Text

Draft Written Exam (Folder 2)
	ML15355A243
Person / Time
Site:	Beaver Valley
Issue date:	11/30/2015
From:	Peter Presby; Operations Branch I
To:	; FirstEnergy Nuclear Operating Co
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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 Applicant Information Name:

Date: Facility/Unit: Beaver Valley Unit 2 Region: I l:&l 11 D 111 D IV D Reactor Type: W [8] CE 0 BW D 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 ZS Points Applicant's Scores Points Applicant's Grade Percent NUREG-1021, Revision 10 FENOC Facsimile r1

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. AS), will alarm in the FIRST 45 seconds AFTER N-43 fails High?

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

A. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into 0Tl1 T trip setpoint calculation. Therefore this alarm will NOT be energized.

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. 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.

D. Incorrect. (1) is plausible if it is not known that N-44 does NOT input into 0Tl1T 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 (Q1)

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 operating at 100% power when the reactor tripped on low PRZR pressure.

[2RCS-Pl472], Pressurizer Relief Tank pressure indicates 35 psig. The crew suspects that a PORV opened inadvertently and is now stuck partially open.

Which of the following confirming indications could be expected if a PORV is stuck partially open?

A. PORV relief line temperature stabilized at 281°F, and the PRZR safety relief line temperatures are slowly rising.

B. PORV relief line temperature stabilized at 259°F, and the PRZR safety relief line temperatures are slowly rising.

C. PORV relief line temperature stabilized at 281°F, and the PRZR safety relief line temperatures indicate ambient temperature and stable.

D. PORV relief line temperature stabilized at 259°F, and the PRZR safety relief line temperatures indicate ambient temperature and stable.

Answer: A 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 system flowpath knowledge determine if the PRZR safety relief line temperatures will sense this temperature change.

A. Correct. 281°F is the saturation temperature corresponding to 50 psia. The safety relief lines would be rising because they share a common discharge line to the PRT with the PORV's.

B. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure = 50 psia). Safety relief line temperatures would be rising because they share a common discharge line to the PRT with the PORVs.

C. Incorrect. 281°F is the saturation temperature corresponding to 50 psia. Plausible to have the safety relief line temperatures ambient temp if it was not known that they share a common discharge line to the PRT with the PORV's.

D. Incorrect. 259°F is approximately the saturation temperature corresponding to 35 psia (35 psig PRT pressure= 50 psia). Plausible to have the safety relief line temperatures ambient temp if it was not known that they share a common discharge line to the PRT with the PORV's.

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 U2 RM-0406-001 Rev. 27 Question Source: Bank- Vision #135831 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 218 Seal Lk Off, 2CHS-FT155B Low, computer address point F0128D)

RCP 21 B Seal Lk Off, 2CHS-FT155B 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 Technical

References:

20M-6.4.AAE Rev. 12 None 20M-7.4.AAH Rev. 22 Question Source: Bank - 2LOT6 NRC Exam (Q4)

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.

A Loss of Charging occurs due to 2CHS*MOV289, 'Normal Charging Header lsol Viv' failing CLOSED

Immediate Operator Actions of AOP-2.7.1, "Loss of Charging or Letdown", have been completed

NO other actions have been completed

Total Seal Injection flow is 20 gpm

Total Seal Water Leakoff is 9 gpm

1) How long will it take to receive A4-1C, Pressurizer Control Level Deviation High/Low alarm?

2) Will the Pressurizer Level Deviation alarm be HIGH or LOW?

A. 1) 22 - 28 minutes

2) High B. 1) 41 - 49 minutes

2) High C. 1) 4 - 6 minutes

2) Low D. 1) 10 - 15 minutes

2) Low Answer: B Explanation/Justification: KJA 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 AK1. 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 6 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 I 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 identifies 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 I 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

6. The plant was operating at 100% power when a main steam line break occurred inside containment.

The following conditions exist:

The Reactor automatically tripped

Containment pressure peaked at 15 psig

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) monitored for proper operation

2) Safety Injection could cause damage to the pump seals due to maximum seal injection flow B. 1) monitored for proper operation

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

2) Safety Injection could cause damage to the pump seals due to maximum seal injection flow D. 1) tripped

2) thermal barrier and motor cooling flow was secured Answer: D 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. With a CIB occurring, monitoring the RCPs would be inappropriate. The correct action is to trip the RCPs. It is not a correct statement regarding Safety Injection initiation since Seal Injection flow is limited by Tech Specs for the explicit purpose of not affecting Safety Injection flow into the RCS when actuated. Plausible distractor if they do not remember that seal injection flow is limited to a maximum of 28 gpm by TSs.

B. Incorrect. With a CIB occurring, monitoring the RCPs would be inappropriate. The correct action is to trip the RCPs. It is correct that a loss of CCP flow to the RCPs has occurred due to the CIB.

C. Incorrect. Tripping the RCPs is correct per E-0 LHP. It is not a correct statement regarding Safety Injection initiation since Seal Injection flow is limited by Tech Specs for the explicit purpose of not affecting Safety Injection flow into the RCS when actuated. Plausible distractor if they do not remember that seal injection flow is limited to a maximum of 28 gpm by Tech Specs.

D. Correct. With Containment pressure peaking at 15 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.

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. 1C Rev.1 pg.45 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 145.61 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 around the PORV setpoint.

D. The plant will remain at power and RCS pressure will cycle around the PRZR Code Safety setpoint.

Answer: C Explanation/Justification: KIA 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 the 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 the LP PORV block at 2185 psig. This will occur only on 456 & 4550 since 455C is manually isolated.

0. Incorrect. Plausible distractor if it is not recognized that the code safety setpoint is 2475 psig, and PORVs 456 & 4550 are available. The PORV 455C was isolated in the question stem due to it being controlled from PK444A. Making the code safety a plausible distractor.

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 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/AlWS".

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 an ATWS:

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 / 45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

9. Given the following conditions:

Rx Tripped due to SGTR on 21A SG

Safety Injection is actuated

E-3, "Steam Generator Tube Rupture" is in progress

The crew is currently isolating flow from the ruptured SG

1) What causes 2SSR*AOV117 A, '21 A SG Slowdown Sample Outside Cnmt Isol Viv' to close on a SG tube rupture?

2) What is this purpose for isolating the normal sample flowpath from the ruptured Steam Generator in E-3?

A. 1) Steam Generator Slowdown Sample Radiation Monitor 2SSR-RQ100 High Alarm

2) To conserve Steam Generator inventory.

S. 1) CIA Signal

2) To minimize radiological releases from the ruptured SG.

C. 1) Steam Generator Slowdown Sample Radiation Monitor 2SSR-RQ100 High Alarm

2) To minimize radiological releases from the ruptured SG.

D. 1) CIA Signal

2) To conserve Steam Generator inventory.

Answer: C 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 closed by 2SSR-RQl100 High Alarm. The reason to isolate the normal flowpath is to minimize radiological releases. Plausible distractor thinking that conserving inventory is the priority in E-3.

B. Incorrect. CIA is not correct. Plausible distractor because 2SSR* AOV117A is an Outside CNMT lsol valve, but it is not a CIA valve that closes when CIA actuates (SI actuated in stem). Minimize radiological releases from the ruptured SG is correct.

