ML102220137

DC Cook, 2010 Initial License Exam, Written Exam with Answers

ML102220137
Person / Time
Site:	Cook
Issue date:	06/21/2010
From:	Division of Reactor Safety III
To:	Indiana & Michigan Electric Co
References
Download: ML102220137 (198)

Text

Cook 2010 NRC Examination

1. 001 001/BOTH/OK/NEW/NEW/000008 AK2.01/2.7/2.7/H/2Given the following conditions on Unit 2:

• The crew is responding to a Reactor Trip and has transitioned to2-OHP-4023-ES-0.1, Reactor Trip Response, from 2-OHP-4023-E-0, Reactor Trip orSafety Injection.

• Following the transition the crew notes the following conditions: o RCS Pressure is 2100 psig and lowering o 2-NRV-152, indicating lights show an intermediate positionWhich ONE of the following describes the actions the operator should take to addressthese conditions: Close 2-NMO-152, PORV Block Valve, to stop discharge into the PRT.Open Pressurizer Spray valves to depressurize the RCS and limit the loss ofReactor Coolant.

Turn on all Pressurizer Heaters to maintain RCS pressure.Isolate air to containment to fail the Pressurizer PORV closed and stop the RCSmass loss.A.B.C.D.ANSWER: AA - CORRECT. Block valve is in series with the Pressurizer PORV. Closing theblock valve will isolate the leak and prevent further depressurizationof the RCS.B - INCORRECT. Mass loss is occurring. Need to isolate the leak by isolating thePORV. Mass loss will continue as long as there is pressure in theRCS and the PORV is leaking.C - INCORRECT. It is highly unlikely that the heaters can maintain pressure with astuck open Pressurizer PORV. Additionally, even if the heaterscould stabilize/raise pressure, the RCS leak would still persist.D - INCORRECT. NRV-152 has Nitrogen Backup supply, so isolating air tocontainment will not force the PORV to fail closed. Page 1 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00202/#RO-C-00202-E9

REFERENCE:

SOD-00202-001

KA - 000008 AK2.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)Knowledge of the interrelations between the Pressurizer Vapor Space Accident and thefollowing:ValvesRO - 2.7 SRO - 2.7 CFR - 41.7 / 45.7 KA Justification - Requires the knowledge of the interrelationship between the stuckopen PZR PORV and the block valve during a vapor space leak. Oneblock valve is in series with each PORV to allow isolation of a leaking,stuck open PORV.Original Question # - NEW Original Question KA - NEW Page 2 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

2. 002 012/BOTH/OK/DIRECT/RO24 AUDIT 76-12/000009 EA2.38/3.9/4.3/H/3Given the following conditions on Unit 2:

• A reactor trip and safety injection have occurred.

• The crew is responding to a Small Break Loss of Coolant Accident (LOCA).

• All Reactor Coolant Pumps are tripped.

• The crew is depressurizing the Reactor Coolant System (RCS) in accordance withStep 13 of 2-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization.

• A PORV is being used to depressurize the RCS.As the depressurization occurs, which one of the following describes the expectedtrend of pressurizer level and the adverse operating condition that may initially occur asa result? Lowering Pressurizer Level; Uncovering Pressurizer heaters.

Rising Pressurizer Level; Water solid conditions in the RCS and Pressurizer.

Rising Pressurizer Level; Upper head region voiding may occur.

Lowering Pressurizer Level; Upper head region voiding may occur. A.B.C.D.ANSWER: CA - INCORRECT. Pressurizer will rise during depressurization versus lower. It isplausible that if the operator does not understand this concept andthey believe Pressurizer level will drop that the heaters wouldbecome uncovered which is an undesirable condition. B - INCORRECT. Although it is correct that Pressurizer level will rise and it is plausiblethat eventually the Pressurizer would go solid, that this would notINITIALLY occur.C - CORRECT. The caution prior to commencing depressurization in ES-1.2 to refillthe Pressurizer, states that a head void may occur as indicated by arising Pressurizer level as water in transferred from the RCS to thePressurizer.D - INCORRECT. Pressurizer level will rise vs. lower. The reason is plausible andtests whether the student correctly understands the importantconcept. Page 3 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP09/#34 & 36

REFERENCE:

02-OHP-4023-ES-1.2 step 13 Background 12-OHP-4023-ES-1.2 -EOP Step #: 13 N1 ERGStep #: 11 C1KA - 000009 EA2.38 Small Break LOCAAbility to determine and interpret the following as they apply to a small break LOCA:Existence of head bubble RO - 3.9 SRO - 4.3 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires ability to determine the indications that represent a bubbleforming in the reactor vessel head during post LOCA cooldown anddepressurization.Original Question # - Cook RO24 Audit - 076-12 Original Question KA - EPE009.EA2.04 Page 4 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

3. 003 002/BOTH/OK/MODIFIED/WATTSBAR - MAY2009/000011 EA1.05/4.3/3.9/F/3Given the following conditions on Unit 1:

• A Large Break LOCA has occured.

• Safety Injection has actuated.Which ONE of the following describes:

1) how the Centrifugal Charging Pump suction swaps to the RWST when a SafetyInjection is initiated -AND-

2) how the CHARGING PUMP SUCTION swapover to the containment sump iscompleted in accordance with 1-OHP-4023-ES-1.3, Transfer to Cold LegRecirculation?Note: VCT valves = QMO-451/QMO-452 RWST valves = IMO-910/IMO-9111) The VCT valves will start to close AFTER one of the RWST valves have traveledto the full open position.2) The RWST valves will AUTOMATICALLY close after IMO-340, Charging PpSuction from East RHR Hx has been opened. 1) The VCT valves will start to close AFTER one of the RWST valves have traveledto the full open position.2) The RWST valves will be MANUALLY closed after IMO-340, Charging PpSuction from East RHR Hx has been opened.1) The VCT valves will start to close AS SOON AS one of the RWST valves start toopen.2) The RWST valves will AUTOMATICALLY close after IMO-340, Charging PpSuction from East RHR Hx has been opened.1) The VCT valves will start to close AS SOON AS one of the RWST valves start toopen.2) The RWST valves will be MANUALLY closed after IMO-340, Charging PpSuction from East RHR Hx has been opened.A.B.C.D. Page 5 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. Plausible, since the valves from the VCT starting to close when thevalves from the RWST get fully open is correct.B - CORRECT. The valves from the VCT will start to close when the valves from theRWST get fully open and after transfer to the containment sump thevalves are placed in the A-Auto position.C - INCORRECT. Plausible, because the ECCS valves used to swapover to thecontainment sump do travel together during the transfer asdescribed in column (1). Handswitches positions is plausible sincethat is the normal position for the switches.D - INCORRECT. Plausible, because the ECCS valves used to swapover to thecontainment sump do travel together during the transfer asdescribed in column (1) and the handswitches being in A Autoposition is correct.LESSON PLAN/OBJ: RO-C-00300-E13, RO-C-EOP09/#22

REFERENCE:

RO-C-00300 pg. 50, OHP-4023-ES-1.3, OP-2-98271KA - 000011 EA1.05 Large Break LOCAAbility to operate and/or monitor the following as they apply to a Large Break LOCA:Manual and/or automatic transfer of suction of charging pumps to borated sourceRO - 4.3 SRO - 3.9 CFR - 41.7 / 45.5 / 45.6

KA Justification - Stem conditions include a large break LOCA, and the question testsunderstanding of charging pump suction alignment during the event.Original Question # - WattsBarMay2009 Original Question KA - 011 EA1.05 Page 6 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

4. 004 006/BOTH/OK/DIRECT-REPEAT/NRC EXAM 2007-6/000015 AK1.04/2.9/3.1/H/3Given the following conditions on Unit 2:

• Unit tripped from 29% power.

• RCP 21 breaker tripped open when the busses swapped.Which one of the following describes the response of Thot and Tcold in Loop 1?Tcold rises to approximately equal Thot.

Thot lowers to approximately equal Tcold.

Tcold lowers, Thot remains approximately stable.

Thot rises, Tcold remains approximately stable.A.B.C.D.ANSWER: B A - INCORRECT. Tcold remains approximately the same, at low power nearsaturation for SG. Plausible due to lack of forced circulationprevents S/G from transferring heat to Main Steam Headerassuming normal RCS flow direction, so RCS loop heats up to ThotB - CORRECT. Loss of RCS flow in 1 loop, reverse flow in that loop will cause Thotto drop (no more forced circulation in that loop) to the Tcold value orslightly below. C - INCORRECT. Tcold remains approximately the same, at low power nearsaturation for SG. Thot lowers since the core exit flow is not forcedinto the loop. Plausible due to reverse flow in RCS loop, from loss ofRCP, allows Tcold to enter Steam Generator which removedadditional energy thus lowering Tcold additionally.D - INCORRECT. Thot lowers since the core exit flow is not forced into the loop.Plausible due to the Turbine Steam Demand did not change sooverall Reactor Power will remain constant so the core will producethe same power from the remaining Steam Generators. Assumingthe student believes normal RCS flowpath (at a reduced rate withno forced circulation) and if Tcold remains the same the onlymethod to increase power from the remaining Steam Generators isto increase Thot. Page 7 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-TRANS4\4A.4

REFERENCE:

RO-C-TRANS4, RCS Loop Flow Transients KA - 000015 AK1.04 017 Reactor Coolant Pump (RCP) Malfunctions Knowledge of the operational implications of the following concepts as they apply toReactor Coolant Pump Malfunctions:

Basic steady state thermodynamic relationship between RCS loops and S/Gs resultingfrom unbalanced RCS flow RO - 2.9 SRO - 3.1 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires knowledge of the thermodynamic relationship of thetemperature response of an idle loop following a malfunction of anRCP causing the RCP to trip.Original Question # - NRC EXAM 2007-6, INPO # 23126 Salem Unit 1 - 11/4/2002Original Question KA - 015.AA1.09 Page 8 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

5. 005 024/BOTH/OK/DIRECT/RO26 AUDIT-24/000025 AA1.09/3.2/3.1/F/2Given the following conditions on Unit 2:

• Reactor Coolant System (RCS) is in mid-loop condition* The following indications are fluctuating on the running Residual Heat Removal(RHR): amps, flow, and discharge pressureWhich ONE of the following statements is correct regarding the standby RHR pump? The standby RHR Pump should:NOT be immediately started because air entrainment could cause a loss of bothRHR trains.

be immediately started because following a loss of RHR flow, an RCSpressurization may occur precluding gravity feed makeup. be immediately started because under certain loss of RHR conditions, coreuncovery or core voiding can occur within 15 to 20 minutes.NOT be immediately started because starting an idle RHR pump under mid-loopconditions could cause an unacceptable reduction in reactor shutdown margin.A.B.C.D.ANSWER: AA - CORRECT. The ARG provides a clear guidance which includes industryexperience of why operation of an RHR pump operating with airentrapment should be evaluated because it could lead to pumpdamage. Starting the other RHR pump could transfer the problem tothe other pump leading to a complete loss of the system. B - INCORRECT. Plausible because gravity makeup is a required action but startingthe RHR pump is not an option immediately. C - INCORRECT. While it is plausible that core uncovery or voiding could occur in arelatively short period of time, it is not correct that the RHR pump bestarted in this plant condition. D - INCORRECT. Although it is correct that the pump should not be started it is notcorrect that SDM would be affected in this condition since it wasalready verified procedurally to meet TS plant conditions. Plausiblesince SDM is a concern during cooldown & RHR flow helps ensureproper mixing & SDM. Page 9 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D16/#RO-C-AOP0430412-E3

REFERENCE:

OHP-4022-017-001 KA - 000025 AA1.09 Loss of Residual Heat Removal System (RHRS)Ability to operate and/or monitor the following as they apply to the Loss of ResidualHeat Removal System:

LPI pump switches, ammeter, discharge pressure gauge, flow meter, and indicatorsRO - 3.2 SRO - 3.1 CFR - 41.7 / 45.5 / 45.6 KA Justification - Requires the ability to monitor RHR pumps amps and flow duringmid-loop operations and determine based on the conditions whatactions should be taken due to the abnormal conditions.Original Question # - RO26 AUDIT-24 Original Question KA - APE035 AK2.02 Page 10 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

6. 006 001/BOTH/OK/NEW/NEW/000027 2.2.40/3.4/4.7/H/3Given the following conditions on Unit 1:

• Unit is operating at 100% power

• 1-NPP-151, PZR Press Channel 1, fails high

• The operator has taken manual control of Pressurizer pressure control and stabilizedpressure at 2085 psig.Following completion of the procedure for response to a malfunction of a pressurizerpressure instrument, what will be the status of the CVCS/Charging system?The ______ CCP will be INOPERABLE with the associated emergency leakoff valvedeenergized in the ______ position.East; open East; closed West; open West; closedA.B.C.D.ANSWER: AA - CORRECT. Channel 1 affects QMO-225 for the East CCP. The Emergencyleakoff is racked out in the open position to ensure minimum flow inthe event of an SI. B - INCORRECT. The Emergency leakoff is racked out in the OPEN position to ensureminimum flow in the event of an SI. C - INCORRECT. Channel 1 affects QMO-225 for the East CCP, QMO-226 is for theWest CCP from channel 2. D - INCORRECT. Channel 1 affects QMO-225 for the East CCP, QMO-226 is for theWest CCP from channel 2. The Emergency leakoff is racked out inthe OPEN position to ensure minimum flow in the event of an SI. Page 11 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D6/#RO-C-AOP0330412-E3

REFERENCE:

1-OHP-4022-013-009, Pressurizer Pressure InstrumentMalfunctionKA - 000027 2.2.40 Pressurizer Pressure Control (PZR PCS) Malfunction Equipment ControlAbility to apply Technical Specifications for a system.RO - 3.4 SRO - 4.7 CFR - 41.10 / 43.2 / 43.5 / 45.3 KA Justification - Requires the operator to know the implications of a failedPressurizer Pressure instrument on the operability of the CCPs andthe required procedural actions to address TS concerns.Original Question # - NEW Original Question KA - NEW Page 12 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

7. 007 001/BOTH/OK/NEW/NEW/000029 EK2.06/2.9/3.1/H/3Given the following conditions on Unit 2:

• The operators are implementing 2-OHP-4023-FR-S.1, Response to Nuclear PowerGeneration/ATWS

• Both Reactor Trip Breakers remain closed.

• The operators have completed steps of 2-OHP-4023-FR-S.1 through opening theMG Set output Breakers to shutdown the reactor.Which ONE of the following describes the impact and potential consequences of theReactor Trip Breakers remaining closed?main steam line isolation signal will NOT occur to prevent excessive reactivityduring the trip due to rapid RCS cooldown.

feedwater isolation signal will NOT actuate to prevent excessive reactor coolantsystem cooldown from the overfeeding of the steam generators.main generator trip signal will NOT be generated preventing transfer of busses toreserve feed.

feedwater flow conservation signal will NOT occur to ensure equal distribution ofwater to the steam generators.A.B.C.D.ANSWER: BA - INCORRECT. Main steam line isolation does not depend on the status of theReactor Trip Breakers. This signal will occur as designed onappropriate containment pressure or SG parameters.B - CORRECT. P-4 (Rx Trip Breaker Position) feeds the feedwater isolation signal. Either breaker being open will cause an isolation of flow to the SGs. However with neither breaker open, all FMOs and FRVs will remainopen and Main Feedpumps will not trip. This will cause excessiveflow to the SGs, which could lead to an overcooling of the RCS.C - INCORRECT. Main Generator trip is caused by a sensed Turbine Trip (All TurbineSteam Valves Closed). There is no interlock between the ReactorTrip Breakers and the generator trip. D - INCORRECT. Flow conservation is often confused with feedwater isolation. Flowconservation will start the Aux Feedwater Pumps and open the AFWPump discharge valves. There is no interlock between the ReactorTrip Breakers and Flow Conservation. Page 13 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01100/#4

REFERENCE:

RO-C-01100 KA - 000029 EK2.06Anticipated Transient Without Scram (ATWS)Knowledge of the interrelations between the ATWS and the following:Breakers, relays, and disconnects RO - 2.9 SRO - 3.1 CFR - 41.7 / 45.7 KA Justification - Requires knowledge of the interrelationship between an ATWS (RxTrip Breakers NOT open) an the relays associated with feedwaterisolation.

Original Question # - NEW Original Question KA - NEW Page 14 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

8. 008 002/BOTH/OK/NEW/NEW/000051 AA1.04/2.5/2.5/H/3Given the following conditions on Unit 1:

• The Unit is operating at 75% power with all systems in automatic.

• Main Turbine DCS is in "MW IN" in preparation for turbine valve testing.

• Condenser vacuum is lowering.Assuming no action has been taken by the crew, which ONE of the following describesthe response of the rod control system to this event? Control rods will automatically: insert due to the rise in Tavg from the rise in steam flow.insert due to the rise in Tavg from the lowering in steam flow.withdraw due to the drop in Tavg from the rise in steam flow.withdraw due to the a drop in Tavg from the lowering steam flow.A.B.C.D.ANSWER: CA - INCORRECT. Controls rods will withdraw. See Answer C. Plausible since duringnormal power escalations Tavg and Steam flow both rise.B - INCORRECT. Controls rods will withdraw. See Answer C. Plausible due to designof system to insert rods on rising Tavg.C - CORRECT. With the Main Turbine in "MW IN" control, the turbine valves areallowed to reposition to try to maintain Main Generator Load. Asvacuum lowers, the turbine will become less efficient, causing moresteam flow for the same MW output. As steam flow rises, RCS Tavewill lower below Tref. With rods in AUTO, the rods will withdraw tominimize the Tave-Tref deviation. D - INCORRECT. Steam flow will rise. See Answer C. Plausible since during normalpower reductions Tavg and Steam flow both lower. Page 15 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05003/#RO-C-05003-E16, RO-C-01200/#RO-C-01200-E7

REFERENCE:

SOD-01200-003, TS3000 Fig. 6-1 KA - 000051 AA1.04 Loss of Condenser VacuumAbility to operate and/or monitor the following as they apply to the Loss of CondenserVacuum:Rod position RO - 2.5 SRO - 2.5 CFR - 41.7 / 45.5 / 45.6 KA Justification - Requires the ability to determine the proper operation of the controlrod system during a transient caused by a Loss of CondenserVacuum.Original Question # - NEW Original Question KA - NEW Page 16 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

9. 009 004/BOTH/OK/DIRECT/RO23 AUDIT-066-6/000054 AK3.04/4.4/4.6/F/3Given the following conditions on Unit 1:

• Unit 1 was operating at 100% power when a condensate system transient causedboth Main FW pumps to trip.

• The turbine and reactor failed to trip automatically. In accordance with the immediate actions of 1-OHP-4023-FR-S.1, Response toNuclear Power Generation/ATWS, the operators will:

1. Manually trip the Reactor, if it fails to trip insert control rods. 2. Manually actuate AMSAC.

3. Manually trip the Turbine, if it fails to trip, then runback the turbine. Which ONE of the following describes the bases for these immediate actions in1-OHP-4023-FR-S.1?

The safeguards systems are designed assuming that the only heat being added to theRCS is from _____________ . For an ATWS event with a loss of normal feedwater, aTurbine trip within 30 seconds will _____________________ .fission product decay and RCP heat; prevent challenging the Pressurizer PORV's.

fission product decay and RCP heat; maintain S/G inventory.

5% power; maintain S/G inventory.

5% power; prevent challenging the Pressurizer PORV'sA.B.C.D. Page 17 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. Turbine is tripped to maintain SG water inventory. Plausible due tofirst portion of answer is correct and second portion of answer isrealistic possibility due to imbalance of energies from the Primary tothe Secondary plants.B - CORRECT. Per FR-S.1 Background Document, the assumed heat generation isfrom decay heat and RCP heat. Turbine is tripped to maintain SGwater inventory.C - INCORRECT. Per FR-S.1 Background Document, the assumed heat generation isfrom decay heat and RCP heat. Plausible due to commonacceptance of 5% power being limitation of ESF system.D - INCORRECT. Per FR-S.1 Background Document, the assumed heat generation isfrom decay heat and RCP heat. Turbine is tripped to maintain SGwater inventory. Plausible due to common acceptance of 5% powerbeing limitation of ESF system and second portion of answer isrealistic possibility due to imbalance of energies from the Primary tothe Secondary plants.LESSON PLAN/OBJ: RO-C-EOP04/#15

REFERENCE:

01-OHP-4023-FR-S.1, Response to Nuclear PowerGeneration/ATWS, Step 1-3 Background, RO-C-EOP04 pg. 11KA - 000054 AK3.04 Loss of Main Feedwater (MFW)

Knowledge of the reasons for the following responses as they apply to the Loss of MainFeedwater (MFW):Actions contained in EOPs for loss of MFWRO - 4.4 SRO - 4.6 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Question tests knowledge of the reasons for FR S.1 Steps andactions associated with a Loss of FW ATWS.Original Question # - RO23 AUDIT-066-6 Original Question KA - EPE 029 EK1.01 Page 18 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

10. 010 002/BOTH/OK/DIRECT/PNTBEACH2002/000055 EK3.02/4.3/4.6/F/3Given the following conditions on Unit 2:

• A loss of all AC Power occurred due to severe weather conditions and failure ofemergency diesel generators to start and supply safeguard buses.

• The operating crew is carrying out actions of 2-OHP-4023-ECA-0.0, Loss of All ACPower.* The operators are at a point where they are to commence cooldown anddepressurization of the steam generators to 190 psig.Based on these conditions, which ONE of the following statements describes thereason why a secondary depressurization is directed?To prevent a challenge to the Core Cooling Safety Function Status Tree whichis being monitored for implementation.

To remove stored energy in the steam generators to limit the potential ofchallenging RCS integrity.

To remove available energy in the steam generators and thus minimizing anychallenges to the containment structure if a Faulted S/G were to occur.To minimize RCS inventory loss through the RCP seals, which maximizes time tocore uncovery.A.B.C.D.ANSWER: DA - INCORRECT. Status Trees are monitored for information only. ECA 0.0 has builtinto the mitigating strategy to manage all actions addressed byFRP's as well as the FRP's are for information use only in ECA-0.0.B - INCORRECT. SG PORVS should limit SG pressures but the primary concern isunrecoverable loss of RCS inventory. Plausible due to reduction inrisk of having an RCS integrity challenge which is not a concern inthis eventC - INCORRECT. SG PORVS should limit SG pressures but the primary concern isunrecoverable loss of RCS inventory. Plausible due to reduction inrisk of having a challenge to the Containment structure due to areduction in the available energy inside ContainmentD - CORRECT. The primary concern is a loss of RCS inventory with no way torecover level. This could lead to core uncovery. The SGs aredepressurized to lower RCS temperature and pressure to slow theloss of inventory. Page 19 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP14/#5 & 9

REFERENCE:

2-OHP-4023-ECA-0.0, Step 19 and Note 1 BackgroundKA - 000055 EK3.02 Loss of Offsite and Onsite Power (Station Blackout)

Knowledge of the reasons for the following responses as they apply to the StationBlackout:Actions contained in EOP for loss of offsite and onsite powerRO - 4.3 SRO - 4.6 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Question tests knowledge of the reason the SGs are depressurized inECA-0.0.Original Question # - INPO # 20572 Point Beach 1 - 2/2/2002Original Question KA - 055.EK3.02 Page 20 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

11. 011 002/BOTH/OK/DIRECT-REPEAT/NRC EXAM 2007-041/000056 AK1.03/3.1/3.4/H/3Given the following conditions on Unit 2:

• Unit has just tripped due to a Loss of Offsite power.

• Both EDGs started and energized the required loads.

• All equipment responded as designed.The following conditions exist:

• Containment parameters are normal

• Average core exit thermocouple (CET) temperature is stable.Which ONE of the following combination of RCS pressure and average CETtemperature verifies the MINIMUM required subcooling to AVOID Safety Injection per2-OHP-4023-ES-0.2, Natural Circulation Cooldown?

Reference Provided: Steam Tables600 psig, 590 o F500 psig, 460 o F450 psig, 430 o F375 psig, 400 o FA.B.C.D.ANSWER: D A - INCORRECT. RCS is saturated - Tsat is 489 o FB - INCORRECT. RCS is 10 oF subcooled - Tsat is 470 o FC - INCORRECT. RCS is 30 oF subcooled - Tsat is 460 o FD - CORRECT. 2-OHP-4023-ES-0.2 , Foldout Page (FOP) needs >40 oF ofsubcooling, or requires that a SI be actuated. Tsat fo 400 psia (375psig + 15 psi) is 444.6 oF. Based on the conditions provided, 44.6 o Fof subcooling exists, exceeding the 40 oF requirement. Page 21 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP03/#18

REFERENCE:

2-OHP-4023-ES-0.2, Natural Circulation Cooldown FoldoutPage, RO-C-EOP03, Plant Trips, Diagnosing Accidents, NaturalCirculation Cooldown, E-0 Series EOPs, and BackgroundInformation pg. 89Reference Provided: Steam Tables KA - 000056 AK1.03 Loss of Offsite Power Knowledge of the operational implications of the following concepts as they apply toLoss of Offsite Power:

Definition of subcooling: use of steam tables to determine itRO - 3.1 SRO - 3.4 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires the use of the steam table to determine subcooling during aloss of offsite power event on natural circ cooldown.Original Question # - 21521-KEWAUNEE02, NRC2004-41-2 Page 22 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

12. 012 003/BOTH/OK/MODIFIED/RO25 AUDIT-12/APE057 AA1.02/3.8/3.7/H/4Given the following conditions on Unit 1:

• Unit is operating at 60% power

• Pressurizer Level Control is in MANUAL

• Pressurizer LEVEL CTRL SELECTOR switch is in the Channel 2-3 position

• CRID 3 power supply fails Assuming no operator action, which ONE of the following statements describes theeffect of this failure on the CVCS and PZR level control system? QRV-251, CCP Disch Flow Control fails OPEN Letdown Isolates Actual Pressurizer Level Rises QRV-251, CCP Disch Flow Control fails CLOSED Letdown Isolates Actual Pressurizer Level Rises QRV-251, CCP Disch Flow Control fails CLOSED QRV-200, Charging Header Pressure Control Valve fails OPENActual Pressurizer Level Lowers QRV-251, CCP Disch Flow Control fails OPEN QRV-200, Charging Header Pressure Control Valve fails CLOSEDActual Pressurizer Level LowersA.B.C.D.ANSWER: AA - CORRECT. Loss CRID 3 causes a Loss of PZR level Channel NLP-153 whichwill result in an indicated low Pressurizer level. This will cause thePZR Level control to Close QRV-112 and Open QRV-251. (Notethat QRV-251 will also fail Open & QRV-112 will Close from Loss ofCRID 3 per OHP-4021-082-008 Table 3. i.e. - either NLP-153 failingor loss of power causes same effects)B - INCORRECT. Channel 3 will fail low. QRV-251 will fail open.C - INCORRECT. Channel 3 will fail low. QRV-251 will fail open. Level will rise.D - INCORRECT. Channel 3 will fail low. Level will rise. Page 23 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00300/#RO-C-00300-E9, RO-C-00202/#RO-C-00202-E14

REFERENCE:

SOD-00202-003, Pressurizer Level Control;1-OHP-4021-082-008Modified: Changed to CRID 3 from 4 (Answer D to A) & Changed Distractor C.KA - 000057 AA1.02 Loss of Vital AC Electrical Instrument BusAbility to operate and/or monitor the following as they apply to the Loss of Vital ACInstrument Bus:

Manual control of PZR level RO - 3.8 SRO - 3.7 CFR - 41.7 / 45.5 / 45.6 KA Justification - Requires the ability to monitor the response of pressurizer levelcontrol to a failure of a CRID (AC Power) while operating in manualcontrol. Student must be able to identify correct failure position formultiple valves failing in different directions and evaluate the impacton the Pressurizer level.Original Question # - Audit RO22-BOTH-59 (#52), RO25 AUDIT-12Original Question KA - KA SYS 004K2.06, APE057 AA1.02 Page 24 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

13. 013 001/BOTH/OK/NEW/NEW/000058 2.4.20/3.8/4.3/F/2Which ONE of the following describes the reason for disabling AMSAC, de-energizingDCS Inverters, and stopping all DC powered Turbine Oil Pumps in2-OHP-4023-ECA-0.0, Loss of All AC Power ?Allow turbine oil systems to be de-energized and drain to the main turbine lube oiltank.

