Text

2010 Perry Plant Initial License Examination Proposed Written Examination
	ML102170193
Person / Time
Site:	Perry
Issue date:	07/18/2010
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NRC Exam - 2010 QUESTION RO 1 The Plant is operating at 95% rated thermal power. Feedwater heater 6B level switch, 1N25-N0263B, failed, causing 6B Feedwater heater to isolate.

This will cause Reactor Power to __(1)__. The following instructions will be used to mitigate the consequences of any reactivity changes that have occurred __(2)__.

__(1)__ __(2)__

A. rise ONI-N36, Loss Of Feedwater Heating and FTI-B002 Control Rod Movements B. rise ONI-C51, Unplanned Changes in Reactor Power or Reactivity and SOI-B33 Reactor Recirculation System C. lower ONI-N36, Loss Of Feedwater Heating and FTI-B002 Control Rod Movements D. lower ONI-C51, Unplanned Changes in Reactor Power or Reactivity and SOI-B33 Reactor Recirculation System

NRC Exam - 2010 QUESTION RO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.43 Importance Rating 4.1 K&A: Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Generic Explanation: Answer B - FW Heater 6B isolation will cause a rise in Rx power. ONI-C51 is entered due to the rise in Rx power and directs lowering Rx power to less than the initial power level per SOI-B33.

A & C - incorrect - No control rod movement (cram rods) is required for a FW heater isolation.

C & D - incorrect - Rx power will rise due to a loss of FW heating Technical Reference(s): ONI-C51 Flow Chart rev I Reference Attached: ONI-C51 Flow Chart (partial)

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-08(LP) A2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 2 The following conditions exist:

The plant has been shutdown for a refuel outage

Fuel shuffle is in progress

RHR Shutdown Cooling has been removed from service per the outage schedule for maintenance Which of the following is a responsibility of the Reactor Operator ATC during core alterations?

A. Monitor reactor coolant temperature 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 - 2010 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 - IOI-9, Refueling, requires moderator temperature to be maintained greater than 68 F whether fuel is irradiated or new. Tech Spec Rounds (TS 3.9.8 C.2) monitors for this condition when no RHR loop is in operation.

B and D - incorrect - this is the responsibility of the Unit Supervisor.

C - incorrect - this is the responsibility of the Refueling Supervisor and the Fuel Handling Supervisor Technical Reference(s): TS 3.9.8 C.2 & IOI-9 rev 26 Reference Attached: TS 3.9.8 C.2 p 3.9.11 &

IOI-9 p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-12(LP)-E Question Source: Bank #

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

NRC Exam - 2010 QUESTION RO 3 The following sequence of events occurred in rapid succession:

1. The plant was operating at 100% rated thermal power

2. An MSIV isolation occurred

3. Reactor pressure peaked at 1115 psig

4. The reactor scrammed on high drywell pressure

5. Safety Relief Valves (SRVs) cycled to relieve pressure during this event

6. No other scram signals have been actuated Based on these conditions, RPS __(1)__ OPERABLE and the SRVs __(2)__ OPERABLE.

__(1)__ __(2)__

A. is are not B. is are C. is not are not D. is not are

NRC Exam - 2010 QUESTION RO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.37 Importance Rating 3.6 K&A: Ability to determine operability and/or availability of safety related equipment.

Generic Explanation: Answer D - RPS is inop due to no scram on High Rx Pressure and no scram on MSIV closure. SRVs are operable due to opening when RPV pressure > 1103 psig.

A & B - incorrect - RPS is inoperable (two functions)

A & C - incorrect - SRVs required to open on Relief at 1103 +/- 15 psig Technical Reference(s): TS 3.3.1.1 & 3.3.6.4 and ARI- Reference Attached: TS 3.3.1.1 pp. 3.3-1&8 H13-P680-A8 rev 12, & ARI-H13-P680-A6 rev 12 3.3.6.4 pp. 3.3-68-69 and ARI-H13-P680-A8 p12, & ARI-H13-P680-A6 p 15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C71-P.E & OT-COMBINED-B21_N11-H Question Source: Bank # Peach Bottom 2008 Modified Bank #

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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 4 The following conditions exist:

A plant startup is in progress

Reactor power is above the Low Power Setpoint

SVI-C11-T1022, Rod Pattern Control System - Rod Withdrawal Limiter is in progress

The Rod Withdrawal Limiter failed to inhibit withdrawal at the 4-notch limit Which one of the following describes the required actions to be taken?

A. Immediately suspend control rod withdrawal.

B. Within one (1) hour take action to insert all control rods.

C. Immediately suspend control rod movement except by scram only.

D. Within one (1) hour take action to be in MODE 2 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

NRC Exam - 2010 QUESTION RO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.39 Importance Rating 3.9 K&A: Knowledge of less than or equal to one hour Technical Specification action statements for systems.

Generic Explanation: Answer A - per TS 3.3.2.1 Condition A, this is the correct action.

B - incorrect - this is the required action for more than 2 SRMs INOP in modes 3 or 4 - this is not a required action for an INOP RWL C - incorrect - this is the required action for an INOP rod pattern controller D - incorrect - this is the required action when the 1-hour action has not been met for TS 3.3.4.2 &

3.3.6.1 - not a required action for INOP RWL Technical Reference(s): TS 3.3.2.1 & svo-C11-T1022 Reference Attached: TS 3.3.2.1 p 3.3-15 &

rev 11 SVI-C11-T1022 p 2 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11_RCIS-J.5 Question Source: Bank # INL-0971 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 - 2010 QUESTION RO 5 The plant was operating at rated power when a scram was inserted.

An Anticipated Transient Without a Scram (ATWS) occurred resulting in the following conditions:

Injection Systems: Terminated and Prevented IAW EOP 1A Level Power Control

SRVs: 8 ADS valves are open IAW EOP 04-2 Emergency Depressurization

Reactor Pressure: 200 psig and lowering

Reactor Level: -20" and lowering

Reactor Power: 2% and lowering Which of the following identifies if a Technical Specification Safety Limit value has been exceeded and if adequate core cooling currently exists?

A Technical Specification Safety Limit value __(1)__ been exceeded and IAW EOP Bases, adequate core cooling __(2)__ exist.

__(1)__ __(2)__

A. has does B. has does not C. has not does D. has not does not

NRC Exam - 2010 QUESTION RO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.42 Importance Rating 3.9 K&A: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Generic Explanation: Answer A - The Reactor Vessel Water Level SL (2.1.1.3) is applicable in all Modes. RPV water level must be maintained greater than or equal to zero inches. Greater than 0 is also considered Adequate Core Cooling. EOPs direct lowering RPV water level, but the SL is still violated. Per EOP Bases, Adequate Core Cooling is also achieved by maintaining RPV pressure > Minimum Steam Cooling pressure. With 8 SRVs open, this would be any pressure higher than 140 psig. The Safety Limit is violated but Adequate Core Cooling currently exists.

B & D - incorrect - Adequate Core Cooling currently exists C & D - incorrect - the Reactor Vessel Water Level SL has been exceeded Technical Reference(s): Tech Spec 2.1, TS Reference Attached: Tech Spec 2.1 p 2.0-1, 2.1.1.3 Bases rev 4 & EOP-1A Bases rev 1 TS 2.1.1.3 Bases p B 2.0-4 & EOP-1A Bases pp. 55-56 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-08 B Question Source: Bank # Hatch 2009 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 6 As the Field Supervisor, you must assign non-licensed operators to perform a surveillance test on the Fuel pool Cooling Cleanup system (FPCC). Per PAP-0114, Radiation Protection Program, you must maintain dose ALARA.

The following conditions exist for a job to be performed.

The general area radiation levels are 10 mrem/hr in the room

The hot spot in the heat exchanger room is a pipe elbow that has a radiation level of 100 mrem/hr

The job will be performed near the hot spot area Additional information:

All 4 cases below have the same transition time to and from destinations

Dose rate for all shielding placement and removal is at 100 mrem/hr

Dose rate for the hot spot with shielding in place is 10 mrem/hr Which method complies with ALARA procedural requirements for performance of the surveillance?

A. The job is performed using 2 operators for 3 hrs each on the job at the hot spot.

B. The job is performed using 3 operators for 1 hr each on the job at the hot spot and a fourth operator reading instructions in the general room area for 1 hr.

C. The job is performed using 2 operators for 2 hrs each on the job at the hot spot and a third operator reading instructions in the general room area for 2 hrs.

D. The placement and removal of lead shielding on the hot spot is performed by 2 Radiation Protection personnel in 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The job is performed after the lead shielding is in place using 2 operators for 3 hrs each on the job.

NRC Exam - 2010 QUESTION RO 6 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 B - > The job is performed by 3 operators for 1 hr each on the job at the hot spot and a fourth operator reading instructions in the general room area for 1hr. (3 operators x 100 mrem/hr x 1 hr) + (1 operator x mrem/hr x 1 hr) = 310 mrem A - incorrect - The job is performed using 2 operators for 3 hrs each on the job at the hot spot - (2 operators x 100 mrem/hr x 3 hrs) = 600 mrem C - incorrect - The job is performed by 2 operators for 2 hrs each on the job at the hot spot and a third Operator reading instructions in the general room area for 2 hrs - (2 operators x 100 mrem/hr x 2 hr) + (1 operators X 10 mrem/hr) = 420 mrem D - incorrect - Two RP personnel hang and remove lead shielding on the hot spot in 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months - The job is performed after the lead shielding is in place using 2 operators for 3 hrs each on the job. (2 RP techs x 100 mrem/hr x 1.5hrs) + (2 operators x 10 mrem/hr x 3 hr) = 360 mrem Technical Reference(s): PAP-0114 rev 15 Reference Attached: PAP-0114 p 15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): None Question Source: Bank # Hope Creek 2009 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 7 The plant is operating at 100% power.

Offgas Post Treatment Radiation monitors 1D17-K601A and K601B have alarmed on a High-High Radiation signal (OG ISOL OG POST-TREAT PRCS RAD A/B 3XHI -

H13-P604-0001-A5).

Which one of the following describes the effect on Offgas and the Main Condenser?

A. Only the Offgas charcoal adsorbers inlet valves will isolate causing a loss of main condenser vacuum.

B. Offgas will shift into the Bypass Mode of operation causing a loss of main condenser vacuum.

C. Offgas will shift to Treat Mode allowing main condenser vacuum to remain constant.

D. Offgas System will isolate causing a loss of main condenser vacuum.

NRC Exam - 2010 QUESTION RO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.15 Importance Rating 2.9 K&A: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Generic Explanation: Answer D - a Hi-Hi signal from both rad monitors causes the Offgas system to isolate.

When OG isolates, main condenser vacuum will degrade.

A - incorrect - OG Discharge Isolation, Cooler Condenser, Prefilter Inlet, and Holdup Line drain valves all isolate B - incorrect - OG will not auto shift to Bypass Mode C - incorrect - OG should be shifted from Auto Mode to Treat Mode at 5% power Technical Reference(s): ARI-H13-P604-001-A5 rev 5 Reference Attached: ARI-H13-P604-001-A5 p 13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N64-M.6 Question Source: Bank # River Bend 2008 Modified Bank #

New Question History: Previous NRC Exam: River Bend 2008 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 - 2010 QUESTION RO 8 The following conditions exist:

The plant scrammed on high reactor pressure

The reactor is shutdown

All feedwater pumps tripped

Level control is on Reactor Core Isolation Cooling

Pressure control is on the Bypass Valves The assigned level band should be __(1)__, but can be expanded if other __(2)__ actions have a higher priority.

__(1)__ __(2)__

A. 150 to 219 EOP B. 150 to 219 ONI C. 178 to 219 EOP D. 178 to 219 ONI

NRC Exam - 2010 QUESTION RO 8 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.14 Importance Rating 3.8 K&A: Knowledge of general guidelines for EOP usage.

Generic Explanation: Answer C - both EOP-1 and ONI-C71-1 direct a level band of 178-219. Only the EOP allows expanding the level band (General Guidelines). Scram on high reactor pressure is an EOP entry.

A - incorrect - this is the expected level band when pressure control is on SRVs.

B - incorrect - ONIs do not allow expansion of the level band, they direct entry into the EOPs.

D - incorrect - correct level band, but the ONIs do not allow expansion of the level band.

Technical Reference(s): EOP-1 Bases rev 1 & EOP-1 Reference Attached: EOP-1 Bases p 27 &

Chart rev B. EOP-1 Chart (partial)

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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 9 The reactor is at 96% rated thermal power with a plant down power in progress for monthly surveillance testing.

The following occurs:

ANN PWR SUPPLY FAIL is illuminated

Alarms that were locked in on H13-P601 have deactivated The Control Room actions would be to dispatch an operator to __(1)__ to investigate and __(2)__.

1 2 A. D-1-A suspended power maneuvering B. D-1-A continue lowering reactor power C. ED-1-A suspended power maneuvering D. ED-1-A continue lowering reactor power

NRC Exam - 2010 QUESTION RO 9 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.32 Importance Rating 3.6 K&A: Knowledge of operator response to loss of all annunciators.

Generic Explanation: Answer A - a loss of D-1-A-06 will cause ANN PWR SUPPLY FAIL window to illuminate and other annunciator windows to deactivate. Operator actions are to maintain the plant stable - suspend power maneuvering.

B & D - incorrect -the action to continuing to lower power is wrong C & D - incorrect - ED-1-A does supply optical isolators which could cause the annunciator windows on H13-P601 to deactivate, but it will not cause ANN PWR SUPPLY FAIL window to illuminate Technical Reference(s): ARI-H13-P680-007-E5 rev 17 & Reference Attached: ARI-H13-P680-007-E5 ONI-R61 rev 3 p 137 & ONI-R61 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-05(LP)-A.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 10 Which one of the following identifies a portion of a Plant Communication system that is reserved for emergency and Control Room communications, per PAP-0202, Communications?

A. Extension 5634 of the Private Branch Exchange (PBX) Phone System.

B. Channel 5-G of the RMT (CEI 800-MHx Trunked) System.

C. Line 5 of the Public Address (PA) System.

D. Channel F2 of the Plant Radio System.

NRC Exam - 2010 QUESTION RO 10 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.43 Importance Rating 3.2 K&A: Knowledge of emergency communications systems and techniques.

Generic Explanation: Answer C - per PAP-0202, Section 3.11, the PA system line 5 is reserved.

A - incorrect - This is the PBX extension for the RO-ATC desk, which is used for all types of communications (not just emergency communications)

B - incorrect - This communication system is used by Off-Site Radiation Monitoring Teams for communications with the TSC and EOF D - incorrect - This is the Radio frequency reserved for use by Maintenance personnel Technical Reference(s): PAP-0202 rev 4 Reference Attached: PAP-0202 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3039-01.B Question Source: Bank # Perry 2003 Modified Bank #

New Question History: Previous NRC Exam: Perry 2003 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 - 2010 QUESTION RO 11 The plant was operating at 100% rated thermal power when both Reactor Recirculation Pumps down shifted.

The plant is now operating in the Immediate Exit Required region of the OPRM Operable - Two Loop Power - Flow Map.

Which of the following will raise the likelihood of power oscillations?

A. Raise core flow.

B. Control rod insertion.

C. Feedwater temperature reduction.

D. Lower pressure regulator setpoint.

NRC Exam - 2010 QUESTION RO 11 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295001 AK1.04 Importance Rating 2.5 K&A: Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

Limiting cycle oscillation: Plant-Specific Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Explanation: Answer C - lowering feedwater temperature can cause increase in reactor power and increase the likelihood of power oscillations. For this reason, RPV level is lowered during an ATWS.

A - incorrect - raising core flow lowers the possibility of power oscillations B - incorrect - inserting control rods to lower power lowers the possibility of power oscillations D - incorrect - reducing the pressure regulator setpoint will not have any effect on power oscillations Technical Reference(s): EOP Bases rev 0 Reference Attached: EOP Bases p 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x Question Source: Bank # INL-1312 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 - 2010 QUESTION RO 12 A Total Loss of AC power (TLAC) is in progress.

ONI-SPI D1, Maintaining System Availability, directs that the Division 3 Battery Room door be opened within two (2) hours.

Which of the following describes the location and the specific reason given for performing this action?

A. Control Complex 620: Dissipate Heat B. Control Complex 620: Prevent Hydrogen build up C. Control Complex 638: Dissipate Heat D. Control Complex 638: Prevent Hydrogen build up

NRC Exam - 2010 QUESTION RO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295003 AK2.01 Importance Rating 3.2 K&A: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: Station batteries Partial or Complete Loss of AC / 6 Explanation: Answer A - per ONI-SPI- D-1, battery room doors are opened to dissipate heat.

B & D - incorrect - hydrogen buildup is a concern when designing battery room ventilation, but it is not the reason given in ONI-SPI D1 for opening the door within two hours C & D - incorrect - Control Complex 638 has Div 1 & 2 battery rooms, but not the Div 3 battery room Technical Reference(s): ONI-SPI-D-1 rev 0 & ONI-R10 Reference Attached: ONI-SPI-D-1 p 3 & ONI-flowchart rev B R10 flowchart Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-18(LP) A.4 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 13 The plant was operating at 80% rated thermal power.

Then, due to a system fault, power is lost to 125 VDC Bus ED-1-A.

Which of the following loses power and what is the immediate operator action?

A. High Pressure Core Spray (HPCS) Logic - Verify RCIC Operable B. Low Pressure Core Spray (LPCS) Logic - Rack-out breaker EH1111 for LPCS pump C. Residual Heat Removal (RHR) C Logic - Place RHR C loop on Alternate Keep-fill D. Reactor Recirculation Breaker 3 Control Logic - Insert a manual reactor scram

NRC Exam - 2010 QUESTION RO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295004 AK3.03 Importance Rating 3.1 K&A: Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Reactor SCRAM:

Partial or Total Loss of DC Pwr / 6 Explanation: Answer D - ED-1-A supplies control logic for Recirc Breakers 3A & 3B. When power is lost, Recirc breakers 5A & 5B trip open. Per ONI-C51, with no Recirc pumps running, place Mode Switch in Shutdown.

A - incorrect - HPCS logic is powered from ED1C B - incorrect - LPCS logic is powered from ED1A, however, there are no required operator actions to rack-out the LPCS pump breaker C - incorrect - RHR C logic is powered from ED1B Technical Reference(s): ONI-R42-1 rev 7 & ONI-C51 Reference Attached: ONI-R42-1 p 4 & ONI-rev 24 C51 p 6 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R42-F & OT-3035-05(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 14 With the reactor at 100% power, which of the following conditions will result in a reactor scram and a direct automatic transfer of the Recirculation Pumps from fast speed to slow speed?

A. Main turbine trip B. Reactor feedwater pump trip C. Drywell pressure high - 1.68 psig D. Reactor water level high - Level 8

NRC Exam - 2010 QUESTION RO 14 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295005 AA1.01 Importance Rating 3.1 K&A: Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Recirculation system:

Main Turbine Generator Trip / 3 Explanation: Answer A - A main turbine trip from >38% power will initiate a reactor scram and EOC-RPT logic will initiate a downshift of Recirc pumps. This is based on the MT Stop valve position (direct)

B - incorrect - RFPT trip will cause a FCV runback when RPV hits L4 C - incorrect - DW pressure high will cause a Rx scram, but not a direct down shift of Recirc pumps - the subsequent lowering of feedwater flow will cause a RR Pump downshift after a time delay D - incorrect - RPV water level high will cause a Rx scram, but not a direct down shift of Recirc pumps -

the subsequent lowering of feedwater flow will cause a RR Pump downshift after a time delay Technical Reference(s): ONI-N32 rev 9 & ARI-H13- Reference Attached: ONI-N32 p 4 & ARI-P680-004-A3 rev 14 H13-P680-004-A3 p 9 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-E.3 Question Source: Bank # INL-1294 Modified Bank #

New Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 15 The following plant conditions exist:

RCIC was running for testing

RHR loop A was running in Suppression Pool Cooling mode to support RCIC testing The following then occurred:

The Control Room was evacuated due to toxic gas

All Immediate Actions of ONI-C61, Evacuation of the Control Room, have been completed

No other operator actions have been performed.

Which of the following describes the Suppression Pool Temperature instrumentation available on the Remote Shutdown Panel(s) that is(are) indicating accurate Suppression Pool Temperature, if any, at this time?

A. No Remote Shutdown Panel Suppression Pool temperature instrumentation.

B. Only the Division 1 Remote Shutdown Panel Suppression Pool temperature instrumentation.

C. Only the Division 2 Remote Shutdown Panel Suppression Pool temperature instrumentation.

D. Both the Division 1 and Division 2 Remote Shutdown Panel Suppression Pool temperature instrumentation.

NRC Exam - 2010 QUESTION RO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295016 AA2.04 Importance Rating K&A: Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Suppression pool temperature Control Room Abandonment / 7 Explanation: Answer C - Division 2 Remote Shutdown Panel instruments do not have transfer switches and are normally energized.

A - incorrect - Div 2 panel instrumentation is providing accurate information B - incorrect - Div 1 panel instruments must be transferred to energize the instruments, this action has not been done yet D - incorrect - same as B Technical Reference(s): SDM-C61 rev 6 Reference Attached: SDM-C61 p 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-13(LP)-A.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 16 The following plant conditions existed at 08:00:

The plant was operating at 100% rated thermal power

NCC pumps A & B were running

NCC pump C was tagged out for maintenance

NCC C heat exchanger was in dry layup

NCC B heat exchanger was removed from service due to indications of a tube leak At 09:30, NCC temperature control valve (NCC HX A TCV) P43-F006A failed.

The Unit Supervisor entered ONI-P43, Loss Of Nuclear Closed Cooling.

The following conditions currently exist:

Validated SPDS Drywell temperature is rising at 0.5oF/minute.

Average Drywell temperature is 138oF

Highest Drywell temperature is 148oF

Suppression Pool Average temperature is 96oF rising slowly

Suppression Pool Level is 18.1 rising slowly

Containment pressure is 0.3 psig stable

Drywell pressure is 0.6 psig rising slowly

Containment Average temperature is 87oF rising slowly Based on these conditions, the following action are required immediately.

A. Insert a manual reactor scram.

B. Continue with ONI-P43 actions only.

C. Suspend ONI-P43 actions and enter EOP-2 Primary Containment Control.

D. Continue with ONI-P43 actions and enter EOP-2 Primary Containment Control.

NRC Exam - 2010 QUESTION RO 16 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295018 2.4.8 Importance Rating 3.8 K&A: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Partial or Total Loss of CCW / 8 Explanation: Answer D - Entry into EOP-2 is required based on Suppression Pool temperature. EOPs are the higher tier documents, but ONI actions should be taken if they don not interfere with EOPs.

A - incorrect - this action is only appropriate if all NCC flow was lost after lowering flow to 58 mlbm/hr B - incorrect -entry condition for EOP-2 is met C - incorrect - it is inappropriate to suspend ONI-P43 actions since entry conditions for ONI-P43 still exist Technical Reference(s): EOP Bases rev 0 & EOP-2 Reference Attached: EOP Bases p 27 &

Bases rev 0 EOP-2 Bases pp. 6 & 15 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-16(LP)-A.3 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 17 The following plant conditions exist:

The reactor is operating at 50% power

The Service Air and Instrument Air Systems are in their normal lineup An air leak occurs resulting in the following:

Instrument Air receiver pressure is 85 psig and lowering

Service Air receiver pressure is 95 psig and lowering Which of the following describes how the Service Air/Instrument Air Cross-Connect Valves, 1P52-F050 & 2P52-F050, respond to these conditions, including the bases for this response?

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

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

NRC Exam - 2010 QUESTION RO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295019 AA1.04 Importance Rating 3.3 K&A: Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service air isolations valves:

Partial or Total Loss of Inst. Air / 8 Explanation: Answer B - the cross-connect valves will close when Instrument Air receiver pressure is

<90 psig to protect the Instrument Air system from a leak in the Service Air system.

A - incorrect - check valves around the P52-F050 valves allow Service Air to continue to supply Instrument Air when the F050 valves are closed C - incorrect - P52-F050 valves are closed - Service air can still supply instrument air header. Therefore, they are not completely isolated from each other D - P52-F050 valves are closed. There are no automatic actions at 80 psig in the IA receiver.

Technical Reference(s): SOI-P51/52 rev 25 & Lesson Reference Attached: SOI-P51/52 p 4 &

Plan OT-COMBINED-P51_P52 rev 1 Lesson Plan OT-COMBINED-P51_P52 p 14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P51_P52-E.7 Question Source: Bank # Perry 2002 Modified Bank #

New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 18 The following conditions exist:

It is day 20 of a refuel outage

Fuel Pool Cooling Cleanup (FPCC) is providing decay heat removal

Fuel shuffle is complete

Core verification is in progress

Div 2 electrical outage in progress

Reactor coolant temperature is 85oF FPCC A pump then trips on a ground fault.

Based on the above conditions, determine time to boil in the core?

Reference Provided: PBD-A016 & PDB-A017 A. 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months B. 8.5 hrs C. 20 hrs D. 35 hrs

NRC Exam - 2010 QUESTION RO 18 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 D - FPCC is the alternate decay heat removal (SDC) system. The decay heat load is 20 MBTU/hr based on curves for decay after 248 bundles discharged and water level at 23 above the RPV flange, the time to boil is approximately 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months . With Div 2 outage in progress, B FPCC pump is not available.

A - incorrect - plausible if use time to boil curve at vessel flange and decay heat curve prior to bundle discharge B - incorrect - plausible if use time to boil curve at vessel flange and decay heat curve after bundle discharge C - incorrect - plausible if use time to boil curve at 23 above flange and decay heat curve prior to bundle discharge Technical Reference(s): PBD-A016 rev 9 & PDB-A017 Reference Attached: PBD-A016 pp. 3 & 5 &

rev 10 PDB-A017 pp 9 & 11 Proposed references to be provided to applicants during examination: PBD-A016 & PDB-A017 Learning Objective (As available): OT-3035-11(LP)-A.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 19 The following plant conditions exist:

A full core offload is in progress The following then occurred:

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

Bubbles are observed coming from the irradiated fuel bundle

The Fuel Handling Building Evacuation Alarm Sounded Based on these conditions, in order to minimize unnecessary exposure, immediate evacuation of ____ from the Fuel Handling Building is required.

1. Platform Operator

2. FME Coordinator

3. Fuel Handling Building Crane Operator

4. Fuel Handling Supervisor

5. Radiation protection Technician

6. Site Protection Officer

7. Spotter A. 1, 3, & 7 B. 1, 4, & 6 C. 2, 5, & 6 D. 3, 5, & 7

NRC Exam - 2010 QUESTION RO 19 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295023 AK3.01 Importance Rating 3.6 K&A: Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS: Refueling floor evacuation Refueling Acc / 8 Explanation: Answer C - Per ONI-J11-2 Necessary Personnel are defined as those personnel necessary to place the equipment or fuel in a safe condition. SOI-F11, fuel handling Platform, identifies personnel required for fuel handling in the FHB. At Perry, necessary personnel are the FH Supervisor, the Platform Operator, and the Spotter - all other personnel are to be evacuated.

