOR COOLANT ACTIVITY > 300 gCI/gm DOSE EQUILAVENT IODINE-131 EQUILAVENT IODINE-131 REVISION 29 PAGE 30 OF 207 3.0 SECONDARY CONTAINMENT

EMERGENCY EPIP-l CLASSIFICATION SECTION IF 3.0SECONDARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT TEMESRIPTURE RDECIATION z"

Ct 3.2-A yo e o owing hgh radiation alarms on Panel 9-3:

RA-90-1A, Fuel Pool Floor Area RA-90-250A, Reactor, Turbine, Refuel Exhaust RA-90-142A, Reactor Zone Exhaust t*

RA-90-140A, Refueling Zone Exhaust AND Confirmation by Refuel Floor personnel that irradiated fuel damage may have occurred.

OPERATING CONDITION:

-All 3.1-S K 3.2-S K An unisolable Primary System leak is discharging into An unisolable Primary System leak is discharging into Secondary Containment Secondary Containment Any area temperature exceeds the Maximum Safe Any area radiation level at or above the Maximum Operating Temperature limit listed in Table 3.1. Safe Operating area Radiation limit listed in Table 3.2. M

z OPERATING CONDITION: OPERATING CONDITION:

-Model -Mode3 -Model -Mode3

- Mode 2 - Mode 2 3.1-G Ki3.2-G Ki An unisolable Primary System leak is discharging into An unisolable Primary System leak is discharging into Secondary Containment Secondary Containment AND AND Any area temperature exceeds the Maximum Safe Any area radiation level at or above the Maximum Operating Temperature limit listed in Table 3.1. Safe Operating area Radiation limit listed in Table 3 2 AND AND Any indication of potential or significant fuel failure Any indication of potential or significant fuel failure exists. Refer to Table 3.1-G/3.2-G. exists. Refer to Table 3.1-G/3.2-G.

OPERATING CONDITION: OPERATING CONDITION:

-Mode I - Mode 3 -Mode 1 -Mode33

- Mode 2 - Mode 2 3.0 SECONDARY PAGE 31 OF 207 REVISION 29 CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 3.0 SECONDARY PROCEDURE EVENT CLASSIFICATION MATRIX CONTAINMENT THIS PAGE INTENTIONALLY BLANK REVISION 29 PAGE 32 OF 207 3.0 SECONDARY CONTAINMENT

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 4.0 RADIACTIVITY PROCEDURE EVENT CLASSIFICATION MATRIX RELEASE RADIOACTIVITY RELEASES 4.0 4.0 RADIACTWITY PAGE 33 OF 207 REVISION 30 RELEASE

EPIP-I EMERGENCY 4.0 RADIACTWITY SECTION II CLASSIFICATION RELEASE EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

NOTE 4.1-U Prior to making this emergency classification based ulon the WRGERMS indication, assess the release by either ofthe following I. Actual field measurements exceed the limits in Table 4.1-U

2. S14.8.B.I.&I Release Fraction exceeds 2.0 Ifneithar assessrent cant be oandxcted within 60 rm, untesthet the declaatimnmustbe male omthe valid WRGERMS iadnag NOTE 4.1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1.Actual field measurements exceed the limits in Table 4.1 -A

2. SI 4.8.B.1.a.1 Release Fraction exceeds 200 lfnetherida entcanbecondxtedwithin 15 mntmutes then the declarafionmustbeniade en the valid WROERMS eairg NOTE 4.1-S Prior to making this emergency classification based upon the Gaseous Release Rate indication, assess the release by either ofthe following methods:

1. Actual fieldimeasurements exceedthe limits in Table 4.1-S.

2. Projected orActual DoseAssessments exceed 100 inern TEDE or 500 ntern CDE.

If neither assessment can be conducted within 15 minutes then the declaration must be made based onthe valid WRGERMS NOTE 4.1-G Priorto making this emergency classification based upon the Gaseous Release Rate indication, assess the release by either ofthe following methods:

1.Actual field measurements exceed the limits in Table 4.1-0.

2. Projected or Actual Dose Assessments exceed 1000 mrem TEDE or 5000 nmern CDE.

Ifneither assessment can be conducted within 15 minutes then the declaration must be made based on the valid WRGERMS reading.