Prevent inadvertent actuation of control systems and auto start of pumps.Extend the DC battery life for N Train and BOP batteries.Limit overheating of cabinets and pump motor overload.A.B.C.D.ANSWER: CA - INCORRECT. Plausible since oil systems will drain back to the main turbine LOtank and student may think this will help reduce risk of fire.B - INCORRECT. Plausible due to concerns for smart shorts of equipment duringemergencies.C - CORRECT. ECA-0.0, Step 17 Note states that DC Loads are shed to extend thelife of the DC batteries associated with the loads..D - INCORRECT. Plausible since other control room cabinets are opened in ECA-0.0to preclude overheating of instrumentation. LESSON PLAN/OBJ: RO-C-EOP14/#11

REFERENCE:

2-OHP-4023-ECA-0.0, Step 17 KA - 000058 2.4.20 Loss of DC Power Emergency Procedures/Plan Knowledge of operational implications of EOP warnings, cautions, and notes.RO - 3.8 SRO - 4.3 CFR - 41.10 / 43.5 / 45.13 KA Justification - Question tests operational implication (plant impact) of note in loss ofDC procedure.Original Question # - NEW Original Question KA - NEW Page 25 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

14. 014 004/BOTH/OK/DIRECT/CALLAWAY2007-59/000059 AK1.02/2.6/3.2/F/3Given the following conditions:

• An accidental spill of the Monitor Tank has occurred in the Aux Building.

• Radiation levels in the area of the spill are 40 mRem per hour at 30 cm.

• Contamination levels based on smear on the floor around the tank are 1.2 x10 4dpm/100 cm 2 beta-gamma.Which ONE of the following describes how the area will be posted in accordance withPMI-6010, Radiation Protection Plan?Radiation Area ONLY.Contamination Area ONLY.Radiation Area AND Contamination Area.High Radiation Area AND Contamination Area.A.B.C.D.ANSWER: CA - INCORRECT. Greater than >1000 dpm/100 cm 2 is a contamination area. Plausiblebecause this answer is only partially correct.B - INCORRECT. Greater than 100 mrem/ hr is a high radiation area. Plausiblebecause this answer is only partially correct.C - CORRECT. This area should be posted as a radiation area (>5 mrem in 1 hourand <100 mrem/hr) and a contamination area (>1000 dpm/100 cm 2).D - INCORRECT. Greater than 100 mrem/ hr is a high radiation area. Area is <100mrem/ hr. Page 26 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-RP02/#4 & 7

REFERENCE:

PMI-6010, Section 4.7, RO-C-RP02 KA - 000059 AK1.02Accidental Liquid Radwaste ReleaseKnowledge of the operational implications of the following concepts as they apply toAccidental Liquid Radwaste Release:Biological effects on humans of various types of radiation, exposure levels that areacceptable for nuclear power plant personnel, and the units used for radiation-intensitymeasurements and for radiation exposure levels RO - 2.6 SRO - 3.2 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires knowledge of the biological implications to workers relatedto the and postings requirements for an accidental release (spill ofMonitor Tank contents) in the Aux Building. In addition, requiresknowledge of the units and acceptable levels of radiation/contamination for these conditions. Original Question # - CALLAWAY2007-59 Original Question KA - 000059 AK1.02 Page 27 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

15. 015 005/BOTH/OK/NEW/NEW/000061 2.2.36/3.1/4.2/H/3Given the following condition:

• 1-MRV-213, Unit 1 SG11 PORV, was locally isolated due to excessive leakby.Which ONE of the following describes the status of the SG11 PORV RadiationMonitors and the required actions, if any?

Reference Provided: U1 TRM 8.3.8, Radiation Monitoring InstrumentationChannel MRA-1601 is Inoperable. Perform surveys of the SG PORV area every 24hours. Implement administrative controls to initiate an alternate method ofmonitoring.

Channel MRA-1601 is Inoperable. Restore to Operable Status within 7 days.Channel MRA-1601 is still Operable provided it is indicating approximately thesame as MRA-1602 (acceptable channel check).

Channel MRA-1601 is Inoperable. Implement administrative controls to initiate analternate method of monitoring within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A.B.C.D.ANSWER: BA - INCORRECT. These are actions B and E. B - CORRECT. TRO 8.3.8 requires 1 Channel per loop to be Operable. The SGPORV Monitor is required to be declared Inoperable if the PORV isclosed or isolated. Declaring the SG PORV Radiation monitorinoperable requires that Function 2.b Action C be applied.C - INCORRECT. The SG PORV Radiation Monitor is declared inoperable if thePORV Is isolated.D - INCORRECT. This is action E which is only required if an accident involving arelease is in progress. The monitor must be restored within 7 days. Page 28 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05103/#4

REFERENCE:

TRM 8.3.8, RO-C-05103 pg. 13, 26-27 Reference Provided: U1 TRM 8.3.8, Radiation Monitoring InstrumentationKA - 000061 2.2.36Area Radiation Monitoring (ARM) System AlarmsEquipment ControlAbility to analyze the effect of maintenance activities, such as degraded power sources,on the status of limiting conditions for operations.RO - 3.1 SRO - 4.2 CFR - 41.10 / 43.2 / 45.13

KA Justification - Question tests ability to determine the status of LCO on PORV RadMonitors due to maintenance activity on the PORV.Original Question # - NEW Original Question KA - NEW Page 29 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

16. 016 008/BOTH/OK/MODIFIED/NRC EXAM 2004-130-3/000065 2.4.8/3.8/4.5/F/3Given the following conditions on Unit 2:

• Unit was operating at 100% power when a malfunction of the Control Air systemoccurs.

• The Control Air header rapidly depressurizes and cannot be restored.Which ONE of the following describes the correct operator response?Immediately trip the Reactor and implement:2-OHP-4023-E-0, Reactor Trip or Safety Injection. 2-OHP-4022-064-002, Loss Of Control Air Recovery, may be performedconcurrently after transitioning to 2-OHP-4023-ES-0.1, Reactor Trip Response.2-OHP-4023-E-0, Reactor Trip or Safety Injection. 2-OHP-4022-064-002, Loss Of Control Air Recovery, is NOT needed since the EOPnetwork may be performed without reliance on Control Air.2-OHP-4023-E-0, Reactor Trip or Safety Injection.

2-OHP-4022-064-002, Loss Of Control Air Recovery, may NOT be performed untilcompletion of 2-OHP-4023-ES-0.1, Reactor Trip Response.2-OHP-4022-064-002, Loss Of Control Air Recovery, until restoration of Control Airfrom any source.

Perform 2-OHP-4023-E-0, Reactor Trip or Safety Injection steps as time allows.A.B.C.D.ANSWER: A A - CORRECT. OHI-4023, Abnormal/Emergency Procedure User's Guide allowsAbnormal Procedures to be implemented concurrently withNon-Accident (ES-0.1, 0.2 or 0.3) Emergency Procedures after theimmediate actions are complete at US discretion.B - INCORRECT. Performance of 02-OHP-4023-E-0 is required upon the reactor trip,but the operators must continue to perform 02-OHP-4022-064-002 toaddress the loss of Control Air.C - INCORRECT. User's Guide allows Abnormal Procedures to be implementedconcurrently with Non-Accident (ES-0.1, 0.2 or 0.3) EmergencyProcedures.D - INCORRECT. The Unit Supervisor should direct action of 02-OHP-4023-E-0, first,NOT as time allows. 02-OHP-4023-E-0 actions take priority over02-OHP-4022-064-002. Page 30 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP01/#25

REFERENCE:

OHI-4023 Abnormal/Emergency Procedure User's Guide,Attachment 2, Step 3.0.KA - 000065 2.4.8 Loss of Instrument Air Emergency Procedures/Plan Knowledge of how abnormal operating procedures are used in conjunction with EOPs.RO - 3.8 SRO - 4.5 CFR - 41.10 / 43.5 / 45.13 K/A Justification - Requires the knowledge of how to use Loss of Control Air (AbnormalOperating Procedure) in conjunction with the Emergency OperatingProcedures.Question Source: NRC EXAM 2004-130-3 Page 31 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

17. 017 001/BOTH/OK/DIRECT/CM-AS17-41643/000067 AK1.02/3.1/3.9/H/3Given the following conditions:

• A Main Transformer fire has occurred on Unit 1.

• The Unit 1 Reactor and Main Turbine/Generator have been tripped.

• The Turbine AEO has reported that the Main Transformer deluge system hasactuated.*The Outside Tour AEO has reported that all three fire water pumps are running.The reported status of the Fire Water System is ____________ for this event. TheMain Transformer deluge system ___________ expected to automatically actuate oncethe Main Generator is tripped, ___________ fire water pumps are expected to berunning.abnormal; is; but only 2 abnormal; is not; but 3 normal; is; and 3 abnormal; is not; and only 2A.B.C.D.ANSWER: AA - CORRECT. For any given actuation of the fire system, the maximum number ofpumps running should be 2. All three pumps running are anindication of a piping rupture. The deluge valve will notautomatically actuate until the main transformer is de-energized.B - INCORRECT. Deluge will automatically actuate. Only 2 pumps should be running.Plausible due to a fire requires activation of the deluge to put out aswell as student must know all 3 pumps are not expected to berunning.C - INCORRECT. Condition is not normal. Only 2 pumps should start oncetransformer is de-energized. Plausible due to student must know all3 pumps are not expected to be running.D - INCORRECT. Deluge will automatically actuate. Plausible due to a fire requiresactivation of the deluge to put out as well as student must know only2 pumps are expected to be running. Page 32 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AS17/#3

REFERENCE:

1-OHP-4024-101, ANNUNCIATOR #101 RESPONSE: PLANTFIRE SYSTEM, Drop 2KA - 000067 AK1.02 Plant Fire on Site Knowledge of the operational implications of the following concepts as they apply toPlant Fire on Site:

Fire fighting RO - 3.1 SRO - 3.9 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires an operational knowledge of the normal configuration of firefighting equipment during a normal actuation of the fire protectionsystem.Original Question # - CM-AS17 - 41643 Original Question KA - 086 A3.02 Page 33 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

18. 018 004/BOTH/OK/NEW/NEW/000068 AK2.07/3.3/3.4/F/3Given the following conditions on Unit 2:

• Reactor Power is at 100% when a fire occurs in the Control Room Cable Vault.

• A large amount of smoke accumulates in the Control Room,

• The Shift Manager determines that the main control room must be evacuated inaccordance with 2-OHP-4025-001-001, Emergency Remote Shutdown.Which ONE of the following describe the operation of the Emergency DieselGenerators (EDGs)?Start both EDGs prior to control room evacuation.

Locally control EDGs from LSI panels as required. Trip both EDG HEAs prior to leaving control room.

Restore EDGs per 2-OHP-4025-R-15, if required Leave EDG control switches as is.

Locally Trip and Isolate EDGs in accordance with 2-OHP-4025-LTI-3, if required.Leave EDG control switches as is.

Locally control EDGs from LSI panels as required.A.B.C.D.ANSWER: CA - INCORRECT. EDGs are isolated after control room evacuation is complete. Thereis no local control for EDGs from the LSI panels. Plausible toprepare the EDG's for loading when needed and provides a logicalplace for control of the EDG outside of the Control RoomB - INCORRECT. EDGs are left as is. Trip, Isolation, and restoration is performedfollowing evacuation per appendix R procedures. Plausible toprevent inadvertent operation of the EDG's complicatingmanagement of power sourcesC - CORRECT. EDGs are left as is. Trip, Isolation, and restoration is performedfollowing evacuation per appendix R procedures.D - INCORRECT. See B. Additionally, there is no local control for EDGs from the LSIpanels. Plausible because it provides a logical place for control ofthe EDG outside of the Control Room Page 34 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EC02\#4

REFERENCE:

OHI-4023 Section 4.2.3 & Attachment 3 Section 4.1KA - 000068 AK2.07 Control Room Evacuation Knowledge of the interrelations between the Control Room Evacuation and thefollowing:

ED/G RO - 3.3 SRO - 3.4 CFR - 41.7 / 45.7 KA Justification - Question addresses how the EDG operation is addressed during aControl Room Evacuation due to fire.Original Question # - NEW Original Question KA - NEW Page 35 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

19. 019 004/BOTH/OK/DIRECT/NRC EXAM 2004-051-3/000077 AA2.07/3.6/4.0/H/3Given the following conditions on Unit 1:

• Unit is in Mode 3.

• The 4160 VAC distribution system is being supplied by the Reserve AuxiliaryTransformers (RATs).

• Due to a system disturbance, indicated voltage on the safeguards buses drops. The following conditions now exist:T11A Voltage Indication is 112 VoltsT11B Voltage Indication is 114 VoltsT11C Voltage Indication is 113 VoltsT11D Voltage Indication is 114 VoltsWhich ONE of the following describes the FINAL plant response if voltage remains atthese values for an extended period?All safeguards busses will be energized by their respective EDG.T11A and T11C busses will be energized by their respective EDG.T11A and T11B busses will be energized by its respective EDG.Only T11A bus will be energized by its respective EDG.A.B.C.D.ANSWER: CA - INCORRECT. T11 C and T11D will NOT deenergize since T11D is > 113V.Plausible if setpoint not known by studentB - INCORRECT. T11C will NOT deenergize since T11D is > 113V. Plausible due toactual setpoint known by student but alignment not understood bystudent for bus stripping.C - CORRECT. An Undervoltage condition of 113 V will energize 62-1 T11A. After a111 Second delay it will open T11A9 and T11B1 causing T11 A andT11B to lose power. This will cause the EDG to start and energizeT11A and T11B.D - INCORRECT. T11B will also recieve a trip signal and be energized by the EDG.Plausible if alignment not understood by student for bus stripping. Page 36 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-08201/#6

REFERENCE:

RO-C-08201, Engineered Safety Systems Electrical pg. 29,32-33, and Att.3. Annunciator #121 Response, Drop 78 Train BAux Buses Undervoltage pg. 178-182, SOD-08201-001KA - 000077 AA2.07 Generator Voltage and Electric Grid DisturbancesAbility to determine and interpret the following as they apply to Generator Voltage andElectric Grid Disturbances:

Operational status of engineered safety features RO - 3.6 SRO - 4.0 CFR - 41.5 and 43.5 / 45.5 / 45.7, and 45.8 KA Justification - Requires determination of the status of electrical buses following anelectrical grid disturbance resulting in degraded grid voltage.Original Question # - NRC EXAM 2004-051-3 Page 37 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

20. 020 002/BOTH/OK/DIRECT/RO24 AUDIT-12/003000 K2.02/2.5/2.6/H/3Given the following conditions on Unit 1:

• Reactor power is 100%.

• West CCP is in operation with the East CCP in standby.

• West CCW pump is tagged out for maintenance. Which ONE of the following describes the immediate operator actions required for aloss of Bus T11D:Trip reactor because the RCP seals will overheat without Component Coolingflow.

Trip reactor because there is NO charging flow to replace letdown.Initiate a controlled shutdown because the Charging pump will overheat withoutComponent Cooling flow.

Initiate a controlled shutdown because the RCP seals will overheat without chargingflow.A.B.C.D.ANSWER: A A - CORRECT. Per Loss of CCW procedure, Trip Reactor and Then trip RCPs.B - INCORRECT. Plausible since Letdown is isolated to conserve level, a trip is notrequired due to loss of letdown.C - INCORRECT. Plausible since CCPs must be shutdown but shutdown is requiredwithin 1-2 minutes so a controlled Shutdown is not warranted.D - INCORRECT. Plausible since RCP seals will overheat when charging is stopped.An attempt is made to crosstie to the opposite unit. The concern withRCP motor bearings is more severe and requires immediate trip. Page 38 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01600 / #RO-C-01600-E6,RO-C-AOP-D14\#RO-C-AOP0560412-E3

REFERENCE:

1-OHP-4022-016-004, Loss of Component Cooling Water,RO-C-AOP-D14, RO-C-01600 pg. 13-14KA - 003000 K2.02 Reactor Coolant Pump System (RCPS)

Knowledge of bus power supplies to the following:

CCW pumps RO - 2.5 SRO - 2.6 CFR - 41.7 KA Justification - Requires the knowledge of the bus power supplies to the CCWpumps and the effect a loss of the CCW pumps will have on theRCPs.Original Question # - INPO - DIRECT 20154, COOK02-052-1,R39,S45, RO24Audit-12Original Question KA - 003000-K2.02, 062.A2.01 Page 39 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

21. 021 004/BOTH/OK/DIRECT/RO26 AUDIT-4/004000 2.4.47/4.2/4.2/H/3Given the following conditions on Unit 2:

• Reactor is at 100% power.

• All control systems are in normal alignment.

• Letdown flow is aligned with a flow of 120 gpm at QFI-301.The following parameters are now noted on the CVCS system:* Seal Return Flows are 3 gpm per RCP

• Charging flow is 137 gpm and rising.

• 2-QTA-160, Regen HX Outlet Temp - Letdown, has lowered 5°F from its steady statevalue.

• VCT level is 33% and lowering.

• PZR level is 55% and lowering slowly.

• RCS temperature is 574°F and stable. Which ONE of the following describes the effect on the unit and the action required toaddress the conditions? RCS leakage is from the letdown line between the orifices and the letdowncontainment isolation valves. Isolate Letdown.

RCS leakage is from the charging line on the RCS side of the regenerative heatexchanger. Isolate Charging and Letdown.

RCS leakage is from the letdown line on the CVCS side of the regenerative heatexchanger. Initiate an investigation to determine if the leak is isolable. RCS leakage is from the charging line on the CVCS side of the regenerative heatexchanger. Isolate Charging and Letdown. A.B.C.D. Page 40 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. Leak is in the charging header. This leakage would cause a rise inletdown temperature. B - CORRECT. If Regen Hx Outlet temperature is lowering, then more charging flowis going through the Regen Hx. This means the leak is downstreamof the RHX in containment. Since Charging flow increased from anormal value of 132 GPM to 137 GPM with no other changes theleak rate is approximately 5 GPM and will required isolation of thecharging header to isolate the leak. C - INCORRECT. Leak is in the charging header. This leak would raise regenerativeHX letdown temperature.D - INCORRECT. Wrong location for leak. This leakage would cause a rise in letdowntemperature. LESSON PLAN/OBJ: RO-C-AOP-D1/#RO-C-AOP0160412-E1

REFERENCE:

SOD-00300-001 KA - 004000 2.4.47 Chemical and Volume Control System (CVCS)

Emergency Procedures/PlanAbility to diagnose and recognize trends in an accurate and timely manner utilizing theappropriate control room reference material.

RO - 4.2 SRO - 4.2 CFR - 41.10 / 43.5 / 45.12 KA Justification - Requires use of reference material (instrument readings and drawing)to determine the location RCS leakage on in the CVCS system.Original Question # - RO26 AUDIT-4, SEQ2007 Original Question KA - 004000 K6.07 Page 41 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

22. 022 004/BOTH/OK/MODIFIED/NRC EXAM 2007-7/004000 K3.08/3.6/3.8/H/3Given the following conditions:

• QRV-200, RCP Seal Backpressure Valve, is operating at 30% open. Assuming QRV-251, Charging Line Flow Control Valve is NOT adjusted, IF QRV-200fails to 60% open, THEN: Charging Pump RCP Seal Charging Flow to Discharge Press Injection Flow Regen Hx Lowers Rises Lowers Rises Lowers Rises Rises Rises Lowers Lowers Lowers RisesA.B.C.D.ANSWER: DA - INCORRECT. Seal injection flow lowers and Charging Flow rises. Plausible due torelationship between RCP Seal Injection flow and Charging is correctbut the directions are wrong. Student must know how the valvefailures impact system operation.B - INCORRECT. CCP discharge pressure lowers and Charging Flow rises. Plausibledue to relationship between RCP Seal Injection flow and Charging iscorrect but the impact on Charging pump discharge pressure iswrong. Student must know how the valve failures impact systemoperation.C - INCORRECT. CCP discharge pressure lowers, Seal injection flow lowers, andCharging Flow rises. Plausible due to relationship between RCPSeal Injection flow and Charging is correct but the directions arewrong but the impact on Charging pump discharge pressure iswrong. Student must know how the valve failures impact systemoperation.D - CORRECT. QRV-200 will cause a lower backpressure on the CCP dischargeand seal injection line, resulting in lower CCP discharge pressureand less flow to the RCP seals. In addition this action will raisecharging flow. Page 42 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00300/#3

REFERENCE:

SOD-00300-001 MODIFIED: Changed stem to go from 30% open to 60% open. Changed thecorrect answer to "D"KA - 004000 K3.08 Chemical and Volume Control System (CVCS)

Knowledge of the effect that a loss or malfunction of the CVCS will have on thefollowing:

RCP seal injection RO - 3.6 SRO - 3.8 CFR - 41.7 / 45.6 KA Justification - Requires the knowledge of a malfunction of a CVCS component(QRV-251) will have on RCP Seal Injection.Original Question # - INPO # 28845 Indian Point Unit 2 - 12/9/2004, NRC EXAM 2007-7Original Question KA - 000022 AK1.02 Page 43 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

23. 023 002/BOTH/OK/DIRECT/NRC EXAM 2006-029-45/005000 K6.03/2.5/2.6/H/3Given the following conditions:

• Unit 2 is in Mode 4 during cooldown per 2-OHP-4021-001-004, Plant Cooldown from Hot Standby to Cold Shutdown

• West RHR Pump and Heat Exchanger are operating, aligned to the cooldown paththrough injection lines to Cold Legs Loops 2 & 3

• RCS temperature is 300°F and stable

• RCS pressure is 335 psig and stableThe air supply line to IRV-320, West RHR Hx Outlet Valve, breaks, causing a completeloss of Instrument Air to the valve.

Which ONE of the following describes the effect on the plant and the action that couldbe taken to mitigate the transient?RHR Flow through the West HX will be lost. Throttle open IRV-311 RHR HX Bypassto maintain greater than 3000 gpm RHR flow.

RHR Flow through the West HX will be lost. Stop the West RHR pump immediatelyto prevent overpressurizing letdown.

RHR Flow through the West HX will rise. Throttle ICM-111, RHR Discharge to ColdLeg 2 & 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.RHR Flow through the West HX will rise. Throttle ICM-321, West RHR Injection toLoops 2 & 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.A.B.C.D.ANSWER: DA - INCORRECT. IRV-320 fails open so flow will raise. Plausible since RHR flow ismaintained > 3000 gpm to minimize vibrations & cavitation throughthe piping.B - INCORRECT. IRV-320 fails open so flow will raise. Plausible since RHR flow tothe letdown system taps off before the IRV-320 and if IRV-320closed it may raise letdown pressure.C - INCORRECT. The ICM-111, RHR Discharge to Cold Leg 2 & 3 is in the NORMALCooldown Path This would be correct if the Normal Cooldown pathwas in service. (2-OHP-4021-017-002, Step 4.13)D - CORRECT. IRV-320 fails open on loss of air. This will raise RHR flow throughthe HX. ICM-321 can be throttled closed to reduce total RHR flowand IRV-311 can be throttled open to allow more flow to bypass theHX in order to control RCS cooldown. Page 44 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01700/#RO-C-01700-E4, RO-C-01700-E6

REFERENCE:

2-OHP-4021-017-002, Placing in Service the RHR System;2-OHP-4022-064-002, Loss of Control Air Recovery (Step 43 &Att. B-9), SOD-01700-001KA - 005000 K6.03 Residual Heat Removal System (RHRS)

Knowledge of the effect of a loss or malfunction of the following will have on the RHRS:RHR heat exchanger RO - 2.5 SRO - 2.6 CFR - 41.7 / 45.7 KA Justification - Requires knowledge of how to control RCS cooldown rate following amalfunction of the RHR Hx Out valve full open.Original Question # - Cook 2006 NRC Exam - 029-45Original Question KA - 005000 A2.01 Page 45 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

24. 024 003/BOTH/OK/DIRECT/NRC EXAM 2004-069-5/006000 K1.11/2.8/3.2/F/3Given the following conditions:

• Unit 2 has experienced a loss of both CCW pumps in MODE 3* NEITHER Unit 2 CCW pump can be restarted.

• CVCS crosstie from Unit 1 is NOT available.

• BOTH Unit 2 CCPs are running because a CCP swap was in progress.

• 2-OHP-4022-016-004, Loss of Component Cooling Water, is in progress.Which ONE of the following describes the procedural requirements for CCP operationbased on these conditions?Immediately stop both CCPs.

Immediately stop one CCP; stop the second CCP within 1-1/2 minutes of the event.Stop BOTH CCPs within 1-1/2 minutes of the event.

Immediately stop one CCP; run the second CCP as long as it continues to operate.A.B.C.D.ANSWER: DA - INCORRECT. One pump should be run as long as possible to allow time to alignSeal injection crosstie. Plausible due to knowledge that the CCP'swill fail in a short time frame without cooling.B - INCORRECT. One pump should be run as long as possible to allow time to alignSeal injection crosstie. (The pump may trip after 1.5 minutes)Plausible due to knowledge that the CCP's will fail in a short timeframe without cooling and the allowable time to operate with nocooling is 90 seconds. This saves one pump for a later period.C - INCORRECT. One pump should be run as long as possible to allow time to alignSeal injection crosstie. (The pump may trip after 1.5 minutes)Plausible due to knowledge that the CCP's will fail in a short timeframe without cooling and the allowable time to operate with nocooling is 90 seconds. This saves both pumps for a later period.D - CORRECT. 02-OHP-4022-016-004 has a note prior to step 4 that describes thepossible damage that may occur to a CCP on the loss of CCW. Thenote and procedure directs that one CCP be saved until CCW isrestored. The other pump should be run as long as possible to allowtime to align Seal injection crosstie. Page 46 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D14\#RO-C-AOP0560412-E3

REFERENCE:

2-OHP-4022-016-004, Loss of Component Cooling WaterKA - 006000 K1.11 Emergency Core Cooling System (ECCS)

Knowledge of the physical connections and/or cause-effect relationships between theECCS and the following systems:

CCWS RO - 2.8 SRO - 3.2 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Requires knowledge of the cause and effect relationship between thea loss of CCW to the CCP (ECCS Pump) and the ability of the CCPto continue to operate.Original Question # - Master AOP1CAOP5.13, NRC Exam 2004-069-5Original Question KA - Unknown Page 47 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

25. 025 004/BOTH/OK/NEW/NEW/007000 A4.10/3.6/3.8/F/3Given the following conditions:

• The Plant has just completed a Heatup to Normal Operating Temperature andPressure.* Operators suspect a small leak through Pressurizer Safety Valve SV-45B.What indication combinations are available to help the operator determine if this valveis faulted?A significant PRT Temperature Rise (>200 oF)The Common Safety Valve Tailpipe Temperature indicatorThe Common Safety Valve line acoustic monitor A significant PRT Temperature Rise (>200 oF)The Safety Valve SV-45B Tailpipe Temperature indicatorThe Safety Valve SV-45B line acoustic monitor A slight PRT Temperature Rise (<50 oF)The Safety Valve SV-45B Tailpipe Temperature indicatorThe Common Safety Valve line acoustic monitor A slight PRT Temperature Rise (<50 oF)The Safety Valve SV-45B Tailpipe Temperature indicatorThe Safety Valve SV-45B line acoustic monitorA.B.C.D. Page 48 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: DA - INCORRECT. A large temperature indication of > 200 oF would only be expected ifthe safety was failed mostly open. PORVs share indicators andsafeties have separate indicators. Plausible but due to constantenthalpy from throttling process the temperature change can not be200 oF as well as the Safeties have individual temp and acousticmonitors unlike the PORV's with a common set of temp andacoustic monitors.B - INCORRECT. A large temperature indication of > 200 oF would only be expected ifthe safety was failed mostly open. It is correct that the safeties haveseparate indicators. - Plausible but due to constant enthalpy fromthrottling process the temperature change can not be 200 oF theremaining portion of the answer is correct.C - INCORRECT. PORVs share indicators and safeties have separate indicators forthe acoustic line. Plausible due to correct anticipated change intemperature but the remaining portion of the answer is not correctfor the Safeties - the PORV's have the common set of temperatureand acoustic monitors.D - CORRECT. For a small leak into the PRT there should not be a significant risein PRT temperature. PORVs share indicators and safeties haveseparate indicators for the acoustic line.LESSON PLAN/OBJ: RO-C-00202/#4, RO-C-EOP09/#22

REFERENCE:

RO-C-00202 pg. 41 KA - 007000 A4.10 Pressurizer Relief Tank/Quench Tank System (PRTS)Ability to manually operate and/or monitor in the control room:Recognition of leaking PORV/code safety RO - 3.6 SRO - 3.8 CFR - 41.7 / 45.5 to 45.8 KA Justification - Question tests the ability of the operator to monitor (by identifyingexpected PRT Temperature trend and available indications) the PRTTemperature and associated connections (PORV/SAFETY lines) tohelp determine which Safety is leakingOriginal Question # - New Original Question KA - New Page 49 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

26. 026 004/BOTH/OK/MODIFIED/NRC EXAM 2007-40/013000 K5.02/2.9/3.3/H/4Given the following conditions:

• Containment pressure instrument Channel #1, 2-PPP-303, declared inoperable.

