ML16034A193

Final Written Examination with Answer Key (401-5 Format)(Folder 2)

ML16034A193
Person / Time
Site:	Beaver Valley
Issue date:	12/10/2015
From:	Peter Presby Operations Branch I
To:	Gibson D FirstEnergy Nuclear Operating Co
Shared Package
ML15161A011	List: IR 05000412/2016301 ML15194A008 ML15334A437 ML15345A016 ML15355A238 ML15355A239 ML15355A241 ML15355A242 ML15355A243 ML15355A244 ML16034A138 ML16034A193 ML16034A218 ML16034A473 ML16042A442
References
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Download: ML16034A193 (114)

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ES-401 Form ES-401-7 Site Specific RO Written Examination Cover Sheet U. S. Nuclear Regulatory Commission Site Specific RO Written Examination BV2LOT15 RO Written Exam Applicant Information Name:

Date: Facility/Unit: Beaver Valley Unit 2 Region: I lR1 11 0 111 0 IV 0 Reactor Type: W [R] CE 0 BW 0 GE 0 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results Examination Values 15 Points Applicant's Scores Points Applicant's Grade Percent NUREG-1021, Revision 10 FENOC Facsimile r2

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

1. The plant is at 100% power with all systems in normal alignment EXCEPT:

• Power Range Channel N-44 has been declared inoperable

• Power Range Channel N-44 has been removed from service IAW AOP-2.2.1 C, "Power Range Channel Malfunction" Power Range Channel N-43 NOW fails HIGH.

• All systems function as designed

• No Operator Actions have been taken Which, of the below listed First Out Annunciators (ANN. A5), will alarm in the FIRST 45 seconds AFTER N-43 fails High?

(1) A5-1 D 2/3 Loops Overtemp AT Reactor Trip (2) A5-2A Reactor Protection System Train A Trouble (3) A5-5G Reactor Trip Due To Turbine Trip (4) A5-68 Turbine Anti-Motoring Turbine Trip (5) A5-6D Turbine Trip Due To Reactor Trip (6) A5-7D Generator Trip Due To Turbine Trip A. 3, 5, 6 ONLY B. 2, 4, 6 ONLY C. 1, 3, 5, 6 ONLY D. 1, 2, 3, 4 ONLY Answer: A Explanation/Justification: K/A is met because the candidate must determine which First Out Annunciators on the A5 panel (Rx trip status panel) will alarm 45 seconds after a Rx Trip occurs.

A. Correct. The reactor will trip due to actions of N-44 failing having been completed which places N-44 bistable to TRIP, then N-43 fails High meeting the required 2/4 PR high Rx trip coincidence. These 3 are normal Rx trip annunciators, and reasons for the other annunciators are not correct are given in the other explanations.

B. Incorrect. (2) is plausible because candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent will energize on the trip. (4) Anti-motoring would alarm if the output breakers did not open. Plausible if the stem of the question didn't state that all systems functioned as designed.

C. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into OT6T trip setpoint calculation. Therefore this alarm will NOT be energized.

D. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into OT6T trip setpoint calculation. (2) is plausible because candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent will energize on the trip. (4) Anti-motoring would alarm if the output breakers did not open. Plausible if the stem of the question didn't state that all systems functioned as designed.

Sys # System Category KA Statement 000007 Reactor Trip /1 EK2 Knowledge of the interrelations between a reactor trip and the following: Reactor trip status panel KIA# EK2.03 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

1.4.AAD, 26.4.AAI and 35.4.AAF UFSAR Fig. 7.3-8 Rev. 14 Question Source: Bank - 2LOT6 NRC Exam (01) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

2. The plant was at 100% power when the reactor tripped on low PRZR pressure. The crew suspects that a PORV opened inadvertently and is now stuck partially open.

Current conditions:

• PRZR pressure is 1885 psig and stable

• [2RCS-P1472], Pressurizer Relief Tank pressure indicates 35 psig.

1) Based on the conditions given, which of the following confirming indications could be expected if a PORV is stuck partially open?

2) High pressure on which of the following tanks will prevent 2RCS-MOV523, Pressurizer Relief Tank Drain Valve from opening?

A. 1) PORV relief line temperature stabilized at 259°F

2) Pressurizer Relief Tank B. 1) PORV relief line temperature stabilized at 281°F

2) Pressurizer Relief Tank C. 1) PORV relief line temperature stabilized at 259°F

2) Primary Drains Transfer Tank D. 1) PORV relief line temperature stabilized at 281°F

2) Primary Drains Transfer Tank Answer: D Explanation/Justification: KIA is met because the partially open PORV will be discharging PRZR vapor space to the PRT. The candidate will have to evaluate the condition to determine what the temperature is at the outlet of a throttled valve (PORV) based on the determined saturation pressure, then using knowledge of the interlocks associated with the PRT drain valve, determine high pressure in the Primary Drains Transfer Tank will prevent the valve from opening.

A. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure= 50 psia). Plausible distractor because PRT high pressure will prevent the PRT spray valve from opening on high pressure.

B. Incorrect. 281°F is the saturation temperature corresponding to 50 psia. Plausible distractor because PRT high pressure will prevent the PRT spray valve from opening on high pressure.

C. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure = 50 psia). PRT drain valve will not open if the Primary Drains Transfer Tank pressure is high.

D. Correct. 281°F is the saturation temperature corresponding to 50 psia. PRT drain valve will not open if the Primary Drains Transfer Tank ressure is hi h.

Sys# System Category KA Statement 000008 Pressurizer (PZR) Vapor Space AK1. Knowledge of the operational Thermodynamics and flow characteristics of open Accident (Relief Valve Stuck Open) I 3 implications of the following concepts as or leaking Valves they apply to a Pressurizer Vapor Space Accident:

KIA# AK1 .01 KIA Importance 3.2 Exam Level RO References provided to Candidate steam Tables Technical

References:

Steam Tables 20M-6.1.D Rev. 3 pg.8 Question Source: Bank- Vision #135831 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

3. The plant is at 100% power.

• Annunciator A2-5C, Reactor Coolant Pump Vibration Alert/Danger Alarms

• 'B' RCP shaft vibration is 16 mils and stable

• 'B' RCP frame vibration is 1 mil and stable

• The crew enters AOP-2.6.8, "Abnormal RCP Operation" While performing the actions of AOP-2.6.8, the following additional alarms and indications are received:

• A2-5D, Reactor Coolant Pump Seal Vent Pot Level High/Low (RCP 21 B Seal Pot Lvl High, computer address point L0508D)

• A2-4D, Reactor Coolant Pump Seal Trouble (RCP 21 B Seal Lk Off, 2CHS-FT1558 Low, computer address point F0128D)

• RCP 21 B Seal Lk Off, 2CHS-FT1558 is 0.80 gpm and stable Based on these alarms and indications, which 'B' RCP seal has failed?

A. #1 seal B. #2 seal C. #3 seal D. Low pressure seal Answer: B Explanation/Justification: KIA is met because the 'B' RCP has a high vibration condition which leads to a coolant pump seal failure. The candidate must analyze the indications given to determine which seal has failed due the malfunction.

A. Incorrect. If #1 seal had failed seal leak-off flow would be high NOT low.

B. Correct. IAW 20M-7.4.AAH, 6.4.AAE and AOP-2.6.8 C. Incorrect. If #3 seal had failed the seal vent pot level low would be indicated NOT high.

D. Incorrect. The low pressure seal is not functional when the motor is coupled to the pump.

Sys# System Category KA Statement 000015/0 Reactor AK2. Knowledge of the interrelations between the Reactor Coolant RCP seals 00017 Coolant Pump Malfunctions (Loss of RC Flow) and the following:

Pump (RCP)

Malfunctions I 4

KIA# AK2.07 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.AAE Rev. 12 20M-7.4.AAH Rev. 22 Question Source: Bank - 2LOT6 NRC Exam (04) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

4. The plant is at 100% power.

• 2CHS*MOV289, 'Normal Charging Header lsol Viv' fails CLOSED

• The Immediate Operator Actions of the appropriate AOP have been performed

• NO other actions have been completed

• Total Seal Injection flow is 20 gpm

• Total Seal Water Leakoff is 9 gpm

1) When will A4-1C, Pressurizer Control Level Deviation High/Low alarm?

2) Will the alarm result from high or low deviation?

A. 1) 22 - 28 minutes

2) High

8. 1) 41 - 49 minutes

2) High C. 1) 4 - 6 minutes

2) Low D. 1) 10 - 15 minutes

2) Low Answer: B Explanation/Justification: KIA is met because the candidate must determine how the PRZR level will respond to a loss of charging and letdown (due to AOP IOAs). The candidate must realize that with the charging pump still in service (given that seal injection flow is 20 gpm) and seal leakoff flow, PRZR level will still change even with a Loss of Reactor Coolant Makeup flowpath.

Net charging= (Charging flow)+ (Total seal inj flow) - (Letdown flow) - (Total seal leakoff), 1% PRZR level =100 gal.

PRZR level deviation setpoint is +/-5% above/below program level A. Incorrect. It will take 25 minutes to receive the HIGH deviation if it is assumed that Letdown is isolated and all seal injection flow is entering the RCS/PRZR. The misconception is that 9 gpm is seal leakoff going back to charging pump suction ..

B. Correct. It will take 45 minutes to receive the HIGH deviation. (0 charging) + (20 seal inj) - (Letdown) - (9 leakoff) = 11 gpm net charging.

500 gal/11 gal=45 min to raise PRZR 5% above program level.

C. Incorrect. It would take -5 minutes to receive the LOW deviation if it is not recognized that Letdown was isolated as an IOA and misconceptions of seal injection volume entering the RCS. 0 + 20 -105- 9 = (500/94 gpm) = 5.3.

D. Incorrect. It would take -10 minutes to receive the LOW deviation if it is not recognized that Letdown was isolated (only 1 orifice in service) as an IOA and misconceptions of seal injection volume entering the RCS. 0 + 20 9 = (500/49 gpm)= 10.2, or using 45 gpm orifice (500/34)=14.7.

Sys# System Category KA Statement 000022 Loss of Reactor AK 1. Knowledge of the operational implications of the following Relationship between charging flow and Coolant Makeup I 2 concepts as they apply to Loss of Reactor Coolant Makeup: PZR level KIA# AK1 .03 KIA Importance 3.0 Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2. 7.1 Rev 7 None 20M-6.4.AAL Rev. 8 20M-6.1.C Rev. 5 pg.12 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

5. The following conditions exist:

• Plant is in Mode 5

• A1-2G, INCORE INSTR ROOM/CNMT SUMP LEVEL HIGHNALVE NOT RESET is in alarm

• RHR HX A INLET TEMP is 113°F and STABLE

• 2RHS*FCV605A, RHS TRAIN A HX BYPASS FLOW CONTROL valve has OPENED an additional 5% in response to the leak Based on the given conditions:

1) Which of the following could identify the location of the leak in the RHR system?

2) Which of the following procedures would be used to isolate the affected train of RHR?

A. 1) RHS*MOV720A, 'RHS Train Return to B Loop Isolation' INLET flange

2) AOP-2.10.1, "Loss of Residual Heat Removal Capability" B. 1) RHS*MOV720A, 'RHS Train Return to B Loop Isolation' INLET flange

2) AOP-2.6.5, "Shutdown LOCA" C. 1) 2RHS-E21A, "A' RHR HX' OUTLET flange

2) AOP-2.10.1, "Loss of Residual Heat Removal Capability" D. 1) 2RHS-E21A, "A' RHR HX' OUTLET flange

2) AOP-2.6.5, "Shutdown LOCA" Answer: C Explanation/Justification: KIA is met by the following. Based on the response of the Bypass Flow Control valve, the candidate must determine the leak location within the RHR system. Once the location is determined, the candidate must decide which AOP would be used to mitigate and isolate the leak in the RHR system.

A. Incorrect. A leak at the inlet of RHS-MOV720A is downstream of (FT605A) flow element, therefore the loss in flow would not cause FCV-605A to respond because it is not seen by the flow element. AOP-2.10.1 is the correct procedure for plant conditions.

B. Incorrect. A leak at the inlet of RHS-MOV720A is downstream of (FT605A) flow element, therefore the loss in flow would not cause FCV-605A to respond because it is not seen by the flow element. AOP-2.6.5 is used for a loss of coolant accident when in mode 3 (after the accumulators are isolated) or mode 4.

C. Correct. A leak at the outlet of the RHR Hx will cause flow to be lower at (FT605A) flow element downstream of the Hx and FCV. This reduced flow will cause FCV-605A to open to raise flow back to the desired setpoint. AOP 2.10.1 is the correct procedure when there is a loss of coolant accident in mode 5 & 6.

D. Incorrect. This is the correct leak location, but the incorrect procedure for the plant conditions. AOP-2.6.5 is used for a loss of coolant accident when in mode 3 (after the accumulators are isolated) or mode 4.

Sys# System Category KA Statement 000025 Loss of Residual Heat AA2. Ability to determine and interpret the following as they Location and isolability of leaks Removal System apply to the Loss of Residual Heat Removal System:

(RHRS) / 4 KIA# AA2.04 KIA Importance 3.3* Exam Level RO References provided to Candidate None Technical

References:

RM-0410-001 Rev. 16 20M-53C.4.2.10.1 rev. 12 pgs.1 & 14 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

6. The plant was operating at 100% power when a Loss of Coolant Accident occurred.

The following conditions exist:

• The Reactor automatically tripped

• RCS pressure is 1190 psig and lowering

• All SG pressures are 950 psig and lowering

• Containment pressure peaked at 13 psig and is currently 9.5 psig and lowering

• No HHSI is indicated

• The crew is performing the actions of E-0, "Reactor Trip and Safety Injection" In accordance with E-0, what is the expected crew response for the Reactor Coolant Pumps (RCPs), and why?

1) The RCPs should be ~~~~~~~~

2) Because _ _ _ _ _ __

A. 1) left running

2) to continue pumping the water/steam mixture through the core B. 1) tripped

2) D/P between RCS pressure and highest SG pressure trip criteria has been met C. 1) tripped

2) thermal barrier and motor cooling flow was secured D. 1) tripped

2) the RCP Seal Water Return Isolation Valves, 2CHS*MOV378 and 381 are isolated Answer: C Explanation/Justification: KIA is met because the candidate must know that E-0 gives the guidance to trip the RCPs if a CIB occurs. The reason for the trip is that primary component cooling water is isolated to the thermal barrier and motor.

A. Incorrect. Plausible distractor is the candidate does not recognize that a CIB has occurred and isolated CCP to the RCP motor. The RCP operation in a water/steam 2 phase condition possibly could occur in FR-C.1, but not E-0.

B. Incorrect. Tripping the RCP is correct, but for the incorrect reason. Trip criteria is 220 psid with adverse conditions and HHSI flow exists. The stem states no HHSI flow, and DP is 240 psid.

C. Correct. With Containment pressure peaking at 13 psig, CIB occurred and isolated (CCP) to both 'A' & 'B' containment headers. This will remove cooling water to the thermal barrier, upper & lower bearing cooler, and the stator cooler. In accordance with E-0 LHP, it is required to trip the RCPs on a loss of CCP flow to the RCPs.

D. Incorrect. Tripping the RCPs is correct, but it is not a correct statement to trip because RCP Seal Water Return Isolation Valves have closed on CIA. Seal Return Header Relief 2CHS-RV382A provides protection for the low pressure RCP seal return piping when this header is isolated (CIA),

and relieves to the PRT.

Sys # System Category KA Statement 000026 Loss of Component AK3. Knowledge of the reasons for the following responses as Guidance actions contained in EOP for Loss of Cooling Water they apply to the Loss of Component Cooling Water: CCW (CCW)/ 8 KIA# AK3.03 KIA Importance 4.0 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.E-O lss. 2 Rev.1 LHP 20M-53B.5.Gl-6 lss. 2 Rev.a pg. 51 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 / 45.6 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

7. The plant is at 75% power with all systems in normal alignment for this power level EXCEPT

[2RCS*MOV535] PORV 455C MOTOR OPERATED ISOL VLV is CLOSED due to PORV seat leakage.

• [2RCS*PK444A] PRZR PRESS CONTROL output fails to 10% in Automatic

• No Operator action is taken What is the status of the plant 15 minutes after this event?

A. The plant will trip due to a PRZR Pressure LOW reactor trip.

B. The plant will trip due to a PRZR Pressure HIGH reactor trip.

C. The plant will remain at power and RCS pressure will cycle between the PORV lift and PORV lift reset setpoints.

D. The plant will remain at power and RCS pressure will cycle between the PORV lift and PRZR Low Pressure PORV Block Interlock setpoints.

Answer: C Explanation/Justification: K/A is met by the knowledge required to determine that the purpose of 2RCS*PK444A is to control PRZR pressure and how the controller will function when it fails to 10%. This knowledge will encompass the response of the PORVs in the non-affected portion of the Pressurizer Pressure Control System, and the control function side with the heaters and spray valves, and the resulting 455C PORV being isolated.

A. Incorrect. Plausible if 2RCS*PK444 output failed high, causing both spray valves to open fully and depressurize the RCS until LP Rx trip occurs.

B. Incorrect. Plausible distractor because pressure would rise to 2375 psig and cause a HP Rx trip if the PORVs didn't lift at 2335 psig. Candidate must know that the Rx trip setpoint is higher than the PORV opening setpoint, and that the PORV will prevent the trip setpoint from being reached.

C. Correct. The plant will remain at power and the pressure will cycle between the PORV auto open setpoint of 2335 psig and close setpoint of 2315 psig. This will occur only on PORV 456 & 4550 since 455C is manually isolated.

D. Incorrect. Plausible distractor if it is not recognized that the 456 & 4550 PORVs will auto close at 2315 psig, and they think that pressure will reduce to 2185 psig causing the block valves to close.

Sys# System Category KA Statement 000027 Pressurizer Pressure Control Generic Knowledge of the purpose and function of major System (PZR PCS) system components and controls.

Malfunction I 3 KIA# 2.1.28 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF Att.2 Rev.13 2om-6.2.B Rev. 13 pg.8 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)1 Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

8. The crew is implementing FR-S.1, "Response to Nuclear Power Generation/ATWS".

Which of the following combinations of breaker positions indicate that the reactor has been tripped?

Note: The Motor Generator INPUT Breakers are both closed.

• RTA =Reactor Trip Breaker A

• RTB =Reactor Trip Breaker B

• BYA =Reactor Trip Bypass Breaker A

• BYB =Reactor Trip Bypass Breaker B

• MG21 =MG21 output Breaker

• MG22 =MG22 output Breaker LEGEND: X =CLOSED; 0 =OPEN RTA RTB BYA BYB MG21 MG22 A. x x 0 0 x 0 B. x 0 0 x 0 x C. 0 x x 0 x x D. x 0 x 0 x x Answer: D Explanation/Justification: KIA is met by the knowledge required to determine which combination of Rx trip, bypass, or motor generator output breaker positions will trip the reactor during an ATWS.

A. Incorrect. Only one MG set breaker is open. Both required to be open to trip reactor.

B. Incorrect. RTA and BYB closed will provide flowpath to the rod coils. Only one MG set breaker is open. Both required to be open to trip reactor.

C. Incorrect. RTB and BYA closed will provide flowpath to the rod coils.

D. Correct. With both RTB and BYB open, the MG set output supply to the rod coils is interrupted, which will result in the rods dropping into the core (a reactor trip).

Sys# System Category KA Statement 000029 Anticipated EK2 Knowledge of the interrelations between the and the following Breakers, relays, and disconnects Transient anATWS:

Without Scram (ATWS)/ 1 KIA# EK2.06 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

Westinghouse 2001.409-001-018 Rev. L Question Source: Bank - DC Cook 2012 NRC Exam (046)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 I 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

9. The plant is operating at 50% power with the crew implementing AOP-2.6.4, "Steam Generator Tube Leakage" for a 50 GPO tube leak on 'A' SG.

A Steam leak inside CNMT occurs requiring a Rx Trip.

• E-0, Reactor Trip or Safety Injection" has been entered

• Safety Injection is actuated Assuming no operator actions other than E-0 immediate operator actions, which of the following describes why 2SSR*AOV117A, '21A SG Slowdown Sample Outside Cnmt lsol Viv' would automatically close?

A. Low RCS pressure, to provide containment isolation.

B. Low SG pressure, to limit excess steam demand effect.

C. High containment pressure, to establish CNMT barrier integrity.

D. High secondary activity, to minimize radiological release.

Answer: D Explanation/Justification: KIA is met by the knowledge required to determine that 2SSR*AOV117A '21A SG Slowdown Sample Outside Cnmt lsol Viv' will automatically close due to a High radiation monitoring alarm on 2SSR-RQ100, and the reason for the isolation is to minimize radiological releases from the ruptured SG.

A. Incorrect. 2SSR*AOV117A is not closed by SI (CIA) signals which actuate on RCS low pressure. Plausible distractor with the steam leak, and a tube leak identified in the stem which both could possibly cause RCS pressure to lower to 1856 psig actuate SI, but has no effect on AOV117A.

B. Incorrect. 2SSR*AOV117A is not closed by SI, CIA, MSI signals which actuate on low SG pressure. Plausible distractor with the steam leak identified in the stem which could cause SG pressure to lower to 500 psig and actuate SI or MSI, but has no effect on AOV117A.

C. Incorrect. 2SSR*AOV117 A is not closed by SI, CIA, MSI, or CIB actuation signals which actuate on cnmt pressure. Plausible distractor with the steam leak inside cnmt identified in the stem which could cause cnmt pressure to raise and actuate SI or MSI, but has no effect on AOV117A.

D. Correct. 2SSR*AOV117A is closed by 2SSR-RQl100 High Alarm. Minimize radiological releases from the ruptured SG is the correct reason.

Sys# System Category KA Statement 000038 Steam Generator EK3 Knowledge of the reasons for the following responses as the Automatic actions associated with high Tube Rupture apply to the SGTR: radioactivity in S/G sample lines.

(SGTR) / 3 KIA# EK3.03 KIA Importance 3.6* Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.E-3 lss. 2 Rev. 4 20M-43.1.E Rev. 6 U2 RM-0414A-001 Rev. 18 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5/41.10145.61 45.13)

Objective: 2SQS-43.1 Obj. 7 Describe the control, protection and interlock functions for the control room components associated with the Radiation Monitoring System, including automatic functions, and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

10. The following conditions exist:

• The plant is at 75% power with all systems in normal alignment for this power level

• Charging pump 2CHS*P21A is RUNNING

• Condensate pumps 2CNM-P21A and Bare RUNNING A loss of 4160V Bus 20 occurs with all equipment operating as designed.

What procedure will the crew use to mitigate this event?

A. E-0, "Reactor Trip Or Safety Injection" B. AOP 2.24.1, "Loss Of Main Feedwater" C. AOP 2.7.1, "Loss Of Charging Or Letdown" D. AOP 2.36.2, "Loss of 4KV Emergency Bus" Answer: B Explanation/Justification: KIA is met by having the candidate determine what equipment will be lost based on interpreting the conditions given, after which they will determine that only 1 MFW pump was lost, and Loss Of Main Feedwater AOP is the appropriate procedure.

A. Incorrect. Plausible because the candidate may think a RCP will be lost, or a condensate pump would trip ('C' is powered by 'D' Bus}, or if they are thinking that we lost the 'B' MFP, without knowing that the AOP states <80% power, then reduce power to <52%.

B. Correct. With the plant at 75% power, AOP 2.24.1 is the correct procedure to mitigate the loss of 1 MFP when at power and <80%. The candidate should know by AOP-2.24.1 IOAs that the Rx should not be tripped if the plant is <80% power.

C. Incorrect. Plausible if the candidate thinks 2CHS-P21A is lost due to the bus loss, but this is incorrect.

D. Incorrect. Plausible if the candidate does not think the 'OF' bus will be energized by the 2-2 EOG. This is incorrect because the stem states that all equipment operated as designed.

Sys# System Category KA Statement 000054 Loss of Main AA2. Ability to determine and interpret the following as they apply to Differentiation between loss of all MFW and trip of Feedwater the Loss of Main Feedwater (MFW): one MFW pump (MFW}/4 KIA# AA2.02 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.24.1 Rev. 6 pg. 2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

11. The following conditions exist:

• The plant has tripped due to a loss of all 4Kv power

• The crew is performing ECA-0.0, "Loss of All AC Power"

• The BOP Operator is placing equipment in PTL in accordance with step 16 of ECA-0.0 Which of the following components will NOT be placed in PTL during this step, and what is the basis for not removing this component from service?

A. Auxiliary Feedwater pump; provide sufficient water to maintain an effective secondary heat sink.

B. Primary Component Cooling Pump; provide cooling to the Reactor Coolant Pump thermal barrier.

C. Charging Pump; provide cooling to the Reactor Coolant Pump seals.

D. Service Water Pump; provide cooling to the Emergency Diesel Generator.

Answer: D Explanation/Justification: KIA is met by the knowledge that ECA-0.0 places equipment to PTL to defeat automatic loading of large loads on the AC emergency bus with the exception of the Service Water Pumps. The knowledge of the bases of the SW pump remaining available to load on a diesel start to provide diesel cooling is expected RO knowledge.

A. Incorrect. MDAFW pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. During ECA-0.0, sufficient AFW flow is provided by the turbine driven AFW pump, so heat sink is not a concern.

B. Incorrect. CCP pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. Providing cooling flow to the hot thermal barrier is not required at this time.

C. Incorrect. Charging pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. Providing cooling flow to the hot RCP seals could cause thermal shock to the seals & shaft.

D. Correct. Service water pump auto start is desired to provide cooling to the EOG in the event it is locally restored. This is stated as a caution prior to step 13 and switches are placed to Auto in step 1 of Att. A-1.5 for starting the diesel locally.

Sys# System Category KA Statement 000055 Loss of Offsite and Generic Knowledge of the specific bases for EOPs.

Onsite Power (Station Blackout) I 6 KIA# 2.4.18 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ECA-O.O lss. 2 Rev. 3 pgs. 5 & 10 20M-53B.4.ECA-O.O lss. 2 Rev. 3 pg. 124 20M-53A.1.A-1.5 lss. 1C Rev. 5 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.1 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

12. The plant has been operating at 100% power with all systems in NSA for the past 100 days.

• An inadvertent reactor trip occurs coincident with a loss of offsite power

• All systems function as designed

• The crew is implementing the actions of ES-0.2, "Natural Circulation Cooldown"

• RCS temperature is 400°F lowering at 20°F/hr

• RCS Subcooling is 165°F

• RCS Pressure 1200 psig and stable Alarm A11-5G, CROM Shroud Fan Auto-Start/Auto-Stop is received. ALL CROM shroud fans have tripped and cannot be restarted.

What ramifications will the loss of these CROM Shroud Fans have on the continued performance of ES-0.2, "Natural Circulation Cooldown"?

A. Further RCS cooldown (below 350°F) cannot continue UNTIL a suitable RX vessel head soak has been performed.

B. Further RCS depressurization (below 1200 psig) cannot continue UNTIL a suitable RX vessel head soak has been performed.

C. Immediately INCREASE RCS pressure 100 psig to RAISE RCS subcooling.

D. Immediately DECREASE RCS pressure 100 psig to LOWER RCS subcooling.

Answer: B Explanation/Justification: KIA is met with the knowledge of ES-0.2 major action step of the requirement to cooldown and depressurize RCS with no upper head void growth when RCS forced flow is lost. Knowledge that the CROM cooling units offer alternative cooling to the upper vessel head region during NC cooldown since core bypass flow existing with forced convection is lost to the upper head region is important to cooling down the head.

A. Incorrect. With the CROM fans lost, the RCS pressure is held at 1200 psig while the RCS is cooled down below 350F. At these conditions there is approx. 200F of subcooling is required for the RCS due to the heat buildup in the head. The 9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> soak time will ensure the head cools off sufficiently since the CROM fans were lost B. Correct. IAW ES-0.2 step 15. Without any CROM fans running, subcooling requirements will be more than 3 times greater (200F). The CROM fans provide alternate cooling to the upper vessel head region during NC CID since core bypass flow existing with forced convection CID is lost in the upper vessel head region. Since this forced cooling is lost, a 9 hr soak is required to ensure the head cools to less than saturation temp for 400 psig.

C. Incorrect. Raising pressure 100 psig is a technique employed by ES-0.4 natural circulation procedure when the head void growth becomes too large.

D. Incorrect. Minimizing Subcooling is a technique employed when RX vessel stresses are the concern but NOT when RX vessel head voids are the concern. Decreasing pressure may actually cause a void to form.

Sys # System Category KA Statement 000056 Loss of AK1. Knowledge of the operational implications of the following Principle of cooling by natural convection Offsite Power concepts as they apply to Loss of Offsite Power:

16 KIA# AK1 .01 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-0.2 lss. 2 Rev. 1 step 15 20M-53B.4.ES-0.2 lss. 2 Rev. 1 pg. 51 Question Source: Bank-2LOT6 NRC Exam (Q27) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

13. Reactor power is at 27% with all systems in normal alignment EXCEPT:

• Rod control is in MANUAL A loss of Vital Bus 2 occurs.

Which of the following indications on Panel 308, PRI PLANT PARAMETERS STATUS panel will be LIT due to the loss of Vital Bus 2?

1) PWR RNG N41 TO P-8

2) PWR RNG N42 HIGH POWER SP

3) RCS LOOP A OT~T RX TRIP

4) SOURCE RNG N32 RX TRIP A. 1 and 3 B. 2 and 4 C. 1and4 D. 2 and 3 Answer: B Explanation/Justification: KIA is met by the candidate interpreting the effects of a loss of vital bus 2 will have on the RPS system, and determining which RPS panel alarms and trip indications will illuminate due to the loss of power/equipment out of service.

A. Incorrect. N41 to P-8 is normally lit when >30% power, but power is 27% in the stem and it is powered from VB1. RCS LOOP A OTL1T RX TRIP is CH1 powered from VB1 and not effected.

B. Correct. PWR RNG N42 HIGH POWER SP and SOURCE RNG N32 RX TRIP are both normally off when not above a setpoint condition. They are both CH2, so in this case, a loss of VB2 will cause the bistable lights to illuminate.

C. Incorrect N41 to P-8 is normally lit when >30% power, but power is 27% in the stem and it is powered from VB1. SOURCE RNG N32 RX TRIP will be lit due to the loss of VB2.

D. Incorrect. PWR RNG N42 HIGH POWER SP is CH2, so in this case, a loss of VB2 will cause the bistable light to illuminate, but RCS LOOP A OTL1T RX TRIP is CH1 powered from VB1 and not effected.

Sys# System Category KA Statement 000057 Loss of Vital AC AA2. Ability to determine and interpret the following as they RPS panel alarm annunciators and trip indicators Electrical apply to the Loss of Vital AC Instrument Bus:

Instrument Bus I 6 KIA# AA2.03 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.38.1B Rev. 6, pgs. 1, 18, 21 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

14. The plant is at 100% power.

• 21A Pri Comp Cooling Hx Temp Control Viv 2CCP-TCV100A is in Manual

• 21 B Pri Comp Cooling Hx Temp Control Viv 2CCP-TCV100B is in Manual

• CCS Hx Bypass Temp Control Viv 2CCS-TCV215 is in Auto

• An inadvertent Train 'A' CIA occurs What are the effects on the Primary and Secondary Component Cooling Water Hx outlet temperatures 10 minutes after the inadvertent Train 'A' CIA occurs?

Primary Component Cooling Water Heat Exchanger (CCP) outlet Temperatures will (1)

Secondary Component Cooling Water Heat Exchanger (CCS) outlet Temperatures will (2)

A. 1) rise

2) rise B. 1) rise

2) lower C. 1) lower

2) lower D. 1) lower

2) rise Answer: D Explanation/Justification: KJA is met by demonstrating knowledge of the increased service water flow through the CCP Hxs to the Circ Water pump suction when the secondary portion of service water is isolated. Secondary side of service water is isolated due to 2 Train 'A' CIA isolation valves closing, and isolating both SWS headers to the Secondary Cooling Water Hxs, thus causing a loss of CCS.

A. Incorrect. Temp will lower with more SW through the Hxs. Plausible distractor because they must know that Train 'A' CIA will close both 107A &

C and isolate both Trains of SW to the CCS Hxs, not the CCP HXs (CCP Hxs are isolated on CIB). It is correct that CCS temps will rise.

B. Incorrect. Temp will lower with more SW through the Hxs. Plausible distractor because they must know that Train 'A' CIA will close both 107A &

C and isolate both Trains of SW to the CCS Hxs, not the CCP HXs (CCP Hxs are isolated on CIB). CCS temps will rise with SW isolated.

C. Incorrect. CCP temp will lower due to increased SW flow through the CCP Hxs with the TCVs in manual and the secondary side SW isolated.

Plausible distractor of CCS temperature lowering if candidate thinks only the 'A' SW header is isolated from the CCS Hxs, and CCS loads have reduced due to the CIA.

D. Correct. CCP temp will lower due to increased SW flow through the CCP Hxs with the TCVs in manual and the secondary side SW isolated. The inadvertent Train 'A' CIA will close 2SWS-MOV107A & C, isolating both SW headers from the CCS HX causing CCS temps to rise.

Sys# System Category KA Statement 000062 Loss of Nuclear AK3. Knowledge of the reasons for the following responses as Effect on the nuclear service water discharge flow Service Water I 4 they apply to the Loss of Nuclear Service Water: header of a loss of CCW KIA# AK3.04 KIA Importance 3.5 Exam Level RO References provided to Candidate Technical

References:

20M-30.1.D Rev. 8, pg. 6, None U2 RM -0430-001 and 003 2SQS-30.1 PPNT Rev. 23 Slide 11 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.4, 41.8 I 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

15. The following conditions exist:

• The plant is at 45% power

• Station Air Compressor 2SAS-C21A is on Clearance for motor replacement

• ERF Substation bus 1H has tripped on overcurrent

• A7-3C, 'Turbine Bearing/Autostop Oil Trouble' is in alarm due to a loss of power to Main Turb Bearing Oil Pump 2TML-P207

• No Operator actions have been taken Based on the above indications, which of the following procedures would be entered to mitigate this event?

A. E-0, "Reactor Trip or Safety Injection" B. AOP-2.26.1, "Turbine and Generator Trip" C. AOP-2.34.1, "Loss of Station/Cnmt Instrument Air" D. AOP-2.37.1, "Loss of 480V BUS 2N OR 2P" Answer: C Explanation/Justification: KIA is met by requiring the candidate to recognize the entry conditions for AOP-2.34.1 "Loss of Station/Cnmt Instrument Air" which are given in the stem of the question as bus 1H tripped on overcurrent which supplies bus 2K (power supply for 2SAS-C21 B). With both SACs secured, the next air compressor to start will be 2SAS-C22 at 90 psig, therefore air pressure will be lowering. These 2 conditions are entry level conditions for loss of Station/Cnmt Instrument Air AOP.

A. Incorrect. Plausible is the candidate thinks the Rx will trip due to the Turb Bearing Oil Pump 2TML-P207 being de-energized causing a turbine trip, or if they think DC bus 2-6 (powered by 2K) will de-energize and trip the Rx similar to DC bus 2-1 or 2-2.

B. Incorrect. Plausible distractor with power <49% and receiving Turbine Bearing/Autostop Oil Trouble alarm. The cause of the alarm is due to Main Turb Bearing Oil Pump 2TML-P207 being de-energized. The candidate must realize that P207 is not running when the turbine is online, and will only auto start when Turb bearing oil header pressure is low. The Turbine will not trip with Bus 2K de-energized.

C. Correct. With the loss of the 1H 4160KV bus, a loss of the 480V 2K bus occurs. 2K supplies power to the 2SAS-C21 Bair compressor. When 'B' SAC trips, the stby air compressor 2SAS-C22 won't auto start until 90 psig. An entry condition for AOP 2.34.1 is running SAC trips, and Station Inst. Air pressure will be dropping.

D. Incorrect. Plausible if the candidate thinks the 1H bus supplies either the 2N or 2P 480v bus. Electrical system knowledge is required to not select thisAOP.

Sys# System Category KA Statement 000065 Loss of Generic Ability to recognize abnormal indications for system operating Instrument parameters that are entry-level conditions for emergency and Air/ 8 abnormal operating procedures.

KIA# 2.4.4 KIA Importance 4.5 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.34.1 Rev.19 pgs. 1 & 2 1/20M-53C.4A.58E.1 Rev. 9 pg. 8 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.2 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

16. The plant is at 45% power with all systems in normal alignment for this power level EXCEPT 2FWS-P21A, 'A' Main Feedwater Pump is cleared for bearing replacement.

• The crew is performing AOP-1/2.35.1, "Degraded Grid" due to grid voltage and frequency swings

• Control Rod Group Selector switch is in MANUAL due to N-44 failing LOW. NO other actions have been taken for the N-44 failure

• A 20% Load Rejection has occurred

• Tavg-Tref deviation is 8.5 °F Based on the above conditions, what are the required actions to restore the RCS Tavg-Tref deviation?

A. Manually close Condenser Steam Dumps B. Manually insert Control Rods C. Manually trip the Reactor D. Manually trip the Turbine Answer: B Explanation/Justification: KIA is met by requiring the candidate to recognize that the control rods must be manually operated to lower Tavg due to the group selector switch being in manual when a load rejection occurs during the degraded grid event, and then recognize that the Tavg-Tref deviation of 8.5F is outside the expected band of+/- 2F.

A. Incorrect. Manually closing the steam dumps would cause Tavg to raise higher and create a larger Tavg-Tref deviation.

B. Correct. With a deviation of 8.5F due to a load rejection (Tref lowering & Tavg remaining constant), the rods will have to be driven inward to lower Tavg to Tref. If the rods were in auto, and N-44 was operable the rods would automatically drive inward to reduce the mismatch. AOP 2.35.2 (Load rejection) states to restore Tavg-Tref by manual rod insertion or boration. AOP-2.35.2 is directed by AOP-1/2.35.1.

C. Incorrect. Tripping the Rx is not required or directed. It is a plausible distractor due to the Tavg to Tref deviation. The Transient response Guidelines state to trip the Rx if the mismatch is +/- 1OF and the cause cannot be readily determined. In this case the deviation is only 8.5F and there was a load rejection.

D. Incorrect. Tripping the Turbine is not required or directed. It is a plausible distractor because the initial conditions are <P9, larger than normal Tavg-Tref deviation, and AOP-1/2.35.1 has been implemented.

Sys# System Category KA Statement 000077 Generator Voltage AA 1. Ability to operate and/or monitor the following as they apply to Reactor controls and Electric Grid Generator Voltage and Electric Grid Disturbances:

Disturbances I 6 KIA# AA 1.04 KIA Importance 4.1 Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2.35.2 Rev. 20 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 and 41.10 / 45.5, 45.7, and 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

17. Given the following conditions:

• ECA-1.2, "LOCA Outside Containment", Step 4, directs checking Reactor Coolant System pressure to determine if the break has been isolated by previous actions.

If the break has NOT been isolated, which of the following identifies the effect that a transition to ECA-1.1, Loss of Emergency Coolant Recirculation, has on mitigating the accident?

Actions of ECA-1.1 are taken to ~~~~~~~~~~~~

A. transfer Safeguards Building Sump contents to the Refueling Water Storage Tank as directed by Technical Support Center.

B. increase the injection flow rate to restore Reactor Coolant System pressure.

C. stabilize Reactor Coolant System pressure to prevent the Safety Injection Accumulators from discharging out the break.

D. minimize Refueling Water Storage Tank depletion by reducing total injection flow.

Answer: D Explanation/Justification: KJA is met by placing the candidate into a specific point of an EOP, and requiring them to understand why they will operate the Safety Injection system when in ECA-1.1 to conserve RWST inventory.

A. Incorrect. Plausible because this action could be performed to recover lost sump water, however, it is not directed by ECA-1.1.

B. Incorrect. Plausible if thought RCS pressure is restored, however, actions are taken to restore RCS mass in ECA-1.1.

C. Incorrect. Plausible because multiple EOPs depressurize SI Accumulators, however, ECA-1.1 utilizes SI Accumulator inventory.

D. Correct. Adding makeup to the RWST and reducing injection flow will minimize RWST depletion which is identified as major action category #2.

Sys # System Category KA Statement W/E11 Loss of Emergency Coolant EA 1. Ability to operate and I or monitor the following as they Desired operating results during Recirculation I 4 apply to the (Loss of Emergency Coolant Recirculation) abnormal and emergency situations.

KIA# EA1.3 KIA 3.7 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-53A.1.ECA-1.1 Rev 1 lss 2 pg. 1 Question Source: Bank-Comanche Peak 2013 NRC exam Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.51 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

18. The plant was at 100% power.

• Main Steam Line break inside containment occurs

• Main Steam Line Isolation Valves failed to close

• Transition to ECA-2.1, "Uncontrolled Depressurization Of All Steam Generators" has occurred Current plant conditions:

• SG A, Band C NR Level= 14% and slowly lowering

• CNMT pressure = 5.4 psig and lowering In accordance with ECA-2.1, based on the above conditions:

1) The MINIMUM feed flow that must be maintained to the Steam Generators is (1)

2) The MAXIMUM cooldown rate allowed is (2)

A. 1) 50 gpm each

2) <100 °F/hr B. 1) 50 gpm each

2) <25 °F/hr C. 1) 700 gpm total

2) <100 °F/hr D. 1) 700 gpm total

2) <25 °F/hr Answer: A Explanation/Justification: K/A is met by requiring the candidate to understand the importance of monitoring and operating the plant to limit feed flow when all SGs are faulted. This is required in order to minimize cooldown rate if necessary, prevent overfilling the SGs, control RCS temperature when cooldown stops, and prevent SG tube dyrout. Based on these primary plant behaviors during an event in which all SGs are faulted, the candidate must have an understanding of how to monitor and operate the plant to limit the effects of thermal shock to the SG components.

A. CORRECT. Per ECA-2.1, if SG NR level < 12% (31 % adverse], maintain a minimum of 50 gpm to each SG. With all 3 SGs blowing down in containment, adverse conditions do exist (5.4 psig). Cooldown rate is limited to <100 F/hr.

B. INCORRECT. Per ECA-2.1, if SG NR level< 12% (31% adverse], maintain a minimum of 50 gpm to each SG. Adverse conditions do exist (5.4 psig). 25 F/Hr is incorrect. Plausible distractor because natural circ cooldown is limited to 25F/hr.

C. INCORRECT. 700 gpm total is incorrect, but a plausible distractor because when adverse and <31% SG level, it is the required flow to removed heat generated when in FR-S.1. Cooldown rate is limited to 100 F/hr.

D. INCORRECT. 700 gpm total is incorrect, but a plausible distractor because when adverse and <31% SG level, it is the required flow to removed heat generated when in FR-S.1. 25F/hr cooldown is incorrect. Plausible distractor because natural circ cooldown is limited to 25F/hr.

Sys # System Category KA Statement W/E12 Uncontrolled EA 1. Ability to operate and I or monitor the following as they Operating behavior characteristics of the facility.

Depressurization of all apply to the (Uncontrolled Depressurization of all Steam Steam Generators 14 Generators)

KIA# EA1.2 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ECA-2.1 lss.2 Rev. 0 Question Source: Bank- 2010 Surry NRC Exam (Q18) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

19. A Reactor Trip occurred from 60% power.

25 minutes post-trip the following conditions exist

• N-35 indicates 4 x 10-10 amps, SUR of O DPM

• N-36 indicates 1x10-11 amps, SUR of 0 DPM Which of the following describes the current conditions?

A. N-35 is undercompensated.

N-31 and N-32 must be manually energized.

B. N-35 is undercompensated.

N-31 and N-32 energized automatically.

C. N-35 is overcompensated.

N-31 and N-32 must be manually energized.

D. N-35 is overcompensated.

N-31 and N-32 energized automatically.

Answer: A Explanation/Justification: KJA is met by knowledge required of the effects of compensating voltage on the Intermediate range detectors, and how an undercompensated IR detector will provide inaccurate indications and prevent the source range detectors from automatically energizing.

A. Correct. With N-35 indicating 4 x E-10 25 minutes after the trip, it is undercompensated. Since it did not reach the P-6 setpoint of 1xE-10, SR detectors will not auto energize since the logic is 2/2 below P-6. N-31 and N-32 will have to be manually energized.

B. Incorrect. With N-35 indicating 4 x E-1 O 25 minutes after the trip, it is undercompensated. It is incorrect that SR will auto energize because 2/2 P-6 logic has not been met.

C. Incorrect. If N-35 was overcompensated, it would go below P-6 before N-35, and the 2/2 logic below P-6 would be met. It is correct that N-31 and N-32 will have to be manually energized since the 2/2 P-6 logic has not been met.

D. Incorrect. If N-35 was overcompensated, it would go below P-6 before N-35, and the 2/2 logic below P-6 would be met. It is incorrect that SR will auto energize because the 2/2 P-6 logic has not been met.

Sys# System Category KA Statement 000033 Loss of AK1. Knowledge of the operational implications of the following Effects of voltage changes on performance Intermediate concepts as they apply to Loss of Intermediate Range Nuclear Range Instrumentation:

Nuclear lnstrumentati on/?

KIA# AK1.01 KIA Importance 2.7 Exam Level RO References provided to Candidate Technical

References:

20M-2.1.B, Rev. 3 pg. 3 & 17 None 20M-1.5.B.2 lss. 4 Rev. 0 Question Source: Bank - Surry 2012 NRC Exam (Q22)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

20. The following plant conditions exist:

• A discharge of the STEAM GENERATOR SLOWDOWN EVAPORATOR TEST TANK

[2SGC-TK238] TO UNIT 2 COOLING TOWER SLOWDOWN is ready to start.

• 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> have passed since sampling was completed (Currently 12-14-15 1400)

• Unit 1 is in Mode 4

• Unit 1 is discharging 1LW-TK-6A, LAUNDRY AND CONTAMINATED SHOWER DRAIN TANK

• Unit 1 COOLING TOWER SLOWDOWN FLOW from RCDR-ENV-MON-1 Ch. 3 is 12000 gpm

• 2CWS-FR101 COOLING TOWER SLOWDOWN FLOW recorder is reading 9400 gpm

• No expected reductions in actual cooling tower blowdown flow exists Refer to attached RWDA-L for 2SGC-TK23B and 20M-25.4L, pages 12-14 Based on the conditions above, which of the following statements are correct?

A. Cooling Tower Slowdown Flow is below the minimum allowed for discharge. Discharge can NOT start.

B. The permit is no longer effective due to exceeding the required time since the sample was taken. Discharge can NOT start.

C. Two tanks are not permitted to be discharged at the same time. Discharge can NOT start.

D. All conditions are satisfactory. Tank discharge is allowed.

Answer: C Explanation/Justification: KIA is met by having the candidate interpret a discharge permit, and determine if a liquid radioactive-waste discharge can commence. The Liquid Waste Release (LWR) permit is designed to prevent an UNCONTROLLED release of radioactive materials to the environment in liquid effluents. The amount of dilution needed is based on the activity of the tank to be released. The dilution includes a limit on the tank release rate and the Cooling Tower Slowdown flowrate.

A. Incorrect. Plausible distractor with 2CWS-FR101 indicating 9400 gpm, which is less than the minimum cooling tower blowdown flow of 10000gpm. By using the open reference pages given, it should be determined that RWDA-L required minimum is based on U1 & U2 CT Slowdown flowrate.

B. Incorrect. Plausible distractor if the candidate does not recognize that the RWDA-L is effective for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after tank sample time (20M-25.4.L, P&L I). The initial conditions gave a 2.5 day period to make it appear excessive.

C. Correct. Only one tank may be discharged at a time from the BVPS Unit1/Unit 2 site (20M-25.4.L P&L C). RWDA-L are based on the proper dilution (Cooling Tower Slowdown Flow) for a particular activity in a tank. Since the site looks at the combined Cooling Tower Slowdown Flow, and if another tank discharge was commenced, dilution would be inadequate and an accidental liquid radwaste release would occur.

D. Incorrect. As stated above, two tanks may not be discharged at the same time.

Sys# System Category KA Statement 000059 Accidental AA2. Ability to determine and interpret the following as they apply to The permit for liquid radioactive-waste release Liquid the Accidental Liquid Radwaste Release:

Radwaste Release I 9 KIA# AA2.02 KIA Importance 2.9 Exam Level RO References provided to Candidate RWDA-L for 2SGC-TK23B Technical

References:

20M-25.4.L Rev. 31 20M-25.4.L, pgs. 12-14 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

21. The crew is performing 20M-56C, "Alternate Safe Shutdown from Outside the Control Room" due to a fire in the Control Room.

What is the reason for closing 21C SG AFW Throttle Viv [2FWE*HCV100A] within 40 minutes of Auxiliary Feedwater actuation?

A. To prevent overfill of the 'C' Steam Generator.

B. To prevent runout of the running AFW pump.

C. To ensure adequate AFW for 'A' and 'B' SGs.

D. To ensure RCS cooldown rate is within limits.

Answer: A Explanation/Justification: KIA is met by the required knowledge of the reason for isolating AFW to the 'C' SG when a fire which could degrade control of the plant from the Control Room occurs. In this case, the crew must trip the reactor and achieve cold shutdown within 72 hrs. from outside the control room using 20M-56C, "Alternate Safe Shutdown from Outside the Control Room".

A. Correct. Preventing 'C' SG overfill is a time critical action as stated in 20M-56C. The concern is also stated multiple places in 20M-56C.4.C & D (NCO & NO procedures). De-energizing the DF bus removes the 'B' AFW from service, and closing HCV1 OOA stops all AFW to 'C' SG.

B. Incorrect. Plausible distractor because the 20M-56C procedures only counts on the 'A' MDAFW and the TDAFW pumps, and they may feel that feeding three SGs may cause the pumps to runout. Incorrect because the steaming rate during C/D is within the AFW pump capabilities.

C. Incorrect. Plausible distractor since it may be thought that water could be limited since TK-210 has approx. 130000 gals of water to feed the SGs during the 72 hr. C/D. Incorrect because AFW pumps do have backup supplies available.

D. Incorrect. Plausible distractor because they may feel that the limited CID rate of 25F/Hr may be exceeded if three SGs are fed at once.

Sys# System Category KA Statement 000067 Plant fire on AK3. Knowledge of the reasons for the following responses as they Actions contained in EOP for plant fire on site site/ 9 apply to the Plant Fire on Site:

KIA# AK3.04 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-56C.4.B Rev. 33 pg. 3 20M-56C.4.C Rev. 20 pg. 4 20M-56C.4.D Rev. 24 pg. 3 Question Source: Bank - Vision #254205 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5,41.10 / 45.6 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

22. The plant is at 83% power with all systems in normal alignment for this power level.

• A serious fire breaks out in the office located behind the Shift Manager desk

• Flames erupt from the office, and the Control Room begins to fill with smoke

• Shift manager directs entry into 20M-56C, "Alternate Safe Shutdown From Outside Control Room Operating Procedures" In accordance with 20M-56C.4.C, "NCO Procedure", the Reactor Operator is expected to trip the Reactor from the ~~~~~~~~~

A. Main Control Room B. Alternate Safe Shutdown Panel C. Emergency Shutdown Panel D. Reactor Trip breakers or MG Set Breakers Answer: A Explanation/Justification: KIA is met by the required knowledge of the Reactor Operator responsibilities during performance of 20M-56C.4.C which include manually tripping the Rx from the Control Room prior to evacuating during a fire.

A. Correct. Step 1 of the NCO procedure is manually trip the Rx. The procedure does not have the RO evacuate the CR until 7 steps later. It is assumed in 20M-56C.3.A that an automatic or manual trip will put the plant in Mode 3.

B. Incorrect. Plausible because the procedure entered is called Alternate Safe Shutdown. No means to trip the Rx exist on the ASP.

C. Incorrect. Plausible because it is a panel located outside the CR, used to shutdown the plant. Knowledge that the purpose of the 56C procedures is to perform a safe shutdown from outside the CR without the Emergency Shutdown panel. No means to trip the Rx exists on the ESP.

D. Incorrect. Plausible answer because locally tripping the reactor is expected if a failure from the Control Room occurs. In this case, the first step of the NCO procedure is to manually trip the Rx.

Sys# System Category KA Statement 000068 Control Room AK2. Knowledge of the interrelations between the Control Room Reactor trip system Evacuation I Evacuation and the following:

8 KIA# AK2.02 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-56C.4.C Rev. 20 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

23. The following conditions exist:

• E-1, "Loss of Reactor or Secondary Coolant" is in progress

• Pressurizer Level is 0%

• RVLIS Full Range Level is 45% and steady

• Containment pressure peaked at 46 psig, and is currently 10.5 psig and slowly lowering

• RWST level is 440 inches and lowering

• Containment sump level is 191 inches and rising Based on the above conditions, which of the following is the highest priority procedure transition required by the crew, and what is the reason for the transition?

A. ES-1.3, "Transfer to Cold Leg Recirculation" because Safety Injection is aligned for recirculation to provide for long term cooling B. FR-Z.1, "Response to High Containment Pressure" because containment integrity is challenged due to peak CNMT pressure C. FR-Z.2, "Response to Containment Flooding" because containment integrity is challenged due to high CNMT sump level D. FR-1.2, "Response to Low Pressurizer Level" because PRZR level has lowered below 14%

Answer: C Explanation/Justification: KIA is met by the knowledge required to determine that entry into FR-Z.2, CNMT Flooding is required to prevent a loss of containment integrity due to high sump level. As the water level rises, it might threaten the availability of equipment required for long-term cooling of the core and/or containment. Such a high water level is considered a severe challenge to the cnmt barrier and restoration is FR-Z.2.

A. Incorrect. ES-1.3 is a plausible distractor if the candidate doesn't know the RWST setpoint of <430 inches, stem states level is 440 inches ..

B. Incorrect. With CNMT pressure <11 psig, there are no entry conditions met into FR-Z.1. Plausible distractor with the peak pressure of 46 psig, the candidate must differentiate between peak and current pressure when making the status tree decision. The candidate must know entry conditions to FR-Z.1 Red or Orange path.

C. Correct. FR-Z.2 Orange path for Containment Flooding is the correct procedure based on Status Tree F-0.5 with CNMT pressure <11 psig and CNMT sump level >187 inches. The reason for using FR-Z.2 is to identify and isolate water sources to minimize sump level which could challenge cnmt integrity.

D. Incorrect. With a PRZR level of 0% and RVLIS level at 45%, this is a plausible distractor with for Inventory (with RVLIS) Status Tree paths.

Sys # System Category KA Statement 000069 Loss of Containment AK3. Knowledge of the reasons for the following responses Guidance contained in EOP for loss of Integrity I 5 as they apply to the Loss of Containment Integrity: containment integrity KIA# AK3.01 KIA Importance 3.8* Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.F-0.5, lss. 2 Rev. 0 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 I 45.6 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

24. Plant was at 100% power when the Rx Tripped due to a Faulted SG.

• The Steam Generator has blown dry

• The crew has transitioned to ES-1.1, SI TERMINATION

• The crew has just secured "B" Charging Pump

• "A" Charging Pump is running

• RCS pressure is rising

• PRZR level is rising

• RWST level is lowering Which of the following describes the actions that will be taken next per ES-1.1, SI TERMINATION, and why should a reduction in SI flow be done expeditiously?

A. 1) Establish letdown flow.

2) To preserve RWST inventory.

B. 1) Establish normal charging flow.

2) To preserve RWST inventory.

C. 1) Establish letdown flow.

2) To prevent the pressurizer from going solid.

D. 1) Establish normal charging flow.

2) To prevent the pressurizer from going solid.

Answer: D Explanation/Justification: KJA is met by the candidate assessing the CR indications, and with knowledge of the SI termination procedure, determine that normal charging must be established prior to letdown. The candidate must also demonstrate knowledge why SI should be terminated quickly as to not take the PRZR solid.

A. Incorrect. Establishing letdown flow is plausible if candidate thinks that securing one charging pump (HHSI) pump in SI term has returned to a normal charging flowpath and restoring UD would be the next logical step to perform. Securing SI expeditiously is not based on RWST depletion, but it is plausible with RWST level lowering.

B. Incorrect. Establish normal charging flow is correct. Securing SI expeditiously is not based on RWST depletion, but it is plausible with RWST level lowering.

C. Incorrect. Establishing letdown flow is plausible if candidate thinks that securing one charging pump (HHSI) pump in SI term has returned to a normal charging flowpath and restoring UD would be the next logical step to perform. Preventing the pressurizer from going solid is the correct reason for securing SI expeditiously.

D. Correct. After isolating HHSI flow, establishing normal charging flow is correct. Restoring UD is identified later in ES-1.1, but it would be incorrect to establish UD without having charging flow restored to prevent flashing downstream of the UD orifices. Preventing the pressurizer from going solid is the correct reason for securing SI expeditiously.

