Text

Final Written Examination with Answer Key (401-5 Format) (Folder 2)
	ML14246A237
Person / Time
Site:	Beaver Valley
Issue date:	07/01/2014
From:	; FirstEnergy Nuclear Operating Co
To:	Brian Fuller; Operations Branch I
Shared Package
ML14058A120	List: ML14183A844; ML14246A231; ML14246A232; ML14246A233; ML14246A234; ML14246A235; ML14246A236; ML14246A237; ML14246A240; ML14247A021; ML15008A076; ML15008A101;
References
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Beaver Valley Unit 1 NRC Written Exam (1LOT14)

1. The plant is at 40% power with all systems in normal alignment for this power level.

The 'A' RCP breaker trips open on overcurrent.

The following conditions exist:

RCS Tavg is 552 oF and lowering

RCS Pressure is 2150 psig and rising

'A' Rx Trip breaker is Open

'B' Rx Trip breaker is Closed How will the plant respond to these conditions?

A Turbine will fail to trip B. Partial Feedwater isolation will fail to occur C. Steam Dumps will maintain Tavg approximately 547 oF on the Rx Trip controller D. Steam Dumps will maintain Tavg approximately 550 oF on the Load Rejection controller Answer: D Explanation/Justification:

A. Incorrect. The turbine will trip due to SOV-TB20-AST-1 (P-4A) energizing. Plausible distracter due to SOV-TB20-AST-2, SOV-TB20-0PC1 &

OPC2, and SOV-1TB20-ET not energizing open without P-4B.

B. Incorrect. Partial feedwater isolation will occur due to Tavg < 554F and P-4A. Plausible because the candidate must know that the P-4A and P-4B are inputs to solenoid vent valve which are in series to cause the MFWRVs to go closed.

C. Incorrect. Without the P-4B relay, the steam dumps will remain on the Load Rejection controller and control approximately 3F higher than Tref.

Plausible distracter since the candidate must know the inputs to arm and actuate the steam dumps and controllers.

D. Correct. The steam dumps will arm and actuate on the Load rejection Controller and maintain Tavg approximately 3F higher than Tref. Tref will be 547 with the turbine tripped and 0% power, therefore 550F is a correct value. P-4B will not actuate since the B Rx Trip Breaker remained closed.

Sys # System Category KA Statement 000007 Reactor Trip/1 EK2 Knowledge of the interrelations between a reactor trip and the following: Breakers, relays and disconnects KIA# EK2.02 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

10M-21.5.A.24 & 25 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR 41.7 /45.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

2. The plant is at 100% power.

An event occurs resulting in the following indications:

RCS pressure is 1500 psig and LOWERING

RCS temperature is 545 oF and stable

Pressurizer level is 78% and rapidly RISING

CNMT pressure is 3.3 psig and slowly RISING

All Reactor Coolant Pumps are running Which of the below listed valves will have automatically CLOSED as result of these conditions?

1) Boron Recirc To BIT lsol Vlv [TV-1SI-884C]

2) CRDM Shroud Clg Coils CCR Inlet CNMT lsol Vlv [TV-1CC-111A2]

3) LTON Rad Monitor Trip lsol Vlv [TCV-1 CH-201]

4) CNMT Activ Monitor Disch CNMT lsol Vlv [TV-1CV-102-1]

5) VCT Outlet to Chg Pumps Suet Vlv [MOV-1 CH-115E]

6) 1A Chg PP Mini Flow lsol Vlv [MOV-1CH-275A]

A. 1, 2, & 3 only B. 1, 4, & 5 only C. 2, 3, & 6 only D. 4, 5, & 6 only Answer: B Explanation/Justification:

A. Incorrect. Item 1 is correct. Item 2 is wrong since CIB has not actuated. Item 3 is wrong, this valve will close on high inlet temperature.

B. Correct. Conditions in the stem indicate a PRZR vapor space accident in progress with RCS pressure low enough for Sl actuation which will cause CNMT isolation phase A. CNMT pressure is not high enough for CNMT isolation phase B.

C. Incorrect. Item 2 is wrong since CIB has not actuated. Item 3 is wrong, this valve will close on high inlet temperature. Item 6 is wrong, the valve used to close on an Sl signal but now these valves are de-energized open for appendix R considerations.

D. Incorrect. Items 4 and 5 are correct. Item 6 is wrong, the valve used to close on an Sl signal but now these valves are de-energized open for appendix R considerations.

Sys# System Category KA Statement 000008 Pressurizer (PZR) Vapor Space Accident AK2 Knowledge of the interrelations between the Pressurizer Vapor Space Valves (Relief Valve Stuck Open)/3 Accident and the following:

KIA# AK2.01 KIA Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

10M-12.1.D pages 3 & 4, 10M-7.1.D pages 14 &15, 10M-11.1.D page 5 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR41.7/45.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

3. The plant is at 100% power.

A Large Break LOCA occurs

EOP Procedure E-1, Loss Of Reactor Or Secondary Coolant has been entered

Step 15 of E-1, Check If Diesel Generators Should Be Stopped is being performed What action will be taken for the Emergency Diesel Generators (EDGs) and what is the bases for this action?

The crew will stop the EDGs and enable ____ (1 )_ __

The EDGs are secured because - - -(2)- - -

A. (1) a Manual start ONLY (2) the EDGs should not be run extensively unless they are carrying load B. (1) a Manual start ONLY (2) the load sequencer must be allowed to reset prior to reclosing the output breaker C. (1) both a Manual or Automatic start (2) the EDGs should not be run extensively unless they are carrying load D. (1) both a Manual or Automatic start (2) the load sequencer must be allowed to reset prior to reclosing the output breaker Answer: C Explanation/Justification:

A. Incorrect. EDGs are secured and then set up for a Manual or Automatic start. The bases is correct, unloaded diesels are to be secured per manufacturer direction.

B. Incorrect. EDGs are secured and then set up for a Manual or Automatic start. The bases is not correct for unloaded diesels, it is a valid concern for the EDG sequencers. The load sequencer uses electromechanical timers that do not reset instantly but must time-out their cycle.

C. Correct. EDGs are secured and then set up for a Manual or Automatic start. unloaded EDGs are not to be run for an extended period of time per manufacturer recommendation and the EOP step bases.

D. Incorrect. EDGs are secured and then set up for a Manual or Automatic start, the bases is not related to the sequencer operation.

Sys # System Category KA Statement 000011 Large Break EK3 Knowledge of the reasons for the following responses as the Maintaining D/Gs available to provide standby LOCA/3 apply to the Large Br~k LOCA: power KIA# EK3.09 KIA Importance 4.2 Exam Level RO References provided to Candidate Technical

References:

10M-52B.4.E-1 step 15 pg 65 None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.5 I 41.10 I 45.6 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

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

Which of the below listed alarms would indicate a complete failure (open) of the #3 seal on the 1C RCP [1RC-P-1C]?

(Assume #1 and #2 seals are intact)

A. A3-79, Reactor Cool Pump Seal Leakoff Flow Low B. A3-87, Reactor Cool Pump Seal Leakoff Flow High C. A3-117, Reactor Cool Pump 1C Seal Vent Pot Level High D. A3-119, Reactor Cool Pump 1C Seal Vent Pot Level Low Answer: D Explanation/Justification:

A. Incorrect. This would alarm if#2 seal was failed.

B. Incorrect. This would alarm if #1 seal was failed.

C. Incorrect. This would alarm if #1 OR #2 seal was failed.

D. Correct. Seal Vent Pot low level is an indication that flow that normally goes to the Vent Pot is bypassing the pot and leaking through #3 seal.

Sys# System Category KA Statement 000015/ Reactor Coolant Pump AK2 Knowledge of the interrelations between the Reactor Coolant Pump RCP seals 000017 (RCP) Malfunctions/4 Malfunctions (Loss of RC Flow) and the following:

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

References:

1OM-6.4.ABL page 3 probable cause No.2 & page 4 151 two notes Question Source: New Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR 41.7/45.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

5. The plant is at 25% power with all systems in normal alignment for this power level.

The following conditions exist:

PRZR level 27% and lowering

RCS Pressure is slowly lowering

Charging Pump Flow [FI-1CH-122A] is indicating zero (0) gpm

'A' Charging Pump, [1CH-P-1A] discharge pressure [PI-1CH-121] indicates 2600 psig

No plant Radiation Monitor alarms exist

Seal Injection flow has increased from 6 gpm to 9.3 gpm on each RCP Based on the above conditions, which procedure would be REQUIRED to be entered?

A. AOP 1.6.7, Excessive Primary Plant Leakage B. AOP 1.7.1, Loss Of Charging Or Letdown C. AOP 1.6.4, Steam Generator Tube Leakage D. AOP 1.6.8, Abnormal RCP Operation Answer: 8 Explanation/Justification:

A. Incorrect. Plausible since PRZR level and RCS pressure are lowering. The candidate could diagnose a RCS leak, but with no radiation alarms, system leakage does not exist. Inventory is lowering due to letdown flow with no charging.

B. Correct. Initial conditions met for entry into loss of charging AOP, the low charging flow is an entry condition.

C. Incorrect. Plausible distracter with pressure and level lowering, but no radiation alarms exist on the N-16 or Air Ejector radiation monitors.

D. Incorrect. Plausible if RCP seal injection increase is diagnosed as RCP Malfunction.

Sys# System Category KA Statement 000022 Loss of Reactor Coolant Makeup/2 Generic Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

KIA# 2.4.47 KIA Importance 4.2 Exam Level RO

References provided to Candidate None Technical

References:

1OM-53C.4.1.7.1 Entry Conditions Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.12)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

6. The plant is in Mode 5 with RHR loop "A" in service and the following conditions present:

Pressurizer level is 25% and stable

RCS temperature is 190 oF and stable A large pipe break occurs at the discharge of the Residual Heat Removal 1A Pump [1 RH-P-1A]

and the crew enters AOP 1.1 0.1, Loss of Residual Heat Removal Capability.

Which of the below listed annunciators will ALARM as a result of this failure?

A. A1-42, Containment Instrument Pit Level High B. A1-49, Containment Sump Level High C. A3-113, Primary Drains Transfer Tank 1 Level High-Low D. A11-28, Safeguards Area Sump Level High Answer: B Explanation/Justification:

A. Incorrect. Plausible since the water from this break will spill onto the CNMT floor however this sump is located outside the CNMT wall and is intended to identify leaks in the CNMT mat.

B. Correct. The water from this leak will spill onto the CNMT floor and make its way to this sump via CNMT floor drains. In order to answer the question, the candidate will need to have knowledge of the inputs to each of choices and be familiar enough with the names and locations of each choice presented.

C. Incorrect. Plausible since the RHR suction and discharge valves have valve stem leakoff that discharges to this tank. If the leak was a packing leak from these valves, this tank would receive the effluent.

D. Incorrect. Plausible since the CNMT sump overflows to the recirc spray sump which is the ECCS sump. However, the safeguards area sump is located outside CNMT. Additionally, the recirc spray sump level indicator in the CR is labeled Containment sump [LI-RS-151A(B)].

Sys# System Category KA Statement 000025 Loss of Residual Heat AK2. Knowledge of the interrelations between the Loss of Reactor building sump Removal System (RHRS)/4 Residual Heat Removal System and the following:

KJA# AK2.05 KJA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

OM Fig. 9-1 grid F-7 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.7 1 45. 7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

7. Initial plant conditions:

The plant is at 100% power

Master Pressurizer Pressure Controller [PC-1 RC-444J] is in AUTO with a 45% demand signal

Pressurizer 28 and 2E heaters in the AUTO ON (Red Target) position

RCS pressure is stable at 2235 psig

Turbine 1st Stage Press Selector Switch is in the 447 position Then, 1B First Stage Pressure, [PT-1 MS-447] fails to 600 psig.

Assume no operator action is taken

The Master Pressurizer Pressure Controller remains at 45% demand What will be the status of the Pressurizer heaters and Spray valve [PCV-1 RC-455A]?

A. All PRZR Heaters De-energized PRZR Spray valve PCV-1 RC-455A Throttled B. All PRZR Heaters Energized PRZR Spray valve PCV-1 RC-455A Closed C. Only PRZR Heaters 28 and 2E will be De-energized PRZR Spray valve PCV-1 RC-455A is Closed D. Only PRZR Heaters 28 and 2E will be Energized PRZR Spray valve PCV-1 RC-455A is Throttled Answer: D Explanation/Justification:

A. Incorrect. All heaters would be energized if the MPC was in automatic. Spray valve 455A will be throttled with the MPC in manual.

B. Incorrect. All heaters would be energized if the MPC was in automatic. Spray valve 455A will be throttled with the MPC in manual.

C. Incorrect. With the MPC in manual at 45% demand and only 28 & 2E heaters in the on position, they will remain energized and Spray valve 455A will remain throttled.

D. Correct. Must diagnose that the Master Pressure Controller (MPC) output is failed and the heaters and sprays will not change status. With the MPC in manual at 45% demand and only 28 & 2E heaters in the on position, they will remain energized and Spray valve 455A will remain throttled. The failure of PT 447 will generate an in signal for the control rods, rod motion will stop after the temperature input to Rod control responds. RCS temperature and pressure will lower which would energize the PRZR Heaters and close the Spray valve if the system functioned as designed. With the MPC failed as is, the heaters and spray will not respond as designed. With the MPC demand fixed at 45% the SCR controlled heaters are effectively controlled in manual.

Sys # System Category KA Statement 000027 Pressurizer Pressure Control AA 1 Ability to operate and I or monitor the following as they apply to the SCR-controlled heaters System (PZR PCS) Malfunctionl3 Pressurizer Pressure Control Malfunctions: in manual mode KJA# AA 1.02 KJA Importance 3.1

Exam Level RO References provided to Candidate None Technical

References:

10M-6.4.1F, Att. 2 pg 23 10M-6.1.Dpg 10& 11 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

8. Given the following plant conditions:

The Unit has been operating at 100% power for 345 days

A VALID high pressurizer pressure reactor trip signal is received and the reactor DOES NOT automatically trip, and it CANNOT be tripped manually from the control room

The control room operators are performing the actions of FR-S.1, Response to Nuclear Power Generation - ATWS For these conditions, what is the basis for tripping the turbine?

The Turbine t r i p - - - - - - - - - - - - - - -

A. removes a large source of positive reactivity addition B. prevents the main feed pumps from tripping on low suction pressure C. provides an additional reactor trip signal to the reactor protection system D. maintains the pressurizer pressure relief system within its relief capability Answer: A Explanation/Justification:

A. Correct. lAW the bases for step 1 and 5 of FR-5.1. The turbine removes a potential RCS cooldown which would add positive reactivity from the negative MTC. The candidate will need to understand the fundamentals of a negative MTC in order to arrive at the correct answer.

B. Incorrect. Tripping the turbine should improve the feed pump suction pressure. However, this is not the basis for tripping the turbine during an ATWS event. Tripping the turbine also conserves SG water inventory. In the event of a loss of feed induced ATWS conserving water inventory is a primary purpose for tripping the turbine. The candidate may link the loss of feed to the low suction pressure trip on the main feed pumps.

C. Incorrect. The Turbine trip will send an additional Rx trip signal to RPS. However, this is not the basis for tripping the turbine during an ATWS event.

D. Incorrect. For the various analyzed A TWS events, RCS pressure does rise, and the pressure relief system will function to keep RCS pressure within acceptable limits. However, this is not the basis for tripping the turbine during an ATWS event. One of the design criteria for the pressurizer is to keep it operable for a variety of analyzed events. The candidate may believe that this is one of the events that challenges these design criteria and tripping of the turbine is necessary to keep the pressurizer operable.

Sys# System Category KA Statement 000029 Anticipated Transient Without EK1 Knowledge of the operational implications of the Definition of negative temperature coefficient Scram (ATWS)I1 following concepts as they apply to the ATWS: as applied to large PWR coolant systems KIA# EK1.05 KIA Importance 2.8 Exam Level RO References provided to Candidate Technical

References:

1OM-53B.4.FR-S.1, page 64 basis 151 sentence None Question Source: Bank- Vision # 82007 Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

9. The crew is responding to three faulted SGs lAW ECA-2.1, Uncontrolled Depressurization Of All Steam Generators.

The following conditions exist:

RCS cooldown rate is 150 °F/HR

All SG levels are between 30-35% WR

CNMT pressure is 17 psig and lowering The feedwater flow rate REQUIRED lAW ECA-2.1 is _ _ (1 )_ _ , and the basis for this flow rate is (2) ?

A. 1) 50 gpm to each SG

2) minimize thermal shock to SG components B. 1) 50 gpm to each SG

2) to prevent the need to transition to FR-H.1, Response To Loss Of Secondary Heat Sink C. 1) 100 gpm to each SG

2) minimize thermal shock to SG components D. 1) 100 gpm to each SG

2) to prevent the need to transition to FR-H.1, Response To Loss Of Secondary Heat Sink Answer: A Explanation/Justification:

A. Correct. 50 gpm per SG is required to minimize thermal shock.

B. Incorrect. 50 gpm per SG is correct. FR-H.1 basis is incorrect. Plausible distracter if candidate does not know entry conditions into FR-H.1, or candidate thinks the 50 gpm is only required to one SG instead of all three.

C. Incorrect. Plausible distracter with 100 gpm to one SG being a requirement for FR-H.1 ifWR levels are <14%. Correct basis for ECA-2.1 flowrate.

D. Incorrect. Plausible distracter with 100 gpm to one SG being a requirement for FR-H.1 ifWR levels are <14%. FR-H.1 basis is incorrect.

Plausible distracter if candidate does not know entry conditions into FR-H.1, or candidate thinks the 100 gpm is only required to one SG instead of all three.

Sys # System Category KA Statement 000040 Steam Line AK1 Knowledge of the operational implications of the following Effects of feedwater introduction on dry S/G Rupture/4 concepts as they apply to Steam Line Rupture:

KIA# AK1.07 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical

References:

1OM-53A.1 ECA-2.1 Step 3 pg 3 1OM-53B.1.ECA-2.1 Step 3 Caution pg 20 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8/41.10 /45.3)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

10. The plant is in Mode 3, ready for Reactor Startup. All systems in normal alignment for this Mode.

Initial conditions:

RCS Tavg is 547 °F

All RCPs running

1FW-P-1 B is running

1FW-P-1A is secured

Bypass Feedwater Flow Control valves are in automatic Bus 1D Supply ACB 341 B [ACB-1 D6] inadvertently opens. No Operator Actions are taken. All systems respond as designed.

(1) Which Auxiliary Feedwater (AFW) Pump(s) will be operating after the event?

(2) How is cooling supplied to the AFW pump lube oil coolers?

Cooling water to the Lube Oil Cooler is supplied (2) _ __

A. (1) Both Motor Driven AFW pumps ONLY.

(2) continuously from the Component Cooling Water (CCR) system B. (1) Both Motor Driven AFW pumps ONLY.

(2) after the AFW pump starts and establishes flow through the cooler C. (1) Turbine Driven AFW pump ONLY.

(2) continuously from the Component Cooling Water (CCR) system D. (1) Turbine Driven AFW pump ONLY.

(2) after the AFW pump starts and establishes flow through the cooler Answer: B Explanation/Justification:

A. Incorrect. Only the MDAFW pumps will be running due to small amount of steam load on SGs. The SG levels will remain >19.6%, therefore the TDAFW pump will not start on 2/3 Low Low SG level on 113 SG. CCR is not used to cool the lube oil cooler B. Correct. MDAFW pumps will start on the Auto Trip of all running MFW Pumps. Cooling is proved by a minimum flow of 20 gpm off of the discharge piping of the AFW pumps. This flowpath is NSA locked-open back to the PPDWS tank .The 1-2 EDG will start on undervoltage and supply the 'B' MDAFW pump C. Incorrect. Both MDAFW pumps will start, but the TDAFW pump will not since no low level exists on SG. CCR is not used to cool the lube oil cooler.

D. Incorrect. The SG levels will remain >19.6%, therefore the TDAFW pump will not start on 213 Low Low SG level on 113 SG. Correct cooling to the lube oil cooler.

Sys # System Category KA Statement 000054 Loss of Main Feedwater AA 1 Ability to operate and I or monitor the following as AFW auxiliaries, including oil cooling water supply (MFW)I4 they apply to the Loss of Main Feedwater (MFW):

KIA# AA1.03 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

1OM-24.1.0 pgs 2 & 4, RM-0424-002 (03 & 06)

Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

11. The plant is at 100% power.

A Reactor trip coincident with a loss of offsite power occurs

BOTH Emergency Diesel Generators fail to start You have been dispatched to locally start Emergency Diesel Generator (EDG #1) EE-EG-1, by performing EOP Attachment 2-E, Local Action To Restore AC Power.

All start failure alarms have been CLEARED.

lAW EOP Attachment 2-E, HOW will EDG #1 be started?

(Place the below listed steps in the correct sequence for starting EDG #1)

1) Depress the FUEL PRIME pushbutton

2) Place selector switch to LOCAL START position

3) Adjust Governor "SPEED SETTING" to an operating speed of 900 RPM

4) Depress the ENGINE START pushbutton A. 1, 2, 3, 4 B. 1, 2, 4, 3 C. 2, 1, 3, 4 D. 2, 1, 4, 3 Answer: D Explanation/Justification:

A. Incorrect. Local start is selected first, Speed setting is adjusted after the EDG is started.

B. Incorrect. Local start is selected first.

C. Incorrect. Speed setting is adjusted after the EDG is started.

D. Correct. In accordance with EOP Att. 2-E, steps 5 thru 8.

Sys # System Category KA Statement 000056 Loss of Offsite Generic Knowledge of RO tasks performed outside the main control room during an emergency and the Powerl6 resultant operational effects.

KIA# 2.4.34 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

1OM-53A.1.2-E steps 5 thru 8 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

12. The plant was operating at 100% with all systems in NSA EXCEPT PRZR Level Channel Selector Switch is selected to POS 3 461/460.

A loss of Vital Bus I occurs

lAW AOP-38.1A, Loss of Vital Bus I, the crew manually trips the Reactor and enters E-0, Reactor Trip and Safety Injection Why is the Reactor manually tripped on a Loss of Vital Bus I?

In anticipation of an automatic trip w h e n - - - - - - - - - - - - - - - -

A. SG Water level reaches >P-14 setpoint B. SG Water level reaches the low level setpoint C. Pressurizer Water level reaches the high level setpoint D. Pressurizer Pressure reaches the low pressure setpoint Answer: 8 Explanation/Justification:

A. Incorrect. Candidate must realize that with the SG Feedwater pump recirc valve failing open (FCV-1 FW-150B), that there will not be sufficient water to supply the SGs at 100% power, therefore reaching >P-14 is not possible.

B. Correct. When the SG Feedwater pump recirc valve fails open (FCV-*1 FW-150B}, there is not sufficient water available to feed the SGs at 100%.

This is reinforced by the compensatory action in AOP-38.1A Att. 1, that states secure main feed pumps and feed SGs with AFW.

C. Incorrect. PRZR level transmitter LT-459 will lose power, but with LT-461 in service there will be no effect on PRZR level in this case. Plausible distracter because candidate must evaluate the effect on PRZR level control system, and how the Pressurizer level will respond.

D. Incorrect. Candidate must realize that pressure control transmitters are not effected by this loss. PT-444 & 445 are powered from VB 3 & 4.

Sys # System Category KA Statement 000057 Loss of Vital AC Electrical AK3 Knowledge of the reasons for the following responses Actions contained in EOP for loss of vital ac Instrument Bus/6 as they apply to the Loss of Vital AC Instrument Bus: electrical instrument bus KIA# AK3.01 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical

References:

1OM-53C.4.1.38.1A pg 18 step 2 compensatory action Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.5/41.10 /45.6/45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

13. The plant is at 100% power with all systems in NSA EXCEPT for the following:

Battery Breaker #1-2 is OPEN for maintenance

Vital Bus #2 Manual Bypass switch is in the ALTERNATE SOURCE TO LOAD position

Battery Charger #2A is in service Battery Charger #2A AC Input Breaker [B301] TRIPS open.

What impact will this have on the control room indications for #2 DC Bus Volts and N42 Power range indication?

Control room indication for #2 DC Bus Volts will _ _ _ (1 )_ _ _ AND N42 Power range indication will (2) _ __

A. (1) remain as is (2) remain as is B. (1) remain as is (2) drop to ZERO C. (1) drop to ZERO (2) remain as is D. (1) drop to ZERO (2) drop to ZERO Answer: C Explanation/Justification:

A. Incorrect. Part 1 is wrong. The dual battery chargers will not automatically swap over (like the vital bus will swap over) when voltage drops. The Part 2 is correct. Having the manual bypass switch in alternate source to load position will keep the vital bus energized.

B. Incorrect. Part 1 is wrong. The dual battery chargers will not automatically swap over (like the vital bus will swap over) when voltage drops. The Part 2 is wrong. Having the manual bypass switch in alternate source to load position will keep the vital bus energized.

C. Correct. The dual battery chargers will not automatically swap over (like the vital bus will swap over) when voltage drops. Therefore the DC bus will be de-energized. Having the manual bypass switch in alternate source to load position will keep the vital bus energized.

D. Incorrect. Part 1 is correct. Part 2 is wrong. Having the manual bypass switch in alternate source to load position will keep the vital bus energized.