• Required actions per 2-OHP-4022-013-011, Containment InstrumentationMalfunction, have been completed.

• Required Technical Specification Actions have been taken for Channel #1,2-PPP-303.Which ONE of the following describes the SI and CTS, and Containment IsolationPhase A (CIA) and B (CIB) response to a subsequent failure of CRID 4 power supply. SI CTS CIA CIB ACTUATES ACTUATES ACTUATES ACTUATES YES NO YES NO YES YES YES YES NO YES NO YES NO NO NO NOA.B.C.D.ANSWER: D A - INCORRECT. See explanation Below.B - INCORRECT. See explanation Below.

C - INCORRECT. See explanation Below.

D - CORRECT. See explanation Below.

The CTS Actuation Bistable is placed in the BYPASSED condition to preventinadvertent actuation. This changes the remaining channel coincidence to 2/3 insteadof the previous 2/4. Only 3 channels (Channels 2, 3, & 4) feed the SI Actuation. Thebistable for the CTS actuation is placed in the BYPASS condition, making the CTS a2/3 coincidence for the remaining channels (2, 3, and 4). CRID 4 failure will NOT meetthe 2/3 co-incidence for either the SI or CIA. CTS/CIB still required 2/3 to actuate,therefore only one channel will not cause the CTS/CIB. The logic between CTS andCIB is always the same which prevents the student from eliminating distractors basedon obvious distractors; this same logic applies to SI and CIA. This forces the student tohave a full understanding of the question in order to answer the question correctly. Page 50 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01100/#6

REFERENCE:

2-OHP-4022-013-011 Containment Instrumentation MalfunctionMODIFIED: Changed from Channel #4 to Channel #1 (PPP-303). Changed to aCRID 4 failure. Changed correct answer to "D."KA - 013000 K5.02 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the operational implications of the following concepts as they apply to theESFAS:

Safety system logic and reliability RO - 2.9 SRO - 3.3 CFR - 41.5 / 45.7 KA Justification - Requires the knowledge of the logic coincidence for both the SI/CIAand CTS/CIB functions and how the redundancy of instrumentsallows for single failure and will still actuate as required for multiplefailures (reliability). Original Question # - COOK04-037, NRC EXAM 2007-40Original Question KA - 013000 K2.01 Page 51 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

27. 027 001/BOTH/OK/NEW/NEW/008000 A2.05/3.3/3.5/H/3Given the following conditions:

• Unit 1 and Unit 2 are operating at 100% Power.

• North Spent Fuel Pit pump and cooler in service

• Spent Fuel Pit at Normal Level

• Spent Fuel Pit contains 2015 spent fuel assemblies

• Refueling Water Purification aligned to purify Unit 1 RWSTThe 1-CRV-445, CCW from North SFP Hx, control air supply line ruptures causingvalve to fail.

Which ONE of the following conditions describes the impact on the Spent Fuel PoolCooling system and the actions needed to address this condition?SFP Temperature will rise.

Place the South SFP Pump and Hx in service.

SFP Temperature will rise.

Manually control SFP temperature using 1-CRV-445 bypass valve.SFP Temperature will lower.

Place the South SFP Pump and Hx in service.

SFP Temperature will lower.

Isolate 1-CRV-445 and manually control SFP temperature using 1-CRV-445 bypassvalve.A.B.C.D.ANSWER: AA - CORRECT. 1-CRV-445 fails closed on loss of air. Due to the heat load in theSFP, the SFP temperature will rise. Impacts the ability of Unit 1CCW system to provide SFP Cooling. Unit 2 SFP cooling loop willneed to be placed in service.B - INCORRECT. 1-CRV-445 fails closed on loss of air. Due to the heat load in theSFP, the SFP temperature will rise. 1-CRV-445 does not have abypass valve however many air operated valves in the plant dohave bypass valves around them which makes this a validdistractor.C - INCORRECT. Temperature would lower if 1-CRV-445 failed open on loss of air.D - INCORRECT. Temperature would lower if 1-CRV-445 failed open on loss of air. 1-CRV-445 does not have a bypass valve however many airoperated valves in the plant do have bypass valves around themwhich makes this a valid distractor Page 52 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01600/#RO-C-01600-E6

REFERENCE:

OP-1-5135B, SOD-01600-001, COMPONENT COOLINGWATER SYSTEMKA - 008000 A2.05 Component Cooling Water System (CCWS)Ability to (a) predict the impacts of the following malfunctions or operations on theCCWS and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:Effect of loss of instrument and control air on the position of the CCW valves that areair operated RO - 3.3 SRO - 3.5 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - requires the ability to predict the impact of loss of air to 1-CRV-445and the ability of the CCW system to cool the SFP. Based on thisimpact, requires knowledge of the actions required to control theconsequences the malfunction.Original Question # - New Original Question KA - New Page 53 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

28. 028 003/BOTH/OK/MODIFIED/NRC EXAM 2008-23/00WE03 EA1.1/4.0/4.0/H/3Given the following conditions on Unit 1:

• A Small Break LOCA has occurred.

• RCS Wide Range Pressure lowered to 1350 psig and is stable.* Containment pressure has remained less than 2.8 psig.* The actions of 1-OHP-4023-ES-1.2, Post LOCA Cooldown And Depressurization,are in progress.

• Both CCPs are running with suction aligned to the RWST.

• Both RHR Pumps are stopped in Neutral.

• Both SI Pumps are running.

• The crew is ready to depressurize the RCS to refill the Pressurizer. Which ONE of the following is the FIRST method available to the operator tocommence the RCS depressurization?

The operator will open:One PZR PORV to depressurize the RCS.

All Pressurizer PORVs to depressurize the RCS.

The PZR Aux Spray Valve to spray down the PZR steam space. PZR Normal Spray Control valve(s) to spray down the PZR steam space. A.B.C.D.ANSWER: DA. INCORRECT. Since normal spray are available (RCS Wide Range Pressure abovethe RCP Trip Criteria), sprays would be used before one PZR PORV.B. INCORRECT. Opening MORE THAN ONE PORV is NOT an appropriate action. Ifthis option were used, then only one PORV would be used tominimize the potential for a PORV sticking open.C. INCORRECT. This action is a third option in the event that a PORV is not available.D - CORRECT. This is the "normal" method used to depressurize the RCS in ES-1.2.Note: ALL distractors are valid methods that can be used todepressurize the RC and accomplish the intended goal ofrefilling the pressurizer. The student must be able to determinewhich action is the most correct based on plant conditions. Page 54 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP09/#36

REFERENCE:

12-OHP-4023-ES-1.2, Step 13 Background MODIFIED: Removed loss of offsite power and replaced with RCS Wide RangePressure above the RCP Trip criteria from the Foldout Page. Changed correct answer to "D."KA - 00WE03 EA1.1 LOCA Cooldown and DepressurizationAbility to operate and/or monitor the following as they apply to the LOCA Cooldown andDepressurization:

Components, and functions of control and safety systems, including instrumentation,signals, interlocks, failure modes, and automatic and manual featuresRO - 4.0 SRO - 4.0 CFR - 41.7 / 45.5 / 45.6 KA Justification - Requires the ability to depressurize the RCS during a post LOCAcooldown and depressurization and the ability to determine thedepressurization method available based on plant conditions.Original Question # - INPO Bank #30436 - KEWAUNEE-222006, NRC EXAM2008-23Original Question KA - WE03EA1.1 Page 55 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

29. 029 003/BOTH/OK/DIRECT/RO26 AUDIT-61/00WE04 EK2.1/3.5/3.9/H/3The plant was in Mode 1. Reactor trip and safety injection have occurred. Due to highAux Building radiation levels, the crew has entered 2-OHP-4023-ECA-1.2, LOCAOutside Containment. Actions have been taken in an attempt to isolate the break. Given the following plant conditions:

• PZR level is off-scale low

• SI pump flow is 0 GPM

• RCS pressure is 1700 psig and rising.

• Aux Building Radiation Monitors are in alarm Which ONE of the following describes the status of the leak based on the requirementsof 2-OHP-4023-ECA-1.2? The leak is isolated based on SI flow of 0 GPM The leak is isolated based on RCS pressure rising. The leak is NOT isolated based on PZR level indication not rising. The leak is NOT isolated based on Aux Building radiation monitor indication. A.B.C.D.ANSWER: BA - INCORRECT. SI flow would be 0 if RCS pressure is above shutoff head of the SIPump. Plausible due to not having any injection flow could meanthat the leak is isolated but this is not what is directed by theprocedureB - CORRECT. RCS pressure is the required parameter for determination ofisolation Incorrect.C - INCORRECT. PZR level is not used, but it will rise after awhile when RCSinventory is restored. Plausible due to Fold out page in severalEOP's require re-initiation of SI if PZR level is at described level instem of question.D - INCORRECT. Plausible since Aux Building radiation is used as an entry conditionto the procedure. Page 56 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP09/#34

REFERENCE:

2-OHP-4023-ECA-1.2, LOCA Outside Containment, RO-C-EOP09 pg. 252KA - 00WE04 EK2.1 LOCA Outside Containment Knowledge of the interrelations between the LOCA Outside Containment and thefollowing:

Components, and functions of control and safety systems, including instrumentation,signals, interlocks, failure modes, and automatic and manual featuresRO - 3.5 SRO - 3.9 CFR - 41.7 / 45.7 KA Justification - Question Addresses a LOCA Outside Containment and how the plant& instrumentation responds (interrelations) based on closure ofvalves (operation of components). Original Question # - RO26 AUDIT-61, GINNA2007 Page 57 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

30. 030 005/BOTH/OK/DIRECT/MASTER 01EOPC1110-7/00WE05 EK3.3/4.0/4.1/F/3The control room operators are responding to a red path on the Heat Sink CSF. Whileattempting to restore feed flow to a SG in accordance with OHP-4023-FR-H.1,Response to Loss of Secondary Heat Sink, conditions degrade to the point that RCSbleed-and-feed must be established.

Under these conditions, RCS bleed-and-feed must be established expeditiously to:prevent a loss of secondary heat sink.

minimize core uncovery and prevent inadequate core cooling. prevent an overpressurization challenge to the reactor vessel. prevent a rapid RCS overpressurization, followed by a rapid RCS depressurizationdue to RCP seal failure.A.B.C.D.ANSWER: BA - INCORRECT. Attempts to restore the heat sink have been unsuccessful. Bleed-and-feed is established to raise the amount of injection flowinto the core and thus minimize the core uncovery. Plausible basedon the purpose of the FRP being used. The goal is to establish andthus prevent a loss of secondary heat sink.B - CORRECT. If the operator cannot restore feedwater flow to the SGs, conditionswill degrade to the point where RCS bleed and feed must beestablished to minimize core uncovery and prevent inadequate corecooling.C - INCORRECT. Based on the FRGs, even if overpressurization were a concern, thepriority of core cooling is higher than PTS concerns. Plausiblebased on the event if no actions were taken the RCS wouldincrease in pressure based on the lack of heat removal from theRCS.D - INCORRECT. Core cooling is the second highest priority in the EOPs(Subcriticality being the highest). Any other concern would be of alower priority than establishing core cooling of some nature, in thiscase bleed-and-feed. Plausible based on the conditions describeddo not state the basis for requiring feed and bleed (the conditionscould be caused by high RCS pressure) and the conditions couldalso exist without any Charging pumps running thus challenging theRCP seals. Page 58 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP11/#10

REFERENCE:

RO-C-EOP11 Study Guide, FR-H.1 Background KA - 00WE05 EK3.3 Loss of Secondary Heat Sink Knowledge of the reasons for the following responses as they apply to the Loss ofSecondary Heat Sink:

Manipulation of controls required to obtain desired operating results during abnormal,and emergency situations RO - 4.0 SRO - 4.1 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Requires knowledge of the reason for implementing bleed-and-feed inthe EOPs during a Loss of Heat Sink event when heat sink cannot berestored.Original Question # - 01EOPC1110-7 Original Question KA - EPE 005 EK3.1 Page 59 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

31. 031 003/BOTH/OK/DIRECT/RO24 AUDIT-023-7/W/E09 EK3.2/3.2/3.6/F/3Given the following conditions:

• Unit 2 Reactor Tripped due to a loss of offsite power* The crew is implementing 2-OHP-4023-ES-0.2, Natural Circulation Cooldown Which ONE fo the following describes the reason for maintaining subcooling greaterthan 90F if ALL CRD fans are running OR greater than 220F if less than ALL CRDfans are running during the cooldown.To collapse any voids formed in the CRD housings. To prevent possible void formation in the upper head. To prevent degradation of reactor coolant pump seals due to steam.To ensure adequate subcooling due to possible degradation of core exit T/Csaccuracy. A.B.C.D.ANSWER: BA - INCORRECT. Plausible since loss of CRD fans would lead to high temperatures inCRD housing cooling, but this is not the reason for requiring highersubcooling.B - CORRECT. 2-OHP-4023-ES-0-2, Natural Circulation Cooldown requires anRCS subcooling of 220°F in the event CRDM fans are NOT runningto preclude void formation in the upper head. Normal naturalcirculation RCS subcooling is 90°F.C - INCORRECT. Plausible since overheating and steam formation in the RCP sealsis a concern on loss of all AC. D - INCORRECT. Plausible since TCs are compensated for inaccuracies, but this isnot the reason for requiring higher subcooling. Page 60 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP03/#12

REFERENCE:

2-OHP-4023-ES-0-2, Natural Circulation Cooldown, step 14KA - 00WE09 EK3.2 Natural Circulation Operations Knowledge of the reasons for the following responses as they apply to the NaturalCirculation Operations:

Normal, abnormal and emergency operating procedures associated with NaturalCirculation Operations RO - 3.2 SRO - 3.6 CFR - 41.5 / 41.10 / 45.6 / 45.13 KA Justification - Requires knowledge of the reason for the requirement to maintainadequate subcooling margin during a natural circ cooldown event.Original Question # - RO24 AUDIT-023-7 Original Question KA - W/E09 EK2.1 Page 61 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

32. 032 003/BOTH/OK/DIRECT/CATAWBA2005/00WE11 EA2.2/3.4/4.2/F/3Given the following conditions:

• 1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation, has just beenentered* Refueling Water Storage Tank (RWST) level is 5.5%Which ONE of the following procedure actions is performed FIRST?Initiate makeup to the U-1 RWST from the Boric Acid Blender.Start one reactor coolant pump.

Initiate makeup to the U-1 RWST from the U-2 RWST.

Secure all ECCS and CTS pumps taking a suction from the RWST.A.B.C.D.ANSWER: DA - INCORRECT. Makeup is started in step 7. Plausible as this is a step in theprocedure but does not occur prior to the requirement off the foldout page.B - INCORRECT. RCPs are not started until later. Plausible as the procedure doesstart RCP's but does not occur prior to the requirement off the foldout page.C - INCORRECT. Makeup from the Opposite unit RWST is not used until step 7.Plausible as the procedure does accomplish this action but not untilafter the fold out page has been implemented .D - CORRECT. The Foldout Page has actions to secure RHR & CTS when level is< 11% and CCP & SI pumps when level is <7%. This would be thefirst action taken upon entering the procedure. Page 62 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP09\#35

REFERENCE:

1-OHP-4023-ECA-1.1 Foldout Page. (Note that this action is alsolisted as a critical task for ECA-1.1 RO-C-EOP09 pg. 106)KA - 00WE11 EA2.2 Loss of Emergency Coolant RecirculationAbility to determine and interpret the following as they apply to the Loss of EmergencyCoolant Recirculation:Adherence to appropriate procedures and operation within the limitations in the facility'slicense and amendments RO - 3.4 SRO - 4.2 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the Ability to determine and interpret proceduralrequirements for stopping ECCS and CTS pumps on low RWST levelto keep plant operating within the limitations in the facility's licenseand amendments.Original Question # - RO25 AUDIT-17, CATAWBA2005 Original Question KA - 00WE11 EA2.2 3.4/4.2 CFR 43.5/45.13 Page 63 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

33. 033 002/BOTH/OK/DIRECT/SEQ2007/00WE12 EK1.2/3.5/3.8/H/3Operators are performing 2-OHP-4023-ECA-2.1, Uncontrolled Depressurization of AllSteam Generators due to a steam leak inside containment along with failure of all SGstop valves to close.

Given the following plant conditions:

• Containment pressure is 3 psig.

• The crew has taken action to minimize the plant cooldown.

• Steam Generator AFW flow indicates 25x10 3 pph each SG.

• T-hots are slowly lowering.

• The following alarms are received: Ann. 213 Drop 5, STEAM GEN #1 WATER LEVEL LOW-LOW Ann. 213 Drop 35, STEAM GEN #2 WATER LEVEL LOW-LOW Ann. 214 Drop 5, STEAM GEN #3 WATER LEVEL LOW-LOW Ann. 214 Drop 35, STEAM GEN #4 WATER LEVEL LOW-LOWWhich ONE of the following actions is required in accordance with2-OHP-4023-ECA-2.1? Adjust AFW flow to 60x10 3 pph on each Steam Generator. The minimum NR level,per 2-OHP-4023-ECA-2.1, is 28%.

Adjust AFW flow to 60x10 3 pph on each Steam Generator. The minimum NR level ,per 2-OHP-4023-ECA-2.1 , is 50%.

Maintain AFW flow at its current value. If T-hot starts to rise, raise AFW flow tostabilize RCS temperature.

Maintain AFW flow at its current value. If SG levels continue to lower, raise AFWflow to maintain SG levels >13% to prevent a transition to 2-OHP-4023-FR-H.1,Response to Loss of Secondary Heat Sink. A.B.C.D. Page 64 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: CA. INCORRECT. After throttling to minimize RCS cooldown, even if levels are low,AFW remains throttled until Thot begins to rise. At that point, AFWis throttled just enough to stabilize temperature. Credible because28% is the lower limit that level is maintained for AdverseContainment. (Containment is NOT Adverse) B. INCORRECT. After throttling to minimize RCS cooldown, even if levels are low,AFW remains throttled until Thot begins to rise. At that point, AFWis throttled just enough to stabilize temperature. Credible because50% is the upper limit that level in most of the EOPs. C. CORRECT. AFW Flow is maintained at a minimum amount since the levels arelow and T-hot is lowering.D. INCORRECT. Flow is maintained per at 25x10 3 pph per the procedure. If thetransition to FR H.1 is reached Step 1 will return the Crew toECA-2.1. Plausible as Operator should know they have the ability toavoid a Red path on heat sink but avoiding this red path is inviolation of ECA 2.1 requirements and the red path addresses theintentional reduction in AFW flow.Note:

• Level setpoint for SG Low-Low is 22% - TDAFP start setpoint.

• Since this an operator induced reduction of AFW flow, FR-H. 1 actionswould not be performed even if the transition was made.LESSON PLAN/OBJ: RO-C-05100\#12, RO-C-EOP07/#8

REFERENCE:

2-OHP-4024-213 & 214 Drops 5 & 35, 2-OHP-4023-ECA-2.1,RO-C-EOP07KA - 00WE12 EK1.2 Uncontrolled Depressurization of all Steam GeneratorsKnowledge of the operational implications of the following concepts as they apply to theUncontrolled Depressurization of all Steam Generators:Normal, abnormal and emergency operating procedures associated with UncontrolledDepressurization of all Steam Generators RO - 3.5 SRO - 3.8 CFR - 41.8 / 41.10 / 45.3 KA Justification - Requires knowledge of the operational implications of the throttlingAFW flow to the SGs during an Uncontrolled Depressurization of allSteam Generators and the provisions for when to raise flow rates.Original Question # - RO26 AUDIT-64, SEQ2007 Original Question KA - 00WE12 2.4.31 Page 65 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

34. 034 002/BOTH/OK/MODIFIED/NRC EXAM 2008-21/00WE14 EK2.2/3.4/3.8/H/3Given the following conditions on Unit 2:

• A LOCA occurred 60 minutes ago.

• Containment Pressure has risen to 5 psig.

• The crew has completed steps of 2-OHP-4023-ES-1.3, Transfer to Cold LegRecirculation, to align RHR/CTS suctions to the recirculation sump and the CCP/SIsuctions to RHR Discharge.

• ONLY the Train A CCP, SI, RHR, and CTS pumps are operating.* The next step of 2-OHP-4023-ES-1.3 directs the crew to "Check if RHR Spray isRequired". Based on the indications above, which ONE of the following would best describe therequired action AND the reason for the decision?Place RHR spray in service NOW since ALL of the requirements are met.Place RHR spray in service ONLY if the CTS pump trips.Do NOT place RHR spray in service because the RHR pump suction is NOTaligned to the RWST.

Do NOT place RHR spray in service because ONLY one RHR pump is operating.A.B.C.D. Page 66 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: AA - CORRECT. RHR has injected for 50 minutes. (A LOCA occurred on Unit 2 sixtyminutes ago.)B - INCORRECT. RHR is required if only 1 CTS pump is operating. After RHR hasinjected for 50 minutes the core is sufficiently cooled to allow RHRto be diverted to support spray functions. Plausible since thestudent should know some form of containment pressuresuppression is required that action would be required if the onlyrunning CTS pump trips. Distractor requires the student to knowthe requirement for pressure and time since accident to determinethe correct answer.C - INCORRECT. RHR spray is required 50 minutes after the accident. It is assumedthat RHR will be on Recirculation at this time. Plausible as concernfor available NPSH to the running RHR pump is a concern foraccident mitigation. Distractor requires the student to know thewater level in containment supports the NPSH requirements forestablishing RHR spray.D - INCORRECT. After RHR has injected for 50 minutes the core is sufficiently cooledto allow RHR to be diverted to support spray functions. Plausibledue to concern for placing the running RHR pump in a runoutcondition from the increase in flow thru the pump. Page 67 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP13 / #13, RO-C-EOP09 / #36

REFERENCE:

OHP-4023-FR-Z-1, Response To High Containment PressureStep 4 & Background, OHP-4023-ES-1.3, Transfer to Cold LegRecirculation Step 17 & BackgroundModified: Changed stem to 60 minutes which makes A correct Answer. ChangedDistractor B (former correct answer) from wait until 50 minutes toONLY required if CTS trips. - SWP 1-21-10KA - 00WE14 EK2.2 High Containment Pressure Knowledge of the interrelations between the High Containment Pressure and thefollowing:

Facility's heat removal systems, including primary coolant, emergency coolant, thedecay heat removal systems, and relations between the proper operation of thesesystems to the operation of the facility RO - 3.4 SRO - 3.8 CFR - 41.7 / 45.7 KA Justification - Questions tests the knowledge of how and when the heat removalsystem is used to aid in controlling a high containment pressure.Original Question # - Cook NRC Exam 2002-026-1, 01EOPC1313-2, NRC EXAM2008-21Original Question KA - 00WE14 EK2.2 Page 68 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

35. 035 005/BOTH/OK/DIRECT/NRC EXAM 2004-074-5/00WE16 EA2.2/3.0/3.3/F/4Chemistry had confirmed two leaking fuel rods on Unit 1 when a Small Break LOCAoccurred 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago.

The following conditions exist on Unit 1:

• All Red and Orange Paths have been addressed.

• Containment pressure is 1.0 psig.

• Containment air temperature is 215ºF.

• Lower Containment high range area monitors, (VRA-1310/1410) are reading10 R/HR

• 1-OHP-4023-FR-Z.3, Response to High Containment Radiation Level, is entered.In accordance with 1-OHP-4023-FR-Z.3, which ONE of the following must be verified?Both Containment Recirculation Fans (CEQ) are running.Upper and Lower Containment Ventilation Fans (CUV/CLV) are running.Containment Ventilation Isolation has occurred.

Control Room Ventilation System is in ISOLATE.

A.B.C.D.ANSWER: C A - INCORRECT. Containment Recirculation Fans are run to help reduce HydrogenBuildup. They are NOT run in 1-OHP-4023-FR-Z.3.B - INCORRECT. Containment Ventilation fans are tripped on a Containment Isolationsignal. Plausible as circulating the containment atmosphere is alogical action to aid in the removal of the radiation.C - CORRECT. 1-OHP-4023-FR-Z.3 requires the crew to verify ContainmentVentilation Isolation.D - INCORRECT. Control Room Ventilation is aligned during a SI but is not addressedin 1-OHP-4023-FR-Z.3. Plausible as it is a required action to ensurethe Control Room staff limits accident dose rates but this action waspreviously completed and is not addressed in Z.3. Page 69 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP13/#6

REFERENCE:

1-OHP-4023-FR-Z.3, Response to High Containment RadiationLevel pg. 2KA - 00WE16 EA2.2 High Containment RadiationAbility to determine and interpret the following as they apply to the High ContainmentRadiation:Adherence to appropriate procedures and operation within the limitations in the facility'slicense and amendments RO - 3.0 SRO - 3.3 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the ability to determine the major action category (requiredactions) associated with 1-OHP-4023-FR-Z.3, Response to HighContainment Radiation Level.Original Question # - NRC EXAM 2004-074-5 Original Question KA - 000061 2.4.6 Page 70 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

36. 036 004/BOTH/OK/MODIFIED/RO25 AUDIT-34/010000 A2.01/3.3/3.6/H/3Given the following conditions:

• Unit 1 in Mode 4 cooling down to Mode 5

• Pressurizer level is 80%

• 11PHC Pressurizer heater groups are in MANUAL and ON

• Reactor coolant pumps (RCP) #13 and #14 are running

• An electrical fault results in the loss of RCP Bus 1D and T11DFive minutes later it is reported the pressurizer outflow cannot be verified.Which of the following actions will reinitiate and then maintain a continuous pressurizeroutflow?Verify pressurizer heaters from 11PHC output current and close NRV-164 Loop 4PZR Spray Control valve Energize pressurizer heaters from 11PHA and close NRV-163 Loop 3 PZR SprayControl valve Raise the demand on NRV-164 Loop 4 PZR Spray Control valveAdjust charging and letdown to raise pressurizer level to 85%A.B.C.D.ANSWER: BA - INCORRECT. The loss of T11D causes the loss of PZR Heaters 11PHC.B - CORRECT. The loss of RCP Bus 1D causes the loss of RCP #3. The loss ofT11D causes the loss of PZR Heaters 11PHC. The operator shouldenergize the 11PHA heaters and close the PZR Spray valveassociated with the tripped RCP.C - INCORRECT. The loss of T11D causes the loss of PZR Heaters 11PHC. Raisethe demand on the Spray valve will lower pressure and causeinflow.D - INCORRECT. Raising level will cause an inflow. Page 71 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00202\#14, RO-C-NOP2\#RO-C-NOP2-E13

REFERENCE:

RO-C-NOP2, OHP-4021-001-004 pg. 18 (Note before 4.15),SOD-00202-002MODIFIED: Changed Heater group energized, & loss of Power to RCP #14.Modified distractors to make B correct & Changed A.KA - 010000 A2.01 Pressurizer Pressure Control System (PZR PCS)Ability to (a) predict the impacts of the following malfunctions or operations on the PZRPCS and (b) based on those predictions, use procedures to correct, control, or mitigatethe consequences of those malfunctions or operations:Heater failures RO - 3.3 SRO - 3.6 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question requires operator to predict response (determine status ofcomponents available) of the PZR pressure control system due to aloss of power (and heaters) and use procedures/actions to correct theconditions resulting from the failure.Original Question # - RO25 AUDIT-34, modified from CATAWBA2005Original Question KA - SYS010 A2.01 Page 72 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

37. 037 004/BOTH/OK/NEW/NEW/011000 K5.05/2.8/3.1/H/4Unit 2 was operating at 100% power when an instrument failure caused PressurizerLevel to rise to 73%.

The Crew has restored the level control instrumentation and has stabilized chargingand letdown The following conditions exist at 1300 hour0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />s:

• 2-QRV-160 and 2-QRV-161, Letdown Orifice Valves are Open

• Letdown Hx Outlet Flow QFI -301 118 gpm

• Charging Header Flow QFI -200 120 gpm

• Total seal flow to RCPs QFI -210 to 240 32 gpm

• Pressurizer Level NLP-151 to 153 73%Assuming no operator actions, design RCP seal return flows, and pressurizer volumeof 75 gallons/% at what time will the Pressurizer Level be returned to the 100%program value?1348 hrs.

1406 hrs.

1522 hrs.

