ML17096A744

2017 Perry Nuclear Power Plant Initial Licensed Operator Examination Administered Written Examinations

ML17096A744
Person / Time
Site:	Perry
Issue date:	04/05/2017
From:	David Reeser NRC/RGN-III/DRS/OLB
To:	NRC/RGN-III/DRS/OLB, FirstEnergy Nuclear Operating Co
Shared Package
ML16242A242	List: ML17096A729 ML17096A732 ML17096A737 ML17096A738 ML17096A744
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Download: ML17096A744 (212)

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NRC 2017 Exam QUESTION RO 1 The Reactor Mode Switch is in SHUTDOWN.

Reactor Coolant Temperature is 140°F.

Which of the following describes the minimum additional condition(s) which will result in a Mode change from MODE 4, COLD SHUTDOWN to MODE 5, REFUEL, per Technical Specifications?

A. The Reactor Mode Switch is placed in REFUEL.

B. All head closure studs are completely de-tensioned.

C. The first head closure stud is less than fully tensioned.

D. The Reactor Mode Switch is placed in REFUEL and the first head closure stud is less than fully tensioned.

1

NRC 2017 Exam QUESTION RO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.1.41 Importance Rating 2.8 K&A: Knowledge of the refueling process.

Generic Explanation: Answer C - TS Table 1.1-1 specifies that REFUEL Mode exists with the Mode Switch in SHUTDOWN or REFUEL with any head closure stud less than fully tensioned.

A - Incorrect - Plausible, the Mode Switch is normally in REFUEL during refueling, but this is not a prerequisite for MODE 5, REFUEL.

B - Incorrect - Plausible, all head closure studs are de-tensioned in preparation for refueling, the Mode change occurs with the first stud less than fully tensioned. Not the MINIMUM additional condition D - Incorrect - Plausible, but is not the MINIMUM additional condition.

Technical Reference(s): Tech Spec 1.1 Reference Tech Specs p 1.0-5 & 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-02.A.2 Question Source: Bank # 2007 Audit Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 2

NRC 2017 Exam QUESTION RO 2 The plant is operating in MODE 3 when the following events occur at Shift Turnover:

• At 06:00 the off-going Shift Manager is informed that only one of the three on-coming licensed Reactor Operators will be able to report for work.

• The off-going Shift Manager immediately starts taking action to call-in replacement licensed Reactor Operators.

• At 06:45 Shift Relief and Turnover is completed and the entire off-going shift leaves for home.

• At 08:30 two replacement licensed Reactor Operators reports to the Control Room.

Which of the following describes if the requirements of NOP-OP-1002, Conduct of Operations, were followed, including the reason for your decision?

A. All requirements were followed because the replacement licensed Reactor Operators arrived within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. All requirements were followed because only one licensed Reactor Operator is required in MODE 3.

C. All requirements were not followed because three licensed Reactor Operators are required in MODE 3.

D. All requirements were not followed because one licensed Reactor Operator from the off-going shift should have been held over until a replacement licensed Reactor Operator arrived.

3

NRC 2017 Exam QUESTION RO 2 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.1.4 Importance Rating 3.3 K&A: Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

Generic Explanation: Answer D - Per NOP-OP-1002, Section 4.1.13, Shift crew composition may be one less than minimum manning requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty crew members. This provision does not allow any shift crew position to be unmanned upon shift change due to an oncoming shift crew person being late or absent. Per Attachment 4, two (2) licensed Reactor Operators are required to be on-shift in Mode 3 for the Control Room.

A - Incorrect - All requirements of NOP-OP-1002, Section 4.1.13 were not followed even though the replacement did arrive within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B - Incorrect - Per NOP-OP-1002, Attachment 4, only 2 Reactor Operators are required in Mode 3. Note: the FBL does not have to be a licensed Reactor Operator. Note: Per Attachment 4, the Field Supervisor position is not required.

C - Incorrect - Per NOP-OP-1002, Attachment 4, only 2 Reactor Operators are required in Mode 3. Note: the FBL does not have to be a licensed Reactor Operator. Note: Per Attachment 4, the Field Supervisor position is not required.

Technical Reference(s): NOP-OP-1002 Rev 11 Reference Attached: NOP-OP-1002 pp 22 & 100 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-K Question Source: Bank # RQL-41989 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 4

NRC 2017 Exam QUESTION RO 3 An I&C tech is performing a surveillance that inserts multiple 1/2 scram signals.

The I&C tech discussed the expected 1/2 scram alarms with only the US.

Which of the following describes the alarm response expectations per NOP-OP-1002, Conduct of Operations?

The ATC ____.

A. is not required to announce expected 1/2 scram annunciators B. communicates alarms to I&C tech without US involvement C. informs the US each time a 1/2 scram annunciator is received D. informs the BOP each time a 1/2 scram annunciator is received 5

NRC 2017 Exam QUESTION RO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.1.17 Importance Rating 3.9 K&A: Ability to make accurate, clear, and concise verbal reports.

Generic Explanation: Answer C - Per NOP-OP-1002, The receipt of expected alarms shall be announced as such to the Command SRO A - Incorrect - Plausible since receipt of expected alarms with US concurrence are only verified to be expected with another RO or work group individual. However, for an alarm to be considered expected, it must be pre-discussed between the US and RO.

B - Incorrect - Plausible if this was pre-discussed between the US and RO.

D - Incorrect - Plausible if this was pre-discussed between the US and RO.

Technical Reference(s): NOP-OP-1002 Rev 11 Reference Attached: NOP-OP-1002 pp 55-56 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-I Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 6

NRC 2017 Exam QUESTION RO 4 MOV testing was performed on 1P45-F130A, ESW PUMP A DISCH VALVE.

The post maintenance test for 1P45-F130A specified stroke timing the valve IAW SVI-P45-T2001, ESW Pump A and Valve Operability Test.

What is the proper method for obtaining valve stroke time?

A. Simultaneously take the valve control switch to OPEN and start the stopwatch.

Stop the stopwatch when the GREEN light extinguishes.

B. Simultaneously take the valve control switch to OPEN and start the stopwatch.

Stop the stopwatch when valve movement has ceased as reported by field observation.

C. Take the valve control switch to OPEN.

Start the stopwatch when the RED indicating light illuminates.

Stop the stopwatch when the GREEN light extinguishes.

D. Take the valve control switch to OPEN.

Start the stopwatch when the RED indicating light illuminates.

Stop the stopwatch when valve movement has ceased as reported by field observation.

7

NRC 2017 Exam QUESTION RO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.2.21 Importance Rating 2.9 K&A: Knowledge of pre- and post-maintenance operability requirements.

Generic Explanation: Answer A - Per SVI-P45-T2001 (and all valve stroke SVIs) stroke time is measured from initiation of control signal (take switch to OPEN) to receipt of desired position indication (GREEN) light extinguishes.

B - Incorrect - Stroke timing is performed using control room indications. Plausible since some throttle valves are positioned based on field observation and may confuse with PI Verification.

C - Incorrect - The stopwatch is started at the same instant the control switch is manipulated. Plausible since the RED light illuminates when control switch taken to OPEN for solenoid valves.

D - Incorrect - The stopwatch is started at the same instant the control switch is manipulated and stroke timing is performed using control room indications.

Technical Reference(s): SVI-P45-T2001 Rev 30 Reference Attached: SVI-P45-T2001 p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3303-01-29 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 8

NRC 2017 Exam QUESTION RO 5 What is the title of the position assigned responsibility for issuing Clearances and keeping Control Room personnel informed of all plant configuration changes prior to establishing or removing a Clearance?

A. Clearance Holder B. Clearance Authority C. Work Group Supervisor D. Operating Representative 9

NRC 2017 Exam QUESTION RO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.2.13 Importance Rating 4.1 K&A: Knowledge of tagging and clearance procedures.

Generic Explanation: Answer B - IAW NOP-OP-1001, the Clearance Authority is responsible for authorizing and issuing clearances and keeping the control room informed.

A - Incorrect - The Clearance Holder accepts the clearance.

C - Incorrect - The Work Group supervisor is responsible for reviewing the clearance, not approving it.

D - Incorrect - The Operating Representative performs clearance duties such as hanging/removing clearances.

Technical Reference(s): NOP-OP-1001 Rev 23 Reference Attached: NOP-OP-1001 p 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): TAGCLRAUTH_FEN-Clearance Authority Question Source: Bank # Perry 2007-2 # RO-69 Modified Bank #

New Question History: Previous NRC Exam Perry 2007-2 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 10

NRC 2017 Exam QUESTION RO 6 Which of the following is an acceptable method to alert the Operator of Control Room annunciators that have been removed from service?

A. Danger Tag B. Information Tag C. Temporary Modification Tag D. Minor Deficiency Monitoring (MDM) Tag 11

NRC 2017 Exam QUESTION RO 6 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.2.43 Importance Rating 3.0 K&A: Knowledge of the process used to track inoperable alarms.

Generic Explanation: Answer B - IAW PAP-1404, Info tags or Caution tags are to be used to identify Control Room annunciators that are removed from service.

A - Incorrect - Although the Caution Tags can be used to track annunciators removed from service, Danger Tags are not used. Plausible if operator not very familiar with tagging procedure.

C- Incorrect - Although the Temp Mod procedure controls annunciators removed from service, TM tags are not used. Additionally, Not-in-Service stickers are no longer allowed to be used to identify OOS annunciators in the Control Room. Plausible if operator not very familiar with TM procedure.

D - Incorrect - The MDM Process is for the management of maintenance deficiencies whose significance is so minor that it would not be prudent to remove the equipment from service to repair. Not tracking of annunciators.

Plausible if operator confuses these tags with Repair Tags.

Technical Reference(s): PAP-1404 Rev 7 & NOP-OP- Reference Attached: PAP-1404 p 4 & NOP-OP-1014 Rev 4 1014 p 26 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-02-F Question Source: Bank # Perry 2015 # RO-06 Modified Bank #

New Question History: Previous NRC Exam Perry 2015 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 12

NRC 2017 Exam QUESTION RO 7 The Reactor has been shutdown in order to replace a defective fuel bundle.

RPV Pressure is 100 psig with a cooldown in progress.

Radiation Protection tech reports that a Containment atmosphere air sample indicates iodine levels are at 0.5 DAC.

How should Containment Vessel and Drywell Purge System (M14) be operated?

A. in the Refuel Mode B. in Containment Venting C. in the Intermittent Mode D. in Single Train Drywell Ventilation Operation 13

NRC 2017 Exam QUESTION RO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.3.14 Importance Rating 3.4 K&A: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Generic Explanation: Answer C - SOI-M14 P&L 2.5 CVDWP operation when iodine levels > 0.3 DAC. With the RPV pressure at 100 psig, the Rx is in Mode 3. CVDWP can only be operated in Intermittent Mode in Mode 3.

A - Incorrect - Refuel Mode not available in Mode 3.

B - Incorrect - Containment Venting shall not be used if Intermittent Mode is available.

D - Incorrect - Single Train Drywell Operation not available in Mode 3.

Technical Reference(s): SOI-M14 Rev 25 Reference Attached: SOI-M14 pp 4 & 64 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M14-H Question Source: Bank # Perry 2007 # SRO-05 Modified Bank #

New Question History: Previous NRC Exam Perry 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 14

NRC 2017 Exam QUESTION RO 8 A transient has occurred that requires venting containment.

IAW EOP-02, Primary Containment Control, Containment venting should only be performed if ____.

A. Primary Containment Limit will be exceeded B. Pressure Suppression Pressure will be exceeded C. Evacuation of affected areas has been completed D. Containment Spray cannot control containment temperature 15

NRC 2017 Exam QUESTION RO 8 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.3.11 Importance Rating 3.8 K&A: Ability to control radiation releases.

Generic Explanation: Answer A - IAW EOP-02, containment venting is performed prior to exceeding PCL.

B - Incorrect - PSP is 15 psig for Perry and preparation for containment venting occurs at this pressure. PCL is much higher. No venting is required for exceeding PSP.

C - Incorrect - While venting should be coordinated with evacuation, there is no requirement to have evacuation complete prior to venting.

D - Incorrect - Venting is done if containment sprays cannot maintain containment pressure, not containment temperature.

Technical Reference(s): EOP-2 Bases Rev 3 Reference Attached: EOP-2 Bases pp 68 - 70 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-09-B Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 16

NRC 2017 Exam QUESTION RO 9 What is the definition of Minimum Steam Cooling Pressure?

MSCP is the lowest RPV pressure at which ____.

A. the covered portion of the reactor core will generate sufficient steam to prevent fuel clad temperature in the uncovered portion of the core from exceeding 2200°F B. the covered portion of the reactor core will generate sufficient steam to prevent fuel clad temperature in the uncovered portion of the core from exceeding 1500°F C. steam flow through open SRVs is sufficient to preclude fuel clad temperature from exceeding 2200°F even if the core is not completely covered D. steam flow through open SRVs is sufficient to preclude fuel clad temperature from exceeding 1500°F even if the core is not completely covered 17

NRC 2017 Exam QUESTION RO 9 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.4.17 Importance Rating 3.9 K&A: Knowledge of EOP terms and definitions.

Generic Explanation: Answer D - This is the definition per EOP Bases.

A - Incorrect - This is a combination of the definition of Minimum Zero Injection RPV Water Level and the General Design Criteria Peak Cladding Temperature.

B - Incorrect - This is the definition of Minimum Zero Injection RPV Water Level.

C - Incorrect - This is the wrong temperature for the MSCP.

Technical Reference(s): EOP Bases Rev 6 Reference Attached: EOP Bases p 43 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-01-C.27 Question Source: Bank # Columbia 2009 # RO-74 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 18

NRC 2017 Exam QUESTION RO 10 A plant startup is in progress with the following conditions:

• Rx Power is at 18%.

• Main condenser vacuum at 4.5HgA.

• Feedwater shift from MFP to RFPT B just commenced.

Chart recorder N21-R183, Main Condenser Shell Vacuum indicates vacuum is degrading at 0.5HgA per minute.

With no operator action, which of the following is the expected time interval until SRVs operate?

A. 7 minutes B. 14 minutes C. 31 minutes D. 34 minutes 19

NRC 2017 Exam QUESTION RO 10 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.4.47 Importance Rating 4.2 K&A: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Generic Explanation: Answer C - This is the time that corresponds to Bypass Valve closure. At this point pressure control needs to be transferred to SRVs.

A - Incorrect - This is the time that corresponds to the Main Turbine trip. At 18% Rx power, Bypass valves will open and control RPV pressure.

B - Incorrect - This is the time that corresponds to a RFPT trip. Since a FW shift is in progress, the Motor Feed Pump will continue to run for level control.

D - Incorrect - This is the time that corresponds to MSIV closure. Pressure control will need to be on SRVs, but this is not the earliest time.

Technical Reference(s): ONI-N62 Rev 10 Reference Attached: ONI-N62 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-10(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 20

NRC 2017 Exam QUESTION RO 11 The plant was at full power when Reactor Recirculation Pump A tripped.

Conditions have been established for single loop operation.

What is the minimum value of the MCPR Safety Limit allowed by Tech Specs?

A. 0.99 B. 1.10 C. 1.13 D. 1.34 21

NRC 2017 Exam QUESTION RO 11 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295001 AK1.03 Importance Rating 3.6 K&A: Knowledge of the operational implications of the following concepts as they apply to Partial Or Complete Loss Of Forced Core Flow Circulation: Thermal limits Partial or Complete Loss of Forced Core Flow Circulation Explanation: Answer C - TS 2.1.1.2 requires MCPR to be 1.13 for single loop operation. Therefore, this is the minimum value. MCPR is also a thermal limit. The 3-D Monicor computer monitors Recirc pump breaker position and automatically adjusts MCPR limit for a tripped pump.

A - Incorrect - Plausible since MFLCPR must be 1.0.

B - Incorrect - Plausible, as this is true for two loop operation.

D - Incorrect - Plausible, as this is the MCPR F and MCPR P power and flow operating limits for a certain type of fuel loaded in the core.

Technical Reference(s): Tech Spec 2.1.1.2 Reference Tech Specs p 2.0-1 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-03-A Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments:

22

NRC 2017 Exam QUESTION RO 12 The plant is operating at 90% rated power.

Interbus Transformer LH-1-B is tagged out for performance of a deluge test when the following occurs:

• Interbus Transformer LH-1-C trips on Neutral Ground Overcurrent.

With no operator action, which of the following describes the status of the plant two minutes after LH-1-C trips?

A. The plant will scram. RPV level will be controlled by HPCS and RCIC.

B. The plant will scram. RPV level will be controlled by the Motor Feed Pump.

C. The plant will not scram. Rx power will remain the same and RFPTs will control RPV level.

D. The plant will not scram. However, Rx power will lower and RFPT B and the Motor Feed Pump will control RPV level.

23

NRC 2017 Exam QUESTION RO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295003 AA2.02 Importance Rating 4.2 K&A: Ability to determine and/or interpret the following as they apply to Partial Or Complete Loss Of A.C. Power: Reactor power / pressure / and level Partial or Complete Loss of AC Explanation: Answer A - With LH-1-B tagged out, all BOP loads will be on LH-1-C. When LH-1-C trips, the RFBPs trip resulting in a loss of RFPTs and the MFP. The loss of feed causes a reactor scram and the auto start of RCIC and HPCS on L2. (This occurs in about 35 seconds) This was run in the simulator on 4/29/16. RCIC and HPCS will shutoff at L8 in about 4 minutes.

B - Incorrect - Plausible since the MFP is fed from a 13.8 KV bus. However, the loss of RFBPs cause a trip on the MFP.

C - Incorrect - Plausible if operator does not recall the Bus H11 is normally fed from LH-1-B and now transferred to LH-1-C.

C - Incorrect - Plausible since initially, the Recirc system runs back. However, the loss of feed will cause a Rx scram.

Technical Reference(s): ARI-H13-P870-01 Rev 15, PDB- Reference Attached: ): ARI-H13-P870-01 p 47, H06 Rev 0, ONI-R22-2 Rev 10 PDB-H06 pp 3 & 4, ONI-R22-2 pp 4 & 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-10(LP)-A.3 Question Source: Bank #

Modified Bank # Limerick 2012 # RO-02 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 24

NRC 2017 Exam QUESTION RO 13 RHR B Pump is running on Minimum Flow.

What impact does losing ED-1-B have on RHR Pump B breaker?

A. Remote breaker tripping is prevented.

Local manual tripping capability remains.

Protective relaying is lost.

B. Remote breaker tripping is prevented.

Local manual tripping capability remains.

Protective relaying will function.

C. Remote and local breaker tripping is prevented.

Protective relaying will function.

D. Remote and local breaker tripping is prevented.

Protective relaying is lost.

25

NRC 2017 Exam QUESTION RO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295004 2.2.37 Importance Rating 3.6 K&A: Ability to determine operability and/or availability of safety related equipment.

Partial or Total Loss of DC Pwr Explanation: Answer A - With the RHR B pump running, and a loss of ED-1-B, remote (control room) tripping capability is lost. Also, ED-1-B provides power to the protective relaying for this pump. The pump can be tripped by the manual trip pushbutton on the front of the breaker.

B - Incorrect - Plausible if operator thinks protective relaying is provided from safety related 120VAC. Some of the protective relaying has AC inputs, but is powered from DC.

C - Incorrect - Local tripping is still available. Plausible is operator does not know the opening springs are charged when the breaker is closed. And, protective relaying will not function.

D - Incorrect - Local tripping is still available. Plausible is operator does not know the opening springs are charged when the breaker is closed.

Technical Reference(s): ONI-R42-2 Rev 7 and SDM-R10 Reference Attached: ONI-R42-2 pp 5 & 8 and Rev 12 SDM-R10 p 57 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-05(LP)-A.2 Question Source: Bank # Nine Mile 2 2012 # RO-75 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 26

NRC 2017 Exam QUESTION RO 14 The plant was operating at 25% rated power with the following conditions:

• The Motor Feed Pump was tagged out for a lube oil leak.

• RFPT A is in AUTO

• RFPT B is shutdown Then a main turbine trip occurred.

Two minutes later the following conditions exist:

• The Reactor Mode Switch is in SHUTDOWN

• No other Operator actions were performed

• RFPT A is in AUTO on Setpoint Setdown

• RPV Water Level is 185

• RPV Pressure is at 900 psig What is the status of RPV level control when RPV pressure lowers to 775 psig?

RFPT A is ____.

A. feeding the vessel with RFPT speed stable, and RPV level is rising B. feeding the vessel with RFPT speed decreasing, and RPV level is rising C. not feeding the vessel with RFPT speed stable, and RPV level is lowering D. not feeding the vessel with RFPT speed decreasing, and RPV level is lowering 27

NRC 2017 Exam QUESTION RO 14 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295005 AK1.03 Importance Rating 3.5 K&A: Knowledge of the operational implications of the following concepts as they apply to Main Turbine Generator Trip: Pressure effects on reactor level Main Turbine Generator Trip Explanation: Answer A - With Setpoint Setdown activated and RPV level at 185 and RFPT in AUTO, it will not be feeding the RPV and the turbine speed will be ~3300 RPM with discharge pressure about 800 psig. Once RPV pressure lowers to <800 psig, the RFPT will start to feed, but speed will remain @ 3300 rpm. Since pressure is <940 psig, the bypass valves will be closed resulting in minimal inventory loss and RPV level rising. This response was verified in the simulator on 7/11/16 - RJT.

B - Incorrect - RFPT speed will remain the same.

C - Incorrect - RFPT will be feeding the RPV. Plausible that level is lowering if not feeding.

D - Incorrect - RFPT will be feeding the RPV and RFPT speed will remain the same. Plausible that level is lowering if not feeding.

Technical Reference(s): OAI-1703 Rev 27 and SOI-C34 Reference Attached: OAI-1703 p 40 and SOI-C34 Rev 35 p 35 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N27-C.6 & OT-3035-01(LP)-A.1 Question Source: Bank # Perry 2009 # RO-15 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 28

NRC 2017 Exam QUESTION RO 15 The plant is starting up with main generator synchronization in progress per IOI-3, Power Changes.

I&C is troubleshooting 1A MSR Drain Tank Normal and Alternate drain valves.

Both drain valves stick closed.

Then, annunciator MOISTURE SEPARATOR DRN TANK 1A LVL HIGH alarms.

