25 sro questions ver5 wednesday, february 29, 2012 page 1 ... · wednesday, february 29, 2012 page...

25 SRO QUESTIONS Ver5Wednesday, February 29, 2012 of 25

76. 003AG2.2.22 076/NEW//C/A 4.0/4.7/003AG2.2.22/TS 3.1.4//IUnit 1 is at 70% power with the following conditions:

• The OATC pulls Control Rods out 2 steps and noles thefollowing indications:- Control Bank D rod P8 drops to 174 steps.- All Control Bank D group step counters indicate 188 steps.- All Control Bank D DRPI, except rod P8, indicate 186 steps.

• An investigation has determined that there is a problem in theRod Control Power Cabinet for Control Rod P8.

Which one of the followingREQUIRED ACTiONS, if any, are applicable perTech Spec 3.1.4, Rod Group Alignment Limits?

REFERENCE PROVIDED

A. CONDITION A only

B’ CONDITION B only

C. Both CONDITION A and CONDITION B

D. Neither CONDITION A or CONDITION B

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25 SRO QUESTIONS Ver5Wednesday, February 29, 2012 Page 1 of 30

76. 003AG2.2.22 076/NEW//C/A 4.0/4.7/003AG2.2.22/TS 3.1.4//I

When a control rod drops out of alignment with the other rods in its bank, an evaluationhad to be made as to why the rod is where it is. The rod can drop due to a problem withthe grippers during mVñent. It may drop to the bottom of the core, or it may becaught again by the grippers. There is CE from our plant where the rod did not fullyinsert. The rod dropped partially into the core and was stuck in that position (previousCE on rod drop testing when coming up out of an outage as a result of control rod tip“swelling”. We had to cool back down and replace some of the older control rods).For the conditions given, a candidate has to evaluate:1) do I have a stuck rod

Management interpretation of the application of Tech Specs is that if a rod dropspartially, we docall it a stuck rod until an investigation is completed to determine if ftis a problem with rod control, oraprobiemwftharodthatis not moveable (stuck). The

it is proven that it is inoperkThititiä[ässmentis that the rod is still operable, but at some point later it will be moved to validate thisassessment. With this question, the candidate has to apply TS bases and managementexpectations for an initial assessment of the TS requirements of 3.1.4.

A. Incorrect Plausible because it has been identified that there is a problem with thatrod in the associated Power cabinet, and a candidateyssume thatthe rod is not trippable as a result. TS 3.1.4 Bases states “Mechanical orelectrical failures may cause a control rod to beöme inoperable or tobecome misaligned from its group.” A candidate may incorrectly interpretthat to mean that an electrical failure may make the rod inoperable(untrippable). Also, the candidate may assume that the rod has to beconsidered stuck (untrippable), until proven otherwise. In addition, theLCO states that “All shutdown and control rods shall be OPERABLE” andin alignment. The candidate may determine that since the rod won’t moveproperly, it is INOPERABLE and CONDITION A is required.

B. Correct Rod P8 DRPI must be aligned within 12 steps of the group step counterindication, and is currently aligned within 14 steps. It is aligned w/i 12steps of the DRPI indications. CONDITION B applies.

C. Incorrect See A for CONDITION A plausibility. CONDITION B is correct, butCONDITION A is not correct.

D. Incorrect A candidate may properly assess that CONDITION A is not required. Hemay also determine that CONDITION B is not required because thedifference between DRPI for the dropped rod and DRPI for other rods inthe bank is within 12 steps. TS 3.1 .4 requires the DRPI indication bewithin 12 steps of the group step counters, not DRPI of the dropped rodwithin 12 steps of the other rods’ DRPI.

Each answer choice above may or may not be correct with a minor adjustment of plantconditions. For example, if the rod was placed at <12 steps then no LCO would be in

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effect. If the investigation revealed that the problem was a rod that would not movethen Condition A and B would be correct. If the rod was placed at <12 steps and theinvestigation revealed that the problem was a rod that would not move, then ConditionA only would be correct.

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76. 003AG2.2.22 076/NEW//C/A 4.0/4.7/003AG2.2.22/TS 3.1.4///

003AG2.2.22Dropped Control Rod

2.2.22 Knowledge of limiting conditions for operations and safety limits.(CFR: 41.5 / 43.2 I 45.2)IMPORTANCE RO 4.0 SRO 4.7

Importance Rating: 4.0 4.7

Technical Reference: Tech Specs, v186Tech Specs Bases, v53

References provided: Tech Spec 3.1.4 page 3.1.4-1 and 3.1.4-2

Learning Objective: RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with Rod Control System, and attendantequipment, to include the following: (OPS-62201 EQ 1):1OCFR55.43 (b) 23.1.4, Rod Group Alignment Limits3.1.5, Shutdown Bank Insertion Limits3.1.6, Control Bank Insertion Limits3.1.7, Rod Position Indication3.1.8, PHYSICS TESTS Exceptions

Question origin: New question.

Basis for meeting K/A: Question tests the knowledge and application of LCO’s andBases for a partially dropped (which is also misaligned)Control Rod.

SRO justification: 10 CFR 55.43(b)(2):Application of Required Actions (Section 3) and SurveillanceRequirements (SR) (Section 4) in accordance with rules ofapplication requirements (Section 1).

2012 NRC examFrom the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour Tech Specs.2) can NOT be answered by knowing information listed “above-the-line”.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:

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• Application of Required Actions (Section 3) and Surveillance Requirements (SR)(Section 4) in accordance with rules of application requirements (Section 1) andknowledge of Tech Spec bases and management expectations regarding RODoperability.

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77. 005G2.4.30 077/NEW/IC/A 2.7/4.1 /005G2.4.30/EIP-8 SECTION/NRC NOTIFY/IThe following conditions exist on Unit 1:

At 10:00:

• RCS temperature is 250°F.• 1 B RHR pump is in service for the RCS cooldown.• All RCS to RHR loop suction isolation MOVs are open.

At 10:05:

• 1 B RHR pump trips due to overcurrent.

At 10:15:

• 1A RHR pump is started and core cooling is restored.

Which one of the following completes the statements below?

(1) INOPERABLE per Tech Spec 3.5.3, ECCS Shutdown.

An 8 hour notification to the NRC Operations Center (NRCOC) (2) requiredper EIP-8.0, Non-Emergency Notifications.

REFERENCE PROVIDED

(1) (2)

A. Only I B RHR pump is IS

B. Only 1 B RHR pump is is NOT

C Both IA and lB RHR pumps are IS

D. Both 1A jjç 1 B RHR pumps are is NOT

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77. 005G2.4.30 077/NEW//C/A 2.7/4.1 /005G2.4.30/EIP-8 SECTION/NRC NOTIFY/I

Step 324 of SOP-7.0:Operation of a train of RHR aligned in cooldown operation when RCStemperature is >225°F will result in the associated train of ECCS being declaredinoperable. One train of ECCS must be operable in Mode 4 (TS 3.5.3).(RER 0101206101)

A. Incorrect 1) Incorrect, 1A RHR pump is also considered inoperable. TS Basesinformation has to be evaluated in addition to SOP-7.0 (step 3.24)information on RHR operability issues. TS Bases states that RHR can beconsidered operable for the ECCS function if it is capable of beingmanually reali ned-totheECG&nic of operation and is not otherwiseinoperable. A ineering evaluatio )has shown that when an RHRpump is in the cooldown alignmen , with its suction water temp > 225°F, ifthe suction is realigned to the lower pressure RWST, flashing and voidingmay occur in the suction line. The RHR pump has to be consideredINOPERABLE if the pump is running in a cooldown alignment with itssuction temp > 225° F. Plausible because IA RHR pump can operate withno issues in the cooldown alignment, and is considered operable per TS3.4.6, RCS Loops - Mode 4, but is not operable for TS 3.5.3, ECCS -

Shutdown, because of the the suction temperature issue.2) Correct, EIP-8.0, step 13.4, requires a NRC notification within 8 hoursfor any event or condition that at the time of discovery could haveprevented the fulfillment of the safety function of structures or systemsthat are needed to (2) Remove residual heat, or (4) Mitigate theconsequences of an accident.ECCS function while in Mode 4, so this event requires a report to theNR000 within 8 hours.A loss of shutdown cooling has occurred for 10 minutes. It is plausiblethat a candidate could determine that NRC notification is required due tothe loss of shutdown cooling and consider that only one RHR pump isinoperable.

B. Incorrect 1) Incorrect, seeA.1.2) Incorrect, See A.2. It is plausible that a candidate would state that NRCnotification is not required with only one RHR pump inoperable.

C. Correct 1) Correct, see A.1.2) Correct, see A.2.

D. Incorrect 1) Correct, see A.1.2) Incorrect, see A.2. It is plausible that a candidate would state that NRCnotification is not required with both RHR pumps inoperable because corecooling was re-established and is being maintained by the running pump.

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77. 005G2.4.30 077/NEW//C/A 2.7/4.1 1005G2.4.3OIEIP-8 SECTION/NRC NOTIFY/I005G2.4.30Residual Heat Removal System

2.4.30 Knowledge of events related to system operation/status that must be reported tointernal organizations or external agencies, such as the State, the NRC, or thetransmission system operator.(CFR: 41.10 /43.5 /45.11)IMPORTANCE RD 2.7 SRO 4.1


Technical Reference: Tech Specs Bases, v53FNP-1-SOP-7.0, vlOOFNP-0-EIP-8.0, V108

References provided: EJP-8.O section 13.0 page 9and 10. (reference providedis that the SRO consult

4 and 8 hour report guidance when making notificationdecisions)

Learning Objective: RECALL AND APPLY the information from the LCD BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLANCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the Residual Heat Removal Systemcomponents and attendant equipment alignment, to includethe following (OPS-62101K01): 1OCFR55.43 (b) 23.5.3, ECCS — Shutdown

Given a set of plant conditions, EVALUATE those conditionsand DETERMINE the emergency responses actions theSM/ED should take to mitigate the consequences of theevent. (OPS-63002A01).


Basis for meeting K/A: Question tests the knowledge of required NRC notificationdue to loss of Residual Heat Removal I Accident mitigationequipment. An 8 hour Non-emergency report is required.

SRO justification: 10 CFR 55.43(b)(2):Knowledge of TS bases that is required to analyze TSrequired actions and terminology. (where amplifyinginstruction is found within the PROCEDURES; Knowledge ofrecently identified design/operational flaws)Although TS allows an RHR train to be considered operablewhen it is aligned in a cooldown alignment, additional

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information is required to properly assess the operability ofthe train of RHR.

The SRO is solely responsible for evaluating Emergencyclassifications and Non-Emergency evaluations fornotification per EIP-8.0.

2012 NRC exam

From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour Tech Specs.2) can NOT be answered by knowing information listed “above-the-line”.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:• Knowledge of TS bases that is required to analyze TS required actions and

terminology. (where amplifying instruction is found within the PROCEDURES.)Bases knowledge of other conditions that can make the RHR system inoperable.

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78. 006A2.02 078/NEW/IC/A 3 .9/4.3/006A2.02/N///A Large Break LOCA has occurred on Unit 1 and Cold leg recirculation has beenestablished per ESP-1.3, Transfer to Cold Leg Recirculation.

Several hours after the return to EEP-1 .0, Loss of Reactor or SecondaryCoolant, a LOSP occurs and B Train emergency busses remain de-energized.

Subsequently, the Control Room team is performing ESP-1 .4, Transfer toSimultaneous Cold and Hot Leg Recirculation.


Simultaneous cold and hot leg recirculation (1) be established perESP-1.4.

At the step to “Check simultaneous cold and hot leg recirculation inprogress”, the Control Room team is required to (2)

(1) (2)

A” CAN return to EEP-1.0

B. can NOT returntoEEP-1.0

C. CAN remain in ESP-1.4

D. can NOT remain in ESP-1 .4

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78. 006A2.02 078/NEW//C/A 3 .9/4.3/006A2.02/N///

For Hot Leg Recirc to be established, at least one train of HHSI is aligned to the HotLegs and at least one train of LHSI aligned to the Cold Legs Q at least one train ofHHSI aligned to the Cold Legs and at least one train of LHSI aligned to the Hot Legs.

In this event with no B Tm power available, MOV-8889 will not have power since it ispowered from FV-B2, a B Tm MCC. MOV-8889 is required for placing either Train ofLHSI on recirc to the Hot Legs, therefore LHSI cannot be aligned to the Hot Legs. LHSIwill have to be aligned to the Cold Legs.

For HHSI to bjJned to the Hot Legs, MOV 88848886 has to be powered up andopened. Both OV 888 and 8886 are normally closd, andpuly. M - 886 has an A Tm power supply so it can be opened. Also MOVs8803A and B must be closed (to prevent runout of the HHSI pumps) since they are inparallel alignment. MOV-8803A can be closed and 8803B has a B Train power supply.However, it can also be aligned to A Train power for this particular case. (MOV-8803Acan NOT be aligned to B Train power).

ESP-1 .4 procedure flowpath can be performed, and will control the alignment to ensureLHSI is aligned to the Cold Leg recirculation and HHSI is aligned to the Hot Leg recirc.This is the only way that the Simultaneous Cold and Hot Leg injection can be aligned.

At step 4 to “Check simultaneous cold and hot leg recirculation in progress by one ofthe following:” The RNO column says if it is not established, then consult TSC andreturn to step 1. Step 6 says to return to procedure and step in affect. The stem showsthe user is in EEP-1 .0 when the event started so the return would be back here.

A. Correct 1) HOT LEG Recirculation can be established since the LHSI system canbe aligned.2) Correct, Since hot and cold leg recirc is aligned a return to EEP-1 .0 isrequired per ESP-1 .4.

B. Incorrect 1) Incorrect, see A.1

Plausibility: Several issues have to be resolved. One is power available to place RHRto the hot legs, Since the answer is no, is power available or required toplace RHR on the cold legs. The answer is that RHR can be aligned tothe cold legs with A Train power available. Next, is power available toplace HHSI on the hot legs. MOV-8886 has power available butMOV-8884 does not. ONLY one of these valves is required to place HHSIon the hot legs. Then the next question asks - is power available toremove HHSI from going to the cold legs. Since MOV-8803A has onlyone power supply and it is A Train, then it can be closed, MOV-8803Bhas two power supplies and has to be be aligned to the alternate A Trainpower supply to enable closing this valve. Therefore HHSI can be placedon the hot leg. Minute detailed system and/or procedure knowledge isrequired to figure out the status of the ECCS system.

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25 SRO QUESTIONS VerSWednesday, February 29, 2012 Page 5 of 30

2) Correct, See A.2.

C. Incorrect 1) Correct, seeA.1.2) Incorrect, ESP-1 .4 would require one to remain in this procedure if hotand cold leg recirc was not established.

Plausibility: If hot and cold leg recirc were established, a candidate may believe thatremaining in this procedure is warranted. Both this procedure andEEP-1 .0 turn over control to the TSC but for different reasons. ESP-1 .4turns over control to the TSC and remains in this procedure if HL/CLrecirc cannot be established and in EEP-1 .0 control is returned to theTSC since there are no other procedural actions.

D. Incorrect 1) Incorrect, see B.1.2) Incorrect, see C.2.

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78. 006A2.02 078/NEW//C/A 3 .9/4.3/006A2.02/N///006A2.02Emergency Core Cooling System (ECCS)

Ability to (a) predict the impacts of the following malfunctions or operations on theECCS; and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:A2.02 Loss of flow path 3.9 4.3


Technical Reference: FNP-1-ESP-1.4, v15FNP-1-EEP-1.0, v30

References provided: None

Learning Objective: ASSESS the facility conditions associated with the (1)ESP-1.3, Transfer to Cold Leg Recirculation; (2) ESP-1.4,Transfer to Simultaneous Cold Leg and Hot LegRecirculation, and based on that assessment:(OPS-62531 GOl)• SELECT the appropriate procedures during normal,abnormal and emergency situations. 1 0CFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required


Basis for meeting K/A: Question tests the ability of a candidate to predict the effectsof the loss of B Train emergency power and therefore a lossof one train of ECCS flowpath (LHSI hot leg recirc is notpossible), and based on that prediction, either remain inESP-1 .4 or transition to EEP-1.

SRO justification: 10 CFR 55.43(b)(5)Assessing plant conditions (normal, abnormal, oremergency) and then selecting a procedure or section of aprocedure to mitigate, recover, or with which to proceed.

2012 NRC exam

From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart for 10 CFR 55.43(b)(5):Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then2) selecting a procedure or section of a procedure to mitigate, recover, or with

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which to proceed.U sing the flowchart, this question can:• NOT be answered solely by knowing “systems knowledge”, i.e., how the system

works, flowpath, logic, component location.• NOT be answered solely by knowing immediate operator actions.• NOT be answered solely by knowing entry conditions for AOPs or plant parameters

that require direct entry to major EOPs.• NOT be answered solely by knowing the purpose, overall sequence of events, or

overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:— Assessing plant conditions (normal, abnormal, or emergency) and then

selecting a procedure or section of a procedure to mitigate, recover, or with whichto proceed.

— Knowledge of diagnostic steps and decision points in the EOPs that involvetransitions to event specific sub procedures or emergency contingency procedures.

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78. 006A2.02 078/NEW//C/A 3 .9/4.3/006A2.02/N///006A2.02Emergency Core Cooling System (ECCS)

Ability to (a) predict the impacts of the following malfunctions or operations on theECCS; and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:A2.02 Loss of flow path 3.9 4.3


Technical Reference: FNP-1-ESP-1.4, v15FNP-1-EEP-1 .0, v30


Learning Objective: ASSESS the facility conditions associated with the (1)ESP-1.3, Transfer to Cold Leg Recirculation; (2) ESP-1.4,Transfer to Simultaneous Cold Leg and Hot LegRecirculation, and based on that assessment:(OPS-62531 GOl)• SELECT the appropriate procedures during normal,abnormal and emergency situations. 1OCFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required


Basis for meeting K/A: Question tests the ability of a candidate to predict the effectsof the loss of B Train emergency power and therefore a lossof one train of ECCS flowpath (LHSI hot leg recirc is notpossible), and based on that prediction, either remain inESP-1 .4 or transition to EEP-1.


2012 NRC exam

From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart for 10 CFR 5543(b)(5):Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:


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which to proceed.Using the flowchart, this question can:• NOT be answered solely by knowing “systems knowledge”, i.e., how the system





— Knowledge of diagnostic steps and decision points in the EOPs that involvetransitions to event specific sub procedures or emergency contingency procedures.

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79. 008A2.01 079/NEW//C/A 3.313.6/008A2.O1/TS 3.7.7//IUnit I is in Mode 3, preparing for Reactor start up, with the following conditions:

At 10:00:

• Both trains of CCW are in operation.• ‘B’ Train of COW is the ON-SERVICE Train.• During a review of STP-23.3, 10 Component Cooling Water Pump Quarterly

Inservice Test, it was discovered that the 10 COW pump failed the STP due tohigh vibrations.

• Repairs to 10 COW pump are expected to take 48 hours.

At 11:00:

• ON-SERVICE Trains have been swapped and now ‘A’ Train of COW isthe ON-SERVICE Train.

At 11:30:

• DFO4, 10 COW PUMP, breaker is racked out.

Which one of the following provides the earliest time that Mode 2 can be entered,while still ensuring compliance with Tech Specs?