1618 hrs.A.B.C.D.ANSWER: CA - INCORRECT. Time is based on time to Unit 2 Pressurizer level of 54.1% assuming30 gpm mismatch (Letdown + Seal Inj - Charging) = 47.25 minutesB - INCORRECT. Time is based on time to Unit 1 Pressurizer level of 46.6% assuming30 gpm mismatch (Letdown + Seal Inj - Charging) = 66 minutesC - CORRECT. Unit 2 100% PZR level is 54.1%. With charging flow at 120 gpm andletdown at 118 with 12 gpm from the seals, a net of 10 gpm is beingremoved from the RCS system. Based on this and a conversion of~75 gallons/% level in the PZR (either unit), level will reach theprogram level setpoint of 54.1% at 1522 hrs (T+141.75 minutes -18.9%=1417.5 gallons). D - INCORRECT. Time is based on Unit 1 Initial Pressurizer Level of 46.6% with 10GPM mismatch = 198 minutesNote: Pressurizer volume based on 16.6 ft 3/% x 1lbm/.0267 ft3 x 1gal/8.35lbm Page 73 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00202/#RO-C-00202-E12

REFERENCE:

SOD-00202-003, SOD-00300-001

KA - 011000 K5.05 Pressurizer Level Control System (PZR LCS)

Knowledge of the operational implications of the following concepts as they apply to thePZR LCS:

Interrelation of indicated charging flow rate with volume of water required to bring PZRlevel back to programmed level hot/cold RO - 2.8 SRO - 3.1 CFR - 41.5 / 45.7 KA Justification - Question tests the operational knowledge of how long it will take torestore the pressurizer level to program based on net charging flow.Original Question # - New Original Question KA - New Page 74 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

38. 038 005/BOTH/OK/DIRECT/NRC EXAM 2004-103-4/012000 A3.04/2.8/2.9/H/3Given the following conditions:

• Pressurizer Pressure Channel #1 has failed and has been placed in the trippedcondition.

• Reactor trip breaker testing was taking place at 75% power.

• Pressurizer Pressure Channel #2 has spiked low causing an inadvertent SafetyInjection Actuation and a reactor trip on Unit 1.

• Pressurizer Pressure Channel #2 has returned to a normal reading.The following conditions currently exist:

• Reactor trip breaker A: OPEN* Reactor trip bypass breaker A: OPEN* Reactor trip breaker B: OPEN* Reactor trip bypass breaker B: CLOSEDWhich ONE of the following describes the impact (if any) this condition will have onrestoring the plant to stable conditions?The Train B Safety Injection signal will NOT be able to be reset. Train B equipmentwill have to be placed in Pull-to-Lockout to stop it.The Train B Safety Injection signal will reset but Auto Safety Injection Actuation willNOT be blocked.

The Safety Injection signal will NOT be able to be reset on either train. Safeguardsequipment will have to be placed in Pull-to-Lockout to stop it.The Safety Injection signal will reset on both trains. Auto Safety Injection Actuationwill be blocked. A.B.C.D. Page 75 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: B A - INCORRECT. The Safety Injection signal will reset. B - CORRECT. The SI reset and P-4 block features are train specific. With a failureof Train B reactor Trip Bypass Breaker to open a P-4 signal is notgenerated on Train B. Since the cause of the SI was a pressurechannel spike the SI signal is NOT preventing Train B from beingreset. The SI will reset but the Auto SI blocking function of P-4 willNOT function on Train B.C - INCORRECT. The Safety Injection signal will reset. The SI reset and P-4 blockfeatures are train specific.D - INCORRECT. The SI reset and P-4 block features are train specific so Train B autoSI will NOT be blocked.LESSON PLAN/OBJ: RO-C-01100 / #6

REFERENCE:

OP-2-98512-21 Safeguard actuation & Reactor Trip SignalsLogic DiagramKA - 012000 A3.04 Reactor Protection SystemAbility to monitor automatic operation of the RPS, including:Circuit breaker RO - 2.8 SRO - 2.9 CFR - 41.7 / 45.5 KA Justification - Question tests ability of operator to determine if auto action (SI) hasoccurred and the impact that a circuit breaker (RTB) will have on thisaction/RPS system. The Operator needs to determine that RTB didNOT automatically Open as expected and that it also provides inputto the P-4 Interlocks.Original Question # - NRC Exam 2004-103-4 Original Question KA - 000007 EA2.03 Page 76 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

39. 039 002/BOTH/OK/DIRECT/NRC EXAM 2006-41/012000 K6.02/2.9/3.1/H/3Given the following conditions on Unit 1:

• A reactor startup is in progress.

• The reactor is critical in the source range.

• N41 Power Range channel has failed and been removed from service with allbistables placed in the trip condition.

• A loss of power to the CRID 2 bus occurs. Which ONE of the following actions will occur?Reactor trips and N32 Source Range channel is de-energized. N31 Source Range channel is still in operation.

The reactor is critical and BOTH source range channels are de-energized.The reactor is critical and N32 Source Range channel is de-energized. N31 Source Range channel is still in operation.

Reactor trips and BOTH source range channels are de-energized.A.B.C.D.ANSWER: DA - INCORRECT. P-10 will be met, both SR's will de energize. B - INCORRECT. Reactor trips on a number of PR/SR trip setpoints.C - INCORRECT. Reactor trips on a number of PR/SR trip setpoints. Also, P-10 willturn off both SR's.D - CORRECT. A loss of CRID 2 causes a loss of power to N42. This loss alsocauses a loss of power to RPS channel 2. This will cause a tripcondition for Power range trips for channel 2. Since N41 is alreadyremoved from service its bistable are in the tripped condition. Thismeets the 2/4 logic to cause a reactor trip. Additionally the signal for2/4 power range channels above P-10 will cause the SR channels todeenergize. Page 77 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01101/#6, RO-C-01300/#RO-C-01300-E12

REFERENCE:

RO-C-01100, TP-52; SOD-01300-002; SOD-01300-004KA - 012000 K6.02 Reactor Protection System Knowledge of the effect of a loss or malfunction of the following will have on the RPS:Redundant channels RO - 2.9 SRO - 3.1 CFR - 41.7 / 45.7 KA Justification - Question tests the knowledge of how the loss of redundant Channels(N42 and N32) will impact the RPS (RX Trip and De-energize theother SR channel N31).

Original Question # - RO26 Audit-16, Cook 2006 NRC Exam -41 Question AUDITRO22-BOTH-23 Q#20Original Question KA - SYS 015K4.01 Page 78 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

40. 040 002/BOTH/OK/DIRECT/MASTER 01056C0007-8/013000 K4.04/4.3/4.5/F/3Given the following conditions on Unit 2: Reactor power is 3% East Main Feedwater Pump is in service Both MDAFW Pumps have been stopped with control switches in NEUTRAL Which ONE of the following signals will cause an automatic start of the MDAFWPumps?AMSAC East Main Feedwater Pump Trip Safety Injection East Main Feedwater Pump Trip Safety Injection Blackout Sequence Blackout Sequence Steam Generator Low Level of 26% on 1 of 4 SGsA.B.C.D.ANSWER: CA - INCORRECT. AMSAC Bypassed at <40% power. Plausible as both are auto startfeatures of AFW but they are not in service with the described plantconditions.B - INCORRECT. MFP Auto Start only available in AUTO. Plausible as both are autostart features of AFW but they are not in service with the describedplant conditions.C - CORRECT. Safety Injection and Blackout will start AFW Pps in Neutral or AUTO.D - INCORRECT. Requires 1/4 SG Levels low-low (<22%) for AUTO start. Plausible asboth are auto start features of AFW but one of the signals set pointsis not correct. Page 79 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05600/#7

REFERENCE:

RQ-C-KNOW KA - 013000 K4.04 Engineered Safety Features Actuation System (ESFAS)

Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for thefollowing:Auxiliary feed actuation signalRO - 4.3 SRO - 4.5 CFR - 41.7 KA Justification - Question tests knowledge of which ESFAS signals (Interlocks) willcause AFW actuation.Original Question # - Master Bank 01056C0007-8 Original Question KA - Unknown Page 80 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

41. 041 002/BOTH/OK/DIRECT/NRC EXAM 2006-002-5/014000 A4.01/3.3/3.1/H/3During a power ascension, with reactor power at 48%, Control Bank C - Group 1 rodB-8 drops. Prior to the drop it was at 230 steps. While restoring the rod, a control rodurgent failure alarm occurs.

Which one of the following explains why the alarm actuated?All other Bank C - Group 1 rod lift coil disconnect switches are open.All Bank C - Group 2 rod lift coil disconnect switches are open.The step counter of the pulse to analog (P/A) converter was not reset to 0.Group C rod moving with group D rods withdrawn.A.B.C.D.ANSWER: BA - INCORRECT. While all other Bank C Group 1 rods lift coils deenergized, the Alarmis generated from the failure of Group 2 movement (System monitorscurrent through the lift coils - Since Bank C group 1 rod B-8 still hascurrent the alarm is from group 2)B - CORRECT. Since the dropped rod is completely inserted, the lift coil disconnectswitches for all operable rods within the affected bank are opened.An Urgent failure will occur when the misaligned rod begins to move.This is caused by the non-movement of the group without themisaligned rod.C - INCORRECT. While the P/A Converter is reset during rod recovery, failure to do sowould not cause an urgent failure.D - INCORRECT. Group C is moved in the bank select mode. This would not cause anurgent failure alarm. Page 81 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D8/#Ro-C-AOP0240412-E1, RO-C-01200 /#4

REFERENCE:

2-OHP-4024-210, Annunciator #210 Response: Flux Rod, Drop26 Rod Control Urgent Failure, 2-OHP-4022-012-005KA - 014000 A4.01 Rod Position Indication System (RPIS)Ability to manually operate and/or monitor in the control room:Rod selection control RO - 3.3 SRO - 3.1 CFR - 41.7 / 45.5 to 45.8 KA Justification - Requires the ability to monitor and verify proper rod position responseduring a dropped rod recovery based on the rod bank/disconnectswitch alignment (rod selection control). Original Question # - Cook 2006 NRC Exam -002-5, INPO # 27278 Ginna1-4/27/2004Original Question KA - 000003 AA1.02, 000003AK2.05 Page 82 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

42. 042 004/BOTH/OK/DIRECT/MASTER AOP1CAOP2.1-4/016000 K1.04/2.7/2.7/F/3Which set of the following describes the response of the reactor protection system to acontrolling feedwater flow instrument failing low with no operator action from 100%power conditions?

1) Turbine/Reactor trip on Low-Low level in associated steam generator2) Turbine/Reactor trip on High-High level in associated steam generator3) Feedwater Conservation

4) Feedwater Isolation2 AND 31 AND 32 AND 41 AND 4A.B.C.D.ANSWER: CA - INCORRECT. The Turbine will trip on High-High level but a FW conservation signalwill not be received. Plausible since the first portion of the distractoris correct and if student believes failure initiates feedwaterconservation (logical since indication shows low feedflow) then this isa logical answer.B - INCORRECT. The SG level will not go low. Plausible since the FW flow goes lowand a steam flow failure would cause this response. Plausible ifstudent believes failure actually results in feedwater flow reduction(logical since indication shows low feedflow) then this is a logicalanswer.C - CORRECT. The FW valves will open when the FW flow instrument fails low,causing actual level to rise to the High-High SG setpoint causing aTurbine trip and FW Isolation.D - INCORRECT. The SG level will not go low. Plausible since the FW flow goes lowand a steam flow failure would cause this response. Plausible ifstudent believes actual reduction in feedwater flow for first portion ofdistractor and second portion of distractor occurs in a Steam Flowtransmitter failure (which is very similar failure). Page 83 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D10/#RO-C-AOP0390412-E1

REFERENCE:

SOD-05100-001, SGWLC KA - 016000 K1.04 Non-Nuclear Instrumentation System (NNIS)

Knowledge of the physical connections and/or cause-effect relationships between theNNIS and the following systems:

MFW System RO - 2.7 SRO - 2.7 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Question tests knowledge of NNIS (SG level & Flow) response &actuations caused by a FW system malfunction.

Original Question # - Master Bank MASTER AOP1CAOP2.1-4, CM-1127-31944 Page 84 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

43. 043 003/BOTH/OK/DIRECT/NRC EXAM-2006-045-5/022000 K3.02/3.0/3.3/H/3Given the following conditions:

• Unit 2 is in Mode 3 at rated temperature and pressure awaiting a startup.

• Lower Containment Cooling NESW supply is throttled to all ventilation units.

• A power failure causes a loss of 4 of the 8 Lower Containment Vent units.

• Average containment temperature rises from 100°F to 119°F.

• Charging Flow Control is in MANUAL

• Assume RCS Pressure and Temperature remain Constant. Which ONE of the following describes the change in indicated Pressurizer level due tothe rise in Containment temperature?

Density lowering in the _______ leg causes indicated pressurizer level to read________ than actual level.reference; higher reference; lower variable; higher variable; lowerA.B.C.D.ANSWER: AA - INCORRECT. Pressurizer Level uses a wet reference leg DP level indicator. Thiscompares the pressure of the full reference leg with the pressure ofthe actual water in the pressurizer. When these are equal the levelindicates 100%. As the temperature in Containment and thereforethe reference leg rises the density & weight of the reference leglowers. This means that the level in the pressurizer will indicatehigher for the same initial actual level.B - CORRECT. Indicated level will be higher than actual level.C - INCORRECT. Reference leg density lowers.

D - INCORRECT. Indicated level will be higher than actual level. Reference Legdensity lowers. Page 85 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-GF27/9d

REFERENCE:

RO-C-GF27, Sensors and Detectors pg. 51 & 52KA - 022000 K3.02 Containment Cooling System (CCS)

Knowledge of the effect that a loss or malfunction of the CCS will have on the following:Containment instrumentation readings RO - 3.0 SRO - 3.3 CFR - 41.7 / 45.6 KA Justification - Requires the knowledge of the effect a malfunction of thecontainment cooling system will have on the pressurizer levelinstruments located in containment.Original Question # - INPO # 27486 Harris 1 - 3/24/2004, Similar to Cook NRCExam -2006-045-5 : INPO # 26772 Kewaunee, Unit 1 -

2/2/2004 Page 86 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

44. 044 001/BOTH/OK/NEW/NEW/025000 A1.03/2.5/2.5/H/3Given the following conditions on Unit 1:

• Unit is in Mode 1 at 100% power

• A control air leak has resulted in isolation of glycol to containment.

• WIN team states that it will take approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to repair the control air leak.Which ONE of the following describes the operating implications of the loss of glycol tocontainment.Immediately declare the Ice Bed Inoperable. Monitor Ice Bed temperatures toensure they remain < 27°F every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.Monitor Ice Bed temperatures to ensure they remain at an acceptable level. Enterappropriate Tech Spec actions if Ice Bed temperature rises to > 27°F Start all Unit 1 Air Handling Units (AHUs). Monitor Ice Bed Temperature locallyonce per hour until glycol system is restored.

Maximize Containment cooling. If glycol cannot be restored within ONE hourdeclare the Ice Bed Inoperable.A.B.C.D.ANSWER: BA. INCORRECT. Technical Specifications requires that temperatures are maintained

<27°F but the loss of glycol alone does not require Tech Spec entryB. CORRECT. Monitoring temperatures to ensure that they remain <27 is all that isrequired. The loss of glycol alone does not require Tech Spec entryC. INCORRECT. The AHUs are generally stopped if glycol is lost, They would NOT bestarted. Monitoring temperatures to ensure that they remain <27 isrequired but a one hour frequency is NOT required and temperaturesmay be monitored from the Control Room.D. INCORRECT. Maximizing Containment Cooling may help slightly but is not required.Technical Specifications requires that temperatures are maintained <27°F but the loss of glycol alone does not require Tech Spec entryNote: The Ice Condenser is sufficiently subcooled and insulated suchthat a significant temperature rise will not be observed forseveral days following the loss of cooling. Page 87 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-TS01/#11

REFERENCE:

TS 3.6.11; SD-01000, pages 9 &10 KA - 025000 A1.03 Ice Condenser SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding designlimits) associated with operating the Ice Condenser System controls including:Glycol flow to ice condenser air handling units RO - 2.5 SRO - 2.5 CFR - 41.5 / 45.5 KA Justification - Questions tests operator knowledge of what actions are required andwhat monitoring is required based on loss of glycol cooling. There isalso an element of prediction in that the operator needs to predict therate of temperature rise and realize that immediate TS actions are notrequired.Original Question # - NEW Original Question KA - NEW Page 88 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

45. 045 001/BOTH/OK/NEW/NEW/026000 K2.02/2.7/2.9/H/3Given the following conditions on Unit 1:

• A Large Break LOCA has occurred 10 minutes ago

• 600 VAC buses 11C and 11D de-energized on the trip

• All other systems function as desired.

• Containment pressure is 5.0 psig and risingWhich ONE of the following describes the current status of the CTS Pump discharge valves?ALL CTS Pump Discharge valves are OPEN IMO-210, East CTS Pump Discharge is CLOSED IMO-211, East CTS Pump Discharge is CLOSED Both West CTS Pump Discharge valves are OPEN IMO-220, West CTS Pump Discharge is CLOSED IMO-221, West CTS Pump Discharge is CLOSED Both East CTS Pump Discharge valves are OPEN IMO-211, East CTS Pump Discharge is CLOSED IMO-210, East CTS Pump Discharge is OPEN IMO-221, West CTS Pump Discharge is CLOSED IMO-220, West CTS Pump Discharge is OPENA.B.C.D.ANSWER: BA - INCORRECT. The Loss of 600VAC Bus 11D will cause the discharge valves forthe East CTS pump to lose power will prevent them from opening.This would be true if the valves were initially open.B - CORRECT. The Loss of 600VAC Bus 11D will cause the discharge valves forthe East CTS pump to lose power will prevent them from opening. C - INCORRECT. The East Valves will not be open. This would be true if Bus 11A waslostD - INCORRECT. Only the East train valves have lost power. This is plausible sincethe CTS pumps have two discharge valves in parallel and theoperators may assume they have crossed power supplies to ensurea flowpath. Page 89 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00900/RO-C-00900-E9, RO-C-00900-E12

REFERENCE:

RO-C-00900 pg. 10 & 20, SOD-00900-001 KA - 026000 K2.02 Containment Spray System (CSS)

Knowledge of bus power supplies to the following:

MOVs RO - 2.7 SRO - 2.9 CFR - 41.7 KA Justification - Question tests knowledge of which CTS MOVs are impacted by theloss of a bus power supply.Original Question # - NEW Original Question KA - NEW Page 90 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

46. 046 002/BOTH/OK/NEW/NEW/029000 A2.03/2.7/3.1/H/3Given the following conditions in Unit 1:

• Unit is in Mode 5

• Containment pressure is negative 0.3 psig.

• RP has requested that Containment Purge be placed in service.1-OHP-4021-028-005, Operation Of The Containment Purge System, provides adefined sequence of operation due to a concern with Containment Pressure.Which ONE of the following describes this sequence and the reason for the concern while starting up the Containment Purge System? Start one Purge Supply fan then open the supply fan valve since TechnicalSpecifications require Containment pressure to be < 0 PSIG at all times.Prior to starting the fans, open the Upper Containment Purge Supply valves toprevent Ice Condenser doors from opening when initiating containment purge. Prior to starting the fans, open the Lower Containment exhaust fan valves toprevent Ice Condenser doors from buckling when initiating containment purge.Start one Purge Exhaust fan then open the exhaust fan valve to prevent a positivepressure from adversely affecting the radiation monitor operations.A.B.C.D.ANSWER: BA - INCORRECT. T.S. 3.6.1.4, Internal Pressure requires pressure to be -1.5 psig to.03 psig. This action will not maintain pressure low.B - CORRECT. A low pressure in upper containment with respect to lowercontainment will cause the Ice Condenser Doors to open.1-OHP-4021-028-005 Attachment 1 step 4.7.4 is performed toraise/equalize upper containment pressure.C - INCORRECT. The buckling concern for the Ice Condenser doors is due to unevenfloor cooling not ventilation fan operation.D - INCORRECT. The radiation monitors will not be affected by minor pressurevariations. Page 91 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-02800/RO-C-02800-T1

REFERENCE:

1-OHP-4021-028-005, Operation Of The Containment PurgeSystem, Attachment 1KA - 029000 A2.03 Containment Purge System (CPS)Ability to (a) predict the impacts of the following malfunctions or operations on theContainment Purge System and (b) based on those predictions, use procedures tocorrect, control, or mitigate the consequences of those malfunctions or operations:Startup operations and the associated required valve lineupsRO - 2.7 SRO - 3.1 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - The question tests the ability of the operator to determine to correctprocedural actions required and the consequences of not followingthose actions (Predicts impacts of incorrect purge operations andprevents impacts through correct sequence)Original Question # - NEW Original Question KA - NEW Page 92 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

47. 047 002/BOTH/OK/NEW/NEW/035000 K6.02/3.1/3.5/H/3Given the following condition on Unit 2:

• Unit is operating at 100% power when an inadvertent Steam Line Isolation occurred.

• Immediately following the isolation, and resultant plant response, the operators notethat all the SG PORVs failed to open.One minute following the Steam Line Isolation, which ONE of the following describesthe maximum expected SG pressures?1025 psig 1040 psig 1065 psig 1085 psigA.B.C.D.ANSWER: CA - INCORRECT. This is the Normal PORV Setpoint and would be the expectedpressures of the other SGs.B - INCORRECT. This is the PORV setpoint used in the SG tube rupture proceduresfor the faulted SG.C - CORRECT. Following the Steam Line Isolation a Rx trip would be expected dueto OTDT. SG pressures would initially surge opening most of thesafeties but as the RCS cooled down pressures would stabilize onthe lowest safety valve setpoint (1065 psig) due to the reduction inReactor Power and Decay heat during the initial 30 seconds of theevent. D - INCORRECT. The pressures may initially surge to this level but would quickly (lessthan 30 seconds) drop after the Rx trip.Note - SG Safety valve setpoints:

SV-1A 1065 psig SV-1B 1065 psig SV-2A 1075 psig SV-2B 1075 psig SV-3 1085 psig Page 93 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05103/#RO-C-05103 -E2, #RO-C-05103-E6

REFERENCE:

RO-C-05103 pg. 11-12 KA - 035000 K6.02 Steam Generator System (S/GS)

Knowledge of the effect of a loss or malfunction of the following will have on the SGs:Secondary PORV RO - 3.1 SRO - 3.5 CFR - 41.7 / 45.7

KA Justification - Requires knowledge of how the loss of a PORV and subsequent SGStop Valve closure will impact the SG pressure. Higher order based on a requirement to determine that the Rx will tripon SLI due to OTDT.Original Question # - NEW Original Question KA - NEW Page 94 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

48. 048 004/BOTH/OK/DIRECT/MAST 01RXTC1021-13/039000 K5.08/3.6/3.6/F/2Given the following conditions:

• Reactor power is 50% with a negative moderator temperature coefficient

• Control Rods are in manual. If steam flow is raised by 5 percent, which ONE of the following statements bestdescribes how reactor power will respond to the change? Reactor power will:decrease to a new lower value.

increase temporarily, then return to its initial value.increase to a new higher value.

decrease temporarily, then return to its initial value.A.B.C.D.ANSWER: CA - INCORRECT. This would be the response of RCS temperatureB - INCORRECT. This would be true for a rise in Rod position C - CORRECT. The increased Steam flow will cause RCS temperature to lower andadd positive reactivity due to the negative MTCD - INCORRECT. This would be true for a rise Boron ConcentrationLESSON PLAN/OBJ: RO-C-GF10/#21

REFERENCE:

RO-C-GF10 KA - 039000 K5.08 Main and Reheat Steam System (MRSS)

Knowledge of the operational implications of the following concepts as they apply to theMRSS:

Effect of steam removal on reactivity RO - 3.6 SRO - 3.6 CFR - 41.5 / 45.7 KA Justification - Question tests knowledge of power rise (operational Implication) dueto a rise in steam flow causing positive reactivity feedback.Original Question # - Master Bank 01RXTC1021-13 Page 95 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

49. 049 002/BOTH/OK/DIRECT-REPEAT/NRC EXAM 2007-52/041000 K2.01/2.8/2.9/F/3Which ONE of the following power supply failures would allow the steam dump systemto continue to operate?CRID II CRID III 250 VDC Bus VDAB 250 VDC Bus VDCDA.B.C.D.ANSWER: BA - INCORRECT. CRID II powers the Steam Dump ControllersB - CORRECT. CRID III does Not supply power to the Steam Dumps or relays.C - INCORRECT. 250 VDC Bus VDAB powers 1 train of Steam Dump SolenoidsD - INCORRECT. 250 VDC Bus VDCD powers 1 train of Steam Dump SolenoidsLESSON PLAN/OBJ: RO-C-05200/#4

REFERENCE:

RO-C-05200 Steam Dump System pg. 15-16 KA - 041000 K2.01 Steam Dump System (SDS) and Turbine Bypass Control Knowledge of bus power supplies to the following:

ICS, normal and alternate power supply RO - 2.8 SRO - 2.9 CFR - 41.7 KA Justification - Question tests knowledge of all of the Steam dump power supplies byrequiring the operator to identify the one that doesn't supply power.Original Question # - Master Bank 01052C0002-4, NRC EXAM 2007-52Original Question KA - 041000 K2.01 Page 96 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

50. 050 004/BOTH/OK/MODIFIED/PRARILND-4232004-14/056000 2.2.44/4.2/4.4/F/3Unit 1 has just completed a Heatup and is preparing for a Reactor Startup.You have been directed to open the FW pump Emergency Leak Offs (ELOs).Which ONE of the following describes the indications/systems that you should checkprior to opening these valves and why?Ensure that the FW pump is reset to allow the ELOs to be opened.Ensure that the FW pump oil system is operating and has been warmed tominimize the effects of cold seal water on FW pump bearings. Ensure that the FW pump oil system is operating to prevent damage to the FWpumps due to condensate flow spinning the pumps.

Ensure that the FW pump has been removed from turning gear to prevent damageto the turning gear motor.A.B.C.D.ANSWER: CA - INCORRECT. The FW turbine needs to be tripped (Stop valve closed) to allow theELOs to be fully closed. They will position based on flow if the stopvalve is opened. Plausible as this is a true statement but notrequired for the action being described.B - INCORRECT. The oil system operation would have minimal impact on the sealwater temperature. Plausible as the oil system is required to be inservice for this evolution but not for the described purpose.C - CORRECT. Placing flow though the FW pumps (opening recirculation valves)causes the turbine and pump to rotate at > 100 rpm and so the oilsystem is required for bearing protection.D - INCORRECT. The FW pump will roll off the turning gear (become disengaged)when the ELOs are opened and the motor will not be damaged.Plausible to prevent equipment damage. Page 97 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05500-E8, RO-C-NOP7-E2

REFERENCE:

2-OHP-4021-055-003 KA - 056000 2.2.44 Condensate System Equipment ControlAbility to interpret control room indications to verify the status and operation of asystem, and understand how operator actions and directives affect plant and systemconditions.

RO - 4.2 SRO - 4.4 CFR - 41.10 / 43.5 / 45.12 KA Justification - Question tests ability to determine what the condition of the systemsmust be prior to aligning the FW ELOs and how the required lineupimpacts the equipment.Original Question # - PRARILND-4232004-14 Original Question KA - 056 K1.03 Page 98 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

51. 051 004/BOTH/OK/DIRECT/RO26 AUDIT-41/059000 A1.07/2.5/2.6/H/3The Unit 2 FW Pump Discharge Header Pressure Transmitter 2-FPC-250A slowly driftsLOW during normal plant operation.