What is the expected plant response if the 1A MSR Drain Tank drain valves cannot be reopened?

A. Reactor scram only B. Main turbine trip only C. Reactor scram and main turbine trip D. No reactor scram or main turbine trip 29

NRC 2017 Exam QUESTION RO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295006 AK2.04 Importance Rating 3.6 K&A: Knowledge of the interrelations between SCRAM and the following: Turbine trip logic SCRAM Explanation: Answer B - With gen synch in progress, Rx power is ~15-18%. A high level in any MSR level instrument will cause a turbine trip. However, since Rx power is within the capability of the bypass valves, the Rx does not scram.

A - Incorrect - Plausible is operator believes Rx will scram but no turbine trip will occur if generator not synchronized to grid.

C - Incorrect - No Rx scram will occur. Plausible if operator assumes a turbine trip at this power level will cause a Rx scram.

D - Incorrect - Plausible since most other logic trip systems are at least 1 out of 2, not single input.

Technical Reference(s): ARI-H13-P870-05 Rev 4 & ONI- Reference Attached: ARI-H13-P870-05 p 3 & ONI-N32 Rev 11, Dwg 302-111 Rev HH, SDM-N36 Rev 9 N32 p 4, Dwg 302-111, SDM-N36 pp 16-17 & 22 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-09(LP)-B.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 30

NRC 2017 Exam QUESTION RO 16 SVI-P57-T2001, Safety-Related Instrument Air Motor Operated Valve Operability Test was in progress with P57-F015A in the closed position when the following occurred:

• A fire occurred in the Control Room

• Immediate actions per ONI-C61 were taken

• SVI-P57-T2001 was suspended as is

• The Unit Supervisor directed performance of Control Room Isolation per IOI-11.

• The Remote Shutdown switches were placed to EMER (Emergency) for valves:

• P57-F015A, CNTMT ADS SUPPLY OTBD ISOL VALVE

• P57-F020A, DW ADS SUPPLY OTBD ISOL VALVE What is the status of P57-F015A and P57-F020A after placing the Remote Shutdown Switches in EMER?

P57-F015A P57-F020A A. open open B. open shut C. shut open D. shut shut 31

NRC 2017 Exam QUESTION RO 16 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295016 AK2.02 Importance Rating 4.0*

K&A: Knowledge of the interrelations between Control Room Abandonment and the following: Local control stations: Plant-Specific Control Room Abandonment Explanation: Answer A - Transferring P57-F015 & F020 to EMERGENCY on the MCC, results in the valves being opened even if they were previously closed.

B - Incorrect - P57-F020A will also be open.

C - Incorrect - P57-F015A will also be open.

D - Incorrect - Both P57-F015A and P57-F020A will be open.

Technical Reference(s): SVI-P57-T2001 Rev 7, SVI-P57- Reference ONI-C61 SVI-P57-T2001 p 3, SVI-P57-T2003 Rev 3, IOI-011 Rev 36 & Dwg 208-199 SH-01 Rev T2003 pp 4 & 6, IOI-011 p 119 & Dwg 208-199 N & SH-03 Rev R SH-01 & SH-03 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C61-F.1 Question Source: Bank # Perry 2013 # RO-16 Modified Bank #

New Question History: Previous NRC Exam Perry 2013 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 32

NRC 2017 Exam QUESTION RO 17 The plant is operating at 100% power.

• TBCC HX OUTLET TEMP HIGH alarm is received on panel H13-P870

• ONI-P44, Loss of Turbine Building Closed Cooling, has been entered

• TBCC Heat Exchanger Outlet Temperature Control Valve, 1P41-F003, was confirmed to have failed in the close position Which of the following describes the plant response to the loss of TBCC if no operator actions are taken?

A. The running Service Air Compressor will trip when its lube oil temperature reaches 158 °F.

B. The Main Turbine will trip when the Main Lube Oil Cooler outlet temperature reaches 125 °F.

C. The Rx Feed Pump Turbines will trip when the RFPT lube oil cooler outlet temperature reaches 135 °F.

D. The running Isolated Phase Bus Cooling Fan will trip when the Isolated Phase Bus Duct temperature reaches 185 °F.

33

NRC 2017 Exam QUESTION RO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295018 AK2.01 Importance Rating 3.3 K&A: Knowledge of the interrelations between Partial Or Complete Loss Of Component Cooling Water and the following: System loads Partial or Total Loss of CCW Explanation: Answer D - The running IsoPhase Bus cooling fan will trip at 185°F in the bus duct.

A - Incorrect - Plausible since the Service Air compressor trips at 158°F LO temperature but is cooled by NCC not TBCC.

B - Incorrect - Plausible since Main Turbine LO alarms at 125°F, but it will not trip. The MT will trip on a stator water run back.

C - Incorrect - Plausible since RFPTs LO alarms at 135°F, but it will not trip.

Technical Reference(s): ONI-P44 Rev 11, ARI-H13- Reference Attached: ONI-P44 pp 3-4, ARI-H13-P680-07 Rev 26, ARI-H13-P870-08 Rev 7, & ARI-H13- P680-07 p 115, ARI-H13-P870-08 p 7, & ARI-H13-P680-15 Rev 6 P680-15 p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P44-J.2 & OT-3035-02(LP)-A.1 Question Source: Bank # Perry 2001 # RO-20 Modified Bank #

New Question History: Previous NRC Exam Perry 2001 Retake Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 34

NRC 2017 Exam QUESTION RO 18 With the plant operating at rated power, an air leak occurs resulting in the following:

• Unit 1 and Unit 2 Instrument Air receiver pressures are 85 psig and lowering

• Unit 1 and Unit 2 Service Air receiver pressures are 95 psig and lowering Which of the following describes how the Service Air/Instrument Air Cross-Connect Valves, 1P52-F050

& 2P52-F050, respond to these conditions, including the bases for this response?

The Service Air/Instrument Air Cross-Connect Valves ____.

A. close to completely isolate the Service Air headers from the Instrument Air headers B. close to prevent a leak in the Service Air header from impacting the Instrument Air header C. remain open. However, they will close if Service Air receiver pressure lowers to 90 psig in order to completely isolate the Service Air and Instrument Air headers D. remain open. However, they will close if Instrument Air receiver pressure lowers to 80 psig in order to prevent a leak in the Service Air header from impacting the Instrument Air header 35

NRC 2017 Exam QUESTION RO 18 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295019 AK3.03 Importance Rating 3.2 K&A: Knowledge of the reasons for the following responses as they apply to Partial Or Complete Loss Of Instrument Air: Service air isolations Partial or Total Loss of Inst. Air Explanation: Answer B - The cross-connect valves will close when Instrument Air receiver pressure is <90 psig to protect the Instrument Air system from a leak in the Service Air system.

A - Incorrect - Check valves around the P52-F050 valves allow Service Air to continue to supply Instrument Air when the F050 valves are closed C - Incorrect - P52-F050 valves are closed - Service air can still supply instrument air header. Therefore, they are not completely isolated from each other D - Incorrect - P52-F050 valves are closed. Plausible, as some valves in various systems begin to reposition to their fail positions at 80 psig.

Technical Reference(s): SOI-P51/52 Rev 31 & Lesson Reference Attached: SOI-P51/52 p 4 & Lesson Plan OT-COMBINED-P51_P52 Rev 4 Plan OT-COMBINED-P51_P52 p 11 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P51_52 #17 Question Source: Bank # Perry 2010 # RO-17 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 36

NRC 2017 Exam QUESTION RO 19 The plant is cooling down for a maintenance outage with the following conditions:

• RHR Pump 'B' is operating in Shutdown Cooling Mode at 7000 gpm

• RPV level is 200 inches

• RPV pressure is 15 psig Then, RPV level lowered to 175 inches.

Given these conditions, which of the following describes the impact, if any, to the RHR System?

NOTE - Valve Functional Locations are as follows:

• 1E12-F008, SHUTDOWN COOLING OUTBD SUCT ISOL VLV

• 1E12-F009, SHUTDOWN COOLING INBD SUCT ISOL VLV

• 1E12-F053B, RHR B TO FEEDWATER S/D COOLING RTRN VLV A. Only 1E12-F009 closes and RHR pump B trips B. RHR Pump 'B' continues to operate at 7000 gpm C. 1E12-F008 and 1E12-F009 close and RHR pump B trips D. 1E12-F053B closes; RHR Pump 'B' continues to runs on Min Flow 37

NRC 2017 Exam QUESTION RO 19 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295021 AA1.02 Importance Rating 3.5 K&A: Ability to operate and/or monitor the following as they apply to Loss Of Shutdown Cooling: RHR/shutdown cooling Loss of Shutdown Cooling Explanation: Answer C - When RPV level lowers to 178, a L3 SDC isolation occurs closing all SDC valves.

A - Incorrect - Plausible if the operator incorrectly believes that F009 (Div 2 valve) does not receive an auto closure signal with RHR Pump 'B' (Div 2 Pump) running.

B - Incorrect - Plausible if the examinee incorrectly believes that the low RPV water level isolation setpoint has not been exceeded and that RHR Pump 'B' is therefore unaffected by the conditions listed in the stem.

D - Incorrect - Plausible if the examinee incorrectly believes that the Min Flow Valve will open in a SDC lineup.

Technical Reference(s): ARI-H13-P680-05 Rev 15, ONI- Reference Attached: ARI-H13-P680-05 p 27, ONI-E12-2 Rev 36, and SDM-E12 Rev 3 E12-2 p 17, and SDM-E12 p 51 Proposed references to be provided to applicants during examination: None Learning Objective (As available): to-combined-E12-F & OT-3035-11(LP)-A.1 Question Source: Bank # Clinton 2013 # RO-12 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 38

NRC 2017 Exam QUESTION RO 20 The plant is shutdown for a refuel outage with the following conditions:

• Fuel movement is in progress in the Spent Fuel Pool.

• An irradiated fuel assembly has just been loaded on the Fuel Handling Bridge main hoist and raised to the Full Up position.

Then, a seismic event results in the following:

• The Fuel Handling Bridge main hoist cannot be moved.

• The common suction line from the Fuel Pool Cooling and Cleanup pumps completely ruptures.

Which of the following describes the Spent Fuel Pool water level response to this event and the availability of Spent Fuel Pool makeup?

The fuel assembly on the hoist will be (1) .

Makeup water to the Spent Fuel Pool is available from (2) .

(1) (2)

A. fully submerged Condensate Transfer System B. fully submerged Fire Water System using hoses C. partially uncovered Condensate Transfer System D. partially uncovered Fire Water System using hoses 39

NRC 2017 Exam QUESTION RO 20 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295023 AA2.02 Importance Rating 3.4 K&A: Ability to determine and/or interpret the following as they apply to Refueling Accidents:

Fuel pool level Refueling Acc Explanation: Answer B - With a rupture on the FPCC pump line, water in the Spent Fuel Pool will only lower a few inches to the siphon breakers. The main hoist will not raise the fuel higher than 8 3 below the Fuel Handling Building Floor. With the FPCC pump line rupture, makeup will be from the Fire Water system as the Condensate transfer system puts water into the surge tanks.

A & C (2nd part) - Incorrect - CTS will fill the surge tanks, not the Fuel Pools without the FPCC pumps.

C & D (1st part) - Incorrect - The fuel will remain fully submerged. Plausible if the operator does not recall that the siphon breakers will limit pool level loss to a few inches.

RO level justification - At Perry, RO will supervise bridge movements in the Fuel Handling Building.

Technical Reference(s): ONI-E12-2 Rev 36, SOI-F11 Rev Reference Attached: ONI-E12-2 pp58-59, SOI-F11 18, SDM-G41 Rev 7, Dwg 302-654 Rev T & Dwg 302- p 5, SDM-G41 pp 12-13, Dwgs 302-654 & 655 655 Rev Z Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-F11_F15-H, OT-COMBINED-G41-R, OT-3035-11(LP)-A.1 Question Source: Bank #

Modified Bank # Nine Mile 2013 # RO-10 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 40

NRC 2017 Exam QUESTION RO 21 The following conditions exist:

• A small break Loss of Coolant Accident (LOCA) has occurred

• HCPS has automatically initiated

• Drywell pressure peaked at 3.0 psig

• RPV water level has lowered to -10 inches

• The operators are restoring plant parameters at this time How would the HPCS Initiation logic be reset in order to place HPCS in Standby Readiness?

A. Manually after the low RPV level signal clears B. Automatically upon RPV level reaching Level 8 C. Manually after the high DW pressure signal clears D. Automatically after both high DW pressure and low RPV level signals clear 41

NRC 2017 Exam QUESTION RO 21 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295024 2.1.28 Importance Rating 4.1 K&A: Knowledge of the purpose and function of major system components and controls.

High Drywell Pressure Explanation: Answer A - If level is >L2, HPCS initiation logic can be reset even if a High DW Pressure condition still exists. If HPCS is not needed for level control, it may be placed in Standby Readiness.

B - Incorrect - Plausible since HPCS will stop injecting at L8, but the Initiation Logic will not be reset. There is no automatic reset for the HPCS initiation logic.

C - Incorrect - RPV level must be > 129 to reset HPCS initiation logic. Plausible if operator doesnt recall HPCS can be reset only after RPV level recovers above L2 and not by lowering DW pressure..

D - Incorrect - There is no automatic reset for the HPCS initiation logic.

Technical Reference(s): Lesson Plan OT-COMBINED- Reference Attached: E22 Lesson Plan pp 11-12, E22 Rev 4, SOI-E22A Rev 36, & SDM-E22A Rev 8 SOI-E22A p 16, & SDM-E22A p 22 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E22A-F.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 42

NRC 2017 Exam QUESTION RO 22 Following a 300 day run, a failure in the Steam Bypass and Pressure Regulating System caused a reactor scram due to high Rx pressure, last night.

Current conditions are as follows:

• RHR A is running in Shutdown Cooling

• RPV level is 250 stable

• RPV temperature is 135°F stable

• MSIVs are shut

• No Rx Recirc pumps are running Based on expected decay heat load, which of the following describes the response if 1E12-F003A, RHR A HX OUTLET VALVE is throttled closed for 5 seconds?

Bulk reactor water temperature will ____.

A. lower until equal with ESW A Loop temperature B. lower until equal with ambient drywell temperature C. rise until boiling occurs and Rx pressure stabilizes at atmospheric pressure D. rise until boiling occurs causing Rx pressure to rise above atmospheric pressure 43

NRC 2017 Exam QUESTION RO 22 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295025 EK1.04 Importance Rating 3.6 K&A: Knowledge of the operational implications of the following concepts as they apply to High Reactor Pressure: Decay heat generation High Reactor Pressure Explanation: Answer D - With RPV temperature stable, closing the E12-F003 for 5 seconds will significantly reduce amount of cooling provided by RHR. Decay heat generation <1 day after shutdown is between 70-80 MBTUs/Hr. Therefore, temperature will rise and with MSIVs shut, RPV pressure will rise above atmospheric pressure.

A - Incorrect - Plausible if operator believes closing the F003 valve will provide more cooling (i.e. confusing with the F048, Bypass Valve)

B - Incorrect - Plausible if operator assumes decay heat load is low enough (at < 1 day) to be handled by ambient loss.

C - Incorrect - RPV pressure will rise above atmospheric pressure. Plausible if operator assumes head vent will be sufficient to prevent pressure rise.

Technical Reference(s): PDB-A16 Rev 15 & ONI-E12-2 Reference Attached: PDB-A16 p 4 & ONI-E12-2 p Rev 36 83 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-11(LP)-A.1 Question Source: Bank # Peach Bottom 2013 # RO-13 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 44

NRC 2017 Exam QUESTION RO 23 The US has entered ONI-B21-1, SRV INADVERTENT OPENING / STUCK OPEN.

Refer to the attached SUPR POOL TEMP VALIDATION SPDS screen printout.

What does this indicate concerning average suppression pool temperature?

Average Suppression Pool Temperature ____.

Attachment provided:

A. has exceeded the SPDS Alarm High setpoint B. is approaching an EOP Entry value C. is NOT VALIDATED on SPDS D. is VALIDATED on SPDS 45

NRC 2017 Exam QUESTION RO 23 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 EK2.04 Importance Rating 2.5 K&A: Knowledge of the interrelations between Suppression Pool High Water Temperature and the following: SPDS/ERIS/CRIDS/GDS Suppression Pool High Water Temp.

Explanation: Answer C - SP Temp is one of eleven control parameters in SPDS. The SP Temp reading on the Validation screen is the AVERAGE SP Temp. If an SPDS control parameter reading has an invalid input, the box displaying the value turns from cyan to yellow. This indicates that the reading is no longer validated.

A - Incorrect - Plausible since some selected values change color when approaching an EOP setpoint.

B - Incorrect - Plausible, as EOP entry condition values change color (RED) when exceeded D - Incorrect - Plausible if the operator does not recall that a yellow box indicates the reading is not valid.

Technical Reference(s): OT-COMBINED-C91 Lesson Reference Attached: OT-COMBINED-C91 Lesson Plan Rev 1 & SPDS Users Manual Appendix H Rev G Plan p 14 & SPDS Users Manual Appendix H pp 33, 39, 178 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C91-I & O Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 46

NRC 2017 Exam QUESTION RO 24 In accordance with EOP Bases, why is Emergency Depressurization required to be performed prior to reaching the Containment Design Temperature?

A. To preclude containment failure following initiation of containment sprays.

B. The environmental qualification temperature limit of SRV solenoids may be exceeded.

C. The environmental qualification temperature for safety related electrical equipment may be exceeded.

D. The Pressure Suppression function may no longer be able to absorb the energy from a loss of coolant accident.

47

NRC 2017 Exam QUESTION RO 24 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295027 EK3.01 Importance Rating 3.7 K&A: Knowledge of the reasons for the following responses as they apply to High Containment Temperature (Mark III Containment Only): Emergency depressurization: Mark-III High Containment Temperature Explanation: Answer C - The containment design temperature limit of 185°F is based on not exceeding the environmental qualifications of safety related electrical equipment. Emergency depressurizing prior to reaching 185°F will maintain equipment operability for as long as possible.

A - Incorrect - This is the bases for initiating containment spray in the SAFE region of the CSIL graph.

B - Incorrect - This is the bases for the Drywell temperature limit.

D - Incorrect - The Pressure suppression function is based on suppression pool temperature and level.

Technical Reference(s): EOP Bases Rev 6 & EOP-2 Reference Attached: EOP Bases p 66 & EOP-2 Bases Rev 3 Bases pp 77-78 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-07-C Question Source: Bank # INL-0778 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 48

NRC 2017 Exam QUESTION RO 25 The following conditions exist:

• The reactor was scrammed from 25% rated power

• A low power ATWS is in progress

• Emergency Depressurization was required

• All SRVs failed to open

• RCIC suction is on the suppression pool

• RCIC is providing RPV level and pressure control

• Suppression Pool level is lowering 1 inch per minute

• Current suppression pool level is 15 feet 2 inches Per EOP Bases, the earliest that continued operation of RCIC will be threatened due to possible RCIC equipment damage is in ____ minutes.

A. 11 B. 35 C. 95 D. 113 49

NRC 2017 Exam QUESTION RO 25 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295030 EA1.02 Importance Rating 3.4 K&A: Ability to operate and/or monitor the following as they apply to Low Suppression Pool Water Level: RCIC Low Suppression Pool Wtr Lvl Explanation: Answer C - With suppression pool level lowering at 1/minute, it will take 95 minutes to reach 7.25. At this level, pump damage from operation below the vortex limit becomes a concern.

A - Incorrect - This is the time to the Suppression Pool ED level limit B - Incorrect - This is the time to uncover the horizontal vents D - Incorrect - This is the time when damage to RHR and LPCS may occur Technical Reference(s): EOP Bases Rev 6 Reference Attached: EOP Bases pp 60-61 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-01-B.2 Question Source: Bank # Perry 2010 # RO-25 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 50

NRC 2017 Exam QUESTION RO 26 The plant was operating at rated power when a total loss of Feedwater occurred.

The only operator action taken was to place the Mode Switch in Shutdown.

No control rod motion occurred.

Given these conditions, which of the following describes Redundant Reactivity Control System response to lower reactor power?

When RPV level reaches ____.

A. Level 3, the Rx Recirc Pumps will always trip to OFF B. Level 2, the Rx Recirc Pumps will always trip to OFF C. Level 3, the Rx Recirc Pumps will trip to OFF if APRMs are not down scale in 25 seconds D. Level 2, the Rx Recirc Pumps will trip to OFF if APRMs are not down scale in 25 seconds 51

NRC 2017 Exam QUESTION RO 26 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295031 EA2.02 Importance Rating 4.0 K&A: Ability to determine and/or interpret the following as they apply to Reactor Low Water Level: Reactor power Reactor Low Water Level Explanation: Answer B - When RPV level lowers to 2, RRCS causes Recirc Pumps to trip to OFF.

A - Incorrect - At RPV L3, RRCS causes pumps to down shift to slow speed.

C - Incorrect - Plausible if operator mistakes L2 actions for L3 actions.

D - Incorrect - This is the correct action for a Hi Rx Pressure signal, not low Rx water level.

Technical Reference(s): ARI-H13-P680-05 Rev 15 Reference Attached: ARI-H13-P680-05 pp 5 & 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C22-F Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 52

NRC 2017 Exam QUESTION RO 27 The plant was operating at rated power with the High Pressure Core Spray Pump out of service for a motor replacement.

Then a loss of offsite power occurred.

Immediate operator actions for a Reactor Scram were completed.

Five minutes after the loss of power the following conditions exist:

• Division 3 Diesel Generator is the only DG operating

• RCIC auto started and is injecting

• RPV pressure is cycling between approximately 930 psig and 1080 psig

• SRV B21-F051C is cycling open and close

• SRV B21-F051D opened and remains open Which of the following abnormal and/or emergency procedures/charts should have been/will be entered based on the conditions above?