Mode 2 entry is allowed

REFERENCE PROVIDED

A. at 10:00

B. at 11:00

C’ at 11:30

D. only afterrepair are-oGrnp anci the IC COW pump is returnedo. OPERABLE tatus

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79. 008A2.01 079/NEW//C/A 3.3/3.6/008A2.01/TS 3.7.7//I

The CCW system has two trains and 3 pumps. The 1A COW pump is always B train,the 1 C CCW pump is always A train, and the 1 B COW pump can be A train or B train,but will always be aligned to the “on service train”. Initially, B train is the “on servicetrain” (lB pump is the backup to 1A pump), so when IC pump (on A train) becomesinoperable, there is one train of CCW inoperable. At 11:00 when the train swap iscompleted, lB COW pump is on the same train as the inoperable IC COW pump andcan provide a backup function. The A train of COW is still inoperable untHthekerfoi1COcW. pump is r d out, to ensure that the lB pump will start in place of the I Cpump.

A. Incorrect A Train COW is inoperable at this time, per SR 3.0.1, since it isdiscovered that the pump failed its Surveillance Test. The SRO isrequired to call the pump inoperable at that time.Plausible that a candidate may consider the pump operable since it is stillrunning. LCO 3.0.4 prevents a Mode change while one train of COW isinoperable, so Mode 2 entry is not allowed. In addition, since repairs canbe completed within the COMPLETION TIME allowed by the REQUIREDACTIONS (48 hours for repair vs 72 hours for COMPLETION TIME), thecandidate may think that actions may continue to make a mode change.

B. Incorrect A Train COW is inoperable at this time since the pump that will be startedautomatically on an Sl/LOSP is the 10 COW pump. Per SOP-23.0, P&L2.4, the A train of COW is still inoperable. Either a) 10 COW pumpbreaker must be racked out, or b) the cell switch should be jumpered anda link opened to prevent 10 pump from starting. Either of these actionswill restore the train to OPERABLE.Plausible that a candidate may consider the A Train CCW operable at thispoint since the train swap is completed. The 1 B pump will startautomatically if IC pump trips, so it is plausible that a candidate wouldconsider that train OPERABLE.LCO 3.0.4 prevents a Mode change while one train of COW is inoperable,so Mode 2 entry is not allowed.

C. Correct Once the on service train has been swapped, and the IC COW pumpbreaker racked out, both trains of COW are operable. I B COW pump willbe running on A train and will auto start as required on an Sl/LOSP signal.See SOP-23.0, P&L 2.3.

D. Incorrect Both A train and B train are operable at 11:30. Tech Specs provides nolimitations on Mode changes with both COW trains operable. 10 COWpump is still inoperable, so it is plausible that a candidate may determinethat an LCO still exists and LCO 3.0.4 will prevent any Mode change untilafter repairs are complete.Plausible since an ADMIN LCO is required to be written by plantprocedures. The ADMIN LCO is for tracking purposes but could bemistaken as a reason to not change modes until the Admin LCO is clear.

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79. 008A201 079/NEW//C/A 3.3/3.6/008A2.O1/TS 3.7.7//I008A2.0 1Component Cooling Water System (CCWS)

Ability to (a) predict the impacts of the following malfunctions or operations on theCOWS, and (b) based on those predictions, use procedures to correct, control,or mitigate the consequences of those malfunctions or operations:(CFR: 41.5 / 43.5 / 45.3 / 45.13)A2.O1 Loss of COW pump 3.3 3.6


Technical Reference: FNP-1 -SOP-23.0, v91 .1FNP-1-STP-23.3, v43.1Tech Specs, v186

References provided: Tech Spec 3.7.7 page 3.7.7-1

Learning Objective: RECALL AND APPLY the information from the LOO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENOE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the COW System components andattendant equipment alignment, to include the following(OPS-621 02G0 1): 1 00FR55.43 (b) 23.7.7 Component Cooling Water System


Basis for meeting K/A: For the COW system, this question tests the ability of acandidate to predict the impact of the loss of a COW pump(impact related to TS and Mode changes), and based onthat prediction, use the guidance of TS to determine when aMode change can be performed.

SRO justification: 10 CFR 55.43(b)(2) meeting all of the following:• Application of generic LCO requirements (3.0.4)• Knowledge of TS bases that is required to analyze TS

required actions and terminology (where amplifyinginstruction is found within the PROCEDURES.)

2012 NRC exam

10 CFR 55.43(b)(2)

Facility operating limitations in the TS and their bases.From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart:

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1) can NOT be answered by knowing less than 1 hour Tech Specs.2) can NOT be answered solely by knowing information listed11above-the-line”.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:

• Application of Required Actions (Section 3) and Surveillance Requirements(Section 4) in accordance with rules of application requirements (Section 1).

• Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR40.1 thru 4.0.4)Requires assessment of TS 3.0.4 mode change allowance TS.

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80. 01 0G2.4.45 080/MOD/SURRY 2009/C/A 4.1/4.3/01 0G2.4.45/N///Unit I has performed a rapid ramp down from 85% to 55% power With thefollowing conditions:

• A malfunction occurs resulting in PCV-444B, PRZR PORV, going open.• PCV-444B handswitch has been taken to CLOSE, the GREEN light is LIT.• PRT parameters are stable.• RCS pressure is 1875 psig and slowly t.• Multiple alarnisare in, with the following annunciators being evaluated:

- GB2, PRZRLO PRESS RX TRIP- HCI, PRZR PRESS HI-LO- HC2, PRZR HI-LO PRESS ALERT


AReactorTrip (1) required.

PCV-444B (2) OPERABLE per Tech Spec 3.4.11, Pressurizer PowerOperated Relief Valves (PORVs).

(1) (2)

A’ IS IS

B. IS is NOT

C. is NOT IS

D. isNOT isNOT

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80. 01 0G2.4.45 080/MOD/STJRRY 2009/C/A 4.1/4.3/01 0G2.4.45/N///

EEP-0, contains the following information:The following are symptoms of a reactor trip:a. Any reactor trip annunciator lit.b. Rapid decrease in neutron level indicated by nuclear instrumentation.c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.

A. Correct 1) Correct, A reactor trip is required per EEP-0 based on having anannunciator in alarm requiring a reactor trip. This alarm is on the GAnnunciator panel (First Out Panel), and is a symptom that a Reactor triphas occurred or is called for. This annunciator will only come in when the2 out of 3 transmitters Pressurizer low pressure coincidence is met. Thealarm is fed from the Solid State Protection System (SSPS).2) Correct, knowledge of TS Bases information is required to evaluate thePORV operability with the handswitch in the closed position and a failurethat will cause the PORV to NOT function automatically.

B. Incorrect 1) Correct, see A.1.2) Incorrect, the PORV is operable. Plausible because the handswitch isin the CLOSE position and the PORV can no longer operateautomatically. TS Bases knowledge is required to determine operabilitystatus of the PORV.

C. Incorrect 1) Incorrect, with annunciator GB2 in alarm, a reactor trip should haveoccurred. It is plausible that a candidate would say a reactor trip is notrequired since the actual Pressurizer pressure is above the reactor tripsetpoint of 1865 psig, put the reactor trip signal is rate compensated anda trip may be demanded when pressure is higher than the actual setpoint.In addition, some annunciators will illuminate when I out of 3 transmittersindicate that trip criteria is met.2) Correct, see A.2.

D. Incorrect 1) Incorrect, see C.1.2) Incorrect, see B.2.

Page8of3o


80. 010G2.4.45 080/MOD/SURRY 2009/C/A 4.1/4.3/010G2.4.45/N///01 0G2.4.45Pressurizer Pressure Control System (PZR PCS)

2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.(CFR: 41.10/43.5/45.3/45.12)IMPORTANCE RO 4.1 SRO 4.3


Technical Reference: Tech Specs Bases, v53FNP-1-EEP-0 v43FNP-1-ARP-1.7 v33


Learning Objective: EVALUATE plant conditions and DETERMINE if entry into(1) EEP-0, Reactor Trip or Safety Injection and/or (2)ESP-0.0, Rediagnosis is required. (OPS-52530A02).

RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the Pressurizer System components andattendant equipment alignment, to include the following(OPS-62101E01): IOCFR55.43 (b) 23.4.9, Pressurizer3.4.10, Pressurizer Safety Valves3.4.11, Pressurizer Power Operated Relief Valves13.4.2, Pressurizer13.4.4, Safety Valves - Shutdown

Question origin: Modified from Surry 2009 NRC SRO Exam question #13.

Basis for meeting K/A: Question tests the ability of the candidate to evaluate alarmsand determine if a reactor trip is required. Although thePressurizer pressure is not indicating it is at the setpoint fora reactor trip (the signal is rate compensated), anannunciator is indicating that a reactor trip is required.Candidate has to prioritize and interpret the significance ofthe alarm for Pressurizer pressure control.

SRO justification: 10 CFR 55.43(b)(2):Knowledge of TS bases that is required to analyze TSrequired actions and terminology.Knowledge of TS Bases is required to evaluate operability ofthe PORV with the handswitch in the CLOSE position. Even

Page9of5l


with no Auto capability, the PORV is considered operable.



terminology.

Page 10 of 51

SRO Portion of Exam13. Unit 1 initial conditions:

Time = 1000Reactor power = 100%1 -RC-PORV-1 455C (Pressurizer Pressure PORV) indicates openBoth Pzr Spray valves indicate openRCS Pressure = 2200 psig decreasing1 -AP-31 .00 (Increasing or Decreasing RCS Pressure) initiated

Current conditions:Time = 1001Reactor Power = 97%RCS Pressure = 2100 psig increasingSpray valve in MANUAL and closed1-RC-PORV-1455C in MANUAL and closed

Based on the above conditions, which ONE of the following correctly states: (1) thecomponent that failed high and (2) the status of 1 -RC-PORV-1 4550 operability perTechnical Specifications?

A. (1) 1-RC-PT-1 444 (Pressurizer Pressure Control)(2) PORV is considered OPERABLE

B. (1) 1-RC-PT-1444 (Pressurizer Pressure Control)(2) PORV is INOPERABLE

C. (1) 1-RC-PT-1445 (Pressurizer Pressure Control)(2) PORV is considered OPERABLE

D. (1) 1-RC-PT-1445 (Pressurizer Pressure Control)(2) PORV is INOPERABLE

13


81. 01 1EA2.02 08 I/NEW//C/A 3.3/3.7/011 EA2.02/N///Unit 1 has experienced a LOCA with the following conditions:

• EEP-1.O, Loss of Reactor or-Secondary-Coolant;isinprogress.• The following annunciators are in alarm:

- CH2, RWST LVL A TRN LO- CH3, RWST LVL B TRN LO

• The Safety Injection signal can NOT be reset.


Transition to (1) is required for these conditions.

Once the procedure transition is made, if a Red or Orange path CSF occurs,Functional Restoration Procedures are required to be implemented (2)

A’ 1) ESP-1 .3, Transfer to Cold Leg Recirculation

2) when directed by ESP-1 .3

B. 1) ESP-1.3, Transfer to Cold Leg Recirculation

2) immediately

C. 1) ECP-1 .1, Loss of Emergency Coolant Recirculation

2) when directed by ECP-1 .1

D. 1) ECP-1.1, Loss of Emergency Coolant Recirculation

2) immediately

Page 6 of 25


81. 01 1EA2.02 081/NEW//C/A 3.3/3.7/0 1 1EA2.02/N///

SI reset is normally performed in most of the ERG procedures. Reset of SI allows anoperator to regain control of equipment that actuated due to the SI signal. Most of thetime, reset of SI is required to reposition a valve (to place it in other than its SI requiredposition), and pumps can be started or stopped regardless of the SI signal. fçjjjRHR systern,pumps are started by the SI signa butvalvesarealready inthirrequired position for Safety Injection If the candidate thinks that some of the valvesreceive an SI signal, he will assume that they cannot be repositioned for ESP-1 .3, andmay choose to go to ECP-1 .1 for loss of recirculation capability.

A. Correct 1) Correct, an SI signal does not affect any of the valves required toestablish Cold Leg recirculation.2) Correct, FRP’s will be implemented after step 7 of ESP-1 .3, as directedby a note in ESP-1 .3. This is acceptable per SOP-O.8, TransientResponse Procedure User’s Guide, but is contrary to the normalimplementation of FRP’s. Detailed procedure knowledge of ESP-1 .3 isrequired.

B. Incorrect 1) Correct, see A.1.2) Incorrect, for most procedures, a red or orange path FRP will beimplemented immediately when it occurs. ESP-1 .3 has a note that statesFRP’s will not be implemented until after step 7 of the procedure.

C. Incorrect 1) Incorrect, Plausible if candidate believes that failure to reset SI willaffect the operation of MOV’s required to be repositioned for Cold Legrecirculation.2) Incorrect, ECP-1 .1 is not the appropriate procedure, but plausiblebecause some procedures delay implementation of FRP’s upon entry.

D. Incorrect 1) Incorrect, see C.1.2) Correct, if the candidate chose ECP-1 .1, the FRP’s would beimplemented immediately.

Page9of3o


81. 01 1EA2.02 08 1/NEW//C/A 3.3/3.7/011 EA2.02/N///

O1IEA2.02Large Break LOCA

Ability to determine or interpret the following as they apply to a Large Break LOCA:(CFR 43.5 /45.13)EA2.02 Consequences to RHR of not resetting safety injection 33* 37*


Technical Reference: FNP-1-ESP-1.3, v22FNP-1-ARP-1.3, v28.1FNP-0-SOP-0.8, v20


Learning Objective: ASSESS the facility conditions associated with the (1)ESP-1.3, Transfer to Cold Leg Recirculation; (2) ESP-1.4,Transfer to Simultaneous Cold Leg and Hot LegRecirculation, and based on that assessment:(OPS-62531 GOl)• SELECT the appropriate procedures during normal,abnormal and emergency situations. 10CFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required

Explain and apply the foldout page requirements duringemployment of the Emergency Response Procedures(ERPs) (OPS52301 B06).


Basis for meeting K/A: Question tests the candidate’s knowledge of the effects ofnot being able to reset an SI signal during a Large BreakLOCA. Candidate has to determine how the failure to resetSI affects the operation of RHR valves required forestablishing Cold Leg recirculation.

SRO justification: 10 CFR 55.43(b)(5) Knowledge of administrative proceduresthat specify hierarchy, implementation, and/or coordinationof plant normal, abnormal, and emergency procedures.

- Knowledge of how Functional Restoration Procedures areimplemented using the guidance of Administrativeprocedure - SOP-0.8, Transient Response Procedure User’sGuide is an SRO function. Normally an FRP would beimplemented immediately during an ESP or ECP, butdetailed procedure knowledge is required to know that

Page 11 of 51


FRP’s are not implemented until after directed in ESP-1 .3.This is allowed per the guidance of SOP-0.8.

2012 NRC exam

From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart for 10 CFR 55.43(b)(5):Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, andemergency procedures.This question tests knowledge of when to implement Functional RestorationProcedures.

Using the flowchart, this question can:• NOT be answered solely by knowing “systems knowledge”, i.e., how the system



overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:—Knowledge of administrative procedures that specify hierarchy, implementation,

and/or coordination of plant normal, abnormal, and emergency procedures.

Page 12 of 51


82. 024AG2.4.46 082/NEW//C/A 4.2/4.2/024AG2.4.46/COLR FIG I//IUnit I is stable at 55% power, after a rapid ramp down, with the following conditions:

• Control Bank D is at 115 steps.• Control Bank D Rod K1O is stuck at 144 steps.• Control Rods have been placed in Manual.• STP-29.5, Shutdown Margin Calculation in Modes I and 2 (TAVG 547°F),

shows current Shutdown Margin is 1.67% deltaKlK.• An Emergency Boration is in progress.


FF2, CONT ROD BANK POSITION LO-LO, (1) in alarm.

Per the Bases of TRM 13.1.1, Shutdown Margin (SDM) - Modes 1 and 2, themost limiting accident for Shutdown Margin requirements is based on (2)

REFERENCE PROVIDED

A’ 1) is NOT

2) an EOLTaLnoioad temperature, Steam Line Break accident

B. 1)is NOT

2) a BOL, 100% power, Control Rod Ejection accident

C. 1)IS

2) an EOL, Tavg at no load temperatureSteam Line Break accident

D. 1)IS

2) a BOL, 100% power, Control Rod Ejection accident

Page 7 of 25


82. 024AG2.4.46 082/NEW//C/A 4.2/4.2/024AG2.4.46/COLR FIG 1//I

EEl, CONT ROD BANK POSITION LO, alarm comes in at 10 steps> than the LO LOalarm. FE2, CONT ROD BANK POSITION LO LO alarm comes in based on the RILcomputer which corresponds to the COLR.

A. Correct 1) Correct, Control Rods are still above their Rod Insertion Limit per theCOLR.2) Correct, per TRM 13.1.1 Bases - “SHUTDOWN MARGIN requirementsvary throughout core life as a function of fuel depletion, RCS boronconcentration, and RCS Tavg. The most restrictive condition occurs atEOL, with Tavg at no load operating temperature, and is associated witha postulated steam line break accident and resulting uncontrolled RCScooldown.”

B. Incorrect 1) Correct, see A.1.Plausible since there are 2 alarms with different setpoints and thecandidate has to know how to use the graph, correlate the reading to oneof 2 alarm setpoints to make the decision.2) Incorrect, plausible because Bases of TS 3.1.1 talks about both thesteam line break accident and the rod ejection accident, but the steamline break is the most limiting.

C. Incorrect 1) Incorrect, Control Rods are still above their Rod Insertion Limit per theCOLR.2) Correct, see A.2.

D. Incorrect 1) Incorrect, see C.1.2) Incorrect, see B.2.

Page lOof3O


82. 024AG2.4.46 082/NEW//C/A 4.2/4.2/024AG2.4.46/COLR FIG 1//I024AG2.4.46Emergency Boration

2.4.46 Ability to verify that the alarms are consistent with the plant conditions.I(CFR:41.1O/43.5/45.3/45.12)IMPORTANCE RO 4.2 SRO 4.2


Technical Reference: FNP-1 -STP-29.5, v5Technical Requirements Manual, v24Technical Requirements Manual Bases, v9COLR, v24-1

References provided: Figure 1, Rod Bank Insertion Limits versus Rated ThermalPower from the COLR (Curve for RIL)

Learning Objective: DETERMINE AND APPLY the information from the LCOBASES sections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with AOP-19, Malfunction of Rod ControlSystem, components and attendant equipment.(OPS-62520S01) 10CFR55.43 (b) 2


Basis for meeting K/A: An Emergency Boration is in progress due to inadequateShutdown Margin. This question tests the candidate’s abilityto determine if the annunciator for Control Rods beinginserted below the Rod Insertion Limit (RIL) should be in atthis time. An emergency boration is required for eitherinadequate Shutdown Margin or Rods inserted below theRIL.

SRO justification: 10 CFR 55.43(b)(2):Knowledge of TS bases that is required to analyze TSrequired actions and terminology.Same items listed above for the Technical RequirementsManual (TRM) and Offsite Dose Calculation Manual(ODCM).- The second part of the question tests the candidate’sknowledge of TRM 13.1.1 Bases for why an emergencyboration is required to restore Shutdown Margin.

2012 NRC exam

Page 13of51

25 SRO QUESTIONS Ver5Wednesday, February 29, 2012 Page l4of 51

10 CFR 55.43(b)(2)Facility operating limitations in the TS and their bases.From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart:

1) can NOT be answered by knowing less than 1 hour Tech Specs.2) can NOT be answered solely by knowing information listed “above-the-line”.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:

• Knowledge of TS bases that is required to analyze TSrequired actions and terminology.