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

0. Incorrect. CIA is not correct. Plausible distractor because 2SSR* AOV117A is an Outside CNMT lsol valve, but it is not a CIA valve that closes when CIA actuates (SI actuated in stem). Plausible distractor thinking that conserving inventory is the priority in E-3.

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. 1C Rev. 19 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 I 41.10 I 45.6 I 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%.

8. 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 13 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 Alt. 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 Rev. 1 lss. 2 pg. 4 & 9 20M-53B.4.ECA-O.O Rev. 1 lss. 2 pg 84 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 350°F and stable

RCS Subcooling is 200°F and stable

RCS Pressure 1200 psig and stable

RCS cooldown rate is 20°F/hr 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 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months 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)

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 46% 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) PRZR LOW PRESS RX TRIP CHAN Ill

4) SOURCE RNG N32 RX TRIP

5) RCS LOOP A OT~ T RX TRIP

6) RCS LOOP B OP~T RX TRIP A. 1, 3, 5 B. 2, 4, 6 C. 1, 4, 5 D. 2, 3, 6 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 (also CH1 powered from VB1), PRZR LOW PRESS RX TRIP CHAN Ill is powered from VB3 and not effected, RCS LOOP A 0Tt1T RX TRIP is CH1 powered from VB1 and not effected.

B. Correct. PWR RNG N42 HIGH POWER SP, SOURCE RNG N32 RX TRIP, and RCS LOOP B 0Pt1T RX TRIP are all normally off when not above a setpoint condition. They are all 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 (also CH1 powered from VB1),, SOURCE RNG N32 RX TRIP will be lit due to the loss of VB3, and LOOP A OHT RX TRIP is CH1 powered from VB1.

D. Incorrect. PRZR LOW PRESS RX TRIP CHAN Ill is powered from VB3 and not effected, PRZR LOW PRESS RX TRIP CHAN Ill is powered from VB3 and not effected, RCS LOOP B OP!1T RX TRIP will be lit due to the loss ofVB2.

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

References:

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

218 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: KIA is met by demonstrating knowledge of the increased service water flow through the CCP Hxs to the Gire 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.

8. 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 None Technical

References:

20M-30.1.D Rev. 8, pg. 6, 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: (CFR41.4,41.8/45.7)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

15. The plant is at 45% power with all systems in normal alignment for this power level EXCEPT Station Air Compressor 2SAS-C21A is on Clearance for motor replacement.

The following conditions exist:

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 2J (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 2J bus occurs. 2J supplies power to the 2SAS-C21 B air 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 Instrument system operating parameters that are entry-level Air/ 8 conditions for emergency and 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.18 pg. 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 / 45.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

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 and STABLE Based on the above conditions, what are the required actions to restore the RCS Tavg-Tref deviation?

A. Manually Withdrawal Control Rods 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 +I- 2F.

A. Incorrect. Rod withdrawal would only make a larger Tavg-Tref deviation. Plausible distractor if there is a misunderstanding of which temperature the rods affect.

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# AA1.04 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.35.2 Rev. 20 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 and 41.10 I 45.5, 45.7, and 45.8)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

17. The plant has tripped 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago from 100% power due to a RCS leak.

The EOP network has been entered and plant conditions required a transition to ECA-1.1, "Loss Of Emergency Coolant Recirculation"

Step 18, Establish Minimum SI Flow to Remove Decay Heat is being performed Refer to attached ECA-1.1step18, and Attachment A-4.7, Minimum SI Flow Versus Time After Trip What is the minimum REQUIRED Safety Injection flow and what injection flowpath will be used?

The minimum REQUIRED SI flow is (1)

The injection flowpath that will be used is the _ _____._(2"'"'")___ flowpath.

A. (1) 380 gpm (2) normal charging B. (1) 380 gpm (2) high head safety injection C. (1) 220 gpm (2) normal charging D. (1) 220 gpm (2) high head safety injection Answer: D Explanation/Justification: KIA is met by placing the candidate into a specific point of an EOP, and requiring them to determine the minimum injection flow and flowpath by which they will operate the Safety Injection system when in ECA-1.1, "Loss Of Emergency Coolant Recirculation".

A. Incorrect. If the curve lower axis is read as hours, the flow requirement would be 380 gpm, however this exceeds the capacity of the normal charging flowpath.

B. Incorrect. If the curve lower axis is read as hours, the flow requirement would be 380 gpm, this would require the use of the HHSI flowpath.

C. Incorrect. The flow required is 220 gpm, however this is above the capacity of the normal charging flowpath.

D. Correct. Requires the use of the SI flow curve based upon time after the Reactor trip and an understanding of the injection flowpath capacities.

Provided the value used as a determinant since this is not required information from memory. The time after trip for 3 hrs is 180 minutes, the curve is at 220 gpm at this point. Above 150 GPM, the HHSI flowpath is used. Step 18, Establish Minimum SI Flow to Remove Decay Heat will establish flow through either the Normal Charging Header with flow control from the Control Room or flow will be established through the HHSI injection flowpath with local operator action required to throttle flow. Even through the step wording is for SI flow to remove decay heat, injection is not always via the SI flowpath.

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 20M-53A.1.A-4. 7 R2 11 c Technical

References:

20M-53A.1.ECA-1.1 Rev 1 lss 2 20M-53A.1. ECA-1.1 Rev 1 20M-53A.1.A-4. 7 Rev 2 lss 1C lss 2 pg 14 - 17 Question Source: Bank-1LOT14 NRG Exam (018)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 / 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 requirement for 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: KIA 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/E 12 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 /4 Generators)

KIA# EA 1.2 KIA Importance 3.6 Exam Level RO References provided to Candidate Technical

References:

20M-53A.1.ECA-2.1 lss.2 Rev. 0 None Question Source: Bank - 2010 Surry NRC Exam (018) 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 0 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-10 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 /7 KIA# AK1.01 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

20M-2.1.B, Rev. 3 pg. 3 & 17 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 / 41.10 / 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-TK23B] TO UNIT 2 COOLING TOWER SLOWDOWN is ready to start.

60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months 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 18-20 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after tank sample time (20M-25.4.N, P&L J). 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.N 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 19 KIA# AA2.02 KIA Importance 2.9 Exam Level RO References provided to Candidate RWDA-L for 2SGC-TK21A Technical

References:

20M-25.4.N Rev. 16 20M-25.4.N, pgs. 18-20 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 21 C SG AFW Throttle Viv [2FWE*HCV1 OOA] within 40 minutes of Auxiliary Feedwater actuation?

A. To prevent 'C' SG overfill.

B. To prevent runout of the running AFW pump.

C. To insure adequate supply of AFW to 'A' and 'B' SGs.

D. To prevent rapid cooldown of the RCS.

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 OF 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 CID 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. CID. Incorrect because AFW pumps do have backup supplies available.

D. Incorrect. Plausible distractor because they may feel that the limited C/D 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 Technical

References:

20M-56C.4.B Rev. 32 pg. 3 None 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 I 45.6 I 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 Cable Spreading Room (CB-2)

Shift manager determines that Control Room evacuation is likely to occur, and directs entering 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. Emergency Shutdown Panel B. Alternate Safe Shutdown Panel C. Main Control Room D. Reactor Trip breakers or MG Set Breakers Answer: C 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. 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.