1. ONI-R10-1, Loss of AC Power 2 ONI-R10-2, Station Blackout

3. ONI-B21-1, SRV Inadvertent Opening/Stuck Open

4. EOP-01, RPV Control

5. EOP-1A, Level Power Control A. 1 and 3 B. 2 and 3 C. 2, 3, and 4 D. 2, 4, and 5 53

NRC 2017 Exam QUESTION RO 27 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295037 2.4.4 Importance Rating 4.5 K&A: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

SCRAM Condition Present and Power Above APRM Downscale or Unknown Explanation: Answer D - With a loss of off-site power, and the HPCS DG the only DG running, but the pump is unavailable, then ONI-R10 TLAC is entered. Since RCIC auto started, RPV level lowered to <L2 then EOP-1 is also entered. And, since one SRV is open and one SRV is cycling, the reactor is still producing power and above 4% (1 SRV5%) EOP-1A is also entered.

A - Incorrect - Plausible since the Div 3 DG is running. But, without the HPCS pump, TLAC is entered. Also since 1 SRV is open and 1 is cycling, ONI-B21-1 is also plausible.

B - Incorrect - Plausible since 1 SRV is open and 1 is cycling. But, this indicates the Rx is still making power.

C - Incorrect - Plausible since 1 SRV is open and 1 is cycling. But, this indicates the Rx is still making power.

Technical Reference(s): ONI-B21-1 Rev 11, ONI-R10 Reference Attached: ONI-B21-1 p 12, ONI-R10 pp Rev 13, EOP-1 Bases Rev 6, EOP-1A Bases Rev 8, 3, 8 & 14, EOP-1 Bases p 8 , EOP-1A Bases p 8 ,

PYBP-POS-30 Rev 3 PYBP-POS-30 p 9 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-18(LP)-A.4, OT-3035-07(LP)-E, OT-3402-02-B & F Question Source: Bank #

Modified Bank # River Bend 2003 # SRO-77 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 54

NRC 2017 Exam QUESTION RO 28 What is the EOP Bases for restarting Heater Bay Building ventilation when operating in EOP-5, Radioactivity Release Control?

A. Ensures that Turbine Building air is filtered prior to releasing to the environment.

B. Ensures that Turbine Building air is monitored prior to releasing to the environment.

C. Allows for continued access to the Turbine Building Steam Tunnel without exceeding the maximum safe operating radiation level.

D. Allows for continued access to the Turbine Building Steam Tunnel without exceeding the maximum safe operating temperature level.

55

NRC 2017 Exam QUESTION RO 28 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295038 EK1.02 Importance Rating 4.2 K&A: Knowledge of the operational implications of the following concepts as they apply to High Off-Site Release Rate: Protection of the general public High Off-site Release Rate Explanation: Answer B - Per the EOP-5 Bases, HB Ventilation is restarted to ensure that radioactivity released in the turbine building is discharged through a monitored release point.

A - Incorrect - Heater Bay ventilation does not contain a charcoal filter. Plausible since the ventilation systems on EOP-3 contain charcoal filters.

C - Incorrect - The Max Safe Operating Conditions are monitored in EOP-3, not EOP-5. Plausible since both the Steam Tunnel and the Heater Bay are connected to the turbine building, but the ventilation systems are not connected. Plausible since the HB Exhaust takes a suction on the TB & HB.

D - Incorrect - The Max Safe Operating Conditions are monitored in EOP-3, not EOP-5. Plausible since both the Steam Tunnel and the Heater Bay are connected to the turbine building, but the ventilation systems are not connected. Plausible since the HB Exhaust takes a suction on the TB & HB.

Technical Reference(s): EOP-3 Bases (EOP-5) Rev 5 Reference Attached: EOP-3 Bases (EOP-5) p 72 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-15-C Question Source: Bank # RQL-41692 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 56

NRC 2017 Exam QUESTION RO 29 A fire has been detected in the Lube Oil Storage/Purifier Room.

In order to automatically initiate the installed CO 2 fire suppression system, how many detectors must activate and how long is the discharge delayed for?

One detector in (1) must activate.

The discharge is delayed by (2) seconds.

(1) (2)

A. 2 zones 60 B. 2 zones 20 C. 1 zone 60 D. 1 zone 20 57

NRC 2017 Exam QUESTION RO 29 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 600000 AK2.01 Importance Rating 2.6 K&A: Knowledge of the interrelations between Plant Fire On Site and the following: Sensors

/ detectors and valves Plant Fire On Site Explanation: Answer A - The MTLO CO 2 system requires 1 detector in each zone and is delayed by 60 seconds to allow personnel to evacuate the area.

B - Incorrect - Plausible since Halon system discharge is delayed by 20 seconds.

C - Incorrect - Requires 2 detectors to go into alarm. Plausible since the first detector causes a panel alarm.

D - Incorrect - Plausible since Halon system discharge is delayed by 20 seconds. And, requires 2 detectors to go into alarm. Plausible since the first detector causes a panel alarm.

Technical Reference(s): SOI-P54(GAS) Rev 8 Reference Attached: SOI-P54(GAS) p 16 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P54(CO 2 )-F Question Source: Bank #

Modified Bank # Perry 2015 # RO-74 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 58

NRC 2017 Exam QUESTION RO 30 The plant is operating at rated power.

Then, SCC informs the control room that a Degraded Grid Condition exists.

Currently, PY-EL LINE, MAIN XFMR, & BUS 1 indicate 345kv and stable.

This condition will require entry into (1) .

In order to minimize the consequences of a Degraded Voltage condition, the Perry Control Room should (2) .

(1) (2)

A. ONI-P56-4, Grid Threat maintain plant parameters stable B. ONI-P56-4, Grid Threat commence a normal plant shutdown C. ONI-S11, Hi/Low Voltage maintain plant parameters stable D. ONI-S11, Hi/Low Voltage commence a normal plant shutdown 59

NRC 2017 Exam QUESTION RO 30 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 700000 AK3.02 Importance Rating 3.6 K&A: Knowledge of the reasons for the following responses as they apply to Generator Voltage And Electric Grid Disturbances: Actions contained in abnormal operating procedure for voltage and grid disturbances Generator Voltage and Electric Grid Disturbances Explanation: Answer C - Notification by SCC that a Degraded Grid Condition exists is an entry condition for ONI-S11. If a degraded grid condition exists, but voltage is stable, shutting down the plant would likely result in a degraded voltage condition. Therefore, maintaining the plant on line with parameters as stable as possible is the desired course of action A - Incorrect - Plausible since the name of the ONI implies this could be correct.

B - Incorrect - Plausible since the name of the ONI implies this could be correct. However, a plant shutdown could result in a degraded voltage condition. 2nd part plausible since shutting down the plant usually puts the plant in a safer condition.

D - Incorrect - A plant shutdown could result in a degraded voltage condition. . 2nd part plausible since shutting down the plant usually puts the plant in a safer condition.

Technical Reference(s): ONI-S11 Rev 10, Reference Attached: ONI-S11 p 3 & 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-18(LP)-A.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 60

NRC 2017 Exam QUESTION RO 31 With the plant operating at rated power, which of the following alarms would result in a degradation of main condenser vacuum?

A. BYPASS VLV SHUT OG POST-TREAT PRCS RAD A/B HI B. OG ISOL OG POST-TREAT PRCS RAD MON A/B 3XHI C. MAIN STEAM LINE RADIATION HI HI/INOP D. OG PRE-TREAT PRCS RAD MON RAD HIGH 61

NRC 2017 Exam QUESTION RO 31 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295002 AK3.06 Importance Rating 2.9 K&A: Knowledge of the reasons for the following responses as they apply to Loss Of Main Condenser Vacuum: Air ejector flow Loss of Main Condenser Vac Explanation: Answer B - Flow from the SJAEs is directed to the Offgas system. Receipt of a OG Post-Treat Prcs Rad Mon A/B 3XHI alarm will isolate Offgas. When Offgas is isolated, flow from the SJAEs is isolated and air and non-condensable gases will buildup in the main condenser causing a loss of main condenser vacuum.

A - Incorrect - Plausible since this alarm is an early indication of a potential fuel problem and this causes the absorber bypass valve to shut.

C - Incorrect - Plausible since this alarm is an early indication of a potential fuel problem and a MSL rad high will cause a trip of the hoggers if running.

D - Incorrect - Plausible since this alarm is an early indication of a potential fuel problem. However, no automatic isolations occur.

Technical Reference(s): ARI-H13-P604-01 Rev 6, ARI- Reference Attached: ARI-H13-P604-01 pp 3, 7, &

H13-P601-19 Rev 19, SDM-N64 Rev 0 13, ARI-H13-P601-19 p 25, SDM-N64 pp 5 & 41 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D17A-I.1 Question Source: Bank # Grand Gulf 2014 # RO-21 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 62

NRC 2017 Exam QUESTION RO 32 The plant is at rated power with the following conditions:

• SVI-M51-T2003A Combustible Gas Mixing System A Operability Test is in progress with the A CGMC running

• The Containment Vessel And Drywell Purge System is running in Intermittent Mode

• Backup Drywell purge valves M51-F090 COMB GAS DW PURGE INBD ISOL and M51-F110 COMB GAS DW PURGE OTBD ISOL are open Then the following alarms are received simultaneously on H13-P680:

• DW PRESS HI/LO

• AIRBORNE RAD P804 The BOP operator reports Drywell Gas Rad Monitor, D17-K676 has a HIGH alarm flashing What automatic actions should have occurred when the above alarms were received?

A. Comb Gas Mix Sys A DW Isol Valve M51-F010A will close and the Combustible Gas Mixing Compressor will trip B. All DW RAD MON INBD & OTBD SUCT & DISCH ISOL valves (D17-F071A/B and D17-F079A/B) will close C. The Containment Vessel And Drywell Purge System dampers will close and fans will trip D. The Backup Drywell purge valves M51-F090 & M51-F110 will close 63

NRC 2017 Exam QUESTION RO 32 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295010 AA1.05 Importance Rating 3.1 K&A: Ability to operate and/or monitor the following as they apply to High Drywell Pressure:

Drywell/suppression vent and purge High Drywell Pressure Explanation: Answer D - A HIGH alarm on the DW Rad Monitor Gas channel will cause the B/U DW Purge valves to isolate.

A - Incorrect - Plausible since a BOP LOCA occurs at 1.68 psig in DW and will cause the M51-F010A to close and the compressor to trip. However, the DW PRESS HI/LO alarm comes in at 1.5 psig but can be confused with the RPS DW Press HI alarm at 1.68 psig.

B - Incorrect - Plausible since a BOP LOCA occurs at 1.68 psig in DW and will cause the rad monitor DW isolation valves to close. However, the DW PRESS HI/LO alarm comes in at 1.5 psig but can be confused with the RPS DW Press HI alarm at 1.68 psig.

C - Incorrect - Plausible since a BOP LOCA occurs at 1.68 psig in DW and will cause the CV&DWP dampers to close and the fans to trip. However, the DW PRESS HI/LO alarm comes in at 1.5 psig but can be confused with the RPS DW Press HI alarm at 1.68 psig.

Technical Reference(s): ARI-H13-P680-05 Rev 15 & ARI- Reference Attached: ARI-H13-P680-05 p 55 &

H13-P680-07 Rev 26 ARI-H13-P680-07 pp 12 & 13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M51_M56-1.7 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 64

NRC 2017 Exam QUESTION RO 33 The plant was operating at rated power with the following conditions:

• An SRV inadvertently opened

• All appropriate ONI and ARI actions were completed

• SUPR POOL TEMP A/B HIGH annunciators in alarm on H13-P601

• RHR loop A was placed in Suppression Pool Cooling

• SPDS is not available How would you monitor suppression pool temperature to evaluate the effectiveness of Suppression Pool Cooling?

Use the Post Accident Monitoring System (PAMS) ____.

A. Recorders on H13-P883 since they display the average Suppression Pool Temperature B. Meters on H13-P601 since they display the average Suppression Pool Temperature C. Meters on H13-P601 since they automatically display the highest Suppression Pool Temperature D. Recorders on H13-P883 or PAMS Meters H13-P601 since they both automatically display the highest Suppression Pool Temperature 65

NRC 2017 Exam QUESTION RO 33 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295013 AA1.01 Importance Rating 3.9 K&A: Ability to operate and/or monitor the following as they apply to High Suppression Pool Temperature: Suppression pool cooling High Suppression Pool Temp.

Explanation: Answer A - The PAMS recorders display eight individual SP temperatures on channels 1-8 and the SP Average temperature on channel 9.

B - Incorrect - The PAMS meters on P601 display is controlled by selector switches manually positioned and located on P883 and can only display the selected channel. It cannot select channel 9.

C - Incorrect - The PAMS meters on P601 display is controlled by selector switches manually positioned and located on P883 and can only display the selected channel. If will not automatically switch to the highest channel.

D - Incorrect - The PAMS meter on P601 does not automatically display the highest SP temperature.

Technical Reference(s): ARI-H13-P601-17 Rev 15, SDM- Reference Attached: ARI-H13-P601-17 p 31, SDM-D23 Rev 3, SVI-D23-T1213 Rev 8 D23 pp 4, 9 & 17, SVI-D23-T1213 pp 10-11, Pic of P883(part)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D23-F Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 66

NRC 2017 Exam QUESTION RO 34 The plant was operating at rated power when an automatic scram occurred.

The SCRAM VALVES pushbutton on P680 is backlit red.

Upon depressing the SCRAM VALVES pushbutton all control rods have green LEDs illuminated on the full core display except for control rod 30-19.

Based on the above information, what is the correct status of the scram valves?

A. Control rod 30-19 is the only rod that has both scram valves open.

B. Control rod 30-19 is the only rod that does not have both scram valves open.

C. All scram valves are open since the SCRAM VALVE pushbutton is backlit red.

D. Control rod 30-19 has one scram valve open while all other control rod scram valves are closed.

67

NRC 2017 Exam QUESTION RO 34 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295015 AA2.02 Importance Rating 4.1 K&A: Ability to determine and/or interpret the following as they apply to Incomplete Scram:

Control rod position Incomplete SCRAM Explanation: Answer B - IAW SOI-C11(RC&IS) Section 7.9.5, if SCRAM VALVES is backlit RED, it indicates that not all scram valves are in the same position (i.e. not all open or not all closed). All control rods with both scram valves open will be indicated by the green LED lit on the full core display.

The lack of a green LED with the SCRAM VALVES pushbutton depressed indicates one or both of the scram valves are closed.

A - Incorrect - This is the opposite and is based on the misconception that a lack of lights has both scram valves open.

C - Incorrect - Red backlight means that not all scram valves are in the same position, it does not mean they are open.

D - Incorrect - Green is typically used for valve closed indications, except for scram valves.

Technical Reference(s): SOI-C11(RC&IS) Rev 29 & Reference Attached: SOI-C11(RC&IS) p 35 &

SDM-C11(RC&IS) Rev 9 SDM-C11(RC&IS) p 36 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11-RC&IS-1.16 Question Source: Bank # Perry 2007-1 #RO-34 Modified Bank #

New Question History: Previous NRC Exam Perry 2007-1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 68

NRC 2017 Exam QUESTION RO 35 The plant was operating at 100% power with Annulus Exhaust Gas Treatment System Fan A in operation.

The following conditions are present:

• A steam leak in the annulus

• ALERT and HIGH alarms on ANNULUS EHAUST GAS TREATMENT RADIATION Monitor A

• ALERT alarm on appropriate PLANT VENT GAS Radiation Monitor Based on these conditions, what is the status of the associated (D19) Post Accident Radiation Monitor and what ONI IMMEDIATE ACTIONS is/are required?

The associated Post Accident Radiation Monitor (1) running and require (2) .

(1) (2)

A. is evacuation of the affected area and a reactor scram B. is evacuation of the affected area only C. is not evacuation of the affected area and a reactor scram D. is not evacuation of the affected area only 69

NRC 2017 Exam QUESTION RO 35 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295017 2.2.44 Importance Rating 4.2 K&A: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

High Off-site Release Rate Explanation: Answer D - The associated D19 rad monitor will start upon the receipt of a High alarm on the Plant Vent rad monitor. Also this requires entry into ONI-D17. Only an evacuation of the affected area is required.

A - Incorrect - The D19 rad monitor will not start on an ALERT on the Plant Vent D17. Plausible if operator thinks a HIGH on the AEGT D17 will start the D19. Also, Rx scram only required for a steam leak in the Offgas system, not the annulus B - Incorrect - The D19 rad monitor will not start on an ALERT on the Plant Vent D17. Plausible if operator thinks a HIGH on the AEGT D17 will start the D19.

C - Incorrect - Plausible since a Rx scram is required for a steam leak in the Offgas system.

Technical Reference(s): ONI-D17 Rev 18, ARI-H13- Reference Attached: ONI-D17 pp 3-5, ARI-H13-P680-07 Rev 26 P680-07 pp 11-13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D17-O, OT-3035-17(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 70

NRC 2017 Exam QUESTION RO 36 A transient occurred resulting in entry into EOP-02, Primary Containment Control.

Suppression Pool water level is approaching the SRV Tail Pipe Level Limit (SRVTPLL).

Which of the following actions would improve the margin to the SRVTPLL?

A. Operate RHR in the Suppression Pool Cooling mode.

B. Initiate the Suppression Pool Makeup System.

C. Lower Suppression Pool water level.

D. Raise RPV pressure.

71

NRC 2017 Exam QUESTION RO 36 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295029 EK1.01 Importance Rating 3.4 K&A: Knowledge of the operational implications of the following concepts as they apply to High Suppression Pool Water Level: Containment integrity High Suppression Pool Wtr Lvl Explanation: Answer C - Lowering Suppression Pool water level improves the margin to SRVTPLL, which is necessary to preserve containment integrity.

A - Incorrect - Lowering SP water temperature has no effect on SRVTPLL. Plausible since this action would improve margin to HCL.

B - Incorrect - Dumping SPMU would further raise SP water level and decrease the margin to SRVTPLL.

Plausible since this action would improve margin to HCL.

D - Incorrect - RPV pressure needs to be lowered to improve margin to SRVTPLL. Plausible if operator does not correctly recall relationship between RPV pressure and SRVTPLL Technical Reference(s): EOP-2 Bases Rev 3 & EOP-SPI Reference Attached: EOP-2 Bases p 48 & EOP-Supplement Rev 6 SPI Supplement p 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-04A-G Question Source: Bank # Perry 2003 #RO-29 Modified Bank #

New Question History: Previous NRC Exam Perry 2003 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 72

NRC 2017 Exam QUESTION RO 37 The plant has experienced a LOCA and the following plant conditions exist:

• Reactor Level - 25

• Time Reactor Level below TAF 20 minutes

• Containment Pressure 10 psig

• Containment Hydrogen Concentration 7.5%

• Drywell Hydrogen Concentration 8.5%

You have been directed to energize the Hydrogen Igniters per the Hardcard.

Should the Hydrogen Igniters be energized?

Reference Provided:

A. no, because Containment HDOL is in the UNSAFE region B. yes, because Containment HDOL is in the SAFE region C. yes, because RPV level has been maintained > TAF D. no, because Drywell HDOL has been exceeded 73

NRC 2017 Exam QUESTION RO 37 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 500000 EK2.03 Importance Rating 3.3 K&A: Knowledge of the interrelations between High Containment Hydrogen Concentrations the following: Containment Atmosphere Control System High CTMT Hydrogen Conc.

Explanation: Answer B - When plotted using the HDOL curve, H2 concentration is in the SAFE region.

Therefore the H2 igniters should be started.

D - Incorrect - DW HDOL is a constant 9% and has not been exceeded. This info has been eliminated on the provided reference.

A - Incorrect - Containment HDOL has not been exceeded but must be determined by using the provided reference.

C - Incorrect - Perry License Commitments allow 30 minutes to start H 2 Igniters after going <TAF. However, information given indicates RPV Level < TAF.

Technical Reference(s): EOP-1 Bases Rev 6, EOP- Reference Attached: EOP-1 Bases p 50, EOP-Supplement Rev 6, & OAI-1703 Rev 27 Supplement p 12, & OAI-1703 p 52 Proposed references to be provided to applicants during examination: Modified EOP-SPI Supplement Figure #7 HDOL Learning Objective (As available): OT-COMBINED-M51_M56-1.8 & 1.15 Question Source: Bank # Perry 2009 # RO-69 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 74

NRC 2017 Exam QUESTION RO 38 The plant was operating at rated power with RCIC tagged out for oil replacement.

A transient occurred resulting in the following sequence of events:

• All ECCS pumps auto started

• HPCS pump tripped and cannot be restarted

• LPCS pump tripped and cannot be restarted

• Emergency Depressurization was performed

• RHR A, B, & C are injecting

• RPV level is 10 and rising

• Annunciator RHR B SUCTION PRESSURE LOW alarms RHR B pump discharge flow and discharge pressure are observed to be lower than normal and fluctuating.

Which of the following actions is required?

A. Obtain US concurrence then trip RHR B pump.

B. Notify the Shift Manager then trip RHR B pump.

C. Immediately trip RHR B pump then update the crew.

D. Maintain RHR B pump running as it is needed for adequate core cooling.

75

NRC 2017 Exam QUESTION RO 38 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 203000 2.4.8 Importance Rating 3.8 K&A: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

RHR/LPCI: Injection Mode Explanation: Answer A - EOP-1 and EOP-1A would require injection from all RHR pumps. However, since RHR B shows signs of misoperation in automatic as evidenced by the alarm and fluctuating flow and pressure, the RO can stop the pump only after obtaining permission from the Unit Supervisor.

B - Incorrect - Notification to the SM is not required, only the US C - Incorrect - Immediately tripping an ECCS pump without first obtaining concurrence from the US is only allowed if approved in the EOPs. Non ECCS pumps can be immediately tripped without US concurrence.

D - Incorrect - With RPV level at 10, ACC is achieved, at least for now. IAW PAP-0205, the US is responsible for the decision to override a system that may jeopardize ACC.

Technical Reference(s): ARI-H13-P601-17 Rev 15, ONI- Reference Attached: ARI-H13-P601-17 p 69, ONI-E12-1 Rev 11, PAP-0205 Rev 21, & EOP Bases Rev 6 E12-1 p 5, , PAP-0205 p 12, & EOP Bases p 28 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-02-G Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 76

NRC 2017 Exam QUESTION RO 39 The plant is shutdown for a refueling outage.

RHR A & B are operating in Refuel Mode of shutdown cooling.

Then RPS Bus B loses power.

RPS A Bus remains energized.

What is the consequence, if any, of losing RPS Bus B?