A - incorrect - bridge driver and spotter are necessary personnel B - incorrect - bridge driver and FH Supervisor are necessary personnel D - incorrect - the spotter is considered necessary personnel Technical Reference(s): ONI-J11-2 rev 13 and SOI-F11 Reference Attached: ONI-J11-2 pp. 5-7 and rev 10 SOI-F11 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-14(LP)-A.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 20 The following plant conditions exist:

The plant is operating at 100% power

No equipment is out of service A large-break LOCA occurs in the Drywell and the following conditions exist:

Drywell pressure is at 12 psig and lowering

Reactor pressure is at 140 psig and lowering

Reactor water level is at 5 inches and rising Which of the following describes the loading sequence to limit starting transients on the 4160 V Emergency Buses?

A. RHR C starts 5 seconds from initiation signal HPCS starts 10 seconds from initiation signal LPCS starts 15 seconds from initiation signal B. RHR C starts 5 seconds from initiation signal HPCS starts 14 seconds from initiation signal LPCS starts 19 seconds from initiation signal C. RHR C starts immediately after initiation signal HPCS starts 10 seconds from initiation signal LPCS starts 15 seconds from initiation signal D. RHR C starts immediately after initiation signal HPCS starts 14 seconds from initiation signal LPCS starts 19 seconds from initiation signal

NRC Exam - 2010 QUESTION RO 20 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295024 EA1.10 Importance Rating 3.4 K&A: Ability to operate and/or monitor the following as they apply to HIGH DRYWELL PRESSURE: A.C. distribution High Drywell Pressure / 5 Explanation: Answer C - a high DW pressure signal will initiate HPCS, LPCS, & RHR. Per Tech Spec Bases, pump starting is sequenced due to the high starting currents to control emergency bus and transformer loading.

A & B - incorrect - RHR C starts immediately.

B & D - incorrect - HPCS starts 10 seconds from initiation. Nineteen seconds for LPCS and 14 seconds for HPCS is plausible because on a loss of bus power the respective pump, start relays are energized for 19 & 14 seconds upon restoration of power.

Technical Reference(s): TS 3.3.5.1 Bases rev 7, SDM- Reference Attached: TS 3.3.5.1 Bases p E12 rev 9, SDM-E21 rev 10, & SDM-E22 rev 7 B3.3-97, SDM-E12 p 37, SDM-E21 p 21, &

SDM-E22 p 22 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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 21 The plant was operating at rated power when a transient occurred that resulted in the following plant parameters:

Reactor pressure: 1040 psig

RPV level: 50

Drywell temperature: 185oF

Drywell pressure: 4 psig

Suppression Pool level: 18.5 feet

Suppression Pool temperature: 114oF

Containment pressure: 2.5 psig

Containment temperature: 125oF Which one of the following conditions will cause the margin to the Heat Capacity Limit (HCL) to improve?

A. RPV level rises B. RPV pressure lowers C. Suppression Pool level lowers D. Suppression Pool temperature rises

NRC Exam - 2010 QUESTION RO 21 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295025 EA2.03 Importance Rating 3.9 K&A: Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Suppression pool temperature High Reactor Pressure / 3 Explanation: Answer B - lowering reactor pressure will move farther away from HCL.

A - incorrect - plausible due to misconception of effect of RPV level on HCL. Raising RPV level will have no effect on HCL margin C - incorrect - lowering suppression pool level will reduce margin to HCL D - incorrect - raising suppression pool temperature lowers margin to HCL Technical Reference(s): EOP Bases rev 0 Reference Attached: EOP bases p 63 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-06-C.1 Question Source: Bank #

Modified Bank # Peach Bottom 2008 New Question History: Previous NRC Exam: Peach Bottom 2008 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 - 2010 QUESTION RO 22 The following plant conditions exist:

Reactor power: 85%

Suppression Pool Average temperature: 76oF The following then occurs:

Safety Relief valve 1B21-F047G inadvertently opens

ONI-B21, SRV Inadvertent Opening/Stuck Open, actions were not successful in closing the SRV

Suppression Pool Average Temperature is rising 3oF/minute

No other operator actions are being performed A Technical Specification requirement to immediately scram the reactor will first be met in ____.

A. 7 minutes B. 10 minutes C. 12 minutes D. 15 minutes

NRC Exam - 2010 QUESTION RO 22 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 2.2.22 Importance Rating 4 K&A: Knowledge of limiting conditions for operations and safety limits.

Suppression Pool High Water Temp. / 5 Explanation: Answer C - with a heat-up rate of 3 deg/min, suppression pool temperature will exceed 110oF in just over 11 minutes. Per tech specs, a scram is required if suppression pool temp exceeds 110oF.

A - incorrect - this is the time to exceed 95oF, which is the requirement to enter TS 3.6.2.1 with no testing in progress adding heat to suppression pool B - incorrect - this is the time to exceed 105oF, which is the requirement to enter TS 3.6.2.1 with testing in progress adding heat to suppression pool D - incorrect - this is the time to exceed 120oF, which is requirement to depressurize the reactor Technical Reference(s): Tech Spec 3.6.2.1 Reference Attached: Tech Spec 3.6.2.1 pp.

3.6 38 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-10.B Question Source: Bank #

Modified Bank # Oyster Creek 2008 New Question History: Previous NRC Exam: Oyster Creek 2008 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 - 2010 QUESTION RO 23 EOP-2, Containment Control, was entered due to a scram discharge volume rupture.

Containment Spray was initiated per EOP-SPI 3.1, Containment Spray Operation.

In accordance with EOP-2, Containment Spray operation is required to be terminated ____.

A. before Containment Pressure is reduced below 0.0 psig B. before Containment Pressure is reduced below 0.5 psig C. after Containment Temperature is reduced below 95°F D. after Containment Temperature is reduced below 185°F

NRC Exam - 2010 QUESTION RO 23 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295027 EK1.03 Importance Rating 3.8 K&A: Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY):

Containment integrity High Containment Temperature / 5 Explanation: Answer A - Containment Spray must be terminated prior to reaching 0 psig in containment in order to maintain a margin to the negative design pressure of containment.

B - incorrect - 0.5 psig is a rule of thumb when securing from containment spray operation C - incorrect - 95oF is the entry condition for containment temperature leg, where containment sprays can be started D - incorrect - 185oF is the containment design limit for containment temperature for ED Technical Reference(s): EOP-2 Bases rev 0 Reference Attached: EOP-2 Bases p 10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-07-C.2 Question Source: Bank # Perry 2007-1 Modified Bank #

New Question History: Previous NRC Exam: Perry 2007-1 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 - 2010 QUESTION RO 24 The following conditions exist:

EOP-2, Containment Control, has been entered

Suppression Pool level is 17.7 feet and lowering slowly

Suppression Pool Average temperature is 83oF

Containment Average temperature is 96oF and stable

Drywell Average temperature is 146oF and rising

Drywell pressure is 0.6 psig and stable Based on these conditions, the required EOP-2 action(s) is(are) to ____.

A. Maximize Suppression Pool Cooling only B. Operate all available Drywell Cooling only C. Operate all available Drywell Cooling and Operate all available Containment Cooling D. Maximize Suppression Pool Cooling and Operate all available Drywell Cooling and Operate all available Containment Cooling

NRC Exam - 2010 QUESTION RO 24 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295028 EK2.04 Importance Rating 3.6 K&A: Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Drywell ventilation High Drywell Temperature / 5 Explanation: Answer C - Containment temperature and DW temperature exceeds the value for operating all containment and DW cooling. Per the EOP bases, the determination that DW temp can not be maintained can be made prior to exceeding the value.

A - incorrect - SP temperature is 13 deg F from the limit B - incorrect - this is only partially correct, containment temperature exceeds the value to operate cooling D - incorrect - SP temperature is 13 deg F from the limit and is rising slowly Technical Reference(s): EOP Bases rev 0 & EOP-2 Reference Attached: EOP Bases p 36 &

Bases rev 0 EOP-2 Bases pp 17 and 50 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-8.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 25 A CAUTION in the EOPs warns the operator about operating RCIC with low Suppression Pool water level (<7.25 feet).

Select the statement below that describes the application of this limit.

Immediate and catastrophic equipment failure ____.

A. is expected. Operation of RCIC at a Suppression Pool level less than 7.25 feet is not permitted.

B. is not expected. Operation of RCIC at a Suppression Pool level less than 7.25 feet is not permitted.

C. is expected. Operation of RCIC at a Suppression Pool level less than 7.25 feet is permitted under certain circumstances.

D. is not expected. Operation of RCIC at a Suppression Pool level less than 7.25 feet is permitted under certain circumstances.

NRC Exam - 2010 QUESTION RO 25 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295030 EK3.03 Importance Rating 3.6 K&A: Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: RCIC operation Low Suppression Pool Wtr Lvl / 5 Explanation: Answer D - Equipment failure is not expected to be immediate and this limit may be disregarded if core cooling is threatened.

A & B - incorrect - This limit may be disregarded if core cooling is threatened.

A & C - incorrect - Equipment failure is not expected to be immediate.

Technical Reference(s): EOP Bases rev 0 Reference Attached: EOP Bases p 48 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-1-b.2 Question Source: Bank # Perry 2001 Modified Bank #

New Question History: Previous NRC Exam: Perry 2001 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 - 2010 QUESTION RO 26 The following conditions exist:

A LOCA in the drywell occurred

Bus EH11 experienced a lockout

High Pressure Core Spray failed to initiate

Validated RPV water level is 25 inches and lowering

Validated RPV pressure is 900 psig and lowering

Drywell pressure is 1.8 psig and rising

1B21-N695B, RPV Level - Lvl 3, failed high

All other systems are functioning as designed.

Which of the following describes the operation of the Automatic Depressurization System (ADS) under the current conditions?

A. ADS SRVs can be opened by their individual key-lock switches only.

B. ADS logic automatically initiates to open ADS SRVs after 105 second time delay relay times out C. ADS SRVs can be opened by arming and depressing the ADS LOGIC MANUAL INITIATION switches.

D. ADS logic automatically initiates to open ADS SRVs 105 seconds after the RPV level set point is reached.

NRC Exam - 2010 QUESTION RO 26 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295031 EA1.06 Importance Rating 4.4 K&A: Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: Automatic depressurization system Reactor Low Water Level / 2 Explanation: Answer C - the Arm & Depress switches will work anytime the respective LPCI/LPCS pump is providing sufficient discharge pressure.

A - incorrect - the manual SRV switches are not the only way to open the valves B & D - incorrect - the ADS B system will not automatically initiate with the Level 3 instrument failed high and ADS A will not automatically initiate with no LPCI/LPCS pump running Technical Reference(s): PDB-I-005 rev 9 Reference Attached: PDB-I005 pp.13-14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21C E.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 27 The plant was operating at 85% rated power with the following equipment out of service:

CRD B pump tagged out due to an oil leak A manual reactor scram was inserted.

The following conditions now exist:

Reactor Scram Hardcard actions were completed

Reactor power is at 5%

Reactor pressure is at 940 psig

The GP1A light above the A manual scram switch is illuminated

Bus EH11 locked out

Full Core Display indicates numerous control rods withdrawn at various positions

SCRAM VLV AIR HEADER PRESS LO annunciator (H13-P680-5A-D6) has alarmed Which of the following methods would successfully insert control rods for the above listed conditions?

A. EOP-SPI 1.1, PULLING SCRAM FUSES B. EOP-SPI 1.3, MANUAL ROD INSERTION C. EOP-SPI 1.5, VENTING CRD OVERPISTON VOLUMES D. EOP-SPI 1.6, INCREASED COOLING WATER P

NRC Exam - 2010 QUESTION RO 27 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295037 EA2.05 Importance Rating 4.2 K&A: Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Control rod position SCRAM Condition Present and Power Above APRM Downscale or Unknown / 1 Explanation: Answer C - CRD B pump is tagged and CRD A bump is not running due to EH11 bus lockout. Therefore, venting the over-piston volumes is the only method that would work.

A - incorrect - RPS did not fail to de-energize, therefore, this method would not be successful B & D - incorrect - a CRD pump must be running to use these sections Technical Reference(s): EOP-SPI 1.1 rev 0, EOP-SPI Reference Attached: EOP-SPI 1.1 p 2, EOP-1.3 rev 1, EOP-SPI 1.5 rev 0, EOP-SPI 1.6 rev 1 SPI 1.3 p 2, EOP-SPI 1.5 p 2, EOP-SPI 1.6 p 2

Proposed references to be provided to applicants during examination: None Learning Objective (As available): 3402-03-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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 28 The following conditions exist:

The plant is operating at full power

Fuel inspections are in progress in the Fuel Handling Building

FHB HVAC Supply Fan A (M40-C001A) is in operation

FHB HVAC Exhaust Fans A and B (M40-C002A(B)) are in operation An event occurs that damages numerous Spent Fuel Bundles with the following results:

All FHB Ventilation Exhaust Airborne Radiation Monitor (D17-K710) module indications are offscale high.

Spent Fuel Pool (D21-K332) and Fuel Prep Pool (D21-K322) Area Radiation Monitor indications are reading 12,000 mR/hr.

The radiation release rate is at the General Emergency level

The Unit Supervisor entered EOP-3, Secondary Containment Control and EOP-5, Radioactivity Release Control The off-site release path is through the __(1)__. The Control Room Operator would __(2)__.

__(1)__ __(2)__

A. Unit 1 Plant Vent verify FHB HVAC Supply Fan A is tripped B. Unit 1 Plant Vent scram the Reactor and Emergency Depressurize C. Unit 2 Plant Vent verify FHB HVAC Supply Fan A is tripped D. Unit 2 Plant Vent scram the Reactor and Emergency Depressurize

NRC Exam - 2010 QUESTION RO 28 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295038 2.4.31 Importance Rating 4.2 K&A: Knowledge of annunciator alarms, indications, or response procedures.

High Off-site Release Rate / 9 Explanation: Answer A - Release path from FHB is through the UNIT 1 Plant Vent, EOP-3 requires verification of supply fans that should have tripped trip.

B & D - incorrect - this would be true if the leak was from a primary system per the EOPs C & D - incorrect - unit 2 plant vent is not release pathway Technical Reference(s): ARI-H13-P902-001-B1 rev 4 & Reference Attached: ARI-H13-P902-001-B1 EOP-3 Bases rev 0 p 7 & EOP-3 Bases p 12 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-17-D.2, OT-COMBINED-M40-B.3, & OT-3402-15-C.2 Question Source: Bank # Perry 2007-1 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 29 The following conditions exist:

The plant is operating at 100% power

SAS reports that Zone 1 and Zone 2 heat detectors for Reactor Recirculation Pump A are in alarm

The Unit Supervisor enters ONI-P54, Fire Which one of the following actions is required for a fire in Reactor Recirculation Pump A per ONI-P54, Fire?

A. Operator confirms the CO2 System has automatically dumped CO2 B. Operator opens 1P54-F395, DW CO2 SUPPLY OTBD ISOL to commence CO2 dump C. Operator opens 1P54-F340, CTMT CO2 SUPPLY OTBD ISOL to commence CO2 dump D. Fire Brigade member opens 1P54-F3590, RX RCIRC PMP A SELECTOR VLV to commence CO2 dump

NRC Exam - 2010 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 C - Per ONI-P54 Immediate operator action for Recirculation pump fire is to open 1P54F340.

A - incorrect - 2 heat detectors automatically initiate CO2, but system will not dump with 1P54F340 closed B - incorrect - wrong valve, valve is already open D - incorrect - Fire Brigade is not required to open the selector valve for a Recirculation Pump fire Technical Reference(s): ONI-P54 rev 15 Reference Attached: ONI-P54 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED P54-CO2-D.1.B, OT-3035-05(LP)-A.9 Question Source: Bank # Perry 2007-1 Modified Bank #

New Question History: Previous NRC Exam: Perry 2007-1 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 - 2010 QUESTION RO 30 SCC informs the Control Room of a degraded grid condition and the following occurs:

1. ONI-S11, Hi/Low Voltage, is entered

2. Grid voltage and generator voltage swings are observed What is the nominal bus voltage at which the undervoltage relays will drop out AND what is the basis for this protection?

A. 3000 Vac (~75% nominal), protection is provided to prevent damage to running safety related equipment.

B. 3800 Vac (~95% nominal), protection is provided to prevent jeopardizing the reliability of starting a diesel generator.

C. 3800 Vac (~95% nominal), protection is provided to prevent jeopardizing the reliability of starting safety related motors.

D. 3000 Vac (~75% nominal), protection is provided to prevent jeopardizing the reliability of starting safety related motors

NRC Exam - 2010 QUESTION RO 30 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 700000 AK2.01 Importance Rating 3.1 K&A: Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following: Motors Generator Voltage and Electric Grid Disturbances / 6 Explanation: Answer D - 75% voltage is to protect starting safety related motors A - incorrect - 3000 vac is protection for starting motors, not running equipment B - incorrect - Bus voltage does not effect the reliability of the D/G starting B & C - incorrect - 3800 volts is to protect running equipment Technical Reference(s): OT-COMBINED-R10 (LP) rev 0 Reference Attached: OT-COMBINED-R10 (LP) pp. 49-50 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R10-D.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 31 The plant is operating at 70% power when Inboard MSIV, B21-F022A, inadvertently closes.

Reactor pressure rises 20 psig and stabilizes

No RPS setpoints are exceeded

No operator actions are taken Which one of the following describes the response of the reactor to this event?

Reactor power initially ____.

A. lowers and then stabilizes at a lower value.

B. lowers and then returns to its original value.

C. rises and then stabilizes at a higher value.

D. rises and then returns to its original value.

NRC Exam - 2010 QUESTION RO 31 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295007 AA2.02 Importance Rating 4.1 K&A: Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor power High Reactor Pressure / 3 Explanation: Answer C -Rx pressure rising causes voids to collapse causing a positive reactivity addition causing Rx power rises. Power will stabilize at a higher value due to the higher pressure.

A & B - incorrect - Rx power will initially rise due to void collapse, not lower D - incorrect - Rx power will not return to its original value Technical Reference(s): OT-3301-04 (LP) rev 4 (GFE-Rx Reference Attached: OT-3301-04 (LP) p49 Theory Chaper 4) and IOI-3 rev 44 and IOI-3 p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3301-04-7 & 9 Question Source: Bank # Perry 2003 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 32 The following conditions exist:

The plant is operating at 50% power

RFPT A is out of service for an oil leak RFPT B tripped.

The following then occurred:

The reactor operator inserted a reactor scram

Reactor pressure peaked at 1045 psig

The Motor Feed pump tripped on an electrical fault

All other systems functioned as designed

EOP-1, RPV Control, was entered Based on the above conditions which of the following valves will be closed?

1. 1B21-F019 MSL DRN & MSIV BYP OTBD ISOL

2. 1D17-F081B CNTMT RAD MON INBD SUCT ISOL

3. 1E12-F073B RHR B HXS SECOND VENT TO SUPR POOL

4. 1G42-F010 SPCU PUMP FIRST SUCT ISOL

5. 1P43-F055 NCC CNTMT SUPPLY OTBD ISOL

6. 1P51-F150 SA SUPPLY HDR CNTMT ISOL.

A. 3, 4, & 5 B. 2, 3, & 6 C. 1, 5, & 6 D. 1, 2, & 4

NRC Exam - 2010 QUESTION RO 32 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295009 2.4.2 Importance Rating 4.5 K&A: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

Low Reactor Water Level / 2 Explanation: Answer B - based on the given conditions, EOP-1 should be entered when vessel hits Level 2. The Level 2 and Level 3 isolations will occur. Additionally, HPCS and RCIC will initiate on L2 to restore level and not allow level to drop to Level 1.

A & C - incorrect - 1P43-F055 isolates on L1 not L2 C & D - incorrect - 1B21-F019 isolates on L1 not L2 Technical Reference(s): EOP-1 Bases rev 1 & OAI-1703 Reference Attached: EOP-1 Bases p22 &

rev 3 OAI-1703 pp. 16 & 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-02-B & OT-COMBINED-B21-NS4-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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 33 The plant is operating at 100% power when a trip of Containment Vessel Chilled Water Chiller A occurred.

Containment temperature and pressure are slowly rising

Drywell temperature and pressure are stable

Alarm CONTAINMENT TEMP A HIGH H13-P601-020-F4 has been received

Alarm CONTAINMENT TEMP B HIGH H13 P601-017-D2 has been received

No EOP Entry Conditions exist Which one of the following conditions will occur if Containment temperature and pressure continue to rise with no operator action taken?

A. Indicated Containment Upper Pool level will lower.

B. Indicated Suppression Pool level will rise.

C. Containment Vacuum Breakers will open.

D. Drywell Vacuum Breakers will open.

NRC Exam - 2010 QUESTION RO 33 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295011 AK1.01 Importance Rating 4 K&A: Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT TEMPERATURE (MARK III CONTAINMENT ONLY):

Containment pressure: Mark-III High Containment Temp / 5 Explanation: Answer D - with stable DW pressure, as containment pressure exceeds the DW vacuum breaker opening setpoint, the DW vacuum breakers will open A - These conditions will have no effect on indicated Containment Upper Pool level.

B - These conditions will cause indicated Suppression Pool level to lower, not raise C - These conditions will cause Containment Vacuum breakers to remain closed, not open.

Technical Reference(s): ARI H13-P601-20(E4) & (F4) & Reference Attached: ARI H13-P601-20(E4)

SDM-M16 rev 4 (F4) pp. 71 & 83 and & SDM-M16 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-T23-K10.D & OT-COMBINED-D23-F.4 Question Source: Bank # Perry 2003 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 34 A plant startup is in progress after completion of a refuel outage.

Plant conditions are as follows:

Mode Switch is in STARTUP

APRM Power is 3%

IRMs are on Range 8 To protect the reactor from an inadvertent reactivity addition due to a control rod withdrawal accident the primary scram signal is ____.

A. SRM High-High Flux B. IRM Neutron Flux-High C. APRM Neutron Flux-High D. APRM Neutron Flux-High Setdown

NRC Exam - 2010 QUESTION RO 34 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295014 AK2.01 Importance Rating 3.9 K&A: Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following: RPS Inadvertent Reactivity Addition / 1 Explanation: Answer B - per Tech Spec Bases, IRM Neutron Flux-High trip is the primary scram signal.

A - incorrect - SRM scram is bypassed for RFO12 startup C - incorrect - this APRM scram is active in Mode 1 D - incorrect - this is a back-up or secondary scram to the IRM scram for the given condition.

Technical Reference(s): Tech Spec Bases 3.3.1.1 rev 0 Reference Attached: Tech Spec Bases 3.3.1.1 p B 3.3-4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C71-F.2 Question Source: Bank # Perry 2007-2 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 35 The following conditions exist:

The plant is operating at rated power

Surveillance SVI-B21-T0061-A, RPV LOW LEVEL 1 AND 2 CHANNEL A FUNCTIONAL FOR 1B21-N681A, is in progress The I&C tech just dialed down 1B21-N681A, RPV LEVEL 1, until it tripped.

If 1B21-N681B, RPV LEVEL 1, now fails downscale, the effect would be a ____.

A. direct reactor scram to preserve the integrity of the fuel cladding B. Main Steam Line drain valve isolation only to prevent off-site dose limits from being exceeded C. RWCU valve isolation only to ensure peak fuel cladding temperatures remain below the limits of 10 CFR 50.46 D. MSIV closure and subsequent reactor scram in anticipation of the complete loss of the normal heat sink and subsequent over-pressurization transient

NRC Exam - 2010 QUESTION RO 35 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295020 AK3.01 Importance Rating 3.8 K&A: Knowledge of the reasons for the following responses as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor SCRAM Inadvertent Cont. Isolation / 5 & 7 Explanation: Answer D - the reactor scrams on a MSIV isolation in anticipation of the complete loss of the normal heat sink and subsequent over-pressurization transient. Channels A & B give full MSIV isolation.

A - incorrect - simultaneous trips on B21-N0681 A & B channels cause all MSIVs to isolate which causes a reactor scram on MSIV position B & C - incorrect - more than the stated actions occur Technical Reference(s): PDB-I005 rev 9, SDM-B21(NS4) Reference Attached: PDB-I005 pp. 1, 17-20, rev 6, & Tech Spec Bases 3.3.1.1 rev 0 & 32,SDM-B21(NS4) 46, & TS Bases p B 3.3-14 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21(NS4)-E, I & J Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 36 The High Pressure Core Spray (HPCS) Pump, 1E22-C001, is being operated in the CST to CST Mode for testing.

During the test, actual Suppression Pool level rose to 18.5.

Which one of the following statements describes the effect on HPCS operation?

A. HPCS operation will be unaffected B. HPCS pumps the Suppression Pool to the CST C. HPCS Pump operates on minimum flow CST to Suppression Pool D. HPCS Pump operates on minimum flow with suction on the Suppression Pool

NRC Exam - 2010 QUESTION RO 36 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295029 EA1.02 Importance Rating 3.1 K&A: Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: HPCS High Suppression Pool Wtr Lvl / 5 Explanation: Answer D - Suppression Pool high level causes suction shift to the Suppression Pool. Both Test Valves Close. This loss of flow path causes the Minimum Flow Valve to open.

A - incorrect - misconception when operating in TEST, if in suppression pool test mode (vice CST to CST test mode) no affect B - incorrect - misconception of CST Test Return Valve closure logic C - incorrect - misconception of Suction Valve transfer logic in TEST Technical Reference(s): ARI-H13-P601-016-G5 Rev 12 Reference Attached: ARI-H13-P601-016-G5 and SDM-E22A rev 7 p 83 and SDM-E22A pp.67 & 69 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E22A-E.3 Question Source: Bank #

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

NRC Exam - 2010 QUESTION RO 37 The following conditions exist:

The plant is shutdown for a refuel outage

Fuel shuffle is in progress

Containment Vessel and Drywell Purge is operating in Refuel Mode A fuel bundle was dropped and the following annunciators alarmed on 1H13-P680:

CNTMT VENT EXH RAD HI

CNTMT VENT EXH RAD A/D HI HI/INOP Observation of radiation monitors in the Control Room reveal various monitors with elevated readings but only the Containment Ventilation Exhaust Radiation Monitors D17-K609A and D17-K609D have reached their alarm setpoints.