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

C. 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.

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 plant has tripped due to a large break LOCA.

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 37.1 psig, and is currently 10.5 psig and 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

0. 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 ..

8. Incorrect. With CNMT pressure <11 psig, there are no entry conditions met into FR-Z.1. 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 None Technical

References:

20M-53A.1.F-0.5, lss. 2 Rev. 0 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.5,41.10 145.61 45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

24. Reactor trip and Safety Injection actuated 20 minutes ago due to faulted Steam Generators.

The following conditions exist:

RCS Pressure is 1550 psig and rising

PRZR level is 95% and rising

SIG A NR level reads 0%

SIG B NR level reads 0%

SIG C NR level reads 35% and slowly rising

RCS cold leg temperature indicates 230°F and slowly lowering

RVLIS Full Range level is 105%

SI flow indicates 400 gpm and stable

Total AFW flow is 180 gpm total

Containment pressure reads 27 psig and stable

All RCPs have been secured

The Crew has just completed isolation of 'A' and 'B' SIG's in E-2, "Faulted Steam Generator Isolation" Which of the following is the next procedure to be performed, and what is the procedural basis for this action?

A. FR-1.1, "Response to High Pressurizer Level" to restore Pressurizer level to normal.

B. FR-1.1, "Response to High Pressurizer Level" to minimize an overpressure condition.

C. FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition" to restore Pressurizer level to normal.

D. FR-P.1, "Response to Imminent Pressurized Thermal Shock Condition" to minimize an overpressure condition.

Answer: D Explanation/Justification: KIA is met by the candidate assessing the parameters given to decide that entry into safety function FR-P.1 (Orange path) is required, and they required to know that a mitigative strategy of FR-P.1 is to Terminate SI because it is a significant contributor to cold leg temperature decreases and intact RCS over pressurization.

A. Incorrect. Plausible distractor due to przr level at 95% and rising, and entry into F-0.6 (Inventory with RVLIS) is przr level >92%. The above conditions do meet Yellow path FR-1.1, but the FR-P.1 conditions are met for an Orange path at minimum, and rules of usage dictates FR-P.1 entry. Correct purpose of FR-1.1, but not for these conditions.

B. Incorrect. Plausible distractor due to przr level at 95% and rising, and entry into F-0.6 (Inventory with RVLIS) is przr level >92%. The above conditions do meet Yellow path FR-1.1. Preventing overpressure condition is the purpose of FR-P.1.

C. Incorrect. FR-P.1 conditions are met for an Orange path at minimum. Incorrect purpose.

D. Correct. FR-P.1 conditions are met for an Orange path at minimum, and rules of usage dictates FR-P.1 entry over a Yellow path Inventory.

Correct purpose of FR-P.1. The RO candidate is required to know that a mitigative strategy of FR-P.1 is to Terminate SI because it is a significant contributor to cold leg temperature decreases and intact RCS overpressurization.

Sys # System Category KA Statement W/E02 SI Termination / Generic Knowledge of the parameters and logic used to assess the status of safety functions, such as 3 reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

KIA# 2.4.21 KIA Importance 4.0 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1-0.4 lss.2 Rev. O 20M-53B.4.FR-P.1 lss. 2 Rev. 0 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41. 7 I 43.51 45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

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).

KIA# EK1.2 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.ES-1.2 lss. 2 Rev. 1 pg.11 Question Source: Bank - Farley 2011 NRC Exam (070)

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

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

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 manually closed with all block valves closed to minimize RCS leakage.

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

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

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

Answer: C Explanation/Justification: K/A 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/EO? 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 / 45.5145.6)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2Lor1s)

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' SGs.

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' SGs.

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

Answer: C Explanation/Justification: K/A 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 C/D 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.

K/A# 2.1.23 K/A Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-H.2 lss. 2 Rev.O 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 28, 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 28 System Service Transformer. When PCB-94 opens SSST 28 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 Technical

References:

20M-1.5.B.1 Rev. 2 pg. 2 None 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.F, "Station 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-condensable gases.

In accordance with 20M-7.4.F, "Degassing the Reactor Coolant System From The Volume Control Tank", which of the following completes the statements below?

(1) gas is used for purging the RCS of non-condensable gasses.

The non-condensable gasses from the VCT will be purged to the (2)

A. 1) Hydrogen

2) Primary Plant Sample System B. 1) Hydrogen

2) Boron Recovery System C. 1) Nitrogen

2) Primary Plant Sample System D. 1) Nitrogen

2) Boron Recovery System Answer: D Explanation/Justification: KJA 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 KJA statement states from the przr bubble space, it would not be possible to meet the KJA due to purging the gasses from the przr to the sample system bypasses the CVCS system which is the KJA 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 None Technical

References:

20M-7.4.F, Rev. 19 pg. 3 & 4 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: KIA 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)

KIA# K2.03 KIA 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 214 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 pumps (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 is the 'A' Charging pump capable of starting 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) capable of starting without the Aux Oil Pump running.

A. 1) to bypass

2) is B. 1) to bypass

2) is NOT C. 1) through

2) is D. 1) through

2) is 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.

0. 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 is the initial action taken to lower PRT temperature?

A. Lower RCS pressure by opening the Pressurizer spray valves [2RCS*PCV455A/B].

B. Open the Primary Water system supply valves [2RCS-MOV516 and 2RCS-AOV519].

C. Vent the PRT to the Degasifiers [2BRS-EV21A/B].

D. Drain the PRT to the Primary Drains Tank [2DGS-TK21].

Answer: B Explanation/Justification: KIA 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. Incorrect. With the safeties and PORVs closed, lowering RCS pressure will not lower PRT pressure if the assumption is that the PRZR and PRT are interconnected.

B. Correct. Temperature is lowered by opening 2RCS-MOV516 and 2RCS-AOV519 and spraying the PRT.

C. Incorrect. Pressure and level are normal therefore venting is not necessary. Venting to the Degasifiers is directed in the ARP for high pressure.

D. Incorrect. Pressure and level are normal therefore draining is not necessary. Draining to TK21 is directed in the ARP for high level.

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 KIA 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 (Q6)

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. Given the following conditions:

The plant is in Mode 3

Cooling down to Mode 5 for an inoperable 'A' CCP Train (LCO 3.7.7 Cond. B)

'B' Train of CCP becomes inoperable due to an unisolable leak on the 'B' supply header Based on the above information, what is the MINIMUM temperature the crew must stop the cooldown with both CCP Trains inoperable?

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 I 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 VLV will 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*AOV1 O?'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*AOV1 O?'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. 1 pg. 4 20M-53C.4.2.6.8 Rev. 12 pg 1 & 2 Question Source: Bank - Farley 2012 NRC Exam (014)

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*PCV455B 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: KJA 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.

B. 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 KJA 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)

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: K/A 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 2/4 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 2/4 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:

K/A# K6.02 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

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

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 / 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: KJA is met by demonstrating knowledge of the containment pressure channel inputs to both Safety injection and Cnmt lsol 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 2/4 and only High-1 is tripped.

D. Incorrect. Plausible if the logic for both are 2/4 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 (Q40)

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

8. 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: KIA 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.