A. Both RHR A & B pumps trip B. RHR A & B pumps continue to run C. Only E12-F037B UPPER POOL COOLING ISOL closes.

D. Only the RHR B pump trips and E12- F037B UPPER POOL COOLING ISOL closes.

77

NRC 2017 Exam QUESTION RO 39 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 205000 K1.06 Importance Rating 3.2 K&A: Knowledge of the physical connections and/or cause-effect relationships between Shutdown Cooling System (RHR Shutdown Cooling Mode) and the following: A.C. electrical power Shutdown Cooling Explanation: Answer A - RPS logic is arranged such that a loss of either RPS bus will result in loss of both loops of SDC based on Rx High Pressure (135 psig). This will cause INBD/OTBD SDC Suction isolation valves to shut and both RHR pumps to trip on a loss of suction path. This also causes the F037 valves to isolate, B - Incorrect - Plausible if operator fails to recall that RPS powers the isolation logic for SDC.

C - Incorrect - All of the RHR SDC valves will isolate on a loss of RPS.

D - Incorrect - All of the RHR SDC valves will isolate and both pumps trip on a loss of either RPS bus.

Technical Reference(s): SDM-E12 Rev 3, SDM-B21(NS4) Reference Attached: SDM-E12 pp 50-51, SDM-Rev 7, & ONI-C71-2 Rev 9 B21(NS4) p 58, & ONI-C71-2 pp 9 & 12 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21(NS4)-F.4 OT-3035-03(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 78

NRC 2017 Exam QUESTION RO 40 Low Pressure Core Spray (LPCS) is running in Test Mode at 3000 gpm.

What would result if a loss of Bus ED-1-A occurs?

Low Pressure Core Spray _____.

A. can be manually aligned to inject from the Control Room B. Pump will trip and be unavailable for operation C. will automatically realign on a LOCA signal D. will continue to operate in Test Mode 79

NRC 2017 Exam QUESTION RO 40 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209001 K2.03 Importance Rating 2.9 K&A: Knowledge of electrical power supplies to the following: Initiation logic LPCS Explanation: Answer D - Loss of initiation logic power causes the system to fail as-is.

A - Incorrect - A loss of ED-1-A prevents opening the injection valve from the control room.

B - Incorrect - ED-1-A supplies control power to the LPCS pump breaker. The breaker cannot trip.

C - Incorrect - A loss of ED-1-A prevents initiation.

Technical Reference(s): PDB-H1 Rev 2 & Dwgs. 208-060 Reference Attached: PDB-H1 pp 18, 52 & 53 &

series Dwgs. 208-060 Sh 4 & Sh 11 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E21-F OT-3035-05(LP)-A.1 Question Source: Bank # Perry 2009 # RO-41 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 80

NRC 2017 Exam QUESTION RO 41 The plant is in Mode 1. The following conditions exist:

• An inadvertent initiation of High Pressure Core Spray (HPCS) occurs.

• However, the HPCS INJECTION VALVE, 1E22-F004 fails to open.

If an ATWS occurs, what can be expected regarding Standby Liquid Control (SLC) and Alternate Boron Injection System (ABI)?

If initiated, SLC (1) inject.

If required, the ABI System will be (2) for injection.

(1) (2)

A. will available B. will not available C. will unavailable D. will not unavailable 81

NRC 2017 Exam QUESTION RO 41 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209002 K3.02 Importance Rating 3.3 K&A: Knowledge of the effect that a loss or malfunction of the High Pressure Core Spray System (HPCS) will have on following: Standby liquid control system HPCS Explanation: Answer C - SLC will inject as it connects downstream of the manual shutoff valve, E22-F036 and ABI is not available as it connects upstream of the injection valve E22-F004.

A - Incorrect - ABI will not be available as it connects upstream of the Injection Valve.

B - Incorrect - SLC connects in downstream of the Injection Check Valve, so it will inject and ABI will not be available as it connects upstream of the Injection Valve.

D - Incorrect - SLC connects in downstream of the Injection Valve, so it will inject.

Technical Reference(s): Dwg 302-701 Rev KK Reference Attached: Dwg 302-701 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C41-B.1 & C41-L-1.5 Question Source: Bank #

Modified Bank # Perry 2009 # RO-43 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 82

NRC 2017 Exam QUESTION RO 42 A SLC pump is protected from over-pressurization by a relief valve that discharges to the ____.

A. SLC Test Tank, C41-A002 B. SLC Storage Tank, C41-A001 C. Containment Floor Drain Sump D. pump suction line, C41-C001A(B) 83

NRC 2017 Exam QUESTION RO 42 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 211000 K4.10 Importance Rating 2.8 K&A: Knowledge of Standby Liquid Control System design feature(s) and/or interlocks which provide for the following: Over pressure protection SLC Explanation: Answer D - The SLC relief valve C41-F029 relieves back to the suction of its respective pump.

A - Incorrect - Plausible, as the SLC Test Tank is used for the quarterly SLC surveillance.

B - Incorrect - Plausible since this is the suction source for the SLC pumps.

C - Incorrect - Plausible since dirty water from RWCU are piped to the Containment FDS.

Technical Reference(s): Dwg 302-691 Rev Z & SDM-C41 Reference Attached: Dwg 302-691 & SDM-C41 p 8 Rev 9 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C41-O.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 84

NRC 2017 Exam QUESTION RO 43 A plant startup is in progress per IOI-1, Cold Startup.

• APRM DOWNSCALE lights on panel P680 have just extinguished

• IRM / APRM overlap checks are in progress

• Main Turbine Shell warm-up is in progress

• Turbine first stage pressure is 180 psig and slowly increasing Which scram signal is active based on current plant conditions?

A. RPV Level 8 B. MSIV Closure C. IRM Neutron Flux High D. Turbine Control Valve Fast Closure 85

NRC 2017 Exam QUESTION RO 43 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 212000 K5.02 Importance Rating 3.3 K&A: Knowledge of the operational implications of the following concepts as they apply to Reactor Protection System: Specific logic arrangements RPS Explanation: Answer C - For the conditions given, the plant is in Mode 2 with the Mode Switch in STARTUP/STANDBY. In this condition, IRM Neutron Flux would be the only active scram signal.

A - Incorrect - This signal is bypassed with the Mode Switch not in RUN.

B - Incorrect - This signal is bypassed with the Mode Switch not in RUN.

D - Incorrect - This signal is not active with turbine 1st stage pressure <212 psig (equivalent to ~ 38% RTP) when Stop Valves are closed.

Technical Reference(s): PDB-I05 Rev 10 Reference Attached: PDB-I05 p 1 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C71-1.7 Question Source: Bank # INL-235083 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 86

NRC 2017 Exam QUESTION RO 44 The plant is in Mode 2 with all IRMs on range 3.

What would be the effect if the High Voltage Power Supply input to the IRM D detector went to 60 VDC?

In this plant condition, a High Voltage input of 60 VDC to IRM D detector would cause an IRM D ____.

A. INOP condition only B. Downscale indication only C. Downscale indication and generate a Rod Block signal only D. Downscale indication and INOP condition and generate Rod Block & Half Scram signals 87

NRC 2017 Exam QUESTION RO 44 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215003 K6.04 Importance Rating 3.0 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Intermediate Range Monitor (IRM) System: Detectors IRM Explanation: Answer D - When an IRM detectors output voltage fails low, it will cause an IRM INOP condition in the IRM system. This then outputs to the RPS and RC&IS systems to cause a 1/2 SCRAM and a Rod Block.

A - Incorrect - An INOP Trip will occur, but this is not the only result of low output from the HV power supply.

B - Incorrect - A Downscale will come in, but this is not the only result of low output from the HV power supply.

C - Incorrect - Both conditions are true, but these are not the only results of low output from the HV power supply.

Technical Reference(s): SDM-C51(IRM) Rev 8 & ARI- Reference Attached: SDM-C51(IRM) pp 11, 17, &

H13-P680-06 Rev 9 45 & ARI-H13-P680-06 p 71 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51_IRM-1.9 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 88

NRC 2017 Exam QUESTION RO 45 A plant startup is in progress with all IRM's on ranges 1 and 2.

Refer to the attached picture to answer the following.

Given the status of the SRMs and IRMs indicated on the attached picture, Attachment Provided:

A. 1) Control Rods cannot be withdrawn.

2) A Short Period condition is currently active.

B. 1) Control Rods cannot be withdrawn.

2) A Short Period condition was previously active, but may now be clear.

C. 1) SRM's A & B cannot be withdrawn.

2) A Short Period condition is currently active.

D. 1) SRM's A & B cannot be withdrawn.

2) A Short Period condition was previously active, but may now be clear.

89

NRC 2017 Exam QUESTION RO 45 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215004 A1.06 Importance Rating 3.1 K&A: Ability to predict and/or monitor changes in parameters associated with operating the Source Range Monitor (SRM) System controls including: Lights and alarms Source Range Monitor Explanation: Answer A - The Retract Permit lights off indicates that the associated SRM's are <100 CPS. This will insert a Rod Block and not allow control rod withdrawal. The Period light indicates that a short period condition currently exists.

B - Incorrect - The Period Light indicates that a short period condition currently exists. Plausible since the back-panel Period Light locks in until the RESET switch is turned.

C - Incorrect - SRM's can be withdrawn at any time. However, if IRM's are not on Rang 3 or greater, the Retract Permit will generate a Rob Block signal.

D - Incorrect - SRM's can be withdrawn at any time. The Period Light indicates that a short period condition currently exists. Plausible since the back-panel Period Light locks in until the RESET switch is turned.

Technical Reference(s): SDM-C51(SRM) Rev 8 Reference Attached: SDM-C51(SRM) pp 15,17-18, 25, & 30 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51(SRM)-1.5 & 1.14 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Attach Picture of SRM's on H13-P680 to question 90

NRC 2017 Exam QUESTION RO 46 A plant startup is in progress.

The Reactor Operator is adjusting control rods to control reactor period.

All IRM's are on range 8.

What is the effect if SRM A fails high?

A. Half scram signal is generated on RPS A.

B. Control rod adjustment may continue.

C. Control rod withdrawal is blocked.

D. The reactor Scrams.

91

NRC 2017 Exam QUESTION RO 46 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215004 K1.01 Importance Rating 3.6 K&A: Knowledge of the physical connections and/or cause-effect relationships between Source Range Monitor (SRM) System and the following: Reactor protection system Source Range Monitor Explanation: Answer B - Shorting links for the RPS Scram are installed (normal configuration) preventing a scram and with the IRM's on range 8, the Rod Block is also bypassed.

A - Incorrect - Plausible since this would be true if any IRM were to fail.

C - Incorrect - Plausible if the IRM's were < Range 8 D - Incorrect - Plausible if shorting links are removed as this would be a non-coincident scram.

Technical Reference(s): SDM-C51(SRM) Rev 8 & ARI- Reference Attached: SDM-C51(SRM) p 2 & ARI-H13-P680-06 Rev 9 H13-P680-06 p 29 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51(SRM)-1.6 Question Source: Bank #

Modified Bank # Fermi 2013 # RO-38 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 92

NRC 2017 Exam QUESTION RO 47 A reactor startup is in progress with the REACTOR MODE SWITCH in STARTUP/STANDBY.

The following is the present status of the APRM versus LPRM inputs, and the indicated power.

APRM: A B C D E F G H LPRMs:

D Level Inputs: 4 5 3 4 4 4 6 6 C Level Inputs: 4 3 4 3 6 2 4 4 B Level Inputs: 3 4 4 3 4 4 6 4 A Level Inputs: 3 3 4 3 6 4 1 2 Indicated Power: 11% 10% 11% 11% 10% 10% 11% 10%

What will be the consequences and what is required to mitigate the plant response to the conditions above?

A. Only rod block - Bypass appropriate APRM B. Full Scram - Perform Immediate Actions of ONI-C71 Reactor Scram C. Only half scram - Bypass appropriate APRM and reset the half scram per SOI-C71 RPS Power Supply Distribution D. Rod block and half scram - Bypass appropriate APRM and reset the half scram per SOI-C71 RPS Power Supply Distribution 93

NRC 2017 Exam QUESTION RO 47 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215005 A2.03 Importance Rating 3.6 K&A: Ability to (a) predict the impacts of the following on the Average Power Range Monitor/Local Power Range Monitor System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inoperative Trip (all causes)

APRM / LPRM / OPRM Explanation: Answer D - APRM D will cause 1/2 scram and rod block (<14 LPRM inputs).

A - Incorrect - Plausible since this is partially correct - will also get 1/2 scram.

B - Incorrect - Plausible misconception that < 2 LPRM (APRM G) inputs will cause 1/2 scram - this is an administrative INOP condition not scram signal.

C - Incorrect - Plausible since this is partially correct - will also get a rod block.

Technical Reference(s): ARI-H13-P680-06 Rev 9, ONI- Reference Attached: ARI-H13-P680-06 pp 75-76, C11-1 Rev 16 ONI-C11-1 p 15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51(AP_OPRM)-1.11 Question Source: Bank # Perry 2009 # RO-50 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 94

NRC 2017 Exam QUESTION RO 48 The Reactor was operating at rated power with the following conditions:

• Reactor Recirculation FCVs are at 70% open

• APRM A is in Bypass Then, APRM E ramped to 125%.

No operator actions were taken.

The FCVs are now (1) open .

If APRM E reading returned to normal, the FCVs will (2) position.

(1) (2)

A. 30% remain at the runback B. 30% return to the pre-runback C. 60% remain at the runback D. 60% return to the pre-runback 95

NRC 2017 Exam QUESTION RO 48 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215005 K3.02 Importance Rating 3.5 K&A: Knowledge of the effect that a loss or malfunction of the Average Power Range Monitor/Local Power Range Monitor System will have on following: Reactor recirculation system: BWR-5,6 APRM / LPRM / OPRM Explanation: Answer B - With APRM A bypassed, a failure of APRM E will cause the AFDL circuitry to close the FCVs 40% from the original starting point. The FCVs will return to the original position if no operator action is taken, i.e. locking up FCVs. APRM A & E are on same bypass joystick, so it would not be possible to bypass both at same time.

A & C 2nd part - Incorrect - If the FCV runback was not caused by AFDL, then the FCVs would remain in the current position C & D 1st part - Incorrect - This is correct if the FCVs would have started at 100% open - plausible if the candidate thinks FCVs will close 40% from 100%

Technical Reference(s): SDM-B33 Rev 11 Reference SDM-B33 p 43 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-F Question Source: Bank # Perry 2010 # RO-66 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 96

NRC 2017 Exam QUESTION RO 49 Reactor Core Isolation Cooling automatically initiated.

Several minutes later, the following indication was received on H13-P601.

Which of the following valves will remain open based on this indication?

A. 1E51-F013, RCIC INJECTION VLV B. 1E51-F019, RCIC MIN FLOW VALVE C. 1E51-F045, RCIC STEAM SHUTOFF D. 1E51-F510, RCIC TURBINE TRIP THRT V LATCH 97

NRC 2017 Exam QUESTION RO 49 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 A3.01 Importance Rating 3.5 K&A: Ability to monitor automatic operations of the Reactor Core Isolation Cooling System (RCIC) including: Valve operation RCIC Explanation: Answer C - The light indicates that a Div 2 isolation signal is active. All the valves close on a RCIC isolation except for 1E51-F045. This valve only closes in response to a L8 or if E51-F068 is not open.

A - Incorrect - E51-F013 closes in response to the closing of E51-F510, which gets a close signal on any divisional isolation.

B - Incorrect - E51-F019 closes in response to the discharge pressure decay following the turbine trip.

D - Incorrect - E51-F510 closes in response to any isolation signal.

Technical Reference(s): SDM-E51 Rev 13 & ARI-H13- Reference Attached: SDM-E51 pp 19-20 & 25-26 &

P601-21 Rev 15 ARI-H13-P601-21 p 53 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E51-F.2 Question Source: Bank # INL-0863 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 98

NRC 2017 Exam QUESTION RO 50 Which of the following alarms would allow for Manual Initiation of RCIC?

1 RCIC ISOL DIAPHRAGM RUPTURED, (H13-P601-0021-B1) 2 STEAM TUNNEL LD AMB TEMP P632, (H13-P601-0019-G4) 3 RCIC TURBINE OIL COOLER OUT TEMP HIGH, (H13-P601-0021-C4) 4 RCIC SUPR POOL SUCT VLV OPEN SUPR PL LVL HI, (H13-P601-0021-G5)

A. 1, 2, & 3 B. 2, 3, & 4 C. 1, 3, & 4 D. 1, 2, & 4 99

NRC 2017 Exam QUESTION RO 50 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 K4.06 Importance Rating 3.5 K&A: Knowledge of reactor core isolation cooling system (RCIC) design feature(s) and/or interlocks which provide for the following: Manual initiation RCIC Explanation: Answer B - Only combination that does not contain RCIC ISOL DIAPHRAGM RUPTURED. This annunciator will cause a trip of the RCIC turbine trip throttle valve and a RCIC isolation.

STEAM TUNNEL LD AMB TEMP P632 has a 29 minute time delay. Hi LO temp does not cause an isolation. RCIC can still be initiated with a high SP water level.

A - Incorrect - Contains RCIC ISOL DIAPHRAGM RUPTURED.

C - Incorrect - Contains RCIC ISOL DIAPHRAGM RUPTURED.

D - Incorrect - Contains RCIC ISOL DIAPHRAGM RUPTURED Technical Reference(s): ARI-H13-P601-021 Rev 15, ARI- Reference Attached: ARI-H13-P601-021 pp 19, 39, H13-P601-19 Rev 19, & SOI-E31 Rev 8 & 89, ARI-H13-P601-19 p 111, & SOI-E31 p19 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E51-I Question Source: Bank #

Modified Bank # Perry 2015 # RO-49 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 100

NRC 2017 Exam QUESTION RO 51 The plant is operating at rated power when the following annunciators on H13-P601 alarm:

• LPCS AUTO START RECEIVED

• LPCS & LPCI A RX LEVEL LO L1

• LPCI A AUTO START RECEIVED

• ADS A PERMISSIVE LPCS / RHR A RUN

• ADS A TIME DELAY LOGIC TIMER RUNNING

• ADS A TIMER 90 SEC & RUNNING Refer to the attached SPDS picture for current plant conditions In accordance with ONI-E12-1, INADVERTENT INITIATION OR ECCS/RCIC, which of the following IMMEDIATE ACTIONS is/are required?

Attachment Provided:

A. Only place ADS A LOGIC INHIBIT Keylock Switch in INHIBIT B. Place ADS A and B LOGIC INHIBIT Keylock Switches in INHIBIT C. Depress the ADS A and B LOGIC SEAL IN RESET pushbuttons and only place ADS A LOGIC INHIBIT Keylock Switch in INHIBIT D. Depress the ADS A and B LOGIC SEAL IN RESET pushbuttons and place both ADS A and B LOGIC INHIBIT Keylock Switches in INHIBIT 101

NRC 2017 Exam QUESTION RO 51 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 218000 A4.12 Importance Rating 4.2 K&A: Ability to manually operate and/or monitor in the control room: Reactor vessel water level ADS Explanation: Answer C - With RPV level and pressure normal, this is an inadvertent initiation of ADS. IAW ONI-E12-1, permissives for ADS A are met. Therefore, the ADS A and B Logic Seal In Reset PB are depressed and with the logic met for initiation, A Logic Inhibit Keylock switch must be placed in Inhibit.

A - Incorrect - This is not the only Immediate Action required by ONI-E12-1. The Seal-in Reset pushbuttons must also be depressed.

B - Incorrect - The permissives for ADS B are not satisfied - ONI says inhibit only the affected logic channel.

D - Incorrect - ONI-E12-1 directs operator to inhibit only the channel associated with the inadvertent initiation.

Technical Reference(s): ONI-E12-1 Rev 11, ARI-H13- Reference Attached: ONI-E12-1 p 5-6, ARI-H13-P601-19 Rev 19 P601-19 pp 71 & 103 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21C-F, I.1, & J.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Attach SPDS 9-Trend picture to question 102

NRC 2017 Exam QUESTION RO 52 The plant is in power ascension following a refuel outage.

Engineering reports that incorrect spring tensions were set on multiple Containment Isolation MOVs that may render them unable to close under accident conditions.

Which of the following pairs of valves, if inoperable, would require Tech Spec Actions to isolate the penetrations within one hour?

A. P50-F140, CVCW INBD RETURN MOV ISOL VALVE and P50-F150, CVCW OTBD RETURN MOV ISOL VALVE B. P43-F055, NCC CNTMT SUPPLY OTBD ISOL and P43-F215, NCC CNTMT RETURN INBD ISOL C. E12-F064C, RHR PUMP C MIN FLOW and E12-F105, RHR C SUPP POOL SUCTION VALVE D. P52-F200, IA CNTMT ISOL VLV and P52-F646, INST AIR DRYWELL SHUTOFF 103

NRC 2017 Exam QUESTION RO 52 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 2.2.40 Importance Rating 3.4 K&A: Ability to apply Technical Specifications for a system.

PCIS/Nuclear Steam Supply Shutoff Explanation: Answer A - IAW TS 3.6.1.3, Condition B states, One or more penetration flow paths with two PCIVs inoperable except due to leakage, requires isolation of the penetration flow path within I hour. P50-F140 and P50-F150 are INBD and OTBD valves on the same penetration.

B - Incorrect - Plausible since one valve is INBD and one is OTBD, but not on the penetration. One is supply and one is return.

C - Incorrect - Plausible since both valves are on RHR C and do not have an INBD isolation. However, both valve lines terminate below suppression pool surface.

D - Incorrect - Plausible since one valve is OTBD of containment and one is INBD of containment. However, P52-F646 is a DW isolation valve. Additionally, the action to isolate a Drywell valve is 8-hours.

Technical Reference(s): TS 3.6.1.3 PDB-G01 Rev 4, Reference Attached: TS 3.6.1.3 pp 3.6-9 & 11, Dwgs 913-08 Rev S, & 302-244 Rev M PDB-G01 pp 6 & 9, Dwgs 913-08, & 302-244 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21(NS4)-K.1 OT-3037-10-A Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 104

NRC 2017 Exam QUESTION RO 53 The plant is operating in EOP-1A, Level Power Control, with the following conditions:

• RHR A pump is operating in Suppression Pool Cooling

• RHR B & C pumps have tripped and cannot be restarted

• The US determined Emergency Depressurization is required Based on this information, loss of which Bus would prevent performing Emergency Depressurization from H13-P601?

