Text

Initial Exam 2014-302 Draft SRO Written Exam
	ML15013A238
Person / Time
Site:	Harris
Issue date:	01/13/2015
From:	; NRC/RGN-II
To:	; Progress Energy Carolinas
Shared Package
ML15013A407	List: IR 05000400/2014302; ML15013A231; ML15013A235; ML15013A238; ML15013A241; ML15013A245; ML15013A249; ML15013A251; ML15013A262; ML15013A267; ML15013A271; ML15013A275;
References
	50-400/OL-14
	Download: ML15013A238 (72)
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QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

76. 2014 NRC SRO 001/NEW/C/A//CSFST, FR-C.1/NONE/EARLY/008AA2.30/SATGiven the following plant conditions: - The unit was operating at 100% power

- A Reactor trip and a Safety Injection occurred due to low PZR pressureCurrent conditions are: - All RCPs are off

- RVLIS Full Range is indicating 33% and stable

- PZR level is off scale high

- PZR pressure is 1360 psig and stable - Containment pressure is 4.2 psig and rising - All SG levels are 45% and stable - Core Exit T/Cs are 755°F and stable - RCS WR Hot Leg temperatures are 678°F and stableBased on these plant condtions, which ONE of the following completes the statementsbelow?A (1) break has occurred.To allow the injection of the RHR pumps, in accordance with EOP-FR-C.1, Responseto Inadequate Core Cooling, (2) is required to depressurize the RCS. (1) PZR steam space (2) opening of the Reactor Head Vent valves (1) PZR steam space (2) the depressuization of all intact SGs at maximum rate (1) RCS hot leg (2) opening of the Reactor Head Vent valves (1) RCS hot leg (2) the depressuization of all intact SGs at maximum rate A.B.C.D.Monday, September 29, 2014 7:41:35 AM

215 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

The RCS is superheated and in excess of 730

°F (Tc temps)and RCS inventory has lowered below 39% on RVLIS. The actions of EOP-FR-C.1 willdirect the operator to depressurize the SGs at maximum rate and as a result the RCS pressure will lower allowing the injection of the RHR pumps. The break is in the PRZsteam space as indicated by the pressurizer being off scale high.

A. Incorrect. The first part is correct. The second part is plausible since the RCS isvented by opening the Reactor Head Vent valves

in response to RCStemperature going above 1200

°F, however this is incorrect since the CETtemperature is stable at 755

°F this method to depressuize the RCS in thatsituation is not required.B. Correct.

C. Incorrect. The first part is plausible since RCS pressure has lowered however this isincorrect since the PZR level has not lowered to in dicate the leak locationis not in the PZR. The second part is plausible since the RCS is ventedby opening the Reactor Head Vent valves

in response to RCStemperature going above 1200

°F, however this is incorrect since the CETtemperature is stable at 755

°F this method to depressuize the RCS in thatsituation is not required.D. Incorrect. The first part is plausible since RCS pressure has lowered however this isincorrect since the PZR level has not lowered to in dicate the leak location is not in the PZR. Th e second part is correct.Monday, September 29, 2014 7:41:35 AM

216 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000008 Pressurizer Vapor Space Accident / 3008AA2.30; Ability to determine and interpret Inadequate core cooling as it applies tothe Pressurizer Vapor Space Accident. (Verified)(CFR 43.5 / 45.13)Importance Rating:

4.7Technical

Reference:

EOP-CSFST Rev. 11, Page 4References to be provided: NoneLearning Objective: EOP-LP-3.10, Objective 4.a Question Origin: Modified Bank OIT Development, 2009B Audit SRO 1, to create a new question with different second part and stem conditions.Comments: K/A match because individual must analyze plantconditions to determine location of break, determine thattemperature indications support superheated conditionsand that entry requirements for FR-C.1 have been met.Tier/Group: T1/G1 SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item.Monday, September 29, 2014 7:41:35 AM 217 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

77. 2014 NRC SRO 002/NEW/C/A//E-1, MCB TEXT/NONE//011EG2.2.44/Given the following plant conditions: - 'A' RHR pump is under clearance

- A Large Break LOCA occurred

- The crew is implementing EOP-E-1, Loss of Reactor or Secondary Coolant, and areevaluating if the 'B' RHR train is capable of cold leg recirculation - 1SI-341, Low Head SI train B to cold leg, red and green lights are not lit Which ONE of the following completes the statements below?

The position of 1SI-341 can be determined by checking the status of the (1)

. IF this redundant indication determines that 1SI-341 is OPEN, the CRS shouldtransition to (2)

. (1) monitor light box (MLB-3B) (2) ECA-1.1, Loss Of Emergency Coolant Recirculation (1) monitor light box (MLB-3B) (2) ES-1.3, Transfer To Cold Leg Recirculation (1) status light box (SLB-2)

(2) ECA-1.1, Loss Of Emergency Coolant Recirculation (1) status light box (SLB-2) (2) ES-1.3, Transfer To Cold Leg Recirculation A.B.C.D.Monday, September 29, 2014 7:41:35 AM 218 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: MONITOR LIGHT BOXES (MLBs) MLBs provide indicationsfor all pumps and valves which are required to function as a part of the EmergencySafeguard Features (ESF) System. E-1 Attachment 2 provides the list of componentsrequired to manually align the RHR system for cold leg recirculation. 1SI-341 is required to be in the open position to support the recirculation alignment. With 1SI-341open Train B is capable of cold leg recirculation and the crew should remain in E-1 until the RWST level lowers to 23.4% at which time the crew will transition to ES-1.3,Transfer To Cold Leg Recirculation A. Incorrect. The first part is correct. The second part is plausible since the RHR Apump is under clearance the Train A is not capable of Cold LegRecirculation with 1SI-341 not in the correct position not all of therequired components woul d be available to align the RHR system to ColdLeg Recirculation, however this is incorrect because 1SI341 is in thecorrect position.B. Correct.

C. Incorrect. The first part is plausible since the status light boxes provide MCBindications for various valves and dampers similarly to the MLBs, howeverthis is incorrect because the status light boxes primarily provideindications for secondary plant equipment and not ESF relatedequipment. The second part is plausible since the RHR A pump is underclearance the Train A is not capable of Cold Leg Recirculation with1SI-341 not in the correct position not all of the required componentswould be available to align the RHR system to Cold Leg Recirculation,however this is incorrect because 1SI341 is in the correct position. D. Incorrect. The first part is plausible since the status light boxes provide MCBindications for various valves and dampers similarly to the MLBs, howeverthis is incorrect because the status light boxes primarily provideindications for secondary plant equipment and not ESF relatedequipment. The second part is correct. Monday, September 29, 2014 7:41:35 AM

219 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000011 Large Break LOCA / 3011EG2.2.44; Ability to interpret control room indications to verify the status andoperation of a system, and understand how operator actions and directives affect plantand system conditions. (Verified)(CFR 41.5 / 43.5 / 45.12)

Importance Rating: 4.4 Technical

Reference:

E-1, page 16, 18 and 30, Rev. 0 MCB Student Text, page 10, Rev. 2 References to be provided: None Learning Objective: MCB Student Text, Objective 1.aQuestion Origin: NewComments: K/A match because individual determine which locationon the MCB will provide the redundant indication of the status for 1SI-341 and that position of the valve is in thecorrect position to support cold leg recirculation. Tier/Group: T1/G1 SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item.Monday, September 29, 2014 7:41:35 AM 220 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

78. 2014 NRC SRO 003/BANK/C/A//OST-1022, TS 3.3.3.6/NONE//025AG2.4.3/Given the following plant conditions: - The unit is cooling down and RHR pump 'A' was just placed in Shutdown CoolingModeSubsequently the following occurs: - PRZ level lowers rapidly

- PRZ pressure lowers rapidly

- RHR pump 'A' trips on overcurrent - RCS tempurature is 352°F and risingWhich ONE of the following completes the statements below?PRZ level indicators 460 and (1) are the Post Accident indicators that should beused to monitor PRZ level. Based on the conditions above, if ONE PRZ level channel of the Post Accidentindicators is NOT operable / available AND is unable to be restored the ACTIONrequired by Technical Specifation 3.3.3.6, Accident Monitoring Instrumentation, is to be in Hot Shutdown within (2) hours.(1) 461 (2) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (1) 461 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (1) 462 (2) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (1) 462 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A.B.C.D.Monday, September 29, 2014 7:41:35 AM 221 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer Analysis Reason answer is correct: OST-1022, Attachment 3 identifies PI-459, PI-460, andPI-461 satisfy Technical Specification surveillance 4.3.3.6.5 for the post accidentinstrumentation monitoring of the PRZ level. For the current conditions the plant is in Mode 3 Hot Standby. Technical Specification surveillance 3.3.3.6 action a states withthe number of OPERABLE accident monitoring instrumentation channels. except InCore Thermocouples and Reactor Vessel Level. less than the Total Required Numberof Channels requirements shown in Table 3.3-10 restore the inoperable channel(s) toOPERABLE status within 7 days. or be in at least Hot Standby within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sand in at least Hot Shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Because the unit is currentlyin Hot Standby the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance to reach that mode does not apply and the unitmust be placed in Hot Shutdown in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the level instrument was not restored to the operable status within 7 days.

A. Correct. B. Incorrect. The first part is correct.

The second part is plausible since action aallows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to achieve Hot Standby, however this is incorrect sincethe unit is currently in Hot Standby the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance to reach thatmode does not apply C. Incorrect. The first part is plausible since the RHR system was placed in Shutdowncooling and PI-462 is the Cold Calibrated PRZ level indicator, howeverthis is incorrect since PI-462 is calibrated for normal cold shutdownconditions not accident conditions. The second part is correct. D. Incorrect. The first part is plausible since the RHR system was placed in Shutdowncooling and PI-462 is the Cold Calibrated PRZ level indicator, howeverthis is incorrect since PI-462 is calibrated for normal cold shutdownconditions not accident conditions. The second part is plausible sinceaction a allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to achieve Hot Standby, however this is incorrectsince the unit is currently in Hot Standby the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance to reachthat mode does not apply Monday, September 29, 2014 7:41:35 AM

222 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000025 Loss of RHR System / 4025AG2.4.3: Ability to identify post-accident instrumentation. (Verified)(CFR 41.6 / 45.4)Importance Rating: 3.9 Technical

Reference:

OST-1022, Attachment 3 pg 22 Tech Specs 3.3.3.6 action aReferences to be provided: NoneLearning Objective: PRZ Student Text, Objective 11.d Question Origin: Bank, 2012 St. Lucie NRC Exam Question modified for HNP terminologyComments:

NoneTier/Group: T1/G1SRO Justification:

10 CFR Part 55 Content -

43(b)(2): Facility operatinglimitations in the Technical Specifications and their bases. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by testing the SRO knowlegde on the application of Limiting Condition for Operation (LCO)requirements (LCO 3.3.3.6)Monday, September 29, 2014 7:41:35 AM 223 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

79. 2014 NRC SRO 004/NEW/FUNDAMENTAL//AOP-025/NONE//056AA2.54/Given the following plant conditions: - The unit is operating at 100% power

- EDG 1B-SB is under clearance Subsequently, a loss of offsite power occursOne minute later DP-1A-SA is lost Which ONE of the following completes the statements below?Breaker indications are available on the MCB for (1)

.The procedural action required based on the current plant conditions is to dispatch anoperator to (2)

? (1) B-SB equipment ONLY (2) locally close breaker 106 (1) B-SB equipment ONLY (2) manually control EDG 1A-SA load / speed (1) BOTH A-SA and B-SB equipment (2) locally close breaker 106 (1) BOTH A-SA and B-SB equipment (2) manually control EDG 1A-SA load / speed A.B.C.D.Monday, September 29, 2014 7:41:35 AM 224 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

On a loss of DC control power, components lose remoteindication and control, so operators must use local indication and control. Therefore forthese conditions only the B-SB components will have remote indicati ons. Becauseoffsite power was lost first EDG 'A' will start and sequence on loads using program A.