Sys# System Category KA Statement W/E02 SI Termination I Generic Ability to interpret control room indications to verify the status and operation of a system, and 3 understand how operator actions and directives affect plant and system conditions.

KIA# 2.2.44 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.1 lss.2 Rev. 0 pg. 9 20M-538.4.E-O lss. 2 Rev. 1 pg. 12 20M-7.2.A Rev. 17 pg. 3 Question Source: Bank- VC Summer 2011 NRC Exam- Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12))

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

25. The following conditions exist:

• A LOCA has occurred

• The crew is performing ES-1.2, "Post LOCA Cooldown and Depressurization"

• RCS cooldown to cold shutdown is in progress

• All RCPs are shutdown

• The crew is reducing RCS pressure to refill the pressurizer Which of the following would indicate to the crew that voiding in the RCS is occurring?

A. 2RCS-Ll460, PRZR CHANNEL 2 LEVEL, rapidly increasing.

B. 2RCS-Pl402, RX CLNT SYSTEM WIDE RNG PRESS, rapidly increasing.

C. 2SIS*Fl943, HHSI TRN B, rapidly decreasing.

D. UPS011, PSMS Average lncore TIC Temp, rapidly decreasing.

Answer: A Explanation/Justification: KIA is met by the required knowledge that the Reactor Operator must understand the implications of depressurizing the RCS when directed by ES-1.2, "Post LOCA Cooldown and Depressurization", and recognize available Control Room indications which indicate a void in the vessel head is occurring.

A. Correct - Voiding causes water to be displaced in the RCS which shows up as an increase in pressurizer level.

B. Incorrect - Increasing RCS pressure would suppress voiding in the RCS. Plausible for the same reason as C.

C. Incorrect - Decreasing SI flow would be indicative of a pressure increase which is inconsistent with voiding in the RCS. Plausible because candidate may think that the expansion of a void bubble and PRZR level rising would cause pressure to rise, thereby reducing HHSI flow.

D. Incorrect - Plausible because the candidate may think that during void formation, there will be less heat transfer and temperature will go down rapidly. Due to saturation conditions during void formation, temperature should stay approximately the same.

Sys # System Category KA Statement W/E03 LOCA Cooldown and EK1. Knowledge of the operational implications of the Normal, abnormal and emergency operating Depressurization I 4 following concepts as they apply to the (LOCA Cooldown and procedures associated with (LOCA Cooldown and Depressurization) Depressurization).

KJA# EK1 .2 KJA Importance 3.6 Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.ES-1.2 lss. 2 Rev. 1 pg.11 None Question Source: Bank - Farley 2011 NRC Exam (070)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

26. While performing actions of FR-C.3, "Response to Saturated Core Cooling", which of the following identifies the condition that the PRZR PORVs and block valves are required to be in?

(Assume no previous PRZR PORV failures)

A. Three PORVs closed with all block valves closed to minimize RCS leakage.

B. Two PORVs closed and one PORV open with associated block valve open for RCS pressure control.

C. Three PORVs closed with at LEAST one block valve open for RCS pressure control.

D. One PORV closed with two PORVs open to depressurize the RCS to facilitate SI Accumulator Injection.

Answer: C Explanation/Justification: KJA met by knowledge of FR-C.3, "Response to Saturated Core Cooling" major action step to check for open RCS vent paths, and the ability to monitor the PORVs and Block valves in the required system configuration.

A. Incorrect. Plausible to prevent RCS inventory loss since core cooling is degraded already, but not IAW FR-C.3.

B. Incorrect. Plausible if they assumed that the PORV was being used to lower RCS pressure as is a mitigative strategy in FR-C.1.

C. Correct. The candidate must be familiar with the basic purpose, overall sequence of events or overall mitigative strategy of Saturated Core Cooling. With knowledge of the FR-C procedures. the RO demonstrates the ability to operate the plant and obtain desired operating results during these emergency plant conditions. The major action categories for FR-C.3 is to establish SI flow to maintain minimum RCS subcooling, and check for open RCS vent paths.

D. Incorrect. This action could be performed in FR-C.1 to depressurize RCS but not performed for yellow condition FR-C.3.

Sys# System Category KA Statement W/E07 Saturated EA 1. Ability to operate and I or monitor the following as they apply to Components. and functions of control and safety Core Cooling the (Saturated Core Cooling) systems, including instrumentation, signals, 14 interlocks, failure modes. and automatic and manual features.

KIA# EA1.1 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-C.3 lss. 2 Rev. 0 Question Source: Bank - Ginna 2011 NRC Exam (062) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

27. Given the following plant conditions:

• 'A' SIG Pressure is 1150 psig

• 'A' SIG Narrow range level is 82%

• RCS hot leg temperatures are 563 °F

• FR-H.2, "Response to Steam Generator Overpressure" has been entered

• All MSIVs are CLOSED

• 2SVS*PCV101A, 21A SG ATM STM DUMP has failed CLOSED

• 2SVS*HCV104 Residual Heat Release Viv is CLOSED and will not open

• 2FWE*P22, Turbine Driven AFW pump is out of service for bearing replacement Which of the following describes the preferred method to reduce "A" SIG pressure in accordance with FR-H.2?

A. Feed 'A' SG with AFW and commence an RCS cooldown to less than 534°F using 'S' & 'C' Steam Generators.

S. Feed 'A' SG with AFW and establish Slowdown from the 'A' Steam Generator.

C. Isolate AFW to the 'A' SG and commence RCS cooldown to less than 534°F using 'S' & 'C' Steam Generators.

D. Isolate AFW to the 'A' SG and establish Slowdown from the 'A' Steam Generator.

Answer: C Explanation/Justification: KIA met by candidates ability to perform FR-H.2, "Response to Steam Generator Overpressure" major action steps of Controlling the affected SG pressure and initiate CID using the unaffected SGs. In the stem of the question steam dumps are unavailable for 'A' SG, therefore candidate must recognize that cooldown is required on the unaffected SGs.

A. Incorrect. Feeding the A SG is not permitted (or procedurally driven) because feed may be the cause of the overpressure. Cooling the RCS using the unaffected SGs is the correct answer if steam cannot be dumped from the affected SG.

B. Incorrect. Feeding the A SG is not permitted (or procedurally driven) because feed may be the cause of the overpressure. Establishing blowdown from A SG is not procedurally driven.

C. Correct. Major action steps of FR-H.2 are Control the affected SG pressure and initiate CID using the unaffected SGs. Given the initial conditions in the stem it is determined that there are no means to control pressure in the affected SG, therefore it will be necessary to isolate AFW to the 'A' SG and cooldown using the other SG by dumping steam using B and/or C ADV.

D. Incorrect. Isolating AFW is correct, but establishing blowdown from A SG is not procedurally driven.

Sys# System Category KA Statement W/E13 Steam Generator Generic Ability to perform specific system and integrated Overpressure I 4 plant procedures during all modes of plant operation.

KIA# 2.1.23 KIA Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-H.2 lss. 2 Rev.a Question Source: Bank - Surry 2012 NRC (024)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.2 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

28. During a plant startup, the following conditions exist:

• Reactor is at 13% power

• Offsite power is supplying all 4KV busses

• All 3 Reactor Coolant Pumps (RCPs) are running The "2B STA SERVICE FEEDER 138KV BREAKER PCB-94" spuriously tripped open.

Which RCP(s) will lose power, and will the Reactor automatically trip due to this loss of the RCP(s)?

(1) will lose power.

The reactor _ _ _(_2).__ _ automatically trip due to this loss of the RCP(s).

A. 1) Only 'C' RCP

2) will B. 1) Only 'C' RCP

2) will NOT C. 1) Both 'A' and 'B' RCPs

2) will D. 1) Both 'A' and 'B' RCPs

2) will NOT Answer: B Explanation/Justification: KIA is met by the candidates ability to analyze the loss of power to the system service transformer 2B, and the loss of power to only 'C' RCP.

A. Incorrect. It is correct that 'C' RCP will lose power. It is incorrect that the Rx will trip. Plausible distractor because the student needs to know that a loss of 1 RCP when power is <P-8 (30%) will not trip the Rx. This coincidence is easy to confuse because a loss of 2/3 RCPs >P-7 (10%) will trip the Rx.

B. Correct. 'C' RCP is powered from 4160 Bus 2C, which is supplied from the 2B System Service Transformer. When PCB-94 opens SSST 2B is de-energized, which de-energizes 4160 bus 2C & 20. There are no RCPs on bus 20. It is correct that the Rx will not trip on the loss of 1 RCP since the plant was <PS (30% power).

C. Incorrect. 'A' & 'B' RCPs are powered from SSST 2A, which is not effected by PCB-94 opening. Plausible distractor because the candidate must know which off-site feed supplies the SSSTs and ultimately the RCP busses. It is incorrect that the Rx will trip. Plausible distractor because the student needs to know that a loss of 1 RCP when power is <P-8 (30%) will not trip the Rx. This coincidence is easy to confuse because a loss of 2/3 RCPs >P-7 (10%) will trip the Rx.

D. Incorrect. 'A' & 'B' RCPs are powered from SSST 2A, which is not effected by PCB-94 opening. It is correct that the Rx will not trip.

Sys # System Category KA Statement 003 Reactor Coolant Pump K2 Knowledge of bus power supplies to the following: RCPS System (RCPS)

KIA# K2.01 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

20M-1.5.B.1 Rev. 2 pg. 2 RE-0001 DH Rev. 4 & RE-0001 E Rev. 9 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

29. The plant is shutting down for a refueling outage in accordance with 20M-52.4.R.1 Shutdown From 100% Power To Mode 5".

• Plant is at 30% power

• All systems in normal alignment for this power level Chemistry has requested a purge of the VCT to remove non-cond In accordance with 20M-7.4.F, "Degassing the Reactor Co nt System From The Volume Control Tank", which of the following completes the state ents below?

At this power level, _ ___._(1"""')_ __ r purging the RCS of non-condensable gasses.

The non-condensable gasses from the V.. will be purged to the _ __....(2.....)_ __

A. 1) Hydrogen

2) Primary Plant Sample S B. 1) 2)

C. 1)

2) Plant Sample System D. itrogen Boron Recovery System Answer: D Explanation/Justification: KIA is met with the knowledge of nitrogen gas being used in conjunction with raising and lowering VCT level to purge the Hydrogen and non-condensable gasses from the VCT (CVCS) during RCS degassing when performing a plant shutdown.

Reducing gas concentration at BVPS is accomplished by reducing Hydrogen and non-condensable gasses from the RCS via the VCT, and venting the PRZR to the sample system. Although the KIA statement states from the przr bubble space, it would not be possible to meet the KIA due to purging the gasses from the przr to the sample system bypasses the CVCS system which is the KIA required system tie. Purging the PRZR is a Chemistry procedure.

A. Incorrect. Plausible because Hydrogen is the normal cover gas maintained on the VCT during operation. The goal during degas is to reduce Hydrogen and non-condensable gasses. Primary Sample System is a plausible distractor because this is an approved method of continually degasifying the przr, but the stem asked about purging the VCT to remove non-condensable gasses from the RCS.

B. Incorrect. Plausible because Hydrogen is the normal cover gas maintained on the VCT during operation. The goal during degas is to reduce Hydrogen and non-condensable gasses. The gasses will be purged from the VCT to the Degasifiers in the Boron Recovery System.

C. Incorrect. Nitrogen is used as the cover gas when removing Hydrogen and non-condensable gasses from the RCS. Primary Sample System is a plausible distractor because this is an approved method of continually degasifying the przr, but the stem asked about purging the VCT to remove non-condensable gasses from the RCS.

D. Correct. Nitrogen is used as the cover gas when removing Hydrogen and non-condensable gasses from the RCS due to chemistry requirements to reduce Hydrogen and non-condensable gasses prior to an outage. The gasses will be purged from the VCT to the Degasifiers in the Boron Recove S stem.

Sys# System Category KA Statement 004 Chemical and K5 Knowledge of the operational implications of the following Reduction process of gas concentration in RCS: vent Volume concepts as they apply to the eves: accumulated non-condensable gases from PZR Control bubble space, depressurized during cooldown or by System alternately heating and cooling (spray) within allowed pressure band (drive more gas out of solution)

KIA# K5.14 KIA Importance 2.5 Exam Level RO References provided to Candidate Technical

References:

20M-7.4.F, Rev. 19 pg. 3 & 4 None U2 RM-0407-002, RM-0409-002, RM-0408-001 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

30. Which of these components can be supplied power from either the 'A' OR 'B' Train of 480VAC (selectable)?

A. 2RHS*MOV702A, RHS Train A Supply Isolation Valve B. 2RHS*MOV702B, RHS Train B Supply Isolation Valve C. 2RHS*MOV720A, RHS Train Return to B Loop Isolation Valve D. 2RHS*MOV720B, RHS Train Return to C Loop Isolation Valve Answer: A Explanation/Justification: K/A is met by the knowledge required to determine which of the RHR pressure boundary MOVs are supplied by dual power supplies.

A. Correct. 2RHS*MOV702A can be powered by either MCC*2-E05 or E06. All other valves listed below are plausible because they are RCS pressure boundary MOVs, but only have one power supply.

B. Incorrect. 2RHS*MOV702B is only powered from MCC*2-E06.

C. Incorrect. 2RHS*MOV720A is only powered from MCC*2-E05.

D. Incorrect. 2RHS*MOV720B is only powered from MCC*2-E06.

Sys# System Category KA Statement 005 Residual Heat K2 Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves Removal System (RHRS)

K/A# K2.03 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

20M-10.1.D rev. 0 lss. 4 Question Source: Bank - Vision #240045 Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

31. Given the following conditions:

• The plant was at 80% power when a LOCA occurred

• All ESF equipment operated as designed

• RCS pressure is now 50 psig

• RWST level is 360 inches and lowering Which of the following describes the Safety Injection System alignment for these conditions?

A. LHSI pumps taking a suction from the RWST and discharging to the RCS Cold Legs.

B. HHSI pumps taking a suction from the RWST and discharging to the RCS Cold Legs.

C. LHSI pumps stopped and suction isolated from the RWST.

D. LHSI pumps taking a suction from the Containment Sump and discharging to the HHSI pump suction.

Answer: C Explanation/Justification: KIA is met by the candidates ability to monitor the given plant conditions, and recognize that when RWST level is <369",

the ECCS system will transfer into Cold Leg Recirc mode, at which time the LHSI pumps will automatically trip and the RWST suction valves will close.

KIA statement was changed from RHR pumps to LHSI pumps after discussion with the Chief Examiner based on the fact that at BVPS2 RHR is not an ECCS system. By changing the system name only, the intent of the KIA was preserved.

A. Incorrect. Plausible distractor because this is the lineup prior to the Cold Leg Recirc mode at 369" in RWST. When the transfer occurs, 2SIS-8809A & B close to isolate the RWST, and the LHSI pumps trip.

B. Incorrect. Plausible distractor because this is the lineup prior to the Cold Leg Recirc mode at 369" in RWST. When the transfer occurs, 2CHS-MOV115B & D close to isolate the RWST, and 2SIS-MOVMOV863A & B open to align the suction to the LHSI discharge piping for the RS pump.

C. Correct. When the RWST level lowers to less than 369" on 2/4 channels coincident with a Safety Injection signal, the ECCS system will transfer into Cold Leg Recirc mode. This will cause Recirc Spray pumps C & D to start, and align their discharge to the suction of the HHSI pumps. The LHSI pumps will trip and their suction valves to the RWST will close.

D. Incorrect. Plausible distractor because LHSI pumps are vital pumps and there is a misconception that they take suction from the sump.

Sys# System Category KA Statement 006 Emergency A3 Ability to monitor automatic operation of the ECCS, including: RHR pllmps (LHSI Pumps)

Core Cooling System (ECCS)

KIA# A3.07 KIA Importance 3.6* Exam Level RO References provided to Candidate None Technical

References:

20M-11.1.D rev.1 20M-11.2.B rev. 5 Question Source: Bank - Vision #123771 Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

32. The plant is at 100% power.

• 'B' Charging pump is RUNNING

• 'A' Charging pump is in STBY

• Normal 480 VAC MCC-2-13 Cub 80 tripped on overcurrent causing the following:

o 2CHS-SOV150A, 'CHG PP 21A Lube Oil Temp Solenoid VLV' is de-energized o 2CHS-P21A-1, 'Charging Pump Auxiliary Lube Oil Pump' is de-energized How will 2CHS-TCV150A, 'CHG PP 21A Lube Oil Temp Control Valve' respond to this failure, and will the 'A' Charging pump start without the Aux Oil Pump running?

1) 2CHS-TCV150A will direct all oil flow _ __....(1_)___ the lube oil cooler.

2) The 'A' Charging pump _ ___._(2~)___ start without the Aux Oil Pump running.

A. 1) to bypass

2) will B. 1) to bypass

2) will NOT C. 1) through

2) will D. 1) through

2) will NOT Answer: C Explanation/Justification: KIA is met by the knowledge demonstrated of the design feature of the centrifugal charging (HHSI) pumps lube oil TCV to divert all oil flow through the LO cooler to provide max cooling to the pump bearings on a loss of air.

A. Incorrect. TCV150A will direct all oil through the cooler to maximize cooling. It is correct the charging pump will start without the AOP running.

B. Incorrect. TCV150A will direct all oil through the cooler to maximize cooling. It is incorrect to state that the Charging pump will not start.

Plausible distractor because the design feature to start the Aux Oil Pump at 14 psig, and stop the AOP when the shaft driven pump raises pressure to normal. This feature could lead to thinking that the AOP must be running to start the Charging Pump.

C. Correct. When 2CHS-SOV150A is de-energized, it vents air from 2CHS-TCV150A causing it to divert all flow through the lube oil cooler for maximum cooling to bearings and gears. The Aux Oil Pump has a design feature to provide lubrication and cooling to the charging pump, but it is not required to be running in order to start the Charging Pump.

D. Incorrect. It is correct that TCV150A will divert all flow through the lube oil cooler. It is incorrect to state that the Charging pump will not start.

Plausible distractor because the design feature to start the Aux Oil Pump at 14 psig, and stop the AOP when the shaft driven pump raises pressure to normal. This feature could lead to thinking that the AOP must be running to start the Charging Pump.

Sys # System Category KA Statement 006 Emergency Core Cooling K4 Knowledge of ECCS design feature(s) and/or Cooling of centrifugal pump bearings System (ECCS) interlock(s) which provide for the following:

KIA# K4.01 KIA Importance 2.6 Exam Level RO References provided to Candidate Technical

References:

20M-7.3.C Rev. 15 pg. 2 None U2 TLD-007-096-01 & 02 Rev. 5 U2 LSK-026-001G Rev. 13 & 001A Rev. 14 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

33. Annunciator A4-3H, Pressurizer Relief Tank (PRT) Trouble has alarmed.

• The operator reports it is due to HIGH PRT temperature

• Level and pressure are within the normal range

• PRZR Safety valves and PORVs are closed In accordance with 20M-6.4.AAY, PRT Trouble ARP, which of the following choices identifies ALL of the valves listed below that would be opened to reduce PRT temperature?

1. PRT Pri Grade M/U Wtr valves [2RCS-MOV516 and 2RCS-AOV519]

2. PRT Vent Viv [2RCS-MOV549]

3. PRT Drain Viv [2RCS-MOV523]

A. 1 only B. 1 and 2 only C. 1 and 3 only D. 1 and 2 and 3 Answer: A Explanation/Justification: K/A is met by the required knowledge of the Pressurizer Relief Tank (PRT) system design of the ability to spray down the PRT with primary water to cool the tank.

A. Correct. With high PRT temperature the ARP states to open 2RCS-MOV516 and 2RCS-MOV519 to spray the tank internally and reduce temperature.

B. Incorrect. Plausible distractor because sect. B of the ARP for reducing tank pressure requires spraying the PRT to reduce pressure, then if pressure does not reduce <8psig, vent the tank by opening 2RCS-MOV549. Stem states level and pressure are within the normal range.

C. Incorrect. Plausible distractor if the candidate thinks you must lower PRT level prior to spraying it down to reduce temperature. The stem states level and pressure are within the normal range.

D. Incorrect. Plausible distractor to open all 3 valves if it is thought that a flushing of the PRT must occur to reduce temperature. This could work, but is not in accordance with the ARP.

Sys# System Category KA Statement 007 Pressurizer K4 Knowledge of PRTS design feature(s) and/or interlock(s) which Quench tank cooling Relief provide for the following:

Tank/Quench Tank System (PRTS)

KIA# K4.01 K/A Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.AAY Rev. 10 Question Source: Bank- Diablo Canyon 2012 NRC Exam (06) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

34. RCS temperature is 495°F with a cooldown in progress to comply with the LCO action requirements (LCO 3.7.7 Cond B) for an inoperable 'A' CCP train. Then, an unisolable leak on the 'B' supply header renders the 'B' CCP train inoperable.

As the crew continues the cooldown, which one of the following temperatures is the MINIMUM RCS temperature that is allowed by Technical Specifications for the given conditions?

A. 350° F B. 201° F C. 200° F D. 137°F Answer: B Explanation/Justification: KIA is met by the 1 hr. and less LCO conditions associated with TS.3.7.7 Component Cooling Water System (CCW), and the requirement of not entering mode 5 without Primary Component Colling water available for RHR to be placed in service.

A. Incorrect. This is the minimum temperature of Mode 3. 2 RCS loops are still required to operable. Procedurally RHS is not available to use until S350F.

B. Correct. This is the minimum temperature for Mode 4. Entry into Mode 5 is not permissible with both trains of CCP inoperable as stated by the note in TS 3.7.7 cond. C (immediate completion time).

C. Incorrect. This is maximum temperature of Mode 5 entry. Entry into Mode 5 is not permissible with both trains of CCP inoperable as stated by the note in TS 3.7.7 cond. C (immediate completion time). Without CCP, RHS is inoperable (TS 3.4.7) and 1 RHS loop is required to be in operation in Mode 5.

D. Incorrect. This is the minimum temperature to operate 3 RCPs. Plausible distractor if the candidate thinks that the RCS loops must remain operable when RHS is not operable due to both Trains of CCP inoperable and Mode 5 is required.

Sys# System Category KA Statement 008 Component Cooling Generic Knowledge of conditions and limitations in the Water System (CCWS) facility license KIA# 2.2.38 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

TS. 3.7.7 A278/161 TS. Table 1.1-1 Def. of modes Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.1 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

35. Given the following conditions:

• The plant is at 100% power

• A rupture of the 21A RCP Thermal Barrier occurs

• No annunciators are in alarm Which of the following completes the statements below?

1) 2CCP*AOV107A, 21A RCP THERMAL BARRIER OUTLET ISOL VLVwill automatically close at (1)

2) In accordance with AOP-2.6.8, "Abnormal RCP Operation", a shutdown of the 'A' RCP (2) required for this failure.

(1) (2)

A. 122 psig is NOT B. 122 psig is C. 50 gpm is NOT D. 50gpm is Answer: A Explanation/Justification: KIA is met by demonstrating the knowledge that RCPs can still operate with a loss of CCP thermal barrier flow as long as RCP seal injection flow is available.

A. Correct. 2CCP*AOV107's auto close at 122 psig and/or 58 gpm. The AOP does not require a shutdown of the RCP as long as there is still Seal Injection. The stem of the question does not state any problems which would lead to seal injection failure and there are no annunciators in alarm.

B. Incorrect. This is the correct pressure which auto closes 2CCP*AOV107's. It is incorrect that the AOP requires a RCP shutdown. There are no indications of a loss of seal injection.

C. Incorrect. 50 Gpm is less than the setpoint of 58 gpm required to auto close 2CCP*AOV107's. It is correct that a shutdown of the RCP is not required.

D. Incorrect. 50 Gpm is less than the setpoint of 58 gpm required to auto close 2CCP*AOV107's. It is incorrect that the AOP requires a RCP shutdown. There are no indications of a loss of seal injection.

Sys# System Category KA Statement 008 Component K3 Knowledge of the effect that a loss or malfunction of the CCWS RCP Cooling will have on the following:

Water System (CCWS)

KIA# K3.03 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-15.2.B Rev. 1pg.4 20M-53C.4.2.6.8 Rev. 12 pg 1 & 2 Question Source: Bank - Farley 2012 NRC Exam (Q14)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: None Listed Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

36. The plant is at 100% power.

• PRZR Channel 1 Press 2RCS-PT455 has failed HIGH

• The control room crew has tripped all associated bistables IAW 20M-6.4.IF, "Instrument Failure Procedure" PRZR Control Pressure [2RCS-PT445] THEN fails HIGH.

What will be the INITIAL plant response to this additional failure?

A. PRZR Spray Valve 2RCS*PCV455A & 2RCS*PCV4558 will OPEN.

B. PRZR PORV 2RCS-PCV455C will OPEN.

C. PRZR PORVs 2RCS-PCV455D & 2RCS-PCV456 will OPEN.

D. High PRZR Pressure Reactor Trip will ACTUATE.

Answer: C Explanation/Justification: KIA is met by demonstrating the knowledge of how PRZR pressure will respond to a pressure control channel failing high, and the knowledge that 2 PORVS (PCV455D & PCV456) will automatically open.

A. Incorrect. This would be the INITIAL response if 2RCS-PT444 failed High.

8. Incorrect. This would be the next response if 2RCS-PT444 failed High.

C. Correct. IAW 20M-6.4.IF attachment 2.

D. Incorrect. Failures are one control channel and one protection channel, therefore NO reactor trip.

Sys# System Category KA Statement 010 Pressurizer Pressure A3 Ability to monitor automatic operation of the PZR PZR pressure Control System (PZR PCS, including:

PCS)

KIA# A3.02 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF attachment 2 Rev. 13 Question Source: Bank - 2LOT6 NRC Exam (036) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

37. Given the following conditions:

• A reactor startup is in progress

• The reactor is critical in the source range

• N42 Power Range channel has failed and has been removed from service with all bistables placed in the trip condition

• A loss of Vital Bus 1 occurs

• A 12-1 H, NOT P-7 changed state after the loss of Vital Bus occurred What is the condition of the reactor, and source range detectors after Vital Bus 1 is lost?

A. Reactor trips N31 Source Range channel is de-energized.

N32 Source Range channel is still in operation.

B. Reactor remains critical BOTH source range channels are de-energized.

C. Reactor remains critical N31 Source Range channel is de-energized.

N32 Source Range channel is still in operation.

D. Reactor trips BOTH source range channels are de-energized.

Answer: D Explanation/Justification: KIA is met by requiring knowledge of a loss of a 2"d (redundant) PR NI channel due to the loss of vital bus 1, and the effects it has on both the Rx Protection System causing a Rx trip and de-energizing both SR channels.

A. Incorrect. It is correct that the Rx will trip due to 214 PR high setpoints. N31 is de-energized by the loss of vital bus 1. N32 will not be in operation due to P-10 auto de-energizing both SR detectors B. Incorrect. Reactor trips on a number of PR/SR trip setpoints. It is correct that both SR detectors will be de-energized C. Incorrect. Reactor trips on a number of PR/SR trip setpoints. N31 is de-energized by the loss of vital bus 1. N32 will not be in operation due to P-10 auto de-energizing both SR detector.

D. Correct. A loss of Vital 1 causes a loss of power to N41. This loss also causes a loss of power to RPS channel 1. This will cause a trip condition for Power range trips for channel 1. Since N42 is already removed from service its bistable are in the tripped condition. This meets the 2/4 logic to cause a reactor trip. N31 is de-energized by the loss of vital bus 1. Additionally the signal for 214 power range channels above P-10 will cause the SR channels to auto de-energize causing N32 to de-energize.

Sys# System Category KA Statement 012 Reactor Protection K6 Knowledge of the effect of a loss or malfunction of the Redundant channels System (RPS) following will have on the RPS:

KIA# K6.02 KIA Importance 2.9 Exam Level RO References provided to Candidate Technical

References:

20M-2.2.A Rev. 1 (P&L 9)

None UFSAR Fig. 7.3-8 and 7.3-9 20M-2.3.C Rev. 5 pg. 3 Question Source: Bank - DC Cook 2010 NRC Exam (039)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45/7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

38. Given the following conditions:

• The plant is at 100% power

• Containment Pressure Channel IV pressure indication was oscillating and has been removed from service IAW 20M-1.4.IF, "Instrument Failure Procedure" Which of the following identifies the logic associated with the HIGH 1 and HIGH 3 Containment Pressure actuations after the Channel IV is removed from service?

HIGH 1 SI Actuation HIGH 3 CIB Actuation A. 1/2 2/3 B. 1/2 1/3 C. 1/3 2/3 D. 1/3 1/3 Answer: A Explanation/Justification: K/A is met by demonstrating knowledge of the containment pressure channel inputs to both Safety injection and Cnmt Isol phase B actuation logics, and how these inputs are removed from service for the reliability of the actuation coincidence.

A. Correct. Channel IV (2LMS-PT953) was removed from service iaw 20M-1.4.IF. This procedure and Tech Specs has the bistable tripped for High 1 (SI) which then makes the logic 1/2. The bistable for the High 3 (CIB) is required to be placed in bypass, which then requires a 2/3 coincidence to initiate a CIB. Both of these bistable configurations satisfies redundancy requirements.

B. Incorrect. Plausible if both bistables are tripped. High-1 is normally a 2/3 logic but changes to 1/2 when one of the logic channels are tripped. For High-3, 2/4 logic, the channel is bypassed, so 2/3 is required.

C. Incorrect. Plausible if the logic for both are 214 and only High-1 is tripped.

D. Incorrect. Plausible if the logic for both are 214 and both are tripped.

Sys# System Category KA Statement 013 Engineered K5 Knowledge of the operational implications of the following Safety system logic and reliability Safety concepts as they apply to the ESFAS:

Features Actuation System (ESFAS)

KIA# K5.02 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-1.4.IF Att. 1 Rev. 9 BVPS TS Bases pg. B3.3.2-38 & 39 Question Source: Bank - Harris 2012 NRC Exam (040)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

39. The plant is at 100% power when a rupture of the Chilled Water supply header to Containment occurs. The crew has isolated the rupture.

1) Which of the following components will be effected by the loss of Chilled Water?

2) What is the design backup for this system?

A. 1) Control Rod Drive Mechanism (CROM) Fans

2) Primary Component Cooling Water B. 1) Containment Air Recirculation (CAR) Fans

2) Service Water C. 1) Control Rod Drive Mechanism (CROM) Fans

2) Service Water D. 1) Containment Air Recirculation (CAR) Fans

2) Primary Component Cooling Water Answer: B Explanation/Justification: K.JA was met by having the candidate predict which Cnmt cooling equipment will lose cooling capabilities when Chilled Water system is lost, and knowledge of the Service Water system as a design backup.

K.JA statement is a loss of service water. At BVPS Chill Water is the normal system aligned the Containment Air Recirc Fans, and Service Water is the backup supply. The question is written to meet the intent of the K.JA as designed at BV.

A. Incorrect. CROM coolers are supplied by CCP and no backup cooling is available. CCP is the normal cooling for the CROM coolers.

B. Correct. CAR fan coolers are supplied by chilled water, and a loss of chilled water can effect cnmt temperature. Service Water is the emergency backup cooling source for the CAR fan coolers.

C. Incorrect. CROM coolers are supplied by CCP and no backup cooling is available. Service water is the emergency backup for the CAR fan coolers.

D. Incorrect. CAR fan coolers are supplied by chilled water, and a loss of chilled water can effect cnmt temperature. CCP is not the backup for CAR fan coolers. It is the primary cooling for CROM coolers.

Sys# System Category KA Statement 022 Containment A2 Ability to (a) predict the impacts of the following malfunctions or Loss of service water Chilled Water Cooling operations on the CCS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (CCS) malfunctions or operations:

KIA# A2.04 KIA Importance 2.9* Exam Level RO References provided to Candidate Technical

References:

U2 RM-0429-004 Rev. 14 None 2SQS-44C.1 PPNT Rev. 12 pgs. 7 & 8 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 / 45.3 /

45.13)

Objective: 2SQS-44C.1 EL0-1 Describe the function of the Containment Ventilation System and the associated major components as documented in Operating Manual Chapter 20M-44C.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

40. The plant is at 100% power.

Based on the information provided on the attached PCS screen, what is the status of Tech Spec LCOs 3.6.4, Containment Pressure and 3.6.5, Containment Air Temperature?

LCO 3.6.4, Containment Pressure _ __._(1"'"'")_ _ _ met, AND LCO 3.6.5, Containment Air Temperature (2) met.

Refer to attached PCS screen A. (1) is NOT (2) is NOT B. (1) is NOT (2) is C. (1) is (2) is NOT D. (1) is (2) is Answer: D Explanation/Justification: KIA is met by having the candidate analyze a Plant Computer Screen printout as seen in the Control Room, and evaluate CNMT temperatures and Pressures for Tech Spec above the line LCO specs.

A. Incorrect. See correct answer explanation.

B. Incorrect. See correct answer explanation.

C. Incorrect. See correct answer explanation.

D. Correct. Cnmt pressures are both less than 14.2 psia required by Tech Specs. The pressures indicated are not red on the PCS screen to indicate they are greater than TS limits, but to display that they are above the high pressure alarm setpoint of 13.9 psia. Average temperature is 99.1 F which is below the LCO required 108F. One of the temperatures were purposely placed above the 108 °F allowed value to ensure the candidates awareness of the LCO being the average, and not any one temperature. To answer this question the candidate will need to assess the computer data provided and determine the status of the LCOs.

Sys # System Category KA Statement 022 Containment A4 Ability to manually operate and/or monitor in the control room: Containment readings of temperature, pressure, Cooling and humidity system.

System (CCS)

KIA# A4.05 KIA Importance 3.8 Exam Level RO References provided to Candidate CNMT PCS screen shot Technical

References:

TS 3.6.4 & 3.6.5 20M-54.3.L5 Rev. 80 Question Source: Bank - 1LOT14 NRC Exam (054) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

41. Given the following conditions:

• The plant was at 100% power when all 3 Steam Generators Faulted in Containment

• Containment Pressure is 31 psig and RISING

• Quench Spray Pumps [2QSS*P21A and P21 B] failed to start

1) What is(are) the minimum required Engineered Safety Features Actuation System (ESFAS) switch manipulations required to start BOTH Quench Spray Pumps?

A. 1 B. 2 C. 3 D. 4 Answer: B Explanation/Justification: KIA is met by the candidates recognizing that GIB did not actuate and the knowledge of how many ESF actuation switches must be operated to ensure both trains of Quench Spray Pumps start and prevent CNMT pressure from exceeding design pressure.

A. Incorrect. Plausible distractor if candidate thinks Safety Injection ESF actuation will start the Quench Spray pumps (CIB is required for QS).

B. Correct. 2 GIB switches on the same train will actuate GIB on both trains. There are a total of 4 GIB switches on the Bench Board. 2 switches per train.

C. Incorrect. Plausible distractor if candidate thinks BOTH a GIB and Safety Injection ESF actuation is needed to start the Quench Spray pump.

D. Incorrect. Plausible distractor if candidate thinks BOTH trains of CIB are required to start BOTH Quench Spray pumps. There are a total of 4 GIB switches on the Bench Board, 2 for each train, and 1 train will actuate both GIB trains.

Sys# System Category KA Statement 026 Containment A1 Ability to predict and/or monitor changes in parameters (to Containment pressure Spray prevent exceeding design limits) associated with operating the CSS System controls including:

(CSS)

KIA# A1.01 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

USFAR Fig. 7.3-13 Rev. K USFAR Fig. 7.3-62 Rev. K Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

42. RWST Temperature is indicating 43°F. Outside air temperature is 26°F.

In accordance with 20M-13.4.D, Maintaining RWST Temperature, which of the following choices identifies ALL available actions to raise RWST temperature.

1. Isolate Chilled Water from Refueling Water Storage Tank Coolers.

2. Increase Chilled Water Temperature if plant conditions allow.

3. Start an additional Refueling Water Cooling Pump.

4. Run a Quench Spray Pump on recirculation.

A. 1 only B. 1 & 2 only C. 1, 2, & 3 only D. 1, 2, 3, & 4 Answer: C Explanation/Justification: K/A is met by demonstrating the candidates knowledge of reviewing and evaluating the Quench Spray Pump surveillance test, and determining equipment operability.

A. Incorrect. See correct answer justification.

B. Incorrect. See correct answer justification.

C. Correct. Isolating CW from the coolers, Increasing CW temp, and starting a Refuel Water Cooling pump are 3 of the 4 methods of raising RWST temperature iaw 20M-13.4.D. The fourth method is to run a LHSI pump on recirculation.

D. Incorrect. Running a Quench Spray pump on recirculation is incorrect, but plausible due to the pump recirc flowpath does go back to the RWST.

The procedure states that a LHSI pump may be operated on recirc.

Sys# System Category KA Statement 026 Containment Generic Ability to perform specific system and integrated Spray plant procedures during all modes of plant System operation.

(CSS)

KIA# 2.1.23 KIA Importance 4.3 Exam Level RO References provided to Candidate Technical

References:

20M-13.4.D Rev. 16 pgs. 3 & 4 None Question Source: Bank - Vision 123823 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.2 / 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

43. Given the following conditions:

• Unit 2 has just entered MODE 1

• Rx power is 6%

• Power is being raised slowly in preparation for rolling the Main Turbine

• 'A' MFP is in service

• Feedwater Bypass Control Valves are in AUTOMATIC

• ALL AFW pumps are aligned for normal standby operation A spurious MSLI actuation occurs.

Which of the following describes the effect the MSLI will have on the Auxiliary Feedwater pumps with NO operator action?

A. ONLY the MDAFW pumps will start.

B. ALL AFW pumps will remain in standby.

C. ONLY the TDAFW pump will start.

D. ALL AFW pumps will start.

Answer: B Explanation/Justification: KIA met by demonstrating knowledge of the integrated plant response to MS IVs inadvertently closing at low power conditions, and the response of the AFW pumps to these changing conditions.

A. Incorrect. Plausible if it is thought that the MDAFW pump started on 2/3 lo-lo SG water level on 2/3 SGs due to a Rx trip. Also, it could be thought that MDAFW pump started due to an auto trip of the MFP.

B. Correct. This is the expected plant response from a low reactor power level. SIG water level shrink is not as severe as a high power level trip.

Additionally, the MFP will remain running and provides more than enough capacity to maintain S/G water levels above the lo-lo SG water level setpoint which would trip the reactor and auto start AFW pumps.

C. Incorrect. Plausible if it is thought that the TDAFW pump started on 2/3 lo-lo SG water level on 1/3 SGs due to a Rx trip.

D. Incorrect. Plausible if it is thought that a RX trip will occur and due to lo-lo SG water level <20.5% all AFW pumps would start. This is not the case since a Rx Trip or low SG water level will not occur with these condition.

Sys# System Category KA Statement 039 Main and K3 Knowledge of the effect that a loss or malfunction of the MRSS AFWpumps.

Reheat will have on the following:

Steam System (MRSS)

KIA# K3.03 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical

References:

UFSAR Instrumentation and control system logic diagram sheet Figure 7.3-19, Rev 7 Question Source: Bank- BVPS 2LOT6 NRC P.O. Practice (2009)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

44. The Main Unit Generator output is 650 MWE.

• An Inadvertent Turbine trip occurs

• 2TMS-GV2, GOVERNOR VLV NO. 2 indicates DUAL

• 2TMS-TV3, THROTTLE VLV NO. 3 Indicates OPEN

• All other TVs and GVs are CLOSED Based on the above conditions:

1) What is the correct procedure to enter?

2) What additional actions would be required to be taken by the procedure?

A. 1) Enter AOP 2.26.1, TURBINE AND GENERATOR TRIP

2) Manually initiate a Steam Line Isolation.

B. 1) Enter AOP 2.26.1, TURBINE AND GENERATOR TRIP

2) Place BOTH Turb EH Fluid Pumps in Pull-to-Lock.

C. 1) Enter E-0, REACTOR TRIP OR SAFETY INJECTION

2) Manually initiate a Steam Line Isolation.

D. 1) Enter E-0, REACTOR TRIP OR SAFETY INJECTION

2) Place BOTH Turb EH Fluid Pumps in Pull-to-Lock.

Answer: C Explanation/Justification: KJA is met by demonstrating the required knowledge to recognize that a turbine trip occurred above P9 which will cause an automatic Rx trip and entry into E-0. Candidate should also recognize that one main steam throttle and governor valve did not close, and must take IOA RNO action steps.

A. Incorrect. With power at -65%, entry into the AOP would be inappropriate because the Rx would have already tripped, and the purpose of the AOP clearly states to stabilize the unit after a turbine and generator trip below the P-9 setpoint. SLI would be correct if the correct procedure were entered (E-0). The AOP does have IOAs for the TV & GVs not being closed, manually trip the turbine, but SLI is not an option.

B. Incorrect. With power at -65%, entry into the AOP would be inappropriate because the Rx would have already tripped, and the purpose of the AOP clearly states to stabilize the unit after a turbine and generator trip below the P-9 setpoint. Placing both EH pumps to PLT would cause the TVs & GVs to close if oil pressure was maintaining them open, but this is no longer an approved method of closing the valves, and it is not procedurally permitted.

C. Correct. With power at -65% when the turbine tripped, the Rx would have tripped due to being >P9 (49% power). Even if they thought the Rx was still critical, E-0 would still be the correct procedure to enter. To respond to All TV &/or GV closed in E-0 step 2 IOA, this is a correct action in the RNO of step 2.

D. Incorrect. With power at -65% when the turbine tripped, the Rx would have tripped due to being >P9 (49% power). Placing both EH pumps to PL T would cause the TVs & GVs to close if oil pressure was maintaining them open, but this is no longer an approved method of closing the valves, and it is not procedurally permitted.

Sys # System Category KA Statement 039 Main and Reheat Steam Generic Knowledge of EOP entry conditions and System (MRSS) immediate action steps.

KJA# 2.4.1 KJA Importance 4.6 Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.E-O, Rev. 1 lss. 2 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15) 45.

The plant is at 90% power The Balance of the Plant Operator observed the above indications for 30 seconds.

What are the required actions for the crew?

A. Trip the Rx.

B. Maintain power level and initiate a CR.

C. Immediately initiate a power reduction to less than 50%.

D. Take MANUAL control of the MFRVs and maintain SG water level.

Answer: A Explanation/Justification: KIA is met by candidate demonstrating the ability to monitor control room indications, and determine that a MFW pump motor has tripped, then based on plant conditions, take IOA of tripping the Rx per the loss of MFW AOP.

KIA statement was changed from MFW turbine trip indication to MFW trip indication after discussion with the Chief Examiner based on the fact that BV2 does not have turbine driven MFW pumps. By removing turbine from the KIA statement, the intent of the KIA was preserved.

A. Correct. The immediate operator actions of AOP-2.24.1 {loss of Main Feedwater) requires that the Rx be tripped if less than 2 MFPs are running when >80%. The candidate must know that if one pump motor is running and the other motor trips, a trip of the running motor will occur.

B. Incorrect. The bright white light on the pump indicates the motor tripped on 2FWS-112182. The candidate must know that if one pump motor is running and the other motor trips, a trip of the running motor should occur within 1.5 seconds, therefore they should trip the motor iaw conduct of Operations procedure. At 90% pwr a Rx trip is required making this answer incorrect.

C. Incorrect. The appropriate action would be to enter AOP-2.24.1 {Loss of Main Feedwater) which would give direction to lower reactor power to

<52% IF initial conditions were <80% power. Therefore, since the plant is at 90%, a manual Rx trip is required, and pwr reduction is incorrect.

D. Incorrect. The candidate may feel that with one motor still running that the feed pump is capable to maintain feed flow but at a reduced capability.

This may lead them to believe that manual FRV control would be required.

Sys # System Category KA Statement 059 Main Feedwater A4 Ability to manually operate and monitor in the control MFW tl:ffbiffe trip indication (MFW) System room:

KIA# A4.01 KIA Importance 3.1* Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2.24.1 Rev. 6 pg. 2 None 20M-24.1.D Rev. 6 pg. 10 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

46. The following conditions exist:

• The plant is at 17% power

• The crew is raising power in accordance with 20M-52.4.A, 'Raising Power From 5% to Full Load Operation"

• All Feedwater Bypass Control valves are in AUTO maintaining SG levels within the control band

• The controller for 2FWS-FCV489, '21 B Feedwater Bypass Control Valve' fails to 10%

open causing 'B' SG NR water level to lower

1) What is the correctAOP to respond to this event when annunciator A6-10E "SG 21B LEVEL DEVIATION FROM SETPOINT' alarms?

2) If a reactor trip due to SG low-low level occurs, which Auxiliary Feed Pump(s) will automatically start?

A. 1) AOP-2.24.1 'LOSS OF MAIN FEEDWATER'

2) TURBINE Driven Auxiliary Feedwater Pump B. 1) AOP-2.24.1 'LOSS OF MAIN FEEDWATER'

2) MOTOR Driven Auxiliary Feedwater Pumps C. 1) AOP-2.4.1 'PROCESS CONTROL FAILURE'

2) TURBINE Driven Auxiliary Feedwater Pump D. 1) AOP-2.4.1 'PROCESS CONTROL FAILURE'

2) MOTOR Driven Auxiliary Feedwater Pumps Answer: C Explanation/Justification: KIA is met by demonstrating the ability to predict the MFW system response to a MFRV controller malfunction, and determining which procedure would be used to mitigate the event. Then state which AFW pump will automatically start if the MFRV failure is not corrected, based on knowledge of the AFW pump auto start coincidences.

At the discretion of the Chief Examiner, this KIA was changed from A2.01 to A2.12.

A. Incorrect. AOP-2.24.2 is a plausible distractor is it is thought that this event constitutes a loss of main feedwater, but in this case a controller has failed and the SGs are still being fed. TDAFW pump is the correct pump to start when 2/3 SGWL detectors in only 1 SG reaches 20.5%.

B. Incorrect. AOP-2.24.2 is a plausible distractor is it is thought that this event constitutes a loss of main feedwater, but in this case a controller has failed and the SGs are still being fed. MDAFW pump is incorrect because they will start when 2/3 SGWL detectors in 2/3 SGs reaches 20.5%.

C. Correct. Correct AOP to use when a process parameter is not being controlled within its normal control band with the control in auto. TDAFW pump is the correct pump to start when 2/3 SGWL detectors in only 1 SG reaches 20.5%.

D. Incorrect. Correct AOP to use when a process parameter is not being controlled within its normal control band with the control in auto. MDAFW pump is incorrect because they will start when 2/3 SGWL detectors in 2/3 SGs reaches 20.5%.

Sys # System Category KA Statement 059 Main Feedwater A2 Ability to (a) predict the impacts of the following malfunctions or operations on Failure of feedwater (MFW) System the MFW; and (b) based on those predictions, use procedures to correct, control, regulating valves or mitigate the consequences of those malfunctions or operations:

KIA# A2.12 KIA Importance 3.1

• Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.4.1 Rev. 1 pg. 1 USFAR figure 7.3-19 Rev. 7 20M-24.4.AAP Rev. 5 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

47. Given the following plant conditions:

• The plant is in Mode 3 following a reactor trip from 100%

• The crew has completed ES-0.1, Reactor Trip Response, and transitioned to 20M-52.4.R.1. F, Station Shutdown from 100% Power to Mode 5

• A plant cooldown of 50°F/hr is commenced using Condenser Steam Dumps

• Steam Generator (SG) Narrow Range levels are maintained stable at 44% during the cooldown, using the Auxiliary Feedwater pumps Which one of the following describes the trend of Auxiliary Feedwater flow requirements as the plant cools down to Mode 5?

A. AFW flow requirements are constant as long as SG level remains constant.

B. AFW flow requirements are constant as long as the cooldown rate remains constant.

C. More AFW flow is required to maintain SG level due to a rise in SG water density as it cools.

D. Less AFW flow is required to maintain SG level because heat input to the SGs lowers as the cooldown continues.

Answer: D Explanation/Justification: KIA is met by determining sufficient AFW flow is available to provide decay heat removal, and the knowledge that decay heat load will be larger after a Rx trip from higher power levels.

A. Incorrect. Plausible if heat input to the SG did not change. Heat input lowers due to less decay heat as the cooldown progresses.

B. Incorrect. Plausible if the effects of less decay heat are not considered.

C. Incorrect. Water density does not have any impact at this temperature in the SG.

D. Correct. Heat input lowers as the cooldown progresses due to less decay heat from the reactor.

Sys# System Category KA Statement 061 Auxiliary I Emergency K5 Knowledge of the operational implications of the Decay heat sources and magnitude Feedwater (AFW) System following concepts as the apply to the AFW:

KIA# K5.02 KIA Importance 3.2 Exam Level RO References provided to Candidate Technical

References:

GO-GPF.R8 A Rev. 1 None Question Source: Bank-Callaway 2013 NRC Exam (046)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 45. 7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

48. Plant is at 38% power, raising power to 100% in accordance with 20M-52.4.A, Raising Power From 5% To Full Load Operation.

• Charging pump 2CHS*P21A is RUNNING

• Main Feedwater Pump 2FWS-P21 Bis RUNNING

• Service Water is in NSA

• Diesel Generator 2-1 is on clearance

• All Tech Spec actions for DG 2-1 have been completed Load Tap Changer for Bus 2A SS Serv Tfmr 2A is in auto and drifts.

Bus voltage is 3800 VAC (108.5 VAC indicated) and stable.

Load Tap Changer will not respond in Auto or Manual.

Based on these conditions, assuming Bus 2A voltage remains at 3800 VAC, and assuming no operator actions have yet been taken:

1) Which of the following correctly describes plant status or expected actions?

2) Procedural guidance in effect 5 minutes after bus voltage drifts to 3800 VAC?

A. 1) The reactor will have automatically tripped due to effects of the abnormal bus voltage.

2) E-0, "Reactor Trip Or Safety Injection" B. 1) Bus 2AE will have automatically de-energized.

2) AOP 2.36.2, "Loss of 4KV Emergency Bus" C. 1) The reactor will be manually tripped IAW the bus abnormal voltage ARP.

2) E-0, "Reactor Trip Or Safety Injection" D. 1) Bus 2AE will be manually de-energized.

2) AOP 2.36.2, "Loss of 4KV Emergency Bus" Answer: B Explanation/Justification: KJA is met by candidate predicting the effect low voltage on Bus 2A will have on the emergency bus 2AE (undervoltage condition will strip Bus 2AE, and the EOG is not available), then respond using the appropriate abnormal operating procedure.

A. Incorrect. Plausible distractor because they may feel that the 'A' RCP would trip (<75% undervoltage, stem is -91 %) due to lowered 'A' bus voltage. With the plant being >P-8 (30%) and 1/3 RCP tripping would cause the Rx to trip. Neither the Rx, nor the 'A' RCP will trip.

B. Correct. With DG 2-1 on clearance and Bus voltage dropping below 93.4% (3885 VAC with 90 sec.TD) the emergency power undervoltage protection will strip and isolate the 2AE bus. The bus will be de-energized and the correct procedure is AOP 2.36.2 for the loss of 2AE.

C. Incorrect. Plausible distractor if the candidate feels as though voltage is too low and the 2A bus should have tripped, which would cause an RCP to trip. Per the ARP bus voltage must be 75% of nominal bus voltage (-3120V), but the stem has voltage -91 %) therefore this answer is incorrect.

D. Incorrect. Plausible distractor if it is recognized that the voltage dropped below the 93.4%. The question stated that 5 minutes have elapsed, therefore bus 2AE would have automatically tripped after a 90 second time delay, and no manual actions would be required.

Sys # System Category KA Statement 062 AC Electrical A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Consequences of exceeding Distribution ac distribution system; and (b) based on those predictions, use procedures to correct, voltage limitations System control, or mitigate the consequences of those malfunctions or operations:

KJA# A2.08 KJA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

LSK-022-0058 rev. 8 SPD-27-VE3200AB rev. 1 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

49. Which of the following is correct regarding a loss of Vital Bus Inverter 2-1 normal power supply?

If Vital Bus 2-1 normal power supply ___(1) ___ is de-energized, Vital Bus Inverter 2-1 will automatically be supplied by ___(2) ___ without affecting the regulated AC output to Vital Bus 2-1.

A. 1) MCC2-E13

2) MCC2-E05 B. 1) MCC2-E07

2) DC SWBD 2-1 C. 1) MCC2-E13

2) DC SWBD 2-1 D. 1) MCC2-E07

2) MCC2-E05 Answer: C Explanation/Justification: KIA is met by demonstrating knowledge of the physical connections between AC and DC supplies to the UPS units, and demonstrating an understanding of the cause and effect relationship between AC source being lost, DC source will pick up the load.

A. Incorrect. Normal power supply is E-13. Incorrect answer of MCC2-E05 being the backup power supply if normal power is lost. Plausible distractor because MCC2-E05 is the backup regulated voltage supply to Vital Bus 2-1 if the inverter is removed from service.

B. Incorrect. Plausible distractor because MCC2-E07 is the backup regulated voltage supply to Vital Bus 2-3. DC SWBD 2-1 is the normal backup to the inverter.

C. Correct. Normal power supply is E-13, with DC SWBD 2-1 being the normal backup to the inverter.

D. Incorrect. Plausible distractor because MCC2-E07 is the backup regulated voltage supply to Vital Bus 2-3 and MCC2-E05 is the backup regulated voltage supply to Vital Bus 2-1 if the inverter is removed from service.

Sys# System Category KA Statement 063 DC Electrical K1 Knowledge of the physical connections and/or cause effect AC electrical system Distribution relationships between the DC electrical system and the following System systems:

KIA# K1.02 KIA Importance 2.7 Exam Level RO References provided to Candidate Technical

References:

20M-38.1.B Rev. 1, pg. 2 None RE-0001AW Rev.21 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 I 45.7 to 45.8)

Objective: 3SQS-38.1 Rev. 8, Obj. 2 From memory, describe the Normal System Arrangement for the Emergency 120 VAC Distribution Systems, including distribution paths, status of feeder breakers, loads, bus transfer switches, power train, and bus designation.

Beaver Valley Unit 2 NRC Written Exam (2LOT1S)

50. A large air break at the outlet of EOG 2-1 air receiver [2EGA*TK21A] is depressurizing the air receiver.

1) What is the minimum Tech Spec required Air Receiver pressure?

2) If air receiver [2EGA*TK21A] fully depressurizes, EOG 2-1 _ __...._(1;...r..)___ start upon receipt of an auto start signal.

A. 1) ~165 psig

2) will B. 1) ~380 psig

2) will C. 1) ~165 psig

2) will NOT

0. 1) ~380 psig

2) will NOT Answer: B Explanation/Justification: KIA is met by requiring knowledge of the EOG air system configuration and lineup, and the effects that a loss of one air receiver will have on the starting capabilities of the EOG.

A. Incorrect. Tech Spec minimum air pressure is 380 psig. BV1 & 2 use combined Tech Specs which identify both unit air pressures on the same page, this makes 165 psig a plausible distractor at BV. DG will start.

B. Correct. TS limit for air pressure is ~380 psig. The knowledge of the correct value is gained through performing OSTs, tech specs, and log taking. DG will start even with a rupture at the outlet of TK-21A due to there being 2 air systems/receivers which are not cross tied. This allows the pressurized receiver to admit air to 6 cylinders (1 /2) and start the diesel.

C. Incorrect. Tech Spec minimum air pressure is 380 psig. BV1 & 2 use combined Tech Specs which identify both unit air pressures on the same page, this makes 165 psig a plausible distractor at BV. DG will start as explained in the correct answer.

D. Incorrect. TS limit for air pressure is ~380 psig. DG will start as explained in the correct answer.

Sys# System Category KA Statement 064 Emergency Diesel K6 Knowledge of the effect of a loss or malfunction of the Air receivers Generator (ED/G) following will have on the ED/G system:

System KIA# K6.07 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

20M-36.1.C Rev. 4 pg. 8 20ST-36.1 rev. 71, pg. 9 U2 RM-0436-003 Rev. 20 Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective: 2SQS-36.2 Rev. 21 Obj. 9 Identify the EOG field instruments, subsystems and components that are required to be operable by the Technical Specifications. 20JT-1.36

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

51. The plant is at 75% power with all systems in normal alignment for this power level.

• Leak Collection Ventilation Radiation monitor 2RMR-RQl301 fails HIGH In response to this failure, where will the Contiguous areas exhaust be discharged?

The Contiguous areas exhaust will be directed through the filter banks and discharged A. through the Ventilation Vent to atmosphere B. through the Elevated Release to atmosphere C. to the Auxiliary Building D. to the Containment Building Answer: B Explanation/Justification: KIA is met by demonstrating knowledge that the normally unfiltered ventilation system realigns to filter the contiguous area exhaust before releasing it to the atmosphere when a process rad monitor malfunction occurs.

