ML16138A103

South Texas Project-2016-05 Draft Written Exam

ML16138A103
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Site:	South Texas
Issue date:	05/04/2016
From:	Vincent Gaddy Operations Branch IV
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Download: ML16138A103 (207)

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3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following describes the BACKUP power supply to Instrument Air Compressor #14upon loss of power to LC 1U? A. TSC diesel generator which energizes LC 1U B. TSC diesel generator which energizes MCC 1G5 C. BOP diesel generator which energizes LC 1U D. BOP diesel generator which energizes MCC 1G5 Answer:DBOP diesel generator which energizes MCC 1G5Exam Bank No.:96Last used on an NRC exam:2013RO Sequence Number:1Page 1 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:078 K2.02Tier:2Group/Category:1Knowledge of the bus power supplies to the following:Emergency air compressor.STP Lesson:LOT 202.26Objective Number:25610LIST the systems that interface with the Instrument Air and Service Air systems and STATE the function of each interface.Attached Reference

Reference:

LOT201.35 PowerPoint

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Source:BankDistractor JustificationA:INCORRECT - LC is credible because LC 1U normally energizes MCC 1G5 (which is the supply to IA #14).  TSC DG is credible because it is another non-class DG which (despite its name) supplies power to plant components such as the CVCS PDP, transformer cooling fans, select plant lighting, etc.B:INCORRECT - TSC DG is credible because it is another non-class DG which (despite its name) supplies power to plant components such as the CVCS PDP, transformer cooling fans, select plant lighting, etc.C:INCORRECT - Credible because the BOP DG supplies IA compressor #14, but it does so by energizing MCC 1G5 which is normally powered through LC 1U.D:CORRECT - The BOP DG powers up MCC 1G5 (which supplies IA compressor 14) upon a loss of power to the MCC.Question Level:FQuestion Difficulty2Justification:The applicant must combine the knowledge of the instrument air system and the backup diesel generators to eliminate the incorrect distractors and select the correct answer.RO Importance:3.3NRC Reference Req'd

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Modified fromExam Bank No.:9610CFR

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55.41(b)(4)Page 2 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe power supply for the Turbine Driven Auxiliary Feedwater Pump Steam Inlet Valve, MS-MOV-0143, is __________. A. Non-1E 480 VAC MCC 1J1 B. Non-1E 125 VDC SWBD 1A C. ESF 120 VAC Panel DP-1202 D. ESF 125 VDC bus E1D11 Answer:DESF 125 VDC bus E1D11Exam Bank No.:486Last used on an NRC exam:2007RO Sequence Number:2Page 3 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:061 K2.01Tier:2Group/Category:1Knowledge of bus power supplies to the following:AFW system MOVsSTP Lesson:LOT 202.28Objective Number:92049List the typical loads on the Class 1E 125 VDC System.Attached Reference

Reference:

LOT 202.28, lesson plan page 45

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Source:BankDistractor JustificationA:INCORRECT: Plausible because Main Steam is as a whole is not a safety related system and a majority fo the components have non-class power supplies.B:INCORRECT: Plausible because Main Steam is as a whole is not a safety related system and a majority fo the components have non-class power supplies.C:INCORRECT: Plausible because many components in the AFW system have alternating current power.D:CORRECT: Power supply to AFW steam supply valve is Class 1E 125 VDC.Question Level:FQuestion Difficulty2Justification:Requires the fundamental knowledge that the power supply to turbine driven aux feed pump MOV is different from the motor driven pumps (DC vs AC)RO Importance:3.2NRC Reference Req'd

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Modified fromExam Bank No.:48610CFR

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55.41(b)(8)Page 4 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMIn accordance with 0POP08-FH-0009, Core Refueling, which of the following is an administrative task for a Licensed Control Room Operator during refueling operations? A. Inform the Core Load Supervisor of the next core location to have a fuel assembly loaded. B. Operate the remote television monitoring equipment used to observe refueling activities. C. Monitor the Core Monitoring NI channels during and following insertion of each fuel assembly. D. Inform the Refueling Machine Operator when they can disengage from a newly seated fuel assembly. Answer:CMonitor the Core Monitoring NI channels during and following insertion of each fuel assembly.Exam Bank No.:1101Last used on an NRC exam:2009RO Sequence Number:3Page 5 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.1.40Tier:3Group/Category:1Knowledge of refueling administrative reqirements.STP Lesson:LOT 201.43Objective Number:66407DESCRIBE the procedural requirements of the fuel handling equipment operating procedure(s) to include purpose, scope, precautions and limitationsAttached Reference

Reference:

0POP08-FH-0009, Core Refueling

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Source:BankDistractor JustificationA:INCORRECT: Credible because move sheets are available to and monitored by the RO during fuel movement, but not an assigned responsibility.B:INCORRECT: Credible because the equipment is available to the operator, but not an assigned responsibilityC:CORRECT: per 0POP08-FH-0009, Core RefuelingD:INCORRECT: Credible because the operator must report his observation of the instruments before unlatching can occur, but he does not give permission.Question Level:FQuestion Difficulty2Justification:The student must have knowledge of the administrative responsibilities for Licensed Operators during refueling.RO Importance:2.8NRC Reference Req'd

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Modified fromExam Bank No.:110110CFR

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55.41(b)(10)Page 6 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMTurbine Impulse Pressure is giving a Temperature Reference (Tref) of 590°F to the Rod Control System. What would this reading equate to in percent Reactor Power? A. 98% B. 92% C. 88% D. 82% Answer:B92%Exam Bank No.:2508Last used on an NRC exam:NeverRO Sequence Number:4Page 7 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:045 K4.01Tier:2Group/Category:2Knowledge of MT/G system design feature(s) and/or interlock(s) which provide for the following:Programmed controller for relationship between steam pressure at T/G inlet (impulse, first stage) and plant power level.STP Lesson:LOT 201.18Objective Number:86061DESCRIBE the instrumentation and controls available to monitor and operate the Rod Control System.Attached Reference

Reference:

LOT 201.18 Lesson Plan on Rod Control.

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Source:NewDistractor JustificationA:INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.B:CORRECT: Tref spans from 567 to 592 degrees F with a corresponding percent power of 0% to 100%. This calculates out to 1/4 Degree F/% power. A Tref of 590 degrees F would then calculate to 92% power.C:INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.D:INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.Question Level:HQuestion Difficulty2Justification:The student must have knowledge of the percent power vs Tref and be able to calculate the two for different power levels. NOTE: Turbine Impulse pressure can be read in an actual pressure, however, an ICS computer point calculates the pressure to a Temperature Reference. For operational validity the question uses Tref.RO Importance:2.7NRC Reference Req'd

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Modified fromExam Bank No.:250810CFR

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55.41(b)(7)Page 8 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is cooling down after a SGTR per 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill, with the following conditions: RHR Pump A and C are in service. RCS temperature is 310ºF and slowly lowering. RCS pressure is being maintained between 325 and 400 psig. Subsequently: ESF 4.16KV 'C' loses power due to an overcurrent lockout on the BUS. Which of the following would be the correct action for the crew to perform? A. Suspend 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and perform 0POP09-AN-01M2, C-8, 'RHR PUMP C TRIP', ONLY. B. Suspend 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and perform 0POP09-AN-01M2, C-8, 'RHR PUMP C TRIP', AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus. C. Continue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, 'RHR PUMP C TRIP', ONLY.D. Continue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, 'RHR PUMP C TRIP', AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus. Answer:DContinue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, 'RHR PUMP C TRIP', AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus.Exam Bank No.:2489Last used on an NRC exam:NeverRO Sequence Number:5Page 9 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 025 G2.4.8Tier:1Group/Category:1Loss of RHR System:Knowledge of how abnormal operating procedures are used in conjunction with EOPs.STP Lesson:LOT 504.04Objective Number:92283GIVEN a set of conditions and the occurrence of a Red, Orange, or Yellow path CSF, STATE the action required per 0POP01-ZA-0018, EOP Users Guide.Attached Reference

Reference:

LOT 504.04 Lesson Plan on EOP Introduction and 0POP01-ZA-0018, Emergency Operating Procedure User's Guide (step 4.26.4)

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the student has to remember that Emergency Operating Procedures have precedence over Off-Normal/Annunciator Response Procedures. Also, the student would need to realize that 0POP04-AE-0001 is equally important to 0POP09-AN-1M02 because without addressing the issue ESF DG on Train C would be running without cooling water.B:INCORRECT: Plausible because the student has to remember that Emergency Operating Procedures have precedence over Off-Normal/Annunciator Response Procedures.C:INCORRECT: Plausible because the student would need to realize that 0POP04-AE-0001 is equallyimportant because without addressing the issue ESF DG on Train C would be running without cooling water.D:CORRECT: Per 0POP01-ZA-0018, Emergency Operating Procedure User's Guide, Off-Normal and Annunciator Response procedures can be performed if they do not conflict with the EOP and adequate resources are available. In this case if resources available then both 0POP09-AN-1M02 and 0POP04-AE-0001 should be performed to restart RHR pump B and secure ESF DG on Train C.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given conditions and have knowledge of the rules of usage for off-normal and emergency procedures to determine the correct answer.RO Importance:3.8NRC Reference Req'd

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Modified fromExam Bank No.:248910CFR

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55.41(b)(10)Page 10 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA Reactor Trip and Safety Injection have occurred with the following conditions: SI has been reset. Due to some confusion with the event, the Unit Supervisor has entered 0POP05-EO-ES00, Rediagnosis. Subsequently a Reactor Operator observes the following: RCS Pressure: ..............................1710 psig and slowly lowering RCS Subcooling: ..........................60oF and slowly lowering Pressurizer Level:.........................20% and slowly lowering All SG Pressures: .........................1175 psig and slowly lowering SG A NR Level: ...........................8% and slowly rising SG B NR Level: ...........................10% and slowly rising SG C NR Level: ...........................17% and stable SG D NR Level: ...........................19% and stable Total AFW Flow: .........................400 gpm Containment Pressure:---..-4 psig and slowly rising Containment Radiation Level:......4 R/HR and stable Which procedure should the crew transition to? A. 0POP05-EO-FRZ1, Response to High Containment Pressure B. 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink C. 0POP05-EO-EO20, Faulted Steam Generator Isolation D. 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Answer:D0POP05-EO-EO10, Loss of Reactor or Secondary CoolantExam Bank No.:2493Last used on an NRC exam:NeverRO Sequence Number:6Page 11 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:W/E01 EA2.1Tier:1Group/Category:2Ability to determine and interpret the following as they apply to the (Reactor Trip and Safety Injection Rediagnosis):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.STP Lesson:LOT 504.09Objective Number:81187DISCUSS the indications available to determine plant status during a loss of primary or secondary coolant accident.Attached Reference

Reference:

LOT 504.09 Lesson Plan on 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant and 0POP05-EO-ES00, Rediagnosis.

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Source:ModifiedDistractor JustificationA:INCORRECT: Credible because this transition may be done with a higher containment pressure (9.5 psig).B:INCORRECT: Credible because this transition is required if adverse containment conditions existed or if all SG levels were less than 14%.C:INCORRECT: Credible because this transition is required if a SG level and pressure are lowering in an uncontrolled manner. Pressure is not given but can assume to be controlled if levels are as indicated.D:CORRECT: With PZR level and pressure slowly lowering and given the other parameters and trends, 0POP05-EO-EO10 would be the correct procedure to enter.Question Level:HQuestion Difficulty3Justification:The applicant must analyze the given conditions and apply their knowledge of SI termination and reinitiation requirements and the loss of heat sink and integrity transitions in order to eliminate the incorrect responses and choose the correct response.RO Importance:3.2NRC Reference Req'd

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Modified from52Exam Bank No.:249310CFR

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55.41(b)(10)Page 12 of 150 3/6/2016Print DateSTP LOT-19 NRC RO EXAMA Reactor Trip and Safety Injection have occurred with the following conditions: SI has been reset Operators have just completed step 1 of 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Due to some confusion with the event, the Unit Supervisor also enters 0POP05-EO-ES00, Rediagnosis. Subsequently a Reactor Operator observes the following: RCS Pressure: ..............................1830 psig RCS Subcooling: ..........................60oF Pressurizer Level:.........................20% SG A NR Level: ...........................8% SG B NR Level: ...........................10% SG C NR Level: ...........................17% SG D NR Level: ...........................19% Total AFW Flow: .........................400 gpm Containment Pressure:---..-4 psig Containment Radiation Level:......5 R/HR Which procedure should the crew transition to? A. 0POP05-EO-FRZ1, Response to High Containment Pressure. B. 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink. C. 0POP05-EO-FRI2, Response to Low Pressurizer Level.

D. 0POP05-EO-ES11, SI Termination. Answer:DTransition to 0POP05-EO-ES11, SI TerminationExam Bank No.:52Last used on an NRC exam:2013RO Sequence Number:5Page 9 of 150 3/6/2016Print DateSTP LOT-19 NRC RO EXAMK/A Catalog Number:EPE W/E02 EA2.2Tier:1Group/Category:2Ability to determine and interpret the following as they apply to the (SI Termination): Adherance to appropriate procedures and operation within the the limitations in the facility's license and amendments.STP Lesson:LOT 504.09Objective Number:81187DISCUSS the indications available to determine plant status during a loss of primary or secondary coolant accident.Attached Reference

Reference:

0POP05-EO-EO10, Conditional Information Page

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Source:BankDistractor JustificationA:INCORRECT: Credible because this transition may be done with a higher containment pressure (9.5 psig).B:INCORRECT: Credible because this transition is required if adverse containment conditions existed or if all SG levels were less than 14%.C:INCORRECT: Credible because transition may be done with a lower pressurizer level (17%).D:CORRECT: The given conditions would allow transition to ES11 which would be the expected action for the Crew.Question Level:HQuestion Difficulty3Justification:The applicant must analyze the given conditions and apply their knowledge of SI termination and reinitiation requirements and the loss of heat sink and integrity transitions in order to eliminate the incorrect responses and choose the correct response.RO Importance:3.5NRC Reference Req'd

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Modified fromExam Bank No.:5210CFR

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55.41(b)(10)Page 10 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA Large Break LOCA has occurred with the following plant conditions: Containment pressure is 10.2 psig Containment temperature is 240 F Containment radiation is 3.0E + 03R/HR The Crew has entered 0POP05-EO-FRZ3, Response to High Containment Radiation Level. Which of the following actions is required in accordance with 0POP05-EO-FRZ3, Response to High Containment Radiation Level? A. Verify Containment Phase A Isolation has occurred. B. Ensure all Reactor Containment Fan Coolers running. C. Verify Containment Ventilation Isolation has occurred. D. Ensure all Containment Spray Pumps running. Answer:CVerify Containment Ventilation Isolation has occurred.Exam Bank No.:16Last used on an NRC exam:1999RO Sequence Number:7Page 13 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:W/E16 EA1.3Tier:1Group/Category:2Ability to operate and/or monitor the following as they apply to the High Containment Radiation: Desired operating results during abnormal and emergency situations.STP Lesson:LOT 504.42Objective Number:T50442Without using reference material unless provided, the student will be able to use 0POP05-EO-FRZ3 to correctly respond to a high containment radiation level.Attached Reference

Reference:

0POP05-EO-F005, 0POP05-EO-FRZ3, Step 1

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Source:BankDistractor JustificationA:INCORRECT: Plausible because containment phase A isolation should have occurred but other procedures verify phase A and this is not an action directed by 0POP05-EO-FRZ3.B:INCORRECT: Plausible because it is all RCFCs should be in operation but this action is not directed by 0POP05-EO-FRZ3.C:CORRECT: CVI should be ensured to prevent release.D:INCORRECT: Plausible because it is desirable to have CS Pumps running however in this conditiononly 2 would be running and this action is not directed by 0POP05-EO-FRZ3.Question Level:HQuestion Difficulty3Justification:Student must analyze the given conditions and apply the mitigating strategies for a high radiation condition in containment to the given conditions and determine whether any operating limits have been exceeded for components contained within the distractors.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:1610CFR

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55.41(b)(10)Page 14 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMPer 0POP05-EO-EC11, Loss of Emergency Coolant Recirculation, which of the following is the correct SEQUENCE to stop backflow from the RWST to the Containment Sump? Stop the LHSI, HHSI and CS Pumps, then ....

A. 1. Open the RWST to SI Suction Header Valves 2. Open the SI Pump Mini Flow Valves 3. Close the Containment Sump Suction Valves 4. Start the LHSI, HHSI and CS Pumps as necessary B. 1. Close the Containment Sump Suction Valves 2. Open the RWST to SI Suction Header Valves

3. Open the SI Pump Mini Flow Valves 4. Start the LHSI, HHSI and CS Pumps as necessary C. 1. Open the RWST to SI Suction Header Valves 2. Close the Containment Sump Suction Valves 3. Open the SI Pump Mini Flow Valves

4. Start the LHSI, HHSI and CS Pumps as necessary D. 1. Open the SI Pump Mini Flow Valves

2. Close the Containment Sump Suction Valves 3. Open the RWST to SI Suction Header Valves 4. Start the LHSI, HHSI and CS Pumps as necessary Answer:B1. Close the Containment Sump Suction Valves; 2. Open the RWST to SI Suction Header Valves; 3. Open the SI Pump Mini Flow Valves; 4. Start the LHSI, HHSI and CS Pumps as necessaryExam Bank No.:18Last used on an NRC exam:2010RO Sequence Number:8Page 15 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:W/E11 EA1.2Tier:1Group/Category:1Ability to operate and / or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation): Operating behavior characteristics of the facility.STP Lesson:LOT 504.27Objective Number:82598DESCRIBE the indications and anticipated readings used to determine that there is no backflow from the RWST to the emergency sump.Attached Reference

Reference:

0POP05-EO-EC11 Rev 18, LOT 201.10

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Source:BankDistractor JustificationA:INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.B:CORRECT: The containment sump valves must be closed before the RWST and the SI Pump mini flow valves can be openedC:INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.D:INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the data given and have knowledge of systems interlocks to determine the proper sequence for this valve re-alignment.RO Importance:3.5NRC Reference Req'd

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Modified fromExam Bank No.:1810CFR

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55.41(b)(7)Page 16 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMFollowing a Loss of Offsite Power (LOOP), which ONE of the following statements describes the Pressurizer heater groups that will be available to maintain Pressurizer pressure? A. Backup heater groups A and B ONLY B. Backup heater groups D and E ONLY C. All Backup heater groups EXCEPT control heater group C D. All Backup heater groups Answer:ABackup heater groups A and B ONLYExam Bank No.:27Last used on an NRC exam:2009RO Sequence Number:9Page 17 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:011 K2.02Tier:2Group/Category:2Knowledge of bus power supplies to the following:PZR heatersSTP Lesson:LOT 201.14Objective Number:8860List the power supplies to the pressurizer heaters.Attached Reference

Reference:

LOT201.14 Lesson Plan on PZR Level and Pressure Control

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Source:BankDistractor JustificationA:CORRECT: Backup heater groups A and B are supplied power from Class 1E Load Centers E1A1 and E1C1 respectively. These loads are backed by their respective ESF D/Gs. All other heaters are non-safety related and do NOT have any backup power.B:INCORRECT: Plausible because students have to remember which to PZR heater groups are backed by an ESF D/G.C:INCORRECT: Plausible because of the importance of having PZR heaters available and the student believing that all heaters would have ESF Backup power except for the control group.D:INCORRECT: Plausible because of the importance of having PZR heaters available and the student thinking that all heaters would be backedup by ESF D/Gs.Question Level:FQuestion Difficulty3Justification:The candidate must know which heater groups are safety related (i..e ESF diesel powered).RO Importance:3.1NRC Reference Req'd

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Modified fromExam Bank No.:2710CFR

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55.41(b)(7)Page 18 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMUnit 1 was stable at 100 % power when a piping failure in the Generator Hydrogen Supply Header caused Main Generator hydrogen pressure to lower. After the leak was isolated, the following conditions exist: Generator hydrogen pressure - 45 psig Generator output - 1300 MWe Generator reactive load - 300 MVARs OUT Based on these conditions, and referring to Figure 7.1 (attached), which of the following represents the HIGHEST ALLOWABLE amount the generator load should be reduced to? A. 1225 MWe and 240 MVARs OUT. B. 1225 MWe and 100 MVARs OUT. C. 1175 MWe and 340 MVARs OUT. D. 1175 MWe and 200 MVARs OUT. Answer:D1175 MWe and 200 MVARs OUT.Exam Bank No.:594Last used on an NRC exam:NeverRO Sequence Number:10Page 19 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.1.25Tier:3Group/Category:1Ability to interpret reference materials, such as graphs, curves, tables, etc.STP Lesson:LOT 202.17Objective Number:3872DISCUSS the relationship between Main Generator load and Generator Hydrogen gas pressure. Include in the discussion how generator capacity varies with reduced Hydorgen gas pressure and what gas pressure would require a generator shutdown.Attached Reference

Reference:

LOT 202.17 Lesson Plan on the Main Generaotr and Main Generator Capability Curve- Unit 1 Plant Curve Book Figure 7.1

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Main Generator Capability Curve- Unit 1 Plant Curve Book Figure 7.1Source:BankDistractor JustificationA:INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.B:INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.C:INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.D:CORRECT: 1175 MWE and 200 MVARS OUT would represent the only loading that would meet the restrictions of the Main Generator Capability curve with the conditions given.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the conditions given and be able to use the Main Generator Capability Curve.RO Importance:3.9NRC Reference Req'd

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Modified fromExam Bank No.:59410CFR

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55.41(b)(6)Page 20 of 150

3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is in Mode 6 with the following conditions: Source Range N31 has failed LOW. Audio Count Rate was lost in the Control Room. Which of the following describes the actions necessary to restore the Control Room Audio Count Rate? A. Place the LEVEL TRIP switch for SR N31 to the 'BYPASS' position ONLY. B. Place the CHANNEL SELECTOR switch on the front of the Audio Count Rate drawer to the 'SR N32' position. C. Place the LEVEL TRIP switch for SR N31 to the 'BYPASS' position AND push the 'RESET' button on the Scaler Timer. D. Place the AMPLIFIER SELECT switch on the rear of the Audio Count Rate drawer assembly to the 'A2' position. Answer:BPlace the CHANNEL SELECTOR switch on the front of the audio count rate drawer to the SR N32 position.Exam Bank No.:715Last used on an NRC exam:NeverRO Sequence Number:11Page 21 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 032 AK2.01Tier:1Group/Category:2Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following: Power supplies, including proper switch positions.STP Lesson:LOT 201.16Objective Number:4886DESCRIBE all the interlocks, trips, permissives, alarms and/or indications associated with the Nuclear Instrument System, including setoints and coincidences.Attached Reference

Reference:

LOT 201.16 Lesson Plan on Excore Nis

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Source:BankDistractor JustificationA:INCORRECT: This distractor is credible because placing the Level Trip switch to Bypass is an action to take for a failed Source Range Channel but it will not restore the Audio Count Rate.B:CORRECT: The Audio Count Rate is feed from one or the other Source Range detectors and is selected from the Channel Selector switch on the front of the Audio Count Rate drawer.C:INCORRECT: This distractor is credible because placing the Level Trip switch to Bypass is an action to take for a failed Source Range Channel but it will not restore the Audio Count Rate even if the Scaler Timer is reset.D:INCORRECT: This distractor is credible because the Amplifier Select switch is part of the Audio Count Rate circuit but it only determines how the audio Count Rate speakers in the Control Room and the Containment are aligned.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition and determine the action necessary to restore the Control Room Audio Count Rate.RO Importance:2.7NRC Reference Req'd

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Modified fromExam Bank No.:71510CFR

Reference:

55.41(b)(7)Page 22 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power when the following events occur: A Control Rod System malfunction causes Control Bank D rods to rapidly insert. The RO places Rod Control in MANUAL and rod motion STOPS. Subsequently the following alarms come in: BANK INSERT LO alarm is LIT. TREF/AUCT TAVG DEV alarm is LIT. The Unit Supervisor has directed that NO further rod motion be attempted until the cause of the rod control problem has been determined. Which of the following actions, ALONE, will cause BOTH of the above listed annunciators to CLEAR? A. Lower Turbine Load B. Raise Turbine Load C. Borate the RCS D. Dilute the RCS Answer:ALower Turbine LoadExam Bank No.:773Last used on an NRC exam:NeverRO Sequence Number:12Page 23 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:001 K1.04Tier:2Group/Category:2Knowledge of the physical connections and/or cause-efect relationships between the CRDS and the following systems:

RCSSTP Lesson:LOT 201.18Objective Number:86061Describe the instrumentation and controls available to monitor and operate the Rod Control System.Attached Reference

Reference:

LOT 201.18 Lesson Plan on Rod Control

Attachment:

Source:BankDistractor JustificationA:CORRECT: Lowering Turbine Load will raise RCS temperature which is lower than Tref and will also lower RX power which will lower the Rod Insertion Limit.B:INCORRECT: Plausible because the student must remember how Rod Motion affects the RCS. In this case raising Turbine Load would not help clear either alarm.C:INCORRECT: Plausible because borating the RCS would lower temperature which would lower power some and lower the Rod Insertion limit but lowering temperature would further cause a deviation from Tref.D:INCORRECT: Plausible because diluting the RCS would raise temperature of the RCS to help clear the Tavg/Tref deviation alarm but it would also raise power some (if turbine is in Imp Out) or none at all (if turbine is in Imp In) which would not help the Rod Insertion Limit.Question Level:HQuestion Difficulty3Justification:The student must be able to anlayze the given conditions and have fundamental knowledge of how the RCS is affected by Control Rod movement.RO Importance:3.2NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:77310CFR

Reference:

55.41(b)(7)Page 24 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following is the indication used in 0POP05-EO-EC12, LOCA Outside Containment, that ensures actions taken to isolate the leak have been successful and that the procedure can be exited? A. FHB Area Radiation alarms clearing. B. RCS pressure rising. C. FHB SI/CS Pump Sump level alarms clearing. D. RCS Hot Leg temperatures lowering. Answer:BRCS pressure rising.Exam Bank No.:816Last used on an NRC exam:NeverRO Sequence Number:13Page 25 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:W/E04 EK1.2Tier:1Group/Category:1LOCA Outside ContainmentKnowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment):Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).STP Lesson:LOT 504.46Objective Number:82657From Memory STATE/IDENTIFY indications and trends used to determine that the break is isolated in accordance with POP05-EO-EC12.Attached Reference

Reference:

0POP05-EO-EC12, Rev. 9, Steps 4j. and 6a.