After DP-1A-SA is lost the EDG 'A' will continue to run and the EDG will default tocontrol of the mechanical portion of the governor actuator (EGB-35P) on the enginewhich would result in the engine running at ~63 Hz if operating in the isochronomousmode on the emergency bus or in an overload condition if operating in parallel withoffsite power on the emergency bus. This will require taking manual control locally.

A. Incorrect. The first part is correct. The second part is plausible since offsite powerwas lost the candidate may misunderstand the expected sequence ofevents and and determine that the EDG 'A' started but the loss ofDP-1A-SA requires local operation of the output breaker (breaker 106) torestore power to in accordance with EOP-ECA-0.0, however this isincorrect because the EDG output breaker closes within 10 seconds of aloss of power (UV) condition to restore power to the Emergency Bus andthe DP-1A-SA power is lost 60 seconds after the loss of offsite powertherefore breaker 106 will be closed and the 6.9KV bus will be energized. B. Correct.

C. Incorrect. The first part is plausible since the breaker 480 volt MCC breakers localindications are powered which is powered by 2 phases of the 3 main linephases which remains available, however this is incorre ct since the MCBbreaker indications and control power are supplied by DC power. Thesecond part is plausible since offsite power was lost the candidate maymisunderstand the expected sequence of events and and determine thatthe EDG 'A' started but the loss of DP-1A-SA requires local operation ofthe output breaker (breaker 106) to restore power to in accordance withEOP-ECA-0.0, however this is incorrect because the EDG output breakercloses within 10 seconds of a loss of power (UV) condition to restorepower to the Emergency Bus and the DP-1A-SA power is lost 60 secondsafter the loss of offsite power therefore breaker 106 will be closed and the6.9KV bus will be energized. D. Incorrect. The first part is plausible since the breaker 480 volt MCC breakers localindications are powered which is powered by 2 phases of the 3 main linephases which remains available, however this is incorre ct since the MCBbreaker indications and control power are supplied by DC power. Thesecond part is correct. Monday, September 29, 2014 7:41:35 AM 225 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000056 Loss of Off-site Power / 6056AA2.54; Ability to determine and interpret breaker position (remote and local) as itapplies to the Loss of Offsite Power. (Verified)(CFR 43.5 / 45.13)Importance Rating: 3.0 Technical

Reference:

AOP-025, Section 3.3, page 45, Rev 38References to be provided: NoneLearning Objective: DCP Student Text, Objective 8 AOP-LP-3.25, Objecive 4Question Origin: NewComments:

NoneTier/Group: T1/G1SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answerthe written test item.Monday, September 29, 2014 7:41:35 AM 226 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

80. 2014 NRC SRO 005/NEW/C/A//AOP-017, OP-151.01/NONE//065AG2.2.44/Given the following plant conditions: - The plant is operating at 100% power - Instrument Air is aligned in SEQUENCE 2 with all Air Compressors availableThe course of the next minute the following indications are observed: - ALB-002-8-5, Computer Alarm Air Systems

- ALB-002-8-1, Instrument Air Low Press

- Instrument Air pressure is 73 psig and slowly loweringAOP-017, Loss Of Instrument Air, is being performed. An AO reports the instrument line to the Condenser Steam Dump valves on the westside of the condenser has ruptured. Which ONE of the following (1) describes the expected operation of the Instrument AirCompressors AND (2) identifies the Attachment that should be performed? (1) 1A and 1B running ONLY (2) Attachment 2, Positioning MCB Controllers (1) 1A and 1B running ONLY (2) Attachment 3, Reducing Instrument Air Header Loads (1) 1A, 1B, and 1C running (2) Attachment 2, Positioning MCB Controllers (1) 1A, 1B, and 1C running (2) Attachment 3, Reducing Instrument Air Header Loads A.B.C.D.Monday, September 29, 2014 7:41:35 AM 227 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

Computer Alarm Air Systems identifies an issue with the airsystem either system pressure is too high, too low or too many compressors arerunning. Service and Instrument Air Low Press indicate the system pressure is 75 psigand 1A and 1B air compressors will be started by CAS and 1C air compressor will startwhen system header pressure lowers to 101 psig. AOP-017, Attachment 3systematically reduces the air loads on the IA system based on im pact isolating eacharea will have on the plant.

A. Incorrect. The first part is plausible since 1C air compressor is not controlled by theCAS in sequence 2 and only 1A and 1B air compressors are controlled byCAS. The second part is plausible since the report from the field is thatthe airline to condenser steam dump valve has ruptured and the steamdump controller is repositioned to manual and 0% output as part ofAttachment 2, however this is incorrect because Attachment 2 is notimplemented until the Instrument Air system header pre ssure is below 35psig and the conditions currently have pressure at 70 psig. B. Incorrect. The first part is plausible since 1C air compressor is not controlled by theCAS in sequence 2 and only 1A and 1B air compressors are controlled byCAS. The second part is correct.C. Incorrect. The first part is correct. The second part is plausible since the report fromthe field is that the airline to condenser steam dump valve has rupturedand the steam dump controller is repositioned to manual and 0% outputas part of Attachment 2, however this is incorrect because Attachment 2is not implemented until the Instrument Air system header pressure isbelow 35 psig and the conditions currently have pressure at 70 psig. D. Correct. Monday, September 29, 2014 7:41:35 AM 228 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000065 Loss of Instrument Air / 8065AG2.2.44; Ability to interpret control room indications to verify the status andoperation of a system, and understand how operator actions and directives affect plantand system conditions. (Verified)(CFR 41.5 / 43.5 / 45.12)

Importance Rating: 4.4Technical

Reference:

AOP-017, Attachment 3, Page 57, Rev. 39 OP-151.01 Precaution and Limitation 6, Page 8, Rev. 87,References to be provided: None Learning Objective: Instrument and Service Air Objective 2.f AOP-LP-3.17, Rev. 5, Objective 4Question Origin:

NEWComments: NoneTier/Group: T1/G1SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-onlyQuestions, Rev 1 this question meets the SRO level ofknowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item.Monday, September 29, 2014 7:41:35 AM 229 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

81. 2014 NRC SRO 006/BANK/C/A///NONE//WE04EA2.1/Given the following plant conditions: - A Reactor Trip and Safety Injection has occured - EOP-E-0, Reactor Trip Or Safety Injection, is being implemented and SI has been resetThe current conditions are as follows: - RCS Pressure is 1500 psig and stable - PZR level is off scale low

- Subcooling is 3°F and improving

- Containment pressure 0.2 psig and rising - RM-1RR-3597, RHR Pump 1B, is in HIGH alarm and trending up - MLB-4A-SA-6-3 and MLB-4B-SB-6-3, RAB Equip C/D Sump Alert Lvl, status lights are lit - SG levels are: A = 23%, B = 24%, C = 15%

- Total AFW flow has been reduced to 215 KPPHBased on this information, which ONE of the following procedures will be implementedwhen exiting EOP-E-0?EOP-FR-H.1, Response to Loss of Secondary Heat SinkEOP-ECA-1.2, LOCA Outside ContainmentEOP-ES-1.1, SI TerminationEOP-ES-1.2, Post LOCA Cooldown and Depressurization A.B.C.D.Monday, September 29, 2014 7:41:35 AM 230 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

The transition to ECA-1.2 is correct. The Radiation Monitorin alarm and sump level alert lights indicate that the leak is in the B RHR Pump Room. LOCA outside containment. Transition to ECA-1.2 would occur at step 60 of E-0.

A. Incorrect.

Plausible since S/G levels are all less than 25%, which meet FR-H.1 entryconditions (Containment conditions normal), however this is not correctbecause total feed flow must also be less than 210 KPPH.B. Correct.

C. Incorrect.

Plausible if the candidate misundertands SI termination criteria anddetermines that positive subcooling satisfies the criteria (+ 3°F andimproving), however this is incorrect because PZR level must also beabove 10% and the current indication is offscale low which does not meetthe requirement for SI termination.D. Incorrect.

Plausible since this is the procedure that would be implemented for thequestion conditions if Auxiliary Building radiation levels were normal,however this is not correct because the RHR B pump room rad monitor isin High alarm and both channels of the RAB equipment drain sump are inAlert Alarm indicating that the sump level is rising.

Monday, September 29, 2014 7:41:35 AM

231 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal NotesW/E04 LOCA Outside Containment /3WE04EA2.1; Ability to determine and interpret facility conditions and selection ofappropriate procedures during abnormal and emergency operations as they apply tothe (LOCA Outside Containment). (Verified)(CFR 43.5 / 45.13)

Importance Rating: 4.3Technical

Reference:

EOP-E-0 Step 60References to be provided: None Learning Objective: EOP-LP-2.3/3.3 Objective 1.dQuestion Origin: Bank OIT Exam Bank, EOP-3.3 (01D) 1Comments:

NoneTier/Group: T1/G1SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SROlevel of knowledge by ensuring that the candidate mustknow procedural knowledge of diagnostic steps anddecision points in the emergency operating procedures(EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.Monday, September 29, 2014 7:41:35 AM

232 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

82. 2014 NRC SRO 007/NEW/FUNDAMENTAL//TS 3.1.3.1 BASES/TS 3.1.3.1//003G2.2.22/Given the following plant conditions: - The unit is operating at 100% power

- Control Bank D group step counters indicate 218 stepsSubsequently the following occurs: - Control Bank D Rod P-8 drops to 145 steps. - It will take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to re-align rod P-8 with the Control Bank D rodsWhich ONE of the following statements describes (1) the MAXIMUM Thermal powerlevel the unit is required to maintain in accordance with Tech Spec 3.1.3.1, MovableControl Assemblies - Group Height, One rod not within alignment limits, AND (2) theBasis for reducing Thermal Power?

(Reference provided) (1) 85% (2) Provides assurance of fuel rod integrity during continued operation. (1) 85% (2) Minimizes the effect of a Control Rod ejection accident. (1) 75% (2) Provides assurance of fuel rod integrity during continued operation. (1) 75%

(2) Minimizes the effect of a Control Rod ejection accident.