A. Incorrect. Plausible distractor because they must know that the Ventilation Vent is the normal discharge path for the Contiguous areas, but that it does not normally pass through the filter banks.

B. Correct. When 2RMR-RQl301 fails high, 2HVS-MOD201A&B will close isolating the Ventilation Vent flowpath, and 2HVS-MOD202A&B opens to align the contiguous areas to the filter banks. The only flowpath from the filter banks is through Leak Collection Filtered Exhaust fans to the Elevated Release to atmosphere.

C. Incorrect. Plausible distractor because they may think that the ventilation lineup will re-align the discharge of filtered exhaust to the surrounding area of the filter banks, which is the Auxiliary Building.

D. Incorrect. Plausible distractor because it could be thought that the ventilation would re-align the discharge of filtered exhaust to the containment building via the purge supply or exhaust.

Sys# System Category KA Statement 073 Process K3 Knowledge of the effect that a loss or malfunction of the PRM Radioactive effluent releases Radiation system will have on the following:

Monitoring (PRM)

System KIA# K3.01 KIA Importance 3.6 Exam Level RO References provided to Candidate Technical

References:

20M-16.1.D Rev.2 pg. 2 None 2SQS-16.1 PPNT Rev. 12 slide 6 Question Source: Bank- Vision #124119 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

52. Given the following plant conditions:

• The Unit was operating at 100% power with all systems in NSA

• An event occurred that caused containment pressure to peak at 6 psig

• Offsite Power has remained available for the duration of the event

• All System functions as designed Based on these plant conditions, which of the following combinations of reactor and turbine building components will have service water flow for temperature control?

CCP HX's = Primary Component Cooling Water Heat Exchangers CCS HX's = Secondary Component Cooling Water Heat Exchangers

=

EDG's Emergency Diesel Generators RSS HX's = Recirculation Spray Heat Exchangers CCP HX's CCS HX's EDG's RSS HX's A. YES YES YES YES B. YES YES YES NO C. NO NO NO NO D. YES NO YES NO Answer: D Explanation/Justification: K/A is met by the candidate predicting which of the listed components will have cooling water supplied after a an SI and CIA occur. The K/A statement is met by identifying that the CCP HXs (Rx plant CCW) will have temperature control capabilities and the CCS HXs (Turbine plant CCW) will not have temperature control.

A. Incorrect. CCS HX will isolate on SI/CIA. RSS HX's will be isolated until CIB actuates at 11.1 psig containment pressure.

B. Incorrect. CCS HX will isolate on SI/CIA.

C. Incorrect. CCP HX's will not isolate until CIB at 11.1 psig containment pressure so therefore will be providing flow and temperature control. EOG will have cooling even though they will be running unloaded in this plant configuration.

D. Correct. At> 5 psig containment pressure, SI and CIA have actuated. 2SWS*MOV107A-D close isolating CCS HX's, therefore there will be no cooling or temperature control to the CCS HX's. The SI signal will start EDGs and open 2SWS*MOV113A&D, therefore providing cooling to EDG's. CIB does not actuate until 11.1 psig, so therefore 2SWS*MOV106A&B will remain open providing cooling and therefore temperature control to the CCP HX's. 2SWS*MOV103A&B remain shut and do not open until containment pressure reaches 11.1 psig (CIB).

Sys# System Category KA Statement 076 Service A 1 Ability to predict and/or monitor changes in parameters (to Reactor and turbine building closed cooling water Water prevent exceeding design limits) associated with operating the SWS temperatures.

System controls including:

(SWS)

KIA# A1.02 KIA Importance 2.6* Exam Level RO References provided to Candidate None Technical

References:

20M-30.1.D, Rev. 8 2808-30.1 PPT, Rev. 23 Question Source: Bank - 2LOT7 NRC Exam (053) (2011)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

53. Given the following conditions:

• The plant is at 100% power

• Containment Instrument Air is being supplied by Station Instrument Air

• A Large Break Loss of Coolant Accident occurs

• All systems function as designed

• No operator actions have been taken Based on these plant conditions, which valve(s) will need to be reopened to restore instrument air to the containment?

1. 21AC-MOV130, CNMT Instrument Air lsol Viv.

2. 21AC-MOV131, CNMT Instrument Air Backup Supply Viv.

3. 21AC-MOV133, CNMT Instrument Air lsol Viv.

4. 21AC-MOV134, CNMT Instrument Air lsol Viv.

A. 1 ONLY.

B. 1AND2 ONLY.

C. 3 AND 4 ONLY.

D. 1, 2, AND 3.

Answer: A Explanation/Justification: KIA met with the required knowledge that CNMT instrument air is supplied from station instrument air, and that a CIA signal will close 21AC-MOV130 and isolate air to CNMT.

A. Correct. 21AC-MOV131 and 21AC*130 are open at 100% power to supply instrument air from instrument air compressors into containment. BVPS Unit 2 no longer uses containment air compressors. Upon a large break LOCA and SI, the subsequent CIA signal will auto close 21AC*130. In order to restore instrument air to containment, this valve needs to be reopened only.

B. Incorrect. Correct that 21AC*MOV130 needs to be reopened. Plausible if the candidate does not know that 21AC-MOV131 does not receive a CIA signal or believes this valve is affected by this signal. The EOP directs both of these valves opened, however, the EOP deals with all modes of operation and in the stated plant mode, the candidate must know it is not necessary to reopen 21AC-MOV131.

C. Incorrect. 21AC*MOV133 & 134 both receive a CIA signal and close. This was the old configuration when running CNMT IAC instrument air to containment. Opening these valves will not restore IA to containment.

D. Incorrect. All three of these valves receive a CIA signal and close from their NSA open positions. The candidate may believe that these valves all need to be reopened to restore instrument air.

Sys# System Category KA Statement 078 Instrument K1 Knowledge of the physical connections and/or cause-effect Containment air Air System relationships between the IAS and the following systems:

(IAS)

KIA# K1.03 KIA Importance 3.3* Exam Level RO References provided to Candidate None Technical

References:

20M-34.1.D Rev. 4 pg. 7 U2 RM-0434-003 rev. 17 20M-53A.1.E-O, Issue 2, Rev. 1, pg. 20 Question Source: Bank - 2LOT8 NRC Exam (053) (2012)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

54. When in Mode 1, what is the NSA required position of 2HVR*DMP206, Containment Vacuum Breaker Ball Valve, to prevent inadvertently breaking containment vacuum?

2HVR*DMP206 is CLOSED ~~~~~~~~~

A. with Instrument Air Isolated B. and De-energized C. and Chain Locked D. with Shorting Bar removed Answer: C Explanation/Justification: KIA is met by the knowledge of that CNMT vacuum breaker is chain locked closed to prevent inadvertent breaking of CNMT vacuum when in Modes 1-4. This is an administrative interlock.

Manually operated 2HVR*DMP206 has remote position indication on the Building Service Control Panel in the Control Room. With this indication available in the CR, it helps to make all incorrect distractors plausible as the candidate may think it is an electrically operated valve.

A. Incorrect. Plausible means of failing an air operated valve closed.

B. Incorrect. Plausible means of failing a motor operated valve in a desired position.

C. Correct. IAW 20ST-48.7, 2HVR*DMP206 is required to be chain locked closed in Modes 1-4.

D. Incorrect. Plausible means of removing power from the contactor to prevent valve movement.

Sys# System Category KA Statement 103 Containment K4 Knowledge of containment system design feature(s) and/or Vacuum breaker protection System interlock(s) which provide for the following:

KIA# K4.01 KIA Importance 3.0* Exam Level RO References provided to Candidate None Technical

References:

20M-44C.4.A Rev. 23 pg 4 RM-0444C-002 Rev. 7 20ST-48.7 Rev. 41 pg 16 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

55. The plant is at 50% power.

• Annunciator A 1-1 E, Containment Air Pressure High/Low has alarmed

• 2CVS-Pl101A1 CNMT PRESS is reading 14.1 psia

• No other Control Room annunciators are in alarm Which of the actions below will clear annunciator A 1-1 E in accordance with 20M-12.4.AAA, "Containment Air Pressure High/Low"?

A. Align the Containment Vacuum Ejector for use B. Shutdown the CNMT Vacuum Pumps C. Shutdown an operating CNMT Air Recirc (CAR) Fan D. Start a CNMT Vacuum Pump Answer: D Explanation/Justification: KJA is met by demonstrating the ability to recognize a CNMT high pressure condition from the Control room, and respond by manually starting the CNMT vacuum pump to restore pressure.

A. Incorrect. This is a plausible means of lowering cnmt pressure, but it is used to draw initial cnmt vacuum. It is not an approved method iaw 20M-12.4.AAA to lower cnmt pressure.

B. Incorrect. This is a plausible distractor if it is thought that cnmt pressure is low due to operating the cnmt vacuum pump. Incorrect because cnmt pressure is high.

C. Incorrect. Plausible distractor if it is thought that cnmt pressure is low due to low temperature, and must be raised back into normal band.

Probable cause #5 of ARP states is pressure is low due to temperature being low, the stop CAR fan. Incorrect because cnmt pressure is high.

D. Correct. Must recognize that cnmt pressure is high and must be lowered to 13.4-13.6 psia using the vacuum pumps. This is directed by 20M-12.4.E "Maintaining the Containment Vacuum" which 20M-12.4.AAA references. This knowledge of the setpoint range can be determined by the above the line RO knowledge for TS 3.6.4 pressure limits of ~12.8 - S14.2 psia.

Sys# System Category KA Statement 103 Containment A4 Ability to manually operate and/or monitor in the control room: Containment vacuum system System KIA# A4.09 KIA Importance 3.1* Exam Level RO References provided to Candidate Technical

References:

20M-12.4.AAA Rev. 5 pg. 6 None 20M-12.4.E Rev. 4 pg. 2 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

56. The plant is at 75% power with all systems in normal alignment for this power level EXCEPT PRZR Pressure Controller 2RCS*PK444A is in MANUAL controlling pressure at 2235 psig.

• All plant parameters are on program

• A 20% Load Rejection occurs

• No operator actions have occurred As compared to the initial conditions, what is the status of the PRZR level and pressure 5 minutes after the Load Rejection occurred?

PRZR Level PRZR Pressure A. Lower Lower B. Lower Higher C. Higher Lower D. Higher Higher Answer: A Explanation/Justification: K/A is met by predicting the effect that the control rods inserting during a load rejection will have on both PRZR level and pressure.

A. Correct. With a 20% load rejection Tref will be at a lower value than initial. Rod control will drive the rods in to get Tavg down to within 1.5F of Tref. This will cause Tavg to lower. PRZR program level control (22-53% program) is based on Tavg (547-574F), therefore, PRZR level will be lower. With 2RCS*PK444A in manual (approx. 42% demand for 2235 psig), when the LR occurs PRZR pressure will lower, since the spray valves remain open (where as if in auto they would close at 40.6%), pressure will drive lower than expected, and it will take longer for heaters to recover pressure because all Backup heaters will not energize. (all Backup heaters turn on at 9.4% demand if in auto).

8. Incorrect. Plausible distractor if it is thought that PRZR Pressure Controller in manual will allow pressure to rise about initial pressure and remain higher.

C. Incorrect. Plausible distractor if the candidate does not have a thorough understanding of PRZR level controller programming and Rod Control Temperature control. Pressure will be lower than initial.

D. Incorrect. Plausible distractor if the candidate does not have a thorough understanding of PRZR level controller programming and Rod Control Temperature control. Pressure would be lower than initial.

Sys# System Category KA Statement 001 Control Rod A1 Ability to predict and/or monitor changes in parameters (to PZR level and pressures Drive System prevent exceeding design limits) associated with operating the CRDS controls including:

KIA# A1.04 KIA Importance 3.7 Exam Level RO References provided to Candidate Technical

References:

20M-1.1.B Rev. 6 pg 10 None 20M-1.5.A.48 lss. 1 Rev. 1 20M-6.4.IF Rev. 13 Pgs. 24 & 25 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

57. The following conditions exist:

• The Plant is at 100% power

• PRZR PRESS CONTROL 2RCS*PK444A is in AUTO set at 2235 psig

• PRZR HEATERS CONTROL GROUP C [2RCP-H2C] CS has a RED target

• PRZR HEATERS BACKUP GROUP B & D [2RCP-H2B & H2D] CS have RED targets

• PRZR HEATERS BACKUP GROUP A & E [2RCP-H2A & H2E] CS have GREEN targets

[2RCS*PCV455B] 'B' PRZR SPRAY VALVE Fails OPEN.

Which of the following is the correct order of automatic actions that occur as RCS pressure is lowering?

1. PRZR Backup Heaters ON

2. 'A' PRZR Spray Valve [2RCS*PCV455A] CLOSED

3. PRZR Heaters Control Group C [2RCP-H2C] ON A. 2, 3, 1 B. 2, 1, 3 C. 3, 2, 1 D. 3, 1, 2 Answer: A Explanation/Justification: KJA is met by the candidates ability to identify automatic actions which occur as RCS pressure is lowering, this includes automatic operation of PRZR heaters which raise PRZR pressure, and PRZR spray flow control on the non-faulted spray valve.

A. Correct. The Master Pressure Controller will respond to the lowering pressure by controlling the spray valve and heaters. The MPC output will be driving to 0 as pressure is lowering. PCV455A will close at 40.6, Control Htrs will come on at 34.4, B/U htrs will energize at 9.4 (2210 psig).

Candidate must have the knowledge of normal plant operations where as some spray flow is desired because the balancing of spray flow and heaters results in a normal, constant Przr outsurge.

B. Incorrect. See correct explanation.

C. Incorrect. See correct explanation.

D. Incorrect. See correct explanation.

Sys# System Category KA Statement 002 Reactor Coolant A3 Ability to monitor automatic operation of the RCS, Pressure, temperatures, and flows System (RCS) including:

KIA# A3.03 KIA Importance 4.4 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF Att. 2 Rev. 13 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

58. The plant is at 100% power.

• PRZR Level Control Channel Selector is in Channel I & II (LT459 & 460)

• PRZR level is on program The reference leg for PRZR Channel I Level [2RCS*LT 459] develops a leak.

How will 2RCS-Ll459 initially respond and, if the event continues with NO operator action, what automatic trip signal will initiate the reactor trip?

Rx trip 2RCS*Ll459 Signal A. Rise High Pressurizer Level B. Lower High Pressurizer Level C. Rise Low Pressurizer Pressure D. Lower Low Pressurizer Pressure Answer: A Explanation/Justification: KIA is met by demonstrating the knowledge to understand the effects a reference leg failure on one of the post accident monitor PRZR level indicators will have of the PRZR Level Control System, and the overall function that the PAM PRZR level indicators have as reactor trip inputs. At BVPS all three of the PAM przr level instruments are used for the level control system.

A. Correct. A reference leg leak will cause the affected PAM channel (LT459) to indicate high. Since LT459 is the controlling channel, it will cause 2CHS-FCV122 to close to minimum flow, thus causing PRZR level to initially lower until letdown isolates at 14% on LT460. After UD isolates, PRZR level will rise due to charging flow (minimum flow of 25 gpm when FCV122 is automatically closed) and seal injection, until PRZR level reaches 92% on 2/3 indicator when > 10% power. This will generate a rx trip.

B. Incorrect. Plausible if the candidate doesn't understand the difference between a reference leg and a variable leg leak. Because a variable leg leak on LT459 would cause indicated level to lower. A rx trip would be generated on high przr level due to LT460 isolating UD.

C. Incorrect. A reference leg leak will cause the affected PAM channel (LT459) to indicate high. Plausible distractor of Rx trip on low PRZR pressure if candidate feels that that a leak will cause a low pressure Rx trip. The reference leg is 3/4" line with a 3/8" flow restrictor in line.

D. Incorrect. Plausible if the candidate doesn't understand the difference between a reference leg and a variable leg leak. Because a variable leg leak on LT459 would cause indicated level to lower. Plausible distractor of Rx trip on low PRZR pressure if candidate feels that that a leak will cause a low pressure Rx trip. The reference leg is 3/4" line with a 3/8" flow restrictor in line.

Sys # System Category KA Statement 011 Pressurizer Level Control K6 Knowledge of the effect of a loss or malfunction on the Function of PZR level gauges as post System (PZR LCS) following will have on the PZR LCS: accident monitors KIA# K6.05 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

GO-GPF.C7 Rev. 4 pg. 55, 20M-1.5.B.1 Rev. 2 20M-6.4.IF, attachment 1, rev 13 U2 RM-0406-003 Rev. 6 Question Source: Bank - Vision #131762 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

59. The following conditions exist.

• A large break LOCA has occurred

• TSC has been activated

• Annunciator A 1-28, Hydrogen Level High/High-High is in alarm Complete the following statements. Assume the TSC has been c decision.

The High-High Hydrogen concentration in Containm (1)

In accordance with the Hydrogen Level High/H" -High ARP, the crew will _ _-->,.;;;(2_,_)_ _ _ in response to the High-High Hydrogen level i ontainment.

A. 1) 2.5%

2) intentionally ignite the C B. 1) 2.5%

2) start [2HCS-F ] Containment Atmosphere Purge Blower C. 1)

2) ally ignite the Containment atmosphere D. .5%

start 2HCS-FN21 Answer: D Explanation/Justification: BVPS2 has retired the H2 recombiners and explosive H2 concentration in the Cnmt is beyond design based accident.

We discussed the KIA with Chief Examiner who stated to stay focused on the purge control portion of the system.

KIA is met knowledge of the operation of the containment purge system in the event of a High-High Hydrogen concentration in the containment.

A. Incorrect. 2.5% was chosen to be a realistic choice verses 0.5% which is the High H2 alarm setpoint. Plausible distractor of igniting the atmosphere since this is an option in our severe accident management guidelines (20M-53E.1.SAG-7), but this is not the correct response at this H2 level or iaw the ARP.

8. Incorrect. 2.5% was chosen to be a realistic choice verses 0.5% which is the High H2 alarm setpoint. This is the correct response iaw the ARP.

C. Incorrect. Correct setpoint for the High-High alarm. Plausible distractor of igniting the atmosphere since this is an option in our severe accident management guidelines (20M-53E.1.SAG-7), but this is not the correct response at this H2 level or iaw the ARP.

D. Correct. Correct setpoint for the High-High alarm. Correct actions iaw the ARP.

Sys # System Category KA Statement 028 Hydrogen Recombiner and K5 Knowledge of the operational implications of the Explosive hydrogen concentration Purge Control System (HRPS) following concepts as they apply to the HRPS:

K/A# K5.01 KJA Importance 3.4 Exam Level RO References provided to Candidate Technical

References:

20M-46.4.ABD Rev. 3 pg. 5 None Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: 2SQS-46.1 Obj. 15 Describe the control, protection and interlock functions for the control room components associated with Post OBA Hydrogen Control System, including automatic functions, setpoints and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

60. Given the following:

• The plant is in Mode 5

• A Forced Containment Purge through the SLCRS Unfiltered flow path is in progress

• The Personnel Airlock and Equipment Hatches are closed

• The Cnmt Purge Supply lsol Damper 2HVR*MOD25A inadvertently CLOSES Assuming NO Operator actions, how will containment pressure compare to pre-event conditions 5 minutes after 2HVR*MOD25A closes, and what will be the status of 2HVS-FN263B, Leak Collection Normal Exhaust Fan?

Containment pressure will be _ ___._(1"-J.)___ it was before 2HVR*MOD25A closed.

2HVS-FN263B, Leak Collection Normal Exhaust Fan will be (2)

A. 1) the same as

2) tripped B. 1) the same as

2) running C. 1) lower than

2) tripped D. 1) lower than

2) running Answer: D Explanation/Justification: KIA is met with the knowledge of the Containment Purge system being aligned for forced purge, and the Cnmt Purge Supply lsol Damper goes closed and the effect this failure will have on containment pressure.

A. Incorrect. Cnmt pressure will lower due to the air being drawn from the cnmt by 2HVS-FN263B. Plausible distractor if candidate thinks that 2HVR-MOD23A closes on interlock with the MOD25A closing, thus causing the containment to be isolated from the purge supply and exhaust flow paths. Plausible distractor because FN263B is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed.

B. Incorrect. Cnmt pressure will lower due to the air being drawn from the cnmt by 2HVS-FN263B. Plausible distractor if candidate thinks that 2HVR-MOD23A closes on interlock with the MOD25A closing, thus causing the containment to be isolated from the purge supply and exhaust flow paths. It is correct that FN263B will continue to run.

C. Incorrect. Cnmt pressure will lower, but 2HVS-FN263B will not be tripped. Plausible distractor because FN263B is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed. Candidate must know the difference between FN263A and FN263B, and which is used for the forced purge lineup.

D. Correct. When 2HVR*MOD25A closes it will isolate the flowpath from 2HVP-ACU211 B, PAB NC Unit which is the forced air side of the cnmt purge. This will leave only 2HVS-FN263B drawing airflow from the cnmt causing cnmt pressure to lower. 2HVS-FN263B will be running because it is not effected by the closure of Cnmt Purge Supply lsol Damper. It is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed not MOD25A.

Sys # System Category KA Statement 029 Containment K3 Knowledge of the effect that a loss or malfunction of the Containment parameters Purge Containment Purge System will have on the following:

System (CPS)

KIA# K3.01 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

U2 RM-0444C-002 Rev 7, U2 RM-0444D-001 Rev. 8 U2 RM-0416-001 rev. 12 20M-44C.4.A Rev. 23 pg. 5 20M-16.1.D Rev. 2 pg. 3 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

61. Which of the following completes the statement below?

During refueling operations the fuel assemblies are removed from the core using the Manipulator crane (1) hoist which is equipped with an automatic overload interlock that is set at a maximum of (2) pounds.

A. 1) main

2) 2700

8. 1) main

2) 3000 C. 1) auxiliary

2) 2700 D. 1) auxiliary

2) 3000 Answer: A Explanation/Justification: K/A is met with the knowledge of the functions of the refueling equipment and the hoist overload protection associated with the manipulator crane main hoist..

A. Correct. The main hoist is raises and lowers the gripper tube and removes fuel assemblies from the core. The overload limit is set at S2700 pounds.

B. Incorrect. The main hoist is raises and lowers the gripper tube and removes fuel assemblies from the core, but the overload limit is not set at 3000 pounds.

C. Incorrect. The auxiliary hoist is located on the bridge monorail for handling accessory refueling equipment, not fuel assemblies. The overload limit is set at S2700 pounds.

D. Incorrect. The auxiliary hoist is located on the bridge monorail for handling accessory refueling equipment, not fuel assemblies, and the overload limit is not set at 3000 pounds.

Sys# System Category KA Statement 034 Fuel Handling K4 Knowledge of design feature(s) and/or interlock(s) which provide Overload protection Equipment for the following:

System (FHES)

KIA# K4.03 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

LP 3SQS-6.13 Rev. 6 pg 12 2RP-3.3, Rev. 5 lss. 0, pg 3 & 21 Question Source: Bank - Harris 2011 NRC Exam (061)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective: 3SQS-6.13 Rev. 6 (1) Describe the function of the following Fuel Handling equipment as documented in the Refueling Procedures:

Manipulator crane, Manipulator crane auxiliary hoist. (2) Describe the control, protection and interlock functions for the fuel handling equipment, including automatic functions, setpoints and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

62. 1) What is the Steam Generator level setpoint for the Auxiliary Feedwater Pumps to automatically start?

2) In accordance with E-0 "Reactor Trip or Safety Injection", what is the basis for Auxiliary Feedwater automatic initiation?

A. 1) 20.5% Narrow Range Level

2) Provide a secondary heat sink.

B. 1) 20.5% Narrow Range Level

2) Prevent Steam Generator dryout.

C. 1) 19.6% Narrow Range Level

2) Provide a secondary heat sink.

D. 1) 19.6% Narrow Range Level

2) Prevent Steam Generator dryout.

Answer: A Explanation/Justification: KIA is met by the candidate demonstrating the knowledge of the design feature of the AFW pump start setpoints, and that the reason AFW starts is to supply feed to the SG for decay heat removal.

A. Correct. 2/3 detectors indicating 20.5% NR level (low-low setpoint) on 2/3 SGs will start the MDAFW, (2/3 on 1 SG for TDAFW) pumps to keep the tubes covered for secondary heat removal.

B. Incorrect. Plausible because the setpoint is correct. Incorrect basis. It could be thought that aux feedwater was provided just to prevent SG dryout, however, the reason is to keep the tubes covered for secondary heat removal.

C. Incorrect. Plausible level because 19.6% is the Unit 1 setpoint. Correct bases.

D. Incorrect. Plausible because 19.6% is the Unit 1 setpoint. Incorrect bases. It could be thought that aux feedwater was provided just to prevent SG dryout, however, the reason is to keep the tubes covered for secondary heat removal.

Sys# System Category KA Statement 035 Steam K4 Knowledge of S/GS design feature(s) and/or interlock(s) which Amount of reserve water in SIG Generator provide for the following:

System (S/GS)

KIA# K4.05 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.E-O lss. 2 Rev. 1 pg. 16 20M-24.1.D Rev. 6 pg. 16 20M-24.2.B Rev. 16 pg. 4 Question Source: Bank - Comanche Peak 2013 NRC Exam (035)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

63. Plant conditions have been established to perform 20ST-26.8, "Main Turbine Overspeed Trip Test".

• Reactor power is stable at 13%

• The generator is NOT synchronized to the grid

• Overspeed Protection Test Switch is in the "lnservice" position

• As Turbine speed is raised, Annunciator A7-2E "Turbine Overspeed Prot Controller Operating" alarms

• SIG pressures are 1000 psig and stable When Turbine speed reaches the overspeed protection controller setpoint, which Turbine EHC valves will close?

(IV =Intercept valves, GV =Governor valves, TV =Throttle valves)

A. Only IVs and GVs Close B. Only GVs, and TVs Close C. Only TVs and IVs Close D. ALL IVs, GVs, and TVs Close Answer: A Explanation/Justification: KJA is met by the understanding that when testing the turbine overspeed, and the annunciator for "Turbine Overspeed Prot Controller Operating" alarms, that only the IVs and GVs will be effected by the OPC.

A. Correct. At 103% (1844-1864 rpm) the Overspeed Protection Controller actuates causing all 4 GVs and all 4 IVs to close. The TVs are not closed by the OPC. Position indication of Turbine valves can be monitored on BB-C.

B. Incorrect. The TVs are not closed by the OPC which makes the distractor incorrect.

C. Incorrect. It is correct that the IVs will close, but the TVs are not closed by the OPC.

D. Incorrect. The TVs are not closed by the OPC which makes the distractor incorrect.

Sys# System Category KA Statement 045 Main Turbine A4. Ability to manually operate and/or monitor in the control room: Turbine valve indicators (throttle, governor, Generator control, stop, intercept), alarms, and annunciators (MT/G)

System KIA# A4.01 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

20M-26.4.AAU lss.1 Rev. 5 20ST-26.8 Rev. 16 pg.15 Question Source: Bank - Vision #138661 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

64. The following conditions exist:

• Plant is at 80% power

• Control Room ACU Outside Air Intake and Exhaust Dampers 2HVC*MOD201 B & D automatically CLOSED.

• Control Room Emergency Supply Fan [2HVC*FN241 B] automatically STARTS 120 seconds after the Control Room isolation occurs Which of the following radiation monitors, and setpoint would cause the above ventilation lineup?

A. Control Room Area [2RMC*RQ201] radiation monitor above the ALERT setpoint.

B. Control Room Area [2RMC*RQ202] radiation monitor above the HIGH setpoint.

C. Control Room Airborne Particulate [2RMC-RQ301A] radiation monitor above the ALERT setpoint.

D. Control Room Airborne Gas [2RMC-RQ301 B] radiation monitor above the HIGH setpoint.

Answer: B Explanation/Justification: KIA is met by re-aligning the Control Room ventilation system and having the candidate demonstrate the ability to determine which radiation monitor alarm would cause the automatic ventilation alignment.

A. Incorrect. Plausible distractor because RQ201 does initiate CR isolation at the HIGH setpoint, but not at the alert setpoint.

B. Correct. RQ202 does initiate a CR isolation when at High setpoint.

C. Incorrect. Plausible distractor with it being a CR rad monitor. RQ301A will not initiate CR isolation.

D. Incorrect. Plausible distractor with it being a CR rad monitor. RQ301 B will not initiate CR isolation.

Sys# System Category KA Statement 072 Area Generic Ability to verify that the alarms are consistent with Radiation the plant conditions.

Monitoring (ARM)

System KIA# 2.4.46 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

20M-43.4.ADB Rev.7 pg. 2 20M-43.1.C Rev. 5 pg. 24 1/20ST-43.17D Rev. 44 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.3 /

45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

65. Initial conditions:

• The plant is at 100% power

• 20ST-30.6B, Service Water Pump [2SWS*P21C] Test on Train B Header is in progress

• 2SWS*P21 C, 'C' SWS is RUNNING on 2DF Bus

• 2SWS*P21 B, 'B' SWS is racked on the bus with the CS in Auto-After Stop Current Conditions:

• A Loss of Offsite Power occurred following a design basis earthquake

• The Control Room crew is performing actions of E-0, "Reactor Trip or Safety Injection"

• A8-2B, 4160V EMER BUS 2AE ACB 2E7 OVERCURRENT TRP is LIT With NO Operator action, one minute after the Loss of Offsite Power, which of the following statements describe the status of the Service Water Pumps?

A. 2SWS*P21A and 2SWS*P21 Bare running B. 2SWS*P21A and 2SWS*P21C are running C. Only 2SWS*P21 B running D. Only 2SWS*P21C running Answer: C Explanation/Justification: KIA is met by demonstrating the knowledge of the available power and pump start interlocks of the essential SWS pumps following a loss of offsite power coincident with a failure of 2AE emergency bus to load.

A. Incorrect. 'A' SWS pump will not be energized due to bus 2AE being de-energized with A8-2B annunciator lit. 2-1 DG will start but 2E10 will not close due to the overcurrent trip on 2E7. The candidate must know the effects of this annunciator condition. 'B' SWS will auto start even though the CS is in Auto-After Stop and the 'C' SWS is racked in on the DF bus.

B. Incorrect. 'A' SWS pump will not be energized due to bus 2AE being de-energized with A8-2B annunciator lit. 2-1 DG will start but 2E10 will not close due to the overcurrent trip on 2E7. The candidate must know the effects of this annunciator condition. 'C' SWS will not auto start even though power is available and it was running, due the 'B' SWS pump being racked onto the 2DF bus with the CS in Auto-After Stop. It is still the priority pump and will be loaded on the diesel.

C. Correct. With the CS in Auto-After Stop, 'B' SWS pump is the priority pump and will be loaded on the diesel even though the 'C' SWS is racked onto the 2DF bus.

D. Incorrect. Even though 'C' SWS pump has power available and was running, 'B' SWS pump will auto start and load onto the diesel because it the riorit um .

Sys# System Category KA Statement 075 Circulating K2 Knowledge of bus power supplies to the following: Emergency/essential SWS pumps Water System KIA# K2.03 KIA Importance 2.6* Exam Level RO References provided to Candidate Technical

References:

20M-30.1.D Rev. 8, Pgs. 2-4 None U2 LSK-017-001A Rev. 14 20M-36.4.ACD Rev. 3 pg. 3 3SQS-36.1 PPNT U2 Rev. 12 lss. 1 Slide 10 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

66. The plant is in Mode 5 preparing to enter Mode 4.

• Valve alignments are being performed on a Safety-Related system

• The REQUIRED NSA position of a manually operated globe valve is OPEN

• The valve must be in this position PRIOR to Mode 4 entry

• The Independent Verifier will receive 8 mR performing the Independent Verification (IV)

IAW the guidance provided in NOP-OP-1002, Conduct of Operations, how could the Independent Verification for this valve be addressed?

A. The Operations Manager has the authority to waive the IV for equipment concerns.

B. The IV may be performed by using the Plant Computer System (PCS) if "Not Closed" is indicated.

C. The Shift Manager can waive the IV due to dose limits.

D. The IV may be performed by a functional test that can prove the valve is open.

Answer: D Explanation/Justification: KIA is met by demonstrating the knowledge of alternative independent verification means for a valve lineup in accordance with the Conduct of Operations manual.

A. Incorrect. Plausible distractor because it may be assumed that the Operations Manager would have this authority, but that is not correct.

B. Incorrect. Plausible distractor but not a reliable alternative to hands on verification. The valve is required to be OPEN, but the remote indication of NOT-CLOSED only means that the valve is not fully closed.

C. Incorrect. Plausible distractor because the SM can waive the IV if it would result in a radiation exposure greater than 1O mRem.

D. Correct. A functional test may be used for an IV iaw NOP-OP-1002 sect. 4.18.2.6.

Sys# System Category KA Statement N/A N/A Generic Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc KIA# 2.1.29 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-1002 Rev. 10 pg. 81 Question Source: Bank - 2LOT6 NRC Exam (066) Modified (2009)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 45.1 I 45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

67. Given the following plant conditions:

• The plant is operating at 90%

• VCT level is 38% and stable Then VCT Level Transmitter, 2CHS*LT112 fails HIGH Which of the following completes the statements below?

(Assume NO operator action)

1) VCT level on 2CHS*LT115 will ( 1)

2) VCT Auto Makeup will be (2)

A. 1) lower

2) available B. 1) lower

2) unavailable C. 1) remain unchanged

2) available D. 1) remain unchanged

2) unavailable Answer: A Explanation/Justification: KJA is met by the candidate recognizing how the system will respond to one VCT level control indication failing high, and validating how the other VCT control indication will respond to this failure.

A. Correct. With LT112 failing high, both LCV112 and LCV115A will reposition to divert and lower the actual level. As VCT level lowers to 20%,

LT115 will start Auto Makeup and try to maintain level 20-40%.

B. Incorrect. Actual level will lower. LT115 is unaffected, therefore as actual level lowers, Auto Makeup will start to maintain level 20-40%.

C. Incorrect. Actual level remaining unchanged would be true if LT112 failed low. It is correct that auto makeup will be available.

D. Incorrect. Actual level remaining unchanged would be true if LT112 failed low. It is incorrect that auto makeup will be available.

Sys# System Category KA Statement N/A N/A Generic Ability to identify and interpret diverse indications to validate the response of another indication.

KIA# 2.1.45 KIA Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-7.4.IF Rev. 3 Att. 1 Question Source: Bank- Harris 2012 NRC Exam (Q67)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 43.51 45.4)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

68. You are performing a procedure out in the plant and you note a typographical error in the step you are about to perform. In accordance with NOP-LP-2601, 'Procedure/Work Instruction Use and Adherence', what action are you required to perform.

A. Have a second qualified Operator peer check the typographical error, continue with the activity, and inform your supervisor upon completion.

B. Contact your supervisor, identify the typographical error, have the supervisor annotate issue in the procedure, and then continue with the activity.

C. Contact your supervisor, identify the typographical error, and perform a Limited Use Change.

D. Contact your supervisor, identify the typographical error, and Revise the procedure.

Answer: B Explanation/Justification: KIA is met with the knowledge of the expected response when a procedure is found to have a typographical error, and how to make the necessary changes to the procedure prior to completing work.

A. Incorrect. By continuing on in the procedure without discussing it with the authorizing authority or responsible supervisor would be a violation of NOP-OP-2601. A peer check by another qualified Operator does not meet the site expectations.

B. Correct. Per NOP-LP-2601, if a typo is discovered, the performer must stop the work, ensure equipment is in a safe condition, and contact they're supervisor. Clearly identify typo by annotating the procedure and then continue with the activity.

C. Incorrect. A Limited Use Change is not required for a typographical error D. Incorrect. A procedure Revision is not required for a typographical error.

Sys# System Category KA Statement NIA N/A Generic Knowledge of the process for making changes to procedures.

KIA# 2.2.6 KIA Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

NOP-LP-2601 rev.5, pg.15 & 16 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.3 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

69. Given the following plant conditions:

• The Unit is operating at 100%.

• You have just returned from a day off and are reviewing the narrative logs.

• 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago, a valve was repositioned out of NSA and selected as an OPEN item using the Short Term Configuration Change Process.

Which of the following statements correctly describes requirements of NOP-OP-1014, "Plant Status Control?

A. A clearance will be necessary if restoration does not occur within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. A system status print sheet will be necessary if restoration does not occur within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. A clearance should have been posted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

D. A system status print sheet should have been issued 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

Answer: C Explanation/Justification: KIA is met by the ability to determine a valve has exceeded the short term configuration change process, and identify the correct actions that should have been take in accordance with NOP-OP-1014, Plant Status Control procedure.

A. Incorrect. Refer to correct answer explanation. The candidate may believe the requirement is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as opposed to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Incorrect. Refer to incorrect choice D explanation. Plausible and balanced distractor.

C. Correct. According to NOP-OP-1014, if a component is not restored to its normal configuration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then a clearance is hung to provide a plant status control tracking method and documentation of the deviation from the components normal alignment. A clearance should have been posted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

D. Incorrect. A System Status Print is required to be filled out at all times reflecting system status conditions, if the system is deemed necessary by the Ops Manager. If it was not deemed necessary, the system status print would not be required. If it was deemed necessary, then it should have been filled out 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago.

Sys# System Category KA Statement NIA N/A Generic Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

KIA# 2.2.15 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-1014, Rev. 4, pg. 14 Question Source: Bank - 1LOTS NRC EXAM (Q96) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.3 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

70. The plant is at 100% power.

Which of the following conditions or events (considered individually) will require Technical Specification action(s) to be performed within one hour or less?

A. RWST borated water temperature drops to 50 °F.

B. One Containment Pressure Transmitter fails to zero.

C. RWST borated water volume drops to 840,200 gallons.

D. BOTH Train "A" - Phase B (CIB) manual Control Switches are declared inoperable.

Answer: C Explanation/Justification: KIA is met by the knowledge required to recognize the RWST level is below Tech Spec require level and is as; 1 hr. TS action statement.

A. Incorrect. TS 3.5.4 Surveillance requires RWST borated water temperature to be <!45 F ands; 65 F. therefore there is no TS LCO entry required for this distractor.

B. Incorrect. TS 3.3.2 Condition D & E apply. The channel is required to be placed in trip/bypass within 72 hrs.

C. Correct. TS 3.5.4 Condition B states that if RWST is inoperable for reasons other than boron concentration or temperature (Condition A), then a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statement is applicable. SR 3.5.4.2 requires Unit 2 RWST level to be <! 859248 gallons. If this surveillance is not met then TS LCO actions apply. RO's are required to knows; 1hour TS LCO's from memory.

D. Incorrect. TS 3.3.2 Condition B applies. This is a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> action statement.

Sys# System Category KA Statement N/A N/A Generic Knowledge of less than or equal to one hour Technical Specification action statements for systems.

KIA# 2.2.39 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

BVPS TS pg. 3.5.4.1 & 2 Amend. 278/161 Question Source: Bank- 1LOTS NRC Exam (041)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.2 /

45.13)

Objective: 2SQS-13.1 Rev. 18 Obj. 18 - For a given set of plant conditions, determine if the condition meets the criteria for entry into a one hour or less action statement in accordance with the Technical Specifications.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

71. You are going into a contaminated area, which has the following radiological characteristics to perform a valve lineup.

• Your current exposure for the year is 938 mrem

• The RWP states:

o General area dose rate =30 mrem/hr o Airborne contamination concentration =10.0 DAC

• The valve lineup will take you 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if you wear a full-face respirator.

• The valve lineup will only take you 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if you do NOT wear the respirator.

1) Which of the following choices for completing this job would maintain your exposure within the station administrative requirements and the principles of ALARA?

2) Why is this action appropriate?

A. 1) You must wear the respirator.

2) You will exceed DAC limits if you do NOT wear a respirator.

B. 1) You must wear the respirator.

2) Your calculated TEDE dose received will be less than if you do NOT wear a respirator.

C. 1) You should NOT wear the respirator.

2) Your calculated TEDE dose received will be less than if you do wear a respirator.

D. 1) You should NOT wear the respirator.

2) Your dose received wearing a respirator will exceed the site annual personnel dose limits.

Answer: C Explanation/Justification: K/A is met by demonstrating the ability to comply with an RWP to determine dose received with or without a respirator to achieve the lowest possible dose for a job.

A. Incorrect. This answer is plausible if the applicant does not understand the concept of DAC-hours and DAC-hour limits.

B. Incorrect. This answer is plausible if the applicant incorrectly calculates the exposure.

C. Correct. Without respirator: TEDE = 30 mrem/hr x 1 hr= 30 mrem, From airborne contamination: TEDE = 10 DACx1 hr x 2.5 mrem/DAC-hr = 25 mrem, TEDE = 30 + 25 = 55 mrem from job, Total exposure for year= 938 + 55 = 993 mrem With respirator, TEDE = 30 mrem/hr x 2 hr= 60 mrem TEDE = 60 mrem, Total exposure for year= 938 + 60 = 998 mrem TEDE = 60 mrem-vs-55 mrem = do not use a respirator D. Incorrect. This answer is plausible if the applicant miscalculates the dose.

Sys # System Category KA Statement N/A NIA Generic Ability to comply with radiation work permit requirements during normal or abnormal conditions.

KIA# 2.3.7 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-4201 Rev. 2 pg. 20 FENRWT Rev 3 CNRR 08-08-14 Handout pg.26 & 64 Question Source: Bank-McGuire 2012 NRC Exam (Q72)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.12 I 45.10)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

72. The following conditions exist:

• Plant is operating at 100% power

• Radiation level is 1,324 mrem/hr at 30 centimeters from the Letdown piping

• You have been assigned to enter the Letdown cubicle and hang a clearance Which of the following identifies the radiation area posting at the cubicle entrance, and the minimum approval authority for entry in accordance with NOP-OP-4101, 'Access Controls for Radiologically Controlled Areas'?

Letdown Cubicle Posting Minimum Approval A. High Radiation Area (HRA) Radiation Protection Manager B. High Radiation Area (HRA) Radiation Protection Supervisor C. Locked High Radiation Area (LHRA) Radiation Protection Manager D. Locked High Radiation Area (LHRA) Radiation Protection Supervisor Answer: D Explanation/Justification: KIA is met by identifying the area as a LHRA and determine who must give permission to enter the LHRA in order to hang a clearance.

A. Incorrect. HRA is An accessible area in which radiation levels could result in an individual receiving a deep-dose equivalent in excess of ;?100 mrem/hr at a distance of 30 centimeters or more from a radiation source or from any surface that the radiation penetrates. The RPM approval is only required if the gen area dose was >2.5 rem/hr, or it was a Very High Rad Area.

B. Incorrect. For posting, see explanation above. Radiation Protection Supervisor is the correct authorization.

C. Incorrect. LHRA is the correct posting. It is incorrect that the RPM must give permission. The RPM approval is only required if the gen area dose was >2.5 rem/hr, or it was a Very High Rad Area.

D. Correct. LHRA is A locked area with an accessible area to individuals, in which radiation levels could result in dose rates ;?1,000 mrem/hr at a distance of 30 centimeters from a radiation source or from any surface that the radiation penetrates. The RP Supervisor must give approval for entry into the LHRA as long as the general area dose rate is <2.5Rem/hr.

Sys # System Category KA Statement NIA N/A Generic 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.

KIA# 2.3.13 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical

References:

NOP-OP-4101 Rev. 11 Pg. 5 & 17 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.12 I 43.4 I 45.9 I 45.10)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

73. The plant is operating at 100% power.

• A Small Break LOCA occurs.

• The crew is performing the actions of ES-1.2, "Post LOCA Cooldown and Depressurization".

• All SI pumps are running.

• All RCPs are running.

• RCS cooldown via Condenser Steam Dumps is ongoing.

• RCS Tavg is 510°F and lowering at a rate of 50°F/Hr.

• RCS pressure is 1350 psig and stable.

• Pressurizer (PRZR) level indicates 38% and rising.

Which of the following describes the NEXT MAJOR action to be implemented in the EOP to mitigate the current conditions?

A. Depressurize the RCS using normal spray to minimize RCS subcooling.

B. Stop the cooldown. Energize all PRZR heaters to collapse voids and stabilize PRZR level.

C. Transition to ES-1.1, "SI Termination" and begin the SI flow reduction sequence by stopping ECCS pumps.

D. Stop RCP's NOT needed for PRZR Spray and begin the SI flow reduction sequence by stopping ECCS pumps.

Answer: D Explanation/Justification: KIA is met by demonstrating the knowledge of the major action steps of ES-1.2, Post LOCA Cooldown and Depressurization to mitigate the event.

A. Incorrect. This is the fifth major action step (EOP step 23) performed after normal charging has been re-established. Plausible because the plant has just been depressurized to raise przr level to >31 % (EOP step 15) by understanding the stem information.

B. Incorrect. ES-0.1 cools the plant down to mode 5 condition so there is no need to stop cooldown unless 1OOFlhr was exceeded. No voids exist at the current time with RCPs running. This would be performed if a void existed in ES-0.2 or ES-0.3.

C. Incorrect. ES-1.1 is a plausible distract since it terminates SI. In the case of ES-1.2, the steps to terminate Si are incorporated in EOP steps 17-21.Candidate must know that reducing SI is a major action of the procedure.

D. Correct. The depressurization to raise przr level to >31% is complete (major action step 2), therefore the next major action step is to stop all but one RCP and reduce RCS injection flow (steps 3 & 4)

Sys# System Category KA Statement NIA NIA Generic Knowledge of EOP mitigation strategies.

KIA# 2.4.6 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.2 lss. 2, Rev. 1, steps 16 & 17 Question Source: Bank - 2LOT8 Audit Exam (058)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

74. Given the following conditions:

• A small fire was discovered in the Unit 2 Control Room

• AOP-2.33.1A, "CONTROL ROOM INACCESSIBILITY" has been implemented

• All Control Room actions are complete

• All equipment operated as expected In accordance with AOP-2.33.1A, what is the Unit 2 Balance of Plant (BOP) role during this event?

A. Emergency Squad B. Communicator I N0#3 C. Alternate Shutdown Panel (ASP)

D. Emergency Shutdown Panel (SOP)

Answer: B Explanation/Justification: KIA is met by demonstrating the knowledge of the licensed operator rules during a fire in the control room in accordance with the Control Room Inaccessibility AOP.

A. Incorrect. Emergency Squad is the required role of the Turbine & PAB Operators.

B. Correct. BOP Operator is required to be the Communicator/Nuclear Operator #3 in accordance with Attachment 6 of AOP-2.33.1A.

C. Incorrect. Alternate Shutdown Panel is not manned during this event. It could be applicable if 20M-56C was implemented.

D. Incorrect. Emergency Shutdown Panel is manned by the Unit Supervisor and Reactor Operator.

Sys# System Category KA Statement N/A N/A Generic Knowledge of fire protection procedures.

KIA# 2.4.25 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.33.1A Rev. 15 pg. 35 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

75. The Plant was operating at 100% power:

• At time 1205 a Reactor Trip and several events occurred

• At time 1215 an ALERT is declared by the Shift Manager

• At time 1225 the Initial Notification Form is completed and approved by the Shift Manager Which of the following identifies the LATEST time that the initial notification to State and County officials is due?

A. 1220 B. 1230 C. 1235 D. 1240 Answer: B Explanation/Justification: KIA is met with Licensed Operator knowledge of the CR Communicator responsibilities and the required times to complete the initial notification to state and county officials.

A. Incorrect. This is the time at which the declaration must be made by the Shift Manager (SM).

B. Correct. Per 1/2-EPP-IP-1.1, Initial Notifications are to be made to the first six (6) listed Agencies of the Emergency Notification Call List (State and County), and MUST be made within 15 minutes of the event declaration.

C. Incorrect. The SM has 15 minutes to declare the event and then 15 minutes from declaration to notify the state and counties. This theoretically gives them 30 minutes to make a notification. However, since the declaration was made at 1215 the notification must be made by 1230. This distractor is based on 30 minutes from 1205.

D. Incorrect. This distractor is based on 15 minutes incorrectly added to the time the INF form was completed and approved. The notification must be made within 15 minutes of the event declaration.

Sys# System Category KA Statement NIA N/A Generic Knowledge of RO responsibilities in emergency plan implementation.

KIA# 2.4.39 KIA Importance 3.9 Exam Level RO References provided to Candidate none Technical

References:

1/2-EPP-IP-1.1 Rev. 51 1/2-EPP-IP-1.1.F02 Rev.19 Question Source: Bank - Robinson NRC 2011 (074)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 45.11)

Objective:

ES-401 Form ES-401-8 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific SRO Written Examination BV2LOT15 SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Beaver Valley Unit 2 Region: I l:R1 11 0 111 0 IV 0 Reactor Type: W l:R1 CE 0 BW 0 GE 0 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> If you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results RO I SRO-Only I Total Examination Values 75 /__ 25 l 100 Points Applicant's Scores I I Points Applicant's Grade I I Percent NUREG-1021, Revision 10 FENOC Facsimile r2

(SRO ONLY}

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

76. The crew is performing ES-0.1, 'Reactor Trip Response' after an inadvertent Reactor trip from 100% power.

10 minutes after the Reactor Trip:

• A 200 gpm small break LOCA occurs

• The ATC operator notes PRZR level is at 20% and lowering

• 2CHS*FCV122 'Charging Pumps Disch Flow Control Viv' is in MANUAL and Full OPEN

• 2CHS*FI 122 indicates 150 gpm and steady

• Net Charging on PCS indicates 60 gpm

• Assume RCS Pressure remains constant during the event

• No automatic ESF actuation conditions are met

• All systems operate as designed

1) With NO Operator action, approximately, how long before the PRZR level indicates 0%?

2) The Unit Supervisor will transition from ES-0.1 to which of the following procedures?

A. 1) 15 minutes

2) E-1, "Loss of Reactor or Secondary Coolant" B. 1) 15 minutes

2) E-0, "Reactor Trip or Safety Injection" C. 1) 45 minutes

2) E-1, "Loss of Reactor or Secondary Coolant" D. 1) 45 minutes

2) E-0, "Reactor Trip or Safety Injection"

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 76 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant status and determine which procedure to transition too based upon the LHP criteria of the procedure. Detailed knowledge of the procedure is required to select the correct procedure actions. The SRO is required to review the Left Hand Page information periodically during procedure implementation and direct the crew to implement actions when conditions are met.

KIA is met by determining how the PRZR level control system will operate during a SBLOCA, and determine how long it will take for the PRZR to indicate empty. SRO level of knowledge of LHP to initiate SI at 4% and transition to E-0.

A. Incorrect. This would be the time if letdown didn't isolate at 14%. Incorrect procedure transition. Plausible procedure choice if candidate is thinking of the LHP requirements of ES-1.1 SI Termination, which states to manually start SI and transition to E-1, Loss of Reactor or Secondary Coolant at 17% PRZR level.

B. Incorrect. This would be the time if letdown didn't isolate at 14%. Correct procedure transition per ES-0.1 LHP.

C. Incorrect. This is the correct time. Incorrect procedure transition. Plausible procedure choice if candidate is thinking of the LHP requirements of ES-1.1 SI Termination, which states to manually start SI and transition to E-1, Loss of Reactor or Secondary Coolant at 17% PRZR level.

D. Correct. Candidate must evaluate the initial net charging and the leak rate to determine the RCS is losing 140gpm. After letdown isolates at 14%, net charging will rise to 165 gpm, with RCS losing 35gpm. 20-14% (600gal)@ 140gpm=4.3 min until UD isolates, then 14-0% (1400 gal)@

35 gpm = -40 min. In ES-0.1 LHP states to actuate SI and go to E-0 if PRZR level cannot be maintained >4%.

Sys # System Category KA Statement 000009 Small Break EA2 Ability to determine or interpret the following as they apply to a The time available for action before PZR is empty, LOCA I 3 small break LOCA: given the rate of decrease of PZR level KIA# EA2.05 KIA Importance 3.9 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.ES-0.1 lss. 2 Rev. 3 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

77. The plant was operating at 100% power when a large break LOCA occurred coincident with a Loss of 4KV Bus DF.

The follow conditions exist:

• 'A' Quench Spray Pump [2QSS-P21A] tripped on startup

• 3 Max CETs indicate 810°F

• RCS is superheated

• CNMT Pressure is 31 psig

• CNMT Temperature is 240°F

• All RCPs have been tripped

• RVLIS Full Range indicates 35%

Based on the above conditions, which answer below completes the following statement?

The required EAL classification is based upon the _ _ _ _ _ __

A. LOSS of one fission product barrier and POTENTIAL Loss of another barrier B. LOSS of two fission product barriers C. LOSS of two fission product barriers and a POTENTIAL loss of a third barrier D. LOSS of all three fission product barriers

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 77 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Classifications. This is a SRO position function only.

KIA is met by demonstrating the knowledge to determine the event classification based on the conditions given using the provided EPP classification chart.

Candidate will have to recognize a General Emergency would be declared based on the following conditions which they will have to interpret from the conditions given in the stem.

FC - Loss due to FR-C.1 Red Path Entry RCS - Loss RCS leak rate greater than available makeup capacity as indicated by RCS subcooling < 46° F adverse containment.

CT - Potential Loss due to Cnmt pressure >11 psig AND less than one full train of depressurization equipment operating.

A. Incorrect. Plausible if it is not recognized that FR-C.1 entry conditions have been met for Fuel Clad failure, or a Loss based on RCS Leak Rate.

B. Incorrect. Plausible if it is not recognized that Potential Loss due to Cnmt pressure >11 psig AND less than one full train of depressurization equipment operating C. Correct. GE based on answer explanation above.

D. Incorrect. Plausible if it is not recognized that containment barrier is a potential loss, and not a loss. This would identify a weakness of CT-8 based on pressure response not consistent with LOCA conditions.

Sys # System Category KA Statement 000011 Large Break Generic Knowledge of the emergency action level LOCA I 3 thresholds and classifications.

KIA# 2.4.41 KIA Importance 4.6 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.F-0.2 lss. 2 Rev. 1 EPP Chart EPP-1-1b.F01 Rev. 0 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.11)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

78. Given the following conditions:

• A reactor trip and safety injection have occurred

• The crew is performing the actions of E-2, "Faulted Steam Generator Isolation" due to the uncontrolled depressurization of 'A' SG.

• The crew is evaluating if SI flow should be reduced.

• The following conditions exist:

o RCS temperature is 460°F o RCS pressure is 1650 psig and slowly rising o Containment pressure is 23 psig o SG 218 and 21C NR levels are 15% and rising o AFW flow is 375 gpm o PRZR level is 20%

Based on the conditions above, when may the crew enter ES-1.1, "SI Termination"?

A. Immediately.

B. After transition to E-1, when RCS subcooling criteria is met.

C. After transition to E-1, when PRZR level criteria is met.

D. After transition to E-1, when Secondary heat sink criteria is met.

Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant status and determine if the conditions are met to terminate Safety Injection and the required procedure transitions within the EOP network. Detailed knowledge of the procedure is required to select the correct transition and the requirements for SI termination.

KIA is met by interpreting the given conditions to determine when SI Termination is permitted. Detailed knowledge of SI Terminations and procedural transitions is required for the SRO.

A. Incorrect. PRZR level criteria is not high enough for the adverse CNMT conditions (38% req.). Plausible distractor because transition to ES-1.1 from E-2 occurs immediately after checking PRZR level. Evaluation of adverse CNMT must be determined.

B. Incorrect. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since it is a faulted SG, subcooling requirements were easily met, but PRZR level is not.

C. Correct. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since PRZR level is still low for adverse CNMT (38% req.) a transition to ES-1.1 must wait until PRZR level is met in E-1.

D. Incorrect. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since SG level does not meet the adverse requirement of 31 % this is a plausible distractor.

Heat sink is met with AFW flow >340 gpm, but PRZR level is not.

Sys # System Category KA Statement 000040 Steam Line AA2 Ability to determine and interpret the following as they apply to When ESFAS systems may be secured Rupture I 4 the Steam Line Rupture:

KIA# AA2.05 KIA Importance 4.5 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.E-2 lss. 2 Rev 0 None 20M-53A.1.E-1lss.2 Rev. 2 Question Source: Bank - 2LOT5 NRC Exam (047)(2005)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

79. The plant is in Mode 3 with all systems in normal alignment for this Mode.

• Battery Breaker 2-1 [BAT*BKR2-1] is on Clearance for Electrical Maintenance to replace the Battery Breaker for maintenance.

• Annunciator A8-9A, "125V DC Bus 2-1 TROUBLE" re-flashes due to Battery Charger 2-1 AC Input Breaker tripping open.

Which of the following Tech Spec LCOs will be applicable?