Based on these conditions, __(1)__ Containment Vessel and Drywell Purge fans will be running and __(2)__ of the Containment Vessel and Drywell Purge containment isolation valves will be closed.

__(1)__ __(2)__

A. no all B. no half C. all none D. some half

NRC Exam - 2010 QUESTION RO 37 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295033 EA2.01 Importance Rating 3.8 K&A: Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Area radiation levels High Secondary Containment Area Radiation Levels / 9 Explanation: Answer B - the CVDWP radiation monitors are located in the Intermediate Building (secondary containment area) and will initiate an isolation signal to prevent a spread of radiation/contamination into secondary containment. A HI/HI alarm on D17-K609A and D radiation monitor causes an M14 Outboard Isolation while a HI/HI on the B and C radiation monitors causes an iM14 inboard isolation. The isolation causes of loss of suction path trip for the fans. All fans trip but only 1/2 (the outboard) valves go closed.

A - incorrect - the inboard isolation valves remain open C - incorrect - all of the fans trip and half of the isolation valves close D - incorrect - all of the fans trip Technical Reference(s): ARI-H13-P680-007 rev 18 and Reference Attached: ARI-H13-P680-007 pp EOP-3 Bases rev 0 27, 29-30 and EOP-3 Bases p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3407-17-C Question Source: Bank #

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 - 2010 QUESTION RO 38 The following conditions exist:

The plant is in MODE 4.

RHR Loop A is running in the Shutdown Cooling Mode.

RHR Loop B is running in the Suppression Pool Cooling Mode for surveillance testing Which one of the following describes the automatic response of the RHR system if a valid RPV Level 1 Reactor water level condition occurs?

A. RHR Pump A trips.

RHR Loop B realigns to the LPCI mode.

B. RHR Pump A continues to operate in the Shutdown Cooling Mode.

RHR Loop B realigns to the LPCI mode.

C. RHR Pumps A and B trip.

RHR Loop B then shifts to the LPCI mode and RHR Pump B restarts.

D. RHR Pump A trips.

RHR Loop B continues to operate in the Suppression Pool Cooling mode.

NRC Exam - 2010 QUESTION RO 38 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 203000 K4.09 Importance Rating 3.1 K&A: Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC)design feature(s) and/or interlocks which provide for the following: Surveillance for all operable components RHR/LPCI: Injection Mode Explanation: Answer A - RHR pump A trips due to RHR-SDC valves F008 and F009 going close on the L3 signal. RHR B loop will realign for LPCI injection at L1. There is no automatic realignment of RHR from SDC to LPCI.

B - incorrect - RHR pump A trips due to RHR-SDC valves F008 and F009 going close on the L3 signal C - incorrect - RHR Pump B does not trip and then restart D - incorrect - RHR B loop will realign for LPCI injection at L1 Technical Reference(s): ARI-H13-P601-017-C4 rev 12, Reference Attached: ARI-H13-P601-017-C4 p ARI-H13-P601-020-F6 rev 15, PDB-I-005 rev 9, & SOI- 37, ARI-H13-P601-020-F6 p 89, PDB-I005 E12 rev. 46 p 35, & SOI-E12 pp.9-10 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E12-H.1.A Question Source: Bank # INL-36851 Modified Bank #

New Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 39 RHR Loop A has just been placed into the Shutdown Cooling Mode of operation using the Normal Return Path.

The cooldown rate is excessive.

In accordance with SOI-E12, Residual Heat Removal System, which of the following actions will lower the cooldown rate?

A. Throttle close the RHR A HXS BYPASS VALVE, E12-F048A, and throttle open the RHR A HXS OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 7000-7100 gpm.

B. Throttle open the RHR A HXS BYPASS VALVE, E12-F048A, and throttle close the RHR A HXS OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 7000-7100 gpm.

C. Throttle open the RHR A HXS BYPASS VALVE, E12-F048A, and throttle close the RHR A HXS OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 2575-7100 gpm.

D. Throttle ESW flow through the RHR Heat Exchanger using RHR A HXS ESW OUTLET VALVE, P45-F068A.

NRC Exam - 2010 QUESTION RO 39 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 205000 K5.03 Importance Rating 2.8 K&A: Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): Heat removal mechanisms Shutdown Cooling Explanation: Answer B - this is the approved method for adjusting the cooldown rate while in Normal Mode SDC.

A - incorrect - this action will raise the cooldown rate C - incorrect - this is the required flow rate band when using the alternate return path via E12-F042A D - incorrect -not an approved method to control the cooldown rate per SOI-E12 Technical Reference(s): SOI-E12 rev 26 Reference Attached: SOI-E12 pp 26, 31-32 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E12-B.4.C Question Source: Bank # Perry 2001 Modified Bank #

New Question History: Previous NRC Exam: Perry 2001 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 - 2010 QUESTION RO 40 The following conditions exist:

A plant startup is in progress

Reactor power 25%

Reactor Recirculation Flow Control Valves are full open

APRM A is in Bypass OPRM A receives a Core Flow signal from APRM __(1)__ and if power oscillations occur, OPRM A __(2)__ trip RPS.

__(1)__ __(2)__

A. A will not B. E will not C. A will D. E will

NRC Exam - 2010 QUESTION RO 40 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215005 K6.04 Importance Rating 3.2 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the OSCILATING POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: Trip units OPRM Explanation: Answer D - when the APRM is bypassed, the associated OPRM automatically receives a core flow signal from the other APRM in the same panel. The OPRM trip is enabled at > 23.3 % power and drive flow < 63.5%.

A & C - incorrect - OPRM A receives core flow signal from APRM E A & B - incorrect - the OPRM trip is enabled and will input a trip signal into RPS Technical Reference(s): SOI-C51(APRM) rev 9 and ARI- Reference Attached: SOI-C51(APRM) p 14 H13-P601-006-A2 rev 7 and ARI-H13-P601-006-A2 p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51-AP_OPRM-E.2 & 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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 41 The Plant is operating at 75% rated power.

One hour ago the blue pressure permissive light for the LPCS Injection Valve, 1E21-F005, extinguished.

Control Room Operators confirmed the blue light bulb was good.

Which one of the following describes the operation of the LPCS Injection Valve control logic if a Loss-of Coolant Accident now occurs?

The LPCS Injection Valve ____.

A. remains closed and cannot be opened with its control switch until RPV pressure lowers to 600 psig B. automatically opens, irrespective of RPV pressure, due to the LPCS LOCA initiation signal.

C. remains closed and cannot automatically open until RPV pressure lowers to 530 psig D. remains closed and cannot automatically open until RPV pressure lowers to 600 psig

NRC Exam - 2010 QUESTION RO 41 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209001 A1.03 Importance Rating 3.8 K&A: Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURECORE SPRAY SYSTEM controls including: Reactor water level LPCS Explanation: Answer D - the Blue permissive light indicates that pressure downstream of the injection valve is <600 psig and will allow the injection valve to automatically open.

A - incorrect - the injection valve control switch bypasses the pressure permissive B - incorrect - the blue light off indicates pressure permissive is not met no automatic opening C - incorrect - 600 psig is LPCS injection valve open permissive, 530 psig for RHR Technical Reference(s): ARI-H13-P601-021-A6 rev 13 Reference Attached: ARI-H13-P601-021-A6 and SDM-E21 rev 10 p 17 and SDM-E21 pp.23-25 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E21-E.1 & 3 Question Source: Bank # Perry 2001-2 Modified Bank #

New Question History: Previous NRC Exam: Perry 2001-2 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 - 2010 QUESTION RO 42 HPCS is operating in CST to CST Mode for PMT

The operator has adjusted HPCS flow to 600 gpm to shutdown HPCS to Standby Readiness per SOI-E22A Section 7.10.18, HPCS Full flow Test to CST

The HPCS Minimum flow valve 1E22-F012 remains closed As the Reactor Operator continues to shutdown the pump, in order to prevent HPCS pump damage caused by __(1)__, the required operator action is to __(2)__.

__(1)__ __(2)__

A. runout open the HPCS PUMP MIN FLOW VALVE, 1E22-F012 B. runout lower flow until 1E22-F0012 automatically opens C. operating at shutoff head open the HPCS PUMP MIN FLOW VALVE, 1E22-F012 D. operating at shutoff head lower flow until 1E22-F0012 automatically opens

NRC Exam - 2010 QUESTION RO 42 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 209002 A2.10 Importance Rating 2.7 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: Valve openings: BWR-5,6 HPCS Explanation: Answer C - HPCS min flow valve should auto open as flow is lowered to 725 gpm to prevent the pump from running at shutoff head. SOI-E22A P&L says to verify min flow valve open prior to stopping flow in CST to CST mode. NOP-OP-1002, it is the responsibility of the operator to take manual actions if automatic actions fail to occur.

A & B - incorrect - runout is a condition of excess pump flow - in this case the pump does not have the minimum required flow B & D - incorrect - all sections of SOI-E22A that contain min flow valve operations say to verify min flow valve open - none discuss lowering pump flow below the minimum flow requirements Technical Reference(s): SOI-E22A rev 25 & NOP-OP- Reference Attached: SOI-E22A pp.5 and 45 1002 rev 5 & NOP-OP-1002 p 43 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E22A-E.3 & OT-3039-01-B Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 43 The following conditions exist:

An ATWS occurred

Standby Liquid Control (SLC) was initiated

The SLC A Squib valve, 1C41-F004A, failed to fire

The SLC B Pump, 1C41-C001B, failed to start

Reactor Pressure is 1000 psig Considering these failures, which one of the following describes the indications for SLC Pump A and B discharge pressure that would be observed on 1H13-P601.

SLC Pump A discharge pressure will be approximately __(1)__ and SLC Pump B discharge pressure will be approximately __(2)__.

__(1)__ __(2)__

A. 1100 psig 1100 psig B. 1100 psig 0 psig C. 1300 psig 1300 psig D. 1300 psig 0 psig

NRC Exam - 2010 QUESTION RO 43 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 211000 A3.01 Importance Rating 3.5 K&A: Ability to monitor automatic operations of the STANDBY LIQUID CONTROL SYSTEM including: Pump discharge pressure SLC Explanation: Answer A - both instruments show discharge pressure even though only one pump starts.

Since one squib valve is open, a flow path exists, thus pressure will be slightly above reactor pressure.

B - incorrect - due to the cross-connect line between the pumps, both meters will show discharge pressure C & D - incorrect - since the pumps are rated at 1250 psig, a 1300 psig is an indication that the relief valve (1C41-F029A/B) is lifting, which is not the case since one squib valve fired Technical Reference(s): SOI-C41 rev 18, Dwg. 302-691 Reference Attached: SOI-C41 pp 6 & 9, Dwg.

rev V (Partial), & SDM-C41 rev 8 302-691 rev V (Partial), & SDM-C41 p 8 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C41-C.3 & OT-3402-D.2 Question Source: Bank # RQL-0047 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 - 2010 QUESTION RO 44 At 10:00 the following conditions exist:

The plant is operating at 100% rated power At 10:01 the following conditions exist:

A load rejection with a loss of turbine EHC occurred

Safety/Relief valves responded as designed At 10:02 the following conditions exist:

Motor Feedwater pump is running on minimum flow

Reactor power is 5%

Scram Hardcard actions have been completed At 10:10 the following conditions exist:

All rods are in

Reactor pressure is being controlled on Bypass Valves

Setpoint Setdown has been reset

Motor Feedwater pump is running in manual, feeding the RPV

The Unit Supervisor is operating in ONI-C71-1, Reactor Scram At this time, the Redundant Reactivity Control System logic __(1)__ be reset and RPS logic __(2)__ be reset.

__(1)__ __(2)__

A. can can B. can not can C. can can not D. can not can not

NRC Exam - 2010 QUESTION RO 44 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 212000 2.4.11 Importance Rating 4 K&A: Knowledge of abnormal condition procedures.

RPS Explanation: Answer B - at time 10:10 there are no RPS scram signals nor RRCS initiation signals present. Therefore, RPS can be reset. However, ONI-C71-1, Reactor Scram contains a NOTE that informs the operator that RRCS logic contains a 12 minute timer, which does not allow reset of the RRCS logic until the timer times out.

A & C - incorrect - RRCS logic can not be reset due to the 12 minute timer C & D - incorrect - RPS logic can be reset as there are no scram signals present Technical Reference(s): ONI-C71-2 rev 8 Reference Attached: ONI-C71-2 p 7 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT3035-11-(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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 45 The plant is operating at 100% power with the Reactor Protection System MG SET TRANSFER switch in ALT A.

The following occurs:

Numerous Control Room Alarms are received

Half scram is indicated

Outboard BOP isolation has occurred This is an indication of a loss of power from Bus ____.

A. F1B08 B. F1C08 C. F1C12 D. F1D12

NRC Exam - 2010 QUESTION RO 45 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 212000 K1.01 Importance Rating 3.4 K&A: Knowledge of the physical connections and/or cause/effect relationships between REACTOR PROTECTION SYSTEM and the following: A.C. electrical distribution RPS Explanation: Answer B - with RPS MG Set Transfer switch in ALT A position, RPS bus A is powered from F1C08. Loosing RPS bus A causes an outboard isolation and 1/2 scram.

A - incorrect - this is the normal power supply for RPS A C - incorrect - this is the normal power supply for RPS B D - incorrect - this is the alternate power supply for RPS B Technical Reference(s): PDB-H014 rev 1, PDB-H15 Reference Attached: PDB-H15 p 5 &

rev 2, PDB-H16 rev 1, & SDM-C71 rev 10 SDM-C71 p 78 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C71 D.1 Question Source: Bank #

Modified Bank # Perry 2007-2 New Question History: Previous NRC Exam: Perry 2007-2 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 - 2010 QUESTION RO 46 What is the power source for the high voltage power supply for IRM H?

A. Division 1 ATWS Dist. Panel 1R14-S014 B. Division 2 ATWS Dist. Panel 1R14-S015 C. RPS A Bus, C71-P001 D. RPS B Bus, C71-P002

NRC Exam - 2010 QUESTION RO 46 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215003 K2.01 Importance Rating 2.5 K&A: Knowledge of electrical power supplies to the following: IRM channels/detectors IRM Explanation: Answer D - RPS B supplies IRM H.

A & B - incorrect - the ATWS panels supply the APRMs not the IRMs C - incorrect - RPS A supplies IRMs A, C, E, & G Technical Reference(s): SOI-C71 rev 18 & SDM- Reference Attached: SOI-C71 p 111 & SDM-C51(IRM) rev 7 C51(IRM) p 53 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51(IRM)-B.3.E Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 47 The following plant conditions exist:

A reactor startup is in progress following replacement of all fuel bundles

Reactor Protection System shorting links are removed

Reactor power is rising with a stable positive period of 150 seconds

SRM Channel A detector is stuck and will not withdraw

SRM Channel A indication rises to 2.5x105 cps

No operator actions are performed Which one of the following subsequently describes SRM Channel A indicated reactor power and reactor period?

Indicated reactor power will __(1)__. Reactor period will __(2)__

__(1)__ __(2)__

A. lower become negative B. lower remain stable and positive C. continue to rise become shorter D. continue to rise remain stable and positive

NRC Exam - 2010 QUESTION RO 47 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215004 K3.04 Importance Rating 3.7 K&A: Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following: Reactor power and indication Source Range Monitor Explanation: Answer A - with shorting links removed, SRM trips are active and non-coincident. As power rises above 2E+5, the reactor will scram on high flux and reactor period will be negative.

B & D - incorrect - Rx period will not remain stable and positive C & D - incorrect - Rx power will lower due to the scram Technical Reference(s): SOI-C51(SRM) rev 6 Reference Attached: SOI-C51(SRM) p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C51(SRM)-D.4 & OT-COMBINED -C71-O.5 Question Source: Bank # Perry 2002 Modified Bank #

New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 48 The following conditions exist:

The plant is operating at 48% rated power

Total Core Flow is 50 MLbs/Hr

Both Recirculation are operating in fast speed What is the current Upscale Thermal Power Trip setpoint rounded to the nearest %?

Reference Provided: PDB-A0012, Recirc Drive Flow vs. Total Core Flow A. 107%

B. 108%

C. 111%

D. 113%

NRC Exam - 2010 QUESTION RO 48 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215005 K4.07 Importance Rating 3.7 K&A: Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:

Flow biased trip setpoints APRM / LPRM Explanation: Answer C - the APRM Upscale Thermal Power Trip setpoint is calculated using the formula Setpoint=0.628Wr + 61% (where Wr is % drive flow) and is clamped at 111%.

A - incorrect - this is the current APRM Upscale Thermal Power Alarm setpoint B - incorrect - this is the clamped value for the APRM Upscale Thermal Power Alarm setpoint D - incorrect - this is the calculated setpoint, however, it is clamped at 111%

Technical Reference(s): SDM-C51 (PRM & OPRM) Reference Attached: SDM-C51 (PRM &

rev 10 and PDB A12 rev 15 OPRM) pp 23-25 and PDB-A12 p 6 Proposed references to be provided to applicants during examination: PDB-A0012, Recirc Drive Flow vs.

Total Core Flow Learning Objective (As available): OT-COMBINED-C51(AP_OPRM)-D.8 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 49 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 The receipt of which annunciator would indicate an immediate threat to RCICs ability to maintain RPV level?

A. RCIC ISOL DIAPHRAGM RUPTURED, (H13-P601-0021-B1)

B. STEAM TUNNEL LD AMB TEMP P632, (H13-P601-0019-G4)

C. RCIC TURBINE OIL COOLER OUT TEMP HIGH, (H13-P601-0021-C4)

D. RCIC SUPR POOL SUCT VLV OPEN SUPR PL LVL HI, (H13-P601-0021-G5)

NRC Exam - 2010 QUESTION RO 49 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 K5.07 Importance Rating 3.1 K&A: Knowledge of the operational implications of the following concepts as they apply to REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Assist core cooling RCIC Explanation: Answer A - an exhaust diaphragm rupture will cause an immediate RCIC turbine trip.

B - incorrect - the steam tunnel high temperature has a 29 minute time delay to isolate C - incorrect - high RCIC lube oil temp does not cause an immediate loss of the turbine D - incorrect - a high suppression level will cause a suction shift, but the suppression pool valve opens fully before the CST valve closes - in this case it doesnt matter which source is lined up.

Technical Reference(s): ARI-H13-P601-021 rev 13, ARI- Reference Attached: ARI-H13-P601-021 H13-P601 rev 11, & SOI-E31 rev 8 pp. 19, 39, & 89, ARI-H13-P601 p111, & SOI-E31 p 19 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E51-D.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 50 The following plant conditions exist:

The reactor is operating at 100% power

A loss of ED-1-A occurs Which of the following describes the ability of the SRVs to function in the Pressure-Relief Mode and in the ADS Mode?

In the Pressure-Relief Mode, __(1)__ SRVs will function and in the ADS Mode, actuation of the Div. II ADS logic __(2)__ open the ADS valves.

__(1)__ __(2)__

A. no will B. no will not C. all will D. all will not

NRC Exam - 2010 QUESTION RO 50 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 218000 K6.06 Importance Rating 3.4 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM: D.C. power:

ADS Explanation: Answer C - the Pressure Relief mode will function using the B solenoids and the ADS function will function using the Div 2 logic A & B - incorrect - the pressure relief mode will still function utilizing the B solenoids.

D - incorrect - Div 2 ADS logic is independent of Div 1 power supply and can still open the ADS valves Technical Reference(s): PDB-I005 rev 9 and SDM-B21C Reference Attached: PDB-I005 p 14 and rev 6 SDM-B21 p 22 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21C-E.1 & OT-COMBINED-R42-H Question Source: Bank # Perry 2003 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 51 An isolation signal was generated that resulted in the following light indications on panel H13-P622, Div 2 Aux Relay Panel Inboard Valves:

NS4 MSL DRN ISOL INBD LOGIC TEST B21H-DS1B light on

NS4 BOP ISOL INBD LOGIC TEST B21H-DS3B light on

RX WTR SMPL VLV B33-F019 LOGIC TEST B21H-DS7B light off

RHR ISOLATION INBD LOGIC TEST B21H-DS5B light off Based on these indications, select the isolation that occurred.

Reference Provided: Picture of panel H13-P622 A. BOP isolation B. RWCU isolation C. RHR LOCA isolation D. RHR Radwaste Valve isolation

NRC Exam - 2010 QUESTION RO 51 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 A1.01 Importance Rating 3.5 K&A: Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: System indicating lights and alarms PCIS/Nuclear Steam Supply Shutoff Explanation: Answer D - the RHR Radwaste Valve isolation is Level 3 and 1.68 psig. The light goes out if isolation signal present. The candidate needs to know if the lights are on or off for the isolation.

A - incorrect - the BOP isolation is actuated on Level 2 and 1.68 psig in DW - the light would be off B - incorrect - the RWCU isolation is actuated on Level 2 and Leak Detection signals - the BOP light would be off if hit L2 C - incorrect - the RHR LOCA isolation is actuated on Level 1 and 1.68 psig in DW - the light would be off Technical Reference(s): IOI-18 rev 6 and DWGS 208- Reference Attached: IOI-18 p 82 and DWGS 013-008 rev P & 208-013-012 rev FF 208-013-008 & 208-013-012 Proposed references to be provided to applicants during examination: Picture of panel H13-P622 Learning Objective (As available): OT-COMBINED-B21(NS4)-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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 52 The plant is operating at full power.

An indication that an SRV is open is __(1)__.

In accordance with ONI-B21-1, SRV Inadvertent Opening/Stuck Open, do not attempt to close the SRV until reactor power is lowered to __(2)__ percent.

__(1)__ __(2)__

A. Main generator electrical output lowers 90 B. Main generator electrical output lowers 95 C. Indicated steam flow on the affected 90 steam line rises D. Indicated steam flow on the affected 95 steam line rises

NRC Exam - 2010 QUESTION RO 52 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 239002 A2.03 Importance Rating 4.1 K&A: Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open SRV SRVs Explanation: Answer A - an indication that an SRV is open is lowering of main generator MWe - per ONI-B21-1, Rx power is lowered to 90%

B & D - incorrect - reactor power is to be lowered to less than 90% prior to attempting to close an SRV C & D - incorrect - the flow orifices are down stream of the SRVs and indicated steam flow will lower not rise Technical Reference(s): ONI-B21-1 rev 9 Reference Attached: ONI-B21-1 pp.3-4 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 x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 53 The following conditions exist:

The plant is at 100% power

Digital Feedwater Operator Rx Level Setpoint was set at 200 inches A manual scram is now inserted.

Which of the following describes the response of the DFWLCS following the scram?

A. Upon receipt of the scram signal, the level demand signal will be 196 inches for 10 seconds and then lower to 178 inches.

B. Upon receipt of the scram signal, the level demand signal will be 200 inches for 10 seconds and then lower to 178 inches.

C. When level reaches 178 inches, the level demand signal will be 196 inches for 10 seconds and then lower to 178 inches.

D. When level reaches 178 inches, the level demand signal will be 200 inches for 10 seconds and then lower to 178 inches.

NRC Exam - 2010 QUESTION RO 53 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 259002 A3.06 Importance Rating 3.0 K&A: Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: Reactor water level setpoint setdown following a reactor scram: Plant-Specific Reactor Water Level Control Explanation: Answer D - with the Operator Rx Level Setpoint set at 200, when RPV level drops below L3, Setpoint Setdown logic demands the Operator Rx Level Setpoint setting for 10 seconds then lowers to 178 A & B - incorrect - the scram signal does not initiate Setpoint Setdown logic A & C - incorrect - the Operator Rx Level Setpoint was set at 200, therefore the Setpoint Setdown logic demands 200 not 196 Technical Reference(s): OT-COMBINED-C32 LP rev 4 Reference Attached: OT-COMBINED-C32 LP p 32 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C34-F.5.c Question Source: Bank # INL-3034 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 - 2010 QUESTION RO 54 The following plant conditions exist:

The reactor is operating at 90% power

Annulus Exhaust Gas Treatment System (AEGTS) Train B is in service

An unplanned gaseous radioactive release occurs in the Annulus Which one of the following Airborne Radiation Monitors (ABRM) would detect this radioactive release in the Annulus?

A. Unit 1 (U1) Plant Vent, 1D17-K0780 B. Unit 2 (U2) Plant Vent, 2D17-K0780 C. Intermediate Bldg Vent, D17-K730 D. Aux Bldg Vent, 1D17-K700

NRC Exam - 2010 QUESTION RO 54 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 261000 A4.01 Importance Rating 3.2 K&A: Ability to manually operate and/or monitor in the control room: Off-site release levels SGTS Explanation: Answer B - The discharge of each AEGTS fan branches into two headers. One header recirculates air flow back into the annulus. The other header exhausts air to the unit's vent, AEGTS subsystem A to the Unit 1 Vent, AEGTS subsystem B to the Unit 2 Vent.

A - incorrect - AEGTS A exhausts to U1 plant vent C & D - plausible due to misconception that AEGTS exhausts to these ventilation systems Technical Reference(s): Dwg 912-605 rev W Reference Attached: Dwg 912-605 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M15-B.1 Question Source: Bank #

Modified Bank # Perry 2002 New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 55 The plant is operating at full power with Division 1 Diesel Generator operating in parallel with the grid when the following valid alarms were received:

BUS EH11 VOLTAGE DEGRADATION (H13-P877-0001-B1)

BUS EH12 VOLTAGE DEGRADATION (H13-P877-0002-B1)

BUS EH12 BUS STRIPPED UNDERVOLTAGE (H13-P877-0002-C1)

What additional condition is required to initiate the LOOP logic?

A. 3 seconds elapses.

B. 12 seconds elapses.

C. Bus EH11 frequency at 58 hertz.

D. Bus EH12 frequency at 57 hertz.

NRC Exam - 2010 QUESTION RO 55 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262001 2.4.45 Importance Rating 4.1 K&A: Ability to prioritize and interpret the significance of each annunciator or alarm.

AC Electrical Distribution Explanation: Answer C - the LOOP logic requires both bus EH11 & bus EH12 to have low voltage (3KV) or low frequency (59 Hz). The low voltage is made up for the EH12 bus, so when frequency on bus EH11 lowers below 59 Hz, the logic will be made up.

A - incorrect - this conditions is already fulfilled when the Bus Stripped alarm is received B - incorrect - this conditions is true if a LOCA signal were present D - incorrect - the LOOP logic is already made up for bus EH12, therefore this signal will have no effect Technical Reference(s): SOI-S11 rev 5 Reference Attached: SOI-S11 pp. 3-4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R10-D.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 56 A fire in the Control Room forced all personnel to abandon the Control Room.

A reactor scram could not be initiated prior to evacuating the Control Room.

Which one of the following describes the preferred method for initiating a reactor scram, including the bases for this method?

Cycle the specified ____.