CURVES/TABLES:

TYPE MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 2.88 X 10 ' ýtCi/sec 1 HOUR GASEOUS RELEASE RATE S14.8.B.l.a.1 RELEASE FRACTION 2.0 1 HOUR SITE BOUNDARY RADIATION READING FIELD ASSESSMENT TEAM 0.10 MREM/HR Gamma 1 HOUR

. ......... RELEASE LINTS WFOR AL TYPE MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 2.88 X 10 9 IiCi/sec 15 MINUTES GASEOUS RELEASE RATE SI 4.8.B.l.a.1 RELEASE FRACTION 200 15 MINUTES SITE BOUNDARY RADIATION READING FIELD ASSESSMENT TEAM 10 MREM/HR Gamma 15 MINUTES TYPE I MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 5.9 X 10 Ci/see 15 MINUTES SITE BOUNDARY RADIATION READING FIELD ASSESSMENT TEAM 100 MREM/HR Gamma 1 HOUR SITE BOUNDARY IODINE-131 FIELD ASSESSMENT TEAM 3.9 X 107 ptCI/cm 3 1 HOUR xX e................................

.............. .... l 4-............ . .. .......

TYPE MONITORING METHOD LIMIT DURATION GASEOUS RELEASE RATE STACK NOBLE GAS (WRGERMS) 5.9 X 10" tCi/sec 15 MINUTES SITE BOUNDARY RADIATION READING FIELD ASSESSMENT TEAM 1000 MREM/HR Gamma I HOUR SITE BOUNDARY IODINE-131 FIELD ASSESSMENT TEAM 3.9X 10.6 ItCI/cmr I HOUR REVISION 30 PAGE 34 OF 207 4.0 RADIACTIVITY RELEASE

EMERGENCY EPIP-1 CLASSIFICATION SECTION 11 4.0 RADIOACTIVE PROCEDURE EVENT CLASSIFICATION MATRIX RELEASE A

A A DESCRIPTION DESCRIPTION 4.1-U4.2-U Gaseous release exceeds ANY limit and duration in Main Steam Line break outside Primary Containment Table 4.1-U. with isolation.

OPERATING CONDITION: OPERATING CONDITION:

-All -Model -Mode 3

- Mode 2 4.1A K Gaseous release exceeds ANY limit and duration in Table 4.1 -A.

OPERATING CONDITION:

-All 4.1-S K R.I EITHER of the following conditions exists: Unisolable Main Steam Line break outside Primary

"*Gaseous release exceeds or is expected to exceed Containment.

ANY limit and duration in Table 4.1-S.

"*Dose assessment indicates actual or projected dose consequences above 100 imern TEDE or 500 mrem thyroid CDE.

OPERATING CONDITION:

OPERATING CONDITION: - Mode 1 - Mode 3

-All - Mode 2 4.1-G KNN EITHER of the following conditions exists:

" Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-G.

" Dose assessment indicates actual or projected dose consequences above 1000 mrrem TEDE or 5000 mrem thyroid CDE.

OPERATING CONDITION:

-All 4.0 RADIACTIVITY PAGE 35 OF 207 REVISION 30 RELEASE

EPIP-I EMERGENCY 4.0 RADIACTIVITY SECTION II CLASSIFICATION RELEASE EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

REVISION 30 PAGE 36 OF 207 4.0 RADIACTIVITY RELEASE

EMERGENCY EPLP-l CLASSIFICATION SECTION 11 4.0 RADIOACTIVE PROCEDURE EVENT CLASSIFICATION MATRIX RELEASE 6 I DESCRIPTION DESCRIPTION 4.3-U Liquid release rate exceeds 20 times ECL as determined by chemistry sample AND Release duration exceeds or will exceed 60 minutes.

OPERATING CONDITION:

- All 4.3-A Liquid release rate exceeds 2000 times ECL as determined by chemistry sample r AND Release duration exceeds or will exceed 15 minutes.

OPERATING CONDITION:

-All ra

z

z 4.0 RADIACTIVITY PAGE 37 OF 207 REVISION 30 RELEASE

EPIP-1 EMERGENCY 4.0 RADIACTIVITY SECTION H CLASSIFICATION RELEASE EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 30 PAGE 38 OF 207 4.0 RADIACTIVITY RELEASE

EMERGENCY EPIP-1 CLASSIFICATION SECTION 11 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MATRIX POWER EMERGNCY PIP-LOSS OF POWER 5.0 5.0 LOSS OF PAGE 39 OF 207 "REVISION 29 1 POWER

EPIP-1 EMERGENCY 5.0 LOSS OF SECTION H CLASSIFICATION POWER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

5.1-U Loss of normal and alternate supply voltage implies inability to restore voltage from any qualified source to normal or alternate feeder for at least one of the unit specific boards within 15 minutes. At least two boards must be energized from Diesel power to meet this classification. If only one board can be energized and that board has only one source of power then refer to EAL 5.1-Al or 5.1-A2.