KIA 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 KIA 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 / 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 is _ ___._(1_,_)_ _ _ met AND LCO 3.6.5, Containment Air Temperature is (2) met.

Refer to attached PCS screen A. (1) not being (2) not being B. (1) not being (2) being C. (1) being (2) not being D. (1) being (2) being Answer: B Explanation/Justification: K/A 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. Correct. Cnmt pressures are both above the 14.2 psia required by Tech Specs. The pressure 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 101.9 which is below the LCO required 108F. The CNMT PCS screen shot was assembled to purposely make the average temp and pressure a BAD reading so the candidate would have to average all of the temperatures manually to determine average temperature then compare this to the LCO requirement w/o reference to the LCO. 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.

C. Incorrect. See correct answer explanation.

D. Incorrect. See correct answer explanation.

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

C. Incorrect. Plausible distractor if candidate thinks BOTH a CIB 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 CIB switches on the Bench Board, 2 for each train, and 1 train will actuate both CIB trains.

Sys# System Category KA Statement 026 Containment A 1 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. 20ST-13.1, "Quench Spray Pump [2QSS*P21A] Test" (Comprehensive Pump Test) was just completed.

Review the attached Data sheets 28 and 4, and determine the operability status of the

'A' Quench Spray Pump.

What is the status of the 'A' Quench Spray Pump?

'A' Quench Spray Pump is _ _ _ _ _ __

A. inoperable due to vibration data only B. inoperable due to pump differential pressure only C. inoperable due to both vibration data and pump differential pressure D. operable Answer: C Explanation/Justification: KIA 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. The pump is inoperable due to V1 pump outbd axial vibration being high above the alert range which makes it inoperable per the note on Data Sheet 2B. It is also inoperable due to the pump DIP of 134.4 psid based on the note on Data sheet 4.

D. Incorrect. See correct answer justification.

Sys# System Category KA Statement 026 Containment Generic Knowledge of surveillance procedures.

Spray System (CSS)

KIA# 2.2.12 KIA Importance 3.7 Exam Level RO References provided to Candidate 20ST-13.1 Data sheet Technical

References:

20ST-13.1 Rev. 32 2B&4 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 45.13)

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 MSIVs inadvertently closing at low power conditions, and the response of the AFVI/ pumps to these changing conditions.

A. Incorrect. Plausible if it is thought that the MDAFVI/ pump started on 213 lo-lo SG water level on 213 SGs due to a Rx trip. Also, it could be thought that MDAFVI/ 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 SIG water levels above the lo-lo SG water level setpoint which would trip the reactor and auto start AFVI/ pumps.

C. Incorrect. Plausible if it is thought that the TDAFVI/ pump started on 213 lo-lo SG water level on 113 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 AFVI/ 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 AFVI/ pumps.

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 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 plant is at 65%.

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: K/A 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 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.

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

KIA# 2.4.1 KIA Importance 4.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.E-O, Rev. 1 lss. 2 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) 45.

The plant is at 90% power The Balance of the Plant Operator observes the above indications. What are the required actions for the crew?

A. Trip the Rx B. Restart the MFP motor C. Commence Unplanned Power Reduction AOP 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. The candidate must know that if one pump motor is running and the other motor trips, a trip of the running motor will occur.

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 condition were <80% power. Unplanned Power Reduction AOP is plausible if candidate didn't realize AOP-2.24.1 would lower power and if the plant initial power had been <80%.

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 tumiRe 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 correct AOP to respond to this event when annunciator A6-1 OE "SG 21 B 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: KJA 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.

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 Feedwater actuation of AFW (MFW) System the MFW; and (b} based on those predictions, use procedures to correct, control, system or mitigate the consequences of those malfunctions or operations:

KIA# A2.01 KIA Importance 3.4* 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 / 43.5 / 45.3 /

45.13)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15) 4 7. The reactor has tripped from 100% power due to a small RCS leak. E-0, "Reactor Trip or Safety Injection" has been exited, and ES-1.1, "SI Termination" is being performed.

The following conditions exist:

Containment pressure is 1.2 psig

Pressurizer pressure is 1752 psig and slowly rising

RCS T avg is 534 °F and slowly lowering

AFW flow has been throttled to limit the cooldown

ES-1.1 Step 21, Check Intact Steam Generator Levels is being performed The Steam Generator parameters are as follows:

SG Narrow Range Level AFW Flow

'A' 10% 116 gpm

'B' 5% 114 gpm

'C' 8% 119 gpm

1) What is the current condition of the Red Path Heat Sink Status tree?

2) Which of the following power levels would generate more decay heat after the Reactor Trip?

Entry into FR-H.1, "Response to Loss of Secondary Heat Sink" _ _ (1 )_ _ required, and more decay heat load after the Reactor Trip is generated from (2) reactor power.

A. 1) is

2) 25%

B. 1)isNOT

2) 25%

C. 1) is

2) 100%

D. 1) is NOT 2 100%

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 distractor with all SG levels <12%. See correct answer below.

8. Incorrect. H-1 entry is not required, but 25% power is incorrect for magnitude of decay heat generation. See correct answer below.

C. Incorrect. Plausible distractor with all SG levels <12%. See correct answer below.

D. Correct. With total AFW flow being >340 gpm, there is sufficient decay heat removal capability even though all SG NR levels are <12%. To meet the KIA for decay heat magnitude, the decision of which power level produces the highest decay heat has to be made. 100% power produces more decay heat after a Rx trip than 25% power.

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 None Technical

References:

20M-53A.1.F-0.3 12 RO, GO-GPF.R8 A Rev. 1 Question Source: Bank-1LOT14 (047) Modified 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. The plant is at 15% power in preparation for turbine startup 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 low.

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

Load Tap Changer will not respond in Auto or Manual.

Based on the above conditions:

1) What will be the status of the plant 5 minutes after bus voltage drifts to 3800 VAC?

2) What procedure (if any) will be used to mitigate this condition?

A. 1) The plant will be in Mode 3 due to an Automatic Rx Trip

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

2) AOP 2.36.2, "Loss of 4KV Emergency Bus" C. 1) Main Feedwater will be lost

2) AOP 2.24.1, "Loss Of Main Feedwater" D. 1) The plant will remain at power

2) No procedural guidance is required.

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-7 (10%) Rx trip would be a possibility if 2/3 RCPs tripped. 1 RCP tripping >P-8 (30%) will trip the Rx. 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. Main Feedwater will not be lost since 'B' MFP was running and it is powered from normal bus 2C & 20. At this point in the procedure only one MFP would be running. A second MFP would not be started until feed flow is 6.3-6.8 mpph.

D. Incorrect. Plausible distractor if it is not recognized that the voltage dropped below the 93.4% and the 2AE bus was lost. Most people remember the 75% voltage drop for 1 second, and may feel that nothing will occur to the plant.

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:

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

References:

LSK-022-005B 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 K 1 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 / 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 (2LOT15)

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_)___ 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: KJA 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. 19 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 _ _ _ _ _ __

A. discharged through the Ventilation Vent to atmosphere B. discharged through the Elevated Release to atmosphere C. discharged to the Auxiliary Building D. discharged 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 GIB 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 GIB 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. GIB 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 (GIB).