This will cause the MFP Speed Control System to generate an indicated FW Delta-Psignal ____(1)_____ than required, causing the main feed pump(s) to______(2)_____. Note: Assume FPC-250A is not identified as failed by DCS. (1) (2) larger speed up larger slow down smaller speed up smaller slow downA.B.C.D.ANSWER: C A - INCORRECT. Steam to FW discharge pressure DP will be smaller. Plausible if thestudent does not know how the DP is being derived.B - INCORRECT. Steam to FW discharge pressure DP will be smaller. The controllerwill raise FW pump Speed. Plausible if the student does not knowhow the DP is being derived.C - CORRECT. The Main FW Pump Speed control compares the UPC-102A/B(highest) steam header pressure to the FW pump Dischargepressure FPC-250A/B (lowest). The speed control attempts tomaintain the Main FW Pump speed such that the FW header toSteam Header DP is on Program. When the FW DischargePressure drifts Low, it will appear that a smaller DP exists which willraise FW pump speed to try to raise FW pump Discharge headerpressure. D - INCORRECT. The controller will raise FW pump Speed. Plausible if the studentdoes not know how the impact of DP effects SGFP controls. Page 99 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05501/#RO-C-05501-E5

REFERENCE:

TS3000, DCS BODD, Page 4-7 KA - 059000 A1.07 Main Feedwater (MFW) SystemAbility to predict and/or monitor changes in parameters (to prevent exceeding designlimits) associated with operating the MFW System controls including:Feed Pump speed, including normal control speed for ICSRO - 2.5 SRO - 2.6 CFR - 41.5 / 45.5 SCLR - 3SPK K/A Justification - Question tests ability to predict changes associated with the FWPump Speed.Original Question # - Modified from NRC EXAM 2007-054 (UPC failure to FPCfailure, Failure to Low & Updated due to DCS), Cook 2006 NRCExam -COOK06-54 , Master Bank 01055C0008-5, RO26 AUDIT-41Original Question KA - 059 K4.05 Page 100 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

52. 052 003/BOTH/OK/DIRECT/RO24 AUDIT 044-12/059000 A4.03/2.9/2.9/F/2Which ONE of the following describes the functional relationship with respect tocontrolling steam generator (SG) levels between the Main Feedwater Pumps (MFPs)and the Main Feedwater Regulating Valves (MFRVs) when the unit is ramping from50% to 100% power? The MFPs maintain a variable differential pressure across the MFRVs, while theMFRVs throttle to maintain a constant SG water level. The MFPs maintain a constant differential pressure across the MFRVs, while theMFRVs throttle to maintain a variable SG water level. The MFPs maintain a variable differential pressure across the MFRVs, while theMFRVs throttle to maintain a variable SG water level.The MFPs maintain a constant differential pressure across the MFRVs, while theMFRVs throttle to maintain a constant SG water level. A.B.C.D.ANSWER: AA - CORRECT. The design of the SGWLC system is to maintain a constant level inthe SGs at all power levels. The MFW control system howevervaries the programming to maintain an optimum DP across theMFRVs.B - INCORRECT. The DP is not constant it varies with program while the level is heldconstant. Plausible due to second portion of the distractor beingcorrect and the student must know the SGFP varies the D/PC - INCORRECT. The Level is held constant. Plausible as the first portion of thedistractor is correct and the student must know SG level is constantfor all power levels which is unique in Westinghouse plantsD - INCORRECT. The DP is not constant it varies with program. Plausible as thesecond portion of the question is correct and the student must knowthe SGFP varies the D/P. Page 101 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05100/#3&6

REFERENCE:

SOD-05100-001, RO-C-05100 Steam Generator SystemKA - 059000 A4.03 Main Feedwater (MFW) SystemAbility to manually operate and/or monitor in the control room:Feedwater control during power increase and decreaseRO - 2.9 SRO - 2.9 CFR - 41.7 / 45.5 to 45.8 KA Justification - Requires knowledge of how to monitor and control feedwater flow tothe SGs during a power escalation.Original Question # - RO24 AUDIT 044-12 Original Question KA - 059 A3.02 Page 102 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

53. 053 002/BOTH/OK/DIRECT/RO26 AUDIT-44/061000 K3.02/4.2/4.4/H/3Given the following conditions on Unit 2:

• Unit is in Mode 1.

• The TDAFW Pump is tagged out of service.

• A Loss of Feedwater causes a reactor trip.

• Coincident with the trip, T21D Differential trip actuates. Which ONE of the following describes the Auxiliary Feedwater alignment andapproximate flow rates? 1 and 4 SGs being fed at 120 x 10 3 pph each 1 and 4 SGs being fed at 240 x 10 3 pph each 2 and 3 SGs being fed at 120 x 10 3 pph each ALL SGs being fed at 120 x 10 3 pph each A.B.C.D.ANSWER: AA - CORRECT. The T21D Differential causes a loss of T21D Bus. With a loss ofT21D, Only the West MDAFW Pump is available. Capacity is ~240 x 10 3 pph, and it is aligned to automatically feed 1 and 4 SGs.B - INCORRECT. Capacity of TDAFW aligned to 2 SGs C - INCORRECT. West MDAFW would be aligned to 1 & 4 SGs, not 2 & 3 SGs(East)D - INCORRECT. This would be the alignment if the TDAFW Pump was the onlyoperating pump. Page 103 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05600/#2 & 3

REFERENCE:

RO-C-05600 Auxiliary Feedwater System pg. 12-14,SOD-08201-001KA - 061000 K3.02Auxiliary / Emergency Feedwater (AFW) SystemKnowledge of the effect that a loss or malfunction of the AFW System will have on thefollowing:

S/G RO - 4.2 SRO - 4.4 CFR - 41.7 / 45.6 K/A Justification - Question asks candidate to identify the amount of AFW flow to thespecific Steam Generators from a single AFW pump. Original Question # - RO26 AUDIT-44 from SEQ2007 Original Question KA - 061000 K6.02 Page 104 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

54. 054 006/BOTH/OK/DIRECT/NRC EXAM 2006-057-5/062000 K1.03/3.5/4.0/H/3The operator incorrectly opens the breaker labeled "7.5 KVA Static Inverter ChannelIV" on 250 VDC distribution panel "MCAB". The operator realizes the mistake andimmediately recloses the breaker.

Which ONE of the following describes the effect of these actions, if any?The alternate power source to the CRID Inverter will be lost when the breaker isreclosed. The CRID will transfer to the 120 VAC from the Regulating Transformer.The alternate power source to the CRID Inverter will be lost. No automatic actionwill occur when the breaker is reclosed. The auto transfer lockout must be reset atthe inverter.

The normal power source to the CRID Inverter will be lost so it will auto transfer tothe alternate source. When the breaker is reclosed, it will auto transfer to thenormal source.

The normal power source to the CRID Inverter will be lost so it will auto transfer tothe alternate source. When the breaker is reclosed, the auto transfer lockout mustbe reset at the inverter.A.B.C.D.ANSWER: CA - INCORRECT. The Alternate source will not be lost. The normal DC supply will berestored and the Inverter will re-transfer to the normal source.Plausible due to continuity of power remains to the panel through outthe evolution B - INCORRECT. The Alternate source will not be lost. The normal DC supply will berestored and the Inverter will re-transfer to the normal source. Plausible assuming student believes the panel will lose power uponinadvertent operation of the supply breaker which is a validassumption with a DC vital breaker.C - CORRECT. The static transfer switch provides a virtual zero time transfer to thealternate source in case of inverter failure. Thirty seconds after thestatic switch transfer event ceases and all system parameters arenormal, the static switch automatically re-transfers the load to theinverter, without power interruption. D - INCORRECT. The normal DC supply will be restored when the breaker is closedand the Inverter will re-transfer to the normal source. Plausible asthe restoration of power is correct if the logic contained and autotransfer lockout which does exist on several other plant electricalcomponents. Page 105 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-08203/#3e, #3g

REFERENCE:

RO-C-08203, Instrumentation Electrical SystemKA - 062000 K1.03A.C. Electrical Distribution SystemKnowledge of the physical connections and/or cause-effect relationships between theA.C. Distribution System and the following systems:DC distribution RO - 3.5 SRO - 4.0 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Question tests for knowledge of how the DC system connects andsupports the AC distribution system (feeds CRID) and thecause-effect relationship due to breaker manipulations.Original Question # - Cook 2006 NRC Exam - 057-5, Bank 01082C0303-2, CM-7852-38509Original Question KA - 063 K4.01, 062000 A3.04 Page 106 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

55. 055 005/BOTH/OK/DIRECT/DIABLO2007-050-1/063000 K4.04/2.6/2.9/H/2Which ONE of the following describes the effect on a closed Circulating Water pumpbreaker if DC control power is lost to the breaker?The breaker immediately trips open and cannot be reclosed until control power isrestored.

The breaker can be tripped from the Control Room but automatic trip functions arenot operable.

Automatic trips are not operable and tripping the breaker from the Control Room isnot possible.

Automatic breaker trips are operable but tripping the breaker from the ControlRoom is not possible. A.B.C.D.ANSWER: CA - INCORRECT. The breaker has stored energy in the spring, but it can not bereleased due to the loss of power. Plausible since many signals(RPS) require power to maintain contacts open.B - INCORRECT. The breaker has stored energy in the spring, but it can not bereleased due to the loss of power. Plausible since manysignals(RPS) require power to maintain contacts open and generatetrip on loss of power.C - CORRECT. A loss of DC control power will prevent breaker operations with thecontrol switch (and trip functions)D - INCORRECT. While it is true that the spring has stored energy, the spring releasemechanism can not release the spring to cause the trip. Page 107 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05700-E11, RO-C-08204-E1, RO-C-AOP0550412-E1

REFERENCE:

RO-C-05700 TP-36 KA - 063000 K4.04 D.C. Electrical Distribution System Knowledge of D.C. Electrical System design feature(s) and/or interlock(s) which providefor the following:

Trips RO - 2.6 SRO - 2.9 CFR - 41.7 KA Justification - Question requires knowledge of the design features provided by DCpower for generating a Trip signal.Original Question # - DIABLO2007-050-1 Original Question KA - 063 K4.04 Page 108 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

56. 056 003/BOTH/OK/DIRECT/RO26 AUDIT-48/064000 2.4.50/4.2/4.0/H/3Given the following conditions on Unit 1:

• Bus T11B normal supply breaker has opened.

• DG1AB Diesel Generator has started and is tied to the bus.* Ann. 118, Drop 53, DG1AB TRIPS DISABLED is LITWhich ONE of the following conditions will automatically trip the diesel generator?Engine Speed of 590 rpm CO 2 actuating in the EDG RoomMain Bearing Temperature 198 FLow Lube Oil Pressure 23 psigA.B.C.D.ANSWER: AA - CORRECT. EDG is in Emergency Mode so Overspeed Trip is the only oneavailable, 590 rpm is 114.7% of Normal 514 rpm - Trip at 110%.B - INCORRECT. CO 2 is trip but not in emergency mode.C - INCORRECT. Main bearing temp of >195 is normal tripD - INCORRECT. Lube oil pressure of <25 psig is normal tripNote: LOOP or SI places EDG in Emergency Mode and blocks 7non-emergency trips. Page 109 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-03200/#10

REFERENCE:

RO-C-03200 Emergency Diesel Generators KA - 064000 2.4.50 Emergency Diesel Generator (ED/G) System Emergency Procedures/PlanAbility to verify system alarm setpoints and operate controls identified in the alarmresponse manual.

RO - 4.2 SRO - 4.0 CFR - 41.10 / 43.5 / 45.3 KA Justification - Requires the ability to determine proper diesel trip setpoint andunderstand (monitor) conditions that will trip the diesel generator forthe given plant conditions.Original Question # - RO23 Audit -059-5 (Q#54), RO26 Audit-48Original Question KA - 064000 K4.01 Page 110 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

57. 057 001/BOTH/OK/NEW/NEW/064000 K4.04/3.1/3.7/F/3Which ONE of the following describes a condition that would cause a loadconservation signal to be generated for the DG1AB.Train A SI and Train A Load Shed Train B SI and Train B CTS Train A SI and Train A CTS Train B SI with a Loss of Offsite PowerA.B.C.D.ANSWER: DA - INCORRECT. Train A would cause a Load Conservation on DG1CD.B - INCORRECT. Should be LOOP or Load Shed with SI. Plausible since startingCTS provides additional loads and NESW Pump response isdifferent with CTS actuation.C - INCORRECT. Train A would cause a Load Conservation on DG1CD. Should beLOOP or Load Shed with SI. Plausible since starting CTS providesadditional loads and NESW Pump response is different with CTSactuation.D - CORRECT. An SI with a LOOP (and subsequent Load Shed) will generate aLoad Conservation signal. Train B is associated with DG1AB.LESSON PLAN/OBJ: RO-C-08201/#RO-C-08201-E6

REFERENCE:

RO-C-08201 KA - 064000 K4.04 Emergency Diesel Generator (ED/G) System Knowledge of ED/G System design feature(s) and/or interlock(s) which provide for thefollowing:

Overload ratings RO - 3.1 SRO - 3.7 CFR - 41.7 KA Justification - A load conservation signal is generated to prevent EDG overloading. This question requires knowledge of the design feature (loadconservation) that prevents overloading of the EDG.Original Question # - New Original Question KA - New Page 111 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

58. 058 002/BOTH/OK/DIRECT/RO25 AUDIT-59/068000 K4.01/3.4/4.1/F/3Which ONE of the following lists the two conditions that will independently causetermination of a liquid release by closing 12-RRV-285, Liquid Waste Disposal EffluentDischarge Header Shutoff Valve, and/or tripping the operating monitor tank pump?Low circulating water flow High radiation sensed in the release header Low circulating water flow High radiation sensed in the circulating water flow Low release header radiation monitor sample flow High radiation sensed in the release header High release header radiation monitor sample flow High radiation sensed in the circulating water flowA.B.C.D.ANSWER: CA - INCORRECT. Even though there is a requirement to have adequate circulatingwater flow, there is no trip for RRV-285 due to low flow conditions.B - INCORRECT. Even though there is a requirement to have adequate circulatingwater flow, there is no trip for RRV-285 due to low flow conditions. The radiation monitor senses radiation levels on the actual releaseline, not the Circ Water system.C - CORRECT. RRS-1001 High alarm sensed on the actual release line willenergize R18-AUX & R18-AUX1 which closes RRV-285 and tripsthe monitor tank pumps. In addition, either high or low sample flow(less than 20% or greater than 90%) will energize R18-AUX &R18-AUX1.D - INCORRECT. The radiation monitor senses radiation levels on the actual releaseline, not the Circ Water system. Page 112 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-02200/ #5, #8

REFERENCE:

12-OHP-4024-139, Drop 18, OP-12-98810, OP-12-98313,OP-12-98276KA - 068000 K4.01 Liquid Radwaste System (LRS)

Knowledge of Liquid Radwaste System design feature(s) and/or interlock(s) whichprovide for the following:

Safety and environmental precautions for handling hot, acidic, and radioactive liquidsRO - 3.4 SRO - 4.1 CFR - 41.7 KA Justification - Question tests knowledge of Liquid Radwaste System design featuresand interlocks which provide for the isolation of radioactive liquids toprevent excessive radioactive discharge to the environment.Original Question # - AUDIT RO22-SRO-9, RO25 AUDIT-59Original Question KA - 068000 K4.01 3.4/4.1 CFR 41.7 Page 113 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

59. 059 003/BOTH/OK/MODIFIED/NRC EXAM 2007-19/072000 A3.01/2.9/3.1/F/3Which ONE of the following lists the Unit 1 Control Room Ventilation system damperalignment for operation during a high alarm on ERS-7401, U1 Control Room AreaRadiation Monitor? 1-HV-ACR-DA-1/1A 1-HV-ACR-DA-2 1-HV-ACR-DA-2A 1-HV-ACR-DA-3 Outside air to CR Outside air to CR PRZN Outside air to CR PRZN CR air to PRZN OPEN CLOSED PARTIAL OPEN OPEN CLOSED PARTIAL OPEN CLOSED OPEN OPEN PARTIAL OPEN CLOSED CLOSED CLOSED CLOSED PARTIAL OPEN CLOSEDA.B.C.D.ANSWER: BA - INCORRECT. Dampers 1/1A will be closed on an ERS-7401 high alarm.B - CORRECT. On an ERS-7401 high alarm: Damper 1/1A will be closed; Damper2 will be partially open; Damper 3 opens. C - INCORRECT. Damper 1/1A will be closed and Damper 3 will remain open on anERS-7401 high alarm.D - INCORRECT. Damper 3 will remain open on an ERS-7401 high alarm.LESSON PLAN/OBJ: RO-C-02801A/#8

REFERENCE:

SOD-02801A-001 Modified: Changed stem to radiation alarm (vs. fire) which changed the correctanswer to B (vs. D)KA - 072000 A3.01Area Radiation Monitoring (ARM) SystemAbility to monitor automatic operation of the ARM system, including:Changes in ventilation alignment RO - 2.9 SRO - 3.1 CFR - 41.7 / 45.5 KA Justification - Question tests ability to monitor changes in the Control RoomVentilation dampers caused by a high radiation alarm.Original Question # - MASTER 01028C01A02-6, NRC EXAM 2007-19Original Question KA - 000067 AA2.02 Page 114 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

60. 060 002/BOTH/OK/DIRECT/RO23-AUDIT-074-3/073000 A2.01/2.5/2.9/F/3Given the following conditions on Unit 2:

• The East CCW HX is in service with the West CCW Pump running.

• CCW Surge Tank level is stable.

• CRS-4301, East CCW HX Radiation Monitor, generates an External Failure Alarmdue to a faulty power supply Which ONE of the following describes the response of the CCW system for the givenconditions and the subsequent operator actions required?No automatic actions will occur since the West CCW pump is running.Notify RP of the failed CRS-4301, East CCW HX Radiation Monitor.No automatic actions will occur since the CRS-4401, West CCW HX RadiationMonitor is still functioning.

Split the CCW Trains with Misc Header on the West Train and isolate the EastTrain.

2-CMO-420, West CCW HX Outlet, opens and 2-CMO-410, East CCW HX Outlet,closes.

Remove CRS-4301, East CCW HX Radiation Monitor, from service and re-alignCCW flow through the West CCW Hx ONLY.

2-CRV-412, CCW Surge Tank Vent Valve, will automatically close.Notify RP to remove CRS-4301, East CCW HX Radiation Monitor, from service,then reopen 2-CRV-412.A.B.C.D. Page 115 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: DA - INCORRECT. Either radiation monitor will close the CCW vent valve, regardless ofthe operating pump. Plausible based on assumption that the activeportion of the system is being monitored.B - INCORRECT. An EXTERNAL FAIL in either CRS-4301 (East CCW Header) ORCRS-4401 (West CCW Header) will close the CCW Vent valve.Plausible based on assumption that the active portion of the systemis being monitored.C - INCORRECT. CCW Rad Monitors do not cause auto re-alignment of the CCWsystem. Plausible based on maintaining flow in the active portion ofthe system.D - CORRECT. An EXTERNAL FAIL in either CRS-4301 (East CCW Header) ORCRS-4401 (West CCW Header) will close the CCW Vent valve.LESSON PLAN/OBJ: RO-C-01600/RO-C-01600-E6

REFERENCE:

12-OHP-4024-139 #29 KA - 073000 A2.01 Process Radiation Monitoring (PRM) SystemAbility to (a) predict the impacts of the following malfunctions or operations on the PRMSystem and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:Erratic or failed power supply RO - 2.5 SRO - 2.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Question tests ability to predict the impact of the loss of power on aprocess rad monitor, and the actions the operator should take inresponse to the failure.Original Question # - RO23-AUDIT-074-3 Original Question KA - SYS 073 A2.02 Page 116 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

61. 061 002/BOTH/OK/DIRECT/01EOPC1412-1/076000 2.4.6/3.7/4.7/F/2The operators are attempting to energize emergency bus T21A from DG2AB during aloss of all AC power.

Which ONE of the following states the equipment switch that would NOT be placed inPULL TO LOCKOUT and the reason why?ESW pump to ensure diesel cooling.

CCW pump to ensure cooling to vital loads.

MDAFW pump to ensure an adequate heat sink is maintained.CTS pump to ensure containment integrity is not challenged.A.B.C.D.ANSWER: AA - CORRECT. Since the EDG is the probably source of power, ESW to the EDGneeds to be available immediately to maintain the EDG cooled onceit starts and is loaded.B - INCORRECT. CCW Pumps are locked out. Plausible as students know CCW isrequired to maintain RCP seals cooled and prevent SBLOCA.C - INCORRECT. MDAFW Pumps are locked out. Plausible as students know Heatsink is a significant contributor to risk.D - INCORRECT. CTS Pumps are locked out. Plausible as students know a LOCA isa risk due to a loss of LOOP, and CTS would be required toaddress Containment Pressure.LESSON PLAN/OBJ: RO-C-EOP14/12

REFERENCE:

OHP-4023-ECA-0.0, ATT A KA - 076000 2.4.6 Service Water System (SWS)

Emergency Procedures/Plan Knowledge of EOP mitigation strategies.

RO - 3.7 SRO - 4.7 CFR - 41.10 / 43.5 / 45.13 KA Justification - Question requires knowledge of how the ESW system (SWS) isoperated during power restoration during an EOP event (mitigation ofa Loss of ALL AC).Original Question # - 01EOPC1412-1 Original Question KA - EPE:055 EK3.02 Page 117 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

62. 062 002/BOTH/OK/DIRECT/RO25 AUDIT-37/076000 A1.02/2.6/2.6/F/3A malfunction of the NESW system has resulted in a partial loss of NESW flow to thecontainment ventilation units.

Under this condition which ONE of the following sets of containment readings wouldresult in Unit 2 containment still being within Technical Specification limits?Pressure is 0.18 PSIG, upper containment temperature 97°F, lower containment temperature 115°F.

Pressure is 0.35 PSIG, upper containment temperature 105°F, lower containment temperature 97°F.

Pressure is 0.15 PSIG, upper containment temperature 105°F, lower containment temperature 105°F.

Pressure is -1.55 PSIG, upper containment temperature 97°F, lower containment temperature 115°F.A.B.C.D.ANSWER: AA - CORRECT. LCO 3.6.4 Containment pressure shall be > -1.5 psig and < +0.3psig. LCO 3.6.5 Containment average air temperature shall be: a. >60°F and < 100°F for the containment upper compartment and b. >60°F and < 120°F for the containment lower compartment.B - INCORRECT. Pressure too high & Upper temp too high, plausible since uppertemp is below lower temp limit & Pressure is still reasonable and/or student confuses allowable values for containment temperatureC - INCORRECT. Upper Temp too high. Plausible if candidate doesn't know thatUpper limit is lower and/or if student confuses allowable values forcontainment temperatureD - INCORRECT. Pressure too low. Plausible since temps are in range, pressure isusually a concern only for higher temps and/or if student does notknow allowable value for containment pressure Page 118 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-03400/#9

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T.S. 3.6.4, 3.6.5, 2-OHL-4030-SOM-041 KA - 076000 A1.02 Service Water System (SWS)Ability to predict and/or monitor changes in parameters (to prevent exceeding designlimits) associated with operating the SWS controls including:Reactor and turbine building closed cooling water temperaturesRO - 2.6 SRO - 2.6 CFR - 41.5 / 45.5 KA Justification - Question tests ability to predict/monitor Containmentparameters/temperatures (identify those within design limits)associated with isolation of NESW (SWS) cooling to containment.Original Question # - RO23-090-4-Q72, RO25 Audit-37Original Question KA - SYS022 A1.04 Page 119 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

63. 063 006/BOTH/OK/MODIFIED/NRC EXAM 2004-93-4/078000 A3.01/3.1/3.2/F/3Procedure 2-OHP-4022-064-001, Control Air Malfunction, has been entered basedupon a 100 psi Control Air Pressure alarm.

Which ONE of the following is the correct sequence of events that will occurautomatically as air pressures lower? 95 psig at PPS-10 (20) Standby PAC starts 90 psig CAS wet receiver pressure CAC starts 85 psig at PPS-11 (21) Plant air header isolates

98 psig at PPS-10 (20) Standby PAC starts 97 psig CAS wet receiver pressure CAC starts 95 psig Control Air pressure Plant air header isolates 97 psig CAS wet receiver pressure CAC starts 95 psig at PPS-10 (20) Standby PAC starts 90 psig at PPS-11 (21) Plant air header isolates

95 psig CAS wet receiver pressure CAC starts 90 psig at PPS-11 (21) Plant air header isolates 85 psig at PPS-10 (20) Standby PAC starts A.B.C.D.ANSWER: A AIR SYSTEM SETPOINTS125 psig Air receiver safety valves open 104 psig at PAC discharge PAC surge protection-unloader opens100 psig CAS wet receiver pressure CAC unloads 98 psig at PPS-11(21) Plant air header un-isolates 97 psig Turbine Building air header Alarm "PAC failure / low pressure"95 psig at XPA-100 (100# header) Control air pressure low alarm95 psig at PPS-10 or 20 Stand-by PAC starts 93 psig CAS wet receiver pressure CAC loads 90 psig CAS wet receiver pressure Associated CAC auto-starts85 psig at PPS-11(21) Plant air header isolates 80 psig Control Air Pressure Manual Reactor TripA - CORRECT. See pressures and order aboveB - INCORRECT. Setpoints wrong C - INCORRECT. Setpoints and Order wrong D - INCORRECT. Setpoints and Order wrong Page 120 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-06401/#4

REFERENCE:

SOD-06401-002, PLANT AIR SYSTEM KA - 078000 A3.01 Instrument Air System (IAS)Ability to monitor automatic operation of the IAS, including:Air pressureRO - 3.1 SRO - 3.2 CFR - 41.7 / 45.5 KA Justification - Question tests ability to monitor automatic actions that will occur asair pressure drops.Original Question # - Cook RO24 Audit - 054-13, NRC02-105-2, 04-93-4 Page 121 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

64. 064 004/BOTH/OK/DIRECT/10098-DIA-1999-160/086000 A1.03/2.7/3.2/F/3Given the following conditions on Unit 1:

• Plant heatup is in progress with RCS temperature at 420°F.

• Fire system engineer reports that the fire door at the entrance to the positivedisplacement charging pump and the centrifugal charging pump rooms isnonfunctional.

• Additionally the engineer reports that no other fire system impairments exist. Which ONE of the following is the minimum required action?Establish an hourly fire watch patrol within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.Enter Tech Spec 3.0.3 due to no operable charging pumps.Verify, by inspection, the operability of the manual fire fighting equipment within 1hour.

Close the door and establish a continuous fire watch on at least one side of the firedoor within 15 minutes.A.B.C.D.ANSWER: AA - CORRECT. Hourly fire watch is required since no other impairments exist. Equipment in the area is still operable. Action A.1.1 and A.1.2B - INCORRECT. Hourly fire watch is required. Equipment in the area is stilloperable.C - INCORRECT. No inspection is required. An hourly firewatch meets TRMrequirements.D - INCORRECT. Action A.3.1 and A.3.2 could be performed (review HVAC and closedoor) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OR a continuous fire watch established within 1hour (A.4) but both are not required within 15minutes. Page 122 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-TS01/#11&#13; RO-C-ADM05/#9

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TRM 8.7.10 Fire Rated Assemblies KA - 086000 A1.03 Fire Protection System (FPS)Ability to predict and/or monitor changes in parameters (to prevent exceeding designlimits) associated with operating the Fire Protection System controls including:Fire doors RO - 2.7 SRO - 3.2 CFR - 41.5 / 45.5 KA Justification - Requires the ability to predict the operability and actions required inthe event that a fire door (Recip/CCP Room Door) becomesinoperable.Original Question # - 10098-DIA-1999-160 Page 123 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

65. 065 004/BOTH/OK/DIRECT/AOP2604-1/103000 K1.02/3.9/4.1/H/2Given the following conditions on Unit 1:

• Reactor has tripped.

• RO notes that two large sections of the containment isolation panel valves have justgone closed. Which ONE of the following sets of conditions would indicate that this is due to aspurious Phase A signal?

CNTMT STEAM LINE SG PZR PRESSURE DELTA-P PRESSURE PRESSURE 2.9 psig 80 psid 1005 psig 2000 psig 0.2 psig 20 psid 450 psig 1810 psig 0.5 psig 15 psid 600 psig 1900 psig 1.7 psig 110 psid 800 psig 1850 psigA.B.C.D.ANSWER: CA - INCORRECT. Containment pressure is high enough (>1.0 psig) for an SI/Phase A.B - INCORRECT. SG Pressures are below the SI/Phase A setpoint.C - CORRECT. All parameters listed are within values to prevent an SI/Phase Aactuation.D - INCORRECT. Containment pressure is high enough (>1.0 psig) and Steam LineDelta-P is high enough (>100 psid) for an SI/Phase A actuation. Page 124 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D6\#RO-C-AOP0520412-E1

REFERENCE:

OHP-4022-034-003 KA - 103000 K1.02 Containment System Knowledge of the physical connections and/or cause-effect relationships between theContainment System and the following systems:

Containment isolation/containment integrity RO - 3.9 SRO - 4.1 CFR - 41.2 to 41.9 / 45.7 to 45.8KA Justification - Question tests knowledge of the conditions whichwill cause a Containment Isolation.Original Question # - AOP2604-1 Page 125 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

66. 066 003/BOTH/OK/DIRECT/MAST OPMED-19/194001 2.1.4/3.3/3.8/F/3A licensed individual is planning to undergo some medical evaluations and a testutilizing radioisotopes. It has been determined that this test will not affect judgment orfitness for duty in any way.

Which ONE of the following describes the procedural requirements for theseconditions?

The licensed individual:Does not need to report this condition as a potentially disqualifying condition since itis not a fitness for duty issue.

Must report this situation to the fitness-for-duty liaison for independent verificationthat it is not a fitness for duty issue, prior to assuming license duties.Must notify the Plant Manager who will evaluate the condition, prior to assuminglicense duties.