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because the presence of FHB Area Radiation alarms are used to indicate that a LOCA has ocurred in the FHB but they are not used in reverse to allow exiting the procedure.B:CORRECT: Once the RCS leak is isolated, RCS pressure should begin to rise and the procedure can be exited.C:INCORRECT: Plausible because the presence of FHB Sump alarms are used to indicate that a LOCA has ocurred in the FHB but they are not used in reverse to allow exiting the procedure.D:INCORRECT: Plausible because Hot leg temperatures lowering would be desirable and indicative of restoring RCS conditions but it is not an indication used in 0POP05-EO-EC12 that the LOCA outside containment has been isolated.Question Level:FQuestion Difficulty3Justification:Student must have fundamental knowledge of the steps in 0POP05-EO-EC12, LOCA Outside Containment.RO Importance:3.5NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:81610CFR

Reference:

55.41(b)(10)Page 26 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is in Mode 6 with the following conditions: A spent fuel assembly is being moved from the reactor to the Upender. The spent fuel assembly is dropped to the bottom of the Transfer Canal. Gas bubbles are observed rising from the damaged spent fuel assembly to the surface of the Transfer Canal. Which of the following correctly describes the primary radiation hazard to personnel in the immediate vicinity AND the related source products released from the damaged spent fuel assembly? A. Alpha radiation from activated hydrogen gas. B. Gamma radiation from fission product gases Xenon and Krypton. C. Alpha radiation from Reactor Coolant particulate fission products. D. Gamma radiation from particulate fission products and corrosion products. Answer:BGamma radiation from fission product gases Xenon and Krypton.Exam Bank No.:826Last used on an NRC exam:2011RO Sequence Number:14Page 27 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.3.14Tier:3Group/Category:3Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.STP Lesson:LOT 204.01Objective Number:20401Given plant or system conditions, predict the response of the plant and/or systemsAttached Reference

Reference:

UFSAR 15.7.4.2.2

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Hydrogen gas within a fuel rod acts to counter balance external pressure on the rod when the fuel is in the reactor at operating pressure. 'Activated Hydrogen' is Tritium which decays by beta emmision. Beta radiation is a relatively strong ionizer and therefore only travels a short distance. It can be easily sheilded and so will not present a significant radiation hazard unless it's ingested.B:CORRECT: Spent fuel assemblies have some amount of fission gases within the fuel rods depending on the degree of use of the fuel assembly. Fission yeilds for Xenon and Krypton are higher than most other elements therefore these gases will be present in larger amounts than other gases. Both of these gases decay by gamma emmision thereby creating the radiation hazard if released from the clad. Because they are gases, they will rise to the surface of the Spent Fuel Pool where they will enter the surrounding atmosphere and create a radiation hazard due to their decay.C:INCORRECT: Fission products predominantly undergo beta, gamma decay. Only the heavier isotopes in the fuel undergo Alpha decay and these are typically not fission products but fuel isotopes or other heavy isotopes created as a result of neutron activation.D:INCORRECT: Gamma radiation from particulate fission products and corrosion products can be a significant radiation source, however since the damage to the spent fuel assembly occurred at the bottom of the Transfer Canal, these products will remain in the water and not be released to the surrounding environment as gases will be.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of the potential radiation hazards associated with spent fuel assemblies.RO Importance:3.4NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:82610CFR

Reference:

55.41(b)(12)Page 28 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following groups includes ONLY events which would cause Area Radiation Monitors to alarm? A. Steam Generator Tube Rupture, Feed water line leak inside RCB, new fuel handling accident in FHB, high radiation at solid waste processing. B. LOCA inside RCB, rupture of charcoal beds (GWPS), spent fuel handling accident in FHB, RCS leakage at incore instrumentation seal table. C. LOCA inside RCB, Main Steamline break, RCS to CCWS leak, new fuel handling accident in FHB. D. Steam Generator Tube Rupture, Main Steamline break, high radiation at solid waste processing, high radiation in the primary sample room or post accident sample room. Answer:BLOCA inside RCB, rupture of charcoal beds (GWPS), spent fuel handling accident in FHB, RCS leakage at incore instrumentation seal table.Exam Bank No.:1023Last used on an NRC exam:2001RO Sequence Number:15Page 29 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 061 G2.4.31Tier:1Group/Category:2ARM System Alarms:Knowledge of annunciator alarms, indications, or response procedures.STP Lesson:LOT 202.42Objective Number:92125PREDICT the probable ARMS alarm(s) that would be energized under the following conditions:A. Loss-of-Coolant Accident (LOCA)B. Steam Generator Tube Rupture (SGTR)C. Main Steamline Break (MSLB)

D. Reactor Coolant System (RCS) to Component Cooling Water System (CCWS) leak E. Gas Storage Tank (GST) Rupture (RHDS) or rupture of the charcoal beds (GWPS)

F. Fuel Handling Accident in the Fuel Building G. High radiation in the drumming station (SWPS)H. RCS leakage at the incore instrumentation seal tableI. High radiation in the Primary Sample Room or Post Accident Sampling RoomAttached Reference

Reference:

LOT202.42, section 3.2

Attachment:

Source:BankDistractor JustificationA:INCORRECT: The SGTR event cannot be effectively detected by an ARM because the activity is released to secondary process systems instead of the atmosphere within an enclosed space. A feed water line break does not, by itself, cause a release of radioactivity, thus cannot be detected by an ARM. A fuel handling accident involving new fuel would not cause and ARM alarm.B:CORRECT: All of these events have the potential to release significant radioactivity to an enclosed area resulting in the radiation levels in those areas rising and thus detectable by an area rad monitorC:INCORRECT: A main steamline break does not, by itself, cause a release of radioactivity, thus cannot be detected by an ARM. A fuel handling accident involving new fuel would not cause and ARM alarm.D:INCORRECT: The SGTR event cannot be effectively detected by an ARM because the activity is released to secondary process systems instead of the the atmosphere within an enclosed space. A main steamline break does not, by itself, cause a release of radioactivity, thus cannot be detected by an ARM.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze each event/accident the Area Rad Monitors would be effective in detecting.RO Importance:4.2NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:102310CFR

Reference:

55.41(b)(6)Page 30 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit was in Mode 1 when a loss of offsite power occurred with the following condition: ESF DG #11 failed to start. Before operator actions are taken, 125 VDC Bus E1A11 will- A. continue to supply its loads from its Battery Bank for a minimum of 12 hrs. B. continue to supply its loads automatically from its Standby Battery Charger. C. not be supplied by a Battery Charger until power is restored to Train 'A' 4160 v Bus. D. not be supplied by a Battery Charger until power is restored to Train 'A' 4160 v Bus AND a Battery Charger is manually realigned. Answer:Cnot be supplied by a Battery Charger until power is restored to Train 'A' 4160 v Bus.Exam Bank No.:1053Last used on an NRC exam:2009RO Sequence Number:16Page 31 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 058 AK1.01Tier:1Group/Category:1Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentationSTP Lesson:LOT 201.37Objective Number:63901GIVEN a loss of power, PREDICT the operation of the class 1E 125 VDC Electrical Distribution System to include automatic actions and interlocks.Attached Reference

Reference:

POP02-EE-0001, LOT 201.37, Class 1E 125 VDC System

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because the batteries would continue to supply the bus but are rated for 2 hrs. minimum, not 12.B:INCORRECT: Plausible because there is a second charger but it is also supplied via 4KV bus "A" and must be placed in service manually when it is used.C:CORRECT: The battery charger that was in service before the LOOP will remain aligned and be returned to service when power is restored to the Train A 4160 V Bus.D:INCORRECT: Plausible if the student believes a charger that was previously in service would need to be realigned to the DC Bus after power is restored.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition to determine how the Vital DC distribution system will react.RO Importance:2.8NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:105310CFR

Reference:

55.41(b)(8)Page 32 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following conditions concerning the Personnel Air Lock would cause a loss of CONTAINMENT INTEGRITY? A. The outer and inner doors are opened simultaneously for a normal transit entry into containment while in MODE 4. B. One air lock door fails acceptance test criteria while the plant is in MODE 6. C. Welding cables are laid through both airlock doors while the plant is in MODE 5. D. The outer door is opened for a normal transit entry into containment while in MODE 3. Answer:AThe outer and inner doors are opened simultaneously for a normal transit entry into containment while in MODE 4.Exam Bank No.:1189Last used on an NRC exam:2005RO Sequence Number:17Page 33 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:103 K3.02Tier:2Group/Category:1Knowledge of the effect that a loss or malfunction of the containment system will have on the following:Loss of containment integrity under normal operationsSTP Lesson:LOT 503.01Objective Number:92101From memory, DEFINE terms used in the Technical Specifications and the Technical Requirements Manual (TRM).Attached Reference

Reference:

LOT 503.01 Lesson Plan for Tech Specs and TRM

Attachment:

Source:BankDistractor JustificationA:CORRECT: CONTAINMENT INTEGRITY is required in Modes 1 to 4 and if both Air Lock doors are open at the same time then CONTAINMENT INTEGRITY would be affected.B:INCORRECT: Plausible because a failed surveillance would constitute a loss of CONTAINMENT INTEGRITY if the plant was in MODE 1 to 4.C:INCORRECT: Plausible because this would constitute a loss of CONTAINMENT CLOSURE if in Mode 6 but not CONTAINMENT INTEGRITY.D:INCORRECT: Plausible because it would be reasonalble to believe that with just one door open that CONTAINMENT INTEGRITY could be affected while in the higher Mode.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of the definition of CONTAINMENT INTEGRITY and what modes of operation it applies.RO Importance:3.8NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:118910CFR

Reference:

55.41(b)(7)Page 34 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA Reactor Trip and Safety Injection have occurred from 100% power with the following conditions: SG 'A' level lowered to 22% but now is slowly rising due to a SGTR. SG 'B', 'C' & 'D' levels lowered to 10% and are now slowly rising due to AFW flow. The Secondary Operator is directed to CLOSE 'A' SG AFW OCIV but the valve will not stay closed. Which of the following signal(s) need to be reset for 'A' SG AFW OCIV to operate properly? 1. Phase 'A' Isolation 2. SG Lo-Lo Level Actuation 3. Safety Injection A. 1 ONLY B. 2 and 3 C. 3 ONLY D. 1 and 3 Answer:B2 and 3Exam Bank No.:1355Last used on an NRC exam:NeverRO Sequence Number:18Page 35 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:013 K1.07Tier:2Group/Category:1Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems:

AFW SystemSTP Lesson:LOT 202.28Objective Number:43805DESCRIBE the AFW system controls and indications the the MCR.Attached Reference

Reference:

Lesson Plan LOT 202.28 Auxiliary Feedwater

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because many containment isolation valves are controled by Phase A.B:CORRECT: The condition states that an SI has actuated and a SG LO-LO level actuatiion would have also occurred on a trip from 100% power. It takes a reset of Safety Injection and SG LO-LO Level Actuation to be able to operate the AFW OCIV.C:INCORRECT: Plausible because the condition states that SG A level did not go below the SG LO-LO actuation setpoint of 20% but the other SGs would have so it takes a reset of both Safety Injection and SG LO-LO Level Actuation.D:INCORRECT: Plausible because many containment isolation valves are controled by Phase A and the condition states that a Safety Injection has occurred.Question Level:HQuestion Difficulty3Justification:The student must analyze the given conditions to determine the correct ESFAS signal to reset.RO Importance:4.1NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:135510CFR

Reference:

55.41(b)()Page 36 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAM(1) Which one of the following correctly describes the Fire Protection System used in the EAB 35' Relay Room? AND (2) Why is this type of fire protection equipment used in the EAB 35' Relay Room? A. (1) Dry Pipe Deluge System (2) Because use of a gas (e.g. Halon or CO2) could render the Relay Room and adjoining Control Room uninhabitable. B. (1) Halon System (2) Because a water-deluge system could cause electrical faults that may result in additional equipment malfunctions. C. (1) Wet Pipe Deluge System (2) Because use of a gas (e.g. Halon or CO2) could render the Relay Room and adjoining Control Room uninhabitable.

D. (1) Carbon Dioxide (CO2) System (2) Because a water-deluge system could cause electrical faults that may result in additional equipment malfunctions. Answer:B(1) Halon System(2) Because a water-deluge system could cause electrical faults that may result in additional equipment malfunctions.Exam Bank No.:1567Last used on an NRC exam:NeverRO Sequence Number:19Page 37 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:086 K5.03Tier:2Group/Category:2Knowledge of the operational implications of the following as they apply to the Fire Protection System: Effect of water spray on electrical components.STP Lesson:LOT 201.29Objective Number:14206Given a Class C fire situation, select the best method for fire suppression.Attached Reference

Reference:

LOT201.29 Fire Protection Rev 6

Attachment:

Source:BankDistractor JustificationA:INCORRECT: the Dry Pipe Deluge system is a water system and therefore should not be used on an electrical fire.B:CORRECT: Early in a fire situation, the switchgear may not have been de-energized. If water is used during this time, additional electrical problems could result.C:INCORRECT: the Wet Pipe Deluge system is a water system and therefore should not be used on an electrical fire.D:INCORRECT: CO2 could be used to fight an electrical fire, but in this instance, it is not. The Switchgear rooms have Halon protection instead.Question Level:FQuestion Difficulty2Justification:Candidate must have knowledge of the hazards of using water on energized electrical systems during fire fighting activities.RO Importance:3.1NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:156710CFR

Reference:

55.41(b)(6)Page 38 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 45% Reactor Power when a total loss of Main Feedwater occurs with the following conditions: The Reactor does NOT automatically trip. The Reactor can NOT be tripped manually. Which of the following correctly describes other automatic action(s) provided for this condition? A. 2/4 SG NR levels in one (1) SG lowering to <15% will actuate Main Turbine relays that will automatically trip the Main Turbine. This condition will also actuate Steam Dumps, SG PORVs, and Pressurizer PORVs to remove heat from the RCS. B. SG NR level in three (3) SGs lowering to <15% will actuate Main Turbine relays that will automatically trip the Main Turbine. This condition will also actuate Auxiliary Feedwater to ensure a sufficient secondary heat sink to remove heat from the RCS. C. Low feedwater flow in three (3) of the feedwater lines to the SGs will separately actuate relays to automatically trip the Main Turbine. This condition will also actuate Auxiliary Feedwater to ensure a sufficient secondary heat sink to remove heat from the RCS. D. SG NR low level in three (3) SGs plus low feedwater flow in three (3) feedwater lines will actuate relays that will actuate Auxiliary Feedwater. This will allow Reactor Power to be lowered to 3% to 4% until Main Feedwater can be restored. Answer:BSG NR level in three (3) SGs lowering to <15% will actuate Main Turbine relays that will automatically trip the Main Turbine. This condition will also actuate Auxiliary Feedwater to ensure a sufficient secondary heat sink to remove heat from the RCS.Exam Bank No.:1579Last used on an NRC exam:NeverRO Sequence Number:20Page 39 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 029 EK2.06Tier:1Group/Category:1Knowledge of the interrelations between ATWS and the following: Breakers, Relays, and DisconnectsSTP Lesson:LOT 201.40Objective Number:91840STATE how the control room operator is aware of AMSAC condition.Attached Reference

Reference:

LOT 201.40 Rev 8

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because the Steam Dumps, SG PORVs, and Pressurizer PORVs may open to remove heat but they are not actuated by AMSAC. Also the SG level condition given is a RXTrip signal and NOT an AMSAC signal.B:CORRECT: This question refers to an ATWS condition where AMSAC is actuated. AMSAC is actuated when greater than 30% Turbine impulse pressure (2/2) and when SG NR levels (3/4) <15% for >25 sec. AMSAC automatically trips the Main Turbine and actuates AFW (also, secures SG blowbown and samples) to minimize secondary inventory loss and ensure a secondary heat sink is maintained.C:INCORRECT: Plausible because low feedwater flow could be indicative of a loss of Main Feedwaterbut it is NOT the signal to actuate AMSAC.D:INCORRECT: Plausible because low feedwater flow could be indicative of a loss of Main Feedwaterbut it is NOT the signal to actuate AMSAC. Also, with an AMSAC condition it would not be desirable to maintain the RX at power.Question Level:HQuestion Difficulty3Justification:The student must analyze the given condition to determine the event and have fundamental knowledge of the basis for AMSAC and how the system works.RO Importance:2.9NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:157910CFR

Reference:

55.41(b)(7)Page 40 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is in Mode 5 with the following conditions: Train 'A' CCW and ECW Pumps are in service (Train Selector Switch in RUN). Train 'B' CCW and ECW Pumps are secured (Train Selector Switch in OFF). Train 'C' CCW and ECW Pumps are secured (Train Selector Switch in STANDBY). Subsequently the following alarms are received in the Control Room and no automatic actions occurred: ECW PUMP 1A TRIP ECW PUMP 1A DISCH PRESS LO (1) Which of the following correctly describes the required action(s)? AND (2) What is the reason why the action(s) is(are) necessary? A. (1) Start Train 'C' ECW Pump ONLY (2) It should have started when the breaker for ECW Pump 1A tripped. B. (1) Start Train 'C' CCW Pump and ECW Pump (2) They should have started when the breaker for ECW Pump 1A tripped. C. (1) Start Train 'C' ECW Pump ONLY (2) It should have started on low ECW pump discharge pressure. D. (1) Start Train 'C' CCW Pump and ECW Pump (2) They should have started on low ECW pump discharge pressure. Answer:D(1) Start Train 'C' CCW Pump and ECW Pump(2) They should have started on low ECW pump discharge pressure.Exam Bank No.:1584Last used on an NRC exam:NeverRO Sequence Number:21Page 41 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:008 G2.4.50Tier:2Group/Category:1Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.STP Lesson:LOT 201.12Objective Number:5213GIVEN a plant or system condition, PREDICT the operation of the Component Cooling Water System.Attached Reference

Reference:

LOT 201.12 Lesson on CCW and 0POP09-AN-02M3

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because it would be reasonable to believe that only Train C ECW pump would need to be started to replace the tripped Train A ECW Pump. Also, ECW Pump breaker position does not feed the auto pump start for another train like some other sets of pumps have.B:INCORRECT: Plausible because the ECW Pump breaker position does not feed the auto pump start for another train like some other sets of pumps have.C:INCORRECT: Plausible because it would be reasonable to believe that only Train C ECW pump would need to be started to replace the tripped Train A ECW Pump.D:CORRECT: With no ECW Pumps running low discharge pressure for the ECW Pumps will send a signal to the standby Train of CCW and ECW to start.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition to determine the correct system response.RO Importance:4.2NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:158410CFR

Reference:

55.41(b)(7)Page 42 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the below methods of forming a Pressurizer (PZR) Steam Bubble would have the greatest likelihood of causing an observable Pressurizer Relief Tank (PRT) level change? Forming a PZR Steam Bubble with the RCS in a -. A. solid plant condition due to the potential of lifting a Pressurizer PORV. B. vacuum-filled condition due to the potential of leakage through the Reactor Vessel Head Vent Valves. C. solid plant condition due to the potential of lifting a Pressurizer Safety Valve. D. vacuum-filled condition due to the potential of leakage through the Pressurizer PORV or Safety Valves. Answer:Asolid plant condition due to the potential of lifting a Pressurizer PORV.Exam Bank No.:1689Last used on an NRC exam:NeverRO Sequence Number:22Page 43 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:007 K5.02Tier:2Group/Category:1Knowledge of the operational implications of the following concepts as they apply to the PRTS:Method of forming a steam bubble in the PressurizerSTP Lesson:LOT 201.04Objective Number:91039DESCRIBE the procedure for formation of a pressurizer bubble.Attached Reference

Reference:

0POP03-RC-0100, RCS Vacuum Fill; 0POP03-ZG-0001, Plant Heatup, LOT 201.04 Lesson Plan on PRZ, PRT and RCDT

Attachment:

Source:BankDistractor JustificationA:CORRECT: The greatest likelihood of causing a PRT level change would be from the PZR PORV during solid plant conditions due to the lift setting for the PZR PORV being lower due to COMS.B:INCORRECT: This distractor is credible because the head vent valves could leak by under a vacuum but the likelihood is small. In addition, the flow would have to be from the PRT, which is lower in the RCB, to the Rx Vessel. It may cause a slight change in PRT level but not an observable change.C:INCORRECT: This distractor is credible because PZR Safety Valves could lift, however, during solid plant conditions, the Pressurizer PORVs have a lower setpoint through the COMS control, thusthey will lift at a significantly lower pressure. The Pressurizer Safety valves will lift at much higher pressure so the potential for them lifting is not as great as that for the PORVs.D:INCORRECT: This distractor is credible because the PZR PORV and Safety Valves could leak by under a vacuum but the likelihood is small. In addition, the flow would have to be from the PRT, which is low in the RCB, to the top of the PZR. It may cause a slight change in PRT level but not an observable change.Question Level:HQuestion Difficulty3Justification:The student must be able to evaluate the different senarios given for forming a PZR Bubble and determine which method and case given would be the most likely to cause an observable level change in the PRT.RO Importance:3.1NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:168910CFR

Reference:

55.41(b)(5)Page 44 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMUnit 1 is shutdown in Mode 3 with the following conditions: RCS Pressure = 990 psig Containment Pressure = 0.1 psig SG pressures =450 psig Subsequently a LOCA inside Containment occurs and the following conditions now exist: RCS pressure = 800 psig Containment pressure = 2.5 psig SG pressures = 425 psig (1) Which of the following correctly describes the response of the ESF actuation logic? AND (2) What is the appropriate procedure actions the crew will implement? A. (1) No ESF actuation signal would occur. (2) Manually initiate Safety Injection and perform the actions for a Reactor Trip/Safety Injection. B. (1) No ESF actuation signal would occur. (2) Perform the actions for a Shutdown LOCA. C. (1) A High Containment Pressure SI will occur. (2) Perform the actions for a Shutdown LOCA. D. (1) A High Containment Pressure SI will occur. (2) Perform the actions for a Reactor Trip/Safety Injection. Answer:B (1) No ESF actuation signal would occur.(2) Perform the actions for a Shutdown LOCA.Exam Bank No.:1710Last used on an NRC exam:NeverRO Sequence Number:23Page 45 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:013 A2.03Tier:2Group/Category:1Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:Rapid depressurization.STP Lesson:LOT 201.20Objective Number:507227Given a description of plant conditions, ANALYZE the conditions and PREDICT how the Solid State Protection System will respond.Attached Reference

Reference:

0POP04-RC-0006 Rev 16, LOT 201.20, SSPS

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because if RCS pressure was above 1000 psig then the actions for a RX Trip/SI would have been appropriate.B:CORRECT: Under the given plant conditions, no ESF signal would be generated. Because RCS pressure is below 1000 psig, the EOP's do not apply. The conditions given meet the entry conditions for a Shutdown LOCA.C:INCORRECT: Plausible because a High Containment pressure would still generate an SI but the setpoint is greater than or equal to 3.0 psig in containment.D:INCORRECT: Plausible because a High Containment pressure would still generate an SI but the setpoint is greater than or equal to 3.0 psig in containment. Also, the conditions given meet the entryconditions for a Shutdown LOCA.Question Level:HQuestion Difficulty3Justification:Candidate must be able to analyze the given conditons and determin whether an ESF actuation has occurred based on the given information. Additionally, he/she must be able to determine the appropriate procedure actions based on the given conditions.RO Importance:4.4NRC Reference Req'd

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Modified fromExam Bank No.:171010CFR

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55.41(b)(5)Page 46 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is operating at 100% power with the following conditions: ECW/CCW Train Mode Selector Switches are aligned as follows: o 'A' in RUN; 'B' in STBY; 'C' in OFF All ECW Pumps are running Subsequently an electrical failure causes a loss of power to ESF 4.16 KV Bus Train C. ECW Pump 1C will be stripped and-.. A. must be started manually. Once started, it will automatically supply cooling water flow to #13 ESF D/G and Train 'C' CCW HX ONLY. B. then sequenced on, automatically supplying cooling water flow to #13 ESF D/G and Train 'C' CCW HX ONLY. C. must be started manually. Once started, it will automatically supply cooling water flow to #13 ESF D/G, Train 'C' CCW Hx, Train 'C' Essential Chiller and Train 'C' CCW Pump Supplementary Cooler. D. then sequenced on, automatically supplying cooling water flow to #13 ESF D/G, Train 'C' CCW Hx, Train 'C' Essential Chiller, and Train 'C' CCW Pump Supplementary Cooler. Answer:Dsequenced on, automatically supplying cooling water flow to #13 ESF D/G, Train 'C' CCW Hx, Train 'C' Essential Chiller, and Train 'C' CCW Pump Supplementary Cooler.Exam Bank No.:1755Last used on an NRC exam:2009RO Sequence Number:24Page 47 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:076 A3.02Tier:2Group/Category:1Ability to monitor automatic operation of the SWS, including: Emergency heat loads.STP Lesson:LOT 201.13Objective Number:91201GIVEN a plant or system condition, PREDICT the operation of the Essential Cooling Water System.Attached Reference

Reference:

LOT201.13 ECW System, LOT201.41 ESF Sequencers

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Source:BankDistractor JustificationA:INCORRECT: Plausible because the student must remember that the ECW/CCW Mode Selector Switch does not affect ECW Pump starts from the sequencer. ECW Pump 1C will be automatically sequenced on. Manual start of the pump is not required. When ECW Pump 1C is running it supplys cooling water flow to #13 ESF D/G, Train C CCW Hx, Train C Essential Chiller, and Train C CCW Supplemnetary Cooler.B:INCORRECT: Plausible because the student must remember which cooling loads are supplied from the ECW Sytem. ECW Pump 1C is automatically sequenced on, however when it is running it will supply cooling water flow to #13 ESF D/G, Train C CCW Hx, Train C Essential Chiller, and Train C CCW Supplemnetary Cooler.C:INCORRECT: Plausible because the student must remember that the ECW/CCW Mode Selector Switch does not affect ECW Pump starts from the sequencer. ECW Pump 1C will be automatically sequenced on. Manual start of the pump is not required.D:CORRECT: ECW Pump 1C will be automatically sequenced on. ESF D/G #13, Train C CCW HX, Train C Essential Chiller, and Train C CCW Pump Supplementary Cooler are all the loads that will be supplied. These loads are always alligned with manual valves to allow cooling water flow. Therefore when the ECW Pump is running cooling water flow is automatically supplied to the loads.Question Level:HQuestion Difficulty3Justification:The student must analyze the given conditions and have an understanding of how a loss of the standby bus affects ECW operation along with a knowledge of the loads supplied by the ECW trains and that all loads are supplied when the ECW Pump in that respective train is running.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:175510CFR

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55.41(b)(7)Page 48 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit 1 is at 100% power with the following conditions: A compressor malfunction has resulted in one Starting Air Receiver on ESF DG #12 to completely depressurize. The second Starting Air Receiver is unaffected and at normal operating pressure. Subsequently: A Unit 1 Standby Transformer lockout occurs. Which of the following correctly describes the effect of the depressurized air receiver on this event? ESF DG #12 will... A. NOT receive a start signal, but IS capable of starting if needed. B. NOT receive a start signal and is NOT capable of starting. C. receive a start signal and WILL start and run. D. receive a start signal, but is NOT capable of starting. Answer:Creceive a start signal and WILL start and run.Exam Bank No.:1833Last used on an NRC exam:2009RO Sequence Number:25Page 49 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:064 K6.07Tier:2Group/Category:1Knowledge of the effect of a loss or malfunction of the following will have on the ED/G System: Air ReceiversSTP Lesson:LOT 201.39Objective Number:98476Given a plant condition and/or various diesel modes of operation, PREDICT the response of the emergency diesels.Attached Reference

Reference:

LOT201.39, ESF Diesel Generator, PowerPoint presentation.

