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08 - Final Written Examination
	ML112700887
Person / Time
Site:	Wolf Creek
Issue date:	08/26/2011
From:	; Division of Reactor Safety IV
To:	; Wolf Creek
References
	Download: ML112700887 (201)
	v • d • e

Cover Page NRC Final

1. 001 AK2.06 4 Given the following plant conditions:

The unit is operating at 88% power

Reactor power is RISING

Tavg is greater than Tref

Pressurizer level is RISING Which ONE of the following would cause the above symptoms to occur?

A. Turbine Load Rejection B. Uncontrolled rod withdrawal C. Failed OPEN S/G Safety valve D. Power Range channel N-43 fails high Justification:

A. Incorrect. Load rejection would justify Tave > Tref and power decrease not increase or PRT.

B. Correct. Continuous rod withdrawal would cause reactor power to increase, Tave to increase, RCS pressure increase due to pressurizer level increase (RCS less dense). Pressure increase could cause PORV to open.

C. Incorrect. RCS pressure would decrease, Tavg would decrease D. Incorrect. Rod withdrawal blocked by C-2 (1/4 >103%). Therefore, none of the operating parameters should change. NI failing high would drive rods in K/A Statement: Knowledge of the interrelations between the Continuous Rod Withdrawal and the following: T-ave./ref. deviation meter (CFR 41.7 / 45.7)

Technical Reference(s): OFN SF-011, LO1732421 Proposed references to be provided to applicants during examination: None Learning Objective: R3/R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 001 AK1.03 [3.9/4.0] {41.8, 41.10}

001 AA2.05 [4.4/4.6] {43.5}

Est Time: 4 min Incorporated Boyle comments. Changed to Rev. 4, 6/18/11 Answer: B

2. 002 K4.01 6 While draining the Reactor Coolant System to mid-loop for nozzle dam removal, wide range level indications BB LI-53A indicates 65" and BB LI-54A indicates 80".

What should the Control Room Operator now do?

A. Continue the drain-down until the NR loop level indicators are on scale and then resolve level discrepancy.

B. Stop the drain-down and use the tygon hose to resolve the level discrepancy and determine the actual level.

C. Continue the drain-down maintaining a difference in readings of less than 15" until the NR loop level indicators are on scale.

D. Stop the drain-down and verify level by comparing NR level indicators BB LI-53B &

BB LI-54B to PZR cold cal level indicator BB LI-462.

Justification:

A. Incorrect, will stop at any point during draindown if anomoly is found, incorrect to continue with discrepancy B. Correct. Acceptable difference on wide range indicators is 10 inches or 3" by NPIS. Step 4.1, 4.2, 4.5.8, Attachment B C. Incorrect, 15" is greater than 10", incorrect to continue with discrepancy D. Incorrect, cold cal 462 is not listed as another indicator to use, discrepancy is verified against tygon hose level.

Cold cal level is used during the draindown K/A Statement: Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following: Filling and draining the RCS (CFR: 41.7)

Technical Reference(s): GEN 00-008, LO1732108 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _12354______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Changed to Rev. 6 7/26/11, incorporated NRC comments. Changed 85" to 80". "D" is credible because we are draining to mid-loop not at mid loop. Requires candidate to know the ranges of cold cal as related to mid-loop Answer: B

3. 003 A4.05 5 With the plant at full power normal operation, which ONE of the following describes where RCP #1 seal leakoff flow is directed?

A. VCT B. PRT C. RCDT D. WHUT Justification A. Correct.

B. Incorrect, Seal return relief discharges to the PRT, normally aligned to the VCT C. Incorrect, #2 & 3 discharges to the RCDT D. Incorrect, #3 seal puts 400 cc/hr leakoff to Containment sump, WHUT receives filtered/clean water

1 is displayed on recorders K/A Statement: Ability to manually operate and/or monitor in the control room: RCP seal leakage detection instrumentation (CFR: 41.7 / 45.5 to 45.8)

Technical Reference(s): SY1300300, M-12BB03, M-12BG01, M-12BG03 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated NRC comments. Changed to Rev. 5 7/26/11. Modified "D" Answer: A

4. 003 AA2.02 5 Given the following plant conditions:

The unit is at 100% power all systems normal

080C, RPI ROD DEV, alarm LIT

081B, ROD AT BOTTOM, alarm LIT

082F, BANK D FULL OUT ROD STOP, alarm LIT

DRPI for rod D4 indicates "0" steps

The RO places rod control in MANUAL

NPIS indications for Reactor Power are as follows:

N41 - 100.1%

N42 - 103.3%

N43 - 100.1%

N44 - 94.7%

Which ONE of the following interlocks or protective features must be cleared before automatic rod withdrawal is reinstated?

A. NIS power range overpower rod withdrawal rod stop.

B. Rod control system non-urgent failure alarm.

C. Overpower Delta T turbine runback and auto rod withdrawal rod stop.

D. Power range channel deviation alarm.

Justification:

A. Correct. C-2 PR High Flux Rod Stop > 103% on 1/4 PRM (Blocks Auto and Manual Rod Withdrawal)

B. Incorrect, not required to move rods, non-urgent does not impair rod control C. Incorrect, has an input but will not get 2/4 especially for dropped rod D. Incorrect, alarm doesn't block rod movement (NPIS generated alarm)

K/A Statement: Ability to determine and interpret the following as they apply to the Dropped Control Rod: Signal inputs to rod control system (CFR: 43.5 / 45.13)

Technical Reference(s): OFN SF-011, SY1301200, M-744-00026, M-744-00021 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Other K/A {CFR}: 003 AK3.04 (3.8-4.1) [41.5,41.10]

Incorporated Boyle comments, changed to Rev. 5, 6/18/11 Answer: A

5. 003 K2.01 5 Given the following plant conditions:

The unit is at 40% RTP

PA02 bus trips out on an 86 Lockout Which ONE of the following describes the plant response?

A. Reactor trips due to the loss of flow from RCP C and RCP D.

B. Control Rods insert due to both Control Rod Drive M/G Sets losing power.

C. NE02 D/G auto starts due to UV on Bus.

D. Letdown isolates due to loss of power.

Justification:

A. Correct. 2/4 on flow logic for a reactor trip under 35% RTP B. Incorrect. Only one RDMG set loses power. The other RDMG would keep power to rods if RPS logic not met. PG 19 & 20 C. Incorrect. S/U Transformer feeds NB02 stub bus (NB02 should still be energized on a loss of PA02).

D. Incorrect. Letdown isolation valves powered from DC. DC bus would not lose power, but would lose charger K/A Statement: Knowledge of bus power supplies to the following: RCPS (CFR: 41.7)

Technical Reference(s): SYS BB-201, SY1300300, E-11PA02, ALR 00-015A, KD-7496, E-11PK01 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 062 K1.04 [3.7/4.2] {41.7}

003 K2.01 [3.1/3.1] {41.7}

012 A3.06 [3.7/3.7] {41.7}

012 K4.04 [3.2/3.5] {41.7}

Est Time: 3 min Incorporated NRC comments, changed to Rev. 5 7/28/11 Answer: A

6. 004 K3.07 3 Given the following plant conditions:

Reactor was operating at 100% RTP

BB LS-459D, PZR LEV CTRL SEL, is selected to L459/L460

A fault occurred on NN01 causing the bus to trip

The Reactor tripped after the bus trip With NO Operator action, which ONE of the following describes the expected PZR level trends following this event?

A. PZR level will increase initially, and then stabilize until the operating crew begins a cooldown of the RCS via steam dumps.

B. PZR level will increase initially, and then decrease until power is restored to NN01, and letdown is restored.

C. PZR level will decrease initially, and then increase until the operating crew begins a cooldown of the RCS via steam dumps.

D. PZR level will decrease initially, and then increase until power is restored to NN01, and letdown is restored.

Justification:

A. Incorrect. PZR level would initially lower due to RCS cooldown. Then, PZR level would be expected to rise based on the loss of letdown and maximum charging flow until instrument power is restored.

B. Incorrect. PZR level would initially lower due to RCS cooldown. Then, PZR level would be expected to rise based on the loss of letdown and maximum charging flow until instrument power is restored.

C. Incorrect. PZR level would initially lower due to RCS cooldown. Then, PZR level would be expected to rise based on the loss of letdown and maximum charging flow until instrument power is restored.

D. Correct. PZR level would initially lower due to RCS cooldown. Then, PZR level would be expected to rise since Letdown isolated due to the loss of instrument power and cannot be reestablished until restored. Charging has swapped to the RWST at maximum rate due to control channel power failure. PZR level will be rising, and will continue to rise until bus power is restored.

K/A Statement: Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: PZR level and pressure (CFR: 41.7/45/6)

Technical Reference(s): LO1732431, SY1301000, M-12BB02, OFN NN-021 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A {CFR}: 004 K1.01 (3.6/4.0) {41.3, 5, 7}

004 K3.02 (3.7/4.1) {41.7}

004 K3.05 [3.8/4.2] {41.7}

004 K4.04 (3.2/3.1) {41.7}

011 K1.01 [3.6/3.9] {41.3, 5, 7}

011 K6.08 [2.1/2.4] {41.7}

057 AA2.16 (3.0/3.1) {43.5}

Est. Time: 3 min.

Incorporated NRC comments. Changed to Rev. 3, 7/28/11, changed subsequent to after in stem Answer: D

7. 005 A4.01 6 Given the following plant conditions:

Unit is at 100% power

During a surveillance test on RHR Pump B, the following control switch indications are observed after the pump is started:

Red Light - OFF / Green Light - ON / Amber Light - ON Locally at the breaker:

Red light - OFF / Green Light - ON / White Light - ON / Blue Light - OFF

Annunciator 00-050A, RHR Pump Trouble, is in alarm Which ONE of the following describes the status of RHR Pump B?

A. Tripped on overcurrent B. Tripped on differential C. Not running due to loss of control power D. Not running after a local stop Justification A. Correct. Pump is not running and blue light OFF indicates the pump trip is caused by an over current relay at the pump breaker.

B. Incorrect, This is not a bus electrical trip -- Blue light OFF indicates a 186 instantaneous overcurrent lockout on the breaker not the bus.

C. Incorrect, CR and Local breaker indication: pump is not running. Control power is available at the breaker.

D. Incorrect, Although the pump has a local start at the breaker, the breaker will not close with a 186 relay actuated ALR 00-050A, RHR PUMP TROUBLE 2.0 SYMPTOMS OR ENTRY CONDITIONS 2.1 When the annunciator is in the alarm state, it will reflash should any other initiating condition occur.

2.2 This procedure is entered when any of the following occurs:

RHR Pump A motor current greater than 100 amps

RHR Pump A trip on overcurrent

RHR Pump B motor current greater than 100 amps

RHR Pump B trip on overcurrent 2.3 The following is a list of the instrumentation:

o Relay 151/X, Time Overcurrent Relay o Relay 186/M, Instantaneous Overcurrent Lockout Relay

K/A Statement: Ability to manually operate and/or monitor in the control room: Controls and indication for RHR pumps (CFR: 41.7 / 45.5 to 45.8)

Technical Reference(s): SY1300500, E-13EJ01, ALR 00-050A, STS EJ-100B Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _46737______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___2008 DCPP_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 6, 8/1/11 Answer: A

8. 005 AK3.01 6 Given the following plant conditions:

A reactor trip without SI has occurred

The Operators are implementing EMG E-0, REACTOR TRIP OR SAFETY INJECTION

RCS temperature is 552°F and decreasing slowly

RCS pressure is 2215 psig

A check of rod position indications shows that one control rod has not fully inserted

The Operators continue with the procedure without taking action concerning the stuck rod Given the conditions above, why is Emergency Boration NOT required while performing the Immediate Actions of EMG E-0?

A. Emergency boration will be required after the transition to EMG ES-02, REACTOR TRIP RESPONSE.

B. Maintaining adequate shutdown margin is ONLY a concern following a reactor trip with RCS temperature less than 551°F and decreasing.

C. Emergency boration will be required after the transition to EMG FR-S2, RESPONSE TO LOSS OF CORE SHUTDOWN.

D. Verifying that the remaining rods are fully inserted ensures adequate Shutdown Margin is present.

Justification A. Incorrect, only one rod stuck, rods checked in ES-02 B. Incorrect, SDM is always a concern with stuck rods. 551 is the Min. Temp for Criticality C. Incorrect, SDM is assured with only 1 stuck rod incorrect transition based on plant conditions D. Correct, SDM is assumed to have the most reactive rod out K/A Statement: Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: Boration and emergency boration in the event of a stuck rod during trip or normal evolutions (CFR 41.5,41.10 / 45.6 / 45.13)

Technical Reference(s): BD-EMG E-0, EMG E-0, BD-EMG ES-02, ES-02, LO1732315, USAR 4.3.1.5 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _5, 10_

55.43 _____

Comments:

Other K/A: 007 EK3.01 4.0/4.6 Incorporated NRC comments, changed to Rev. 6, 7/27/11 Answer: D

9. 005 K2.03 4 Which ONE of the following is the power supply to BB PV-8702B, RHR Suction from Loop 4 HL?

A. NG01B B. NG02B C. NG03C D. NG04C Justification:

A. Incorrect, see below B. Correct.

C. Incorrect, see below D. Incorrect, see below BB PV-8702A (NG02B) and EJ HV-8701A (NG01B), RCS Hot Leg 1 BB PV-8702B (NG02B) and EJ HV-8701B (NG01B), RCS Hot Leg 4 NG03C Auxiliary Building Ventilation Loads NG04C Auxiliary Building Space Heater Loads The electrical power supplies for the RHR pumps are as follows:

NB01 RHR pump A, (Breaker NB0101).

NB02 RHR pump B, (Breaker NB0204).

K/A Statement: Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves (CFR: 41.7)

Technical Reference(s): SY1300500, M-12BB01, E-13BB12B, CKL BB-110 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Est. Time: 2 min Incorporated Brendan comments, changed to Rev. 4, 6/9/11 Answer: B

10. 006 K3.01 4 Given the following plant conditions:

Recovery from a Large Break LOCA is in progress Which ONE of the following components will have the GREATEST impact on LONG TERM core cooling?

Loss of . . .

A. Safety Injection Pumps.

B. Reactor Coolant Pumps.

C. Centrifugal Charging Pumps.

D. Residual Heat Removal Pumps.

Justification:

A. Incorrect. SIPs have ~600 gpm/pump flow. RHR has more capacity.

B. Incorrect. RCPs would be OFF due to low pressure. If used due to FR-C1, they could provide some cooling by increasing vapor flow through tube side of S/G. Radiant to conduction to convective heating less efficient than convective heating from RHR liquid flow.

C. Incorrect. CCPs have ~3-400 gpm/pump flow. RHR has more capacity.

D. Correct. 3800 gpm/pump, suction realigned to CTMT Recirc Sump on RWST LOLO level (long term)

K/A Statement: Knowledge of the effect that a loss or malfunction of the ECCS will have on the following: RCS (CFR: 41.7 / 45.6)

Technical Reference(s): USAR 6.3, 15.6.5, EMG Executive Volume, SY1300500, SY1300600 Proposed references to be provided to applicants during examination: None Learning Objective: R1, R3, R5 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 006 K6.13 [2.6/2.9] {41.7}

Est Time: 3 min Incorporated Boyle comments, changed to Rev. 4, 6/18/11 Answer: D

11. 007 2.1.32 4 Which ONE of the following is the reason for maintaining a nitrogen blanket on the Pressurizer Relief Tank (PRT)?

A. Minimize lining corrosion of the PRT due to oxygen entrainment from RCS.

B. To minimize the possibility of forming an explosive mixture of hydrogen and oxygen in the PRT.

C. To ensure NPSH when circulating water from the PRT through the Reactor Coolant Drain Tank HX.

D. To reduce the amount of hydrogen released to Containment if overpressure causes rupture of the rupture disks.

Justification A. Incorrect, RCS is maintained with hydrogen or nitrogen so that oxygen in minimized B. Correct. The cover gas prevents air in-leakage and minimizes formation of explosive hydrogen-oxygen content, and isolates (HV-8026 & 8027) on a Containment Isolation Signal Phase A (CIS-A).

C. Incorrect, level basis, N2 does not improve NPSH due to dissolved gas D. Incorrect, PRT not a major source of Hydrogen, although it will be a source if rupture disk breached, N2 does not reduce amount K/A Statement: Ability to explain and apply system limits and precautions. (CFR: 41.10 /

43.2 / 45.12)

Technical Reference(s): SY1300200, M-12BB02, SYS BB-202 (4.4, 4.4.1, 4.4.2)

Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _10___

55.43 __2___

Comments:

Other K/A: 007 2.1.28 [3.2/3.3] {41.7}

007 K1.04 [2.1/2.3] {41.3}

Est Time: 2 min Incorporated NRC comments, changed to Rev. 4, 7/28/11 Answer: B

12. 007 A2.01 7 Given the following plant conditions:

The plant was at 100% power

The reactor is tripped but the Pressurizer Safety Valve has not reseated properly

Pressurizer Relief Tank (PRT) pressure is 30 psig, rising slowly

PRT temperature is 175°F, rising slowly Which ONE of the following choices completes the following statement?

PRT pressure will continue to rise until . . . .

A. rupture disk fails. Vent the PRT to Waste Gas to reduce pressure then cool the PRT by filling and draining the PRT.

B. rupture disk fails. Vent the PRT to Waste Gas to reduce pressure then cool the PRT using the RCDT Heat Exchanger.

C. BB HIS-8026, PRT N2 SPLY INNER CTMT ISO VLV, opens to reduce pressure.

Manually initiate Reactor Makeup Water as desired to lower pressure and verify the RCDT Pump(s) maintains PRT level in the desired band.

D. BB HIS-8026, PRT N2 SPLY INNER CTMT ISO VLV, opens to reduce pressure.

Manually initiate Reactor Makeup Water as desired to lower pressure and open PRT Drain to Containment Sump Valve BB HIS-8037A to control PRT level.

Justification:

A. Incorrect. Fill and drain is not used unless Waste Gas and RCD Subsystems are not available.

B. Correct. Pressure will rise unless the safety valve reseats. When cooling the PRT using the RCDT Heat Exchanger, do not allow RCDT pressure to exceed 50 psig. Exceeding 50 psig will overpressurize the Reactor Coolant Drain Subsystem. Pressure has to lowered first then cool the PRT. If annunciator 00-034D, PRT TEMP HI is received, the contents of the PRT should be cooled by recirculating the tank through the Reactor Coolant Drain Tank (RCDT) Heat Exchanger.

C. Incorrect. A leaking Safety valve will continue to leak until it reseats, pressure will continue to rise unless action taken. HIS-8026 does not auto open to reduce pressure. Wrong action to implement.

D. Incorrect. A leaking Safety valve will continue to leak until it reseats, pressure will continue to rise unless action taken. HIS-8026 does not auto open to reduce pressure. Wrong action to implement.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck-open PORV or code safety (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Technical Reference(s): ALR 00-034D, ALR 00-034E, M-12BB02, SYS BB-202 (Prec. 4.1 & 4.6, Sec 6.4), SY1301000

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: R1/R2/R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/28/11 Answer: B

13. 007 EK2.03 6 Given the following plant conditions:

The reactor tripped from 100% power

ALL PR HI FLUX bistables are LIT on SB069, PARTIAL TRIP STATUS PERMISS/BLOCK PANEL

ALL SG LOLO bistables are LIT on SB069, PARTIAL TRIP STATUS PERMISS/BLOCK PANEL Which ONE of the following describes the MINIMUM bistable conditions required to close and maintain closed reactor trip breakers?

A. Reset 4/4 PR HI FLUX and 3/4 SG LOLO bistables clear on 4/4 SG.

B. Reset 4/4 PR HI FLUX and 4/4 SG LOLO bistables clear on 4/4 SG.

C. Reset 3/4 PR HI FLUX and 4/4 SG LOLO bistables clear on 4/4 SG.

D. Reset 3/4 PR HI FLUX and 3/4 SG LOLO bistables clear on 4/4 SG.

Justification:

A. Incorrect, would work but is not the minimum.

B. Incorrect, would work but is not the minimum.

C. Incorrect, would work but is not the minimum.

D. Correct. This is the minimum required.

K/A Statement: Knowledge of the interrelations between a reactor trip and the following:

Reactor trip status panel (CFR 41.7 / 45.7)

Technical Reference(s): SY1301200 (Table 1), TS Basis 3.3.1, SY1301100, M-744-00019, M-744-00020, M-744-00024, ALR 00-085B, ALR 00-085A Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 6, 8/1/11 Answer: D

14. 008 AK2.02 7 Given the following plant conditions:

The plant is in MODE 3 at normal operating temperature and pressure.

NORMAL CHRG PMP FLOW CTRL, BG FK-462 is in AUTO, maintaining 27% PZR level.

The reference (upper) leg tap for Pressurizer level transmitter BB LT-460 breaks off of the Pressurizer, and RCS pressure starts decreasing.

Which ONE of the following describes the INITIAL responses of PZR level instruments BB LI-459 and BB LI-460 to this event?

BB LI-459 BB LI-460 PZR Level Indication PZR Level Indication A. Increasing trend Off-scale high B. Decreasing trend Off-scale low C. Decreasing trend Off-scale high D. Increasing trend Off-scale low Justification A. Incorrect. BB LI-459 controls charging flow. BB LI-459 sense change in actual level and charging flow increases. BB LI-459 would decrease, BB LI-460 would be off-scale high due to break B. Incorrect, BB LI-459 would decrease, BB LI-460 would be off-scale high due to break C. Correct, based on simulator run and validation BB LI-459 would decrease, BB LI-460 is off-scale high due to break.