Sys# System Category KA Statement 076 Service A1 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 2SQS-30.1 PPT, Rev. 23 Question Source: Bank - 2LOT7 NRG Exam (053)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 I 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 NRG Exam (053)

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:

KJA# K4.01 KJA Importance 3.0* Exam Level RO References provided to Candidate Technical

References:

20M-44C.4.A Rev. 23 pg 4 None 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 annunciators are in the Control Room 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. Isolate Cooling Water flow to the operating CNMT Air Recirc (CAR) Fans C. Shutdown an operating CNMT Air Recirc (CAR) Fan D. Start a CNMT Vacuum Pump Answer: D Explanation/Justification: KIA 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 low temperature, and must be raised back into normal band.

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.

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. 4 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: KIA 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).

B. 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. Safety Injection Actuation

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

4. PRZR Pressure Low Rx Trip

5. PRZR Heaters Control Group C [2RCP-H2C] ON A. 3, 5, 1, 4, 2 B. 3, 1, 5, 2, 4 C. 5, 3, 1, 4, 2 D. 5, 1, 3, 2, 4 Answer: A Explanation/Justification: KIA 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 temperature, and PRZR spray flow control on the non-faulted spray valve, as well as automatic SI flow actuation into the RCS.

A. Correct. The Master Pressure Controller will respond to the lowering pressure by controlling the spray valve and heaters. PRZR Pressure Protection channels will control the Low pressure Rx trip, and the SI Initiation. The MPC output will be driving to O 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), Rx trip at 1945, and SI at 1856 psig.

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 & 11 (LT459 & 460)

PRZR level is on program The reference leg for PRZR Channel I Level [2RCS*LT459] develops a leak.

With NO operator action, how will the following PRZR level indicators in the Control Room respond 5 minutes after the leak develops?

If the event continues with NO operator action, a Rx trip signal _ _ _ _ _ be generated.

Rx trip 2RCS*Ll459 2RCS*Ll460 Signal A. Rise Lower will B. Lower Rise will NOT C. Rise No change will D. Lower No change will NOT Answer: A Explanation/Justification: K/A 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 all 3 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%. 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. The operable (LT460) level instruments would indicate a higher PRZR level in this case because charging flow would be maximized. A rx trip would be generated on high przr level due to LT459 isolating UD.

C. Incorrect. A reference leg leak will cause the affected PAM channel (LT459) to indicate high. The No change indications on the operable PAM indicators is plausible if the candidate thinks that LT459 is NOT the controlling channel for 2CHS-FCV122, and will have no effect on charging.

Plausible distractor of Rx trip if candidate feels that LT459 and LT461 both rise to the przr level trip setpoint.

D. Incorrect. Plausible because a variable leg leak on LT459 will cause indicated level to lower. The No change indications on the operable PAM indicators is plausible if the candidate thinks that LT459 is NOT the controlling channel for 2CHS-FCV122, and will have no effect on charging.

Rx not tripping would be correct for the assumed failure conditions of this answer.

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 Technical

References:

GO-GPF.C7 Rev. 4 pg. 55, 20M-1.5.B.1 Rev. 2 None 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 I 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 1-2B, Hydrogen Level High/High-High is in alarm Complete the following statements. Assume the TSC has been contacted and concurs with the decision.

The High-High Hydrogen concentration in Containment setpoint is (1)

In accordance with the Hydrogen Level High/High-High ARP, the crew will ----""'(2;;..r..)___ in response to the High-High Hydrogen level in Containment.

A. 1) 0.5%

2) intentionally ignite the Containment atmosphere B. 1) 0.5%

2) start [2HCS-FN21] Containment Atmosphere Purge Blower C. 1) 4.5%

2) intentionally ignite the Containment atmosphere D. 1) 4.5%

2) start [2HCS-FN21] Containment Atmosphere Purge Blower 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. 0.5% is the High H2 alarm. Plausible distractor of igniting the atmosphere since this is an option in our severe accident management guidelines, but this is not the correct response at this H2 level or iaw the ARP.

B. Incorrect. 0.5% is the High H2 alarm. 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, 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:

KIA# K5.01 KIA 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

An Induced Containment Purge through the SLCRS Unfiltered flow path is in progress

The Personnel Airlock and Equipment Hatch are closed

The Cnmt Purge Supply lsol Damper 2HVR*MOD258 is inadvertently CLOSED

NO other components reposition Which of the following containment parameters will be the FIRST to be affected by this failure?

A. Pressure B. Temperature C. Radiation Level D. Humidity Level Answer: A Explanation/Justification: KIA is met with the knowledge of the Containment Purge system being aligned for induced purge, and the Cnmt Purge Supply lsol Damper goes closed, what containment parameter will be effected due to this misalignment.

A. Correct. The Induced purge flowpath has Leak Collection Normal Exhaust Fan taking suction on the Cnmt exhaust line. When the Containment Purge Supply lsol Damper is closed, the cnmt pressure will lower.

B. Incorrect. Temperature will remain constant since the Induced Purge flowpath does not provide a cooling function C. Incorrect. Radiation levels would only rise to cause a purge isolation, they would not rise because of an isolation D. Incorrect. Humidity is a function of the containment temperature and dewpoint, which are unaffected by purge control operation.

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 Question Source: Bank - 1LOT7 NRC Exam (Q28)

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. The following conditions exist:

The plant is in Mode 6

Core off-load is in progress

A4-5D, NIS Source Range High Flux at Shutdown is in alarm

Both SR channels indicate a rising Neutron rate and positive startup rate Based on the above conditions, Fuel Handling personnel are required to stop fuel movement and evacuate the (1)

The High Flux at Shutdown setpoint is ----"'(2_,,)_ _ _ decade above the SR detector background count rate.

A. 1) Fuel Building

2) 1/2 B. 1) Fuel Building

2) 1 C. 1) Containment Building

2) 1/2 D. 1) Containment Building

2) 1 Answer:C Explanation/Justification: KIA is met with the knowledge of the cause and effect relationship between the NIS Source Range High Flux at Shutdown alarm being in alarm, and what effect it will have on the fuel movement in containment.

A. Incorrect. Plausible distractor is the candidate thinks the High Flux at Shutdown alarm is monitoring the core offload in the Fuel Pool and the setpoint was based on the SR values in the core. 1 decade is an incorrect value for the setpoint.

B. Incorrect. Plausible distractor is the candidate thinks the High Flux at Shutdown alarm is monitoring the core offload in the Fuel Pool and the setpoint was based on the SR values in the core. 1/2 is the correct setpoint.

C. Correct. When both SR channels indicate a rising Neutron rate and positive startup rate, the CR is directed to evacuate the Cnmt Building iaw.

A4-5D ARP. In accordance with 20M-2.4.H, the setpoint is set at Y:z times (3.16) SR background.

D. Incorrect. This is the correct action for the conditions. It is an incorrect value for the setpoint.

Sys# System Category KA Statement 034 Fuel Handling K1 Knowledge of the physical connections and/or cause effect NIS Equipment relationships between the Fuel Handling System and the following System systems:

(FHES)

KIA# K1.04 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

20M-2.4.H Rev.4 20M-2.4.AAG Rev. 1 Question Source: New 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 (2Lor1s)

62. Given the following conditions:

'C' MAIN FEEDWATER REG VALVE 2FWS*FCV498 failed CLOSED

Reactor tripped on Low-Low Steam Generator level

1) What is the Steam Generator level setpoint for the Motor Driven Auxiliary Feedwater Pumps to automatically start?

