ML12299A182

Beaver Valley Unit 2 - Draft Written Exam (Folder 2)

ML12299A182
Person / Time
Site:	Beaver Valley
Issue date:	08/03/2012
From:	Rudolph W J FirstEnergy Nuclear Operating Co
To:	David Silk Operations Branch I
Jackson D E
Shared Package
ML12136A033	List: ML12142A240 ML12263A060 ML12298A500 ML12299A089 ML12299A110 ML12299A182 ML12299A202 ML12299A226 ML12299A229 ML12299A276 ML12299A280 ML12307A237 ML12314A266
References
TAC MEU01850
Download: ML12299A182 (100)

Text

Beaver Valley Unit 2 NRC Written E:Kam (2LOT8) 1. Given the following plant conditions:

• Reactor Power is 85%, steady state, all systems in NSA.

• A transient occurred resulting in a reactor trip and safety injection.

• RCS pressure is 1045 psig and LOWERING.

• RCS temperature is 545 of.

• Pressurizer level is 78% and RISING.

• Reactor Coolant Pumps are tripped. The Control Room Team is performing E-O, "Reactor Trip or Safety Injection" when the following plant conditions develop:

• RCS pressure is 1200 psig and slowly RISING.

• RCS temperature is 545 of.

• Pressurizer level is 32% and LOWERING.

Which ONE of the following is the cause of these changing plant conditions?

A. An open PORV has reseated.

B. A faulted steam generator has boiled dry. C. The size of the RCS leak has increased.

J. The turbine failed to trip and the MSIVs were closed. Answer: A Explanation/Justification: Correct. The candidate must have knowledge of the interrelations between a PRZR vapor space accident and sensors and detectors.

During situations where a steam vent path is established from the PRZR vapor space and where subcooling is not indicated, PRZR level may not be a true indication of RCS inventory.

The candidate must sort through the various indications provided by sensors and detectors and analyze these indications to determine a PORV has lifted and is no longer lifting (vapor space accident).

They must understand the interrelations of these indications to rule out the other choices. Incorrect.

PRZR level would act in the opposite way if the faulted S/G boiled dry. Incorrect.

RCS pressure would drop if the RCS leak size increased. Incorrect.

If the turbine failed to trip, PRZR level would act in the opposite direction due to initial plant cooldown until the MSIVs closed. Sys # System Category KA Statement 000008 Pressurizer Vapor AK2. Knowledge of the interrelations between the Pressurizer Sensors and detectors Space Accident Vapor Space Accident and the following:

KlA# AK2.02 KIA 2.7 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-53B.5.GI-11, Issue 2, Rev. 0, pg. 7 Simulator Response.

Question Source: Bank -Vision #46480 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55; Content: (CFR 41.7 145.7) Objective:

2S0S-6.4 41. Given a change in plant conditions due to a system or component failure, analyze the PRZR and PRZR Relief System to determine what failure has occurred.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 2. Given the following plant conditions: The Unit was operating at Full Power with all systems in NSA. A LOCA occurred and the Control Room Team transitioned to E-1, "Loss of Reactor or Secondary Coolant".

The following plant conditions exist: oRCS pressure is 550 psig and slowly DROPPING.

o Core Exit Thermocouple Temperatures are 472 OF and slowly DROPPING.

oRCS loop cold leg temperatures are 442 OF and slowly DROPPING.

o S/G pressures are 375 psig and slowly DROPPING.

oRCS 11 T is indicating UPSCALE. RCP's are NOT running. Based on these conditions, which ONE of the following identifies the source(s) of SI flow providing core cooling, AND what is the status of natural circulation?

A. High Head SI flow ONLY; Natural Circulation is occurring.

B. High Head AND Low Head SI flow; Natural Circulation is NOT occurring.

C. High Head SI Flow AND SI Accumulators; Natural Circulation is occurring.

J. High Head SI flow AND SI Accumulators; Natural Circulation is NOT occurring.

Answer: C Explanation/Justification: Incorrect.

Partially correct that High Head SI flow is a source of cooling, however, SI Accumulators are also a source. Correct natural circulation conclusion (refer to correct answer explanation) Incorrect.

Incorrect that natural circulation is not occurring.

Low Head SI flow is NOT a source of SI flow (refer to correct answer explanation) Correct. With RCS pressure at 550 psig, the High Head SI pumps &SI Accumulators (Begin to inject when RCS pressure drops < 600 psig) will be supplying SI flow for core cooling. The shutoff head for the Low Head SI pumps is about 1 i'8 psig so therefore will not be providing flow. Natural circulation is occurring because of the conditions specified on Attachment A-1. 7 are met. Both ES-1.1 & 1.2 reference Attachment A-1. 7 for verification of natural circulation flow in the LOCA series procedures making this question operational relevant. Incorrect.

Correct sources of SI flow. Incorrect natural circulation conclusion (refer to correct answer explanation) System Category KA Statement Small Break EA2 Ability to determine or interpret the following as they apply to a Existence of adequate natural circulation LOCA small break LOCA: KlA# EA2.37 KIA Importance 4.2 Exam RO References provided to Candidate None Technical 20M-53A.1.A-1.7, Issue 1C. Rev. 1, Pg. 2 2SQS-11.1, Rev. 16, Pg. 6 20M-53A.1.ES-1.1.

Issue 1C. Rev. 12, pg. 14 20M-53A1.ES-1.2.

Issue 1C. Rev. 10, Pg. 14 Question Source: Modified Bank -1 LOT8 NRC Exam Q#66 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR 43.5 I 45.13) Objective:

11. State from memory five conditions which indicate that natural circulation is occurring, lAW BVPS EOP Executive Volume.

Beaver Valley Unit 2 NRC Written E)(am (2LOT8) 3. Given the following plant conditions:

• The Unit is at Full Power with all systems in NSA.

• The thermal barrier heat exchanger for the "21A" RCP develops a 75 gpm tube leak.

• All systems function as designed.

Which ONE of the following describes the effect this leak will have on the Primary Component Cooling Water System (CCP)? The "21A" RCP thermal barrier ____ A. outlet isolation valve automatically closes on high pressure.

B. inlet & outlet isolation valves automatically close on high flow. C. outlet isolation valve automatically closes on high flow with the inlet isolated by a check valve. D. inlet isolation valve automatically closes on high flow with the outlet isolated by a check valve. Answer: C Explanation/Justification: Incorrect.

Correct that "A" RCP outlet thermal barrier isolation valve isolates, however, the isolation is on flow versus pressure.

Plausible because RCS pressure is higher than CCP. B. Incorrect.

There is only an outlet thermal barrier isolation valve that auto Correct. The candidate must understand the system design features and system interrelationship between CCP and a thermal barrier leak. accordance with OM Chapter 15/AOP at 58 gpm (significant tube leak -loss of Reactor Coolant Flow from system into CCP), the thermal barrier outlet isolation valve associated with the effected RCP will auto close. Incorrect.

Opposite of correct configuration.

Plausible if the candidate does not know the system interrelations.

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

KlA# AK2.08 KIA Importance 2.6 Exam Level RO 20M-15.1.0, Issue 4, Rev 1, pg. 13 & 14 References provided to None Technical

References:

20M-15.5, OP Manual Figure 15-3, Rev. 9 Candidate 20M-53C.4.2.6.8, pg. 1, Rev. 8 Question Source: Bank -Vision #33301 Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7/45.7)

Objective:

2S0S-15.1

27. Given a condition of excessive Reactor Coolant System RCP/CCP flow, summarize how the system will respond to the condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 4. Given the following plant conditions: The Unit is shutdown and cooled down to 235 of where Tavg is STABLE. All systems are aligned for these plant conditions. "A" RCP is running. The running charging pump experiences an overcurrent trip. The Control Room Team enters AOP 2.7.1, "Loss of Charging or Letdown".

Which ONE of the following actions will be REQUIRED within the next hour? Isolate letdown and establish excess letdown. Initiate boration to restore shutdown margin within limits. Perform seal injection surveillance to ensure seal injection flow meets TS 3.5.5, "Seal Injection Flow" requirements. Restore a charging pump to functional status to meet LRM 3.1.2, "Boration Flow Operating" requirements.

Answer: 0 Explanation/Justification: Incorrect.

AOP 2.7.1 does direct letdown to be isolated so therefore this is a correct action. It is also plausible that excess letdown is placed in service although not procedurally required.

There is also no 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit for this action to occur. Incorrect.

This is not an action directed by AOP 2.7.1, however, TS 3.1.1 does have a less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statement to initiate boration restore SDM within limits. Although plausible this is not an action that is necessitated based on the plant conditions provided since SDM not have changed nor have been effected since Tavg is Incorrect.

AOP 2.7.1 does direct action to check RCP seal injection flow. The TS 3.5.5 required action is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> action statement.

Also this applies in Mode 1,2, &3 and is not applicable in Mode Correct. The candidate must know that the plant is currently in Mode 4. The RO candidate is also required to know TS/LRM actions which are one hour or less from memory. LRM 3.1.2 is applicable in Mode 4 and requires the flowpath from tile refueling water storage tank via one charging pump to the RCS to be functional.

Condition C allows one hour to restore this flowpath from the RWST to functional status. The candidate must also have knowledge of AOP-2.7.1 actions in order to analyze and rule out the alternate choicl9S.

With RCS Temperature

< 240 F (enable temperature) only one charging pump is functional/operable and an alternate charging pump will need to be made functional.

Sys # System Category KA Stcltement 000022 Loss of Reactor Coolant Makeup Generic KnowlE!dge of less than or equal to one hour Technical Specification action statements for systems. KlA# 2.2.39 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.

7.1, Rev. 4, pg. 1-3 LRM BVPS Unit 2, Rev. 67, pg. 3.1.2-1 &2 LRM BVPS Unit 2, Rev. 67, pg. B 3.1.1-3.1.8 TS BVPS Unit 1 & 2, Amend 278/161, pg. 3.5.5-1 TS BVPS Unit 1 & 2, Amend 278/161, pg. 3.3.3-1 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 / 41.10 / 43.2 /45.13) Objective:

2S0S-7.1 24. For a given set of plant conditions, determine if the condition meets the criteria for entry into a less than or equal to one hour action statement in accordance with Technical Specifications.

Beaver Valley Unit 2 NRC Written E:xam (2LOT8) 5. Given the following plant conditions: The plant is in Mode 5 with the Pressurizer (PRZR) solid. The following indications occur: A1-2G. "INCORE INST ROOM/CNMT SUMP LEVEL HIGHNALVE NOT RESET' annunciated. 2DAS*Ll220. "Reactor CNMT Sump Level" reads 5.1" and is RISING. 2DAS*Ll222, "Reactor CNMT Sump Level" reads 5.1" and is RISING. 2RCS-Ll-462. "PRZR Cold Calib Level" is 70% and is rapidly DROPPING.

Based on these indications.

which ONE of the following procedures will be entered AND what action will be taken? (AOP 2.6.5. "Shutdown (AOP 2.10.1. "Loss of Residual Heat Removal A. Enter AOP 2.6.5 and actuate Safety Injection.

B. Enter AOP 2.6.5 and start all charging pumps. C. Enter AOP 2.10.1 and isolate letdown/known drain paths. D. Enter AOP 2.10.1 and check RCS Inventory and then go to AOP 2.6.5. \nswer: C Explanation/Justification: Incorrect.

Incorrect procedure since the plant is in Mode 5 versus Mode 4. Incorrect plausible action that is checked but not directed by this procedure. Incorrect.

Incorrect procedure since the plant is in Mode 5 versus Mode 4. Correct action directed by this procedure. Correct. The candidate must recognize from the indications provided that a leak into containment from the RHR or RCS is occurring.

They must also have knowledge based on these indications of the actions taken in accordance with the applicable procedure.

Both AOP 2.6.5 and 2.10.1 have entry conditions for uncontrollable PRZR level drop. AOP 2.10.1 is applicable in Mode 5 when not in a reduced inventory mid loop condition and would be entered. One of the actions taken is to isolate letdown and known drain paths to attempt stopping the loss of inventory.

AOP 2.6.5 is applicable in Mode 3 & 4 ONLY. The RO is required to know AOP entry conditions and understand overall mitigative strategies or sequence of events which occur. Incorrect.

Correct procedure entry with a plausible correct action to check inventory, however, since the plant is in Mode 5 a transition to AOP 2.6.5 is not applicable and therefore will not occur making this choice incorrect.

Sys # System Category KA Statement 000025 Loss of RHR AA 1. Ability to operate and 1 or monitor the following as they apply to Reactor building sump level indicators System the Loss of Residual Heat Removal System: KlA# AA 1.11 KIA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.1 0.1, Rev. 11, pg. 1,2,3,5, & 9 20M-53C.4.2.6.5, Rev. 18, pg. 1 & 2 20M-9.4.AAA, Rev. 4, pg 2, 5-7 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 145.5/45.6)

Objective:

2SQS-53C 7. Given a set of conditions, apply the correct AOP.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 6. Given the following plant conditions and sequence of events:

• The Plant is operating at 50% power.

• Control rods are in MANUAL.

• Pressurizer (PRZR) 2A & 2B Backup Heaters are in the ON position.

• The Pressurizer (PRZR) Master pressure controller output failed AS IS.

• A load rejection (step load decrease of 10%) occurs.

• No operator action is taken. Based on these plant conditions, what will be the impact of the load rejection on PRZR Spray Valve [2RCS*PCV455A]

position AND the two groups of energized PRZR Backup Heaters [2A & 2B]? 2RCS*PCV455A willlNITIALL Y _ (1) _. Energized PRZR heaters will _ (1) open (2) de-energize B. (1) further open (2) remain energized C. (1) remain as is (2) de-energize (1) remain as is 2 remain ener ized Answer: D Explanation/Justification:

(2) Incorrect.

Incorrect spray valve response.

Plausible if candidate does not recognize the impac.t of the Master pressure Controller failure. PRZR heaters will remain energized.

Plausible if the candidate believes PRZR level drops to 14% which cuts off PRZR heaters by interlock. Incorrect.

Incorrect spray valve response.

Without the malfunction it is correct that the valve would further open. Correct PRZR Backup Heater response. Incorrect.

Correct spray valve response.

PRZR heaters will remain energized.

Plausible because PRZR heaters are designed to turn off with increasing PRZR pressure. Correct. A load rejection results in an increase in RCS temperature. (Plant will not trip due to reactor power level) The Tavg increase will cause an expansion of water into the PRZR (Insurge) compressing the vapor space which in turn will increase PRZR pressure.

On increasing PRZR pressure, the Master Pressurizer Control System is designed to open the Spray Valves to lower PRZR pressure back to NOP (2235 psig). However, since the Master Pressure Controller has failed as is, it will not respond to the system parameters and therefore will not reposition Spray Valves open. Backup PRZR heaters 2A & 2B will remain energized because they are energized on and the master pressure controller has failed at a setpoin! which will not cause them to turn off regardless of what happens to PRZR pressure following the transient.

A 10% drop in power will result in a 3.S% PRZR level increase which is insufficient to turn all PRZR heaters ON (5% level increase needed). Sys # System 000027 Pressurizer Pressure Control System Malfunction KlA# AA1.01 References provided to Candidate Category AA 1. Ability to operate and I or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions:

KIA Importance 4.0 Exam Level Technical

References:

KA Statement PZR heaters, sprays, and PORVs RO 20M-SA. IF, Rev. 13, pg. 12,24 & 25 Question Source: Bank Question Cognitive Level: Objective:

2S0S-S.4 1 LOT8 NRC Exam 0#8 Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.7 145.5 I 45.S) 40. Given a specific plant condition, predict the response of the PRZR and Pressure Relief System control room indication and control loops, including all autornatic functions and changes in equipment status, for either a change in plant conditions or for an off-normal condition (ie: Process instrument failure).

Beaver Valley Unit 2 NRC Written Exam (2LOT8) You have been sent to LOCALLY open reactor trip/bypass breakers during an ATWS event due to the reactor not tripping from the control room. Which ONE of the following combinations of breaker positions will indicate the reactor is NOT tripped when you arrive at the Reactor Trip Breaker Panel? RTA = Reactor Trip Breaker "A" RTB = Reactor Trip Breaker "B" BYA = Reactor Trip Bypass Breaker "A" BYB = Reactor Trip Bypass Breaker "B" RTA BYA RTB BYB CLOSED OPEN CLOSED OPEN OPEN OPEN CLOSED CLOSED CLOSED OPEN OPEN OPEN OPEN CLOSED OPEN OPEN Answer: A Explanation/Justification:

\. Correct. The candidate must know the interrelationship between the Reactor Trip and Bypass Breakers.

During an A 1WS condition operators are sent to locally trip the reactor. The operators must know what combinations will successfully rElsult in a reactor trip in order to mitigate the adverse effect caused by an A 1WS condition.

All of the other combinations will indicate the reactor is t:ripped.

This is the only combination that indicates the reactor is NOT tripped. B. Incorrect.

This breaker configuration will result in a reactor trip. (Refer to correct answer explanation)

C. Incorrect.

This breaker configuration will result in a reactor trip. (Refer to correct answer explanation)

D. Incorrect.

This breaker configuration will result in a reactor trip. (Refer to correct answer explanation)

Sys# System KA Statement 000029 A1WS EK2 Knowledge of the interrelations between the and the following an A 1WS: Breakers, relays, and disconnects KJA# EK2.06 KIA Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

BVPS UFSAR Unit 2, Rev. 0, pg. 7.2-1 & 7.2-2 BVPS UFSAR Unit 2, Figure 7.1-1 & 7.1-7 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7 145.7) Objective:

3SQS-1.2 3. Describe the control, protection and interlock functions for the field components associated with Reactor Protection System Hardware, including automatic functions, setpoints and changes in equipment status as applicable.

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Beaver Valley Unit 2 NRC Written Exam (2LOT8) 8. Given the following plant conditions:

• A Steam Generator Tube Rupture in the "A" S/G has occurred.

• The Control Room Team is performing E-3, "Steam Generator Tube Rupture" actions.

• All systems function as designed.

Which ONE of the following will be the reason for verifying

[2BDG*AOV100A1], "21A S/G Blowdown Outside CNMT Isolation Valve" automatically CLOSED? To minimize A. S/G tube creep. B. radiological releases.

C. 8P between ruptured and non-ruptured S/G's. D. time to cover ruptured S/G U-tubes (promote thermal stratification).

Answer: B Explanation/Justification: Incorrect.

This is the reason for checking S/G water level during an ICC condition prior to starting RCPs. The candidate may confuse background document steps and isolating BID will increase S/G water level faster making this a plausible choice. Correct. The affected S/G BID Isolation valve automatically closes on a high radiation level. The candidate must know the reason for this isolation according to the E-3 BIG document.

This is the only automatic action in E-3 that has an automatic isolation signal provided by a process radiation monitor. Incorrect.

This is opposite of another reason for isolation of flow from the ruptured SfG. DP should be maximized as opposed to minimized. Incorrect.

Plausible that limiting blowdown will decrease the time to fill the S/G. E-3 does require S/G U-tubes covered prior to isolation of feedwater flow to the ruptured S/G. Establishing and maintaining water level above the U-tubes in the ruptured S/G promotes thermal stratification to prevent ruptured S/G depressuization.

According to the BIG this is the reason for checking ruptured S/G level, not for isolating BID. System Category KA Statement Steam Generator EK3 Knowledge of the reasons for the following responses as Automatic actions provided by each PRM Tube Rupture the apply to the SGTR: KlA# EK3.04 KIA Importance 3.9 Exam Level RO References provided to Candidate Technical

References:

20M-43A.AEF, Rev. 7, pg. 2 20M-53A.1.E-3, Issue 1C, Rev. 16, pg 4 & 5 20M-53BA.E-3, Issue 1C. Rev. 16, pg. 57 -59 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5 141.10 I 45.61 45.13) Objective:

3S0S-53.3

3. State from memory the basis and sequence for the major action steps of each EOP lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written (2LOT8) 9. Given the following plant conditions: The Unit is operating at 100% power with all systems in NSA. A steam line break occurs outside containment upstream of 21A Main Steam Isolation Valve (MSIV). A Main Steam Line Isolation (MSLI) signal occurs. Assume all systems function as designed and no operator action occurs. Given this event, which of the following is(are) the purpose(s) for Main Steam Line Isolation? To terminate the event as soon as MSLI occurs. To limit the blowdown to only one steam generator. To ensure a supply of steam is available to the Terry Turbine. A. 1 ONLY. B. 2 ONLY. C. 1,2, AND 3. D. 2 AND 3 ONLY. Answer: D 'xplanation/Justification:

A. Incorrect.

The MSLI will not terminate the event if the break is upstream of the MSIVs until complete S/G blowdown occurs. B. Incorrect.

Correct reason for MSIV closure according to references, however, not the only correct answer. C. Incorrect.

This is incorrect because #1 is incorrect.

Refer to A explanation. Correct. Correct reason for MSIV closure according to references.

Sys# System Category KA Statement 000040 Steam Line Rupture -AK3. Knowledge of the reasons for the following Operation of steam line isolation valves Excessive Heat Transfer responses as they apply to the Steam Line Rupture: KlA# AK3.01 KIA 4.2 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-21.1.B, Issue 4, Rev. 0, pg. 1 BVPS Unit 1 & 23.7.2 TS and Bases Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5,41.10/45.6/45.13) Objective:

2S0S-21.1

36. Describe the design basis for the Main Steam Supply System and associated major components as documented in the UFSAR.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 10. Given the following plant conditions: The plant has been operating at 100% power for 200 days with all systems in NSA. A complete Loss of Main Feedwater occurred. The Control Room Team has transitioned to FR-H.1, "Response to Loss of Secondary Heat Sink" and initiated Bleed and Feed. Subsequently, an AFW pump has been started and it is desired to recover Steam Generator (S/G) water level. S/G Primary side temperature is 550 of and all S/G WR levels are 5%. Which ONE of the following will be the method of recovering S/G water level AND associated reason why? S/G Water Level will be recovered by initially feeding _ (1) _ to ensure S/G _ (2)_, A. (1) all three S/Gs at s 50 gpm (2) tubes remain wetted and are fully covered before raising flowrate.

B. (1) all three S/Gs at s 100 gpm (2) thermal stress is minimized.

C. (1) only one S/G s 50 gpm (2) tubes remain wetted and are fully covered before raising flowrate.

). (1) only one S/G s 100 gpm (2) thermal stress is minimized.

Answer: 0 Explanation/Justification: Incorrect.

Incorrect number of S/Gs and flowrate.

The reason and flowrate are plausible if the candidate confuses the FR-H.1 background with ECA-2.1 background. Incorrect.

Correct flowrate and reason but incorrect number of S/Gs. Incorrect.

Correct number of S/Gs with incorrect flowrate.

The reason and flowrate are plausible if the candidate confuses the FR-H.1 background with ECA-2.1 background.background. Correct. The candidate must know the background document bases for FR-H.1 information on how to restore feedwater flow to a hot dry S/G and understand the operational effect if this is not performed properly.

Specifically.

step 28 background states that a hot dry S/G is defined as having primary side ofthe S/G temperature>

525 F and <14% WR level. The background further states that the SIG water level should be restored to one S/G at a time and at a minimal flowrate not to exceed 100 gpm to minimize thermal stress. Sys # System Category KA Statement 000054 Loss of Main AK 1. Knowledge of the operational implications of the following Effects of feedwater introduction on dry S/G Feedwater concepts as they apply to Loss of Main Feedwater (MFW): KlA# AK1.02 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.FR-H.1, Issue lC, Rev. 9. pg 85 & 86 20M-53B.4.ECA-2.1, Issue le. Rev. 11, pg 19 Question Source: New Question Cognitive Level: Lower-Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.8/41.10 I 45.3) Objective:

3S0S-53.3

3. State from memory the basis and sequence of major Action steps lAW BVPS EOP Executive Volume.

Beaver Valley Unit 2 NRC Written E)cam (2LOT8) 11. Given the following plant conditions:

The Unit is operating at 100% power when a Station Blackout causes a reactor Twenty five (25) minutes after the trip, power has been restored to Emergency Bus 2AE The Control Room team has transitioned to ECA-0.1, "Loss of All AC Power Recovery SI

• All Steam Generator (S/G) pressures are 1000 psig and STABLE.

• Reactor Coolant System (RCS) pressure is 2220 psig and slowly RISING.

• T-hot is 585°F in all three (3) loops and slowly RISING.

• Core exit thermocouples indicate 590°F and RISING.

• T-cold is 555°F in all three (3) loops and STABLE.

• All systems function as designed.

Based on these conditions, what is the status of RCS natural circulation heat removal? Natural Circulation cooling is _____ A. occurring and is being maintained by Condenser Steam Dumps. B. occurring and is being maintained by S/G Atmospheric Steam Dumps. C. NOT occurring and may be established by opening the S/G Atmospheric Steam Dumps. NOT occurring but forced cooling may be established by starting the "A" Reactor Answer: C Explanation/Justification: Incorrect.

Natural circulation conditions do not exist lAW Attachment A-1.7. Condenser Steam dumps are unavailable. Incorrect.

Natural circulation conditions do not exist lAW Attachment A-1. 7. Atmospheric steam dumps are not maintaining heat removal. Correct. Tcold is too hot for existing steam pressure.

Steam temperature and Tcold should be about the same if natural circulation is present. Without power to condenser cooling tower pumps, the condenser is unavailable and therefore atmospheric steam dumps must be used to increase steaming rate and thus establish natural circulation of the RCS through SIG cooling. Incorrect.

Correct that natural circulation does not exist, however, "A" RCP in not available since only Bus 2AE is available.

Sys # System Category KA Statement 000055 Station EK1 Knowledge ofthe operational implications ofthe 'following Natural circulation cooling Blackout concepts as they apply to the Station Blackout:

KlA# EK1.02 KIA Importance 4.1 Exam Level RO References provided to Candidate Steam Tables (Red) Technical

References:

20M-53A.1.ECA-0.1, Rev. 7. pg. 11 20M-53A.1.A-1.7.

Issue 1C, Rev. 1, pg. 2 20M-53A.1.A-5.1.

Rev. 1. pg. 1 Question Source: New Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR 41.8/41.10/45.3)

Objective:

12. State from memory the five conditions which indicate natural circulation is occurring lAW BVPS EOP Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 12. Given the following plant conditions:

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

• AS-3A, "REACTOR PROTECTION SYSTEM TRAIN B TROUBLE" has annunciated.

• [2CCP-FI107A], "21A RCP Thermal Barrier Flow is reading 48 gpm. * [2CCP-FI107B], "21 B RCP Thermal Barrier Flow is reading 0 gpm. * [2CCP-FI107C], "21C RCP Thermal Barrier Flow is reading 0 gpm. * [2CHS-FI130A], "Seal Injection Flow" to "A" RCP is reading 8.5 gpm. * [2CHS-FI124A], "Seal Injection Flow" to "B" RCP is reading 8 gpm. * [2CHS-FI127A], "Seal Injection Flow" to "C" RCP is reading 8 gpm. Based on these indications, which ONE of the following procedures will be entered? A. AOP 2.38.1A, "Loss of Vital Bus 1". B. AOP 2.38.1 B, "Loss of Vital Bus 2". C. AOP 2.38.1 D, "Loss of Vital Bus 4". D. E-O, "Reactor Trip or Safety Injection".

Answer: B Explanation/Justification:

Incorrect.

Incorrect but plausible distractor if the candidate believes Train B RPS is powered from Vital Bus 1. The "A" RCP Thermal Barrier Valve is powered from VAC Vital Bus 1. Correct. The candidate must analyze the stated plant conditions and determine which procedure to enter based on these conditions.

Train B Reactor Protection System Trouble can be caused by Loss of Power. Train B is supplied by either Vital AC Bus 2 or 4. An additional indication that the candidate needs to correctly narrow down the correct procedure is that AC Vital Bus 2 supplies power to the "B" & "C" RCP thermal barrier isolation valves which explains why they are closed. (The valve indications are powered by associated Vital Bus so therefore flow was used as opposed to valve position).

The RO is required to know AOP entry conditions. Incorrect.

Plausible power supply because the Train B RPS Trouble can be caused by either Vital Bus 2 or 4. Incorrect.

Plausible because if a loss of aU seal cooling occurs concurrently with a loss of CCP cooling to the thermal barrier, a reactor trip Sys # System Category KA Statement 000057 Loss of Vital AC Generic Ability to recognize abnormal indications for system operating Electrical Instrument Bus parameters that are entry-level conditions for emergency and abnormal operalting procedures.