D. Incorrect, first part incorrect, BB LI-460 would be off-scale high due to break Charging flow can be selected to any channel. BB LT-460 controls only the isolation valve BG LCV-460.

Since BB LT-460 has failed high, the valve BG LCV-460 remains open.

The PZR level selector switch, BB LS-459, allows 459 or 461 only as the controlling channel.

460 goes off-scale high due to break D/P vs level. The tap is 3/4" to condensate pot only the instruments have the flow restrictors. 459 goes down due to 3/4" LOCA > charging pump capacity.

A makeup flow rate from one centrifugal charging pump is adequate to sustain pressurizer level at 2,250 psia for a 0.375-inch-diameter hole. This break results in a loss of approximately 17.5 lb/sec (127 gpm at 130°F and 2,250 psia). USAR Ch 6 K/A Statement: Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Sensors and detectors (CFR 41.7 / 45.7)

Technical Reference(s): SY1301000; BD-OFN SB-008, Att J; M-12BB02, USAR Ch. 6 Proposed references to be provided to applicants during examination: None

Learning Objective: R10/R5 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/28/11 Answer: C

15. 008 K1.02 7 Given the following plant conditions:

The plant is in MODE 3 with a cooldown in progress

CCW Train "B" is supplying the service loop

Annunciator 053D, CCW SRG TK B LEV HILO, alarms

CCW Surge Tank "B" water level is slowly decreasing

Diverse indications show the makeup valve is full open and the Aux Watch reports there are no signs of system leakage Which ONE of the following describes the probable cause of the above conditions?

A. Excessive primary plant cooldown rate.

B. A leak exists in an RCP thermal barrier heat exchanger.

C. A tube leak exists in the seal water return heat exchanger.

D. CCW surge tank vent is clogged.

Justification:

A. Incorrect, CCW not affected by cooldown as much as SW or ESW. RHR not in service in Mode 3 B. Incorrect, Pressure would be higher than CCW pressure resulting in inleakage.

C. Correct. VCT pressure is < CCW D. Incorrect, clogged vent would not cause a decreased level on increased pressure.

Tank is vented, vent only closes on hi rad.

K/A Statement: Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Loads cooled by CCWS (CFR: 41.2 to 41.9 / 45.7 to 45.9)

Technical Reference(s): ALR 00-053D, OFN EG-004, SY1400800, LO1732414, M-12EG01 Proposed references to be provided to applicants during examination: None Learning Objective: R2/R4 Question Source: Bank # _18237______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _2, 4, 7, 8_

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/28/11 Answer: C

16. 009 EK2.03 7 Given the following plant conditions:

A small break LOCA has occurred

The condenser is not available to receive steam

All S/Gs have been determined to be intact

PZR level indicates zero and RVLIS indicates a bubble in the reactor vessel

Reactor vessel level is decreasing slowly

RCS pressure is greater than all S/G pressures

ECCS is running with 100 gpm flow indicated

RCPs are secured

S/G pressures are slowly increasing

RCS hot leg temperatures are slowly increasing Which ONE of the following describes core cooling mechanisms presently available?

A. Natural circulation cooling removing almost ALL heat from core, Break flow removing ALL heat from primary B. Boiling removing almost ALL heat from the core, Break flow removing ALL heat from primary C. Boiling removing almost ALL heat from the core, Condensation of steam in U-tubes occurs D. Natural circulation cooling removing almost ALL heat from core, Condensation of steam in U-tubes occurs Justification:

At this point core is boiling; most heat removed by heat of vaporization, heat is removed by SI injection; break flow and S/Gs (reflux).

A. Incorrect. Natural circulation does not exist. S/Gs do contribute, natural circulation is not primary heat removal B. Incorrect. First part correct, Break flow only removes some heat from the primary must have SI flow C. Correct.

D. Incorrect. Natural circulation does not exist, S/Gs do contribute Attachment B of EMG ES-04 K/A Statement: Knowledge of the interrelations between the small break LOCA and the following: S/Gs (CFR 41.7 / 45.7)

Technical Reference(s): ERG FR-C1 Bkgd, TS Basis 3.5.2, USAR 15.6.5, West. MCD Ch. 2, LO1610706

Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Est Time: 3 min Modified based on NRC comments. Changed to Rev. 7, 7/28/11, reformatted choices, made each choice balanced.

Answer: C

17. 010 K5.01 7 Given the following plant conditions:

Pressurizer pressure 985 psig

Pressurizer Relief Tank (PRT) pressure 5 psig

PRT temperature 90°F Assume ambient heat losses are negligible and the steam quality in the Pressurizer bubble is 100%. Also assume Pressurizer and PRT conditions do NOT change.

Which ONE of the following PORV downstream temperatures and fluid state would be caused by a leaking Pressurizer PORV?

A. 230°F, two phase B. 230°F, saturated vapor C. 300°F, superheated vapor D. 353°F, saturated vapor JUSTIFICATION:

A. Incorrect. Candidate used saturation temperature for 20 psia.

B. Incorrect. Candidate used saturation temperature for 20 psia.

C. Correct. Enthalpy for 1000 psia is 1193 btu/lbm. Throttling is an isenthalpic process. Using Mollier diagram, this will give a temperature of 300°F in the superheated region.

D. Incorrect. Enthalpy for 1000 psia is 1193 btu/lbm. Throttling is an isenthalpic process. Using Mollier diagram, this will give a temperature of 353°F in the superheated region. This choice is on the opposite end of the saturation curve.

K/A Statement: Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables (CFR: 41.5 / 45.7)

Technical Reference(s): Mollier Diagram and/or Steam Tables, LO1131121, LO1131115 Proposed references to be provided to applicants during examination: Mollier Diagram and/or Steam Tables Learning Objective: R20/21 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 008 AK1.01 (3.2/3.7) {41.8, 10}

193003 K1.25 (3.3/3.4) {41.14}

Est. Time: 3 min.

Modified based on Jane comments. Changed to Rev. 7, 7/2/11 Answer: C

18. 011 EA2.04 4 Given the following plant conditions:

Reactor trip and SI have actuated

RCS pressure is 300 psig and DECREASING

PZR level is offscale low

RCS temperature is 504°F and DECREASING

Containment pressure is 6 psig and INCREASING slowly

S/G pressures are 690 psig and DECREASING slowly

Crew is performing EMG E-0, REACTOR TRIP OR SAFETY INJECTION Which ONE of the following caused these conditions?

A. PZR Safety Valve failed OPEN B. Main Feed Line Break inside Containment C. Cold Leg Break D. Main Steam Line Break inside Containment Justification:

A. Incorrect. A failed safety valve would result in Pressurizer level increase. An operator who does not understand the difference between a vapor space LOCA and a LOCA may choose this distractor.

B. Incorrect. A main feedline break would result in S/G pressures much lower than the saturation pressure for the RCS temperature of 504°F. An operator who does not understand the difference between a feedline break and a LOCA may choose this distractor. S/G pressure does not come down until the feed line is isolated or the S/G is dry.

C. Correct. S/G pressures are at ~ Tsat for RCS temperature. RCS pressure < S/G pressures.

Therefore, RCS cooldown is causing S/G pressure decrease, no faulted S/Gs.

D. Incorrect. A main steamline break would result in S/G pressures much lower than the saturation pressure for the RCS temperature of 504°F and pressure would decrease rapidly in the faulted S/G.

An operator who does not understand the difference between a MSLB and a LOCA may choose this distractor.

K/A Statement: Ability to determine or interpret the following as they apply to a Large Break LOCA: Significance of PZR readings (CFR 43.5 / 45.13)

Technical Reference(s): EMG E-0, Steam Tables, BD-EMG E-0, LO1610500, LO1732313, USAR Chap. 15 Sections 15.6.1, 15.6.5, 15.1.5, 15.2.8 Proposed references to be provided to applicants during examination: None Learning Objective: R3/4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Other K/A: 011 EA2.13 [3.7/3.7] {43.5}

002 K1.11 [4.1/4.2] [41.2, 4, 5, 7}

002 K3.03 [4.2/4.6] {41.7}

002 K5.11 [4.0/4.2] {41.5}

Est. Time: 3 min.

Incorporated Brendan comments, changed to Rev. 4, 6/9/11 Answer: C

19. 011 K6.04 6 Given the following plant conditions:

Unit is at 100% power

All control systems are in automatic

Pressurizer Level Control Selector Switch, BB LS-459D, is in the L459/L460 position

Pressurizer Level Channel II, BB LT-460, fails LOW Which ONE of the following lists ALL plant responses due to the failures with no Operator actions?

A.

Letdown isolates

Pressurizer level increases

All Pressurizer Heaters shut off B.

No effect on charging flow

No Letdown Isolation

Pressurizer Backup Heaters turn on C.

Charging flow to maximum

No Letdown Isolation

Pressurizer Backup Heaters turn on D.

Charging flow to maximum

Letdown Isolates

All Pressurizer Heaters shut off Justification A. Correct.

B. Incorrect, charging flow decreases, letdown isolates C. Incorrect, letdown isolates D. Incorrect, all heaters turn off, charging will decrease.

BB LI-460 is not the controlling channel. Ran on simulator Level increases, Ltdn isolates, B/U htrs turn off K/A Statement: Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS: Operation of PZR level controllers (CFR: 41.7 / 45.7)

Technical Reference(s): SY1301000, BD-OFN SB-008, ALR 00-032D, E-13BB24, E-13BB22, M-12BB02, M-744-00028, M-744-00029 Proposed references to be provided to applicants during examination: None Learning Objective: R5/R10 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A:

011 K2.03 (2.4 - 2.5) 011 K4.02 (3.3 - 3.4) 011 K4.04 (3.0 - 3.3) 011 K6.04 (3.1 - 3.1) 011 A1.01 (3.5 - 3.6) 011 A2.11 (3.4 - 3.6)

Modified based on NRC comments. Changed to Rev. 6, 7/28/11 Answer: A

20. 012 K4.08 5 Given the following plant conditions:

Unit is at 40% power

Solid State Protection System (SSPS) Train B Actuation Logic testing is being performed

Train B SSPS Mode Selector Switch is in the TEST position

Train B SSPS Input Error Inhibit Switch is in the INHIBIT position Which ONE of the following describes the status of the Reactor if one 48-volt power supply were lost on Train A SSPS and why?

A. Reactor Trip with a GENERAL WARNING for BOTH Train A and Train B SSPS and all red First Out annunciators illuminated.

B. Reactor remains at 40% power with a GENERAL WARNING for Train B SSPS ONLY.

C. Reactor Trip with GENERAL WARNING for BOTH Train A and Train B SSPS and one red First Out annunciator illuminated.

D. Reactor remains at 40% power with a GENERAL WARNING for Train A SSPS ONLY.

Justification A. Incorrect, Plausible because a Reactor Trip is generated, but a First Out annunciator will not occur because there is no condition for a first out B. Incorrect, Plausible because a GENERAL WARNING is generated for a loss of either 48 VDC power supply, however, performing testing on the other Train generates a GENERAL WARNING for both Trains and the Unit trips.

C. Correct. Testing on one train of SSPS generates a GENERAL WARNING. A loss of any of the four DC power supplies in the other Train of SSPS also generates a GENERAL WARNING and opens the Reactor Trip Breakers.

D. Incorrect, Plausible because a GENERAL WARNING is generated for a loss of either 48 VDC power supply, however, performing testing on the other Train generates a GENERAL WARNING for both Trains and the Unit trips.

K/A Statement: Knowledge of RPS design features and/or interlocks which provide for the following: Logic matrix testing (CFR: 41.7)

Technical Reference(s): EMG E-0, SY1301100, M-744-00019, ALR 00-075A, ALR 00-076A Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _NRC Bank______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___CPNPP March 2009 NRC RO_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated NRC comments, changed to Rev. 5. 7/28/11 Answer: C

21. 013 K4.09 4 Which ONE of the following is an EXCEPTION to the bistable de-energize to actuate design criteria of the Engineered Safeguards Actuation System?

A. Steam line low pressure negative rate MSLIS.

B. Safety Injection actuation.

C. Containment Isolation Phase A (CISA).

D. Containment spray actuation.

Justification A. Incorrect, de-energize to actuate (ESF signal)

B. Incorrect, de-energize to actuate (ESF signal)

C. Incorrect, de-energize to actuate (ESF signal)

D. Correct, energize to actuate (ESF signal)

RWST is also energize to actuate K/A Statement: Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Spurious trip protection (CFR: 41.7)

Technical Reference(s): SY1301301, M-744-00025 Proposed references to be provided to applicants during examination: None Learning Objective: R5/R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated Jane comments, changed to Rev. 4. 6/18/11 Answer: D

22. 014 A1.04 6 Given the following plant conditions:

Plant is at 90% power

AFD is at -1% on all four channels

The RO attempts to insert control rods 8 steps to maintain Tavg/Tref The following was observed by the RO:

Group step counters counted down 8 steps

DRPI indication for the affected rods did not change Based on these conditions, what are the expected diverse indications assuming the DRPI System is functioning properly?

A.

AFD would go more negative

Power Range NIs would decrease

No radial flux tilt change B.

AFD would go less negative (towards 0%)

Power Range NIs would not change

A small radial flux tilt will be noted C.

AFD would not change

Power Range NIs would decrease

A small radial flux tilt will be noted D.

AFD would not change

Power Range NIs would not change

No Radial flux tilt change Justification A. Incorrect. Rods did not move. Group step counters get signal from Rod Control demanding motion.

No changes would be noted on diverse indications B. Incorrect. Rods did not move. Group step counters get signal from Rod Control demanding motion.

No changes would be noted on diverse indications C. Incorrect. Rods did not move. Group step counters get signal from Rod Control demanding motion.

No changes would be noted on diverse indications D. Correct K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including: Axial and radial power distribution Technical Reference(s): ALR 00-079D, SY1301501, SY1300100, SY1301400 Proposed references to be provided to applicants during examination: None

Learning Objective: R1/R6 Question Source: Bank # _GSP Bank______

Modified Bank # _______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Modified based on Brendan comments. Changed to Rev. 5, 6/9/11 Answer: D

23. 015 K3.02 5 Given the following plant conditions:

Startup in progress

Reactor power 50%

Control Bank D rods are at 170 steps in Auto

Power range monitor NI44 fails off scale HIGH over 30 seconds Which ONE of the following is the expected INITIAL response of the Rod Control System without Operator action?

A. Rods drive in at max rate until rate input is zero.

B. Rods drive out at max rate until rate input is zero.

C. Rods drive in until C-5 permissive setpoint is reached.

D. Rods drive out until C-11 permissive setpoint is reached.

JUSTIFICATION:

A. Correct; Rod Control Power Mismatch circuit (rate of change difference between primary power, NIS, vs secondary power, turbine impulse pressure), once the failed instrument stops- upper or high scale range peg, the mismatch change or rate input is 'zero'. Max rate is 72 spm.

B. Incorrect, opposite rod direction. Once the failed instrument stops- upper or high scale range peg, the mismatch change or rate input is 'zero'. Max rate is 72 spm.

C. Incorrect, right rod direction, C-5 is wrong. Once the failed instrument stops- upper or high scale range peg, the mismatch change or rate input is 'zero'. Max rate is 72 spm. C-5 is at 15% by impulse D. Incorrect, opposite rod direction. Once the failed instrument stops- upper or high scale range peg, the mismatch change or rate input is 'zero'. Max rate is 72 spm.

K/A Statement: Knowledge of the effect that a loss or malfunction of the NIS will have on the following: CRDS (CFR: 41.7 / 45.6)

Technical Reference(s): SY1301501, SY1300100, BD-OFN SB-008 - Att. R, M-744-00020, M-744-00021, M-744-00026 Proposed references to be provided to applicants during examination: None Learning Objective: R7 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 001 K1.05 (4.5/4.4) {41.2, 5, 6, 7}

001 K6.05 [2.4/2.7] {41.7}

015 K1.03 [3.1/3.1] {41.2, 5, 6, 7}

Est. Time: 3 min.

Modified based on Brendan comments. Changed to Rev. 5, 6/6/11 Answer: A

24. 015/017 AA2.01 5 Given the following plant conditions:

The unit is at 32% power

The following SB069 windows are lit:

- LO FLOW L3 FB434A

- LO FLOW L3 FB435A

- LO FLOW L3 FB436A

"C" RCP indications are:

- BB II-3, RCP C AMPS - '0' amps

- Green and amber lights above the handswitch are lit

The other RCPs indications are normal Which ONE of the following identifies the condition causing the C RCP indications and the required Operator action?

Condition causing RCP Trip Required Operator Action A. Under voltage Trip the reactor and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION B. Under voltage Shutdown the plant IAW GEN-004, POWER OPERATION C. Overcurrent Trip the reactor and go to EMG E-0, REACTOR TRIP OR SAFETY INJECTION D. Overcurrent Shutdown the plant IAW GEN-004, POWER OPERATION Justification The RCPs have protection from undervoltage and underfrequency via relays. Both conditions can cause trips of the RCPs and if conditions and logic is met, the reactor would automatically trip.

A. Incorrect, The conditions in the stem describe the trip of RCP C from an overcurrent condition not an underfrequency/undervoltage trip, however NO automatic trip would be initiated at the stated power level (below P8). Plausible because the condition causing the RCP trip is a valid tripping conditions and if the power level had been greater than P8, an automatic reactor trip should have occurred. Tripping the reactor is not procedurally required.

B. Incorrect, The conditions in the stem describe the trip of RCP C from an overcurrent condition not an underfrequency/undervoltage trip. With the reactor power level below P-8, no automatic reactor trip would be initiated from low flow condition in a single loop.

C. Incorrect, The conditions in the stem describe the trip of RCP C from an overcurrent condition. Below P-8 a reactor trip is NOT a required Operator action.

D. Correct, The conditions in the stem describe the trip of RCP C from an overcurrent condition not an underfrequency/undervoltage trip. The underfrequency trip of the RCP requires 2/4 logic to be made in the SSPS which then will trip the reactor and all RCPs. In MODE 1, 3 RCP operation is NOT permitted by TS 3.4.4 and Shutdown is required.

RCPC UV bistable in the same column would be lit (there is also a RCPC UF in that column). Ran this on the simulator and the RCPC RLY TROUBLE UV/UF did not light but the RCPC UV bistable did.

SY1300300 Objective 6, Section 6.3 page 31 states the Trouble light actuates when there is a problem ONLY with the relay. IF NK bus power is lost (power feed to the RCP relays) the TROUBLE light would light, but the RCP would continue to run.

K/A Statement: Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Cause of RCP failure (CFR 43.5 /

45.13)

Technical Reference(s): GEN 00-004, SY1300300, TS 3.4.4, TS Basis 3.4.4 Proposed references to be provided to applicants during examination: None Learning Objective: R6 Question Source: Bank # __GSP Bank_____

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Modified based on NRC comments. Changed to Rev. 5, 7/28/11.

Answer: D

25. 016 K5.01 7 Which ONE of the following explains the purpose of the Isolation Amplifier associated with the Steam Generator Pressure Instrument Loop?

A. Amplifies and integrates the pressure output signal between Containment and the instrument racks.

B. Isolates and integrates the pressure transmitter from the impacts of changing Containment pressures.

C. Protects the control signal from a perturbation due to backfeed from a disturbance in the Reactor Protection System.

D. Protects the Reactor Protection signal from a perturbation due to backfeed from a disturbance in the control circuit.

Justification A. Incorrect, RVLIS amplifier, not an isolation amp, has an isolation bellows, no integration circuits B. Incorrect, EQ issue, no integration circuits C. Incorrect, opposite response D. Correct.

K/A Statement: Knowledge of the operational implication of the following concepts as they apply to the NNIS: Separation of control and protection circuits (CFR: 41.5 / 45.7)

Technical Reference(s): SY1301100 - Section 1.2, USAR 7.1.2.2.1, SY1301200 Proposed references to be provided to applicants during examination: None Learning Objective: R1/R5 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Other K/A: 010 K1.01 [3.9/4.1] {41.7}

010 2.1.28 [3.2/3.3] {41.7}

Est Time: 2 min.

Modified based on NRC comments. Changed to Rev. 7, 7/28/11 Answer: D

26. 022 A1.04 6 Given the following plant conditions:

The unit is at 100% power

XNB01 has a fault causing 018B, NB01 BUS UV, to alarm All equipment operates as designed when the following conditions occurs:

ESW Pump Discharge pressure is 105 psig

CCW HX Outlet temperatures are RISING

Annunciator 060E, CTMT SUMP A/B LEV HI, is LIT

Containment temperature is slowly rising Which ONE of the following is causing the Containment conditions?

A. RCS leak inside Containment B. CCW Thermal Barrier Leak C. Containment Cooler supply line break D. Steam leak inside Containment JUSTIFICATION:

A. Incorrect; No rad monitor readings/alarms to indicate a RCS leak. CTMT temperature rising could indicate a primary leak, however without a corresponding change in humidity or radiation an RCS leak is not indicated.

B. Incorrect; No CCW low pressure/surge tank alarms to indicate CCW leakage. CCW temps would be rising on a TB leak, No effect on Ctmt temp or change in rad monitors.

C. Correct; ESW pressure is below the alarm setpoint indicates system leakage as a possible cause of the low header pressure. That, coupled with containment sump alarm and increasing temperature indicates that the leakage is from ESW supply line to coolers.

D. Incorrect; Sump alarms would be in for a steam leak condensation, containment temps would be rising on a steam leak, does not explain ESW pressure low.