A.B.C.D.Monday, September 29, 2014 7:41:35 AM 233 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: A single Control Bank D rod has dropped to 145 steps andis > 12 steps out of alignment with its "Demanded position". TS 3.1.3.1 will require aDownpower to <75% within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if the rod cannot be restored to within alignmentwithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The bases of 3.1.3.1 states "Misalignment of a rod requiresmeasurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. Inaddition, those safety analyses affected by a misaligned rod are reevaluated to confirmthat the results remain valid during future operation." A. Incorrect. The first part is plausible since the Neutron High Flux setpoint must bereduced to 85% within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to comply with action d.3.d. However this is incorrect since the Thermal Power is required to bereduced below 75%. Th e second part is correct.B. Incorrect. The first part is plausible since the Neutron High Flux setpoint must bereduced to 85% within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to comply with action d.3.d. However this is incorrect since the Thermal Power is required to bereduced below 75%. The second plausible because one of the reasonsfor maintaining Control Rod alignment within limits is to ensure the effectsof a rod ejection accident are within analyzed limits. For this condition, ifthe rod ejection occurred on the dropped rod, the reactivity effects wouldbe more pronounced at 75% power than at 100% power. This is not thereason that Bases requires reducing power to < 75%.C. Correct.

D. Incorrect. The first part is correct. The second plausible because one of thereasons for maintaining Control Rod alignment within limits is to ensurethe effects of a rod ejection accident are within analyzed limits. For thiscondition, if the rod ejection occurred on the dropped rod, the reactivityeffects would be more pronounced at 75% power than at 100% power.This is not the reason that Bases requires reducing power to < 75%. Monday, September 29, 2014 7:41:35 AM

234 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000003 Dropped Control Rod / 1003G2.2.22; Knowledge of limiting conditions for operations and safety limits. (Verified)(CFR 41.5 / 43.2 / 45.2)Importance Rating: 4.7 Technical

Reference:

Technical Specification 3.1.3.1 Bases, page B3/4 1-4 (page 26) Technical Specification 3.1.3.1, page 3/4 1-15 (page 95)References to be provided: Technical Specification 3.1.3.1 Learning Objective: RODCS Student Text Objective 16.aQuestion Origin: NewComments:

NoneTier/Group: T1/G2SRO Justification:

10 CFR Part 55 Content - 43(b)(2) because applicationof Technical Specification below the line application of required actions considered to be SRO knowledge levelTechnical Specification items and the basis for the actions. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level ofknowledge by testing the SRO knowlegde on theapplication of Limiting Condition for Operation (LCO) requirements (LCO 3.1.3.1 and the basis for the actionstatement).Monday, September 29, 2014 7:41:35 AM 235 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

83. 2014 NRC SRO 008/NEW/C/A//AOP-009, OST-2044/NONE//060AG2.2.12/Given the following plant conditions: - A crane lift above the WGDTs is planned - A sample of the 'A' GDT had been obtained to release the tank later todayWhich ONE of the following completes the statements below?In accordance with OST-2044, for a planned crane lift above the WGDTs theMAXIMUM time allowed to complete the crane lift AFTER GDT sampling andactivity calculations are performed is within (1) hours. In accordance with PLP-114, the concern with the amount of radioactivity in theWaste Gas System is based on not exceeding a MAXIMUM of (2)

Remexposure to the public in the event of an uncontrolled release as a result of acrane mishap over the GDT.Note: Procedure titles are as follows -OST-2044, Radwaste Daily Operations Surveillance Test Modes: At All TimesPLP-114, Relocated Technical Specifications and Design Basis Requirements (1) 12 (2) 2 (1) 24 (2) 2 (1) 12 (2) 0.5 (1) 24 (2) 0.5 A.B.C.D.Monday, September 29, 2014 7:41:35 AM 236 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

The time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is for the Admin Limit for Crane Lifts Above the Gas Decay Tanks (GDT) after performing the total curie calculations. Thesecalculations are required after sampling of a GDT. The concern is that something beinglifted by the crane could drop on the GDTs and cause the tanks to rupture accidentlyreleasing the contents of the tanks. This check of quantity of radioactivity contained ineach gas storage tank will met the limit required in PLP-114 Attachment 5 to ensurethat in the event of an unco ntrolled release of the tank's contents the resulting wholebody exposure to a member of the public will not exceed 0.5 rem. Because of the potential for an accidental release sampling of the GDTs will be performed perrequirements of OP-120.07 and RST-211. In accordance with OP-120.07 the crane liftactivities may be performed above the GDTs and valve gallery if the OP test is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after a GDT has been sampled due to the addition ofradioactive gas and Radioactivy in the Waste GDT check showed the total activity to be

< 20,000 curies.

A. Incorrect. The first part is correct. The second part is plausible since the limit of 2.0Rem is the yearly limit for Total Effe ctive Dose Equivalent (TEDE) to thewhole body found in the occupational annual dose limites ofPD-RP-ALL-0001 (Rev. 3)B. Incorrect. The first part is plausible since the ODCM requires grab samples to beanalyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after they are taken, howeverthis is incorrect because OST-2044 requires the GDT activity to becalculated provided the GDT was sampled within the last 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Thesecond part is plausible since the limit of 2.0 Rem is the yearly limit forTotal Effective Dose Equivalent (TEDE) to the whole body found in theoccupational annual dose limites of PD-RP-ALL-0001 (Rev. 3)C. Correct.

D. Incorrect. The first part is plausible since the ODCM requires grab samples to beanalyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after they are taken, howeverthis is incorrect because OST-2044 requires the GDT activity to becalculated provided the GDT was sampled within the last 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The second part is correct Monday, September 29, 2014 7:41:35 AM

237 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000060 Accidental Gaseous Radwaste Rel. / 9060AG2.2.12; Knowledge of surveillance procedures. (Verified)(CFR 41.10 / 45.13)Importance Rating: 4.1 Technical

Reference:

OST-2044, Page 11, Rev. 29 PLP-114, Attachment 5, Page 15, Rev. 23 References to be provided: NoneLearning Objective: GWPS Objective 6.b Question Origin: NEWComments: NoneTier/Group: T1G2 SRO Justification:

10 CFR Part 55 Content -

43(b)(4): Radiation hazardsthat may arise during normal and abnormal situations,including maintenance activies and variouscontamination conditions.Monday, September 29, 2014 7:41:35 AM 238 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

84. 2014 NRC SRO 009/BANK/C/A///NONE/EARLY/061AA2.05/SATGiven the following plant conditions: - The unit is in Mode 6 with core offload in progress

- A fuel assembly has just been unlatched in the Containment Upender - The Fuel Handling SRO has reported that Refueling Cavity Level is rapidly lowering - The crew has entered and is implementing AOP-031, Loss of Refueling CavityIntegrity - The CRS has directed that the fuel assembly be returned to the Reactor VesselPrior to relatching the assembly, the following occurs: - ALL CNMT Ventilation Isolation radiation monitors have just gone into alarm - HP reports radiation levels on the Manipulator Crane of 154 mR/hr and increasingIn accordance with AOP-031, what action must be taken in Containment? Place the fuel assembly in the Reactor Vessel and evacuate ONLY unnecessarypersonnel Leave the fuel assembly in the upender and evacuate ONLY unnecessarypersonnelPlace the fuel assembly in the Reactor Vessel and evacuate ALL personnel Leave the fuel assembly in the upender and evacuate ALL personnel A.B.C.D.Monday, September 29, 2014 7:41:35 AM

239 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

In accordance with AOP-031, when radiation levels on theManipulor Crane area radiation monitor exceed 150 mR/hr all personell must beevacuated from the Containment and Security is directed to verify all personnel areclear. The personnel handling the fuel assembly would not move the assembly to a safe location. The assembly would be left in the upender or where ever it happened tobe located at the time of the high radiation condition. IF the radiation level on this arearadiation monitor were < 150 mR/hr THEN the fuel assembly would have been movedfrom the upender to a sefe storage location defined as: the Reactor vessel, a SpentFuel Pool fuel rack, or an area physically isolated from the Refu eling Cavity. Leavingthe fuel assembly in the upender is NOT an option unless radiation levels are highenough to warrent an immediate evaucation.

Additionally, a note in AOP-031 states: CNMT Ventilation Isolation radiation monitor sare set to alarm at less than or equal to 150 mR/hr for fuel movement. (theContainment Ventilation Isolation rad monitors are also area radiaiton monitors)

A. Incorrect.

The first part is plausible because AOP-031 directs this action IF the radlevels were < 150 mR. The second part is plausible becauseunnecessary personnel would be evacuated while the Fuel Handlersstored the fuel assembly but only if radiation levels were <150 mR/hr.

B. Incorrect. The first part is correct the fuel assembly should be left in the upenderdue to radiation levels being > 150 mR/hr BUT all personnel should be evacuated.C. Incorrect.

The first part is plausible because AOP-031 directs this action IF the radlevels were < 150 mR. The second part is plausible becauseunnecessary personnel would be evacuated while the Fuel Handlersstored the fuel assembly if radiation levels were <150 mR/hr but withradiation level greater 150 mR/hr ALL personnel should be evacuatedprior to storing the fuel assembly. D. Correct. Monday, September 29, 2014 7:41:35 AM 240 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes000061 ARM System Alarms / 7061AA2.05; Ability to determine and interpret the need for area evacuation; checkagainst existing limits as it applies to the Area Radiation Monitoring (ARM) SystemAlarms. (Verified)(CFR 43.5 / 45.13)

Importance Rating: 4.2Technical

Reference:

AOP-031, Attachment 1, Page 36, Rev 18References to be provided: None Learning Objective: AOP-031 Objective 4Question Origin: Bank OIT Exam Bank AOP-031 (04) 1Comments:

SRO question application tie to 10 CFR Part 55 Content -

43(b)(7) alsoTier/Group: T1/G2SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment of facility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SROlevel of knowledge by ensuring that the candidate mustknow procedural knowledge of diagnostic steps anddecision points in the abnormal operating procedures(AOP) that involve transitions to event specific attachments within the procedure.Monday, September 29, 2014 7:41:35 AM 241 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

85. 2014 NRC SRO 010/BANK/C/A//FR-Z.1, ECA-1.1/NONE/EARLY/WE14EA2.2/SATGiven the following plant conditions: - The unit was operating at 100% power

- The plant experienced a LOCA several hours ago - Only one Containment Spray Pump is running due to actions taken by the crew inaccordance with EOP-ECA-1.1, Loss Of Emergency Coolant Recirculation Subsequently, a transition has just been made to EOP-FR-Z.1, Response to HighContainment Pressure - Current Containment pressure is 14 psig and slowly lowering - RWST level is 47% and slowly lowering - 3 Containment Fan Coolers are available and runningWhich ONE of the following actions should be taken for the current plant conditions?Start the second Containment Spray Pump if Containment pressure does not lowerbelow 10 psig before exiting EOP-FR-Z.1.Start the second Containment Spray Pump per EOP-FR-Z.1 since pressure isabove 10 psig. Continue operation with one Containment Spray pump per EOP-ECA-1.1unless Containment pressure exceeds 45 psig, then start the second pump.

Continue operation with one Containment Spray pump per EOP-ECA-1.1 unlessContainment pressure begins rising, then start the second pump.

A.B.C.D.Monday, September 29, 2014 7:41:35 AM

242 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: In accordance with FR-Z.1 with Containment pressure >10psig and with ECA-1.1 is in effect then during the implementation of FR-Z.1 theContainment Spray pumps are to be operated using ECA-1.1. With current plantconditions of RWST level 47% and Containment Pressure of 14 psig the requirement per step 7 table is to have 1 Containment Spray pump running. ECA-1.1 will continueto loop back to step 1 from step 33 until RWST level is < 3%. Prior to RWST depletionto <3% if Containment pressure exceed 45 PSIG (Containment design pressure) then asecond Spray Pump would be requir ed to be in operation.