2) What is the basis for Auxiliary Feedwater automatically initiating?

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 MDAFW 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 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 S/G 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: Higher - Comprehension or Analysis 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 Prat Controller Operating" alarms

SIG pressures are 1000 psig and stable When Turbine speed reaches 1870 RPM, how will the Reactor and Turbine EHC valves respond?

(IV= Intercept valves, GV =Governor valves, TV= Throttle valves)

Reactor Status Turbine EHC valves A. At Power Only IV's and GV's Close B. At Power ALL IV's, GV's, and TV's Close C. Not at power Only IV's and GV's Close D. Not at power ALL IV's, GV's, and TV's Close Answer: A Explanation/Justification: KIA 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. The Rx will remain at power. 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. The Rx will remain at power.

C. Incorrect. It is correct that only the IVs and GVs will close. It is incorrect to believe that the Rx will trip. Plausible distractor because a Turbine trip generates a Rx trip when >P-9. Power was set at 13% to be >P-10 to support the plausibility of the distractor.

D. Incorrect. The TVs are not closed by the OPC which makes the distractor incorrect. It is incorrect to believe that the Rx will trip. Plausible distractor because a Turbine trip generates a Rx trip when >P-9. Power was set at 13% to be >P-10 to support the plausibility of the distractor.

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 Technical

References:

20M-26.4.AAU lss.1 Rev. 5 None 20ST-26.8 Rev. 16 pg.15 Question Source: Bank- Vision #138661 Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 145.5 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*FN241B] 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-43.170 Rev. 44 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 45.3 /

45.12)

Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

65. Given the following conditions:

The plant was operating at Full Power with all systems in NSA

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"

Emergency Diesel Generator #1 has failed to start

All other systems function as designed Which of the following describes the status of power to 2SWS*P21A and P21 B, "Service Water Pumps"?

2SWS*P21A has (1) 2SWS*P21 B has (2)

A. (1) power (2) power B. (1) no power (2) power C. (1) power (2) no power D. (1) no power (2) no power Answer: B Explanation/Justification: KIA is met by demonstrating the knowledge of the available power to the essential SWS pumps following a loss of offsite power coincident with an EOG start failure.

A. Incorrect. Incorrect that 2SWS*P21A has power (refer to correct answer explanation) Correct 2SWS*P21 B status.

B. Correct. 2SWS*P21A is normally powered from Bus 2AE. Since there is no offsite power and EOG 2-1 failed to start, 2SWS*P21A has no power.

2SWS*P21 Bis powered from Bus 20F. Since all other systems functioned as designed, EOG 2-2 started and is supplying Bus 20F so therefore 2SWS*P21 B does have power.

C. Incorrect. Incorrect that 2SWS*P21A has power (refer to correct answer explanation) Incorrect 2SWS*P21 B status. Plausible if the candidate does not understand integrated plant status and power supplies.

D. Incorrect. Correct that 2SWS*P21A has no power. Incorrect 2SWS*P21 B status 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-30.3.C Rev. 15, Pg. 8 & 10 None 3SQS-36.1 PPNT U2 Rev. 12 lss. 1 Slide 10 Question Source: Bank - 1LOTS NRC Exam (Q64)

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 10 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 NIA 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. 9 pg. 82 Question Source: Bank - 2LOT6 NRC Exam (066) Modified 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%

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: KIA 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 NIA 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 Alt. 1 Question Source: Bank- Harris 2012 NRC Exam (067)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 43.5 / 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. No action is required, continue with the procedure, 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.

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 N/A 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 I 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ago, a valve was repositioned out of NSA and selected as an OPEN item using the Short Term Configuration Change Process.

Based on the requirements of NOP-OP-1014, "Plant Status Control", does this comply with the Short Term Configuration Change Process?

A. Yes; a clearance will only be necessary if restoration does not occur within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. Yes; a system status print sheet will be necessary if restoration does not occur within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. No; a clearance should have been posted 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ago.

D. No; a system status print sheet should have been issued 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ago.

Answer: C Explanation/Justification: K/A 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months as opposed to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ago.

D. Incorrect. Correct that it does not comply with the short term configuration control process. 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ago.

Sys# System Category KA Statement N/A 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. 3, pg. 14 Question Source: Bank - 1LOTS NRC EXAM (096)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 43.3 I 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 a s 1 hr. TS action statement.

A. Incorrect. TS 3.5.4 Surveillance requires RWST borated water temperature to be ~45 F and s 65 F. therefore there is no TS LCO entry required for this distractor.

8. 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months action statement.

Sys# System Category KA Statement NIA NIA 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if you wear a full-face respirator.

The valve lineup will only take you 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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: KIA 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 NIA 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) MOD Question Cognitive Level: Higher - Comprehension or Analysis 10CFR 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 NIA 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.11Pg.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: KJA 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 stern information.

B. Incorrect. ES-0.1 cools the plant down to mode 5 condition so there is no need to stop cooldown unless 1OOF/hr 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 N/A N/A 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 (Q58)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 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 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 NIA NIA 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 N/A 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.18 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 IBJ 11 D 111 D IV D Reactor Type: W [8] CE D BW D GE D 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete the combined examination, and 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 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 L100 Points Applicant's Scores I I Points Applicant's Grade I I Percent NUREG-1021, Revision 10 FENOC Facsimile r1

(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) 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) 45 minutes

2) E-0, "Reactor Trip or Safety Injection" C. 1) 15 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.

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 is the correct time with UD isolating at 14%. Correct procedure transition per ES-0.1 LHP.

C. Incorrect. This would be the time if letdown didn't isolate at 14%. 20-0% (2000 gal)@ 140 gpm=14.3 min. 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 None Technical

References:

20M-53A.1.ES-0.1 lss. 2 Rev. 3 Question Source: New 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)

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 statements?

The required EAL classification is (1)

An Offsite Protective Action Recommendation (PAR) _ _--"'(2;:;.,o.)___ required.

A. 1) Site Area Emergency

2) is B. 1) Site Area Emergency

2) is NOT C. 1) General Emergency

2) is D. 1) General Emergency

2) is NOT

(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 and PAR requirements. 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.

General Emergency declared based on the following conditions.

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. PAR is required.

B. Incorrect. Plausible if it is not recognized that FR-C.1 entry conditions have been met for Fuel Clad failure. Plausible that PAR is not required if Site Area Emergency is the declaration.

c. Correct. GE based on answer explanation above. IAW 1/2-EPP-IP-4.1, when a GE is declared, a PAR must be provided to State/County within 15 minutes.

D. Incorrect. GE is correct based on answer explanation above. Plausible that a PAR is not required if the candidate thinks that a radiological release had not occurred.

Sys# System Category KA Statement 000011 Large Break Generic Knowledge of the emergency action level LOCAi 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-1-1 b.F01 Rev. 0 1/2-EPP-IP-4.1, Offsite Protective Actions rev. 31 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 / 43.5 / 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 21 B and 21 C N R 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 None Technical

References:

20M-53A.1.E-2 lss. 2 Rev 0 20M-53A.1.E-1 lss. 2 Rev. 1 Question Source: Bank- 2LOT5 NRC Exam (Q47)

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.