5.1-Al Only one source of power (Diesel or Offsite) is available to any one of the listed unit specific 4KV Shutdown Boards. No power is available to the three remaining boards.

5.1-A2 Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in operation 5.1-S would apply.

5.1-S Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only. Determination of the event classification depends on the affected unit operating mode. For units in Shutdown or Refuel 5. 1-A2 would apply.

5.1-G Loss of voltage to all unit specific 4KV Shutdown Boards applies to those boards which normally supply emergency AC power to the affected unit only.

CURVES/TABLES:

UNT4K$HUTDON BOARDAPLABiY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT I A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 3A, 3B, 3C, and 3D REVISION 29 PAGE 40 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIF-1 CLASSIFICATION SECTION II 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MATRIX POWER LOS OF AC POWE DESCRIPTION DESCRIPTION 5.1-U b z

Loss of normal and alternate supply voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes r AND At least two Diesel Generators supplying power to unit specific 4KV Shutdown Boards listed in Table 5.1. z OPERATING CONDITION:

-All 5.1-Al KK 5.1-A2 I Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV Shutdown Shutdown Boards from Table 5.1 for greater than 15 Boards from Table 5.1 for greater than 15 minutes.

minutes AND Only one source of AC power available to the remaining board.

OPERATING CONDITION: OPERATING CONDITION:

-Mode I -Mode 3 -Mode 4 -Defuieled

- Mode 2 -Mode 5 5.1-s SKh*

Loss of voltage to ALL unit Specific 4KV Shutdown Boards from Table 5.1 for greater than 15 minutes.

0

z OPERATING CONDITION:

-Model -Mode3

-Mode2 5.1-G KK 0j

z Loss of voltage to ALL unit specific 4KV Shutdown Boards From Table 5.1 EITHER of the following conditions exists:

"* Restoration of at least one 4KV Shutdown Board is NOT likely within three hours

"*Adaipate oae ooling CANNOTkbeauA OPERATING CONDITION:

-Model -Mode 3

- Mode 2 5.0 LOSS OF PAGE 41 OF 207 REVISION 29 POWER

EPIP-1 EMERGENCY 5.0 LOSS OF SECTION II CLASSIFICATION POWER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

5.2 250V DC bus voltage of less than 248 volts on any feeder to any referenced board constitutes a loss of voltage for that feeder; thus, a loss of DC control voltage to the referenced board. The voltage readings are obtained at the 250V Battery Boards feeding the referenced boards.

CURVES/TABLES:

TABLE 5.2-U UNIT 4KV SHUTDOWN BOARD APPLICABILITY APPLICABLE UNIT APPLICABLE 4KV SHUTDOWN BOARDS UNIT 1 A, B, C, and D UNIT 2 A, B, C, and D UNIT 3 3A, 3B, 3C, and 3D TABLE 5.2-S CRITICAL DC POWER AND ESSENTIAL SYSTEMS COMBINATION CRITICAL 250V DC POWER ESSENTIAL SYSTEMS (UNIT SPECIFIC UNLESS OTHERWISE NOTED) 4KV UNIT BD, A, B, and C CONTROL POWER MAIN CONDENSER AND AND 480V UNIT BOARD A and B CONTROL POWER EHC PUMPS AND AND PANEL 9-9 CABINET 1 REACTOR FEED PUMPS IV III 250V DC RMOV BD A HPCI 250V DC RMOV BD C RCIC IV 250V DC RMOV BDs A, B, and C > 4 MSRVs AND AND 4KV SHUTDOWN BDs A, B, C, and D CONTROL POWER 1 RHR PUMP (3A, 3B, 3C, and 3D FOR UNIT 3) OR 1 CORE SPRAY PUMP REVISION 29 PAGE 42 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIP-l CLASSIFICATION SECTION II 5.0 LOSS OF PROCEDURE EVENT CLASSIFICATION MATRIX POWER SSI O '

DESCRIPTION DESCRIPTION 5.2-U K K Unplanned loss of250V DC control power to ALL unit spjmic 4KV Shutdown Boards from table 5.2-U for greater dhan 15 minutes OR Unplanned loss of250V DC control power to unit specific 480V Shutdown Boards A and B for greater than 15 minutes.