A. ED-1-A B. ED-1-B C. EV-1-A D. EV-1-B 105

NRC 2017 Exam QUESTION RO 53 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 239002 K2.01 Importance Rating 2.8 K&A: Knowledge of electrical power supplies to the following: SRV solenoids SRVs Explanation: Answer A - Bus ED-1-A powers the SRV A solenoids and the ADS A logic. With neither RHR B nor C pumps running, the ADS B logic is not satisfied. Therefore, ED must be performed from the back panel H13-P631 to energize the ADS B solenoids.

B - Incorrect - Plausible since the ADS B logic is still energized, but no RHR pumps are running to satisfy the logic.

C - Incorrect - Plausible since this is a Division 1 Vital power supply.

D - Incorrect - Plausible if power supplies to the SRV solenoids is recalled incorrectly.

Technical Reference(s): ELI-R42 Rev 8, Dwgs 208-011 Reference Attached: ELI-R42 pp 3-4, Dwgs 208-Sh 4 Rev M. Sheet 5 Rev J 011 Sheets 4 & 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21C-C OT-3035-05(LP)-A.1 Question Source: Bank #

Modified Bank # Perry 2013 # RO-13 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 106

NRC 2017 Exam QUESTION RO 54 The plant is at 25% power following a refueling outage with the following conditions:

• RFPT A is in AUTO on DFWCS

• RFPT B is at 1100 RPM

• Main Generator output is 300 MW What would be the consequence if Level transmitters 1C34-N004A, RX LEVEL A and 1C34-N004C, RX LEVEL C failed high?

A. Only Main Turbine trip B. Only Main Turbine and RFPT A trip C. Only Main Turbine and both RFPTs trip D. Main Turbine and both RFPTs trip and Rx Scram on Level 8 107

NRC 2017 Exam QUESTION RO 54 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 259002 K3.06 Importance Rating 2.8 K&A: Knowledge of the effect that a loss or malfunction of the Reactor Water Level Control System will have on following: Main turbine Reactor Water Level Control Explanation: Answer C - If 2 out of 3 of the C34 RPV level instruments fail high, the trip logic for the main turbine and Feedwater pumps is satisfied.

A - Incorrect - Both RFPTs also trip.

B - Incorrect - RFPT B also trips.

D - Incorrect - The RPV Level 8 comes from different transmitters. The Rx will scram, but on Level 3.

Technical Reference(s): ARI-H13-P680-03 Rev 15 & Reference Attached: ARI-H13-P680-03 p 23 &

PDB-I05 Rev 10 PDB-I05 p 51 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C34-1.14 OT-3035-04(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 108

NRC 2017 Exam QUESTION RO 55 Given the following Annulus Gas Exhaust Treatment System, components:

• HEPA Filter

• Roughing Filter

• Charcoal Adsorber Which component(s) is/are designed to remove:

1) Particulate

2) Iodine A. 1) Roughing Filter only

2) HEPA Filter and Charcoal Adsorber only B. 1) HEPA Filter only

2) Charcoal Adsorber only C. 1) Roughing and HEPA Filters

2) HEPA Filter and Charcoal Adsorber D. 1) Roughing and HEPA Filters

2) Charcoal Adsorber only 109

NRC 2017 Exam QUESTION RO 55 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 261000 K4.04 Importance Rating 2.7 K&A: Knowledge of Standby Gas Treatment System design feature(s) and/or interlocks which provide for the following: Radioactive particulate filtration SGTS (AEGTS)

Explanation: Answer D - The Roughing Filter, and the HEPA Filter both remove particulate. The Charcoal Adsorber removes iodine.

A - Incorrect - The HEPA filter does not remove iodine and the roughing filter is not the only component that removes particulate.

B - Incorrect - The HEPA filter is not the only component that removes particulate.

C - Incorrect - The HEPA Filter does not remove iodine.

Technical Reference(s): SDM-M15 Rev 7 Reference Attached: SDM-M15 pp 1, 3-5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M15-C Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 110

NRC 2017 Exam QUESTION RO 56 The plant is at rated power with the Auxiliary Transfer Switch on H13-P870 in the following position:

What is the operational implication of this switch configuration?

A. Allows normal and alternate supply breakers H1201 and H1202 to be closed and remain closed simultaneously.

B. Allows normal and alternate supply breakers L1006 and L1102 to be closed and remain closed simultaneously.

C. Prevents normal and alternate supply breakers H1101 and H1102 from being closed and remaining closed simultaneously.

D. Prevents normal and alternate supply breakers L1003 and L1004 from being closed and remaining closed simultaneously.

111

NRC 2017 Exam QUESTION RO 56 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262001 K5.02 Importance Rating 2.6 K&A: Knowledge of the operational implications of the following concepts as they apply to A.C. Electrical Distribution: Breaker control AC Electrical Distribution Explanation: Answer B - The Aux Transfer Switch (ATS) in the OFF position allows breakers L1006 and L1102 to be closed simultaneously. This is done for certain operations such as backfeeding an L Bus.

A - Incorrect - The ATS only affects breaker control for the L-Buses. Common misconception that the ATS affects the H-Bus breaker controls.

C - Incorrect - The ATS only affects breaker control for the L-Buses. Common misconception that the ATS affects the H-Bus breaker controls.

D - Incorrect - If he ATS was in the AUTO position, this would be true.

Technical Reference(s): SDM-R10 Rev 12 Reference Attached: SDM-R10 pp 23-24 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R10 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 112

NRC 2017 Exam QUESTION RO 57 A transient occurred and all ECCS systems started automatically.

Then a loss of off-site power occurred.

Which of the following describes the start sequence for the LPCS and RHR Pumps to prevent bus overloading?

Upon closing the respective diesel generator output breaker, ____.

A. LPCS and RHR C start immediately, RHR A and B start after a 5 second time delay B. LPCS and RHR B start immediately, RHR A and C start after a 5 second time delay C. RHR A and C start immediately and LPCS and RHR B start after a 5 second time delay.

D. LPCS and RHR C start immediately and RHR A and B starts after a 10 second time delay 113

NRC 2017 Exam QUESTION RO 57 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262001 A1.02 Importance Rating 3.1 K&A: Ability to predict and/or monitor changes in parameters associated with operating the A.C. Electrical Distribution controls including: Effects of loads when energizing a bus AC Electrical Distribution Explanation: Answer A - When a LOOP occurs, if an ECCS initiation signal is present when power is restored to the EH Buses, the RHR C pump will start immediately and LPCS pump breaker remains closed (no UV trip) so it starts immediately. RHR A & B start after a 5 second time delay.

B - Incorrect - RHR B starts after a 5 second time delay and RHR C starts immediately after the bus is re-energized C - Incorrect - RHR A starts after a 5 second time delay and LPCS starts immediately.

D - Incorrect - RHR A & B start after a 5 second time delay - not 10.

Technical Reference(s): SDM-E12 Rev 3 & SDM-E21 Reference Attached: SDM-E12 p 37-38 & SDM-Rev 1 E21 p 25 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R43_48-N.1 Question Source: Bank #

Modified Bank # Perry 2015 # RO-27 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 114

NRC 2017 Exam QUESTION RO 58 QUESTION DELETED - It was identified during post-exam review that the question was inaccurate and incomplete technical information, and not enough information is provided in the stem to obtain a correct answer.

Annunciator INVERTER DB-1-A TROUBLE just alarmed on H13-P870.

A(An) (1) condition will cause an Auto Transfer to the Alternate Source.

If the condition causing the alarm clears in 10 minutes, the loads will (2) back to the inverter.

(1) (2)

A. DC ground fault automatically transfer B. DC ground fault need to be manually transferred C. inverter LOW output voltage automatically transfer D. inverter LOW output voltage need to be manually transferred 115

NRC 2017 Exam QUESTION RO 58 QUESTION DELETED Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262002 K6.01 Importance Rating 2.7 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Uninterruptable Power Supply (A.C./D.C.) : A.C. electrical power UPS (AC/DC)

Explanation: Answer C - Low AC output is annunciated for the BOP inverter, not the ATWS inverter. If the alarm condition clears, the Static Transfer will automatically transfer the load back to the inverter. One condition causes the Static Transfer Switch to latch after 15 minutes.

A - Incorrect - Plausible since DC GROUND FAULT is an alarm condition for the TSC-UPS inverter.

B - Incorrect - Plausible since DC GROUND FAULT is an alarm condition for the TSC-UPS inverter. For the BOP inverter, if the alarm condition clears, the Static Transfer will automatically transfer the load back to the inverter.

D - Incorrect - If the alarm condition clears, the Static Transfer will automatically transfer the load back to the inverter.

Technical Reference(s): ARI-H13-P870-01 Rev 15 & Reference Attached: ARI-H13-P870-01 p 39, &

SDM-R14_15 Rev 2 SDM-R14_15 pp 7-8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R14_15-10 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 116

NRC 2017 Exam QUESTION RO 59 Annunciator DIV 2 BATTERY DC SYSTEM TROUBLE alarmed on H13-P877.

Use the attached picture of observed readings on H13-P877.

An NLO Reported the following indications from EFD-1-B 125VDC Battery Charger:

• Charger DC Voltage is 123 VDC

• Charger DC Current is 400 Amps

• FLOAT/EQUALIZE switch mis-positioned to EQUALIZE

• Red DC VOLTS LOW light is lit

• White AC ON light is lit With no operator action, which of the following describes the expected Bus ED-1-B voltage trend and the reason for that trend?

Bus ED-1-B voltage will ____.

Attachment Provided:

A. lower because the float voltage is low out of band B. rise because an equalizing charge is being provided C. rise because the bus load is less than the charger capacity D. lower because the bus load is greater than the charger capacity 117

NRC 2017 Exam QUESTION RO 59 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 263000 A1.01 Importance Rating 2.5 K&A: Ability to predict and/or monitor changes in parameters associated with operating the D.C. Electrical Distribution controls including: Battery charging/discharging rate DC Electrical Distribution Explanation: Answer D - The charger capacity is 400 amps but the ED1-1B bus current is indicated at ~50 amps as shown by the ammeter on P870 in the DISCHARGE region. The total current draw is

~ 450 amps. This will cause ED-1-B voltage to lower.

A - Incorrect - Voltage is low, but with the battery in EQUALIZE the voltage should be higher.

B - Incorrect - This would be true if load current did not exceed charger capacity.

C - Incorrect - This is the opposite, but plausible if the meter indications are misunderstood.

Technical Reference(s): ARI-H13-P877-02 Rev 13, SOI- Reference Attached: ARI-H13-P877-02 p 79, SOI-R42 (Div 2) Rev 11, SDM-R42 Rev 10 R42 (Div 2) p 47, SDM-R42 pp 7-8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R42-33 & 34 Question Source: Bank # River Bend 2003 # RO-65 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Attach Panel H13-P877 Meters to question 118

NRC 2017 Exam QUESTION RO 60 Div 1 Diesel Generator is running in parallel with Bus EH11 for the monthly surveillance test.

Grid conditions are stable.

Then, annunciator DIESEL GENERATOR OUT OF SERVICE H13-P877 alarms.

The NLO in the Div. 1 DG room reports the following:

• Local annunciator 125V DC TROUBLE is in alarm.

• DC Control Power to H51-P054A has been lost.

The ability to adjust Div 1 Diesel Generator voltage and load (1) be impacted.

If necessary, shutdown of the Div 1 Diesel Generator can be performed (2) .

(1) (2)

A. will locally by pushing the DIESEL Push To STOP/Pull To RUN valve B. will not locally by pushing the DIESEL Push To STOP/Pull To RUN valve C. will from P877 by taking DIESEL GENERATOR control switch to PULL-TO-LOCK D. will not from P877 by taking DIESEL GENERATOR control switch to PULL-TO-LOCK 119

NRC 2017 Exam QUESTION RO 60 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 264000 A2.01 Importance Rating 3.5 K&A: Ability to (a) predict the impacts of the following on the Emergency Generators (Diesel/Jet) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Parallel operation of emergency generator EDGs Explanation: Answer A B - With a loss of DC Control Power, the DG speed governor and voltage regulator will cannot be adjusted; therefore, DG load will remain constant unless grid frequency changes, and reactive load will not change unless grid voltage changes. With the DG in parallel with the grid (Bus EH11), DG speed and voltage will remain unchanged upon a loss of DC Control Power. Also, shutdown from P877 is not available. The only method to S/D the DG with a loss of control power is to press the Run/Stop valve. (This was recent OE at Perry)

B A - Incorrect - With a loss of DC Control Power, the DG governor and voltage regulator cannot be adjusted. Speed and voltage will not increase/decrease with grid conditions stable.

C - Incorrect - With a loss of DC Control Power, the DG governor and voltage regulator cannot be adjusted. Speed and voltage will not increase/decrease with grid conditions stable. Normally, DG S/D from P877 is the quickest method. However, shutdown from P877 is not available with a loss of control power.

D - Incorrect - Normally, DG S/D from P877 is the quickest method. However, shutdown from P877 is not available with a loss of control power.

Technical Reference(s): ARI-H13-P877-01 Rev 12, ARI- Reference Attached: ARI-H13-P877-01 pp 13-14, H51-P054A Rev 15, NOBP-OP-1002 Rev 2, SOI-R43 ARI-H51-P054A pp 59-60, NOBP-OP-1002 p 19, Rev 45, & ONI-R42-1 Rev 7 SOI-R43 pp 43-44, & ONI-R42-1 pp 7-8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R43_48-F.7 & OT-3039-01-L Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x It was identified, by an exam proctor during the exam administration, that the answer key had not been revised when the question was changed due to reviewer comments. This change was made prior to grading the exam.

120

NRC 2017 Exam QUESTION RO 61 A Non-Licensed Operator reports that the refrigeration unit for in-service Instrument Air (IA) Dryer 1P52-D003A is not operating.

If this condition is left uncorrected, (1) will be introduced into the Instrument Air system.

To mitigate this condition, (2) .

(1) (2)

A. water droplets shift from malfunctioning IA Dryer 1P52-D003A to the standby IA Dryer 1P52-D003B B. water droplets open the IA Desiccant Air Dryer Bypass Valve to bypass malfunctioning IA Dryer 1P52-D003A C. desiccant particles shift from malfunctioning IA Dryer 1P52-D003A to the standby IA Dryer 1P52-D003B D. desiccant particles open the IA Desiccant Air Dryer Bypass Valve to bypass malfunctioning IA Dryer 1P52-D003A 121

NRC 2017 Exam QUESTION RO 61 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 300000 A2.01 Importance Rating 2.9 K&A: Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions Instrument Air Explanation: Answer A - Water is the correct contaminant. Shifting dryers would maintain a low dew point.

B - Incorrect - Bypassing the malfunctioning IA Dryer would not correct the problem.

C - Incorrect - Plausible since the dryer beds contain desiccant and the after-filters can remove particles.

However, a failure of the refrigerant unit would cause the IA dew point to go up resulting in the potential introduction of water into the IA System.

D - Incorrect - Plausible since the dryer beds contain desiccant and the after-filters can remove particles.

However, a failure of the refrigerant unit would cause the IA dew point to go up resulting in the potential introduction of water into the IA System. Bypassing the malfunctioning IA Dryer would not correct the problem.

Technical Reference(s): Dwg 302-241 Rev DD & SDM- Reference Attached: Dwg 302-241 & SDM-P51/52 P51/52 Rev 2 p8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P51_52-8 Question Source: Bank # Perry 2009 # RO-62 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 122

NRC 2017 Exam QUESTION RO 62 The plant was operating at 75% rated thermal power with the following conditions:

• Transformer LH-2-C is tagged out for deluge testing.

• Loads supplied by LH-2-C were transferred to the Alternate source.

The following then occurred:

• Transformer LH-2-B experienced a lockout What is the consequence of this electrical transient?

A. Service Water Pump D, P41-C001D, will trip if running B. Control Complex Chiller C, P47-B001C, cannot be started C. Nuclear Closed Cooling Pump C, P43-C001C, cannot be started D. Unit 2 Instrument Air Compressor, 2P52-C001, will trip if running 123

NRC 2017 Exam QUESTION RO 62 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 300000 K2.01 Importance Rating 2.8 K&A: Knowledge of electrical power supplies to the following: Instrument air compressor Instrument Air Explanation: Answer D - A lockout on an LH transformer will cause the buses supplied from the transformer to transfer to the Alternate or Normal supply. U2 IAC is powered from Bus H22 which is normally powered from LH-2-C. But, since LH-2-C is tagged out, H22 is on its Alternate source - LH-2-B.

The U2 IAC will trip if running.

A - Incorrect - Plausible since SWP C would trip.

B - Incorrect - Would be true of Lockout was on LH-2-A.

C - Incorrect - Would be true of Lockout was on LH-2-A.

Technical Reference(s): ARI-2H13-P870-01 Rev 8, ELI- Reference Attached: ARI-2H13-P870-01 p 33, ELI-R22 Rev 9, & Dwg. 256-016 Rev U R22 pp.30-31, & Dwg. 256-016 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P51_52-27 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 124

NRC 2017 Exam QUESTION RO 63 The plant is operating at rated power.

Annunciator SW PUMP DISCH HEADER PRESSURE LOW on H13-P970 alarmed.

Service Water pump discharge header pressure indicates 36 psig.

What action will restore Service Water pump discharge header pressure to normal?

A. Throttle P41-F390, TBCC HX SW TCV BYP in the OPEN direction B. Throttle P41-F390, TBCC HX SW TCV BYP in the CLOSE direction C. Throttle P41-F400, NCC HX SW BYPASS VLV in the OPEN direction D. Throttle P41-F400, NCC HX SW BYPASS VLV in the CLOSE direction 125

NRC 2017 Exam QUESTION RO 63 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 A4.01 Importance Rating 3.1 K&A: Ability to manually operate and / or monitor in the control room: CCW indications and control Component Cooling Water Explanation: Answer D - P41-F400 controls bypass flow around the NCC HXs and controls SW discharge header pressure. Throttling closed on P41-F400 lowers bypass flow around the NCC HXs and raises SW discharge header pressure.

A - Incorrect - This action is required for a malfunction of the TBCC HX TCV valve and will not raise SW header pressure.

B - Incorrect - Plausible since this valve can affect SW flow if open, but this valve is typically closed and is in parallel to the TBCC HX TCV.

C - Incorrect - This is the opposite action that is required.

Technical Reference(s): ARI-H13-P970-01 Rev 23 & Reference Attached: ARI-H13-P970-01 pp 29-30 SDM-P41 Rev 10 SDM-P41 pp 22 & 41 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P41-I.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 126

NRC 2017 Exam QUESTION RO 64 The plant was operating at 75% rated power when a transient occurred.

Following the transient the Reactor Operator plotted Rx power and core flow on the attached Power Flow Map.

Based on this information, what is the required action?

Attachment Provided:

A. Insert Control Rods IAW Pull Sheets until Reactor Power is approximately 45%

B. Insert Cram Rods until Reactor Power is approximately 35%

C. Raise Core Flow with FCV A to >42 Mlbm/hr D. Insert a Manual Reactor Scram 127

NRC 2017 Exam QUESTION RO 64 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201003 2.1.25 Importance Rating 3.9 K&A: Ability to interpret reference materials, such as graphs, curves, tables, etc.

Control Rod and Drive Mechanism Explanation: Answer B - With power and flow plotted inside the Immediate Exit Region, the operator must insert Cram Rods to 35% reactor power IAW ONI-C51 Immediate Actions and FTI-B02.

A - Incorrect - While this action will exit the Immediate Exit Region, it is not the prescribed method in ONI-C51.

C - Incorrect - While this action will exit the Immediate Exit Region, it is not the prescribed method in ONI-C51.

D - Incorrect - This action would be correct if OPRM's were inoperable. However, nothing indicates OPRM's are INOP.

Technical Reference(s): PDB-A06 Rev 15, ONI-C51 Rev Reference Attached: PDB-A06 p 3. ONI-C51 p 6, &

27, & FTI-B02 Rev 15 FTI-B02 p 16-18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-05(LP)-A.10 & OT-3039-01-O Question Source: Bank #

Modified Bank # Quad Cities 2009 # SRO-91 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 128

NRC 2017 Exam QUESTION RO 65 A malfunction with the rod position indication for control rod 30-31 necessitated bypassing the rod in the Rod Action Control System (RACS) per SOI-C11(RCIS), Rod Control and Information System.

How can verification that the correct rod is bypassed be performed from H13-P680?

A. Ensuring the POSITION BYPASS pushbutton is back lit.

B. Ensuring rod 30-31 will not move when selected and given a withdraw command in IND DRIVE MODE.

C. Depressing the POSITION BYPASS pushbutton and observing rod 30-31 has a green LED lit on the full core display.

D. Depressing the SUBST POSITION pushbutton and observing rod 30-31 has a red LED lit on the full core display.

129

NRC 2017 Exam QUESTION RO 65 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201005 K1.05 Importance Rating 3.5 K&A: Knowledge of the physical connections and/or cause-effect relationships between Rod Control And Information System (RCIS) and the following: Rod action control system RCIS Explanation: Answer C - IAW SOI-C11(RCIS), depressing the POSITION BYPASS pushbutton will cause a green LED to illuminate next to any control rod bypassed in RACS.

A - Incorrect - The POSITION BYPASS pushbutton will be back lit for any control rod that is bypassed in RACS.

B - Incorrect - This would be correct if the rod was bypassed in RGDS.

D - Incorrect - Depressing the SUBST POSITION pushbutton will only indicate those rods with substitute date entered, not those that are bypassed.

Technical Reference(s): SOI-C11(RC&IS) Rev 30 Reference Attached: SOI-C11(RC&IS) pp 54 & 57 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11_RC&IS-1.15 Question Source: Bank # Clinton 2009 #RO-56 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 130

NRC 2017 Exam QUESTION RO 66 The following plant conditions exist:

• The reactor is operating at 45% power during a plant startup

• A loss of Bus H11 occurs

• A Reactor scram was inserted

• Setpoint Setdown activated and maintained RPV level stable Which of the following describes the status of the Reactor Recirculation System prior to taking any Reactor Scram Hardcard actions?