K/A 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 EDG 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.

B. 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 / 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:

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 just occurred, resulting in damage to the switchyard and a Loss of Offsite Power Based on these conditions and events:

1) What procedural transition from ECA-1.2 is REQUIRED?

2) Which of the following Abnormal Operating Procedures will be performed in conjunction with 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.

KIA 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 are not met for entry into this AOP.

B. Correct. If RCS pressure is not rising, then IAW 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, none of which are given.

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 20M-53C.4A.75.3 Rev. 19, pg. 1 Question Source: Bank - 2LOT6 NRG Exam (Q81) Modified 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, apply adverse criteria and select 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 None Technical

References:

FR-H.1 pg 2 Rev. 1 lss. 2 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. 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.

1) What is the MAXIMUM number of rods capable of being placed on DC Hold?

2) What is the OPERABILITY status of the rods when the rods are on DC Hold?

A. 1) 4 rods

2) Operable B. 1) 8 rods

2) Operable C. 1) 4 rods

2) Inoperable D. 1) 8 rods

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.

K/A 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). 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 None Technical

References:

20M-1.4.0 Rev. 0 lss. 1 pg. 1 TS Bases pg. B 3.1.4-5 rev. 0 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. Initial conditions:

The reactor failed to trip after receiving a trip signal

SRO transitioned from E-0, Reactor Trip or Safety Injection, to FR-S.1, Response to Nuclear Power Generation/AlWS

Reactor Power is 23% and decreasing 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 SID 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. 0 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 43.51 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

/4 operations.

KIA# EA2.1 KIA Importance 4.2 Exam Level SRO References provided to Candidate Technical

References:

20M-53A.1.FR-C.2 lss. 2 Rev. 2 None 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 Which of the following procedures will be entered and what is the MAXIMUM allowable RCS Cooldown rate?

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: B 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 C/D 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. Correct. With pressure <1950psig and Tho! <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 only one train of RVLIS is OOS, therefore RVLIS is available.

The cooldown rate in ES-03 is 100F/hr.

C. 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. Cooldown rate is correct for ES-0.4 when >500F ..

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-0.3 lss. 2 Rev. 1 pg.3 Question Source: Bank- 1LOT14 NRC Exam (085) 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)

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: 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 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.

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.

8. 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 Technical

References:

20M-53C.4.2.6.8 Rev. 12 None 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 has SLOWLY trended up to 100.1 %.

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-401Attachment2 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.

K/A 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:

K/A# A2.33 KIA Importance 3.3 Exam Level SRO References provided to Candidate None Technical

References:

20M-53C.4.2.51.2 Rev. 2 pg 1 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 / 45/5)

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 the Safety Injection Signal, how would the operators determine that a single reactor trip breaker failed to open following the trip?

Annunciator A12-1C, Auto Safety Injection Blocked will (1)

(Assume SI signal has been reset and both Rx Trip Breakers are OPEN)

Immediately after the Safety Injection Signal has been reset, a Loss of Offsite Power occurs and RCS pressure lowers below the Safety Injection setpoint.

How will the Safety Injection Signal actuate?

The Safety Injection Signal will be _ _-->,;:(2::..<,)___ actuated.

A. (1) intermittently flash (2) automatically B. ( 1) intermittently flash (2) manually C. (1) not illuminate (2) automatically D. 1) not illuminate (2) manually

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 88 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 procedure caution related to the ESFAS signals and action required to initiate the ESF signal per the procedure bases. The caution alters the SRO that the SI signal would have to be manually initiated to restart the safeguards equipment should the EOGs start and energize the bus, after the SI signal has been reset. The knowledge of the plant response to a failure of a single train of SI to reset is background knowledge in the EOP.

KIA is met by demonstrating the knowledge of the Caution in ES-1.1 regarding manual actuation of Safety Injection if a LOOP event occurs.

A. Incorrect, the annunciator response is correct. The SI signal will not automatically actuate and load SI equipment on the EOG.

B. Correct, The Caution prior to step 1 of procedure ES-1.1 alerts the operator that a SI signal will not Automatically occur on a LOOP, manual action to initiate SI or load equipment will be required to start equipment when the EOG re-energizes the bus. The operator must have the knowledge of how the SI blocked annunciator responds if a single RTB does not open on the trip. When resetting the SI signal the alarm will flash in and out as one train of SI is not reset.

C. Incorrect, the annunciator response is not correct, the alarm will flash as the two trains of SSPS compare inputs. The manual SI signal will allow the EOG to start ESF loads.

0. Incorrect, the annunciator response is not correct, the alarm will flash as the two trains of SSPS compare inputs. The SI signal will not automatically actuate and load SI equipment on the EOG 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 20M-53B.4.ES-1.1 lss. 2 Rev. 0 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 (2Lor1s)

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 annunciator is LIT

[2SWS-PI 113A and Pl 1138], Service Water Header Pressures are 48 psig and stable

[2SWS*MOV107A, B, C, DJ, 'Sec Comp Clg Wtr Hx Serv Water Supply Hdr lsol Vlvs' are OPEN

NO other Annunciators are currently LIT

AOP-2.30.1, Service Water/Main Intake Structure Loss has been entered Based on the above conditions, what could cause this Service Water System condition, and what would be the correct procedure to mitigate the condition?

The above conditions could indicate a _ _ _(_1)..___ _ , and would be mitigated by (2)

A. 1) Service Water Pump trip

2) continuing in AOP-2.30.1, Service Water/Main Intake Structure Loss ONLY B. 1) Service Water Pump trip

2) manually tripping the reactor, and entering E-0, Reactor Trip or Safety Injection C. 1) Service Water System leak

2) continuing in AOP-2.30.1, Service Water/Main Intake Structure Loss ONLY D. 1) Service Water System leak

2) manually tripping the reactor, and entering E-0, Reactor Trip or Safety Injection

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15) question 89 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 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. 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.

B. 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.

C. 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, A1-4F, '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).

D. 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.

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 None Technical

References:

20M-53C.4.2.30.1 Rev. 9 Question Source: New Question Cognitive Level: Higher - Comprehension or Analysis 10 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 (1) within 10 minutes of entering 20M-56C.4.B, "Unit Supervisor Procedure".

The reason for this action is to ~~~~~~~~---.._~~~~~~~-

2 A. 1) Closed

2) ensure the valve remains in the required position B. 1) Closed

2) prevent thermal shock to the RCP seals C. 1) Open

2) provide an RCS inventory control flow path D. 1) Open

2) supply seal injection flow to the running RCP seals 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 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.

KIA 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. Incorrect. Closed is the incorrect failure position. Plausible distractor since air operated valves are place in their desired positions to prevent spurious fire induced operation.

B. Incorrect. Closed is the incorrect failure position. Preventing thermal shock is a plausible distractor, but 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. 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.

D. Incorrect. Failing 2CHS*HCV186 open is correct. Seal injection flow is plausible if the candidate thinks the RCPs are running and seal inj flow is required. The 3rt1 CR action for the ATC Operator is to trip the RCPs, therefore there are no RCPs running.