OPERATING CONDITION:

-Mode4

- Mode 5 5.2-S KhK Loss of 250V DC power to ALL combinations (I, II, m, and IV) of essential systems from Table 5.2-S forM greater than 15 minutes.

OPERATING CONDITION:

-Model -Mode 3

- Mode 2

z 5.0 LOSS OF PAGE 43 OF 207 REVISION 29 POWER

EPIP-1 EMERGENCY 5.0 LOSS OF SECTION II CLASSIFICATION POWER EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 29 PAGE 44 OF 207 5.0 LOSS OF POWER

EMERGENCY EPIP-1 CLASSIFICATION SECTION H PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS HAZARDS 6.0 6.0 HAZARDS PAGE 45 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE

- I NOTES:

CURVES/TABLES:

REVISION 30 PAGE 46 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS DESCRIPTION DESCRIPTION 6.1-U Valid, unexpected increase of ANY in plant ARM reading to 1000 mrem/hr (except TIP Room).

ýz CI)

OPERATING CONDITION:

-All 6.1 -Al 6. 1 -A2 Valid, unexpected increase of ANY in plant ARM Control Room radiation levels >15 mreo/hr.

reading to 1000 nrem/hr (except TIP Room).

AND >r Personnel actions required in the affected area(s).

OPERATING CONDITION: OPERATING CONDITION:

-All -All rA

z

z 6.0 HAZARDS PAGE 47 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION H CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

REVISION 30 PAGE 48 OF 207 6.0 HAZARDS

EMERGENCY EPLP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS

$ T RO M TURBI NE 6VCATO FAILR DESCRIPTION DESCRIPTION

__________________________________________________________ I I -

6.3 -U Turbine failure resulting in casing penetration OR Significant damage to turbine or generator seals during operation.

r" t7' OPERATING CONDITION:

-Model

- Mode 2 6.2-A 6.3-A ]

Control Room Abandonment from entry into AOI-100- Turbine failure resulting in visible structural damage to 2 or SSI-16 for ANY Unit Control Room. or penetration of ANY of the following structures from missiles:

Reactor Building Diesel Generator Building Intake Structures OPERATING CONDITION: Control Bay

-All OPERATING CONDITION:

-Mode l

- Mode 2 6.2-S Control Room Abandonment from entry into AOI-100-2 or SSI-16 forANY Unit Control Room.

AND Control of Reactor water level, Reactor pressure, and Reactor power (for Modes 1,2, & 3) or decayheat removal (for Modes 4&5)perAOI-00-2 or SSI-16 as applicable CANNOT be established within 20 minutes after evacuation is initiated.

OPERATING CONDITION:

-All 6.0 HAZARDS PAGE 49 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

APIICALEPLAT.REA REACTOR BUILDING REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATTERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATIONS)

TURBINE BUILDING (ALL ELEVATIONS IN OR ADJACENT TO AREAS CONTAINING SAFE SHUTDOWN EQUIPMENT INTAKE PUMPING STATION (ALL ELEVATIONS)

RADWASTE BUILDING (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING

..................... A.E..

REACTOR BUILDING REFUEL FLOOR 4KV SHUTDOWN BOARD ROOMS 4KV SHUTDOWN BOARD BATTERY ROOMS 480V SHUTDOWN BOARD ROOMS 3A and 3B RMOV BOARD ROOMS 4KV BUS TIE BOARD ROOM CONTROL BAY ELEVATION 593', 606' and 617' DIESEL GENERATOR BUILDINGS (ALL ELEVATIONS)

INTAKE PUMPING STATION (ALL ELEVATIONS)

CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING REVISION 30 PAGE 50 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS NmRw/wcuwgkNrg DESCRIPTION DESCRIPTION 6.4-Ul K 6.4-U2 Confirmed fire in)ANY plant area listed in Table 6A-Ul Unanticipated explosion within the protected area AND resulting in visible damage to ANY permanent NOT extinguished within 15 minutes. structure or equipment.

ýz OPERATING CONDITION: OPERATING CONDITION:

-All -All 6.4-A K Fire or explosion in ANY plant area listed in Table 6.4-A affecting safety system performance OR Fire or explosion causing visible damage to permanent structures or safety systems in ANY area listed in Table 6.4-A.