KlA# 2.4.4 KIA Importance 4.5 Exam RO Level References provided to Candidate None Technical 20M-1.4.AAJ, Hev. 3, pg. 4 & 6

References:

20M-53C.4.2.38.1A , Rev. 4, pg. 1 & 2 20M-53C.4.2.38.1 B, Rev. 5, pg. 1 20M-53C.4.2.38.1 D, Rev. 2, pg. 1 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10 / 43.2 / 45.6) Objective:

2SQS-53C.1

2. State from memory the conditions or symptoms that would require entry in the AOPs.

Beaver Valley Unit 2 NRC Written Exam 13. Given the following plant conditions:

• The following annunciators are received:

o AS-9B, "125V DC Bus 2-2 Trouble" is acknowledged.

o A1-1C, "Vital Bus Inverter OperationlTrouble"

• DC BUS 2-2 VOLTS reads zero (O).

• All systems functions as designed.

• No operator action has yet occurred.

What will be the status of Battery Breaker [2BAT-BKR-2-2]

AND Battery Charger 2-2] control room indication? (Note this is not an all inclusive list of alarms present) Battery Breaker [2BAT-BKR-2-2]

__ (1) __Battery Charger [2BAT-CHG-2-2]

__ (2) __GREEN LIGHT RED LIGHT A. (1) NOT LIT (2) NOT LIT B. (1) NOT LIT (2) NOT LIT LIT C. (1) NOT LIT LIT (2) NOT LIT D. (1) NOT LIT NOT LIT (2) NOT LIT NOT LIT Answer: 0 Explanation/Justification: Incorrect.

This would be the indication for a Loss of Vital AC condition only. Incorrect.

Incorrect battery charger breaker indication, Incorrect battery breaker indication, Incorrect, Incorrect battery charger breaker indication, Incorrect battery breaker indication, Correct The candidate must be able to determine the operational implications of battery charger equipment and instrumentation as applied to a Loss of DC power. In order to have a loss of DC power both the battery charger and battery must be divorced from its associated bus, The candidate must analyze the annunciators and indications in the question stem and determine what the battery charger and battery breaker indications will be from the control room which is operationally relevant.

Both the battery breaker and battery charger will have no indication from the control room. All distractors are plausible based on various combinations of light configurations which are valid indications for other operational implications with the battery or battery charger equipment.

Sys # System Category 000058 Loss of DC AK 1. Knowledge of the operational implications of the following Power concepts as they apply to Loss of DC Power: KlA# AK1.01 KIA Importance 2.8 Exam Level References provided to Candidate Technical

References:

Ouestion Source: New KA Statement Battery charger equipment and instrumentation RO 20M-38.4.AAA, Rev. 7. pg. 2, 4. 9 &10 20M-39.4.AAE, Rev. 7, pg 2 -4 20M-53C.4.2.39,1B, Rev. 3, pg. 1 3S0S-39.1 Powerpoint Slides "'uestion Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR 41.8/41.10 I 45.3) Jbjective:

3S0S-38,1

18. Given a 125VDC configuration, and without reference material, describe the 125VDC control room response to the following malfunctions, including automatic functions and changes in plant status: Loss of AC Power, Loss of Station Battery, Loss of DC Power.

---

Beaver Valley Unit 2 NRC Written E:(am (2LOT8) 14. Given the following plant conditions: The plant was operating at 100% power with all systems in NSA. The Control Room Team manually tripped the plant based on excessive Steam Generator Tube Leakage. Currently they are performing actions in E-3, "Steam Generator Tube Rupture". While performing E-3 actions, the RO reports there are no Station Air Compressors running and instrument air pressure is dropping rapidly. Which ONE of the following is the reason for restoring instrument air compressors according to the E-3 background document?

To ensure A. normal letdown and charging are available.

B. excess letdown and alternate charging are available.

C. intact S/G Main Steam Isolation Valves can be closed. D. ruptured S/G Blowdown Isolation Valves can be closed. Answer: A ;xplanation/Justification: Correct. The candidate must know the reason why Instrument Air Compressors are verified running while in E-3. Without instrument air compressors a loss of instrument air will occur and the result is the inability to use normal letdown and charging as well as other support systems. Incorrect.

The BIG document specifically refers to restoring normal charging and normal letdown. Plausible because the EOP does allow alternate use of excess letdown if normal letdown in unavailable and alternate charging if normal charging is unavailable.

Incorrect alternate charging does not require instrument Incorrect.

The MSIVs are designed to fail closed on a loss of instrument air. It is plausible that the intact S/G MSIVs are used to support E-3 actions and could be construed as a support system. Since MSIV are already closed, it is not necessary to restore air to close these valves. Incorrect.

Unit 2 S/G blowdown valves are supplied by Containment Instrument air which is supplied via station instrument air. It is plausible that the ruptured S/G blowdown valves are used for support system restoration in ES-3.2. These valves require air to open versus close. Sys# System Category KA Statement 000065 Loss of AK3. Knowledge of the reasons for the following responses as Actions contained in EOP for loss of instrument air Instrument Air they apply to the Loss of Instrument Air: KlA# AKS.08 KIA Importance S.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53B.4.E-3, Issue 1C, Rev. 16, pg. 79 & 80 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.5,41.10 145.6/45.13) Objective:

3SQS-53.3

3. State from memory the basis and sequence of major action steps of each EOP procedure, lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2l0T8) 15. Given the following plant conditions and sequence of events: The Unit is operating at 100% power with all systems in NSA EXCEPT: There was a hydrogen gas leak on the Main Generator. The hydrogen leak has been isolated. Main Generator hydrogen gas pressure is 60 psig and STABLE. The Control Room has been notified by the DLC System Operations Control Center of possible grid instability and requests the control room maintain current power factor with maximum permissible megawatts. The US entered AOP Y:z.35.1, "Degraded Grid". The following Main Generator parameters are provided: Power Factor =.97 MVAR's OUT = 230 Using Figure 35-14, "Main Generator Capability Curve, what will be the MAXIMUM permissible megawatt output for the Main Generator? (Reference Provided)

The maximum permissible megawatt output for the Main Generator is ___ A. 790MW B. 850MW " '. 930MW D. 950MW Answer: C Explanation/Justification: Incorrect.

Plausible if the candidate incorrectly applies MVARs and correctly uses the 60 psig hydrogen curve. Incorrect.

Plausible if the candidate incorrectly applies MVARs and incorrectly uses the 45 psig hydrogen pressure curve. Correct. The RO requires the knowledge of how to implement the Main Generator Capability Curve as specified by the AOP for degraded Incorrect.

This is plausible if the candidate does not understand which side of the curve they must operate to protect the Main Generator.

In order to arrive at this number they would correctly apply the hydrogen pressure curve and have of MVARs. Sys # System 000077 Generator Voltage and Electric Grid Disturbances KlA# AA2.03 Category AA2. Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

KIA Importance 3.5 Exam Level References provided to Candidate 20M-35.5A 14 Technical

References:

Question Source: New Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: Objective:

2SQS-53C.1

7. Given a set of conditions, apply the correct AOP. KA Statement Generator current outside the capability curve RO Y:..OM-53CAA.35.1.

Rev. 8. pg. 4 20M-35.5A 14, Rev. 3, pg. 2 (CFR: 41.5 and 43.5/45.5,45.7, and 45.8)

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 16. Given the following plant conditions: A LOCA outside containment has occurred. The Control Room team is performing the actions in ECA-1.2, "LOCA Outside Containment" . Which ONE of the following indications will be used to determine if the leak has been isolated in accordance with ECA-1.2? A. RCS Pressure INCREASING.

B. RVLlS Indication INCREASING.

C. Safety Injection Flow DECREASING.

D. Aux Building Radiation Level INCREASING.

Answer: A Explanation/Justification: Correct In accordance with ECA-1.2, RCS pressure is used as an indication of whether leak isolation has occurred and determines transition to E-1 or ECA 1.1 will Incorrect.

Other E-1 series procedures use RVLlS as an indication but other factors would also change level. Incorrect.

As RCS pressure rises, ECCS flow drops, but indication is not used in ECA-1.2. Incorrect.

Auxiliary building or safeguards radiation levels are specified in ECA-1.2 but would be decreasing if the LOCA outside containment was isolated.

Sys# System Category KA Statement W/E04 LOCA Outside Generic Ability to perform specific system and integrated plant Containment proc.edures during all modes of plant operation.

KlA# 2.1.23 KIA 4.3 Exam Level RO Importance References provided to Candidate None Technical 20M-53A.1.ECA-1.2, Issue 1C, Rev. 1, pg. 4

References:

20M-53B.4.ECA-1.2, Issue 1C, Rev. 1, pg. 2, 5, & 6 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5/45.2 f 45.6) Objective:

3S0S-53.3

3. State from memory the basis and sequence for the major actions steps of each EOP procedure, lAW BVPS Executive Volume.

---

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 17. Given the following plant conditions and sequence of events: The Unit was operating at 100% power with all systems in NSA. A Complete Loss of Main Feedwater occurs resulting iln a reactor trip. The transient has also resulted in a break in the RCS which caused containment pressure to rise to 6.2 psig. The Control Room Team is performing FR-H.1, "Response to Loss of Secondary Heat Sink" actions. Which ONE of the following describes when Bleed and Feed is required to be initiated in accordance with FR-H.1? As soon as A. Wide Range in ALL S/Gs lowers to s 14%. B. Narrow Range Level in ALL S/Gs lowers to 0%. C. Wide Range in any TWO S/Gs lowers to s 32%. D. Narrow Range Level in any TWO S/Gs lowers to s 14%. Answer: C -xplanation/Justification: Incorrect.

Plausible but incorrect.

This criteria does meet bleed and feed criteria, however, this is not the soonest value. Incorrect.

Plausible if the candidate mistakes narrow range and wide range S/G water level and believes that bleed and feed is required as as NR S/G water level goes off scale Correct. The candidate must have the ability to operate or monitor operating behavior characteristics of the facility as applied to the loss of secondary heat sink. The candidate must know the continuous action criteria for establishing Bleed &Feed because of its importance as an alternative heat sink to prevent core uncovery and inadequate core cooling. WR S/G water level in two or more S/Gs is Bleed and Feed initiation criteria lAW FR-H.1 step 2 when WR S/G level lowers to s 32%. The candidate must recognize that> 5 psig containment pressure is an adverse containment number. Incorrect.

The candidate may confuse WR and NR S/G water level and also not recognize that adverse containment criteria exists. Sys System Category KA Statement Loss of EA 1. Ability to operate and / or monitor the following as they apply to Operating behavior characteristics of the facility.

Secondary the (Loss of Secondary Heat Sink) Heat Sink KJA# EA1.2 KJA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-H.1, Issue 1 C, Rev. 9, pg. 2 20M-53B.4.FR-H.1, Issue 1C, Rev. 9, pg. 48 & 49 Question Source: Bank -Vision #46864 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7/45.5/45.6)

Objective:

3. State from memory the basis and sequence of for the major action steps of each EOP procedure lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 18. Given the following plant conditions: A LOCA has occurred. Due to multiple equipment failures, the Control Room Team is performing actions of ECA-1.1, "Loss of Emergency Coolant Recirculation". Two (2) Charging/HHSI pumps and two (2) LHSI pumps are running. One (1) Quench Spray pump is running. Containment pressure is 13 psig and SLOWLY dropping. RWST Level is 20 inches and SLOWLY dropping.

Which ONE of the following describes the required action in accordance with ECA-1.1? STOP ALL pumps taking suction from the RWST and verify no backflow from the RWST to CNMT sump. STOP ALL pumps taking suction from the RWST and initiate secondary depressurization to facilitate SI accumulator injection. STOP ONLY ONE (1) HHSI and ONLY ONE (1) LHSI pump and initiate secondary depressurization to facilitate SI accumulator injection.

Secure the Quench Spray pump. STOP BOTH LHSI pumps and ONE (1) HHSI pump. Maintain Quench Spray pump running until containment pressure is < 11 psig and then add makeup to RCS from alternate sources. mswer: B Explanation/Justification: Incorrect.

Correct that all pumps are stopped. Incorrect plausible action, Correct. The RO candidate must know the overall mitigative strategy of ECA-1.1 and sequence of events. ECA 1.1 directs the operator to secure all pumps taking suction from the RWST when level is < 30 inches. Once stopped the procedure directs the operator to check if all intact S/Gs should be depressurized. Incorrect.

Incorrect but plausible that one HHSI and one LSHI pump are secured because one of the procedural mitigating strategies is to conserve RWST water and in fact the procedure does directs action to secure pumps. Correct that the crew will initiate secondary depressurization to facilitate SI accumulator injection.

Also correct that the Quench Spray pump is secured. Incorrect.

Incorrect but plausible action to maintain one pump running as noted above. Also plausible but incorrect that the Quench Spray pump is maintained running until containment pressure is < 11 psig. Normally by procedure this would be a correct action, Sys # System Category KA Stateme'nt W/E 11 Loss of Emergency EA2. Ability to determine and interpret the Adherence to appropriate procedures and operation within the Coolant Recirculation following as they apply to the (Loss of limitations in the Facility's license and amendments, Emergency Coolant Recirculation):

KlA# EA2.2 KIA Importance 3.4 Exam RO Level References provided to Candidate None Technical

References:

20M-53A.1.ECA-1.1, Issue lC, Rev. 10, pg. 1,38. 18 8. 19 20M-53A.l.ECA-1.1, Issue 1e, Rev. 10, pg, 1-4,6,14,578.

59 Question Source: Bank -2LOT5 NRC Exam Q# 56 Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR: 43.5/45,13)

Objective:

35Q5-53.3 State from memory the basis and sequence for the Major Action Steps of each EOP procedure lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 19. Given the following plant conditions: The plant is in Refueling Mode with systems aligned for core off-load. While lowering a spent fuel assembly into the Spent Fuel Pool, the assembly ruptures and releases ALL of the gasses from ALL of the rods in that assembly ONLY. [2RMF-RQ202], "Fuel Pit Bridge Area Radiation Monitor" goes into HIGH alarm. Based on these plant conditions, what will be the status of [2RMF-RQ301], "Fuel Building Ventilation Radiation Monitor AND reason why? [2RMF-RQ301

] (1) __ be in HIGH alarm because (2)_. A. (1) will (2) the iodine and xenon released from the fuel assembly will NOT be sufficiently scrubbed out by the water above the assembly.

B. (1) will (2) [2RMF-RQ301]

is designed to detect gamma radiation (GM tube). C. (1) will NOT (2) the iodine and xenon released from the fuel assembly WILL be sufficiently scrubbed out by the water above the assembly.

D. (1) will NOT (2) [2RMF-RQ301]

is designed to detect beta radiation (scintillation).

Answer: A Explanation/Justification: Correct. The candidate must be able to determine and interpret the occurrence of a fuel handling incident as it applies to a fuel handling accident.

Specifically to answer the question they must have fundamental knowledge of the type of detEJclors in the Fuel Pool building and associated design. They must also have knowledge of the UFSAR accident analysis regarding Fuel Handling Accidents.

The UFSAR analysis states quite clearly that activity will be released into the buildings (Fuel or Containment) for the a fuel handling accident of this type. Therefore both detectors will be in HIGH alarm. Incorrect.

Correct 2RMF-RQ301 response, however, the detector type is incorrect.

2RMF-301 is NOT a GM tube. Opposite of correct Incorrect.

The UFSAR analysis states quite clearly that activity will be release into the for a fuel handling accident of this type. Incorrect.

Even though the listed monitor type is correct (2RMF-301 is a scintillation detector), the fact that the area monitor went into an condition implies that enough activity was released into the area to raise the activity level sensed in the ventilation Sys # System Category KA Statement 000036 Fuel Handling AA2. Ability to determine and interpret the following as they apply to Occurrence of a fuel handling incident Incidents the Fuel Handling Incidents:

KlA# AA2.02 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical

References:

20M-43.4.ADF, Rev. 6, pg. 3 & 4 20M-43.4, Issue 1, Rev. 4, pg 1 BVPS Unit 2 UFSAR, Rev. 16, pg.15.7-2-4 GO-ATA-4.3, Rev. 6 pg. 57 & 58 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 43.51 45.13) Objective:

GO-ATA-4.3

2. Identify the "worst" case of initial conditions or, given a parameter, identify which direction of its magnitude would be "worse" for initial conditions for each listed accident.

Beaver Valley Unit 2 NRC Written E:Kam (2LOT8) Which ONE of the following RADIATION MONITOR detectors when in HIGH alarm will result in a Control Room Alarm and subsequent AUTOMATIC action? [2RMC*RQ201], "Control Room Area". [2RMR-RQ203], "Manipulator Crane Area". [2RMR*RQ202A], "Outside Personnel Hatch Area". [2RMR*RQ206], "In Containment High Range Area". Answer: A Explanation/Justification: Correct. Since every ARM detector provides an alarm in the control room, the interrelationship between the ARM alarms and detectors at location is an automatic action. The control room area monitor is the only area monitor that provides any automatic Incorrect.

This is an area radiation detector that has no automatic actions but does provide an alarm to the control room. Incorrect.

This is an area radiation detector that has no automatic actions but does provide an alarm to the control room. Incorrect.

This is an area radiation detector that has no automatic actions but does provide an alarm to the control room. Sys# System Category KA Statement 000061 ARM System AK2. Knowledge of the interrelations between the Area Radiation Detectors at each ARM system location Alarms Monitoring (ARM) System Alarms and the following:

KiA# AK2.01 KiA Importance 2.5* Exam Level RO References provided to Candidate None Technical

References:

20M-43.1.B, Issue 4, Rev. 1, pg. 4-6 20M-43.4.ADB, Rev. 7, pg. 2 20M-43.5.B.3, Rev. 2, pg. 2 20M-43.1.C, Rev. 4, pg. 25, & 49 luestion Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7 145.7) Objective:

8. Given a specific plant condition, predict the response of the Radiation Monitoring System control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off-normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 21. Given the following plant conditions: The Unit is operating at 100% Power with all systems in NSA. Annunciator A11-1B, "CABLE VAULT & ROD CONT AREA CABLE TUNNEL FIRE" alarms. A serious fire in the East Cable Vault and Rod Control Area 735' 6" is confirmed. Assume all automatic fire suppression systems function as designed.

Based on these plant conditions, which ONE of the following describes the impact on Fire Brigade personnel?

The major concern to Fire Brigade personnel entering the East Cable Vault is __ (1) __ due to __ (2) __ used to automatically extinguish the fire in this area. A. (1) asphyxiation from displacement of oxygen (2) Halon and C02 B. (1) flooding and subsequent electrocution (2) C02 and Water C. (1) asphyxiation from displacement of oxygen (2) C02 ONLY D. (1) flooding and subsequent electrocution (2) Water ONLY Answer: C Explanation/Justification: Incorrect.

Correct that asphyxiation is a major concern. Incorrect that halon is used in this space (refer to correct answer). Incorrect.

Water is not used in the East Cable Vault as part of any automatic suppression fire fighting systems and therefore flooding is not major Correct The candidate must know that C02 is the fire fighting agent used in the East Cable Vault to automatically distinguish fires. Water or Halon are NOT used in this area. The operational implications of a serious fire in the East Cable Vault is that C02 is a major concern when entering this space due to the safety hazards it may cause (ie: cardiac arrest or nervous system effects).

Note that there are manual water hose stations in the area, however, the question is asking about automatic actions, Also in Unit 2 there are other annunicators which will alarm, however, they are excluded from the question stem to preclude guiding the candidate toward the correct answer. Incorrect.

Water is not used in the East Cable Vault as part of any automatic suppression fire fighting systems and therefore flooding is not major concern. Plausible that water is used as a fire extinguishing agent and flooding would then become a Sys# System Category KA Statement 000067 Plant Fire AK1. Knowledge of the operational implications of the following Fire fighting On-site concepts as they apply to Plant Fire on Site: KlA# AK1.02 KIA Importance 3,1 Exam Level RO References provided to Candidate None Technical 20M-33.4,ACK.

Rev. 1, pg. 2, 3, & 5 20M-33.1.B, Rev, 5, pg. 4 20M-33.5.B.a, Issue 4, Rev, 0, pg. 1 20M-33.4.ACT, Rev. 2, pg. 2 & 3 Question Source: Bank -1 LOTS NRC Exam Q#22 Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.S/41, 10/45.3) Objective:

4. Given a change in plant conditions, describe the response of the fire protection system field indication and control loops, including all automatic functions and changes in equipment status. 11. Given a fire protection system alarm condition and using the ARP, determine the appropriate alarm response, including automatic and operator actions in the control room, Beaver Valley Unit 2 NRC Written Exam (2LOT8) 22. Given the following plant conditions: The Plant is operating at Full Power with all systems in NSA. A small fire develops in the Control Room requiring Control Room Evacuation. The Control Room Crew carries out control room actions in accordance with AOP 2.33.1A, "Control Room Inaccessibility". One of these actions included tripping [2FWS-P21A], "Main Feedwater Pump" prior to evacuation. Following evacuation AOP 2.33.1A directs locally tripping the running "B" Main Feedwater Pump after Auxiliary Feedwater flow is verified.

Which ONE of the following describes how this task will be accomplished according to AOP 2.33.1A? Go to Switchgear Room and trip open ____ pump control circuit 125VOC Bkr 8-1 on [PNL-OC2-08]

ONL local cubicle test switch on Bus 2C [4160 VAC Cub 2C1] ONLY. local cubicle test switch on Bus 20 [4160 VAC Cub 201] ANI2 Pump control circuit 125VOC Bkr 8-1 on [PNL-OC2-04] [2FWS-P21B1]

Pump Motor Breaker on Bus 2C [4160 VAC Cub 2C1] AND [2FWS-P21B2]

Pump Motor Breaker on Bus 20 [4160 VAC Cub 201]. Answer: D Explanation/Justification: Incorrect.

Plausible that DC power needs to be secured. This power is for the "AU MFW Pump. Incorrect.

This control switch will only operate if the breaker is in test, so therefore will not be successful nor is it directed by procedure. Incorrect.

incorrect control switch with plausible correct DC power to B MFW Pump, however, not required to be opened lAW AOP 2.33.1A. Correct. The candidate must have knowledge of how to trip the MFW pumps during a control room evacuation situation.

The AOP is enough to state that both motor breakers need to be opened and expects the operator has understanding of which breakers need to be The competent operator must fundamentally know that each MFW Pump has two motors and that each is tripped. The AOP has no action to condensate Sys # System Category KA Statement 000068 Control Room AA 1. Ability to operate and I or monitor the following as they apply to Local trip of main feed pumps and Condensate Evacuation the Control Room Evacuation:

pumps KlA# AA1.27 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical

References:

20M-53CA.2.33.1A , Rev.12, pg. 1, 5, &9 20M-24.3.C, Rev. 16,6 & 8 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR 41.7 145.5! 45.6) Objective:

2SQS-53C.1

7. Given a set of conditions, apply the correct AOP.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 23. Given the following plant conditions: A rapid load reduction from 100% to 65% power was performed approximately three hours ago. [2CHS*RQ101A], "Reactor Coolant Low Range Monitor" is in HIGH alarm. [2CHS*RQ1 01 B], "Reactor Coolant High Range Monitor" has just reached its HIGH alarm setpoint. Actions of 20M-43.4.AAC, "Radiation Monitoring Level High" have been completed. Actions of AOP 2.6.6, "High Reactor Coolant Activity" have been completed.

Which ONE of the following reflects the desired plant line-up as specified in AOP 2.6.6? A. Letdown demineralizers automatically bypassed.

B. Letdown automatically isolated on high radiation.

C. [2CVS-P21 AlB], "CNMT Vacuum Pumps" are BOTH running. D. [2DGS-P21A1B], "Primary Drain Transfer Pumps" are BOTH in PTL. Answer: D Explanation/Justification: Incorrect.

Plausible that the candidate may believe that letdown flow has been increased and this leads to higher temperature which auto bypasses the letdown demineralizers.

AOP 2.6.6 directs putting more demins in service and rE!ducing letdown flow which is the desired lineup. Incorrect.

Plausible that letdown is isolated on a high radiation condition.

The candidate must have knowledge that there is no automatic function provided by the radiation monitors which are in high alarm. Incorrect.

Plausible incorrect action which is opposite of that specified in AOP 2.6.6 which stops the containment vacuum pumps. The candidate may have a misconception about maintaining a negative pressure inside the containment to keep radiation from leaking to the outside environment.

They must understand that this action although it would maintain a more negative pressure would draw some of the activity to the outside making matters worse, while the overall intent of the procedure is to minimize exposur,e and reduce radiological concerns. Correct. The candidate must understand the desired plant lineup as specified in AOP 2.6.6 for a High RCS Coolant Activity condition.

One of the actions specified in AOP 2.6.6 is to place both PDT pumps in PTL. The RO is required to know the overall mitigative strategies of procedures.

Sys # System Category KA Statement 000076 High Reactor Coolant Activity Generic Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. KlA# 2.1.31 KIA Importance 4.6 Exam Level RO References provided to Candidate None Technical

References:

20M-53C.4.2.6.6, Rev. 3, pg. 1 -5 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 / 45.12) Objective:

2SQS-53C.1

7. Given a set of conditions, apply the correct AOP.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 24. Given the following plant conditions: The STA reports a Yellow path on CORE COOLING exists. The Unit Supervisor announces a transition to FR-C.3, "Response to Saturated Core Cooling".

Which ONE of the following is a mitigating strategy in A. Start RCPs and open all RCS vent paths. B. Depressurize SIGs to depressurize the RCS. C. Ensure RCPs stopped and open all ReS vent paths. D. Establish SI flow to maintain minimum RCS subcooling.

Answer: D Explanation/Justification: Incorrect.

This is one of the major action categories for FR-C.1 versus FR-C.3. FR-C.3 checks for open paths but does not open them. Incorrect.

This is a major action category for FR-C.1 &2, but not for FR-C.3. Incorrect.

Opening RCS vent is a major action category for FR-C.1. It is plausible that RCPs are stopped before running them with minimal RCS pressure. Correct. The candidate must be familiar with the basic purpose, overall sequence of events or overall mitigative strategy of Saturated Cooling, With knowledge of these procedures the RO demonstrates the ability to operate the plant and obtain desired operating results these emergency plant conditions.

A major action category for FR-C.3 is to establish SI flow and maintain minimum RCS

..ys # System Category KA Statement W/E07 Saturated EA 1. Ability to operate and 1 or monitor the following as they apply to Desired operating results during abnormal and Core Cooling the (Saturated Core Cooling):

emergency situations.

KlA# EA1.3 KIA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.FR-C,3, Issue 1C, Rev. 2, pg. 1 20M-53A.1.FR-C.1, Issue 1C, Rev. 5, pg. 1 20M-53A.1.FR-C.2, Issue 1C, Rev. 4, pg. 1 Question Source: New Question Cognitive Level: Lower-Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 I 45.51 45.6) Objective:

3. Explain from memory the basis and sequence for the Major Action Steps of each EOP procedure, lAW EOP Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 25. Given the following plant conditions: The plant was operating at Full Power with all systems in NSA. An unisolable Steam Line Break has necessitated a transition to FR-P.1, "Response to Imminent Pressurized Thermal Shock". This transition was based on an Orange CSF Status Tree condition.

Which ONE of the following describes the action that will be taken in FR-P.1 and the reason for this action? A. Depressurize the RCS to maximize SI flow to the core. B. Stabilize RCS pressure to minimize SI flow to the core. C. Depressurize the RCS to minimize pressure stresses on the reactor vessel. D. Stabilize RCS pressure to minimize pressure stresses on the pressurizer.

Answer: C Explanation/Justification: Incorrect.

Maximizing SI flow would increase the cooldown and increase temperature stresses on the reactor vessel. Depressurizing the RCS is a correct action for the incorrect reason but correct effect. Incorrect.

SI flow is terminated if possible, however reactor vessel pressure is reduced to a minimum, decreasing the pressure stresses on reactor Correct. The candidate must have knowledge of the actions to reduce pressure and temperature effects in FR-P.i and reasons for these actions. Specifically one of the major action categories is to depressurize the RCS to minimize pressure stress. According to the background document the reason for this action is to decrease pressure stress on the reactor vessel wall as much as possible. Incorrect.

System pressure is reduced to a minimum to decrease the pressure stresses on the reactor vessel versus pressurizer.

