ML12278A324

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

ML12278A324
Person / Time
Site:	Calvert Cliffs
Issue date:	09/12/2012
From:	Moore R Constellation Energy Nuclear Group
To:	D'Antonio J Operations Branch I
Jackson D
Shared Package
ML12067A049	List: IR 05000317/2012301 ML12102A010 ML12216A019 ML12277A367 ML12277A369 ML12277A371 ML12277A382 ML12278A020 ML12278A305 ML12278A324 ML12278A404 ML12348A184 ML15043A710
References
U01849
Download: ML12278A324 (245)

Text

1 b 34 d 67 a 2 b 35 a 68 b 3 c 36 deleted 69 c 4 a 37 a 70 a 5 d 38 b 71 b 6 d 39 c 72 d 7 a 40 c 73 c 8 d 41 d 74 d 9 b 42 a 75 a 10 c 43 a 76 b 11 d 44 d 77 d 12 b 45 b 78 c 13 b 46 d 79 b 14 a 47 b 80 a 15 d 48 a or b 81 c 16 a 49 b 82 c 17 c 50 d 83 b 18 c 51 c 84 d 19 b 52 a 85 d 20 c 53 d 86 b or c 21 d 54 c 87 a 22 a 55 d 88 a 23 c 56 c 89 d 24 a 57 c 90 b 25 b 58 b 91 b 26 d 59 a 92 c 27 c 60 c 93 a 28 c 61 b 94 c 29 d 62 c 95 a 30 a 63 a 96 c 31 c 64 b 97 a 32 d 65 a 98 d 33 c 66 d 99 b 100 d

CALVERTCLIFFS NUCLEAR POWER PLANT 2012 NRC INITIAL LICENSED OPERATOR SRO WRITTEN EXAM KEY Page 1 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Page 2 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Page 3 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit-1 is performing a reactor startup at 300 MWD/MTU. Critical data has been recorded and reactor power stabilized at the POAH with Group 4 CEAs at 90 inches.

The TBV controller, 1-PIC-4056, output signal fails to 10% in automatic resulting in a plant cooldown. The RO monitoring the reactor reports the following:

• Reactor power is below 1OE-1 % and continuing to lower

• SUR is negative

• RCS T COLD is 530 of and lowering slowly As the CRS, which ONE of the following actions would you direct the crew to perform?

A. Withdraw Regulating Group CEAs to restore RCS T COLD.

B. Trip the reactor and implement EOP-O.

C. Place the TBV controller in manual at 0% output.

D. Fully insert Regulating Group 4 CEAs in manual sequential.

Answer: B Page 4 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Answer Explanation:

A. Incorrect - OP-2 states the following precaution: Primary plant anomalies caused by secondary plant transients are rarely, if ever, successfully mitigated by adding positive reactivity, especially by withdrawing CEAs. Do NOT use CEAs to control RCS temperature without an approved procedure.

Events have occurred in the industry where CEAs have been withdrawn to reestablish critical conditions. Conditions indicate the reactor has gone subcritical and AOP-7K Section IV Actions require a reactor trip and implement EOP-O.

B. Correct - Per AOP-7K, which is entered due to overcooling event and plant is in MODE 2, this is the correct action based on reactor conditions provided.

C. Incorrect - Although this is part of the recovery action to restore from overcooling event, conditions indicate the reactor has gone subcritical and AOP-7K Section IV Actions require a reactor trip and implement EOP-O.

D. Incorrect - OP-2 directs with conditions of reactor above, to FULLY insert ALL regulating CEAs not just Group 4. However, an overcooling event has occurred and actions of AOP-7K are required and operators will trip the reactor and implement EOP-O.

Page 5 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY

• Actions required when Rx goes subcritical from overcoolinn event in Mode 2 Tier/Group: 3

• 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.11 _. Knowledge of abnormal condition procedures.

SRO Importance: 4.2 Proposed references to be None provided to applicant:

Given an overcooling event in progress, determine and Learning Objective: implement the applicable actions to mitigate the event per plant operating procedures.

10 CFR Part 55 Content: 55.43(b)(5}

Question source:

Cognitive level: r:g] Comprehension/Analysis Last NRC Exam used on:

Exam Bank History:

AOP-7K, Overcooling Event in Modes 1 and 2; Technical references:

Startup from Hot Standby to Minimum Load None Page 6 of 61

2012 NRC SRO EXAM MASTER KEY During a Steam Generator Tube Rupture event, Pressurizer level is maintained or lowered to between 101 to 120 inches if backfill of the RCS is anticipated.

Which ONE of the following describes the reason for limiting the Pressurizer level to a maximum of 120 inches?

A. Minimizes the loss of primary fluid to the secondary.

B. Minimizes the potential for a Pressurized Thermal Shock Event.

C. Ensures RCS Pressure and Inventory control is established.

D. Allows additional inventory to be added to the RCS with minimal impact.

Answer: D Answer Explanation:

A. Incorrect - This is the basis for maintain subcooling at the low end of the band.

B. Incorrect - The basis for throttling minimizes the potential for a Pressurized Thermal Shock Event.

C. Incorrect - This is a basis for minimum Pressurizer level (101 inches) in conjunction with minimum subcooling. "If backflow from the affected S/G is anticipated, maintaining a lower pressurizer level will allow additional inventory to be added to the RCS with minimal impact."

D. Correct - As stated in the EOP-6 Technical Basis document, "If backflow from the affected S/G is anticipated, maintaining a lower pressurizer level will allow additional inventory to be added to the RCS with minimal impact".

Page 7 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: EOP-6 Technical Basis I Tier/Group: 1/2 037 - Steam Generator Tube Leak KIA Info: 2.4 - Emergency Procedures / Plan 2.4.18 - Knowledge of specific bases for EOPs.

SRO Importance: 4.0 Proposed references to be

. None

• provided to applicant:

Given EOP-6, determine the basis for maintaining PZR Learning Objective: level band prior to backfill into the ReS.

10 CFR Part 55 Content:

• Cognitive level: C8J Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any exam LOI 2006-2 Remediation EOP/AOP Basis exam (10/08)

Technical references: EOP-6 Step I. 4 and Technical Bases Comments: None Page 8 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following conditions on Unit-1:

• Core Reload is in progress per FH-305, Core Alterations

• The Refueling Machine Operator is inserting a new fuel assembly into the core with current hoist readout at 200 inches

• Refueling Control Room Operator reports an unexpected increase in count rate on two of the four wide range NI channels

• Audible count rate in the containment is rising As the Fuel Handling Supervisor, which ONE of the following actions should be completed first?

A. Notify the Shift Manager immediately.

B. Observe behavior of the affected Nls.

C. Withdraw the assembly from the core.

D. Stop insertion and allow counts to stabilize.

Answer: C Answer Explanation:

A. Incorrect - These are the actions per FH-305 for a sustained rising count rate, on two or more Nls, after an assembly has been inserted.

B. Incorrect - This is a partial action per FH-305 being taken for a single wide range NI channel that may be unreliable. Question stem states 2 of 4 channels have increased unexpectedly.

C. Correct - This is the proper action to take as stated in FH-305 for an unexpected increase in count rate on more than one wide range NI channel.

D. Incorrect - Stopping insertion is prudent but FH-305 requires that fuel assembly be withdrawn.

Page 9 of 61

2012 NRC SRO EXAM MASTER KEY Topic: Inadvertent dilution during Core Alts Tier/Group: 2/2

! 034 - Fuel Handling

• K 1 - Knowledge of the physical connections and/or KIA Info: cause-effect relationships between the Fuel Handling System and the following systems:

• K1.04-NIS SRO Importance: 3.5 Proposed references to be None provided to applicant:

Determine the proper location for a fuel assembly during Learning Objective:

an Inadvertent Dilution in Modes 3, 4, 5 or 6.

10 CFR Part 55 Content: 55.43(b)(7)

Cognitive level:

~ Comprehension or Analysis Last NRC Exam used on: No record of use on any exam

• Exam Bank History: LOR 11**6F Biennial written exam (12/12)

Technical references: FH-305, Core Alterations Comments: None 10 of 61

2012 NRC SRO EXAM MASTER KEY Which ONE of the following conditions challenges the Core and RCS Heat Removal safety function during EOP-O and which Optimal Recovery procedure should be entered?

A. 1-CVC-506-CV (RCP Bleed-Off Inboard Isol) fails closed due to a broken airline; EOP-1, Reactor Trip.

B. Unable to start any Component Cooling Pump; EOP-2, Loss Of Offsite Power/Loss Of Forced Circulation.

C. 11A RCP middle seal and vapor seal failed and 11A RCP was secured; EOP-5, Loss of Coolant Accident.

D. RCS pressure lowers to 1350 PSIA with containment parameters normal; EOP-6, Steam Generator Tube Rupture.

Answer: B Answer Explanation:

A. Incorrect - Inboard RCP bleed-off isolation failing closed. No requirement to secure ALL RCPs as bleed-off RV lifts in containment to maintain a flowpath with RCPs operating. To enter EOP-1, ALL safety functions are complete (met).

Core and RCS Heat Removal would be met as at least one RCP is operating in a loop with a S/G available.

B. Correct - Per EOP-O Vital Auxiliaries if unable to start a CCW pump all RCPs must be secured. Core and RCS Heat Removal requires at least one RCP operating in a loop with a S/G available for heat removal and NO RCPs would be operating.

C. Incorrect - Two RCPs are secured due to trip strategy in EOP-O but Core and RCS Heat Removal per EOP-O is complete (met) as at least one RCP is operating in a loop with a S/G available. A loss of the vapor seal results in an RCS leak to the containment.

D. Incorrect - Core and RCS Heat Removal would be met as at least one RCP is operating in a loop with a S/G available. Two RCPs would be tripped based on SIAS actuation. HPSI Pumps are not injecting flow into the RCS so a cooldown is not occurring at this pressure value.

Page 11 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: RCP Malfunctions Tier/Group: 1/1 015/017 - Rep Malfunctions

• 2.4 - Emergency Procedures / Plan

• 2.4.21 - Knowledge of the parameters and logic used KIA Info:

to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

SRO Importance: 4.6 Proposed references to None be provided to applicant:

Given plant conditions, assess the status of Core and RCS Learning Objective:

Heat Removal safety function.

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: D Memory/Fundamental Last NRC Exam used on: New question Exam Bank History: None Technical references: EOP-O and EOP-O Diagnostic Flowchart 1C07-ALM, Chemical & Volume Control Alarm Manual, window F-07 Comments: None Page 12 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Using Provided Reference(s):

Given the following condition:

• Unit-1 reactor was manually tripped as directed by AOP in use and the immediate actions of EOP-O have been completed.

Which ONE of the following optimal recovery procedure recommendations would you make to the Shift Manager?

A. EOP-6, Steam Generator Tube Rupture B. EOP-5, Loss of Coolant Accident C. EOP-4, Excess Steam Demand D. EOP-2, Loss of Offsite Power/Loss of Forced Circulation Answer: A A. Correct - Based on S/G level trends with containment pressure normal, both S/G pressures normal, and RCS pressure and level trends EOP-6 is appropriate procedure to recommend.

B. Incorrect - Although SIAS has actuated, the Containment pressure is normal and S/G levels are mismatched and trends are different which when diagnosed EOP-5 would NOT be recommended.

C. Incorrect - Although SIAS has actuated, S/G pressures are at 850 PSIA and stable thus indicating EOP-4 would NOT be recommended.

D. Incorrect -There are indications to support a SGTR and EOP-6 addresses a SGTR coincident with a loss of offsite power. RCS pressure and level trends along with S/G level trends support a SGTR is occurring. EOP-2 would not be recommended as the RCPs are ON using the Pressure/Inventory and Core/RCS Heat Removal pages provided indicating forced circulation.

Page 13 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Assessment of SGTR using SPDS Tier/Group: 1/1 038 -Steam Generator Tube Rupture 2.1 - Conduct of Operations KIA Info:

• 2.1.19 - Ability to use plant computers to evaluate system or component status.

SRO Importance: 3.8 Proposed references to be None provided to applicant:

Using SPDS assess EOP-O Safety Function status and Learning Objective: using the EOP-O Diagnostic flowchart, determine the appropriate (optimal or functional) EOP to enter.

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

D Memory or Fundamental Cognitive level:

[g] Comprehension or Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2008 Plant Computer, SPDS (01/09)

EOP-O Safety Function Status Checks and Diagnostic Flowchart None Page 14 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following conditions on Unit 2:

• Reactor power is 100%.

• A loss of Instrument Bus 2Y02 has occurred.

(1) Which ONE of the following component responses is observed and (2) What actions would you direct as the Unit CRS?

A. (1) ONLY Two (2) TCBs open; (2) Refer to alarm manual to determine cause and required corrective actions.

B. (1) ONLY Two (2) TCBs open; (2) Implement EOP-O, Post-Trip Immediate Actions.

C. (1) ONLY Four (4) TCBs open and RPS Channel B is deenergized; (2) De-energize RPS Channel B in preparation for power restoration.

D. (1 ) ONLY Four (4) TCBs open and RPS Channel B is deenergized; (2) De-energize Actuation Logic Cabinet BL and Sensor Cabinet ZD for power restoration.

Answer: C Answer Explanation:

A. Incorrect - Loss of a single 120V AC Vital instrument bus opens 4 TCBs.

Referring to the alarm manual would be appropriate for any TCBs opening.

B. Incorrect - Loss of a single 120V AC Vital instrument bus opens 4 TCBs (two trip paths de-energize) but does not trip the reactor.

C. Correct - This is response observed in the control room. Alarm manual would be referenced as part of crew response directing them to AOP-7 J which provides direction for de-energizing the RPS channel.

D. Incorrect - This is response observed in the control room. Logic cabinet referenced is correct. However, sensor cabinet ZD is powered from 2Y01.

Page 15 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic:

Tier/Group: 1/1 057 - Loss of Vital AC Inst. Bus

• • AA2 - Ability to determine and interpret the following as

. KIA Info: they apply to the Loss of Vital AC Instrument Bus:

• AA2.03 - RPS panel alarm annunciators and trip indicators SRO Importance: 3.9 Proposed references to None be provided to applicant:

• Recall the expected response of RPS upon a loss of a 120V Learning Objective: Vital AC Instrument Bus with respect to final condition of Trip Path Relays and TCBs.

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: cg] Memory/Fundamental Last NRC Exam used on: No record of use Exam Bank History: None Technical references: Loss Of 120 Volt Vital AC or 125 Volt Vital DC Power Comments: Modified from Q20182 Page 16 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit-1 was operating at 100% power when a plant transient caused a reactor trip.

EOP-O, Post-Trip Immediate Actions, was implemented and the following conditions were observed:

• CEA #1 indicates fully withdrawn

• Amber lights are energized for all other CEAs except CEA # 52, whose green light is energized

• 11 S/G Pressure at 920 PSIA

• 12 S/G Pressure at 800 PSIA

• 11 S/G level at (-)115 inches

• 12 S/G level at (-)165 inches

• Condenser vacuum at 19.5 inches Hg

• Containment pressure at 0.8 PSIG

• No automatic safety system actuations have occurred Which action is the first required by the operating crew (assuming standard safety function hierarchy is used) and which EOP-O, Post-Trip Immediate Actions, block step would direct this action?

A Shut the MSIVs as directed by "Ensure Turbine Trip".

B. Borate the RCS to 2300 PPM as directed by "Verify the Reactivity Control Safety Function is Satisfied".

C. Start an AFW Pump as directed by "Verify the Core and RCS Heat Removal Safety function is satisfied".

D. Place all Containment Air Coolers (CACs) in pull-to-Iow and open the Emergency Outlet valves for the operating CACs as directed by "Verify the Containment Environment Safety Function is Satisfied".

Answer: C 17 of 61

2012 NRC SRO EXAM rvlASTER KEY Answer Explanation:

A. Incorrect - "Verify the Core and RCS Heat Removal Safety function is Satisfied" provides direction to shut the MSIVs should S/G pressure drop to 800 PSIA.

"Ensure Turbine Trip" does provide guidance to shut the MSIVs, but the guidance is based on turbine valve failures, turbine speed and loss of power effects.

S. Incorrect - Soration of the RCS is required only if "more than one CEA is not fully inserted". The EOP-O basis document states "A CEA is considered fully inserted if the rod drop light (amber) or the lower electrical limit light (green) is energized.

C. Correct - Main Feedwater flow has been lost due to the SGFPs tripping on low condenser vacuum and is directed by "Verify the Core and RCS Heat Removal Safety function is satisfied".

D. Incorrect - The "Verify the Containment Environment Safety Function is Satisfied" does not direct placing the CACs in pull-to low.

2012 NRC SRO EXAM MASTER KEY EOP-O, Post-Trip Immediate Actions, hierarchy to initiate Topic:

AFW Tier/Group: 1/1 054 - Loss of Main Feedwater

• AA2 - Ability to determine and interpret the following as KiA Info: they apply to the Loss of Main Feedwater (MFW):

• AA2.03 - Conditions and reasons for AFW startup SRO Importance: 4.2 I Proposed references to None be provided to applicant.

Learning Objective:

10 CFR Part 55 Content:

Question source:

Cognitive level: ~ Comprehension/Analysis Last NRC Exam used on: New question Page 19 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Following a plant trip from 100% power, Pressurizer (pzr) level lowered to 90 inches before recovering. The crew implemented EOP-1, Reactor Trip, as ALL safety functions were met.

The Pzr level final acceptance criteria in EOP-1, Reactor Trip, has an operating band of 130 to 180 inches and trending to 160 inches.

Which ONE of the following choices below is (1) the basis for this operating band and (2) an administrative post-trip action requirement?

A. Allows some tolerance from the normal band assuming a standard reactor trip with charging and letdown isolated; Entry into the T. S. LCO for the Pzr being inoperable because two emergency banks of Pzr heaters deenergized when Pzr level fell below 101 inches.

B. Actual level outside this band means it is challenging the Pressure and Inventory Control safety function; Entry into the T. S. LCO for the Pzr being inoperable because Pzr level fell below the minimum operating level, following the trip.

C. Allows some tolerance from the normal band assuming a standard reactor trip with charging and letdown remaining in service; Recording backup Charging Pump(s) start/stop times per EN-1-115, Recording of Plant Transients/Operational Cycles.

D. Ensures that pressurizer heaters remain covered and allows a band of

(+) or (-) 25 inches from programmed pressurizer level; Recording occurrence of the reactor trip per EN-1-115, Recording of Plant Transients/Operational Cycles.

Answer: B Page 20 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Answer Explanation:

A. Incorrect - Isolation of charging and letdown in EOP-1, indicate something more than a standard reactor trip has occurred; Although the heaters are deenergized when level is below 101 inches (an interlock), the LCO for Pzr being inoperable based on emergency heaters is not entered as power remained to emergency heater banks defined in tech specs during this E!Vent.

B. Correct - This is per EOP-1 basis Step IV.D; EOP Att. 13 states ensure any LCOs that have NOT been met during the event are entered AND all appropriate log entries have been made. Pzr level went below minimum operating level of 133 inches per LCO 3.4.9 and this entry is required.

C. Incorrect - Actual level outside this band means it is challenging the Pressure and Inventory Control safety function. The backup Charging pump(s) will operate to return Pzr level to the specified band. Although charging pumps start and stop, no entries per EN-1-115 are required since charging was never lost based on stem statement that all safety functions were met.

D. Incorrect - Programmed level at 0% power is 160 inches, so high limit is only

(+) 20 inches but lower limit is (-) 30 inches from program. The reactor trip transient log entry is required per EN-1-115.

Page 21 of 61 Rev. 3

2012 NRC SRO EXAM rvlASTER KEY Basis for Pzr level in EOP-1, Reactor Trip and admin Post Topic:

trip actions Tier/Group: 1/1 ICE E02 - Reactor Trip Recovery KIA Info:

• 2.2.42 - Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

SRO Importance: 4.6 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.43(b)(2)

• Cognitive level: D Memory/Fundamental Last NRC Exam used on: New question None Tech Spec LCO 3.4.9 Pressurizer EOP-1, Reactor Trip and Technical Bases

• Technical references: EOP Att. 13: AdministrativE~ Post-Trip Actions EN-1-115, Recording Of P!lant Transients/Operational Cycles ments: . None Page 22 of 61 Rev. 3

2012 NRC SRO EXAM MIASTER KEY With Unit-1 at 100% power, TSV-3942 failed open resulting in a reactor trip. EOP-O, Post-Trip Immediate Actions, was implemented and alternate actions taken.

Given the following conditions in EOP-O:

• RCS Soration in progress due to loss of power E~ffects (LOPE)

• 11 4KV Bus is energized from offsite

• 14 4KV Bus is faulted

• All 125VDC bus voltages indicate 124 VDC

• Radiation Levels External to Containment (RLEC) alternate actions were taken due to loss of power effects

• PRZR pressure is 1950 PSIA and slowly lowering

• PRZR level is 70 inches and slowly lowering

• TcoLD is 516°F and slowly lowering

• 11 S/G pressure is 780 PSIA and continues to lower

• 12 S/G pressure is 880 PSIA and slowly rising

• 13 AFW pump is operating to restore S/G levels

• 11 S/G level is minus (-) 150 inches and lowering

• 12 S/G level is minus (-) 110 inches and rising

• Containment pressure is 1.5 PSIG and rising

• Containment temperature is 140°F and rising

• Containment RMS is unchanged Which ONE of the following will be implemented based on plant parameters and conditions?

A. EOP-8, Functional Recovery Procedure B. EOP-6, Steam Generator Tube Rupture.

C. EOP-5, Loss of Coolant Accident D. EOP-4, Excess Steam Demand Event Answer: D Page 23 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Answer Explanation:

A. Incorrect - Based on T COLD and S/G pressure/levels lowering an ESDE is occurring. There is only one event occurring so EOP-8 is not required to be entered. Plausible based on multiple degraded parameters.

B. Incorrect - Based on T COLD and S/G pressure/levels lowering an ESDE is occurring. A SGTR can be eliminated based on S/G level and pressure responses. Plausible based on Pzr pressure and rising SG level.

C. Incorrect - Based on T COLD and S/G pressure/I!vels lowering an ESDE is occurring. A LOCA can be eliminated based on containment RMS response.

Plausible based on Pzr pressure and level and containment parameters.

D. Correct - Based on TCOLD and S/G pressures lowering an ESDE is occurring.

LOCA and SGTR can be eliminated based on S/G level and pressure responses.

Page 24 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: EOP-4 Excess Steam Demand Tier/Group: 1/1 CE/E05 Excess Steam Demand

• EA2 - Ability to determine and interpret the following as they apply to the (Excess Steam Demand)

KIA Info:

• EA2.1 - Facility conditions and selection of appropriate procedures during abnormal and emergency operations SRO Importance: 4.0 Proposed references to None be provided to applicant:

Given plant conditions and/or parameters, determine which Learning Objective: optimal recovery procedure is the correct one for the condition/parameters given.

10 CFR Part 55 Content: 55.43(b)(5)

Cognitive level: [2J Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2008 AOP/EOP exam (04/10)

Technical references: EOP-O Diagnostic Flowchart and EOP-4, Excess Steam Demand Event entry conditions Comments: None Page 25 of 61 Rev. 3

2012 NRC SRO EXAM rvlASTER KEY Unit-2 is operating at 60% power when a loss of 4KV Bus 22 occurs.

(1) What effect does this condition have on plant operation?

(2) What is the correct action to address this condition?

A. (1) Loss of 22 and 23 Condensate Pumps; (2) Commence a Rapid Power Reduction, to lower Condensate Header flow to less than 8000 GPM.

B. (1) Loss of lube oil to both SGFPs; (2) Trip the Reactor, implement EOP-O.

C. (1) Loss of 21 and 22 Condensate Booster Pumps; (2) Trip the Reactor, implement EOP-O.

D. (1) Loss of 21 and 22 Condensate Booster Pumps; (2) Commence a Rapid Power Reduction, to lower Condensate Header flow to less than 8500 GPM.

Answer: D Answer Justifications:

A. Incorrect - 22 and 23 Condensate Pps are powered from 4KV Bus 23 and remain in operation. Stated actions would be correct for a loss of 4KV Bus 23.

B. Incorrect - Each SGFP has an Oil Pp powered from MCC-206 and one powered from MCC-216; therefore lube oil will not be lost with a loss of MCC-206 (22 4KV bus).

C. Incorrect - The listed loads are in fact lost. Tripping the Reactor and implementation of EOP-O would be correct actions if Reactor power were greater than 70%.

D. Correct - 21 and 22 Condensate Booster Pps are lost necessitating a power reduction to get Condensate Header flow to less than the capacity of a single Condensate Booster Pp.

Page 26 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: Loss of 22 4KV Bus effects Tier/Group: 3 2.1 - Conduct of Operations KIA Info: 2.1.20 - Ability to interpret and execute procedure steps.

SRO Importance: 4.6 Proposed references to None be provided to applicant Learning Objective:

Question source:

Cognitive level: D Memory/Fundamental [gJ Comprehension/Analysis Last NRC Exam used on: N/A AOP-71-2, Loss Of 4kv, 480 Volt Or 208/120 Volt Instrument Bus Power Comments:

Page 27 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Using provided reference(s):

Unit-2 enters AOP-2A due to an RCS leak which required a reactor trip.

Prior to the trip RCS boron was 813 PPM Given the following post-trip conditions:

• Both BAST concentrations are 7.25%

• CEAs 38 and 46 are stuck at 120 inches withdrawn

• The Pressurizer emptied in EOP-O

• SIAS was verified in EOP-O

• RCS pressure is 1380 PSIA and continuing to lower

• The Crew transitioned to the appropriate Optimal Recovery Procedure fifteen (15) minutes after entering EOP-O Fifteen (15) minutes after requested, Plant Chemistry reports the RCS boron sample result is 1100 ppm.

Which ONE of the following represents: (1) The status of the boron concentration for Shutdown Margin (SOM) and (2) The required action for the existing plant conditions?

A. (1) Present boron concentration meets required SOM; (2) Align Charging pump suction to the RWT.

B. (1) Present boron concentration is below required SOM; (2) Borate until BAST volume or Charging Pp run time requirement is met.

C. (1) Present boron concentration is below required SOM; (2) Borate until SOM requirement for both EOP-O and the optimal EOP is met.

O. (1) Present boron concentration meets required SOM; (2) Align Charging pump suction to VCT after SIAS has been reset.

Answer: C Page 28 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Answer Explanation:

A. Incorrect - The status of boron concentration is incorrect. However, based on question stem two CEAs are stuck out and NEOP-23 Fig. 2-11-A.5 requires a boron concentration of ~ 2300 PPM.

B. Incorrect - The present boron concentration does not meet the requirement of SDM for two stuck CEAs. Using Fig. 1 of 01-2B determines gallons of boric acid needed to reach 2300 PPM are 8,812. Applying IEOP-5 requirements for BAST volume or charging pump run times adds the following:

• 134 inches X 58.8 gallons / inch (Fig. 2 of 01-2C provided) =7879 gallons

• =

60 minutes X 132 gallons/minute 7920 gallons Second part is plausible if examinee falls to recognize that two stuck CEAs requires ~ 2300 PPM for SDM.

C. Correct - Boration during the LOCA must continue until boron concentration is ~

2300 PPM per NEOP-23 Fig. 2-II-A-5 for two stuck CEAs. Using Fig. 1 of 01-2B determines gallons of boric acid needed to reach 2300 PPM are 8,812. Applying EOP-5 requirements for BAST volume or charging pump run times adds the following:

• 134 inches X 58.8 gallons / inch (Fig. 2 of OI-2C provided) = 7879 gallons

• 60 minutes X 132 gallons/minute = 7920 gallons D. Incorrect - EOP-5 does not direct continue to borate until SIAS is verified and reset. There are specific criteria to meet required SDM for 2 stuck CEAs. If SIAS is verified and reset, one of the paths to realign charging pump suction to is the VCT.

Page 29 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Question 86 (Q97063)

Topic: SDM requirement for SGTR and two stuck CEAs Tier/Group: 1/2 024 - Emergency Boration KIA Info:

• AA2 - Ability to determine and interpret the following as they apply to the Emergency Boration:

• AA2.05 - Amount of boron to add to achieve required SDM SRO Importance: 3.9 01-2B, Figure 1(Boration Volume (RCS Not On SDC)

Proposed references to be 01-2C, Figure 2 Boric Acid Storage Tank volume provided to applicant: EOP-5, Step IV.H. Commence RCS Boration NEOP-23, Figures 2-1 I. A. 1 & 2-11.A.3 Given plant parameters, identify the appropriate response for Learning Objective:

Loss of Coolant Accident (LOCA) per EOP-5.