Sys# System Category KA Statement 000058 Loss of DC AK1 Knowledge of the operational implications of the following Battery charger equipment and instrumentation Powerl6 concepts as they apply to Loss of DC Power:

KIA# AK1.01 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

LP 3SQS-38.1 Unit 1 ppt slide 28, 1OM-39.4.AAL page 2 of 5 step 5; LP 3SQS-39.1 Unit 1 ppt slide 7 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8 I 41.10 I 45.3)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

14. The plant is at 100% power.

Breaker 4 KV Bus 1AE to 1A [ACB 1E7] has spuriously tripped open

Emergency Diesel Generator (EDG #1) EE-EG-1 has been manually started and has energized the 1AE emergency bus

Procedure AOP 1.36.2, Loss of 4 KV Emergency Bus has been entered

Emerg Gen 1 Wattmeter is 730 KW and stable

Reactor Plant River Water [1WR-P-1A] Pump bright WHITE indicating light is lit

Reactor Plant River Water [1WR-P-1 B] Pump WHITE indicating light is lit

CCR Heat Ex River Water Press A Header is 0 psig

CCR Heat Ex River Water Press B Header is 0 psig

NO operator actions have been taken lAW AOP 1.36.2, Loss of 4 KV Emergency Bus Attachment 3, Local Action to Restore AC Power- Train A, what is the MAXIMUM ALLOWABLE time Emergency Diesel Generator EE-EG-1 can continue to operate under these conditions?

A 1 Minute B. 5 Minutes C. 15 Minutes D. 30 Minutes Answer: 8 Explanation/Justification:

A. Incorrect. This is the time limit for the EDG Sequencer to start loads.

B. Correct. Must determine that RPRW has been lost to the EDG and then identify the maximum run time allowed by procedure. The AOP would require action to start pumps or secure the EDG, however no operator action is credited in the question. The AOP limits the loaded or unloaded run time to 5 minutes without cooling water.

C. Incorrect. 15 minutes was selected for question symmetry per NRC request.

D. Incorrect. 30 minutes is the 40 CFR 63 RICE requirement time from EDG start to synchronization.

Sys# System Category KA Statement 000062 Loss of Nuclear AA2 Ability to determine and interpret the following The length of time after the loss of SWS flow to a Service Water/4 as they apply to the Loss of Nuclear Service Water: component before that component may be damaged KIA# AA2.06 KIA Importance 2.8* Exam Level RO References provided to Candidate None Technical

References:

1OM-53C.4.1.36.2 Att 3 pg 16 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

15. The plant is at 100% power.

Station Air Compressor [1 SA-C-1 B] is on clearance for maintenance

Station Air Compressor [1SA-C-1A] is running

[PI-11A-106A], CNMT Instrument Air (lA) Header Pressure is 92 psig and slowly DROPPING

[PI-11A-106], Station Instrument Air Header Pressure is 98 psig and slowly DROPPING

[PI-1SA-101], Station Air Header Pressure is 98 psig and slowly DROPPING Subsequently, with NO Operator action, the following occurs:

[PI-11A-106A], CNMT Instrument Air (lA) Header Pressure lowers to 90 psig AND begins to slowly RISE

[PI-11A-106], Station Instrument Air Header Pressure lowers to 94 psig AND begins to slowly RISE

[PI-1SA-101], Station Air Header Pressure lowers to 94 psig AND begins to slowly RISE (Assume all systems function as designed)

(1) Where was the air leak?

(2) How was the leak isolated?

A. (1) Station Air Leak Outside CNMT.

(2) Station Air Header Isolation Valve [TV-1SA-105] automatically CLOSED.

B. (1) Station Air Leak Outside CNMT.

(2) lnst Air To CNMT lnst Air lsol [TV-11A-400] automatically CLOSED.

C. (1) Instrument Air Leak Inside CNMT.

(2) Station Air Header Isolation Valve [TV-1SA-105] automatically CLOSED.

D. (1) Instrument Air Leak Inside CNMT.

(2) lnst Air To CNMT lnst Air lsol [TV-11A-400] automatically CLOSED.

Answer: A Explanation/Justification:

A. Correct. Conditions stated in the stem are entry conditions for loss of station instrument air. Candidate will need to evaluate conditions given and understand the station air, instrument air and CNMT instrument air systems. Candidate will also need to have knowledge of automatic actions of TV-105 and 400. 1SA-105 auto closes at 95 psig to isolate the leak.

B. Incorrect. Location is correct. Wrong automatic isolation. TV-11A-400 will auto close on CIB signal or loss of air. The loss of air given in the stem is not severe enough to cause the valve to close.

C. Incorrect. Wrong location. Correct automatic isolation. If TV-1SA-105 closes and the leak is inside CNMTthen [PI-11A-106] and [PI-1SA-101]

would rise as stated in the stem. However, [PI-11A-1 06A] would continue to drop.

D. Incorrect. Wrong location. Wrong automatic isolation. TV-11A-400 will auto close on CIB signal or loss of air. The loss of air given in the stem is not severe enough to cause the valve to close.

Sys # System Category KA Statement 000065 Loss of Instrument AA2 Ability to determine and interpret the following as they apply to the Loss of Location and isolation of leaks Air/8 Instrument Air:

KIA# AA2.03 KIA Importance 2.6 Exam Level RO References provided to Candidate Technical

References:

OM Figs. 34-1,2 ,3, 4, & 6 None 10M-34.2.8 Page 3 of3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

16. The plant is at 100% power with all systems in NSA EXCEPT Emergency Diesel Generator (EDG #2) EE-EG-2 is OOS for maintenance on the engine driven fuel oil pump.

The DLC System Operations Control Center notifies the control room of a significant threat to grid stability

Grid voltage swings are verified in the control room

The crew enters AOP 1/2.35.1, Degraded Grid lAW AOP 1/2.35.1, Degraded Grid, what if any, conservative actions will be preemptively taken with respect to Emergency Diesel Generator (EDG #1) EE-EG-1?

A. NO actions required. Allow EDG #1 to Auto start if necessary.

B. Manually start EDG #1, flash the field AND manually CLOSE [ACB 1E9] Emerg Gen1 Circuit Breaker.

C. Manually start EDG #1, flash the field AND manually OPEN [ACB 1A10] 4KV Bus 1A to 1AE Circuit Breaker.

D. Manually start EDG #1, flash the field, synchronize the EDG to the 1AE bus, assume load, AND then manually OPEN [ACB 1E7] 4KV Bus 1AE to 1A Circuit Breaker.

Answer: C Explanation/Justification:

A. Incorrect. These are the correct actions for other AOPs that could possibly threaten offsite electrical power such as a tornado or high winds.

B. Incorrect. Plausible since it makes sense to get the EDG ready for service with manual action. However, this procedure will specifically direct getting the EDG off of any parallel operations and then performing choice C.

C. Correct. This is a major action step of this procedure to divorce safety related equipment from offsite power and establish the EDG as power to safety related equipment.

D. Incorrect. This will accomplish the task of divorcing safety related equipment from offsite power and establishing the EDG as the power supply to safety related equipment. However, this is not how the procedure directs this to be accomplished and violates the OM 36 P&L to not parallel EDGs with offsite power when anticipating a loss of offsite power.

Sys# System Category KA Statement 000077 Generator Voltage and Electric AA2 Ability to determine and interpret the following as they apply to Operational status of Grid Disturbances/6 Generator Voltage and Electric Grid Disturbances: emergency diesel generators KIA# AA2.09 KIA Importance 3.9 Exam Level RO References provided to Candidate Technical

References:

1/20M-53C.4A.35.1 step 15 RNO None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.5 and 43.5/45.5, 45.7, and 45.8)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

17. Following a reactor trip and safety injection, the crew is performing the diagnostic steps of E-0, Reactor Trip Or Safety Injection.

The following conditions exist:

All SG pressures are 1000 psig and stable

All SG NR levels are 25% and rising

Total AFW is 900 gpm and stable

RCS pressure is 1000 psig

RCS temperature is 545 oF and stable

[RM-1VS-110], CNMT/SLCRS Exhaust Monitor SPING 4 is in alarm

[RM-1VS-112], CNMT/SLCRS Exhaust Monitor SA 9/10 is in alarm

NO other Radiation Monitors are in alarm

Containment pressure is minus (-) 1.2 psig Which of the following procedures contains the procedural guidance to mitigate this event?

A. ES-0.2, Natural Circulation Cooldown B. ES-1.2, Post-LOCA Cooldown And Depressurization C. ES-1.1, Sl Termination D. ECA-1.2, LOCA Outside Containment Answer: D Explanation/Justification:

A. Incorrect. Transition from ES-0.1 or ES-1.1 would be required, but the stem indicates an Sl should have occurred, causing E-0 usage past the transition to ES-0.1 and ES-1.1 criteria is not met. Plausible since RCS and SG pressures are 1000 psig and RCP Trip criteria has been met iaw E-0 LHP. Without RCPs, Natural Circulation C/D is a plausible distractor.

B. Incorrect. E-1 would need to be implemented first before any of the mitigating strategies of ES-1.2 will be effective, and neither E-1 nor ES-1.2 contain the appropriate mitigating steps for a LOCA outside containment.

C. Incorrect. RCS pressure is stable, a heat sink is available, BUT no subcooling exists. Therefore, ES-1.1 is not appropriate.

D. Correct. This is a borderline SRO question since it seems as though the question is asking what procedure entry is required. Which JAW guidance provided in the white paper, would be SRO knowledge. However, in order to answer the question, the candidate will need to have E-0 diagnostic knowledge to diagnosis the event in progress (JAW E-0) and then only have overall purpose/mitigation knowledge of each of the procedures presented in the choices. This overall purpose/mitigation knowledge is RO knowledge JAW the SRO only white paper guidance.

Sys # System Category KA Statement W/E04 LOCA Outside Generic Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate Containment/3 control room reference material.

KIA# 2.4.47 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical

References:

1OM-53A.1.E-O step 22 Question Source: Bank- 2005 BVPS-1 NRC Exam Q#54 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.12)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

18. 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 20, Prepare to establish Minimum Sl Flow to Remove Decay Heat is being performed Refer to attached ECA-1.1 step 20, and Attachment 5-G, Minimum Sl Flow Versus Time After Trip What is the minimum REQUIRED Safety Injection flow and what injection flowpath will be used?

The minimum REQUIRED Sl 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:

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 Sl 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 200 GPM, the HHSI flowpath is used. Step 20, Prepare To Establish Minimum Sl 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 Sl flow to remove decay heat, injection is not always via the Sl flowpath.

Sys # System Category KA Statement WIE11 Loss of Emergency EA 1 Ability to operate and I or monitor the following as they *Desired operating results during abnormal Coolant Recirculationl4 apply to the (Loss of Emergency Coolant Recirculation) and emergency situations.

KIA# EA1.3 KIA Importance 3.7 Exam Level RO References provided to Candidate 10M-53A.1.5-G pg 1 & 2 Technical

References:

10M-53A.1.ECA-1.1 pg 14 and 15 and Att 5-G 10M-53A.1.ECA-1.1 pg 14 and 15 Question Source: New Question Cognitive Level: Low- Fundamental 10 CFR Part 55 Content: (CFR: 41.7 I 45.51 45.6)

Objective:

BVPS-EOP 1OM-53A.1.ECA-1.1 (ISS2)

Number Title ECA-1.1 Loss Of Emergency Coolant Recirculation Issue 2 Revision 1 STEP II ACTION/EXPECTED RESPONSE tI RESPONSE NOT OBTAINED 20 Pregare To Establish Minimum SI Flow To Remove Decay Heat

a. Verify suction of charging pumps aligned to RWST:

1) Verify [MOV-lCH-1158 or D] -

OPEN

2) Verify [MOV-1CH-115C or E] -

CLOSED

b. Stop any running LHSI Pumps AND place in AUTO.

c. Determine minimum SI flow using Attachment 5-G.

d. Determine SI flow path:

For minimum required SI flow less than 200 GPM, GO TO Step 20.e

-OR-For minimum required SI flow greater than 200 GPM, GO TO Step 21.

e. Establish normal charging flow: e. IF flow can NOT remain less than 200 GPM, THEN realign SI

1) Close [FCV-1CH-122] flow to BIT AND GO TO Step 21.

2) Open [MOV-lCH-310]

3) Open [MOV-lCH-289]

4) Close [MOV-1SI-867A and 8]

5) Close [MOV-1SI-867C and D]

6) Adjust [FCV-lCH-122] to maintain required flow according to Attachment 5-G

7) GO TO Step 28 1ECA11 314/2014 14 of29

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

19. The plant is at 100% power.

The following Annunciators are received:

A4-11, Pressurizer Control Press Low

A4-12, Pressurizer Control Low Press Deviation

A4-59, NIS Power Range Low Setpoint Flux Deviation or Auto Defeat

A4-67, NIS Power Range High Setpoint Flux Deviation or Auto Defeat

A4-68, NIS Power Range High Comparator Deviation

A4-126, Rod Bottom Rod Drop

A4-76, Computer Alarm Rod Deviation I SEQ NIS Power Range Tilt For these plant conditions, what is the effect on QPTR and how is the Quadrant Power Tilt Ratio (QPTR) determined?

The Upper Detector QPTR calculation will be effected ___(1) ___ the Lower Detector QPTR calculation.

AND QPTR will be determined by comparing the (2)___ Upper or Lower detector output to the AVERAGE of the Upper or Lower detector output.

A. ( 1) more than (2) minimum B. (1) more than (2) maximum C. (1) the same as (2) minimum D. (1) the same as (2) maximum Answer: D Explanation/Justification:

A. Incorrect. The rod effects the upper and lower core the same, the QPTR is the max to average ratio not minimum.

B. Incorrect. The rod effects the upper and lower core the same, the QPTR is the max to average ratio.

C. Incorrect. The rod effects the upper and lower core the same, however the QPTR is the max to average ratio, not minimum.

D. Correct. The alarms indicate that a Control Rod is dropped and not just misaligned. The effect on QPTR will be the same since the Rod is fully inserted into the core. The definition of QPTR states the Max upper or lower detector output is compared to the upper or lower average. The rod effects the upper and lower the same, the QPTR is the max upper or lower to the average of that section of the core.

Sys# System Category KA Statement 000003 Dropped AK1 Knowledge of the operational implications of the following concepts as they Definitions of core quadrant power tilt Control Rod/1 apply to Dropped Control Rod:

KIA# AK1.10 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

Tech Spec Definition QPTR Pg 1.1-5, Simulator NIS response to dropped rod Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR 41.8/41.10 /45.3)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

20. A reactor startup is in progress with all systems aligned for this plant condition.

Control Room indications are as follows:

Control Bank D Group 1 and 2 step counters indicate 25 steps

Control Bank D Group 2 Control Rod K1 0 position indication is 0 steps

All other Control Bank D individual rod indications are 25 steps

Annunciator A4-105 Rod Control System Urgent Alarm is NOT lit In addition to the Rod Bottom Light, which of the following alarms (if any) would be received in the control room?

1) Annunciator A4-116 Rod Control Bank D Low

2) Annunciator A4-124 Rod Control Bank D Low-Low

3) Annunciator A4-126 Rod Bottom Rod Drop A. None B. 1 and 2 ONLY C. 2 and 3 ONLY D. 1, 2, and 3 Answer: A Explanation/Justification:

A. Correct. Must evaluate the indications and determine that the rod is physically stuck since there are no control system alarms and the other control bank "D" rods are aligned with the groups step counters. The dropped Rod Alarm Bypass Relay is still enabled until 35 steps withdrawn, so the Dropped Rod Annunciator will not alarm. The Bank D Low and Low-Low alarms would be indications of control bank insertion near the calculated RIL setpoints however at 0% power there is no limit. Control Banks A and B have fixed RIL setpoints. The rod is stuck based upon the indications. The Dropped Rod Annunciator, which alarms when a control rod reaches 20 steps, is bypassed by the Rod Bottom Bypass Relay until the bank reaches 35 steps, the bank is currently at 25 steps. This matches the KIA by assessing the candidates knowledge of initial rod withdraw, and ability to diagnose the stuck rod. With this knowledge, understands the function of the Rod Bottom Bypass Relay, which is why a Rod Bottom Rod Drop annunciator does not alarm when <35 steps.

B. Incorrect. For Control Bank D The Control Bank Low and Low-Low alarms are driven from a calculated Rod Insertion Limit based upon power. At 0% power there is no RIL for this bank.

C. Incorrect. The Dropped Rod Annunciator is blocked until Bank D reaches 35 steps. The Bank D Low-Low alarm is not enabled by the RIL calculation at 0% power.

D. Incorrect. The Bank Low-Low alarm setpoint is calculated based upon power, at 0% there is no limit, the Low alarm setpoint is 10 steps higher than the RIL setpoint. The Dropped Rod alarm is bypassed until the bank reaches 35 steps.

Sys # System Category KA Statement 000005 Inoperable/Stuck AK2 Knowledge of the interrelations between the Inoperable I Breakers, relays, disconnects, and control room Control Rod/1 Stuck Control Rod and the following: switches KIA# AK2.02 KIA Importance 2.5 Exam Level RO References provided to Candidate None Technical

References:

10M-1.1.B pg 171 51 & 2"a paragraphs, 10M-1.2.B pg 2, 1SQS-1.4 Rev. 14, pg. 11 Question Source: New Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR 41.7/45.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

21. The plant is at 100% power.

Pressurizer Level Transmitter [LT-1 RC-460] has failed low.

Assume no operator action is taken.

What will be the response of Regenerative Heat Exchanger Charging Outlet Temperature?

Regenerative Heat Exchanger Charging Outlet Temperature [TI-1 CH-123] will

_ _(1) _ __

AND (2) What temperature limitation may be exceeded for this event?

A. (1) rise (2) Charging line penetration to the RCS piping temperature differential.

B. (1) lower (2) Charging line penetration to the RCS piping temperature differential.

C. (1) rise (2) Pressurizer normal spray line differential temperature.

D. (1) lower (2) Pressurizer normal spray line differential temperature.

Answer: B Explanation/Justification:

A. Incorrect. Letdown will isolate so preheating of the CHS will be lost, temperatures will lower, correct thermal limit concern.

B. Correct. Must evaluate the LT 460 failure and determine that letdown isolates. This will result in lowering charging temperature since the Regen Hx will not provide preheating. Letdown temp lowers due to a loss of flow. The temperature concern is for the charging line to the RCS piping, as this is a function of the Regen Hx. Simulator response over 5 minutes CHS Temp lowers from 433Fto 100F and Letdown Temp lowers from 256 oF to 244 °F. Letdown will isolate and preheating of the CHS will be lost, temperatures will lower, due to the loss of flow and ambient heat loss.

The function of the Regen Hxer is to limit the temperature thermal limits on the Charging to RCS system penetrations. The Charging line penetration thermal limit is a concern when the preheating is lost.

C. Incorrect. Temperatures will not rise. The normal spray line !:::,.Tis not affected since the normal spray line water comes from the RCS loops, this water is mixed on the system so the temperature change is minimal. The student may select this piping due to the t::.. T concerns when collapsing the PRZR bubble.

D. Incorrect. Temperatures will lower. The normal spray line t::.. Tis not affected since the normal spray line water comes from the RCS loops.

Sys # System Category KA Statement 000028 Pressurizer (PZR) Level AA 1 Ability to operate and I or monitor the following as they apply Regenerative heat exchanger and Control Malfunctionl2 to the Pressurizer Level Control Malfunctions: temperature limits KIA# AA1.04 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

10M-7.1.C pg 6, 10M-6.4.1F pg 7 Question Source: New Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

22. The plant is at 100% power.

Procedure 10M-19.4.1, "Troubleshooting Gas Leaks", is being performed due to increasing radiation levels in the Auxiliary Building. A gas leak has been identified on Primary Drains Tank No.2 [1 DG-TK-2]. The tank has been isolated, which has stopped the leak.

lAW with 10M-19.4.1, which ofthefollowing evolutions CANNOT be performed with [1DG-TK-2]

isolated?

A. Vent the Charging Pump suction piping.

B. Vent the Safety Injection Accumulators.

C. Drain the Pressurizer Relief Tank.

D. Lower the level in the Safeguards Area Sump .

Answer: A Explanation/Justification:

A. Correct. Isolation of the Primary Drains Tank Number 2 will prevent the venting of the Charging Pump Suction piping. The precaution and limitation in 1OM-19.4.1, "Troubleshooting Gas Leaks" notes that the vent path is unavailable to [1 DG-TK-2] while performing the procedure, isolation of the tank would prevent venting the charging pump suction ..

B. Incorrect. The Safety Injection Accumulators vent to either Containment or the Nitrogen supply header line via a drain valve.

C. Incorrect. The Pressurizer Relief Tank is drained to the other Primary Drains Tank NO. 1 [1DG-TK-1] not Primary Drains Tank No. 2 [1DG-TK-2].

Plausible if they confuse the drain path of the PRT and the correct drains tank.

D. Incorrect. The Safeguards Area Sump discharges to waste drain header, not to Primary Drains Tank No. 2 [1 DG-TK-2]

Sys # System Category KA Statement 000060 Accidental Gaseous AA2 Ability to determine and interpret the following as they The effects on the power plant of isolating a given Radwaste Release/9 apply to the Accidental Gaseous Radwaste: radioactive-gas leak KIA# AA2.04 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

10M-19.4.1 Rev 1 pg 2. VOND RM 09-1,09-2, 06-2 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 43.5 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

23. The plant is at 100% power.

Which of the following actions will occur if Control Room Area Radiation Monitor [RM-1 RM-218A] spuriously fails HIGH?

Assume only this channel has failed.

Control Room Air Intake Damper [1VS-D-40-1A] will (1)_ __

Control Room Air Intake Damper [1VS-D-40-1B] will (2) _ __

Control Room Air Exhaust Damper [1VS-D-40-1C] will __(3)_ __

Control Room Air Exhaust Damper [1VS-D-40-1 D] will __(4) _ __

A. 1) close

2) close

3) close

4) close B. 1) remain as is

2) close 3 remain as is

4) close C. 1) close

2) remain as is

3) close

4) remain as is D. 1) remain as is

2) remain as is

3) remain as is

4) remain as is Answer: C Explanation/Justification:

A. Incorrect. The High alarm from RM 218A will close the Train A Components, dampers 1VS-D-40-1A & 1C only.

B. Incorrect. The High alarm from RM 218A will close the Train A Components, not dampers 1VS-D-40-1 B & 1D (Train B).

C. Correct. Must evaluate the action from a single Area RM in High, and whether a single train of dampers or both trains are affected, as well as the Automatic action for the intake and/or exhaust damper. The High alarm from RM 218A will close the Train A Components, which are dampers 1VS-D-40-1A & 1C.

D. Incorrect. The High alarm from RM 218A will close the Train A Components only, One intake damper 1VS-D-40-1A and one exhaust damper 1VS-D-40-1 C receive the High signal. If student assumes one train Rad Monitor actuates both trains then this answer is plausible.

Sys# System Category KA Statement 000061 Area Radiation Monitoring (ARM) AA 1 Ability to operate and I or monitor the following as they apply to Automatic actuation System Alarms/7 the Area Radiation Monitoring (ARM)System Alarms:

KIA# AA1.01 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

10M-43.5.B.2 Pg 3, 10M-43.4.ADP pg 3, 10M-44A.1.D Pg 10 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR 41.7 I 45.5 I 45.6)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

24. A fire has been identified in the East Cable Vault, procedure 10M-56B.4.H, Safe Shutdown Following a Serious Fire in Safeguards, has been implemented.

Which of the following lists the Safe Shutdown Equipment that is located in Fire Area CV-2, East Cable Vault?

1. 480V Emergency Motor Control Center [MCC 1-E5]

2. 480V Emergency Motor Control Center [MCC1-E6]

3. 480V Emergency Motor Control Center [MCC1-E10]

4. 480V Emergency Motor Control Center [MCC1-E12]

5. Backup Indicating Panel [PNL-BIP]

6. Control Room AIC Unit [1VS-1AIC-1A]

A. 1, 3, 5 ONLY B. 2, 4, 5 ONLY C. 3, 4, 6 ONLY D. 1, 2, 6 ONLY Answer: 8 Explanation/Justification:

A. Incorrect. MCC-E5 is located in the West CV, MCC-E10 is in OF SWGR, the BIP is in the East CV.

B. Correct. MCC-E6 is located in the East CV, MCC-E12 is in East CV, the BIP is in the East CV.

C. Incorrect. MCC-E10 is located in DF Swgr, MCC-E12 is in the East CV, the CR A/C Unit is in the Control Bldg.

D. Incorrect. MCC-E5 is located in the West CV, MCC-E6 is located in the East CV, the CR A/C Unit is in the Control Bldg.

Sys # System Category KA Statement 000067 Plant fire on AA2 Ability to determine and interpret the following as they apply to the Location of vital equipment within fire zone site/9 Plant Fire on Site:

KIA# AA2.12 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

10M-56B.4.H pgs 65, 83, 167 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 43.5/ 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

25. The plant was operating at 100% power.

A Reactor Trip and Safety Injection occur due to a Loss Of Coolant Accident

Plant conditions require entry into procedure FR-C.2, Response To Degraded Core Cooling

Step 13, Depressurize All Intact SGs To 110 psig, is being performed, the cooldown rate is limited to 100 °F/Hr The following table shows the trend of Cold Leg Temperature and RCS Pressure:

TIME Cold Leg Temp RCS Pressure 10:00 330 °F 1750 psig 10:30 310 °F 1200 psig 11:00 250 OF 940 psig 11:30 230 °F 800 psig Refer to attached EOP Attachment 5-A, RCS Cooldown Limits -Technical Specifications At 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months , which of the following set of plant conditions is allowable per the procedure?