Sys# System Category KA Statement W/E08 RCS Overcooling

-EK3. Knowledge of the reasons for the following Facility operating characteristics during transient conditions, PTS responses as they apply to the (Pressurized Thermal coolant chemistry and the effects of temperature, Shock): pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

KlA# EK3.1 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical 20M-53A.1.F-O.4, Issue 1C, Rev. 0, pg. 1

References:

20M-53A.1.FR-P.1,lssue iC, Rev. 7, pg. 1, 12, & 19 20M-53B.4.FR-P.1, Issue iC, Rev. 7, pg. 37 & 38 Question Source: Bank -2LOT 4 NRC Exam 0#48 Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/41.10,45.6,45.13) Objective:

3808-53.3

3. 8tate from memory the basis and sequence for the major actions steps of each EOP procedure, lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 26. Given the following plant conditions: The plant was operating at 85% Power with all systems in NSA. A Loss of Offsite Power occurred resulting in a Unit trip. Both 4160VAC Emergency Busses are energized. Both Trains of RVlIS are available. Immediate actions of E-O, "Reactor Trip or Safety Injection" are complete and a transition to ES-0.1, "Reactor Trip Response" has been made. It is desired to begin and maintain a plant cooldown to Mode 5 at s 25 of/hr. Given these plant conditions, which ONE of the following procedures will be required to achieve Mode 5? A. Remain in ES-0.1. B. ES-0.2, "Natural Circulation Cooldown".

C. ES-0.3, "Natural Circulation Cooldown with Steam Void in Vessel (with RVlIS)". D. Applicable portions of 20M-52.4.

R.1. F, "Station Shutdown from 100% to Mode 5". Answer: B Explanation/Justification:

\. Incorrect.

This procedure is applicable if forced circulation exists. Plausible if the candidate does not recognize these plant conditions. Correct. The candidate must be able to determine facility conditions (ie: no RCPs available and natural circulation operations are applicable) and then determine which procedure should be selected to cool the plant down to Mode 5 at a specified cooldown rate. This is appropriate RO level knowledge because they are required to know the basic purpose and overall sequence of events that will occur or the overall mitigative strategy of a procedure. Incorrect.

This procedure will achieve Mode 5 ,however, is only entered when cooldown rate F/hr. RVLlS available was added to question stem to increase plausibility.

This is incorrect because C/D is to be maintained s25 F/hr. Incorrect.

This procedure will achieve Mode 5, however, will only be entered at the end of ES-O.1 if forced ReS cooling exists which it does not. Sys # System Category KA Statement E09 Natural EA2. Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Circulation the (Natural Circulation Operations) procedures during abnormal and emergency operations.

KlA# EA2.1 KIA Importance 3.1 Exam RO References provided to Candidate None Technical 20M-53A.1.ES-O.l, Issue 1 C, Rev. 8, pg. 1 & 13 20M-53A.l.ES-O.2, Issue 1C, Rev. 10, pg. 1 & 9 20M-53A.1.ES-O.3, Issue lC, Rev. 6, pg. 1 Question Source: New Question Cognitive Level: Higher -

10 CFR Part 55 Content: (CFR: 43.5 I 45.13) Objective:

3S0S-53.3

6. Given a set of conditions, locate and apply the proper EOP lAW BVPS Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 27. Given the following plant conditions:

• The Unit was operating at Full Power with all systems in NSA

• A Large Break LOCA occurred.

• The Control Room Team transitioned to E-1, "Loss of Reactor or Secondary Coolant".

• All systems function as designed.

• Containment pressure peaked at 35 psig and is now 10 psig and SLOWLY DROPPING.

• Indicated RWST Level is 380 inches and DROPPING.

• Assume no operator action related to equipment stated below. Which ONE of the following describes what equipment currently will be in service for the containment pressure reduction?

A Recirc Spray ONLY. B. Quench Spray ONLY. C. Recirc Spray AND Quench Spray. D. Quench Spray, Recirc Spray, AND Containment Air Recirculation Fans. Answer: C Explanation/Justification:

1\. Incorrect Correct that Recirc spray is in service. Incorrect that it is the only equipment in service. (refer to correct answer explanation) Incorrect.

Correct that Quench spray is in service. Incorrect that it is the only equipment in service. (refer to correct answer explanation) Correct. The candidate must analyze plant conditions and determine based on these conditions what components are functioning to automatically reduce high containment pressure caused by a LBLOCA. Specifically, the candidate must know that the Quench Spray pumps will start on a CIB signal which is caused when containment pressure increases above 11 psig. Note that current containment pressure is below 11 psig, but without operator action, these pumps will continue to run. E-1 directs the operator to secure these pumps at < 8.5 psig. The RSS pumps start upon an CIB signal plus RWST level < 381". At 380" these pumps are designed to AUTO start and therefore are running. The containment air recirculation fan automatically trips on a SI and/or sump level. Incorrect.

Containment air recirculation fans are tripped due to CIA actuation.

The other two are correct. (refer to correct answer explanation) System Category KA Statement High EK2. Knowledge of the interrelations between the (High Containment Components, and functions of control and safety Containment Pressure) and the following:

systems, including instrumentation, signals, Pressure interlocks, failure modes, and automatic and manual features.

KlA# EK2.1 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical

References:

20M-13.1.B, Rev. 3, pg 2 25Q5-13.1 Powerpoint, Rev. 17 Question Source: Bank 1 LOT7 NRC Exam #14 Question Cognitive Level: Higher Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 145.7) Objective:

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 28. Given the following plant conditions: The Unit is operating at Full Power with all systems in NSA. 21 B RCP UBLO CLG WTR DISCH FLOW, [2CCP-FT104B]

indicates 185 gpm and is SLOWLY RISING. 21 B RCP LBLO CLG WTR DISCH FLOW, [2CCP-FT106B]

indicates 9.6 gpm and is SLOWLY RISING. Containment temperatures as indicated on [2LMS*TI1 00-1,2,7,13,14, 15] are STABLE. [2CCP-Ll100A(B)], "CCP Surge Tank Level" is SLOWLY DROPPING.

If the leak is on the inlet flange side of the 21 B RCP Stator Aiir Cooler Heat Exchanger, which of the following control room indications will confirm this leak location?

21 B RCP Stator Clg Water Disch Flow [2CCP-FT -1 21 B RCP Stator Winding temperatures on

[2CCP-FT-1 [2RCS-TR-448Bl A.

decreasing B.

decreasing C.

remains the same r remains the same Answer: C Explanation/Justification: Incorrect.

If the candidate believes the flow transmitter is located on the inlet side of the HX th'9n it is plausible that 2CCP-FT-105B flow will be increasing.

They may also have a misconception that if flow is increasing that stator temperature will decrease. Incorrect.

Correct flow indication.

Incorrect but plausible stator temperature if the leak was on the outlet side of the HX where there would be more actual flow through the air cooler. Correct. The candidate must be able to predict or monitor changes to 21 B Rep stator winding temperature and flow based on a leak upstream of the stator air cooler HX. They must understand the system layout and location of the flow tram;mitter in relation to the air cooler HX to derive the correct answer. Based on system design a large amount of air flow through the RCP Stator Air Cooler is from the Containment Recirculation Fans. The stem of the question states that containment temperature remains stable, so therefore the impact of this system leak is minimal. Increasing is not used a distractor for stator winding temperature because it is difficult to asceltain the exact amount of cooling loss that would be needed to cause a stator temperature increase and could be challenged. Incorrect.

If the candidate believes the flow transmitter is located on the inlet side of the HX th'9n it is plausible that 2CCP-FT-1 05B flow will be increasing.

Potentially correct stator winding temperature response.

Refer to correct answer explanation.

Sys # System Category 003 Reactor A 1 Ability to predict and/or monitor changes in parameters (to Coolant prevent exceeding design limits) associated with operating the Pump RCPS controls including:

KlA# A 1.03 KIA Importance 2.6 Exam Level References provided to Candidate None Technical

References:

Question Source: New KA Statement RCP motor stator winding temperatures RO Op Manual Fig. 15-3 20M-6.1.E, Rev. 6, pg. 51 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/45.5)

Objective:

2S0S-6.4 37. Given a specific plant condition, predict the response of the Reactor Coolant Pump and support system control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off-normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 29. Given the following plant conditions:

• Both Units are operating at 100% Power with all systems in NSA.

• A Loss of Offsite Power coincident with a Reactor Trip.

• The 2-1 EDG did NOT auto start. 138 KV Bus 1 & Bus 2 are verified de-energized.

• All systems function as designed.

• No operator actions have occurred.

Which ONE of the following will be the status of power to [2WTD-P23A1B], "Demineralized Water Pumps"? [2WTD-P23Al A. Has power B. Has power C. Has NO Power D. Has NO Power Answer: A :xplanation/Justiflcation:

[2WTD-P23B]

Has power Has NO Power Has power Has NO Power Correct. The candidate must analyze the plant conditions provided and determine the status Ctf power to the Demineralized Water Pumps (Primary Makeup Pumps). 2Wro-P23A is powered from MCC-2-23 which is powered from Bus 1 G. 2 WTD-P23B is powered from MCC-2-26 which is powered from Bus 1 H. On a Loss of Offsite Power, the ERF Black EDG will start and auto close onto both busses (1 H & 1G). Incorrect.

Correct that 2WTD-P23A has power. Plausible that 2WTD-P23B does not have power if the candidate does not understand the ERF Substation operations or is distracted by the 2-1 EDG not starting. Incorrect.

Correct that 2WTD-P23B has power. Plausible that 2WTD-P23A does not have power if the candidate does not understand the ERF Substation operations or is distracted by the 2-1 EDG not starting. Incorrect.

Plausible if the candidate does not know the power supplies or understand the impact of stated plant conditions.

Sys# System Category 004 Chemical and K2 Knowledge of bus power sup plies to the following:

Volume Control KlA# K2.02 KIA 2.9 Exam Level Importance References provided to Candidate None Technical

References:

Question Source: New KA Statement Makeup pumps RO 20M-32.3.C, Rev. 2, pg. 4 &5 3SQS-58E.1 Powerpoint.

Rev. 9 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55* Content: (CFR: 41. 7) Objective:

24. Given an under voltage condition, predict the response of the ERF Electrical Distribution System and how the plant configuration will change as a result of the electrical systems actions.

Beaver Valley Unit 2 NRC Written (2LOT8) 30. Given the following plant conditions:

• The Unit is in Mode 3. * [2CHS*LCV115A], "VCT Level Cont Divert to Degas" Controller is in AUTO. * [2CHS*LCV115A], "VCT Level Control Valve" control switch is in AUTO. * [2CHS*LCV112], "VCT Level Cont Divert to CLNT RCVY Valve" Controller is in AUTO.

• VCT level Control Setpoint for [2CHS*LCV115A]

is at 0% (HIC).

• VCT Level is currently 75% and DROPPING.

• The system is functioning as designed with no operator intervention.

Which ONE of the following will be the current status of [2CHS*LCV115A] [2CHS*LCV112]

based on system [2CHS*LCV115A]

will be __ (1) __ AND [2CHS*LCV112]

will be __ (2) __. A. (1) partially diverted to the Degasifer (2) fully diverted to Unit 1 Coolant Recovery Tanks. B. (1) fully diverted to the VCT (2) fully diverted to Unit 1 Coolant Recovery Tanks. C. (1) fully diverted to the Degasifer (2) partially diverted to Unit 1 Coolant Recovery Tanks. ). (1) partially diverted to the VCT (2) partially diverted to Unit 1 Coolant Recovery Tanks. Answer: A Explanation/Justification: Correct. The candidate must be familiar with CVCS design and interlocks associated with VCT level and VCT Diversion valve positions as well as system flowpaths to answer this question.

By design 2CHS*LCV115A will start to divert @ 65% and Full Divert @ 80%, so therefore at 75% is partially diverted to the Unit 1 Degasifier and partially diverted to the VCT. 2CHS*LCV112 will start to divert @ 55% and will full divert at 70%, so therefore will be full diverted @ 70% to the Unit 1 Coolant Recovery Tanks. Incorrect.

Correct that 2CHS*LCV115A is diverted to the VCT but only partially.

Correct that 2CHS*LCV112 is fully diverted to the Unit 1 CRTs. Incorrect.

2CHS*LCV115A is partially diverted to the degasifier.

2CHS*LCV112 position VCT level 55 -70%. Incorrect.

Correct 2CHS*LCV115A valve position.

2CHS*LCV112 position represents VCT level 55 -70%. Sys # System Category 004 Chemical and K4 Knowledge of CVCS design feature(s) and/or interlock(s)

Volume Control which provide for the following:

KlA# K4.14 KIA Importance 2.8* Exam Level References provided to Candidate None Technical

References:

Question Source: New KA Statement Control interlocks on letdown system (letdown tank bypass valve) RO 2SQS-7.1 Powerpoint Slides Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR: 41.7) Objective:

2SQS-7.1 18. Describe the control and interlock functions for the control room components associated with the CVCS system including automatic functions, setpoints and changes in equipment status as applicable:

HI-LO VCT level, RCS Makeup controls.

Beaver Valley Unit 2 NRC Written (2LOT8) 31. Given the following plant conditions: The Plant is in Mode 4 cooling down for a Refueling Outage. 20M-10.4.A, "Residual Heat Removal System Startup" is in progress. RHR Inlet Temperature as read on [2RCS-TR604A], "RHR Hx DiffTemp Recorder" is reading 290°F. Prior to Starting [2RHS*P21A], "A" RHS Pump you review procedure Precautions and Limitations.

Which ONE of the following will be the MINIMUM allowable suction pressure AND maximum allowable system flow AND reason for these limitations?

tE!.gure 10-11 Provided)

The minimum allowable suction pressure will be The maximum allowable system flow will be The reason for these limitations will be to ensure A. (1) 80 psig (2) 3550 gpm which excludes RHR Pump mini-flow ONLY. (3) adequate flow to the core to ensure core cooling. B. (1) 90 psig (2) 4000 gpm which includes RHR Pump mini-flow ONLY. (3) adequate flow to the core to ensure core cooling. (1) 80 psig (2) 4000 gpm which includes RHR Pump mini-flow and letdown flow. (3) adequate available NPSH for continuous RHR pump operation.

D. (1) 90 psig (2) 3550 gpm which excludes RHR Pump mini-flow and letdown flow. (3) adequate Ciiyailable NPSH for continuous RHR pump operation.

Answer: C Explanation/Justification: Incorrect.

Correct minimum allowable pressure.

Incorrect plausible maximum flow limit. This is the limit for two pump operation if vented. Incorrect that RHR pump mini flow is excluded from the flow limits. Also incorrect but plausible reason f':lr these limitations. Incorrect.

Incorrect plausible value if the candidate misreads 300 F or thinks they need to be cn the right side of the curve. Correct maximum flow. Correct that RHR pump mini flow is included in the flow limits, however, letdown flow is also included.

Also incorrect but plausible reason for these limitations. Correct. The candidate must be familiar with RHS precautions and limitations reasons and be able to apply them. Specifically, they must be able to determine the minimum suction pressure is 80 psig when RHR HX DT is 290 F. The maximum system flow to ensure adequate NPSH is 4000 gpm which does include RHR pump mini flow and letdown flow as part of the flow limits. The reason for these limitations is to ensure adequate NPSH for continuous RHR pump operation according to 20M*10.4.A. Incorrect.

Incorrect plausible value if the candidate misreads 300 F or thinks they need to be on the right side of the curve. Incorrect maximum flow. This is the limit for two pump operation if vented. Incorrect reason for limitations (opposite of correct Sys # System Category KA Statement 005 Residual Heat Generic Ability to explain and apply system limits and Removal precautions.

KlA# 2.1.32 KIA 3.8 Exam Level RO Importance

'teferences provided to 20M-10.5.A.11 , Figure 10*11 Technical

References:

20M-10.4.A, Rev. 38, pg. 3 &4 ;andldate 20M-10.5.A.11 , Issue 4, Rev. 0, pg. 1 Question Source: Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR: 41.10/43.2/45.12Objective:

2505-10.1

8. Given a set of plant conditions and appropriate procedure(s), apply the operational sequence, parameter precautions and limitations, and cautions & notes applicable to the completion of the task activities in the field.

Beaver Valley Unit 2 NRC Written E)(am (2LOT8) 32. Given the following plant conditions: The plant is in Mode 4 cooling down for a refueling outage. Both trains of Residual Heat Removal System (RHS) are in service. RCS Temperature is 320 of and DROPPING. A 12-1 D, "Safety Injection Signal" has annunciated. The Control Room Team confirms an Inadvertent Saf*!ty Injection Actuation occurred and enters the appropriate procedure. All systems function as designed.

Which ONE of the following will be the status of the RHS system flow? (assume no operator action) A. RHS flow is affected because BOTH RHS pumps trip. B. RHS flow continues to return to RCS "A" & "C" cold leg loops. C. RHS flow continues to return to RCS "B" & "C" cold leg loops. D. RHS flow is affected because all four inlet isolation valves AUTO close. Answer: C Explanation/Justification:

Incorrect.

This would be correct if a CIS were to occur. Since an inadvertent 81 occurred, there, is no reason to believe a high containment pressure condition exists. Incorrect.

Correct that RHS flow continues, however, the candidate may not know the SIS interrelationship is back to the B versus A cold common Correct. An SIS will not impact RHR system flow directly unless containment pressure were to increase greater than 11 psig (CIB) or a manual CIB signal were generated.

AOP 2.6.9, "Inadvertent SI Actuation

< 350 F does check RHR status because CCP flow will be effected to RHS because SIS causes a CIA which causes an isolation of CCP cooling to RHS HX's. RHS flow to the common loops where SIS and RCS tie together is however unaffected.

The candidate must have knowledge of specific SIS signal calJse/effect relationships as well as the physical connection where they tie into a common SI return line to the RCS. Incorrect.

The inlet isolation valves do have auto close signals on high system pressure as opposed to an SI signal. Sys # System KA Statement 005 Residual Heat K1 Knowledge of the physical connections and/or cause/effect SIS Removal relationships between the RHRS and the following systems: KlA# K1.13 KIA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

2SQS-10.1 Powerpoint OP Manual Fig. 10-1, Rev. 16 20M-53C.4.2.6.9, Rev. 0, pg. 7 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.2 to 41.9/45.7 to 45.8) Objective:

18. Given a specific plant condition, predict the response of RHR System control room indications and control loops, including all automatic functions and changes in equipmeint status, for either a change in plant condition or off normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) Which ONE of the following ECCS actuation signals will directly result in a trip of the [2FWS-P21A21, Main Feedwater Pump Motor? Train "8" Feedwater Isolation Signal. Train "8" Partial Feedwater Isolation Signal. Train "A" Low Steam Line Pressure Safety Injection Signal. Train "A" Low Pressurizer Pressure Safety Injection Signal. Answer: A Explanation/Justification: Correct. According to UFSAR Logic and 20M-24.1, A FWI signal from Train B will directly trip the 2FWS-P21A2 motor. This FWI signal is generated by ECCS actuation.

The KJA is met because the candidate must know that the ECGS actuation signal has a cause/effect relationship with the MFW system. Incorrect.

A partial FWI signal will only close the MFRV's but will not trip the FW pump. Incorrect.

The "A" Low Steam Line Pressure SI signal will indirectly trip the 2FWS-P21A1 mot()r not the 2FWS-P21A2 motor. (Train specific) Incorrect.

The "A" Low PRZR Pressure SI signal will indirectly trip the 2FWS-P21A1 motor not the 2FWS-P21A2 motor. (Train specific)

Sys# System Category KA Statement 006 Emergency K1 Knowledge of the physical connections and/or cause/effect MFWSystem Core Cooling relationships between the ECCS and the following systems: K1A# K1.07 KJA Importance 2.9* Exam Level RO References provided to Candidate None Technical

References:

20M-24.1.D, Rev. 6, pg. 10 UFSAR BVPS Unit 2 Figure 7.3-13 & 18 'uestion Source: New ..luestion Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9/45.7 to 45.8) Objective:

2S0S-24.1

35. Given a MF, StU FW, AFW or SGWLC system configuration and without referenced material, describe the associated system control room response to the following actucltion signals, including automatic functions and changes in equipment status as applicable:

Safety Injection.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 34. Given the following plant conditions: The plant was operating at 100% power with all systems in NSA. A 25% load rejection occurred. A4-3H, "PRESSURIZER RELIEF TANK TROUBLE" annunciates. PRT Temperature is 150 of and RISING. PRT Pressure is 18 psig and SLOWLY RISING. The RO suspects a PRZR PORV or Safety Valve opened and is now stuck partially OPEN. Which ONE of the following indications will confirm a Safety Valve is the cause of leakage AND what AUTOMATIC action will reduce PRT pressure if NO OPERATOR action were to occur? PRZR Safety Relief line temperatures

_ PRT Pressure will be reduced __ (2) A. (1) are RISING.

(2) when the PRT Rupture disc(s) blow. B. (1) are consistent with PRZR temperature.

(2) when the PRT Relief Valve(s) open. C. (1) are RISING. (2) when [2RCS-AOV519], "Makeup Water to PRT Valve" opens ONLY. D. (1) are consistent with PRZR temperature.

(2) when BOTH [2RCS-AOV519]

AND [2RCS-MOV516]

PRT Spray Valves open. Answer: A Explanation/Justification: Correct. The candidate must be able to analyze the conditions provided and apply system knowledge.

For the conditions provided a leaking PRZR or Safety Valve do discharge to the PRT and rising temperatures are indicative of discharge to the PRT. Correct that PRZR Safety Relief line temperature is rising. The PRT rupture disk(s) will automatically rupture to reduce PRT Incorrect.

Incorrect but plausible misconception of isenthalpic processes (TMI). Correct that the PRT relief valve functions to lift before the rupture disk(s) is (are) blown. Incorrect.

Correct PRZR Safety line temperature response.

2RCS-AOV519 is procedurally used to lower pressure.

however, it is not an automatic action in Unit 2. Incorrect.

Incorrect PRZR Safety line temperature response.

2RCS-AOV519 and 2RCS-MOV516 are procedurally used to lower however, it is not an automatic action in Unit Sys# System Category KA Statement 007 Pressurizer A3 Ability to monitor automatic operation of the PRTS, Components which discharge to the PRT Relief/Quench Tank including:

KlA# A3.01 KIA Importance 2.7* Exam RO References provided to Candidate None Technical 2SQS-6.4 Powerpoin!

Diagram, Rev. 15 20M-6.1.C, Rev. 5, pg. 34 & 35 20M-6.1.0, Rev. 3, pg. 7 -9 20M-6.4.AAY, Rev. 10, pg. 2 & 3 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7 I 45.5) Objective:

2SQS-6.4 10. Given a specific plant condition, predict the response of the PRZR & PRZR Relief System control room indications and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or off-normal condition:

excessive primary plant leakage.

Beaver Valley Unit 2 NRC Written E:xam (2LOT8) 35. Given the following plant conditions: The Unit is at Full power with all systems in NSA. A6-1H, "PRIMARY COMPONENT COOLING WATER SYSTEM TROUBLE" is received and acknowledged. [2CCP-Ll-100A

& 100B] "CCP Surge Tank level" is 70 inches and slowly RISING. A2-5F, "REACTOR COOLANT PUMP COOLING WATER TROUBLE" is received several minutes later. No operator actions occur and plant systems function as designed.

Which ONE of the following CCP flow indications confirms the cause of these plant conditions?

A. 2CCP-FT106A, "RCP 21A LBLO CLG FLW LOW". B. 2CCP-FT107A, "RCP 21A TH BARR FLW HIGH". C. 2CCP-FT104A, "RCP 21A UBLO CLG FLW LOW". D. 2CCP-FT106A, "RCP 21A LBLO CLG FLW HIGH". Answer: B Explanation/Justification: Incorrect.

This is a plausible alarm which will cause A2-SF to annunciate, however, A6-1H annunciated due to high surge tank level and a low flow condition will not cause this condition and is more indicative of a system rupture in which case CCP surge tank level would be reading low. Correct. The candidate must be able to predict based on the set of annunciators and paramet'9r provided that the only correct CCW flowrate which can cause this set of conditions is high thermal barrier flow rate. AS-1 H has various inputs. The candidate must deduce that the high surge tank level was the cause of this alarm. although several other conditions can cause AS-1 H to annunciate.

They must also understand that the only condition related to the RCP which will cause level to increase is a leak from the RCP thermal barrier into the CCP system. Incorrect This is a plausible alarm which will cause A2-5F to annunciate, however. A6-1H annunciated due to high surge tank level and a low flow condition will not cause this condition and is more indicative of a system rupture in which case CCP surge tank level would be reading low. Incorrect.

A high flow condition will not bring A2-SF into alarm. The candidate must understand that even if there were a high flow condition. this flow rate would not impact the CCP surge tank level and is not due to an intersystem Sys# System Category KA Statement 008 Component A 1 Ability to predict and/or monitor changes in parameters (to CCW flow rate Cooling prevent exceeding design limits) associated with operating the Water CCWS controls including:

KlA# A1.01 KIA Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

20M-15.4.AAC, Rev. 6, pg. 2, 4 & 5 20M-S.4.AAG, Rev. 9 pg. 2 -4 OP Manual Figure 15-1 & 3 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5 f 45.5) Objective:

2505-15.1 Given a condition of excessive reactor coolant system RCP CCP flow, summarize how the system will respond.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 36. Given the following plant conditions: The Unit is at 100% power with all systems NSA EXCIEPT: 2RCS-PCV-455A, "PRZR Spray Valve" is in its FAIL position due to a broken air line AND the Proportional Heater is in PTL. Reactor Coolant System (RCS) Pressure is 2235 psig and STABLE. RCS Tavg is 578°F and STABLE. 2RCS*PT444, "Pressurizer (PRZR) Control Channel", fails HIGH over a ONE (1) minute period. With no operator action, which ONE of the following describe-s how the PRZR Pressure Control System will respond? A. TWO (2) PRZR PORVs will be OPEN. B. ONE (1) PRZR PORV and ONE (1) PRZR Spray Valve will be OPEN. C. ALL PRZR BtU heaters will be ON and BOTH PRZR Spray Valves will be CLOSED. D. ALL PRZR BtU heaters will be OFF and ONE (1) PRZR Spray Valve will be in NSA position.

Answer: B Explanation/Justification:

1\. Incorrect.

These indications are indicative of 2RCS'PT445 failing in the high (lirection.

Plausible if the candidate confuses the PRZR pressure control system inputs or does not understand the impact of the items which are OOS for this system. Correct. A failure of 2RCS'PT444 in the high direction will typically result in 2RCS-PCV-455C opening and both PRZR spray Valves failing open. 2RCS-PCV-455A fails closed on a loss of air, therefore only 2RCS-PCV-455B will Incorrect.

This is indicative of 2RCS'PT444 failing in the low direction. Incorrect.

Correct PRZR BIU heater response except that NSA two heaters will be ON, so therefore not all heaters will be OFF. At 100% power, 2RCS-PCV-455A is typically slightly open. Since it is failed closed, it is not NSA. The other spray valve will be opening and it is plausible based on the integral impact of the failure on the master pressure controller which makes it difficult to ascertain the exact valve position.

Sys# System Category KA Statement 010 Pressurizer KS Knowledge of the effect of a loss or malfunction of the PZR sprays and heaters Pressure Control following will have on the PZR PCS: KlA# KS.03 KIA Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

20M-S.4.IF , Rev. 13, pg. 1S -21 &24 Question Source: New Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR: 41.7/45.7)

Objective:

2S0S-S.4 27. Given a change in plant conditions, describe the response of the PRZR and Pressure Relief System field indication and control loops, including all automatic functions and changes in equipment status.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 37. Given the following plant conditions: The Unit is operating at 45% power with all systems in NSA for this power level. Testing is in progress on Train "B" of SSPS. The "B" Reactor Trip Bypass Breaker is racked-in and closed. An instrument malfunction results in a reactor trip from the Train "A' Reactor Protection System ONLY. All systems function as designed. No operator action occurs. Which ONE of the following will be the response of the Reactor Protection System AND the effect on the Turbine Generator? Only the "A" Reactor Trip Breaker will open; The Turbine Generator will trip. Only the "A" Reactor Trip Breaker will open; The Turbine Generator will NOT trip. Both the "A" Reactor Trip Breaker AND the "B" Reactor Trip Bypass Breaker will open; The Turbine Generator will trip. Both the "A" Reactor Trip Breaker AND the "B" Reactor Trip Bypass Breaker will open; The Turbine Generator will NOT trip. Answer: C Explanation/Justification: Incorrect.

Plausible if the candidate does not understand the RPS logic that they may believe only the "A' RTB opens. Correct turbine status. Incorrect.