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Cooling water flow (CFR: 41.5 / 45.5)

Technical Reference(s): ALR 00-055A, OFN EF-33, LO1732443 Proposed references to be provided to applicants during examination: None Learning Objective: R2/3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Other K/A: 022 K1.01 [3.5/3.7] {41.4, 5, 7}

022 A2.05 (3.1/3.5) {41.5, 43.5}

022 2.4.47 [3.4/3.7] {41.10, 43.5}

022 2.4.48 [3.5/3.8] {43.5}

076 2.4.47 [3.4/3.7] {41.10, 43.5}

076 2.4.48 [3.5/3.8] {43.5}

062 AA2.01 [2.9/3.5] {43.5}

Est. Time: 4 min.

Modified based on NRC comments. Changed to Rev. 6, 7/28/11 Answer: C

27. 022 A3.01 3 Given the following plant conditions:

A LOCA has occurred Which ONE of the following describes 1) the cooling water alignment to the Containment Coolers and 2) the speed required per cooler?

A. 1) Service Water

2) Fast B. 1) Service Water

2) Slow C. 1) ESW

2) Slow D. 1) ESW

2) Fast Justification A. Incorrect, wrong system, wrong speed B. Incorrect, wrong system, right speed C. Correct D. Incorrect, right system, wrong speed/flow The Containment Fan Coolers are finned-tube type coolers supplied by the Essential Service Water System. Refer to Figure 4: Containment Coolers. The fans normally run in Fast speed, the sequencer starts them in Slow for accident conditions The minimum required flow rate through each cooler is 1,000 gpm per cooler (2000 gpm per group) during normal operation and post-accident.

K/A Statement: Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation (CFR: 41.7 / 45.5)

Technical Reference(s): SY1302600, USAR 6.2.2.2.2.3, M-12EF02, M-12GN01 Proposed references to be provided to applicants during examination: None Learning Objective: R8 Question Source: Bank # _______

Modified Bank # _58877______

New _______

Question History: Last NRC Exam ___2007 Callaway_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 022 K4.02 [3.1/3.4] {41.7}

013 K1.03 [3.8/4.1] {41.4, 5, 7, 7, 8, 9}

Est. Time: 3 min.

Incorporated Brendan comments, changed to Rev. 3, 6/6/11 Answer: C

28. 022 AK1.02 3 Given the following plant conditions:

The unit is at 100% power

BG HCV-182, CHG HDR BACK PRESS CTRL, is closed slightly by the Reactor Operator Which ONE of the following describes RCS Makeup flow response to the above condition?

RCP Seal Charging Flow to Injection Flow Regen Hx A. Increases Decreases B. Decreases Increases C. Increases Increases D. Decreases Decreases Justification A. Correct.

B. Incorrect, seal flow increases as BG HCV-182 is closed, charging flow to regen hx decreases as BG HCV-182 is closed.

C. Incorrect, seal injection flow increases, charging flow decreases as BG HCV-182 is closed.

D. Incorrect, seal flow increases as BG HCV-182 is closed, charging flow to regen hx decreases as BG HCV-182 is closed.

Downstream of BG FE-121, a line branches from the charging header to supply the reactor coolant pump seals. RCP seal flow is controlled with BG HCV-182, Charging Header Flow Control Valve. This air-operated valve provides backpressure on the common discharge header to control the amount of water forced through the seal injection lines. Seal injection flow is set using throttle valves located in the South Mechanical penetration room. The further closed these throttle valves are the further closed BG HCV-182 must be to provide adequate seal flow. These throttle valves are set using a Tech. Spec.

Surveillance and have caused problems with CCPs maintaining PZR level at 120 GPM Letdown. BG HCV-182 is positioned from the MCB panel RL001 by adjusting a potentiometer. This valve fails open on a loss of instrument air or control power.

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Reactor Coolant Pump Makeup: Relationship of charging flow to pressure differential between charging and RCS (CFR 41.8 / 41.10 / 45.3)

Technical Reference(s): SY1300400, ALR 00-042E, M-12BG03

Proposed references to be provided to applicants during examination: None Learning Objective: R3/4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X___

10 CFR Part 55 Content:

55.41 _8, 10_

55.43 _____

Comments:

Incorporated NRC comments, changed to Rev . 3, 7/28/11 Answer: A

29. 025 AK3.02 6 Given the following plant conditions:

The plant is in MODE 4

2 RCPs are running

RCS temperature is 342°F

RCS pressure is 325 psig

PZR level is 25%

"A" train RHR is in service Which ONE of the following describes the effect on the RHR System operation and required Operator response if RCS pressure rises to 450 psig?

RHR System Effect Operator Response A. RHR Suction Relief lifts Open a PZR PORV B. RHR Suction Relief lifts Open a PZR Spray Valve C. RHR Isolates Open a PZR PORV D. RHR Isolates Open a PZR Spray Valve Justification A. Incorrect, first part correct, PORV not directed, below LTOP for 342F B. Correct.

C. Incorrect, first part see below, second part not directed D. Incorrect, first part see below, second part correct There are no closing interlocks or any automatic signals associated with Hot Leg Suction valves. These valves fail as is.

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: Isolation of RHR low-pressure piping prior to pressure increase above specified level (CFR 41.5, 41.10 / 45.6 / 45.13)

Technical Reference(s): ALR 00-049B, SY1300500, M-12EJ01, M-12BB04, GEN 00-002 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _5, 10_

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 6, 7/28/11 Answer: B

30. 026 2.4.2 6 Given the following plant conditions:

A pipe break has occurred in the "A" train Component Cooling System (CCW)

The CRS enters OFN EG-004, CCW SYSTEM MALFUNCTIONS

CCW Train "A" is supplying the Service loop

Surge Tank "A" level is 18% and DOWN FAST

Annunciator 060E, CTMT SUMP A/B LEV HI, is lit Which ONE of the following describes the Operator response to this event?

A. Trip the Reactor when Surge Tank "A" level decreases to 14%.

B. Isolate CCW Return from RCP Thermal Barriers by closing EG HV-62, to stop a leak identified on the Containment Building Return Header.

C. Stop and lockout all "A" train ECCS pumps and the Containment Spray pump due to loss of seal cooling.

D. Verify ESW makeup is automatically aligned to CCW Surge Tank "A" due to the normal makeup being inadequate.

Justification A. Correct, per OFN foldout B. Incorrect, Surge Tank level would be increasing for a Thermal Barrier leak, closing return does not affect supply C. Incorrect, Ctmt Spray pumps not placed in PTL D. Incorrect, ESW alignment is a manual action in OFN RNO, Step 3 K/A Statement: Loss of Component Cooling Water. Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7

/ 45.8)

Technical Reference(s): ALR 00-051D, OFN EG-004, LO1732414, SY1400800, M-12EG01 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 026 AK3.03 [4.0 / 4.2]

Modified based on Jane comments. Changed to Rev. 6, 7/2/11 Answer: A

31. 026 A2.07 4 Given the following plant conditions:

A LOCA has occurred

The crew has transferred suctions to the Containment Recirc. Sump

EJ LI-7, CTMT RECIRC SUMP, Train A indicates 1998' 0 decreasing

EJ LI-8, CTMT RECIRC SUMP, Train B indicates 2000' 6" decreasing

Containment Spray pump A amps are oscillating

Containment Spray pump B amps are steady Which ONE of the following describes the impact of these levels and the Operator response IAW EMG C-13, CONTROL ROOM RESPONSE TO SUMP BLOCKAGE?

Impact Response A. Recirc Sump "A" and B are low Place both Ctmt Spray pump switches in Stop B. Recirc Sump A and "B" are low Monitor both Ctmt Spray pumps for signs of cavitation C. Recirc Sump "A" ONLY is low Place "A" Ctmt Spray pump switch in Stop D. Recirc Sump "A" ONLY is low Monitor "A" Ctmt Spray pump for signs of cavitation Justification:

A. Incorrect, first part correct, wrong response, should place ONLY "A" switch in STOP. PTL is used to prevent pumps from starting with an inadequate suction or wrong plant conditions.

B. Incorrect, sump level is adequate, should place ONLY "A" switch in STOP. PTL is used to prevent pumps from starting with an inadequate suction or wrong plant conditions.

C. Correct. Meets conditions for Step 1 RNO b. for one less than 1999 7 and BOTH trains decreasing for RHR and meets foldout page criterion for securing CTMT Spray Pumps.

D. Incorrect, sump level is inadequate, should place ONLY "A" switch in STOP. PTL is used to prevent pumps from starting with an inadequate suction or wrong plant conditions. Monitoring for cavitation is not the Operator response required for the given conditions in C-13.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Technical Reference(s): EMG C-13, BD-EMG C-13, LO1732336 Proposed references to be provided to applicants during examination: None Learning Objective: R3/R4

Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Incorporated NRC comments, changed to Rev. 4 7/28/11 Answer: C

32. 027 2.2.22 5 Given the following plant conditions:

The unit is at 63% power

The Reactor Operator is responding to a Pressurizer pressure control malfunction

The plant is approaching DNBR limits

The RO recognizes that an action which is outside Technical Specifications or license conditions must be taken Which ONE of the following describes the correct Operator action IAW AP 21-003, OPERATIONS?

A. Immediately take appropriate actions necessary and inform the SM when time permits.

B. Must get agreement and obtain a peer check prior to taking action.

C. Must obtain approval from a licensed SRO prior to taking action.

D. May only take action with SM concurrence.

Justification:

A. Incorrect. Response is correct for taking immediate actions, but this is a 10CFR50.54(x, y) situation.

B. Incorrect. Correct answer for a normal procedural situation, but this is a 10CFR50.54(x, y) situation.

C. Correct. SRO approval required per 10CFR50.54(x).

D. Incorrect. Certain emergencies in the EOPs do require SM approval.

K/A Statement: Knowledge of limiting conditions for operations and safety limits. (CFR:

41.5 / 43.2 / 45.2)

Technical Reference(s): AP 21-003 (6.6), 10CFR50.54 x & y, LO1733203, TS 2.0 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __5___

55.43 __2___

Comments:

Other K/A: 027 G2.1.2 [3.0 - 4.0] {41.10, 45.13}

Modified based on NRC comments. Changed to Rev. 5, 7/28/11 Answer: C

33. 027 K2.01 5 Given the following plant conditions:

Containment Atmosphere Control Fan CGR01A has been placed in service

A total loss of offsite power has occurred Which ONE of the following describes how the loss of offsite power affects CGR01A operation?

A. De-energized due to load shed of non-vital equipment.

B. De-energized since 480v bus PG19 has lost power.

C. Energized from 480v bus NG01 and still in service.

D. Energized from 480v bus PG19 and still in service.

Justification A. Incorrect, right response, wrong action, no load shed B. Correct.

C. Incorrect, opposite response, wrong bus D. Incorrect, opposite response right bus CGR01A is powered from 480v bus PG19, this is a non-vital bus that will not have power K/A Statement: Knowledge of bus power supplies to the following: Fans (CFR: 41.7)

Technical Reference(s): SYS GR-120, SY1302800, CKL GR-120, M-12GR01, E-13GR01 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _46895______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___2008 DCPP_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 5, 7/28/11 Answer: B

34. 028 A2.01 4 Calculate the "B" Hydrogen Recombiner initial power setting using the attached curve assuming the following information:

Pre LOCA CTMT Temperature (GNU001) = 105°F

Post LOCA CTMT Pressure (GNP936) = 8.3 psig

Reference Power Plaque setting = 46.9 (REFERENCE PROVIDED)

A. 63.32 B. 64.25 C. 64.72 D. 65.19 Justification A. Incorrect, 120 @1.35 B. Correct C. Incorrect, 90 @1.38 D. Incorrect, 75 @1.39 8.3 psig + 14.7 = 23 psig, use 23 psig on reference provided to determine pressure factor (Cp), CP = 1.37 6.3.3 Record from Plant Operating Logs the Pre-LOCA Containment temperature.

Determine pressure factor (Cp), using FIGURE 1, DRY CONTAINMENT RECOMBINER CORRECTION FACTOR VS. CONTAINMENT PRESSURE.

Containment Pressure ____________ Cp (1.37)

Multiply the reference power (plaque on Recombiner Control Panels) by Cp to determine required recombiner power setting, as indicated on the power meter.

SGS01A Reference Power _46.9______ X Cp __1.37______ = __64.25_______ Power Setting

PRESSURE FACTOR (CP) 1.6 1.5 1.4 1.3 PRE-LOCA CONTAINMENT TEMPERATURE 60 °F 1.2 90 °F 1.1 120 °F 1

14 16 18 20 22 24 26 28 POST-LOCA CONTAINMENT PRESSURE (PSIA)

K/A Statement: Malfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Hydrogen recombiner power setting, determined by using plant data book (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Technical Reference(s): SYS GS-120, SY1302800 Proposed references to be provided to applicants during examination: Step 6.3.3 and Figure 1 Learning Objective: R5

Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Modified based on Jane comments. Changed to Rev. 4, 6/22/11 Answer: B

35. 028 AK2.03 6 Given the following plant conditions:

The unit is at 100% power

The controlling PZR level transmitter (BB LT-459) fails at 57% level Assuming NO Operator action, which ONE of the following describes the INITIAL effect if a turbine runback to 80% occurs?

A. Charging flow increases and actual PZR level increases.

B. Charging flow decreases and actual PZR level decreases.

C. Charging flow remains constant and actual PZR level increases.

D. Charging flow remains constant and actual PZR level decreases.

Justification:

A. Incorrect, flow remains constant and PZR level increases (RCS Temp. increases).

B. Incorrect, wrong flow response due to failure, opposite PZR level direction.PZR level initially increases as Tavg goes up, then decreases as rods and dumps drive temp. down w/o charging flow control.

C. Correct. Level indication higher than program level, transient failure prevents repositioning of flow control valve and PZR level increases (RCS Temp. increases).

D. Incorrect, right charging flow response, opposite level direction initially increases due to swell.

K/A Statement: Knowledge of the interrelations between the Pressurizer Level Control Malfunctions and the following: Controllers and positioners (CFR 41.7 / 45.7)

Technical Reference(s): OFN SB-008, BD-OFN SB-008, M-12BB02, LO1732418, SY1301000 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7___

55.43 _____

Comments:

Other K/A: 028 AA1.02 (3.4/3.4) {41.7}

011 K1.01 [3.6/3.9] {41.3, 7}

011 K3.01 [3.2/3.4] {41.7}

011 K4.04 [3.0/3.3] {41.7}

011 K6.09 [2.4/2.6] {41.7}

011 A3.02 [2.6/2.8] {41.7}

011 A3.03 [3.2/3.3] {41.7}

Est Time: 4 min Modified based on NRC comments. Changed to Rev. 6, 7/28/11 Answer: C

36. 029 A4.01 4 Given the following plant conditions:

The unit is at 50% RTP

A Containment depressurization is in progress per SYS GT-120, CONTAINMENT MINI-PURGE SYSTEM OPERATIONS Which ONE of the following describes the effect on the operating purge equipment following a loss of NK01?

A. Train B dampers go closed, fans stay running.

B. Train B dampers go closed, fans stop.

C. Train A dampers go closed, fans stay running.

D. Train A dampers go closed, fans stop.

Justification A. Incorrect, "A" train dampers close, "B" Train dampers remain open. Fans unaffected by NK01 loss B. Incorrect, "A" train dampers close, "B" Train dampers remain open. Fans unaffected by NK01 loss C. Correct. "A" train dampers close due to NK01 loss, fans remain running D. Incorrect, "A" train dampers close due to NK01 loss, fans remain running "A" Train GT HZ-5 & 12 NK01 power supply. If NK01 lost "A" train dampers close "B" Train GT HZ-4 & 11 NK04 power supply. If NK01 lost "B" train dampers unaffected CGT02 power supply is PG25H - not affected by NK01/4 loss SGT02 power supply is PG19N - not affected by NK01/4 loss Mini-purge supply & Ctmt exhaust isolation dampers GT HZ-4 & 11 close. Ctmt mini-purge air supply unit, SGT02, remains running & Ctmt mini-purge exhaust fan CGT02, remains running.

Mini-purge supply & Ctmt exhaust isolation dampers GT HZ-5 & 12 close. Ctmt mini-purge air supply unit, SGT02, remains running & Ctmt mini-purge exhaust fan CGT02, remains running.

K/A Statement: Ability to manually operate and/or monitor in the control room:

Containment purge flow rate (CFR: 41.7 / 45.5 to 45.8)

Technical Reference(s): OFN NK-020, LO1732430, SY1302800 (R7), M-12GT01 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 029 K2.04 [2.1/2.3] {41.7}

063 K3.02 [3.5/3.7] {41.7}

058 AA2.02 [3.5/3.9] {43.5}

058 AK3.02 [4.0/4.2] {41.5, 10}

Est Time: 4 min Modified based on NRC comments. Changed to Rev. 4, 7/28/11 Answer: C

37. 034 2.2.12 5 Given the following plant conditions:

The unit is in MODE 6

"A" train RHR is in service

B train RHR is in standby

Refueling Pool level is 23' above the fuel on BB LI-53A, RCS LEVEL WR COLD CAL LOOP 4

Control Rod Shaft Unlatching is the next evolution

All Refueling personnel are on station Which ONE of the following describes whether or not unlatching can begin and why or why not?

A. Yes, level and RHR are adequate.

B. Yes, ONLY proper level is required.

C. No, RHR loops are inadequate.

D. No, level is inadequate.

Justification A. Correct, level is adequate, RHR (one train operating & in-service) adequate.

B. Incorrect, water level is adequate and RHR pump operation is required.

C. Incorrect, RHR loops are adequate for button unlatching evolution.

D. Incorrect, level is adequate, must be 23' above flange to move fuel.

STS CR-002, Att. A, P 30/44 Data Sheet 1 Step A23 K/A Statement: Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)

Technical Reference(s): LO1732109, GEN 00-009, STS CR-002, TR 3.9.7, TS 3.9.6 Proposed references to be provided to applicants during examination: None Learning Objective: R5 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __10_

55.43 _____

Comments:

Modified based on Jane comments. Changed to Rev. 5, 6/18/11 Answer: A

38. 035 A3.01 5 Given the following plant conditions:

The unit is operating at 100% power

The "A" SG controlling pressure transmitter fails HIGH Which ONE of the following describes the INITIAL feedwater flow response and the INITIAL Operator action that is procedurally directed?

Initial Response Operator Action A. Feedwater flow would decrease. Place the "A" MFRV in MANUAL and restore level to program.

B. Feedwater flow would increase. Place the "A" MFRV in MANUAL and restore level to program.

C. Feedwater flow would decrease. Leave the "A" MFRV in AUTO because SG level is the dominant control signal and will restore feed flow to match steam flow.

D. Feedwater flow would increase. Leave the "A" MFRV in AUTO because SG level is the dominant control signal and will restore feed flow to match steam signal.

JUSTIFICATION:

A. Incorrect; Compensating pressure transmitter failing high will cause steam flow to indicate high, which will cause an increase in feed flow. The operator response is correct.

B. Correct. Compensating pressure transmitter failing high will cause steam flow to indicate high, which will cause an increase in feed flow. The operator response is correct.

C. Incorrect; Compensating pressure transmitter failing high will cause steam flow to indicate high, which will cause an increase in feed flow. The operator response is incorrect based on OFN. C and D distractor are valid because actual SG level is the dominant control signal and if the LCV were left in auto, it should match FF/SF after some time delay, however OFN provide direction for transferring to Man control.

D. Incorrect; Compensating pressure transmitter failing high will cause steam flow to indicate high, which will cause an increase in feed flow. The operator response is incorrect based on OFN K/A Statement: Ability to monitor automatic operation of the S/G including: S/G water level control. (CFR: 41.7 / 45.5)

Technical Reference(s): OFN SB-008 Att. C, LO173241, BD-OFN SB-008, ALR 00-111C

Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 035 A2.04 [3.6/3.8] {41.5, 43.5}

035 A2.03 [3.4/3.6] {41.5, 43.5}

035 A4.01 [3.7/3.6] {41.7}

054 AA2.08 [2.9/3.3] {43.5}

Est. Time: 2 min.

Modified based on Jeff comments. Changed to Rev. 5, 8/1/11 Answer: B

39. 037 AA1.04 1 The plant is at 100% power when a Hi-Hi radiation alarm is received on GE RE-92, Main Condenser Vacuum Pump Vent Radiation Monitor.

What reset action, if any, is necessary to open SG DNSTRM OUTSIDE CTMT SAMPLE ISO, valves BM HV-5, 6, 7 and 8?

A. No reset required.

B. Reset BSPIS.

C. Reset AFAS-M.

D. Reset SGBSIS.

Justification A. Correct.

B. Incorrect, once the rad monitor signal clears BSPIS has to be reset, does not control these valves C. Incorrect, rad monitor does not cause AFAS-M signal D. Incorrect, cannot individually reset SGBSIS, done vis AFAS-M K/A Statement: Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: Condensate air ejector exhaust radiation monitor and failure indicator (CFR 41.7 / 45.5 / 45.6)

Technical Reference(s): OFN BB-07A, SY1407300, LO1732436 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _16430______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _X__

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Replaced K/A based on NRC comments, 7/29/11 Answer: A

40. 038 EA1.15 6 Given the following plant conditions:

A steam generator tube rupture has occurred on Loop "D" SG

The TDAFW pump is running, providing 700 gpm flow

Both MDAFW pumps are running, providing 250 gpm per pump

EMG E-3, STEAM GENERATOR TUBE RUPTURE, has been implemented by the operating crew

The cooldown to target T/C of 484°F is in progress

CST is currently at 90% level Which ONE of the following is the expected usable CST volume AFTER cooldown has been completed?