OPERATING CONDITION:

-All t7 0*

z3 6.0 HAZARDS PAGE 51 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION H CLASSI]ICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

TABLE 6.5/6.6 APPLICABLE PLANT AREA REACTOR BUILDINGS REFUEL FLOOR CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING REVISION 30 PAGE 52 OF 207 6.0 HAZARDS

EMERGENCY EPTP-1 CLASSIFICATION SECTION 11 PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS T A DESCRIPTION DESCRIPTION 6.5-UK EITHER of the following conditions exists:

"Normaloperations impeded due to access restrictions caused by toxic gas concentrations within any building or structure listed in Table 6.5/6.6.

"Confirmedreport by Local, County, or State Officials that a large offsite toxic gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or above the Permissible Exposure Limit (PEL) causing an evacuation of any site personnel.

OPERATING CONDITION:

-All 6.5-A ALL of the following conditions exists:

INA 0 Plantipersonnelreporttoxic gas within any building or structure listed inTable 6.5/6.6.

" Plant personnel report severe adverse health reactions due to toxic gas (i.e., burning eyes, throat, or dizziness) or r Sampling results by Fire Protection or Industrial Safety personnel indicate levels above M the Permissible Exposure Limit (PEL).

"* Determination by the Site Emergency Director that plant personnel would be unable to perform actions necessary to establish and maintain cold shutdown conditions while utilizing appropriate personnel protective equipment.

OPERATING CONDITION:

-ALL rV tz C'

6.0 HAZARDS PAGE 53 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

APPLCABVIPr AREA:::.,.::.

REACTOR BUILDINGS REFUEL FLOOR CONTROL BAY DIESEL GENERATOR BUILDINGS TURBINE BUILDING INTAKE PUMPING STATION RADWASTE BUILDING CABLE TUNNEL (INTAKE TO TURBINE BUILDING)

STANDBY GAS TREATMENT BUILDING REVISION 30 PAGE 54 OF 207 6.0 HAZARDS

EMERGENCY EPJP-1 CLASSIFICATION SECTION It PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS FLAMMIBTO GASES DESCRIPTION DESCRIPTION 6.6-UK EITHER of the following conditions exists:

Release of flammable gas within the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation.r Confinned report by Local, County, or State Officials that a large offsite flammable gas release has occurred within one mile of the site with potential to enter the site boundary in concentrations at or above 25% of the Lower Explosive Limit (LEL).

OPERATING CONDITION:

-All 6.6-A K Release of flanmnable gases within any building or structure listed in Table 6.5/6.6 in concentrations at or above 25% of the Lower Explosive Limit (LEL) for any three readings obtained in a 10 ft. triangular area as indicated by Fire Protection or Industrial Safety personnel using appropriate monitoring instrumentation.

OPERATING CONDITION:

-All rn Iz 03

,4 6.0 HAZARDS PAGE 55 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE I I NOTES:

CURVES/TABLES:

REVISION 30 PAGE 56 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS SECURIT I

DESCRIPTION DESCRIPTION 6.7-U ANY of the following conditions exist:

Bomb device discovered within the plant protected area but NOT within a vital area Attempted or imminent attempt by a hostile force to penetrate the plant protected area barrier Civil disturbance ongoing on the owner controlled property outside the protected area that threatens to r interrupt plant operations Hostage/Extortion situation that threatens to interrupt plant operations.

A credible site-specific security threat notification.

OPERATING CONDITION:

-All 6.7-A Bomb device discovered within ANY plant vital area OR Actual intrusion into the plant protected area by a hostile force.

OPERATING CONDITION:

-All 6.7-S Intrusion into ANY plant vital area by a hostile force.

OPERATING CONDITION:

-All 6.7-G Intrusion by a hostile force into Control Rooms, backup control areas, or plant vital areas which results ina loss of physical control of equipment or functions required to reach and maintain safe shutdown or remove decay heat from any unit.

OPERATING CONDITION:

-All 6.0 HAZARDS PAGE 57 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

REVISION 30 PAGE 58 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION 11 PROCEDURE EVENT CLASSIFICATION MATRIX 6.0 HAZARDS VOEHICLE CLRAA DESCRIPTION DESCRIPTION 6.8-U Vehicle crash (for example; aircraft or barge) into plant structures or systems within the protected area boundary.

OPERATING CONDITION:

-AII 6.8-A ]

Vehicle crash (for example; aircraft or barge) into ANY Plant vital area.