Plausible if the candidate does not understand the RPS logic that they may believe only the UN RTB opens. Incorrect turbine status. Correct. The candidate must have knowledge of the effect that a malfunction of the RPS (Train A functions ONLY) and resultant effect on the Turbine Generator.

The candidate must have knowledge of the RPS breaker configuration (RPS KIA). In the current plant configuration an "An Train reactor trip signal will open both the "A" RTB and BYB. The turbine will trip because the reactor tripped. The candidate may confuse the P-9 logic and because we are below P-9 (49%), a turbine trip will not result in a reactor trip however this is not true in reverse. Incorrect.

Correct RPS status. Incorrect turbine status (refer to correct answer explanation).

Sys # System Category KA Statement 012 Reactor K3 Knowledge of the effect that a loss or malfunction of the RPS will TIG Protection have on the following:

KlA# K3.02 KIA Importance 3.2* Exam Level RO References provided to Candidate None Technical

References:

UFSAR Logic Figure 7.3.7 & 7.3-20 3SQS-1.1 Powerpoint Slide Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 5S Content: (CFR: 41.7/45.6)

Objective:

3SQS-1.1 10. Given a specific plant condition, predict or describe the response of the RPS & ESF control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off normal condition.

Beaver Valley Unit 2 NRC Written E:l<am (2LOT8) 38. Given the following plant conditions:

• The plant is stable in Mode 3 following a reactor trip.

• Containment Pressure Transmitter 2LMS*PT950 has failed high.

• All required actions directed by the Instrument Failure Procedure were completed.

• Subsequently, 2LMS*PT951 fails high. What will be the CIB and Safety Injection response, if any? A. Both CIB and Safety Injection actuate. B. CIB actuates but Safety Injection doesn't. C. Neither CIB or Safety Injection actuate. D. Safety Injection actuates but CIB doesn't. Answer: C Explanation/Justification: A. Incorrect.

Refer to correct answer explanation, B. Incorrect.

Refer to correct answer explanation. C. Correct. The candidate must recognize that the initial failure was on CH-1. The stem states that all required actions of the IF procedure have been completed, which means that the CH-I input to the CIB actuation circuitry has been bypElssed which changes the actuation logic from 2/4 to 2/3. Upon a subsequent failure of a 2nd channel, (CH-II), NO CIB actuations will occur because only 1 of 3 Channels have seen the failure. CH I does not provide input to Safety Injection actuation circuitry, so therefore when CH II fails it does not satisfy the 2/3 logic required for Safety Injection to actuate, therefore no SI actuation occurs. All distractors are plausible if the candidate does not know SSPS logics or impacts of these failures upon the system. D. Incorrect.

Refer to correct answer explanation.

Sys# System Category KA Statement 013 Engineered Safety K6 Knowledge of the effect of a loss or malfunction on the Sensors and detectors Features Actuation following will have on the ESFAS: KlA# K6.01 KIA Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

20M-1.4.IF , Rev. 9, pg. 4-6 Question Source: Bank -Vision Question Cognitive Level: Higher -Application 10 CFR Part 55 Content: (CFR: 41.7/45.5 to 45.8) Objective:

3S0S-1.1 10. Given a specific plant condition, predict or describe the response of the reactor protection system trip logics & ESFAS control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off-normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 39. Given the following plant conditions:

• The Unit is operating at Full Power with all systems in NSA to ensure train separation.

• [2HVR*FN201A], "CNMT Air Recirc Fan" is running. * [2HVR*FN201 B], "CNMT Air Recirc Fan" is secured for maintenance.

• [2HVR*FN201 C], "CNMT Air Recirc Fan" is running.

• A Loss of Bus 2P occurs.

• No operator action has occurred and all systems function as designed.

Which ONE of the following will be the CURRENT status of [2HVR*FN201NC]

Containment Air Recirculation Fans? [2HVR*FN201 A] A. RUNNING B. RUNNING C. NOT RUNNING D. NOT RUNNING Answer: B xplanation/Justification:

[2HVR*FN201 C] RUNNING NOT RUNNING NOT RUNNING RUNNING Incorrect.

Correct that 2HVR*FN201A is funning, however, 2HVR*FN201 C is tripped. Plausible because 2HVR*FN201 C can be selected to either power supply. Correct. 2HVR*FN201A is powered from Bus 2N, 2HVR*FN2018 is powered from Bus 2P ancl2HVR*FN201C can be powered from either 2N or 2P. In the stated plant conditions 2HVR*FN201C is running. Since 2HVR*FN201A is being supplied from 2N, NSA would dictate that 2HVR*FN201C would be aligned to the 2P bus to allow train separation.

If 2P is lost then 2HVR*FN201A will be the only running containment air recirc fan. Incorrect.

Correct that 2HVR*FN201C is not running. Incorrect that 2HVR*FN201A is not running. D. Incorrect.

Opposite of the correct fan status. Sys# System Category 022 Containment K2 Knowledge of power supplies to the following:

Cooling KlA# K2.01 KIA Importance 3.0* Exam Level References provided to Candidate None Technical

References:

Question Source: New KA Statement Containment cooling fans RO 20M-44C.3.C, Rev. 10, pg. 5 -8 2SQS-44C.l PPNT, Rev.11 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7) Objective:

2. Identify the power supplies for the components identified on the Normal System Arrangement System flowpath drawing which are powered from the class 1 E electrical distribution system.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 40. Given the following plant conditions: A Large Break LOCA resulted in a Unit Trip from 100 0 A) power with all systems in NSA. While progressing through the EOP set, RWST level dropped below 400 inches and the Control Room Team has transitioned to ES-1.3, "Transfer to Cold Leg Recirculation". Automatic actuations on Attachment A-0.7, "Cold Leg Recirculation Actuation" are being verified. All systems function as designed.

Which ONE of the following Attachment A-0.7 automatic actuations by design does NOT prevent radioactive release from the containment to the RWST? When the LHSI pumps trip, their associated suction valves [2SIS*MOV8809A1B]

close. When the LHSI pumps trip, their associated mini flow valves [2SIS*MOV8890AlB]

close. When LHSI to HHSI valves [2SIS*MOV863A1B]

open, the RWST to HHSI suction valves [2CHS*LCV115B/D]

close. When "C"/D" RSS Pump LHSI Header valves [2SIS*MOV8811A1B]

open, associated spray header isolation valves [2RSS*MOV156C/D]

close. Answer: 0 "planation/Justification: Incorrect.

This is a correct design to prevent backflow from the containment to RWST upon transfer to recirculation, so therefore this is an incorrect answer (refer to correct answer explanation) Incorrect.

This is a correct design to prevent backflow from the containment to RWST upon transfer to recirculation, so therefore this is incorrect answer (refer to correct answer Incorrect.

This is a correct design to prevent backflow from the containment to RWST upon transfer to recirculation, so therefore this is an incorrect answer (refer to correct answer explanation) Correct. The candidate must know the design features for cold leg recirculation swapover and be aware of how these realignments by design impact minimizing the escape of radioactivity from the containment to the RWST. The question is asking which of the alignments does not prevent release, so therefore the candidate must sort through each alignment before deducing that although 2SIS*MOV8811A1B open and 2RSS*MOV156C/D close, this realignment does not preclude release to the RWST but rather is more to align flow to the LHSI header to maximize recirculation of sump water back to the core for cooling. Sys# System KA Statement 026 Containment K4 Knowledge of CSS design feature(s) and/or Prevention of path for escape of radioactivity from containment to Spray interlock(s) which provide for the following:

the outside (interlock on RWST isolation after swapover).

KlA# K4.09 KIA Importance 3.7* Exam Level RO References provided to Candidate None Technical BVPS-2 UFSAR, Rev. 17, pg. 6.3.3 &4 2S0S-11.1 Powerpoint Figures 20M-53A1.ES**1.3, Issue 1C, Rev. 6, pg. 4 20M-53A1.A-O.7, Rev. 3, pg. 2-4 Question Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7) Objective:

14. Given a specific plant condition, predict the response of the containment depressurization system control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant conditions or for an off-normal condition.

Beaver Valley Unit 2 NRC Written E:Kam (2LOT8) 41. Given the following plant conditions: The Control Room Team is performing 20M-52.4.R.1.F, "Station Shutdown from 100% Power to Mode 5". The procedure directs RCS cooldown by dumping steam by adjusting

[2MSS-PK464], Main Stm Manifold Press Control in AUTO or MANUAL. Current RCS temperature is 495 OF and DROPPING.

The maximum allowable ADMINISTRATIVE C/O rate allowed by 20M-52.4.R.1.F is _ (1)_ AND the reason for this C/O limit is to ensure _ (2) _. A. (1) 50°F/hr (2) reactor vessel brittle fracture margins are maintained.

B. (1) 60°F/hr (2) reactor vessel brittle fracture margins are maintained.

C. (1) 90°F/hr (2) TS & LRM limits are not exceeded.

D. (1) 100°F/hr (2) TS & LRM limits are not exceeded.

Answer: C .=xplanation/Justification: Incorrect.

50 F/hr is a cool-down rate specified in accordance with 20M-52.4.R.1 F to initially begin RCS C/O but is not the maximum allowable rate. Correct reason for limit. Incorrect.

60 F/hr is the maximum allowed heatup rate in accordance with 20M-52.4.R.1 F. Correct reason for limit. Correct. 90 F/hr is the TS maximum allowed administrative C/O limit allowed by 20M-52.4.R.1 F. The basis of this C/O rate is so the RCS is not operated under conditions that can result in brittle fracture of the RCPB. Violating LCO limits places the reactor vessel outside the bounds of the stress analyses.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary.

So therefore the limit is to ensure TS & LRM limits are not exceeded. Incorrect.

According to LRM Section 5.2, 100 F/hr is the maximum allowed RCS cooldown Correct CIO rate bases. Sys# System Category KA Statement 039 Main and Reheat K5 Knowledge of the operational implications of the following Bases for RCS cooldown limits Steam concepts as the apply to the MRSS: KlA# K5.05 KIA Importance 2.7 Exam Level RO References provided to Candidate None Technical BVPS TS 3.4.3, Amend 278/161 LRM 5.2.1.1, Rev. 62, pg. 5.2-1,2,13, & 17 BVPS TS 3.4.3 Bases, Rev. 0 20M-52.4.R.1.F, Rev. 23, pg 29 -33 & 79 Question Source: New Question Cognitive Level: Lower-Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

3SQS-ITS.007

2. State the purpose of each TS 3.4 specification as described in the Applicable Safety Analysis section of the TS Bases.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 42. Given the following plant conditions and sequence of events:

• The plant is operating at 100% power. * [2FWE*P23A], "A Motor Driven Auxiliary Feedwater Pump" is OOS.

• A Loss of Offsite power coincident with a turbine trip occurs.

• Bus 2DF has an overcurrent lockout.

• All systems function as designed.

With no operator action, which ONE of the following describes the response of the Auxiliary Feedwater (AFW) System? A total AFW flow of approximately

__ Generators through the __ (2) __ A. (1)375 (2) "A" Header ONLY. B. (1) 700 (2) liB" Header ONLY. C. (1)700 (2) "A" Header ONLY. D. (1) 900 (2) "A" AND "B" Headers. Answer: C Explanation/Justification:

(1) __ GPM will bH provided to ALL Steam Incorrect.

Incorrect capacity.

Correct header. Plausible if the candidate does not know the capacities or misunderstands the initial plant conditions.

One validator chose this dlstractor based on confusing the AFW pump capacities. Incorrect.

Correct capacity.

Incorrect header. Plausible if the candidate believes NSA is to the "8" header or believes the impact of 2FWE*P23A is realignment of 2FWE-P22 to the "8" header. Correct. A loss of offsite power coincident with a turbine trip results in a reactor trip and subsequent loss of both MFW pumps. The EDGs are designed to start on a loss of power to AE and OF bus which will power both electric AFW pumps. In the stated conditions, with an overcurrent condition on the OF bus, 2FWE*P23B will not have power. Since 2FWE*P23A is already OOS, only 2FWE-P22 (Turbine Driven AFW pump) will start to provide approximately 700 gpm AFW flow. The AFW system is designed to feed all three S/G based on NSA alignment requirements.

NSA has 2FWE-P22 aligned to the "An Header. Incorrect.

Correct capacity.

If the candidate does not know the capacities or understand the impact based on initial plant conditions, then it is plausible that AFW flow would be provided through the "A" header by 2FWE-P22 and the "B" header by 2FWE*P23B.

In this case the total flow will be 900 gpm based on limiting orifices which limit flow to 300 gpm per S/G. Incorrect because 2FWE*P23B has no power. Sys # System Category 059 Main K3 Knowledge of the effect that a loss or malfunction of the MFW will Feedwater have on the following:

KlA# K3.02 KIA Importance 3.6 Exam Level References provided to Candidate None Technical

References:

Question Source: Bank -1 LOT8 NRC Exam 0#42 KA Statement AFW system RO 20M-24.1.C, Rev.2, pg. 58.6 2S0S-24.1, Rev. 24 PPNT slide. Question Cognitive Level: Higher Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7/45.6)

Objective:

16. Given a Main Feedwater, Startup Feedwater, Auxiliary Feeclwater System or Steam Generator Water Level Control System configuration and without referenced material, describe the associated system's control room response to the following off-normal conditions, including autorratic functions and changes in equipment status as applicable.

Loss of instrument air or Loss of electrical power.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 43. Given the following plant conditions:

• The plant is in Mode 3 following a reactor trip from an extended 450 day run.

• A 50 °F/hr cool down from 547 of has just begun. S/G NR Water Levels are being maintained constant at 44% using AFW. Which ONE of the following describes the required AFW flow trend required to maintain a constant RCS cool down rate to Mode 5? A. AFW flow requirements will be constant as long as the C/O rate remains constant.

B. More AFW flow will be required to maintain S/G Water Level due to decreased density. C. AFW flow requirements will be constant as long as S/G Water Level remains constant.

D. Less AFW flow will be required to maintain S/G Water Level because heat input to S/Gs drops. Answer: 0 Explanation/Justification: Incorrect.

Eventually as the cool down continues, RHS will be placed in service which will reduce the steaming rate if the cool down rate maintained constant.

Less steam requires less AFW Incorrect.

Density will increase versus decrease, Plausible that to maintain a constant cool down, that the steaming rate must increase, therefore, more AFW flow seems Incorrect.

It requires less feedwater to maintain a constant S/G water level as the RCS cooldown continues.

Otherwise, AFW pumps would adequate for full power Correct. This is an operational fundamental question that requires the candidate to simply understand that decay heat rate drops over Therefore less AFW flow is required as the amount of heat transfer from the RCS to S/Gs Sys # System Category KA Statement 061 Auxiliaryl K5 Knowledge of the operational implications of the Relationship between AFW flow and RCS heat transfer Emergency Feedwater following concepts as the apply to the AFW: KiA# KiA Importance 3.6 Exam Level RO References provided to Candidate Technical ThermodynamidReactor Theory Fundamentals

References:

Question Source: Bank -2LOT5 NRC Exam #20 Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

17. Given a specific plant condition, predict the response of the MF, S/U Feed, AFW, or SGWLC system's control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant conditions or for an off normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 44. Given the following plant conditions and sequence of events: The Unit was operating at 45% with AMSAC removed from service for testing. A reactor trip occurred one minute ago. Safety Injection did NOT actuate and is NOT required. ALL Steam Generator Narrow Range level indicators dropped to the following levels before recovering: "A" S/G 33% "8" S/G 11% "C" S/G 22% Containment conditions are normal and all equipment functioned as No operator action has yet After completion of E-O, "Reactor Trip or Safety Injection" 10As, which ONE of the following will be the Auxiliary Feedwater Pump(s) status and associated flow requirements, if any? A. Turbine driven pump running with no flow requirements.

8. Turbine driven pump running with flow> 340 gpm required.

C. Motor and Turbine driven pumps running with no flow requirements.

O. Motor and Turbine driven pumps running with flow> 340 gpm required.

Answer: A Explanation/Justification: Correct The candidate must be able to analyze plant conditions and apply the conditions to the automatic start status of the AFW pumps. Specifically they must monitor changes in S/G water level as well as other parameters to determine which AFW pumps have started and also determine heat sink requirements based on these S/G levels. None of the start permissives have been met to auto start the motor driven AFW pumps. The Turbine AFW pump has started on low S/G water level in "8" S/G due to water level being < 20.5% (2/3 lo-lo S/G levels on 1/3 S/Gs). Since "A" & "c" S/G have >12% (31 %) with no adverse containment, then no flow is required to meet heat sink requirements.

All distractors are plausible if the candidate does not properly apply the auto start permissives for the AFW pumps (Motor Driven AFW Pumps require 2/3 lo-lo S/G levels on 2/3 S/Gs). Unit 2 requires >12% in any S/G or 340 gpm total feedwater flow for heat sink lAW F-0.3 Heat Sink. Incorrect Correct pump status. Incorrect heat sink requirement (refer to correct answer explanation) Incorrect.

Incorrect pump status. Correct heat sink requirement (refer to correct answer explanation) Incorrect.

Incorrect pump status. Incorrect heat sink requirement (refer to correct answer explanation)

Sys # System Category KA Statement 061 Auxiliary/ A 1 Ability to predict and/or monitor changes in parameters (to S/G level Emergency Feedwater prevent exceeding design limits) associated with operating the AFW controls including:

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

References:

20M-24.1.D, Rev. 6, pg. 16 -18 20M-53A1.F-0.3, Issue 1C, Rev. 2, pg. 1 20M-24.2.B, Rev. 16, pg. 2 Question Source: Bank -Vision # 51709 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/45.5)

Objective:

2S0S-24.1

40. Given a specific plant condition, predict the response of the MFW, SUFW, AFW, or SGWLC systems control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off-normal condition.

---

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 45. Given the following plant conditions: The Unit is operating at Full Power with all systems in NSA. The Control Room Crew has just synchronized the 2-1 Emergency Diesel Generator (EDG) to the grid for surveillance lAW 20ST-36.1, "EDG [2EGS*EG2-1], "Monthly Test". The 2-1 EDG is paralleled to the grid, carrying about 50% load. The RO places the 2-1 EMERG GEN VOLTAGE ADJUST control switch to RAISE. Which ONE of the following describes the result of placing the 2-1 EMERG GEN VOLTAGE ADJUST control switch to RAISE? Indicated A. grid voltage rises; EDG speed remains the same. B. grid voltage rises; EDG speed droops as load rises. C. reactive load rises; EDG speed remains the same. D. reactive load rises; EDG speed droops as a function of load. Answer: C Explanation/Justification:

'\. Incorrect.

Parallel operation of an EDG with a large power grid is such that adjusting voltage of the smaller machine will have no impact on the larger power grid. Correct that EDG speed remains the same. Incorrect.

Parallel operation of an EDG with a large power grid is such that adjusting voltage of the smaller machine will have no impact on the larger power grid. Plausible that the EDG speed droops since it is in the Droop mode when paralleled with off site power sources. Correct. The candidate must understand from performing this OST in the simulator or in plant the impact of making adjustments during parallel operations of the EDG and offsite power sources. As the voltage adjust control switch is taken to raise, KVARs increase and there is no impact on EDG speed even though it is operating in the Speed Droop mode because it is paralled with an infinite power source. Incorrect.

Correct that reactive load rises. Incorrect but plausible since the EDG is in the speed droop mode and if candidate confuses mode of operation.

Sys# System Category KA Statement 062 AC Electrical A4 Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different ac Distribution supplies KlA# A4.07 KIA Importance 3.1* Exam Level RO References provided to Candidate None Technical

References:

20ST-36.1, Rev. 66, pg. 35 -38 2S0S-36.2 Powerpoint, Rev. 20 Question Source: Bank -Vision #17367 Question Cognitive Level: Lower-Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7/45.5/

to 45.8) Objective:

2S0S-36.2

28. Describe the control, protection and interlock functions for the field components associated with EDG, including automatic functions, setpoints and changes in equipment status, as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 46. Given the following plant conditions:

• The Unit is operating at Full Power with all systems in NSA.

• A1-1C, "VITAL BUS INVERTER OPERATIONITROUBLE" annunciator is received.

• Computer address V01 01 D indicates Vital Bus 2-1 INV TROUBLE.

• Nuclear Instrumentation on NI-41 rack is energized.

• Once dispatched the NLO reports Inverter failure is indicated on [UPS*VITBS2-1].

Assuming the system functioned as designed and no operator action has occurred, which ONE of the following will be the impact on Vital Bus 2-1 Loads? Vital Bus 2-1 Loads ____ A. are being supplied by 125 VDC SWBD 2-1. B. are being supplied by an alternate source (MCC2-E05) via static switch. C. are being supplied by an alternate source (MCC2**E08) via static switch. D. are NOT being supplied until the static switch is manually transferred.

Answer: B Explanation/Justification:

\. Incorrect.

Plausible because this would be the supply of power for a rectifier failure. Indications provided do not support this as the cause. Since A 1-1 C is a common annunciator, the plant computer would have a different message specifying Vital Bus 2-1 Batt Operating which indicates a rectifier failure as opposed to a inverter failure has occurred. Correct The candidate must know the system design for the UPS power sources. They must understand the indications provided in the question stem as noted above. On a loss of UPS inverter, a static switch is designed to automatically swap over to an alternate source of power which will be from MCC2-E05. Incorrect.

Correct that Vital Bus 1 will be supplied by an alternate source via a static switch, hc)wever, the candidate must know that MCC2-E07 is the alternate power supply to Vital Bus 4 versus Bus 1. Incorrect.

Plausible if a complete loss of power to Vital Bus 1 occurred, however, the candidate must recognize that with NI41 energized there is still power to rule out this Sys # System Category KA Statement 062 AC Electrical K4 Knowledge of ac distribution system design feature{s) and/or Uninterruptable ac power sources Distribution interlock{s) which provide for the following:

KlA# K4.10 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

20M-38.4.AAA, Rev. 7, pg. 2, 7 & 8 3S0S-38.1 Powerpoint Slides, Rev. 6 20M-38.1.B, Issue 4, Rev. 1, pg. 2-3 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7) Objective:

4. From memory, describe the control, protection, and interlock functions associated with the 120 VAC Distribution System operation for the following:

as applicable include automatic functions, setpoints, and changes in equipment status: Static Switches.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 47. Given the following plant conditions: The plant is operating at Full Power with all systems in NSA. Multiple control room alarms and indications simultaneously occur. The following indications exist: All Rod Bottom Lights are LIT. All Main Steam Isolation Valves are closed. Loss of benchboard indicating lights for loads on 4KV Bus 2AE and 4S0V Bus SN. Letdown flow indicates ZERO (0). All systems function as designed. No operator actions have yet occurred.

Based on these plant conditions, which ONE of the following is the cause of these plant conditions?

A. A Loss of 4S0V Bus 2N occurred.

B. A Loss of 125VDC Bus 2-1 occurred.

C. A Loss of 125VDC Bus 2-2 occurred.

D. A Loss of 120VAC Vital Bus 1 occurred. ,nswer: B Explanation/Justification: Incorrect.

Plausible since it has some commonalities.

The reactor would not trip for this condition. Correct. The candidate must be able to evaluate the plant conditions and make an operational judgment based on indications provided related to the DC Electrical System. The judgment necessary to be made is the cause of plant performance or conditions provided in the stem. The candidate must recognize that the reactor is tripped based on all rod bottom lights lit. A Loss of DC Bus 1 will result in an automatic reactor trip due to MSIV's closing. An automatic letdown isolation will occur. Another non direct symptom provided is Loss of Benchboard lights for loads on 2AE and Bus 8N. Incorrect.

Plausible since all of the indications are the same for a Loss of DC Bus 2 with the exception of the opposite bus 2DF and 9P Incorrect.

Plausible if the candidate confuses a Loss of Vital Bus 1 with a Loss of DC Bus. Letdown will isolate in both cases. Sys# System Category KA Statement 063 DC Electrical Generic Ability to evaluate plant performance and make operational Distribution judgments based on operating characteristics, reactor behavior, and instrument interpretation.

KlA# 2.1.7 KIA Importance 4.4 Exam Level RO References provided to None Technical

References:

20M*53C.4.2.38.1A , Rev. 4, pg 1 Candidate 20M*53C.4.1.39.1A , Rev 3, pg. 1,2, & 7 20M*53C.4.1.39.1B , Rev. 3, pg. 1 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/43.5/45.12

/45.13) Objective:

3SQS*39.1

20. Given a change in plant conditions due to a system/component failure, analyze the 125VDC Distribution System to determine what failure occurred.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 48. Given the following plant conditions: The Unit is operating at 100% power with all systems in NSA. A8-10A, "125V DC Bus 2-1 Ground" is received and acknowledged. The Control Room Team references 20M-39.4.F, Grounds (125 VDC Buses 2-1 and 2-2)". Which ONE of the following is the impact if more than one ground exists AND what action will be taken to preclude this impact according to this procedure?

The impact of multiple DC grounds is that __ (1) __To preclude this impact __ A. (1) inadvertent actuations may occur. (2) de-energize DC Bus 2-1 until grounds are located. B. (1) inadvertent actuations may occur. (2) open knife switches or breakers prior to resetting relays. C. (1) control functions may not occur when called upon. (2) de-energize DC Bus 2-1 until grounds are located. D. (1) control functions may not occur when called upon. (2) the unit must be shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if grounds are not isolated.

Answer: B Explanation/Justification: Incorrect.

Correct impact according to 20M-39.4.F.

Incorrect but plausible action. Grounds located by isolating individual components supplied by DC-Bus 2-1. If the entire bus were de-energized it would be difficult to locate and isolate the ground. Correct. The candidate must be able to predict the impacts of multiple DC grounds on DC Busi 2-1. According to 20M-39.4.F, the impact of multiple grounds is that inadvertent actuations may occur. BVPS has had some actual OE regarding this issue. The candidate must also be able to use procedures to control the impact of multiple grounds. According to the precautions and limitations of the same reference, the method of control is to open knife switches or breakers prior to resetting relays. Incorrect.

Correct impact not referenced in our procedure, however, from research it is also possible with a DC ground that control functions may not operate when called upon depending on the resistance of the circuit. Incorrect but plausible action as explained in A above. Incorrect.

Correct impact not referenced in our procedure, however, from research it is also possible with a DC ground that control functions may not operate when called upon depending on the resistance of the circuit. Plausible incorrect action. TS 3.8.4 actions if a battery charger is inoperable is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action. The RO is required to know:;; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions from memory. Sys# System Category KA Statement 063 DC Electrical A2 Ability to (a) predict the impacts of the following malfunctions or Grounds Distribution operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

KlA# A2.01 KIA Importance 2.5 Exam Level RO References provided to Candidate None Technical

References:

20M-39.4.AAJ, Issue 4, Rev. 1, pg. 2 20M-39.4.F, Rev. 5, pg. 2 -4 NETA World 2008, pg. 2 Question Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/43.5/45.3/45.13) Objective:

3505-39.1

19. Given a 125 VDC Distribution System alarm condition and using the ARP determine the appropriate alarm response, including automatic and operator actions in the control room.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 49. Given the following plant conditions and sequence of events: The Unit suffered a Loss of Off-Site Power. Both Emergency Diesel Generators (EDGs) are supplying emergency busses. Grid stability is confirmed and the Operations Manager has granted permission to return to the grid. The Control Room Team is performing 20M-36.4.E, 'Transferring 4KV Emergency Bus 2AE to Bus 2A". EDG 2-1 is being synchronized to the grid and ACB 2E7 is closed. Upon ACB 2E7 closure, the following annunciator sequence occurs: A8-2B, "4160V EMER BUS 2AE ACB 2E7 OVERCURRENT TRIP" received. A8-2A, "4160V EMER BUS 2AE ACB 2A10/2E'7 AUTO TRIP" received. A8-2B, "4160V EMER BUS 2AE ACB 2E7 OVERCURRENT TRIP" clears. Which ON E of the following describes the impact on EDG 2-1? EDG 2-1 will ____ cooling water available.

A. trip with B. trip without C. continue to run with ). continue to run without Answer: C Explanation/Justification: Incorrect.

EDG 2-1 does not trip but remains running. Plausible that the EDG would trip on an overcurrent condition, however, protection in this scenario is provided by ACB 2E7. Correct that cooling is still available. Incorrect.