A. 140 °F, 400 psig B. 160 °F, 550 psig C. 170 °F, 450 psig D. 180 °F, 600 psig Answer: C Explanation/Justification:

A. Incorrect. The cooldown will exceed the 100 "F/ hr limit, but is in the Acceptable region of the curve.

B. Incorrect. The cooldown will not exceed the 100 "F/ hr limit, but is in the Unacceptable area of the curve, it is below the 50 "F/hr limit line.

C. Correct. Must recognize the 100 "F/hr limitations and extrapolate that information on the C/D curve. The cooldown is less than 100 "F/Hr and in the Acceptable region for this rate.

D. Incorrect. The cooldown will not exceed the 100 "F/ hr limit, but is in the Unacceptable region of the curve at the common point for the lower C/D Rates.

Sys # System Category KA Statement W/E06 Degraded Core Cooling/4 Generic Ability to interpret reference materials, such as graphs, curves, tables, etc.

KIA# 2.1.25 KIA Importance 3.9 Exam Level RO References provided to Candidate 1OM-53A.1.5-A Rev 2 Technical

References:

1OM-53A.1. FR-C.2 pg 7, 1OM-53A.1.5-A Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.12)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

26. The plant was at 100% power.

A Reactor Trip and Safety Injection occur due to a Faulted Steam Generator The following conditions exist:

The Reactor trip occurred 30 minutes ago

RCS Tavg is 310 °F

RCS Cold Leg Temperature is 280 oF

RCS Pressure is 350 psig Refer to attached EOP Attachment 5-D, PTS-Operational Limits Curve Which of the following describes the bases for the operating limits that exist for the current plant conditions?

A. Vessel integrity IS challenged because pressure exceeds the Cold Overpressure Limit B. Vessel integrity is NOT challenged, even at system operating pressure.

C. A vessel wall flaw IS assumed to grow, vessel failure is imminent.

D. A vessel wall flaw is NOT assumed to grow, operator action is required to maintain plant conditions from exceeding limits.

Answer: D Explanation/Justification:

A. Incorrect. The COPS setpoint of 395 psig is not exceeded.

B. Incorrect. The temperature must be above 300 oF for there not to be a challenge to vessel integrity, temperature is below this value.

C. Incorrect. This response is the bases if the Limit A line is exceeded. This is not the case for these conditions.

D. Correct. Plant conditions are less than 300 oF line and to the right of Limit A, if Pressure rises or Temperature lowers the limit could be exceeded.

The conditions listed are to the right of the limit A line and to the left of the 300 oF line.

Sys # System Category KA Statement WIEOB Pressurized EK3 Knowledge of the reasons for the Facility operating characteristics during transient conditions, including coolant Thermal following responses as they apply to chemistry and the effects of temperature, pressure, and reactivity changes and Shock/4 the (Pressurized Thermal Shock) operating limitations and reasons for these operating characteristics.

KIA# EK3.1 KIA Importance 3.4 Exam Level RO References provided to Candidate 1OM-53A.1.5-D Technical

References:

10M-53B.4.F-0.4 step 4 Bases pg 16 1OM-53A.1.5-D Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: (CFR: 41.5 I 41.10 I 45.6 I 45.13)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

27. The plant was at 100% power when a reactor trip occurred.

The following conditions exist:

All Main Steam Line Isolation Valves are closed

The 1A Steam Generator pressure is 1130 psig and slowly rising

Steam Generator Atmospheric Dump Valve [PCV-1MS-101A] and Residual Heat Release Valve [HCV-1MS-104] are failed closed and cannot be manually operated

Steam Generator 1B and 1C pressures are 1035 psig and stable

All Feedwater flow to the 1A SG is isolated

Narrow Range Steam Generator level in all SGs is 10%

Hot Leg Temperatures are 555 oF and slowly rising

Total AFW flow to the 1B and 1C SGs is 400 gpm

The crew enters procedure FR-H.2, Response To Steam Generator Overpressure Which of the following actions will be taken per procedure FR-H.2, Steam Generator Overpressure, to mitigate the event?

A. Manually actuate Safety Injection, open at least two PORVs.

B. Cooldown the RCS by dumping steam from the 1B and 1C Steam Generators.

C. Establish 100 gpm AFW flow to the 1A Steam Generator.

D. Stop Reactor Coolant Pump [1 RC-P-1A].

Answer: 8 Explanation/Justification:

A. Incorrect. This action is taken for a Loss of Heat Sink, however AFW flow is adequate.

B. Correct. Overpressure procedure directs reestablishing steam dump capability or if Thot exceeds 525 *F, dumping steam from the other SGs.

This action will cooldown the RCS, including the affected SG. Heat removal is via the RCS piping to the other unaffected SGs. lAW FR-H.2 for a SG Overpressure event, the unaffected SGs are used to lower Hot Leg temperatures below 525 *F.

C. Incorrect. This is the AFW established to a hot, dry Steam Generator.

D. Incorrect. This action would lower heat input to the 1A SG, however there is no procedural direction to operate the RCP.

Sys# System Category KA Statement W/E13 Steam EK2 Knowledge of the interrelations between Facility's heat removal systems, including primary coolant, emergency Generator the (Steam Generator Overpressure) and the coolant, the decay heat removal systems, and relations between the Overpressure/4 following: proper operation of these systems to the operation of the facility.

KIA# EK2.2 KIA Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

10M-53A.1.FR-H.2 pg 1 & 3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7/45.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

28. Preparations are being made to start the 1B Reactor Coolant Pump [1 RC-P-1 B] for post maintenance testing.

The 1B RCP Control Switch is taken to the START position, the following indications exist:

1B RCL Bypass lsol Vlv [MOV-1 RC-586] is OPEN

1B RCL Hot Leg lsol Vlv [MOV-1 RC-592] is CLOSED

1B RCL Cold Leg lsol Vlv [MOV-1 RC-593] is CLOSED

1B RCP Bearing Lift Pump [1-RC-P-1 B1] is RUNNING Which of the following INTERLOCKS will PREVENT the RCP from starting?

A. Lift Oil Pump [1 RC-P-1 B 1] STOPS after running for 170 seconds.

B. 1B 4KV Bus Frequency lowers to 59.0 Hertz.

C. 1B RCL Cold Leg lsol Vlv [MOV-1 RC-593] drifts OPEN.

D. RCP 1B Number 1 Seal Differential Pressure [PI-1CH-155A] lowers to 190 psid.

Answer: C Explanation/Justification:

A. Incorrect. This is the correct configuration for the lift oil pump the time delay of running for 120 seconds then the pumps stops 50 seconds later.

B. Incorrect. This frequency is low {<60 Hz) but above the interlock setpoint of 57.5 hz C. Correct. Must know the interlocks associated with the RCPs and setpoints, including the physical conditions of components. The RCP start requires a flowpath & adequate lift oil pump pressure for 120 seconds. The low frequency trip of the RCPs is based on a 2/3 bus coincidence.

The Cold leg valve must be closed, not open with the Bypass valve open to start the pump. The pump start circuit requires both valves in this alignment or the interlock to allow the pump start is not made.

D. Incorrect. This seal differential pressure is below the 200 psid limit, however this is not an interlock that will prevent the pump from starting. This is the correct PI, for the 8 RCP Sys# System Category KA Statement 003 Reactor Coolant Pump K6 Knowledge of the effect of a loss or malfunction on the following will have on the Starting requirements System (RCPS) RCPS:

KIA# K6.14 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

1OM-6.1.0 pg 1 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 45.5)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

29. Given the following plant conditions and sequence of events:

A Load reduction is in progress at 1% per minute

Reactor power is 22% and preparations are being made to take the turbine off-line

The Main Feed Regulating Bypass Valves have been transferred to AUTO

The 1B Reactor Coolant Pump (RCP) unexpectedly trips

No operator actions have been taken and the plant responds as designed Which ONE of the following will be the INITIAL effects of the RCP shutdown?

1B SG Steam Flow will __ (1) __.

1B SG Pressure will _ _ (2) _ _

A. (1) decrease (2) decrease B. (1) increase (2) decrease C. (1) decrease (2) increase D. (1) increase (2) remain the same Answer: A Explanation/Justification:

A. Correct. A loss of a single RCP below P-8 (30%) will NOT result in a reactor trip. The immediate effects of the tripped RCP in the effected loop is a decrease in steam flow (other two loops pick up flow). SG pressure will drop since loop Tavg is lower.

B. Incorrect. Opposite Steam Flow Response. Steam flow will drop in effected loop but will increase in unaffected loops. Correct steam pressure response.

C. Incorrect. Correct that steam flow decreases. SG pressure drops versus increases.

D. Incorrect. All parameter responses are incorrect. Plausible if the candidate does not understand RCP trip effects on these parameters.

Sys # System Category KA Statement 003 Reactor Coolant Pump K3 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: S/G System (RCPS)

KIA# K3.02 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

G03ATA 3.2 U1 PPNT Abnormal Tra*nsients, Rev. 4 slides 60 & 62 Question Source: Bank - 1LOTS NRC Exam April 2012 Q 28 Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

30. The plant is at 100% power.

Annunciator A3-54, Volume Control Tank Pressure High-Low alarms

The cause of the alarm is Low Hydrogen Gas supply pressure, due to empty Hydrogen bottles, at the Hydrogen Farm

An operator is dispatched to align new Hydrogen bottles The gas bottle supply vendor INADVERTANTLY filled the Hydrogen Bottles with Nitrogen, and Volume Control Tank Pressure is restored, using these improperly filled gas bottles.

What will be the potential impact of this inadvertent Nitrogen bottle alignment?

A. An INCREASE in RCS silica concentration.

B. An INCREASE in RCS dissolved Oxygen concentration.

C. A DECREASE in charging pump available NPSH.

D. A DECREASE in RCS dissolved ammonia concentration.

Answer: B Explanation/Justification:

A. Incorrect. Nitrogen cover gas will have no impact on silica.

B. Correct. At BVPS the design feature for controlling oxygen in the RCS is the maintenance of a Hydrogen cover gas pressure on the VCT. The candidate will need to have knowledge of this design feature and the potential impact of using the wrong gas as the cover pressure. Both hydrogen and nitrogen are used at different times. During SID conditions, nitrogen is used through the same pressure regulator.

C. Incorrect. Nitrogen cover gas will have no impact on charging pump available NPSH. Both cover gases will impact NPSH in the same manner.

Since BVPS has had gas bubble formation at the Charging pump suction (due to improper venting practices}, the candidate may believe the nitrogen will form gas bubbles at the charging pump suction.

D. Incorrect. Ammonia concentrations would increase as a result of additional nitrogen in the RCS.

Sys # System Category KA Statement 004 Chemical and Volume K4 Knowledge of CVCS design feature(s) and/or Oxygen control in RCS Control System (CVCS) interlock(s) which provide for the following:

KIA# K4.01 KIA Importance 2.8 Exam Level RO References provided to Candidate Technical

References:

10M-7.1.C page 10 151 paragraph None Question Source: New Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.7)

Objective:

Beaver Valley Unit 1 NRC Written Exam (1LOT14)

31. The plant is in Mode 4 with all systems in normal alignment for this condition.

T avg is 300 °F and stable

Following maintenance, the Residual Heat Removal System (RHR) is about to be filled, in preparation for Starting Up the system

RWST Boron concentration is within Tech Spec Limits

RCS Boron concentration is 1600 ppm

Background Source Range Counts are 4.0 x 102 CPS and stable The RHR system is filled, using the RWST and then placed in service.

How will the Source Range Counts change as a result of placing RHR in service?

Prior to RHR start up, what will be the High Flux at Shutdown alarm setpoint?

(1) Source Range Counts will _ _ _(1) _ _ _ as a result of placing RHR in service.

(2) Prior to RHR start up, the High Flux at Shutdown alarm setpoint will be _ _ _(2)_ __

A. (1) decrease (2) 1.26 x 103 CPS B. (1) decrease (2) 9.0 x 102 CPS C. (1) increase (2) 1.26 x 103 CPS D. (1) increase (2) 9.0 x 102 CPS Answer: A Explanation/Justification:

A. Correct. In order to answer the question, the candidate will need to know that the RWST TS boron cone. for Mode 4 is a value greater than the given RCS value. This will cause an increase in the RCS boron cone. when the RHR system is placed in service. This increase in boron cone. will result in SR counts decreasing. The Hi at SD alarm should be set Y. decade above background (3.16 times background).

B. Incorrect. Correct SR response. Wrong Hi <1> at SD alarm setpoint. Plausible if candidate interrupts Y. decade to be 500 cps.

C. Incorrect. Wrong SR response. Correct Hi at SD alarm setpoint.

D. Incorrect. Wrong SR response. Wrong Hi at SO alarm setpoint. Plausible if candidate interrupts Y. decade to be 500 cps.

Sys # System Category KA Statement 005 Residual Heat Removal K5 Knowledge of the operational implications of the following concepts Reactivity effects of RHR fill water System (RHRS) as they apply to the RHRS:

KIA# K5.03 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:
TS pages 3.5.4-1 & 2, Applied Fundamental knowledge, 10ST-2.3A page 20 NOTE Question Source: New Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
32. The plant is at 100% power with all systems in NSA EXCEPT RVLIS is NOT available.
A Large Break LOCA occurs coincident with a loss of offsite power
Emergency Diesel Generator (EDG #2) EE-EG-2 FAILS to start and CANNOT be started
RCS Pressure is 35 psig and stable
CNMT pressure is 25 psig and rising
CNMT Sump level is 55 inches and slowly rising
RWST level is 14.0 FT and slowly lowering
5 hottest core exit T/Cs are 400 oF and slowly rising
NO Quench Spray pumps are running
NO Recirculation Spray Pumps are running
All other equipment has operated as designed Based on these plant conditions, what component is at the highest risk and why?
A. Reactor Vessel, due to pressurized thermal shock concerns.
B. Reactor Core, due to high temperature.
C. Containment, due to high pressure.
D. SIS Accumulator Discharge Valve, due to high sump water level.
Answer: C Explanation/Justification:
A. Incorrect. With a Large Break LOCA in progress there will be no pressure stress on the Reactor Vessel, even with the cooldown.
B. Incorrect. Core Cooling is not yet being challenged, temperatures are rising, however the current conditions are not challenging this function.
w/o RVLIS core temperatures are not at the limit.
C. Correct. The conditions challenge the containment integrity because pressure is high and there is no ECCS flow from the Quench Spray system.
The containment fission product barrier is at risk due to leakage from the building at high pressure.
D. Incorrect. The SIS Accumulator valve is a component in Containment that could be affected by flooding however it is normally de-energized open to allow flow even if water level is high ..
Sys# System Category KA Statement 006 Emergency Core Cooling K3 Knowledge of the effect that a loss or malfunction of Containment System (ECCS) the ECCS will have on the following:
KIA# K3.03 KIA Importance 4.2 Exam Level RO References provided to Candidate Technical
References:
10M-53B.4.FR-Z.1 Pg 1 None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
33. The plant is at 100% power.
The control room crew is performing 10ST-6.2, Reactor Coolant System Water Inventory Balance
Test Duration time was 30 minutes
The following Pressurizer Relief Tank (PRT) data was taken in support of procedure:
PRT average temperature was 92 oF
Initial Average PRT level was 70.0%
Final Average PRT level was 74.5%
Based on this data, what is the Mass Leakage Rate INTO the PRT?
Refer to attached 1OST - 6.2 Attachment 4, PRT Mass Leakrate A. 0.0046 Ibm/min B. 0.0326 Ibm/min C. 125.8 Ibm/min D. 892.9 Ibm/min Answer: C Explanation/Justification:
A. Incorrect. This would be the number if the candidate uses v0 instead of vr in the steam table look-up.
B. Incorrect. This would be the number if the candidate multiplies vr instead of dividing.
C. Correct. Using Att. 4 & Steam Tables along with the data provided in the stem. Vf-Vi+=60.79 ft3, L\V!TD=2.03 ft3/min, N= .016106 ft3/lbm, VLR/v=125.81bm/min.
D. Incorrect. This would be the number if the candidate uses v0 instead of vr in the steam table look-up and multiplies instead of divides.
Sys# System Category KA Statement 007 Pressurizer Relief Tank/Quench Tank System (PRTS) Generic Knowledge of surveillance procedures.
KIA# 2.2.12 KIA Importance 3.7 Exam Level RO References provided to Candidate 1OST -6.2 Attachment 4 & Technical
References:
1OST -6.2 Attachment 4 & Steam Tables Steam Tables Question Source: New Question Cognitive Level: High - Application 10 CFR Part 55 Content: (CFR: 41.10 /45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
34. The plant is at 100% power.
PRZR PORV Relief Valve [PCV-1RC-455C] begins LEAKING.
Troubleshooting efforts to stop the leaking PORV reveal that the valve will NOT manually cycle AND is STUCK - 1% open.
Pressurizer Relief Tank Pressure [PI-1 RC-472] has RISEN to 50 psig.
(1) What will be the STABLE PORV relief line temperature, for these conditions?
(2) IF PRZR PORV Relief Valve [PCV-1RC-455C] is declared INOPERABLE, what Tech. Spec.
Actions will be REQUIRED within 1 Hour?
A. (1)281 °F (2) Close and MAINTAIN power to the associated block valve.
B. (1)281 OF (2) Close and REMOVE power to the associated block valve.
C. (1)298°F (2) Close and MAINTAIN power to the associated block valve.
D. (1) 298 OF (2) Close and REMOVE power to the associated block valve.
Answer: D Explanation/Justification:
A. Incorrect. Wrong Relief line temperature, this is the relief line temperature for 50 psia NOT 50 psig. Wrong TS action, this is the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action if the PORV is capable of being manually cycled and the stem of the question clearly indicates that it cannot be cycled.
B. Incorrect. Wrong Relief line temperature, this is the relief line temperature for 50 psia NOT 50 psig. Correct TS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action.
C. Incorrect. Correct Relief line temperature. Wrong TS action, this is the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action if the PORV is capable of being manually cycled and the stem of the question clearly indicates that it cannot be cycled.
D. Correct. lAW fundamental knowledge and use of steam tables and TS page 3.4.11-1.
Sys # System Category KA Statement 007 Pressurizer Relief A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Stuck-open PORV Tank/Quench Tank PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate or code safety System (PRTS) the consequences of those malfunctions or operations:
KlA# A2.01 KJA Importance 3.9 Exam Level RO References provided to Candidate Steam Tables with Mollier Diagram Technical
References:
Steam Tables, TS page 3.4.11-1 Question Source: New Question Cognitive Level: Low _ Fundamental 10 CFR Part 55 Content: (CFR: 41.51 43.5 1 45.3 1 45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
35. The plant is at 100% power.
CCR 1C Pump [1 CC-P-1 C] is tagged out for maintenance
CCR 1A Pump [1CC-P-1A] is running
CCR 18 Pump [1 CC-P-1 B] is in standby CCR 1A Pump [1CC-P-1A] TRIPS on overcurrent and CCR 18 Pump [1CC-P-1B] CANNOT be started.
Which of the below listed loads will have LOST cooling water as a result of this failure?
A. Main Feed Pumps B. Gaseous Waste Compressors C. Charging Pump Lube Oil Coolers D. CNMT Air Recirculation Fans Answer: B Explanation/Justification:
A. Incorrect. MFPs are cooled by turbine plant CCW.
B. Correct. These are cooled by Primary Plant CCW. The trip of the running pump and start failure of the standby pumps results in a loss of CCW system flow. The 1C pumps can be aligned to replace either pump, however it is unavailable. The loss of cooling flow will effect the gaseous waste compressors since they are cooled by CCW.
C. Incorrect. Charging Pump Lube Oil Coolers are cooled by river water.
D. Incorrect. CNMT Air Recirculation Fans are cooled by chilled water.
Sys # System Category KA Statement 008 Component Cooling K3 Knowledge of the effect that a loss or malfunction of the CCWS will have on Loads cooled by CCWS Water System (CCWS) the following:
KIA# K3.01 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical
References:
OM Fig. 15-2 (A6) and (A2)
None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)*
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
36. A plant heatup is in progress with the following conditions:
Pressurizer pressure is 1800 psig and stable
All systems are in normal alignment for these conditions Pressurizer Chan 2 Press [PI-1 RC-445] FAILS HIGH.
What impact, if any, will this failure have on the PRZR PORVs and block valves?
A. ONLY PRZR PORV lsol MOVs [MOV-1RC-535, 536, & 537] will AUTO CLOSE.
B. ONLY PRZR PORV Relief Valves [PCV-1 RC-456 & 4550] will OPEN.
C. ONLY PRZR PORV Relief Valve [PCV-1RC-455C] will OPEN.
D. NO PORVs will OPEN and NO PRZR PORV lsol MOVs will CLOSE.
Answer: D Explanation/Justification:
A. Incorrect. The P-11 permissive provides an interlock for opening the PORV but not for the PORV block valves. The P-11 permissive is provided by the protection channels not the control channels. The failed transmitter is a control channel. P-11 does provide an auto closed signal to the PORV block valves (interlock similar to Unit 2) but this does not exist at Unit 1.
B. Incorrect. This would be the correct answer if PRZR pressure was already above 2000 psig and the P-11 interlock was met.
C. Incorrect. This would be the correct answer if PRZR pressure was already above 2000 psig, the P-11 interlock was met, and channel1 had failed.
D. Correct. In order to answer the question, the candidate must evaluate each potential impact presented in the choices. The candidate must have knowledge of the PORV bloc~ valve logic to rule out choice A. The candidate must also have knowledge of the PORV auto open circuitry to rule out choices B & C. Specifically, the candidate must have knowledge where the P-11 interlock is generated and how P-11 impacts PORV automatic operations. In this particular case the failed transmitter will have no impact on the PORVs or their block valves since the P-11 permissive has not been met.
Sys# System Category KA Statement 010 Pressurizer Pressure Control A4 Ability to manually operate and/or monitor in the control room: PORV and block valves System (PZR PCS)
KIA# A4.03 KIA Importance 4.0 Exam Level RO References provided to Candidate None Technical
References:
1OM-6.1.0 page 7 2"d to last paragraph, 1OM-6.4.1F pages 22 & 23 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5 to 45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
37. The plant is at 100% power.
The following conditions exist:
Both PRZR Spray Valves are CLOSED
PRZR Pressure is 2220 psig and stable
Regenerative HX Charging Outlet Temp [TI-1CH-123] is 430 oF and stable Which of the below listed actions will result in the LARGEST thermal shock to the PRZR spray nozzle?
A. Closing valve [1 RC-51] Byp Spray Man Throttle THEN 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months later fully opening the associated PRZR spray valve.
B. Opening valve [1 RC-51] Byp Spray Man Throttle THEN 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months later fully opening the associated PRZR spray valve.
C. Volume Control Tank level drops to 4%.
D. Energizing a PRZR backup heater.
Answer: A Explanation/Justification:
A. Correct. The BVPS design feature that provides spray valve warm-up is accomplished by throttling the PRZR Spray Valve Bypass valves. The line warm-up is procedurally controlled by throttling the manual bypass valves around the PRZR spray valves which provides a continuous small amount of spray flow. 1OM-6.4.M accomplishes this design concern. In order to answer the question, the candidate must have knowledge of how BVPS limits thermal stress/shock on the PRZR spray nozzle, and must be aware that the NSA position of 1RC-51 is throttled open. Since the valve is normally throttled open to limit stress/shock, closing it then opening the spray valve defeats the purpose of the valve and will result in the largest thermal shock to the PRZR spray nozzle.
B. Incorrect. The initial opening of 1RC-51 will put some stress on the nozzle, but this valve is only a% valve and the stress will be smaller than choice A. Then opening the spray valve will result in little to no change in any thermal stress.
C. Incorrect. VCT level drop will result in an automatic swap of the charging pump suction to the cold RWST water. However, in NSA this will result in a thermal shock to the charging line penetration but not the PRZR spray nozzle. IF MOV 1CH-311 aux. spray valve were open at this time, this would result in thermal shock to the PRZR spray nozzle.
D. Incorrect. This will result in a larger~T across the PRZR spray nozzle. However, the temperature rise is very gradual and the KW rating of the heater is not large enough to cause thermal shock of the PRZR nozzle.
Sys # System Category KA Statement 010 Pressurizer Pressure Control K4 Knowledge of PZR PCS design feature(s) and/or interlock(s) which Spray valve warm-up System (PZR PCS) provide for the following:
KIA# K4.01 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical
References:
1OM-6.4.M page 2 purpose and page 5 step 7a Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
39. Given the following conditions:
Rx Power is 2 x 1o- 11 amps on both IR channels
All systems are in normal alignment for this power level
Intermediate Range (IR) Channel N-35 indication becomes erratic
Maintenance decides to perform a calibration of IR channel N-35
In support of this calibration, the control room crew places theIR channel N-35 Level Trip Switch to the BYPASS position
However, theIR channel N-35 Level Trip Switch malfunctions and NO BYPASSES are instated When the IR channel N-35 Level Trip Switch was placed to the BYPASS position, which of the below listed bypasses SHOULD HAVE been instated?
1. Train A High IR Flux Rod Stop (C-1)
2. Train A High IR Flux Trip
3. Train A IR Neutron Flux (P-6) Interlock A. 1 & 3 ONLY B. 1 & 2 ONLY C. 2 & 3 ONLY D. 1, 2, & 3 Answer: B Explanation/Justification:
A. Incorrect. Plausible if the candidate believes the switch includes a P-6 bypass. P-6 is generated from a separate bistable that does not have block or bypass capabilities.
B. Correct. The candidate must know that the rod stop is also bypassed with the level trip switch, which is not apparent in the switch name.
Candidate may believe theIR Trip block switches (BB-B) are the only switches that will also block the rod stop.
C. Incorrect. Plausible if the candidate believes the level trip switch does not block the rod stop and does block P-6.
D. Incorrect. Plausible if the candidate believes the level trip switch also blocks P-6. P-6 is generated from a separate bistable that does not have block or bypass capabilities.
Sys # System Category KA Statement 012 Reactor Protection K6 Knowledge of the effect of a loss or malfunction of the following will Bistables and bistable test equipment System have on the RPS:
KIA# K6.01 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical
References:
UFSAR Fig. 7.2-1 SH 2, 3, & 4 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7/45.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
38. Given the following conditions:
Rx Power is 2 x 1o-8 amps on both IR channels
All systems are in normal alignment for this power level
A Loss of Vital Bus IV occurs (Bus is de-energized)
Which of the below listed Rx Trip BISTABLES will be TRIPPED in the Reactor Protection System as a DIRECT result of this failure?
A. Pressurizer Low Pressure B. Power Range High Flux Low C. Loop 3 RCS Flow Low D. Intermediate Range N-36 High Flux Answer: B Explanation/Justification:
A. Incorrect. Some trip circuits that only have 3 signals do not use Channell as the power supply, rather they use Channel IV. This 3 loop circuit is powered from Channels I, II, & Ill.
B. Correct. A loss of Vital Bus IV will de-energize Power Range Channel N-44 causing the Power Range High Flux Low bistable to trip. Power is
< P1 0 (1 0%) in the stem of the question.
C. Incorrect. Some trip circuits that only have 3 signals do not use Channell as the power supply, rather they use Channel IV. This 3 loop circuit is powered from Channels I, II, & Ill.
D. Incorrect. Plausible if candidate believes N-36 is channel IV. Since there are only 2 channels, the power distribution could have used channel IV for this instrument loop. N-36 is Channel II powered.
Sys# System Category KA Statement 012 Reactor Protection System K2 Knowledge of bus power supplies to the following: RPS channels, components, and interconnections KIA# 1<2.01 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical
References:
1OM-53C.4.1.38.1D pg 5 step 6a & pg 14 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
40. Which of the following conditions would generate an Auto Start signal for Inside Recirculation Spray pump [1 RS-P-1A]?
Containment Pressure of ___(1 )_ _ _ WITH a Refueling Water Storage Tank Level of
_ _(2) _ _,
A. (1) 8.3 psig (2) 12 feet 7 inches B. (1) 13.3 psig (2) 26 feet 6 inches C. (1) 10.3 psig (2) 13 feet 9 inches D. (1) 12.4 psig (2) 28 feet 4 inches Answer: B Explanation/Justification:
A. Incorrect. The CNMT CIB setpoint is not exceeded, SIS and MSLI setpoints are exceeded. The RWST level is below setpoint for Transfer to Recirc.
B. Correct. For BVPS ESFAS Safety related Containment Cooling is provided by RS and QS pumps. RS pump starts on CIB signal (11.1 psig) in CNMT and a RWST Low level of 27 feet 7.5 inches. The ESFAS signal effect on Containment Cooling is to start the RS pumps when the CIB signal and RWST level setpoints are reached. The CNMT CIB setpoint is exceeded, the RWST level is below setpoint. The student must evaluate the combination of conditions that meet the RSS pump start logic of the CIB (ESFAS) signal in combination with low RWST level.
C. Incorrect. The CNMT CIB setpoint is not exceeded SIS and MSLI setpoints are exceeded, the RWST level is below setpoint for Transfer to Recirc.
D. Incorrect. The CNMT CIB setpoint is exceeded, the RWST level is not below setpoint.
Sys # System Category KA Statement 013 Engineered Safety Features K1 Knowledge of the physical connections and/or cause effect relationships CCS Actuation System (ESFAS) between the ESFAS and the following systems:
KIA# K1.03 KIA Importance 3.8 Exam Level RO References provided to Candidate None Technical
References:
10M-13.2.B pg 2, 10M-13.1.D pg 5 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.2 to 41.9/45.7 to 45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
41. The plant is at 100% power.
Which of the following conditions would cause Control Rod Drive Mechanism Shroud Fan
[1VS-F-2A] to auto STOP?
Consider each condition separately.
A. 4KV bus 1OF is de-energized B. Containment Isolation Phase A signal (CIA)
C. Reactor Coolant System pressure lowers to 1750 psig D. Containment Pressure rises to 13.0 psig Answer: D Explanation/Justification:
A. Incorrect. The 1DF 4 KV bus powers the 1P 480 VAC bus and Stub bus. The 2A CRDM fan is powered from the 1N bus.
B. Incorrect. The fans do not receive a trip signal from a CIA, nor is the bus power supply affected.
C. Incorrect. RCS pressure this low would generate a SIS signal, the fans are not affected by this signal.
D. Correct. BVPS uses CAR and CRDM shroud fans for Containment cooling. The fans are powered from the emergency 480 VAC busses via the Stub Busses. These busses are stripped on a CIB signal. The fans are stopped on the initiation of the Safeguards CIB signal via the bus power loss. This high of a CNMT pressure would generate a CIB signal, which will open the 480 VAC Stub Bus supply breaker, stopping the Fan.
Sys # System Category KA Statement 022 Containment Cooling A3 Ability to monitor automatic operation of the CCS, Initiation of safeguards mode of operation System (CCS) including:
KIA# A3.01 KIA Importance 4.1 Exam Level RO References provided to Candidate None Technical
References:
10M-44C.1.E pg 7 10M-37.1.D pg 9 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 /45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
42. A Large Break LOCA occurred eight (8) hours ago, all systems function as designed.
Which of the following identifies the flowpath of core cooling, number of spray pumps in operation and the suction water supply, as established by the Emergency Operating Procedures?
HHSI is High Head Safety Injection LHSI is Low Head Safety Injection HHSI Injection LHSI Injection Spray Pumps Suction Source Condition 1 Cold Legs Hot Legs 2 Recirc Spray CNMT Sump Condition 2 Hot Legs Cold Legs 2 Quench Spray RWST Condition 3 Hot Legs Cold Legs 4 Recirc Spray CNMT Sump Condition 4 Cold Legs Hot Legs 2 Quench Spray RWST A. Condition 1 B. Condition 2 C. Condition 3 D. Condition 4 Answer: A Explanation/Justification:
A. Correct. For a large Break LOCA, the plant will align for transfer to recirc in ES-1.3 then Simultaneous HoUCold leg recirc by 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the event. Injection will be to the Cold Legs from the HHSI pumps and Hot legs from the LHSI pumps. After 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months procedure ES-1.4 transfer to Simultaneous Hot and Cold Leg Injection will align injection via this lineup, Two Recirc Spray pumps will be secured in ES-1.3. Cooling via spray water mixing in the CNMT Sump.
B. Incorrect. This lineup is for the initial injection phase following a large Break LOCA, after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the QS pumps will have been secured and flow established from the CNMT sump.
C. Incorrect. The HHSI pumps inject via the Cold Legs and the LHSI via the Hot legs, only 2 RS pumps will be in service by procedure at eight hours after the event.
D. Incorrect. The pump injection points are correct, however 2 RS pumps would be in service not 2 QS pumps, pump suction source would be the CNMT sump, not the RWST.
Sys # System Category KA Statement 026 Containment Spray System (CSS) K1 Knowledge of the physical connections and/or cause/effect relationships ECCS between the CSS and the following systems:
KIA# K1.01 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical
References:
E-1 pgs 15 & 21, ES-1.3 pg 3, ES-1.4 pg 3, 1SQS-11.1 Rev. 13, PPNT slide 106 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 I 45.7 to 45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
44. Which of the following conditions will REQUIRE Technical Specification action within one hour or less? Consider each situation independently.
A. Plant is at 100% Power, QPTR is calculated to be 1.03.
B. Plant is in Mode 3, two Steam Generator Atmospheric Steam Dump Valves [PCV-1MS 101A and 101 B] are BOTH INOPERABLE.
C. Plant is at 30% Power, 1A First Stage Pressure [PT-1MS-446] fails Low (not selected for control).
D. Plant is at 3% power, 125 VDC Battery# 1 (Bat #1) is declared OOS.
Answer: C Explanation/Justification:
A. Incorrect. Per 3.2.4 Action is to reduce power from RTP 3% for every 1% in 2 hrs.
B. Incorrect. Per TS 3.7.4 Atmospheric valve failures are 24 hr action statement in Modes 1, 2, 3 & 4 if SGs are required.
C. Correct. Per TS 3.3.1 function 17.f action P, 1 hr action to verify the P-7 interlock is in required state for plant conditions.
D. Incorrect. Per TS 3.8.4 Action B, 2 hrs is allowed to restore the battery to operable status in Modes 1-4.
Sys # System Category KA Statement 039 Main and Reheat Generic Knowledge of less than or equal to one hour Technical Steam System (MRSS) Specification action statements for systems.
KIA# 2.2.39 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
TS 3.3.1 Function 17.f Action P Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7/41.10 /43.2/45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
45. The plant has tripped from 100% power coincident with a loss of offsite power.
Both Emergency Diesel Generators have failed to start
Procedure ECA-0.0, Loss of All Emergency 4KV AC Power has been entered
Instrument Air Pressure is 0 psig
Chemistry reports that there is high RCS activity due to failed fuel (1) IF a Steam Generator Tube Rupture were to subsequently occur, which Radiation Monitors would be used to identify the ruptured Steam Generator?
AND (2) IF the Steam Generator pressure rises due to decay heat, with no operator actions, what is the highest pressure that will exist in the ruptured Steam Generator, assuming all systems function as designed? .
A. (1) N-16 Steam Generator Leak Monitor (RM-1MS-102A, B, C)
(2) 1060 psig B. (1) N-16 Steam Generator Leak Monitor (RM-1MS-102A, B, C)
(2) 1075 psig C. (1) Steam Relief Monitors (RM-1MS-100A, B, C)
(2) 1060 psig D. (1) Steam Relief Monitors (RM-1MS-100A, B, C)
(2) 1075 psig Answer: D Explanation/Justification:
A. Incorrect. The N-16 monitors are used at power and are located downstream of the MSIVs which will be closed, this setpoint is for the Atmospherics and they are not available due to the loss of air.
B. Incorrect. The N-16 monitors are used at power and are located downstream of the MSIVs which will be closed, this setpoint is for the lowest Safety valve.
C. Incorrect. The MS safety valve RM is upstream of the MSIVs and is a high range monitor, his setpoint is for the Atmospherics and they are not available due to the loss of air.
D. Correct. The MS safety valve RM is upstream of the MSIVs and is a high range monitor, this setpoint is for the lowest Safety valve.
Sys # System Category KA Statement 039 Main and Reheat A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding Main steam line radiation Steam System (MRSS) design limits) associated with operating the MRSS controls including: monitors KIA# A 1.09 KIA Importance 2.5* Exam Level RO References provided to Candidate None Technical
References:
1OM-43.1.C pg 6 & 7, 1OM-21.2.8 pg 2 &3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
46. The plant is at 100% power.
The following conditions exist:
1C SG STM Flow Sig Sel FC 1FW 498 is in the FM 494 (Channel 3) position
Steam Pressure transmitter [PT-1 MS-495] (Channel 3) is slowly failing downscale What INITIAL effect will this failure have on the 1C SG Main FW Feed Reg Vlv [FCV-1 FW-498]
and how will the Main Feedwater Regulating valve be controlled once 10M-24.1F, Instrument Failure Procedure is complete?
Main Feed Regulating Valve [FCV-1FW-498] will INITIALLY (1) , and once the actions of the Instrument Failure Procedure are complete the Main Feedwater Regulating valve will be controlled in (2) _ _
A. (1) open (2) manual B. (1) open (2) auto C. (1) close (2) manual D. (1) close (2) auto Answer: D Explanation/Justification:
A. Incorrect. The Steam pressure failure response is the same as steam flow, the valve closes, per the IF the valve will be returned to Auto.
B. Incorrect. The Steam pressure failure response is the same as steam flow, the valve closes, per the IF the valve will be returned to Auto.
C. Incorrect. The Steam pressure failure response closes the MFRV, per the IF the valve will be returned to Auto after selecting the alternate channel.
D. Correct. The Steam pressure failure response closes the MFRV, per the IF the valve will be returned to Auto after selecting the alternate channel.
Sys# System Category KA Statement 059 Main Feedwater A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Failure of feedwater control (MFW) System MFW; and (b) based on those predictions, use procedures to correct, control, or system mitigate the consequences of those malfunctions or operations:
KIA# A2.11 KIA Importance 3.0* Exam Level RO References provided to Candidate Technical
References:
10M-24.4.1F pg 28 None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.5/ 43.5 I 45.3/ 45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
47. The reactor has tripped from 100% power due to a small RCS leak. Procedure E-0, Reactor Trip or Safety Injection has been exited, and procedure ES-1.1, Sl 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 20, Check Intact Steam Generator Levels is being performed The Steam Generator parameters are as follows:
Steam Generator Narrow Range Level AFWFiow 1A 25% 115 gpm 18 28% 114 gpm 1C 33% 105 gpm (1) What was the effect on the AFW pump discharge PRESSURE as the throttle valves were throttled closed? (Assume AFW pump recirc valves remain shut throughout)
2) What is the current condition of the Red Path Heat Sink Status tree?
Auxiliary Feedwater pump discharge PRESSURE (1) as the throttle valves were closed, and the entry into FR-H.1, Response to Loss of Secondary Heat Sink
_ _ _(2) required.
A. (1) increased (2) is not B. (1) increased (2) is C. (1) decreased (2) is not D. (1) decreased (2) is Answer: A Explanation/Justification:
A. Correct. Requires an understanding of Pump flow curve and an assessment of the Steam Generator and plant conditions to evaluate the status of the Heat Sink status tree. Discharge Pressure I Head increases as the pump discharge valve is closed. Heat sink is met, adverse CNMT is not applicable and level requirement is 31% in at least one SG B. Incorrect. Pump Discharge Pressure I Head increases, Heat sink requires 370 GPM of flow total or NR 31% in at least one SG, no entry required.
C. Incorrect. Pump Discharge Pressure I Head does not decrease. Heat sink is met requires NR 31% in at least one SG. If they believe the PT is downstream of the throttle valves, then the pressure would lower as the SG pressure lowers.
D. Incorrect. Pump Discharge Pressure I Head does not decrease. Heat sink requires 370 GPM of flow total or NR 31% in at least one SG Sys# System Category KA Statement 061 Auxiliary I Emergency K5 Knowledge of the operational implications of Pump head effects when control valve is shut Feedwater (AFW) System the following concepts as the apply to the AFW:
KIA# K5.03 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical
References:
ES-1.1 Step 20, FR-H.1 pg 2, GO-GPF.C2 Rev. 2 pg 29 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR: 41.51 45. 7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
48. The plant has tripped due to a loss of all 4160 VAC power.
Procedure ECA-0.0, Loss of All Emergency 4KV AC Power was performed.
Emergency Diesel Generator (EDG #1) EE-EG-1 is currently loaded to 2400 KW
The crew has transitioned to ECA-0.2, Loss of All AC Power Recovery With Sl Required
The crew is evaluating starting the additional equipment loads listed below:
COMPONENT KWLoad
a. PZR-HTR-A, Pressurizer Backup Heaters 215
b. PZR-HTR-0, Pressurizer Backup Heaters 270