1) 3.8.1, AC Sources - Operating

2) 3.8.7, Inverters - Operating

3) 3.8.9, Distribution Systems - Operating A. None B. 1 &2 ONLY C. 2 & 3 ONLY D. 1, 2, 3 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine which TS are applicable. Detailed knowledge of the bases is required to determine the impact of the loss of the power supplies and which TS are applicable.

KIA is met by demonstrating Tech Spec bases knowledge for the effected equipment when a loss of DC bus occurs. DC bus loss will effect Inverters, and the EOG start capabilities. This is TS bases knowledge.

A. Incorrect. Plausible distractor because TS 3.8.4, DC Sources Operating was intentionally omitted from the above list. Candidate must know the TS bases for all the listed TSs to correctly answer the question.

8. Incorrect. Plausible distractor if the bases for TS. 3.8.9 is not known. The bases states that DC subsystems require the associated buses and distribution panels to be energized to their correct voltage from either the associated battery or charger.

C. Incorrect. Plausible distractor if the bases for TS 3.8.1 is not known. The bases states that each DG must be capable of starting and loading.

With DC bus 1 de-energized, the diesel starting circuits and load sequencer are not capable of performing their function.

D. Correct. All of the TSs are applicable. The bases for TS 3.8.1 and 3.8.9 are described above. TS 3.8.7 bases states that an inverter can be supplied from an internal AC source via a rectifier as long as the battery is available. However, in the stem it stated that the 2-1 battery breaker was on clearance for maintenance.

Sys# System Category KA Statement 000058 Loss of DC Generic Knowledge of the bases in Technical Power/6 Specifications for limiting conditions for operations and safety limits.

KIA# 2.2.25 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

U2 RE-0001AR Rev. 22 TS bases 3.8.1, 3.8.7, 3.8.9 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 41.7 I 43.2)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

80. The plant was operating at 100% power.

• A LOCA OUTSIDE containment occurs

• At step 21 of E-0, Reactor Trip Or Safety Injection, the crew enters ECA-1.2, LOCA Outside Containment

• At the completion of ECA-1.2, the crew has been UNABLE to locate and isolate the break The following plant conditions exist:

• Offsite Power has been lost

• All SG pressures are 800 psig and stable

• All SG NR levels are 35% and slowly rising

• All Secondary radiation monitors are consistent with pre-event values

• CNMT parameters are consistent with pre-event

• RCS Subcooling is 40°F and slowly dropping

• RCS Pressure is 1125 psig and slowly dropping

• PRZR level is 12% and slowly dropping

• Auxiliary Building Radiation levels are rising

• Auxiliary Building sump levels are rising

• A seismic event of 0.07g has been recorded Based on these conditions and events:

1) What procedural transition from ECA-1.2 is REQUIRED?

2) Of the choices listed below, which of the Abnormal Operating Procedures will be performed in conjunction with the EOP network?

A. 1) ECA-1.1, Loss Of Emergency Coolant Recirculation

2) AOP-2.36.1, Loss Of All AC Power When Shutdown B. 1) ECA-1.1, Loss Of Emergency Coolant Recirculation

2) AOP-1/2.75.3, Acts of Nature - Seismic Event C. 1) E-1, Loss Of Reactor Or Secondary Coolant

2) AOP-2.36.1, Loss Of All AC Power When Shutdown D. 1) E-1, Loss Of Reactor Or Secondary Coolant

2) AOP-1/2.75.3, Acts of Nature - Seismic Event

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 80 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the ineffective break isolation steps. This evaluation requires detailed knowledge of the EOP procedure flow-paths of sub procedures. Detailed knowledge of the procedure is required to select the correct transition.

K/A is met by demonstrating knowledge of entry conditions into abnormal operating procedures based on given indications while responding to a LOCA Outside Containment. In the question the SRO is given a seismic indication greater than the Alarm Response Procedure (ARP) entry setpoint, and must determine that the ARP is an entry condition into the Seismic Event Abnormal Operating Procedure. This AOP is performed in conjunction with the EOP network.

A. Incorrect. Correct EOP transition. Incorrect AOP. AOP-2.36.1, Loss Of All AC Power When Shutdown is a plausible distractor with the crew performing ECA-1.2, LOCA Outside CNMT, then losing offsite power. The candidate must determine that entry conditions of Rx vessel defueled or RHR being used to control RCS temperature, are not met for entry into this AOP.

B. Correct. If RCS pressure is not rising, then JAW ECA-1.2 step 4 RNO transition must be made to ECA 1.1. Seismic Event AOP is correct due to the event registered 0.07g which is greater than ARP setpoints (A 10-5H). Entry into the Seismic Event AOP is required based upon annunciator A 10-5H, a report from NEIC, or felt or observed ground movement by plant personnel.

C. Incorrect. Plausible since E-1 would be the appropriate entry if RCS pressure were rising. Incorrect AOP as explained in answer 'A'.

D. Incorrect. Plausible since E-1 would be the appropriate entry if RCS pressure were rising. Correct AOP entry.

Sys # System Category KA Statement W/E04 LOCA Outside Containment I 3 Generic Knowledge of abnormal condition procedures.

KIA# 2.4.11 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ECA-1.2 lss. 2 Rev. 0, pg. 3 20M-45B.4.AAA Rev. 8, pg. 3 1/20M-53C.4A.75.3 Rev. 19, pg. 1 Question Source: Bank- 2LOT6 NRC Exam (081) Modified (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

81. The plant is at 100% power.

• A loss of all Feedwater occurs

• The crew enters FR-H.1, Response to Loss of Secondary Heat Sink The following conditions now exist:

• FR-H.1 Step 6 Stop All RCPs has just been completed

• 'A' SG Wide Range Level is 10%, pressure is 1000 psig and stable

• 'B' SG Wide Range Level is 19%, pressure is 975 psig and stable

• 'C' SG Wide Range Level is 12%, pressure is 600 psig and lowering

• Containment Pressure is 4.0 psig and stable (1) Which of the following actions are REQUIRED based upon these indications?

(2) Per Tech. Specs., with the plant in Mode 3, what is the MINIMUM water level required to consider a Steam Generator OPERABLE as a heat sink?

A. (1) Transition to E-2, Faulted Steam Generator Isolation (2) 12% Narrow Range B. (1) Initiate RCS Bleed and Feed (2) 12% Narrow Range C. (1) Transition to E-2, Faulted Steam Generator Isolation (2) 15.5% Narrow Range D. (1) Initiate RCS Bleed and Feed (2) 15.5% Narrow Range Answer: D Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. 11.E pg 6 and SRO level knowledge of TS bases for the Surveillance requirements. The first part requires an understanding of the EOP mitigation strategy which is RO level knowledge, however the SRO must assess plant conditions, determines if adverse criteria is applicable, and selects the section of the procedure to mitigate the event. The TS minimum level is SRO level knowledge since the level required for operability is NOT addressed in the LCO rather is addressed in the bases and the surveillance requirement.

KIA is met by interpreting the given conditions of a loss of secondary heat sink, then determining the appropriate procedural actions based on the conditions.

A. Incorrect. This transition is possible since the 'C' SG pressure is lowering, however Bleed and Feed criteria are met. EOP Rules of usage does not allow for exit until FR-H.1 is complete. This SG NR level is the minimum in the EOP network not TS.

B. Incorrect. Bleed and Feed criteria are met. This SG NR level is the minimum in the EOPs not TS C. Incorrect. This transition is possible since the 'C' SG pressure is lowering, however Bleed and Feed criteria are met. EOP Rules of usage does not allow for exit until FR-H.1 is complete. Correct TS SG level.

D. Correct. Bleed and Feed criteria are met per continuous action step 3, which states WR level in at least 2 SGs <14% go to the RNO for Feed and Bleed actions. Correct TS bases setpoint per SR 3.4.5.2 bases.

Sys # System Category KA Statement W/E05 Loss of EA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Secondary the (Loss of Secondary Heat Sink) procedures during abnormal and emergency Heat Sink I 4 operations.

KIA# EA2.1 KIA Importance 4.4 Exam Level SRO References provided to Candidate Technical

References:

FR-H.1 pg. 2 Rev .. 1 lss. 2 None TS Bases 3.4.5.2 pg b 3.4.5-5 Question Source: Bank - 1LOT14 NRC Exam (080)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

82. The plant is at 100% power.

• l&C is requesting Operations place control rods on DC Hold in accordance with 20M-1.4. 0, "Placing a Control Rod Power Cabinet Group on DC Hold", to perform maintenance in a rod control power cabinet.

• The crew has been briefed on compensatory measures in the event of a load rejection.

1) What is the MAXIMUM number of control rod group(s) capable of being placed on DC Hold?

2) If a load rejection were to occur, what is the Tech Spec OPERABILITY status of the rods that are on DC Hold?

A. 1) 1 Group

2) Operable B. 1) 2 Groups

2) Operable C. 1) 1 Group

2) Inoperable D. 1) 2 Groups

2) Inoperable Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the Operability of the Control Rods while they are on DC hold. The OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. Detailed knowledge of the bases is required to determine the impact of the loss of the power supplies and Operability of the Control Rods.

KIA is met by analyzing the effect of the rods being placed on DC hold for maintenance will have on the operability of the rods. Rods on DC hold are still operable (trippable) per the Tech Spec Bases.

A. Correct. The maximum number of rods is 4 (1 group at BV). Tech Spec bases defines a rod as operable if it is trippable. The DC hold cabinet is in parallel with the rod control power cabinets, both being powered through the reactor trip bkrs. When the Rx trip bkrs open, the rods will insert.

B. Incorrect. Plausible is the candidate thinks DC Hold can maintain a bank of rods (2 groups). It is correct that they are operable.

C. Incorrect. The maximum number of rods is 4 (1 group). Inoperable is not correct because when the DC Hold cabinet loses power (ie. Rx trip) the rods will insert.

D. Incorrect. Plausible is the candidate thinks DC Hold can maintain a bank of rods (2 groups). Inoperable is not correct because when the DC Hold cabinet loses power (ie. Rx trip) the rods will insert.

Sys # System Category KA Statement 000003 Dropped Generic Ability to analyze the effect of maintenance Control Rod I activities, such as degraded power sources, on 1 the status of limiting conditions for operations.

KIA# 2.2.36 KIA Importance 4.2 Exam Level SRO References provided to Candidate Technical

References:

20M-1.4.0 Rev. 0 lss. 1 pg. 1 None TS Bases pg. B 3.1.4-5 rev. 0 3SQS-1.3 Rev 7 lss. 1 pg. 13 Question Source: New Question Cognitive Level: Lower_ Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.2 I 45.13)

Objective: 3SQS-1.3 Obj 9 Explain the function, operation, location and limitations of the DC Hold Cabinet.

3SQS-1.3 Obj. 28 Using a copy of Technical Specifications or the Licensing Requirements Manual, assess a given set of plant conditions for compliance with the licensing requirements, including the determination of equipment operability and applicable action statements.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

83. The plant was at 100% power when the following occurs:

• The reactor failed to trip after receiving a valid trip signal

• SRO transitioned from E-0, Reactor Trip or Safety Injection, to FR-S.1, Response to Nuclear Power Generation/A1WS Current conditions:

• Emergency Boration was initiated

• Safety injection did not actuate

• Reactor power is 3% and decreasing

• Intermediate range channels indicate negative SUR

• Operators are verifying the reactor subcritical at step 7 of FR-S.1 Based on the current plant conditions:

(1) Boration _ _ _ _ _ _ _ required to continue after verifying the reactor is subcritical.

(2) Which of the following describes the required procedural flowpath?

A. 1) is

2) Return to E-0.

B. 1) is

2) Remain in FR-S.1 C. 1) is not

2) Return to E-0.

D. 1) is not

2) Remain in FR-S.1

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 83 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the existing power level and SUR. This evaluation requires detailed knowledge of the EOP procedure content and transition criteria. Additionally, the SRO must decide what action is required related to continuing the boration flow. Knowledge of the procedure steps is required to make the decision and select the correct transition.

KIA is met with the EOP background knowledge that emergency boration is required to continue to ensure adequate shutdown margin during future cooldown. The candidate must also determine if conditions are satisfied to transition back to E-0, or stay in FR-S.1.

A. Correct: In FR-S.1, after verifying the Rx is subcritical in step 7, step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions." Per the background this is to ensure adequate S/D margin during the future plant cooldown. When power<

5% and negative IR SUR is achieved in FR-S.1, step 7d directs returning to the procedure and step in effect which is E-0.

B. Incorrect: Boration is required to continue to obtain adequate shutdown margin during subsequent actions. It is not required to remain in FR-S.1 once it has been verified that the reactor is subcritical. Step 7d directs returning to the procedure and step in effect which is E-0.

C. Incorrect: step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions. Returning to E-0 is correct since the reactor is subcritical.

D. Incorrect: step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions. It is not required to remain in FR-S.1 once it has been verified that the reactor is subcritical. Step 7d directs returning to the procedure and step in effect which is E-0.

Sys# System Category KA Statement 000024 Emergency Generic Ability to perform specific system and integrated Boration / 1 plant procedures during all modes of plant operation.

KIA# 2.1.23 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.FR-S.1 lss. 2 Rev. O 20M-53B.4.FR-S.1 lss. 2 Rev. 0 Question Source: Bank- Surry 2010 NRC Exam (082) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 / 45.2 I 45.6)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

84. Initial conditions:

• Core Cooling CSFST is ORANGE

• FR-C.2, Response to Degraded Core Cooling is in progress Current conditions:

• Steam Generator depressurization to 100 psig is in progress

• A validated Orange Path on the CSFSTs points to FR-P.1, Response to Imminent Pressurized Thermal Shock Condition

1) What is the purpose of depressurizing all intact SGs to 100 psig in FR-C.2?

2) How must the Unit Supervisor respond to the Orange path on FR-P.1?

A. 1) To assist in core recovery by injecting the Safety Injection Accumulators.

2) Remain in FR-C.2 until completion, then transition to FR-P.1.

B. 1) To assist in core recovery by injecting the Safety Injection Accumulators.

2) Immediately transition to FR-P.1.

C. 1) To assist in core recovery by injecting using the Low Head Safety Injection Pumps.

2) Remain in FR-C.2 until completion, then transition to FR-P.1.

D. 1) To assist in core recovery by injecting using the Low Head Safety Injection Pumps.

2) Immediately transition to FR-P.1.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 84 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the rules of use and hierarchy for the Function Restoration Procedures. This evaluation requires detailed knowledge of the EOP procedure flow-paths. Additional knowledge of the procedure step bases is required, beyond the high level action steps for FR-C.2.

KIA is met with the knowledge of the bases for depressurizing the SGs in Orange path FR-C.2, Response to Degraded Core Cooling, and accessing the transition to an Orange path FR-P.1 cause by SI accumulators injecting.

A. Correct. Depressurization of the SGs to 100 psig is required to lower RCS pressure low enough to inject SI Accumulators and cover the core. It is correct to remain in FR-C.2 if a Orange path in FR-P.1 is created when the SI accumulators inject. It is an expected condition stated by a CAUTION prior to SG depressurization step.

B. Incorrect. It is correct that depressurization of the SGs to 100 psig is required to lower RCS pressure low enough to inject SI Accumulators and cover the core. It would be incorrect to immediately transition to FR-P .1 due to the note prior to the depressurization step. This is a plausible distractor if candidate has a misconception of the hierarchy for the Function Restoration Procedures.

C. Incorrect. Plausible because after the accumulators are isolated at 100 psig SG pressure, continued SG depressurization to atmospheric pressure allows the RCS pressure to be low enough for LHSI to inject into the core (step 17). It is correct to complete FR-C.2 prior to going to FR-P.1.

D. Incorrect. Plausible because after the accumulators are isolated at 100 psig SG pressure, continued SG depressurization to atmospheric pressure allows the RCS pressure to be low enough for LHSI to inject into the core (step 17). It would be incorrect to immediately transition to FR-P.1 due to the note prior to the depressurization step. This is a plausible distractor if candidate has a misconception of the hierarchy for the Function Restoration Procedures.

Sys# System Category KA Statement W/E06 Degraded EA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Core Cooling the (Degraded Core Cooling) procedures during abnormal and emergency 14 operations.

KIA# EA2.1 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.FR-C.2 lss. 2 Rev. 2 20M-53B.4.FR-C.2 lss. 2 Rev. 2 Question Source: Bank - Vogtle 2012 NRC exam (099) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

85. The plant is at 100% power.

• A reactor trip occurs coincident with a loss of all offsite power

• The operators have verified natural circulation flow and are cooling down the plant per ES-0.2, Natural Circulation Cooldown

• Train A RVLIS is OOS for Maintenance The following plant conditions now exist:

• RCS Pressure is 1940 psig and stable

• RCS Hot Leg temperatures are 540 °F and lowering

• RCS Cooldown rate based upon Cold Leg temperatures is currently 30 °F/Hr and CANNOT be reduced

• PSMS Data Processing Unit B indication is "NO OFFLINE" Which of the following procedures will be entered and what is the MAXIMUM allowable RCS Cooldown ratV i.ri ft..e pr-oc~c.<,Yf!- to b~ <!J.1t'(_ve.J. ~ ~J~

A. ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS); 50 °F/Hr B. ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS); 100 °F/Hr C. ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS); 50 °F/Hr D. ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS); 100 °F/Hr Answer: C Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. 11.E pg 7 which requires the knowledge of diagnostics steps and decision points in EOPs that involve transitions to event specific sub-procedures The SRO must be aware of sub-procedures for Natural Circulation Cooldown, if the CID rate cannot be maintained less than 25 °F/Hr. Detailed procedure knowledge is required for CID rate.

KIA is met by interpreting the conditions given in the question, then based on this knowledge, transition to the appropriate procedure due to Cooldown rate limitations and RVLIS availability.

A. Incorrect. Correct procedure. Cooldown rate is incorrect. ES-03 allows a cooldown rate of <100F/hr.

B. Incorrect Procedure With pressure <1950psig and Thot <550F, conditions are met to maintain 25F/hr cooldown rate. If <25F/hr cannot be maintained, the RNO step transitions the crew to ES-03 with RVLIS. In the stem of the question both trains of RVLIS are OOS, therefore RVLIS is not available.

C. Correct. Procedure is correct. With pressure <1950psig and Thot <550F, conditions are met to maintain 50 F/hr cooldown rate. In the stem of the question both trains of RVLIS are OOS, therefore RVLIS is not available. The cooldown rate in ES-04 is 50F/hr until temperature is less than 450F, then the rate is raised to 100F/hr.

D. Incorrect. Procedure is incorrect. Plausible distractor with one train of RVLIS OOS. The SRO must know that one train on RVLIS is still available, and a transition to ES-04 would not be correct. 1OOF/hr is the correct cooldown rate for ES-04 when Thot is between 500-450F.

Sys # System Category KA Statement W/E09 Natural EA2 Ability to determine and interpret the following as they apply Adherence to appropriate procedures and Circulation to the (Natural Circulation Operations) operation within the limitations in the facility*s Operations I 4 license and amendments.

KIA# EA2.2 KIA Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ES-0.2 lss. 2 Rev. 1 pg.16 20M-53A.1.ES-0.4 lss .. 2 Rev. 1 pg.3 20M-5.D.1.D lss. 4 Rev 0 pg. 15-16 Question Source: Bank - 1LOT14 NRC Exam (Q85) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

86. Given the following conditions:

• The plant is operating at 20% power

• AOP 2.6.8, 'Abnormal RCP Operation' has been entered due to rising temperatures on the 'B' RCP The following conditions exist:

RCS*P21B MTR LWR RCS*P21 B MTR UPR Time RADIAL [T0435Al THRUST [T0434Al 1000 181°F 184°F 1005 189°F 188°F 1010 197°F 194°F 1015 204°F 201°F

1) Which Motor Bearing reaches the RCP trip setpoint FIRST in accordance with AOP-2.6.8?

2) What actions will be directed by the Unit Supervisor?

A. 1) Motor Lower RADIAL Bearing

2) Shutdown 'B' RCP, go to AOP-2.51.1, Unplanned Power Reduction, and perform a controlled plant shutdown.

B. 1) Motor Lower RADIAL Bearing

2) Trip the reactor, go to E-0, complete the IOAs, then shutdown 'B' RCP.

C. 1) Motor Upper THRUST Bearing

2) Shutdown 'B' RCP, go to AOP-2.51.1, Unplanned Power Reduction, and perform a controlled plant shutdown.

D. 1) Motor Upper THRUST Bearing

2) Trip the reactor, go to E-0, complete the IOAs, then shutdown 'B' RCP.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 86 Answer: 8 Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures and transitions between Abnormal and EOPs. The SRO must evaluate the plant conditions and determine which setpoint has been exceeded for continued RCP operation. Then the SRO must determine the specific sequence of actions to take when securing the RCP, the sequence of actions are listed as sub-steps in the Abnormal Operating Procedure. Additionally directing the action to secure the pump is to occur following completion of the IOAs, which is SRO knowledge of the AOP procedure content. Per the EOP users guide, the Continuous Actions are on the fold out page, the Reader (US) is responsible for reviewing and monitoring the CA page and informing the crew when conditions are met to apply the action.

KIA is met by demonstrating the ability to predict the impact of a rising RCP bearing temperature, then based on reaching a required RCP immediate shutdown setpoint, chose the appropriate procedure to shutdown the Rx and the RCP. This is an abnormal RCP shutdown sequence in that the Rx is tripped, then the RCP is tripped. Normally RCP shutdowns occur prior to the Rx being critical during plant heat up, or after plant cooldown.

A. Incorrect. Correct bearing. Incorrect RCP shutdown sequence and procedure for shutting down the plant. Plausible distractor because tripping of an RCP when power is <30% (P-8) does not generate a Rx trip, and a controlled shutdown would be plausible, but not permitted.

B. Correct. IAW the AOP, motor bearing temperature setpoint for trip criteria is >195F which is met at 1010 by the MTR LWR RADIAL BEARING at 197F. AOP-2.6.8 Continuous action step 1 directs tripping the Rx, E-0, IOAs, then tripping RCP.

c. Incorrect. Incorrect bearing. Incorrect RCP shutdown sequence and procedure for shutting down the plant. Plausible distractor because tripping of an RCP when power is <30% (P-8) does not generate a Rx trip, and a controlled shutdown would be plausible, but not permitted.

D. Incorrect. Incorrect bearing. Correct Rx trip, IOAs, and RCP shutdown sequence.

Sys# System Category KA Statement 003 Reactor A2 Ability to (a) predict the impacts of the following malfunctions or operations Conditions which exist for an abnormal Coolant Pump on the RCPS; and (b) based on those predictions, use procedures to correct, shutdown of an RCP in comparison to a System (RCPS) control, or mitigate the consequences of those malfunctions or operations: normal shutdown of an RCP KIA# A2.02 KIA Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

20M-53C.4.2.6.8 Rev. 12 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5/ 45.3 / 45/13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

87. Given the following initial conditions:

• Rx power is 100% and stable

• Control rods are in MANUAL for l&C maintenance

• Chemistry requested Cation bed demineralizer [2CHS-DEMIN22] be placed in service to lower RCS pH

• Cation bed demineralizer [2CHS-DEMIN22] was placed in service in accordance with 20M-7.4.C2, "Lowering RCS PH" One hour after the Cation bed demineralizer was placed in service, the Reactor Operator reports Reactor power is 100.1 % and slowly rising.

(Assuming the demineralizer was the cause)

Which of the following is the reason for the power rise, AND the appropriate procedure for the Unit Supervisor to implement?

A Cation Demineralizer was placed in service with a _ ___._(1~)___ boron concentration than the RCS.

The Unit Supervisor will implement (2)

A. 1) LOWER

2) AOP-2.51.2, "Reactor Overpower" B. 1) LOWER

2) 20M-52.4.B.1, "Turbine Load Changes" C. 1) HIGHER

2) AOP-2.51.2, "Reactor Overpower" D. 1) HIGHER

2) 20M-52.4.B.1, "Turbine Load Changes"

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 87 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures related to coordination with normal procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure to implement. The slow power rise warrants the use of the normal Turbine Load Change procedure versus the Abnormal Reactor Overpower procedure.

Detailed knowledge of the content is required to select the correct procedure. The first part of the question is a fundamental knowledge of the effect of a dilution event due to the cation demineralizer operation.

KIA is met by predicting the effect of placing a cation demineralizer in service with a lower boron concentration than the RCS (dilution event), and determine the correct procedure to use to mitigate the power change.

A. Incorrect. Correct that the demineralizer had a lower boron concentration than the RCS. Incorrect to use Reactor Overpower because power was not rapidly rising.

B. Correct. The given conditions indicate there is an RCS dilution in progress. If a Cation Demineralizer were placed in service with a boron concentration lower than the RCS, it would remove boron from the RCS resulting in a dilution event. With it being a slow rise in power the SRO would use the Turbine Load Changes procedure to control power. A note in the Reactor Overpower states this procedure is intended for use when power is rapidly rising. Conditions given had power rise 0.4% over an hour.

C. Incorrect. If the demineralizer was higher than the RCS, power would decrease, not rise as the conditions given. Incorrect to use Reactor Overpower because power was not rapidly rising.

D. Incorrect. If the demineralizer was higher than the RCS, power would decrease, not rise as the conditions given. Correct procedure to use for slowly rising power.

Sys# System Category KA Statement 004 Chemical and A2 Ability to (a) predict the impacts of the following malfunctions or Fact that isolating cation demineralizer stops Volume operations on the eves; and (b) based on those predictions, use boron dilution and enables restoration of normal Control procedures to correct, control, or mitigate the consequences of those boron concentration System malfunctions or operations:

KIA# A2.33 KIA Importance 3.3 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.51.2 Rev. 2 pg 1 None 20M-52.4.B.1 Rev. 1 pg. 3 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/ 43/5 / 45/3 I 4515)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

88. The plant has tripped from 100% power, due to an inadvertent Safety Injection.

When resetting both trains of the Safety Injection Signal in ES-1.1, SI Termination, step 1, Annunciators A 12-1 C, Auto Safety Injection Blocked and A 12-1 D Safety Injection Signal are intermittently flashing.

What action will the crew take in response to these conditions?

The operators will _ _ _ _ ___._1'-"------

If, after taking this action, SI initiation parameter setpoints are exceeded, the affected Safety Injection equipment (2) operate automatically.

A. (1) perform Attachment A-1.25, ESF Signal Reset by Alternate Method (2) will B. ( 1) close, then open the reactor trip breakers per ES-1.1, SI Termination (2) will C. (1) perform Attachment A-1.25, ESF Signal Reset by Alternate Method (2) will NOT D. (1) close, then open the reactor trip breakers per ES-1.1, SI Termination (2) will NOT

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 88 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the procedure caution related to the ESFAS signals and action required to manually operate the ESF equipment. The second caution in Attachment A-1.25 alerts the SRO that ESF equipment will have to be to be manually started. The knowledge of the plant response to a failure of a single train of SI to reset is detailed EOP knowledge of the Attachment. A-1.25. The operator must have the knowledge of how the SI blocked annunciator responds if a single train does not reset. After resetting the SI signal the alarms will flash in and out as one train of SI is not reset KIA is met by demonstrating the knowledge of the Caution in A-1.25 regarding manual actuation of ESF equipment.

A. Incorrect, The operator action is correct. The SI equipment will not automatically actuate after the SSPS train is disabled.

B. Incorrect, This action is plausible, Step 31 of procedure ES-1.1 enables an Automatic Safety Injection by cycling the reactor trip breakers. The equipment will automatically actuate if both trains of Safety Injection were capable of being reset.

C. Correct, The operator action is correct and equipment must be manually operated. The Caution prior to step 1 of procedure A-1.25 alerts the operator that a SI signal will not Automatically operate after performing the procedure steps, manual action to operate equipment will be required.

D. Incorrect, This action is plausible, Step 31 of procedure ES-1.1 enables an Automatic Safety Injection by cycling the reactor trip breakers. The equipment would automatically actuate if both trains of Safety Injection were capable of being reset, however the equipment will not operate automatically if the correct actions were taken.

Sys# System Category KA Statement 013 Engineered Safety Features Generic Knowledge of the operational implications of EOP Actuation System (ESFAS) warnings, cautions, and notes.

KIA# 2.4.20 KIA Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.1 lss. 2 Rev. 0 pg. 2 20M-53A.1.A-1.25 Rev 0 pg. 2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

89. The plant is operating at 50% power with all systems in normal alignment for this power level.

The following conditions exist:

• A1-4H, Service Water System Trouble is the only LIT overhead annunciator

• [2SWS-PI 113A and Pl 1138], Service Water Header Pressures are 48 psig and stable

• [2SWS*MOV107A, B, C, D], 'Sec Comp Clg Wtr Hx Serv Water Supply Hdr lsol Vlvs' are OPEN Based on the above conditions, what could cause this Service Water condition, and which of the following procedures listed below would be the correct procedure to mitigate this condition?

The above conditions could indicate a _ ___._(1. . ). .___ , and would be mitigated by (2)

A. 1) Service Water System leak

2) AOP-2.30.1, Service Water/Main Intake Structure Loss B. 1) Service Water System leak

2) E-0, Reactor Trip or Safety Injection C. 1) Service Water Pump trip

2) AOP-2.30.1, Service Water/Main Intake Structure Loss D. 1) Service Water Pump trip

2) E-0, Reactor Trip or Safety Injection

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 89 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine that the SWS pressure is low due to a leak and not due to the pump trip. Since pressure remains high, reactor trip is not warranted, the response is to continue with the AOP to respond to the leak. Detailed knowledge of the content is required to select the correct procedural direction.

KIA is met by the ability to predict a Service Water leak malfunction based on lower than normal service water header pressure and other conditions given, and use the Service Water AOP to mitigate the lower Service Water header pressure.

A. Correct. This is an indication of a service water leak due to pressure being lower and equal in both headers (headers cross tied in NSA and normal pressure -70 psig). Pressure is not, and was not low enough to start a stby SW pump (34 psig) which would cause Ann. A1-5F, 'Stby SW Pump Auto start/Auto stop' to alarm. Also, A 1-4F, 'SW Pump Auto start/Auto stop' is not LIT. The correct procedural guidance is in AOP-2.30.1 because pressure is not below 34 psig or CCS is not isolated (107s are open).

B. Incorrect. This is an indication of a service water leak. With SW pressure >34 psig or CCS is not isolated (107s are open) entry conditions to E-0 do not exist per the AOP.

C. Incorrect. No indication of pump trip exists in the stem. Pressure is lower, but annunciators which indicate a pump trip or a stby pump start do not exist. The AOP is the correct procedure.

D. Incorrect. No indication of pump trip exists in the stem. Pressure is lower, but annunciators which indicate a pump trip or a stby pump start do not exist. With SW pressure >34 psig or CCS is not isolated (107s are open) entry conditions to E-0 do not exist per the AOP.

Sys # System Category KA Statement 076 Service A2 Ability to (a) predict the impacts of the following malfunctions or Service water header pressure Water operations on the SWS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (SWS) malfunctions or operations:

KIA# A2.02 KIA Importance 3.1 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.30.1 Rev. 9 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 1o CFR Part 55 Content: (CFR: 41.5 / 43.5 I 45131 45/13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

90. A fire has occurred in the Cable Tunnel. The Unit Supervisor directs an Operator to perform 20M-56C.4.D, "Nuclear Operator #1 Procedure".

20M-56C.4.D requires 2CHS*HCV186, 'RCP Seal Hdr Flow Control Valve' to be failed by isolating air within 10 minutes of entering 20M-56C.4.B, "Unit Supervisor Procedure".

The reason for this action is to ~~~~~~~~~~~~~~~~~

A. provide a reactor coolant system inventory control flow path B. avoid thermal shock to the reactor coolant pump seals C. prevent the valve from opening due to a fire induced short circuit D. maximize flowrate through the charging header Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the methodology of the alternate shutdown procedures intent and methodology. The bases for the actions taken in this procedure are specific to the SRO position. Detailed knowledge of the procedure content is required.

K/A is met with the knowledge that a Licensed Operator will be isolating Instrument Air to fail open 2CHS-HCV186 during the performance of Alternate Safe Shutdown From Outside Control Room, and identify the operational effects of this evolution.

A. Correct. Locally failing 2CHS*HCV186 open within 10 minutes is the correct action taken by the BOP Operator when performing 20M-56C.4.D.

Providing an inventory flow path is identified in the Intent and Methodology procedure 20M-56C.4.A. The valve is failed open to prevent spurious fire induced operation.

B. Incorrect. Plausible distractor if the candidate thinks HCV186 is failed closed, but HCV186 is failed open, and there is a caution in 20M-56C.4.B (US procedure) stating that thermal shock to the seal may occur and cause increased RCP seal leak rates.

C. Incorrect. Plausible distractor since air operated valves are place in their desired positions to prevent spurious fire induced operation, however the intent of isolating air to the valve is to cause it to fail open, not prevent it from opening.

D. Incorrect. Plausible distractor if the candidate thinks HCV186 is failed closed, as this would add to increased charging flow capabilities.

Sys# System Category KA Statement 078 Instrument Air Generic Knowledge of RO tasks performed outside the main control System (IAS) room during an emergency and the resultant operational effects.

K/A# 2.4.34 KIA Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

20M-56C4.A rev. 14 pg.2 & 4 20M-56C4.D rev. 24 pg.2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

91. Given the following plant conditions:

• Unit 2 is currently in Mode 6

• Fuel movement is in progress

• Spent Fuel Pool (SFP) boron concentration (Cb) sample results have significantly dropped since last sample and are currently at the Technical Specification limit of 2000 PPM If SFP Cb continues to drop, what is the impact on shutdown margin?

A 5% (Keff < .95) shutdown margin will be _ _ _ _ _ __

A. no longer maintained regardless of SFP Cb B. maintained as long as SFP Cb > 495 PPM C. maintained as long as SFP Cb > 350 PPM D. maintained regardless of SFP Cb

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 91 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine the impact on SDM per the TS bases and TS action required. Detailed knowledge of the bases is required to determine the impact of the boron dilution.

KIA is met by predicting the impact of SFP Boron concentration lowering on the SFP shutdown margin.

A. Incorrect. Keff cannot be maintained < .95 for a credible dilution event. (refer to correct answer explanation)

B. Correct. According to TS 3.7.16 and its associated bases, the >2000 PPM limit conservatively assures Keff is maintained within the limit (Keff

<.95) for the worst case misplaced fuel assembly accident. In addition, this limit ensures no credible boron dilution event will reduce Cb < 495 ppm required during non-accident conditions to maintain Keff < .95.

c. Incorrect. Keff will not be maintained at this low of a boron concentration. This is a common value in Tech Specs so it is plausible that this choice may be selected.

D. Incorrect. It is incorrect that Keff will be maintained at any boron concentration in the SFP.

Sys# System Category KA Statement 033 Spent Fuel A2 Ability to (a) predict the impacts of the following malfunctions or Inadequate SOM Pool Cooling operations on the Spent Fuel Pool Cooling System ; and (b) based System on those predictions, use procedures to correct, control, or mitigate (SFPCS) the consequences of those malfunctions or operations:

KIA# A2.01 KIA Importance 3.5 Exam Level SRO References provided to Candidate Technical

References:

TS 3.7.16 Bases, Rev. 5 None Question Source: Bank - 2LOT7 NRC Exam (Q93) Modified (2011)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 I 45.3 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

92. The plant was operating at 100% power when a SGTR occurred on 'B' Steam Generator.

• The reactor was manually tripped

• Condenser Vacuum is 18" Hg Vac and stable

• The crew has progressed up through step 6, 'Initiating RCS Cooldown' of E-3, "Steam Generator Tube Rupture"

• All previous EOP steps, including local Operator actions, have been completed How many separate valve controllers are available to MANUALLY cooldown the RCS in accordance with E-3 step 6?

A. 1 B. 2 C. 3 D. 4 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the EOP procedures. The loss of condenser vacuum will disable the condenser steam dumps as a possible flowpath. The procedure steps in E-3 will isolate the ruptured SG steam supply to the common RHR valve so it will be available at this time. Also the procedure cautions against use of the ruptured SG atmospheric, so 3 steam flowpaths are available. Detailed knowledge of the procedure actions to the step to cooldown is required to select the correct answer.

KIA is met by demonstrating the ability to cooldown the plant after a SGTR in conjunction with a low vacuum condition.

The low vacuum condition must be identified by the candidate as a valid Condenser NOT Available (C-9) alarm causing condenser steam dumps from being available.

A. Incorrect. Plausible distractor if candidate doesn't recognize that C-9 is in, and thinks the condenser steam dumps are available. This is the preferred steam relief flowpath.

B. Incorrect. Plausible distractor if candidate thinks only the 2 atmospheric valves from the intact SGs are available.

C. Correct. 2 atmospheric dumps (A & C) from the intact SGs, and the RHR valve can be used since it was isolated from the rupture SG in step 4b.

The condenser steam dumps are not available due to condenser vacuum at 18" is above the C-9 setpoint of 19.5" and annunciator A 12-4C

'Condenser Unavailable (C-9)' would be lit. C-9 blocks condenser steam dump operation.

D. Incorrect. Plausible distractor if candidate thinks 2 atmospheric dump valves, the RHR valve, and the condenser steam dumps are available or the ruptured SG could be used. Step 6 states go to ECA-3.1 is the ruptured SG must be used.

Sys# System Category KA Statement 041 Steam Dump System Generic Ability to verify system alarm setpoints and (SDS)/Turbine Bypass Control operate controls identified in the alarm response manual.

KIA# 2.4.50 KIA Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.E-3 lss. 2 Rev. 4, pgs. 4 & 10 20M-1.5.B.3 rev. 1 pg. 2 20M-26.4.ABM Rev. 2 pg. 3 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.3)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

93. The plant is operating at 100% power.

• The crew is implementing AOP-2.34.1 "LOSS OF STATION/CNMT INSTRUMENT AIR"

• [2SAS-C21A] STATION AIR COMPRESSOR TRIPPED

• [2SAS-C22] CONDENSATE POLISHING AIR COMPRESSOR is RUNNING

• [21AS-C21] DIESEL DRIVEN AIR COMPRESSOR failed to start

• [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV is OPEN

[21AS-Pl106] STA INSTR AIR HEADER PRESSURE is currently 72 psig and lowering

1) What is the next action required to be taken in accordance with AOP-2.34.1?

2) At what pressure will a Manual Reactor Trip, and transition to E-0, "Reactor Trip or Safety Injection" be required?

A. 1) CLOSE [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV

2) 55 psig

8. 1) ISOLATE [21AS-DRY23A & 238] INSTR AIR DRYERS

2) 55 psig C. 1) CLOSE [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV

2) 65 psig D. 1) ISOLATE [21AS-DRY23A & 238] INSTR AIR DRYERS

2) 65 psig

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 93 Answer: C Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. II E. SRO is required to have knowledge of the content of the procedure versus knowledge of the overall mitigative strategy or purpose, Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures. The knowledge of the procedural sequence is required based on the continuous action step 5 which checks SIA header pressure >86 psig, since it is not, the RNO of the step must be completed. This directs the closing of 2SAS-AOV105. The knowledge of tripping the Rx at 65 psig is contained within the AOP step RNO and is detailed procedural knowledge for the SRO. It is stated in a NOTE in the AOP attachments that at 65 psig, the MFW Regulating Valves will fail closed.

KIA is met the ability to predict the effects of an air leak in the station air system with a failure of redundant air compressors, and recognize that the station to Instrument Air header cross connection valve has failed to close. Then determine that based on an air pressure <65 psig, that a Rx trip and entry into E-0 is required .

.*A. Incorrect. It is correct to close AOV105, but incorrect Inst Air pressure for manually tripping the Rx and transitioning to E-0. 55 psig is a plausible distractor because this was the old AOP value.

B. Incorrect. Isolation of the Station Air system from the Inst Air header is performed prior to bypassing around and isolating the Inst Air Dryers procedurally. 55 psig is a plausible distractor because this was the old AOP value.

C. Correct. Closing AOV105 is required per the AOP because it should have automatically closed at 86 psig and isolated Station Air from Inst Air.

65 psig is the correct value requiring a manual Rx trip and transition to E-0, per continuous action step 5 when 21AS-Pl106 is S65 psig. This basis is stated in a NOTE in the AOP attachments stating that 65 psig the MFW Regulating Valves will fail closed.

D. Incorrect. Isolation of the Station Air system from the Inst Air header is performed prior to bypassing around and isolating the Inst Air Dryers procedurally. Correct air pressure for manually tripping Rx and transitioning to E-0.

Sys # System Category KA Statement 079 Station Air A2 Ability to (a) predict the impacts of the following malfunctions or operations on Cross-connection with IAS System the SAS; and (b) based on those predictions, use procedures to correct, control, or (SAS) mitigate the consequences of those malfunctions or operations:

KIA# A2.01 KIA Importance 3.2 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.34.1 Rev. 19 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

94. The plant is operating at 100% power when an Operator filling a Tech Spec required watch position becomes ill and leaves the site.

Per TS 5.2.2 and 10CFR50.54, the crew composition may remain less than the minimum for a period of time not to exceed (1)

If the vacant position is NOT refilled within the required time, the crew will (2)

A. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) take action to place the unit in MODE 5 B. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) maintain current power level C. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) take action to place the unit in MODE 5 D. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) maintain current power level Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .A page 17 third bullet. SRO is required to have knowledge of the TS section 5 and 6 actions related to plant staffing. Additionally the SRO is required to know the administrative procedure content related to not meeting staffing requirements.

K/A is met with the knowledge of how long an on-duty shift position may be unfilled, and the requirements of Tech Specs and Plant Procedures to maintain safe plant operation.

A. Incorrect. See correct answer.

B. Incorrect. See correct answer.

C. Incorrect. See correct answer.

D. Correct. In accordance with TS 5.2.2, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the expected time to fill the required position provided immediate action is taken to restore shift composition to minimum. BVPS has incorporated Licensing position on TS 5.2.2 into NOP-OP-1002, which states it is not conservative to place the plant into a transient due to staffing, therefore maintain the unit in a steady state condition and continue calling out personnel.

Sys# System Category KA Statement N/A NIA Generic Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc.

K/A# 2.1.4 KIA Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

T.S. 5.2.2 Amend 278/161 pg. 5.2-1 NOP-OP-1002 Rev. 10 sect. 4.1.13 Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.2)

Objective: 3SQS-48.1 Obj. 3 From memory, describe the required actions if less than the minimum shift staffing complement exists.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

95. The plant is operating in Mode 6 with all systems in normal alignment for this Mode.

• Core Off-Loading activities are in progress and core off-load is half complete

• Source Range Channel N31 fails LOW

• Source Range Channel N32 remains OPERABLE

• Refuel cavity water clarity is murky

• Gamma Metrics detectors N52A and N52B are Out of Service Which of the following activities can be performed WITHOUT violating the Technical Specification required actions for Source Range Instrumentation?

A. Latch and move a spent fuel assembly from the upender to the Spent Fuel Pool.

B. Latch and move a spent fuel assembly from the core to the Spent Fuel Pool.

C. Install a temporary secondary source into a core location.

D. Add Hydrogen Peroxide mixed with primary grade water to the refueling cavity for cleanup.

Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 first and third bullet.

SRO must have knowledge of the TS bases to answer this question. Specifically, SRO must know and apply the TS definition of core alteration and be familiar with the TS bases discussion on what is allowed and not allowed, with respect to compliance with the action statements. Additionally, the SRO must be knowledgeable of the "safe" locations defined in TSs and will be responsible for directing the operator actions to comply with the TS actions.

K/A is met with the knowledge of permissible actions iaw Tech Specs during core alterations when a Source Range detector is inoperable.

A. Correct. The SRO must understand that a loss of N31 puts them into AOP-2.2.1A 'SR Channel Malfunction' and TS 3.9.2. Both the AOP & TS direct that core alterations are immediately suspended. Core alterations are defined as movement of any fuel, sources, or reactivity components, within the reactor vessel with the vessel head removed and with fuel in the vessel. The SRO must have knowledge of the administrative requirements associated with refueling activities and have knowledge of TS bases. In order to answer this question the SRO must know the definition of Core Alterations and be able to apply this definition to a set of plant conditions. The movement of a Spent Fuel Assembly from the upender to the SFP is allowable because it is not within the reactor vessel.

B. Incorrect. Latching and moving a fuel assembly from the core would not be allowable by definition. Removing the assembly would not be considered placing the assembly in a safe location. This is plausible because some TS such as TS 3.9.4 LCO preclude core onload but do allow core offload to continue.

C. Incorrect. Plausible that operationally an alternative source could be installed, however, it is not allowed by the definition for what constitutes a Core Alteration.

D. Incorrect. Plausible that hydrogen peroxide is added to the water for clarity and cleanliness. However, the addition of primary grade water into the RCS would violate the second part of TS 3.9.2 since primary grade water could reduce boron concentration and is not allowed.

Sys# System Category KA Statement N/A Generic Conduct Of Operations Knowledge of procedures and limitations involved in core alterations.

KIA# 2.1.36 KIA Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

20M-53C.4.2.2.1A, Rev. 9, Pg. 8; TS Definitions Pg. 1.1-2; TS B3.9.2 Pg. B3.9.2-2 Question Source: Bank - 1LOTS NRC Exam (095)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 / 43.6 I 45.7)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

96. Given the following:

• Following a turbine runback, the crew is stabilizing the plant in accordance with the appropriate procedure.

• Control Bank "D" Group Counters are at 180 steps.

• On DRPI, one Control Bank "D" rod indicates 196 steps; all others indicate 182 steps.

• The affected rod has a blown movable gripper fuse and has been determined to be trippable.

Which of the following describes the technical specification implications of this event?

A. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.

B. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

C. The rod is INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.

D. The rod is INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine the impact on Rod alignment and TS action required. Detailed knowledge of the bases is required to determine which TS actions are applicable. This item meets the 10CFR55.43 (b) 2 SRO criteria because it requires the applicant to apply technical specification action with knowledge of the bases for that action.

K/A is met by demonstrating the ability to recognize that a rod exceeds alignment limits, but is still operable due to it being trippable. Knowledge of the TS bases is required to know why this is an undesired condition.

A. Correct. Since the rod is trippable it is operable. Restore rod to within alignment limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition B.

Misalignment limits are based on impact on power distribution limits iaw with TS 3.1.4 bases.

B. Incorrect. Correct, the rod is operable, but the concern for the situation presented is not shutdown margin.

C. Incorrect. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition A, but rod is not inoperable if it is trippable iaw TS 3.1.4 bases. If the rod were untrippable, then SOM would be affected. Power distribution limits are the correct reason for misaligned rods.

D. Incorrect. Would be true if the rod were untrippable.

Sys # System Category KA Statement N/A N/A Generic Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

KIA# 2.2.42 KIA Importance 4.6 Exam Level SRO References provided to Candidate None Technical

References:

TS 3.1.14, condition B, and basis Question Source: Bank-1LOT7 NRC Exam (091)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 /41.10/43.2/

43.3 I 45.3)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

97. In accordance with 1/2-EPP-IP-5.3, "Emergency Exposure Criteria And Control", whose authorization is required to exceed the emergency exposure limits of 10 CFR 20 "Standards For Protection Against Radiation" to save a life during an emergency, and what TEDE limit is this authorization limited to?

1) Who is authorized to grant exceeding the 10 CFR 20 limits during a declared emergency?

2) What is the maximum TEDE limit this individual may authorize?

A. 1) Emergency Director

2) 10 Rem TEDE B. 1) Emergency Director

2) 75 Rem TEDE C. 1) Emergency Recovery Manager

2) 10 Rem TEDE D. 1) Emergency Recovery Manager

2) 75 Rem TEDE Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and position responsibilities for the Emergency Director. This is a SRO position function only.

KIA met with knowledge of 10CFR20 emergency limits, and who may authorize emergency exposure limits, and what the limit is for saving a human life.

A. Incorrect. The ED is the only individual authorized to grant exceeding 10CFR20 emergency limits up to 75 Rem at which the Senior Vice President must give concurrence. 10 Rem is a plausible distractor because it is the 10CFR20 emergency exposure limit for preventing the failure of equipment necessary to protect the public health and safety.

B. Correct. The ED is the only individual authorized to grant exceeding 10CFR20 emergency limits up to 75 Rem at which time the Senior Vice President must give concurrence.

C. Incorrect. Plausible because the Emergency Recovery Manager has many responsibilities, and will discuss information with the ED, but ONLY the ED is authorized to grant exceeding 10CFR20 emergency limits. 10 Rem is a plausible distractor because it is the 10CFR20 emergency exposure limit for preventing the failure of equipment necessary to protect the public health and safety.

D. Incorrect. Plausible because the Emergency Recovery Manager has many responsibilities, and will discuss information with the ED, but ONLY the ED is authorized to grant exceeding 10CFR20 emergency limits. The expose limit of 75 Rem is the maximum allowed to be granted by the ED when the above stated conditions exist. Above 75 Rem requires the Senior Vice Presidents concurrence.

Sys# System Category KA Statement N/A NIA Generic Knowledge of radiation exposure limits under normal or emergency conditions.

KIA# 2.3.4 KIA Importance 3.7 Exam Level SRO References provided to Candidate none Technical

References:

1/2-EPP-IP-5.3, Rev. 11 pg. 3, 4, 9 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.12 I 43.4 I 45. 10)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

98. The Plant is operating at 100% power.

• Unit 2 is discharging the contents of the Gaseous Waste Storage tanks IAW 1/20M-19.4A.B, 'Unit 2 GW Storage Tk Disch To Unit 1 Atmos Vent'

• Rad Monitor RM-1GW-1088, Gaseous Waste Gas fails downscale and is declared inoperable

• The crew terminates the discharge In order to re-start the discharge, what 1/2-0DC-3.03, 'ODCM: Controls for RETS and REMP Programs' actions will be REQUIRED?

(Refer to attached reference)

A. The system/process flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (or assumed to be at the ODCM design value).

B. At least two independent samples of the tank's content are analyzed and at least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.

C. Grab samples (or local monitor readings) are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If grab samples are taken, these samples are to be analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Samples are continuously collected with auxiliary sampling equipment as required in ODCM Control 3.11.2.1, Table 4.11-2, or sampled and analyzed once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II D page 20 first bullet. SRO is required to have knowledge of the Offsite Dose Calculation Manual and actions required for failed monitoring equipment.

This is a SRO position function only.

This question requires the candidate to use their knowledge of the provided ODCM (84 pgs.), and determine which attachment is applicable for determining the status of RM-1GW-108B. Then determine whether a continuous or batch discharge is in progress. Based on that decision, follow up with any actions associated with the release.

KIA is met by identifying what controls must be used for radioactive releases if a Gaseous Waste gas detector is inoperable during a Gaseous Waste discharge.

A. Incorrect. Plausible distractor because this is a required action if FR-GW-108 is OOS (Action 28A) not RM-GW-108B.

B. Correct. IAW ODCM 1/2-0DC-3.03 Att.F page 38 and action 27 on page 42.

C. Incorrect. Plausible distractor because this is the required action for all continuous releases thru this pathway. (Action 29)

D. Incorrect. Plausible distractor because this is the required action for continuous releases if the alt channel 109 is also not available (Action 32).

For Batch release alt. RM-1GW-109 shall not be used as a comparable alternate monitoring channel iaw Att. F page 2.

Sys # System Category KA Statement NIA N/A Generic Ability to control radiation releases.

KIA# 2.3.11 KIA Importance 4.3 Exam Level SRO References provided to Candidate Y:z-ODC- 3.03 Technical

References:

ODCM Y:z-ODC-3.03 Rev. 13 Att.F pages 38 & 42 Question Source: Bank- 2LOT8 NRC Exam (Q98) (2012)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.11/43.4 / 45.10)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

99. Given the following plant conditions:

• Large Break LOCA occurred

• A release is in progress via a small breach in the Containment Equipment Hatch

• An ALERT was declared at 1500

• A Site Area Emergency was declared at 1530 In accordance with 1/2-EPP-IP-3.2, SITE Assembly and Personnel Accountability;

1) What is the latest time that a Site Accountability report must be completed?

2) What is the location to which Non-Emergency personnel will be directed for evacuation?

A. 1) 1530

2) Emergency Operations Facility

8. 1) 1600

2) Emergency Operations Facility C. 1) 1530

2) Hookstown Grange D. 1) 1600

2) Hookstown Grange Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and position responsibilities for the Emergency Director. This is a SRO position function only.

KIA is met with the knowledge of site accountability requirements and evacuation locations of site personnel.

A. Incorrect. 1530 is plausible because accountability must be completed within 30 minutes of a SAE. If the applicant assumes in error that accountability is required for Alert then 1500 + 30 min = 1530. The EOF is plausible because it is an offsite location where emergency response personnel report.

B. Incorrect. 1600 is the correct time, but non-emergency personnel will be directed to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C. The EOF is plausible because it is an offsite location where emergency response personnel report.

C. Incorrect. 1530 is plausible because accountability must be completed within 30 minutes of a SAE. If the applicant assumes in error that

=

accountability is required for Alert then 1500 + 30 min 1530. The ED may direct non-emergency personnel to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C.

D. Correct. IAW Y.-EPP-IP-3.2 sect. 8.2.6, site accountably must be completed within 30 minutes of a Site Area Emergency. The ED may direct non-emergency personnel to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C.

Sys# System Category KA Statement NIA N/A Generic Knowledge of the emergency plan.

KIA# 2.4.29 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

1/2-EPP-IP-3.2 rev. 19 pg. 7 & 13 Question Source: Bank - VC Summer 2013 SRO NRC Exam (Q25)

Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.11)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15) 100. Given the following plant conditions:

• The Emergency Director declared a Site Area Emergency at 1235

• The initial report to state and local government was completed at 1250

• An upgrade to General Emergency was declared at 1245

• The Initial Protective Action Recommendation (PAR) was made without a dose projection

• At 1255 a Valid dose projection is available which requires an upgraded PAR The Initial AND Upgraded (PAR) to the State/County Agencies must be given by which of the following times?

INITIAL UPGRADED A. 1245 1300 B. 1245 1310 C. 1300 1305 D. 1300 1310 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and notification requirements. This is a SRO position function only.

K/A is met with the knowledge of the time requirements of the initial PAR, and the upgraded PAR after a dose projection is available.

A. Incorrect. Plausible if it is determined that the Initial PAR must be declared at the time of the GE declaration (1245), and that the Upgraded PAR is declared 15 minutes after that (1300)

B. Incorrect. Plausible if it is determined that the initial PAR must be declared within 15 minutes of declaration of a GE (1245). Correct time for the Upgraded PAR determination, it must be completed within 15 minutes of assessment being available (ie: dose projection) at 1310 C. Incorrect. Plausible it is correct that the initial PAR must be declared within 15 minutes of declaration of a GE (1300). Incorrect time for the Upgraded PAR determination, if it assumed that the Upgraded PAR is required within 15 minutes of the notification time to the states and counties of the Site Area Emergency (1305)

D. Correct. The Initial PAR must be declared within 15 minutes of declaration of a GE. The upgraded PAR does not change emergency classification status. Upgraded PAR determination must be completed within 15 minutes of assessment being available (ie: dose projection) This is SRO level knowledge only.

Sys# System Category KA Statement N/A NIA Emergency Procedures/Plan Knowledge of emergency plan protective action recommendations.

KIA# 2.4.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

1/2-EPP-IP-4.1, Rev. 31, pg. 10 & 13 Question Source: Bank - 1LOTS NRC Exam (#100) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/ 41.12 /43.5/

45.11)

Objective:

KEY Beaver Valley Unit 2 - 2015 SRO NRC \Vritten Exam 1 A 35 A 69 c 2 D 36 c 70 c 3 B 37 D 71 c 4 B 38 A 72 D 5 c 39 B 73 D 6 c 40 D 74 B 7 c 41 B Ace¥f B trfb~ 75 B

.+ i!-/..a-,'lt-8 D 42 c 76 D 9 D 43 B 77 c 10 B 44 c 78 c 11 D 45 A 79 D 12 B 46 c 80 B 13 B 47 D 81 D 14 D 48 B 82 A 15 c 49 c 83 A 16 B 50 B 84 A 17 D 51 B 85 c 18 A 52 D 86 B 19 A 53 A 87 B 20 c 54 c 88 c 21 A 55 D 89 A 22 A 56 A 90 A 23 c 57 A 91 B 24 D 58 A 92 c 25 A 59 D lu-k.k~~ 93 c 26 c 60 D ~I- 94 D 27 c 61 A 95 A 28 B 62 A 96 A

2§ D lu.l-t:t<- a -..2.. 63 A 97 B

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30 A 64 B 98 B 31 c 65 c 99 D 32 c 66 D 100 D 33 A Ac.c.0 A (;'(" C... 67 A

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34 B 68 B Page 1of1

ES-401 Form ES-401-7 Site Specific RO Written Examination Cover Sheet U. S. Nuclear Regulatory Commission Site Specific RO Written Examination BV2LOT15 RO Written Exam Applicant Information Name:

Date: Facility/Unit: Beaver Valley Unit 2 Region: I lR1 11 0 111 0 IV 0 Reactor Type: W [R] CE 0 BW 0 GE 0 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results Examination Values 15 Points Applicant's Scores Points Applicant's Grade Percent NUREG-1021, Revision 10 FENOC Facsimile r2

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

1. The plant is at 100% power with all systems in normal alignment EXCEPT:

• Power Range Channel N-44 has been declared inoperable

• Power Range Channel N-44 has been removed from service IAW AOP-2.2.1 C, "Power Range Channel Malfunction" Power Range Channel N-43 NOW fails HIGH.