EDG 2-1 does not trip but remains running. Plausible that the EDG would trip on an overcurrent condition, however, protection in scenario is provided by ACB 2E7. Also incorrect that cooling water is not Correct. For the given conditions, EDG 2-1 is running paralleled to the grid. An overcurrent condition was caused by the closure of ACB2E7 and results in ACB 2A10 & 2E7 automatically opening. Upon ACB-2E7 opening, the overcurrent condition clears which is indicative oflhe problem being downstream of ACB 2E7. The EDG will continue to run with cooling since ACB2E10 remains closed and EDG cooling would be maintained from the running "A" Train SW pump being supplied by the AE Bus powered by the EDG. It is not RO knowledge to select procedures so therefore only the first part of the higher cognitive KIA was tested. For purposes of satisfying the KIA, the Aux Feeder Breaker is the 2E7 breaker to the AE bus. Incorrect.

Correct that EDG 2-1 remains running, incorrect that it is running without cooling. Plausible if the candidate believes the overcurrent trip opens ACB 2E10 and does not recognize or understand the RW system configuration.

If the EDG did trip the opposite train SW cooling would need to be manually aligned. Sys# System Category KA Statement 064 Emergency A2 Ability to (a) predict the impacts of the following malfunctions or Consequences of opening auxiliary feeder bus Diesel operations on the ED/G system; and (b) based on those predictions, (ED/G sub supply) Generator use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

KlA# A2.13 KIA Importance 2.6* Exam Level RO References provided to Candidate None Technical

References:

20M-36.4.E, Rev. 10, pg. 3 &4 20M-36.4.ACC, Rev. 8, pg. 3 &4 20M-36.4.ACD, Rev. 3, pg. 3 4KV and SW Powerpoint Slides 1uestion Source: Bank -1 LOT8 NRC Exam 0#48 ..luestion Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/43.5/45.3/45.13) Objective:

13. Given an EDG configuration and without referenced material, describe the EDG control room response to the following actuation signals, including automatic functions and changes in equipment status as applicable:

SI or Bus UV.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 50. Given the following plant conditions:

The Unit is operating at Full Power with all systems in NSA EXCEPT: 2-2 Emergency Diesel Generator Air Compressor

[2EGA*C21 B] control switch is in OFF. [2EGA*C21 B] is being placed on clearance for maintenance. While posting the clearance

[2EGA *C22B] control switch was inadvertently taken from the AUTO to OFF position.

Based on this plant configuration, which ONE of the foliowin£1 will be the first control room indication(s), if any? A. A2-3H, "Safety System Train A Inoperable".

B. A2-4H, "Safety System Train B Inoperable".

C. DG 2-2 Starting Air Pressure indication slowly dropping.

D. There will be no control room indication for this plant configuration.

Answer: B Explanation/Justification: Incorrect.

Plausible incorrect answer. This alarm would be received the opposite scenario were to occur. Correct. The candidate must have knowledge of the impact in the control room of remote operation of the EDG air compressor switches.

If local air compressor control switches are in OFF for the associated EDG, then the control room will receive a BISI alarm (Safety System Train Incorrect.

This indication is local versus in the control room. Plausible if the candidate does ne,t know what EDG indications are in the Incorrect.

Plausible that the candidate may believe there is no control room indication when operating the EDG remote air compressor switches.

Sys # System Category KA Statement 064 Emergency Diesel A4 Ability to manually operate and/or monitor in the control Remote operation of the air compressor switch Generator room: (different modes) KlA# A4.04 KIA 3.2* Exam Level RO Importance References provided to Candidate None Technical

References:

20M-36.4.ADF, Rev. 5, pg 2 & 12 20M-36.4.ADC, Rev. 7, pg 2 & 3 20M-36.4.AEI, Rev. 16, pg. 2 20M-36.1.D, Issue 4, Rev. 3, pg. 23 & 24 20M-36.1.C, Rev. 4, pg. pg. 8 & 9 Question Source: New Question Cognitive Level: Higher -Comprehension 10 CFR Part 55 Content: (CFR: 41.7/45.5 to 45.8) Objective:

36. Describe the control, protection and interlock functions for control room components associated with the EDG, including automatic functions, setpoints and changes in E!quipment status as applicable.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 51. Given the following plant conditions: The plant is operating at 100% with all systems in NSA A liquid waste discharge is in progress to the Unit 1 Cooling Tower Blowdown. [2SWS-RQI102], "Component Cooling HX SW" fails upscale HIGH. [2CCP-RQI100], "Component Cooling Water" is reading normal and is unchanged. The following alarms are received: A4-5A, "RADIATION MONITORING SYSTEM TROUBLE" A4-5C, "RADIATION MONITORING LEVEL HIGH" On RM-11, it is confirmed that COMPONENT COOLING HX SW [2SWS-RQI102]

is blinking RED. No other alarms are present and no operator action has occurred. All systems function as designed.

What will be the impact of this process monitor failure on the effluent release in progress?

A. The release will automatically terminate immediately.

B. The release will continue and IS required to be manually terminated.

C. The release will automatically terminate after a short time delay. D. The release will continue and is NOT required to be manually terminated. "nswer: D Explanation/Justification: Incorrect.

Plausible if the candidate confuses this monitor with 2SGC-RQ1 00 which would reslJlt in auto termination if an upscale failure occurred, Incorrect.

Correct that release will continue and plausible but incorrect that the release must be manually terminated, The candidate must understand system interrelationships.

With CCP in normal there is no reason to believe there is any radiation coming from the CCP system into the SW system. Incorrect.

Plausible if the candidate confuses this monitor with 2SGC-RQ100 which would reslJlt in auto termination if an upscale failure occurred and they also confuse or do not know that there is no time delay with this failure. Correct. The candidate must understand how the failure of 2SWS-RQI1 02 will impact the effluent release in progress.

There is no automatic action associated with this radiation monitor. Therefore an upscale failure will have no impact on the release. The candidate must also know the system interrelationships between CCP and SW. Ifthere were a leak from the RCS into CCP, then there would be an alarm from 2CCP-RQI100.

The ARP does not require any release to be terminated.

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

KlA# K3.01 KIA 3.6 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-43.4.AAA, Rev. 8, pg 2,3, & 8 20M-43.4.AAC, Rev. 1, pg. 2 & 3 20M-43.4.ACG, Rev. 5, pg. 2 20M-43.4.AEI, Rev. 7, pg. 2 20M-43.4.ACO, Rev. 7, pg. 2 20M-43.1.C, Rev. 4, pg. 55 & 59 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7/45.6)

Objective:

8. Given a specific plant condition, predict the response of the RM system control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 52. Given the following plant conditions:

• The Unit is operating at 80% power with all systems in NSA.

• A LOCA results in a Reactor Trip and SAFETY INJECTION. Immediately following the safety injection, [2SWS*P21A], Service Water Pump trips.

• All systems function as designed and NO operator actions occur. Which of the following will have Service Water available for cooling? 1. [2CHS*P21A], HHSI Pump [2CHS*P21 B], HHSI Pump 3. [2SIS*P21 B], LHSI Pump [2HVC*ACU201B], Control Room Ventilation A. 2 AND 4 ONLY. B. 3 AND 4 ONLY. C. 1 AND 2 ONLY. I. 1,2, AND 4. Answer: 0 Explanation/Justification: Incorrect Both B HHSI pump and B Control Room Ventilation will have cooling. This is incorrect because the A HHSI pump will also have cooling. Plausible if the candidate does not know about the auto start feature of A ESW for A header restoration. Incorrect.

LHSI pumps are cooled by local air cooling and require no service water cooling. Pla,usible if the candidate confuses LHSI and pump cooling. Correct that B Control Room Ventilation has SW Incorrect.

Correct that A & B HHSI pump wi" have cooling. Plausible if the candidate knows that LHSI pumps do not receive SW cooling and is unaware of the cooling medium to the B Control Room Ventilation, Correct. Service Water provides cooling to the charging pump lube oil coolers. These pump become the High Head SI pumps on a SI When the A" SW Pump tripped, a low pressure condition occurred resulting in an AUTO start clf the "A>> Emergency Service Water Pump restores cooling to the "A" SW header. Therefore both High Head SI Pumps will have SW available, Control Room Ventilation is emergency heat load which will have SW Sys# System Category KA Statement 076 Service Water A3 Ability to monitor automatic operation of the SWS, including:

Emergency heat loads KlA# A3,02 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-30.1.B, Rev, 6, pg, 5-7 OP Manual Figure 30-1, 1A, & 2 Question Source: New Question Cognitive Level: Higher -Comprehension 10 CFR Part 55 Content: (CFR: 41.7 I 45.5) Objective:

2SQ8-30,1

36. Describe the control, protection and interlock functions for the control room components associated with the SW system, including automatic functions, setpoints, and changes in equipment status as applicable, Beaver Valley Unit 2 NRC Written Exam (2LOT8) 53. Given the following plant conditions:

• The Unit is at 100% power.

• Containment Air is being supplied by Station Instrument Air.

• A Large Break Loss of Coolant Accident occurs.

• All systems function as designed.

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

1. 2IAC*MOV130, "CNMT Instrument Air Isol Vlv." 2. 2IAC-MOV131, "CNMT Instrument Air Backup Supply Vlv." 3. 2IAC*MOV133, "CNMT Instrument Air Isol Vlv." 4. 2IAC*MOV134, "CNMT Instrument Air Isol Vlv." A. 1 ONLY. B. 1 AND 2 ONLY. C. 3 AND 4 ONLY. D. 1.2. AND 3 . *nswer: A ExplanationlJustification: Correct. 2IAC-MOV131 and 21AC*130 are open at 100% power to supply instrument air from instrument air compressors into containment.

BVPS Unit 2 no longer uses containment air compressors.

Upon a large break lOCA and 51 and subsequent CIA signal will auto close 2IAC*130.

In order to restore instrument air to containment, this valve needs to be reopened only. Incorrect.

Correct that 21AC*MOV130 needs to be reopened.

Plausible if the candidate does not know that 2IAC*MOV131 does not receive a CIA signal or believes this valve is affected by this signal. The EOP directs both of these valves opened, however, the EOP deals with all modes of operation and in the stated plant mode, the candidate must know it is not necessary to reopen 2IAC-MOV131. Incorrect.

2IAC*MOV133

& 134 both receive a CIA Signal and close. This was the old configuration when running CNMT lAC instrument air to containment.

Opening these valves will not restore IA to containment. Incorrect.

All three of these valves receive a CIA Signal and close from their NSA open positions.

The candidate may believe that these valves all need to be reopened to restore instrument air. Sys # System Category KA Statement 078 Instrument K1 Knowledge of the physical connections andlor cause-effect Containment air Air relationships between the lAS and the following systems: KlA# K1.03 KIA Importance 3.3 Exam level RO References provided to Candidate None Technical

References:

2SQS-34.1, Rev. 18, pg. 3 2SQ8-34.1 Power-point slide 20M-53A.1.E-0, Issue 1C, Rev. 8, pg. 12 Question Source: New Question Cognitive Level: lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.2 to 41.9) Objective:

14. Given a Unit 2 Compressed Air configuration and without referenced material, describe the compressed air system control room response to the following off-normal conditions, including automatic functions and changes in equipment status as applicable:

Containment Isolation Signal Phase A (CIA)

Beaver Valley Unit 2 NRC Written Exam (2LOT8) Given the following plant conditions:

• The Unit was operating at Full Power with all systems in NSA.

• A steam line break occurred outside containment.

• An automatic reactor trip and safety injection occurred from Train "Au ONLY. Which ONE of the following will be the status of the Containment penetration lines for the Phase "A" (CIA) and Phase "B" (CIB) isolation valves AND what operator action is required, if any? All CIA & CIB valves close. No operator action is required. Train "A" CIA & CIB valves close. Operators must manually isolate Train "B" CIA & CIB valves. All CIA valves close. All CIB valves do NOT reposition.

Operators must manually isolate CIB valves. Train "A" CIA valves close. All CIB valves do NOT reposition.

Operators must manually isolate Train "B" CIA valves ONLY. Answer: 0 Explanation/J ustification: Incorrect Plausible if the candidate believes that either train will isolate both trains CIA & CIB isolation valves in which case there would be no need for operator action. (refer to correct answer explanation)

Incorrect.

Correct that Train A CIA valves are closed. Incorrect that Train A CIB valves are closed. Plausible action that the operators would close the Train B valves if they failed to isolate. Incorrect.

Plausible if the candidate believes only one train isolates all CIA valves & CIB requires both trains in which case there would be a need for operators to close CIS valves. Either manual CIA will isolate both trains. (refer to correct answer explanation) Correct. The candidate must be able to analyze the stated plant conditions and be able to app:iy knowledge of how a Train A SI will effect CIA and CIS. A Train A SI signal will actuate the Train A CIA valves ONLY since it is train specific (unle:ss manually actuated).

They must also understand the impact of the SLB outside containment on CIB. Since the break is outside containment, no CIB actuation will occur, The correct action if all CIA valves do not isolate is for the control room team to ensure Train B CIA valves are isolatect E-O directs the operator to perform Attachment 0.11, Verification of Automatic Actions which directs the operators to attempt manual isolation, Sys# System Category KA Statement 103 Containment A2 Ability to (a) predict the impacts of the following malfunctions or Phase A and B isolation operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations KlA# A203 KIA Importance 3.5* Exam Level RO References provided to Candidate None Technical

References:

UFSAR Logic Diagram Figure 7.3-13 20M-53A.1.E-0, Issue 1, Rev. 8, pg. 4 20M-53A, 1.A-O.11, Rev. 6, pg. 6 3SQS-1.1 Powerpoint, Rev. 7 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41,5/43.5/45,3/45.13) Objective:

3SQS-1, 9. Given a Reactor Protection System Trip Logics & ESF configuration and without referenced material, describe the RPS & ESF control room response to the following actuation signals, including automatic functions and changes in plant equipment status as applicable:

Main Steam Line Break Accident Beaver Valley Unit 2 NRC Written Exam (2LOT8) 55. Given the following plant conditions: A Large Break LOCA occurred. An ORANGE path has developed on the Containment CSF Status Tree due to an abnormal rise in containment sump level. The Control Room Team transitions to FR-Z.2, "Response to Containment Flooding".

Which ONE of the following describes the mitigating strategy of this procedure?

The mitigating strategy of this procedure is to ____ A. verify containment isolation and heat removal. B. check for and isolate a faulted steam generator.

C. identify unexpected sources of sump water and isolate. D. isolate additional safety injection flow beyond what is required.

Answer: C Explanation/Justification: Incorrect.

Incorrect but plausible strategy related to FR-Z.1 versus FR-Z.2. Incorrect.

Incorrect but plausible strategy related to FR-Z.1 versus FR-Z.2. Correct. The RO candidate must have knowledge of the mitigation strategies related to the containment.

They must have knowledge of FR-Z.2 major action categories.

The major action category for this procedure is to identify unexpected sources of sump water and isolate. They must know that to get to FR-Z.2 that the sources of water must be beyond that provided by Safety Injection. Incorrect.

Incorrect but plausible if the candidate does not know the overall purpose and strategy of FR-Z.2. Sys # System Category KA Statement 103 Containment Generic Knowledge of EOP mitigation strategies.

KJA# 2.4.6 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1 .FR-Z.1. Issue 1C. Rev. 2. pg. 1 20M-53A.1.FR-Z.2.lssue 1C, Rev. 2, pg. 1 & 2 20M-538.4.FR-Z.2, Issue 1C, Rev. 2, pg 1, 2, & 4 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/43.5/45.13) Objective:

3. State from memory the basis and sequence for the major action steps of each EOP procedure, lAW 8VPS EOP Executive Volume.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) Which ONE of the following describes the sequence of components from power supply to the Control Rod Drive Mechanism (CRDM's)? (RTB's = Reactor trip (RDMG's = Rod Drive Motor 480 VAC Substation 8N &9P, RDMG's, RTB's, Power Cabinets. 480 VAC Substation 8N & 9P, Power Cabinets, RDMG's, RTI3's. 480 VAC Substation 2-1 & 2-2, RDMG's, RTB's, Power Cabinets. 480 VAC Substation 2-1 & 2-2, Power Cabinets, RDMG's, RTB's. Answer: C Explanation/Justification: Incorrect.

Plausible incorrect emergency power supply with correct flowpath. Incorrect.

Plausible emergency power supply with incorrect flowpath. Correct. The candidate must know the power supply to the Motor Generator Sets and have understanding of the f10wpath of this power to the Control Rod Drive Mechanisms.

480 VAC Substation 2-1 supplies power to 2RDS-MG21 and 480 VAC Substation 2-2 supplies power to MG22. The proper flowpath is via the RDMGs via the RTBs through the power cabinets to the CRDMs. Incorrect.

Correct power supply with plausible incorrect f1owpath.

Sys # System KA Statement 001 Control Rod Drive System K2 Knowledge of bus power supplies to the following:

MIG sets "/A# K2.05 KIA Importance 3.1* Exam Level RO provided to Candidate None Technical

References:

20M-1.3.C, Rev. 18, pg. 13 20M-1.3 Powerpoint, Rev. 6 Question Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7) Objective:

3S0S-1.3 3. Describe how power is supplied to the Rod Drive Motor Gene,rator sets, Logic/Power cabinets, DC Hold Cabinet, and the Control Rod Drive Mechanism coils.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 57. Given the following plant conditions:

• The Unit is operating at 100% power with all systems in NSA. 120 VAC Vital Bus II de-energizes.

Which ONE of the following describes an IMMEDIATE consequence associated with the Loss of 120 VAC Vital Bus II? A. MANUAL rod withdrawal is blocked. B. All Atmospheric Steam Dump Valves failed closed. C. RCS low flow reactor trip logic changes from 2/3 in 1/3 loops to 1/2 in 2/3 loops. D. All Power Range NI 2/4 logic is reduced to 2/3 until the required bistables are tripped. Answer: A Explanation/Justification: Correct. High Power Rod Stop logic is y.. A loss of Vital Bus II causes a loss of control power which feeds bistables which when perform their function.

The NIS Power Range High Setpoint Overpower Rod Stop Block Rod Withdrawal functions to prevent manual withdrawal.

Auto rod withdrawal has been Incorrect.

The atmospheric steam dump valves will become unavailable for a Loss of Vital Bus 1 and are not affected by the Loss of Vital Bus II. Incorrect.

P-8 logic becomes 1/2 in 1/3 loops (above P-8). Incorrect.

The Vital Bus II loss results in the associated bistables tripping which results in a 1/:1 remaining logic. Sys # System Category KA Statement 015 Nuclear Instrumentation Generic Ability to operability and/or availability of safety related eqLlipment.

KlA# 2.2.37 KIA Importance 3.6 Exam RO Level References provided to Candidate None Technical 3SQS-2.1 Powerpoint Slide

References:

20M-2.5.A.4, Rev. 5, pg. 2 20M-2.1.C, Rev. 2, pg. 15 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7/43.5/45.12) Objective:

3SQS-2.1 16. Given a specific plant condition, predict the response of the NIS, including all automatic functions and changes in equipment status, for a change in plant conditions.

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Beaver Valley Unit 2 NRC Written Exam (2LOT8) An internal fault (short circuit) occurs in the PRZR Press Control [2RCS-PK444A]

Controller.

Which ONE of the following describes the effect this fault will have on the Reactor Protection System? The CONTROLLER fault could NOT directly feed back into the protection circuit due to use of isolation devices. directly feed back into the protection circuit, causing the SELECTED channels to trip. directly feed back into the protection circuit, preventing the SELECTED channels from tripping. NOT directly feed back into the protection circuit since separate transmitters are used for control and protection.

Answer: 0 Explanation/Justification: Incorrect This is the design feature for PRZR Level control circuit but is not the same for the pressure control/protection circuit Incorrect.

PRZR pressure control is separate from the protection and do not have isolation amplifiers.

The candidate could confuse PRZR circuitry with PRZR pressure circuitry.

The transmitters do share common taps off of the same' instrument line. A PRZR pressure controller failure will indirectly have an impact on actual PRZR pressure so therefore will impact RPS Incorrect.

PRZR pressure control is separate from the protection and do not have isolation amplifiers.

The candidate could confuse PRZR Level circuitry with PRZR pressure circuitry.

The transmitters do share common taps off of the same, instrument line. The candidate may focus on whether the fault causes reference pressure to fail high or low which makes B & C distractors plausible. Correct. The candidate must have knowledge of the operational implications of separation of control and protection circuits for non-nuclear instrumentation.

Specifically, they must have knowledge of how a fault on the NNIS (control side of PRZR pressure) impacts RPS (protection side of PRZR pressure).

PRZR pressure uses separate transmitters so therefore the NNIS does not directly feedback into the Reactor Protection circuitry.

Indirectly a failure of the reference PRZR pressure controller could impact actual plant pressure and therefore indirectly effect RPS. Sys# System Category KA Statement 016 Non-nuclear K5 Knowledge of the operational implication of the following Separation of control and protection circuits Instrumentation concepts as they apply to the NNIS: KlA# K5.01 KIA Importance 2.7* Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.1F , Rev. 13, pg. 23 & 24 Ops Manual Figures 6-35 & 36 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

2SQS-6.4 17. Describe the control, protection, and interlock functions for the control room components associated with the PRZR & PRZR Relief System, including automatic functions, sE!tpoints and changes in plant equipment status as applicable.

Beaver Valley Unit 2 NRC Written (2LOT8) 59. Given the following plant conditions:

• A Reactor Trip from Full power occurred due to low ReS pressure.

• RCS pressure is currently 885 psig and DROPPING.

• Containment pressure is 6 psig and RISING.

• All ESF equipment functioned as designed.

• Core Exit Thermocouples (CET) are currently 532 of and STABLE. Based on these parameter trends, what will happen to CET neliability and what is the current condition of the RCS? CET indication will __ (1) __ AND the RCS is currently

__ (2) __. A. (1) remain reliable (2) superheated.

B. (1) remain reliable (2) saturated.

C. (1) become less reliable since adverse conditions exist (2) saturated.

D. (1) become less reliable since adverse conditions exist (2) superheated.

Answer: B Explanation/Justification: Incorrect.

Correct status of CET's. Incorrect RCS condition.

Misapplication of the conversion from psig to psia is a common fundamental problem. Correct. The candidate must be able to predict or monitor changes in core exit temperature as the containment conditions degrade. They must also be able to apply the thermocouple reading to obtain correct condition of the RCS. 900 psia corresponds to 532 F. which means the RCS is in a saturated condition.

Misapplication of the conversion will result in a different end result. Incorrect.

Plausible as containment conditions become more adverse that instrumentation will become less accurate.

The CETs are designed operate in this type of environment.

Correct RCS Incorrect.

Plausible as containment conditions become more adverse that instrumentation will become less accurate.

The CETs are designed operate in this type of environment.

Incorrect but plausible RCS condition, if the candidate does not know how to use steam Sys# System 017 In-Core Temperature Monitor System (ITM) KlA# A1.01 KA Statement A1 Ability to predict and/or monitor changes in parameters (to pmvent exceeding design limits) associated with operating the ITM system controls including:

KIA 3.7 Exam Level RO Importance Core exit temperature References provided to Candidate None Technical

References:

20M-2.3.1, Issue 1, Rev. 2, pg. 2 3S0S-3.1, Rev. 5, pg. 12, 13, & 16 Question Source: Bank -Vision #68041 Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.5/45.7)

Objective:

3SQS-3.1 8. Describe the response of a thermocouple readout to an adverse containment environment.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 60. Given the following plant conditions:

• A Plant Startup is in progress.

• Control Rods are in MANUAL.

• Reactor power is currently at 23%. * [2MSS-PK464], "Main Stm Manifold Pressure Control" is in AUTO with Zero (0) demand.

• Steam Dumps are in the Tavg Mode.

• Main Feedwater Regulating Bypass valves are in AUTO.

• An Inadvertent Turbine Trip occurs. Which ONE of the following describes the steam dump and S/G NR Water Level response to the turbine trip, assuming NO operator actions? Steam Dumps will S/G NR Water Level will_ (2) A. (1) open and close. (2) drop and then rise. B. (1) open and remain open. (2) rise and then drop. C. (1) open and remain open. (2) drop and then rise. D. (1) remain closed. (2) remain at programmed level. Answer: C Explanation/Justification: Incorrect.

Correct that steam dumps open, however, the candidate must understand that with control rods in manual, that the increased Tavg caused by the turbine trip will not be reduced until the operators insert control rods. Therefore the steam dumps will remain open until Tavg is reduced to Tref. Correct S/G water level response. Incorrect.

Correct steam dump response.

Opposite S/G water level response. Correct. The candidate must analyze the plant conditions and understand steam dump operation and S/G water level response.

With steam dumps in Tavg Mode initially, they would be closed. When the turbine trips, Tavg increases and the steam dumps will open based on temperature difference between Tref as sensed by first stage pressure and Tavg (C-7A Load Rejection Arms). The steam dumps will remain open until the operator inserts control rods to lower Tavg. S/G water level will initially drop to due to a decrease in steam demand (turbine trip). Once the steam dumps open in response to increasing Tavg, the S/G water level will begin to increase due to increase in steam demand which leads to S/G swell. Also note that there are other dynamics in play such as SGWLC. With Bypass valves in auto, the SGWLC system will respond by opening the bypass valves which will increase feedwater flow and also increase S/G water level. The tie between steam dumps and SIG water level is the change in steam demand. Incorrect.

Incorrect steam dump response if the candidate believes they will not operate based on stated plant conditions.

It is plausible that if the Steam dumps do not open that S/G water level would be unaffected since there would be no change in steam demand. Sys# System Category KA Statement 041 Steam Dump System K1 Knowledge of the Physical connections and/or cause-effect S/G level (50S) and Turbine relationships between the SOS and the following systems: Bypass Control KlA# K1.02 KIA 2.7 Exam Level RO Importance References provided to Candidate None Technical

References:

20M-21.5.A.12 , Rev. 3, pg. 2 20M-21.5.A.13 , Rev. 3, pg. 2 20M-24.1.0, Rev. 6, pg. 2-6 luestion Source: New Question Cognitive level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.2 to 41.91 45.7 to 45.8) Objective:

2SQS-21.1

12. Given a change in plant conditions predict the response of the MSSS control room indications and control loops, including all automatic functions and changes in equipment status, for either a change in plant conditions or for an off normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) Which ONE of the following is the minimum required Fuel Storage Pool Boron Concentration to ensure adequate shutdown margin (Keff s 0.95) in accordance with Technical Specification 3.7.16? (assume current rack configuration) 0 ppm ;::: 495 ppm ;::: 1050 ppm ;::: 2000 ppm Answer: 0 Explanation/Justification: Incorrect.

Plausible because it is the TS 4.0 design features value mentioned in section 4.3.1.2.b.

Unit 2 can only maintain Keff 1.0 without crediting soluble boron. Incorrect.

Plausible because it is the TS 4.0 design features value mentioned in section 4.3.1.c. Incorrect.

Plausible since it is the TS 3.7.16 Unit 1 number. Correct. The candidate must have knowledge of the Spent Fuel Pool Cooling Design features/interlocks which ensures adequate SID margin (Cb concentration).

Unit 2 is currently undergoing a major rerack project. Some of these numbers 11ave been incorporated into our Technical Specifications and therefore even though the project may not complete by the time 2LOT8 takes the IL T exam. we are testing current TS's. According to TS 3.7.16, Unit 2 requires 2000 ppm to ensure Keff 0.95. This is a conservativE' value to ensure no credible boron dilution event will reduce boron concentration below 450 ppm. This is RO level of knowledge since it tests LCO knowledge.

Sys# System Category KA Statement 033 Spent Fuel K4 Knowledge of design feature(s) and/or interlock(s) which provide Adequate 80M (boron concentration)

Pool Cooling for the following:

JA# K4.05 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

BVPS TS 3.7.16, Amend 278/161, pg. 3.7.16-1 BVPS TS 4.3, Amend 278/161, pg 4.0-1-4 UFSAR, Rev. 13, pg 3.1-44 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7) Objective:

2S0S-20.1

30. Describe the deSign basis for the Fuel Pool Cooling and Purification System and the associated major components as documented in the UFSAR. 28. For a given set of plant conditions, determine if the condition meets the criteria for entry into a one hour or less action statement in accordance with TS's.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 62. Given the following plant conditions:

• Unit 2 is in Mode 6 during a refueling outage.

• The Containment Equipment Hatch is closed.

• Fuel Movement is in progress.

• Containment Purge is in operation and NO features have been defeated.

• [2HVR*RQ104B].

Containment Purge Radiation Monitor fails upscale HIGH. * [2HVR*RQ104A].

Containment Purge Radiation Monitor is unaffected.

• All systems function as designed.

• No operator action has occurred.

What will be the impact on Containment Purge? Containment Purge will ______ A. automatically isolate with no time delay. B. automatically isolate after a short time delay. C. be unaffected and will require manual isolation.