Must notify the Ops Training Manager of a potential disqualifying medical conditionA.B.C.D.ANSWER: DA - INCORRECT. This condition could affect a person's ability to perform requiredtasks in the Aux Building..B - INCORRECT. No independent review by the fitness -for-duty liaison is required..C - INCORRECT. The Plant Manager is not responsible for this item. Reporting needsto be made to the Ops Training Manager.D - CORRECT. The described condition does not affect judgment nor is it a fitnessfor duty issue. Since a medical test that utilizes radioisotopes wouldimpact an individual's ability to enter and exit the auxiliary building, itlimits the individual's ability to perform licensed duties. It is thereforereportable to the Ops Training Manager using Data Sheet 1. Page 126 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-ADM06/Terminal and #3

REFERENCE:

OHI-2071 KA - 194001 2.1.4 Generic Conduct of Operations Knowledge of individual licensed operator responsibilities related to shift staffing, suchas medical requirements, 'no-solo' operation, maintenance of active license status,10CFR55, etc.

RO - 3.3 SRO - 3.8 CFR - 41.10 / 43.2 KA Justification - Question tests knowledge of the operators responsibility for makingnotifications regarding changes to medical condition.Original Question # - Master Bank OpMED-19 Original Question KA - Unknown Page 127 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

67. 067 004/BOTH/OK/DIRECT/NRC EXAM 2006-094-12/194001 2.1.5/2.9/3.9/F/2Given the following conditions on Unit 2:

• Reactor power is stable at 100%.

• A Reactor Operator and the Unit Supervisor are in the Control Room.

• A high vibration alarm is received on the Heater Drain pump requiring someone to gobehind the panel to check the indications.Which ONE of the following describes the procedurally accepted method of checkingthe indications?The Unit Supervisor can go behind the panel to check the vibration.The Reactor Operator can go behind the panel to check the vibration.Both the Reactor Operator and the Unit Supervisor are allowed to go behind thepanel to check the vibration as long as all controls are in automatic.Neither the Reactor Operator or the Unit Supervisor can go behind the panels. They must get another operator to check the vibration.A.B.C.D.ANSWER: AA - CORRECT. The Unit Supervisor must be in the Control Room but may gobehind the panels. The RO must remain in the view of the panels.B - INCORRECT. The RO must remain in the view of the panels.C - INCORRECT. The RO must remain in the view of the panels.D - INCORRECT. The SRO may go behind the panels. Page 128 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-ADM01/#1

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OHI- 4000, Conduct of Operations, Attachment 22 (Shift Staffing)KA - 194001 2.1.5 Generic Conduct of OperationsAbility to use procedures related to shift staffing, such as minimum crew complement,overtime limitations, etc.

RO - 2.9 SRO - 3.9 CFR - 41.10 / 43.5 / 45.12 KA Justification - Question tests ability to understand and use procedures governingstaffing and At-the-Controls areas.Original Question # - Cook 2006 NRC Exam-094-12 : AUDIT RO22-BOTH-25Original Question KA - 2.1.4 Page 129 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

68. 068 001/BOTH/OK/NEW/NEW/194001 2.1.39/3.6/4.3/F/2Conservative Decision making states that :

"WHEN faced with unexpected or uncertain conditions, THEN personnel must promptlyidentify a transition point at which efforts to keep the unit on-line or on schedule are nolonger conservative, nor reasonable."

Once this point is reached :the reactor must be tripped immediately.

senior management must be notified to determine course of action.actions must be taken to place the unit in a safe condition without hesitation.the NRC must be notified and actions taken to address the problem within onehour.A.B.C.D.ANSWER: CA - INCORRECT. tripping the reactor is only one of the options available and may notbe the prudent choice based on the transition point determined.B - INCORRECT. Action must be taken without hesitation.C - CORRECT. OHI-4000, Att. 5, Step 3.4 states "WHEN faced with unexpected oruncertain conditions, THEN personnel must promptly identify atransition point at which efforts to keep the unit on-line or onschedule are no longer conservative, nor reasonable. Once thispoint is reached, actions to place the unit in a safe condition byreducing power, tripping the reactor, or suspending core alterationsmust be taken without hesitation.D - INCORRECT. NRC Notification is not required and action must be taken withouthesitation. Page 130 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-ADM14/#ADM14-3

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OHI-4000, Attachment 5 KA - 194001 2.1.39 Generic Conduct of Operations Knowledge of conservative decision making practices.RO - 3.6 SRO - 4.3 CFR - 41.10 / 43.5 / 45.12 KA Justification - Question tests Knowledge of conservative decision making practices.Original Question # - New Original Question KA - New Page 131 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

69. 069 003/BOTH/OK/DIRECT/MAST 01ADMC0302-1/194001 2.2.20/2.6/3.8/F/3In accordance with administrative procedures, which ONE of the following conditionswould permit troubleshooting to be performed on Technical Specification equipment?Equipment is operating.

Equipment is declared inoperable.

Equipment is logged operable but degraded.

Equipment is operable but LCO time is not exceeded.A.B.C.D.ANSWER: BA - INCORRECT. The fact that TS equipment is operating would not make itacceptable to perform trouble shooting activities.B - CORRECT. Troubleshooting activities are allowed if the equipment is declaredinoperable.C - INCORRECT. Equipment may be logged as operable but degraded but this is nota factor in determining if troubleshooting can be performed.D - INCORRECT. There is no provision to perform activities within the LCO time limits,The equipment must be inoperable or restored with an approvedprocedure.

PMP-2291-TRS-001:

3.1.5 Troubleshooting activities shall not be performed on TechnicalSpecification equipment that results in the equipmentbecoming inoperable -UNLESS-The troubleshooting is performed in conjunction with anapproved procedure that returns the equipment to an operable status, AND Permission is obtained from the Shift Manager.

3.1.6 Troubleshooting activities may be performed on TechnicalSpecification equipment that is out of service or inoperable. Page 132 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-ADM03/RO-C-ADM03-E2, RO-C-ADM03-E7

REFERENCE:

PMP-2291-TRS-001, Step 3.1.5 & 3.1.6.

KA - 194001 2.2.20 Generic Equipment Control Knowledge of the process for managing troubleshooting activities.RO - 2.6 SRO - 3.8 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Question tests knowledge of the troubleshooting process (whenallowed on TS equipment).Original Question # - MASTER 01ADMC0302-1 Original Question KA - 2.1.12 Page 133 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

70. 070 002/BOTH/OK/DIRECT/NRC EXAM 2004-064-5/194001 2.2.21/2.9/4.1/F/3A maintenance visual inspection requires momentarily placing a 'B' Train pump controlswitch in PULL-TO-LOCKOUT. The Unit condition is such that BOTH trains arerequired to auto start.

Which ONE of the following describes the status of the affected ESF system?The 'B' Train pump is INOPERABLE until:the control switch is independently verified in its normal position.the pump's monthly surveillance has been performed.

the pump's auto start function is tested.

the pump is manually started.A.B.C.D.ANSWER: A A - CORRECT. The B train pump may be considered OPERABLE after beingreturned to the correct position and being independently verified.B - INCORRECT. Surveillance does NOT need to be performed to declare B trainequipment OPERABLE.C - INCORRECT. Once returned to the correct position and being independentlyverified train B is considered OPERABLE - a test of the pump's autostart function is NOT required.D - INCORRECT. Once returned to the correct position and being independentlyverified train B is considered OPERABLE - a functional test (manualstart) is NOT required. Page 134 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-O-0005/#ADM0170301

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OHI-4043, Technical specification Open Items LogKA - 194001 2.2.21 Generic Equipment Control Knowledge of pre- and post-maintenance operability requirements.RO - 2.9 SRO - 4.1 CFR - 41.10 / 43.2 KA Justification - Requires knowledge of the requirements for declaring a pumpOPERABLE following maintenance activity.Original Question # - NRC EXAM 2004-064-5 Original Question KA - GENERIC 2.2.24 Page 135 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

71. 071 002/BOTH/OK/DIRECT/RO26 AUDIT-98/194001 2.3.4/3.2/3.7/F/3Given the following conditions on Unit 2:

• A LOCA has occurred

• The TSC has been fully staffed and activated

• An individual is needed for lifesaving activities during which 30 Rem of TEDEexposure is expected to be receivedWhich ONE of the following is correct concerning this lifesaving activity?The individual is required to be a volunteer and the Site Emergency Coordinator isrequired to approve the exposure.

The individual is required to be a volunteer and the Operations Shift Manager isrequired to approve the exposure.

The individual is NOT required to be a volunteer and the Site EmergencyCoordinator is required to approve the exposure.

The individual is NOT required to be a volunteer and the Operations Shift Manageris required to approve the exposure.A.B.C.D.ANSWER: AA - CORRECT. Once in a lifetime doses in excess of 25 REM require a person tobe a volunteer. Any extension above 10CFR20 limits requires SECapproval.B - INCORRECT. The SEC approves extensions above 10CFR20 limits. (SM hasbeen relieved of SEC duties since TSC is activated.C - INCORRECT. A volunteer is required if the dose is in excess of 25 REM. D - INCORRECT. The SEC approves extensions above 10CFR20 limits.(SM has beenrelieved of SEC duties since TSC is activated. Page 136 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-RP02/#RO-C-RP02-E6

REFERENCE:

RO-C-RP02, RMT-2080-TSC-001, Attachment 13KA - 194001 2.3.4 Generic Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions.RO - 3.2 SRO - 3.7 CFR - 41.12 / 43.4 / 45.10 SCLR - 1P K/A Justification - Requires knowledge of the conditions required to allow for anemergency dose during an accident.Modified Question to raise dose level to 30 R. Changed correct Answer to A vs. C.Original Question - RO25 Audit-93, CATAWBA2005, RO26 Audit-98 Page 137 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

72. 072 003/BOTH/OK/DIRECT/NRC EXAM 2004-099-3/194001 2.3.7/3.5/3.6/H/3Given the following conditions:

• Units 1 and 2 are at 100% power.

• Unit 2 has experienced several fuel pin failures.

• A leak must be repaired on a pipe at the end of the Aux. Bldg. 601 ft. elev. pipetunnel.* The general area dose rate in the location of the repair is 600 mrem/hr.

• In order to reach the location of the repair the worker must transit through a 6Rem/hr high radiation area for 2 minutes and return via the same path.

• The worker currently has an accumulated annual dose of 400 mrem. Which ONE of the following is the maximum allowable time that the worker canparticipate in the repairs and NOT exceed the TEDE Administrative Dose Limit?70 minutes 120 minutes 140 minutes 160 minutesA.B.C.D.ANSWER: BA - INCORRECT. Based on using a limit of 1500 versus correct ADL (2000). B - CORRECT. The candidate should determine that the ADL is 2000 mrem.Transient exposure is 400 mrem (6000mrem/hr x 4/60hr). (transit to and from the job). (Current) 400 mrem + (transit) 400mrem = 800 mrem ADL of 2000 mrem - 800 mrem = 1200 mremallowable before reaching ADL.

1200 mrem /600 mrem/hr = 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C - INCORRECT. Based on calculating using a one-way transit dose. D - INCORRECT. Based on using ADL (2000) and NO transit dose. Page 138 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-RP02/#5

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RO-C-RO02 10CFR20 and Radiation Protection Attachment pg. 1KA - 194001 2.3.7 Generic Radiation ControlAbility to comply with radiation work permit requirements during normal or abnormalconditions.

RO - 3.5 SRO - 3.6 CFR - 41.12 / 45.10 KA Justification - Question tests knowledge of stay time, transition, time, and doselimits required for compliance with an RWP.Original Question # - NRC Exam 2004-099-3, AUDIT02-SRO-6Original Question KA - GENERIC 2.3.1 Page 139 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

73. 073 002/BOTH/OK/DIRECT/NRC EXAM 2004-049-5/194001 2.3.11/3.8/4.3/H/3Given the following conditions on Unit 1:

• Containment Purge System is operating in the VENTILATION MODE.

• A HIGH alarm on VRS-1505, Auxiliary Building Ventilation Noble Gas ActivityMonitor, occurs (unplanned).Which ONE of the following describes the required operator response for theContainment Ventilation System to the failure alarm?Stop the Containment Purge and consult with Radiation Protection prior torestarting the system.

Continue the Purge as long as VRS-1101, Containment Normal Range AreaRadiation Monitor still indicating as expected.

Verify the following:

• Containment ventilation isolation valves 1-VCR-101 through 1-VCR-107 close;

• 1-HV-CPS-1/2, Containment Purge Supply Fans 1 and 2, trip;* 1-HV-CPX-1/2, Containment Purge Exhaust Fans 1 and 2, trip;* 1-HV-CPR-1, Containment Pressure Relief Fan, trips;* 1-HV-CIPS-1, Containment Instrument Room Purge Supply Fan, trips.Verify the following:

• Containment ventilation isolation valves 1-VCR-201 through 1-VCR-207 close;

• 1-HV-CPS-1/2, Containment Purge Supply Fans 1 and 2, trip;* 1-HV-CPX-1/2, Containment Purge Exhaust Fans 1 and 2, trip;* 1-HV-CPR-1, Containment Pressure Relief Fan, trips;* 1-HV-CIPX-1, Containment Instrument Room Purge Exhaust Fan, trips.A.B.C.D. Page 140 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: A A - CORRECT. When the Containment Purge system is operating in the VentilationMode, the automatic isolation signals are blocked. The procedurerequires the Purge to be stopped and radiation protectionconcurrence prior to restarting the system.B - INCORRECT. The procedure requires the Purge to be stopped and radiationprotection notified. Plausible as the Containment radiation monitoris still operable monitoring for any releaseC - INCORRECT. When the Containment Purge system is operating in the VentilationMode, the automatic isolation signals are blocked. Plausible asthese are functions from containment isolation signal actuationresults.D - INCORRECT. When the Containment Purge system is operating in the VentilationMode, the automatic isolation signals are blocked. Plausible asthese are functions from containment isolation signal actuationresults.LESSON PLAN/OBJ: RO-C-02800/#9

REFERENCE:

1-OHP-4021-028-005, Att. 2 KA - 194001 2.3.11 Generic Radiation ControlAbility to control radiation releases.RO - 3.8 SRO - 4.3 CFR - 41.11 / 43.4 / 45.10 KA Justification - Question tests ability to control releases through identification ofconditions requiring manual termination.Original Question # - NRC Exam 2004-049-5, AUDIT02-BOTH31Original Question KA - GENERIC 2.3.9 Page 141 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

74. 074 004/BOTH/OK/DIRECT/22430-DIAB2002-63/194001 2.4.2/4.5/4.6/F/3Given the following conditions on Unit 1:

• A reactor trip occurred from 15% power.

• Safety Injection was NOT actuated and was NOT required.

• 1-OHP-4023-E-0, Reactor Trip or Safety Injection, has been performed, and atransition to 1-OHP-4023-ES-0.1, Reactor Trip Response, has been made. The following conditions exist: Tavg is STABLE at 547°F Pressurizer level is 11% and lowering slowly RCS subcooling is 32°F and lowering slowly All NR SG levels are 28 - 30%; AFW flows indicate 0 klb/hr Containment pressure is 0.7 psig and rising slowly Which ONE of the following describes the appropriate actions for these conditions? Actuate SI and return to 1-OHP-4023-E-0 step 1.

Go to 1-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink.Go to 1-OHP-4023-FR-I.2, Response to Low Pressurizer Level.Manually start ECCS pumps and continue with 1-OHP-4023-ES-0.1.A.B.C.D.ANSWER: A A - CORRECT. The foldout page directs an SI and return to 1-OHP-4023-E-0 ifsubcooling is < 40°FB - INCORRECT. SG levels are adequate for Heat Sink even with no AFW flow.Plausible due to loss of subcooling the student could want toincrease cooling to the S/G as well as missing the Narrow Rangelevel.C - INCORRECT. Initiating SI and returning to 1-OHP-4023-E-0 would takeprecedence over 1-OHP-4023-FR-I.2. Plausible due to the desire ofthe student to recover PZR level.D - INCORRECT. This may be the action required in other emergency procedures(ES-1.2) but a return to 1-OHP-4023-E-0 is required to verify properalignment of ECCS equipment. ES-0.1 is considered a'non-accident' EOP. Plausible due to this action being elsewhere inthe EOP network. Page 142 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP03/#18 & 23

REFERENCE:

1-OHP-4021-ES-0.1 (Foldout Page)

KA - 194001 2.4.2 Generic Emergency Procedures/Plan Knowledge of system set points, interlocks and automatic actions associated with EOPentry conditions.

RO - 4.5 SRO - 4.6 CFR - 41.7 / 45.7 / 45.8 KA Justification - Requires a knowledge of conditions(setpoints) that would require atransition from Reactor Trip Response to re-entry into the ReactorTrip/SI EOP.Original Question # - 22430-DIAB2002-63 Original Question KA - E05.2.4.21 Page 143 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

75. 075 002/BOTH/OK/DIRECT/RO22 AUDIT-BOTH-13/194001 2.4.6/3.7/4.7/F/3Given the following conditions on Unit 2:

• A Safety Injection (SI) has ocurred.

• The Immediate Action steps of 2-OHP-4023-E-0, Reactor Trip Or Safety Injection,have just been completed.The following steam generator conditions are noted:

• SG 21 pressure is 740 psig and lowering slowly.

• SG 22 pressure is 450 psig and lowering rapidly

• SG 23 pressure is 735 psig and lowering slowly.

• SG 24 pressure is 745 psig and lowering slowly.

• Main Steam header pressure is 700 psig and lowering slowly.Which ONE of the following actions should be promptly performed to mitigate theevent? Transition to 2-OHP-4023-E-1, Loss of Reactor or Secondary Coolant.Transition to 2-OHP-4023-E-2, Faulted Steam Generator Isolation,Close all the SG stop valves and continue with 2-OHP-4023-E-0, Reactor Trip OrSafety Injection.

Close SG 22 Stop Valve and verify steam supply available to the Turbine DrivenAxiliary Feedwater Pump (TDAFP)

. A.B.C.D. Page 144 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: CA - INCORRECT. Step transition to E-1 is not performed until later in procedures. Inaddition, conditions given indicate a steam leak (E-2) condition.Plausible due to E-0 does direct going to E-1.B - INCORRECT. Plausible due to conditions described indicate a Faulted S/G but thetransition is not allowed at this point in E-0. The transition to E-2 islater in the procedure.C - CORRECT. Steamlines should be isolated for any number of reasons, such as Automatic steamline isolation failure, RCS temperature lowering(procedural steps)and personnel protection. OHI-4023 allowsprudent actions to trip the SG Stop Valves when it is apparent that asteam line leak has occured for personnel protection and inresponse to automatic action failures.D - INCORRECT. All SG Stop valves should be closed when taking the prudentactions in OHI-4023. Plausible due to student diagnosing theproblem and taking actions only to isolate the affected S/G. LESSON PLAN/OBJ: RO-C-EOP01/#17, RO-C-EOP07/#3

REFERENCE:

OHI-4023, Abnormal/Emergency Procedure User's Guide, Step4.7.3.b.5.KA - 194001 2.4.6 Generic Emergency Procedures/Plan Knowledge of EOP mitigation strategies.

RO - 3.7 SRO - 4.7 CFR - 41.10 / 43.5 / 45.13 KA Justification - Requires the knowledge of the mitigating strategy for limiting thecooldown in the EOPs during a Steam Line break before thediagnostic steps have been reached (OHI-4023 allowance to closeSG Stop Valves).Original Question # - RO22 Audit-Both-13 Original Question KA - APE 040 AK3.04 Page 145 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

76. 076 005/SRO/OK/MODIFIED/RO26 AUDIT-95/000005 AA2.03/3.5/4.4/H/3Given the following conditions on Unit 2:

• Unit was at 90% power

• Control Rods were in AUTO

• CBD began to step out with no mismatch signal.

• Rods were taken to MANUAL and rod motion ceased.The following conditions now exist:

• CBD Bank Demand position is now 222 Steps.

• Group 2 Bank D RPIs ALL indicate 222 Steps.

• Group 1 Bank D RPIs indicate as follows:o Rod D4: 205 Steps. o Rod D12: 222 Stepso Rod M12: 207 Stepso Rod M4: 222 StepsReactor Engineering has determined that all CBD rods are free to move and hasprovided the following information:

• R is 1.041

• CFQ is 2.335

• K(Z) is .95 (at 10 feet)

• is 2.174 (at 10 feet))Z (F W QWhich ONE of the following identifies the Technical Specification Action Condition(s)that must be entered?

Reference Provided: Unit 2 TS 3.1.4 Rod Group Alignment Limits 3.1.4.A Only 3.1.4.B Only 3.1.4.A and 3.1.4.B Only 3.1.4.B and 3.1.4.D OnlyA.B.C.D. Page 146 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: DA - INCORRECT. Rods are free to move, as stated in the stem, thus they are alloperable.B - INCORRECT. Using the numbers provided with Figure 3.1.4-1, rods which are <14 steps misaligned meet alignment requirements. This is thecorrect action with only 1 misaligned rod.C - INCORRECT. Rods are free to move and are operable.D - CORRECT. Both D4 and M12 are unacceptably misaligned. D4 is 17 steps offand M12 is 15 steps off.LESSON PLAN/OBJ: RO-C-TS01/#11

REFERENCE:

TS 3.1.4, Rod Group Alignment Limits Reference Provided: Unit 2 TS 3.1.4 Rod Group Alignment Limits MODIFIED: Changed to have 2 rods misaligned. Changed answer to D from B.KA - 000005 AA2.03 Inoperable/Stuck Control RodAbility to determine and interpret the following as they apply to the Inoperable/StuckControl Rod:

Required actions if more than one rod is stuck or inoperable.RO - 3.5 SRO - 4.4 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the ability to determine and interpret the required TS actionswith 2 misaligned rods.Original Question # - RO26 AUDIT-95 Original Question KA - 194001 2.1.7 Page 147 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

77. 077 008/SRO/OK/NEW/NEW/000007 EA2.04/4.6/4.4/F/3Given the following conditions on Unit 1:

• A power escalation is in progress

• At approximately 35% power the Main Turbine Trips The following conditions now exist:

• All 4 Turbine Stop Valve Status Lights are lit

• All RPIs have lost power

• Train A Reactor Trip Breaker is Open

• Train B Reactor Trip Breaker is Closed

• WR Log Power = 7%

• WR Startup Rate = 0.0 DPM and stableWhich ONE of the following actions is required?Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection. Following completion of Immediate Actions, transition to 1-OHP-4023-ES-0.1,Reactor Trip Response.

Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection.During verification of Reactor Trip, transition to Implement 1-OHP-4023-FR-S.1,Response to Nuclear Power Generation/ATWS, and manually insert control rods.Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection).When directed go to 1-OHP-4023-FR-S.1, Response to Nuclear PowerGeneration/ATWS.

Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection).Upon Turbine Trip Verification, go to 1-OHP-4023-E.0, Reactor Trip or SafetyInjection. A.B.C.D. Page 148 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. A transition to FR-S.1 is required. Plausible as E-0 is the correctentry procedure and student may consider actions in E-0 capable ofresolving the problem.B - CORRECT. Due to flux being greater than 5% and not lowering, E-0, Step 1RNO will require a transition to FR-S.1. Since Reactor trip cannot beverified (flux NOT lowering), manual control rod insertion will berequired. C - INCORRECT. E-0 should be implemented since a Turbine Trip greater than P-8should result in a Reactor trip. Plausible as student may key in oncurrent power level and go to procedure governing for a loss of loadwith a reactor power level below the reactor trip setpoint.D - INCORRECT. E-0 should be implemented since a Turbine Trip greater than P-8should result in a Reactor trip. E-0 takes precedence over AOPs.Plausible as student may key in on current power level and go toprocedure governing for a loss of load with a reactor power levelbelow the reactor trip setpoint.LESSON PLAN/OBJ: RO-C-EOP03/#13, #14, RO-C-EOP04/#13

REFERENCE:

1-OHP-4023-E-0, Reactor Trip or Safety Injection,1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWSKA - 000007 EA2.04 Reactor Trip - StabilizationAbility to determine and interpret the following as they apply to a reactor trip:If reactor should have tripped but has not done so, manually trip the reactor and carryout actions in ATWS EOP RO - 4.6 SRO - 4.4 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the ability to determine that a Reactor Trip is required andthat the reactor is not tripped (FR-S.1 Entry), and then the immediateactions needed to make the reactor subcritical (inserting controlrods).Original Question # - RO23 AUDIT 011-4 Page 149 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

78. 078 003/SRO/OK/MODIFIED/NRC EXAM 2004-020-2/000022 2.4.50/4.2/4.0/F/3Given the following conditions on Unit 2:

• Reactor power was 100% power when an RCS leak developed.

• The Unit Supervisor is implementing 2-OHP 4022-002-020, Excessive RCSLeakage.The following conditions now exist:

• Letdown flow is isolated.

• East and West Charging pumps are operating.

• Charging flow is 180 gpm.

• Pressurizer level is 51% and constant.

• VCT makeup is in Auto.

• VCT level is 22% and lowering.

• Containment pressure is 0.5 psig and constant.Which ONE of the following describes the required operator action and why (assumeall control systems function as designed)?Verify that CCP suction automatically aligns to the RWST at 14.0% VCT level andperform a controlled rapid shutdown per 2-OHP-4022-001-006 Rapid PowerReduction Response, to maintain RCS Tavg-Tref.

Verify VCT auto makeup begins at 14.0 % and then restore 75 gpm letdown toensure proper regen heat exchanger warming of the charging flow.Verify that CCP suction automatically aligns to the RWST at 7.0% VCT level andperform a controlled rapid shutdown per 2-OHP-4022-001-006 Rapid PowerReduction Response since RCS leakage is greater than the Technical SpecificationLimit.

Trip the reactor and transition to 2-OHP-4023-E-0, Reactor Trip or Safety Injectionsince VCT level can NOT be maintained with VCT auto makeup.A.B.C.D. Page 150 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: D A - INCORRECT. The procedure directs a Reactor Trip. Temperature control wouldbe extremely difficult. The VCT Refueling Water SequenceActuation and alarm come in at 2.5% VCT level. Plausible sinceVCT low alarm setpoint is 14.0% and RWST alignment wouldimpact temperature control.B - INCORRECT. Letdown was isolated to allow Pressurizer level to be stabilized.Auto Makeup will start at 24% but flow will not be sufficient to makeup for 180 gpm leak. C - INCORRECT. The procedure directs a Reactor Trip. The VCT Refueling WaterSequence Actuation and alarm come in at 2.5% VCT level.Plausible since a VCT low alarm setpoint is at 7.0% and thisleakage would exceed TS.D - CORRECT. Leakage in excess of VCT makeup will lead to eventual loss of CCPsuction. This would be mitigated by the refueling water sequenceswapover to the RWST suction source but this would result inexcessive boration of the RCS. Lowering level in excess of automakeup capability require a RX trip per 2-OHP-4022-002-020. Page 151 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D1/RO-C-AOP0160412-E3

REFERENCE:

2-OHP-4022-002-020, Excessive Reactor Coolant Leakage Step3, 2-OHP-4024-209, Drops 48, 49, 50 KA - 000022 2.4.50 Loss of Reactor Coolant Makeup Emergency Procedures/PlanAbility to verify system alarm setpoints and operate controls identified in the alarmresponse manual.

RO - 4.2 SRO - 4.0 CFR - 41.10 / 43.5 / 45.3 KA Justification - Requires the ability to determine that with the stated conditions VCTMakeup will be insufficient to maintain VCT level. Further lowering ofVCT level will eventually lead to the Refueling Water Sequence alarmsetpoint and actuation. Based on these conditions, the SRO isrequired to assess the plant conditions and determine that a reactortrip and entry into the EOPs is required.Modified by changing VCT Makeup in the Stem (was operating at maximum) to inAuto. Changed distractors A & C to add RWST swapover setpoints.and B & D to include VCT Makeup occurring or not.Original Question # - NRC 2004-020-2 Original Question KA - 002000 A2.01 Page 152 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

79. 079 002/SRO/OK/MODIFIED/NRC EXAM 2008-77/000026 AA2.02/2.9/3.6/H/3Given the following conditions on Unit 2:

• Unit was operating at 100% power

• 2-OHP-4022-016-001, Malfunction of the CCW System is being implemented due toindications of a lowering CCW Surge tank level.

• The Crew has started the West CCW pump, split the East and West Headers, and aligned the Miscellaneous Services Header to the East Header.

• An AEO has reported that a CCW leak of approximately 150 gpm has beenidentified in the Aux Building 609' elevation, flowing toward the passenger elevator.The following Surge Tank Level Recorder conditions exist: Train 'A' Train 'B' CLR-410 CLR-411Reading 48" 18"Trend Stable LoweringWhich ONE of the following describes the leak location and the required actions?The leak is located on the:Miscellaneous Services Header. Trip the Reactor, Stop both CCW Pumps, andImplement 2-OHP-4022-016-004, Loss of CCW along with 2-OHP-4023-E-0,Reactor Trip or Safety Injection.

East Safeguards Header. Shutdown the East CCW pump and align theMiscellaneous Services Header to the West Safeguards Header.West Safeguards Header. Shutdown the West CCW pump and the equipmentcooled by the West Header.

