ML13066A354

Final Written Examination with Answer Key (401-5 Format) (Folder 3)

ML13066A354
Person / Time
Site:	Indian Point
Issue date:	02/05/2013
From:	Entergy Nuclear Operations
To:	D'Antonio J Operations Branch I
Jackson D
References
TAC U01866
Download: ML13066A354 (202)

Text

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000008K202 Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: - Sensors and detectors Importance 2.7 2.7 Question # 1 Operators suspect a vapor space leak through either a Pressurizer Safety or PORV. What indication combinations are available to help the operator determine which valve is faulted?

ACOUSTIC MONITOR TAILPIPE TEMPERATURE A. each safety each safety each PORV common PORV line B. common safety line common safety line each PORV each PORV C. each safety each safety common PORV line common PORV line D. common safety each safety line common PORV line each PORV Answer: A Exp lanationlJustification:

A. Correct.

B. Incorrect. Plausible because the PORVs have a single common temperature sensor and each PORV and Safety Valve has its own Acoustic Monitor.

C. Incorrect. Plausible because the PORVs have a single common temperature sensor and each PORV and Safety Valve has its own Acoustic Monitor.

D. Incorrect. Plausible because the PORVs have a single common temperature sensor and each PORV and Safety Valve has its own Acoustic Monitor.

Technical

References:

Syst Desc 1.4 Proposed References to be provided: None Learning Objective 13LP-ILO-RCSPZR 2 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000009K324 Knowledge of the reasons for the following responses as they apply to the small break LOCA: - ECCS throttling or termination criteria Importance 4.1 4.6 Question # 2 The crew is performing ECCS reduction in ES-1.2, Post LOCA Cooldown and Depressurization. One SI pump has been stopped.

Which of the following describes how the required subcooling changes and the basis for the amount of subcooling required for securing the second SI pump?

A. Required Subcooling increases to compensate for the anticipated void formation after the second pump is stopped.

B. Required Subcooling increases to ensure the RCS will remain subcooled after the second SI pump is stopped.

C. Required Subcooling decreases due to decreased break flow after stopping the second pump.

D. Required Subcooling decreases to allow pressure to decrease sufficiently after stopping the second pump to place RHR cooling in service.

Answer: B ExplanationlJustification:

A. Incorrect. Plausible because required subcooling increases; however, no void formation is expected or should exist because the RCS is subcooled.

B. Correct. because throttling SI flow is not practical, pumps must be stopped. As pumps are stopped, a step decrease in flow is experienced. RCS pressure will decrease to reach equilibrium with break flow and ECCS 'flow. RCS subcooling

and pressure control must be maintained as the reduction takes place. As flow is reduced, a large allowance is needed.

C. Incorrect. Plausible because breakflow does decrease after stopping each pump; however, subcooling requirement increases.

D. Incorrect Plausible because the most likely exit from ES-1.2 is starting an RHR pump in injection mode; however, subcooling requirement increases.

Technical

References:

2-ES-1.2 3-ES-1.2 Proposed References to be provided: None Learning Objective 12LP-ILO-EOPS121 Question Source: Bank Question History: NRC Exam IPEC Unit 22012 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000011A208 Ability to determine and interpret the following as they apply to a Large Break LOCA: - Conditions necessary for recovery when accident reaches stable phase Importance 3.4 3.9 Question # 3 Given the following:

• A large break LOCA has occurred.

• The 31 spray pump is out of service.

• All other ESF equipment started on demand.

• The crew has entered ES-1.3, Transfer to Cold Leg Recirculation.

Which of the following correctly describes the containment spray pump operation when recirc switch 1 is placed to on during this transfer?

A. Spray pump 32 will continue to operate and must be manually tripped when RWST level reaches 1.5 feet.

8. Spray pump 32 will trip when recirc switch 1 is placed to on. The pump must be restarted and must be manually tripped when RWST level reaches 1.5 feet.

C. Spray pump 32 will continue to operate and will automatically trip when RWST level reaches 1.5 feet.

D. Spray pump 32 will trip when recirc switch 1 is placed to on. The pump must be restarted and will automatically trip when RWST level reaches 1.5 feet.

Answer: A Explanation/Justification:

A. Correct

B. Incorrect. Plausible because 32 Spray Pumps is normally secured when Recirc Switch 1 is place to on (assuming both pumps are running). If 31 spray pump is not running then 32 spray pump is not secured. It must be manually secured when RWST level is 1.5 feet.

C. Incorrect. Plausible because the pump will continue to operate but it will not automatically trip.

D. Incorrect. Plausible because 32 Spray Pumps is normally secured when Recirc Switch 1 is place to on (assuming both pumps are running). If 31 spray pump is not running then 32 spray pump is not secured. It must be manually secured when RWST level is 1.5 feet.

Technical

References:

3-ES-1.3 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE10 22 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000056A110 Ability to operate and/or monitor the following as they apply to the Loss of Offsite Power:

Auxiliary/emergency feedwater pump (motor driven)

Importance 4.3 4.3 Question # 4 The plant is at 25% power, on chemistry hold for secondary cation concentration, Action Level 1. A loss of offsite power occurs. The plant remains on-line and all equipment operates as designed.

What is the configuration of the AFW system?

A. 32 and 33 AFW pumps are running, supplying auxiliary feedwater to all steam generators.

B. 31 and 32 AFW pumps are running, only 31 and 32 Steam Generators are being supplied auxiliary feedwater.

C. 32 and 33 AFW pumps are running, only 33 and 34 Steam Generators are being supplied auxiliary feedwater.

D. 31, 32 and 33 AFW pumps are running, supplying auxiliary feedwater to all steam generators.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because 32 and 33 AFW pumps are running; however, only 33 and 34 SGs are being supplied AFW from 33 AFW pump.

B. Incorrect. Plausible because 32 SG would be running. Candidates must remember the power supplies to the motor driven AFW pumps and which SGs each one feed.

C. Correct. 32 and 33 AFW pumps start on Non-SI Blackout signal from bus 6A; AFW pump 33 feeds SGs 33 and 34.

D. Incorrect. Plausible because candidate must remember which buses remained energized on loss of oft-site power, and the electric plant configuration at 25% power.

Technical

References:

Logic Unit 3 Sheet 7 Logic Unit 3 Sheet 7a Logic Unit 3 Sheet 8 Logic Unit 3 Sheet 8A Logic Unit 3 Sheet 8B Proposed References to be provided: None Learning Objective 13LP-ILO-AFW001 2 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000022K306 Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: - RCP thermal barrier cooling Importance 3.2 3.3 Question # 5 Which of the following identifies the first pump checked, and if necessary started, in 3-AOP-480V-1, Loss of Normal Power to Any 480V Bus and why this pump is addressed first?

A. A charging pump to ensure RCS inventory is adequate to prevent letdown isolation B. A charging pump to ensure RCP Seal cooling if CCW to thermal barrier also lost C. A Service Water Pump to ensure cooling is maintained to Hydrogen Seal Oil Coolers D. A Service Water Pump to ensure cooling to Emergency Diesel Generators Answer: B Explanation/Justification:

A. Incorrect. Plausible because letdown isolation will occur after at least 32 minutes. Also letdown isolation will protect necessary equipment (heaters).

B. Correct. Historically, 625 has closed when standby CCW pump starts. This could result in a loss of Seal Injection and Thermal Barrier Cooling.

C. Incorrect. Plausible because essential service water supplies Hydrogen Seal Oil Coolers and personnel safety is a concern if cooling is lost to hydrogen seal oil coolers.

D. Incorrect. Plausible because essential service water does supply the emergency diesel which should have started to energize the bus Technical

References:

3-AOP-480V-1 Proposed References to be provided: None Learning Objective 13LP-ILO-RCSRCP F Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000025K101 Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System:

- Loss of RHRS during all modes of operation Importance 3.9 4.3 Question # 6 Given

• Unit 3 is in a refueling outage.

• Unit 3 had been operating at 100% power for the past year.

• All activities occur on schedule for a 22 day outage.

• RHR is lost.

Which of the following initial plant conditions will result in the shortest time to boil?

A. Just after securing the final RCP following crud burst.

B. During Reactor Vessel Stud detensioning.

C. Just after vacuum is met to start vacuum refill.

D. Right after the Reactor Head is initially lifted.

Answer: C Explanation/Justification:

A. Incorrect but plausible since this choice has the highest decay heat.

B. Incorrect but plausible since this is close to the shortest time to boil for the outage.

C. Correct answer per outage risk analysis

D. Incorrect but plausible since this has a short time to boil and very high risk significance.

The outage risk analysis for the upcoming 22 day outage identifies vacuum refill as the shortest time to boil activity. This has been the case for every outage since using vacuum refill was implemented at IP3.

Technical

References:

Proposed References to be provided: None Learning Objective 13LP-ILO-RHR001 6 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KJA# 000026A201 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: - Location of a leak in the CCWS Importance 2.9 3.5 Question # 7 Given the following:

• Unit is at 100% power.

• CCW SURGE TANK #31/#32 LEVEL alarms are lit

• A Low level condition is present.

• CCW surge tanks level are DECREASING.

• The makeup valve is OPEN.

• There is NO indication of relief valve leakage in any CCW-cooled components Which ONE of the following is the location of the leak?

A. RCS sample coolers during sampling B. Non-Regenerative heat exchanger C. Thermal barrier heat exchanger D. Seal return heat exchanger Answer: D Explanation/J ustification:

All coolers are cooled by CCW. Candidate must know which fluid pressure is higher for each cooler. CCW Pressure is approximately 110 psig.

A. Incorrect. RCS> CCW B. Incorrect. Letdown upstream of PCV-135 is > CCW C. Incorrect. Thermal Barrier is RCS pressure> CCW D. Correct. Seal Return HX approximately the same as VCT pressure < CCW Technical

References:

3-AOP-CCW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPCCW G Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 K1A# 00WE12K103 Knowledge of the operational implications of the following concepts as they apply to the Uncontrolled Depressurization of all Steam Generators:

Annunciators and conditions indicating signals, and remedial actions associated with the Uncontrolled Depressurization of all Steam Generators Importance 3.4 3.7 Question # 8 The plant has sustained a Main Steam Line Break affecting all 4 Steam Generators.

The crew has just entered ECA 2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS.

The crew has throttled AFW to 100 gpm to each steam generator to minimize the RCS cooldown.

The following conditions exist:

SG Level Pressure SG 31 20%WR 360 psig STABLE SG32 19%WR 320 psig DECREASING SG33 18%WR 310 psig DECREASING SG 34 26%WR 380 psig INCREASING Which of the following describes the action that should be taken and the reason for the action?

A. Transition to E-2, Faulted Steam Generator Isolation, because there is an intact Steam Generator available.

B. Transition to FR-H.1, Loss of Secondary Heat Sink, because there is a RED condition on the Heat Sink Status Tree.

C. Transition to E-3, Steam Generator Tube Rupture, because there is an unexplained increase in Steam Generator Level.

D. Continue with ECA 2.1, Uncontrolled Depressurization of All Steam Generators, because Safety Injection termination is not complete.

Answer: A ExplanationlJustification:

A. Correct. The foldout page criteria for E-2 Transition is met.

B. Incorrect. Plausible because all SG levels are below "adequate heat sink" level of 9%; however, there is adequate AFW flow.

C. Incorrect. Plausible because 34 SG level is higher than the remaining SGs which could be caused by a SGTR.

D. Incorrect. Plausible because if SI termination is started in ECA-2.1, then transition to E-2 is not performed until it is complete. (Foldout Page criteria).

Technical

References:

3-ECA-2.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE20 8 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000029K206 Knowledge of the interrelations between the A TWS and the following: - Breakers, relays, and disconnects Importance 2.9 3.1 Question # 9 Reactor Trip and Bypass Breakers have been aligned to support testing of Reactor Protection Train A when an inadvertent Safety Injection signal is generated on Safety Injection Train A. Which of the following describes the Reactor Protection System response?

A. An ATWS occurs because Reactor Trip Breaker A and Reactor Trip Bypass Breaker B remain closed.

B. An ATWS occurs because Reactor Trip Breaker B and Reactor Trip Bypass Breaker A remain closed.

C. The Reactor trips. Reactor Trip Breaker A, Reactor Trip Breaker B, and Reactor Trip Bypass Breaker A are open. Reactor Trip Bypass Breaker B is locked open.

D. The Reactor trips. Reactor Trip Breaker A, Reactor Trip Breaker B, and Reactor Trip Bypass Breaker B are open. Reactor Trip Bypass Breaker A is locked open.

Answer: C Expla nation!Justification:

Meets KA 000029EK2.06 because the KA calls for knowledge of interrelations between trip breakers and ATWS. Since the question tests the knowledge of whether or not this breaker configuration can lead to an ATWS, the KA is met.

A. Incorrect but plausible because an operator may think that SI from each train only goes to one train of RPS B. Incorrect but plausible because an operator may think that SI from each train only goes to one train of RPS C. Correct D. Incorrect but plausible if an operator does not recall how trip and bypass breakers are paired.

Technical

References:

Syst Desc 28 Proposed References to be provided: None Learning Objective 12LP-ILO-ICROD 9 13LP-ILO-ICROD 2 Question Source: Bank Question History: IPEC Unit 2 2010 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 6 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000038K103 Knowledge of the operational implications of the following concepts as they apply to the SGTR: - Natural circulation Importance 3.9 4.2 Question # 10 Given:

• The Unit was operating at 100% power.

• A Steam Generator Tube Ruptured occurred in 34 SG.

• Off-site power was lost coincident with the reactor trip.

Which of the following describes how the E-3 Cooldown will be controlled?

A. Limited to 0.4 E6 Ibm/hr to prevent false Integrity Status Tree Entry while on natural circulation.

B. Maximum achievable rate using all intact SG atmospheric dump valves.

C. Limited to 25 °F/hr to ensure final Natural Circulation T-cold will remain above 320°F D. Maximum achievable rate using available condenser steam dumps.

Answer: B Explanation/Justification:

A. Incorrect. Plausible because 0.4 E6lbm/hr is the limit on cooldown rate using the condenser steam dumps. The condenser dumps are not available for this condition.

B. Correct. Procedure RNO states

C. Incorrect. Plausible because the value 320 0 is correct to prevent Integrity Status Tree entry.

D. Incorrect.. Plausible because cooldown using the condenser steam dumps.

is not available. The loss of off-site power resulted in loss of condenser (circulating water pumps).

Technical

References:

3-E-3 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE30 7 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 0000542120 Conduct of Operations - Ability to interpret and execute procedure steps.

Importance 4.6 4.6 Question # 11 Given the following plant conditions:

• Plant is operating at 81 % power

• No equipment is out of service

• 31 and 32 Condensate Pumps operating

• 31 Condensate Pumps trips due to a winding short Which of the below statements describes the required operator actions to mitigate the transient?

A. Verify Turbine Runback to < 745 MWe occurs to maintain MBFP suction pressure greater than 265 psig.

B. Verify 31 and 33 AFW Pumps automatically start; verify MBFP suction pressure greater than 265 psig.

C. Start 33 Condensate Pump, reduce turbine load as necessary to maintain MBFP suction pressure greater than 265 psig.

D. Trip the Reactor, Go To E-O, Reactor Trip or Safety Injection.

Answer: C Explanation/Justification:

A. Incorrect. Plausible this is correct for Unit 2. The condensate pump must be started to successfully recover from this transient.

B. Incorrect. Plausible because tripping of a main boiler feed pump would cause the AFW pumps to automatically start NOT tripping of a condensate pump.

C. Correct. It is necessary to start 33 condensate pump to recover from this event.

D. Incorrect. Plausible because tripping of a feed pump at -;:: 80% power requires tripping the reactor and going to E-O.

Technical

References:

3-AOP-FW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPFW1 3 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 1940012218 Equipment Control- Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Importance 2.6 3.9 Question # 12 Unit 3 is in Mode 5 on RHR with 31 and 33 CCW Pumps in service. Automatic safety injection was disabled shortly after entering Mode 5. All 3 EDGs are operable. Maintenance wants to perform an activity that will disable the auto start of 31 CCW Pump. Which of the following would you expect regarding this activity?

A. The activity will be allowed since auto starts of CCW pumps during a station blackout are already disabled when automatic safety injection is disabled.

B. The activity will be allowed since auto starts of 31 CCW pump would not occur during a station blackout since 31 CCW is already in service.

C. The activity will not be allowed. 31 CCW is required to be operable since RHR is in service. Loss of auto start capability will render 31 CCW pump inoperable.

D. The activity may be allowed after assessing for risk impact of having to manually restart 31 CCW during a station blackout.

Answer: 0 Expla nation/J ustification:

KA match because this tests the effect of a maintenance activity on the plant/operator response to a station blackout during shutdown operations.

A. Incorrect because normally the pump would auto start after the EDG ties onto the bus. Plausible since an operator could misunderstand the relationship between blackout and SI auto starts.

B. Incorrect but plausible if an operator misunderstands the blackout starting circuit.

C. Incorrect but plausible since and operator may believe this pump is required to be operable to support RHR operability.

D. Correct answer. Job would almost assuredly be allowed following evaluation.

The system response would be that Technical

References:

Syst Desc 4.1 Proposed References to be provided: None Learning Objective 13LP-ILO-CCW001 4a Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 4 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000057A103 Ability to operate and/or monitor the following as they apply to the Loss of Vital AC Instrument Bus:

- Feedwater pump speed to control pressure and level in S/G Importance 3.6 3.6 Question # 13 The plant is at normal 100% power lineup with Channel B selected for feed flow and steam flow control. Which set of actions will be required per 3-AOP-IB-1, Loss of Power to an Instrument Bus, for a loss of 31 and 31A Instrument Busses?

I - Swap feed flow and steam flow to the A channel II - Place MBFPs in Manual control III - Place MFRVs in Manual control A. I and II B. I and III C. II and III D. I, II, and III Answer: A Expla nation/Justification:

Plausibility/Explanation:

Selection III being needed is plausible because it is a required action for a loss of 34 Instrument Bus.

Selection II not being required is plausible because only one MBFP is affected by this failure and it goes into track and hold. However, the procedure specifies going to manual.

Selection I not being needed is plausible because the plant does not trip if this is not done. However, the procedure specifies this action.

A is the correct answer per 3-AOP-IB-1 Technical

References:

3-AOP-IB-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPIB1 E Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 0000622120 Conduct of Operations - Ability to interpret and execute procedure steps.

Importance 4.6 4.6 Question # 14 Given the following plant conditions:

• The Unit is at 20% power

• A loss of ALL normal Service Water Pumps has occurred due large amount of debris on the screens

• No Circ Water Pumps are available due to the debris Which of the following describes the required operator action for the above condition?

A. Trip the Reactor, DO NOT SHUT the MSIVs and initiate E-O, Reactor Trip or Safety injection.

B. Trip the Reactor, SHUT the MSIVs and initiate E-O, Reactor Trip or Safety injection.

C. Immediately align Backup Service Water Pumps, 37, 38 and 39 to the essential service water header.

D. Commence a rapid plant shutdown as long as temperatures remain below the trip setpoint.

Answer: B Explanation/Justification:

A. Incorrect. Plausible because tripping the reactor is correct; however, shutting the MSIVs is necessary to prevent admitting steam to the condenser with no circ water.