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.

KIA# 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 / 43.5 / 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, AND what will be the Technical Specification required action?

A 5% (Keff < .95) shutdown margin will be (1)

The Technical Specification required action will be that fuel movement in the SFP (2)

A. (1) maintained regardless of SFP Cb (2) can continue as long as fuel is moved into proper SFP regions B. (1) no longer maintained regardless of SFP Cb (2) must be immediately suspended and boron concentration restored within limits C. (1) maintained as long as no credible boron dilution event reduces SFP Cb < 495 PPM (2) can continue as long as fuel is moved into proper SFP regions D. (1) maintained as long as no credible boron dilution event reduces SFP Cb< 495 PPM (2) must be immediately suspended and boron concentration restored within limits

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

Question 91 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 the impact on SOM per the TS bases and TS action required. Detailed knowledge of the bases is required to determine the impact of the boron dilution and which TS actions are applicable.

KIA is met by predicting the impact of SFP Boron concentration lowering on the SFP shutdown margin, and in accordance with Tech Specs, stop fuel movement and raise Boron concentration.

A. Incorrect. Keff cannot be maintained < .95 for a credible dilution event. (refer to correct answer explanation)

B. Incorrect. It is not correct that Keff < .95 is maintained regardless of SFP Cb in Unit 2. The second part of the statement is correct.

C. Incorrect. First part is correct (refer to correct answer explanation). Second part is incorrect action statement for Unit 2 but plausible for Unit 1.

D. 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. TS 3.7.16 required action is to immediately restore SFP Cb to> 2000 ppm and suspend fuel movement within 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 None Technical

References:

TS 3.7.16, Amend. 278/161 TS 3.7.16 Bases, Rev. 5 Question Source: Bank - 2LOT7 NRC Exam (Q93)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5 / 43.5 / 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 '8' 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 steam relief flow paths are available to MANUALLY cooldown the RCS in accordance with E-3 step 6?

A. 1

8. 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.

K/A 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.

K/A# 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 43.5 / 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

[21AS-Pl106] STA INSTR AIR HEADER PRESSURE is 80 psig and LOWERING

[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. 18 None 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:

(SRO ONLY)

Beaver Valley Unit 2 NRC Written Exam (2LOT15)

94. Crew composition may be less than the minimum requirement of 10CFR50.54 for a period of time not to exceed (1) , in order to accommodate an unexpected absence of an on-duty shift crew member.

If you are unable to fill the vacant position within the required time, plant shutdown (2) required in accordance with Tech Spec 5.2.2.

A. 1) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2) is B. 1) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2) is not C. 1) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

2) is D. 1) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

2) is not 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.

KJA 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

KIA# 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: 41.1o1 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 N528 are Out of Service Which of the following activities can be performed WITHOUT violating the Technical Specification required actions for Source Range Instrumentation?

A. Install a temporary secondary source into a core location.

8. Latch and move a spent fuel assembly from the core to the Spent Fuel Pool.

C. Latch and move a spent fuel assembly from the upender to the Spent Fuel Pool.

D. Add Hydrogen Peroxide mixed with primary grade water to the refueling cavity for cleanup.

Answer: C Explanation/Justification: Meets the requirements of the SRO only guidance of ES-401 Attachment 2 per section II .8 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.

KJA is met with the knowledge of permissible actions iaw Tech Specs during core alterations when a Source Range detector is inoperable.

A. Incorrect. Plausible that operationally an alternative source could be installed, however, it is not allowed by the definition for what constitutes a Core Alteration.

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. 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.

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 83.9.2 Pg. 83.9.2-2 Question Source: Bank - 1LOTS NRC Exam (Q95)

Question Cognitive Level: Higher - Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 I 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 INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure acceptable power distribution limits are maintained.

B. The rod is INOPERABLE.

Realign the rod within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure Shutdown Margin is maintained.

C. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure acceptable power distribution limits are maintained.

D. The rod is OPERABLE.

Realign the rod within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure Shutdown Margin is maintained.

Answer: C 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.

KIA 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. Incorrect. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 SDM would be affected. Power distribution limits are the correct reason for misaligned rods.

B. Incorrect. Would be true if the rod were untrippable.

C. Correct. Since the rod is trippable it is operable. Restore rod to within alignment limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

D. Incorrect. Correct, the rod is operable, but the concern for the situation presented is not shutdown margin.

Sys # System Category KA Statement N/A NIA 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 / 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) This authorization is limited to what maximum TEDE limit?

A. 1) Emergency Director

2) 10 Rem TEDE B. 1) Emergency Director

2) 75 Rem TEDE C. 1) Radiological Controls Coordinator

2) 10 Rem TEDE D. 1) Radiological Controls Coordinator

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 Radiological Controls Coordinator will advise 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 Radiological Controls Coordinator will advise 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 NIA N/A 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If grab samples are taken, these samples are to be analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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.

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-1 GW-109 shall not be used as a comparable alternate monitoring channel iaw Alt. 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)

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. A Site Area Emergency has been declared. In accordance with 1/2-EPP-IP-3.2, 'Site Assembly and Personnel Accountability', which of the following members of the Emergency Response Organization can order a Site Assembly and/or Personnel Accountability?

A. Emergency Director B. Emergency/Recovery Manager C. Security Coordinator D. Operations Coordinator 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 Emergency Plan and position responsibilities for the Emergency Director. This is a SRO position function only.

KJA is met with the knowledge of whose authority is required to order site assembly and/or personnel accountability.

A. Correct. ED responsible for ordering Site Assembly and Site Accountability when a Site Area or General Emergency which mandates Personnel Accountability is declared. The ED may also order Accountability if the situation warrants.

8. Incorrect. ERM is stationed at the EOF at a SAE or GE, and will assume primary responsibility for offsite emergency response activities by FENOC personnel during an emergency.

C. Incorrect. Security Coordinator is responsible for the actions pertaining to Site Assembly and Site Accountability.

D. Incorrect. Operations Coordinator remains cognizant of control room and in-plant activities, and matters concerning plant operations.

Sys# System Category KA Statement NIA NIA Generic Knowledge of the emergency plan.

KJA# 2.4.29 KJA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

1/2-EPP-IP-3.2 rev. 19 Question Source: New Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.51 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 1215

The initial report to state and local government was completed at 1227

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 1255 B. 1300 1310 C. 1300 1345 D. 1327 1355 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 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. All distractors are plausible but incorrect as they are intervals of the given time in the question stem.

B. 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.

C. Incorrect. All distractors are plausible but incorrect as they are intervals of the given time in the question stem.

D. Incorrect. All distractors are plausible but incorrect as they are intervals of the given time in the question stem.

Sys# System Category KA Statement NIA N/A 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)

Question Cognitive Level: Lower - Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 41.12 I 43.51 45.11)

Objective:

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

Date: Facility/Unit: Beaver Valley Unit 2 Region: I l:&l 11 D 111 D IV D Reactor Type: W [8] CE 0 BW D 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 ZS Points Applicant's Scores Points Applicant's Grade Percent NUREG-1021, Revision 10 FENOC Facsimile r1

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. AS), will alarm in the FIRST 45 seconds AFTER N-43 fails High?