• All systems function as designed

• No Operator Actions have been taken Which, of the below listed First Out Annunciators (ANN. A5), will alarm in the FIRST 45 seconds AFTER N-43 fails High?

(1) A5-1 D 2/3 Loops Overtemp AT Reactor Trip (2) A5-2A Reactor Protection System Train A Trouble (3) A5-5G Reactor Trip Due To Turbine Trip (4) A5-68 Turbine Anti-Motoring Turbine Trip (5) A5-6D Turbine Trip Due To Reactor Trip (6) A5-7D Generator Trip Due To Turbine Trip A. 3, 5, 6 ONLY B. 2, 4, 6 ONLY C. 1, 3, 5, 6 ONLY D. 1, 2, 3, 4 ONLY Answer: A Explanation/Justification: K/A is met because the candidate must determine which First Out Annunciators on the A5 panel (Rx trip status panel) will alarm 45 seconds after a Rx Trip occurs.

A. Correct. The reactor will trip due to actions of N-44 failing having been completed which places N-44 bistable to TRIP, then N-43 fails High meeting the required 2/4 PR high Rx trip coincidence. These 3 are normal Rx trip annunciators, and reasons for the other annunciators are not correct are given in the other explanations.

B. Incorrect. (2) is plausible because candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent will energize on the trip. (4) Anti-motoring would alarm if the output breakers did not open. Plausible if the stem of the question didn't state that all systems functioned as designed.

C. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into OT6T trip setpoint calculation. Therefore this alarm will NOT be energized.

D. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into OT6T trip setpoint calculation. (2) is plausible because candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent will energize on the trip. (4) Anti-motoring would alarm if the output breakers did not open. Plausible if the stem of the question didn't state that all systems functioned as designed.

Sys # System Category KA Statement 000007 Reactor Trip /1 EK2 Knowledge of the interrelations between a reactor trip and the following: Reactor trip status panel KIA# EK2.03 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

1.4.AAD, 26.4.AAI and 35.4.AAF UFSAR Fig. 7.3-8 Rev. 14 Question Source: Bank - 2LOT6 NRC Exam (01) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

2. The plant was at 100% power when the reactor tripped on low PRZR pressure. The crew suspects that a PORV opened inadvertently and is now stuck partially open.

Current conditions:

• PRZR pressure is 1885 psig and stable

• [2RCS-P1472], Pressurizer Relief Tank pressure indicates 35 psig.

1) Based on the conditions given, which of the following confirming indications could be expected if a PORV is stuck partially open?

2) High pressure on which of the following tanks will prevent 2RCS-MOV523, Pressurizer Relief Tank Drain Valve from opening?

A. 1) PORV relief line temperature stabilized at 259°F

2) Pressurizer Relief Tank B. 1) PORV relief line temperature stabilized at 281°F

2) Pressurizer Relief Tank C. 1) PORV relief line temperature stabilized at 259°F

2) Primary Drains Transfer Tank D. 1) PORV relief line temperature stabilized at 281°F

2) Primary Drains Transfer Tank Answer: D Explanation/Justification: KIA is met because the partially open PORV will be discharging PRZR vapor space to the PRT. The candidate will have to evaluate the condition to determine what the temperature is at the outlet of a throttled valve (PORV) based on the determined saturation pressure, then using knowledge of the interlocks associated with the PRT drain valve, determine high pressure in the Primary Drains Transfer Tank will prevent the valve from opening.

A. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure= 50 psia). Plausible distractor because PRT high pressure will prevent the PRT spray valve from opening on high pressure.

B. Incorrect. 281°F is the saturation temperature corresponding to 50 psia. Plausible distractor because PRT high pressure will prevent the PRT spray valve from opening on high pressure.

C. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure = 50 psia). PRT drain valve will not open if the Primary Drains Transfer Tank pressure is high.

D. Correct. 281°F is the saturation temperature corresponding to 50 psia. PRT drain valve will not open if the Primary Drains Transfer Tank ressure is hi h.

Sys# System Category KA Statement 000008 Pressurizer (PZR) Vapor Space AK1. Knowledge of the operational Thermodynamics and flow characteristics of open Accident (Relief Valve Stuck Open) I 3 implications of the following concepts as or leaking Valves they apply to a Pressurizer Vapor Space Accident:

KIA# AK1 .01 KIA Importance 3.2 Exam Level RO References provided to Candidate steam Tables Technical

References:

Steam Tables 20M-6.1.D Rev. 3 pg.8 Question Source: Bank- Vision #135831 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

3. The plant is at 100% power.

• Annunciator A2-5C, Reactor Coolant Pump Vibration Alert/Danger Alarms

• 'B' RCP shaft vibration is 16 mils and stable

• 'B' RCP frame vibration is 1 mil and stable

• The crew enters AOP-2.6.8, "Abnormal RCP Operation" While performing the actions of AOP-2.6.8, the following additional alarms and indications are received:

• A2-5D, Reactor Coolant Pump Seal Vent Pot Level High/Low (RCP 21 B Seal Pot Lvl High, computer address point L0508D)

• A2-4D, Reactor Coolant Pump Seal Trouble (RCP 21 B Seal Lk Off, 2CHS-FT1558 Low, computer address point F0128D)

• RCP 21 B Seal Lk Off, 2CHS-FT1558 is 0.80 gpm and stable Based on these alarms and indications, which 'B' RCP seal has failed?

A. #1 seal B. #2 seal C. #3 seal D. Low pressure seal Answer: B Explanation/Justification: KIA is met because the 'B' RCP has a high vibration condition which leads to a coolant pump seal failure. The candidate must analyze the indications given to determine which seal has failed due the malfunction.

A. Incorrect. If #1 seal had failed seal leak-off flow would be high NOT low.

B. Correct. IAW 20M-7.4.AAH, 6.4.AAE and AOP-2.6.8 C. Incorrect. If #3 seal had failed the seal vent pot level low would be indicated NOT high.

D. Incorrect. The low pressure seal is not functional when the motor is coupled to the pump.

Sys# System Category KA Statement 000015/0 Reactor AK2. Knowledge of the interrelations between the Reactor Coolant RCP seals 00017 Coolant Pump Malfunctions (Loss of RC Flow) and the following:

Pump (RCP)

Malfunctions I 4

KIA# AK2.07 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.AAE Rev. 12 20M-7.4.AAH Rev. 22 Question Source: Bank - 2LOT6 NRC Exam (04) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

4. The plant is at 100% power.

• 2CHS*MOV289, 'Normal Charging Header lsol Viv' fails CLOSED

• The Immediate Operator Actions of the appropriate AOP have been performed

• NO other actions have been completed

• Total Seal Injection flow is 20 gpm

• Total Seal Water Leakoff is 9 gpm

1) When will A4-1C, Pressurizer Control Level Deviation High/Low alarm?

2) Will the alarm result from high or low deviation?

A. 1) 22 - 28 minutes

2) High

8. 1) 41 - 49 minutes

2) High C. 1) 4 - 6 minutes

2) Low D. 1) 10 - 15 minutes

2) Low Answer: B Explanation/Justification: KIA is met because the candidate must determine how the PRZR level will respond to a loss of charging and letdown (due to AOP IOAs). The candidate must realize that with the charging pump still in service (given that seal injection flow is 20 gpm) and seal leakoff flow, PRZR level will still change even with a Loss of Reactor Coolant Makeup flowpath.

Net charging= (Charging flow)+ (Total seal inj flow) - (Letdown flow) - (Total seal leakoff), 1% PRZR level =100 gal.

PRZR level deviation setpoint is +/-5% above/below program level A. Incorrect. It will take 25 minutes to receive the HIGH deviation if it is assumed that Letdown is isolated and all seal injection flow is entering the RCS/PRZR. The misconception is that 9 gpm is seal leakoff going back to charging pump suction ..

B. Correct. It will take 45 minutes to receive the HIGH deviation. (0 charging) + (20 seal inj) - (Letdown) - (9 leakoff) = 11 gpm net charging.

500 gal/11 gal=45 min to raise PRZR 5% above program level.

C. Incorrect. It would take -5 minutes to receive the LOW deviation if it is not recognized that Letdown was isolated as an IOA and misconceptions of seal injection volume entering the RCS. 0 + 20 -105- 9 = (500/94 gpm) = 5.3.

D. Incorrect. It would take -10 minutes to receive the LOW deviation if it is not recognized that Letdown was isolated (only 1 orifice in service) as an IOA and misconceptions of seal injection volume entering the RCS. 0 + 20 9 = (500/49 gpm)= 10.2, or using 45 gpm orifice (500/34)=14.7.

Sys# System Category KA Statement 000022 Loss of Reactor AK 1. Knowledge of the operational implications of the following Relationship between charging flow and Coolant Makeup I 2 concepts as they apply to Loss of Reactor Coolant Makeup: PZR level KIA# AK1 .03 KIA Importance 3.0 Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2. 7.1 Rev 7 None 20M-6.4.AAL Rev. 8 20M-6.1.C Rev. 5 pg.12 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 / 41.10 / 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

5. The following conditions exist:

• Plant is in Mode 5

• A1-2G, INCORE INSTR ROOM/CNMT SUMP LEVEL HIGHNALVE NOT RESET is in alarm

• RHR HX A INLET TEMP is 113°F and STABLE

• 2RHS*FCV605A, RHS TRAIN A HX BYPASS FLOW CONTROL valve has OPENED an additional 5% in response to the leak Based on the given conditions:

1) Which of the following could identify the location of the leak in the RHR system?

2) Which of the following procedures would be used to isolate the affected train of RHR?

A. 1) RHS*MOV720A, 'RHS Train Return to B Loop Isolation' INLET flange

2) AOP-2.10.1, "Loss of Residual Heat Removal Capability" B. 1) RHS*MOV720A, 'RHS Train Return to B Loop Isolation' INLET flange

2) AOP-2.6.5, "Shutdown LOCA" C. 1) 2RHS-E21A, "A' RHR HX' OUTLET flange

2) AOP-2.10.1, "Loss of Residual Heat Removal Capability" D. 1) 2RHS-E21A, "A' RHR HX' OUTLET flange

2) AOP-2.6.5, "Shutdown LOCA" Answer: C Explanation/Justification: KIA is met by the following. Based on the response of the Bypass Flow Control valve, the candidate must determine the leak location within the RHR system. Once the location is determined, the candidate must decide which AOP would be used to mitigate and isolate the leak in the RHR system.

A. Incorrect. A leak at the inlet of RHS-MOV720A is downstream of (FT605A) flow element, therefore the loss in flow would not cause FCV-605A to respond because it is not seen by the flow element. AOP-2.10.1 is the correct procedure for plant conditions.

B. Incorrect. A leak at the inlet of RHS-MOV720A is downstream of (FT605A) flow element, therefore the loss in flow would not cause FCV-605A to respond because it is not seen by the flow element. AOP-2.6.5 is used for a loss of coolant accident when in mode 3 (after the accumulators are isolated) or mode 4.

C. Correct. A leak at the outlet of the RHR Hx will cause flow to be lower at (FT605A) flow element downstream of the Hx and FCV. This reduced flow will cause FCV-605A to open to raise flow back to the desired setpoint. AOP 2.10.1 is the correct procedure when there is a loss of coolant accident in mode 5 & 6.

D. Incorrect. This is the correct leak location, but the incorrect procedure for the plant conditions. AOP-2.6.5 is used for a loss of coolant accident when in mode 3 (after the accumulators are isolated) or mode 4.

Sys# System Category KA Statement 000025 Loss of Residual Heat AA2. Ability to determine and interpret the following as they Location and isolability of leaks Removal System apply to the Loss of Residual Heat Removal System:

(RHRS) / 4 KIA# AA2.04 KIA Importance 3.3* Exam Level RO References provided to Candidate None Technical

References:

RM-0410-001 Rev. 16 20M-53C.4.2.10.1 rev. 12 pgs.1 & 14 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

6. The plant was operating at 100% power when a Loss of Coolant Accident occurred.

The following conditions exist:

• The Reactor automatically tripped

• RCS pressure is 1190 psig and lowering

• All SG pressures are 950 psig and lowering

• Containment pressure peaked at 13 psig and is currently 9.5 psig and lowering

• No HHSI is indicated

• The crew is performing the actions of E-0, "Reactor Trip and Safety Injection" In accordance with E-0, what is the expected crew response for the Reactor Coolant Pumps (RCPs), and why?

1) The RCPs should be ~~~~~~~~

2) Because _ _ _ _ _ __

A. 1) left running

2) to continue pumping the water/steam mixture through the core B. 1) tripped

2) D/P between RCS pressure and highest SG pressure trip criteria has been met C. 1) tripped

2) thermal barrier and motor cooling flow was secured D. 1) tripped

2) the RCP Seal Water Return Isolation Valves, 2CHS*MOV378 and 381 are isolated Answer: C Explanation/Justification: KIA is met because the candidate must know that E-0 gives the guidance to trip the RCPs if a CIB occurs. The reason for the trip is that primary component cooling water is isolated to the thermal barrier and motor.

A. Incorrect. Plausible distractor is the candidate does not recognize that a CIB has occurred and isolated CCP to the RCP motor. The RCP operation in a water/steam 2 phase condition possibly could occur in FR-C.1, but not E-0.

B. Incorrect. Tripping the RCP is correct, but for the incorrect reason. Trip criteria is 220 psid with adverse conditions and HHSI flow exists. The stem states no HHSI flow, and DP is 240 psid.

C. Correct. With Containment pressure peaking at 13 psig, CIB occurred and isolated (CCP) to both 'A' & 'B' containment headers. This will remove cooling water to the thermal barrier, upper & lower bearing cooler, and the stator cooler. In accordance with E-0 LHP, it is required to trip the RCPs on a loss of CCP flow to the RCPs.

D. Incorrect. Tripping the RCPs is correct, but it is not a correct statement to trip because RCP Seal Water Return Isolation Valves have closed on CIA. Seal Return Header Relief 2CHS-RV382A provides protection for the low pressure RCP seal return piping when this header is isolated (CIA),

and relieves to the PRT.

Sys # System Category KA Statement 000026 Loss of Component AK3. Knowledge of the reasons for the following responses as Guidance actions contained in EOP for Loss of Cooling Water they apply to the Loss of Component Cooling Water: CCW (CCW)/ 8 KIA# AK3.03 KIA Importance 4.0 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.E-O lss. 2 Rev.1 LHP 20M-53B.5.Gl-6 lss. 2 Rev.a pg. 51 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 / 45.6 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

7. The plant is at 75% power with all systems in normal alignment for this power level EXCEPT

[2RCS*MOV535] PORV 455C MOTOR OPERATED ISOL VLV is CLOSED due to PORV seat leakage.

• [2RCS*PK444A] PRZR PRESS CONTROL output fails to 10% in Automatic

• No Operator action is taken What is the status of the plant 15 minutes after this event?

A. The plant will trip due to a PRZR Pressure LOW reactor trip.

B. The plant will trip due to a PRZR Pressure HIGH reactor trip.

C. The plant will remain at power and RCS pressure will cycle between the PORV lift and PORV lift reset setpoints.

D. The plant will remain at power and RCS pressure will cycle between the PORV lift and PRZR Low Pressure PORV Block Interlock setpoints.

Answer: C Explanation/Justification: K/A is met by the knowledge required to determine that the purpose of 2RCS*PK444A is to control PRZR pressure and how the controller will function when it fails to 10%. This knowledge will encompass the response of the PORVs in the non-affected portion of the Pressurizer Pressure Control System, and the control function side with the heaters and spray valves, and the resulting 455C PORV being isolated.

A. Incorrect. Plausible if 2RCS*PK444 output failed high, causing both spray valves to open fully and depressurize the RCS until LP Rx trip occurs.

B. Incorrect. Plausible distractor because pressure would rise to 2375 psig and cause a HP Rx trip if the PORVs didn't lift at 2335 psig. Candidate must know that the Rx trip setpoint is higher than the PORV opening setpoint, and that the PORV will prevent the trip setpoint from being reached.

C. Correct. The plant will remain at power and the pressure will cycle between the PORV auto open setpoint of 2335 psig and close setpoint of 2315 psig. This will occur only on PORV 456 & 4550 since 455C is manually isolated.

D. Incorrect. Plausible distractor if it is not recognized that the 456 & 4550 PORVs will auto close at 2315 psig, and they think that pressure will reduce to 2185 psig causing the block valves to close.

Sys# System Category KA Statement 000027 Pressurizer Pressure Control Generic Knowledge of the purpose and function of major System (PZR PCS) system components and controls.

Malfunction I 3 KIA# 2.1.28 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF Att.2 Rev.13 2om-6.2.B Rev. 13 pg.8 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)1 Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

8. The crew is implementing FR-S.1, "Response to Nuclear Power Generation/ATWS".

Which of the following combinations of breaker positions indicate that the reactor has been tripped?

Note: The Motor Generator INPUT Breakers are both closed.

• RTA =Reactor Trip Breaker A

• RTB =Reactor Trip Breaker B

• BYA =Reactor Trip Bypass Breaker A

• BYB =Reactor Trip Bypass Breaker B

• MG21 =MG21 output Breaker

• MG22 =MG22 output Breaker LEGEND: X =CLOSED; 0 =OPEN RTA RTB BYA BYB MG21 MG22 A. x x 0 0 x 0 B. x 0 0 x 0 x C. 0 x x 0 x x D. x 0 x 0 x x Answer: D Explanation/Justification: KIA is met by the knowledge required to determine which combination of Rx trip, bypass, or motor generator output breaker positions will trip the reactor during an ATWS.

A. Incorrect. Only one MG set breaker is open. Both required to be open to trip reactor.

B. Incorrect. RTA and BYB closed will provide flowpath to the rod coils. Only one MG set breaker is open. Both required to be open to trip reactor.

C. Incorrect. RTB and BYA closed will provide flowpath to the rod coils.

D. Correct. With both RTB and BYB open, the MG set output supply to the rod coils is interrupted, which will result in the rods dropping into the core (a reactor trip).

Sys# System Category KA Statement 000029 Anticipated EK2 Knowledge of the interrelations between the and the following Breakers, relays, and disconnects Transient anATWS:

Without Scram (ATWS)/ 1 KIA# EK2.06 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

Westinghouse 2001.409-001-018 Rev. L Question Source: Bank - DC Cook 2012 NRC Exam (046)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 I 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

9. The plant is operating at 50% power with the crew implementing AOP-2.6.4, "Steam Generator Tube Leakage" for a 50 GPO tube leak on 'A' SG.

A Steam leak inside CNMT occurs requiring a Rx Trip.

• E-0, Reactor Trip or Safety Injection" has been entered

• Safety Injection is actuated Assuming no operator actions other than E-0 immediate operator actions, which of the following describes why 2SSR*AOV117A, '21A SG Slowdown Sample Outside Cnmt lsol Viv' would automatically close?

A. Low RCS pressure, to provide containment isolation.

B. Low SG pressure, to limit excess steam demand effect.

C. High containment pressure, to establish CNMT barrier integrity.

D. High secondary activity, to minimize radiological release.

Answer: D Explanation/Justification: KIA is met by the knowledge required to determine that 2SSR*AOV117A '21A SG Slowdown Sample Outside Cnmt lsol Viv' will automatically close due to a High radiation monitoring alarm on 2SSR-RQ100, and the reason for the isolation is to minimize radiological releases from the ruptured SG.

A. Incorrect. 2SSR*AOV117A is not closed by SI (CIA) signals which actuate on RCS low pressure. Plausible distractor with the steam leak, and a tube leak identified in the stem which both could possibly cause RCS pressure to lower to 1856 psig actuate SI, but has no effect on AOV117A.

B. Incorrect. 2SSR*AOV117A is not closed by SI, CIA, MSI signals which actuate on low SG pressure. Plausible distractor with the steam leak identified in the stem which could cause SG pressure to lower to 500 psig and actuate SI or MSI, but has no effect on AOV117A.

C. Incorrect. 2SSR*AOV117 A is not closed by SI, CIA, MSI, or CIB actuation signals which actuate on cnmt pressure. Plausible distractor with the steam leak inside cnmt identified in the stem which could cause cnmt pressure to raise and actuate SI or MSI, but has no effect on AOV117A.

D. Correct. 2SSR*AOV117A is closed by 2SSR-RQl100 High Alarm. Minimize radiological releases from the ruptured SG is the correct reason.

Sys# System Category KA Statement 000038 Steam Generator EK3 Knowledge of the reasons for the following responses as the Automatic actions associated with high Tube Rupture apply to the SGTR: radioactivity in S/G sample lines.

(SGTR) / 3 KIA# EK3.03 KIA Importance 3.6* Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.E-3 lss. 2 Rev. 4 20M-43.1.E Rev. 6 U2 RM-0414A-001 Rev. 18 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5/41.10145.61 45.13)

Objective: 2SQS-43.1 Obj. 7 Describe the control, protection and interlock functions for the control room components associated with the Radiation Monitoring System, including automatic functions, and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

10. The following conditions exist:

• The plant is at 75% power with all systems in normal alignment for this power level

• Charging pump 2CHS*P21A is RUNNING

• Condensate pumps 2CNM-P21A and Bare RUNNING A loss of 4160V Bus 20 occurs with all equipment operating as designed.

What procedure will the crew use to mitigate this event?

A. E-0, "Reactor Trip Or Safety Injection" B. AOP 2.24.1, "Loss Of Main Feedwater" C. AOP 2.7.1, "Loss Of Charging Or Letdown" D. AOP 2.36.2, "Loss of 4KV Emergency Bus" Answer: B Explanation/Justification: KIA is met by having the candidate determine what equipment will be lost based on interpreting the conditions given, after which they will determine that only 1 MFW pump was lost, and Loss Of Main Feedwater AOP is the appropriate procedure.

A. Incorrect. Plausible because the candidate may think a RCP will be lost, or a condensate pump would trip ('C' is powered by 'D' Bus}, or if they are thinking that we lost the 'B' MFP, without knowing that the AOP states <80% power, then reduce power to <52%.

B. Correct. With the plant at 75% power, AOP 2.24.1 is the correct procedure to mitigate the loss of 1 MFP when at power and <80%. The candidate should know by AOP-2.24.1 IOAs that the Rx should not be tripped if the plant is <80% power.

C. Incorrect. Plausible if the candidate thinks 2CHS-P21A is lost due to the bus loss, but this is incorrect.

D. Incorrect. Plausible if the candidate does not think the 'OF' bus will be energized by the 2-2 EOG. This is incorrect because the stem states that all equipment operated as designed.

Sys# System Category KA Statement 000054 Loss of Main AA2. Ability to determine and interpret the following as they apply to Differentiation between loss of all MFW and trip of Feedwater the Loss of Main Feedwater (MFW): one MFW pump (MFW}/4 KIA# AA2.02 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.24.1 Rev. 6 pg. 2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

11. The following conditions exist:

• The plant has tripped due to a loss of all 4Kv power

• The crew is performing ECA-0.0, "Loss of All AC Power"

• The BOP Operator is placing equipment in PTL in accordance with step 16 of ECA-0.0 Which of the following components will NOT be placed in PTL during this step, and what is the basis for not removing this component from service?

A. Auxiliary Feedwater pump; provide sufficient water to maintain an effective secondary heat sink.

B. Primary Component Cooling Pump; provide cooling to the Reactor Coolant Pump thermal barrier.

C. Charging Pump; provide cooling to the Reactor Coolant Pump seals.

D. Service Water Pump; provide cooling to the Emergency Diesel Generator.

Answer: D Explanation/Justification: KIA is met by the knowledge that ECA-0.0 places equipment to PTL to defeat automatic loading of large loads on the AC emergency bus with the exception of the Service Water Pumps. The knowledge of the bases of the SW pump remaining available to load on a diesel start to provide diesel cooling is expected RO knowledge.

A. Incorrect. MDAFW pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. During ECA-0.0, sufficient AFW flow is provided by the turbine driven AFW pump, so heat sink is not a concern.

B. Incorrect. CCP pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. Providing cooling flow to the hot thermal barrier is not required at this time.

C. Incorrect. Charging pumps are not required immediately after power restoration and are considered a large load. The goal of this step is to avoid potential overload of the energized emergency bus. Providing cooling flow to the hot RCP seals could cause thermal shock to the seals & shaft.

D. Correct. Service water pump auto start is desired to provide cooling to the EOG in the event it is locally restored. This is stated as a caution prior to step 13 and switches are placed to Auto in step 1 of Att. A-1.5 for starting the diesel locally.

Sys# System Category KA Statement 000055 Loss of Offsite and Generic Knowledge of the specific bases for EOPs.

Onsite Power (Station Blackout) I 6 KIA# 2.4.18 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ECA-O.O lss. 2 Rev. 3 pgs. 5 & 10 20M-53B.4.ECA-O.O lss. 2 Rev. 3 pg. 124 20M-53A.1.A-1.5 lss. 1C Rev. 5 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.1 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

12. The plant has been operating at 100% power with all systems in NSA for the past 100 days.

• An inadvertent reactor trip occurs coincident with a loss of offsite power

• All systems function as designed

• The crew is implementing the actions of ES-0.2, "Natural Circulation Cooldown"

• RCS temperature is 400°F lowering at 20°F/hr

• RCS Subcooling is 165°F

• RCS Pressure 1200 psig and stable Alarm A11-5G, CROM Shroud Fan Auto-Start/Auto-Stop is received. ALL CROM shroud fans have tripped and cannot be restarted.

What ramifications will the loss of these CROM Shroud Fans have on the continued performance of ES-0.2, "Natural Circulation Cooldown"?

A. Further RCS cooldown (below 350°F) cannot continue UNTIL a suitable RX vessel head soak has been performed.

B. Further RCS depressurization (below 1200 psig) cannot continue UNTIL a suitable RX vessel head soak has been performed.

C. Immediately INCREASE RCS pressure 100 psig to RAISE RCS subcooling.

D. Immediately DECREASE RCS pressure 100 psig to LOWER RCS subcooling.

Answer: B Explanation/Justification: KIA is met with the knowledge of ES-0.2 major action step of the requirement to cooldown and depressurize RCS with no upper head void growth when RCS forced flow is lost. Knowledge that the CROM cooling units offer alternative cooling to the upper vessel head region during NC cooldown since core bypass flow existing with forced convection is lost to the upper head region is important to cooling down the head.

A. Incorrect. With the CROM fans lost, the RCS pressure is held at 1200 psig while the RCS is cooled down below 350F. At these conditions there is approx. 200F of subcooling is required for the RCS due to the heat buildup in the head. The 9 hour1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> soak time will ensure the head cools off sufficiently since the CROM fans were lost B. Correct. IAW ES-0.2 step 15. Without any CROM fans running, subcooling requirements will be more than 3 times greater (200F). The CROM fans provide alternate cooling to the upper vessel head region during NC CID since core bypass flow existing with forced convection CID is lost in the upper vessel head region. Since this forced cooling is lost, a 9 hr soak is required to ensure the head cools to less than saturation temp for 400 psig.

C. Incorrect. Raising pressure 100 psig is a technique employed by ES-0.4 natural circulation procedure when the head void growth becomes too large.

D. Incorrect. Minimizing Subcooling is a technique employed when RX vessel stresses are the concern but NOT when RX vessel head voids are the concern. Decreasing pressure may actually cause a void to form.

Sys # System Category KA Statement 000056 Loss of AK1. Knowledge of the operational implications of the following Principle of cooling by natural convection Offsite Power concepts as they apply to Loss of Offsite Power:

16 KIA# AK1 .01 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-0.2 lss. 2 Rev. 1 step 15 20M-53B.4.ES-0.2 lss. 2 Rev. 1 pg. 51 Question Source: Bank-2LOT6 NRC Exam (Q27) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

13. Reactor power is at 27% with all systems in normal alignment EXCEPT:

• Rod control is in MANUAL A loss of Vital Bus 2 occurs.

Which of the following indications on Panel 308, PRI PLANT PARAMETERS STATUS panel will be LIT due to the loss of Vital Bus 2?

1) PWR RNG N41 TO P-8

2) PWR RNG N42 HIGH POWER SP

3) RCS LOOP A OT~T RX TRIP

4) SOURCE RNG N32 RX TRIP A. 1 and 3 B. 2 and 4 C. 1and4 D. 2 and 3 Answer: B Explanation/Justification: KIA is met by the candidate interpreting the effects of a loss of vital bus 2 will have on the RPS system, and determining which RPS panel alarms and trip indications will illuminate due to the loss of power/equipment out of service.

A. Incorrect. N41 to P-8 is normally lit when >30% power, but power is 27% in the stem and it is powered from VB1. RCS LOOP A OTL1T RX TRIP is CH1 powered from VB1 and not effected.

B. Correct. PWR RNG N42 HIGH POWER SP and SOURCE RNG N32 RX TRIP are both normally off when not above a setpoint condition. They are both CH2, so in this case, a loss of VB2 will cause the bistable lights to illuminate.

C. Incorrect N41 to P-8 is normally lit when >30% power, but power is 27% in the stem and it is powered from VB1. SOURCE RNG N32 RX TRIP will be lit due to the loss of VB2.

D. Incorrect. PWR RNG N42 HIGH POWER SP is CH2, so in this case, a loss of VB2 will cause the bistable light to illuminate, but RCS LOOP A OTL1T RX TRIP is CH1 powered from VB1 and not effected.

Sys# System Category KA Statement 000057 Loss of Vital AC AA2. Ability to determine and interpret the following as they RPS panel alarm annunciators and trip indicators Electrical apply to the Loss of Vital AC Instrument Bus:

Instrument Bus I 6 KIA# AA2.03 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.38.1B Rev. 6, pgs. 1, 18, 21 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

14. The plant is at 100% power.

• 21A Pri Comp Cooling Hx Temp Control Viv 2CCP-TCV100A is in Manual

• 21 B Pri Comp Cooling Hx Temp Control Viv 2CCP-TCV100B is in Manual

• CCS Hx Bypass Temp Control Viv 2CCS-TCV215 is in Auto

• An inadvertent Train 'A' CIA occurs What are the effects on the Primary and Secondary Component Cooling Water Hx outlet temperatures 10 minutes after the inadvertent Train 'A' CIA occurs?

Primary Component Cooling Water Heat Exchanger (CCP) outlet Temperatures will (1)

Secondary Component Cooling Water Heat Exchanger (CCS) outlet Temperatures will (2)

A. 1) rise

2) rise B. 1) rise

2) lower C. 1) lower

2) lower D. 1) lower

2) rise Answer: D Explanation/Justification: KJA is met by demonstrating knowledge of the increased service water flow through the CCP Hxs to the Circ Water pump suction when the secondary portion of service water is isolated. Secondary side of service water is isolated due to 2 Train 'A' CIA isolation valves closing, and isolating both SWS headers to the Secondary Cooling Water Hxs, thus causing a loss of CCS.

A. Incorrect. Temp will lower with more SW through the Hxs. Plausible distractor because they must know that Train 'A' CIA will close both 107A &

C and isolate both Trains of SW to the CCS Hxs, not the CCP HXs (CCP Hxs are isolated on CIB). It is correct that CCS temps will rise.

B. Incorrect. Temp will lower with more SW through the Hxs. Plausible distractor because they must know that Train 'A' CIA will close both 107A &

C and isolate both Trains of SW to the CCS Hxs, not the CCP HXs (CCP Hxs are isolated on CIB). CCS temps will rise with SW isolated.

C. Incorrect. CCP temp will lower due to increased SW flow through the CCP Hxs with the TCVs in manual and the secondary side SW isolated.

Plausible distractor of CCS temperature lowering if candidate thinks only the 'A' SW header is isolated from the CCS Hxs, and CCS loads have reduced due to the CIA.

D. Correct. CCP temp will lower due to increased SW flow through the CCP Hxs with the TCVs in manual and the secondary side SW isolated. The inadvertent Train 'A' CIA will close 2SWS-MOV107A & C, isolating both SW headers from the CCS HX causing CCS temps to rise.

Sys# System Category KA Statement 000062 Loss of Nuclear AK3. Knowledge of the reasons for the following responses as Effect on the nuclear service water discharge flow Service Water I 4 they apply to the Loss of Nuclear Service Water: header of a loss of CCW KIA# AK3.04 KIA Importance 3.5 Exam Level RO References provided to Candidate Technical

References:

20M-30.1.D Rev. 8, pg. 6, None U2 RM -0430-001 and 003 2SQS-30.1 PPNT Rev. 23 Slide 11 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.4, 41.8 I 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

15. The following conditions exist:

• The plant is at 45% power

• Station Air Compressor 2SAS-C21A is on Clearance for motor replacement

• ERF Substation bus 1H has tripped on overcurrent

• A7-3C, 'Turbine Bearing/Autostop Oil Trouble' is in alarm due to a loss of power to Main Turb Bearing Oil Pump 2TML-P207

• No Operator actions have been taken Based on the above indications, which of the following procedures would be entered to mitigate this event?

A. E-0, "Reactor Trip or Safety Injection" B. AOP-2.26.1, "Turbine and Generator Trip" C. AOP-2.34.1, "Loss of Station/Cnmt Instrument Air" D. AOP-2.37.1, "Loss of 480V BUS 2N OR 2P" Answer: C Explanation/Justification: KIA is met by requiring the candidate to recognize the entry conditions for AOP-2.34.1 "Loss of Station/Cnmt Instrument Air" which are given in the stem of the question as bus 1H tripped on overcurrent which supplies bus 2K (power supply for 2SAS-C21 B). With both SACs secured, the next air compressor to start will be 2SAS-C22 at 90 psig, therefore air pressure will be lowering. These 2 conditions are entry level conditions for loss of Station/Cnmt Instrument Air AOP.

A. Incorrect. Plausible is the candidate thinks the Rx will trip due to the Turb Bearing Oil Pump 2TML-P207 being de-energized causing a turbine trip, or if they think DC bus 2-6 (powered by 2K) will de-energize and trip the Rx similar to DC bus 2-1 or 2-2.

B. Incorrect. Plausible distractor with power <49% and receiving Turbine Bearing/Autostop Oil Trouble alarm. The cause of the alarm is due to Main Turb Bearing Oil Pump 2TML-P207 being de-energized. The candidate must realize that P207 is not running when the turbine is online, and will only auto start when Turb bearing oil header pressure is low. The Turbine will not trip with Bus 2K de-energized.

C. Correct. With the loss of the 1H 4160KV bus, a loss of the 480V 2K bus occurs. 2K supplies power to the 2SAS-C21 Bair compressor. When 'B' SAC trips, the stby air compressor 2SAS-C22 won't auto start until 90 psig. An entry condition for AOP 2.34.1 is running SAC trips, and Station Inst. Air pressure will be dropping.

D. Incorrect. Plausible if the candidate thinks the 1H bus supplies either the 2N or 2P 480v bus. Electrical system knowledge is required to not select thisAOP.

Sys# System Category KA Statement 000065 Loss of Generic Ability to recognize abnormal indications for system operating Instrument parameters that are entry-level conditions for emergency and Air/ 8 abnormal operating procedures.

KIA# 2.4.4 KIA Importance 4.5 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.34.1 Rev.19 pgs. 1 & 2 1/20M-53C.4A.58E.1 Rev. 9 pg. 8 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.2 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

16. The plant is at 45% power with all systems in normal alignment for this power level EXCEPT 2FWS-P21A, 'A' Main Feedwater Pump is cleared for bearing replacement.

• The crew is performing AOP-1/2.35.1, "Degraded Grid" due to grid voltage and frequency swings

• Control Rod Group Selector switch is in MANUAL due to N-44 failing LOW. NO other actions have been taken for the N-44 failure

• A 20% Load Rejection has occurred

• Tavg-Tref deviation is 8.5 °F Based on the above conditions, what are the required actions to restore the RCS Tavg-Tref deviation?

A. Manually close Condenser Steam Dumps B. Manually insert Control Rods C. Manually trip the Reactor D. Manually trip the Turbine Answer: B Explanation/Justification: KIA is met by requiring the candidate to recognize that the control rods must be manually operated to lower Tavg due to the group selector switch being in manual when a load rejection occurs during the degraded grid event, and then recognize that the Tavg-Tref deviation of 8.5F is outside the expected band of+/- 2F.

A. Incorrect. Manually closing the steam dumps would cause Tavg to raise higher and create a larger Tavg-Tref deviation.

B. Correct. With a deviation of 8.5F due to a load rejection (Tref lowering & Tavg remaining constant), the rods will have to be driven inward to lower Tavg to Tref. If the rods were in auto, and N-44 was operable the rods would automatically drive inward to reduce the mismatch. AOP 2.35.2 (Load rejection) states to restore Tavg-Tref by manual rod insertion or boration. AOP-2.35.2 is directed by AOP-1/2.35.1.

C. Incorrect. Tripping the Rx is not required or directed. It is a plausible distractor due to the Tavg to Tref deviation. The Transient response Guidelines state to trip the Rx if the mismatch is +/- 1OF and the cause cannot be readily determined. In this case the deviation is only 8.5F and there was a load rejection.

D. Incorrect. Tripping the Turbine is not required or directed. It is a plausible distractor because the initial conditions are <P9, larger than normal Tavg-Tref deviation, and AOP-1/2.35.1 has been implemented.

Sys# System Category KA Statement 000077 Generator Voltage AA 1. Ability to operate and/or monitor the following as they apply to Reactor controls and Electric Grid Generator Voltage and Electric Grid Disturbances:

Disturbances I 6 KIA# AA 1.04 KIA Importance 4.1 Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2.35.2 Rev. 20 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 and 41.10 / 45.5, 45.7, and 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

17. Given the following conditions:

• ECA-1.2, "LOCA Outside Containment", Step 4, directs checking Reactor Coolant System pressure to determine if the break has been isolated by previous actions.

If the break has NOT been isolated, which of the following identifies the effect that a transition to ECA-1.1, Loss of Emergency Coolant Recirculation, has on mitigating the accident?

Actions of ECA-1.1 are taken to ~~~~~~~~~~~~

A. transfer Safeguards Building Sump contents to the Refueling Water Storage Tank as directed by Technical Support Center.

B. increase the injection flow rate to restore Reactor Coolant System pressure.

C. stabilize Reactor Coolant System pressure to prevent the Safety Injection Accumulators from discharging out the break.

D. minimize Refueling Water Storage Tank depletion by reducing total injection flow.

Answer: D Explanation/Justification: KJA is met by placing the candidate into a specific point of an EOP, and requiring them to understand why they will operate the Safety Injection system when in ECA-1.1 to conserve RWST inventory.

A. Incorrect. Plausible because this action could be performed to recover lost sump water, however, it is not directed by ECA-1.1.

B. Incorrect. Plausible if thought RCS pressure is restored, however, actions are taken to restore RCS mass in ECA-1.1.

C. Incorrect. Plausible because multiple EOPs depressurize SI Accumulators, however, ECA-1.1 utilizes SI Accumulator inventory.

D. Correct. Adding makeup to the RWST and reducing injection flow will minimize RWST depletion which is identified as major action category #2.

Sys # System Category KA Statement W/E11 Loss of Emergency Coolant EA 1. Ability to operate and I or monitor the following as they Desired operating results during Recirculation I 4 apply to the (Loss of Emergency Coolant Recirculation) abnormal and emergency situations.

KIA# EA1.3 KIA 3.7 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-53A.1.ECA-1.1 Rev 1 lss 2 pg. 1 Question Source: Bank-Comanche Peak 2013 NRC exam Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.51 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

18. The plant was at 100% power.

• Main Steam Line break inside containment occurs

• Main Steam Line Isolation Valves failed to close

• Transition to ECA-2.1, "Uncontrolled Depressurization Of All Steam Generators" has occurred Current plant conditions:

• SG A, Band C NR Level= 14% and slowly lowering

• CNMT pressure = 5.4 psig and lowering In accordance with ECA-2.1, based on the above conditions:

1) The MINIMUM feed flow that must be maintained to the Steam Generators is (1)

2) The MAXIMUM cooldown rate allowed is (2)

A. 1) 50 gpm each

2) <100 °F/hr B. 1) 50 gpm each

2) <25 °F/hr C. 1) 700 gpm total

2) <100 °F/hr D. 1) 700 gpm total

2) <25 °F/hr Answer: A Explanation/Justification: K/A is met by requiring the candidate to understand the importance of monitoring and operating the plant to limit feed flow when all SGs are faulted. This is required in order to minimize cooldown rate if necessary, prevent overfilling the SGs, control RCS temperature when cooldown stops, and prevent SG tube dyrout. Based on these primary plant behaviors during an event in which all SGs are faulted, the candidate must have an understanding of how to monitor and operate the plant to limit the effects of thermal shock to the SG components.

A. CORRECT. Per ECA-2.1, if SG NR level < 12% (31 % adverse], maintain a minimum of 50 gpm to each SG. With all 3 SGs blowing down in containment, adverse conditions do exist (5.4 psig). Cooldown rate is limited to <100 F/hr.

B. INCORRECT. Per ECA-2.1, if SG NR level< 12% (31% adverse], maintain a minimum of 50 gpm to each SG. Adverse conditions do exist (5.4 psig). 25 F/Hr is incorrect. Plausible distractor because natural circ cooldown is limited to 25F/hr.

C. INCORRECT. 700 gpm total is incorrect, but a plausible distractor because when adverse and <31% SG level, it is the required flow to removed heat generated when in FR-S.1. Cooldown rate is limited to 100 F/hr.

D. INCORRECT. 700 gpm total is incorrect, but a plausible distractor because when adverse and <31% SG level, it is the required flow to removed heat generated when in FR-S.1. 25F/hr cooldown is incorrect. Plausible distractor because natural circ cooldown is limited to 25F/hr.

Sys # System Category KA Statement W/E12 Uncontrolled EA 1. Ability to operate and I or monitor the following as they Operating behavior characteristics of the facility.

Depressurization of all apply to the (Uncontrolled Depressurization of all Steam Steam Generators 14 Generators)

KIA# EA1.2 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ECA-2.1 lss.2 Rev. 0 Question Source: Bank- 2010 Surry NRC Exam (Q18) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

19. A Reactor Trip occurred from 60% power.

25 minutes post-trip the following conditions exist

• N-35 indicates 4 x 10-10 amps, SUR of O DPM

• N-36 indicates 1x10-11 amps, SUR of 0 DPM Which of the following describes the current conditions?

A. N-35 is undercompensated.

N-31 and N-32 must be manually energized.

B. N-35 is undercompensated.

N-31 and N-32 energized automatically.

C. N-35 is overcompensated.

N-31 and N-32 must be manually energized.

D. N-35 is overcompensated.

N-31 and N-32 energized automatically.

Answer: A Explanation/Justification: KJA is met by knowledge required of the effects of compensating voltage on the Intermediate range detectors, and how an undercompensated IR detector will provide inaccurate indications and prevent the source range detectors from automatically energizing.

A. Correct. With N-35 indicating 4 x E-10 25 minutes after the trip, it is undercompensated. Since it did not reach the P-6 setpoint of 1xE-10, SR detectors will not auto energize since the logic is 2/2 below P-6. N-31 and N-32 will have to be manually energized.

B. Incorrect. With N-35 indicating 4 x E-1 O 25 minutes after the trip, it is undercompensated. It is incorrect that SR will auto energize because 2/2 P-6 logic has not been met.

C. Incorrect. If N-35 was overcompensated, it would go below P-6 before N-35, and the 2/2 logic below P-6 would be met. It is correct that N-31 and N-32 will have to be manually energized since the 2/2 P-6 logic has not been met.

D. Incorrect. If N-35 was overcompensated, it would go below P-6 before N-35, and the 2/2 logic below P-6 would be met. It is incorrect that SR will auto energize because the 2/2 P-6 logic has not been met.

Sys# System Category KA Statement 000033 Loss of AK1. Knowledge of the operational implications of the following Effects of voltage changes on performance Intermediate concepts as they apply to Loss of Intermediate Range Nuclear Range Instrumentation:

Nuclear lnstrumentati on/?

KIA# AK1.01 KIA Importance 2.7 Exam Level RO References provided to Candidate Technical

References:

20M-2.1.B, Rev. 3 pg. 3 & 17 None 20M-1.5.B.2 lss. 4 Rev. 0 Question Source: Bank - Surry 2012 NRC Exam (Q22)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

20. The following plant conditions exist:

• A discharge of the STEAM GENERATOR SLOWDOWN EVAPORATOR TEST TANK

[2SGC-TK238] TO UNIT 2 COOLING TOWER SLOWDOWN is ready to start.

• 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> have passed since sampling was completed (Currently 12-14-15 1400)

• Unit 1 is in Mode 4

• Unit 1 is discharging 1LW-TK-6A, LAUNDRY AND CONTAMINATED SHOWER DRAIN TANK

• Unit 1 COOLING TOWER SLOWDOWN FLOW from RCDR-ENV-MON-1 Ch. 3 is 12000 gpm

• 2CWS-FR101 COOLING TOWER SLOWDOWN FLOW recorder is reading 9400 gpm

• No expected reductions in actual cooling tower blowdown flow exists Refer to attached RWDA-L for 2SGC-TK23B and 20M-25.4L, pages 12-14 Based on the conditions above, which of the following statements are correct?

A. Cooling Tower Slowdown Flow is below the minimum allowed for discharge. Discharge can NOT start.

B. The permit is no longer effective due to exceeding the required time since the sample was taken. Discharge can NOT start.

C. Two tanks are not permitted to be discharged at the same time. Discharge can NOT start.

D. All conditions are satisfactory. Tank discharge is allowed.

Answer: C Explanation/Justification: KIA is met by having the candidate interpret a discharge permit, and determine if a liquid radioactive-waste discharge can commence. The Liquid Waste Release (LWR) permit is designed to prevent an UNCONTROLLED release of radioactive materials to the environment in liquid effluents. The amount of dilution needed is based on the activity of the tank to be released. The dilution includes a limit on the tank release rate and the Cooling Tower Slowdown flowrate.

A. Incorrect. Plausible distractor with 2CWS-FR101 indicating 9400 gpm, which is less than the minimum cooling tower blowdown flow of 10000gpm. By using the open reference pages given, it should be determined that RWDA-L required minimum is based on U1 & U2 CT Slowdown flowrate.

B. Incorrect. Plausible distractor if the candidate does not recognize that the RWDA-L is effective for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after tank sample time (20M-25.4.L, P&L I). The initial conditions gave a 2.5 day period to make it appear excessive.

C. Correct. Only one tank may be discharged at a time from the BVPS Unit1/Unit 2 site (20M-25.4.L P&L C). RWDA-L are based on the proper dilution (Cooling Tower Slowdown Flow) for a particular activity in a tank. Since the site looks at the combined Cooling Tower Slowdown Flow, and if another tank discharge was commenced, dilution would be inadequate and an accidental liquid radwaste release would occur.

D. Incorrect. As stated above, two tanks may not be discharged at the same time.

Sys# System Category KA Statement 000059 Accidental AA2. Ability to determine and interpret the following as they apply to The permit for liquid radioactive-waste release Liquid the Accidental Liquid Radwaste Release:

Radwaste Release I 9 KIA# AA2.02 KIA Importance 2.9 Exam Level RO References provided to Candidate RWDA-L for 2SGC-TK23B Technical

References:

20M-25.4.L Rev. 31 20M-25.4.L, pgs. 12-14 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

21. The crew is performing 20M-56C, "Alternate Safe Shutdown from Outside the Control Room" due to a fire in the Control Room.

What is the reason for closing 21C SG AFW Throttle Viv [2FWE*HCV100A] within 40 minutes of Auxiliary Feedwater actuation?

A. To prevent overfill of the 'C' Steam Generator.

B. To prevent runout of the running AFW pump.

C. To ensure adequate AFW for 'A' and 'B' SGs.

D. To ensure RCS cooldown rate is within limits.

Answer: A Explanation/Justification: KIA is met by the required knowledge of the reason for isolating AFW to the 'C' SG when a fire which could degrade control of the plant from the Control Room occurs. In this case, the crew must trip the reactor and achieve cold shutdown within 72 hrs. from outside the control room using 20M-56C, "Alternate Safe Shutdown from Outside the Control Room".

A. Correct. Preventing 'C' SG overfill is a time critical action as stated in 20M-56C. The concern is also stated multiple places in 20M-56C.4.C & D (NCO & NO procedures). De-energizing the DF bus removes the 'B' AFW from service, and closing HCV1 OOA stops all AFW to 'C' SG.

B. Incorrect. Plausible distractor because the 20M-56C procedures only counts on the 'A' MDAFW and the TDAFW pumps, and they may feel that feeding three SGs may cause the pumps to runout. Incorrect because the steaming rate during C/D is within the AFW pump capabilities.

C. Incorrect. Plausible distractor since it may be thought that water could be limited since TK-210 has approx. 130000 gals of water to feed the SGs during the 72 hr. C/D. Incorrect because AFW pumps do have backup supplies available.

D. Incorrect. Plausible distractor because they may feel that the limited CID rate of 25F/Hr may be exceeded if three SGs are fed at once.

Sys# System Category KA Statement 000067 Plant fire on AK3. Knowledge of the reasons for the following responses as they Actions contained in EOP for plant fire on site site/ 9 apply to the Plant Fire on Site:

KIA# AK3.04 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-56C.4.B Rev. 33 pg. 3 20M-56C.4.C Rev. 20 pg. 4 20M-56C.4.D Rev. 24 pg. 3 Question Source: Bank - Vision #254205 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5,41.10 / 45.6 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

22. The plant is at 83% power with all systems in normal alignment for this power level.

• A serious fire breaks out in the office located behind the Shift Manager desk

• Flames erupt from the office, and the Control Room begins to fill with smoke

• Shift manager directs entry into 20M-56C, "Alternate Safe Shutdown From Outside Control Room Operating Procedures" In accordance with 20M-56C.4.C, "NCO Procedure", the Reactor Operator is expected to trip the Reactor from the ~~~~~~~~~

A. Main Control Room B. Alternate Safe Shutdown Panel C. Emergency Shutdown Panel D. Reactor Trip breakers or MG Set Breakers Answer: A Explanation/Justification: KIA is met by the required knowledge of the Reactor Operator responsibilities during performance of 20M-56C.4.C which include manually tripping the Rx from the Control Room prior to evacuating during a fire.

A. Correct. Step 1 of the NCO procedure is manually trip the Rx. The procedure does not have the RO evacuate the CR until 7 steps later. It is assumed in 20M-56C.3.A that an automatic or manual trip will put the plant in Mode 3.

B. Incorrect. Plausible because the procedure entered is called Alternate Safe Shutdown. No means to trip the Rx exist on the ASP.

C. Incorrect. Plausible because it is a panel located outside the CR, used to shutdown the plant. Knowledge that the purpose of the 56C procedures is to perform a safe shutdown from outside the CR without the Emergency Shutdown panel. No means to trip the Rx exists on the ESP.

D. Incorrect. Plausible answer because locally tripping the reactor is expected if a failure from the Control Room occurs. In this case, the first step of the NCO procedure is to manually trip the Rx.

Sys# System Category KA Statement 000068 Control Room AK2. Knowledge of the interrelations between the Control Room Reactor trip system Evacuation I Evacuation and the following:

8 KIA# AK2.02 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-56C.4.C Rev. 20 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

23. The following conditions exist:

• E-1, "Loss of Reactor or Secondary Coolant" is in progress

• Pressurizer Level is 0%

• RVLIS Full Range Level is 45% and steady

• Containment pressure peaked at 46 psig, and is currently 10.5 psig and slowly lowering

• RWST level is 440 inches and lowering

• Containment sump level is 191 inches and rising Based on the above conditions, which of the following is the highest priority procedure transition required by the crew, and what is the reason for the transition?

A. ES-1.3, "Transfer to Cold Leg Recirculation" because Safety Injection is aligned for recirculation to provide for long term cooling B. FR-Z.1, "Response to High Containment Pressure" because containment integrity is challenged due to peak CNMT pressure C. FR-Z.2, "Response to Containment Flooding" because containment integrity is challenged due to high CNMT sump level D. FR-1.2, "Response to Low Pressurizer Level" because PRZR level has lowered below 14%

Answer: C Explanation/Justification: KIA is met by the knowledge required to determine that entry into FR-Z.2, CNMT Flooding is required to prevent a loss of containment integrity due to high sump level. As the water level rises, it might threaten the availability of equipment required for long-term cooling of the core and/or containment. Such a high water level is considered a severe challenge to the cnmt barrier and restoration is FR-Z.2.

A. Incorrect. ES-1.3 is a plausible distractor if the candidate doesn't know the RWST setpoint of <430 inches, stem states level is 440 inches ..

B. Incorrect. With CNMT pressure <11 psig, there are no entry conditions met into FR-Z.1. Plausible distractor with the peak pressure of 46 psig, the candidate must differentiate between peak and current pressure when making the status tree decision. The candidate must know entry conditions to FR-Z.1 Red or Orange path.

C. Correct. FR-Z.2 Orange path for Containment Flooding is the correct procedure based on Status Tree F-0.5 with CNMT pressure <11 psig and CNMT sump level >187 inches. The reason for using FR-Z.2 is to identify and isolate water sources to minimize sump level which could challenge cnmt integrity.

D. Incorrect. With a PRZR level of 0% and RVLIS level at 45%, this is a plausible distractor with for Inventory (with RVLIS) Status Tree paths.

Sys # System Category KA Statement 000069 Loss of Containment AK3. Knowledge of the reasons for the following responses Guidance contained in EOP for loss of Integrity I 5 as they apply to the Loss of Containment Integrity: containment integrity KIA# AK3.01 KIA Importance 3.8* Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.F-0.5, lss. 2 Rev. 0 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 I 45.6 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

24. Plant was at 100% power when the Rx Tripped due to a Faulted SG.

• The Steam Generator has blown dry

• The crew has transitioned to ES-1.1, SI TERMINATION

• The crew has just secured "B" Charging Pump

• "A" Charging Pump is running

• RCS pressure is rising

• PRZR level is rising

• RWST level is lowering Which of the following describes the actions that will be taken next per ES-1.1, SI TERMINATION, and why should a reduction in SI flow be done expeditiously?

A. 1) Establish letdown flow.

2) To preserve RWST inventory.

B. 1) Establish normal charging flow.

2) To preserve RWST inventory.

C. 1) Establish letdown flow.

2) To prevent the pressurizer from going solid.

D. 1) Establish normal charging flow.

2) To prevent the pressurizer from going solid.

Answer: D Explanation/Justification: KJA is met by the candidate assessing the CR indications, and with knowledge of the SI termination procedure, determine that normal charging must be established prior to letdown. The candidate must also demonstrate knowledge why SI should be terminated quickly as to not take the PRZR solid.

A. Incorrect. Establishing letdown flow is plausible if candidate thinks that securing one charging pump (HHSI) pump in SI term has returned to a normal charging flowpath and restoring UD would be the next logical step to perform. Securing SI expeditiously is not based on RWST depletion, but it is plausible with RWST level lowering.

B. Incorrect. Establish normal charging flow is correct. Securing SI expeditiously is not based on RWST depletion, but it is plausible with RWST level lowering.

C. Incorrect. Establishing letdown flow is plausible if candidate thinks that securing one charging pump (HHSI) pump in SI term has returned to a normal charging flowpath and restoring UD would be the next logical step to perform. Preventing the pressurizer from going solid is the correct reason for securing SI expeditiously.

D. Correct. After isolating HHSI flow, establishing normal charging flow is correct. Restoring UD is identified later in ES-1.1, but it would be incorrect to establish UD without having charging flow restored to prevent flashing downstream of the UD orifices. Preventing the pressurizer from going solid is the correct reason for securing SI expeditiously.

Sys# System Category KA Statement W/E02 SI Termination I Generic Ability to interpret control room indications to verify the status and operation of a system, and 3 understand how operator actions and directives affect plant and system conditions.