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Plausible because in a normal Unit 1 lineup if the affected ESF D/G was 11 or 13 then they would NOT receive a start signal. With ONLY one receiver out of service the ESF DG is capable of starting.B:INCORRECT: Plausible because in a normal Unit 1 lineup if the affected ESF D/G was 11 or 13 then they would NOT receive a start signal. With ONLY one receiver out of service the ESF DG is capable of starting.C:CORRECT: In a normal electrical lineup 13.8 KV Standby BUS 1G feeds ESF 4.16 KV BUS Train B. If the Unit 1 Standby Transformer is lost then ESF D/G #12 will receive a start signal and start as long as at least one Starting Air Receiver is available.D:INCORRECT: Plausible if the student believes that both Starting Air Recivers are needed to start the ESF D/G.Question Level:HQuestion Difficulty3Justification:Candidate must analyze the effect of the loss on the transformer on the diesel and then determine the effect of the depressurized receiver on the start capability.RO Importance:2.7NRC Reference Req'd

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Modified fromExam Bank No.:183310CFR

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55.41(b)(7)Page 50 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMPer the basis for 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink, why are the Reactor Coolant Pumps secured? Prevent - A. uncovering the Reactor Core B. seal damage to the RCPs C. heat input from the RCPs D. over pressurizing the Reactor Coolant System Answer:Cheat input from the RCPsExam Bank No.:2502Last used on an NRC exam:NeverRO Sequence Number:26Page 51 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:W/E05 EK3.1Tier:1Group/Category:1Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink)Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.STP Lesson:LOT 504.33Objective Number:83013GIVEN a step, note or caution from 0POP05-EO-FRH1, STATE its basis.Attached Reference

Reference:

LOT 504.33 Lesson Plan on 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink

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Source:NewDistractor JustificationA:INCORRECT: Plausible because uncovering the core is the basis for a step in 0POP05-EO-FRH1. Performing the RCS Feed and Bleed.B:INCORRECT: Plausible because preventing seal damage to the RCPs the basis for a steps in other EOPs that have the operators trip Reactor Coolant Pumps.C:CORRECT: Securing the RCPs while performing 0POP05-EO-FRH1 removes the heat input form the RCPs and lengthens the time to dry out the SGs when feedwater flow is unavailable.D:INCORRECT: Plausible because over pressurizing the RCS would be a concern while performing 0POP05-EO-FRH1.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of the reasons for performing emergency actions.RO Importance:3.4NRC Reference Req'd

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Modified fromExam Bank No.:250210CFR

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55.41(b)()Page 52 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMUnit 1 is in a normal electrical alignment when the following occurred: An electrical disturbance caused the loss of the NORTH Bus in the Switchyard. The SOUTH Bus remains energized. Which of the following correctly describes the operational status of offsite power available to Unit 1? Unit 1-A. Auxiliary and Standby Transformers are both ENERGIZED. B. Auxiliary and Standby Transformers are both DE-ENERGIZED. C. Auxiliary Transformer is ENERGIZED, but the Standby Transformer is DE-ENERGIZED. D. Auxiliary Transformer is DE-ENERGIZED, but the Standby Transformer is ENERGIZED. Answer:CAuxiliary Transformer is ENERGIZED, but the Standby Transformer is DE-ENERGIZED.Exam Bank No.:1904Last used on an NRC exam:2010RO Sequence Number:27Page 53 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 077 AA2.05Tier:1Group/Category:1Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Operational status of offsite circuit.STP Lesson:LOT 201.30Objective Number:91662Given control room indications associated with the Offsite Electrical Distribution system, EVALUATE plant conditions.Attached Reference

Reference:

LOT 201.31, POP02-AE-0001, Rev. 23

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Source:BankDistractor JustificationA:INCORRECT: All distractors are plausible because the answer would be different in Unit 2. The Standby Transformer is supplied by the North Bus and will be de-energized under these conditions.B:INCORRECT: All distractors are plausible because the answer would be different in Unit 2. The Aux Transformer normally supplied via the breaker and a half lignup whould be unaffected by the loss of the North Bus.C:CORRECT: The U1 Standby Transformer is supplied from the North Bus and would now be de-energized. The Aux Transformer normally supplied via the breaker and a half lignup whould be unaffected by the loss of the North Bus.D:INCORRECT: All distractors are plausible because the answer would be different in Unit 2. The Standby Transformer is supplied by the North Bus and will be de-energized under these conditions. The Aux Transformer normally supplied via the breaker and a half lignup whould be unaffected by the loss of the North BusQuestion Level:HQuestion Difficulty3Justification:The student must be able to analyze the given plant condition and be able to determine the status of offsite power sources.RO Importance:3.2NRC Reference Req'd

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Modified fromExam Bank No.:190410CFR

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55.41(b)(7)Page 54 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMIn accordance with 0POP05-EO-FO02, Core Cooling Critical Safety Function Status Tree, which of the following temperature sensors are used to determine if an Inadequate Core Cooling Condition exists? A. Wide range Hot Leg RTDs (Thot) B. Wide range Cold Leg RTDs (Tcold)

C. Wide range Hot Leg RTDs and Cold Leg RTDs (Tave)

D. Core Exit Thermocouples (CETs) Answer:DCore Exit Thermocouples (CETs)Exam Bank No.:1909Last used on an NRC exam:2010RO Sequence Number:28Page 55 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 074 EK2.08Tier:1Group/Category:2Knowledge of the interrelationships between the Inadequate Core Cooling and the following:Sensors and DetectorsSTP Lesson:LOT 504.04Objective Number:92282STATE the individual parameter(s) used in each Critical Safety Function Status Tree.Attached Reference

Reference:

0POP05-EO-FO02, rev 2

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Source:BankDistractor JustificationA:INCORRECT: Plausible because this indication (Thot) is readily available to the operator and is indicative of RCS conditions, it is not to be used in accordance with procedure.B:INCORRECT: Plausible because this indication (Tcold) is readily available to the operator and is indicative of RCS conditions, it is not to be used in accordance with procedure.C:INCORRECT: Plausible because this indication (Tavg) is readily available to the operator and is indicative of RCS conditions, it is not to be used in accordance with procedure.D:CORRECT: The procedure directs the use of CETs when determining the status of Core Cooling.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of what indication is used to determine the status of the Core Cooling Safety Function.RO Importance:2.5NRC Reference Req'd

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Modified fromExam Bank No.:190910CFR

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55.41(b)(2)Page 56 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 25% power when the following occurs: RCP B breaker trips due to an instrument failure of the under voltage relay. Prior to the operator manually tripping the Reactor, what is the impact on RCS LOOP B flow? LOOP B FLOW FI-0427A would read zero in about _____(1)_____ and _____(2)_____. A. (1) 5 seconds (2) slowly rise to about 25% B. (1) 5 seconds (2) remain at 0% C. (1) 30 seconds (2) slowly rise to about 25% D. (1) 30 seconds (2) remain at 0% Answer:C (1) 30 seconds(2) slowly rise to about 25%Exam Bank No.:2497Last used on an NRC exam:NeverRO Sequence Number:29Page 57 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:003 K5.02Tier:2Group/Category:1Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP Coastdown on RCS Parameters.STP Lesson:LOT 201.02Objective Number:96651Given plant conditions, ANALYZE the conditions and acccurately PREDICT Reactor Coolant System response.Attached Reference

Reference:

LOT 201.05 Lesson Plan on RCPs and LOT 201.02 Lesson plan on the RCS.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because 5 seconds would be indicative of an overcurrent locked rotor trip or a mechanical issue with the pump motor.B:INCORRECT: Plausible because 5 seconds would be indicative of an overcurrent locked rotor trip or a mechanical issue with the pump motor. Also, plausible because the student must remember that back flow would occur in the effected LOOP.C:CORRECT: the RCP flywheel would cause the motor and pump to coast down even against back flow in the affected LOOP and then the flow would rise back up about 25% as back flow is established from the running RCPs.D:INCORRECT: Plausible because the student must remember that back flow would occur in the effected LOOP.Question Level:HQuestion Difficulty3Justification:The student must analyze the given condition and determine the effect on the RCS parameters.RO Importance:2.8NRC Reference Req'd

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Modified fromExam Bank No.:249710CFR

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55.41(b)(5)Page 58 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit tripped from 100% power with the following conditions: RCS Tave is 567 oF and stable. Pressurizer pressure is 1737 psig and lowering. Pressurizer level is 45% and rising. Containment pressure is 0.1 psig and stable. PRT pressure is 20 psig and rising. Which of the following events has likely occurred?

A. Steam Generator feedwater line break outside Containment. B. Charging flow control valve, FCV-0205, failed open. C. A Pressurizer PORV has failed open. D. Steam Generator steamline break outside Containment. Answer:CA Pressurizer PORV has failed open.Exam Bank No.:2096Last used on an NRC exam:2011RO Sequence Number:30Page 59 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 008 AK2.01Tier:1Group/Category:1Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:Valves.STP Lesson:LOT 501.21Objective Number:501215Given a set of conditions or event description, be able to PREDICT the sequence of events and trends of plant parameters for a transient or accident involving a decrease in Reactor Coolant Inventory.Attached Reference

Reference:

LOT j501.21

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Source:BankDistractor JustificationA:INCORRECT: a SG feedline break outside of containment would be an overcooling type of event, however RCS temperature is stable, not going down. Additionally, Pzr. level should be lowering if anovercoolig were occurring. Instead, it's going up.B:INCORRECT: If the charging flow control valve failed open it would normally cause a rise in Pressurizer level. However, based on the given conditions, a Safety Injection has occurred and the charging line has been isolated by Phase 'A' Isolation.C:CORRECT: Based on RCS pressure lowering with RCS temperature stable, the basic event going on is a loss of coolant and not an overcooling. With a Pressurizer PORV open, a low pressure area exists in the top of the Pzr causing RCS water to expand into the Pzr. raising Pzr. Level. There is no Containment pressure response because the PORV discharges to the PRT.D:INCORRECT: a SG steamline break outside of containment would be an overcooling type of event, however RCS temperature is stable, not going down. Additionally, Pzr. level should be lowering if anovercooling were occurring. Instead, it's going up.Question Level:HQuestion Difficulty3Justification:Student must be able to determine the event that has occurred based on the given plant conditions.RO Importance:2.7NRC Reference Req'd

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Modified fromExam Bank No.:209610CFR

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55.41(b)(5)Page 60 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is operating at 100% power with the following conditions: A leak in the CCW system develops causing CCW Surge Tank level to lower. CCW Surge Tank level is currently at 63%. Based on the given conditions, CCW flow _____(1)_____ been isolated to the RCP's. Maintaining CCW flow to the RCP's is important to prevent damage to the _____(2)_____ of any operating RCP.

(1) (2) A. HAS Thermal barrier B. has NOT Motor bearings C. has NOT Thermal Barrier D. HAS Motor bearings Answer:Bhas NOT; Motor bearingsExam Bank No.:2097Last used on an NRC exam:2011RO Sequence Number:31Page 61 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 015/017AK3.01Tier:1Group/Category:1Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Potential damage from high winding and/or bearing temperatures.STP Lesson:LOT 201.05Objective Number:97119Given plant conditions, ANALYZE the conditions and accurately PREDICT Reactor Coolant Pump response.Attached Reference

Reference:

LOT 201.05, LOT 201.12

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Source:BankDistractor JustificationA:INCORRECT: CCW flow has NOT yet been isolated to the RCP's. Additionally, the RCP Thermal Barriers are not at risk for damage because their CCW cooling is only important if normal seal injection flow is lost.B:CORRECT: A Surge Tank level of 63% is below the 'first level isolation' so CCW flow has been isolated to some components, but not the RCP's. If CCW Surge Tank level continued to lower, CCW flow to the RCP's would be isolated at a level of 61.5%. If it is isolated and the RCP continues to run, the motor bearings will damaged due to the loss of cooling.C:INCORRECT: CCW flow has not been isolated to the RCP's, as stated. However, the RCP Thermal Barriers are not at risk for damage because their CCW cooling is only important if normal seal injection flow is lost.D:INCORRECT: CCW flow has NOT yet been isolated to the RCP's. If it is isolated and the RCP continues to run, the motor bearings will damaged due to the loss of cooling.Question Level:HQuestion Difficulty3Justification:Students must be able to determine how the CCW system has been affected based on the current plant conditions and have knowledge of the RCP components that could be potentially damaged if CCW is lost.RO Importance:2.5NRC Reference Req'd

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Modified fromExam Bank No.:209710CFR

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55.41(b)(7)Page 62 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMAn ECO has been hung on an ECW Pump for maintenance that is expected to last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. As a result of this, which of the following is required in accordance with 0POP01-ZQ-0022, Plant Operations Shift Routines? A. - Make an entry in the Control Room Logbook OR Operability Assessment System log. - Update the completed Safety Function Checklist performed at the beginning of shift. B. - Make an entry in the Control Room Logbook AND Operability Assessment System log. - Update the completed Safety Function Checklist performed at the beginning of shift. C. - Make an entry in the Control Room Logbook OR Operability Assessment System log. - Add the LCO to the Shift Turnover Checklist. D. - Make an entry in the Control Room Logbook AND Operability Assessment System log. - Add the LCO to the Shift Turnover Checklist. Answer:D-Make an entry in the Control Room Logbook AND Operability Assessment System log; - Add the LCO to the Shift Turnover ChecklistExam Bank No.:2368Last used on an NRC exam:2014RO Sequence Number:32Page 63 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.2.36Tier:3Group/Category:2Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operationsSTP Lesson:LOT 507.01Objective Number:92184Given the title of an administrative procedure, identify the actions that are performed by the control room operator.Attached Reference

Reference:

0POP01-ZQ-0022 steps 3.4.1.7 and 6.4.2.2

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Source:BankDistractor JustificationA:INCORRECT: Plausible because both methods are used for tracking and are somewhat redundant, however both are required and the Safety Function Checklist is used to verify the appropriate equipment is available, but it is only performed once per shift (at the beginning) and not updated.B:INCORRECT: Plausible because the Safety Function Checklist is used to verify the appropriate equipment is available, but it is only performed once per shift (at the beginning) and not updated.C:INCORRECT: Plausible because both methods are used for tracking and are somewhat redundant, however both are required.D:CORRECT: Per the referenced procedure, in this situation the CR log, OAS log (if greater than one shift) and turnover checklist must be updated to alllow tracking the activity.Question Level:HQuestion Difficulty3Justification:The student must apply procedural requirements to the given situation.RO Importance:3.1NRC Reference Req'd

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Modified fromExam Bank No.:236810CFR

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55.41(b)(10)Page 64 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMAn on shift crew was performing 0PSP03-RC-0013, Reactor Makeup Water to PZR Relief Tank Check Valve Test, when it was noticed that the surveillance failed to meet the Acceptance Criteria. The _____(1)_____ will determine Operability status and LCO Action entry requirements and the _____(2)_____ will ensure that a condition report is initiated. A. (1) Engineering Manager (2) Test Coordinator B. (1) Shift Manager (2) Test Coordinator C. (1) Engineering Manager (2) Plant Surveillance Coordinator D. (1) Shift Manager (2) Plant Surveillance Coordinator Answer:B (1) Shift Manager(2) Test CoordinatorExam Bank No.:2442Last used on an NRC exam:NeverRO Sequence Number:33Page 65 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 022 G2.2.12Tier:1Group/Category:1Loss of Reactor Coolant Makeup:Knowledge of Surveillance Procedures.STP Lesson:LOT 507.01Objective Number:92183Given the title of an administrative procedure, IDENTIFY the individuals (by job title) with specific responsibilities in the procedure.Attached Reference

Reference:

Administrative Procedures Lesson LOT 5.7.01, 0PGP03-ZE-0004, Plant Surveillance Program

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the Engineering Manager does determine Maintenance Rule issues but not Operability and LCO Action issues.B:CORRECT: The Shift Manager determines Operability and LCO requirements and the Test Coordinator would ensure a condition report is written.C:INCORRECT: Plausible because the Engineering Manager does determine Maintenance Rule issues but not Operability and LCO Action issues and the Plant Surveillance Coordinator does implement the Surveillance Program but would not necessarily be present during a surveillance test.D:INCORRECT: Plausible because the Plant Surveillance Coordinator does implement the Surveillance Program but would not necessarily be present during a surveillance test.Question Level:FQuestion Difficulty3Justification:Student must have fundamental knowledge of the Plant Surveillance Procedure.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:244210CFR

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55.41(b)(10)Page 66 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMPer 0ERP01-ZV-IN03, Emergency Response Organization Notifications, which of the following is the LOWEST Emergency Classification Level at which the ENS Communicator must activate the Emergency Notification Response System (ENRS)? A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer:AUnusual EventExam Bank No.:2446Last used on an NRC exam:NeverRO Sequence Number:34Page 67 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.4.29Tier:3Group/Category:4Knowledge of the emergency plan.STP Lesson:LOT 507.01Objective Number:68900Maintain required Mode 1 logs, records, charts, printouts and status boards in accordance with 0POP01-ZQ-0022.Attached Reference

Reference:

0ERP01-ZV-SH01, Shift Manager, 0ERP01-ZV-IN03, Emergency Response Organization Notification

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Source:NewDistractor JustificationA:CORRECT: The ENRS is activated at an Unusual Event even though it is the lowest activation level.The ENS Communicator, normally a Reactor Operator, is responsible for this E-Plan Duty.B:INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.C:INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.D:INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.Question Level:FQuestion Difficulty2Justification:The student must have knowledge of ENRS and when it needs to be activated.RO Importance:3.1NRC Reference Req'd

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Modified fromExam Bank No.:244610CFR

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55.41(b)(10)Page 68 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following is the reason during a SGTR that primary pressure is equalized with secondary pressure in the ruptured SG? A. Reduce the primary to secondary leakage. B. Ensure the Pressurizer is not overfilled. C. Reduce the risk of rupturing another SG. D. Ensure the Reactor Vessel Head remains full. Answer:AReduce the primary to secondary leakage.Exam Bank No.:2466Last used on an NRC exam:NeverRO Sequence Number:35Page 69 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 038 EK3.01Tier:1Group/Category:1Knowledge of the reasons for the following responses as they apply to the SGTR:Equalizing pressure on the primary and secondary sides of ruptured S/G.STP Lesson:LOT 504.15Objective Number:92408GIVEN a copy of a step from 0POP05-EO-EO30, STATE/IDENTIFY how the action is performed and the basis for the action to include the action itself, its purpose and the result.Attached Reference

Reference:

LOT 504.15 Lesson Plan for SGTR and WOG Background document for SGTR

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Source:NewDistractor JustificationA:CORRECT: Per the WOG background document the reason for equalizing pressure in the ruptured SG is to reduce primary to secondary leakage.B:INCORRECT: Plausible because refilling the PZR is part of depressurizing the RCS but overfiling the PZR is a reason for stopping the depressurization early.C:INCORRECT: Plausible because reducing RCS pressure could lower the risk of another tube failure but is not the reason for equalizing pressure in the ruptured SG.D:INCORRECT: Plausible because it is desired to not have voids in the Reactor Vessel Head which could come from depressurizing the RCS but depending on plant conditions this could be an unavoidable consequence. A separate EOP would be used for this condition.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of the reasons for performing steps in EOPs.RO Importance:4.1NRC Reference Req'd

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Modified fromExam Bank No.:246610CFR

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55.41(b)(5)Page 70 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring a control room evacuation due to a fire, in accordance with 0POP04-ZO-0001, Control Room Evacuation, why is the Reactor/Main Turbine tripped before leaving the control room? Because of- A. the possibility of spurious operation or unavailability of supporting equipment. B. a loss of EHC control to the Main Turbine. C. a loss of Main Feedwater. D. the lack of control functions available at the Auxiliary Shutdown Panel. Answer:A the possibility of spurious operation or unavailability of supporting equipment.Exam Bank No.:2469Last used on an NRC exam:NeverRO Sequence Number:36Page 71 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 068 AK3.02Tier:1Group/Category:2Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation:System response to turbine trip.STP Lesson:LOT 505.01Objective Number:92105STATE the purpose of, and DESCRIBE the scope of the referenced procedure.Attached Reference

Reference:

LOT 505.01 Lesson Plan on Off-Normal procedures and 0POP04-ZO-0001, Control Room Evacuation

Attachment:

Source:NewDistractor JustificationA:CORRECT: Even though the control room could be evacuated for reasons other than a fire, in the basis for 0POP04-ZO-0001 all of the steps performed prior to leaving the control room are based on a fire causing spurious operation of equipment. This includes tripping the reactor which would causea main turbine trip.B:INCORRECT: Plausible because losing EHC control would be a reason for tripping the main turbine but EHC is not specifically shutdown during a control room evacuation like main steam is.C:INCORRECT: Plausible because losing Main Feedwater would be a reason for tripping the main turbine but Feedwater is not specifically shutdown during a control room evacuation like main steam is.D:INCORRECT: Plausible because there is limited controls at the auxiliary shutdown panel but that is not the reason for tripping the Reactor/Main Turbine and performing other steps prior to leaving the control room in accordance with 0POP04-ZO-0001.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of the basis for steps performed during a control room evacuation.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:246910CFR

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55.41(b)(7)Page 72 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMUsing the RWP attached, which of the following Operations activities would require the Operator to request a separate RWP? The Operator is tasked with- A. replacing a CVCS Reactor Coolant filter. B. hooking up a temporary Pressure Gauge on a Centrifugal Charging pump. C. removing a drain hose from one contaminated area and putting it in another. D. performing a valve lineup in a MAB valve pit that is posted 'CONTAMINATED AREA.' Answer:Areplacing a CVCS Reactor Coolant filter.Exam Bank No.:2488Last used on an NRC exam:NeverRO Sequence Number:37Page 73 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.3.7Tier:3Group/Category:3Ability to comply with radiation work permit requirements during normal or abnormal conditions.STP Lesson:LOT 507.01Objective Number:92186Given the title of an administrative procedure, DISCUSS the requirements associated with the referenced procedure.Attached Reference

Reference:

LOT 507.01 Lesson on Administrative Procedures - 0PGP03-ZR-0051, Radiological Access Controls/Standards

Attachment:

RWP 2016-0-0003Source:NewDistractor JustificationA:CORRECT: Replacing a reactor coolant filter would be in the definition of a process filter which would not be allowed to be changed out with the given RWP.B:INCORRECT: All distractors are palusible because it requires the student to apply their knowledge of operations activities and the requirements of RWPs.C:INCORRECT: All distractors are palusible because it requires the student to apply their knowledge of operations activities and the requirements of RWPs.D:INCORRECT: All distractors are palusible because it requires the student to apply their knowledge of operations activities and the requirements of RWPs.Question Level:HQuestion Difficulty2Justification:The student must be able to analyze each operations task and apply it to the requirements of the given RWP to determine which task would require requesting a different RWP.RO Importance:3.5NRC Reference Req'd

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Modified fromExam Bank No.:248810CFR

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55.41(b)(12)Page 74 of 150

3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMFor a Steam Generator Tube Rupture (SGTR), which of the following identifies Mitigation Strategies (Major Actions) specific to a SGTR that will need to be performed? A. Check for Main Steam Line Isolation Check at least one Steam Generator NOT Faulted B. Identify and Isolate the affected Steam Generator Depressurize the RCS C. Check for Main Steam Line Isolation Identify and Isolate the affected Steam Generator D. Check at least one Steam Generator NOT Faulted Depressurize the RCS Answer:B Identify and Isolate affected Steam GeneratorDepressurize RCS to restore InventoryExam Bank No.:2215Last used on an NRC exam:NeverRO Sequence Number:38Page 75 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.4.6Tier:3Group/Category:4Knowledge of EOP Mitigation StrategiesSTP Lesson:LOT 504.15Objective Number:80942From memory, STATE/IDENTIFY all Control Room Instrumentation available for determining what actions need to be taken to minimize RCS to secondary leakage during a SGTR.Attached Reference

Reference:

LOT 504.15 Lesson plan on procedure for SGTR - 0POP05-EO-EO30, Steam Generator Tube Rupture

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Source:BankDistractor JustificationA:INCORRECT: Plausible because these are major actions but they are specific for a faulted steam generator.B:CORRECT: Both major actions are specific to a SGTR.C:INCORRECT: Plausible because main steam line isolation is a major action but for a faulted steam generator. Identifying and Isolating the affected steam generator is a major action for both a SGTR and a faulted steam generator.D:INCORRECT: Plausible because checking at least one steam generator not faulted is a major action but specific to a faulted steam generator.Question Level:FQuestion Difficulty3Justification:The student needs to have fundamental knowledge of Emergency procedure Major Action Steps/Mitigation Strategies.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:221510CFR

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55.41(b)(10)Page 76 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhen heating up the Secondary Plant and performing a Plant Startup, the operating procedures have cautions about avoiding Hydraulic Transients (Water Hammer) when operating the Main Steam and Reheat Steam Systems. Which of the following describes a cause and definition of a Hydraulic Transient? CAUSE DEFINITION A. During a manual Cold Start of MSRs, rapidly initiating Main Steam through the Reheat Control Valves. An increase in steam demand with a resultant pressure reduction. B. When at NOP/NOT, opening a Main Steam Isolation Valve with downstream pressure 60 psig lower. An increase in steam demand with a resultant pressure reduction. C. During a manual Cold Start of MSRs, rapidly initiating Main Steam through the Reheat Control Valves. The shock imposed on piping from initiating steam flow through pipes containing liquid condensate. D. When at NOP/NOT, opening a Main Steam Isolation Valve with downstream pressure 60 psig lower. The shock imposed on piping from initiating steam flow through pipes containing liquid condensate. Answer:C During a manual Cold Start of MSRs, rapidly initiating Main Steam through the Reheat Control Valves. - The shock imposed on piping from initiating steam flow through pipes containing liquid condensate.Exam Bank No.:2252Last used on an NRC exam:2014RO Sequence Number:39Page 77 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:039 K5.01Tier:2Group/Category:1Knowledge of the operational implications of the following concepts as they apply to the MRSS:Definition and causes of steam/water hammer.STP Lesson:LOT 102.57Objective Number:N99862Explain operational implications of water (fluid) hammer.Attached Reference

Reference:

LOT 102.57 Lesson Plan Handout Page 46 and Procedures 0POP03-ZG-0003, Secondary Plant Startup, and 0POP02-MS-0001, Main Steam System.