B. Correct. Shutting the MSIVs is necessary to prevent admitting steam to the condenser with no circ water.

C. Incorrect. Plausible because backup service water takes a suction from a different location (discharge canal); however, they only supply nuclear essential loads NOT all essential loads.

D. Incorrect. Plausible because this action is directed if Service water pressure cannot be maintained greater than 50 spig.

Technical

References:

2-AOP-SW-1 3-AOP-SW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-SW001 6 Question Source: Bank Question History: IPEC Unit 3 NRC 2006 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000077AK302 Knowledge of the reasons for the following responses as they apply to Generator Voltage and Electrical Grid Disturbances:

Actions contained in abnormal operating procedure for voltage and grid disturbances Importance 3.6 3.9 Question # 15 Unit 3 is in a normal 100% power lineup when the "480 VOLT SAFEGUARD BUS. UNDERVOLTAGE" alarm annunciates and immediately clears. The cause is a grid disturbance that affected 138 KV Feeder 95331. Which of the following describes the direction specified in ARP-005, Panel SBF SAFEGUARDS?

A. Per AOP-138KV-1, Loss of Power to 6.9 KV Bus 5 and/or 6, check status of SFP pumps because historically the standby pump has started due to grid disturbances.

B. Per AOP-138KV-1, Loss of Power to 6.9 KV Bus 5 and/or 6, check status of SFP pumps because historically the running pump has tripped due to grid disturbances.

C. Per a checklist in SOP-EL-5, Operation of On-Site Power Sources, check status of SFP pumps because historically the standby pump has started due to grid disturbances.

D. Per a checklist in SOP-EL-5, Operation of On-Site Power Sources, check status of SFP pumps because historically the running pump has tripped due to grid disturbances.

Answer: D Explanation/J ustification:

A. Incorrect. Plausible because historically, the Spent fuel pool pumps are affected (tripped not started) on this condition. AOP-138kV-1 is not the correct procedure.

B. Incorrect. Plausible because historically, the Spent fuel pool pumps are affected (tripped) on this condition. AOP-138kV-1 is not the correct procedure.

C. Incorrect. Plausible because historically, the Spent fuel pool pumps are affected (tripped not started) on this condition. The ARP is the correct procedure.

D. Correct.

Technical

References:

3-AOP-138KV-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOP138 1 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 00WE04A202 Ability to determine and interpret the following as they apply to the LOCA Outside Containment:

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Importance 3.6 4.2 Question # 16 The crew has completed the actions of ECA-1.2, LOCA Outside Containment.

• Pressurizer level is 35% and stable.

• Steam Generator narrow range levels are 30%.

• Subcooling is 35°F.

• RCS pressure is 1400 psig and slowly lowering.

Which ONE (1) of the following procedures will be performed next?

A. ES-1.1, SI Termination.

B. E-O, Reactor Trip or Safety Injection.

C. ECA-1.1, Loss of Emergency Coolant Recirculation.

D. E-1, Loss of Reactor or Secondary Coolant.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because SI Termination criteria are satisfied with the exception of RCS pressure; however no direct transition to ES-1.1 from ECA-1.2

B. Incorrect. Plausibe because ECA-1.2 can be entered from E-O (step 22) and candidate may believe that the transition is back to procedure and step in effect; however, there is no transition to E-O.

C. Correct, because pressure is lowering, transition is to ECA-1.1 to address potential for inadequate inventory inside VC for recirculation E-1. All SI termination are satisfied, so from E-1, the transition to ES-1.1 will be made.

D. Incorrect. Plausible a transition to E-1 is correct if the leak is isolated and pressure is rising.

A. Incorrect. Plausible because SI Termination criteria are satisfied with the exception of RCS pressure; however no direct transition to ES-1.1 from ECA-1.2 B. Incorrect. Plausibe because ECA-1.2 can be entered from E-O (step 22) and candidate may believe that the transition is back to procedure and step in effect; however, there is no transition to E-O.

C. Correct, because pressure is rising, transition is to E-1. All SI termination are satisfied, so from E-1, the transition to ES-1.1 will be made.

D. Incorrect. Plausible a transition to ECA-1.1 if lowering is made if pressure is lowering.

Technical

References:

2-ECA-1.2 3-ECA-1.2 Proposed References to be provided: None Learning Objective 12LP-ILO-EOPC125 Question Source: Modified Question History: IPEC Unit 22012 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KfA# OOWE05K202 Knowledge of the interrelations between the Loss of Secondary Heat Sink and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Importance 3.9 4.2 Question # 17 The plant is in an emergency condition, and the control room operators are responding to a loss of secondary heat sink. RCS bleed-and-feed is in progress, and RCS temperatures are stable. The operators restore a feedwater source and prepare to feed the S/Gs, which are hot and dry. The SM directs them to re establish FW flow to only one S/G rather than to all of the S/Gs.

The primary reason for feeding only one S/G under these conditions is to:

A. Prevent a rapid cooldown of the RCS that could lead to a pressurized thermal shock condition.

B. Ensure that if S/G failure occurs due to excessive stresses, the failure is isolated to one S/G.

C. Ensure pump run out conditions will not occur for the FW source until adequate heat sink established D. Prevent excessive cooldown which could cause a void in the SG U tubes and loss of natural circulation.

Answer: B

Explanation/Justification:

A. Incorrect. Plausible because feeding all SGs could result in a PTS condition; however, the thermal stresses could result in failures in multiple SGs.

B. Correct.

C. Incorrect. Plausible because depending on the feedwater source, pump runout could occur if all SGs are feed at the same time.

D. Incorrect. Plausible because excessive cooldown can cause a loss of natural circulation flow.

Technical

References:

3-FR-H.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRH 7 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# OOWE11A103 Ability to operate and/or monitor the following as they apply to the Loss of Emergency Coolant Recirculation: - Desired operating results during abnormal and emergency situations Importance 3.7 4.2 Question # 18 Given:

• A LOCA has occurred.

• Operators were performing ECA-1.1, Loss of Emergency Coolant Recirculation when Containment pressure is noted to be 22.3 psig.

• The decision of whether to remain in ECA-1.1 or transition to another procedure was properly made.

Which of the following describes how the Containment Spray system will be operated, and why?

The Containment Spray System is operated as directed in ...

A. ECA-1.1 because it establishes minimum required containment spray flow and conserves RWST inventory.

B. FR-Z.1, Response to High Containment Pressure, since restoration of the critical safety function takes precedence.

C.* ECA-1.1 since FRPs (Functional Restoration Procedures) are NOT implemented during the performance of ECA-1.1.

D. FR-Z.1 because high containment pressure over time will result in off site doses in excess of analyzed values.

Answer: A Explanation/Justification:

A. Correct B. Incorrect. Plausible because FR procedure typically have a higher priority than other emergency procedures. There are two procedures (ES-1.3 Transfer to Cold Leg Recirculation and ECA-O.O Loss of All AC Power) that take priority over FRPs. The concept that a procedure may take prioity over the FRPs is not unrealistic. ECA-1.1 addresses a "best guess" approach to containment conditions where FR-Z.1 addresses a "worst case" approach. The need to conserver RWST inventory for core cooling is a higher priority C. Incorrect. Plausible becauseECA-1.1 is a special case and candidates may believe that no FRPs are implemented. FRP are implemented in ECA-1.1 if the condition occurs with the exception of FR-Z.1 due to the need to conserve inventory for core cooling.

D. Incorrect. Plausible because containment is the final fission product barrier.

In general actions in the SAMGs are focused on maintaining containment intact.

Technical

References:

3-ECA-1.1 3-FR-Z.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRZ 7 Question Source: Bank Question History: IPEC Unit 2 2010 Question Cognitive Level: Fundamental Knowleqge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KJA# 000033K302 Knowledge of the reasons for the following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

Guidance contained in EOP for loss of intermediate-range instrumentation Importance 3.6 3.9 Question # 19 Given the following:

• The plant was initially at 95% power.

• The reactor tripped 20 minutes ago.

• INTERMEDIATE RANGE NO.1 LOSS OF COMPENSATE VOLTAGE alarm is annunciated.

Which ONE (1) of the following describes the response of Intermediate Range N 35 to the loss of compensating voltage and what actions are given for this condition?

A. Indicates LOW; Press appropriate Intermediate Range Permissive Defeat Button to re-energize Source Range Channel N31 B. Indicates HIGH; Press both Source Range AlB Logic Trip Block Buttons to re-energize both Source Range Channels C. Indicates LOW; Press Source Range A Logic Trip Block Button to re energize Source Range Channel N31 D. Indicates HIGH; Press both Intermediate Range Permissive Defeat Buttons to re-energize both Source Range Channels

Answer: 0 ExplanationlJustification:

Loss of compensating voltage for the intermediate range will cause the channel to indicate high. Both Intermediate Range Channels must indicate less than 3.7 E-10 amps to automatically reenergize the Source range Nls. Both push buttons must be depressed simultaneously to re-energize the source ranges manually.

A. Incorrect. Plausible because candidate must remember which direction indicated power will fail with a failure of compensating voltage. The correct button description is used; however, both buttons must be pressed simultaneously.

B. Incorrect. Plausible because indicated power level is correct; however the wrong pushbuttons are listed to re-energize the Source Range Nls.

C. Incorrect. Plausible because candidate must remember which direction indicated power will fail with a failure of compensating voltage. the wrong pushbuttons are listed to re-energize the Source Range Nls.

D. Correct.

Technical

References:

3-ES-O.1 Syst Desc 13 Proposed References to be provided: None Learning Objective 13LP-ILO-ICEXC 8 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 000036A104 Ability to operate and/or monitor the following as they apply to the Fuel Handling Incidents: - Fuel handling equipment during an incident Importance 3.1 3.7 Question # 20 Given the following conditions:

• Plant is in refueling performing a core reload

• Transfer tube gate valve is opened

• An irradiated assembly has just removed from the Containment Upender and is in the mast

• The following occurs: Reactor pit sump pump indicates ON

• High level Reactor Pit alarm annunciates

• Containment and Recirculation sump levels are rising slowly

• R-2 and R-7 are in alarm and indications are rising Which one of the following actions would be directed by AOP-FH-1 "Fuel Damage OR Loss of SFP/Refueling Cavity Level"?

Place the Fuel Assembly A. in the upender and lower to horizontal position B. directly west of the RCC Change Fixture and unlatch the assembly from the gripper C. leave assembly in the mast, lower to the bottom of the cavity, and maintain gripper engaged D. in the RCC Change Fixture Answer: A Explanation/Justification:

Duplicated from question no 2082 made changes based on validator comments before U3 2013 NRC Exam A. Correct answer per AOP-FH-1 B. Incorrect but plausible since this would be the location picked if the upender is not available. Also incorrect because we would not unlatch the assembly.

C. Incorrect because the procedure does not have an assembly sent to the FSB. Plausible because this would be a safe location.

D. Incorrect because the procedure does not specify this action. Plausible because the procedure has a similar action for placing assembly near RCC change fixture.

Technical

References:

3-AOP-FH-1 Proposed References to be provided: None Learning Objective 13LP-ILO-FHD001 m Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 0000682434 Emergency Procedures/Plan Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Importance 4.2 4.1 Question # 21 Which of the following describes the Reactor Operator's task for safe shutdown outside the control room and reason for this task?

A. Establish charging within 30 minutes of the event initiation to ensure pressurizer level remains in the indicating range.

B. Establish charging with 60 minutes of the event initiation to ensure adequate shutdown margin is maintained.

C. Establish AFW flow within 30 minutes of the event initiation to ensure steam generators do not dry out.

D. Establish AFW flow within 60 minutes of the event initiation to ensure RCS pressure does not exceed PORV setpoints.

Answer: C Explanation/Justification:

A. Incorrect but plausible since charging is established early in the event for this reason, but not by the RO.

B. Incorrect but plausible since this is the correct time for establishing charging.

C. Correct answer per AOP-SSD-1 and OAP-115.

D. Incorrect but plausible since the RO does establish AFW flow. The time and reason are incorrect, but plausible Technical

References:

3-AOP-SSD-1 OAP-115 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPSSD F Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 000076K201 Knowledge of the interrelations between the High Reactor Coolant Activity and the following:

- Process radiation monitors Importance 2.6 3 Question # 22 The operating team has entered AOP-HIACT-1, High RCS Activity, due to elevated RCS activity as noted by Chemistry. Which of the following is true regarding R-63 (Gross Failed Fuel Detector) readings per AOP-HIACT-1?

A. If R-63 is in alarm, verify that RCS sample isolation valves has automatically closed.

B. If R-63 is in alarm, verify that letdown has automatically diverted through the standby mixed bed.

C. If activity on R-63 is > 50 \-ICi/cc, then no samples of the RCS are to be taken.

D. If activity on R-63 is > 50 \-ICi/cc, then consideration should be given to using the Post-Accident Sampling System.

Answer: D Explanation/Justification:

A. Incorrect but plausible because there are radiation monitors that automatically isolate valves.

B. Incorrect but plausible because this action is specified by the procedure.

It's just not an automatic action.

C. Incorrect but plausible because 50 \-ICi/cc is a procedure decision level.

D. Correct answer per the note before step 4.3 of AOP-HIACT-1.

Technical

References:

3-AOP-HIACT-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPACT 3 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# OOWE02A201 Ability to determine and interpret the following as they apply to the SI Termination: - Facility conditions and selection of appropriate procedures during abnormal and emergency operations Importance 3.3 4.2 Question # 23 Given:

• The team is performing the actions of ES-1.1, "SI Termination"

• At step 4 of ES-1.1, The team secured all SI and RHR Pumps

• At step 8 of ES-1.1, the team secured all but one charging pump

• RCS Pressure and Pressurizer Level begin to lower after the charging pumps are secured Which one of the following describes the required operator actions in accordance with ES-1.1?

A. Monitor RCS Pressure and Pressurizer Level, if pressure and level continue to lower, restart SI pumps and return to procedure previously in effect.

B. Manually re-initiate SI and Transition to E-O, "Reactor Trip or Safety Injection" step 1.

C. Adjust charging flow and if Pressurizer Level cannot be maintained, then restart SI pumps as required and transition to E-1, "Loss of Reactor or Secondary Coolant".

D. Restart one charging pump and if Pressurizer Level cannot be maintained transition to E-1, "Loss of Reactor or Secondary Coolant."

Answer: C Expla nationlJustification:

Duplicated 'from question no 16400 A. Incorrect. Plausible because candidate may believe that returning to the procedure previously in effect will resume mitigation strategies for that event.

B. Incorrect. Plausible because candidate may believe that a new event has occurred and that restarting in E-O is appropriate.

C. Correct. These are the actions of step 9 in ES-1.1 D. Incorrect. Plausible because the transition to E-1 is correct; however, the SI pumps must be started first.

Technical

References:

2-E-1 2-ES-1.1 3-E-1 3-ES-1.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE10 21 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 000028K202 Knowledge of the interrelations between the Pressurizer Level Control Malfunctions and the following: - Sensors and detectors Importance 2.6 2.7 Question # 24 Given the following Plant conditions;

• The plant is operating at 100% power, steady state.

• The Pressurizer Level Defeat switch is in the Defeat 3 position The operator notes the following indications 3 minutes after a failure occurred:

• All Pressurizer heaters are on.

• Letdown isolation valve LCV-459 is open.

• Charging pump speed slowly lowering.

Which ONE of the following describes a cause for these indications?

A. Pressurizer level channel 1 detector, LT-459, failed high.

B. Pressurizer level channel 2 detector, LT-460, failed high.

C. Pressurizer level channel 1 detector, LT-459, failed low.

D. Pressurizer level channel 2 detector, LT-460, failed low.

Answer: B Explanation/Justification:

Duplicated from question no 18947 Plausibility:

A Channel failing high will cause these indications, but a channel failing low can also cause some of the same indications as well. By not stating the condition of LCV-460 in the stem, a channel failing low remains plausible although incorrect.

Confusing which level channel is paired with which LCV is also a very plausible mistake.

The correct answer is B. The controlling channel is 2/460 when channel 3 is defeated. If the controlling channel fails high, all heaters energize and charging speed decreases, and LCVs remain open.

Technical

References:

3-AOP-INST-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPINT 4 13LP-ILO-AOPINT 5 13LP-ILO-AOPINT 6 Question Source: Bank Question History: Not on NRC exam Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 00WE06K102 Knowledge of the operational implications of the following concepts as they apply to the Degraded Core Cooling:

Normal, abnormal and emergency operating procedures associated with Degraded Core Cooling Importance 3.5 4.1 Question # 25 An INADEQUATE CORE COOLING condition is present during a small break loss of coolant accident. The following conditions are noted:

• No safety injection pumps are available

• Both RHR pumps are running

• No RCPs are running

• Both motor driven AFW pumps are operating

• RCS pressure is 1100 psig

• CETs are 765°F and rising slowly

• RVLlS is 31% and lowering slowly

• Steam generator pressure is 1050 psig Which of the following is true regarding recovery actions?

A. A secondary depressurization will cause enough heat rejection to the steam generators that safety injection system flow will no longer be required to combat the event.

B. A secondary depressurization will lower reactor coolant system pressure, allowing safety injection accumulators to recover the core.

C. Opening power-operated relief valves (PORVs) will take the reactor coolant system to saturation, allowing latent heat removal to bring down core temperature.

D. Opening power-operated relief valves (PORVs) will depressurize the reactor coolant system low enough to allow the RHR system to cool the core.

Answer: B Explanation!Justification:

A Incorrect. Plausible because a secondary depressurization is performed; however, the explanation is not correct.

B. Correct.

C. Incorrect. Plausible because opening PORVs is an action in FR-C.1; however, the explanation is incorrect.

C. Incorrect. Plausible because opening PORVs is an action in FR-C.1; however, the explanation is incorrect.

Technical

References:

2-FR-C.1 2-FR-C.1 BG 3-FR-C.1 Proposed References to be provided: None Learning Objective 13SG-LOR-EOP0081 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 00WE14K103 Knowledge of the operational implications of the following concepts as they apply to the High Containment Pressure:

Annunciators and conditions indicating signals, and remedial actions associated with the High Containment Pressure Importance 3.3 3.6 Question # 26 Given the following conditions:

• The reactor is tripped

• A Safety Injection signal has been generated

• All RCPs are running

• Average CETs indicate 515°F

• RCS Pressure is 1800 psig and stable

• CNMT Pressure is 24 psig and slowly increasing

• Pressurizer level is off scale low

• All ESF/ECCS equipment operated as designed Which ONE of the following describes why the RCPs should or should not be tripped?

A. RCP trip foldout page criteria are met The RCPs should be tripped Prevent exceeding peak clad temperature limits B. RCP trip foldout page criteria are met The Reps should continue to run Provide forced cooling.

C. RCP trip foldout page criteria are NOT met The RCPs should be tripped Prevent damage to pumps D. RCP trip foldout page criteria are NOT met

The RCPs should remain running Ensure adequate core cooling Answer: C Explanation/Justification:

A. Incorrect. Plausible because candidate must calculate subcooling and consider adverse containment conditions to recognize that trip criteria is NOT met and the RCPs should be tripped due to Phase B isolation.