A. Incorrect. Plausible since ECA-1.1 would direct the operator to verify that theContainment Spray pumps are running. In accordance with the HNPEOP Users Guide General Information section 6.7 definition of "verify" : AtHNP, the operator is frequently directed to "verify" or "check" a condition.

In several instances the operator is also directed to "actuate" componentsor signals. The term "verify" means to observe the condition, and if it isnot satisfied, attempt to satisfy it using whatever means are available (e.g.manually from the MCB, locally by an AO, etc.). If the condition is still notestablished, the operator should go to the contingency action or continueon as appropriate. Therefore with only one Containment Spray pumprunning the Operator would start the second spray pump. But, sinceECA-1.1 is in effect the Containment Spray pumps are ran in accordancewith the ECA not the FR procedure.B. Incorrect. Plausible since ECA-1.1 would direct the operator to verify that theContainment Spray pumps are running. In accordance with the HNPEOP Users Guide General Information section 6.7 definition of "verify" : AtHNP, the operator is frequently directed to "verify" or "check" a condition.

In several instances the operator is also directed to "actuate" componentsor signals. The term "verify" means to observe the condition, and if it isnot satisfied, attempt to satisfy it using whatever means are available (e.g.manually from the MCB, locally by an AO, etc.). If the condition is still notestablished, the operator should go to the contingency action or continueon as appropriate. Therefore with only one Containment Spray pumprunning the Operator would start the second spray pump. But, sinceECA-1.1 is in effect the Containment Spray pumps are ran in accordancewith the ECA not the FR procedure.C. Correct.

D. Incorrect. Plausible since an increase in Containment pressure would indicate thatthe current Containment pressure reduction strategy is ineffective but inaccordance with ECA-1.1 starting the second pump is not required unlessContainment pressure exceeded design pressure (45 psig).Monday, September 29, 2014 7:41:35 AM 243 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal NotesW/E14 High Containment PressureWE14EA2.2; Ability to determine and interpret adherence to appropriate proceduresand operation within the limitations in the facility's license and amendments as theyapply to the (High Containment Pressure). (Verified)(CFR 43.5 / 45.13)

Importance Rating: 3.8Technical

Reference:

EOP-FR-Z.1, Step 3, Page 4, Rev. 0 EOP-ECA-1.1, Step 7 table, Page 6, Rev. 0 References to be provided: None Learning Objective: LP EOP-3.13 Objective 4.bQuestion Origin: Bank, OIT Development Bank, EOP-2.3/3.3-S.2 (1)Comments:

NoneTier/Group: T1/G2SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level ofknowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item.Monday, September 29, 2014 7:41:35 AM 244 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

86. 2014 NRC SRO 011/BANK/C/A//OWP-RP, TECH SPECS/NONE/EARLY/012G2.4.11/SATGiven the following plant conditions: - The unit is operating at 100% power

- Reactor Engineering is performing a flux map - OWP-RP-01, Reactor Protection, is in place for Loop B, Tavg/-T Channel II, dueto a failed bistable - Repair of Loop B, Tavg/-T Channel II, will not be completed for another 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months I&C discovers the bistables for Power Range channel N-41 are set higher than theallowable values of Technical Specifications and repairs will take 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months sWhich ONE of the following identifies the LEAST restrictive requirement that must becompleted for these conditions in accordance with OWP-RP-01 and TechnicalSpecifications?Bypass Power Range N-41 until repairs are completed on Loop B Tavg/-TPrepare to shutdown in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in Hot Shutdown within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sTrip the bistables for Power Range N-41 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Place the unit in Hot Standby within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months A.B.C.D.Monday, September 29, 2014 7:41:35 AM 245 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: Requires evaluation of T.S. Table 3.3-1 Action 2.a anddetermining that based on plant procedures it can not be complied with, resulting inT.S. 3.0.3 entry.Taking the instrument out of service will produce a plant trip, Action 2.acan not be performed. Therefore the unit must be place in Mode 3 (Hot Standby). TheLCO requirements are place a channel in trip in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Action 2.a) and since thechannel cannot be placed in a trip condition TS 3.0.3 applies. The unit must be placedin Mode 3 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . The total time that the operators have to get the unit from thecurrent condition to Mode 3 (Hot Standby) is therefore 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

A.Incorrect. This is plausible because Action 2.b. says the inoperable channel may bebypassed during the performance of the surveillance testing. Bypassingthe channel is only allowed during this time period for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .B. Incorrect. This is plausible because Action 2.a can not be satisfied withoutgenerating a reactor trip and with this action not satisfied entry into TS3.0.3 is required. T.S. 3.0.3 allows 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to prepare to shut the unitdown then requires the unit to be placed in Hot Standby within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sand Hot Shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Entry into TS 3.0.3 is notrequired until the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months time for action 2.a has el apsed. This is not theleast restrictive required action.C. Incorrect. This is plausible because Action 2.a requires tripping bistables but theprecautions of the OWP-RP require checking bistables when takinginstruments out of service to preclude a plant trip.

In the giveninformation the combination of instruments already failed and thediscovery of N-41 failure would cause a Reactor Trip. D. Correct. Monday, September 29, 2014 7:41:35 AM 246 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes012 Reactor Protection012G2.4.11; Knowledge of abnormal condition procedures. (Verified)(CFR: 41.10 / 43.5 / 45.13 )Importance Rating: 4.2 Technical

Reference:

OWP-RP, Tech Spec 3.0.3 pg 3/4 0.1, 3.3.1 pg 3/4 3-2, 3/4 3-6&7(pages 91, 138, 142, and 143)

References to be provided: None Learning Objective: RPS Objective 14Question Origin: Bank OIT Exam Bank, RPS (14) 4Comments:

NoneTier/Group: T2/G1SRO Justification:

10 CFR Part 55 Content - 43(b)(2) because applicationof Technical Specification below the line application of required actions considered to be SRO knowledge levelTechnical Specification items and the basis for the actions. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level ofknowledge by testing the SRO knowlegde on theapplication of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4). Monday, September 29, 2014 7:41:35 AM 247 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

87. 2014 NRC SRO 012/NEW/C/A//EP-EAL, EAL MATRIX/EAL MATRIX//039G2.1.19/Given the following plant conditions: - The plant was operating at 40%

- 0600, A 15 gpm tube leak develops in the 'A' SG - 0604, ALB 022-9-1, Start XFMR-B Lockout Trip Or Trouble, alarms - 0607, ALB-014-8-5, Computer Alarm Steam Generators, alarms - Computer Point ID ZMS1254A, 1MS-58 SG A PORV, indicates NOT SHUT - Attempts to shut 1MS-58 from the MCB are not successful - 0610, ALB-026-1-4, Annun Sys 1 Power Supply Failure, alarmsSubsequently the following occurs: - 0625, An AO reports 1MS-59, MS Line A PORV Isol Vlv, is stuck open - 0630, 'A' SG tube leakage degrades and a Reactor Trip and Safety Injection are initiatedWhich ONE of the following identifies the FIRST required classification for theconditions above?(Reference provided)

FU1.1 SU1.1 SU5.1 SU8.1 A.B.C.D.Monday, September 29, 2014 7:41:35 AM 248 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: FU1.1 Unusual Event is defined as Any loss or any potentialloss of Containment (Table F-1). The computer alarm indicates the A PORV is not shutand with the SG leakage provides a path to the environment, the A PORV is normallyisolable by shutting 1MS-59 locally, but this valve is stuck open and this meets theclassification criteria for a potential loss of containment. With the RCS leakage into ASG at 15 gpm the threshold for FU1.1 has been exceeded and the classification is met.

A. Correct. B. Incorrect. The answer is plausible since the B SUT has lost power, however this isincorrect because the A SUT is available, both sources of offsite power have not been lost to the safety buses.C. Incorrect. The answer is plausible since the annunciator for the lost of power tosystem 1 has alarmed, however this is incorrect because the loss of oneannunciator system will not result in exceeding 75% of the availableannunciation. D. Incorrect. The answer is plausible since the classification is based on unidentified orpressure boundary leakage > 10 gpm OR Identified leakage > 25 gpm.The candidate may misunderstand the difference in pressure boundaryand identified leakage and determine that the tube leak is pressureboundary leakage. However is incorrect because Identified leakage isdefined as RCS leakage through a steam generator to the SecondaryCoolant System (primary-to-secondary leakage). Monday, September 29, 2014 7:41:35 AM 249 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes039 Main and Reheat Steam039G2.1.19; Ability to use plant computers to evaluate system or component status.(Verified)(CFR: 41.10 / 45.12)Importance Rating: 3.8 Technical

Reference:

EP-EAL, page 230 and 289, Rev. 12References to be provided: EAL MatrixLearning Objective: EP-LP-2.0, Rev. 1, Objective 10 Question Origin: NewComments:

The K/A is met because the candidate must use thestatus of the computer alarm point to determine if an EAL is classification is met for the FPB matrix.Tier/Group: T2/G1SRO Justification:

10 CFR Part 55 Content -

43(b) (5) The Site EmergencyCoordinator is an SRO only job function, and thus theSEC actions, and responsibilities for determining theEAL classification during an event is SRO-only knowledge. Per NRC Clarification Guidance forSRO-only Questions, Rev 1 this question meets the SROlevel of knowledge by ensuring that the candidate musthave procedural knowledge of diagnostic steps and decision points in the EAL Matrix that involve declaringan event classification.Monday, September 29, 2014 7:41:35 AM 250 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

88. 2014 NRC SRO 013/NEW/C/A//TS 3.6.2.1, 3.6.2.3/TS 3.6.2.1, 3.6.2.3//022A2.04/Given the following plant conditions: - The unit was operating at 100% power

- Containment Spray Pump 'A' is under clearance

- 1A3-SA, cubicle 4A, Emerg MCC 1A34-SA Supply Brkr, is under clearance Subsequently a Service Water leak develops downstream of 1SW-91, CNMT FanCooler AH-2 Inlet CIV, which requires Service Water to be isolated. Which ONE of the following identifies (1) the impact of isolating Service Water to AH-2AND (2) a required action in accordance with Technical Specification 3.6.2.3,Containment Cooling System? (Reference provided) (1) Service Water is isolated to ONLY AH-2 1A-SA (2) restore the inoperable spray system to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be inat least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (1) Service Water is isolated to ONLY AH-2 1A-SA (2) restore the inoperable train of fan coolers to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months orbe in at least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (1) Service Water is isolated to BOTH AH-2 1A-SA and 1B-SA (2) restore the inoperable spray system to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be inat least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (1) Service Water is isolated to BOTH AH-2 1A-SA and 1B-SA (2) restore the inoperable train of fan coolers to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months orbe in at least Hot Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A.B.C.D.Monday, September 29, 2014 7:41:35 AM

251 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: Four containment fan coolers (AH-1, AH-2, AH-3, and AH-4)shall be OPERABLE with one of two fans in each cooler capable of operation at lowspeed. Train SA consists of AH-2 and AH-3. Train SB consists of AH-1 and AH-4. 1A34-SA is the power supply to AH-3 which is inoperable without power. With AContainment Spray Pump under clearance it is also inoperable. Isolating service waterflow to AH-2 will make both AH-2 fans inoperable and with the current inoperableequipment require entry into LCO 3.6.2.3 action c. One inioperable train ofContainment Spray must be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HSB within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months A. Incorrect. The first part is plausible since each Air handler has two fans in each unitand the fan coolers have individual isolations, however this is incorrectbecause the service water flow is common to each fan on a cooler unitand will isolate cooling to both fans making the Fan cooler inoperable.