c. 1VS-F-22A, Diesel Generator Bldg Exh Fan 28
d. BAT-CHG-1, 3 Station Battery Charger 19
e. 1LO-M-7, Turbine Bearing Lube Oil Pump 39
f. 1LO-M-5, Main Turbine Turning Gear Motor 39 (1) Per ECA-0.2, Loss of All AC Power Recovery With Sl Required, which of the following lists of components may be loaded on the Emergency Diesel Generator without exceeding the MAXIMUM KW load limit?
AND (2) What is the consequence of exceeding the Maximum Emergency Diesel Load limit?
A. (1)a,b,c,e (2) The electrical bus may be lost B. (1) a, b, c, e (2) Excessive carbon buildup in the exhaust system C. (1)a,c,d,f (2) The electrical bus may be lost D. (1) a, c, d, f (2) Excessive carbon buildup in the exhaust system Answer: C Explanation/Justification:
A. Incorrect. This amounts to 2952 KW and would exceed the load limit of 2750 KW, the potential exists for the bus to fail per ECA-0.2 background.
B. Incorrect. This amounts to 2952 KW and would exceed the load limit of 2750 KW. Excessive carbon buildup is a concern for low loading.
C. Correct. Per ECA-0.2 Caution the maximum EDG load limit is 2750 KW, per the OM the maximum load for 0.5 hrs is 3050 and 2950 for 168 hrs per year. If these values are assumed to be the procedural maximums, then the choices with the highest load would be selected. The EOP bases lists the potential of bus failure as a consequence, low load has the potential for exhaust buildup per 1OM-36.4.AG pg 10. This amounts to 2701 KW and would not exceed the load limit of 2750 KW, the potential exists for the bus to fail per ECA-0.2 background.
D. Incorrect. This amounts to 2701 KW and would not exceed the load limit of 2750 KW. Excessive carbon buildup is a concern for low loading.
Sys # System Category KA Statement 062 AC Electrical A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding Significance of DIG load Distribution System design limits) associated with operating the ac distribution system controls including: limits KIA# A 1.01 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical
References:
1OM-53A.1.ECA-0.2 pg 3 & Att. 2-F pg 4, 1OM-538.4.ECA-0.2 pg 11 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
49. The plant is at 100% power.
Which of the following combinations of Control Room indications would REQUIRE a reactor trip?
A. Battery Charger# 1 GREEN Light NOT Lit and RED light Lit.
Battery Breaker# 1 GREEN Light Lit and RED Light NOT Lit.
B. Battery Charger# 2 GREEN Light Lit and RED light NOT Lit.
Battery Breaker# 2 GREEN Light Lit and RED Light NOT Lit.
C. Battery Charger# 3 GREEN Light Lit and RED light NOT Lit.
Battery Breaker# 3 GREEN Light Lit and RED Light NOT Lit.
D. Battery Charger# 5 GREEN Light NOT Lit and RED light Lit.
Battery Breaker# 5 GREEN Light NOT and RED Light Lit.
Answer: 8 Explanation/Justification:
A. Incorrect. The Battery Charger will supply power to 125 VDC Bus #1, there is no need to trip the reactor.
B. Correct. A loss of either 125 VDC Bus 1 & 2 will result in a Reactor Trip due to the associated RCP trip, feedwater valves will also fail closed. The loss of 125 VDC bus 3 or 5 will not result in a reactor trip. The student must evaluate the breaker indicating lights and determine where the DC bus has power, then evaluate which busses, if lost, would require a reactor trip. The 125 VDC Bus #2 is deenergized since both the Battery and Charger supply breakers are open. The reactor is tripped because the loss of 125 VDC Bus 2 results in a loss of CCR to all RCPs, requiring them to be secured. Also the MFRVs and Bypass vlvs will fail closed.
C. Incorrect. These indications would result in a loss of power to 125 VDC Bus #3, however no reactor trip signal is generated or required for a loss of 125 VDC Bus #3.
D. Incorrect. The Battery will supply power to 125 VDC Bus #5, and there is no need to trip the reactor for a loss of 125 VDC Bus #5.
Sys# System Category KA Statement 063 DC Electrical Distribution A3 Ability to monitor automatic operation of the DC electrical Meters, annunciators, dials, System system, including: recorders, and indicating lights KIA# A3.01 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical
References:
1OM-53C.4.1.39.1 B pg 1 & 9 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
50. The plant is in Mode 3 with all systems in normal alignment for this Mode.
125V DC Bus 1 Voltage is ZERO
Subsequently, a Loss of Offsite power occurs How will these conditions impact #1 EDG start capabilities?
EDG #1 (1) be MANUALLY started from the control room by depressing both STOP pushbuttons, then selecting EXERCISE, then depressing the START pushbutton AND EDG #1 (2) be LOCALLY started by placing the AUTO/LOCAL switch to LOCAL, then depressing the ENGINE START pushbutton.
A. (1) CAN (2) CANNOT B. (1) CAN (2) CAN C. (1) CANNOT (2) CANNOT D. (1) CANNOT (2) CAN Answer: C Explanation/Justification:
A. Incorrect. Plausible if candidate believes that only the local manual start circuitry is impacted by these conditions.
B. Incorrect. Plausible if candidate believes that only the Auto start feature is impacted by these conditions.
C. Correct. The EDG start solenoids will not have power and cannot be started from the CR or locally. The sequence for all of the manual/local starts are all correct for starting the EDG if there was power available to the solenoids.
D. Incorrect. Plausible if candidate believes that only the CR manual start circuitry is impacted by these conditions.
Sys# System Category KA Statement 064 Emergency Diesel K1 Knowledge of the physical connections and/or cause/effect relationships DC distribution system Generator (ED/G) System between the ED/G system and the following systems:
KIA# K1.04 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical
References:
1OM-53C .4 .1 .39 .1 A Att. 1 page 20 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 I 45.7 to 45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
51. The plant is at 100% power with all systems in NSA EXCEPT for the following:
1OST-24.4, Steam Turbine Driven Auxiliary Feed Pump Test [1 FW-P-2] is being performed
Unit 1 Waste Gas Decay Tank [GW-TK-1 B] is being discharged Based on these conditions, which of the below Radiation Monitor High-High alarms will cause the Main Filter Banks to automatically realign to filter the discharging effluents?
A. Gaseous Waste Particulate [RIS-GW1 08A]
B. Condenser Air Ejector Vent [RIS-SV1 00]
C. 1FW-P-2 Exhaust [RIS-MS-1 01]
D. Aux Bldg Vent Sys B Gas [RIS-VS102B]
Answer: D Explanation/Justification:
A. Incorrect. Hi-Hi- alarm will automatically isolate the discharge but will not re-align the discharge thru the MFB.
B. Incorrect. Hi-Hi- alarm will automatically isolate the normal discharge flowpath and align the discharge to CNMT, but will not re-align the discharge thru the MFB.
C. Incorrect. This monitor does not have any automatic functions. If the Hi-Hi- alarm is received, the effluent will continue to discharge. The monitor is purely for monitoring purposes and offsite dose projections.
D. Correct. The MFB will filter the effluent and reduce the release to values that are within acceptable design limits for 10CFR20.
Sys # System Category KA Statement 073 Process Radiation A 1 Ability to predict and/or monitor changes in parameters (to prevent exceeding Radiation levels Monitoring (PRM) System design limits) associated with operating the PRM system controls including:
KIA# A1.01 KIA Importance 3.2 Exam Level RO References provided to Candidate None Technical
References:
1OM-43.1.C page 11 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.5 I 45. 7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
52. The plant is at 100% power.
Which of the following conditions will result in an automatic CLOSURE of CCR HX RW Series lsol Vlv [MOV-1 RW-1148]?
A. PRZR Pressure drops to 1200 psig B. CNMT pressure rises to 12 psig C. CCR HX RW Pressure drops to 12 psig D. CNMT Radiation level rises to 1.2 X 105 R/Hr Answer: 8 Explanation/Justification:
A. Incorrect. This would cause Sl and CIA but not Cl8.
B. Correct. MOV-1 RW-1148 isolates RW to the CCR (aux bldg supply) and allows all RW to be directed to the recirc spray HXs for core cooling and CNMT cooling.
C. Incorrect. Candidate may have the misconception that MOV-1 RW-1148 has an isolation feature to auto close when low pressure (a leak downstream) is detected.
D. Incorrect. This would be indicative of Adverse CNMT conditions, but not an auto close signal to MOV-1RW-1148.
Sys# System Category KA Statement 076 Service Water K4 Knowledge of SWS design feature(s) and/or Conditions initiating automatic closure of closed cooling System (SWS) interlock(s) which provide for the following: water auxiliary building header supply and return valves KIA# K4.01 KIA Importance 2.5* Exam Level RO References provided to Candidate None Technical
References:
10M-30.1.D page 10 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
53. The plant is in Mode 3 with all systems in normal alignment for this Mode.
1B Station Air compressor is running and 1A Station Air compressor is in Standby
A loss of offsite power occurs
All systems function as designed in response to these conditions EXCEPT Emergency Diesel Generator (EDG #2) EE-EG-2 FAILS to start Which, if any, Air compressors will be supplying compressed air to the Instrument Air Header?
(ASSUME NO OPERATOR ACTIONS HAVE BEEN TAKEN).
A. ONLY 1A Station Air Compressor.
B. 1A Station Air Compressor AND the Diesel Driven Air Compressor.
C. ONLY the Diesel Driven Air Compressor.
D. NO Air Compressors will be running.
Answer: C Explanation/Justification:
A. Incorrect. 1A Station Air compressor is powered by the 1B 480V normal bus. Plausible since the A Train of emergency power will be available based on the conditions stated in the stem. If the 1A Station Air compressor was running, pressure would not drop low enough to reach the auto start setpoint for the diesel driven air compressor.
B. Incorrect. 1A Station Air compressor is powered by the 1B 480V normal bus. Plausible since the A Train of emergency power will be available based on the conditions stated in the stem. The diesel driven air compressor is correct.
C. Correct. Requires knowledge of the power supplies for all of the motor driven air compressors to select this answer. The Diesel Air Compressor is self-powered and will supply the system. With a loss of off-site power there will be no normal power available to the SACs.
D. Incorrect. 1A Station Air compressor is powered by the 1B 480V normal bus. 1B Station Air compressor is powered by the 1C 480V normal bus.
The diesel driven air compressor has an auto start feature at 95 psig, whereas the old diesel driven air compressor had to be manually started.
Sys # System Category KA Statement 078 Instrument Air System (lAS) K2 Knowledge of bus power supplies to the following: Instrument air compressor KIA# K2.01 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical
References:
10M-34.4.U Page 2 of 18 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
54. The plant is at 100% power.
Based on the information provided on the attached IPC screens, 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 IPC screens 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: D Explanation/Justification:
A. Incorrect. See correct answer explanation.
B. Incorrect. See correct answer explanation.
C. Incorrect. See correct answer explanation.
D. Correct. The CNMT IPC screen shots were assembled to purposely make the average temp. 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.
Several 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. The CNMT pressure screen has both psia and psig numbers which is intended to verify candidate awareness of the LCO requirement being in psia and not psig. To answer this question the candidate will need to assess the computer data provided and determine the status of the LCOs.
Sys# System Category KA Statement 103 Containment System Generic Ability to use plant computers to evaluate system or component status.
KIA# 2.1.19 KIA Importance 3.9 Exam Level RO References provided to Candidate CNMT IPC screen shots Technical
References:
CNMT IPC screen shots, TS LCOs 3.6.4 & 3.6.5, 10M-54.3.L5 pg 27 & 28 Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: (CFR: 41.10 /45.12)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
55. The plant is at 50% power with all systems in normal alignment for this power level.
Containment pressure transmitter PT-1 LM-1 OOD failed HIGH
The instrument failure procedure for PT-1 LM-1 OOD has been completed A Main Steam line leak is in progress resulting in the following indications:
Containment pressure transmitter PT-1 LM-1 OOA indicates 9.5 psig and is slowly rising
Containment pressure transmitter PT-1 LM-1 008 indicates 11.5 psig and is slowly rising
Containment pressure transmitter PT-1LM-100C indicates 12.0 psig and is slowly rising
The plant remains at power What will be the response of Containment pressure to these conditions?
A. CNMT Pressure will continue to rise until CIA is automatically actuated.
B. CNMT Pressure will continue to rise until CIB is automatically actuated.
C. CNMT Pressure will lower when CIA is manually actuated.
D. CNMT Pressure will lower when CIB is manually actuated.
Answer: D Explanation/Justification:
A. Incorrect. SIS I CIA should have automatically actuated when CNMT pressure reached 5.0 psig, the signal is failed, CIA actions will not reduce CNMT pressure. The current CNMT pressures indicate that CIA and CIB have failed to actuate.
B. Incorrect. CIB should have actuated by this point when 214 channels reach 11.1 psig. Plausible if the incorrect setpoint is used.
C. Incorrect. CIA will not reduce CNMT pressure after it is manually actuated.
D. Correct. Candidate must recognize that the initial failure was on CH-IV. The stem states that the required actions of the IF procedure have been completed. Candidate must have knowledge of the procedure used to mitigate the first failure, in this case the channel is bypassed by placing the CIB bistable to the TRIP position, the CIB signal is NOT generated. This changes the CIB actuation logic from 2/4 to 213 of the remaining channels. The CIB setpoint of 11.1 psig has been reached so the CIB signal has failed to automatically actuate and manual action is required to initiate containment spray to reduce CNMT pressure. The emergency procedures contain steps to verify that the CIA and CIB signals are actuated, if not, then the operator is directed to actuate both trains, also the Conduct of Operation procedure has direction to actuate ESF signals that have failed to automatically actuate. The CIB signal will initiate a start of the Quench Spray pumps and lower CNMT pressure.
Sys# System Category KA Statement 103 Containment A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Phase A and B isolation System containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
KIA# A2.03 KIA Importance 3.5* Exam Level RO References provided to Candidate None Technical
References:
10M-1.4.1F pages 5 & 7, UFSAR Fig. 7.2-1 Sh 8 NOP-OP-1 002 Page 17 Question Source: Bank- Modified Vision# 88140 Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 43.51 45.3 I 45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
56. The plant is at 75% power with all systems in normal alignment for this power level EXCEPT Turb 1st Stage Press Sel SW [HSS-1 PM-446] is selected to the 446 position.
For these plant conditions, which of the following failures will cause the Rod Control System to automatically insert the rods?
A. 1 B First Stage Pressure Transmitter [PT-1 MS-447] fails LOW.
B. Loop 3 Protection T avg [TI-1 RC-4320] fails LOW.
C. Power Range Channel [N-44] fails HIGH.
D. Loop 1 Narrow Range Cold leg [TRB-1RC-412C] fails HIGH.
Answer: C Explanation/Justification:
A. Incorrect. PT 447 failure will cause steam dumps to arm on a load rejection signal, which would drive rods for an actual load rejection. However, there is not AT present so the steam dumps will not actuate. CN-1 00 will open which will cause power to rise which could cause a rod insertion but the power rise is caused by cold Tavg which is a stronger outward signal in rod control. Therefore, the rods will not automatically insert. If the selector SW was in 447 position this would be correct.
B. Incorrect. The Tavg signal to rod control is auctioneered.
C. Correct. Requires knowledge of the Nuclear Power Anticipatory circuit as it is referenced to the Turbine Power circuit. Understanding the purpose of this instrument input to Rod Control is required to select this answer.
D. Incorrect. There is only one Tc transmitter, per loop, that is used for developing the Tavg signal and it is not auctioneered in developing the Tavg signal. This could lead to the belief that this failure will lead to rod insertion. However, the output of the developed Tavg signal is auctioneered with the other loops before inputting to the rod control circuit.
Sys# System Category KA Statement 001 Control Rod K6 Knowledge of the effect of a loss or malfunction on the following Purpose and operation of sensors feeding into the Drive System CRDS components: CRDS KIA# K6.02 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical
References:
10M-53C.4.1.2.1C step 3, OM Fig. 1-51 and 1-52 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7/45.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
57. The plant is at 50% power with all systems in normal alignment for this power level.
Loop Protection T avgs are as follows:
Loop 1 Protection T avg [TI-1 RC-412D] is 562 °F
Loop 2 Protection T avg [TI-1 RC-422D] is 562 OF
Loop 3 Protection Tavg [TI-1 RC-432D] is 561 oF THEN Loop 1 Protection T avg [TI-1 RC-412D] fails HIGH What impact will this failure have on annunciator A4-46, Tavg Deviation from Tref AND how will Chg Flow To Regen Hx Inlet Control Vlv [FCV-1CH-122] respond to this failure?
Annunciator A4-46, Tavg Deviation from TrefWill (1) _ __
Chg Flow To Regen Hx Inlet Control Vlv [FCV-1CH-122] will _ _ _ (2)_ _ _ to control PRZR level at the programmed value.
A. (1) remain dark (2) remain as is B. (1) remain dark (2) open further C. (1) alarm (2) remain as is D. (1) alarm (2) open further Answer: A Explanation/Justification:
A. Correct. The alarm and input to PRZR level program are both downstream of the median selector and is not be impacted by this channel failure.
B. Incorrect. Alarm status is correct. Wrong FCV-1 CH-122 response. Since the PRZR level program is developed downstream of the median selector.
C. Incorrect. FCV-1CH-122 response is correct. A4-42 will alarm for loop Tavg deviation but not A4-46.
D. Incorrect. Wrong FCV-1CH-122 response. Since the PRZR level program is developed downstream of the median selector. A4-42 will alarm for loop Tavg deviation but not A4-46.
Sys # System Category KA Statement 011 Pressurizer Level Control A1 Ability to predict and/or monitor changes in parameters (to prevent exceeding T-ave System (PZR LCS) design limits) associated with operating the PZR LCS controls including:
KIA# A1.04 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical
References:
10M-6.4.1F page 12 and page 40 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5/45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
58. The plant is at 80% power with all systems in normal alignment for this power level EXCEPT the IPC is out of service for computer group data updates.
lAW Tech. Spec. 3.1.7.1, Unit 1 Rod Position Indication, which of the following conditions will REQUIRE action within 15 minutes to "Verify the affected rod position by measuring the RPI channel primary voltage"?
Bank D Rod Bank D Group Bank D Group 1 Bank D Group 2 Motion History Demand Position Rod H2 Indicated Rod K6 Indicated Position on VB Position on VB RPI RPI Condition 1 Rods have 212 195 195 moved in the last 30 minutes Condition 2 Rods have 187 185 205 moved in the last 45 minutes Condition 3 Rods have NOT 210 215 225 moved in the last 90 minutes Condition 4 Rods have NOT 200 190 210 moved in the last 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months A. Condition 1 B. Condition 2 C. Condition 3 D. Condition 4 Answer: C Explanation/Justification:
A. Incorrect. Both the group 1 and 2 rods are not indicating W/1 12 steps of demand. However, since the rods were moved in the last 30 minutes there is no requirement to be W/1 12 steps.
B. Incorrect. The group 2 rod is not indicating W/1 12 steps of demand. However, since the rods were moved in the last 30 minutes there is no requirement to be W/1 12 steps. The rods are more than 12 steps from each other but W/1 12 steps of demand which could lead to confusion over the requirement.
C. Correct. A one hour stabilization period is allowed following rod motion until the RPI voltages are required to be verified. RO only knowledge is required for the less than one hour action statement, since the note is below the line, ROs must know this applies to the one hour action.
D. Incorrect. The rods are more than 12 steps from each other but W/1 12 steps of demand which could lead to confusion over the requirement.
Sys # System Category KA Statement 014 Rod Position Indication System (RPIS) Generic Ability to determine operability and/or availability of safety related equipment.
KIA# 2.2.37 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical
References:
TS pages 3.1.7.1-1 and 3.1.7.1-4 Question Source: New Question Cognitive Level: High -Analysis 1 0 CFR Part 55 Content: (CFR: 41.7 /43.5/45.12)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
59. The plant is at 100% power.
One of the two control power fuses on the Power Range N41 drawer fails OPEN (blown)
(1) How will this failure impact the N41 drawer meter reading?
AND (2) What will be the impact in the Reactor Protection System (RPS)?
The N41 drawer meter reading will (1 )_ and Bench Board B Status light panel 176 A 16, Power Range Hi Set Pt light will _ _(2)_ __
A. (1) drop to 0%
(2) NOT be LIT B. (1) drop to 0%
(2) be LIT C. (1) remain at 100%
(2) NOT be LIT D. (1) remain at 100%
(2) be LIT Answer: D Explanation/Justification:
A. Incorrect. Wrong meter reading. Control power will not impact meter reading since it is developed from instrument power. Wrong status light status. Loss of control power will cause the hi flux trip bistables in RPS to trip.
B. Incorrect. Wrong meter reading. Control power will not impact meter reading since it is developed from instrument power. Correct status light status.
C. Incorrect. Correct meter reading. Wrong status light status. Loss of control power will cause the hi flux trip bistables in RPS to trip.
D. Correct. In accordance with AOP-1.2.1 C, the control power fuses are pulled to trip the hi flux trip bistables in RPS. Control power will not impact meter reading since it is developed from instrument power. The status light for hi flux on the drawer will not be lit since it is powered from this circuit. However, the SSPS input will see the hi flux condition since it is de-energize to go.
Sys# System Category KA Statement 015 Nuclear Instrumentation K3 Knowledge of the effect that a loss or malfunction of the NIS will have on the following: RPS System (NIS)
KIA# K3.01 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
lAW 1OM-53C.4.1.2.1 C step 9 Question Source: New Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.7 I 45.6)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
60. The plant is at 100% power.
The average of the 5 hottest core exit thermocouples, as displayed on the Inadequate Core Cooling Monitor (ICCM), is 610 oF and stable.
The heater controlling the incore thermocouple reference junction box MALFUNCTIONS AND reference junction box temperature RAISES 20 °F.
Based on this malfunction, what will be the ICCM display for the average of the 5 hottest core exit thermocouples?
A. Four purple asterisks B. 590 OF D. 630 °F Answer: C Explanation/Justification:
A. Incorrect. This is the SPDS readout for bad TIC data.
B. Incorrect. Without the automatic compensation feature of ICCM, this would be the reading for reference temperature rising.
C. Correct. At BVPS the RTD temperature compensates for changes in junction box temperature.
D. Incorrect. If candidates believes as reference temp increase, so wiiiiCCM reading, they will choose this answer.
Sys# System Category KA Statement 016 Non-Nuclear Instrumentation A3 Ability to monitor automatic operation of the Relationship between meter readings and actual System (NNIS) NNIS, including: parameter value KIA# A3.02 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical
References:
Fundamental knowledge of TIC and reference junction boxes. 1OM-3.1.C Page 8 of 8, 3SQS-3.1 Rev 6 pg 11-13 & slide 36 Question Source: New Question Cognitive Level: High - Comprehension 1 0 CFR Part 55 Content: (CFR: 41.7 I 45.5)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
61. The plant is in Mode 6. A fuel assembly is being lowered into the core.
IF the fuel assembly "BINDS" against another fuel assembly, downward motion of the hoist will be automatically stopped to prevent fuel assembly damage.
What manipulator crane interlock provides this protection?
A. Overload B. Underload C. Tube Down D. Bridge-Trolley Hoist Answer: B Explanation/Justification:
A. Incorrect. Overload will stop UPWARD motion if an assembly is binding while moving upward.
B. Correct. In accordance with LP 3SQS-6.13 slide 99 and 1RP-3.3.
C. Incorrect. Tube down interlock will stop hoist downward motion when the hoist is all the way down.
D. Incorrect. Bridge-Trolley Hoist interlock will only allow motion/movement in one direction at a time.
Sys # System Category KA Statement 034 Fuel Handling Equipment K4 Knowledge of design feature(s) and/or interlock(s) which Fuel protection from binding and System (FHES) provide for the following: dropping KIA# K4.01 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical
References:
LP 3SQS-6.13 Rev. 5 pg 12 & Slide 98, 1RP-3.3, Rev. 8 pg 7 Question Source: Bank- Vision # 82020 (NRC Exam BVPS Unit 2- 2LOT7 Q20)
Question Cognitive Level: Low_ Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
62. The plant is in Mode 3 with all systems in normal alignment for this mode.
The Condenser Steam Dump system is in Auto controlling Tavg at 547 oF in the Steam Pressure Mode
A non-isolable piping failure in the Instrument Air System has occurred Which of the following describes the response of RCS T avg and the action REQUIRED per AOP 1.34.1 Loss of Station Instrument Air, to control RCS temperature?
RCS Tavg will (1) , and temperature control will be maintained by operating the Atmospheric Steam Dumps [PCV-1MS-101A, B, C) (2)_ __
A. (1) rise (2) locally B. (1) rise (2) from the Control Room C. (1) lower (2) locally D. (1) lower (2) from the Control Room Answer: A Explanation/Justification:
A. Correct. The Condenser Steam dumps fail closed on a low of air, RCS Tavg will rise. The Atmospheric Steam Dumps also fail closed on a loss of air, but can only be operated locally by procedure.
B. Incorrect. The Condenser Steam dumps fail closed on a low of air, RCS Tavg will rise. Atmospherics cannot be operated from the Control Room since they also will fail closed on the loss of Air.
C. Incorrect. The Condenser Steam dumps fail closed on a loss of air, this will cause Tavg to rise. Plausible if the candidate assumes the steam dumps fail open. The Atmospherics can only be operated locally without air.
D. Incorrect. The Condenser Steam dumps fail closed on a loss of air, this will cause Tavg to rise. Plausible if the candidate assumes the steam dumps fail open. The Atmospherics can only be operated locally without air.
Sys# System Category KA Statement 041 Steam Dump System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; Loss of lAS (SDS)fTurbine Bypass and (b) based on those predictions use procedures to correct, control, or mitigate the Control consequences of those malfunctions or operations:
KIA# A2.03 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical
References:
1OM-53C.4.1.34.1 Step 12 Pg 7 & 17 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 43.5 I 45.3 I 45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
63. The plant is at 75% power with all systems in normal alignment for this power level.
The following indications are received:
Annunciator A7-1, Condensate Pump Auto Start-Stop
Annunciator A7-5, Condensate Pump Disch Press Low
Annunciator A7-6, Steam Generator Feed Pump Suet Press Low
Steam Generator Feed Pump suction pressure is 275 psig and stable
Condensate Pumps 1CN-P-1A and 1B, Bright WHITE indicating lights are lit
Attempts to restart the Condensate pumps are unsuccessful
1A Main Feedwater Pump [1 FW-P-1A], RED indicating lights are lit
1B Main Feedwater Pump [1 FW-P-1 B], Bright WHITE indicating lights are lit (1) What impact will these conditions have on Heater Bypass Valve [TV-1CN-100]?
AND (2) What action will be REQUIRED based upon these conditions?
A. (1) Heater Bypass Valve [TV-1CN-100] will OPEN.
(2) Trip the Reactor and enter Procedure E-0, Reactor Trip or Safety Injection.
B. (1) Heater Bypass Valve [TV-1CN-100] will remain CLOSED.
(2) Trip the Reactor and enter Procedure E-0, Reactor Trip or Safety Injection.
C. (1) Heater Bypass Valve [TV-1CN-100] will OPEN.
(2) Reduce power to s 50%.
D. (1) Heater Bypass Valve [TV-1CN-100] will remain CLOSED.
(2) Reduce power to s 50%.
Answer: B Explanation/Justification:
A. Incorrect. The Heater Bypass valve does not open on feed flow changes, the valve will remain closed. The correct response for Condensate and FW pump trip above 70% power is to trip the reactor.