Attachment:

Source:BankDistractor JustificationA:INCORRECT: This distractor is credible because it describes the definition of SWELL which is related to thermodynamic processes.B:INCORRECT: This is a credible distractor because it decribes a condition in 0POP03-ZG-0003, where opening the MSIV with a differential pressure of greater than 50 psig can cause the phenomenom of SWELL in the corresponding SG but at the given differential of 60 psig it would not cause water hammer. Also, it describes the definition of SWELL which is related to thermodynamic processes.C:CORRECT: Manually rapidly opening an MSR Reheat Control Valve during a Cold Start (described in 0POP02-MS-0001) can cause water hammer and thereby system damage. Procedure requires valves to be throttled to raise temperature no greater than 100 degrees F per hour. The definition of water hammer includes the shock imposed on piping from initiating steam flow through pipes containing liquid condensate.D:INCORRECT: This is a credible distractor because it decribes a condition in 0POP03-ZG-0003, where opening the MSIV with a differential pressure of greater than 50 psig can cause the phenomenom of SWELL in the corresponding SG but at the given differential of 60 psig it would not cause water hammer.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of thermodynamics and procedures covering Secondary Plant Startup from cold conditions to 100% power.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:225210CFR

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55.41(b)(5)Page 78 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit was at 100% power when the following occurred: A Steam Line break on SG 1A in the IVC up stream of the Main Steam Isolation Valve. One (1) of the Steam Pressure Transmitters on SG 1A is damaged causing it to read off-scale HIGH on Control Board and QDPS indications. The damaged Steam Pressure Transmitter was selected for control with the associated Steam Flow Transmitter for SG Water Level Control. (1) The Controlling Channel of Steam Flow will indicate _____(1)_____ than actual steam flow? AND (2) Will a Main Steam Isolation (MSI) automatically occur if required? (1) (2) A. HIGHER NO B. LOWER YES C. LOWER NO D. HIGHER YES Answer:D Indicated Steam Flow will read HIGHER than actual Steam Flow. - If needed, an MSI will AUTOMATICALLY actuate.Exam Bank No.:2305Last used on an NRC exam:NeverRO Sequence Number:40Page 79 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 040 AK2.02Tier:1Group/Category:1Knowledge of the interrelations between the Steam Line Rupture and the following:Sensors and detectors.STP Lesson:LOT 202.02Objective Number:12768Given a plant or system condition, PREDICT the operation of the Main Steam System.Attached Reference

Reference:

LOT 202.02 Lesson Plan

Attachment:

Source:BankDistractor JustificationA:INCORRECT: This distractor is credible because the steam pressure channel did fail high but when the other two steam pressure channels lower to the actuation setpoint then an MSI will automaticallyactuate.B:INCORRECT: This distractor is credible because the failed high steam pressure channel does affect the associated steam flow channel but the pressure compensation causes the steam flow channel to read higher than actual steam flow.C:INCORRECT: This distractor is credible because the failed high steam pressure channel does affect the associated steam flow channel but the pressure compensation causes the steam flow channel to read higher than actual steam flow. Also, the steam pressure channel did fail high but when the other two steam pressure channels lower to the actuation setpoint then an MSI will automatically actuate.D:CORRECT: With a high failure of a controlling steam pressure channel, the controlling steam flow channel will indicate higher than actual steam flow due to the pressure compensation. One of the three channels of Steam Pressure feeding the MSI actuation will not affect the ability for an automatic actuation on 2 of the remaining 3 Steam Pressure Channels.Question Level:HQuestion Difficulty3Justification:The student must be able to evaluate the given conditions to determine the affects on indicated steam flow and MSI actuation.RO Importance:2.6NRC Reference Req'd

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Modified fromExam Bank No.:230510CFR

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55.41(b)(7)Page 80 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following is: (1) The design feature that will prevent the Pressurizer (PZR) Heaters from being uncovered during an instantaneous 10% addition of turbine load? AND (2) The PZR level at which the PZR Heaters will FIRST de-energize? (1) (2) A. Design Water and Steam Volume of the PZR. 17% PZR Level B. Design Water and Steam Volume of the PZR. 8% PZR Level C. Design Control Response of Charging Flow Control Valve CV-FV-0205. 17% PZR Level D. Design Control Response of Charging Flow Control Valve CV-FV-0205. 8% PZR Level Answer:ADesign Water and Steam Volume of the PZR. - 17% PZR LevelExam Bank No.:2321Last used on an NRC exam:NeverRO Sequence Number:41Page 81 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:010 K4.02Tier:2Group/Category:1Knowledge of the PZR PCS design feature(s) and/or interlock(s) which provide for the following:Prevention of uncovering PZR heaters.STP Lesson:LOT 201.04Objective Number:91011DESCRIBE the basis for the sizing of the Pressurizer and the PRT.Attached Reference

Reference:

LOT 201.04 Lesson Plan on PZR, PRT and RCDT.

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Source:BankDistractor JustificationA:CORRECT: The design water and steam volume of the PZR will prevent uncovering the PZR Heaters on a 10% load increase and the PZR Heaters will de-energize at 17% PZR level.B:INCORRECT: This distractor is credible because the PZR level given (8%) is used in many of the emergency procedures to direct the operator to restart SI pumps and/or initiate SI.C:INCORRECT: This distractor is credible because the charging flow control valve will open on a lowering PZR level but it is not sized to prevent uncovering the PZR Heaters on a step load increaseof 10%.D:INCORRECT: This distractor is credible because the charging flow control valve will open on a lowering PZR level but it is not sized to prevent uncovering the PZR Heaters on a step load increaseof 10%. Also, the PZR level given (8%) is used in many of the emergency procedures to direct the operator to restart SI pumps and/or initiate SI.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of PZR Pressure and Level Control system design features and control interlocks.RO Importance:3.0NRC Reference Req'd

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Modified fromExam Bank No.:232110CFR

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55.41(b)(7)Page 82 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power with the following conditions: The Component Cooling Water (CCW) 'MODE SEL' switches are in the following position: o Train A is in 'RUN' o Train B is in 'AUTO' o Train C is in 'OFF' CCW Pumps Control Room handswitches are as follows: o CCW Pump 1A is in 'AUTO' and is running. o CCW Pump 1B and 1C are in 'AUTO' and are NOT running. Reactor Containment Fan Cooler (RCFC) Control Room handswitches are as follows: o RCFCs Trains A and B are in 'AUTO' and running with RCB Chill Water aligned. o RCFC Train C is in 'AUTO' and NOT running with RCB Chill Water aligned. Subsequently: A Loss of Offsite Power occurs. Which of the following manual actions will the Control Room Operator have to perform in order to establish all three trains of cooling to Containment? Manually- A. START CCW Pump 1C. B. START RCFCs on Train C C. CLOSE the RCB Chill Water to RCFC Motor Operated isolation valves. D. OPEN the CCW to RCFC Motor Operated isolation valves. Answer:DOPEN the CCW to RCFC Motor Operated isolation valves.Exam Bank No.:2323Last used on an NRC exam:NeverRO Sequence Number:42Page 83 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:022 A4.04Tier:2Group/Category:1Ability to manually operate and/or monitor in the control room:Valves in the CCS.STP Lesson:LOT 201.12Objective Number:5213GIVEN a plant or system condition, PREDICT the operation of the Component Cooling Water System.Attached Reference

Reference:

LOT 201.12 Lesson Plan PPT

Attachment:

Source:BankDistractor JustificationA:INCORRECT: This distractor is credible because even though the MODE SEL switch on Train C is in the 'OFF' position the CCW Pump 1C will still get a start signal.B:INCORRECT: This distractor is credible because even though the MODE SEL switch on Train C is in the 'OFF' position the RCFCs on Train C will still get a start signal.C:INCORRECT: This distractor is credible because even though the CCW to RCFCs Motor Operated isolation valves do not get a signal to auto open, the RCB Chill Water to RCFCs Motor Operated isolation valves do get an Auto close signal.D:CORRECT: The CCW to RCFC Motor Operated isolation valves do not get a signal to auto open and must be manually opend to establish containment cooling on all 3 trains.Question Level:HQuestion Difficulty3Justification:The student must be able to evaluate the given conditions to determine which manual action needs to be performed.RO Importance:3.1NRC Reference Req'd

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Modified fromExam Bank No.:232310CFR

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55.41(b)(7)Page 84 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMPlant procedures put a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit on loading over 6050kw when running an ESF Diesel Generator. Running the Diesel Generator overloaded could result in a Generator _____(1)_____ condition causing the Diesel Generator to trip during _____(2)_____ mode of operation. A. (1) DIFFERENTIAL (2) NORMAL OR EMERGENCY B. (1) OVERCURRENT (2) ONLY NORMAL C. (1) DIFFERENTIAL (2) ONLY NORMAL D. (1) OVERCURRENT (2) NORMAL OR EMERGENCY Answer:B(1) OVERCURRENT(2) ONLY NORMALExam Bank No.:2494Last used on an NRC exam:NeverRO Sequence Number:43Page 85 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:062 A1.01Tier:2Group/Category:1Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:Significance of D/G load limitsSTP Lesson:LOT 201.39Objective Number:45057STATE the Emergency Diesel Generator trips in the emergency mode and in the test modeAttached Reference

Reference:

LOT 201.39 Lesson Plan and 0POP09-AN-0102, B-8

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the generator differential is similar to the generator overcurrent in that it trips the DG and opens the output breaker but the generator differential would be indicative of generator damage not necessarily caused by overloading the generator. Also the generator overcurrent trip is only active in the normal mode.B:CORRECT: Overloading the DG would cause a generator overcurrent trip and is only active in the normal mode. In addition this condition would likely only happen while the DG is paralled to offsite power. When paralled to offsite power the DG is running in the normal mode.C:INCORRECT: Plausible because the generator differential is similar to the generator overcurrent in that it trips the DG and opens the output breaker but the generator differential would be indicative of generator damage not necessarily caused by overloading the generator.D:INCORRECT: Plausible, however, the generator overcurrent trip is only active in the normal mode.Question Level:FQuestion Difficulty3Justification:The applicant must have knowledge of the trips availables during the different modes of ESF DG operation and now the difference between a Generator Overcurrent and a Generator Differential Trip.RO Importance:3.4NRC Reference Req'd

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Modified fromExam Bank No.:249410CFR

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55.41(b)(7)Page 86 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power with the following conditions: A fuel handling accident has occurred in the FHB involving a spent fuel assembly. FHB radiation levels are rising. A high alarm has been received on RT-8035 and RT-8036 resulting in the required FHB HVAC actuation. (1) Which of the following identifies the expected trend for RT-8035 and RT-8036? AND (2) What action must the operators perform in accordance with 0POP04-FH-0001, Fuel Handling Accident? (1) (2) A. Continue to rise. Check only one train of FHB Exhaust Fans (Main and Booster) in operation. B. Continue to rise. Place Control Room Envelope HVAC in Emergency Mode. C. Begin to lower. Check only one train of FHB Exhaust Fans (Main and Booster) in operation. D. Begin to lower. Place Control Room Envelope HVAC in Emergency Mode. Answer:B Continue to rise; Place Control Room Envelope HVAC in Emergency ModeExam Bank No.:2332Last used on an NRC exam:NeverRO Sequence Number:44Page 87 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:073 A1.01Tier:2Group/Category:1Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: Radiation levelsSTP Lesson:LOT 202.38Objective Number:2207LIST all the systems that interface with the Fuel Handling HVAC System and state the function of each interfaceAttached Reference

Reference:

LOT202.38 PowerPoint, 0POP04-FH-0001 steps 19 and 24

Attachment:

Source:BankDistractor JustificationA:INCORRECT: Credible because the POP04 has the operator check only 1 filter train in service (and 2 sets of fans).B:CORRECT: Since the rad monitors are upstream of the filters, rad levels will continue to rise. The POP04 requires operators to place CRE HVAC in emergency mode.C:INCORRECT: Credible because the high rad alarm will put filters in service which will lower the rad levels in the HVAC exhaust, so the applicant must understand system design to determine indicated rad will still rise. Action is credible because the POP04 has the operator check only 1 filter train in service (and 2 sets of fans).D:INCORRECT: Credible because the high rad alarm will put filters in service which will lower the rad levels in the HVAC exhaust, so the applicant must understand system design to determine indicated rad will still rise.Question Level:HQuestion Difficulty3Justification:The applicant must evaluate the given information and use knowledge of system design to determine the proper trend. Knowledge of procedural requirements is also required.RO Importance:3.2NRC Reference Req'd

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Modified fromExam Bank No.:233210CFR

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55.41(b)(11)Page 88 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power with the following condition: ESF D/G #11 is paralleled to the ESF 4.16 KV BUS E1A to support 0PSP03-DG-0001, Standby Diesel 11(21) Operability Test. Subsequently: A Loss of Offsite Power (LOOP) occurred. ESF DG #13 failed to start and cannot be manually started. If needed, which component(s) listed below would NOT have auto started but could be manually started? A. BOTH Centrifugal Charging Pumps 1A and 1B B. ONLY Centrifugal Charging Pump 1A C. BOTH HHSI Pumps 1A and 1B D. ONLY HHSI Pump 1B Answer:CBOTH HHSI Pumps 1A and 1BExam Bank No.:2333Last used on an NRC exam:NeverRO Sequence Number:45Page 89 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:Tier:2Group/Category:1Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:ED/G (manual loads)STP Lesson:LOT 201.41Objective Number:98035Given a plant or system condition, predict the operation of the ESF Load Sequencer.Attached Reference

Reference:

LOT 201.41 Lesson Plan PPT on ESF Load Sequencers

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Source:BankDistractor JustificationA:INCORRECT: This distractor is credible because Centrifugal Charging pumps do not Auto start on a Mode II signal but CCP 1A is powered from Train C and will not be able to be manually started either.B:INCORRECT: This distractor is credible because Centrifugal Charging pumps do not Auto start on a Mode II signal but CCP 1A is powered from Train C and will not be able to be manually started either.C:CORRECT: A LOOP will cause the ESF DG Sequencers to start on a Mode II. However, a Mode II signal will not Auto start the HHSI Pumps. Therefor HHSI Pump 1A and 1B would have to be manualy started if needed.D:INCORRECT: This distractor is credible because other surviellances use SI signals to emergency start the ESF DG for testing and if the student believed that a Mode I (SI) was present on Train A when the Mode II (LOOP) occurred then only HHSI Pump 1B would be manually started if needed.Question Level:HQuestion Difficulty3Justification:The student must be able to evaluate the given conditions to determine which components, if needed, did not auto start and could be manually started.RO Importance:3.6NRC Reference Req'd

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Modified fromExam Bank No.:233310CFR

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55.41(b)(7)Page 90 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMAfter a Reactor Trip from 100% power, indicated Reactor power will lower to _____(1)_____ power in 2 to 3 seconds and then lower to a subcritical equilibrium level at a rate of _____(2)_____ decades per minute. A. (1) 6% (2) - 0.3 B. (1) 10% (2) - 0.1 C. (1) 6% (2) - 0.1 D. (1) 10% (2) - 0.3 Answer:A(1) 6%(2) - 0.3Exam Bank No.:2439Last used on an NRC exam:NeverRO Sequence Number:46Page 91 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 007 EK1.04Tier:1Group/Category:1Knowledge of the operational implications of the following concepts as they apply to the reactor trip:Decrease in reactor power following reactor trip (prompt drop and subsequent decay).STP Lesson:LOT 101.25Objective Number:N99751Explain the shape of the curve of reactor power versus time after a reactor trip.Attached Reference

Reference:

Fundamental Lesson Plan LOT 101.25

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Source:NewDistractor JustificationA:CORRECT: After a reactor trip power makes a prompt drop to about 6% at which time delayed neutrons slow the decrease at a calculated rate of minus 0.3 decades per minute.B:INCORRECT: Plausible because 10% reactor power is a basis for other reactor limits such as when certain reactor trips need to be blocked/activated. (6% power is close to the 5% basis for the size of Aux Feedwater.) Also, a minus 0.1 decades per minute lowering of reactor power would be indicative of a failed intermediate range instrument which would require manually energizing source range instruments. (A similar event happened on a recent reactor trip in Unit 1 CR-16-1227)C:INCORRECT: Plausible because a minus 0.1 decades per minute lowering of reactor power would be indicative of a failed intermediate range instrument which would require manually energizing source range instruments. (A similar event happened on a recent reactor trip in Unit 1 CR-16-1227)D:INCORRECT: Plausible because 10% reactor power is a basis for other reactor limits such as when certain reactor trips need to be blocked/activated. (6% power is close to the 5% basis for the size of Aux Feedwater.)Question Level:HQuestion Difficulty2Justification:Stundent must analyze the condition given and have the fundamental concept of how reactor power reacts after a reactor trip.RO Importance:3.6NRC Reference Req'd

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Modified fromExam Bank No.:243910CFR

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55.41(b)(1)Page 92 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA Reactor Trip and Safety Injection have occurred with the following conditions: Reactor Coolant System (RCS) Pressure is stable at 1500 psig. RCS Temperature is stable at 525ºF. Safety Injection has been reset. All Reactor Coolant Pumps (RCPs) are running. HHSI Pumps A and B are running. HHSI Pump C failed to start due to an overcurrent. All LHSI Pumps are secured and in AUTO. Subsequently: HHSI Pump B trips on overcurrent. RCS Pressure lowers to 390 psig and stabilizes. RCS Temperature lowers to 445ºF and stabilizes. Per procedural requirements, RCPs will be ______(1)______ and LHSI Pumps will be _____(2)______. (1) (2) A. running secured in AUTO B. secured secured in AUTO C. running running D. secured running Answer:Dsecured - runningExam Bank No.:2440Last used on an NRC exam:NeverRO Sequence Number:47Page 93 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 011 EA2.01Tier:1Group/Category:1Ability to determine or interpret the following as they apply to a Large Break LOCA:Actions to be taken, based on RCS temperature and pressure - saturated and superheatedSTP Lesson:LOT 504.05Objective Number:92218From memory, STATE/IDENTIFY the criteria on the conditional information page of POP05-EO-EO00 to include operator response, initiating parameter(s) and values.Attached Reference

Reference:

CIP of 0POP05-EO-EO00 and 0POP05-EO-EO10

Attachment:

Source:NewDistractor JustificationA:INCORRECT: Plausible because if no HHSI Pumps are running then the RCPs would be required to stay running and even though RCS pressure has stabilized at about LHSI Pump shut off head the procedure requires the pumps to be started if pressure drops below 415 psig.B:INCORRECT: Plausible because even though RCS pressure has stabilized at about LHSI Pump shut off head the procedure requires the pumps to be started if pressure drops below 415 psig.C:INCORRECT: Plausible because if no HHSI Pumps are running then the RCPs would be required to stay running.D:CORRECT: With the given conditions the RCPs would be required to be manually secured and the LHSI Pumps would be manually started.Question Level:HQuestion Difficulty3Justification:Student must be able to analyze the given conditions to determine the correct answer.RO Importance:4.2NRC Reference Req'd

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Modified fromExam Bank No.:244010CFR

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55.41(b)(10)Page 94 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring a SBLOCA the following parameters are given: RCS pressure is indicating 1650 psig. PZR pressure is 1700 psig. RCS Hot Leg temperatures are indicating 590ºF. Max Quad TC Avg temperature is indicating 600ºF. Which of the following would be the correct indication for subcooling from QDPS?

A. 10ºF B. 14ºF C. 20ºF D. 24ºF Answer:A10 degrees FExam Bank No.:2441Last used on an NRC exam:NeverRO Sequence Number:48Page 95 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:EPE 009 EA1.16Tier:1Group/Category:1Ability to operate and monitor the following as they apply to a small break LOCA:Subcooling margin monitorsSTP Lesson:LOT 202.44Objective Number:91674List the plant systems/components controlled by QDPS.Attached Reference

Reference:

QDPS Lesson Plant LOT 202.44

Attachment:

Source:NewDistractor JustificationA:CORRECT: QDPS uses Max Quad TC average and RCS pressure to calculate subcooling.B:INCORRECT: Plausible because answer reflects using PZR pressure in the calculation instead of RCS pressure.C:INCORRECT: Plausible because answer reflects using Hot Leg temperatures in the calculation instead of Max Quad TC Avg.D:INCORRECT: Plausible because answer reflects using Hot Leg temperatures in the calculation instead of Max Quad TC Avg and reflects using PZR pressure in the calculation instead of RCS pressure.Question Level:HQuestion Difficulty3Justification:Student must be able to know how subcooling is calcualted using QDPS.RO Importance:4.2NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:244110CFR

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55.41(b)(7)Page 96 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit was at 100% power when a SI occurred due to a LOCA in Containment with the following condition: Containment pressure is at 7 psig and rising. Subsequently: The normal feeder breaker to Train B 4.16 KV ESF bus tripped open causing a loss of power on the bus. AT THE SAME TIME that the Train B Diesel Generator output breaker closed in, re-energizing the bus, a Containment Spray (CS) Actuation Signal was generated from containment pressure. (1) Which of the following states the times at which Train B CS Pump started? AND (2) Which of the following states when the Train B CS Pump discharge valve started to open? (1) (2) A. Immediately 1 second B. Immediately 15 seconds C. 15 seconds 15 seconds D. 15 seconds 1 second Answer:D15 seconds, 1 secondExam Bank No.:2444Last used on an NRC exam:NeverRO Sequence Number:49Page 97 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:026 A3.01Tier:2Group/Category:1Ability to monitor automatic action of the CSS, including:Pump starts and correct MOV positioning.STP Lesson:LOT 201.11Objective Number:2009GIVEN a plant or system condition, PREDICT the operation of the Containment Spray System.Attached Reference

Reference:

LOT 201.11 Lesson Plan Containment Spray

Attachment:

Source:NewDistractor JustificationA:INCORRECT: Plausible because the CS Actuation Signal does start the CS Pumps but in this case since the ESF D/G just energized the BUS there is a 15 second time delay before the CS Pump gets a start signal.B:INCORRECT: Plausible because the CS Actuation Signal does start the CS Pumps but in this case since the ESF D/G just energized the BUS there is a 15 second time delay before the CS Pump gets a start signal. Also many pumps in the plant have their discharge valves start to open when a pump starts but in this case the CS Pump discharge valves start to open on the CS Actuation signal.C:INCORRECT: Plausible because many pumps in the plant have their discharge valves start to open when a pump starts but in this case the CS Pump discharge valves start to open on the CS Actuation signal.D:CORRECT: Per the given conditions the CS Pump would start 15 seconds after the ESF BUS is energized but the CS Discharge valve would start to open on the CS Actuation signal. (1 second is for load center breakers to close after the DG output breaker closes)Question Level:HQuestion Difficulty3Justification:The candidate must have a knowledge of the conditions required to actuate both the spray pumps and discharge valves. This knowledge must then be applied to the conditions given to determine the correct response.RO Importance:4.3NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:244410CFR

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55.41(b)(7)Page 98 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAM The Component Cooling Water (CCW) normal makeup valve, CC-LV-4501, should be _____(1)_____ and the CCW Surge Tank low level alarm should be _____(2)_____ given the indicated CCW Surge Tank level. A. (1) Closed (2) Lit B. (1) Closed (2) Extinguished C. (1) Open (2) Lit D. (1) Open (2) Extinguished Answer:C (1) Open - (2) LitExam Bank No.:2465Last used on an NRC exam:NeverRO Sequence Number:50Page 99 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 026 AA1.05Tier:1Group/Category:1Ability to operate and/or monitor the following as they apply to the Loss of Component Cooling Water:The CCWS surge tank, including level control and level alarms, and radiation alarm.STP Lesson:LOT 201.12Objective Number:80198DESCRIBE the Instrumentation and Controls available to monitor and operate the CCW System.Attached Reference

Reference:

LOT 201.12 Lesson on CCW and 0POP04-CC-0001, Addendum 1

Attachment:

Source:NewDistractor JustificationA:INCORRECT: All distractors are plausible given the fact that the CCW Surge Tank high level alarm comes in at > 83.3% which is outside of the green area of the indication. In otherwords the fact that the CCW Surge tank level indication is lower than the green area of the indication is not in itself an indication of when level controls and alarms are actuated. The student must have knowledge of these setpoints.B:INCORRECT: All distractors are plausible given the fact that the CCW Surge Tank high level alarm comes in at > 83.3% which is outside of the green area of the indication. In otherwords the fact that the CCW Surge tank level indication is lower than the green area of the indication is not in itself an indication of when level controls and alarms are actuated. The student must have knowledge of these setpoints.C:CORRECT: The indicated CCW Surge Tank level is 65%. The normal makeup valve opens at < 68.75% and the low level alarm comes in at < 66.7%.D:INCORRECT: All distractors are plausible given the fact that the CCW Surge Tank high level alarm comes in at > 83.3% which is outside of the green area of the indication. In otherwords the fact that the CCW Surge tank level indication is lower than the green area of the indication is not in itself an indication of when level controls and alarms are actuated. The student must have knowledge of these setpoints.Question Level:HQuestion Difficulty2Justification:The student must be able to anlayze the given CCW Surge Tank level indication to determine the correct status of level controls and level alarms. NOTE: CCW radiation monitor does not have any ties to CCW surge tank level.RO Importance:3.1NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:246510CFR