B. Incorrect. Plausible because candidate must calculate subcooling and consider adverse containment conditions to recognize that trip criteria is NOT met. The candidate may recognize that a steam break accident is indicated and keeping RCPs running during excessive heat removal is desired.

C. Correct. Containment pressure is high resulting in a Phase B isolation. With no CCW cooling to RCPs the pumps must be tripped.

D. Incorrect. Plausible because foldout page trip criteria are not met; however the pumps should be tripped.

Technical

References:

3-E-0 Proposed References to be provided: None Learning Objective 12LP-ILO-AOPRCP 10 13LP-ILO-AOPRCP 10 Question Source: Modified Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 K1A# OOWE10K301 Knowledge of the reasons for the following responses as they apply to the Natural Circulation with Steam Void in Vessel with/without RVLlS: - Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics Importance 3.4 3.7 Question # 27 Following a plant trip, RCS cold leg temperatures are 540°F with no RCPs in service. A natural circulation cooldown is to be performed without RVLlS being available. Which of the following describes the requirement for boration for this cooldown?

A Once boration is started, cooldown may begin.

B. Cold Shutdown boron concentration must be achieved before beginning cooldown.

C. Cooldown to 400°F may begin once Hot Shutdown boron concentration is achieved.

D. Since RVLlS is not available, Cold Shutdown boron concentration times 1.28 must be achieved prior to starting cooldown.

Answer: B Explanation/Justification:

Per ES-0.2, boration to cold shutdown concentration is completed prior to starting a natural circulation cooldown.

A. Incorrect but plausible since it is reasonable that once boration is started, cooldown could begin.

B. Correct answer.

C. Incorrect. Plausible, since there is a hot shutdown boron concentration requirement in normal plant operating procedures. It is reasonable that the plant could be cooled down part of the way to CSD with this amount of boration.

D. Incorrect. Plausible since this is required for cooldown with SG backfill. RVLlS not being available could cause increased concerns for RCS mixing since void growth is harder to detect.

Technical

References:

3-ES-0.2 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPEOO 12 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 003000A205 Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Effects of VCT pressure on RCP seal leakoff flows Importance 2.5 2.8 Question # 28 GIVEN:

• The unit experienced a reactor trip coincident with a loss of off site power.

• The crew entered 3-ES-0.2, Natural Circulation Cooldown

• Lift oil pump is running

• RCS pressure is 350 psig

• Seal Injection flow is 8 gpm to each RCP

• Seal Leakoff flow is 0.19 gpm from each RCP

• Number 1 seal leakoff temperature is 100°F

• All bearing temperatures are less than 150°F

• VCT pressure is 50 psig

• Both Spray valves are in AUTO and closed

• 138kV power is available from Buchanan Switchyard

• The crew is making preparations to start RCP 34 Which of the following actions should the CRS direct the operator to perform?

A. Direct the operator to start RCP 34, all conditions are satisified.

B. Direct the operator to start RCP 34, all conditions are not satisified but are not required.

C. Direct the operator to decrease VCT pressure to increase sealleakoff flow, then RCP 34 may be started.

D. Direct the operator to open RCP No.1 Seal Bypass Valve, 246 to increase sealleakoff flow, then RCP 34 may be started.

Answer: C ExplanationlJustification:

A. Incorrect. Plausible because lower seal return flow conditions are expected and allowed at low RCS pressure.

B. Incorrect. Plausible because under certain conditions, the RCPs are started without proper support conditions.

C. Correct D. Incorrect. Plausible because conditions to open seal bypass valve 246 exist with the exception of high temperatures.

Technical

References:

3-S0P-RCS-001 Proposed References to be provided: None Learning Objective 13LP-ILO-RCSRCP B Question Source: Bank Question History: DC Cook 2007 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 004000K629 Knowledge of the effect of a loss or malfunction on the following CVCS components: - Reason for excess letdown and its relationship to CCWS Importance 2.7 3.1 Question # 29 Unit 3 is at 100% power when a CCW leak is found on AC-TCV-130, Non Regenerative Heat Exchanger CCW Outlet Temperature Control Valve. Repairs are expected to only take a few hours, but will require isolating CCW to AC-TCV 130.

Which of the following is correct for these conditions?

A. The plant will have to proceed to Mode 5 to make repairs since isolation of this leak will cause a loss of cooling to letdown and RCP seal return.

B. The plant will have to proceed to placing RHR in service before isolation can occur because isolation will require removing all of CCW Loop 2 from service.

C. Repairs can occur on line, but letdown and all charging will be isolated.

OM permission will be required to operate with thermal barrier cooling as the only means of cooling RCP seals.

D. Repairs can occur on line. Excess letdown will be placed in service with normal letdown secured to allow isolation of AC-TCV-130.

Answer: D Explanation!Justification:

A. Incorrect but plausible since the NRHX and Seal Return HX are physically close together and both cooled by CCW.

B. Incorrect but plausible since Loop 2 of CCW cools RCPs operators know that RCP have to be secured to isolate this loop.

C. Incorrect but plausible if an operator does not completely understand the function and allowable times to use excess letdown.

D. Correct answer. Valves AC-81 0 and AC-814 are available for isolation of CCW to AC-TCV-130 without disabling other portions of CCW. 3-S0P-CVCS-2 provides guidance for removing normal letdown from service and placing excess letdown in service. This is clearly the action the station would take for this issue.

Technical

References:

3-S0P-CVCS-002 Syst Desc 4.1 Proposed References to be provided: None Learning Objective 13LP-ILO-CVC001 8 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 4 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 005000K101 Knowledge of the physical connections and/or cause-effect relationships between the RHRS and the following systems:

CCWS Importance 3.2 3.4 Question # 30 Given:

• RHR was placed in service one hour ago

• 34 RCP is in service

• RCS Pressure is 375 psig

• 31 RHR Heat Exchanger is in service

• A tube in 31 RHR Heat Exchanger fails Which of the following indications would the operator expect to see?

A. CCW surge tank level would decrease and RCS inventory would increase B. CCW surge tank level would decrease and RHR boron concentration would decrease C. CCW surge tank level would increase and CCW radiation levels would increase D. CCW surge tank level would increase and 822A, RHR HX CCW Shutoff Valve would auto close.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because the indications would be correct if CCW pressure was higher than RHR pressure.

B. Incorrect. Plausible because the indications would be correct if CCW pressure was higher than RHR pressure.

C. Correct.

D. Incorrect. Plausible because tank level will increase, valve 625, Thermal Barrier Heat Exchanger Return will auto close if a leak occurs. 822A does not auto close.

Technical

References:

3-AOP-UCCW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-CCW001 1 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KfA# 005000K203 Knowledge of bus power supplies to the following: - RCS pressure boundary motor-operated valves Importance 2.7 2.8 Question # 31 Unit 3 is on RHR with 34 RCP running and AC-MOV-730 and AC-MOV-731 energized. A loss of offsite power occurs and 31 EDG fails to start. 32 and 33 EDGs start and automatically power their 480V busses. During the loss of offsite power, RCS wide range pressure increased to 600 psi on both channels and remains at this pressure. Without operator action, what is the status of 730 and 731 after the 480V busses re-energize?

A. 730 and 731 are both closed or going closed.

B. 730 and 731 are both open and not going closed.

C. 730 is open and not going closed, 731 is closed or going closed.

D. 731 is open and not going closed, 730 is closed or going closed.

Answer: A Explanation/Justification:

A. Correct answer. 730 and 731 are powered from busses that are supplied by 32 and 33 EDGs. They will regain power without operator action. Since RCS pressure is above 550 psi, both valves will automatically close.

B. Incorrect but plausible. There are a number of possible system misunderstandings that could lead an operator to believe that both valves will remain open. These include: 1) Not knowing the auto-closure setpoint. 2)

Believing that the valves only have an open permissive interlock like IP2. 3) Not knowing that the MCCs that supply these valves will re-energize without operator action.

C. Incorrect but plausible if an operator does not know the power supply to 730.

D. Incorrect but plausible if an operator does not know the power supply to 731.

Technical

References:

Syst Desc 4.2 Proposed References to be provided: None Learning Objective 13LP-ILO-RHR001 4/5 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 006000K605 Knowledge of the effect of a loss or malfunction of the following will have on the ECCS: - HPI/LPI cooling water Importance 3 3.5 Question # 32 Given:

• The plant experienced a LOCA coincident with loss of off-site power

• Safety Injection is actuated

• RCS pressure remains above the shutoff head of the RHR pumps

• The conditions necessary to secure the RHR pumps are not satisfied Which of the following describes the RHR pump operations for this condition?

A. Auxiliary Component Cooling Water Pumps will circulate CCW through RHR pump seal and thermal barrier heat exchangers. Pumps can continue to operate indefinitely.

B. With no Component Cooling Water to the RHR pumps, the pumps can operate for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> together followed by a single pump for an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Shaft driven pumps will circulate CCW through RHR pump seal and thermal barrier heat exchangers. Pumps can continue to operate indefinitely.

D. With no Component Cooling Water to the RHR pumps, the pumps can operate for 45 minutes together followed by a single pump for an additional 45 minutes.

Answer: D

Exp lanation/Justification:

The following CAUTION exists in 3-S0P-RHR-1:

If one or both RHR Pumps are in normal operation with suction fluid temperature exceeding 150 degrees F and CCW flow to the Pump(s) is lost, promptly restore cooling flow and/or perform the applicable steps of 3-AOP-CCW-1. If the RHR Pumps are in "miniflow recirculation" operation and CCW flow to the Pump(s) is lost, then two Pumps can run for a maximum of 45 minutes followed by single Pump operation for a maximum of an additional 45 minutes.

A. Incorrect. Plausible because Auxiliary Component Cooling Water Pumps circulate CCW through the internal Recirc Pumps.

B. Incorrect. Plausible because this answer is similar to the correct answer with a different time.

C. Incorrect. Plausible because Safety Injection Pumps have shaft driven pumps that circulate CCW and they can operate indefinitely.

D. Correct.

Technical

References:

3-S0P-ESP-001 3-S0P-RHR-001 Proposed References to be provided: None Learning Objective 13LP-ILO-RHR001 2 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 3 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 006000K508 Knowledge of the operational implications of the following concepts as they apply to the ECCS: - Operation of pumps in parallel Importance 2.9 3.1 Question # 33 Unit 3 experienced a LOCA and has transferred to cold leg recirculation when it is determined that inadequate low head flow exists due to high RCS pressure.

The team transitions to the attachment to establish HHSI recirculation. Which of the following actions will be taken per this attachment?

A. Both Recirc Pumps will be operated to ensure adequate suction flow to the HHSI Pumps while maintaining recirc spray flow.

B. Only one Recirc Pump will be operated to ensure that no Recirc Pump is operating with its discharge check valve closed by flow from the other pump.

C. All HHSI Pumps will be operated to ensure adequate core recirculation flow is established.

D. Only One HHSI Pump (31 or 33) will be operated.

Answer: B Explanation/Justification:

A. Incorrect but plausible since spray flow could be required in addition to recirc flow.

B. Correct answer per ES-1.3 Att. 4 step 1. Although the reason for the action is not listed, this was a change to the procedure about 5 years ago to address strong pump - weak pump concerns.

C. Incorrect but plausible since all three HHSI pumps would actually provide less recirc flow than would be delivered by one recirc pump during a larger size LOCA. Incorrect since the procedure will only run 2 pumps.

D. Incorrect but plausible since 32 has some special actions for it in the recirc procedures. E.g. manual valve 898 is opened. 32 is actually only used if 31 or 33 are not available. The special actions for 32 SI pump make this a plausible misconception.

Technical

References:

3-ES-1.3 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE10 19 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 8 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 0070002132 Conduct of Operations - Ability to explain and apply all system limits and precautions.

Importance 3.8 4 Question # 34 Which of the following describes the original basis for the volume of the PRT and water level high and low level limits?

A. Cool and condense the steam from PORV actuation due to the insurge following a complete loss of load without exceeding rupture disk setpoint.

B. Cool and condense the steam from PORV actuation due to the insurge following a loss of feed ATWS without exceeding rupture disk setpoint.

C. Cool and condense steam volume from a complete depressurization of RCS from normal operating pressure to RHR system pressure without exceeding rupture disk setpoint.

D. Cool and condense steam volume from depressurization of RCS during SGTR to SI Termination without exceeding rupture disk setpoint.

Answer: A Explanation/Justification:

A. Correct. This is the assumption based on the maximum possible insurge into the pressurizer.

B. Incorrect. Plausible because because PORVs will lift on loss offeed ATWS; rupture disk mayor may not rupture.

C. Incorrect. Plausible because a complete loss of offsite power may require depressurization using PORVs only to RHR conditions.

D. Incorrect. Plausible because SG tube rupture is an accident that requires actions to terminate SI to satisfy off site dose projections.

Technical

References:

Syst Desc 1.4 Proposed References to be provided: None Learning Objective 13LP-ILO-RCSPZR 2 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 K1A# 008000K101 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: - SWS Importance 3.1 3.1 Question # 35 Given the following plant conditions:

• Reactor Coolant System temperature is 185°F.

• Reactor Coolant System pressure is 100 psig.

• RHR cooldown operations have been established with both RHR pumps and heat exchangers in service.

• The Shift Manager has directed the crew to establish a cooldown.

Which of the following describes how this is accomplished?

A. Opening AC-820 AlB is not limited. These valves provide coarse temperature control.

Opening SWN-35-1/2 is limited to 27.5° and 27° respectively. These valves provide fine temperature control.

B. Opening AC-820 AlB is limited to 27.5° and 27° respectively. These valves provide coarse temperature control.

Opening SWN-35-1/2 is not limited. These valves provide fine temperature control.

C. Opening AC-820 AlB is not limited. These valves provide fine temperature control.

Opening SWN-35-1/2 is not limited. These valves provide coarse temperature control.

D. Opening AC-820 AlB is limited to 27.5° and 2r respectively. These valves provide coarse temperature control.

Opening SWN-35-1/2 is not limited. These valves provide fine temperature control.

Answer: A Explanation/Justification:

Duplicated from question no 18653 Position of 820AlB is not limited; however if moved from normal 75° position the system engineer must be notified. SWN 35-1 and 35-1 are not limited when temperature is between 350 and 200 0 , but they are limited when temperature is <

200.

A. Correct.

B. Incorrect.

C. Incorrect.

D. Incorrect.

Technical

References:

2-S0P-4.2.1 3-S0P-RHR-001 Proposed References to be provided: None Learning Objective 13LP-ILO-POP002 4 Question Source: Modified Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 010000K403 Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the following:

Over pressure control Importance 3.8 4.1 Question # 36 The controlling pressurizer pressure channel PT-455 has failed low. Which of the following describes plant response to heaters automatically energizing? (Assume No Operator Action)

A. Pressure increase will be limited by pressurizer sprays.

B. Plant pressure will be controlled by ONE PORV cycling.

C. Plant pressure will be controlled by BOTH PORVs cycling.

D. Pressure increases until the plant trips on high pressure.

Answer: B Explanation/Justification:

The controlling pressurizer pressure channel PT-455 has failed low. Which of the following describes plant response? (Assume No Operator Action)

A. All heaters energize. Pressure increase will be limited by pressurizer sprays B. All heaters energize. Plant pressure will be controlled by one PORV cycling C. Backup heaters energize. Pressure increases until the plant trips on high pressure

D. Backup heaters energize. Pressure will increase until a PORV lifts, and the plant will subsequently trip on low pressure.

Technical

References:

Syst Desc 1.4 Proposed References to be provided: None Learning Objective 13LP-ILO-ICPZPC 2 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 010000A107 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including:

RCS pressure Importance 3.7 3.7 Question # 37 A rapid load reduction has just been completed. You observe the following:

• pressurizer sprays partially open

• modulating heaters full off

• backup heaters on Which of the following would cause these indications?

A. Controlling pressurizer pressure channel failed high B. Pressurizer level insurge from the downpower C. Controlling pressurizer level channel failed low D. Pressurizer program level decrease from the downpower Answer: B ExplanationlJustification:

A. Incorrect but plausible. This instrument failure would lead to all of these conditions except sprays would be full open.

B. Correct answer per the system description. This satisfies the KA because this controller feature is ensure RCS pressure stays within design limits during an expected power maneuver.

C. Incorrect but plausible. Plausible since these conditions would occur if the level channel failed high.

D. Incorrect but plausible. Plausible since the only indication that should not be present is that sprays would not be open.

Technical

References:

Syst Desc 1.4 Proposed References to be provided: None Learning Objective 12LP-ILO-RCSPZR 4 13LP-ILO-RCSPZR 3 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 012000A307 Ability to monitor automatic operation of the RPS, including:

Trip breakers Importance 4 4 Question # 38 Given the following plant conditions:

• Unit 3 is at 80% power with all systems aligned normal.

• PT-412A, Turbine First Stage Pressure, fails high.

For the above conditions, the control rods will be continuously (1) . If the rods continue to move after controls are shifted to MANUAL the operator will initiate a reactor trip, and then confirm that the reactor is tripped by observing the (1 ) (2)

A. Inserting Reactor Trip Breakers GREEN lights LIT, and the Bypass Breaker indicating lights DARK.

B. Inserting GREEN lights LIT on both the Reactor Trip Breakers and Bypass Breakers.

C. Withdrawing Reactor Trip Breakers GREEN lights LIT, and the Bypass Breaker indicating lights DARK.

D. Withdrawing GREEN lights LIT on both the Reactor Trip Breakers and Bypass Breakers.

Answer: D

ExplanationlJustification:

A. Incorrect. Plausible because actual steam header pressure decreases as power increases, but first stage pressure increases as power increases. The failure will cause rods to withdraw. Additionally, the bypass breakers are racked out which may cause candidate to believe the lights are not lit.

B. Incorrect. Plausible because actual steam header pressure decreases as power increases, but first stage pressure increases as power increases. The failure will cause rods to withdraw. Additionally, the bypass breaker indications are correct.

C. Incorrect. Plausible because actual steam header pressure decreases as power increases, but first stage pressure increases as power increases. Rod withdrawal is correct for this condition. Additionally, the bypass breakers are racked out which may cause candidate to believe the lights are not lit.

D. Correct. The rods will withdraw for this condition and all breaker lights will indicate green.

Technical

References:

Syst Desc 16.1 Syst Desc 28 Proposed References to be provided: None Learning Objective 13LP-ILO-ICRXP 1 Question Source: Bank Question History: Watts Bar 2009 Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 6 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KJA# 012000K402 Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following: -

Automatic reactor trip when RPS setpoints are exceeded for each RPS function; basis for each Importance 3.9 4.3 Question # 39 Which of the following describes the automatic actions and basis for these actions if an under-frequency condition exists on 6.9 KV Busses 1-4 with Unit 3 at 100% power?

A. 6.9 KV Busses 1-4 de-energize. This causes a reactor trip to ensure that KW/ft limits are not violated.

B. 6.9 KV Busses 1-4 de-energize. This causes a reactor trip prior to reaching conditions that could lead to lifting a pressurizer safety on a trip.