The second part is correct.B. Incorrect. The first part is plausible since each Air handler has two fans in each unitand the fan coolers have individual isolations, however this is incorrectbecause the service water flow is common to each fan on a cooler unitand will isolate cooling to both fans making the Fan cooler inoperable. The second part is plausible since this is the correct action if both trains offan coolers are inoperable with both containment spray pumpsinoperable, however this is not correct since the loss of the electrical MCC only results in the loss of AH-3 and not the loss of both trains of Fan coolers.C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since this is thecorrect action if both trains of fan coolers are inoperable with bothcontainment spray pumps inoperable, however this is not correct sincethe loss of the electrical MCC only re sults in the loss of AH-3 and not theloss of both trains of Fan coolers. Monday, September 29, 2014 7:41:35 AM 252 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes 022 Containment Cooling System (CCS) 022 A2.04; Ability to (a) predict the impacts of the following malfunctions or operationson the CCS; and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations: Loss of service water(CFR: 41.5 / 43.5 / 45.3 / 45.13

)Importance Rating: 3.2Technical

Reference:

Technical Specification 3.6.2.1 Technical Specification 3.6.2.3References to be provided: Technical Specification 3.6.2.1 Technical Specification 3.6.2.3Learning Objective: CCS Student Text, Objective 11.b Question Origin: NewComments: Unable to write a question to SRO level for original K/A063 A2.02 due to the lack procedural actions or an SROtask for the original K/A Replace with randomly selected K/A 022 A2.04 by David Lanyi 5/09/2014

Tier/Group: T2/G1SRO Justification:

10 CFR Part 55 Content - 43(b)(2) because applicationof Technical Specification below the line application of required actions considered to be SRO knowledge levelTechnical Specification items and the basis for theactions. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by testing the SRO knowlegde on the application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru4.0.4). Monday, September 29, 2014 7:41:35 AM 253 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

89. 2014 NRC SRO 014/MODIFIED/C/A//AOP-022, TS 3.7.4/NONE//076A2.01/Given the following plant conditions: - The unit is operating at 85% power

- The following annunciator is received in the Control Room:

- ALB-002-7-2, Serv Wtr Pumps Discharge Low Press - Cooling Tower Basin Level is lowering rapidly - Service Water header pressure is 50 psig and loweringOne minute later the following conditions exist: - Service Water header pressure is 35 psig and rising - The Cooling Tower Basin Level is stable- The RAB AO reports that a large volume of water is gushing from the upstreamflange of 1SW-274, Header B Return To Normal Header valve Which ONE of the following completes the statements below?

The leak is located in the (1) system. In accordance with Technical Specification 3.7.4, Emergency Service Water, the basesfor the Limiting Condition of Operation is to ensure that sufficient cooling capacity isavailable for continued operation of safety related equipment during (2) conditions. (1) Normal Service Water (2) normal AND accident (1) Normal Service Water (2) ONLY accident (1) Emergency Service Water (2) normal AND accident (1) Emergency Service Water (2) ONLY accident A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct:

Based on SW pressure indications, ESW should haveautomatically started and isolated ESW from NSW. Cooling Tower level stabilizesindicates the leak is isolated from the NSW system therefore indicates a leak on ESW.

Additionally the report is the leak is on the ustream side of the ESW Train B return header and theref ore is isolated from the NSW system. The bases for technicalspecification 3.7.4 states the OPERABILITY of the Emergency Service Water SystemMonday, September 29, 2014 7:41:35 AM

254 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal ensures that sufficient cooling capacity is available for continued operation ofsafety-related equipment during normal and accident conditions. The redundant coolingcapacity of this system, assuming a single failure is consistent with the assumptionsused in the safety analyses.

A. Incorrect. The first part is plausible if candidate misunderstands that ESW wouldautomatically initiated and isolated from NSW in response to the loweringNSW pressure. The second part is correct.B. Incorrect. The first part is plausible if candidate misunderstands that ESW wouldautomatically initiated and isolated from NSW in response to the loweringNSW pressure. The second part is plausible because the ESW systemdoes provide cooling capacity for accident conditions to safety relatedequipment, however the normal system line up of the ESW header whichis supplied by the NSW system during normal operation and componentson the ESW header such as the CSIPs and CCW pumps are required tobe cooled during these normal operating conditions. C. Correct.

D. Incorrect. The first part is correct The second part is plausible because the ESWsystem does provide cool ing capacity for accident conditions to safetyrelated equipment, however the normal system line up of the ESW headerwhich is supplied by the NSW system during normal operation andcomponents on the ESW header such as the CSIPs and CCW pumps arerequired to be cooled during these normal operating conditions. Monday, September 29, 2014 7:41:35 AM

255 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes076 Service Water System (SWS)076 A2.01 Ability to (a) predict the impacts of the following malfunctions or operationson the SWS; and (b) based on those predictions, use procedures to correct, control,or mitigate the consequences of those malfunctions or operations: Loss of SWS(CFR: 41.5 / 43.5 / 45/3 / 45/13) (Verified)Importance Rating: 3.7 Technical

Reference:

AOP-022, Section 3.2, Rev 35, pg 30 ALB-002-7-2, Rev. 48 , pg 33 Technical Specification 3.7.4 Bases, Amendment 102,pg B 3/4 7-3References to be provided: NoneLearning Objective: AOP-022 Objective 4Question Origin: Modified Previous 2013 NRC SRO radomly selected Comments: NoneTier/Group: T2G1SRO Justification:

10 CFR Part 55 Content - 43(b)(2) because applicationof Technical Specification below the line application of required actions considered to be SRO knowledge levelTechnical Specification items and the basis for theactions. Per NRC Clarification Guidance for SRO-onlyQuestions, Rev 1 this question meets the SRO level of knowledge by testing the SRO knowlegde on the application of Limiting Condition for Operation (LCO) requirements (LCO 3.7.4 and the basis for the LCO). Monday, September 29, 2014 7:41:35 AM 256 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

90. 2014 NRC SRO 015/NEW/FUNDAMENTAL//TS 3.9.2, ALB-13-2-1/NONE//103A2.04/The following conditions are occurring with the unit in Mode 6 and refueling activities inprogress:

- Source range audible counts are unexpectedly rising - Annunciator ALB-013-2-1, Source Range High Flux Lvl at Shutdown, has gone intoalarm due to both NI-31 and NI-32 responsesWhich ONE of the following completes the statements below?

The Containment local evacuation alarm (1) actuated. Subsequently: - It was determined that both Source Range detectors were not responding correctlyTechnical Specifications requires the RCS boron concentration to be determined within (2) hours. (1) should be manually (2) 4 (1) should be manually (2) 12 (1) will be automatically (2) 4 (1) will be automatically (2) 12 A.B.C.D.Monday, September 29, 2014 7:41:35 AM 257 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: The High Flux At Shutdown annuciator ALB-13-2-1 alarmwill alarm when Source Range levels are > 1.9 times background with the plant in Mode5 for refueling. This level is required to provide an alarm for a significant positivereactivity addition during a dilution heatup, cooldown, core alterations...ect. Tech SpecAction 3.9.2 action b would apply for this situation. With both of the Source Rangedetectors inoperable or not operating core alterations must be immediately suspendedand take actions to restore one Source Range to operable status. RCS boronconcentration must be determined within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

A. Incorrect. The first part is plausible since actions of both AOP-013 and AOP-31 is toverify all unnecessary personnel are evacuated from Containment bysounding the local evacuation alarm and making a PA announcement. But the Containment evacuation alarm will automatically sound whenHigh Flux at Shutdown occurs. The second part is correct. B. Incorrect. The first part is plausible since actions of both AOP-013 and AOP-31 is toverify all unnecessary personnel are evacuated from Containment bysounding the local evacuation alarm and making a PA announcement. But the Containment evacuation alarm will automatically sound whenHigh Flux at Shutdown occurs. The second part is plausible since theTech Spec action is to obtain the RCS boron concentration every 12hours after the initial concentration is determined within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of theentry into the action statement. C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since the Tech Specaction is to obtain the RCS boron concentration every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after theinitial concentration is determined within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the entry into theaction statement. Monday, September 29, 2014 7:41:35 AM 258 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes103 Containment103A2.04; Ability to (a) predict the impacts of Containment evacuation (includingrecognition of the alarm) on the containment system and (b) based on thosepredictions, use procedures to correct, control, or mitigate the consequences. (Verified)(CFR: 41.5 / 43.5 / 45.3 / 45.13)Importance Rating: 3.6Technical

Reference:

ALB-013-2-1, Rev. 34, Tech Spec section 3.9.2References to be provided: NoneLearning Objective: AOP-LP-3.13, Objective 5Question Origin: New Comments: NRC Chief Examiner David Lanyi on 9/18/2014determine it was satisfactory to address the (b) part ofthis K/A by having the candidate apply the actionrequirements of T.S. 3.9.2 Tier/Group: T2/G1SRO Justification:

10 CFR Part 55 Content - 43(b)(2) because applicationof Technical Specification below the line application of required actions considered to be SRO knowledge levelTechnical Specification items and the basis for the actions. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level ofknowledge by testing the SRO knowlegde on theapplication of Limiting Condition for Operation (LCO) requirements (LCO 3.9.2 and the basis for the LCO). Monday, September 29, 2014 7:41:35 AM 259 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

91. 2014 NRC SRO 016/NEW/C/A//OWP-RP, TECH SPECS/NONE//015G2.1.23/Given the following plant conditions: - A Reactor startup is in progress at 10

-8 amps in accordance with GP-004, ReactorStartup (Mode 3 To Mode 2) - Power Range N-42 instrument power fuse was found in the blown condition Which ONE of the following identifies (1) the procedure required to be implemented toremove Power Range N-42 from service AND (2) the mode change requirement for thecurrent conditions? (1) OWP-RP, Reactor Protection (2) Entry in to Mode 1 is allowed. (1) OWP-RP, Reactor Protection (2) Must be returned to Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (1) OP-105, Excore Nuclear Instrumentation (2) Entry in to Mode 1 is allowed. (1) OP-105, Excore Nuclear Instrumentation (2) Must be returned to Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A.B.C.D.Monday, September 29, 2014 7:41:35 AM

260 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: OWP-RP-24 for Power Range N-42 provides the LCOaction required due to inoperable Channel 2 Power Range Nuclear Instrumentation.