Page 14 of 51


83. 025AA2.05 083/NEW//C/A 3.1/3.5/025AA2.05/N///Unit I is shutdown with the following conditions:

• The RCS level is 1 foot below the flange.• Reactor Head stud detensioning is in progress.• 1A RHR pump begins to cavitate and is shutdown.• Actions to restore RHR are in progress.

Which one of the following states the RHR minimum flow requirement tobeand the basis for that requirement per TS 3 9 5, Residual heat

Removal (RHR) and Coolant Circulation -- Low Water Level?

RHR flow is required to be restored to a minimum of (1) inorder to (2)

(1) (2)

A. 1750 gpm prevent thermal and boron stratification in the core

B 3000 gpm prevent thermal and boron stratification in the core

C. 3000 gpm prevent exceeding 110% of RHR design dischargepressure during an RCS over pressurization event

D. 1750 gpm prevent exceeding 110% of RHR design dischargepressure during an RCS over pressurization event

Page8of25


83. 025AA2.05 083/NEW//C/A 3.1/3.5/025AA2.05/N///

TS SR 3.9.5.1 requires 3000 gpm minimum RHR flow during Mode 6 for “sufficientdecay heat removal capability and to prevent thermal and boron stratification in thecore”. There is no TS minimum RHR flow requirement in Mode 5, but there are otherminimum flow requirements per SOP-7.0 Precaution & Limitations.

P&L 3.17 states - In order to prevent exceeding 110% of RHR design dischargepressure during an RCS over pressurization event; the minimum allowable flowrate, forUnit 1, is 1750 gpm when the RHR loop is aligned to the RCS. This limitation would notbe applicable when RCS over pressurization is not feasible (i.e., with the reactorvessel head removed, midloop, Reactor Vessel Cover installed, etc.Reference: Westinghouse letter ALA-95-580).

A. Incorrect 1) Incorrect, 1750 gpm is a minimum RHR flow requirement to preventexceeding 110% of the design discharge pressure during an RCSoverpressurization event - SOP-7.0, P&L 3.17.2) Correct, TS 3.9.5, SR 3.9.5.1 Bases states that this flow is required toprovide “sufficient decay heat removal capability and to prevent thermaland boron stratification in the core.”

B. Correct 1) Correct, Tech Specs SR 3.9.5.1 requires > 3000 gpm RHR flow whilein Mode 6.2) Correct, See A.2.

C. Incorrect 1) Correct. see B.1.2) Incorrect, plausible because this is a RHR requirement for Mode 5,along with the 1750 gpm. The plant is currently in Mode 6. See P&L 3.17of SOP-7.0.

D. Incorrect 1) Incorrect, see A.1.2) Incorrect, see C.2.This combination would be a correct answer for a minimum flowrate whilein Mode 5, although not a TS minimum flow rate.

Page 11 of 30


83. 025AA2.05 083/NEW//C/A 3.1/3 .5/025AA2.05/N///025AA2.05Loss of Residual Heat Removal System (RHRS)

Ability to determine and interpret the following as they apply to the Loss of ResidualHeat Removal System:(CFR: 43.5 I 45.13)AA2.05 Limitations on LPI flow and temperature rates of change 3.1* 35*


Technical Reference: Tech Specs, v186FNP-1-SOP-7.0 vlOO


Learning Objective: RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the Residual Heat Removal Systemcomponents and attendant equipment alignment, to includethe following (OPS-62101K01):3.9.5, Residual Heat Removal (RHR) and CoolantCirculation - Low Water Level


Basis for meeting K/A: A loss of RHR has occurred and the candidate is questionedon the Tech Spec requirements and limitations forrestoration of RHR flow. In addition, the candidate isquestioned on the bases for the RHR minimum flow while inMode 6. When RCS level is low, TS limits require aminimum amount of flow for core cooling and maintaining ahomogenous mixture to prevent thermal and boronstratification.


2012 NRC exam

10 CFR 55.43(b)(2)Facility operating limitations in the TS and their bases.

Page 15of51

25 SRO QUESTIONS Ver5Wednesday, February 29, 2012 Page 16of51

From the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart:


• Application of Required Actions (Section 3) and Surveillance Requirements(Section 4) in accordance with rules of application requirements (Section 1)

Page 16of51


84. 026A2.03 084/BANK/WATTS BAR 2009/C/A 4.1/4.4/026A2.03/N///Unit I is at 100% power with the following conditions:

• PT-950, CTMT PRESS (Channel 1), failed and is out of service with thechannel bistables positioned as required by Tech Specs

• The Surveillance Test for PT-953, CTMT PRESS (Channel 4), is now due.

Which one of the following completes the statements below for performing theSurveillance Test on PT-953?

PT-950 bistables are required to be placed in the (1) position.

Subsequent testing of PT-953 will (2)

(1) (2)

A. trip still allow a valid automatic Containment Sprayactuation to occur.

B’ bypass still allow a valid automatic Containment Sprayactuation to occur.

C. trip prevent a valid automatic Containment Sprayactuation from occurring.

D. bypass prevent a valid automatic Containment Sprayactuation from occurring.

Page 9 of 25


84. 026A2.03 084/BANK/WATTS BAR 2009/C/A 4.1/4.4/026A2.031N///

PT-950 only has one function associated with the PT and that is Hi-3. PT-951, 952 and953 have Hi-I, 2 and 3 associated with the PTs and this can be confusing. Hi-I and 2are placed in the trip condition and if 2 PTs were placed in the trip condition, Hi-l and 2conditions would result. Since PT-950 has no Hi-I and 2 bistables, these functions arenot affected. Also these PTs have a Hi-3 bistable which is bypasses vs. tripped.

Bases B page 3.3.2-14This Function requires the bistable output to energize to perform its required action. It is notdesirable to have a loss of power actuate containment spray, since the consequences of aninadvertent actuation of containment spray could be serious. Note that this Function also hasthe inoperable channel placed in bypass (disabled) rather thanrip to decrease thejpbabjUtyiñädièitêñt actuation The Containment Pressure High 3 instrument Functionconsists of a two-out-of-four logic configuration. Since containment pressure is not used forcontrol, this arrangement exceeds the minimum redundancy requirements. Additionalredundancy is warranted because this Function is energize to trip. Containment Pressure —

High 3 must be OPERABLE in MODES 1, 2, and 3 when there is sufficient energy in theprimary and secondary sides to pressurize the containment following a pipe break.

Condition E - page 3.3.2-36The Required Actions are modified by a Note that allows one additional channel to be bypassedfor up to 12 hours for surveillance testing. Placing a second channel in the bypass condition forup to 12 hours for testing purposes is acceptable based on the results of Reference 11.

A. Incorrect 1) Incorrect, bistables are required to be placed in bypass for the Hi-3function. Plausible because most of the time bistables are placed in tripfor a PT failure, and in fact normally they are placed in trip for the Hi-I orHi-2 functions of a Containment Pressure channel.2) Correct, For the Hi-3 function of Containment pressure, PT-950 andPT-953 will both be bypassed, but PT-951 and PT-952 will still beavailable to initiate a CS and Phase B actuation on a 2 out of 4 signal.

B. Correct 1) Correct, Tech Specs requires going to bypass within 72 hours on thefailed channel for the Hi-3 function, TS 3.3.2, Cond E. In addition, a notein Cond E allows one additional channel to be bypassed for up to 12hours for surveillance testing.2) Correct, see A.2.

C. Incorrect l)Incorrect, seeA.I.2) Incorrect, a CS and Phase B actuation can still occur. Plausible if acandidate does not know that the Hi-3 actuation occurs on 2 out of 4transmitters, as most actuations are on a 2 out of 3 transmitters.

D. Incorrect 1) Correct, see B.1.2) Incorrect, see C.2.

Page 12 of 30


84. 026A2.03 084/BANK/WATTS BAR 2009/C/A 4.1/4.4/026A2.03/N///026A2.03Containment Spray System (CSS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CSS;and (b) based on those predictions, use procedures to correct, control,or mitigate the consequences of those malfunctions or operations:(CFR: 41.5 /43.5 /45.3/45.13)A2.03 Failure of ESF 4.1 4.4


Technical Reference: Tech Specs, v186


Learning Objective: RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with Reactor Protection System (RPS), andattendant equipment, to include the following:(OPS-62201 101): 1 OCFR55.43 (b) 23.3.1 Reactor Trip System (RTS) Instrumentation3.3.2 Engineered Safety Feature Actuation System(ESFAS) Instrumentation

Question origin: Same as Watts Bar 2009 NRC Exam question #88

Basis for meeting K/A: Applicant must determine how the bistables will beconcurrently configured on a failed Containment SpraySystem actuation ESF transmitter and an ESF transmitterwhich is required to be tested. Candidate has to be able todetermine the impact of the failure and how the function ismaintained as identified in the Technical Specificationbases.

SRO justification: 10 CFR 55.43(b)(2)Knowledge of TS bases and notes that is required toanalyze TS required actions and terminology.

2012 NRC exam

10 CFR 55.43(b)(2)

Facility operating limitations in the TS and their bases.From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart:

Page 17of51




Page 18 of 51

I Watts Bar 2009Friday, January 20, 2012 Page 1 of 3

1. 026 A2.03 088/NEW//HIGHERJ/SRO/WATTS BAR/i 1/2009/NOGiven the following:

- Unit 1 at 100% power.- Containment Pressure Transmitter “1-PDT-30-43” (Channel 1W) failed and

is out of service with the channel bistables positioned as required byTechnical Specifications.

- The Surveillance Instruction for “1-PDT-30-44” (Channel II) is now due.

Which ONE of the following describes the required action for performing theSurveillance Instruction on “1-PDT-30-44,” and what is the impact onContainment Spray actuation?

A. “1-PDT-30-43” is required to be placed in the TRIPPED position;subsequent testing of “1 -PDT-30-4411will still allow a valid automaticContainment Spray actuation to occur.

B. “1-PDT-30-43” is required to be placed in the TRIPPED position;subsequent testing of “1 -PDT-30-44” will prevent a valid automaticContainment Spray actuation from occurring.

C” “1-PDT-30-43” is required to be placed in the BYPASS position;subsequent testing of “1 -PDT-30-44” will still allow a valid automaticContainment Spray actuation to occur.

D. “1-PDT-30-43” is required to be placed in the BYPASS position;subsequent testing of “1-PDT-30-44” will prevent a valid automaticContainment Spray actuation from occurring.

Page 1 of 3


DIS TRACTOR ANAL YSIS:

A. Incorrect, 1-PD T-30-43 will not be placed to the f position and subsequent testingof 1-PDT-30-44 will still allow vaild automatic actuation of the containment sprayduring the surveillance test because the other 2 channels still provide for theprotection of the function. Plausible because the positioning of bistables to the tripposition is a normal process performed to meet Tech Spec requirements to ensurea function can be met and in this COT the Hi containment pressure SI bistable forthis instrument would be placed in the Trip position unless another channel ws tobe tested.

B. Incorrect, 1-PD T-30-43 will not be placed to the position and subsequent testingof 1-PDT-30-44 will not prevent a valid automatic actuation of the containmentspray during the surveillance test because the other 2 channels still provide for theprotection of the function. Plausible because the positioning of bistables to the tripposition is a normal process performed to meet Tech Spec requirements to ensurea function can be met (and in this COT the Hi containment pressure SI bistable forthis instrument would be placed in the Trip position unless testing of anotherchannel was required.

C. Correct, 1-PD T-30-43 would be placed in the bypass position as identified in theTech Spec 3.3.2 Bases. The Required Action for LCO 3.3.2 Condition E has a Notethat allows a channel to be bypassed for up to 12 hours for surveillance testing.This Note is explained in the Tech Spec Bases. There are 4 channels provided forthe Hi-Hi containment function to actuate and it takes 2 of the 4 to generate thesignal (and 2 channels remain in service.

D. Incorrect, 1-PD T-30-43 would be placed in the bypass position but the testing of1-PD T-30-44 will not prevent a valid containment spray actuation from occurringeven though the HI-HI bistables would be tested in the bypass position.

Page 2 of 3


Question Number: 88

Tier: 2 Group 1

K/A: 026 A2.03Containment Spray System (CSS)Ability to (a) predict the impacts of the following malfunctions or operationson the CSS; and (b) based on those predictions, use procedures tocorrect, control, or mitigate the consequences of those malfunctions oroperations: Failure of ESF

Importance Rating: 4.1 /4.4

10 CFR Part 55: 41.5! 43.5/45.3 / 45.13

IOCFR55.43.b: 2,5

KIA Match: Applicant must determine how the bistables will be concurrentlyconfigured on a failed Containment Spray System actuation ESFtransmitter and an ESF transmitter which is required to be tested inorder to run the surveillance instruction and how the function ismaintained as identified in the Technical Specification bases.SRO because the question requires knowledge of Tech Spec basesthat is required to analyze Tech Spec required actions andterminology.

Technical Reference: Tech Spec LCO 3.3.2, ESFASlnstrumentation(Amendment 68)and Bases (Revision 90)

1-47W611-88-1, R23

Proposed references Noneto be provided:

Learning Objective: 3-OT-5Y5072A20. Explain Tech Spec bases for Containment Spray

components and parameters governed by TechSpecs.

Question Source:

New XModified BankBank

Question History: New Question

Comments:

Page 3of3


85. 026AG2.2.25 085/MOD/FNP-NO NRC/C/A 3 .2/4.2/026AG2.2.251N///Unit I is in Mode 3 with the following conditions:

• ‘A’ train is the ON-SERVICE Train.• HV-3096A, CCW TO/FROM EVAP PKGS & H2 RECOMB, is stuck open

and will NOT close.• HV-3096B, CCW TO/FROM EVAP PKGS & H2 RECOMB, is functioning

properly.


HV-3096A & B are required to isolate upon receipt of an (1)

The A Train COW system is (2)

A. 1)Sl signal ONLY

2) OPERABLE; Either HV3096A OR B must close to maintain operability.

B. I) SI signal ONLY

2) NOT OPERABLE; BOTH HV3096A AND B must close to maintain operability.

0. 1) SI signal OR LO LO Surge Tank level

2) OPERABLE; Either HV3096A OR B must close to maintain operability.

D’ I) SI signal OR LO LO Surge Tank level

2) NOT OPERABLE; BOTH HV3096A AND B must close to maintain operability.

Page 100f25


85. 026AG2.2.25 085/MOD/FNP-NO NRC/C/A 3 .2/4.2/026AG2.2.25/N///

The FSD (Functional System Description) documents are manuals that were created and aremaintained by the Engineering Department. They provide a component by componentdescription of most of the major systems for the plant. It lists the equipment required for thesystem to perform its safety function in one section, and non safety equipment in anothersection. It is a document that the SRO can use to help him determine operability of a systemwhen there is a component malfunction, and it is unclear if it affects operability of the system.

The FSD requires that HV3096A & B must both close to isolate the seismic rated piping fromthe non-seismic rated pipTh of theCCW system to prevent a loss of COW due to a pipe breakin the non-seismic portion of the system. HY3096A& B allow COW flow to the Waste GasHydrogen Recombiners, Waste Evaporator, Recycle Evaporator, and Waste Gas compressors.These are all non-safety reted components. HV3096B isolates the inlet line to thesecomponents and HV3096A isolates the outlet from these components. If there were a leak onany of these non-seismic components,both valves would have to close to completely isolatethe leak. This is different from the normal convention in that normally there are two valves inseries that will provide isolation if either one closes.

HV3096A & B are in the COW system as part of the Miscellaneous Header. The train thatsupplies the Miscellaneous Header is the “On-service train”. For this question, ‘A’ Train is theon service train, and since HV3096A can not be closed, then A train must be declaredINOPERABLE until HV3096A is closed.

TS Basis for 3.7.7 states:1. A CCW train is considered OPERABLE when:

a. The pump and associatecLsurge tank section are OPERABLE; andb. The associated piping,alv> heat exchanger, and instrumentation and controlsreqirecto perform the saftrreiated funclion are OPERABLE

2. The isolation of COW from other components or systems not required for safety may renderthose components or systems inoperable but does not affect the OPERABILITY of theCCW System.

These bases comments may seem to be opposing, but are actually talking about two differentthings. The first one (1 .b) is saying that (among other equipment) the safety related isolationvalves have to be operable - HV3096A & B have to be able to close.

The second bases statement is saying that the Isolation of CCW from other systems does notmake the CCW system inoperable, BUT for our question we are looking at a Failure to Isolate.A failure to isolate falls under the first statement and not the second statement. For example,the second statement is saying that if COW is isolated to a Letdown Heat Exchanger, it maymake the Letdown Heat Exchanger inoperable, but it does not make the COW systeminoperable.

The Isolation of components does not make the COW system inoperable, but a Failure toIsolate will make the COW system inoperable.

A181000, COW FSD 3.13.1 Basic FunctionsThe nonessential user isolation valves shall provide positive automatic isolation ofthe hydrogen recombiners, waste gas compressors, waste evaporator package, and recycleevaporator package on low-low surge tank level, safety injection actuation signal (SIAS),

Page 13 of 30


section 3.13.2.3 continues to state:The valves shall fail closed to isolate the Seismic Category I portion of the CCWSfrom the non-seismic portion (Reference 6.7.05).

FSAR 9.2.2 states that All portions of the component cooling system that are safety relatedare Seismic Category I design. Valves in the supply and return lines for non-safety-relatedequipment will be automatically closed.

A. Incorrect 1) Incorrect, an SI signal will close the valves, but since the area isolatedby the HV3096 valves is non-seismic, the HV3096 valves will also goclosed on a LO LO Surge tank level.2) Incorrect, OPERABLE is incorrect because the seismic portion can notbe isolated from the non-seismic portion of the system, this “SAFETYFUNCTION” is not satisfied, therefore the system is INOPERABLE untilHV3096A is closed. Plausible because normally there are two valves inseries and only one is required to close to provide isolation.

B. Incorrect 1)lncorrect,seeA.1.2) Correct, the A train is not operable since it is the “on-service” train andHV3096A is failed.

C. Incorrect 1) Correct, HV3096 A & B will isolate on either an SI or LO LO Surge tanklevel.2) Incorrect, see A.2.

D. Correct 1) Correct, see C.1.2) Correct, see B.2.

Page 14 of 30


85. 026AG2.2.25 085/MOD/FNP-NO NRC/C/A 3.2/4.2/026AG2.2.25/N///026AG2.2.25Loss of Component Cooling Water (COW)

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions foroperations and safety limits.(CFR: 41 .5/41.7/43.2)IMPORTANCE RO 3.2 SRO 4.2


Technical Reference: Tech Spec Bases, v53A181000, Functional System Description v24FSAR 9.2.2 v24


Learning Objective: RECALL AND APPLY the information from the LCD BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the CCW System components andattendant equipment alignment, to include the following(OPS-62102G01): 1OCFR55.43 (b) 23.7.7 Component Cooling Water System

Question origin: Modified FNP Bank CCW-62102G02 01

Basis for meeting K/A: This question requires a knowledge of TS Bases as itrelates to a piping break in a non-Seismic portion of theCOW system. A piping break, with a failure of the isolationvalves to close would cause a loss of the COW system, anda resultant loss of support of its associated ESF equipment.This question requires assessing system operability (LOSSTHEREOF); Systems level knowledge is required regardingthe signal which isolates this non-seismic, non-essentialportion of the system as well as SRO level knowledge of theTS Basis to assess operability.

SRO justification: 10 CFR 55.43(b)(2):Knowledge of TS bases that is required to analyze TSrequired actions and terminology.