C. All RCPs trip. This prevents damage to RCPs to ensure they are available if needed for a loss of cooling event.

D. All RCPs trip. This causes a reactor trip to ensure that DNBR limits are not violated.

Answer: D ExplanationlJustification:

A. Incorrect but plausible. Incorrect because the busses do not de-energize.

Plausible because the reactor trip is what is ultimately needed and KWIft is similar to DNBR.

B. Incorrect but plausible. Incorrect because the busses do not de-energize.

Plausible because the reactor trip is what is ultimately needed and preventing lifting a pressurizer safety from lifting is the reason for other protection features.

C. Incorrect but plausible. All RCPs do trip. Protecting RCPs for later use in core cooling is plausible because this is an action FR-C2.

D. Correct answer per System Description 28.

Technical

References:

Syst Desc 28 Proposed References to be provided: None Learning Objective 13LP-ILO-ICRXP 1 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 013000K410 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: -

Safeguards equipment control reset Importance 3.3 3.7 Question # 40 A Safety Injection Signal occurred on both Trains A and B due to I&C testing.

The operating crew reset the SI signal in accordance with emergency operating procedures. Non-SI Blackout Logic Defeat has not been reset. All components functioned as designed. Subsequently, 6.9 KV bus 5 normal feed breaker tripped on overcurrent. What is the status of CCW Pumps?

A. All CCW Pumps are secured B. 31 and 32 CCW Pumps are running 33 CCW Pump is secured.

C. All CCW Pumps are running D. 31 and 32 CCW Pumps are secured, 33 CCW pump is running.

Answer: D Explanation/Justification:

The automatic start signals for CCW pumps are:

1) Non-SI Blackout on the respective bus (5A, 2A, 6A). This signal was blocked when SI occurred and has NOT been reinstated.

2) Safety Injection Signal with off-site power available. This signal has been removed when SI was reset 3} Low CCW header pressure. This signal was blocked during SI or Blackout condition and has not been reset.

A. Incorrect. Plausible because a "blackout" will cause all running pumps on that bus to trip. Candidate must realize that the blackout condition did not occur on bus 6A and 33 CCW pump did not stop.

B. Incorrect. Plausible because the 31 and 32 pumps would have started if Non SI Blackout Defeat signal had been reset; however it was not. Additionally, 33 CCW pump would not trip on a loss of 6.9 kV buses 5, 1, and 2.

C. Incorrect. Plausible because if the Non-SI Blackout Defeat had been reset, all CCW pumps would have been running.

D. Correct. 31 and 32 pumps would trip when 480V buses 5A and 2A de energized due to loss of 6.9 kV bus 5. There is currently no auto start signal so the pumps remain de-energized. CCW pump 33 was running before 6.9 bus 5 lost power and will continue to run throughout the event.

Technical

References:

Syst Desc 10 Syst Desc 27.4 Proposed References to be provided: None Learning Objective 13LP-ILO-CCW001 4 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KfA# 061000A202 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of air to steam supply valve Importance 3.2 3.6 Question # 41 Due to a catastrophic failure of the instrument air system in the aux boiler feed pump building, all air and backup nitrogen pressure has been lost.

Which of the following describes the impact on 32 Auxiliary Boiler Feed Pump and what actions are necessary.

A. Both PCV-1139 and HCV-1118 fail open on loss of air. Local manual control of valves is necessary to control pump speed.

B. Both PCV-1139 and HCV-1118 fail closed on loss of air. Local manual control of valves is necessary to control pump speed.

C. PCV-1139 fails closed and HCV-1118 fails open. Only local manual control of PCV-1139 is necessary to control speed.

D. PCV-1139 fails open and HCV-1118 fails closed. Only local manual control of HCV-1118 is necessary to control speed.

Answer: A ExplanationlJustification:

A. Correct. Both valve fail fully open (as do the 405 flow control valves) resulting in maximum flow to all SGs on a catastrophic loss of IA.

B. Incorrect. Plausible because 32 ABFP starts at minimum speed when air/nitrogen pressure are present. Manual action is required to establish flow to the steam generators.

C. Incorrect. Plausible because PCV-1139 limits down stream steam pressure.

Candidate may believe valve fails closed to prevent lifting 700 pound relief valve.

D. Incorrect. Plausible because 32 ABFP starts at minimum speed when air/nitrogen pressure are present. Manual action is required to establish flow to the steam generators.

Technical

References:

Syst Desc 21.2 Proposed References to be provided: None Learning Objective 13LP-ILO-AFW001 1 13LP-ILO-AFW001 2 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 K1A# 022000A104 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: - Cooling water flow Importance 3.2 3.3 Question # 42 Consider the following possible actions for SWN-TCV-1104/1105:

1. Fail OPEN on a loss of air to ensure adequate containment cooling is available in during an accident.

2. Fail CLOSED on a loss of air to ensure runout of ESW pump(s) does not occur on a loss of instrument air.

3. Get open signal SI circuit

4. Get open signal from High Containment Pressure circuit

5. Throttle to maintain temperature following SI reset From the above list, which of the following contains actions that are correct for SWN-TCV-1104/1105?

A. 1,4,5 B. 2,4 C. 1,3 D. 2,3,5 Answer: C Explanation/Justification:

Discussion of choices:

1. Correct action per System Description 24

2. Incorrect but plausible because if IA fails to these valves under non-SI conditions, ESW pressure will drop very low.

3. Correct action per System Description 24

4. Incorrect but plausible since there is only a need for increased ESW flow to FCUs if there is high pressure in containment.

5. Incorrect but plausible because this is how SWN-FCV-1176/11767A (ESW to EDGs) work.

Technical

References:

Syst Desc 24 Proposed References to be provided: None Learning Objective 13LP-ILO-SW001 6 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 4 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 026000K301 Knowledge of the effect that a loss or malfunction of the CSS will have on the following: - CCS Importance 3.9 4.1 Question # 43 The following plant conditions exist:

• Fan Cooler #32 is tagged out for maintenance.

• A Large Break Loss of Coolant Accident occurred coincident with a Los of Off Site Power

• Emergency Diesel # 33 fails to start

• Local efforts to start the diesel by the NPO were unsuccessful Based only on the actions taken above, what containment cooling equipment is expected to be running when the crew exits E-O?

A. #31, #33 and #34 Fan Cooler Units, #32 Spray Pump B. #33, #34,and #35 Fan Cooler Units, #31 Spray Pump C. #34 and #35 Fan Cooler Units, #31 Spray pump D. #34 and #35 Fan Cooler Units, #32 Spray Pump Answer: D Explanation/Justification:

Bus 5A emergency power 33 EDG Bus 5A supplies 31 and 33 Fan Cooler Units and 31 Spray Pump Bus 2Al3A emergency power 31 EDG Bus 2A supplies 32 Fan Cooler Unit and bus 3A supplies 34 Fan Cooler Unit Bus 6A emergency power 32 EDG

Bus 6A supplies 35 Fan Cooler Unit and 32 Spray Pump A. Incorrect.

B. Incorrect C. Incorrect D. Correct Technical

References:

Logic Unit 3 Sheet 7 Logic Unit 3 Sheet 8 Logic Unit 3 Sheet 8B Proposed References to be provided: None Learning Objective 13LP-ILO-VCCARC 1 13LP-ILO-VCCARC 3 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 039000K508 Knowledge of the operational implications of the following concepts as they apply to the MRSS: - Effect of steam removal on reactivity Importance 3.6 3.6 Question # 44 The Unit is operating at 80% power with all systems in automatic. One Group of condenser steam dump valves fails full OPEN. Assuming that NO operator action occurs, what will be the resulting Rx power level?

A. 0%

B. 70%

C. 80%

D. 90%

Answer: D Explanation/Justification:

The HP Steam Dump System has 12 Steam Dump Valves divided into 4 groups of 3 valves. The capacity of the HP Steam Dump System is approximately 40%.

A. Incorrect. Plausible because a high steam flow condition coincident with low steam pressure or Tavg will result in Safety Injection and Steam Line Isolation resulting in reactor trip.

B. Incorrect. Plausible candidate may believe that less steam flow through the turbine will result in lower reactor power.

C. Incorrect. Plausible candidate may believe that 10% less steam is flowing through the turbine" and the total steam load remains the same.

D. Correct. Turbine load will remain constant and additional steam flow will increase reactor power by approximately 10%.

Technical

References:

2-AOP-UC-1 3-AOP-UC-1 Proposed References to be provided: None Learning Objective 13LP-ILO-SDS001 4 Question Source: Bank Question History: IPEC UNIT 2 NRC 2004 Question Cognitive Level: Com prehension 10 CRF Part 55 Content: 55.41 (b) 6 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 059000A103 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW System controls including:

Power level restrictions for operation of MFW pumps and valves Importance 2.7 2.9 Question # 45 3-AOP-FW-1, Loss of Feedwater, requires tripping the reactor if both MBFPs are not operating when reactor power is above a specified value. Select the statement below that best describes the basis for this step.

The reactor is tripped if power is >  % because a single MBFP can only sustain SG level at power level no higher than  %.

A. 70,70 B. 80,70 C. 80,80 D. 75, 75 Answer: B Explanation/Justification:

Plausibility: B is the correct answer per AOP-FW-1 (trip value) and the lesson plan (max sustainable level backed by simulator). Because it is easy to

.confuse when the 70% and 80% values apply, all distracters are plausible.

Also, before power uprate, 80% was a sustainable value making choice C and D more plausible.

Technical

References:

3-AOP-FW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPFW1 2 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 4 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 059000A403 Ability to manually operate and/or monitor in the control room: -

Feedwater control during power increase and decrease Importance 2.9 2.9 Question # 46 The plant is shutting down for a refueling outage and is currently at 250 MWe about to perform a planned Reactor Trip. All Main Feedwater equipment is available and all steps of POP-2.1 and POP-3.1 have been satisfied at this point in the shutdown. Which of the following is the expected configuration of Main Feedwater prior to the trip?

A. Both MBFPs are in service Low Flow Bypass Feed Regulation Valves are in service B. One MBFPs is in service, One MBFP is in Standby Low Flow Bypass Feed Regulation Valves are in service C. Both MBFPs are in service Main Feed Regulation Valves in service D. One MBFPs is in service, One MBFP is in Standby Main Feed Regulation Valves in service Answer: D Explanation/Justification:

Plausibility: Both MBFPs could stay in service, but POP-3.1 specifies that one pump is placed in standby at 400 MWe even if a trip is planned. MFRVs are placed in service at about the power level that the plant is at, but they are removed from service at 200 MWe and only when we are not tripping the unit.

The correct answer is D based on SOP-FW-OO 1.

Technical

References:

3-S0P-FW-001 Proposed References to be provided: None

Learning Objective 13LP-ILO-POP0073 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 061000K503 Knowledge of the operational implications of the following concepts as they apply to the AFW System: - Pump head effects when control valve is shut Importance 2.6 2.9 Question # 47 During the startup of 31 Auxiliary Boiler Feedwater pump as part of a plant shutdown, the operator inadvertently places the recirculation valve control switch in close instead of automatic. What affect will this have on the Auxiliary Boiler Feedwater pump?

A. The low flow trip is disabled; the pump will overheat if flow decreases too low.

B. The low follow trip remains enabled, the pump will trip if flow drops below 40 gpm for >15 seconds.

C. The low flow alarm only is disabled, but the pump will still trip if flow remains below 130 gpm for >15 seconds.

D. The low flow alarm is enabled; an orificed line to the CST will maintain minimum flow.

Answer: A Explanation/J ustification:

A. Correct. Both alarm and trip functions are disabled if the valve switch is in CLOSE or OPEN (Not automatic)

B. Incorrect.

C. Incorrect. Plausible because the low flow trip timer starts if flow is < 40 gpm and the trip occurs if the flow is not greater than 130 gpm in 15 seconds.

D. Incorrect. Plausible because 32 AFW pump has an orificed recirculation line to the CST Technical

References:

3-ARP-006 Proposed References to be provided: None Learning Objective 13LP-ILO-AFW001 5 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 4 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 062000K302 Knowledge of the effect that a loss or malfunction of the A.C.

Distribution System will have on the following: - ED/G Importance 4.1 4.4 Question # 48 A grid disturbance in the Buchanan Switchyard resulted in a unit trip and loss of off site power. The BOP identifies buses 5A and 6A are energized from their respective diesel generators.

Of the following conditions, which will prevent the automatic closure of the 31 Emergency Diesel Generator breaker (52/EG-1)7 A. Fault on 480 VAC Bus 3A B. Normal feed breaker for Bus 2A (52/2A) closed C. Bus 2A-3A tie breaker is closed.

D. Normal feed breaker for Bus 3A (52/3A) closed

{

Answer: B ExplanationlJustification:

A. Incorrect. Plausible because 31 EDG will energize bus 3A through the tie breaker if closed. The tie breaker will not close if bus 3A is faulted.

B. Correct. Normal feed breaker for bus 2A should have opened when voltage was lost.

C. Incorrect. Plausible because tie breakers between other buses will prevent auto closure of the respective EDG output breakers. This breaker will auto close to energize bus 3A.

D. Incorrect. Plausible because 31 EDG will energize bus 3A through the tie breaker if closed. The tie breaker will not close if bus 3A feeder breaker is closed.

Technical

References:

Syst Desc 27.4 Proposed References to be provided: None Learning Objective 13LP-ILO-EDS480 2 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 063000A403 Ability to manually operate and/or monitor in the control room:

Battery discharge rate Importance 3 3.1 Question # 49 Given:

• The plant experienced a steam break in the turbine building.

• All equipment functioned as designed.

• All required procedural actions were completed

• When the crew entered ES-1.1, SI Termination, off-site power was temporarily lost.

Which of the following describes actions (if any) related to the electrical lineup?

A. MCCs with battery chargers must be aligned and reset within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. MCCs with battery chargers must be aligned and reset within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. MCCs with battery chargers do not strip for Blackout, no alignment is necessary.

D. MCCs previously reset will remain energized; no further action necessary.

Answer: A ExplanationlJustification:

A. Correct answer per electrical distribution and T.S. 3.8.4 bases. Three of four safeguards battery chargers are powered from MCCs that will strip and not automatically re-energize.

B. Incorrect but plausible if the candidate does not remember the discharge time of the batteries

C. Incorrect but plausible since it is believable that action should not be necessary to keep battery chargers energized.

D. Incorrect but plausible since MCCs were probably aligned and reset with offsite power prior to reaching ES-1.1.

Technical

References:

3-S0P-EL-015 Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-EDS 125 2 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KJA# 1940012403 Emergency Procedures/Plan Ability to identify post-accident instrumentation.

Importance 3.7 3.9 Question # 50 The fire main in the 480V Switchgear room ruptured resulting in a loss of all 480V AC buses. The crew entered ECA-O.O, Loss of All AC Power and started the Appendix R Diesel Generator. Bus 312 is energized and all Appendix R equipment is available.

Which of the following properly states the post-accident instrumentation available with power from the Appendix R Diesel Generator?

A. 31,32,33 SG Narrow Range Level Pressurizer Level Pressurizer Pressure CST Level R-27 Core Exit Thermocouples B. 31, 32, 33 SG WR Level Pressurizer Level Pressurizer Pressure 31 Loop WR Thot 31 Loop WR Tcold N-38 C. 31,32,33 SG Narrow Range Level Pressurizer Level Pressurizer Pressure CST Level R-27 31 Loop WR Tcold D. 31,32,33 SG WR Level Pressurizer Level Containment Pressure

31 Loop WR Thot RWST Level N-38 Answer: B Explanation/Justification:

A. Incorrect. Plausible because all instrumentation except R-27 is Post Accident Instrumentation per Tech Specs; however only Pressurizer Level and Pressure can be energized from the Appendix R EDG.

B. Correct C. Incorrect. Plausible because all instrumentation except R-27 is Post Accident Instrumentation per Tech Specs; however only 31 Loop Tcold, Pressurizer Level and Pressure can be energized from the Appendix R EDG.

D. Incorrect. Plausible because all instrumentation is Post Accident Instrumentation per Tech Specs; however only SG WR Levels, Pressurizer Level 31 Loop Thot, and N-38 can be energized from the Appendix R EDG.

Technical

References:

3-AOP-SSD-1 Proposed References to be provided: None Learning Objective 13LP-ILO-EDSAPR 2 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 073000A402 Ability to manually operate and/or monitor in the control room:

Radiation monitoring system control panel Importance 3.7 3.7 Question # 51 R-19 (SG Blowdown Activity) is in alarm and has caused an auto-closure of SG sample isolation valves. Unit 3 remains on line. Which of the following actions will be necessary to re-open these valves to sample SGs?

A. Acknowledge R-19 alarm on RM-23 panel to re-open valves.

B. Purge R-19 or raise alarm setpoint to re-open valves.

C. Place individual SG sample isolation valves in Rad Bypass position.

D. Press individual valve reset buttons on Panel SMF.

Answer: C Explanation/Justification:

A. Incorrect but plausible since there is an acknowledge button on the RM-23 panel for R-19.

B. Incorrect but plausible since this would actually work, but it is not the specified action.

C. Correct answer D. Incorrect but plausible since this action is taken for opening valves after Phase A isolation.

Technical

References:

Syst Desc 9 Proposed References to be provided: None

Learning Objective 13LP-ILO-RMSPRM B Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 076000A202 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Service water header pressure Importance 2.7 3.1 Question # 52 A large leak has developed on the Essential Service Water Header causing SW Header pressure to drop to 40 psig. All available SW Pumps are operating and component temperatures are elevated but stable below alarm setpoints. The plant is currently at 25% power.

Which of the following describes the required operator action for the above condition?

A. Isolate Service Water Header and restore header to operable within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in Mode 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B. Maintain present power level unless turbine generator bearing oil drain temperatures cannot be maintained below 180°F then trip the turbine.

C. Maintain present power level unless oil temperature from the MBFP bearings cannot be maintained below 190°F then trip the affect MBFP(s).

D. Initiate a plant shutdown to Mode 3 using POP-3.1, Plant Shutdown from 45% Power as long as adequate cooling can be maintained.

Answer: D Explanation/Justification:

A. Incorrect. Plausible because depending on power level, main turbine bearing temperatures may rise rapidly; however, at 25% power this is not expected.

B. Incorrect. Plausible because bearing high temperature alarm is 150°F and the trust bearing trip is at 210°F.

C. Incorrect. Plausible because the Boiler Feed Pump Lube Oil Separator will shutdown at 180°F D. Correct.

Technical

References:

3-AOP-SW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-SW001 1 Question Source: Modified Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 076000K201 Knowledge of bus power supplies to the following: - Service water Importance 2.7 2.7 Question # 53 Given the following:

• The plant is at 100% power

• 32, 33, and 36 Service Water Pumps are running

• 31,34, and 35 Service Water Pumps are in AUTO

• The Service Water Pump Mode Selector Switch is in the 4-5-6 position

• The tie breaker between 2A and 3A (2AT3A) is OOS for maintenance.

Assuming no operator action, which one of the following describes the status of the SW Pumps after a Reactor Trip coincident with a Loss of Oft-site power occurs?

A. 32, 34, and 36 SW Pumps will be running B. 35 and 36 SW Pumps will be running C. 34,35, and 36 SW Pumps will be running D. 34 and 36 SW Pumps will be running Answer: 0 ExplanationlJustification:

A. Incorrect. Plausible because 32 is powered from 2A, however it is not selected for essential header so it would not start. 34 and 36 pumps running is correct.