Assume the following:

Cooldown will take 25 minutes

AFW pump discharge flow remains constant during the cooldown (REFERENCE PROVIDED)

A. 362,662 gallons B. 377,317 gallons C. 384,989 gallons D. 407,317 gallons Justification A. Incorrect, 80% level B. Correct, See below C. Incorrect. 85% level.

D. Incorrect, 90% level CST @ 90% = 407,317 gal (25 min) (700 + 250 + 250) = 30,000 gal CST (after) = 377,317 gal K/A Statement: Ability to operate and monitor the following as they apply to a SGTR:

AFW source level and capacity (chart) (CFR 41.7 / 45.5 / 45.6)

Technical Reference(s): EMG E-3, TAP01, BD-EMG E-3, LO1732325 Proposed references to be provided to applicants during examination: TAP01 Learning Objective: R6 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on Jane comments. Changed to Rev. 6, 7/2/11 Answer: B

41. 039 K5.08 4 Given the following plant conditions:

Unit is at 20143 MWD/MTU

The unit is critical at 10-10 amps with all systems and components normal

S/G Atmospheric Relief Valve AB PV-4 fails OPEN Which ONE of the following describes the effect the failure will have on Reactor power and RCS pressure?

Power Pressure A. Increase Increase B. Decrease Decrease C. Decrease Increase D. Increase Decrease Justification:

A. Incorrect. Pressure will not increase since Tavg decrease.

B. Incorrect. Power will increase since MTC is large and negative.

C. Incorrect. Temperature will decrease since heat removal > heat production - pressure will decrease, power will increase.

D. Correct. Temperature decrease will insert positive reactivity. Thus power increase. Pressure will decrease since Tc decreases causing pressurizer level (due to density changes) to lower causing pressure to decrease.

K/A Statement: Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity (CFR: 441.5 / 45.7)

Technical Reference(s): LO1610500, LO1732451, OFN AB-041 Proposed references to be provided to applicants during examination: None Learning Objective: R3/R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Other K/A: 039 K3.05 [3.6/3.7] {41.7}

192008 K1.10 [3.3/3.4] {41.1}

Est Time: 4 min Incorporated NRC comments. Changed to Rev. 4, 7/29/11 Answer: D

42. 054 AK1.02 5 Given the following plant conditions:

The crew is responding to a loss of secondary heat sink event caused by a loss of Normal and Auxiliary Feedwater

SG wide range levels are all 5%

SG pressures are all 1060 psig

Bleed and Feed cooling has been initiated per EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK

RCS hot leg temps are 556°F

CETC are 586°F and stable

Containment pressure is 3.6 psig

Containment radiation is 1 R/hr Which ONE of the following describes the INITIAL AFW flowrate and why?

AFW Flow is to be . . .

A. above 35,000 lbm/hr to ONE SG to prevent an excessive cooldown B. below 35,000 lbm/hr to ONE SG to prevent an excessive cooldown C. below 35,000 lbm/hr to ALL SGs to prevent an uncontrolled S/G pressure decrease D. above 35,000 lbm/hr to ALL SGs to prevent an uncontrolled S/G pressure decrease Justification:

A. Incorrect. Adverse Feed/bleed number from foldout page, not termination level, purpose is cooldown not thermal stress B. Correct. Basis for Step 41 see below C. Incorrect, basis for controlling cooldown rate, purpose is cooldown not pressure decrease D. Incorrect, basis for controlling cooldown rate, purpose is cooldown not pressure decrease Need to determine adverse ctmt or not (No)

When feedwater flow is restored to a steam generator, high flow rates may cause RCS temperature to rapidly decrease. This is especially applicable when main feedwater is used to restore steam generator inventory. Although rapid cooldown of the RCS can produce temperature induced stress with the potential for loss of vessel integrity under high pressure conditions, it is likely that the vessel would have already been subjected to a rapid cooldown prior to reestablishing feedwater flow, due to the injection of relatively cold SI water during bleed and feed. After reestablishing feedwater flow, minimizing the rate of RCS cooldown is more beneficial for operator controllability of the plant (by minimizing coolant shrinkage and pressure transients) than integrity of the vessel. Guidance for preventing excessive RCS cooling is included in FR-H1 to enhance operator control of the transient.

It should be noted that increasing the time to termination of bleed and feed by minimizing feedwater flow is not a concern, since the greatest threat to vessel integrity has already occurred and the containment

atmosphere has already been subjected to a mass release. Once feedwater flow is restored, its effectiveness as a heat sink for the RCS can be determined by decreasing RCS temperatures (cold legs, hot legs, core exit thermocouples) with increasing steam generator level. These parameters should be monitored when controlling feedwater flow, to prevent excessive RCS cooldown. After feedwater flow has been restored, feedwater flow should be maintained on scale and controlled as necessary to maintain increasing steam generator level and slowly decreasing RCS temperatures.

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G (CFR 41.8 / 41.10 / 45.3)

Technical Reference(s): EMG FR-H1, BD-EMG FR-H1, LO1732346 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _8, 10_

55.43 _____

Comments:

Modified based on Jane comments. Changed to Rev. 5, 7/6/11 Answer: B

43. 051 2.4.31 3 Given the following plant conditions:

The unit is at 95% power

AC PI-204, LP TURBINE A EXHAUST PRESSURE indicated 5.2 HgA and slowly rising

Annunciator 116B, COND A VAC LO, has alarmed

The crew has entered ALR 00-116B Which ONE of the following describes appropriate crew response if condition is not met?

A. Check exhaust pressure <15.1 inches HgA Manually trip the turbine and enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION B. Check exhaust pressure <5.0 inches HgA Manually trip the reactor and enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION C. Check exhaust pressure <5.0 inches HgA Address condenser vacuum IAW OFN AF-025, UNIT LIMITATIONS D. Check exhaust pressure <15.1 inches HgA Address condenser vacuum IAW OFN AF-025, UNIT LIMITATIONS Justification A. Incorrect, reactor is tripped not the turbine, reactor trip is only for < 30% power, 15.1 is for FW pump turbines.

B. Incorrect, reactor trip is only for < 30% power.

C. Correct. Step F3 b and RNO.

D. Incorrect, 15.1 is for FW Pump Turbines.

K/A Statement: Knowledge of annunciator alarms, indications, or response procedures.

(CFR: 41.10 / 45.3)

Technical Reference(s): ALR 00-116B, OFN AF-025, SY1505500, LO1732435 Proposed references to be provided to applicants during examination: None Learning Objective: R5/R3 Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10__

55.43 _____

Comments:

Incorporated NRC comments, changed to Rev. 3 8/1/11 Answer: C

44. 040 AK1.03 4 Given the following plant conditions:

The unit is in MODE 3

The operating crew observes abnormal conditions on:

RCS temperature

RCS pressure

Pressurizer level

Containment pressure Which ONE of the following parameters is used to differentiate between the early stages of a moderately sized RCS Pressure Boundary Leak or a moderately sized Secondary Steam Break?

A. RCS Pressure B. Pressurizer Level C. RCS Temperature D. Containment Pressure Justification:

A. Incorrect. RCS pressure would reduce during both events.

B. Incorrect. Pressurizer level would decrease during both events, one is a shrink the other is a loss of inventory.

C. Correct. RCS temperature should follow faulted S/G pressure, temperature does not decrease on an RCS break.

D. Incorrect. Containment pressure would increase during both events.

K/A Statement: Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: RCS shrink and consequent depressurization (CFR 41.8 / 41.10 / 45.3)

Technical Reference(s): BD-EMG E-1, BD-EMG E-0, LO1610500, BD-ERG E-1 (Figs. 1, 2, 3, 4, 15, 16, 17, 18, 19), LO1732320 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _8, 10_

55.43 _____

Comments:

Est Time: 3 min.

Modified based on Brendan comments. Changed to Rev. 4, 6/6/11 Answer: C

45. 055 EA1.01 7 Given the following plant conditions:

The plant experienced a Station Blackout

The crew entered EMG C-0, LOSS OF ALL AC

The TSC diesel has failed to start and maintenance personnel are troubleshooting

After 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months the diesel has not been started Which ONE of the following describes where the Operator would be able to monitor core temperatures?

A. NPIS terminal B. Reference Junctions C. Incore Thermocouple Cabinet D. Hot Leg RTD meters on the MCB Justification:

A. Incorrect. NPIS terminal will have no power after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Incorrect. Not accessible during power ops.

C. Correct. The analysis for not using the Hot Leg RTD meters in BD-EMG C-0, refers to BD-EMG F-0.

BD-EMG F-0 does not allow Hot Leg RTD meters to be used because they are unreliable.

D. Incorrect. EMG bases document and F-0 requires at least 5 thermocouples; one near the center of the core and the hottest in each quadrant, not Thot. Hot Leg RTD meters not used because they are unreliable.

K/A Statement: Ability to operate and monitor the following as they apply to a Station Blackout: In-core thermocouple temperatures (CFR 41.7 / 45.5 / 45.6)

Technical Reference(s): EMG C-0, BD-EMG F-0, BD-EMG C-0, SY1301700, SY1408301 Proposed references to be provided to applicants during examination: None Learning Objective: R4R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated Jane comments. Changed to Rev. 7, 7/2/11 Answer: C

46. 059 K4.17 5 Which ONE of the following describes the design feature to limit feed water flow following a reactor trip?

A. High level (68%) in 2/4 steam generators.

B. Lo-Lo level (25.3%) in 2/4 steam generators.

C. P-4 signal with 2/4 low Tavg signals.

D. Both main feedwater pumps trip.

Justification A. Incorrect, level setpoint is 78% (carryover)

B. Incorrect, setpoint is 23.5% (heat sink)

C. Correct.

D. Incorrect, AFP start signal (anticipatory)

K/A Statement: Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Increased feedwater flow following a reactor trip (CFR: 41.7)

Technical Reference(s): SY1505900, SY1505902, M-744-00030, M-12AE01, M-761-00098, M-744-00022, TS Basis 3.3.2 for P4 Proposed references to be provided to applicants during examination: None Learning Objective: R6 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 007 EA1.02 (3.8-3.7) 013 K1.15 (3.4-3.8)

Incorporated NRC comments, changed to Rev. 5, 7/29/11 Answer: C

47. 061 AK3.02 4 Given the following plant conditions:

The unit is steady-state at 75% power

Annunciator 00-062A, AREA RAD HIHI, has just alarmed

The unexpected alarm SD RI-29, HOT MACHINE SHOP, is valid and reads 65 mr/hr In addition to determining the area affected, which ONE of the following RO actions must be performed?

A. Direct Health Physics to survey the Hot Machine Shop and evacuate the entire radiologically controlled area.

B. Dispatch an Auxiliary Operator to investigate the Hot Machine Shop high radiation alarm.

C. Announce over the PA system to evacuate personnel from the Hot Machine Shop.

D. Notify the SM to evacuate the Auxiliary Building and classify the event.

Justification:

A. Incorrect. HP is notified of alarm only, no evacuation of the entire RCA.

B. Incorrect. Dispatching an AO would unnecessarily expose him to high radiation. Not required by ALR.

C. Correct. ALR 00-062A.

D. Incorrect. SM permission not required to evacuate the AB; classification of the event not required by ALR 00-062A and the EALs.

K/A Statement: Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system (CFR 41.5,41.10 / 45.6 / 45.13)

Technical Reference(s): ALR 00-062A, SY1407200 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _5, 10_

55.43 _____

Comments:

Other K/A: 2.4.10 [3.0/3.1] {41.10, 43.5}

Est Time: 2 min Incorporated NRC comments, changed to Rev. 4, 7/29/11 Answer: C

48. 061 K6.01 5 Given the following plant conditions:

TDAFW Pump testing is in progress per STS AL-103, TDAFW PUMP INSERVICE PUMP TEST

The BOP opens FC HV-312A, Trip and Throttle Valve, and the TDAFW Pump speed stabilizes at 1100 RPM and will NOT increase Which ONE of the following describes the probable cause of the TDAFW Pump failing to achieve the required speed?

A. Loss of 48 VDC Control Power.

B. Loss of hydraulic pressure to the governor valve.

C. Failure of the oil pump to CLOSE the governor valve.

D. Failure of Limit Switch 6 on FC HV-312A to make-up.

Justification A. Incorrect, valves would fail close on loss of power. Would result in overspeed B. Incorrect, GV is normally open, oil pressure closes GV.

C. Incorrect, opposite response. Would result in overspeed D. Correct. LS-6 starts the ramp generator to increase from 1100 rpm K/A Statement: Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners (CFR: 41.7 / 45.7)

Technical Reference(s): STS AL-103, SY1406100, E-13FC24, M-12FC02 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated NRC comments, changed to Rev. 5, 8/1/11 Answer: D

49. 062 2.1.28 4 Given the following plant conditions:

The plant is operating at 100% power

ESW Pump Discharge pressure indicates 106 psig on both trains

Strainer D/P is 4 psid

ESW load temperatures are rising

A cold front has quickly moved through the area

Lake temperature is 33°F

Air temperature is 21°F

Time is currently 1830 hours0.0212 days 0.508 hours 0.00303 weeks 6.96315e-4 months

The crew has entered OFN EF-033, LOSS OF ESSENTIAL SERVICE WATER Which ONE of the following describes the cause of the above conditions and the component(s) available to restore load cooling?

A. Trash Rack blocked ESW Warming Line B. Trash Rack blocked ESW Pump Recirculation lines C. Clogged Pump Strainer ESW Warming Line D. Clogged Pump Strainer ESW Pump Recirculation lines Justification A. Correct.

B. Incorrect, recirc lines are for pump protection not ice prevention C. Incorrect, strainer D/P is within limits, correct component D. Incorrect, strainer D/P is within limits, wrong component K/A Statement: Knowledge of the purpose and function of major system components and controls. (CFR: 41.7)

Technical Reference(s): ALR 00-055A, ALR 00-055B, OFN EF-033, SYS EF-205, SY1006002 Proposed references to be provided to applicants during examination: None Learning Objective: R4/R6

Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Incorporated NRC comments. Changed to Rev. 4, 7/29/11 Answer: A

50. 062 A1.03 6 Given the following plant conditions:

The unit is at 100% RTP

Operators are swapping NN04 power supply to XNN05 Sola Transformer Which ONE of the following describes the effect on the plant of making this transfer before swapping control channels in SYS NN-200, TRANSFERRING NN BUSSES BETWEEN POWER SOURCES?

A. SG level low-low reactor trip B. Pressurizer level decreasing C. VCT swapover to the RWST D. BG FCV-462, NCP Flow Control Valve, fails open JUSTIFICATION:

A. Incorrect, with the loss of NN04 this input is not a controlling channel.

B. Incorrect, with the loss of NN04 this input is not a controlling channel.

C. Correct. BG LT112 failing low would cause the RWST to Charging pump suction valve BN LV112D to open and VCT outlet valve BG LCV112A to close causing a boration. Power restoration alone would not realign these valves requiring Operator action. Aligning the RWST would provide a constant boration to the RCS providing negative reactivity, lowering RCS temperature.

D. Incorrect, with the loss of NN04 this component is not affected.

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies (CFR:

41.5 / 45.5)

Technical Reference(s): OFN NN-021, SYS NN-200, E-11010, LO1732431, SY1506300, SY1300400 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Other K/A: 062 A2.01 [3.4/3.9] {41.5, 43.5}

062 K3.01 [3.5/3.9] {41.7}

057 AA1.01 [3.7/3.7] {41.7}

Est. Time: 3 min Incorporated NRC/validation comments, changed to Rev. 6, 8/1/11 Answer: C

51. 063 2.1.31 4 Given the following sequence of events:

Plant is at 100% power all systems normally aligned

Loss of Vital DC Bus NK04 occurs Which ONE of the following is the expected indication that correctly reflects current plant lineup?

A. Both MDAFW pumps start and no indication available for B train, and all MFRV and bypass valves fail closed.

B. "B" train P-14 solenoids on all MFRV and bypass valves deenergize causing valves to fail closed.

C. Steam line pressure transmitters to "B" and "D" ARV's deenergize and "B" and "D" ARVs indicate closed.

D. Turbine Driven AFW pump starts and AFW flow can be throttled to all S/Gs.

JUSTIFICATION:

A. Incorrect - A MDAFW pumps start ONLY and MFRV and Bypass valves fail closed when the FWI solenoid powered from NK04 is de-energized closing the valves, indication lost from NK04 UV.

B. Correct - 125VDC power is required to maintain the P-14 solenoids open. Closing the solenoid removes control air from the valve and it fails shut.

C. Incorrect - D ARV pressure transmitter and controller is de-energized ONLY D. Incorrect - A MDAFWP starts and the TDAFWP starts when SG level shrinks below 23.5% in more than 1 SG, A Train AFW smart valves can be throttled, however flow is unaffected because the TDAFWP discharge valves are failed full open in this condition.

K/A Statement: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. (CFR: 41.10 / 45.12)

Technical Reference(s): E-13AE06, E-13AE07, E-13RL03, M-12AE01, SY1505900, OFN NK-020 Proposed references to be provided to applicants during examination: None Learning Objective: R10/R5 Question Source: Bank # _22283______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10__

55.43 _____

Comments:

Incorporated NRC comments. Changed to Rev. 4 7/29/11 Answer: B

52. 063 K3.02 4 Which ONE of the following describes the effects of a loss of DC control power to 4160 VAC breaker NB0112, NB01 MN FDR BKR FROM XNB01? (Assume that this breaker is the only component affected by this loss of DC power.)

A. Breaker will fail in its current position and can be opened, but not closed, from the MCB.

B. Breaker will trip automatically and can be closed, but not opened from the MCB.

C. Breaker will fail in its current position and cannot be tripped or closed from the MCB.

D. Breaker will trip automatically and cannot be closed from the MCB.

Justification A. Incorrect, breaker cannot be opened from the MCB B. Incorrect, no operations from the MCB C. Correct, OFN NK-020 D. Incorrect, stays in its current position K/A Statement: Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: Components using DC control power (CFR: 41.7 /

45.6)

Technical Reference(s): SYS NK-131, OFN NK-020, E-11NK01, E-13NB12, E-13NB13, LO1732430 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _18590______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A: 058 AA2.03 [3.5/3.9] {43.5}

012 K4.07 [3.0/3.2] {41.7}

Modified based on NRC comments. Changed to Rev. 4, 7/29/11

Answer: C

53. 064 K6.08 7 Given the following plant conditions:

The plant is in MODE 1

STS KJ-011A, DG NE01 24 HOUR RUN, is in progress

It started at 0830

EDG Fuel Oil Storage Tank, TJE01A, level indicates as follows:

0800 90%

1200 87%

1600 84%

2000 81.5%

2400 78.5%

Annunciator 089B, DG FUEL TK A LEV LO, alarmed at ~1145 Which ONE of the following describes the correct Technical Specification and the LATEST entry time?

(REFERENCE PROVIDED)

A. TS 3.8.1, AC Sources - Operating, 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months B. TS 3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air, 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months C. TS 3.8.1, AC Sources - Operating, 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months D. TS 3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air, 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months Justification A. Incorrect, wrong LCO, approximately 80%, or 85,300 gallons is minimum TS required level, may consider EDG inoperable B. Incorrect, wrong time approximately 80%, or 85,300 gallons is minimum TS required level C. Incorrect, wrong LCO, may consider EDG inoperable.

D. Correct. Approximately 80%, or 85,300 gallons is minimum TS required level.

~660 gals/hr at full load 90% - 95,236 87.4% is alarm ~93,150 gals, ALR initiated action to fill tank 87% - 92,596 84% - 89,956 81.5% - 87,316 79.5% - 85,336, sometime around 2330 reach TS limit 78.5% - 84,676 ALR NOTE At 80% level, the diesel generator may continue to run for approximately 7 days at rated capacity before running out of fuel.

K/A Statement: Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks

Technical Reference(s): TS 3.8.3, SR 3.8.3.1, WCRE-03, TJE01A & B, SY1406400, ALR 00-089B Proposed references to be provided to applicants during examination: TJE01A & B, TS 3.8.1/3.8.3 Learning Objective: R10 Question Source: Bank # _______

Modified Bank # _58957______

New _______

Question History: Last NRC Exam ___2007 Callaway_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/29/11 Answer: D

54. 069 AK1.01 4 Which ONE of the following transients is ANALYZED to result in the highest Containment pressure AND the greatest leakage out of Containment?

Highest Pressure Greatest Total Leakage A. Design basis LOCA Design basis LOCA B. Design basis LOCA Design basis Steam Line Break inside Containment C. Design basis Steam Line Design basis Steam Line Break inside Containment Break inside Containment D. Design basis Steam Line Design basis LOCA Break inside Containment Justification A. Incorrect, most energy and volume of liquid, highest peak pressure, most leakage, opposite responses B. Incorrect, small leak size limits containment pressure C. Incorrect, first part correct, limited pressure rise, less leakage D. Correct, large spike in pressure, but limited amount of volume K/A Statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Containment Integrity: Effect of pressure on leak rate (CFR 41.8 /

41.10 / 45.3)

Technical Reference(s): USAR, Figures 6.2.1-1, 6.2.1-2, 6.2.1-81, 6.2.1-82, LO1610722, LO1610720, LO1610721 Proposed references to be provided to applicants during examination: None Learning Objective: R4/R5/R7 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _8, 10_

55.43 _____

Comments:

Incorporated NRC comments, changed to Rev. 4, 7/29/11 Answer: D

55. 073 2.1.30 6 Given the following plant conditions:

The unit is at 75% power

A spike occurs on Radiation monitor GT RE-21B, Unit Vent Wide Range Gaseous Monitor (WRGM) and peaked at 2 x 105 Ci/cm3

Radiation monitor GT RE-21B is now reading 2.5 x 10-1 Ci/cm3.