10 CFR Part 55 Content: 55.43(b)(5)

Cognitive level: o Memory/Fundamental Last NRC Exam used on: No record of use Technical references: NEOP-23, Figs. 2-1 I. A. 1, Soluble Boron Concentration versus burnup NEOP-23, Figs. 2-11.A.3: Shutdown Boron Concentration for All Rods In NEOP-23 Fig. 2-11.A.5: Shutdown Boron Concentration for More than One CEA Stuck EOP-5 Step H and Technical Bases Q25815 Page 30 of 61

2012 NRC SRO EXAM MASTER KEY Unit-1 is at 75% power with TAVE at 558 of when 12 Hot Leg RTD, TE-121X, fails high.

Reactor Regulating System (RRS) channel selector switch, 1-HS-5600, is selected to RRS-X.

Which ONE of the following (1) describes the impact of the instrument failure on the Pressurizer (pzr) level control system and (2) is the direction provided to the RO?

A. (1) Pzr level setpoint increases, all Charging Pumps start, letdown flow goes to minimum; (2) Place the appropriate (S1 or S2) switch to off on RRS channel X and Y.

B. (1) Pzr level setpoint decreases, selected Charging Pump remains in operation, letdown flow goes to maximum; (2) Use 01-7, Reactor Regulating System, to determine failed TE actions.

C. (1) Pzr level setpoint decreases, selected Charging Pump remains in operation, letdown flow goes to maximum; (2) Place the appropriate (S 1 or S2) switch to off on RRS channel X.

D. (1) Pzr level setpoint increases, all Charging Pumps start, letdown flow goes to minimum; (2) Place RRS channel selector switch, 1-HS-5600, to RRS-Y position.

Answer: A Answer Explanation:

A. Correct - The Pzr level setpoint is generated from a TAVE signal between 30 and 95% power. At 75% TAVE is -558 OF. The failed TE causes TAVE to fail to its maximum value. This results in the Pzr level control system sending a signal to start all charging pumps and reduce UD to minimum. It is necessary to place the S2 switch in both RRS channels to off to remove failed TE input.

B. Incorrect - As stated above, Pzr level setpoint increases; 01-7 is the correct procedure to reference per the alarm manual response.

C. Incorrect - Setpoint does not lower and placing S1 or S2 switch to off in Channel X only does not remove failed input that still exists in channel Y.

D. Incorrect - This is the correct response to TE failing high. Switching to Channel Y without removing the failed TE input will not return Pzr level setpoint to the proper value.

Page 31 of 61 Rev. 3

2012 NRC SRO EXAM MIASTER KEY Topic: Failed TE inputto RRS Tier/Group: 1/2 028 - Pressurizer Level Control Malfunction AA2 - Ability to determine and interpret the following as they KIA Info: apply to the Pressurizer Level Control Malfunctions:

• AA2.08 - PZR level as a function of power level SRO Importance: 3.5 Proposed references to None be provided to applicant:

Given the following conditions, determine as an RO/CRO and/or direct as the SRO the following actions needed:

Learning Objective:

a. Pzr level response to failure of TE input to RRS and actions per 01-7, Reactor Regulating System operation.

10 CFR Part 55 Content: 55.43(b)(5)

Cognitive level: D Memory/Fundamental [2J Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: None Technical references: 01-7, Reactor Regulating System Alarm Response Manual 1C05, window D-40 Comments: Modified from Q14428 Page 32 of 61 Rev. 3

2012 NRC SRO EXAM MIASTER KEY U-2 is in a Refueling Outage and is currently being defueled. The Refueling Machine operator has just begun lowering a fuel assembly from the core into the upender. A freshly burned Fuel Assembly is in the Inspection Stand in the Spent Fuel Pool (SFP).

The Containment Outage Door (COD) is in place and is open for equipment move in.

A large truck carrying scaffold has backed into the COD and caused damage which prevents dogging the COD shut.

Which ONE of the following correctly describes the required actions?

A. Place the fuel assembly in a safe location, suspend movement of irradiated fuel assemblies within the Containment, and Install the Equipment Hatch with a minimum of 4 bolts.

B. Place the fuel assembly in a safe location, suspend movement of irradiated fuel assemblies within the Containment and the SFP, and Install the Equipment Hatch with a minimum of 4 bolts.

C. Fuel movement may continue provided the Equipment Hatch is installed with at least 4 bolts within the Time to Boil, or place the fuel assembly in a safe location and suspend movement of irradiated fuel assemblies within the Containment.

D. No actions are required if the Equipment Hatch is available to be installed in less than the Time to Boil.

Answer: A Answer justification:

A. Correct - Per AOP-4A, Loss of Containment Closure and T.S. 3.9.3.

B. Incorrect - T.S. 3.9.3 specifies "suspend movement of irradiated fuel assemblies within containment". These actions are not extended to Spent Fuel Pool activities with irradiated fuel.

C. Incorrect - The fuel assembly must be placed in a safe location and movement of irradiated fuel assemblies within containment must be suspended until the equipment hatch is installed with a minimum of a least 4 bolts.

D. Incorrect - The fuel assembly must be placed in a safe location and movement of irradiated fuel assemblies within containment must be suspended until the equipment hatch is installed with a minimum of a least 4 bolts.

Page 33 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Question 88 (Q97055)

Topic: Loss of Containment Integrity during Fuel Handling Tier/Group: Generic Knowledge and Abilities 2.1 - Conduct of Operations KIA Info:

• 2.1.35 - Knowledge of the fuel-handling responsibilities of SHOs.

SRO Importance: 3.9 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.43(b)(7)

Question source: ~ Bank Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No previous NRC Exam use Exam Bank History: None AOP-4A, Loss of Containment Integrity Technical references:

T.S. 3.9.3, Containment Penetrations Comments: None Page 34 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following conditions on Unit-1:

• The RO is performing STP-O-29, CEA Free Movement Test, when a Shutdown Group CEA cannot be withdrawn from 127.5 inches, after insertion

• Electrical Maintenance determines the CEA is mechanically stuck

• System Engineering has declared CEA untrippable Which ONE of the following actions is required based on the report from Electrical Maintenance?

A. Perform a rapid shutdown per OP-3, Appendix B, Rapid Power Reduction; upon turbine trip, borate the RCS at 2: 40 GPM of at IE!ast 2300 PPM until SOM is met.

S. Insert the remaining CEAs within the group to realign with the stuck CEA to clear the CEA Motion Inhibit (CMI) while maintaining power level.

C. If unable to realign the CEA after two hours, then trip the reactor and implement EOP-O, Post Trip Immediate Actions.

o. Shutdown and place the unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per OP-3, Normal Power Operation.

Answer: 0 Answer Explanation:

A. Incorrect - These are actions required per AOP-1B for two or more untrippable CEAs and subsequent boration required.

B. Incorrect - Examinee recognizes realigning other CEAs in group with stuck CEA will remove CEA group deviation. However, this does not clear the CMI. CIVIl remains in because Regulating Group CEAs are unable to move (MIRG) as Shutdown CEAs are less than 129 inches withdrawn.

C. Incorrect - These are the actions required when two or more CEAs are misaligned by > 15 inches within their group.

o. Correct - For a single untrippable CEA, AOP-1 B directs the plant be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, per OP-3.

Page 35 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: Actions for untrippable CEA (stuck)

Tier/Group: 1/2 005 - Inoperable/Stuck Control Rod 2.4 - Emergency Procedums / Plan KIA Info:

• 2.4.11 - Knowledge of abnormal condition procedures.

SRO Importance: 4.2 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 ContelntI55.43(b)(5)

Cognitive level: [2J Comprehension/Analysis t NRC Exam used on: New question None None Page 36 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Due to emergent equipment issues, the GS-Shift Operations directs a change be to the Safe Shutdown Summary Schedule (S4).

Per NO-1-103, Conduct of Lower Mode Operations, which ONE of the following satisfies the S4 change review requirements?

A. The Shutdown Safety Review Board.

B. The designated SRO and a second independent SRO.

C. Outage Management Outage Specialist and the GS-Shift Operations.

D. The designated SRO and the GS-Shift Operations.

Answer: B Answer justification:

A. Incorrect - The Shutdown Safety Review Board (SSRB) is comprised of one SRO, one Senior Leadership team member and a member from the PRA group or Engineering.

B. Correct - per NO-1-103, the review must be performed by an SRO appointed (designated) by the GS-SO and a second independent SRO.

C. Incorrect - per NO-1-103, the review must be performed by an SRO appointed (designated) by the GS-SO and a second independent SRO. Since the GS-SO is directing the change he would not be considered an independent SRO reviewer.

D. Incorrect - per NO-1-103, the review must be performed by an SRO appointed (designated) by the GS-SO and a second independent SRO. Since the GS-SO is directing the change he would not be considered an independent reviewer.

Page 37 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY QUE~$ti()n 90~(Q97090)

Topic: Approval of a change to the S4 Tier/Group: 3 2.2 - Equipment Control KIA Info:

• 2.2.18 - Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

SRO Importance: 3.9 Proposed references to None

• be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: rg] Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No previous use Exam Bank History: Last used in LOR Session quiz - 1/11 Technical references: NO-1-103, Conduct of Lower Mode Operations Comments: None Page 38 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit-2 has entered the appropriate Optimal Recovery Procedure for a Loss of Coolant Accident. The following conditions exist:

• HPSI and LPSI pumps are in Pull To Lock to m~et throttling criteria

• ALL Charging pumps are operating to maintain Pressurizer level within the desired band

• 11 Band 12A RCPs are operating

• A plant cooldown is in progress to reach SDC cooling initiation

• RCS Pressure is being lowered to maintain RCS subcooling low in the band The STA reports present RCS pressure trend will challenge continued Rep operation Which ONE of the following is occurring and what is the required action to maintain RCPs operating?

A. Cooldown rate is too excessive; Adjust the ADVs, to reduce the cooldown rate, which will raise subcooling.

B. Aux Spray is in use; Secure Aux Spray by reopening charging header isolations and shut the Aux Spray isolation.

C. Aux Feedwater feed rate is excessive; Reduce feed rate to S/Gs to lower cooldown rate and stabilize RCS pressure.

D. Aux Spray is in use; Secure all but one charging pump to reduce RCS depressurization.

Answer: B Answer Explanation:

A. Incorrect - Cooldown rate is not too excessive as P:zr level is being maintained with all charging pumps running. Shutting ADVs allows RCS to heatup resulting in RCS subcooling becoming even smaller and further challenge continued RCP operation.

B. Correct - This is why RCS subcooling is lowering as RCS pressure is lowered.

Reopening charging header stops and shutting Aux Spray isolation will stop subcooling from continuing to lower and maintain RCP operation.

C. Incorrect - Lowering AFW feed rate will cause RCS to heatup as this is primary method of heat removal since HPSI pumps are secured. This will further challenge subcooling limit for continued RCP operation.

D. Incorrect - This is why RCS pressure is lowering. Securing 2 of 3 charging pumps will slow depressurization however subcooling will continue to be lowered and pressurizer level will begin to lower as HPSI pumps are secured per throttling criteria. Information provided in question stem states all 3 charging pumps operating are maintaining Pzr level with cooldown in progress.

Page 39 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Actions required to recover Pzr level in EOP-5 Tier/Group: 2/1 004 - CVCS

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based KiA Info: on those predictions, use procedures to correct. control, or mitigate the consequences of those malfunctions or operations:

• A2.17 - Low PZR pressure

• SRO Importance: 3.7 Proposed references to None be provided to applicant:

Given RCS parameters, identify the appropriate response

• Learning Objective:

for Loss of Coolant Accident (LOCA) per EOP-5.

10 CFR Part 55 Content:

Question source:

Cognitive level: D Memory/Fundamental [gJ Comprehension/Analysis Last NRC Exam used on: New question Exam Bank History: None EOP-5, Loss of Coolant Accident Step J None Page 40 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit -1 is at 100% power when the following control room alarm annunciates:

EAST ECCS PP RM LVL HI

• The ABO looking through the Component Cooling Room access hatch observes a significant amount of water in the room and continuing to rise

• The CRO observes 11 Refueling Water Tank (RWT) level is 462 inches and lowering Which ONE of the following groups represents ALL affected components and what directions should be provided to the crew?

A. 12 and 13 HPSI, 12 LPSI, and 12 Containment Spray pumps; Shut the RWT outlet MOV on "A" train ECCS header, place affected pumps in Pull To Lock.

B. 11 and 12 HPSI, 11 LPSI, and 11 Containment Spray pumps; Shut the RWT outlet MOV on "B" train ECCS header, place affected pumps in Pull To Lock.

C. 11 and 12 HPSI, 11 LPSI, and 11 Containment Spray pumps; Shut the RWT outlet MOV on "A" train ECCS header, place affected pumps in Pull To Lock.

D. 12 and 13 HPSI, 12 LPSI, and 12 Containment Spray pumps; Shut the RWT outlet MOV on "B" train ECCS header, place affected pumps in Pull To Lock.

Answer: C Page 41 of 61 Rev. 3

2012 NRC SRO EXAM MIASTER KEY Explanation:

A. Incorrect - 12 HPSI Pump is located in the affected room for the alarm provided but the other components are located in the West ECCS room. Actions are for the "A" train and the leak is on the "A" train header.

B. Incorrect - Components provided are located in room with leak occurring.

Actions provided are correct to address the leak.

C. Correct - All components provided are located in room with leak occurring.

Actions provided are correct to address the leak.

D. Incorrect - 12 HPSI Pump is located in the affected room for the alarm provided but the other components are located in the West ECCS room. These are actions to isolate the "B" train ECCS components. The leak is on "A" train ECCS header.

Page 42 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Tier/Group: 2/1 006 - EGGS

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the EGGS; and (b) based KIA Info: on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.11 - Rupture of EGCS header.

SRO Importance: 4.4 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: D Memory/Fundamental IS] Comprehension/Analysis Last NRC Exam used on: New question Exam Bank History: None Injection and Containment Spray Page 43 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit-2 has just completed a startup following a refueling outage. Given the following events and conditions:

• Reactor power is 4%

• 22 SGFP is out of service for emergent repairs 21 SGFP is operating on Main Steam when the following occurs:

• 21 SGFP speed lowers from 3300 rpm to 1200 RPM

• 21 and 22 S/G levels lower to minus (-) 20 inches and continue slowly lowering Which ONE of the following statements (1) correctly dE!Scribes your direction to the operators to restore S/G water levels and (2) when a rE~actor trip would be ordered?

A. (1) Reduce power to less than 1%, initiate AFW flow to S/Gs and allow S/G levels to slowly recover while maintaining T COLD within 2°F of program; (2) Trip the reactor if S/G levels are approaching minus (-) 40 inches B. (1) Immediately align the Auxiliary Steam supply and slowly restore S/G water levels. Withdraw CEAs to maintain T COLD above 515 OF; (2) Trip the reactor if T COLD lowers to 515 OF.

C. (1) Immediately align the Auxiliary Steam supply and maximize feedwater flow to restore S/G water levels; (2) Trip the reactor if S/G levels are approaching minus (-) 40 inches.

D. (1) Reduce power to less than 1% and maximizc3 AFW flow to S/Gs to restore S/G levels. Withdraw CEAs to maintain T COLD above 515 OF; (2) Trip the reactor if TCOLD lowers to 515 OF.

Answer: A Page 44 of61 Rev. 3

A. Correct - This is the correct sequence of actions required by AOP-3G which would be implemented based on conditions listed in question stem.

B. Incorrect - Per OI-12A, this is a controlled evolution and will take several minutes between each adjustment of the Aux Steam Supply valve. Doing this would allow S/G levels to continue lowering and reach trip criteria. Withdrawing CEAs is not one of the methods provided to control RCS temperature but it will raise reactor power and may cause a plant trip on high power. MTC is very low at BOL and the effects of withdrawing CEAs will be to raise power substantially while raising T COLD relatively slowly. If the examinees are not familiar with the 1995 LER for S/G overfeed event, these are the, actions that were taken during that event complicating crew response resulting in an automatic reactor trip.

C. Incorrect - Promptly shifting back to the auxiliary steam supply will overs peed the SGFP and overfeed the S/G causing TCOLD to lower. Per OI-12A, this is a controlled evolution and will take several minutes between each adjustment of the Aux Steam Supply valve. Partially correct as AOP-3G requires tripping the reactor if SG level approaches -40 inches D. Incorrect - Withdrawing CEAs is not one of the methods provided to control RCS temperature but it will raise reactor power and may cause a plant trip on high power. 515 OF is the minimum temperature for critical operations.

Page 45 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: Shifting SGFP steam supplies at low power Tier/Group: 2/1 059 - Main Feedwater 2.4 - Emergency Procedun~s / Plan KIA Info:

• 2.4.4 - Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

,SRO Importance: 4.2 Proposed references to None be provided to applicant:

Identify the actions taken upon the failure of a SGFP with Learning Objective:

reactor power less than 5%.

10 CFR Part 55 Content: 55.41 (b)(1 0)

Question source:

Cognitive level: D Memory/Fundamental o Comprehension/Analysis Last NRC Exam used on: 2006 SRO (08/06)

Exam Bank History: LOI-2010 1C03 Exam (08/'11) reflect current procedure actions 46 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following conditions on Unit-1 at 100%:

• The following alarms are received on control panel 1C34:

o Window U-13: 11, 12125V DC BUS UN o Window U-15: 11, 12,23,24 125V BATT CHGR FAILURE

• DC Bus 11 voltage indication on panel 1C24 is '122 VDC and lowering slowly Which ONE of the following describes (1) the failure that has occurred, and (2) the operability of DC Bus 11 in accordance with Technical Specifications?

A. (1) 11 and 23 Battery Chargers have failed; (2) DC Bus 11 operability will be restored when bus voltage is restored to > 125 VDC by BOTH Battery Chargers being restored to the bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. (1) 12 and 24 Battery Chargers have failed; (2) DC Bus 11 operability will be restored when bus voltage is restored to > 125 VDC by BOTH Battery Chargers being restored to the bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. (1) 11 and 23 Battery Chargers have failed; (2) DC Bus 11 operability will be restored when bus voltage is restored to > 125 VDC by EITHER Battery Charger being restored to the bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. (1) 12 and 24 Battery Chargers have failed; (2) DC Bus 11 operability will be restored when bus voltage is restored to> 125 VDC by EITHER Battery Charger bein~J restored to the bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Answer: C Answer Explanation:

A. Incorrect Only battery charger 11 normally supplies the Bus. Operability, per TS. 3.8.4, requires a SINGLE battery charger on the Bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not BOTH.

B. Incorrect - 12 Charger does not supply DC Bus 11. Operability, per TS. 3.8.4, requires a SINGLE battery charger on the Bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not BOTH.

C. Correct - Listed battery chargers are those that normally supply the Bus.

TS. 3.8.4 will be met when EITHER battery charger is restored to the DC bus and voltage is > 125 VDC within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. Incorrect - Neither battery charger supplies 11 DC Bus. 11 DC Bus operability will not be restored using either battery charger 12 and 24.

Page 47 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: Battery Chargers inoperability on DC bus Tier/Group: 2/1 063 - DC Electrical Distribution KiA Info:

• 2.2 - Equipment Control

• 2.2.42 - Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

SRO Importance: 4.6 Proposed references to None be provided to applicant:

Given plant conditions, determine if 125 VDC busses are Learning Objective:

operable per appropriate tE~ch specs.

10 CFR Part 55 Content: 55.43(b)(2)

Question source:

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2006 Audit Remediation (11/08)

Tech Spec 3.8.4 - DC Sources, Operating 1C34-ALM, HVAC Systems Control windows U-13 and U-15 Comments: None 48 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following:

• Core Alts are in progress

• The Containment Purge system is in operation

• RI-5316A (Containment Area Monitor) exhibited erratic operation and the ESFAS Sensor Channel ZD CRS Sensor module was pulled to comply with the action of applicable Tech Spec Currently, which ONE of the following explains the effect of the Out Of Service Containment Area Monitor on (1) CRS/Containment Purge operation, and (2) fuel handling status per applicable Tech Spec?

A. (1) CRS actuation logic is reduced to 1 out of 3 logic and Containment Purge may remain in operation; (2) Fuel handling may continue.

R (1) ALL CRS sensor channels must be operable" therefore, immediately secure Containment Purge; (2) Immediately suspend fuel handling within containment.

C. (1) CRS actuation is reduced to 2 out of 3 logic and Containment Purge may remain in operation; (2) Fuel handling may continue.

D. (1) CRS actuation requires a 2 out of 4 logic, therefore, immediately secure Containment Purge; (2) Immediately suspend fuel handling within containment.

Answer: A Answer Explanation:

A. Correct - Since channel removed from service (Le. tripped), CRS requires 1 of remaining 3 channels to trip and actuate to secure Containment Purge. Fuel handling may continue in this case.

B. Incorrect - This is true, however, the tech spec actions allow continued operation of Containment Purge; second part would only occur if unable to place channel in trip within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Incorrect - Examinee may forget fact that first RMS channel is tripped requiring only one more channel to trip to actuate CRS and secure Containment Purge.

D. Incorrect - The effect of the OOS sensor has provided one of the required 2 out of 4 trip logic to actuate CRS. Containment Purge remains in operation and fuel handling continues within containment.

Page 49 of 61

2012 NRC SRO EXAM MASTER KEY Topic: RMS channel OOS for Containment Radiation Signal Tier/Group: 2/2

.029 - Containment Purge

• 2.2 - Equipment Control KIA Info:

• 2.2.37 - Ability to determine operability and/or availability of safety related equipment.

SRO Importance: 4.6 Proposed references to None be provided to applicant:

Learning Objective:

• 10 CFR Part 55 Content: 55.43(b)(5)

• Cognitive level: D Memory/Fundamental [gJ Comprehension/Analysis Last NRC Exam used on: No record of use None ech Spec 3.3.7 - Containment Radiation Signal Comments: Modified from Q50710 Page SO of 61

2012 NRC SRO EXAM MASTER KEY

. *******1 Unit-1 was operating at 100% power when a loss of instrument air occurred. Given the following events and conditions:

• The operators enter AOP-7D, Loss of Instrument Air

• Instrument air pressure is 50 PSIG and loweringl at a rapid and continuous rate Which statement correctly describes (1) the effect on the plant (2) direction provided to the crew?

A. (1) The TBVs will not quick-open below 40 PSIG; (2) Trip the reactor at 40 PSIG and lowering; B. (1) The FRVs will fail as-is at 40 PSIG; (2) Trip the reactor at 40 PSIG and lowering; C. (1) The TBVs will not quick-open below 50 PSIG; (2) Trip the reactor at 50 PSIG and lowering; D. (1) The FRVs will fail-as-is at 50 PSIG; (2) Trip the reactor at 50 PSIG and lowering; Answer: C Answer Explanation:

A. Incorrect - The TBV's will not quick open below 50 PSIG and AOP-7D specifies a reactor trip at 50 PSIG IfA Header pressure not 40 PSIG to ensure the FRVs ramp shut and TBVs are allowed to quick open upon a reactor trip to remove heat.

B. Incorrect - The FRVs fail-as-is at 40 PSIG. AOP-7D specifies a reactor trip at 50 PSIG itA Header pressure not 40 PSIG.

C. Correct - AOP-7D initial actions are to start the Saltwater Air Compressors (SWACs) which provide air to the ADVs. The 50 PSIG trip value was chosen to enable FRVs and TBVs post-trip response. The TBVs are able to quick open fully at 50 PSIG. The FRVs ramp shut, removing the immediate need to trip the SGFPs due to overfeeding effects on the RCS and provide opportunity to maintain normal heat removal methods as long as possible.

D. Incorrect - The FRVs fails as-is at 40 PSIG. AOP-7D specifies a reactor trip at 50 PSIG IfA Header pressure.

Page 51 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY

• Tier/Group: 2/2 041 - Steam DumplTurbine Bypass Control

• A2 - Ability to (a) predict the impacts of the malfunctions or operations on the SDS; and (b) based on KIA Info:

those predictions or mitigate the consequences of those malfunctions or operations:

• A2.03 - Loss of lAS SRO Importance: 3.1 Proposed references to None be provided to applicant:

Determine the Operator actions for a loss of Instrument Air Learning Objective:

in the following situations: Modes 1 and 2 10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: D Memory/Fundamental cg] Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: None Technical references: AOP-7D, Loss of Instrument Air EOP-O, Post-Trip Immediate Actions Comments: None Page 52 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following conditions on Unit 1:

• Reactor Startup in progress

• Reactor power is 10%

• The following alarm is received in the control room:

• CNDSR EXH HOOD TEMP HI VAC LO

• The crew enters AOP-7G, Loss of Condenser Vacuum

• All Condenser Air Removal Units are verified running

• Condenser vacuum indicates 23.5 inches Hg and is lowering Rapidly Which ONE of the following actions should be directed?

A. Trip the Reactor and implement EOP-O, Post Trip Immediate Actions.

B. Insert CEAs to reduce reactor power to less than 1%.

C. Trip the Turbine and implement EOP-O, Post Trip Immediate Actions.

D. Initiate RCS boration to reduce reactor power to less than 1%.

Answer: A Answer Explanation:

A. Correct - Per AOP-7G, Loss of Condenser Vacuum, requirements, Condenser vacuum has reached the low vacuum trip setpoint of 23.5 inches Hg, requiring a reactor trip and implementation of EOP-O, Post-Trip Immediate Actions B. Incorrect - This step from the AOP is applicablE! for an initial power level of

<5%.

C. Incorrect - Question stem does not indicate the Main Turbine is paralleled to the grid or being warmed up. If vacuum reaches 22.5 inches Hg, the turbine trip automatically and the reactor will not trip automatically as the Loss of Load trip is disabled < 14% power.

D. Incorrect - This step from the AOP is applicablE! for an initial power level of

<5%.

Page 53 of 61 Rev. 3

2012 NRC SRO EXAM MIASTER KEY

~

',\;' :,~

'<';~~/:,~i :,

~;(,~ij!

\c" . f(l~lt " Qu~~tiol{9j;:lnfo,(Q~991 )

<>~~,; . " '" :-' ~ '- "'~" "

,:",;\,,"/ ,!;;5ft<f' ,

Topic: Actions for a loss of Condenser Vacuum Tier/Group: Generic 2.4 - Emergency Procedufies/Plan KIA Info:

• 2.4.1 - Knowledge of EOP entry conditions and immediate action steps.

SRO Importance: 4.8 Proposed references to None be provided to applicant:

Given a loss of condenser vacuum and/or plant conditions Learning Objective:

and parameters, determine the correct operator response(s).

10 CFR Part 55 Content: 55.43(b)(5)

"",:"",,:  :," " /,' ,~~';"Y;',E'*""';;~:::o,/"" "".\1,' ,

y,Y, '"

":y'" '*"W<<L';', "',C' "'<'j;;~~'" ,';',&5%,7:,.,,;,;,; ."

Question source: ~ Bank D Modified DNew Cognitive level: ~ Memory/Fundamental ID Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2006 Audit Exam

~. ~: ,,' ,:'",Y';\,:,,',

'Y{~i%2 <,:;::f?~i"C<

,'/)1(:';;,." "

',' ';;';::; ".:Ll,/::Y;i}

Technical references: AOP-7G, Loss of Condenser Vacuum Comments: None Page 54 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Unit-2 is at 90% power. Given the following events and conditions:

• RCS activity is at normal values

• A 30 GPO tube leak develops in 22 S/G Which ONE of the following statements correctly describes the response of (1) 2-RIC-5422A (22 MAIN STM N-16 RAO MON) and 2-RIC-5422 (22 MAIN STM EFFL RAO MON), and (2) Required action?

A. (1) 2-RIC-5422A and 2-RIC-5422 show no increase; (2) Current leak rate does not meet any AOP entry criteria, continue to monitor.

B. (1) 2-RIC-5422A shows observable increase and 2-RIC-5422 shows no increase; (2) Implement AOP-2A, Excess RCS Leakage.

C. (1) 2-RIC-5422A and 2-RIC-5422 show observable increase; (2) Place the unit in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Cold Shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

O. (1) 2-RIC-5422A and 2-RIC-5422 show observable increase; (2) Implement AOP-1 0, Abnormal Secondary Chemistry Conditions.

Answer: 0 Answer Explanation:

A. Incorrect - Above 50% power, 2-RIC-5422A (N**16 gamma monitor) and 2-RIC 5422 (Main Steam Effluent rad monitor) will be in service and see an increase.

Each are able to detect a 5 GPO tube leak at normal operating temperature. 5 GPO through anyone S/G is criteria for entering! AOP-10.

B. Incorrect - Above 50% power, 2-RIC-5422A (N**16 gamma monitor) and 2-RIC 5422 (Main Steam Effluent rad monitor) will be in service and see an increase.