KIA# 2.2.44 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.1 lss.2 Rev. 0 pg. 9 20M-538.4.E-O lss. 2 Rev. 1 pg. 12 20M-7.2.A Rev. 17 pg. 3 Question Source: Bank- VC Summer 2011 NRC Exam- Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.12))

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

25. The following conditions exist:

• A LOCA has occurred

• The crew is performing ES-1.2, "Post LOCA Cooldown and Depressurization"

• RCS cooldown to cold shutdown is in progress

• All RCPs are shutdown

• The crew is reducing RCS pressure to refill the pressurizer Which of the following would indicate to the crew that voiding in the RCS is occurring?

A. 2RCS-Ll460, PRZR CHANNEL 2 LEVEL, rapidly increasing.

B. 2RCS-Pl402, RX CLNT SYSTEM WIDE RNG PRESS, rapidly increasing.

C. 2SIS*Fl943, HHSI TRN B, rapidly decreasing.

D. UPS011, PSMS Average lncore TIC Temp, rapidly decreasing.

Answer: A Explanation/Justification: KIA is met by the required knowledge that the Reactor Operator must understand the implications of depressurizing the RCS when directed by ES-1.2, "Post LOCA Cooldown and Depressurization", and recognize available Control Room indications which indicate a void in the vessel head is occurring.

A. Correct - Voiding causes water to be displaced in the RCS which shows up as an increase in pressurizer level.

B. Incorrect - Increasing RCS pressure would suppress voiding in the RCS. Plausible for the same reason as C.

C. Incorrect - Decreasing SI flow would be indicative of a pressure increase which is inconsistent with voiding in the RCS. Plausible because candidate may think that the expansion of a void bubble and PRZR level rising would cause pressure to rise, thereby reducing HHSI flow.

D. Incorrect - Plausible because the candidate may think that during void formation, there will be less heat transfer and temperature will go down rapidly. Due to saturation conditions during void formation, temperature should stay approximately the same.

Sys # System Category KA Statement W/E03 LOCA Cooldown and EK1. Knowledge of the operational implications of the Normal, abnormal and emergency operating Depressurization I 4 following concepts as they apply to the (LOCA Cooldown and procedures associated with (LOCA Cooldown and Depressurization) Depressurization).

KJA# EK1 .2 KJA Importance 3.6 Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.ES-1.2 lss. 2 Rev. 1 pg.11 None Question Source: Bank - Farley 2011 NRC Exam (070)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

26. While performing actions of FR-C.3, "Response to Saturated Core Cooling", which of the following identifies the condition that the PRZR PORVs and block valves are required to be in?

(Assume no previous PRZR PORV failures)

A. Three PORVs closed with all block valves closed to minimize RCS leakage.

B. Two PORVs closed and one PORV open with associated block valve open for RCS pressure control.

C. Three PORVs closed with at LEAST one block valve open for RCS pressure control.

D. One PORV closed with two PORVs open to depressurize the RCS to facilitate SI Accumulator Injection.

Answer: C Explanation/Justification: KJA met by knowledge of FR-C.3, "Response to Saturated Core Cooling" major action step to check for open RCS vent paths, and the ability to monitor the PORVs and Block valves in the required system configuration.

A. Incorrect. Plausible to prevent RCS inventory loss since core cooling is degraded already, but not IAW FR-C.3.

B. Incorrect. Plausible if they assumed that the PORV was being used to lower RCS pressure as is a mitigative strategy in FR-C.1.

C. Correct. The candidate must be familiar with the basic purpose, overall sequence of events or overall mitigative strategy of Saturated Core Cooling. With knowledge of the FR-C procedures. the RO demonstrates the ability to operate the plant and obtain desired operating results during these emergency plant conditions. The major action categories for FR-C.3 is to establish SI flow to maintain minimum RCS subcooling, and check for open RCS vent paths.

D. Incorrect. This action could be performed in FR-C.1 to depressurize RCS but not performed for yellow condition FR-C.3.

Sys# System Category KA Statement W/E07 Saturated EA 1. Ability to operate and I or monitor the following as they apply to Components. and functions of control and safety Core Cooling the (Saturated Core Cooling) systems, including instrumentation, signals, 14 interlocks, failure modes. and automatic and manual features.

KIA# EA1.1 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-C.3 lss. 2 Rev. 0 Question Source: Bank - Ginna 2011 NRC Exam (062) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

27. Given the following plant conditions:

• 'A' SIG Pressure is 1150 psig

• 'A' SIG Narrow range level is 82%

• RCS hot leg temperatures are 563 °F

• FR-H.2, "Response to Steam Generator Overpressure" has been entered

• All MSIVs are CLOSED

• 2SVS*PCV101A, 21A SG ATM STM DUMP has failed CLOSED

• 2SVS*HCV104 Residual Heat Release Viv is CLOSED and will not open

• 2FWE*P22, Turbine Driven AFW pump is out of service for bearing replacement Which of the following describes the preferred method to reduce "A" SIG pressure in accordance with FR-H.2?

A. Feed 'A' SG with AFW and commence an RCS cooldown to less than 534°F using 'S' & 'C' Steam Generators.

S. Feed 'A' SG with AFW and establish Slowdown from the 'A' Steam Generator.

C. Isolate AFW to the 'A' SG and commence RCS cooldown to less than 534°F using 'S' & 'C' Steam Generators.

D. Isolate AFW to the 'A' SG and establish Slowdown from the 'A' Steam Generator.

Answer: C Explanation/Justification: KIA met by candidates ability to perform FR-H.2, "Response to Steam Generator Overpressure" major action steps of Controlling the affected SG pressure and initiate CID using the unaffected SGs. In the stem of the question steam dumps are unavailable for 'A' SG, therefore candidate must recognize that cooldown is required on the unaffected SGs.

A. Incorrect. Feeding the A SG is not permitted (or procedurally driven) because feed may be the cause of the overpressure. Cooling the RCS using the unaffected SGs is the correct answer if steam cannot be dumped from the affected SG.

B. Incorrect. Feeding the A SG is not permitted (or procedurally driven) because feed may be the cause of the overpressure. Establishing blowdown from A SG is not procedurally driven.

C. Correct. Major action steps of FR-H.2 are Control the affected SG pressure and initiate CID using the unaffected SGs. Given the initial conditions in the stem it is determined that there are no means to control pressure in the affected SG, therefore it will be necessary to isolate AFW to the 'A' SG and cooldown using the other SG by dumping steam using B and/or C ADV.

D. Incorrect. Isolating AFW is correct, but establishing blowdown from A SG is not procedurally driven.

Sys# System Category KA Statement W/E13 Steam Generator Generic Ability to perform specific system and integrated Overpressure I 4 plant procedures during all modes of plant operation.

KIA# 2.1.23 KIA Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-H.2 lss. 2 Rev.a Question Source: Bank - Surry 2012 NRC (024)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.2 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

28. During a plant startup, the following conditions exist:

• Reactor is at 13% power

• Offsite power is supplying all 4KV busses

• All 3 Reactor Coolant Pumps (RCPs) are running The "2B STA SERVICE FEEDER 138KV BREAKER PCB-94" spuriously tripped open.

Which RCP(s) will lose power, and will the Reactor automatically trip due to this loss of the RCP(s)?

(1) will lose power.

The reactor _ _ _(_2).__ _ automatically trip due to this loss of the RCP(s).

A. 1) Only 'C' RCP

2) will B. 1) Only 'C' RCP

2) will NOT C. 1) Both 'A' and 'B' RCPs

2) will D. 1) Both 'A' and 'B' RCPs

2) will NOT Answer: B Explanation/Justification: KIA is met by the candidates ability to analyze the loss of power to the system service transformer 2B, and the loss of power to only 'C' RCP.

A. Incorrect. It is correct that 'C' RCP will lose power. It is incorrect that the Rx will trip. Plausible distractor because the student needs to know that a loss of 1 RCP when power is <P-8 (30%) will not trip the Rx. This coincidence is easy to confuse because a loss of 2/3 RCPs >P-7 (10%) will trip the Rx.

B. Correct. 'C' RCP is powered from 4160 Bus 2C, which is supplied from the 2B System Service Transformer. When PCB-94 opens SSST 2B is de-energized, which de-energizes 4160 bus 2C & 20. There are no RCPs on bus 20. It is correct that the Rx will not trip on the loss of 1 RCP since the plant was <PS (30% power).

C. Incorrect. 'A' & 'B' RCPs are powered from SSST 2A, which is not effected by PCB-94 opening. Plausible distractor because the candidate must know which off-site feed supplies the SSSTs and ultimately the RCP busses. It is incorrect that the Rx will trip. Plausible distractor because the student needs to know that a loss of 1 RCP when power is <P-8 (30%) will not trip the Rx. This coincidence is easy to confuse because a loss of 2/3 RCPs >P-7 (10%) will trip the Rx.

D. Incorrect. 'A' & 'B' RCPs are powered from SSST 2A, which is not effected by PCB-94 opening. It is correct that the Rx will not trip.

Sys # System Category KA Statement 003 Reactor Coolant Pump K2 Knowledge of bus power supplies to the following: RCPS System (RCPS)

KIA# K2.01 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

20M-1.5.B.1 Rev. 2 pg. 2 RE-0001 DH Rev. 4 & RE-0001 E Rev. 9 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

29. The plant is shutting down for a refueling outage in accordance with 20M-52.4.R.1 Shutdown From 100% Power To Mode 5".

• Plant is at 30% power

• All systems in normal alignment for this power level Chemistry has requested a purge of the VCT to remove non-cond In accordance with 20M-7.4.F, "Degassing the Reactor Co nt System From The Volume Control Tank", which of the following completes the state ents below?

At this power level, _ ___._(1"""')_ __ r purging the RCS of non-condensable gasses.

The non-condensable gasses from the V.. will be purged to the _ __....(2.....)_ __

A. 1) Hydrogen

2) Primary Plant Sample S B. 1) 2)

C. 1)

2) Plant Sample System D. itrogen Boron Recovery System Answer: D Explanation/Justification: KIA is met with the knowledge of nitrogen gas being used in conjunction with raising and lowering VCT level to purge the Hydrogen and non-condensable gasses from the VCT (CVCS) during RCS degassing when performing a plant shutdown.

Reducing gas concentration at BVPS is accomplished by reducing Hydrogen and non-condensable gasses from the RCS via the VCT, and venting the PRZR to the sample system. Although the KIA statement states from the przr bubble space, it would not be possible to meet the KIA due to purging the gasses from the przr to the sample system bypasses the CVCS system which is the KIA required system tie. Purging the PRZR is a Chemistry procedure.

A. Incorrect. Plausible because Hydrogen is the normal cover gas maintained on the VCT during operation. The goal during degas is to reduce Hydrogen and non-condensable gasses. Primary Sample System is a plausible distractor because this is an approved method of continually degasifying the przr, but the stem asked about purging the VCT to remove non-condensable gasses from the RCS.

B. Incorrect. Plausible because Hydrogen is the normal cover gas maintained on the VCT during operation. The goal during degas is to reduce Hydrogen and non-condensable gasses. The gasses will be purged from the VCT to the Degasifiers in the Boron Recovery System.

C. Incorrect. Nitrogen is used as the cover gas when removing Hydrogen and non-condensable gasses from the RCS. Primary Sample System is a plausible distractor because this is an approved method of continually degasifying the przr, but the stem asked about purging the VCT to remove non-condensable gasses from the RCS.

D. Correct. Nitrogen is used as the cover gas when removing Hydrogen and non-condensable gasses from the RCS due to chemistry requirements to reduce Hydrogen and non-condensable gasses prior to an outage. The gasses will be purged from the VCT to the Degasifiers in the Boron Recove S stem.

Sys# System Category KA Statement 004 Chemical and K5 Knowledge of the operational implications of the following Reduction process of gas concentration in RCS: vent Volume concepts as they apply to the eves: accumulated non-condensable gases from PZR Control bubble space, depressurized during cooldown or by System alternately heating and cooling (spray) within allowed pressure band (drive more gas out of solution)

KIA# K5.14 KIA Importance 2.5 Exam Level RO References provided to Candidate Technical

References:

20M-7.4.F, Rev. 19 pg. 3 & 4 None U2 RM-0407-002, RM-0409-002, RM-0408-001 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

30. Which of these components can be supplied power from either the 'A' OR 'B' Train of 480VAC (selectable)?

A. 2RHS*MOV702A, RHS Train A Supply Isolation Valve B. 2RHS*MOV702B, RHS Train B Supply Isolation Valve C. 2RHS*MOV720A, RHS Train Return to B Loop Isolation Valve D. 2RHS*MOV720B, RHS Train Return to C Loop Isolation Valve Answer: A Explanation/Justification: K/A is met by the knowledge required to determine which of the RHR pressure boundary MOVs are supplied by dual power supplies.

A. Correct. 2RHS*MOV702A can be powered by either MCC*2-E05 or E06. All other valves listed below are plausible because they are RCS pressure boundary MOVs, but only have one power supply.

B. Incorrect. 2RHS*MOV702B is only powered from MCC*2-E06.

C. Incorrect. 2RHS*MOV720A is only powered from MCC*2-E05.

D. Incorrect. 2RHS*MOV720B is only powered from MCC*2-E06.

Sys# System Category KA Statement 005 Residual Heat K2 Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves Removal System (RHRS)

K/A# K2.03 K/A Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

20M-10.1.D rev. 0 lss. 4 Question Source: Bank - Vision #240045 Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

31. Given the following conditions:

• The plant was at 80% power when a LOCA occurred

• All ESF equipment operated as designed

• RCS pressure is now 50 psig

• RWST level is 360 inches and lowering Which of the following describes the Safety Injection System alignment for these conditions?

A. LHSI pumps taking a suction from the RWST and discharging to the RCS Cold Legs.

B. HHSI pumps taking a suction from the RWST and discharging to the RCS Cold Legs.

C. LHSI pumps stopped and suction isolated from the RWST.

D. LHSI pumps taking a suction from the Containment Sump and discharging to the HHSI pump suction.

Answer: C Explanation/Justification: KIA is met by the candidates ability to monitor the given plant conditions, and recognize that when RWST level is <369",

the ECCS system will transfer into Cold Leg Recirc mode, at which time the LHSI pumps will automatically trip and the RWST suction valves will close.

KIA statement was changed from RHR pumps to LHSI pumps after discussion with the Chief Examiner based on the fact that at BVPS2 RHR is not an ECCS system. By changing the system name only, the intent of the KIA was preserved.

A. Incorrect. Plausible distractor because this is the lineup prior to the Cold Leg Recirc mode at 369" in RWST. When the transfer occurs, 2SIS-8809A & B close to isolate the RWST, and the LHSI pumps trip.

B. Incorrect. Plausible distractor because this is the lineup prior to the Cold Leg Recirc mode at 369" in RWST. When the transfer occurs, 2CHS-MOV115B & D close to isolate the RWST, and 2SIS-MOVMOV863A & B open to align the suction to the LHSI discharge piping for the RS pump.

C. Correct. When the RWST level lowers to less than 369" on 2/4 channels coincident with a Safety Injection signal, the ECCS system will transfer into Cold Leg Recirc mode. This will cause Recirc Spray pumps C & D to start, and align their discharge to the suction of the HHSI pumps. The LHSI pumps will trip and their suction valves to the RWST will close.

D. Incorrect. Plausible distractor because LHSI pumps are vital pumps and there is a misconception that they take suction from the sump.

Sys# System Category KA Statement 006 Emergency A3 Ability to monitor automatic operation of the ECCS, including: RHR pllmps (LHSI Pumps)

Core Cooling System (ECCS)

KIA# A3.07 KIA Importance 3.6* Exam Level RO References provided to Candidate None Technical

References:

20M-11.1.D rev.1 20M-11.2.B rev. 5 Question Source: Bank - Vision #123771 Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

32. The plant is at 100% power.

• 'B' Charging pump is RUNNING

• 'A' Charging pump is in STBY

• Normal 480 VAC MCC-2-13 Cub 80 tripped on overcurrent causing the following:

o 2CHS-SOV150A, 'CHG PP 21A Lube Oil Temp Solenoid VLV' is de-energized o 2CHS-P21A-1, 'Charging Pump Auxiliary Lube Oil Pump' is de-energized How will 2CHS-TCV150A, 'CHG PP 21A Lube Oil Temp Control Valve' respond to this failure, and will the 'A' Charging pump start without the Aux Oil Pump running?

1) 2CHS-TCV150A will direct all oil flow _ __....(1_)___ the lube oil cooler.

2) The 'A' Charging pump _ ___._(2~)___ start without the Aux Oil Pump running.

A. 1) to bypass

2) will B. 1) to bypass

2) will NOT C. 1) through

2) will D. 1) through

2) will NOT Answer: C Explanation/Justification: KIA is met by the knowledge demonstrated of the design feature of the centrifugal charging (HHSI) pumps lube oil TCV to divert all oil flow through the LO cooler to provide max cooling to the pump bearings on a loss of air.

A. Incorrect. TCV150A will direct all oil through the cooler to maximize cooling. It is correct the charging pump will start without the AOP running.

B. Incorrect. TCV150A will direct all oil through the cooler to maximize cooling. It is incorrect to state that the Charging pump will not start.

Plausible distractor because the design feature to start the Aux Oil Pump at 14 psig, and stop the AOP when the shaft driven pump raises pressure to normal. This feature could lead to thinking that the AOP must be running to start the Charging Pump.

C. Correct. When 2CHS-SOV150A is de-energized, it vents air from 2CHS-TCV150A causing it to divert all flow through the lube oil cooler for maximum cooling to bearings and gears. The Aux Oil Pump has a design feature to provide lubrication and cooling to the charging pump, but it is not required to be running in order to start the Charging Pump.

D. Incorrect. It is correct that TCV150A will divert all flow through the lube oil cooler. It is incorrect to state that the Charging pump will not start.

Plausible distractor because the design feature to start the Aux Oil Pump at 14 psig, and stop the AOP when the shaft driven pump raises pressure to normal. This feature could lead to thinking that the AOP must be running to start the Charging Pump.

Sys # System Category KA Statement 006 Emergency Core Cooling K4 Knowledge of ECCS design feature(s) and/or Cooling of centrifugal pump bearings System (ECCS) interlock(s) which provide for the following:

KIA# K4.01 KIA Importance 2.6 Exam Level RO References provided to Candidate Technical

References:

20M-7.3.C Rev. 15 pg. 2 None U2 TLD-007-096-01 & 02 Rev. 5 U2 LSK-026-001G Rev. 13 & 001A Rev. 14 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

33. Annunciator A4-3H, Pressurizer Relief Tank (PRT) Trouble has alarmed.

• The operator reports it is due to HIGH PRT temperature

• Level and pressure are within the normal range

• PRZR Safety valves and PORVs are closed In accordance with 20M-6.4.AAY, PRT Trouble ARP, which of the following choices identifies ALL of the valves listed below that would be opened to reduce PRT temperature?

1. PRT Pri Grade M/U Wtr valves [2RCS-MOV516 and 2RCS-AOV519]

2. PRT Vent Viv [2RCS-MOV549]

3. PRT Drain Viv [2RCS-MOV523]

A. 1 only B. 1 and 2 only C. 1 and 3 only D. 1 and 2 and 3 Answer: A Explanation/Justification: K/A is met by the required knowledge of the Pressurizer Relief Tank (PRT) system design of the ability to spray down the PRT with primary water to cool the tank.

A. Correct. With high PRT temperature the ARP states to open 2RCS-MOV516 and 2RCS-MOV519 to spray the tank internally and reduce temperature.

B. Incorrect. Plausible distractor because sect. B of the ARP for reducing tank pressure requires spraying the PRT to reduce pressure, then if pressure does not reduce <8psig, vent the tank by opening 2RCS-MOV549. Stem states level and pressure are within the normal range.

C. Incorrect. Plausible distractor if the candidate thinks you must lower PRT level prior to spraying it down to reduce temperature. The stem states level and pressure are within the normal range.

D. Incorrect. Plausible distractor to open all 3 valves if it is thought that a flushing of the PRT must occur to reduce temperature. This could work, but is not in accordance with the ARP.

Sys# System Category KA Statement 007 Pressurizer K4 Knowledge of PRTS design feature(s) and/or interlock(s) which Quench tank cooling Relief provide for the following:

Tank/Quench Tank System (PRTS)

KIA# K4.01 K/A Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.AAY Rev. 10 Question Source: Bank- Diablo Canyon 2012 NRC Exam (06) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

34. RCS temperature is 495°F with a cooldown in progress to comply with the LCO action requirements (LCO 3.7.7 Cond B) for an inoperable 'A' CCP train. Then, an unisolable leak on the 'B' supply header renders the 'B' CCP train inoperable.

As the crew continues the cooldown, which one of the following temperatures is the MINIMUM RCS temperature that is allowed by Technical Specifications for the given conditions?

A. 350° F B. 201° F C. 200° F D. 137°F Answer: B Explanation/Justification: KIA is met by the 1 hr. and less LCO conditions associated with TS.3.7.7 Component Cooling Water System (CCW), and the requirement of not entering mode 5 without Primary Component Colling water available for RHR to be placed in service.

A. Incorrect. This is the minimum temperature of Mode 3. 2 RCS loops are still required to operable. Procedurally RHS is not available to use until S350F.

B. Correct. This is the minimum temperature for Mode 4. Entry into Mode 5 is not permissible with both trains of CCP inoperable as stated by the note in TS 3.7.7 cond. C (immediate completion time).

C. Incorrect. This is maximum temperature of Mode 5 entry. Entry into Mode 5 is not permissible with both trains of CCP inoperable as stated by the note in TS 3.7.7 cond. C (immediate completion time). Without CCP, RHS is inoperable (TS 3.4.7) and 1 RHS loop is required to be in operation in Mode 5.

D. Incorrect. This is the minimum temperature to operate 3 RCPs. Plausible distractor if the candidate thinks that the RCS loops must remain operable when RHS is not operable due to both Trains of CCP inoperable and Mode 5 is required.

Sys# System Category KA Statement 008 Component Cooling Generic Knowledge of conditions and limitations in the Water System (CCWS) facility license KIA# 2.2.38 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

TS. 3.7.7 A278/161 TS. Table 1.1-1 Def. of modes Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.1 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

35. Given the following conditions:

• The plant is at 100% power

• A rupture of the 21A RCP Thermal Barrier occurs

• No annunciators are in alarm Which of the following completes the statements below?

1) 2CCP*AOV107A, 21A RCP THERMAL BARRIER OUTLET ISOL VLVwill automatically close at (1)

2) In accordance with AOP-2.6.8, "Abnormal RCP Operation", a shutdown of the 'A' RCP (2) required for this failure.

(1) (2)

A. 122 psig is NOT B. 122 psig is C. 50 gpm is NOT D. 50gpm is Answer: A Explanation/Justification: KIA is met by demonstrating the knowledge that RCPs can still operate with a loss of CCP thermal barrier flow as long as RCP seal injection flow is available.

A. Correct. 2CCP*AOV107's auto close at 122 psig and/or 58 gpm. The AOP does not require a shutdown of the RCP as long as there is still Seal Injection. The stem of the question does not state any problems which would lead to seal injection failure and there are no annunciators in alarm.

B. Incorrect. This is the correct pressure which auto closes 2CCP*AOV107's. It is incorrect that the AOP requires a RCP shutdown. There are no indications of a loss of seal injection.

C. Incorrect. 50 Gpm is less than the setpoint of 58 gpm required to auto close 2CCP*AOV107's. It is correct that a shutdown of the RCP is not required.

D. Incorrect. 50 Gpm is less than the setpoint of 58 gpm required to auto close 2CCP*AOV107's. It is incorrect that the AOP requires a RCP shutdown. There are no indications of a loss of seal injection.

Sys# System Category KA Statement 008 Component K3 Knowledge of the effect that a loss or malfunction of the CCWS RCP Cooling will have on the following:

Water System (CCWS)

KIA# K3.03 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-15.2.B Rev. 1pg.4 20M-53C.4.2.6.8 Rev. 12 pg 1 & 2 Question Source: Bank - Farley 2012 NRC Exam (Q14)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: None Listed Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

36. The plant is at 100% power.

• PRZR Channel 1 Press 2RCS-PT455 has failed HIGH

• The control room crew has tripped all associated bistables IAW 20M-6.4.IF, "Instrument Failure Procedure" PRZR Control Pressure [2RCS-PT445] THEN fails HIGH.

What will be the INITIAL plant response to this additional failure?

A. PRZR Spray Valve 2RCS*PCV455A & 2RCS*PCV4558 will OPEN.

B. PRZR PORV 2RCS-PCV455C will OPEN.

C. PRZR PORVs 2RCS-PCV455D & 2RCS-PCV456 will OPEN.

D. High PRZR Pressure Reactor Trip will ACTUATE.

Answer: C Explanation/Justification: KIA is met by demonstrating the knowledge of how PRZR pressure will respond to a pressure control channel failing high, and the knowledge that 2 PORVS (PCV455D & PCV456) will automatically open.

A. Incorrect. This would be the INITIAL response if 2RCS-PT444 failed High.

8. Incorrect. This would be the next response if 2RCS-PT444 failed High.

C. Correct. IAW 20M-6.4.IF attachment 2.

D. Incorrect. Failures are one control channel and one protection channel, therefore NO reactor trip.

Sys# System Category KA Statement 010 Pressurizer Pressure A3 Ability to monitor automatic operation of the PZR PZR pressure Control System (PZR PCS, including:

PCS)

KIA# A3.02 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF attachment 2 Rev. 13 Question Source: Bank - 2LOT6 NRC Exam (036) (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

37. Given the following conditions:

• A reactor startup is in progress

• The reactor is critical in the source range

• N42 Power Range channel has failed and has been removed from service with all bistables placed in the trip condition

• A loss of Vital Bus 1 occurs

• A 12-1 H, NOT P-7 changed state after the loss of Vital Bus occurred What is the condition of the reactor, and source range detectors after Vital Bus 1 is lost?

A. Reactor trips N31 Source Range channel is de-energized.

N32 Source Range channel is still in operation.

B. Reactor remains critical BOTH source range channels are de-energized.

C. Reactor remains critical N31 Source Range channel is de-energized.

N32 Source Range channel is still in operation.

D. Reactor trips BOTH source range channels are de-energized.

Answer: D Explanation/Justification: KIA is met by requiring knowledge of a loss of a 2"d (redundant) PR NI channel due to the loss of vital bus 1, and the effects it has on both the Rx Protection System causing a Rx trip and de-energizing both SR channels.

A. Incorrect. It is correct that the Rx will trip due to 214 PR high setpoints. N31 is de-energized by the loss of vital bus 1. N32 will not be in operation due to P-10 auto de-energizing both SR detectors B. Incorrect. Reactor trips on a number of PR/SR trip setpoints. It is correct that both SR detectors will be de-energized C. Incorrect. Reactor trips on a number of PR/SR trip setpoints. N31 is de-energized by the loss of vital bus 1. N32 will not be in operation due to P-10 auto de-energizing both SR detector.

D. Correct. A loss of Vital 1 causes a loss of power to N41. This loss also causes a loss of power to RPS channel 1. This will cause a trip condition for Power range trips for channel 1. Since N42 is already removed from service its bistable are in the tripped condition. This meets the 2/4 logic to cause a reactor trip. N31 is de-energized by the loss of vital bus 1. Additionally the signal for 214 power range channels above P-10 will cause the SR channels to auto de-energize causing N32 to de-energize.

Sys# System Category KA Statement 012 Reactor Protection K6 Knowledge of the effect of a loss or malfunction of the Redundant channels System (RPS) following will have on the RPS:

KIA# K6.02 KIA Importance 2.9 Exam Level RO References provided to Candidate Technical

References:

20M-2.2.A Rev. 1 (P&L 9)

None UFSAR Fig. 7.3-8 and 7.3-9 20M-2.3.C Rev. 5 pg. 3 Question Source: Bank - DC Cook 2010 NRC Exam (039)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 45/7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

38. Given the following conditions:

• The plant is at 100% power

• Containment Pressure Channel IV pressure indication was oscillating and has been removed from service IAW 20M-1.4.IF, "Instrument Failure Procedure" Which of the following identifies the logic associated with the HIGH 1 and HIGH 3 Containment Pressure actuations after the Channel IV is removed from service?

HIGH 1 SI Actuation HIGH 3 CIB Actuation A. 1/2 2/3 B. 1/2 1/3 C. 1/3 2/3 D. 1/3 1/3 Answer: A Explanation/Justification: K/A is met by demonstrating knowledge of the containment pressure channel inputs to both Safety injection and Cnmt Isol phase B actuation logics, and how these inputs are removed from service for the reliability of the actuation coincidence.

A. Correct. Channel IV (2LMS-PT953) was removed from service iaw 20M-1.4.IF. This procedure and Tech Specs has the bistable tripped for High 1 (SI) which then makes the logic 1/2. The bistable for the High 3 (CIB) is required to be placed in bypass, which then requires a 2/3 coincidence to initiate a CIB. Both of these bistable configurations satisfies redundancy requirements.

B. Incorrect. Plausible if both bistables are tripped. High-1 is normally a 2/3 logic but changes to 1/2 when one of the logic channels are tripped. For High-3, 2/4 logic, the channel is bypassed, so 2/3 is required.

C. Incorrect. Plausible if the logic for both are 214 and only High-1 is tripped.

D. Incorrect. Plausible if the logic for both are 214 and both are tripped.

Sys# System Category KA Statement 013 Engineered K5 Knowledge of the operational implications of the following Safety system logic and reliability Safety concepts as they apply to the ESFAS:

Features Actuation System (ESFAS)

KIA# K5.02 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-1.4.IF Att. 1 Rev. 9 BVPS TS Bases pg. B3.3.2-38 & 39 Question Source: Bank - Harris 2012 NRC Exam (040)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

39. The plant is at 100% power when a rupture of the Chilled Water supply header to Containment occurs. The crew has isolated the rupture.

1) Which of the following components will be effected by the loss of Chilled Water?

2) What is the design backup for this system?

A. 1) Control Rod Drive Mechanism (CROM) Fans

2) Primary Component Cooling Water B. 1) Containment Air Recirculation (CAR) Fans

2) Service Water C. 1) Control Rod Drive Mechanism (CROM) Fans

2) Service Water D. 1) Containment Air Recirculation (CAR) Fans

2) Primary Component Cooling Water Answer: B Explanation/Justification: K.JA was met by having the candidate predict which Cnmt cooling equipment will lose cooling capabilities when Chilled Water system is lost, and knowledge of the Service Water system as a design backup.

K.JA statement is a loss of service water. At BVPS Chill Water is the normal system aligned the Containment Air Recirc Fans, and Service Water is the backup supply. The question is written to meet the intent of the K.JA as designed at BV.

A. Incorrect. CROM coolers are supplied by CCP and no backup cooling is available. CCP is the normal cooling for the CROM coolers.

B. Correct. CAR fan coolers are supplied by chilled water, and a loss of chilled water can effect cnmt temperature. Service Water is the emergency backup cooling source for the CAR fan coolers.

C. Incorrect. CROM coolers are supplied by CCP and no backup cooling is available. Service water is the emergency backup for the CAR fan coolers.

D. Incorrect. CAR fan coolers are supplied by chilled water, and a loss of chilled water can effect cnmt temperature. CCP is not the backup for CAR fan coolers. It is the primary cooling for CROM coolers.

Sys# System Category KA Statement 022 Containment A2 Ability to (a) predict the impacts of the following malfunctions or Loss of service water Chilled Water Cooling operations on the CCS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (CCS) malfunctions or operations:

KIA# A2.04 KIA Importance 2.9* Exam Level RO References provided to Candidate Technical

References:

U2 RM-0429-004 Rev. 14 None 2SQS-44C.1 PPNT Rev. 12 pgs. 7 & 8 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 / 45.3 /

45.13)

Objective: 2SQS-44C.1 EL0-1 Describe the function of the Containment Ventilation System and the associated major components as documented in Operating Manual Chapter 20M-44C.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

40. The plant is at 100% power.

Based on the information provided on the attached PCS screen, what is the status of Tech Spec LCOs 3.6.4, Containment Pressure and 3.6.5, Containment Air Temperature?

LCO 3.6.4, Containment Pressure _ __._(1"'"'")_ _ _ met, AND LCO 3.6.5, Containment Air Temperature (2) met.

Refer to attached PCS screen A. (1) is NOT (2) is NOT B. (1) is NOT (2) is C. (1) is (2) is NOT D. (1) is (2) is Answer: D Explanation/Justification: KIA is met by having the candidate analyze a Plant Computer Screen printout as seen in the Control Room, and evaluate CNMT temperatures and Pressures for Tech Spec above the line LCO specs.

A. Incorrect. See correct answer explanation.

B. Incorrect. See correct answer explanation.

C. Incorrect. See correct answer explanation.

D. Correct. Cnmt pressures are both less than 14.2 psia required by Tech Specs. The pressures indicated are not red on the PCS screen to indicate they are greater than TS limits, but to display that they are above the high pressure alarm setpoint of 13.9 psia. Average temperature is 99.1 F which is below the LCO required 108F. One of the temperatures were purposely placed above the 108 °F allowed value to ensure the candidates awareness of the LCO being the average, and not any one temperature. To answer this question the candidate will need to assess the computer data provided and determine the status of the LCOs.

Sys # System Category KA Statement 022 Containment A4 Ability to manually operate and/or monitor in the control room: Containment readings of temperature, pressure, Cooling and humidity system.

System (CCS)

KIA# A4.05 KIA Importance 3.8 Exam Level RO References provided to Candidate CNMT PCS screen shot Technical

References:

TS 3.6.4 & 3.6.5 20M-54.3.L5 Rev. 80 Question Source: Bank - 1LOT14 NRC Exam (054) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

41. Given the following conditions:

• The plant was at 100% power when all 3 Steam Generators Faulted in Containment

• Containment Pressure is 31 psig and RISING

• Quench Spray Pumps [2QSS*P21A and P21 B] failed to start

1) What is(are) the minimum required Engineered Safety Features Actuation System (ESFAS) switch manipulations required to start BOTH Quench Spray Pumps?

A. 1 B. 2 C. 3 D. 4 Answer: B Explanation/Justification: KIA is met by the candidates recognizing that GIB did not actuate and the knowledge of how many ESF actuation switches must be operated to ensure both trains of Quench Spray Pumps start and prevent CNMT pressure from exceeding design pressure.

A. Incorrect. Plausible distractor if candidate thinks Safety Injection ESF actuation will start the Quench Spray pumps (CIB is required for QS).

B. Correct. 2 GIB switches on the same train will actuate GIB on both trains. There are a total of 4 GIB switches on the Bench Board. 2 switches per train.

C. Incorrect. Plausible distractor if candidate thinks BOTH a GIB and Safety Injection ESF actuation is needed to start the Quench Spray pump.

D. Incorrect. Plausible distractor if candidate thinks BOTH trains of CIB are required to start BOTH Quench Spray pumps. There are a total of 4 GIB switches on the Bench Board, 2 for each train, and 1 train will actuate both GIB trains.

Sys# System Category KA Statement 026 Containment A1 Ability to predict and/or monitor changes in parameters (to Containment pressure Spray prevent exceeding design limits) associated with operating the CSS System controls including:

(CSS)

KIA# A1.01 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

USFAR Fig. 7.3-13 Rev. K USFAR Fig. 7.3-62 Rev. K Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

42. RWST Temperature is indicating 43°F. Outside air temperature is 26°F.

In accordance with 20M-13.4.D, Maintaining RWST Temperature, which of the following choices identifies ALL available actions to raise RWST temperature.

1. Isolate Chilled Water from Refueling Water Storage Tank Coolers.

2. Increase Chilled Water Temperature if plant conditions allow.

3. Start an additional Refueling Water Cooling Pump.

4. Run a Quench Spray Pump on recirculation.

A. 1 only B. 1 & 2 only C. 1, 2, & 3 only D. 1, 2, 3, & 4 Answer: C Explanation/Justification: K/A is met by demonstrating the candidates knowledge of reviewing and evaluating the Quench Spray Pump surveillance test, and determining equipment operability.

A. Incorrect. See correct answer justification.

B. Incorrect. See correct answer justification.

C. Correct. Isolating CW from the coolers, Increasing CW temp, and starting a Refuel Water Cooling pump are 3 of the 4 methods of raising RWST temperature iaw 20M-13.4.D. The fourth method is to run a LHSI pump on recirculation.

D. Incorrect. Running a Quench Spray pump on recirculation is incorrect, but plausible due to the pump recirc flowpath does go back to the RWST.

The procedure states that a LHSI pump may be operated on recirc.

Sys# System Category KA Statement 026 Containment Generic Ability to perform specific system and integrated Spray plant procedures during all modes of plant System operation.

(CSS)

KIA# 2.1.23 KIA Importance 4.3 Exam Level RO References provided to Candidate Technical

References:

20M-13.4.D Rev. 16 pgs. 3 & 4 None Question Source: Bank - Vision 123823 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.2 / 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

43. Given the following conditions:

• Unit 2 has just entered MODE 1

• Rx power is 6%

• Power is being raised slowly in preparation for rolling the Main Turbine

• 'A' MFP is in service

• Feedwater Bypass Control Valves are in AUTOMATIC

• ALL AFW pumps are aligned for normal standby operation A spurious MSLI actuation occurs.

Which of the following describes the effect the MSLI will have on the Auxiliary Feedwater pumps with NO operator action?

A. ONLY the MDAFW pumps will start.

B. ALL AFW pumps will remain in standby.

C. ONLY the TDAFW pump will start.

D. ALL AFW pumps will start.

Answer: B Explanation/Justification: KIA met by demonstrating knowledge of the integrated plant response to MS IVs inadvertently closing at low power conditions, and the response of the AFW pumps to these changing conditions.

A. Incorrect. Plausible if it is thought that the MDAFW pump started on 2/3 lo-lo SG water level on 2/3 SGs due to a Rx trip. Also, it could be thought that MDAFW pump started due to an auto trip of the MFP.

B. Correct. This is the expected plant response from a low reactor power level. SIG water level shrink is not as severe as a high power level trip.

Additionally, the MFP will remain running and provides more than enough capacity to maintain S/G water levels above the lo-lo SG water level setpoint which would trip the reactor and auto start AFW pumps.

C. Incorrect. Plausible if it is thought that the TDAFW pump started on 2/3 lo-lo SG water level on 1/3 SGs due to a Rx trip.

D. Incorrect. Plausible if it is thought that a RX trip will occur and due to lo-lo SG water level <20.5% all AFW pumps would start. This is not the case since a Rx Trip or low SG water level will not occur with these condition.

Sys# System Category KA Statement 039 Main and K3 Knowledge of the effect that a loss or malfunction of the MRSS AFWpumps.

Reheat will have on the following:

Steam System (MRSS)

KIA# K3.03 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical

References:

UFSAR Instrumentation and control system logic diagram sheet Figure 7.3-19, Rev 7 Question Source: Bank- BVPS 2LOT6 NRC P.O. Practice (2009)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

44. The Main Unit Generator output is 650 MWE.

• An Inadvertent Turbine trip occurs

• 2TMS-GV2, GOVERNOR VLV NO. 2 indicates DUAL

• 2TMS-TV3, THROTTLE VLV NO. 3 Indicates OPEN

• All other TVs and GVs are CLOSED Based on the above conditions:

1) What is the correct procedure to enter?

2) What additional actions would be required to be taken by the procedure?

A. 1) Enter AOP 2.26.1, TURBINE AND GENERATOR TRIP

2) Manually initiate a Steam Line Isolation.

B. 1) Enter AOP 2.26.1, TURBINE AND GENERATOR TRIP

2) Place BOTH Turb EH Fluid Pumps in Pull-to-Lock.

C. 1) Enter E-0, REACTOR TRIP OR SAFETY INJECTION

2) Manually initiate a Steam Line Isolation.

D. 1) Enter E-0, REACTOR TRIP OR SAFETY INJECTION

2) Place BOTH Turb EH Fluid Pumps in Pull-to-Lock.

Answer: C Explanation/Justification: KJA is met by demonstrating the required knowledge to recognize that a turbine trip occurred above P9 which will cause an automatic Rx trip and entry into E-0. Candidate should also recognize that one main steam throttle and governor valve did not close, and must take IOA RNO action steps.

A. Incorrect. With power at -65%, entry into the AOP would be inappropriate because the Rx would have already tripped, and the purpose of the AOP clearly states to stabilize the unit after a turbine and generator trip below the P-9 setpoint. SLI would be correct if the correct procedure were entered (E-0). The AOP does have IOAs for the TV & GVs not being closed, manually trip the turbine, but SLI is not an option.

B. Incorrect. With power at -65%, entry into the AOP would be inappropriate because the Rx would have already tripped, and the purpose of the AOP clearly states to stabilize the unit after a turbine and generator trip below the P-9 setpoint. Placing both EH pumps to PLT would cause the TVs & GVs to close if oil pressure was maintaining them open, but this is no longer an approved method of closing the valves, and it is not procedurally permitted.

C. Correct. With power at -65% when the turbine tripped, the Rx would have tripped due to being >P9 (49% power). Even if they thought the Rx was still critical, E-0 would still be the correct procedure to enter. To respond to All TV &/or GV closed in E-0 step 2 IOA, this is a correct action in the RNO of step 2.

D. Incorrect. With power at -65% when the turbine tripped, the Rx would have tripped due to being >P9 (49% power). Placing both EH pumps to PL T would cause the TVs & GVs to close if oil pressure was maintaining them open, but this is no longer an approved method of closing the valves, and it is not procedurally permitted.

Sys # System Category KA Statement 039 Main and Reheat Steam Generic Knowledge of EOP entry conditions and System (MRSS) immediate action steps.

KJA# 2.4.1 KJA Importance 4.6 Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.E-O, Rev. 1 lss. 2 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15) 45.

The plant is at 90% power The Balance of the Plant Operator observed the above indications for 30 seconds.

What are the required actions for the crew?

A. Trip the Rx.

B. Maintain power level and initiate a CR.

C. Immediately initiate a power reduction to less than 50%.

D. Take MANUAL control of the MFRVs and maintain SG water level.

Answer: A Explanation/Justification: KIA is met by candidate demonstrating the ability to monitor control room indications, and determine that a MFW pump motor has tripped, then based on plant conditions, take IOA of tripping the Rx per the loss of MFW AOP.

KIA statement was changed from MFW turbine trip indication to MFW trip indication after discussion with the Chief Examiner based on the fact that BV2 does not have turbine driven MFW pumps. By removing turbine from the KIA statement, the intent of the KIA was preserved.

A. Correct. The immediate operator actions of AOP-2.24.1 {loss of Main Feedwater) requires that the Rx be tripped if less than 2 MFPs are running when >80%. The candidate must know that if one pump motor is running and the other motor trips, a trip of the running motor will occur.

B. Incorrect. The bright white light on the pump indicates the motor tripped on 2FWS-112182. The candidate must know that if one pump motor is running and the other motor trips, a trip of the running motor should occur within 1.5 seconds, therefore they should trip the motor iaw conduct of Operations procedure. At 90% pwr a Rx trip is required making this answer incorrect.

C. Incorrect. The appropriate action would be to enter AOP-2.24.1 {Loss of Main Feedwater) which would give direction to lower reactor power to

<52% IF initial conditions were <80% power. Therefore, since the plant is at 90%, a manual Rx trip is required, and pwr reduction is incorrect.

D. Incorrect. The candidate may feel that with one motor still running that the feed pump is capable to maintain feed flow but at a reduced capability.

This may lead them to believe that manual FRV control would be required.

Sys # System Category KA Statement 059 Main Feedwater A4 Ability to manually operate and monitor in the control MFW tl:ffbiffe trip indication (MFW) System room:

KIA# A4.01 KIA Importance 3.1* Exam Level RO References provided to Candidate Technical

References:

20M-53C.4.2.24.1 Rev. 6 pg. 2 None 20M-24.1.D Rev. 6 pg. 10 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

46. The following conditions exist:

• The plant is at 17% power

• The crew is raising power in accordance with 20M-52.4.A, 'Raising Power From 5% to Full Load Operation"

• All Feedwater Bypass Control valves are in AUTO maintaining SG levels within the control band

• The controller for 2FWS-FCV489, '21 B Feedwater Bypass Control Valve' fails to 10%

open causing 'B' SG NR water level to lower

1) What is the correctAOP to respond to this event when annunciator A6-10E "SG 21B LEVEL DEVIATION FROM SETPOINT' alarms?

2) If a reactor trip due to SG low-low level occurs, which Auxiliary Feed Pump(s) will automatically start?

A. 1) AOP-2.24.1 'LOSS OF MAIN FEEDWATER'

2) TURBINE Driven Auxiliary Feedwater Pump B. 1) AOP-2.24.1 'LOSS OF MAIN FEEDWATER'

2) MOTOR Driven Auxiliary Feedwater Pumps C. 1) AOP-2.4.1 'PROCESS CONTROL FAILURE'

2) TURBINE Driven Auxiliary Feedwater Pump D. 1) AOP-2.4.1 'PROCESS CONTROL FAILURE'

2) MOTOR Driven Auxiliary Feedwater Pumps Answer: C Explanation/Justification: KIA is met by demonstrating the ability to predict the MFW system response to a MFRV controller malfunction, and determining which procedure would be used to mitigate the event. Then state which AFW pump will automatically start if the MFRV failure is not corrected, based on knowledge of the AFW pump auto start coincidences.

At the discretion of the Chief Examiner, this KIA was changed from A2.01 to A2.12.

A. Incorrect. AOP-2.24.2 is a plausible distractor is it is thought that this event constitutes a loss of main feedwater, but in this case a controller has failed and the SGs are still being fed. TDAFW pump is the correct pump to start when 2/3 SGWL detectors in only 1 SG reaches 20.5%.

B. Incorrect. AOP-2.24.2 is a plausible distractor is it is thought that this event constitutes a loss of main feedwater, but in this case a controller has failed and the SGs are still being fed. MDAFW pump is incorrect because they will start when 2/3 SGWL detectors in 2/3 SGs reaches 20.5%.

C. Correct. Correct AOP to use when a process parameter is not being controlled within its normal control band with the control in auto. TDAFW pump is the correct pump to start when 2/3 SGWL detectors in only 1 SG reaches 20.5%.

D. Incorrect. Correct AOP to use when a process parameter is not being controlled within its normal control band with the control in auto. MDAFW pump is incorrect because they will start when 2/3 SGWL detectors in 2/3 SGs reaches 20.5%.

Sys # System Category KA Statement 059 Main Feedwater A2 Ability to (a) predict the impacts of the following malfunctions or operations on Failure of feedwater (MFW) System the MFW; and (b) based on those predictions, use procedures to correct, control, regulating valves or mitigate the consequences of those malfunctions or operations:

KIA# A2.12 KIA Importance 3.1

• Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.4.1 Rev. 1 pg. 1 USFAR figure 7.3-19 Rev. 7 20M-24.4.AAP Rev. 5 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

47. Given the following plant conditions:

• The plant is in Mode 3 following a reactor trip from 100%

• The crew has completed ES-0.1, Reactor Trip Response, and transitioned to 20M-52.4.R.1. F, Station Shutdown from 100% Power to Mode 5

• A plant cooldown of 50°F/hr is commenced using Condenser Steam Dumps

• Steam Generator (SG) Narrow Range levels are maintained stable at 44% during the cooldown, using the Auxiliary Feedwater pumps Which one of the following describes the trend of Auxiliary Feedwater flow requirements as the plant cools down to Mode 5?

A. AFW flow requirements are constant as long as SG level remains constant.

B. AFW flow requirements are constant as long as the cooldown rate remains constant.

C. More AFW flow is required to maintain SG level due to a rise in SG water density as it cools.

D. Less AFW flow is required to maintain SG level because heat input to the SGs lowers as the cooldown continues.

Answer: D Explanation/Justification: KIA is met by determining sufficient AFW flow is available to provide decay heat removal, and the knowledge that decay heat load will be larger after a Rx trip from higher power levels.

A. Incorrect. Plausible if heat input to the SG did not change. Heat input lowers due to less decay heat as the cooldown progresses.

B. Incorrect. Plausible if the effects of less decay heat are not considered.

C. Incorrect. Water density does not have any impact at this temperature in the SG.

D. Correct. Heat input lowers as the cooldown progresses due to less decay heat from the reactor.

Sys# System Category KA Statement 061 Auxiliary I Emergency K5 Knowledge of the operational implications of the Decay heat sources and magnitude Feedwater (AFW) System following concepts as the apply to the AFW:

KIA# K5.02 KIA Importance 3.2 Exam Level RO References provided to Candidate Technical

References:

GO-GPF.R8 A Rev. 1 None Question Source: Bank-Callaway 2013 NRC Exam (046)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 45. 7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

48. Plant is at 38% power, raising power to 100% in accordance with 20M-52.4.A, Raising Power From 5% To Full Load Operation.

• Charging pump 2CHS*P21A is RUNNING

• Main Feedwater Pump 2FWS-P21 Bis RUNNING

• Service Water is in NSA

• Diesel Generator 2-1 is on clearance

• All Tech Spec actions for DG 2-1 have been completed Load Tap Changer for Bus 2A SS Serv Tfmr 2A is in auto and drifts.

Bus voltage is 3800 VAC (108.5 VAC indicated) and stable.

Load Tap Changer will not respond in Auto or Manual.

Based on these conditions, assuming Bus 2A voltage remains at 3800 VAC, and assuming no operator actions have yet been taken:

1) Which of the following correctly describes plant status or expected actions?

2) Procedural guidance in effect 5 minutes after bus voltage drifts to 3800 VAC?

A. 1) The reactor will have automatically tripped due to effects of the abnormal bus voltage.

2) E-0, "Reactor Trip Or Safety Injection" B. 1) Bus 2AE will have automatically de-energized.

2) AOP 2.36.2, "Loss of 4KV Emergency Bus" C. 1) The reactor will be manually tripped IAW the bus abnormal voltage ARP.

2) E-0, "Reactor Trip Or Safety Injection" D. 1) Bus 2AE will be manually de-energized.

2) AOP 2.36.2, "Loss of 4KV Emergency Bus" Answer: B Explanation/Justification: KJA is met by candidate predicting the effect low voltage on Bus 2A will have on the emergency bus 2AE (undervoltage condition will strip Bus 2AE, and the EOG is not available), then respond using the appropriate abnormal operating procedure.

A. Incorrect. Plausible distractor because they may feel that the 'A' RCP would trip (<75% undervoltage, stem is -91 %) due to lowered 'A' bus voltage. With the plant being >P-8 (30%) and 1/3 RCP tripping would cause the Rx to trip. Neither the Rx, nor the 'A' RCP will trip.

B. Correct. With DG 2-1 on clearance and Bus voltage dropping below 93.4% (3885 VAC with 90 sec.TD) the emergency power undervoltage protection will strip and isolate the 2AE bus. The bus will be de-energized and the correct procedure is AOP 2.36.2 for the loss of 2AE.

C. Incorrect. Plausible distractor if the candidate feels as though voltage is too low and the 2A bus should have tripped, which would cause an RCP to trip. Per the ARP bus voltage must be 75% of nominal bus voltage (-3120V), but the stem has voltage -91 %) therefore this answer is incorrect.

D. Incorrect. Plausible distractor if it is recognized that the voltage dropped below the 93.4%. The question stated that 5 minutes have elapsed, therefore bus 2AE would have automatically tripped after a 90 second time delay, and no manual actions would be required.

Sys # System Category KA Statement 062 AC Electrical A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Consequences of exceeding Distribution ac distribution system; and (b) based on those predictions, use procedures to correct, voltage limitations System control, or mitigate the consequences of those malfunctions or operations:

KJA# A2.08 KJA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

LSK-022-0058 rev. 8 SPD-27-VE3200AB rev. 1 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 45.3 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

49. Which of the following is correct regarding a loss of Vital Bus Inverter 2-1 normal power supply?

If Vital Bus 2-1 normal power supply ___(1) ___ is de-energized, Vital Bus Inverter 2-1 will automatically be supplied by ___(2) ___ without affecting the regulated AC output to Vital Bus 2-1.

A. 1) MCC2-E13

2) MCC2-E05 B. 1) MCC2-E07

2) DC SWBD 2-1 C. 1) MCC2-E13

2) DC SWBD 2-1 D. 1) MCC2-E07

2) MCC2-E05 Answer: C Explanation/Justification: KIA is met by demonstrating knowledge of the physical connections between AC and DC supplies to the UPS units, and demonstrating an understanding of the cause and effect relationship between AC source being lost, DC source will pick up the load.

A. Incorrect. Normal power supply is E-13. Incorrect answer of MCC2-E05 being the backup power supply if normal power is lost. Plausible distractor because MCC2-E05 is the backup regulated voltage supply to Vital Bus 2-1 if the inverter is removed from service.

B. Incorrect. Plausible distractor because MCC2-E07 is the backup regulated voltage supply to Vital Bus 2-3. DC SWBD 2-1 is the normal backup to the inverter.

C. Correct. Normal power supply is E-13, with DC SWBD 2-1 being the normal backup to the inverter.

D. Incorrect. Plausible distractor because MCC2-E07 is the backup regulated voltage supply to Vital Bus 2-3 and MCC2-E05 is the backup regulated voltage supply to Vital Bus 2-1 if the inverter is removed from service.

Sys# System Category KA Statement 063 DC Electrical K1 Knowledge of the physical connections and/or cause effect AC electrical system Distribution relationships between the DC electrical system and the following System systems:

KIA# K1.02 KIA Importance 2.7 Exam Level RO References provided to Candidate Technical

References:

20M-38.1.B Rev. 1, pg. 2 None RE-0001AW Rev.21 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 I 45.7 to 45.8)

Objective: 3SQS-38.1 Rev. 8, Obj. 2 From memory, describe the Normal System Arrangement for the Emergency 120 VAC Distribution Systems, including distribution paths, status of feeder breakers, loads, bus transfer switches, power train, and bus designation.

Beaver Valley Unit 2 NRC Written Exam (2LOT1S)

50. A large air break at the outlet of EOG 2-1 air receiver [2EGA*TK21A] is depressurizing the air receiver.

1) What is the minimum Tech Spec required Air Receiver pressure?

2) If air receiver [2EGA*TK21A] fully depressurizes, EOG 2-1 _ __...._(1;...r..)___ start upon receipt of an auto start signal.

A. 1) ~165 psig

2) will B. 1) ~380 psig

2) will C. 1) ~165 psig

2) will NOT

0. 1) ~380 psig

2) will NOT Answer: B Explanation/Justification: KIA is met by requiring knowledge of the EOG air system configuration and lineup, and the effects that a loss of one air receiver will have on the starting capabilities of the EOG.

A. Incorrect. Tech Spec minimum air pressure is 380 psig. BV1 & 2 use combined Tech Specs which identify both unit air pressures on the same page, this makes 165 psig a plausible distractor at BV. DG will start.

B. Correct. TS limit for air pressure is ~380 psig. The knowledge of the correct value is gained through performing OSTs, tech specs, and log taking. DG will start even with a rupture at the outlet of TK-21A due to there being 2 air systems/receivers which are not cross tied. This allows the pressurized receiver to admit air to 6 cylinders (1 /2) and start the diesel.

C. Incorrect. Tech Spec minimum air pressure is 380 psig. BV1 & 2 use combined Tech Specs which identify both unit air pressures on the same page, this makes 165 psig a plausible distractor at BV. DG will start as explained in the correct answer.

D. Incorrect. TS limit for air pressure is ~380 psig. DG will start as explained in the correct answer.

Sys# System Category KA Statement 064 Emergency Diesel K6 Knowledge of the effect of a loss or malfunction of the Air receivers Generator (ED/G) following will have on the ED/G system:

System KIA# K6.07 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

20M-36.1.C Rev. 4 pg. 8 20ST-36.1 rev. 71, pg. 9 U2 RM-0436-003 Rev. 20 Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective: 2SQS-36.2 Rev. 21 Obj. 9 Identify the EOG field instruments, subsystems and components that are required to be operable by the Technical Specifications. 20JT-1.36

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

51. The plant is at 75% power with all systems in normal alignment for this power level.

• Leak Collection Ventilation Radiation monitor 2RMR-RQl301 fails HIGH In response to this failure, where will the Contiguous areas exhaust be discharged?

The Contiguous areas exhaust will be directed through the filter banks and discharged A. through the Ventilation Vent to atmosphere B. through the Elevated Release to atmosphere C. to the Auxiliary Building D. to the Containment Building Answer: B Explanation/Justification: KIA is met by demonstrating knowledge that the normally unfiltered ventilation system realigns to filter the contiguous area exhaust before releasing it to the atmosphere when a process rad monitor malfunction occurs.

A. Incorrect. Plausible distractor because they must know that the Ventilation Vent is the normal discharge path for the Contiguous areas, but that it does not normally pass through the filter banks.

B. Correct. When 2RMR-RQl301 fails high, 2HVS-MOD201A&B will close isolating the Ventilation Vent flowpath, and 2HVS-MOD202A&B opens to align the contiguous areas to the filter banks. The only flowpath from the filter banks is through Leak Collection Filtered Exhaust fans to the Elevated Release to atmosphere.