1) can NOT be answered by knowing less than 1 hour Tech Specs.

Page 19 of 51


2) can NOT be answered by knowing information listed “above-the-lin&’.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:• Knowledge of TS bases that is required to analyze TS required actions and

terminology. (where amplifying instruction is found within the PROCEDURES.)

Page200f5l

026AA2.03Thursday, March 01, 2012 Page 1 of 1

1. CCW-62 1 02G02 001 /HLT!LOCT/SRO/C/A 2.6/2.9/APEO26AA2.03/3 .7.7//!Unit 1 is in Mode 3, and A train is in service. The following conditions are discovered:

• HV-3096A, CCW from the Waste and Recycle Evaporator Packages and WasteGas System Components, is stuck open and will not close.

• HV-3096B, CCW to the Waste and Recycle Evaporator Packages and WasteGas System Components, operates properly.

• No other valve manipulations are made.

Which one of the following describes required operation of HV-3096A and states theoperability of A train CCW?

HV-3096A is required to close after a (1) signal.

The A train CCW system is (2)

(1) (2)

A. Phase A Operable; Either HV3096A OR B must close tomaintain operability.

B. Phase A NOT operable; BOTH HV3096A AND B must close tomaintain operability.

C. Safety Injection Operable; Either HV3096A OR B must close tomaintain operability.

D” Safety Injection NOT operable; BOTH HV3096A AND B must close tomaintain operability.

Page 1 of 1


86. 034K4.01 086/MOD/ROBINSON 2007/C/A 2.6/3.4/034K4.01/N///Unit 1 is in a refueling outage with the following conditions:

• Due to a problem with the Gripper Engaged Limit Switch on the ManipulatorCrane, TS-2, HOIST INTERLOCKS BY-PASS switch, has been taken to ON.

• A fuel assembly is being lowered into the Reactor vessel.• The Manipulator Crane operator releases the HOIST JOG lever and the bottom

of the fuel assembly is 2” from the lower core plate.


Per FHP-7.O, Limitations and Precautions for Handling Fuel Assemblies, the(1) is REQUIRED to give permission to place the HOIST INTERLOCKS

BY-PASS switch to ON.

When the GRIPPER switch is 1aken to the DISENG position, the gripper (2)disengage from the fuel assembly.

(1) (2)

A. Fuel Handling Supervisor WILL

B” Fuel Handling Supervisor will NOT

C. Shift Manager WILL

D. Shift Manager will NOT

Page 11 of 25


86. 034K4.0 1 086/MOD/ROBINSON 2007/C/A 2.6/3 .4/034K4.0 1/N//I

When transferring fuel, for the gripper to unlatch, two things have to happen.I )The Slack Cable interlock has to be made up to indicate that the fuel assembly is fulldown. This enables an air supply solenoid to operate to supply air to the gripper todisengage it.2) About 300 pounds of mechanical force has to be applied to the top of the fuelassembly to disengage a mechanical interlock on the gripper.

For the Slack Cable interlock to function properly, the Underload interlock isautomatically disabled in the last two inches of travel when lowering a fuel assembly.This lends plausibility to the gripper operating when the fuel assembly is two inchesfrom the lower core plate. In addition, the Hoist Interlocks By-Pass bypasses theunderload circuit. The underload circijtwW stop hoist downwthnacJjJjflanunderload is sensed.tThE candidate may thikJhf cetfieUnderload interlock is - V

inches of travel or by the Hoist Interlocks By-Pass, the fuelassembly can be unlatchedJ

A. Incorrect 1) Correct, per FNP-0-FHP-7.0 - “No electrical bypass switch operation ofthe manipulator crane shall be allowed without the permission of the FuelHandling Supervisor”.2) Incorrect, the gripper will not disengage until the fuel assembly is fullyseated, due to the mechanical latch that prevents releasing a fuelassembly until sufficient downward force is exerted on the latch by theManipulator Crane mast.

B. Correct I) Correct, Fuel Handling Supervisor permission must be obtained prior togoing to bypass in every situation.2) Correct, the Manipulator Crane gripper will not disengage, see A.2.

C. Incorrect 1) Incorrect, Shift Manager may give permission, but Fuel HandlingSupervisor permission is still required with or without Shift Managerapproval.2) Incorrect, see A.2.

D. Incorrect I) Incorrect. See C.I.2) Correct. See B.2.

Page 15of30


86. 034K4M 1 086/MOD/ROBINSON 2007/C/A 2.6/3.4/034K4.0 1 IN//I034K4.01Fuel Handling Equipment System (FHES)

Knowledge of design feature(s) and/or interlock(s) which provide for the following:(CFR: 41.7)K4.O1 Fuel protection from binding and dropping 2.6 3.4


Technical Reference: FNP-0-FHP-7.0, v23FNP-1-FHP-5.13, v22


Learning Objective: RECALL AND DISCUSS the Precautions and Limitations(P&L), Notes and Cautions (applicable to the “ReactorOperator”) found in the following procedures(OPS-521 08D05):FHP-1 .0 Refueling OperationsFHP-3.0 Receipt and Storage of New FuelFHP-4.0 Transfer of New Fuel to Spent Fuel PitFHP-7.0 Limitations And Precautions For Handling FuelAssemblies

Question origin: Modified from Robinson 2007 NRC Exam Question #97

Basis for meeting K/A: This question requires knowledge of the mechanicalinterlock on the Refueling Manipulator Crane that willprevent dropping a fuel assembly when the gripper islatched and the fuel assembly is suspended. The interlockwill prevent dropping an assembly due to a loss of air, lossof power, or inadvertent operation of the in correcthandswitch.

SRO justification: 10 CFR 55.43(b)(7):Refuel floor SRO responsibilities.

2012 NRC exam

Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)]

Some examples of SRO exam items for this topic include:• Refuel floor SRO responsibilities.• Assessment of fuel handling equipment surveillance requirementacceptance criteria.• Prerequisites for vessel disassembly and reassembly.

Page 21 of 51


• Decay heat assessment.• Assessment of surveillance requirements for the refueling mode.• Reporting requirements.• Emergency classifications.

Page 22 of 51

HLC-06 NRC Replacement Exam97.

Who, by title, has approval authority for bypassing interlocks on Refueling equipment inaccordance with OMM-OO1 -1 8, Outages?

A. Operations Outage Coordinator (OOC)

B. Operations Shift Manager (OSM)

C. Superintendent - Shift Operations (SSO)

D. Refueling SRO

97


87. 056AG2.1.7 087/BANK/WATTS BAR 2009/M 4.4/4.7/056AG2.1.7/N///

The foNowing conditions exist on Unit 1:

• ECP-0.0, Loss of AN AC Power, is in progress.• Depressurization of ALL intact Steam Generators has been initiated.


The Steam Generator depressurization is required to be stopped if (1)

When transitioning from ECP-0.0, a parameter that will be monitored todetermine the correct recovery procedure is (2)

A. 1) Pressurizer level falls to 12%, to prevent Reactor Vessel Head voiding

2) RCS Pressure

B. 1) all intact SG narrow range levels fall to 30%, to prevent a loss of adequateheat transfer capability

2) RCS Pressure

C. 1) Pressurizer level falls to 12%, to prevent Reactor Vessel Head voiding

2) RCS Subcooling

D” 1) all intact SG narrow range levels fall to 30%, to prevent a loss of adequateheat transfer capability

2) RCS Subcooling

Page 12 of 25


87. 056AG2. 1.7 087/BANK/WATTS BAR 2009/M 4.414.71056AG2. 1 .7/N//I

A. Incorrect 1) Incorrect, Plausible because Pressurizer level is monitored frequentlyin the ERGs and alternate actions are taken as a result of Pressurizerlevel getting too low (see step 31 of ESP-O.O). Preventing Reactor VesselHead voiding is a concern, especially in the ECP-O series procedures.There is a note prior to step 17 of ECP-O.O, that discusses Pressurizerlevel and vessel upper head voiding.2) Incorrect, Plausible because RCS pressure is monitored frequently inthe ERGs and alternate actions are taken if RCS pressure is not keptwithin limits.

B. Incorrect 1) Correct, per ECP-O.O Bases. Natural circulation may be interrupted ifSG levels are not maintained high enough to keep the SG tubes fullycovered.2) Incorrect, see A.2.

C. Incorrect 1) Incorrect, seeA.1.2) Correct, per ECP-O.O. RCS subcooling provides transition criteria atstep 31 of ECP-O.O.

D. Correct 1) Correct, see B.1.2) Correct, see C.2.

Page 16 of 30

25 SRO QUESTIONS Ver5Wednesday, February 29, 2012

87. 056AG2. 1.7 O87JBANKJWATTS BAR 20091M 4.4/4.7!056AG2. 1 .7/NI/I

56AG2.1 .7Loss of Offsite Power

Page 23 of 51

2.1.7 Ability to evaluate plant performance and make operational judgments based onoperating characteristics, reactor behavior, and instrument interpretation.(CFR: 41.5/43.5/45.12/45.13)IMPORTANCE RO 4.4 SRO 4.7

Importance Rating: 4.4

Technical Reference: ECP-0.0, v25


Learning Objective:

Question origin:

Basis for meeting K/A:

SRO justification:

ASSESS the facility conditions associated with the (1)ECP-0.0, Loss of All AC Power; (2) ECP-0.1, Loss of All ACPower Recovery, Without SI Required; (3) ECP-0.2, Loss ofAll AC Power Recovery, With SI Required, and based onthat assessment: (OPS-62532A01)• SELECT the appropriate procedures during normal,abnormal and emergency situations. 10CFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required• DETERMINE if the critical safety functions are satisfied

Same as Watts Bar December 2009 NRC Exam, Question#99

A Loss of Offsite Power has occurred and the candidate hasto evaluate conditions and determine (by instrumentinterpretation) when stopping the SG depressurization wouldbe required. In addition, the candidate has to has to be ableto evaluate instrumentation required to transition to theapppriaterecoveryprocedure.

10 CFR 55.43(b)(5)Assessing plant conditions (normal, abnormal, oremergency) and then selecting a procedure or section of aprocedure to mitigate, recover, or with which to proceed.

The applicant’s knowledge can be evaluated at the level of10 CFR55.43(b)(5) by ensuring that the additionalknowledge of the procedure’s content is required to correctlyanswer the written test item, for example:• Knowledge of diagnostic steps and decision points in theemergency operating procedures (EOP) that involvetransitions to event specific subprocedures or emergency

4.7

Page 23 of 51


contingency procedures.

2012 NRC exam

From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart for 10 CFR 55.43(b)(5):Assessment of facility conditions and selection of appropriate procedures duringnormal, abnormal, and emergency situations [10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then2) selecting a procedure or section of a procedure to mitigate, rcover, or with




overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:— Knowledge of diagnostic steps and decision points in the emergency operating

procedures (EOP) that involve transitions to event specific subprocedures oremergency contingency procedures.

Page 24 of 51

11/2009 Watts Bar SRO NRC Exam - As submitted10/2/2009

99. G 2.4.18 099

Given the following Unit I plant conditions:

- ECA-0.0, Loss of All AC Power is in effect.- Depressurization of all intact SGs at the ‘maximum rate’ is in progress.

Which ONE of the following identifies...(1) the parameter used to determine the maximum rate and its basis

and(2) a parameter used to determine which recovery procedure will be

implemented when ECA-0.0 is completed?

A. (1) Ability to maintain pressurizer level above 10% to prevent loss of pressurecontrol resulting in reactor vessel head voiding.

(2) RCS pressure trend

B. (1) Ability to maintain pressurizer level above 10% to prevent loss of pressurecontrol resulting in reactor vessel head voiding.

(2) RCS Subcooling value

C. (1) Ability to maintain at least one SG narrow range level above 29% to preventloss of sufficient heat transfer capability from the RCS.

(2) RCS pressure trend

D’ (1) Ability to maintain at least one SG narrow range level above 29% to preventloss of sufficient heat transfer capability from the RCS.

(2) RCS Subcooling value

Page 71


88. 067AA2.08 088/NEW//C/A 2.9/3.6/067AA2.08/N/EMERG CLASS!!

Unit 1 is at 100% power with the following conditions:

The 1A Control Room Air Conditioning (CRACS) system is Tagged Out.

At 10:00:

• The Rover reports a fire in the lB CRACS system.

At 10:14:

• The fire is extinguished.• The 1 B CRACS system is damaged and cannot be restarted.


An emergency classification declaration (1) required per NMP-EP-1 10,Emergency Classification Determination and Initial Action.

The FSAR design limit maximum temperature for the Control Roomequipment is (2)

(1) (2)

A. IS 104°F

B. is NOT 104°F

C’ IS 120°F

D. is NOT 120°F

Page 13 of 25


88. 067AA2.08 088/NEW//C/A 2.9/3 .6/067AA2.08/N/EMERG CLASS/I

Per NMP-EP-1 10 - the criteria for ALERT is:HA2 - FIRE OR EXPLOSION Affecting the Operability of Plant Safety SystemsRequired to Establish or Maintain Safe Shutdown (pg. 69)1. FIRE OR EXPLOSION AND affected system parameter indications show degradedperformance OR plant personnel report VISIBLE DAMAGE to permanent structures ORsafety related equipment in any of the following VITAL AREAs:Containment I Auxiliary Building I Control RoomStorage Pond dike and dam - Service Water PondCondensate Storage Tank (CST)Service Water Intake Structure (SWIS)Refueling Water Storage Tank (RWST)Diesel Generator Building / Pond Spillway Structure

The criteria for a NOUE is:HU2 - FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutesof Detection (pg. 77).1. FIRE in buildings OR areas contiguous to any of the following areas NOTextinguished within 15 minutes of control room notification OR verification of a controlroom alarm unless disproved by personnel observation within 15 minutes of the alarm:Containment I Auxiliary BuildingCondensate Storage Tank (CST)Service Water Intake Structure (SWIS)Refueling Water Storage Tank (RWST)Diesel Generator Building Control Room

Per FSAR 9.4.1.1 -

Two separate and redundant air conditioning systems are provided to maintain thetemperature in the control room at approximately 78°F (db). Safety-related componentsin the control room are designed to withstand a maximum environmental temperatureof 120°F.

A. Incorrect 1) Correct, An emergency classification declaration is required byNMP-EP-1 10. An ALERT is required for a fire in a vital area that causesdamage to safety related equipment.2) Incorrect, Plausible because STP-63.3, EQ Area TemperatureMonitoring, acceptance criteria is for area temperatures to be < 104°F.

B. Incorrect 1) Incorrect, plausible because for a fire that has been extinguished within15 minutes, with no equipment damage, no classification is required.2) Incorrect, see A.2.

C. Correct 1) Correct, see A.1.2) Correct, per the FSAR, 120°F is the design limit for equipmentoperability in the Control Room.

D. Incorrect 1) Incorrect, see B.1.

Page 17 of 30


2) Correct, see 0.2.

Page 18of30


88. 067AA2.08 088/NEW//C/A 2.9/3 .6/067AA2.08/N/EMERG CLASS/I

067AA2.08Plant Fire On Site

Ability to determine and interpret the following as they applyto the Plant Fire on Site:(CFR: 43.5 / 45.13)AA2.08 Limits of affected area 2.9 3.6


Technical Reference: NMP-EP-110-GLOI, v2FSAR, v24


Learning Objective: Using plant procedures/references, ANALYZE a set of plantconditions and DETERMINE the proper classification of theemergency condition as being a NOUE, Alert, Site Area, orGeneral Emergency. (OPS-63002C01).

Using plant procedures/references, DETERMINE theappropriate actions that are to be performed by the SM/EDduring a NOUE, Alert, Site Area, or General Emergencyincluding developing PARs, and the consequences ofinadequate actions. (OPS-63002C02).


Basis for meeting K/A: A fire has occurred in the Auxiliary Building. Candidate hasto display ability to determine and interpret the limits forapplication of the Emergency Plan (is an emergencyclassification required). In addition, candidate has to haveknowledge of the Control Room temperature limit associatedwith the loss of Control Room HVAC, an area affected bythe fire.

SRO justification: III. Justification for Plant Specific ExemptionsUNIQUE to the SRO position:

Justification: A question that is not tied to one of the10 CFR 55.43(b) items can still be classified as“SRO-only” provided the licensee has documentedevidence to prove that the knowledge/ability is “unique tothe SRO position” at the site. An example of documentedevidence includes:• The question is linked to a learning objective that isspecifically labeled in the lesson plan as being SRO-only

Page 25 of 51


(e.g., some licensee lesson plans have columns in themargin that differentiate AC, RO, and SRO learningobjectives) [NUREG 1021, ES-401, Section D.2.d}AND/OR• A question is linked to a task that is labeled as anSRO-only task, and the task is NOT listed in the RO tasklist.

The SRO is solely responsible for determiningClassifications at FNP, both objectives listed above areSRO only objectives.

2012 NRC exam

HI. Justification for Plant Specific Exemptions UNIQUE to the SRO position:

DOES NOT MATCH one of the 10 CFR 55.43(b) items but FNP has classified theknowledge/ability as “unique to the SRO position” as documented within SAT processas ties the knowledge/ability to the licensee’s SRO job position duties.

Page 26 of 51


89. 07202.2.36 089/MOD/FARLEY 2011/M 3.1 /4 .2/07202.2 .36/N//I

Maintenance is preparing to de-energize R-5, SEP RM, for troubleshooting andrepair on Unit 1.

At 10:00:

• An electrical lead is lifted that disables the alarm function of R-5.

At 10:30:

• The power supply to R-5 is de-energized and Tagged Out.


Per TRM 13.3.4, Radiation Monitoring Instrumentation, R-5 becomesnonfunctional at (1)

De-energizing R-5 (2) cause the Penetration Room Filtration Systemto automatically start.

(1) (2)

A. 10:00 WILL

B 10:00 will NOT

C. 10:30 WILL

D. 10:30 will NOT

Page 14 of 25


89. 072G2.2.36 089/MOD/FARLEY 2011 /M 3. l/4.2/072G2.2.36/N///

Bases of TR 13.3.4 Radiation Monitoring InstrumentationThe FUNCTIONALITY of the radiation monitoring channels ensures that:

1) the radiation levels are continually measured in the areas served by the individualchannelsand2) the alarm is initiated when the radiation level trip setpoint is exceeded.

A. Incorrect 1) Correct, the alarm function is required for R-5 to be FUNCTIONAL perTRM 13.3.4.2) Incorrect, PRF will j auto start on a High Rad signal on R-5, SFProom monitor, but plausible because it will autostart on R-25, SPENTFUEL BLDG EXH. Both monitors are providing radiation level informationof the Spent Fuel Pool area.