Which ONE of the following describes the operation of GT RE-21A, Unit Vent Effluent Monitor?

A. A valid reading is available on the Unit Vent Effluent Monitor GT RE-21A - Iodine channel.

B. Unit Vent Effluent Monitor GT RE-21A - Iodine channel, readings are valid because GT RE-21B are now normal.

C. A valid reading is available on the Unit Vent Effluent Monitor GT RE-21A -

Particulate channel.

D. Unit Vent Effluent Monitor GT RE-21A channels are no longer valid. The Accident Isolate mode must be reset locally.

Justification A. Incorrect, effluent is NOT being monitored at this point B. Incorrect, GT RE-21A remains in Accident Isolate Mode (purge air from room) until reset at the monitor. GT RE-21B decreasing radioactivity will not automatically restore GT RE-21A C. Incorrect, effluent is NOT being monitored at this point D. Correct, see step 4.8 of SYS SP-121 SY1407300 The function of the WRGM is to detect, indicate and alarm the presence of gaseous radioactivity in their respective airborne sample streams during normal and accident conditions.

The integral components of each WRGM are a sample conditioning skid, sample detection skid, an RM-80, and Control Room electronics.

GT RE-21B:

Sample detection skid: The sample detection skid consists of two lead shielded detector assemblies.

One assembly houses the low range radioactivity (high-flow) beta scintillation detector. The other assembly houses the mid-high range radioactivity (low-flow) cadmium telluride (CdTe), chlorine-doped, solid-state sensor detectors. Each CdTe detector output is coupled to a preamplifier in the detector assembly.

This skid also contains a power controller for the WRGM system, two pumps (low and high range sample pumps), valves, flow transducers, flow control valves, and lead shielded electronics. The sample detection skid, under the control of the RM-80, extracts a gas sample from the effluent stack that is representative of the effluent conditions. The low range sample line uses a 1/4 HP, 120VAC vacuum pump for sample flow. The mid/high range sample line uses a 1/40 HP, 120VAC vacuum pump for sample flow. Upon exiting the detector skid, the sample returns to the effluent plenum exhaust.

WRGM detector(s) total range is 1E-7 to 1E+5 mCi/cm3. During normal operation only the low radiation range path is used, and the mid/high radiation range path is shutdown. The mid/high radiation range path is automatically placed into operation when radioactivity in the sample meets a predetermined level. As the radioactivity level continues to increase, the low radiation range path is shutdown at a predetermined level.

A placard on the WRGM states that GT RE-21A will go into the Accident Isolate Mode of operation when the mid-high range pump is ON.

WRGM Control Room electronics:

The Control Room electronics of each WRGM are located on panel SP010, located the behind the Control Room desk. The assembly contains an RM-23 for control of the RM-80. Also, in the same assembly, are controls for the sample conditioning skid filters and grab sample collection Unit Vent effluent monitor (a Particulate Iodine monitor), GT RE-21A, has an Accident Isolation mode of operation. Upon receipt of an Alert gaseous radioactivity alarm from its WRGM (GT RE-21B), the PI monitor airborne sample is drawn from the room its located in (purge), rather than the effluent stack it monitors.

The Alert gaseous radioactivity alarm causes the effluent stack sample inlet motor operated valve to close, and the purge air motor operated valve to open. The Accident Isolation mode of operation is indicated at the monitor by a lit red light. The Control Room does not have indication at the RM-23 (at the SP010 panel) that the monitor has gone into the Accident Isolate mode. The RM-11R will indicate gray for the PI monitor and indicates the monitor is in purge mode, but no alarms are generated. The Accident Isolate mode of operation protects the particulate and iodine detectors in GT RE-21A from exposure to high levels of radioactivity. Once it is in Accident Isolate mode, the Unit Vent Effluent is not being monitored for particulate and iodine radioactivity.

K/A Statement: Ability to locate and operate components, including local controls. (CFR:

41.7 / 45.7)

Technical Reference(s): SY1407300, SYS SP-121 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _60037______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on Jane/NRC comments. Changed to Rev. 6, 7/29/11

Answer: D

56. 073 K1.01 3 Given the following plant conditions:

The unit is at 100% power, EOL

Fuel handling is in progress for outage preps

A depleted fuel assembly is dropped

Gas bubbles are seen rising to the surface Which ONE of the following describes the radiation monitor in alarm, and the status of the Emergency Exhaust fans and Fuel Building Supply fans?

Radiation Monitor Emergency Fuel Building in Alarm Exhaust Fans Supply Fan A. GG RE-27 Running Stopped B. GG RE-27 Stopped Running C. GT RE-22 Running Stopped D. GT RE-22 Stopped Running Justification A. Correct: RE-27 or 28 will initiate action B. Incorrect: Right monitor, opposite fans status C. Incorrect: Wrong process radiation monitor 22/33 are for Ctmt Purge, GG RE-27 and GG RE-28 are the process radiation monitors for the Fuel Building. Correct fan response D. Incorrect: Wrong process radiation monitor 22/33 are for Ctmt Purge, GG RE-27 and GG RE-28 are the process radiation monitors for the Fuel Building. Opposite fans response.

K/A Statement: Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Technical Reference(s): ALR 00-062D, SY1301301 Proposed references to be provided to applicants during examination: None Learning Objective: R2/R3 Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _3, 4_

55.43 _____

Comments:

Incorporated Jane comments. Changed to Rev. 3, 7/2/11 Answer: A

57. 076 A1.02 3 Given the following plant conditions:

Plant startup in progress

Lake temperature 33°F

Normal system alignments Which ONE of the following describes temperature control to prevent exceeding design limits in the Turbine Building Closed Cooling Water System (EB)?

A. Service Water flow through both heat exchangers are controlled by EA HV-030, Turbine Bldg CLCW HX 1A/1B Service Water Return.

B. CLCW flow through heat exchanger EB01A is controlled using the CLCW HX Outlet isolation, EB V-008.

C. The heat exchanger bypass valve EB TV-009, CLCW Temp Control Valve, bypasses Service Water around the in-service heat exchanger.

D. The heat exchanger bypass valve EB TV-009, CLCW Temp Control Valve, bypasses Closed Cooling Water around the in-service heat exchanger.

Justification:

A. Incorrect: EA HV-030 is full open isolation valve under normal alignment and fails open on loss of air.

B. Incorrect: Wrong outlet valve for "A" Hx, open to allow EB TV-009 to control temperature.

C. Incorrect: The HX bypass valve bypasses CLCW around the HX for temperature control and can bypass up to 90% of full flow.

D. Correct Objective 1 - SY1507400 Heat is transferred to service water via the Closed Cooling Water heat exchanger. Service water temperature ranges from a maximum expected 90ºF to 33°F and is supplied to the Closed Cooling Water heat exchangers. The Closed Cooling Water System provides a continuous supply of water and corrosion-inhibitor, at a maximum expected temperature of 105ºF, to cool Turbine Building loads.

Objective 2 - SY1507400 An 8-inch bypass line around the heat exchangers contains a temperature flow control valve, EB TV-

9. This valve controls heat exchanger bypass flow to maintain the desired closed cooling water system outlet temperature via temperature indicating controller, EB TIC-9. As outlet temperature increases, the valve will reduce bypass flow to increase closed cooling water flow through the heat exchanger.

Closed Cooling Water heat exchangers are designed for continuous operation at a maximum closed cooling water flow rate of approximately 930 gpm. Even at full normal system flow and maximum service water temperature, some closed cooling water bypasses the heat exchanger. During low service water temperature conditions, 90% of the closed cooling water flow will bypass the heat exchangers.

K/A Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures. (CFR: 41.5 / 45.5)

Technical Reference(s): SYS EB-120 section 6.1, M-12EB01, SY1507400 Proposed references to be provided to applicants during examination: None Learning Objective: R1/R2 Question Source: Bank # _______

Modified Bank # _58975______

New _______

Question History: Last NRC Exam ___2007 Callaway_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __5__

55.43 _____

Comments:

Modified based on Lee/Jane comments. Changed to Rev. 3, 3/22/11 Answer: D

58. 077 AA2.09 1 Given the following plant conditions:

The plant is at 100% power

DG A has been paralleled with bus NB01 and is carrying 5.8 MWe of load in accordance with STS KJ-015A, MANUAL/AUTO FAST START SYNCHRONIZATION AND LOADING OF EDG NE01

ANN 19E, NB01 BUS DGRD VOLT is lit

55 seconds later, Brkr 13-48, NORMAL FEEDER FOR XNB01, opens Which ONE of the following describes the status of the A Train Safeguards Power system with NO operator actions?

A. NB01 Emergency Supply Breaker will OPEN, NE01 will stop.

B. NB01 Normal Feeder Breaker will remain CLOSED, NE01 will stop.

C. NB01 Emergency Supply Breaker will OPEN, NE01 will remain running.

D. NB01 Normal Feeder Breaker will remain CLOSED, NE01 will remain running.

Justification A. Incorrect, breaker will remain closed, D/G will not stop, last part incorrect.

B. Incorrect, right breaker position, D/G will not stop, LOCA seq will not start.

C. Incorrect, right breaker position, right NE01 status, LOCA seq will not start D. Correct.

Comments: ALR 00-019E states that if the EDG is in parallel then to stop parallel operation. OFN NB-042, LOSS OF OFFSITE POWER TO NB01 (NB02) WITH EDG PARALLELED requires removing power from the switchyard first and then opening the EDG breaker to cause the UV on the bus for emergency operation.

K/A Statement: Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Operational status of emergency diesel generators (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

Technical Reference(s): OFN NB-042, LO1732444 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _45856______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___2009 Callaway_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Replaced K/A based on NRC comments, 7/29/11 Answer: D

59. 078 A3.01 5 Given the following plant conditions:

Instrument Air (IA) Compressor "A" is operating as the LEAD compressor.

IA Compressor "B" is in an AUTO-START condition as the LAG compressor.

IA Compressor "C" is in STANDBY.

The following sequence of events occur:

At 1415, 092A, COMPRESS AIR PRESS LO, alarms as pressure drops to 112 psig.

At 1416, 091A, INSTR AIR DRYER PRESS LO, alarms as pressure drops to 100 psig.

All other Control Room alarms related to the IA System remain clear.

At 1420, a stuck-open pre-filter relief valve on dryer Train B reseats.

At 1422, alarm 091A has cleared, 092A still in, header pressure is 100 psig At 1424, assuming NO additional Operator actions and with IA Compressor "A" running and loaded, which ONE of the following is the status of IA Compressors "B" and "C"?

IA Compressor "B" is _____1)_____ and IA Compressor "C" is _____2)_____.

A. 1) running and unloaded; 2) running and unloaded B. 1) running and loaded; 2) shutdown C. 1) running and unloaded; 2) shutdown D. 1) running and loaded; 2) running and loaded Justification A. Incorrect, see below.

B. Incorrect, see below.

C. Incorrect, see below.

D. Correct, From SY1407800:

The air header pressure at which the compressors will start is as follows:

The normal operating compressor (lead compressor) cycles between 116 and 125 psig.

The first standby (lag) compressor starts when system pressure drops to 114 psig, and unloads at 123 psig.

The second standby compressor starts at 112 psig, and unloads at 121 psig.

With the stated conditions in the stem all KA Comp. are running and loaded.

K/A Statement: Ability to monitor automatic operation of the IAS, including: Air pressure Technical Reference(s): SYS KA-120, 121, SY1407800, ALR 00-091A, ALR 00-092A, M-12KA06 Proposed references to be provided to applicants during examination: None

Learning Objective: R5 Question Source: Bank # _47197______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___CPNPP 2009_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Modified based on Jane comments. Changed to Rev. 5, 7/2/11 Answer: D

60. 078 K1.04 2 Given the following plant conditions:

KA HSS-310, Sequencer Selector Switch is in position B-C-A

"B" Instrument Air compressor is running

The turbine watch opens KA-V1475, AIR COMP CKA01B RTN HDR DRAIN, causing a very high flow condition

Annunciator 00-092E, AIR CMPSR B TROUBLE, is in alarm Which ONE of the following describes the cause for the trouble alarm?

The compressor tripped due to . . . .

A. high temperature.

B. low oil pressure.

C. high oil pressure.

D. motor overload.

Justification A. Correct. High dP thru EF HV-044 sensed by EF PDT-44 isolates cooling water B. Incorrect, loss of cooling water should not cause low oil pressure C. Incorrect, loss of cooling water should not cause high oil pressure D. Incorrect, loss of cooling water should not cause motor overload, but is plausible due to motor effects K/A Statement: Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Cooling water to compressor (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Technical Reference(s): ALR 00-092E, SY1407800, SY1408900, M-12EF01, M-12KA01 Proposed references to be provided to applicants during examination: None Learning Objective: R3/R4 Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _2-9_

55.43 _____

Comments:

Incorporated Lee/Jane comments, Replaced K/A 5/27/11 due to not applicable to WC, changed to Rev. 2, 6/9/11 Answer: A

61. 103 A2.04 8 Given the following plant conditions:

Refueling is in progress

Spent Fuel Pool level is -8 inches

OFN KE-018, FUEL HANDLING ACCIDENT, was just entered due to a Containment Air Radiation High alarm

An alarm inside Containment has just been sounded Which ONE of the following 1.) Identifies an alarm inside Containment and 2.) What action is required per OFN KE-018, FUEL HANDLING ACCIDENT?

A. 1.) Alarm is a "wailing" tone.

2.) Commence filling the Refueling pool.

B. 1.) Alarm is a "pulsing" tone.

2.) Evacuate all personnel.

C. 1.) Alarm is a "wailing" tone.

2.) Evacuate all personnel.

D. 1.) Alarm is a "pulsing" tone.

2.) Commence filling the Refueling pool.

Justification:

This is the correct tone for a CTMT Evacuation Alarm. This is the required action per OFN KE-018.

A. Distracter - Plausible because this is the correct tone, however, Filling the Refueling pool is only necessary if pool level is <23 above the fuel.

B. Distracter - Plausible, however, it is a wailing tone.

C. Correct.

D. Distracter - Plausible but a Site Evacuation Alarm. Filling the Refueling pool is only necessary if pool level is <23 above the fuel.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations : Containment evacuation (including recognition of the alarm) (CFR: 41.5 /

43.5 / 45.3 / 45.13)

Technical Reference(s): OFN KE-018, ALR 00-076D, LO1732428 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _47198______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___CPNPP 2009_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Replaced question with bank question 47198. Changed to Rev. 8, 7/29/11 Answer: C

62. 2.1.18 4 Given the following plant conditions:

A reactor trip occurring three days ago was originally attributed to and logged as being caused by rod drive MG breaker failures

After significant trouble shooting the actual cause of the trip was determined to be from several dropped rods Which ONE of the following describes the process for addressing the original autolog log entry?

A. A late entry should be made stating the correct cause of the trip and the original entry should be annotated to refer to the new entry by page and line number.

B. The original entry should be corrected and initialed by the person who made that entry, with the date and time of the correction indicated.

C. The original entry should be corrected and initialed by the person who determined the correct cause, with the date and time of the correction indicated.

D. A late entry should be made stating the actual cause of the trip and refer to the original entry with the date, time and person.

Justification A. Incorrect, first part correct, original entries are not changed IAW AP 21-001, 6.5.3, you would use a late entry B. Incorrect, original entries are not changed IAW AP 21-001, 6.5.3, you would use a late entry C. Incorrect, original entries are not changed IAW AP 21-001, 6.5.3, you would use a late entry D. Correct.

Corrections to electronic log entries may be made prior to archiving by the individual that made the initial entry. Corrections to others log entries shall include whose log entry was corrected along with the date the entry was corrected.

Logging of information and/or events should be done in chronological order. If an "after the fact" entry is made, it should be logged using the actual time of the event, if during that shift.

If an after the fact log entry is made for a shift before the current one, the Late Entry checkbox in AUTOLOG should be checked.

K/A Statement: Ability to make accurate, clear, and concise logs, records, status boards, and reports. (CFR: 41.10 / 45.12 / 45.13)

Technical Reference(s): AP 21-001, LO1733211 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _10__

55.43 _____

Comments:

Modified based on Brendan comments. Changed to Rev. 4, 6/11/11.

Answer: D

63. 2.1.36 7 Given the following plant conditions:

Refueling is in progress

Fuel movement is being restarted after a long delay Which ONE of the following would require suspension of fuel movement?

A. Only one RHR pump is in operation.

B. Containment air temperature is 110°F.

C. 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months have elapsed since the last communication surveillance check has been performed between the control room and refueling personnel.

D. STS CR-002, SHIFT LOG FOR MODES 4 5 AND 6, was completed except that GK RE-4, Control Room Air In Gas Act, is out of service, and GK RE-5 is out of calibration.

Justification A. Incorrect, normally only one in service TS 3.9.5 B. Incorrect, not a limit to move fuel TS 3.6.5 C. Incorrect, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is required per Attachment A, STS CR-002, A.27, checked every 8 by the checklist D. Correct, RE-4 and 5 required to be operable, A. 21 K/A Statement: Knowledge of procedures and limitations involved in core alterations.

(CFR: 41.10 / 43.6 / 45.7)

Technical Reference(s): LO1732109, GEN 00-009, STS CR-002, TS table 3.3.7-1 Proposed references to be provided to applicants during examination: None Learning Objective: R2/R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _10___

55.43 __6___

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/29/11 Answer: D

64. 2.1.45 4 Given the following plant conditions:

The plant has experienced a large break LOCA

All safeguards equipment operated as designed

The crew has transitioned to EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION

Annunciator 00-047C, RWST LEVEL LOLO2, is illuminated

Indications are as follows:

1. EJ LI-7 CTMT RECIRC SUMP LEV = 2002' 5"

2. EJ LI-8 CTMT RECIRC SUMP LEV = 2002' 6"

3. LF LI-9 CTMT NORM SUMP LEV = 2001'

4. LF LI-10 CTMT NORM SUMP LEV = 2001'

5. BN LI-930 RWST LEV = 11%

6. BN LI-931 RWST LEV = 11%

7. BN LI-932 RWST LEV = 13%

8. BN LI-933 RWST LEV = 12%

Which ONE of the following identifies parameters used to accurately determine if the alarm is valid?

A. 1, 3, 4, 7 B. 2, 3, 5, 8 C. 2, 4, 6, 7 D. 1, 2, 5, 6 Justification A. Incorrect. 3 and 4 are reading low, 7 is reading high compared to the diverse indications.

B. Incorrect. 3 is reading low & 8 is reading high compared to the diverse indications.

C. Incorrect. 4 is reading low & 7 is reading high compared to the diverse indications.

D. Correct. 1, 2, 5, 6 are accurate indications and diversely show that the alarm is valid.

K/A Statement: Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7 / 43.5 / 45.4)

Technical Reference(s): EMG ES-12, EMG E-1, TBN 01, EMG C-31 Figure 1, WC Tank Setpoint Document, LO1732322 Proposed references to be provided to applicants during examination: None Learning Objective: R3

Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7___

55.43 __5___

Comments:

Incorporated NRC comments, changed to Rev. 4, 7/29/11 Answer: D

65. 2.2.41 5 One Operations off-normal procedure (OFN) requires a local operator to close EP HV-8808 series (Safety Injection Accumulator Isolation) valves by installation of a jumper from Wire 1 to Wire 21 at terminal board points 4 and 5.

What does this jumper do to the valve circuit?

(REFERENCE PROVIDED)

A. Deenergizes the 42C Coil allowing the 42O coil to energize.

B. Jumpers the 42C/a contact energizing the 42C coil.

C. Deenergizes 42 C coil allowing the 42 C/b contact to close.

D. Jumpers the SIS contacts to prevent SI actuation of the valve.

Justification A. Incorrect,42C coil is energized closing the valve.

B. Correct.

C. Incorrect, The jumper energizes the 42C coil closing the 42C/a contacts, closing the valve.

D. Incorrect, it does not jumper the SIS contact since it is a normally closed contact. It jumpers the main control board HIS close pushbutton.

OFN RP-014, step 18b E-13EP02A -- See attached drawing E-13EP02A (partial)

K/A Statement: Ability to obtain and interpret station electrical and mechanical drawings. (CFR: 41.10 / 45.12 / 45.13)

Technical Reference(s): OFN RP-014, LO1732424, E-13EP02A Proposed references to be provided to applicants during examination: E-13EP02A Learning Objective: R3 Question Source: Bank # _18564______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 5, 8/1/11 Answer: B

66. 2.2.43 4 A Main Control room "F" tier annunciator alarm has been alarming and resetting at a rate greater than sixteen times in an hour.

What is the proper response to this condition per AP 21-001, CONDUCT OF OPERATIONS?

A. Allow the alarm to time out, place the annunciator number on the White Board with explanation, and designate an individual to continuously monitor the alarm section with the timed out alarm.

B. Initiate a Work Request/Condition Report for the problem, make an Instrument-Out-Of-Service sticker and affix to the alarm window, then allow the alarm to time out.

C. Allow the alarm to time out, place the annunciator number on the white board with explanation, and then pull the alarm card in the RK racks to return the MCB alarms to Black Board.

D. Initiate an Instrument-Out-Of-Service log entry and affix a sticker to the alarm window, write a Work Request/Condition Report, and then allow the window to time out.