OPERATING CONDITION:

-All C-n

z C) 6.0 HAZARDS PAGE 59 OF 207 REVISION 30

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 6.0 HAZARDS EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 30 PAGE 60 OF 207 6.0 HAZARDS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTS NATURAL EVENTS 7.0 7.0 NATURAL PAGE 61 OF 207 REVISION 28 EVENTS

EPIP-1 EMERGENCY 7.0 NATURAL SECTION II CLASSIFICATION EVENTS EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

REVISION 28 PAGE 62 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION 1I 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTS rDESCRIPTION S7.1-U EART QUAKE m mOmNADOMII I 7.2-U ]

DESCRIPTION H WIS z

Valid annunciation in Unit One Control Room, Panel Report by plant personnel of Tornado striking within Ct 1-XA-55-4B, Window 29, START OF STRONG the protected area boundary.

MOTION ACCELEROGRAPH AND r

Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

z OPERATING CONDITION: OPERATING CONDITION:

-All -All 7.1-A 7.2-A Any ofthe following annunciations in Unit One Cntrol Tornado striking plant vital area Room, Panel l-XA-55-4B: OR

-Window 22, SEISMIC TRIGGER A Onsite wind speed above 90 MPH as indicated using

Window 23, SEISMIC TRIGGER B the meteorological data screen of the Integrated

Window 30, SEISMIC TRIGGER C Computer System (ICS).

AND Assessment by Unit One and Two Control Room personnel that an earthquake has occurred.

OPERATING CONDITION: OPERATING CONDITION:

-All -All Ct 7.0 NATURAL PAGE 63 OF 207 REVISION 28 EVENTS

EPLP-1 EMERGENCY 7.0 NATURAL SECTION 11 CLASSIFICATION EVENTS EVENT CLASSIFICATION MATRIX PROCEDURE EPIP- ENC NOTES:

CURVES/TABLES:

REVISION 28 PAGE 64 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II 7.0 NATURAL PROCEDURE EVENT CLASSIFICATION MATRIX EVENTS I

ES 7.3-U Wheeler Lake level greater than elevation 565 FT. zi AND Ct-Water entering permanent plant structures due to flooding.

OPERATING CONDITION:

-All 7,3-A Wheeler Lake level greater than elevation 565 FT.

AND Either of the following conditions exists:

" Breech or failure of any water-tight structure causing flooding of the structure.

" Affecting equipment required for safe shutdown.

OPERATING CONDITION:

-All rA H.

0<

z nZ 7.0 NATURAL PAGE 65 OF 207 REVISION 28 EVENTS

EPIP-1 EMERGENCY 7.0 NATURAL SECTION II CLASSIFICATION EVENTS EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 28 PAGE 66 OF 207 7.0 NATURAL EVENTS

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CLASSIFICATION MATRIX 8.0 OTHER EMERGENCY DIRECTOR JUDGEMENT 8.0 8.0 OTHER PAGE 67 OF 207 REVISION 29

EPIP-1 EMERGENCY SECTION H CLASSIFICATION 8.0 OTHER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

CURVES/TABLES:

................ T.A &VLE LOSS....

.. O...

... O..

ONSITE COMMUNICATION OFFSITE COMMUNICATIONS PLANT PHONE SYSTEM NODE 1 BELL LINES TWO WAY RADIO (CH Fl, F2, F3, F4, and F5) DIGITAL MICROWAVE SOUND POWER PHONES NRC (FTS-2000)

CELLULAR PHONES (IF AVAILABLE)

HEALTH PHYSICS RADIO NETWORK REVISION 29 PAGE 68 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION 11 PROCEDURE EVENT CALSSIFICATION MATRIX 8.0 OTHER I AT C M A DESCRIPTION DESCRIPTION 8.1-U 8.2-U KI Inability to reach required shutdown condition (Mode 3 Unplanned loss of onsite communication listed in z

or Mode 4) within Technical Specification Limiting Table 8.2-U that defeats the Plant Operations Conditions for Operation (LCO) limits. Staffs ability to perform routine operations m.

OR r Unplanned loss of ALL offsite communication listed in Table 8.2-U.

OPERATING CONDITION:

-Model -Mode3 OPERATING CONDITION:

- Mode 2 -All r

Ct U

0 zC,

I-0 z

U 0

z C,

I REVISION 29 8.0 OTHER 8.0 PAGE 69 PAGE 69 OFOF 207 207 REVISION 29

EPIP-1 EMERGENCY SECTION [I CLASSIFICATION 8.0 OTHER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

8.3 Significant Transient is an unplanned event involving one or more of the following: (1) Automatic turbine runback >25% thermal reactor power: (2) Electrical load reduction >25% full electrical load: (3) Reactor scram: (4) Valid ECCS initiation: or (5) thermal power oscillations > 10%..