B. Correct 1) Correct, see A.1.2) Correct, PRF will not auto start on a High Rad signal on R-5, SFP roommonitor.

C. Incorrect 1) Incorrect, the alarm function is required for R-5 to be FUNCTIONAL perTRM 13.3.4, so the rad monitor was non FUNCTIONAL at 10:00.2) Incorrect, see A.2

D. Incorrect 1) Incorrect, see C.1.2) Correct, see B.2.

Page 19 of 30


89. 072G2.2.36 089/MOD/FARLEY 201 1/M 3.l/4.2/072G2.2.36/N///

072G2.2.36Area Radiation Monitoring (ARM) System

2.2.36 Ability to analyze the effect of maintenance activities, such as degraded powersources, on the status of limiting conditions for operations.(CFR: 41.10 /43.2 /45.13)IMPORTANCE RO 3.1 SRO 4.2


Technical Reference: Technical Requirements Manual, v24Technical Requirements Manual Bases, v9


Learning Objective: RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with Radiation Monitoring System componentsand attendant equipment alignment, to include the following(OPS-62106D01): IOCFR55.43 (b) 23.3.3, Post Accident Monitoring (PAM) Instrumentation3.3.6, Containment Purge and Exhaust IsolationInstrumentation3.3.7, Control Room Emergency Filtration/PressurizationSystem (CREFS) Actuation Instrumentation3.3.8, Penetration Room Filtration (PRF) System ActuationInstrumentation3.4.15, RCS Leakage Detection Instrumentation13.3.4, Radiation Monitoring Instrumentation

Question origin: Modified from Farley 2011 NRC Exam Question #86

Basis for meeting K/A: This question tests the ability of the candidate to evaluatethe effects of Maintenance personnel activities on an AreaRad Monitor. Maintenance sometimes lifts leads to disablethe alarm function to prevent it being a nuisance while workis in progress in an area. Using TRM bases information,candidate has to determine the FUNCTIONALITY of the radmonitor per TRM 13.3.4, at different points in themaintenance process.

SRO justification: 10 CFR 55.43(b)(2):Knowledge of TS bases that is required to analyze TSrequired actions and terminology.

Page 27 of 51


Knowledge of TRM Bases is required to evaluatefunctionality of R-5.

2012 NRC examFrom the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart for 10 CFR 55.43(b)(2)


terminology.

Page28of5l

2011 HLT-34 QuestionThursday, March01, 2012 Page 1 of 1

1. 061 AA2.04 086/NEW/SRO/MEM 3.5/4.2/061 AA2.04/N/3/HF REVIEW-WHY/UNSAT


The alarm function(s) of R-5, SEP RM, (1) required for FUNCTIONALITY per TR

13.3.4, Radiation Monitoring Instrumentation.

R-5 (2) automatically trip the Fuel Handling Area Supply and Exhaust Fans on a

HIGH radiation condition.

(1) (2)

A. IsNOT willNOT

B. is NOT WILL

C’ IS will NOT

D. IS WILL

Page 1 of I


90. 086A2.02 090/MOD/FAR 2010/VOGT 2011/C/A 3.0/3.3/086A2.02/N///

FNP-0-FSP-203.1, Motor Driven Fire Pump Functional Test, has been completed withthe following conditions:

• The Motor Driven Fire pump (MDFP) started automatically at its required lowFire Protection System header setpoint.

• The MDFP could NOT maintain header pressure as required by FSP-203.1.• The #1 Diesel Driven Fire Pump is Tagged Out.


The MDFP low pressure auto start setpoint is (1) psig.

The Fire Suppression Water System (2) OPERABLE per FSAR section 9B.

(1) (2)

A. 70 is NOT

90 is NOT

C. 70 IS

D. 90 IS

Page 15of25


90. 086A2.02 090/MOD/FAR 2010/VOGT 2011/C/A 3.0/3.3/086A2.02/N///

FSP-203.1 is a test to verify operability of the Motor Driven Fire Pump. The FSP checksthat the pump autostarts at 90 psig and maintains Fire Protection Header pressure > 70psig. Normally, there are two back up Diesel Driven Fire Pumps that autostart at 80psig and 70 psig respectively In this situation the #1 DDFP is Tagged Out Per theFSAR, section 9B, if any two of the three Fire Pumps are operable, the FireSuppression Water System is operable If less than two pumps are operable, the FireSuppression Water System is inoperable, and actions are required.

A. Incorrect 1) Incorrect, the MDFP autostart setpoint is 90 psig. FPS header pressureis required to be maintained > 70 psig.2) Correct, the Fire Water Suppression system is inoperable with twoinoperable Fire Pumps.

B. Correct 1) Correct, autostart setpoint is 90 psig.2) Correct, see A.2.

C. Incorrect 1) Incorrect, see A.1.2) Incorrect, the Fire Water Suppression system is inoperable with twoinoperable Fire Pumps.

D. Incorrect 1) Correct, see B.1.2) Incorrect, see 0.2.

Page 20 of 30


90. 086A2.02 090/MOD/FAR 2O10IVOGT 2011/C/A 3.0/3.3/086A2.02/N///086A2.02Fire Protection System (FPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the FireProtection System; and (b) based on those predictions, use procedures to correct,control, or mitigate the consequences of those malfunctions or operations:(CFR: 41.5 /43.5/45.3 /45.13)A2.02 Low FPS header pressure 3.0 3.3


Technical Reference: FNP-0-FSP-203.1 vSFSAR 9B.C-2, ver 24FNP-0-SOP-0.4 v83.2


Learning Objective: OPS-62108K01Recall and Apply the information from the [...] T.S.-FireProtection Program: Renewed License No.NPF-2-Amendment No.175- Fire Protection Program asdescribed in FSAR which implements fire protectionrequirements of 100FR5O.48, 1OCFR5O, and Appendix R.

Question origin: Modified from Vogtle 2011 NRC Exam, question #93 andFarley 2010 NRC Exam #90

Basis for meeting K/A: The K/A calls for the ability to predict the impacts of the LowFPS header pressure. In this question, the low FPS headerpressure causes this pump to be inoperable per the FSPand the candidate has to evaluate the remaining pumpsavailable and determine if the Fire Suppression WaterSystem is operable per the FSAR.

SRO justification: Conditions and limitations in the facility license.10 CFR 55.43(b)(1)

Administration of fire protection program requirementssuch as compensatory actions associated with inoperablesprinkler systems, fire doors, etc.

Requires knowledge of limitations within the Facility License(FSAR 9B) required to analyze OPERABILITY of the FireSuppression Water system. In addition, SOP-0.4, FireProtection Program Administration Procedure, is used toprovide guidance on LCO requirements for a Fire ProtectionLCO as it applies to the FSAR section 9B.

Page 29 of 51


2012 NRC exam

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)}Some examples of SRO exam items for this topic include:

• Administration of fire protection program requirements such ascompensatory actions associated with inoperable sprinkler systems, firedoors, etc.

Page3Oof5l


91. G2.1.8 091/BANK/WATTS BAR 20101M 3.4/4.1/G2.1.8/N///

Unit I has tripped from 100% power with the following conditions:

• FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress.• A Site Area Emergency has been declared by the Emergency Director.• TheJScJs fully staffed and has been declared OPERATIONAL.• A radioactive release is in progress.• ARe-entry is planned to reset the TDAFW trip/throttle valve.

In accordance with EIP-14.0, Personnel Movement, Relocation, Re-entry, and SiteEvacuation, which one of the following identifies the actions required to coordinate anddispatch a team to reset the TDAFW trip/throttle valve?

A. A team, will be dispatched from the assembly area in the Main Control Room.The TSC/OSC will be notified that the team has been dispatched and theteam’s intended location and action.

B. A team will be dispatched from the assembly area in the Main Control Room.The TSC/OSC notification is NOT required because the Re-entry is beingtracked by the Shift Manager.

C’ After being notified of the task to be performed, the TSC/OSC staff willassign a team to perform the task. The team will be briefed, dispatched, andtracked by the TSC/OSC.

D. After being notified of the task to be performed, the TSC/OSC staff will dispatch ateam to the Main Control Room to be briefed prior to being sent to perform the task.The team will be tracked by the TSC/OSC.

Page 16 of 25


91. G2.l.8 091/BANK/WATTS BAR 2010/M 3.4/4.1/G2.1 .8/N//I

A. Incorrect Plausible because the Shift Manager can dispatch someone from theControl Room for a Relocation after the TSC is staffed, but not aRe-entry. This answer could be correct for a Relocation.

B. Incorrect Plausible because the Shift Manager can dispatch someone from theControl Room for a Relocation after the TSC is staffed, but not aRe-entry. This answer could be correct for a Relocation.

C. Correct Per the guidance of FNP-O-EIP-14.O, the Re-entry team will beestablished, briefed, approved, dispatched, and tracked by the OSC/TSC.The ED can give verbal approval for the Re-entry and all other actions areperformed by the OSC Mgr./ HP Supervisor I Maintenance Supervisor.

D. Incorrect Per the guidance of FNP-O-EIP-14.O, the Re-entry team will beestablished, briefed, approved, dispatched, and tracked by the OSCITSC.The individual does not have to report to the Control Room to be briefed.The ED can give verbal approval for the Re-entry and all other actions areperformed by the OSC Mgr.I HP Supervisor I Maintenance Supervisor.

Page 21 of 30


91. G2.l.8 091/BANK/WATTS BAR 2010/M 3.4/4.1/G2.1 .8/N/I!

G2.1.8

Ability to coordinate personnel activities outside the control room.(CFR: 41.10/45.5/45.12/45.13)IMPORTANCE RO 3.4 SRO 4.1


Technical Reference: FNP-0-EIP-14.0, v26


Learning Objective: Given a set of plant conditions, EVALUATE those conditionsand DETERMINE the emergency responses actions theSM/ED should take to mitigate the consequences of theevent. (OPS-63002A01).

Given an emergency scenario, EVALUATE plant conditionsand DETERMINE if a reentry, relocation or movement isrequired. EVALUATE the conditions and DETERMINE doselimits for personnel involved. (OPS-63002D01).

Question origin: Same as Watts Bar 2010 NRC Exam question #95

Basis for meeting KJA: Ability to coordinate personnel activities outside the controlroom is displayed by knowledge of EIP-14.0 and therequirements for performing a Re-Entry after the TSC is fullystaffed and OPERATIONAL.

SRO justification: Ill. Justification for Plant Specific ExemptionsUNIQUE to the SRO position:

Justification: A question that is not tied to one of the 10 CFR55.43(b) items can still be classified as “SRO-only” providedthe licensee has documented evidence to prove that theknowledge/ability is “unique to the SRO position” at the site.An example of documented evidence includes:• The question is linked to a learning objective that isspecifically labeled in the lesson plan as being SRO-only(e.g., some licensee lesson plans have columns in themargin that differentiate AO, RO, and SRO learningobjectives) [NUREG 1021, ES-401, Section D.2.d}AND/OR• A question is linked to a task that is labeled as anSRO-only task, and the task is NOT listed in the RO tasklist.

Page 31 of 51


The SRO is solely responsible for direction andimplementation of the Emergency Plan at FNP, bothobjectives listed above are SRO only objectives.

2012 NRC exam

Ill. Justification for Plant Specific Exemrtions UNIQUE to the SRO position:


Page32ofSl

Laska Question Watts Bar 2010Thursday, March 01,2012 Page 1 of 1

I. G 2.1.8 195/NEW//LOWER!/SRO/WATTS BAR!08!20 10/NO

Given the following:

- An ALERT has been declared on Unit 1.- All emergency centers are activated.- Conditions require AUO action to manually isolate 1-ISV-70-700, RCP OIL

COOLER CCS RETURN ISOLATION [A4N EL. 710 U-I Penetrationroom] in accordance with E-0, “Reactor Trip Safety Injection,” AttachmentB3.

In accordance with EPIP-7, “Activation and Operation of the Operations SupportCenter (OSC),” which ONE of the following identifies the actions to coordinateand dispatch a team to isolate the valve?

A. OSC Team A stationed in the MCR will be sent to perform the task. TheTSC/OSC will be notified that the team has been dispatched and the team’sintended location and action.

B. OSC Team A stationed in the MCR will be sent to perform the task.TSC/OSC notification is not required because the OSC Team A is beingtracked as assigned to the MCR.

C” After being notified of the task to be performed, the OSC staff will assign ateam to perform the task. The team will be briefed, dispatched, and trackedby the OSC.

D. After being notified of the task to be performed, the OSC staff will dispatch ateam to the MCR to be briefed prior to being sent to perform the task. Theteam will be tracked by the OSC.

Page 1 of 1


92. G2.2. 1 O92JBANKJFARLEY 2008/C/A 4.5/4.4/G2.2. I /N///

Unit 1 is in MODE 3 preparing to do an initial startup after a core reload lAW UOP-1 .2,Startup of Unit from Hot Standby to Minimum Load.

The following conditions exist:

• The current RCS boron sample is 1540 ppm.• Both Shutdown banks have been withdrawn.• All RCPs are running and the RCS is at normal operating pressure and

temperature.• An Infrequently Performed Test or Evolution (IPTE) brief will be performed for

the startup.• FT-I 68, TOTAL MAKEUP FLOW TO CHG/VCT, has FAILED LOW.

The crew is at the step in UOP-I .2 to dilute the RCS to the Hot Shutdown boronconcentration.

(1) Which one of the following is the correct method used for diluting per SOP-2.3,

Chemical and Volume Control System Reactor Makeup Control System, and

(2) Which one of the following is the correct person to conduct the IPTE brief lAW

NM P-AD-006, Infrequently Performed Tests and Evolutions?

A. 1) Perform a Manual Makeup to the charging pump suction.Verify the dilution AUTOMATICALLY stops when batch integrator setpointis reached.

2) Shift Supervisor

B. 1) Perform a Manual Makeup to the charging pump suction.Verify the dilution AUTOMATICALLY stops when batch integrator setpointis reached.

2) Engineering Director

C. 1) Verify a RMW pump runningOpen FCVI 14B, RMW to BlenderOpen FCVI I3B, MKUP TO CHG PUMP SUCTIONManually secure dilution when complete

2) Shift Supervisor

D” (1)Verifya RMW pump runningOpen FCV1 14B, RMW to BlenderOpen FCV1 13B, MKUP TO CHG PUMP SUCTIONManually secure dilution when complete

(2) Engineering Director

Page 17of25


92. G2.2. I 092/BANK/FARLEY 2008/C/A 4.5/4.4/G2.2. 1 /N///

A. Incorrect 1) Incorrect, Plausible because the MANUAL method of Makeup is usedfor a variety of boration/dilution/blended flow evolutions. In the situationgiven, FT-I 68 which is inoperable, would not provide the actual flow tothe “Flow Indicating Switch” FIS-168. FIS-168 would thus not indicatetotal gallons properly or automatically secure the dilution as required bythis method.2) Incorrect, Plausible, since the SS normally conducts the pre-job briefs.The SS is required to conduct a PJB after the IPTE brief to fill in thecoordination details and additional instructions lAW NMP-AD-006.

Also the Affected Shift Supervisor signs Attachment 1 of NMP-006 givingpermission to start the test which ensures that the signatures above forthe Senior line manager brief and the test/coordinator brief are correctand conducted lAW with the NMP and are the proper people.

B. Incorrect 1) Incorrect, see A.I2) Correct, A Senior Line Manager is required to conduct the IPTE brief.The personnel listed in step 4.2 of NMP-AD-006, qualify:4.2 Senior Line Manager - The functions of the Senior Line Manager willbe performed by one of the following:• Site Vice President• Plant Manager• Engineering Director• Site Support Manager• Department head / line manager - preferred• Operations Superintendent• Maintenance Assistant Manager

C. Incorrect 1) Correct, SOP-2.3, App. D, “Operation of the Chemical and VolumeControl System Reactor Makeup Control System with FIS 168 Total FlowBatch Integrator Unreliable” provides procedure guidance for thissituation. It is unique, since all other dilution evolutions utilize the FIS 168flow integrator. The applicant must realize that if the Flow transmitter isout of commission, so the Flow Indicating Switch will not work either.2) Incorrect, see A.2.

D. Correct 1) Correct, see Ci.2) Correct, see B.2.

Page 22 of 30


92. G2.2. 1 092/BANK!FARLEY 2008/C/A 4.5/4.4/G2.2. 1/N/I!

G2.2.1

Ability tofo -strt cedu for the facility, including operating thosecontrols associated with plant equipment that could affect reactivity.(CFR:41.5141.10143.5143.6145.l)IMPORTANCE RD 4.5 SRO 4.4


Technical Reference: FNP-1-SOP-2.3, v58FNP-1-UOP-1.2, V102NMP-AD--006, VI 1


Learning Objective: Define an infrequently performed test or evolution(0PS40502C09).

RELATE AND IDENTIFY the operational characteristicsincluding design features, capacities and protectiveinterlocks for the components associated with the ReactorMakeup Control and Chemical Addition System, to includethe following (OPS-40301 G02):Boric Acid Transfer PumpsReactor Makeup Water PumpsMakeup to Charging Pump Suction Header, FCV-l I 3BMakeup to VCT, FCV-1 14AReactor Makeup Water to Boric Acid Blender, FCV-1 14BBoric Acid to Blender, FCV-1 I 3AEmergency Borate Valve, MOV-8104Manual Emergency Boration valve, V-8439Chemical Addition SystemChemical Mixing Tank and OrificeBoric Acid BlenderReactor Makeup Water Storage TankReactor Makeup Water Storage TankBoric Acid Storage TanksInter connections with other systemsEmergency Boration flow pathsMakeup Mode Selector and Control SwitchesBoric Acid Batch Integrator, FIS-1 13Total Flow Batch Integrator, FIS-168Boric Acid Makeup Flow Controller, FK-113Primary Water Makeup Flow Controller, FK-1 68

Question origin: Same as Farley 2008 NRC Exam question #95,FNP Bank question PLT COMM-40502C09 02

Page33of5l


Basis for meeting K/A: UOP-1 .2 directs a dilution prior to the Reactor startup whichwill affect reactivity, and SOP-2.3 gives guidance for thistask. This question requires knowledge of the dilutionprocedure and manipulation of specific controls which willaffect reactivity for a condition in which the normal methodof dilution does not work.

In addition, UOP-1 .2 (a pre-startup procedure) states that“This procedure has been identified as involving aninfrequently performed test or evolution (IPTE) requiringSenior Line Manager oversight.” The question requiresknowledge of NMP-AD-006 to decide who will give the IPTEbriefing. This is an SRO function and knowledge to makesure the proper administrative requirements are met.

SRO justification: 10 CFR 55.43(b)(5)Knowledge of administrative procedures that specifyhierarchy, implementation, and/or coordination of plantnormal, abnormal, and emergency procedures.

NMP-AD-006 provides guidance on performing IPTE briefs.UOP-1 .2 states that an IPTE brief is required. SROknowledge is required to implement and coordinate therequirements of NMP-AD-006, while performing UOP-1 .2 fora Reactor startup. Conducting an IPTE brief, when an IPTEbrief is required, and qualifications for conducting an IPTEbrief are all part of NMP-AD-006 and are SRO knowledge.

2012 NRC exam






overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:— Knowledge of administrative procedures that specify hierarchy, implementation,

Page 34 of 51



Page 35 of 51

Original Bank 2008 Farley NRC QuestionThursday, March 01, 2012 Page 1 of 1

1. PLT COMM-40502C09 002!HLT!SRO/C/A 4.5!4.4/G2.2. 11//l

Unit 1 is in MODE 3 preparing to do an initial startup after a core reload lAW UOP-1 .2,Startup of Unit from Hot Standby to Minimum Load, and STP-1 01, Zero Power ReactorPhysics Testing.

The following conditions exist:

• The current RCS boron sample is 1540 ppm.• Both Shutdown banks have been withdrawn.• All RCPs are running and the RCS is at normal operating pressure and

temperature.• FT-i 68, TOTAL MAKEUP FLOW TO CHG/VCT, has failed LOW.

The crew is at the step in UOP-i .2 to dilute the RCS to the Hot Shutdown boronconcentration.The crew is also preparing to conduct a brief for an infrequently performed test orevolution (IPTE).

Which one of the following is the correct method used for diluting per SOP-2.3, CVCSReactor Makeup Control System, and which one of the following is the correct personto conduct the IPTE brief lAW NMP-AD-006, Infrequently Performed Tests andEvolutions?