A. ATWS UPS distribution panel breakers since this will not cause a MSIV closure B. ATWS UPS distribution panel breakers since this will not cause a loss of LPRMs/APRMs C. RPS power distribution panel breakers since this will not cause a MSIV closure D. RPS power distribution panel breakers since this will not cause a loss of LPRMs/APRMs

NRC Exam - 2010 QUESTION RO 56 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 262002 K1.19 Importance Rating 2.9 K&A: Knowledge of the physical connections and/or cause/effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: Power range neutron monitoring system: Plant-Specific UPS (AC/DC)

Explanation: Answer A - cycling the ATWS breakers is preferred because this method deenergizes the APRM to cause a scram and allows the MSIVs to remain open.

B - incorrect - this will cause the APRMs to lose power C - incorrect this will cause the MSIVs to close D - incorrect - this is not the preferred method Technical Reference(s): ONI-C61 rev 6 Reference Attached: ONI-C61 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): x Question Source: Bank # Perry 2002 Modified Bank #

New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 57 Which of the following is the power supply for the listed load?

1N43-C007, Turb Emg Bearing Oil Pump RHR B/C Logic Main Generator Trip Logic A. D1A ED1A D1B B. D1B ED1B D1B C. D1A ED1A D1A D. D1B ED1B D1A

NRC Exam - 2010 QUESTION RO 57 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 263000 K2.01 Importance Rating 3.1 K&A: Knowledge of electrical power supplies to the following: Major D.C. loads DC Electrical Distribution Explanation: Answer D - DC bus D1B supplies Emergency Bearing oil pump, D1A supplies the main generator trip logic, and ED1B supplies RHR B/C logic.

A - incorrect - all three selections are incorrect B - incorrect - D1B does not supply main generator trip logic C - incorrect - wrong selections for emergency bearing oil pump and RHR logic Technical Reference(s): PDB-H002 rev 2, PDB-H004 rev Reference Attached: PDB-H002 p 12, PDB-3, & PDB-H005 rev 2 H004 pp. 13-14, & PDB-H005 p 2 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R42-B.2.C Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 58 The following conditions exist:

A Loss of Off-Site Power (LOOP) occurred

Divisional diesel generators are supplying their respective buses

DIESEL GENERATOR CONTROL TRANSFER switch is in CONT RM on 1H51-P055A

DIESEL GENERATOR CONTROL TRANSFER switch is in LOCAL on 1H51-P055B The following alarms are then received on H13-P877:

Division 1

DIV 1 DIESEL GENERATOR TROUBLE, (H13-P877-0001-D5)

DG TRIP* LUBE OIL TEMP HIGH, (H13-P877-0001-C3)

Division 2

DIV 2 DIESEL GENERATOR TROUBLE, (H13-P877-0002-D5)

DG TRIP DIFF RELAY LOCKOUT, (H13-P877-0002-E4)

What will be the status of Buses EH11 and EH12?

EH11 EH12 A. Energized Energized B. Energized De-energized C. De-energized Energized D. De-energized De-energized

NRC Exam - 2010 QUESTION RO 58 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 264000 K3.02 Importance Rating 3.9 K&A: Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: A.C. electrical distribution EDGs Explanation: Answer B - the Lube Oil high temperature trip is bypassed with a LOOP signal present when the DG CONTROL TRANSFER switch is in CR position and the DIFF RELAY LOCKOUT trip (Div

2) is always active A - incorrect - EH12 will be deenergized C & D - incorrect - EH11 will be energized Technical Reference(s): SOI-R43 rev 36, ARI-H13-P877- Reference Attached: SOI-R43 p 6, ARI-H13-001 rev 10, ARI-H13-P877-002 rev 11, & ARI-H51- P877-001 pp 31& 44, ARI-H13-P877-002 P054B rev 7 pp. 44 & 53, & ARI-H51-P054B p 97 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-R43_48-D.9 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 59 Which of the following will automatically trip on high temperature on a complete loss of Nuclear Closed Cooling?

A. Control Complex Chiller B, P47-B001B B. Control Rod Drive Pump A, 1C11-C001A C. Reactor Recirculation Pump A, 1B33-C001A D. Unit 1 Instrument Air Compressor, 1P52-C001

NRC Exam - 2010 QUESTION RO 59 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 300000 K4.03 Importance Rating 2.8 K&A: Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following: Securing of IAS upon loss of cooling water Instrument Air Explanation: Answer D - NCC cools the IAC lube oil. On a loss of NCC, lube oil will heat up to the trip setpoint.

A - incorrect -CC chiller B is supplied from ECC, CC chiller C is supplied from NCC B & C - incorrect - CRD and Recirc pumps do not have automatic trips on high temp Technical Reference(s): ARI-H13-P904 rev 9, ARI-H13- Reference Attached: ARI-H13-P904 p 53, P680-004 rev 14, & ONI-P43 rev 11 ARI-H13-P680-004 pp. 107 & 129 & ONI-P43 p 17 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P43-H & OT-3035-16(LP)-A.3 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 60 The following conditions exist:

The plant is operating at 75% reactor power.

SVI-C71-T0039, MSIV Closure Channel Functional, is in progress

MSL B INBD MSIV B21-F022B control switch is in the TEST position.

The Control Room Operator depresses the MSL B INBD MSIV TEST pushbutton 1B21H-S3B.

Which one of the following describes the response of MSL B INBD MSIV B21-F022B?

A. Safety-Related Instrument Air bleeds off the bottom portion of the MSIV air cylinder and the top portion of the MSIV air cylinder is pressurized to stroke the MSIV closed in 3-5 seconds.

B. Instrument Air bleeds off the bottom portion of the MSIV air cylinder and the top portion of the MSIV air cylinder is pressurized to stroke the MSIV closed in 3-5 seconds.

C. Safety-Related Instrument Air bleeds off the bottom portion of the MSIV air cylinder causing the MSIV to slowly close.

D. Instrument Air bleeds off the bottom portion of the MSIV air cylinder causing the MSIV to slowly close.

NRC Exam - 2010 QUESTION RO 60 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 300000 K1.05 Importance Rating 3.1 K&A: Knowledge of the connections and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: Main Steam Isolation Valve air Instrument Air Explanation: Answer D - air is bled off the MSIV air cylinder allowing it to slow close.

A & B - incorrect - bleeding air off the MSIV allows it to close slowly - not 3-5 seconds which is the normal stroke time A & C - incorrect - Safety Related Instrument Air (P57) is not the source of operating air to the MSIVs Technical Reference(s): SDM-B21/N11 rev 10 Reference Attached: SDM-B21/N11 pp.25 &

58 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21_N11-E.2 Question Source: Bank # Perry 2001 Modified Bank #

New Question History: Previous NRC Exam: Perry 2001 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 - 2010 QUESTION RO 61 What is the power supply to Nuclear Closed Cooling Pump C?

A. H11 B. H21 C. XH11 D. XH21

NRC Exam - 2010 QUESTION RO 61 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 K2.01 Importance Rating 2.9 K&A: Knowledge of electrical power supplies to the following: CCW pumps Component Cooling Water Explanation: Answer D - XH2102 from bus XH21 is power supply for NCC pump C A & D - incorrect - not the correct power supply C - incorrect - this is the supply to NCC pump A Technical Reference(s): ELI-R22 rev 8 Reference Attached: ELI-R22 p 25 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P43-C.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 62 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 - 2010 QUESTION RO 62 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 KKKK Reference Attached: Dwg 206-021 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C41-E.1, OT-COMBINED-R10-C.17 Question Source: Bank # Perry 2007-1 Modified Bank #

New Question History: Previous NRC Exam: Perry 2007-1 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 - 2010 QUESTION RO 63 The following conditions exist:

The plant is operating at 100% power

Lake water temperature is 55oF

TBCC Temperature Control Valve, 1P44-F300 is at midposition The lake rolls over causing lake water temperature to rise to 63oF.

The rise in lake water temperature causes the temperature control valve to ____.

A. raise the TBCC flow through the heat exchanger B. lower the TBCC flow through the heat exchanger C. raise the Service Water flow through the heat exchanger D. lower the Service Water flow through the heat exchanger

NRC Exam - 2010 QUESTION RO 63 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 400000 A1.02 Importance Rating 2.8 K&A: Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including: CCW temperature Component Cooling Water Explanation: Answer A - the TBCC temperature control valve splits the TBCC flow through the HX and bypassing the HX. As the lake temperature rises, SW temperature rises. This causes the TCV to bypass less TBCC flow around the HX to maintain TBCC outlet temperature constant. This is a recent modification to the plant to throttle TBCC vs. SW flows.

B - incorrect - this is true if lake water temperature lowers C & D - incorrect - SW is full flow through the TBCC HXs - plausible is candidate does not understand which medium is throttled Technical Reference(s): SDM-P44 rev 6 Reference Attached: SDM-P44 pp 4 & 23 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-P44-B.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 64 The plant is operating at full power with the following conditions:

A scram signal is present

The Scram Inlet Valve, EP-126, for control rod 30-31 fails to open Which of the following describes how control rod 30-31 is inserted?

The Control Rod Drive ball check valve repositions directing ____ the drive piston to insert the control rod.

A. reactor water above B. reactor water below C. accumulator water above D. accumulator water below

NRC Exam - 2010 QUESTION RO 64 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201003 K1.02 Importance Rating 2.9 K&A: Knowledge of the physical connections and/or cause/effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following: Reactor water Control Rod and Drive Mechanism Explanation: Answer B - when the scram outlet valve opens, the ball check valve will reposition since it sences accumulator pressure < Rx pressure and directs Rx water under the drive piston to insert the control rod.

A - incorrect - Rx water is directed below the piston not above C & D - incorrect - the ball check valve does not reposition to allow accumulator water to drive the rod -

the accumulator is normally lined up to drive the rod Technical Reference(s): SDM-C22(CRDH) rev 5 Reference Attached: SDM-C22(CRDH) pp. 6

& 37 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11(CRDM)-C.1 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 65 Which of the following describes the operational implications of maintaining control rods within designated withdrawal distances during control rod movements above the high power setpoint?

A. Establish a 2 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

B. Establish a 4 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

C. Establish a 2 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

D. Establish a 4 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

NRC Exam - 2010 QUESTION RO 65 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201005 K5.09 Importance Rating 3.5 K&A: Knowledge of the operational implications of the following concepts as they apply to ROD CONTROL AND INFORMATION SYSTEM (RCIS): High power setpoints RCIS Explanation: Answer C - RC&IS enforces the limits of the rod withdrawal limiter above the HPSP and limits rod withdrawal to 2 notches to mitigate the consequences of a rod withdrawal error.

A & B - incorrect - the rod pattern controller of RC&IS is designed to mitigate the consequences of a control rod drop accident B & D - incorrect - the 4-notch limit is enforced between the LPSP and the HPSP Technical Reference(s): Tech Spec Bases 3.3.2.1 rev 0 Reference Attached: Tech Spec Bases

& SOI-C11(RCIS) rev 26 3.3.2.1 pp. B 3.3-42 & 43 & SOI-C11(RCIS) p 6

Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-C11(RCIS)-G & J.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 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 66 The following conditions exist:

Reactor is operating at rated power

Reactor Recirculation FCVs are at 70% open

APRM A is in Bypass

APRM E ramped to 125%

No operator actions were taken The FCVs are now __(1)__ open .

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

__(1)__ __(2)__

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

NRC Exam - 2010 QUESTION RO 66 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 202002 A3.01 Importance Rating 3.6 K&A: Ability to monitor automatic operations of the RECIRCULATION FLOW CONTROL SYSTEM including: Flow control valve operation: BWR-5,6 Recirculation Flow Control Explanation: Answer B - the FCVs will runback 40% from the original starting point. The FCVs will return to the original position if no operator action is taken.

A & C - incorrect - if the runback was not caused by AFDL, then the FCVs would remain in the current position C & D - incorrect - this is correct if the FCVs would have started at 100% open - plausible if the candidate thinks FCVs will close 40% from 100%

Technical Reference(s): SDM-B33 rev 9 Reference Attached: SDM-b33 p 43 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B33-E.16 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 67 The following conditions exist:

The control room was evacuated per ONI-C61, control Room Evacuation

The Plant Computer lost power

Preparation in progress to place Shutdown Cooling in service in accordance with IOI-11, Shutdown from Outside Control Room While lowering reactor pressure to place Shutdown Cooling in service the following reactor pressure readings were obtained from 1C61-R011, Reactor Pressure, on the Remote Shutdown Panel, C61-P001, at the indicated times:

Time Rx pressure 1200 1000 psig 1300 425 psig 1400 100 psig 1500 25 psig Which one of the following choices completes the following statement?

The reactor cooldown rate specified in IOI-11 ____.

Reference provided - Steam Tables A. has not been exceeded B. was exceeded between 1200 and 1300 C. was exceeded between 1300 and 1400 D. was exceeded between 1400 and 1500

NRC Exam - 2010 QUESTION RO 67 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 216000 K4.01 Importance Rating 3.6 K&A: Knowledge of NUCLEAR BOILER INSTRUMENTATION design feature(s) and/or interlocks which provide for the following: Reading of nuclear boiler parameters outside the control room Nuclear Boiler Inst.

Explanation: Answer C - With the Plant computer out of service, pressure readings must be obtained from installed instrumentation on either Div 1 or Div 2 Remote Shutdown panels. Tech Spec and IOI-11 requires cooldown at <100oF, cooldown from 1000-425 psig is approximately 92oF, from 425-100 psig is approximately 116oF, and 100-25 psig is approximately 71oF per ABB steam tables. Similar readings were obtained using RPV saturation curve of IOI-11.

A - incorrect - cooldown exceeded 100oF/hr between 1300 and 1400 B & D - incorrect - cooldown rate not exceeded Technical Reference(s): IOI-11 rev 19 Reference Attached: IOI-11 p 4 Proposed references to be provided to applicants during examination: Steam Tables Learning Objective (As available): OT-3035-05(LP)-A.10 & OT-3037-08-B Question Source: Bank # Brunswick 2008 Modified Bank #

New Question History: Previous NRC Exam: Brunswick 2008 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 - 2010 QUESTION RO 68 A high Drywell pressure scram occurred.

Eleven (11) minutes later, the following plant conditions exist:

Reactor pressure is 400 psig and lowering

Reactor water level is +12 inches and stable

Drywell pressure is 4 psig and slowly rising

Containment pressure is 6 psig and slowly rising

No operator actions have been performed Which one of the following describes the operating condition of RHR Loop A?

RHR Loop A is ____.

A. spraying Containment B. operating on minimum flow; Containment Spray mode can be manually initiated C. injecting into the rector vessel; Containment Spray mode can be manually initiated D. operating on minimum flow; Containment Sprays mode cannot be manually initiated

NRC Exam - 2010 QUESTION RO 68 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 226001 A1.07 Importance Rating 3.1 K&A: Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE controls including: System pressure RHR/LPCI: CTMT Spray Mode Explanation: Answer B - the high DW pressure signal will initiate RHR A. However, Rx pressure is too high for LPCI injection and Containment Spray mode will not auto initiate until containment pressure exceeds 8 psig.

A - incorrect - Containment Spray auto initiate logic is not complete C - incorrect - LPCI injection valve opens at 530 psig but system injection doesnt start until ~280 psig due to the discharge pressure of RHR pumps D - incorrect - RHR Containment Spray mode can be manually initiated when drywell pressure is above 1.68 psig.

Technical Reference(s): ARI-H13-P601-020 rev 15, SOI- Reference Attached: ARI-H13-P601-020 p E12 rev 46, & SDM-E12 rev 9 11, SOI-E12 p 9, & SDM-E12 p 97 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-E12-G.1 & G.2 Question Source: Bank # Perry 2002 Modified Bank #

New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 69 Refueling operations are in progress and the Inclined Fuel Transfer System (IFTS) is in operation.

The IFTS Carriage Assembly has just been raised to the RAISE FILL/DRAIN STOP position in the Fuel Handling Building.

The Bottom Valve and Drain Valve have closed.

Which one of the following describes what the Control Room operator will observe on H13-P601 meter, G43-R022A P601, POOL LEVEL A-SEPARATION?

The Upper Containment Pool water level will initially ____.

A. lower when the IFTS Transfer Tube is filled with water; water level must be manually restored with makeup water from the Condensate Transfer and Storage System B. lower when the IFTS Transfer Tube is filled with water; water level is restored when water from the FPCC Surge Tanks are subsequently pumped back to the Upper Containment Pool C. rise due to the displacement of water by the IFTS Carriage Assembly; water level is restored when the IFTS Carriage Assembly is subsequently lowered to the Fuel Handling Building D. rise due to the displacement of water by the IFTS Carriage Assembly; water level is automatically restored via an automatic drain valve to the Fuel Storage Pool in the Fuel Handling Building

NRC Exam - 2010 QUESTION RO 69 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 233000 A4.04 Importance Rating 2.9 K&A: Ability to manually operate and/or monitor in the control room: Pool level Fuel Pool Cooling/Cleanup Explanation: Answer B - First, upper pool level lowers due to IFTS tube filling, then level is restored when the FTT drain pumps pump water into the FPCC surge tanks and water is returned to the upper pools.

A - incorrect - The Upper Containment Pool level is restored via the Fuel Transfer Tube Drain Tank Pump to the FPCC Surge Tanks and then pumped back to the Upper Containment Pool C & D - incorrect - Upper containment pool level will initially lower as the transfer tube is filled (until the FPCC Upper Pool return can restore pool level)

Technical Reference(s): SDM-G41 rev 5 Reference Attached: SDM-G41 pp.5 & 50 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-G41-B.1, B.2.c, & B.2.d Question Source: Bank # Perry 2002 Modified Bank #

New Question History: Previous NRC Exam: Perry 2002 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 - 2010 QUESTION RO 70 Select the statement that describes how the Low-Low Set (LLS) function of the Safety Relief Valves (SRVs) automatically operates following a main steam isolation from high power, to maintain RPV Pressure?

A. 1B21-F051C opens at 1103 psig arming LLS which opens five LLS SRVs at 1113 psig.

1B21F051C cycles between 1073 psig and 936 psig.

B. 1B21-F051C and 1B21-F051D open at 1103 psig arming LLS.

Four LLS SRVs cycle between 1113 psig and 936 psig.

1B21F051C and 1B21-F051D cycle between 1073 psig and 926 psig.

C. 1B21-F051D opens at 1103 psig arming LLS which opens 1B21-F051C.

Four LLS SRVs cycle between 1113 psig and 946 psig.

1B21-F051C and 1B21-F051D cycle between 1033 psig and 976 psig.

D. 1B21-F051D opens at 1103 psig arming LLS which opens 1B21-F051C.

Four LLS SRVs cycle between 1113 psig and 946 psig.

1B21-F051C closes at 936 psig.

1B21-F051D cycles between 1033 psig and 926 psig.

NRC Exam - 2010 QUESTION RO 70 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 239001 K5.09 Importance Rating 3.4 K&A: Knowledge of the operational implications of the following concepts as they apply to MAIN AND REHEAT STEAM SYSTEM: Decay heat removal Main and Reheat Steam Explanation: Answer D - 1B21F051D opens at 1103 psig arming LLS, which opens 1B21F051C and F047F, F051A, F051B, F051G. SRVs 47F, 51A, 51B, and 51G cycle between 946 and 1113 psig. F051C cycles between 936 and 1073 psig and F051D cycles between 926 and 1033 psig.

A - incorrect - F051C does not arm LLS logic B - incorrect - cycle pressures for all LLS SRVs is not correct C - incorrect - cycle pressure for F051C and D is not correct Technical Reference(s): ONI-B21-1 rev 9, SOI-B21 Reference Attached: ONI-B21-1 p 11, SOI-rev 16, SDM-B21 rev 10 B21 p 4, SDM-B21 pp. 21 & 29 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21/N11-E.1 and E.2 Question Source: Bank # Perry 2007-1 Modified Bank #

New Question History: Previous NRC Exam: Perry 2007-1 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 - 2010 QUESTION RO 71 The following conditions exist:

A plant startup is in progress

The main turbine is reset If Main Turbine speed rises from 100% to 107% of rated, which of the following describes the response of the Main Turbine Control Valves (CVs) and Intercept Valves (IVs)?

A. The CVs and IVs start to throttle closed together B. The IVs throttle closed and the CVs remain open C. The IVs throttle closed first followed by the CVs D. The CVs throttle closed first followed by the IVs

NRC Exam - 2010 QUESTION RO 71 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 241000 A1.13 Importance Rating 2.7 K&A: Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including: Main turbine speed Reactor/Turbine Pressure Regulator Explanation: Answer D - The CVs will throttle closed first and will be fully closed at 105.5% of rated speed. Then the master IVs will throttle closed and all CVs will be fully closed at 107.5% of rated speed.

A - incorrect - the CVs AND IVs will not throttle together B - incorrect - the IVs will NOT throttle closed first - the CVs will throttle closed first and will be fully closed at 105.5% of rated speed, then the master IVs will throttle closed AND all IVs will be fully closed at 107.5% of rated speed C - incorrect - the CVs will NOT remain open - the CVs will throttle closed first and will be fully closed at 105.5% of rated speed, then the master IVs will throttle closed and all IVs will be fully closed at 107.5% of rated speed Technical Reference(s): SDM-N32/C85 rev 6 Reference Attached: SDM-N32/C85 pp. 34-35 & 115 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N32/C85-G Question Source: Bank # Nine Mile Point 2002 Modified Bank #

New Question History: Previous NRC Exam: Nine Mile Point 2002 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 - 2010 QUESTION RO 72 The following conditions exist:

The plant is operating at approximately 63% power

RFPT B on 3 element (3E) control in AUTO

RFPT B is operating at the suction flow limit of 23.1 Kgpm per SOI-C34, Feedwater Control System

The Motor Feed pump is shutdown to Casing Warmup

RFPT A is shutdown on the turning gear The plant then entered ONI-P52, Loss Of Service And/Or Instrument Air, due to an air leak which resulted in the following:

The source of the air leak was isolated from the Instrument Air header by closing 1P52F591, Instrument Air To Heater Bay Bldg Isol

Instrument Air pressure in the Heater Bay is 50 psig and dropping steadily Which one of the following describes the required operator response to the above conditions?

A. Take RECIRC FLOW CONTROLLER MINIMUM OUTPUT, 1N27-K162B_OP1 controller to 0% to close the RFP B RECIRC CONTROL VALVE and maintain vessel level 192 to 200 inches.

B. Start the Motor Feed pump per SOI-C34, Feedwater Control System, and maintain vessel level 192 to 200 inches.

C. Scram the Reactor due to lowering vessel level and enter ONI-C71-1, Reactor Scram.

D. No action required RFPT B will recover vessel level in automatic.

NRC Exam - 2010 QUESTION RO 72 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 259001 A2.05 Importance Rating 3 K&A: Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of applicable plant air systems Reactor Feedwater Explanation: Answer C - the RFPT recirc valve fails open on a loss of air. Since the RFPT is operating at the suction limit to maintain RPV level, the > 7000 gpm going to recirc will cause RPV level to lower.

Vessel level will continue to lower until power is reduced to the lower feed capacity. Scram prior to auto scram at level 3. Lowering power would not be an option since it would require a R/R pump downshift.

A - incorrect - this will have no effect as the RFPT recirc valve fails open on a loss of air B - incorrect - motor feed pump will start but control valves will not open to provide flow to vessel due to the loss of air D - incorrect - RFPT B will respond in auto but it is already operating at 95%, will add recirc flow of 7000+ gpm, vessel level will lower Technical Reference(s): ONI-P52 rev 14, ARI-H13-P680- Reference Attached: ONI-P52 p 30, ARI-H13-003 rev 11, & SOI-C34 rev 26 P680-003 p 103, & SOI-C34 p 4 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N27-K.4 Question Source: Bank # RQL-16886 Modified Bank #

New Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 73 Which of the following will cause Offgas After Filter flow to rise?

A. Moisture in adsorber trains B. Excessive steam flow rate to SJAEs C. Low level in condenser seal troughs D. Prefilter Line Drain Loop Seal Level Low

NRC Exam - 2010 QUESTION RO 73 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 271000 A4.02 Importance Rating 2.9 K&A: Ability to manually operate and/or monitor in the control room: System flows Offgas Explanation: Answer C - a low level in the condenser seal troughs can cause increased air in-leakage into the condensers. Increased air leakage will result in higher Offgas system flow.

A - incorrect - moisture in the adsorber trains can actually lower after filter flow by blocking the flowpath B - incorrect - excessive steam flow to the SJAE can cause high cooler condenser temperatures, but not a change in Offgas flow D - incorrect - a prefilter line drain loop seal low level can cause a spread of contamination, but not an increase in after filter flow Technical Reference(s): ARI-H13-P845-001 rev 9 & ARI- Reference Attached: ARI-H13-P845-001 H13-P680-002 rev 8 p 107 & ARI-H13-P680-002 p 3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-N62-K.2 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 x 55.43 Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION RO 74 The following conditions exist:

The plant is shutdown for a refueling outage

M25/26 Control Room Ventilation train A is running in NORMAL mode

M25/26 Control Room Ventilation train B is in STANDBY A loss of input from Control Room Atmosphere Gas Module D17-K776 to the M25/26 logic has been experienced due to an electrical switching error on buses K1A and K1N.

What is the current status of the Control Room HVAC system?

Train A Train B A. running in NORMAL mode remains in STANDBY B. running in NORMAL mode running in EMERGENCY RECIRC mode C. running in EMERGENCY RECIRC remains in STANDBY mode D. running in EMERGENCY RECIRC running in EMERGENCY RECIRC mode mode

NRC Exam - 2010 QUESTION RO 74 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 290003 K6.01 Importance Rating 2.7 K&A: Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROOM HVAC: Electrical power Control Room HVAC Explanation: Answer D - Both trains initiate in ER on loss of radiation monitor signal.

A - incorrect - plausible - loss of power will not have any affect B&C - incorrect - both trains initiate in ER Technical Reference(s): SOI-M25/26 Rev 18 & dwg Reference Attached: SOI-M25/26 p 5 & dwg 208-117 sh 11 rev S 208-117 sh 11 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-M25/26 F & G Question Source: Bank # Perry 2009 Modified Bank #

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

NRC Exam - 2010 QUESTION RO 75 The following conditions exist:

The plant is in Mode 4

Reactor cooldown is in progress

SVI-B21-T1176, RCS Heatup And Cooldown Surveillance is in progress

Current RCS temperature is 69oF What is the consequence of cooling down the Reactor Coolant System below 70°F?

The probability of ____ rises.

Reference Provided: SVI-B21-T1176 partial - modified A. brittle fracture B. ductile failure C. erosion corrosion D. stress corrosion cracking

NRC Exam - 2010 QUESTION RO 75 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 290002 2.2.12 Importance Rating 3.7 K&A: Knowledge of surveillance procedures.

Reactor Vessel Internals Explanation: Answer A - operating to the left of the heatup/cooldown curve increases the probability of brittle fracture.

B & D - incorrect - A high RCS temperature increases the probability of these occurring, not a low RCS temperature.

C - incorrect - The probability of this occurring is maximum at 270°F, not below 70°F Technical Reference(s): SVI-B21-T1176 rev 12, Tech Reference Attached: SVI-B21-T1176 pp. 9 &

Spec 3.4.11 Bases rev 8, & GFE(LP) Chapt. 10 rev 4 18, Tech Spec 3.4.11 Bases p B 3.4-56, &

GFE(LP) Chapt. 10 p 23 Proposed references to be provided to applicants during examination: SVI-B21-T1176 partial - modified Learning Objective (As available): x Question Source: Bank # Perry 2003 Modified Bank #

New Question History: Previous NRC Exam: Perry 2003 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 - 2010 QUESTION SRO 1 A staff Senior Reactor Operator had an Active and Valid license on December 31, 2009.

He then stood the following 12-hour shifts:

Jan 12, 2010 Shift Engineer

Jan 31, 2010 ATC

Feb 21, 2010 ATC

Feb 22, 2010 BOP

Feb 23, 2010 ATC

Mar 23, 2010 Field Supervisor

Apr 3, 2010 Unit Supervisor No additional shifts were stood as of April 30, 2010.

As of May 1, 2010 the staff SRO ____.

A. maintained proficiency as SRO B. maintained proficiency as RO only C. maintained proficiency as Shift Engineer D. did not maintain proficiency as a licensed operator

NRC Exam - 2010 QUESTION SRO 1 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.1 Importance Rating 4.2 K&A: Knowledge of conduct of operations requirements.

Generic Explanation: Answer D - The SRO is required to stand 5 12-hour watches per calendar quarter in a SRO position to maintain proficiency. Alternatively, he may stand 1 SRO watch and the remaining 4 may be in a Tech Spec required RO or SRO position.

A - incorrect - plausible if candidate uses quarter rather than calendar quarter B - incorrect - the Field Supervisor is not a required TS position - only 4 RO watches can be credited C - incorrect - the Shift Engineer must also maintain an Active RO or SRO license per PYBP-POS-1-5 Technical Reference(s): TMA-4206 rev 12 & Reference Attached: TMA-4206 p 17 &

PYBP-POS-1-5 rev 3 PYBP-POS-1-5 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-2600-01-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 55.43b.1 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 2 Refuel outage is in progress with the following conditions:

Core alterations are in progress

Div 2 electrical outage in progress

RHR A loop is operating in Shutdown Cooling Mode

Reactor cavity water level is above the top of the weir wall What condition below will require the Unit Supervisor to suspend movement of irradiated fuel assemblies within the RPV immediately?

A. Cavity water level is at 2210 above the fuel B. Refuel position one-rod-out interlock is inoperable C. RHR A is shutdown to replace breaker control power fuses D. Control Rod 30-31 position indication channel 2 is inoperable

NRC Exam - 2010 QUESTION SRO 2 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.1.36 Importance Rating 4.1 K&A: Knowledge of procedures and limitations involved in core alterations.

Generic Explanation: Answer A - per TS 3.9.6 or 3.9.7, RPV water level needs to be 229 above the RPV flange. This is much lower.

B - incorrect - this is TS 3.9.2 and action is to suspend control rod withdrawal C - incorrect - TS 3.9.8 allows RHR SDC to be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period D - incorrect - this is TS 3.9.4 - but per TS bases, channel 1 is the required channel Technical Reference(s): TS 3.9.6, 3.9.7, 3.9.2, 3.9.8, Reference Attached: TS 3.9.6 p 3.9-8, 3.9.7 p 3.9.4, & TS Bases 3.9.4 rev 0 3.9-9, 3.9.2 p 3.9-2, 3.9.8 p 3.9-10, 3.9.4 p 3.9-5, & TS Bases 3.9.4 p B 3.9-13 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-13-A, B, & C Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 3 The plant is operating at full power.

A problem develops with the Residual Heat Removal A pump and the Unit Supervisor declares RHR A INOPERABLE at 0500 on June 1

On June 3, a fire occurs in breaker EH1208 for RHR B pump and RHR B is declared INOPERABLE at 1700

RHR A is returned to OPERABLE status three hours later at 2000 If RHR B cannot be returned to OPERABLE status, what is the latest time that the plant must be in MODE 3?

Reference provided: Tech Spec 3.5.1 A. June 8, 0500 B. June 9, 0500 C. June 9, 1700 D. June 10, 1700

NRC Exam - 2010 QUESTION SRO 3 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.23 Importance Rating 4.6 K&A: Ability to track Technical Specifications limiting conditions for operations Generic Explanation: Answer C - per TS 1.3, this qualifies for a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time extension - the plant is required to be in Mode 3 by June 9, 1700 A - incorrect - this is the original completion time for RHR A inop B - incorrect - this is the time to enter TS 3.5.1 condition D - not Mode 3 D - incorrect - this is the time to enter Mode 4 Technical Reference(s): TS 3.5.1 and TS 1.3 Reference Attached: TS 3.5.1 pp 3.5-13.5-3 and TS 1.3 pp. 1.0-11&12 Proposed references to be provided to applicants during examination: Tech Spec 3.5.1 Learning Objective (As available): OT-3037-02-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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 4 The following plant conditions exist:

Reactor Mode Switch in SHUTDOWN

Reactor Coolant Temperature is 250°F

Unit 2 Div. 1 battery (2R42-S002) is INOPERABLE

Unit 1 Div.1 battery (1R42-S002) is on float

RCIC System is in Secured Status While performing weekly battery voltage surveillance, Electrical Maintenance discovered that the Unit 1 Division 1 battery (1R42-S002) terminal voltage to be 105 VDC.

As the Unit Supervisor, select the statement that describes the actions required.

Reference provided: Technical Specifications 3.8.4 & 3.8.5 A. Restore the Unit 1, Div. 1 battery to OPERABLE within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B. Restore either Div. 1 battery to OPERABLE within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. Restore both Div.1 batteries to OPERABLE within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in Cold Shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D. Declare affected required feature(s) INOPERABLE, immediately

NRC Exam - 2010 QUESTION SRO 4 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.2.36 Importance Rating 4.2 K&A: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Generic Explanation: Answer B - with RCS temperature > 250 deg. F, TS 3.8.4 is applicable. LCO Bases states either the Unit 1 or Unit 2 battery.

A - incorrect - while this cycles the pump breaker, this is not the minimum requirement B - incorrect - this also cycles the pump breaker, but is not the minimum requirement C - incorrect - this is performed if keep-fill is lost or voiding is suspected Technical Reference(s): Tech Spec 3.8.4 & 3.8.5 and TS Reference Attached: Tech Spec 3.8.4 & 3.8.5 Bases 3.8.4 rev 1 and TS Bases 3.8.4 p B 3.8-53 Proposed references to be provided to applicants during examination: Technical Specifications 3.8.4 &

3.8.5 Learning Objective (As available): OT-3037-12-C Question Source: Bank # Perry 2009 Audit 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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 5 The following plant conditions exist:

A plant transient results in fuel damage

The Shift Manager declared a Site Area Emergency 18 minutes ago

It is necessary for an operator to enter the Containment to terminate the offsite radioactive release

Radiation Protection estimates the expected exposure for this task is 3 Rem

The operator volunteering for the task has 500 mrem TEDE for the year As the Shift Manager acting as the Emergency Coordinator, describe the occupational dose limit(s) that will be exceeded, including the required approval(s) for this task?

A. Only the Site Administrative Dose Limit will be exceeded; only your approval is required.

B. Only the Site Administrative Dose Limit will be exceeded; your approval and the Radiation Protection Managers approval are required.

C. The Site Administrative Dose Limit and the Federal Dose Limit will be exceeded; only your approval is required.

D. The Site Administrative Dose Limit and the Federal Dose Limit will be exceeded; your approval and the Radiation Protection Managers approval are required.

NRC Exam - 2010 QUESTION SRO 5 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.4 Importance Rating 3.7 K&A: Knowledge of radiation exposure limits under normal or emergency conditions.

Generic Explanation: Answer A - based on the 3 rem exposure estimate and the 500 mr dose received to date, it is not expected to exceed federal Dose limits, but it does exceed Perry admin dose limits of 1000 mr.

since the TSC has not been activated, only the Shift Managers approval is necessary for the increased exposure.

B & D - incorrect - RP managers approval is not required C & D - incorrect - the federal dose limits will not be exceeded Technical Reference(s): HPI-B-003 rev 26 & NOP-OP- Reference Attached: HPI-B-003 pp. 4-5 &

4201 rev 0 NOP-OP-4201 p 18 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-03-I 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 55.43b.4 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 6 The plant is operating at rated power.

At 0700 on 3/1/2010, while reviewing the Tech Spec Rounds, the Unit Supervisor discovers the following:

The last time that a Channel Check was performed on Control Room Airborne radiation monitor D17-K776 was 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ago Per Technical Specifications, D17-K776 is __(1)__. The surveillance is required to be performed no later than __(2)__.

Reference provided: Tech Spec 3.3.7.1

__(1)__ __(2__

A. INOPERABLE 1300 on 3/2/2010 B. OPERABLE 1300 on 3/2/2010 C. INOPERABLE 0700 on 3/2/2010 D. OPERABLE 0700 on 3/2/2010

NRC Exam - 2010 QUESTION SRO 6 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.3.15 Importance Rating 3.1 K&A: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Generic Explanation: Answer D - TS SR 3.0.3 allows up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to complete a missed surveillance. TS SR 3.0.2 - 25% extension is not allowed for a missed surveillance.

A & B - incorrect -TS SR 3.0.3 allows up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to complete a missed surveillance - 3.0.2 - 25%

extension is not allowed for a missed surveillance A & C - incorrect -TS SR 3.0.3 states if a surveillance is missed then immediate entry into the LCO is not required Technical Reference(s): Tech Spec 3.0.3, Tech Spec Reference Attached: Tech Spec 3.0.3 p 3.0-4, Bases 3.0.3 rev 4, and Tech Spec 3.3.7.1 Tech Spec Bases 3.0.3 pp. B 3.0-12 to 14, and Tech Spec 3.3.7.1 p 3.3-70 to 73 Proposed references to be provided to applicants during examination: Tech Spec 3.3.7.1 Learning Objective (As available): OT-3037-04-J & K Question Source: Bank # Grand Gulf 2009 Modified Bank #

New Question History: Previous NRC Exam: Grand Gulf 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.4 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 7 The following conditions exist:

An Alert has been declared

The TSC is not yet activated Which of the below listed duties are allowed to be delegated by the Shift Manager?

1. Provide information and assistance to the Public Information Organization.

2. Determine the emergency classification including reclassification or termination.

3. Recommend protective actions for the general public to State and local County Officials.

4. Coordinate and direct the actions necessary to terminate or mitigate the effects of the emergency.

5. Request mobilization of the Corporate Planning Assistance Center (CPAC) in accordance with NOBP-LP-5001

6. Provide an interface with FirstEnergy Corporation organizational management and senior levels of outside organizations.

A. 1,2,3, B. 2,3,4 C. 3,4,5 D. 4,5,6

NRC Exam - 2010 QUESTION SRO 7 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 2.4.40 Importance Rating 4.5 K&A: Knowledge of SRO responsibilities in emergency plan implementation.

Generic Explanation: Answer D - these duties are allowed to be delegated by the Emergency Coordinator who, in this case, is the Shift Manager A & B - incorrect - Classification of the event can not be delegated B & C - incorrect - Protective Action Recommendations can not be delegated Technical Reference(s): EPI-A2 rev 14 Reference Attached: EPI-A2 pp 3-5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): EPL-0801-01-3 Question Source: Bank #

Modified Bank #

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

NRC Exam - 2010 QUESTION SRO 8 The following conditions exist:

The plant is shutdown for a refuel outage

Core Alterations are in progress

High Pressure Core Spray Pump is out of service for a motor change out

Division 1 ECCS pumps are out of service for LLRT testing

Buses EH11, EH12, & EH13 are aligned to LH-1-A

The Perry Transmission Yard East Bus (Bus # 1) is de-energized for maintenance A lockout occurs on Unit 1 Startup transformer Ten (10) minutes after the Unit 1 transformer lockout the following conditions exist:

Division 1 Diesel Generator is operating supplying Bus EH11

Division 2 Diesel Generator failed to start

Division 3 Diesel Generator is operating supplying Bus EH13 As the Unit Supervisor, which abnormal procedure will you enter to mitigate the consequences of the conditions above?

A. ONI-R10, LOSS OF AC POWER - SBO B. ONI-R10, LOSS OF AC POWER - TLAC C. ONI-R10, LOSS OF AC POWER - LOOP D. ONI-R22, Loss OF AN ESSENTIAL AND/OR A STUB 4.16KV BUS

NRC Exam - 2010 QUESTION SRO 8 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295003 AA2.05 Importance Rating 4.2 K&A: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Whether a partial or complete loss of A.C. power has occurred Partial or Complete Loss of AC / 6 Explanation: Answer B - With HPCS pump and Div 1 ECCS pumps out of service, having the Div 1 and Div 3 DGs start is irrelevant. The plant is still in a TLAC not an SBO.

A - incorrect - SBO would be entered if HPCS pump was available C - incorrect - LOOP would be entered if either Bus EH11 or EH12 with associated ECCS pumps available D - incorrect - ONI-R22-1 directs entry into ONI-R10 (and exit of ONI-R22-1)

Technical Reference(s): ONI-R10 rev 9 & ONI-R22-1 Reference Attached: ONI-R10 p 3 & ONI-rev 8 R22-1 p 5 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-18 (LP) A-4 Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 9 Given the following conditions:

D-1-A Voltage 125 VDC

D-1-B Voltage 100 VDC

ED-1-A Voltage 100 VDC

ED-1-B Voltage 0 VDC

ED-1-C Voltage 90 VDC

Reactor Coolant Temperature 180F

All control rods are inserted Entry into Off-normal instruction(s) __(1)__ is(are) required and, the Emergency Plan classification will be __(2)__.

Reference Provided: EPI-A1 Attachments 1 and 2 (1) (2)

A. ONI-R42-2, LOSS OF DC BUS ED-1-B and ES-1 ONI-R42-3, LOSS OF DC BUS ED-1-C B. ONI-R42-2, LOSS OF DC BUS ED-1-B and EU-1 ONI-R42-3, LOSS OF DC BUS ED-1-C C. ONI-R42-2, LOSS OF DC BUS ED-1-B ES-1 D. ONI-R42-2, LOSS OF DC BUS ED-1-B EU-1

NRC Exam - 2010 QUESTION SRO 9 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295004 2.4.11 Importance Rating 4.2 K&A: Knowledge of abnormal condition procedures.

Partial or Total Loss of DC Pwr / 6 Explanation: Answer D - the Entry Condition for all R42 ONIs is Bus Voltage zero. The other entry conditions in the ONI describe the effects of the loss of the DC bus. Therefore, only ONI-R42-2 should be entered. Additionally, Since the Plant is in Mode 4, EU-1 is the correct E-plan classification.

A & B - incorrect - voltage on ED-1-C is greater than the ONI-R42-3 Entry Condition voltage A & C - incorrect - ES-1 would be correct if plant was in Mode 1, 2, or 3.

Technical Reference(s): ONI-R42-2 rev 6 & EPI-A001 Reference Attached: ONI-R42-2 p 3 & EPI-rev 21 A001 p 14 Proposed references to be provided to applicants during examination: EPI-A1 Attachments 1 and 2 Learning Objective (As available): OT-3035-05(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 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 10 Surveillance SVI-C71-T0046, TURBINE STOP VALVE CLOSURE AND TURBINE CONTROL VALVE FAST CLOSURE CHANNEL FUNCTIONAL FOR 1C71-N006A, B, C, D, E, F, G, H AND 1C71-N005A, B, C, D commenced at 12:00 on June 2.

The Unit Supervisor signed the SVI for channel INOPERABILITY at 14:00 on June 2 At 16:00 on June 2, Turbine Stop Valve #3 failed to initiate a half scram signal to RPS when tested.

The Technical Specification Required Actions are to place channel in trip by __(1)__ on June 3.

The bases of the Turbine Stop Valve closure scram is to ensure the __(2)__.

Reference Provided: Technical Specification 3.3.1.1 - partial - modified

__(1)__ __(2)__

A. 04:00 fuel peak cladding temperature remains below the limits of 10CRF50.46 B. 08:00 fuel peak cladding temperature remains below the limits of 10CRF50.46 C. 04:00 Minimum Critical Power Ratio (MCPR) Safety Limit is not exceeded D. 08:00 Minimum Critical Power Ratio (MCPR) Safety Limit is not exceeded

NRC Exam - 2010 QUESTION SRO 10 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295005 AA2.03 Importance Rating 3.1 K&A: Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Turbine valve position Main Turbine Generator Trip / 3 Explanation: Answer C - On a turbine trip with Rx power > 38%, the Turbine Stop valves generate a Rx scram signal when they go closed. Testing of the Stop Valves includes cycling the TSV to ensure a 1/2 scram signal is generated when each valve is < 93% open. Time of Discovery of an INOP condition starts the clock for Required Actions. In this case, Time of Discovery was 16:00. RA is to place channel in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The bases of the TSV closure scram is to prevent exceeding the MCPR safety limit.

A & B - incorrect - the bases of the TSV scram is to not exceed MCPR B & D - incorrect - the Required Actions must be completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of time of discovery - this answer is plausible if candidate believes that the 6-hour note for surveillance testing applies to completing the Required Actions Technical Reference(s): Tech Spec 3.3-1, Tech Spec Reference Attached: Tech Spec 3.3-1 1.3, & Tech Spec Bases B 3.3.1.1 rev 0 pp. 3.3-1&3, Tech Spec 1.3 p 1.0-11, & Tech Spec Bases B 3.3.1.1 pp. B 3.3-14&16 Proposed references to be provided to applicants during examination: Technical Specification 3.3.1.1 -

partial - modified Learning Objective (As available): OT-3037-02-E 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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 11 Which one of the following identifies a condition that would require declaration of an ALERT only?

Reference Provided: EPI-A1 Attachments 1 and 2 A.

RPV level is lowering slowly

Fifty Control Rods are withdrawn at various positions

Reactor Power is 10%

Mode Switch is in SHUTDOWN B.

RPV level at 175 inches and lowering

Control Rods 30-31, 18-55, 14-15, 46-15, 46-31, & 46-47 are at position 48

Nuclear Instruments are fully inserted

IRMs are indicating on range 4 C.

Mode Switch is placed in SHUTDOWN

Control Rod 46-15 is at position 48

Nuclear Instruments are fully inserted

Power is indicating middle of the source range D.

RPV level at 179 inches and lowering

Mode Switch is placed in SHUTDOWN

All Control Rods indicate 00

Nuclear Instruments are fully inserted

Power is indicating middle of the source range

NRC Exam - 2010 QUESTION SRO 11 Level: RO SRO Tier # 3 Examination Outline Cross-Reference Group #

K/A# 295006 2.4.41 Importance Rating 4.6 K&A: Knowledge of the emergency action level thresholds and classifications.

SCRAM / 1 Explanation: Answer B - this is a CA1 - failure to auto scram and power below 4%

A - incorrect - this should be a CS1 C - incorrect - no auto scram setpoint exceeded - Rx is S/D under all conditions without boron D - incorrect - no auto scram setpoint exceeded - no EAL entry Technical Reference(s): EPI-A1 rev 21 Reference Attached: EPI-A1 pp. 14 & 23 Proposed references to be provided to applicants during examination: EPI-A1 Attachments 1 and 2 Learning Objective (As available): EPL-0804-01-4 Question Source: Bank # Grand Gulf 2004 Modified Bank #

New Question History: Previous NRC Exam: Grand Gulf 2004 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 12 Refueling activities are in progress.

The following conditions exist on the Refuel Floor:

A fuel bundle just arrived from the Fuel Handling Building with the IFTS Upender vertical

A fuel bundle is in transit from the RPV to the fuel storage racks via F15 Bridge

An unexplained drop in upper pool level occurs Which of the following actions is required concerning the status of the two bundles?

A. Incline the IFTS Upender and return the fuel bundle on the F15 Bridge back to any open vessel location.

B. Incline the IFTS Upender and continue the fuel movement with the F15 Bridge to the fuel storage racks.

C. Continue fuel movement with the F15 Bridge to the fuel storage racks, then, transfer the fuel bundle in IFTS to the fuel storage racks.

D. Transfer the fuel bundle in IFTS down to the Fuel Handling Building and return the fuel bundle on the F15 Bridge back to the vessel location from which it was removed.

NRC Exam - 2010 QUESTION SRO 12 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295023 AA2.02 Importance Rating 3.7 K&A: Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: Fuel pool level Refueling Acc / 8 Explanation: Answer B - > The stem contains entry conditions described for ONI-E12-2 entry. The ONI also describes Safe Conditions. FTI-D009 further restricts placing a bundle back into the Rx after it has been off-loaded.

A - incorrect - Moving the bundle back to the core is an unsafe act/condition C - incorrect - An IFTS transfer to the FHB is an unsafe act/condition D - incorrect - An IFTS transfer to the FHB is an unsafe act/condition - moving the bundle back to the core is an unsafe act/condition Technical Reference(s): ONI-E12-2 rev 25 & FTI-D009 Reference Attached: ONI-E12-2 pp. 3-7 &

rev 13 FTI-D009 p 9 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-11(LP)-A.2 & OT-3602-01-D.4 & E.2 Question Source: Bank # RQL-0748 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 55.43b.7 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 13 The plant was operating at 100% power when a LOCA occurred.

The following conditions now exist:

LPCI A and LPCS pumps are injecting

Adequate core cooling is being maintained

NO other ECCS pumps are available

Suppression Pool temperature is rising Based on the information from the attached SPDS screen print, select the statement below that describes the use of RHR A Loop for Suppression Pool cooling.

Reference Provided: SPDS screen print A. LPCI A must be diverted to Suppression Pool Cooling to ensure that Suppression Pool temperature is maintained below the Heat Capacity Limit, since LPCS can maintain adequate core cooling through spray cooling alone.

B. LPCI A may be diverted to Suppression Pool Cooling as long as LPCS is able to maintain RPV water level above -25 inches.

C. LPCI A must be diverted to Suppression Pool Cooling, irrespective of adequate core cooling, when neither Suppression Pool temperature nor Reactor pressure can not be maintained below the Heat Capacity Limit (HCL)

D. LPCI A may be diverted to Suppression Pool Cooling as long as LPCS is able to maintain RPV water level above -42.5 inches

NRC Exam - 2010 QUESTION SRO 13 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 1 K/A# 295026 2.1.19 Importance Rating 3.8 K&A: Ability to use plant computers to evaluate system or component status.

Suppression Pool High Water Temp. / 5 Explanation: Answer B - The Print shows the scram was successful and ED was performed. Injection sources needed for adequate core cooling are not allowed to be diverted to suppression pool cooling.

However, adequate core cooling is defined as maintaining RPV level > -25 with injection. So, if LPCS can maintain RPV level > -25 adequate core cooling is assured.

A & C - incorrect - no directions are given tin the EOPs to divert injection sources required for adequate core cooling to suppression pool cooling to avoid exceeding HCL.

D - incorrect - the - 42.5 is allowable if no injection to the RPV is available Technical Reference(s): EOP-Bases rev 0 & EOP-2 Reference Attached: EOP-Bases pp 33-35 &

Bases rev 0 EOP-2 Bases pp. 22-23 Proposed references to be provided to applicants during examination: SPDS screen print Learning Objective (As available): 3402-01-C.1, C.30, C.1.c Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 14 While the Unit Supervisor was reviewing Technical Specification Rounds he discovered the following:

Containment Average Air Temp 78oF

Suppression Pool validated level 17.2 feet

Half of the suppression pool temperature detectors on H13P868 and H13-P869 read 79oF

Half of the suppression pool temperature detectors on H13P868 and H13-P869 read 85oF Based on the readings from Tech Spec Rounds, select the correct statement below.

A. All Suppression Pool temperature instruments are OPERABLE B. The Suppression Pool temperature instruments reading 79oF are OPERABLE C. The Suppression Pool temperature instruments reading 85oF are OPERABLE D. None of the Suppression Pool temperature instruments are OPERABLE

NRC Exam - 2010 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 Explanation: Answer C - Eight of sixteen SP level instruments are uncovered and exposed to the containment atmosphere when SP level drops below 17.33. In order for the SP level instruments to perform their intended function (Operable) they must be submerged.

A - incorrect - the instruments not submerged are not operable B - incorrect - these instruments are reading containment air temperature and are not operable D - incorrect - the instruments still submerged are operable - plausible is candidate is confused about which sets of instruments are reading accurately Technical Reference(s): Tech Spec 1.0, Tech Spec Reference Attached: Tech Spec 1.0 p 1.0-5, Bases 3.3 rev 1, & Dwg 240-082 rev K Tech Spec Bases 3.3 p B 3.3-54 & Dwg 240-082 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-D23.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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 15 The following conditions exist:

Drywell pressure is 2.0 psig

Reactor pressure is 870 psig

Reactor is shutdown Given these conditions, EOP-2 Primary Containment Control contains steps to have the Unit Supervisor direct prevention of LPCS and LPCI injection if not needed for adequate core cooling.

What is the basis for this action?

A. Facilitate RPV level control.

B. Facilitate RPV pressure control.

C. Reduce thermal stress on the Reactor vessel caused by cold water injection.

D. Prevent the possibility of a reactor power excursion large enough to severely damage the core.

NRC Exam - 2010 QUESTION SRO 15 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295010 2.4.18 Importance Rating 4.0 K&A: Knowledge of the specific bases for EOPs.

High Drywell Pressure / 5 Explanation: Answer A - > Per the bases, low pressure systems are prevented so as not to complicate RPV water level control.

B - incorrect - plausible since injection of cold water can also affect RPV pressure, however, not per the Bases C - incorrect - controlling cooldown rates controls thermal stress - not per the Bases D - incorrect - this is true if the reactor is not shutdown - Rx s/d given in stem Technical Reference(s): EOP-2 Bases rev 0 Reference Attached: EOP-2 Bases pp. 38-39 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-01-C.44 & B.1 Question Source: Bank #

Modified Bank #

New Monticello 2002 Question History: Previous NRC Exam: Monticello 2002 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 16 The following conditions exist:

The plant is operating at rated power

A loss of drywell cooling occurs

Drywell average air temperature has reached 150°F and continues to slowly rise Per the Bases for Technical Specification 3.6.5.5, Drywell Air Temperature, if this condition is not corrected, then exceeding the ____ during a design basis LOCA can not be ensured.

A. limit for hydrogen concentration in the drywell B. drywell pressure suppression capability C. heat capacity of the suppression pool D. drywell design temperature limit

NRC Exam - 2010 QUESTION SRO 16 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295012 AA2.01 Importance Rating 3.9 K&A: Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell temperature.

High Drywell Temperature / 5 Explanation: Answer D - > The Tech Spec Bases for high DW temperature states the design drywell temperature of 330 degrees cannot be ensured following a design basis LOCA if drywell temperature is not maintained below 145 degrees.

A - incorrect - this is the containment HDOL limit B - incorrect - the DW pressure suppression capability is a function of DW bypass leakage C - incorrect - the Heat capacity limit is a function of RPV pressure and suppression pool temperature Technical Reference(s): Tech Spec Bases 3.6.5.5 rev 7 Reference Attached: Tech Spec Bases 3.6.5.5 p B 3.6-148 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3037-10-B Question Source: Bank # Grand Gulf 2009 Modified Bank #

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

NRC Exam - 2010 QUESTION SRO 17 The following plant conditions exist:

Reactor water level is -10 inches

Reactor pressure is 900 psig

Reactor power is 5%

Drywell pressure is + 1.1 psig

Drywell temperature is 130oF

Containment pressure is +1.5 psig.

Containment temperature is 100oF

Suppression Pool temperature is 116oF

Suppression Pool level is 24.8 feet

All appropriate EOPs have been entered As the Unit Supervisor, which one of the following actions should you direct, including the bases for this action?

Reference Provided: Technical Specification 3.3.1.1 A. Open the Main Turbine Bypass Valves and depressurize the reactor; the RPV cannot be permitted to remain at pressure if operation of SRVs may cause the SRV discharge line or associated components to fail.

B. Open the Main Turbine Bypass Valves and depressurize the reactor; if primary containment water level rises above the elevation of the SRV solenoids, the SRVs may no longer be operable.

C. Open the ADS valves and depressurize the reactor; the RPV cannot be permitted to remain at pressure if operation of SRVs may cause the SRV discharge line or associated components to fail.

D. Open the ADS valves and depressurize the reactor; if primary containment water level rises above the elevation of the SRV solenoids, the SRVs may no longer be operable.

NRC Exam - 2010 QUESTION SRO 17 Level: RO SRO Tier # 1 Examination Outline Cross-Reference Group # 2 K/A# 295029 2.1.32 Importance Rating 4 K&A: Ability to explain and apply system limits and precautions.

High Suppression Pool Wtr Lvl / 5 Explanation: Answer C - > The SRV Tail Pipe Level Limit is 24.5. Operation of SRVs with SP level above this could result in damage to the SRV discharge lines and direct pressurization of containment.

Since the reactor is not shutdown, Anticipate ED is not allowed. ED by SRV is required.

A & B - incorrect - using the turbine bypass valves to anticipate ED is not allowed when Rx not shutdown D - incorrect - the high SP level can affect the MSIV drain valves not the SRV solenoids Technical Reference(s): EOP-2 Bases rev 0 Reference Attached: EOP-2 Bases p 35 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-05-C Question Source: Bank # Perry 2003 Audit 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 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 18 The following conditions exist:

First startup after refueling outage in progress

The IRM Linearity Check was performed The table below shows the IRM linearity data after final adjustments by I&C.

IRM LINEARITY TABLE IRM Range 6 Range 6 Divided By Range 7 Sat / Unsat (40 to 75) 3.1623 +/- 3/125 A 65 20.5547 21 S/U E 58 18.3411 19 S/U C 49 15.4951 19 S/U G 55 17.3924 14 S/U B 45 14.2301 13 S/U F 65 20.5547 22 S/U D 50 15.8113 16 S/U H 51 16.1275 20 S/U Based on the above data, which of the following actions is required?

Reference provided: Technical Specification 3.3.1.1 - Partial (Modified)

A. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , insert a half scram.

B. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , either place one IRM channel in trip or insert a half scram.

C. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , place the plant in MODE 3.

D. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , either place one IRM channel in trip or insert a half scram.

NRC Exam - 2010 QUESTION SRO 18 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 215003 A2.05 Importance Rating 3.5 K&A: Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/system IRM Explanation: Answer D - IRM channels C, G, & H are out of spec. Channels G & H are on trip system B.

TS require 3 of 4 channels per trip system. TS 3.3.1.1 Condition A is entered since only Trip System B is below minimum number of channels. The channel is placed in trip IAW the Trip Instructions. Or the trip system is placed in trip (half scram inserted)

A - incorrect - This the action from Condition H which is not required at this time - incorrect use of the table B - incorrect - This is the action from Condition B - it is a misconception that both trip systems are affected; Trip System B still has three channels D - incorrect - This is the action from Condition C, which is not entered since no loss of function has occurred Technical Reference(s): IOI-1 rev 30 & Tech Spec Reference Attached: IOI-1 28-29 & Tech 3.3.1.1 Spec 3.3.1.1p 3.3-1 Proposed references to be provided to applicants during examination: Technical Specification 3.3.1.1 -

Partial (Modified)

Learning Objective (As available): OT-COMBINED-C51_IRM-J.1 Question Source: Bank # RQL-0934 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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 19 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 Technician Specification function of RCIC is to __(1)__.

RCIC is __(2)__.

Reference Provided: 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 either automatically or manually following a loss OPERABLE of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level D. operate either automatically or manually following a loss INOPERABLE of coolant flow from the feedwater system to provide adequate core cooling and control of RPV water level

NRC Exam - 2010 QUESTION SRO 19 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 217000 2.1.27 Importance Rating 4 K&A: Knowledge of system purpose and/or function.

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 and Tech Spec Bases 3.5.1 rev 0 B 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 #

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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 20 The following conditions exist:

The plant is at 100% power

I&C is performing surveillance SVI-B21-T0072-C, MAIN STEAM LINE LOW PRESSURE CHANNEL C CALIBRATION FOR 1B21-N076C

All other MSL Low Pressure instruments are OPERABLE

The technician reports to you that he is unable to adjust the 1B21-N076C instrument within the allowable value.

Based on this information, ____.

Reference Provided: Technical Specification 3.3.6.1 - partial - modified A. enter TS 3.3.6.1 Condition A and place channel in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. enter TS 3.3.6.1 Condition A and place channel in trip within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. no action is required at this time; the allowable value is not the required condition D. no action is required at this time; only 2 of the 4 channels are required to be operable

NRC Exam - 2010 QUESTION SRO 20 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 223002 A2.08 Importance Rating 3.1 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: Surveillance testing PCIS/Nuclear Steam Supply Shutoff Explanation: Answer B - per Tech Spec Bases, the appropriate condition and required action must be entered if any one of the 4 MSL low presure transmitters is INOP. Twenty-four hours is correct because this is for Function 1b A - incorrect - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is for Functions 2b, 5b, & 5d C - incorrect -- plausible if misinterprets Allowable Value for Leave As Is Zone D - incorrect - required channels are 2 per trip system, with 4 channels total in 2 trip systems 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 4, SVI-B21-T0072-C rev 5 48 & 54 and Tech Spec Bases p B 3.3-142 Proposed references to be provided to applicants during examination: Technical Specification 3.3.6.1 -

partial - modified Learning Objective (As available): OT-COMBINED-B21(NS4)-l Question Source: Bank #

Modified Bank #

New x Question History: Previous NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 21 The following conditions exist:

The reactor scrammed due to a small-break LOCA

The only available injection source is from the Condensate Transfer system

To maximize injection, Emergency Depressurization was initiated approximately 20 minutes ago and all ADS SRVs were verified open

The SRV OPEN annunciator just reset You have directed the Reactor Operator to verify the status of the ADS SRVs.

The ADS SRVs are reported to be ____.

closed based on stable SRV tailpipe temperatures. You would direct the panel A.

operators to open SRVs using their control switches.

closed based on SRV tailpipe temperatures slowly lowering. You would direct the B.

panel operators to use alternate methods of depressurizing the reactor vessel.

open based on SRV tailpipe temperatures of approximately 250°F and stable.

C.

Injection is occurring. You would direct operators to monitor reactor vessel level.

open based on SRV tailpipe temperatures of approximately 330°F and slowly rising D.

due to the lack of injection. You would direct operators to open additional SRVs and continue to monitor for injection.

NRC Exam - 2010 QUESTION SRO 21 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 239002 2.4.46 Importance Rating 4.2 K&A: Ability to verify that the alarms are consistent with the plant conditions.

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 250oF which corresponds to reactor pressure of ~25 psig.

A & B incorrect - SRVs are still open D - incorrect - tailpipe temperature of 330oF corresponds to normal reactor pressure Technical Reference(s): ARI-H13-P601-019 rev 11 & Reference Attached: ARI-H13-P601-019 p 17 ABB Steam Tables Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-COMBINED-B21_N11-H Question Source: Bank # Perry 2005 Modified Bank #

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

NRC Exam - 2010 QUESTION SRO 22 The following conditions exist:

The Mode Switch is in SHUTDOWN

All control rods are inserted

Reactor coolant temperature is 250°F

Buses EH11 and EH13 are aligned to the Preferred Source

Bus EH12 is aligned to the Alternate Preferred Source

Division 2 Diesel generator is out of service and tagged out

AEGT System A is in operation A grid disturbance then occurs resulting in the following:

Loss of power to the Unit 2 Startup Transformer

A non-valid Division 2 RHR-LOCA signal is generated No operator action have been performed.

The status of the A and B AEGT systems is __(1)__, requiring you to enter __(2)__

3.6.4.3 AEGTS LCO in accordance with OAI-1701, Tracking Of LCOs.

(1) (2)

A. AEGTS A is running a Potential or NO AEGTS B is running B. AEGTS A is running an Active AEGTS B is not running C. AEGTS A is not running a Potential or NO AEGTS B is running D. AEGTS A is not running an Active AEGTS B is not running

NRC Exam - 2010 QUESTION SRO 22 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 1 K/A# 261000 A2.07 Importance Rating 2.8 K&A: Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C.

electrical failure SGTS Explanation: Answer B - A loss of the Unit 2 S/U transformer results in the loss of Bus EH12 since Div. 2 DG is tagged out. Therefore, AEGT train (B) will not start on a LOCA signal. Since the LOCA signal is a Div 2 signal, it will not affect the Div 1 AEGT fan. Although the plant is shutdown, RCS temperature is

>200°F which is Mode 3. Therefore, TS 3.6.4.3 is applicable.

A - incorrect - AEGTS B train is Not running due to the loss of 480VAC power - this is an active LCO C - incorrect - AEGTS A train continues to run and AEGT B can not run - this is an active LCO D - incorrect - AEGTS A train continues to run Technical Reference(s): oai-1701 rev 11, Tech Spec Reference Attached: Tech Spec 3.6.4.3 p 3.6.4.3 and Tech Spec Bases B 3.6.4.3 rev 7 3.6.56 and Tech Spec Bases p B 3.6-119 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 55.43b.2 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 23 The following conditions exist:

The plant is operating at 100% power with a high load line

FMEOD (Fraction of Maximum Extended Operating Domain) at 0.985.

During a planned power reduction to 80% using the Recirc Flow Control Valves, the SCC Dispatcher requests Perry to hold power at 90%.

A short time later, the Reactor Operator reports that FMEOD is 0.992.

The Unit Supervisor is required to restore FMEOD to 0.990 by ____.

A. only inserting Cram Rods B. only inserting control rods per Reactor Engineering direction C. entering ONI-C51, Unplanned Change in Reactor Power or Reactivity and inserting Cram Rods D. entering ONI-C51, Unplanned Change in Reactor Power or Reactivity and inserting control rods per ONI direction

NRC Exam - 2010 QUESTION SRO 23 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 201005 2.1.32 Importance Rating 4.0 K&A: 2.1.32 Ability to explain and apply system limits and precautions.

RCIS Explanation: Answer B - Per IOI-3, P&L 2.2, if FMEOD is > 0.990, then insert control rods to restore FMEOD to <0.990 per Reactor Engineering direction.

A - incorrect - IOI-3, P&L 2.2 does not direct use of Cram Rods C & D - incorrect - per ONI-C51Entry Conditions and IOI-3, P&L 2.2, ONI-C51 is only entered when FMEOD is > 1.000 Technical Reference(s): IOI-3 rev 24 & ONI-C51 Chart Reference Attached: ONI-C51 p 5 & ONI-C51 rev I flowchart Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3035-08(LP)-B.4.B Question Source: Bank # Perry 2007-1 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 55.43b.1 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 24 The plant was operating at rated power, when a transient occurred, resulting in the following conditions:

All control rods are inserted

Drywell pressure is 3.0 psig and stable

Reactor pressure is 950 psig and stable

RPV water level is (zero) 0 inches and lowering slowly

No injection sources are currently available

RWCU Isolations have been bypassed per EOP-SPI 2.6, Bypass Of RWCU Isolations

RWCU blowdown rate is 60 gpm and stable

RWCU Recirculation and Blowdown Modes are being used to control RPV pressure

The Unit Supervisor entered EOP-4-3, Steam Cooling Based on these conditions, Emergency Depressurization should __(1)__ delayed. And, with regard to the RWCU system, the Unit Supervisor will direct the Reactor Operator to __(2)__.

__(1)__ __(2)__

A. be Secure Blowdown Mode and maintain Recirculation Mode B. be Maintain Blowdown Mode and secure Recirculation Mode C. not be Secure Blowdown Mode and maintain Recirculation Mode D. not be Maintain Blowdown Mode and secure Recirculation Mode

NRC Exam - 2010 QUESTION SRO 24 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 204000 A2.12 Importance Rating 2.8 K&A: Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Excessive drain flow rates RWCU Explanation: Answer A - Based on RPV level trend, Emergency Depressurization will be required prior to reaching -25 inches. RWCU is draining 60 gpm from vessel inventory. This is resulting in lowering RPV level. When RPV level drops below 0 with no injection sources operating, Steam Cooling is entered.

When in Steam Cooling, it is appropriate to secure RWCU from blowdown mode to preserve RPV inventory. With no injections sources available, ED should be delayed as long as possible.

B & D - incorrect - blowdown mode should be secured C & D - incorrect - with no injection sources available, ED should be delayed as long as possible Technical Reference(s): EOP-1 Bases rev 1 Reference Attached: EOP-1 Bases pp. 47, 55, & 76 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OT-3402-01-C-29 & 02-F, Question Source: Bank #

Modified Bank # Dresden 2009 New Question History: Previous NRC Exam: Dresden 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43b.5 x Comments: Level of Difficulty = x

NRC Exam - 2010 QUESTION SRO 25 The following conditions exist:

The reactor is shutdown

Control Rod Drive Hydraulic system is shutdown Which of the following conditions require system manipulations to be directly supervised by the Shift Manager or Unit Supervisor per OAI-0201, Operations General Instructions and Operating Practices?

A. Adjusting FPCC Upper Pool flow while in MODE 3 B. Placing a RWCU Filter/Demin in service while in MODE 4 C. Shifting NCC heat exchangers from A to C while in MODE 4 D. Placing RHR B loop in Suppression Pool Cooling with RHR A loop is in Shutdown Cooling while in MODE 3

NRC Exam - 2010 QUESTION SRO 25 Level: RO SRO Tier # 2 Examination Outline Cross-Reference Group # 2 K/A# 216000 2.1.9 Importance Rating 4.5 K&A: 2.1.9 Ability to direct personnel activities inside the control room.

Nuclear Boiler Inst.

Explanation: Answer D - With CRDH shutdown, reference leg purge is out of service. With reference leg purge out of service, accuracy of the RPV water level instruments is questioned. OAI-0201 requires compensatory measures when performing activities with a potential to drain the vessel with the reference leg purge is out of service while in Modes 1, 2, or 3. With RHR in SDC, valve manipulations on any part of the RHR system must be supervised by the SM or US.

A - incorrect - with plant not in Mode 5, adjusting upper pool flow does not affect RPV level B - incorrect - OAI-0201 specifies RWCU system manipulations in Mode 3 must be supervised - not Mode 4 C - incorrect - per Risk Management, if the plant was in Mode 5 and NCC was affecting Decay Heat removal, this would require supervisors oversight Technical Reference(s): OAI-0201 rev 20 Reference Attached: OAI-0201 pp. 13-14 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 55.43b.5 x Comments: Level of Difficulty = x

RO Question 18 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0016

Title:

Use Category:

Decay Heat Curve In-Field Reference Revision: Page:

9 1 of 7 DECAY HEAT CURVE Functional Location (1J11)

Plant Data Book Effective Date: 1-29-09 Preparer: Patrick Curran / 11-24-08 Date

RO Question 18 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0016

Title:

Use Category:

Decay Heat Curve In-Field Reference Revision: Page:

9 2 of 7

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 1 of 19 POOL HEATUP CURVES Functional Location (J11)

Plant Data Book Effective Date: 2-18-09 Preparer: A. Widmer / 1-18-09 Date

RO Question 18 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 2 of 19 1.0 PURPOSE Provide plant personnel the information with the time for either the reactor or fuel handling building pools to reach a specified temperature. Given the plant configuration, decay heat load, and location of the fuel from the previous operating cycle, the time it would take to reach bulk coolant saturation temperature with no Decay Heat Removal (or Spent Fuel Pool Cooling) systems in operation. In the case of the Fuel Handling Building pools, information is provided regarding time to reach 150oF as well as 212oF.

2.0 REFERENCES

2.1 Discretionary Calculation G41-038 Revision 10 Calculation G41-042 Revision 10 Plant Data Book PDB-A0016 PAP-1925 Shutdown Defense In Depth Assessment and Management NSAC/175L - EPRI Outage Risk Assessment and Management (ORAM)

Program Safety Assessment of BWR Risk During Shutdown Operations 2.2 Obligations None Commitments addressed in this document:

None

RO Question 18 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 3 of 19 3.0 DETAILS In the event that Decay Heat Removal system is lost, the following graphs can be utilized to determine the time it takes to get to a specified temperature. The graphs are decay heat vs. time. The decay heat value may be determined from Plant Data Book PDB-A0016 for a given plant configuration and day after shutdown. The decay heat value can be used herein to determine the applicable heat-up time given the initial water temperature is known.

4.0 SCOPE OF REVISION Rev. 10 1. Updated curves to reflect decay heat vs. time.

2. Revised format in accordance with NOP-SS-3007.

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 4 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Reactor Vessel @ Normal Water Level 3.5 Initial Temperature:

70 80 3.0 90 100 110 2.5 120 130 Time to Boil (hours) 140 2.0 1.5 1.0 0.5 0.0 85 80 75 70 65 60 55 50 45 40 35 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 5 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Reactor Vessel @ Normal Water Level 25 Initial Temperature:

70 80 20 90 100 110 120 Time to Boil (hours) 130 15 140 10 5

0 35 30 25 20 15 10 5 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 6 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level at 250" above TAF 3.5 Initial Temperature:

70 3.0 80 90 100 110 2.5 120 130 Time to Boil (hours) 140 2.0 1.5 1.0 0.5 0.0 85 80 75 70 65 60 55 50 45 40 35 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 7 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level at 250" above TAF Initial Temperature:

70 20 80 90 100 110 120 15 Time to Boil (hours) 130 140 10 5

0 35 30 25 20 15 10 5 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 8 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level at Reactor Flange 4.5 Initial Temperature:

70 4.0 80 90 100 3.5 110 120 3.0 Time to Boil (hours) 130 140 2.5 2.0 1.5 1.0 0.5 85 80 75 70 65 60 55 50 45 40 35 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 9 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level at Reactor Flange 30 Initial Temperature:

70 80 25 90 100 110 120 20 130 Time to Boil (hours) 140 15 10 5

0 35 30 25 20 15 10 5 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 10 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level 23' above Reactor Flange 18 Initial Temperature:

70 16 80 90 100 14 110 120 12 Time to Boil (hours) 130 140 10 8

6 4

2 85 80 75 70 65 60 55 50 45 40 35 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 11 of 19 TIME-TO-BOIL CURVES FOR REACTOR VESSEL Water Level 23' above Reactor Flange 125 Initial Temperature:

70 80 90 100 100 110 120 Time to Boil (hours) 130 75 140 50 25 0

35 30 25 20 15 10 5 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 12 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS TO 150°F Time Needed to Heat FHB Pools in Communication to 150 oF Volume 62485 ft3 20 Initial Temperature:

70 F 80 F 90 F 100 F 15 110 F 120 F Hours 10 5

0 80.00 70.00 60.00 50.00 40.00 30.00 20.00 Decay Heat (mBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 13 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS TO 150OF Time to Heat Fuel Pools in Communication to 150oF Volume 62485 ft3 300.00 Initial Temperature:

250.00 70 80 90 100 200.00 110 120 150.00 Hours 100.00 50.00 0.00 20 15 10 5 0 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 14 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS TO 212OF o

Time Needed to Heat FHB Pools in Communication to 212 F 3

Volume 62485 ft 30 Initial Temperature:

25 70 F 80 F 90 F 20 100 F 110 F 120 F Hours 15 10 5

0 80.00 70.00 60.00 50.00 40.00 30.00 20.00 Decay Heat (mBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 15 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS TO 212OF Time to Heat FHB Pools in Communication to 212oF Volume 62485 ft3 350 Initial Temperature:

300 70 80 90 100 250 110 120 200 Hours 150 100 50 0

20 18 16 14 12 10 8 6 4 2 0 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 16 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS (INCLUDING CASK POOL) TO 150OF o

Time Needed to Heat FHB Pools Including the Cask Pool to 150 F 3

Volume 69905 ft 20 Initial Temperature:

70 F 15 80 F 90 F 100 F 110 F 120 F Hours 10 5

0 80.00 70.00 60.00 50.00 40.00 30.00 20.00 Decay Heat (mBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 17 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS (INCLUDING CASK POOL) TO 150OF Time needed to Heat FHB Pools Including Cask Pool to 150oF Volume 69905 ft3 300.00 Initial Temperature:

70 250.00 80 90 100 110 200.00 120 150.00 Hours 100.00 50.00 0.00 20 15 10 5 0 Decay Heat (MBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 18 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS (INCLUDING CASK POOL) TO 212OF o

Time Needed to Heat FHB Pools Including the Cask Pool to 212 F Volume 69905 ft3 30 Initial Temperature:

25 70 F 80 F 90 F 20 100 F 110 F Hours 120 F 15 10 5

0 80.00 70.00 60.00 50.00 40.00 30.00 20.00 Decay Heat (mBTU/hr)

RO Question 18 Number:

PERRY NUCLEAR POWER PLANT PDB-A0017

Title:

Use Category:

Pool Heatup Curves In-Field Reference Revision: Page:

10 19 of 19 TIME TO HEAT FUEL HANDLING BUILDING POOLS (INCLUDING CASK POOL) TO 212OF o

Time needed to Heat FHB Pools Including the Cask Pool to 212 F 3

Volume 69905 ft 400.00 Initial Temperature: 350.00 70 80 300.00 90 100 110 250.00 120 200.00 Hours 150.00 100.00 50.00 0.00 20 18 16 14 12 10 8 6 4 2 0 Decay Heat (MBTU/hr)

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 1 of 6 RECIRC DRIVE FLOW VS. TOTAL CORE FLOW Functional Location (B33)

Plant Data Book Effective Date: 9-20-07 Preparer: P. Curran / 8-24-07 Date

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 2 of 6

1.0 REFERENCES

1.1 Discretionary None 1.2 Obligations SVI-F41-T3008, Reactor Recirculation System Flow Data Verification performed May 2005 and June/July 2007.

SVI-B33-T1160, Jet Pump Operability Technical Specification SR 3.4.3.1.b ICS/SDS Displays RFPERF and RSPERF.

Commitments addressed in this document:

None 2.0 SCOPE OF REVISION Rev. 15 1. Revised format in accordance with NOP-SS-3007.

2. Updated references to include the June/July 2007 performance of SVI-F41-T3008. There were no changes to the graphs.

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 3 of 6 A0012 Slow Speed (Kgpm)

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 4 of 6 A0012 Slow Speed (%)

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 5 of 6 A0012 Fast Speed (Kgpm)

RO Question 48 Procedure Number:

PERRY NUCLEAR POWER PLANT PDB-A0012

Title:

Use Category:

Recirc Drive Flow vs. Total Core Flow General Skill Reference Revision: Page:

15 6 of 6 A0012 Fast Speed (%)

RO Question 51 H13-P622 NS4 MSL DRN ISOL NS4 BOP ISOL RX WTR SMPL VLV RHR ISOLATION INBD LOGIC TEST INBD LOGIC TEST B33-F019 LOGIC TEST INBD LOGIC TEST B21H-DS1B B21H-DS3B B21H-DS7B B21H-DS5B Light on Light on Light off Light off

RO Question 75 SRO Question 3 SRO Question 3 SRO Question 3 SRO Question 4 SRO Question 4 SRO Question 4 SRO Question 4 SRO Question 4 SRO Question 4 SRO Question 4 SRO Question 6 SRO Question 6 SRO Question 6 SRO Question 6 SRO QUESTIONS 9 & 11 INITIATING CONDITION INDEX PNPP No. 8852 Rev. 3/26/10 EPI-A1 M M M M O O O O EVENT CATEGORY D E

UNUSUAL EVENT D E

ALERT D E

SITE AREA EMERGENCY D GENERAL EMERGENCY E

1 1 1 1 2 Fuel clad degradation 2 Any loss or challenge to the Fuel Clad 2 Loss of RPV water level that has or will 2 Loss of two barriers, AND a loss or challenge to the 3 3 3 4

3 barrier. uncover fuel. third barrier.

4 5 5 A: FISSION PRODUCT Page 16 - AU1 Page 19 (FPB Matrix) - AA1 Page 18 - AS1 Page 19 (FPB Matrix) - AG1 BARRIER 1 1 1 Either a challenge or loss of both the DEGRADATION 2 Reactor Coolant System leakage. 2 Any loss or challenge to the Reactor 2 3 3 3 Fuel Clad barrier AND Reactor Coolant Coolant System barrier.

System barrier.

Page 17 - AU2 Page 19 (FPB Matrix) - AA2 Page 20 (FPB Matrix) - AS2 1 1 Challenge to either the Fuel Clad barrier 2 Any loss or challenge to the Containment 2 3 3 OR Reactor Coolant System barrier, AND barrier.

the loss of any additional barrier.

Page 19 (FPB Matrix) - AU3 Page 20 (FPB Matrix) - AS3 1

Inability to maintain plant in COLD 2 Complete loss of functions needed to B: LOSS OF DECAY 3 NOT APPLICABLE 4 SHUTDOWN achieve COLD SHUTDOWN HEAT REMOVAL NOT APPLICABLE 5

FUNCTIONS Page 20 - BA1 Page 21 - BS1 1 Inability to reach required shutdown within 1 1 1 C: LOSS OF SHUTDOWN 2 2 Failure to initiate or complete an Failure to initiate or complete an automatic Failure to initiate or complete a successful shutdown, 3 Technical Specification limits FUNCTIONS OR FAILURE automatic Reactor Scram once an RPS Reactor Scram once an RPS function is AND indication of an extreme challenge to the ability Page 22 - CU1 function is required. required, AND a manual Scram was NOT to cool the core.

TO SHUTDOWN successful.

Page 23 - CA1 Page 26 - CG1 Page 24 - CS1 1 1 1 1 2 Loss of all offsite power to Division 1 and 2 2 Power capability to Division 1 and 2 EH 2 2 Prolonged loss of all offsite power AND onsite power 3 3 3 Loss of all offsite power AND onsite power 3 EH Essential Busses for greater than 15 Essential Busses reduced to a single to Division 1 and 2 EH Essential Busses, AND D: A.C. POWER LOSS 4 5 minutes.

to Division 1 and 2 EH Essential Busses for D

power source for greater than 15 greater than 15 minutes. continuing degradation of core cooling capability.

Page 28- DU1 minutes, such that any additional single Page 32 - DG1 failure would result in a Station Blackout. Page 31 - DS1 Page 29 - DA1 Loss of all offsite power AND onsite power to Division 1 and 2 EH Essential 4

5 Busses for greater than 15 minutes.

D Page 30 - DA2 1

Degradation of Division 1 and 2 essential 2 DC power for greater than 15 minutes. Degradation of Division 1 and 2 essential E: D.C. POWER 4 3

DC power for greater than 15 minutes.

NOT APPLICABLE NOT APPLICABLE DEGRADATION 5 Page 34 - EU1 Page 35 - ES1 1 1 2 Fire within a Safe Shutdown Building NOT 2 Fire OR explosion affecting the F: FIRE OR 3 extinguished within 15 minutes. 3 operability of plant safety systems 4 4 NOT APPLICABLE NOT APPLICABLE EXPLOSION 5 D

5 required to establish or maintain safe D

Page 36 - FU1 shutdown.

1 2

3 Explosion affecting a Safe Shutdown 4 Building. Page 38 - FA1 5

D Page 37 - FU2 1 1 2 Unexpected increase in plant radiation 2 Increases in radiation levels within Safe G: INCREASED PLANT 3 4 levels. 3 4 Shutdown Buildings that impede RADIATION LEVELS 5 D

5 D

operation of systems required to Page 40 - GU1 maintain safe operations OR to 1 establish or maintain COLD 2 Uncontrolled fuel pool or Reactor Cavity SHUTDOWN.

3 water level decrease with irradiated fuel NOT APPLICABLE NOT APPLICABLE 4

5 remaining covered. Page 42 - GA1 D

1 Page 41 - GU2 2

3 Damage to irradiated fuel.

4 5

D Page 43 - GA2

INITIATING CONDITION INDEX PNPP No. 8852 Rev. 3/26/10 EPI-A1 M M M M O O O O EVENT CATEGORY D UNUSUAL EVENT D ALERT D SITE AREA EMERGENCY D GENERAL EMERGENCY E E E E 1 Any unplanned release of gaseous 1 Any unplanned release of gaseous 1 Site Boundary dose resulting from an 1 Site Boundary dose resulting from an actual or 2

H: INCREASED 3 radioactivity to the environment that 2

3 radioactivity to the environment that 2

3 actual or imminent release of gaseous 2

3 imminent release of gaseous radioactivity that RADIATION RELEASE TO 4 5 exceeds two times the ODCM Control 4

5 exceeds 200 times the ODCM Control 4 5

radioactivity that exceeds 100 mRem 4 exceeds 1000 mRem TEDE OR 5000 mRem CDE 5

THE ENVIRONMENT D limit for 60 minutes or greater. D limit for 15 minutes or greater. D TEDE dose OR 500 mRem CDE Child D Child Thyroid dose for the actual or projected duration Thyroid dose for the actual OR projected of the release.

Page 46 - HU1 Page 48 - HA1 duration of the release.

1 Any unplannned release of liquid 1 Any unplanned release of liquid Page 50 - HS1 Page 51 - HG1 2 2 3 radioactivity to the environment that radioactivity to the environment that 3

4 exceeds two times the ODCM Control 4 exceeds 200 times the ODCM Control 5 5 limit for 60 minutes or greater. D limit for 15 minutes or greater.

D Page 47 - HU2 Page 49 HA2 1 1 Control Room Evacuation has been Control Room evacuation has been I: CONTROL ROOM 2 2 initiated. initiated, AND plant control CANNOT be EVACUATION NOT APPLICABLE 3 3 4 4 established within 15 minutes. NOT APPLICABLE 5 Page 52 - IA1 5 D D Page 53 - IS1 1 1 Loss of most annunciators or indication 1 Inability to monitor a significant transient in J: LOSS OF Loss of most annunciators or indication 2 2 2

in the Control Room for greater than 15 in the Control Room with either : (1) a progress.

ANNUNCIATORS OR 3 3 3 minutes. significant transient in progress; OR (2) NOT APPLICABLE INDICATIONS compensatory indications are NOT Page 56 - JS1 Page 54 - JU1 available.

Page 55 - JA1 1

K: LOSS OF Loss of onsite OR in-plant 2

communications capabilities.

COMMUNICATIONS 3 4

5 Page 58 - KU1 D

NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE 1 Significant degradation of offsite 2

3 communications capabilities.

4 Page 59 - KU2 5

D 1 1 L: NATURAL OR Natural OR destructive phenomena Natural OR destructive phenomena 2 2 affecting the Protected Area boundary. affecting Safe Shutdown Buildings.

DESTRUCTIVE 3 3 NOT APPLICABLE NOT APPLICABLE 4 4 PHENOMENA 5 5 D

Page 60 - LU1 D Page 61 - LA1 1

Release of toxic OR flammable gasses 1 Release of toxic OR flammable gases M: RELEASE OF 2 2 3 affecting the Protected Area boundary within a Safe Shutdown Building which TOXIC OR 4 deemed detrimental to the safe 3

jeopardizes operation of systems required FLAMMABLE GAS 5 operation of the plant.

4 to maintain safe operations OR to NOT APPLICABLE NOT APPLICABLE 5

D D establish or maintain COLD SHUTDOWN.

Page 62- MU1 Page 63- MA1 1 1 Confirmed SECURITY CONDITION or HOSTILE ACTION resulting in loss of physical control N: SECURITY EVENTS 2 2 threat which indicates a potential 3 of the facility.

3 4 degradation in the level of safety of NOT APPLICABLE NOT APPLICABLE 4 5 the plant. 5 D

D Page 64 - NU1 1 1 Page 70 - NG1 HOSTILE ACTION within the PROTECTED 2 Notification of an airborne attack threat. 2 3 3 AREA 4 4 5

5 Page 66 - NA2 D D Page 69 - NS2 1

2 HOSTILE ACTION within the 3

4 OWNER CONTROLLED AREA 5

D Page 67 - NA3 O: EMERGENCY 1 Other conditions existing, which in the 1 Other conditions existing, which in the 1 Other conditions existing, which in the 1 Other conditions existing, which in the judgement of 2 2 2 2 judgement of the Emergency Coordinator, judgement of the Emergency Coordinator, judgement of the Emergency Coordinator, the Emergency Coordinator, warrant declaration of a COORDINATOR'S 3 3 warrant declaration of an Alert.

3 warrant declaration of a Site Area 3

4 warrant declaration of an Unusual Event. 4 4 4 General Emergency.

JUDGEMENT 5 5 5 Emergency. 5 D D D D Page 72- OU1 Page 73 - OA1 Page 74- OS1 Page 75 - OG1

Category A: Fission Product Barrier Degradation Initiating Conditions Entry Criteria AU1 AU1 U

N U

Fuel clad degradation High Offgas pretreatment air activity Reactor Coolant System sample indicates S greater than the Technical activity greater than Technical U Specification 3.7.5. Specification 3.4.8 limits. A L

E V

E N

Applicable Modes: T 1 2 3 4 5 NOTE Fuel clad degradation is NOT an issue when the Reactor is defueled. Damage to fuel in spent fuel pools is addressed in GU1.

Initiating Conditions Entry Criteria AU2 AU2 Greater than 10 gpm Greater than 30 gpm total Greater than 30 gpm total U unidentified leakage in leakage in Drywell leakage in Drywell. N Drywell. averaged over the previous U Reactor Coolant 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. S System leakage. U A

Greater than 2 gpm L increase in unidentified leakage within the previous E 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. V E

N Applicable Modes: T 1 2 3

Initiating Conditions Entry Criteria AS1 S

I T

AS1 RPV water level CANNOT be maintained greater than 0 inches. E A

R E

Loss of RPV water level Reactor is shutdown under all conditions without boron. A that has or will uncover fuel E M

E R

G E

N C

Y Applicable Modes:

1 2 3 4 5 NOTE AS1 is applicable only to non-ATWS situations in which RPV level was NOT intentionally lowered per

<EOP-01A> as a means of power control. Refer to Event Category C for classification under ATWS conditions in which RPV water level is intentionally lowered.

PNPP No. 9855 Rev. 8/26/08 FISSION PRODUCT BARRIER MATRIX EPLAN/EPI-A1/PSI-0019 INITIATING CONDITIONS REACTOR PRESSURE REACTOR COOLANT EMERGENCY COORDINATOR UNUSUAL EVENT VESSEL LEVEL DRYWELL RADIATION SYSTEM ACTIVITY JUDGEMENT INSTRUCTIONS AU3 Any loss or challenge to the Entry into SAG-1, Primary Containment Drywell radiation monitor Sample activity is equal to or Containment barrier.

LOSS Any condition that, in the judgment Flooding.

3 reading greater than greater than 300 uCi/gm dose of the Emergency Coordinator 1. For each of the three barriers, Modes: 1 2 3 FUEL CLAD 4,000 Rem/hr. equivalent iodine-131.5 indicates loss of the Fuel Clad determine if any LOSS or CHALLENGE CRITERIA barrier.1 criteria have been met. ALERT BARRIER AA1 Any loss or challenge to

2. Compare the barrier LOSS(es) and Fuel Clad barrier.

CHALLENGE RPV level is less RPV level Any condition that, in the judgment Modes: 1 2 3 CANNOT be CHALLENGE(s) to the initiating than 0 inches. 2 of the Emergency Coordinator determined. NOT APPLICABLE NOT APPLICABLE conditions listed, and make the AA2 Any loss or challenge to the indicates a challenge to the Fuel CRITERIA appropriate event declaration. Reactor Coolant System barrier.

Clad barrier.1 Modes: 1 2 3 SITE AREA EMERGENCY REACTOR PRESSURE DRYWELL RADIATION REACTOR COOLANT EMERGENCY COORDINATOR AS2 Either a challenge or loss of both REACTOR PRESSURE CONTROL DRYWELL PRESSURE VESSEL LEVEL SYSTEM BYPASS JUDGEMENT the Fuel Clad barrier AND Reactor Coolant System barrier.

RPV level less An SRV is Drywell pressure MSL break outside containment Any condition that, in the Drywell radiation SRV stuck Emergency Modes: 1 2 3 REACTOR COOLANT SYSTEM than 0 inches. 2 being cycled to greater than 1.68 psig. exceeding one or more MSIV judgment of the Emergency monitor reading greater open. Depressurization control RPV Tech. Spec. allowable values. Coordinator, indicates loss of than 135 Rem/hr. is required. AS3 Challenge to either the Fuel Clad Containment penetration does the RCS barrier.1 LOSS pressure. barrier OR Reactor Coolant NOT isolate on a valid closure System barrier, AND the loss of signal. any additional barrier.

CRITERIA BARRIER Sample activity is equal to Indication of RCS leakage Immediate Operator actions in Modes: 1 2 3 or greater than 300 uCi/gm inside the Drywell. the Control Room are NOT 5

dose equivalent Iodine-131. successful in isolating affected GENERAL EMERGENCY penetration. Loss of two barriers, AND a loss AG1 CHALLENGE or challenge to the third barrier.

Unisolable primary system Any condition that, in the Modes: 1 2 3 discharging outside judgment of the Emergency NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE CRITERIA Containment per either Coordinator, indicates a EOP-03 or EOP-05. 4 challenge to the RCS barrier.1 REACTOR PRESSURE CONTAINMENT CONTAINMENT CONTAINMENT VESSEL LEVEL RADIATION CONTAINMENT ISOLATION EMERGENCY COORDINATOR JUDGEMENT HYDROGEN PRESSURE Entry into SAG-1, Containment Intentional venting of Intentional venting of Unisolable primary system Primary Containment Any condition that, in the judgment of the Emergency Containment required per Containment required per penetration does discharging outside Flooding 3 Containment per either Coordinator, indicates loss of the Containment SAG-2 EOP-02 NOT isolate on a EOP-03 or EOP-05. 4 barrier.1 valid closure signal.

NOT APPLICABLE Immediate Operator (Loss of the Containment barrier may include a rapid LOSS actions in the Control unexplained decrease in Containment pressure CONTAINMENT Room are NOT following an initial increase.)

CRITERIA BARRIER successful in isolating affected penetration.

Pathway to the environment exists via penetration.

Containment CHALLENGE Any condition that, in the judgment of the Emergency Containment radiation In the UNSAFE pressure is Coordinator, indicates a challenge to the Containment monitor reading greater region on the HCL greater than NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE barrier.1 CRITERIA than 20,000 Rem/hr. figure. 15 psig and increasing.

NOTE: If the criteria for loss has been met, then by definition, the challenge criteria has been met for that same fission product barrier.

FOOTNOTES : 1. Those thresholds for which a LOSS or CHALLENGE is determined to be IMMINENT (i.e., within the next 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ), classify as though the threshold(s) has been exceeded.

2. RPV level is less than 0 inches is both a FUEL CLAD BARRIER CHALLENGE CRITERIA and a REACTOR COOLANT SYSTEM BARRIER LOSS CRITERIA.

3. Entry into SAG-1, Primary Containment Flooding is both a FUEL CLAD BARRIER LOSS CRITERIA and a CONTAINMENT BARRIER LOSS CRITERIA.

4. Unisolable primary system discharging outside containment per EOP-03 and/or EOP-05 is both a REACTOR COOLANT SYSTEM BARRIER CHALLENGE CRITERIA and a CONTAINMENT BARRIER LOSS CRITERIA.

5. Sample activity is equal to or greater than 300 uCi/gm dose equivalent Iodine-131 is both a FUEL CLAD BARRIER LOSS CRITERIA and a contributor to a REACTOR COOLANT SYSTEM BARRIER LOSS CRITERIA.

Category B: Loss of Decay Heat Removal Functions Initiating Conditions Entry Criteria Loss of Shutdown Cooling Mode function for RHR loop A.

BA1 BA1 Loss of Shutdown Cooling Mode function for RHR loop B.

A Inability to maintain L plant in COLD RCS temperature exceeds COLD Uncontrolled temperature rise E SHUTDOWN SHUTDOWN limit of 200°F per approaching 200°F RCS temperature. R Technical Specification Table 1.1-1. T Applicable Modes:

4 5 NOTES

The IC remains applicable for situations in which an increase in RCS temperature greater than 200°F results in a change to Mode 3.

The above criteria is met as soon as it becomes known that sufficient cooling CANNOT be restored to maintain temperature below 200°F regardless of the current temperature. The intent of BA1 is NOT to classify based on an unplanned excursion above 200°F when heat removal capability is available.

Uncontrolled means that RCS temperature increase is NOT the result of planned actions by plant staff.

Initiating Conditions Entry Criteria BS1 RHR Loops A and B are NOT capable of lowering RPV temperature.

S I

T E

A R

E BS1 A Complete loss of E functions needed to M achieve COLD E SHUTDOWN R The plant is operating in the UNSAFE Region on the HCL figure. G E

N C

Y Applicable Modes:

1 2 3

Category C: Loss of Shutdown Functions or Failure to Shutdown Initiating Conditions Entry Criteria CU1 U

CU1 Plant is NOT brought to the required operating mode within the Technical N Specification Required Action Completion Time following entry into an LCO. U Inability to reach S required shutdown U within Technical A Specification limits L E

V E

N Applicable Modes: T 1 2 3 NOTE Declaration should be made because of equipment failures that prevent the performance of an orderly shutdown or failure to meet the shutdown completion time from the time discovered and a required action being entered.

Declaration of an Unusual Event is based on the time at which the specified completion time period elapses and is NOT related to how long a condition may have existed before it was discovered.

Initiating Conditions Entry Criteria CA1 CA1 Actuation of RPS has occurred or should Actuation of RRCS has occurred or have occurred. should have occurred.

Failure to initiate or A complete an automatic L Reactor Scram once an The reactor is NOT shutdown under all conditions without boron. E RPS function is R required T Applicable Modes:

1 2 NOTES

CA1 is applicable if either Mode 1 or 2 existed when the transient started and NOT the mode which exists at the time of classification.

Entry criteria is applicable for actions taken by an Operator to manually initiate either RPS or RRCS prior to or after exceeding an automatic actuation setpoint.

Initiating Conditions Entry Criteria Actuation of RPS has occurred or should Actuation of RRCS has occurred or CS1 have occurred. should have occurred.

The reactor is NOT shutdown under all conditions without boron. S CS1 I T

Failure to initiate or Manual operator actions Reactor power CANNOT Suppression Pool E complete an automatic taken at 1H13-P680 to be determined. temperature is greater than Reactor Scram once an insert control rods were 110°F. A RPS function is NOT successful in R required, AND a lowering Reactor power to E manual Scram was less than 4%. A NOT successful E

M E

R G

E N

C Applicable Modes: Y 1

NOTE Refer to next page.

NOTES

CS1 is applicable if Mode 1 existed when the transient started and NOT the mode which exists at the time of classification. Refer to CA1 for Mode 2 applicability.

Manual Operator actions are defined as any set of actions by the Reactor Operator at 1H13-P680 which results in a scram signal. These actions include placing the Reactor Mode Switch in the SHUTDOWN position, arming and depressing the RPS Manual Scram push buttons, and arming and depressing the RRCS Manual ARI push buttons. Injection of boron is NOT considered in reducing reactor power below 4%.

A concurrent challenge to the ability to cool the core would escalate this event to General Emergency per CG1.

Initiating Conditions Entry Criteria Actuation of RPS has occurred or should Actuation of RRCS has occurred or CG1 have occurred. should have occurred.

G E

CG1 N The reactor is NOT shutdown under all conditions without boron. E Failure to initiate or R complete a successful Manual operator actions taken at Reactor power Suppression Pool A shutdown, AND 1H13-P680 to insert control rods CANNOT be temperature is greater L indication of an extreme were NOT successful in lowering determined. than 110°F.

challenge to the ability Reactor power to less than 4%. E to cool the core M E

R G

Any of the following conditions exist: E N

Entry into <SAG-1>, Primary Containment Flooding. C

In the UNSAFE region on the HCL figure. Y Applicable Modes:

1 NOTES

CG1 is applicable if Mode 1 existed when the transient started and NOT the mode which exists at the time of classification. Refer to CA1 for Mode 2 applicability.

Entry criteria is applicable for actions taken by an Operator to manually initiate either RPS or RRCS prior to or after exceeding an automatic actuation setpoint.

(INTENTIONALLY BLANK)

Category D: A. C. Power Loss Initiating Conditions Entry Criteria DU1 ONI-R10 entered for a Loss of Off-site Power (LOOP). U N

DU1 U S

Loss of all offsite power U to Division 1 and 2 EH Either of the following power sources CANNOT be made available within 15 minutes A Essential Busses for for energizing bus EH11: L greater than 15 minutes

Normal Preferred E

Alternate Preferred V E

N Either of the following power sources CANNOT be made available within 15 minutes T for energizing bus EH12:

Normal Preferred Applicable Modes:

Alternate Preferred 1 2 3 4 5 D NOTE Failure of either bus EH11 or EH12 to be supplied from its respective diesel generator is evaluated for escalation to an Alert under DA1 for Modes 1, 2 and 3. Failure of both busses EH11 and EH12 to be supplied from their respective diesel generators (Station Black Out) is evaluated for escalation to an Alert under DA2 for Modes 4 and 5 and to a Site Area Emergency under DS1 for Modes 1, 2 and 3.

Initiating Conditions Entry Criteria Essential AC power reduced to only one of the following power sources for greater DA1 DA1 than 15 minutes:

Power capability to

Normal Preferred Division 1 and 2 EH

Alternate Preferred Essential Busses

Division 1 Diesel Generator A reduced to a single

Division 2 Diesel Generator L power source for E greater than R 15 minutes, such that T any additional single failure would result in Station Loss of the single remaining power source will result in a loss of AC power to both Blackout busses EH11 and EH12.

Applicable Modes:

1 2 3 NOTES

Escalation to a Site Area Emergency is evaluated under DS1, for Operating Modes 1, 2 and 3, based on a total loss of AC power to both busses EH11 and EH12.

A total loss of AC power to busses EH11 and EH12 while in Operating Modes 4 and 5 is classified as an Alert under DA2. No escalation path exists to a Site Area Emergency for Operating Modes 4 and 5.

Initiating Conditions Entry Criteria Both busses EH11 and EH12 CANNOT be energized from the Normal Preferred DA2 DA2 source within 15 minutes.

Loss of all offsite power A AND onsite power to Both busses EH11 and EH12 CANNOT be energized from the Alternate Preferred L Division 1 and 2 EH source within 15 minutes. E Essential Busses for R greater than T 15 minutes. Both busses EH11 and EH12 CANNOT be energized from the Associated Diesel Generator source within 15 minutes.

Applicable Modes:

4 5 D

Initiating Conditions Entry Criteria DS1 Both busses EH11 and EH12 CANNOT be energized from the Normal Preferred source within 15 minutes. S I

DS1 T E

Loss of all offsite power AND onsite power to Both busses EH11 and EH12 CANNOT be energized from the Alternate Preferred A Division 1 and 2 EH source within 15 minutes. R Essential Busses for E greater than 15 minutes A E

Both busses EH11 and EH12 CANNOT be energized from the Associated Diesel M Generator source within 15 minutes. E R

G E

N Applicable Modes: C Y

1 2 3 NOTE Escalation to a General Emergency is evaluated under DG1 for Modes 1, 2 and 3, based on a continuing degradation of core cooling capability.

Initiating Conditions Entry Criteria Both busses EH11 and EH12 CANNOT be energized from the Normal Preferred DG1 source.

G Both busses EH11 and EH12 CANNOT be energized from the Alternate Preferred E source. N DG1 E R

Prolonged loss of all Both busses EH11 and EH12 CANNOT be energized from the Associated Diesel A offsite power AND Generator source. L onsite power to Division 1 and 2 EH E Busses, AND continuing Restoration of power to either of the RPV water level RPV water level M degradation of core following busses is NOT likely in less less than 0 inches. CANNOT be E cooling capability than four hours: determined. R G

EH11 E

EH12 N C

Applicable Modes: Y 1 2 3

(INTENTIONALLY BLANK)

Category E: D. C. Power Degradation Initiating Conditions Entry Criteria EU1 U

EU1 Voltage on ED-1-A buss is less than 105 VDC for greater than 15 minutes. N U

Degradation of S Division 1 and 2 U essential DC power for A greater than 15 minutes L E

Voltage on ED-1-B buss is less than 105 VDC for greater than 15 minutes. V E

N Applicable Modes: T 4 5 NOTE The same set of conditions as described in this EAL would be classified as Site Area Emergency under ES1 if they occurred during Modes 1, 2, or 3.

Initiating Conditions Entry Criteria ES1 S

I Voltage on ED-1-A buss is less than 105 VDC for greater than 15 minutes. T E

ES1 A

Degradation of R Division 1 and 2 E essential DC power for A greater than 15 minutes E

M E

R Voltage on ED-1-B buss is less than 105 VDC for greater than 15 minutes. E N

G C

Y Applicable Modes:

1 2 3

Category F: Fire or Explosion Initiating Conditions Entry Criteria Fire within any Safe Shutdown Building. FU1 U

Fire CANNOT be extinguished within Fire CANNOT be extinguished within N FU1 15 minutes of the verification of alarm. 15 minutes of the notification received in U the Control Room from plant personnel S Fire within a Safe that a fire exists. U Shutdown Building A NOT extinguished L within 15 minutes E

V E

N Applicable Modes: T 1 2 3 4 5 D NOTE Verification in this context means those actions taken in the Secondary Alarm Station (SAS) to determine that the alarm is NOT spurious. Verification includes the receipt of multiple or independent alarms or confirmation of a single detector by visual inspection of the affected area by a first responder. List of Safe Shutdown Buildings is found in Section 3 Definitions, sub-step 3.10.

Initiating Conditions Entry Criteria FU2 U

N FU2 Report by plant personnel confirming the occurrence of an explosion within the U Protected Area resulting in visible damage to a Safe Shutdown Building. S Explosion affecting a U Safe Shutdown Building A L

E V

E N

Applicable Modes: T 1 2 3 4 5 D NOTE No attempt is made to assess the magnitude of the damage. The occurrence of the explosion with reports of damage (deformation/scorching) is sufficient for declaration. Actual damage to safe shutdown equipment is covered under Alert FA1. List of Safe Shutdown Buildings is found in Section 3 Definitions, sub-step 3.10.

Initiating Conditions Entry Criteria Either of the following has been confirmed: FA1 FA1

Fire in a Safe Shutdown Building. A Fire OR explosion

Explosion in a Safe Shutdown Building. L affecting the operability E of plant safety systems Plant personnel at the scene report visible Affected safe shutdown system indicates R required to establish or damage to safe shutdown equipment or degraded performance. T maintain safe shutdown components.

Applicable Modes:

1 2 3 4 5 D NOTE The inclusion of a report of visible damage should NOT be interpreted as mandating a lengthy damage assessment prior to classification. NO attempt is made in this EAL to assess the actual magnitude of damage beyond the immediate area. The occurrence of the explosion or fire with reports of evidence of damage (e.g., deformation, scorching) is sufficient for declaration. List of Safe Shutdown Buildings is found in Section 3 Definitions, sub-step 3.10.

NOTE Safe Shutdown System/Equipment refers to equipment identified in the Safe Shutdown Capability Report. This is the minimum list of equipment required to achieve and maintain COLD SHUTDOWN (including all auxiliary equipment such as AC/DC power, cooling water and instrumentation). A detailed list is provided in the <Appendix R Evaluation - Safe Shutdown Capability Report>.

Safe Shutdown System/Equipment list: (Division 1 and 2 only)

Reactor Protection System Control Rod Drive Hydraulics Automatic Depressurization System/SRV Reactor Core Isolation Cooling Low Pressure Core Spray Low Pressure Coolant Injection - A/B/C Suppression Pool Cooling Shutdown Cooling Safety-Related Instrument Air Emergency Service Water Emergency Service Water Screen Wash Emergency Service Water Pump House Ventilation ECCS Pump Room Cooling Systems Diesel Generator Building Ventilation Stand-by Diesel Generator (DG)

DG Fuel Oil Storage/Transfer Electrical Power Distribution Emergency Closed Cooling Pump Area Cooling Emergency Closed Cooling Control Complex Chilled Water MCC, Switchgear and Miscellaneous Electrical Equipment Areas HVAC System Battery Room Exhaust Control Room HVAC and Emergency Recirculation System

(

Reference:

<NUMARC/NESP-007> (Rev. 2), Unusual Event HA2)

Category G: Increased Plant Radiation Levels Initiating Conditions Entry Criteria A valid area radiation monitor (D21) Health Physics surveys indicate an increase by GU1 reading increases by a factor of 1000 a factor of 1000 times over normally expected over normal levels. area radiation levels. U N

U S

GU1 U A

Unexpected increase in L plant radiation levels E

In-plant radiation level increase CANNOT be attributed to any of the following: V E

the start-up and operation of plant equipment or systems within design parameters. N

the planned movement of radioactive materials. T

the planned movement of shielding (i.e., plugs, lead shot, etc.)

Applicable Modes:

1 2 3 4 5 D NOTE Normal area radiation levels can be considered as the highest reading in the past 24-hour period, excluding the current peak value.

Initiating Conditions Entry Criteria Uncontrolled decrease in one or more of the following fuel pools containing irradiated GU2 fuel:

Reactor Cavity U

FHB Fuel Storage and Preparation Pool N

FHB Fuel Transfer Pool U GU2

FHB Spent Fuel Storage Pool S

FHB Cask Pit U

CNTMT Fuel Storage Pool A Uncontrolled fuel pool or Reactor Cavity water

CNTMT Fuel Transfer Pool L level decrease with irradiated fuel E remaining covered. V E

N T

Applicable Modes:

1 2 3 4 5 D

Initiating Conditions Entry Criteria Area radiation levels of Area radiation levels of greater than 6000 mRem/hr in a GA1 greater than 15 mRem/hr in Safe Shutdown Building, as determined by either:

any of the following areas:

area radiation surveys

Control Room

installed or portable radiation monitors GA1

Central Alarm Station Increases in radiation levels within Safe Shutdown Buildings Access is required to maintain safe operation or perform A that impede operation a safe shutdown, as determined by the Shift Manager. L of systems required to E maintain safe R operations OR to T establish or maintain COLD SHUTDOWN Applicable Modes:

1 2 3 4 5 D NOTE This IC addresses increased radiation levels that impede necessary access to operating stations or other areas containing equipment that must be operated manually in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant.

Initiating Conditions Entry Criteria Damage to irradiated fuel Water level observed to be below top of GA2 A valid HIGH alarm on one or more of the gate sill separating any of the GA2 the following radiation monitors: following containing irradiated fuel:

Damage to

SPENT FUEL POOL D21-K332

Reactor Cavity A irradiated fuel

UPPER POOL AREA 1D21-K083

FHB Fuel Storage and Preparation L

FUEL PREP POOL D21-K322 Pool E

FHB VENT EXH GAS D17-K716

FHB Fuel Transfer Pool R

CNTMT ATMOS GAS 1D17-K686

FHB Spent Fuel Storage Pool T

FHB Cask Pit Applicable Modes:

CNTMT Fuel Storage Pool

CNTMT Fuel Transfer Pool 1 2 3 4 5 D NOTE The intent of this EAL is to allow observations from plant personnel to be factored into the declaration decision and is not intended to direct an entry into an area solely to observe pool level. The gate sill is the lip between the pools where the bottom of the gate would sit if installed.

NOTE This EAL is only applicable to emergency conditions and is NOT applicable to pre-planned evolutions such as a reactor drain down to remove or re-install the reactor head during refueling operations.

NOTE Damage to irradiated fuel is defined as; a degraded fuel bundle that results in the release of fission product gasses normally present in the fuel rod gap to the surrounding environment.

(INTENTIONALLY BLANK)

Category H: Increased Radiation Release to the Environment Initiating Conditions Entry Criteria A valid reading greater than TWO times Routine or as required sample analysis HU1 the HIGH alarm setpoint on one or more indicates a release rate greater than two of the following plant gaseous effluent times ODCM 3.11.2.1 limits.

monitors lasting greater than or equal to 60 minutes:

HU1

PLANT VENT GAS 1D17-K786 U

OG VENT PIPE GAS 1D17-K836 The release lasts for greater than or equal N Any unplanned release

TB/HB VENT GAS 1D17-K856 to 60 minutes. U of gaseous radioactivity

PLANT VENT GAS 2D17-K786 S to the environment that U exceeds two times the A ODCM Control limit Chemistry sample analysis methods L for 60 minutes or CANNOT confirm within 60 minutes of greater. receipt of the HIGH alarm, on one or E more of the plant gaseous effluent V monitors, that effluent levels are less than E two times ODCM 3.11.2.1 limits. N T

Applicable Modes:

1 2 3 4 5 D NOTE It is NOT intended that the release be averaged over 60 minutes. Further, the Emergency Coordinator should NOT wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release will exceed TWO times the ODCM Control 3.11.2.1 limit for greater than 60 minutes.

Initiating Conditions Entry Criteria A valid reading greater than A valid reading greater Routine or as required HU2 1.2E3 cpm above background than 20 times the sample analysis indicates for one or more of the HIGH-HIGH alarm a release rate greater than HU2 following liquid process setpoint on NCC HX and two times ODCM monitors lasting greater than or RW EFF Common Control 3.11.1.1 limits. U equal to 60 minutes: Process Radiation N Recorder 0D17-R0170. U

EMERGENCY S Any unplanned release SERVICE WATER U of liquid radioactivity LOOP A D17-K604 A to the environment that L exceeds two times the

EMERGENCY ODCM Control limit SERVICE WATER E for 60 minutes or LOOP B D17-K605 V greater. E Chemistry sample analysis Release CANNOT be The release lasts for N methods CANNOT confirm terminated within greater than or equal to T within 60 minutes of receipt of 60 minutes of exceeding 60 minutes.

the HIGH-HIGH alarm, on NCC HX and RW EFF either ESW Loop A or B Common Process radiation monitors, that liquid Radiation Recorder release levels are less than two HIGH-HIGH alarm Applicable Modes: setpoints.

times the ODCM 1 2 3 4 5 D Control 3.11.1.1 limits.

NOTE It is NOT intended that the release be averaged over 60 minutes. Further, the Emergency Coordinator should NOT wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release will exceed TWO times the ODCM Control 3.11.1.1 limit for greater than 60 minutes.

Initiating Conditions Entry Criteria A valid reading greater A valid offscale high Routine or as Portable HA1 than 200 times the reading on one or more required sample survey HA1 HIGH alarm setpoint on of the following plant analysis indicates instruments Any unplanned one or more of the gaseous effluent a release rate indicate release of gaseous following plant gaseous monitors:* greater than radiation radioactivity to the effluent monitors:* 200 times ODCM levels of A environment that Control 3.11.2.1 greater than L

PLANT VENT GAS 1D17-K786 exceeds 200 times the

OG VENT PIPE GAS 1D17-K836 limits. 10 mRem/hr E

PLANT VENT GAS 1D17-K786 ODCM Control limit

TB/HB VENT GAS 1D17-K856

OG VENT PIPE GAS 1D17-K836 at the Site R

PLANT VENT GAS 2D17-K786 for 15 minutes or

TB/HB VENT GAS 1D17-K856 Boundary for T

PLANT VENT GAS 2D17-K786 greater greater than or equal to The valid reading lasts The release lasts 15 minutes.

Applicable Modes: for greater than or equal for greater than or to 15 minutes. equal to 1 2 3 4 5 D 15 minutes.

Perform an emergency dose calculation (i.e., CADAP run) within 15 minutes to determine if the Site Area Emergency entry criteria is met.

NOTE 1 ODCM [coolant activity] methodology is used to determine the threshold for this EAL. The Site Area Emergency threshold for EAL, HS1, is established using a clad damage source term. As a result, if the EAL threshold for the Alert is met AND there is clad damage, then the Site Area Emergency thresholds may be exceeded.

NOTE 2 It is NOT intended that the release be averaged over 15 minutes. Rather, the Emergency Coordinator should declare the event as soon as it is determined that the release will exceed 200 times the ODCM Control 3.11.2.1 limit for greater than 15 minutes.

Initiating Conditions Entry Criteria A valid reading greater than A valid Reading Routine or as required HA2 1.2E5 cpm above background for greater than sample analysis indicates one or more of the following liquid 2000 times the a release rate greater than process monitors: HIGH-HIGH alarm 200 times ODCM HA2 setpoint on NCC HX Control 3.11.1.1 limits.

EMERGENCY SERVICE and RW EFF Any unplanned release WATER Common Process of liquid radioactivity to LOOP A D17-K604 Radiation Recorder A the environment that 0D17R0170. The release lasts for L exceeds 200 times the

EMERGENCY SERVICE greater than or equal to E ODCM Control limit WATER 15 minutes. R for 15 minutes or LOOP B D17-K605 Release CANNOT be T greater terminated within 15 minutes of exceeding NCC HX and RW EFF The reading lasts greater than or Common Process equal to 15 minutes. Radiation Recorder HIGH-HIGH alarm setpoints.

Applicable Modes:

1 2 3 4 5 D NOTE It is NOT intended that the release be averaged over 15 minutes. Rather, the Emergency Coordinator should declare the event as soon as it is determined that the release will exceed 200 times the ODCM Control 3.11.1.1 limit for greater than 15 minutes.

Initiating Conditions Entry Criteria A valid indication greater than the listed Emergency dose Field survey HS1 reading for one or more of the following calculations, using results indicate HS1 plant gaseous effluent monitors: actual meteorology that one or more of indicate that one or the following have S Site Boundary dose

PLANT VENT GAS 1D19-K300 3.8E-1 Ci/cc more of the been met at the I resulting from an actual

OG VENT GAS 1D19-K400 2.2E0 Ci/cc following are met at Site Boundary: T or imminent release of

TB/HB VENT GAS 1D17-K856 1.6E4 cpm the Site Boundary: E gaseous radioactivity

PLANT VENT GAS 2D19-K300 6.0E-1 Ci/cc

Greater than that exceeds 100 mRem

Greater than 100 mRem/hr A TEDE dose OR 100 mRem Whole Body R 500 mRem CDE Child TEDE

Greater than E Thyroid dose for the

Greater than 500 mRem A actual or projected 500 mRem CDE Child duration of the release CDE Child Thyroid E Thyroid M E

R G

Emergency dose calculations CANNOT Dose rates are E confirm, within 15 minutes of exceeding expected to N limit, that levels at the Site Boundary are continue for C less than 100 mRem TEDE and 500 mRem greater than or Y CDE Child Thyroid dose using actual equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

meteorology.

Applicable Modes:

1 2 3 4 5 D

Initiating Conditions Entry Criteria A valid indication greater than the reading Emergency dose Field survey results listed for one or more of the following calculations, using indicate that one or HG1 plant gaseous effluent monitors: actual meteorology more of the indicate that one or following have been HG1

PLANT VENT GAS 1D19-K300 3.8E0 µCi/cc more of the met at the Site

OG VENT GAS 1D19-K400 2.2E1 µCi/cc following are met Boundary:

Site Boundary dose

TB/HB VENT GAS 1D17-K856 1.6E5 cpm at the Site G resulting from an actual

PLANT VENT GAS 2D19-K300 6.0E0 µCi/cc Boundary:

Greater than E or imminent release of 1000 mRem/hr N gaseous radioactivity

Greater than Whole Body E that exceeds 1000 mRem

Greater than R 1000 mRem TEDE dose TEDE 5000 mRem A OR 5000 mRem CDE

Greater than CDE Child L Child Thyroid dose for 5000 mRem Thyroid the actual or projected CDE Child E duration of the release Thyroid M E

Emergency dose calculations CANNOT Dose rates are R confirm, within 15 minutes of exceeding expected to G above limit, that levels at the Site continue for greater E Boundary are less than 1000 mRem TEDE than or equal to N and 5000 mRem CDE Child Thyroid dose 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . C Applicable Modes: using actual meteorology. Y 1 2 3 4 5 D NOTE Exceeding the entry criteria for HG1 may require the initiation of an RPV emergency depressurization per

<EOP-05>. Ensure Shift Manager is notified immediately whenever the above entry criteria for a General Emergency is met.

Category I: Control Room Evacuation Initiating Conditions Entry Criteria IA1 IA1 Control Room Entry into <ONI-C61>. A evacuation has been L initiated E R

Applicable Modes: T 1 2 3 4 5 D NOTE An inability to establish plant control from outside the Control Room will escalate this event to a Site Area Emergency per IS1.

Initiating Conditions Entry Criteria IS1 S

IS1 Entry into <ONI-C61>. I T

E Control Room A evacuation has been Within 15 minutes of entry into <ONI-C61>, Operator(s) located at the remote R initiated, AND plant shutdown controls CANNOT establish control of one or more of the following E control CANNOT be parameters per <IOI-11>: A established within 15 minutes.

RPV level E

RPV pressure M

Suppression Pool temperature E

Reactor power R

Decay heat removal, if required G E

N C

Applicable Modes: Y 1 2 3 4 5 D NOTE A maximum 15 minute time frame for the physical transfer of control of required systems was established by

<NUMARC/NESP-007>. Control at the Remote Shutdown Areas is accomplished by the repositioning of control transfer switches per <IOI-11>. Control is assumed unless indication of the absence of control is present.

Category J: Loss of Annunciators or Indication Initiating Conditions Entry Criteria Unplanned loss of most Control Room Unplanned loss of most Control Room JU1 annunciators for greater than 15 minutes. indication for greater than 15 minutes.

U JU1 N U

Loss of most S annunciators or U indication in the A Control Room for L greater than 15 minutes In the Shift Managers opinion, increased surveillance is warranted to safely operate E the plant. V E

N Applicable Modes: T 1 2 3 NOTE Quantification of most is left to the Shift Manager. It is NOT intended that plant personnel perform a detailed count of the instrumentation lost, but rather make a judgment call with approximately 75% being the threshold.

Initiating Conditions Entry Criteria JA1 JA1 Unplanned loss of most Control Room Unplanned loss of most Control Room Loss of most annunciators for greater than 15 minutes. indication for greater than 15 minutes.

annunciators or indication in the Control Room with A either: (1) a L significant transient in E progress, OR In the Shift Managers opinion, increased surveillance is warranted to safely operate R (2) compensatory the plant. T indicators are NOT available.

A significant plant transient is in progress. Compensatory indications, i.e., Integrated Computer System (ICS), are NOT Applicable Modes: available.

1 2 3 NOTES

Quantification of most is left to the Shift Manager. It is NOT intended that plant personnel perform a detailed count of the instrumentation lost, but rather make a judgment call with approximately 75% being the threshold.

A significant transient includes response to automatic OR manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injection, or thermal power oscillations of 10% or greater.

Initiating Conditions Entry Criteria Loss of most Control Room Loss of most Control Room indication. JS1 annunciators.

S I

Compensatory indicators, i.e., Integrated Computer System (ICS), are NOT available. T JS1 E A

Inability to monitor a R significant transient in A significant transient is in progress. E progress A E

M E

R G

E N

Sufficient indication is NOT available to directly monitor plant critical safety C parameters for EOPs entered due to the transient. Y Applicable Modes:

1 2 3

(INTENTIONALLY BLANK)

Category K: Loss of Communications Initiating Conditions Entry Criteria Loss of all five Plant Public Address System channels. KU1 U

Loss of all of the following Plant Radio System channels: N KU1 U

Channel 1 S Loss of onsite OR in-

Channel 2 U plant communications

Channel 3 A capabilities L E

V E

N T

Applicable Modes:

1 2 3 4 5 D

Initiating Conditions Entry Criteria Loss of the State and County Notification Circuit (5-way) reported to the Control KU2 Room.

U Loss of offsite long distance calling capability on three or more of the following N KU2 systems circuits for greater than 15 minutes: U S

Significant degradation

Control Room private (259-) lines U of offsite

Private Branch Exchange, Service Building (5000) Switch A communications

Private Branch Exchange, Warehouse Building (6000) Switch L capabilities

Company Off-Premise Exchange E

V E

N T

Applicable Modes:

1 2 3 4 5 D NOTES

A loss of the 5 Way Circuit refers to the inability to contact one or more of the four offsite contacts: the State of Ohio, and the counties of Ashtabula, Geauga, and Lake. Testing to determine 5-Way operability or to initiate circuit restoration actions are governed under <PSI-0007>.

Direct (259-) off-site calling capability from the Control Room via private lines refer to: autodialer at the US console, and private (259-) lines on the superphones and line at the SAS console.

Category L: Natural or Destructive Phenomena

Initiating Conditions Entry Criteria Control Room receives Report by High Indications in the Control LU1 report from plant plant sustained Room of a Main Turbine personnel who felt an personnel winds greater Trip or failure earthquake. confirming than 70 mph either of the for equal to following or greater U within the than N LU1 Protected 15 minutes. U Area S Natural OR destructive boundary: U phenomena affecting WHITE AMBER Turbine Catastrophic A the Protected Area event light(s) on

tornado casing damage to L boundary indicator Seismic strike penetration. generator light on Monitoring

plane or seals. E local Panel train crash V Seismic 0H13-P969. E Monitoring N Panel T 0H51-P021.

Applicable Modes:

1 2 3 4 5 D

Initiating Conditions Entry Criteria Control Room receives Report of Sustained Greater Report by report from plant visible damage high winds than plant LA1 personnel who felt an to any Safe with a <EOP-03> personnel earthquake. Shutdown velocity Maximum confirming a Building caused greater than Safe turbine by any of the 90 mph for Operating failure which following: 15 minutes Value for results in LA1 Either of the following or longer. Area Water penetration indications present:

tornado Level of the turbine Natural OR destructive

aircraft, (internal casing.

phenomena affecting

YELLOW seismic barge or train flooding) A Safe Shutdown switch indicator crash Missile gen- L Buildings light on local

other natural erated from E Seismic Monitoring or destructive the turbine R Panel OH51-P021. phenomena failure result- T ing in either:

RED light on Seismic Monitoring

damage to Panel OH13-P969. Safe Shut-down equipment

penetration of a Safe Shutdown Bldg.

Applicable Modes:

1 2 3 4 5 D

Category M: Release of Toxic or Flammable Gases Initiating Conditions Entry Criteria Toxic or flammable gas concentrations Report by local, county, or State MU1 detected within the Protected Area. officials for a potential evacuation of site personnel based on an offsite event. U MU1 N U

Release of toxic OR S flammable gases U affecting the Protected A Area boundary deemed L detrimental to the safe operation of the E plant Normal operation of the plant is impeded due V to access restrictions. E N

Applicable Modes: T 1 2 3 4 5 D NOTE A toxic or flammable gas release is considered to be impeding normal operations due to access restrictions if it is of sufficient magnitude that access to areas normally accessed to plant operator rounds is restricted. It also includes releases where access to these areas is possible only through the use of protective equipment, such as respirators since this limits the operators visibility and mobility thereby affecting normal plant operations.

Initiating Conditions Entry Criteria MA1 MA1 Entry of toxic or flammable gases into Safe Shutdown Buildings or Areas.

Release of toxic OR flammable gases within Toxic gas in concentrations Flammable gas estimated or Plant personnel NOT a Safe Shutdown considered life-threatening determined to be in explosive able to perform actions A Building which concentrations necessary to establish L jeopardizes operation of and maintain Mode 4 E systems required to while utilizing R maintain safe appropriate protective T operations OR to equipment.

establish or maintain COLD SHUTDOWN Applicable Modes:

1 2 3 4 5 D NOTE This IC addresses increased toxic or flammable gas levels that impede necessary access to operating stations or other areas containing equipment that must be operated manually in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant.

Category N: Security Events Initiating Conditions Entry Criteria NU1 U

NU1 N U

Confirmed SECURITY SECURITY CONDITION A credible site specific A validated notification S CONDITION OR that does NOT involve a security threat from NRC providing U threat which indicates a HOSTILE ACTION as notification. information of an aircraft A potential degradation in reported by the Security threat. L the level of safety of the Shift Supervisor.

plant E V

E N

T Applicable Modes:

1 2 3 4 5 D

(INTENTIONALLY BLANK)

Initiating Conditions Entry Criteria NA2 NA2 A

Notification of an A validated notification from the NRC of an airliner attack threat less than 30 minutes L Airborne Attack Threat away. E R

T Applicable Modes:

1 2 3 4 5 D NOTE Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant. The status and size of the plane may be provided by NORAD through the NRC.

Initiating Conditions Entry Criteria NA3 NA3 A

HOSTILE ACTION A HOSTILE ACTION is occurring or has occurred within the OWNER L within the Owner CONTROLLED AREA as reported by the Security Shift Supervisor. E Controlled Area R T

Applicable Modes:

1 2 3 4 5 D NOTES

HOSTILE ACTION is defined as an act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area).

(INTENTIONALLY BLANK)

Initiating Conditions Entry Criteria NS2 S

I T

E NS2 A

HOSTILE ACTION A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as R within the reported by the Security Shift Supervisor. E A

PROTECTED AREA E

M E

R G

E N

C Applicable Modes: Y 1 2 3 4 5 D NOTES

HOSTILE ACTION is defined as an act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area).

Initiating Conditions Entry Criteria NG1 G

E N

E NG1 R A HOSTILE ACTION has occurred such A HOSTILE ACTION has caused failure A HOSTILE ACTION L resulting in loss of that plant personnel are unable to operate of spent fuel cooling systems and physical control of the equipment required to maintain safety imminent fuel damage is likely for freshly functions. off-loaded reactor core in the pool. E facility M E

R G

E N

C Y

Applicable Modes:

1 2 3 4 5 D NOTES

HOSTILE ACTION is defined as an act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area).

(INTENTIONALLY BLANK)

Category O: Emergency Coordinators Judgment Initiating Conditions Entry Criteria OU1 OU1 U N

Other conditions Events are in process or have occurred which indicate a potential degradation of the U existing, which in the level of safety of the plant. S judgment of the U Emergency A Coordinator, warrant L declaration of an Unusual Event E V

E N

Applicable Modes: T 1 2 3 4 5 D NOTE For those cases where the degradation in the level of safety of the plant is tied to equipment or system malfunctions, the decision that the component is degraded should be based upon its functionality and NOT its operability.

Initiating Conditions Entry Criteria OA1 OA1 Other conditions Events are in progress or have occurred which indicate an actual or potential existing, which in the degradation of systems needed for the protection of the public and which warrant judgment of the increased monitoring of plant functions. A Emergency L Coordinator, warrant E declaration of an Alert R T

Applicable Modes:

1 2 3 4 5 D NOTE This IC is intended to address unanticipated conditions NOT addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the Alert emergency class. This includes a determination by the Emergency Coordinator that additional assistance similar to that provided by the TSC and OSC staffs, including a transfer of the Emergency Coordinator responsibilities to the TSC, is necessary for the event to be effectively mitigated. Transfer of Emergency Coordinator duties for classification, offsite notifications and PAR decisions, is used as an initiator since an event significant enough to warrant transfer of command and control is a substantial reduction in the level of safety of the plant. It is not intended to declare an Alert if Emergency Coordinator responsibilities are transferred out of the Control Room for convenience.

Initiating Conditions Entry Criteria OS1 S

I T

E OS1 A

Other conditions Other conditions exist which indicate an actual or likely major failure of plant R existing, which in the functions needed for protection of the public. E judgment of the A Emergency Coordinator, warrant E declaration of a Site M Area Emergency E R

G E

N C

Applicable Modes: Y 1 2 3 4 5 D

Initiating Conditions Entry Criteria OG1 G

E OG1 N E

Other conditions Other conditions exist which indicate an Potential for an Potential for an R existing, which in the actual or imminent substantial core uncontrolled release uncontrolled release A judgment of the degradation with the potential loss of which can which can L Emergency Containment integrity. reasonably be reasonably be Coordinator, warrant expected to be expected to be E declaration of a General greater than 1 Rem greater than 5 Rem M Emergency TEDE at the Site CDE Child Thyroid E Boundary. at the Site R Boundary. G E

N C

Applicable Modes: Y 1 2 3 4 5 D

SRO Question 10 SRO Question 13 8

NO YES 5.9 YES 5.4

SRO Question 17 SRO Question 17 SRO Question 17 SRO Question 17 SRO Question 18 SRO Question 18 SRO Question 19 PMT Data

1. Suction Valve Post Maintenance Tests: 1E51-F010 (EO, EC, ST),

1E51-F031 (EO, EC, ST).

a. Close 1E51-F010, RCIC PUMP CST SUCTION VALVE, measure stroke time and record the following:

1) Exercise Closed Results:

SAT UNSAT

2) Stroke Time - Closed direction.

__24.8___ seconds Reference Value = 25.0 seconds Acceptable Range = 21.3 to 28.7 seconds SAT UNSAT

b. Open 1E51-F031, RCIC PUMP SUPR PL SUCT ISOL, measure stroke time and record the following:

1) Exercise Open Results:

SAT UNSAT

2) Stroke Time - Open direction.

__19.5___ seconds Reference Value = 16.9 seconds Acceptable Range = 14.4 to 19.4 seconds SAT UNSAT

c. Close 1E51-F031, RCIC PUMP SUPR PL SUCT ISOL, measure stroke time, and record the following:

1) Exercise Closed Results:

SAT UNSAT

2) Stroke Time - Closed direction.

__18.9 __ seconds Reference Value = 17.0 seconds Acceptable Range = 14.5 to 19.5 seconds SAT UNSAT

SRO Question 19 PMT Data

d. Open 1E51-F010, RCIC PUMP CST SUCTION VALVE, measure stroke time, and record the following:

1) Exercise Open Results:

SAT UNSAT

2) Stroke Time - Open direction.

__23.8 __ seconds Reference Value = 24.2 seconds Acceptable Range = 20.6 to 27.8 seconds SAT UNSAT

SRO Question 20 SRO Question 20 SRO Question 20 SRO Question 20 SRO Question 20

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