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55.41(b)(7)Page 100 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit was at 100% power when a control rod dropped into the core to the 'Rod Bottom' position. The Power Range NIs now read the following: NI-0041B 98% NI-0042B 98% NI-0043B 97% NI-0044B 98% The control rod that dropped into the core was at the _____(1)_____ of the core. AND As the dropped control rod is being recovered its Differential Rod Worth will be _____(2)_____ as it approaches the mid-plane of the core. A. (1) center (2) less B. (1) edge (2) less C. (1) center (2) more D. (1) edge (2) more Answer:C(1) center - (2) moreExam Bank No.:2467Last used on an NRC exam:NeverRO Sequence Number:51Page 101 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 003 AK1.19Tier:1Group/Category:2Knowledge of the operational implications of the following concepts as they apply to Dropped Control Rods:

Differential Rod Worth.STP Lesson:LOT 101.22Objective Number:N99697Define control rod worth, differential rod worth and intergral rod worth.Attached Reference

Reference:

LOT 101.22 Lesson Handout on Control Rod Worths

Attachment:

Source:NewDistractor JustificationA:INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.B:INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.C:CORRECT: With a single control rod dropping into the center of the core the power range NIs wouldequally lower. As the control rod is being recovered the DRW would be more as it approched the mid-plane of the core.D:INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.Question Level:HQuestion Difficulty3Justification:The student must analyze the given NI power range readings to determine what part of the core the control rod dropped and have fundamental knowledge of control rod worths for control rods at different locations/heights in the core.RO Importance:2.8NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:246710CFR

Reference:

55.41(b)(1)Page 102 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring a fast load reduction from 100% power the Control Rods were placed in MANUAL and the following DRPI indication was observed: Control Rod(s) _____(1)_____ is(are) misaligned. The crew will _____(2)_____. A. (1) D4 and M12 (2) place the Unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. (1) D4 and M12 (2) trip the Reactor C. (1) D4 (2) place the Unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. (1) D4 (2) trip the Reactor Answer:A (1) D4 and M12 - (2) place the Unit in Mode 3 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sExam Bank No.:2468Last used on an NRC exam:NeverRO Sequence Number:52Page 103 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 005 AA1.05Tier:1Group/Category:2Ability to operate and/or monitor the following as they apply to the Inoperable/Stuck Control Rod:RPISTP Lesson:LOT 201.19Objective Number:93001Given a plant or system condition, PREDICT the operation of the Rod Position Indication System.Attached Reference

Reference:

LOT 201.19 Lesson Plan and 0POP04-RS-0001

Attachment:

Source:NewDistractor JustificationA:CORRECT: Control Rods D4 and M12 are both misaligned from the other control rods by more than 12 steps. Per 0POP04-RS-0001, the Unit will be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.B:INCORRECT: Plausible because for 2 dropped control rods the crew would trip the reactor.C:INCORRECT: Plausible because the indicated positions of Control Rods D4 and M12 are different and the student must have the knowledge of how many steps different from other control rods would be considered a misalignment.D:INCORRECT: Plausible because the indicated positions of Control Rods D4 and M12 are different and the student must have the knowledge of how many steps different from other control rods would be considered a misalignment. Also the student must have knowledge of procedural requirements.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition to determine which Control rods are misaligned and the required procedural action to take.RO Importance:3.4NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:246810CFR

Reference:

55.41(b)()Page 104 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power with the following conditions: CVCS Cation Bed 1A was used 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago for RCS lithium control. Subsequently: An Alert Alarm is received on Failed Fuel Monitor RT-8039. The Crew is responding per 0POP04-RC-0001, High Reactor Coolant System Activity. Chemistry has requested that Cation Bed 1A be placed back in service to help control RCS activity. Failed Fuel Monitor RT-8039 readings can be validated by _____(1)_____. AND A consequence of placing Cation Bed 1A in service is that _____(2)_____ would lower. (1) (2) A. obtaining a current RCS sample RCS pH B. obtaining a current RCS sample boron concentration C. trending the GWPS inlet Radiation Monitor RCS pH D. trending the GWPS inlet Radiation Monitor boron concentration Answer:A obtaining a current RCS sample - RCS pHExam Bank No.:2470Last used on an NRC exam:NeverRO Sequence Number:53Page 105 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 076 G2.2.44Tier:1Group/Category:2High Reactor Coolant Activity:Ability to interpret control room indications to verify the status of a system and understand how operator actions and directives affect plant and system conditions.STP Lesson:LOT 505.01Objective Number:92109GIVEN a plant condition, DESCRIBE and/or INTERPRET the reqirements and/or limits of a precaution or step of a referenced procedure.Attached Reference

Reference:

LOT 505.01 Lesson Plan and 0POP04-RC-0001, High RCS Activity

Attachment:

Source:NewDistractor JustificationA:CORRECT: Per 0POP04-RC-0001 an alternate method for verifing High RCS activity when RT-8039 is trending upward is to get a current RCS sample. A caution in the procedure also warns the operator that placing a CVCS Cation bed in service will lower RCS lithium thus lowering RCS pH and possibly cause a Crud Burst.B:INCORRECT: Plausible because placing a CVCS Cation Bed in service can lower boron concentration, however, it was stated in the conditions of the question that the CVCS Cation Bed was placed in service 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago which indicates that its boron concentration is equal to the RCS.C:INCORRECT: Plausible because in 0POP04-RC-0001 GWPS is monitored because VCT purge flow rate is raised but GWPS is not monitored from a stand point of verifing RCS activity.D:INCORRECT: Plausible because in 0POP04-RC-0001 GWPS is monitored because VCT purge flow rate is raised but GWPS is not monitored from a stand point of verifing RCS activity. Also, placing a CVCS Cation Bed in service can lower boron concentration, however, it was stated in the conditions of the question that the CVCS Cation Bed was placed in service 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago which indicates that its boron concentration is equal to the RCS.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given conditions to determine what affect an action will have on the RCS and have knowledge of alternate ways to verify high RCS activity.RO Importance:4.2NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:247010CFR

Reference:

55.41(b)(7)Page 106 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following 480 Volt Motor Control Centers (MCC) supplies power to Boric Acid Transfer Pump 1B? A. MCC E1A4 B. MCC E1C4 C. MCC 1G8 D. MCC 1K3 Answer:AMCC E1A4Exam Bank No.:2501Last used on an NRC exam:NeverRO Sequence Number:54Page 107 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:004 K2.01Tier:2Group/Category:1Knowledge of bus power supplies to the following:Boric acid makeup pumps.STP Lesson:LOT 201.07Objective Number:91054DESCRIBE the boric acid pumps to include: A. Function and locartion B. Pump flow and pressure capacity C. Electrical power supply D. Pump controls and meaning of associated alarms.Attached Reference

Reference:

LOT 201.07 Lesson Plan on Reactor Makeup.

Attachment:

Source:NewDistractor JustificationA:CORRECT: The Boric Acid Transfer Pumps power supply comes from Class 1E 480 Volt MCCs. BAT Pump 1B is powered from MCC E1A4. (NOTE: Same Train of power as CCP 1B)B:INCORRECT: Plausible because the power for the Boric Acid pumps is not Train specific and MCC E1C4 is the power for BAT Pump 1A.C:INCORRECT: Plausible because some safety significant equipment is powered from NC MCC 1G8 like the Positive Displacement Charging Pump. Plus MCC 1G8 is backed by the TSC DG.D:INCORRECT: Plausible because some some safety support equipment is powered from NC MCC 1K3 like the RWST Purification Pump. Plus MCC 1K3 is physically close to the Boric Acid Transfer Pumps.Question Level:FQuestion Difficulty2Justification:The student has to have fundamental knowledge of power supplies for safety significant and safety related equipment.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:250110CFR

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55.41(b)(7)Page 108 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring a plant heatup, which of the following represents the HIGHEST ALLOWABLE flow rate through one CVCS demineralizer? A. 350 gpm B. 300 gpm C. 250 gpm D. 200 gpm Answer:C250 gpmExam Bank No.:2473Last used on an NRC exam:NeverRO Sequence Number:55Page 109 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:004 A4.14Tier:2Group/Category:1Chemical and Volume Control:Ability to manually operate and/or monitor in the control room:

Ion exchangers and demineralizers.STP Lesson:LOT 201.06Objective Number:70033STATE the maximum specified purification flow rates through the CVCS System Demineralizers including the reasons for the limit.Attached Reference

Reference:

LOT 201.06 Lesson Plan for CVCS and 0POP03-ZG-0001, Plant Heatup

Attachment:

Source:NewDistractor JustificationA:INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.B:INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.C:CORRECT: The limit on flow rate through one CVCS demineralizer during a plant heatup is 250 gpm.D:INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.Question Level:FQuestion Difficulty2Justification:The student must have knowledge of CVCS demin flow requirements.RO Importance:2.8NRC Reference Req'd

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Modified fromExam Bank No.:247310CFR

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55.41(b)(7)Page 110 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit has been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and mid-loop operations have just commenced with the following conditions: RHR Pumps A and C are in service. RHR pump B is in standby. RCS Hot Leg level is +9 inches. Subsequently: RHR Pump A receives alarm 'RHR PUMP A CURRENT LO'. The alarm came in and out 2 times and is now locked in. RCS Hot Leg level is observed at +4 inches. (1) With the conditions given, the 'RHR PUMP A CURRENT LO' alarm is indication of RHR Pump _____(1)_____. AND (2) The Crew should _____(2)_____. (1) (2) A. RUNOUT secure RHR Pump A & C and refill RCS using a LHSI Pump B. RUNOUT lower RHR Pump A flow to less than 1000 gpm C. VORTEXING secure RHR Pump A & C and refill RCS using a LHSI Pump D. VORTEXING lower RHR Pump A flow to less than 1000 gpm Answer:C VORTEXING - secure RHR Pump A & C and refill RCS using a LHSI PumpExam Bank No.:2503Last used on an NRC exam:NeverRO Sequence Number:56Page 111 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:005 A2.01Tier:2Group/Category:1Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, or mitigate the consequences of those malfunctions or operations:Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation.STP Lesson:LOT 201.09Objective Number:4245GIVEN a plant or system condition, PREDICT the operation of the Residual Heat Removal System.Attached Reference

Reference:

LOT 201.09 Lesson Plan on RHR and 0POP04-RH-0001, Loss of Residual Heat Removal

Attachment:

Source:NewDistractor JustificationA:INCORRECT: Plausible because pump runout is a condition where pump current is effected but for pump runout current would be high.B:INCORRECT: Plausible because pump runout is a condition where pump current is effected but for pump runout current would be high. Also plausible because lowering RHR Pump flow would be helpful but lowering flow less than 1000 gpm would be an incorrect action because the pump trips at925 gpm. The procedure states to lower flow for some conditions to between 1000 and 1500 gpm.C:CORRECT: With the given conditions the CURRENT LO alarm would indicate vortexing. 0POP04-RH-0001 will have the crew stop all running RHR pumps and refill the RCS.D:INCORRECT: Plausible because lowering RHR Pump flow would be helpful but lowering flow less than 1000 gpm would be an incorrect action because the pump trips at 925 gpm. The procedure states to lower flow for some conditions to between 1000 and 1500 gpm.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition to determine the correct procedural action and have knowledge of the RHR Pump lcurrent low alarm.RO Importance:2.7NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:250310CFR

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55.41(b)(10)Page 112 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring a Large Break LOCA, if the Safety Injection Accumulators failed to inject, what would be the effect on water being used to cool the core? Without the Safety Injection Accumulators injecting, the ECCS water injected to cool the core would have a _____(1)_____ boron concentration and a _____(2)_____ pH. A. (1) higher (2) lower B. (1) lower (2) lower C. (1) higher (2) higher D. (1) lower (2) higher Answer:D (1) lower - (2) higherExam Bank No.:2475Last used on an NRC exam:NeverRO Sequence Number:57Page 113 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:006 K6.02Tier:2Group/Category:1Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:Core flood tanks (accumulators).STP Lesson:LOT 201.10Objective Number:29419GIVEN a plant condition, PREDICT the operation of the ECCS to include automatic actuations, interlocks and/or trips.Attached Reference

Reference:

LOT 201.10 Lesson Plan on ECCS and LOT 201.11 Lesson Plan on Containment Spray

Attachment:

Source:NewDistractor JustificationA:INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.B:INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.C:INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.D:CORRECT: Without the water from the accumulators injecting into the RCS during a LB LOCA it would make the water being used for cooling the core at a lower boron concentration and a higher PH.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of the contents of the SI Accumulators and how they would effect system operation during a LB LOCA.RO Importance:3.4NRC Reference Req'd

Attachment:

Modified fromExam Bank No.:247510CFR

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55.41(b)(7)Page 114 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMDuring surveillance testing of the Pressurizer Power Operated Relief Valves (PORV) which alarm listed should be monitored per 0PSP03-RC-0010, Pressurizer PORV Operability Test? A. RCDT PRESS HI B. PRT PRESS HI C. RCDT LEVEL HI-HI/LO-LO D. PRT LEVEL HI/LO Answer:BPRT PRESS HIExam Bank No.:2476Last used on an NRC exam:NeverRO Sequence Number:58Page 115 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:010 A3.01Tier:2Group/Category:1Ability to monitor automatic operation of the PZR PCS, including:PRT temperature and pressure during PORV testing.STP Lesson:LOT 201.14Objective Number:4460STATE the interfaces between the pressurizer pressure and level control systems and the following: A. CVCS B. RCS C. RPS D. Pressure relief tank.Attached Reference

Reference:

LOT 201.14 Lesson Plan on PZR level and pressure control and 0PSP03-RC-0010, Pressurizer PORV Operability Test

Attachment:

Source:NewDistractor JustificationA:INCORRECT: Plausible because there is a connection to the RCDT from the PRT but it comes from the PZR PORV MOV Block valve packing leakoff line and it is not likely to pass steam or water unless the valve is required to be closed and that would only happen if the PZR PORV were to stick open.B:CORRECT: 0PSP03-RC-0010 has the operator monitor PRT PRESS HI because when the PZR PORV opens during testing a steam mass could quickly raise pressure in the PRT.C:INCORRECT: Plausible because there is a connection to the RCDT from the PRT but it comes from the PZR PORV MOV Block valve packing leakoff line and it is not likely to pass steam or water unless the valve is required to be closed and that would only happen if the PZR PORV were to stick open.D:INCORRECT: Plausible because PRT level would rise but not very much because the PZR PORV would not stay open long and the volume of water in the steam mass would be minimal.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of notes in the surveillance procedure for the PZR PORVs.RO Importance:3.0NRC Reference Req'd

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Modified fromExam Bank No.:247610CFR

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55.41(b)(7)Page 116 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Reactor will trip at a specified Reactor Coolant Pump frequency setpoint. If Reactor Coolant Pump frequency became an issue and the Reactor trip setpoint had drifted _____(1)_____ the Reactor would be more likely to exceed _____(2)_____ limits. A. (1) high (2) DNB B. (1) high (2) Fuel Integrity C. (1) low (2) DNB D. (1) low (2) Fuel Integrity Answer:C (1) low - (2) DNBExam Bank No.:2477Last used on an NRC exam:NeverRO Sequence Number:59Page 117 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:012 A1.01Tier:2Group/Category:1Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including:

Trip setpoint adjustment.STP Lesson:LOT 202.20Objective Number:3832DESCRIBE the reactor protection system control and permissive interlocks including inputs, setpoints, coincidences, and functions.Attached Reference

Reference:

LOT 201.20 Lesson Plan on SSPS and Technical Specifications bases for the reasons for the different Reactor Trips.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because it is reasonalble to believe the Reactor Trip setpoint could drift high and cause the Reactor to be more likely to exceed DNB limits.B:INCORRECT: Plausible because it is reasonalble to believe the Reactor Trip setpoint could drift high and plausible because Fuel Integrity is a concern from the reactor but Fule Integrity is protected by the OPDT Reactor Trip.C:CORRECT: The Reactor will trip on low Reactor Coolant Pump frequency. If the setpoint were to drift low then the Reactor would be more likely to exceed DNB limits.D:INCORRECT: Plausible because Fuel Integrity is a concern from the reactor but Fule Integrity is protected by the OPDT Reactor Trip.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of Reactor Trip setpoints and the reason for the trips.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:247710CFR

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55.41(b)(7)Page 118 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 28% power when the following occurs: RCP C develops a sheared shaft. Prior to the Reactor Operator tripping the Reactor, what automatic response will occur associated with the Main Feedwater System? A. SG C MFRV will start to open to restore level in SG C. B. SU SGFP #14 will start and restore level in SG C. C. ALL MFRVs will start to open to restore level in ALL SGs. D. Main Feedwater will isolate and Auxiliary Feedwater will restore level in ALL SGs. Answer:ASG C MFRV will start to open to restore level in SG C.Exam Bank No.:2478Last used on an NRC exam:NeverRO Sequence Number:60Page 119 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:059 K4.13Tier:2Group/Category:1Knowledge of the MFW design feature(s) and/or interlocks which provide for the following:Feedwater fill of S/G upon loss of RCP.STP Lesson:LOT 202.13Objective Number:20359GIVEN plant/system conditions, PREDICT the operation of the Feedwater System.Attached Reference

Reference:

LOT 202.13 Lesson Plan on Main Feedwater

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Source:NewDistractor JustificationA:CORRECT: A loss of RCP flow will cause a loss of load in the affected loop. SG level will quickly lower and the SG MFRV will start to oen to try to restore level. An automatic reactor trip will not occur because reactor power is below 40% power. (P-8)B:INCORRECT: Plausible because the SU SGFP can supply extra feedwater but it will only auto start on a loss of a SGFPT.C:INCORRECT: Plausible because if the student believed that a sheared shaft on 1 RCP could effect level in ALL of the SGs then ALL MFRVs would open.D:INCORRECT: Plausible because if the student believed that the reactor auto tripped for this event then MFW would isolate and AFW would auto start to feed all SGs.Question Level:HQuestion Difficulty3Justification:The student must analyze the given condition to determine how MFW will respond.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:247810CFR

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55.41(b)(7)Page 120 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following conditions will cause an automatic shutdown of Auxiliary Feedwater Pump #14? A. Shaft speed is at 4250 rpm. B. The 'LOW OIL PRESSURE' red light is illuminated. C. The 'MAIN OIL PUMP PRESSURE NORMAL' amber light is extinguished. D. The difference in temperature between low point steam line drains and Tsat for the current Psat in 'D' Main Steam Line is 35ºF. Answer:AShaft speed is at 4250 rpm.Exam Bank No.:2479Last used on an NRC exam:NeverRO Sequence Number:61Page 121 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:061 K4.07Tier:2Group/Category:1Knowledge of AFW design Feature(s) and/or interlock(s) which provide for the following:Turbine trip including overspeed.STP Lesson:LOT 202.28Objective Number:43847DISCUSS the following elements associated wht the AFW turbine driven Pump: A. Signals that will trip the trip and throttle valve B. How to reset the trip and throttle valve C. Lubrication system D. How to determine status E. Exhaust and drain paths F. Reason for warming up steam lines before startup G. Transfer switch operation to transfer operation to ASP H. Associated Alarms.Attached Reference

Reference:

LOT 202.28 Lesson Plan on AFW and 0POP02-AF-0001, Auxiliary Feedwater.

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Source:NewDistractor JustificationA:CORRECT: The overspeed condition is the only design feature that will automatically shutdown AFWP #14.B:INCORRECT: Plausible because this having the 'low oil pressure' red light on is a reason to shutdown AFWP #14.C:INCORRECT: Plausible because having the 'main oil pump pressure normal' amber light extinguished is a reason to shutdown AFWP #14.D:INCORRECT: Plausible because the differential temperature between the D Main Steam line and the steam line drains has to less than 25 degrees F but this only applies to Operability and starting AFWP #14.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of the AFWP #14 design feature for overspeed.RO Importance:3.1NRC Reference Req'd

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Modified fromExam Bank No.:247910CFR

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55.41(b)(7)Page 122 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA Large Break LOCA has occurred with the following conditions: Containment pressure is 17 psig and slowly lowering. The crew has just transitioned to 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant. (1) Which Control Room Panel can the RO operate the Containment Spray (CS) Pumps? AND (2) How many CS Pumps should be running? A. (1) CP-001 (2) 3 CS Pumps B. (1) CP-001 (2) 2 CS Pumps C. (1) CP-002 (2) 3 CS Pumps D. (1) CP-002 (2) 2 CS Pumps Answer:D(1) CP-002 - (2) 2 CS PumpsExam Bank No.:2480Last used on an NRC exam:NeverRO Sequence Number:62Page 123 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:103 G2.1.31Tier:2Group/Category:1Containment:Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.STP Lesson:LOT 201.11Objective Number:2009GIVEN a plant or system condition, PREDICT the operation of the Containment Spray System.Attached Reference

Reference:

LOT 201.11 Lesson Plan on Containment Spray and 0POP05-EO-EO00, Reactor Trip or Safety Injection, CIP.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because CP-001 also has safety related equipment that would need to be operated under these conditions. Also, with the given containment pressure it would be reasonable to believe that all CS Pumps should remain running.B:INCORRECT: Plausible because CP-001 also has safety related equipment that would need to be operated under these conditions.C:INCORRECT: Plausible because with the given containment pressure it would be reasonable to believe that all CS Pumps should remain running.D:CORRECT: CS Pumps Controls are located on CP-002 and the CIP for 0POP05-EO-EO00, Reactor Trip or Safety Injection, states that for RWST conservation a 3rd CS Pump should be stopped.Question Level:HQuestion Difficulty2Justification:The student must analyze the given conditions to determine the correct system lineup and have knowledge of where to operate the CS Pumps.RO Importance:4.6NRC Reference Req'd

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Modified fromExam Bank No.:248010CFR

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55.41(b)(10)Page 124 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 75% power with the following conditions: Control Bank 'D' is at 230 steps. Rod Control is in Auto. Subsequently: Power Range Channel N42 fails high. What effect will this failure have on the Rod Control system? Control Bank 'D' will step- A. OUT and continue to step out until stopped by C-11 (248 steps). B. IN and continue to step in unless the RO places Rod Control in Manual. C. OUT ONLY as N42 failed high and then step back in to lower Tavg back to Tref. D. IN ONLY as N42 failed high and then stop; C-2 (PR high rod stop) will prevent rods from stepping back out. Answer:DIN ONLY as N42 failed high and then stop; C-2 (PR high rod stop) will prevent rods from stepping back out.Exam Bank No.:2481Last used on an NRC exam:NeverRO Sequence Number:63Page 125 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:015 K3.06Tier:2Group/Category:2Knowledge of the effect that a loss or malfunction of the NIS will have on the following:Reactor regulating system.STP Lesson:LOT 201.18Objective Number:2410GIVEN a change in plant or system conditions, PREDICT how the rod control system will respond.Attached Reference

Reference:

LOT 201.18 Lesson Plan on Rod Control

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Source:NewDistractor JustificationA:INCORRECT: Plausible because if the student believed that control rods would continue to step out then they would stop at 248 steps.B:INCORRECT: Plausible because if the student believed that Control Rods would continue to step in then there would be not immediate interlocks to stop the control rods and the RO would place control rods in Manual.C:INCORRECT: Plausible because if the student believed that the control rods would only step out during the time of the failed channel then rods would try to step back in to recover any temperature deviation between Tavg and Tref.D:CORRECT: Control Rod Bank 'D' will step in only as N42 failed high to try to lower what the system believes is a high reactor power. When there is no change between Reator power and turbine power then the control rods will stop. Since it only takes one PR NI channel to bring in the PR high Rod Stop (C-2) the control rods will not try to step back out to restore Tavg to Tref.Question Level:HQuestion Difficulty3Justification:The student must be able to anlayze the given condition to determine the response of Rod Control.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:248110CFR

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55.41(b)(7)Page 126 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 75% Reactor power with the following conditions: PZR level control is in automatic controlling PZR level on program. LOOP 'A' Tavg is 586ºF. LOOP 'B' Tavg is 586ºF. LOOP 'C' Tavg is 584ºF. LOOP 'D' Tavg is 585ºF. Subsequently: LOOP 'B' Tavg Channel fails high. Which LOOP Tavg is now determining PZR program level? A. LOOP 'A' Tavg. B. LOOP 'B' Tavg. C. LOOP 'C' Tavg. D. LOOP 'D' Tavg. Answer:BLOOP 'B' Tavg.Exam Bank No.:2482Last used on an NRC exam:NeverRO Sequence Number:64Page 127 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:016 A3.01Tier:2Group/Category:2Ability to monitor automatic operation of the NNIS, including:Automatic selection of NNIS inputs to control systems.STP Lesson:LOT 201.14Objective Number:92779GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control system.Attached Reference

Reference:

LOT 201.14 Lesson Plan on PZR pressure and level control systems.

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Source:NewDistractor JustificationA:INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.B:CORRECT: PZR program level setpoint is automatically set off of the highest channel of RCS LOOP Tavg.C:INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.D:INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.Question Level:HQuestion Difficulty2Justification:The student must analyze the given condition and have knowledge of the PZR level control system to determine which channel would set PZR program level setpoint.RO Importance:2.9NRC Reference Req'd

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Modified fromExam Bank No.:248210CFR

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55.41(b)(7)Page 128 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAM(1) What temperature limit would be exceeded to enter an Inadequate Core Cooling Safety Function? AND (2) Which indication would be monitored and controlled? A. (1) 1200ºF (2) Core Exit Thermocouples B. (1) 1200ºF (2) Wide Range Hot Leg Temperature C. (1) 2200ºF (2) Core Exit Thermocouples D. (1) 2200ºF (2) Wide Range Hot Leg Temperature Answer:A (1) 1200ºF - (2) Core Exit ThermocouplesExam Bank No.:2483Last used on an NRC exam:NeverRO Sequence Number:65Page 129 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:017 A1.01Tier:2Group/Category:2Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ITM system controls including:

Core exit thermocouples.STP Lesson:LOT 202.44Objective Number:7667GIVEN a change in plant or system condition, EXPLAIN the operation and indications of the QDPS System.Attached Reference

Reference:

LOT 202.44 Lesson Plan on QDPS and 0POP05-EO-FO02, Core Cooling Critical Safety Function Status Tree.