Each monitor is able to detect a 5 GPO tube leak at normal operating temperature. At this point leak rate is not exceeding any Tech Spec limits so placing plant in Hot Standby and subsequently Cold Shutdown is not warranted.

C. Incorrect - With power level above 50%, 2-RIC-*5422A (N-16 gamma monitor) and 2-RIC-5422 (Main Steam Effluent rad monitor) will be in service and see an increase for this RCS leak. Each can detect a 5 GPO tube leak at normal operating temperature. Entry into AOP-2A is required when S/G leakage reaches 50 GPO through anyone S/G.

o. Correct - Both these monitors see an observable increase based on this 30 GPO tube leak and power level above 50%. 5 GPO through anyone S/G is criteria for entering AOP-1 0 and continuing to monitor per Att. 2.

Page 55 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic: Main Steam Line RMS response based on Rx Power Tier/Group: 3 2.3 - Radiation Control KIA Info:

• 2.3.11 - Ability to control radiation releases SRO Importance: 4.3 Proposed references to None be provided to applicant:

Identify the Radiation Monitors that have a control interface Learning Objective:

with another system and State their control functions.

10 CFR Part 55 Content:

Question source:

Cognitive level: o Memory/Fundamental [2J Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: Remediation LOI 2010 Panel Comp (01/12)

Technical references: AOP-10, Abnormal Secondary Chemistry Conditions and bases Comments: None Page 56 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Given the following 22B RCP seal parameters at 100% power:

• Middle seal pressure 2000 PSIA

• Upper seal pressure 130 PSIA

• VCT pressure 40 PSIA

• Controlled Bleedoff pressure 52 PSIA

• Lower seal Temperature 195°F

• Controlled Bleedoff flow 2.7 GPM (1) Which of the following describes the impact on plant operation and (2) What direction will you provide the crew?

A. (1) Lower Seal has degraded and Upper Seal has failed requiring increased monitoring of the RCP seal parameters; (2) Direct the OWC to immediately contact the system engineer to evaluate continued operation of the RCP.

B. (1) Two RCP seals have failed requiring commencement of an expeditious plant shutdown; (2) Commence cooldown of the RCS to less than 350 0 F per OP-5, then secure the RCP.

C. (1) Two RCP seals have failed requiring the reactor be tripped immediately; (2) Verify reactivity control safety function then secure the RCP based on controlled bleed-off flow greater than normal.

D. (1) Lower Seal has degraded and Upper Seal has failed requiring contacting GS-SO for continued RCP operation; (2) Direct the OWC to immediately contact the system engineer to evaluate continued operation of the RCP.

Answer: B 57 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Answer Explanation:

A. Incorrect - Per the alarm manual these are the actions to take based on one seal failed. Parameters given indicate two seals have failed NOT degraded.

Controlled Bleedoff flow higher than normal confirms the lower and upper seals have failed per criteria stated in OI-1A with less than 300 PSID across each seal stage.

R Correct - Per the alarm manual this is the action to take based on two seals failed. RCP would be secured during plant cooldown when RCS temperature is 0

below 350 F (per OP-5 this is the temperature when the first two RCPs are secured during plant cooldown). Controlled Bleedoff flow higher than normal confirms the lower and upper seals have failed per criteria in OI-1A with less than 300 PSID across each seal stage.

C. Incorrect - Lower seal temperature is not trip criteria for any RCP. Controlled Bleedoff flow higher than normal confirms both the lower and upper seals have failed per criteria stated in OI-1A with less than ~~OO PSID across each seal stage.

D. Incorrect - Two seals are NOT degraded but have failed which requires RCP be shutdown per OI-1A and the alarm manual actions. Controlled Bleedoff flow higher than normal confirms the lower and upper seals have failed per criteria stated in OI-1A with less than 300 PSID across each seal stage.

Page 58 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Topic:

Tier/Group: 2/1 003 - Reactor Coolant Pump System

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based KJA Info: on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.01 - Problems with RCP seals, especially rates of seal leak-off SRO Importance: 3.9 Proposed references to None be provided to applicant:

Determine the actions required for single or multiple RCP Learning Objective:

seal failures.

10 CFR Part 55 Content: 55.43(b)(5)

• Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use LOI-2008 1C06 & Reactor Reg (04/09) 1C06 ALM, RCS Control Alarm Manual Technical references:

OI-1A, Reactor Coolant System and Pump Operations Comments: None Page 59 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY Using provided reference:

Unit-1 and Unit-2 are operating at 100% power. Given the following events and conditions:

• Maintenance requested to take the 1A Diesel Generator (DG) out of service for surveillance.

• 01-49 (Operability Verification) was performed on Unit 1 ZB train equipment.

• All other DGs and offsite power sources were verified to be operable.

Which ONE of the following statements correctly and completely describes the impact of this maintenance on the status of 11 HPSI Pump?

A. 11 HPSI pump is considered operable while the 1A DG is out of service regardless of the status of the remaining HPSI pumps.

B. 11 HPSI pump is considered NOT operable while the 1A DG is out of service regardless of the status of the remaining HPSI pumps.

C. 11 HPSI pump is considered operable while the 1A DG is out of service unless both the 12 and 13 HPSI pumps are declared to be inoperable.

D. 11 HPSI pump is considered operable while the 1A DG is out of service unless the 13 HPSI pump is declared to be inoperable.

Answer: D Answer Explanation:

A. Incorrect - If examinee is unfamiliar with how to apply the requirements of LCO 3.8.1 action 8.3 this may be selected. The 11 HPSI pump would be inoperable only if 13 HPSI pump became inoperable for these conditions.

8. Incorrect - If examinee is unfamiliar with how to apply the requirements of LCO 3.8.1 action B.3 this may be selected. The 11 HPSI pump would be inoperable only if 13 HPSI pump became inoperable for these conditions.

C. Incorrect - The 11 HPSI pump would be inoperable only if 13 HPSI pump became inoperable for these conditions. 12 HPSI pump is NOT qualified as a HPSI pump in the safety analysis because it is mechanically aligned to the 11 loop but electrically aligned to 14 4KV bus.

D. Correct - This is the correct interpretation of LCO 3.8.1 action 8.3.

Page 60 of 61 Rev. 3

2012 NRC SRO EXAM MASTER KEY HPSI Pp operability with DG out of service

• Tier/Group: 3 2.2 - Equipment Control KJA Info:

• 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance: 4.2 Proposed references to

'd d T.S 3,8.1 be provi e to applicant:

Given a Mode of operation and a set of equipment conditions, identify applicable Technical Specifications (TS)

Learning Objective:

Conditions and Technical Requirement Manual (TRM)

Non-Conformances.

10 CFR Part 55 Content: 55.43(b)(2)

Question source:

Cognitive level: o Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: LOI-2006 (08/06)

Exam Bank History: LOI-2006 Recovery Exam (10/08)

Tech Spec 3.8.1 Action B.3 Technical references:

01-49, Operability Verification page 18 Comments: None Page 61 of 61 Rev. 3

CLIFFS NUCLEAR POWER PLANT 2012 NRC INITIAL LICENSED OPERATOR RO WRITTEN EXAM KEY Page 1 of 162

2012 NRC RO EXAM MASTER KEY or "B"to say a licensed op~rator e<"!",.. ....,rn Page 2 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY nanCeIIJlenl- D~lete(ft:l;re word "eNE~' from the qtt~stion Page 3 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Shortly after a reactor trip, when reactor power indicates 10-3 %, a stable negative SUR is attained. Reactor power will decrease to 10-4% in approximately _ _ _ _._ _

seconds.

A. 90 B. 180 C. 360 D. 540 Answer: B Answer Explanation:

A. Incorrect - Following a Rx trip, the SUR is -1/3 DPM and it will take approximately 3 minutes (180 seconds) not 90 seconds to lower power to 10E-4%.

B. Correct - Following a Rx trip, the SUR is -1/3 DPM and it will take approximately 3 minutes (180 seconds).

C. Incorrect - Following a Rx trip, the SUR is -1/3 DPM and it will take approximately 3 minutes (180 seconds) not 6 minutes (360 seconds) to lower power to 10E-4%.

D. Incorrect - Following a Rx trip, the SUR is -1/3 DPM and it will take approximately 3 minutes (180 seconds) not 9 minutes (540 seconds) to lower power to 10E-4%.

Page 40f162 Rev. 3

2012 NRC RO EXAM Mi\STER KEY Topic: Which RPS response is correct for a reactor trip?

Tier/Group: 1/1 EPE - 007 Reactor Trip

• EK1 - Knowledge of the operational implications of KIA Info: the following concepts as they apply to the reactor trip:

• EK1.04 - Decrease in reactor power following reactor trip (prompt drop and subsequent decay)

RO Importance: 3.6 Proposed references to be None provided to applicant:

Learning Objective: LOI-58-1-01 10 CFR Part 55 Content: 55.41 (b)(8)

Cognitive level: [gI Comprehension/Analysis Last NRC Exam used on:

Exam Bank History: 11-60 Biennial written exam (12/11)

Technical references: EOP-O Technical Bases page 12 Comments: None 5 of 162 Rev. 3

2012 NRC RO EXAM MJ\STER KEY Given the following:

• Unit-1 is at 100% power

• RCS Pressure Control is in AUTO

• Pressurizer Backup Heaters are in AUTO

• RCS Pressure is 2250 PSIA What is the IMMEDIATE plant response if the selected Pressurizer Pressure controller setpoint fails to 2500 PSIA?

A. Spray valve controller goes to minimum output, proportional heaters output goes to maximum, and all backup heaters energize.

B. Spray valve controller goes to minimum output, proportional heaters output goes to maximum, and all backup heaters remain off.

C. Spray valve controller goes to maximum output, proportional heaters output goes to maximum, and all backup heaters deenergize.

D. Spray valve controller goes to minimum output, proportional heaters output goes to minimum, and all backup heaters remain off.

Answer: B Answer Explanation:

A. Incorrect - Spray valves remain closed and Backup Heaters remain off until actual pressure lowers to 2200 PSIA. Proportional Heaters go to maximum.

Spray will collapse the Pressurizer bubble causing Pressurizer level to rise.

B. Correct - The Pressurizer Spray valves would remain closed, Proportional Heaters energize to maximum to raise PZR pressure to setpoint, and Backup Heaters remain off until actual pressure lowers to 2200 PSIA.

C. Incorrect - The Pressurizer Spray valves remain closed and the Backup Heaters remain off until actual pressure lowers to 2200 PSIA.

D. Incorrect - The Pressurizer Spray valves remain closed and the Proportional Heaters would go to maximum output.

Page 6 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Plant response to a changE~ in the Pzr pressure controller setpoint.

Tier/Group: 1/1 027 - Pressurizer Pressure Control System (PZR PCS)

Malfunction:

• KIA Info:

• AK2 - Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following:

• AK2.03 - Controllers and positioners RO Importance: 2.6 Proposed references to None be provided to applicant:

Learning Objective: LOI-064A2-1 10 CFR Part 55 Content: 55.41 (b)(7)

Cognitive level: D Memory/Fundamental ~ Comprehension/ Analysis Last NRC Exam used on: N/A Exam Bank History: LOR11-6B Biennial Written Exam (11/11)

System Description - 0640, RCS Instrumentation; Technical references:

ALM-1 C06, RCS Control Comments: Modified version of Q92862

2012 NRC RO EXAM Mi\STER KEY Given the following conditions on Unit 1:

• Reactor power is 100%.

• The following annunciator window alarms are received in the sequence listed:

• 1C03, C-28, 11 SGFP DISCH PRESS HI

• 1C03, C-38, 11 SG FW CONTR CH LVL

• 1C03, C-39, 12 SG FW CONTR CH LVL

• 1C03, C-44, 11 SGFPT SPD CONTR SYS TROUBLE The eRO observes the following at the control panel:

• 11 SGFPT speed is lowering

• 11 SGFP discharge pressure is 1352 PSIG and lowering

• 11 and 12 SG levels are (+) 32 inches and slowly rising

• Main Feed Reg Valves are responding as expected Which ONE of the following describes the status of the Feedwater system; and the action required for the plant conditions?

A. 11 SGFPT discharge pressure has ONLY exceeded the setpoint for SGFPT setback (Runback);

Trip the reactor, trip 11 SGFP, and perform EOP-O, Post-Trip Immediate Actions.

B. 11 SGFPT discharge pressure has exceeded the setpoint for SGFPT setback (Runback) AND SGFPT trip; Trip 11 SGFP and reduce SG levels using the guidance in AOP-3G, Malfunction of Main Feedwater System.

C. 11 SGFPT discharge pressure has ONLY exceeded the setpoint for SGFPT setback (Runback);

Operate 11 SGFP in manual to reduce speed and restore SG levels per AOP-3G, Malfunction of Main Feedwater System.

D. 11 SGFPT discharge pressure has exceeded the setpoint for SGFPT setback (Runback) AND SGFPT trip; Trip the reactor, trip 11 SGFP, and perform EOP-O, Post-Trip Immediate Actions.

Answer: C Page 8 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Incorrect -. S/G level trip setpoint (+50 inches) not yet reached but discharge pressure above 1350 PSIG initiates the setback circuit. S/G level at +32 inches will actuate a S/G level control channel alarm. Not necessary to trip reactor until attempt made to control 11 SGFP speed manually which, if successful, will restore S/G levels.

B. Incorrect - SGFPT trip setpoint not reached (1450 PSIG); Tripping SGFP at 100% power results in being unable to control S/G levels. A rapid downpower would be necessary to continue operating at power but being successful to control S/G levels would most likely cause an automatic reactor trip.

C. Correct - Alarm response manual validates that automatic runback signal will initiate whenever pressure exceeds 1350 psig and automatically start to lower SGFP speed. Since MFRVs are closing to compensate for high levels, it is necessary for operator to take manual control and adjust speed to restore S/G levels.

D. Incorrect - SGFPT Setback is initiated but SGFPT trip not reached. Tripping SGFP at 100% power with setback initiated would result in a more rapid drop in S/G levels causing a reactor trip before attempts made to control SGFP speed in manual and restore S/G levels. EOP-O not required, level is not above trip setpoint (+50 inches)

Page 9 of 162 Rev. 3

2012 NRC RO EXAM M)\STER KEY Topic: Main Feedwater Tier/Group: 2/1 059 - Main Feedwater

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based KJA Info: on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.03 - Overfeeding event RO Importance: 2.7 Proposed references to be None provided to applicant:

Recall the actions taken for a SGFP speed controller Learning Objective:

failure.

10 CFR Part 55 Content:

Question source:

Cognitive level: D Memory/Fundamental [gj Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2006 Audit Remediation Exam (11/08)

AOP-3G, Main Feedwater Malfunctions; Technical references:

ALM-1 C03, Condensate and Feedwater Control Comments: None Page 10 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Unit-1 is operating at 100% power. The following RCP parameters are being monitored:

11B RCP 12A RCP 40 PSIG 40 PSIG 870 PSIA 400 PSIA Middle seal 1750 PSIA 1325 PSIA Lower cavity seal 120 of 124 of temperature Bleedoff Flow 2.2 GPM 0.0 GPM Controlled Bleedoff 122 of 145 of Temperature Which ONE of the following statements correctly describes the condition of the RCP seals?

A. 11 B RCP lower seal degraded; 12A RCP upper seal degraded with vapor seal failed.

B. 11 B RCP middle seal degraded; 12A RCP upper and vapor seal failed.

C. 11 B RCP seals are normal; 12A RCP middle seal degraded.

D. 11 B RCP lower seal failed; 12A RCP vapor seal failed.

Answer: A Page 11 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Correct - Based on 11 B RCP middle and upper seal pressures the lower seal is degraded. Based on 12A RCP middle seal pressure higher than normal the upper seal is degraded with vapor seal failed based on controlled bleedoff flow.

B. Incorrect - 11 B middle seal is higher as it is breaking down % of remaining pressure drop; 12A RCP upper seal has not completely failed yet as lower and middle seal are reducing RCS pressure by % the current value. Per OI-1A, there is > 300 PSID across upper seal so it is reducing pressure but not by %

the value.

C. Incorrect- Normal pressures are % the value (Le. 1500/750); 12A RCP middle seal pressure is adjusting due to degraded upper seal. Also since 12A RCP controlled bleedoff flow is 0.0, the vapor seal has failed.

D. Incorrect - The 11 B lower seal is degraded not failed. The 12A RCP upper seal is degraded and the vapor seal has failed.

Page 12 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY (Q97()01) 003 Reactor Coolant Pump System (RCPS)

• A4 - Ability to manually operate and/or monitor in the KIA Info: Control Room:

• A4.04 - RCP seal differential pressure instrumentation RO Importance: 3.1 Proposed references to be None provided to applicant:

Given a set of RCP seal indications, determine the status of Learning Objective:

the seal(s).

10 CFR Part 55 Content: 55.41 (b)(7)

Question source: Bank D New Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use on NRC exam Exam Bank History: None Technical references: OI-1A, Reactor Coolant System And Pump Operations Comments: Modified Q28840 to add 2nd RCP seal conditions.

Page 13 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions on Unit-1 in MODE 5:

• RCS Temperature is 190 of with a plant heatup in progress per OP-1

• PZR level being maintained at 150 inches

• Shutdown Cooling (SOC) has just been secured

• RCS Pressure is being maintained at 290 PSIA lAW OP-1

• 11A and 12B RCPs Oil Lift pumps have been started and operated for at least one minute

• 11A RCP was started five (5) minutes ago

• Just prior to starting 12B RCP, the "OIL LIFT PP PRESS LO" annunciator alarms and will not clear

• 12B RCP Upper and lower oil reservoir levels checked using plant computer indicate normal values Which ONE of the following is appropriate action based on current conditions?

A. Raise RCS pressure to allow single RCP operation using applicable pump operating curve per OI-1A, RCS and Pump operations.

B. Start 128 RCP, after 30 seconds ensure oil lift pump stops automatically and check clear the "OIL LIFT PP PRESS LO" alarm.

C. Stop 128 Oil lift pump, start 12A RCP oil lift pump and operate for at least one minute, then start 12A RCP.

O. Secure 11A RCP, lower RCS pressure, and reinitiate SOC operation lAW OP-1, Plant Startup from Cold Shutdown.

Answer: 0 Answer Justification:

A. Incorrect - This condition is not allowed per OP-1 to operate a single RCP to commence initial plant heatup.

B. Incorrect - 12B RCP will not start as oil lift pressure is interlocked with the RCP starting circuit.

C. Incorrect - Per OP-1, when selecting a pair of RCPs to start, the second pump must be started within 5 minutes of first one started per Caution prior to step in OP-1. 12A RCP oil lift pump must operate for one minute before starting RCP and this exceeds time limit between RCP starts of OP-1.

O. Correct - Per OP-1, it states if a pair of RCPs cannot be started, then reinitiate SOC. One of first steps is to lower RCS pressure before opening SOC Header return isolations.

Page 14 of 162 Rev. 3

2012 NRC RO EXAM M/\STER KEY

• Tier/Group: 2/1 003 Reactor Coolant Pump System (RCPS)

KIA Info:

• K6 - Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

• K6.14 - Starting requirements RO Importance: 2.6 Proposed references to None be provided to applicant:

Determine which set of RCPs are the preferred set for initial Learning Objective:

starting and identify the initial RCP starting criteria.

10 CFR Part 55 Content: 55.41 (b)(7)

Cognitive level: D Memory/Fundamental k8J Comprehension/Analysis Last NRC Exam used on: New Question None OP-1, Plant Heatup from Cold Shutdown Technical references:

01-1 A, Reactor Coolant System And Pump Operations Comments: None Page 15 of 162 Rev. 3

2012 NRC RO EXAM MJ~STER KEY Given the following plant conditions on Unit-2:

• A small break LOCA has occurred

• EOP-O actions have been completed

• The appropriate Optimal Recovery Procedure has been implemented

• Containment pressure peaked at 3.1 PSIG and is slowly lowering

• Both S/Gs levels at (-) 70 inches and rising slowly

• RCS T COLD is 520 0 F and lowering

• Pressurizer (Pzr) level is 140 inches and rising rapidly

• Aux Spray is initiated and PZR pressure is 1100 PSIA and lowering

• CET subcooling is 45 OF

• RVLMS lights 1 and 2 are illuminated Which ONE of the following are the appropriate actions per conditions stated?

A. Reduce charging flow to a single pump then secure aux spray to stabilize RCS pressure.

B. Secure both HPSI pumps simultaneously and adjust cooldown to stabilize RCS temperature.

C. Slow the cooldown rate and secure Auxiliary Spray to maintain subcooling in the specified band.

D. Reduce HPSI flow by throttling HPSI header valves or stopping HPSI Pumps one at a time to maintain Pzr level in the specified band.

Answer: D Page 16 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Justifications:

A. Incorrect - This is allowed but is only done ~fter HPSI flow has been secured. Reducing Aux Spray will stop lowering RCS pressure, however, the biggest rise in PZR level is attributed to HPSI flow into RCS.

B. Incorrect - Securing both HPSI pumps together would stop rise in PZR level immediately. Step for HPSI throttling/termination specifically states when conditions met to stop HPSI Pumps one at a time or throttle HPSI header valves. Stabilizing RCS temperature would only stop RCS depressurization, however, charging flow would still be injecting into RCS causing PZR level to continue rising.

C. Incorrect - Subcooling is not being jeopardized at the current value or with the current trends. Examinee must know the subcooling limits for the given condition and perform an analysis to eliminate this distracter. These actions may affect HPSl flow but are not the EOP-5 actions.

D. Correct - ALL conditions are met to throttle/terminate HPSI flow which is the most correct action to take for plant conditions.

Page 17 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Actions to control PZR leVE!!

Tier/Group: 1/1 009 Small Break LOCA / 3 KiA Info: EK3 - Knowledge of the reasons for the following responses as they apply to the small break LOCA:

• EK3.24 - ECCS throttling or termination criteria RO Importance: 4.1 Proposed references to None be provided to applicant:

Given plant conditions, det,ermine actions to take for ECCS Learning Objective:

throttling/termination criteria.

10 CFR Part 55 Content: 55.41 (b)(5)(10)

Question source:

Cognitive level: D Memory/Fundamental ~ Comprehension/ Analysis Last NRC Exam used 011: New Question None EOP-5, Loss of Coolant Accident, and Technical Bases 1L;()mlments: None Page 18 of 162

2012 NRC RO EXAM MASTER KEY Considering an ESDE and a LOCA, both which cause containment pressure to peak at 30 PSIG, which ONE of the following conditions can be used to differentiate between the accidents?

A. RCS subcooling conditions may be at saturation during the LOCA.

B. Total hydrogen generation is less during the LOCA.

C. S/Gs are a major contributor to heat removal during the LOCA.

D. The Containment High Range monitor will alarm during the LOCA.

Answer: A Answer Explanation:

A. Correct - Due to loss of inventory from the RCS, subcooling may reach saturation during a LOCA. During an ESDE, subcooling is increased as RCS cools down from faulted S/G, no inventory is lost.

B. Incorrect - Hydrogen is generated during the ESDE and the LOCA. During the LOCA the amount of hydrogen produced depends on the duration of core uncovery and the maximum core temperature reached. During the ESDE hydrogen is produced, however, no RCS fluid is released into the containment to add to hydrogen being produced.

C. Incorrect - During a large break LOCA the RCS and S/Gs are uncoupled.

During a small break LOCA the S/Gs become a significant contributor to RCS heat removal. Since containment pressure peaked at 30 PSIG this represents a medium to large break LOCA has occurred and the SGs are uncoupled from the primary.

D. Incorrect - The EOP-4 and 5 technical bases state "The containment high range radiation monitor is not expected to be received though, except in extreme situations. Calculations have shown that this alarm will not alarm even during LOCA conditions unless fuel failure has occurred." The question stem does not provide any indication that fuel failure is present or has occurred. Also peak containment pressure value given is not the design large break LOCA analyzed per the UFSAR.

Page 19 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Actions to control PZR level Tier/Group: 1/1 011 Large Break LOCA /3

• EA2 - Ability to determine or interpret the following KIA Info: as they apply to a Large Break LOCA:

• EA2.13 - Difference between overcooling and LOCA indications RO Importance: 3.7 Proposed references to be None provided to applicant:

Compare the following plant parameters response to Learning Objective: differentiate between the design basis accidents, ESDE and a LOCA, occurring:

10 CFR Part 55 Content: 55.43(b)(5)

Question source:

Cognitive level: [gJ Memory/Fundamental 0 Comprehension/Analysis Last NRC Exam used on: New Question EOP-5, Loss of Coolant Accident and Technical Bases None 20 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions:

• Unit-1 is performing a plant cooldown

• PORVs are in Variable MPT Enable

• An RCS overpressure condition occurred

• The cause of the high pressure condition is corrected Which ONE of the following provides the complete operator response to this condition, if any?

A. No action required, the PORVs will close automatically.

B. Place and maintain PORV Override handswitches in override to close.

C. Shut both PORV block valves, when PORVs reseat reopen the block valves.

D. Place PORV Override handswitches in override to close; return to auto when PORVs are shut.

Answer: D Answer Explanation:

A. Incorrect - Plausible as during normal operation these valves will reclose.

When in single or variable MPT enable must plaice in override to close and when PORVs are shut return to auto.

B. Incorrect - When in single or variable MPT enable must place in "override to close" to shut the PORVs but must be returned to AUTO to restore MPT overpressure protection.

C. Incorrect - Plausible as PORVs will reclose when not in LTOP conditions once block valves have been shut. Per alarm manual, this action is only required if a PORV fails to close or has opened due to a failE~d transmitter (each PT operates only one PORV for MPT).

D. Correct - When in single or variable MPT enable must place in override to close but to restore overpressure protection must be returned to AUTO.

Page 21 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Actions to control PZR level Tier/Group: 1/1 008 Pressurizer Vapor Space Accident / 3

• AK2 - Knowledge of the interrelations between the KIA Info: Pressurizer Vapor Space Accident and the following:

• AK2.01 - Valves RO Importance: 2.7 Proposed references to be None provided to applicant:

• Given PORV HS positions, RCS temperature and RCS Learning Objective: pressure, determine whether PORVs are enabled or disabled for MPT.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: D Memory/Fundamental Last NRC Exam used on: New Exam Bank History:

Technical references: 1C06-ALM, RCS Control window E-21 Comments:

Page 22 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given a Reactor trip on High Pressurizer Pressure:

Which ONE of the following specifically identifies that the Diverse Scram System (DSS) has actuated to automatically trip the reactor?

A. Window D-05: "Prot Ch Trip" B. Window D-46: "MG Set No Output" C. Window D-16: "Pzr Press Hi Ch Pre-Trip" D. Window D-45: "Reactor Trip Bus UN Relay Trip" Answer: B Answer Explanation:

A. Incorrect - This alarm occurs when anyone of the ten RPS trip units reaches the trip setpoint and would annunciate in the event of DSS tripping the reactor. Post trip conditions can result in receipt of this alarm due to normal plant response to a reactor trip (low S/G level, TM/LP, etc.). DSS is monitored by ESFAS and provides a "DSS TRIP" alarm on 1C05 which is not provided in question stem.

B. Correct - Whenever DSS actuates each CEDM MG set main load contactor (3M) is opened and this annunciator window alarms along with "DSS TRIP" alarm which is not provided in question stem.

C. Incorrect - Examinee may assume this alarm occurs when PZR pressure reaches 2335 PSIA, which is significantly below DSS trip setpoint (2435 to 2460 PSIA). ESFAS sensor channel trips (if not already in alarm) would alert operator of impending DSS condition.

D. Incorrect - This alarm, by itself, would not identify a DSS trip. It can occur as the result of anyone of the following conditions:

• RPS generated trip

• Manual Rx trip

• DSS generated Rx trip

• A single faulted UV relay.

Page 23 of 162 Rev. 3

2012 NRC RO EXAM M}~STER KEY Topic: Determining when DSS has actuated Tier/Group: 1/1 029 - Anticipated Transient Without Scram

• 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.45 - Ability to prioritize and interpret the significance of each annunciator / alarm.

RO Importance: 4.1 Proposed references to be None provided to applicant:

Identify the cause and effect of the following Learning Objective:

Control Element Drive System (CEDS) ...

10 CFR Part 55 Content: 55.41(b)(10)

Question source:

Cognitive level: k8J Memory/Fundamental Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI 2010 Panel Comp remediation (01/12)

Technical references: 1C05-ALM, Reactivity Control Alarm Manual 01-34, Engineered Safety Features Actuation System, Appendix 0 Comments: Modified from 036552 Page 24 of 162 Rev. 3

2012 NRC RO EXAM M)~STER KEY Using Provided

Reference:

Unit-2 is at 100% power when the containment sump annunciator alarms 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from last alarm. Frequency of alarm prior to this current alarm was every 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. All other containment parameters remain stable.