Technical Specfication 3.3.1 for the Power Range NI's requires three detectors to beoperable and if not the apply action two of the LCO. Action 2 states with the number ofOPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION may proceed providedthe following conditions are satisfied:a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Provided N-42 is placed in the tipped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months continued operation isallowed and therefore entry into Mode 1 is allowed per Technical Specification 3.0.4 aswell. A. Correct. B. Incorrect. The first part is correct. he second part is plausible since the unit is inmode 2 at 10-8 amps and the action statement allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to place thechannel in trip the candidate may misunderstand the application of theaction statement and assume placing the unit in a condition when theaction statement doesn't apply (Mode 3) is the correct action to perform,however is incorrect because Technical Specifications allow continuedoperation of the unit with one Power Range NI removed from service. C. Incorrect. The first part is plausible since OP-105 is used to adjust the gain on thegain of the Power Range NI's during normal operations, however this isincorrect because performing a gain adjustment does not require the NI tobe removed from operation. Th e second part is correct. D. Incorrect. The first part is plausible since OP-105 is used to adjust the gain on thegain of the Power Range NI's during normal operations, however this isincorrect because performing a gain adjustment does not require the NI tobe removed from operation. The second part is plausible since the unit isin mode 2 at 10-8 amps and the action statement allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to placethe channel in trip the candidate may misunderstand the application of theaction statement and assume placing the unit in a condition when theaction statement doesn't apply (Mode 3) is the correct action to perform,however is incorrect because Technical Specifications allow continuedoperation of the unit with one Power Range NI removed from service.Monday, September 29, 2014 7:41:35 AM

261 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes015 Nuclear Instrumentation015G2.1.23; Ability to perform specific system and integrated plant procedures duringall modes of plant operation. (Verified)(CFR: 41.10 / 43.5 / 45.2 / 45.6)Importance Rating: 4.4 Technical

Reference:

Technical Specification 3.0.4 Technical Specification 3.3.1 OWP-RPReferences to be provided: None Learning Objective: TS-LP-2.0, Objective 3.a NIS Student Text, Objective 12Question Origin: New Comments:

NoneTier/Group: T2/G2SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-onlyQuestions, Rev 1 this question meets the SRO level ofknowledge by ensuring that the additional knowledge ofthe procedure's selection is required to correctly answerthe written test item.

Monday, September 29, 2014 7:41:35 AM

262 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

92. 2014 NRC SRO 017/NEW/C/A//AOP-041, AOP-041-BD/NONE//033G2.4.18/Given the following plant conditions: - Spent Fuel Pool temperature is rising

- Spent Fuel Pool level is loweringWhich ONE of the following in accordance with AOP-041, Spent Fuel Pool Events,identifies (1) the FIRST action required to be taken to address the above conditionsAND (2) the basis for the action? (1) Restore Spent Fuel Pool level using Attachment 4, SFP Makeup Water SourceAvailability. (2) To prevent boiling of the Spent Fuel Pools that could result in boron plating onthe fuel assemblies. (1) Restore Spent Fuel Pool level using Attachment 4, SFP Makeup Water SourceAvailability. (2) Adequate cooling to the Spent Fuel Pools will likely be provided through thefeed and bleed process. (1) Reduce Spent Fuel Pool temperature using Attachment 5, Aligning SFPCooling with Cross-Connected Pump and HX. (2) To prevent boiling of the Spent Fuel Pools that could result in boron plating onthe fuel assemblies. (1) Reduce Spent Fuel Pool temperature using Attachment 5, Aligning SFPCooling with Cross-Connected Pump and HX. (2) Adequate cooling to the Spent Fuel Pools will likely be provided through thefeed and bleed process.

A.B.C.D.Monday, September 29, 2014 7:41:35 AM 263 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct: SFP Cooling is a longer-term concern, as long as fuel poolinventory remains normal. In the event of an ongoing loss of inventory, providing makeup water will likely provide adequate cooling through the feed and bleed process.If necessary, deliberate feed and bleed can be initiated to provide long term SFPcooling. With SFP level lowering the temperature section directs the candidate toreturn to section 3.1 and address the lowering level first. The Section 3.1 step 10 RNOactions refer the candidate to Attachment 4 to determine the methods availalbe torestore SFP level.

A. Incorrect. The first part is correct. The second part is plausible since the lowerin glevel will reduce the total height of the water in the pool which reduces themargin to boiling. As the water boils off from the pool which causes ahigher boron concentration in the SFP. A higher boron concentration willresult a higher probability of the boron coming out of solution andadhering to the hotter localized areas pl ating out on the fuel assembly. B. Correct.

C. Incorrect. The first part is plausible since AOP-041 addresses both lowering leveland rising temperature. The SFP Cooling system is used to provide bothmakeup inventory and cooling capacity therefore the candidate may havea misconception on the prority required to address each condition. Thesecond part is plausible since the lowering level will cause a higher boronconcentration in the SFP. A higher boron concentration will result ahigher probability of the boron coming out of solution and adhering to thehotter localized areas plating out on the fuel assembly.D. Incorrect. The first part is plausible since AOP-041 addresses both lowering leveland rising temperature. The SFP Cooling system is used to provide bothmakeup inventory and cooling capacity therefore the candidate may havea misconception on the prority required to address each condition.Thesecond part is correct. Monday, September 29, 2014 7:41:35 AM 264 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes033 Spent Fuel Pool Cooling033G2.4.18; Knowledge of the specific bases for EOPs. (Verified)(CFR: 41.10 / 43.1 / 45.13)Importance Rating: 4.0 Technical

Reference:

AOP-041, Section 3.1 step 10 RNO, page 7, 8 Rev. 2 AOP-041-BD, Section 3.2 step 1 Note, page 12, Rev 1 References to be provided: NoneLearning Objective: AOP-LP-3.41, Objective 4 Question Origin: Modified Bank OIT Exam Bank, AOP-041 (03) 1, tocreate a new question with different second part and stem conditions.Comments:

8/29/2014 Dave Lanyi (NRC Chief Examiner) says it isOK to write question to AOP-041 basis since the EOP's don't have any relationship to the Spent Fuel PoolCooling system.

Tier/Group: T2/G2 SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item. Monday, September 29, 2014 7:41:36 AM 265 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

93. 2014 NRC SRO 018/NEW/C/A//OP-120.07, ODCM/ODCM 3.3.3.11//071A2.06/A release of WGDT 'E' is in progress.Subsequently: - REM-1WV-3546, WPB Stack 5 PIG Monitor has gone into HIGH ALARM - 3WG-229, WG Decay Tanks E&F To Plant Vent Valve, failed to shut Which ONE of the following completes the statement below?

In accordance with AOP-005, Radiation Monitoring System (1) is required to beimplemented to stop the release in progress.In accordance with ODCM 3.3.3.11, Radioactive Gaseous Effluent MonitoringInstrumentation, (2) is (are) required to resume the release.

(Reference provided) (1) OP-100, Reactor Coolant System (2) the initiatation of the preplanned alternate method of monitoring the appropriateparameter(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (1) OP-100, Reactor Coolant System (2) samples, release rate calcs, and an Independent Verification of the valveline-up (1) OP-120.07, Waste Gas Processing (2) the initiatation of the preplanned alternate method of monitoring the appropriateparameter(s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (1) OP-120.07, Waste Gas Processing (2) samples, release rate calcs, and an Independent Verification of the valveline-up A.B.C.D.Monday, September 29, 2014 7:41:36 AM 266 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisReason answer is correct:

In accordance with AOP-005, Radiation Monitoring System if a Waste Gas Decay Tank the release is in progress and accidentia lrelease of waste gas is occurring in the Waste Process Building (WPB) then tostopping the release would be accomplished using OP-120.07, Waste Gas Processing. WASTE PROCESSING BUILDING VENT STACK 5 functional unit a.1 Noble GasActivity Monitor (PIG) requires one monitor at all times or performance of actions 45and 51. With the malfunction of 3WG-229 to automatically shut only the Stack 5 PIG isinoperable therefore only action 45 applies. Prior to initiating the release at least 2independent samples of the tank's contents are analyzed and at least 2 tech qualifiedmembers of the facility staff independently verify the release rate calc and dischargevalve lineup.

A. Incorrect. The first part is plausible since in AOP-009 Attachment 1 step 2 wouldapply if the PRT was being vented (which is vented to the WGDT) then itwould require the use of OP-100, Reactor Coolant System to stop theventing process which would stop th e release. The second part isplausible since this is partially corre ct if both the Stack 5 PIG and WRGMare inoperable, however this is incorrect because 3WG-229 onlyautomatically shuts due to a High alarm from the PIG. Both the Stack 5PIG and WRGM high alarms functioned therefore the automatic functionof the PIG did not operate correctly which make only the PIG inoperable. B. Incorrect. The first part is plausible since in AOP-009 Attachment 1 step 2 wouldapply if the PRT was being vented (which is vented to the WGDT) then itwould require the use of OP-100, Reactor Coolant System to stop theventing process which would stop th e release. The second part iscorrect.C. Incorrect. The first part is correct. The second part is plausible since this is partiallycorrect if both the Stack 5 PIG and WRGM are inoperable, however this isincorrect because 3WG-229 only automatically shuts due to a High alarmfrom the PIG. Both the Stack 5 PIG and WRGM high alarms functionedtherefore the automatic function of the PIG did not operate correctly whichmake only the PIG inoperable. D. Correct. Monday, September 29, 2014 7:41:36 AM 267 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes071 Waste Gas Disposal071A2.06: Ability to (a) predict the impacts of a supply failure to the isolation valve onthe Waste Gas Disposal System; and (b) based on those predictions, use proceduresto correct, control, or mitigate the consequences. (Verified)(CFR: 41.5 / 43.5 / 45.3 / 45.13)

Importance Rating: 2.5Technical

Reference:

AOP-005, Attachment 3, page 15, Rev. 30, AOP-009, Attachment 1, Page 6, Rev. 18 ODCM 3/4.3.3.11 Radioactive Gaseous Effluent Monitoring InstrumentationReferences to be provided: NoneLearning Objective: Lesson Plan AOP-LP-3.5 (AOP-005) Objective 4 GWPS Student Text, Objective 9 Question Origin: NEWComments:

NoneTier/Group: T2/G2 SRO Justification:

10 CFR Part 55 Content -

43(b)(5): Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by ensuring that the additional knowledge ofthe procedure's content is required to correctly answer the written test item.Monday, September 29, 2014 7:41:36 AM 268 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

94. 2014 NRC SRO 019/BANK/C/A//TS 3.4.8/NONE/EARLY/G2.1.34/SATGiven the following plant information: - The unit was operating at 79% power when a Turbine runback occurred

- The crew has stabilized the plant at 58% power - Fifteen minutes after stabilization, annunciator ALB 026-2-1, Gross Failed Fuel DetTrouble, alarmed - An RCS activity sample was taken with the following results:- Gross (beta-gamma) specific activity has been calculated to be 165/

E- Dose-equivalent I-131 specific activity (DEI) is 130 µCi/gmWhich ONE of the following conditions would (1) satisfy Technical Specificationsrequirements, AND (2) what is the basis for the requirement? (1) Be in Mode 3 with T avg <500°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (2) To minimize the potential release of activity in the event of a SG Tube Rupture. (1) Be in Mode 3 with T avg <500°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (2) To reduce the production of fission products until DEI is restored to the TechSpec required value of

< 0.35 µCi/gm. (1) Reduce power below 50% within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . (2) To minimize the potential release of activity in the event of a SG Tube Rupture. (1) Reduce power below 50% within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . (2) To reduce the production of fission products until DEI is restored to the TechSpec required value of

< 0.35 µCi/gm.A.B.C.D.Monday, September 29, 2014 7:41:36 AM

269 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer Analysis Reason answer is correct: Requires evaluation of T.S. 3.4.8 Action b and determiningthe bases for that action. Tech Spec 3.4.8 limit for specific activity of the RCS in Modes1, 2, 3, 4, and 5 is: a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT 1-131. and b. Less than or equal to 100/E microCuries per gram of gross radioactivity.The Chemistry results provided for the RCS activity samples have exceeded the TechSpec limits. With the limit exceeded the action required is be in at least HOT STANDBY(Mode 3) with Tavg <500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Tech Spec basis for reducing Tavg to < 500°F is that with the RCS at or below500°F the RCS saturation pressure during a SGTR event will be below the lift pressureof the atmospheric steam relief valves.