A. Reactor Recirculation Pump A is OFF, Reactor Recirculation Pump B is running in SLOW B. Reactor Recirculation Pump A is running in SLOW, Reactor Recirculation Pump B is OFF C. Both Reactor Recirculation Pumps are running in FAST D. Both Reactor Recirculation Pumps are running in SLOW 131

NRC 2017 Exam QUESTION RO 66 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 202001 K2.02 Importance Rating 3.2 K&A: Knowledge of electrical power supplies to the following: MG sets Recirculation Explanation: Answer A - Rx Recirc pump A will trip from Fast to OFF when RPV lowers to L3 due to the loss of power to the MG Set for SLOW speed operation. Setpoint Setdown activates at L3.

B - Incorrect - Recirc pump B will be running in Slow. Plausible if operator incorrectly recalls power supply to MG set.

C - Incorrect - Plausible since pumps are up-shifted ~35%. When RPV level lowers to L3 Recirc pumps will transfer from Fast to Slow.

D - Incorrect - Plausible since both should be running in Slow. But, without power to A MG, Recirc pump A till trip to OFF.

Technical Reference(s): ARI-H13-P870-01 Rev 15, SDM- Reference Attached: ARI-H13-P870-01 p 57, SDM-B33 Rev 11, SDM-C34 Rev 3, & SDM-R10 Rev 12 B33 p 35, SDM-C34 p 28, & SDM-R10 p 69 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-L.2 Question Source: Bank #

Modified Bank # Perry 2002 # RO-78 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 132

NRC 2017 Exam QUESTION RO 67 The plant is operating at rated power.

Temperature Switch 1G33-N008, RWCU Non Regen Heat Exchanger Temp-Hi has failed low.

Which of the following Reactor Water Cleanup (RWCU) valves, if open, could have an adverse effect on Rx water quality?

A. 1G33-F028, RWCU BLWDN HDR INBD ISOL B. 1G33-F039, RWCU RETURN HDR OTBD ISOL C. 1G33-F042, RWCU HX OUTLET THROTTLE D. 1G33-F107, RWCU HX SHELL SIDE BYPASS 133

NRC 2017 Exam QUESTION RO 67 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 204000 K3.01 Importance Rating 3.2 K&A: Knowledge of the effect that a loss or malfunction of the Reactor Water Cleanup System will have on following: Reactor water quality RWCU Explanation: Answer D - Temperature switch G33-N008 provides an isolation signal to G33-F004 to prevent damage to the resin on high demin inlet temperature. If demin inlet temp rises >140°F, the resin can breakdown and release collected contaminants. If G33-F107 fails open, the Regenerative HX is bypassed causing the demin inlet temp to rise.

A - Incorrect - At rated power both the INBD and OTBD isolation valves are closed. Therefore, opening one would not have any effect.

B - Incorrect - This valve is open at power with RWCU in service.

C - Incorrect - This valve is open when RWCU in service. Plausible since it may be confused with G33-F044.

Technical Reference(s): ARI-H13-P680-01 Rev 13, Reference Attached: ARI-H13-P680-01 p 29, Lesson Plan OT-COMBINED-G33_36 Rev 4, & SDM- Lesson Plan OT-COMBINED-G33_36 slide 14, &

G33 Rev 9 SDM-G33 pp 16 & 33 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-G33_36-I.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 134

NRC 2017 Exam QUESTION RO 68 During a DBA LOCA, how many loops of Containment Spray are required to be operating to lower containment pressure and what is/are the design function(s) of Containment Spray?

During a DBA LOCA, (1) of Containment Spray must be operating to lower containment pressure.

Containment Spray is designed to provide (2) .

(1) (2)

A. 1 loop Containment cooling only B. 1 loop Containment cooling and fission product removal C. 2 loops Containment cooling only D. 2 loops Containment cooling and fission product removal 135

NRC 2017 Exam QUESTION RO 68 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 226001: K4.02 Importance Rating 2.8 K&A: Knowledge of RHR/LPCI: Containment Spray System Mode design feature(s) and/or interlocks which provide for the following: Redundancy RHR/LPCI: CTMT Spray Mode Explanation: Answer B - Per the USAR, the design bases of Containment Spray (CS) is to have 2 redundant means to spray into containment to lower containment pressure below design limits.

Additionally, CS provides cooling and scrubbing of fission products.

A - Incorrect - CS provides scrubbing of fission products as well cooling.

C - Incorrect - One loop of CS is sufficient to lower containment pressure below the containment design limit for pressure. And, CS provides scrubbing of fission products as well cooling.

D - Incorrect - One loop of CS is sufficient to lower containment pressure below the containment design limit for pressure.

Technical Reference(s): USAR C-5 Rev 12 & Lesson Reference Attached: USAR pp 5.4-41, 6.5-9-11 &

Plan OT-COMBINED-E12 Rev 4 E12 Lesson Plan p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-Combined-E12-D Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 136

NRC 2017 Exam QUESTION RO 69 A full core offload is in progress The following then occurred:

• An irradiated fuel bundle is being unloaded from the Fuel Handling Building IFTS Upender

• A noticeable decrease in Fuel Pool water level is observed

• The Fuel Handling Building Evacuation Alarm sounded Based on these conditions, in order to minimize unnecessary exposure, immediate evacuation of ____

from the Fuel Handling Building is required, if present.

1. FME Coordinator

2. Site Protection Officer

3. Fuel Handling Building Crane Operator

4. Radiation Protection Technician A. 1 and 3 only B. 1, 2, and 3 only C. 1, 3, and 4 only D. 1, 2, 3, and 4 137

NRC 2017 Exam QUESTION RO 69 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 234000 K5.03 Importance Rating 2.9 K&A: Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Equipment: Water as a shield against radiation Fuel Handling Equipment Explanation: Answer D - Lowering of the fuel pool water level increases radiation exposure to people in the Fuel Handling Building. This requires evacuating unnecessary personnel from the FHB. Per ONI-J11-2 Necessary Personnel are defined as those personnel necessary to place the equipment or fuel in a safe condition. SOI-F11, fuel handling Platform, identifies personnel required for fuel handling in the FHB. At Perry, necessary personnel are the FH Supervisor, the Platform Operator, and the Spotter - all other personnel are to be evacuated.

A - Incorrect - Site Protection officer and RP Tech are NOT necessary personnel.

B - Incorrect - RP Tech is NOT necessary personnel.

C - Incorrect - Site Protection officer is NOT necessary personnel.

Technical Reference(s): ONI-E12-2 Rev 36 and SOI-F11 Reference Attached: ONI-E12-2 p 6 and SOI-F11 p Rev 18 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-14(LP)-A4, OT-3035-11(LP)-A1 Question Source: Bank # Perry 2010 # RO-19 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 138

NRC 2017 Exam QUESTION RO 70 The plant was operating at 30% rated power when the main turbine tripped due to a loss of main condenser vacuum.

Following the turbine trip, what is the status of the Combined Intermediate Valves and the Positive Assist Non-return Check valves?

The Combined Intermediate Valves are (1) .

The Positive Assist Non-return Check valves are (2) .

(1) (2)

A. open open B. shut shut C. open shut D. shut open 139

NRC 2017 Exam QUESTION RO 70 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 239001 K6.08 Importance Rating 3.3 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Main And Reheat Steam System: Main condenser vacuum Main and Reheat Steam Explanation: Answer B - On a main turbine trip, the CIVs close to isolate Reheat steam to the LP turbine preventing an over-speed condition. Also, The PACVs close to prevent an over-speed condition.

A - Incorrect - Both sets of valves close to protect the main turbine. Plausible misconception that these valves remain open since no reactor scram will occur at this power level.

C - Incorrect - The CIVs are also shut.

D - Incorrect - The PACVs are also shut. Plausible misconception that the PACV open similar to the drain valves on a turbine trip.

Technical Reference(s): ONI-N32 Rev 11, SDM- Reference Attached: ONI-N32 pp 3-4, SDM-N31/N11A/39 Rev 2, & SDM-N36/25/26 Rev 9 N31/N11A/39 pp 14a,-15, & SDM-N36/25/26 pp 14-15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-09(LP)-B.1, OT-COMBINED-N31-F, OT-COMBINED-N36_25_26-F.1 Question Source: Bank #

Modified Bank # Hatch 2013 # RO-26 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 140

NRC 2017 Exam QUESTION RO 71 Power ascension is in progress following a refueling outage.

Reactor power is 50%.

It is observed that as Reactor power is increased, the difference between Reactor pressure and Turbine Throttle pressure is becoming larger.

This condition is (1) because the EHC system controls (2) .

(1) (2)

A. not expected Turbine Throttle pressure to maintain it within 30 psig of Reactor pressure B. not expected Reactor pressure to maintain it and Turbine Throttle pressure in a 30 psi regulation band C. expected Turbine Throttle pressure to maintain it in a 30 psi regulation band D. expected Reactor pressure to maintain it in a 30 psi regulation band and the lower Turbine Throttle pressure results from Main Steam line headloss 141

NRC 2017 Exam QUESTION RO 71 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 241000 A1.02 Importance Rating 4.1 K&A: Ability to predict and/or monitor changes in parameters associated with operating the Reactor/Turbine Pressure Regulating System controls including: Reactor power Reactor/Turbine Pressure Regulator Explanation: Answer C - The EHC system regulates Turbine Throttle pressure in a 30 psi band as Rx pressure increases. The Turbine Throttle pressure rises from 940 to 970 psig at a 3.33% steam flow per 1 psig rise as Reactor pressure raises from 940 to 1025 psig. Reactor pressure raises more due to increased differential pressure caused by the MSL pressure drop as steam line flow increases.

A - Incorrect - Plausible if candidate confuses the Turbine Throttle pressure regulation band (940-970 psig) verses Reactor pressure (940-1025 psig) relationship.

B & D - Incorrect - Plausible if candidate confuses the Turbine Throttle pressure regulation band vs. Reactor pressure relationship. EHC senses the pressure averaging manifold pressure to maintain it in a 30 psig regulation band, not Reactor pressure Technical Reference(s): SDM-N32 Rev 6 Reference Attached: SDM-N32 pp 15-17, 22-23, 121 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N32_C85-A.2 Question Source: Bank # Limerick 2012 # RO-61 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 142

NRC 2017 Exam QUESTION RO 72 Plant soft shutdown is in progress with Rx power at 17%.

• Motor Feed Pump in AUTO on DFWCS

• RFPT A shutdown to 1100 RPM in progress and RFPT B shutdown Then an earthquake occurs and the following condition exists:

• An NLO reports the Instrument Air supply line to the Heater Bay is severed Which of the following describes the effect to the Feedwater system and actions required to mitigate?

(1) (2)

A. MFP FCV M/A station transfers to Use MFP to control RPV level 192-200 Manual inches B. DFWCS transfer to 1-Element (1E) Use MFP to control RPV level 192-200 control inches C. MFP FCV M/A station transfers to Lock the Mode Switch in SHUTDOWN Manual D. DFWCS transfer to 1-Element (1E) Lock the Mode Switch in SHUTDOWN control 143

NRC 2017 Exam QUESTION RO 72 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 259001 A2.07 Importance Rating 3.7 K&A: Ability to (a) predict the impacts of the following on the Reactor Feedwater System ;

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reactor water level control system malfunctions Reactor Feedwater Explanation: Answer C - Loss of Instrument Air (IA) to the Heater Bay causes a loss of IA to the MPF FCVs which will initiate the MFP Freeze circuit. This will prevent the MFP FCVs from moving.

Additionally, it causes the MFP Flow controller to shift to manual. It also causes the MFP Recirc valve to fail open causing ~4000 gpm flow to be diverted from the RPV. This will cause RPV level to lower to the L3 scram setpoint since the MFP was at its flow limit. Per ONI-C34, Immediate Actions, a Rx scram is required before hitting L3.

A - Incorrect - The operator not be able to control RPV level since the MFP is at its limit and the recirc valve failed open.

B - Incorrect - This will not cause DFWCS to shift to 1E, nor is the operator able to control RPV level.

D - Incorrect - This will not cause DFWCS to shift to 1E control Technical Reference(s): ARI-H13-P680-03 Rev 15, ONI- Reference Attached: ARI-H13-P680-03 p 105, C34 Rev 9, ONI-P52 Rev 18, SOI-C34 Rev 35 ONI-C34 pp 5-6, ONI-P52 p 31, SOI-C34 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C34-1.7 & 1.13 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 144

NRC 2017 Exam QUESTION RO 73 A plant startup is in progress with Rx power at ~ 1%.

During control rod withdrawal the following annunciators alarm:

• MAIN STEAM LINE RADIATION HIGH

• MAIN STEAM LINE RADIATION HI HI/INOP The main steam line rad monitors indicate as follows:

• D17-K610A 1500 mrem/hr and increasing

• D17-K610B 1550 mrem/hr and increasing

• D17-K610C 10 mrem/hr and Stable

• D17-K610D 1520 mrem/hr and increasing What automatic actions will occur and what manual action will need to be performed?

Automatic action performed (1) .

Manual action required (2) .

(1) (2)

A. 1B33-F020, REACTOR WATER Close 1B33-F019, REACTOR WATER SAMPLE ISOL valve closes SAMPLE ISOL valve B. 1B33-F020, REACTOR WATER Stop the Mechanical Vacuum pump SAMPLE ISOL valve closes C. All MSIVs close Close 1B33-F019, REACTOR WATER SAMPLE ISOL valve D. All MSIVs close Stop the Mechanical Vacuum pump 145

NRC 2017 Exam QUESTION RO 73 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 272000 A2.01 Importance Rating 3.7 K&A: Ability to (d) predict the impacts of the following on the Radiation Monitoring System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Fuel element failure Radiation Monitoring Explanation: Answer A - An increasing MSL radiation can be caused by fuel element failure. When the rad level exceeds the setpoint for each rad monitor, certain automatic actions occur or should occur such as isolation of both INBD and OTBD Rx sample isolation valves and tripping of the hoggers. The actions are based on the combination of channels receiving the high rad signal.

Since rad monitor C did not trip, the OTBD Rx sample isolation valve did not close automatically and must be closed manually per ARI-H13-P601 and NOP-OP-1002.

B - Incorrect - The hogger will trip on either channel A or C receiving a High signal.

C - Incorrect - The MSIVs no longer isolate automatically on high MSL rads.

D - Incorrect - The MSIVs no longer isolate automatically on high MSL rads and the hogger will trip on either channel A or C receiving a High signal.

Technical Reference(s): ARI-H13-P601-19 Rev 19 and Reference Attached: ARI-H13-P601-19 pp 25 & 49 NOP-OP-1002 Rev 11 and NOP-OP-1002 p 62 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D17A-I.1 Question Source: Bank #

Modified Bank # Perry 2007 # RO-25 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 146

NRC 2017 Exam QUESTION RO 74 A Dry Cask Storage campaign is in progress in the Fuel handling Building.

Then annunciator COMMON AIRBORNE P902 on H13-P680 alarms.

All Fuel Handling Building Vent Exhaust D17 Radiation Monitors have HIGH alarms locked in.

Which of the following is the expected Fuel Handling Building Ventilation lineup?

1 2 3 4 FHB SUPP FAN A OFF OFF ON ON FHB SUPP FAN B OFF OFF OFF ON FHB EXH HTR A ON ON OFF OFF FHB EXH HTR B ON ON OFF OFF FHB EXH HTR C OFF ON OFF ON FHB EXH FAN A ON ON ON OFF FHB EXH FAN B ON ON ON OFF FHB EXH FAN C OFF ON OFF ON A. 1 B. 2 C. 3 D. 4 147

NRC 2017 Exam QUESTION RO 74 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 288000 A3.01 Importance Rating 3.8 K&A: Ability to monitor automatic operations of the Plant Ventilation Systems including:

Isolation/initiation signals Plant Ventilation Explanation: Answer A - Both FHB supply fans trip. All other components remain as is.

B - Incorrect - Plausible since AEGTS and M26 heaters automatically start when fans auto start.

C - Incorrect - The running supply fan trips and the heaters remain energized. Plausible if confuses requirements for supply vs. exhaust fans running.

D - Incorrect - The supply fans trip and exhaust fans continue to run. Plausible if operator incorrectly recalls auto start logic.

Technical Reference(s): ARI-H13-P680-08 Rev 15, ONI- Reference Attached: ARI-H13-P680-08 pp 3-4, D17 Rev 18, SOI-M40 Rev 10 ONI-D17 p 18, SOI-M40 pp 3, 5-6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M40-B.4 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 148

NRC 2017 Exam QUESTION RO 75 The plant is operating at rated power when the following annunciator alarms on H13-P680:

• RWCU ISOL PUMP A/B RM TEMP HIGH How does the reactor operator determine which area is causing this alarm?

Reference Provided:

A. Use the Riley Room Temperature Monitor, on H13-P800 B. Use the Riley Room Temperature Indicator on H13-P904 C. Use the NUMAC Leak Detection Monitors on H13-P632/P642 D. Use the RWCU TEMP SELECTOR SW, G33-N601 on H13-P680.

149

NRC 2017 Exam QUESTION RO 75 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 290001 A4.02 Importance Rating 3.3 K&A: Ability to manually operate and/or monitor in the control room: Reactor building area temperatures Secondary CTMT Explanation: Answer C - Per the ARI, this alarm is caused by the NUMAC monitors. These monitors are located on H13-P632 and P642. SOI-E31 gives directions for trending temperatures using the NUMACs A - Incorrect - Plausible since steam tunnel temps can be monitored using this method.

B - Incorrect - Plausible since these modules causes other alarms in the control room.

D - Incorrect - Plausible since this switch is used to monitor various points within the RWCU system, but not the room temps.

Technical Reference(s): ARI-H13-P680-01 Rev 13, SOI- Reference Attached: ARI-H13-P680-01 p 37, SOI-E31 Rev 8 E31 p 9 Proposed references to be provided to applicants during examination: EOP-SPI Supplement Figure 13 - Modified Learning Objective (As available): OT-COMBINED-E31-C, E.1, Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x 150

NRC 2017 Exam QUESTION SRO 1 Per NOP-OP-1002, Conduct of Operations, what is the Command SRO required to do during a plant transient?

A. Announce when entering or exiting an ONI or EOP and when transitioning to a different EOP.

B. Provide peer checks for the ATC RO during component manipulations when the BOP RO is not available.

C. Assist the ATC RO by performing panel manipulations when multiple system manipulations are required.

D. Direct all Annunciator Response Instruction steps performed by the ATC RO while executing ONIs and EOPs.

151

NRC 2017 Exam QUESTION SRO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.1.6 Importance Rating 4.8 K&A: Ability to manage the control room crew during plant transients.

Generic Explanation: Answer A - Per NOP-OP-1002, Conduct of Operations: The Command SRO shall clearly announce to the crew when entering or exiting AOP/ONIs, and when transitioning to different sections/flow charts of the EOPs.

B - Incorrect - Plausible since during normal conditions (non-transient), the Command SRO is required to enforce and utilize peer checks. NOP-OP-1002 4.10.6.1: SROs should not be used for a peer check of component manipulations.

C - Incorrect - Plausible since the Command SRO prioritizes execution of procedures, but is prohibited from making panel manipulations. NOP-OP-1002 4.10.2.5: SROs shall not operate plant equipment and must remain in their roles, at all times.

D - Incorrect - Plausible since the Command SRO directs ONI and EOP actions, but ARI actions are owned by ROs. NOP-OP-1002 4.9.2.9: ARP/ARIs shall be owned by the reactor operators. This allows the Command SRO to maintain an oversight role during the execution of specified actions.

Technical Reference(s): NOP-OP-1002 Rev 11 Reference Attached: NOP-OP-1002 pp 56, 59-60,

& 65 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-1A & OT-3035-01(LP)-D.1.A Question Source: Bank # Perry 2005 Audit Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b Comments: SRO justification - Justification for Plant Specific Exemption for task unique to the SRO position.

A question is linked to a task that is labeled as an SRO-only task, and the task is NOT listed in the RO task list.

Tasks 344-507-05-02 & 344-508-05-02 152

NRC 2017 Exam QUESTION SRO 2 Reactor power must be lowered from 100% to 60% to repair a steam leak.

The Shift Manager and Shift Engineer are attending a meeting in SB-318 In accordance with NOP-OP-1004, Reactivity Management, who must authorize the Evolution Specific Reactivity Plan prior to use?

A. Ops Manager B. Shift Engineer C. Command SRO D. Reactor Engineer 153

NRC 2017 Exam QUESTION SRO 2 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.1.37 Importance Rating 4.6 K&A: Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Generic Explanation: Answer C - The Unit Supervisor is the Command SRO. All reactivity plans must be authorized by the Command SRO. With the other two shift SROs outside the control room, the Unit Supervisor must be the Command SRO A - Incorrect - Plausible if the Ops Manager is approval authority for the reactivity plan.

B - Incorrect - Plausible since the SE authorizes most work to be performed.

D - Incorrect - Plausible since the RE prepares the Evolution Specific Reactivity Plan and signs it and the Control Rod Movement Sheets are authorized by a Reactor Engineer.

Technical Reference(s): NOP-OP-1002 Rev 11 & NOP- Reference Attached: NOP-OP-1002 p 13 & NOP-OP-1004 Rev 13 OP-1004 p 31 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-N Question Source: Bank #

Modified Bank # Grand Gulf 2013 - # 95 New Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.6 Comments: SRO justification - ES-401 Att2 - Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]

154

NRC 2017 Exam QUESTION SRO 3 The plant has been operating at rated power for six months.

On July 6th at 0900 it is discovered that SR 3.5.1.1, with a required frequency of 31 days, has not been completed since June 2nd at 0800 for LPCS.

Given only the above conditions, when is the latest the surveillance can be completed for the system status to remain in compliance with Technical Specifications?

Reference Provided:

A. July 7th at 0900 B. July 11th at 0200 C. August 6th at 0900 D. August 14th at 0200 155

NRC 2017 Exam QUESTION SRO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.2.22 Importance Rating 4.7 K&A: Knowledge of limiting conditions for operations and safety limits.

Generic Explanation: Answer B - It has been 34 days since SR 3.5.1.1 was last successful completed. SR 3.5.1.1 has a 31 day frequency. However, with the SR 3.0.2 grace period the required completion date is extended 7.75 days (31x1.25=38.75) and as such, the required surveillance must be completed by July 11th @ 0200.