B. Incorrect. Plausible because 35 is selected for essential; however, there is no power to the pump with 3A de-energized.

C. Incorrect. Plausible because all pumps are selected for the essential; however, 35 does not have power with 3A de-energized.

D. Correct. With the 2AT3A OOS the EDG will start and energize 2A. 3A will not energize and 35 SW pump will not start.

Technical

References:

Proposed References to be provided: None Learning Objective 12LP-ILO-SW001 4 12LP-ILO-SW001 6 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 078000A301 Ability to monitor automatic operation of the lAS, including:

Air pressure Importance 3.1 3.2 Question # 54 Unit 3 is at 100% power with 31 Instrument Air Compressor (lAC) in HAND and 32 lAC in AUTO. SOV-1142, Station Air Makeup to Instrument Air is initially closed when a rupture occurs somewhere in the Instrument Air System. What indications do you expect to see in the Control Room for this event?

A. At 105 psig a red light for 32 lAC and at 100 psig a red light for SOV-1142 B. At 100 psig a red light for 32 lAC and at 95 psig a red light for SOV-1142 C. At 95 psig a red light for 32 lAC and at 90 psig a red light for SOV-1142 D. At 90 psig a red light for 32 lAC and at 85 psig a red light for SOV-1142 Answer: C Explanation/Justification:

A. Incorrect but plausible. 105# is the unload point for the IA compressors, so it is plausible that an operator could confuse this with 95#. 100# is 5# below this which would be the logical setpoint for 1142 if the operator thought 105# was the auto start setpoint for the standby compressor.

B. Incorrect but plausible. 100# is the load setpoint for an IA compressor in HAND. For the same reasons as A, this would make these setpoints plausible.

C. Correct answer per system description and procedures.

D. Incorrect but plausible. 90# is the actual setpoint for 1142, so these setpoints are plausible.

Technical

References:

3-S0P-IA-001 Syst Desc 29.2 Proposed References to be provided: None Learning Objective 13LP-ILO-IA001 7 Question Source: Bank Question History: Not on NRC Exam Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KfA# 103000K301 Knowledge of the effect that a loss or malfunction of the Containment System will have on the following: - Loss of containment integrity under shutdown conditions Importance 3.3 3.7 Question # 55 Unit 3 is in a forced maintenance outage with RCS temperature being maintained at 300°F while repairs are being made to 32 AFW Pump. Material is being brought into containment through the 95 ft Airlock, when the inner door is damaged. The door cannot be closed. No other equipment was damaged. What is the operational impact of this event?

A There is no impact in the current mode. Repairs will be necessary prior to entering a higher mode.

B. If the outer door is verified closed within one hour and locked closed within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, operation can continue in the current mode.

C. The airlock is inoperable. The airlock must be declared operable within one hour or the plant will have to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

D. The containment system is inoperable. The system must be declared operable within one hour or the plant will have to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Answer: B Explanation/Justification:

A. Incorrect but plausible since the plant is in Mode 4 which removes some containment restrictions, but not this one.

B. Correct answer per LCO 3.6.2

C. Incorrect but plausible. The system is required to be operable per the LCO statement, so it is reasonable that going to a mode where the LCO does not apply would be required.

D. Incorrect but plausible since this is the exact requirement if the containment system is inoperable.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-VCVCB 7 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 001000K205 Knowledge of bus power supplies to the following: - M/G sets Importance 3.1 3.5 Question # 56 Given:

• The unit is at 100% power

• 31 EDG is OOS for a breaker PM

• A fault occurred in the Buchanan switchyard resulting in a loss of 138KV power.

What is the immediate impact on the plant?

A. The reactor will trip due to loss of both Rod Drive MG Sets when bus 138kV power is lost.

B. Since 31 Rod Drive MG set was required to be secured due to 31 EDG being OOS; the reactor trips when bus 138kV power is lost.

C. The reactor will remain at approximately1 00% power with only 31 Rod Drive MG set energized.

D. The reactor will remain at approximately100% power with both Rod Drive MG Sets energized.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because candidate my believe the rod drive MG sets are powered from 5A and 6A which both loose power when offsite power is lost.

Note: Buses 2A and 3A do not loose power and 31 Rod Drive MG Set will continue to operate (powered from 2A)

B. Incorrect. Plausible because a precaution in 3-S0P-EL-1, states "If and EDG is sole power supply for a 480V bus and 2 rod drive MG seta are in operation then MG se supplied by EDG should be shutdown." Candidate may believe 31 RDMG set is already shutdown.

C. Correct.

D. Incorrect. Plausible because candidate may believe the MG set nywheel will maintain voltage until EDG re-energies the bus. The MG set supply breaker will trip on undervoltage.

Technical

References:

Proposed References to be provided: None Learning Objective 13LP-ILO-ICROD 4 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 002000A204 Ability to (a) predict the impacts of the following malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Loss of heat sinks Importance 4.3 4.6 Question # 57 Control Room operators are responding to a loss of secondary heat sink.

Attempts to restore feedwater flow to the steam generators have been unsuccessful, so the operators are attempting to provide "Bleed and Feed Cooling" to the RCS. Only two Charging Pumps have been started. During performance of these actions there is a delay in starting SI Pumps and the STA recommends opening the PZR PORVs while waiting for SI flow to be established.

Which of the following is your assessment of the STA's recommendation?

A. Agree. With two Charging Pumps running it is acceptable to open the PORVs while continuing with the steps necessary to establish SI flow.

B. Agree. The procedure Caution states that due to a loss of secondary heat sink, the steps to implement bleed and feed should be immediately initiated.

C. Disagree. The correct action would be to start the third Charging Pump then open both PORVs while waiting for SI flow to be established.

D. Disagree. Opening the PORV before SI Pumps are operating could result in uncovery and severe damage to the core.

Answer: D Explanation/Justification:

A. Incorrect. Plausible since RCS pressure at this point is above the shut off head of the HHSI pumps. Establishing a bleed path is essential but not done without feed established.

B. Incorrect. Plausible since this is an actual caution.

C. Incorrect. Plausible since all charging pumps are started prior to bleed and feed.

D. Correct answer per FR-H.1 Technical

References:

2-FR-H.1 3-FR-H.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRH 7 Question Source: Bank Question History: Not on NRC Exam Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 011000K515 Knowledge of the operational implications of the following concepts as they apply to the PZR LCS: - PZR level indication when RCS is saturated Importance 3.6 4 Question # 58 Given the following conditions:

• The unit was operating at 100% power

• A malfunction has caused a Pressurizer Safety Valve to open and stick open

• A Reactor Trip and Safety Injection occurred

• The RCS rapidly depressurized to saturation conditions

• Pressurizer level initially dropped and then began to increase rapidly What is the relationship between pressurizer level and RCS inventory under these conditions?

A. Level is not an accurate indication of inventory, because the pressurizer level channel reference legs will flash causing an indicated increase in level B. Level is not an accurate indication of inventory, because RCS voiding may result in a rapidly increasing pressurizer level.

C. Level is an accurate indication of level, because hydraulic pressure would force any voids into the pressurizer steam space and out the safety valve.

D. Level is an accurate indication of inventory, because voiding would occur in the pressurizer prior to reaching saturation conditions in the RCS.

Answer: B Explanation/Justification:

A. Incorrect. Plausible because in general reference leg flashing will cause increase in indicated level; however, pressurizer reference legs are sealed and cannot "flash" B. Correct C. Incorrect. Plausible because steam voids will go to the highest point in the system; however, the steam voids will hold water in the pressurizer causing level to indicate high.

D. Incorrect. Plausible because the statement is true; however, the dynamics of RCS cause water to be drawn into and held in the pressurizer by the steam passing to the open safety vavle.

Technical

References:

3-E-1 Background Proposed References to be provided: None Learning Objective IOLP-ILO-TAA001 14 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KfA# 015000A302 Ability to monitor automatic operation of the NIS, including:

Annunciator and alarm signals Importance 3.7 3.9 Question # 59 With the plant operating at 90% power, the following alarms have annunciated:

• NIS POWER RANGE OVER POWER ROD STOP

• NIS POWER RANGE SINGLE CHANNEL HIGH RANGE TRIP

• NIS POWER RANGE LOSS OF DETECTOR VOLTAGE

• NIS POWER RANGE DROPPED ROD ROD STOP All Power Range Channels are indicating 90% power. Which of the following are correct, with no operator action?

A. Instrument power fuse has blown, Control rods will not move B. Instrument power fuse has blown, Control rods can only be inserted C. Control power fuse has blown, Control rods will not move D. Control power fuse has blown, Control rods can only be inserted Answer: D Expla nationlJustification:

Modified Watts Bar May 2009 Question 37

It is plausible that the instrument power fuses have failed because of the loss of detector voltage alarm being up. It is plausible that all rod motion is stopped because the over power alarm locks out manual out motion.

Correct answer is D based simulator response. While the system description, lesson plan, and ARP would indicate that a control power fuse failure will cause these symptoms, there is not a concise reference for the effects of a failed control power fuse.

Technical

References:

3-AOP-NI-1 3-S0P-NI-001 Proposed References to be provided: None Learning Objective 13LP-ILO-ICEXC 6 Question Source: Bank Question History: Watts Barr May 2009 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 6 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 017000A402 Ability to manually operate and/or monitor in the control room:

Temperature values used to determine RCS/RCP operation during inadequate core cooling (Le., if applicable, average of five highest values)

Importance 3.8 4.1 Question # 60 Select the answer which describes when an RCP would be restarted while performing FR-C.1, "RESPONSE TO INADEQUATE CORE COOLING".

A. SG Blowdown to Atmospheric pressure has failed to reduce CET temperatures below 1200°F.

B. All three HHSI pumps have failed to start and CET temperatures are above 1200°F.

C. Opening of PORVs has been unsuccessful in reducing CET temperatures below 1200°F.

D. RCPs must be started any time CET temperatures exceed 1200°F.

Answer: A Exp lanation/Justification:

A. Correct answer per FR-C.1 B. Incorrect since more steps in FR-C.1 would have to be completed to start RCPs. Plausible since 1200°F is the setpoint.

C. Incorrect. Plausible since this strategy is used in FR-H.1.

D. Incorrect. Plausible since 1200°F is the setpoint.

Technical

References:

3-FR-C.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRC 13 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 K1A# 041000K106 Knowledge of the physical connections and!or cause-effect relationships between the SDS and the following systems:

Condenser Importance 2.6 2.9 Question # 61 A Normal Plant Shutdown is in progress. You are at 200 MWe preparing to shift the steam dumps from temperature mode to pressure mode. When in pressure mode, what interlocks must be made up to open the High Pressure Steam Dumps?

A. Vacuum greater than 25" and the circ water pump in service for that condenser sextant.

B. Vacuum greater than 29" and condensate running through the condenser.

C. One condensate pump running and condenser vacuum greater than 27".

D. One circulating water pump per condenser running and Turbine Trip as sensed by relay P5 Answer: A Explanation!Justification:

A. Correct.

B. Incorrect. Plausible because both a minimum value of condenser vacuum is necessary and condensate flow through the condenser. The value of condenser

vacuum is incorrect and condensate flow is indicated by circ water pump breaker position.

C. Incorrect. Plausible because Condensate flow path A contains the air ejector condensers which are necessary to maintain vacuum. The value of condenser vacuum is incorrect.

D. Incorrect. Plausible because the circ water pump for a condenser sextant must be operating for its corresponding valves to operate. The turbine trip (P5) relay changes the temperature control mode 'from load rejection to turbine trip.

Technical

References:

Syst Desc 18.1 Proposed References to be provided: None Learning Objective 13LP-ILO-SDS001 2 13LP-ILO-SDS001 3 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 029000K403 Knowledge of Containment Purge System design feature(s) and/or interlock(s) which provide for the following: - Automatic purge isolation Importance 3.2 3.5 Question # 62 Given the following plant conditions on Unit 3:

• The unit is in Mode 6 with refueling activities in progress.

• Containment purge is in service.

• A fuel element accidentally dropped into the cavity.

• R-11, R-12 and R-14 have a HIGH alarm.

• R-25 and R-26, VC High Range Area Monitors, have a HIGH alarm.

Which ONE of the following actions would occur, assuming that operators follow the required actions of 3-AOP-FH-1, "Fuel Damage or Loss Of SFP/Refueling Cavity Level", and all equipment responds as designed?

A. Containment evacuation alarm sounds automatically. Containment purge stops automatically.

B. Containment evacuation alarm is manually actuated by the control room operator. Containment purge stops automatically.

C. Containment evacuation alarm sounds automatically. Containment purge is stopped manually by the control room operator.

D. Containment evacuation alarm is manually actuated by the control room operator. Containment purge is stopped manually by the control room operator.

Answer: A

Explanation/J ustification:

DC Cook 2008 Question 66 Modified A. Correct.

B. Incorrect. Plausible because R-25 and R-25 do not have auto functions. Also procedure directs the operator to make an announcement to evacuate containment.

C. Incorrect. Plausible because Containment evacuation alarm does sound automatically; however, containment purge will automatically stop from R-11/12.

D. Incorrect. Plausible because R-25 and R-25 do not have auto functions.

Technical

References:

3-AOP-FH-1 3-S0P-RM-010 Proposed References to be provided: None Learning Objective 13LP-ILO-ONPRM2 2 Question Source: Bank Question History: DC Cook NRC 2008 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 034000K602 Knowledge of the effect of a loss or malfunction of the following will have on the Fuel Handling System: - Radiation monitoring systems Importance 2.6 3.3 Question # 63 Given the following conditions:

• Fuel handling is in progress in Containment and the Spent Fuel Pool (SFP).

• FSB Fans are operating in NORMAL mode.

• Radiation Monitor R 5 SPENT FUEL POOL AREA MONITOR fails LOW.

• All other radiation monitors are operable.

What is the impact of this failure?

A. Only recently irradiated fuel assembly movement in the FSB must be suspended.

B. All fuel assembly movement in the FSB AND Containment must be suspended.

C. All fuel assembly movement in the FSB must be suspended.

D. Fuel assembly movement may continue in the FSB and Containment provided R-14, Plant Vent Gas Activity and R-27, Wide Range Plant Vent Gas Activity, monitors are operable.

Answer: C Explanation/Justification:

A Incorrect. Plausible since it would be reasonable that the requirement only apply to recently irradiated fuel assemblies. Tech Specs only apply to recently irradiated fuel.

B. Incorrect. Plausible since it would be reasonable that all fuel movement would be suspended when R-5 is unavailable.

C. Correct answer per the TRM D. Incorrect. Plausible since it is reasonable that compensatory radiation monitoring would be needed if R-5is OOS.

Technical

References:

3-0NOP-RM-1 Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-ONPRM2 2 Question Source: Bank Question History: IP32006 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KfA# 0450002127 Conduct of Operations Knowledge of system purpose and or function.

Importance 3.9 4 Question # 64 The plant is at 100% power. You have just adjusted Main Turbine Generator VARs from 100 MVAR Leading to 100 MVAR Lagging. Based on Flight Panel indications, which of the following list of parameter changes describes what you expect following this adjustment? Assume all systems are in automatic and functioning properly.

A. Generator Voltage has increased, Exciter Field Amps have increased, Gross MWe has observably decreased.

B. Generator Voltage has increased, Exciter Field Amps have increased, Gross MWe has NOT observably changed.

C. Generator Voltage has decreased, Exciter Field Amps have decreased, Gross MWe has observably increased.

D. Generator Voltage has decreased, Exciter Field Amps have decreased, Gross MWe has NOT observably changed.

Answer: B Explanation!Justification:

The correct answer is B. Based on plant experience and the simulator readings.

For the simulator, Generator Voltage goes from 21.1 KV to 21.9 KV, Exciter Field Amps go from 127 Amps to 166 Amps, and there is no change in MWe out to the

0.1 place. The flight panel is a 0-1200 dial meter, so there is never an observable change in MWe for a voltage adjustment at full power.

There are two sets of wrong information in the distracters that are both plausible.

The first is that changing voltage and field amps will cause a noticeable change in MWe. This is not the case. The second is that a candidate could confuse how lagging and leading VARs correspond to voltage.

Technical

References:

Syst Desc 27.2 Proposed References to be provided: None Learning Objective 13LP-ILO-EDS22K 3 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KfA# 086000A105 Ability to predict and!or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including: - FPS lineups Importance 2.9 3.1 Question # 65 Given the following:

• The unit is at 100% power

• Fire Protection header pressure is lowering Which one of the following has the alternate pumps listed in the proper sequence of automatic start?

A. Standby Jockey Pump at 115 psig, Motor Driven Fire Pump at 110 psig; Diesel Fire Pump at 105 psig without time delay B. Standby Jockey Pump at 110 psig; Motor Driven Fire Pump at 105 psig; Diesel Fire Pump at 95 psig without time delay C. Standby Jockey Pump at 115 psig; Motor Driven Fire Pump at 110 psig; Diesel Fire Pump when <105 psig for more than 10 seconds D. Standby Jockey Pump at 110 psig; Motor Driven Fire Pump at 105 psig; Diesel Fire Pump when <95 psig for more than 10 seconds Answer: B Explanation!Justification:

Below are the correct Auto Start pressure values for the Fire Pumps in this question:

Diesel Fire Pump 95 psig Motor Driven Fire Pump 105 psig Jockey Fire Pump Lead 115 psig Jockey Fire Pump Stby 110 psig A. Incorrect. Plausible because the values used are correct for different pumps.

B. Correct.

C. Incorrect. Plausible because the values used are correct for different pumps.

D. Incorrect. Plausible because the values used are correct for different pumps.

Technical

References:

3-S0P-FP-001 Proposed References to be provided: None Learning Objective 13LP-ILO-FPS001 2 Question Source: Bank Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012105 Conduct of Operations - Ability to use procedures related to shift staffing such as minimum crew compliment, overtime limitations, etc.

Importance 2.9 3.9 Question # 66 Unit 3 is in a refueling outage with fuel movement in progress. Per EN-OP-115, Conduct of Operations, what is the minimum Unit 3 on-duty control room compliment for this condition? (Assume Fire Brigade Leader is staffed by Unit 2 individual and a Unit 3 SRO is staffed as Refueling SRO)

A. ONE Shift Manager, ONE Control Room Supervisor, TWO Reactor Operators, ONE Shift Technical Advisor.

B. ONE Shift Manager, ONE Control Room Supervisor, TWO Reactor Operators.

C. ONE Shift Manager, TWO Reactor Operators, ONE Shift Technical Advisor D. ONE Shift Manager, ONE Control Room Supervisor, ONE Reactor Operator.

Answer: B Explanation/Justification:

A. Incorrect since STA not required. STA is required in MODE 1-4 and plausible that STA could be required during fuel movement.

B. Correct answer per EN-OP-115.

C. Incorrect since CRS is required. Plausible since at one time the SRO could become the RSRO per T.S.

D. Incorrect since two ROs are needed. Plausible to believe only one is required for minimum staffing shut down.

Original Question Kewaunee Unit 1 2/2/06 What is the MINIMUM on-duty shift complement when the plant is in COLD SHUTDOWN?