Miscellaneous Services Header. Trip the Reactor, Trip the RCPs, and isolate theMiscellaneous Services Header while performing 2-OHP-4023-E-0, Reactor Trip orSafety Injection.A.B.C.D. Page 153 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: CA - INCORRECT. During the initial train split, the Misc. Header is aligned to the EastSafeguards Header. Miscellaneous Header is isolated from the leak.Plausible if the student believes the leak is on the Misc header. Differs from distractor D as to the additional actions listed. Thisanswer stops the CCW pumps and implements loss of CCW whileperforming E-0 which is logical for mitigating the event however it isNOT in full compliance with the procedural requirements.B - INCORRECT. During the initial train split, the Misc. Header is aligned to the EastSafeguards Header. The East CCW Header is isolated from theleak. Plausible if the student determines the leak is on the incorrectheader.C - CORRECT. Initial train separation places the Miscellaneous Header on the EastTrain. The conditions presented indicate that the leak is on the WestSafeguards Header which can be isolated from the East Headeramd Miscellaneous Header.D - INCORRECT. Miscellaneous Header is aligned to the East CCW Train. Miscellaneous Header does not need to be isolated. Plausible if thestudent believes the leak is on the Misc header. Differs fromdistractor A as to the additional actions listed. This answer trips thereactor, RCP's, and isolates the Misc header while performing E-0which is logical for mitigating the event however it is NOT in fullcompliance with the procedural requirements. Page 154 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D8\RO-C-AOP0420412-E3, RO-C-01600-E3

REFERENCE:

2-OHP-4022-016-001, Malfunction of the CCW System,SOD-01600-001, RO-C-AOP-D8Modified: Stem surge tank levels (Swapped) and leakage location. This changedthe correct answer from D to C.KA - 000026 AA2.02 Loss of Component Cooling Water (CCW)Ability to determine and interpret the following as they apply to the Loss of ComponentCooling Water:

The cause of possible CCW loss RO - 2.9 SRO - 3.6 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Question requires Candidate to determine the leak location andidentify the equipment lost and actions required.Original Question # - NRC EXAM 2008-77 Original Question KA - 000026 AA2.02 Page 155 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

80. 080 004/SRO/OK/DIRECT/RO24 AUDIT-021-7/000032 2.1.7/4.4/4.7/H/3Given the following conditions on Unit 2:

• A reactor startup is in progress with the reactor just critical.

• The operator has just stopped moving control rods.

• Intermediate Range Power slowly rises above 2 x 10-10 amps.

• ONE source range (SR) nuclear instrumentation channel (N-31) fails LOW.

• Remaining power indications stabilize. Which ONE of the following actions, if any, is required for compliance with TechnicalSpecifications?

Reference Provided: Unit 2 TS 3.3.1, Reactor Trip System InstrumentationNo action required, source range not required to be operable.

Trip the reactor and enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection. Conduct a reactor shutdown and restore both SR channels to operability prior tonext startup.

Suspend all operations involving positive reactivity changes until both SR channelsare restored to operability. A.B.C.D.ANSWER: A A - CORRECT. TS 3.3.1 (Instrumentation) establishes that above P-6, the SR NIare not required by TS and will shortly be de-energized byprocedure. Since there are no TS implications, the startup mayproceed.B - INCORRECT. No entry conditions are met for a reactor trip. (plausiblemisconception) Plausible as this would be a conservative action toplace the plant in a known safe, stable condition but does notanswer the actual question.C - INCORRECT. A plant shutdown is not required unless both SR channels are lostduring reactor startup. Plausible since shutdown is performed forseveral startup inconsistencies (ECC wrong, conditions change,etc.) and this is an action for several reactor start up issues.D - INCORRECT. SR channels can be blocked. Reactor power is above P-6.Plausible since this statement would be correct if reactor power was< P-6 and this is an action associated with the Source RangeDetectors in Technical Specifications. Page 156 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-01300/#21

REFERENCE:

TS 3.3.1, RTS Instrumentation, Table 3.3.1-1 Reference Provided: Unit 2 TS 3.3.1, Reactor Trip System InstrumentationKA - 000032 2.1.7 Loss of Source Range Nuclear Instrumentation Conduct of OperationsAbility to evaluate plant performance and make operational judgments based onoperating characteristics, reactor behavior, and instrument interpretation.RO - 4.4 SRO - 4.7 CFR - 41.5 / 43.5 / 45.12 / 45.13 KA Justification - Requires an evaluation of the plant status due to a loss of sourcerage instruments above P-6 setpoint and to determine per TS anyactions required.Original Question # - RO24 AUDIT-021-7 Original Question KA - APE.032GEN2.1.20 Page 157 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

81. 081 010/SRO/OK/NEW/NEW/000038 EA2.09/4.2/4.2/H/3Given the following conditions on Unit 2:

• A SGTR has occurred coincident with a Loss of Offsite Power* 2-OHP-4023-E-3, Steam Generator Tube Rupture, is being performed.

• The Unit Supervisor is at Step 41, Select Appropriate Post-SGTR CooldownProcedure.The following conditions now exist:

• RCS Wide Range Pressure is 800 psig and stable.

• RCS Incore Thermocouples are 475 oF and slowly lowering

• RCS T-hots are 470 and slowly lowering

• RCS T-colds are 445 and stable

• SG Pressures are 387 psig and stable.

• PZR level is 25% and slowly rising. Which ONE of the following describes the status of natural circulation and theappropriate procedural transition for the Unit Supervisor?Natural Circulation exists.

Transition to 2-OHP-4023-ES-3.1, Post-SGTR Cooldown Using Backfill.Natural Circulation does NOT exist.

Transition to 2-OHP-4023-ES-3.2, Post-SGTR Cooldown Using BlowdownNatural Circulation does NOT exist.

Transition to 2-OHP-4023-ES-3.3, Post-SGTR Cooldown Using Steam Dump.Natural Circulation exists.

Transition to 2-OHP-4023-ECA-3.1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired.A.B.C.D. Page 158 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: AA - CORRECT. Based on the conditions Natural Circ exists: T/Cs, T-hot, and SGPress. (Stable or Lowering); RCS cold leg temperatures are atsaturation Temp for SG Press; Subcooling is 51 oF (>40 oF).2-OHP-4023-ES-3.1, Post-SGTR Cooldown Using Backfill, is anappropriate transition form E-3. B - INCORRECT. Based on the conditions Natural Circ DOES exist. Plausible as thisprocedure is a valid transition from E-3 but incorrect if the studentcan not validate the status of natural circulation.C - INCORRECT. Based on the conditions Natural Circ DOES exist. Plausible as thisprocedure is a valid transition from E-3 but incorrect if the studentcan not validate the status of natural circulation.D - INCORRECT. ECA-3.1, SGTR With Loss Of Reactor Coolant - SubcooledRecovery Desired, transition is from the foldout page when eithersubcooling or PZR Level cannot be maintained. PZR level andsubcooling are both adequate. Plausible since Natural Circulationrequires subcooling and student knows that cooldown of the rupturedSG will be somewhat impeded under natural circulation conditions.LESSON PLAN/OBJ: RO-C-EOP08/#15, #22

REFERENCE:

2-OHP-4023-E-3, Steam Generator Tube RuptureKA - 000038 EA2.09 Steam Generator Tube Rupture (SGTR)Ability to determine and interpret the following as they apply to a SGTR:Existence of natural circulation, using plant parametersRO - 4.2 SRO - 4.2 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires knowledge of the conditions that indicate natural circ duringa SGTR event and (SRO) the appropriate transition based on thatdetermination.Original Question # - New Original Question KA - New Page 159 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

82. 082 003/SRO/OK/DIRECT/NRC EXAM 2006-080-20/000040 2.4.4/4.5/4.7/H/3Given the following conditions on Unit 2:

• Steam Gen 1/2/3/4 Steam Line Flow High Alarms - LIT

• Steam Gen 1/2/3/4 SF>FWF Flow Mismatch Alarms - LIT

• RCS Tavg is 561 oF and lowering

• Turbine load is lowering

• Rods are stepping out

• Steam flows are 3.6 x 10 6 lbm/hr and stable.

• FW flows are 2.1 x 10 6 lbm/hr and rising.Which ONE of the following correctly describes the cause and required action to betaken for the above conditions?A steam line break exists. Direct the operators to perform a Reactor Trip and MainSteamline Isolation.

A feed line break exists. Direct the operators to perform a Reactor Trip and MainFeedwater Isolation.

Feedwater Pump Delta-P is too Low. Direct the operator to raise FW Pump Speedand FW pump flow.

MPC-253 has failed LOW. Direct the operators to perform actions for failed FirstStage Turbine Impulse Pressure Transmitter.A.B.C.D.ANSWER: AA - CORRECT. Based on the conditions presented a steam line break hasoccurred. Steam flow is indicating at the 97 to 98% power range.Tavg is 13 oF Low for 98% power. A reactor trip and Steam Lineisolation is warranted.B - INCORRECT. If a FW break existed RCS temperature would be rising. Plausibleas the alarms are associated with Feed and well as steam.C - INCORRECT. If FW Flow was low RCS Temperature would be rising. Plausible asfeed is less than steam which can be caused by SGFP low DP.D - INCORRECT. If MPC-253 failed low the alarms would come in (Steam flow higherthan calculated power) but rods would step out and SF/FWFmismatch would not be this high. Plausible as this failure woulddrive some of the alarms being received. Page 160 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP07/#5

REFERENCE:

RO-C-EOP07 KA - 000040 2.4.4 Steam Line Rupture Emergency Procedures/PlanAbility to recognize abnormal indications for system operating parameters that areentry-level conditions for emergency and abnormal operating procedures.RO - 4.5 SRO - 4.7 CFR - 41.10 / 43.2 / 45.6 KA Justification - Requires knowledge of the conditions that would require entry into thereactor trip procedure and the actions needed to address thesteamline break (Main Steam Isolation)Original Question # - Cook 2006 NRC Exam-080-20, Modified from NRC EXAM2004-073-1Original Question KA - 000040 AA2.01 Page 161 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

83. 083 007/SRO/OK/NEW/NEW/000060 2.4.45/2.9/3.1/H/3Given the following conditions:

• A Gas Decay Tank release is in progress through the Unit 1 plant vent.

• A High Alarm occurs on VRS-2505, Unit 2 Vent Low Range Noble Gas RadiationMonitor.* RP determines that VRS-2505 is INOPERABLE due to a failed high channel.The effects of the VRS-2505 High Alarm are that 12-RRV-306, Vent Stack ReleaseValve, __________________.

Additionally, which ONE of the following is required by PMP-6010-OSD-001, Off-siteDose Calculation Manual?

Reference Provided: PMP-6010-OSD-001, Off-site Dose Calculation Manual,Attachment 3.4 (pages 57-59)will automatically close Grab samples must be taken at least once per shift and analyzed for gross activitywithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

must be manually closed Grab samples must be taken at least once per shift and analyzed for gross activitywithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

will automatically close No actions are required as long as VRS-1505 remains OPERABLE.must be manually closed No actions are required as long as VRS-1505 remains OPERABLE.A.B.C.D. Page 162 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: AA - CORRECT. A high alarm on VRS-2505 will automatically close 12-RRV-306.PMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment3.4, Action 6 requires grab samples to be taken at least once pershift and analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for continuedeffluent release through the vent header. B - INCORRECT. A high alarm on VRS-2505 will automatically close 12-RRV-306.Plausible as the ODCM requirements are correct.C - INCORRECT. With VRS-2505 INOPERABLE, the ONE channel that monitors theUnit 2 Plant Vent is INOPERABLE. In accordance withPMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment3.4, Action 6 is applicable for this condition. Plausible as theautomatic closure of 12-RRV-306 is correct and a logicaljustification for no actions is provided.D - INCORRECT. A high alarm on VRS-2505 will automatically close 12-RRV-306.With VRS-2505 INOPERABLE, the ONE channel that monitors theUnit 2 Plant Vent is INOPERABLE. In accordance withPMP-6010-OSD-001, Off-site Dose Calculation Manual, Attachment3.4, Action 6 is applicable for this condition. Plausible as a logicaljustification for taking no actions is providedLESSON PLAN/OBJ: RO-C-02300/#3

REFERENCE:

PMP-6010-OSD-001, Off-site Dose Calculation Manual,Attachment 3.4 (pages 57-59) 12-OHP-4024-139, Drop 5Reference Provided: PMP-6010-OSD-001, Off-site Dose Calculation Manual,Attachment 3.4 (pages 57-59)KA - 000060 2.4.45Accidental Gaseous Radwaste ReleaseEmergency Procedures/PlanAbility to prioritize and interpret the significance of each annunciator or alarm.RO - 4.1 SRO - 4.3 CFR - 41.10 / 43.5 / 45.3 / 45.12 KA Justification - Requires the ability to determine the actions that are required for analarm on vent stack rad monitor and the ability to use the Off-siteDose Calculation Manual, to determine the appropriate response tothe rad monitor alarm.Original Question # - New Original Question KA - New Page 163 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

84. 084 008/SRO/OK/NEW/NEW/000062 2.4.20/3.8/4.3/F/3Given the following conditions on Unit 2:

• Unit is implementing 2-OHP-4023-ECA-0.0, Loss of all AC Power

• Power has just been restored from the U2 AB EDG

• Cooling flow to the U2 AB EDG is being checked per Step 32* ONLY the normal ESW supply to the U2 AB EDG is open and providing flow.Which ONE of the following describes the actions the Unit Supervisor should direct andthe reason for those actions?Open the Alternate Supply to the AB EDG Maximizes flow through the EDG to compensate for maximum loading.Leave the Alternate Supply to the AB EDG closed Prevents a loss of ESW cooling to both trains of equipment due to silt and mudbuild-up in the component's heat exchangers.

Open the Alternate Supply to the AB EDG Ensures adequate flow in the event of loss of the normal supply path.Leave the Alternate Supply to the AB EDG closed Limits the amount of flow to the EDG to ensure that other components in the ESWtrain receive adequate flow to support safety functions.A.B.C.D. Page 164 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. Alternate supply should remain closed. See Answer B forjustification. Plausible based on known loading for coming out ofLoss of all AC with only one EDG.B - CORRECT. Step 32 Note states "The alternate ESW cooling supply to theEDGs should remain isolated unless an EDG is running AND the normal ESW supply is NOT available." This Note was added to prevent a loss of ESW cooling from occurring to bothtrains of equipment due to slit and mud build-up in the component's heat exchangers if the ESW system trains arecross-tied via the alternate cooling supplies to the EDGs.C - INCORRECT. Alternate supply should remain closed. See Answer B forjustification. Plausible to ensure cooling to the only operating EDGfor the unit.D - INCORRECT. See Answer B for justification. Plausible based on ensuringcooling is provided to other required equipment.LESSON PLAN/OBJ: RO-C-EOP14/#11

REFERENCE:

12-OHP-4023-ECA-0.0, EOP Step 32 (ERG Step N/A) Note 1BackgroundKA - 000062 2.4.20 Loss of Nuclear Service Water Emergency Procedures/Plan Knowledge of operational implications of EOP warnings, cautions, and notes.RO - 3.8 SRO - 4.3 CFR - 41.10 / 43.5 / 45.13 KA Justification - Requires SRO knowledge of actions to direct concerning ESW supplyto the EDGs and the reason (operational implications) of the EOPnote that directs how to maintain the ESW supply valves to the EDG.

Original Question # - New Original Question KA - New Page 165 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

85. 085 003/SRO/OK/DIRECT/CM-7746-38403/000069 AA2.01/3.7/4.3/H/3Given the following conditions on Unit 1:

• Unit is preparing for a reactor start up following a refueling outage.

• Tavg is 515°F with a heatup in progress.

• During the outage testing was performed per the Containment Leak Rate TestingProgram* At 0200, a Station Engineer reports that a mistake had been made in analyzing therequired Containment Leak Rate Test results that were conducted prior to exceeding200°F.* The initial calculated Type A leakage had been recorded as 0.5 L a* Re-calculation indicates that the Type A leakage is actually 0.8 L a* The re-calculated values have been verified and reviewed by the Shift ManagerWhich ONE of the following actions, if any, is required by Technical Specifications inresponse to this situation?

Reference Provided: Unit 1 TS 3.6.1, Containment and its Bases Continue with the heatup. Entry into Tech Spec 3.6.1 is not required.Continue with the heatup. Do not enter Mode 2 until the leak test is re-performedEnter Tech Spec 3.0.3. Be in MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.Enter Tech Spec 3.6.1. Be in MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.A.B.C.D. Page 166 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: DA - INCORRECT. Entry into TS 3.6.1 is required. See Answer D. Plausible if studentdoes not recognize the required Tech Spec entry.B - INCORRECT. TS 3.6.1 is Applicable in Modes 1-4. Must comply with actions.Plausible if the student applies Mode change restraints to ensuringall required components are operable prior to changing Modes.C - INCORRECT. This is not a TS 3.0.3 issue; actions are provided in TS 3.6.1.Plausible if the student can not correctly apply tech Spec's butknows something is not correctD - CORRECT. TS 3.6.1 required the Containment to be OPERABLE in Modes 1 -4. If the Containment is INOPERABLE, then return to OPERABLEstatus in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 5 in the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (37 hour4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />sto Mode 5)

TS 3.6.1 Bases: Containment OPERABILITY is maintained bylimiting leakage to <1.0 La, except prior to the first startup afterperforming a required Containment Leakage Rate Testing Programleakage test. As left leakage prior to the first startup afterperforming a required Containment Leakage Rate Testing Programleakage test is required to be <0.6 La for combined Type B and Cleakage, and <0.75 La for overall Type A leakage. LESSON PLAN/OBJ: RO-C-TS01/#11

REFERENCE:

TS 3.6.1, Containment, B 3.6.1, Containment Reference Provided: Unit 1 TS 3.6.1, Containment and its Bases KA - 000069 AA2.01 Loss of Containment IntegrityAbility to determine and interpret the following as they apply to the Loss of ContainmentIntegrity:

Loss of containment integrity RO - 3.7 SRO - 4.3 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the ability to determine (in accordance with TS) whethercontainment integrity exists.Original Question # - CM-7746-38403 Original Question KA - SYS 103 K3.02 Page 167 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

86. 086 003/SRO/OK/DIRECT-REPEAT/NRC EXAM 2007-094/005000 2.4.21/4.0/4.6/F/3Given the following conditions on Unit 1:

• A LOCA occurred 30 minutes ago

• RCS pressure is 125 psig

• RCS Core Exit TCs read 380°F

• RCS Cold Leg temperatures are 250°F

• 1N SI Pump is running providing 650 gpm flow

• 1E RHR Pump is running providing 3000 gpm flowWhat is the appropriate action taken in response to the above conditions?Entry into 1-OHP-4023-FR-P.1, Response to Pressurized Thermal Shock Condition, is:made but NO actions are implemented before returning to procedure in effect.made and cooldown will continue within a limit of 50°F in any 60 minute period.made and a RCS temperature soak for a ONE hour period will be completed.NOT required since RCS pressure is below 300 psig.A.B.C.D.ANSWER: A A - CORRECT. Entry into FR-P.1 is required due to the Orange Path with RCS at<285°F. The first step of P.1 checks RCS pressure at greater than300 psig. Since Pressure is less than 300 psig and RHR flow is>400 gpm, no actions are performed and the operator is directedback to the procedure & step in effect.B - INCORRECT. Cooldown is not limited since the RCS has already experienced alarge break. Plausible if the student does not recognize the exitcriteria for the Red path with a LOCA.C - INCORRECT. A soak is not required since the RCS has already experienced alarge break. Plausible if the student remembers the Red path has asoak requirement.D - INCORRECT. Entry into the procedure is still required and a pressure and flowcheck is made within the procedure. Plausible if the student knowsthe Red Path does not get implemented but incorrect in that thetransition must still be made. Page 168 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP12/#29

REFERENCE:

1-OHP-4023-FR-P.1 KA - 005000 2.4.21 Residual Heat Removal System (RHRS)

Emergency Procedures/Plan Knowledge of the parameters and logic used to assess the status of safety functions,such as reactivity control, core cooling and heat removal, reactor coolant systemintegrity, containment conditions, radioactivity release control, etc.RO - 4.0 SRO - 4.6 CFR - 41.7 / 43.5 / 45.12 KA Justification - Requires the ability to assess the status of PTS and recognize that fora LB LOCA (RHR Flow >400 gpm), that implementation of FR-P.1 isnot required.Original Question # - Cook NRC Exam 2007-094, INPO # 19406 Kewaunee, Unit 1 -12/11/2000, RO26 Audit-80Original Question KA - KA - 000011 EA2.14 Page 169 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

87. 087 010/SRO/OK/DIRECT/RO23 AUDIT-016-3/006000 A2.11/4.0/4.4/H/3Given the following conditions on Unit 1:

• Unit is responding to a LOCA using 1-OHP-4023-E-1, Loss of Reactor or SecondaryCoolant.

• The Unit Supervisor is at Step 11, Initiate Evaluation Of Plant Status.The following plant conditions now exist:

• RWST Level is 42% and lowering

• Containment pressure is 0.3 psig and stable

• Containment Recirc Sump Minimum Recirc Level Lights - NOT LIT

• East RHR Pump Compartment Sump Annunciator - LIT

• East RHR Pump Discharge pressure is 600 psig

• Aux Building area radiation monitors are in alarm Which ONE of the following procedures should the Unit Supervisor transition into from1-OHP-4023-E-1?1-OHP-4023-ECA-1.2, LOCA Outside Containment 1-OHP-4023-ECA-1.3, Sump Blockage Control Room Procedure1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation1-OHP-4023-ES-1.3, Transfer To Cold Leg RecirculationA.B.C.D.ANSWER: AA - CORRECT. Based on RHR Pump Compartment Sump and Aux Building Radalarms being in (along with the AEO report of the leak), transition toECA-1.2, LOCA Outside Containment, would be required.B - INCORRECT. Plausible since the Minimum Sump Recirc Lights are not lit, but thetransition to ECA-1.3 is made only if the Red vortex alarm is lit.C - INCORRECT. Transition to ECA-1.1 is only made in E-1 if one Train combinationof a Recirc Sump Valve and its associated RHR pump is not verifiedavailable. Plausible if the student remembers a transition to thisprocedure exists from E-1 and some of the conditions exist tosupport that transition (however not all of the conditions exist).D - INCORRECT. Transition to ES-1.3 is made per the foldout page when RWST levelis less than 30%. Plausible as ES 1.3 is one of the highest levelEOP procedures and the students know that this procedure willaddress 2 of the other 3 distractors by transitions from ES 1.3. Page 170 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP09/#40, 42

REFERENCE:

1-OHP-4023-E-1 KA - 006000 A2.11 Emergency 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:Rupture of ECCS header RO - 4.0 SRO - 4.4 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Requires the SRO to predict that a LOCA outside Containment existsand that transition to ECA-1.2, LOCA Outside Containment,procedure is required.Original Question # - RO23 AUDIT-016-3 Original Question KA - WE 04 EK3.2 Page 171 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

88. 088 004/SRO/OK/MODIFIED/RO23 AUDIT-068-4/008000 A2.07/2.5/2.8/H/3Given the following conditions on Unit 1:

• The Unit Supervisor entered 1-OHP-4022-016-001, Malfunction Of The CCWSystem, 5 minutes ago.The following plant conditions now exist:

• Reactor Power is 8%

• All RCP Motor Bearing Lube Oil CCW flow low annunciators are lit.

• All flows and temperatures are STABLE as follows: RCP RCP RCP RCP #11 #12 #13 #14o Upper Motor Bearing temps 189 oF 190 oF 205 oF 191 o Fo Motor Bearing Lube Oil CCW flow 4 gpm 4 gpm 3 gpm 3 gpmWhich ONE of the following describes the sequence of actions the Unit Supervisor willprovide to the panel operators?1) Perform a rapid shutdown per 1-OHP-4021-001-006, Rapid Power Reduction2) Trip #13 RCP immediately after the reactor is shutdown.3) Close NRV-163, Loop 13 PZR Spray Control

1) Trip the reactor

2) Go to 1-OHP-4023-E-0, Reactor Trip or Safety Injection3) Trip #13 RCP after Reactor Trip is verified

4) Close NRV-163, Loop 13 PZR Spray Control

1) Trip the reactor

2) Go to 1-OHP-4023-E-0, Reactor Trip or Safety Injection3) Trip all RCPs after Reactor Trip is verified

4) Close NRV-163 and NRV-164, Loop 13 and 14 PZR Spray Control1) Perform a rapid shutdown per 1-OHP-4021-001-006, Rapid Power Reduction2) Trip all RCPs after the reactor is shutdown.

3) Close NRV-163 and NRV-164, Loop 13 and 14 PZR Spray ControlA.B.C.D. Page 172 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: BA - INCORRECT. Reactor and affected RCP must be tripped. Rapid shutdown is notappropriate. Plausible as the rapid down power will not take longfrom 8% power and the temperature of the RCP is stable. Studentmust know this temperature is above the trip setpoint.B - CORRECT. 1-OHP-4022-016-001, Caution prior to Step 1 states that theaffected RCP must be removed from service by performing Step 17if motor bearing temperatures exceed 200 oF. Step 17 andassociated note states that the order is to Trip the Reactor, go toE-0, verify reactor trip, then trip the AFFECTED RCP.C- INCORRECT. Loss of CCW requires trip of Reactor and ALL RCPs, but CCWMalfunction only requires tripping the AFFECTED RCP (RCP #13with the temperature >200 oF). Plausible as all actions are correctexcept stopping all RCP's. This is incorrect and will complicate themitigation strategy of the EOPS but is a fairly routine activity in thesimulator for the students to stop all RCP's vs. just stopping oneRCP.D - INCORRECT. Reactor and affected RCP must be tripped. Rapid shutdown is notappropriate. Combined the plausibility of answers A and C.LESSON PLAN/OBJ: RO-C-AOP-D14/#RO-C-AOP0140412-E3,RO-C-AOP-D8/#RO-C-AOP0420412-E2

REFERENCES:

1-OHP-4022-016-001, Malfunction Of The CCW System FirstCaution, 1-OHP-4021-002-001, RCP Malfunction, Step 1KA - 008000 A2.07 Component Cooling Water System (CCWS)Ability to (a) predict the impacts of the following malfunctions or operations on theCCWS and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:Consequences of high or low CCW flow rate and temperature; the flow rate at whichthe CCW standby pump will start RO - 2.5 SRO - 2.8 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Requires the ability to predict the impact of a loss of CCW flow willhave on the RCPs and to direct the actions required based on thecondition of the RCP following the loss of CCW flow.Original Question # - RO23 Audit-068-4 Original Question KA - APE 026 G2.4.47 Page 173 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

89. 089 003/SRO/OK/DIRECT/RO25 AUDIT-82/034000 K1.01/2.5/3.2/F/3Given the following conditions:

• The refueling crew has just set a new fuel assembly next to an irradiated fuelassembly in the Unit 1 core.

• The manipulator crane operator observes the refueling cavity level lowering rapidly.

• The SRO-CA and Control Room begin implementing 1-OHP-4022-002-006, Loss ofRefueling Water Level during Refueling Operations.The following conditions exist Twenty Minutes Later:

• The Transfer Tube Gate Valve has been closed

• The Weir Gate could NOT be closed

• RCS level is 614' and lowering

• RWST to RHR makeup is in progressBased on these conditions, which ONE of the following describes the correct actionsfor the Control Room SRO?Go to 12-OHP-4022-018-001, Loss of SFP Cooling.

Go to 1-OHP-4022-017-001, Loss of RHR Cooling Direct the SRO-CA to verify integrity of the Refueling Cavity SealDirect the SRO-CA to check for misalignment of the Refueling Cavity DrainsA.B.C.D.ANSWER: BA - INCORRECT. Since the Transfer Tube is Isolated, The Spent Fuel Pool has beenisolated even if the Weir gate is not isolated. Plausible as this is atransition from the procedure in use.B - CORRECT. If RCS Level can NOT be maintained >614' then an RCS leak is inProgress and 1-OHP-4022-017-001 Loss of RHR Cooling Must beInitiated. C - INCORRECT. Even though the Refueling Cavity Drains/Piping are below the 621'elevation, the physical construction fo the Refueling cavity will notallow an inadvertent draining of the Refueling Cavity to lower levelbelow the Reactor Vessel Flange (~621' elevation). Plausible as thisis a likely source of leakage.D - INCORRECT. The Seal is > 620" elevation so even if it were failed the level dropwould have stopped at 620'. Plausible as this is a likely source ofleakage Page 174 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D9/#RO-C-AOP0130412-E3

REFERENCE:

RO-C-AOP-D9; 1-OHP-4022-002-006 MODIFIED: Remove distractor relating to nozzle dams and replaced withdistractor pertaining to Refueling cavity Drains.KA - 034000 K1.01 Fuel Handling Equipment System (FHES)

Knowledge of the physical connections and/or cause-effect relationships between theFuel Handling System and the following systems:

RCS RO - 2.5 SRO - 3.2 CFR - 41.2 to 41.9 / 45.7 to 45.8 KA Justification - Requires the knowledge of the interconnections between the RCSwater inventory and the Transfer Canal water inventory. Based onthese interrelationships, question requires the SRO to determineappropriate procedure for a leak that affects both water inventories.Original Question # - RO25 AUDIT-82 Original Question KA - 000036 2.4.4 Page 175 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

90. 090 003/SRO/OK/DIRECT/ITS FINAL 3.7-21/039000 2.2.37/3.6/4.6/H/3Given the following conditions on Unit 2:

• Unit is in MODE 4 and preparing to startup following a refueling outage.