The following conditions exist:

• 21 and 23 Charging Pumps are running

• REGEN HX OUT temperature, TE-221, is rising

• PZR level has lowered from 216 inches to 214 inches over the last two minutes and continues to lower slowly

• RCS temperatures are stable

• Letdown flow is starting to lower

• The appropriate AOP has been implemented Based on plant conditions, which ONE of the following actions is required?

A. Shut 2-CVC-182, CHG PP HDR XCONN, and start ONLY 22 or 23 Charging pump to initiate flow through 2-CVC-518 and 2-CVC-519, LOOP CHG ISOLs.

B. Shut 2-CVC-182, CHG PP HDR XCONN, and start ONLY 21 Charging pump to initiate flow through 2-CVC-269-MOV, SI TO CHG HDR valve to the Aux HPSI header.

C. Shut 2-CVC-183, REGEN HX CHG INLET, and start 21,22, or 23 Charging pump to initiate flow through 2-CVC-269-MOV, 81 TO CHG HDR valve to the Aux HPSI header.

D. Shut 2-CVC-183, REGEN HX CHG INLET, and start 21 Charging pump to initiate flow through 2-CVC-518 and 2-CVC-519, LOOP CHG ISOLs.

Answer: C 25 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Incorrect -Based on the containment sump alarm the leak is on the charging header downstream of 2-CVC-183 and AOP-2A directs closing this valve and establishing flowpath to Aux HPSI header with any Charging pump. Shutting 2-CVC-182 assumes that leak is upstream of this valve and starting 22 or 23 charging pump will restart the leak on charging header.

B. Incorrect - Based on the containment sump alarm the leak is on the charging header downstream of 2-CVC-183 and AOP-2A directs closing this valve.

Shutting 2-CVC-182 is the action taken when leak is determined to be upstream of 2-CVC-183 to establish flowpath using only 21 Charging pump to the Aux HPSI header.

C. Correct - Based on the containment sump alarm the leak is on the charging header downstream of 2-CVC-183 and AOP-2A directs closing this valve.

Shutting 2-CVC-183 will isolate the leak and AOP-2A direct starting any charging pump to the Aux HPSI header when 2-CVC-183 is shut through 2-CVC-269 MOV, SI to CHG HDR.

D. Incorrect - Shutting 2-CVC-183 will isolate the leak and remove the flowpath using the normal charging header stops. Only path available is aligning charging to the Aux HPSI header through 2-CVC-269-MOV, SI to CHG HDR.

Page 26 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Restoring letdown flow Tier/Group: 1/1 022 - Loss of Reactor Coolant Makeup

• AK3 - Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant KIA Info: Makeup:

• AK3.02 - Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging. and abnormal charging RO Importance: 3.5 Proposed references to be Simplified drawing of cves showing ONLY charging pump provided to applicant: flowpath Given a charging header leak, determine the actions Learning Objective: required per AOP-2A to reestablish charging flow into the RCS.

10 CFR Part 55 Content:

Question source:

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of previous use Technical references: AOP-2A-2, Excess RCS Leakage Comments: Resampled KIA and modified from Q50859 Page 27 of 162 Rev. 3

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2012 NRC RO EXAM MASTER KEY Following a plant transient resulting in a loss of ALL AC power and reactor trip, you are directed to verify Natural Circulation.

Which ONE of the following would be observed when comparing Core Exit Thermocouple (CET) response to RCS temperature trends?

A. CET temperatures are approximately TAVE but trend with T HOT.

B. CET temperatures are always slightly lower than THOT but trend with T HOT.

C. CET temperatures trend consistent with T COLD which is constant or lowering.

D. CET temperatures trend consistent with T HOT which is constant or lowering.

Answer: D Answer Explanation:

A. Incorrect - Per EOP-7 Basis step K, CET temperatures trend consistent with T HOT.

Hot leg RTD temperature should be consistent with core exit thermocouples.

Adequate natural circulation flow ensures core exit thermocouple temperatures will be approximately equal to the hot leg RTDs tempE~rature within the bounds of the instruments' inaccuracies. Generally speaking, the, CET temperatures will be somewhat higher than T HOT.

B. Incorrect - Per EOP-7 Basis step K, CET temperatures trend consistent with T HOT.

Hot leg RTD temperature should be consistent with core exit thermocouples.

Adequate natural circulation flow ensures core exit thermocouple temperatures will be approximately equal to the hot leg RTDs tempE!rature within the bounds of the instruments' inaccuracies. Generally speaking, the CET temperatures will be somewhat higher than T HOT.

C. Incorrect - Per EOP-7 Basis step K, CET temperatures trend consistent with T HOT.

Hot leg RTD temperature should be consistent with core exit thermocouples.

Adequate natural circulation flow ensures core exit thermocouple temperatures will be approximately equal to the hot leg RTDs tempE!rature within the bounds of the instruments' inaccuracies. Generally speaking, the CET temperatures will be somewhat higher than T HOT.

D. Correct - Per EOP-7 Basis .step K, CET temperatures trend consistent with T HOT.

Hot leg RTD temperature should be consistent with core exit thermocouples.

Generally speaking, the CET temperatures will be somewhat higher than T HOT.

Page 28 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY Tier/Group: 1/1 055 Station Blackout / 6 KIA Info:

• EA 1 - Ability to operate and monitor the following as they apply to a SBO:

• EA1. a1 - In-core thermocouple temperatures 1 RO Importance: 13.7 i Proposed references to be None provided to applicant:

Recall the plant parameters used to verify natural Learning Objective:

circulation is occurring or being maintained.

10 CFR Part 55 Content: 55.41 (b)(7) i k8J Memory/Fundamental 0 Comprehension/ Analysis New Question Technical references:

Comments: iNone Page 29 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY Unit 1 was operating at 100% power when a reduction in instrument air header pressure occurred. Given the following events and conditions:

• Instrument Air header pressure lowered to 87 PSIG

• Plant Air header pressure lowered to 84 PSIG Which of the following conditions should have occurred?

(1) The standby instrument air compressor started (2) CNTMT IA SUPPLY CV, 1-IA-2085-CV, shuts (3) Plant air header automatic isolation valve (PA-2059-CV) closed (4) Plant air to instrument air cross connect valve (PA-2061-CV) opened A. Actions 2 and 3 only B. Actions 1,3, and 4 only C. Actions 1 and 4 only D. Actions 2, 3, and 4 only Answer: B Answer Explanation:

A. Incorrect - Action 3 has occurred but Action 2 has not. Action 2 occurs at 75 PSIG IA pressure, and Action 3 occurred at 85 PSIG PA pressure.

B. Correct - Action 1 occurred at 93 PSIG IA pressure, Action 3 occurred at 85 PSIG PA pressure, and Action 4 occurred at 88 PSIG IA pressure.

C. Incorrect - Action 1 occurred at 93 PSIG IA pressure and Action 4 occurred at 88 PSIG IA pressure, however, Action 3 also occurs.

D. Incorrect - Action 2 has not occurred based on IA pressure value. Actions 1, 3, and 4 have occurred based on IA and PA pressure.

Page 30 of 162 Rev. 3

2012 NRC RO EXAM MJ~STER KEY Question :12 (Q9'l'00a:)

Topic: Lowering IA pressure effects Tier/Group: 1/1 065 - Loss of Instrument Air /8 KfA Info:

• AA1- Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air:

• AA1.04 - Emergency Air Compressor RO Importance: 3.5 Proposed references to be None provided to applicant:

Given lowering instrument air conditions, determine the Learning Objective:

actions needed and why.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

[gJ Memory or Fundamental Cognitive level:

D Comprehension or Analysis Last NRC Exam used on: LOI-2008 RO (06/08)

Exam Bank History: None Technical references: AOP-70-1, Loss of Instrument Air Page 5 Comments: Modified Q50747 - added IIA pressure value to each distractor along with basis behind action.

Page 31 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions:

• Both Units have tripped from 100% power due to loss of offsite power

• Both 4KV ESF buses on each unit are reenergizlad from the dedicated DG

• 13 and 23 AFW Pumps are operating to recover S/G levels with the following flow rates observed in the EOP implemented from EOP-O:

• 11 S/G is 290 GPM

• 12 S/G is 280 GPM

• 21 S/G is 270 GPM

• 22 S/G is 260 GPM Which ONE of the following represents the status of the Total AFW and individual unit AFW Flow limits and expected action?

A. Total AFW flow is below maximum allowed, Unit -1 and 2 AFW flows are below maximum allowed; Monitor S/G levels and adjust AFW flows to maintain within band.

B. Total AFW flow is below maximum allowed, Unit-1 AFW flow is below maximum allowed, Unit-2 AFW flow limit is 300 GPM; Reduce Unit-2 AFW flows to 150 GPM per S/G.

C. Total AFW flow is above maximum allowed, Unit-1 and Unit-2 AFW flow limits are 300 GPM; Reduce Unit-1 and Unit-2 AFW flows to 150 GPM per S/G.

D. Total AFW flow is above maximum allowed, Unit-1 AFW flow is limited to 300 GPM, Unit-2 AFW flow is below maximum allowed; Reduce Unit-1 AFW flows to 150 GPM per S/G.

Answer: B Page 32 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Incorrect - Total Flow and Unit-1 AFW are correct; Unit-2 AFW flow is limited to 300 GPM as 23 AFW Pp is being powered from 2B DG.

B. Correct - Status of each flow limit is correct; R:educing Unit-2 AFW flow to 150 GPM to each S/G ensures that the 2B DG is not overloaded.

C. Incorrect - Only Unit-2 is exceeding AFW flow limits, only need to reduce flow to Unit-2 S/Gs to 150 GPM to ensure the 2B DG is not overloaded.

D. Incorrect - Unit-2 AFW flow is limited to 300 GPM as 23 AFW Pp is being powered from 2B DG, reducing flow to 150 GPM per S/G ensures the 2B DG is not overloaded.

Page 33 of 162 Rev. 3

2012 NRC RO EXAM MJ\STER KEY Question 13 (q9700SI)

Topic: AFW flow limits during LOOP on each unit Tier/Group: 1/1 056 Loss of Off-site Power / 6 KIA Info:

• 2.2 - Equipment Control

• 2.2.3 - Knowlledge of the design, procedural, and operatiol1al differences between units.

RO Importance: 3.8 Proposed references to be None provided to applicant:

Given plant conditions, determine if 13(23) AFW flow limits Learning Objective:

are being met.

10 CFR Part 55 Content: 55.41(b)(10)

Question source:

Cognitive level: D Memory/Fundamental [gJ Comprehension/Analysis Last NRC Exam used on: New Question Exam Bank History: None OI-32A-1 & 2, Auxiliary Feledwater System Technical references: EOP-2-1 & 2, Loss of Offsite Power/Loss of Forced Circulation Comments: None Page 34 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions on Unit 1:

• Reactor has tripped

• The crew has transitioned to EOP-4, Excess Steam Demand Event, due to 12 S/G pressure lowering uncontrollably

• 12 ADV is shut and control has been transferred to 1C43

• The following conditions are indicated:

• 11 S/G pressure is 700 PSIA

• 12 S/G pressure is 200 PSIA

• RCS pressure is 1350 PSIA

• Core Exit Thermocouple temperatures are 445°F

• RCS Loop 12 TcoLD is 410°F Which ONE of the following provides the needed response, and reason for the response, as the faulted S/G continues to blowdown?

A. Reduce pressure in 11 S/G to 500 PSIA; Maintains heat removal to limit the possibility of a PTS transient.

B. Reduce pressure in 11 S/G to 500 PSIA; Prevents a steam generator tube rupture once 12 S/G is dry.

C. Reduce pressure in 11 S/G to 350 PSIA; Maintains heat removal to limit voids forming in the unaffected S/G.

D. Reduce pressure in 11 S/G to 350 PSIA; Minimizes flP between S/Gs to limit the possibility of a PTS transient.

Answer: A Page 35 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Correct - This pressure is within 25°F of CETs and does not add to the RCS cooldown rate as S/G saturation temperature is established above the CET temperatures and limits the possibility of a PTS transient following an excessive cooldown of the RCS.

B. Incorrect - This pressure is within 25 OF of CETs and does not add to the RCS cooldown rate as S/G temperature is established above the CET temperatures but the reason is wrong.

C. Incorrect - Lowering 11 S/G pressure to this value (432°F) adds to the RCS cooldown from faulted S/G although it is within 25°F of CETs. Actions of EOP 4 are concerned about RCS voids forming during this event but this is not the reason for lowering unaffected S/G pressure.

D. Incorrect - Lowering 11 S/G pressure to this value (432°F) adds to RCS cooldown from faulted S/G although it is within 25 OF of CETs. Lowering unaffected S/G pressure is not to limit the L\P between the S/Gs. Actions are to maintain within 25°F of CETs to restrict ReS heatup following blowdown to limit possibility of PTS transient.

Page 36 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY

• Topic: In-core Thermocouple temperatures trend Tier/Group: 1/1 040 Steam Line Rupture - Excessive Heat Transfer

/4

• AK1 - Knowledge of the operational implications of KIA Info: the following concepts as they apply to Steam Line Rupture:

• AK1.03 - ReS shrink and consequent depressurization I RO Importance: 3.8 Proposed references to be None provided to applicant:

Given conditions and/or parameters associated with an Learning Objective:

ESDE, determine the appropriate operator actions.

10 CFR Part 55 Content:

Cognitive level: D Memory/Fundamental IS] Comprehension/Analysis Last NRC Exam used on: No record of previous use LOI-2008 Audit (11108)

Technical references: EOP-4 and EOP-4 Technical Bases Comments: None Page 37 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following condition with Unit-1 at 100% power:

• An electrical perturbation occurs resulting in ALL Unit-1 Annunciator lights being de-energized (Status Panels remain energized)

What is (1) the minimum bus lost and (2) the required actions expected to be performed?

A. (1) DC Bus 21; (2) Verify the PRZR LVL CH SEL Switch is in the "110X" position, PZR HTR LO LVL CUT-OFF SEL Switch in the "X" position and resetthe Proportional Htrs by placing the handswitches to OFF and return to AUTO, and RRS CH SEL Switch is in the RRS-X position.

B. (1) DC Bus 11; (2) Verify the PRZR LVL CH SEL Switch is in the "110Y" position, PZR HTR LO LVL CUT-OFF SEL Switch in the "Y" position and reset the Proportional Htrs by placing the handswitches to OFF and return to AUTO, RRS CH SEL Switch is in the RRS-Y position.

C. (1)DCBus21; (2) Perform the alternate actions to trip the reactor by deenergizing the MG sets from panel1C17as the pushbuttons at 1C05 and 1C15 are inoperable and implement EOP-O.

D. (1) DC Bus 11; (2) Dispatch an operator to the Unit-1 Main Turbine front standard and when notified operator is stationed at front standard trip the reactor from 1C05, then immediately trip the main turbine from the front standard and implement EOP-O.

Answer: A

2012 NRC RO EXAM Mf\STER KEY Answer Explanation:

A. Correct - DC Bus 21 has been lost based on all annunciator lights being de energized on Unit-1. These are the immediate stabilizing actions performed by the operators from the Immediate Actions Plaque at 100% power prior to implementing AOP-7J.

B. Incorrect - DC Bus 11 remains energized. These actions are wrong as PZR pressure and level, and RRS channel "Y" instruments have been deenergized based on DC Bus 21 lost.

C. Incorrect - Bus lost is correct and all Reactor Trip Circuit Breakers are able to be tripped using either set of push buttons at 1C05 or at 1C 15. There is no need to perform Reactivity Control Alternate Actions to deenergize the CEDM MG sets from panel 1C 17 per EOP-O.

D. Incorrect - DC Bus 11 remains energized and the Unit-1 Main Turbine can be tripped manually at 1C02. There is no need to dispatch an operator to the front standard to trip the main turbine upon a reactor trip from 1C05.

Page 41 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Loss of Vital Bus

• Tier/Group: 1/1 058 Loss of DC Power / 6

• AA1 - Ability to operate and / or monitor the following KIA Info: as they apply to the Loss of DC Power:

• AA1.03 - Vital and battery bus components

• I RO Importance: 3.1 Proposed references to None be provided to applicant:

Given a loss of any 125VDC bus on Unit-1 or Unit-2,

• Learning Objective:

i determine the actions requil"ed per applicable AOP-7J.

10 CFR Part 55 Content:

! Cognitive level: D Memory/Fundamental cg] Comprehension/Analysis Last NRC Exam used on: NEW None AOP-7J, Loss of 120 Volt Vital AC or 125 Volt Vital DC Comments:

Page 42 of 162 Rev. 3

AOP-7J Rev 19/Unit 1 Page 78 of 102 XIII. 21 125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS RESPOND TO A LOSS OF 21 125 VOLT DC BUS.

Verify that the PRZR PRESS CH SEL Switch is in the X position.

Verify that the RRS CH SEL Switch is in the RRS-X position.

Verify that the PRZR LVL CH SEL Switch is in the 110X position.

Verify that the PZR HTR LO LVL CUT-OFF SEL Switch is in the X position.

NOTE Switch S2 is located inside the RRS Test Panel Drawer at 1C31

5. Isolate RCS Loop 12 instruments to RRS Channel X by placing switch S2 to OFF.

(continue)

AOP-7J Rev 19/Unit 1 Page 79 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A. (continued)

CAUTION The Turbine will NOT automatically trip AND Main Feedwater will NOT reconfigure to the post trip state when the Reactor is

(

tripped due to ESFAS BL Actuation Cabinet erglzea.

6. IF the Reactor is critical, THEN perform the following actions:

,... I

a. Station personnel at 1C05 and 1CO2.

b. Trip the Reactor.

c. WHEN the Reactor is tripped, THEN immediately trip the Turbine.

d. Perform the Reactivity Control i

~

immediate actions of EOP-O, POST TRIP IMMEDIATE ACTIONS.

e. Isolate the 13 KV Bus power supplies to ALL RCPs:

(1 ) Place Unit 1 RCP Bus Feeder 'l Breaker Control Switch, 1-CS-252-1201, in PULL TO LOCK.

(2) Place Unit 1 RCP Bus Feeder Breaker Control Switch, 2-CS-252-2202 in PULL TO LOCK.

f. IMPLEMENT EOP-O, POST TRIP IMMEDIATE ACTIONS.

.,;4~

(continue)

AOP-7J Rev 19/Unit 1 Page 80 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A. (continued)

7. IF the Reactor is NOT critical, THEN isolate the 13 KV Bus power supplies to ALL RCPs:

a. Place Unit 1 RCP Bus Feeder Breaker Control Switch, 1-CS-252-1201, in PUll TO lOCK.

b. Place Unit 1 RCP Bus Feeder Breaker Control Switch, 2-CS-252-2202 in PUll TO LOCK.

c. Determine the appropriate emergency response actions PER the ERPIP.

8. The following components will be affected bv thA,'L.* Olli'

~ ~ ALL Un~ 1 Annunciator ti9h:----"'"

~

deenergized (Status Panels remain energized)

II.

D

• Normal power supply to the 11 B and 12B RCPs

• 13 and 144 KV Buses

• 13A, 13B, 14A and 14B 480 Volt Buses

• 11 Band 12B RCPs are untrippable from 1C06

• CC CNTMT RETURN, 1-CC-3833-CV, fails shut

• Loss of SRW to the Turbine Building

• IA and PA may be lost due to loss of SRW to the Turbine Building

• Channel B ESFAS and AFAS Actuation Cabinets de-energized (continue)

AOP-7J Rev 19/Unit 1 Page 81 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A.8 (continued)

• Channel A CSAS and SGIS will NOT actuate the following:

• 11 SGFP

• 12 SGFP

• 11 COND BSTR PP

• 12 COND BSTR PP

• 13 COND BSTR PP

• 11 HTR DRN PP

• 12 HTR DRN PP

• Channel ZE ESFAS and AFAS Sensor Cabinets de-energized

• Channel B RPS Cabinet de-energized

• Channel B PAMS de-energized

• 11 SG AFW STM SUPP & BYPASS valves,1-MS-4070-CV, 1-MS-4070A-CV fail shut

• loss of quick open signal to the Turbine Bypass Valves AND loss of quick open signal and auto control of ADVs

• 12 CC and 12 ECCS Pump Room HX SW outlet valves fail open

• 12 SRW HX SW valves fail to their full HX flow position

• loss of letdown, due to 1-CVC-516-CV failing shut

• 11 and 12 SFP Heat Exchangers lose cooling flow due to SRW inlet CVs failing shut

• AFW Turbine Driven Train Flow Control Valves fail open:

• (11 SG) 1-AFW-4511-CV

• (12 SG) 1-AFW-4512-CV (continue)

AOP-7J Rev 19/Unit 1 Page 82 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A.8 (continued)

• PORV-404 inoperable in MPT ENABLE

• CNTMT Area Rad Monitor, 1-RI-5316B, out of service

• 12 Main Steam Effluent Rad Monitor, 1-RIC-5422, out of service

• 12 MSIV loses position indication, but can still be closed from 1C03

• RCP BLEED-OFF ISOL valve, 1-CVC-505-CV, fails shut CAUTION If the difference between the PRZR WTR TEMP and CHG OUT TEMP is greater than 4000 F, then TRM 15.4.2 must be complied with.

9. Operate Pressurizer HTRs as necessary 9.1 IF RCS pressure is greater than 2275 to maintain RCS pressure between 1850 PSIA, and 2275 PSIA. THEN initiate AUX SPRAY.

a. IF 12 Proportional Heater is to be a. Record the following information:

turned off, THEN locally trip NO. 12 PZR Heater

• PI~ZR WTR TEMP (1-TI-101)

Proportional Controller Breaker,

• CHG OUT TEMP (1-TI-229) 52-1430.

b. Open the AUX SPRAY valve, 1-CVC-517-CV.

c. Operate the LOOP CHG valves as necessary to adjust AUX SPRAY flow:

• 1-CVC-518-CV

• 1-CVC-519-CV

d. Shift the PRESSURIZER SPRAY VLV CONTROLLER, 1-HIC-100, to MANUAL.

(continue) (continue)

AOP-7J Rev 19/Unit 1 Page 83 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A.9 (continued) A.9.1 (continued)

e. Shut the PRZR SPRAY VLVs by adjusting the output of 1-HIC-100 to 0%:

• ~1-RC-1 OOE-CV

• 1-RC-100F-CV

f. WHEN AUX SPRAY is NO longer needed, THEN perform the following actions:

(1 ) Open LOOP CHG valves:

• 1-CVC-518-CV

• 1-CVC-519-CV (2) Shut AUX SPRAY valve, 1-CVC-517-CV.

10. Operate Charging Pumps to maintain PZR level between 80 and 180 inches:

a. IF 12 Charging Pump is to be stopped, THEN locally trip 12 Charging Pump Breaker 52-1415.

b. IF 13 Charging Pump is supplied from 14480 Volt Bus, THEN locally trip 13 Charging Pump Breaker 52-1404 and align its power supply from the 11A 480 Volt Bus PER 01-270, STATION POWER 480 VOLT SYSTEM.

c. Shut the UO CNTMT ISOL valves:

• 1-CVC-515-CV

• 1-CVC-516-CV

11. Place the CC CNTMT RETURN, 1-HS-3833 to CLOSE.

(continue)

AOP-7J Rev 19/Unit 1 Page 84 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A. (continued)

12. Start 11 and 12 SALTWATER AIR COMPRs.

13. Throttle 12 SW Pump Discharge Valve.

1-SW-108, to maintain 12 SW Pump discharge pressure between 15 and 30 PSIG as indicated at the pump discharge pressure gauge.

14. Trip 12 IA Compressor Breaker 52-1418.

NOTE Due to 11 SG AFW STM SUPP & BYPASS valves, 1-MS-4070-CV and 1-MS-4070A-CV failing shut, there may be a difference in SG pressures. Under certain conditions an AFAS BLOCK signal could be generated.

NOTE AFW Steam Train Flow Control Valves, 1-AFW-4511 and 1-AFW-4512. fail open on loss of power.

15. Initiate Motor train AFW flow: 15.1 Initiate Steam train AFW flow:

a. Start 13 AFW PP. a. Shut the inlet isolation valves to the steam train flow control valves:

b. Maintain SG levels between (-)170 and (+)30 inches. * (1-AFW-4511-CV) 1-AFW-162

• (1-AFW-4512-CV) 1-AFW-164

c. IF the AFW Steam train is in operation.

THEN secure the AFW Steam train. b. Open the 12 SG AFW STM SUPP &

BYPASS valves. 1-MS-4071-CV, 1-MS-4071 A-CV.

(continue) (continue)

AOP-7J Rev 19/Unit 1 Page 85 of 102 XIII. 21125 VOLT DC BUS ACTIONS ALTERNATE ACTIONS A.15 (continued) A.15.1 (continued)

c. Throttle open the bypass valves to the steam train flow control valves to maintain SG levels between

(-)170 and (+)30 inches.

• (1-AFW-4511-CV) 1-AFW-163

• (1-AFW-4512-CV) 1-AFW-165 NOTE The 12 and 13 COND PPs, 13 COND BSTR PP, 12 HTR DRN PP, and 12 SGFP have lost ALL protective trips and remote trip functions.

16. Secure Main Feedwater System lineup:

a. Trip 11 SGFP.

b. Locally trip 12 SGFP.

c. Locally trip 12 HTR DRN PP at Breaker 152-1306 by depressing the TRIP pushbutton.

d. Stop 11 HTR DRN PP.

e. Locally trip 13 Condensate Booster Pump at Breaker 152-1304 by depressing the TRIP pushbutton.

f. Place ALL COND BSTR PP handswitches in PULL TO LOCK.

g. Locally trip 12 and 13 Condensate Pumps at their respective breakers by depressing the TRIP pushbuttons:

• (12 COND PP) 152-1307

• (13 COND PP) 152-1308

h. Operate 11 COND PP as necessary.

i. Place 12 and 13 COND PP handswitches in PULL TO LOCK.

(continue)

2012 NRC RO EXAM M)~STER KEY Given the following conditions:

• Unit-1 is in Mode 3

• RCS temperature is 532 OF and steady

• RCS pressure is 2250 PSIA and steady

• VCT level is steady

• ALL RCPs are running

• The following control room annunciators have alarmed:

• "CCW FLOW LO" and "CCW TEMP HI" for each RCP

• "cc HEADTK LVL" with level at 38 inches and lowering

• "CNTMT NORMAL SUMP LVL HI" Which ONE of the following is the required action by the control room?

A. Implement AOP-7C, Loss of Component Cooling Water; secure ALL RCPs and then implement EOP-2, Loss of Offsite Power/Loss of Forced Circulation.

B. Implement AOP-2A, Excess RCS Leakage; isolate CC to Letdown HX by placing 1-TIC-223 in manual with 100% output signal and shut 1-CC-266.

C. Implement AOP-7C, Loss of Component Coolin~1 Water; secure ALL RCPs and concurrently implement AOP-3E, Loss of ALL Rep Flow.

D. Implement AOP-2A, Excess RCS Leakage; secure ALL RCPs and shut CC Containment Supply and Return Valves.

Answer: C Answer Explanation:

A. Incorrect - Implementing AOP-7C and securing ALL RCPs is correct, however, AOP-7C does not direct implementing EOP-2 based on current mode.

B. Incorrect - Stem statement has the unit in MODE 3 at 532 OF. Implementing AOP-2A does not apply here. Since VCT level is steady an RCS leak does NOT exist. There is no indication of leakage into CC from the Letdown HX. These are the actions taken if letdown isolation valves were shut and stopped the RCS leak.

RCS leakage into CC system from letdown would cause the head tank level to rise NOT lower.

C. Correct - Securing ALL RCPs per AOP-7C and implementing AOP-3E are required actions to take.

D. Incorrect - Stopping RCPs and isolating CC to containment are needed actions, however, AOP-2A does not apply here. VCT level is steady per question stem indicating no leak exists from the RCS into the CC system. These actions would be correct if letdown isolation valves were shut and RCS leak was not isolated.

RCS fluid leaks into CC system from the RCPs would cause the head tank level to rise NOT lower.

Page 43 of 162 Rev. 3

2012 NRC RO EXAM Mi\STER KEY Topic: Effect on the CCW flow header of a loss of CCW Tier/Group: 1/1 026 - Loss of Component Cooling Water

• AK3 - Knowledge of the reasons for the following responses as they apply to the Loss of Component KIA Info:

Cooling Water:

• AK3.04 - Effect on the CCW flow header of a loss ofCCW RO Importance: 3.5 Proposed references to be None provided to applicant:

Given a loss of CC system, diagnose the event and take

. Learning Objective:

appropriate actions.