CURVES/TABLES:

REACTOR POWER REACTOR PRESSURE REACTOR LEVEL SUBCRITICALLITY DRYWELL TEMPERATURE DRYWELL PRESSURE SUPPRESSION CHAMBER PRESSURE SUPPRESSION POOL TEMPERATURE SUPPRESSION POOL LEVEL REVISION 29 PAGE 70 OF 207 8.0 OTHER

EMERGENCY EPIP-l CLASSIFICATION SECTION 11 PROCEDURE EVENT CALSSIFICATION MATRIX 8.0 OTHER A CAA I DESCRIPTION DESCRIPTION 8.3-U loss of most or all safety system annunciators or indicators which causes a significant loss Z of plant assessment capability for greater than 15 minutes C1 AND Compensatory non-alarming safety system indications are available (SPDS, ICS)

AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant.

OPERATING CONDITION:

-Model -Mode3

-Mode2 8.3-A Fh Unplanned loss of most or all safety system annunciators or indicators which causes a significant oss of plant assessment capability for greater than 15 minutes AND In the opinion of the Shift Manager, increased surveillance is required to safely operate the plant AND EITTER of the following conditions exists:

(SPDS, ICS)

Compensatory non-alarming safety system indications are NOT available A significant transient is in progress.

OPERATING CONDITION:

-Model -Mode 3

- Mode 2 8.3-S rs i o annunciators associated with safety systems AND Compensatory non-alarming safety system indications are NOT available (SPDS, ICS)

AND Indications needed to monitor safety functions are NOT available (Refer to Table 8.3-S)

AND A significant transient is in progress OPERATING CONDITION:

-Model -Mode 3

-Mode2 8.0 OTHER PAGE 71 OF 207 KILV1Ir4U Z9

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 8.0 OTHER EVENT CLASSIFICATION MATRIX PROCEDURE NOTES:

8.4-U Table 8.4-U contains only eanple events hat mayjustify Unuual Eventdassification This evern classification is intended to address unanticipW conditions not explicitly addressed elswhere, but warrant declaration ofan emergency because conditions existswhich theEmergemn.Director&b1ievesto fMI under the Unusual Event Classification. Additionally this EAL should be considered in making emergency classifications regarding challenges to fissionproduct innieas not specifically addre elewheminte EAL matrix 8.4-A This event classification is intended to address unaripated conditions not explicitly addressed elewhere, but that warraint declaration ofan ernagencybecause conditions exist which the Site EmergencyDirector believes to fall under the Alert classification. Additionalythis EAL should be considered in making emergency classifications regarding challenges to fission product baniers not specifically address elmvWer in the EAL mati, 8.4-S This event classification is intended to address unanticipated conditions not explicity addressd elswhere, but that xarant declaration ofan energency because conditions exist which the Site Emergy Directorbelieves to fall under the Site Area Ernergency classification. Additionaly this EAL should be considered in niakng emergen classifications regarding challenges to fission productbariers not specifically address elsewvre in the EAL mtrx 8.4-G This event classification is intended to address unanticipted conditions not eVxhcidtly addressd elswere but that wrrant declaration ofan emergency becaus conditions exist which the Site Emergency Director beieves to fall under the General EmergenLy classification. Additionally this EAL should be considered in malkng emegemn classifications regarding challenges to fission productbarners not specifically address elcwere in the EAL matrix CURVES/TABLES:

0T1 5............... ......

EXAMPLE UNUSUAL EVENTS PLANT TRANSIENT RESPONSE UNEXPECTED OR NOT UNDERSTOOD UNANALYZED SAFETY SYSTEM SONFIGURATION AFFECTING, THREATENING SAFE SHUTDOWN INADEQUATE PERSONNEL TO ACHIEVE OR MAINTAIN SAFE SHUTDOWN DEGRADED PLANT CONDITIONS BEYOND LICENSE BASIS THREATENING SAFE OPERATION OR SAFE SHUTDOWN EMERGENCY PROCEDURES NOT ADEQUATE TO MAINTAIN SAFE OPERATION OR ACHIEVE SAFE SHUTDOWN REVISION 29 PAGE 72 OF 207 8.0 OTHER

EMERGENCY EPIP-1 CLASSIFICATION SECTION II PROCEDURE EVENT CALSSIFICATION MATRIX 8.0 OTHER OTHE DESCRIPTION DESCRIPTION 8.4-U Other events are in process or have occurred which indicate a potential degradation in the level of safety of the plant No radioactive releases are expected whichrequire offsite response. Refer to Table 8.4-U.