B. Correct. The Heater Bypass valve does remain closed, it opens on a large load rejection not a loss of feed flow, however bypassing the heater string is a plausible response to a loss of condensate and feedwater flow event. The correct response is to trip the reactor since power is above 70% and the conditions indicate that the pumps have tripped. The AOP requires two FW pumps to be operating above 70%.
C. Incorrect. The Heater Bypass valve does not open on feed flow changes, the valve will remain closed. The power reduction would be applicable if the plant was below 70% power when the pump tripped.
D. Incorrect. The Heater Bypass valve does remain closed, it opens on a large load rejection not a loss of feed flow, however bypassing the heater string is a plausible response to a loss of condensate and feedwater flow event. The power reduction would be applicable if the plant was below 70% power when the pump tripped.
Sys# System Category KA Statement 056 Condensate A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Loss of condensate System Condensate System; and (b) based on those predictions, use procedures to correct, control, pumps or mitigate the consequences of those malfunctions or operations:
KIA# A2.04 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical
References:
1OM-53C.4.1.24.1 Page 2, 1OM-22.4.AAC pg 1, 1OM-22.1.0 pg 4 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.5 I 43.51 45.3 I 45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
64. The plant is at 100% power.
Annunciator A6-65, Cooling Tower Slowdown Flow Low 10.000 GPM is in alarm Per the Alarm Response Procedure, which of the following pump operations will cause Cooling Tower Slowdown flow to DECREASE?
1. Reactor Plant River Water Pump [1WR-P-1A] Start
2. Reactor Plant River Water Pump [1WR-P-1A] Stop
3. Turbine Plant River Water Pump [1WR-P-6A] Start
4. Turbine Plant River Water Pump [1WR-P-6A] Stop
5. Cooling Tower Pump [1CT-P-1A] Start
6. Cooling Tower Pump [1CT-P-1A] Stop A. 1,2,3,4,5,6 B. 2, 4, 5, 6 ONLY C. 1, 3, 5 ONLY D. 2, 3, 6 ONLY Answer: B Explanation/Justification:
A. Incorrect. RP and TP RW pump starts will increase overall CW flow which increases blowdown flow. The RW pump stops will reduce blowdown flow, as will the starting or stopping of the Cooling Tower pumps. It will reduce CW basin causing blowdown flow to reduce.
B. Correct. Stopping the RW pumps will reduce blowdown flow by reducing overall CW flow. Cooling Tower pump starts or stops will also reduce blowdown flow since the system is open to atmosphere at the cooling tower and pump operation will affect basin level, reducing the amount of blowdown flow.
C. Incorrect. River Water pump starts will increase overall CW flow which increases blowdown flow. Cooling Tower pump start is correct. It will reduce CW basin causing blowdown flow to reduce.
D. Incorrect. TP River Water pump start will increase overall CW flow which increases blowdown flow. RP RW pump stop and CT pump stop will reduce blowdown flow.
Sys # System Category KA Statement 075 Circulating Water System A4 Ability to manually operate and/or monitor in the control room: Emergency/essential SWS pumps KIA# A4.01 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical
References:
1OM-31.4.AAA pg 2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5 to 45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
65. The plant is at 100% power.
TWO Heat Actuating devices (HD-1 FP-6-1 & HD-1 FP-6-2) on Unit Station Transformer 1C fail high and generate the following indication:
Annunciator AB-60, Unit Sta Serv Trans 1C FIRE alarms Assuming all systems function as designed, what impact, if any, will these conditions have on the operation of USST 1C?
The USST 1C Deluge valve will (1)_ _ and 4KV Busses 1A and 1B
___(2) transfer to Offsite Power.
A. (1) OPEN (2) will B. (1) not OPEN (2) will C. (1) OPEN (2) will not D. (1) not OPEN (2) will not Answer: C Explanation/Justification:
A. Incorrect. The undervoltage will not occur due to only the transformer deluge system actuation.
B. Incorrect. The differential will not occur due to only the transformer deluge system actuation, and is generated from the Main Generator circuit.
C. Correct. The HADs actuate Deluge only, the Over current, Differential and Undervoltage signals are generated from other devices which transfer the busses to offsite. The HADs actuate the Deluge system, however this only sprays the transformer with water, no other actuations will occur.
D. Incorrect. The overcurrent will not occur due to only the transformer deluge system actuation, only active on Backfeed.
Sys# System Category KA Statement 086 Fire Protection K5 Knowledge of the operational implication of the following Effect of water spray on electrical components System (FPS) concepts as they apply to the Fire Protection System:
KIA# K5.03 KIA Importance 3.1 Exam Level RO References provided to Candidate Technical
References:
1OM-33.4.AAE pg 2, 1OM-33.1.0 Pg 3, None 10ST.33.10.1 pg 9 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.5/45.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
66. The plant is at 100% power.
lAW NOP-OP-1 002, Conduct of Operations, which of the following lists the UNIT 1 NORMAL shift staffing requirements and the MAXIMUM time allowed to fill a position due to an UNEXPECTED absence?
Senior Reactor #Reactor Operator(s) Maximum Time to Fill Operator(s) the Position Condition 1 1 1 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Condition 2 2 2 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Condition 3 1 2 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Condition 4 2 2 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months A. Condition 1 B. Condition 2 C. Condition 3 D. Condition 4 Answer: D.
Explanation/Justification:
A. Incorrect. Per NOP, must have 2 SROs (not 1 SRO), 2 ROs, and 1 hr limit is not the maximum allowed time.
B. Incorrect. Per NOP, 2 SROs and 2 ROs is correct, however, 1 hr limit is not the maximum allowed time.
C. Incorrect. Per NOP, must have 2 SROs (not 1 SRO), 2ROs and 2 hr limit is allowed to fill a vacancy.
D. Correct. Per NOP in Modes 1-4, 2 SROs, 2R0s, and 2 hr limit is allowed to fill a vacancy.
Sys # System Category KA Statement N/A N/A Generic Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
KJA# 2.1.5 KJA Importance 2.9* Exam Level RO References provided to Candidate None Technical
References:
NOP-OP-1 002 Att 4 pg 97, Section 4.1.13 pg 21 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.12)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
67. The plant is performing a startup in accordance with 1OM-52.4.A, Raising Power From 5% to Full Load Operation.
The local operator has requested approval to omit and mark N/A the following step due to noise level concerns:
d. +Check that governor valves open freely, with no hesitation, locally AND indicated at Turbine E-H Control Panel, Benchboard-Section C.
Initial lAW the guidance in the 1OM-52.4.A and NOP-LP-2601, Procedure Use and Adherence can this step be omitted?
The step may_________________________________________________
A. be omitted as N/A, provided two SROs concur and document the decision on the procedure B. be omitted as N/A, by the procedure performer with supervisor approval C. NOT be omitted since the step does not have a designated option as N/A D. NOT be omitted since it is identified as a step that must be performed Answer: D Explanation/Justification:
A. Incorrect. The step is marked with a + which indicates that it cannot be marked N/A, two SROs are required to approve N/A of steps that do not have a conditional statement that allows for N/A.
B. Incorrect. The step is marked with a + which indicates that it cannot be marked N/A, for Non-Operations procedures, the step must be evaluated for plant impact and approved by the supervisor.
C. Incorrect. The step does not have to have a conditional statement that allows for N/A.
D. Correct. Those procedural steps NOT marked by a diamond ( +) sign may be omitted or performed out of sequence at the discretion of the Shift Manager, provided a brief written explanation of justification for the change is given. The steps marked by a(+) sign can NOT be omitted, but may be started out of sequence at the discretion of the Shift Manager. The step is marked with a + which indicates that it must be performed, however it may be started out of sequence.
Sys # System Category KA Statement N/A N/A Generic Ability to interpret and execute procedure steps.
KIA# 2.1.20 KIA Importance 4.6 Exam Level RO References provided to Candidate None Technical
References:
10M-52.4.A pg 29, NOP-LP-2601 pg 10 & 11, 1/20M-48.2.C pg 10 Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: (CFR: 41.10 /43.5/45.12)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
68. Per the Operating Manual, which of the following is a function of the Chemical and Volume Control System?
A. Collects radioactive reactor coolant and separates it into stripped liquid and gas, evaporates the liquid to reclaim boric acid and primary grade water.
B. Control water chemistry conditions, activity level, makeup water and degasification, of the RCS.
C. Collects potentially radioactive fluids and gases from various systems and discharges them to the Gaseous, Liquid or Boron Recovery systems.
D. Receive, process and dispose of aerated liquid waste.
Answer: B Explanation/Justification:
A. Incorrect. This is a function of the Boron Recovery.
B. Correct. RO Knowledge for system function from for the CVCS system. Distractors are systems that are similar in the processing of liquid and gases.
C. Incorrect. This is a function of the Reactor Plant Vents and Drains System.
D. Incorrect. This is a function of the Liquid Waste System.
Sys # System Category KA Statement N/A N/A Generic Knowledge of system purpose and/or function.
KIA# 2.1.27 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
1OM 7.1.A (#4 on pg 1)
Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
69. The plant is in Mode 3 with all systems in normal alignment for this mode.
The following activities are in progress:
Maintenance is required on Emergency Diesel Generator (EDG #1) EE-EG-1
The scope of the work is to replace the Air Start Motors
The work is scheduled to complete within the next two shifts Which of the following Technical Specifications are REQUIRED to be entered for these conditions?
1. TS 3.8.1 AC Sources - Operating
2. TS 3.8.2 AC Sources- Shutdown
3. TS 3.8.7 Inverters- Operating
4. TS 3.8.8 Inverters - Shutdown A. 1 Only B. 1 & 3 Only C. 2 Only D. 2 & 4 Only Answer: A Explanation/Justification:
A. Correct. In Mode 3 two EDGs are required to be OPERABLE.
B. Incorrect. The plant is not in Mode 5/6 when this TS is applicable.
C. Incorrect. The plant is in Mode 3, the inverters are required to be OPERABLE, however they are energized from their energized sources so the TS is not required to be entered.
D. Incorrect. The plant is not in Mode 5/6 when this TS is applicable, and the Inverters are still energized.
Sys# System Category KA Statement N/A N/A Generic Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
KIA# 2.2.36 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical
References:
TS 3.8.1 Question Source: New Question Cognitive Level: High - Comprehension 1 0 CFR Part 55 Content: (CFR: 41.10 /43.2/45.13)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
70. The plant is in Mode 5 in preparation for a Refueling Outage.
Which of the following components is considered AVAILABLE per NOP-OP-1 005, Shutdown Defense in Depth?
A. Low Head Safety Injection Pump 1SI-P-1A, with a Red Danger Tag clearance on the motor supply breaker.
B. Charging Pump 1CH-P-1A, with the motor supply breaker removed from the bus cubicle.
C. Emergency Diesel Generator 1EE-EG-1, with a flange on the fuel oil piping not installed.
D. Auxiliary Feedwater Pump 1FW-P-3A, with the manual suction isolation valve closed.
Answer: D Explanation/Justification:
A. Incorrect. SSCs that are red danger tagged are not considered Available.
B. Incorrect. SSCs that require installing breakers in cubicles are not considered Available.
C. Incorrect. SSCs that require reconnecting pipe are not considered Available.
D. Correct. Meets the definition of Available, opening a manual valve is a reasonable action. Per the definition of Available- The status of a system, structure, or component (SSC) that is in service or can be placed in service within a reasonable period of time by immediate manual or automatic actuation. Minimal actions taken by operators to close or rack in breakers, align valves, and start pumps are allowed under the terms of a reasonable period of time. Reconnecting pipe, installing breakers in cubicles, or installing valves is not considered reasonable for a system to be Available. Meets the definition of Available, opening a manual valve is a reasonable action.
Sys # System Category KA Statement N/A N/A Generic Ability to determine operability and/or availability of safety related equipment.
KIA# 2.2.37 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical
References:
NOP-OP-1005 pg 3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.7 /43.5/45.12)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
71. The plant is at 100% power.
The following conditions exist:
The power supply breaker to PRZR HTRS 2C Controlling Group TRIPS on overcurrent
Pressurizer water level is 90%
Spent Fuel Pool level is 22 feet above the storage racks
RCS pressure is 2750 psig
The Quadrant Power Tilt Ratio is 1.01 Based on these conditions, which of the below listed Tech. Spec. LCOs are NOT being Met?
1. LCO 3.4.9, Pressurizer
2. LCO 3.7.15, Fuel Storage Pool Water Level
3. LCO 2.1.2, Reactor Coolant System Pressure Safety Limit
4. LCO 3.2.4, Quadrant Power Tilt Ratio A. 1 & 2 Only B. 2 & 3 Only C. 3 &4 Only D. 1 &4 Only Answer: B Explanation/Justification:
A. Incorrect. LCO 3.4.9 is met. LCO 3. 7.15 is not met. (LCO 3.4.9- PRZR level s 92 and 2 sets of PRZR heaters with each set~ 150 kw and powered from an emergency bus). PRZR HTRS 2C are powered from normal power. All other heaters are available.
B. Correct. Both are not met. (LCO 3.7.15- FP level shall be~ 23ft.), (LCO 2.1.2- RCS Pressure shall be s 2735 psig).
C. Incorrect. LCO 3.2.4 is met. LCO 2.1.2 is not met. (LCO 3.2.4- QPTR shall be s 1.02)
D. Incorrect. Both LCO 3.4.9 and LCO 3.2.4 are being met.
Sys # System Category KA Statement N/A N/A Generic Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
KIA# 2.2.42 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical
References:
TS pages 3.7.15-1 and 2.0-1 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: (CFR: 41.7 I 41.10 I 43.2 I 43.3 I 45.3)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
72. An operator is required to enter the Blender Cubicle to perform several valve manipulations.
The following radiological conditions exist for Blender Cubicle:
General area dose rate levels range from 15 - 40 mr/Hr
The measurements taken in the cubicle are:
Point 1 is 150 mr/Hr at 30 em
Point 2 is 625 mr/Hr at 30 em
Point 3 is 1225 mr/Hr at 30 em lAW NOP-OP-4101, Access Controls for Radiologically Controlled Areas what is the radiological posting REQUIRED for the Blender cubicle?
A. Radiation Area B. High Radiation Area C. Locked High Radiation Area D. Very High Radiation Area Answer: C Explanation/Justification:
A. Incorrect. Radiation areas are > 5mr/hr at 30 em.
B. Incorrect. High Radiation area is defined as .::,1 00 mr/hr at 30 em. but< 1000 mr/hr.
C. Correct. LHRA- 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.
D. Incorrect. Very high radiation levels are > 500 rads/hr at 1 meter.
Sys # System Category KA Statement N/A N/A Generic Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
KIA# 2.3.12 KIA Importance 3.2 Exam Level RO References provided to Candidate None Technical
References:
NOP-OP-41 01 page 5 item 3.6 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: (CFR: 41.12/45.9/45.10)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
73. lAW the EOP background documents, which of the following procedures, when implemented, could result in high radiation in the safeguards and auxiliary buildings?
A. FR-S.2, Response To Loss Of Core Shutdown B. E-3, Steam Generator Tube Rupture C. ECA-0.0, Loss of All Emergency 4KV AC Power D. ES-1.3, Transfer To Cold Leg Recirculation Answer: D Explanation/Justification:
A. Incorrect. Loss of core SID may lead to the reactor going critical possible high neutron radiation levels in containment but not in the safeguards and auxiliary buildings.
B. Incorrect. SGTR could lead to high radiation levels along the steam lines (which are in the safeguards building) but this does not include the auxiliary building and is not specifically addressed in the backgrounds.
C. Incorrect. With no emergency 4KV power available, it is possible to believe that high radiation levels will result.
D. Correct. During a LOCA, water from the RCS with higher than normal activity will be transferred from the break in the RCS to the containment sump. When the plant is switched over to the recirculation mode, these higher activity levels may cause higher than normal radiation in the safeguards and auxiliary buildings. EOP backgrounds specifically address this condition since the resultant radiation levels could be extremely high.
Sys# System Category KA Statement N/A N/A Generic Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
KIA# 2.3.14 KIA Importance 3.4 Exam Level RO References provided to Candidate Technical
References:
10M-53B.4.ES-1.3 step 1 Caution #3 None Question Source: New Question Cognitive Level: Low - Fundamental 10 CFR Part 55 Content: (CFR: 41.12/43.4/45.10)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
74. Which of the following conditions will REQUIRE entry into E-0, Reactor Trip or Safety Injection?
Rx Power RCS Loop 1 RCS Loop 2 RCS Loop 3 Turbine Status level Flows Flows Flows Condition 1 5% 82% 92% 86% Tripped Condition 2 18% 95% 80% 93% Latched Condition 3 35% 100% 88% 89% Latched Condition 4 48% 93% 94% 96% Tripped A. Condition 1 B. Condition 2 C. Condition 3 D. Condition 4 Answer: C Explanation/Justification:
A. Incorrect. 2/3 loop flows are below trip setpoint (90.2%), but power is below P-7 (1 0%). When <P7 3/3 loop flows below trip setpoint would be required to enter E-0.
B. Incorrect. 1/3 loop flows is below trip setpoint (90.2%), and power is above P-7 (10%), but below P-8 (30%). Between P-7 and P-8 there would have to be 2/3 loop flows low to enter E-0.
C. Correct. RO must analyze each condition to determine if a Rx trip setpoint has been exceeded. Condition 3 has 2/3 loop flows below trip setpoint (90.2%) with power above P-8 (30%) setpoint. This condition requires entry into E-0.
D. Incorrect. All flows are above trip setpoint (90.2%), but the turbine is tripped. However power is below P-9 (49%). If power was >P9 and the turbine tripped, E-0 would be entered.
Sys # System Category KA Statement N/A N/A Generic Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.
KIA# 2.4.2 KIA Importance 4.5 Exam Level RO References provided to Candidate None Technical
References:
UFSAR fig. 7.2-1 sh 5, 10M-1.2.B page 5 Question Source: New Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: (CFR: 41.7/45.7 /45.8)
Objective:
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
75. The plant is at 100% power.
A Large Break LOCA occurs, the Reactor trips and Sl is actuated
All systems function as designed
The crew enters E-0, Reactor Trip or Safety Injection and transitions to E-1, Loss Of Reactor or Secondary Coolant While in E-1, the following VALID annunciator and associated indication are received:
A6-81, Cooling Tower Pump Auto Stop
Cooling Tower 1A Pump [1CT-P-1A] has a BRIGHT WHITE light lit and the RED light NOT Lit lAW the guidance in NORM-OP-1002, Conduct Of Operations Handbook, how is the RO/ATC REQUIRED to respond to this alarm?
A. Announce the unexpected alarm to the command SRO; perform the ARP actions in conjunction with the execution of the EOPs, without interfering with EOP execution.
B. There is no requirement to announce the unexpected alarm; perform the ARP actions in conjunction with the execution of the EOPs, without interfering with EOP execution.
C. There is no requirement to announce the unexpected alarm; leave the alarm flashing so long as other alarms are not masked.
D. Announce the unexpected alarm by performing a crew update; leave the alarm flashing so long as other alarms are not masked.
Answer: 8 Explanation/Justification:
A. Incorrect. Announcing is required for unexpected alarms when not in transient response. Correct actions.
B. Correct. In order to answer the question, the RO must know when transient response guidelines are implemented and the recognize the CT pump trip as not significant then apply the rules as stated in the NORM.
C. Incorrect. Correct announcement requirement. Wrong action, this is the action required for nuisance alarms.
D. Incorrect. Wrong announcement requirement, announcing is required for unexpected alarms when not in transient response. Wrong action, this is the action required for nuisance alarms.
Sys # System Category KA Statement N/A N/A Generic Knowledge of annunciator alarms, indications, or response procedures.
KIA# 2.4.31 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical
References:
NORM-OP-1002 page 75 151 paragraph and paragraph 4 on page 75 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: (CFR: 41.10 /45.3)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
76. The plant is at 100% power.
A Small break LOCA occurs and all systems function as designed
The Crew has transitioned to E-1, Loss of Reactor or Secondary Coolant and are currently at step 9, Check if Sl flow should be reduced The following plant conditions exist:
RCS pressure is 1185 psig and stable
All SG NR levels are 35% and stable
Core exit TICs are 452 oF and slowly dropping
All SG pressures are 950 psig and slowly dropping
CNMT pressure is 3.5 psig and stable
Total AFW flow is 500 gpm and stable
PRZR level is 18% and slowly rising
Annunciator A8-109, 4160V Bus 1AE Undervoltage has illuminated
4160V Bus 1AE Voltmeter indicates 0 Volts Based on these conditions, what procedural transitions are REQUIRED?
Transition to______________________________
A. ES-1.1, Sl Termination and perform AOP 1.36.2, Loss of 4KV Emergency Bus in PARALLEL B. ES-1.1, Sl Termination and perform AOP 1.36.1, Loss of All AC Power When Shutdown in PARALLEL C. AOP 1.36.2, Loss of 4KV Emergency Bus, THEN transition to ES-1.2, Post LOCA Cooldown and Depressurization D. AOP 1.36.1, Loss of All AC Power When Shutdown, THEN transition to ES-1.1, Sl Termination Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper page 8 Figure 2 last bullet. SRO is required to have knowledge of admin procedures that specify the hierarchy and coordination of Emergency and Abnormal procedures. SROs are required to understand procedure content and interpret the plant conditions that require transition to EOP sub-procedures.
A. Correct. SIS Termination criteria is met, however the 1AE bus is also de-energized, AOP 1.36.2 will attempt to restore power.
SRO ONLY The SRO is required to implement procedures and understand the importance hierarchy and Transient Response Guidelines.
B. Incorrect. SIS termination criteria is met. AOP 1.36.1 is not applicable until RHR is in service.
C. Incorrect. The EOP network is not exited to perform AOPs. ES-1.2 is plausible if the candidate does not realize that SIS termination criteria is met.
D. Incorrect. The EOP network is not exited to perform AOPs. AOP-1.36.1 is not applicable until RHR is in service. SIS termination criteria is met.
Sys # System Category KA Statement 000009 Small Break LOCN3 Generic Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
KIA# 2.4.8 KIA Importance 4.5 Exam Level SRO References provided to Candidate None Technical
References:
E-1, Step 9 pg 8, AOP-1.36.2, BVBP-OPS-0024, Section (4.14) pg 14 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
{SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
77. The plant is at 100% power.
Annunciator A6-37, Pri Comp Cool WTR Surge Tank Level High-Low is lit
CCR Surge Tk Level Control Vlv [LCV-1CC-100A] is opening in AUTO
Reactor Plant Component Cooling Surge Tank Level on [LI-1CC-100] is 24 inches and slowly lowering
Auxiliary Building South Sump level is slowly rising Which of the following could be the source of the system leakage and what actions are REQUIRED per AOP 1.15.1, Loss of Primary Component Cooling Water, if CCR Surge Tank level continues to lower?
___(1)_ _ _ is leaking and when CCR Surge Tank level is off scale low, the SRO will direct the crew to (2) _ __
A. (1) Degassifier Recirculation Pump (1BR-P-7A) Relief Valve [RV-1CC-147A]
(2) Trip the CCR pumps, then Trip the Reactor, enter E-0 and perform IOAs B. (1) Fuel Pool Heat Exchanger (1FC-E-1A) Relief Valve [RV-1CC-111A]
(2) Trip the CCR pumps, then Trip the Reactor, enter E-0 and perform IOAs C. ( 1) Degassifier Recirculation Pump ( 1BR-P-7A) Relief Valve [RV-1 CC-14 7A]
(2) Trip the Reactor, enter E-0 and perform IOAs, then trip the RCPs D. (1) Fuel Pool Heat Exchanger (1 FC-E-1A) Relief Valve [RV-1CC-111A]
(2) Trip the Reactor, enter E-0 and perform IOAs, then trip the RCPs Answer: C Explanation/Justification: Meets the requirements for SRO only Section II.E first paragraph page 6. The first part is RO knowledge of the system flowpaths and sump alignments. The second part is SRO knowledge requires assessing plant conditions (Abnormal) and then selecting a section of the procedure to mitigate the event. The SRO must know detailed procedure steps in the AOP to then select the correct action.
A. Incorrect. This Relief Vlv discharges to the Aux Bldg South Sump, correct sump. Wrong sequence per AOP-1.15.1.
B. Incorrect. This Relief Vlv discharges to the Fuel Bldg Sump, incorrect sump. Wrong sequence per AOP-1.15.1.
C. Correct. This Relief Vlv discharges to the Aux Bldg South Sump. Per AOP 1.15.1 continuous action step 1 RNO, If Surge Tank level drops offscale low, Trip Reactor, then when IOAs are complete, stop RCPs, then stop CCR pumps.
SRO Only based on detailed knowledge of AOP.1.15.1 and determination of procedure actions.
D. Incorrect. This Relief Vlv discharges to the Fuel Bldg Sump, incorrect sump. Correct sequence per AOP-1.15.1.
Sys # System Category KA Statement 000026 Loss of Component AA2 Ability to determine and interpret the following as they apply to the The cause of possible CCW loss Cooling Water (CCW)/8 Loss of Component Cooling Water:
KIA# AA2.02 KIA Importance 3.6 Exam Level SRO References provided to Candidate None Technical
References:
AOP 1.15.1 pg 3 and 12, OM 9 Fig 9-4, Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
78. The plant was at 100% power when a Steam Generator Tube Rupture occurs in the 1C Steam Generator requiring a Reactor Trip and Safety Injection.
All systems respond as designed, EXCEPT all 4KV Normal Busses failed to transfer to Offsite Power
The crew transitioned to procedure E-3, Steam Generator Tube Rupture and then to ES-3.1, Post SGTR Cooldown Using Backfill Subsequently, Pressurizer Level has lowered to 5%, requiring the crew to manually align pumps and valves and transition to ECA-3.1, SGTR With Loss Of Reactor Coolant - Subcooled Recovery Desired.
RVLIS indication is 95% Full Range
The crew is implementing Step 20, Check If An RCP Should be Started
Offsite Power has been restored and all support conditions exist to start the RCPs Per step 20 of ECA-3.1, which RCP will be started and why is this pump selected?
A 1A RCP, to collapse the steam void in the reactor vessel head.
B. 1A RCP, to preclude an inadvertent criticality from occurring.
C. 1C RCP, to collapse the steam void in the reactor vessel head.
D. 1C RCP, to preclude an inadvertent criticality from occurring.
Answer: B Explanation/Justification: Meets the SRO only white paper Section II.E first paragraph pg 6. Requires assessing plant conditions and application of the procedure Caution in decision process for starting RCPs, detailed understanding of the EOP sub-procedure sequence is required. SRO must assess plant conditions and decide the correct course of action. Normal RCP start sequence is C-A-B, however due to dilution concerns post backfill, the C RCP will not be started. The vessel void collapse is normally a priority however since backfill was employed earlier in the sequence of events, dilution is the concern at this point in the procedure, and void collapsing not a consideration in selecting which RCP is started.
A. Incorrect. 1A RCP is correct. Wrong reason, the EOP background states that a dilution event may occur from a backfilled loop.
B. Correct. Normal RCP start sequence is C-A-B, however due to dilution concerns post backfill, the C RCP will not be started. The vessel void collapse is normally a priority however since backfill was employed earlier in the sequence of events, dilution is the concern at this point in the procedure, and void collapsing not a consideration in selecting which RCP is started.
C. Incorrect. 1C RCP is incorrect, The caution prior to Step 20 states not to start the pump in the ruptured loop following the Backfill recovery.
Wrong reason, the EOP background states that a dilution event may occur from a backfilled loop.
D. Incorrect. 1C RCP is incorrect, The caution prior to Step 20 states not to start the pump in the ruptured loop following the Backfill recovery.
Correct reason, the EOP background states that a dilution event may occur from a backfilled loop.
Sys # System Category KA Statement 000038 Steam Generator Tube Generic Ability to explain and apply system limits and Rupture (SGTR)/3 precautions.