10 CFR Part 55 Content:

Cognitive level: ~ Comprehension/Analysis Last NRC Exam used on: No record of previous use Exam Bank History: LOR 11-5E Session 5 weekly quiz (10/11)

Technical references:

Comments: . Enhanced stem statement conditions and changed distractors to 2 X 2 format with answer explanations.

Page 44 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions on Unit-1:

• A S/G tube leak developed and the crew performed the required AOP actions

• The Reactor was tripped per the AOP trip criteria and EOP-O entered

• During EOP-O actions 1Y10 deenergized due to an electrical fault The Crew transitioned to the appropriate Optimal Recovery Procedure. The following conditions exist:

• RCS pressure is stable at 800 PSIA and approximately equal to ruptured S/G

• SIAS and SGIS were blocked during RCS depressurization and cooldown

• RCS cooldown continuing to initiate shutdown cooling

• The Ruptured S/G is isolated and its level is being maintained between a and

+50 inches

• A Pressurizer bubble exists and Pressurizer level is being maintained between 101 and 180 inches Which ONE of the following represents the status of charging and letdown flow paths for the EOP in use?

A. Charging flowpath is ONLY thru Aux Spray valve; Letdown flowpath was isolated for inventory control but is available.

B. Charging flowpath is ONLY thru Loop Charging valves; Letdown flow has been restored.

C. Charging flowpath is thru LOOP CHG valves .Q! Aux Spray valve; Letdown flowpath was isolated for inventory control and remains unavailable.

D. Charging flowpath is through the Aux HPSI Header.

Letdown flowpath remains unavailable due to a loss of Instrument Air.

Answer: C Page 45 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Incorrect - Wrong, both Loop CHG valves or Aux Spray valve are able to maintain charging flow path; Letdown was isolated in AOP-2A and remains unavailable as 1-CVC-515-CV fails shut due to loss of 1Y10.

B. Incorrect - Wrong, both Loop CHG valves or Aux Spray valve are able to maintain charging flow path; Prior to EOP-6 entry, letdown was isolated per AOP-2A or EOP-O actions and remains unavailable as 1-CVC-515-CV fails shut on loss of 1Y10.

C. Correct - Both paths are available to maintain charging flow and letdown was isolated and remains unavailable as 1-CVC-S15-CV fails shut on loss of 1Y10.

D. Incorrect - EOP-6 does not provide guidance for Charging via the Aux HPSI Header and 1-CVC-515-CV is closed due to the loss of 1Y10, not a loss of IIA.

Page 46 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY qdestion Charging and Letdown flow paths during a SGTR i Tier/Group: 1/1 038 - Steam Generator Tube Rupture

• EA2 - Ability to determine or interpret the following as KIA Info: they apply to a SGTR:

• EA2.10 - Flow path for charging and letdown flows RO Importance: 3.1 Proposed references to be I'

. provided to applicant: None Learning Objective:

10 CFR Part 55 Content:

Cognitive level: o Memory/Fundamental [gJ Comprehension/Analysis Last NRC Exam used on: New Question Exam Bank History: None

.Technical references: i EOP-6, Steam Generator Tube Rupture I Comments: None Page 47 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions:

• Both units are at 100% power

• Both units generator frequency begin to slowly oscillate between 59.5 HZ and 60.3 HZ

• Unit-1 Main Generator terminal voltage begins to slowly oscillate between 24 KV and 26 KV

• Unit -1 "MAIN GEN EXCTR AUTO TO MANUAL TRANSFER" alarm annunciates Which ONE of the following is the required operator action?

A. Trip both units and implement EOP-O.

B. Place 11 GEN VOLT REG SEL 1-CS-43 to MAN and notify SO-TSO.

C. Transfer 21 GEN VOLT REG SEL 2-CS-90 to TEST (MAN) and notify SO-TSO.

D. Commence a rapid down power on both units to maintain grid stability.

Answer: B Answer Explanation:

A. Incorrect - Although grid disturbances are occurring, the frequency and voltage oscillations have not reached trip criteria on either unit.

B. Correct - Alarm response manual directs operator to transfer voltage regulator to MANUAL upon receipt of this alarm to allow controlling main generator voltage as needed. Alarm actuating implies the regulator has shifted to manual but matching HS to manual clears the alarm and allows operator to control main generator voltage using 11 GEN MANUAL VOLT CONTR, 1-CS-70, handswitch.

C. Incorrect - Stem statement alarm condition informs examinee that the Unit-1 voltage regulator has determined a need to shift to manual. AOP-7M directs checking U-2 Voltage Regulator in AUTO since main generator is paralleled to the grid. Stem statement provided no indication of voltage swings occurring on Unit-2. Since no voltage oscillations are occurring on Unit-2, it is prudent to keep the voltage regulator in AUTO.

D. Incorrect - The AOP only directs a load reduction if partial grid losses have occurred and it is needed to lower frequency. Per stem statement this has not occurred. Grid losses are not occurring just voltage swings that resulted in Unit-1 voltage regulator transferring from Auto to Manual.

Page 48 of 162

2012 NRC RO EXAM MASTER KEY Topic: Major Grid Malfunctions Tier/Group: 1/1 077 - Generator Voltage and Electric Grid Disturbances

• 2.1 - Conduct of Operations KIA Info:

• 2.1.23 - Ability to p*erform specific system and integrated plant procedures during all modes of plant operation.

RO Importance: 4.3 Proposed references to be None provided to applicant:

Given voltage and frequency parameters for Unit 1 and Unit 2 Learning Objective: main generators, evaluate for entry conditions of AOP-7M and take the appropriate actions.

10 CFR Part 55 Content:

Question source:

Cognitive level: ~ Memory/Fundamental Last NRC Exam used on: No record of previous use LOR 11-3R weekly remediation exam (08/11)

AOP-7M, Major Grid Disturbances and Technical Bases document Technical references:

ALM-1 C01: Main Generator And Switchyard Control Alarm Manual Comments: Added answer explanations for distractors. Added alarm condition to stem statement and revised two distractors to be enhance difficulty.

1C01-ALM MAIN GENERATOR AND SWiTCHYARD CONTROL Rev. 40 ALARM MANUAL Page 72 of 82 DEVICE SETPOINT WINDOW A-51 RTXA N/A MAIN GEN EXCTR AUTO TO MANUAL TRANSFER POSSIBLE CAUSES

• Potential Transformer voltage imbalance

• Potential Transformer malfunction

• Exciter field overcurrent

• Failure of maximum excitation limiter

• Generator Field Breaker open with control switch in normal position AUTOMATIC ACTIONS Alarm on 1E01, A1-4. REG TRIP TO MANUAL, will annunciate.

CAUTION A Main Turbine trip may occur if automatic corrective action is not effective.

CONDITION RESPONSE

1. U1 Main Generator Exciter transfer 1. Perform the following:

from auto to manual control.

a. IE U 1 Reactor trips, THE:N IMPLEMENT EOP-O, Post Trip Immediate Actions.

b. TRJliNSFER voltage regulator to manual.

c. CONTROL U1 Main Generator voltage using manual control.

d. MONITOR Main Generator and Exciter operation closely while the regulator is in manual.

e. NOTIFY the SO-TSO. [B0614] IQ400J (continued)

AOP-7M Rev 1 Page 15 of 20

v. MAJOR GRID DISTURBANCES ACTIONS ALTERNATE ACTIONS IE. RESPOND TO GRID INSTABILITY. ,

1. IF EITHER Main Generator is paralleled, THEN perform the following:

a. Check the Main Generator Voltage Regulator(s) in AUTO.

\;AUIIUN Upon notification, action should be taken as soon as practical, NOT to exceed 30 minutes.

b. Coordinate with the SO-TSO to maintain Main Generator MVARS AND frequency.

NOTE If partial grid losses have occurred, Main Generator load reduction may be required to lower frequency.

~AUTION Upon notification, action should be taken as soon as practical, NOT to exceed 30 minutes.

c. Adjust Reactor Power as necessary PER OP-3, NORMAL POWER OPERATION.

2. Perform the following:

• IF the grid disturbance reaches plus or minus 2% voltage or frequency, THEN notify Engineering and Licensing to ensure a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report is generated to DOE/NERC.

• Initiate a Condition Report.

• The Transient Undervoltage (TUR) setpoint of 3.71 I<V ensures adequate voltage to start all safety-related equipment. The setpoint was determined by considering the minimum voltage needed at the 4KV bus to ensure that there is at least 75% voltage at the terminals of the safety-related equipment. Safety-related equipment is purchased to start at 75% of nominal voltage. There is a time delay of 8 seconds associated with this setpoint.

• The Loss of Voltage (LOV) setpoint of 2.45 KV is considered a loss of bus voltage and warrants immediate Emergency Diesel Generator start to immediately supply plant safety systems.

There is a time delay of 2 seconds associated with tllis setpoint.

A concern of SOER 99-01 was the loss of individual components due to tripping on overcurrent caused by low voltage conditions requiring local operation of equipment even after restoration of the electrical bus to a nominal condition. This condition was evaluated at CCNPP and is protected by the Transient Undervoltage and the Steady State Undervoltage setpoints contained in the ESFAS.

There is a 2 of 4 logic for each of these setpoints on each 4KV Vital Bus. If the design function fails, actions for this step instruct the operator to take protective actions for the individual components for that bus train (including the lower buses) by de-energizing the associated 4KV Vital Bus. The associated safety-related Emergency Diesel Generator breaker is specifically verified open first, to ensure the step does not interfere with the Loss of Voltage relay design function automatically re energizing the bus. The operator should then check that Undervoltage Actuation has occurred. If Undervoltage Actuation has failed, proper load shed, blocking and sequencing is not assured. In this condition, if a DG is manually closed onto the bus, dElsign loading may be exceeded immediately or at some later time. (

Reference:

SOER 99-01, Loss of Grid)

The objective of this step is to maintain grid stability. Auto operation of the Main Generator Voltage Regulators is desirable, and it was assumed that if these regulators are not already in auto, that manual operation was required by a preexisting condition. Operational limitations should be understood for the existing mode of operation. The cautions to make adjustments within 30 minutes are to ensure notification to the SO-TSO of the inability to control grid parameters. (

Reference:

SOER 99-01, Loss of Grid) m ***

2. If a grid disturbance is experienced, a condition report is !Jenerated to ensure proper investigation and reports are generated and communicated with the SO-TSO and NERC reliability coordinator. If the disturbance is reaches +/-2% voltage or frequency, then a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report is required to the DOE/NERC. (

Reference:

SOER 99-01, Loss of Grid)

I AOP-7M Major Grid Disturbances 9 of 12 Rev. 1/U-1 & 2

2012 NRC RO EXAM MASTER KEY Unit-1 has tripped from 100% power. The following conditions exist:

• A momentary loss of Control Room lighting occurred on BOTH Units

• 1C03 annunciator window C-25, "SGFP(S) SUeT PRESS LO" in alarm

• 1C04 annunciator window W-03, "MOTOR SYS NO Flow" in alarm

• 1C04 annunciator window W-04, "TURB SYS NO Flow" in alarm

• Diesel Generators have automatically started with output breakers closed

• Both MSIVs were shut per EOP-O Alternate Actions Assuming ALL EOP-O actions are complete, which procedure would be implemented for plant conditions?

A. EOP-8, Functional Recovery Procedure B. EOP-4, Excess Steam Demand Event C. EOP-3, Loss of ALL Feedwater D. EOP-2, Loss of Offsite Power I Loss of Forced Circulation Answer: C Answer Explanation:

A. Incorrect - Although a LOOP has occurred with a Loss of All Feedwater EOP-8 is not the appropriate EOP to implement. EOP-3 addresses a LOOP within it actions.

B. Incorrect - MSIVs being shut are required manual actions due to LOOP as an alternate action for Turbine Trip to secure MSR lineup. Since non-vital power is unavailable, the MSIVs must be shut. No other conditions indicate an ESDE is in progress.

C. Correct - This is Optimal EOP to implement. Bullets 2 thru 4 conditions are indicative that a loss of all feedwater has occurred with a LOOP event as well.

EOP-3 addresses a loss of offsite power in the major actions to allow crew to restore a source of feedwater. Motor system no flow alarm indicates issue with 13 AFW pump lost or system lineup (power is available to 4KV buses 11 and 14).

Turbine system no flow alarm indicates issue with 11 and 12 AFW Pumps or system lineup.

D. Incorrect - Although a LOOP has occurred as l:lvidenced by bullets 1, 2, 5 and 6 implementing EOP-2 does NOT address actions needed to restore feedwater to S/Gs which EOP-3 specifically provides.

Page 50 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Loss of ALL Feedwater

• Tier/Group: 1/1 CE/E06 Loss of Feedwater /4

• EK1. Knowledge of the operational implications of the, following concepts as they apply to the (Loss of '

KIA Info: Feedwater)

• EK1.2 - Normal, abnormal and emergency operating procedures associated with (Loss of Feedwater)

RO Importance: 3.2 Proposed references to None I be provided to applicant:

Given various plant conditions and EOP-O actions complete, Learning Objective:

implement the appropriate EOP.

10 CFR Part 55 Content: 55.41 (b)(10)

Question source:

Cognitive level: D Memory/Fundamental fZI Comprehension/Analysis Last NRC Exam used on: New Question Exam Bank History: None Technical references: EOP-3, Loss of ALL Feedwater Comments: None Page 51 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY Unit-2 is at 100% power. STP 0-8B-2 is in progress witlh the 2B DG paralleled to the 24 4KV bus and has been at full load for 30 minutes.

A transient occurs resulting in a "21 SRW HDR PRESS LO" and "U-2 4KV ESF MOTOR OVERLOAD" alarms. 21 SRW header pressure indicates 30 PSIG and steady.

The following temperatures exist:

• Main Turbine Thrust Bearing Metal is 145°F

• Main Turbine Journal Bearing Metal is 180°F

• Generator Hydrogen temperature is 46°C Which ONE of the following actions should be taken first?

A. Immediately trip the 2B DG.

B. Immediately trip the reactor and implement EOP-O.

C. Commence a power reduction per OP-3.

D. Start 23 SRW pump after verifying it is aligned to 21 SRW header.

Answer: D Answer Explanation:

A. Incorrect - AOP-7B, step V.B.2 states "If 2A or 2B DG is affected by loss of its associated SRW header, then with SM/CRS permission, shutdown the DG PER the appropriate procedure being used at the time~ of event initiation". The 2B DG is unaffected by the transient as it is cooled by 22 SRW header and if examinee believes the 2B DG could be affected this action is incorrect. The DG would be shutdown using the OI/STP used to start and parallel onto the bus.

B. Incorrect - No trip criteria have been exceeded for the thrust or journal bearing temperatures and the generator hydrogen temperature although they are rising.

C. Incorrect - A load reduction would not be required yet based on the given parameters. SRW loads in the Turbine Building are cross connected and although the temperatures on equipment would rise, an immediate load reduction would not be required. Coordination with system operator is to reduce MVARs to zero to minimize heating on main generator.

D. Correct - Indications are provided that the 21 SRW pump has tripped due to an electrical issue. AOP-7B directs that the swing SRW pump be mechanically aligned and started on the affected SRW header.

Page 52 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: SRW leak and isolation Tier/Group: 1/1 062 Loss of Nuclear Svc Water / 4 AA2 - Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water:

KIA Info:

• AA2.03 - The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition RO Importance: 2.6 Proposed references to None be provided to applicant:

Given ANY of the following alarms, determine the cause and Learning Objective: corrective actions required to clear the alarm(s):

• 21(22) SRW HEAD TK LVL 10 CFR Part 55 Content: 55.41(b)(10) i Cognitive level: Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: I LOR-09E Biennial Written exam (12/09)

Technical references: AOP-7B, Loss of Service Water 2C13-ALM, SRW And Misc Station Services Alarm Manual None Page 53 of 162 Rev. 3

AOP-7B Rev 12/Unit 2 Page 8 of 50

v. MODES 1 OR2 ACTIONS ALTERNATE ACTIONS Is. REDUCE SRW HEAT LOAD.

I f<il0TE Reducing Main Generator Reactive Load will reduce Main Generator heating.

CAUTION Reducing Reactive Load on Unit 2 may cause Unit 1 Main Generator or DG limits to be exceeded.

CAUTION Rapid changes in Main Generator Reactive Load require coordination with the SO-TSO to minimize Electric System perturbations and alarms.

1. IF the Main Generator is paralleled, THEN coordinate with the SO-TSO to reduce the Main Generator MVARs to zero.

~F 2A or 2B DG is affected by loss of ~ }If, rvfl:. H

(, associated SRW header, THEN with SM/CRS permission, shutdown the DG PER the appropriate procedure

\

) I,"nt~

~ng used at the time of event initiati:/

3. Commence power reduction PER OP-3, 1-Mp2J5 ]}t.

NORMAL POWER OPERATION, as required.

Ic. ATTEMPT TO RESTORE SRW FLOW.

I

1. IF the loss of SRW is due to a system leak or rupture, THEN PROCEED to step D, Page 14.

(continue)

AOP-7B Rev 12/Unit 2 Page 90f50

v. MODES 1 OR 2 ACTIONS ALTERNATE ACTIONS C. (continued)

2. IF an operating SRW PP has failed, THEN perform the following actions:

a. Place the handswitch for the failed SRW PP in PULL TO LOCK.

b. IF a Saltwater Header is removed from service PER 01-15 SERVICE WATER SYSTEM, THEN PROCEED to step C.2.d.

Page 10.

c. IF the backup SRW PP is available. I c.1 IF 23 SRW pp needs to be aligned to 21 \

THEN ensure that the backup SRW SRWHeader.

PP is mechanically aligned to the affected header. ~ ~

THEN perform the following actions:

,-1\ r:u.-~handswitch for 23 SRW

, PP in PULL TO LOCK.

(2) Lock shut 23 SRW PP Suction and Discharge valves to 22 SRW Header:

'. 2-SRW-119

• 2-SRW-120

• 2-SRW-123

• 2-SRW-124 (3) LocI< open 23 SRW pp, Suction and Dischi:1rge valve:; to 21 SRW,.

Header:

• 2-SRW-117

• 2-SRW-118

• 2-SRW-121

• 2-SRW-122 (continue) (continue)

12 NRC RO EXAM MASTER KEY An End-of-Cycle (EOC) reactor start-up on Unit-1 is in progress 4 days after a forced outage shutdown. The following conditions exist:

• Boron equalization is in progress

• Critical data has been recorded at 1 x 10E4% power

• The RO withdraws Reg. Group 4 CEAs to establish a sustained positive SUR of 0.8 DPM to raise power to the Point of Adding Heat (POAH).

• Shortly after CEA withdrawal is terminated the following are observed:

• 1C05 annunciator window D-15: "Power Lvi Rate Hi Ch Pre-Trip" alarms

• CEA outward motion is observed with CEDS in "OFF"

• S/G pressures are 910 PSIA and slowly rising Which ONE of the following statements describes the required response?

A. Trip the Reactor and implement EOP-O, Post-Trip Immediate Actions per AOP-1 B, CEA Malfunctions.

B. Place TBVs in MANUAL and lower output signal and insert CEAs or BORATE the RCS to lower SUR to zero to stabilize power.

C. Insert CEAs using Manual Sequential to lower SUR below 1.0 DPM as an excessive CEA withdrawal event has occurred.

D. Commence fast boration to the RCS to raise boron to 2300 PPM, trip the reactor, and implement EOP-O.

Answer: A Answer Explanations:

A. Correct - This is the required action of AOP-1B since the CEDS control system was in OFF based on 2 nd bullet of stem statement and it is malfunctioning. It is apparent that an uncontrolled CEA withdrawal is occurring.

B. Incorrect - Examinee notes that S/G pressures rising means T COLD is rising and an RCS cooldown is NOT occurring. This is action directed from alarm manual condition #3.

C. Incorrect - If examinee believes an excessive withdrawal means CEAs are continuing to move OUT. Since SUR continued to rise after original SUR established, using CEDS to insert CEAs will most likely be unsuccessful.

Once again this is action from alarm manual condition #1.

D. Incorrect - This is the action taken when the Reactor has gone critical below ZPDIL which is not the case. Since reactor is already critical actions of AOP 1A are to borate as needed and/or insert CEAs to control power if examinee assumes a boron dilution event is occurring.

Page 54 of 162

2012 NRC RO EXAM MASTER KEY Topic: Uncontrolled CEA Withdrawal Event

• Tier/Group: 1/2 001 - Continuous Rod Withdrawal

• AA2 - Ability to determine and interpret the following KIA Info:

as they apply to the Continuous Rod Withdrawal:

• AA2.04 - Reactor power and its trend RO Importance: 4.2 Proposed references to None be provided to applicant:

Given a CEA Malfunction the examinee will be able to Learning Objective: identify, understand the basis and take appropriate actions per plant operating procedures to mitigate the event.

10 CFR Part 55 Content: 55.41 (b}(1 O}

Question source:

Cognitive level: r:g] Comprehension/Analysis Last NRC Exam used on:

Exam Bank History: LOI-2006 Trip I Setpoint Criteria (09/08)

Technical references:

1C05-ALM, Reactivity Control Alarm Manual windows 0-15 Modified version of Q42232, enhanced stem statement and Comments: strengthened distractors a nd explanations to reflect 0-15 window response.

Page 55 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY Given that AOP-9A is implemented:

Which ONE of the following statements explains why the Fairbanks Morse Diesel Generators are shutdown?

A. To prevent overloading the Diesel Generators as equipment starts because the Shutdown Sequencers may be inoperabIE~.

B. To ensure fuel is conserved for continued extended operation of the 1A and DC DGs.

C. To prevent engine damage due to the non-essential trips being bypassed with an active UV signal.

D. The DC DG is aligned to power a Unit-1 and Unit-2 safety related 4KV bus simultaneously.

Answer: C Answer Explanation:

A. Incorrect - The ESFAS sequencers are powered from 12DVAC vital busses that are provided power from the DC busses.

B. Incorrect - Although tripping these DGs will conserve fuel oil they normally use, the 1A DG is not aligned as a power source, ONLY the DC DG is aligned to 11 and 24 4KV busses of each unit to provide power and it has its own fuel supply.

C. Correct - This is per AOP-9A Rev. 12 page 3 bases document.

D. Incorrect - This is a true statement in and of itself. It is not, however, the reason for securing the Fairbanks Morse Engines. Plausibility lies in the fact that it could make sense to run just one DG rather than two, given the higher load capacity of the SACM engine.

Page 56 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY AOP-9A bases for actions i Tier/Group: 1/2 067 - Plant Fire On-site

• AK3 - Knowledge of the reasons for the following KIA Info: responses as they apply to the Plant Fire on Site:

• AK3.04 - Actions contained in EOP for plant fire on site RO Importance: 3.3 Proposed references to None be provided to applicant:

Given AOP-9A and the Technical Bases, list the actions Learning Objective: performed by each watchstander and determine the bases for those actions.

10 CFR Part 55 Content: 55.41(b)(10)

Question source:

Cognitive level: l8J Memory/Fundamental o Comprehension/Analysis Last NRC Exam usedl on: I LOI-2004 RO (04/04)

Exam Bank History: No record of previous use Technical references: AOP-9A and Technical Bases document Add to bank Page 57 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following occurs on a loss of power to the Control Room ventilation RMS (O-RI-5350) monitor?

A. Control Room kitchen and toilet exhaust fan STOPS with gravity damper SHUT; BOTH post-LOCI filter fans START.

B. Control Room outside air supply and common exhaust dampers SHUT; BOTH post-LOCI filter fans START.

C. Operating Control Room HVAC outside air supply damper SHUTs; Selected post-LOCI filter fan STARTS.

D. Operating Control Room HVAC dampers shift to recirculation mode; Selected post-LOCI filter fan STARTS.

Answer: A Answer Explanation:

A. Correct - Since Control Room ventilation normal lineup is in a recirculation mode, only actions stated occur automatically.

B. Incorrect - Control Room outside air dampers are already shut as recirculation mode is the normal ventilation lineup.

C. Incorrect - All actions listed are wrong.

D. Incorrect - System is already in recirculation mode per OI-22F. Loss of power causes both Post-LOCI filter fans to start not just one fan to start.

2012 NRC RO EXAM MP\STER KEY Topic: Loss of power to Control Room Ventilation RMS Tier/Group: 1/2 061 - ARM System Alarms

• AA1 - Ability to operate and / or monitor the following as KIA Info: they apply to the Area Radiation Monitoring (ARM)

System Alarms:

• AA1.01 - Automatic actuation RO Importance: 3.6 Proposed references to be None provided to applicant:

DESCRIBE the design features that provide for the following during operation of the CR Ventilation and Chilled Learning Objective: Water System:

• 100% recirculation during high radiation conditions (or loss of power to RMS) 10 CFR Part 55 Content: 55.41 (b)(10)

Question source:

Cognitive level: L8J Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI-2010 1C22/34 exam (09/11) 01-35, Radiation Monitoring System Technical references:

01-22F, Control Rm and Cable Spreading Rm Vent Comments: IMctdifiE3d flrom Q24745 Page 59 of 162 Rev. 3

2012 NRC RO EXAM Mt\STER KEY During a Unit-2 initial startup after refueling, the following exist:

• Power was stabilized at 30% for NI CAL

• CEA withdrawal recommences to raise power when a Regulating Group 4 CEA drops fully to the bottom Which ONE of the following actions is the initial response required?

A. Commence realignment of the dropped CEA and borate the RCS as needed to keep power constant.

B. Adjust turbine load to maintain T COLD on program and stop reactor power from continuing to lower.

C. Withdraw remaining CEAs in steps as needed to maintain TCOLD on program and stabilize reactor power.

D. Initiate boration to counter effects of TCOLD lowering causing reactor power to rise above level prior to CEA drop.

Answer: B Answer Explanations:

A. Incorrect - This action is not the initial response per AOP-1 B.

This is only done after plant is stabilized by adjusting turbine load, TBVs or ADVs, or initiating boration.

B. Correct - At BOL a positive MTC exists. A dropped CEA adds negative reactivity causing T COLD to lower which with a positive MTC will add more negative reactivity causing T COLD to continue to lower. Lowering turbine load will stabilize T COLD and reactor power.

C. Incorrect - CEAs will add positive reactivity to compensate for TCOLD continuing to lower with positive MTC. However, CEAs shall NOT be used to control T COLD per caution of AOP-1 B.

D. Incorrect - If examinee doesn't recognize positive MTC exists (adds negative reactivity and causes T COLD to continue to lower) may assume power will rise due to drop in TCOLD and initiate boration to prevent reactor power from exceeding power level prior to CEA drop.

Page 60 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY i Dropped CEA actions at power and SOL Tier/Group: 1/2 003 Dropped Control Rod I 1 AK1 - Knowledge of the operational implications of the KIA Info:

following concepts as they apply to Dropped Control Rod:

• AK1.16 - MTC RO Importance: 2.9 Proposed references to be provided to applicant:

Learning Objective:

10 CFR Part 55 Content:

Question source:

Cognitive level: ~ Comprehension/Analysis Last NRC Exam used on:

Modified from Q37647 Page 61 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given an RCS fill evolution in progress during Shutdown Cooling Ops:

(1) Which ONE of the following criteria defines adequate mixing and (2) Indicates a boron dilution event may be in progress?

A. (1) At least 1500 GPM flow thru the core AND at least 500 GPM flow through at least one S/G; (2) Unexpected slow rise in SDC temperature.

B. (1) At least 1500 GPM flow thru the core AND at least 500 GPM flow through both S/Gs; (2) Audible countrate, in the Control Room, increases.

C. (1) At least 3000 GPM flow thru the core AND at least 500 GPM flow throLlgh at least one S/G; (2) An unexpected rise in RCS level.

D. (1) At least 3000 GPM flow thru the core AND at least 500 GPM flow through both S/Gs; (2) An unexpected rise in countrate on Nuclear Instrumentation.

Answer: D Answer Explanation:

A. Incorrect - Both parts are wrong ... 1500 GPM is the minimum SDC flow surveillance requirement for Mode 6 (logged in CRO logs), the 500 GPM flow is requirement per S/G. SDC temperature would not rise during any boron dilution event when shutdown. If anything it will lower as fill water is added and operator needs to adjust to maintain temperature.

B. Incorrect - 1500 GPM is the minimum SDC flow surveillance requirement for Mode 6 (logged in CRO logs), the 500 GPM flow is requirement per S/G; audible popper noise becoming more frequent indicates counts are rising on Nls which could be expected during inadvertent dilution event.