C. Incorrect. Plausible distractor because they may think that the ventilation lineup will re-align the discharge of filtered exhaust to the surrounding area of the filter banks, which is the Auxiliary Building.

D. Incorrect. Plausible distractor because it could be thought that the ventilation would re-align the discharge of filtered exhaust to the containment building via the purge supply or exhaust.

Sys# System Category KA Statement 073 Process K3 Knowledge of the effect that a loss or malfunction of the PRM Radioactive effluent releases Radiation system will have on the following:

Monitoring (PRM)

System KIA# K3.01 KIA Importance 3.6 Exam Level RO References provided to Candidate Technical

References:

20M-16.1.D Rev.2 pg. 2 None 2SQS-16.1 PPNT Rev. 12 slide 6 Question Source: Bank- Vision #124119 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

52. Given the following plant conditions:

• The Unit was operating at 100% power with all systems in NSA

• An event occurred that caused containment pressure to peak at 6 psig

• Offsite Power has remained available for the duration of the event

• All System functions as designed Based on these plant conditions, which of the following combinations of reactor and turbine building components will have service water flow for temperature control?

CCP HX's = Primary Component Cooling Water Heat Exchangers CCS HX's = Secondary Component Cooling Water Heat Exchangers

=

EDG's Emergency Diesel Generators RSS HX's = Recirculation Spray Heat Exchangers CCP HX's CCS HX's EDG's RSS HX's A. YES YES YES YES B. YES YES YES NO C. NO NO NO NO D. YES NO YES NO Answer: D Explanation/Justification: K/A is met by the candidate predicting which of the listed components will have cooling water supplied after a an SI and CIA occur. The K/A statement is met by identifying that the CCP HXs (Rx plant CCW) will have temperature control capabilities and the CCS HXs (Turbine plant CCW) will not have temperature control.

A. Incorrect. CCS HX will isolate on SI/CIA. RSS HX's will be isolated until CIB actuates at 11.1 psig containment pressure.

B. Incorrect. CCS HX will isolate on SI/CIA.

C. Incorrect. CCP HX's will not isolate until CIB at 11.1 psig containment pressure so therefore will be providing flow and temperature control. EOG will have cooling even though they will be running unloaded in this plant configuration.

D. Correct. At> 5 psig containment pressure, SI and CIA have actuated. 2SWS*MOV107A-D close isolating CCS HX's, therefore there will be no cooling or temperature control to the CCS HX's. The SI signal will start EDGs and open 2SWS*MOV113A&D, therefore providing cooling to EDG's. CIB does not actuate until 11.1 psig, so therefore 2SWS*MOV106A&B will remain open providing cooling and therefore temperature control to the CCP HX's. 2SWS*MOV103A&B remain shut and do not open until containment pressure reaches 11.1 psig (CIB).

Sys# System Category KA Statement 076 Service A 1 Ability to predict and/or monitor changes in parameters (to Reactor and turbine building closed cooling water Water prevent exceeding design limits) associated with operating the SWS temperatures.

System controls including:

(SWS)

KIA# A1.02 KIA Importance 2.6* Exam Level RO References provided to Candidate None Technical

References:

20M-30.1.D, Rev. 8 2808-30.1 PPT, Rev. 23 Question Source: Bank - 2LOT7 NRC Exam (053) (2011)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

53. Given the following conditions:

• The plant is at 100% power

• Containment Instrument Air is being supplied by Station Instrument Air

• A Large Break Loss of Coolant Accident occurs

• All systems function as designed

• No operator actions have been taken Based on these plant conditions, which valve(s) will need to be reopened to restore instrument air to the containment?

1. 21AC-MOV130, CNMT Instrument Air lsol Viv.

2. 21AC-MOV131, CNMT Instrument Air Backup Supply Viv.

3. 21AC-MOV133, CNMT Instrument Air lsol Viv.

4. 21AC-MOV134, CNMT Instrument Air lsol Viv.

A. 1 ONLY.

B. 1AND2 ONLY.

C. 3 AND 4 ONLY.

D. 1, 2, AND 3.

Answer: A Explanation/Justification: KIA met with the required knowledge that CNMT instrument air is supplied from station instrument air, and that a CIA signal will close 21AC-MOV130 and isolate air to CNMT.

A. Correct. 21AC-MOV131 and 21AC*130 are open at 100% power to supply instrument air from instrument air compressors into containment. BVPS Unit 2 no longer uses containment air compressors. Upon a large break LOCA and SI, the subsequent CIA signal will auto close 21AC*130. In order to restore instrument air to containment, this valve needs to be reopened only.

B. Incorrect. Correct that 21AC*MOV130 needs to be reopened. Plausible if the candidate does not know that 21AC-MOV131 does not receive a CIA signal or believes this valve is affected by this signal. The EOP directs both of these valves opened, however, the EOP deals with all modes of operation and in the stated plant mode, the candidate must know it is not necessary to reopen 21AC-MOV131.

C. Incorrect. 21AC*MOV133 & 134 both receive a CIA signal and close. This was the old configuration when running CNMT IAC instrument air to containment. Opening these valves will not restore IA to containment.

D. Incorrect. All three of these valves receive a CIA signal and close from their NSA open positions. The candidate may believe that these valves all need to be reopened to restore instrument air.

Sys# System Category KA Statement 078 Instrument K1 Knowledge of the physical connections and/or cause-effect Containment air Air System relationships between the IAS and the following systems:

(IAS)

KIA# K1.03 KIA Importance 3.3* Exam Level RO References provided to Candidate None Technical

References:

20M-34.1.D Rev. 4 pg. 7 U2 RM-0434-003 rev. 17 20M-53A.1.E-O, Issue 2, Rev. 1, pg. 20 Question Source: Bank - 2LOT8 NRC Exam (053) (2012)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

54. When in Mode 1, what is the NSA required position of 2HVR*DMP206, Containment Vacuum Breaker Ball Valve, to prevent inadvertently breaking containment vacuum?

2HVR*DMP206 is CLOSED ~~~~~~~~~

A. with Instrument Air Isolated B. and De-energized C. and Chain Locked D. with Shorting Bar removed Answer: C Explanation/Justification: KIA is met by the knowledge of that CNMT vacuum breaker is chain locked closed to prevent inadvertent breaking of CNMT vacuum when in Modes 1-4. This is an administrative interlock.

Manually operated 2HVR*DMP206 has remote position indication on the Building Service Control Panel in the Control Room. With this indication available in the CR, it helps to make all incorrect distractors plausible as the candidate may think it is an electrically operated valve.

A. Incorrect. Plausible means of failing an air operated valve closed.

B. Incorrect. Plausible means of failing a motor operated valve in a desired position.

C. Correct. IAW 20ST-48.7, 2HVR*DMP206 is required to be chain locked closed in Modes 1-4.

D. Incorrect. Plausible means of removing power from the contactor to prevent valve movement.

Sys# System Category KA Statement 103 Containment K4 Knowledge of containment system design feature(s) and/or Vacuum breaker protection System interlock(s) which provide for the following:

KIA# K4.01 KIA Importance 3.0* Exam Level RO References provided to Candidate None Technical

References:

20M-44C.4.A Rev. 23 pg 4 RM-0444C-002 Rev. 7 20ST-48.7 Rev. 41 pg 16 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

55. The plant is at 50% power.

• Annunciator A 1-1 E, Containment Air Pressure High/Low has alarmed

• 2CVS-Pl101A1 CNMT PRESS is reading 14.1 psia

• No other Control Room annunciators are in alarm Which of the actions below will clear annunciator A 1-1 E in accordance with 20M-12.4.AAA, "Containment Air Pressure High/Low"?

A. Align the Containment Vacuum Ejector for use B. Shutdown the CNMT Vacuum Pumps C. Shutdown an operating CNMT Air Recirc (CAR) Fan D. Start a CNMT Vacuum Pump Answer: D Explanation/Justification: KJA is met by demonstrating the ability to recognize a CNMT high pressure condition from the Control room, and respond by manually starting the CNMT vacuum pump to restore pressure.

A. Incorrect. This is a plausible means of lowering cnmt pressure, but it is used to draw initial cnmt vacuum. It is not an approved method iaw 20M-12.4.AAA to lower cnmt pressure.

B. Incorrect. This is a plausible distractor if it is thought that cnmt pressure is low due to operating the cnmt vacuum pump. Incorrect because cnmt pressure is high.

C. Incorrect. Plausible distractor if it is thought that cnmt pressure is low due to low temperature, and must be raised back into normal band.

Probable cause #5 of ARP states is pressure is low due to temperature being low, the stop CAR fan. Incorrect because cnmt pressure is high.

D. Correct. Must recognize that cnmt pressure is high and must be lowered to 13.4-13.6 psia using the vacuum pumps. This is directed by 20M-12.4.E "Maintaining the Containment Vacuum" which 20M-12.4.AAA references. This knowledge of the setpoint range can be determined by the above the line RO knowledge for TS 3.6.4 pressure limits of ~12.8 - S14.2 psia.

Sys# System Category KA Statement 103 Containment A4 Ability to manually operate and/or monitor in the control room: Containment vacuum system System KIA# A4.09 KIA Importance 3.1* Exam Level RO References provided to Candidate Technical

References:

20M-12.4.AAA Rev. 5 pg. 6 None 20M-12.4.E Rev. 4 pg. 2 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

56. The plant is at 75% power with all systems in normal alignment for this power level EXCEPT PRZR Pressure Controller 2RCS*PK444A is in MANUAL controlling pressure at 2235 psig.

• All plant parameters are on program

• A 20% Load Rejection occurs

• No operator actions have occurred As compared to the initial conditions, what is the status of the PRZR level and pressure 5 minutes after the Load Rejection occurred?

PRZR Level PRZR Pressure A. Lower Lower B. Lower Higher C. Higher Lower D. Higher Higher Answer: A Explanation/Justification: K/A is met by predicting the effect that the control rods inserting during a load rejection will have on both PRZR level and pressure.

A. Correct. With a 20% load rejection Tref will be at a lower value than initial. Rod control will drive the rods in to get Tavg down to within 1.5F of Tref. This will cause Tavg to lower. PRZR program level control (22-53% program) is based on Tavg (547-574F), therefore, PRZR level will be lower. With 2RCS*PK444A in manual (approx. 42% demand for 2235 psig), when the LR occurs PRZR pressure will lower, since the spray valves remain open (where as if in auto they would close at 40.6%), pressure will drive lower than expected, and it will take longer for heaters to recover pressure because all Backup heaters will not energize. (all Backup heaters turn on at 9.4% demand if in auto).

8. Incorrect. Plausible distractor if it is thought that PRZR Pressure Controller in manual will allow pressure to rise about initial pressure and remain higher.

C. Incorrect. Plausible distractor if the candidate does not have a thorough understanding of PRZR level controller programming and Rod Control Temperature control. Pressure will be lower than initial.

D. Incorrect. Plausible distractor if the candidate does not have a thorough understanding of PRZR level controller programming and Rod Control Temperature control. Pressure would be lower than initial.

Sys# System Category KA Statement 001 Control Rod A1 Ability to predict and/or monitor changes in parameters (to PZR level and pressures Drive System prevent exceeding design limits) associated with operating the CRDS controls including:

KIA# A1.04 KIA Importance 3.7 Exam Level RO References provided to Candidate Technical

References:

20M-1.1.B Rev. 6 pg 10 None 20M-1.5.A.48 lss. 1 Rev. 1 20M-6.4.IF Rev. 13 Pgs. 24 & 25 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

57. The following conditions exist:

• The Plant is at 100% power

• PRZR PRESS CONTROL 2RCS*PK444A is in AUTO set at 2235 psig

• PRZR HEATERS CONTROL GROUP C [2RCP-H2C] CS has a RED target

• PRZR HEATERS BACKUP GROUP B & D [2RCP-H2B & H2D] CS have RED targets

• PRZR HEATERS BACKUP GROUP A & E [2RCP-H2A & H2E] CS have GREEN targets

[2RCS*PCV455B] 'B' PRZR SPRAY VALVE Fails OPEN.

Which of the following is the correct order of automatic actions that occur as RCS pressure is lowering?

1. PRZR Backup Heaters ON

2. 'A' PRZR Spray Valve [2RCS*PCV455A] CLOSED

3. PRZR Heaters Control Group C [2RCP-H2C] ON A. 2, 3, 1 B. 2, 1, 3 C. 3, 2, 1 D. 3, 1, 2 Answer: A Explanation/Justification: KJA is met by the candidates ability to identify automatic actions which occur as RCS pressure is lowering, this includes automatic operation of PRZR heaters which raise PRZR pressure, and PRZR spray flow control on the non-faulted spray valve.

A. Correct. The Master Pressure Controller will respond to the lowering pressure by controlling the spray valve and heaters. The MPC output will be driving to 0 as pressure is lowering. PCV455A will close at 40.6, Control Htrs will come on at 34.4, B/U htrs will energize at 9.4 (2210 psig).

Candidate must have the knowledge of normal plant operations where as some spray flow is desired because the balancing of spray flow and heaters results in a normal, constant Przr outsurge.

B. Incorrect. See correct explanation.

C. Incorrect. See correct explanation.

D. Incorrect. See correct explanation.

Sys# System Category KA Statement 002 Reactor Coolant A3 Ability to monitor automatic operation of the RCS, Pressure, temperatures, and flows System (RCS) including:

KIA# A3.03 KIA Importance 4.4 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.IF Att. 2 Rev. 13 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41. 7 I 45.5)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

58. The plant is at 100% power.

• PRZR Level Control Channel Selector is in Channel I & II (LT459 & 460)

• PRZR level is on program The reference leg for PRZR Channel I Level [2RCS*LT 459] develops a leak.

How will 2RCS-Ll459 initially respond and, if the event continues with NO operator action, what automatic trip signal will initiate the reactor trip?

Rx trip 2RCS*Ll459 Signal A. Rise High Pressurizer Level B. Lower High Pressurizer Level C. Rise Low Pressurizer Pressure D. Lower Low Pressurizer Pressure Answer: A Explanation/Justification: KIA is met by demonstrating the knowledge to understand the effects a reference leg failure on one of the post accident monitor PRZR level indicators will have of the PRZR Level Control System, and the overall function that the PAM PRZR level indicators have as reactor trip inputs. At BVPS all three of the PAM przr level instruments are used for the level control system.

A. Correct. A reference leg leak will cause the affected PAM channel (LT459) to indicate high. Since LT459 is the controlling channel, it will cause 2CHS-FCV122 to close to minimum flow, thus causing PRZR level to initially lower until letdown isolates at 14% on LT460. After UD isolates, PRZR level will rise due to charging flow (minimum flow of 25 gpm when FCV122 is automatically closed) and seal injection, until PRZR level reaches 92% on 2/3 indicator when > 10% power. This will generate a rx trip.

B. Incorrect. Plausible if the candidate doesn't understand the difference between a reference leg and a variable leg leak. Because a variable leg leak on LT459 would cause indicated level to lower. A rx trip would be generated on high przr level due to LT460 isolating UD.

C. Incorrect. A reference leg leak will cause the affected PAM channel (LT459) to indicate high. Plausible distractor of Rx trip on low PRZR pressure if candidate feels that that a leak will cause a low pressure Rx trip. The reference leg is 3/4" line with a 3/8" flow restrictor in line.

D. Incorrect. Plausible if the candidate doesn't understand the difference between a reference leg and a variable leg leak. Because a variable leg leak on LT459 would cause indicated level to lower. Plausible distractor of Rx trip on low PRZR pressure if candidate feels that that a leak will cause a low pressure Rx trip. The reference leg is 3/4" line with a 3/8" flow restrictor in line.

Sys # System Category KA Statement 011 Pressurizer Level Control K6 Knowledge of the effect of a loss or malfunction on the Function of PZR level gauges as post System (PZR LCS) following will have on the PZR LCS: accident monitors KIA# K6.05 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

GO-GPF.C7 Rev. 4 pg. 55, 20M-1.5.B.1 Rev. 2 20M-6.4.IF, attachment 1, rev 13 U2 RM-0406-003 Rev. 6 Question Source: Bank - Vision #131762 Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

59. The following conditions exist.

• A large break LOCA has occurred

• TSC has been activated

• Annunciator A 1-28, Hydrogen Level High/High-High is in alarm Complete the following statements. Assume the TSC has been c decision.

The High-High Hydrogen concentration in Containm (1)

In accordance with the Hydrogen Level High/H" -High ARP, the crew will _ _-->,.;;;(2_,_)_ _ _ in response to the High-High Hydrogen level i ontainment.

A. 1) 2.5%

2) intentionally ignite the C B. 1) 2.5%

2) start [2HCS-F ] Containment Atmosphere Purge Blower C. 1)

2) ally ignite the Containment atmosphere D. .5%

start 2HCS-FN21 Answer: D Explanation/Justification: BVPS2 has retired the H2 recombiners and explosive H2 concentration in the Cnmt is beyond design based accident.

We discussed the KIA with Chief Examiner who stated to stay focused on the purge control portion of the system.

KIA is met knowledge of the operation of the containment purge system in the event of a High-High Hydrogen concentration in the containment.

A. Incorrect. 2.5% was chosen to be a realistic choice verses 0.5% which is the High H2 alarm setpoint. Plausible distractor of igniting the atmosphere since this is an option in our severe accident management guidelines (20M-53E.1.SAG-7), but this is not the correct response at this H2 level or iaw the ARP.

8. Incorrect. 2.5% was chosen to be a realistic choice verses 0.5% which is the High H2 alarm setpoint. This is the correct response iaw the ARP.

C. Incorrect. Correct setpoint for the High-High alarm. Plausible distractor of igniting the atmosphere since this is an option in our severe accident management guidelines (20M-53E.1.SAG-7), but this is not the correct response at this H2 level or iaw the ARP.

D. Correct. Correct setpoint for the High-High alarm. Correct actions iaw the ARP.

Sys # System Category KA Statement 028 Hydrogen Recombiner and K5 Knowledge of the operational implications of the Explosive hydrogen concentration Purge Control System (HRPS) following concepts as they apply to the HRPS:

K/A# K5.01 KJA Importance 3.4 Exam Level RO References provided to Candidate Technical

References:

20M-46.4.ABD Rev. 3 pg. 5 None Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 45.7)

Objective: 2SQS-46.1 Obj. 15 Describe the control, protection and interlock functions for the control room components associated with Post OBA Hydrogen Control System, including automatic functions, setpoints and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

60. Given the following:

• The plant is in Mode 5

• A Forced Containment Purge through the SLCRS Unfiltered flow path is in progress

• The Personnel Airlock and Equipment Hatches are closed

• The Cnmt Purge Supply lsol Damper 2HVR*MOD25A inadvertently CLOSES Assuming NO Operator actions, how will containment pressure compare to pre-event conditions 5 minutes after 2HVR*MOD25A closes, and what will be the status of 2HVS-FN263B, Leak Collection Normal Exhaust Fan?

Containment pressure will be _ ___._(1"-J.)___ it was before 2HVR*MOD25A closed.

2HVS-FN263B, Leak Collection Normal Exhaust Fan will be (2)

A. 1) the same as

2) tripped B. 1) the same as

2) running C. 1) lower than

2) tripped D. 1) lower than

2) running Answer: D Explanation/Justification: KIA is met with the knowledge of the Containment Purge system being aligned for forced purge, and the Cnmt Purge Supply lsol Damper goes closed and the effect this failure will have on containment pressure.

A. Incorrect. Cnmt pressure will lower due to the air being drawn from the cnmt by 2HVS-FN263B. Plausible distractor if candidate thinks that 2HVR-MOD23A closes on interlock with the MOD25A closing, thus causing the containment to be isolated from the purge supply and exhaust flow paths. Plausible distractor because FN263B is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed.

B. Incorrect. Cnmt pressure will lower due to the air being drawn from the cnmt by 2HVS-FN263B. Plausible distractor if candidate thinks that 2HVR-MOD23A closes on interlock with the MOD25A closing, thus causing the containment to be isolated from the purge supply and exhaust flow paths. It is correct that FN263B will continue to run.

C. Incorrect. Cnmt pressure will lower, but 2HVS-FN263B will not be tripped. Plausible distractor because FN263B is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed. Candidate must know the difference between FN263A and FN263B, and which is used for the forced purge lineup.

D. Correct. When 2HVR*MOD25A closes it will isolate the flowpath from 2HVP-ACU211 B, PAB NC Unit which is the forced air side of the cnmt purge. This will leave only 2HVS-FN263B drawing airflow from the cnmt causing cnmt pressure to lower. 2HVS-FN263B will be running because it is not effected by the closure of Cnmt Purge Supply lsol Damper. It is tripped if 2HVR*MOD23A, Cnmt Purge Discharge lsol Damper were to go closed not MOD25A.

Sys # System Category KA Statement 029 Containment K3 Knowledge of the effect that a loss or malfunction of the Containment parameters Purge Containment Purge System will have on the following:

System (CPS)

KIA# K3.01 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

U2 RM-0444C-002 Rev 7, U2 RM-0444D-001 Rev. 8 U2 RM-0416-001 rev. 12 20M-44C.4.A Rev. 23 pg. 5 20M-16.1.D Rev. 2 pg. 3 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

61. Which of the following completes the statement below?

During refueling operations the fuel assemblies are removed from the core using the Manipulator crane (1) hoist which is equipped with an automatic overload interlock that is set at a maximum of (2) pounds.

A. 1) main

2) 2700

8. 1) main

2) 3000 C. 1) auxiliary

2) 2700 D. 1) auxiliary

2) 3000 Answer: A Explanation/Justification: K/A is met with the knowledge of the functions of the refueling equipment and the hoist overload protection associated with the manipulator crane main hoist..

A. Correct. The main hoist is raises and lowers the gripper tube and removes fuel assemblies from the core. The overload limit is set at S2700 pounds.

B. Incorrect. The main hoist is raises and lowers the gripper tube and removes fuel assemblies from the core, but the overload limit is not set at 3000 pounds.

C. Incorrect. The auxiliary hoist is located on the bridge monorail for handling accessory refueling equipment, not fuel assemblies. The overload limit is set at S2700 pounds.

D. Incorrect. The auxiliary hoist is located on the bridge monorail for handling accessory refueling equipment, not fuel assemblies, and the overload limit is not set at 3000 pounds.

Sys# System Category KA Statement 034 Fuel Handling K4 Knowledge of design feature(s) and/or interlock(s) which provide Overload protection Equipment for the following:

System (FHES)

KIA# K4.03 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

LP 3SQS-6.13 Rev. 6 pg 12 2RP-3.3, Rev. 5 lss. 0, pg 3 & 21 Question Source: Bank - Harris 2011 NRC Exam (061)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective: 3SQS-6.13 Rev. 6 (1) Describe the function of the following Fuel Handling equipment as documented in the Refueling Procedures:

Manipulator crane, Manipulator crane auxiliary hoist. (2) Describe the control, protection and interlock functions for the fuel handling equipment, including automatic functions, setpoints and changes in equipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

62. 1) What is the Steam Generator level setpoint for the Auxiliary Feedwater Pumps to automatically start?

2) In accordance with E-0 "Reactor Trip or Safety Injection", what is the basis for Auxiliary Feedwater automatic initiation?

A. 1) 20.5% Narrow Range Level

2) Provide a secondary heat sink.

B. 1) 20.5% Narrow Range Level

2) Prevent Steam Generator dryout.

C. 1) 19.6% Narrow Range Level

2) Provide a secondary heat sink.

D. 1) 19.6% Narrow Range Level

2) Prevent Steam Generator dryout.

Answer: A Explanation/Justification: KIA is met by the candidate demonstrating the knowledge of the design feature of the AFW pump start setpoints, and that the reason AFW starts is to supply feed to the SG for decay heat removal.

A. Correct. 2/3 detectors indicating 20.5% NR level (low-low setpoint) on 2/3 SGs will start the MDAFW, (2/3 on 1 SG for TDAFW) pumps to keep the tubes covered for secondary heat removal.

B. Incorrect. Plausible because the setpoint is correct. Incorrect basis. It could be thought that aux feedwater was provided just to prevent SG dryout, however, the reason is to keep the tubes covered for secondary heat removal.

C. Incorrect. Plausible level because 19.6% is the Unit 1 setpoint. Correct bases.

D. Incorrect. Plausible because 19.6% is the Unit 1 setpoint. Incorrect bases. It could be thought that aux feedwater was provided just to prevent SG dryout, however, the reason is to keep the tubes covered for secondary heat removal.

Sys# System Category KA Statement 035 Steam K4 Knowledge of S/GS design feature(s) and/or interlock(s) which Amount of reserve water in SIG Generator provide for the following:

System (S/GS)

KIA# K4.05 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.E-O lss. 2 Rev. 1 pg. 16 20M-24.1.D Rev. 6 pg. 16 20M-24.2.B Rev. 16 pg. 4 Question Source: Bank - Comanche Peak 2013 NRC Exam (035)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

63. Plant conditions have been established to perform 20ST-26.8, "Main Turbine Overspeed Trip Test".

• Reactor power is stable at 13%

• The generator is NOT synchronized to the grid

• Overspeed Protection Test Switch is in the "lnservice" position

• As Turbine speed is raised, Annunciator A7-2E "Turbine Overspeed Prot Controller Operating" alarms

• SIG pressures are 1000 psig and stable When Turbine speed reaches the overspeed protection controller setpoint, which Turbine EHC valves will close?

(IV =Intercept valves, GV =Governor valves, TV =Throttle valves)

A. Only IVs and GVs Close B. Only GVs, and TVs Close C. Only TVs and IVs Close D. ALL IVs, GVs, and TVs Close Answer: A Explanation/Justification: KJA is met by the understanding that when testing the turbine overspeed, and the annunciator for "Turbine Overspeed Prot Controller Operating" alarms, that only the IVs and GVs will be effected by the OPC.

A. Correct. At 103% (1844-1864 rpm) the Overspeed Protection Controller actuates causing all 4 GVs and all 4 IVs to close. The TVs are not closed by the OPC. Position indication of Turbine valves can be monitored on BB-C.

B. Incorrect. The TVs are not closed by the OPC which makes the distractor incorrect.

C. Incorrect. It is correct that the IVs will close, but the TVs are not closed by the OPC.

D. Incorrect. The TVs are not closed by the OPC which makes the distractor incorrect.

Sys# System Category KA Statement 045 Main Turbine A4. Ability to manually operate and/or monitor in the control room: Turbine valve indicators (throttle, governor, Generator control, stop, intercept), alarms, and annunciators (MT/G)

System KIA# A4.01 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

20M-26.4.AAU lss.1 Rev. 5 20ST-26.8 Rev. 16 pg.15 Question Source: Bank - Vision #138661 Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

64. The following conditions exist:

• Plant is at 80% power

• Control Room ACU Outside Air Intake and Exhaust Dampers 2HVC*MOD201 B & D automatically CLOSED.

• Control Room Emergency Supply Fan [2HVC*FN241 B] automatically STARTS 120 seconds after the Control Room isolation occurs Which of the following radiation monitors, and setpoint would cause the above ventilation lineup?

A. Control Room Area [2RMC*RQ201] radiation monitor above the ALERT setpoint.

B. Control Room Area [2RMC*RQ202] radiation monitor above the HIGH setpoint.

C. Control Room Airborne Particulate [2RMC-RQ301A] radiation monitor above the ALERT setpoint.

D. Control Room Airborne Gas [2RMC-RQ301 B] radiation monitor above the HIGH setpoint.

Answer: B Explanation/Justification: KIA is met by re-aligning the Control Room ventilation system and having the candidate demonstrate the ability to determine which radiation monitor alarm would cause the automatic ventilation alignment.

A. Incorrect. Plausible distractor because RQ201 does initiate CR isolation at the HIGH setpoint, but not at the alert setpoint.

B. Correct. RQ202 does initiate a CR isolation when at High setpoint.

C. Incorrect. Plausible distractor with it being a CR rad monitor. RQ301A will not initiate CR isolation.

D. Incorrect. Plausible distractor with it being a CR rad monitor. RQ301 B will not initiate CR isolation.

Sys# System Category KA Statement 072 Area Generic Ability to verify that the alarms are consistent with Radiation the plant conditions.

Monitoring (ARM)

System KIA# 2.4.46 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

20M-43.4.ADB Rev.7 pg. 2 20M-43.1.C Rev. 5 pg. 24 1/20ST-43.17D Rev. 44 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.3 /

45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

65. Initial conditions:

• The plant is at 100% power

• 20ST-30.6B, Service Water Pump [2SWS*P21C] Test on Train B Header is in progress

• 2SWS*P21 C, 'C' SWS is RUNNING on 2DF Bus

• 2SWS*P21 B, 'B' SWS is racked on the bus with the CS in Auto-After Stop Current Conditions:

• A Loss of Offsite Power occurred following a design basis earthquake

• The Control Room crew is performing actions of E-0, "Reactor Trip or Safety Injection"

• A8-2B, 4160V EMER BUS 2AE ACB 2E7 OVERCURRENT TRP is LIT With NO Operator action, one minute after the Loss of Offsite Power, which of the following statements describe the status of the Service Water Pumps?

A. 2SWS*P21A and 2SWS*P21 Bare running B. 2SWS*P21A and 2SWS*P21C are running C. Only 2SWS*P21 B running D. Only 2SWS*P21C running Answer: C Explanation/Justification: KIA is met by demonstrating the knowledge of the available power and pump start interlocks of the essential SWS pumps following a loss of offsite power coincident with a failure of 2AE emergency bus to load.

A. Incorrect. 'A' SWS pump will not be energized due to bus 2AE being de-energized with A8-2B annunciator lit. 2-1 DG will start but 2E10 will not close due to the overcurrent trip on 2E7. The candidate must know the effects of this annunciator condition. 'B' SWS will auto start even though the CS is in Auto-After Stop and the 'C' SWS is racked in on the DF bus.

B. Incorrect. 'A' SWS pump will not be energized due to bus 2AE being de-energized with A8-2B annunciator lit. 2-1 DG will start but 2E10 will not close due to the overcurrent trip on 2E7. The candidate must know the effects of this annunciator condition. 'C' SWS will not auto start even though power is available and it was running, due the 'B' SWS pump being racked onto the 2DF bus with the CS in Auto-After Stop. It is still the priority pump and will be loaded on the diesel.

C. Correct. With the CS in Auto-After Stop, 'B' SWS pump is the priority pump and will be loaded on the diesel even though the 'C' SWS is racked onto the 2DF bus.

D. Incorrect. Even though 'C' SWS pump has power available and was running, 'B' SWS pump will auto start and load onto the diesel because it the riorit um .

Sys# System Category KA Statement 075 Circulating K2 Knowledge of bus power supplies to the following: Emergency/essential SWS pumps Water System KIA# K2.03 KIA Importance 2.6* Exam Level RO References provided to Candidate Technical

References:

20M-30.1.D Rev. 8, Pgs. 2-4 None U2 LSK-017-001A Rev. 14 20M-36.4.ACD Rev. 3 pg. 3 3SQS-36.1 PPNT U2 Rev. 12 lss. 1 Slide 10 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

66. The plant is in Mode 5 preparing to enter Mode 4.

• Valve alignments are being performed on a Safety-Related system

• The REQUIRED NSA position of a manually operated globe valve is OPEN

• The valve must be in this position PRIOR to Mode 4 entry

• The Independent Verifier will receive 8 mR performing the Independent Verification (IV)

IAW the guidance provided in NOP-OP-1002, Conduct of Operations, how could the Independent Verification for this valve be addressed?

A. The Operations Manager has the authority to waive the IV for equipment concerns.

B. The IV may be performed by using the Plant Computer System (PCS) if "Not Closed" is indicated.

C. The Shift Manager can waive the IV due to dose limits.

D. The IV may be performed by a functional test that can prove the valve is open.

Answer: D Explanation/Justification: KIA is met by demonstrating the knowledge of alternative independent verification means for a valve lineup in accordance with the Conduct of Operations manual.

A. Incorrect. Plausible distractor because it may be assumed that the Operations Manager would have this authority, but that is not correct.

B. Incorrect. Plausible distractor but not a reliable alternative to hands on verification. The valve is required to be OPEN, but the remote indication of NOT-CLOSED only means that the valve is not fully closed.

C. Incorrect. Plausible distractor because the SM can waive the IV if it would result in a radiation exposure greater than 1O mRem.

D. Correct. A functional test may be used for an IV iaw NOP-OP-1002 sect. 4.18.2.6.

Sys# System Category KA Statement N/A N/A Generic Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc KIA# 2.1.29 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-1002 Rev. 10 pg. 81 Question Source: Bank - 2LOT6 NRC Exam (066) Modified (2009)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 45.1 I 45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

67. Given the following plant conditions:

• The plant is operating at 90%

• VCT level is 38% and stable Then VCT Level Transmitter, 2CHS*LT112 fails HIGH Which of the following completes the statements below?

(Assume NO operator action)

1) VCT level on 2CHS*LT115 will ( 1)

2) VCT Auto Makeup will be (2)

A. 1) lower

2) available B. 1) lower

2) unavailable C. 1) remain unchanged

2) available D. 1) remain unchanged

2) unavailable Answer: A Explanation/Justification: KJA is met by the candidate recognizing how the system will respond to one VCT level control indication failing high, and validating how the other VCT control indication will respond to this failure.

A. Correct. With LT112 failing high, both LCV112 and LCV115A will reposition to divert and lower the actual level. As VCT level lowers to 20%,

LT115 will start Auto Makeup and try to maintain level 20-40%.

B. Incorrect. Actual level will lower. LT115 is unaffected, therefore as actual level lowers, Auto Makeup will start to maintain level 20-40%.

C. Incorrect. Actual level remaining unchanged would be true if LT112 failed low. It is correct that auto makeup will be available.

D. Incorrect. Actual level remaining unchanged would be true if LT112 failed low. It is incorrect that auto makeup will be available.

Sys# System Category KA Statement N/A N/A Generic Ability to identify and interpret diverse indications to validate the response of another indication.

KIA# 2.1.45 KIA Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-7.4.IF Rev. 3 Att. 1 Question Source: Bank- Harris 2012 NRC Exam (Q67)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 43.51 45.4)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

68. You are performing a procedure out in the plant and you note a typographical error in the step you are about to perform. In accordance with NOP-LP-2601, 'Procedure/Work Instruction Use and Adherence', what action are you required to perform.

A. Have a second qualified Operator peer check the typographical error, continue with the activity, and inform your supervisor upon completion.

B. Contact your supervisor, identify the typographical error, have the supervisor annotate issue in the procedure, and then continue with the activity.

C. Contact your supervisor, identify the typographical error, and perform a Limited Use Change.

D. Contact your supervisor, identify the typographical error, and Revise the procedure.

Answer: B Explanation/Justification: KIA is met with the knowledge of the expected response when a procedure is found to have a typographical error, and how to make the necessary changes to the procedure prior to completing work.

A. Incorrect. By continuing on in the procedure without discussing it with the authorizing authority or responsible supervisor would be a violation of NOP-OP-2601. A peer check by another qualified Operator does not meet the site expectations.

B. Correct. Per NOP-LP-2601, if a typo is discovered, the performer must stop the work, ensure equipment is in a safe condition, and contact they're supervisor. Clearly identify typo by annotating the procedure and then continue with the activity.

C. Incorrect. A Limited Use Change is not required for a typographical error D. Incorrect. A procedure Revision is not required for a typographical error.

Sys# System Category KA Statement NIA N/A Generic Knowledge of the process for making changes to procedures.

KIA# 2.2.6 KIA Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

NOP-LP-2601 rev.5, pg.15 & 16 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.3 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

69. Given the following plant conditions:

• The Unit is operating at 100%.

• You have just returned from a day off and are reviewing the narrative logs.

• 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago, a valve was repositioned out of NSA and selected as an OPEN item using the Short Term Configuration Change Process.

Which of the following statements correctly describes requirements of NOP-OP-1014, "Plant Status Control?

A. A clearance will be necessary if restoration does not occur within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. A system status print sheet will be necessary if restoration does not occur within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. A clearance should have been posted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

D. A system status print sheet should have been issued 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

Answer: C Explanation/Justification: KIA is met by the ability to determine a valve has exceeded the short term configuration change process, and identify the correct actions that should have been take in accordance with NOP-OP-1014, Plant Status Control procedure.

A. Incorrect. Refer to correct answer explanation. The candidate may believe the requirement is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> as opposed to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Incorrect. Refer to incorrect choice D explanation. Plausible and balanced distractor.

C. Correct. According to NOP-OP-1014, if a component is not restored to its normal configuration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then a clearance is hung to provide a plant status control tracking method and documentation of the deviation from the components normal alignment. A clearance should have been posted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

D. Incorrect. A System Status Print is required to be filled out at all times reflecting system status conditions, if the system is deemed necessary by the Ops Manager. If it was not deemed necessary, the system status print would not be required. If it was deemed necessary, then it should have been filled out 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago.

Sys# System Category KA Statement NIA N/A Generic Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

KIA# 2.2.15 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-1014, Rev. 4, pg. 14 Question Source: Bank - 1LOTS NRC EXAM (Q96) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.3 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

70. The plant is at 100% power.

Which of the following conditions or events (considered individually) will require Technical Specification action(s) to be performed within one hour or less?

A. RWST borated water temperature drops to 50 °F.

B. One Containment Pressure Transmitter fails to zero.

C. RWST borated water volume drops to 840,200 gallons.

D. BOTH Train "A" - Phase B (CIB) manual Control Switches are declared inoperable.

Answer: C Explanation/Justification: KIA is met by the knowledge required to recognize the RWST level is below Tech Spec require level and is as; 1 hr. TS action statement.

A. Incorrect. TS 3.5.4 Surveillance requires RWST borated water temperature to be <!45 F ands; 65 F. therefore there is no TS LCO entry required for this distractor.

B. Incorrect. TS 3.3.2 Condition D & E apply. The channel is required to be placed in trip/bypass within 72 hrs.

C. Correct. TS 3.5.4 Condition B states that if RWST is inoperable for reasons other than boron concentration or temperature (Condition A), then a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statement is applicable. SR 3.5.4.2 requires Unit 2 RWST level to be <! 859248 gallons. If this surveillance is not met then TS LCO actions apply. RO's are required to knows; 1hour TS LCO's from memory.

D. Incorrect. TS 3.3.2 Condition B applies. This is a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> action statement.

Sys# System Category KA Statement N/A N/A Generic Knowledge of less than or equal to one hour Technical Specification action statements for systems.

KIA# 2.2.39 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

BVPS TS pg. 3.5.4.1 & 2 Amend. 278/161 Question Source: Bank- 1LOTS NRC Exam (041)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.2 /

45.13)

Objective: 2SQS-13.1 Rev. 18 Obj. 18 - For a given set of plant conditions, determine if the condition meets the criteria for entry into a one hour or less action statement in accordance with the Technical Specifications.

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

71. You are going into a contaminated area, which has the following radiological characteristics to perform a valve lineup.

• Your current exposure for the year is 938 mrem

• The RWP states:

o General area dose rate =30 mrem/hr o Airborne contamination concentration =10.0 DAC

• The valve lineup will take you 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if you wear a full-face respirator.

• The valve lineup will only take you 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if you do NOT wear the respirator.

1) Which of the following choices for completing this job would maintain your exposure within the station administrative requirements and the principles of ALARA?

2) Why is this action appropriate?

A. 1) You must wear the respirator.

2) You will exceed DAC limits if you do NOT wear a respirator.

B. 1) You must wear the respirator.

2) Your calculated TEDE dose received will be less than if you do NOT wear a respirator.

C. 1) You should NOT wear the respirator.

2) Your calculated TEDE dose received will be less than if you do wear a respirator.

D. 1) You should NOT wear the respirator.

2) Your dose received wearing a respirator will exceed the site annual personnel dose limits.

Answer: C Explanation/Justification: K/A is met by demonstrating the ability to comply with an RWP to determine dose received with or without a respirator to achieve the lowest possible dose for a job.

A. Incorrect. This answer is plausible if the applicant does not understand the concept of DAC-hours and DAC-hour limits.

B. Incorrect. This answer is plausible if the applicant incorrectly calculates the exposure.

C. Correct. Without respirator: TEDE = 30 mrem/hr x 1 hr= 30 mrem, From airborne contamination: TEDE = 10 DACx1 hr x 2.5 mrem/DAC-hr = 25 mrem, TEDE = 30 + 25 = 55 mrem from job, Total exposure for year= 938 + 55 = 993 mrem With respirator, TEDE = 30 mrem/hr x 2 hr= 60 mrem TEDE = 60 mrem, Total exposure for year= 938 + 60 = 998 mrem TEDE = 60 mrem-vs-55 mrem = do not use a respirator D. Incorrect. This answer is plausible if the applicant miscalculates the dose.

Sys # System Category KA Statement N/A NIA Generic Ability to comply with radiation work permit requirements during normal or abnormal conditions.

KIA# 2.3.7 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

NOP-OP-4201 Rev. 2 pg. 20 FENRWT Rev 3 CNRR 08-08-14 Handout pg.26 & 64 Question Source: Bank-McGuire 2012 NRC Exam (Q72)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.12 I 45.10)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

72. The following conditions exist:

• Plant is operating at 100% power

• Radiation level is 1,324 mrem/hr at 30 centimeters from the Letdown piping

• You have been assigned to enter the Letdown cubicle and hang a clearance Which of the following identifies the radiation area posting at the cubicle entrance, and the minimum approval authority for entry in accordance with NOP-OP-4101, 'Access Controls for Radiologically Controlled Areas'?

Letdown Cubicle Posting Minimum Approval A. High Radiation Area (HRA) Radiation Protection Manager B. High Radiation Area (HRA) Radiation Protection Supervisor C. Locked High Radiation Area (LHRA) Radiation Protection Manager D. Locked High Radiation Area (LHRA) Radiation Protection Supervisor Answer: D Explanation/Justification: KIA is met by identifying the area as a LHRA and determine who must give permission to enter the LHRA in order to hang a clearance.

A. Incorrect. HRA is An accessible area in which radiation levels could result in an individual receiving a deep-dose equivalent in excess of ;?100 mrem/hr at a distance of 30 centimeters or more from a radiation source or from any surface that the radiation penetrates. The RPM approval is only required if the gen area dose was >2.5 rem/hr, or it was a Very High Rad Area.

B. Incorrect. For posting, see explanation above. Radiation Protection Supervisor is the correct authorization.

C. Incorrect. LHRA is the correct posting. It is incorrect that the RPM must give permission. The RPM approval is only required if the gen area dose was >2.5 rem/hr, or it was a Very High Rad Area.

D. Correct. LHRA is A locked area with an accessible area to individuals, in which radiation levels could result in dose rates ;?1,000 mrem/hr at a distance of 30 centimeters from a radiation source or from any surface that the radiation penetrates. The RP Supervisor must give approval for entry into the LHRA as long as the general area dose rate is <2.5Rem/hr.

Sys # System Category KA Statement NIA N/A Generic 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.

KIA# 2.3.13 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical

References:

NOP-OP-4101 Rev. 11 Pg. 5 & 17 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.12 I 43.4 I 45.9 I 45.10)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

73. The plant is operating at 100% power.

• A Small Break LOCA occurs.

• The crew is performing the actions of ES-1.2, "Post LOCA Cooldown and Depressurization".

• All SI pumps are running.

• All RCPs are running.

• RCS cooldown via Condenser Steam Dumps is ongoing.

• RCS Tavg is 510°F and lowering at a rate of 50°F/Hr.

• RCS pressure is 1350 psig and stable.

• Pressurizer (PRZR) level indicates 38% and rising.

Which of the following describes the NEXT MAJOR action to be implemented in the EOP to mitigate the current conditions?

A. Depressurize the RCS using normal spray to minimize RCS subcooling.

B. Stop the cooldown. Energize all PRZR heaters to collapse voids and stabilize PRZR level.

C. Transition to ES-1.1, "SI Termination" and begin the SI flow reduction sequence by stopping ECCS pumps.

D. Stop RCP's NOT needed for PRZR Spray and begin the SI flow reduction sequence by stopping ECCS pumps.

Answer: D Explanation/Justification: KIA is met by demonstrating the knowledge of the major action steps of ES-1.2, Post LOCA Cooldown and Depressurization to mitigate the event.

A. Incorrect. This is the fifth major action step (EOP step 23) performed after normal charging has been re-established. Plausible because the plant has just been depressurized to raise przr level to >31 % (EOP step 15) by understanding the stem information.

B. Incorrect. ES-0.1 cools the plant down to mode 5 condition so there is no need to stop cooldown unless 1OOFlhr was exceeded. No voids exist at the current time with RCPs running. This would be performed if a void existed in ES-0.2 or ES-0.3.

C. Incorrect. ES-1.1 is a plausible distract since it terminates SI. In the case of ES-1.2, the steps to terminate Si are incorporated in EOP steps 17-21.Candidate must know that reducing SI is a major action of the procedure.

D. Correct. The depressurization to raise przr level to >31% is complete (major action step 2), therefore the next major action step is to stop all but one RCP and reduce RCS injection flow (steps 3 & 4)

Sys# System Category KA Statement NIA NIA Generic Knowledge of EOP mitigation strategies.

KIA# 2.4.6 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.2 lss. 2, Rev. 1, steps 16 & 17 Question Source: Bank - 2LOT8 Audit Exam (058)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

74. Given the following conditions:

• A small fire was discovered in the Unit 2 Control Room

• AOP-2.33.1A, "CONTROL ROOM INACCESSIBILITY" has been implemented

• All Control Room actions are complete

• All equipment operated as expected In accordance with AOP-2.33.1A, what is the Unit 2 Balance of Plant (BOP) role during this event?

A. Emergency Squad B. Communicator I N0#3 C. Alternate Shutdown Panel (ASP)

D. Emergency Shutdown Panel (SOP)

Answer: B Explanation/Justification: KIA is met by demonstrating the knowledge of the licensed operator rules during a fire in the control room in accordance with the Control Room Inaccessibility AOP.

A. Incorrect. Emergency Squad is the required role of the Turbine & PAB Operators.

B. Correct. BOP Operator is required to be the Communicator/Nuclear Operator #3 in accordance with Attachment 6 of AOP-2.33.1A.

C. Incorrect. Alternate Shutdown Panel is not manned during this event. It could be applicable if 20M-56C was implemented.

D. Incorrect. Emergency Shutdown Panel is manned by the Unit Supervisor and Reactor Operator.

Sys# System Category KA Statement N/A N/A Generic Knowledge of fire protection procedures.

KIA# 2.4.25 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.33.1A Rev. 15 pg. 35 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

75. The Plant was operating at 100% power:

• At time 1205 a Reactor Trip and several events occurred

• At time 1215 an ALERT is declared by the Shift Manager

• At time 1225 the Initial Notification Form is completed and approved by the Shift Manager Which of the following identifies the LATEST time that the initial notification to State and County officials is due?

A. 1220 B. 1230 C. 1235 D. 1240 Answer: B Explanation/Justification: KIA is met with Licensed Operator knowledge of the CR Communicator responsibilities and the required times to complete the initial notification to state and county officials.

A. Incorrect. This is the time at which the declaration must be made by the Shift Manager (SM).

B. Correct. Per 1/2-EPP-IP-1.1, Initial Notifications are to be made to the first six (6) listed Agencies of the Emergency Notification Call List (State and County), and MUST be made within 15 minutes of the event declaration.

C. Incorrect. The SM has 15 minutes to declare the event and then 15 minutes from declaration to notify the state and counties. This theoretically gives them 30 minutes to make a notification. However, since the declaration was made at 1215 the notification must be made by 1230. This distractor is based on 30 minutes from 1205.

D. Incorrect. This distractor is based on 15 minutes incorrectly added to the time the INF form was completed and approved. The notification must be made within 15 minutes of the event declaration.

Sys# System Category KA Statement NIA N/A Generic Knowledge of RO responsibilities in emergency plan implementation.

KIA# 2.4.39 KIA Importance 3.9 Exam Level RO References provided to Candidate none Technical

References:

1/2-EPP-IP-1.1 Rev. 51 1/2-EPP-IP-1.1.F02 Rev.19 Question Source: Bank - Robinson NRC 2011 (074)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 45.11)

Objective:

ES-401 Form ES-401-8 Cover Sheet U. S. Nuclear Regulatory Commission Site-Specific SRO Written Examination BV2LOT15 SRO Written Examination Applicant Information Name:

Date: Facility/Unit: Beaver Valley Unit 2 Region: I l:R1 11 0 111 0 IV 0 Reactor Type: W l:R1 CE 0 BW 0 GE 0 Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> If you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results RO I SRO-Only I Total Examination Values 75 /__ 25 l 100 Points Applicant's Scores I I Points Applicant's Grade I I Percent NUREG-1021, Revision 10 FENOC Facsimile r2

(SRO ONLY}

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

76. The crew is performing ES-0.1, 'Reactor Trip Response' after an inadvertent Reactor trip from 100% power.

10 minutes after the Reactor Trip:

• A 200 gpm small break LOCA occurs

• The ATC operator notes PRZR level is at 20% and lowering

• 2CHS*FCV122 'Charging Pumps Disch Flow Control Viv' is in MANUAL and Full OPEN

• 2CHS*FI 122 indicates 150 gpm and steady

• Net Charging on PCS indicates 60 gpm

• Assume RCS Pressure remains constant during the event

• No automatic ESF actuation conditions are met

• All systems operate as designed

1) With NO Operator action, approximately, how long before the PRZR level indicates 0%?

2) The Unit Supervisor will transition from ES-0.1 to which of the following procedures?

A. 1) 15 minutes

2) E-1, "Loss of Reactor or Secondary Coolant" B. 1) 15 minutes

2) E-0, "Reactor Trip or Safety Injection" C. 1) 45 minutes

2) E-1, "Loss of Reactor or Secondary Coolant" D. 1) 45 minutes

2) E-0, "Reactor Trip or Safety Injection"

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 76 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant status and determine which procedure to transition too based upon the LHP criteria of the procedure. Detailed knowledge of the procedure is required to select the correct procedure actions. The SRO is required to review the Left Hand Page information periodically during procedure implementation and direct the crew to implement actions when conditions are met.

KIA is met by determining how the PRZR level control system will operate during a SBLOCA, and determine how long it will take for the PRZR to indicate empty. SRO level of knowledge of LHP to initiate SI at 4% and transition to E-0.

A. Incorrect. This would be the time if letdown didn't isolate at 14%. Incorrect procedure transition. Plausible procedure choice if candidate is thinking of the LHP requirements of ES-1.1 SI Termination, which states to manually start SI and transition to E-1, Loss of Reactor or Secondary Coolant at 17% PRZR level.

B. Incorrect. This would be the time if letdown didn't isolate at 14%. Correct procedure transition per ES-0.1 LHP.

C. Incorrect. This is the correct time. Incorrect procedure transition. Plausible procedure choice if candidate is thinking of the LHP requirements of ES-1.1 SI Termination, which states to manually start SI and transition to E-1, Loss of Reactor or Secondary Coolant at 17% PRZR level.

D. Correct. Candidate must evaluate the initial net charging and the leak rate to determine the RCS is losing 140gpm. After letdown isolates at 14%, net charging will rise to 165 gpm, with RCS losing 35gpm. 20-14% (600gal)@ 140gpm=4.3 min until UD isolates, then 14-0% (1400 gal)@

35 gpm = -40 min. In ES-0.1 LHP states to actuate SI and go to E-0 if PRZR level cannot be maintained >4%.

Sys # System Category KA Statement 000009 Small Break EA2 Ability to determine or interpret the following as they apply to a The time available for action before PZR is empty, LOCA I 3 small break LOCA: given the rate of decrease of PZR level KIA# EA2.05 KIA Importance 3.9 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.ES-0.1 lss. 2 Rev. 3 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

77. The plant was operating at 100% power when a large break LOCA occurred coincident with a Loss of 4KV Bus DF.

The follow conditions exist:

• 'A' Quench Spray Pump [2QSS-P21A] tripped on startup

• 3 Max CETs indicate 810°F

• RCS is superheated

• CNMT Pressure is 31 psig

• CNMT Temperature is 240°F

• All RCPs have been tripped

• RVLIS Full Range indicates 35%

Based on the above conditions, which answer below completes the following statement?

The required EAL classification is based upon the _ _ _ _ _ __

A. LOSS of one fission product barrier and POTENTIAL Loss of another barrier B. LOSS of two fission product barriers C. LOSS of two fission product barriers and a POTENTIAL loss of a third barrier D. LOSS of all three fission product barriers

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 77 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Classifications. This is a SRO position function only.

KIA is met by demonstrating the knowledge to determine the event classification based on the conditions given using the provided EPP classification chart.

Candidate will have to recognize a General Emergency would be declared based on the following conditions which they will have to interpret from the conditions given in the stem.

FC - Loss due to FR-C.1 Red Path Entry RCS - Loss RCS leak rate greater than available makeup capacity as indicated by RCS subcooling < 46° F adverse containment.

CT - Potential Loss due to Cnmt pressure >11 psig AND less than one full train of depressurization equipment operating.

A. Incorrect. Plausible if it is not recognized that FR-C.1 entry conditions have been met for Fuel Clad failure, or a Loss based on RCS Leak Rate.

B. Incorrect. Plausible if it is not recognized that Potential Loss due to Cnmt pressure >11 psig AND less than one full train of depressurization equipment operating C. Correct. GE based on answer explanation above.

D. Incorrect. Plausible if it is not recognized that containment barrier is a potential loss, and not a loss. This would identify a weakness of CT-8 based on pressure response not consistent with LOCA conditions.

Sys # System Category KA Statement 000011 Large Break Generic Knowledge of the emergency action level LOCA I 3 thresholds and classifications.

KIA# 2.4.41 KIA Importance 4.6 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.F-0.2 lss. 2 Rev. 1 EPP Chart EPP-1-1b.F01 Rev. 0 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.11)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

78. Given the following conditions:

• A reactor trip and safety injection have occurred

• The crew is performing the actions of E-2, "Faulted Steam Generator Isolation" due to the uncontrolled depressurization of 'A' SG.

• The crew is evaluating if SI flow should be reduced.

• The following conditions exist:

o RCS temperature is 460°F o RCS pressure is 1650 psig and slowly rising o Containment pressure is 23 psig o SG 218 and 21C NR levels are 15% and rising o AFW flow is 375 gpm o PRZR level is 20%

Based on the conditions above, when may the crew enter ES-1.1, "SI Termination"?

A. Immediately.

B. After transition to E-1, when RCS subcooling criteria is met.

C. After transition to E-1, when PRZR level criteria is met.

D. After transition to E-1, when Secondary heat sink criteria is met.

Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant status and determine if the conditions are met to terminate Safety Injection and the required procedure transitions within the EOP network. Detailed knowledge of the procedure is required to select the correct transition and the requirements for SI termination.

KIA is met by interpreting the given conditions to determine when SI Termination is permitted. Detailed knowledge of SI Terminations and procedural transitions is required for the SRO.

A. Incorrect. PRZR level criteria is not high enough for the adverse CNMT conditions (38% req.). Plausible distractor because transition to ES-1.1 from E-2 occurs immediately after checking PRZR level. Evaluation of adverse CNMT must be determined.

B. Incorrect. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since it is a faulted SG, subcooling requirements were easily met, but PRZR level is not.

C. Correct. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since PRZR level is still low for adverse CNMT (38% req.) a transition to ES-1.1 must wait until PRZR level is met in E-1.

D. Incorrect. It is correct that a transition to E-1 is required from E-2, because ES-1.1 requirements were not met at the step in E-2. E-1 continuous action step 8 is the only transition to ES-1.1 from E-1. Since SG level does not meet the adverse requirement of 31 % this is a plausible distractor.

Heat sink is met with AFW flow >340 gpm, but PRZR level is not.

Sys # System Category KA Statement 000040 Steam Line AA2 Ability to determine and interpret the following as they apply to When ESFAS systems may be secured Rupture I 4 the Steam Line Rupture:

KIA# AA2.05 KIA Importance 4.5 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.E-2 lss. 2 Rev 0 None 20M-53A.1.E-1lss.2 Rev. 2 Question Source: Bank - 2LOT5 NRC Exam (047)(2005)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

79. The plant is in Mode 3 with all systems in normal alignment for this Mode.

• Battery Breaker 2-1 [BAT*BKR2-1] is on Clearance for Electrical Maintenance to replace the Battery Breaker for maintenance.

• Annunciator A8-9A, "125V DC Bus 2-1 TROUBLE" re-flashes due to Battery Charger 2-1 AC Input Breaker tripping open.

Which of the following Tech Spec LCOs will be applicable?