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Source:NewDistractor JustificationA:CORRECT: Core Exit Thermocouples reading 1200 degrees F would require entry into the Inadequate Core Cooling Safety Function.B:INCORRECT: Plausible because RCS Wide Range temperatures are monitored for entry into Safety Function procedures (Integrity).C:INCORRECT: Plausible because 2200 degrees F is an important temperature to reactor core safety but it is associated with cladding temperature.D:INCORRECT: Plausible because 2200 degrees F is an important temperature to reactor core safety but it is associated with cladding temperature. Also, RCS Wide Range temperatures are monitored for entry into Safety Function procedures (Integrity).Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of the parameters monitored and controlled to prevent entry into a Core Cooling Safety Function.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:248310CFR

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55.41(b)(7)Page 130 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power with the following condition: A recent Chemistry sample of RCB Atmosphere Noble Gas is reading 6.50E-4 µCi/cc resulting in a purge permit notification level of 9.75E-4 µCi/cc. (1) What are the consequences of performing a containment purge under this condition? AND (2) What procedural actions should be performed prior to starting the purge? A. (1) An ALERT alarm ONLY will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors. (2) Raise the isolation setpoint on RT-8012 and RT-8013. B. (1) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors. (2) Raise the isolation setpoint on RT-8012 and RT-8013. C. (1) An ALERT alarm ONLY will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors. (2) Verify Containment Ventilation Actuation is BLOCKED. D. (1) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors. (2) Verify Containment Ventilation Actuation is BLOCKED. Answer:B (1) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.(2) Raise the isolation setpoint on RT-8012 and RT-8013.Exam Bank No.:2484Last used on an NRC exam:NeverRO Sequence Number:66Page 131 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:029 A2.04Tier:2Group/Category:2Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System, and (b) based on those predictions, use procedures to correct, or mitigate the consequences of those malfunctions or operations:Health physics sampling of containment atmosphere.STP Lesson:LOT 202.33Objective Number:97097Discuss 0POP02-HC-0003 including: A. Purpose and scope B. Precautions C. Notes.Attached Reference

Reference:

LOT 202.33 Lesson Plan on RCB HVAC and 0POP02-HC-0003, Supplemental Containment Purge.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the student has to be aware of the isolation alarm (High Alarm) set point that would cause an ESF Containment Ventalation Isolation (CVI).B:CORRECT: RT-8011 reading is above the level which will cause an isolation alarm (High Alarm) on RT-8012 and RT-8013 (5.00E-4). The procedural action is to raise the RT-8012 and RT-8013 isolation alarm set point (High Alarm) to prevent an ESF actuation prior to performing the purge.C:INCORRECT: Plausible because the student has to be aware of the isolation alarm (High Alarm) set point that would cause an ESF Containment Ventalation Isolation (CVI). Also, you can block CVI but the procedural action to verify that CVI is blocked is for Mode 5 and 6 and Defueled only.D:INCORRECT: Plausible because you can block CVI but the procedural action to verify that CVI is blocked is for Mode 5 and 6 and Defueled only.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition to determine the correct procedural actions and have knowledge of the ESF actuation setpoint for CVI. NOTE: Chemistry performs samples of contaiment at STP. NOT Health Physics.RO Importance:2.5NRC Reference Req'd

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Modified fromExam Bank No.:248410CFR

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55.41(b)(10)Page 132 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power when the following alarm annunciates: 'RWST LEVEL HI/LO' on CP-001 'ACC TK 1A(2A) PRESS HI/LO' on CP-001 'CNTMT PRESS HI/LO' on CP-002 'PZR DNBR PRESS LOW' on CP-004 Considering each one separately, which alarm based on its setpoints would require a one hour entry into Technical Specifications? A. 'RWST LEVEL HI/LO' on CP-001 B. 'ACC TK 1A(2A) PRESS HI/LO' on CP-001 C. 'CNTMT PRESS HI/LO' on CP-002 D. 'PZR DNBR PRESS LOW' on CP-004 Answer:C 'CNTMT PRESS HI/LO' on CP-002Exam Bank No.:2486Last used on an NRC exam:NeverRO Sequence Number:67Page 133 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.2.39Tier:3Group/Category:2Knowledge of less than or equal to one hour Technical Specification action statements for systems.STP Lesson:LOT 503.01Objective Number:80056GIVEN a system condition, DETERMINE the applicable Technical Specification and/or the Technical Requirements Manual (TRM) and APPLY the specification(s).Attached Reference

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LOT 503.01 Lesson Plan On TSs, especially those that require the unit to take action wihtin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the alarm is yellow indicating a higher priority, is related to safety systems and TSs apply but not based just on alarm setpoints.B:INCORRECT: Plausible because the alarm is yellow indicating a higher priority, is related to safety systems and TSs apply but not based just on alarm setpoints.C:CORRECT: Containment Pressure HI/LO alarm is the only one listed that has setpoints at the TS limits. (for hi or lo limit) With this alarm in containment pressure has to be restored to within limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or a shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is required. With the other alarms the setpoints for the alarms are within the limits of the TS for that parameter.D:INCORRECT: Plausible because the DNBR alarm does require a TS entry but it is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> TS.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of alarms that would cause a direct entry in to Technical Specifications.RO Importance:3.9NRC Reference Req'd

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Modified fromExam Bank No.:248610CFR

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55.41(b)(10)Page 134 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMIn regards to 480 V Load Center 12L; (1) Which Unit(s) supplies power to the Load Center? AND (2) Which Unit(s) has a trouble alarm for the Load Center? (1) (2) A. Unit 1 ONLY Both Unit 1 and Unit 2 B. Both Unit 1 and Unit 2 Both Unit 1 and Unit 2 C. Unit 1 ONLY Unit 1 ONLY D. Both Unit 1 and Unit 2 Unit 1 ONLY Answer:AUnit 1 ONLY - Both Unit 1 and Unit 2Exam Bank No.:2487Last used on an NRC exam:NeverRO Sequence Number:68Page 135 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.2.3Tier:3Group/Category:2Knowledge of the design, procedural, and operational differences between units.STP Lesson:LOT 203.21Objective Number:N0056STATE the design differences between Unit 1 and Unit 2.Attached Reference

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LOT 203.21 Lesson Plan on Unit Differences.

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Source:NewDistractor JustificationA:CORRECT: For LC 12L only Unit 1 supplies power to this single ended load center but both units have alarms for the bus.B:INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.C:INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.D:INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.Question Level:FQuestion Difficulty2Justification:the student must have fundamental knowledge of Unit differences.RO Importance:3.8NRC Reference Req'd

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Modified fromExam Bank No.:248710CFR

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55.41(b)(7)Page 136 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMThe Unit is at 100% power. Prior to a containment entry the crew should start _____(1)_____ to lower containment iodine levels. AND The crew should start _____(2)_____ . A. (1) Normal Containment Purge (2) ONE supply and ONE exhaust fan B. (1) Normal Containment Purge (2) TWO supply and TWO exhaust fans C. (1) Containment Carbon Filter Unit (2) ONE fan per train D. (1) Containment Carbon Filter Unit (2) TWO fans per train Answer:C(1) Containment Carbon Filter Unit(2) ONE fan per trainExam Bank No.:2490Last used on an NRC exam:NeverRO Sequence Number:69Page 137 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:027 A4.03Tier:2Group/Category:2Ability to manually operate and/or monitor in the control room:CIRS fans.STP Lesson:LOT 202.33Objective Number:80242DESCRIBE the instrumentation and controls available to monitor and operate the RCB-HVAC systems.Attached Reference

Reference:

LOT 202.33 Lesson Plan on RCB HVAC and 0POP02-HC-0001, Containment HVAC

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Source:NewDistractor JustificationA:INCORRECT: Plausible because a purge can be used to lower iodine levels in containment prior to a containment entry but in Modes 1 to 4 Supplemental Purge must be used not Normal purge.B:INCORRECT: Plausible because a purge can be used to lower iodine levels in containment prior to a containment entry but in Modes 1 to 4 Supplemental Purge must be used not Normal purge. Also, if using the Normal Purge only one supply and one exhaust fan are started because one set has 100% capacity.C:CORRECT: Containment Carbon Unit Filters can be used when requested by Health Physics to lower iodine levels in containment prior to a containment entry. Only one fan in one unit is started.D:INCORRECT: Plausible because there are two Containment Carbon Filter Units that are 100% capacity each and the fans associated with the Units are also rated at 100% capacity and only require one to be started.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of the procedural requirements for starting the Containment Carbon Units and when they would be needed.RO Importance:3.3NRC Reference Req'd

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Modified fromExam Bank No.:249010CFR

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55.41(b)(7)Page 138 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich Area Radiation Monitor has a 'Loss of Communications?' A. RT-8053 B. RT-8088 C. RT-8092 D. RT-8097 Answer:DRT-8097Exam Bank No.:2492Last used on an NRC exam:NeverRO Sequence Number:70Page 139 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:072 G2.1.19Tier:2Group/Category:2Ability to use plant computers to evaluate system or component status.STP Lesson:LOT 202.41Objective Number:68793DESCRIBE the meaning of colors on the RM-11 display.Attached Reference

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LOT 202.41 Lesson Plan on the Radiation Monitoring System.

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Source:NewDistractor JustificationA:INCORRECT: All distractors are plausible because they are in some type of alarm status.B:INCORRECT: All distractors are plausible because they are in some type of alarm status.C:INCORRECT: All distractors are plausible because they are in some type of alarm status.D:CORRECT: For this case RT-8097 is showing a loss of communication (magenta).Question Level:HQuestion Difficulty2Justification:The student must analyze the data on the RM-11 computer screen to determine which Area Radiation Monitor has the highest priority.RO Importance:3.9NRC Reference Req'd

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Modified fromExam Bank No.:249210CFR

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55.41(b)()Page 140 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMIn accordance with 0POP04-IA-0001, Loss of Instrument Air, which of the following is the FIRST Instrument Air pressure reached that would require a manual Reactor Trip? A. 90 psig B. 80 psig C. 70 psig D. 60 psig Answer:D60Exam Bank No.:2116Last used on an NRC exam:2011RO Sequence Number:71Page 141 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 065 AA2.06Tier:1Group/Category:1Ability to determine and interpret the following as they apply to the Loss of Instrument Air:When to trip reactor if instrument air pressure is decreasing.STP Lesson:LOT 505.01Objective Number:92108Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.Attached Reference

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0POP04-IA-0001, Rev.15

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Source:BankDistractor JustificationA:INCORRECT: Plausible because 90 psig is a setpoint but its when the yard isolation valve closes.B:INCORRECT: Plausible because 80 psig is a setpoint but its when the dryer bypass valve opens.C:INCORRECT: Plausible because 70 psig is the approximate setpoint when the main feed reg valves begin to close.D:CORRECT: If air pressure goes below 60 psig, a manual Reactor trip is required by 0POP04-IA-0001. 55 psig is the FIRST pressure indicating less than 60 psig.Question Level:FQuestion Difficulty3Justification:Student must have fundamental knowledge of the procedural requirement to perform a manual Reactor Trip on lowering Inst. Air pressure.RO Importance:3.6NRC Reference Req'd

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Modified fromExam Bank No.:211610CFR

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55.41(b)(7)Page 142 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following correctly describes how control room operators should obtain E2A11 Battery Discharge current? This information can - A. be obtained from CP-003 using the BATT CUR indicator. B. be obtained from CP-010 using the BATT CUR indicator. C. ONLY be obtained from one of the QDPS plasma displays. D. ONLY be obtained locally. A Plant Operator must be dispatched. Answer:Abe obtained from CP-003 using the BATT CUR indicator.Exam Bank No.:2197Last used on an NRC exam:2013RO Sequence Number:72Page 143 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:063 A4.03Tier:2Group/Category:1Ability to manually operate and/or monitor in the control room: Battery discharge rateSTP Lesson:LOT 201.37Objective Number:92986DESCRIBE the local and MCR instrumentation available to monitor the Class 1E 125 VDC SystemAttached Reference

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LOT201.37 PowerPoint slide 55

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Source:BankDistractor JustificationA:CORRECT: This information is available on CP-003B:INCORRECT: Credible because CP-010 is also an electrical panel, however this indication is not located on it.C:INCORRECT: Credible because the QDPS computer provides safety related system information, but not this. NOTE: This information is available on the ICS computer system (in the control room).D:INCORRECT: Credible because many plant parameters are only available through local indication. Tha applicant must be familiar with what indication is on the control panels to correctly respond.Question Level:FQuestion Difficulty2Justification:The applicant must have knowledge of the indications available in the control room for the batteries.RO Importance:3.0NRC Reference Req'd

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Modified fromExam Bank No.:219710CFR

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55.41(b)(7)Page 144 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA large leak of radioactive water is occurring in the Mechanical Auxiliary Building (MAB). Securing the sump pumps for which sump would prevent an unmonitored release directly to the environment? A. Component Cooling Water Sump B. MAB Floor Drain Sump #2 C. Essential Cooling Water Sump D. MAB Elevator Sump Answer:CEssential Cooling Water SumpExam Bank No.:2211Last used on an NRC exam:2013RO Sequence Number:73Page 145 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:G2.3.11Tier:3Group/Category:3Ability to control radiation releases.STP Lesson:LOT 203.10Objective Number:98076Given a set of conditions, PREDICT the effect(s) and/or response(s) on the Equipment and Floor Drains system.Attached Reference

Reference:

LOT 203.10 lesson on Equipment and Floor Drains

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Source:BankDistractor JustificationA:INCORRECT: This is a credible distractor because Component Cooling Water Sump could collect radioactive effluent but the sump is pumped to the Condensate Polisher Regeneration Waste Collection Tank (CPRWCT) and then to a WHT where the water is processed and monitored prior to release.B:INCORRECT: This is a credible distractor because MAB Sump #2 could collect radioactive effluent but is pumped to the Floor Drain Tank and then to a WHT where the water is processed and monitored prior to release.C:CORRECT: Essential Cooling Water Sump pumps directly to the enviroment through the Circ Water system without being monitored.D:INCORRECT: This is a credible distractor because MAB Elevator Sump could collect radioactive effluent but is pumped to MAB Sump #3, then to the Floor Drain Tank and then to a WHT where the water is processed and monitored prior to release.Question Level:FQuestion Difficulty3Justification:In order to control a potential release, the student needs fundemental knowledge of where MAB sumps discharge.RO Importance:3.8NRC Reference Req'd

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Modified fromExam Bank No.:221110CFR

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55.41(b)(5)Page 146 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMA loss of which Class 1E vital 120 VAC power distribution channel will result in a loss of Train 'A' accident monitoring described in TRM 3.3.3.6? A. Channel I B. Channel II C. Channel III D. Channel IV Answer:AChannel IExam Bank No.:2443Last used on an NRC exam:NeverRO Sequence Number:74Page 147 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:APE 057 G2.4.3Tier:1Group/Category:1Loss of Vital AC Electrical Instrumentation Bus:Ability to identify post-accident instrumentationSTP Lesson:LOT 201.38Objective Number:91527LIST typical loads supplied by Class 1E Vital 120 VAC.Attached Reference

Reference:

0POP02-AE-0004 Addendum 2, LOT 201.38, LOT 201.27, TRM 3.3.3.6

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Source:ModifiedDistractor JustificationA:CORRECT: Per 0POP02-AE-0004, 120 VAC ESF Vital Distribution Power Supplies, Addendum 2, Loss of Distribution Panel DP 001 Failures, a loss of DP 001 will de-energize train A containment hydrogen monitoring. Per powerpoint for LOT 201.38, Channel I of Class 1E Vital 120 VAC provides power to DP 001. Also LOT 201.27 states the monitor power supply as hydrogen monitor CM-AIT-4102-DP001. Per TRM 3.3.3.6, the only listed accident instrument is containment hydrogen monitoring. The other options are plausible if the applicant does not remember which channel supplies the A Containment Hydrogen Monitor. Because it requires the applicant to determine what is required accident monitoring in order to answer the question, it meets the K/A.B:INCORRECT: Plausible if student does not remember which channel supplies Train A accident monitoring equipment.C:INCORRECT: Plausible if student does not remember which channel supplies Train A accident monitoring equipment.D:INCORRECT: Plausible if student does not remember which channel supplies Train A accident monitoring equipment.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of power supplies for accident monitoring equipment.RO Importance:3.7NRC Reference Req'd

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Modified from2435Exam Bank No.:244310CFR

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55.41(b)(6)Page 148 of 150 3/6/2016Print DateSTP LOT-20 NRC RO EXAMA loss of which Class 1E vital 120 VAC power distribution channel will result in a loss of Train 'C' accident monitoring described in TRM 3.3.3.6? A. Channel I B. Channel II C. Channel III D. Channel IV Answer:DChannel IVExam Bank No.:2435Last used on an NRC exam:2015RO Sequence Number:42Page 79 of 146 3/6/2016Print DateSTP LOT-20 NRC RO EXAMK/A Catalog Number:062 G2.4.3Tier:2Group/Category:1062 AC Electrical Distribution2.4.3 Ability to identify post-accident instrumentationSTP Lesson:LOT 201.38Objective Number:91527LIST typical loads supplied by Class 1E Vital 120 VAC.Attached Reference

Reference:

0POP02-AE-0004 Addendum 2 Rev. 58, LOT 201.38 Rev. 13, LOT 201.27 Rev. 11, TRM 3.3.3.6

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Source:NewDistractor JustificationA:INCORRECT: Plausible if student does not remember which channel supplies Train C accident monitoring equipment.B:INCORRECT: Plausible if student does not remember which channel supplies Train C accident monitoring equipment.C:INCORRECT: Plausible if student does not remember which channel supplies Train C accident monitoring equipment.D:CORRECT: Per 0POP02-AE-0004, 120 VAC ESF Vital Distribution Power Supplies, Addendum 3, Loss of Distribution Panel DP 002 Failures, a loss of DP 002 will de-energize train C containment hydrogen monitoring. Per powerpoint for LOT 201.38, Channel IV of Class 1E Vital 120 VAC provides power to DP 002. Also LOT 201.27 states the monitor power supply as hydrogen monitor CM-AIT-4105-DP002. Per TRM 3.3.3.6, the only listed accident instrument is containment hydrogen monitoring. The other options are plausible if the applicant does not remember which channel supplies the C Containment Hydrogen Monitor. Because it requires the applicant to determine what is required accident monitoring in order to answer the question, it meets the K/A.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of power supplies for accident monitoring equipment.RO Importance:3.7NRC Reference Req'd

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Modified fromExam Bank No.:243510CFR

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55.41(b)(7)Page 80 of 146 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMWhich of the following would have a direct effect on the performance of the Reactor Coolant Pumps? A. Reactor Containment Fan Coolers B. Reactor Containment Cubicle Exhaust Fans C. Containment Isolation Phase A D. Containment Isolation Phase B Answer:DContainment Isolation Phase BExam Bank No.:2471Last used on an NRC exam:NeverRO Sequence Number:75Page 149 of 150 3/4/2016Print DateSTP LOT-20.1 NRC RO EXAMK/A Catalog Number:003 K1.08Tier:2Group/Category:1Knowledge of the physical connections and/or cause-effect relationships between the RCPs and the following systems:

Containment isolation.STP Lesson:LOT 201.05Objective Number:50785DESCRIBE the physical relationship between the reactor collant pumpsand thefollowing: A. RCP Seals B. Chemical Volume and Control System C. Containment Isolation System D. Reactor Coolant System E. Component Cooling Water.Attached Reference

Reference:

LOT 201.05 Lesson Plan for RCPs

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the RCPs are in containment and the RCFCs help cool containment but they do not have any direct cooling connections to the RCPs.B:INCORRECT: Plausible because the Reactor Containment Cubicles Exhaust fans do help cool components in selected areas of containment but not the RCPs.C:INCORRECT: Plausible because Containment Isolation Phase A isolates many systems to containment but not any to the RCPs.D:CORRECT: If a Containemnt Isolation Phase B were to occur then CCW to the RCPs would be lost and the RCPs would hav eto be tripped.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of systems that directly interface with the RCPs.RO Importance:2.7NRC Reference Req'd

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Modified fromExam Bank No.:247110CFR

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55.41(b)(7)Page 150 of 150 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit just completed a rapid load reduction from 100% power due to an equipment problem in the secondary plant and have stabilized the plant at 80% power. Assuming Reactor power is to remain at 80%, the Unit Supervisor will be required to have chemistry collect an RCS sample to- A. perform an Isotopic Analysis for Iodine. B. perform a Radiochemical Analysis for determination. C. determine Chloride level. D. determine Dissolved Oxygen level. Answer:A perform an Isotopic Analysis for Iodine.Exam Bank No.:2461Last used on an NRC exam:NeverSRO Sequence Number:76Page 1 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.2.42Tier:3Group/Category:2Ability to recognize system parameters that are entry level conditions for Technical SpecificationsSTP Lesson:LOT 503.01Objective Number:SRO92102Given the topic or title of a specification included in the Technical Specifications, or the Technical Requirements Manual (TRM), DESCRIBE the general requirements of the specification to include components or administrative requirements affected, limitations, major time frames involved, major surveillance in order to comply, and the bases for the specification.Attached Reference

Reference:

TS 3/4.4.8

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Source:NewDistractor JustificationA:CORRECT: An analysis for Iodine is the only one required after a power change of 15% or more in a one hour period.B:INCORRECT: All distractors are plausible because they are samples required by TSs or the TRM.C:INCORRECT: All distractors are plausible because they are samples required by TSs or the TRM.D:INCORRECT: All distractors are plausible because they are samples required by TSs or the TRM.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of Technical Specification requirements.SRO Importance:4.6NRC Reference Req'd

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Modified FromExam Bank No.:246110CFR Reference or SRO Objective:55.43(b)(2)Page 2 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMWhile performing diagnostic steps in 0POP05-EO-EO00, Reactor Trip or Safety Injection, the following conditions are observed: RCS Tave is 530°F and lowering RCS pressure is 1650 psig and lowering PRZR level is 54% and lowering All SG pressures are 900 psig and lowering All SG Wide Range levels are 55% and lowering CNTMT Temperature is 175F and rising CNTMT radiation levels are normal Auxiliary Building radiation levels are normal All MSIVs are open Which of the following would cause this event and the procedure the Unit Supervisor would next enter? Cause of Event Procedure A. LOCA INSIDE Containment 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant B. Steam Line Break INSIDE Containment 0POP05-EO-EO20, Faulted Steam Generator Isolation C. LOCA OUTSIDE Containment 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant D. Steam Line Break OUTSIDE Containment 0POP05-EO-EO20, Faulted Steam Generator Isolation Answer:BSteam Line Break Inside Containment - 0POP05-EO-EO20, Faulted Steam Generator IsolationExam Bank No.:647Last used on an NRC exam:2015SRO Sequence Number:77Page 3 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 040 AA2.01Tier:1Group/Category:1Ability to determine and interpret the following as they apply to the Steam Line Rupture:Occurrence and location of a steam line rupture from pressure and flow indications.STP Lesson:LOT 504.13Objective Number:81261STATE/IDENTIFY the indications and anticipated readings used to determine that a faulted Steam Generator exists and which Steam Generator(s) is/are faulted.Attached Reference

Reference:

LOT 504.13 Lesson Plan on Faulted Steam Generators and 0POP05-EO-EO20, Faulted Steam Generator Isolation.