C. Incorrect - Flow thru core is right but flow must be thru both S/Gs not just one.

Rise in RCS level compared to change in RWT level may indicate another source of water is entering RCS causing an inadvertent dilution event.

D. Correct - Flow thru core and S/Gs meets criteria; NI counts rising is an indication that a boron dilution event may be occurring.

Page 62 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Reactivity effects of SDC fill water into RCS Tier/Group: 2/1 005 Residual Heat Removal System (RHRS)

K5 - Knowledge of the operational implications of the

• KIA Info:

following concepts as they apply the RHRS:

• K5.03 - Reactivity effects of RHR fill water RO Importance: 2.9 Proposed references to None be provided to applicant:

Identify RCS dilution limitation including requirements for Learning Objective:

adequate mixing.

10 CFR Part 55 Content: 55.41 (b)(5)

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use on any exam

• Exam Bank History:

Technical references: AOP-1A, Inadvertent Dilution Event OP-7, Shutdown Operations Comments: Modified from Q25960 Page 63 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following condition on Unit-2 in Mode 4:

• The SDC header is placed in recirculation through the SIT recirculation leak-off isolation valves, 2-SI-463 and 2-SI-455, flow path return to the RWT Which ONE of the following procedural controls ensures Containment Integrity is reestablished?

A. An approved Contingency Plan is activated to re**establish Containment Integrity when the evolution has been completed.

B. Containment Closure Deviation Sheets are used to ensure Containment Integrity is re-established.

C. A dedicated operator is stationed in continuous communication with the Control Room to restore valves to locked shut condition.

D. A Component Manipulation Form (CMF) is completed and closed out when the lineup is secured.

Answer: C Answer Explanation:

A. Incorrect - Contingency plans are used in lower modes to address plant situations where maintenance or equipment situcJttions challenge the MEEL or for High Risk evolutions.

B. Incorrect - Containment Closure Deviation tracking sheets are only used in Modes 5 and 6.

C. Correct - These are manual valves administratively controlled per NO-1-205 and per T.S.3.6.3. Per 01-3B there is a step to open these valves then shut and lock them by stationing a dedicated operator in continuous communication when this path is in use.

D. Incorrect - This activity is covered by an Operating Instruction. A CMF is not required.

Page 64 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY Topic: Containment Isolation valves Tier/Group: 2/1 069 Loss of Containment Integrity /5 AA1. Ability to operate and / or monitor the following as they KIA Info:

apply to the Loss of Containment Integrity:

• AA1.03 - Fluid systems penetrating containment RO Importance: 2.9 Proposed references to None be provided to applicant:

Recall the actions established to take per NO-1-205 when Learning Objective: operating administratively controlled containment isolation valves during Recirc of SDC to RWT.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: ~ Memory/Fundamental Last NRC Exam used on: LOI-2002 RO (07/02)

Exam 8ank History: LOI-2006 RO/SRO Audit Remediation Exam (05/08) 01-38, Section 6.2 Recirculation of SDC header Technical references:

NO-1-114, Containment Closure Comments: None Page 6S of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which of the following statements describes the relationship of RCS activity to the Process Radiation Monitor, 1(2)-RI-202, installed in the Letdown line sample path?

A. ONLY RCS gross activity is monitored.

B. ONLY activity associated with a specific isotope related to fuel failure events is monitored.

C. Gross activity and activity associated with a specific isotope related to fuel failure events are monitored.

D. Gross activity and activity associated with a specific isotope related to fuel failure events are monitored continuously, during all accident conditions.

Answer: C Answer Explanation:

A. Incorrect - The PRM detects both gross activity and activity associated with a specific isotope.

B. Incorrect - The PRM detects both gross activity and activity associated with a specific isotope.

C. Correct - This is the purpose of the monitor in the letdown system. If both are increasing it identifies that a fuel failure event is occurring rather than just a crud burst.

D. Incorrect - The PRM detects both gross activity and activity associated with a specific isotope. However, the PRM is isolatE~d when SIAS is actuated and can no longer be relied upon.

Page 66 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Process Rad Monitor relationship to RCS activity Tier/Group: 2/1 076 High Reactor Coolant Activity /9

• AK2. Knowledge of the interrelations between the High KIA Info:

Reactor Coolant Activity and the following:

• AK2.01 - Process radiation monitors RO Importance: 2.6 Proposed references to None be provided to applicant:

Learning Objective: Identify the purpose of the Process Radiation Monitor.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: ~ Memory/ Fundamental 0 Comprehension/Analysis Last NRC Exam used on: LOI-2004 RO (02/04)

Exam Bank History: None Technical references: System Description (SD) 041 - CVCS Comments: Modified from Q14577 Page 67 of 162 Rev. 3

2012 NRC RO EXAM MJ\STER KEY Unit-2 is in Mode 5 with the following conditions:

• The plant has been shut down for 3 days 0

• RCS temperature is 130 F

• RCS is at atmospheric pressure with the Pzr manway installed

• Pressurizer level is 120" and lowering at approximately 20 inches per minute

• The appropriate AOP has been entered Which ONE of the following actions will restore RCS level in accordance with the AOP?

A. Start all available Charging pumps.

8. Open a LPSI Pump normal suction valve.

C. Start a HPSI pump and throttle flow to less than 210 GPM.

o. Start a HPSI pump and maintain RCS pressure less than 260 PSIA.

Answer: 0 Answer Explanation:

A. Incorrect - Starting all available Charging Pumps is the first step specified in AOP-38. However, information provided, in the stem of the question, indicates Charging Pumps alone will not restore RCS level at this leak rate requiring that a HPSI Pump be started per Att. 7, Filling the RCS.

B. Incorrect - While opening the LPSI Pp Normal Suction may supply makeup water to the RCS (if the RWT Outlet MOV is open), AOP-38 does not direct this action. This action requires local operator action outside of the control room.

C. Incorrect - Per AOP-38 Attachment 7, flow into the RCS is limited to less than 210 GPM unless a leak exists. Indications of a leak are provided in the stem of the question.

A. Correct - Per AOP-38 Attachment 7, Filling the RCS: When RCS temperature is 0

less than 365 F AND the RCS vent opening is less than 2.6 square inches, flow into the RCS is limited to less than 210 GPM unless a leak exists. If a leak exists, flow may exceed 210 GPM as long as pressure is maintained less than 380 PSIA (or 260 PSIA if the SOC Header Return Isolation valves, 1-SI-651-MOVand 1-SI 652-MOV, are open).

Page 68 of 162

2012 NRC RO EXAM Mf~STER KEY Topic: IRCS leakage into CCW and cannot be isolated Tier/Group: 1/2 CE/A 16 - Excess RCS Leakage

• 2.1 - Conduct of Operations KiA Info:

• 2.1.20 - Ability to interpret and execute procedure steps.

RO Importance: 4.6

. Proposed references to None be provided to applicant:

Given various AOPs and bases documents with a set of

. Learning Objective: plant conditions, navigate the procedures correctly to mitigate the effects of various malfunctions 10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: o Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: No record of use on any exam Technical references: AOP-3B, Abnormal Shutdown Cooling Conditions IComments: None Page 69 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY The Miscellaneous Waste Monitor Tank (MWMT) is being discharged per an approved release permit when the LlQUI D WASTE DISCH RMS monitor, O-RIC 2201, alarms high. Upon investigation, the Control Room observes the LIQUID WASTE DISCH CVs, 0-MWS-2201-CV and 0-MWS-2202-CV, have not shut a utomatica lIy.

Which ONE of the following is the expected operator response?

A. Verify valves O-MWS-2201-CV and O-MWS-2202-CV shut.

B. Stop the MWMT pump being used to discharge the MWMT.

C. Ensure valves 0-MWS-103 and O-MWS-105 are shut to isolate the Unit-2 SG Blowdown overboard discharge path.

D. Continue discharge of MWMT using the procedure for O-RE-2201 not available and energized.

Answer: A Answer Explanation:

A. Correct - Per 1C22-ALM, RMS Alarm Manual, this is the appropriate action.

Verify means to make it happen if it hasn't. In this case, placing the handswitches for O-MWS-2201-CV and O-MWS-2202-CV in close would cause the valves to shut, terminating the accidental liquid waste release.

B. Incorrect - This action is specified by AOP-6B, Accidental Liquid Waste Release, if the discharge CVs fail to shut when handswitches placed to shut position.

C. Incorrect -This action is specified by AOP-68, Accidental Liquid Waste Release, if the discharge CVs fail to shut when handswitches placed to shut position.

D. Incorrect - This action per Alarm Manual due to an RMS failure. Question Stem stated due to high alarm.

Page 70 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY nl6Q~lnn 30~(g97,019) 7:::'"~':'7 Topic: Liquid Waste Monitor, 0-RIC-2201, automatic actions Tier/Group: 1/2 059 Accidental Liquid RadVVaste Release / 9 AK2 - Knowledge of the intE~rrelations between the KIA Info:

Accidental Liquid Radwaste Release and the following:

• AK2.01 - Radioactive-liquid monitors RO Importance: 2.7 Proposed references to None be provided to applicant:

Determine the automatic actions upon a high alarm on 0 Learning Objective:

RIC-2201.

10 CFR Part 55 Content: 55.41 (b)(7)

Cognitive level: [g] Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: None AOP-SB, Accidental Liquid Waste Release Technical references:

1C22 Alarm Response Manual Window 032 Comments: Modified from Page 71 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which one of the following conditions would require the implementation of EOP-B, Functional Recovery procedure?

A. Reactivity Control safety function cannot be met in EOP-O due to no power available to CEA indications.

B. A loss of offsile power results in a reactor trip, and the EOP-O flowchart recommends EOP-6 implementation.

C. The EOP-O flowchart recommends implementing both EOP-3 and EOP-7 and single event diagnosis is not possible.

D. EOP-4 is implemented but the Final Safety Function Acceptance Criteria is not being met.

Answer: C Answer Explanation:

A. Incorrect - The EOP-O Diagnostic flowchart would recommend considering EOP-7, Station Blackout in this case.

S. Incorrect - The EOP-O Diagnostic flowchart would recommend considering EOP-2 and EOP-6, Steam Generator Tube Rupture. EOP-6 is written to address a LOOP coincident with a SGTR.

C. Correct - These are the conditions needed to enter EOP-B.

D. Incorrect - Final acceptance criteria not bein~1 met is incorrect. EOP-B would be implemented if the Intermediate Safety Function Status Check(s) is/are not met.

Page 72 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: EOP-8 entry conditions Tier/Group: 1/2 CE/E09 - Functional Recovery

• EK 1- Knowledge of the operational implications of the following concepts as they apply to the KIA Info: (Functional Recovery)

• EK1.2 - Normal, abnormal and emergency operating procedures associated with (Functional Recovery)

RO Importance: 3.2 Proposed references to be None provided to applicant:

Learning Objective: Determine conditions when EOP-8 may be entered 10 CFR Part 55 Content: 55.41 (b)(10)

Question source:

[SJ Memory or Fundamental Cognitive level:

D Comprehension or Analysis Last NRC Exam used on: 2010 RO Recertification Test Exam Bank History: LOI-2006 SRO practice (03/08)

Technical references: EOP-8, Functional Recovery Procedure Comments: None Page 73 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY RCS boration is in progress using the blended batch makeup mode flowpath when a loss of instrument air occurs.

Which ONE of the following flow paths remains available to continue boration?

A. Borate Makeup Mode to the charging pump suction in MODE 1.

B. Manual Makeup mode to the VCT (CVCS).

C. Borate Makeup Mode to the VCT (CVCS).

D. RWT to charging pump suction.

Answer: D Answer Explanation:

A. Incorrect - This flowpath remains unavailable as it requires the use of DI water FIC-210X-CVand Boric Acid Flow FIC-210Y-CVand CVC-512-CV (VCT Inlet) which failed closed based on the loss of IA occurring per the question stem.

B. Incorrect - This flowpath remains unavailable as it requires the use of the Boric Acid Flow and DI Water Flow control valves, FIC**210X-CV and FIC-210Y-CV, and CVC-512-CV (VCT Inlet) which failed closed based on the loss of IA occurring per the question stem.

C. Incorrect - This flowpath remains unavailable as it requires the use of the Boric Acid Flow control valve, FIC-210Y-CV and CVC-512-CV (VCT Inlet) which failed closed based on the loss of IA per the question stem.

D. Correct -This flowpath uses only an MOV which requires no air to operate only power available and a manual isolation valve locked open directly to charging pump suction.

Page 74 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY s of IA effects on when borating to VCT Tier/Group: 2/1 004 - Chemical and Volume Control

• K6 - Knowledge of the effect of a loss or malfunction on.

KIA Info: the following CVCS components: .

• K6.13 - Purpose and function of the boration/dilution batch controller RO Importance: 3.1 Proposed references to be provided to applicant:

i Explain how CVCS responds to the following conditions:

Learning Objective:

Loss of Instrument Air 10 CFR Part 55 Content: 55.41 (b){7)

Question source:

Cognitive level: ~ Memory/ Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use

• Technical references: .01-2B, CVCS Boration, Dilution, and Makeup Operations

'page 70 AOP-1A, Inadvertant Boron Dilution Attachment 1 pages 2, 3, &4 Comments: Modified from Q20598 Page 75 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following plant conditions:

• Unit-2 is in Mode 5

• Shutdown Cooling (SOC) is in operation

• S/G Nozzle Dams are installed

• A Component Cooling leak requires placing ALL Component Cooling Pumps in Pull To Lock Which ONE of the following actions is required per the applicable Tech Spec?

A. Initiate action to restore both SOC loops to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least one in operation immediately.

B. Initiate action to restore one SOC loop to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or align spent fuel pool cooling to supplement shutdown cooling.

C. Initiate action to restore one SOC loop to operable status immediately or initiate action to restore the SGs to operable status.

O. Initiate action to restore one SOC loop to operable status and operation immediately.

Answer: 0 Answer Explanation:

A. Incorrect - Examinee may expect an hour is allowed to initiate action to return loop to operable status. Based on conditions, S/Gs are unavailable as nozzle dams installed, therefore, RCS loops not filled and TS LCO 3.4.8 Action B is applicable.

B. Incorrect - Examinee may expect an hour is allowed to initiate action to return loop to operable status; however, no provision is made for use of SFP cooling to supplement SOC. Based on conditions, S/Gs are unavailable as nozzle dams installed, therefore, RCS loops not filled and TS LCO 3.4.8 Action B is applicable.

C. Incorrect - Examinee may know immediate action is required to return loop to operable status; however, no provision is made in the LCO for returning the S/Gs to operable status as the RCS loop(s) are not filled.

O. Correct - Per LCO 3.4.8 Action B is the required operator response.

Page 38 of 162 Rev. 3

12 NRC RO EXAM MP\STER KEY Topic: Knowledge of T.S. LCOs of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less Tier/Group: 1/1 025 Loss of RHR System 14

• AK2 - Knowledge of the interrelations between the Loss of Residual Heat Removal System and the KIA Info:

following:

• AK2.03 - Service water or closed cooling water pumps RO Importance: 2.7 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: r.sJ Memory/Fundamental o Comprehension/Analysis Last NRC Exam used on: No record of previous use Exam Bank History: None Technical references: Tech Spec 3.4.8 Comments: Enhanced question by adding bullets for conditions and ensured each distractor had same wording Page 39 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Unit-1 is in Mode 6 with the RCS drained to the 37.5 Foot elevation.

Which ONE of the following prerequisites must be established prior to shifting LPSI pumps in this condition?

A. Raise RCS level to at least 38 feet and Reduce SOC flow to 800 GPM.

B. Adjust 1-FIC-306 and 1-HIC-3657 to maintain SOC flow at 1500 GPM.

C. Reduce SOC flow to 800 GPM using 1-FIC-306 and 1-HIC-3657.

O. Verify the designated LPSI header MOVs are throttled to limit SOC flow to 1700 GPM if a loss of power occurs to 1-SI-306-CV.

Answer: C Answer Explanation:

A. Incorrect - Raising the RCS level to 38' would place the plant in a condition where the idle LPSI Pp could be started and the previously running LPSI Pp could be stopped without SOC flow limitations.

B. Incorrect - This prerequisite is associated with SOC flowrate limitations prior to draining the RCS to below the 37.6 ft. elevation.

C. Correct 3B states: PLACE the SOC FLOW CONTR, 1-FIC-306 in MANUAL ANO REDUCE SDC Flow to approximately 800 GPM by adjusting the SDC FLOW CONTR, 1-FIC-306 and SDC TEMP CONTR, 1-HIC-3657.

O. Incorrect - 1700 GPM is the limit, established in OP-7, when the UGS is installed, to prevent damage to the ICI thimbles.

Page 76 of 162 Rev. 3

2012 NRC RO EXAM MP~STER KEY Topic: Determine the prerequisites for shifting LPSI pumps.

Tier/Group: 2/1 005 - Residual Heat Removal System (RHRS)

• A4 - Ability to manually operate and/or monitor in the KIA Info:

control room:

• A4.02 - Heat exchanger bypass flow control RO Importance: 3.4 Proposed references to None be provided to applicant:

i Learning Objective:

10 CFR Part 55 Content:

• Cognitive level: D Comprehension/Analysis Last NRC Exam used on: LOI-2010 1C08, 09, & 10 mid-term Exam Bank History: None 01-3B, Shutdown Cooling Technical references:

OP-7, Shutdown Operations Comments: None Page 77 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY When initiating SOC, OP-5 specifies the RCS cooldown should be stopped and new baseline temperatures obtained after SOC is initiated.

Which ONE of the following is the reason for obtaining new baseline temperature data?

A. The indicated temperature difference between the SOC HX outlets and the hot legs prevents accurate cooldown determination when SOC is initiated.

B. The indicated temperature difference between the hot and cold legs prevents accurate cooldown determination when SOC is initiated.

C. The indicated temperature difference between the hot legs prevents accurate cooldown determination when SOC is initiated.

O. The indicated temperature difference between the CETs and TR-351 prevents accurate cooldown determination when SOC is initiated.

Answer: 0 Answer Explanation:

A. Incorrect - This is accurate once flow is directed thru the SOC HXs and this is the return temperature to the RCS.

B. Incorrect - The temperature difference between hot and cold legs is accurate once natural circulation has been established after the RCPs are secured and prior to SOC initiation.

C. Incorrect - Although SOC suction is only from one hot leg, there is no temperature difference observed between hot legs.

O. Correct -OP-5 states: "Due to the indicated temperature differential between the CETs and 1-TR-351, accurate cooldown determination is not available when SOC is initiated. When initiating SOC, the cooldown should be stopped and new baseline temperatures obtained after SOC is initiated.

Page 78 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Why new set of baseline data is obtained prior to Topic:

reinitiating SOC

• Tier/Group: 2/1 006 Emergency Core Cooling K5 - Knowledge of the operational implications of the following concepts as they apply to ECCS:

KIA Info:

• K5.07 - Expected temperature levels in various locations of the RCS due to various plant conditions RO Importance: 2.7 Proposed references to be None provided to applicant:

Initiate SOC during a plant cooldown upon securing Learning Objective:

RCPs 10 CFR Part 55 Content: 55.41 (b)(5)

~ Memory or Fundamental Cognitive level:

D Comprehension or Analysis Last NRC Exam used on: No record of use on any exam None Technical references: OP-5, Section 6.3 step G note, page 41 None Page 79 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY A loss of load transient resulted in a Unit-1 trip that reached a maximum RCS pressure of 2425 PSIA. The following conditions exist:

• "CNTMT NORMAL SUMP LVL HI" annunciator in alarm

• "QUENCH TK *TEMp*LVL*PRESS" annunciator in alarm Which ONE of the following has occurred? Assume NO other events in progress A. Quench Tank Rupture Disk is failed.

B. Reactor Coolant Drain Tank header relief valve lifted.

C. A Reactor Coolant System Code Safety Valve lifted.

D. The Quench Tank Drain valve, 1-RC-401-CV, leaked by.

Answer: A Answer Explanation:

A. Correct - The sump alarm with the quench tank alarm indicates that the rupture disk has failed as quench tank would then discharge to the containment normal sump.

B. Incorrect - This valve relieves to the quench tank not the containment sump.

Additionally, temperature of water in the RCDT is cooled by CCW and is below RCS temperature and would not cause rupture disk barrier to be lost if it continued to relieve to quench tank.

C. Incorrect - Question stem stated that RCS pressure peaked at 2425 PSIA; therefore, neither code safety valve would have lifted. Code safety lift setpoints are between ~ 2475 and :s 2525 PSIA for RV-200 and between ~ 2540 and

s 2590 PSIA for RV-201. Also these valves initially discharge to the Quench Tank not the Containment Sump.

D. Incorrect - The Quench Tank drain valve when opened is connected to the RC Drain Tank. It does not go directly to the Containment Sump.

Page 80 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY IQuench Tank rupture disk failing

. Tier/Group: 2/1 007 - Pressurizer Relief Tank/Quench Tank System

• K1 - Knowledge of the physical connections and/or KIA Info: cause/effect relationships between the PRTS and the following systems:

• K1.03 - RCS i

. RO Importance:

IProposed references to None i be provided to applicant:

Identify indications that a quench tank rupture disk has Learning Objective:

failed.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: o Memory/ Fundamental [gI Comprehension/ Analysis Last NRC Exam used on: NEW Tech Spec 3.4.10 - Pressurizer Safety Valves Page 81 of 162 Rev. 3

j 1C06-ALM RCS CONTROL ALARM MANUAL Rev. 49/Unit 1 Page 6 of 144 DEVICE SETPOINT WINDOW E-01 1-TIA-116 120°F (117to 123°F) 1-PIA-116 10 PSIG (9 to 11 PSIG)

QUENCHTK 1-LlA-116 30.5 in. (29.5 to 31.5 in,)(High)

.TEMP.LVL 26,S in. (25,S to 27,S in,)(Low)

• PRESS POSSIBLE CAUSES

• Lifting or leaking:

• Reactor Coolant System safety valve(s), 1-RC-200-RV or 1-RC-201-RV

• Reactor Coolant System PORV(s), 1-RC-402-ERVor 1-RC-404-ERV

• Safety Injection System recirculation line relief valve, 1-SI-446-RV

• Reactor Coolant Drain Tank header relief valve, 1-RCW-4252,HV

• Leaking or open:

• Demineralized water valve, 1-DW-S460-CV

• Nitrogen supply valve, O-N2-238

• Drain to Reactor Coolant Drain Tank, 1-RC-401-CV

• Sample valve, 1-PS-6531-CV

• Pressurizer vent solenoid valves, 1-RC-105-SV and 1-RC-106-SV

• Reactor Vessel vent solenoid valves, 1-RC-103-SV and 1-RC-1 04-SV AUTOMA'*IC ACTIONS None (continued)

1C06-ALM RCS CONTROL ALARM MANUAL Rev. 49/Unit 1 Page 7 of 144 (continued) WINDOW E-01 CONDITION RESPONSE

1. Quench Tank parameter in alarm. 1. Perform the following:

a. SHUT any open valves listed under leaking or open Possible Causes.

b. IF a PORV is leaking or open and fails to shut when RCS pressure is reduced below its lift setpoint, THEN CONSIDER placing the applicable PORV Override handswitch, 1-HS-1402 or 1-HS-1404, in OVERRIDE, OR SHUT PORV Block, 1-RC-403-MOV or 1-RC-405-MOV.

c. RETURN parameter to within normal limits by venting, filling, draining or feed and bleed as neCE!Ssary PER 01-1 B, Quench Tank OpElrations.

d. REFER to Technical Specifications 3.4.11 and 3.4.12 for PORV operability requirements.

ANNUNCIATOR COMPENSATORY ACTIONS

• MONITOR Quench Tank parameters at least hourly

• IF alarm card E-01 is removed from the alarm panel in the cable spreading room, THEN LOG the following alarm windows out of service: E-05, E-07, E-08, E-29, E-33, E-34, E-35 and E-36 REFERENCES None

Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Two pressurizer safety valves shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with all RCS cold leg temperatures> 365°F (Unit 1),

> 301°F (Unit 2).

---

NOTE ---------------------------

The lift settings are not required to be within Limiting Condition for Operation limits during MODE 3 > 365°F (Unit I), > 301°F (Unit 2) for the purpose of setting the pressurizer safety valves under ambient (hot) conditions.

This exception is allowed for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following entry into MODE 3 > 365°F (Unit I), > 301°F (Unit 2) provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer A.l Restore valve to 15 minutes safety valve OPERABLE status.

inoperable.

CALVERT CLIFFS - UNIT 1 3.4.10-1 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201

Pressurizer Safety Valves 3.4.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Reduce all RCS cold 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> leg temperatures to Two pressurizer ~ 365°F (Unit I),

safety valves ~ 30 (Unit 2).

inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is In accordance OPERABLE in accordance with the Inservice with the Testing Program. The lift settings shall be Inservice within limits as specified below: Testing Program As Found As Left Valve lift Setting (psia) Lift Setting (psia)

RC-200 ~ 2475 and ~ 2550 ~ 2475 and ~ 2525 RC-201 ~ 2514 and ~ 2616 ~ 2540 and ~ 2590 CALVERT CLIFFS - UNIT 1 3.4.10-2 Amendment No. 227 CALVERT CLIFFS - UNIT 2 Amendment No. 201

2012 NRC RO EXAM MASTER KEY Given containment pressure on Unit-2 has reached 5.0 PSIG during an event, which ONE of the following valves requires manual action to close if open?

A. IA CONTAINMENT ISOLATION, 2-IA-2080-MOV B. RCS SAMPLE ISOL valve, 2-PS-5464-CV C. OW CNTMT ISOL valve, 2-0W-5460-CV O. SRW SUPP TO 22 BO HX, 2-SRW-1640-CV Answer: C A. Incorrect - IA CONTAINMENT ISOLATION, 2-IA-2080-MOV automatically closes on receipt of a CIS (Containment Pressure greater than 2.8 PSIG).

B. Incorrect - RCS SAMPLE ISOL valve, 2-PS-5464-CV automatically closes on receipt of a SIAS (Containment Pressure greater than 2.8 PSIG or RCS pressure less than 1725 PSIA).

C. Correct - This valve receives no automatic ESFAS signal to close. It is an administratively controlled valve. CIS verification checklist (EOP Att. 4 Page 1) directs shutting this valve if open.

O. Incorrect - SRW SUPP TO 22 BO HX, 2-SRW-1640-CV automatically closes on receipt of a CSAS (Containment Pressure greater than 4.25 PSIG).

2012 NRC RO EXAM MASTER KEY Topic: Restore quench tank temperature Tier/Group: 2/1 007 Pressurizer Relief TanklQuench Tank System

• A4 - Ability to manually ()perate and/or monitor in the KIA Info:

control room:

• A4.01 - PRT spray supply valve RO Importance: 2.7 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: IZI Memory/FundamentaJ ILJ Last NRC Exam used on: No record of use on any exam Exam Bank History: None Technical references: EOP Attachments 2,3, and 4 Comments: None Page 83 of 162 Rev. 3

12 NRC RO EXAM MJ\STER KEY A loss of offsite power has occurred causing the Unit-2 reactor to trip. Given the following:

• Prior to the trip, 21 Component Cooling pump was running

• RCS pressure is 1700 PSIA and slowly lowering

• 480V bus 24A is deenergized

• No operator actions have been performed for the Vital Auxiliaries safety function Which combination of annunciator windows, in alarm, indicate that 23 Component Cooling pump is operating?

A. "ACTUATION SYS SIAS TRIP" and "ccw PPS

• SIAS BLOCKED

• AUTO START" B. "cc PPS DISCH PRESS LO" and "ACTUATION SIGNAL BLOCKED" C. "U-2 4KV ESF MOTOR OVERLOAD" and "SEQUENCER INITIATED" D. "U-2 480V ESF UN TRIP" and "23 CC PP BKR LlU IMPR" Answer: A Answer Explanation:

A. Correct - PZR pressure at 1700 PSIA generates a SIAS signal sent to ALL 3 CCW Pump starting circuits and 480V bus 24A deenergized provides a UN condition to actuate the second alarm. This bus also supplies power to 22 CCW Pump which is unavailable, therefore, 23 CCW Pump will start (since normally aligned to 480V bus 24B) after one second upon receipt of SIAS signal when 22 CCW Pump fails to start.

S. Incorrect - The first alarm indicates NO CCW Pumps are running and second alarm occurs upon a LOOP and SIAS indicating all LOCI sequencer steps have not been completed for train A and/or B (this alarm clears when all steps timeout once power is restored).

C. Incorrect - The Second alarm occurs for conditions stated in stem statement.

First alarm is wrong as CCW Pumps are 480V loads not 4KV loads.