Justification A. Correct. Section 6.3 B. Incorrect, monitoring required C. Incorrect, maintenance required to repair D. Incorrect, all are things to do but not in this case K/A Statement: Knowledge of the process used to track inoperable alarms. (CFR: 41.10 /

43.5 / 45.13)

Technical Reference(s): AP 21-001, LO1733211 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _46942______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___2008 DCPP_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Modified based on Brendan comments. Changed to Rev. 4, 6/11/11 Answer: A

67. 2.3.12 6 Issuance of a key for entry into a Very High Radiation Area requires the permission of the _____________ in accordance with AP 25A-200, ACCESS TO LOCKED HIGH OR VERY HIGH RADIATION AREAS.

A. HP Shift Technician ONLY B. HP Supervisor and HP Shift Technician C. Shift Manager and Radiation Protection Manger D. Radiation Protection Manger and Plant Manager Justification A. Incorrect, not complete answer B. Incorrect, not complete answer C. Correct, 5.2 & 5.3, 6.3.1 D. Incorrect, PM responsible for the process K/A Statement: Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 / 45.10)

Technical Reference(s): AP 25A-200, LO1733204 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _12__

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 6, 7/29/11 Answer: C

68. 2.3.7 3 What posting is required for areas accessible to personnel with radiation levels greater than or equal to 1000 mr/hr at 30 cm. (12 in.), but less than 500 Rad/hr at 1 meter from the radiation source?

A. Radiation Areas B. High Radiation Areas C. Very High Radiation Areas D. Locked High Radiation Areas Justification A. Incorrect, see below B. Incorrect, see below C. Incorrect, see below D. Correct, see below AP 25A-200, Section 4.0, Step 4.5, Locked High Radiation Area (LHRA) 4.5.1 Areas accessible to personnel with radiation levels equal to or greater than 1,000 mREM/hr at 12 in (30 cm.) from the radiation source, but less than 500 Rads at 1 meter (3.1.3).

K/A Statement: Ability to comply with radiation work permit requirements during normal or abnormal conditions. (CFR: 41.12 / 45.10)

Technical Reference(s): AP 25A-200, LO1733204, AP 25A-001 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _17523______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _12__

55.43 _____

Comments:

Modified based on Lee/Jane comments. Changed to Rev. 2, 3/23/11 Answer: D

69. 2.4.21 2 Given the following plant conditions:

At 100% power a main steamline break on "D" S/G occurred

SI actuated and EMG E-0, REACTOR TRIP OR SAFETY INJECTION, was entered

The crew transitioned to EMG E-2, FAULTED STEAM GENERATOR ISOLATION, and completed all actions required by the procedure

The crew transitioned to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, as required and at step 2, Check If S/G's Are Not Faulted, they find the following:

Secondary Rad Monitors - NORMAL

Containment pressure - 28 psig

RCS Cold Legs - 225°F and decreasing

RCS Hot Legs - 235°F and decreasing

RCS pressure - 1000 psig and decreasing

S/G "A", "B", 'C" pressures - 700 psig and stable

S/G "D" - 200 psig and decreasing Which ONE of the following describes the expected crew actions?

A. Continue and complete all of EMG E-1 since all actions of EMG E-2 have already been completed.

B. Immediately transition to EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE due to Orange path.

C. Immediately transition to EMG FR-P1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK due to Orange path.

D. Go to EMG E-2, FAULTED STEAM GENERATOR ISOLATION to re-perform the procedure.

Justification A. Incorrect, must exit and go to any red/orange path that exists, status trees are continuously monitored B. Incorrect, FR-P series has higher priority than Z series C. Correct.

D. Incorrect, E2 is not re-entered unless indications of an additional faulted S/G are identified K/A Statement: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12)

Technical Reference(s): EMG F-0, LO1732338, EMG FR-P1

Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _18493______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___2006 WC_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __7___

55.43 __5___

Comments:

Modified question based on NRC comments, changed to Rev. 2 7/29/11 Answer: C

70. 2.4.47 2 Given the following plant conditions:

A natural circulation cooldown is in progress using EMG ES-06, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN THE VESSEL (WITH RVLIS)

Pressurizer level begins rising rapidly

RVLIS indication is 68% and decreasing Which ONE of the following describes the expected crew actions?

A. Re-pressurize the RCS until RVLIS indication is >76%.

B. Stop the cooldown and energize backup heaters to saturate the PZR.

C. Initiate SI and transition to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

D. Transition to EMG FR-I3 RESPONSE TO VOIDS IN REACTOR VESSEL.

Justification A. Correct. Repressurizing will cause the saturation temperature to change. By repressurizing the steam void in the vessel will collapse, increasing RX vessel level.

B. Incorrect, RNO to establish RCP start C. Incorrect, foldout page would transition to E-0, not E-1 D. Incorrect, I3 sends you right back to ES, incorrect transition K/A Statement: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 /

45.12)

Technical Reference(s): EMG ES-06, LO1732317 Proposed references to be provided to applicants during examination: None Learning Objective: R11 Question Source: Bank # _14084______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Incorporated Jane comments, changed to Rev. 2 6/27/11 Answer: A

71. 2.4.9 3 Given the following plant conditions:

The unit is currently in Mode 4

RCS temperature is 290°F with RHR in service

PRT level has started to rise

RCS pressure indicates 285 psig

Pressurizer level has started to lower in an uncontrolled manner

Containment pressure and radiation levels are normal Which ONE of the following describes the mitigative action required by OFN BB-031, SHUTDOWN LOCA?

A. Lower RCS pressure to reseat the RHR suction relief valve.

B. Stop the affected RHR pump(s) if pressurizer level is less than 6%.

C. Open an RWST suction to the running RHR pump to increase PZR level.

D. Place both RHR pumps in Pull-to-Lock when RCS subcooling is less than 45°F.

Justification A. Incorrect, pressure < LTOP setpoint 290°F = ~450 psig suction relief B. Correct, foldout page RHR Pump Stopping Criteria C. Incorrect, incorrect action for this temperature, would have to be less than 225°F to be correct.

(27 RNO)

D. Incorrect, SP is 30°F, 45°F is the adverse SP K/A Statement: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies. (CFR: 41.10 /

43.5 / 45.13)

Technical Reference(s): OFN BB-031, LO1732425 Proposed references to be provided to applicants during examination: None Learning Objective: R7 Question Source: Bank # _______

Modified Bank # _18624______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Incorporated NRC comments, changed to Rev. 3 7/29/11 Answer: B

72. E01 EA2.2 3 Given the following plant conditions:

The unit was initially at 100% power, normal lineup, NE01 Emergency Diesel out of service

Reactor trip and SI have occurred

Offsite power was lost concurrent with the trip

All systems respond as designed

S/G level in "D" S/G was increasing uncontrollably

Crew initially transitioned to EMG E-3, STEAM GENERATOR TUBE RUPTURE

Crew has determined the high "D" S/G level was due to AL HV-5, MOTOR DRIVEN AFWP DISCH HDR TO SG D ISO Valve, failure and have entered EMG ES-01, REDIAGNOSIS

"D" S/G pressure is stable at 1010 psig At this time:

"B" CCW pump has just suffered a seized bearing. The pump breaker fails to trip but the upstream supply breaker (the breaker supplying the bus which supplies the pump) does trip, isolating the fault

The running Emergency Diesel generator did NOT trip during the transient Which ONE of the following is the correct EMG transition to make in response to the above conditions? (Assume another Operator is addressing any OFN's that apply as a result of the above conditions.)

A. Return to EMG E-3 and terminate SI in EMG E-3, STEAM GENERATOR TUBE RUPTURE.

B. Transition to EMG C-0, LOSS OF ALL AC POWER.

C. Transition to EMG E-2, FAULTED STEAM GENERATOR ISOLATION.

D. Transition to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

Justification:

A. Incorrect. The high level in D SG was due to the AL HV-5 failure and not a SGTR. Return to E-3 would provide no assistance in mitigating the event.

B. Correct. EDG's are not supplying busses NB01/02. CCW pump is a 4160V breaker (NB0206).

NB02 is lost, NB01 is de-energized on LOOP.

C. Incorrect. No indications that a faulted SG as D SG at 1010# and stable.

D. Incorrect. With NB01/02 de-energized, correct procedure transition is EMG C-0.

K/A Statement: Ability to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection Rediagnosis) Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments. (CFR: 43.5 /

45.13)

Technical Reference(s): EMG ES-01, E-13EG01C, LO1732314 Proposed references to be provided to applicants during examination: None Learning Objective: R1/R2/R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Other K/A: E01 EA2.1 [3.2/4.0] {43.5}

E01 EK1.1 [3.1/3.5] {41.8, 10}

E01 EK1.2 [3.4/4.0] {41.8, 10}

Est Time: 3 min.

Modified based on Lee/Jane comments. Changed to Rev. 3, 3/23/11 Answer: B

73. E04 EA1.2 2 Given the following plant conditions

A LOCA outside Containment has occurred

The crew is performing the actions in EMG C-12, LOCA OUTSIDE CONTAINMENT

RCS Cooldown is NOT in progress Which ONE of the following indications is used to determine if the leak has been isolated in accordance with EMG C-12?

A. RVLIS indication, because when the break is isolated, vessel head voiding will immediately be reduced.

B. Pressurizer level, because when the break is isolated, RCS inventory will rapidly rise.

C. Safety injection flow, because when the break is isolated, it is the first parameter that will change.

D. RCS pressure, because when the break is isolated, SI flow will repressurize the RCS.

Justification A. Incorrect. RVLIS may indicate 100% at the start, so may not provide indication of isolation at all B. Incorrect. RCS inventory will increase, but may not immediately show up on PZR level C. Incorrect. SI Flow is a good confirmatory indication when RCS pressure rises, because it will be reduced, but RCS pressure rise is the only immediate indication D. Correct. RCS pressure is the primary means of determining whether the leak is isolated.

K/A Statement: Ability to operate and / or monitor the following as they apply to the (LOCA Outside Containment) Operating behavior characteristics of the facility (CFR:

41.7 / 45.5 / 45.6)

Technical Reference(s): EMG C-12, LO1732333 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _59101______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___Callaway 2007_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 __7__

55.43 _____

Comments:

Other K/A E04 EA2.2 Incorporated Jane comments, changed to Rev. 2 6/27/11 Answer: D

74. E05 EK3.2 1 Which ONE of the following is the primary reason for stopping all RCP's in EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK?

A. They are secured to prevent the heat added by the RCPs from adversely affecting indications used to determine whether or not RCS bleed and feed will be required.

B. They are secured to reduce the heat added from the RCPs, thereby delaying the need for bleed and feed and gaining time to establish a means of supplying FW to a S/G.

C. This will reduce RCS pressure enough to ensure bleed and feed is adequate for RCS cooling requirements.

D. This will establish natural circulation conditions and will tend to mitigate the loss of secondary heat sink by establishing a delta T across the core.

Justification:

A. Incorrect, these indications are designed for these conditions, not a basis for pump trip.

B. Correct. Basis for step 7.

C. Incorrect, this alone will not be adequate for heat sink, this only gives and additional 20-30 minutes.

Step 2 basis.

D. Incorrect, this is a natural outcome but not the basis for doing so.

K/A Statement: Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink). (CFR: 41.5 / 41.10, 45.6, 45.13)

Technical Reference(s): BD-EMG FR-H1, LO1732346, EMG FR-H1 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _59099______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___Callaway 2007_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _5, 10_

55.43 _____

Comments:

Replaced question based on Brendan comments, 6/11/11 Answer: B

75. E11 EK3.2 6 Given the following plant conditions:

While performing EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION, an ORANGE path condition is identified for the Containment Status Tree Which ONE of the following reasons describes why the Containment Spray Pumps are operated within the guidelines of EMG C-11 instead of using the guidelines contained in EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE?

EMG C-11 ensures . . .

A. a more rapid means of verifying automatic actuation of the containment spray system.

B. that the maximum heat removal system capacity that is available is used to reduce the containment pressure.

C. that the pump operating criteria requires less pumps, permitting reduced containment spray operation to conserve RWST inventory.

D. that the pump operating criteria requires more pumps, ensuring continuous containment spray system operation to reduce containment pressure.

JUSTIFICATION:

A. Incorrect; this is from EMG FR-Z1 Basis Document see below.

B. Incorrect; This is the requirement from EMG FR-Z1, for coolers and spray.

C. Correct.

D. Incorrect; EMG C-11 uses a less restrictive criteria, which permits reduced spray pump operation depending on RWST level and containment pressure.

EMG FR-Z1 basis - This procedure specifies maximum available heat removal system operability in order to reduce containment pressure. EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION, uses a less restrictive criteria, which permits reduced spray pump operation depending on RWST level, containment pressure and number of emergency fan coolers operating. The less restrictive criteria for containment spray operation is used in EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION, since recirculation flow to the RCS is not available and it is very important to conserve RWST water, if possible, by stopping containment spray pumps.

K/A Statement: Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation) Normal, abnormal and emergency operating procedures associated with (Loss of Emergency Coolant Recirculation). (CFR:

41.5 / 41.10, 45.6, 45.13)

Technical Reference(s): EMG C-11, EMG FR-Z1, BD-EMG C-11, BD-EMG FR-Z1, LO1732350, LO1732332

Proposed references to be provided to applicants during examination: None Learning Objective: R3/R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _5, 10_

55.43 _____

Comments:

Modified based on NRC comments. Changed to Rev. 6, 7/29/11 Answer: C

76. S 006 2.4.6 4 Given the following plant conditions:

Safety injection pump train "A" has been tagged out for motor bearing replacement

A Safety Injection occurs due to a Large Break LOCA inside Containment

Two hours later, the "B" SI pump fails

All other equipment functions as designed for the duration of the accident Which ONE of the following describes how the loss of both SI pumps affect the ability of the crew to mitigate this accident in the long term?

A. Align both trains of ECCS to hot leg recirculation when required, cold leg recirculation flowpath is still available, loss of SI pumps has no long term effect.

B. Both trains of ECCS will remain in injection mode during alignment per EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION.

C. Action to restore one SI pump to service must be completed prior to realignment per EMG ES-13, TRANSFER TO HOT LEG RECIRCULATION.

D. Transition to EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION, will be required even if RCS pressure is less than 300 PSIG.

Justification:

A. Correct.

B. Incorrect. SI pumps does not determine RHR alignment.

C. Incorrect, RHR is the long term cooling supply.

D. Incorrect, ES-11 entry criteria will not be met.

K/A Statement: Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)

Technical Reference(s): EMG ES-13, BD-EMG ES-13, LO1732323 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _12602______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Incorporated Jeff comments, changed to rev. 4, 8/2/11 Answer: A

77. S 008 A2.05 4 Given the following plant conditions:

The unit is at 100% power

Instrument air is lost to BG TCV-130, LTDN HX OUTLET TEMP CTRL, valve

Annunciator 052A, CCW TO RCP FLOW LO, actuates

ALR 00-052A actions are performed Which ONE of the following describes the INITIAL response of letdown temperature and CCW return flow from the letdown heat exchanger and what compensatory action would be taken?

Response Compensatory Action A. Letdown temperature will increase Secure Normal letdown and place Excess Letdown in service per SYS BG-310, SECURING NORMAL LETDOWN.

CCW return flow will decrease B. Letdown temperature will decrease Secure Normal letdown and place Excess Letdown in service per SYS BG-310, SECURING NORMAL LETDOWN.

CCW return flow will increase C. Letdown temperature will increase Divert letdown flow to the RHUT BG LCV-112A, Divert Valve per OFN KA-019, LOSS OF INSTRUMENT AIR.

CCW return flow will decrease D. Letdown temperature will decrease Divert letdown flow to the RHUT BG LCV-112A, Divert Valve per OFN KA-019, LOSS OF INSTRUMENT AIR.

CCW return flow will increase Justification:

A. Incorrect. Opposite system response, correct action. Plausible since high temperatures on a failed valve would require removing normal letdown from service.

B. Correct. Low temperature on demineralizers would initiate a reactivity addition; removing normal letdown from service would be the correct response.

C. Incorrect. Opposite system response, incorrect action. Plausible because high temperature diversion would prevent a reactivity addition or excursion, however OFN KA-019 does not direct diverting letdown flow.

D. Incorrect. Correct system response. Incorrect action and procedure. Plausible because low temperature diversion would prevent a reactivity addition or excursion, however OFN KA-019 does not direct diverting flow.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those malfunctions or operations: Effect of loss of instrument and control air on the position of the CCW valves that are air operated (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Technical Reference(s): OFN KA-019, ALR 00-052A, SYS BG-310, M-12BG02, SYS BG-208, LO1732429, AP 21-001 Step 6.1.3 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Other K/A: 026 AA1.07 [2.9/3.0] {41.7}

008 G2.4.11 [4.0/4.2] {41.10, 43.5}

Est Time: 3 min Modified based on Brendan comments. Changed to Rev. 4, 6/11/11 Answer: B

78. S 010 A2.03 7 Given the following plant conditions:

The unit is at 100% power

A Pressurizer PORV is leaking by the seat to the PRT at a rate of 1 gpm

All other system components are normal Which ONE of the following describes the correct Technical Specification Limiting Condition of Operation (LCO) and procedural entry requirements to mitigate?

A. LCO 3.4.11 - Pressurizer Power Operated Relief Valves (PORVs).

OFN BB-007, RCS LEAKAGE HIGH B. LCO 3.4.11 - Pressurizer Power Operated Relief Valves (PORVs).

GEN-004, POWER OPERATION C. LCO 3.4.13 - RCS Operational LEAKAGE GEN-004, POWER OPERATION D. LCO 3.4.13 - RCS Operational LEAKAGE OFN BB-007, RCS LEAKAGE HIGH Justification A. Correct. LCO, Action, procedure to mitigate B. Incorrect, PORV is inop, within limits no SD required, incorrect procedure C. Incorrect, wrong leakage (not > 10 gpm), no SD required, incorrect procedure, < 10 gpm D. Incorrect, wrong leakage (not > 10 gpm), right action, right procedure, < 10 gpm K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

PORV failures (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Technical Reference(s): TS 3.4.13, TS Definitions, SY1300200, LO1732417, OFN BB-007, TS 3.4.11 Proposed references to be provided to applicants during examination: None Learning Objective: R8 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Incorporated NRC comments, changed to Rev. 7, 7/29/11 Answer: A

79. S 029 EA2.01 1 Given the following plant conditions:

A large LOCA has occurred resulting in a plant trip

The following plant conditions exist:

Reactor Power 10% and lowering

Pressurizer Level 0%

Pressurizer Pressure 1300 psig and lowering

RVLIS - Nat Circ 38%

Core Exit TCs 1250°F and rising

Containment Temp 175°F Which ONE of the following would be the correct implementation of the Emergency Operating Procedures after implementation of EMG E-0, REACTOR TRIP OR SAFETY INJECTION?

A. EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION, to SACRG-1, SEVERE ACCIDENT CR GUIDELINE INITIAL RESPONSE.

B. EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, to FR-C1, RESPONSE TO INADEQUATE CORE COOLING.

C. EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT, to SACRG-1, SEVERE ACCIDENT CR GUIDELINE INITIAL RESPONSE.

D. EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION, to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING.

Justification A. Correct.

B. Incorrect, EMG FR-S1 required, E-1 plausible due to LOCA.

C. Incorrect, EMG FR-S1 required, E-1 plausible due to LOCA, C.1 plausible due to CET.

D. Incorrect, EMG FR-S1 required, C.1 plausible due to CET.

K/A Statement: Ability to determine or interpret the following as they apply to a ATWS:

Reactor trip alarm (CFR 43.5 / 45.13)

Technical Reference(s): EMG FR-S1, LO1732339 Proposed references to be provided to applicants during examination: None Learning Objective: R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Replaced K/A based on NRC comments, 7/29/11 Answer: A

80. S 033 2.2.25 6 Given the following plant conditions:

Normal Reactor start up in progress

Power is 3% RTP

Intermediate Range Channel NI36 is INOPERABLE

Maintenance has determined it will take at least three hours to correct Which ONE of the following describes the correct operational CRS decision, Technical Specification action, and the basis for that action?

A. Increase power to > P-10 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Provides protection by minimizing the time spent below P-10 and removes the requirement for the intermediate range channel to be operable B. Maintain 3% RTP until NI36 is corrected.

Ensures that protection is provided for multiple rod drop accidents C. Maintain between 5% and 10% RTP until NI36 is corrected.

Ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture D. Startup increase may continue without restriction or time limit.

Provides protection for control rod withdrawal from subcritical and control rod ejection events.

Justification A. Correct. 3.3.1 Condition F B. Incorrect, wrong TS action, Neg rate basis, wrong basis C. Incorrect TS action, High positive rate basis, wrong basis D. Incorrect, wrong TS action, SR basis, wrong basis, 24 hr time limit to be above P-10 K/A Statement: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2)

Technical Reference(s): TS, TS Basis, SY1301501 Proposed references to be provided to applicants during examination: None Learning Objective: R13 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _5, 7__

55.43 __2___

Comments:

Modified based on NRC comments. Changed to Rev. 6, 7/29/11 Answer: A

81. S 045 A2.17 4 Given the following plant conditions:

The unit is at 45% power

While conducting STS AC-001, MAIN TURBINE VALVE CYCLE TEST, during the performance of step B.1.2, the Operator presses and holds the CIV-2 TEST pushbutton

TURB STOP VLV2 bistable light is LIT

Upon releasing the CIV-2 TEST pushbutton the test solenoids fail to de-energize The CIV-2 and ISV are 1 and you should direct an Operator action to 2 and enter procedure 3 .