OR Any loss or any potential loss of containment OPERATING CONDITION:

-All 8.4-A K Other events are in process or have occned which involve an actual or potential substantial degradation in the level ofsafety ofthe plant Radioactive releases are expected to be within a small fiaction of the EPA guidelines.

OR Any loss or potential loss of fuel cladding or RCS pressure boundary OPERATING CONDITION:

-All 8.4-S K Other events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Radioactive releases are NOT expected to result in exposure levels that exceed EPA guidelines except near the site boundary.

OR Any loss or potential loss of both fuel cladding and RCS pressure boundary OR Potential loss of either fuel cladding or RCS pressure boundary and loss of any additional barrier OPERATING CONDITION:

-ALL 8.4-G K Other events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Radioactive releases are expected to exceed EPA guidelines for exposure levels offsite beyond the site boundary.

OR Loss of any two barriers and potential loss of third barrier OPERATING CONDITION:

-All -M Ci 8.0 OTHER PAGE 73 OF 207 REVISION 29

EPIP-1 EMERGENCY SECTION II CLASSIFICATION 8.0 OTHER EVENT CLASSIFICATION MATRIX PROCEDURE THIS PAGE INTENTIONALLY BLANK REVISION 29 PAGE 74 OF 207 8.0 OTHER

QUESTIONS REPORT for SRO 2002-301 LXRBank

99. G2.4.3 001 Per Regulatory Guide 1.97 post accident instrumentation must be appropriately identified in control rooms to provide information required by the control room operators during accident conditions.

Which ONE of the following describes how RPV level instruments are designated as post accident monitoring and which instruments are used?

A. Black labels are placed on the Emergency Systems Range instruments only.

B. Blue labels are placed on the Post Accident Flood Range instruments only.

C. Black labels are placed on both the Emergency Systems Range and Post Accident Flood range instruments.

D. Blue labels are placed on both the Post Accident Flood Range and the Shutdown Vessel Flood Range instruments.

Re ferences: OPI171.003 Rev.15 pg 24 and 26 Tech Spec Bases B 3.3.3.1 pg B 3.3-84 Note: Modified stem and answers slightly. On last exam O A. Incorrect since on more than the instruments listed.

B. Incorrect since wrong color and on more than the instruments listed.

C. Correct answer.

D. Incorrect since wrong color and wrong instruments listed.

RO Tier: T3 SRO Tier: T3 Keyword: POST ACCIDENT Cog Level: MEM 3.5/3.8 Source: B Exam: BF02301 Test: C Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 99

QUESTIONS REPORT for SRO 2002-301 LXRBank 100. G2.4.41 001 Units 1 and 2 Control Room has become engulfed with smoke and at 1415 the SM orders the control room abandoned. The status of the Units is as follows:

Unit I is defueled.

Unit 2 is in Mode 1.

Unit 3 control room is unaffected.

At 1438 control of Unit 2 is established from Panel 2-25-32 with water level at -75" and slowly lowering.

Which ONE of the following describes the Emergency Classification and who is responsible for EPIP implementation?

A. SM declares an ALERT and he implements EPIP.

B. SM declares an ALERT and Unit 1 US implements EPIP.

C. SM declares a SITE AREA EMERGENCY and SSS implements EPIP.

Df SM declares a SITE AREA EMERGENCY and Unit 3 US implements EPIP.

References:

2-AOI-100-2 Rev. 47 pg 3 Enabling Objective OPL1 71.208 #4 A. Incorrect since this is classified as a Site Emergency due to backup control from Panel 2-25-32 is NOT established within 20 minutes.

B. Incorrect since this is classified as a Site Emergency due to backup control from Panel 2-25-32 is NOT established within 20 minutes.

C. Incorrec since Unit 3 US assumes responsibility for EPIP implementation.

D. Correct answer since control from Panel 2-25-32 is NOT, established within 20 minutes.

Note: Changed stem as far as time it takes to establish cont rol from Panel 2-25-32 which changes the answer.

RO Tier: SRO Tier: T3 Keyword: CONTROL ROOM Cog Level: C/A 2.3 /4.1 Source: M Exam: BF023C'1 Test: S Misc: TCK Tuesday, January 21, 2003 07:30:58 AM 100

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