D. Incorrect - The first alarm will occur when 480V bus 24A is deenergized. The second alarm indicates 23 CCW Pp is different from the standard lineup of one breaker racked in with its associated disconnect shut.

Page 84 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY Topic: The "standby" feature for CCW pumps Tier/Group: 2/1 008 Component Cooling Water System

• K4 - Knowledge of CCV\fS design feature(s) and/or KIA Info:

interlock(s) which provide for the following:

• K4.09 - The "standby" feature for the CCW pumps RO Importance: 2.7 Proposed references to None be provided to applicant:

Given plant conditions, detElrmine the status of "standby" I Learning Objective:

CCWpump 10 CFR Part 55 Content: 55.41 (b)(7)

Cognitive level: D Memory/Fundamental [S] Comprehension/Analysis Last NRC Exam used on: No record of use on any exam LOI-2008 RO Audit (05/10)

Technical references: 2C08-ALM, ESFAS 21 Alarm Manual 2C13-ALM, SRW and Misc Station Services Alarm Manual 2C17-ALM, 4KV & 480V Normal FOR BKR Alarm Manual 1C19-ALM, 13KV & 4KV Essential Feeder Bkrs Control Board Alarm Manual Comments: Modified from Q92262

2012 NRC RO EXAM MJ~STER KEY A Loss of Offsite Power has occurred with the 1B Diesel Generator failing to start.

Assuming no electrical buses are tied, which of the following is correct?

A. Pressurizer backup heater banks 1 and 3 are available from 1C43 only.

B. Pressurizer backup heater bank 1 is available from 1C43 only.

C. Pressurizer backup heater banks 1 and 3 are available from 1C06 and 1C43.

D. Pressurizer backup heater bank 3 is available from 1C43 only.

Answer: B Answer Explanation:

A. Incorrect - Pressurizer Backup Heater banks 1 & 3 are powered from 480V Busses 11 Band 14B respectively. The stem states the 1B DG failed to start which means Pressurizer Backup Heater bank ~~ is NOT available.

B. Correct .. Pressurizer Backup Heater banks 1 & 3 are powered from 480V Busses 11 Band 14B respectively. The stem states the 1B DG failed to start which means only Pressurizer Backup Heater bank 1 is available. Because 1Y10 de-energizes, as a result of the 1B DG Start Failure, all Pressurizer Heaters receive a signal to turn off. Operation of the Pressurizer Heater(s) under these conditions requires transferring control to 1C43 via a local keyswitch.

C. Incorrect - Pressurizer Backup Heater banks 1 & 3 are powered from 480V Busses 1 'I Band 14B respectively. The stem states the 1B DG failed to start which means Pressurizer Backup Heater bank 3 is NOT available. Because 1Y10 de-energizes, as a result of the 1B DG Start Failure, all Pressurizer Heaters receive a signal to turn off. Operation of the Pressurizer Heater(s) under these conditions requires transferring control to 1C43 via a local keyswitch.

D. Incorrect - Pressurizer Backup Heater banks 1 & 3 are powered from 480V Busses 11 Band 14B respectively. The stem states the 1B DG failed to start which means Pressurizer Backup Heater bank 3 is NOT available.

Page 86 of 162 Rev. 3

2012 NRC RO EXAM M,l~STER KEY EDG power to the pressurizer heaters Tier/Group: 2/1 010 - Pressurizer Pressure Control KIA Info:

• K2 - Knowledge of bus power supplies to the following:

• K2.04 - Pzr Heaters RO Importance: 3.0 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.41 (b )(7)

Question source:

Cognitive level:

Last NRC Exam used on:

I Technical references: AOP-71, Loss Of 4kv, 480 Volt Or 208/120 Volt Instrument

Bus Power Comments: None Page 87 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Unit-1 is at 100% power when a malfunction occurs in RPS Channel B causing the Power Trip Test Interlock (PTTI) to actuate.

Which ONE of the following describes the effect on the Control Rod Drive System?

A. 1 of 2 required pre-trips from VOPT .Q! APD for CEA Motion Inhibit is met.

B. 1 of 4 required pre-trips from APD or HI-SUR for CEA Withdrawal Prohibit is met.

C. 1 of 2 required pre-trips from VOPT or TM/LP for CEA Withdrawal Prohibit is met.

D. 1 of 4 required pre-trips for HI SUR or TM/LP for CEA Motion Inhibit is met.

Answer: C Answer Explanation:

A. Incorrect - These trips occur in RPS from PTTI; however, they do NOT provide an input to CMI circuit for CEAs.

B. Incorrect - ONLY APD is tripped due to PTTI; however, APD does not provide an input into CWP circuit to prevent outward movement of CEAs.

C. Correct - BOTH of these trips occur from PTTI and each provides 1 of 2 required pre-trips to CWP circuit to prevent outward movement of CEAs.

D. Incorrect - ONLY TM/LP is tripped on PTTI; however, it does not provide an input to CMI circuit for CEAs.

Page 88 of 162 Rev. 3

2012 NRC RO EXAM MP,STER KEY Topic: Effect of PTTI occurring to CEAs Tier/Group: 2/1 012 Reactor Protection

• K3 - Knowledge of the effect that a loss or malfunction of KIA Info:

the RPS will have on the following:

• K3.01 - CRDS RO Importance: 3.9 Proposed references to None be provided to applicant:

Determine the effect on the Control Rod Drive System when Learning Objective:

the Power Trip Test Interlock occurs.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: D Memory/ Fundamental IZI Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: None 1C05-ALM, Reactivity Control Alarm Panel Technical references:

SD-058, Reactor Protective System Description Comments: None Page 89 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY During a Unit -1 power escalation lAW OP-3, the annunciator window "HI POWER TRIP RESET DEMAND" alarm is received on 1C05.

(1) What condition has caused the alarm to actuate?

(2) What are the consequences of taking NO actions for this alarm?

A (1) Actual Reactor Power is 8.4% away from the Reactor Trip setpoint.

(2) A "POWER LVL HI CH PRE-TRIP" alarm will be received if Reactor Power is allowed to rise an additional 2.6%.

B. (1) Actual Reactor Power is 5.8% away from the Reactor Trip setpoint.

(2) A "POWER LVL HI CH PRE-TRIP" alarm will be received if Reactor Power is allowed to rise an additional 1.5%.

C. (1) Actual Reactor Power is 2.6% away from the Reactor Trip setpoint.

(2) A Reactor Trip will occur if Reactor Power is allowed to rise an additional 2.6%.

D. (1) Actual Reactor Power is 1.5% from the Reactor Trip setpoint.

(2) A Reactor Trip will occur if Reactor Power is allowed to rise an additional 1.5%.

Answer: C Answer Explanation:

A Incorrect - Examinee may not understand the Variable Overpower Trip (VOPT) setpoint and how it is measured with respect to current reactor power. 8.4% is the margin gained between reset and the new trip setpoint. A 2.6% rise from the last reset will not cause an alarm. See explanation for correct answer.

B. Incorrect - Power has to rise approximately 5.8(7'0 from the last VOPT reset for the "HI POWER TRIP RESET DEMAND" alarm to annunciate. A power rise of 2.6% will not give this alarm.

C. Correct - When the VOPT reset pushbutton is depressed the high power trip setpoint is increased to a power level that is approximately 8.4% higher than current power, As power continues to rise the "HI POWER TRIP RESET DEMAND" will annunciate at approximately 2.6% away from the trip setpoint with the pre-trip occurring approximately 1.5% away from the trip setpoint. If the VOPT setpoint is not reset, the reactor will trip.

D. Incorrect - The High power pre-trip alarm annunciates at approximately 1.5%

away from the trip setpoint. 1.5% away from the trip setpoint. If the VOPT setpoint is not reset, the reactor will trip.

Page 90 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Significance of RPS alarm and NO action taken Tier/Group: 2/1 012 - Reactor Protection

• 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.

RO Importance: *4.2 Proposed references to

. None be provided to applicant:

• Identify the source of the VOPT Reset Demand alarm, and Learning Objective:

determine the effect on plant if NO action taken.

10 CFR Part 55 Content:

Question source:

i Cognitive level: o Memory! Fundamental [g] Comprehension/Analysis Last NRC Exam used on: No record of use on any exam LOI-2010 RPS, AOP-7H, Pwr Dist. Tech Specs (05/11)

.Alarm Manual 1COS Alarm window 0-12 Page 91 of 162 Rev. 3

2012 NRC RO EXAM Mt\STER KEY Unit-1 is operating at 100% power when the following sequence of events occurs:

Time 0 11 S/G Pressure is 860 PSIA 12 S/G Pressure is 860 PSIA 11 S/G Level is 0 inches 12 S/G Level is 0 inches Time +1 Min 11 S/G Pressure is 856 PSIA 12 S/G Pressure is 740 PSIA 11 S/G Level is minus (-) 120 inches 12 S/G Level is minus (-) 175 inches Time +2 Min 30 seconds 11 SG Pressure is 800 PSIA 12 SG Pressure is 740 PSIA 11 SG Level is minus (-) 100 inches 12 SG Level is minus (-) 180 inches Assuming NO operator actions, what is the current status of Auxiliary Feed Water?

A. AFW is supplying 11 S/G ONLY B. AFW is supplying 12 S/G ONLY C. AFW is supplying neither S/G D. AFW is supplying 11 & 12 S/G Answer: 0 Answer Explanation:

A. Incorrect - AFAS has initiated based on 12 S/G levels below -170 inches for>

20 seconds and AFAS block did occur to 12 S/G initially but has cleared since block valves remain in AUTO and reopen when condition no longer exists. Thus AFW is supplying BOTH S/Gs.

B. Incorrect - AFAS has initiated and AFW is being supplied to both S/Gs.

C. Incorrect - Conditions for generating an AFAS have lasted for 30 seconds and AFW is being supplied to both S/Gs.

D. Correct - Based on timeline, AFAS has initiated and AFW is supplying both S/Gs.

Page 92 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: AFAS / AFAS Block with NO operator action Tier/Group: 2/1 013 Engineered Safety Features Actuation

• K5 - Knowledge of the operational implications of the KIA Info:

following concepts as they apply to the ESFAS:

• K5.02 - Safety system logic and reliability RO Importance: 2.9

. Proposed references to None be provided to applicant:

Explain the initiating plant conditions and predict the AFAS response actions for the following:

Learning Objective:

• AFAS Start

• AFAS Block 10 CFR Part 55 Content:

Question source:

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: LOI-2006 RO (06/08)

Exam Bank History: LOI-2008 SRO Audit (05/10)

Technical references: 1C03-ALM, Condensate and Feedwater Control Alarm Manual EOP-O, Post Trip Immediate Actions Comments: None Page 93 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY ESFAS channel ZD sensor module for SIAS CP (Containment Pressure) is erratic. The sensor channel has been bypassed for troubleshooting. During troubleshooting I&C technicians remove the SIAS CP module from the sensor cabinet.

Which ONE of the following is the effect on ESFAS when this occurs?

A. The SIAS CP is no longer bypassed and each logic cabinet receives a trip input signal.

B. The SIAS CP is no longer bypassed and ONLY logic cabinet A receives a trip input signal.

C. The SIAS CP trip remains bypassed preventing a trip input signal from being sent to each logic cabinet.

D. The SIAS CP is no longer bypassed and ONLY logic cabinet B receives a trip input signal.

Answer: A Answer Explanation:

A. Correct - The bypass key only works as long as sensor module keeps continuity of circuit. Since module withdrawn, power path is broken and trip input signal is sent to each logic cabinet for SIAS CPo B. Incorrect - First part is true, however, each logic cabinet receives a SIAS CP trip input signal.

C. Incorrect - Even thoUgh bypass key is still installed, the power to circuit was removed thus allowing a SIAS CP trip input signal sent to each logic cabinet.

D. Incorrect - First part is true, however, each logic: cabinet receives a SIAS CP trip input signal.

Page 94 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: ESFAS Sensor Module Maintenance Bypass Circuit Tier/Group: 2/1 013 - Engineered Safety Features Actuation (ESFAS)

• A3 - Ability to monitor automatic operation of the ESFAS KIA Info:

including:

• A3.01 - Input channe'ls and logic RO Importance: 3.7 Proposed references to None be provided to applicant:

Recall the operation of ESFAS that includes:

Learning Objective:

• Sensor module maintenance bypass channel circuit 10 CFR Part 55 Content: 55.41 (b )(7)

Question source:

Cognitive level: D Memory/Fundamental IX] Comprehension/Analysis Last NRC Exam used on: New question Exam Bank History: None I TE~chlnicj31 references: 01-34, ESFAS Fig. 1, Sensor Maintenance Bypass Circuit Comments: None Page 95 of 162 Rev. 3

2012 NRC RO EXAM MA~STER KEY Given the following conditions on Unit-1 at 100% power:

• Containment Cooling System is in a normal lineup with ALL Containment Air Coolers (CACs) available

• 11, 12, and 13 CAC Fans operating in FAST speed

• 11 CAC Emergency SRW Outlet valve is open An event occurs resulting in a Reactor Trip with the following conditions:

• All equipment functions as designed upon the trip

• RCS pressure is 1910 PSIA and lowering

• Containment pressure 0.7 PSIG and rising

• Containment humidity for Dome and Rx Cavity are respectively 38%

and 52% and both rising

• Containment temperature is 110°F and rising Which ONE of the following describes the required operation of the Containment Air Coolers for Containment Environment Safety Function in EOP-O?

A Start 14 CAC in FAST and ensure open ALL CAC Emergency SRW Outlet valves.

B. Start 14 CAC in FAST with ALL CAC Normal SRW Outlet valves open.

C. Open the Emergency SRW Outlet valves on 12 and 13 CACs.

D. No additional manipulation of the CACs is required.

Answer: A Answer Explanation:

A Correct - Since containment pressure is degrading, alternate actions of EOP o require that ALL CACs be started and the Emergency SRW Outlet valves opened.

B. Incorrect - First part is required action, however, the Emergency SRW Outlet valves are opened to assist in lowering pressure and temperature.

C. Incorrect - EOP-O specifically states ensure open ALL CAC Emergency SRW Outlet valves for containment pressure> 0.7 psig or containment temperature

> 120 OF.

D. Incorrect - Taking no actions does not meet expectations of EOP-O based on parameter trends.

2012 NRC RO EXAM MASTER KEY Containment Air Cooler operation in EOP-O Tier/Group: 2/1 022 - Containment Cooling

• A 1 - Ability to predict and/or monitor changes in KIA Info: parameters (to prevent exceeding design limits) associated with operating the CCS controls including:

• A 1.01 - Containment temperature RO Importance: 3.6 Proposed references to None be provided to applicant:

Recall the purpose of each of the safety function boxed Learning Objective:

steps of EOP-O.

10 CFR Part 55 Content: 55.41 (b)(5)

Cognitive level: cg] Comprehension/Analysis Last NRC Exam used on:

Technical references: EOP-O Containment Environment Safety Function Comments:

Page 97 of 162 Rev. 3

2012 NRC RO EXAM MP\STER KEY Following a Unit-2 plant trip from 100% power, a LOCA has occurred. CIS and SIAS have actuated.

Which condition represents the response of the Component Cooling (CC) system valves and equipment to the current ESFAS signals? Consider ONLY the Component Cooling side of the system.

A. Each CC HX outlet valve opens, the CC containment isolation valves shut, and all CCW pumps start.

B. Each CC HX outlet valve opens, each SOC HX inlet valve opens, and 21 and 22 CC pumps start.

C. Each SOC HX outlet valve opens, the CC containment isolation valves shut, and all CC pumps start.

O. Each SOC HX outlet valve opens, the CC containment isolation valves shut, and 21 and 22 CC pumps start.

Answer: 0 Answer Explanation:

A. Incorrect - ONLY the CC containment isolation valves response is correct. ALL CC pumps receive a SIAS start signal but 13 (23) pump only starts if 120r 22 CC Pp does not start within one second after receiving a SIAS.

B. Incorrect - The CC HX outlet valves do NOT receive any ESFAS signal but remaining response is correct.

C. Incorrect - First two responses are correct, and third response as stated in A above does not occur.

O. Correct - All 3 responses are expected actions of CC upon a SIAS and CIS.

Page 98 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY CC system response to SIAS and CIS

• Tier/Group: 2/1 026 - Containment Spray

• A3 - Ability to monitor automatic operation of the CSS, KIA Info: including:

• A3.02 - Verification that cooling water is supplied to the containment spray heat exchanger

• RO Importance: 3.6 Proposed references to None be provided to applicant:

Determine the response on CCW system when a SIAS, CIS, Learning Objective:

and CSAS occur 10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: D Memory/Fundamental rg] Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI-2008 RO Audit (11/08)

Technical references: EOP Attachment 2 page 4 of 5 EOP Attachment 4 page 1 of 2 EOP Attachment 6 Page 99 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given a total loss of Component Cooling, which of the following actions ensures core cooling is maintained during a LOCA with Recirculation Actuation Signal (RAS) in progress?

A. Secure one HPSI pump, align one Containment Spray pump for injection, and throttle HPSI flow to minimum allowed per EOP attachment.

B. Align one Containment Spray Pump for injection and stop ALL running HPSI pumps.

C. Stop ALL running HPSI pumps, start a LPSI pump using RAS override for injection and THEN throttle LPSI flow to minimum allowed per EOP attachment.

D. Align BOTH Containment Spray Pumps for injection and stop ALL running HPSI pumps.

Answer: B Answer Explanation:

A. Incorrect - Securing one HPSI pump with NO CC does not protect remaining HPSI pump from overheating. The Containment Spray pump can operate without CCW flow as ECCS pump room air coolers provide SW cooling into room. Flow may be throttled through LPSI header valves.

B. Correct - These are required actions when CCW flow cannot be restored during RAS per EOP-5 Block Step S.1.f.2 C. Incorrect - Use of a LPSI pump during RAS is not allowed as the large flow initially may lead to clogging the sump screens with debris resulting in loss of NPSH for other pumps taking suction from the sump.

D. Incorrect - Aligning both spray pumps for safety injection would lower flow to cool the containment possibly preventing a lowering of containment temperature and pressure.

Page 100 of 162 Rev. 3

2012 NRC RO EXAM MP~STER KEY Topic: Containment Spray Pp purpose during LOCA Tier/Group: 2/1 026 Containment Spray KIA Info:

• 2.1 - Conduct of Operations

• 2.1.28 - Knowledge of the purpose and function of major system components and controls.

RO Importance: 4.1 Proposed references to None be provided to applicant:

Learning Objective: Given EOP-5 implemented, verify RAS actions.

i 10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: D Memory/Fundamental [XI Comprehension/Analysis Last NRC Exam used New None Technical references: EOP-5 Step S.1.f.2 and Technical Bases document Comments: None Page 101 of 162 Rev. 3

2012 NRC RO EXAM MA~STER KEY Unit-1 was operating at 100% power when the following occurs:

• A loss of 1Y10 occurs, and

• A total loss of Condenser Vacuum How should the ADVsffBVs respond immediately upon the reactor trip?

A. ADVs ramp open, TBVs ramp open B. ADVs quick open, TBVs ramp open C. ADVs ramp open, TBVs remain shut D. ADVs quick open, TBVs remain shut Answer: 0 Answer Explanation:

A. Incorrect - A loss of vacuum tripping the main turbine also makes TBVs inoperable. ADVs initially quick open upon trip from 100% power.

B. Incorrect - ADVs quick open, however, as stated in A TBVs are inoperable due to loss of vacuum.

C. Incorrect - First part is wrong and TBVs remain shut upon the reactor trip.

D. Correct - This is response to trip at 100% power with a loss of vacuum that trips the main turbine.

Page 102 of 162

2012 NRC RO EXAM MASTER KEY AOVITBV response on loss of vacuum and 1Y10 Tier/Group: 2/1 039 - Main and Reheat Steam (MRSS)

• K3 - Knowledge of the effect that a loss or malfunction of KIA Info:

the MRSS will have on the following:

• K3.06 - SOS RO Importance: 2.8 Proposed references to None be provided to applicant:

Evaluate AOVrrBV operation upon Loss of 1Y10 and loss of Learning Objective:

vacuum causing a reactor trip.

10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: ~ Comprehension/Analysis Last NRC Exam used on:

Exam Bank History: 1-2006 RO/SRO Audit Remediation (05/08)

Technical references: AOP-7G-1, Loss of Vacuum Comments: None Page 103 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following:

• A loss of Service Water resulted in a Unit -1 trip and loss of the Instrument Air compressors 30 minutes ago.

• 13 AFW Pump is unavailable (1) How is the AFW system affected and; (2) What operator actions are required to maintain Steam Generator levels?

A. (1) The operating AFW pump trips on overspeed; (2) Adjust the local speed adjust knob to minimum, reset the overspeed trip device, raise AFW Pump discharge pressure to 100 PSI above SfG pressure.

B. (1) The operating AFW pump speed will rise to the maximum governor setting; (2) Adjust the local speed adjust knob to maintain AFW Pump discharge pressure 100 PSI greater than SfG pressure.

C. (1) The operating AFW pump speed will lower to the minimum governor setting; (2) Adjust the AFW Pump Speed Controller, at 1C04, to obtain the desired AFW flow rate.

D. (1) SfG levels rise due to the flow control valves failing open; (2) Align the Liquid N2 System to supply SfG FLOW CONTR valves via the AFW System Air Accumulators.

Answer: B Answer Explanation:

A. Incorrect - The AFW Pump(s) run up to max speed, they do not trip. Actions taken would be correct if AFW Pump(s) did trip.

B. Correct - Effect of loss of IfA is as noted and AOP-7D provides direction to perform actions to locally control AFW Pump speed C. Incorrect - AFW pump speed goes to maximum due to the loss of IfA. The AFW Pump Speed Controller at 1C04 has no effect on AFW Pp speed due to the loss of IfA. Examinee may think AFW Pp speed control (governor) is supplied by the AFW air accumulators that provide a source of air to other AFW components in an extended loss of Instrument Air situation.

D. Incorrect - The AFW Flow Control CVs will not fail open due to being supplied air via the AFW air accumulators (good for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). SfG level would be controlled by maintaining AFW Pp speed 100 PSI above SfG pressure.

Controlling FCVs thru use of liquid N2 is directed by EOP-7, Station Blackout which assumes the AFW air accumulators have been depleted.

Page 104 of 162 Rev. 3

12 NRC RO EXAM Mt\STER KEY Effects to Unit-2 AFW valves on loss of Instrument Air 2/1 061 - Auxiliary Feedwater

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on KIA Info: those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

• A2.07 - Air or MOV failure RO Importance: 3.4 Proposed references to be provided to None applicant:

Given a loss of any 125 VDC Vital Bus, evaluate the effect on Learning Objective:

each unit and required actions.

10 CFR Part 55 55.41 (b)(5)

Content:

I ul.lestion source:

Cognitive level: o Memory/Fundamental [2:J Comprehension/Analysis Last NRC Exam used New question on:

Exam Bank History: None Technical references: AOP-7D-2, Loss of Instrument Air and bases pages 7 and 12 Comments: None Page 105 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Following a reactor trip, which ONE of the following bus losses would require operator actions to maintain the Core and RCS heat removal safety function per EOP-O?

A. 2Y09 B. MCC-107 C. 13B 480V Bus D. 12 4KV Bus Answer: B Answer Explanation:

A. Incorrect - Loss of 2Y09 major effect would be ALL Low Pressure Feedwater Heater High Level Dumps fail open and challenge MFW operation when operating at power. Since the reactor has tripped these high level dumps receive a signal to open on the trip and loss of 2Y09 during EOP-O has little affect on MFW thus will not challenge the Core and RCS heat removal safety function.

S. Correct - MCC-107 lost results in tripping off ALL Unit-1 Circ Water Pumps. This leads to a loss of vacuum and a trip of the SGFPs and loss of Turbine Bypass Valves. Initiation of AFW will be the alternate action necessary in EOP-O for Core and RCS Heat Removal and ADVs will be used to control RCS temperature.

C. Incorrect - 13B 480V bus loss will result in the loss of MCC-116. This will result in potential loss of 2 Condensate pumps. A singlE~ Condensate pump will be able to support MFW requirements and AFW will not be needed in EOP-O.

D. Incorrect - The Main Feed system continues to operate, just at a reduced capacity as only 1 Condensate pump and 2 Condensate Booster pumps have been lost.

Page 106 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY AC Electrical Distribution - Bus loss causing LONHR Tier/Group: 2/1

• 062 - AC Electrical Distribut~on

• K3 - Knowledge of the effect that a loss or malfunction of KIA Info:

the ac distribution system will have on the following:

• K3.01 - Major system loads RO Importance: 3.5 Proposed references to None be provided to applicant:

• Given an electrical bus malfunction, diagnose the event and Learning Objective:

take appropriate actions per AOP-71.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: LOR 11-6B Biennial Exam (11/11)

Technical references: AOP-7I, Loss of 4KV, 480 Volt or 208/120 Volt Instrument

• Bus Power EOP-O, Post Trip Immediate Actions Comments: None Page 107 of 162 Rev. 3

12 NRC RO EXAM MP\STER KEY Given the following conditions on Unit-1:

• A Station Blackout is in progress.

• EOP-7, Station Blackout, has been implemented.

Which ONE of the following describes why the Plant Computer Inverter, 1Y05A, is deenergized?

A. Removes a large DC load from 11 DC Bus allowing the bus to meet the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> discharge rate.

B. Removes a large DC load from 12 DC Bus allowing the bus to meet the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> discharge rate.

C. Removes a large DC load from 12 DC Bus allowing the bus to meet the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> discharge rate.

D. Removes a large DC load from 11 DC Bus allowing the bus to meet the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> discharge rate.

Answer: B Answer Explanation:

A. Incorrect - 12 DC Bus has minimal load on it during normal operation. With SBO occurring, the load does not change. 1Y05A is not powered from 11 125V DC Bus.

B. Correct - Per EOP-7 Step J basis. Removing this load was identified during PRA that would allow the bus to be maintained for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> just on battery.

C. Incorrect - Once again 12 DC Bus has minimal load on it during normal operation. Calculations performed verify that during a SBO each battery can carry required loads for at least one hour and most likely 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. Incorrect - Per UFSAR each station battery is designed to last at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, however, EOP-7 states that removing this load will allow 12 DC Bus to meet a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> discharge rate. 1Y05A is powered from 12 125V DC Bus.

Page 108 of 162 Rev. 3

2012 NRC RO EXAM Mi\STER KEY Shedding Computer Inverter load during SBO

, Tier/Group: 2/1 063 - DC Electrical Distribution

• A 1 - Ability to predict and/or monitor changes in parameters associated with operating the DC electrical KIA Info:

system controls including:

• A 1.01 - Battery capacity as it is affected by discharge rate RO Importance: 3.6 Proposed references to None i be provided to applicant:

STATE the electrical performance and design attributes of Learning Objective:

the 125 VDC, and 120 VAC Vital Busses.

10 CFR Part 55 Content: 55.41 (b)(5)

, Cognitive level: o Comprehension/Analysis i Last NRC Exam used on:

,Technical references: EOP-7, Station Blackout ancl Technical Bases Comments: None Page 109 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given a Loss of Offsite Power to both units, the following conditions exist:

• 1A Diesel Generator is out of service for maintenance

• 2B Diesel Generator did not load due to a faulted 4KV bus Which ONE of the following statements is correct?

A. 11 DC bus is being supplied ONLY by 11 battery charger.

B. 21 DC bus is being supplied ONLY by 21 battery charger.

C. 12 DC bus is being supplied by 24 battery chargHr.

D. 22 DC bus is being supplied by 22 battery charger.

Answer: D Answer Explanation:

A. Incorrect - 11 Bus will receive power from 23 battery charger. 11 Battery Charger is not available due to the unavailability of the 1A DG.

B. Incorrect - 21 battery charger is powered from 24A 480V Bus, which remains deenergized as the 2B DG did not load.

C. Incorrect - 24 battery charger is powered from 24B 480V Bus which remains deenergized as the 2B DG did not load.

D. Correct - 22 battery charger is powered from 21 B 480V Bus which is reenergized from 2A Diesel Generator.

Page 110 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Emergency DG and DC busses Tier/Group:

064 - Emergency Diesel Gemerator

• K 1 - Knowledge of the physical connections and/or KIA Info: cause/effect relationships between the ED/G system and the following systems:

• K1.04 - DC distribution system RO Importance: 3.6 Proposed references to None i be provided to applicant:

Recall the purpose of each of the safety function boxed ILearning Objective:

steps of EOP-O.