A & C 1st part - Incorrect - Plausible due to incorrect application of SR 3.0.3. A is 24 hrs. after discovery and C is 31 days after discovery. These options would be permitted assuming the applicant believed the surveillance was not completed within the specified frequency. But as of July 6th 0900 it is still within the specified frequency time with application of SR 3.0.2 grace period.

D- Incorrect - Plausible for incorrect combination of SR 3.0.2 and 3.0.3.

Technical Reference(s): TS 3.5.1, TS 3.0, & PDB-R002 Reference Attached: TS 3.5.1 p 3.5-4, TS 3.0 pp Rev 1. 3.0-4 &-5, & PDB-R002 p 35.

Proposed references to be provided to applicants during examination: TS SR 3.0.1 - 3.0.4 Learning Objective (As available): OT-3037-02-F, I & J Question Source: Bank # Monticello 2016 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

156

NRC 2017 Exam QUESTION SRO 4 Which of the following requires a 10CFR50.59, (Changes, tests, and experiments), evaluation?

A. Removal of floor plugs in Aux-620 per a Maintenance Work Order for one month to support Turbine Bldg. Chill Water system work.

B. Change of responsibility from Shift Manager to Unit Supervisor for approving Liquid Radwaste Discharge permits.

C. Installation of a jumper directed by SVI-B21-T0246A, ATWS-RPT Logic System Functional Test For Division 1.

D. Installation of a leak sealant device on 1G33-F107, RWCU HX SHELL SIDE BYPASS VALVE to stop a 5 gpm leak that will remain in place for at least 4 months.

157

NRC 2017 Exam QUESTION SRO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.2.5 Importance Rating 3.2 K&A: Knowledge of the process for making design or operating changes to the facility.

Generic Explanation: Answer D - IAW NOBP-LP-4003A, installation of a leak sealant device requires a 50.59 evaluation since it is not a temp alt supporting maintenance.

A - Incorrect - Removal of floor plugs for less than 90 days do not require a 50.59 eval. If it was for >90 days, an eval would be required.

B - Incorrect - While some managerial changes require a 50.59 eval, this is specifically exempted in NOBP-LP-4003A C - Incorrect - Installation of jumpers to support maintenance do not require a 50.59 evaluation unless it were to be left installed >90 days. In this case, the jumpers would be installed less than one shift.

Technical Reference(s): NOBP-LP-4003A Rev 8 Reference Attached: NOBP-LP-4003A pp 6, 10-11, 16-17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-02-B Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.3 Comments: SRO justification = Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

• 10 CFR 50.59 screening and evaluation processes.

158

NRC 2017 Exam QUESTION SRO 5 The Plant is shutdown for a refueling outage.

FDST B discharge is in progress IAW SVI-G50-T5266, Liquid Radwaste Release Permit.

Rad Monitor D17-K606, LRW TO ESW RAD MONITOR fails downscale and is declared inoperable.

The crew terminates the discharge In order to re-start the discharge, what does the ODCM (Offsite Dose Calculation Manual) require?

Reference Provided:

A. Analyze at least two samples of the tanks content and have at least two technically qualified members of the facility staff independently verify the release rate calculation.

B. Verification by at least two members of the facility staff of the discharge valve lineup and that the discharge valve position corresponds to the desired flow rate.

C. Obtain and analyze grab samples for gross radioactivity at least every twelve hours.

D. Estimate the flow rate at least every four hours during the actual release.

159

NRC 2017 Exam QUESTION SRO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.3.11 Importance Rating 4.3 K&A: Ability to control radiation releases.

Generic Explanation: Answer A - D17-K606 is the rad monitor identified in table 4.3.7.9-1 of the ODCM. ACTION 110 must be completed to restart the discharge.

B - Incorrect - Plausible since this is a required Action if G50-N445, Radwaste High Flow Discharge Header Flow rad monitor is OOS (Action 112).

C - Incorrect - Plausible since this is a required Action if D17-K604, Emergency Service Water Loops rad monitor is OOS (Action 111).

D - Incorrect - Plausible since this is a required Action if P41-N442, Service Water Discharge Header Flow rad monitor is OOS (Action 113).

Technical Reference(s): ODCM Rev 20 Reference Attached: ODCM pp 114-118 Proposed references to be provided to applicants during examination: ODCM Learning Objective (As available): OT-COMBINED-D17A-K.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

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

160

NRC 2017 Exam QUESTION SRO 6 Which one of the following identifies a condition that would require declaration of an ALERT only?

Reference Provided:

A.

• RPV level is lowering slowly

• Fifty Control Rods are withdrawn at various positions

• Reactor Power is 10%

• Mode Switch is in SHUTDOWN B.

• RPV level at 175 inches and lowering

• Control Rods 30-31, 18-55, 14-15, 46-15, 46-31, & 46-47 are at position 48

• Nuclear Instruments are fully inserted

• IRMs are indicating on range 4 C.

• Mode Switch is placed in SHUTDOWN

• Control Rod 46-15 is at position 48

• Nuclear Instruments are fully inserted

• Power is indicating middle of the source range D.

• RPV level at 17.9 inches and lowering

• Mode Switch is in SHUTDOWN

• All Control Rods indicate 00

• Nuclear Instruments are fully inserted

• Power is indicating middle of the source range 161

NRC 2017 Exam QUESTION SRO 6 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.4.41 Importance Rating 4.6 K&A: Knowledge of the emergency action level thresholds and classifications.

Generic Explanation: Answer B - This is a CA1 - failure to auto scram and power below 4%. RPS should have actuated at 178 and reactor is not shutdown under all conditions without boron.

A - Incorrect - This is a CS1 C - Incorrect - No EAL entry criteria listed. Reactor shutdown under all conditions without boron. No EAL entry.

D - Incorrect - RPV water level not below TAF. No EAL entry.

Technical Reference(s): EPI-A1 Rev 26 Reference Attached: EPI-A1 pp 14 & 23 Proposed references to be provided to applicants during examination: EPI-A1, Attachments 1 & 2 Learning Objective (As available): EPL-0804-01-4 Question Source: Bank # Perry 2010 # SRO-11 Modified Bank #

New Question History: Previous NRC Exam: Perry 2010 # SRO-11 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

162

NRC 2017 Exam QUESTION SRO 7 The Plant is operating at rated power.

Which one of the following conditions, if it were to occur, requires an Hourly Fire Watch Patrol?

Reference Provided:

A. RCIC Pump Room Wet-Pipe Sprinkler will not deliver water.

B. Heat Detection for Reactor Recirculation Pump B is out of service.

C. Unit 1 Division 1 Cable Spreading Pre-Action Spray System will not deliver water.

D. General area smoke detectors in Containment are functional but the detection system will not transmit an alarm to SAS.

163

NRC 2017 Exam QUESTION SRO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# Generic 2.4.26 Importance Rating 3.6 K&A: Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage.

Generic Explanation: Answer D - This equipment is required to be functional and transmit an alarm or establish hourly fire watch to protect plant equipment.

A - Incorrect - Continuous fire watch required.

B - Incorrect - Fire watch not required, remote monitoring required.

C - Incorrect - Continuous fire watch with each area inspected every 15 minutes.

Technical Reference(s): PAP-1910 Rev 34 Reference Attached: : PAP-1910 pp 58, 60, 64, 65, 77-78, 81, & 83 Proposed references to be provided to applicants during examination: PAP-1910 Body & Attachment #3 Learning Objective (As available): OT-3039-03-H Question Source: Bank # Perry 2009 #SRO-04 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 #SRO-04 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

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

164

NRC 2017 Exam QUESTION SRO 8 Plant startup in progress with plant in Mode 2.

Multiple annunciators on panel H13-P877 were received two minutes ago.

Refer to attached picture of partial panel H13-P877 for current conditions.

What Tech Spec LCO(s) Condition(s) must be entered, if any?

Attachment Provided:

Reference Provided:

A. Enter T.S. 3.8.4 Condition A only B. Enter T.S. 3.8.7 Condition B only C. Enter T.S. 3.8.4 Condition A and T.S. 3.8.7 Condition B D. Neither T.S. 3.8.4 Condition A nor T.S. 3.8.7 Condition B are entered 165

NRC 2017 Exam QUESTION SRO 8 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295004 AA2.02 Importance Rating 3.9 K&A: Ability to determine and/or interpret the following as they apply to Partial Or Complete Loss Of D.C. Power: Extent of partial or complete loss of D.C. power Partial or Total Loss of DC Pwr Explanation: Answer C - With the readings on the Div 1 Battery/Bus meters, the charger output is not sufficient to maintain Bus ED1A voltage >125V TS 3.8.4 Bases, charger must be operable. TS 3.8.4 Condition A is entered. Additionally, since ED1A voltage is less than required by Bases table B 3.8.7-1, TS 3.8.7 Condition B is also entered.

A - Incorrect - TS 3.8.7 Condition B must also be entered since ED1A Bus voltage < 125V. Plausible if candidate doesnt recall that TS 3.8.7 checks bus voltage.

B - Incorrect - TS 3.8.4 Condition A must also be entered since the charger is not maintaining bus voltage 129V. Plausible if candidate sees battery voltage in normal range and assumes only distribution is inop.

D - Incorrect - Both TS 3.8.4 & 3.8.7 LCOs are not met. Plausible if candidate does not recall the minimum bus voltage for ED1A.

Technical Reference(s): TS 3.8.4, TS 3.8.7, TS 3.8.4 Reference Attached: TS 3.8.4 p 3.8-24, TS 3.8.7 p Bases Rev 7, & TS 3.8.7 Bases Rev 1 3.8-26, TS 3.8.4 Bases B 3.8-53, & TS 3.8.7 Bases pp 73, 78, 80 Proposed references to be provided to applicants during examination: TS 3.8.4 (partial) and TS 3.8.7 (partial)

Learning Objective (As available): OT-3037-12 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

166

NRC 2017 Exam QUESTION SRO 9 The Plant is in a forced outage with the following conditions:

• RHR A in operating in SDC

• RCS temperature band 110°F to 120°F.

• RPV water level band is 200 to 210

• SOI-E12 Residual Heat Removal System, Section 7.40, Protecting Shutdown Cooling Operation was not performed.

While performing maintenance, RHR A Min Flow Valve 1E12-F064A was inadvertently opened.

RPV level lowered to 172 before 1E12-F064A was able to be reclosed.

Based on the above conditions only, which of the following describes the NRC Notifications required for this event?

Reference Provided:

A. Only a 4 Hour Notification B. Only a 8 Hour Notification C. 1 Hour Notification and 4 Hour Notification D. 1 Hour Notification and 8 Hour Notification 167

NRC 2017 Exam QUESTION SRO 9 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295021 2.4.30 Importance Rating 4.1 K&A: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Loss of Shutdown Cooling Explanation: Answer B - When RPV level lowered to <178, a valid RPS signal was generated causing an isolation which resulted in a loss of shutdown cooling. An 8-hour notification is required for a valid RPS signal with reactor not critical and the valid isolation signal.

A - Incorrect hour notification not required. Plausible if Rx was critical at time of RPS actuation.

C - Incorrect hour notification not required. Plausible if candidate assumes entry into E-plan required for loss of SDC. Only required if approaching or exceeding 200°F.

D - Incorrect hour notification not required. Plausible if Rx was critical at time of RPS actuation.

Technical Reference(s): NOBP-OP-1015 Rev 3 Reference Attached: NOBP-OP-1015 pp 33, 64 &

78 Proposed references to be provided to applicants during examination: NOBP-OP-1015 Learning Objective (As available): OT-3039-01-A & F Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• The SRO is responsible for making NRC Notifications.

168

NRC 2017 Exam QUESTION SRO 10 Refueling activities are in progress.

The following conditions exist on the Refuel Floor:

• A fuel bundle just arrived from the Fuel Handling Building with the IFTS Upender vertical

• A fuel bundle is in transit from the RPV to the fuel storage racks via Refueling Bridge

• An unexplained drop in upper pool level occurs Which of the following actions is required concerning the status of the two bundles?

A. Incline the IFTS Upender and continue the fuel movement with the Refueling Bridge to the fuel storage racks.

B. Incline the IFTS Upender and return the fuel bundle on the Refueling Bridge back to any open vessel location.

C. Continue fuel movement with the Refueling Bridge to the fuel storage racks, then, transfer the fuel bundle in IFTS to the fuel storage racks.

D. Transfer the fuel bundle in IFTS down to the Fuel Handling Building and return the fuel bundle on the Refueling Bridge back to the vessel location from which it was removed.

169

NRC 2017 Exam QUESTION SRO 10 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295023 AA2.02 Importance Rating 3.7 K&A: Ability to determine and/or interpret the following as they apply to Refueling Accidents:

Fuel pool level Refueling Acc Explanation: Answer A - The stem contains entry conditions for ONI-E12-2. The ONI also describes Safe Conditions for a fuel bundle. FTI-D009 further restricts placing a bundle back into the Rx after it has been off-loaded. The SRO is responsible for all the fuel moves on the refuel floor.

B - Incorrect - Moving the bundle back to the core is an unsafe act/condition and not allowed by FTI-D09.

C - Incorrect - An IFTS transfer to the FHB is an unsafe act/condition and not allowed by FTI-D-09.

D - Incorrect - An IFTS transfer to the FHB is an unsafe act/condition - moving the bundle back to the core is an unsafe act/condition, both are not allowed by FTI-D09 Technical Reference(s): ONI-E12-2 Rev 36 & FTI-D009 Reference Attached: ONI-E12-2 pp 4-7 & FTI-Rev 18 D009 pp 11-12 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-11(LP)-A.1 & OT-3602-01-D.4 & E.2 Question Source: Bank # Perry 2010 # SRO-12 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 # SRO-12 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.7 Comments: SRO justification = Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)]

• Refuel floor SRO responsibilities.

• Decay heat assessment.

170

NRC 2017 Exam QUESTION SRO 11 The plant was operating at rated power with Suppression Pool level at 18.3 feet.

Then a transient occurred.

Refer to the attached sheet for current plant conditions.

If the Drywell pressure trend continues, indicated Suppression Pool level will (1) .

Per the Bases for Technical Specification 3.6.2.2, Suppression Pool Water Level, if this trend is not stopped, suppression pool level could result in (2) in the event of a LOCA.

Attachment Provided:

(1) (2)

A. raise excessive hydrodynamic loads on submerged structures during SRV and horizontal vent steam discharges B. raise voiding the analysis for maximum drain down of the suppression pool C. lower excessive hydrodynamic loads on submerged structures during SRV and horizontal vent steam discharges D. lower voiding the analysis for maximum drain down of the suppression pool 171

NRC 2017 Exam QUESTION SRO 11 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295024 EA2.03 Importance Rating 3.8 K&A: Ability to determine and/or interpret the following as they apply to High Drywell Pressure: Suppression pool level High Drywell Pressure Explanation: Answer A - As Drywell pressure increases, indicated Suppression Pool level will raise due to the SP area of the DW being connected to the SP area of the containment via the horizontal vents.

Also, TS 3.6.2.2 Bases states document states the upper limit is based, in part on precluding excessive dynamic loading on the S/RV.

C & D 1st part - Incorrect - Plausible since this will cause the SL level in the DW to lower, but not the indicated SP level.

B & D 2nd part - Incorrect - Plausible since this is the basis for minimum suppression pool level. However, with DW pressure increasing, SP level is going up.

Technical Reference(s): TS 3.6.2.2 Bases Rev 7 and Reference Attached: TS 3.6.2.2 Bases p B 3.6-76 Lesson Plan OT-3037-10 Rev 3 & Lesson Plan OT-3037-10 p 18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-10-B Question Source: Bank #

Modified Bank # Perry 2013 # SRO-16 New Question History: Previous NRC Exam Perry 2013 # SRO-16 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that are required to analyze TS required actions and terminology.

172

NRC 2017 Exam QUESTION SRO 12 The following conditions exist:

• SVI-E51-T2001, RCIC Pump and Valve Operability test is in progress following maintenance

• SVI-D23-T1213, Suppression Pool Average Temperature is being performed by an I&C tech Based on the information from the attached SPDS screen print, select the required action.

Attachment and Reference Provided:

A. Trip the RCIC turbine Restore Suppression Pool Average Temperature to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Verify Suppression Pool Average Temperature 110°F once per 30 minutes B. Trip the RCIC turbine Verify Suppression Pool Average Temperature 110°F once per hour Restore Suppression Pool Average Temperature to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. Initiate Suppression Pool Cooling Restore Suppression Pool Average Temperature to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Verify Suppression Pool Average Temperature 120°F once per 30 minutes D. Initiate Suppression Pool Cooling Reduce Reactor Power to 1% RTP in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Verify Suppression Pool Average Temperature 110°F once per hour 173

NRC 2017 Exam QUESTION SRO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 2.2.22 Importance Rating 4.7 K&A: Knowledge of limiting conditions for operations and safety limits.

Suppression Pool High Water Temp.

Explanation: Answer B - With a suppression pool average temperature > 105° F, testing adding heat to the pool must be suspended immediately (trip the RCIC turbine). Once testing is stopped, TS Bases states that the SP temperature must be lowered to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A - Incorrect - Verifying SP temp 110°F is once per hour C & D - Incorrect - Initiating SP Cooling is not required D - Incorrect - Lowering Rx power is not a correct requirement Technical Reference(s): TS 3.6.2.1 & TS 3.6.2.1 Bases Reference Attached: TS 3.6.2.1 pp 36-38 & TS Rev 1 3.6.2.1 Bases pp 70-71 Proposed references to be provided to applicants during examination: SPDS screen print and Technical Specification 3.6.2.1 Learning Objective (As available): OT-3037-10-B Question Source: Bank # Perry 2010 # SRO-13 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 # SRO-13 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Knowledge of TS bases that are required to analyze TS required actions and terminology.

174

NRC 2017 Exam QUESTION SRO 13 The plant was operating at rated power with Containment Vessel Chiller A in service when the following occurred:

Containment Vessel Chiller A tripped on low refrigerant pressure Containment temperature rose to 100°F.

Containment temperature continues to rise slowly.

Based on the above information, which of the following procedures provides the required actions that mitigate these plant conditions?

A. EOP-2, Primary Containment Control and SOI-P50, Containment Vessel Chilled Water System B. EOP-2, Primary Containment Control and EOP-SPI 2.2, Bypass of CVCW Isolation C. EOP-1, RPV Control and SOI-M11, Containment Vessel Cooling System D. EOP-1, RPV Control and ONI-C71, Reactor Scram 175

NRC 2017 Exam QUESTION SRO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295027 EA2.01 Importance Rating 3.7 K&A: Ability to determine and/or interpret the following as they apply to High Containment Temperature (Mark III Containment Only): Containment temperature High Containment Temperature Explanation: Answer A - EOP-2 is entered when containment temperature exceeds 95°F. Since the containment vessel chiller tripped on low refrigerant pressure, SOI-P50 contains actions to start another chiller.

B - Incorrect - Although EOP-2 directs EOP-SPI 2.2, no isolation has occurred.

C - Incorrect - EOP-1 entry is directed from EOP-2, however there is sufficient margin to EOP-1 entry to not require entry at this time. SOI-M11 contains a section to Maximize Containment Cooling and EOP-2 directs Maximizing Containment Cooling. But, without a chiller, starting more cooling fans will have no effect.

D - Incorrect - If containment temperature rose high enough, EOP-1 and ONI-C71-1 would be entered.

However, there is sufficient margin to EOP-1 entry to not require scramming at this time.

Technical Reference(s): EOP-2 Chart Rev D, SOI-P50 Reference Attached: EOP-2 chart (partial), SOI-Rev 12, & ARI-H13-P904-01 Rev 10 P50 pp, & ARI-H13-P904-01 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-7-A & -C Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.

176

NRC 2017 Exam QUESTION SRO 14 An OBE earthquake caused a rupture of a Radwaste tank.

• Multiple Radwaste Building Area radiation monitors have Alert and High alarms locked in

• The RWB VENT EXH GAS (D17-K727) radiation monitor has Alert and High alarms locked in

• Chemistry reports contamination in the Underdrain System Which of the following actions would the Unit Supervisor direct?

1 Startup all available Emergency Service Water loops IAW SOI-P45/49, Emergency Service Water System 2 Shutdown all running Emergency Service Water loops IAW SOI-P45/49, Emergency Service Water System 3 Startup all available Underdrain pumps IAW SOI-P72, Plant Foundation Underdrain System 4 Shutdown all running Underdrain pumps IAW SOI-P72, Plant Foundation Underdrain System A. 1 and 3 B. 2 and 3 C. 1 and 4 D. 2 and 4 177

NRC 2017 Exam QUESTION SRO 14 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295038 2.4.50 Importance Rating 4.0 K&A: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

High Off-site Release Rate Explanation: Answer C - A rupture of a Radwaste tank can migrate to the underdrain system. The underdrain pumps are shutdown to minimize the release of radioactive water. If the radioactive water migrates to the underdrain system as confirmed by Chemistry sampling, and the water level in the underdrain system rises, the ESW pumps are started to provide a dilution source as the water is released off site.

A - Incorrect - Starting underdrain pumps is plausible if candidate thinks this action will control and divert the release.

B - Incorrect - Stopping ESW and Starting underdrain pumps is plausible if candidate believes this will better control the release.

D - Incorrect - Stopping both the ESW pumps and the underdrain pumps is plausible if candidate believes this will prevent any release.

Technical Reference(s): USAR C-15 Rev 13 & ONI-D17 Reference Attached: USAR pp 15.7-1719 & ONI-Rev 18 D17 pp 3, 9-10, & 14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): 3035-01(LP)A.3 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.4 Comments: SRO justification = Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. [10 CFR 55.43(b)(4)]

• Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

178

NRC 2017 Exam QUESTION SRO 15 A transient is in progress.

Refer to the attached sheet for current plant conditions.

Which of the following procedures contains the specific actions to mitigate this transient?

Attachment Provided:

A. ARI-H13-P660-03-A9, RX LEVEL HI/LO L7/L4 B. ARI-H13-P660-03-B7, FEED FLOW STEAM FLOW MISMATCH C. ONI-C34, Feedwater Flow Malfunction D. ONI-C71-1, Reactor Scram 179

NRC 2017 Exam QUESTION SRO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295008 AA2.01 Importance Rating 3.9 K&A: Ability to determine and/or interpret the following as they apply to High Reactor Water Level: Reactor water level High Reactor Water Level Explanation: Answer C - With the indications given, Feedwater flow is excessive for indicated steam demand (flow). This indicates there is a feed control malfunction. ONI-C34 contains the steps to control feed flow to match steam flow.

A - Incorrect - Plausible since this alarm is received for the indicated conditions. However, this ARI does not contain specific steps to control Feedwater flow.

B - Incorrect - Plausible since this alarm is received for the indicated conditions. However, this ARI does not contain specific steps to control Feedwater flow.

D - Incorrect - Plausible, as ONI-C71-1 does contain a general step to control RPV water level and a reactor scram will occur if feed flow is not controlled. However, for the given conditions, ONI-C71-1 would not be entered yet.

Technical Reference(s): ONI-C34 Rev 9, ONI-C71-1 Rev Reference Attached: ): ONI-C34 pp 3, 5-6, & 8, 20, & ARI-H13-P680-03 Rev 15 ONI-C71-1 p 6, & ARI-H13-P680-03 pp 25-26 &

33-34 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-04(LP) A1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

180

NRC 2017 Exam QUESTION SRO 16 The plant was operating at rated power when a transient occurred resulting in a reactor scram.

Current plant conditions are as follows:

• Appropriate EOPs have been entered

• 2 SRVs are open

• RPV pressure is 900 psig and stable

• Suppression Pool temperature is 115°F

• Termination and Prevention of injection into the RPV has been completed

• RPV water level is 89 inches and lowering

• Drywell pressure is 1.9 psig Based on the above information, when would you to direct the RO to recommence feeding the RPV?

A. All SRVs are closed B. RPV level lowers to 50 inches C. Drywell pressure lowers below 1.68 psig D. APRM Downscale lights illuminate on P680 181

NRC 2017 Exam QUESTION SRO 16 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295015 AA2.01 Importance Rating 4.3 K&A: Ability to determine and/or interpret the following as they apply to Incomplete SCRAM:

Reactor power Incomplete SCRAM Explanation: Answer D - With 2 SRVs open and RPV pressure stable at 900 psig, Rx power is approximately 8-10%. This will require the SRO to direct action from EOP-1A, which requires injection remain Terminated and Prevented until one of several conditions are met. APRM's Downscale is a standalone condition that allows for recommencement of injection to stabilize RPV level.

A - Incorrect - Plausible since reinjection is allowed if SRVs are closed and Drywell pressure <1.68 psig.

B - Incorrect - Would be correct if level lowered to 16.5 inches. Plausible since this is the lower end of the level band if Suppression Pool temperature was <110°F.

C - Incorrect - This is partially correct. However, SRVs must remain closed to recommence injection.

Technical Reference(s): EOP-1A Chart Rev F Reference Attached: EOP-1A Chart Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-04B-A Question Source: Bank # Clinton 2002 #117 Modified Bank #

New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

182

NRC 2017 Exam QUESTION SRO 17 The plant was operating at rated power when an earthquake occurred.

Subsequently the following annunciators alarmed.

• RHR B PUMP ROOM SUMP LEVEL HIGH

• RHR C PUMP ROOM SUMP LEVEL HIGH

• AUX BLDG. 568 EL WATER LVL HIGH

• SUPR POOL LEVEL A HI/LO

• SUPR POOL LEVEL B HI/LO An NLO reports there is about 10 inches of water in Aux 568.

Suppression Pool level is lowering slowly.

Which of the following actions has the highest priority?

A. Perform Emergency Depressurization.

B. Operate RHR B and C sump cubicle drains.

C. Transition to EOP-1, RPV Control and scram the reactor.

D. Commence normal plant shutdown IAW IOIs or ONI-C71-1 183

NRC 2017 Exam QUESTION SRO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295036 2.4.20 Importance Rating 4.3 K&A: Knowledge of operational inplications of EOP warnings, cautions, and notes.

Secondary Containment High Sump/Area Water Level Explanation: Answer D - Per a NOTE in EOP-3, when any area is inaccessible due to a hazard and the associated sump level annunciator is in alarm then the Max Safe operating Limit for that area is considered exceeded. With RHR B & C sump level alarms in and water in Aux 568, RHR B & C rooms are inaccessible and 2 max safes are exceeded. Since there is no indication of a primary system leaking, the SRO must commence a normal plant shutdown.

A - Incorrect - Plausible since ED is required if 2 or more areas of same parameter exceeded max safe if primary system discharging.

B - Incorrect - Plausible, since this is a required action. However, since there is water on Aux 568, operation of the cubical drain valves would not be possible without extraordinary measures. Additionally, 2 max safes are already exceeded.

C - Incorrect - Plausible since this is the required action if it was a primary system leaking.

Technical Reference(s): EOP-3 Chart Rev E, EOP-3 Reference Attached: EOP-3 Chart partial, EOP-3 Bases Rev 5, ARI-H13-P601-18 Rev 16, & ARI-H13- Bases p 49, ARI-H13-P601-18 pp 37-38 & 41-42, P870-03 Rev 10 and ARI-H13-P870-03 pp 5-6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-17-C & 17-D Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.

184

NRC 2017 Exam QUESTION SRO 18 The plant was operating at rated power when an ATWS occurred.

Current conditions are as follows:

• Reactor power is 5% and stable

• RPV level is 80 inches and stable

• RPV pressure is 820 psig lowering slowly

• SLC Pumps A & B control switches in ON

• SLC A Squib Continuity light is ON

• SLC B Squib Continuity light is OFF

• SLC Pumps A & B Discharge Pressures are 1400 psig Based on these conditions the SLC system is ____.

A. Injecting boron into the RPV at 86 gpm. Direct crew to continue injecting boron until SLC tank level reaches 200 gallons Per EOP-1A B. Injecting boron into the RPV at 43 gpm. Direct crew to continue injecting boron until SLC tank level reaches 200 gallons Per EOP-1A C. Not injecting boron into the RPV. Direct crew to perform SOI-C41, SLC Transfer System Emergency Preparation/Transfer D. Not injecting boron into the RPV. Direct crew to perform EOP-SPI 1.8 Alternate Boron Injection 185

NRC 2017 Exam QUESTION SRO 18 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 211000 A2.04 Importance Rating 3.4 K&A: Ability to (a) predict the impacts of the following on the Standby Liquid Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow SLC Explanation: Answer D - With RPV pressure at ~800 psig, SLC discharge pressure should be only slightly greater than 800 psig. Discharge pressure of 1400 psig indicates the pump relief valves are lifting.

EOP-SPI-1.8 contains direction for Alternate Boron Injection.

A - Incorrect - Plausible as 86 gpm is the normal flow rate for 2 pumps. However, SLC is not being injected into the RPV. SLC discharge pressure should be only slightly above Rx pressure. 2nd part - correct action if injecting.

B - Incorrect - Plausible as 43 gpm is the normal flow rate for 1 pump. However, SLC is not being injected into the RPV. SLC discharge pressure should be only slightly above Rx pressure. 2nd part - correct action if injecting.

C - Incorrect - This procedure would be used to refill the SLC storage tank.

Technical Reference(s): GMI-008 Rev 23, SOI-C41 Rev Reference Attached: GMI-008 p18, SOI-C41 pp 9 20, SDM-C41 Rev 10, Dwg. 302-691 Rev Z, EOP-1A & 60, SDM-C41 pp 8 & 11,Dwg. 302-691 partial, Chart Rev G EOP-1A Chart partial Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED--C41-G & O.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.

186

NRC 2017 Exam QUESTION SRO 19 The plant was operating at 15% power in preparation of shutdown.

The HPCS pump was tagged out for motor replacement when a LOOP occurred.

Below are the current conditions:

• Div 1 DG tripped on overspeed.

• Div 2 DG failed to start.

• RCIC is injecting at 500 gpm.

• RPV level is 5 inches and stable.

• Containment temperature is 186°F and rising 1°F per minute.

• Maintenance expects repairs to Div 2 DG to be complete in 30 minutes.

How should the Unit Supervisor direct cool down?

Depressurize the RPV (1) to (2) psig.

(1) (2)

A. disregarding cooldown rate <135 B. disregarding cooldown rate between 150 to 250 C. maintaining Tech Spec cooldown rate <135 D. maintaining Tech Spec cooldown rate between 150 to 250 187

NRC 2017 Exam QUESTION SRO 19 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 2.4.9 Importance Rating 4.2 K&A: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

RCIC Explanation: Answer B - With containment temperature at 186°F and rising 1°/min, ED is required by EOP-2.

The SRO needs to transition from EOP-2 to EOP-1 then to EOP-4-2 (ED). Since RCIC is injecting

@ 500 and RPV level is stable, RCIC is required for adequate core cooling. A NOTE in EOP-1 tells the US to terminate ED to maintain RCIC if required for ACC and directs a pressure band of 150 to 250 psig until RCIC no longer required for ACC.

A - Incorrect - Cooldown to <135°F is required to start SDC. However, based on containment H/U rate, ED will be required before SDC could be placed in service.

C - Incorrect - Maintaining TS cooldown rate is not required when ED is required.

D - Incorrect - Maintaining TS cooldown rate is not required when ED is required.

Technical Reference(s): EOP-1 Chart Rev F, EOP-2 Reference Attached: EOP-1 Chart, EOP-2 Chart, &

Chart Rev D, & EOP-4-2 Chart Rev E EOP-4-2 Chart Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-02-F, C, & 12-C.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

188

NRC 2017 Exam QUESTION SRO 20 The plant was operating at 35% power when a transient occurred.

A scram was inserted.

The following alarms were received on H13-P680:

• RPS RX LEVEL LO L3

• RRCS RX LEVEL LO L2 HPCS and RCIC started automatically.

Refer to the attached SPDS printouts for RPV level and valve isolation data.

Based on these conditions, the (1) valves should also have isolated.

Direct required isolation actions from (2) .

Attachments Provided:

(1) (2)

A. Reactor Water Clean Up IOI-18, Emergency Operating Procedure and and Isolation Restoration Reactor Sampling B. Reactor Water Clean Up EOP-1, RPV Control and Reactor Sampling C. only Reactor Water Clean Up IOI-18, Emergency Operating Procedure and Isolation Restoration D. only Reactor Water Clean Up EOP-1, RPV Control 189

NRC 2017 Exam QUESTION SRO 20 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 2.4.46 Importance Rating 4.2 K&A: Ability to verify that the alarms are consistent with the plant conditions.

PCIS/Nuclear Steam Supply Shutoff Explanation: Answer B - The level trace shows RPV level lowered below L2. This would cause both RWCU and Rx Sample valves to isolate. The direction to isolate valves that didnt properly isolate is directed from EOP-1, Verify Isolations and Actuations A & C 2nd part - Incorrect - Plausible since this is the procedure to recover from isolations of these valves.

C & D 1st part - Incorrect - Both RWCU and Rx Sample valves should isolate on L2 Technical Reference(s): ARI-H13-P680-05 Rev 15, ARI- Reference Attached: ARI-H13-P680-05 pp 5 & 27, H13-P601-19 Rev 19, EOP-1 Chart Rev G EOP-1 Bases ARI-H13-P601-19 p 39, Chart (Partial) EOP-1 Rev 7 Bases p 29 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21(NS4)-9, -11 & -21 and OT-3403-01A(SG)-F.3 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

190

NRC 2017 Exam QUESTION SRO 21 The following conditions exist:

• The reactor scrammed due to a small-break LOCA

• The only available injection source is from the Condensate Transfer system

• To maximize injection, Emergency Depressurization was initiated approximately 20 minutes ago and all ADS SRVs were verified open

• RPV level is -10 inches and rising slowly

• The SRV OPEN annunciator just reset You have directed the Reactor Operator to verify the status of the ADS SRVs.

The ADS SRVs are reported to be (1) . You would direct the panel operators to (2) .

Reference Provided:

(1) (2) closed based on the GREEN light ON bypass Instrument Air Isolations IAW A.

above the SRV switches EOP-SPI 2.8 to open non-ADS SRVs closed based on the GREEN light ON open additional SRVs IAW EOP 4-2 B.

above the SRV switches Emergency Depressurization open based on SRV tailpipe restore and maintain RPV level greater C.

temperatures of approximately 250°F than 16.5 IAW EOP 4-1 ALTERNATE and stable. LEVEL CONTROL open based on SRV tailpipe perform Containment Spray IAW EOP-D.

temperatures of approximately 250°F SPI 3.1 and stable.

191

NRC 2017 Exam QUESTION SRO 21 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 239002 A2.05 Importance Rating 3.4 K&A: Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor pressure SRVs Explanation: Answer C - The US is in EOP 4-2 for ED. With ED occurring 20 minutes ago, RPV pressure will decay to <30 psig. This will cause the SRV OPEN annunciator to reset and the SRV Open/Close lights to change state. Since the only injection source is CTS, RPV pressure cannot be >30 psig wirh RPV level @ -10. The SRVs are verified open by observing SRV tailpipe temperature of 250°F which corresponds to reactor pressure of ~25 psig. Since RPV level is <16.5, the US will transition to EOP-4-1 and direct the RO to restore and maintain RPV level using CTS.

A - Incorrect - SRVs are still open. The green light indicates the pressure in < 30 psig in the line. No need to bypass IA isolations to open additional SRVs B - Incorrect - SRVs are still open. Based on given conditions, actions to open additional SRVs is inappropriate D - Incorrect - Containment Spray is only performed if RPV water level cannot be maintained > -25.

Technical Reference(s): ARI-H13-P601-019 Rev 19, Reference Attached: ARI-H13-P601-019 p 17 &

EOP-4-1 Chart Rev F & ABB Steam Tables EOP-4-1 chart (partial)

Proposed references to be provided to applicants during examination: Steam Tables Learning Objective (As available): OT-COMBINED-B21_N11-F & I.1 and OT-3402-02-F &

Question Source: Bank # Perry 2010 # SRO-21 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 # SRO-21 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.

192

NRC 2017 Exam QUESTION SRO 22 The plant is in MODE 2.

Lake Erie temperature is 45°F.

Then, annunciator ESW B SLUICE GATE POWER LOSS alarmed.

A walkdown of ESW B Sluice Gate, revealed that the Sluice Gate MCC disconnect was charred and in need of repair.

What is/are the required action(s), if any, for maintaining the safety function of the ESW Systems?

A. Manually open ESW B Sluice Gate only.

B. Verify ESW A Sluice Gate is capable of opening.

C. Align all loops of ESW to the swale and open & deactivate ESW A Sluice Gate.

D. No action required unless ESW Forebay temperature approaches maximum design limit.

193

NRC 2017 Exam QUESTION SRO 22 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 2.2.38 Importance Rating 4.5 K&A: Knowledge of conditions and limitations in the facility license.

Component Cooling Water Explanation: Answer C - Per Tech Spec Bases, 3.7.1, if a sluice gate is being tested or repaired, the safety function of ESW is maintained by locking open one sluice gate and aligning all ESW loops to the swale.

A - Incorrect - The ESW loops also need to be aligned to the swale to maintain safety function.

B - Incorrect - One sluice gate must be locked open, not just capable of opening unless ESW B is declared inop.

D - Incorrect - Misconception that no action required until the forebay warms up. Monitoring ESW forebay temp is a condition of this alignment. However, it will not maintain the safety function.

Technical Reference(s): TS 3.7.1 Bases Rev 5 & ARI- Reference Attached: TS 3.7.1 Bases pp B 3.7-3 &

H13-P601-17 Rev 15 3a and ARI-H13-P601-17 p 117 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P45-H & -K.2 and OT-3037-11-C Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

194

NRC 2017 Exam QUESTION SRO 23 The plant is operating at rated power.

The Instrument Air Line to in-service CRD Flow Control Valve becomes severed at the valve.

The scram function (1) maintained and the US would direct the RO to (2) .

(1) (2)

A. is shift flow control valves per SOI-C11(CRDH),

Control Rod Drive Hydraulics System B. is take manual control of CRD HYDRAULICS FLOW CONTROL and control flow per NOP-OP-1002, Conduct of Ops C. is not shift flow control valves per SOI-C11(CRDH),

Control Rod Drive Hydraulics System D. is not take manual control of CRD HYDRAULICS FLOW CONTROL and control flow per NOP-OP-1002, Conduct of Ops 195

NRC 2017 Exam QUESTION SRO 23 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201001 A2.04 Importance Rating 3.9 K&A: Ability to (a) predict the impacts of the following on the Control Rod Drive Hydraulic System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Scram condition CRD Hydraulic Explanation: Answer A - With a loss of Instrument Air to the FCV, the FCV fails closed, but, the charging water header pressure is maintained >1520 psig (~1700 psig). Therefore, with Rx at rated pressure and charging water header pressure >1520 psig, the scram will occur if needed. To mitigate this condition, the FCVs must be shifted per SOI-C11(CRDH).

B - Incorrect - Per NOP-OP-1002 if a system is not operating properly in AUTO, the controller may be taken to MANUAL. Taking manual control of the flow controller will not have any effect on the high flow as the controller is already calling for 0 flow.

C - Incorrect - Scram function is maintained.

D - Incorrect - Scram function is maintained and taking manual control of the flow controller will not have any effect on the high flow as the controller is already calling for 0 flow.

Technical Reference(s): SOI-C11(CRDH) Rev 25, SDM- Reference Attached: SOI-C11(CRDH) pp 21-23, C11-CRDH Rev 8, TS 3.1.5 Bases Rev 1 SDM-C11-CRDH p 43, TS 3.1.5 Bases p B 3.1-32 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11-L.1 & K.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: SRO justification = Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

196

NRC 2017 Exam QUESTION SRO 24 The plant is operating at rated power.

The vendor reported a 10CFR Part 21 defect for the hydrogen igniter power supplies.

The US has declared both divisions of hydrogen igniters inoperable.

Which of the following is required to maintain the hydrogen control function of primary containment?

A. Only one division of Hydrogen Recombiners B. Both divisions of Combustible Gas Mixing Systems C. Two Hydrogen Recombiners or one Combustible Gas Mixing System D. One Hydrogen Recombiner and one Combustible Gas Mixing System 197

NRC 2017 Exam QUESTION SRO 24 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 223001 2.2.25 Importance Rating 4.2 K&A: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Primary CTMT and Aux.

Explanation: Answer D - Per TS 3.6.3.2 Bases, if both divisions of H2 Igniters are inop, the H2 control function is maintained by verifying 1 H2 recombiner and 1 combustible gas mixing system is operable.

A - Incorrect - Plausible since if both recombiners are inop, only one division of H2 igniters will maintain alternate H2 control capabilities per ORM 6.4.12.

B - Incorrect - Not correct per TS bases. Plausible if operator fails to remember that one division of recombiners and mixing system is also required.

C - Incorrect - Both are required. Plausible if operator fails to remember that both are required.

Technical Reference(s): TS 3.6.3.2 Bases Rev 5 Reference Attached: TS 3.6.3.2 Bases p B 3.6-98 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M51_M56-1.11 & OT-3037-10-B Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

198

NRC 2017 Exam QUESTION SRO 25 The plant was operating at rated power with Control Room Ventilation operating in the NORMAL mode when indication was lost on H13-P904 for M26-D001B, CONT RM EMG RCIRC ELECT HTR.

The NLO reports the control power fuse is blown in MCC EF1C09-H for M26-D001B.

Is entry into any Tech Spec LCO(s) required?

A. No Tech Spec entry is required.

B. Tech Spec 3.7.3, Control Room Emergency Recirculation (CRER) System only.

C. Tech Spec 3.7.4, Control Room Heating Ventilation and Air Conditioning (HVAC)

System only.

D. Both Tech Spec 3.7.3, Control Room Emergency Recirculation (CRER) System and Tech Spec 3.7.4, Control Room Heating Ventilation and Air Conditioning (HVAC)

System.

199

NRC 2017 Exam QUESTION SRO 25 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 288000 2.2.42 Importance Rating 4.6 K&A: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Plant Ventilation Explanation: Answer B - In accordance with TS Bases, the electrical heater is required for operability of the Emergency Recirc System, but not the normal HVAC system.

A - Incorrect - TS Bases for 3.7.3 specifically specifies the heater as a requirement for operability. Plausible since CR HVAC in in Normal mode and not ER.

C - Incorrect - TS Bases for 3.7.4 specifically exempts the heater as a requirement for operability. Plausible since the heater is energized if a high humidity condition occurs in the Control Room D - Incorrect - TS Bases for 3.7.4 specifically exempts the heater as a requirement for operability. Plausible if operator thinks heater is also required in Normal mode for high humitity.

Technical Reference(s): TS 3.7.3, TS 3.7.4, TS 3.7.3 Reference Attached: TS 3.7.3 p 3.7-4,, TS 3.7.4 p Bases Rev 10, & TS 3.7.4 Bases Rev 4 3.7-8, TS Bases pp 3.7-11a, & 3.7-18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M25_M26-K.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: SRO justification = Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

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

200

Attachment for Question RO 23 Attachment for Question RO 45 Question RO 51 Attachment Question RO 59 Attachment Partial panel picture of H13-P877

Number:

PERRY NUCLEAR POWER PLANT PDB-A0006

Title:

Use Category:

Power Flow Map In-Field Reference Revision: Page:

15 3 of 10

QUESTION SRO 8 Partial Panel Picture of H13-P877

Question SRO 11 Attachment 11 / 23 / 16 11 : 25 : 35 106 YES 0.7 ES 0.0

Question SRO 15 H13-P680 Partial Instrument displays on Section 3 and Section 5

Question SRO 20 Attachment Question SRO 20 Attachment 2017 NRC ILO Exam Provided References A. EOP-SPI EOP-SPI Supplement - Figure #7 RO-37 B. EOP-SPI EOP-SPI Supplement - Back Panel Locations RO-75 C. TS 3.0 SR Tech Spec Surveillance Requirement Applicability SRO-03 D. ODCM Offsite Dose Calculation Manual SRO-05 E. EPI-A1 Emergency Action Levels SRO-06 F. PAP-1910 Fire Protection Program SRO-07 G. TS 3.8.4 DC Systems - Operating SRO-08 H. TS 3.8.7 Distribution Systems - Operating SRO-08 I. NOBP-OP-1015 Event Notifications SRO-09 J. TS 3.6.2.1 Suppression Pool Average Temperature SRO-12 K. Steam Tables