1. Valve Position 2. Operator Action 3. Procedure in Effect A. Closed trip the reactor EMG E-0, REACTOR TRIP OR SAFETY INJECTION B. Closed evaluate TR 3.3.14 GEN 00-004, POWER OPERATION Turbine Overspeed Protection C. Open reduce turbine load OFN MA-038, RAPID PLANT to 30% SHUTDOWN D. Open reduce turbine load ALR 00-083C, RX PARTIAL TRIP to 40%

Justification A. Incorrect, only if trip criteria met, not under these conditions B. Correct.

C. Incorrect, valves remain closed D. Incorrect, valves remain closed.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Malfunction of electrohydraulic control (CFR: 41.5 / 43.5 / 45.3 / 45.5)

Technical Reference(s): STS AC-001, TR 3.3.14, GEN 00-004, SY1504600 Proposed references to be provided to applicants during examination: None Learning Objective: R8 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Modified based on NRC comments. Changed to Rev. 4, 7/29/11 Answer: B

82. S 056 AA2.03 5 Given the following plant conditions:

Preparations are being made to establish charging flow in EMG ES-03, SI TERMINATION, when offsite power is lost

The Emergency Diesel Generators start and power both safety buses

Fifteen seconds later the STA observes SI Pumps are NOT running

SI pumps were running before loss of offsite power Should SI pumps restart automatically by this point in time and what procedural action is directed by the CRS?

A. Yes, the LOCA Sequencers should have started the SI pumps 10 seconds ago.

Direct the crew to manually start the SI pumps.

B. Yes, the Shutdown Sequencer should have started the SI pumps 10 seconds ago.

Direct the crew to manually start the SI pumps.

C. No, the LOCA Sequencer did not actuate and will not start the SI pumps.

Direct the crew to manually start the SI pumps after the S/D sequencer has timed out.

D. No, the LOCA Sequencer will not start the SI pumps for another 10 seconds.

Direct the crew to monitor the pumps and start them if the sequencer does not.

Justification A. Incorrect, the LOCA seq was not actuated because SI was reset and SS sequencer was actuated by LOOP B. Incorrect, the SS does not start SI pumps C. Correct. Crew should restart the pumps until procedurally directed to secure them D. Incorrect, the LOCA seq was not actuated because SI was reset SI reset caution prior to step 1. SI is reset in step 1, Charging is adjusted in steps 4-6, 10-14 SI Pumps stopped at step 15 K/A Statement: Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational status of safety injection pump (CFR: 43.5 / 45.13)

Technical Reference(s): BD-EMG ES-03, EMG ES-03, E-12NF01, SY1406401, LO1732316 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _18031______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Modified based on NRC comments. Changed to Rev. 5, 7/29/11.

Answer: C

83. S 057 2.4.8 5 Given the following plant conditions:

A Reactor startup is in progress

Source range channels NI31 and NI32 indicate 104 cps

Intermediate range channels NI35 and NI36 indicate 5 X10-11 amps

Annunciator 025A, NN01 INST BUS UV, has just alarmed Which ONE of the following describes the actions and identifies the required implementation of procedures for this event?

A. Commence a reactor shutdown to insert all control banks, AND Isolate Instrument Inverter NN11.

Implement OFN NN-021, LOSS OF VITAL 125 VAC INSTRUMENT BUS, and refer to GEN 00-003, HOT STANDBY TO MINIMUM LOAD, as necessary.

B. Verify reactor trip, AND Restore power to NN01 from alternate AC power source.

Implement EMG E-0, REACTOR TRIP OR SAFETY INJECTION, and have an Operator perform OFN NN-021, LOSS OF VITAL 125 VAC INSTRUMENT BUS.

C. Commence a reactor shutdown to insert all control banks, AND Restore power to NN01 from alternate AC power source.

Implement OFN MA-038, RAPID PLANT SHUTDOWN, and have an Operator perform OFN NN-021, LOSS OF VITAL 125 VAC INSTRUMENT BUS.

D. Verify reactor trip, AND Isolate Instrument Inverter NN11.

Implement EMG E-0, REACTOR TRIP OR SAFETY INJECTION, then transition to OFN NN-021, LOSS OF VITAL 125 VAC INSTRUMENT BUS.

Justification A. Incorrect, reactor has tripped, wrong procedures for plant conditions B. Correct, reactor has tripped, plant will need to be stabilized using E-0, plant components will have to be realigned with OFN, AP 15C-003 providing guidance for branching C. Incorrect, reactor has tripped, wrong procedure for plant conditions D. Incorrect, reactor has tripped, plant components using OFN. EMG E-0 does not allow transition to OFN AP 15C-003, Section 6.10, details guidance for referring to a procedure branching and transitioning (CRS knowledge and supervisory activities)

K/A Statement: Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10 / 43.5 / 45.13)

Technical Reference(s): ALR 00-025A, OFN NN-021, EMG E-0, AP 15C-003, M-744-00019, M-744-00020, E-13SE01, E-13SE02, E-13NN01, OFN MA-038, GEN 00-003, LO1732103,

LO1733203, SY1301501, LO1732313, LO1732431 Proposed references to be provided to applicants during examination: None Learning Objective: R18 Question Source: Bank # _45815______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Other K/A: 057 AK3.01 Modified based on Brendan comments. Changed to Rev. 5, 6/11/11 Answer: B

84. S 058 2.4.3 6 Given the following plant conditions:

The plant is at 100% power

RCS Wide Range pressure channel BB PI-405 is out of service

BB PT-403 loses power Based on these conditions, what (if any) LCO's are affected and what actions are appropriate?

(REFERENCE PROVIDED)

A. BB PT-403 is a post-accident instrument. In accordance with TS Table TS 3.3.3-1 for RCS WR pressure indication, a 7-day LCO is entered to restore the channel to OPERABLE.

B. BB PT-403 is a non-qualified instrument, and therefore is NOT covered in Technical Specifications, write a Condition Report to repair.

C. BB PT-403 is a post-accident instrument. NO LCO is entered since BB PT-403 does NOT have an input to any required accident monitoring functional unit, write a Condition Report to repair.

D. BB PT-403 provides a safeguards actuation input. In accordance with TS Table 3.3.2-1 for RCS WR Pressure indication, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to place the associated Safety Injection actuation bistables in TRIP or be in MODE 3 in 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months , and be in MODE 4 in 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months .

Justification A. Correct. Instrument/Safety function identified, correct TS and actions B. Incorrect, PT-403 is a post-accident qualified instrument. Incorrect TS application C. Incorrect, PT-403 is a post-accident qualified instrument. Incorrect TS application.

D. Incorrect, Wide range pressure instruments do provide a safeguards input, LTOP. Incorrect TS application BB PT-403 is powered from NN02 (NK02), BB PT-406 are powered from NN04, Condition A for inop PT-405, Cond C for 2 inop PT-403/405 - restore 1 within 7 days K/A Statement: Ability to identify post-accident instrumentation. (CFR: 41.6 / 45.4)

Technical Reference(s): TS 3.3.3, SY1300202, M-12BB04, E-13BB15, M-761-00043, M-761-00101, OFN SB-008 Att. S Proposed references to be provided to applicants during examination: Table 3.3.3-1, Table 3.3.2-1 Learning Objective: R1 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __6__

55.43 _____

Comments:

Modified based on Jane comments. Changed to Rev. 6, 6/22/11 Answer: A

85. S 062 2.1.25 7 Given the following plant conditions:

The unit is operating at 80%

Generator power is 1050 Megawatts

Annunciator 00-130E, GEN AUX TROUBLE, is in alarm

ALR 408-06A, MACH GAS PRESSURE HIGH-LOW, is in alarm

CC PI-1, MACHINE GAS PRESSURE, reads 41 psig

Reactive Load is outgoing at 200 MVA

Minor leakage from the Main Generator housing is reported Which ONE of the following is the MINIMUM hydrogen pressure needed to ensure the Generator has enough cooling (assume hydrogen cooling aligned normal) and what actions should the CRS to direct?

(REFERENCE PROVIDED)

A. 75 psig for MVARs up to 750 If leakage cannot be isolated, trip the unit and emergency depressurize the Main Generator using SYS CC-321, GENERATOR HYDROGEN EMERGENCY DEPRESSURIZATION B. 60 psig for MVARs up to 600 Isolate the leak in accordance with OFN AF-025, UNIT LIMITATIONS C. 48 psig for MVARs up to 650 Isolate the leak in accordance with OFN AF-025, UNIT LIMITATIONS D. 45 psig for MVARs up to 500 If leakage cannot be isolated, trip the unit and emergency depressurize the Main Generator using SYS CC-321, GENERATOR HYDROGEN EMERGENCY DEPRESSURIZATION Justification A. Incorrect, would be outside curve, correct procedural direction B. Incorrect, would be outside curve, incorrect procedural action, leak is isolated IAW ALR 408 C. Incorrect, do not interpolate curves, incorrect procedural action, leak is isolated IAW ALR 408 D. Correct. ALR 408, Att. P step P4 RNO

K/A Statement: Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12)

Technical Reference(s): OFN AF-025, Figure 1, ALR 408 Att. P, SYS CC-321, LO1732435, SY1502300, SY1504502

Proposed references to be provided to applicants during examination: Figure 1 Learning Objective: R2/R9/R7 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/29/11 Answer: D

86. S 064 A2.06 8 Given the following plant conditions:

The unit is in MODE 4

NE01 D/G is in a post maintenance run

It ran for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months at 1.15 MW

Maintenance has requested that you secure the D/G Which ONE of the following describes 1) the impact on the EDG and 2) the required action?

A. 1) Oil will buildup in the exhaust

2) Increase load on NE01 to half load for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months then full load prior to securing the D/G IAW SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR A B. 1) Sludge will build up in the oil

2) Increase load on NE01 to half load for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months then full load prior to securing the D/G IAW SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR A C. 1) Oil will buildup in the exhaust

2) Decrease load and secure the D/G as requested IAW SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR A D. 1) Sludge will build up in the oil

2) Decrease load and secure the D/G as requested IAW SYS KJ-123, POST MAINTENANCE RUN OF EMERGENCY DIESEL GENERATOR A Justification A. Correct, SYS KJ-123 P&L 4.4 B. Incorrect, During engine operation, a portion of the combustion air is used to drive an ejector. The ejector is designed to maintain a negative pressure in the crankcase.

C. Incorrect, first part correct, wrong action, right procedure D. Incorrect, During engine operation, a portion of the combustion air is used to drive an ejector. The ejector is designed to maintain a negative pressure in the crankcase. wrong action, right procedure.

The Diesel Generator should not be operated at loads less than 20% (1.25 MW) for greater than 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to minimize buildup of oil in the exhaust. If the engine has been ran for an extended amount of time at low load, the engine should be slowly loaded approximately 400 kW per minute, to half load (3 MW) and held there for one hour then loaded to full load.

K/A Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Operating unloaded, lightly loaded, and highly loaded time limit (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Technical Reference(s): SY1406400, SYS KJ-123 Proposed references to be provided to applicants during examination: None Learning Objective: R8 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Incorporated NRC comments. Changed to Rev. 8, 7/29/11.

Answer: A

87. S 065 AA2.01 6 Given the following plant conditions:

Reactor power 65% stable

RCS pressure 2245 psig rising (5# every 3 mins)

RCS temperature 576°F stable

Letdown is isolated

Instrument Air header pressure 85 psig decreasing (1# every 5 mins)

Annunciator 091A, INST AIR DRYER PRESS LO, is LIT

Annunciator 092A, COMPRESS AIR PRESS LO, is LIT Which ONE of the following identifies the required action(s) and implementation of procedures for this event?

A. Attempt to identify and isolate the instrument air system leakage Implement OFN KA-019, LOSS OF INSTRUMENT AIR B. Take manual control of Pressurizer pressure Stabilize the plant using Attachment K (PZR PRESSURE MALFUNCTION) of OFN SB-008 INSTRUMENT MALFUNCTIONS C. Trip the Reactor Implement EMG E-0, REACTOR TRIP OR SAFETY INJECTION D. Perform a shutdown per OFN MA-038, RAPID PLANT SHUTDOWN Refer to OFN KA-019, LOSS OF INSTRUMENT AIR, to identify and isolate the leak Justification A. Correct, step 6 RNO, correct action and procedure implementation B. Incorrect, air pressure above 70 psig, action is for pressure reaching the high pressure trip setpoint, incorrect procedure implementation C. Incorrect, air pressure above 70 psig, will eventually get there if correct actions not done.

Incorrect action, incorrect procedure implementation D. Incorrect, Incorrect action, correct procedure implementation K/A Statement: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Cause and effect of low-pressure instrument air alarm (CFR: 43.5 /

45.13)

Technical Reference(s): OFN KA-019, EMG E-0, OFN MA-038, SY1407800, LO1732429 Proposed references to be provided to applicants during examination: None Learning Objective: R2/R3

Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Modified based on NRC comments. Changed to Rev. 6, 7/29/11 Answer: A

88. S 068 2.1.20 7 During a release of Waste Monitor Tank (WMT), with the release in progress, HB RE-18, Liquid Radwaste Discharge monitor becomes inoperable.

What are the procedurally required actions to be taken?

A. Release in progress may continue up to 30 minutes with the current release permit and take actions for HB RE-18 being INOPERABLE.

B. Terminate the release, have Chemistry reanalyze samples taken of the WMT, and update the release permit, and then recommence the release.

C. Release in progress may continue up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with the current release permit and take actions for HB RE-18 being INOPERABLE.

D. Terminate the release, request Chemistry perform a new analysis, issue a new permit and verify the release path, then recommence the release.

Justification A. Incorrect, samples must be taken and analyzed B. Incorrect, results must be verified by qualified people C. Incorrect, samples must be taken and reviewed before continuing D. Correct, see below All plant releases are monitored for radioactive content. Discharges from the Waste Monitor Tanks are monitored by effluent radiation monitor HB RE-18 (located in the southeast portion of the 1976 level of the Radwaste Building). Detection of high level of radioactivity (Main Control Board alarm Process Rad HiHi will actuate) will cause HB RV-18 to close, automatically terminating a release. The RM 80 for HB RE-18 is powered from non-safety 480 VAC MCC, PG17L.

ODCM, Action 31 Action 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Section 2.1.2, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

Samples shall be taken at the initiation of effluent flow and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter while the release is occurring. To be representative of the liquid effluent, the sample volume shall be proportioned to the effluent stream discharge volume. The ratio of sample volume to effluent discharge volume shall be maintained constant for all samples taken for the composite sample.

2.4.1.2 With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-2. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next annual Radioactive Effluent Release Report, why this inoperability was not corrected within the time specified.

K/A Statement: Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 /

45.12)

Technical Reference(s): AP 07B-003, ALR 00-061C, LO1732420, LO1733209, Table 2-2 Action 31, AP 07B-003, AP 07B-001 Proposed references to be provided to applicants during examination: None Learning Objective: R7/R4 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Modified based on NRC comments. Changed to Rev. 7, 7/29/11 Answer: D

89. S 074 EA2.01 9 Given the following plant conditions:

The unit has tripped from 100% power due to a Small Break Loss of Coolant Accident

RCPs are secured

Pressurizer pressure is 1535 psig

S/G pressure are 1000 psig

CETCs indicate 722°F

RVLIS Natural Circulation Range indicates 42%

The crew is currently in EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT Which ONE of the following describes the amount of superheat and what is the required procedure implementation path?

A. 122°F Transition to EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING B. 177°F Transition to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING C. 177°F Transition to EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING D. 122°F Transition to EMG FR-C1, RESPONSE TO INADEQUATE CORE COOLING Justification:

A. Incorrect, wrong amount of superheat (used 722-600 = 122F superheat). C1 is the required procedure to implement, may select because CETC are < 1200F B. Incorrect. 177 is using S/G pressures, F-0 being monitored should transition to C1 C. Incorrect. 177 is using S/G pressures, C1 is the required procedure to implement may select because CETC are < 1200F.

D. Correct. (used 722°F-600 = 122). Correct procedure.

Sat. temp for 1550 psia (1535 psig) = 600F K/A Statement: Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: Subcooling margin (CFR 43.5 / 45.13)

Technical Reference(s): Steam Tables, BD-EMG F-0, EMG F-0, EMG FR-C1, LO1732341

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Other K/A: 074 EA1.13 [4.3/4.6] {41.7}

193003 K1.17 [3.0/3.2] {41.14}

193003 K1.25 [3.3/3.4] {41.14}

Est Time: 5 min.

Modified based on NRC comments. Changed to Rev. 9, 7/29/11 Answer: D

90. S 001 A2.04 1 Given the following condition at 80%:

A stator cooling water runback has occurred

Annunciator 081C, ROD BANK LOLO LIMIT is lit

Control Bank D rods are inserted to 55 steps

Control Bank C rods are inserted to 171 steps

Reactor power is 61% and stable Which ONE of the following describes: 1) the most immediate concern and 2) what is the proper action to mitigate the situation?

A. 1.) The reactivity worth of a stuck rod in one fully withdrawn group could exceed COLR limits.

2.) Withdraw Control Rod Bank D.

B. 1.) Control Bank D overlap with Control Bank C is greater than 115 steps.

2.) Insert Control Bank C to establish overlap of 115 steps.

C. 1.) Available SHUTDOWN MARGIN may be inadequate.

2.) Initiate boration within one hour.

D. 1.) Axial Flux Difference has been driven too low in the core.

2.) Withdraw Control Bank D as Axial Offset allows.

Justification:

a. Incorrect, Plausible because Rod Insertion Limits are also to minimize worth of an ejected rod, however, not for a stuck rod.

b. Incorrect, Plausible because Rod Bank Overlap is an issue when withdrawing or inserting rods but in this case the overlap is still less than 115 steps and insertion of Control Bank C would cause a further reduction in SHUTDOWN MARGIN.

c. Correct.

d. Incorrect, Plausible because insertion of Control Bank D will drive AFD lower in the core but SHUTDOWN MARGIN is the major concern.

Because Control Rod Insertion Limits have been violated available SHUTDOWN MARGIN may be inadequate. SHUTDOWN MARGIN must be verified within one hour.

3.1.6, CONTROL BANK INSERTION LIMITS Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR in Mode 1 and in Mode 2 with the reactor critical. This LCO is not applicable during the performance of OPT-106A/B.

Control bank insertion limits are required, in addition to shutdown bank insertion, axial flux difference and quadrant power tilt ratio limits, to prevent power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring a reactor trip. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and ensure the required shutdown margin is maintained. Proper control bank sequence and overlap preserve power distribution and reactivity rate insertion assumptions. IPO- 002A/B, Plant Startup from Hot Standby, requires verification that control banks are above insertion limits during a reactor startup prior to achieving criticality. Control bank insertion, sequence and overlap limits are verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per OPT-102A/B. These limits are listed in the COLR.

If control bank insertion is not within limits, adequate shutdown margin must be established and verified by performing a reactivity balance calculation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and the control banks must be restored to within limits in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . If the control banks are not restored to within limits in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the reactor must be shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If control bank sequence or overlap limits are not met, the same action is required.

K/A Statement: Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Positioning of axial shaping rods and their effect on SDM (CFR: 41.5/43.5/45.3/45.13)

Technical Reference(s): TS 3.1.6, SY1300100 Proposed references to be provided to applicants during examination: None Learning Objective: R9 Question Source: Bank # _47238______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___CPNPP 2009_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 __5___

55.43 __5___

Comments:

Replaced K/A based on NRC comments, 7/29/11 Answer: C

91. S 2.1.35 5 Given the following plant conditions:

The unit is in MODE 6

Fuel is being off-loaded from the core

Refueling canal boron concentration is 2230 ppm

Both Source Range indications are in operation with 2 visual and 1 audio indication

The Reactor has been subcritical for 86 hours9.953704e-4 days 0.0239 hours 1.421958e-4 weeks 3.2723e-5 months

A maintenance worker is stationed at the personnel air lock

The equipment hatch is closed and held in place with 4 bolts

2 air hoses are run through the Emergency Escape hatch airlock doors with quick disconnects at the airlock doors with a person stationed for Admin controls

The Reactor vessel water level is 24 feet above the top of the reactor vessel flange Which ONE of the following LCOs, if any, would pertain to the given plant conditions?

A. No LCOs should be entered B. LCO 3.9.1 (Boron Concentration) should be entered C. LCO 3.9.3 (Nuclear Instrumentation) should be entered D. LCO 3.9.4 (Containment Penetrations) should be entered Justification A. Incorrect, see below B. Correct, boron < 2300 ppm C. Incorrect, both SR instruments are operable D. Incorrect, LCO requirements are met COLR 2.11 Boron Concentration (LCO 3.9.1) The refueling boron concentration shall be greater than or equal to 2300 PPM.

4.35 The reactor shall be determined to have been subcritical for at least 76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. With the reactor subcritical for less than 76 hours8.796296e-4 days 0.0211 hours 1.256614e-4 weeks 2.8918e-5 months , suspend all operations involving movement of irradiated fuel in the reactor vessel. (3.1.15)

PIR 2005-2672, Time To Offload K/A Statement: Knowledge of the fuel-handling responsibilities of SROs. (CFR: 41.10 /

43.7)

Technical Reference(s): TS, GEN 00-009, COLR, LO1732109 Proposed references to be provided to applicants during examination: None

Learning Objective: R3/R5 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _X__

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _10___

55.43 __7___

Comments:

Modified based on NRC comments. Changed to Rev. 5, 7/29/11 Answer: B

92. S 2.1.4 5 Given the following plant conditions:

The plant is in MODE 4

Two Licensed SROs and Two Licensed Reactor Operators are on shift

One RO becomes seriously ill and must be taken to the hospital

There are three hours left until shift change Which ONE of the following describes the required actions?