A. • Perform a Manual Makeup to the charging pump suction. Verify the dilutionautomatically stops when batch integrator setpoint is reached.

• Shift Supervisor.

B. • Perform a Manual Makeup to the charging pump suction. Verify the dilutionautomatically stops when batch integrator setpoint is reached.

• Engineering Director

C. • Verify a RMW pump running; open RMW to Blender, FCV1 14B; Open MKUPTO CHG PUMP SUCTION, FCVI i3B; estimate the flow by VCT level rise; andcalculate time for dilution, manually secure dilution when complete.

• Shift Supervisor.

• Verify a RMW pump running; open RMW to Blender, FCV1 14B; Open MKUPTO CHG PUMP SUCTION, FCV1 1 3B; estimate the flow by VCT level rise; andcalculate time for dilution, manually secure dilution when complete.

• Engineering Director

Page 1 of 1


93. G2.2. 15 0931M0D/FAP1LEY2008/M 3 .9/4.3/G2.2. 15/N//I

During a system 2ragout, a Motor-Operated Valve (MOV) was manually closed (locallyat the valve) and its breaker was turned OFF. (Both the handwheel and the breakerwere included on the Tagout.)

The MOV is a normally closed valve, whose function is to automatically open on aSafety Injection signal.

The maintenance was completed and the Tagout was subsequently cleared.The following conditions currently exist:

• The MOV is closed.• No work was performed on the MOV.• The MOV power supply is on and the GREEN valve position light is

illuminated.

Which one of the following identifies the status of the MOV and the post maintenancetesting requirements in accordance with SOP-O.O, General Instructions to OperationsPersonnel?

A’ The MOV is NOT OPERABLE. The MOV must be electrically stroked one full cycle.

B. The MOV is OPERABLE. Remote observation of valve stroking is allowed.

C. The MOV is OPERABLE. Local observation of valve stroking is required.

D. The MOV is NOT OPERABLE. A system flow test is required.

Page 18of25


93. G2.2. 15 093/MOD/FARLEY 2008/M 3 .9/4.3/G2.2. 15/N/I!

A. Correct The valve is inoperable until electrically cycling the valve one full cycle,and that will allow clearing the LCO per FNP-O-SOP-O.O, Step 15.5.7.

B. Incorrect Plausible since during refueling outages, the MOV team performs stroketests on many MOV’s. If no adjustments are made to the valve, it isoperable after power is restored. The Post Maintenance test requirementis to remotely observe the proper valve stroke.

C. Incorrect Plausible, because this would be true for some maintenance or preventivemaintenance activities. During refueling outages, all valve strokes areobserved locally for the appropriate Surveillance Test. If the valve was notmanually stroked, and no work was performed on the valve or operator, itwould be considered operable when power was restored. De-energizingand then re-energizing a MOV in and of itself, does not make the valveinoperable.

D. Incorrect Plausible because the valve is not operable, and some Surveillance testsfor MOV’s do require a system flow test. Under these conditions a systemflow test is not required.

Page23of3O

25 SRO QUESTIONS Ver5Thursday, March 01, 2012 Page 1 of 2

93. G2.2.15 093/MOD/PARLEY 2008/M 3.9/4.3/G2.2.15/N///

G2.2.15

2.2.15 Ability to determine the expected plant configuration using design andconfiguration control documentation, such as(CFR: 41.10/43.3/45.13)IMPORTANCE RD 3.9 SRO 4.3


Technical Reference: FNP-0-SOP-0.0 Version 149.1


Learning Objective: Given a set of Plant Conditions ASSESS those conditionsand DETERMINE the ability of plant equipment andstructures to meet their intended, designated function(OPS-52302A06)

Question origin: Modified from Farley 2008 NRC Exam question #96,FNP Bank question INTRO TS-52302A06 04

Basis for meeting K/A: This question tests an SRO’s ability to determine theconfiguration (operability) of a valve from actions taken on aconfiguration control document (Tagout), as it relates to thevalve performing its design safety function.

SRO justification: 10 CFR 55.43(b) (5)Assessment of facility conditions and selection ofappropriate procedures during normal, abnormal, andemergency situations.

Knowledge of administrative procedures that specifyhierarchy, implementation, and/or coordination of plantnormal, abnormal, and emergency procedures.

This question tests the SRO knowledge of the administrativerequirements (from SOP-0.0) that assist an SRO indetermining operability of a component, and whetheradditional Post Maintenance Testing actions are required toestablish operability. A TS evaluation has to be performed todetermine operability of the component, but an in depthknowledge of SOP-0.0 is required since it is not informationthat is provided by the TS Bases. Manually stroking an MOVmakes the MOV inoperable, and the SRO has to determinethe actions required to clear the LCD and return the MOV toOPERABLE status. This is an SRO only function, since itrequires an OPERABILITY determination.

Page 1 of 2

25 SRO QUESTIONS Ver5Thursday, March01, 2012 Page 2 of 2

2012 NRC exam






overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:— Knowledge of administrative procedures that specify hierarchy, implementation,


Page 2of2


94. G2.3. 11 094/NEW/fM 3.8/4.3/G2.3.1 1/N//I

Unit I is at 100% power with the following conditions:

At 10:00:

• R-14, PLANT VENT, radiation monitor is INOPERABLE.• #4 Waste Gas Decay Tank (WGDT) release is in progress.• #8 WGDT has high activity and is at 60 psig.

At 10:15:

• #8 WGDT relief valve fails open.• The crew closes RCV-14.


For release of #4 WGDT with R-14 INOPERABLE, the ODCM requiressampling by analyzing (1)

The #8 WGDT relief discharge (2) isolated when RCV-14, GDT DISCHTO PLANT VENT, is closed.

A. 1) an initial sample prior to the release and then a grab sample once every8 hours when a release is in progress

2) IS

B. 1) an initial sample prior to the release and then a grab sample once every8 hours when a release is in progress

2) is NOT

C. 1) two independent samples prior to release

2) IS

ED” 1) two independent samples prior to release

2) is NOT

Page 19of25


94. G2.3. 11 094/NEW/!M 3.814.3/G2.3. 11/N//I

Per the ODCM, Table 3-1, Condition 2a, either R-14 or R-22 has to be operable for aContinuous release from the Plant Vent Stack. That is met by R-22 being operable, soAction 37 is not required -

ACTION 37 - Wfth the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, effluent releases via this pathway maycontinue provided grab samples are taken at least once per 8 hours and these samplesare analyzed for gross activity within 24 hours.

ODCM, Table 3-1, Condition 3 is not met due to R-14 being out of service, so Action 35is required when releasing a WGDT -

ACTION 35 - With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, the contents of the tank(s) may bereleased to the environment provided that prior to initiating the release:a. At least two independent samples of the tank’s contents are analyzed, andb. At least two technically qualified members of the Facility Staff independently verifythe discharge line valving, and(1) Verify the manual portion of the computer input for the release ratecalculations performed on the computer, or(2) Verify the entire release rate calculations if such calculations are performedmanually.Otherwise, suspend release of radioactive effluents via this pathway.

This question targets the sampling requirements per the ODCM for releasing a WGDT,along with systems knowledge of the relief flowpath.

A. Incorrect 1) Incorrect, this describes requirements similar to the ODCM for actionsof R-14 AND R-22 inoperable, and a continuous release from the PlantVent Stack.2) Incorrect, the relief on the #8 WGDT goes directly to the Plant VentStack. There are no isolation valves in the line. If a relief lifts, it will not beisolated by RCV-14. Plausible since #1-6 WGDTs have a relief thatdischarges to the #8 WGDT. This relief is then routed to just downstreamof RCV-14 isolation valve.

B. Incorrect 1) Incorrect, seeA.1.2) Correct, the release from #8 WGDT will not be stopped.

C. Incorrect 1) Correct, the ODCM Action 35 requires “At least two independentsamples of the tank’s contents are analyzed”.2) Incorrect, see A.2.

D. Correct 1) Correct, See C.1.2) Correct, See B.2.

Page 24 of 30


94. G2.3. 11 O94JNEW//M 3 .8/4.3/G2.3. 11/N//I

G2.3.1 I

2.3.11 Ability to control radiation releases.(CFR: 41.11 /43.4 /45.10)IMPORTANCE RD 3.8 SRO 4.3


Technical Reference: ODCM, Version 24Drawing D175042 sh. 6 v33


Learning Objective: RECALL AND APPLY the information from the LCD BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with Waste Gas System components andattendant equipment alignment, to include the following(OPS-62106B01): 1OCFR55.43 (b) 213.12.1, Waste Gas Monitoring Instrumentation13.12.3, Waste Gas Monitoring13.12.4, Gas Storage Tanks


Basis for meeting K/A: The SRO must display knowledge of effective actionsrequired to terminate a release when a high radiationcondition occurs during a release. In addition, SROknowledge is required for actions to initiate a release when arad monitor is inoperable. The ODCM provides actionsrequired to perform a release when a radiation monitorbecomes inoperable. This question provides a scenario inwhich the release cannot be continued until ODCM actionsare accomplished as determined by the SRO. The applicantmust display ability to perform actions for a radioactiverelease.


Same items listed above for the Technical RequirementsManual (TRM) and Offsite Dose Calculation Manual(ODCM).

Page 38 of 51


Requires application of REQUIRED ACTIONS of theODCM.

2012 NRC examFrom the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010flowchart for 10 CFR 55.43(b)(2)

1) can NOT be answered by knowing less than 1 hour Tech Specs.2) can NOT be answered by knowing information listed “above-the-line”.3) can NOT be answered by knowing the TS Safety Limits or their bases.4) Does involve one or more of the following for TS, TRM or ODCM:• Application of Required Actions (Section 3) and Surveillance Requirements (SR)

(Section 4) in accordance with rules of application requirements (Section 1)..

Page 39 of 51


95. 02.3.4 095/BANKJHAkRIS 2009/M 3.2/3.7/02.3.4/N//I

A General Emergency has been declared for the Plant Site.


The MAXI MUM exposure th a—diation--worker--can-be-required- to receivefor a life saving mission during a declared emergency is (1) REM TEDE.

The- LOWEST level of-author1ty---thatcan-authori-ze--this-dose is the (2)

(1) (2)

A. 10 HP Supervisor

B. 25 HP Supervisor

C” 25 Emergency Director (ED)

D. 10 Emergency Director (ED)

0

Page 20 of 25


95. G2.3.4 095/BANK/HARRIS 2009/M 3.2/3.7/G2.3.4/N///

A. Incorrect 1) Incorrect, 10 REM TEDE is plausible because this is the limit forProtecting Valuable Property, but for saving a life, 25 REM can berequired.2) Incorrect, Plausible because the HP Supervisor can give approval toexceed plant administrative dose limits, but not IOCFR2O limits.

B. Incorrect 1) Correct, per EIP-14.2) Incorrect, see A.2.

C. Correct 1) Correct, see B.1.2) Correct, per EIP-14, the ED has to approve exceeding 1OCFR2O limits.

D. Incorrect 1) Incorrect, seeA.1.2) Correct, see C.2.

Page 25 of 30


95. G2.3.4 095/BANKIHARRIS 2009/M 3.2/3 .71G2.3 .4/NI/I

G2.3.4

Knowledge of radiation exposure limits under normal or emergency conditions.(CFR: 41.12/43.4/45.10)IMPORTANCE RO 3.2 SRO 3.7


Technical Reference: FNP-0-EIP-14.0, v26


Learning Objective: Given an emergency scenario, EVALUATE plant conditionsand DETERMINE if a reentry, relocation or movement isrequired. EVALUATE the conditions and DETERMINE doselimits for personnel involved. (OPS-63002D01).

Question origin: Same as Harris December 2009 NRC Exam question #97

Basis for meeting K/A: Knowledge of Emergency radiation exposure limits perEIP-14.0 is demonstrated by this question.


Justification: A question that is not tied to one of the 10 CFR55.43(b) items can still be classified as “SRO-only” providedthe licensee has documented evidence to prove that theknowledge/ability is “unique to the SRO position” at the site.An example of documented evidence includes:• The question is linked to a learning objective that isspecifically labeled in the lesson plan as being SRO-only(e.g., some licensee lesson plans have columns in themargin that differentiate AO, RO, and SRO learningobjectives) [NUREG 1021, ES-401, Section D.2.d}AND/OR• A question is linked to a task that is labeled as anSRO-only task, and the task is NOT listed in the RO tasklist.

The SRO is solely responsible for direction andimplementation of the Emergency Plan at FNP, includingapproving emergency dose limits. The objective listedabove is an SRO only objective.

2012 NRC exam

________

Page 40 of 51


Ill. Justification for Plant Specific Exemptions UNIQUE to the SRO position:


Page 41 of 51


96. G2.4.18 096/BANKIVOGTLE 201 1IM 3.3/4.0/G2.4.18/N///

FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress on Unit 1 with thefollowing conditions:

• The operators are initiating RCS Bleed and Feed.• PCV-444B, PRZR PORV, and its associated Block Valve are OPEN.• PCV-445A, PRZR PORV, will NOT open.• Restoration of SG feed capability is imminent.


Per FRP-H.1, opening ONLY one PORV (1) provide sufficient RCS bleedflow to permit adequate RCS heat removal.

After SG feed flow is restored (2)

(1) (2)

A. WILL immediately go to EEP-1, Loss of Reactor orSecondary Coolant

B. will NOT immediately go to EEP-1, Loss of Reactor orSecondary Coolant

C. WILL remain in FRP-H.1 until at least one SG narrowrange level is >31%

D” will NOT remain in FRP-H.1 until at least one SG narrowrange level is >31%

Page 21 of 25


96. G2.4.18 096/BANK!VOGTLE 201 1IM 3.3/4.01G2.4.181N!//

A. Incorrect 1) Incorrect, FRP-H.1 requires opening Rx Head Vents if only one PORVis opened. Per the background document for H.1, opening two PORV’sensures that sufficient depressurization of the RCS can occur for HighHead Injection flow to provide core cooling.2) Incorrect, Plausible because once SG feed flow is restored, thecandidate has to take actions to stop the depressurization. There iseffectively a LOCA in progress, so EEP-1 would be a logical transition. Inaddition, in FRP-H1, EEP-1 is one of the transitions out of H.1 if RCSpressure is not restored after closing PORV’s and Head Vents.

B. Incorrect 1) Correct, per FRP-H.1 Bases. Opening two PORV’s is required toensure adequate Core cooling is maintained.2) Incorrect, see A.2.

C. Incorrect 1) Incorrect, seeA.1.2) Correct, per FRP-H.1 the operator is required to stay in Hi until atleast one SG level is above 31%. This is to ensure that conditions forentering H.1 are not met as soon as it is exited.

D. Correct 1) Correct, see Bi.2) Correct, see C.2.

Page 26 of 30


96. G2.4.18 096/BANKIVOGTLE 201 1IM 3.3/4.0/G2.4.18!N!//

G2.4.18

Knowledge of the specific bases for EOPs.(CFR: 41.10 /43.1 / 45.13)IMPORTANCE RO 3.3 SRO 4.0


Technical Reference: FNP-1-FRP-H.1, v27FNP-0-FRB-H.1, v3


Learning Objective: ASSESS the facility conditions associated with the (1)FRP-H.1, Response to Loss of Secondary Heat Sink; (2)FRP-H.2, Response to SG Overpressure; (3) FRP-H.3,Response to SG High Level; (4) FRP-H.4, Response toLoss of Normal Steam Release Capabilities; (5) FRP-H.5,Response to SG Low Level, and based on that assessment:(OPS-62533F01)• SELECT the appropriate procedures during normal,abnormal and emergency situations. 1OCFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required• DETERMINE if the critical safety functions are satisfied

Question origin: Same as Vogtle 2011 NRC Exam, Question #99

Basis for meeting K/A: The candidate is presented with a scenario where RCSBleed and Feed is required with the inability to open onePORV. Background knowledge of FRP-H.1 is required toknow that opening one PORV does not provide sufficientRCS depressurization to ensure that adequate core coolingcan be maintained. Per the WOG Background documents,both PORV’s must be open to ensure adequate heatremoval is provided.


The applicant’s knowledge can be evaluated at the level of10 CFR55.43(b)(5) by ensuring that the additionalknowledge of the procedure’s content is required to correctlyanswer the written test item, for example:• Knowledge of diagnostic steps and decision points in the

Page 42 of 51


emergency operating procedures (EOP) that involvetransitions to event specific subprocedures or emergencycontingency procedures.

2012 NRC exam






overall mitigative strategy of a procedure.• be answered with knowledge of ONE or MORE of the following:— Knowledge of diagnostic steps and decision points in the emergency operating

procedures (EOP) that involve transitions to event specific subprocedures oremergency contingency procedures.

Page 43 of 51

99. G2.4. 18 003/31N/AIEOP BASISJ4.O CIAJNEWIINRCIGCW

19231-C, “FR-Hi, Response To Loss of Secondary Heat Sink” is in effect.

- All SG WR levels are - 8%.

- PORV 455 and associated Block Valve are open.

- PORV 456 Block Valve is shut and cannot be opened.

- All Reactor Vessel Head Vent valves are open.

- All CCPs and SIPs are running.

- CETs are 562°F and stable.

- Restoration of feed capability is imminent.

Which ONE of the following is CORRECT regarding:

1) the current RCS bleed path, and

2) the proper procedural action to take when feed flow is restored?

RCS Bleed Path Feed Restoration

A. adequate remain in 19231-C until at least one SG NR level >10%

B. adequate immediately exit to 19010-C, “Loss of Reactor orSecondary Coolant”

C not adequate remain in 19231-C until at least one SG NR level> 10%

D. not adequate immediately exit to 19010-C, “Loss of Reactor orSecondary Coolant”

Page: 196 of 1994/13/2011

2.4 Emergency Procedures I Plan

2.4.18 Knowledge of the specific bases for EOPs.(CFR: 41.101 431 l 4513>

KIA MATCH ANALYSIS

The candidate is presented with a scenario where RCS feed and bleed is required withthe inability to open one PORV and RCS Head Vents are in open. In addition, whenfeed capability is restored with Hot Dry SG’s he must determine how many SGs to feed.The information regarding this is found in the WOG Background Documents for FR-HiLoss of Secondary Heat Sink.

SRQ 1OCFR55.43 (b5)

ANSWER I DISTRACTOR ANALYSIS

A. Incorrect. Per WOG BG documents, both PORVs must be open for adequate heatremoval. Crew must remain in 19231 until feed and bleed is terminated.

8. Incorrect. Per WOG BG documents, both PORVs must be open for adequate heatremoval. Crew must remain in 19231 until feed, and bleed is terminated.

C. Correct. Per WOG BG documents, both PORVs must be open for adequate heatremoval. Crew must remain in 19231 until feed and bleed is terminated and SG NRlevel is> 10%.

D. Incorrect. Bleed path is inadequate and Crew must remain in 19231 until feed andbleed is terminated.

REFERENCES

FR-Hi, WOG Background Documents for Loss of Secondary Heat Sink

V-LO-HO-37051, Loss of Secondary Heat Sink

VEGP learning objectives:

LO-LP-37051 -05, State th precautions which should be taken in feeding a hot, drysteam following recovery from a loss of heat sink accident.

LO-LP-37051-08, Using EOP 19231 as a guide, briefly describe how each major step isaccomplished. Describe the bases for each. (commitment).