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Source:BankDistractor JustificationA:INCORRECT: Plausible distractor in that containment conditions are not normal and may lead the student into believing a LOCA inside containment is in progress.B:CORRECT: With the given conditions and especially high temperature in the RCB with a lack of radiation alarms, the event is a Main Steam line break in the RCB.C:INCORRECT: Plausible in that the RCS low pressure conditions may lead the student into selecting a LOCA outside containment.D:INCORRECT: Plausible distracter in that the student may see only the SG low pressures and level and believe a feedbreak is occurring.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given conditions to determine the type of event.SRO Importance:4.7NRC Reference Req'd

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Modified FromExam Bank No.:64710CFR Reference or SRO Objective:55.43(b)(5)Page 4 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMA LOCA with core damage has occurred with the following conditions: An SAE has been declared. The TSC and EOF are activated. A MAB entry is required at the 41' containment penetration area. Projected dose rate in the area is 1.16E+5 mR/hr. Duration of the exposure is expected to be 3 minutes. Who must authorize this exposure? A. Plant General Manager B. STPNOC Vice President C. Radiological Director D. Emergency Director Answer:DEmergency DirectorExam Bank No.:661Last used on an NRC exam:2014SRO Sequence Number:78Page 5 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.3.4Tier:3Group/Category:3Knowledge of radiation exposure limits under normal or emergency conditions.STP Lesson:LOT 803.14Objective Number:SRO-65180Given a description of responsibilities related to an Emergency Response Organization position that interfaces with the Emergency Director, DETERMINE the responsible individual by title.Attached Reference

Reference:

0ERP01-ZV-IN06; 0PGP03-ZR-0050

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Source:BankDistractor JustificationA:INCORRECT: Plausible because the Plant General Manager is responsible for authorizing exposures in excess of 2 Rem at STP or 3 Rem Total during NORMAL operating conditions.B:INCORRECT: Plausible because the STPNOC Vice President is responsible for authorizing exposures in excess of 2 Rem at STP or 4 Rem Total during NORMAL operating conditions.C:INCORRECT: Plausible because the Radiological Director is responsible for authorizing exposures above 2 Rem but less than 5 Rem when responding to EMERGENCY conditions.D:CORRECT: Emergency Director is responsible for authorizing exposures above 5 Rem when responding to EMERGENCY conditions. With a projected dose rate of 1.16E+5 mR/hr the total dose to respond to this emergency condition is 5.8 Rem (1.16E+5 mR/hr / 60minutes x 3 minutes =

5.8 R)Question Level:HQuestion Difficulty3Justification:The student must accurately determine the projected dose and then use that knowledge to determine the approval authority based on the applicable exposure limits for each position.SRO Importance:3.7NRC Reference Req'd

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Modified FromExam Bank No.:66110CFR Reference or SRO Objective:55.43(b)(4)Page 6 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following condition: Pressurizer Heater Group 'C' has lost power. Pressurizer Pressure will ____(1)_____. AND The Unit Supervisor should _____(2)_____. (1) (2) A. lower enter TS 3.4.3, Pressurizer, AND 0POP04-RP-0001, Loss of Pressurizer Pressure Control B. lower ONLY enter 0POP04-RP-0001, Loss of Pressurizer Pressure Control C. remain the same enter TS 3.4.3, Pressurizer, AND 0POP04-RP-0001, Loss of Pressurizer Pressure Control D. remain the same ONLY enter 0POP04-RP-0001, Loss of Pressurizer Pressure Control Answer:B lower - ONLY enter 0POP04-RP-0001, Loss of Pressurizer Pressure ControlExam Bank No.:2505Last used on an NRC exam:NeverSRO Sequence Number:79Page 7 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:010 A2.01Tier:2Group/Category:1Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failuresSTP Lesson:LOT 507.01Objective Number:92106Given plant conditions/symptoms, EVALUATE the conditions/symptoms and STATE whether or not the referenced procedure is to be used.Attached Reference

Reference:

0POP04-RP-0001 and TS 3.4.3

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Source:NewDistractor JustificationA:INCORRECT: Plausible because if the student has to remember which sets of PZR heaters are covered by Technical Specifications. It would be reasonable for a student to believe that PZR Heater Group C is the controlling group and covered by TSs.B:CORRECT: PZR Heater Group C is the controlling group of heaters but it is Groups A and D that are covered by TS so only the procedure would be entered.C:INCORRECT: Plausible because it would be reasonable for the student to believe that on eof the other PZR Heater Groups is the controlling group. Also plausible because if the student has to remember which sets of PZR heaters are covered by Technical Specifications. It would be reasonable for a student to believe that PZR Heater Group C is the controlling group and covered by TSs.D:INCORRECT: Plausible because it would be reasonable for the student to believe that on eof the other PZR Heater Groups is the controlling group and the procedure would be entered for any failure of a component of the PZR pressure control system even though the pressure my not actually change.Question Level:HQuestion Difficulty3Justification:The student must determine what affect the loss of the control group of heaters will have on how the master controller will maintain pressurizer pressure and that entry requirements for the off-normal procedure have been met but that TSs have not.SRO Importance:3.6NRC Reference Req'd

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Modified FromExam Bank No.:250510CFR Reference or SRO Objective:55.43(b)(5)Page 8 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit was in Mode 3 when a Control Room Evacuation occurred with the following conditions: The crew has stationed themselves at their designated locations. Train 'A' 480V LC's are de-energized. (1) Describe the operational effect if these 480V LC's are not re-energized. AND (2) What action should the Shift Manager (SM) perform? A. (1) Two trains of RHR would be unavailable. (2) The SM should direct the Secondary RO to re-energize the LC's per 0POP04-ZO-0001, Control Room Evacuation. B. (1) One train of RHR would be unavailable. (2) The SM should direct the Secondary RO to re-energize the LC's per 0POP04-ZO-0001, Control Room Evacuation. C. (1) Two trains of RHR would be unavailable. (2) The SM should transition to 0POP04-AE-0001, First Response to Any or All 13.8 kV or 4.16 kV Bus, and direct the Primary RO to re-energize the LC's. D. (1) One train of RHR would be unavailable. (2) The SM should transition to 0POP04-AE-0001, First Response to Any or All 13.8 kV or 4.16 kV Bus, and direct the Secondary RO to re-energize the LC's. Answer:A (1) Two trains of RHR would be unavailable.(2) The SM should direct the Secondary RO to re-energize the LC's per 0POP04-ZO-0001, Control Room Evacuation.Exam Bank No.:1714Last used on an NRC exam:NeverSRO Sequence Number:80Page 9 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.4.34Tier:3Group/Category:4Knowledge of RO Tasks performed outside the main control room during emergency and the resultant operational effects.STP Lesson:LOT 505.01Objective Number:92108Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.Attached Reference

Reference:

0POP04-ZO-0001, Rev. 32

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Source:BankDistractor JustificationA:CORRECT: All RHR train suction valves have cross-train power supplies, thus loss of any one train of power will render 2 RHR trains inoperable. 0POP04-ZO-0001, Control Room Evacuation providesthe instructions to re-energize these LC's since the control is transferred to a local control station. The Secondary RO is stationed in the 'A' Train Switchgear Room and so would be the operator that would carry out the action to re-energize the LC's.B:INCORRECT: All RHR train suction valves have cross-train power supplies, thus loss of any one train of power will render 2 RHR trains inoperable, not just one.C:INCORRECT: 0POP04-ZO-0001, Control Room Evacuation is the priority procedure for the US to use at this time. It provides the instructions to re-energize these LC's since the control is transferred to a local control station. Also, it would be the Secondary RO, not the Primary RO that performs the action.D:INCORRECT: 0POP04-ZO-0001, Control Room Evacuation is the priority procedure for the US to use at this time. It provides the instructions to re-energize these LC's since the control is transferred to a local control station. All RHR train suction valves have cross-train power supplies, thus loss of any one train of power will render 2 RHR trains inoperable.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given data and have fundamental knowledge of where the RO stations are during a Control Room Evac, that the suction valves of each RHR train are powered from different 480V LC's effect of losing power to the Train 'A' and that 0POP04-ZO-0001 is the highest priority procedure to implement under the given conditions.SRO Importance:4.1NRC Reference Req'd

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Modified FromExam Bank No.:171410CFR Reference or SRO Objective:55.43(b)(5)Page 10 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following condition: SFP level is within the requirements of Technical Specifications. An electrical malfunction results in the loss of the load center supplying the in service Spent Fuel Pool Cooling Pump, causing Annunciator 22M2-F6, SFP TROUBLE, to alarm. With no operator actions, which of the following correctly describes the impact of the loss AND the action to be taken by the Unit Supervisor? Spent Fuel Pool Temperature _____(1)_____. AND The Unit Supervisor will enter _____(2)_____. (1) (2) A. will NOT rise to the boiling point the annunciator response procedure for the SFP TROUBLE alarm and start the standby pump B. WILL rise to the boiling point the annunciator response procedure for the SFP TROUBLE alarm and start the standby pump C. will NOT rise to the boiling point 0POP04-FC-0001, Loss of Spent Fuel Pool Level or Cooling, and add water to the Spent Fuel Pool using a LHSI pump to maintain temperature since the loss of the load center also disabled the standby pump D. WILL rise to the boiling point 0POP04-FC-0001, Loss of Spent Fuel Pool Level or Cooling, and add water to the Spent Fuel Pool using a LHSI pump to maintain temperature since the loss of the load center also disabled the standby pump Answer:BWILL rise to the boiling point. Enter the annunciator response procedure for the SFP TROUBLE alarm and start the standby pump.Exam Bank No.:1766Last used on an NRC exam:2009SRO Sequence Number:81Page 11 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:033 A2.02Tier:2Group/Category:2Ability to (a) predict the impacts of the following malfunctions on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:Loss of SFPCSSTP Lesson:LOT 201.42Objective Number:92051GIVEN a plant or system condition, PREDICT the operation of the Spent Fuel Pool Cooling and Cleanup System.Attached Reference

Reference:

Tech Spec 3.9.11.1 bases; 0POP09-AN-22M2 (Rev 21), window F-6

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Source:BankDistractor JustificationA:INCORRECT: Requirements for water level in the SFP are for dose considerations, not mass considerations for cooling (temperature will rise to boiling). The annunciator response for SFP Trouble directs the operators to start the standby pump upon trip of a running pump. The standby pump is powered from a different source.B:CORRECT: Requirements for water level in the SFP are for dose considerations, not mass considerations for cooling. The annunciator response for SFP Trouble directs the operators to start the standby pump upon trip of a running pump. The standby pump is powered from a different source.C:INCORRECT: The loss of the indicated Load Center will not disable the standby pump. Requirements for water level in the SFP are for dose considerations, not mass considerations for cooling (temperature will rise to boiling). While the POP04 does contain directions for using a LHSI pump to maintain SFP level, it is only used if normal means are not sufficient.D:INCORRECT: The loss of the indicated Load Center will not disable the standby pump. Requirements for water level in the SFP are for dose considerations, not mass considerations for cooling. While the POP04 does contain directions for using a LHSI pump to maintain SFP level, it is only used if normal means are not sufficient.Question Level:HQuestion Difficulty3Justification:The student must assess plant conditions and using system and procedure knowledge, predict the impact on the plant and the action required to help mitigate the situation.SRO Importance:3.0NRC Reference Req'd

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Modified FromExam Bank No.:176610CFR Reference or SRO Objective:55.43(b)(5)Page 12 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMUnit 1 is at 100% power with the following conditons: The Gaseous Waste Processing System (GWPS) is in service for normal operations. Subsequently: A GWPS automatic shutdown occurs. (1) Which of the following is consistent with an automatic shutdown of GWPS? AND (2) What action will the Unit Supervisor take?

A. (1) RT-8031, GWPS Inlet Rad Monitor, has reached its HIGH alarm setpoint. (2) Enter 0POP04-RA-0001, Radiation Monitoring System Alarm Response, and verify GWPS Inlet Header Valve, FV-4657, and GWPS Discharge Flow Valve, FV-4671, are closed. B. (1) RT-8031, GWPS Inlet Rad Monitor, has reached its HIGH alarm setpoint. (2) Enter TRM 3/4.3.3.11, Explosive Gas Monitoring Instrumentation, to initiate grab sampling once Waste Gas Processing System operation is re-established. C. (1) RT-8032, GWPS Outlet Rad Monitor, has reached its HIGH alarm setpoint. (2) Enter 0POP04-RA-0001, Radiation Monitoring System Alarm Response, and verify GWPS Inlet Header Valve, FV-4657, and GWPS Discharge Flow Valve, FV-4671, are closed. D. (1) RT-8032, GWPS Outlet Rad Monitor, has reached its HIGH alarm setpoint. (2) Enter TRM 3/4.3.3.11, Explosive Gas Monitoring Instrumentation, to initiate grab sampling once Waste Gas Processing System operation is re-established. Answer:C(1) RT-8032, GWPS Outlet Rad Monitor, has reached its HIGH alarm setpoint.(2) Enter 0POP04-RA-0001, Radiation Monitoring System Alarm Response, and verify GWPS Inlet Header Valve, FV-4657, and GWPS Discharge Flow Valve, FV-4671, are closed.Exam Bank No.:1804Last used on an NRC exam:NeverSRO Sequence Number:82Page 13 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.4.46Tier:3Group/Category:4Ability to verify that the alarms are consistent with the plant conditions.STP Lesson:LOT 203.14Objective Number:91649Discuss the following operations (general): C. Abnormal ConditionsAttached Reference

Reference:

0POP04-RA-0001, Addendum 8

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Source:BankDistractor JustificationA:INCORRECT: RT-8031 will not initiate an automatic system shutdown. However, the action specified is correct for the rad monitor that will cause a system shutdown (RT-8032)B:INCORRECT: RT-8031 will not initiate an automatic system shutdown. Also, the O2 monitor does NOT become inoperable on a system shutdown so TRM entry is not req'd.C:CORRECT: RT-8032 will initiate a system shutdown at the HIGH alarm setpoint. This will cause both the inlet and outlet valves to close to the GWPS. Appropriate action is given in 0POP04-RA-0001.D:INCORRECT: RT-8032 will initiate a system shutdown at the HIGH alarm setpoint. This will cause both the inlet and outlet valves to close to the GWPS. However, the O2 monitor does NOT become inoperable on a system shutdown so TRM entry is not req'd.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given data and have knowledge of the rad monitor conditons that will initiate an automatic system shutdown and the appropriate procedural action based on automatic equipment operation.SRO Importance:4.2NRC Reference Req'd

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Modified FromExam Bank No.:180410CFR Reference or SRO Objective:55.43(b)(5)Page 14 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: PZR pressure channel PT-458 is in BYPASS for maintenance. PZR pressure channel PT-457 is the CONTROLLING channel. Subsequently: PZR pressure channel PT-457 FAILS HIGH The Unit Supervisor will ensure the Reactor Operator _____(1)_____ to control PZR pressure and will enter _____(2)_____. (1) (2) A. deselects PT-457 TS 3.3.1 Action 6 and TS 3.3.2 Action 20; place PT-457 in TRIPPED condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. deselects PT-457 TS 3.0.3; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to shutdown the Unit and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C. places PZR pressure control in manual TS 3.3.1 Action 6 and TS 3.3.2 Action 20; place PT-457 in TRIPPED condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> D. places PZR pressure control in manual TS 3.0.3; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to shutdown the Unit and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Answer:B deselects PT-457 - TS 3.0.3; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to shutdown the Unit and be in HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sExam Bank No.:2445Last used on an NRC exam:NeverSRO Sequence Number:83Page 15 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 027 AA2.15Tier:1Group/Category:1Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Actions to be taken if PZR pressure instrument fails high.STP Lesson:LOT 201.14Objective Number:92779GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control systems.Attached Reference

Reference:

LOT 201.14 Lesson on PZR pressure and level control systems. TS 3.0.3.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the TS action describe is an action taken but for one failed channel. With one channel already in BYPASS, however, that's makes 2 channels inoperable requiring the ACTIONS of TS 3.0.3.B:CORRECT: Since the PZR pressure channel failed high the correct response is to deselct the channel per 0POP04-RP-0001. When a channel is in BYPASS the channel is still considered inoperable. With another channel failed, PT-457, then the correct action is to enter TS 3.0.3.C:INCORRECT: Plausible because if PT-457 fails low then the action would be to place PZR pressure control in manual. Also, plausible because the TS action describe is an action taken but for one failed channel. With one channel already in BYPASS, however, that's makes 2 channels inoperable requiring the ACTIONS of TS 3.0.3.D:INCORRECT: Plausible because if PT-457 fails low then the action would be to place PZR pressure control in manual.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given PZR control system conditions to determine the correct answer.SRO Importance:4.0NRC Reference Req'd

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Modified FromExam Bank No.:244510CFR Reference or SRO Objective:55.43(b)(5)Page 16 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMUnit 2 is at 100% power with a normal electrical lineup. Subsequently: A lock-out occurs on the switch yard SOUTH Bus. The Unit 2 Unit Supervisor will enter Technical Specification _____(1)_____ and direct the actions of _____(2)_____ to restore lost electrical power once the switchyard SOUTH Bus is restored.

A. (1) 3.8.1.1.a due to loss of ONE required offsite circuit (2) 0POP04-AE-0002, Loss of One or More 13.8 KV Auxiliary or Non-Class 4.16 KV Bus D B. (1) 3.8.1.1.a due to loss of ONE required offsite circuit (2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby Bus C. (1) 3.8.1.1.e due to loss of TWO required offsite circuit (2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby Bus D. (1) 3.8.1.1.e due to loss of TWO required offsite circuit (2) 0POP04-AE-0002, Loss of One or More 13.8 KV Auxiliary or Non-Class 4.16 KV Bus D Answer:C(1) 3.8.1.1.e due to loss of TWO required offsite circuit(2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby BusExam Bank No.:2447Last used on an NRC exam:NeverSRO Sequence Number:84Page 17 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 056 G2.4.11Tier:1Group/Category:1Loss of Off-Site Power:Knowledge of abnormal condition procedures.STP Lesson:LOT 201.31Objective Number:62351GIVEN a plant or system condition, PREDICT the operation of the Non-Class 1E 13.8 to 4.16 volt AC distribution system.Attached Reference

Reference:

LOT 201.31 Lesson Plan on Non-Class 13.8 and 4.16 KV power and TS 3.8.1.1.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because losing one swityard bus constitues losing one offsite circuit. It is reasonable for a student to forget that the Stanby Bus G would also be effected. Also plausible because for this condition the Unit Supervisor would start directing actions per 0POP04-AE-0001 and then have to make a determination of which procedure to use next based on the extent of the loss of power.B:INCORRECT: Plausible because losing one swityard bus constitues losing one offsite circuit. It is reasonable for a student to forget that the Stanby Bus G would also be effected.C:CORRECT: Under a normal electrical lineup, if the South Bus was lost, it would constitue losing TWO required offsite sources because 13.8 KV Standby Bus 2G would also be effected. 0POP04-AE-0001 would first be entered and then 0POP04-AE-0003 would be used to restore power to Unit 2 13.8 KV Standby Bus 2G once the South switch yard Bus was restored.D:INCORRECT: Plausible because for this condition the Unit Supervisor would start directing actions per 0POP04-AE-0001 and then have to make a determination of which procedure to use next based on the extent of the loss of power.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the condition to determine the extent of the loss of power and have knowledge of normal electrical lineups and of off-normal electrical procedures.SRO Importance:4.2NRC Reference Req'd

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Modified FromExam Bank No.:244710CFR Reference or SRO Objective:55.43(b)(2)Page 18 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: PZR level control is selected to 465/466 for control Subsequently: Vital AC Bus DP-1201 losses power PZR level will begin to _____(1)_____. The Unit Supervisor will enter _____(2)_____. (1) (2) A. lower 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permits B. lower 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control. The US will also enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, if manpower permits C. rise 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permits D. rise 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control. The US will also enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, if manpower permits Answer:C rise until the RO takes manual control - 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permitsExam Bank No.:2448Last used on an NRC exam:NeverSRO Sequence Number:85Page 19 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 057 AA2.12Tier:1Group/Category:1Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:PZR level controller, instrumentation, and heater operation.STP Lesson:LOT 201.14Objective Number:92779GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control systems.Attached Reference

Reference:

LOT 201.14 Lesson on PZR pressure and level control systems.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the student has to know how the failure affects the PZR level control system.B:INCORRECT: Plausible because the student has to know how the failure affects the PZR level control system. Also it would be plausible to enter 0POP04-RP-0002 first because the failure would directly affect PZR level control.C:CORRECT: A loss of power to DP-1201 would cause the controlling PZR level channel, LT-465, to fail low. This would cause actual level to rise as CV-FV-0205 would begin to open until manual control was made by the RO. The US would enter 0POP04-VA-0001 and only enter 0POP04-RP-0002 if manpower was available.D:INCORRECT: It would be plausible to enter 0POP04-RP-0002 first because the failure would directly affect PZR level control.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given PZR control system conditions to determine the correct answer.SRO Importance:3.7NRC Reference Req'd

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Modified FromExam Bank No.:244810CFR Reference or SRO Objective:55.43(b)(5)Page 20 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMA Unit trip has occurred from 100% power with the following conditions: No AFW Pumps can be started All SG levels are 10% narrow range and lowering Based on these conditions which have been observed for 15 minutes, what emergency action level would the Emergency Director declare? A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer:CSite Area EmergencyExam Bank No.:2449Last used on an NRC exam:NeverSRO Sequence Number:86Page 21 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:W/E05 G2.4.41Tier:1Group/Category:1Loss of Secondary Heat Sink:Knowledge of emergency action level thresholds and classifications.STP Lesson:LOT 803.14Objective Number:SRO-74026GIVEN an emergency condition and a copy of the emergency classification tables from 0ERP01-ZV-IN01, Emergency Classification, CLASSIFY the emergency condition.Attached Reference

Reference:

LOT 803.14 Lesson Plan on Emergency Classifications

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0ERP01-ZV-IN01, Emergency ClassificationSource:NewDistractor JustificationA:INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.B:INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.C:CORRECT: The student must be able to recognize that the given condition represents a loss of heat sink RED PATH. This causes a potential loss of both the RCS and Fuel Cladding. Therefore the correct classification is a Site Area Emergency.D:INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given conditions and apply the emergency classification tables and procedure.SRO Importance:4.6NRC Reference Req'd

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Modified FromExam Bank No.:244910CFR Reference or SRO Objective:Objective SRO-74026Page 22 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit experienced an ATWS and the crew is performing 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS, with the following condition: Emergency Boration has been aligned to the RCS per Step 4 The crew is in the process of completing Step 5, ENSURE Containment Ventilation Isolation. Subsequently Operators make the following observations: Core Exit TCs are 680ºF and lowering Extended Range NIs are 4% and lowering Extended Range NIs Startup Rate is -0.1 What procedural steps should the Unit Supervisor perform NEXT? A. TRANSITION TO 0POP05-EO-EO00, Reactor Trip or Safety Injection. B. TRANSITION TO 0POP05-EO-FRC2, Response to Degraded Core Cooling. C. GO TO Step 18 and then SECURE Emergency Boration. D. GO TO Step 6 and CHECK SI Status. Answer:CGO TO step 18 and then secure emergency boration.Exam Bank No.:2450Last used on an NRC exam:NeverSRO Sequence Number:87Page 23 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:EPE 029 G2.1.20Tier:1Group/Category:1ATWS:Ability to interpret and execute procedure steps.STP Lesson:LOT 504.28Objective Number:83681DESCRIBE the indications and anticipated readings used to determine that all dilution paths are isolated.Attached Reference

Reference:

LOT 504.28 Lesson on Response to Nuclear Power Generation - ATWS

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Source:NewDistractor JustificationA:INCORRECT: Plausible because transitioning to 0POP05-EO-EO00 would eventually be performed but it is not the NEXT step. Emergency Boration must be secured first.B:INCORRECT: Plausible because the indicated CET reading of 680 degrees F is close to the conditon of 708 degrees F to enter 0POP05-EO-FRC2.C:CORRECT: The conditions given would indicate that the reactor is being shutdown and that emergency boration can be secured. Going to step 18 is a CIP step.D:INCORRECT: Plausible because it would be reasonable to believe that conditions warrant to continue with the procedure to the next step.Question Level:HQuestion Difficulty3Justification:The student must analyze the given conditions to determine what the next required procedure step needs to be taken.SRO Importance:4.6NRC Reference Req'd

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Modified FromExam Bank No.:245010CFR Reference or SRO Objective:55.43(b)(5)Page 24 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMA Fast Load Reduction was being performed. At 80% power Control Rods were placed in Manual when the following was observed: One Control Rod in Control Bank D, Group 2, H8, was at 216 steps. All other Control Rods in Bank D were at 200 steps. Control Bank D, Group 2, Control Rod H8-A. remains OPERABLE if power is reduced to 75% within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. B. remains OPERABLE if it is determined that Shutdown Margin is satisfied within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.C. is INOPERABLE. Be in HOT Standby within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. D. is INOPERABLE. Restore the misaligned Control Rod H8 to within +/-12 steps within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Answer:D is INOPERABLE. Restore the misaligned Control Rod H8 to within +/-12 steps within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.Exam Bank No.:2451Last used on an NRC exam:NeverSRO Sequence Number:88Page 25 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 005 G2.2.37Tier:1Group/Category:2Inoperable/Stuck Control Rod:Ability to determine operability and/or availability of safety related equipment.STP Lesson:LOT 503.01Objective Number:80056GIVEN a system scenario, DETERMINE the applicable Technical Specifications and/or Technical Requirements Manual (TRM) for the system and APPLY the specification(s).Attached Reference

Reference:

LOT 503.01 Technical Specifications TS 3.1.3.1

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Source:NewDistractor JustificationA:INCORRECT: Plausible because there is a requirement in TS 3.1.3.1 to reduce power to less than 75 percent but the Control Rod is INOPERABLE because it is misaligned by more than 12 steps.B:INCORRECT: Plausible because there is a requirement in TS 3.1.3.1 to determine shutdown margin but the Control Rod is INOPERABLE because it is misaligned by more than 12 steps.C:INCORRECT: Plausible because the control rod is INOPERABLE but a requirement to be in HOT STANDBY is only for a stuck control rod and there is no indication that the control rod is stuck. That must come from I/C. Plus the requirement to be in HOT STANDBY allows 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.D:CORRECT: With the given conditions the Control Rod is INOPERABLE because it is misaligned by more than 12 steps. One of the actions to restore OPERABILITY would be to realign the Control Rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given control rod condition to determine Operability and TS action.SRO Importance:4.6NRC Reference Req'd

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Modified FromExam Bank No.:245110CFR Reference or SRO Objective:55.43(b)(2)Page 26 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: Steam Generators A B C D Steam Line Radiation 1.7E-2 uCi/cc 1.5E-2 uCi/cc 1.4E-2 uCi/cc 3.9E-1 uCi/cc Blowdown Radiation 2.7E-4 uCi/cc 2.4E-4 uCi/cc 2.3E-4 uCi/cc 4.6E-2 uCi/cc N-16 Monitors 9.0 gpd 0.2 gpd 0.1 gpd 77.0 gpd Chemistry reports total current primary to secondary leak rate is 79 gpd. The rate of increase is about 1 to 2 gpd/hr. (1) Which Steam Generator(s) have tube leak(s)?

AND (2) What action will the Unit Supervisor take per 0POP04-RC-0004, Steam Generator Tube Leakage? A. (1) ONLY Steam Generator 'D' has a tube leak. (2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. B. (1) ONLY Steam Generator 'D' has a tube leak. (2) Transition to 0POP04-TM-0005, Fast Load Reduction, and reduce power to <50% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. C. (1) Steam Generators 'A' and 'D' have tube leaks. (2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D. (1) Steam Generators 'A' and 'D' have tube leaks. (2) Transition to 0POP04-TM-0005, Fast Load Reduction, and reduce power to <50% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Answer:A (1) ONLY Steam Generator 'D' has a tube leak.(2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.Exam Bank No.:2452Last used on an NRC exam:NeverSRO Sequence Number:89Page 27 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 037 AA2.01Tier:1Group/Category:2Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:Unusual readings of the monitors; steps needed to verify readings.STP Lesson:LOT 505.01Objective Number:92108Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.Attached Reference

Reference:

LO T 505.01 Lesson Plans for Off-Normal Procedures. 0POP04-RC-0004, Steam Generator Tube Leakage.