• During the outage, all 5 Main Steam Safety Valves (MSSVs) associated with SG21were removed and inspected.

• The maintenance was satisfactory and the MSSVs will function if required.

• The Inservice Testing of the MSSV setpoints has NOT been performed.

• A risk assessment has NOT been performed.Without reliance on SR 3.0.3, which ONE of the following describes if the reactorstartup can proceed to MODE 3 with the MSSVs in this condition?Reference Provided: Unit 2 TS Section 1.4, Frequency, Section SR 3.0,Surveillance Requirements, and TS 3.7.1, Main Steam Safety ValvesNo. The TS ACTIONS must be immediately entered and all portions of the posttesting must be completed before entering MODE 3.

No. An alternate method of setpoint verification must be used and MSSVOPERABILITY must be demonstrated before entering MODE 3.Yes. However, when the unit reaches MODE 2, the test must be at least startedwithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.

Yes. However, when test conditions can be established, the test must becompleted prior to MODE 2.A.B.C.D.ANSWER: DA - INCORRECT. Entry into Mode 3 is permitted by the Note prior to SR 3.7.1.1.Plausible as the LCO applicability is Modes 1-3.B - INCORRECT. Entry into Mode 3 is permitted by the Note prior to SR 3.7.1.1.Plausible as the LCO applicability is Modes 1-3 and distractorprovides a logical method of compliance.C - INCORRECT. Surveillance requirement must be complete PRIOR to entry intoMode 2. Plausible as this applies the missed Surveillancerequirement to the distractor.D - CORRECT. SR 3.7.1.1 NOTE states that the Surveillance Requirements areonly required to be performed in MODES 1 and 2. Per Example1.4-5, this note allows entry into Mode 3 to perform the surveillance. Page 176 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05103/#RO-C-05103-E11

REFERENCE:

TS SR 3.0.1 and Note to SR 3.7.1.1; TS Example 1.4-5Reference Provided: Unit 2 TS Section 1.4, Frequency, Section SR 3.0,Surveillance Requirements, and TS 3.7.1, Main Steam Safety ValvesKA - 039000 2.2.37 Main and Reheat Steam System (MRSS)

Equipment ControlAbility to determine operability and/or availability of safety related equipment.RO - 3.6 SRO - 4.6 CFR - 41.7 / 43.5 / 45.12 KA Justification - Requires SRO to apply TS for MSSVs and determine operability andrestrictions based on the status of the valves.Original Question # - ITS FINAL 3.7-21 Original Question KA - 2.1.12 Page 177 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

91. 091 007/SRO/OK/NEW/NEW/045000 2.4.35/3.8/4.0/F/3Given the following conditions on Unit 2:

• A Main Generator fire has been confirmed by an AEO who reported that flames arecoming out of the Unit 2 Main Generator Shaft.After directing a Reactor/Turbine Trip and entering 2-OHP-4023-E-0, Reactor TripResponse, which ONE of the following would you direct the AEO to perform first?Dispatch an AEO to:Shut down the Main Generator Seal Oil System per 2-OHP-4021-080-002Operation Of Shaft Seal Oil System Depressurize and Purge the Main Generator per 2-OHP-4022-053-002 EmergencyDegassing Of The Electrical Generator Start All Fire Pumps per 12-OHP-4021-066-001, Fire Protection System (Water)Operation.

Shut down the Generator Condition Monitor at 2-GCM-AARP-11, GeneratorCondition Monitor Auto Alarm Remote Panel A.B.C.D.ANSWER: BA - INCORRECT. Eliminate fuel for fire (H2) first. Plausible as this removes a sourceof ignition from the fire area but this would allow H2 to escape intothe fire area.B - CORRECT. The major concern is the fire. The source of the fire is the H2 gasfrom the generator. The first action must be to eliminate the sourceof the H2 by degassing the main generator.C - INCORRECT. Eliminate fuel for fire (H2) first. Plausible as this would aid in puttingthe fire out but all pumps are not requiredD - INCORRECT. Eliminate fuel for fire (H2) first. Plausible as this is a required actionwhen degassing the generator. Page 178 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-AOP-D14/RO-C-AOP0510412-E2

REFERENCE:

2-OHP-4022-053-002, Emergency Degassing Of The ElectricalGeneratorKA - 045000 2.4.35 Main Turbine Generator (MT/G) System Emergency Procedures/Plan Knowledge of local auxiliary operator tasks during an emergency and the resultantoperational effects.

RO - 3.8 SRO - 4.0 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Justification - Requires the knowledge of emergency task (emergency degas ofmain generator) associated with the Main Turbine Generator (MT/G)System.Original Question # - New Original Question KA - New Page 179 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

92. 092 003/SRO/OK/DIRECT/ITS FINAL 3.7-3/061000 A2.04/3.4/3.8/H/3Given the following conditions on Unit 2:

• A plant heatup is in progress following a forced shutdown

• Reactor coolant average temperature at 450

°F when the following AuxiliaryFeedwater (AFW) trains become inoperable:o 0100 on July 7, TDAFW train is declared inoperable due to 2-MCM-221 steamsupply valve being inoperable.o 1830 on July 8, 2E AFW train is declared inoperable.o 1900 on July 8, 2-MCM-221 steam supply valve is restored to OPERABLEstatus.Including any extensions permitted by TS, and without re-entering a TechnicalSpecification condition requiring a plant shutdown, the 2E AFW train must be restoredto OPERABLE status by __________________.

Reference Provided: Unit 2 TS 3.7.5 Auxiliary Feedwater System, TS Section 1.3Completion Times0100 on July 10.

0100 on July 17.

1830 on July 11.

1830 on July 12.A.B.C.D.ANSWER: CA - INCORRECT. This is 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) from the first event. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock forthe 2E AFW pump started at the time the 2E pump was declaredINOPERABLE.B - INCORRECT. This is applying the TDAFP 10 days from discovery of failure tomeet LCO.C - CORRECT. 2E Pump 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock started when the 2E pump becameINOPERABLE. The clock does not get reset with the OPERABILITY of the TDAFP.D - INCORRECT. This applies the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock for the 2E pump plus a 24 hourextension per TS Section 1.3, Completion Times. Page 180 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-05600/#13, RO-C-TS01/RO-C-TS01-E11

REFERENCE:

TS 3.7.5, Required Action B.1; TS Example 1.3-3Reference Provided: Unit 2 TS 3.7.5 Auxiliary Feedwater System, TS Section 1.3,Completion TimesKA - 061000 A2.04Auxiliary / Emergency Feedwater (AFW) SystemAbility to (a) predict the impacts of the following malfunctions or operations on the AFWSystem and (b) based on those predictions, use procedures to correct, control, ormitigate the consequences of those malfunctions or operations:pump failure or improper operation RO - 3.4 SRO - 3.8 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Requires TS knowledge and ability to determine operability actionstimes for AFW pumps failures and to determine the required time perTS and procedures that an AFW pump must be made OPERABLE.Original Question # - ITS FINAL 3.7-3 Original Question KA - 2.1.12 Page 181 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

93. 093 005/SRO/OK/DIRECT/RO24 AUDIT-096-8/011000 A2.11/3.4/3.6/H/3Given the following conditions on Unit 1:

• The plant is operating at 6% power preparing for Turbine roll.

• NLP-151, PZR Level Channel 1 failed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ago. The bistables have been trippedand all actions are complete as per 1-OHP-4022-013-010, Pressurizer LevelInstrument Malfunction.

• PZR level is currently 40% on the remaining PZR Level channels.Which ONE of the following describes the effects on the plant if NLP-153, PZR LevelChannel 3 fails low and the affect on Unit Supervisor's decision to trip bistables for theChannel 3 failure?

Note: Assume NO operator actions.

Reference Provided: Unit 1 TS Section 3.0 Limiting Condition for OperationApplicability and TS 3.3.1 Reactor Trip System InstrumentationLetdown will Isolate and heaters will de-energize.

Bistables may be tripped without causing a reactor trip. Power must remain lessthan 10%.

Letdown will remain in service and heaters will de-energize.Bistables should NOT be tripped since a reactor trip will be generated. Power mustremain less than 10%.

Letdown will remain in service and heaters will de-energize.Bistables may be tripped without causing a reactor trip. Power must be reduced toless than 5%.

Letdown will Isolate and heaters will de-energize.

Bistables should NOT be tripped since a reactor trip will be generated. Power mustbe reduced to less than 5%.A.B.C.D. Page 182 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: AA - CORRECT. With Channel 1 NLP-151 in the tripped condition the High level RxTrip signal will be made up for 1 channel (1/2 coincidence onremaining channels). The level control selector switch for thepressurizer is in the 2/3 position with channel 3 NLP-153 as thecontrolling channel. When it fails low letdown will isolate and theheaters will de-energize. When the bistables are tripped, a reactortrip signal will be generated but it is blocked by P-7 (Reactor andTurbine power both below 10%). Plant startup can NOT continue.Power must be maintained below 10% (P-7).B - INCORRECT. Letdown will isolate. Reactor will not trip (See Answer A). Plausibleas student could be focused on Technical Specification impacts andnot consider operational impacts causing Letdown isolation.C - INCORRECT. Letdown will isolate. Power does not need to be reduced. Powerjust must remain less than P-7 (10%). Plausible as student couldconfuse Mode change requirements with Tech Spec compliance.D - INCORRECT. Reactor will not trip (See Answer A). Power does not need to bereduced. Power just must remain less than P-7 (10%). Plausible asstudent could believe a trip signal could be generated with theadditional bistable actuation. Page 183 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00202/#RO-C-00202-E17; RO-C-TS01/RO-C-TS01-E11

REFERENCE:

SOD-00202-003, Pressurizer Level Control TS 3.3.1 actions D & N. Table 3.3.1-1 Function 9 Reference Provided: Unit 1 TS Section 3.0 Limiting Condition for OperationApplicability and TS 3.3.1 Reactor Trip System InstrumentationKA - 011000 A2.11 Pressurizer Level Control System (PZR LCS)Ability to (a) predict the impacts of the following malfunctions or operations on the PZRLCS and (b) based on those predictions, use procedures to correct, control, or mitigatethe consequences of Failure of the PZR level instrument - low RO - 3.4 SRO - 3.6 CFR - 41.5 / 43.5 / 45.3 / 45.13 KA Justification - Requires the ability to predict the impacts of multiple PZR Levelchannel failures (including a low failure), and to use the AOPs todetermine the appropriate actions.Original Question # - INPO - DIRECT 5314, NRC02-045-3 (SRO41), RO24Audit-096-8Original Question KA - 000028 - G2.2.22 Page 184 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

94. 094 003/SRO/OK/DIRECT/RO25 AUDIT-78/194001 2.1.23/4.3/4.4/H/4Given the following conditions on Unit 1:

• Unit is in Mode 6

• Refueling cavity is filled to 24.2 feet above flange

• Core reload is in progress

• RCS temperature is 82°F

• 1W RHR is OOS due to an Oil Leak

• 1E RHR train is in operationA leak has been reported on the 1E RHR pump mechanical seal heat exchanger. Torepair the leak, the RHR pump must be stopped. Maintenance estimates it will take 40minutes to complete repairs.

1. How does this affect the ability to continue core reload?2. What is the basis for having one RHR loop in operation in this condition?1. Core reload must be stopped.

2. Provides for adequate RCS mixing and control of reactor coolant temperature.1. Core reload must be stopped.

2. Ensures that a core Keff of less than or equal to 0.95 is maintained during fuelhandling operations.1. Core reload may continue provided no operations are permitted that would dilutethe refueling cavity boron concentration.2. Provide for adequate RCS mixing and control of reactor coolant temperature.1. Core reload may continue provided no operations are permitted that would dilutethe refueling cavity boron concentration.2. Ensures that a core Keff of less than or equal to 0.95 is maintained during fuelhandling operations.A.B.C.D. Page 185 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: AA - CORRECT. Because both pumps are inoperable, core loading must stop. If thepump were not inoperable, core loading is allowed to continuewithout RHR for up to one hour provided no change in BoronConcentration. This allows loading near the edges of the corewhere flow may interfere with setting fuel assemblies. Theindication and control of temperature is one of the bases for thisLCO. Maintaining core Keff </=0.95 is the bases of LCO 3.9.1which is boron concentration during refueling operations. Theconfusion is that the RHR bases is for MIXING of the borated waterto prevent potential criticality.B - INCORRECT. The bases is for RCS Temperature control. Plausible as the firstportion of the answer is correct and the second portion is the basisfor maintaining the Mode.C - INCORRECT. Core Reload Must be Stopped. Plausible as this combines some ofthe actions in the applicable spec but the actions are not in line withthe required work activities.D - INCORRECT. Core Reload Must be Stopped & The bases is for RCSTemperature control. Plausible as this combines portions ofanswers B and C.LESSON PLAN/OBJ: RO-C-01700/#13 & 15

REFERENCE:

Tech Specs & Bases 3.9.4 KA - 194001 2.1.23 Generic Conduct of OperationsAbility to perform specific system and integrated plant procedures during all modes ofplant operation.

RO - 4.3 SRO - 4.4 CFR - 41.10 / 43.5 / 45.2 / 45.6KA Justification - Requires the ability to determine if refueling cancontinue based on equipment availability (Integrated Plant Operation)and the basis for limiting operations.Original Question # - CATAWBA2005, RO25 AUDIT-78Original Question KA - 000025 2.2.25 Page 186 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

95. 095 095/SRO/OK/DIRECT/KEWAUNEE-962002/194001 2.1.43/4.1/4.3/H/3Given the following conditions on Unit 2:

• A normal plant startup and power escalation to 100% was initiated following arefueling outage.

• The reactor achieved 100% rated power (3468 MWth) approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ago.

• MTI has informed the Unit Supervisor the Blowdown flow instrument for SG21 isreading approximately 100 gpm higher than actual blowdown flow. The Unit Supervisor should: direct the control room operator to raise power slightly since actual thermal power isless than PPC calculated thermal power of 3468 MWth. direct an immediate power reduction to ensure the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average does not exceed3468 MWth.

order that no reactor power adjustments be made for the next 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> so anaccurate 1-hour power average is obtained.

order that no reactor power adjustments be made for the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and thenmake adjustments to power as required. A.B.C.D.ANSWER: B A - INCORRECT. The blowdown flow error will result in PPC derived thermal power tobe nonconservatively low. This may result in actual thermal poweralready being greater than 3468 MWth. A power rise would onlymake matters worse.B - CORRECT. The blowdown flow error will result in PPC derived thermal power tobe nonconservatively low. This may result in actual thermal poweralready being greater than 3468 MWth. Power must be reduced toensure plant is operating that less than 3468 MWth actual power.C - INCORRECT. Holding power stable for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> will only keep the plant operatingwith actual power greater than 3468 MWth for a longer period oftime. Power must be reduced to less than 3468 MWth.D - INCORRECT. Holding power stable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> will only keep the plant operatingwith actual power greater than 3468 MWth for a longer period oftime. Power must be reduced to less than 3468 MWth. Page 187 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-ADM02/ #24

REFERENCE:

2-OHP-4021-011-001, At-Power Operation, Including LoadSwingsKA - 194001 2.1.43 Generic Conduct of OperationsAbility to use procedures to determine the effects on reactivity of plant changes, suchas reactor coolant system temperature, secondary plant, fuel depletion, etc.RO - 4.1 SRO - 4.3 CFR - 41.10 / 43.6 / 45.6 KA Justification - Requires ability to determine the affect on reactivity form anerroneous calculation on secondary flow (SG Blowdownmiscalculated) and (SRO) determine the required actions based onthe reactivity anomaly.Original Question # - KEWAUNEE-962002, RO26 AUDIT-DRAFT-95Original Question KA - 194001 2.1.7 Page 188 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

96. 096 005/SRO/OK/DIRECT/CM-1140-31956/194001 2.2.40/3.4/4.7/H/3Given the following conditions in Unit 2:

• The unit was stable at 95% power with Rod Control in AUTO* Control Bank D is at 220 steps with AFD at -6.5%

• A HIGH failure of Power Range channel N41 occurred* The reactor operator responded by placing Rod control in MANUAL 15 seconds afterthe event.Which ONE of the following may require prompt crew actions to ensure continuedcompliance with Technical Specifications?

Reference Provided: U2 COLR (Cycle 18) and TDB 2-Figure 13.1, Target Bandand ARMAxial Flux Difference (AFD)

Quadrant Power Tilt Ratio (QPTR)

Rod Insertion Limits (RIL)

Shutdown Margin (SDM)A.B.C.D. Page 189 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: A A - CORRECT. AFD is close to the -7.2 limit of TDB 2-Figure 13.1, for 95% power. Based on the The operator response time and the maximum rodspeed of 72 steps/minute, the rods would travel approximately 18steps into the core. Rod motion would place the AFD outside thetarget band which is has a 15 minute requirement to restore AFDwithin the band or be below 90% in the next 15 minutes. TS 3.2.3,Axial Flux Difference (AFD)B - INCORRECT. Failure will cause rod motion. Rod motion is a symmetrical eventwhich should not lead to a QPTR concern. Plausible as there areactions required per Technical Specifications for QPTR but they arenot prompt (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).C - INCORRECT. Rod insertion Limit at 95% is approximately 180 steps on controlbank D. With rods starting at 220 and only traveling 18 steps, theminimum rod height for this event is 202 steps on control bank D,which is above the RIL requirements of Technical Specifications.Plausible RIL are a concern with any rod insertion with the unit athigh power.D - INCORRECT. As long as rods are above the RIL then SDM should be maintained. See Answer C justification. Plausible as students confuse SDMwith rod motion and students know the SDM is short time frameLCO.LESSON PLAN/OBJ: RO-C-AOP-D7/#RO-C-AOP0300412-E2

REFERENCE:

U2 COLR (Cycle 18) and TDB 2-Figure 13.1, TS 3.2.3, Axial FluxDifference (AFD)Reference Provided: U2 COLR (Cycle 18) and TDB 2-Figure 13.1, Target Bandand ARMKA - 194001 2.2.40 Generic Equipment ControlAbility to apply Technical Specifications for a system.RO - 3.4 SRO - 4.7 CFR - 41.10 / 43.2 / 43.5 / 45.3

KA Justification - Requires the ability to apply TS 3.2.3, Axial Flux Difference (AFD).Original Question # - CM-1140-31956 Original Question KA - 2.1.11 Page 190 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

97. 097 002/SRO/OK/MODIFIED/NRC EXAM 2008-97/194001 2.3.5/2.9/2.9/H/3A LOCA that resulted in significant core damage occurred at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />. ContainmentPressure and Radiation levels were recorded as follows: Radiation Pressure Time (R/Hr) (psig) 1600 420,000 6.2 1630 420,000 6.2 1700 350,000 5.8 1730 280,000 5.3 1800 260,000 4.8 1830 120,000 4.6 1900 90,000 4.3 1930 90,000 4.0 2000 90,000 3.9 At 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, while performing Emergency Operating Procedures, a step isencountered which states "Check PZR level - GREATER THAN 20% [24%ADVERSE]".

Which ONE of the following describes the required Pressurizer level and why?20% because the Containment Radiation levels are no longer above the Adversesetpoint requirement.

24% because adverse values must be used until evaluated for lasting effectsbecause the integrated dose limit has been exceeded.24% because adverse containment exists due to the current containment radiationdose rate.

20% because the Containment Pressure is no longer above the Adverse setpointrequirementA.B.C.D. Page 191 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination ANSWER: B A - INCORRECT. The integrated dose is (1,015,000 R) which is greater than 10 6 R, soadverse containment values must be used. Plausible as the instantvalues for dose do not require the use of Adverse numbersB - CORRECT. Adverse containment values are required to be used whencontainment pressure is >5 psig or >105 R/Hr. When pressurelowers to <5 psig normal values may be used as long as theintegrated dose is <10 6 R. The integrated dose is (1,015,000 R)which is greater than 10 6 R, so adverse containment values must beused.C - INCORRECT. The current Dose Rate is <105 R/Hr. Plausible as the instant valuefor Containment pressure is close to the adverse setpointD - INCORRECT. Pressure is <5 psig. Plausible as the instant value for containmentpressure is close to the adverse setpoint.LESSON PLAN/OBJ: RO-C-EOP01/#8 & #9

REFERENCE:

OHI-4023, Abnormal / Emergency Procedure User's Guide,Attachment 2, Step 6, RO-C-EOP01Modified: Raised dose rates so that the integrated dose becomes >106 Rchanging the correct answer to B. Added one more hour of readings.Changed Distractor A & D.KA - 194001 2.3.5 Generic Radiation ControlAbility to use radiation monitoring systems, such as fixed radiation monitors andalarms, portable survey instruments, personal monitoring equipment, etc.RO - 2.9 SRO - 2.9 CFR - 41.11 / 41.12 / 43.4 / 45.9 KA Justification - Requires the ability to use fixed rad monitor (Containment HighRange) to determine whether adverse containment conditions exist.Original Question # - Cook NRC Exam 2004 118-2(SRO87), NRC EXAM 2008-97Original Question KA - 194001 2.3.5 Page 192 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

98. 098 003/SRO/OK/DIRECT/AOP1CAOP7.10-1/194001 2.3.14/3.4/3.8/F/3Which ONE of the following responses correctly reflects the bases for Reactor CoolantSpecific Activity in Technical Specifications?The short-lived radioactive isotope fission products will have decayed prior to anyfuel movement.

Limitations on specific activity in the RCS reduces corrosion product activation andsubsequent RCS integrity challenge.

During a LOCA the dose will NOT exceed the 10CFR20 limits at the site boundary.During a steam generator tube rupture the release of activity through theatmospheric relief valves will NOT exceed 10CFR100 limits.A.B.C.D.ANSWER: DA - INCORRECT. See Answer D. Plausible as this is a concern during refuelingevolutions for dose to employees moving fuel.B - INCORRECT. See Answer D. Plausible as this identifies a challenge to corrosionaffecting the metal in the RCS for long term RCS operation.C - INCORRECT. See Answer D. Plausible for accident dose concerns but the CFRreference is in correct.D - CORRECT. The maximum dose to the whole body and the thyroid that anindividual at the site boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during anaccident is specified in 10 CFR 100 (Ref. 1). The limits on specificactivity ensure that the doses are held to a small fraction of the 10CFR 100 limits during analyzed transients and accidents.The RCS specific activity LCO limits the allowable concentrationlevel of radionuclides in the reactor coolant. The LCO limits areestablished to minimize the offsite radioactivity dose consequencesin the event of a steam generator tube rupture (SGTR) accident. Page 193 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-00200/# RO-C-00200-E10,RO-C-AOP-D8/#RO-C-0030500401-E3

REFERENCE:

TS 3.4.16 Bases KA - 194001 2.3.14 Generic Radiation Control Knowledge of radiation or contamination hazards that may arise during normal,abnormal, or emergency conditions or activities.

RO - 3.4 SRO - 3.8 CFR - 41.12 / 43.4 / 45.10 KA Justification - Requires the Bases (SRO) knowledge for the TS limits on RCSActivity relating the radiological hazards of a SGTR and release to theatmosphere.Original Question # - CM-1166-31980, AOP1CAOP7.10-1Original Question KA - 2.1.12 Page 194 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination

99. 099 003/SRO/OK/DIRECT/NRC EXAM 2002-025-12/194001 2.4.4/4.5/4.7/F/3Given the following conditions on Unit 2:

• A LOCA has occurred

• The STA is monitoring the Critical Safety Functions and notes the followingindications: o WR log power 0% o WR startup rate Negative o Containment Pressure 13 psig o CETC's 5 highest 760°F o RVLIS Narrow Range 76% o Pressurizer Level 0% o RCS Pressure 480 psig o AFW Flow 300 x103 pph SG21 SG22 SG23 SG24 o Narrow Range SG Levels 12% 15% 16% 12%Given the conditions described above, to which ONE of the following proceduresshould the SRO transition?2-OHP-4023-FR-C.2, Response to Degraded Core Cooling2-OHP-4023-FR-I.2, Response to Low Pressurizer Level2-OHP-4023-FR-Z.1, Response to High Containment Pressure2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat SinkA.B.C.D.ANSWER: C A - INCORRECT. 2-OHP-4023-FR-C-2 is identified by RCS Temp >752°F and RVLIS>46% but it is an ORANGE path so it has a lower priority.B - INCORRECT. 2-OHP-4023-FR-I-2 is indicated by Pressurizer level <17% but it is aYELLOW path so it has a lower priority.C - CORRECT. Containment Pressure of >12 psig is a RED path requiring2-OHP-4023-FR-Z-1.D - INCORRECT. 2-OHP-4023-FR-H-1 is not indicated. ALL SGs are <24% (adverse)but AFW flow is sufficient and so the only H series procedure wouldbe a YELLOW path for 2-OHP-4023-FR-H-5. Page 195 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP01/#23

REFERENCE:

OHI-4023, Abnormal /Emergency Procedure User's Guide,Attachment 5, Critical Safety Function Status TreesKA - 194001 2.4.4 Generic Emergency Procedures/PlanAbility to recognize abnormal indications for system operating parameters that areentry-level conditions for emergency and abnormal operating procedures.RO - 4.5 SRO - 4.7 CFR - 41.10 / 43.2 / 45.6 KA Justification - Question tests the ability of the SRO to evaluate the plant conditionsand determine the required procedural transition.Original Question # - NRC Exam 2002-025-12 Page 196 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination 100. 100 004/SRO/OK/DIRECT/RO24 AUDIT 085-11/194001 2.4.5/3.7/4.3/F/3Given the following conditions on Unit 2:

• The unit has tripped and experienced a safety injection.

• While performing 2-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization,an ORANGE path condition was noted for the Core Cooling Critical Safety Function.

• 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, was entered. While performing steps of this procedure, the Shift Technical Advisor reports thefollowing:

• RED path condition exists for Core Cooling Critical Safety Function.

• RED path condition exists for Containment Critical Safety Function.

• NO other abnormal conditions were noted. Based on these plant conditions, which ONE of the following is the appropriate actionfor the Unit Supervisor to take? Complete actions of 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling,and then transition to 2-OHP-4023-FR-Z.1, Response to Containment HighPressure.

Complete actions of 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling,and then transition to 2-OHP-4023-FR-C.1.

Stop performing 2-OHP-4023-FR-C.2, Response to Degraded Core Cooling, andimmediately transition to 2-OHP-4023-FR-Z.1, Response to Containment HighPressure.

Stop performing 02-OHP-4023-FR-C.2, Response to Degraded Core Cooling, andimmediately transition to 2-OHP-4023-FR-C.1, Response to Inadequate CoreCooling.

A.B.C.D.ANSWER: DA - INCORRECT. FR-C-1 is higher Priority than both FR-C.2 and FR-Z.1.B - INCORRECT. FR-C-1 is higher Priority than FR-C.2C - INCORRECT. FR-C.2 is higher priority than FR-Z.1D - CORRECT. Per OHI-4023, Rules of usage, when a Red path is encountered,immediately initiate the FRP. Because FR-C. 1 is a higher prioritythan FR-Z. 1, the US should proceed to FR-C. 1 vs. FR-Z. 1. Page 197 of 198 Test Date: 7/2/2010 Cook 2010 NRC Examination LESSON PLAN/OBJ: RO-C-EOP01/#22

REFERENCE:

OHI-4023 Abnormal/Emergency Procedure User's Guide,Attachment 5 Section 5KA - 194001 2.4.5 Generic Emergency Procedures/Plan Knowledge of the organization of the operating procedures network for normal,abnormal, and emergency evolutions.

RO - 3.7 SRO - 4.3 CFR - 41.10 / 43.5 / 45.13 KA Justification - Requires knowledge of the priority (organization) of the FunctionalRestoration Procedures (FRPs) and the ability to apply thisknowledge to determine the appropriate procedure to implement.Original Question # - RO24 AUDIT 085-11 Original Question KA - W/E03.EA2.1 Page 198 of 198 Test Date: 7/2/2010