Page: 197 of 199 4/13/2011


97. G2.4.32 097/MOD/TURKEY PT 2010/C/A 3 .6/4.0/G2.4.32/N/EMERG CLASS//Unit 2 is at 100% power with the following conditions:

• A problem in the Cable Spreading Room has caused an unplanned lossof ALL Main Control Board Annunciators

• AOP-35.0, Loss of Main Control Board Annunciators, is in progress.• An Emergency declaration is being evaluated per NMP-EP-110,

Emergency Classification Determination and n itial Action.


A NOUE classification is required when there is an unplanned loss of jJ,MOB Annunciators for a MINIMUM of (1) minutes.

Upon a loss of annunciator MHI, FIRE, AOP-35.0 requires establishing(2) within one hour.

(1) (2)

A. 15 hourly fire watches in all affected zones

B 15 a Pyrotronics panel firewatch

C. 30 hourly fire watches in all affected zones

D. 30 a Pyrotronics panel firewatch

Page 22 of 25


97. G2.4.32 097/MOD/TURKEY PT 2010/CIA 3 .6/4.0/G2.4.32/N/EMERG CLASS//

NOUE classification is required for the following:UNPLANNED loss of most OR all (approximately 75% of the MCBannunciators) OR indicators associated with safety systems for greater than15 minutes.

AOP-35.0, Loss of Main Control Board Annunciators -

8.0 IF FIRE annunciator MH1 is inoperable, THEN establish pyrotronics panelfirewatch within one hour.

A. Incorrect 1) Correct, a loss of all Annunciators for> 15 minutes would be a NOUE.2) Incorrect, plausible because this is similar to the requirement for thePyrotronics panel being out of service. If the Pyrotronics panel is removedfrom service, hourly and continuous fire watches have to be established inthe zones covered by the Pyrotronics panel.

B. Correct 1) Correct, see A.1.2) Correct, a Pyrotronics panel firewatch is required within one hour perAOP-35, step 8.0.

C. Incorrect 1) Incorrect, a loss of all Annunciators for> 15 minutes would be a NOUE.Plausible because there is no transient in progress and someclassifications do have a 30 minute requirement - HA4 #2 from the hotmatrix, CGI #2 from the cold matrix, etc.2) Incorrect, see A.2.

D. Incorrect 1) Incorrect, see C.1.2) Correct, see B.2.

Page 27 of 30


97. G2.4.32 097/MOD/TURKEY PT 2010/C/A 3.6/4.0/G2.4.32/N/EMERG CLASS//G2.4.32

Knowledge of operator response to lossof1àii)nnunciators.(CFR:41.10/43.5/45.13)IMPORTANCE RO 3.6 SRO 4.0


Technical Reference: NMP-EP-110-GLO1, v2FNP-0-SOP-0.4, v83.2FNP-2-AOP-35.0 v8


Learning Objective: Using plant procedures/references, ANALYZE a set of plantconditions and DETERMINE the proper classification of theemergency condition as being a NOUE, Alert, Site Area, orGeneral Emergency. (OPS-63002C01).

ANALYZE plant conditions and DETERMINE the successfulcompletion of any step in AOP-35.0, Loss of MCBAnnunciators. (OPS-52521 M07)

Question origin: Modified from Turkey Point 2010 NRC Exam, Question #99.

Basis for meeting K/A: Question tests the knowledge of the SRO response to a lossof all annunciators. A loss of annunciators is criteria for anEmergency classification and the SRO is tested on hisknowledge of when NOUE conditions are met. In addition,the candidate is tested on the specific actions of a loss ofannunciator MH1, FIRE, per AOP-35.0.


Justification: A question that is not tied to one of the 10CFR 55.43(b) items can still be classified as “SRO-only”provided the licensee has documented evidence to provethat the knowledge/ability is “unique to the SRO position”at the site. An example of documented evidence includes:• The question is linked to a learning objective that isspecifically labeled in the lesson plan as being SRO-only(e.g., some licensee lesson plans have columns in themargin that differentiate AO, RO, and SRO learningobjectives) [NUREG 1021, ES-401, Section D.2.d]AND/OR• A question is linked to a task that is labeled as an

Page 44 of 51


SRO-only task, and the task is NOT listed in the RO tasklist.

The SRO is solely responsible for determiningClassifications at FNP, the first objective listed above isan SRO only objectives.

In addition, 10 CFR 55.43(b)(1)Conditions and limitations in the facility license.

Administration of fire protection program requirementssuch as compensatory actions associated with inoperablesprinkler systems, fire doors, etc.

There are compensatory administrative actions requiredwithin one hour for the inoperable annunciator MH1,FIRE.

2012 NRC exam

III. Justification for Plant Specific Exemptions UNIQUE to the SRO position:


10 CFR 55.43(b)(1)Conditions and limitations in the facility license.• Administration of fire protection program requirements such as compensatory actionsassociated with inoperable sprinkler systems, firedoors, etc.

Page 45 of 51

TURKEY POINT 2010 NRC EXAM

(I)

A. 15 minutes

B. 15 minutes

C. 30 minutes

D. 30 minutes

(2)

Hot Ring Down

Emergency Notification System

Hot Ring Down

Emergency Notification System

Question 99

Which ONE of the following completes both of the following statements?

An Unusual Event is an unplanned loss of Safety System Annunciation orIndication in the Control Room for at least (1) or longer.

If an Unusual Event is declared, the Communicator will use the (2)Phone to contact the State Warning Point.

TURKEY POINT 2010 NRC EXAM

Question 99

K/A G2.4.32Knowledge of operator response to loss of all annunciators

CFR/IR43.5 3.6/4.0

Reference:1. Turkey Point EAL Classification Tables — Hot Conditions p. 2 rev.

07/09/102. 0-EPIP-20134 section 4.0 rev. 11/10/08

Classification required by 1230 and should be made as soon as determinedthe loss of annunciators will exceed 15 minutes. The HRD is used to notifythe SWP; the ENS is used to notify the NRC.

Question history:New

Correct answer: C

A. Incorrect AW above discussion. Plausible — PTN implemented thenew classifications this year. Before that, the expectation was the SMwould wait the required times before making a classification.

B. Incorrect lAW above discussion. Plausible — PTN implemented thenew classifications this year. Before that, the expectation was the SMwould wait the required times before making a classification. The HRDand ENS are listed in EPIP-20134.

C. Correct lAW above discussion.

D. Incorrect lAW above discussion. Plausible — The HRD and ENS arelisted in EPIP-20134.

Cog / LOD: Comprehension or Analysis / 2

Lesson and objective: 6900041 E04


98. WEO4EG2A.9 098/NEW//C/A 3 .8/4.2/WIEO4EG2.4.9/TS 3.4.6//IThe following conditions exist on Unit 1:

At 10:00:

• lB RHR pump is running in a cooldown alignment.• 1A RHR pump is in standby.• RCS temperature is 300°F.• All Reactor Coolant Pumps have been Tagged Out.

At 10:10:

• Water is leaking into the 1 B RHR pump room.• NE2, lB RHR PUMP RM SUMP LVL HI-HI OR TRBL, is in alarm

on the BOP Panel.• RCS pressure is going down.• Pressurizer level is going down.


The leak will be terminated by performing the actions of (1)

Actions required by Tech Spec 3.4.6, RCS Loops - Mode 4, afterprocedural actions are complete, are to be in Mode 5 in a maximumof (2)

REFERENCE PROVIDED

(1) (2)

A. ECP-1 .2, LOCA Outside Containment 24 hours

B” AOP-12.0, Residual Heat Removal 24 hoursSystem Malfunction

C. ECP-1 .2, LOCA Outside Containment 26 hours

D. AOP-12.0, Residual Heat Removal 26 hoursSystem Malfunction

Page 23 of 25


98. WEO4EG2.4.9 098/NEW//C/A 3 .8/4.2/W/EO4EG2.4.9/TS 3.4.6//I

TS 3.4.6 note:NOTES

1. All reactor coolant pumps (RCPs) and RHR pumps may not be inoperation for = 2 hours per 8 hour period provided:a. No operations are permitted that would cause reduction of theRCS boron concentration; andb. Core outlet temperature is maintained at least 10°F belowsaturation temperature.

A. Incorrect 1) Incorrect, per SOP-0.8, EEP-0 is not applicable in Mode 4. Plausiblebecause EEP-0 does not specify applicable modes and is normally theentry point when a major RCS leak is encountered. In addition, TS 3.3.2,ESF Actuation System Instrumentation, requires Manual initiation of SI tobe available until Mode 5 (<200°F) is entered. SI is available for thecurrent conditions.2) Correct, TS 3.4.6 requires being in Mode 5 within 24 hours.

B. Correct 1) Correct, AOP-1 2 actions are applicable for an RCS leak or RHR pumptrip in Mode 4.2) Correct, see A.2.

C. Incorrect 1) Incorrect, see A.1.2) Incorrect, Plausible because TS 3.4.6 has a Note that allows all RHRand RCP’s to be secured for 2 hours per 8 hour period. A candidatecould determine that Mode 5 is required in 26 hours instead of 24 hours.TS Bases knowledge is required to know that the note specifically appliesto performance of Surveillance tests.


Page 28 of 30


98. WEO4EG2.4.9 098/NEW//C/A 3 .8/4.21W/EO4EG2.4.9/TS 3.4.6//IWEO4EG2.4.9LOCA Outside Containment

2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolantaccident or loss of residual heat removal) mitigation strategies.(CFR: 41.10/43.5/45.13)IMPORTANCE RO 3.8 SRO 4.2


Technical Reference: FNP-1 -AOP-1 2.0, v24FNP-1-EEP-0.0, v43FNP-0-SOP-0.8, v20Tech Specs, v186Tech Spec Bases, v53

References provided: Tech Spec 3.4.6 page 3.4.6-1 and 3.4.6-2

Learning Objective: RECALL AND APPLY the information from the LCO BASESsections: BACKGROUND, APPLICABLE SAFETYANALYSIS, ACTIONS, SURVEILLENCE REQUIREMENTS,for any Technical Specifications or TRM requirementsassociated with the Residual Heat Removal Systemcomponents and attendant equipment alignment, to includethe following (OPS-62101K01): 1OCFR55.43 (b) 23.4.3, RCS Pressure and Temperature (P/T) Limits3.4.6, RCS Loops — MODE 43.4.7, RCS Loops - MODE 5, Loops Filled3.4.8, RCS Loops - MODE 5, Loops Not Filled3.4.12, Low Temperature Overpressure Protection (LTOP)

System3.4.14, RCS Pressure Isolation Valve (PIV) Leakage3.5.2, ECCS — Operating3.5.3, ECCS — Shutdown3.9.4, Residual Heat Removal (RHR) and CoolantCirculation - High Water Level3.9.5, Residual Heat Removal (RHR) and CoolantCirculation - Low Water Level13.5.1, Emergency Core Cooling System (ECCS)


Basis for meeting K/A: For this question, a LOCA outside Containment hasoccurred. The candidate has to determine the properprocedure application for the current conditions, which isdifferent than the procedural flowpath for normal, 100%power applications - knowledge of accident mitigation

Page 46 of 51


strategies during low power/shutdown conditions. Accidentmitigation strategies change when changing from Mode 3 toMode 4.

SRO justification: 10 CFR 55.43(b)(2)Know’edge of TS bases that is required to analyze TSrequired actions and terminology.

TS 3.4.6 has a Note that allows all RHR and RCP’s to besecured for < 2 hours per 8 hour period. A candidate coulddetermine that Mode 5 is required in 26 hours instead of 24hours. TS Bases knowledge is required to know that thenote specifically applies to performance of SurveillanceTests.

2012 NRC exam

10 CFR 55.43(b)(2)

Facility operating limitations in the TS and their bases.From the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010flowchart:



Page 47 of 51


99. WE 1 OEA2.2 099/BANK/FARLEY 2010/C/A 3.4/3 .9/W/E 1 OEA2.2/ESP-0.2 ATTACH3///Unit I is performing a natural circulation cooldown in accordance with ESP-0.2, NaturalCirculation Cooldown to Prevent Reactor Vessel Head Steam Voiding. The followingconditions exist:

• RCS cold leg temperatures are 525°F.• RCS pressure is 1900 psig.• 1A and lB CRDM fans are running.• RCPs 1A, IB, and IC are tripped and can NOT be restarted.• CST level is 10.0 ft.• AFW flow is 350 gpm• RCS cooldown rate is 5°F/hr.PZieveLisstableat25%.• RVLIS is NOT available

Which one of the following is the correct response lAW ESP-0.2 and the MAXIMUMpermitted cooldown rate to lower TcoLD to 500° F?

Procedure names are as follows:

ESP-0.2, Natural Circulation Cooldown to Prevent Reactor Vessel Head SteamVoiding

ESP-0.4, Natural Circulation Cooldown With Allowance For Reactor VesselHead Steam Voiding(Without RVLIS) ‘\

REFERENCE PROVIDED

A. • Continue with ESP-0.2.

• The maximum cooldown rate is < 50°F/hr.

B. • Continue with ESP-0.2.

• The maximum cooldown rate is 100°F in any 60 minute period.

C’ • Transition to ESP-0.4.

• The maximum cooldown rate is < 50°F/hr.

D. • Transition to ESP-0.4.


Page24of25


99. WE1OEA2.2 099/BANK/FARLEY 2010/C/A 3.4/3.9/W/E1OEA2.2/ESP-0i ATTACH3///

A. Incorrect 1) Incorrect, ESP-0.2 maximum allowed CID rate is 25°F/hr and with anRCS temp of 525°F, a minimum of 1 7ft of CST inventory is required toremain in ESP-0.2. An accelerated cooldown is required thereforeESP-0.3 or ESP-0.4 must be used. Plausible because the referenceprovided demonstrates that an increased cooldown is required due toCST inventory concerns. Without knowledge of the C/D limit of ESP-0.2,remaining in ESP-0.2 might be thought the preferred procedure, when noindication of a void is present (STABLE PZR level).2) Correct, since there is insufficient CST inventory, and RVLIS isunavailable ESP-0.4 must be used. Plausible because in ESP-0.4, whileRCS temperature is being reduced to 500°F the maximum allowedcooldown rate is 50°F/hr.

B. Incorrect 1) Incorrect, see A.1.2) Incorrect, plausible since 100 °F in any 60 minute period is the normallimit for RCS cooldown and would also be correct if ESP-0.4 wereimplemented at any Tcojd <500°F, OR if ESP-0.3 were the procedureimplemented (RVLIS available).

C. Correct 1) Correct, 25°F/hr cooldown is allowed by ESP0.2 but in this case thereduced CST volume requires a cooldown rate of greater than 25°F/hr (inthis case 75°F/hr) ESP-0.2 limits cooldown to 25°F/hr therefore step 10RNO must be implemented to go to ESP-0.3 or ESP-0.4, and with RVLISunavailable, ESP-0.4 is required.2) Correct, this is the correct C/D limit until RCS temp is reduced to 500°Fin ESP-0.4.

D. Incorrect 1) Correct, see C.12) Incorrect, see B.2. Plausible: A faster cooldown rate of <100°F in any60 minute period is permitted by ESP-0.4 when RCS temp is <500°F andRCS pressure is 1850 psig.

Page 29 of 30


99. WE1OEA2.2 099/BANK/FARLEY 2010/C/A 3.4/3.9/W/E1OEA2.2/ESP-0.2 ATTACH3///

WE1OEA2.2Natural Circulation with Steam Void in Vessel with/without RVLIS

EA2. Ability to determine and interpret the following as they apply to the (NaturalCirculation with Steam Void in Vessel with/without RVLIS) (CFR: 43.5 / 45.13)EA2.2 Adherence to appropriate procedures and operation within the limitations in thefacility*s license and amendments.IMPORTANCE RO 3.4 SRO 3.9


Technical Reference: FNP-1-ESP-0.4, v21FNP-1-ESP-0.2, v19

References provided: Attachment 3 of ESP-0.2

Learning Objective: ASSESS the facility conditions associated with the (1)ESP-0.2, Nat Circ CID to Prevent Reactor Vessel HeadSteam Voiding; (2) ESP-0.3, Nat Circ C/D with Allowance forReactor Vessel Head Steam Voiding (with RVLIS); (3)ESP-0.4, Nat Circ CID with Allowance for Reactor VesselHead Steam Voiding (without RVLIS), and based on thatassessment: (OPS-62531 COl)• SELECT the appropriate procedures during normal,abnormal and emergency situations. I 0CFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required• DETERMINE if the critical safety functions are satisfied

Question origin: Same as Farley 2010 NRC Exam Question #98

Basis for meeting K/A: 1) ESP-0.4 and the mitigation strategy is chosen to addressWE1O portion of the K/A; this procedure is only applicable inNatural Circ cooldown without RVLIS when conditionsrequire a faster cooldown than allowed by ESP-0.4. Somevoiding in the Head is expected in ESP-0.4.2) Procedural adherence: is satisfied by knowledge of theappropriate procedure for the given conditions.3) Operation within the limitations of the Facility license isdemonstrated by challenging the knowledge of actions forlow CST inventory and cooldown rate limitations.


Page 48 of 51


2012 NRC exam








Page49of5l

98. Unit us performing a natural circulation cooldown in accordance with ESP-0.2, NaturalCirculation Cooldown to Prevent Reactor Vessel Head Steam Voiding. The followingconditions exist:

• RCS cold leg is 525°F.• RCS pressure is 1900 psig.• IA and lB CRDM fans are running.• RCPs 1A, IB, and 1C are tripped and cannot be restarted.• CST level is 10.0 ft.• AFW flow is 350 gpm• RCS cooldown rate is 5°F/hr.• PZR level is stable at 25%.• RVLIS is not available.

Which one of the following is the correct response lAW ESP-0.2 and the maximumpermitted cooldown rate to lower TCOLD to 500° F?

REFERENCE PROVIDED

Procedure names are as follows:

ESP-0.2, Natural Circulation Cooldown to Prevent Reactor Vessel Head SteamVoiding

ESP-0.4, Natural Circulation Cooldown With Allowance For Reactor VesselHead Steam Voiding (Without RVLIS)

A. • Continue with ESP-0.2.

• The maximum cooldown rate is <50°F/hr.

B. • Continue with ESP-0.2.


C. Transition to ESP-0.4.

• The maximum cooldown rate is <50°F/hr.

D. • Transition to ESP-0.4.



100. WEI 1EA2. 1100/NEW//C/A 3.4/4.2/WE 11 EA2. 1/N//IUnit I has experienced a Large Break LOCA and Cold Leg Recirculation has beenestablished. The following conditions exist:

At 10:0O• IA RHR pump is Tagged Out.• ECP-1 .3, Loss of Emergency Coolant Recirculation Caused by Sump

Blockage, is entered.

At 10:30:• CETCs are 705°F and rising.

At 10:45:• IA RHR pump Tag Out has been cleared.• 1A RHR pump has been started with stable amps and flow.• CETCs are 695°F and trending down.


At 10:30, FRP-C.2, Response to Degraded Core Cooling, (1) be performed. -

/1- V4Atter core cooling is restored, (2)

(1) (2)

A. WILL remain in ECP1.3

B” will NOT remain in ECP-I.3

C. WILL transition to EEP-1 .0, Loss of Reactor orSecondary Coolant

D. will NOT transition to EEP-I .0, Loss of Reactor orSecondary Coolant

Page 25 of 25


100. WE1 1 EA2. 1100/NEW//C/A 3.4/4.2/WE 1 1EA2. 1/N//I

A. Incorrect 1) Incorrect, plausible because normally this is the correct action to take.If the transition was made to FRP-C.2, and core cooling was restored,step 8.3 would exit to procedure and step in effect. Detailed knowledge ofECP-1 .3 is required to know that there is a note at the beginning ofECP-1 .3 that states “CSF Status Trees should be monitored forinformation only. FRPs should not be implemented.”2) Correct, ECP-1 .3 states “There is no return to the ERP network oncethis procedure is entered.’. The operator will remain in ECP-1 .3 once it isentered.