D. be unaffected and will NOT require manual isolation.

Answer: A KplanationlJustification: Correct. The candidate must have knowledge of the impact of a radiation monitor upscale failure high. Specifically, if 2 HVR*RQ1 048 fails high. it will cause an automatic isolation of containment purge. This is the only radiation monitor which has any automatic functions associated with fuel handling.

Also note that there is no actual high radiation condition, rather an IF condition exists. The stem of the question states that Containment Purge is in operation and no auto functions have been defeated.

This is to alleviate any confusion based on procedure flexibility which allows auto isolation features to be defeated at the end of 20M-44.C.4.A (Containment Purge Startup). Incorrect.

Correct that containment purge auto isolates but not after a short time delay. Plausible because some of BVPS radiation monitors have time delays. Incorrect.

Plausible if the candidate believes the logic is 2/2 for containment purge to isolate and that manual isolation is required due to impact on fuel handling operations. Incorrect.

Plausible if the candidate believes there is no auto isolation and isolation is NOT required which is correct for these plant conditions. (ie: has no impact on fuel handling operations)

Sys# System Category KA Statement 034 Fuel Handling K6 Knowledge ofthe effect of a loss or malfunction on the following Radiation monitoring systems Equipment will have on the Fuel Handllng System: KlA# K6.02 KIA Importance 2.6" Exam Level RO References provided to Candidate None Technical

References:

20M43.1.C, Rev. 4, pg 18 BVPS TS 3.9.3, Amend 278/161, pg. 3.9.3-1 BVPS BIG TS 3.9.3, Rev. 0, pg. B3.9.3-1-4 BVPS LRM 3.9.3. Rev. 52. pg. 3.9.3-1 20M-53C.4.2.49.1, Rev. 9, pg. 2 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7/45.7)

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 63. Given the following plant conditions: The plant is operating at 100 % power with all systems in NSA. High Radiation is confirmed on 2ARC-RQI-100, "Condenser Air Ejector Discharge". The crew enters AOP 2.6.4, "Steam Generator Tube Leakage" and determines a 75 gpd Steam Generator Tube Leak is in progress. All systems function as designed.

With no operator action, which ONE of the following describes the alignment of the Unit 2 Condenser Air Ejector Off-Gas? A. Air Ejector discharge is AUTO aligned to Unit 2 containment.

B. 2MSS*SOV120, "Common Header Isolation Downstream" AUTO OPENS. C. Air Ejector discharge is AUTO aligned through the Charcoal Delay Beds. D. No AUTO action occurs, discharge continues to atmosphere until manual action occurs. Answer: 0 ExplanationlJ ustification: Incorrect.

Plausible if the candidate confuses Unit 2 with Unit 1 since Unit 1 does AUTO align to the containment. Incorrect.

Plausible because one of the ARP actions is to manually align 2MSS*SOV120.

This valve does auto open on an SI signal. C. Incorrect.

The air ejector discharge is aligned to the charcoal delay beds however, this is not an AUTO ). Correct. The ARP directs the air ejector discharge be aligned to the gaseous waste system through the delay beds in accordance with 19.4.H due confirmed high radiation level from the condenser air removal system which is indicative of a S/G tube leak. AOP 2.6.4 also directs the alignment through the delay beds. Sys # System Category KA Statement 055 Condenser K1 Knowledge of the physical connections and/or cause/effect PRM system Air Removal relationships between the CARS and the following systems: KlA# K1.06 KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

20M-43.1.C, Rev. 4, pg. 8 20M-43.4.ACN, Rev. 5, pg. 2 &3 20M-53C.4.2.6.4, Rev. 26, pg. 21 20M-19.4.H, Rev. 14, pg. 2 -4 Question Source: New Question Cognitive Level: Higher* Comprehension 10 CFR Part 55 Content: (CFR: 41.2 to 41.9 145.7 to 45.8) Objective:

Given a specific plant condition, predict the response of the RM system control room indications and control loops, including any automatic functions and changes in equipment status for either a change in plant conditions or an normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 64. Given the following plant conditions: The Unit is operating at 100% power with all systems in NSA. The Control Room Team is performing 20M-19.4.G, "Filling the Unit 2 Gaseous Waste Storage Tanks from Unit 2 Surge Tank". Oxygen concentration has been verified < 2% by sample. Which ONE of the following components and/or indications have the capability of being operated or monitored from the control room to perform this evolution lAW 20M-19.4.G? [2GWS-OA 1 OOA] , "Oxygen Analyzer" sample flow. [2GWS-SOV125A1 through 125G1]. Tank 25A through 26G Inlet Isolation Valves". [2GWS-AOV108], "Gaseous Waste Storage TK Inlet Header Isolation Valve". Gaseous Waste Storage Tank Pressures A. 3 ONLY. B. 3 &4 ONLY. C. 1. 2, & 4 ONLY. D. 1, 3, & 4 ONLY. Answer: B t:.xplanation/Justification: Incorrect.

Correct that this valve is operated from the control room, however, it is not the only indication/valve provided. Correct. The candidate must have knowledge of monitoring or manually operating valves, indications, sample line and gas decay tanks associated with the Waste Gas Disposal System. In order to operate or monitor, they must have knowledge of where valves and indications are located. In accordance with 20M-19.4.G, 2GWS-AOV10S is operated from BS-A and Gaseous Waste Storage tank Pressures are monitored from either the computer or at local indications.

All indications/valves provided are plausible since they are called out by the referenced procedure. Incorrect.

Oxygen Analyzers are operated from the auxiliary building.

Since Oxygen Concentration may be obtained by sample or computer point, sample flow is specified which is locally obtained only. Waste gas Storage Tanks Inlet isol Valves are also operated locally in the PAS at 2GWSTP. Gaseous Waste Storage Tank Pressures may be obtained locally or in the control room. Incorrect.

Refer to previous discussions.

1 & 2 are incorrect.

3 & 4 are correct. Sys # System Category KA Statement 071 Waste Gas A4 Ability to manually operate and/or monitor in the control room: Gas decay tanks, including valves, indicators, and Disposal sample line KlA# A4.05 KIA Importance 2.6* Exam Level RO References provided to Candidate None Technical

References:

20M-19.4.G, Rev. 4, pg. 2-4 2S0S-191 , Rev. 17 Powerpoint Slides Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.7 / 45.5 to 45.S) Objective:

2S0S-19.1

10. Describe the control, protection, and interlock functions for the control room components associated with the GWDS, including automatic functions, setpoints and changes in plant equipment status as applicable.

Beaver Valley Unit 2 NRC Written (2LOT8) 65. Given the following plant conditions: The Plant has been operating at 100% power with all systems in NSA for 500 days. A downpower is currently in progress.

Reactor Power iis currently 55%. The following alarms are received within several minutes of each other. A6-3E, "Cooling Tower Pump Trouble". A6-5G, "Condenser Vacuum Low/Low-Low". A5-5B, "Condenser Vacuum Low Turbine Trip" [2CNM-PR103], "Main Condenser Side A & B Vacuum Recorder" is reading 24 IN-VAC and DROPPING. All systems function as designed. No operator actions have yet occurred.

Which ONE of the following will be the status of the reactor AND safety injection?

A reactor trip __ (1) __ AND safety injection

___ (2) __ as a result of these plant conditions.

A. (1) has occurred (2) has actuated B. (1) has occurred (2) has NOT and will NOT actuate C. (1) will occur when vacuum drops below 22 IN-VAC (2) has NOT and will NOT actuate D. (1) has occurred (2) has NOT actuated but will actuate when PRZR pressure drops below 1845 psig. Answer: B Explanation/Justification: Incorrect Correct reactor response.

Incorrect 51 response. (refer to correct answer explanation) Correct. The candidate must be able to analyze the stated plant conditions and based on these indications apply knowledge of RPS &ESFAS functions.

If systems function as designed, the turbine will trip on low condenser vacuum at 24 IN-VAC. Because the reactor was operating>

P-9 (49%) power, a turbine trip will result in a reactor trip. Without circulating water or condenser vacuum the candidate must understand that eventually the condenser steam dumps will be lost and at EOC life, maximum decay heat, the plant will not actuate safety injection.

AFW will auto start and provide feedwater and the plant will cycle on the atmospheric dump valves with no operator action to remove decay heat Plant parameters will not approach any 51 setpoint A reactor trip does not result in an SIS. TS 3.3.1, Function 8 states that P-4 is an E5FAS function. Incorrect.

Incorrect reactor response.

Plausible if the candidate does not know the Low Vacuum reactor trip setpoint A reactor trip has occurred.

Correct that SI will not Incorrect.

Correct reactor response.

Incorrect that SI will occur. Correct SI setpoint (refer to correct answer explanation).

5ys # System Category KA Statement 075 Circulating K3 Knowledge of the effect that a loss or malfunctions.

of the ESFAS Water circulating water system will have on the following:

KlA# K3.07 KIA Importance 3.4* Exam Level RO References provided to Candidate None Technical 20M-31.4.AAB, Rev. 9, pg. 2 20M-26.4.AAK, Rev. 13, pg. 3 20M-26.4.AAB, Rev. 1, pg. 2 20M-1.5.B.1, Rev. 2, pg. 3 20M-1.5.B.4F , Issue 4, Rev. 0, pg. 1 & 2 BVPS TS, Amendments 278/164, Table 3.3.2-1 350S*1.1 Powerpoint, Rev. 7 Question Source: New Question Cognitive Level: Higher-Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7/45.6)

Objective:

3S0S-1.1 10. Given a set of plant conditions, predict or describe the response of the RPS & ESFAS control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off normal condition.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 66. Given the following plant conditions:

• The US provides you a working copy of 20ST-45.11 , "Cold Weather Verification".

• You note that this procedure has not been annotated as the latest approved procedure.

Which ONE of the following will be the required method, if any, to validate 20ST-45.11 is the latest revision in accordance with NOP-LP-2601, "Procedure Use and Adherence"?

A. NOT required to be validated prior to use. B. MUST be validated by comparing to FileNet prior to use. C. MUST be validated by comparing to Control Room Copy prior to use. D. ONLY required to be validated by comparing to FileNet at least once every two days. Answer: B Explanation/Justification: Incorrect.

Plausible incorrect answer. During emergency operations and drills, the documents in the emergency facilities may be used without validating to File Net. Correct. The candidate must have knowledge of how a controlled copy of an operating procedure is verified.

In accordance with this is the requirement for all other procedures other than emergencies or safeguard information which is not viewable in Incorrect.

Plausible incorrect answer. Only required to compare with control room copies if File!Net is unavailable.

Correct that the procedure required to be validated prior to Incorrect.

Plausible because NOP-LP-2601 does require procedures other than emergencies

<)r drills to be validated every three days thereafter.

The procedure is also required to be validated prior to use. Sys # System Category KA Statement N/A N/A Generic Ability to verify the controlled procedure copy. KlA# 2.1.21 KIA Importance 3.5* Exam Level RO References provided to Candidate None Technical

References:

NOP-LP-2601, Rev. 4, pg. 20 &21 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/45.10/45.13) Objective:

3505-48.1

11. From memory, explain the requirements of adherence to and familiarization with operations procedures.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 67. Given the following plant conditions: The plant was operating at 100% power with all systems in NSA. A reactor trip and safety injection occurred. The US directs the BOP to perform Attachment A-O.11, "Verification of Automatic Actions".

Which ONE of the following conditions in accordance with Attachment A-0.11 will require the BOP to direct action outside the control room? A. SWS Pumps seal water pressure LOW. B. Two Hydrogen Analyzers NOT running. C. Two Service Water pumps NOT running. D. All Train "B" orange CIA marks are LIT. (several Train "A" CIA marks are NOT LIT) Answer: A Explanation/Justification: Correct. The candidate must be able to coordinate personnel activities outside the control room. In order to direct these actions they must be aware of conditions which require local action. The competent RO must be able to verify automatic actions following safety injection lAW Attachment A-O.11 and be able to coordinate local actions in the event that automatic actions have not occurred.

Of all of the conditions provided, a SWS pump low seal water pressure condition requires the BOP to dispatch an NLO to the intake structure to investigate this plant condition. Incorrect.

Although Attachment A-O.11 does check two hydrogen analyzers running. the required action is to start the analyzers performed in the control room as opposed to outside the control room. Incorrect.

Although Attachment A-O.11 does check two service water pumps running, the required actions for this condition are all performed the control room as opposed to outside the control Incorrect.

No action is required outside the control room for this condition.

As long as all of the valves are closed in one train the redundant valves do not need to be locally Sys# System KA Statement N/A N/A Generic Ability to coordinate personnel activities outside the control room. KlA# 2.1.8 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical

References:

20M-53A.1.E-O, Issue 1C, Rev. 8 pg. 4 20M-53A.1.A-0.11 , Rev. 6, pg. 5 & 6 20M-53A.1.A-0.2, Issue 1 C, Rev. 0, pg. 2 Question Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 145.5/45.12/45.13)

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 68. Given the following plant conditions:

• The Unit is operating at 85% power with all systems in NSA for this power level.

• The following Annunciators are received:

o A4-3F, "LOOP Tavg DEVIATION".

o A4-4C, "LOOP l::.T DEVIATION".

o A4-3C, "Tavg DEVIATION FROM Tref'. * [2RCS*TI412A], "LOOP 1 PROT l::.Tn on V8-B has decreased.

• [2RCS*TI412D], "LOOP 1 PROT Tavg" on VB-B has decreased.

Which ONE of the following diverse computer point indications will confirm the cause of these alarms and indications?

A. [T0401A], "RCL A NR THOT 1 2RCS*TE412B1" has failed LOW. B. [T0401A], "RCL A NR THOT 1 2RCS*TE412B1" has failed HIGH. C. [T0405A], "RCL A NR TCOLD 1 2RCS*TE412C/D" has failed LOW. D. [T0405A], "RCL A NR TCOLD 1 2RCS*TE412C/D" has failed HIGH. Answer: A 'l(planation/Justification:

..... Correct The candidate must be able to interpret diverse indication to validate the response of ather indications.

Specifically, they must be able to analyze the alarms and indications and apply knowledge of the OTfTavg functional diagram well as IF procedure knowledge to correctly deduce that the indications provided can be validated on the plant computer to confirm the cause. If OT decreases, this can be caused by lowering TH or increasing TC. ForTAVG to also decrease, than either TH or TC must decrea!;e.

The common issue between the two indications is TH decreasing which can be confirmed by plant computer.

All distractors are plausible if the, candidate does not have the knowledge of how to confirm or misapplies these valid plant conditions.

B. Incorrect.

Refer to correct answer explanation.

C. Incorrect.

Refer to correct answer explanation.

D. Incorrect.

Refer to correct answer explanation.

Sys # System Category KA Statement NIA NIA Generic Ability to identify and interpret diverse indications to validate the response of another indication.

KiA# 2.1.45 KiA Importance 4.3 Exam Level RO References provided to Candidate None Technical

References:

20M-6.4.1F , Rev. 13, pg. 34 -39 2SQS-6.5 Powerpoint, Rev. 17 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.7/43.5 145.4) Objective:

28Q8-6.5 21.Given a specific plant condition, predict the response of the Reactor Coolant System control room indication and control loops, including all automatic functions and changes in equipment status, for either a change in plant condition or for an off-normal condition.

Process Instrument Failure Beaver Valley Unit 2 NRC Written Exam (2LOT8) 69. Given the following plant conditions: The plant was operating at 75% power with all systems in NSA for this power level. A Load Rejection to 45% power occurs. RCS Pressure rose to 2290 psig and then dropped to 2185 psig and is currently 2200 psig and SLOWLY RISING. All Four Delta Flux indicators read -11% on BB-B. A4-9D, "ROD CONTROL BANK D LOW-LOW' is acknowledged. No operator actions have yet occurred.

Which ONE of the following Technical Specification (TS) LCO requires entry, based on these plant conditions? LCO 3.1.6, Control Bank Insertion Limits LCO 3.2.3, Axial Flux Difference (AFD) LCO 3.4.1, RCS Pressure, Temperature, & Flow Departure from Nucleate Boiling Limits A. 3 ONLY. B. 1 AND 2 ONLY. C. 1 AND 3 ONLY. O. 1,2, AND 3. Answer: C Explanation/Justification: Incorrect Correct that this is applicable but it is not the only applicable TS LCO. (refer to correct answer explanation) Incorrect.

Correct that control bank insertion limit is applicable however the candidate must recognize that below 50% power AFD is not to be tracked. Plausible because -11 would require entry at 100% power. (refer to correct answer Correct. The RO is required to know LCO statements and associated applicability information (ie: the information above the double line separating the actions from the LCO and associated statements).

The candidate must analyze the plant conditions provided and be able to recognize which TS LCOs are applicable, in order to be able to track them. They must recognize that RCS pressure did drop below 2214 psia and although it is now back to normal will still need to be tracked. Also, they must recognize the significance of the Bank D Low-Low RIL and that LCO entry and tracking must occur. AFD is not required to be tracked because reactor power is < 50%. Incorrect.

AFD entry not required so no tracking is necessary. (refer to correct answer explanation)

Sys # System Category KA Statement N/A N/A Generic Ability to track Technical Specification limiting conditions for operations.

KlA# 2.2.23 KIA Importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

BVPS TS 3.1.6/3.2.3/3.4.1 BVPS LRM COLR Cycle 16, pg. 5.1-5, 5.1-9 &5.1-11 20M-1.4.AAM, Rev. 4, pg. 3 Question Source: New Question Cognitive Level: Higher -Comprehension or Analysis 10 CFR Part 55 Content: (CFR: 41.10/43.2/45.13) Objective:

3SQS-ITS.01

1. Given plant conditions, apply the rules of ITS section 3.0 to ensure compliance with technical specification.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 70. Given the following plant conditions and sequence of events: The Unit is in Mode 6 with core reload in progress. The crew has completed Train swap to the "A" Train aiter completing operability run on the 2-1 Emergency Diesel Generator (EDG). 2-2 EDG has been placed on clearance for maintenance activities. Several hours later the control room receives a report that after reviewing the maintenance work order for 2-1 EDG, incorrect gasket material installation makes 2-1 EDG inoperable.

Which of the following TS 3.8.2, "AC Sources -Shutdown" LCO(s) action(s) is (are) immediately required? Suspend Core Alterations. Suspend operations involving positive reactivity additions that could result in a loss of shutdown margin or boron concentration. Initiate action to restore required EDG to operable status. A. 1 ONLY. B. 1 AND 2 ONLY. C. 2 AND 3 ONLY. ). 1,2, AND 3. Answer: 0 Explanation/Justification: Incorrect. (refer to correct answer explanation) Incorrect. (refer to correct answer explanation) Incorrect. (refer to correct answer explanation) Correct. The RO candidate must be able to analyze the effect of maintenance activities on the EDG and determine the status of LCOs for TS 3.8.2. The ROs are expected to know the LCO statements and associated applicability information (ie: the information above the double lines separating actions from the LCO and associated applicability statements).

ROs are also required to know s: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements.

Based on stated plant conditions TS 3.8.2 requires one EDG capable of supplying one train of the onsile, Class 1 E AC electrical power distribution subsystems during modes 5 & 6. Since there are no operable EDGs, TS 3.8.2 Condition B requires all of the actions above to be immediately performed.

Sys# System Category KA Statement NIA NIA Generic Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operation.

KlA# 2.2.36 KIA Importance 3.1 Exam Level RO References provided to Candidate Technical

References:

TS 3.8.2 Amend. 278/161, pg. 3.8.2-1 & 3. Question Source: New Question Cognitive Level: Lower Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/43.2/45.13) Objective:

3SQS*ELECT ITS 4. Given plant conditions that constitute non-compliance with any electrical power systems LCO, or LRM, determine the applicable condition(s), required action(s), and associated completion times.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) Which ONE of the following plant conditions/evolutions can result in significantly higher radiation levels in the Safeguards Building? Venting an idle charging pump lAW 20M-7.4.AK, "Venting of Idle Charging Pump". Performing the Low Head SI Pump Test lAW 20ST-11.1, "LHSI Pump [2SIS*P21A]

Test". Transferring to Cold Leg Recirculation lAW ES-1.3, "Transfer to Cold Leg Recirculation". Placing the deborating demineralizer in operation lAW 20M-7.4AM, "Mixed Bed/Deborating Demineralizer Operation".

Answer: C Explanation/Justification: Incorrect.

This is a plausible evolution which is a radiation hazard and requires RP assistance due to the potential for high radioactive release. This hazard is in the PAS as opposed to the safeguards area. This evolution has a potential to result in EPP Incorrect.

LHSI Pumps are located in Safeguards and this evolution recirculates the RWST through the safeguards which makes this plausible.

However, this evolution should not increase radiation levels in Correct. The candidate must have knowledge of radiation or contamination hazards that may ,lfise during any plant activity.

Specifically, they must sort through a list of valid situations and determine that transfer to cold leg recirculation during a LOCA has the greatest potential to increase Safeguards and/or PAS radiation levels. ES-1.3 has a caution that warns the operator of this hazard. Incorrect.

This evolution has a potential to increase radiation levels, however, the procedure is more concerned with the potential reactivity event which could occur as a result of this evolution.

Increased radiation levels would be more of a concern in the PAS as opposed to Safeguards.

Sys # System KA Statement N/A N/A Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

KlA# 2.3.14 KIA Importance 3.4 Exam Level RO References provided to Candidate None Technical 20M-53A.1.ES-1.3, Issue 1C, Rev. 6, pg. 2 20M-7.4.AK , Rev. 14, pg. 3 -5 20M-7.4.AM , Rev. 15, pg, 2 & 3 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41,12 143.4 145.10)

Beaver Valley Unit 2 NRC Written Exam (2LOT8) According to NOP-OP-4202, "Declared Pregnant Workers" and NOP-OP-4201, "Routine External Exposure Monitoring".

which ONE of the following Beaver Valley Occupational Dose Limits are required? (Assume no extensions or planned special exposures)

The Embryo/Fetus Dose Equivalent (EFDE) Limit for a Declared Pregnant Worker over entire gestation period is _ The Site Administrative Control Level Dose Limit (TEDE) for an individual working at the Nuclear Facility is (2) (1) 100 mr/term (2) 2000 mr/year (1) 500 mr/term (2) 1000 mr/year (1) 100 mr/term (2) 1000 mr/year (1) 500 mr/term (2) 2000 mr/year Answer: B Explanation/Justification:

Incorrect Incorrect but plausible DPW limit. The DPW is limited to 100 mr/month but may up to 500 mr for the entire term or gestation period. The Annual Administrative Limit for an individual working at a nuclear facility is incorrect because it reflects a BVPS Administrative Limit where extensions are involved.

The stem of the question specifically excludes extensions or planned special exposures. Correct. According to NOP-OP-4201 Attachment B, when a worker declares pregnancy, she will have an administrative level of 500 mr for the term of pregnancy.

This ensures the dose to the unborn child is minimized.

The federal limit is also 500 mrem for the pregnancy period or term. NOP-OP-4101 refers the reader to NOP-OP-4202 which defines a declared pregnant worker a:nd specifies the occupational dose limit for the entire period of declared pregnancy is 500 mrem (100 mrem/month)

NOP-OP-4101 states on ,l\Uachment A that BVPS Administrative Control Limit for TEDE is 1000 mr/year. Incorrect Incorrect DPW value but plausible as described above. Correct BVPS Administrative Control Limit for TEDE. Incorrect Correct value for DPW. Incorrect value for BVPS Administrative Control Limit for TEDE, however plausible because it reflects the initial annual ACL limit for TEDE when dealing with extensions.

Sys# System Category KA Statement N/A N/A Generic Knowledge of the radiation exposure limits under normal or emergency conditions.

KlA# 2.3.4 KIA Importance 3.2 Exam RO Level References provided to Candidate None Technical NOP-OP-4201, Rev. 1, pg. 3, 11, 14 & 20

References:

NOP-OP-4202, Rev. 0, pg. 3 & 4 Question Source: Bank -1 LOT8 Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.12/43.4/45.10 Beaver Valley Unit 2 NRC Written E)cam (2LOT8) 73. The Unit Two Control Room has been evacuated.

Which of the following indications will be directly available at the Emergency Shutdown Panel for post accident monitoring?

1. RCS Tavg 2. RCS Wide Range Pressure 3. Steam Generator Wide Range Water Level 4. PRZR Level A. 1 AND 2 ONLY. B. 2 AND 4 ONLY. C. 1 AND 3 ONLY. D. 2, 3, AND 4 ONLY. Answer: D Explanation/Justification: Incorrect.

Refer to correct answer explanation. Incorrect.

Refer to correct answer explanation. Incorrect.

Refer to correct answer explanation. Correct. The candidate must be able to identify which instrumentation provides indication for post accident monitoring at the emergency shutdown panel for a control room inaccessibility situation.

RCS Tavg is NOT directly available, however, Tc & Th indications are available and Tavg is procedurally derived from these two indications.

S/G wide range level as opposed to narrow range level is available.

RCS Wide Range pressure and PRZR Level are available at the alternate SID panel and can be directly read at this location.

Sys # System KA Statement NIA NIA Ability to identify post-accident instrumentation.

KlA# 2.4.3 KIA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

8VPS TS Amend 278/161, pg. 3.3.3-3 & 4 & 83.3.3-1 20M-53C.4.2.33.1A , Rev. 12, pg. 8 -10 & 12 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.6 I 45.4) Objective:

3S0S-53.5 Describe the actions for control room inaccessibility.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 74. Given the following plant conditions:

• An unexpected automatic Reactor Trip and Safety Injection from 100% power occurred.

• All systems responded normally to actuation signals.

• E-O, "Reactor Trip or Safety Injection", Step 4 is being implemented.

• The BOP opens [2CCS-AOV118], "Domestic Water to Station Air Compressor Valve". Which ONE of the following describes the action taken by the BOP? According to BVBP-OPS-0024, "Transient Response Guidelines" this action was ____ A. allowed. B. NOT allowed at this time. C. NOT allowed until first transition.

D. allowed by obtaining US/SM concurrence during IOAs. Answer: B Explanation/Justification: Incorrect.

Refer to correct answer explanation.

Plausible if the candidate does not know his role in performing pre-emptive actions. Correct. The RO must know his roles and responsibilities during EOP usage. The candidate must know that opening 2CCS-AOV118 is an allowable pre-emptive action. This action should not be performed during lOA's and shall not be performed with out SM/US concurrence.

Since the crew is performing Step 4 of E-O, lOA's have not been completed.

Incorrect.

Refer to correct answer explanation.

Plausible if the candidate confuses rules of usage for FRP implementation. Incorrect.

The SRO may not provide concurrence during lOA's. IOAs are not complete until read and verified.

The SRO may assign pre-emptive actions prior to reactor trip however, since the reactor trip was unexpected automatic there was no time for this assignment to be made in the circumstances provided.

Sys# System Category KA Statement N/A N/A Generic Knowledge of crew roles and responsibilities during EOP usage. KlA# 2.4.13 KIA Importance 4.0 Exam Level RO References provided to Candidate Question Source: New Question Cognitive Level: Objective:

380S-48.1 None Technical 1/20M-53B.2, Issue 1C, Rev. 7, pg. 7 & 29 BVBP-OPS-0024, Rev. 4, pg. 9 Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10/45.12) 24. Explain from memory all of the Operations Managers Expectations.

Beaver Valley Unit 2 NRC Written Exam (2LOT8) 75. Given the following plant conditions: A serious fire in the cable spreading room has been reported. The Shift Manager determines actions of 20M-56C, "Alternate Safe Shutdown From Outside the Control Room" are necessary. The SM directs the RO to perform actions of 20M-56C.4.C, "NCO Procedure".

Which ONE of the following will be a time critical action performed by the Reactor Operator (RO) outside the Control Room AND reason why? In accordance with 20M-56C.4.C, the RO will _____ A. open [2RCS*PCV456], "PRZR PORV" to reduce RCS pressure.

B. trip [2FWE*P23B], "B" AFW Pump" to prevent Steam Generator overfill.

C. start [2QSS*P2'1 B], "Quench Spray Pump" to reduce containment pressure.

D. close [2CHS*HCV186], "RCP Seal Hdr Flow Control Valve" to protect RCP seals. Answer: B Explanation/Justification: Incorrect.

This is an action performed but is not time critical.

There is a time critical action related to closing the PORV if it spuriously opens. This action is performed to preclude the PORV from spuriously lifting and is directed in the US proc:edure vs. RO procedure.