Answer: ONE Shift Manager (SRO), TWO Licensed Reactor Operators, TWO Nuclear Auxiliary Operators, and ONE Radiation Technologist.

Dist-1: ONE Shift Manager (SRO), ONE Licensed Reactor Operator, TWO Nuclear Auxiliary Operator, ONE Radiation Technologist.

Dist-2: ONE Shift Manager (SRO), ONE Licensed Reactor Operator, TWO Nuclear Auxiliary Operator, ONE Radiation Technologist and the STA within 10 minutes of control room.

Dist-3: ONE Shift Manager (SRO), TWO Licensed Reactor Operators, TWO Nuclear Auxiliary Operator, ONE Radiation Technologist and the STA within 10 minutes of control room.

Technical

References:

EN-OP-115 Proposed References to be provided: None Learning Objective IOLP-ILO-ADM01 3 Question Source: Bank Question History: Kewaunee Unit 1 2006 Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 0000542123 Conduct of Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance 4.3 4.4 Question # 67 The plant is at 100% power with all systems in automatic when both heater drain pumps trip. Which of the following actions will the operator take in order to stabilize the plant?

A. Immediately run the turbine back to 70% power to prevent a main feed pump trip.

B. Reduce turbine load as needed to match steam flow and feed flow while maintaining MBFP suction pressure above 265 psig.

C. Reduce turbine load to 900 MW and ensure MBFP suction pressure remains greater than 265 psig at the new load.

D. If MBFP suction pressure decreases to 265 psig, reduce turbine load until suction pressure is greater than 265 psig.

Answer: B Explanation/Justi'fication:

A. Incorrect. Plausible because 70% is the approximate power level the procedure suggests if ONE heater drain tank pump is lost.

B. Correct.

C. Incorrect. Plausible because a power reduction is necessary and maintaining suction pressure greater than 265 is necessary; however, 900 MWe will be inadequate,

D. Incorrect. Plausible because maintaining suction pressure greater than 265 is necessary; however, it is not necessary to wait for pressure to decrease to <

265 to initiate the load reduction.

Technical

References:

3-AOP-FW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPFW1 3 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012137 Conduct of Operations Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Importance 4.3 4.6 Question # 68 The plant had been operating at 100% power for the last 100 days, when a reactor trip occurred 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> ago.

Conditions prior to the trip were:

• Core burnup is 10,000 MWD/MTU (MOL)

• Boron concentration was 700 ppm

• Control Bank D was at 230 steps The estimated critical position is Control Bank D at 120 steps. The operators have just completed pulling the shutdown banks. The source range counts (in CPS) while withdrawing the shutdown banks are:

SR# Initial CPM SB"A" SB"B" SB"C" SB"D" t 31 165 200 280 300 350 t 32 180 220 260 300 340 Which action below should crew take?

A. Suspend rod withdrawal, re-evaluate all inputs and mathematics used in the ECP, and direct Chemistry to obtain an RCS boron sample.

B. Fully insert all CONTROL RODS, and re-evaluate all inputs and mathematics used in the ECP.

C. Obtain new baseline counts for both source range channels and commence pulling CONTROL BANKS. Criticality is expected at about the 5th doubling.

D. Calculate the amount and borate so that criticality will occur at 120 steps

on Control Bank D and resume CONTROL RODS to go critical. Criticality is expected at about the 5th doubling.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because procedure operator guidelines direct action if the counts exceed 2 doublings during shutdown bank withdrawal. Counts did not double; however, they are close.

B. Incorrect. Plausible because procedure operator guidelines direct action if the counts exceed 2 doublings during shutdown bank withdrawal. Counts did not double; however, they are close.

C. Correct.

D. Incorrect. Plausible because counts are higher than normally expected for withdrawal of shutdown banks.

Technical

References:

3-POP-1.2 Proposed References to be provided: None Learning Objective IOWKB-ILO-ADMOO 12 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KJA# 1940012201 Equipment Control - Ability to perform pre-startup procedures for the facility. including operating those controls associated with plant equipment that could affect reactivity.

Importance 4.5 4.4 Question # 69 Given the following conditions:

• A reactor startup is in progress.

• A malfunction of the Rod Control Bank Overlap Unit has caused Control Bank D Rods to begin withdrawing 20 steps earlier than designed.

Which ONE (1) of the following describes the impact of this malfunction if it were allowed to continue during the power increase?

A. QPTR limits may be challenged.

B. Shutdown margin may be reduced if a reactor trip occurs.

C. Power Peaking factors may rise to unacceptable values.

D. The Safety Analysis for ejected rod worth may be invalid.

Answer: C Explanation/Justification:

A. Incorrect. Any change should be similar in all quadrants.

B. Incorrect. There is more potential reactivity insertion from this position.

C. Correct. TS Basis.

D. Incorrect. Safety Analysis for ejected rod worth is a fixed value that will not change due to rod mispositioning or improper bank overlap.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 12LP-ILO-ICROD 14 12LP-ILO-ICROD 9 13LP-ILO-ICROD B 13LP-ILO-ICROD I Question Source: Bank Question History: Not NRC Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KfA# 1940012202 Equipment Control - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Importance 4.6 4.1 Question # 70 Per POP-1.1, Plant Heatup from Cold Shutdown Condition, when are the Over Pressure System (OPS) states links required to be OPENED?

A. Prior to closing the final RCS vent path.

B. Prior to starting the first RCP.

C. Prior to exceeding 200°F.

D. Prior to exceeding 350°F.

Answer: D Expla nation!Justification:

The states links are required to be closed per T.S. below 3360F. T.S. never requires the links to be open. Unit 3 somewhat uniquely operates with these links open when the plant is above Mode 4. Since an operator could be confused as to the function of these links, all answers are plausible since they are heatup milestones.

Technical

References:

3-POP-1.1 Proposed References to be provided: None Learning Objective 13LP-ILO-POP005 3 Question Source: New

Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

K1A# 1940012236 Equipment Control - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance 3.1 4.2 Question # 71 The following plant conditions exist:

• 100 % power

• 32 EDG is out of service for preventative maintenance The maintenance supervisor requests a work permit for 31 AFW pump to be worked now due to man power availability. The estimated completion time for the work is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Should 31 AFW pump be taken out of service? Select the proper action to be taken with the justification for your choice.

A. Yes. Technical Specifications allow for one MDAFW pump to be out of service, provided it is returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the remaining pumps are demonstrated operable.

B. Yes. Technical Specifications allows for one MDAFW pump to be out of service, for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided the remaining pumps are demonstrated operable.

C. No. Technical Specifications requires all 3 AFW pumps with their associated piping and valves to be operable if any EDG is inoperable.

D. No. Technical Specifications state that if one EDG is out of service then the other 2 EDG's and their associated safeguards equipment should be maintained operable.

Answer: D Explanation!Justification:

A. Incorrect. Plausible because the turbine driven (32 AFW) pump is a 100%

capacity pump that supplies flow to all 4 SG. A single MDAFW Pump may be Inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, however, with 32 EDG OOS (emergency supply for second MDAFW) this would make 33 MDAFW inoperable also.

B. Incorrect. Plausible because the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowable time is used in Tech Specs and is shorter than the standard AOTs. For example in MODE 3 HHSI pump made incapable of injecting pursuant to LCO 3.4.12 LTOP is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Incorrect. Plausible because 33 MDAFW pump need not be declared inoperable simpley because its EDG is inoperable. When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. If a loss of safety function is determined to exist by this program, the pump should be declared inoperable.

D. Correct.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective Question Source: Modified Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 7 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012312 Radiological Controls Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance 3.2 3.7 Question # 72 The following conditions exist:

• The PAB Sump Tank Room is posted as a "Locked High Radiation Area".

• The area is at the MINIMUM required radiation level for the posting in all posted areas.

• Your present annual exposure is 500 mRem.

• You are required to perform a retest within the posted area.

What would be your maximum calculated stay time in this area be in order to avoid exceeding the Entergy Routine Annual Limit for radiation exposure per EN RP-201, Dosimetry Administration?

Assume 80% of dose limit guidelines are not applicable for this evolution.

A. 30 minutes B. 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> C. 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> D. 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Answer: B Explanation/Justification:

The Entergy routine administrative limit per EN-RP-201 is 2000 mR/year and the minimum dose rate for a locked high radiation area is 1000 mR/hr. Therefore the correct answer is 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (500 mR + 1.5 hrs x 1000 mR/hr = 2000 mR). 30 minutes is plausible since the candidate could think the admin limit is 1000 mR.

2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is plausible since the candidate could think the limit is 3000 mR. 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is plausible since the candidate could think the limit is 5000 mR.

Original Question from Millstone Unit 2 12/4/2002 INPO 25161 The plant has been shut down due to a leak in the CVCS Regenerative Heat Exchanger. You have been directed to open and red tag the vents and drains associated with the heat exchanger. The following conditions exist:

- The area around the CVCS Regenerative Heat Exchanger has been posted a

'Locked High Radiation Area'.

- The area is at the MINIMUM required radiation level for the posting.

- Your present annual exposure is 500 mRem.

- All the valves you have been assigned to operate are inside the posted area.

What would be your maximum calculated stay time in this area in order to avoid exceeding the Millstone administrative limit?

Answer: 30 minutes Distracter 1: 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Distracter 2: 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Distracter 3: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> IPEC's answer is different since there are differences in administrative dose limits.

Technical

References:

EN-RP-201 Proposed References to be provided: None Learning Objective IOLP-ILO-ADM01 4 Question Source: Bank Question History: Millstone Unit 2 12/4/2002 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 9 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KJA# 1940012313 Radiological Controls Knowledge of radiological safety procedures pertaining to licensed operator duties, such as radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance 3.4 3.8 Question # 73 Unit 3 is at power and CH-LCV-459, Letdown Isolation, has failed closed.

Maintenance believes the reason the valve failed closed is due to a problem with a solenoid on the valve that can be repaired in 5 minutes. Dose rates are 2000 mR/hr at the job site. Which of the following describes the requirements per OAP-007, Containment Entry and Egress, that need to be satisfied to make this repair?

(Assume all other procedural requirements for the repair and entry are met and workers will not exceed annual exposure limits from this job.)

A. Reactor power must be lowered until dose rates are less than 1000 mR/hr at the job site prior to allowing work to begin.

B. Since the expected dose is less than 1000 mRem, only normal OAP-007 SM authorization is required prior to allowing job to begin.

C. Unit AOM and RP Manager approvals are required prior to allowing job to begin.

D. Since the valve is outside the crane wall, only normal OAP-007 SM authorization is required prior to allowing job to begin.

Answer: C

ExplanationlJustification:

A. Incorrect but plausible since there is guidance in OAP-007 to request Operations to consider reducing power for some high dose entries.

B. Incorrect but plausible since extra approval is required, just not at this level. Some administrative dose extension waivers do require Site VP approval adding to the plausibility of this choice.

C. Correct answer per OAP-007.

D. Incorrect but plausible since this would be true if the valve were outside the crane wall. It is plausible that the valve would be there since the orifice valves are outside the crane wall.

Technical

References:

OAP-007 Proposed References to be provided: None Learning Objective IOLP-ILO-ADM01 4 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 9 Comments

Exam Outline Cross

Reference:

Level RO 5RO Tier# 3 Group#

KlA# 1940012405 Emergency Procedures/Plan Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Importance 3.7 4.3 Question # 74 An AOP is being implemented when it is determined that a reactor trip and 51 are required. Per OAP-015, AOP and ONOP User's Guide, what is the earliest time the CR5 can direct that actions in the AOP be resumed (assuming these actions will not detract from performance of the EOP)?

A. Only when called for by the EOP in use.

B. When E-O has been exited.

C. When 51 is reset.

D. After completion of step 4 of E-O.

Answer: D Explanation/Justification:

Per OAP-015, step 4.1.18, actions can be taken after completion of step 4 of E-O.

A. Plausible since actions of the EOP do take precedence over AOP actions.

B. Plausible since early diagnostic steps are important and it reasonable to assume these will be done before taking AOP actions.

C. Plausible since prior to 51 reset equipment may not start or stop as desired.

D. Correct answer

Technical

References:

OAP-015 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPROU 19 Question Source: New Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KfA# 1940012411 Emergency Procedures/Plan Knowledge of abnormal condition procedures.

Importance 4 4.2 Question # 75 Security has just confirmed that a Security CODE RED exists. Armed intruders have gained entry into the protected area.

Which one of the following states the correct actions in the proper order?

A. Trip the reactor, ensure AFW pumps start, perform E-O in parallel with AOP-SEC-1 B. Start AFW Pumps, trip the reactor, perform E-O in parallel with AOP-SEC-1 C. Perform a normal plant shutdown in parallel with AOP-SEC-1 D. Maintain plant in a stable condition while performing AOP-SEC-1 Answer: B Explanation/Justification:

A. Incorrect because AFW is started before trip. Plausible that this would be the order of operations.

B. Correct answer per AOP-SEC-1 C. Incorrect because trip is directed. Plausible to assume that a shutdown would be performed.

D. Incorrect because trip is directed. Plausible since this is the action for an Orange Path and it is reasonable that the plant would not be put through a transient in these conditions.

Technical

References:

0-AOP-SEC-1 Proposed References to be provided: None Learning Objective 10LP-ILO-AOPSEC 10 Question Source: Modified Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.41 (b) 10 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000015A208 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions: - When to secure RCPs on high bearing temperature Importance 3.4 3.5 Question # 76 Given:

• The plant is operating at 100% power

• 34 RCP Shaft Vibration is 16 mils and oscillating +/- 0.6 mils.

• 34 RCP Upper Bearing temperature is 1850 and increasing at approximately 10 Ihr for the last 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

• The cause of the vibration and temperature increases is under investigation

• The team has initiated a plant shutdown Which of the following describes the operating strategy for 3-AOP-RCP-1 for the above conditions?

A. Perform a normal shutdown (POP-2.1 Operation at Greater than 45%

Power; POP-3.1 Plant Shutdown from 45% Power) and secure 34 RCP prior to reaching high vibrations limit B. Perform a rapid (ONOP-TG-3 Rapid Shutdown) shutdown and secure 34 RCP prior to reaching high vibration limit.

C. Perform a normal shutdown (POP-2.1 Operation at Greater than 45%

Power; POP-3.1 Plant Shutdown from 45% Power) and secure 34 RCP prior to reaching high bearing temperature limit.

D. Perform a rapid (ONOP-TG-3 Rapid Shutdown) shutdown and secure 34 RCP prior to reaching high bearing temperature limit

Answer: C ExplanationlJustification:

A. Incorrect. Plausible because a normal shutdown is performed if vibrations are greater than 15 and increasing such that they would reach the trip setpoint in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. Incorrect. Plausible because vibrations are above the alert limit however they are not increasing.

C. Correct. If bearing temperature reaches 185°Fand continues to increase, a shutdown is required in the procedure.

D. Incorrect. Plausible because the unit/pump could be shutdown due to high bearing temperature; however, we do not use ONOP-TG-3 for the shutdown.

Technical

References:

3-AOP-RCP-001 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPRCP C 13LP-ILO-AOPRCP H Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000022A204 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

- How long PZR level can be maintained within limits Importance 2.9 3.8 Question # 77 The plant is in a normal 100% alignment with 31 Charging Pump out of service for a major overhaul. While expanding the tagging boundaries, a mistake occurred which rendered the suction path to all charging pumps unavailable due to a gas buildup. Operators responded by securing charging and letdown and entered AOP-CVCS-1, Chemical and Volume Control System Malfunctions.

Restoration of the suction path is expected to take two hours.

Based on these conditions, what will the mitigating strategy of AOP-CVCS-1 be?

A. Pressurizer level will lower to less than allowed for heater operation in the time that repairs are expected to take. Therefore, AOP-CVCS-1 directs a normal shutdown for this condition.

B. Pressurizer level will lower to less than allowed for heater operation in the time that repairs are expected to take. Therefore, the AOP directs a shutdown using ONOP-TG-3, Rapid Shutdown.

C. Pressurizer level will NOT lower to less than allowed for heater operation in the time that repairs are expected to take. However, the AOP directs a shutdown using ONOP-TG-3, Rapid Shutdown whenever charging flow cannot be immediately re-established.

D. Pressurizer level will NOT lower to less than allowed for heater operation in the time that repairs are expected to take. Therefore, AOP-CVCS-1 allows for continued operation until charging is re-established.

Answer: 0

Explanation/Justification:

The KA is to determine how rapidly PZR level will drop without charging.

Differentiating AlB vs. C/D satisfies the KA. Differentiating C vs. D satisfies the requirement for an SRO question because it requires knowledge of requirements of the AOP well beyond initial actions.

If a candidate believes the pressurizer will lower below the heater trip level in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (it will actually take 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) than either A or B is plausible since both shutdown procedures are specified throughout other AOPs that direct shutdowns.

Choice C is plausible because it would be undesirable to operate with RCS water cooling No. 1 Seal vs. filter charging water.

Choice D is correct since the AOP never directs a shutdown or power reduction for this condition.

Technical

References:

3-AOP-CVCS-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPCVC 2.0 Question Source: New Question History: None Question Cognitive Level: Analysis Synthesis Evaluation 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 0000292123 Conduct of Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance 4.3 4.4 Question # 78 Given the following:

• The crew is responding to an A TWS.

• FR-S.1, Response to Nuclear Power Generation/ATWS has been entered and the Immediate Actions are complete.

• The reactor is tripped and all control rods are inserted.

• Neutron flux is decreasing in the Intermediate Range.

• Safety Injection is NOT actuated.

Which of the following identifies the condition concerning boration as addressed in FR-S.1 based on the above indications?

A. With the reactor tripped and neutron flux decreasing, boration is NOT initiated in FR-S.1; additional boration may be required by subsequent procedure in effect.

B. Boration must be initiated in FR-S.1, but may be terminated in FR-S.1 once the sources of positive reactivity addition to the RCS have been eliminated.

C. Boration must be initiated in FR-S.1, and the Cold Shutdown boron concentration is NOT required to exit FR-S.1 D. Boration must be initiated in FR-S.1, and the Cold Shutdown boron concentration IS required to exit FR-S.1.

Answer: C Explanation/Justification:

A. Incorrect. Plausible because the reactor is subcritical, and in general actions of FRPs are completed prior to exit.

B. Incorrect: Plausible because "emergency boration" is secured when sources of positive reactivity are eliminated; normal boration is initiated to achieve cold shutdown concentration prior to exiting FR-S.1.

C. Correct: Normal boration is initiated to achieve cold shutdown concentration prior to exiting FR-S.1.

D. Incorrect. Plausible because "emergency boration" is secured when reactor is verified subcritical; normal boration is initiated to achieve cold shutdown concentration prior to exiting FR-S.1.

Technical

References:

2-FR-S.1 3-FR-S.1 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRS 11 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 0000382411 Emergency Procedures/Plan Knowledge of abnormal condition procedures.

Importance 4 4.2 Question # 79 The following conditions exist at Unit 3:

• Plant was operating at 100% when a 5 gpm SGTL developed on 34 SG.

• The plant was shutdown in response to this condition.

• Plant cooldown is in progress with RCS temperature at 499°F.

• 34 SG tube leakage has just increased to 275 gpm.

Which of the following describes the correct procedure in place at this time and the appropriate actions?

A. Cooldown is being performed per AOP-SG-1, Steam Generator Tube Leak. Initiate SI and go to E-O, Reactor Trip or Safety Injection.

B. Cooldown is being performed per AOP-SG-1, Steam Generator Tube Leak. Start SI Pumps and go to E-3, Steam Generator Tube Rupture.

C. Cooldown is being performed per POP-3.3, Plant Cooldown - Hot to Cold Shutdown. Initiate SI and go to E-O, Reactor Trip or Safety Injection.

D. Cooldown is being performed per POP-3.3, Plant Cooldown - Hot to Cold Shutdown. Start SI Pumps and go to E-3, Steam Generator Tube Rupture.

Answer: A Explanation/Justification:

AOP-SG-1 governs plant cooldown to Mode 5, but it is completely plausible that the cooldown would be performed per the POP. AOP-SG-1 speci'fies initiating SI and going to E-O despite initial plant mode if PZR level cannot be maintained. Going directly to E-3 is plausible because it would actually work to

mitigate the event in a timely manner. It is also plausible that a candidate may misunderstand that SI never was actuated in the AOP-SG-1 response.

Technical

References:

3-AOP-SG-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPSG1 C Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 000058A203 Ability to determine and interpret the following as they apply to the Loss of DC Power: - DC loads lost; impact on to operate and monitor plant systems Importance 3.5 3.9 Question # 80 The plant is in a 100% normal lineup when 480V Bus 5A is lost due to a short that causes severe bus damage. Which of the following statements describes how DC loads will be addressed for this event?

A. Alternate power options to 31 DC Power Panel are not allowed for this condition prior to the plant being placed in Mode 5. During the shutdown/cooldown, 31 DC Power Panel will be lost when 31 Battery depletes, however 31 Instrument Bus will remain energized.

B. Alternate power options to 31 DC Power Panel are not allowed for this condition prior to the plant being placed in Mode 5. During the shutdown/cooldown, 31 DC Power Panel and 31 Instrument Bus will be lost when 31 Battery depletes.

C. 3-AOP-DC-1, Loss of Power to a 125V DC Panel, will cross connect 31 and 32 DC Power Panels to maintain power to 31 DC Power Panel during the shutdown/cooldown.

D. 3-AOP-DC-1, Loss of Power to a 125V DC Panel, will align 35 Battery Charger to maintain power to 31 DC Power Panel during the shutdown/cooldown.

Answer: A Explanation/Justification:

A. Correct answer. Above Mode 5, the DC and 480V safeguard busses cannot be cross-connected. When the battery depletes, 31 IB would automatically

transfer to an alternate supply from another 480V bus. Also, the 3-AOP-480V-1 will manually do this before the battery depletes.

B. Incorrent but plausible. Incorrect because the 31 IB will not be lost. Plausible because the candidate may not realize that the power supply for reverse power of the IB Static Inverter comes off a different bus. This answer would actually be correct for IP-2.

C. Incorrect but plausible. Incorrect because this is not allowed as discussed in A. Plausible because the system could physically do this.

D. Incorrect but plausible. Incorrect because this is not allowed as discussed in A. Plausible because the system could physically do this.

Technical

References:

3-AOP-480V-1 3-AOP-DC-1 3-S0P-EL-003 Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-AOP480 4.0 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 1 KlA# 0000252409 Emergency Procedures/Plan Knowledge of low power /

shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

Importance 3.8 4.2 Question # 81 INITIAL PLANT CONDITIONS:

• Plant is in MODE 5 on the 31 Train of RHR

• 480V Bus 6A Electrical Distribution Train is de-energized for maintenance, and cannot be immediately restored

• Vacuum refill of the RCS has been completed.

• RCS temperature is 180°F

• RCS pressure is 175 psi

• All Steam Generators are in Wet-Layup

• No RCPs are running 31 RHR Pump fails due to a motor bearing failure.

Which of the following actions are required in accordance with AOP-RHR-1, Loss of RHR?

A. Establish Natural Circulation; control SG level and open the steam generator atmospheric dump valves.

B. Establish bleed and feed; open both PORVs, fill the RCS using one SI Pump from the RWST.

C. Establish bleed and feed; open both PORVs, fill the RCS using one Charging Pump from the RWST.

D. Start one Reactor Coolant Pump, and open the steam generator atmospheric dump valves.

Answer: A Explanation/Justification:

A. Correct. The SGs are coupled and can be used as a heat sink since vacuum refill is completed B. Incorrect. Plausible because these actions are directed if RCS level is < 62' 6" and a more rapid increase in RCS level is desired.

C. Incorrect. Plausible because these actions are directed if RCS level is <62'6" and charging from 2 pumps can adequately increase RCS level.

D. Incorrect. Plausible because the procedure directs these actions if conditions exist that support starting an RCP. Pressure at 200 psig is too low.

Technical

References:

3-AOP-RHR-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPRHR 2 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 0000032244 Equipment Control - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Importance 4.2 4.4 Question # 82 Given the following plant conditions:

• One Shutdown Bank-A Group 2 Rod has dropped into the core.

• The crew is recovering the dropped rod.

• ROD CONTROL URGENT FAILURE, alarm is received.

Which ONE (1) of the following describes the reason for the alarm and the expected response?

A. The alarm is NOT expected for the recovery of a shutdown bank rod.

Actions to Correct Stepping Sequence will be initiated per SOP-RC-1, Full Length Rod Control and RPI System Operation, prior to proceeding with rod recovery.

B. The alarm is NOT expected for the recovery of a shutdown bank rod.

Actions to reset the Master Cycler will be initiated per the contingency attachment of AOP-ROD-1, Rod Control and Indication System Malfunction prior to proceeding with rod recovery.

C. The alarm is expected due to Shutdown Bank-A Group 1 rods not moving.

Prior to proceeding, the AOP requires verifying the alarm at the appropriate power cabinet.

D. The alarm is expected due to Shutdown Bank-A Group 1 rods not moving.

Alarm is anticipated by AOP and recovery may continue with no further action.

Answer: D Explanation!Justification:

A similar question is in our Unit 2 bank that only tests system knowledge.

Question was modified to make it an SRO level question and improved validity of distractors. Question has not been used on recent IPEC NRC exam.

A. Incorrect (the alarm is expected) but plausible because this statement would be true for a SD Bank C or D rod. SOP-RC-1 does have a section for Correcting Stepping Sequence B. Incorrect (the alarm is expected) but plausible because this statement would be true for a SD Bank C or D rod, and contingency attachments are contained in reactor startup procedures.

C. Incorrect but plausible because a candidate could think the unaffected rod positions are checked.

D. Correct answer. Knowledge of steps beyond initial actions is required to confidently answer this question.

Technical

References:

3-AOP-ROD-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPROD 9 Question Source: Modified Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 K1A# 0OOO37A216 Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:

Pressure at which to maintain RCS during SIG cooldown Importance 4.1 4.3 Question # 83 The plant is being cooled down following a SGTR using blowdown. Given the following plant conditions:

• Pressurizer Level is 53% and increasing

• Ruptured SG NR Level is 35% and decreasing

• No other event is in progress

• Shutdown margin is adequate What action is required per E-3, Steam Generator Tube Rupture, based on these conditions?

A. Increase RCS Makeup Flow B. Turn ON Pressurizer Heaters C. Depressurize the RCS using normal spray D. Increase cooldown rate using SG atmospherics Answer: B ExplanationlJustification:

A. Incorrect. Plausible because increasing makeup would slowly raise RCS pressure stopping the backfill and this would be used if Pressurizer level was low.

B. Correct.

C. Incorrect. Plausible because this action is used if SG level is increasing and Pressurizer Level is decreasing.

D. Incorrect. Plausible because this action will cause a Decrease in RCS Pressure not an Increase.

Technical

References:

2-ES-3.2 3-ES-3.2 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE30 16 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KlA# 00WE032237 Equipment Control - Ability to determine operability and/or availability of safety related equipment.

Importance 3.6 4.6 Question # 84 IPEC Unit 3 was at 100% power with 33 EDG tagged out for maintenance when a small break loss of coolant accident occurred coincident with a loss of offsite power. The team is in ES-1.2, Post LOCA Cooldown and Depressurization, and has just secured one HHSI pump. Which of the following describes how accumulators will be addressed?

A. Accumulators will be left in service until all HHSI pumps are secured.

8. All accumulators will be vented.

C. SI-MOV-894B and SI-MOV-894D will be closed, 31 and 33 Accumulators will be vented.

D. SI-MOV-894A and SI-MOV-894C will be closed, 32 and 34 Accumulators wi II be vented.

Answer: C Explanation/Justification:

A. Incorrect but plausible. Accumulators are isolated with HHSI pumps still in service, but a candidate could believe they are kept in service until break flow has been reduced enough to secure all HHSI pumps.

B. Incorrect but plausible. The procedure could direct venting all accumulators. In fact 33 is always vented for LOCAs due to valve height.

C. Correct. 894B and D will have power and be energized and closed. 894A and C will not have power due to 33 EDG being unavailable.

D. Incorrect but plausible. Plausible if a candidate does not know power supply for the valves. Also, 894C is not closed for LOCAs, but a candidate could believe that for a SBLOCA it is acceptable to close the valve since VC water level should be less than 894C level if SI can be reduced.

Technical

References:

3-ES-1.2 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE10 17 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 1 Group# 2 KJA# 00WE10A201 Ability to determine and interpret the following as they apply to the Natural Circulation with Steam Void in Vessel with/without RVLlS: - Facility conditions and selection of appropriate procedures during abnormal and emergency operations Importance 3.2 3.9 Question # 85 Given the following conditions:

• A reactor trip has occurred concurrent with a loss of offsite power.

• The Team is performing actions of ES-0.2, "Natural Circulation Cooldown".

• Train "A" of RVLlS is out of service.

• The Team has commenced RCS cooldown and depressurization.

• RCS pressure is 1380 psig and trending DOWN.

• Qualified CETs are 468°F and trending DOWN.

• Pressurizer level is 38% and trending UP slowly.

Due to secondary inventory concerns, RCS cooldown rate MUST be performed at approximately 40°F/Hr.

Which one of the following actions will be required?

A. Repressurize the RCS to minimize void growth in accordance with ES-0.2.

B. Stop the depressurization to reestablish subcooling in accordance with ES-0.2.

C. Transition to ES-0.3, "Natural Circulation Cooldown With Steam Void In Vessel (With RVLlS)".

D. Transition to ES-O.4, "Natural Circulation Cooldown With Steam Void In Vessel (Without RVLlS)".

Answer: C Explanation!Justification:

A. Incorrect. Plausible because candidate may believe that increasing pressure will allow a larger cooldown rate.

B. Incorrect. Plausible because candidate may believe that increased subcooling may allow a larger cooldown rate.

C. Correct. Only one train of RVLlS is necessary to continue in ES-0.3.

D. Incorrect. Plausible because candidate may believe that both trains of RVLlS are required to remain in ES-0.3.

Technical

References:

2-ES-0.3 2-ES-0.3 BG 3-ES-0.3 Proposed References to be provided: None Learning Objective 12LP-ILO-EOPS03 2 13LP-ILO-EOPEOO 8 Question Source: Bank Question History: IPEC Unit 3 NRC 2006 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 0610002132 Conduct of Operations - Ability to explain and apply all system limits and precautions.

Importance 3.8 4 Question # 86 Given:

• The plant is at 50% power

• N2 MAKE UP INST. AIR LOW PRESS. AUX. B.F.P. ROOM has annunciated

• Nitrogen Pressure is 1837 psig.

The ARP directs a dedicated operator be stationed immediately. What is the reason for stationing a dedicated operator?

A. The operator is stationed near the SG Atmospherics. Backup nitrogen must be lined up to the Atmospheric Steam Dumps in the event of a loss of instrument air.

B. The operator is stationed near 1310AfB. Backup nitrogen bottles must be lined up to 131 OAfB to maintain the valves open in the event of a loss of instrument air.

C. The operator is stationed near the MSIVs. Local manual action is required to maintain the MSIVs open in the event of a loss of instrument air.

D. The operator is stationed near the Aux Boiler Feed Pumps. Local manual action is required to prevent a pump runout condition in the event of a loss of instrument air.

Answer: D Explanation/Justification:

A. Incorrect. Plausible because a local N2 bottle must be lined up to operate the Atmospheric Steam Dumps; however this is not an immediate concern.

B. Incorrect. Plausible because 1310AlB have a backup supply from the Nitrogen bottle bank; and the AFW system is the concern for this action; howver the valves fail open.

C. Incorrect. Plausible because instrument air holds the MSIVs open against spring pressure. An accumulator for each valve hold the valve open for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a loss of instrument air.

D. Correct. This is an "Operable Compensatory Measure". When instrument air and nitrogen backup are lost, the motor driven pump runout protection is lost.

Technical

References:

3-ARP-006 Proposed References to be provided: None Learning Objective 13LP-ILO-AFW001 9 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 0070002238 Equipment Control* Knowledge of conditions and limitations in the facility license.

Importance 3.6 4.5 Question # 87 The plant is cooling down in preparation for refueling. RCS Temperature is 250 0F and lowering as per the cooldown schedule with all systems in the expected configuration. All actions to make the Low Temperature Over Pressure System (LTOPS) operable were completed when directed by the POP. Personnel performing work on top of the pressurizer inadvertently damage the nitrogen line for PORV PCV455C rendering it inoperable. Which of the following describes the required action regarding LTOPS?

A. No action is required since ONLY one PORV is needed to satisfy the LTOPS LCO.

B. Restore PCV-455C to operable status or place the plant in a condition that does not require LTOPS within 7 days.

C. Immediately place ALL HHSI pump switches to trip-pull-out.

D. Immediately close and de-energize ALL accumulator isolation valves.

Answer: B Explanation/J ustification:

A. Incorrect but plausible because the allowed outage time is 7 days, the plant will probably be brought to a condition where LTOPS is not required before expiration of the AOT.

B. Correct answer per T.S. 3.4.12 Action 0 C. Incorrect since HHSI pumps are disabled prior to reaching 3300F, but plausible since this action is required per action A.1 of T.S. 3.4.12.

D. Incorrect since this action is most likely taken prior to reaching 3300F, but plausible since this action is required per action B.1 of T.S. 3.4.12.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-RCSPZR 10 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 013000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-LOCA Importance 4.6 4.8 Question # 88 Given:

• A Large Break LOCA has occurred.

• Reactor Trip Breaker A failed to open and could not be locally opened by the NPO.

• All other equipment functioned as designed.

During the performance of RO-1, BOP Operator Actions During the Performance of EOPs, the BOP depresses both SI Reset Pushbuttons.

Which of the following describes the plant response and procedural requirements?

A. Neither Train of SI will reset.

Place the SI Block Key Switches to "Defeat" After 2 minutes has elapsed, depress the SI Reset Pushbuttons again.

B. Train A of SI will not reset Place SI Block Key Switches to "Defeat" Press Pin Resets on SI Master Relays in G racks C. Neither Train of SI will reset.

Place the SI Block Key Switches to "Defeat'

Press Pin Resets on SI Master Relays in G racks D. Train A of SI will not reset Place Train A SI Block Unblock switch to "Block"

After 2 minutes has elapsed, depress the SI Reset Pushbuttons again Answer: B Expla nationlJustification:

A. Incorrect. Plausible because only one train of SI will fail to reset; placing the key switches to Defeat is correct. With the Key switch in defeat and 2 minutes passed, this action would reset SI; however, it is NOT what the procedure directs.

B. Correct. Only one train will fail to reset; however, the procedure directs using the key switches and the pin resets to reset SI.

C. Incorrect. Plausible because one train of SI will fail to reset; placing the key switches to Defeat is correct as is using the pin resets to reset SI D. Incorrect. Plausible because one train will fail to reset; however, the actions to reset are incorrect.

Technical

References:

3-E-O Proposed References to be provided: None Learning Objective 13LP-ILO-EOPEOO 12 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KlA# 062000A216 Ability to (a) predict the impacts of the following malfunctions or operations on the A.C.

Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Degraded system voltages Importance 2.5 2.9 Question # 89 The plant is operating at 100% power with a normal electrical plant alignment and all TS equipment OPERABLE.

A drop in offsite system voltage causes 480V Bus 5A voltage to drop to 418V causing the "480 VOLT SAFEGUARD BUS UNDERVOLTAGE" alarm to annunciate. System voltage is subsequently restored to normal one minute later.

Which of the following describes the effect on Bus 5A and the actions required?

A. Load breakers on 5A for non-essential equipment open, but the bus stays energized. When voltage returns to normal, loads are restored using an AOP.

B. Load breakers on 5A for non-essential equipment open, but the bus stays energized. When voltage returns to normal, loads are restored using SOPs.

C. The normal supply breaker to Bus 5A opens, load breakers for non essential equipment open, and power to the bus is supplied by 33 EDG.

When voltage returns to normal, normal power and loads are restored using an AOP.

D. The normal supply breaker to Bus 5A opens, load breakers for non essential equipment open, and power to the bus is supplied by 33 EDG.

When voltage returns to normal, normal power and loads are restored using SOPs.

Answer: C Explanation/Justification:

Degraded Grid Voltage protection setpoint is set at 425 Volts. If this condition exists for> 36 seconds (no SI signal) or 3 seconds (with SI signal), the normal feeder breaker will open and the EDG will then start on Under Voltage and the breaker will automatically close when conditions are met.

A. Incorrect. It is plausible that a way of protecting vital equipment from low voltage would be to drop non-essential loads to increase bus voltage. However, this is not how the system actually works.

B. Incorrect, plausible for same reasons as A.

C. Correct answer.

D. Incorrect. Since the initiating problem almost instantly rights itself, it is plausible that SOPs would be used vs. the AOP. This is actually the case for service water and CCW malfunctions. In fact, there is adequate SOP guidance to restore the plant to normal configuration. However, entering AOP-480V-1 is the correct required action.

Technical

References:

3-AOP-480V-1 3-ARP-005 Proposed References to be provided: None Learning Objective 13LP-ILO-EDS480 2 13LP-ILO-AOP480 4 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 1 KiA# 073000A202 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure Importance 2.7 3.2 Question # 90 Given the following conditions:

• The plant is operating at 100% power with all systems in normal alignments.

• A liquid release of 31 Monitor Tank is in progress.

• The following annunciators are received in the Control Room o CHANNEL FAILURE-o R-18 LlQ. EFF."

• R-18, Liquid Waste Effluent monitor is alarming with the indication showing off-scale low.

• The liquid release remains in progress.

Which ONE (1) of the following describes effect on the plant and the actions required?

A. The radiation monitor has failed. Request that HP recheck calculations and provide recommendations on action to be taken.

B. The radiation monitor has failed. The release may continue provided 2 independent samples are taken and the activity is verified to be below ODCM limits.

C. The discharge should have automatically stopped. Stop the discharge, direct Chemistry to sample the tank and refer to the ODCM for further actions.

o The discharge should have automatically stopped. Stop the discharge.

The release CANNOT be continued until R-18 is operable.

Answer: C Explanation/Justification:

A. Incorrect but plausible. Plausible since the monitor has failed and although the release is not being monitored, the tank was sampled.

B. Incorrect but plausible. The release must be terminated. This is plausible since the release can go on after these actions are taken.

C. Correct per SOP D. Incorrect but plausible. It is plausible that this monitor has to be operable.

Technical

References:

3-S0P-WDS-014 Proposed References to be provided: None Learning Objective 13LP-ILO-RMSPRM E Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 001000A218 Ability to (a) predict the impacts of the following malfunctions or operations on the CROS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

- Incorrect rod stepping sequence Importance 3.2 3.8 Question # 91 Given:

• The plant was at 95% power with all control systems in automatic.

• Control Bank 0 rods were at 220 steps.

• PT-412A First Stage Pressure Transmitter failed high

• Step counters for Control Bank 0 indicated 236 steps when Rod Control was placed in manual.

• The ATC immediately inserted rods until CBO step counters indicated 220 step.

Which of the following describes the current status of the rod control system and what actions are required to correct this condition?

A. Actual rod position is lower than indicated on the step counters 3-S0P-RC-001 Full Length Rod Control and RPI System Operation will properly restore rod alignment and sequence.

B. Actual rod position is lower than indicated on the IRPls.

3-AOP-ROO-1, Rod Control and Indication System Malfunctions will realign the rods and reset the PIA converter.

C. Actual rod position is higher than indicated on the step counters.

Rods are properly aligned and sequenced, only the Rod Control Logic needs to be reset per 3-S0P-RC-001 Full Length Rod Control and RPI System Operation o Actual rod position is higher than indicated on the IRPls.

3-AOP-ROD-1, Rod Control and Indication System Malfunctions will realign the rods and reset the Bank Overlap Unit.

Answer: A Explanation/Justification:

A. Correct. When the ATC began inserting rods, actual rod height was 231 step with 236 on the step counters. The step counters indicate 5 steps higher than actual rod position. SOP-RC-001 will correct the Stepping Sequence.

B. Incorrect. Plausible because there is a 5 step difference between actual and demand position; however, IRPI will indicate actual rod position. The actions for rod realignment in the AOP are not correct for this condition.

C. Incorrect. Plausible because actual rod position will be lower than indicated on the step counters. In addition, the rods are not properly aligned and sequenced.

D. Incorrect. Plausible because there is a 5 step difference between actual and demand position; however, IRPI will indicate actual rod position. In addition more than just the Bank Overlap Unit needs to be addressed during correcting the stepping sequence.

Technical

References:

3-S0P-RC-001 Proposed References to be provided: None Learning Objective 13LP-ILO-ICROD H Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 0020002142 Conduct of Operations Knowledge of new and spent fuel movement procedures.

Importance 2.5 3.4 Question # 92 Given the following conditions:

• Unit 3 is in a refueling outage

• Core unload is in progress

• The fuel transfer cart is in containment

• RWST Level is 1 ft

• Mansell Level Indication begins lowering rapidly

• Reactor Sump Level begins increasing rapidly Which one of the following statements describes the mitigating actions for this event per AOP-FH-1, Fuel Damage of Loss of SFP/Refueling Cavity Level?

A. Ensure the fuel transfer gate valve is closed to isolate the SFP from the Reactor Cavity. Align RHR to take suction on the Recirc Sump Level when level is above 48' 2".

B. Ensure the fuel transfer gate valve is closed to isolate the SFP from the Reactor Cavity. Align a Recirc Pump to take suction on the Recirc Sump Level when level is above 48' 2".

C. Ensure the fuel transfer gate valve is open to provide increased inventory while establishing recirculation flow. Align RHR to take suction on the Recirc Sump Level when level is above 48' 2".

D. Ensure the fuel transfer gate valve is open to provide increased inventory while establishing recirculation flow. Align a Recirc Pump to take suction on the Recirc Sump Level when level is above 48' 2".

Answer: B

Exp lanation/J ustification:

A. Incorrect but plausible if operator believes RHR can be aligned to the Recirc Sump vs. Containment Sump.

B. Correct answer per 3-AOP-FH-1 C. Incorrect but plausible if operator believes RHR can be aligned to the Recirc Sump vs. Containment Sump. Also plausible if the operator believes it is better have more water in containment at the expense of losing SFP level.

D. Incorrect but plausible if the operator believes it is better have more water in containment at the expense of losing SFP level.

Technical

References:

3-AOP-FH-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPFH1 E Question Source: Modified Question History: MCGuire 2005 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 2 Group# 2 KlA# 034000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Dropped fuel element Importance 3.6 4.4 Question # 93 Core off-load is in progress when you are notified by the Fuel Handling Supervisor that an irradiated fuel assembly was dropped in the spent fuel pool.

The assembly fell into the correct pool location. R-5 and R-27 readings are increasing. R-5 is in ALARM.

Based on these conditions, what actions are required after suspending fuel handling operations in the FSB by 3-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level?

A. Evacuate ONLY non-essential personnel from the FSB, place FSB ventilation in service and monitor R-27.

B. Evacuate ONLY non-essential personnel from the FSB, ensure SFP purification is in service with maximum flow and monitor R-27.

C. Evacuate ALL personnel from the FSB, ensure ALL FSB doors are closed and monitor R-27.

D. Evacuate ALL personnel from the FSB, ensure SFP purification is in service with maximum flow and monitor R-27.

Answer: C ExplanationlJustification:

A. Incorrect but plausible. All personnel are evacuated and the procedure does not direct placing ventilation in service. However only evacuating non essential personnel and placing ventilation in service are plausible.

B. Incorrect but plausible. Maximizing purification is plausible, but not in the procedure.

C. Correct answer per AOP-FH-1 D. Incorrect but plausible. Plausible for same reason as B.

Original Question Indian Point 2 12/9/2004 The following conditions exist:

"There is a core off-load in progress.

"The fuel handler was moving irradiated fuel to a location in the spent fuel pool.

"You are notified that the spent fuel bundle was accidentally dropped in the spent fuel pool.

liThe fuel handler reports the fuel bundle fell into the correct pool location.

"R44, Plant Vent radiation monitor reads 4E-3 :Ci/cc and is steady.

What actions, if any, are required?

A.Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level. Evacuate non-essential personnel from the FSB, place FSB ventilation in service and monitor R44.

REnter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level. Suspend ALL fuel handling operations in FSB, Evacuate non-essential personnel from the FSB, secure FSB ventilation and monitor R44.

C.Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level. Suspend ALL fuel handling operations in FSB, Evacuate ALL personnel from the FSB, dispatch an operator to close ALL FSB doors and monitor R44.

D.Enter procedure 2-AOP-FH-1, Fuel Damage or Loss of SFP/Refueling Cavity Level. Suspend ALL fuel handling operations in FSB, Evacuate ALL personnel from the FSB and dispatch an operator to independently verify proper fuel bundle location and monitor R44.

Answer C Technical

References:

3-AOP-FH-1 Proposed References to be provided: None

Learning Objective 13LP-ILO-FHD001 M Question Source: Bank Question History: Indian Point 2 Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012140 Conduct of Operations Knowledge of refueling administrative requirements.

Importance 2.8 3.9 Question # 94 The following conditions exist at Unit 3:

• Core offload is in progress.

• Nuclear Source Range N-31 is selected as the input for the Audio Count Rate Drawer.

• N-38 and N-39 are out of service due to outage electrical work, but are expected to be returned to service in 30 minutes.

• N-31 has just become inoperable.

Which ONE of the following describes the proper action for this condition?

A. Core alterations may continue for up to one hour provided the Audio Count Rate Drawer is switched to N-32 within this time.

B. Immediately stop any core alterations. Core alterations may continue when N-38 or N-39 is placed in service and the Audio Count Rate Drawer is switched to an operable channel.

C. Immediately stop any core alterations. Core alterations may continue when N-31 is restored to operable status. N-38 or N-39 cannot be used to satisfy requirements for core alterations.

D. Core alterations may continue for up to one hour provided N-38 or N-39 is placed in service and the Audio Count Rate Drawer is switched to an operable channel within this time.

Answer: B Explanation/J ustification:

A. Incorrect but plausible. One instrument could be considered sufficient provided audio count rate indication is available. The one hour allowance is plausible since this provision exists for RHR flow.

B. Correct answer.

C. Incorrect but plausible because it is possible that N-38/39 would not be allowed per Tech Specs D. Incorrect but plausible since the one hour provision is allowed in RHR specs.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-FHD001 N Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KJA# 1940012145 Cond uct of Operations - Ability to identify and interpret diverse indications to validate the response of another indication.

Importance 4.3 4.3 Question # 95 While transferring 31 Static Inverter from Reverse Transfer to Forward Transfer following a surveillance test, 31 Main Boiler Feed Pump Speed Control Trouble alarm annunciated. The flight panel indications for 31 and 32 MBFPs are:

Indication 31 MBFP 32 MBFP Turbine RPM 4726 4706 Percent Startup 94.8 94 Percent Feedwater 67.9 65.2 Percent Hold 66 Blank Failure Hold LED ON OFF The I&C technician states the cause of the alarm was due to a momentary voltage spike on 31 Instrument Bus.

How is the voltage spike theory confirmed, and what actions are required?

A. %FW signal for 31 MBFP is within 4% of %FW for 32 MBFP, Press Reset button on Flight Panel per ARP for 31 MBFP Speed Control Trouble B. %FW for 31 MBFP is within 4% of % Hold signal, Press Reset button on Flight Panel per AOP-FW-1 Loss of Feedwater C. %FW signal for 31 MBFP is within 4% of %FW for 32 MBFP, Press Reset button on Flight Panel per AOP-FW-1 Loss of Feedwater D. If %FW for 31 MBFP is within 4% of % Hold signal, Press Reset button on Flight Panel per ARP for 31 MBFP Speed Control Trouble

Answer: B Explanation/Justification:

A. Incorrect. Plausible because the procedure will direct resetting Track & Hold if the %FW signal and the %Hold signal are within 4% of each other.

B. Correct.

C. Incorrect. Incorrect. Plausible because the procedure will direct resetting Track & Hold if the %FW signal and the %Hold sjgnal are within 4% of each other. Also resetting Track and Hold is accomplished using the Flight Panel button per the AOP.

D. Incorrect. Plausible because the %FW signal and %Hold signals within 4% of each other does confirm the instrument bus spike theory. Also Track and Hold is reset using the flight panel button; however this is directed in the AOP not the ARP Technical

References:

3-AOP-FW-1 Proposed References to be provided: None Learning Objective 13LP-ILO-AOPFW1 C Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012222 Equipment Control - Knowledge of limiting conditions for operations and safety limits.

Importance 4 4.7 Question # 96 On 8/24/12 at 0800, the plant commenced a reduction in power from 100% to just below the C-20 (AMSAC) setpoint due to AMSAC being inoperable and its LCO condition expiring. Reactor power was below 50% at 1100 and the load reduction was completed at 1230. During and after the load reduction, there were problems keeping L\I in the target band Out of In Out of In Quadran Band Band Band Band t

41 1020 1210 42 1015 1105 43 1030 1050 1130 1230 44 1055 1135 TIme Lme:

41 I----l~::;::::::.;;::::::::+::~:::::::::;:::;::;::::::::;:.;:~----I

~o,ll I ;*J~3-o--~--+---~--r---~~-~-~112LI Assuming AMSAC gets repaired, what is the earliest time reactor power can be raised above 50%?

A. 1110 on 8/25/12 B. 1035 on 8/25/12 C. 1030 on 8/25/12 D. 1020 on 8/25/12

Answer: B ExplanationlJustification:

Penalty minutes start at 10:20 (when 2nd channel exceeded the allowable limit).

40 penalty minutes were accumulated at 1 for 1 up to 11 :00.

35 penalty minutes were accumulated at .5 for 1 from 11 :00 to 12: 10 75 penalty minutes total.

Penalty minutes start to be removed at 10:20 on 8/25/12.

At 10:35 penalty minutes equal 60 in the last 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and power can be increased above 50%

A. Incorrect.

B. Correct C. Incorrect D. Incorrect Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 13LP-ILO-ICEXC 10 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 0560002225 Equipment Control- Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Importance 3.2 4.2 Question # 97 What is the basis for requiring a minimum of 360,000 gallons of water in the Condensate Storage Tank?

Ensures adequate volume to:

A. maintain Hot Zero Power for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and then cooldown to MODE 4

<350°F.

B. maintain MODE 4 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with at minimum of 2 SGs greater than 9%

NR level.

C. Cooldown from Trip concurrent with loss of off-site power to MODE 4

<350°F on RHR.

D. remove decay heat while in MODE 3 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a reactor trip from 102% RTP.

Answer: D ExplanationlJustification:

To satisfy accident analysis assumptions, the CST must contain sufficient cooling water to remove decay heat while in MODE 3 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a reactor trip from 102% RTP.

A. Incorrect. Plausible because the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time is correct; however, the action and plant conditions are not correct

B. Incorrect. Plausible because the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time is correct; however, the action and plant conditions are not correct C. Incorrect. Plausible because the cooldown with LOOP will be performed using the Atmospheric steam dumps. The CST inventory will be consumed.

D. Correct.

Technical

References:

Tech Specs Proposed References to be provided: None Learning Objective 12LP-ILO-CND01 3 13LP-ILO-CND001 9 Question Source: Modified Question History: None Question Cognitive Level: Fundamental Knowledge 10 CRF Part 55 Content: 55.43 (b) 2 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012305 Radiological Controls - Ability to use radiation monitoring systems, such as fixesd radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance 2.9 2.9 Question # 98 Given the following:

• A LBLOCA occurred 40 minutes ago

• Internal recirculation and spray have been established

• No equipment was out of service, and all equipment started as required

• Containment temperature, pressure, and sump levels are as expected for these conditions

• R-25/26 are noted to be reading 20 R/hr and stable In addition to possible discretionary entry into FR-Z.3, Response to High Containment Radiation Level, which of the following describes the required actions due to R-25/26 readings?

A. Remain in current EOP. Declare a General Emergency based on these radiation monitor readings.

B. Transition to ECA-1.1, Loss of Emergency Coolant Recirculation. Declare a General Emergency based on these radiation monitor readings.

C. Remain in current EOP. Declare a Site Area Emergency based on these radiation monitor readings.

D. Transition to ECA-1.1, Loss of Emergency Coolant Recirculation. Declare a Site Area Emergency based on these radiation monitor readings.

Answer: C Explanation/Justification:

The correct answer is C. 17 R/hr is the SAE reading (unless containment was failed) and 68 R/hr reading. It is very plausible that a candidate would confuse these numbers. There is no required transition out of the current EOP (ES-1.3) due to high radiation. The statement in the stem about FR-Z.3 is to prevent a contention. FR-Z.3 is a yellow path procedure, so entry is discretionary. It is plausible that a candidate may think ECA-1.1 entry would be required because there is something causing the high radiation levels. However this is not the case.

Technical

References:

IP-EP-120 Proposed References to be provided: None Learning Objective IOLP-ILO-ERT0025 Question Source: New Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.41 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

K1A# 1940012422 Emergency Procedures/Plan Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

Importance 3.6 4.4 Question # 99 The STA has advised the CRS that an ORANGE path exists for CORE COOLING and no other higher ORANGE or RED paths exist.

While depressurizing the SGs to 175 psig lAW 3-FR-C.2, "Response To Degraded Core Cooling", the STA reports that you are now in a RED path on the INTEGRITY CSF.

Which of the following describe the appropriate action to be taken?

A. Complete the depressurization of all SGs to 175 psig, then transition to FR P.1, "Response to Imminent Pressurized Thermal Shock".

B. Immediately transition to FR-P.1, "Response to Imminent Pressurized Thermal Shock".

C. Complete FR-C.2, "Response to Degraded Core Cooling", then transition to FR-P.1, "Response To Imminent Pressurized Thermal Shock".

D. Continue with FR-C.2, "Response To Degraded Core Cooling", and concurrently perform actions in in FR-P.1, "Response To Imminent Pressurized Thermal Shock", that do not conflict with FR-C.2 Answer: C Explanation/Justification:

A. Incorrect. Plausible because a CAUTION directs the operators to "complete the procedure". The candidates may believe they should complete the STEP then transition.

B. Incorrect. Plausible because in general when an ORANGE path procedure is in progress and a RED path condition occurs, the ORANGE procedure is suspended and the RED is implemented unless the procedure specifically states otherwise.

C. Correct. The CAUTION prior to the SG depressurization step specifically sates to complete FR-C.2 before a transition to FR-P.1.

D. Incorrect. Plausible because some procedures (e.g. AOPs) can be performed concurrently with EOPs as long as the actions do not conflict with the EOP in progress.

Technical

References:

3-FR-C.2 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPFRC 12 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

Exam Outline Cross

Reference:

Level RO SRO Tier# 3 Group#

KlA# 1940012423 Emergency Procedures/Plan Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Importance 3.4 4.4 Question # 100 The plant is responding to a large break LOCA. The operators have just transitioned to ES-1.3, "Transfer to Cold Leg Recirculation" on RWST Low-Low level. SI has been reset.

While aligning the plant for cold leg recirculation, containment pressure is noted to be 14 psig and rising slowly.

If pressure continues to rise to 22 psig, the team should ...

A. Exit ES-1.3 at step in progress and transition to FR-Z.1, "Response to Containment High Pressure".

B. Complete ALL steps in ES-1.3 prior to transitioning to FR-Z.1, "Response to Containment High Pressure".

C. Complete steps in ES-1.3 to establish cold leg recirculation prior to transitioning to FR-Z.1, "Response to Containment High Pressure".

D. Remain in ES-1.3. Functional Restoration Procedures are not implemented once ES-1.3 has been entered.

Answer: C Explanation/Justification:

A. Incorrect but plausible because in almost all EOPs transition would occur at this point.

B. Incorrect but plausible since you do need to complete all steps to establish cold leg recirculation.

C. Correct per note at beginning of ES-1.3 D. Incorrect but plausible because this is true while recirculation is being established.

Technical

References:

3-ES-1.3 Proposed References to be provided: None Learning Objective 13LP-ILO-EOPE10 24 Question Source: Bank Question History: None Question Cognitive Level: Comprehension 10 CRF Part 55 Content: 55.43 (b) 5 Comments

IPEC Unit 2 NRC Written Exam Answer Key February 20, 2013 1 A 26 C 51 C 76 C 2 B 27 B 52 D 77 D 3 A 28 I C 53 D 78 C I

4 C 29 D 54 C 79 A I

5 B 30 C 55 B 80 A I

6 C 31 A 56 C 81 A 7 D 32 D 57 D 82 D 8 33 58 83 I A B B B 9 C 34 A 59 D 84 C I

10 B 35 A 60 A 85 C 11 C 36 i B 61 A 86 D 12 D 37 I B 62 A 87 B 13 A 38 I D 63 C 88 B 14 B 39 D 64 B 89 C 15 D 40 D 65 B 90 C 16 C 41 A 66 B 91 A 17 B 42 C 67 B 92 B 18 A 43 D 68 C 93 C 19 D 44 D 69 C 94 B 20 A 45 B 70 D 95 B i

21 i C 46 D 71 D 96 B 22 D 47 A 72 B 97 D 23 C 48 B 73 C 98 C 24 49 74 I B A D 99 i C 25 B 50 B 75 B 100 C