KIA# 2.1.32 KIA Importance 4.0 Exam Level SRO References provided to Candidate None Technical
References:
E-3, ES-3.1, ECA-3.1 pg 17, ECA-3.1 Bases Step 20 pg 64 Question Source: New Question Cognitive Level: High - Comprehension 1 0 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
79. The plant is at 100% power.
A loss of all 4KV AC power occurs
The crew enters ECA-0.0, Loss of All Emergency 4KV AC Power
Station Instrument Air Header Pressure [PI-11A-1 06] is 0 psig
Diesel Driven Air Compressor [11A-C-4] has failed to start
All Steam Generator Levels are 25% Narrow Range and slowly decreasing
Total AFW flow is 230 GPM
Containment pressure is 5.5 psig Which of the following describes how the AFW Pump Recirculation valves will fail and the actions that will be taken to control Steam Generator Water level per Attachment 2-S, Monitoring AFW Pump Performance During Loss Of Station Instrument Air?
The AFW Recirc Valves will fail _ _ (1 )_ _ _ and the _ _ _(2) _ _ _ will be operated to control SG water level.
A. (1) OPEN (2) Turbine Driven AFW [1 FW-P-2] Pump B. (1) CLOSED (2) Turbine Driven AFW [1 FW-P-2] Pump C. (1) OPEN (2) Aux FeedwaterThrottle valves [MOV-1FW-151A-F] (locally)
D. (1) CLOSED (2) Aux Feedwater Throttle valves [MOV-1 FW-151A-F] (locally)
Answer: B Explanation/Justification: Meets the SRO only white paper Section II.E pg 7 first bullet. The first part of the determination for the recirc valve failure is RO knowledge, the actions to cycle the Turbine driven AFW pump and control at the higher level setpoint are detailed sub-procedure knowledge from the Attachment. Normal Steam Generator minimum level is 31%, the Attachment requires a higher control setpoint if a loss of air occurs.
A. Incorrect. The Recirc Vlvs fail Closed. The TDAFW pump is cycled to control level at 67% not 31%.
B. Correct. The Recirc Vlvs fail Closed on a loss of Air. Attachment 2-S manually starts and stops the Turbine Driven AFW pump to control SG level at 67% to preclude pump cycling whenever AFW flow is less than 250 GPM.
C. Incorrect. The Recirc Vlvs fail Closed. The AFW throttle valves are not used to control level to 67%. AFW throttle valves could be locally throttled to control level however, the procedure specifies using the TDAFW pump.
D. Incorrect. The Recirc Vlvs do fail Closed, however, Attachment 2-S manually starts and stops the Turbine Driven AFW Pump, level is controlled at 67% not 31%. AFW throttle valves could be locally throttled to control level however, the procedure specifies using the TDAFW pump.
Sys# System Category KA Statement 000055 Loss of Offsite and Onsite Power EA2 Ability to determine or interpret the following Existing valve positioning on a loss of instrument
{Station Blackout)/6 as they apply to a Station Blackout: air system KIA# EA2.01 KIA Importance 3.7 Exam Level SRO References provided to Candidate None Technical
References:
10M-53A.1.ECA-O.O step 22 pg 15, EOP Att 2-S Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43{b){5)
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
80. 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
1A Steam Generator Level is 10% Wide Range, pressure is 1000 psig and stable
1B Steam Generator Level is 19% Wide Range, pressure is 975 psig and stable
1C Steam Generator Level is 12% Wide Range, 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) 31% Narrow Range B. (1) Initiate RCS Feed and Bleed (2) 31% Narrow Range C. (1) Transition to E-2, Faulted Steam Generator Isolation (2) 28% Narrow Range D. (1) Initiate RCS Feed and Bleed (2) 28% Narrow Range Answer: D Explanation/Justification: Meets the SRO only white paper Section II.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.
A. Incorrect. This transition is possible since the 1C 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 1C 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 level.
D. Correct. Bleed and Feed criteria are met per continuous action step 3, which states WR level in at least 2 SGs <14%. Correct TS bases setpoint per SR 3.4.5.2 bases.
Sys# System Category KA Statement W/E05 Loss of Secondary EA2 Ability to determine and interpret the following as Facility conditions and selection of appropriate Heat Sink/4 they apply to the (Loss of Secondary Heat Sink) procedures during abnormal and emergency operations.
KIA# EA2.1 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical
References:
FR-H.1 pg 2, TS Bases 3.4.5.2 pg b 3.4.5-5 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY}
Beaver Valley Unit 1 NRC Written Exam (1LOT14) 81 . The plant is at 100% power.
A Steam Generator Tube Leak of 0.1 GPM has been identified in the 1A Steam Generator
The Reactor has tripped coincident with a Loss of Offsite Power
All equipment has operated as designed and the crew has entered procedure E-2~
Faulted Steam Generator Isolation The following conditions exist:
Pressurizer Level is 25% and slowly decreasing
RCS Pressure is 1425 psig and decreasing
Containment pressure is 13.0 psia and stable
No Radiation Monitors are in alarm
Total AFW flow is 420 GPM RCS LOOI;! TCold T Hot 1A 494 oF and decreasing 525 oF and decreasing 18 480 oF and decreasing 520 oF and decreasing 1C 502 oF and decreasing 529 oF and decreasing SG SG Level SG Pressure SG"A" 20% WR and stable 320 psig decreasing SG"B" 22% WR and stable 310 psig decreasing SG"C" 26% WR and stable 350 psig decreasing Which of the following procedures will be implemented to address the event in progress?
A. ES-0.2 Natural Circulation Cooldown 1
B. ES-1.1 Sl Termination I
C. ECA-3.1 SGTR With Loss of Reactor Coolant- Subcooled Recovery Desired I
D. ECA-2.1 Uncontrolled Depressurization Of All Steam Generators I
Answer: D Explanation/Justification: Meets the SRO only white paper Section II.E first paragraph pg 6. Requires SRO knowledge of the entry into sub procedures in the EOP network and diagnostic analysis of plant parameters. The SRO is required to assess plant conditions and select a procedure to mitigate the event A. Incorrect. The loss of all AC power will result in Natural Circulation, however conditions exist for a SIS, so this transition is incorrect.
B. Incorrect. SIS termination criteria is not met due to lowering RCS Pressure, however it is a potential transition from E-2.
C. Incorrect. The SGTL is not of a size that requires entry into the SGTR EOP sequence, potential transition to E-3 series procedure if diagnosis is incorrect. The SGTL was pre-event and NOT the reason for the transient based on no other indications that the leak has degraded to a rupture.
D. Correct. All Steam Generator pressures are dropping in an Uncontrolled manner, Tc temps lead the Tsat (Psat) temps indicating the steam generators are faulted.
Sys# System Category KA Statement W/E12 Uncontrolled Depressurization of all Steam Generic Ability to diagnose and recognize trends in an accurate and timely Generators/4 manner utilizing the appropriate control room reference material.
KIA# 2.4.47 KIA Importance 4.2 Exam Level SRO References provided to Candidate Steam Table Technical
References:
E-2 Step 3 Page 3 Question Source: New Question Cognitive Level: High - Comprehension 1 0 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
82. The plant is at 100% power.
The following alarms are received:
A4-88, Steam Gen N-16 Monitor Alert/High Trouble
A4-71, Radiation Monitoring High
A4-72, Radiation Monitoring High-High
The Condenser Air Ejector Vent Radiation Monitor [RM-1 SV-1 00] has both the Amber and Red lights lit (1) How will these indications affect the Condenser Air Ejector discharge?
(2) What is the Tech. Spec. bases for limiting Primary to Secondary Leakage?
The High-High Radiation Monitor signal will CLOSE Air Ejector Discharge to Gaseous Waste valve [TV-1SV-100B] (1) _ _
The Tech. Spec. bases for Primary to Secondary leakage limits ensures the accident dose contributions at the site boundary is limited to fractions of the (2) limits.
A. (1) and OPEN Air Ejector to Containment valve [TV-1 SV-1 OOA]
(2) 10CFR 50.67 B. (1) ONLY.
(2) 10CFR 50.67 C. (1) and OPEN Air Ejector to Containment valve [TV-1SV-100A]
(2) 10CFR 20 D. (1)0NLY (2) 10CFR 20 Answer: A Explanation/Justification: Meets the SRO only white paper Section II.B 3rd bullet pg 3. The first part is RO knowledge for the changes in system alignment when the Rad Monitor is in alarm. The Leakage limit is per the TS bases after an accident is SRO only knowledge. An understanding of the TS bases is required to analyze the terminology in the second part.
A. Correct. The High- High limit closes the Gaseous Waste discharge and aligns flow to CNMT. TS bases 10 CFR-50.67 limits the dose rate.
The N-16 monitor is the most sensitive monitor in detecting a SGTL, the condenser air ejector monitor is a second method for detecting SGTL.
The blowdown rad monitor and chemistry sampling are additional methods for confirming SGTL.
B. Incorrect. Both valves (TV-1SV-100B & TV-1SV-100A) are aligned on a High-High alarm. Correct reference to 10CFR50.67.
C. Incorrect. The High- High limit closes the Gaseous Waste discharge and aligns flow to CNMT. TS limit is not based on 10CFR 20.
D. Incorrect. Both valves (TV-1SV-100B & TV-1SV-100A) are aligned on a High-High alarm. TS limit is not based on 10CFR 20.
Sys # System Category KA Statement 000037 Steam Generator AA2 Ability to determine and interpret the following as they System status, using independent readings from (S/G) Tube Leak/3 apply to the Steam Generator Tube Leak: redundant Condensate air ejector exhaust monitor KIA# AA2.09 KIA Importance 3.4* Exam Level SRO References provided to Candidate None Technical
References:
1OM-53C.4.1.6.4 Pg 2, TS Bases 3.4.13 pg B 3.4.13-2 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
83. The plant is at 100% power.
A liquid waste discharge (RWDA-L) is progress with the Liquid Waste Effluent Monitor RM-1 LW-1 04 high setpoint set to 2.0 x 102 cpm and the High-High setpoint is 4.0 x 103 cpm.
At 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months an operator reports that the discharge line has ruptured and radiation levels are rising in the Auxiliary Building.
The following radiation monitor readings are indicated:
Ventilation Vent Liquid Waste Effluent TIME RM-1VS-109 ch 5 RM-1LW-104 1015 0.2 E 1 cpm 0.5 E 1 cpm 1030 3.1 E 3 cpm 1.7 E 2 cpm 1045 7.1 E 3 cpm 3.6 E 2 cpm 1100 2.3 E 4 cpm 2.1 E 3cpm 1115 1.1 E 4 cpm 3.5 E 3 cpm 1130 2.6 E 4 cpm 8.1 E 5 cpm 1145 1.1 E 5 cpm 8.3 E 5 cpm 1200 2.3 E 5 cpm 8.4 E 5 cpm Based upon the above times and indications:
(1) What is the HIGHEST Emergency Plan classification level?
(2) What is the LATEST Time that an Initial Notification must be made?
(Reference Provided)
A. (1) Unusual Event (2) 1145 B. (1) Alert (2) 1145 C. (1) Unusual Event (2) 1200 D. (1) Alert (2) 1200 Answer: B Explanation/Justification: Meets the SRO only white paper Section 11.0 second bullet. Analysis of plant radiation activity levels and the selection of emergency procedures. At BVPS SROs are responsible for EPP declarations. Requires the candidate to evaluate the control room indications and apply the criteria in the EPP chart for multiple conditions. Two classification criteria are met with different notification times which are limited to 15 minutes after criteria for classification are met.
A. Incorrect. UE level is initially met at 1130 hrs for RM-1VS-109 exceeding the limit (2.94E3) for 60 minutes, but this is not the highest level. The notification time is correct.
B. Correct. The Alert level is exceeded at 1145 hrs for RM-1 LW-1 04 exceeding the 200 times High-High alarm limit (8E5) for 15 minutes. The UE notification is required at 1145, since the limit for RM-1 VS-1 09 exceeded 2.94E3 for 60 minutes at 1130 hrs.
C. Incorrect. UE level is initially met at 1130 hrs, but is not the highest level. The time is incorrect, this would be the Notification time for the Alert.
D. Incorrect. Alert level is initially met at 1145 hrs. The time is incorrect, this would be the Notification time for the Alert, however the UE notification time is required by 1145 hrs.
Sys # System Category KA Statement 000059 Accidental Liquid Radwaste Release/9 Generic Knowledge of the emergency action level thresholds and classifications.
KIA# 2.4.41 KIA Importance 4.6 Exam Level SRO References provided to Candidate EPP Chart Technical
References:
EPP IP TAB RA1 and RU1 Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 55.43(b)(4)
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
84. The plant is at 100% power.
Radiation monitor alarms are received on Reactor Coolant Letdown Monitors [RM-1 CH-1 01A and B), due to increased RCS activity.
How will Pressurizer level respond three minutes after the required AOP actions are taken, and what are the accident assumptions for Tech. Spec. 3.4.16, RCS Specific Activity?
Pressurizer level will _ _ (1)_ _ _ and the Tech. Spec. accident assumptions are for a
_ _(2) _ __
A. (1) Rise (2) Main Steam Line Break and a Steam Generator Tube Rupture B. (1) Lower (2) Main Steam Line Break and a Steam Generator Tube Rupture C. (1) Rise (2) Steam Generator Tube Rupture ONLY D. (1) Lower (2) Steam Generator Tube Rupture ONLY Answer: A Explanation/Justification: Meets the SRO only white paper Section 11.8 third bullet. Determining PRZR level response is RO system response knowledge, however understanding the AOP procedure steps and knowing that letdown will be reduced is an SRO function. Additionally, knowledge of the TS bases for the Specific Activity Safety Analysis is a SRO responsibility. The Main Steam Break assumes an existing Steam Generator Tube Leak exists, so the offsite dose is impacted when the SLB occurs. The Steam Generator Tube Rupture assumes that there is also an offsite release, so the limit on specific activity limits offsite dose W/1 10CFR50.67 limits.
A. Correct. The AOP procedure directs reducing letdown flow, with PRZR level control in auto and a reduction in letdown flow, PRZR level will rise for approximately seven minutes before turning to restore level to program. The TS bases references the SLB and SGTR accidents.
B. Incorrect. Reduction in letdown flow causes PRZR level to Rise not lower. Correct accidents.
c. Incorrect. Reduction in letdown flow does cause PRZR level to Rise. Incorrect accident analysis because both the SLB and SGTR are analyzed per TS bases.
D. Incorrect. Reduction in letdown flow causes PRZR level to Rise not lower. Incorrect accident analysis because both the SLB and SGTR are analyzed per TS bases.
Sys # System Category KA Statement 000076 High Reactor AA2 Ability to determine and interpret the following as they apply Response of PZR LCS to changes in the letdown Coolant Activity/9 to the High Reactor Coolant Activity: flow rate KIA# AA2.06 KIA Importance 2.5 Exam Level SRO References provided to Candidate Technical
References:
10M-7.4.1F pg 12, 10M-53C.4.1.6.6 pg 4, None TS 3.4.16 bases pg B 3.4.16-2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
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
ICCM Train A is OOS for Maintenance The following plant conditions now exist:
RCS Pressure is 1940 psig and stable
RCS Hot Leg temperatures are 540 oF and lowering
RCS Cooldown rate based upon Cold Leg temperatures is currently 30 °F/Hr and CANNOT be reduced
Annunciator A3-71, ICC Malfunction Train B is in alarm due to a processor halt Which of the following procedures will be entered and what is the MAXIMUM allowable RCS Cooldown rate while reducing RCS Hot Leg temperatures to 500 oF?
A. ES-0.3, Natural Circulation with Steam Void in Vessel (With RVLIS); 50 °F/Hr B. ES-0.3, Natural Circulation with Steam Void in Vessel (With RVLIS); 100 °F/Hr C. ES-0.4, Natural Circulation with Steam Void in Vessel (Without RVLIS); 50 °F/Hr D. ES-0.4, Natural Circulation with Steam Void in Vessel (Without RVLIS); 100 °F/Hr Answer: C Explanation/Justification: Meets the SRO only white paper Section II.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, if the C/D rate cannot be maintained less than 25 °F/Hr, and RVLIS (ICCM both trains) are OOS. Detailed procedure knowledge is required for C/D rate above 500 oF the rate is 50 F/Hr, when less than 450 oF the rate rises to 100 F/ Hr. The alarm for Train B of ICCM(processor halt) must be recognized by the SRO that Train B of RVLIS is now also OOS, and then utilize this information to make the appropriate procedure transition.
A. Incorrect. Procedure is incorrect, the alarm on ICCM Train B renders both trains of RVLIS OOS. Correct Cooldown rate.
B. Incorrect. Procedure is incorrect, the alarm on ICCM Train B renders both trains of RVLIS OOS. Incorrect Cooldown rate.
C. Correct. Procedure is correct, the alarm on ICCM Train B renders both trains of RVLIS OOS, therefore ES-0.4 must be entered. Correct Cooldown rate.
D. Incorrect. Procedure is correct, the alarm on ICCM Train B renders both trains of RVLIS OOS. Incorrect Cooldown rate.
Sys # System Category KA Statement W/E10 Natural Circulation with Steam Void Generic Ability to verify system alarm setpoints and operate controls identified in the alarm in Vessel with/without RVLIS/4 response manual.
KIA# 2.4.50 KIA Importance 4.0 Exam Level SRO References provided to Candidate None Technical
References:
10M-53A.1.ES-0.2 step 15 pg 16, 10M-53A.1.ES-0.4 step 3 pg 3 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
86. The plant is at 50% power with all systems in normal alignment for this power level.
A load rejection has just occurred
Annunciator A4-46, "TAVG DEVIATION FROM TREF" is in alarm
Annunciator A4-124, "ROD CONTROL BANK D LOW-LOW" is in alarm
While performing 10M-7.4.S, "Emergency Boration" attempts to open Emergency Boration Isolation valve [MOV-1CH-350] have failed (1) What procedural action will be taken to address the LOW-LOW Rod Control Bank D position?
(2) If Emergency Boration Isolation valve [MOV-1 CH-350] was opened and per Tech. Spec 3.1.1 Shutdown Margin Bases, what BORATION flow rate as indicated on Emergency Boration Flow [FI-1CH-110] is established to restore SDM to within limits?
A. (1) Initiate boration from the blender (2) 30 gpm B. (1) Initiate boration from the blender (2) 105 gpm C. (1) Raise Turbine load (2) 30 gpm D. (1) Raise Turbine load (2) 105 gpm Answer: A Explanation/Justification: Meets the SRO only white paper Section 11.8 third bullet. Determining the procedural response to the failed emergency boration valve is potentially RO level knowledge if system piping knowledge is used to evaluate the situation. The detailed actions taken per the Emergency Boration procedure is SRO knowledge. Additionally, knowledge of the TS bases for Shutdown Margin Action A.1 is a SRO responsibility.
A. Correct. The Load Rejection will raise RCS temperature and Control Rods will insert. The Low-Low Rod position procedure directs establishing emergency boration flow. With the Emergency Boration Valve [MOV-1CH-350) failed closed, an alternate flowpath is required using the Blender. This boration flowpath will allow for rod withdrawal above the RIL setpoint to comply with Tech Specs.
Correct flow rate per Tech. Spec 3.1.1 bases to restore SDM to within limits.
B. Incorrect. This is the correct action to take per the procedure. The second part is incorrect, the 105 gpm flowrate is the normal charging header flow established when using the Refueling Water Storage Tank (RWST) to establish a boration flowpath. This is plausible ifthey remember the 105 gpm flow rate from the procedure.
C. Incorrect. Raising Turbine load would allow for withdrawal of the control rods to correct the Low-Low bank position, however this action is not per the ARP procedure and should not be performed until the cause of the load rejection is determined. The flow rate is correct D. Incorrect. Raising Turbine load would allow for withdrawal of the control rods to correct the Low-Low bank position, however this action is not per ARP procedure and should not be performed until the cause of the load rejection is determined. The flow rate is incorrect Sys# System Category KA Statement 004 Chemical and A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; Emergency Boration Volume Control and (b) based on those predictions, use procedures to correct, control, or mitigate the System consequences of those malfunctions or operations:
KIA# A2.14 KIA Importance 3.9 Exam Level SRO References provided to Candidate Technical
References:
10M-1.4.ABF Rev. 6, 10M-7.4.S Rev. 7, None TS 3.1.1 bases pg B 3. 1 .1-4 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(2)
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
87. The plant is at 100% power.
The following conditions exist:
Annunciator A 11-25, Containment Air Recirc Fan Auto Stop is in alarm
Containment Air Recirc Fan [1VS-F-1A], indicating light is Bright WHITE
The control switch for Containment Air Recirc Fan [1VS-F-1A] has a RED target
CNMT temperature rises to 100 oF and stabilizes For these conditions:
(1) What will be the impact on Peak Containment Pressure following a Design Basis LOCA?
Peak Containment Pressure following a DBA LOCA will (1 )_ __
The crew then implements 1OM-44C.4.E, Changing Containment Air Recirculation Fan 1C Power Supply.
(2) lAW 10M-44C.4.E, what procedural action(s) will be REQUIRED?
The procedurally REQUIRED action(s) of 1OM-44C.4.E is/are to _ _ _ _(2) _ _ __
A. (1) NOT exceed accident analysis limits (2) align ONLY the Containment Air Recirc Fan [1VS-F-1 C] to the 1N Bus B. (1) exceed accident analysis limits (2) align ONLY the Containment Air Recirc Fan [1VS-F-1 C] to the 1N Bus C. (1) NOT exceed accident analysis limits (2) align BOTH the Containment Air Recirc Fan [1VS-F-1C] and Control Rod Drive Mechanism Fan [1VS-F-2C] to the 1N Bus D. (1) exceed accident analysis limits (2) align BOTH the Containment Air Recirc Fan [1VS-F-1 C] and Control Rod Drive Mechanism Fan [1VS-F-2C] to the 1N Bus Answer: C Explanation/Justification: Meets the SRO only white paper Section II.E first paragraph pg 6, and II.B 3rt1 bullet on page 3, requires the SRO to assess plant conditions and then select the appropriate procedure action. There will be no effect on Peak Containment pressure as the CAR fans are secured on a CIB signal, they are not required for accident conditions. The 1C CAR fan is powered from either of the 480V emergency busses. The 2C CRDM fan is to be aligned to the same bus as the 1C CAR fan to ensure that Train A and B common raceways and penetrations are separated. The SRO is required to have knowledge of specific procedure details and know the TS safety Analysis bases.
A. Incorrect. It is correct that peak containment pressure following a DBA LOCA will not exceed accident analysis limits since cnmt temperature is within TS range of 70-1 OBF. It is incorrect that only the CAR fan is aligned since due to electrical separation concerns both of the swing powered fans must be aligned to the same bus.
B. Incorrect. Peak containment pressure following a DBA LOCA will not exceed accident analysis limits since cnmt temperature is within TS range of 70-108F. It is incorrect that only the CAR fan is aligned since due to electrical separation concerns both of the swing powered fans must be aligned to the same bus.
C. Correct. It is correct that peak containment pressure following a DBA LOCA will not exceed accident analysis limits since cnmt temperature is within TS range of 70-1 08F, and due to electrical separation concerns, both of the swing powered fans must be aligned to the same bus.
D. Incorrect. Peak containment pressure following a DBA LOCA will not exceed accident analysis limits since cnmt temperature is within TS range of 70-1 08F. Due to electrical separation concerns, both of the swing powered fans must be aligned to the same bus.
Sys# System Category KA Statement 022 Containment A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Fan motor over-current Cooling System CCS; and (b) based on those predictions, use procedures to correct, control, or (CCS) mitigate the consequences of those malfunctions or operations:
KIA# A2.01 KIA Importance 2.7 Exam Level SRO References provided to Candidate None Technical
References:
1OM-44C.4.E pg 2 & 3, TS bases page B3.6.5-1 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
88. The plant is at 100% power.
The 125 VDC breaker [INV-VITBUS1-1-B1] to UPS VIT BUS 1-1 has tripped
All systems function as designed Per Tech. Spec. 3.8.7 Inverters- Operating and Tech. Spec. 3.8.9 Distribution Systems-Operating, how will this failure affect system OPERABILITY?
The Inverter is (1)_ __
The Electrical Distribution system is (2) _ __
A. (1) OPERABLE (2) OPERABLE B. (1) OPERABLE (2) INOPERABLE C. (1) INOPERABLE (2) OPERABLE D. (1) INOPERABLE (2) INOPERABLE Answer: C Explanation/Justification: Meets the SRO only white paper Section II.B 3rd bullet pg 3, requires system knowledge of the breaker input to the inverter, which is RO knowledge, however the determination of what constitutes Operable subsystems per TS 3.8. 7 and 3.8.9 requires knowledge of the TS bases.
A. Incorrect. The inverter is not operable when the DC supply is unavailable. The breaker that opened supplies DC to the Inverter. The Electrical Distribution system is still energized and remains operable.
B. Incorrect. The inverter is not operable when the DC supply is unavailable. The breaker that opened supplies DC to the Inverter. The Electrical Distribution system is not inoperable.
C. Correct. The inverter is not operable when the DC supply is unavailable. The breaker that opened supplies DC to the Inverter. The Electrical Distribution system is still energized and remains operable.
D. Incorrect. The inverter is not operable when the DC supply is unavailable. The breaker that opened supplies DC to the Inverter. The Electrical Distribution system is still energized and remains operable.
Sys# System Category KA Statement 063 DC Electrical Distribution System Generic Ability to determine operability and/or availability of safety related equipment.
KIA# 2.2.37 KIA Importance 4.6 Exam Level SRO References provided to Candidate None Technical
References:
TS 3.8.7 & Bases pg B 3.8.7-2, TS 3.8.9 Bases pg B 3.8.9-2 Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
89. The plant is in Mode 4 preparing to enter Mode 3. All systems are in normal alignment for this Mode, with Train B as the protected train.
The outside operator reports the following Emergency Diesel Generator fuel oil inventories:
Emergency Diesel Generator (EDG #1) EE-EG-1 is 14,500 gal
Emergency Diesel Generator (EDG #2) EE-EG-2 is 17,000 gal Based on these conditions, what is the Tech. Spec. status of the required EDG(s) and what condition will exit any Tech. Spec. Required Action?
(Reference Provided)
A. EDG #2 is Inoperable. Plant in Mode 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .
B. EDG #1 is Inoperable. BOTH Emergency Diesel Generator fuel oil inventories are restored to
~ 17,500 gal within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
C. EDG #1 AND EDG #2 are Inoperable. Plant in Mode 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .
D. EDG #2 is Inoperable. ONLY EDG #2 fuel oil inventory is restored to
~ 17,500 gal within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 151 and 3m bullet on page 3. The SRO candidate must predict the impact of the reported fuel oil levels on the operability status of the EDGs and then use the provided TS pages to determine what actions are required. The candidate must also have knowledge of the TS bases to determine if the EDGs are operable. Specifically, the SRO must apply knowledge of the TS bases where 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is allowed to restore fuel oil levels before the EDG must be declared inoperable, as long as level is not below 15,000 gal.
A. Incorrect. Plausible since this would be the TS required action if both EDGs were inoperable. Since the B train is protected, the candidate may assume that the single protected train EDG is required to be the only operable EDG in Mode 4 and that TS 3.0.3 is applicable. Mode 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months is the TS 3.0.3 action.
B. Correct. In order to answer the question the candidate must predict the impact of the reported fuel oil levels on the operability status of the EDGs and then use the provided TS page to determine what actions are required. The candidate must also have knowledge of the TS bases to determine if the EDGs are operable. Inventory must be <17500 gal and ~15000gal to obtain the allowed 48 hr completion time to restore and not declare the EDG inoperable as stated in the TS bases. Based on this, EDG #1 will be Inoperable immediately, and both EDGs have to be restored to ~17500 gal w/in 48 hrs.
C. Incorrect. Plausible if the candidate believes that the low fuel oil level makes both EDGs inoperable and TS 3.0.3 is applicable. Mode 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months is the TS 3.0.3 action.
D. Incorrect. Plausible if the candidate incorrectly believes since we are in Mode 4 and train B protected, this is the only required EDG and correctly assesses that the given fuel oil level does not make this EDG inoperable.
Sys# System Category KA Statement 064 Emergency A2 Ability to (a) predict the impacts of the following malfunctions or operations on Load, VARS, pressure on air Diesel Generator the ED/G system; and (b) based on those predictions, use procedures to correct, compressor, speed droop, (ED/G) System control, or mitigate the consequences of those malfunctions or operations: frequency, voltage, fuel oil level, temperatures KIA# A2.02 KIA Importance 3.3 Exam Level SRO References provided to Candidate TS Page 3.8.3-1 & 2 Technical