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Source:NewDistractor JustificationA:CORRECT: With the given conditions only SG D would have a tube leak and the correct procedure would be 0POP04-RC-0004.B:INCORRECT: Plausible because if the believed to be large enough leak then tripping the reactor would be warranted however total leakage would have to be close to or greater than the capacity of a charging pump. About 200gpm.C:INCORRECT: Plausible because of the elevated N-16 reading for SG A. However, the N-16 reading would be normal on SG A with a leak on SG D because SG A N-16 monitor would pick up some indications from SG D.D:INCORRECT: Plausible because if the believed to be large enough leak then tripping the reactor would be warranted however total leakage would have to be close to or greater than the capacity of a charging pump. About 200gpm. Also, because of the elevated N-16 reading for SG A. However, the N-16 reading would be normal on SG A with a leak on SG D because SG A N-16 monitor would pick up some indications from SG D.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given readings for each SG and determine which SG(s) have tube leaks or tube ruptures and apply the correct procedure.SRO Importance:3.4NRC Reference Req'd

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Modified FromExam Bank No.:245210CFR Reference or SRO Objective:55.43(b)(5)Page 28 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: Main Condenser vacuum is 23" HG and slowly lowering. The low Main Condenser vacuum would be caused by a _____(1)_____ and the Unit Supervisor would direct _____(2)_____. (1) (2) A. Condensate Pump trip and failure of the standby pump to start a fast load reduction per 0POP04-CD-0001, Loss of Condensate Flow B. Condensate Pump trip and failure of the standby pump to start the immediate actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection C. Condenser Air Removal Pump trip and failure of the standby pumps to start a fast load reduction per 0POP04-CR-0001, Loss of Condenser Vacuum D. Condenser Air Removal Pump trip and failure of the standby pumps to start the immediate actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection Answer:CCondenser Air Removal Pump trip and failure of the standby pumps to start - a fast load reduction per 0POP04-CR-0001, Loss of Condenser VacuumExam Bank No.:2453Last used on an NRC exam:NeverSRO Sequence Number:90Page 29 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 051 AA2.01Tier:1Group/Category:2Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:Cause for low vacuum condition.STP Lesson:LOT 202.25Objective Number:33957GIVEN a plant or system condition, PREDICT the operation of the Condenser Air Removal System.Attached Reference

Reference:

LOT 202.25 lesson on Condenser Air Removal

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Source:NewDistractor JustificationA:INCORRECT: Plausible because a trip of a Condensate Pump with a failure of the standby pump to start will cause condenser vacuum to lower, however, with normal condenser vacuum at 29" HG it would only lower to about 27.5" HG. The procedure would be correct if a Condensate Pump tripped with the failure of the standby condensate pumps to start.B:INCORRECT: Plausible because a trip of a Condensate Pump with a failure of the standby pump to start will cause condenser vacuum to lower, however, with normal condenser vacuum at 29" HG it would only lower to about 27.5" HG. The procedure would be correct if a Condensate Pump tripped with the failure of the standby condensate pumps to start. Also, with the current condenser vacuum at 23" HG this would only cause the Low Condenser Vacuum alarm to come in and not low enough to cause a main turbine and reactor trip. That setpoint is 21"HG.C:CORRECT: A trip of a condenser air removal pump with the standby pumps failing to start will cause a significant lowering of vacuum in the main condenser and with the current condition givenof 23"HG and slowly lowering the correct procedure is 0POP04-CR-0001 and perform a fast load reduction per step 6.D:INCORRECT: Plausible because with the current condenser vacuum at 23" HG this would only cause the Low Condenser Vacuum alarm to come in and not low enough to cause a main turbine and reactor trip. That setpoint is 21"HG.Question Level:HQuestion Difficulty2Justification:The student must be able to analyze the given condenser vacuum condition to determine the cause and correct procedure.SRO Importance:2.7NRC Reference Req'd

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Modified FromExam Bank No.:245310CFR Reference or SRO Objective:55.43(b)(5)Page 30 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is in Mode 6 with fuel movement in progress. Subsequently: Containment High Range Monitor RT-8050 brings in a Dark Blue condition on RM-11. Containment High Range Monitor RT-8051 brings in a Magenta condition on RM-11. (1) Which of the following is correct about the condition of RT-8050 and RT-8051? AND (2) What action will the Unit Supervisor take? A. (1) Both RT-8050 and RT-8051 are Non-Functional. (2) Suspend fuel movement. B. (1) Both RT-8050 and RT-8051 are Non-Functional. (2) Suspend Polar Crane operations with loads over the Reactor Cavity. C. (1) ONLY RT-8051 is Non-Functional. (2) Suspend fuel movement. D. (1) ONLY RT-8051 is Non-Functional. (2) Suspend Polar Crane operations with loads over the Reactor Cavity. Answer:A (1) Both RT-8050 and RT-8051 are Non-Functional.(2) Suspend fuel movement.Exam Bank No.:2454Last used on an NRC exam:NeverSRO Sequence Number:91Page 31 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:APE 061 G2.4.45Tier:1Group/Category:2ARM Systems Alarms:Ability to prioritize and interpret the significance of each annunciator or alarm.STP Lesson:LOT 202.41Objective Number:68793DESCRIBE the meanings of the colors on the RM-11 display.Attached Reference

Reference:

LOT 202.41 Lesson Plan on Radiation Monitors

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Source:NewDistractor JustificationA:CORRECT: If both RT-8050 and RT-8051 are non-functional then fuel movement must be suspended or an alternate method for monitoring a radioactive release must be available.B:INCORRECT: Plausible because with some Refueling requirements not met an additonal action is to secure crane movement with loads over fuel. This is not the case, however, with containment high range area radiation monitors.C:INCORRECT: Plausible because the magenta condition is the highest priority condition but both magenta and dark blue cause the radiation monitor to be non-functional.D:INCORRECT: Plausible because the magenta condition is the highest priority condition but both magenta and dark blue cause the radiation monitor to be non-functional. Also, with some Refueling requirements not met an additonal action is to secure crane movement with loads over fuel. This is not the case, however, with containment high range area radiation monitors.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given refueling conditon to determine the correct conditon of the monitors and action to take.SRO Importance:4.3NRC Reference Req'd

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Modified FromExam Bank No.:245410CFR Reference or SRO Objective:55.43(b)(2)Page 32 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMHow many Reactor Coolant Pumps (RCPs) must be in operation in Mode 2 AND the action required if the condition is not met? _____(1)_____ RCPs must be in operation. _____(2)_____ is the action to be taken. A. (1) Two (2) Place the required RCPs in operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Two (2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. C. (1) All (2) Place the required RCPs in operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> D. (1) All (2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Answer:D(1) All(2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.Exam Bank No.:2455Last used on an NRC exam:NeverSRO Sequence Number:92Page 33 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:003 G2.2.22Tier:2Group/Category:1Reactor Coolant Pumps:Knowledge of the limiting conditions for operations and safety limits.STP Lesson:LOT 503.01Objective Number:80056GIVEN a system scenario, DETERMINE the applicable Technical Specifications and/or Technical Requirements Manual (TRM) for the system and APPLY the specification(s).Attached Reference

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TS 3.4.1.1

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Source:NewDistractor JustificationA:INCORRECT: Plausible because two RCP in operation are required in Mode 3 with Reactor Trip Breakers closed. Also, plausible because the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is associated with the operablity of the loop only. Not if the loop is in operation. (RCP running) and is associated with Mode 3 Requiremeints.B:INCORRECT: Plausible because two RCP in operation are required in Mode 3 with Reactor Trip Breakers closed.C:INCORRECT: Plausible because the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is associated with the operablity of the loop only. Not if the loop is in operation. (RCP running) and is associated with Mode 3 Requiremeints.D:CORRECT: Mode 2 requires all RCPs in operation or be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.Question Level:FQuestion Difficulty3Justification:The student must have knowledge of the TS for Reactor Coolant Loops in Mode 2.SRO Importance:4.7NRC Reference Req'd

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Modified FromExam Bank No.:245510CFR Reference or SRO Objective:55.43(b)(2)Page 34 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: A plant transient caused a PZR PORV to open and close several times. PRT pressure is 25 psig. PRT temperature is 120ºF. PRT level is 76%. (1) would be the consequence if PRT parameters continue to rise. AND The Unit Supervisor will FIRST direct the operator to (2) the PRT per the appropriate section of 0POP02-RC-0001, Pressurizer Relief Tank and Reactor Coolant Drain Tank System Operation. A. (1) High flow in the GWPS (2) PUMP DOWN B. (1) PRT rupture disk failure (2) PUMP DOWN C. (1) High flow in the GWPS (2) VENT D. (1) PRT rupture disk failure (2) VENT Answer:D (1) PRT rupture disk failure(2) VENTExam Bank No.:2456Last used on an NRC exam:NeverSRO Sequence Number:93Page 35 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:007 A2.02Tier:2Group/Category:1Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:Abnormal pressure in the PRT.STP Lesson:LOT 201.04Objective Number:91014DESCRIBE the overpressure protection scheme for the PRTAttached Reference

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LOT 201.04 Lesson Plan on the PRT and RCDT and 0POP02-RC-0001, PRT and RCDT System Operation

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the PRT can be vented to the GWPS but is not normal lined up at power. Also, plausible because the PRT HI/LO level alarm would be in alarm with the PRT HI pressure alarm but the PRT Pressure has to be lowered first to prevent lifting the suction relief on the RCDTpumps when lowering level.B:INCORRECT: Plausible because the PRT HI/LO level alarm would be in alarm with the PRT HI pressure alarm but the PRT Pressure has to be lowered first to prevent lifting the suction relief on the RCDTpumps when lowering level.C:INCORRECT: Plausible because the PRT can be vented to the GWPS but is not normal lined up at power.D:CORRECT: With the given conditions the potential impact is to the PRT rupture disk and the PRT pressure must be reduced to below 15 psig and then the procedure would have the operators pump down the PRT. Venting the PRT and Pumping down the PRT are seprate section in 0POP02-RC-0001, PRT and RCDT System Operation. Also with both the PRT HI pressure and the PRT HI/LO level alarms in the SRO has to evaluate which to address first.Question Level:HQuestion Difficulty3Justification:The student must utilize their knowledge of the design features of the PRT and with the conditions provided recognize that pressure in the PRT must be lowered prior to cooling.SRO Importance:3.2NRC Reference Req'd

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Modified FromExam Bank No.:245610CFR Reference or SRO Objective:55.43(b)(5)Page 36 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit has experienced a Reactor Trip and Safety Injection due to a Small Break LOCA has occurred with the following conditions: The Containment Normal Sump is pumping to the Waste Holdup Tank. RCP 1A has a low seal leakoff flow. A 'CCW SURGE TK LVL LO' annunciator is in. A 'TURB L.O. RSVR LVL LO-LO' annunciator is in. Which condition should the Unit Supervisor address FIRST? A. Ensure the Containment Sump penetration is isolated per Addendum 5 of 0POP05-EO-EO00, Reactor Trip or Safety Injection. B. Trip RCP 1A per 0POP04-RC-0002, Reactor Coolant Pump Off Normal. C. Restore the CCW Surge Tank per 0POP04-CC-0001, Component Cooling Water System Leak. D. Restore Turbine Lube Oil Reservoir level per 0POP09-AN-07M3/D-5, 'TURB L.O. RSVR LVL LO-LO.' Answer:A Ensure the Containment Sump penetration is isolated per Addendum 5 of 0POP05-EO-EO00, Reactor Trip or Safety Injection.Exam Bank No.:2457Last used on an NRC exam:NeverSRO Sequence Number:94Page 37 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:013 G2.4.8Tier:2Group/Category:1Engineered Safety Features Actuation:Knowledge of how abnormal operating procedures are used in conjunction with EOPs.STP Lesson:LOT 504.04Objective Number:92283GIVEN a set of conditions and the occurrence of a Red, Orange or Yellow path CSF, STATE the action required per 0POP01-ZA-0018, EOP User's Guide.Attached Reference

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LOT 504.04 Lesson Plan and 0POP01-ZA-0018, EOP User's Guide

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Source:NewDistractor JustificationA:CORRECT: Containment Isolation is an Engineered Safety Features system. Off-normal procedures can only be performed in parrallel with EOPs if they don't preclude taking action of an EOP. The failure of the Containment Penetration to isolate would have to be addressed first.B:INCORRECT: Plausible because tripping a RCP when needed is very important.C:INCORRECT: Plausible because CCW supports the cooling of the ECCS but is still addressed with an off-normal procedure.D:INCORRECT: Plausible because ensuring the main turbine can coast down without damage is very important and is a Lessons Learned at STPNOC.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of how off-normal procedures work with EOPs.SRO Importance:4.5NRC Reference Req'd

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Modified FromExam Bank No.:245710CFR Reference or SRO Objective:55.43(b)(5)Page 38 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power with the following conditions: Train A and Train B RCFCs are running. Train A CCW Pump is running. Subsequently: 13.8KV Standby BUS G loses power. Containment pressure will _____(1)_____ and the Unit Supervisor could use _____(2)_____ to help restore containment pressure. (1) (2) A. rise 0POP02-HC-0003, Supplemental Containment Purge B. lower 0POP02-HC-0003, Supplemental Containment Purge C. rise 0POP02-HC-0002, Normal Containment Purge D. lower 0POP02-HC-0002, Normal Containment Purge Answer:A rise - 0POP02-HC-0003, Supplemental Containment PurgeExam Bank No.:2458Last used on an NRC exam:NeverSRO Sequence Number:95Page 39 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:022 A2.04Tier:2Group/Category:1Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:Loss of service water.STP Lesson:LOT 202.33Objective Number:51319STATE the power supplies for the RCB-HVAC systems.Attached Reference

Reference:

LOT 202.33 Lesson Plan and 0POP02-HC-0002 and 0POP02-HC-0003

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Source:NewDistractor JustificationA:CORRECT: With the given conditions, a loss of 13.8 KV Standby Bus 1G will cause the 4.16 KV ESF Bus 'B' to lose power. This in turn causes a loss of cooling water (chill water) flow to Train 'A' and 'B' RCFCs. ESF DG #12 will start and load the Bus with a Mode II signal. The Mode II signal starts CCW Pump 1B but does not restore any cooling water (CCW or chill water) flow to Train 'A' or 'B' RCFCs. Thus containment pressure will rise. The US would eventually restore cooling water flow but could also use 0POP02-HC-0003, Suplemental Containment Purge to help restore containment pressure.B:INCORRECT: Plausible because CCW Pump 1B would start on the Mode II signal and that could be viewed as additional flow for the RCFCs that would lower containment pressure.C:INCORRECT: Plausible but 0POP02-HC-0002, Normal Containment Purge, is only used in Modes 5 and 6 when the Unit is shutdown.D:INCORRECT: Plausible because CCW Pump 1B would start on the Mode II signal and that could be viewed as additional flow for the RCFCs that would lower containment pressure.Also, 0POP02-HC-0002, Normal Containment Purge, is only used in Modes 5 and 6 when the Unit is shutdown.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given conditions to determine the impact of lossing cooling flow to containment and decide which procedure to use based on the Unit staying on line in Mode 1. In addition the SRO has to have TS knowledge to determine which procedure to use.SRO Importance:3.2NRC Reference Req'd

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Modified FromExam Bank No.:245810CFR Reference or SRO Objective:55.43(b)(5)Page 40 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit was at 100% power when a Loss of Offsite Power occurred with the following conditions: All ESF DG started and loaded normally All SG Pressures 1200 psig and stable The impact to the RCS due to the Loss of Offsite Power is that there will be less cooling to the _____(1)_____. AND The Unit Supervisor will direct a cooldown rate of between ______(2)_____ to control the cooldown of the RCS? A. (1) Reactor Vessel Head (2) 15ºF/hour and 20ºF/hour. B. (1) Reactor Vessel Head. (2) 35ºF/hour and 50ºF/hour. C. (1) Reactor Coolant Pump seals. (2) 15ºF/hour and 20ºF/hour. D. (1) Reactor Coolant Pump seals. (2) 35ºF/hour and 50ºF/hour. Answer:B (1) Less cooling to the Reactor Vessel Head.(2) 35ºF/hour and 50ºF/hour.Exam Bank No.:2459Last used on an NRC exam:NeverSRO Sequence Number:96Page 41 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:002 A2.03Tier:2Group/Category:2Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:Loss of forced circulation.STP Lesson:LOT 504.25Objective Number:92228STATE the basis for maximum cooldown rate associated with natural circulation cooldown.Attached Reference

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LOT 504.25 Lesson Plan and 0POP05-EO-ES02, Natural Circulation Cooldown.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because a cooldown rate of 15 to 20 degrees F/hour would be used if one or more SGs were faulted.B:CORRECT: Loss of forced circulation causes less cooling to the Reactor Vessel Head which can cause a bubble to form in the head during cooldown. The cooldown is limit is based on the condition of the SGs. If all SGs are intact and active then the cooldown limit is 35 to 50 degrees F/hour.C:INCORRECT: Plausible because cooling to the RCP seals is a concern to the RCS because the seals are a part of the RCS pressure boundary. If this were a loss of ALL AC power seal cooling would be lost and a LOCA through the seals is likely. Also, at the beginning of the procedure for a natural circulation cooldown it does have the operator check RCP seal cooling but it is only for a restart of the RCP if it were to become available. Also plausible because a cooldown rate of 15 to 20 degrees F/hour would be used if one or more SGs were faulted.D:INCORRECT: Plausible because cooling to the RCP seals is a concern to the RCS because the seals are a part of the RCS pressure boundary. If this were a loss of ALL AC power seal cooling would be lost and a LOCA through the seals is likely. Also, at the beginning of the procedure for a natural circulation cooldown it does have the operator check RCP seal cooling but it is only for a restart of the RCP if it were to become available.Question Level:HQuestion Difficulty3Justification:The student must analyze the given conditions and have fundamental knowledge of the affect of losing forced circulation and the procedural actions from 0POP05-EO-ES02, Natural Circulation Cooldown.SRO Importance:4.3NRC Reference Req'd

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Modified FromExam Bank No.:245910CFR Reference or SRO Objective:55.43(b)(5)Page 42 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMA Loss of Coolant Accident has occurred with the following: The crew is performing 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant. RCS pressure is 550 psig and slowly lowering. Pressurizer level is 90% and slowly rising. RWST level is 70,000 gallons and slowly lowering. Containment pressure is 9.6 psig and slowly rising. All LHSI pumps are running. The Unit Supervisor can validate the LHSI Control Board indications observed by determining the LHSI Pumps _____(1)_____ and the Unit Supervisor should transition to _____(2)_____. (1) (2) A. ARE providing flow in to the RCS because PZR level is rising 0POP05-EO-ES13, Transfer to Cold Leg Recirculation B. are NOT providing flow into the RCS because RCS pressure is too high 0POP05-EO-ES13, Transfer to Cold Leg Recirculation C. ARE providing flow in to the RCS because PZR level is rising 0POP05-EO-FRZ1, Response to High Containment Pressure D. are NOT providing flow into the RCS because RCS pressure is too high 0POP05-EO-FRZ1, Response to High Containment Pressure Answer:Bare NOT providing flow into the RCS because RCS pressure is too high - 0POP05-EO-ES13, Transfer to Cold Leg RecirculationExam Bank No.:2460Last used on an NRC exam:NeverSRO Sequence Number:97Page 43 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.1.45Tier:3Group/Category:1Ability to identify and interpret diverse indications to validate the response of another indication.STP Lesson:LOT 504.09Objective Number:81103Fom memory STATE/IDENTIFY the criteria on the conditional information page of 0POP05-EO-EO10 to include operator response, initiating parameter(s) and values.Attached Reference

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LOT 504.09 Lesson Plan and 0POP01-ZA-0018, EOP User's Guide

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Source:NewDistractor JustificationA:INCORRECT: Plausible because PZR level rising can be an indication of added inventory but it can also be an indication of a PZR Vapor space leak.B:CORRECT: With the given conditions RCS pressure is too high for LHSI flow to the RCS. When conditions exist for transfer to cold leg recirculation then that procedure is the priority.C:INCORRECT: Plausible because PZR level rising can be an indication of added inventory but it can also be an indication of a PZR Vapor space leak. Also, containment pressure is high enough for a transition to the containment integrity safety function but the transfer to cold leg recirculation takes priority.D:INCORRECT: Plausible because containment pressure is high enough for a transition to the containment integrity safety function but the transfer to cold leg recirculation takes priority.Question Level:HQuestion Difficulty3Justification:The student must be able to analyze the given condition and determine the response of the LHSI Pumps and the procedure that should be entered.SRO Importance:4.3NRC Reference Req'd

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Modified FromExam Bank No.:246010CFR Reference or SRO Objective:55.43(b)(5)Page 44 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMThe Unit is at 100% power. Subsequently: ECW Pump A discharge strainer stopped rotating. The strainer is causing an intermittent ICS alarm. To remove the ICS Alarm from ALARM CHECKING, enter the strainer ICS point in the _____(1)_____. The Technical Specification action is to restore ECW Pump A to OPERABLE status within

_____(2)_____ days. (1) (2) A. "OAS" Log 28 B. "Points Off-Scan" Log 28 C. "OAS" Log 7 D. "Points Off-Scan" Log 7 Answer:D Points Off-Scan Log - 7Exam Bank No.:2462Last used on an NRC exam:NeverSRO Sequence Number:98Page 45 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.2.43Tier:3Group/Category:2Knowledge of the process used to track inoperable alarms.STP Lesson:LOT 503.01Objective Number:SRO92102Given the topic or title of a specification included in the Technical Specifications, or the Technical Requirements Manual (TRM), DESCRIBE the general requirements of the specification to include components or administrative requirements affected, limitations, major time frames involved, major surveillance in order to comply, and the bases for the specification.Attached Reference

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TS 3.7.4 and 0PGP03-ZO-0039, Configuration Mangement

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Source:NewDistractor JustificationA:INCORRECT: Plausible because the OAS Log would be used but it is for tracking the ECW Pump itself not the ICS Alarm. Also, there are some 28 day actions in TS such as for a motor driven Aux feedwater pump.B:INCORRECT: Plausible because there are some 28 day actions in TS such as for a motor driven Aux feedwater pump.C:INCORRECT: Plausible because the OAS Log would be used but it is for tracking the ECW Pump itself not the ICS Alarm.D:CORRECT: The ICS alarm would be tracked in the Points Off-Scan Log and the ECW Pump would have to be returned in 7 days. It would be tracked in the OAS Log.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of 0PGP03-ZO-0039 for dealing with ICS points taken off-scan and fundamental knowledge of actions in TSs.SRO Importance:3.3NRC Reference Req'd

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Modified FromExam Bank No.:246210CFR Reference or SRO Objective:55.43(b)(2)Page 46 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMA fire breaks out behind CP-005 in the Unit 1 Control Room causing a Reactor Trip. The Unit 1 Unit Supervisor's priority would be to use _____(1)_____ to respond. AND Using this procedure the Unit 1 Unit Supervisor will direct the _____(2)_____. (1) (2) A. 0POP04-ZO-0001, Control Room Evacuation Unit 2 Unit Supervisor announce the Control Room Evacuation B. 0POP04-ZO-0008, Fire/Explosion Unit 2 Unit Supervisor announce the Fire/Explosion C. 0POP04-ZO-0001, Control Room Evacuation Unit 1 Secondary RO announce the Control Room Evacuation D. 0POP04-ZO-0008, Fire/Explosion Unit 1 Secondary RO announce the Fire/Explosion Answer:C 0POP04-ZO-0001, Control Room Evacuation - Unit 1 Secondary RO announce the Control Room EvacuationExam Bank No.:2464Last used on an NRC exam:NeverSRO Sequence Number:99Page 47 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:G2.1.14Tier:3Group/Category:1Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.STP Lesson:LOT 505.01Objective Number:92106GIVEN plant conditions/symptoms, EVALUATE the conditions/symptoms and STATE whether or not the referenced procedure is to be used.Attached Reference

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LOT 505.01 Lesson Plan for Off-Normal procedures.

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Source:NewDistractor JustificationA:INCORRECT: Plausible because for a fire where the control room is evacuated the unaffected Unit does NOT make an announcement of the evacuation but they do make the announcement of the fire/explosion because they would implement 0POP04-ZO-0008 per 0POP04-ZO-0001, step 15..B:INCORRECT: Plausible because 0POP04-ZO-0008 is entered by the unaffected unit per 0POP04-ZO-0001, step 15.C:CORRECT: For a fire that requires a control room evacuation the affected Unit's Secondary RO will announce the control room evacuation. The affected Unit's Unit Supervisor will contact the unaffected Unit to enter 0POP04-ZO-0008. 0POP01-ZA-0018,EOP Users Guide, states that if the control room is evacuated then 0POP04-ZO-0001 SHALL take precedence over all EOPs and 0POP04-ZO-0008. Also, for a fire in the control room that has caused unexpected equipment actuation then the Unit Supervisor is required to enter 0POP04-ZO-0001 first.D:INCORRECT: Plausible because 0POP04-ZO-0008 is entered by the unaffected unit per 0POP04-ZO-0001, step 15. Also, in this case, 0POP04-ZO-0001 has the Unit Supervisor direct the Unit 1 secondary RO to announce the Control Room evacuation NOT the fire.Question Level:FQuestion Difficulty2Justification:The student must have fundamental knowledge of who makes announcements during a Fire/Explosion and Control Room Evacuation.SRO Importance:3.1NRC Reference Req'd

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Modified FromExam Bank No.:246410CFR Reference or SRO Objective:55.43(b)(5)Page 48 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMIn accordance with the STPNOC Operating License as listed in the UFSAR, conditions allowing release of liquid radwaste require a/an _____(1)_____ to alarm on high effluent radioactive and automatically _____(2)_____ to stop the liquid radwaste release. A. (1) Area Radiation Monitor (2) secure the effluent release pump B. (1) Area Radiation Monitor (2) divert the release to recirculation C. (1) Process Radiation Monitor (2) secure the effluent release pump D. (1) Process Radiation Monitor (2) divert the release to recirculation Answer:D (1) Process Radiation Monitor - (2) divert the release to recirculationExam Bank No.:2491Last used on an NRC exam:NeverSRO Sequence Number:100Page 49 of 50 3/4/2016Print DateSTP LOT-20.1 NRC SRO EXAMK/A Catalog Number:068 G2.2.38Tier:2Group/Category:2Liquid Radwaste:Knowledge of conditions and limitations in the facility license.STP Lesson:LOT 203.11Objective Number:92083State the purpose of the LWPS Effluent Radiation Monitor (RT-8038) and Discharge Divert Valve (FV-4077).Attached Reference

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LOT 203.11 Lesson Plan on LWPS and USFAR Section 11.2

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Source:NewDistractor JustificationA:INCORRECT: Plausible because are used to detect high radiation in system piping such as for Main Steam they are not used in any of th eprocess efflent release systems. Also, stopping the pump is plausible because this method is used to stop a release from the TGB Sump #1.B:INCORRECT: Plausible because are used to detect high radiation in system piping such as for Main Steam they are not used in any of th eprocess efflent release systems.C:INCORRECT: Stopping the pump is plausible because this method is used to stop a release from the TGB Sump #1.D:CORRECT: Description in the UFSAR matches the conditions - Process Radiation Monitor that diverts the release to recirculation.Question Level:FQuestion Difficulty3Justification:The student must have fundamental knowledge of conditions and limitations of the facility license.SRO Importance:4.5NRC Reference Req'd

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Modified FromExam Bank No.:249110CFR Reference or SRO Objective:55.43(b)(1)Page 50 of 50