B. Correct 1) Correct, FRP’s are not implemented during the performance ofECP-1.3. SeeA.1.2) Correct, see A.2.

C. Incorrect 1) Incorrect, see A.1.2) Incorrect, plausible because many procedures will be exited once theyhave completed their goal. FRP-C.2 could be exited at step 8.3. Acandidate may infer that if core cooling is re-established and the RHRpump was operating with no problems, ECP-1 .3 would no longer berequired and transition to EEP-1 .0 would be appropriate. In this instance,core cooling was restored by starting an unaffected RHR pump. EEP-1 .0is the normal controlling procedure during a Large Break LOFA.


Page 30 of 30


100. WE1 1EA2. 1 100/NEW/IC/A 3.4/4.2/WE 1 1EA2. 1/N//IWEI 1EA2.1Loss of Emergency Coolant Recirculation

EA2. Ability to determine and interpret the following as they apply to the (Loss ofEmergency Coolant Recirculation) (CFR: 43.5 /45.13)EA2.1 Facility conditions and selection of appropriate procedures during abnormal andemergency operations.IMPORTANCE RO 3.4 SRO 4.2


Technical Reference: FNP-1-ECP-1.3, v4FNP-1-FRP-C.2, v17FNP-1-CSF-0, v17


Learning Objective: ASSESS the facility conditions associated with the (1)ECP-1.1, Loss of Emergency Coolant Recirculation; (2)ECP-1 .3, Loss of Emergency Coolant Recirculation, Causedby Sump Blockage, and based on that assessment:(OPS-62532D01)• SELECT the appropriate procedures during normal,abnormal and emergency situations. IOCFR55.43 (b) 5• DETERMINE if transition to another section of theprocedure or to another procedure is required• DETERMINE if the critical safety functions are satisfied


Basis for meeting K/A: A loss of Emergency Recirculation has occurred and actionsare being performed in ECP-1 .3, Loss of EmergencyCoolant Recirculation Caused by Sump Blockage.Conditions have been met for implementation of FunctionalRestoration Procedure FRP-C-2, for an Orange CSF.Candidate is questioned on the selection of the appropriateprocedural flowpath under these conditions. In addition,candidate is questioned on the proper procedural actions totake when core cooling is restored.


Conditions are evaluated (Core cooling is restored) andcandidate is questioned on the proper procedure with which

Page 50 of 51


to proceed. Most of the time a transition is made back to thenormal Emergency Event procedure when conditions arerestored in an ECP.

10 CFR 55.43(b)(5)Knowledge of administrative procedures that specifyhierarchy, implementation, and/or coordination of plantnormal, abnormal, and emergency procedures.

- Knowledge of how Functional Restoration Procedures areimplemented using the guidance of Administrativeprocedure - SOP-0.8, Transient Response Procedure User’sGuide is an SRO function. Normally an FRP would beimplemented immediately during an ESP or ECP, butdetailed procedure knowledge is required to know thatFRP’s are not implemented in ECP-1 .3. This is allowed perthe guidance of SOP-0.8.

2012 NRC exam








— Knowledge of administrative procedures that specify hierarchy, implementation,and/or coordination of plant normal, abnormal, and emergency procedures.

Page 51 of 51

Re4Q,1cc - F(2O h(i 1icry

DOCUMENT ID PAGESFNP-0-AOP-29.0, v40.0 Table I PARTIAL- Pg 1 of 4 & pg 3 of 4

2 Pages

FNP-1-EEP-3.0, v26 PARTIAL- Pg 39 of 541 Page

FNP-0-EIP-8.0, v108 PARTIAL2 Pages

FNP-1-ESP-0.2, v19 Attachment 3 COMPLETE1 Page

COLR UNIT I Cycle 24 Figure 1, Aug 2010 COMPLETE - Pg 9 of 12I Page

TS 3.1.4 v186 PARTIAL2 Pages

TS 3.4.6 v186 COMPLETE2 Pages

TS 3.7.7 v186 PARTIALI Pages

D-175009 Sheet I of—I Page

NOTE: LARGER drawing available upon request (tablet 1 1X17)

FNP-0-AOP-29.0 PLANT FIRE Revision 40

TABLE 1

ACTIONS REQUIRED TO MAINTAIN AT LEAST ONE TRAIN OF SAFE SHUTDOWN

**************************************************************************************

CAUTION: Cables/components can fail in the event of a fire causing a loss ofreactor coolant inventory. The rooms containing thesecables/components are designated with an asterisk (*)•

**************************************************************************************

NOTE: • This table lists various rooms/locations that contain redundantsafe shutdown equipment or could potentially result in loss ofreactor coolant inventory. Appendix R safe shutdown analysisdocuments the basis of safe shutdown for each fire area.

• For each location listed, an appropriate attachment iscross - referenced.

1 For a fire in an area listed in Table 1, implement the actionsrequired by the associated attachment.

NOTE: A fire that is outside an area listed in Table 1 will not affect bothtrains of a safe shutdown function.

2 IF a fire IS NOT in an area listed in Table 1,THEN use other plant procedures for any actions that may be required.

-END-

Page 1 of 4

FNP-O-AOP-29.0 PLANT FIRE Revision 40

Table 1

Unit 1 Fire Areas

Fire Area Location Attachment1-001 UNIT 1 AUXILIARY BUILDING RAD SIDE 121’ AND 100’ 4

PENETRATION ROOMS, 83’ ELEVATION AND 77’ ELEVATION1-004 Zone 1 UNIT 1 AUXILIARY BUILDING 100’ RAD SIDE EXCEPT 5

PENETRATION ROOM AND CHARGING PUMP AREA1-004 Zone 2 UNIT 1 AUXILIARY BUiLDING 121’ AND 127’ RAD SIDE 6

EXCEPT THE 121’ PENETRATION ROOM1-004 Zone 3 UNIT 1 AUXILIARY BUILDING 139’ RAD SIDE EXCEPT PRF 6

ROOM AND THE ELECTRICAL PENETRATION ROOMS1-004 Zone 4 UNIT 1 AUXILIARY BUILDING 155’ ELEVATION AND SFP 7

HVAC ROOM1-005 CHARGING PUMP ROOMS AND HALLWAY AND STORAGE ROOMS 81-006 UNIT 1 NON RAD AUXILIARY BUILDING 100’ ELEVATION 91-006 UNIT 1 MAIN STEAM AND FEEDWATER VALVE ROOM 101-008 UNIT 1 TRAIN A VERTICAL CABLE CHASE 11*1009 UNIT 1 TRAIN B VERTICAL CABLE CHASE 121-0 2 IIT 1 HOT SHUTDOWN PANEL ROOM

1-014 UNIT I COMPUTER ROOM 201-015 UNIT 1 COIThfflNIATIONS ROOM1-016 UNIT 1 B TRAIN AUX BLDG BATTERY ROOM 61-017 UNIT 1 A TRAIN AUX BLDG BATTERY ROOM1018 UNIT 1 TRAIN A DC SWITCHGEAR ROOM AUX BUiLDING 51-019 UNIT 1 TRAIN B DC SWGR ROOM

1-021 UNIT 1 AUX BLDG SWGR ROOM B TRN 121-023 UNIT 1 AUX BLDG CRDM SWGR RM 121-030 UNIT I AUX BLDG B TRAIN CABLE CHASE 121-031 UNITI AUX BLDG A CABLE TUNNEL

*1034 B TRAIN ELECTRICAL PENE ROOM AND PENETRATION ROOM 16FILTRATION ROOM

*1035 A TRAIN ELECTRICAL PENE ROOM 171-041 UNIT 1 139’ 4160V SWGR AND CRDM MG SET ROOM 11

(RM 335,343 AND 346)1-042 UNIT 1 139’ HALLWAY (RM 319, 339, AND 345) 151-075 UNIT I AUX BLDG TO DIESEL BLDG A TRN CABLE TUNNEL 111-076 UNiT I AUX BLDG TO DIESEL BLDG B TRN CABLE TUNNEL 181-081 UNIT I TURBINE BUILDING BATTERY ROOM 381-094 COMBUSTIBLE STORAGE ROOM 167 391-SOl UNIT I NON RAD STAIRWELL 211-Sb UNIT I RAD STAIRWELL EASTSIDE TO 130’ EL 37

1-SVB2 SERVICE WATER VALVE BOX 2 191-SVB4 SERVICE WATER VALVE BOX

1-TB UNIT 1 TURBINE BUILDING 20

1-020 UNIT 1 NON-RAD HALLWAY 121 FT. FT

11

4 19

Page 3 of 4

FNP-1-EEP-3 STEAM GENERATOR TUBE RUPTURE Revision 26

Step Action/Expected Response Response NOT Obtained

I I I****************************** ************************************ ********************

CAUTION: To prevent release of radioactivity to the environment. RCS andruptured SG(s) pressures must be maintained less than the rupturedSG(s) atmospheric relief valve setpoint (1035 psig)

******* * *********** *** ** ** * ** ********************** ** ******** ******* ******************

31 [CAJ Control RCS parameters tominimize RCS to secondaryleakage.

31.1 Perform appropriate action(s)from table.

RUPTURED SG(s) LEVEL

Rising Falling Offscale high

. Raise • Raise • Raisecharging flow, charging flow, charging flow.

Less than25%(50%} • Reduce • Maintain RCS

RCS pressure. and rupturedSG(s) pressuresequal.

PR Between • Reduce • Turn on • Maintain RCSZ 25%{50%} RCS pressure. PRZR and rupturedR and 60%t60%} heaters. SG(s) pressures

equal.LE • Reduce • Turn on • Maintain RCSV Between RCS pressure. PRZR and rupturedE 60%t60%) and heaters. SG(s) pressuresL 73%f66%} • Reduce equal.

charging flow.

Greater than • Reduce • Turn on • Maintain RCS

73%t66%} charging flow. PRZR and rupturedheaters. SG(s) pressures

equal.

Step 31 continued on next page.

Page Completed

Page 39 of 54

10/10/11 12:16:08 FNP-0-EIP-8.0

12.0 Four Hour Independent Storage of Spent Fuel Reports 1OCFR72.75(b)(1).

• 11 NOT reported as a Declaration of an Emergency class or as a onehour non emergency report, THEN the NRCOC and Corporate DutyManager SHALL be notified by the Shift Manager or the FNP DutyManager as soon as practical and in ALL cases within four hours of theoccurrence of ANY of the following: (Figure 1 will be used):

• Notify the NRC resident in coniunction with this notification.

12.1 An action taken in an emergency that departs from a condition or aTechnical Specification contained in a license or certificate of complianceissued under this part, WHEN the action is immediately needed to protectthe public health and safety, AND NO action consistent with license orcertificate of compliance conditions or Technical Specifications that canprovide adequate or equivalent protection is immediately apparent.

13.0 Eight-Hour Reports 1OCFR5O.72(b)(3).

• IF not reported as a Declaration of an Emergency class or as a onehour or four hour non emergency report, THEN the NRCOC andCorporate Duty Manager SHALL be notified by the Shift Manager or theFNP Duty Manager as soon as practical and in JJ cases within eighthours of the occurrence of ANY of the following (Figure 1 will be used):

• Notify the NRC resident in coniunction with this notification.

13.1 ANY event or condition that results in the condition of the nuclear powerplant, including its principal safety barriers being seriously degraded.1 OCFR5O.72(b)(3)(ii)(A)

13.2 ANY event or condition that results in the nuclear power plant being in anunanalyzed condition that significantly degrades plant safety.1 OCFR5O.72(b)(3)(ii)(B)

-9- Version 108

10/10/11 12:16:08 FNP-0-EIP-8.0

13.3 ANY event or condition that results in valid actuation of any of the systemslisted below EXCEPT when the actuation results from and is part of a preplanned sequence during testing or reactor operation. (1OCFR5O.72(b)(3)(iv)(A)

(1) Reactor Protection System (RPS) including reactor trip. Actuation of theRPS when the reactor is critical is reportable under section 10 as a fourhour report.

(2) General containment isolation signals affecting containment isolationvalves in more than one system OR multiple main steam isolation valves(MSIV5).

(3) Emergency core cooling systems (ECCS) including: high-head and low-head injection systems.

(4) Auxiliary feedwater system.

(5) Containment heat removal and depressurization systems, includingcontainment spray NQ fan cooler systems.

(6) Emergency AC electrical power systems, including emergency dieselgenerators (EDG5).

13.4 ANY event or condition that at the time of discovery could have preventedthe fulfillment of the safety function of structures or systems that areneeded to: 1OCFR5O.72(b)(3)(v)

(1) Shut down the reactor and maintain it in a safe shutdown condition.

(2) Remove residual heat.

(3) Control the release of radioactive material.

(4) Mitigate the conseguences of an accident.

• Events covered in this step may include:

1 One or more procedural errors.

2. Equipment failures.

3 Discovery of design, analysis, fabrication, construction, and/orprocedural inadequacies.

• However, individual component failures need NOT be reported pursuantto this step IF redundant equipment in the same system was operableAND available to perform the required safety function.

-10- Version 108

1/6/2012 13:451 NATURAL CIRCULATION COOLDOT TO PREVENT REACTORRevision 19FNP-1-ESP-O.2

I VESSEL HEAD STEAM VOIDING

ATTACHMENT 3

CALCULATION FOR ADEQUATE AVAILABLE CST INVENTORY

RCS Cooldown Rate vs Minimum Required CST Level

10°F/hr

2rFIhr

50°F/hr

75°F/hrI

NOTE: These plots are based O4 an AFW flowrate of 350 gpm. An accurate assessmentof required CST inventory can be made using the following calculation and thatresult used to make a determinahon if a transihon to ESP-0..3 or 0.4 is required.

FPi’ci-350i1x AFW flowrate x6(÷ 125061 + 5.3 = Required cs’r Level

Cooldwn Rate ) J

Basis:a Current Cooldown Rate is in Degrees F per hour.• Current AFW flowrate is in Gallons per minute.• 35O’F = Temperature required to place RHR on-service.• 12500 = Gallons per foot of CST Level.a 60 = Minutes per hour.a 5,3 = Minimum CST level in feet before having to shift AFW suction to SW.

3&) 375 40.) 425 46) 475 &X) 525 5)RCS TCOLD

CORE OPERATING LIMITS REPORT, FNP UNIT 1 CYCLE 24 AUGUST 2010

Figure 1Rod Bank Insertion Limits versus Rated Thermal Power

Fully Withdrawn — 225 to 231 steps, inclusive

225 • rrmr

-

- (.552, 225)

200

(1,187).

175

BankC

150 1125

—(0,114)

100o

_________________________

liii

_______________________________________________

BankD

75

50

25

_____________

(.070, 0)

___________________________________________________________________

0 11111 I

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

Fraction of RATED THERMAL POWER

Fully Withdrawn shall be the condition where control rods are at a position within the interval 225 and231 steps withdrawn.

Note: The Rod Bank Insertion Limits are based on the control bank withdrawal sequence A, B, C, D anda control bank tip-to-tip distance of 128 steps.

Page 9 of 12

Rod Group Alignment Limits3.1.4

3.1 REACTIVITY CONTROL SYSTEMS

3.1.4 Rod Group Alignment Limits

LCO 3.1.4 All shutdown and control rods shall be OPERABLE, with all individualindicated rod positions within 12 steps of their group step counter demandposition.

APPLICABILITY: MODES I and 2.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more rod(s) A.1 .1 Verify SDM to be within 1 houruntrippable. the limits provided in the

COLR.

DR

A.1 .2 Initiate boration to 1 hourrestore SDM to withinlimit.

AN

A.2 Be in MODE 3. 6 hours

B. One rod not within B.1 Restore rod to within 1 houralignment limits, alignment limits.

oa

B.2.1 .1 Verify SDM to be within 1 hourthe limits provided in theCOLR.

DR

(continued)

Farley Units I and 2 3.1.4-1 Amendment No. 146 (Unit 1)Amendment No. 137 (Unit 2)

Rod Group Alignment Limits

ACTIONS

3.1.4


B. (continued) B.2.1.2 Initiate boration to 1 hourrestore SDM to withinlimit.

ANI2

B.2.2 Reduce THERMAL 2 hoursPOWER to RTP.

AN

B.2.3 Verify SDM to be within Once perthe limits provided in the 12 hoursCOLR.

AN

B.2.4 Perform SR 3.2.1.1. 72 hours

AN

B.2.5 Perform SR 3.2.2.1. 72 hours

AND

B.2.6 Re-evaluate safety 5 daysanalyses and confirmresults remain valid forduration of operationunder these conditions.

C. Required Action and C.1 Be in MODE 3. 6 hoursassociated CompletionTime of Condition B notmet.

Farley Units 1 and 2 3.1.4-2 Amendment No. 146 (Unit 1)Amendment No. 137 (Unit 2)

RCS Loops—MODE 43.4.6

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.6 RCS Loops—MODE 4

LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heatremoval (RHR) loops shall be OPERABLE, and one loop shall be inoperation.

NOTES1. All reactor coolant pumps (RCPs) and RHR pumps may not be in

operation for 2 hours per 8 hour period provided:

a. No operations are permitted that would cause reduction of theRCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F belowsaturation temperature.

2. No RCP shall be started with any RCS cold leg temperature 325°Funless:

a. The secondary side water temperature of each steamgenerator (SG) is <50°F above each of the RCS cold legtemperatures; or

b. The pressurizer water volume is less than 770 cubic feet (24%of wide range, cold, pressurizer level indication).

APPLICABILITY: MODE 4.

ACTIONS


A. One required RCS loop A.1 Initiate action to restore a Immediatelyinoperable, second loop to

OPERABLE status.AN

Two RHR loopsinoperable.


RCS Loops—MODE 4

ACTIONS

Farley Units 1 and 2 3.4.6-2

3.4.6

Amendment No. 147 (Unit 1)Amendment No. 138 (Unit 2)


B. One required RHR loop B.1 Be in MODE 5. 24 hoursinoperable.

AND

Two required RCS loopsinoperable.

C. Required RCS or RHR C.1 Suspend all operations Immediatelyloops inoperable, involving a reduction of

RCS boron concentration.OR

ANDNo RCS or RHR loop inoperation. C.2 Initiate action to restore Immediately

one loop to OPERABLEstatus and operation.

SURVEILLANCE_REQUIREMENTS

SURVEILLANCE FREQUENCY

SR 3.4.6.1 Verify one RHR or RCS loop is in operation. 12 hours

SR 3.4.6.2 Verify SG secondary side water levels are 75% 12 hours(wide range) for required RCS loops.

SR 3.46.3 Verify correct breaker alignment and indicated power 7 daysare available to the required pump that is not inoperation.

3.7 PLANT SYSTEMS

3.7.7 Component Cooling Water (CCW) System

CCW System3.7.7

LCO 3.7.7 Two CCW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS


A. One CCW train A.1 NOTEinoperable. Enter applicable

Conditions and RequiredActions of LCO 3.4.6,‘RCS Loops—MODE 4,”for residual heat removalloops made inoperable byCCW.

Restore CCW train to 72 hoursOPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hoursassociated CompletionTime of Condition A not ANflmet.

B.2 Be in MODE 5. 36 hours


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