References:
TS pages 3.8.3-1 & 2, TS Bases page B 3.8.3-3 2nd paragraph Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content:
Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air LCO 3.8.3 The stored diesel fuel oil, lube oil, and starting air subsystem shall be within limits for each required diesel generator (DG).
APPLICABILITY: When associated DG is required to be OPERABLE.
ACTIONS
-NOTE-Separate Condition entry is allowed for each DG.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more DGs with fuel A.1 Restore fuel oil inventory to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months inventory within limits.
~1Z;;§** '"'V
gal{.~
..d.53..
225 g~i:ar1dfitt4li~a2st
,_~,, *',..,.,.~ ***~rtl"ff.w':. &> .'* '~-* ,. *.* " * .,,_~-.;., .
.J-f,.,."'.**>F""
gal (Unit 2),i in storage tank.
B. One or more DGs with lube B.1 Restore lube oil inventory 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months oil inventory < 330 gal and to within limits.
~ 283 gal.
C. One or more DGs with C.1 Restore fuel oil total 7 days stored fuel oil total particulates to within limits.
particulates not within limit.
D. One or more DGs with new D.1 Restore stored fuel oil 30 days fuel oil properties not within properties to within limits.
limits.
Restore starting air 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to Beaver Valley Units 1 and 2 3.8.3- 1 Amendments 278 I 161
Diesel Fuel Oil, Lube Oil, and Starting Air
. 3.8.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Declare associated DG Immediately associated Completion inoperable.
Time not met.
OR One or more DGs with diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than Condition A, B, C, D, or E.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 31 days SR 3.8.3.2 Verify lubricating oil inventory is ;::: 330 gal. 31 days SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance tested in accordance with, and maintained within the with the Diesel limits of, the Diesel Fuel Oil Testing Program. Fuel Oil Testing Program SR 3.8.3.4 31 days SR 3.8.3.5 Check for and remove accumulated water from each 92 days fuel oil storage tank.
Beaver Valley Units 1 and 2 3.8.3- 2 Amendments 278/161
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
90. The plant is at 25% with all systems in normal alignment for this power level.
Containment Pressure is 2.0 psig and slowly rising.
Annunciator A6-110, CNMT lnstr Air System Press Low is received
Containment Instrument Air Header Pressure [PI-11A-1 06A] is 85 psig and slowly dropping
lnst Air To Cnmt lnst Air lsol [TV-11A-400] is OPEN
Station Instrument Air Header Pressure [PI-11A-1 06] is 95 psig and stable
The crew enters AOP 1.34.2, Loss of Containment Instrument Air
IA-90 Bypass Valve For [TV-11A-400] is slowly OPENED lAW guidance in AOP 1.34.2 Shortly after IA-90 is opened, the following annunciator is received:
A 1-56, Steam line Stop Valve Not Fully Open lAW AOP 1.34.2, what actions are REQUIRED?
IA-90 and TV-11A-400 will _ _ _(1) _ _ AND the SRO will direct the crew to
_ _ _ _ _ _ _ _(2) _ _ _ _ _ _ _ __
A. (1) be CLOSED (2) manually trip the reactor and enter E-0, Reactor Trip Or Safety Injection B. (1) remain OPEN (2) manually trip the reactor and enter E-0, Reactor Trip Or Safety Injection C. (1) be CLOSED (2) perform a power reduction lAW AOP 1.51.1, Unplanned Power Reduction D. (1) remain OPEN (2) perform a power reduction lAW AOP 1.51.1, Unplanned Power Reduction Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper paragraph E 3n1 bullet on page ?and 151 paragraph page 6. SRO only in that it requires the implementation of administrative controls necessary to maintain the CNMT penetration operable as well as detailed knowledge of the content of an AOP . Leaving IA-90 open would allow valves in the CNMT to operate as long as possible which is a reasonable action however there is a matter of the TS administrative requirements to maintain the CNMT penetration operable. SRO must be aware of this administrative requirement and be knowledgeable of the procedure content to realize that the procedure will direct closing of IA-90 if ANN A 1-56 is received.
A. Correct. In accordance with AOP-1.34.2 continuous action step 2.d, if A 1-56 is not clear, then perform the RNO step. RNO states to verify closed IA-90 & TV-11A-400, then manually trip the Reactor.
B. Incorrect. IA-90 & TV-11A-400 are directed to be closed. Manually tripping is the correct action.
C. Incorrect. Closing IA-90 & TV-11A-400 is correct. Performing a power reduction would temporarily alleviate the situation and allow more time for the air system to possibly recover however, the AOP does not direct this action.
D. Incorrect. IA-90 & TV-11A-400 are directed to be closed. Performing a power reduction would temporarily alleviate the situation and allow more time for the air system to possibly recover however, the AOP does not direct this action.
Sys # System Category KA Statement 078 Instrument Air Generic Ability to interpret control room indications to verify the status and operation of a system, and System (lAS) understand how operator actions and directives affect plant and system conditions.
KIA# 2.2.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical
References:
AOP 1.34.2 step 2.d RNO Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
91. The plant is performing a startup and is currently at 15% power with all systems in NSA for this power.
The compensating voltage power supply to Intermediate Range Neutron Flux channel N35 fails to zero volts (1) How will this failure affect the Source Range Neutron Flux channel response following a reactor trip?
Following a reactor trip, the Source Range Neutron Flux channels ____ (1 )_ _ __
as flux decreases into the Source Range.
(2) Per AOP 1.2.1 B, "Intermediate Range Channel Malfunction", what is the status of the plant startup?
The plant startup _ _ _ _(2) _ _ _ _continue.
A. (1) will automatically unblock (2) may not B. (1) will automatically unblock (2) may C. (1) must be manually unblocked (2) may not D. (1) must be manually unblocked (2) may Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 3rd bullet on page 3. SRO must have knowledge of the TS bases to answer this question. Specifically, knowledge of when the IR trip is applicable. The loss of the IR Compensating Voltage will require the Source Range channels to be manually unblocked This is RO knowledge. The status of power operation is SRO knowledge from the TS actions and applicable operating conditions.
A. Incorrect. SR detectors will not auto unblock since the logic is 2/2 below P-6 and the PS failure will cause N35 to remain above the P-6 setpoint.
Intermediate Range Neutron Flux trip is blocked above 10% power, power operations may continue.
B. Incorrect. SR detectors will not auto unblock since the logic is 2/2 below P-6 and the PS failure will cause N35 to remain above the P-6 setpoint.
Part 2 is correct.
C. Incorrect. Part 1 is correct. Power operation may continue if the plant is above P-10.
D. Correct. lAW Fundamental knowledge of neutron detection and AOP 1.2.1 B step 6.b & c and TS bases page B 3.3.1-13 item 4. Power Operation may continue per the AOP and TS applicable conditions.
Sys # System Category KA Statement 015 Nuclear A2 Ability to (a) predict the impacts of the following malfunctions or operations Power supply loss or erratic Instrumentation on the NIS; and (b) based on those predictions, use procedures to correct, operation System (NIS) control, or mitigate the consequences of those malfunctions or operations:
KIA# A2.01 KIA Importance 3.9 Exam Level SRO References provided to Candidate None Technical
References:
Fundamental knowledge, AOP 1.2.1 B step 6.b &
c, TS 3.3.1-1 Function 4, TS bases page B 3.3.1-14 item 4.
Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
92. The plant is at 100% power.
Fuel defects are present and RCS activity is elevated
RCS gross activity is 11 0 ~Ci/gm
Dose Equivalent 1-131 is 0.33 ~Ci/g~
The Average Disintegration Energy (E) of the RCS is 0.5
Waste Gas Decay Tank [GW-TK-1C] is aligned for filling
Chemistry reports the total Curie content of GW-TK-1C is 52,500 curies Based on these conditions, what Tech. Spec./ODCM action(s) will be REQUIRED?
(Excluding any reporting requirements)
(Reference Provided)
A Be in Mode 3 with Tavg ~ 500 oF within 6 hours AND reduce GW-TK-1C tank contents to within the limit within 48 hours ONLY.
B. Be in Mode 3 with Tavg ~ 500 oF within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND restore Dose Equivalent 1-131 to within limits within 48 hours ONLY.
C. Immediately suspend all additions of radioactive material to the tank, AND reduce GW-TK-1C tank contents to within the limit within 48 hours ONLY.
D. Immediately suspend all additions of radioactive material to the tank, AND restore Dose Equivalent 1-131 to within limits within 48 hours ONLY.
Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B, 151 and 41h bullet on page 3. SRO must have knowledge of the TS/ODCM to answer this question. Specifically, the SRO must also be able to apply the ODCM actions for high currie content of a WGDT and be able to eliminate the TS actions associated with high gross RCS activity.
A. Incorrect. Plausible distractor if candidate views RCS Specific Activity as the problem and misinterprets TS 3.4.16 as the problem. This would require Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . It is correct to reduce tank to w/in limits in 48 hrs.
B. Incorrect. Plausible distractor if candidate views RCS Specific Activity as the problem and misinterprets TS 3.4.16 as the problem. This would require Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 1-131 is w/in TS limit, therefore reducing 1-131 tow/in limits is incorrect.
C. Correct. Waste Gas Decay tank is greater than the ODCM limit of 52,000 Curies. This will require suspending additions and reducing tank contents to w/in limits in 48 hrs.
D. Incorrect. Plausible distractor if candidate views both 1-131 and WGDT activity as the problem. Suspending additions and reducing tank contents tow/in limits in 48 hrs is correct, but restoring dose equivalent 1-131 tow/in limits in 48 hrs. is a required action for 1-131 exceeding TS limits which is not correct for the conditions stated in the question.
Sys # System Category KA Statement 071 Waste Gas Disposal A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Use of waste gas release System (WGDS) Waste Gas Disposal System; and (b) based on those predictions, use procedures to monitors, radiation, gas correct, control, or mitigate the consequences of those malfunctions or operations: flow rate, and totalizer KIA# A2.02 KIA Importance 3.6 Exam Level SRO References provided to Candidate 1/2-0DC-3.03; TS 3.4_ 16 Technical
References:
1/2-0DC-3.03 page 66 of 83 Att.O item 3.11.2.5, TS 5.5.8, TS 3.4.16 No Bases Provided Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 55.43(b) (1) & (2)
' (SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
93. lAW Tech. Spec. bases, what is the most limiting event (design bases accident) for the Atmospheric Dump Valves (ADVs)?
A. ATWS B. Steam Generator Tube Rupture C. Station Blackout D. Loss of All Feedwater Answer: B Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 3rd bullet on page 3. SRO must have knowledge of the TS bases to answer this question. Specifically, SRO must know the design bases accident for the ADVs. At Beaver Valley, the SAS supplies the lAS and only a few air operated valves are credited in accident analysis and addressed in the TS. Therefore the tie between SAS and TS bases comes through those valves credited in accident analysis that are air operated. The SRO must have knowledge of TS bases for those air operated valves.
A. Incorrect. The ADVs are used during ATWS events since the Main Steamline Isolation valves are closed. However, no accident analysis credit is taken for use of the ADVs during an A TWS event.
B. Correct. The design bases for the ADVs are established by the capability to cool the unit to RHR entry conditions. For the recovery from a design bases SGTR accident, the operator is required to perform a limited C/D to establish adequate subcooling as a necessary step to terminate the primary to secondary break flow into the faulted SG.
C. Incorrect. The ADVs are used during Station blackouts to cool the RCS as quickly as possible to limit RCS inventory loss thru the RCP seals, however this is not the design bases accident for the ADVs.
D. Incorrect. The ADVs are used to reduce SG pressure to allow Condensate to feed the SGs. However, no accident analysis credit is taken for use of the ADVs during this event.
Sys# System Category KA Statement 079 Station Air System (SAS) Generic Knowledge of the bases in Technical 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:
TS bases page B3.7.4-2 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
94. The plant is at 100% power.
At 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 1OST-1.1, "Control Rod Assembly Partial Movement Test" is scheduled for performance.
lAW NOP-OP-1 004, Reactivity Management, which of the items below are the responsibility of the Shift Manager/Unit Supervisor (Command SRO) during the performance of this procedure?
1. Direct control of reactivity including taking conservative actions to safeguard the integrity of the fuel.
2. Initiate a manual reactor trip when in his/her judgment a situation exists which jeopardizes or threatens to jeopardize public or plant safety.
3. Direct conservative actions to respond to unexpected or anomalous core conditions.
4. Consider the experience level of each individual and of the reactivity team when assigning members to perform reactivity manipulations.
5. Performing on-shift risk analyses for emergent conditions that may arise.
A. 1, 2 & 3 only B. 1, 3, & 4 only C. 2, 4, & 5 only D. 3, 4, & 5 only Answer: 8 Explanation/Justification: Meets the requirements of the SRO only white paper paragraph E 3n1 bullet on page 7. SRO only in that it requires the implementation of administrative procedures. Specifically, the SROs conservative decision responsibilities during reactivity manipulations.
A. Incorrect. Items 1 and 3 are correct; Item 2 is the responsibility of the ATC.
B. Correct. In accordance with NOP-OP-1004, section 4.1.9, steps 2, 3, and 8.
C. Incorrect. Item 4 is correct. Item 2 is the responsibility of the ATC and item 5 is the responsibility of the STA. Risk assessment is always the SRO responsibility but not the formal risk analysis.
D. Incorrect. Items 3 and 4 are correct. Item 5 is the responsibility of the STA. Risk assessment is always the SRO responsibility but not the formal risk analysis.
Sys # System Category KA Statement N/A N/A Generic Knowledge of conservative decision making practices.
KIA# 2.1.39 KIA Importance 4.3 Exam Level SRO References provided to Candidate None Technical
References:
NOP-OP-1 004 page 9 step 4.1.9 Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: 55.43(b)(5)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
95. The plant is in Mode 3 preparing to enter Mode 2. All systems are in normal alignment for this condition.
The following indications are present in the control room:
Charging Pump Press [PI-1CH-121] is indicating 2500 psig and stable
RCP1A Seallnj. [FI-1CH-130] is indicating 10.5 gpm and stable
RCP1B Seallnj. [FI-1CH-127] is indicating 9.75 gpm and stable
RCP1C Seallnj. [FI-1CH-124] is indicating 9.0 gpm and stable
RCS pressure is 2235 psig and stable (1) Based on these conditions, is Tech. Spec. LCO 3.5.5 Seal Injection Flow being met?
(2) What is the Tech. Spec. bases for LCO 3.5.5 Seal Injection Flow?
A. (1) Yes (2) Ensures adequate Sl flow B. (1) Yes (2) Ensures RCP Seal Integrity C. (1) No (2) Ensures adequate Sl flow D. (1) No (2) Ensures RCP Seal Integrity Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 3rd bullet on page 3. SRO must have knowledge of the TS bases to answer this question. Specifically, the SRO must know that the analysis performed to verify acceptable Sl flows for small and large break LOCAs was performed assuming this was full open. The SRO must discriminate for the concern for maintaining the integrity of the RCP seals and meeting the Sl flow requirements assumed in the accident analysis. Part 1 is RO knowledge since the position of HCV-186 is addressed above the line in TS 3.5.5. Part 2 is SRO knowledge since it is only addressed in the Bases section of the TS.
A. Incorrect. Wrong LCO assessment. Correct bases.
B. Incorrect. Wrong LCO assessment. Incorrect bases.
C. Correct. Seal injection flow is 29.25 gpm. TS states seal injection flow is required to be 98 gpm with discharge pressure ~457 and FCV full open. TS bases states that flow w/in the limits ensures that the ECCS injection flow (SI flow) stay w/in the safety analysis assumptions. Part 1 can be answered with RO system knowledge and knowledge of TS above the line information. Part 2 is SRO knowledge of TS bases. Candidate must assess indications and controls given in the stem to determine if TS LCO is being met.
D. Incorrect. Correct LCO assessment. Incorrect bases.
Sys# System Category KA Statement N/A N/A Generic Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
KIA# 2.1.31 KIA Importance 4.3 Exam Level SRO References provided to Candidate Technical
References:
TS page 3.5.5-1, TS bases page 3.5.5-3 None Question Source: New Question Cognitive Level: High - Comprehension 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
96. Given the following plant conditions:
The plant is 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 indicates 800 cps and stable
Source Range Channel N32 indicates 800 cps and stable
Refueling cavity water clarity is murky The following control room alarms and indications are then received:
A4-85, NIS Source Range CHI Detector Voltage Trouble
Source Range Channel N31 indicates 0 cps
Loss of Detector Volt. Status light is LIT on the N31 Drawer Based on these conditions, 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.
B. 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 white paper paragraph B 151 & 3rd bullets on page 3. 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 the AOP and will be responsible for directing the operator actions to comply with the TS actions.
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 interpret the CR indications as a loss of N31. Both AOP 1.2.1A for SR Channel Malfunction and TS 3.9.2 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 N/A Generic Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
KIA# 2.2.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical
References:
1OM-53C.4.1.2.1A, Rev. 6, Pg. 1, 2, & 6; TS Definitions Pg. 1.1-2; TS 3.9.2 Pg. 3.9.2-1; TS B3.9.2 Pg. B3.9.2-1 &2; 1/2RP-1.1, Issue 0, Rev.25, Pg.5 Question Source: Bank- Modified Vision# 81994 Question Cognitive Level: High -Analysis 10 CFR Part 55 Content: 55.43(b)(2)
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
97. The plant is in Mode 6 performing the core offload.
A spent fuel assembly is damaged while moving through the transfer canal and the following radiation monitor alarms are received:
Fuel Pool Bridge Area Monitor [RM-1 RM-207] Alarms High
Fuel Bldg. Vent. Exh. Gross Activity [RIS-VS-103A & B] Alarm High
Containment Purge Exhaust Monitor [RM-1VS-104A & 8] Alarm High-High
The Containment Equipment Hatch is open Based on these indications, what action is required per AOP 1.49.1 Irradiated Fuel Damage?
A. MANUALLY Close [1VS-D-4-2A or 28] Leak Coli Sys Filt Bk Dmpr and Open [1VS-D-4-1A and 1B] Leak Coli Sys Filt Bk Bypass Dmprs.
B. CHECK Closed [1VS-D-4-2A or 28] Leak Coli Sys Filt Bk Dmpr and Open [1VS-D-4-1A and 18]
Leak Coli Sys Filt Bk Bypass Dmprs.
C. MANUALLY Open [1VS-D-5-3A and 38] CNMT lso Purge Exh To Main Filt Bk Dmprs and
[1VS-D-5-5A and 58] CNMT lsol Purge Sup Dmprs.
D. CHECK Open [1VS-D-5-3A and 38] CNMT lso Purge Exh To Main Filt Bk Dmprs and [1VS-D-5-5A and 58] CNMT lsol Purge Sup Dmprs.
Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper paragraph D 2nd bullet on page 6. SRO must analyze the radiation conditions given in the stem to determine location of the fuel failure, then select the correct action. The SRO is required to know the plant configuration for Refueling activities, and for the conditions stated the Automatic isolation feature of the Purge and Exhaust dampers is defeated.
A. Incorrect. Fuel Pool Vent RM HIGH-HIGH alarm is not in Alarm so the dampers do not align automatically from this Rad Monitor. lffuel damage location is diagnosed as in the Fuel Pool, then this action is plausible. The CNMT Purge Monitors will align these dampers automatically, however the damper positions are incorrect.
B. Incorrect. Fuel Pool Vent RM HIGH-HIGH alarm is not in Alarm so dampers do not automatically align from this Rad Monitor. If fuel damage location is diagnosed as in the Fuel Pool, then this action is plausible. The dampers would realign from the CNMT Purge Rad monitors, however the positions are incorrect.
C. Incorrect. The automatic closure feature of the Purge and Exhaust Dampers is defeated, they will remain open on the High-High radiation signal from RM-1VS-104 A and 104 B, they do not need to be realigned manually..
D. Correct. Since the Containment Equipment Hatch is open, the automatic closure feature of the Purge and Exhaust Dampers is defeated, they will remain open on the High-High radiation signal from RM-1VS-104 A and 104 8, the procedure will verify that they are open ..
Sys# System Category KA Statement N/A N/A Generic Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
KIA# 2.3.5 KIA Importance 2.9 Exam Level SRO References provided to Candidate None Technical
References:
1OM-53C.1.49.1 pages 2 and 6 Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 55.43(b)(4)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
98. Following a recent refueling outage, the plant is at 25% power with all systems in normal alignment for this power level. Initial power escalation is in progress.
Maintenance is moving the offloaded fuel assemblies in the fuel pool to support modifications to the spent fuel pool level transmitter
The power supply to Unit 1 Control Room Radiation Monitor [RM-1RM-2188]
de-energizes What Tech. Spec./Licensing Requirements Manual (LRM) action(s) will be REQUIRED?
(Reference Provided)
A. Verify the corresponding Unit 2 Control Room Radiation Monitor is functional within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once per 31 days thereafter AND place one CREVS train in emergency pressurization mode of operation within 7 days.
B. Immediately suspend movement of recently irradiated fuel assemblies AND restore required CREVS train to operable status within 7 days.
C. Verify the corresponding Unit 2 Control Room Radiation Monitor is functional within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and once per 31 days thereafter ONLY.
D. Immediately place one CREVS train in emergency pressurization mode of operation ONLY.
Answer: C Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 1st bullet on page 3. SRO must recognize that the PS failure will make the monitor inoperable and apply the appropriate TS. SRO must also be familiar with the TS bases to recognize that the fuel being moved is NOT recently irradiated by the bases definition.
A. Incorrect. Correct actions for recently irradiated fuel.
B. Incorrect. Correct action for 2 channels inoperable.
C. Correct. lAW LRM pages 3.3.14-1 & 2; TS pages 3.3.7-1, 2, &,3; TS pages 3.7.10-1, 2, &,3 and TS bases page 3.7.10-7. Although all of the choices contain 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less action statements (RO Knowledge) the question cannot be answered with information alone. In order to answer this question, the SRO must apply the TS and LRM to arrive at the correct answer. Application of the TS is an SRO knowledge.
D. Incorrect. Partially correct for one channel inoperable and recently irradiated fuel.
Sys # System Category KA Statement N/A N/A Generic Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
KIA# 2.3.15 KIA Importance 3.1 Exam Level SRO References provided to Candidate LRM pages 3.3.14-1 ,2; & 3 Technical
References:
LRM pg 3.3.14-1 & 2, TS pg 3.3.7-1, 2, & 3, TS pg TS pages 3.3.7-1, 2, &,3; 3.7.10-1, 2, & 3, TS bases pg 3.7.10-7 TS pages 3.7.10-1, 2, & 3 Question Source: New Question Cognitive Level: High - Application 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14)
99. lAW the Tech. Spec. bases for Post Accident Monitoring Instrumentation two OPERABLE channels are required for most functions while in Modes 1, 2, and 3.
lAW the Tech. Spec. bases, which of the following functions are EXCEPTIONS to this two channel requirement?
1. Source Range Neutron Flux
2. Pressurizer Water Level
3. Steam Generator Wide Range Water Level
4. Primary Plant Demineralized Water Storage Tank Level
5. Penetration Flow Path Containment Isolation Valve Position
6. High Head Sl Automatic Injection Header Flow A. 1, 2 & 3 only B. 1, 4, & 5 only C. 2, 4, & 6 only D. 3, 5, & 6 only Answer: D Explanation/Justification: Meets the requirements of the SRO only white paper paragraph B 3n1 bullet on page 3. Knowledge of TS bases.
A. Incorrect. Item 3 is correct; Item 1 and 2 only requires one channel for the emergency shutdown panel instrumentation TS, but requires two for PAMS.
B. Incorrect. Item 5 is correct; item 4 may be mistakenly chosen based on the PPDWST containing enough volume for 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months of heat sink which would lead one to mistakenly believe the TS considers this enough time to take corrective actions. Item 1 only requires one channel for the emergency shutdown panel instrumentation TS, but requires two for PAMS.
C. Incorrect. Item 6 is a correct function. Item 2 only requires one channel for the emergency shutdown panel instrumentation TS, but requires two for PAMS, Item 4 may be mistakenly chosen based on the PPDWST containing enough volume for 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months of heat sink which would lead one to mistakenly believe the TS considers this enough time to take corrective actions.
D. Correct. lAW TS bases these functions are exceptions to the two operable channels requirement. The bases explains why each of these exceptions is acceptable. In order to answer this question, the candidate must have knowledge of the TS bases to identify the PAM instrumentation that only requires one channel of instrumentation to satisfy the required function.
Sys # System Category KA Statement N/A N/A Generic Ability to identify post-accident instrumentation.
KIA# 2.4.3 KIA Importance 3.9 Exam Level SRO References provided to Candidate Technical
References:
TS bases pages 8 3.3.3-3 & 4 None Question Source: New Question Cognitive Level: Low- Memory 10 CFR Part 55 Content: 55.43(b)(2)
Objective:
(SRO ONLY)
Beaver Valley Unit 1 NRC Written Exam (1LOT14) 100. A LOCA has occurred coincident with some fuel damage and high CNMT radiation.
A General Emergency has been declared at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months 
The TSC has NOT yet been activated
No radioactive release has occurred or is imminent (within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )
Health Physics has provided the following dose projections:
At the EAB: 25 mRem TEDE; 15 mRem CDE At 5 miles: 4.5 mRem TEDE; 5.0 mRem CDE At 2 miles: 5.0 mRem TEDE; 10 mRem CDE The following meteorological conditions exist:
35' wind direction is from 1ooo at 3 MPH
150' wind direction is OOS
500' wind direction is from 120° at 10 MPH Based on these conditions, what Protective Action Recommendation (PAR) is REQUIRED?
(Reference Provided)
A. Evacuate 0-5 miles, 360 degrees AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer Kl in accordance with the State plan.
B. Evacuate 0-2 miles, 360 degrees AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer Kl in accordance with the State plan.
C. Evacuate 0-2 miles, 360 degrees and 5 mile downwind wedge LMNPQR AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer Kl in accordance with the State plan.
D. Evacuate 0-2 miles, 360 degrees and 5 mile downwind wedge MNPQR AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer Kl in accordance with the State plan.
Answer: A Explanation/Justification: Meets the requirements of the SRO only white paper paragraph E 3m bullet on page 7. SRO only in that it requires the implementation of administrative procedures that specify implementing emergency procedures. Specifically the offsite PAR which at BVPS is an SRO only task.
A. Correct. lAW 1/2-EPP-IP-4.1, Attachment A part 2. SRO must follow the attachment to part 2 and then follow part 2 to select the appropriate PAR.
B. Incorrect. Candidate will select this choice if they incorrectly apply part 2 of the attachment.
C. Incorrect. If the candidate uses the 35" wind speed in place of the OOS 150' wind speed and incorrectly applies part 1 of the attachment they will select this answer.
D. Incorrect. If the candidate incorrectly applies part 1 of the attachment and correctly applies the 500' wind speed to determine the sector, they will select this answer.
Sys # System Category KA Statement N/A N/A Generic Knowledge of emergency plan protective action recommendations.
KIA# 2.4.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate 112-EPP-IP-4. 1, Offsite Protective Technical 1/2-EPP-IP-4.1, Offsite Protective Actions Actions
References:
Attachment A Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 55.43(b)(5)
Objective:

ML14246A237

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