A. Correct. B. Incorrect. The first part is correct. The second part is plausible since Tech Specsrequires restoring DEI to < 0.35

µCi/gm but the time given is within 48hours. C. Incorrect. The first part is plausible since Secondary chemistry limits do not applywhen the unit is operating below 50% power. AOP-016 for SG tubeleakage requires the unit to be reduced to < 50% within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . The second part is correct.D. Incorrect. The first part is plausible since Secondary chemistry limits do not applywhen the unit is operating below 50% power. AOP-016 for SG tubeleakage requires the unit to be reduced to < 50% within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . Thesecond part is plausible since Tech Specs requires restoring DEI to <

0.35 µCi/gm but the time given is within 48 hourMonday, September 29, 2014 7:41:36 AM

270 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.1 Conduct of OperationsG2.1.34 Knowledge of primary and secondary plant chemistry limits. (Verified)(CFR: 41.10 / 43.5 / 45.12)Importance Rating: 3.5 Technical

Reference:

Tech Spec 3.4.8 pg 3/4 4.29, 3.4.8 Bases pg 3/4 4-6 (pages 229 and 60)References to be provided: Tech Spec 3.4.8Learning Objective: RCS Objective 16 Question Origin: Bank OIT Exam Bank RCS (16) 3Comments:

NoneTier/Group: T3 SRO Justification:

10 CFR Part 55 Content -

43(b)(2): Facility operatinglimitations in the technical specifications and their bases.Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by testing the SRO knowlegde of TS bases that are required to analyze TS required actions and terminology.Monday, September 29, 2014 7:41:36 AM 271 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

95. 2014 NRC SRO 020/NEW/C/A//TS 3.3.1/NONE//G2.1.45/Given the following plant conditions: - The unit is operating at 100% power

- Annunciator ALB-009-4-3, Pressurizer Low Level Ltdn Secured and Htrs Off, hasalarmed due to a transmitter failureWhich ONE of the following completes the statements below regarding the systemfailure and Technical Specification requirements? 1CS-231, FK-122.1, Charging Flow demand has increased due to the failure of PRZ level transmitter (1)

. AND The Technical Specifications required action is to place the channel in the trippedcondition within (2) hours. (1) 459 (2) 4 (1) 460 (2) 4 (1) 459 (2) 6 (1) 460 (2) 6 A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct:

With PRZ level control in the normal configuration PRZLT-459 and LT-460 are selected as controlling channels. Although either channelfailing low will cause annunciator ALB-009-4-3 to alarm, secure PRZ heaters andsecure letdown ONLY LT-459 will cause 1CS-231, FK-122.1 to increase demand. LT-459 failing low will cause a level error signal between perceived actual level and programmed level. The level will appear low when actually has not changed. The leve lerror signal will continue to cause excessive Charging flow until operator action placesthe controller in manual and lowers the output demand. A failure of LT-460 low doesnot affect 1CS-231, FK-122.1 since the signal from LT-460 does not input to Chargingflow controls. Identifying that LT-459 has failed low by using diverse indications of what1CS-231 response is validates that LT-459 is the transmitter that failed. Tech Spec3.3.1 requires a minimum Reactor Trip system instrumentation channels to beMonday, September 29, 2014 7:41:36 AM

272 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittalchannels and 2 required to trip and also be operable.

Since LT-459 has failed Action 6applies. The inoperable channel is required to be placed in the tripped conditionswithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A. Incorrect. The first part is correct. The second part is plausible since Tech Spe cAction 6 for the inoperable channel allows the inoperable channel to bebypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels butthis is not the time required to place the channel in the tripped condition.B. Incorrect. The first part is plausible since both LT-459 and LT-460 will cause theannunciator window to alarm. It requires detailed system knowledge todetermine that LT-459 failure is the only possiblity for the increase inCharging flow. If LT-460 had failed low Charging flow would eventuallydecrease since actual level would continue to rise due to letdownsecuring. The actual level sensed by LT-459 and compared to programlevel would cause 1CS-231 to reduce Charging flow to the PRZ based onlevel becoming higher than program level. The second part is plausiblesince Tech Spec Action 6 for the inoperable channel allows theinoperable channel to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillancetesting of other channels but this is not the time required to place the channel in the tripped condition.C. Correct.

D. Incorrect. The first part is plausible since both LT-459 and LT-460 will cause theannunciator window to alarm. It requires detailed system knowledge todetermine that LT-459 failure is the only possiblity for the increase inCharging flow. If LT-460 had failed low Charging flow would eventuallydecrease since actual level would continue to rise due to letdownsecuring. The actual level sensed by LT-459 and compared to programlevel would cause 1CS-231 to reduce Charging flow to the PRZ based onlevel becoming higher than program level. The second part is correct. Monday, September 29, 2014 7:41:36 AM

273 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.1 Conduct of OperationsG2.1.45 Ability to identify and interpret diverse indications to validate the response ofanother indication (Verified)(CFR: 41.7 / 43.5 / 45.4)

Importance Rating: 4.3Technical

Reference:

ALB-009-4-3, Tech Spec 3.3.1, Student Text PZR LevelControlReferences to be provided: None Learning Objective: Lesson Plan PZRLC Objectives 9.c and (SRO ONLY)Objective 11.aQuestion Origin: NEW Comments:

SRO question application tie to 10 CFR Part 55 Content -

43(b)(2) alsoTier/Group: T3SRO Justification:

10 CFR Part 55 Content -

43(b)(5) - Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. The applicant's knowledge can be evaluatedat the level of 10 CFR 55.43(b)(5) by ensuring that theadditional knowledge of the procedure's content is required to correctly answer the written test item. In thisinstance the applicant must know the informationcontained in an annunciator procedure (RO knowledge). To make this question SRO specific a Tech Spec actionwas added to the second part of the question. Monday, September 29, 2014 7:41:36 AM

274 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

96. 2014 NRC SRO 021/BANK/FUNDAMENTAL///NONE/EARLY/G2.2.17/SATGiven the following plant conditions: - The unit is operating at 100% power

- You are working a weekend backshift (Off-Normal Hours) - An instrumentation failure has resulted in the crew entering AOP-001, Malfunctionof Rod Control and Indication System and placing the rod control system to manualIn accordance with WCM-001, On-Line Maintenance Risk Management, which ONE ofthe following personel are RESPONSIBLE for assessing the impact on scheduled workand plant risk? STAWCC-SROWork Week ManagerShift Manager (or designee)

A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct: This question deals with off normal hour shift problems. Inaccordance with WCM-001 the responsibilites of the Shift Manager: When anemergent failure or other non-discretionay condition arises, the impact on scheduledwork and risk should be assessed. The Shift Manager (or designee), is responsible forassessing this impact during off-normal working hours.

A. Incorrect. Plausible since the STA's responsiblity is to determine the blended riskassociated with equipment configuration through the use of EOOS and asdirected by WCM-001 or the Shift Manager. The STA would not berequired to perform WCM-001 Attachment 2. B. Incorrect. Plausible since the WCC-SRO is required to perform a real time riskassessment at the time of work release in accordance withADM-NGGC-0104, Work Management Process but would not beresponsible for assessing the impact of this failure per WCM-001,Attachment 2. C. Incorrect. Plausible since the Work Week Manager would perform a riskassessment of the schedule and any emergent work that arises that mustbe worked in that week. But, they are not reponsible to asses the impactof weekend backshift equipment failures. That is the responsiblilty of theSM. D. Correct. Monday, September 29, 2014 7:41:36 AM 275 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.2 Equipment controlG2.2.17 Knowledge of the process for managing maintenance activities during poweroperations, such as risk assessments, work prioritization, and coordination with thetransmission system operator. (Verified)(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: 3.8Technical

Reference:

WCM-001, SM responsibilites, Page 6 and Attachment 2, Page 28, Rev. 21References to be provided: None Learning Objective: PP-3.6 Objective 5.bQuestion Origin: Bank, OIT Exam Bank, PP-3.6 (05B) 1Comments:

NoneTier/Group: T3SRO Justification:

10 CFR Part 55 Content -

43(b)(3): Facility licenseeprocedures required to obtain authority for design and operating changes in the facility. Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this questionmeets the SRO level of knowledge by use of Siteprocedure that require the Shift Manager specifically toperform the evaluation of plant risk in the event of emergent work release.

Monday, September 29, 2014 7:41:36 AM

276 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

97. 2014 NRC SRO 022/BANK/FUNDAMENTAL///NONE//G2.2.25/Which ONE of the following satisfies the Technical Specification bases requirement foroff-site power distribution with the plant in Mode 1? The requirement can be satisfied by any two separate off-site transmission linesthat can power the SUTs.The requirement can ONLY be satisfied by two off-site transmission lines that feedthe SUTs directly (Cary, Regency Park and Cape Fear North).The requirement can ONLY be satisfied by the off-site transmission lines that do notfeed the respective North or South switchyard bus through a jumper.The requirement is satisfied as long as the switchyard alignment is such that poweris available from the off-site transmission network to both SUTs regardless of thenumber of transmission lines available.

A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct:

The minimum offsite power sources will be maintained sothat at least 2 independent offsite circuits are available. Circuits may consist of any twoof the incoming transmission lines to the SATs (either through the switchyard ordirectly).

A. Correct. B. Incorrect.

Plausible since this would meet the operability requirement, but is notthe only permitted alignment. C. Incorrect.

Plausible since BKR 52-16 can only be used to maintain operability ofTrain B power if the disconnects for 52-17 are closed with the jumpersinstalled, but this is an acceptable source.D. Incorrect.

Plausible since a single transmission line can provide power to bothSUTs, but 2 transmission lines are required for redundancy. Monday, September 29, 2014 7:41:36 AM 277 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.2 Equipment controlG2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions foroperations and safety limits. (Verified)(CFR: 41.5 / 41.7 / 43.2)Importance Rating: 4.2 Technical

Reference:

TS Basis section 3.8, OST-1023References to be provided: NoneLearning Objective: ADEL-2.7-2 Question Origin: Bank, OIT Development Bank, ADEL-2.2-2 (5) Comments: 2011 NRC SRO Q#4Tier/Group: T3 SRO Justification:

10 CFR Part 55 Content -

43(b)(2): Facility operatinglimitations in the technical specifications and their bases.Per NRC Clarification Guidance for SRO-only Questions, Rev 1 this question meets the SRO level of knowledge by testing the SRO knowlegde of TS bases that are required to analyze TS required actions and terminology.Monday, September 29, 2014 7:41:36 AM 278 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

98. 2014 NRC SRO 023/PREVIOUS/C/A///AP-617 ATT. 1//G2.3.14/Given the following plant conditions: - An employee was injured and contaminated

- The employee was transported to Western WakeMed for treatment before he wasde-contaminated - Duke Energy is planning a news release for this eventWhich ONE of the following completes the statements below?