The RO actually is directed to take power off PORV 456 isolation valve 2RCS*MOV536. Correct. The candidate must have knowledge of RO actions performed outside the control room during alternate safe shutdown and the operational effect of this task. According to 20M-56.C.4.C, pg. 5, it is time critical that the RO secure 2FWE*P23B within 40 minutes prevent S/G Incorrect.

There is a time critical action for the RO to secure 20SS*P21 B if it spuriously starts but there is no action to start this pump. Plausible if the candidate is unfamiliar with the procedure or has concepts confused. Incorrect.

There is a time critical action performed by the NLO versus RO to fail open 2CHS*HCV186 versus close this valve. Plausible incorrect action based on similar actions to isolate RCP seals to prevent action such as ECA-O.O. Sys# System Category KA Statement NIA N/A Generic Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. KlA# 2.4.34 KIA Importance 4.2 Exam Level RO References provided to Candidate None Technical 20M-56C.4.C, f;:ev. 18. pg. 5

References:

20M-56C.4.D, Rev. 22, pg. 2 20M-56C.4.B, Rev. 30, pg. 3, 4 &13 Question Source: New Question Cognitive Level: Lower -Memory or Fundamental 10 CFR Part 55 Content: (CFR: 41.10 I 43.5 145.13) Objective:

1. Describe the function of Alternate Safe Shutdown from Outside the Control Room and the associated major components as documented in Operating Manual Chapter 20M-56C.

(SRO Beaver Valley Unit 2 NRC Written E:Kam 76. The plant is at 100% power with all systems in NSA.

• A Reactor trip occurs.

• The FAST bus transfer to offsite power fails to occur and all Normal 4Kv power is lost.

• All other systems respond as designed.

The problem with offsite power has been corrected, and all4Kv power has been restored.

The crew has transitioned to ES-O.1, Reactor Trip Response and is currently at step 12 attempting to start all RCPs. The following plant conditions exist:

• All RCP #1 seal leakoffs are 0.25 gpm and stable. * "A" Rep seal injection flow is 7.0 gpm and stable * "8" Rep seal injection flow is 6.5 gpm and stable * "e" RCP seal injection flow is 5.0 gpm and stable * "A" RCP thermal barrier temperature is 95 OF and stable * "8" RCP thermal barrier temperature is 75 OF and stable * "e" RCP thermal barrier temperature is 80 OF and stable

• All other Rep support conditions are within range for starting the RCPs. 8ased on these conditions and lAW the guidance provided in ES-0.1, what is the order of priority for starting the Reps? Start: A. "C" Rep, then "8" Rep, then "A" Rep B. "8" Rep, then "Al! Rep, then "C" RCP C. "A" Rep, then "8" RCP, DO NOT start "e" Rep D. "C" Rep, then "8" RCP, DO NOT start "An Rep Answer: C Explanation/Justification: Incorrect.

This would be the priority if the candidate does not recognize that normal support conditions do not exist for the "C" RCP and that the "8" pump supplies the "6" spray Incorrect.

This is the starting sequence in FR-C.1 to address ICC conditions. Correct. lAW ES-0.1 step 12 and the preceding note and the bases document for this step and note. The SRO must be familiar with the and bases of ES-0.1 including EOP Attachment A-1.31. SRO Only in that the SRO must assess plant conditions (normal, abnormal, emergency) and then select a procedure or section of a procedure to mitigate, recover, or with which to Incorrect.

This would be the sequence if the candidate does not recognize that the C RCP seal injection flow is too low and believes that the RCP thermal barrier temperature is too high support RCP Sys # System KA Statement 000007 Reactor Trip Generic Knowledge of the operational implications of EOP -Stabilization warnings, cautions, and notes. KlA# 2.4.20 KIA Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

ES-0.1 step 12 and preceding note. ES-0.1 step 12 and preceding note bases; EOP Attachment 1.31. Question Source: New Question Cognitive Level: High -

10 CFR Part 55 Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 77. A Large Break LOCA coincident with some fuel damage. A General Emergency has been declared at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />. A non-routine airborne release of radioactive material as a result of this event is in progress due to 2FWE*P22 [Steam Driven AFW Pump] operation. No radioactive release has occurred or is imminent (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). The TSC has NOT yet been activated. Health Physics has provided the following dose At the EAB: 15 mRem TEDE; 10 mRem At 5 miles: 2.9 mRem TEDE; 3.5 mRem At 2 miles: 5.0 mRem TEDE; 8 mRem The following meteorological conditions exist: 35' wind direction is from 110 0 at 8 MPH. 150' wind direction is from 135 0 at 15 MPH. 500' wind direction is from 150 0 at 20 MPH. Based on these conditions, what Protective Action Recommendation (PAR) is (Refer to attached Evacuate 0-5 miles, 360 degrees and AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer KI in accordance with the State plan. Evacuate 0-2 miles, 360 degrees AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer KI in accordance with the State plan. Evacuate 2 miles, 360 degrees and 5 mile downwind wedge NPQRAB AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer KI in accordance with the State plan. Evacuate 2 miles, 360 degrees and 5 mile downwind wedge MNPQRAB AND shelter the remainder of the 10 mile EPZ AND advise the general public to administer KI in accordance with the State plan. Answer: C Explanation/Justification: Incorrect.

If the candidate incorrectly applies 1 12-EPP-IP-4. 1 , Offsite Protective Actions Attachment A they will select this answer. Incorrect.

If the candidate incorrectly applies 1/2-EPP-IP-4.1, Offsite Protective Actions Attachment A they will select this answer. Correct. lAW 1/2-EPP-IP-4.1, Offsite Protective Actions Attachment A. 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 task. Incorrect.

If the candidate incorrectly applies the 35 foot wind speed to the wedge calculation, they will select this answer. Sys# System KA Statement 000011 Large Break Generic Knowledge of emergency plan protective action LOCA recommendations.

KlA# 2.4.44 KIA Importance 4.4 Exam Level SRO References provided to Candidate 1/2-EPP-IP-4.1, Offsite Technical

References:

1 12-EPP-IP-4. 1 , Offsite Protective Actions Protective Attachment A Question Source: New Question Cognitive Level: High -

10 CFR Part 5f, Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 78. The plant is at 100% power with all systems in NSA. An unisolable leak occurs in the Primary Plant Component Cooling Water (CCP) discharge header. Pri Comp Cooling Surge Tank Level [2CCP*LCV100A and 8] valves are in Auto and full open. CCP Surge Tank Level is slowly dropping. The crew has entered AOP 2.15.1, Loss of Primary Plant Component Cooling Water. lAW the guidance provided in AOP 2.15.1, which of the below listed conditions will require a manual reactor trip? CCP pump flow and amps________._________ A. fluctuating AND CCP Surge Tank Level drops to offscale low B. fluctuating AND CCP Surge Tank Level drops to 3 inches C. steady AND CCP Surge Tank Level drops to offscale low D. steady AND CCP Surge Tank Level drops to 3 inches Answer: A Explanation/Justification: Correct. lAW AOP 2.15.1 Rev.3 step 1 CAS RNO. The SRO must be familiar enough with the contents of the AOP to know what conditions will require a manual reactor trip. The candidate must interrupt the data given in the stem and determine that the conditions are met for directing a reactor trip. SRO Only in that the SRO must assess plant conditions (normal, abnormal, or emergency) and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed. Incorrect.

CCP Surge Tank Level dropping to 3 inches is the setpeint for isolation of the non-essential CCP header Incorrect.

In addition to CCP Surge Tank Level dropping to off scale low. AOP 2.15.1 also has, a requirement to have indication of cavitation the CCP pumps. This choice does not contain indications of Incorrect.

CCP Surge Tank Level dropping to 3 inches is the selpeint for isolation of the non-essential CCP header. There is also a requirement to have indication of cavitation on the CCP pumps. This choice does not contain indications of c;avitation.

Sys # System Category KA Statement 000026 Loss of Generic Ability to interpret and execute procedure steps. Component Cooling Water KlA# 2.1.20 KIA Importance 4.6 Exam Level SRO References provided to Candidate None Technical

References:

AOP 2.15.1 Rev.3 step 1 CAS RNO Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam The Plant is operating in Mode 3 with all systems in normal alignment for this Mode.

• Tavg is 547 of and stable * "A" Charging pump is running. * "B" MFW pump is running. 13S KV Motor Oper Disc SW S9-2A inadvertently opens and cannot be closed. The following annunciators are in alarm:

• AS-2C, 4160V EMERG BUS 2AE UNDERVOLTAGE

• AS-2A, 4160V EMERG BUS 2AE ACB 2A10 AUTO TRIP

• AS-6D, 4S0V EM ERG BUS 2N UNDERVOLTAGE The following breakers have their white indicating lights LIT and their red indicating lights NOT LIT:

• 2-1 Emer Gen Output BKR ACB 2E10

• 2A SS Serv TFMR To 4KV Bus 2A ACB 42A

• 4KV Bus 2A To Emer Bus 2AE ACB 2A10

• 4KV Emer Bus 2AE To 4KV Bus 2A ACB 2E7 The following annnciators are NOT in alarm:

• AS-2B, 4160V EMERG BUS 2AE ACB 2E7 OVERCURRENT TRIP

• A8-4C, DIESEL GEN 2-1 ELECTRICAL FAULT Based on these conditions, what procedure entry is required and what actions will be required? Enter AOP 2.36.1, Loss of All AC Power When Shutdown and attempt to start and load the 2-1 emergency diesel generator. Enter AOP 2.36.1, Loss of All AC Power When Shutdown and DO NOT attempt to start the 2-1 emergency diesel generator and go to AOP 2.37.1, Loss of 4S0 VAC Emergency Bus. Enter AOP 2.36.2, Loss of 4KV Emergency Power and DO NlOT attempt to start the 2-1 emergency diesel generator and go to AOP 2.37.1, Loss of 480 VAC Emergency Bus. Enter AOP 2.36.2, Loss of 4KV Emergency Power and attempt to start and load the 2-1 emergency diesel generator.

Answer: 0 Explanation/Justification: Incorrect AOP 2.36.1, Loss of All AC Power When Shutdown is only applicable if RHS is being used to control RCS temperature.

Even though the title suggests that it is applicable when shutdown.

Correct actions. Incorrect.

AOP 2.36.1, Loss of AI! AC Power When Shutdown is only applicable if RHS is being used to control RCS temperature.

Even though the title suggests that it is applicable when shutdown.

Incorrect actions although this action may be warranted

.. Incorrect.

Correct procedure entry. Incorrect action, although this action would be correct if either AB-2B or AB-4C were in alarm. Correct. lAW AOP 2.36.2 Rev.12 step B. SRO Only in that the SRO must assess plant conditions (normal, abnormal, or emergency) and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed. SRO must determine that ANN AB-2B is no longer lit and it is acceptable to energize the emergency bus. Sys# System Category KA Statement 000056 Loss of AA2. Ability to determine and interpret the following as they apply to Indicators to assess status of ESF breakers Offsite Power the Loss of Offsite Power: (tripped/not-tripped) and validity of alarms (false/not-false)

<lA# AA2.45 KIA Importance 3.9 Exam Level SRO References provided to Candidate Technical

References:

AOP 2.36.2 Rev.13 step 8. Question Source: New Question Cognitive Level: High -

10 CFR Part 55; Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written E:Kam 80. The Unit is operating at 100% power with all systems in NSA A large leak occurs in the Service Water System. The control room receives A1-4H, "SERVICE WATER SYSTEM TROUBLE" followed shortly after by A1-4G, "SERVICE WATER HEADER PRESSURE LOW". "A" &"B" SW Header Pressures BOTH indicate 28 psig and slowly DROPPING. "A" & "B" CCS Water HX Service Water Supply Header Isolation (2SWS*MOV107 NB/CID) automatically isolate AND cannot be AFTER 2SWS*MOV1 07 NB/C/D automatically isolate, "A" "B" SW Header Pressures begin to RISE. (1) Based on these plant conditions, which Service Water System component is leaking? (2) lAW AOP 2.30.1, Service Water/Normal Intake Structure Loss which of the below listed components are required to be tripped? All Station Air Compressors All Main Feed Pumps All Heater Drain Pumps All Condensate Pumps A. (1) The in service Primary Component Cooling Heat Exchangers

[2CCP*E21A, B ,C] (2) ONLY the Main Feed Pumps and Condensate Pumps B. (1) The in service Centrifugal Water Chillers [2CDS-CHL23A, B, C] (2) ONLY the Main Feed Pumps and Heater Drain Pumps C. (1) The in service Primary Component Cooling Heat Exchangers

[2CCP*E21A, B ,C] (2) ONLY Station Air Compressors and Heater Drain Pumps D. (1) The in service Centrifugal Water Chillers [2CDS-CHL23A, B, C] (2) ONLY the Station Air Compressors and Condensate Pumps Answer: B Explanation/Justification: Incorrect.

Incorrect.

leaking component and condensate pumps are not to be tripped in AOP 2.30.1. Correct. Since pressure recovered when 2SWS*MOV107A1B/CID isolated, the leak must be in the secondary side header. The Centrifugal Water Chillers are on the secondary side header and the Primary Component Cooling Heat Exchangers are on the primary side header. lAW AOP 2.30.1 If header pressure cannot be restored above 34 psig the main feed pumps and heater drain pumps are to be stopped. Additionally, the station air compressors have a backup supply of cooling water that can be placed in service and the AOP directs the starting of at least one condensate pump. The first part of the question can be answered with RO knowledge.

The second part is SRO only since it requires specific knowledge of procedure content and cannot be answered with system knowledge alone. The SRO must assess plant conditions (normal, abnormal, or emergency) and then select a procedure or section of a procedure to mitigate, recover, or with which to proceed. Incorrect.

Incorrect leaking component and air compressors are not to be tripped in AOP 2.30.1. Incorrect.

Correct leaking component and air compressors and condensate pumps are not to be tripped in AOP 2.30.1. Sys # System Category KA Statement 000062 Loss of AA2. Ability to determine and interpret the following as they apply to Location of a leak in the SWS Nuclear the Loss of Nuclear Service Water: Service Water KlA# AA2.01 KIA Importance 3.5 Exam Level SRO References provided to Candidate None Technical

References:

AOP 2.30.1 Rev. 9 Steps 7, 9, & 10. Simplified SWS Drawing (2SQS-LP-301 slides 6 and 8). Question Source: Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 10 CFR (SRO Beaver Valley Unit 2 NRC Written Exam 81. The plant was at 100% power with all systems in NSA. A main steam line break affecting all 3 SGs occurred. The crew is currently performing ECA 2.1, Uncontrolled Depressurization Of All Steam Generators. AFW flow has been throttled to 50 gpm to each SG to minimize the RCS cooldown. Safety Injection Termination Criteria have been met. The crew has just stopped all but one charging pump lAW step 16 of ECA-2.1. The following Steam Generator conditions exist: SG SG Pressure SG"A" 20% WR and 320 psig decreasing SG"8" 22% WR and 310 psig decreasing SG"C" 26% WR slowly 420 psig increasing Which of the following describes the required procedure transition, if any, and what is the bases for this decision? Transition to FR-H.1, Loss Of Secondary Heat Sink; there is a RED condition on the Heat Sink Status Tree . Transition to ES-1.1, SI Termination; the SI termination criteria have been met. Transition to E-2, Faulted Steam Generator Isolation; there is an intact SG available. Continue with ECA 2.1, Uncontrolled Depressurization Of All Stearn Generators; Safety Injection termination is not complete.

Answer: D Explanation/Justification: Incorrect.

Both SG level and AFW flow meet the criteria for FR-H.1 entry, However a Caution prior to Step 3 indicates FR-H.1 would not be entered since, Operator action reduced feed. Incorrect.

There are no transitions to ES-1.1 within ECA-2.1. ECA-2.1 has the necessary steps to address SI termination.

Additionally, termination transitions would only occur after transitioning to E-2 Incorrect.

lAW LHP action of ECA-2.1 requires transition to E-2 when anyone SG pressure increases UNLESS SI termination is in progress has not yet been Correct. lAW LHP action of ECA-2.1 requires transition to E-2 when anyone SG pressure increases UNLESS SI termination is in progress has not yet been completed.

SRO Only in that the SRO must assess plant conditions (normal, abnormal, or emergency) and then select procedure or section of a procedure to mitigate, recover, or with which to System Category KA Statement Steam Line EA2. Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Rupture -the (Uncontrolled Depressurization of all Steam Generators) procedures during abnormal and emergency Excessive operations.

Heat Transfer KlA# EA2.1 KIA Importance 4.0 Exam Level SRO provided to Candidate None Technical

References:

ECA-2.1 LHP .luestion Source: New Question Cognitive Level: High -

10 CFR Part 55 Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 82. The Plant is stable at 55% power with all systems in normal alignment for this condition. Tavg is 565°F and stable Control Bank D is at 190 steps. Control Bank D Demand step counters are at 190 steps. Control Rod Group Selector Switch is in the "MAN" position.

Plant Parameters are NOW as follows: Tavg is 562 OF and slowly dropping. RCS Pressure is 2230 psig and slowly dropping. A4-8G, Rod Position Deviation is in alarm. Reactor power has dropped to 51 % and is slowly rising. PR N-43 Negative Rate Trip bistable is LIT All other PR Negative Rate Trip bistables are NOT LIT Control Bank D Demand step counters remain at 190 steps. DRPI indication for Rod D4 is ZERO steps. Based on these conditions:

What procedure contains the REQUIRED guidance to address these plant conditions?

A. E-O, Reactor Trip Or Safety Injection.

B. AOP 2.1.7, Rod Position Indication Malfunction.

C. AOP 2.2.1 C, Power Range Channel Malfunction.

D. AOP 2.1.8, Rod Inoperability.

Answer: D Explanation/Justification: Incorrect.

The PR rate coincidence is 2/4 and only one rate bistable has been actuated.

Power is above P-9 (50%) however since there is only one rate channel above the trip setpoint.

No reactor trip should occur and E-O entry is not required. Incorrect.

At BVPS this procedure may be entered as part of the initial diagnostics, however entry into this procedure is not REQUIRED and procedure will NOT contain the necessary guidance to address the dropped rod situation posed in the question.

Additionally, the section of this procedure address no corresponding power change. In the stem of the question, power has changed. The question asks for the procedure that contains the recovery guidance and NOT what procedure entry is required.

The phrasing of the question makes choice clearly Incorrect.

With power dropping and temp dropping and power recovering there MUST be some negative p being added (dropped rod). The PR indications are therefore consistent with this negative p addition and they are not malfunctioning. Correct. lAW AOP 2.1.8 symptoms the alarms and plant response are conSistent with a dropped rod. In order to properly diagnosis the dropped rod, the SRO must use the excore response along with the loop Tavg response to conclude that there is in fact a dropped rod and this is not than a DRPI malfunction, or a NIS malfunction, and NO trip setpoint has been exceeded.

AOP 2.1.8 addresses a dropped rod in Part A SRO candidate must evaluate the given conditions and those that are NOT present to determine that a rod has dropped, and is in fact at zero steps. SRO Only in that the SRO must assess plant conditions and then select a procedure to mitigClte, recover, or with which to proceed. This is more than just entry conditions for an AOP which would be RO knowledge.

This requires the SRO to have specific knowledge of the procedure content. Sys# System Category KA Statement 000003 Dropped AA2. Ability to determine and interpret the following as they apply to Dropped rod, using in-core/ex-core Control Rod the Dropped Control Rod: instrumentation, in-core or loop temperature measurements KJA# AA2.03 KJA Importance 3.8 Exam Level SRO provided to Candidate None Technical

References:

AOP 2.1.8 Rev. 3 pages 1 & 2 Question Source: New Question Cognitive Level: High Analysis 10 CFR Part 55 Content: 10 CFR 55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 83. The plant was at 100% power with all systems in NSA. A Large Break LOCA occurs coincident with a loss of all LHSI flow. The 5 hottest core exit TICs reach 730 of and the crew Transitions to FR-C.2, Response to Degraded Core Cooling. 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the trip, the following conditions exist: CNMT pressure is 4 psig and slowly dropping. All CNMT spray systems are operating as designed. The 5 hottest core exit TICs are 750 of and slowly rising. In-Containment High Range Area radiation monitors [2RMH-RQ206

& 207] are reading 2.0 X 10 7 mRlhr and stable. All RCPs have been secured. RVLlS Full Range level is 33% and slowly dropping.

Based on these conditions, what Emergency Action Level (EAL) classification is required? (Refer to attached reference)

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: 0 Explanation/Justification: Incorrect.

Candidates would choose this if they only recognized the potential loss of the CNMT barrier. Incorrect.

Candidates would choose this if they only recognized the loss of the fuel clad or ReS barrier. Incorrect.

Candidates would choose this if they failed to recognized the potential loss of the CNMT barrier. Correct. lAW EPPII-1b Attachment A Fission Product Barrier Matrix. SRO only in that it requires the implementation of administrative procedures that specify implementing emergency procedures.

Specifically, implementing the E-Plan. The SRO must utilize the computer data provided (PSMS and DRMS) to determine that a RED path condition (ICC) exist for core cooling and this RED path condition means that a loss of both the fuel clad barrier and the RCS barrier are present as a result. Additionally, the CNMT barrier is also potentially lost as a result of the rad monitor readings and the time since RX trip. Sys # System Category KA Statement 000074 Inadequate Generic Ability to use plant computers to evaluate system Core Cooling or component status. KlA# 2.1.19 KIA Importance 3.8 Exam Level SRO References provided to Candidate EPPII-1 b Attachment A Technical

References:

EPPII-1 b Attachment A Question Source: New Question Cognitive Level: High -

10 CFR Part 55 Content: 10 CFR 55.43(b )(5) Objective:

---

(SRO Beaver Valley Unit 2 NRC Written Exam 84. The plant was operating at 100% power with all systems in NISA 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 8, Check if SI flow should be reduced. The following plant conditions exist: RCS pressure is 1085 psig and stable All S/G NR level are 33% and stable Core exit TICs are 452 OF and slowly dropping All S/G pressures are 435 psig and slowly dropping RWST level is 490 inches and slowly dropping CNMT pressure is 3.5 psig and stable Total AFW flow is 900 gpm and stable PRZR level is 18% and slowly rising Based on these conditions, what procedural transition is Required?

Transition A ECA-2.1, Uncontrolled Depressurization of All Steam Generators B. ES-1.2, Post LOCA Cooldown and Depressurization C. ES-1.3, Transfer to Cold leg Recirculation D. ES-1.1, SI Termination Answer: D Explanation/Justification: Incorrect.

All S/G pressures are dropping, however they are dropping as a result of the RCS cooldown.

The S/Gs are not faulted. Incorrect.

SI termination criteria are met, which negates the need to perform ES-1.2. Incorrect.

RWST level is not low enough to meet the procedure transition (400 inches). Correct. lAW E-1 step 8 and E-1 LHP 51 termination criteria are met. 5RO only since it requires the 5RO to assess plant conditions abnormal, or emergency) and then select a procedure or section of a procedure to mitigate, recover, or with which to System Category KA Statement SI EA2. Ability to determine and interpret the following as they apply to Facility conditions and selection of appropriate Termina-tion the (51 Termination) procedures during abnormal and emergency operations.

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

References:

E-1 step 8 Question Source: New Question Cognitive Level: High -

10 CFR Part Content: 10 CFR55.43(b)(5)

Objective:

(SRO ONLY) Beaver Valley Unit 2 NRC Written Exam (2LOT8) The plant was in Mode 3 with all systems in normal alignment for this Mode and RCS temperature at 54rF and STABLE. A SG tube leak occurred on the 21A SG and the crew has entered AOP 2.6.4, Steam Generator Tube Leakage. Letdown flow has been reduced to 45 gpm. 21A SG has been isolated (Steam flow out and feed flow in are BOTH isolated). An RCS cooldown to 500°F has been initiated. Charging flow is 55 GPM and STABLE. PRZR level is 22% and slowly dropping. 21A SG NR level is 95% and slowly rising All PRZR heaters are OFF. SI has NOT been actuated.

The crew has progressed through AOP 2.6.4 to step 17 "Control RCS pressure and Charging flow to Minimize RCS-to-Secondary leakage" (step 17 is a continuous action step). Based on these conditions, and lAW the guidance in AOP 2.6.4, how will charging flow and RCS pressure be controlled to minimize RCS-to-Secondary leakage and prevent SG overfill that would lead to water entering the steam lines? Lower charging flow and depressurize the RCS. Lower charging flow and equalize RCS and 21A SG pressures. Raise charging flow and depressurize the RCS. Raise charging 'flow and equalize RCS and 21 A SG pressures.

Answer: C Explanation/Justification: Incorrect.

These are the actions from AOP 2.6.4 step 17 if PRZR level is between 50% and 713 % and SG level is rising. Incorrect.

Lowering charging flow would allow RCS pressure to drop which would "backfill" water from the ruptured SG and raise PRZR level. However. this will only work if RCS pressure is allowed to drop. Maintaining RCS and 21A SG pressures equal will result in PRZR level dropping more rapidly Correct. lAW AOP 2.6.4 step 17 chart. SRO only since it requires the SRO to assess plant conditions and then select a section of a with which to proceed. Specifically.

the appropriate actions from step 17 that will prevent SG overfill and thus prevent water entry into Incorrect.

These are the correct actions if 21A SG is "offscale" high. Sys # System Category KA Statement 000037 Steam AA2. Ability to determine and interpret the following as they apply Actions to be taken if S/G goes solid and water Generator (S/G) to the Steam Generator Tube Leak: enters steam lines Tube Leak KlA# AA2.14 KIA Importance 4.4 Exam Level SRO References provided to Candidate None Technical

References:

AOP 2.6.4 step 17 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 10 CFR55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 86. The plant is in Mode 4 with the following plant conditions: All SI Accumulators have been isolated. OPPS is in service with PRZR PORVs [2RCS*PCV455C

& 456] operable with lift settings within the limits specified in the Pressure Temperature Limit Report (PTLR). All RCS cold leg temperatures are below the enable temperature specified in the PTLR. Only one Charging pump is capable of injecting into the ReS. A steam bubble exists in the PZR. RCS pressure is 400 psig and stable. RCS pressure then rises above the variable lift setting pressure specified in the PTLR. Neither PRZR PORV [2RCS*PCV455C

& 456] automatically opens. The RO (ATC) attempts to open PRZR PORV [2RCS*PCV456]

but it will NOT open. The RO (ATC) manually opens PRZR PORV [2RCS*PCV455C]

and reduces RCS below the variable lift setting pressure specified in the PTLR. RCS pressure is STABILIZED at 400 psig. Based on these plant conditions and this sequence of events, what Tech Spec actions will be required? (Refer to attached reference)

A. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> depressurize the RCS and establish an RCS vent of.?: 3.14 in 2. B. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore PRZR PORVs [2RCS*PCV455C

& 456] to operable status. C. Within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> enter Mode 5. D. Within 7 days restore PRZR PORV [2RCS*PCV456]

to operable status. Answer: A Explanation/Justification:

A. Correct. lAW TS 3.4.12 condition G. SRO only since it involves application of Required Actions in Section 3 of the TS. Incorrect.

This would be the required action if the plant was in Mode 5. Incorrect.

This would be the T5 3.0.3 required action if no action statement was available for the conditions in the stem. Incorrect.

This would be the required action if the candidate believes 2RCS*PCV455C was operable since in was manually opened. Sys # System Category KA Statement 010 Pressurizer A2 Ability to (a) predict the impacts of the following malfunctions or PORV failures Pressure operations on the PZR PCS; and (b) based on those predictions, use Control procedures to correct, control, or mitigate the consequences of those System (PZR malfunctions or operations:

PCS) KlA# A2.03 KIA Importance 4.2 Exam Level SRO References provided to Candidate T53.4.12 Technical

References:

T5 3.4.12 Condition G Question Source: New Question Cognitive Level: High Application 10 CFR Part 55 Content: 10 CFR 55.43(b)(2)

Objective:

(SRO Beaver Valley Unit 2 NRC Written E)(am 87. The plant is operating at 100% power with all systems in normal alignment for this power level.