1) 3.8.1, AC Sources - Operating

2) 3.8.7, Inverters - Operating

3) 3.8.9, Distribution Systems - Operating A. None B. 1 &2 ONLY C. 2 & 3 ONLY D. 1, 2, 3 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine which TS are applicable. Detailed knowledge of the bases is required to determine the impact of the loss of the power supplies and which TS are applicable.

KIA is met by demonstrating Tech Spec bases knowledge for the effected equipment when a loss of DC bus occurs. DC bus loss will effect Inverters, and the EOG start capabilities. This is TS bases knowledge.

A. Incorrect. Plausible distractor because TS 3.8.4, DC Sources Operating was intentionally omitted from the above list. Candidate must know the TS bases for all the listed TSs to correctly answer the question.

8. Incorrect. Plausible distractor if the bases for TS. 3.8.9 is not known. The bases states that DC subsystems require the associated buses and distribution panels to be energized to their correct voltage from either the associated battery or charger.

C. Incorrect. Plausible distractor if the bases for TS 3.8.1 is not known. The bases states that each DG must be capable of starting and loading.

With DC bus 1 de-energized, the diesel starting circuits and load sequencer are not capable of performing their function.

D. Correct. All of the TSs are applicable. The bases for TS 3.8.1 and 3.8.9 are described above. TS 3.8.7 bases states that an inverter can be supplied from an internal AC source via a rectifier as long as the battery is available. However, in the stem it stated that the 2-1 battery breaker was on clearance for maintenance.

Sys# System Category KA Statement 000058 Loss of DC Generic Knowledge of the bases in Technical Power/6 Specifications for limiting conditions for operations and safety limits.

KIA# 2.2.25 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

U2 RE-0001AR Rev. 22 TS bases 3.8.1, 3.8.7, 3.8.9 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 41.7 I 43.2)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

80. The plant was operating at 100% power.

• A LOCA OUTSIDE containment occurs

• At step 21 of E-0, Reactor Trip Or Safety Injection, the crew enters ECA-1.2, LOCA Outside Containment

• At the completion of ECA-1.2, the crew has been UNABLE to locate and isolate the break The following plant conditions exist:

• Offsite Power has been lost

• All SG pressures are 800 psig and stable

• All SG NR levels are 35% and slowly rising

• All Secondary radiation monitors are consistent with pre-event values

• CNMT parameters are consistent with pre-event

• RCS Subcooling is 40°F and slowly dropping

• RCS Pressure is 1125 psig and slowly dropping

• PRZR level is 12% and slowly dropping

• Auxiliary Building Radiation levels are rising

• Auxiliary Building sump levels are rising

• A seismic event of 0.07g has been recorded Based on these conditions and events:

1) What procedural transition from ECA-1.2 is REQUIRED?

2) Of the choices listed below, which of the Abnormal Operating Procedures will be performed in conjunction with the EOP network?

A. 1) ECA-1.1, Loss Of Emergency Coolant Recirculation

2) AOP-2.36.1, Loss Of All AC Power When Shutdown B. 1) ECA-1.1, Loss Of Emergency Coolant Recirculation

2) AOP-1/2.75.3, Acts of Nature - Seismic Event C. 1) E-1, Loss Of Reactor Or Secondary Coolant

2) AOP-2.36.1, Loss Of All AC Power When Shutdown D. 1) E-1, Loss Of Reactor Or Secondary Coolant

2) AOP-1/2.75.3, Acts of Nature - Seismic Event

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 80 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the ineffective break isolation steps. This evaluation requires detailed knowledge of the EOP procedure flow-paths of sub procedures. Detailed knowledge of the procedure is required to select the correct transition.

K/A is met by demonstrating knowledge of entry conditions into abnormal operating procedures based on given indications while responding to a LOCA Outside Containment. In the question the SRO is given a seismic indication greater than the Alarm Response Procedure (ARP) entry setpoint, and must determine that the ARP is an entry condition into the Seismic Event Abnormal Operating Procedure. This AOP is performed in conjunction with the EOP network.

A. Incorrect. Correct EOP transition. Incorrect AOP. AOP-2.36.1, Loss Of All AC Power When Shutdown is a plausible distractor with the crew performing ECA-1.2, LOCA Outside CNMT, then losing offsite power. The candidate must determine that entry conditions of Rx vessel defueled or RHR being used to control RCS temperature, are not met for entry into this AOP.

B. Correct. If RCS pressure is not rising, then JAW ECA-1.2 step 4 RNO transition must be made to ECA 1.1. Seismic Event AOP is correct due to the event registered 0.07g which is greater than ARP setpoints (A 10-5H). Entry into the Seismic Event AOP is required based upon annunciator A 10-5H, a report from NEIC, or felt or observed ground movement by plant personnel.

C. Incorrect. Plausible since E-1 would be the appropriate entry if RCS pressure were rising. Incorrect AOP as explained in answer 'A'.

D. Incorrect. Plausible since E-1 would be the appropriate entry if RCS pressure were rising. Correct AOP entry.

Sys # System Category KA Statement W/E04 LOCA Outside Containment I 3 Generic Knowledge of abnormal condition procedures.

KIA# 2.4.11 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ECA-1.2 lss. 2 Rev. 0, pg. 3 20M-45B.4.AAA Rev. 8, pg. 3 1/20M-53C.4A.75.3 Rev. 19, pg. 1 Question Source: Bank- 2LOT6 NRC Exam (081) Modified (2009)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

81. The plant is at 100% power.

• A loss of all Feedwater occurs

• The crew enters FR-H.1, Response to Loss of Secondary Heat Sink The following conditions now exist:

• FR-H.1 Step 6 Stop All RCPs has just been completed

• 'A' SG Wide Range Level is 10%, pressure is 1000 psig and stable

• 'B' SG Wide Range Level is 19%, pressure is 975 psig and stable

• 'C' SG Wide Range Level is 12%, pressure is 600 psig and lowering

• Containment Pressure is 4.0 psig and stable (1) Which of the following actions are REQUIRED based upon these indications?

(2) Per Tech. Specs., with the plant in Mode 3, what is the MINIMUM water level required to consider a Steam Generator OPERABLE as a heat sink?

A. (1) Transition to E-2, Faulted Steam Generator Isolation (2) 12% Narrow Range B. (1) Initiate RCS Bleed and Feed (2) 12% Narrow Range C. (1) Transition to E-2, Faulted Steam Generator Isolation (2) 15.5% Narrow Range D. (1) Initiate RCS Bleed and Feed (2) 15.5% Narrow Range Answer: D Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. 11.E pg 6 and SRO level knowledge of TS bases for the Surveillance requirements. The first part requires an understanding of the EOP mitigation strategy which is RO level knowledge, however the SRO must assess plant conditions, determines if adverse criteria is applicable, and selects the section of the procedure to mitigate the event. The TS minimum level is SRO level knowledge since the level required for operability is NOT addressed in the LCO rather is addressed in the bases and the surveillance requirement.

KIA is met by interpreting the given conditions of a loss of secondary heat sink, then determining the appropriate procedural actions based on the conditions.

A. Incorrect. This transition is possible since the 'C' SG pressure is lowering, however Bleed and Feed criteria are met. EOP Rules of usage does not allow for exit until FR-H.1 is complete. This SG NR level is the minimum in the EOP network not TS.

B. Incorrect. Bleed and Feed criteria are met. This SG NR level is the minimum in the EOPs not TS C. Incorrect. This transition is possible since the 'C' SG pressure is lowering, however Bleed and Feed criteria are met. EOP Rules of usage does not allow for exit until FR-H.1 is complete. Correct TS SG level.

D. Correct. Bleed and Feed criteria are met per continuous action step 3, which states WR level in at least 2 SGs <14% go to the RNO for Feed and Bleed actions. Correct TS bases setpoint per SR 3.4.5.2 bases.

Sys # System Category KA Statement W/E05 Loss of EA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Secondary the (Loss of Secondary Heat Sink) procedures during abnormal and emergency Heat Sink I 4 operations.

KIA# EA2.1 KIA Importance 4.4 Exam Level SRO References provided to Candidate Technical

References:

FR-H.1 pg. 2 Rev .. 1 lss. 2 None TS Bases 3.4.5.2 pg b 3.4.5-5 Question Source: Bank - 1LOT14 NRC Exam (080)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

82. The plant is at 100% power.

• l&C is requesting Operations place control rods on DC Hold in accordance with 20M-1.4. 0, "Placing a Control Rod Power Cabinet Group on DC Hold", to perform maintenance in a rod control power cabinet.

• The crew has been briefed on compensatory measures in the event of a load rejection.

1) What is the MAXIMUM number of control rod group(s) capable of being placed on DC Hold?

2) If a load rejection were to occur, what is the Tech Spec OPERABILITY status of the rods that are on DC Hold?

A. 1) 1 Group

2) Operable B. 1) 2 Groups

2) Operable C. 1) 1 Group

2) Inoperable D. 1) 2 Groups

2) Inoperable Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the Operability of the Control Rods while they are on DC hold. The OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. Detailed knowledge of the bases is required to determine the impact of the loss of the power supplies and Operability of the Control Rods.

KIA is met by analyzing the effect of the rods being placed on DC hold for maintenance will have on the operability of the rods. Rods on DC hold are still operable (trippable) per the Tech Spec Bases.

A. Correct. The maximum number of rods is 4 (1 group at BV). Tech Spec bases defines a rod as operable if it is trippable. The DC hold cabinet is in parallel with the rod control power cabinets, both being powered through the reactor trip bkrs. When the Rx trip bkrs open, the rods will insert.

B. Incorrect. Plausible is the candidate thinks DC Hold can maintain a bank of rods (2 groups). It is correct that they are operable.

C. Incorrect. The maximum number of rods is 4 (1 group). Inoperable is not correct because when the DC Hold cabinet loses power (ie. Rx trip) the rods will insert.

D. Incorrect. Plausible is the candidate thinks DC Hold can maintain a bank of rods (2 groups). Inoperable is not correct because when the DC Hold cabinet loses power (ie. Rx trip) the rods will insert.

Sys # System Category KA Statement 000003 Dropped Generic Ability to analyze the effect of maintenance Control Rod I activities, such as degraded power sources, on 1 the status of limiting conditions for operations.

KIA# 2.2.36 KIA Importance 4.2 Exam Level SRO References provided to Candidate Technical

References:

20M-1.4.0 Rev. 0 lss. 1 pg. 1 None TS Bases pg. B 3.1.4-5 rev. 0 3SQS-1.3 Rev 7 lss. 1 pg. 13 Question Source: New Question Cognitive Level: Lower_ Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.2 I 45.13)

Objective: 3SQS-1.3 Obj 9 Explain the function, operation, location and limitations of the DC Hold Cabinet.

3SQS-1.3 Obj. 28 Using a copy of Technical Specifications or the Licensing Requirements Manual, assess a given set of plant conditions for compliance with the licensing requirements, including the determination of equipment operability and applicable action statements.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

83. The plant was at 100% power when the following occurs:

• The reactor failed to trip after receiving a valid trip signal

• SRO transitioned from E-0, Reactor Trip or Safety Injection, to FR-S.1, Response to Nuclear Power Generation/A1WS Current conditions:

• Emergency Boration was initiated

• Safety injection did not actuate

• Reactor power is 3% and decreasing

• Intermediate range channels indicate negative SUR

• Operators are verifying the reactor subcritical at step 7 of FR-S.1 Based on the current plant conditions:

(1) Boration _ _ _ _ _ _ _ required to continue after verifying the reactor is subcritical.

(2) Which of the following describes the required procedural flowpath?

A. 1) is

2) Return to E-0.

B. 1) is

2) Remain in FR-S.1 C. 1) is not

2) Return to E-0.

D. 1) is not

2) Remain in FR-S.1

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 83 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the existing power level and SUR. This evaluation requires detailed knowledge of the EOP procedure content and transition criteria. Additionally, the SRO must decide what action is required related to continuing the boration flow. Knowledge of the procedure steps is required to make the decision and select the correct transition.

KIA is met with the EOP background knowledge that emergency boration is required to continue to ensure adequate shutdown margin during future cooldown. The candidate must also determine if conditions are satisfied to transition back to E-0, or stay in FR-S.1.

A. Correct: In FR-S.1, after verifying the Rx is subcritical in step 7, step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions." Per the background this is to ensure adequate S/D margin during the future plant cooldown. When power<

5% and negative IR SUR is achieved in FR-S.1, step 7d directs returning to the procedure and step in effect which is E-0.

B. Incorrect: Boration is required to continue to obtain adequate shutdown margin during subsequent actions. It is not required to remain in FR-S.1 once it has been verified that the reactor is subcritical. Step 7d directs returning to the procedure and step in effect which is E-0.

C. Incorrect: step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions. Returning to E-0 is correct since the reactor is subcritical.

D. Incorrect: step 7c states "Continue boration as necessary to obtain adequate shutdown margin during subsequent actions. It is not required to remain in FR-S.1 once it has been verified that the reactor is subcritical. Step 7d directs returning to the procedure and step in effect which is E-0.

Sys# System Category KA Statement 000024 Emergency Generic Ability to perform specific system and integrated Boration / 1 plant procedures during all modes of plant operation.

KIA# 2.1.23 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.FR-S.1 lss. 2 Rev. O 20M-53B.4.FR-S.1 lss. 2 Rev. 0 Question Source: Bank- Surry 2010 NRC Exam (082) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 / 45.2 I 45.6)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

84. Initial conditions:

• Core Cooling CSFST is ORANGE

• FR-C.2, Response to Degraded Core Cooling is in progress Current conditions:

• Steam Generator depressurization to 100 psig is in progress

• A validated Orange Path on the CSFSTs points to FR-P.1, Response to Imminent Pressurized Thermal Shock Condition

1) What is the purpose of depressurizing all intact SGs to 100 psig in FR-C.2?

2) How must the Unit Supervisor respond to the Orange path on FR-P.1?

A. 1) To assist in core recovery by injecting the Safety Injection Accumulators.

2) Remain in FR-C.2 until completion, then transition to FR-P.1.

B. 1) To assist in core recovery by injecting the Safety Injection Accumulators.

2) Immediately transition to FR-P.1.

C. 1) To assist in core recovery by injecting using the Low Head Safety Injection Pumps.

2) Remain in FR-C.2 until completion, then transition to FR-P.1.

D. 1) To assist in core recovery by injecting using the Low Head Safety Injection Pumps.

2) Immediately transition to FR-P.1.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 84 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure transition based upon the rules of use and hierarchy for the Function Restoration Procedures. This evaluation requires detailed knowledge of the EOP procedure flow-paths. Additional knowledge of the procedure step bases is required, beyond the high level action steps for FR-C.2.

KIA is met with the knowledge of the bases for depressurizing the SGs in Orange path FR-C.2, Response to Degraded Core Cooling, and accessing the transition to an Orange path FR-P.1 cause by SI accumulators injecting.

A. Correct. Depressurization of the SGs to 100 psig is required to lower RCS pressure low enough to inject SI Accumulators and cover the core. It is correct to remain in FR-C.2 if a Orange path in FR-P.1 is created when the SI accumulators inject. It is an expected condition stated by a CAUTION prior to SG depressurization step.

B. Incorrect. It is correct that depressurization of the SGs to 100 psig is required to lower RCS pressure low enough to inject SI Accumulators and cover the core. It would be incorrect to immediately transition to FR-P .1 due to the note prior to the depressurization step. This is a plausible distractor if candidate has a misconception of the hierarchy for the Function Restoration Procedures.

C. Incorrect. Plausible because after the accumulators are isolated at 100 psig SG pressure, continued SG depressurization to atmospheric pressure allows the RCS pressure to be low enough for LHSI to inject into the core (step 17). It is correct to complete FR-C.2 prior to going to FR-P.1.

D. Incorrect. Plausible because after the accumulators are isolated at 100 psig SG pressure, continued SG depressurization to atmospheric pressure allows the RCS pressure to be low enough for LHSI to inject into the core (step 17). It would be incorrect to immediately transition to FR-P.1 due to the note prior to the depressurization step. This is a plausible distractor if candidate has a misconception of the hierarchy for the Function Restoration Procedures.

Sys# System Category KA Statement W/E06 Degraded EA2 Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Core Cooling the (Degraded Core Cooling) procedures during abnormal and emergency 14 operations.

KIA# EA2.1 KIA Importance 4.2 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.FR-C.2 lss. 2 Rev. 2 20M-53B.4.FR-C.2 lss. 2 Rev. 2 Question Source: Bank - Vogtle 2012 NRC exam (099) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

85. The plant is at 100% power.

• A reactor trip occurs coincident with a loss of all offsite power

• The operators have verified natural circulation flow and are cooling down the plant per ES-0.2, Natural Circulation Cooldown

• Train A RVLIS is OOS for Maintenance The following plant conditions now exist:

• RCS Pressure is 1940 psig and stable

• RCS Hot Leg temperatures are 540 °F and lowering

• RCS Cooldown rate based upon Cold Leg temperatures is currently 30 °F/Hr and CANNOT be reduced

• PSMS Data Processing Unit B indication is "NO OFFLINE" Which of the following procedures will be entered and what is the MAXIMUM allowable RCS Cooldown ratV i.ri ft..e pr-oc~c.<,Yf!- to b~ <!J.1t'(_ve.J. ~ ~J~

A. ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS); 50 °F/Hr B. ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel (With RVLIS); 100 °F/Hr C. ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS); 50 °F/Hr D. ES-0.4, Natural Circulation Cooldown with Steam Void in Vessel (Without RVLIS); 100 °F/Hr Answer: C Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. 11.E pg 7 which requires the knowledge of diagnostics steps and decision points in EOPs that involve transitions to event specific sub-procedures The SRO must be aware of sub-procedures for Natural Circulation Cooldown, if the CID rate cannot be maintained less than 25 °F/Hr. Detailed procedure knowledge is required for CID rate.

KIA is met by interpreting the conditions given in the question, then based on this knowledge, transition to the appropriate procedure due to Cooldown rate limitations and RVLIS availability.

A. Incorrect. Correct procedure. Cooldown rate is incorrect. ES-03 allows a cooldown rate of <100F/hr.

B. Incorrect Procedure With pressure <1950psig and Thot <550F, conditions are met to maintain 25F/hr cooldown rate. If <25F/hr cannot be maintained, the RNO step transitions the crew to ES-03 with RVLIS. In the stem of the question both trains of RVLIS are OOS, therefore RVLIS is not available.

C. Correct. Procedure is correct. With pressure <1950psig and Thot <550F, conditions are met to maintain 50 F/hr cooldown rate. In the stem of the question both trains of RVLIS are OOS, therefore RVLIS is not available. The cooldown rate in ES-04 is 50F/hr until temperature is less than 450F, then the rate is raised to 100F/hr.

D. Incorrect. Procedure is incorrect. Plausible distractor with one train of RVLIS OOS. The SRO must know that one train on RVLIS is still available, and a transition to ES-04 would not be correct. 1OOF/hr is the correct cooldown rate for ES-04 when Thot is between 500-450F.

Sys # System Category KA Statement W/E09 Natural EA2 Ability to determine and interpret the following as they apply Adherence to appropriate procedures and Circulation to the (Natural Circulation Operations) operation within the limitations in the facility*s Operations I 4 license and amendments.

KIA# EA2.2 KIA Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ES-0.2 lss. 2 Rev. 1 pg.16 20M-53A.1.ES-0.4 lss .. 2 Rev. 1 pg.3 20M-5.D.1.D lss. 4 Rev 0 pg. 15-16 Question Source: Bank - 1LOT14 NRC Exam (Q85) Modified Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 43.5 / 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

86. Given the following conditions:

• The plant is operating at 20% power

• AOP 2.6.8, 'Abnormal RCP Operation' has been entered due to rising temperatures on the 'B' RCP The following conditions exist:

RCS*P21B MTR LWR RCS*P21 B MTR UPR Time RADIAL [T0435Al THRUST [T0434Al 1000 181°F 184°F 1005 189°F 188°F 1010 197°F 194°F 1015 204°F 201°F

1) Which Motor Bearing reaches the RCP trip setpoint FIRST in accordance with AOP-2.6.8?

2) What actions will be directed by the Unit Supervisor?

A. 1) Motor Lower RADIAL Bearing

2) Shutdown 'B' RCP, go to AOP-2.51.1, Unplanned Power Reduction, and perform a controlled plant shutdown.

B. 1) Motor Lower RADIAL Bearing

2) Trip the reactor, go to E-0, complete the IOAs, then shutdown 'B' RCP.

C. 1) Motor Upper THRUST Bearing

2) Shutdown 'B' RCP, go to AOP-2.51.1, Unplanned Power Reduction, and perform a controlled plant shutdown.

D. 1) Motor Upper THRUST Bearing

2) Trip the reactor, go to E-0, complete the IOAs, then shutdown 'B' RCP.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 86 Answer: 8 Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures and transitions between Abnormal and EOPs. The SRO must evaluate the plant conditions and determine which setpoint has been exceeded for continued RCP operation. Then the SRO must determine the specific sequence of actions to take when securing the RCP, the sequence of actions are listed as sub-steps in the Abnormal Operating Procedure. Additionally directing the action to secure the pump is to occur following completion of the IOAs, which is SRO knowledge of the AOP procedure content. Per the EOP users guide, the Continuous Actions are on the fold out page, the Reader (US) is responsible for reviewing and monitoring the CA page and informing the crew when conditions are met to apply the action.

KIA is met by demonstrating the ability to predict the impact of a rising RCP bearing temperature, then based on reaching a required RCP immediate shutdown setpoint, chose the appropriate procedure to shutdown the Rx and the RCP. This is an abnormal RCP shutdown sequence in that the Rx is tripped, then the RCP is tripped. Normally RCP shutdowns occur prior to the Rx being critical during plant heat up, or after plant cooldown.

A. Incorrect. Correct bearing. Incorrect RCP shutdown sequence and procedure for shutting down the plant. Plausible distractor because tripping of an RCP when power is <30% (P-8) does not generate a Rx trip, and a controlled shutdown would be plausible, but not permitted.

B. Correct. IAW the AOP, motor bearing temperature setpoint for trip criteria is >195F which is met at 1010 by the MTR LWR RADIAL BEARING at 197F. AOP-2.6.8 Continuous action step 1 directs tripping the Rx, E-0, IOAs, then tripping RCP.

c. Incorrect. Incorrect bearing. Incorrect RCP shutdown sequence and procedure for shutting down the plant. Plausible distractor because tripping of an RCP when power is <30% (P-8) does not generate a Rx trip, and a controlled shutdown would be plausible, but not permitted.

D. Incorrect. Incorrect bearing. Correct Rx trip, IOAs, and RCP shutdown sequence.

Sys# System Category KA Statement 003 Reactor A2 Ability to (a) predict the impacts of the following malfunctions or operations Conditions which exist for an abnormal Coolant Pump on the RCPS; and (b) based on those predictions, use procedures to correct, shutdown of an RCP in comparison to a System (RCPS) control, or mitigate the consequences of those malfunctions or operations: normal shutdown of an RCP KIA# A2.02 KIA Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

20M-53C.4.2.6.8 Rev. 12 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 / 43.5/ 45.3 / 45/13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

87. Given the following initial conditions:

• Rx power is 100% and stable

• Control rods are in MANUAL for l&C maintenance

• Chemistry requested Cation bed demineralizer [2CHS-DEMIN22] be placed in service to lower RCS pH

• Cation bed demineralizer [2CHS-DEMIN22] was placed in service in accordance with 20M-7.4.C2, "Lowering RCS PH" One hour after the Cation bed demineralizer was placed in service, the Reactor Operator reports Reactor power is 100.1 % and slowly rising.

(Assuming the demineralizer was the cause)

Which of the following is the reason for the power rise, AND the appropriate procedure for the Unit Supervisor to implement?

A Cation Demineralizer was placed in service with a _ ___._(1~)___ boron concentration than the RCS.

The Unit Supervisor will implement (2)

A. 1) LOWER

2) AOP-2.51.2, "Reactor Overpower" B. 1) LOWER

2) 20M-52.4.B.1, "Turbine Load Changes" C. 1) HIGHER

2) AOP-2.51.2, "Reactor Overpower" D. 1) HIGHER

2) 20M-52.4.B.1, "Turbine Load Changes"

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 87 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures related to coordination with normal procedures. Specifically the SRO must evaluate the plant conditions and determine which the procedure to implement. The slow power rise warrants the use of the normal Turbine Load Change procedure versus the Abnormal Reactor Overpower procedure.

Detailed knowledge of the content is required to select the correct procedure. The first part of the question is a fundamental knowledge of the effect of a dilution event due to the cation demineralizer operation.

KIA is met by predicting the effect of placing a cation demineralizer in service with a lower boron concentration than the RCS (dilution event), and determine the correct procedure to use to mitigate the power change.

A. Incorrect. Correct that the demineralizer had a lower boron concentration than the RCS. Incorrect to use Reactor Overpower because power was not rapidly rising.

B. Correct. The given conditions indicate there is an RCS dilution in progress. If a Cation Demineralizer were placed in service with a boron concentration lower than the RCS, it would remove boron from the RCS resulting in a dilution event. With it being a slow rise in power the SRO would use the Turbine Load Changes procedure to control power. A note in the Reactor Overpower states this procedure is intended for use when power is rapidly rising. Conditions given had power rise 0.4% over an hour.

C. Incorrect. If the demineralizer was higher than the RCS, power would decrease, not rise as the conditions given. Incorrect to use Reactor Overpower because power was not rapidly rising.

D. Incorrect. If the demineralizer was higher than the RCS, power would decrease, not rise as the conditions given. Correct procedure to use for slowly rising power.

Sys# System Category KA Statement 004 Chemical and A2 Ability to (a) predict the impacts of the following malfunctions or Fact that isolating cation demineralizer stops Volume operations on the eves; and (b) based on those predictions, use boron dilution and enables restoration of normal Control procedures to correct, control, or mitigate the consequences of those boron concentration System malfunctions or operations:

KIA# A2.33 KIA Importance 3.3 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.51.2 Rev. 2 pg 1 None 20M-52.4.B.1 Rev. 1 pg. 3 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/ 43/5 / 45/3 I 4515)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

88. The plant has tripped from 100% power, due to an inadvertent Safety Injection.

When resetting both trains of the Safety Injection Signal in ES-1.1, SI Termination, step 1, Annunciators A 12-1 C, Auto Safety Injection Blocked and A 12-1 D Safety Injection Signal are intermittently flashing.

What action will the crew take in response to these conditions?

The operators will _ _ _ _ ___._1'-"------

If, after taking this action, SI initiation parameter setpoints are exceeded, the affected Safety Injection equipment (2) operate automatically.

A. (1) perform Attachment A-1.25, ESF Signal Reset by Alternate Method (2) will B. ( 1) close, then open the reactor trip breakers per ES-1.1, SI Termination (2) will C. (1) perform Attachment A-1.25, ESF Signal Reset by Alternate Method (2) will NOT D. (1) close, then open the reactor trip breakers per ES-1.1, SI Termination (2) will NOT

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 88 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the procedure caution related to the ESFAS signals and action required to manually operate the ESF equipment. The second caution in Attachment A-1.25 alerts the SRO that ESF equipment will have to be to be manually started. The knowledge of the plant response to a failure of a single train of SI to reset is detailed EOP knowledge of the Attachment. A-1.25. The operator must have the knowledge of how the SI blocked annunciator responds if a single train does not reset. After resetting the SI signal the alarms will flash in and out as one train of SI is not reset KIA is met by demonstrating the knowledge of the Caution in A-1.25 regarding manual actuation of ESF equipment.

A. Incorrect, The operator action is correct. The SI equipment will not automatically actuate after the SSPS train is disabled.

B. Incorrect, This action is plausible, Step 31 of procedure ES-1.1 enables an Automatic Safety Injection by cycling the reactor trip breakers. The equipment will automatically actuate if both trains of Safety Injection were capable of being reset.

C. Correct, The operator action is correct and equipment must be manually operated. The Caution prior to step 1 of procedure A-1.25 alerts the operator that a SI signal will not Automatically operate after performing the procedure steps, manual action to operate equipment will be required.

D. Incorrect, This action is plausible, Step 31 of procedure ES-1.1 enables an Automatic Safety Injection by cycling the reactor trip breakers. The equipment would automatically actuate if both trains of Safety Injection were capable of being reset, however the equipment will not operate automatically if the correct actions were taken.

Sys# System Category KA Statement 013 Engineered Safety Features Generic Knowledge of the operational implications of EOP Actuation System (ESFAS) warnings, cautions, and notes.

KIA# 2.4.20 KIA Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.1 lss. 2 Rev. 0 pg. 2 20M-53A.1.A-1.25 Rev 0 pg. 2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

89. The plant is operating at 50% power with all systems in normal alignment for this power level.

The following conditions exist:

• A1-4H, Service Water System Trouble is the only LIT overhead annunciator

• [2SWS-PI 113A and Pl 1138], Service Water Header Pressures are 48 psig and stable

• [2SWS*MOV107A, B, C, D], 'Sec Comp Clg Wtr Hx Serv Water Supply Hdr lsol Vlvs' are OPEN Based on the above conditions, what could cause this Service Water condition, and which of the following procedures listed below would be the correct procedure to mitigate this condition?

The above conditions could indicate a _ ___._(1. . ). .___ , and would be mitigated by (2)

A. 1) Service Water System leak

2) AOP-2.30.1, Service Water/Main Intake Structure Loss B. 1) Service Water System leak

2) E-0, Reactor Trip or Safety Injection C. 1) Service Water Pump trip

2) AOP-2.30.1, Service Water/Main Intake Structure Loss D. 1) Service Water Pump trip

2) E-0, Reactor Trip or Safety Injection

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 89 Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 third bullet. SRO is required to have knowledge of the content of the procedures. Specifically the SRO must evaluate the plant conditions and determine that the SWS pressure is low due to a leak and not due to the pump trip. Since pressure remains high, reactor trip is not warranted, the response is to continue with the AOP to respond to the leak. Detailed knowledge of the content is required to select the correct procedural direction.

KIA is met by the ability to predict a Service Water leak malfunction based on lower than normal service water header pressure and other conditions given, and use the Service Water AOP to mitigate the lower Service Water header pressure.

A. Correct. This is an indication of a service water leak due to pressure being lower and equal in both headers (headers cross tied in NSA and normal pressure -70 psig). Pressure is not, and was not low enough to start a stby SW pump (34 psig) which would cause Ann. A1-5F, 'Stby SW Pump Auto start/Auto stop' to alarm. Also, A 1-4F, 'SW Pump Auto start/Auto stop' is not LIT. The correct procedural guidance is in AOP-2.30.1 because pressure is not below 34 psig or CCS is not isolated (107s are open).

B. Incorrect. This is an indication of a service water leak. With SW pressure >34 psig or CCS is not isolated (107s are open) entry conditions to E-0 do not exist per the AOP.

C. Incorrect. No indication of pump trip exists in the stem. Pressure is lower, but annunciators which indicate a pump trip or a stby pump start do not exist. The AOP is the correct procedure.

D. Incorrect. No indication of pump trip exists in the stem. Pressure is lower, but annunciators which indicate a pump trip or a stby pump start do not exist. With SW pressure >34 psig or CCS is not isolated (107s are open) entry conditions to E-0 do not exist per the AOP.

Sys # System Category KA Statement 076 Service A2 Ability to (a) predict the impacts of the following malfunctions or Service water header pressure Water operations on the SWS; and (b) based on those predictions, use System procedures to correct, control, or mitigate the consequences of those (SWS) malfunctions or operations:

KIA# A2.02 KIA Importance 3.1 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.30.1 Rev. 9 None Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 1o CFR Part 55 Content: (CFR: 41.5 / 43.5 I 45131 45/13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

90. A fire has occurred in the Cable Tunnel. The Unit Supervisor directs an Operator to perform 20M-56C.4.D, "Nuclear Operator #1 Procedure".

20M-56C.4.D requires 2CHS*HCV186, 'RCP Seal Hdr Flow Control Valve' to be failed by isolating air within 10 minutes of entering 20M-56C.4.B, "Unit Supervisor Procedure".

The reason for this action is to ~~~~~~~~~~~~~~~~~

A. provide a reactor coolant system inventory control flow path B. avoid thermal shock to the reactor coolant pump seals C. prevent the valve from opening due to a fire induced short circuit D. maximize flowrate through the charging header Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the methodology of the alternate shutdown procedures intent and methodology. The bases for the actions taken in this procedure are specific to the SRO position. Detailed knowledge of the procedure content is required.

K/A is met with the knowledge that a Licensed Operator will be isolating Instrument Air to fail open 2CHS-HCV186 during the performance of Alternate Safe Shutdown From Outside Control Room, and identify the operational effects of this evolution.

A. Correct. Locally failing 2CHS*HCV186 open within 10 minutes is the correct action taken by the BOP Operator when performing 20M-56C.4.D.

Providing an inventory flow path is identified in the Intent and Methodology procedure 20M-56C.4.A. The valve is failed open to prevent spurious fire induced operation.

B. Incorrect. Plausible distractor if the candidate thinks HCV186 is failed closed, but HCV186 is failed open, and there is a caution in 20M-56C.4.B (US procedure) stating that thermal shock to the seal may occur and cause increased RCP seal leak rates.

C. Incorrect. Plausible distractor since air operated valves are place in their desired positions to prevent spurious fire induced operation, however the intent of isolating air to the valve is to cause it to fail open, not prevent it from opening.

D. Incorrect. Plausible distractor if the candidate thinks HCV186 is failed closed, as this would add to increased charging flow capabilities.

Sys# System Category KA Statement 078 Instrument Air Generic Knowledge of RO tasks performed outside the main control System (IAS) room during an emergency and the resultant operational effects.

K/A# 2.4.34 KIA Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

20M-56C4.A rev. 14 pg.2 & 4 20M-56C4.D rev. 24 pg.2 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

91. Given the following plant conditions:

• Unit 2 is currently in Mode 6

• Fuel movement is in progress

• Spent Fuel Pool (SFP) boron concentration (Cb) sample results have significantly dropped since last sample and are currently at the Technical Specification limit of 2000 PPM If SFP Cb continues to drop, what is the impact on shutdown margin?

A 5% (Keff < .95) shutdown margin will be _ _ _ _ _ __

A. no longer maintained regardless of SFP Cb B. maintained as long as SFP Cb > 495 PPM C. maintained as long as SFP Cb > 350 PPM D. maintained regardless of SFP Cb

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 91 Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine the impact on SDM per the TS bases and TS action required. Detailed knowledge of the bases is required to determine the impact of the boron dilution.

KIA is met by predicting the impact of SFP Boron concentration lowering on the SFP shutdown margin.

A. Incorrect. Keff cannot be maintained < .95 for a credible dilution event. (refer to correct answer explanation)

B. Correct. According to TS 3.7.16 and its associated bases, the >2000 PPM limit conservatively assures Keff is maintained within the limit (Keff

<.95) for the worst case misplaced fuel assembly accident. In addition, this limit ensures no credible boron dilution event will reduce Cb < 495 ppm required during non-accident conditions to maintain Keff < .95.

c. Incorrect. Keff will not be maintained at this low of a boron concentration. This is a common value in Tech Specs so it is plausible that this choice may be selected.

D. Incorrect. It is incorrect that Keff will be maintained at any boron concentration in the SFP.

Sys# System Category KA Statement 033 Spent Fuel A2 Ability to (a) predict the impacts of the following malfunctions or Inadequate SOM Pool Cooling operations on the Spent Fuel Pool Cooling System ; and (b) based System on those predictions, use procedures to correct, control, or mitigate (SFPCS) the consequences of those malfunctions or operations:

KIA# A2.01 KIA Importance 3.5 Exam Level SRO References provided to Candidate Technical

References:

TS 3.7.16 Bases, Rev. 5 None Question Source: Bank - 2LOT7 NRC Exam (Q93) Modified (2011)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 I 45.3 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

92. The plant was operating at 100% power when a SGTR occurred on 'B' Steam Generator.

• The reactor was manually tripped

• Condenser Vacuum is 18" Hg Vac and stable

• The crew has progressed up through step 6, 'Initiating RCS Cooldown' of E-3, "Steam Generator Tube Rupture"

• All previous EOP steps, including local Operator actions, have been completed How many separate valve controllers are available to MANUALLY cooldown the RCS in accordance with E-3 step 6?

A. 1 B. 2 C. 3 D. 4 Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .E page 21 second bullet. SRO is required to have knowledge of the content of the EOP procedures. The loss of condenser vacuum will disable the condenser steam dumps as a possible flowpath. The procedure steps in E-3 will isolate the ruptured SG steam supply to the common RHR valve so it will be available at this time. Also the procedure cautions against use of the ruptured SG atmospheric, so 3 steam flowpaths are available. Detailed knowledge of the procedure actions to the step to cooldown is required to select the correct answer.

KIA is met by demonstrating the ability to cooldown the plant after a SGTR in conjunction with a low vacuum condition.

The low vacuum condition must be identified by the candidate as a valid Condenser NOT Available (C-9) alarm causing condenser steam dumps from being available.

A. Incorrect. Plausible distractor if candidate doesn't recognize that C-9 is in, and thinks the condenser steam dumps are available. This is the preferred steam relief flowpath.

B. Incorrect. Plausible distractor if candidate thinks only the 2 atmospheric valves from the intact SGs are available.

C. Correct. 2 atmospheric dumps (A & C) from the intact SGs, and the RHR valve can be used since it was isolated from the rupture SG in step 4b.

The condenser steam dumps are not available due to condenser vacuum at 18" is above the C-9 setpoint of 19.5" and annunciator A 12-4C

'Condenser Unavailable (C-9)' would be lit. C-9 blocks condenser steam dump operation.

D. Incorrect. Plausible distractor if candidate thinks 2 atmospheric dump valves, the RHR valve, and the condenser steam dumps are available or the ruptured SG could be used. Step 6 states go to ECA-3.1 is the ruptured SG must be used.

Sys# System Category KA Statement 041 Steam Dump System Generic Ability to verify system alarm setpoints and (SDS)/Turbine Bypass Control operate controls identified in the alarm response manual.

KIA# 2.4.50 KIA Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

20M-53A.1.E-3 lss. 2 Rev. 4, pgs. 4 & 10 20M-1.5.B.3 rev. 1 pg. 2 20M-26.4.ABM Rev. 2 pg. 3 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 45.3)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

93. The plant is operating at 100% power.

• The crew is implementing AOP-2.34.1 "LOSS OF STATION/CNMT INSTRUMENT AIR"

• [2SAS-C21A] STATION AIR COMPRESSOR TRIPPED

• [2SAS-C22] CONDENSATE POLISHING AIR COMPRESSOR is RUNNING

• [21AS-C21] DIESEL DRIVEN AIR COMPRESSOR failed to start

• [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV is OPEN

[21AS-Pl106] STA INSTR AIR HEADER PRESSURE is currently 72 psig and lowering

1) What is the next action required to be taken in accordance with AOP-2.34.1?

2) At what pressure will a Manual Reactor Trip, and transition to E-0, "Reactor Trip or Safety Injection" be required?

A. 1) CLOSE [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV

2) 55 psig

8. 1) ISOLATE [21AS-DRY23A & 238] INSTR AIR DRYERS

2) 55 psig C. 1) CLOSE [2SAS-AOV105] SAS MAIN HEADER TO SERVICE AIR HEADER AOV

2) 65 psig D. 1) ISOLATE [21AS-DRY23A & 238] INSTR AIR DRYERS

2) 65 psig

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 93 Answer: C Explanation/Justification: Meets NUREG-1021 Rev. 10, Att.2 Sect. II E. SRO is required to have knowledge of the content of the procedure versus knowledge of the overall mitigative strategy or purpose, Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures. The knowledge of the procedural sequence is required based on the continuous action step 5 which checks SIA header pressure >86 psig, since it is not, the RNO of the step must be completed. This directs the closing of 2SAS-AOV105. The knowledge of tripping the Rx at 65 psig is contained within the AOP step RNO and is detailed procedural knowledge for the SRO. It is stated in a NOTE in the AOP attachments that at 65 psig, the MFW Regulating Valves will fail closed.

KIA is met the ability to predict the effects of an air leak in the station air system with a failure of redundant air compressors, and recognize that the station to Instrument Air header cross connection valve has failed to close. Then determine that based on an air pressure <65 psig, that a Rx trip and entry into E-0 is required .

.*A. Incorrect. It is correct to close AOV105, but incorrect Inst Air pressure for manually tripping the Rx and transitioning to E-0. 55 psig is a plausible distractor because this was the old AOP value.

B. Incorrect. Isolation of the Station Air system from the Inst Air header is performed prior to bypassing around and isolating the Inst Air Dryers procedurally. 55 psig is a plausible distractor because this was the old AOP value.

C. Correct. Closing AOV105 is required per the AOP because it should have automatically closed at 86 psig and isolated Station Air from Inst Air.

65 psig is the correct value requiring a manual Rx trip and transition to E-0, per continuous action step 5 when 21AS-Pl106 is S65 psig. This basis is stated in a NOTE in the AOP attachments stating that 65 psig the MFW Regulating Valves will fail closed.

D. Incorrect. Isolation of the Station Air system from the Inst Air header is performed prior to bypassing around and isolating the Inst Air Dryers procedurally. Correct air pressure for manually tripping Rx and transitioning to E-0.

Sys # System Category KA Statement 079 Station Air A2 Ability to (a) predict the impacts of the following malfunctions or operations on Cross-connection with IAS System the SAS; and (b) based on those predictions, use procedures to correct, control, or (SAS) mitigate the consequences of those malfunctions or operations:

KIA# A2.01 KIA Importance 3.2 Exam Level SRO References provided to Candidate Technical

References:

20M-53C.4.2.34.1 Rev. 19 None Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

94. The plant is operating at 100% power when an Operator filling a Tech Spec required watch position becomes ill and leaves the site.

Per TS 5.2.2 and 10CFR50.54, the crew composition may remain less than the minimum for a period of time not to exceed (1)

If the vacant position is NOT refilled within the required time, the crew will (2)

A. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) take action to place the unit in MODE 5 B. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) maintain current power level C. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) take action to place the unit in MODE 5 D. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) maintain current power level Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .A page 17 third bullet. SRO is required to have knowledge of the TS section 5 and 6 actions related to plant staffing. Additionally the SRO is required to know the administrative procedure content related to not meeting staffing requirements.

K/A is met with the knowledge of how long an on-duty shift position may be unfilled, and the requirements of Tech Specs and Plant Procedures to maintain safe plant operation.

A. Incorrect. See correct answer.

B. Incorrect. See correct answer.

C. Incorrect. See correct answer.

D. Correct. In accordance with TS 5.2.2, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the expected time to fill the required position provided immediate action is taken to restore shift composition to minimum. BVPS has incorporated Licensing position on TS 5.2.2 into NOP-OP-1002, which states it is not conservative to place the plant into a transient due to staffing, therefore maintain the unit in a steady state condition and continue calling out personnel.

Sys# System Category KA Statement N/A NIA Generic Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc.

K/A# 2.1.4 KIA Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

T.S. 5.2.2 Amend 278/161 pg. 5.2-1 NOP-OP-1002 Rev. 10 sect. 4.1.13 Question Source: New Question Cognitive Level: Lower- Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.2)

Objective: 3SQS-48.1 Obj. 3 From memory, describe the required actions if less than the minimum shift staffing complement exists.

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

95. The plant is operating in Mode 6 with all systems in normal alignment for this Mode.

• Core Off-Loading activities are in progress and core off-load is half complete

• Source Range Channel N31 fails LOW

• Source Range Channel N32 remains OPERABLE

• Refuel cavity water clarity is murky

• Gamma Metrics detectors N52A and N52B are Out of Service Which of the following activities can be performed WITHOUT violating the Technical Specification required actions for Source Range Instrumentation?

A. Latch and move a spent fuel assembly from the upender to the Spent Fuel Pool.

B. Latch and move a spent fuel assembly from the core to the Spent Fuel Pool.

C. Install a temporary secondary source into a core location.

D. Add Hydrogen Peroxide mixed with primary grade water to the refueling cavity for cleanup.

Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 first and third bullet.

SRO must have knowledge of the TS bases to answer this question. Specifically, SRO must know and apply the TS definition of core alteration and be familiar with the TS bases discussion on what is allowed and not allowed, with respect to compliance with the action statements. Additionally, the SRO must be knowledgeable of the "safe" locations defined in TSs and will be responsible for directing the operator actions to comply with the TS actions.

K/A is met with the knowledge of permissible actions iaw Tech Specs during core alterations when a Source Range detector is inoperable.

A. Correct. The SRO must understand that a loss of N31 puts them into AOP-2.2.1A 'SR Channel Malfunction' and TS 3.9.2. Both the AOP & TS direct that core alterations are immediately suspended. Core alterations are defined as movement of any fuel, sources, or reactivity components, within the reactor vessel with the vessel head removed and with fuel in the vessel. The SRO must have knowledge of the administrative requirements associated with refueling activities and have knowledge of TS bases. In order to answer this question the SRO must know the definition of Core Alterations and be able to apply this definition to a set of plant conditions. The movement of a Spent Fuel Assembly from the upender to the SFP is allowable because it is not within the reactor vessel.

B. Incorrect. Latching and moving a fuel assembly from the core would not be allowable by definition. Removing the assembly would not be considered placing the assembly in a safe location. This is plausible because some TS such as TS 3.9.4 LCO preclude core onload but do allow core offload to continue.

C. Incorrect. Plausible that operationally an alternative source could be installed, however, it is not allowed by the definition for what constitutes a Core Alteration.

D. Incorrect. Plausible that hydrogen peroxide is added to the water for clarity and cleanliness. However, the addition of primary grade water into the RCS would violate the second part of TS 3.9.2 since primary grade water could reduce boron concentration and is not allowed.

Sys# System Category KA Statement N/A Generic Conduct Of Operations Knowledge of procedures and limitations involved in core alterations.

KIA# 2.1.36 KIA Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

20M-53C.4.2.2.1A, Rev. 9, Pg. 8; TS Definitions Pg. 1.1-2; TS B3.9.2 Pg. B3.9.2-2 Question Source: Bank - 1LOTS NRC Exam (095)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 / 43.6 I 45.7)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

96. Given the following:

• Following a turbine runback, the crew is stabilizing the plant in accordance with the appropriate procedure.

• Control Bank "D" Group Counters are at 180 steps.

• On DRPI, one Control Bank "D" rod indicates 196 steps; all others indicate 182 steps.

• The affected rod has a blown movable gripper fuse and has been determined to be trippable.

Which of the following describes the technical specification implications of this event?

A. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.

B. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

C. The rod is INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.

D. The rod is INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

Answer: A Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .B page 17 third bullet. SRO is required to have knowledge of the TS bases. Specifically the SRO must evaluate the plant status and determine the impact on Rod alignment and TS action required. Detailed knowledge of the bases is required to determine which TS actions are applicable. This item meets the 10CFR55.43 (b) 2 SRO criteria because it requires the applicant to apply technical specification action with knowledge of the bases for that action.

K/A is met by demonstrating the ability to recognize that a rod exceeds alignment limits, but is still operable due to it being trippable. Knowledge of the TS bases is required to know why this is an undesired condition.

A. Correct. Since the rod is trippable it is operable. Restore rod to within alignment limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition B.

Misalignment limits are based on impact on power distribution limits iaw with TS 3.1.4 bases.

B. Incorrect. Correct, the rod is operable, but the concern for the situation presented is not shutdown margin.

C. Incorrect. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition A, but rod is not inoperable if it is trippable iaw TS 3.1.4 bases. If the rod were untrippable, then SOM would be affected. Power distribution limits are the correct reason for misaligned rods.

D. Incorrect. Would be true if the rod were untrippable.

Sys # System Category KA Statement N/A N/A Generic Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

KIA# 2.2.42 KIA Importance 4.6 Exam Level SRO References provided to Candidate None Technical

References:

TS 3.1.14, condition B, and basis Question Source: Bank-1LOT7 NRC Exam (091)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 /41.10/43.2/

43.3 I 45.3)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

97. In accordance with 1/2-EPP-IP-5.3, "Emergency Exposure Criteria And Control", whose authorization is required to exceed the emergency exposure limits of 10 CFR 20 "Standards For Protection Against Radiation" to save a life during an emergency, and what TEDE limit is this authorization limited to?

1) Who is authorized to grant exceeding the 10 CFR 20 limits during a declared emergency?

2) What is the maximum TEDE limit this individual may authorize?

A. 1) Emergency Director

2) 10 Rem TEDE B. 1) Emergency Director

2) 75 Rem TEDE C. 1) Emergency Recovery Manager

2) 10 Rem TEDE D. 1) Emergency Recovery Manager

2) 75 Rem TEDE Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and position responsibilities for the Emergency Director. This is a SRO position function only.

KIA met with knowledge of 10CFR20 emergency limits, and who may authorize emergency exposure limits, and what the limit is for saving a human life.

A. Incorrect. The ED is the only individual authorized to grant exceeding 10CFR20 emergency limits up to 75 Rem at which the Senior Vice President must give concurrence. 10 Rem is a plausible distractor because it is the 10CFR20 emergency exposure limit for preventing the failure of equipment necessary to protect the public health and safety.

B. Correct. The ED is the only individual authorized to grant exceeding 10CFR20 emergency limits up to 75 Rem at which time the Senior Vice President must give concurrence.

C. Incorrect. Plausible because the Emergency Recovery Manager has many responsibilities, and will discuss information with the ED, but ONLY the ED is authorized to grant exceeding 10CFR20 emergency limits. 10 Rem is a plausible distractor because it is the 10CFR20 emergency exposure limit for preventing the failure of equipment necessary to protect the public health and safety.

D. Incorrect. Plausible because the Emergency Recovery Manager has many responsibilities, and will discuss information with the ED, but ONLY the ED is authorized to grant exceeding 10CFR20 emergency limits. The expose limit of 75 Rem is the maximum allowed to be granted by the ED when the above stated conditions exist. Above 75 Rem requires the Senior Vice Presidents concurrence.

Sys# System Category KA Statement N/A NIA Generic Knowledge of radiation exposure limits under normal or emergency conditions.

KIA# 2.3.4 KIA Importance 3.7 Exam Level SRO References provided to Candidate none Technical

References:

1/2-EPP-IP-5.3, Rev. 11 pg. 3, 4, 9 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.12 I 43.4 I 45. 10)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

98. The Plant is operating at 100% power.

• Unit 2 is discharging the contents of the Gaseous Waste Storage tanks IAW 1/20M-19.4A.B, 'Unit 2 GW Storage Tk Disch To Unit 1 Atmos Vent'

• Rad Monitor RM-1GW-1088, Gaseous Waste Gas fails downscale and is declared inoperable

• The crew terminates the discharge In order to re-start the discharge, what 1/2-0DC-3.03, 'ODCM: Controls for RETS and REMP Programs' actions will be REQUIRED?

(Refer to attached reference)

A. The system/process flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (or assumed to be at the ODCM design value).

B. At least two independent samples of the tank's content are analyzed and at least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.

C. Grab samples (or local monitor readings) are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If grab samples are taken, these samples are to be analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Samples are continuously collected with auxiliary sampling equipment as required in ODCM Control 3.11.2.1, Table 4.11-2, or sampled and analyzed once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer: B Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II D page 20 first bullet. SRO is required to have knowledge of the Offsite Dose Calculation Manual and actions required for failed monitoring equipment.

This is a SRO position function only.

This question requires the candidate to use their knowledge of the provided ODCM (84 pgs.), and determine which attachment is applicable for determining the status of RM-1GW-108B. Then determine whether a continuous or batch discharge is in progress. Based on that decision, follow up with any actions associated with the release.

KIA is met by identifying what controls must be used for radioactive releases if a Gaseous Waste gas detector is inoperable during a Gaseous Waste discharge.

A. Incorrect. Plausible distractor because this is a required action if FR-GW-108 is OOS (Action 28A) not RM-GW-108B.

B. Correct. IAW ODCM 1/2-0DC-3.03 Att.F page 38 and action 27 on page 42.

C. Incorrect. Plausible distractor because this is the required action for all continuous releases thru this pathway. (Action 29)

D. Incorrect. Plausible distractor because this is the required action for continuous releases if the alt channel 109 is also not available (Action 32).

For Batch release alt. RM-1GW-109 shall not be used as a comparable alternate monitoring channel iaw Att. F page 2.

Sys # System Category KA Statement NIA N/A Generic Ability to control radiation releases.

KIA# 2.3.11 KIA Importance 4.3 Exam Level SRO References provided to Candidate Y:z-ODC- 3.03 Technical

References:

ODCM Y:z-ODC-3.03 Rev. 13 Att.F pages 38 & 42 Question Source: Bank- 2LOT8 NRC Exam (Q98) (2012)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.11/43.4 / 45.10)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

99. Given the following plant conditions:

• Large Break LOCA occurred

• A release is in progress via a small breach in the Containment Equipment Hatch

• An ALERT was declared at 1500

• A Site Area Emergency was declared at 1530 In accordance with 1/2-EPP-IP-3.2, SITE Assembly and Personnel Accountability;

1) What is the latest time that a Site Accountability report must be completed?

2) What is the location to which Non-Emergency personnel will be directed for evacuation?

A. 1) 1530

2) Emergency Operations Facility

8. 1) 1600

2) Emergency Operations Facility C. 1) 1530

2) Hookstown Grange D. 1) 1600

2) Hookstown Grange Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and position responsibilities for the Emergency Director. This is a SRO position function only.

KIA is met with the knowledge of site accountability requirements and evacuation locations of site personnel.

A. Incorrect. 1530 is plausible because accountability must be completed within 30 minutes of a SAE. If the applicant assumes in error that accountability is required for Alert then 1500 + 30 min = 1530. The EOF is plausible because it is an offsite location where emergency response personnel report.

B. Incorrect. 1600 is the correct time, but non-emergency personnel will be directed to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C. The EOF is plausible because it is an offsite location where emergency response personnel report.

C. Incorrect. 1530 is plausible because accountability must be completed within 30 minutes of a SAE. If the applicant assumes in error that

=

accountability is required for Alert then 1500 + 30 min 1530. The ED may direct non-emergency personnel to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C.

D. Correct. IAW Y.-EPP-IP-3.2 sect. 8.2.6, site accountably must be completed within 30 minutes of a Site Area Emergency. The ED may direct non-emergency personnel to the Hookstown Grange, or the Beaver County Community College Golden Dome Offsite Assembly Areas iaw attachment C.

Sys# System Category KA Statement NIA N/A Generic Knowledge of the emergency plan.

KIA# 2.4.29 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

1/2-EPP-IP-3.2 rev. 19 pg. 7 & 13 Question Source: Bank - VC Summer 2013 SRO NRC Exam (Q25)

Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.11)

Objective:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15) 100. Given the following plant conditions:

• The Emergency Director declared a Site Area Emergency at 1235

• The initial report to state and local government was completed at 1250

• An upgrade to General Emergency was declared at 1245

• The Initial Protective Action Recommendation (PAR) was made without a dose projection

• At 1255 a Valid dose projection is available which requires an upgraded PAR The Initial AND Upgraded (PAR) to the State/County Agencies must be given by which of the following times?

INITIAL UPGRADED A. 1245 1300 B. 1245 1310 C. 1300 1305 D. 1300 1310 Answer: D Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II E page 21 third bullet. SRO is required to have knowledge of the Emergency Plan and notification requirements. This is a SRO position function only.

K/A is met with the knowledge of the time requirements of the initial PAR, and the upgraded PAR after a dose projection is available.

A. Incorrect. Plausible if it is determined that the Initial PAR must be declared at the time of the GE declaration (1245), and that the Upgraded PAR is declared 15 minutes after that (1300)

B. Incorrect. Plausible if it is determined that the initial PAR must be declared within 15 minutes of declaration of a GE (1245). Correct time for the Upgraded PAR determination, it must be completed within 15 minutes of assessment being available (ie: dose projection) at 1310 C. Incorrect. Plausible it is correct that the initial PAR must be declared within 15 minutes of declaration of a GE (1300). Incorrect time for the Upgraded PAR determination, if it assumed that the Upgraded PAR is required within 15 minutes of the notification time to the states and counties of the Site Area Emergency (1305)

D. Correct. The Initial PAR must be declared within 15 minutes of declaration of a GE. The upgraded PAR does not change emergency classification status. Upgraded PAR determination must be completed within 15 minutes of assessment being available (ie: dose projection) This is SRO level knowledge only.

Sys# System Category KA Statement N/A NIA Emergency Procedures/Plan Knowledge of emergency plan protective action recommendations.

KIA# 2.4.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

1/2-EPP-IP-4.1, Rev. 31, pg. 10 & 13 Question Source: Bank - 1LOTS NRC Exam (#100) Modified Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/ 41.12 /43.5/

45.11)

Objective:

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