ML21020A097

11 Final Written Examination

ML21020A097
Person / Time
Site:	Palo Verde
Issue date:	12/10/2020
From:	Greg Werner Operations Branch IV
To:	Arizona Public Service Co
References
Download: ML21020A097 (438)

Text

Examination Outline Cross-

Reference:

Level RO SRO K/A: Reactor Trip, Stabilization, Recovery: Ability to Tier 1 determine or interpret the following as they apply to a Group 1 reactor trip: Reactor trip breaker position K/A 007 EA2.03 IR 4.2 Question 1 Each EXTINGUISHED phase current light on Control Room Board 5 (B05) indicates a MINIMUM of ___(1)___ RTCB(s) is(are) open, and a MINIMUM of ___(2)___ phase current light(s) must be extinguished in order for the Reactor to trip.

A. (1) 1 (2) 1 B. (1) 1 (2) 2 C. (1) 2 (2) 1 D. (1) 2 (2) 2

Proposed Answer: B Explanations:

A. First part is correct. The second part is plausible because to trip the Reactor a minimum of 2 breakers need to be opened. A combination of A or C AND B or D need to be opened to trip the Reactor. If A and C or B and D were the only 2 breakers open, only 1 phase current light would be extinguished and the Reactor wouldnt trip.

B. Correct C. First part is plausible if it is thought that there will still be a path for current to the phase current light and only by opening both breakers is power completely isolated to the light. The second part is plausible because to trip the Reactor a minimum of 2 breakers need to be opened. A combination of A or C AND B or D need to be opened to trip the Reactor. If A and C or B and D were the only 2 breakers open, only 1 phase current light would be extinguished and the Reactor wouldnt trip.

D. First part is plausible if it is thought that there will still be a path for current to the phase current light and only by opening both breakers is power completely isolated to the light. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: Plant Protection System LP #NKASYC14907 - Describe how RTCBs are tripped and what indication or trip path status is available

Technical

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Operator Information Manual Technical

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Plant Protection System Tech Manual Examination Outline Cross-

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Level RO SRO K/A: Small Break LOCA: Ability to determine or Tier 1 interpret the following as they apply to a small break Group 1 LOCA: Adequate core cooling K/A 009 EA2.39 IR 4.3 Question 2 Given the following conditions:

A LOCA is in progress The CRS has entered 40EP-9EO03, Loss of Coolant Accident Containment Temperature is 150°F Adequate Core cooling is indicated by a MINIMUM subcooling or MAXIMUM superheat of A. 24°F subcooled B. 0°F C. 44°F superheat D. 60°F superheat

Proposed Answer: C Explanations:

A. 24°F subcooled is the minimum subcooled value for adequate core cooling during an uncomplicated Reactor trip B. 0°F is the value in which water changes from a subcooled condition to a superheated condition C. Correct D. 60°F superheat is the value of adequate core cooling when the Containment is in Harsh Condition Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 14 Reference N Provided:

Learning Objective: Given conditions of LOCA, analyze Core Heat Removal to determine if the SFSC acceptance criteria is satisfied per 40EP-9EO03

Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Technical

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40EP-9EO02, Reactor Trip Examination Outline Cross-

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Level RO SRO K/A: Large Break LOCA: Ability to diagnose and Tier 1 recognize trends in an accurate and timely manner Group 1 utilizing the appropriate control room reference material K/A 011 G 2.4.47 IR 4.2 Question 3 Given the following conditions:

Unit 1 tripped due to a large break LOCA The CRS has entered 40EP-9EO03, Loss of Coolant Accident RWT level is 80% and lowering at a rate of 1%/min As RWT level continues to lower, the crew will be procedurally REQUIRED to shift Charging Pump suction to an alternate source in MAXIMUM of A. 7 minutes B. 30 minutes C. 36 minutes D. 46 minutes

Proposed Answer: A Explanations:

A. Correct B. RWT level lowering to 50% will require the crew to stop 1 charging pump C. RWT level lowering to 44% will require the crew to stop all charging pump D. This value correlates to the auto makeup to the VCT setpoint Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 29739 - Using the current copy of the Standard Appendices, perform Charging Pump Alternate Suction to the SFP / Restoration, per 40EP-9EO10, Appendix 11

Technical

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Examination Outline Cross-

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Level RO SRO K/A: Reactor Coolant Pump Malfunctions: Ability to Tier 1 determine and interpret the following as they apply to Group 1 the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on high stator K/A 015 AA2.09 temperatures IR 3.4 Question 4 Given the following conditions:

Unit 3 was tripped due to a high Motor Stator temperature on 1A RCP Per 40AO-9ZZ04, Reactor Coolant Pump Emergencies, the crew should stop

___(1)___. Stopping the RCP(s) should be performed ___(2)___.

A. (1) ALL RCPs (2) BEFORE the Reactivity Control Safety function is addressed B. (1) ALL RCPs (2) AFTER the Reactivity Control Safety function is addressed C. (1) 1A RCP ONLY (2) BEFORE the Reactivity Control Safety function is addressed D. (1) 1A RCP ONLY (2) AFTER the Reactivity Control Safety function is addressed

Proposed Answer: D Explanations:

A. First part is plausible because during malfunctions that cause a loss of cooling water to RCPs, all RCPs are stopped. Second part is plausible if it is thought that stopping the RCP takes precedence over verifying that the Reactor has tripped since there is a high temperature.

B. First part is plausible because during malfunctions that cause a loss of cooling water to RCPs, all RCPs are stopped. Second part is correct.

C. First part is correct. Second part is plausible if it is thought that stopping the RCP takes precedence over verifying that the Reactor has tripped since there is a high temperature.

D. Correct Question Source: New X Bank Modified Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 10 Reference N Provided:

Learning Objective: 18654 - Given that the ORP is being implemented, describe the use of an AOP or OP when the reactor trips or when performing an EOP, in accordance with 40DP-9AP16, EOP Users Guide

Technical

Reference:

Previous Question on 2016 NRC Exam Technical

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40AO-9ZZ04, Reactor Coolant Pump Emergencies Technical

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EOP Operations Expectations Examination Outline Cross-

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Level RO SRO K/A: Loss of Reactor Coolant Makeup: Knowledge of Tier 1 the operational implications of EOP warnings, cautions, Group 1 and notes K/A 022 G 2.4.20 IR 3.8 Question 5 Given the following conditions:

Unit 2 is operating at 100% power A & B Charging Pumps are operating Subsequently:

The OATC recognizes that charging flow has lowered to 25 gpm An Auxiliary Operator reports to the Control Room that A Charging Pump is partially gas bound and B Charging Pump is completely gas bound (1) A completely gas bound pump should be indicated by a (2) The crew should isolate letdown and stop A. (1) quieter than normal sound (2) both Charging Pumps B. (1) quieter than normal sound (2) ONLY the B Charging Pump and evaluate whether Charging flow restores to normal C. (1) louder than normal sound (2) both Charging Pumps D. (1) louder than normal sound (2) ONLY the B Charging Pump and evaluate whether Charging flow restores to normal

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because in a scenario, if there is a RAS actuation and there is indication of cavitation, the Containment Spray pump will be stopped and conditions evaluated for improvement C. First part is plausible since Charging Pumps are positive displacement pumps they try to maintain a higher discharge pressure. As the pump become more gas bound it will work harder and therefore make a louder than expected noise. Another possibility is that the more gas bound a pump becomes, the more cavitation is occurring because of air in the system. Therefore more cavitation will have a higher than normal noise. Second part is correct.

D. First part is plausible since Charging Pumps are positive displacement pumps they try to maintain a higher discharge pressure. As the pump become more gas bound it will work harder and therefore make a louder than expected noise. Another possibility is that the more gas bound a pump becomes, the more cavitation is occurring because of air in the system. Therefore more cavitation will have a higher than normal noise Second part is plausible because in a scenario, if there is a RAS actuation and there is indication of cavitation, the Containment Spray pump will be stopped and conditions evaluated for improvement Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 3 Reference N Provided:

Learning Objective: 311393 - Explain how gas binding of the charging pumps is mitigated in 40AO-9ZZ05, Loss of Charging or Letdown.

Technical

Reference:

40AO-9ZZ05, Loss of Charging or Letdown Technical

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40AO-9ZZ05, Loss of Charging or Letdown Examination Outline Cross-

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Level RO SRO K/A: Loss of Residual Heat Removal System: Ability to Tier 1 operate and/or monitor the following as they apply to the Group 1 Loss of Residual Heat Removal System: RCS Inventory K/A 025 AA1.02 IR 3.8 Question 6 Given the following conditions:

Unit 3 is in MODE 4 Train A SDC is in-service using the A LPSI pump RCS Pressure is 320 psia RCS Temperature is 250°F Pressurizer level is 40%

Subsequently:

A leak in the SDC loop occurs RCS pressure is 310 psia and slowly lowering Pressurizer level is 35% and lowering (1) With NO operator action, the Pressurizer Low Level alarm should annunciate AS SOON AS Pressurizer level lowers to (2) After the A SDC Cooling Loop is isolated, the crew can shift to B SDC Cooling Loop using A. (1) 10%

(2) ONLY B LPSI Pump B. (1) 10%

(2) B LPSI OR B CS Pump C. (1) 25%

(2) ONLY B LPSI Pump D. (1) 25%

(2) B LPSI OR B CS Pump

Proposed Answer: C Explanations:

A. First part is plausible because 10% pressurizer level represents a value that ensures that there is subcooled liquid in the Pressurizer. During SPTAs, operators will maintain Pressurizer level greater than 10%. Second part is correct.

B. First part is plausible because 10% pressurizer level represents a value that ensures that there is subcooled liquid in the Pressurizer. During SPTAs, operators will maintain Pressurizer level greater than 10%. Second part is plausible because a CS pump can be used if temperature and pressure requirements have been met (<210 psia and <185°F)

C. Correct D. First part is correct. Second part is plausible because a CS pump can be used if temperature and pressure requirements have been met (<210 psia and <185°F)

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 19381 - Describe the purpose and conditions under which Shutdown Cooling System is designed to function

Technical

Reference:

40AL-9RK4A, Panel B04A Alarm Responses Technical

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40EP-9EO11, Lower Mode Functional Recovery Procedure Technical

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40EP-9EO11, Lower Mode Functional Recovery Examination Outline Cross-

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Level RO SRO K/A: Loss of Component Cooling Water: Knowledge of Tier 1 the reasons for the following responses as they apply to Group 1 the Loss of Component Cooling Water: Effect on the CCW flow header of a loss of CCW K/A 026 AK3.04 IR 3.5 Question 7 Given the following conditions:

Unit 1 is operating at 100% power A complete loss of Nuclear Cooling water occurred 5 minutes ago The CRS has directed the BOP to perform 40AO-9ZZ03 Loss of Cooling Water, Appendix A, Cross-connect EW to NC, using Train A EW Per Appendix A the BOP should ensure that a MAXIMUM of ___(1)___ Normal Chiller NC outlet valve(s) is(are) open in order to ___(2)___.

A. (1) one (2) provide additional EW flow to NC priority loads B. (1) one (2) ensure sufficient flow to the A SDCHX in the event of a design basis accident C. (1) two (2) provide additional EW flow to NC priority loads D. (1) two (2) ensure sufficient flow to the A SDCHX in the event of a design basis accident

Proposed Answer: A Explanations: This K/A is a match because there is a loss of NC (CCW) that results in flow lost to NC priority loads. After EW is cross-tied to NC, there is still insufficient flow to NC priority loads so ensuring that a maximum of 1 chiller outlet valve will increase flow.

A. Correct B. First part is correct. Second part is plausible because losing the outlet valve initially will provide more flow to the SDCHX, however if there is an accident and a SIAS, the cross connect valves will close to ensure that there is enough flow to the SDCHX.

C. First part is plausible because during summer months, two large chillers are needed to maintain adequate cooling supply. Second part is correct.

D. First part is plausible because during summer months, two large chillers are needed to maintain adequate cooling supply. Second part is plausible because losing the outlet valve initially will provide more flow to the SDCHX, however if there is an accident and a SIAS, the cross connect valves will close to ensure that there is enough flow to the SDCHX.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 22359 - Given a loss of NC, describe how flow to the RCPs is increased after EW has been cross-tied

Technical

Reference:

40AO-9ZZ03, Loss of Cooling Water, Appendix A, Cross-Connect EW to NC

Technical

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Loss of Cooling Water Lesson Plan NKASMC030305 Examination Outline Cross-

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Level RO SRO K/A: Pressurizer Pressure Control System Malfunction: Tier 1 Knowledge of the reasons for the following responses Group 1 as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in EOP for PZR PCS K/A 027 AK3.03 malfunction IR 3.7 Question 8 Given the following conditions:

Unit 3 is operating at 100% power Subsequently:

Pressurizer Spray Control Valve, RCE-PV-100E, failed open All ARP actions to close the failed valve were unsuccessful The CRS directed the BOP to trip the Reactor when RCS Pressure lowered to 1950 psia During SPTAs, the OATC operated RCPs as directed in the ARP Which of the following describes the ARP directed action for RCP operation and the reason for this direction?

The OATC should trip ___(1)___ RCPs in order to ___(2)___ .

A. (1) ONLY 2 (2) protect RCPs due to insufficient NPSH for 4 RCPs to be in operation B. (1) ONLY 2 (2) reduce DP across the Main Spray valves to allow heaters to restore pressure C. (1) ALL 4 (2) protect RCPs due to insufficient NPSH for 4 RCPs to be in operation D. (1) ALL 4 (2) reduce DP across the Main Spray valves to allow heaters to restore pressure

Proposed Answer: D Explanations:

A. First part is plausible since the Main Spray valves only tap off of two RCS loops and maintaining forced circulation is always preferred, however the ARP directs stopping all 4 RCPs. Second part is plausible as NPSH is degrading as RCS pressure lowers, and RCPs are required to be stopped if pressure lowers to less than minimum NPSH, however that pressure has not been reached and is not the basis for stopping RCPs following the reactor trip.

B. First part is plausible since the Main Spray valves only tap off of two RCS loops and maintaining forced circulation is always preferred, however the ARP directs stopping all 4 RCPs. Second part is correct.

C. First part is correct. Second part is plausible as NPSH is degrading as RCS pressure lowers, and RCPs are required to be stopped if pressure lowers to less than minimum NPSH, however that pressure has not been reached and is not the basis for stopping RCPs following the reactor trip.

D. Correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 14 Reference N Provided:

Learning Objective: 24946 - Describe the response of the Pressurizer Pressure Control System to a failure of an input transmitter

Technical

Reference:

40AL-9RK4A, Panel B04A Alarm Responses Technical

Reference:

40EP-9EO10-002, Appendix 2: Figures Examination Outline Cross-

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Level RO SRO K/A: Anticipated Transient Without Scram: Knowledge Tier 1 of the interrelations between the ATWS and the Group 1 following an ATWS: Breakers, relays, and disconnects K/A 029 EK2.06 IR 2.9 Question 9 Given the following conditions:

A malfunction has caused Pressurizer pressure to rise IF an ATWS occurred and NO OPERATOR ACTION is taken, the SPS should send a trip signal to ___(1)___ AS SOON AS RCS pressure reaches a MINIMUM of

___(2)___ psia.

A. (1) RTCBs ONLY (2) 2383 B. (1) RTCBs ONLY (2) 2409 C. (1) RTCBs and MG Set contactors (2) 2383 D. (1) RTCBs and MG Set contactors (2) 2409

Proposed Answer: D Explanations:

A. First part is plausible because if an RPS Reactor trip setpoint is exceeded, only the RTCBs will open. Second part is plausible because 2383 psia is the RPS Reactor trip setpoint.

B. First part is plausible because if an RPS Reactor trip setpoint is exceeded, only the RTCBs will open. Second part is correct.

C. First part is correct. Second part is plausible because 2383 psia is the RPS Reactor trip setpoint.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 7 Reference N Provided:

Learning Objective: 24948 - Describe the Supplementary Protection System including its function, instrumentation, bases, and setpoint

Technical

Reference:

40AL-9RK5A, Panel B05A Alarm Responses Technical

Reference:

Plant Protection System Tech Manual Technical

Reference:

40AL-9RK5A, Panel B05A Alarm Responses Examination Outline Cross-

Reference:

Level RO SRO K/A: Steam Generator Tube Rupture: Knowledge of the Tier 1 operational implications of the following concepts as Group 1 they apply to the SGTR: Use of steam tables K/A 038 EK1.01 IR 3.1 Question 10 Given the following conditions:

Unit 2 was tripped due to a SG tube rupture on SG #1 The CRS entered 40EP-9EO04, SGTR The BOP should lower Steam Generator pressures to a MAXIMUM of ___(1)___ to ensure that ___(2)___ is at the required temperature prior to isolating SG #1.

A. (1) 950 psia (2) THOT B. (1) 950 psia (2) TCOLD C. (1) 1135 psia (2) THOT D. (1) 1135 psia (2) TCOLD

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because when cooling down for all other events, TCOLD is used to track the cooldown.

C. First part is plausible because 1135 psia is the pressure that the RCS will be lowered to prevent possibly lifting a Main Steam Safety Valve during a SGTR. Second part is correct.

D. First part is plausible because 1135 psia is the pressure that the RCS will be lowered to prevent possibly lifting a Main Steam Safety Valve during a SGTR. Second part is plausible because when cooling down for all other events, TCOLD is used to track the cooldown.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 5 Reference Y Steam Tables Provided:

Learning Objective: 29951 - Given that the SGTR ORP is being performed and the RCS is being cooled to allow SG isolation, state the associated parameter and value of the cooldown target and its basisin accordance with 40EP-9EO04 and the SGTR Technical Guideline

Technical

Reference:

40EP-9EO04, Steam Generator Tube Rupture Technical

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40EP-9EO04, Steam Generator Tube Rupture Examination Outline Cross-

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Level RO SRO K/A: Steam Line Rupture - Excessive Heat Transfer: Tier 1 Knowledge of the interrelations between the (Excess Group 1 Steam Demand) and the following: Facilitys heat removal systems, including primary coolant, emergency K/A 040 EK2.2 coolant, the decay heat removal systems, and relations IR 3.2 between the proper operation of these systems to the operation of the facility Question 11 Given the following conditions:

An unisolable ESD event outside of Containment is in progress in Unit 1 SG #1 pressure is 920 psia and lowering SG #2 pressure is 950 psia and stable The CRS has entered 40EP-9EO05, Excess Steam Demand After dryout conditions have been met on the faulted Steam Generator, the crew should minimize the effects of Pressurized Thermal Shock by stabilizing TCOLD using ___(1)___

and ___(2)___.

A. (1) ADVs (2) throttling closed HPSI Injection valves B. (1) ADVs (2) depressurizing the RCS with Auxiliary Spray Valves C. (1) SBCVs (2) throttling closed HPSI Injection valves D. (1) SBCVs (2) depressurizing the RCS with Auxiliary Spray Valves

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because opening Auxiliary Spray Valves will stop (or limit) the repressurization of the RCS, however the continuing injection of SI flow will continue making PTS a possibility.

C. First part is plausible because SBCVs would normally be used to control RCS temperature post Reactor trip. However since SG pressures have lowered below MSIS setpoints, therefore ADVs will be used. Second part is correct.

D. First part is plausible because SBCVs would normally be used to control RCS temperature post Reactor trip. However since SG pressures have lowered below MSIS setpoints, therefore ADVs will be used. Second part is plausible because opening Auxiliary Spray Valves will stop (or limit) the repressurization of the RCS, however the continuing injection of SI flow will continue making PTS a possibility.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 5 Reference N Provided:

Learning Objective: 25489 - Given a set of plant parameters determine when and how RCS temperature is stabilized during an ESD per 40EP-9EO05, ESD

Technical

Reference:

40AL-9RK5A, Panel B05A Alarm Responses Technical

Reference:

40EP-9EO05, Excess Steam Demand Technical

Reference:

40EP-9EO05, Excess Steam Demand Technical

Reference:

40DP-9AP10, Excess Steam Demand Technical Guideline Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Main Feedwater: Knowledge of the Tier 1 operational implications of the following concepts as Group 1 they apply to the (Loss of Feedwater): Components, capacity, and function of emergency systems K/A 054 EK1.1 IR 3.2 Question 12 Given the following conditions:

Unit 1 tripped due to a complete loss of Main Feedwater.

AFB-P01 has been manually started and aligned to feed both SGs.

Subsequently:

AFAS-1 actuates.

With NO operator action, how should the AFAS-1 affect the current feed lineup?

AFA-P01 should start and feed ___(1)___ and AFB-P01 should be feeding ___(2)___.

A. (1) SG #1 ONLY (2) SG #1 ONLY B. (1) SG #1 ONLY (2) both SGs C. (1) both SGs (2) SG #1 ONLY D. (1) both SGs (2) both SGs

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible since it will align to feed SG 1 and it was already aligned to feed SG 2, however on an AFAS-1, all feed will stop to SG 2 and both AFW Pumps will commence feeding SG 1.

C. First part is plausible since AFA-P01 is drawing steam from both SGs on an AFAS-1 or AFAS-2, however it will only feed the SG with the active AFAS signal. Second part is correct.

D. First part is plausible since AFA-P01 is drawing steam from both SGs on an AFAS-1 or AFAS-2, however it will only feed the SG with the active AFAS signal. Second part is plausible since it will align to feed SG 1 and it was already aligned to feed SG 2, however on an AFAS-1, all feed will stop to SG 2 and both AFW Pumps will commence feeding SG 1.

Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 24499 - Describe the system response to an Auxiliary Feedwater Actuation Signal

Technical

Reference:

Auxiliary Feedwater System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Station Blackout: Knowledge of the reasons for Tier 1 the following responses as they apply to the Station Group 1 Blackout: Actions contained in EOP for loss of offsite and onsite power K/A 055 EK3.02 IR 4.3 Question 13 Per 40EP-9EO08, Blackout, the purpose of actuating a MSIS is to

1. minimize cooldown of the RCS

2. minimize effects of loss of Instrument Air

3. prevent damage to the Main Condenser

4. prevent an inadvertent loss of steam pressure and inventory A. 1 AND 4 ONLY B. 1 AND 3 ONLY C. 2 AND 3 ONLY D. 2 AND 4 ONLY

Proposed Answer: B Explanations:

A. Minimize cooldown of the RCS is correct. Prevent an inadvertent loss of steam pressure and inventory is plausible because an MSIS will minimize pressure and inventory losses but this is not the reason for MSIS. During Blackout inventory will be maintained with AFA-P01 and will not be an issue.

B. Correct C. Minimize the effects of loss of Instrument Air is plausible because all Instrument Air and Service Air Compressors lose power. As Instrument Air pressure lowers, there are valves that will fail open and could potentially affect the RCS. 40EP-9EO08, Blackout does have a step to monitor IA and Nitrogen air pressure, however it is not one of the purposes for actuating MSIS. It is also plausible because the second step in the Loss of Instrument Air AOP is to initiate MSIS if desired.

Prevent damage to the Main Condenser is correct.

D. Minimize the effects of loss of Instrument Air is plausible because all Instrument Air and Service Air Compressors lose power. As Instrument Air pressure lowers, there are valves that will fail open and could potentially affect the RCS. 40EP-9EO08, Blackout does have a step to monitor IA and Nitrogen air pressure, however it is not one of the purposes for actuating MSIS. It is also plausible because the second step in the Loss of Instrument Air AOP is to initiate MSIS if desired.

Prevent an inadvertent loss of steam pressure and inventory is plausible because an MSIS will minimize pressure and inventory losses but this is not the reason for MSIS. During Blackout inventory will be maintained with AFA-P01 and will not be an issue.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference N Provided:

Learning Objective: 26233 - Explain why an MSIS is initiated in accordance with 40EP-9EO08, Blackout

Technical

Reference:

40DP-9AP13, Blackout Technical Guideline Technical

Reference:

40DP-9AP13, Blackout Technical Guideline Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Vital AC Instrument Bus: Knowledge of Tier 1 the reasons for the following responses as they apply to Group 1 the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus K/A 057 AK3.01 IR 4.1 Question 14 Given the following conditions:

Unit 2 is operating at 100% power 120 VAC Class Instrument Bus, PNA-D25, tripped on a fault The crew is required to commence monitoring DNBR/LHR/AZTILT/ASI for ADVERSE trends within a MAXIMUM of ___(1)___ to ensure DNBR/LHR/AZTILT/ASI are within Technical Specification limits due to the loss of ___(2)___.

A. (1) 15 minutes (2) COLSS B. (1) 15 minutes (2) CEAC 1 C. (1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) COLSS D. (1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) CEAC 1

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because CEAC 1 is INOPERABLE with a loss of PNA-D25 C. First part is plausible because a loss of PNA-D25 makes COLSS out of service and it may be assumed that LHR and DNBR are exceeding the Technical Specification limits and is required to be restored within one hour. Second part is correct.

D. First part is plausible because a loss of PNA-D25 makes COLSS out of service and it may be assumed that LHR and DNBR are exceeding the Technical Specification limits and is required to be restored within one hour. Second part is plausible because CEAC 1 is INOPERABLE with a loss of PNA-D25. CEAC can still be monitored from CEAC 2.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 17537 - Given conditions where COLSS is inoperable, monitor DNBR/LHR/ASI with COLSS out of service in accordance with 72ST-9RX03

Technical

Reference:

40AO-9ZZ13, Loss of Class Instrument or Control Power Technical

Reference:

Technical

Reference:

Technical Specifications Technical

Reference:

Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of DC Power: Knowledge of limiting Tier 1 condition for operations and safety limits Group 1 K/A 058 G 2.2.22 IR 4.0 Question 15 Given the following conditions:

Unit 1 is operating in MODE 4 The crew is cooling down and depressurizing the RCS following a LOOP per 40EP-9EO07, Loss of Offsite Power/Loss of Forced Circulation Subsequently:

Class 125 VDC Bus, PKB-M42, tripped on overcurrent Per Technical Specifications, the REQUIRED ACTION(s) of LCO ___(1)___ must be performed and ___(2)___ Auxiliary Spray Valve(s) is(are) available to continue the depressurization.

A. (1) 3.8.4, DC Sources - Operating (2) one B. (1) 3.8.4, DC Sources - Operating (2) both C. (1) 3.8.5, DC Sources - Shutdown (2) one D. (1) 3.8.5, DC Sources - Shutdown (2) both

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because only PKA-M41 and PKB-M42 cause a loss of an Auxiliary Spray Valve. PKC-M43 and PKD-M44 will not cause a loss of an Auxiliary Spray Valve.

C. First part is plausible because the Lower Mode Functional Recovery procedure is used in Mode 4 so it could be assumed that if the LMFRP is being used then LCO 3.8.5 would be applicable.

Second part is correct.

D. First part is plausible because the Lower Mode Functional Recovery procedure is used in Mode 4 so it could be assumed that if the LMFRP is being used then LCO 3.8.5 would be applicable.

Second part is plausible because only PKA-M41 and PKB-M42 cause a loss of an Auxiliary Spray Valve. PKC-M43 and PKD-M44 will not cause a loss of an Auxiliary Spray Valve.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 8 Reference N Provided:

Learning Objective: 21203 - Given a set of plant conditions determine whether or not the LCOs and TLCOs of 3.8 are satisfied in accordance with Tech Spec 3.8

Technical

Reference:

Technical Specifications Technical

Reference:

40AO-9ZZ13, Loss of Class Instrument or Control Power Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Nuclear Service Water: Ability to operate Tier 1 and/or monitor the following as they apply to the Loss of Group 1 Nuclear Service Water (SWS): Loads on the SWS in the control room K/A 062 AA1.02 IR 3.2 Question 16 Given the following conditions:

Unit 2 is operating at 100% power A PW pump is OOS Subsequently:

B PW pump trips on overcurrent (1) Per 40AO-9ZZ03, Loss of Cooling Water, the crew should trip (2) The crew should break vacuum on the Main Turbine A. (1) the Main Turbine ONLY (2) IMMEDIATELY following the Main Turbine trip B. (1) the Main Turbine ONLY (2) as soon as the Main Turbine reaches 1200 RPM C. (1) the Reactor (2) IMMEDIATELY following the Main Turbine trip D. (1) the Reactor (2) as soon as the Main Turbine reaches 1200 RPM

Proposed Answer: D Explanations:

A. First part is plausible because Plant Cooling Water is the heat sink for the Turbine Cooling Water HX, therefore cools portions of the Main Turbine. However since it is also the heat sink for the Nuclear Cooling Water HXs, the Reactor must be tripped. Second part is plausible if it is thought that the quicker the Main Turbine is stopped, the less damage a high temperature condition will cause.

B. First part is plausible because Plant Cooling Water is the heat sink for the Turbine Cooling Water HX, therefore cools portions of the Main Turbine. However since it is also the heat sink for the Nuclear Cooling Water HXs, the Reactor must be tripped. Second part is correct.

C. First part is correct. Second part is plausible if it is thought that the quicker the Main Turbine is stopped, the less damage a high temperature condition will cause.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference N Provided:

Learning Objective: 27138 - Given the Loss of Cooling Water AOP is being performed determine the appropriate mitigating strategies for a loss of plant cooling water in accordance with 40AO-9ZZ03

Technical

Reference:

40AO-9ZZ03, Loss of Cooling Water Technical

Reference:

40AO-9ZZ03, Loss of Cooling Water, Appendix B, Minimize Cooling Load on TC

Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Instrument Air: Ability to operate and/or Tier 1 monitor the following as they apply to the Loss of Group 1 Instrument Air: RPS K/A 065 AA1.05 IR 3.3 Question 17 Given the following conditions:

Unit 3 is operating at 100% power A loss of Instrument Air and Service Air has caused system pressure to lower to 50 psig With NO operator action, the Reactor should AUTOMATICALLY trip on A. Low DNBR B. Variable Overpower C. High Pressurizer Pressure D. Low Steam Generator Water Level

Proposed Answer: D Explanations: The Reactor will trip on Low Steam Generator Water Level because both MFPs trip.

A. Since MFPs trip, steam generator water levels will lower causing the RCS to heat up. Also power will rise from HDP discharge valves closing causing DNBR to lower.

B. HDP discharge valves close and will cause a minor rise in Reactor power but will not rise to an automatic Variable Overpower Reactor trip.

C. SBCV valves fail closed and MFPs will trip sequentially therefore there Pressurizer pressure will rise. Pressurizer spray valves will fail closed once IA pressure lowers to 38-48 psig. Since they will still operate at 50 psig, Pressurizer pressure will not rise to the automatic Reactor trip setpoint D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 25935 - Determine the major effects on plant operation as instrument air pressure degrades

Technical

Reference:

40AO-9ZZ06, Loss of Instrument Air Examination Outline Cross-

Reference:

Level RO SRO K/A: Generator Voltage and Electric Grid Disturbances: Tier 1 Knowledge of the interrelations between Generator Group 1 Voltage and Electric Grid Disturbances and the following: Sensors, detectors, indicators K/A 077 AK2.03 IR 3.0 Question 18 Given the following conditions:

Unit 1 is operating at 100% power Main Generator MVARs are at UNITY Subsequently:

A transmission line relaying has caused grid voltage to rise (1) Main Generator MVARs should initially be (2) Main Generator MVARs should be restored to UNITY A. (1) BUCKING (2) by a manual voltage adjustment B. (1) BUCKING (2) by the Auto Voltage Regulator C. (1) BOOSTING (2) by a manual voltage adjustment D. (1) BOOSTING (2) by the Auto Voltage Regulator

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because the AC Regulator will maintain Generator terminal voltage. As system load changes, the terminal voltage will need to be adjusted with a manual adjustment.

C. First part is plausible because when grid voltage changes, MVARS will no longer be in Unity.

Since grid voltage rises it may be assumed that the Main Generator will react in the same way and will be Boosting. Second part is correct.

D. First part is plausible because when grid voltage changes, MVARS will no longer be in Unity.

Since grid voltage rises it may be assumed that the Main Generator will react in the same way and will be Boosting. Second part is plausible because the AC Regulator will maintain Generator terminal voltage. As system load changes, the terminal voltage will need to be adjusted with a manual adjustment.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 5 Reference N Provided:

Learning Objective: 367777 - Explain the operation of the EX2100e Voltage Regulators

Technical

Reference:

Main Generator Excitation and Regulation (EX2100e) Lesson Plan Technical

Reference:

Main Generator Excitation and Regulation (EX2100e) Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Continuous Rod Withdrawal: Knowledge of the Tier 1 operational implications of the following concepts as Group 2 they to Continuous Rod Withdrawal: Integral rod worth K/A 001 AK1.21 IR 2.9 Question 19 Given the following conditions:

Unit 3 is operating at 50% power The crew is recovering from a loss of the A MFP The OATC is withdrawing CEAs to restore overlap per 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump)

The selected Group is 30 inches withdrawn After the RO lets go of the withdrawal switch, CEA 18 continues to withdraw (1) As CEA 18 continues to withdraw, its integral rod worth available to insert should (2) If all actions to stop CEA 18 were unsuccessful, the crew should A. (1) increase (2) trip the Reactor B. (1) increase (2) manually insert all other CEAs in the selected group C. (1) decrease (2) trip the Reactor D. (1) decrease (2) manually insert all other CEAs in the selected group

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because SDM will still be met if the CEA completely withdraws. It may be assumed that because Reactor power is 50% a single CEA withdrawing will not cause power to exceed our thermal power limit so inserting the remaining CEAs in the group will allow the crew to maintain the Reactor on line and possible troubleshoot the malfunctioning CEA..

C. First part is plausible because as a CEA is withdrawn differential rod worth will eventually decrease, it may be assumed that integral rod worth is the same. However as a CEA withdraws, the available integral rod worth will increase. Second part is correct.

D. First part is plausible because as a CEA is withdrawn differential rod worth will eventually decrease, it may be assumed that integral rod worth is the same. However as a CEA withdraws, the available integral rod worth will increase. Second part is plausible because SDM will still be met if the CEA completely withdraws. It may be assumed that because Reactor power is 50% a single CEA withdrawing will not cause power to exceed our thermal power limit so inserting the remaining CEAs in the group will allow the crew to maintain the Reactor on line and possible troubleshoot the malfunctioning CEA..

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 6 Reference N Provided:

Learning Objective: 25220 - Given conditions of a CEA Malfunction determine when a Reactor trip is required

Technical

Reference:

40AO-9ZZ11, CEA Malfunctions Examination Outline Cross-

Reference:

Level RO SRO K/A: Inoperable/Stuck Rod: Ability to operate and/or Tier 1 monitor the following as they apply to the Group 2 Inoperable/Stuck Control Rod: CRDS K/A 005 AA1.01 IR 3.6 Question 20 Given the following conditions:

Unit 1 is operating at 10% power A power ascension is in progress The OATC is withdrawing Group 5 CEAs in Manual Sequential Group 5 CEAs are currently 138 inches Subsequently:

The OATC withdraws Group 5 CEAs All Group 5 CEAs withdraw to 142.5 inches with the exception of CEA 14 CEA 14 is stuck at 138 inches CEA 14 can be verified stuck at 138 inches using ___(1)___. Once troubleshooting is complete and management concurrence is received, per 40AO-9ZZ11, CEA Malfunctions, the crew should re-align CEAs by ___(2)___.

A. (1) RSPTs (2) withdrawing CEA 14 to 142.5 inches B. (1) RSPTs (2) inserting Group 5 CEAs to 138 inches C. (1) pulse counters (2) withdrawing CEA 14 to 142.5 inches D. (1) pulse counters (2) inserting Group 5 CEAs to 138 inches

Proposed Answer: A Explanations: Pulse counters vs reed switches are commonly confused at PVNGS. It is very plausible for someone to think that the pulse counters are referring to a magnetic pulse that actuates as the rod passes each magnet and that each time the CEDMCS latches and unlatches from upper and lower grippers that the next reed switch in the system is actuated.

A. Correct B. First part is correct. Second part is plausible because inserting CEAs is a possibility. However, 40AO-9ZZ11, CEA Malfunctions directs withdrawing a single CE to align with its group.

C. First part is plausible because pulse counters can normally be used to determine CEA location.

However since a withdraw demand was inputted and CEA position did not change, pulse counter indication will be 142.5 inches instead of 138 inches. RSPTs will indicate 140 inches. Second part is correct.

D. First part is plausible because pulse counters can normally be used to determine CEA location.

However since a withdraw demand was inputted and CEA position did not change, pulse counter indication will be 142.5 inches instead of 138 inches. RSPTs will indicate 140 inches. Second part is plausible because inserting CEAs is a possibility. However, 40AO-9ZZ11, CEA Malfunctions directs withdrawing a single CE to align with its group.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 6 Reference N Provided:

Learning Objective: 22615 - Explain the operation of the RSPTs

Technical

Reference:

Control Element Drive Mechanism Control System Tech Manual Technical

Reference:

40AO-9ZZ11, CEA Malfunctions Examination Outline Cross-

Reference:

Level RO SRO K/A: Accidental Liquid Radwaste Release: Knowledge Tier 1 of the interrelations between the Accidental Liquid Group 2 Radwaste Release and the following: Radioactive-gas monitors K/A 059 AK2.02 IR 2.7 Question 21 Given the following indications:

A leak on Liquid Radwaste Holdup Tank, LRN-T01C, caused a Lo-Lo Level Alarm at the Liquid Radwaste Annunciator Panel, LRN-E01.

(1) When LRN-T01C, Lo-Lo Level alarm annunciates a trip signal should be sent to (2) Airborne radioactivity vented from any Liquid Radwaste Holdup Tank should be detected INITIALLY by ___(2)___

A. (1) ALL Liquid Radwaste Holdup Tank Pumps (2) RU-143, Plant Vent radiation monitor B. (1) ALL Liquid Radwaste Holdup Tank Pumps (2) RU-14, Radwaste Building Ventilation Exhaust Filter Inlet radiation monitor C. (1) ONLY Liquid Radwaste Holdup Tank Pump, LRN-P01C (2) RU-143, Plant Vent radiation monitor D. (1) ONLY Liquid Radwaste Holdup Tank Pump LRN-P01C (2) RU-14, Radwaste Building Ventilation Exhaust Filter Inlet radiation monitor

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible RU-143 is the Tech Spec radiation monitor that is used to monitor planned releases.

B. Correct C. First part is plausible because if the pump suctions were not cross-tied to all of the tanks, it would make sense that only the pump that is pumping down a tank would actually trip. Second part is plausible RU-143 is the Tech Spec radiation monitor that is used to monitor planned releases.

D. First part is plausible because if the pump suctions were not cross-tied to all of the tanks, it would make sense that only the pump that is pumping down a tank would actually trip. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 11 Reference N Provided:

Learning Objective: 31202 - Given a Radiation Monitor number and name describe the purposes and sample points of the monitor

Technical

Reference:

Liquid Radwaste System Tech Manual Technical

Reference:

Radiation Monitoring System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Accidental Gaseous Radwaste Release: Ability to Tier 1 verify system alarm setpoints and operate controls Group 2 identified in the alarm response manual K/A 060 G 2.4.50 IR 4.2 Question 22 During a release of a Waste Gas Decay Tank, a high-high alarm on which of the following Radiation Monitors, INDIVIDUALLY, will require the crew to ensure that the Gaseous Discharge Header Isolation Valves, GR-UV-34A and GR-UV-34B are closed per 74AL-9SQ01, Radiation Monitoring System Alarm Validation and Response?

1. RU-8, Auxiliary Building Ventilation Exhaust Filter Monitor

2. RU-12, Waste Gas Decay Tank Monitor

3. RU-15, Waste Gas System Area Combined Ventilation Exhaust Monitor A. 1 ONLY B. 2 ONLY C. 1 and 3 ONLY D. 2 and 3 ONLY

Proposed Answer: B Explanations:

A. Plausible because the Auxiliary Building is adjacent to the Radwaste Building and if there was a leak from the Waste Gas header, it is possible that RU-8 will detect it. However the high radiation is in the enclosed Waste Gas system, therefore RU-8 will not detect it.

B. Correct C. Plausible because the Auxiliary Building is adjacent to the Radwaste Building and if there was a leak in the Waste Gas header, it is possible that RU-8 will detect it. However the high radiation is in the enclosed Waste Gas system, therefore RU-8 will not detect it. RU-15 will detect a leak from the Waste Gas header however the high radiation is in the enclosed Waste Gas system, therefore RU-15 will not detect it.

D. RU-12 is correct. RU-15 will detect a leak from the Waste Gas header however the high radiation is in the enclosed Waste Gas system, therefore RU-15 will not detect it.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 11 Reference N Provided:

Learning Objective: 18733 - Describe the automatic functions / interlocks with Gaseous Discharge Header Isolation Valves (UV-34A & 34B)

Technical

Reference:

Radiation Monitoring System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Control Room Evacuation: Knowledge of the Tier 1 interrelations between the Control Room Evacuation and Group 2 the following: Auxiliary shutdown panel layout K/A 068 AK2.01 IR 3.9 Question 23 Per 40AO-9ZZ19, Control Room Fire, the crew should actuate MSIS ___(1)___

exiting the Control Room and disconnect switches should be taken to LOCAL on the

___(2)___ Remote Shutdown Panel.

A. (1) prior to (2) A B. (1) prior to (2) B C. (1) after (2) A D. (1) after (2) B

Proposed Answer: B Explanations:

A. First part is correct. The second part is plausible because A Remote Shutdown Panel is similar to the B panel with the exception of the disconnect switches.

B. Correct C. First part is plausible because an MSIS can be actuated from the RSD panels and if there is a Control Room fire it is important to evacuate the CR as quickly as possible. The second part is plausible because A Remote Shutdown Panel is similar to the B panel with the exception of the disconnect switches.

D. First part is plausible because an MSIS can be actuated from the RSD panels and if there is a Control Room fire it is important to evacuate the CR as quickly as possible. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 8 Reference N Provided:

Learning Objective: 26655 - State the indications available to the operator at the Remote Shutdown Panel (RSP)

Technical

Reference:

40AO-9ZZ19, Control Room Fire Technical

Reference:

Simulator Computer Drawing Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Containment Integrity: Knowledge of the Tier 1 reasons for the following responses as they apply to the Group 2 Loss of Containment Integrity: Guidance contained in EOP for loss of containment integrity K/A 069 AK3.01 IR 3.8 Question 24 Given the following conditions:

Unit 3 was manually tripped due to a LOCA on RCP 1B HP Seal Cooler All RCPs were stopped RCP Controlled Bleedoff was isolated The power supply to RCP 1B HP Seal Cooler isolation valves, NHN-M10 faulted When the crew attempted to close NCB-UV-403 NCWS Return Internal Isolation Valve, it failed to close (1) To maintain Containment integrity the crew should close at a MINIMUM (2) Closing the correct NC valve should A. (1) NCB-UV-401 NCWS Supply External Isolation Valve (2) isolate the RCS leak B. (1) NCB-UV-401 NCWS Supply External Isolation Valve (2) restrict the RCS leak to Containment ONLY C. (1) NCA-UV-402 NCWS Return External Isolation Valve (2) isolate the RCS leak D. (1) NCA-UV-402 NCWS Return External Isolation Valve (2) restrict the RCS leak to Containment ONLY

Proposed Answer: D Explanations:

A. First part is plausible because NCB-UV-401 is a Containment Isolation Valve that the crew will attempt to close. However there is a check valve in line with NCB-UV-401 that will isolate the NC Supply. Second part is plausible because the RCS leak will be isolated to Containment. However, there is a relief valve that will lift inside of Containment, therefore the leak will continue inside of Containment.

B. First part is plausible because NCB-UV-401 is a Containment Isolation Valve that the crew will attempt to close. However there is a check valve in line with NCB-UV-401 that will isolate the NC Supply. Second part is correct.

C. First part is correct. Second part is plausible because the RCS leak will be isolated to Containment. However, there is a relief valve that will lift inside of Containment, therefore the leak will continue inside of Containment.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 23790 - Explain the operation of the NC Containment Isolation Valves under normal operating conditions

Technical

Reference:

Nuclear Cooling Water System Technical Manual Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Examination Outline Cross-

Reference:

Level RO SRO K/A: High Reactor Coolant Activity: Ability to determine Tier 1 and interpret the following as they apply to the High Group 2 Reactor Coolant Activity: Location or process point that is causing alarm K/A 076 AA2.01 IR 2.7 Question 25 An RMS alarm on ___(1)___ , which monitor(s) radiation levels of ___(2)___, is(are) the primary RMS indication(s) of high reactor coolant activity and possible fuel failure.

A. (1) Primary Coolant Activity Monitors, RU-150/151 (2) each RCS hot leg B. (1) Primary Coolant Activity Monitors, RU-150/151 (2) one RCS cold leg of each Steam Generator C. (1) Reactor Coolant Letdown Line Radiation Monitor, RU-155D (2) the letdown line at the inlet of the Letdown Heat Exchanger D. (1) Reactor Coolant Letdown Line Radiation Monitor, RU-155D (2) the letdown line between the Letdown Heat Exchanger and the Ion Exchangers

Proposed Answer: D Explanations:

A. RU-150/151 is plausible since they are used to determine activity in the primary under post accident conditions, however the primary indicator for high RCS activity under non post accident conditions is RU-155D. Monitored location is plausible because that is where RU-150/151 monitors radiation levels.

B. RU-150/151 is plausible since they are used to determine activity in the primary under post accident conditions, however the primary indicator for high RCS activity under non post accident conditions is RU-155D. Monitored location is plausible because the first location that high activity coolant will enter from the core is the hot leg.

C. First part is correct. Plausible that RU-155D would detect radiation upstream of the letdown HX and downstream of the letdown containment isolation valve to provide earlier detection of high RCS activity than the actual monitoring point for RU-155D and while allowing for the isolation of letdown to determine if RU-155D was reading actual activity or the RM was providing false indications of high activity.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 11 Reference N Provided:

Learning Objective: 31202 - Given a Radiation Monitor number and name describe the purposes and sample points of the monitor

Technical

Reference:

Radiation Monitoring System Tech Manual Technical

Reference:

Radiation Monitoring System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Natural Circulation: Ability to determine and Tier 1 interpret the following as they apply to the (Natural Group 2 Circulation Operations): Adherence to appropriate procedures and operation within the limitations in the K/A CE A13 AA2.2 Facilitys license and amendments IR 2.9 Question 26 Per 40EP-9EO07, Loss of Offsite Power / Loss of Forced Circulation, natural circulation flow is verified by checking that the RCS is a MINIMUM of ___(1)___

subcooled as indicated by ___(2)___ .

A. (1) 24°F (2) CET Subcooling B. (1) 24°F (2) RCS Subcooling C. (1) 30°F (2) CET Subcooling D. (1) 30°F (2) RCS Subcooling

Proposed Answer: A Explanations:

A. Correct.

B. First part is correct. Second part is plausible since RCS Subcooling is the normal parameter used to verify subcooling, however when all RCPs are secured, CET Subcooling is the correct indication of subcooling.

C. First part is plausible since 30°F is the maximum allowed delta-T between Th and max quadrant CET temp for verifying natural circulation, however the minimum allowable subcooling is 24°F.

Second part is correct.

D. First part is plausible since 30°F is the maximum allowed delta-T between Th and max quadrant CET temp for verifying natural circulation, however the minimum allowable subcooling is 24°F.

Second part is plausible since RCS Subcooling is the normal parameter used to verify subcooling, however when all RCPs are secured, CET Subcooling is the correct indication of subcooling.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 14 Reference N Provided:

Learning Objective: 62760 - Given a loss of forced circulation, identify the parameters used to determine Natural Circulation flow per 40EP-9EO07.

Technical

Reference:

40EP-9EO07, LOOP/LOFC Technical

Reference:

Appendix 2, Figures RCS Subcooling is calculated using Th indication, CET Subcooling is calculated using REPCET for Th.

Examination Outline Cross-

Reference:

Level RO SRO K/A: RCS Overcooling - Pressurized Thermal Shock: Tier 1 Ability to operate and/or monitor the following as they Group 2 apply to the (RCS Overcooling) Components, and functions of control and safety systems, including K/A CE A11 AA1.1 instrumentation, signals, interlocks, failure modes, and IR 3.3 automatic and manual features Question 27 Given the following conditions:

Unit 1 tripped from 100% power due to an ESD outside of containment upstream of the MSIVs for SG #1 SG #1 has been isolated per Appendix 113 - Steam Generator 1 Isolation Using the provided Appendix 2, Figures, on the next page:

The red dot represents the current RCS temperature and pressure after SG #1 has been isolated Which of the following is the FIRST action the crew is required to take?

A. Heatup the RCS ONLY B. Depressurize the RCS ONLY C. Heatup and depressurize the RCS D. Perform an RCS soak for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

Proposed Answer: B Explanations:

A. Plausible because heating up the RCS will eventually put RCS temperature and pressure to the allowable side of the oversubcooled line. However once the RCS is oversubcooled, there should not be any additional thermal stresses placed on the RCS or Reactor internals.

B. Correct C. Plausible because heating up the RCS will eventually put RCS temperature and pressure to the allowable side of the oversubcooled line. However once the RCS is oversubcooled, there should not be any additional thermal stresses placed on the RCS or Reactor internals. Depresurizing is correct, but it will be the only action taken.

D. Plausible because since the RCS cooldown rate was violated, it is an action that will need to be taken. However, the RCS temperature and pressure must be on the allowable side of the oversubcooled curve, prior to performing a soak.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference Y Attached Appendix 2, Figures Provided:

Learning Objective: 25501 - Given that the EOPs are being performed and specific plant conditions are given, determine whether or not the plant is oversubcooled, and if it is what actions must be taken per the appropriate procedure

Technical

Reference:

40EP-9EO05, Excess Steam Demand Examination Outline Cross-

Reference:

Level RO SRO K/A: Reactor Coolant Pump: Ability to (a) predict the Tier 2 impacts of the following malfunctions or operations on Group 1 the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the K/A 003 A2.02 consequences of those malfunctions or operations: IR 3.7 Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP Question 28 Given the following conditions:

Unit 1 was tripped from 100% power due to a malfunction on RCP 1A RCP 1A was manually tripped during SPTAs per 40AO-9ZZ04, RCP Emergencies The CRS has entered 40EP-9EO02, Reactor Trip (1) When securing a RCP per 40AO-9ZZ04, the RCP Oil Lift Pump should automatically start and then (2) If there was NO RCP malfunction and the crew entered 40OP-9ZZ10, Mode 3 to Mode 5 Operations, the FIRST RCP should be stopped once RCS temperature is lowered to a MAXIMUM of A. (1) automatically stop (2) 350°F B. (1) automatically stop (2) 500°F C. (1) will need to be manually stopped (2) 350°F D. (1) will need to be manually stopped (2) 500°F

Proposed Answer: D Explanations:

A. First part is plausible because per 40OP-9RC01, Reactor Coolant Pump Operations, RCN-P02A, RC Pump 1A Oil Lift Pump will automatically stop after an RCP is started within 2 minutes.

Second part is plausible because 350°F is MODE 3 entry and two RCPs are required to be stopped per 40OP-9ZZ10, Mode 3 to Mode 5 Operations.

B. First part is plausible because per 40OP-9RC01, Reactor Coolant Pump Operations, RCN-P02A, RC Pump 1A Oil Lift Pump will automatically stop after an RCP is started within 2 minutes.

Second part is correct.

C. First part is correct. Second part is plausible because 350°F is MODE 3 entry and two RCPs are required to be stopped per 40OP-9ZZ10, Mode 3 to Mode 5 Operations.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 3 Reference N Provided:

Learning Objective: 22613 - Explain the operation of the Reactor Coolant Pumps Lube Oil system under normal operating conditions

Technical

Reference:

40OP-9ZZ10, Mode 3 to Mode 5 Operations Technical

Reference:

40OP-9RC01, Reactor Coolant Pump Operations

Technical

Reference:

Examination Outline Cross-

Reference:

Level RO SRO K/A: Chemical and Volume Control: Knowledge of the Tier 2 effect that a loss or malfunction of the CVCS will have Group 1 on the following: RCPs K/A 004 K3.04 IR 3.7 Question 29 Given the following conditions:

Unit 1 is operating at 100% power CHB-UV-515, Letdown to Regenerative Heat Exchanger Isolation Valve fails closed The CRS enters 40AO-9ZZ05, Loss of Charging or Letdown After the crew isolates Seal Injection, which of the following describes the effect (if any) on RCP temperatures?

A. HP Seal Cooler inlet temperature should remain relatively constant, while all other seal temperatures should rise by about 70°F B. HP Seal Cooler inlet temperature should rise to between 200°F and 220°F, all other seal temperatures should rise by about 70°F C. HP Seal Cooler inlet temperature should rise to between 200°F and 220°F while all other seal temperatures remain normal D. Isolating Seal Injection should have NO impact on seal temperatures

Proposed Answer: C Explanations:

A. Plausible because once Seal Injection is lost, flow stagnates and temperatures will remain the same because heat from the RCP is only being transferred to the water around the seals and not around the temperature indicators. Second part is plausible because Seal Injection will isolate at 70°F. Therefore if it lost and Seal Injection temperature was at its minimum, temperature of seals will rise by approximately 70°F. Also, if NC flow is lost along with Seal Injection all seal temperatures will rise.

B. First part is correct. Second part is plausible because Seal Injection will isolate at 70°F. Therefore if it lost and Seal Injection temperature was at its minimum, temperature of seals will rise by approximately 70°F. Also, if NC flow is lost along with Seal Injection all seal temperatures will rise C. Correct D. Plausible because once Seal Injection is lost, flow stagnates and temperatures will remain the same because heat from the RCP is only being transferred to the water around the seals and not around the temperature indicators.

Question Source: New X Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 3 Reference N Provided:

Learning Objective: 26284 - Given a RCP with seal injection removed, determine the temperature response when seal injection is secured to an RCP in accordance with 40AO-9ZZ04 or 40AO-9ZZ05

Technical

Reference:

40AO-9ZZ04, Reactor Coolant Pump Emergencies Technical

Reference:

Chemical Volume Control System Tech Manual Technical

Reference:

Reactor Coolant System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Chemical and Volume Control: Knowledge of the Tier 2 effect of a loss or malfunction on the following CVCS Group 1 components: Seal injection system and limits on flow range K/A 004 K6.31 IR 3.1 Question 30 Given the following conditions:

Unit 2 is operating at 100% power Seal Injection was isolated when CHB-HV-255, RCP Seal Injection Header Supply Valve, was inadvertently closed The CRS entered 40AO-9ZZ04, RCP Emergencies, and has directed restoring Seal Injection per Appendix H, Restoring RCP Seal Injection The individual Seal Injection Flow Controllers have been placed in MANUAL and the Seal Injection Flow Control Valves have been closed CHB-HV-255, RCP Seal Injection Header Supply Valve, has been reopened In order to restore Seal Injection, the OATC should raise Seal Injection flow by

___(1)___ OUTPUT on the Seal Injection Flow Controllers to achieve a final target Seal Injection flow of ___(2)___ .

A. (1) raising (2) 2.0 - 4.0 gpm B. (1) raising (2) 6.0 - 7.5 gpm C. (1) lowering (2) 2.0 - 4.0 gpm D. (1) lowering (2) 6.0 - 7.5 gpm

Proposed Answer: D Explanations:

A. First part is plausible since raising output on a controller usually results in raising flow, however Seal Injection Flow Controllers are reverse acting. Second part is plausible since 2-4 gpm is the normal flowrate for RCP Seal Bleedoff, however the normal flowrate for RCP Seal Injection is 6-75 gpm.

B. First part is plausible since raising output on a controller usually results in raising flow, however Seal Injection Flow Controllers are reverse acting. Second part is correct.

C. First part is correct. Second part is plausible since 2-4 gpm is the normal flowrate for RCP Seal Bleedoff, however the normal flowrate for RCP Seal Injection is 6-75 gpm.

D. Correct.

Question Source: New X Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 3 Reference N Provided:

Learning Objective: 311397 - Explain Restoration of Seal Injection

Technical

Reference:

40AO-9ZZ04, RCP Emergencies Technical

Reference:

40AO-9ZZ04, RCP Emergencies Technical

Reference:

40AO-9ZZ04, RCP Emergencies Examination Outline Cross-

Reference:

Level RO SRO K/A: Residual Heat Removal: Ability to predict and/or Tier 2 monitor changes in parameters (to prevent exceeding Group 1 design limits) associated with operating the RHRS controls including: Heatup/cooldown rates K/A 005 A1.01 IR 3.5 Question 31 Given the following conditions:

Unit 1 is in MODE 4 The crew is placing SDC in service using the Train A LPSI Pump The CRS directs warming up the Train A SDCHX at the MAXIMUM heat up rate allowed by 40OP-9SI01, Shutdown Cooling Initiation SIA-HV-306, LPSI S/D Cooling HX A Bypass Valve, is 20% open The A LPSI Pump has been started SIA-UV-635, LPSI Header A to RC Loop 1A, is 10% open Based on the trend on the following page, in order to comply with the CRS direction, the A SDCHX heat up rate should be ___(1)___ and the crew can accomplish this by throttling ___(2)___ on SIA-HV-306, LPSI S/D Cooling HX A Bypass Valve.

A. (1) raised (2) open B. (1) raised (2) closed C. (1) lowered (2) open D. (1) lowered (2) closed

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible since HV-306 is throttled open to heatup (or slow the cooldown of) the RCS, however to raise the heatup rate of the SDCHX, HV-306 must be throttled closed.

B. Correct.

C. First part is plausible since the SDCHX is heating up at a rate of ~ 13°F/min, and the C/D rate limit for the RCS in MODE 4 is 100°F/hr (~ 1.6°F/min), which is could be assumed is the same temperature change limit for the SDCHX, however the heatup rate limit for the SDCHX is 19°F/min. Second part is plausible since opening HV-306 would lower the heatup rate, however in this case, HV-306 needs to be throttled closed to raise the heatup rate of the SDCHX.

D. First part is plausible since the SDCHX is heating up at a rate of ~ 13°F/min, and the C/D rate limit for the RCS in MODE 4 is 100°F/hr (~ 1.6°F/min), which is could be assumed is the same temperature change limit for the SDCHX, however the heatup rate limit for the SDCHX is 19°F/min. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference Y Attached picture of the SDC Train A Inlet Temperature Provided:

Learning Objective: 21607 - Describe the temperature requirements and their their bases for initiating and securing SDC.

Technical

Reference:

40OP-9SI01, Shutdown Cooling Initiation Technical

Reference:

Safety Injection System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Residual Heat Removal: Ability to verify that the Tier 2 alarms are consistent with the plant conditions Group 1 K/A 005 G 2.4.46 IR 4.2 Question 32 Given the following conditions:

Unit 2 is in MODE 5 Train A SDC is in service using A LPSI Pump Subsequently:

The A LPSI Pump tripped due to an 86 lockout The crew should be alerted of the loss of the A LPSI Pump by a ___(1)___ on the SESS Panel and annunciator 2B06A, SDC TRAIN A/B FLOW LO, ___(2)___

annunciate.

A. (1) white light AND a blue light (2) SHOULD B. (1) white light AND a blue light (2) should NOT C. (1) white light ONLY (2) SHOULD D. (1) white light ONLY (2) should NOT

Proposed Answer: D Explanations:

A. First part is plausible since the blue SESS alarm indicates that a piece of equipment which should be running is not running, however this is only for ESF equipment which is running due to an ESF actuation. Second part is plausible since the trip of the LPSI pump will result in a loss of SDC flow, however in order for that alarm to annunciate, the SDC pump breaker must be closed, therefore on a loss of flow due to a pump trip, the SDC Train A/B Low Flow alarm does not come in.

B. First part is plausible since the blue SESS alarm indicates that a piece of equipment which should be running is not running, however this is only for ESF equipment which is running due to an ESF actuation. Second part is correct.

C. First part is correct. Second part is plausible since the trip of the LPSI pump will result in a loss of SDC flow, however in order for that alarm to annunciate, the SDC pump breaker must be closed, therefore on a loss of flow due to a pump trip, the SDC Train A/B Low Flow alarm does not come in.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2018 Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 19358 - Discuss the Shutdown Cooling Low Flow Alarms

Technical

Reference:

LOIT Safety Injection Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Core Cooling: Ability to evaluate plant Tier 2 performance and make operational judgments based on Group 1 operating characteristics, reactor behavior, and instrument interpretation K/A 006 G 2.1.7 IR 4.4 Question 33 Given the following conditions:

A LOCA is in progress on Unit 2 The crew has entered 40EP-9EO03, Loss of Coolant Accident Containment pressure is 6.5 psig and rising at 1 psig/min Per 40DP-9AP16, Emergency Operating Procedure Users Guide:

(1) When Containment pressure approaches the CSAS setpoint the crew should (2) When RWT level approaches the RAS setpoint the crew should A. (1) let CSAS actuate automatically (2) let RAS actuate automatically B. (1) let CSAS actuate automatically (2) manually actuate RAS on trend C. (1) manually actuate CSAS on trend (2) let RAS actuate automatically D. (1) manually actuate CSAS on trend (2) manually actuate RAS on trend

Proposed Answer: C Explanations:

A. First part is plausible because a RAS is an ECCS actuation that needs to be automatically actuated. However, the reason for that is to make sure that there is enough inventory in Containment for the RAS. If a CSAS is imminent, it should be manually actuated on trend per EOP Operations Expectations. Second part is correct.

B. First part is plausible because a RAS is an ECCS actuation that needs to be automatically actuated. However, the reason for that is to make sure that there is enough inventory in Containment for the RAS. If a CSAS is imminent, it should be manually actuated on trend per EOP Operations Expectations. Second part is plausible because every other ECCS actuation should be manually actuated prior to the auto setpoint.

C. Correct D. First part is correct. Second part is plausible because every other ECCS actuation should be manually actuated prior to the auto setpoint.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 7 Reference N Provided:

Learning Objective: 24932 - Given conditions of a LOCA, describe the problems associated with initiating a RAS early per 40EP-9EO03

Technical

Reference:

Emergency Operating Procedure Users Guide Technical

Reference:

40DP-9AP08, Loss of Coolant Accident Technical Guideline Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Relief/Quench Tank: Ability to Tier 2 manually operate and/or monitor in the control room: Group 1 Relationships between PZR level and changing levels of the PRT and bleed holdup tank K/A 007 A4.09 IR 2.5 Question 34 Given the following conditions:

Unit 1 Reactor was tripped due to a Pressurizer relief valve stuck full open One minute after the Reactor trip and with NO operator action, Pressurizer level should be ___(1)___ and ___(2)___ level should be rising.

A. (1) rising (2) EDT B. (1) rising (2) RDT C. (1) lowering (2) EDT D. (1) lowering (2) RDT

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because letdown relief valves discharge to the EDT.

Most auxiliary systems will discharge to the EDT while any identified RCS leakage will collect in the RDT.

B. Correct C. First part is plausible because RCS leakage that occurs anywhere but the Pressurizer will cause Pressurizer level to lower. If the leak is anywhere in the steam space, Pressurizer level will rise.

Second part is plausible because letdown relief valves discharge to the EDT. Most auxiliary systems will discharge to the EDT while any identified RCS leakage will collect in the RDT.

D. First part is plausible because RCS leakage that occurs anywhere but the Pressurizer will cause Pressurizer level to lower. If the leak is anywhere in the steam space, Pressurizer level will rise.

Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 5 Reference N Provided:

Learning Objective: 26636 - Given conditions of LOCA, describe how the plant would respond to various types of RCS leaks per 40EP-9EO03

Technical

Reference:

40DP-9AP08, Loss of Coolant Accident Technical Guideline Technical

Reference:

Chemical and Volume Control System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Component Cooling Water: Knowledge of the Tier 2 physical connections and/or cause-effect relationships Group 1 between the CCWS and the following systems: RCS, in order to determine source(s) of RCS leakage into the K/A 008 K1.04 CCWS IR 3.3 Question 35 Given the following conditions:

Unit 2 is operating at 100% power Train A Essential Cooling Water is cross-tied to Nuclear Cooling Water Subsequently:

A small leak occurred in a RCP High Pressure Seal Cooler In this condition, which of the following process radiation monitors should be able to detect the resultant activity?

1. RU-2, Train A Essential Cooling Water

2. RU-3, Train B Essential Cooling Water

3. RU-6, Nuclear Cooling Water A. 1 ONLY B. 1 AND 2 ONLY C. 1 and 3 ONLY D. 1, 2, and 3

Proposed Answer: A Explanations:

A. Correct B. RU-3 is plausible if thought that the trains of Essential Cooling Water shared a common header, however when one train of EW is supplying priority loads, only the associated train will detect activity due to RCS leakage.

C. Plausible since RU-2 will be able to detect activity from an RCS leak, and plausible that RU-6 is located upstream of the NC-EW cross-tie valves, however RU-6 is isolated when the cross-tie is performed.

D. RU-2 is correct. RU-3 is plausible if thought that the trains of Essential Cooling Water shared a common header, however when one train of EW is supplying priority loads, only the associated train will detect activity due to RCS leakage, and plausible that RU-6 is located upstream of the NC-EW cross-tie valves, however RU-6 is isolated when the cross-tie is performed.

Question Source: New X Bank Modified X Previous NRC Exam 2019 Q42 Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 11 Reference Provided: N Learning Objective: 66723 - Given a Radiation Monitor number and name, describe the purposes and sample points of the Radiation Monitors at PVNGS

Technical

Reference:

LOIT Nuclear Cooling Water Lesson Plan Technical

Reference:

This picture is for one train of EW - the A Train uses RU-2, the B Train uses RU-3they do not share any common headers:

Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Pressure Control: Knowledge of PZR Tier 2 PCS design feature(s) and/or interlock(s) which provide Group 1 for the following: Over pressure control K/A 010 K4.03 IR 3.8 Question 36 To aid in protecting the Pressurizer from an over pressure condition, the Main Spray Valves are designed to be FULLY OPEN if RCS pressure rises to a MINIMUM of

___(1)___ and Pressurizer Backup Heaters are designed to trip if RCS pressure rises to a MINIMUM of ___(2)___ .

A. (1) 2300 psia (2) 2285 psia B. (1) 2300 psia (2) 35 psia above the setpoint of Pressure Master Controller, RCN-PIC-100 C. (1) 50 psia above the setpoint of Pressure Master Controller, RCN-PIC-100 (2) 2285 psia D. (1) 50 psia above the setpoint of Pressure Master Controller, RCN-PIC-100 (2) 35 psia above the setpoint of Pressure Master Controller, RCN-PIC-100

Proposed Answer: C Explanations:

A. First part is plausible since the normal setpoint for PIC-100 is 2250 psia, which would result in the Main Spray Valves being full open as soon as RCS pressure reached 2300 psia, however the actual design of the system is for Main Spray Valves to be full open as soon as RCS pressure is 50 psia above the setpoint on PIC-100. Second part is correct.

B. First part is plausible since the normal setpoint for PIC-100 is 2250 psia, which would result in the Main Spray Valves being full open as soon as RCS pressure reached 2300 psia, however the actual design of the system is for Main Spray Valves to be full open as soon as RCS pressure is 50 psia above the setpoint on PIC-100. Second part is plausible since 2285 psia is 35 psia above the normal setpoint of PIC-100, however all backup heaters trip at 2285 psia, regardless of PIC-100 setpoint.

C. Correct.

D. First part is correct. Second part is plausible since 2285 psia is 35 psia above the normal setpoint of PIC-100, however all backup heaters trip at 2285 psia, regardless of PIC-100 setpoint.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 22581 - Describe the automatic features associated with the Pressurizer Pressure Control System Bistables

Technical

Reference:

Pressurizer Pressure Control System Tech Manual Technical

Reference:

Pressurizer Pressure Control System Tech Manual Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Pressure Control: Knowledge of the Tier 2 operational implications of the following concepts as the Group 1 apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables K/A 010 K5.01 IR 3.5 Question 37 Given the following conditions:

Unit 2 is recovering from an ESD on SG #2 Inaction from the crew has caused the Pressurizer to go solid Subsequently:

The crew is drawing a bubble in the Pressurizer per 40EP-9EO05, Excess Steam Demand 1A and 2A RCPs are running Pressurizer pressure is 1800 psia Pressurizer temperature is 610°F (1) The Pressurizer currently (2) Per 40DP-9AP10, Excess Steam Demand Technical Guideline, if RCP flow is not maintained, the most UNDESIRABLE place for bubble formation is in the A. (1) has a bubble (2) Steam Generator U-Tubes B. (1) has a bubble (2) Reactor Vessel Upper Head C. (1) is in a water solid condition (2) Steam Generator U-Tubes D. (1) is in a water solid condition (2) Reactor Vessel Upper Head

Proposed Answer: C Explanations:

A. First part is plausible if a candidate thinks that a saturated system exists when pressure is greater than the pressure listed in the steam tables. Second part is correct.

B. First part is plausible if a candidate thinks that a saturated system exists when pressure is greater than the pressure listed in the steam tables. Second part is plausible because it is not desirable to have voiding in the Reactor Vessel Head. However, it is not a problem if there is RCP flow or natural circulation.

C. Correct D. First part is correct. Second part is plausible because it is not desirable to have voiding in the Reactor Vessel Head. However, it is not a problem if there is RCP flow or natural circulation.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 5 Reference N Provided:

Learning Objective: 25498 - Given the EOPs are being performed, describe how the operator will diagnose water solid conditions per 40EP-9EO05

Technical

Reference:

Steam Tables Technical

Reference:

40EP-9EO05, Excess Steam Demand Technical

Reference:

40DP-9AP10, Excess Steam Demand Technical Guideline Examination Outline Cross-

Reference:

Level RO SRO K/A: Reactor Protection: Knowledge of bus power Tier 2 supplies to the following: RPS channels, components, Group 1 and interconnections K/A 012 K2.01 IR 3.3 Question 38 Continuous power DIRECTLY to RPS Matrix Logic is supplied from ___(1)___ via

___(2)___.

A. (1) 120 VAC Class buses (2) auctioneering diodes B. (1) 120 VAC Class buses (2) static transfer switches C. (1) 125 VDC Class buses (2) auctioneering diodes D. (1) 125 VDC Class buses (2) static transfer switches

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because the power source is PN. PN busses use a static transfer switch to maintain power.

C. First part is plausible because 120 VDC is used for control power for Reactor Trip Circuit Breakers which is also part of the Plant Protection System. Second part is correct.

D. First part is plausible because 120 VDC is used for control power for Reactor Trip Circuit Breakers which is also part of the Plant Protection System. Second part is plausible because the power source is PN. PN busses use a static transfer switch to maintain power.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 8 Reference N Provided:

Learning Objective: 18804 - Describe how matrix logic receives electrical power

Technical

Reference:

Plant Protection System Tech Manual Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Reactor Protection: Knowledge of RPS design Tier 2 feature(s) and/or interlock(s) which provide for the Group 1 following: Logic matrix testing K/A 012 K4.08 IR 2.8 Question 39 Given the following conditions:

Unit 3 is performing logic matrix testing of the RPS system All Channel C RPS parameters have been placed in BYPASS Testing on Channel C is complete If a Channel B parameter is taken to BYPASS BEFORE the corresponding Channel C is removed from BYPASS, the Channel C parameter should ___(1)___ and the Channel B parameter should ___(2)___ .

A. (1) remain in BYPASS (2) be in BYPASS B. (1) remain in BYPASS (2) NOT go to BYPASS C. (1) come out of BYPASS (2) be in BYPASS D. (1) come out of BYPASS (2) NOT go to BYPASS

Proposed Answer: C Explanations:

A. First part is plausible because if Channel A was in bypass it would remain in bypass because it is a higher priority channel. However, since Channel C is a lower priority channel, it will come out of bypass. Second part is correct.

B. First part is plausible because if Channel A was in bypass it would remain in bypass because it is a higher priority channel. However, since Channel C is a lower priority channel, it will come out of bypass. Second part is plausible because if Channel A was in bypass, Channel B would not go to bypass since it is higher priority channel. However, since Channel C is a lower priority channel, Channel B will go into bypass.

C. Correct D. First part is correct. Second part is plausible because if Channel A was in bypass, Channel B would not go to bypass since it is higher priority channel. However, since Channel C is a lower priority channel, Channel B will go into bypass.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 18776 - Describe the Matrix Testing Interlock associated with the RPS

Technical

Reference:

Plant Protection System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Engineered Safety Features Actuation: Ability to Tier 2 predict and/or monitor changes in parameters (to Group 1 Prevent exceeding design limits) associated with operating the ESFAS controls including: Containment K/A 013 A1.02 pressure, temperature, and humidity IR 3.9 Question 40 Given the following conditions A LOCA inside containment is in progress Containment pressure is 5 psig and rising Containment temperature is 130°F and rising Given the current conditions with NO OPERATOR ACTION, a CSAS ___(1)___

occurred and if parameters continue to rise will reach a harsh condition AS SOON AS Containment temperature reaches ___(2)___°F.

A. (1) HAS (2) 170 B. (1) HAS (2) 235 C. (1) has NOT (2) 170 D. (1) has NOT (2) 235

Proposed Answer: C Explanations:

A. First part is plausible because at 3 psig SIAS, CIAS, and MSIS all automatically actuate. Second part is correct.

B. First part is plausible because at 3 psig SIAS, CIAS, and MSIS all automatically actuate. Second part is plausible because 235°F is the expected Containment temperature during a LOCA.

C. Correct D. First part is correct. Second part is plausible because 235°F is the expected Containment temperature during a LOCA.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 7 Reference N Provided:

Learning Objective: 22587 - Describe what automatically initiates the Containment Spray Actuation System (CSAS) and its function

Technical

Reference:

40AL-9RK5B, Panel B05B Alarm Responses Technical

Reference:

EOP Setpoints Document Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment Cooling: Knowledge of the Tier 2 operational implications of EOP warnings, cautions, and Group 1 notes K/A 022 G 2.4.20 IR 3.8 Question 41 Given the following conditions:

Unit 1 was tripped due to a loss of all Feedwater The CRS entered 40EP-9EO01, Standard Post Trip Actions Containment temperature is 120°F and slowly rising The BOP is performing step 9 of SPTAs and reports that NO Containment ACUs and NO Normal Chillers are running Per EOP Operations Expectations, during SPTAs the BOP should start ___(1)___ of Containment ACUs and ___(2)___ Large Normal Chiller(s).

A. (1) one train (2) one B. (1) one train (2) two C. (1) both trains (2) one D. (1) both trains (2) two

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible if it is assumed that because Containment temperature is not meeting the required temperature in SPTAs and it is rising, more than one chiller should be started to aid in restoring temperature.

C. First part is plausible because Containment temperature is not meeting the required temperature in SPTAs and it is rising, therefore more than one Containment ACU should be started to aid in restoring temperature. Second part is correct.

D. First part is plausible because Containment temperature is not meeting the required temperature in SPTAs and it is rising, therefore more than one Containment ACU should be started to aid in restoring temperature. Second part is plausible to start two large Normal Chillers to support two trains of Containment ACUs Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference N Provided:

Learning Objective: 22504 - Given plant conditions following a Reactor trip, analyze whether the Containment Temperature, Pressure and Combustible Gas Control Safety Function is met and what contingency actions are required if it is not in accordance with 40EP-9EO01

Technical

Reference:

EOP Operations Expectations Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment Spray: Knowledge of bus power Tier 2 supplies to the following: MOVs Group 1 K/A 026 K2.02 IR 2.7 Question 42 The feeder breaker to Containment Spray Header Discharge Valve, SIA-UV-672, is located on which of the following panels?

A. PNA-D26 B. PKA-M41 C. PHA-M35 D. PGA-L35

Proposed Answer: C Explanations:

A. Plausible because the Containment Spray discharge valve is very important to safety during a LOCA or ESD, therefore should be powered from an inverter that has a backup ac power supply from a voltage transformer B. Plausible because the Containment Spray discharge valve is very important to safety during a LOCA or ESD, therefore should have a power supply that has a battery charger and a battery backup.

C. Correct D. Plausible because the Containment Spray discharge valve is very important to safety during a LOCA or ESD and can be de-energized by a loss of PBA-S03 or PGA-L35. However, the feeder breaker for the valve is located on PHA-M35.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 8 Reference N Provided:

Learning Objective: 23844 - Identify the power supplies to SI related equipment

Technical

Reference:

40AO-9ZZ12, Degraded Electrical Power Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment Spray: Ability to manually operate Tier 2 and/or monitor in the control room: CSS controls Group 1 K/A 026 A4.01 IR 4.5 Question 43 Given the following conditions:

An inadvertent Train B CSAS has occurred The CRS entered 40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations The B Containment Spray pump was stopped The B Containment Spray header isolation valves were closed Prior to the CSAS being reset:

ESF Service Transformer NBN-X04 tripped B EDG energized PBB-S04 B Containment Spray pump A. starts because the CSAS signal resets.

B. starts because the Train B Load Sequencer goes through Mode 0.

C. does not start because the pump is overridden to the STOP position D. does not start because the breaker remains in the anti-pump condition until control power is cycled

Proposed Answer: B Explanations:

A. Plausible if it is thought that the CSAS signal will be reset if there is a loss of power to class bus that powers all of the Containment Spray equipment B. Correct C. Plausible if this pump was overridden, however it is taken to Stop and is anti-pumped D. Plausible if it is thought that because it is anti-pumped it cannot automatically be restarted, however the load sequencer will restart the pump Question Source: New X Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 23824 - Explain how the Load Sequencer changes between the different modes of operation

Technical

Reference:

BOP ESFAS System Tech Manual Technical

Reference:

LOIT BOP ESFAS Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Main and Reheat Steam: Knowledge of the Tier 2 operational implications of the following concepts as the Group 1 apply to the MRSS: Effect of steam removal on reactivity K/A 039 K5.03 IR 3.6 Question 44 Given the following conditions:

Unit 2 is operating at 100% power Subsequently:

The 6A Feedwater Heater Normal Control Valve has failed closed The 6A Feedwater Heater High Level Control Valve is seized closed With NO operator action, Reactor power should INITIALLY ___(1)___ due to

___(2)___.

A. (1) rise (2) a decrease in feedwater heating B. (1) rise (2) an increase in steam being sent to the Main Turbine C. (1) lower (2) a decrease in feedwater heating D. (1) lower (2) an increase in steam being sent to the Main Turbine

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible since the failures in the stem would result in the extraction steam valve to the 6A heater closing, thus diverting steam to the low pressure turbine, however this will not impact reactor power since the steam leaving the SGs will be unaffected.

C. Plausible since hot water in the 6A heater can no longer be rejected to the condenser (due to the normal level control valve failing closed) which would potentially increase the amount of hot water available to be sent to the SG, however the 6A heater will have steam isolated to it resulting in a lower temperature and a net decrease in feedwater heating.

D. Plausible since the isolation of extraction steam to the 6A heater will result in a lower feedwater temperature, which could cause more extraction steam to be aligned to other heaters to compensate for the reduction in feedwater heating (and thus taking steam which could have gone to the main turbine), however when the extraction steam to the 6A heater is stopped, the steam is diverted to the low pressure turbine.

Question Source: New X Bank Modified X Previous NRC Exam 2018 Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 14 Reference N Provided:

Learning Objective: 18420 - Explain the operation of the High Pressure Feedwater Heaters under normal operating conditions

Technical

Reference:

ICES Report #415578 (Palo Verde Unit 2 - May 2017)

Technical

Reference:

Examination Outline Cross-

Reference:

Level RO SRO K/A: Main Feedwater: Ability to (a) predict the impacts Tier 2 of the following malfunctions or operations on the MFW; Group 1 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those K/A 059 A2.12 malfunctions or operations: Failure of feedwater IR 3.1 regulating valves Question 45 Given the following conditions:

Unit 3 Reactor was tripped for a Refueling Outage TCOLD is 571°F and rising RRS TAVE has failed to 550°F Based on the RRS failure, the DFWCS should automatically ___(1)___ and to mitigate the condition the BOP should ___(2)___.

A. (1) stop feeding (2) adjust MFP speed B. (1) stop feeding (2) take MANUAL control of downcomer valves C. (1) be feeding at the maximum rate (2) adjust MFP speed D. (1) be feeding at the maximum rate (2) take MANUAL control of downcomer valves

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because raising or lowering MFP speed would normally increase/decrease Feedflow, however with this malfunction the downcomer valves will be closed so changing MFP speed will not do anything.

B. Correct C. First part is plausible if RTO is fed from TCOLD and not TAVE. Second part is plausible because raising or lowering MFP speed would normally increase/decrease Feedflow, however with this malfunction the downcomer valves will be closed so changing MFP speed will not do anything.

D. First part is plausible if RTO is fed from TCOLD and not TAVE. Second part is correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 31226 - Describe the response of the Reactor Regulating System to a failure of a Temperature Transmitter input

Technical

Reference:

Feedwater Control System Tech Manual Technical

Reference:

40AL-9RK6A, Panel B06A Alarm Responses Examination Outline Cross-

Reference:

Level RO SRO K/A: Auxiliary/Emergency Feedwater: Knowledge of the Tier 2 effect of a loss or malfunction of the following will have Group 1 on the AFW components: Controllers and positioners K/A 061 K6.01 IR 2.5 Question 46 Given the following conditions:

Unit 1 is in MODE 2 at 1% power during a startup AFN-P01 is feeding both Steam Generators via the Feedwater Isolation bypass valves SGN-HV-1143 and SGN-HV-1145 Subsequently:

An inadvertent SIAS occurs With NO operator action, SGN-HV-1143 and SGN-HV-1145 should fail ___(1)___ and AFN-P01 ___(2)___ running.

A. (1) closed (2) is B. (1) closed (2) is not C. (1) as-is (2) is D. (1) as-is (2) is not

Proposed Answer: D Explanations:

A. First part is plausible because on a SIAS, NC cross-tie valves will fail closed. If valves fail as-is, the Feedrate will no longer be controlled by an operator in the Control Room. It is reasonable that at low power levels when this valve is used, the valve will fail closed until manual operation can be restored..Second part is plausible because if AFB-P01 was running, it would be stripped and then restarted during a SIAS. AFN-P01 is stripped but not restarted.

B. First part is plausible because on a SIAS, NC cross-tie valves will fail closed. If valves fail as-is, the Feedrate will no longer be controlled by an operator in the Control Room. It is reasonable that at low power levels when this valve is used, the valve will fail closed until manual operation can be restored. Second part is correct.

C. First part is correct. Second part is plausible because if AFB-P01 was running, it would be stripped and then restarted during a SIAS. AFN-P01 is stripped but not restarted.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference X Provided:

Learning Objective: 24524 - Describe the Control Room controls associated with the Non Essential Auxiliary Feedwater Pump AFN-P01 including its indications

Technical

Reference:

40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations Technical

Reference:

40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations Technical

Reference:

40AO-9ZZ12, Degraded Electrical Power Examination Outline Cross-

Reference:

Level RO SRO K/A: AC Electrical Distribution: Knowledge of the effect Tier 2 that a loss or malfunction of the AC Distribution system Group 1 will have on the following: DC system K/A 062 K3.03 IR 3.7 Question 47 Assuming the battery room is maintained a minimum of 60°F during a Station Blackout, with NO operator action, the Class 1E batteries should supply DC system loads for a MINIMUM of A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

Proposed Answer: B Explanations:

A. Plausible because in 40EP-9EO08, Blackout there is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time requirement to start and place a Station Blackout Generator on a class bus within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if power is not available from offsite or an EDG.

B. Correct C. Plausible because in 40EP-9EO08, Blackout there is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time requirement to start a cooldown if offsite power or an EDG is not restored to a class bus.

D. Plausible because in 40EP-9EO08, Blackout there is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time requirement to align cooling to the Spent Fuel Pool if power has not been restored to a class bus with offsite power or an EDG.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 8 Reference X Provided:

Learning Objective: 18175 - Explain the operation of the Class 1E 125 VDC Batteries under normal operating conditions

Technical

Reference:

Class 125 VDC Power System Tech Manual Technical

Reference:

40EP-9EO08, Blackout Technical

Reference:

40EP-9EO08, Blackout Technical

Reference:

40EP-9EO08, Blackout Examination Outline Cross-

Reference:

Level RO SRO K/A: DC Electrical Distribution: Knowledge of the Tier 2 physical connections and/or cause effect relationships Group 1 between the DC electrical systems and the following systems: AC electrical system K/A 063 K1.02 IR 2.7 Question 48 Given the following conditions:

Unit 2 is operating at 100% power.

Inverter PNC-N13 Manual Bypass Switch is in the Normal Operation position The supply breaker to inverter PNC-N13 was inadvertently opened at PKC-M43 Based on these conditions, PNC-D27 should A. NOT automatically align to its alternate power supply. Power can be restored by manually pressing the Bypass Source to Load pushbutton.

B. automatically align to its alternate power supply and should automatically transfer back to its normal source when the inverter is re-energized.

C. NOT automatically align to its alternate power supply. Power can be restored by manually placing the Manual Bypass Switch to the Bypass to Load position.

D. automatically align to its alternate power supply and can be manually realigned to its normal source when the inverter is re-energized by pressing the Inverter to Load pushbutton.

Proposed Answer: D Explanations:

A. Plausible that it will NOT auto align to the alternate source because unit 1 once did not have static switches with automatic switching capabilities. Also, the examinee may very well think that the Bypass Source to Load pushbutton reverses the last transfer, which would realign the bus to the normal source.

B. Plausible since it will auto transfer to the alternate source, however it will not auto transfer back to the normal source.

C. Plausible that it will NOT auto align to the alternate source because unit 1 once did not have static switches with automatic switching capabilities.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 378660 - Explain the operation of the Static Switch provided on Ametek Inverters (new)

Technical

Reference:

120 VAC Class 1E Instrument Power (PN) Lesson Plan Technical

Reference:

120 VAC Class 1E Instrument Power (PN) Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Diesel Generator: Ability to monitor Tier 2 automatic operation of the ED/G system, including: Group 1 Load Sequencing K/A 064 A3.07 IR 3.6 Question 49 Given the following conditions:

Unit 1 has tripped due to a LOOP Which of the following loads should automatically start after the EDGs start?

1. A Auxiliary Feedwater Pump

2. B Auxiliary Feedwater Pump

3. N Auxiliary Feedwater Pump A. 1 ONLY B. 2 ONLY C. 1 AND 3 ONLY D. 2 AND 3 ONLY

Proposed Answer: B Explanations:

A. Plausible because AFA-P01 will start on an AFAS. However, the only pump that starts on a LOOP is AFB-P01.

B. Correct C. First part is plausible because AFA-P01 will start on an AFAS. Second part is plausible if it is thought that both electrical pumps start because if there is a LOOP and a LOP on PBB-S04, there will be no auxiliary feed pumps that automatically start.

D. First part is correct. Second part is plausible if it is thought that both electrical pumps start because if there is a LOOP and a LOP on PBB-S04, there will be no auxiliary feed pumps that automatically start.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 23823 - Explain the operation of the ESF Load Sequencer

Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Process Radiation Monitoring: Ability to predict Tier 2 and/or monitor changes in parameters (to prevent Group 1 exceeding design limits) associated with operating the PRM system controls including: Radiation levels K/A 073 A1.01 IR 3.2 Question 50 Given the following conditions:

Unit 1 is operating at 100% power A Steam Generator Tube Leak on SG #1 ___(1)___ cause rising radiation levels on SG #2 RU-142 N-16 Main Steam Line Radiation Monitor and once a downpower is started, INDICATED leak rates on RMS should ___(2)___.

A. (1) SHOULD (2) lower B. (1) SHOULD (2) remain the same C. (1) should NOT (2) lower D. (1) should NOT (2) remain the same

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible if it is thought that because the leak itself has not changed, than indicated leak rate should remain the same.

C. First is plausible because the steam from SG #2 will not have any activity. However the radiation monitor will detect the radioactivity from SG #1. Second part is correct.

D. First is plausible because the steam from SG #2 will not have any activity. However the radiation monitor will detect the radioactivity from SG #1. Second part is plausible if it is thought that because the leak itself has not changed, than indicated leak rate should remain the same.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 11 Reference N Provided:

Learning Objective: 31204 - Explain the basic operation of Process Radiation Monitors

Technical

Reference:

74AL-9SQ01, Radiation Monitoring System Alarm Validation and

Response

Examination Outline Cross-

Reference:

Level RO SRO K/A: Service Water: Knowledge of SWS design Tier 2 feature(s) and/or interlock(s) which provide for the Group 1 following: Conditions initiating automatic closure of closed cooling water auxiliary building header supply K/A 076 K4.01 and return valves IR 2.5 Question 51 Given the following conditions:

Unit 2 is operating at 100% power Train A EW-NC Cross-Tie Supply and Return Valves, EWA-UV-65 and EWA-UV-145, are open in support of Train A EW cross-tied with NC Which of the following conditions, INDIVIDUALLY, should result in the automatic closure of EWA-UV-65 and EWA-UV-145?

1. Inadvertent Train A SIAS

2. Inadvertent Train A CSAS

3. Low Level in the A EW Surge Tank A. 2 ONLY B. 3 ONLY C. 1 and 2 ONLY D. 1 and 3 ONLY

Proposed Answer: D Explanations:

A. Plausible that an A CSAS would be correct since an A CSAS will stop EW flow through the aux building, however the CSAS stops flow through the aux building by closing the NC CIVs, not the EW-NC cross-tie valves.

B. Plausible since a low level in the A EW Surge Tank will close the EW-NC cross-tie valves, however an A SIAS will also close the EW-NC cross-tie valves.

C. Plausible since an A SIAS will close the EW-NC cross-tie valves, and an A CSAS will stop flow through the aux building, however the A CSAS will not close the EW-NC cross-tie valves.

Additionally, low level in the A EW Surge Tank will close the EW-NC cross-tie valves.

D. Correct.

Question Source: New Bank X Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 7 Reference N Provided:

Learning Objective: 18454 - Describe the automatic functions associated with the Essential Cooling Water Cross-tie to Nuclear Cooling Water Valves EWA-UV-145 and EWA-UV-65

Original Question: 2018 NRC Exam Q35 (correct answer was B)

Technical

Reference:

Essential Cooling Water Tech Manual Technical

Reference:

Essential Cooling Water Tech Manual

Technical

Reference:

40AO-9ZZ17, Inadvertent PPS-ESFAS Actuations Examination Outline Cross-

Reference:

Level RO SRO K/A: Service Water: Ability to (a) predict the impacts of Tier 2 the following malfunctions or operations on the SWS; Group 1 and (b) based on those predictions use procedures to correct, control, or mitigate the consequences of those K/A 076 A2.02 malfunctions or operations: Service water header IR 2.7 pressure Question 52 Given the following conditions:

Unit 1 is in MODE 5 SDC is in service using Train A auxiliaries and the A LPSI Pump A tube leak in the A Essential Cooling Water Heat Exchanger has just occurred (1) The tube leak in the EW Heat Exchanger should send water from the (2) If the EW Pump is stopped in response to the tube leak, the in-service SDCHX can be cooled directly from the A. (1) Essential Cooling Water System to the Spray Pond Cooling Water System (2) Nuclear Cooling Water System B. (1) Essential Cooling Water System to the Spray Pond Cooling Water System (2) Spray Pond Cooling Water System C. (1) Spray Pond Cooling Water System to the Essential Cooling Water System (2) Nuclear Cooling Water System D. (1) Spray Pond Cooling Water System to the Essential Cooling Water System (2) Spray Pond Cooling Water System

Proposed Answer: C Explanations: Part (b) of the KA is met by knowing the procedure to use to restore cooling to the SDCHX. The reason the procedure to use was not included in the stem of the question is that the name of the procedure would give the correct answer away (Appendix 243, NC Cross-Tie to EW Train A)

A. First part is plausible since nominal system pressure of the SP system is ~ 50-55 psig compared to EW which has a nominal system pressure of ~ 95 psig, however the EW system is designed such that at the EW heat exchanger, EW pressure is lower than SP pressure to ensure that in the event of an EW HX tube leak, leakage goes from the SP system to the EW system to minimize the potential for environmental contamination. Second part is correct.

B. First part is plausible since nominal system pressure of the SP system is ~ 50-55 psig compared to EW which has a nominal system pressure of ~ 95 psig, however the EW system is designed such that at the EW heat exchanger, EW pressure is lower than SP pressure to ensure that in the event of an EW HX tube leak, leakage goes from the SP system to the EW system to minimize the potential for environmental contamination. Second part is plausible since the Spray Pond system is the cooling medium for EW, and thus the ultimate heat sink for SDC, however the Spray Pond system cannot be directly lined up to the SDCHX.

C. Correct.

D. First part is correct. Second part is plausible since the Spray Pond system is the cooling medium for EW, and thus the ultimate heat sink for SDC, however the Spray Pond system cannot be directly lined up to the SDCHX.

Question Source: New x Bank - question was slightly modified but not to the point where the question can be classified as modified per NUREG 1021 Modified x Previous NRC Exam 2016 Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 18538 - Describe the design characteristics of the Essential Cooling Water Heat Exchangers

Technical

Reference:

Essential Cooling Water System Lesson Plan Technical

Reference:

40EP-9EO11, Lower Mode Functional Recovery Examination Outline Cross-

Reference:

Level RO SRO K/A: Instrument Air: Knowledge of the physical Tier 2 connections and/or cause-effect relationships between Group 1 the IAS and following systems: MSIV air K/A 078 K1.05 IR 3.4 Question 53 Given the following conditions:

Unit 2 is operating at 100% power An Instrument Air rupture has occurred just downstream of the IA compressors IA pressure is at atmospheric pressure throughout the system The nitrogen backup supply valve has failed closed Based on these conditions, the Main Steam Isolation Valves should A. slow close due to the loss of IA B. fast close due to the loss of IA C. remain open and can only be slow closed D. remain open and can only be fast closed

Proposed Answer: D Explanations:

A. Plausible that the MSIVs would fail closed as this is the fail safe position, and the valves are stoked open in slow speed and can be closed in slow speed, however the MSIVs remain open on a loss of instrument air B. Plausible that the MSIVs would fail closed as this is the fail safe position, and the valves are normally closed in fast speed, however the MSIVs remain open on a loss of instrument air.

C. Plausible since the MSIVs will remain open, however slow close is not available on a loss of instrument air.

D. Correct.

Question Source: New X Bank Modified X Previous NRC Exam 2019 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 4 Reference N Provided:

Learning Objective: 25935 - Determine the major effects on plant operation as instrument air pressure degrade

Technical

Reference:

40AO-9ZZ06, Loss of Instrument Air Examination Outline Cross-

Reference:

Level RO SRO K/A: Instrument Air: Ability to monitor automatic Tier 2 operation of the IAS, including: Air pressure Group 1 K/A 078 A3.01 IR 3.1 Question 54 During a leak on the Instrument Air header, the Nitrogen Backup Valve should automatically open AS SOON AS header pressure lowers to ___(1)___ psig and should re-close AS SOON AS header pressure rises to ___(2)___ psig.

A. (1) 85 (2) 105 B. (1) 85 (2) 115 C. (1) 95 (2) 105 D. (1) 95 (2) 115

Proposed Answer: A Explanations:

A. Correct.

B. First part is correct. Second part is plausible since 115 psig is the middle of the control band for normal IA header pressure (109-119), however the backup N2 valve closes when pressure rises to 105 psig.

C. First part is plausible since 95 psig is the setpoint for the IA header low pressure alarm, however the N2 backup valve doesnt open until 85 psig. Second part is correct.

D. First part is plausible since 95 psig is the setpoint for the IA header low pressure alarm, however the N2 backup valve doesnt open until 85 psig. Second part is plausible since 115 psig is the middle of the control band for normal IA header pressure (109-119), however the backup N2 valve closes when pressure rises to 105 psig.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference N Provided:

Learning Objective: 31231 - Describe the automatic functions associated with the Instrument Air System

Technical

Reference:

40AO-9ZZ06, Loss of Instrument Air Technical

Reference:

Instrument Air System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment: Knowledge of the effect that a loss Tier 2 or malfunction of the containment system will have on Group 1 the following: Loss of containment integrity under shutdown conditions K/A 103 K3.01 IR 3.3 Question 55 Given the following conditions:

Unit 3 is in MODE 4 A Containment vent is in progress Subsequently:

A malfunction causes Containment vent valves to be stuck open When the Containment vent valves are closed Containment pressure is -0.5 psig After the vent an AO reports that a Containment air lock inner door window has a crack causing it to be INOPERABLE To maintain compliance with Technical Specifications the crew should raise Containment pressure to a MINIMUM of ___(1)___ psig, and the MINIMUM REQUIRED action(s) is(are) to ___(2)___.

A. (1) -0.3 (2) verify the OPERABLE door is closed in the affected air lock ONLY B. (1) -0.3 (2) verify the OPERABLE door is closed in the affected air lock AND initiate action to evaluate overall containment leakage rate C. (1) 0.25 (2) verify the OPERABLE door is closed in the affected air lock ONLY D. (1) 0.25 (2) verify the OPERABLE door is closed in the affected air lock AND initiate action to evaluate overall containment leakage rate

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because evaluating containment leakage rate is the action take if the air lock is INOPERABLE for reasons other than Condition A or B.

C. First part is plausible because 0.25 is the pressure at which a Containment vent is stopped per 40OP-9CP01, Containment Purge System. Second part is correct.

D. First part is plausible because 0.25 is the pressure at which a Containment vent is stopped per 40OP-9CP01, Containment Purge System. Second part is plausible because evaluating containment leakage rate is the action take if the air lock is INOPERABLE for reasons other than Condition A or B.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 9 Reference N Provided:

Learning Objective: Given a set of plant conditions, apply the one hour or less actions statements of T.S. 3.6 in accordance with Tech Spec 3.6

Technical

Reference:

Technical Specifications Technical

Reference:

40OP-9CP01, Containment Purge System Technical

Reference:

Technical Specifications Technical

Reference:

Technical Specifications Examination Outline Cross-

Reference:

Level RO SRO K/A: Control Rod Drive: Ability to monitor automatic Tier 2 operation of the CRDS, including: RCS temperature and Group 2 pressure K/A 001 A3.06 IR 3.9 Question 56 Given the following conditions:

A MFP has tripped causing a RPCB All required CEA subgroups have been verified to have fully inserted into the core TAVG is 582°F TREF is 576°F Assuming TREF remains constant, Group 3 CEAs are currently inserting at a

___(1)___ rate and should STOP inserting AS SOON AS TAVG is less than ___(2)___

°F.

A. (1) low (2) 579 B. (1) low (2) 580.5 C. (1) high (2) 579 D. (1) high (2) 580.5

Proposed Answer: C Explanations:

A. First part is plausible because the CEAs will insert at a low rate once TAVG-TREF deviation is less than 4.5°F. The current deviation of 6°F will cause CEAs to insert at a high rate. Second part is correct.

B. First part is plausible because the CEAs will insert at a low rate once TAVG-TREF deviation is less than 4.5°F. The current deviation of 6°F will cause CEAs to insert at a high rate. Second part is plausible because at 580.5°F the CEAs will stop inserting at a high rate, however they will still be inserting at a low rate.

C. Correct D. First part is correct. Second part is plausible because at 580.5°F the CEAs will stop inserting at a high rate, however they will still be inserting at a low rate.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 6 Reference N Provided:

Learning Objective: 19485 - Describe the automatic functions/interlocks associated with CEDMCS

Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Reactor Coolant: Ability to manually operate Tier 2 and/or monitor in the control room: Indications Group 2 necessary to verify natural circulation from appropriate level, flow, and temperature indications and valve K/A 002 A4.02 positions upon loss of forced circulation IR 4.3 Question 57 Given the following conditions:

Unit 1 tripped from 100% power due to a loss of off-site power.

The crew is verifying natural circulation has been established.

As natural circulation flow develops, the crew should expect to see loop T indicating

___(1)___ 65°F and should expect a delay of approximately ___(2)___ before the RCS temperature response of feeding and steaming adjustments can be verified.

A. (1) less than (2) 1 to 2 minutes B. (1) less than (2) 5 to 15 minutes C. (1) greater than (2) 1 to 2 minutes D. (1) greater than (2) 5 to 15 minutes

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible since frequent adjustments of steaming and feeding are needed when controlling in manual (as is the case in a LOOP/LOFC) in order to maintain parameters within post-trip control bands, however in natural circulation conditions, the plant response to these adjustments will not be seen for ~ 5 to 15 minutes.

B. Correct C. First part is plausible since the driving head in natural circulation is developed by the difference in density between the hot and cold legs, therefore a higher delta-T than with forced circulation is plausible, however delta-T must be < 65°F (full power delta-T) in natural circulation conditions.

Second part is plausible since frequent adjustments of steaming and feeding are needed when controlling in manual (as is the case in a LOOP/LOFC) in order to maintain parameters within post-trip control bands, however in natural circulation conditions, the plant response to these adjustments will not be seen for ~ 5 to 15 minutes.

D. First part is plausible since the driving head in natural circulation is developed by the difference in density between the hot and cold legs, therefore a higher delta-T than with forced circulation is plausible, however delta-T must be < 65°F (full power delta-T) in natural circulation conditions.

Second part is correct.

Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 14 Reference N Provided:

Learning Objective: 26275 - Explain the difference between single phase and two phase natural circulation

Technical

Reference:

40EP-9EO07, Loss of Off Site Power / Loss of Forced Circulation Technical

Reference:

40DP-9AP12, Loss of Off Site Power / Loss of Forced Circulation Technical Guideline

Examination Outline Cross-

Reference:

Level RO SRO K/A: Rod Position Indication: Ability to locate control Tier 2 room switches, controls, and indications, and to Group 2 determine that they correctly reflect the desired plant lineup K/A 014 G 2.1.31 IR 4.6 Question 58 Given the following conditions:

Unit 3 was operating at 100% power when a Reactor Power Cutback occurred The CRS has entered 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump)

All automatic CEA motion has stopped The CRS has just directed the OATC to restore CEA overlap Current CEA positions are as follows:

Per 40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump), prior to commencing the restoration of CEA overlap, the OATC should ensure that the CEDMCS Mode Selector Switch is selected to ___(1)___ and the FIRST CEA Reg Group to be withdrawn should be ___(2)___ .

A. (1) Manual Group (2) Reg Group 3 B. (1) Manual Group (2) Reg Group 4

C. (1) Manual Sequential (2) Reg Group 3 D. (1) Manual Sequential (2) Reg Group 4

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible since Group 3 CEAs were the last to insert so it would make sense that they would be first to withdraw, however prior to withdrawing Group 3 CEAs, Group 4 CEAs must be withdrawn until they are within 95 inches of Group 3 CEAs.

B. Correct.

C. First part is plausible since CEAs will be restored to an ARO condition using manual sequential control, however when re-establishing CEA group overlap, manual group is used. Second part is plausible since Group 3 CEAs were the last to insert so it would make sense that they would be first to withdraw, however prior to withdrawing Group 3 CEAs, Group 4 CEAs must be withdrawn until they are within 95 inches of Group 3 CEAs.

D. First part is plausible since CEAs will be restored to an ARO condition using manual sequential control, however when re-establishing CEA group overlap, manual group is used. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 6 Reference N Provided:

Learning Objective: 19515 - Explain how electric control of the CEDMs is achieved

Technical

Reference:

40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump)

Technical

Reference:

40AO-9ZZ09, Reactor Power Cutback (Loss of Feedpump)

Examination Outline Cross-

Reference:

Level RO SRO K/A: In-Core Temperature Monitor: Knowledge of the Tier 2 effect of a loss or malfunction of the following ITM Group 2 system components: Sensors and detectors K/A 017 K6.01 IR 2.7 Question 59 Given the following conditions:

Unit 3 is operating at 100%

One Core Exit Thermocouple (CET) sensor has just failed out of range low The failure of this CET should be indicated on QSPDS by ___(1)___ and the input from the failed CET into the overall CET calculation should be ___(2)___ .

A. (1) NO DATA (2) ignored by QSPDS B. (1) NO DATA (2) replaced by a canned value C. (1) question marks (2) ignored by QSPDS D. (1) question marks (2) replaced by a canned value

Proposed Answer: C Explanations:

A. First part is plausible since invalid inputs on Plant PI are indicated by NO DATA, however QSPDS uses a string of question marks to indicate a failed sensor. Second part is correct.

B. First part is plausible since invalid inputs on Plant PI are indicated by NO DATA, however QSPDS uses a string of question marks to indicate a failed sensor. Second part is plausible as a canned value is used in the DFWCS for inputs that are out of range, however CET data that is out of range is simply ignored by QSPDS.

C. Correct.

D. First part is correct. Second part is plausible as a canned value is used in the DFWCS for inputs that are out of range, however CET data that is out of range is simply ignored by QSPDS.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 2 Reference N Provided:

Learning Objective: 19076 - Describe the Control Room indications associated with the QSPDS system

Technical

Reference:

QSPDS Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment Iodine Removal: Knowledge of the Tier 2 physical connections and/or cause effect relationships Group 2 between the CIRS and the following systems: CSS K/A 027 K1.01 IR 3.4 Question 60 The amount of gaseous iodine in the containment atmosphere is minimized during normal conditions by the use of ___(1)___ filters and is minimized during a LOCA by maintaining pH of the water in containment ___(2)___ 7.0.

A. (1) HEPA (2) less than B. (1) HEPA (2) greater than C. (1) charcoal (2) less than D. (1) charcoal (2) greater than

Proposed Answer: D Explanations:

A. First part is plausible since HEPA filters are used in several air filtration units throughout the plant and filter our micro particles from the air, however the iodine is filtered by use of charcoal filters.

Second part is plausible since the water injected into the core during a LOCA is a boric acid solution, and boric acid has a pH less than 7.0, however in order to maintain iodine in solution, trisodium phosphate is added to the water to raise the pH to greater than 7.0.

B. First part is plausible since HEPA filters are used in several air filtration units throughout the plant and filter our micro particles from the air, however the iodine is filtered by use of charcoal filters.

Second part is correct.

C. First part is correct. Second part is plausible since the water injected into the core during a LOCA is a boric acid solution, and boric acid has a pH less than 7.0, however in order to maintain iodine in solution, trisodium phosphate is added to the water to raise the pH to greater than 7.0.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 13 Reference N Provided:

Learning Objective: Explain the operation of the Containment Building Pre-Access Filtration AFUs (HCNF01A and B) under normal operating conditions

Technical

Reference:

Containment HVAC System Tech Manual Technical

Reference:

LOIT Safety Injection System Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Hydrogen Recombiner and Purge Control: Tier 2 Knowledge of bus power supplies to the following: Group 2 Hydrogen recombiners K/A 028 K2.01 IR 2.5 Question 61 The power supply to Hydrogen Recombiners is a ___(1)___ and ___(2)___.

A. (1) 480V Class Bus (2) is hardwired to the recombiners B. (1) 480V Class Bus (2) must be manually connected to the recombiners C. (1) 4.16 kV Class Bus (2) is hardwired to the recombiners D. (1) 4.16 kV Class Bus (2) must be manually connected to the recombiners

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because another component that is used for Containment Hydrogen during an accident, Hydrogen Analyzers, are hardwired into the electrical system.

B. Correct C. First part is plausible because the power supply is Class power, however it is only 480 VAC.

Second part is plausible because another component that is used for Containment Hydrogen during an accident, Hydrogen Analyzers, are hardwired into the electrical system.

D. First part is plausible because the power supply is Class power, however it is only 480 VAC.

Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 8 Reference N Provided:

Learning Objective: 18259 - Describe the response of the Class AC Distribution System to an abnormal/emergency operating condition

Technical

Reference:

LOIT Hydrogen Control System Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Steam Dump/Turbine Bypass Control: Knowledge Tier 2 of the effect that a loss or malfunction of the SDS will Group 2 have on the following: RCS K/A 041 K3.02 IR 3.8 Question 62 Given the following conditions:

Unit 2 is operating at 100% power NNN-D11 is de-energized Subsequently:

The Main Turbine trips The Reactor should trip on ___(1)___ and the crew should control RCS temperature with ___(2)___.

A. (1) High Pressurizer pressure - RPS (2) ADVs B. (1) High Pressurizer pressure - RPS (2) SBCV-1007 and 1008 C. (1) High Pressurizer pressure - SPS (2) ADVs D. (1) High Pressurizer pressure - SPS (2) SBCV-1007 and 1008

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because there are malfunction (loss of vacuum) that will allow for the use of SBCV-1007 and 1008. However the loss of NNN-D11 will cause a loss of power to SBCS and no valves can be operated automatically or manually.

C. First part is plausible because if the Reactor failed to trip on High Pressure at 2383 psia, the SPS trip will trip the Reactor at 2409 psia. Second part is correct.

D. First part is plausible because if the Reactor failed to trip on High Pressure at 2383 psia, the SPS trip will trip the Reactor at 2409 psia. Second part is plausible because there are malfunction (loss of vacuum) that will allow for the use of SBCV-1007 and 1008. However the loss of NNN-D11 will cause a loss of power to SBCS and no valves can be operated automatically or manually Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 4 Reference N Provided:

Learning Objective: 25753 - Given a loss of non-class instrument power, describe how the loss impacts the operation of SBCS in accordance with 40AO-9ZZ14

Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Main Turbine Generator: Ability to (a) predict the Tier 2 impacts of the following malfunctions or operation on Group 2 the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the K/A 045 A2.08 consequences of those malfunctions or operations: IR 2.8 Steam dumps are not cycling properly at low load, or stick open at higher load (isolate and use atmospheric reliefs when necessary)

Question 63 Given the following conditions:

Unit 3 is operating at 100% power Core life is MOC Subsequently:

The Main Turbine tripped 10 minutes after the Main Turbine Trip Reactor Power stabilizes at 60%

SBCV-1001 and SBCV-1004 are both FULL open (1) If automatic control of SBCV-1001 is lost and SBCV-1001 Mode Selector Switch is taken to OFF, the FIRST set of valves to modulate to pick up steam load is (2) With NO operator action, over the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the SBCVs that modulated open after SBCV-1001 failed should modulate in the A. (1) SBCV-1002 & SBCV-1005 (2) open direction B. (1) SBCV-1002 & SBCV-1005 (2) closed direction C. (1) SBCV-1003 & SBCV-1006 (2) open direction D. (1) SBCV-1003 & SBCV-1006 (2) closed direction

Proposed Answer: D Explanations:

A. The first part is plausible because SBCV-1002 & SBCV-1005 are the next valves numerically.

Second part is plausible because eventually xenon will decay away and the SBCVs will modulate open. However over the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, xenon will be building in.

B. The first part is plausible because SBCV-1002 & SBCV-1005 are the next valves numerically.

Second part is correct.

C. First Part is correct. Second part is plausible because eventually xenon will decay away and the SBCVs will modulate open. However over the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, xenon will be building in.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 1 Reference N Provided:

Learning Objective: 378910 - Describe the overall system operation of the Steam Bypass Control System

Technical

Reference:

Operator Information Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Waste Gas Disposal: Knowledge of design Tier 2 feature(s) and/or interlock(s) which provide for the Group 2 following: Isolation of waste gas release tanks K/A 071 K4.04 IR 2.9 Question 64 Given the following conditions:

Unit 2 is venting the RDT to the Waste Gas system Subsequently:

An inadvertent CIAS occurs The Waste Gas header should be isolated by ___(1)___ Containment Isolation valve(s) and if header pressure downstream of the Containment Isolation Valve(s) rises, there is a relief valve that should lift and relieve ___(2)___.

A. (1) one (2) DIRECTLY to the Radwaste Building Exhaust B. (1) one (2) to the Radwaste Building Exhaust via the Gaseous Discharge Header Release path C. (1) two (2) DIRECTLY to the Radwaste Building Exhaust D. (1) two (2) to the Radwaste Building Exhaust via the Gaseous Discharge Header Release path

Proposed Answer: C Explanations:

A. First part is plausible because there are other systems that are isolated by only one Containment Isolation valves (e.g RDT Makeup Valve CHA-UV-580). Second part is correct.

B. First part is plausible because there are other systems that are isolated by only one Containment Isolation valves (e.g RDT Makeup Valve CHA-UV-580). Second part is plausible because the Waste Gas Decay tanks are released through this path.

C. Correct D. First part is correct. Second part is plausible because the Waste Gas Decay tanks are released through this path.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 9 Reference N Provided:

Learning Objective: 20262 - Describe the Gaseous Release Flowpath

Technical

Reference:

Gaseous Radwaste System Tech Manual Technical

Reference:

LOIT Gaseous Radwaste Lesson Plan Examination Outline Cross-

Reference:

Level RO SRO K/A: Area Radiation Monitoring: Knowledge of the Tier 2 operational implications of the following concepts as Group 2 they apply to the ARM system: Radiation theory, including sources, types, units, and effects K/A 072 K5.01 IR 2.7 Question 65 Control Room Area Radiation Monitor, RU-18, measures ___(1)___ radiation and when it rises to the alarm setpoint ___(2)___ auto actuate CREFAS.

A. (1) neutron (2) SHOULD B. (1) neutron (2) should NOT C. (1) gamma (2) SHOULD D. (1) gamma (2) should NOT

Proposed Answer: D Explanations:

A. First part is plausible because neutron radiation is highly hazardous, however RU-18 measures gamma radiation. Second part is plausible because RU-18 does monitor radiation levels around the Control Room, however only RU-29 and RU-30 directly will automatically actuate CREFAS.

B. First part is plausible because neutron radiation is highly hazardous, however RU-18 measures gamma radiation. Second part is correct.

C. First part is correct. Second part is plausible because RU-18 does monitor radiation levels around the Control Room, however only RU-29 and RU-30 directly will automatically actuate CREFAS.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 11 Reference N Provided:

Learning Objective: 23911 - Explain the basic operation of Area Radiation Monitors

Technical

Reference:

LOIT Radiation Monitoring Lesson Plan Technical

Reference:

Radiation Monitoring System Tech Manual Technical

Reference:

Radiation Monitoring System Tech Manual Examination Outline Cross-

Reference:

Level RO SRO K/A: Conduct of Operations: Knowledge of shift or Tier 3 short-term relief turnover practices Group K/A G 2.1.3 IR 3.7 Question 66 Given the following conditions:

You are preparing to take the shift as the OATC The last shift you worked was 5 days ago Per 40DP-9OP33, Shift Turnover:

(1) PRIOR to turnover, you must review the Unit Logs going back a MINIMUM of (2) AFTER turnover, how much more of the Unit Logs, if any, must be reviewed?

A. (1) 3 days (2) No additional Unit Logs review is required B. (1) 3 days (2) 2 additional days of Unit Logs review is required C. (1) 5 days (2) No additional Unit Logs review is required D. (1) 5 days (2) 2 additional days of Unit Logs review is required

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible since no additional log review would be required if the last shift was within the last 3 days, however since the last shift was 5 days ago, an additional 2 days of log review is required.

B. Correct.

C. First part is plausible since 5 days of logs are required to be reviewed, however the minimum requirement for log review prior to turnover is 3 days or until the last shift, whichever is SHORTER. Second part is plausible as it would be correct if 5 days of log review was required prior to turnover, however since only 3 days of log review were required in part 1, the answer is incorrect.

D. First part is plausible since 5 days of logs are required to be reviewed, however the minimum requirement for log review prior to turnover is 3 days or until the last shift, whichever is SHORTER. Second part is plausible since the requirement for log review after turnover is since the last shift worked or 7 days, but the requirement is the shorter of those two options, which in this case would be 5 days.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 25772 - Given the conditions associated with Control Room relief, describe the required review of operating logs prior to this relief in accordance with 40DP-9OP33

Technical

Reference:

40DP-9OP33, Shift Turnover Technical

Reference:

40DP-9OP33, Shift Turnover

Examination Outline Cross-

Reference:

Level RO SRO K/A: Conduct of Operations: Knowledge of the Tier 3 administrative requirements for temporary management Group directives, such as standing orders, night orders, Operations memos, etc K/A G 2.1.15 IR 2.7 Question 67 Given the following conditions:

An Operational Decision Making Issue (ODMI) has been issued for a Pressurizer safety valve that is leaking by The crew is calculating RCS leakage from the Pressurizer Safety valve every hour to determine if 1 GPM is exceeded and additional action needs to be taken The ODM Action Plan is approved by the ___(1)___ and if 1 GPM is exceeded the crew should refer to the ___(2)___ point section of the ODMI.

A. (1) Plant Manager (2) hold B. (1) Plant Manager (2) trigger C. (1) Operations Director (2) hold D. (1) Operations Director (2) trigger

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because a hold point is a term used for radiation exposure to maintain individual and collective doses ALARA and prevent exceeding dose limits.

B. Correct C. First part is plausible because the Operations Director oversees Operations for all 3 units, however the Plant Manager approves the Action Plan. Second part is plausible because a hold point is a term used for radiation exposure to maintain individual and collective doses ALARA and prevent exceeding dose limits.

D. First part is plausible because the Operations Director oversees Operations for all 3 units, however the Plant Manager approves the Action Plan. Send part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 203144 - Describe the Operational Decision Making process

Technical

Reference:

01DP-0ZZ01, Operational Decision Making Technical

Reference:

01DP-0ZZ01, Operational Decision Making Examination Outline Cross-

Reference:

Level RO SRO K/A: Equipment Control: Knowledge of Surveillance Tier 3 procedures Group K/A G 2.2.12 IR 3.7 Question 68 Per 40ST-9ZZM1, Operations Mode 1 Surveillance Logs:

(1) Appendix B, Mode 1 SHIFTLY Surveillance Logs Data Sheets, must have the Acceptance Review completed NO LATER THAN (2) Appendix C, Mode 1 DAILY Surveillance Logs Data Sheets, is directed to be performed during A. (1) 0800 on day shift and 2000 on night shift (2) day shift B. (1) 0800 on day shift and 2000 on night shift (2) night shift C. (1) 1100 on day shift and 2300 on night shift (2) day shift D. (1) 1100 on day shift and 2300 on night shift (2) night shift

Proposed Answer: D Explanations:

A. First part is plausible since 0800 is the earliest the shiftly logs can be completed and reviewed, however the latest is 1100. Second part is plausible since some of the Mode 1 daily surveillances are done on the day shift (i.e. ISFSI daily checks), however the Daily Surveillance Logs Data Sheets are done on the night shift.

B. First part is plausible since 0800 is the earliest the shiftly logs can be completed and reviewed, however the latest is 1100. Second part is correct.

C. First part is correct. Second part is plausible since some of the Mode 1 daily surveillances are done on the day shift (i.e. ISFSI daily checks), however the Daily Surveillance Logs Data Sheets are done on the night shift.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 10 Reference N Provided:

Learning Objective: 27053 - Describe the responsibilities of the Reactor Operator with respect to logkeeping

Technical

Reference:

40ST-9ZZM1, Operations Mode 1 Surveillance Logs Technical

Reference:

40ST-9ZZM1, Operations Mode 1 Surveillance Logs Technical

Reference:

40ST-9ZZM1, Operations Mode 1 Surveillance Logs Examination Outline Cross-

Reference:

Level RO SRO K/A: Equipment Control: Knowledge of tagging and Tier 3 clearance procedures Group K/A G 2.2.13 IR 4.1 Question 69 Per 40DP-9OP29, Power Block Clearance and Tagging, double valve isolation is REQUIRED for systems that are greater than a MINIMUM of ___(1)___ °F OR greater than a MINIMUM of ___(2)___ psig.

A. (1) 200 (2) 385 B. (1) 200 (2) 500 C. (1) 212 (2) 385 D. (1) 212 (2) 500

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because greater than 385 psia double valve isolation of the RCS from the SDC system is required.

B. Correct C. First part is plausible because 212°F is the temperature that water boils. Second part is plausible because greater than 385 psia double valve isolation of the RCS from the SDC system is required.

D. First part is plausible because 212°F is the temperature that water boils. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 10 Reference N Provided:

Learning Objective: 27256 - Describe the special Clearance precautions used when establishing isolation boundaries on fluid or gas systems that operate at high temperatures or pressures

Technical

Reference:

40DP-9OP29, Power Block Clearance and Tagging Technical

Reference:

40OP-9SI01, Shutdown Cooling Initiation Examination Outline Cross-

Reference:

Level RO SRO K/A: Equipment Control: Knowledge of the process for Tier 3 managing maintenance activities during power Group operations, such as risk assessments, work prioritization, and coordination with the transmission K/A G 2.2.17 system operator IR 2.6 Question 70 Given the following conditions:

All Units are operating at 100% power There is required maintenance on Westwing #1 transmission line An Auxiliary Operator needs to access the Switchyard to hang a clearance Per 40DP-9OP34, Switchyard Administrative Controls:

(1) Auxiliary Operator access to the Switchyard may be granted by (2) In order to hang the clearance, a Switching Order is REQUIRED to be provided by A. (1) ANY of the Unit SMs (2) the Energy Control Center (ECC) ONLY B. (1) ANY of the Unit SMs (2) the Energy Control Center (ECC) AND Salt River Project (SRP)

C. (1) the Unit 1 SM ONLY (2) the Energy Control Center (ECC) ONLY D. (1) the Unit 1 SM ONLY (2) the Energy Control Center (ECC) AND Salt River Project (SRP)

Proposed Answer: C Explanations:

A. First part is plausible because each SM will oversee switching of their respective generator output breakers, including associated 525kV MOD and Start-up Transformer Secondary 13.8kV Disconnects. Second part is correct.

B. First part is plausible because each SM will oversee switching of their respective generator output breakers, including associated 525kV MOD and Start-up Transformer Secondary 13.8kV Disconnects. Second part is plausible because SRP is involved in the coordination of switching orders, however, the ECC provides the procedures.

C. Correct D. First part is correct. Second part is plausible because SRP is involved in the coordination of switching orders, however, the ECC provides the procedures Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 26287 - Identify the responsibilities of the Unit 1 Shift Manager concerning switchyard operations

Technical

Reference:

40DP-9OP34, Switchyard Administrative Controls Technical

Reference:

40DP-9OP34, Switchyard Administrative Controls Examination Outline Cross-

Reference:

Level RO SRO K/A: Radiation Control: Knowledge of radiological Tier 3 safety procedures pertaining to licensed operator duties, Group such as response to radiation monitor alarms, containment entry requirements, fuel handling K/A G 2.3.13 responsibilities, access to locked high-radiation areas, IR 3.4 aligning filters, etc Question 71 Per 40DP-9ZZ01, Containment Entry in Mode 1 Through Mode 4, prior to a Containment Entry in MODE 1, Containment must be purged if Containment Atmosphere (1) H2 Concentration is GREATER than or equal to a MINIMUM of OR (2) Containment Atmospheric O2 Concentration is LESS than a MAXIMUM of A. (1) 0.04%

(2) 19.5%

B. (1) 0.04%

(2) 20.3%

C. (1) 0.15%

(2) 19.5%

D. (1) 0.15%

(2) 20.3%

Proposed Answer: C Explanations:

A. First part is plausible since if H2 concentration is > 0.15%, purge is required until H2 concentration is < 0.04%, however purge is not required if initial H2 concentration is < 0.15%. Second part is correct.

B. First part is plausible since if H2 concentration is > 0.15%, purge is required until H2 concentration is < 0.04%, however purge is not required if initial H2 concentration is < 0.15%. Second part is plausible since if two consecutive O2 samples are less than 19.5%, purge is required until O2 concentration is > 20.3%, however purge is not required if initial O2 concentration is > 19.5%.

C. Correct.

D. First part is correct. Second part is plausible since if two consecutive O2 samples are less than 19.5%, purge is required until O2 concentration is > 20.3%, however purge is not required if initial O2 concentration is > 19.5%.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 9 Reference N Provided:

Learning Objective: 27001 - Given that a confined space entry must be made, determine the necessary requirements per 01DP-01SI2

Technical

Reference:

40DP-9ZZ01, Containment Entry in Mode 1 Through Mode 4 Examination Outline Cross-

Reference:

Level RO SRO K/A: Radiation Control: Knowledge of radiological Tier 3 safety principles pertaining to licensed operator duties, Group such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, K/A G 2.3.12 aligning filters, etc IR 3.2 Question 72 When working at heights in the RCA, fall protection is required for any work being performed above a MINIMUM height of ___(1)___ feet, and RP must be contacted to evaluate the need to perform a survey for any work being performed above a MINIMUM height of ___(2)___ feet.

A. (1) 4 (2) 6 B. (1) 4 (2) 7 C. (1) 6 (2) 6 D. (1) 6 (2) 7

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because 6 feet was the old requirement and was recently changed to 7 feet.

B. Correct C. First part is plausible since 6 feet was the minimum height which required fall protection until 2015, however the current minimum height requiring fall protection is 4 feet. Second part is plausible because 6 feet was the old requirement and was recently changed to 7 feet.

D. First part is plausible since 6 feet was the minimum height which required fall protection until 2015, however the current minimum height requiring fall protection is 4 feet. Second part is correct.

Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 12 Reference N Provided:

Learning Objective: 228365 - Describe operations expectations when it comes to Safety in accordance with ODP-1, Operations Principles and Standards

Technical

Reference:

75DP-0RP01, Radiological Posting and Labeling Technical

Reference:

01DP-01S20, Safety at Heights - Fall Protection Examination Outline Cross-

Reference:

Level RO SRO K/A: Radiation Control: Knowledge of radiation Tier 3 exposure limits under normal or emergency conditions Group K/A G 2.3.4 IR 3.4 Question 73 Per 10CFR20.1201, Occupational Dose Limits, the annual limit for dose to the lens of the eye is ___(1)___ rem and to extremities is ___(2)___ rem.

A. (1) 12 (2) 40 B. (1) 12 (2) 50 C. (1) 15 (2) 40 D. (1) 15 (2) 50

Proposed Answer: D Explanations:

A. First part is plausible because 12 rem is an administrative dose limit. Second part is plausible because 40 rem is an administrative dose limit.

B. First part is plausible because 12 rem is an administrative dose limit. Second part is correct.

C. First part is correct. Second part is plausible because 40 rem is an administrative dose limit.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.41: 12 Reference N Provided:

Learning Objective: State the Federal Dose Limits for lens of the eye and extremities

Technical

Reference:

75DP-9RP01, Radiation Exposure and Access Control Technical

Reference:

75DP-9RP01, Radiation Exposure and Access Control Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Procedures/Plan: Knowledge of the Tier 3 emergency plan Group K/A G 2.4.29 IR 3.1 Question 74 Given the following conditions:

Unit 1 SM has just declared an ALERT for an event in progress The STSC communicator duties are normally performed by a(n) ___(1)___ and offsite notifications are required to be made within a MAXIMUM of ___(2)___ minutes of the declaration.

A. (1) Reactor Operator (2) 15 B. (1) Reactor Operator (2) 30 C. (1) Auxiliary Operator (2) 15 D. (1) Auxiliary Operator (2) 30

Proposed Answer: C Explanations:

A. First part is plausible because the ENS Communicator is normally a Reactor Operator. Second part is coorect.

B. First part is plausible because the ENS Communicator is normally a Reactor Operator. Second part is plausible because site Accountability is required within 30 minutes of a Site Area Emergency.

C. Correct D. First part is correct. Second part is plausible because site Accountability is required within 30 minutes of a Site Area Emergency.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 28129 - Identify actions to be taken as STSC Communicator

Technical

Reference:

40DP-9OP02, Conduct of Operations Technical

Reference:

Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Procedures/Plan: Knowledge of facility Tier 3 protection requirements, including fire brigade and Group portable firefighting equipment usage K/A G 2.4.26 IR 3.1 Question 75 Per 40DP-9OP02, Conduct of Operations, there will be one Fire Team Advisor (FTA) assigned to ___(1)___ and the LOWEST QUALIFICATION level he/she is required to be qualified is ___(2)___.

A. (1) each Unit (2) Reactor Operator B. (1) each Unit (2) Auxiliary Operator C. (1) the entire Site (2) Reactor Operator D. (1) the entire Site (2) Auxiliary Operator

Proposed Answer: C Explanations:

A. First part is plausible because each unit will have their own STSC and ENS communicator during an emergency. Second part is correct.

B. First part is plausible because each unit will have their own STSC and ENS communicator during an emergency. Second part is plausible because an Auxiliary Operator fills the role of STSC Communicator and it is reasonable to think that an Auxiliary Operator will be able to respond to fire faster.

C. Correct D. First part is correct. Second part is plausible because an Auxiliary Operator fills the role of STSC Communicator and it is reasonable to think that an Auxiliary Operator will be able to respond to fire faster.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.41: 10 Reference N Provided:

Learning Objective: 445256 - Describe the Duties and Responsibilities of the Fire Team Advisor

Technical

Reference:

40DP-9OP02, Conduct of Operations Technical

Reference:

40DP-9OP02, Conduct of Operations qExamination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Vapor Space Accident: Knowledge of Tier 1 the operational implications of EOP warnings, cautions, Group 1 and notes K/A 008 G 2.4.20 IR 4.3 Question 76 (1) Per 40EP-9EO03, LOCA, one indication of voiding in the RCS occurs AS SOON AS RVLMS indicates a vessel level of less than (2) Per the EAL Hot Chart, a POTENTIAL LOSS of the Fuel Cladding Barrier occurs AS SOON AS RVLMS indicates a vessel level of less than A. (1) 16% in the RVUH (2) 16% in the RVUH B. (1) 16% in the RVUH (2) 21% in the plenum C. (1) 100% in the RVUH (2) 16% in the RVUH D. (1) 100% in the RVUH (2) 21% in the plenum

Proposed Answer: D Explanations:

A. First part is plausible because greater than 16% in the RVUH means that the Inventory Control Safety Function is met and with the plenum full it could be assumed that there is no voiding.

Second part is plausible because if there is not 16% in the RVUH then the Inventory Control Safety Function would net be met and there could potentially be voiding.

B. First part is plausible because greater than 16% in the RVUH means that the Inventory Control Safety Function is met and with the plenum full it could be assumed that there is no voiding.

Second part is correct.

C. First part is correct. Second part is plausible because if there is not 16% in the RVUH then the Inventory Control Safety Function would net be met and there could potentially be voiding.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 4 Reference N Provided:

Learning Objective: 297491 - Demonstrate RCS Void Control per Standard Appendices Appendix 15, RCS Void Control

Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Technical

Reference:

EAL Hot Chart Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Large Break LOCA: Knowledge of the bases in Tier 1 Technical Specifications for limiting conditions for Group 1 operations and safety limits K/A 011 G 2.2.25 IR 4.2 Question 77 Per Technical Specifications, in order for Safety Injection Tanks to be OPERABLE, they must have a MINIMUM boron concentration of ___(1)___ ppm in order to ensure

___(2)___ in the event a LOCA.

A. (1) 2300 (2) the Reactor will remain subcritical following the injection of relatively colder SIT water volume into the RCS B. (1) 2300 (2) back leakage from the RCS into the SITs during normal operations will not dilute the SITs to less than the minimum required boron concentration in the safety analysis C. (1) 4000 (2) the Reactor will remain subcritical following the injection of relatively colder SIT water volume into the RCS D. (1) 4000 (2) back leakage from the RCS into the SITs during normal operations will not dilute the SITs to less than the minimum required boron concentration in the safety analysis

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible as this is the basis for the minimum boron concentration in the RWT.

B. Correct.

C. First part is plausible since this is the minimum required boron concentration for the RWT.

Second part is plausible as this is the basis for the minimum boron concentration in the RWT.

D. First part is plausible since this is the minimum required boron concentration for the RWT.

Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 21211 - Identify the basis of Technical Specifications LCOs and TLCOs for Section 3.5

Technical

Reference:

Tech Specs LCO 3.5.1, SITs Technical

Reference:

Tech Specs LCO 3.5.5, RWT

Technical

Reference:

Tech Spec Bases for LCO 3.5.1, SITs Technical

Reference:

Tech Spec Bases for LCO 3.5.5, RWT

Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Component Cooling Water: The normal Tier 1 values and upper limits for the temperatures of the Group 1 components cooled by CCW K/A 026 AA2.04 IR 2.9 Question 78 Given the following conditions:

Unit 2 is operating at 100% power Subsequently:

A loss of all Nuclear Cooling Water occurs The CRS enters 40AO-9ZZ03, Loss of Cooling Water The crew should cross-tie Train A EW to NC to prevent RCP HP Seal Cooler inlet temperature to prevent exceeding the procedural driven RCP trip setpoint of

___(1)___ °F. After the cross-tie is complete, Train A EW is considered ___(2)___

per Technical Specifications.

A. (1) 250 (2) OPERABLE B. (1) 250 (2) INOPERABLE C. (1) 300 (2) OPERABLE D. (1) 300 (2) INOPERABLE

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because there are no components or valves that are out of service, however, since a manual valve has been throttled in the closed direction for the SDCHX and the cross connect valves are out of position, the EW System is INOPERABLE.

B. Correct C. First part is plausible because to maintain LPSI seal life, SDC may not be placed in service until RCS temperature is less than 300°F. Second part is plausible because there are no components or valves that are out of service, however, since a manual valve has been throttled in the closed direction for the SDCHX and the cross connect valves are out of position, the EW System is INOPERABLE.

D. First part is plausible because to maintain LPSI seal life, SDC may not be placed in service until RCS temperature is less than 300°F. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 22357 - Given the status of NC and RCP seal injection, describe the limitations on RCP operation without NC in accordance with 40AO-9ZZ03

Technical

Reference:

40AO-9ZZ04, Reactor Coolant Pump Emergencies Technical

Reference:

40OP-9SI01, Shutdown Cooling Initiation Technical

Reference:

40AO-9ZZ03, Loss of Cooling Water, Appendix A, Cross-connect EW to NC

Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Main Feedwater: Knowledge of limiting Tier 1 conditions for operations and safety limits Group 1 K/A 054 G 2.2.22 IR 4.7 Question 79 Per Technical Specification Basis for LCO 3.3.1, RPS Instrumentation - Operating, which of the following RPS trips mitigates a Feedwater Line Break?

1. Departure from Nucleate Boiling Low

2. Containment Pressure High

3. Pressurizer Pressure High A. 2 ONLY B. 3 ONLY C. 1 AND 2 ONLY D. 1 AND 3 ONLY

Proposed Answer: B Explanations:

A. Plausible because a Feedwater line break is high energy release into containment. However only a LOCA and ESD are mitigated by a High Containment pressure trip per Technical Specification Bases B. Correct C. First part is plausible because a Feedwater line break is a heatup event (loss of Feedwater will cause a diminished heat sink and RCS temperature will rise), therefore DNBR will lower. Second part is plausible because a Feedwater line break is high energy release into containment.

However only a LOCA and ESD are mitigated by a High Containment pressure trip per Technical Specification Bases.

D. First part is plausible because a Feedwater line break is a heatup event (loss of Feedwater will cause a diminished heat sink and RCS temperature will rise), therefore DNBR will lower. Second part is correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 22620 - Identify the basis of Technical Specification LCOs and TLCOs for section 3.3 in accordance with Tech Spec 3.3 basis

Technical

Reference:

Technical Specifications Basis Technical

Reference:

Technical Specifications Basis Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Station Blackout: Ability to determine or interpret Tier 1 the following as they apply to a Station Blackout: Faults Group 1 and lockouts that must be cleared prior to re-energizing buses K/A 055 EA2.06 IR 4.1 Question 80 Given the following conditions:

Unit 1 is in a blackout condition The B EDG is OOS and will not be available for the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Per Appendix 55, Restore DG A to PBA-S03, which of the following faults can the crew attempt to reset in order to restore power to PBA-S03?

1. Overspeed trip of the A EDG

2. Generator Differential trip of the A EDG

3. Overcurrent trip of the A EDG Output Breaker A. 2 ONLY B. 3 ONLY C. 1 AND 2 ONLY D. 1 AND 3 ONLY

Proposed Answer: C Explanations:

A. Generator Differential is correct, however Overspeed trip is also correct.

B. Plausible because overcurrent of the EDG output breaker can be reset to re-energize the bus, however it cannot be reset using Appendix 55. There is a separate Relay Resetting procedure that would be used.

C. Correct D. Overspeed trip is correct. Overcurrent of the EDG output can be reset to re-energize the bus, however it cannot be reset using Appendix 55. There is a separate Relay Resetting procedure that would be used.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 26234 - Given conditions of a Blackout and plant status, determine the flowpath of power available to energize a vital bus in accordance with 40EP-9EO08

Technical

Reference:

40EP-9EO10-055, Appendix 55: Restore DG A to PBA-S03 Technical

Reference:

40EP-9EO10-055, Appendix 55: Restore DG A to PBA-S03 Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Offsite Power: Ability to determine and Tier 1 interpret the following as they apply to the Loss of Group 1 Offsite Power: EDG indicators for the following:

voltage, frequency, load, load-status, and closure of the K/A 056 AA2.37 bus-tie breakers IR 3.8 Question 81 Given the following conditions:

Unit 1 was manually tripped in preparation for a refueling outage SPTAs are complete and the CRS has transitioned to 40EP-9EO02, Reactor Trip Subsequently:

A LOOP occurred On the LOOP the following occurred:

o A EDG tripped on overspeed o NNN-D12 tripped on a fault The BOP reports that there is no frequency indication on B EDG B EDG frequency should be determined ___(1)___ and entry into 40EP-9EO09, Functional Recovery is ___(2)___.

A. (1) locally (2) REQUIRED B. (1) locally (2) NOT required C. (1) in the Control Room (2) REQUIRED D. (1) in the Control Room (2) NOT required

Proposed Answer: D Explanations:

A. First part is plausible because speed is indicated locally and there is no direction in SPTAs to energize the synchroscope if frequency indication is lost. Second part is correct.

B. First part is plausible because speed is indicated locally and there is no direction in SPTAs to energize the synchroscope if frequency indication is lost. Second part is plausible if it is thought that because there are multiple events with the LOOP and loss of NNN-D11, the Functional Recovery procedure is required to recover. However the MVAC Safety is met in the LOOP ORP.

C. First part is correct. Second part is plausible if it is thought that because there are multiple events with the LOOP and loss of NNN-D11, the Functional Recovery procedure is required to recover.

However the MVAC Safety is met in the LOOP ORP.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 23195 - Analyze MVA to determine if the SFSC acceptance citeria is satisfied

Technical

Reference:

40DP-9AP06, Standard Post Trip Actions Technical Guidelines Technical

Reference:

40EP-9EO07, Loss of Offsite Power / Loss of Forced Circulation Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Dropped Control Rod: Ability to recognize Tier 1 abnormal indications for system operating parameters Group 2 that are entry-level conditions for emergency and abnormal operating procedures K/A 003 G 2.4.4 IR 4.7 Question 82 Given the following initial conditions:

Unit 2 is operating at 100% power B Boric Acid Makeup pump is OOS Subsequently:

A Seismic event occurs A 4-finger CEA dropped to the bottom of the core Reactor power stabilized at 97% following the dropped CEA The CRS entered 40AO-9ZZ11, CEA Malfunctions An Auxiliary Operator reports that the A Boric Acid Makeup pump is severely damaged and CANNOT be used The CRS should direct the crew to lower Reactor power to a MAXIMUM of ___(1)___

within the first hour from the CEA drop and direct the crew to borate the RCS using

___(2)___ .

A. (1) 77%

(2) 40AO-9ZZ11, CEA Malfunctions, Appendix J, Boration for Power Reduction B. (1) 77%

(2) 40AO-9ZZ01, Emergency Boration C. (1) 80%

(2) 40AO-9ZZ11, CEA Malfunctions, Appendix J, Boration for Power Reduction D. (1) 80%

(2) 40AO-9ZZ01, Emergency Boration

Proposed Answer: D Explanations:

A. First part is plausible because it can be assumed that the downpower of 20% is from the power level after the CEA drops. Second part is plausible because if a BAMP is available then 40AO-9ZZ11, CEA Malfunctions, Appendix J can be used.

B. First part is plausible because it can be assumed that the downpower of 20% is from the power level after the CEA drops. Second part is correct.

C. First part is correct. Second part is plausible because if a BAMP is available then 40AO-9ZZ11, CEA Malfunctions, Appendix J can be used.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 25223 - Describe what action is taken to commence a Tech Spec required power reduction due to a dropped/slipped CEA

Technical

Reference:

40AO-9ZZ11, CEA Malfunctions Technical

Reference:

40AO-9ZZ11, CEA Malfunctions, Appendix J, Boration for Power Reduction

Technical

Reference:

40AO-9ZZ11, CEA Malfunctions Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Level Control Malfunction: Ability to Tier 1 determine and interpret the following as they apply to Group 2 the Pressurizer Level Control Malfunction: Letdown flow indicator K/A 028 AA2.06 IR 2.8 Question 83 Given the following conditions:

Unit 3 is operating at 100% power A PLCS malfunction has caused a sudden increase in letdown flow The increase in letdown flow caused PSV-354, Low Pressure Letdown Relief Valve, to come off its closed seat ERFDADS indicates the leak rate through PSV-354 is 5 gpm and stable PSV-354 should relieve to the ___(1)___ and per LCO 3.4.14 RCS Operational Leakage, this ___(2)___ considered RCS Leakage.

A. (1) EDT (2) IS B. (1) EDT (2) is NOT C. (1) RDT (2) IS D. (1) RDT (2) is NOT

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because the leakage does register as RCS leakage on ERFDADS and inventory is actually being lost from the RCS however intersystem leakage.

B. Correct C. First part is plausible because the RDT is also connected to the CVCS system and collects various leakages. Second part is plausible because the leakage does register as RCS leakage on ERFDADS and inventory is actually being lost from the RCS.

D. First part is plausible because the RDT is also connected to the CVCS system and collects various leakages. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 22681 - Given conditions when an LCO is not met, apply Tech Spec Section 3.4.14 (RCS Operational Leakage) in accordance with Tech Spec 3.4.14

Technical

Reference:

Chemical Volume Control System Tech Manual Technical

Reference:

LOIT Excessive RCS Leakrate Lesson Plan Technical

Reference:

LOIT Excessive RCS Leakrate Lesson Plan Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Loss of Source Range Nuclear Instrumentation: Tier 1 Ability to recognize system parameters that are entry- Group 2 level conditions for Technical Specifications K/A 032 G 2.2.42 IR 4.6 Question 84 Given the following conditions:

Unit 1 is in MODE 6 Core reload is in progress Subsequently:

Audible indication of count rate for the Startup Range Monitors (SRMs) is lost inside Containment Audible and visual SRM indications remain available in the Control Room Based on these indications, the core reload A. MAY continue provided audible AND visual source range indications are available in the control room B. MUST be suspended in accordance with LCO 3.3.12 Boron Dilution Alarm System, Condition A, for two required SRMs inoperable C. MAY continue provided audible source range indication is available in the control room AND the Refueling machine maintains constant communications with the control room D. MUST be suspended and action must be taken to restore audible indication in Containment in accordance with LCO 3.9.2 Nuclear Instrumentation, Conditions A and B for two required SRMs inoperable

Proposed Answer: D Explanations:

A. Plausible since visual and audible indications will be maintained in the control room, and there is nothing to indicate that the SRM is not functioning (i.e. only the speaker in Containment is faulted), however in order for the SRM to be operable, audible indications are required in both the control room and containment.

B. Plausible that LCO 3.3.12 would not be met in this situation as inoperability of an SRM normally makes BDAS inoperable, however if SRMs are inoperable SOLELY due to the loss of audible indication, BDAS remains operable.

C. Plausible since audible indications will be maintained in the control room, and maintaining constant communication with the refueling machine could be interpreted as meeting the requirement for audible indication in containment, however communication from the control room to the refueling machine is not credited for meeting the operability requirement of LCO 3.9.2.

D. Correct Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 25278 - Given a set of plant conditions identify whether or not LCO 3.9.2 is satisfied and any actions or surveillance requirements that would prevent core alterations per Tech Spec 3.9 and its basis

Technical

Reference:

Technical Specifications Technical

Reference:

Technical Specifications Basis Technical

Reference:

SRO Level flowchart Examination Outline Cross-

Reference:

Level RO SRO K/A: Steam Generator Tube Leak: Magnitude of Tier 1 atmospheric radioactive release if cooldown must be Group 2 completed using steam dump or atmospheric reliefs K/A 037 AA2.15 IR 4.2 Question 85 Given the following conditions:

Unit 1 was tripped due to a Design Basis Steam Generator Tube Rupture event on SG #1 On the trip, offsite power was lost The crew is commencing a cooldown using ADVs to meet conditions required to isolate SG #1 The use of ADVs for the INITIAL cooldown ___(1)___ considered a loss of the Containment Barrier, and the release in progress ___(2)___ exceeding federally approved limits.

A. (1) IS (2) IS B. (1) IS (2) is NOT C. (1) is NOT (2) IS D. (1) is NOT (2) is NOT

Proposed Answer: D Explanations:

A. First part is plausible because for the initial RCS cooldown, there will be a release to the environment. However, since it is not an unisolable fault (e.g stuck open MSSV), this is not a loss of the Containment Barrier. Second part is plausible because for the initial RCS cooldown, there will be a release to the environment. However, per 0903, Accident Assessment, is not a release that exceeds Federal limits.

B. First part is plausible because for the initial RCS cooldown, there will be a release to the environment. However, since it is not an unisolable fault (e.g stuck open MSSV), this is not a loss of the Containment Barrier. Second part is correct.

C. First part is correct. Second part is plausible because for the initial RCS cooldown, there will be a release to the environment. However, per 0903, Accident Assessment, is not a release that exceeds Federal limits.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 4 Reference N Provided:

Learning Objective: 248390 - Determine whether a radioactive release is in progress

Technical

Reference:

EP-0901, Classifications Basis Technical

Reference:

EP-0903, Accident Assessment, Appendix A - Release Evaluation Flowchart

Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Core Cooling: Knowledge of EOP Tier 2 mitigation strategies Group 1 K/A 006 G 2.4.6 IR 4.7 Question 86 Given the following conditions:

Unit 1 was tripped from 100% power due to a Pressurizer Safety lifting and sticking open SPTAs have been performed and the CRS has entered 40EP-9EO03, LOCA The RCS is 15°F subcooled and stable RCS TCOLD is 565°F and slowly lowering Indicated Pressurizer level is 95% and slowly rising Both SGs are 20% NR and slowly rising, being fed from AFB-P01 QSPDS indicates 41% in the upper head Containment Temperature is 140°F and slowly rising Containment High Range Radiation Monitors RU-148 and RU-149 indicate 6.5 x 102 mR/hr and slowly rising The RCS Heat Removal Safety Function is ___(1)___ and the CRS should implement

___(2)___ to lower pressurizer level.

A. (1) MET (2) Appendix 15, RCS Void Control B. (1) MET (2) Appendix 2, Figures: HPSI Throttle Criteria C. (1) NOT met (2) Appendix 15, RCS Void Control D. (1) NOT met (2) Appendix 2, Figures: HPSI Throttle Criteria

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because all the parameters meet HPSI Throttle Criteria with the exception of RCS subcooling.

C. First part is plausible Steam Generator water levels are not in band. However to meet the Safety Function Feedwater only needs to be restoring SGWL back in band. Second part is correct.

D. First part is plausible Steam Generator water levels are not in band. However to meet the Safety Function Feedwater only needs to be restoring SGWL back in band. Second part is plausible because all the parameters meet HPSI Throttle Criteria with the exception of RCS subcooling.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference Provided:

Learning Objective: 24922 - Given conditions of LOCA, analyze RCS Heat Removal to determine if the SFSC acceptance criteria is satisfied

Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Relief / Quench Tank: Ability to (a) Tier 2 predict the impacts of the following malfunctions or Group 1 operations on the PRTS; and (b) based on those predictions, use procedures to correct, control, or K/A 007 A2.04 mitigate the consequences of those malfunctions or IR 2.9 operations: Overpressurization of the waste gas vent header Question 87 Given the following conditions:

Gaseous Radwaste Radiation Monitor RU-12 has failed off scale high and is alarming on RMS The Gaseous Radwaste header pressure is slowly rising A Waste Gas Decay Tank release is required Which of the following describes the required action(s) in order to perform the release as planned?

In order for the release to be performed, ___(1)___ as required by ___(2)___.

A. (1) the valve galleries associated with the release path must be posted as a high radiation area (2) the Offsite Dose Calculation Manual B. (1) the valve galleries associated with the release path must be posted as a high radiation area (2) 74RM-9EF41, Radiation Monitoring System Alarm Response C. (1) at least two technically qualified personnel must independently verify the discharge valve lineup (2) the Offsite Dose Calculation Manual D. (1) at least two technically qualified personnel must independently verify the discharge valve lineup (2) 74RM-9EF41, Radiation Monitoring System Alarm Response

Proposed Answer: C Explanations:

A. First part is plausible since there is guidance in the alarm response procedure to evaluate changing the radiation postings in the event of a radiation monitor alarm, and since RU-12 has failed high, it would be reasonable to raise the postings as a conservative approach to ALARA, however this is not required in order for the release to commence. Second part is correct.

B. First part is plausible since there is guidance in the alarm response procedure to evaluate changing the radiation postings in the event of a radiation monitor alarm, and since RU-12 has failed high, it would be reasonable to raise the postings as a conservative approach to ALARA, however this is not required in order for the release to commence. Second part is plausible since the ARP provides contingency actions for alarming or failed RMs, however there are no requirements in the ARP related to gaseous releases.

C. Correct D. First part is correct. Second part is plausible since the ARP provides contingency actions for alarming or failed RMs, however there are no requirements in the ARP related to gaseous releases.

Question Source: New X Bank Modified X Previous NRC Exam 2018 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 386147 - As an SRO describe what actions to take during an effluent release if RU-12, Waste Gas Decay Tank Monitor, goes inoperable

Technical

Reference:

Technical

Reference:

74RM-9EF41, Radiation Monitoring System Alarm Response Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Engineered Safety Features Actuation: Knowledge Tier 2 of annunciator alarms, indications, or response Group 1 procedures K/A 013 G 2.4.31 IR 4.1 Question 88 Given the following conditions:

Unit 3 was tripped due to an unisolable small break LOCA into Containment Given the SEIS panel drawing on the following page, the CRS should enter ___(1)___

and direct the crew to cooldown and depressurize the RCS to establish long term cooling via ___(2)___.

A. (1) 40EP-9EO09, Functional Recovery (2) LPSI injection B. (1) 40EP-9EO09, Functional Recovery (2) Shutdown Cooling C. (1) 40EP-9EO03, Loss of Coolant Accident (2) LPSI injection D. (1) 40EP-9EO03, Loss of Coolant Accident (2) Shutdown Cooling

Train A Train B Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because if the RCS leak was isolable, the crew would cooldown/depressurize to SDC entry conditions. Also, there is currently no injection flow due to the loss of both HPSI pumps. The crew will rapidly cooldown/depressurize to achieve LPSI injection.

C. First part is plausible because there is a LOCA and the SPTA diagnostic flow chart says to consider LOCA. However, because the RCS Inventory safety function will not be met, the CRS will enter Functional Recovery. Second part is correct.

D. First part is plausible because there is a LOCA and the SPTA diagnostic flow chart says to consider LOCA. However, because the RCS Inventory safety function will not be met, the CRS will enter Functional Recovery. Second part is plausible because if the RCS leak was isolable, the crew would cooldown/depressurize to SDC entry conditions. Also, there is currently no injection flow due to the loss of both HPSI pumps. The crew will rapidly cooldown/depressurize to achieve LPSI injection.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference Y Provided:

Learning Objective: 25046 - Given conditions of LOCA, analyze the RCS Inventory Control to determine if the SFSC acceptance criteria is satisfied per 40EP-9EO03

Technical

Reference:

40EP-9EO03, Loss of Coolant Accident Technical

Reference:

40EP-9EO09, Functional Recovery Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Containment Cooling: Ability to (a) predict the Tier 2 impacts of the following malfunctions or operations on Group 1 the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the K/A 022 A2.02 consequences of those malfunctions or operations: IR 2.6 Loss of CCS Pump Question 89 Given the following conditions following a large break LOCA inside Containment:

Containment Pressure is 20 psig and slowly rising Both CS Pumps have tripped The CRS has entered 40EP-9EO09, Functional Recovery Per 40EP-9EO09, Functional Recovery, the crew should align a ___(1)___ pump to the CS Spray Header and should verify the CTPC Safety Function is met by

___(2)___.

A. (1) LPSI (2) CS flow indication in the Control Room B. (1) LPSI (2) indicated Containment pressure either lowering or stabilizing C. (1) HPSI (2) CS flow indication in the Control Room D. (1) HPSI (2) indicated Containment pressure either lowering or stabilizing

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because when CS is aligned to the spray header, there is indicated flow. However when LPSI is aligned to the spray header, there will be no indicated flow.

B. Correct C. First part is plausible because HPSI will be used for Hot Leg Injection. However LPSI pumps are used for CS when CS pumps are not available. Second part is plausible because when CS is aligned to the spray header, there is indicated flow. However when LPSI is aligned to the spray header, there will be no indicated flow.

D. First part is plausible because HPSI will be used for Hot Leg Injection. However LPSI pumps are used for CS when CS pumps are not available. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 25485 - Given conditions of LOCA, analyze Containment Temperature and Pressure Control to determine if the SFSC acceptance criteria is satisfied per 40EP-9EO03

Technical

Reference:

40EP-9EO09, Functional Recovery Technical

Reference:

LOIT Functional Recovery Procedure Lesson Plan Technical

Reference:

40EP-9EO10-100, Appendix 100: Hot Leg Injection Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Emergency Diesel Generator: Knowledge of local Tier 2 auxiliary operator tasks during an emergency and the Group 1 resultant operational effects K/A 064 G 2.4.35 IR 4.0 Question 90 Given the following conditions:

Unit 1 is operating at 100% power Subsequently:

A loss of offsite power occurs B EDG trips on low lube oil pressure The crew performs SPTAs The CRS enters 40EP-9EO07, Loss of Offsite Power/Loss of Forced Circulation The ECC has informed the Control Room that estimated time for restoration of offsite power is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> There is no thunderstorm activity in the area Per 40EP-9EO07, Loss of Offsite Power/Loss of Forced Circulation, the crew should direct the ___(1)___ to energize NAN-S07 with an SBOG and the CRS should direct an RO to perform ___(2)___.

A. (1) Outside Area Operator (2) 40EP-9EO10-081, Appendix 81 - Align SBOG to PBB-S04 (BO)

B. (1) Outside Area Operator (2) 40EP-9EO10-054, Appendix 54 - Energize Switchyard Loads From the SBOGs C. (1) Control Building Operator (2) 40EP-9EO10-081, Appendix 81 - Align SBOG to PBB-S04 (BO)

D. (1) Control Building Operator (2) 40EP-9EO10-054, Appendix 54 - Energize Switchyard Loads From the SBOGs

Proposed Answer: B Explanations:

A. First part is correct. Second part is plausible because PBB-S04 is de-energized, however since the MVAC safety function is met, there is no reason to align power to PBB-S04. Additionally, Appendix 81 is not directed in LOOP/LOFC. It only exists in the Blackout EOP.

B. Correct C. First part is plausible because the Control Building Operator will check on a diesel, but it is the A and B EDGs. Second part is plausible because PBB-S04 is de-energized, however since the MVAC safety function is met, there is no reason to align power to PBB-S04. Additionally, Appendix 81 is not directed in LOOP/LOFC. It only exists in the Blackout EOP.

D. First part is plausible because the Control Building Operator will check on a diesel, but it is the A and B EDGs. Second part is correct.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: Given conditions of a LOOP, analyze these conditions to determine if switchyard loads should be energized by a SBOG per 40EP-9EO07

Technical

Reference:

40EP-9EO07, Loss of Offsite Power / Loss of Forced Circulation Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Pressurizer Level Control: Ability to (a) predict the Tier 2 impacts of the following malfunctions or operations on Group 2 the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the K/A 011 A2.07 consequences of those malfunctions or operations: IR 3.3 Isolation of letdown Question 91 Given the following conditions:

Unit 3 is operating at 12%

The crew is preparing to sync the Main Generator to the Grid CHE-P01 is aligned to Train B Charging Pump Mode Selector switch, CHN-HS-4, is in the 2-3-1 position Subsequently:

PBB-S04 trips on overcurrent With NO operator action, Pressurizer level should FIRST exceed the ___(1)___ level LCO 3.4.9 Tech Spec limit and the Unit will be required to be in MODE 3 within a MAXIMUM of ___(2)___ hours from the time that the LCO 3.4.9 limit is exceeded.

A. (1) low (2) 6 B. (1) low (2) 7 C. (1) high (2) 6 D. (1) high (2) 7

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because if pressurizer level lowers to 25%, both class Pressurizer heater banks will be INOPERABLE resulting in LCO 3.0.3. However, per LCO 3.4.9 the unit will be in a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> action once Pressurizer level lowers to 27%.

C. First part is plausible because if the Charging Pump Mode Selector switch is in the 1-2-3 position, CHA-P01 will remain running, letdown will isolate and the LCO 3.4.9 Pressurizer level high Tech Spec value of 56% will be exceeded.

D. First part is plausible because if the Charging Pump Mode Selector switch is in the 1-2-3 position, CHA-P01 will remain running, letdown will isolate and the LCO 3.4.9 Pressurizer level high Tech Spec value of 56% will be exceeded. Second part is plausible because as Pressurizer level rises, Pressurizer pressure will also rise. Once Pressurizer pressure rises to 2285 psia, both class Pressurizer heater banks will be INOPERABLE resulting in LCO 3.0.3.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 22676 - Given conditions when an LCO is not met, apply Tech Spec Section 3.4.9 (Pressurizer) in accordance with tech pec 3.4.9

Technical

Reference:

Operator Information Manual Technical

Reference:

Operator Information Manual Technical

Reference:

Technical Specifications Technical

Reference:

Operator Information Manual Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Fuel Handling Equipment: Ability to predict and/or Tier 2 monitor changes in parameters (to prevent exceeding Group 2 design limits) associated with operating the Fuel Handling System controls including: Water level in the K/A 034 A1.02 refueling canal IR 3.7 Question 92 Given the following conditions:

A crack in the Spent Fuel Pool liner has caused the level to lower Spent Fuel Pool level is 23 ft 8 in above the irradiated fuel assemblies Level is lowering at a rate of 1 inch every 5 minutes (1) With NO operator action, Spent Fuel Pool level should reach the Technical Specification MINIMUM level in (2) The basis for the Technical Specification MINIMUM level is to A. (1) 40 minutes (2) shield and minimize the general area dose when the storage racks are filled to their maximum capacity B. (1) 40 minutes (2) maintain Spent Fuel Pool keff < 0.95 assuming the most limiting single fuel mishandling accident C. (1) 60 minutes (2) shield and minimize the general area dose when the storage racks are filled to their maximum capacity D. (1) 60 minutes (2) maintain Spent Fuel Pool keff < 0.95 assuming the most limiting single fuel mishandling accident

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because it is the basis for the Spent Fuel Pool boron concentration. If the Spent Fuel Pool lowers and the temperature rises to the point that some Boron may come out of solution and lower the Boron Concentration.

C. First part is plausible because Spent Fuel Pool level 22 feet 8 inches above irradiated fuel is the Spent Fuel Pool LO-LO alarm setpoint. Second part is correct.

D. First part is plausible because Spent Fuel Pool level 22 feet 8 inches above irradiated fuel is the Spent Fuel Pool LO-LO alarm setpoint. Second part is plausible because it is the basis for the Spent Fuel Pool boron concentration. If the Spent Fuel Pool lowers and the temperature rises to the point that some Boron may come out of solution and lower the Boron Concentration.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 21218 - Given a set of plant conditions, determine whether or not the LCOs and TLCOs of 3.7 are satisfied in accordance with Tech Spec 3.7

Technical

Reference:

Technical Specifications Technical

Reference:

Technical Specifications Basis Technical

Reference:

Technical Specifications Basis Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Condensate: Ability to analyze the effect of Tier 2 maintenance activities, such as degraded power Group 2 sources, on the status of limiting conditions for operations K/A 056 G 2.2.36 IR 4.2 Question 93 Given the following conditions:

Unit 1 is operating at 100% power The CST is being drained for emergent corrective maintenance CST has just dropped below 22 feet Per LCO 3.7.6, Condensate Storage Tank, the crew must INITIALLY verify operability of the RMWT within a MAXIMUM of ___(1)___ hours and to restore the CST to OPERABLE the CST level will need to be raised to a MINIMUM of ___(2)___ feet.

A. (1) 4 (2) 29.5 B. (1) 4 (2) 31 C. (1) 5 (2) 29.5 D. (1) 5 (2) 31

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because 31 feet is the alarm setpoint for CST AT MINIMUM OPERATING LEVEL.

C. First part is plausible because 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> would be allowed if this was a surveillance per SR 3.0.2.

However, there is no extension for an LCO. Second part is correct.

D. First part is plausible because 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> would be allowed if this was a surveillance per SR 3.0.2.

However, there is no extension for an LCO. Second part is plausible because 31 feet is the alarm setpoint for CST AT MINIMUM OPERATING LEVEL.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 17468 - Given conditions when an LCO is not met, apply Tech Spec 3.7.6 in accordance with Tech Specs

Technical

Reference:

Technical Specifications Technical

Reference:

40AL-9RK6A, Panel B06A Alarm Responses Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Ability to interpret reference materials, such as Tier 3 graphs, curves, tables, etc. Group K/A G 2.1.25 IR 4.2 Question 94 Using the Safety Function Tracking Sheet on the following page:

(1) The first Safety Function performed will be (2) After all Challenged and Jeopardized Safety Functions are performed, the next Success Path in use to be verified will be A. (1) Pressure Control (2) MVDC B. (1) Pressure Control (2) Reactivity Control C. (1) Heat Removal (2) MVDC D. (1) Heat Removal (2) Reactivity Control

Proposed Answer: C Explanations:

A. First part is plausible because the Pressure Control safety function is higher in the hierarchy than RCS Heat Removal. However, since its jeopardized it takes priority over a challenged safety function. Second part is correct.

B. First part is plausible because the Pressure Control safety function is higher in the hierarchy than RCS Heat Removal. However, since its jeopardized it takes priority over a challenged safety function. Second part is plausible because Success Path performance is normally done in safety function hierarchy order. However, since the completed column is greyed out (all CEAs inserted),

the first safety function success path verified will be MVDC.

C. Correct D. First part is correct. Second part is plausible because Success Path performance is normally done in safety function hierarchy order. However, since the completed column is greyed out (all CEAs inserted), the first safety function success path verified will be MVDC.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference Y Safety Function Tracking Sheet Provided:

Learning Objective: 27348 - Given the FRP is being performed and specific plant conditions, determine if a specific selected success path is jeopardized or challenged and how that information will be used in accordance with 40EP-9EO09

Technical

Reference:

40DP-9AP14, Functional Recovery Technical Guideline Technical

Reference:

40DP-9AP14, Functional Recovery Technical Guideline Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Knowledge of new and spent fuel movement Tier 3 procedures Group K/A G 2.1.42 IR 3.4 Question 95 The transportation of a dry cask from the Unit 2 Fuel Building to its designated storage location at the ISFSI is complete (1) Ownership of the dry cask while being delivered to the ISFSI is the responsibility of the (2) Ownership of this dry cask concerning the performance of specific conditional surveillances and inspections is the responsibility of the A. (1) Unit 1 Shift Manager (2) Unit 1 Shift Manager B. (1) Unit 1 Shift Manager (2) Unit 2 Shift Manager C. (1) Unit 2 Shift Manager (2) Unit 1 Shift Manager D. (1) Unit 2 Shift Manager (2) Unit 2 Shift Manager

Proposed Answer: C Explanations:

A. First part is plausible since Unit 1 has ownership of the dry cask conditional surveillances and inspections. Second part is correct.

B. First part is plausible since Unit 1 has ownership of the dry cask conditional surveillances and inspections. Second part is plausible because Unit 2 has ownership of the dry cask while it being delivered to the ISFSI.

C. Correct D. First part is correct. Second part is plausible because Unit 2 has ownership of the dry cask while it being delivered to the ISFSI.

Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.43: 7 Reference N Provided:

Learning Objective: 90026 - Describe Operations responsibilities for Dry Cask Storage Operations

Technical

Reference:

2016 NRC Exam Original Question The transportation of a dry cask from the Unit 2 Fuel Building to its designated storage location at the ISFSI is complete Who has the ownership of this dry cask concerning the performance of specific conditional surveillances and inspections?

A. Unit 1 Shift Manager B. Unit 2 Shift Manager C. Unit 1 Control Room Supervisor D. Unit 2 Control Room Supervisor

Technical

Reference:

40DP-9OP02, Conduct of Operations Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Ability to perform pre-startup procedures for the Tier 3 facility including operating those controls associated Group with plant equipment that could affect reactivity K/A G 2.2.33 IR 4.4 Question 96 Given the following conditions:

Unit 1 is conducting a Reactor startup Per 40DP-9OP02 Conduct of Operations, during a Reactor Startup the Reactivity Manager can be ___(1)___ and the EARLIEST they are required to be stationed is when the ___(2)___.

A. (1) an off-watch SRO OR the Unit 1 licensed STA (2) Shutdown Group Bank CEAs are withdrawn B. (1) an off-watch SRO OR the Unit 1 licensed STA (2) Regulating Group 1 CEAs are withdrawn C. (1) an off-watch SRO ONLY (2) Shutdown Group Bank CEAs are withdrawn D. (1) an off-watch SRO ONLY (2) Regulating Group 1 CEAs are withdrawn

Proposed Answer: C Explanations:

A. First part is plausible because the on watch STA has an SRO license however, he cannot perform duties as the STA and the Reactivity Manager at the same time. Second part is correct.

B. First part is plausible because the on watch STA has an SRO license however, he cannot perform duties as the STA and the Reactivity Manager at the same time. Second part is plausible because when Regulating Group CEAs are withdrawn, it is the first time that a reactivity change is observable. However, the Reactivity Manager will be stationed during Shutdown bank withdrawal.

C. Correct D. First part is correct. Second part is plausible because when Regulating Group CEAs are withdrawn, it is the first time that a reactivity change is observable. However, the Reactivity Manager will be stationed during Shutdown bank withdrawal.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 268823 - Determine the control room operators responsibilities with respect to Reactivity Management

Technical

Reference:

40DP-9OP02, Conduct of Operations Technical

Reference:

40OP-9ZZ23, Outage GOP Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Ability to apply Technical Specifications for a Tier 3 system Group K/A G 2.2.40 IR 4.7 Question 97 Given the following conditions:

Unit 1 is exiting an outage with Tcold 345°F Preparations are being made to enter MODE 3 During the outage, maintenance on AFA-P01 was conducted and the governor was replaced All maintenance activities have been completed including all Surveillance Requirements, with the exception of Surveillances needed to be performed at NOP/NOT Based on these conditions, AFA-P01 is considered A. OPERABLE, and SR 3.0.4 allows changing modes only after performing a risk assessment B. OPERABLE, because SR 3.0.1 allows the completion of required surveillances when plant conditions support C. INOPERABLE, however the mode change can be completed and the required surveillances must be completed within a MAXIMUM of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. INOPERABLE, however the mode change can be completed and the required surveillances must be completed within a MAXIMUM of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

Proposed Answer: B Explanations:

A. Plausible since AFA-P01 is operable, and its plausible since 3.0.4 addresses changing modes and when to perform a risk assessment.

B. Correct C. Plausible since not all surveillances on AFA-P01 have been completed. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is plausible since when a surveillance is out of periodicity, the time requirement to complete the surveillance is that surveillances completion time or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, whichever is longer.

D. Plausible since not all surveillances on AFA-P01 have been completed. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is plausible since it is the time requirement to perform SR 3.7.5.3 once at NOT.

Question Source: New X Bank Modified X Previous NRC Exam 2016 Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 2 Reference N Provided:

Learning Objective: 21081 - Concerning Technical Specification, describe the requirements of SR 3.0.1 in accordance with Tech Specs

Technical

Reference:

Technical Specifications Technical

Reference:

Technical Specifications Bases Technical

Reference:

Technical Specifications Technical

Reference:

SRO Level flowchart from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Ability to approve release permits Tier 3 Group K/A G 2.3.2 IR 3.8 Question 98 Given the following conditions:

A large break LOCA has occurred.

A Site Area Emergency has been declared Due to emergency conditions, a gaseous radioactive release from Containment must be performed to relieve pressure in the Containment and bring the plant to a safer condition.

(1) During a SAE, releases ___(1)___ exceed EPA Protective Action Guidelines (PAGs) at the site boundary (2) The SM/CRS ___(2)___ the only personnel that may AUTHORIZE the release A. (1) WILL (2) ARE B. (1) WILL (2) are NOT C. (1) will NOT (2) ARE D. (1) will NOT (2) are NOT

Proposed Answer: C Explanations:

A. First part is plausible because during a SGTR when ADVs must be used for the initial cooldown, federal limits are not exceeded. That scenario would be an Alert declaration. It may be assumed that because an SAE is the next higher declaration, PAGs will be exceeded. Second part is correct.

B. First part is plausible because during a SGTR when ADVs must be used for the initial cooldown, federal limits are not exceeded. That scenario would be an Alert declaration. It may be assumed that because an SAE is the next higher declaration, PAGs will be exceeded. Second part is plausible because the RP Manager and Sr VP Site Operations must acknowledge the release but they cannot authorize the release.

C. Correct D. First part is correct. Second part is plausible because the RP Manager and Sr VP Site Operations must acknowledge the release but they cannot authorize the release.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 4 Reference N Provided:

Learning Objective: 25949 - Describe whose authority is needed to exceed requirements and what reporting is necessary

Technical

Reference:

74RM-9EF20, Gaseous Radioactive Release and Offsite Dose Assessment

Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Knowledge of general guidelines for EOP usage Tier 3 Group K/A G 2.4.14 IR 4.5 Question 99 (1) Per EOP Operations Expectations, during performance of SPTAs, when a step requires going to contingency actions (2) Per EOP Operations Expectations, once an EOP is entered the CRS should ensure that Safety Function Status Checks are completed within a MAXIMUM of A. (1) CRS concurrence is REQUIRED (2) 15 minutes B. (1) CRS concurrence is REQUIRED (2) 30 minutes C. (1) CRS concurrence is NOT required (2) 15 minutes D. (1) CRS concurrence is NOT required (2) 30 minutes

Proposed Answer: A Explanations:

A. Correct B. First part is correct. Second part is plausible because the STA (normally performs SFSCs) will perform Accountability and a Core Damage Assessment within 30 minutes of an EAL being exceeded.

C. First part is plausible because when Reactor trip and ESFAS setpoints are exceeded, CRS concurrence is not required for manual actuation. Second part is correct.

D. First part is plausible because when Reactor trip and ESFAS setpoints are exceeded, CRS concurrence is not required for manual actuation. Second part is plausible because the STA (normally performs SFSCs) will perform Accountability and a Core Damage Assessment within 30 minutes of an EAL being exceeded.

Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: X Memory or Fundamental Knowledge Comprehension or Analysis Level of Difficultly: 2 10CFR55.43: 5 Reference N Provided:

Learning Objective: Given that an ORP, FRP, or LMFRP is in use, describe the performance of the Safety Function Status Checks in accordance with 40DP-9AP16

Technical

Reference:

EOP Operations Expectations Technical

Reference:

EOP Operations Expectations Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Examination Outline Cross-

Reference:

Level RO SRO K/A: Knowledge of the lines of authority during Tier 3 implementation of the emergency plan Group K/A G 2.4.37 IR 4.1 Question 100 Given the following conditions:

It is Thursday afternoon during a non-holiday workday At time = 1300 - An ALERT is declared for an event in progress At time = 1530 - A GENERAL EMERGENCY is declared The Emergency Coordinator is located in the ___(1)___ and the PAR will be performed by the ___(2)___.

A. (1) Control Room (2) Emergency Coordinator B. (1) Control Room (2) Emergency Operations Director (EOD)

C. (1) Technical Support Center (TSC)

(2) Emergency Coordinator D. (1) Technical Support Center (TSC)

(2) Emergency Operations Director (EOD)

Proposed Answer: D Explanations:

A. Plausible because if the General Emergency was declared less than an hour after the Alert, then the Shift Manager will still have EC duties in the Control Room including PARs.

B. First part is plausible because if the General Emergency was declared less than an hour after the Alert, then the Shift Manager will still have EC duties in the Control Room including PARs. Second part is correct.

C. First part is correct. Second part is plausible if the General Emergency was declared less than an hour after the Alert, then the Shift Manager will still have EC duties in the Control Room including PARs.

D. Correct Question Source: X New Bank Modified Previous NRC Exam Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis Level of Difficultly: 3 10CFR55.43: 5 Reference N Provided:

Learning Objective: 64480 - State the purpose and location of the onsite and offsite Emergency Response Facilities

Technical

Reference:

EP-0904, ERO/ERF Activation and Operation Technical

Reference:

EP-0904, ERO/ERF Activation and Operation Technical

Reference:

SRO Level Question Criteria from NUREG-1021 Technical

Reference:

SRO Level Question - SRO Master Task List 1 B 26 A 51 D 76 D 2 C 27 B 52 C 77 B 3 A 28 D 53 D 78 B 4 D 29 C 54 A 79 B 5 A 30 D 55 A 80 C 6 C 31 B 56 C 81 D 7 A 32 D 57 B 82 D 8 D 33 C 58 B 83 B 9 D 34 B 59 C 84 D 10 A 35 A 60 D 85 D 11 A 36 C 61 B 86 A 12 A 37 C 62 A 87 C 13 B 38 A 63 D 88 A 14 A 39 C 64 C 89 B 15 A 40 C 65 D 90 B 16 D 41 A 66 B 91 A 17 D 42 C 67 B 92 A 18 A 43 B 68 D 93 A 19 A 44 A 69 B 94 C 20 A 45 B 70 C 95 C 21 B 46 D 71 C 96 C 22 B 47 B 72 B 97 B 23 B 48 D 73 D 98 C 24 D 49 B 74 C 99 A 25 D 50 A 75 C 100 D