In accordance with AP-617, Reportability Determination And Notification, theEARLIEST required NRC notification of this event is within (1) hours. In accordance with AOP-013, Fuel Handling Accident, (2) is the primaryradiological concern for fuel off-loaded more than 6 months ago because it will NOT bedetected by personal dosimetry or area radiation monitors.(Reference provided) (1) 4 (2) Krypton-85 (1) 4 (2) Iodine-131 (1) 8 (2) Krypton-85 (1) 8 (2) Iodine-131 A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct: The Fuel in the Spent Fuel Pool has been there for morethan 6 months. The personal non-detectable radiation hazard would be Krypton-85which is a beta emitter. AOP-013 has a note stating: Kr-85 is the primary radiological concern for fuel off-loaded more than 6 months ago. Kr-85 is a beta hazard and willNOT be detected by personal dosimetry or area radiation monitors. There is also acaution stating: Airborne radiation may be present and gas bubbles may be visible if afuel assembly is ruptured. Personnel should remain clear until Health Physics hasestablished access controlsThe basis document states the activity of most concern is that which is contained in thevolatile fission product gases contained in the fuel pellet to cladding gap. When a fuelMonday, September 29, 2014 7:41:36 AM

279 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittalpin is damaged, this fission product inventory can be released to the SFP water.Technical Specifications 3.9.10 and 3.9.11 require a minimum water level of 23 feet inthe SFP and Refueling Cavity specifically to reduce the potential dose resulting from afuel handling accident. This amount of water will capture 99% of the assumed 10%iodine activity present in the pellet to clad gap before it breaks the surface of the water.However, although the water is expected to retain a large fraction of this activity, aportion of it will reach the surface and bubble out into the FHB or CNMT atmosphere.

(Since halogens are soluble, a large fraction of these halogens will be retained by thewater, whereas noble gases, being insoluble, will not be retained.) Once in theatmosphere, much of this fission product activity will cause an observed increase inarea radiation levels. (Gases such as Kr-85 which are primarily beta hazards will notbe detectable using installed monitors.)The transportation of a potentially contaminated individual must be reported to the NRC within8 hours per AP-617. However, since a press release regarding the event is planned by Duke Energy the incident must be reported to the NRC within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per Attachment 1 and note 12of AP-617 (Offsite Notification has been or will be made).

A. Correct. B. Incorrect. The first part is correct. The second part is plausible since volatile fissionproduct gases have escaped from the damaged fuel assembly and wouldbe observable as bubbles coming to the surface of the SFP. Iodine 131would be part of the volatile gases. I-131 is gamma emiter which wouldbe detectable with personal dosimetry and therefore would NOT be anon-detectable radiation concern.C. Incorrect. The first part is plausible since under other other circumstances an 8-hourNRC report would be required for transport of a potentially contaminatedindivdual. A 4-hour report applies only because a related offsitenotification to another agency will be made. The second part is correct.D. Incorrect. The first part is plausible since under other other circumstances an 8-hourNRC report would be required for transport of a potentially contaminatedindivdual. A 4-hour report applies only because a related offsitenotification to another agency will be made. The second part is plausible since volatile fission pr oduct gases have escaped from the damaged fuelassembly and would be observable as bubbles coming to the surface ofthe SFP. Iodine 131 would be part of the volatile gases. I-131 is gammaemiter which would be detectable with personal dosimetry and thereforewould NOT be a non-detectable radiation concern. Monday, September 29, 2014 7:41:36 AM 280 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.3 Radiation Control G2.3.14 Knowledge of radiation or contamination hazards that may arise duringnormal, abnormal, or emergency conditions or activities.(CFR: 41.12 / 43.4 / 45.10)Importance Rating: 3.4 3.8 Technical

Reference:

AOP-013-BD Rev. 3 AP-617 Rev 35, pg 13References to be provided: AP-617, Attachment 1, Rev. 35 Learning Objective: LPPP2-17, Obj 5; Lesson Plan AOP3-13, Obj 2aQuestion Origin: Previous 2013 NRC SRO radomly selectedComments:

SRO question application tie to 10 CFR Part 55 Content -

43(b)(1) / (4) alsoTier/Group: T3SRO Justification:

10 CFR Part 55 Content - 43(b)(7): Fuel HandlingFacilities and Procedures.

Per NRC ClarificationGuidance for SRO-only Questions, Rev 1 this questionmeets the SRO level of knowledge because the SROmust have knowledge of such as coolant activity andemergency classifications, (i.e. notifying the NRC) basedon core conditions.

Monday, September 29, 2014 7:41:36 AM

281 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal

99. 2014 NRC SRO 024/BANK/FUNDAMENTAL//OP-120.07/NONE//G2.3.6/Which ONE of the following personnel are RESPONSIBLE for preparing and approvinga Batch Gaseous Effluent Permit?

Prepares Permit

Approves Permit Chemistry Control Room Supervisor Chemistry Shift Manager Radiation Protection Control Room Supervisor Radiation Protection Shift Manager A.B.C.D.FeedbackPlausibility and Answer AnalysisReason answer is correct:

In accordance with OP-120.07, Waste Gas Processing, prior to discharging a gaseous batch release a copy of the Discharge Log is given toChemistry for issuance of the Batch Gaseous Effluent Permit (CRC 853). Chemistry prepares the permit in accordance with CRC 853. Prior to starting the release the CR Sreviews and signs the Batch Gaseous Effluent Permit. Additionally, OMM-001,Operations Administrative Requirements responsibilities for the CRS section 4.1.2.1.a states that the CRS is respons ible to ensure operation of radwaste systems is in a safe,efficient manner per approved operating procedures and within requirements of theNational Pollutant Discharge Elimination System permits.

A. Correct. B. Incorrect. The first part is correct. The second part is plausible since the ShiftManager is the senior licensed on-shift crew member in charge of thestation but is not responsible for approving the release of gaseous permitsfor the station. C. Incorrect. Plausible since the release is a radioactive release and RadiationProtection is responsible for monitoring the dose of station activities and related radioactive conditions but they are not responsible for preparingBatch Gaseous Effluent permits. The second part is correct. D. Incorrect. Plausible since the release is a radioactive release and RadiationProtection is responsible for monitoring the dose of station activities and related radioactive conditions but they are not responsible for preparingBatch Gaseous Effluent permits The second part is plausible since theShift Manager is the senior licensed on-shift crew member in charge ofthe station but is not responsible for approving the release of gaseouspermits for the station. Monday, September 29, 2014 7:41:36 AM 282 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.3 Radiation ControlG2.3.6 Ability to approve release permits(CFR: 41.13 / 43.4 / 45.10)Importance Rating: 3.8 Technical

Reference:

OP-120.07, Section 8.12.2, Rev. 67References to be provided: NoneLearning Objective: Lesson Plan Gaseous Waste Processing SystemObjective 5Question Origin: BankComments: None

Tier/Group: T3 SRO Justification:

10 CFR Part 55 Content - 43(b)(4) Radiation hazardsthat may arise during normal and abnormal situations,including maintenance activies and variouscontamination conditions.

Per NRC ClarificationGuidance for SRO-only Questions, Rev 1 this questionmeets the SRO level of knowledge by questioning the process for gaseous/liquid release approvals, i.e.,release permits.

Monday, September 29, 2014 7:41:36 AM

283 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal 100. 2014 NRC SRO 025/NEW/C/A//PEP-330/NONE//G2.4.38/Given the following plant conditions: - An event has occurred that has resulted in the activation of the EmergencyResponse Organization (ERO) - The Technical Support Center (TSC) has completed turnover from the MCR - A Maintenance Mechanic is standing by to enter the RAB to isolate a release inprogress. This will result in a dose of 11,400 mrem TEDE for the individual Which ONE of the following describes the dose limit basis for the type of entry ANDthe position that must approve the expected dose the Mechanic will receive in accordance with PEP-330, Radiological Consequences?

(NOTE: The titles for the positions are listed below) SEC-TSC Site Emergency Coordinator - Technical Support Center RCD Radiological Control Director

_________Limit Basis__________

____Position____Protection of large populations ONLY SEC-TSCProtection of large populations ONLY RCD Protect valuable property AND large populations SEC-TSC Protect valuable property AND large populations RCD A.B.C.D.Monday, September 29, 2014 7:41:36 AM

284 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal FeedbackPlausibility and Answer AnalysisPEP-330 Attachment 1 identifies that exposures up to 25 REM are allowed forlifesaving or protection of large populations (when a lower dose is not practicable). PEP-230, SEC-TSC Attachment 1 step 29 states the following: Authorize ProgressEnergy emergency worker exposures

> 5 Rem TEDE or entry into fields

> 25 Rem/Hr per PEP-330, Att. 1, as necessary.

A. CorrectB. Incorrect- The first part correct. The second part is incorrect. Plausible since theRCD evaluates the need for and coordinates with the SEC -TSC authorization ofthe allowance of onsite emergency worker exposures > 5 Rem TEDE or entry intofields > 25 Rem/Hr per Attachment 8 of PEP-240.C. Incorrect The first part is plausible since the activities table for dose limitsexceeding 10 Rem contains two seperate ac tivities but the other acti vity is forlifesaving NOT protecting valuable propert

y. The second part is correct. D. Incorrect- The first part is plausible si nce the activities table for dose limitsexceeding 10 Rem contains two seperate ac tivities but the other acti vity is forlifesaving NOT protecting valuable prope rty. The second part is incorrect.Plausible since the RCD evaluates the need for and coordinates with the SEC-TSC authorization of the allowance of onsite emergency worker exposures > 5Rem TEDE or entry into fields > 25 Rem/Hr per Attachment 8 of PEP-240. Monday, September 29, 2014 7:41:36 AM

285 QUESTIONS REPORTfor 2014 NRC RO SRO Written Rev 4 - 45 Day Submittal Notes2.4 Emergency Procedures / PlanG2.4.38 Ability to take actions called for in the facility emergency plan, includingsupporting or acting as emergency coordinator if required.(CFR: 41.10 / 43.5 / 45.11)Importance Rating: 4.4 Technical

Reference:

PEP-330, Page 17, Rev. 12 PEP-240, Attachment 8 - RADIOLOGICAL CONTROLDIRECTOR CHECKLIST, Rev.17, page 43References to be provided: None Learning Objective: Lesson Plan EP-LP-2.0/3.0 SRO objective 1Question Origin: Bank, OIT Development, 2012 Audit SRO 23 Comments:

SRO question application tie to 10 CFR Part 55 Content -

43(b)(4) alsoTier/Group: T3SRO Justification:

10 CFR Part 55 Content - 43(b)(5):

Assessment offacility conditions and selection of appropriateprocedures during normal, abnormal, and emergencysituations. Per NRC Clarification Guidance for SRO-onlyQuestions, Rev 1 this question meets the SRO level ofknowledge by questioning the Emergency Coordinator SRO only job function, and thus the EC actions, andresponsibilities for directing entries into hazardous areas during an emergency are SRO-only knowledge.

Monday, September 29, 2014 7:41:36 AM

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