ML15126A300

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2014 Prairie Island Nuclear Generating Plant Initial License Examination Proposed Written Examination - Senior Reactor Operator
ML15126A300
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 05/19/2014
From:
Region 3 Administrator
To:
Northern States Power Co, Xcel Energy
References
Download: ML15126A300 (53)
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2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR76. P8170L-002 136/015/017 AA2.10/3.7/3.7/3H/YES/P8100/1C3 AOP2/T.S. 3.4.5/2014 ILT NRC S76Given the following conditions: - Unit 1 is in Mode 3, HOT STANDBY. - 11 and 12 RCPs are running. - The RCP indications on Panel B (CVCS Letdown) are as follows: Question continued on next page.Page 1 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR76. P8170L-002 136/015/017 AA2.10/3.7/3.7/3H/YES/P8100/1C3 AOP2/T.S. 3.4.5/2014 ILT NRC S76Question continued on next page.Question continued from previous page. - The CC indications of Panel A (Component Cooling) are as follows: After completing the actions of the appropriate AOP, the Shift Supervisor willdeclare _____________________ INOPERABLE.ONLY the "A" RCS loopONLY the "B" RCS loopBOTH RCS LoopsNEITHER RCS LoopsA.B.C.D.Page 2 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for T.S. 3.4.5.Justifications:a. Incorrect. Plausible as seal injection is loss to the 11 RCP; however, 11 RCP will NOT be securedbecause seal cooling is not lost to 11 RCP as indicated by bearing temperatures and CC flows to 11 RCP.b. Correct. 12 RCP has lost seal cooling (Seal injection and CC to the bearings); therefore, 12 RCP willbe secured per 1C3 AOP2. Once 12 RCP is secured, the "B" RCS Loop is INOPERABLE per T.S. 3.4.5.c. Incorrect. Plausible if examinee incorrectly believes both RCPs will be stopped based on loss of sealinjection flow alone. This is incorrect per 1C12.1 AOP1.d. Incorrect. Plausible if examinee is not familiar or does not recognize 12 RCP has exceeded thebearing water temperature limit of 200F and determines NO RCPs need to be tripped at this time. K/A Number:015/017 Reactor Coolant Pump (RCP) MalfunctionsAA2.10:Ability to determine and interpret the following as they apply to the Reactor Coolant PumpMalfunctions (Loss of RC Flow):When to secure RCPs on loss of cooling or seal injectionTechnical Reference(s): 1C3 AOP2 page 4, 1C12 AOP1 page 4, TS LCO 3.4.5Proposed references to be provided to applicants during examination: NoneLearning Objective: P8170L-002 Obj. 3HQuestion Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 __ ___55.43 __2___Comments:Page 3 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR77. P8172L-001A 133/022 2.4.8/3.8/4.5/7B/YES/P8100/1E-0/1C12.1 AOP1/SWI O-10/2014 ILT NRC S77Given the following conditions: - The crew is performing 1C12.1 AOP1, Loss of RCP Seal Injection. - Unit 1 Reactor is tripped. - Both Unit 1 RCPs are tripped. - Immediate operator actions of 1E-0, Reactor Trip or Safety Injection, are complete. - CC flow to each RCP is 210 gpm. - The crew can NOT restore any Unit 1 Charging Pumps.The SS will...direct the Lead RO to perform Attachment L andenter 1C3 AOP2, Loss of RCP Seal Cooling.direct the Lead RO to perform Attachment L andenter 1C18 AOP1, Makeup or Boration of the RCS Using a Safety Injection Pump.transition to 1ES-0.1, Reactor Trip Recovery andenter 1C3 AOP2, Loss of RCP Seal Cooling.transition to 1ES-0.1, Reactor Trip Recovery andenter 1C18 AOP1, Makeup or Boration of the RCS Using a Safety Injection Pump.A.B.C.D.3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Incorrect. Plausible if examinee incorrectly believes SI will be actuated during E-0 in order to providemakeup to the RCS and incorrectly believes RCP seal cooling is lost.b. Incorrect. Plausible as the SS will enter 1C18 AOP1; however, the SS will NOT direct the Lead RO toperform Attachment L.c. Incorrect. Plausible as the SS will transition to 1ES-0.1; however, the SS will NOT enter 1C3 AOP2.d. Correct.Page 4 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:022 Loss of Reactor Coolant Makeup2.4.8:Knowledge of how abnormal operating procedures are used in conjunction with EOPsTechnical Reference(s): 1C12.1 AOP1 page 4, 1E-0 pages 4 -5, SWI O-10 pgs 5, 10, 14. Proposed references to be provided to applicants during examination: NoneLearning Objective: P8172L-001A Obj. 7BQuestion Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge __ __Comprehension or Analysis __X_10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 5 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR78. P8180L-003 058/025 AA2.07/3.4/3.7/7B/YES/P8100/1C15 AOP1 / AOP2/1C15 AOP3 / D2 AOP1/2014 ILT NRC S78Given the following conditions: - Unit 1 is in Mode 5. - ERCS DP is 151" and stable. - RVLIS is 100% and stable. - An out-plant operator is in the field performing a valve lineup in the Auxiliary Building. - 11 RHR Pump is in standby. - 12 RHR Pump is running with the following indications: - Discharge pressure is oscillating between 0 and 100 psig. - Flow to the RCS is oscillating between 0 and 400 gpm.The Shift Supervisor will enter...1C15 AOP1, RHR Flow Restoration,and stop 12 RHR pump.1C15 AOP2, Loss of Coolant Inventory with RHR in Operation,and isolate the leak.D2 AOP1, Loss of Coolant While In A Reduced Inventory Condition,and make up to the RCS.1C15 AOP3, RHR Operation without CR Instrumentation or Flow Control,and manually throttle CLOSE the 11/12 RHR HX Bypass Flow Valve.A.B.C.D.3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires the knowledge of diagnostic steps and decision points in the EOPsthat involve transitions to event specific sub-procedures or emergency contingency procedures.Justifications:a. Correct. Indications given are showing the 12 RHR Pump cavitating due to loss of suction.b. Incorrect. Plausible if examinee incorrectly believes a loss of level is what caused the loss of RHRflow; however, RVLIS and ERCS DP are stable.c. Incorrect. Plausible if examinee incorrectly believes the RCS is at reduced inventory; however, theRCS is NOT considered at reduced inventory until ERCS DP level is below 52.25 inches (corresponds to3 feet below the reactor vessel flange) per 1D2.d. Incorrect. Plausible as the 12 RHR pump is cavitating; however, NOT due to the flow control valvefailing open.Page 6 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:025 Loss of Residual Heat Removal System (RHRS)AA2.07:Ability to determine and interpret the following as they apply to the Loss of Residual HeatRemoval System:Pump cavitationTechnical Reference(s): 1C15 AOP1 pages 3 & 4, 1C15 AOP2 pages 3 & 4, 1C15 AOP3pages 3 & 4, D2 AOP1 pages 3 & 4, 1D2 page 18.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8180L-003 058 Obj. 7BQuestion Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 7 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR79. P8197L-012 219/E12 2.2.44/4.2/4.4/38/YES/P8100/ECA-2.1//2014 ILT NRC S79Given the following conditions: - Unit 1 has experienced a major secondary system break. - Both MSIVs are OPEN and can NOT be closed remotely. - The crew has entered 1ECA-2.1, Uncontrolled Depressurization of Both Steam Generators. - An Out-plant Operator closes 11 MSIV locally. - 11 SG pressure is 550 psig and RISING. - 12 SG pressure is 575 psig and LOWERING. - 11 SG WR level is 45% and stable. - 12 SG WR level is 48% and slowly lowering. - Secondary radiation is normal. - RWST level is 43% and slowly lowering. - RCS pressure is 1600 psig and slowly lowering.The Shift Supervisor will transition to...1ES-1.2, Transfer To Recirculation.1E-3, Steam Generator Tube Rupture.1E-2, Faulted Steam Generator Isolation.1E-1, Loss of Reactor or Secondary Coolant.A.B.C.D.Page 8 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. This isNOT a direct entry into a major EOP because the SS will have entered E-2, then transitioned to1ECA-2.1, and then transition back to E-2.Justifications:a. Incorrect. Plausible if examinee incorrectly believes the switchover criteria is 43% RWST level;however, switchover criteria is 33% RWST level.b. Incorrect. Plausible as 11 SG pressure rising is an indication of a SG Tube Rupture; however, duringa SG Tube Rupture, level would also rise.c. Correct. ECA-2.1 directs the operator to go to E-2 if one of the SGs pressures start to rise.d. Incorrect. Plausible as RCS pressure is lower than normal; however, transition from 1ECA-2.1 onlyoccurs if RCS is pressure is below 250 psig (no adverse containment).K/A Number:E12 Uncontrolled Depressurization of all Steam Generators2.2.44:Ability to interpret control room indications to verify the status and operation of a system, andunderstand how operator actions and directives affect plant and system conditions.Technical Reference(s): ECA-2.1 pages 5, 6, and information page.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8197L-012 Obj. 38Question Source: Bank # _______Modified Bank # _P8197L-012 046__ New _______Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 9 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR80. P8197L-011 105/055 EA2.03/3.9/4.7/7/YES/P8100/1ECA-0.0//2014 ILT NRC S80Given the following conditions: - The crew has entered 1ECA-0.0, Loss of All Safeguards AC Power. - Offsite power is NOT available. - Bus 15 is locked out. - D2 Diesel Generator is OOS. - The Bus 15 green load rejection lights are LIT. - The Bus 16 green load rejection lights are NOT LIT. - The Unit 2 Safeguard busses are energized from their respective diesels. - 1ECA-0.0 is provided.On which step of 1ECA-0.0 will the Shift Supervisor direct the Lead Reactor Operatorto restore power to a Unit 1 Safeguards Bus?step 6step 7step 10step 11A.B.C.D.2-RIEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Incorrect. Plausible if examinee incorrectly believes a Unit 1 source is available; however, since Bus15 is locked out, D2 OOS, and NO offsite power, there is NO unit 1 source available.b. Incorrect. Plausible as it is a common misconception to state Bus 16 is available; however, it is NOTavailable for sequencer loading. Therefore, ECA-2.1 directs operator to take isolate RCP seals and takeloads to pullout prior to energizing Bus 16 from Unit 2 to prevent block loading the Unit 2 Diesel.c. Incorrect. Plausible if examinee incorrectly believes a Unit 1 source is available.d. Correct. Since the load sequencers are NOT available, Unit 1 sources are NOT available, and Unit 2safeguards buses are energized, the crew will restore power on step 11.Page 10 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:055 Loss of Offsite and Onsite Power (Station Blackout)EA2.03:Ability to determine or interpret the following as they apply to a Station Blackout:Actions necessary to restore powerTechnical Reference(s): 1ECA-0.0 pages 5 - 12Proposed references to be provided to applicants during examination: All steps of 1ECA-0.0,but no background information.Learning Objective: P8140L-247 Obj. 7Question Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 11 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR81. P8197L-011 017/E04 2.4.18/3.3/4.0/A3/YES/P8100/E-0/ECA-1.2/2014 ILT NRC S81Given the following conditions: - A LOCA has occurred on Unit 1. - Containment pressure is 0.1 psig and stable. - RCS subcooling is 92°F and stable. - Total feed flow to SGs is 250 gpm and stable. - RCS pressure is 1840 psig and stable. - Pressurizer level is 0% and stable. - Auxiliary building radiation alarms are the ONLY radiation alarms occurring. - Additional equipment failures result in the break NOT being isolable from the RCS.The Shift Supervisor will transition from 1E-0 to ________________________ and theUnit will be cooled down to Cold Shutdown using _____________________________.1ECA-1.2, LOCA Outside Containment1ECA-1.1, Loss of Emergency Coolant Recirculation1ECA-1.2, LOCA Outside Containment1ES-1.1, Post-LOCA Cooldown and Depressurization1ES-0.2, SI Termination1ES-1.1, Post-LOCA Cooldown and Depressurization1ES-0.2, SI Termination1C1.3, Unit 1 ShutdownA.B.C.D.Page 12 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Correct. Since the LOCA is in the Auxiliary Building, transition is made directly to 1ECA-1.2 based onadverse radiation levels in the Auxiliary Building. The transition to 1ECA-1.1 is made because all RCSwater is going to the Auxiliary Building instead of sump B in containment; therefore, loss of recirccapability.b. Incorrect. Plausible as the SS will transition from 1E-0 to 1ECA-1.2; however, the unit will not becooled down using ES-1.1.c. Incorrect. Plausible if the examinee incorrectly believes SI termination criteria is met and incorrectlybelieves the unit will be cooled down using 1ES-1.1. SI termination criteria is not met because pressurizerlevel is below 7% and RCS pressure is below 2000 psig. During a normal small break LOCA (i.e. insidecontainment), the unit would be cooled down using 1ES-1.1.d. Incorrect. Plausible as the crew would transition to 1ES-0.2 and the unit would be cooled down using1C1.3; however, this procedure flow path would only be used in this situation if SI termination criteria wasmet. SI termination criteria is not met because pressurizer level is below 7% and RCS pressure is below2000 psig.K/A Number:E04 LOCA Outside Containment2.4.18:Knowledge of the specific bases for EOPsTechnical Reference(s): 1E-0 page 11, 1ECA-1.2 pages 1 -4.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8197L-011 Obj. A3Question Source: Bank # _P8197L-011 017_Modified Bank # _______ New _______Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 13 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR82. P8182L-003 030/036 2.4.4/4.5/4.7/3C/YES/P8100/C1.6 AOP1 /D5.2 AOP1/B17 / C17/2014 ILT NRC S82Given the following conditions: - Core reload refueling activities are in progress. - An irradiated fuel assembly is being lowered into the core with the HOIST JOG SWITCH. - The ENTERING CORE SLOW ZONE light has just extinguished. - The manipulator crane operator continues lowering the irradiated fuel assembly into the core, now using the HOIST CONTROL LEVER. - The hoist abruptly stops and the mast support tube is shaking noticeably. - The manipulator crane camera shows the fuel assembly is FULLY inserted. - The following indications are present: - ENTERING CORE SLOW ZONE light is OFF. - INTERMEDIATE CORE ZONE light is ON. - BOTTOM CORE SLOW ZONE light is OFF. - UNDERLOAD light is ON. - SLACK CABLE light is ON. - Gas bubbles are visible rising from the vicinity of the fuel assembly. The _________ _______ control signal has failed. The Containment SRO willimplement _________ __________ AND _________ _________.UNDERLOADD5.2 AOP1, Damaged Fuel AssemblyC1.6 AOP1, Containment EvacuationUNDERLOADD5.2 AOP1, Damaged Fuel AssemblyD5.2 AOP4, Spent Fuel Pool Area Evacuation-RefuelingBOTTOM CORE SLOW ZONED5.2 AOP1, Damaged Fuel AssemblyC1.6 AOP1, Containment EvacuationBOTTOM CORE SLOW ZONEC1.6 AOP1, Containment EvacuationD5.2 AOP4, Spent Fuel Pool Area Evacuation-RefuelingA.B.C.D.Page 14 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(7) Fuel handling facilities and procedures. This questionrequires knowledge of the Refuel Floor SRO responsibilities.Justification:a.Incorrect. This is the wrong control signal failure, since the underload light is illuminated, even thoughthe procedures are correct.b.Incorrect. This has both the wrong control signal failure and incorrect procedures to implement.c.Correct. This is the correct control signal failure and the correct procedures to be implemented.d.Incorrect. Although this is the correct control signal failure, the implemented procedures are notcorrect.K/A Number:036 Fuel Handling Accident2.4.4:Ability to recognize abnormal indications for system operating parameters that are entry levelconditions for emergency and abnormal operating procedures.Technical Reference(s): C1.6 AOP1 page 3, D5.2 AOP1 page 3, B17 pages 10 - 14, C17 page27.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8182L-003 Obj. 3CQuestion Source: Bank #: P8182L-003 030 Modified Bank #: New: Question History: Last NRC Exam: 2012 ILT NRC EXAM Question Cognitive Level: Memory or Fundamental Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41: 55.43: 7 Comments:Page 15 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR83. P8171L-007 071/037 AA2.10/3.2/4.1/6/YES/P8100/TS 3.4.14//2014 ILT NRC S83Given the following conditions: - Unit 1 is at 100% power. - The following RCS leakage indications were determined at 1100 on 8/4/14: - IDENTIFIED leakage is 9.1 gpm. - UNIDENTIFIED leakage is 0.8 gpm. - Primary to Secondary leakage is 432 gallons per day. - T.S. LCO 3.4.14 is provided.Technical Specification LCO 3.4.14 requires Unit 1 to be in Mode 5 by ________ on_______________.23008/5/1407008/6/1423008/7/1403008/8/14A.B.C.D.3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for T.S. 3.4.14.Justifications:a. Correct. Since primary to secondary leakage is 432 gallons per day, LCO 3.4.14 condition D is enteredrequiring the unit to be in Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.b. Incorrect. Plausible if examinee incorrectly believes ALL leakage combined meets TS 3.4.14 for >10gpm IDENTIFIED leakage; however, in this case IDENTIFIED leakage is 9.4 gpm total.c. Incorrect. Plausible if examinee incorrectly believes UNIDENTIFIED leakage is "unidentified" plus "prito sec" at 1.1 gpm AND incorrectly applies Cond B only; however, in this case "unidentified" leakage islimited to 0.8 gpm and Cond A & B would be entered if >1 gpm.d. Incorrect. Plausible if examinee incorrectly believes UNIDENTIFIED leakage is "unidentified" plus "pri tosec" at 1.1 gpm; however, in this case "unidentified" leakage is limited to 0.8 gpm.Page 16 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Statement: 037 Steam Generator Tube LeakAA2.10:Ability to determine and interpret the following as they apply to the Steam Generator TubeLeak:Tech-Spec limits for RCS leakageTechnical Reference(s): TS 3.4.14Proposed references to be provided to applicants during examination: TS 3.4.14Learning Objective: P8171L-007 Obj. 6 Question Source: Bank # ____Modified Bank # _____ New ___X____Question History: Last NRC Exam ____N/A___Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 ___ __55.43 __2___Comments:Page 17 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR84. P8182L-002 001/2.2.38/3.6/4.5/10C/YES/P8100/H4//2014 ILT NRC S84Given the following conditions: - Steam Generator blowdown is aligned to the river. - Secondary coolant specific activity is <0.01uCi/gram DOSE EQUIVALENT I-131. - R-21, CIRC WATER DISCH MONITOR, fails low and is declared inoperable. - Table 2.2 of H4, Offsite Dose Calculation Manual, is provided.The Shift Supervisor will ensure...flow rate is estimated every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.grab samples are collected and analyzed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.grab samples are collected and analyzed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.grab samples are collected and saved for weekly composition and analysis every12 hours.A.B.C.D.2-RIEXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for the Offsite DoseCalculation Manual.Justifications:a. Incorrect. Plausible if examinee incorrectly believes R-21 is a flow monitor for Steam GeneratorBlowdown; however, R-21 is a radiation monitor.b. Correct. Per Table 2.2 of H4, if the discharge canal monitor is inoperable than action 6 is required.c. Incorrect. Plausible if examinee incorrectly believes R-21 is used to measure Steam GeneratorBlowdown effluent; however, R-19 is used to measure blowdown effluent.d. Incorrect. Plausible if examinee incorrectly believes R-21 is the rad monitor used to measure TurbineBuilding Sump effluent; however, there are seperate rad monitors for this.Page 18 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:APE 059 Accidental Liquid Radwaste Release2.2.38:Knowledge of conditions and limitations in the facility license.Technical Reference(s): H4 pages 93 - 94.Proposed references to be provided to applicants during examination: Table 2.2 of H4Learning Objective: P8182L-002 Obj. 10CQuestion Source: Bank # _ X _Modified Bank # _ New _______Question History: Last NRC Exam ___N/A_______Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __2__Comments:Page 19 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR85. P8197L-012 225/E03 EA2.1/3.4/4.2/17/YES/P8100/1ES-1.1//2014 ILT NRC S85Given the following conditions: - A small break LOCA has occurred on Unit 1. - The crew is in 1E-1, Loss of Reactor or Secondary Coolant.Based on the following information: The Shift Supervisor will transition to ___________________________.1ES-0.2, SI Termination1ES-0.1, Reactor Trip Recovery1ECA-1.2, LOCA Outside Containment1ES-1.1, Post LOCA Cooldown and DepressurizationA.B.C.D.Page 20 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires the knowledge of diagnostic steps and decision points in the EOPsthat involve transitions to event specific sub-procedures or emergency contingency procedures.Justifications:a. Incorrect. Plausible if examinee incorrectly believes subcooling is sufficient and CTMT is NOT adverse;however, in this case CTMT is adverse, also RCS pressure and PRZR level are not sufficient to terminateSI.b. Incorrect. Plausible if examinee incorrectly believes a SI did NOT occur and a transition to 1ES-0.1 iswarranted.c. Incorrect. Plausible if examinee incorrectly believes the LOCA is outside containment; however, theLOCA is inside containment as indicated from Containment Pressure and water level.d. Correct. For the given conditions, per step 20 of 1E-1, a transition to 1ES-1.1 is appropriate.K/A Number:E03 LOCA Cooldown and DepressurizationEA2.1:Ability to determine and interpret the following as they apply to the (LOCA Cooldown andDepressurization):Facility conditions and selection of appropriate procedures during abnormal and emergencyoperations.Technical Reference(s): 1E-1 Info Page & page 11, 1ES-1.1 info page & pages 2 & 3.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8197L-012 Obj. 17Question Source: Bank # P8197L-012 225 Modified Bank # New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:Page 21 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR86. P8197L-014 116/003 2.1.7/4.4/4.7/50/YES/P8100/1FR-I.3//2014 ILT NRC S86Given the following conditions: - Both RCPs are secured. - Prior to both RCPs being secured, 12 RCP seal cooling was lost. - A status evaluation of 12 RCP has NOT been completed. - The crew is on step 8 of 1FR-I.3, Response to Voids in Reactor Vessel. - RCS pressure is 1185 psig. - RCS cold leg temperatures are 500°F. - RCS subcooling is 65°F. - Containment Pressure is 3.8 psig. - RVLIS Full Range is 70% and lowering. - Pressurizer level is 92%. - The water in the pressurizer is saturated. - 1FR-I.3 is provided.What is the NEXT action the Shift Supervisor will direct?Start 11 RCP.Start 12 RCP.Block Safety Injection.Dump steam as necessary.A.B.C.D.3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Correct.b. Incorrect. Plausible if examinee disregards Caution on top of page 6 and also because 12 RCP is thepreferred RCP.c. Incorrect. Plausible if examinee incorrectly believes containment is adverse and goes to step 12 perRNO on step 9a. However, containment is not adverse because it is below 5 psig.d. Incorrect. Plausible as step 14 will require dumping steam if subcooling is not greater than 70F;however, this would NOT be the next step performed.Page 22 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:003 Reactor Coolant Pump System (RCPS)2.1.7:Ability to evaluate plant performance and make operational judgments based on operatingcharacteristics, reactor behavior, and instrument interpretation Technical Reference(s): 1FR-I.3 pages 5 - 7.Proposed references to be provided to applicants during examination: 1FR-I.3 (no bases)Learning Objective: P8197L-014 Obj. 50Question Source: Bank # Modified Bank # New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:Page 23 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR87. P8180L-004 024/006 A2.13/3.9/4.2/4B/YES/P8100/1C18 AOP2//2014 ILT NRC S87Given the following conditions: - Unit 1 is in Mode 3, Hot Standby. - RCS Tavg is 380°F and stable. - RCS pressure is 400 psig and stable. - RCS subcooling is 70°F and stable. - PT-945, 1 CNTMT PRESS 1 NARROW RANGE, has failed HIGH. - When going to trip the bistable for PT-945, I&C inadvertently tripped the SI high pressure bistable for PT-947, 1 CNTMT PRESS 3 NARROW RANGE. - Containment pressure is 0 psig on all Control Room indications.The Shift Supervisor will enter _____________________________________ anddirect a Control Room Operator to ______________________________.1ES-0.2, SI Terminationstop SI pumps ONLY1ES-0.2, SI Terminationstop SI and RHR pumps1E-0, Reactor Trip or Safety Injectionperform Attachment L1C18 AOP2, Inadvertent Safety Injection When Shutdownplace running SI pumps in PULLOUTA.B.C.D.3-PEOEXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Incorrect. Plausible as this procedure would be used for inadvertent SI actuation after transition from1E-0; however, in this case 1E-0 entry conditions are NOT met and 1ES-0.2 would not be entered.b. Incorrect. Plausible if examinee incorrectly believes the RHR pumps should be secured and incorrectlybelieves 1ES-0.2 should be used.c. Incorrect. Plausible as this procedure would be used for inadvertent SI actuation at power; however, inthis case 1E-0 entry conditions are NOT met. d. Correct.Page 24 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Statement: 006 Emergency Core Cooling System (ECCS)A2.13:Ability to Ability to (a) predict the impacts of the following malfunctions or operations on theECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations:Inadvertent SIS actuationTechnical Reference(s): 1C18 AOP2 pages 2 & 3, 1E-0 page 2, 1ES-0.2 page 2.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8180L-004 Obj. 4B Question Source: Bank # Modified Bank # P8180L-004 024 New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:Page 25 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR88. P8180L-009H 048/022 2.2.22/4.0/4.7/9B/YES/P8100/T.S. 3.6.5 BASES//2014 ILT NRC S88Given the following conditions: - Unit 2 is at 100% power. - 21 & 23 CFCUs are running in SLOW and aligned to the DOME. - 22 & 24 CFCUs are running in FAST and aligned to the GAP/SUP CLG. - Containment Fan Coil Units (CFCUs) are being shifted per 2C19.2, Containment System Ventilation Unit 2. - 22 CFCU fails to start in SLOW speed. - 22 CFCU is re-started and running in FAST speed. - 23 CFCU fails to start in FAST speed. - 23 CFCU is re-started and running in SLOW speed.The 22 CFCU is _______________ AND the 23 CFCU is _______________. OPERABLEOPERABLEOPERABLEINOPERABLEINOPERABLEOPERABLE INOPERABLEINOPERABLEA.B.C.D.1-BEXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the knowledge of TS bases that is required to analyzeTS required actions and terminology.Justifications:a. Incorrect. Plausible as 23 FCU is operable based on slow speed capability; however, in this case 22FCU is NOT operable.b. Incorrect. Plausible if examinee incorrectly believes that fast speed operation is required by safetyanalysis instead of slow speed.c. Correct.d. Incorrect. Plausible as 22 FCU is inoperable; however, in this case 23 FCU is still operable because itcan fulfill its required slow speed function.Page 26 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:022 Containment Cooling2.2.22:Knowledge of limiting conditions for operations and safety limits.Technical Reference(s): T.S. 3.6.5 BasesProposed references to be provided to applicants during examination: NoneLearning Objective: P8180L-009H Obj. 9BQuestion Source: Bank #: Modified Bank #: ____ New: X__ Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41: 55.43: 2 Comments:Page 27 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR89. P8182L-002 136/039 A2.03/3.4/3.7/3J/YES/P8100/PINGP 1576//2014 ILT NRC S89Given the following conditions: Question continued on next page.Page 28 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR89. P8182L-002 136/039 A2.03/3.4/3.7/3J/YES/P8100/PINGP 1576//2014 ILT NRC S89Question continued on next page.Question continued from previous page. - Radiation levels are expected to remain as shown for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. - PINGP 1576, Emergency Classification Tables, is provided.Based ONLY on the ERCS information given, which of the following EAL classificationswill the Shift Manager declare?RU 1.2RU 2.2RA 1.2RS 1.1A.B.C.D.3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(6) Procedures and limitations involved in alterations in coreconfiguration. This question requires evaluating emergency classifications based on core conditions.Justifications:a. Incorrect. Plausible as conditions are met for this NUE; however, in this case the indications givenexceed the ALERT threshold.b. Incorrect. Plausible as conditions are met for this NUE; however, in this case the indications givenexceed the ALERT threshold.c. Correct.d. Incorrect. Plausible if examinee incorrectly believes indications given meet the SAE threshold.Page 29 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:039 Main and Reheat Steam System (MRSS)A2.03:Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences ofthose malfunctions or operations:Indications and alarms for main steam and area radiation monitors (during SGTR)Technical Reference(s): PINGP 1576Proposed references to be provided to applicants during examination: PINGP 1576Learning Objective: P8182L-002 Obj. 3JQuestion Source: Bank # Modified Bank # New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 6 Comments:Page 30 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR90. P8182L-002 137/073 A2.02/2.7/3.2/5C/YES/P8100/D5.1 AOP1//2014 ILT NRC S90Given the following conditions: - Both Units are at 100% power. - Fuel handling is in progress in the Spent Fuel Pool (SFP) area. - A fuel assembly is dropped. - R-25, Spent Fuel Pool Air Monitor A, fails low. - R-31, Spent Fuel Pool Air Monitor B, is in alarm.The Shift Supervisor will enter __________________________________ and direct _____________________________________________________.D5.1 AOP1, SFP Area Evacuation - Non-Refuelingraising the R-25 test current signal at the radiation monitor racks D5.1 AOP1, SFP Area Evacuation - Non-Refuelingplacing 122 Spent Fuel Special & 21 In-Service Purge Exhaust Fan in START D5.2 AOP4, SFP Area Evacuation - Refuelingraising the R-25 test current signal at the radiation monitor racks C47047 R-25, Spent Fuel Pool Air Monitor Aplacing 122 Spent Fuel Special & 21 In-Service Purge Exhaust Fan in START A.B.C.D.3-SPKEXPLANATION: This question is linked to 10 CFR 55.43(b)(7) Fuel handling facilities and procedures. This questionrequires knowledge of the Refuel Floor SRO responsibilitiesJustifications:a. Correct.b. Incorrect. Plausible as the SS will enter D5.1 AOP; however, the SS will not direct starting 122 SFPSpecial Fan.c. Incorrect. Plausible as the SS will direct raising R-25 test current signal; however, the SS will not enterD5.2 AOP4.d. Incorrect. Plausible as entering C47047 R-25 would occur; however, that procedure will not directstarting 122 SFP Special.Page 31 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:073 Process Radiation Monitoring (PRM) SystemA2.02:Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations:Detector failureTechnical Reference(s): D5.1 AOP1 pages 3 - 5, D5.2 AOP1 pages 1 -5, D5.2 AOP4 page 3,C47047 (R25) pages 1 & 2.Proposed references to be provided to applicants during examination: NoneLearning Objective: P8182L-002 Obj. 5CQuestion Source: Bank #: Modified Bank #: New: X Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41: 55.43: 7 Comments:Page 32 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR91. P8170L-006 058/011 A2.03/3.8/3.9/10C/YES/P8100/T.S. 3.3.1/C51/2014 ILT NRC S91Given the following conditions: - Unit 1 is at 100% power. - 1LT-428, Blue Channel Pressurizer LEVEL, fails LOW. - Actions per 1C51.3, Instrument Failure Guide, are in progress. - Bistables cannot be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. - T.S. LCO 3.3.1 is provided.Technical Specification LCO 3.3.1 requires Unit 1 thermal power to be reduced to...MODE 3 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.less than 10% in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.less than 10% in 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.A.B.C.D.3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for TS 3.3.1.Justifications:a. Incorrect. Plausible if examinee incorrectly believes T.S. LCO 3.3.1 does not apply and therefore, T.S.LCO 3.0.3 must be entered, which requires to be in Mode 3 in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.b. Incorrect. Plausible if examinee incorrectly misapplies T.S. and believes Condition E should beentered.c. Correct.d. Incorrect. Plausible if examinee incorreclty applies the NOTE for Condition K of LCO 3.3.1 andbelieves that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> may be added to the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> requirement.Page 33 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:011 Pressurizer Level Control System (PZR LCS)A2.03:Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS;and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences ofthose malfunctions or operations:Loss of PZR levelTechnical Reference(s): T.S. LCO 3.0.3, T.S. LCO 3.3.1Proposed references to be provided to applicants during examination: T.S. LCO 3.3.1Learning Objective: P8170L-006 Obj. 10CQuestion Source: Bank # _______Modified Bank # _____ New ___X___Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 __ ___55.43 __2___Comments:Page 34 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR92. P8184L-002 081/2.2.12/3.7/4.1/10D/YES/P8100/T.S. 3.3.1/SP-1198/2014 ILT NRC S92Given the following conditions: - SP-1198, NIS Power Range Startup Test, is being performed. - N42, PR Nuclear Instrument, trip function setpoint was found set at 37%. - The as left setpoint was recorded as 25.2%. - T.S. LCO 3.3.1 and SP 1198 are provided.What is the status of N42 and SP-1198? N42 SP-1198 AS FOUND ACCEPTANCE OPERABILITY CRITERIA OPERABLE MET OPERABLE NOT MET INOPERABLE MET INOPERABLE NOT META.B.C.D.3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of surveillance requirements.Justifications:a. Incorrect. Plausible as the as found status of N42 is operable; however, the acceptance criteria forN42 as left is not met.b. Correct.c. Incorrect. Plausible if examinee incorrectly believes as found status must be below 25% and thereforeis inoperable. Also, if examinee incorrectly believes that as left must be below 40% and the acceptancecriteria is therefore met.d. Incorrect. Plausible if the examinee incorrectly believes the as found and as left status must be below25%.Page 35 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:015 Nuclear Instrumentation2.2.12Knowledge of surveillance procedures.Technical Reference(s): T.S. 3.3.1, T.S. 3.3.1 Bases pgs 1-3, SP-1198.Proposed references to be provided to applicants during examination: T.S. 3.3.1, SP-1198.Learning Objective: P8184L-002 Obj. 10D.Question Source: Bank #: __________ Modified Bank #: P8184L-002 081 New: Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41: 55.43: 2 Comments:Page 36 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR93. P8182L-002 089/072 A2.01/2.7/2.9/10C/YES/P8100/TS LCO 3.3.3//2014 ILT NRC S93Given the following conditions: - Both units are at 100% power. - 1R-48, U1 CNTMT HI RNG AREA MON B, is OOS for the past 10 days. - T.S. LCO 3.3.3 Condition A was entered 10 days ago. - 1R-49, U1 CNTMT HI RNG AREA MON A, fails LOW due to a loss of power. - T.S. LCO 3.3.3 is provided.The Shift Supervisor will enter Technical Specification LCO 3.3.3 Condition ____ .C D

H IA.B.C.D.3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for TS 3.3.3.Justifications:a. Incorrect. Plausible if examinee incorrectly believes that a second channel inoperable causesCondition A required action not to be met.b. Correct. Per Table 3.3.3-1, two Containment High Range Area Monitors are required to be operable. With two required channels inoperable, Condition D is entered with required action of restoring onechannel to operable status within 7 days.c. Incorrect. Plausible if examinee implements T.S. 3.3.3 similiar to TS 3.3.1 and 3.3.1 by going to thetable first and entering the condition listed in the table (in this case "I") AND uses containment pressureinstead of containment area radiation.d. Incorrect. Plausible if examinee implements T.S. 3.3.3 similiar to TS 3.3.1 and 3.3.1 by going to thetable first and entering the condition listed in the table (in this case "I").Page 37 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:072 Area Radiation Monitoring (ARM) SystemA2.01:Ability to (a) predict the impacts of the following malfunctions or operations on the ARMsystem- and (b) based on those predictions, use procedures to correct, control, or mitigate theconsequences of those malfunctions or operations:Erratic or failed power supplyTechnical Reference(s): T.S. LCO 3.3.3Proposed references to be provided to applicants during examination: T.S. LCO 3.3.3Learning Objective: P8182L-002 Obj. 10CQuestion Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __2___Comments:Page 38 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR94. P8184L-002 103/2.1.7/4.4/4.7/10C/YES/P8100/T.S. 3.2.4//2014 ILT NRC S94Given the following conditions: - Unit 1 is at 96% power. - Quadrant Power Tilt Ratio (QPTR) is 1.05. - T.S. LCO 3.2.4 is provided.Technical Specification LCO 3.2.4 requires thermal power to bereduced to less than _____ within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of QPTR determination.91%87%85%81%A.B.C.D.3 - SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solelyknowing < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS actions OR by solely knowing the LCO "above the line" information OR by solelyknowing TS Safety Limits. The question requires the application of required actions for TS 3.2.4.Justifications:a. Incorrect. Plausible if examinee incorrectly subtracts 1.02 from 1.05 instead of 1.00 from 1.05. 100 - [3(3)]% = 91%b. Incorrect. Plausible if examinee incorrectly reduces power by the correct amount but from the currentpower level instead of RTP and incorrectly subtracts 1.02 from 1.05 instead of 1.00 from 1.05. 96% - [3(3)]% = 87%c. Correct. 100% - [5(3)]% = 85%d. Incorrect. Plausible if examinee incorrectly reduces power by the correct amount but from the currentpower level instead of RTP. 96% - [5(3)]% = 81%Page 39 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:Conduct of Operations2.1.7Ability to evaluate plant performance and make operational judgments based on operatingcharacteristics, reactor behavior, and instrument interpretation.Technical Reference(s): T.S. 3.2.4Proposed references to be provided to applicants during examination: T.S. 3.2.4Learning Objective: P8184L-002 Obj. 10CQuestion Source: Bank #: Modified Bank #: P8184L-002 099 New: Question History: Last NRC Exam: N/A Question Cognitive Level: Memory or Fundamental Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41: 55.43: 2 Comments:Page 40 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR95. P9150L-024 038/2.2.5/2.2/3.2/2/YES/P8100/FP-E-SE-03/FP-E-MOD-03/2014 ILT NRC S95Given the following conditions: - Unit 2 is at 100% power. - 2HD-4-3, SCAV STM FROM 1B MSR TO 25B FW HEATER, has a body to bonnet leak. - The leak cannot be stopped by torquing the body to bonnet studs. - Furmanite will perform an INJECTION LEAK SEAL to stop the body to bonnet leak. - 2HD-4-3 will be replaced during the Unit 2 scheduled outage. - The Unit 2 scheduled outage is in 135 days.Which of the following is REQUIRED to perform the Furmanite repair to 2HD-4-3? 50.59 Temporary Screening Modification NO NO NO YES YES NO YES YESA.B.C.D.1-FEXPLANATION: This question is linked to 10 CFR 55.43(b)(3) Facility licensee procedures required to obtain authority fordesign and operating changes to the facility. This question requires knowledge of 10CFR50.59 screeningprocess and the administrative process for temporary modifications.Justifications:a. Incorrect. Plausible if examinee is not familiar with the requirements for 50.59 screening and t-modsand incorrectly believes neither is required.b. Incorrect. Plausible as a T-Mod is required; however, a 50.59 screening is also required.c. Incorrect. Plausible as a 50.59 screening is required; however, a T-Mod is also required.d. Correct.Page 41 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:Equipment Control2.2.5Knowledge of the process for making design or operating changes to the facility. Technical Reference(s): FP-E-SE-03 page 19-23, FP-E-MOD-03 page 3.Proposed references to be provided to applicants during examination: NoneLearning Objective: P9150L-024 Obj. 2Question Source: Bank # Modified Bank # __________ New X Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments:Page 42 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR96. P7410L-002 048/2.2.44/4.2/4.4/3/YES/P8100/PINGP 1576/F3-2/2014 ILT NRC S96Given the following conditions: - PINGP 1576, Emergency Classification Tables, is provided.Based ONLY on the ERCS STAT screen above, what is the status of the followingfission product barriers? Fuel Cladding RCS Containment POTENTIAL LOSS LOSS INTACT POTENTIAL LOSS INTACT INTACT INTACT LOSS INTACT INTACT INTACT POTENTIAL LOSSA.B.C.D.Page 43 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(6) Procedures and limitations involved in alterations in coreconfiguration. This question requires evaluating emergency classifications based on core conditions.Justifications:a. Correct. A potential loss of fuel cladding barrier is indicated by the Core Cooling CSF being orangeand by RVLIS full range less than 40% with both RCPs stopped. A loss of RCS is indicated by subcoolingbeing less than 35F (containment is adverse because pressure is greater than 5 psig). Containment isintact because pressure is less than 46 psig. Also both trains of depressurization equipment (CFCU &CS) are operating.b. Incorrect. Plausible as there is a potential loss of fuel cladding; however, the RCS is NOT intact.c. Incorrect. Plausible as there is a loss of RCS; however, fuel cladding is NOT intact.d. Incorrect. Plausible if examinee incorrectly believes RVLIS level is ok because it is green instead ofred on the STAT screen and misses that Core Cooling is orange. Also, if examinee does not notice theRCS is at saturation indicating a LB LOCA condition. Also, if examinee incorrectly believes containmentis potentially loss because containment pressure is above 23 psig; however, this will only causecontainment to be potentially loss if there is less than one full train of depressurization equipment running. According to the STAT screen in this case, both trains are operating.K/A Number:Equipment Control2.2.44:Ability to interpret control room indications to verify the status and operation of a system, andunderstand how operator actions and directives affect plant and system conditions.Technical Reference(s): PINGP 1576, F3-2Proposed references to be provided to applicants during examination: PINGP 1576Learning Objective: P7410L-002 Obj 3Question Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 __ ___55.43 __6___Comments:Page 44 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR97. P8182L-001C 137/2.3.6/2.0/3.1/6/YES/P8100/H4 ODCM/C21.3-10.1/2014 ILT NRC S97Given the following conditions: - Preparations for a gaseous radioactive waste release from 121 Low Level Gas Day Tank are in progress. - C21.3-10.1, Releasing Radioactive Gas from 121 Low Level Gas Decay Tank, is provided.The Shift Supervisor will _________________ the releasebecause _______________________________________.NOT approveprecipitation is occurringNOT approvewind speed is greater than 10 mphapprovewind direction is from 110°approvecooling towers are not in operationA.B.C.D.Page 45 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(4) Radiation hazards that may arise during normal andabnormal situations, including maintenance activities and various contamination conditions. This questionrequires knowledge of the process for gaseous release approvals.Justifications:a. Incorrect. Plausible if examinee incorectly believes that precipitation is occurring.b. Incorrect. Plausible as wind speed is greater than 10 mph; however, the permit would be approved.c. Correct. Four limits apply in C21.3-10.1: Permit SHALL NOT be approved if ALL 3 of the following aremet: 1) ANY CT in operation 2) Wind direction between 330-360 or 0-60 AND 3) wind speed < 10 mph;also 4) permits should NOT be approved if precipitation is occurring.With the given information the examinee must identify that CTs are in operation based on U1 in Mode 1during the April - Oct summer months, wind direction is from 110, wind speed is 15 mph, AND it is NOTraining.d. Incorrect. Plausible as the permit will be approved based on wind speed and direction; however,cooling towers are in operation but would not be a factor.K/A Statement: Radiation Control2.3.6 Ability to approve release permits.Technical Reference(s): H4 ODCM pages 32 & 33, C21.3-10.1 page 3.Proposed references to be provided to applicants during examination: C21.3-10.1Learning Objective: P8182L-001C Obj. 6Question Source: Bank # __P8182L-001C 005__Modified Bank # _______ New _______Question History: Last NRC Exam ___N/A_____Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 _4___Comments:Page 46 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR98. P8197L-013 114/2.3.14/3.3/4.0/7/YES/P8100/1E-3 STEP 7 BASES//2014 ILT NRC S98Given the following conditions: - The crew is on step 7, Initiate RCS Cooldown, of 1E-3, Steam Generator Tube Rupture.The Shift Supervisor will direct the crew to initiate RCS cooldown by releasing steamvia the ________________________ from the _____________ steam generatorbecause this path will _____________________________________________. Condenser Steam Dumprupturedprevent pressurizing the steam generator to RCS pressureCondenser Steam Dumpintactminimize radiological releasePORVrupturedprevent radiological contamination of the Main Steam systemPORVintactensure adequate capacity to cooldown the RCSA.B.C.D.1-BThis question is linked to 10 CFR 55.43(b)(4) Radiation hazards that may arise during normal andabnormal situations, including maintenance activities and various contamination conditions. This questionrequires knowledge interpretation of radiation and activity readings as they pertain to selection ofemergency procedures.Justifications:a. Incorrect. Plausible as steam will be released from the ruptured steam generator as needed; however,in this case it is to prevent overpressurization of the SG. The intact SG would be used for RCS cooldown.b. Correct.c. Incorrect. Plausible as the ruptured steam generatorwill maintain pressure using PORVs as neededAND radiological concerns are appropriate;however, in this case PORVs are to preventoverpressurization of the SG and cooldown would be from the intact SG to condenser. d. Incorrect. Plausible as the intact steam generator is the appropriate choice for the cooldown; however,in this case the PORVs would not be used; also if examinee incorrectly believes capacity of the steamdumps is not adequate.Page 47 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Statement: Radiation Control2.3.14:Knowledge of radiation or contamination hazards that may arise during normal, abnormal, oremergency conditions or activities.Technical Reference(s): 1E-3 Step 7 Bases pages 5 & 6.Proposed references to be provided to applicants during examination: None Learning Objective: P8140L-242 Obj. 7Question Source: Bank #: P8197L-013 114 Modified Bank #: New: Question History: Last NRC Exam 2010 ILT NRC Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:Page 48 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR99. P8197L-011 116/2.4.19/3.4/4.1/7/YES/P8100/1ECA-0.0/SWI O-10/2014 ILT NRC S99Given the following conditions: - A loss of all AC power has occurred on Unit 1. - The crew is performing step 18 of 1ECA-0.0, Loss of All Safeguards AC Power. - Bus 15 is locked out. - Bus 16 is locked out. - Containment pressure is 0.6 psig and stable. - Containment radiation is 1.6 x 100 R/H. - 11 SG indications are as follows: - NR level is 6% and stable. - WR level is 58% and stable. - AFW flow is 50 gpm. - 12 SG indications are as follows: - NR level is 55% and rising rapidly. - WR level is 62% and rising. - AFW flow is 50 gpm. - 1ECA-0.0 is provided.The NEXT action the Shift Supervisor will direct is...establish Battery Room Cooling per step 19.isolate AFW flow to 12 SG per step 17.b RNO.raise total AFW flow to 200 gpm per step 17.a RNO.raise ONLY 11 SG AFW flow to 150 gpm per step 17.b.A.B.C.D.Page 49 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR3-SPREXPLANATION: This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question canNOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operatoractions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry toMAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategyof a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) andthen selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.Justifications:a. Incorrect. Plausible if examinee does not understand that step 17 is a continuous action step.b. Correct. Step 17 is annotated as a continous action step with a triangle. Since 12 SG level is risinguncontrollable, Step 17.b. RNO directs the operator to isolate AFW flow to the ruptured SG.c. Incorrect. Plausible if the examinee incorrectly believes that containment is adverse and usesAttachment E values for required SG levels in step 17.a.d. Incorrect. Plausible if examinee incorrectly believes 11 SG level is too low and level needs to beraised in 11 SG.K/A Number:Emergency Procedures / Plan2.4.19Knowledge of EOP layout, symbols, and icons.Technical Reference(s): 1ECA-0.0 pages 17 & 18, SWI O-10 page 10.Proposed references to be provided to applicants during examination: All steps of 1ECA-0.0,but no background information.Learning Objective: P8140L-247 Obj. 7Question Source: Bank # _______Modified Bank # _______ New __X_____Question History: Last NRC Exam ___N/A_________Question Cognitive Level: Memory or Fundamental Knowledge _____Comprehension or Analysis __X__10 CFR Part 55 Content: 55.41 _____55.43 __5___Comments:Page 50 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATOR100. P9150L-011 003/2.4.40/2.7/4.5/5/YES/P8100/F3-2//2014 ILT NRC S100Given the following conditions: - The Control Room observes indications of a LOCA at 0200. - The Shift Manager declares an Alert at 0205. - The PINGP-577 is completed at 0210. What are the LATEST ALLOWABLE notification times? States and Counties NRC 0215 0300 0220 0305 0220 0310 0225 0310A.B.C.D.1-FEXPLANATION: This question is linked to 10 CFR 55.43(b)(1) Conditions and limitations in the facility license. Thisquestion requires knowledge of government notification requirements per 10CFR50.72.Justifications:a. Incorrect. Plausible if examinee incorrectly believes the clock starts for notifications at time of accidentinstead of time of classification.b. Correct. 10CFR50, App. E requires state and local government notification to be made within 15minutes from time of emergency declaration. Also, 10CFR50.72 (a)(3) requires NRC notification to bemade immediately after the notification of the state and local governments and not later than one hourafter the emergency declaration.c. Incorrect. Plausible if examinee incorrectly believes the NRC notifications clock starts when thePINGP-577 is completed.d. Incorrect. Plausible if examinee incorrectly believes the offsite notification and NRC notification clockstarts when the PINGP-577 is completed.Page 51 2014 NRC INITIAL LICENSE WRITTEN EXAMSENIOR REACTOR OPERATORK/A Number:Emergency Procedures / Plan2.4.40Knowledge of SRO responsibilities in emergency plan implementation. Technical Reference(s): F3-2 pages 9 -11.Proposed references to be provided to applicants during examination: NoneLearning Objective: P9150L-011 Obj. 5Question Source: Bank # Modified Bank # P9150L-011 003 New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 Comments:You have completed the test!Page 52

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