A. The affected operator must not be allowed to leave site until a relief operator arrives.

B. Shift turnover occurs before action is required.

C. Immediate action must be taken to ensure a relief operator arrives within two hours.

D. No action is required given the plants current mode of operation.

Justification A. Incorrect, must leave, must be replaced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , FFD requirements apply B. Incorrect, turnover is not allowed with an unmanned position relief must arrive within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C. Correct. AP requires that immediate action be taken to find a replacement. The position cannot remain unmanned for greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . In M 4, 2 ROs are required to be present in the control room.

D. Incorrect, must be filled in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6.9.2The Shift Crew Composition may be one less than the minimum requirements of ATTACHMENT A, MINIMUM SHIFT CREW COMPOSITION for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements.

1. This provision does not permit any shift crew position to be unmanned upon shift change due to an on-coming shift crew member being late or absent.

K/A Statement: Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2)

Technical Reference(s): TS 5.2.2, AP 21-003 (6.8 & Att. A), LO1733203, 10CFR50.54m Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X_

Comprehension or Analysis ____

10 CFR Part 55 Content:

55.41 _10___

55.43 __2___

Comments:

Modified based on NRC comments. Changed to Rev. 5, 8/1/11 Answer: C

93. S 2.2.38 4 Given the following plant conditions:

The unit is at 85% power

All AFW pumps are inoperable What is the appropriate application of TS concerning entry into MODE 3?

A. Entry into MODE 3 is permitted. The affected LCOs are not required to be entered if the unit proceeds to MODE 4 as required by LCO 3.0.3.

B. Entry into MODE 3 is permitted. The affected LCOs must be declared not met and the Required Actions must be entered upon entry into MODE 3.

C. Entry into MODE 3 is NOT permitted. When 1 train of AFW is restored, then affected LCO permits entering MODE 3.

D. Entry into MODE 3 is NOT permitted. When the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months completion time expires, then LCO 3.0.3 is entered which permits entering MODE 3.

Justification A. Incorrect, Mode 3 is not permitted, LCO's are required to be entered B. Incorrect. Mode changes are not allowed until 1 AFW train is restored. See below.

C. Correct, Mode 3 is not permitted, no delay to restore AFW D. Incorrect, Mode 3 is not permitted, not 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months time - Immediate action to restore AFW train is required LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

K/A Statement: Knowledge of conditions and limitations in the facility license. (CFR:

41.7 / 41.10 / 43.1 / 45.13)

Technical Reference(s): TS 3.7.5, TS 3.0.3, LO1732700 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _12544______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _7, 10_

55.43 __1___

Comments:

Modified based on NRC comments. Changed to Rev. 4, 7/29/11 Answer: C

94. S 2.3.13 3 What are the Shift Manager's responsibilities for ensuring safety when radiography occurs in the protected area?

A. Ensure plant wide announcements are made on the Gaitronics.

B. Verify radiography license and procedure requirements are met.

C. Contact the Health Physics Shift Technician just prior to start.

D. Ensure boundaries are set up in accordance with 29CFR1910, OCCUPATIONAL SAFETY AND HEALTH STANDARDS.

Justification A. Correct, see below B. Incorrect, Radiographer function C. Incorrect, although probably done, not a SM responsibility D. Incorrect, Radiographer function AP 25B-200 5.4 Shift Manager 5.4.1 Ensure an announcement is made on the Gaitronics stating the specific location and time radiography will commence, AND for unauthorized personnel to evacuate and stay clear of the area. [Commitment Step 3.2.1]

5.4.2 Ensure an announcement is made on the Gaitronics when commencing, moving the radiological boundaries OR when radiography is complete and normal access has been restored to the area.

K/A Statement: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 43.4 / 45.9 / 45.10)

Technical Reference(s): AP 25B-200 (section 5.4), LO1733204 Proposed references to be provided to applicants during examination: None Learning Objective: R5 Question Source: Bank # _17522______

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

55.41 _12___

55.43 __4___

Comments:

Incorporated Brendan comments, changed to Rev. 3, 6/13/11 Answer: A

95. S 2.3.5 6 Given the following plant conditions:

The unit is at 90% power Time Event 0 Loop "C" MSIV fails CLOSED 15 seconds Reactor trip 19 seconds Feedwater Isolation occurs 4.0 minutes A Steam Generator tube rupture occurs in one S/G Which ONE of the following describes the method of determining the affected S/G and the appropriate Operator action?

A.

SJ RE-02, SGBD Sampling

Reset FWI signal to allow sampling B.

BM RE-25, SGBD Process

Reset FWI signal to allow sampling C.

GE RE-92, Condenser Air Removal Discharge

Isolate all Feedwater flow to affected S/G D.

Main steam line readings taken by Health Physics

Isolate all Feedwater flow to affected S/G JUSTIFICATION:

A. Incorrect. Steam generator blowdown rad monitor is isolated by feedwater isolation (SGBSIS).

Plausible action to allow sampling.

B. Incorrect. Steam generator blowdown lines are isolated by feedwater isolation (SGBSIS). Plausible action to allow sampling, cannot identify generator with this monitor C. Incorrect. Loop "C" MSIV is closed so no steam from ruptured steam generator is seen by the condenser. correct action to prevent overfill.

D. Correct. After the SGTR, the affected loop steam line atmospheric relief setpoint is raised to minimize cycling, this does not preclude the valve lifting. When the valve starts cycling it results in the rad monitor trending up as a result of steam flow to, past the rad monitor. Appropriate action in EMG E-0.

K/A Statement: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.11 / 41.12 / 43.4 / 45.9)

Technical Reference(s): EMG E-3, BD-EMG E-3, LO1732325, SY1407300 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _______

Modified Bank # _from Practice Exam______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _11, 12_

55.43 ___4__

Comments:

Other K/A: 038 EA2.11 (3.7/3.9) {43.5}

Est. Time: 3 min.

Note: 10CFR55.41.11 includes "Purpose and operation of radiation monitoring systems, including alarms and survey equipment." NUREG-1122 contains only one objective tied to 41.11. This objective is 2.3.5

[2.9/2.9]. Therefore, the actual wording of the 10CFR55.41.11 is being used to categorize this question.

Modified stem and distractors from WE 8 question Modified based on NRC comments. Changed to Rev. 6 7/29/11 Answer: D

96. S 2.4.38 3 Given the following conditions:

A Site Area Emergency has been declared due to S/G tube rupture with offsite release of radiation

The TSC is manned, but NOT activated

The Shift Manager is the acting Site Emergency Manager Which ONE of the following responsibilities of the Shift Manager CANNOT be delegated to the TSC at the current conditions?

A. Determining protective action recommendations to give to the governmental agencies.

B. Implementing actions recommended by the Recovery/Re-entry Team.

C. Directing the search and rescue operations.

D. Deploying environmental monitoring teams.

Justification A. Correct.

B. Incorrect, actions can be delegated to the Operations Team C. Incorrect, can be delegated to the TSC, EOF D. Incorrect, responsibility of other groups The Shift Manager assumes the duties of the Site Emergency Manager upon the declaration of an Emergency Classification. While performing the duties of the Site Emergency Manager, the Shift Manager may not delegate the following responsibilities:

o Emergency Classification o Authorization of Notification of Off-site Authorities o Protective Action Recommendations o Authorization of Emergency Exposure in excess of 10CFR20 Limits EPP 06-006 - The authority to transmit protective action recommendations to the State of Kansas and Coffey County shall not be delegated by the Emergency Manager.

K/A Statement: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. (CFR: 41.10 / 43.5 / 45.11)

Technical Reference(s): EPP 06-001, EPP 06-006, EPP 06-008, GE1135628 Proposed references to be provided to applicants during examination: None Learning Objective: R2 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Incorporated Brendan comments, changed to Rev. 3, 6/13/11 Answer: A

97. S 2.4.45 6 Given the following plant conditions:

Unit is operating at 100% power

The following Control Room indications are observed:

BG TB-127, LTDN Regen HX Outlet, indicates 395°F and rising

RCP Seal Injection flows at 0 gpm on all pumps

BG FI-121A, Chg Hdr Flow, indicates 0 gpm

BG FCV-121, Charging Header Flow Control, is fully OPEN in AUTO

Letdown Flow - 75 gpm

Pressurizer level - 52% - TRENDING DOWN Based on the above indications, which ONE of the following describes the correct procedure(s) to mitigate these conditions?

A. Implement ALR 00-041A, SEAL INJ TO RCP FLOW LOW, ONLY B. Implement ALR 00-042E, CHARGING PMP TROUBLE, ONLY C. Implement ALR 00-041A, SEAL INJ TO RCP FLOW LOW, and transition to OFN BB-005, RCP MALFUNCTIONS D. Implement ALR 00-042E, CHARGING PMP TROUBLE, and transition to OFN BB-005, RCP MALFUNCTIONS Justification A. Incorrect, ALR 41A does not address the plant conditions, it will only restart a charging pump to restore seal injection only, not isolate letdown and restore it properly.

B. Correct, this ALR properly restores the plant procedurally. It isolates letdown, start the CCP, restore seal flow, and restore letdown.

C. Incorrect, wrong ALR, OFN does address the charging system issues, but does not address isolating letdown prior to restoring charging. ALR does not direct transition to OFN BB-005 for the given plant conditions. Plant conditions do not support the transitions to the OFN.

D. Incorrect, correct ALR, ALR does not direct to go to OFN and OFN would not be necessary when the ALR is completed. ALR does not direct transition to OFN BB-005.

2.2 The following are symptoms of no charging pumps running:

a. Loss of seal injection flow to the Reactor Coolant Pumps.

b. Loss of normal charging flow.

c. If charging header flow control valve is in automatic, it will fully open as pressurizer level decreases.

o BG FCV-121

K/A Statement: Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

Technical Reference(s): ALR 00-042E, SY1301000 Proposed references to be provided to applicants during examination: None Learning Objective: R1 Question Source: Bank # _47028______

Modified Bank # _______

New _

Question History: Last NRC Exam ___2008 DCPP_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Modified based on NRC comments. Changed to Rev. 6, 8/1/11 Answer: B

98. S E12 2.4.18 5 Given the following plant conditions:

A large Main Steam line break has occurred

All MSIVs failed to CLOSE

RCS has cooled down to 425°F within 15 minutes following initiation of the event

Narrow Range levels in all S/Gs are off-scale low

EMG C-21, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, is in effect and is directing AFW flow reduction Which ONE of the following describes the Auxiliary Feedwater flow requirements under these conditions and the basis for the flow requirements?

AFW Flow Basis A. 30,000 lbm/hr per S/G Extend the life of the Condensate Storage Tank B. 30,000 lbm/hr per S/G Minimize the RCS cooldown and to prevent S/G dry out C. 270,000 lbm/hr to all S/G's Control the RCS cooldown and to ensure adequate heat sink D. 270,000 lbm/hr to all S/G's Control the RCS cooldown to place RHR in service JUSTIFCATION:

A. Incorrect; BD-EMG C-21 addresses dryout which could affect SG tubes. CST life is not the basis.

Will not be able to terminate the cooldown until the S/G can be isolated.

B. Correct; BD-EMG C-21 C. Incorrect; BD-EMG C-21 states that flow should be reduced to minimum measurable flow to a SG of 30,000lbm/hr. Plausible basis not for these conditions D. Incorrect; BD-EMG C-21 states that flow should be reduced to minimum measurable flow to a SG of 30,000lbm/hr. Plausible basis not for these conditions K/A Statement: Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13)

Technical Reference(s): BD-EMG C-21, LO1732334 Proposed references to be provided to applicants during examination: None Learning Objective: R3 Question Source: Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __1___

Comments:

Other K/A: 040 AK3.04 (4.5/4.7) {41.5, 10}

E12 EK2.2 (3.6/3.9) {41.7}

E12 EK3.2 [3.3/3.9] {41.5, 41.10}

E12 EA1.3 [3.4/3.5] {41.7}

035 K1.01 [4.2/4.5] {41.4, 5, 7}

061 K1.01 [4.1/4.1] {41.4, 5, 7}

Est. Time: 2 min.

Modified based on NRC comments. Changed to Rev. 5, 7/29/11 Answer: B

99. S E13 EA2.2 5 Given the following plant conditions:

Reactor trip and SI occurred from 100% RTP

RCPs are stopped

MSIVs have closed

Containment pressure is 8 psig

B SG pressure is 1240 psig

B SG narrow range level peaked at 93% but has since fallen to 81%

Steam flow of 1.5 MPPH is indicated on "B" SG

The crew has elected to implement EMG FR-H2, RESPONSE TO STEAM GENERATOR OVERPRESSURE Which ONE of the following describes the correct action relative to releasing steam?

A. Steam may be released without restriction since narrow range level is on scale.

B. Steam release should NOT occur since natural circulation flow in other loops may be disrupted.

C. Steam release should NOT occur since damage may occur due to water trapped in the steamline.

D. Steam may be released at less than 50 psi per hour to remain within SG tube differential pressure limits.

Justification A. Incorrect. If the affected SG level has reached the upper tap, then the SG may be filled to the steamline. Decreasing the affected SG level into the narrow range does not ensure that water is removed from the affected steamline. Steamline conditions should be evaluated prior to releasing steam from any SG with level above 93% (79% Adverse) to prevent potential damage to downstream components.

B. Incorrect. Natural circulation flow is unaffected.

C. Correct. If the affected SG level has reached the upper tap, then the SG may be filled to the steamline. Decreasing the affected SG level into the narrow range does not ensure that water is removed from the affected steamline. Steamline conditions should be evaluated prior to releasing steam from any SG with level above 93% (79% Adverse) to prevent potential damage to downstream components. Transition to H3 is appropriate.

D. Incorrect. S/G-RCS differential pressure is not the concern; the concern is overpressurizing the S/G vessel.

K/A Statement: Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure) Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. (CFR: 43.5 / 45.13)

Technical Reference(s): EMG FR-H2, BD-EMG FR-H2, LO1732344

Proposed references to be provided to applicants during examination: None Learning Objective: R3/R4 Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5__

Comments:

Incorporated Jeff comments, changed to Rev. 5 8/2/11. Added RCPs are stopped to the stem. Operator could assume RCP were still running because pressure was above 1400 psig Answer: C

100. S E15 2.4.1 4 Given the following plant conditions:

A large break LOCA has occurred

One Containment Spray Pump was secured per EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT

EMG ES-12, TRANSFER TO COLD LEG RECIRCULATION, has just been completed when the STA reports the following conditions:

Containment Pressure is 26 psig decreasing

Containment Radiation is 10 Rad/hr

Recirculation sump level is 2004 Which ONE of the following describes the immediate Containment concern and the correct procedure to enter?

A. Inadequate suction to the RHR pumps, transition to EMG C-11, LOSS OF EMERGENCY COOLANT RECIRCULATION.

B. Flooding vital equipment in Containment; transition to EMG FR-Z2, RESPONSE TO CONTAINMENT FLOODING.

C. Erroneous instrumentation readings, transition to EMG FR-Z3, RESPONSE TO HIGH CONTAINMENT RADIATION LEVEL, when desired.

D. Containment structural integrity; transition to EMG FR-Z1, RESPONSE TO HIGH CONTAINMENT PRESSURE.

Justification A. Incorrect, Ctmt sump level is adequate, Loss of emergency coolant recirculation is not the procedure that is required to be entered with these conditions.

B. Correct, Containment sump level is high and flooding is a concern and level has reached the threshold value to enter FR-Z2.

C. Incorrect, Radiation levels are high, but FR-Z3 is entered on operator discretion and sump level is a higher priority.

D. Incorrect, Pressure is somewhat high, however with one CTMT Spray Pump running it merits a orange path EMG FR-Z1 K/A Statement: Knowledge of EOP entry conditions and immediate action steps. (CFR:

41.10 / 43.5 / 45.13)

Technical Reference(s): EMG F-0 Figure 6, EMG FR-Z2, BD-EMG FR-Z2, LO1732351 Proposed references to be provided to applicants during examination: None Learning Objective: R2

Question Source: Bank # _______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

10 CFR Part 55 Content:

55.41 _10___

55.43 __5___

Comments:

Incorporated NRC comments, changed to Rev. 4 7/29/11 Answer: B

ANSWER KEY REPORT for NRC Test Form: 0 Answers

ID Points Type 0 1 001 AK2.06 4 1.00 MCS B 2 002 K4.01 6 1.00 MCS B 3 003 A4.05 5 1.00 MCS A 4 003 AA2.02 5 1.00 MCS A 5 003 K2.01 5 1.00 MCS A 6 004 K3.07 3 1.00 MCS D 7 005 A4.01 6 1.00 MCS A 8 005 AK3.01 6 1.00 MCS D 9 005 K2.03 4 1.00 MCS B 10 006 K3.01 4 1.00 MCS D 11 007 2.1.32 4 1.00 MCS B 12 007 A2.01 7 1.00 MCS B 13 007 EK2.03 6 1.00 MCS D 14 008 AK2.02 7 1.00 MCS C 15 008 K 1.027 1.00 MCS C 16 009 EK2.03 7 1.00 MCS C 17 010 K5.01 7 1.00 MCS C 18 011 EA2.04 4 1.00 MCS C 19 011 K6.04 6 1.00 MCS A 20 012 K4.08 5 1.00 MCS C 21 013 K4.09 4 1.00 MCS D 22 014 A1.04 6 1.00 MCS D 23 015 K3.02 5 1.00 MCS A 24 015/017 AA2.01 5 1.00 MCS D 25 016 K5.01 7 1.00 MCS D 26 022 A1.04 6 1.00 MCS C 27 022 A3.01 3 1.00 MCS C 28 022 AK1.02 3 1.00 MCS A 29 025 AK3.02 6 1.00 MCS B 30 026 2.4.2 6 1.00 MCS A 31 026 A2.07 4 1.00 MCS C 32 027 2.2.22 5 1.00 MCS C 33 027 K2.01 5 1.00 MCS B 34 028 A2.01 4 1.00 MCS B 35 028 AK2.03 6 1.00 MCS C 36 029 A4.01 4 1.00 MCS C 37 034 2.2.125 1.00 MCS D 38 035 A3.01 5 1.00 MCS B 39 037 AA1.04 1 1.00 MCS A 40 038 EA1.15 6 1.00 MCS B 41 039 K5.08 4 1.00 MCS D 42 054 AK1.02 5 1.00 MCS B 43 051 2.4.31 3 1.00 MCS C 44 040 AK1.03 4 1.00 MCS C 45 055 EA 1.01 7 1.00 MCS C 46 059 K4.17 5 1.00 MCS C 47 061 AK3.02 4 1.00 MCS C

Answer

ID Points Type 0 48 061 K6.01 5 1.00 MCS D 49 062 2.1.28 4 1.00 MCS A 50 062 A1.03 6 1.00 MCS C 51 063 2.1.31 4 1.00 MCS B 52 063 K3.02 4 1.00 MCS C 53 064 K6.08 7 1.00 MCS D 54 069 AK 1.01 4 1.00 MCS D 55 073 2.1.30 6 1.00 MCS D 56 073 K1.013 1.00 MCS A 57 076 A 1.023 1.00 MCS D 58 077 AA2.09 1 1.00 MCS D 59 078 A3.01 5 1.00 MCS D 60 078 K1.04 2 1.00 MCS A 61 103 A2.04 8 1.00 MCS C 62 2.1.18 4 1.00 MCS D 63 2.1.36 7 1.00 MCS D 64 2.1.45 4 1.00 MCS D 65 2.2.41 5 1.00 MCS B 66 2.2.43 4 1.00 MCS A 67 2.3.12 6 1.00 MCS C 68 2.3.7 3 1.00 MCS D 69 2.4.21 2 1.00 MCS C 70 2.4.47 2 1.00 MCS A 71 2.4.9 3 1.00 MCS B 72 E01 EA2.2 3 1.00 MCS B 73 E04 EA1.2 2 1.00 MCS D 74 E05 EK3.2 1 1.00 MCS B 75 E11 EK3.2 6 1.00 MCS C 76 S 006 2.4.6 4 1.00 MCS A 77 S 008 A2.05 4 1.00 MCS B 78 S 010 A2.03 7 1.00 MCS A 79 S 029 EA2.01 1 1.00 MCS A 80 S 033 2.2.25 6 1.00 MCS A 81 S 045 A2.17 4 1.00 MCS B 82 S 056 AA2.03 5 1.00 MCS C 83 S 057 2.4.8 5 1.00 MCS B 84 S 058 2.4.3 6 1.00 MCS A 85 S 062 2.1.25 7 1.00 MCS DELETED 86 S 064 A2.06 8 1.00 MCS A 87 S 065 AA2.01 6 1.00 MCS A 88 S 068 2.1.20 7 1.00 MCS D 89 S 074 EA2.01 9 1.00 MCS D 90 S 001 A2.04 1 1.00 MCS C 91 S 2.1.35 5 1.00 MCS B 92 S 2.1.4 5 1.00 MCS C 93 S 2.2.38 4 1.00 MCS C 94 S 2.3.13 3 1.00 MCS A

Answer

lD Points Type 0 95 S 2.3.5 6 1.00 MCS D 96 S 2.4.38 3 1.00 MCS A 97 S 2.4.45 6 1.00 MCS B 98 S E12 2.4.18 5 1.00 MCS B 99 S E13 EA2.2 5 1.00 MCS C 100 S E15 2.4.1 4 1.00 MCS B

ML112700887: Difference between revisions

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