• A Loss of 125VDC Bus 2-2 occurs

• All systems function as designed (1) What impact, if any, will this loss of 125VDC Bus 2-2 have on Rx trip breaker status? (2) lAW the guidance provided in AOP 2.39.1 B, Loss of 1:25VDC Bus 2-2, what compensatory action would be necessary to control SG water level? The Reactor trip breakers will (1)__In order to control SG water level it will be necessary to ____(2)____A. (1) open (2) control Auxiliary Feedwater flow B. (1) remain closed (2) place the Startup Feedwater pump in service C. (1) open (2) control SG Feedwater Bypass Control Vlvs [2FWS*FCV 479,489, & 499] D. (1) remain closed (2) control SG Main Feed Reg Vlvs [2FWS*FCV 478,488, &498] Answer: A Explanation/Justification: Correct. lAW AOP 2.39.1 B rev. 3 Attachment 1 page 7. There is no direct Rx trip from loss of this DC bus. SRO must realize that the trip will occur from loss of both the main feed reg valves and the bypass valves. SRO only by ensuring that the additional knowledge of the procedure's content is required; Assessing plant conditions and then selecting a section of a procedure to mitigate, recover, or with which to proceed. Incorrect.

The reactor will trip on low SG water level. Placing the Startup feedwater pump in may help control SG level, but is addressed in the AOP Incorrect.

The reactor will trip on low SG level but the bypass feed reg valves will not be available. Incorrect.

The reactor will trip on low SG water level. The main feed reg valves are not available.

Sys# System Category KA Statement 012 Reactor A2 Ability to (a) predict the impacts of the following malfunctions or Loss of dc control power Protection operations on the RPS; and (b) based on those predictions, use System, procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

KlA# A2.07 KIA Importance 3.7 Exam Level SRO References provided toCandidate None Technical

References:

AOP 2.39.1 B rev. 3 Attachment 1 page 7 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 10 CFR55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 88. The plant is operating at 100% power with all systems in NSA. Maintenance is performing 2MSP-13.01-1, 2QSS-L 104A, Refueling Water Storage Tank TK21 Level Loop Channell Test Maintenance reports the as found setpoint for: 2QSS-LSEL 1 04A RWST Ext-Lo Level SI Switchover Comparator Trip was 29' 6" (Tech Spec Allowable Value is between 31' 8" and 31' 10") AND 2QSS-LSL 104A Recirc Spray Pump Start Interlock Comparator Trip was 32' 9" (Tech Spec Allowable Value is between 32' 8" and 32' 10") If this channel was returned to service in this condition, what would be the status of the following Tech Spec required Engineered Safety Feature Actuation System (ESFAS) Instrumentation? RWST level Extreme low SI Switchover RWST level low Recirc Spray Pump Start Interlock This channel of: _________________

_ RWST level Extreme low AND RWST level low are still OPERABLE ONLY if a risk assessment is performed. RWST level Extreme low is INOPERABLE AND RWST level low is; still OPERABLE ONLY if a risk assessment is performed. RWST level Extreme low is INOPERABLE AND RWST level low is still OPERABLE. BOTH RWST level Extreme low AND RWST level low are INOPERABLE.

Answer: C Explanation/Justification: Incorrect.

Extreme level low is inoperable since it is outside the allowable band on the low side. If it was outside on the high side, it could still be operable dependent on the outcome of an evaluation to determine if it could still perform its function.

Tech Spec 3.0.4 allows these types of risk assessments, but only for mode changes NOT AT POWER. Level low is still in the band and operable.

Candidate may believe the entire level transmitter function is inoperable until the assessment confirms operability. Incorrect.

It is correct that extreme level low is inoperable.

Candidate may believe level low can only be operable dependent on the outcome of an evaluation to determine if it could still perform its function.

Tech Spec 3.0.4 allows these types of risk assessments, but only for mode changes NOT AT POWER Correct. IAWTS page 3.3.2-9 item 2.b.2.and page 3.3.2-13 item 7.b Incorrect.

It is correct that extreme level low is inoperable.

Level low is within band and operable.

Candidate may believe the all level functions are inoperable if anyone level switch Sys # System. Category KA Statement 013 Engineered Generic Ability to determine operability and/or availability Safety of safety related equipment.

Features Actuation System (ESFAS) KlA# 2.2.37 KIA Importance 4.6 Exam Level SRO provided to Candidate None Technical

References:

TS page 3.3.2-9 item 2.b.2.and page 3.3.2-13 item 7.b Question Source: New Question Cognitive.

Level: High -AnalYSis 10 CFR Part 55 Content: 10 CFR 55.43(b)(2)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 89. Given the following conditions:

• The plant is at 100% power with all systems in NSA.

• The RO (ATC) has recently performed a small dilution for Tavg control. Shortly after the dilution, the following conditions exist:

• Power Range NI's are increasing.

• Tavg is decreasing.

• Steam flow and feed flow are slightly elevated.

• Reactor power is 101 % and rising slowly. Which ONE of the following describes the event in progress and the action required?

A. Main steam line leak; reduce power to less than 100% by reducing turbine load as necessary.

B. Inadvertent Res dilution; reduce power to less than 100% by adjusting control rods. C. Main steam line leak; trip the reactor and enter E-O, Reactor Trip Or Safety Injection.

D. Inadvertent RCS dilution; trip the reactor and enter E-O, Reactor Trip Or Safety Injection.

Answer: A Explanation/Justification: Correct. Conditions in the stem indicate a steam break. lAW AOP 2.51.2 step 2 RNO. Diagnosis of the event is RO knowledge.

Selecting the appropriate procedure and directing the appropriate actions is SRO only knowledge.

SRO must evaluate plant conditions and determine appropriate procedural action. in this case power must be brought below 100% and step 2 of the AOP is a continuous action step for the SRO to direct the crew to reduce turbine load as necessary and at rate determined by the SRO. Incorrect.

Candidate may believe that something about the recent dilution has caused these conditions.

The dilution will cause power to rise initially until the MTC feedback offsets the effects. This does not explain why Tavg is dropping and steam/feed are rising. Reducing power is the correct action. but not by rod movement. Incorrect.

Conditions in the stem indicate a steam break. AOP 2.51. 2 does not direct a reactclr trip. If the steam break indications were more severe, then a trip would be warranted based on approaching the high flux trip setpoint. Incorrect.

Candidate may believe that something about the recent dilution has caused these conditions.

The dilution will cause power to rise initially until the MTC feedback offsets the effects. This does not explain why Tavg is dropping and steam/feed are rising. AOP 2.51. 2 does not direct a reactor trip. If the steam break indications were more severe. then a trip would be warranted based on approaching the high flux trip setpoint.

Sys# System Category KA Statement 039 Main and A2 Ability to (a) predict the impacts of the following malfunctions or Increasing steam demand, its relationship to Reheat operations on the MRSS; and (b) based on predictions.

use increases in reactor power Steam procedures to correct, control, or mitigate the consequences of those System malfunctions or operations: (MRSS) KlA# A2.05 KIA Importance 3.6 Exam Level SRO References provided to Candidate None Technical

References:

AOP 2.51.2 Rev.O step 2 CAS and RNO Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 55 Content: 10 CFR55,43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 90. The Plant is operating at 100% power with all systems in NSA EXCEPT: 2MSS*SOV1 05A Turb Driven AFW Pump STM HDR A Supply Isol valve was placed on clearance today (September

1) at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for SOV replacement. At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> today, 2MSS*SOV105E Turb Driven AFW Pump STM HDR B Supply Isol valve Is Declared INOPERABLE. (The valve cannot be cycled open or closed) Based on these conditions, what Tech Spec action(s) will be required? (Refer to attached reference)

A. Be in Mode 3 by 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on September 1 AND Mode 4 by 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> on September

2. B. Restore AFW train to OPERABLE status by 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on September

4. C. Restore 2MSS*SOV105E OR 2MSS*SOV1 05A to OPERABLE status by 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on September

8. D. Restore 2MSS*SOV105E OR 2MSS*SOV105A to OPERABLE status by 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on September

11. Answer: C Explanation/Justification: Incorrect.

This would be the required action if the candidate believes that having two steam supply lines unavailable constitutes two trains of Incorrect.

This would be the required action if the candidate believes that having two steam supply lines unavailable constitutes an steam driven AFW pump. The requirement to re-align the supply headers is already being met by the NSA Correct. lAW T. S. 3.7.S and bases ONLY two of three steam supply lines are required for steam driven AFW pump operability.

The candidate will need to use the TS bases to recognize that ONLY two of three steam supply lines are required for steam driven AFW pump operability.

The Condition A statement specifically says one of the required steam supply lines inoperable.

Therefore, taking 2MSS*SOV10SA out of service would not require any TS action since 2 trains are still available.

When 2MSS*SOV10SE fails to meet the required stroke time, it must be declared inoperable and Condo A action would apply. It is an SRO responsibility to be familiar enough with the operability requirements for the SOV to declare it inoperable based on the performance data presented in the stem. BRO only because it requires Application of Required Actions in Section 3 of the TS, which is an SRO responsibility. Incorrect.

This would be the required action if the candidate believes that the 10 day statement in Condo A provides an additional allowance two inoperable steam supplies.

The bases for this 10 day statement is to limit the time allowed in thi condition when Condo A and B are Sys# System Category KA Statement 061 Auxiliary I Generic Knowledge of operator responsibilities during all Emergency modes of plant operation.

Feedwater (AFW) System KlA# 2.1.2 KIA Importance 4.7 Exam Level SRO References provided to Candidate T. S. 3.7.S and bases Technical

References:

T. S. 3.7.5 Cond A and bases Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 10 CFR 55.43(b)(2)

Objective:

(SRO Beaver Valley Unit 2 NRC Written E:Kam A Plant startup is in progress with the reactor critical at 10 -8 amps on the intermediate range. All systems are in normal alignment for this condition.

• Annunciator A4-4E, NIS Detector/Compensator Trouble alarms

• The Loss of Comp.volt status light is LIT on the N-35 Intermediate Range drawer. IF the reactor were to trip with these conditions, N35 intermediate range indication would be reading __(1) than N36 intermediate range indication.

In order to maintain power operations, the AOP 2.2.1 B, Intermediate Range Channel Malfunction, required actions are to place the N-35 LEVEL TRIP switch to the bypass position AND ______(2)_______ (1) lowe( (2) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> EITHER reduce thermal power to < P-6 OR Raise thermal power to> P-10. (1) higher (2) Place BOTH the Intermediate Range A and B block switches to BLOCK. (1) lower (2) Place BOTH the Intermediate Range A and B block switches to BLOCK. (1) higher (2) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> EITHER reduce thermal power to < P-6 OR Raise thermal power to> P-10. Answer: 0 Explanation/Justification: Incorrect.

Wrong response, Correct action Incorrect.

Correct response.

Wrong action. These are the actions to be taken if power is abov.e P-10. Placing the block switches to block is not procedurally allowed until the P-10 permissive is received. Incorrect.

Wrong response.

Wrong action. These are the actions to be taken if power is above, P-10. Placing the block switches to block is not procedurally allowed until the P*10 permissive is received. Correct. It is an RO fundamental knowledge to predict what impact loss of compensating voltage will have on the IR response.

Lesson plan 35QS-2.1 slide 18 illustrates this response.

Correct action lAW AOP 2.2.1 B step 5. SRO only by ensuring that the additional knowledge of the procedure's content is required; Assessing plant conditions and then selecting a section of a procedure to mitigate, recover, or with which to proceed. Sys # System Category KA Statement 015 Nuclear A2 Ability to (a) predict the impacts of the following malfunctions Faulty or erratic operation of detectors or Instrumentation or operations on the N15; and (b) based on those predictions, use compensating components System. procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

KlA# A2.02 KIA Importance 3.5* Exam Level SRO References provided to Candidate None Technical

References:

AOP 2.2.1 B rev. 3 step 5 Question Source: New Question Cognitive Level: High -Analysis 10 CFR Part 5fi Content: 10 CFR5S.43(b)(S)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 92. The Unit 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. Tube Down B. Underload C. Overload D . Bridge-Trolley-Hoist Answer: B Explanation/Justification: Incorrect.

Tube down interlock will stop hoist downward motion when the hoist is all the way down. Correct. lAW LP 3508-6.13 slide 49. (2RP-3.3)

SROs are responsible for the assessment of fuel handling equipment surveillance requirement acceptance criteria.

and this is a required manipulator crane interlock for fuel movement. Incorrect.

Overload will stop UPWARD motion if an assembly is binding while moving upward. Incorrect.

Bridge-Trolley-Hoist interlock will only allow motion/movement in one direction at a time. ;ys # System Category KA Statement 034 Fuel Handling K4 Knowledge of design feature(s) and/or interlock(s) which provide Fuel protection from binding and dropping Equipment for the following:

System (FHES) KlA# K4.01 KIA Importance 3.4 Exam Level SRO References provided to Candidate Technical

References:

LP 3S0S-6.13 slide 103. (2RP-3.3)

Question Source: BVPS2 Question Cognitive Level: Low-Memory 10 CFR Part 55, Content: 10 CFR (SRO Beaver Valley Unit 2 NRC Written Exam 93. The Plant is operating at 100% power with all systems in NSA. An inadvertent Reactor Trip occurs After the automatic FAST bus transfer, SSST 2A experiences an under frequency condition, which causes 4Kv busses 2A and 2B to de-energize. All other systems function as designed.

The crew enters E-O, Reactor Trip or Safety Injection and has just completed the Immediate Safety Injection has not actuated and is not Current plant conditions are as RCS pressure is 2100 pSig and slowly riSing. S/G pressures are 1000 psig and stable. Tavg is 550 OF and slowly rising Tcold is 542 OF in all three loops and stable. NR S/G levels are 10% and rising. Total AFW flow is 900 gpm and stable. For these plant conditions, what EOP procedural action(s) will be r'equired to AVOID an automatic Safety Injection actuation?

The Unit Supervisor will direct the crew to ____________

_ A Manually actuate Steam line Isolation

3. Place the Steam Dump Control in Steam Pressure Mode C. Throttle total AFW flow to greater than or equal to 300 gpm D. Place the Low Steamline Pressure SI block switches to block Answer: B Explanation/Justification: Incorrect.

Manually actuating SLI will physically prevent SI. However, there is NO EOP procedural guidance to perform this action. Correct. lAW 1I20M-53B.2 Iss 1C Rev. 7 page 7 preemptive action guidelines.

The SM!US should direct this action to avoid the SI. SRO only since the SRO must assess plant conditions and then selecting a section of a procedure to mitigate, recover, or with which to proceed. In this case the SRO must recognize that the setting up of natural circulation with the steam dumps in Tavg mode will result in an inappropriate Sf. The SRO must further recognize that an EOP preemptive action exists for this situation, and direct the crew to place the steam dumps in the steam pressure.

Mode. Incorrect.

Throttling AFW flow is also a preemptive action. However, this is only required to control a cooldown and no cooldown exists. Additionally, throttling AFW flow will not prevent the SI for these conditions. Incorrect.

This would prevent the SI if RCS pressure was below 2000 psig. Since RCS pressure is above 2000 psig SI will not be blocked. Sys # System Category KA Statement 041 Steam Dump Generic Ability to locate control room switches, controls, System. and indications, and to determine that they (SDS) and correctly reflect the desired plant lineup. Turbine Bypass Control KlA# 2.1.31 KIA Importance 4.3 Exam Level SRO References provided to Candidate None Technical

References:

1/20M-53B.2 Iss 1C Rev. 7 page 7 Question Source: New ::!uestlon Cognitive Level: High -Analysis 10 CFR Part 55 Content: 10 CFRS5.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam NOP-LP-4011, FENOC Work Hour Control requires the Unit Supervisor to ensure that no personnel exceed 10 CFR 26 work hour limits without appropriate prior authorization.

Which of the below listed items are 10 CFR 26 work hour limits? (Assume both Units are at 100% power with all systems in NSA) 1. No more than 20 work hours in any 32-hour period. 2. No more than 16 work hours in any 24-hour period. 3. No more than 26 work hours in any 4B-hour period. 4. No more than 72 work hours in any 7-day. 5. No more than 72 work hours in any 16B-hour period. 6. A 34-hour break in any 9-day period. 7. A 40-hour break in any 216-hour period. 1, 2, 3, 4, 6, &7 ONLY 1, 4, 5, 6, & 7 ONLY 2, 3, 4, 5, &6 ONLY 2, 3, 4, 5, 6 &7 ONLY \nswer: C Explanation/Justification: Incorrect Items 1 and 7 are not required. Incorrect.

Items 1 and 7 are not required. Correct. lAW NOP-LP-4011 Rev. 5 pages 15 and 16. NOP-LP-4011 is one of the tools that BV uses to ensure the Tech Spec required minimum staffing requirements are being met. This meets the SRO only requirement for meting conditions and limitations in the facility license as defined in 10CFR 55.43(b)(1).

This is also an SRO only task at BV as stated in the NOP itself. Incorrect.

Item 7 is not required.

Sys# System KA Statement N/A N/A Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc, KlA# 2.1.5 KIA Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

NOP-LP-4011 Rev, 6 page 16 Question Source: New Question Cognitive Level: High Comprehension 10 CFR Part 55 Content: 10 CFR 55.43(b)(1)

ObJective:

(SRO ONLY) Beaver Valley Unit 2 NRC Written Exam (2LOT8) The Unit is in Mode 1 at 89% following a power reduction from 100%. Control Bank "0" Group 1 indicates the following:

• Group step counter position is 196 steps.

• ORPI indicates the following:

o Control Rod H02 at 192 steps. o Control Rod H14 at 204 steps. o Control Rod P08 at 174 steps. o Control Rod B08 at 180 steps. For these plant conditions, the required Tech Spec action is to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. What is the bases for this Tech Spec Action? Shutdown Margin is not met. Rod drop times cannot be met. AFO limits cannot be met. Accident analysis assumptions are not met. Answer: 0 Explanation/Justification: Incorrect.

SOM may be impacted by two misaligned rods. but this is not a given. TS for these conditions also requires SOM verification w/l 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and if necessary initiate boration to restore. It is not the bases for Mode 3 entry wll 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Incorrect.

It is true that one rod may take longer to drop than the rest of the rods in the bank. but this is not the reason for Mode 3 entry wll hours. TS required rod drop times are measured from full out Incorrect.

AFO may be adversely impacted by misaligned rods. However. the AFO actions are to reduce power below 50%. This would not be reason Mode 3 entry wll 6 Correct. lAW TS Bases page B 3.1.4-8 Action 02. The SRO must first recognize that the items in the stem are indicative of two rods that have violated the TS required rod alignment limits. The SRO must then explain why it required to enter Mode 3 wll 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. SRO only since it requires knowledge of TS bases that are required to analyze TS required actions. The SRO must also apply the TS knowledge that misaligned does not necessarily mean INOPERABLE.

Sys# System Category KA Statement NIA NIA Generic Ability to explain and apply system limits and precautions.

KlA# 2.1.32 KIA Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

TS Bases page B 3.1.4-8 Action 02 Question Source: New Question Cognitive Level: High -AnalysiS 10 CFR Part 55 Content: 10 CFR 55.43(b)(2)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 96.

• The Plant is operating at 100% power with all systems in NSA. Preparations are being made to receive New Fuel. The 480V power supply breaker [MCC-2-21 cubicle 2C] to the New Fuel elevator trips while testing the elevator with a "Dummy" fuel assembly.

Electrical Maintenance and System Engineering desire to implement troubleshooting activities on this breaker to determine the cause of the trip. The troubleshooting will NOT result in any unanticipated control room alarms. The troubleshooting will ONLY involve taking voltage/current readings, at the breaker with the breaker energized. The breaker has NO test points for voltage or current. lAW NOP-ER-3001, Problem Solving and Decision Making Process what type of Troubleshooting Plan will be required and what Minimum approval authority will be required? (Refer to attached reference)

A ____(1 )____ troubleshooting plan will be required.

Minimum required approval of this plan will be from ___ (2)___ A. (1) Simple (2) a SRO B. (1) Simple (2) a Manager C. (1) Complex (2) a Manager with concurrence of the Shift Manager D. (1) Complex (2) the Plant Duty Manager with concurrence of the Shift Manager Answer: A Explanation/Justification: Correct. lAW NOP-ER-3001 Rev. 5 Att. 3 pages 26 thru 30. SRO only since it requires the SHO to have a working knowledge of the process for making changes to plant equipment In particular the process for managing troubleshooting a::tivities. Incorrect.

Correct Plan. Wrong approval level. Incorrect.

Incorrect plan with appropriate approval for that plan if the candidate mis-applies the procedure. Incorrect.

Incorrect plan with appropriate approval for that plan if the candidate mis-applies the procedure.

Sys# System Category KA Statement N/A N/A Generic Knowledge of the process for managing troubleshooting activities.

KJA# 2.2.20 KJA Importance 3.8 Exam Level SRO References provided to Candidate NOP-ER-3001 Rev. 5 Technical

References:

NOP-ER-3001 Rev. 5 Att. 3 pages 26 thru 30. Question Source: New Question Cognitive Level: High -Application 10 CFR Part 5!; Content: 10 eFR 55.43(b){3)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 97. Unit 1 and Unit 2 are at 100% power with all systems in NSA. A RWDA-L has been prepared for discharging Steam Generator Blowdown Evaporator Test Tank [2SGC-TK23A]. After the RWDA-L is approved by Radiation Protection, the Unit 2 SM or US is then required to review the RWDA-L to confirm the status of various items as part of the approval process. lAW the guidance in 20M-25.4.L, Discharging Steam Generator Blowdown Evaporator Test Tank [2SGC-TK23A(B)]

Contents to Cooling Tower Blowdown, which of the below items is NOT REQUIRED as part of this review/approval?

A. Verify the effective period for the RWDA-L has NOT expired. B. Verify Unit 2 cooling tower blowdown flow is greater than the minimum flow specified on the permit. C. Verify the tank data is correct. D. Verify all hand calculations are correct. Answer: B Explanation/Justification: Incorrect.

lAW 20M-25.4.L Rev. 29 step IV.A.12 page 16 this is a required item. Correct. lAW 20M-25.4.l Rev. 29 step IV.A.12 page 16 this is NOT a required item. Minimum Cooling tower blowdown flow for liquid discharges is based on the combined flow of Unit 1 and Unit 2. With both Units at full power, the RWDA-L. bases the Minimum Cooling tower blowdown flow on this combined flow. The SM/US is required to verify that the combined U1 and U2 cooling tower blowdown flow is greater than that specified on the permit. Unit 2 flow alone will NEVER meet this requirement.

SRO only in that this an SRO task and involves the process for liquid releases. Incorrect.

IAW.20M-25.4.L Rev. 29 step IV.A.12 page 16 this is a required item. Incorrect.

lAW 20M-25.4.L Rev. 29 step IV.A.12 page 16 this is a required item. Sys# System Category KA Statement N/A N/A Generic Ability to approve release permits. KlA# 2.3.6 KIA Importance 3.8 Exam Level SRO References provided to Candidate None Technical

References:

20M-25.4.L Rev. 30 step IV.A.12 pages 16 & 17 Question Source: New Question Cognitive Level: High -Comprehension 10 CFR Part 56, Content: 10 CFR 55.43(b)(4)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam The Plant is operating at 100% power with all systems in NSA. Unit 2 is discharging the contents of the Gaseous Waste Storage tanks lAW 19.4A.B, Unit 2 GW Storage Tk Disch To Unit 1 Atmos Vent Rad Monitor RM-1 GW-108B, Gaseous Waste Gas fails downscale and is declared inoperable. The crew terminates the discharge.

In order to re-start the discharge, what %-ODC-3.03, ODCM: Controls for RETS and REMP Programs actions Will be required? (Refer to attached reference) The system/process flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (or assumed to be at the ODCM design value). At least two independent samples of the tank's content are analyzed and at least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup. Grab samples (or local monitor readings) are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If grab samples are taken, these samples are to be analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Samples are continuously collected with auxiliary sampling equipment as required in ODCM Control 3.11.2.1 ,.Table 4.11-2, or sampled and analyzed once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Answer: B Explanation/Justification: Incorrect.

This is a required action if FR-GW-108 is DOS NOT RM-GW-10SB. (Action 2SA) Correct. lAW ODCM 'h-ODC-3.03 Att.F page 38 and action 27 on page 42. Incorrect.

This is the required action for all continuous releases thru this pathway. (Action 29) Incorrect.

This is required action 32 for continuous releases if the alt channel 109 ch5 is also not available.

Sys # System Category KA Statement N/A N/A Generic Ability to control radiation releases.

KlA# 2.3,11 KIA Importance 4.3 Exam Level SRO References provided to Candidate

'h-ODC-3.03 Technical

References:

ODCM 'h-ODC-3.03 Rev. 11 Att.F pages 38-43 Question Source: New Question Cognitive Level: High -Application 10 CFR Part 55 Content: 10 CFR 55.43(b)(4)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 99. The Plant is operating at 100% power with all systems in NSA. A loss of all 4KV AC power occurs The crew enters ECA-O.O, Loss of All AC Power After 30 minutes, power is restored to one 4KV Emergency bus and the crew has reached the end of procedure ECA-O.O and are in the process of selecting the appropriate recovery procedure.

The following conditions now exist: RCS subcooling is ZERO degrees PRZR level is 2% and dropping HHSI and LHSI flows are ZERO gpm A RED path condition exists for the heat sink status tree Based on these conditions, what procedure transition is required?

Transition to ____________________________

_ A. FR-H.1, Response To Loss Of Secondary Heat Sink B. ECA-0.1, Loss of All AC Power Recovery Without SI Required C. ECA-0.2, Loss of All AC Power Recovery With SI Required D. ES -0.2, Natural Circulation Cooldown 'nswer: C Explanation/Justification: Incorrect.

FRPs usuaUy have higher priority than ECA procedure.

In tis case they do not since I:::CA-0.1 and 0.2 are structured such that they will restart E5F equipment.

Only after completing these steps will FR-H.1 be implemented. Incorrect.

The conditions in the stem indicate the need for 51. Correct. lAW the NOTE prior to step 1 of ECA-O.O FRPs are not to be implemented while in ECA-O.O. A note prior to step 1 of both ECA-0.1 and 0.2 then reminds the operator to complete steps 1-11 before implementing any FRP. Therefore, the appropriate transition is to ECA-0.2 since the conditions in the stem indicate the need for 51. 5RO only since it requires the additional knowledge of the procedure's content and; assessing plant conditions and then selecting a procedure to mitigate, recover, or with which to proceed. Incorrect.

Without any normal 4KV power, the RCP will not be running and this would be a transition if 4KV emergrncy had not been lost. Sys# System Category KA Statement N/A N/A Generic Knowledge of the operational implications of EOP warnings, cautions, and notes. KlA# 2.4.20 KIA Importance 4.3 Exam Level 5RO References provided to Candidate None Technical

References:

ECA-O.O. step 39 & step 1 NOTE Question Source: New Question Cognitive Level: High -

10 CFR Part 55 Content: 10 CFR55.43(b)(5)

Objective:

(SRO Beaver Valley Unit 2 NRC Written Exam 100. Given following plant conditions:

• Unit 2 core off-load is in progress. The Control Room receives a report that cable separation has occurred on the upender containing an irradiated fuel assembly from the vertical position. The RO reports that [2RMF-RQ301A1Bl, "Fuel Building Vent radiation levels are rising and High alarms are validated. The Control Room has received A4-5A, "Radiation Monitoring System Trouble" AND A4-5C, "Radiation Monitoring Level High".

• No other alarms are present. Have entry conditions been met for the SRO to perform AOP 2.49.1, "Irradiated Fuel Damage" actions AND using the Emergency Plan Procedure provided, does an ALERT classification exist for the present plant conditions? (Excluding ED Judgment)

AOP 2.49.1 entry conditions

__ (1) __ been An ALERT classification

__ (2) __ exist for the stated plant (Refer to attached reference)

A (1) have (2) does 8. (1) have NOT (2) does C. (1) have (2) does NOT D. (1) have NOT (2) does NOT Answer: A Explanation/Justification: Correct. lAW AOP 2.49.1 entry conditions and EPP Tab 6.5. SRO only since it requires classification in the EPP. Incorrect.

AOP entry conditions do exist if 2RMF-RQ301A1B are in high alarm. It is correct that an Alert classification exists. Incorrect.

Correct that AOP entry conditions are met. However, an Alert classification exists in Tab 6.5. Incorrect.

AOP entry conditions do exist if 2RMF-RQ301AIB are in high alarm. Entry conditions would not exist if they were only at the alert level. An Alert classification exists in Tab 6.5. Sys# System Category KA Statement N/A N/A Generic Knowledge of the emergency action level thresholds and classifications.

KlA# 2.4.41 KIA Importance 4.6 Exam Level SRO References provided to Candidate EPP IPs Technical

References:

AOP 2.49.1 Rev. 9 Entry conditions and EPP Tab 6.5 Question Source: Bank Question Cognitive Level: High Application 10 CFR Part Content: 10 CFR 55.43(b)(7)

Objective: