Text

2013 Perry Initial License Examination Administered Written Exam
	ML13113A060
Person / Time
Site:	Perry
Issue date:	04/04/2013
From:	Bielby M; Operations Branch III
To:	; FirstEnergy Nuclear Operating Co
	Bielby M
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ML11354A271	List: ML13113A060; ML13178A368; ML13353A112; ML13353A123; ML13353A127;
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NRC Exam 2013 QUESTION RO 1 IAW NOP-OP-1002, Conduct of Operations, which of the following indicates the minimum Safe Shutdown staffing requirements during Mode 3?

Shift Manager Unit Supervisor Reactor Operator A. 1 0 1 B. 0 1 2 C. 1 1 2 D. 1 1 1

NRC Exam 2013 QUESTION RO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.5 Importance Rating 2.9*

K&A: Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Generic Explanation: Answer C - IAW NOP-OP-1002 Att 4, Safe Shutdown staffing for Modes 1, 2, & 3 is 1 SM, 1 US, & 2 ROs.

A - Incorrect - This is the staffing requirements for Modes 4 & 5 B - Incorrect - Plausible if candidate thinks this is Mode 4 or 5.

D - Incorrect - Plausible if candidate thinks this is Mode 4 or 5 and the US is a required position.

Technical Reference(s): NOP-OP-1002 Rev 7 Reference Attached: NOP-OP-1002 p 99 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-K Question Source: Bank # Perry Bank 40118 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 2 Which one of the following is a responsibility of the Reactor Operator during core alterations?

A. Monitor SRM count rate and period.

B. Authorize commencement of fuel movements C. Verify required refueling surveillances are current D. Ensure the Control Room fuel tag board is maintained current

NRC Exam 2013 QUESTION RO 2 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.44 Importance Rating 3.9 K&A: Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Generic Explanation: Answer A - Per IOI-9 monitor core reactivity.

B and D - Incorrect - These are the responsibility of the Unit Supervisor.

C - incorrect - This is the responsibility of the Refueling Supervisor and the Fuel Handling Supervisor.

Technical Reference(s): IOI-009 Rev. 30 Reference Attached: IOI-009 p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-12(LP)-E Question Source: Bank # Perry 2009 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 #RO-2 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 3 A dollar sign ($) step that does not meet the Acceptance Criteria in a Surveillance test constitutes a failure ____.

A. to adequately prevent preconditioning B. of the preventive maintenance program C. to comply with the applicable Tech Spec LCO D. to meet the requirements of the Surveillance Program

NRC Exam 2013 QUESTION RO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.12 Importance Rating 3.7 K&A: Knowledge of surveillance procedures.

Generic Explanation: Answer C - Per NOP-WM-2003, WM Surveillance Process, a ($) sign step is a surveillance requirement from Tech Specs. TS SR 3.0.1 states that failure to meet the surveillance is failure to meet the LCO.

A - Incorrect - Preconditioning is a term used to describe the act of operating a piece of equipment prior to performing the surveillance and may be selected by the candidate due to its association with surveillances.

B - Incorrect - Not necessarily impacted by the PM program but could be selected by the candidate due to possibly degraded equipment causing failure to meet acceptance criteria.

D - Incorrect - Failure to meet acceptance criteria does not impact the Surveillance Program but may be selected by the candidate due to its reference to surveillances. The Surveillance Program (per NOP-WM-2003) includes aspects of scheduling surveillances, updating databases, tracking surveillances, etc.

Technical Reference(s): NOP-WM-2003 Rev 7, PAP- Reference Attached: NOP-WM-2003 pp 26-0500 Rev 7 & TS SR 3.0.1 28, PAP-0500 p 77 & TS SR p 3.0-4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-04-T Question Source: Bank # Clinton 2007 Modified Bank #

New Question History: Previous NRC Exam Clinton 2007 # RO 19 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 4 The plant was operating at 85% power and 100% loadline, when following occurred:

APRM A failed upscale.

AFDL in Control alarm (ARI-H13-P680-004-E9) was received.

The immediate actions for AFDL in Control were completed.

The plant is currently stable with the following conditions:

JP LOOP TOT FLOW (Loop A) B33-R612A reading 26 Mlb/hr

JP LOOP TOT FLOW (Loop B) B33-R612B reading 37 Mlb/hr

TOTAL JP FLOW B33-R613(R) reading 63 Mlb/hr Which of the following Technical Specification LCO(s) if any are not being met?

A. only 3.4.2 Flow Control Valves.

B. only 3.4.1 Recirculation Loops Operating.

C. all Technical Specification LCO are being met.

D. 3.4.1 Recirculation Loops Operating and 3.4.2 Flow Control Valves.

NRC Exam 2013 QUESTION RO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.40 Importance Rating 3.4 K&A: Ability to apply Technical Specifications for a system.

Generic Explanation: Answer B - T.S. 3.4.1 requires recirculation flow mismatch to be 10% of rated core flow when operating at <70% of rated core flow. Per IOI-3 step 2.4 rated flow is 109.2 Mlbm/hr. Therefore, 70% of rated core flow is 76.3 Mlbm/hr and 10% is 10.92 Mlbm/hr. Since the mismatch is >10% TS 3.4.1 Conditions apply.

A, and D - Incorrect -Locking up the FCVs does not make them inoperable. It is a common misconception that when FCV are locked up they are Inoperable, locking up the FCVs is the requirement for TS 3.4.2 for Inoperable FCVs.

C - Incorrect - A mismatch of >10% flow is depicted with core flow <70% rated core flow. Therefore T.S.

3.4.1 applies.

Technical Reference(s): TS 3.4.1 and 3.4.2, IOI-3 Rev. Reference Attached: TS 3.4.1 pg 3.4.1 and 47 3.4.4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-08-B Question Source: Bank # Perry 2009 RO 11 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 5 You are performing a housekeeping walk-down and observe only a red ty-wrap on valve P43-F523A, NCC A HX DRAIN.

P43-F523A is in the open position NCC A heat exchanger is in dry layup.

Based on this information, Plant procedures require you to ____

A. close P43-F523A and remove the ty-wrap B. report your finding to the Work Week Manager C. notify the Control Room to ensure personnel safety D. review eSOMS to determine if P43-F523A has an active Danger tag

NRC Exam 2013 QUESTION RO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.14 Importance Rating 3.9 K&A: Knowledge of the process for controlling equipment configuration or status.

Generic Explanation: Answer C - IAW NOP-OP-1001 Clearance/Tagging Program, A Component Found With A Red Ty-Wrap Attached Is To Be Treated As A Red Tagged Component Until Proven Otherwise. Prompt notification of the Control Room is necessary for personnel safety.

A - Incorrect - Since The Valve Is Open And The Component Is To Be Treated As A Red Tagged Component, This Would Be A Violation Of The Tagging Program.

B - Incorrect - This is the wrong person to contact in this instance.

D - Incorrect - This is the responsibility of the Clearance Authority Technical Reference(s): NOP-OP-1001 Rev 19 Reference Attached: NOP-OP-1001 p 121 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3304-01-D.5 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 6 A discharge of Waste Sample Tank (WST) A is in progress in accordance with SVI-G50-T5266, Liquid Radwaste Release Permit Annunciator H13-P906-0001-A3, RW DISCH ISOL RADWASTE TO ESW PRCS RAD MON HI alarms.

Based on this information, ____.

A. an additional ESW pump will need to be started B. an additional Service Water pump will need to be started C. the RADWASTE HI FLW DISCH HDR FCV, G50-F153 will isolate automatically D. the RADWASTE HI FLW DISCH HDR FCV, G50-F153 will need to be manually isolated

NRC Exam 2013 QUESTION RO 6 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.5 Importance Rating 2.9 K&A: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Generic Explanation: Answer C - Per ARI-H13-P906-01-A3 the G50-F153 valve will automatically isolate on a high radiation condition on the Radwaste to ESW PRCS Rad Monitor.

A - Incorrect - misconception, Per ARI-H13-P970-01-A8 the G50-F153 valve will automatically isolate on a low discharge tunnel flow. The G50 SVI requires running an ESW pump during discharges to ensure no low flow condition exists during times of high cooling tower makeup.

B - Incorrect - misconception - Per ARI-H13-P970-01-A8 the G50-F153 valve will automatically isolate on a low discharge tunnel flow. The Subsequent Actions of the ARI is to start an additional pump if necessary.

D - Incorrect - The G50-F153 valve will automatically isolate - manual isolation not necessary.

Technical Reference(s): ARI-H13-P906-001 Rev 4 Reference Attached: ARI-H13-P906-001 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D17-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 b.11 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 7 You are required to make a Drywell Entry at 12% reactor power.

Since this is a considered a Very High Radiation Area, you must obtain ____.

1. RP Manager written approval

2. Operations Unit Supervisor approval

3. Operations Shift Manager approval

4. Director of Site Operations approval A. 1, 2, & 3 B. 2, 3, & 4 C. 1, 2, & 4 D. 1, 3, & 4

NRC Exam 2013 QUESTION RO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.13 Importance Rating 3.4 K&A: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Generic Explanation: Answer D - RPM, SM and DSO approval is required for LHRA entry IAW NOP-OP-4101 A, B, & C - Incorrect - The Unit Supervisors approval is not required.

Technical Reference(s): NOP-OP-4101 Rev 8 Reference Attached: NOP-OP-4101 p 18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-03-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 b.12 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 8 Which of the following require notification to the Federal Aviation Administration (FAA)?

1) Unit 2 cooling top red flashing beacons burned out

2) Seven days since last notification to FAA

3) TEC helicopter pad lights out

4) Met tower upper light out

5) Unit 1 cooling tower middle red steady beacons off

6) Microwave tower flashing lights off A. 1, 2, & 5 B. 1, 4, & 6 C. 2, 3, & 5 D. 3, 4, & 6

NRC Exam 2013 QUESTION RO 8 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.30 Importance Rating 2.7 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.

Generic Explanation: Answer B - IAW OAI-0201, notification to the FAA is required for items 1, 4 & 6 only.

A - Incorrect - Notifications must be made every 15 days. And middle steady burning light failure need not be reported.

C - Incorrect - Notifications must be made every 15 days. The TEC helicopter pad lights are not covered.

And middle steady burning light failure need not be reported.

D - The TEC helicopter pad lights are not covered.

Technical Reference(s): OAI-0201 Rev 31, PDB-C08 Reference Attached: OAI-0201 pp 57-58 &

Rev 1, & PERS PDB-C08 p 1 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01-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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 9 The plant is operating at 80% power when the following occurs:

Annunciator H13-P680-08-B6, LOAD SET RUNBACK STATOR CLG alarms The condition that caused this alarm was ____.

A. Stator cooling water inlet flow to main generator at 530 gpm B. Stator cooling water inlet temperature element fails to 85°C C. Stator cooling water outlet from main generator temperature is 76°C D. Stator cooling water inlet pressure to the main generator at 40 psig

NRC Exam 2013 QUESTION RO 9 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.46 Importance Rating 4.2 K&A: Ability to verify that the alarms are consistent with the plant conditions.

Generic Explanation: Answer D - SWC inlet pressure to the main generator < 42.5 will cause this alarm.

A - Incorrect - The low inlet water flow (<531 gpm) causes H2 Seal/STATOR CLG TRBL alarm not the LOAD SET RUNBACK STATOR CLG alarm B - Incorrect - The high inlet water temperature causes H2 Seal/STATOR CLG TRBL alarm not the LOAD SET RUNBACK STATOR CLG alarm C - Incorrect - The inlet temperature setpoint is >81°C not the outlet temperature. The outlet temp gives local alarm causing causes H2 Seal/STATOR CLG TRBL alarm not the LOAD SET RUNBACK STATOR CLG alarm Technical Reference(s): ARI-H13-P680-08 Rev 13 Reference Attached: ARI-H13-P680-08 p 21 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N43-I.1 Question Source: Bank #

Modified Bank # LaSalle 2006 New Question History: Previous NRC Exam LaSalle 2006 # RO-74 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 10 You are the Field Supervisor.

At 09:58 a Chemistry Tech reported an explosion in the Chemistry Lab.

At 10:03 the Security Shift Supervisor notified the Shift Manager that the explosion in the Chemistry Lab was a result of Hostile Action within the plant

At 10:09 the Shift Manager declared a Site Area Emergency

The Shift Manager has called for activation of the OSC at the alternate location.

You must report to __(1)__.

The latest time the OSC should be declared Operational to meet the goal is __(2)__.

__(1)__ __(2)__

A. TSC @ SB-603 10:58 B. TSC @ SB-603 11:09 C. Unit 2 Control Room 10:58 D. Unit 2 Control Room 11:09

NRC Exam 2013 QUESTION RO 10 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.42 Importance Rating 2.6 K&A: Knowledge of emergency response facilities.

Generic Explanation: Answer D - The alternate location for the OSC is the Unit 2 control room for people on site.

Sixty minutes from the time the event is declared is the goal for declaring the OSC Operational.

A - Incorrect - Wrong location for the alternate OSC and incorrect time.

B - Incorrect - Wrong location for the alternate OSC C - Incorrect - Wrong time for meeting the goal Technical Reference(s): EPI-A7 Rev 21 Reference Attached: EPI-A7 p 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): EPL-0804-01 & EPL-0815-01-4 & -6 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 11 The plant was operating at 100% power when reactor power unexpectedly decreased.

ONI-C51, Unplanned Change in Reactor Power or Reactivity, has been entered.

Plant conditions have stabilized with reactor power at 90%.

The following changes in Recirculation System parameters occurred:

Total Core Flow has decreased

Core Plate d/p has decreased

Loop A & B Flow has slightly increased

Loop A Total Jet Pump Flow has decreased

Loop B Total Jet Pump Flow has increased Which of the following has occurred based on these plant conditions?

A. A Jet Pump Riser in Loop A has failed.

B. Flow Control Valve A has drifted closed.

C. A vortexing event in Loop A has occurred.

D. Loop A Discharge Valve has drifted closed.

NRC Exam 2013 QUESTION RO 11 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295001 AK3.02 Importance Rating 3.7 K&A: Knowledge of the reasons for the following responses as they apply to Partial Or Complete Loss Of Forced Core Flow Circulation: Reactor power response Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Explanation: Answer A - The reason reactor power decreased is that recirc flow lowered on the A side as a result of a failed JP riser.

B - Incorrect - If this had occurred, Loop A flow would have decreased, not increased slightly.

C - Incorrect - If this had occurred, reactor power would have increased and return to the pre-transient value, not decreased and stabilized.

D - Incorrect -If this had occurred, Loop A flow would have decreased and recirc pump would have tripped, not increased slightly Technical Reference(s): ONI-C51 Rev J & SEN-105 Reference Attached: ONI-C51 - partial Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-C.9 Question Source: Bank # Perry 2005 Modified Bank #

New Question History: Previous NRC Exam Perry 2005 # RO-75 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 12 The plant is operating at rated power with the following electrical lineup:

Buses EH11 and EH13 are on their Preferred Source.

Bus EH12 is on its Alternate Preferred Source.

Annunciator 2H13-P870-01-D4, INTERBUS XFMR LH-2-A LOCKOUT RELAY alarmed.

What is the status of the Emergency Diesel Generators (EDGs) two minutes after INTERBUS XFMR LH-2-A LOCKOUT RELAY alarm is received?

A. All three EDGs are running and loaded.

B. All three EDGs remain in standby status.

C. Division 2 EDG is running loaded and Division 1 and 3 EDGs remain in standby status.

D. Division 2 EDG remains in standby status and Division 1 and 3 EDGs are running loaded.

NRC Exam 2013 QUESTION RO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295003 AA1.02 Importance Rating 4.2*

K&A: Knowledge of the operational implications of the following concepts as they apply to partial or complete loss of A.C. Power: Emergency generators Partial or Complete Loss of AC / 6 Explanation: Answer C - Tie bus TH21 has lost power from LH2A, so Division 2 DG starts and powers EH12.

A - Incorrect - Plausible if loss of both LH1A & LH2A B - Incorrect - Div 2 DG will auto start on loss of power to EH12 D - Incorrect - Plausible if loss of tie bus TH11 Technical Reference(s): ARI-2H13-P870-01 Rev 6 Reference Attached: ARI-2H13-P870-01 p 39 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R10-C.8 & OT-COMBINED-R48_48-F.10, 11, & 12 Question Source: Bank # Perry 2007 Modified Bank #

New Question History: Previous NRC Exam Perry 2007 # RO-12 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 13 The plant is operating in EOP-1A, Level Power Control with the following conditions:

RHR A pump is operating in Suppression Pool Cooling

A loss of bus ED-1-A occurs

RHR B & C pumps have tripped and can not be restarted

The US determined Emergency Depressurization is required Based on this information, you would perform Emergency Depressurization by ____.

A. arming and depressing both ADS A pushbuttons on P601 B. arming and depressing both ADS B pushbuttons on P601 C. individually operating the ADS SRV control switches on P601 D. individually operating the ADS SRV control switches on P631

NRC Exam 2013 QUESTION RO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295004 AA2.04 Importance Rating 3.2 K&A: Ability to determine and/or interpret the following as they apply to Partial Or Complete Loss Of D.C. Power: System lineups Partial or Total Loss of DC Pwr / 6 Explanation: Answer D - Bus ED-1A powers the ADS A solenoids. Since the A solenoids have no power, and no RHR pumps are running on the B side, the SRVs must be operated individually from P631.

A - Incorrect - plausible; even though a RHR pump is running, the A solenoids have no power.

B - Incorrect - plausible; even though the B solenoids have power no RHR pump is running, C - Incorrect - plausible; if misconception about which solenoids are powered from which bus Technical Reference(s): ONI-B21-1 Rev 11, ELI-R42 Reference Attached: ONI-B21-1 p 12, ELI-Rev 8, Dwgs 208-011 Sh 4 Rev M. Sheet 5 Rev J, R42 pp 3-4, Dwgs 208-011 Sh 4. (partial)

Sheet 5 (partial)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21C-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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 14 The reactor is operating at 40% power.

A Main Turbine trip occurs due to actuation of the Main Generator lockout relay.

The reactor does not automatically scram.

Which one of the following describes the Reactor Protection System response in accordance with Technical Specifications 3.3.1.1, RPS Instrumentation?

A. RPS Instrumentation is OPERABLE; no Required Action(s) need to be completed.

B. RPS Instrumentation Turbine Stop Valve Closure trip only is INOPERABLE, Required Action(s) need to be completed.

C. RPS Instrumentation Turbine Stop Valve Closure and Turbine Control Valve Fast Closure trips are INOPERABLE; Required Action(s) need to be completed.

D. RPS Instrumentation Turbine Stop Valve Closure and Turbine Control Valve Fast Closure trips are not required to be OPERABLE, no Required Action(s) need to be completed.

NRC Exam 2013 QUESTION RO 14 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295005 2.2.37 Importance Rating 3.6 K&A: Ability to determine operability and/or availability of safety related equipment.

Main Turbine Generator Trip / 3 Explanation: Answer C - Since reactor power is >38%, both stop valve and control valve scram signals should have been generated. Because both trips did not occur, they are both inop.

A - Incorrect - RPS is not operable with trips not occurring at 40% power.

B - Incorrect - TCV trip also needs to be operable at >30% RTP D - Incorrect - The Applicability for TSV & TCV trips is 38% RTP. These are required to be operable at

>40% RTP.

Technical Reference(s): TS 3.3.1.1-1 Table Reference Attached: TS 3.3.1.1-1 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-07-F Question Source: Bank # Perry Audit 2003 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 b.7 55.43 Comments: Level of Difficulty = x - The TS table is part of the Applicability statement.

NRC Exam 2013 QUESTION RO 15 The plant was operating at 50% power when the following occurred:

The Mode Switch was placed in Shutdown

All Control Rods inserted except one, which is stuck fully withdrawn

RPV level lowered to 140 inches then recovered using feedwater

SRMs and IRMs have been inserted Shutdown criteria is __(1)__. Entry into __(2)__ is required.

__(1)__ __(2)__

A. met EOP-1A, Level Power Control B. met ONI-C71-1, Reactor Scram C. not met EOP-1A, Level Power Control D. not met ONI-C71-1, Reactor Scram

NRC Exam 2013 QUESTION RO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295006 AK1.03 Importance Rating 3.7 K&A: Knowledge of the operational implications of the following concepts as they apply to SCRAM: Reactivity control SCRAM / 1 Explanation: Answer B - IAW EOP Bases, one rod out meets the criteria for SHUTDOWN. Since IRMs and SRMs are inserted, Rx power is not unknown. Entry into ONI-C71-1 is required.

A - Incorrect - Entry into EOP-1A is not required.

C - Incorrect - Shutdown criteria is net D- Incorrect - Shutdown criteria is met and entry into EOP-1A is not required.

Technical Reference(s): EOP Bases Rev 3 , ONI-C71-1 Reference Attached: EOP Bases pp 46-47 &

Rev 16, EOP-1 Chart Rev D, & EOP-1A Chart Rev D ONI-C71-1 pp 3 & 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-01(LP)-A.1 & OT-3402-11-B Question Source: Bank # Clinton 2002 Modified Bank #

New Question History: Previous NRC Exam Clinton 2002 # RO-46 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 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 you to perform Control Room Isolation per IOI-11.

You placed the Remote Shutdown switches for valves P57-F015A, CNTMT ADS SUPPLY OTBD ISOL VALVE, and P57-F020A, DW ADS SUPPLY OTBD ISOL VALVE, to EMERGENCY.

After placing the Remote Shutdown Switches in EMERGENCY, the position of P57-F015A is __(1)__ and the position of P57-F020A is __(2)__?

__(1)_ __(2)__

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

NRC Exam 2013 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 / 7 Explanation: Answer A - Transferring P57-F015 & F020 locally, results in the valves being opened even if they were previously closed.

B, C, D - Incorrect - Transferring both switches on each MCC causes each valve to open.

Technical Reference(s): SVI-P57-T2001 Rev 5, SVI- Reference Attached: SVI-P57-T2001 p 3, P57-T2003 Rev 3, IOI-011 Rev 26 & Dwg 208-199-01 SVI-P57-T2003 pp 4 & 6, IOI-011 p 120 &

Rev N & SH-03 Rev R Dwg 208-199-01 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 # RQL-0043 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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 17 The plant was operating at rated power when a complete loss of Nuclear Closed Cooling Water occurred.

The reason the reactor must be scrammed expeditiously because the ____.

A. operating CRD pump will automatically trip B. Reactor Recirculation pumps must be secured IAW SOI-B33 C. Reactor Recirculation Hydraulic Power Units automatically trip D. loss of NCC will result in Drywell temperature exceeding the scram setpoint

NRC Exam 2013 QUESTION RO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295018 AK3.02 Importance Rating 3.3 K&A: Knowledge of the reasons for the following responses as they apply to Partial Or Complete Loss Of Component Cooling Water: Reactor power reduction Partial or Total Loss of CCW / 8 Explanation: Answer B - IAW ONI-P43, the plant is scrammed in anticipation of loss of cooling to Recirc Pumps. Following the scram, the recirc pumps are secured.

A - Incorrect - The CRD pump will trip on high temperature after a period of time not automatically on a loss of NCC.

C - Incorrect - Misconception that HPUs are cooled by NCC.

D - Incorrect - Misconception that the Rx will scram on high DW temperature vs. pressure. The loss of NCC will result in the loss of DW cooling and the slow rise in DW temperature and pressure.

Technical Reference(s): ONI-P43 Rev 11 Reference Attached: ONI-P43 pp 4 and 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-16(LP)-A.3 & OT-COMBINED-B33-N.1 Question Source: Bank # River Bend 2007 Modified Bank #

New Question History: Previous NRC Exam River Bend 2007 #RO-7 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 18 The plant is operating at 100% power. The Safety Related Instrument Air Compressor is out of service due to a failed motor.

Air pressures as indicated on ADS AIR STRG PRESS, 1P57-R026A and 1P57-R026B, are 155 psig and slowly lowering.

The if ADS air pressure continues to lower, the __(1)__ MSIVs will be affected. Restore air pressure using __(2)__.

__(1)__ __(2)__

A. inboard Instrument Air (P52)

B. inboard portable air cylinders C. outboard Instrument Air (P52)

D. outboard portable air cylinders

NRC Exam 2013 QUESTION RO 18 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295019 AA1.01 Importance Rating 3.5 K&A: Ability to operate and/or monitor the following as they apply to Partial Or Complete Loss Of Instrument Air: Backup air supply Partial or Total Loss of Inst. Air / 8 Explanation: Answer D - IAW SOI-P57 P&L 2.3, if ADS B air pressure lowers to 90 psig, TS 3.6.1.3 needs to be entered. At <45 psig, the MSIVs may not be leak tight following a DBA LOCA.

A - Incorrect - Inboard MSIVs are not affected by lowering ADS pressure. P57 is normally 160 psig.

Instrument air pressure is normally only ~ 125 psig.

B - Incorrect - Inboard MSIVs are not affected by lowering ADS pressure.

C - Incorrect - P57 is normally 160 psig. Instrument air pressure is normally only ~ 125 psig..

Technical Reference(s): SOI-P57 Rev 16, ARI-H13- Reference Attached: SOI-P57 pp 3 & 13, ARI-P601-019 rev 14 H13-P601-019 p 130 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P57-M Question Source: Bank #

Modified Bank # Perry 2009 New Question History: Previous NRC Exam Perry 2009 #RO-18 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 19 The plant is in a refuel outage with a full core offload in progress with the following conditions:

The RHR A System is providing Shutdown Cooling

The RHR B System is lined up for Fuel Pool Cooling Assist.

Temperatures in the vessel and the spent fuel pools are stable.

A loss of RPS B then occurs.

The reactor coolant temperature will __(1)__. The Spent Fuel Pool temperature will __(2)__.

__(1)__ __(2)__

A. rise rise B. rise remain fairly constant C. remain fairly constant remain fairly constant D. remain fairly constant rise

NRC Exam 2013 QUESTION RO 19 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295021 AK2.05 Importance Rating 2.7 K&A: Knowledge of the interrelations between Loss Of Shutdown Cooling and the following: Fuel pool cooling and cleanup system Loss of Shutdown Cooling / 4 Explanation: Answer B - A loss of either RPS bus will cause the SDC suction valves to isolate. This will cause a rise in Rx coolant temperature. However, a loss of RPS does not affect the Fuel Pool Cooling Assist mode of RHR and the spent fuel pool temperatures will remain fairly constant.

A - Incorrect - cooling is not lost to SFP - temperatures will remain stable.

C - Incorrect - The loss of RPS will cause a loss of SDC.

D - Incorrect - The loss of RPS will cause a loss of SDC and no loss of cooling to the SFP.

Technical Reference(s): SOI-E12 Rev 57 & ONI-C71-2 Reference Attached: SOI-E12 pp 50-53 &

Rev 8 ONI-C71-2 pp 9 & 12 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-03(LP)-A.1 Question Source: Bank # Columbia 2003 Modified Bank #

New Question History: Previous NRC Exam Columbia 2003 #RO-56 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 20 The plant is in a refuel outage with fuel shuffle in progress.

The Refuel SRO informs the control room that bubbles are seen rising from a fuel bundle after it was bumped against the RPV.

Area Radiation Monitor 1D21-K083, Upper Pool Area indicates a HIGH alarm.

1D21-K083 has a red placard with white lettering affixed to it.

This placard identifies that an alarm on this monitor may be a potential ____ entry condition.

A. Emergency Plan (E-Plan)

B. Off-Normal Instruction (ONI)

C. Emergency Operating Procedure (EOP)

D. Offsite Dose Calculation Manual (ODCM)

NRC Exam 2013 QUESTION RO 20 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295023 2.4.41 Importance Rating 2.9 K&A: Knowledge of the emergency action level thresholds and classifications.

Refueling Acc / 8 Explanation: Answer A - IAW PAP-0524, a red background placards with white lettering are used to identify an E-Plan entry condition.

B - Incorrect - While an alarm on this rad monitor may trigger an ONI entry, there are no placards that indicate an ONI entry condition.

C - Incorrect - EOP entry condition placards are orange with white letters D - Incorrect - Plausible since this is a radiation monitor. However, there are no placards in the control room that indicate entry into the ODCM Technical Reference(s): PAP-0524 Rev 10 Reference Attached: PAP-0524 pp 13-14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 21 The maximum internal design pressure of the drywell is __(1)__ and is based on__(2)__.

__(1)__ __(2)__

A. 21 psig a large break LOCA inside drywell B. 21 psig condensing steam in containment following a LOCA C. 30 psig a large break LOCA inside drywell D. 30 psig condensing steam in containment following a LOCA

NRC Exam 2013 QUESTION RO 21 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295024 EK1.01 Importance Rating 4.1 K&A: Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity; Plant-Specific High Drywell Pressure / 5 Explanation: Answer C - The USAR design pressure of the drywell is 30 psig based on a large break LOCA.

A - Incorrect - 21 psid is the external to internal differential pressure limit. And condensing steam following a LOCA is the bases for the 21 psid.

B - Incorrect - 21 psid is the external to internal differential pressure limit.

D - Incorrect - Condensing steam following a LOCA is the bases for the 21 psid not the 30 psig Technical Reference(s): SDM-T23 Rev 12, USAR Table Reference Attached: SDM-T23 p 5, USAR 6.2-1 Rev 14, USAR Table 6.2-6 Rev 12, & USAR Table 6.2-1 p 154, USAR Table 6.2-6 p 161, 6.2.7.3.3 Rev 12 & USAR p 6.2-139 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-T23-D Question Source: Bank # River Bend 2003 Modified Bank #

New Question History: Previous NRC Exam River Bend 2003 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.9 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 22 The following plant conditions exist:

Reactor scrammed on high reactor pressure

MSIVs isolated

RPV water level 230 inches

RPV pressure 900 psig

Suppression Pool temperature 105°F

Suppression Pool level 22 ft Lowering ____ would challenge the margin to the Heat Capacity Limit (HCL).

A. RPV pressure to 700 psig B. RPV water level to 200 inches C. Suppression Pool temperature to 90°F D. Suppression Pool water level to 18.0 feet

NRC Exam 2013 QUESTION RO 22 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 EK2.06 Importance Rating 3.5 K&A: Knowledge of the interrelations between Suppression Pool High Water Temperature and the following: Suppression pool level Suppression Pool High Water Temp. / 5 Explanation: Answer D - Lowering SP level reduces the margin to HCL.

A & C - Incorrect - These actions raise the margin to HCL B - Incorrect - this has no effect on HCL Technical Reference(s): EOP-SPI Supplement Rev 3 Reference Attached: EOP-SPI Supplement p 8

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3432-06-C.1.B Question Source: Bank # Perry 2003 Modified Bank #

New Question History: Previous NRC Exam Perry 2003 #RO-27 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 23 The following conditions exist:

The reactor was scrammed from rated power

An ATWS is in progress

Suppression Pool level is 15 feet 5 inches lowering 1 inch per minute

RPV level is - 8 and lowering 1 inch per minute

Drywell temperature is 280°F and rising 2°F per minute

Containment temperature is 150°F and rising 3°F per minute Per the EOP Bases, without anticipatory action, Emergency Depressurization will first be required based on ____.

A. RPV level B. Drywell temperature C. Suppression Pool level D. Containment temperature

NRC Exam 2013 QUESTION RO 23 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 / 5 Explanation: Answer D - ED is required prior to exceeding 185°F in containment. Based on a 150°F initial temp and rising 3°F per minute, it will be approximately 11 1/2 minutes until ED is required.

A - Incorrect - Based on initial RPV level of -8 inches and lowering 1 inch per min, it will be approximately 17 minutes before ED can be performed.

B - Incorrect - Based on initial DW temp of 280°F and rising 2°F per minute, it will be approximately 25 minutes before ED is required C - Incorrect - Based on initial SP level of 15 feet 5 inches lowering 1 inch per minute, it will be approximately 14 minutes before ED is required.

Technical Reference(s): EOP-2 Chart (Partial) Rev B & Reference Attached: EOP-2 Chart (Partial) &

EOP Bases Rev 3 EOP Bases p 38 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 24 The plant is operating in EOP-2 Primary Containment Control. The following alarms have been received:

DRYWELL AVERAGE TEMP A HI - H13-P601-20-F3

DRYWELL AVERAGE TEMP B HI - H13-P601-17-F5 The Unit Supervisor has directed the BOP Operator to operate all available Drywell cooling.

Then electrical power is lost to the solenoid for the LW DW CLG COOLER SEL VLV, P43-F025 (NCC Supply Valve) for the Lower DW Cooler.

The Drywell Ventilation system will respond to the loss of electrical power to the solenoid as follows:

A. The LW DW CLG COOLER SEL VLV fails to the A cooling coil B. The LW DW CLG COOLER SEL VLV fails to the B cooling coil C. The LW DW CLG TEMP TEMP CONT Valve, P43-F365 automatically opens D. The LW DW CLG COOLER SEL VLV fails shut isolating NCC to the DW Cooler

NRC Exam 2013 QUESTION RO 24 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295028 EA1.02 Importance Rating 3.9 K&A: Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell ventilation system High Drywell Temperature / 5 Explanation: Answer B - On a loss of electrical power to the 3-way selector valve, it will fail to the B cooling coil position.

A - Incorrect - Loss of electrical power or instrument air causes the valve to fail to the B coil.

C - Incorrect - The temp control valve is a MOV and is not affected by loss of power to the selector valve.

D - Incorrect - The selector valve is a 3-way valve that fails to the B position, not closed.

Technical Reference(s): OT-Combined-M13 lesson plan Reference Attached: OT-Combined-M13 Rev 3 & PDB-H027 Rev 0 lesson plan p 13 & PDB-H027 p 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x Question Source: Bank #

Modified Bank # Perry 2009 New Question History: Previous NRC Exam Perry 2009 # RO-25 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 25 Plant Conditions are as follows:

Reactor Power 0%, with 2 rods at position 12

Reactor pressure 900 psig

Reactor water level 210

Suppression Pool temperature 100°F

Suppression Pool level 14.0 feet

Drywell pressure 2.5 psig

Containment pressure 2.0 psig

Containment temperature 110°F What action is required to be performed?

A. Inject Boron B. Emergency Depressurize C. Commence Controlled Cooldown D. Anticipate Emergency Depressurization

NRC Exam 2013 QUESTION RO 25 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295030 EA2.01 Importance Rating 4.1*

K&A: Ability to determine and/or interpret the following as they apply to Low Suppression Pool Water Level: Suppression pool level Low Suppression Pool Wtr Lvl / 5 Explanation: Answer B - At 14.25 SP level ED is required.

A - Incorrect - Boron injection required with SP temp 110°F or >4% power C & D - Incorrect - Can not Cooldown or anticipate Emergency Depressurization in an ATWS ( 2 rods at position 12)

Technical Reference(s): EOP-2 Bases Rev 1 Reference Attached: EOP-2 Bases p 37 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-05-C.1 Question Source: Bank # Perry 2007-2 Modified Bank #

New Question History: Previous NRC Exam Perry 2007-2 #RO-16 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 26 The plant is operating at rated power with the following conditions:

At 09:15, the Unit Supervisor declares, 1B21-N0673C (RPV LVL 2) INOP so Maintenance can work on the transmitter.

At 09:45, 1B21-N0673L (RPV LVL 2) fails upscale and the Shift Engineer determines that HPCS initiation capability has been lost

At 09:50, a loss of Feed Water occurs causing RPV level to lower

At 09:51, the HPCS Manual Initiation pushbutton is armed and depressed The HPCS system must be declared Inoperable by __(1)__.

When the Manual Initiation pushbutton is armed and depressed, HPCS __(2)__ inject.

__(1)__ __(2)__

A. 10:15 will B. 10:15 will not C. 10:45 will D. 10:45 will not

NRC Exam 2013 QUESTION RO 26 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295031 2.2.36 Importance Rating 3.1 K&A: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Reactor Low Water Level / 2 Explanation: Answer C - When the determination is made that HPCS initiation capability has been lost, the RA is to declare HPCS system INOP in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Manual initiation will work regardless of auto initiation capability.

A - Incorrect - One hour is allowed to declare HPCS System INOP.

B - Incorrect - One hour is allowed to declare HPCS System INOP. HPCS will inject. The Manual initiation will work regardless of auto initiation capability. However, this is plausible because the SPMU system (Arm & Depress) requires a back-panel key manipulation if a RHR LOCA signal is not present.

D - Incorrect - HPCS will inject. The Manual initiation will work regardless of auto initiation capability.

However, this is plausible because the SPMU system (Arm & Depress) requires a back-panel key manipulation if a RHR LOCA signal is not present.

Technical Reference(s): TS 3.3.5.1, DWG 208-065 Sh Reference Attached: TS 3.3.5.1 pp 3.3-32, 03 Rev P, PDB-I04 Rev 13, & PDB-I05 Rev 9 33, & 41, DWG 208-065 Sh 03 Partial, &

PDB-I04 p11, & PDB-I05 p 12 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-07-F and OT-COMBINED-E22A-K.1 & F.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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 27 The plant was operating at 75% rated power when the following occurred:

The Rx was manually scrammed

Little rod motion occurred

The BOP operator initiated Standby Liquid Control

SLC pump B is tagged out

SLC Storage tank level is 4860 gallons per SPDS

SLC Storage tank boron concentration is 2.85%

The minimum amount of time to inject the Cold Shutdown Weight of Boron into the reactor is approximately ____ minutes.

A. 54 B. 57 C. 108 D. 113

NRC Exam 2013 QUESTION RO 27 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295037 EK1.05 Importance Rating 3.4 K&A: Knowledge of the operational implications of the following concepts as they apply to Scram Condition Present And Reactor Power Above APRM Downscale Or Unknown:

Cold shutdown boron weight; Plant-Specific SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 Explanation: Answer C - The Cold S/D Weight of Boron is defined in the EOP Bases as the contents of the SLC Storage tank. 202 gallons must be subtracted from the SPDS reading of 4860 gallons to get the usable volume. Since 1 SLC pump is tagged out it will take ~108 minutes @ 43 gpm (the capacity of 1 SLC pump) to inject the volume of the SLC tank A - Incorrect - this is the time for 2 SLC pumps injecting.

B - Incorrect - This is the time for 2 SLC pumps if the 202 gallons is not subtracted from the SPDS value D - Incorrect - This is the time if the 202 gallons is not subtracted from the SPDS value.

Technical Reference(s): EOP-Bases Rev 3, LP-OT- Reference Attached: EOP-Bases pp 47 & 58, COMBINED-C41 Rev 1, & PYBP-POS-0030 Rev 1 LP-OT-COMBINED-C41 p 10, & PYBP-POS-0030 p11 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C41 OT-3402-03-D.2 Question Source: Bank #

Modified Bank # Vermont Yankee 2005 New Question History: Previous NRC Exam Vermont Yankee 2005 #RO-18 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.8 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 28 The plant was operating at 100% power with Annulus Exhaust Gas Treatment System Fan B in operation.

The following conditions are present:

A steam leak in the annulus

ALERT and HIGH alarms on ANN EXH B GAS Radiation Monitor, 1D17-K697B

The Shift Manager has declared an ALERT (HA-1) based on dose rate at the site boundary Entry into EOP(s) __(1)__ is required.

Monitor the __(2)__ Plant Vent Radiation Monitor to track release rate.

__(1)__ __(2)__

A. Secondary Containment Control, EOP-03 only Unit 1 B. Secondary Containment Control, EOP-03 and Unit 1 Radioactivity Release Control, EOP-05 C. Secondary Containment Control, EOP-03 only Unit 2 D. Secondary Containment Control, EOP-03 and Unit 2 Radioactivity Release Control EOP-05

NRC Exam 2013 QUESTION RO 28 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295038 2.4.1 Importance Rating 4.6 K&A: Knowledge of EOP entry conditions and immediate action steps.

High Off-site Release Rate / 9 Explanation: Answer D - Since the SM declared the Alert on Dose at site boundary entry into EOP-5 is required. With AEGTS rad monitor in alarm (HIGH) entry into EOP-3 is required. With AEGTS B fan in operation, the Unit 2 plant vent is the correct release point.

A - Incorrect - EOP-5 needs to be entered, too and incorrect plant vent.

B - Incorrect - Incorrect plant vent.

C - Incorrect - EOP-5 needs to be entered, too.

Technical Reference(s): EOP-03 & EOP 5 chart Rev C Reference Attached: EOP-03 & EOP 5 chart 912-0605 Rev W and EPI-A1 rev 25 partial, 912-0605 Partial, and EPI-A1 p 48 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-15-B & 17-B and OT-COMBINED-M15-B Question Source: Bank # Perry 2009 Modified Bank #

New Question History: Previous NRC Exam Perry 2009 #RO-29 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 29 Overall command responsibility for a fire in the Water Treatment Building resides with __(1)__.

Overall command responsibility for a fire in the Owner Controlled Area that affects Plant safety resides with __(2)__.

__(1)__ __(2)__

A. Fire Brigade Leader Fire Brigade Leader B. Fire Brigade Leader Responding Off-Site Fire Department C. Responding Off-Site Fire Department Fire Brigade Leader D. Responding Off-Site Fire Department Responding Off-Site Fire Department

NRC Exam 2013 QUESTION RO 29 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 600000 AK1.02 Importance Rating 2.9 K&A: Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Fighting Plant Fire On Site / 8 Explanation: Answer A - IAW ONI-P54, the FBL is responsible for fire fighting in the protected area. For fires in the OCA, the FBL is responsible for fires that affect plant safety.

B - Incorrect - The responding offsite fire department is responsible for fires in the OCA that do not affect plant safety or operability are not affected.

C - Incorrect - FBL is responsible for fire fighting in the protected area.

D - Incorrect - FBL is responsible for fire fighting in the protected area. The responding offsite fire department is responsible for fires in the OCA that do not affect plant safety or operability are not affected.

Technical Reference(s): ONI-P54 Rev 16 Reference Attached: ONI-P54 pp 23 & 38 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-05(LP)-A.9 Question Source: Bank # Perry Audit 2009 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

NRC Exam 2013 QUESTION RO 30 The plant is operating at rated power with the following conditions:

The monthly load test SVI for Division 1 DG is in progress with DG load at 5800 KW

Division 2 and Division 3 DGs are in Standby with respective buses lined up to the Preferred Source Subsequently, the following conditions developed:

ONI-S11, Hi/Low Voltage, was entered due to Degraded Grid condition

Annunciator VOLTS TO HERTZ RATIO HI, H13-P680-0009-B2 just alarmed

Main Generator frequency is at 58 hz

Main Generator terminal voltage is at 24,000 volts What is the consequence if this alarm does not clear within a minute?

A. The Main Generator Voltage Regulator shifts to manual only B. The reactor will automatically scram on a main turbine trip C. The breakers feeding Buses EH12 and EH13 will trip on under-frequency.

D. Breaker Div 1 DG output breaker EH11-02 will trip on under-frequency

NRC Exam 2013 QUESTION RO 30 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 700000 AK2.06 Importance Rating 3.9 K&A: Knowledge of the interrelations between Generator Voltage And Electric Grid Disturbances and the following: Reactor power Generator Voltage and Electric Grid Disturbances / 6 Explanation: Answer B - The turbine will trip due to a generator trip. With the plant at rated power a turbine trip will cause a reactor scram.

A - Incorrect - The voltage regulator will shift to manual, but that is not the only thing that will happen C - Incorrect - Plausible because the Preferred Source breakers would trip at 59 hz if DG was running in parallel to the bus. With the DGs not running in parallel to the buses, the under-freq trips are not active.

D - Incorrect -. Plausible because the Preferred Source or Alt-Preferred Source breakers would trip at 59 hz with DG running in parallel to the bus. During the monthly load test, the DG runs in parallel to the Preferred or Alt-Preferred sources.

Technical Reference(s): ARI-H13-P680 -09 Rev 11 & Reference Attached: ARI-H13-P680-09 p 17 ONI-S11 Rev 9 & ONI-S11 p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N41_N51-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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 31 The plant was operating at rated power with both Reactor Recirculation System FCVs indicating 73% open when the following occurred:

RFPT A tripped and RPV level momentarily lowered to 185 inches before being restored to normal.

Five minutes later, RFPT B tripped and RPV level lowered to 140 inches before being restored to normal with the Motor Feed Pump.

Currently, the Reactor Recirculation Pumps are in __(1)__ and the Flow Control Valves are at __(2)__ open.

__(1)__ __(2)__

A. Slow Speed 17%

B. Slow Speed 48%

C. OFF 17%

D. OFF 48%

NRC Exam 2013 QUESTION RO 31 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295009 AK3.01 Importance Rating 3.2 K&A: Knowledge of the reasons for the following responses as they apply to Low Reactor Water Level: Recirculation pump run back; Plant-Specific Low Reactor Water Level / 2 Explanation: Answer A - The reason the FCVs runback is a signal is generated when RPV level goes below L4 concurrent with a trip of the first RFPT in order to lower Rx power. Without operator action, the runback can not be reset. Then, the Recirc Pumps automatically shift to Slow speed due to RPV level lowering to 177 (L3) after the scram or FW flow < 3.43 MLBM/hr for 15 seconds. They do not trip to Off until an RPV level of +130 (L2).

B - Incorrect - This is the flow rate not the valve position.

C - Incorrect - Pumps will not trip to off until level lowers to L2.

D - Incorrect - Pumps will not trip to off until level lowers to L2 and, 48% is the flow rate not the valve position.

Technical Reference(s): SDM-B33 Rev 9 and ARI-H13- Reference Attached: SDM-B33 pp 27 & 44 P680-04 Rev 18 and ARI-H13-P680-04 pp 9 & 47 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-E.2 & E.6 Question Source: Bank # INL-2376 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 32 At 09:00 The plant was manually scrammed due to a small steam leak in the drywell. And the following DW parameters were noted:

DW Temperature 140°F

DW Pressure 1.5 psig At 09:10 the A Upper DW Cooling Fan M13-C003A tripped. And the following DW parameters were noted:

DW Temperature 140°F

DW Pressure 1.8 psig Without taking operator actions, the B Upper DW Cooling Fan, M13-C003B ____.

A. will start and DW temperature can be maintained below EOP-2 entry condition B. will start and DW temperature cannot be maintained below EOP-2 entry condition C. will not start and DW temperature can be maintained below EOP-2 entry condition D. will not start and DW temperature cannot be maintained below EOP-2 entry condition

NRC Exam 2013 QUESTION RO 32 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295012 AK3.01 Importance Rating 3.5 K&A: Knowledge of the reasons for the following responses as they apply to High Drywell Temperature: Increased drywell cooling High Drywell Temperature / 5 Explanation: Answer B - The standby DW cooling fan starts with a low flow in the running fan. With DW pressure >1.68 psig, NCC has been isolated. Without NCC, DW temperature will rise to the EOP-2 entry condition of 145°F.

A - Incorrect - Since NCC is isolated, the DW temperature will rise due to the steam leak.

C - Incorrect - The standby fan will auto start but DW temperature will continue to rise due to the loss of NCC.

D - Incorrect - The standby fan will auto start.

Technical Reference(s): ARI-H13-P800-03 Rev 9, OAI- Reference Attached: ARI-H13-P800-03 p 49, 1703 Rev 14, & EOP-2 Chart Rev B OAI-1703 p 24, & EOP-2 Chart partial Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M13-F, & L.2 and OT-3402-08-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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 33 Plant conditions as follows:

Reactor power 100%

Suppression Pool Level 18.1 ft

Suppression Pool Temperature 85°F Then, two SRVs inadvertently open due to an I&C SVI.

One minute after the SRVs open, the Reactor Operator updates the crew that Suppression Pool temperature by ICS indicates 90°F.

Following the crew update, how much time remains for the operators to close the SRVs, before a reactor scram is required to be inserted IAW Perry Technical Specifications?

A. 1 minute B. 3 minutes C. 4 minutes D. 5 minutes

NRC Exam 2013 QUESTION RO 33 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295013 AA1.02 Importance Rating 3.9 K&A: Ability to operate and/or monitor the following as they apply to High Suppression Pool Temperature: Systems that add heat to the suppression pool High Suppression Pool Temp. / 5 Explanation: Answer C - The pool heat-up rate can be determined to be 5 °F/minute. Five (5) minutes total time puts the pool temperature at 110°F. This is 4 minutes after the announcement is made. EOP-02 requires that EOP-01 be entered, and thus the reactor scrammed, prior to exceeding 110°F. TS 3.6.2.1 requires Mode Switch in shutdown immediately when > 110°F A - Incorrect - This time corresponds to the OAI-1703 Margins and Limits Scram Required time.

B - Incorrect - This time corresponds to the TS LCO for pool temperature during testing which adds heat to the pool - from the initial SRV opening.

D - Incorrect - This time corresponds to 115°F in the pool.

Technical Reference(s): EOP-2 Chart Rev B, ONI-B21-1 Reference Attached: EOP-2 Chart (partial),

Rev 11 & TS 3.6.2.1 ONI-B21-1 p 4 & TS 3.6.2.1 pp 3.6-36 & 37 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-06-C.2 Question Source: Bank # RQL-16027 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 34 A plant transient is in progress. The following indications are observed:

A Reactor Scram signal is present

Pressure set at 940 psig

NO control rods have inserted into the core

Reactor power is approximately 100%

The MSIVs are open

The main turbine is online

Recirculation pumps are in FAST speed

Reactor water level is 196 inches The EOPs direct the operator to down-shift then trip the Recirc pumps.

Which ONE of the following describes the consequences of tripping the Reactor Recirculation pumps rather than down shifting first?

A. A level transient may result in a turbine trip causing Bypass Valves to open and RPV pressure to rise until SRVs open.

B. A level transient may result in a turbine trip causing Bypass valves to open. Bypass valves will control reactor pressure at pressure set.

C. Tripping the recirculation pumps will rapidly increase vessel water level, resulting in a large power increase and possible fuel damage.

D. Tripping the recirculation pumps will result in an immediate power reduction, with a subsequent decrease in reactor pressure, and level lowering to level 2.

NRC Exam 2013 QUESTION RO 34 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295015 AA2.01 Importance Rating 4.1*

K&A: Ability to determine and/or interpret the following as they apply to Incomplete Scram: Reactor power Incomplete SCRAM / 1 Explanation: Answer A - IAW the EOP bases, if Recirc Pumps are tripped from high power, the resultant changes in steam flow, Rx pressure, and RPV level may cause a turbine trip. Since power is above the capacity of the BPVs, SRVs RPV pressure will increase until SRVs open.

B - Incorrect - Since the Rx power is > the capacity of the BPVs, Rx pressure will increase causing SRVs to open.

C - Incorrect - Tripping the recirc pumps results in a power decrease due to the void coefficient.

D - Incorrect - RPV level will initially rise due to tripping the Recirc pumps and RPV pressure will not decrease without inserting rods or injecting boron.

Technical Reference(s): EOP-01A Bases Rev 4 Reference Attached: EOP-01A Bases p 14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-03-C Question Source: Bank # INL-0828 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 35 The plant scrammed yesterday following a 16 month run. The following conditions exist:

RHR "A" is operating in Shutdown Cooling

RPV water level is 230 inches

Reactor coolant temperature is 130°F and stable

Reactor Recirculation Pump B is operating Subsequently, a loss of RPS Bus A occurs.

It is estimated that RPS Bus A can be recovered in two hours.

The effect on Shutdown Cooling is that __(1)__ isolation occurs?

And, in order to comply with Technical Specifications you will __(2)__.

__(1)__ __(2)__

A. only a Division 1 monitor reactor coolant temperature and pressure once per hour B. only a Division 1 verify two alternate methods of decay heat removal are available within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. both a Division 1 and 2 monitor reactor coolant temperature and pressure once per hour D. both a Division 1 and 2 verify two alternate methods of decay heat removal are available within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

NRC Exam 2013 QUESTION RO 35 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295020 2.2.38 Importance Rating 3.6 K&A: Knowledge of conditions and limitations in the facility license Inadvertent Cont. Isolation / 5 & 7 Explanation: Answer D - A loss of either RPS bus will cause both E12-F008 & E12-F009 to isolate. This will cause a loss of both RHR SDC subsystems. IAW TS 3.4.10 Condition A, one method of alternate decay heat removal is necessary for each INOP RHR system. With the common suction isolated, both loops of RHR are INOP.

A - Incorrect - Both divisions isolate. Circulation by an alternate method is only required if Recirc Pump not running.

B - Incorrect - Both divisions isolate.

C - Incorrect - Circulation by an alternate method is only required if Recirc Pump not running.

Technical Reference(s): TS 3.4.10, Dwg 302-642 Rev Reference Attached: TS 3.4.10 pp 3.4-24 &

HH, & ONI-C71-2 Rev 8 25, Dwg 302-642 (partial), & ONI-C71-2 pp 5, 8, & 9 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-08-B & OT-COMBINED-E12-F Question Source: Bank # Perry 2007 Modified Bank #

New Question History: Previous NRC Exam Perry 2007 SRO-17 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 36 The plant was operating at 100% power when a Loss of Feedwater Heating occurred.

The Reactor Engineer reports that Minimum Critical Power Ratio (MCPR) is 1.08.

This value of MCPR is ____.

A. within the Safety Limit; no operator actions are required B. in violation of the Safety Limit; it is required to insert Cram Rods immediately C. in violation of the Safety Limit; it is required to insert all insertable Control Rods within two hours D. in violation of the Safety Limit; it is required to insert all insertable Control Rods within four hours

NRC Exam 2013 QUESTION RO 36 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295014 AK1.05 Importance Rating 3.7 K&A: Knowledge of the operational implications of the following concepts as they apply to INADVERTENT REACTIVITY ADDITION : Fuel thermal limits Inadvertent Reactivity Addition Explanation: Answer C - For Cycle 14, with power at 100%, MCPR Safety Limit is greater than or equal to 1.10 for two Recirc Loops operating. With MCPR below 1.10, it is required to insert all insertable Control Rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months IAW TS 2.2, SL Violations.

A - Incorrect - This would be true if MCPR were >1.10.

B - Incorrect - This would be true if MEOD was exceeded. Not correct action for SL violation.

D - Incorrect - The control rods need to be inserted within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , not 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per TS 2.2.

Technical Reference(s): TS 2.1 & TS 2.2 Amendment Reference Attached: TS p 2.0-1 155 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-03-A Question Source: Bank # Perry 2007-1 #RO-33 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 b.9 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 37 The plant is operating at 96% power in coast-down for a refueling outage with the following condition present:

The IFTS tube blank flange has been removed

Testing of IFTS is in progress

AEGT Fan A, M15-C001A is running

FHB HVAC SUPP Fan B, M40-C001B is running.

FHB HVAC EXH Fans, M40-C002A & C, are running A seismic event then occurs resulting in the following annunciators alarming:

ANNULUS EXH FAN A FLOW LOW, H13-P800-01-A1

AIRBORNE RAD P804, H13-P680-07-A10,

COM AREA & PRCS MON P906, H13-P680-08-A4

COMMON AIRBORNE P902, H13-P680-08-A1

FHB Evacuation Alarm, H13-P902-01-B1 NLO reports AEGTS A flow is 1600 CFM The following radiation monitors have HIGH alarms locked in:

FUEL PREP POOL, D21-K322

SPENT FUEL POOL, D21-K332

ANN EXH A GAS 1D17-K697A

FHB VENT EXH GAS, D17-K716 What is the configuration of the AEGTS and FHB ventilation fans?

AEGT Fan FHB HVAC EXH FHB HVAC SUPP Fan B Fans A & C A. A running running tripped B. B running running tripped C. A running tripped running D. B running tripped running

NRC Exam 2013 QUESTION RO 37 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295033 EK2.02 Importance Rating 3.8 K&A: Knowledge of the interrelations between HIGH SECONDARY Containment Area Radiation Levels and the following: Process radiation monitoring system High Secondary Containment Area Radiation Levels / 9 Explanation: Answer A - AEGT A is running. There is no high rad trip on AEGTS. FHB exhaust fans are running and the supply fan trips on high rads.

B - Incorrect - The AEGT low flow alarm comes in at 1650 cfm and the opposite fan starts at 1500 cfm.

So the AEGT B fan is not running.

C - Incorrect - The FHB exhaust fans do not trip on hi rads, but the supply fan does.

D - Incorrect - The AEGT low flow alarm comes in at 1650 cfm and the opposite fan starts at 1500 cfm.

So the AEGT B fan is not running. The FHB exhaust fans do not trip on hi rads, but the supply fan does.

Technical Reference(s): ARI-H13-P902-01 Rev 5, ARI- Reference Attached: ARI-H13-P902-01 p 7, &

H13-P680-08 Rev 13, ARI-H13-P800-01 Rev 7, & ARI- ARI-H13-P800-01 p 5 H13-P680-07 Rev 20 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M15-E, OT-COMBINED-M40-B.4 Question Source: Bank #

Modified Bank # Clinton 2002 New Question History: Previous NRC Exam Clinton 2002 # RO-69 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 & b.11 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 38 The following plant conditions exist:

A Loss of Coolant Accident has occurred

The RPV is depressurized

All control rods are fully inserted

LPCI A, B, and C are injecting into the reactor at 6800 gpm (each)

RPV water level is 20 inches and rising rapidly Subsequently, LPCI A System flow and pump amps are observed to be rising significantly and discharge pressure has lowered.

LPCI B and C parameters are stable within their normal indications.

Which one of the following describes the condition of the LPCI A Pump, including procedural guidance for continued operation?

The LPCI A Pump is ____.

A. cavitating and may be secured since adequate core cooling exists B. running out and may be secured since adequate core cooling exists C. cavitating and should not be secured since adequate core cooling does not exist D. running out and should not be secured since adequate core cooling does not exist

NRC Exam 2013 QUESTION RO 38 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 203000 A2.12 Importance Rating 2.6*

K&A: Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump runout RHR/LPCI; Injection Mode Explanation: Answer B - High current, high flow, and low discharge pressure are indications of pump runout. The pump may be secured since adequate core cooling exists with RPV level > TAF and rising rapidly. Additionally, IAW ONI-E12-1 the pump should be secured if mis-operation in automatic is confirmed.

A - Incorrect - Cavitation characteristics would include fluctuations in pump amps and discharge pressure.

C - Incorrect - Cavitation characteristics would include fluctuations in pump amps and discharge pressure. The LPCI pump may be secured since adequate core cooling exists and further operation of the pump may cause damage.

D - Incorrect - The LPCI pump may be secured since adequate core cooling exists and further operation of the pump may cause damage.

Technical Reference(s): SOI-E12 Rev 57, OT-3303-02 Reference Attached: SOI-E12 pp 7-8, OT-Lesson Plan Rev 4, and EOP-Bases Rev 3 3303-02 Lesson Plan pp 89-90, and EOP-Bases p 34 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3303-03-02.7 & OT-COMBINED-E12-H Question Source: Bank #

Modified Bank # Perry 2003 New Question History: Previous NRC Exam Perry 2003 #RO-41 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

NRC Exam 2013 QUESTION RO 39 RHR A is running in Refuel Mode Shutdown Cooling.

During performance of a surveillance test, an isolation signal was inadvertently inserted that resulted in the following:

E12-F008, SHUTDOWN COOLING INBD SUCT ISOL Valve automatically isolates

E12-F009, SHUTDOWN COOLING INBD SUCT ISOL Valve automatically isolates

E12-F037A, RHR UPPER POOL COOLING ISOL Valve automatically isolates

E12-F006A, RHR A SHUTDOWN CLG SUCT Valve remains open Which of the following isolation signals would not result in this valve lineup?

A. High Drywell Pressure B. High Reactor Pressure C. Low Reactor Water Level D. High RHR Room Temperature

NRC Exam 2013 QUESTION RO 39 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 205000 A3.01 Importance Rating 3.2 K&A: Ability to monitor automatic operations of the Shutdown Cooling System (RHR Shutdown Cooling Mode) including: Valve operation Shutdown Cooling Explanation: Answer A - A high DW pressure signal is the only isolation signal that will not isolate the SDC valves.

B, C, & D - Incorrect - All of these signals isolate the SDC INBD suct, OTBD suct, & Upper pool isol valves Technical Reference(s): OAI-1703 rev 14 & PYBP-POS- Reference Attached: OAI-1703 pp 24-25 &

027 Rev 2 PYBP-POS-027 p 18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E12-F Question Source: Bank #

Modified Bank # Grand Gulf 2008 New Question History: Previous NRC Exam Grand Gulf 2008 #RO-51 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 40 The plant was operating at rated power.

An inadvertent initiation of Low Pressure Core Spray occurred due to failure of DW pressure trip units.

Only the Immediate Actions of ONI-E12-1, Inadvertent Initiation of ECCS or RCIC were performed and were successful.

Subsequently, a loss of offsite power occurred coincident with a LOCA.

When power is restored to the divisional buses by the diesel generators, LPCS will ____.

A. not automatically restart B. automatically restart immediately C. automatically restart in 10 seconds D. automatically restart in 15 seconds

NRC Exam 2013 QUESTION RO 40 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209001 K2.01 Importance Rating 3.0 K&A: Knowledge of electrical power supplies to the following: Pump power LPCS Explanation: Answer A - The Immediate Action for an inadvertent initiation of LPCS is to override the pump to OFF. Since the LPCS pump has been overridden off, K13 is energized, preventing the pump from restarting.

B - Incorrect - This would be true if the pump had not been previously overridden off.

C - Incorrect - This is the time allowed for the DG to reenergize the bus.

D - Incorrect - This is the normal time delay for the LPCS pump to start without a LOOP.

Post Exam comments revised correct answer to B vice A.

Technical Reference(s): OT-COMBINED-E21 LP Reference Attached: OT-COMBINED-E21 LP (PowerPoint) Rev 1 (PowerPoint) slides 39-41 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E21-F Question Source: Bank #

Modified Bank # Monticello 2009 New Question History: Previous NRC Exam Monticello 2009 #RO-31 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 41 The plant was operating at rated power when a LOCA occurred resulting in the following:

The reactor scrammed

RPV pressure has dropped to 480 psig

RPV level is 14 inches and stable

HPCS and RCIC are injecting to the RPV Based on these conditions, the LPCS Injection Valve, 1E21-F005 is __(1)__ and the LPCS Pump Min Flow Valve, 1E21-F011 is __(2)__.

__(1)__ __(2)__

A. open closed B. open open C. closed closed D. closed open

NRC Exam 2013 QUESTION RO 41 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209001 A4.03 Importance Rating 3.7 K&A: Ability to manually operate and/or monitor in the control room: Injection valves LPCS Explanation: Answer B - LPCS will initiate and the pump will start when RPV level drops <L1 (16.5).

The Injection valve will open upon system initiation as long is the pressure between the injection valve and the RPV is < 600 psig. However, LPCS will not inject until RPV pressure is < 450 psig. This will cause the min flow valve to be open.

A - Incorrect - The min flow valve will be open (no injection flow).

C - Incorrect - The Injection valve will be open and the min flow valve will be open.

D - Incorrect - The min flow valve will be open (no injection flow).

Technical Reference(s): SOI-E21 Rev 28 Reference Attached: SOI-E21 p 13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E21-F Question Source: Bank # LaSalle 2007 Modified Bank #

New Question History: Previous NRC Exam LaSalle 2007 #RO-14 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 42 Which of the following components, if not functioning properly, could result in a water hammer event in the HPCS system?

A. HPCS Pump Discharge Restricting Orifice, E22-D002 B. HPCS Pump CST Suction Check Valve, E22-F002 C. HPCS Testable Check Valve, E22-F005 D. HPCS Water Leg Pump, E22-C002

NRC Exam 2013 QUESTION RO 42 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209002 K1.03 Importance Rating 3.0 K&A: Knowledge of the physical connections and/or cause-effect relationships between High Pressure Core Spray System (HPCS) and the following: Water leg (jockey) pump; BWR-5,6 HPCS Explanation: Answer D - Per SDM-E22A, the waterleg pump prevents water hammer by keeping the injection filled and pressurized between the pump discharge check valve and the injection valve (F004).

A - Incorrect - The HPCS Pump Discharge Restricting Orifice prevents pump runout at low discharge pressures. It will not cause a water hammer.

B - Incorrect - The HPCS Pump CST Suction Check Valve prevents cross-connecting the CST and the Suppression Pool. If it malfunctions, it will allow CST pressure to be felt in the HPCS system preventing water hammer.

C - Incorrect - A malfunction of the HPCS Testable Check Valve would either allow Rx pressure to be felt in the HPCS system or prevent injection if stuck shut.

Technical Reference(s): SDM-E22A, Rev. 8 Reference Attached: SDM-E22A, p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E22A-C.5 Question Source: Bank # River Bend, 2008 Modified Bank #

New Question History: Previous NRC Exam River Bend, 2008, RO Question 17 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 43 Breaker EH1104, 4.16KV TO 480V XFMR EHF-1-A TO BUS EF-1-A, tripped.

Which one of the following has lost power?

A. Suppression Pool Cleanup Pump, 1G42-C001 B. Standby Liquid Control Pump A, 1C41-C001A C. Reactor Water Cleanup Pump A, 1G33 C001A D. Control Complex Chill Water Pump C, P47-C001C

NRC Exam 2013 QUESTION RO 43 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 211000 K2.01 Importance Rating 2.9 K&A: Knowledge of electrical power supplies to the following: SBLC pumps SLC Explanation: Answer B - Standby Liquid Control Pump A, 1C41-C001A is powered from Bus EF-1-A, Breaker EH1104 on Bus EH11 supplies power to Bus EF-1-A A - incorrect - is plausible; powered from F-1-E bus C - incorrect - is plausible; Powered from F-1-C bus D - incorrect - is plausible; powered from EF-2-A Bus Technical Reference(s): Dwg 206-021 Rev SSSS Reference Attached: Dwg 206-021 (partial)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R10-J-2 & OT-COMBINED-C41-F-2 Question Source: Bank # Perry 2010 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 #RO-62 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 44 RPS B motor-generator output voltage rose to 133 volts over several seconds then returned to normal and remained there.

The status of the CVCW OTBD RETURN MOV ISOL VLV, 1P50-F150 is __(1)__ and the CVCW INBD RETURN MOV ISOL VALVE, 1P50-F140 is __(2)__

__(1)__ __(2)__

A. open open B. open closed C. closed open D. closed closed

NRC Exam 2013 QUESTION RO 44 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 212000 K3.02 Importance Rating 3.7 K&A: Knowledge of the effect that a loss or malfunction of the Reactor Protection System will have on following: Primary containment isolation system/nuclear steam supply shut-off; Plant-Specific RPS Explanation: Answer B - A voltage increase 132 volts will cause the RPS EPA to trip and deenergizes RPS B Bus. The loss of the B bus will cause the inboard containment isolation valves to close.

A - Incorrect - The inboard containment isolation valve will be shut.

C - Incorrect - The outboard isolation valves will not be affected by a loss of RPS B, only the inboard valves.

D - Incorrect - The outboard isolation valves will not be affected by a loss of RPS B Technical Reference(s): TS 3.3.8.2 & ONI-C71-2 Rev 8 Reference Attached: TS 3.3.8.2 p 3.3-79 &

ONI-C71-2 pp 5, 7, 9, & 11 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-11(LP)-A.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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 45 The plant is operating at rated power with the following:

RPS Bus A is on its NORMAL power supply

RPS Bus B is on its ALTERNATE power supply The main line fuses in the MCC disconnect for RPS MG Set A opened.

Scram solenoid power (white) indicating lights __(1)__ will extinguish on P680.

The appropriate action before the RPS A half-scram can be reset is to __(2)__.

__(1)__ __(2)__

GP 1A, GP 1B, place the MG SET A TRANSFER A.

GP 3B and GP 3A switch in the ALT A position GP 1A, GP 1B, make RPS MG Set A power B.

GP 3B and GP 3A available again to RPS Bus A GP 1A, GP 2A, place the MG SET A TRANSFER C.

GP 3A and GP 4A switch in the ALT A position GP 1A, GP 2A, make RPS MG Set A power D.

GP 3A and GP 4A available again to RPS Bus A

NRC Exam 2013 QUESTION RO 45 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 212000 A2.02 Importance Rating 3.7 K&A: Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: RPS bus power supply failure RPS Explanation: Answer D - Only one RPS bus can be powered from the alternate power supply at a time due to design of the MG SET TRANSFER switch. Therefore, the only way to reset the A half-scram is by making RPS MG Set A power available again to RPS Bus A. The GP A lights are in the circuits that bring RPS A bus power to all of the A scram solenoids; the GP B lights are in the circuits that bring RPS B bus power to all of the B scram solenoids. When RPS MG Set A tripped (in the stem conditions), de-energizing RPS Bus A, all of the A scram solenoid circuits (including the associated power monitoring lights) lost power.

A - Incorrect - Cannot put RPS A in Alt position due to switch design, also as discussed above the GP A lights will be extinguished.

B - -Incorrect- As discussed above the GP A lights will be extinguished.

C - Incorrect - Cannot put RPS A in Alt position due to switch design, Technical Reference(s): SOI-C71 Rev. 19, SDM C71, Reference Attached: SOI-C71 p 4, SDM C71, Rev 12 p75 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C71-M Question Source: Bank # Grand Gulf 2010 Modified Bank #

New Question History: Previous NRC Exam Grand Gulf 2010 RO #47 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 46 The plant is in Mode 2 with a reactor startup in progress, with the following:

All APRMs are reading 2% power

IRM B is reading 122 on Range 6

Annunciator H13-P680-0006-E3 IRM D/H UPSC TRIP/INOP alarms due to IRM H module becoming unplugged The response of the Reactor Protection System (RPS) and Rod Control and Information System (RC&IS) is to generate a ____.

A. rod block only B. half scram only C. full reactor scram D. rod block and half scram

NRC Exam 2013 QUESTION RO 46 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215003 K4.02 Importance Rating 4.0 K&A: Knowledge of Intermediate Range Monitor (IRM) System design feature(s) and/or interlocks which provide for the following: Reactor SCRAM signals IRM Explanation: Answer D - IRM B upscale trip (120/125) and RB (108/125) would be in with Mode Switch in STARTUP. IRM H INOP would also cause 1/2 scram and Rod Block. IRM B and H RPS trips are both in RPS Trip System B and would only give a 1/2 scram.

A - Incorrect - This would be true if IRM B was above 108/125 and below 120/125 and if IRM H INOP was not in.

B - Incorrect - With IRM B above trip 120/125 it is also above RB 108/125. IRM H INOP gives RB.

C - Incorrect - This would be true if IRM B and H were in different RPS Trip Systems.

Technical Reference(s): ARI-H13-P680-06 Rev. 8, PDB- Reference Attached: ): ARI-H13-P680-06 pp I0005 Rev. 9 33 and 71, PDB-I0005 pp 3 and 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51_IRM-F Question Source: Bank # Nine Mile 2002 Modified Bank #

New Question History: Previous NRC Exam Nine Mile 2002 RO #49 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 47 Regarding Source Range Monitors:

A reduction in SRM detector gas pressure will cause an SRM to read __(1)__ and as the U235 coating depletes in an SRM detector, the SRM will read __(2)__.

__(1)__ __(2)__

A. higher lower B. higher higher C. lower lower D. lower higher

NRC Exam 2013 QUESTION RO 47 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215004 K5.01 Importance Rating 2.6 K&A: Knowledge of the operational implications of the following concepts as they apply to Source Range Monitor (SRM) System: Detector operation Source Range Monitor Explanation: Answer C - The decreased gas pressure, in combination with a smaller amount of uranium, makes the detector much less sensitive to thermal neutron interaction and subsequent ionization of the argon gas.

A - Incorrect - As gas pressure decreases, the amount of interactions taking place with fission fragments decreases, reducing the output current.

B - Incorrect - Decreasing gas pressure, and a smaller amount of uranium, makes the detector much less sensitive to thermal neutron interaction.

D - Incorrect - A smaller amount of uranium, makes the detector much less sensitive to thermal neutron interaction.

Technical Reference(s): SDM OT-COMBINED-C51 IRM Reference Attached: SDM OT-COMBINED-Rev 8 C51 IRM p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x Question Source: Bank # Monticello 2009 Modified Bank #

New Question History: Previous NRC Exam Monticello 2009 #RO-36 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.6 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 48 The plant is operating at rated power.

Refer to the attached LPRM to APRM Channel Assignment sheet (Attachment 2), SOI-C51 (APRM) for current LPRM status.

The LPRMs that are crossed out are Inoperable.

The current LPRM status results in ____ being Inoperable.

Worksheet Attached A. no APRM channels B. 1 APRM channel C. 2 APRM channels D. 3 APRM channels

NRC Exam 2013 QUESTION RO 48 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215005 K6.03 Importance Rating 3.1 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Average Power Range Monitor/Local Power Range Monitor System: Detector APRM / LPRM / OPRM Explanation: Answer C - Per SOI-C51(APRM), 14 LPRM detector inputs are required for each APRM and that each APRM channel needs at least 2 LPRM detector inputs from each axial level. APRM B only has 13 operable LPRM detectors and APRM C only has 1 LPRM detector on the C axial level.

A & B - Incorrect - APRM B and C are Inop D - Incorrect - Only APRM B and C are Inop.

Technical Reference(s): SOI-C51(APRM) Rev 11 Reference Attached: SOI-C51(APRM) p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-07-H & OT-COMBINED-C51AP_OPRM-C.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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 49 The plant is operating in accordance with EOP-1 RPV Control with the following conditions:

The MSIVs have isolated

The MFP is being used to control RPV Level

RCIC is being used to help control RPV pressure IAW EOP-SPI 6.6, RCIC Injection and Pressure Control

SRVs are being cycled when required for RPV pressure control What is the response of the RCIC system if Suppression Pool level rises to 18.5 feet due to the operation of RCIC and SRVs?

A. RCIC suction remains on the CST and RCIC operation remains unchanged B. The RCIC First Test Valve To CST, 1E51-F022 and RCIC Second Test Valve To CST, 1E51-F059 close and RCIC runs on minimum flow with suction from the CST C. The RCIC Pump Supr Pl Suct Isol, 1E51-F031 opens, the RCIC Pump CST Suction Valve, 1E51-F010 closes, and RCIC pumps from the Suppression Pool to the CST D. The RCIC Pump Supr Pl Suct Isol, 1E51-F031 opens, the RCIC Pump CST Suction Valve, 1E51-F010 and the RCIC First Test Valve To CST, 1E51-F022 and RCIC Second Test Valve To CST, 1E51-F059 close and RCIC runs on minimum flow

NRC Exam 2013 QUESTION RO 49 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 A1.07 Importance Rating 3.3 K&A: Ability to predict and/or monitor changes in parameters associated with operating the Reactor Core Isolation Cooling System (RCIC) controls including: Suppression pool level RCIC Explanation: Answer A - > With EOP-SPI performed, the shift from CST to Suppression Pool on a high SP level is bypassed. RCIC will continue to run. (Would shift @ 18.4 if not overridden)

B - Incorrect - If EOP-SPI 6.6 were not performed, the test valves would close. The suction will not swap due to the performance of EOP-SPI 6.6. Misconception - the test valves will not close.

C - Incorrect - If EOP-SPI 6.6 was not performed, the suction would shift and the test valves would close.

D - Incorrect - This is what would happen if EOP-SPI 6.6 was not performed.

Technical Reference(s): EOP-SPI 6.6 Rev 0 Reference Attached: EOP-SPI 6.6 p 2 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E51-B.2:

Question Source: Bank # RQL-0266 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 50 The following conditions exist:

The reactor scrammed due to a loss of the feedwater system

High Pressure Core Spray pump tripped due to a seized shaft

Reactor Core Isolation Cooling is maintaining RPV level at 30 inches

Reactor pressure is 870 psig and lowering slowly Which of the following conditions would not indicate an immediate threat to RCICs ability to maintain RPV level?

A. RCIC Turbine Exhaust Diaphragm rupture.

B. High RCIC steam line flow caused by a steam supply line leak.

C. A high Steam Tunnel temperature resulting in an MSIV isolation.

D. High temperatures in RHR A room resulting in a SDC isolation signal.

NRC Exam 2013 QUESTION RO 50 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 K3.01 Importance Rating 3.7 K&A: Knowledge of the effect that a loss or malfunction of the Reactor Core Isolation Cooling System (RCIC) will have on following: Reactor water level RCIC Explanation: Answer C - The steam tunnel high temperature has a 29 minute time delay to isolate - not an immediate threat.

A - Incorrect - An exhaust diaphragm rupture will cause an immediate RCIC turbine trip.

B - Incorrect - RCIC will isolate after 8 seconds on a high steam flow - immediate.

D - Incorrect - A high temp in the RHR rooms will immediately isolate RCIC.

Technical Reference(s): ARIs H13-P601-20 Rev 16 & Reference Attached: ARIs H13-P601-20 p 57 H13-P601-21 Rev 15, & H13-P601-21 pp 19, 33, & 49 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E51-F.2 Question Source: Bank #

Modified Bank # Perry 2010 New Question History: Previous NRC Exam Perry 2010 #RO-49 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 51 The plant scrammed due to a LOCA. Current conditions are as follows:

RPV water level is - 20 inches and lowering

DW Pressure is 8 psig and slowly rising The BOP operator then initiates ADS by arming and depressing the ADS A TIME DELAY LOGIC MANUAL INITIATION and the ADS A INSTANTANEOUS LOGIC MANUAL INITIATION pushbuttons resulting in only four ADS SRVs opening.

Four ADS SRVs opening is an insufficient number of SRVs for __(1)__.

The action required to mitigate this situation is __(2)__.

__(1)__ __(2)__

A. decay heat removal open additional non-ADS SRVs B. Emergency Depressurization open additional non-ADS SRVs C. decay heat removal override ADS IAW ONI-E12-1 Inadvertent Initiation Of ECCS/RCIC D. Emergency Depressurization override ADS IAW ONI-E12-1 Inadvertent Initiation Of ECCS/RCIC

NRC Exam 2013 QUESTION RO 51 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 218000 A2.04 Importance Rating 4.1 K&A: Ability to (a) predict the impacts of the following on the Automatic Depressurization System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS failure to initiate ADS Explanation: Answer B - IAW the EOP bases, the minimum number of SRVs required for ED (MNSRED) is 5. This is based on MSCP RHR injection makeup. IAW EOP-4-2 ED, open additional SRVs is the action.

A - Incorrect - This is the definition for MNSDHR which is 2 SRVs.

C - Incorrect - This is the definition for MNSDHR which is 2 SRVs. - overriding ADS would not be appropriate nor required by EOP-4-2.

D - Incorrect - Overriding ADS would not be appropriate nor required by EOP-4-2.

Technical Reference(s): EOP Bases Rev 3 Reference Attached: EOP Bases p 43 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x Question Source: Bank #

Modified Bank # Nine Mile 2008 New Question History: Previous NRC Exam Nine Mile 2008 #SRO-87 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 52 The following plant conditions exist:

Drywell pressure 1.3 psig

Reactor water level 105 inches

Main condenser vacuum 25 inches Hg A

Reactor pressure 75 psig The system components that isolated based on these plant conditions are the ____.

A. Reactor Water Sample isolation valves, RWCU isolation valves, MSIVs and MSL Drain isolation valves B. RWCU isolation valves, MSIVs and MSL Drain isolation valves, RCIC steam supply line isolation valves C. MSIVs and MSL Drain isolation valves, NCC Containment & Drywell isolation valves, RWCU isolation valves D. RCIC steam supply line isolation valves, Drywell Floor Drain Sump & Containment Drain Sump isolation valves, Reactor Water Sample isolation valves

NRC Exam 2013 QUESTION RO 52 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 A3.02 Importance Rating 3.5 K&A: Ability to monitor automatic operations of the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off including: Valve closures PCIS/Nuclear Steam Supply Shutoff Explanation: Answer A - The RWCU sample and RWCU isolation valves close on a L2 signal. The MSL and MSL Drain valves close on low vacuum (21.5HgA)

B - Incorrect - RCIC steam supply line isolation condition not met (reactor pressure < 60 psig).

C - Incorrect - NCC isolation valve isolation conditions not met (RPV level < Level 1 or DW pressure >

1.68 psig).

D - Incorrect - RCIC steam supply line isolation condition not met (reactor pressure < 60 psig).

Technical Reference(s): IOI-18 Rev 12 & ONI-N62 Rev 9 Reference Attached: IOI-18 p 105 & ONI-N62 p4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21(NS4)-F (multiple)

Question Source: Bank # Perry 2001 Modified Bank #

New Question History: Previous NRC Exam Perry 2001 #RO-76 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 53 The plant is in a refueling outage.

Containment Vessel and Drywell Purge Exhaust System (M14) is operating in REFUEL MODE

CONT VENT EXH radiation monitor, 1D17-K609B is failed upscale The following then occurs:

CONT VENT EXH radiation monitor, 1D17-K609D fails downscale The Containment Vessel and Drywell Purge Exhaust System.

A. CNTMT & DW PURGE Isolation Dampers shut B. CNTMT PURGE SUPP FAN A & B 1M14-C001A(B) trip C. continues to run in REFUEL MODE with all fans operating D. CNTMT & DW PURGE EXH FAN A & B M14-C003A(B) trip

NRC Exam 2013 QUESTION RO 53 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 K3.18 Importance Rating 3 K&A: Knowledge of the effect that a loss or malfunction of the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off will have on following: Containment ventilation PCIS/Nuclear Steam Supply Shutoff Explanation: Answer C - Failures on both the A & D or both B & C rad monitors will isolate M14 and cause the fans to trip.

A - Incorrect - Wrong combination of rad monitors to cause damper isolation.

B - Incorrect - Wrong combination of rad monitors to cause Supply fans to trip D - Incorrect - Wrong combination of rad monitors to cause Exhaust fans to trip Technical Reference(s): ARI-H13-P680-07 Rev 20 & Reference Attached: ARI-H13-P680-07 pp PDB-I05 Rev 9 67-68 & PDB-I05 p 26 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M14-F OT-COMBINED-D17A-F Question Source: Bank #

Modified Bank # Perry 2007-1 New Question History: Previous NRC Exam Perry 2007 #SRO-17 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 54 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 (H13-P601-21-A2) just reset You have been directed to verify the status of the ADS SRVs.

The ADS SRVs are ____.

A. closed based on stable SRV tailpipe temperatures B. closed based on SRV tailpipe temperatures slowly lowering C. open based on SRV tailpipe temperatures of approximately 250°F and stable D. open based on SRV tailpipe temperatures of approximately 330°F and slowly rising due to decay heat

NRC Exam 2013 QUESTION RO 54 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 239002 A4.07 Importance Rating 3.6 K&A: Ability to manually operate and/or monitor in the control room: Lights and alarms SRVs Explanation: Answer C - With ED occurring 20 minutes ago and the only injection source being CTS, RPV pressure will lower to < 30 psig. This will cause the SRV OPEN annunciator to reset and the SRV Open/Close lights to change state. The SRVs are verified open by observing SRV tailpipe temperature of 250°F which corresponds to reactor pressure of ~25 psig.

A & B - Incorrect - SRVs are still open. The annunciator and indicating lights are activated from the SRV tailpipe pressure switch.

D - Incorrect - Tailpipe temperature of 330°F corresponds to an open SRV at normal reactor pressure.

Technical Reference(s): ARI-H13-P601-019 Rev 14, Reference Attached: ARI-H13-P601-019 p 17 EOP-01 Chart Rev D & ABB Steam Tables & ABB Steam Tables Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21_N11-F Question Source: Bank # Perry 2010 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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 55 You have been directed to perform the field actions for ONI-SPI A-8, LPCS Fire Water.

Connecting the fire hoses to the LPCS system is done at __(1)__. After ONI-SPI A-8 alignment is complete, commence RPV injection by __(2)__.

__(1)__ __(2)__

A. Aux Building 620 East Side starting the LPCS Pump B. Aux Building 620 East Side opening the LPCS Injection Valve C. Aux Building 620 West Side starting the LPCS Pump D. Aux Building 620 West Side opening the LPCS Injection Valve

NRC Exam 2013 QUESTION RO 55 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 259002 2.1.30 Importance Rating 4.4 K&A: Ability to locate and operate components, including local controls Reactor Water Level Control Explanation: Answer B - IAW ONI-SPI A-8, the 5 fire hose is connected to LPCS via a flange on Aux 620 C/07. This is on the east side of the Aux building. This instruction is entered when the LPCS pump is unavailable for injection. Additionally, the LPCS pump breaker is racked out during alignment. Opening the injection valve is the method for adding water to the RPV.

A - Incorrect - The LPCS pump breaker is racked out.

C - Incorrect - The east side is the location for the HPCS flange connection.

D - Incorrect - The east side is the location for the HPCS flange connection. The LPCS pump breaker is racked out.

Technical Reference(s): ONI-SPI A-8 Rev 2 Reference Attached: ONI-SPI A-8 pp 2, 3, 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-02-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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 56 A LOCA occurred resulting in fuel damage and elevated radiation levels in the drywell, containment, and annulus.

Subsequently, a fire ignited in one the AEGTS trains and charcoal beds releasing smoke and contamination into the Intermediate Building.

1D17-K786 & 2D17-K786 (Unit 1 & 2) D17 Plant Vent Radiation Monitors have ALERTS locked in.

1D19-K300 & 2D19-K300 (Unit 1 & 2) D19 Plant Vent Post Accident Radiation Monitors __(1)__.

Any release due to the fire will be monitored by the __(2)__ Plant Vent D19-K300.

__(1)__ __(2)__

A. are currently running Unit 1 B. are currently running Unit 2 C. need to be manually started Unit 1 D. need to be manually started Unit 2

NRC Exam 2013 QUESTION RO 56 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 261000 K1.04 Importance Rating 2.5 K&A: Knowledge of the physical connections and/or cause-effect relationships between Standby Gas Treatment System and the following: High radiation sampling system SGTS Explanation: Answer B - The D19 will auto start on a LOCA signal or a HIGH from the respective D17 rad monitor. Most areas of the IB exhaust to the Unit 2 plant vent with the exception of areas served by the sub-exhaust van (areas with potential for contamination).

A - Incorrect - Only the IB Sub-Exhaust areas go to the Unit 1 plant vent.

C - Incorrect - The high level rad monitors start on a L2 LOCA signal. Only the IB Sub-Exhaust areas go to the Unit 1 plant vent.

C - Incorrect - The high level rad monitors start on a L2 LOCA signal.

Technical Reference(s): ODCM Rev 19, SDM-D19 Rev Reference Attached: ODCM p 33, SDM-D19 6, & Dwg 208-056 Sh. 207 Rev L pp 31-34, & Dwg 208-056 Sh. 207 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D19-F & OT-COMBINED-M33-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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 57 The plant is operating at rated power.

Which of the following combinations of breaker alignments will require entry in to the Required Actions of TS 3.8.1, AC Sources-Operating?

Reference Provided:

L1003 L1004 L1006 L2003 L2004 Main Start-up Bus Tie Start-up Main Start-up Bus Tie Supply Breaker Supply Supply Breaker Breaker Breaker Breaker A. Open Shut Open Open Shut B. Shut Open Shut Open Shut C. Shut Open Open Shut Open D. Open Shut Shut Open Open

NRC Exam 2013 QUESTION RO 57 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262001 K2.01 Importance Rating 3.3 K&A: Knowledge of electrical power supplies to the following: Off-site sources of power AC Electrical Distribution Explanation: Answer D - Breakers L2003 & 2004 are the normal and alternate feeds to Bus L20. With both of these breakers open, one source of off-site power is lost - Unit 2 Startup Transformer.

A - Incorrect - This lineup is possible, but not preferred because the L10 & L20 buses will not auto transfer to the normal source if the alt source is lost. This still meets TS requirements.

B, C - Incorrect - Both of these lineups still meet the TS requirements for 2 off-site sources of power.

Technical Reference(s): SVI-R10-T5227 Rev 7 Reference Attached: SVI-R10-T5227 p 5 Proposed references to be provided to applicants during examination: SVI-R10-T5227 Learning Objective (As available): OT-COMBINED-R10-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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 58 A reactor startup is in progress with the following conditions:

Reactor power approximately 16%

Turbine generator ready to synchronize to the grid Vital inverter DB-1-A experienced a failure. Additionally, the static transfer switch failed to shift to the Alternate Source resulting in a loss of power to Bus V-1-A.

Based on these conditions, other than scramming, control rods can ____.

A. not be inserted or withdrawn B. be inserted using In-Timer-Skip C. only be inserted or withdrawn by single notch D. only be withdrawn using the Continuous Withdraw

NRC Exam 2013 QUESTION RO 58 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262002 K3.07 Importance Rating 2.6 K&A: Knowledge of the effect that a loss or malfunction of the Uninterruptable Power Supply (A.C./D.C.) will have on following: Movement of control rods; Plant-Specific UPS (AC/DC)

Explanation: Answer A - Per ONI-R25-2 if V-1-A is lost the control rods will not be able to be moved except by scram.

B, C & D - Incorrect - Per ONI-R25-2 if V-1-A is lost the control rods will not be able to be moved.

Technical Reference(s): ONI-R25-2 Rev. 10 Reference Attached: ONI-R25-2 p 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R14_R15-A & J.1 Question Source: Bank # Nine Mile Point 2 2010 Modified Bank #

New Question History: Previous NRC Exam Nine Mile Point 2 2010 RO#6 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 59 Switching requiring defeat of the Kirk Key Interlock will be performed on Bus D-1-A.

The purpose of the Kirk Key Interlock is to __(1)__. To defeat the Kirk Key Interlock, breakers __(2)__ must be racked out to Disconnect?

__(1)__ __(2)__

A. prevent removal of the DC Bus Battery D1A08, Reserve Charger Output Fuses under a load Breaker and D1A03, Bus D-1-A Main Breaker B. prevent paralleling the Normal and D1A02, Normal Charger Output Breaker Reserve Battery chargers and D1A08, Reserve Charger Output Breaker C. prevent removal of the DC Bus Battery D1A02, Normal Charger Output Breaker Fuses under a load and D1A03, Bus D-1-A Main Breaker D. prevent paralleling the Normal and D1A08, Reserve Charger Output Reserve Battery chargers Breaker and D1A03, Bus D-1-A Main Breaker

NRC Exam 2013 QUESTION RO 59 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 263000 K4.01 Importance Rating 3.1 K&A: Knowledge of D.C. Electrical Distribution design feature(s) and/or interlocks which provide for the following: Manual/ automatic transfers of control; Plant-Specific DC Electrical Distribution Explanation: Answer C - The Kirk Key Interlock prevents removing the battery fuses with any current present. The Normal Charger and the Main Bus breakers must be opened to isolate the battery fuses.

A - Incorrect - The reserve charger is isolated when the Main Breaker is opened.

B & D - Incorrect - This is not the purpose of the application of the Kirk Key Interlock in this situation.

(This is used in the M40 system). The Reserve Charger breaker does not need to be racked out.

Technical Reference(s): SDM-R42 Rev 8 Reference Attached: SDM-R42 p 13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R42-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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 60 The plant was operating at rated power when a large-break LOCA occurred in the drywell resulting in the following conditions

Drywell Pressure peaked at 8 psig and is currently at 7 psig.

Reactor Pressure is at 110 psig.

Reactor Water Level is at 0 (zero) inches.

Which of the following describes the proper load sequencing of the associated equipment after the accident signal is received?

A. Div 2 DG is ready to load at 13 seconds ESW B pump breaker closes at 10 seconds RHR C pump breaker closes immediately B. Div 2 DG is ready to load at 10 seconds ESW B pump breaker closes at 18 seconds RHR B pump breaker closes immediately C. ESW B pump breaker closes at 18 seconds RHR B pump breaker closes at 5 seconds RHR C pump breaker closes at 5 seconds D. Div 2 DG is ready to load at 10 seconds RHR B pump breaker closes at 5 seconds RHR C pump breaker closes immediately

NRC Exam 2013 QUESTION RO 60 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 264000 K5.06 Importance Rating 3.4 K&A: Knowledge of the operational implications of the following concepts as they apply to Emergency Generators (Diesel/Jet): Load sequencing EDGs Explanation: Answer D - The DG is ready to load 10 sec, RHR B pump starts 5 sec, and RHR C pump starts immediately after the receipt of a LOCA signal. Additionally, ESW pumps start ~18 seconds after a LOCA signal is received.

A - Incorrect - This is the timing for Div 3 DG. ESW pump starts 18 seconds after a LOCA signal. RHR C pump starts immediately.

B - Incorrect - ESW pump starts 18 seconds after a LOCA signal. RHR B pump starts 5 seconds after a LOCA signal.

C - Incorrect - RHR C pump starts immediately after a LOCA signal.

Technical Reference(s): Lesson Plan E12 Rev 2 & SVI- Reference Attached: Lesson Plan E12 p 31 &

R43-T5367 Rev 21 SVI-R43-T5367 pp 129-130 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R43_R48-H.1 Question Source: Bank # Browns Ferry 2008 Modified Bank #

New Question History: Previous NRC Exam Browns Ferry 2008 #RO-22 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 61 The HPCS Pump automatically started on a valid demand signal.

The Division 3 Diesel Generator is now running.

Annunciator DG TRIP LUBE OIL PRESS LOW (H13-P601-16-D2) then alarmed.

This alarm indicates a failure or malfunction of the __(1)__.

The Division 3 Diesel Generator __(2)__ continue to operate.

__(1)__ __(2)__

A. AC Soak Back Pump will B. Main Lube Oil Pump will C. AC Soak Back Pump will not D. Main Lube Oil Pump will not

NRC Exam 2013 QUESTION RO 61 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 264000 K6.03 Importance Rating 3.5 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Emergency Generators (Diesel/Jet) : Lube oil pumps EDGs Explanation: Answer B - Receipt of this alarm indicates a failure or malfunction of the Main oil pump.

The Div 3 DG will trip except during a LOCA when this alarm is received.

A - Incorrect - The Soak Back LO pump loss will not cause this alarm.

C - Incorrect - The Soak Back pump loss will not cause this alarm. During a LOCA the DG trip is bypassed and the DG will continue to run.

D - Incorrect - During a LOCA the DG trip is bypassed and the DG will continue to run.

Technical Reference(s): ARI-H13-P601-16 Rev 17 Reference Attached: ARI-H13-P601-16 p 39 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E22B-I.2.2 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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 62 The Unit 2 Service and Instrument Air Compressors (2P51-C001 and 2P52-C001) lost power due to electrical problems.

Unit 1 Service Air Compressor 1P51-C001 is running in Lead.

Subsequently, a malfunction caused the 1P52-F050, SA/IA XCONN VALVE to close.

The 1P52-F050 closed due to __(1)__.

Air pressure in the Unit 1 Instrument Air System __(2)__.

__(1)__ __(2)__

A. blown control power fuse for 1P52-F050 is maintained at normal air pressure B. blown control power fuse for 1P52-F050 lowers until the Unit 1 Instrument Air Compressor auto starts C. low Unit 1 Instrument Air Header is maintained at normal air pressure pressure D. low Unit 1 Instrument Air Header lowers until the Unit 1 Instrument Air pressure Compressor auto starts

NRC Exam 2013 QUESTION RO 62 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 300000 K6.07 Importance Rating 2.5 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Instrument Air System: Valves Instrument Air Explanation: Answer A - The xP52-F050 valve will close on a loss of control power, a loss of breaker control power, or a low pressure in the receiver. A check valve in parallel with the F050 valve allows air pressure to be maintained in the instrument air system.

B - Incorrect - Normally one compressor will be able to supply the entire service and instrument air systems. The air pressure will not lower causing the IAC to auto start.

C - Incorrect - The auto close of the F050 comes from low pressure in the receiver, not the header.

D - Incorrect - The auto close of the F050 comes from low pressure in the receiver, not the header.

Normally one compressor will be able to supply the entire service and instrument air systems. The air pressure will not lower causing the IAC to auto start.

Technical Reference(s): SOI-P51/52 Rev 26 and ONI- Reference Attached: SOI-P51/52 p 4 and P52 Rev 16 ONI-P52 p 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P51_P52-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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 63 Annunciator NCC SURGE TANK LEVEL HIGH on panel H13-P970 has alarmed.

A leak in the on-service ____ would cause this alarm.

A. NCC Heat Exchanger, P43-B001B B. Containment Vessel Chiller, P50-B001C C. Control Rod Drive Hydraulic Pump, C11-C001A D. Fuel Pool Cooling Cleanup Heat Exchanger, G41-B001B

NRC Exam 2013 QUESTION RO 63 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 A1.04 Importance Rating 2.8 K&A: Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including: Surge Tank Level Component Cooling Water Explanation: Answer D - FPCC HX operates at a higher pressure than NCC and leakage from FPCC to NCC would cause a high surge tank level alarm.

A - Incorrect - Service Water operates at a lower pressure (55-60) than NCC (94-123)

B - Incorrect - CV chiller operates at a lower pressure (~75) and is gas.

C - Incorrect - CRD oil coolers operate a lower pressure (~8)

Technical Reference(s): ARI-H13-P970-001 rev 16 & Reference Attached: ARI-H13-P970-001 rev SDM-P43 rev 10 p 67 & SDM-P43 p 16 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P43-B.2 Question Source: Bank # Perry Audit 2010 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 64 A plant startup is in progress.

The reactor is at the point-of-adding-heat.

When Control Rod 20-25 was fully withdrawn, substitute data was entered for Channel 1 position indication.

Later, Control Rod 20-25 lost its Channel 2 position indication.

To continue the plant startup, first, Control Rod 20-25 ____.

A. position must be bypassed in both RC&IS RACS cabinets B. must be inserted to position 46 from the Operator Control Module C. must have Substitute Data entered for Channel 2 position indication D. must be inserted using the Single Rod Insertion (SRI) switches on the HCU

NRC Exam 2013 QUESTION RO 64 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 201005 K3.02 Importance Rating 3.5 K&A: Knowledge of the effect that a loss or malfunction of the Rod Control And Information System (RCIS) will have on following: Reactor startup; BWR-6 RCIS Explanation: Answer A - Since bad data exists on both Channels 1 and 2, the Rod Pattern Controller (RPC) inserts a rod block. The signals must be bypassed in RACS 1 & 2 to remove the constraints of the RPC.

B - Incorrect - Since there is bad data on both channels, the control rod can not be moved from the OCM.

C - Incorrect - Cannot enter substitute data from Channel I because it contains substitute data.

D - Incorrect - SOI-C11(RCIS) P&L prohibit use Scram Test switches for this condition.

Technical Reference(s): SOI-C11(RCIS) Rev 28 & SDM- Reference Attached: SOI-C11(RCIS) pp 6, C11(RC&IS) Rev 9 11, & 12 and SDM-C11(RC&IS) pp 10 & 19 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11_RC&IS-C, F, H, M Question Source: Bank # River Bend 2003 Modified Bank #

New Question History: Previous NRC Exam River Bend 2003 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 65 The plant was operating at rated power when initiation of a manual scram resulted in an ATWS.

The keylock switches for SLC pumps A & B were taken to ON

Both SQUIB CONTINUITY indicating lights are off (de-energized)

SLC Pump A started normally

SLC Pump B lost indication and failed to start Based on the conditions listed above the direct response of the Reactor Water Cleanup System is that ____.

A. only the 1G33-C001A, RWCU Pump will trip B. both 1G33-C001A and 1G33-C001B, RWCU Pumps will trip C. only the 1G33-F001, RWCU SUCT FM CNTMT INBD ISOL will isolate D. both 1G33-F001, RWCU SUCT FM CNTMT INBD ISOL and 1G33-F004, RWCU SUCT FM CNTMT OTBD ISOL will isolate

NRC Exam 2013 QUESTION RO 65 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 204000 K4.04 Importance Rating 3.5 K&A: Knowledge of Reactor Water Cleanup System design feature(s) and/or interlocks which provide for the following: System isolation upon-receipt of isolation signals RWCU Explanation: Answer D - Taking the SLC A pump control switch to ON causes the G33-F004 valve to close and taking the SLC B pump control switch to ON causes the G33-F001 valve to close, even though the B SLC pump failed to start.

A & B - Incorrect - The RWCU pumps trip on the F001 / 4 valves closing, not directly from initiation of SLC C - Incorrect - Both F001 & F004 valves automatically close.

Technical Reference(s): SOI-G33 Rev 33 & SDM-G33 Reference Attached: SOI-G33 pp 5, 32, & 33 Rev 9 & SDM-G33 p 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-G33_G36-C.1 & F.1 Question Source: Bank #

Modified Bank # Fermi 2006 New Question History: Previous NRC Exam Fermi 2006 #RO-66 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 66 The plant was operating at rated power when the following occurred:

The reactor was scrammed 20 minutes ago due to a problem with the pressure regulator system

During the transient, both Reactor Recirculation Pumps tripped off

Current RPV pressure is 600 psig and lowering at 2.5 psig/minute.

The scram has just been reset

RPV Bottom Head Drain temperature is 445°F

RPV Vessel Head Flange temperature is 500°F Based on these conditions, bulk RPV water temperature currently is approximately__(1)__. If cooldown is allowed to continue at the present rate, Tech Spec cooldown rate __(2)__ Exceeded Reference Provided:

__(1)__ __(2)__

A. 44°F > Bottom Head Drain will B. 16°F < Vessel Head Flange will C. 44°F > Bottom Head Drain will not D. 16°F < Vessel Head Flange will not

NRC Exam 2013 QUESTION RO 66 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 216000 K5.11 Importance Rating 3.2 K&A: Knowledge of the operational implications of the following concepts as they apply to Nuclear Boiler Instrumentation: Indicated vessel temperature response during rapid heatups or cooldowns Nuclear Boiler Inst.

Explanation: Answer C - This is a steam table question. The saturated temperature of 600 psig (~615 psia) is ~490°F (489), which is about 44°> the bottom head temp in this case. Additionally, with pressure lowering at 2.5 psig/min, the pressure at 1 hr after the scram will be 500 psig which corresponds to

~470°F. With initial pressure at 1024 psig, Tsat is 549°F. Thus not exceeding 100°F/hr Each hour thereafter, the cool-down rate is l<100°F/hr A - Incorrect - Cool-down rate will not be exceeded.

B - Incorrect - Cool-down rate will not be exceeded. And, 16°F < Vessel Head Flange is temperature calculated if psia is calculated backwards.

D - Incorrect - 16°F < Vessel Head Flange is temperature calculated if psia is calculated backwards.

Post Exam comments revised correct answer to A vice C.

Technical Reference(s): ABB Steam Tables Reference Attached: x Proposed references to be provided to applicants during examination: Steam Tables Learning Objective (As available): OT-COMBINED-B21(INST)-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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 67 The plant has experienced a LOCA and the following plant conditions exist:

Reactor Level - 25

Time Reactor Level below TAF 28 minutes

Containment Pressure 10 psig

Containment Hydrogen Concentration 7.5%

Drywell Hydrogen Concentration 8.5%

The Hydrogen Igniters failed to energize.

As the RO, you have been directed to startup the Combustible Gas Mixing Compressors per the SOI-M51-56 Combustible Gas Control System and Hydrogen Igniters The Combustible Gas Mixing Compressors should ____.

Reference Provided:

A. not be started because Drywell HDOL has been exceeded B. be started because level has been below TAF for < 30 minutes C. be started because Containment HDOL has not been exceeded D. not be started because level has been below TAF for > 15 minutes

NRC Exam 2013 QUESTION RO 67 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 223001 K6.08 Importance Rating 3.3 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Primary Containment System And Auxiliaries: Containment atmospheric control Primary CTMT and Aux.

Explanation: Answer C - Override LPC/L-3 in EOP-1A directs starting hydrogen gas mixing compressors if both DW H2 concentration is > 1% and containment is below HDOL.

A - Incorrect - DW HDOL is 9%. As given in stem DW HDOL has not been exceeded.

B - Incorrect - The Hydrogen Control Hardcard directs starting the H2 Igniters if <TAF for <30 minutes- it does not address the mixing compressors.

D - Incorrect - Perry License Commitments allow 30 min to start H2 igniters - not compressors.

Technical Reference(s): EOP-1A Bases Rev 4 & EOP- Reference Attached: EOP-1A Bases pp 32-33 SPI Supplement Rev 3 & EOP-SPI Supplement p 10 Proposed references to be provided to applicants during examination: Modified EOP-SPI Supplement Figure #7 HDOL Learning Objective (As available): OT-3402-03-D.2 Question Source: Bank #

Modified Bank # Perry 2009 New Question History: Previous NRC Exam Perry 2009 #RO-69 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 68 A Refueling Outage is in progress.

The Reactor Mode Switch is in the REFUEL position.

Fuel movement is in progress Fuel assembly in reactor location 25-38 is seated in the reactor and grappled.

Which of the following will automatically prevent removing a fuel assembly from the RPV in quadrant A?

A. SRM A not fully inserted.

B. Control Rod 30-31 at position 02.

C. The Hoist Loaded light is extinguished.

D. Zero pounds air pressure in the Refuel Bridge air receivers.

NRC Exam 2013 QUESTION RO 68 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 234000 A1.03 Importance Rating 3.4 K&A: Ability to predict and/or monitor changes in parameters associated with operating the Fuel Handling Equipment controls including: core reactivity level Fuel Handling Equipment Explanation: Answer B - HOIST REFUEL INTERLOCK - If at least one Control Rod not full in and the Main Hoist Loaded indicator on the Operating Console is ON and the platform limit switch indicates the Refuel Platform is over the vessel, then power to the fuel hoist motor is interrupted.

A - Incorrect - This is an administrative requirement, not automatic interlock.

C - Incorrect - The main hoist loaded light going out indicates there is <350 pounds on the hoist. This can be confused with the slack cable interlock which interrupts power to the hoist motor if < 15 pounds on hoist.

D - Incorrect - This will prevent ungrappling the fuel bundle but not lifting it. The hoist is electric motor, not air motor driven.

Technical Reference(s): SOI-F15 Rev 16 Reference Attached: SOI-F15 p 126 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-F11_F15-F Question Source: Bank #

Modified Bank # INL-2165 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 69 The plant scrammed two days ago following a 15 month run.

The cause of the scram was corrected.

A plant startup was in progress with the following conditions:

RPV pressure is 940 psig

Reactor power is 2%

Main Turbine warm-up in progress Then, an automatic reactor scram occurred and all control rods fully inserted.

During the transient the following conditions were observed:

Reactor pressure lowered to 795 psig

Reactor pressure is currently 935 psig and rising

Condenser Vacuum is 20.5 HgA and degrading Based on these conditions, the MSIVs are currently __(1)__. Reactor pressure is being controlled using __(2)__.

__(1)__ __(2)__

A. open Turbine Bypass Valves B. open SRVs C. shut SRVs D. shut RCIC

NRC Exam 2013 QUESTION RO 69 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 239001 A2.08 Importance Rating 3.6 K&A: Ability to (a) predict the impacts of the following on the Main And Reheat Steam System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low condenser vacuum Main and Reheat Steam Explanation: Answer B - With turbine warmup in progress, the Mode Switch will be in Startup. The MSL 807 psig MSIV closure is bypassed. Therefore, the MSIVs will be open. However, since condenser vacuum is >20 inHgA, the Bypass Valves are closed. The Rx Scram Hardcard tells the operator to stabilize pressure using the BPVs or the SRVs. Pressure control will be on the SRVs.

A - Incorrect - BPVs close at 20 HgA.

C & D - Incorrect - MSIVs are open.

Technical Reference(s): OAI-1703 rev 14, ARI-H13- Reference Attached: OAI-1703 p 35, ARI-P601-19 Rev 14 & ONI-N62 rev 9 H13-P601-19 p 7 & ONI-N62 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21_N11-F & OT-3402-02-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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 70 Plant power was lowered from rated to approximately 50% power between noon and 13:00.

Load Set was not adjusted during the power reduction.

A problem with the Stator Cooling Water Temperature Control Valve caused temperature to rise before being corrected.

Annunciator H13-P680-08-B6, LOAD SET RUNBACK STATOR CLG alarmed.

Below is the Stator Water generator outlet temperature trend.

Time Temperature Trend 13:07:00 79.0°C Rising 13:07:30 79.4°C Rising 13:08:00 80.0°C Rising 13:08:30 80.5°C Rising 13:09:00 81.0°C Rising 13:09:30 81.4°C Peak 13:10:00 81.0°C Lowering 13:10:30 80.4°C Lowering 13:11:00 80.0°C Lowering 13:11:30 77.5°C Lowering 13:12:00 77.3°C Stable Based on the data above, generator load at time 13:15 would be approximately ____ MWe A. 650 B. 390 C. 325 D. 0

NRC Exam 2013 QUESTION RO 70 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 241000 A3.17 Importance Rating 3.3 K&A: Ability to monitor automatic operations of the Reactor/Turbine Pressure Regulating System including: Turbine runback Reactor/Turbine Pressure Regulator Explanation: Answer A - A SWC runback is initiated when SWC inlet temp is >81°C and stops when either generator load reaches 9900amps (25% load) or the runback signal clears (temp <81°C). Since Load Set was not adjusted during the power reduction, the runback had to runback the load set motor from ~1450 MWe at a rate of 1%/3 seconds. SW temperature was only >81°C for 60 seconds. Therefore the Load set motor was only run back 20%. - No change in generator load.

B - Incorrect - This corresponds to the 25% no-liquid cooling load.

C - Incorrect - This corresponds to 20% runback if load set were just above generator load when the runback occurred.

D - Incorrect - Plausible because if a runback is initiated from high power, the plant will scram on high Rx pressure.

Technical Reference(s): ARI-H13-P680-08 Rev 13, Reference Attached: ARI-H13-P680-08 pp 21 SDM-N32/C85 Rev 6 & LP OT-COMBINED-N32_C85 and 22, SDM-N32/C85 p 119 & LP OT-Rev 2 COMBINED-N32_C85 pp 9 & 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x 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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 71 The Main Generator synchronization is in the progress IAW IOI-0003, Power Changes.

The following indications are observed on panel H13-P680:

SYNC SELECT SWITCH is in the S610-PY-TIE position

MAIN TRANSFORMER (incoming) S11-R013 346 KV

PY-EL-LINE (running) N41-R120 344 KV

Synchroscope is stopped at the 6:00 position Before the S610-PY-TIE breaker can be closed, the operator must go to __(1)__ on the Auto Voltage Regulator to match voltage. He must also go to __(2)__ on the Load Selector pushbuttons until the Synchroscope is moving slowly in the clockwise direction.

__(1)__ __(2)__

A. LOWER INCREASE B. LOWER DECREASE C. RAISE DECREASE D. RAISE INCREASE

NRC Exam 2013 QUESTION RO 71 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 245000 A4.02 Importance Rating 3.1 K&A: Ability to manually operate and/or monitor in the control room: Generator controls Main Turbine Gen. / Aux.

Explanation: Answer A - With voltage on the Main Transformer at 346 Kv and on the EL line at 344 Kv, the voltage regulator needs to be lowered. With the synchroscope stopped, the generator frequency exactly matches the grid frequency. The Load Selector (Load Set) needs to be increased to cause the turbine to go slightly faster than the grid. This will cause the synchroscope to move in the CW direction.

B, C, & D - Incorrect - See Answer A explanation.

Technical Reference(s): IOI-3 Rev 47 Reference Attached: IOI-3 pp 36-38 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N41_N51-O Question Source: Bank # INL-3127 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 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 72 The plant was operating at 80% power when the following occurred:

Annunciator HOT SURGE TANK LEVEL HI, H13-P680-02-E2 alarmed

Hot Surge Tank level indicates 132 and rising The first action the Reactor Operator needs to take to restore HST level is to adjust the __(1)__.

The valve(s) that respond(s) to restore HST level is/are __(2)__.

__(1)__ __(2)__

A. HOT SURGE TANK LEVEL only the Hot Surge Tank Level Control Controller, 1N21-R475 Valve, 1N21-F230 B. HOT SURGE TANK LEVEL both the Hot Surge Tank Level Control Controller, 1N21-R475 Valve 1N21-F230 and the Hot Surge Tank Level Control Bypass Valve, 1N21-F220 C. HST LVL CV Manual Control only the Hot Surge Tank Level Control switch/potentiometer, 1N21-R708 Valve, 1N21-F230 D. HST LVL CV Manual Control both the Hot Surge Tank Level Control switch/potentiometer, 1N21-R708 Valve 1N21-F230 and the Hot Surge Tank Level Control Bypass Valve, 1N21-F220

NRC Exam 2013 QUESTION RO 72 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 256000 2.1.32 Importance Rating 3.8 K&A: Ability to explain and apply system limits and precautions Reactor Condensate Explanation: Answer A - Training strategy in the simulator reinforces taking manual control of the HST Level Controll with the 1N21-R475 controller. This is reinforced by the ARI. Per SOI-N21 P&L 2.9 the HOT SURGE TANK LEVEL CONTROL 1N21-F230 controller 1N21-R475 controls only the operation of its respective control valve (1N21-F230).

B - Incorrect - The 1N21-F475 controller does not change the position of the Hot Surge Tank Level Control Bypass Valve, 1N21-F220.

C - Incorrect- The first action ARI says to take is to use the 1N21-F475 controller, not the 1N21-R708 potentiometer.

D - Incorrect - The first action ARI says to take is to use the 1N21-F475 controller, not the 1N21-R708 potentiometer. ARI says to control 1N21-F230 valve not both.

Technical Reference(s): SOI-N21 Rev 21, ARI-H13- Reference Attached: SOI-N21 p 5 ARI-H13-P680-02, Rev 10 P680-02 pp 53 and 54 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT- COMBINED-N21_N61-H 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 b.10 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 73 The plant is operating at rated power with HPCS running in CST to CST Mode for a PMT.

A high temperature in the Aux Building exhaust duct results in isolating Aux Building ventilation.

No Aux Building ventilation operating could result in inadequate cooling to the ____.

A. HPCS and RCIC rooms B. RCIC and RWCU Pump rooms C. HPCS room and the Steam Tunnel D. Steam Tunnel and RWCU Pump rooms

NRC Exam 2013 QUESTION RO 73 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 290001 K1.06 Importance Rating 3.4 K&A: Knowledge of the physical connections and/or cause-effect relationships between Secondary Containment and the following: Auxiliary building isolation; BWR-6 Secondary CTMT Explanation: Answer D - IAW with EOP-3 Bases, Perry has an expanded functional Secondary Containment which includes the Intermediate Building, Aux building, Annulus, and Steam Tunnel. Aux Building ventilation takes a suction from the steam tunnel and is the only vent supply for the RWCU pump rooms. SOI-M38/47 P&L 2.2, losing AB ventilation may result in inadequate cooling to the steam tunnel and RWCU pump rooms..

A - Incorrect - Both HPCS and RCIC have room coolers.

B - Incorrect - The RCIC room has its own room cooler supplied from ECC.

C - Incorrect - The HPCS room has its own room cooler supplied from HPCS-ESW.

Technical Reference(s): EOP-3 Bases Rev 3 & SOI- Reference Attached: EOP-3 Bases p 7 &

M38/47 Rev 6 SOI-M38/47 p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M38-H 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 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 74 The plant is operating at rated power with the following conditions:

Control Room HVAC train A is operating in Normal Mode

Control Room HVAC train B is Standby A Control Room HVAC High radiation condition was sensed in the supply duct.

IAW SOI-M25/26, Control Room HVAC And Emergency Recirculating System, the BOP Operator has verified that the CONT RM HVAC TRAIN A and B MODE SELECT switches are selected to the same Mode as the current operating Mode of the Control Room HVAC system.

Which of the following system line-ups is correct for the given conditions?

Supply Fan Return Fan M25-C001A M25-C002A A. On On B. On Off C. Off On D. Off Off

NRC Exam 2013 QUESTION RO 74 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 290003 K1.01 Importance Rating 3.4 K&A: Knowledge of the physical connections and/or cause-effect relationships between Control Room HVAC and the following: Radiation monitors Control Room HVAC Explanation: Answer B - With a High radiation condition was sensed in the supply duct, the CR HVAC system will shift automatically to Emergency Recirc. When in ER, the Supply Fan continues to run, the Return Fan trips, A - Incorrect - The Return Fan trips.

C - Incorrect - The Supply Fan continues to run.

D - Incorrect - The Supply Fan continues to run.

Technical Reference(s): SOI-M25/26 rev 22 & ARI-H13- Reference Attached: SOI-M25/26 p 64 & ARI-P904-02 Rev 10 H13-P904-02 p 71 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M25_26-B Question Source: Bank #

Modified Bank # INL-0552 New Question History: Previous NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.7 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION RO 75 The plant was operating at 100% rated power when a failure of both Pressure Regulators occurred causing the following annunciators to alarm:

RX PRESS HI. H13-P680-0007-D1

RPS RX PRESS HI. H13-P680-0005-A8

RRCS RX PRESS HI. H13-P680-0005-A2 Due to other equipment failures, the SRVs operated only on Spring Set Pressure.

Based on this information __(1)__ were/was exceeded. To control RPV pressure, use the __(2)__.

__(1)__ __(2)__

A. the Reactor Coolant System Pressure SRVs Safety Limit and the Reactor Steam Dome Pressure Technical Specification limit B. the Reactor Coolant System Pressure MAX COMBINED FLOW LIMIT Safety Limit and the Reactor Steam Dome Pressure Technical Specification limit C. only the Reactor Steam Dome Pressure SRVs Technical Specification limit D. only the Reactor Steam Dome Pressure MAX COMBINED FLOW LIMIT Technical Specification limit

NRC Exam 2013 QUESTION RO 75 Level: RO SRO Tier # 2 3 Examination Outline Cross-Reference Group # 2 K/A# 290002 A2.02 Importance Rating 3.6 K&A: Ability to (a) predict the impacts of the following on the Reactor Vessel Internals; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Over pressurization transient Reactor Vessel Internals Explanation: Answer C - Only the TS limit for steam dome pressure (1045 psig) will be exceeded based on annunciator for RRCS pressure high (1083 paig) and the SRVs opening on spring set pressure -

safety mode (highest setpoint -1190 psig). USAR Chapter 15 section 15.2.1.4.2 indiscates the peak pressure for this analyzed event is 1180 psig which is below the Safety Limit value of 1325 psig. SRVs would be manually opened to control RPV pressure.

A - Incorrect - The max RPV pressure should be 1180 psig which is below the SL of 1325 psig.

B - Incorrect - The max RPV pressure should be 1180 psig which is below the SL of 1325 psig.

Additionally, the Max Combined Flow Limit would not work for this type of regulator failure.

D - Incorrect - The Max Combined Flow Limit would not work for this type of regulator failure.

Technical Reference(s): TS 2.1.2, TS 3.4.12, ONI-C85 Reference Attached: TS 2.1.2 p 2.0-1, TS Rev 0, ARI-H13-P680-05 Rev 13, ARI-H13-P680-07 Rev 3.4.12 p 3.4-32, ONI-C85 p 11, ARI-H13-20, & USAR Chapt 15, 15.2.1.4.2 Rev 12 P680-05 pp 7 & 19, ARI-H13-P680-07 pp 89,

& USAR Chapt 15 p 15.2-6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-06(LP)-A.1, OT-3037-03-G and OT-3037-08-B Question Source: Bank #

Modified Bank # Nine Mile 2002 New Question History: Previous NRC Exam Nine Mile 2002 RO-87 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x

NRC Exam 2013 QUESTION SRO 1 A loss of Hot Surge Tank level occurred.

The following conditions now exist:

Operating in EOP-01, RPV Control

RPV level is - 46 inches and lowering

RPV pressure is 5 psig

HPCS pump shaft has broken.

LPCS pump is tagged out for motor replacement with motor removed

RHR A pump is degraded

EH12 has a Bus lockout

No Alternate Injection Subsystems can be lined up

The EOF is operational As the Shift Manager, you would notify the Emergency Response Organization that entry into

__(1)__ is required.

EOP actions are __(2)__ after the SAGs are entered.

__(1)__ __(2)__

A. SAG-1, Primary Containment Flooding Continued B. SAG-1, Primary Containment Flooding Exited C. SAG-2, RPV, Containment, and Continued Radioactivity Release Control D. SAG-2, RPV, Containment, and Exited Radioactivity Release Control

NRC Exam 2013 QUESTION SRO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.20 Importance Rating 4.6 K&A: Ability to interpret and execute procedure steps.

Generic Explanation: Answer B - SRO must know that with the given conditions, EOP-01 requires entry into SAG-1 is required. The shift manager is responsible for notifying the ERO that entry into SAGs is required. EOP actions are exited when SAGs are entered.

A - incorrect - the EOP Actions are discontinued when SAGs are entered.

C - incorrect - SAG-2 entry is not required. EOP actions are exited when SAGs are entered.

D incorrect - SAG-2 entry is not required.

Technical Reference(s): EOP-01 Bases Rev 3 Reference Attached: EOP-01 p 61 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3410-01-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.5 Comments: Level of Difficulty = x - E. 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

NRC Exam 2013 QUESTION SRO 2 You are an off-shift SRO.

Which of the following schedules would allow you to maintain proficiency as an SRO?

All shifts are full shifts with turnovers included.

1 Jul 4 Jul 5 Jul 21 Jul 31 Aug 8 Aug 30 Sep 1 US US SM SE ATC SM US 12 hrs 12 hrs 12 hrs 8 hrs 12 hrs 12 hrs 8 hrs 2 Mar. 31 Apr. 6 Apr. 10 May 1 May 21 Jun 1 Jun 21 US ATC BOP US ATC SE SE 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 3 Jan 3 Jan 8 Feb 15 Feb 16 Feb 28 Mar 18 Apr 1 US ATC BOP ATC BOP SE US 8 hrs 8 hrs 8 hrs 8 hrs 8 hrs 8 hrs 8 hrs 4 Jul 4 Jul 5 Jul 21 Jul 31 Aug 8 Aug 30 Sep 1 FS US ATC BOP ATC BOP FS 8 hrs 12 hrs 12 hrs 12 hrs 12 hrs 12 hrs 8 hrs A. 1 only B. 2 only C. 3 and 4 only D. 1 and 4 only

NRC Exam 2013 QUESTION SRO 2 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.4 Importance Rating 3.8 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 - Correct - Per NUREG 1021, ES-601, need 5-12 hr or 7-8 hr per calendar quarter to maintain proficiency. Also need at least one watch in the SRO position to maintain the SRO portion of the license. Both schedules 1 & 4 contain the correct number of watches.

A - incorrect - schedule #4 would also allow to maintain proficiency B - incorrect - the US watch is in the wrong calendar quarter the Shift Engineer (SE) watch does not count for SRO proficiency.

C - incorrect - schedule #3 does not contain enough watches as the SE does not count for proficiency.

Technical Reference(s): 10CFR55.53.e, NUREG 1021, Reference Attached: 10CFR55.53(e) p 1, ES-605, NOP-OP-1002 Rev 7, TMA-4206 Rev 13, and NUREG 1021, ES-605 p 4, NOP-OP-1002 p TS 5.3 99, and TMA-4206 p 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-2600-01 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: Level of Difficulty = x SRO Only Justification: since this is a proficiency maintenance question for an SRO, the assessment of the combination of watch stations is unique to the SRO position.

SROs are allowed to stand a combination of RO and SRO watches to maintain SRO proficiency.

Whereas ROs must stand only RO position watches.

NRC Exam 2013 QUESTION SRO 3 The Plant is operating at full power.

The NI-Outs NLO reported the following diesel generator air receiver parameters:

Div 1 DG Div 3 DG Right Bank Left Bank Right Bank Left Bank 245 psig 207 psig 225 psig 205 psig Based on the above conditions ____.

Reference Provided:

A. no Tech Spec ACTIONS are required B. the Unit Supervisor would take Tech Spec ACTIONS for Div 1 Starting Air C. the Unit Supervisor would take Tech Spec ACTIONS for Div 3 Starting Air D. the Unit Supervisor would take Tech Spec ACTIONS for Div 1 and Div 3 Starting Air

NRC Exam 2013 QUESTION SRO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.42 Importance Rating 4.6 K&A: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Generic Explanation: Answer C - Per TS Bases, Div 3 requires BOTH air starting air systems for the DG to be operable, while for Div 1 & 2, only one air system is required.

A - incorrect - TS LCO entry for Div 3 DG starting air is required B - incorrect - Div 1 starting air is operable with one receiver > 210 psig D - incorrect - Div 1 starting air is operable with one receiver > 210 psig Technical Reference(s): TS 3.8.3 and TS Bases 3.8.3 Reference Attached: TS 3.8.3 and TS Bases Rev 1, 3, & 7 3.8-41a, 42, & 49 Proposed references to be provided to applicants during examination: TS 3.8.3 (Partial)

Learning Objective (As available): OT-3037-12-C & 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.2 Comments: Level of Difficulty = x - 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).

Knowledge of TS bases that are required to analyze TS required actions and terminology.

NRC Exam 2013 QUESTION SRO 4 While performing a surveillance, the required annunciator did not illuminate.

An I&C technician developed a Simple Troubleshooting Plan to install test equipment to monitor the annunciator relay voltages.

Approval of the Simple Troubleshooting Plan is the responsibility of the ____.

A. Shift Manager B. Plant Manager C. I&C Supervisor D. Unit Supervisor

NRC Exam 2013 QUESTION SRO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.20 Importance Rating 3.8 K&A: Knowledge of the process for managing troubleshooting activities.

Generic Explanation: Answer A - IAW NOP-ER-3001, sect 4.10, the Shift Managers approval is required for the Simple Trouble Shooting Plan.

B - Incorrect - Plant Manger approval is required for more than Simple Troubleshooting Plans C - Incorrect - the I&C supervisor is not responsible for APPROVAL of the troubleshooting plan D - Incorrect - the Unit Supervisor is responsible for normal work release, not troubleshooting.

Technical Reference(s): NOP-ER-3001 Rev 5 Reference Attached: NOP-ER-3001 pp 14-15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): 3039-02-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 55.43 b.5 Comments: Level of Difficulty = x - E. 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.

NRC Exam 2013 QUESTION SRO 5 The following conditions exist:

The plant start up from a forced outage is in progress IAW IOI-1, Cold Startup.

The plant is operating with a known fuel leak.

Reactor Power is on range 8 of the IRMs.

Main condenser vacuum is being maintained between 4 HgA and 5 HgA using the mechanical vacuum pumps.

Annunciator H13-P680-07-A10, AIRBORNE RAD P804 alarmed.

Radiation monitor OG VENT PIPE GAS, D17-K836 exceeded the HIGH alarm setpoint.

The Unit Supervisor will enter __(1)__ and direct __(2)__.

__(1)__ __(2)__

A. ONI-N11, Pipe Break Outside stopping the Mechanical Vacuum Pump Containment IAW SOI-N64/62 Off-Gas/Condenser Air Removal System B. ONI-D17, High Radiation levels Within stopping the Mechanical Vacuum Pump Plant IAW SOI-N64/62 Off-Gas/Condenser Air Removal System C. ONI-N11, Pipe Break Outside isolating the Main Steam Lines IAW Containment SOI-B21 Nuclear Steam Supply Shutoff, Automatic Depressurization And Nuclear Steam Supply Systems D. ONI-D17, High Radiation levels Within isolating the Main Steam Lines IAW Plant SOI-B21 Nuclear Steam Supply Shutoff, Automatic Depressurization And Nuclear Steam Supply Systems

NRC Exam 2013 QUESTION SRO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.11 Importance Rating 4.3 K&A: Ability to control radiation releases.

Generic Explanation: Answer B - IAW ONI-D17, the US should direct the stopping the MVPs when the OG Vent pipe rad monitor exceeds the HIGH alarm setpoint.

A - Incorrect - ONI-N11 is the wrong procedure to enter. This is for a pipe break. No indications of a pipe break were given in the stem. The second part contains the correct action.

C - Incorrect - ONI-N11 is the wrong procedure to enter. This is for a pipe break. No indications of a pipe break were given in the stem. The second part contains the wrong action of closing the main steam lines.

D - Incorrect - the first part is correct, but the second part contains the wrong action of closing the main steam lines.

Technical Reference(s): ONI-D17 Rev 16 and ARI-H13- Reference Attached: ONI-D17 pp 3 & 7 and P680-07 Rev 20 ARI-H13-P680-07 p 15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D17A-M 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: Level of Difficulty = D. 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

NRC Exam 2013 QUESTION SRO 6 The plant was operating at full power, when a transient occurs, resulting in the following conditions:

NO Control Rod movement occurred.

White SCRAM SOL VLS Lights (GP1A, GP1B, GP2A, GP2B, GP3A, GP3B, GP4A, and GP4B) on P680 are ALL OFF.

The Turbine Generator is ON LINE.

Reactor Power is 30%.

ADS is INHIBITED.

The HPCS pump tripped on over-current.

RPV Water Level is at 30 inches, lowering at 4 inches per minute.

Suppression Pool Water Temperature is 90°F and rising, due to RCIC starting.

Which ONE of the following EOP Actions would have highest priority, based on these conditions?

A. Scram And ARI IAW EOP-SPI 1.2 B. Pulling Scram Fuses IAW EOP-SPI 1.1 C. Bypass of MSIV and ECCS Interlocks IAW EOP-SPI 2.3 D. Start one loop of RHR in Suppression Pool Cooling IAW the Hardcard

NRC Exam 2013 QUESTION SRO 6 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.23 Importance Rating 4.4 K&A: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Generic Explanation: Answer C - RPV Level 1 will cause MSIVs to isolate in approximately three minutes. At 30% power, MSIV Closure will remove the main condenser as a heat sink and require dumping steam to the suppression pool. This represents a substantial containment threat as HCL could be reached in a short period of time.

A - Incorrect - plausible; Control Rod insertion is required. However, EOP-SPI 1.2 is used when SOME rod motion has occurred B - Incorrect - plausible; Control Rod insertion is required. SCRAM SOL VLS Lights being OFF indicate Scram Valves are OPEN and efforts to de-energize scram solenoids will be not be effective.

D - Incorrect - Starting Suppression Pool Cooling is not the highest priority. The heat addition from RCIC is minimal compared to the heat addition from SRVs.

Technical Reference(s): EOP-1A Bases Rev 4 Reference Attached: EOP-1A Bases pp 7 &

35 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-04B Question Source: Bank # Fermi 2008 Q100 Modified Bank #

New Question History: Previous NRC Exam Fermi 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 b.5 55.43 Comments: Level of Difficulty = x - E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 7 Consider the following conditions:

Twenty minutes ago the plant was manually scrammed from 50% power due to a fire in bus D-1-A.

Bus D-1-A was de-energized and the fire was out in 15 minutes.

The Shift Manager was able to recruit off-shift day shift personnel for additional control room monitoring.

All control rods inserted.

Reactor level is being controlled using Feedwater at 178.

Reactor pressure is being controlled using Bypass valves.

Unit 1 Plant Vent radiation monitor showed elevated release rates following the scram, but has returned to normal.

Classify the event.

Reference Provided:

A. No classifiable event B. Unusual Event C. Alert D. Site Area Emergency

NRC Exam 2013 QUESTION SRO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.32 Importance Rating 4.0 K&A: Knowledge of operator response to loss of all annunciators.

Generic Explanation: Answer C - a loss of D-1A causes a loss of the majority of the control room annunciators.

The manual scram is considered a significant plant transient and additional monitoring in the control room satisfies the entry conditions for an ALERT.

A - Incorrect - Plausible since the fire is in a non-safe shutdown bldg, the release is less than 60 minutes. And if they dont realize that the annunciators are powered from D-1-A.

B - Incorrect - plausible if they think the fire is in a safe shutdown building.

D - Incorrect - Plausible if they assume sufficient indications are not available to monitor critical plant parameters.

Technical Reference(s): EPI-A1 rev 25 Reference Attached: EPI-A1 p 55 Proposed references to be provided to applicants during examination: EPI-A1 Learning Objective (As available): OT-3035-05(LP)-A.6 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: Level of Difficulty = x - E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 8 The following conditions exist:

The plant is operating at 98% power.

Core flow is 103 Mlbs/Hr.

OPRMs are INOPERABLE

Alternate methods to detect and suppress thermal hydraulic instability oscillations have been initiated IAW 3.3.1.3 OPRM Instrumentation.

Reactor Recirculation Pump B then trips.

In accordance with ONI-C51, Unplanned Changes Reactor Power or Reactivity, the Unit Supervisor will direct ____.

Reference Provided:

A. inserting Cram Rods IAW FTI-B002, Control Rod Movements B. restarting Recirc Pump B IAW SOI-B33, Reactor Recirculation C. inserting a manual reactor scram IAW ONI-C71-1, Reactor Scram D. shutting Recirc Pump B FCV IAW ONI-SPI G-2, Single Pump Operation

NRC Exam 2013 QUESTION SRO 8 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295001 2.1.23 Importance Rating 4.4 K&A: Ability to perform specific system and integrated plant procedures during all modes of plant operation Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Explanation: Answer A - With the plant operating at 103 Mlbm/Hr/ a trip of 1 recirc pump would lower core flow to < 50% rated core flow, putting operation in the Controlled Entry Immediate Exit Region of the Backup Stability Protection (OPRM INOP) P/F map. The US would assess the situation and direct actions IAW the ONI-C51 Flow Chart - inserting Cram Rods to lower power to exit this region. (These are not Immediate Actions (from memory), these are actions directed by the Unit Supervisor.

B - Incorrect - Plausible since ONI-C51 indicates the preferred method to exit the CE/IE region is opposite the way it was entered.

C - Incorrect - Plausible since at lower core flows, a trip of 1 recirc pump will cause entry into the Manual Scram region.

D - Incorrect - Plausible since shutting the Recirc Loop Suction valve is required, but not the FCV.

Post Exam comments DELETED this question.

Technical Reference(s): ONI-C51 Flow Chart Rev J and Reference Attached: ONI-C51 and PDB-A6 p PDB-A6 rev 14 5 Proposed references to be provided to applicants during examination: PDB-A06, Power Flow Map (modified)

Learning Objective (As available): OT-COMBINED-B33-I Question Source: Bank #

Modified Bank # Peach Bottom 2008 # SRO-9 New Question History: Previous NRC Exam Peach Bottom 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 9 Plant startup is in progress with the conditions:

Control Rod withdrawal in progress

Reactor pressure is 280 psig

Turbine Shell Warming is in progress IAW SOI-N32/39/41/51, Main Turbine Generator and Turning Gear System The following occurs:

While throttling Main Stop Valve No. 2 Bypass Valve using the CHEST AND SHELL WARMING INCREASE / DECREASE pushbuttons, the INCREASE pushbutton momentarily stuck depressed

Turbine first stage pressure equalized with reactor pressure

Annunciator H13-P680-05-E5, FAST CLOSE & TURB STOP SCRAM BYP, reset

Automatic actions failed to occur The action the Unit Supervisor would direct to compensate for the failed automatic action is to __(1)__

The basis for the automatic action that failed is to __(2)__

__(1)__ __(2)__

A. Place Mode Switch in SHUTDOWN ensure the Minimum Critical Power Ratio (MCPR) Safety Limit is not exceeded B. Place Mode Switch in SHUTDOWN ensure the fuel peak cladding temperature remains below the limits of 10CRF50.46 C. Depress the CHEST AND SHELL ensure the Main Turbine heat-up rate is WARMING DECREASE pushbutton to not exceeded close the Main Stop Valve No. 2 Bypass Valve D. Depress the CHEST AND SHELL prevent turbine packing damage by WARMING DECREASE pushbutton to limiting turbine speed close the Main Stop Valve No. 2 Bypass Valve

NRC Exam 2013 QUESTION SRO 9 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295006 2.4.50 Importance Rating 4 K&A: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual SCRAM / 1 st Explanation: Answer A - When turbine 1 stage pressure exceeds 212 psig, the annunciator will clear, indicating that power above the LPSP is sensed. With power above the LPSP, a Rx scram should result due to turb stop valves being closed. Since no automatic actions occurred, the US should direct inserting a manual scram. Additionally, the bases for the scram is to prevent exceeding the MCPR safety limit.

B - Incorrect - Correct action, but wrong bases for action.

C - Incorrect - H/U rate limit is 150°F/ hr, however once a scram setpoint limit is exceeded, the action to scram becomes more important than closing the bypass valve.

D - incorrect - Packing damage is a concern if turbine speed exceeds 30 rpm, however once a scram setpoint limit is exceeded, the action to scram becomes more important than closing the bypass valve Technical Reference(s): TS Bases 3.3.1.1 Rev 0, SOI- Reference Attached: TS Bases p B-3.3-16 N32 Rev 25, ARI-H13-P680-05 Rev 13 SOI-N32 p 34 & ARI-H13-P680-05 p 97 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N32_C85-C.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: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

B. 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.

NRC Exam 2013 QUESTION SRO 10 The plant was at rated power when a transient resulted in the following conditions:

Suppression Pool temperature rising

Suppression Pool level rising

Containment temperature rising

Containment pressure rising Conditions continue to deteriorate necessitating containment venting.

Venting Containment disregarding adequate core cooling and irrespective of offsite radioactivity release rate, must be done __(1)__ reaching __(2)__.

__(1)__ __(2)__

A prior to PSP B prior to PCL C after PSP D after PCL

NRC Exam 2013 QUESTION SRO 10 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295038 2.4.21 Importance Rating 4.6 K&A: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

High Off-Site Release Rate Explanation: Answer D - Since this action may result in possible dose to the public, It must be done only after PCL is reached.

A - incorrect - Plausible because EOP-2 venting is required prior to reaching PCL if systems not required for adequate core cooling are available. It also requires ED prior to reaching PSP just prior to venting.

B - incorrect - Plausible because EOP-2 venting is required prior to reaching PCL if systems not required for adequate core cooling are available.

C - incorrect - Plausible because venting containment is required after reaching PSP. However, it only must be done after reaching PCL Technical Reference(s): EOP-2 Rev B, EOP-2 Bases rev Reference Attached: EOP-2 chart, EOP-2 1 Bases pp 57-58 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-05 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: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 11 The plant is operating at 75% power with the following conditions:

Unit 1 Service Air Compressor, 1P51C001 is tagged out for maintenance

Unit 1 Instrument Air Compressor, 1P52C001 is running in Lead Thirty minutes ago Bus H22 experienced a bus lockout.

Currently, Unit 1 Instrument Air Compressor is showing degrading output.

Regarding Service and Instrument Air systems only:

If Unit 1 Instrument Air Compressor is lost, the PSA Risk Category will be __(1)__.

In accordance with NOP-OP-1007, Risk Management, notification of the Duty Team __(2)__

required.

Reference Provided:

__(1)__ __(2)__

A. Green is B. Green is not C. Yellow is D. Yellow is not

NRC Exam 2013 QUESTION SRO 11 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295019 2.1.25 Importance Rating 4.2 K&A: Ability to interpret reference materials, such as graphs, curves, tables, etc.

Partial or Total Loss of Inst. Air / 8 Explanation: Answer C - with Bus H22 locked out, both U2 air compressors are lost. If U1 IAC is lost, all 4 air compressors will be out of service resulting in a loss of air. IAW PDB-C007 this changes the Plant risk from Green to Yellow. IAW NOP-OP-1007, duty team notification is required for a change in risk level.

A - Incorrect - Risk Category will change to Yellow.

B - Incorrect - Risk Category will change to Yellow and Duty Team notification is required.

D - Incorrect - Duty Team notification is required.

Technical Reference(s): NOP-OP-1007 Rev 15, PDB- Reference Attached: NOP-OP-1007 p 23, C0011 Rev 5 PDB-C0011 p 10 Proposed references to be provided to applicants during examination: PDB-C0011, PSA Pre-Solved Configurations for On-Line Risk (Partial)

Learning Objective (As available): OT-3035-14(LP)-A.3 and OT-3060-01-I 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: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 12 The Plant is operating at 90% power.

RCIC is being run in CST to CST mode for a PMT.

The Suppression Pool Average Temperature was recorded as follows:

T1 94.0 °F T2 96.2 °F T3 100.5 °F T4 102.3 °F T5 105.0 °F T6 106.0 °F The RCIC turbine was tripped at T3.

Based on the above data, the Unit Supervisor would enter Tech Spec 3.6.2.1 ____.

Reference Provided:

A. Condition A @ T2 B. Condition A @ T3 C. Condition C @ T5 D. Condition C @ T6

NRC Exam 2013 QUESTION SRO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 EA2.01 Importance Rating 4.2 K&A: Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature: Suppression pool water temperature.

Suppression Pool High Water Temp. / 5 Explanation: Answer B - Per Tech Spec Bases once testing is suspended, the higher SP temperature limit is no longer in effect and Condition A is entered if Temperature is > 95°F.

A - Incorrect - At T2 testing is still being performed, so Condition A does not apply.

C - Incorrect - At T5 Condition A is applicable, but Condition C would be applicable at > 105° if testing were still being performed.

D - incorrect - At T6 Condition C would be applicable at > 105° if testing were still being performed.

Technical Reference(s): TS 3.6.2.1 & TS 3.6.2.1 Bases Reference Attached: TS 3.6.2.1 pp 3.6-36 Rev 1 and 37 & B 3.6-73 Proposed references to be provided to applicants during examination: TS 3.6.2.1 - Partial Learning Objective (As available): TO-3037-07-H 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: Level of Difficulty = x - B. 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.

NRC Exam 2013 QUESTION SRO 13 While operating at rated power, the plant experienced a LOCA and a LOOP resulting in the following conditions:

Drywell pressure 4.0 psig - stable

Drywell temperature 220°F - Stable

Rx Pressure 800-1000 psig on SRVs

The lowest RPV water level 110 inches

RCIC tripped

HPCS did not auto start

HPCS was manually started for level control

RPV water level has been stabilized This event meets the conditions for a/an ____ notification(s) to the NRC Operations Center.

Reference Provided:

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months only B. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months only C. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months only D. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

NRC Exam 2013 QUESTION SRO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295028 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.

High Drywell Temperature / 5 Explanation: Answer D - 1 hr - E-Plan (Hi DW temp & pressure indicate loss of RCS barrier AA2), 4 hr

- ECCS system Injection, & 8 hr - ECCS System actuation A - Incorrect - had ECCS Actuation B - incorrect - had E-Plan entry C - Incorrect - had ECCS system injection Technical Reference(s): EPI-A1 Rev 25, PAP-1604 Rev Reference Attached: EPI-A1 p 19, PAP-1604 24 pp. 26, 29, 30, 33, 35, & 36 Proposed references to be provided to applicants during examination: PAP-1604, Reports Management

& EPI-A1, Emergency Action Levels Learning Objective (As available): OT-3039-01-A Question Source: Bank # Vermont Yankee 2005 Modified Bank #

New Question History: Previous NRC Exam Vermont Yankee 2005 #83 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

NRC Exam 2013 QUESTION SRO 14 The plant was operating at rated power.

A manual Rx scram was inserted due to a loss of main condenser vacuum.

The transient resulted in the following conditions:

An ATWS is in progress

MSIVs are shut

RPV water level is 65 inches and stable

RPV pressure is 960 psig and stable

2 SRVs are open

Suppression Pool Temperature is 116°F and rising slowly

Suppression Pool Level 18.1 and lowering due to a leak

The margin to exceeding HCL is 3°F What is the action that the Unit Supervisor would order first?

A. Open an additional SRV to lower RPV pressure B. Open MSIVs IAW EOP-SPI 9.2, Opening MSIVs C. Transition to EOP 4-2, Emergency Depressurization D. Anticipate Emergency Depressurization IAW EOP-02, Containment Control

NRC Exam 2013 QUESTION SRO 14 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295030 EA2.02 Importance Rating 3.9 K&A: Ability to determine and/or interpret the following as they apply to Low Suppression Pool Water Level: Suppression pool temperature Low Suppression Pool Wtr Lvl / 5 Explanation: Answer A - With SP temp rising and SP level lowering, the margin to HCL is shrinking. IAW EOP-01A, the SRO would direct the RO to lower RPV pressure by opening an additional SRV to maintain margin to HCL.

B - Incorrect - With the main condenser not available, opening the MSIVs would not be appropriate.

C - Incorrect - ED would only be performed if SP temperature could not be restored and maintained below HCL.

D - Incorrect - Anticipating ED is not appropriate during an ATWS.

Technical Reference(s): EOP-1A Chart Rev D Reference Attached: EOP-1A Chart (partial)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-04B-D.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: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

NRC Exam 2013 QUESTION SRO 15 The plant was operating at full power, when the following occurred:

Both Feedwater Pump Turbines tripped.

The Motor Feed Pump failed to start.

The reactor automatically scrammed.

One Control Rod is at position 48.

All other Control Rods are fully inserted.

HPCS initiation raised RPV Water Level from 110 inches.

HPCS was manually overridden OFF as RPV Water Level reached 210 inches.

Current Plant conditions are:

Reactor pressure 700 psig, rising at 10 psig per minute.

MSIVs are open.

The operating CRD Pump tripped.

Over the next ten minutes RPV Water Level will __(1)__.

The procedure used to control RPV Water Level is __(2)__.

__(1)__ __(2)__

A. rise due to swell EOP-1, RPV Control B. rise due to swell EOP-1A, Level Power Control C. lower due to shrink EOP-1, RPV Control D. lower due to shrink EOP-1A, Level Power Control

NRC Exam 2013 QUESTION SRO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295008 AA2.05 Importance Rating 3.1 K&A: Ability to determine and/or interpret the following as they apply to High Reactor Water Level: Swell High Reactor Water Level / 2 Explanation: Answer A - HPCS injected100 inches x 250 gal per inch= 25,000 gallons of cold CST water. As this water is heated, SWELL occurs. Shrink cannot occur because heatup and pressurization of saturated system is in progress with NO steam voids. Since the nominal assigned pressure band is 800-1000 psig, all SRVs and BPVs are shut for the next ten minutes because Reactor Pressure will be below 800 psig. With only 1 control rod out, the reactor is shutdown and EOP-1 is the appropriate procedure.

B - Incorrect - Plausible since EOP-1A is for ATWS and 1 rod is withdrawn. However, SHUTDOWN is defined as all rods in except for 1.

C - Incorrect - RPV level will rise due to swell.

D - Incorrect - RPV level will rise due to swell. And EOP-1A is for ATWS and 1 rod is withdrawn.

However, SHUTDOWN is defined as all rods in except for 1.

Technical Reference(s): GFE Chap. 7 Rev 4, EOP Reference Attached: GFE Chap. 7 p 28, EOP Bases Rev 3, EOP-1A Chart Rev D & EOP-1 Chart Rev Bases pp 46-47, EOP-1A Chart (partial) &

D EOP-1 Chart (partial)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3303-07 & OT-3302-04A Question Source: Bank #

Modified Bank # Fermi 2008 New Question History: Previous NRC Exam Fermi 2008 #83 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: Level of Difficulty = x Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

NRC Exam 2013 QUESTION SRO 16 The plant is operating at rated power.

A non-isolable leak has resulted in suppression pool level being above the upper limit established by the Technical Specifications and the level is still slowly rising.

Per the Bases for Technical Specification 3.6.2.2, Suppression Pool Water Level, with suppression pool level above the upper limit:

A. RCIC may trip on high exhaust back-pressure.

B. The peak drywell design pressure may be exceeded during a design basis LOCA.

C. The peak containment design pressure may be exceeded during a design basis LOCA.

D. There could be excessive hydrodynamic loads on submerged structures during SRV and horizontal vent steam discharges.

NRC Exam 2013 QUESTION SRO 16 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295029 EA2.02 Importance Rating 3.6 K&A: Ability to determine and/or interpret the following as they apply to High Suppression Pool Water Level: Reactor pressure High Suppression Pool Wtr Lvl / 5 Explanation: Answer D - 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.

A - Incorrect - bases document does not state the RCIC turbine may trip with a high suppression pool level but is credible because with a higher SP water level, RCIC back pressure would be higher.

B - Incorrect - the bases document does not state the drywell design pressure could be exceeded with a high suppression pool level but is credible because drywell pressure would be higher given a DBA LOCA and a higher SP water level.

C - Incorrect - the bases document states the containment design pressure would not be exceeded with a high suppression pool level.

Technical Reference(s): TS 3.6.2.2 Bases Rev 7 and Reference Attached: TS 3.6.2.2 Bases p B Lesson Plan OT-3037-002-10 Rev 2. 3.6-76 & Lesson Plan OT-3037-002-10 p 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-10-B Question Source: Bank # Grand Gulf 2008 Modified Bank #

New Question History: Previous NRC Exam Grand Gulf 2008 # SRO1 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: Level of Difficulty = x 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.

NRC Exam 2013 QUESTION SRO 17 The plant was operating at rated power when the following annunciator alarmed:

H13-P680-01-C5, RWCU ISOL PUMP A/B RM TEMP HI Upon investigation, the RO reports the following:

1E31-N700A A1-1, RWCU Pump A Room ventilation differential temperature is tripped and reading 32°F

1E31-N700A A2-1, RWCU Pump A Room ambient temperature is not tripped and reading 133°F and rising slowly Based on this information, the Unit Supervisor will __(1)__.

The Technical Specification basis for the RWCU area high temperature trip function is to __(2)__.

__(1)__ __(2)__

A. Immediately enter EOP-3, Secondary Serve as a backup to RPV Level 2 trip Containment Control B. Immediately enter EOP-3, Secondary Limit offsite dose rates Containment Control C. Delay entry into EOP-3, Secondary Serve as a backup to RPV Level 2 trip Containment Control until the RWCU A pump room exceeds the ambient temperature trip setpoint D. Delay entry into EOP-3, Secondary Limit offsite dose rates Containment Control until the RWCU A pump room exceeds the ambient temperature trip setpoint

NRC Exam 2013 QUESTION SRO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295032 2.4.2 Importance Rating 4.6 K&A: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

High Secondary Containment Area Temperature / 5 Explanation: Answer B - Ventilation cooler T is entry condition for EOP-3. The TS Bases for the ambient temperature high trip is to limit off site dose.

A - Incorrect - misconception that the ambient temperature high trip is a b/u to the RPV Level 2 trip.

C - Incorrect - Plausible if candidate thinks entry should be delayed until the high ambient trip is reached.

Also, misconception that the ambient temperature high trip is a b/u to the RPV Level 2 trip.

D - Incorrect - Plausible if candidate thinks entry should be delayed until the high ambient trip is reached.

Technical Reference(s): TS 3.3.6.1 Bases Rev 3, EOP-3 Reference Attached: TS 3.3.6.1 Bases pp B Chart Rev C. EOP-3 Bases Rev 3, & ARI-H13-P680-01 3.3-156 & 157, EOP-3 Chart partial. EOP-3 Rev 11 Bases pp 8-10, & ARI-H13-P680-01 pp 35-36 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-07-G & OT-3402-17 Question Source: Bank #

Modified Bank # River Bend 2007 New Question History: Previous NRC Exam River Bend 2007 #84 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 b.2 &

Comments: Level of Difficulty = x 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.

NRC Exam 2013 QUESTION SRO 18 The plant is in shutdown with the following conditions:

Average Reactor Coolant temperature is 190°F.

RHR A loop placed in Shutdown Cooling (SDC) Mode of operation IAW SOI-E12 Based on this information, the LPCI mode of RHR A system is ____.

A. NOT affected, since it is NOT required to be OPERABLE with the current plant conditions B. INOPERABLE, since the RHR Minimum Flow Valve is deenergized closed for SDC Operations C. INOPERABLE, since the system must be manually realigned when required D. OPERABLE, provided the system can be manually realigned when required

NRC Exam 2013 QUESTION SRO 18 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 203000 2.2.25 Importance Rating 4.2 K&A: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

RHR/LPCI: Injection Mode Explanation: Answer D - Per TS 3.5.2 Bases, one LPCI subsystem may be considered operable during alignment or operation for decay heat removal in Mode 4 or 5, if capable of being manually realigned to the LPCI mode.

A - Incorrect - LPCI Mode is required to be Operable in this mode.

B - Incorrect - LPCI subsystem is considered operable under this condition.

C - Incorrect - LPCI subsystem is considered operable under this condition.

Technical Reference(s): TS 3.5.2, TS 3.5.2 Bases Rev 7, Reference Attached: TS 3.5.2, TS 3.5.2 Bases p B 3.5-15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-09-B Question Source: Bank #

Modified Bank # LaSalle 2003 New Question History: Previous NRC Exam LaSalle 2003 #102 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: Level of Difficulty = x 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.

HO - need to determine RHR cut in permissive pressure and mode of operation.

NRC Exam 2013 QUESTION SRO 19 The following conditions exist:

The plant is operating at rated power.

RCIC was shutdown to Secured Status yesterday to change the turbine lube oil.

At 10:30 annunciator H13-P601-81-E1, HPCS Pump Room Level High alarmed.

An NLO reports that the HPCS pump seal is spraying water and water is at the floor grating level.

Based on this information the EOP-3, Secondary Containment Control required action is to isolate HPCS and __(1)__.

The Tech Spec requirement to be in Mode 3 must be met no later than __(2)__.

Reference Provided:

__(1)__ __(2)__

A. wait until a second area water level is 22:30 above Max Safe to shutdown the reactor.

B. enter EOP-1, RPV Control and 23:30 shutdown the reactor C. wait until a second area water level is 23:30 above Max Safe to shutdown the reactor.

D. enter EOP-1, RPV Control and 22:30 shutdown the reactor

NRC Exam 2013 QUESTION SRO 19 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209002 A2.07 Importance Rating 3.0 K&A: Ability to (a) predict the impacts of the following on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump seal failure: BWR-5,6 HPCS Explanation: Answer C - The suppression pool is not a primary system. Therefore, IAW EOP-3, it is required to wait until 2 areas of the same parameter is above the Max Safe to shutdown the reactor. The TS required S/D time to Mode 3 is 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months or 23:30.

A - Incorrect - The TS S/D time is plausible if the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to verify RCIC is operable is not taken.

B - Incorrect - If this were a primary system discharging, entering EOP-1 would be appropriate.

D - Incorrect - The TS S/D time is plausible if the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to verify RCIC is operable is not taken. If this were a primary system discharging, entering EOP-1 would be appropriate.

Technical Reference(s): EOP-3 rev C & TS 3.5.1 Reference Attached: EOP-3 Chart & TS 3.5.1 p 3.5-1,2, & 3 Proposed references to be provided to applicants during examination: TS 3.5.1 (partial)

Learning Objective (As available): OT-3037-09-C & OT-3402-17 Question Source: Bank #

Modified Bank # Dresden 2009 New Question History: Previous NRC Exam Dresden 2009 #87 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 & b.2 Comments: Level of Difficulty = x 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.

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).

NRC Exam 2013 QUESTION SRO 20 The following conditions exist:

The plant is operating at full power

Work was performed on the 1E51-F031, RCIC PUMP SUPR PL SUCT ISOL

1E51-F031 is now down powered

RCIC is in standby with suction aligned to the Condensate Storage Tank A Technical Specification function of RCIC is to __(1)__.

RCIC is __(2)__.

See attached PMT data.

__(1)__ __(2)__

A. maintain coolant inventory, as well as vessel level, if a OPERABLE small break occurs in the RPV while the RCS is still pressurized B. maintain coolant inventory, as well as vessel level, if a INOPERABLE small break occurs in the RPV while the RCS is still pressurized C. operate following RPV isolation accompanied by a loss OPERABLE of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level D. operate following RPV isolation accompanied by a loss INOPERABLE of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level

NRC Exam 2013 QUESTION SRO 20 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 2.2.37 Importance Rating 4.6 K&A: Ability to determine operability and/or availability of safety related equipment.

RCIC Explanation: Answer D - The function of RCIC is to provide water for adequate core cooling if feedwater is isolated. The Suppression Pool suction is INOP because it is down powered. The Tech Spec required suction source is the Suppression Pool. With RCIC aligned to the CST and the SP suction INOP, the TS required source is unable to support the TS required function and RCIC is INOP.

A & C - incorrect - RCIC is INOP due to not being on the SP A & B - incorrect - this is the Tech Spec function of HPCS Technical Reference(s): Tech Spec Bases 3.5.3 rev 5 Reference Attached: Tech Spec Bases p B and Tech Spec Bases 3.5.1 rev 0 3.5-21 & B 3.5-1 Proposed references to be provided to applicants during examination: PMT data Learning Objective (As available): OT-3037-09-B Question Source: Bank # Perry 2010 Modified Bank #

New Question History: Previous NRC Exam Perry 2010 #SRO19 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: Level of Difficulty = x 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.

NRC Exam 2013 QUESTION SRO 21 The following conditions exist:

The plant is at 100% power

I&C is performing surveillance SVI-B21-T0073-C, MAIN STEAM LINE LOW PRESSURE CHANNEL C CALIBRATION FOR 1B21-N076C

All other MSL Low Pressure instruments are OPERABLE

At 08:17 on Feb 2nd, the technician reports to you that he is unable to adjust the 1B21-N076C instrument within the allowable value.

At 08:30 on Feb 2nd, while the BOP Operator was performing Tech Spec Rounds, he reports that trip unit B21-N676A is pegged high.

Based on this information, the Main Steam Line Pressure Low function for MSIV isolation

__(1)__ maintained.

The Required Action(s) is to __(2)__.

Reference Provided:

__(1)__ __(2)__

A. is place C channel in trip by 09:17 on Feb 3rd B. is place A channel in trip by 09:30 on Feb 3rd C. is not place either channel in trip by 09:17 on Feb 2nd to restore isolation capabilities D. is not place either channel in trip by 09:30 on Feb 2nd to restore isolation capabilities

NRC Exam 2013 QUESTION SRO 21 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 A2.05 Importance Rating 3.6 K&A: Ability to (a) predict the impacts of the following on the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Nuclear boiler instrumentation failures PCIS/Nuclear Steam Supply Shutoff Explanation: Answer D - per Tech Spec Bases, sufficient channels do not remain Operable such that a valid signal will not isolate the MSIVs. Therefore, the isolation function is not maintained. Trip capability must restored within one hour. Per TS Bases, placing one channel in trip will restore trip capability.

A - Incorrect - Function is lost. This is the Action for Condition A.

B - Incorrect - Function is lost. This is the Action for Condition A, but the incorrect time.

C - Incorrect - This is the incorrect time for placing a channel in trip Technical Reference(s): Tech Spec 3.3.6.1 & Tech Spec Reference Attached: Tech Spec 3.3.6.1 pp.

Bases B 3.3.6.1 rev 3 & 4, PDB-I5 Rev 9 48 & 54 and Tech Spec Bases p B 3.3-142, 164, & 165, and PDB-I5 pp 17-20 Proposed references to be provided to applicants during examination: Technical Specification 3.3.6.1 (partial) and Plant Data Book Tab I - (partial)

Learning Objective (As available): OT-COMBINED-B21(NS4)-l & OT-3037-07-B.5 Question Source: Bank #

Modified Bank # Perry 2010 New Question History: Previous NRC Exam Perry 2010 #SRO-20 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.2 Comments: Level of Difficulty = x 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.

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

NRC Exam 2013 QUESTION SRO 22 The plant is operating at rated power with the following conditions:

Lake temperature is 79°F

SVI-E12-T2001, RHR A Pump And Valve Operability Test is in progress

Annunciator ECC HX A OUTLET TEMP HI/LO, H13-P601-20-B1 is in alarm

The BOP reports ECC A HX Outlet Temperature is 96°F and stable Based on this information, ECC A loop is ____.

A. OPERABLE. SVI-E12-T2001 may continue B. OPERABLE. Direct the BOP to suspend the SVI until the ECC temperature alarm is reset C. INOPERABLE. Enter the applicable Conditions and Required Actions for the associated systems or components.

D. INOPERABLE. Do not enter the applicable Conditions and Required Actions for the associated systems or components.

NRC Exam 2013 QUESTION SRO 22 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 2.1.32 Importance Rating 4.0 K&A: Ability to explain and apply system limits and precautions.

Component Cooling Water Explanation: Answer C - SOI-P42 P&L # 2.4 states that ECC is inoperable at temperatures >95°F. IAW TS LCO 3.0.6 Bases, the LCO 3.0.6 exception to LCO 3.0.2 does NOT apply. The candidate needs to be familiar with OAI-1701 to formulate the correct answer.

A & B - Incorrect - ECC A is not Operable.

D - Incorrect - the LCO 3.0.6 exception to LCO 3.0.2 does not apply Technical Reference(s): SOI-P42 Rev19, TS 3.7.10, TS Reference Attached: SOI-P42 p 4, TS p 3.7-3.0.2 & 3.0.6, and TS 3.0.6 Bases Rev 4 19, TS p 3.0-1 & 3.0-2, and TS Bases pp 3.0-7&8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-02-A.3 & OT-3037-04-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.2 Comments: Level of Difficulty = x 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).

Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

NRC Exam 2013 QUESTION SRO 23 The following scram insertion time measurements were recorded, per SVI-C11-T1006, Control Rod Maximum Scram Insertion Time.

ROD NOTCH POSITION TIME 34-59 43 0.38 sec 06-27 29 0.96 sec 10-27 13 7.05 sec The measurements were conducted with reactor vessel dome pressure at 1024 psig.

Based on the above results, continued operation in Mode 1 is ____.

Reference Provided:

A. not permitted; Immediately enter LCO 3.0.3.

B. permitted; Verify scram times of at least another 18 rods.

C. not permitted; Place the plant in Hot Shutdown in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. permitted; Fully insert rod 10-27 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarm it within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

NRC Exam 2013 QUESTION SRO 23 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201003 2.1.7 Importance Rating 4.7 K&A: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Control Rod and Drive Mechanism Explanation: Answer D - rod 10-27 is Inoperable and not slow per the Note in TS Table 3.1.4-1. This requires entering LCO 3.1.3 Condition C.

A - Incorrect - LCO 3.0.3 not required since a Required Action applies.

B - Incorrect - Placing plant in hot shutdown does not satisfy Tech Specs and is not required.

C - Incorrect - Not required since 10-27 is not considered slow, thus there are not two slow rods adjacent to each other.

Technical Reference(s): SVI-C11-T1006 Rev 16, TS Reference Attached: TS 3.1 pp 12-14 & 7-9 3.1.4 & 3.1.3 Proposed references to be provided to applicants during examination: TS 3.1.4 & 3.1.3 (partial)

Learning Objective (As available): OT-3037-05-E Question Source: Bank # RQL-0536 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: Level of Difficulty = x 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).

NRC Exam 2013 QUESTION SRO 24 The following conditions exist:

The plant has been operating at rated power for 105 days when main turbine vibrations increased. The decision was made to take the turbine offline and insert a balance shot.

A plant down power to 15% Rated Thermal Power is in progress to remove the generator from the grid.

As power was lowered to 22% RTP, annunciator H13-P680-05-C9, PWR BELOW LPSP alarmed.

Based on this information, per Tech Specs, SVI-C11-T1019, Rod Pattern System - Rod Pattern Controller must be performed within one hour of __(1)__.

Per Tech Spec Bases, the Rod Pattern Controller will __(2)__

__(1)__ __(2)__

A. receiving the PWR BELOW LPSP alarm initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the specified sequence B. lowering reactor power to 19% rated initiate control rod withdrawal and insert thermal power blocks when the actual sequence deviates beyond allowances from the specified sequence C. receiving the PWR BELOW LPSP alarm prevent a violation of the MCPR Safety Limit that may result from a single control rod withdrawal error D. lowering reactor power to 19% rated prevent a violation of the MCPR Safety thermal power Limit that may result from a single control rod withdrawal error

NRC Exam 2013 QUESTION SRO 24 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201005 A2.03 Importance Rating 3.2 K&A: Ability to (a) predict the impacts of the following on the Rod Control And Information System (RCIS) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Insert block: BWR-6 RCIS Explanation: Answer B - Per Tech Specs, the RPC (SR 3.3.2.1.4) needs to be performed within one hour of lowering power to 19% RTP. Per TS Bases, the purpose of the RPC is to initiate control rod withdrawal and insert blocks.

A - Incorrect - The alarm comes in based on turbine first stage pressure, not thermal power.

C - Incorrect - The alarm comes in based on turbine first stage pressure, not thermal power. This is the Bases for the RWL.

D - Incorrect - This is the Bases for the Rod Withdrawal Limiter.

Technical Reference(s): ARI-H13-P680-05 Rev 13, TS Reference Attached: ARI-H13-P680-05 p 57, 3.3.2.1, & TS 3.3.2.1 Bases Revs 7 & 1. TS 3.3.2.1 pp 15-19, & TS 3.3.2.1 Bases pp B 3.3-43 & 48 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-07-H 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: Level of Difficulty = x 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).

Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

NRC Exam 2013 QUESTION SRO 25 Core Alterations are in progress. The following conditions exist:

A fuel bundle has been removed from the Upper Containment Fuel Storage Pool (RP-1) per the Fuel Movement Checklist.

That same fuel bundle has just been lowered 6 feet into the core, but the grapple has not been released.

Then, you, the Refuel SRO recognize the fuel bundle is not in its correct core location.

As the Refuel SRO, you are required to ____.

A. remove the bundle from the core and return it an open RP-1 location B. remove the bundle from the core and place it in its correct core location C. seat the bundle in the current core location and contact the Reactor Engineer for further guidance D. maintain the bundle in the current elevation and contact Reactor Engineering for further guidance

NRC Exam 2013 QUESTION SRO 25 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 234000 K1.05 Importance Rating 3.3 K&A: Knowledge of the physical connections and/or cause-effect relationships between Fuel Handling Equipment and the following: Reactor vessel components: Plant-Specific Fuel Handling Equipment Explanation: Answer A - > IAW FTI-D09, Use of the Fuel Movement Checklist, the mis-positioned bundle must be relocated to a vacant non-reactor location (RP-1)

B - Incorrect - This is not allowed per FTI-D09 C & D - Incorrect - A bundle in the incorrect location is a SDM concern. It must be removed from the incorrect location.

Technical Reference(s): FTI-D009 Rev 16 Reference Attached: FTI-D09 pp 12-14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3602-01-D.4 Question Source: Bank #

Modified Bank # Clinton 2009 New Question History: Previous NRC Exam Clinton 2009 #SRO11 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 b.5 Comments: Level of Difficulty = x E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Answer Key Written Examination 2013 Perry NRC ILE March 6, 2013 RO RO RO SRO 1 C 26 C 51 B 1 B 2 A 27 C 52 A 2 D 3 C 28 D 53 C 3 C 4 B 29 A 54 C 4 A 5 C 30 B 55 B 5 B 6 C 31 A 56 B 6 C 7 D 32 B 57 D 7 C 8 B 33 C 58 A 8 A delete 9 D 34 A 59 C 9 A 10 D 35 D 60 D 10 D 11 A 36 C 61 B 11 C 12 C 37 A 62 A 12 B 13 D 38 B 63 D 13 D 14 C 39 A 64 A 14 A 15 B 40 A B 65 D 15 A 16 A 41 B 66 C A 16 D 17 B 42 D 67 C 17 B 18 D 43 B 68 B 18 D 19 B 44 B 69 B 19 C 20 A 45 D 70 A 20 D 21 C 46 D 71 A 21 D 22 D 47 C 72 A 22 C 23 D 48 C 73 D 23 D 24 B 49 A 74 B 24 B 25 B 50 C 75 C 25 A

ML13113A060

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