I 10 CFR Part 55 Content:

! Cognitive level: ~ Memory/Fundamental o Comprehension/Analysis i Last NRC Exam used on: I LOI-2004 RO None AOP-71-1 & 2, Loss of 4KV, 480 Volt or 208/120 Volt

. Instrument Bus Power Never put into bank Page 111 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions:

• 1-RIC-4095 operating as a substitute for 1-RIC-4014 per Ol-BA

• S/G Blowdown is discharging to Unit-1 Circ Water

• The Blowdown Recovery HI-TEMP DUMP, 1-BD-40BB-CV, is shut

• Annunciator window "UNIT 1 S/G BID RECOVERY" has just alarmed at 1C22H due to HIGH alarm setpoint exceeded Which ONE of the following reflects the response of the SG Blowdown system?

(Assume NO operator action) i  !

I BID REC I! BID REC BID REC 11(12)SG DISCH TO DISCH TO DISCH TO BOT BID COND, CW, MWS, CNTMT ISOLs,

• 1-BD-4096- 1-BD-4015- 1-BD-4097- 1-BD-4011-CV CV CV CV 1-BD-4013-CV i I

A. Shut Shut Open Shut B. Open Open Shut Open C. Shut Shut Open Open D. Open Shut Shut Shut Answer: C Answer Explanation:

A. Incorrect - First 3 responses are correct based on alarm actions and system lineup. SG Bottom BD valves must be manually closed when 1-RIC-4095 is substituting for 1-RIC-4014 per Ol-BA Note for Precaution 5.0E.

B. Incorrect - The RMS still provides a close signal to control circuit for each valve preventing operator from opening valves unless placed in RAD TRIP Override.

C. Correct - This is the correct response of system valves with the given system lineup. The operator must manually shut the SG Bottom BD valves since automatic actions to close occur only from RIC-4014 which is OOS.

D. Incorrect - Only BD recovery discharge CW response is correct. All others are wrong. BD does not transfer to Condenser from Circ Water upon a high RMS condition.

Page 112 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY

• Topic: S/G Slowdown response upon RMS alarm Tier/Group: 2/1 073 Process Radiation Monitoring

• K4 - Knowledge of PRM system design feature(s) and/or KIA Info: interlock(s) which provide for the following:

• K4.01 - RE~lease termination when radiation exceeds setpoint RO Importance: 4.0 Proposed references to None be provided to applicant:

Determine the response to S/G Slowdown system valves

'Learning Objective:

upon 1-RIC-4095 high alarm 10 CFR Part 55 Content: 55.41 (b )(7)

• Cognitive level:

D Memory/ Fundamental [g] Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Sank History: LOI-20 10 1C22/1 C34 exam (09/11)

Technical references: OI-8A, S/G Slowdown System Comments: Modified from Q24653 by adding response of S/G Slowdown valves on RIC-4095 high alarm Page 113 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following is the normal bus power alignment for 13 (23)

SRWpumps?

A. 13 Pump - 14 Bus; 23 Pump - 24 Bus.

B. 13 Pump - 14 Bus; 23 Pump - 21 Bus C. 13 Pump - 11 Bus; 23 Pump - 21 Bus D. 13 Pump - 11 Bus; 23 Pump - 24 Bus Answer: A Answer Explanation:

A. Correct - These are the normal power alignments of the SRW pumps per OI-27C. Both of these pumps are capable of being aligned to either 4KV safety related bus when required by using disconnects.

B. Incorrect - Pump breaker power alignment is wrong for 23 SRW pump per OI-27C. Both of these pumps are capable of being aligned to either 4KV safety related bus when required by using disconnects.

C. Incorrect- Bus alignments for 13 and 23 are to 14 and 24 4KV busses respectively. Both of these pumps are capable of being aligned to either 4KV safety related bus when required by using disconnects.

D. Incorrect - 13 SRW Pump breaker normal power alignment is wrong per OI-27C. Both of these pumps are capable of being aligned to either 4KV safety related bus when required by using disconnects.

Page 114 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Service Water Pump power supplies Tier/Group: 2/1 076 - Service Water System (SWS)

KIA Info:

• K2 - Knowledge of bus power supplies to the following:

• K2.01 - Service water RO Importance: 2.7 Proposed references to None be provided to applicant:

Recall the power supply alignment of SRW pumps for each Learning Objective:

unit.

10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: [2J Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOl2006 1C02 Exam (10/07)

Technical references: OI-27C, 4.16 KV SYSTEM Comments: Modified from Q20452; removed reference to headers Page 115 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following conditions:

• Unit-2 is in Mode 1 at 100% Reactor Power

• An electrical perturbation occurs

• The CEAPDS monitor has deenergized as a result of the electrical perturbation What is (1) the minimum bus lost and (2) the immediate stabilizing actions expected to be performed?

A. (1) 2Y09, (2) Insert CEAs as needed to restore TCOlD to below 548 0 F and maintain on program.

S. (1) 2Y10, (2) Shift PZR HTR LO LVL CUTOFF SEL switch to Channel X and manually reenergize ALL Pressurizer Heaters as necessary.

C. (1) 2Y09, (2) Fast borate to reduce rE~actor power and promptly reduce Turbine load to restore T COLD to program.

D. (1) 2Y10, (2) Align Chg Pp suction to the VCT, Reduce Turbine load to restore TCOLD to program and place two Chg Pps in Pull To Lock.

Answer: D Answer Explanation:

A. Incorrect - Conditions given in the stem indicate a loss of 2Y1 0 as a minimum.

These actions apply to the effects of a loss of 2Y09 as 2 nd stage MSR reheat steam inlet valves fail shut on Unit-2 but the Feedwater Heater high level dump control valves also fail open causing a reactor power excursion to lower T COLD requiring operator to adjust turbine load and borate to stop power rise.

B. Incorrect - Conditions given in the stem indicate a loss of 2Y10 as a minimum.

These actions apply to a loss of 2Y02. they are not appropriate for a loss of 2Y10.

C. Incorrect - A loss of 2Y09 does require the immediate actions stated in this distracter because the Feedwater Heater HLDCVs fails open causing a reactor power excursion. However, the CEAPDS monitor is NOT deenergized on a loss of 2Y09. The monitor display indicates ALL CEAs are inserted on a loss of 2Y09.

D. Correct - Loss of CEAPDS indicates 2Y10, as a minimum, is deenergized. The "Immediate Actions" plaque states the Charging Pump suction shifts to the RWT with all Charging Pumps running and directs opening the VCT outlet MOV, shutting the RWT outlet to the Charging Pump suction, adjusting turbine load to maintain T COLD on program and placing two Charging Pumps in PTL.

Page 116 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Loss of 2Y1 0 immediate actions Tier/Group: 2/1 004 - CVCS 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

RO Importance: 4.6 Proposed references to None be provided to applicant:

Mentally develop a methodology for diagnosing electrical Learning Objective: malfunctions in the Control Room by using key control board indications 10 CFR Part 55 Content: 55.41 (b)(5)

Question source: D Modified Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any NRC exam Exam Bank History: LOI-2006 Panel Comp Technical references:

Comments: None Page 117 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Unit-2 is in Mode 3 when an RCS depressurization event occurs causing Pressurizer pressure to lower to 1700 PSIA.

Which ONE of the following occurs based on this event?

A. The Saltwater Air Compressors (SWACs) start and will continue providing air to operate the TBVs.

B. IA Containment isolation, 2-IA-2080-MOV, shuts isolating control system air to containment components.

C. Instrument Air compressors trip on high Aftercooler or Intercooler temperature; Plant Air compressor trips on high discharge or first stage temperature.

D. The B/U IA HDR PCV TO U-2, 2-IA-6301-PCV, will open to supply the IA header from the IA Storage Tanks.

Answer: C Answer Explanation:

A. Incorrect - SWACs do start on SIAS but do NOT provide air to TBVs.

B. Incorrect - Stated conditions do not support actuation of CIS which closes this valve. Instrument Air to containment will be supplied by the Unit -1 Plant Air Compressor once the Unit-2 Plant Air Compressor trips.

C. Correct - Stated conditions support actuation of SIAS which isolates SRW to turbine building and eventually these compressors trip on high temperature conditions listed.

D. Incorrect - 11 Plant Air Compressor will be supplying the U-2 Instrument Air header via the cross-connect valves. Instrument Air header pressure would not lower to the setpoint for opening 2-IA-6301-PCV (85 PSIG).

Page 118 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Loss of SRW to Compressed Air system due to SIAS Tier/Group: 2/1 078 Instrument Air K4 - Knowledge of lAS design feature(s) and/or interlock(s)

KIA Info:

which provide for the following:

• K4.03 - Securing of SAS upon loss of cooling water RO Importance:

Proposed references to None be provided to applicant:

Evaluate the long-term effect of a SIAS on the compressed Learning Objective:

air system.

10 CFR Part 55 Content: 55.41 (b}(7)

Question source:

Cognitive level: D Memory/Fundamental Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI-2008 RO Audit (11/08)

Technical references: Alarm Response Manual 2C 13 Comments: Modified from Q20286 Page 119 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Under which ONE of the following conditions will a stop motion signal be supplied to the group programmer modules? (UCS/LCS = Upper/Lower Computer Stop)

A. ONLY during Manual Group mode withdrawal when highest CEA in group reaches UCS at 130.5 inches.

B. ONLY during Manual Sequential mode insertion when lowest CEA in group reaches LCS at 10 inches.

C. During Manual Sequential or Manual Group mode withdrawal when lowest CEA in group reaches UCS at 135.0 inches.

D. During Manual Sequential or Manual Group mode insertion when highest CEA in group reaches LCS at 6 inches.

Answer: D Answer Explanation:

A. Incorrect - Outward motion is terminated when the lowest (not highest) CEA, in the group, reaches 130.5 inches if selected to manual sequential or manual group mode.

B. Incorrect - Inward motion is terminated when the highest (not lowest) CEA, in the group, reaches 6 inches (vice 10 inches) if selected to manual sequential or manual group mode.

C. Incorrect - Outward motion is terminated when the lowest CEA, in the group, reaches 130.5 inches (not 135.0 inches) if either mode is selected. This is Upper Electrical Limit for reed switch indication.

D. Correct - Inward motion is terminated when the highest CEA, in the group, reaches 6 inches if either mode is selected.

Page 120 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Tier/Group: 2/2 001 - Control Rod Drive

• K4 - Knowledge of CRDS design feature(s) and/or KIA Info:

interlock(s) which provide for the following:

• K4.23 - Rod motion inhibit RO Importance: 3.4 Proposed references to be provided to None applicant:

During withdrawal or insertion, determine condition to stop Learning Objective: CEA group motion when in manual sequential or manual group mode.

10 CFR Part 55 55.41 (b )(7)

Question source:

Cognitive level: ~ Memory/Fundamental o Comprehension/Analysis

• Last NRC Exam used

• No record of use on any exam on:

Exam Bank History: LOI-2010 1C07, AFWand AFAS exam (04/11)

Technical references: 01-42, CEDM System Operation OP-2, Plant Startup From Hot Standby To Minimum Load Comments: *Modified Q25785 to add variation of Manual Sequential and/or Manual Group to each distractor.

Page 121 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY The following conditions exist on Unit-1:

• 100% power with core burnup of 11,000 MWD/MTU

• 1-HIC-5206, 11 CC Hx Saltwater Flow Controller, output signal drifts from 8% to 12%.

• Attempts to return 1-HIC-5206 controller output signal to 8% are unsuccessful (1) Which ONE of the following is the plant response and (2) What action is required per plant procedure?

A. (1) Component Cooling HX outlettemperature rises causing RCS boron to lower and reactor power to rise; (2) Place Letdown Hx Temp. ControIlE~r, 1-TIC-223, in MANUAL to maintain letdown temperature constant.

B. (1) Letdown HX outlet temperature rises causing RCS boron to rise and reactor power to lower; (2) PLACE IX BYPASS, 1-CVC-520-CV, to BYPASS to stop the power reduction.

C. (1) Letdown HX outlet temperature lowers causing RCS boron to lower and reactor power to rise; (2) PLACE IX BYPASS, 1-CVC-520-CV, to BYPASS to stop the power rise.

D. (1) Component Cooling HX outlet temperature lowers causing RCP seal pressure perturbations.

(2) Place 1-TIC-3823, 11 CC HX TEMP CONT BYP, to AUTO to return CC HX outlet temperature to maintain normal operating temperature.

Answer: C Page 122 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Answer Explanation:

A. Incorrect - CC Hx outlet temperature lowers which causes RCS boron to be lowered and raise reactor power. Appropriate action is to bypass the IXs to stabilize reactor power.

B. Incorrect - LID outlet temperature lowers not raises and reactor power would rise not lower; Bypassing IXs will immediately terminate the reactivity addition.

C. Correct - This is the expected response to RCS Boron and power; bypassing IXs will immediately terminate the positive reactivity addition per AOP-1A, Inadvertent Dilution.

D. Incorrect - First part is correct. RCP seals will be affected due to increased flow thru CC Hx, however, the boron effect to the RCS is the immediate concern.

Placing 1-TIC-3823 in AUTO would be a deviation from plant procedures (01-16 or AOP-7C). 01-16 only allows placing this controller in MANUAL and fully opening this valve. AOP-7A or 7C does not provide any action to operate this valve.

Page 123 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Temperature affects on CVCS IX resin Tier/Group: 2/1 008 - Component Cooling Water System

• A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS and (b) based on KIA Info: those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations

• A2.03 - High/low CCW temperature RO Importance: 3.5 Proposed references to be provided to None applicant:

Learning Objective:

10 CFR Part 55 55.41 (b)(5)

Content:

Question source:

Cognitive level: [;g] Comprehension/Analysis Last NRC Exam used New question on:

Exam Bank History: None

! 01-29, Saltwater System AOP-1A, Inadvertent Dilution None Page 124 of 162 Rev. 3

2012 NRC RO EX.AM MASTER KEY PZR level is 10 inches below setpoint. If all systems are in AUTO, what should letdown flow be?

A. 0 GPM B. 24 GPM C. 30 GPM D. 36 GPM.

Answer: C Answer Explanation:

A. Incorrect - The Letdown Stop Valves would have to be shut for this value.

Information in the stem does not support this conclusion.

B. Incorrect - The HIC has a flow limiter which prevents the letdown valves from closing below 30 gpm. Examinee may subtract RCP Bleedoff flow from minimum UD flow to reach a total of 24 GPM.

C. Correct** The HIC has a flow limiter which prevents the letdown valves from closing below 30 gpm.

D. Incorrect - The HIC has a flow limiter which prevents the letdown valves from closing below 30 GPM. Examinee may add RCP Bleedoff flow to minimum UD flow to reach a total of 36 GPM.

Page 125 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Interrelationship between RCS and CVCS Tier/Group: 2/2 002 - Reactor Coolant

• K1 - Knowledge of the physical connections and/or KIA Info: cause-effect relationships between the RCS and the following systems:

• K1.06 - CVCS RO Importance: 3.7 Proposed references to None be provided to applicant:

Determine the minimum letdown flow during CVCS Learning Objective:

operation.

10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

Cognitive level: [gJ Memory/Fundamental D Comprehension/Analysis I Lalst INRC Exam used on: No record of USE:! on any exam I Ex:am Bank History: None Technical references: SD-41 CVCS Comments: None Page 126 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Given the following:

• Reactor power is being raised from 50 to 100%

• TCOLD is on program

• The "Nuclear L\T Power Ch Deviation" alarm is received.

Which ONE of the following actions is required to be performed by Operations to clear this alarm for the current power level?

A. Balance turbine load with reactor power.

B. Calibrate the Ex-core NI Channels.

C. Null the NI Pots to the Delta-T Pots.

D. Adjust the T COLD Calibrate Pot.

Answer: B Answer Explanation:

A. Incorrect - The stem statement identifies that T COLD is on program meaning reactor power and turbine load are balanced for the current power.

B. Correct - The conditions specified in the stem of the question indicate the need, per the Alarm Manual, for calibration of the Excore NI Channels in accordance with 01-30, Nuclear Instrumentation.

C. Incorrect - Nulling NI Pots to L\T Pots can only be performed when reactor power is < 30% per 01-30, Nuclear Instrumentation.

D. Incorrect - The TCOLD Calibrate Pot is not adjusted by Operations personnel.

Page 127 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: NI alarm response Tier/Group: 2/2 015 Nuclear Instrumentation

• A3 - Ability to monitor automatic operation of the NIS, KIA Info:

including:

• A3.02 - Annunciator and alarm signals RO Importance: 3.7 Proposed references to None be provided to applicant:

Learning Objective:

• 10 CFR Part 55 Content:

Question source:

Cognitive level: D Memory/Fundamental

. Last NRC Exam used on: No NRC Exam use Exam Bank History: LOI-2010 RPS (05/11)

Technical references: 01-30, Nuclear Instrumentation 1C05-ALM, Reactivity Control Alarm Manual Comments: None Page 128 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following indicates a Core Exit Thermocouple (CET) input, to the Post Accident Monitoring System, has been bypassed?

A. A blue backlight and a "B" adjacent to the parameter.

B. Parameter will indicate with a magenta backlight.

C. A green "s" adjacent to the parameter.

D. Parameter will indicate with a"??".

Answer: A Answer Explanation:

A. Correct - Per 01-11, Post Accident Monitoring System, a blue backlight and a "B" adjacent to the parameter indicate a bypassed parameter.

B. Incorrect - Per 01-11, Post Accident Monitoring System, failed parameters will indicate with a magenta backlight.

C. Incorrect - Per 01-11, Post Accident Monitoring System, a green "s" indicates a substituted RVLMS Probe.

D. Incorrect - Per 01-11, Post Accident Monitoring System, "??" indicates a parameter that is not valid due to insufficient data to support the parameter.

Page 129 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: PAMS operation with CETs bypassed Tier/Group: 2/2 017 In-Core Temperature Monitor

• K6 - Knowledge of the effect of a loss or malfunction of KIA Info:

the following ITM system components:

• K6.01 - Sensors and detectors RO Importance: 3.6 Proposed references to None be provided to applicant:

Learning Objective: Determine how a bypassed CET is indicated by PAMS.

10 CFR Part 55 Content: 55.41 (b)(5)

Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: None 01-11, Post Accident Monitoring System Technical references:

LOI-114-1-2, Post Accident Monitoring System (slide 35)

Comments: None Page 130 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY Given the following:

• Unit-1 reactor tripped due to a LOCA

• Containment pressure has reached 3.0 PSIG Which ONE of the following describes Containment Iodine Removal Unit operation for existing plant conditions?

A. CIS starts ONLY 11 and 12 Iodine Removal Units.

B. CSAS starts ALL Iodine Removal Un~ts.

C. SIAS starts ALL Iodine Removal Units.

D. CRS starts ONLY 11 and 12 Iodine Removal Units.

Answer: C Answer Explanation:

A. Incorrect - ALL IRUs start on SIAS, not CIS. Both SIAS and CIS actuate at a Containment pressure of 2.8 PSIG.

B. Incorrect - All IRUs start on SIAS not CSAS. SIAS actuates at a Containment pressure of 2.8 PSIG. CSAS actuates at a Containment pressure of 4.25 PSIG.

Stated conditions indicate a CSAS would not be actuated.

C. Correct - AlllRUs start on SIAS. SIAS actuates at a Containment pressure of 2.8 PSIG.

D. Incorrect - AlIlRUs start on SIAS. CRS actuates based on high radiation as indicated on Containment Area Radiation Monitors. These monitors are only for refueling purposes and are disabled during normal power operation. CRS and starting Iodine Removal Units seem a logical fit if the examinee is unsure of the correct answer.

Page 131 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Containment IRU controls Tier/Group: 2/2 027 Containment Iodine Removal i KIA Info:

• A4 - Ability to manually operate and/or monitor in the control room:

RO Importance:

Proposed references to None be provided to applicant:

Recall the purpose of each of the safety function boxed Learning Objective:

steps of EOP-O.

10 CFR Part 55 Content:

Question source:

Cognitive level: r;gJ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI-2002 1coa, 09, and 10 (05/03)

Technical references:

Comments:

Page 132 of 162 Rev. 3

NRC RO EXAM MASTER KEY Unit-2 has tripped from 100% due to a LOCA and loss of offsite power. The following conditions exist:

• The OC DG was out of service prior to the trip

• The 2B DG had a start failure upon the loss of offsite power

• The Crew has implemented EOP-5 The CRS has directed the RO to perform the following action per EOP-5:

"IF hydrogen concentration can NOT be determined, THEN start the Hydrogen Recombiners per OI-41A, HYDROGEN RECOMBINERS."

Which Hydrogen Recombiner(s) have power available?

A. 21 and 22 Hydrogen Recombiners by tying MCCs 204 and 214 B. 21 Hydrogen Recombiner from 480V Bus 21 B C. 21 and 22 Hydrogen Recombiners from 480V Bus 21A and 24B D. 22 Hydrogen Recombiner from 480V Bus 24A Answer: B Answer Explanation:

A. Incorrect - Examinee may believe these loads receive power from MCCs rather than 480V load centers. Tying MCCs together would be an action directed per EOP-5 if a single 4KV bus is lost. Hydrogen recombiners are NOT powered from MCC-204 or 214.

B. Correct - Hydrogen Recombiner 21 is only one available and powered from 480V bus 21 B.

C. Incorrect - Examinee may recognize power supplies are correct, however, 22 is unavailable as 2B DG failed to start and reenergize 4KV Bus 24. Hydrogen Recombiner 21 is powered from Bus 21 B not 21A and Hydrogen Recombiner 22 is powered from 480V Bus 24B which is deenergized due to 2B DG tripped and OC DG being unavailable.

D. Incorrect - Hydrogen Recombiner 22 is unavailable as 2B DG tripped and has NOT reenergized 4KV bus 24, the OC DG is unavailable, and the power supply is 480V Bus 24B not 24A.

Page 133 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Hydrogen Recombiner Power Supplies Tier/Group:

028 Hydrogen Recombiner and Purge Control

• K2 - Knowledge of bus power supplies to the following:

• K2.01 - Hydrogen recombiners RO Importance: 2.5 Proposed references to None

• be provided to applicant:

Learning Objective: *Recall the power supplies to the hydrogen recombiners.

I

• 10 CFR Part 55 Content: 55.41 (b)(7)

Question source:

• Cognitive level: D Memory/Fundamental [gI Comprehension/Analysis i

• Last NRC Exam used on: No record of use on any exam Exam Bank History: LOI-2002 1C08, 09,10 Misc Remediation (06/03)

AOP-71,Section VIII, page 42 and Section XXVII, page 164 .

Modified from Q20688 Page 134 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Unit-1 is recovering from a plant trip after extended full power operation (400 days).

• Reactor power is 30% and holding for NI Calibration

• No CEA motion or boration/dilution operations are in progress

• TBV Controller, 1-PIC-4056, is in auto and the setpoint is set at 900 PSIA

• Turbine Bypass Valve, 1-MS-3944-CV, has failed open Which ONE of the following sets of actions is taken to stabilize the plant?

A. Insert CEAs, as necessary, to return Reactor power to the required value; Maintain turbine load constant and isolate the TBV to restore TCOLD to program.

B. Withdraw CEAs, as necessary, to maintain Reactor power; Maintain turbine load constant and isolate the TBV to restore T COLD to program.

C. Insert CEAs, as necessary, to return Reactor power to the required value; Lower turbine load to restore TCOLD to program.

D. Withdraw CEAs, as necessary, to maintain Reactor power; Lower turbine load to restore T COLD to program.

Answer: C Answer Explanation:

A. Incorrect - Per AOP-7K, Overcooling Event in Mode One or Two, CEAs should be inserted, as necessary, to maintain reactor power constant (later in the core cycle) and the overcooling event is compensated for by adjusting turbine load.

B. Incorrect - Per AOP-7K, Overcooling Event in Mode One or Two, CEAs should be inserted, as necessary, to maintain reactor power constant (later in the core cycle) and the overcooling event is compensated for by adjusting turbine load.

C. Correct - Per AOP-7K, Overcooling Event in Mode One or Two, CEAs should be inserted, as necessary, to maintain reactor power constant (later in the core cycle) and the overcooling event is compensated for by adjusting turbine load.

D. Incorrect - Per AOP-7K, Overcooling Event in Mode One or Two, CEAs should be inserted, as necessary, to maintain reactor power constant (later in the core cycle) and the overcooling event is compensated for by adjusting turbine load.

Page 135 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Main Turbine Generator and MTC relationship Tier/Group: 2/2 045 Main Turbine Generator K5 - Knowledge of the operational implications of the following concepts as the apply to the MT/B System:

• KIA Info:

• K5.17 - Relationship between moderator temperature coefficient and boron concentration in RCS as T/G load increases RO Importance: 2.5 Proposed references to None be provided to applicant:

• Learning Objective:

10 CFR Part 55 Content:

Cognitive level: D Memory/Fundamental ~ Comprehension/Analysis Last NRC Exam used on: No previous NRC Exam use Exam Bank History: None AOP-7K, Overcooling Event in Mode 1 or Two Comments: Modified version of Q92905.

Page 136 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY The selected Pressurizer Level control channel process variable fails low at 100%

power.

Which ONE of the following describes the plant response? (Assume NO operator action is taken)

A. All heaters deenergize, letdown goes to minimum, standby charging pumps start, actual Pzr level I pressure rises and the reactor trips on High Pzr pressure.

B. All heaters energize, letdown goes to maximum, only the selected charging pump runs, actual Pzr level I pressure lowers and the reactor trips on TM/LP.

C. All heaters energize, letdown goes to minimum, actual Pzr level I pressure rise and the reactor trips on High Pressurizer Pressure.

D. All heaters deenergize; actual Pzr level I pressure lowers; the reactor trips on TM/LP.

Answer: A Answer Explanation:

A. Correct - With the level controller failing low, PLCS would respond to an indicated level lower than set point. Letdown valves would throttle back to raise level to setpoint. All charging pumps would start on level deviation. All heaters would deenergize based on pressurizer level being less than 101". Pressurizer bubble would be compressed and RCS pressure will rise until RPS high pressure trip setpoint is reached.

B. Incorrect - heaters will not energize due to failed detector indicating less than 101 inches, letdown does not go to maximum, all charging pumps start heaters will deenergize, but pressure level rises.

C. Incorrect - heaters will not energize due to failed detector indicating less than 101 inches D. Incorrect - heaters will deenergize, but pressure level rises.

12 NRC RO EXAM MASTER KEY Pressurizer Level Control Channel failure 2/2 011 - Pressurizer Level Control System

• A 1 - Ability to predict and/or monitor changes in KJA Info: parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:

• A 1.01 - PZR level and pressure RO Importance: 3.5 Proposed references to None be provided to applicant:

Recall the operating range of the Containment Learning Objective: Hi-Range Radiation Monitors and automatic actions occurring upon alarm setpoint exceeded.

10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: D Memory/Fundamental [2J Comprehension/Analysis Last NRC Exam used on: None LOI-2006 RO Remediation Audit (11/08) 01-35, Radiation Monitoring System; Technical references:

ARM 1(2) C10 annunciator window J-04.

Modified Q74600 Page 138 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following describes the reason why a Circulating Water Pump (CWP) handswitch is returned to AUTO and not held in the START position until starting current lowers to running current?

A. Holding the handswitch in START prevents the motor protective relay circuit from arming and immediately reopens the breaker.

B. Holding the handswitch in START prevents the motor protective relay circuit from arming and the only protection is an ()vercurrent trip.

C. Holding the handswitch in START prevents the starting current from dissipating and causes the motor to trip on overcurrent.

D. Holding the handswitch in START prE~vents the charging spring motor from recharging to allow closing breaker upon subsequent starts.

Answer: B Answer Explanation:

A. Incorrect - First part of statement is true, however, breaker does not trip open immediately.

B. Correct - Per OI-14A, Caution on page 14, prior to starting any CWP this is stated.

C. Incorrect - Starting current will dissipate if handswitch held in START, it does not remain once pump is started. If held in start, only motor overcurrent protection is active to trip breaker open.

D. Incorrect - This does not prevent charging spring motor from recharging. Once breaker is closed the charging spring motor resets for next closing operation.

Page 139 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Circulating Water Tier/Group: 2/2 075 - Circulatin£1 Water

• • 2.1 - Conduct of Operations

• KIA Info:

• 2.1.32 - Ability to explain and apply system limits and precautions.

RO Importance: 3.8 Proposed references to be None provided to applicant:

Apply all system limits (cautions and notes) and precautions i Learning Objective:

when starting or stopping a Circulating Water Pump.

10 CFR Part 55 Content: 55.41 (b){10)

Question source:

Cognitive level: o Comprehension/Analysis

. Last NRC Exam used on:

Exam Bank History:

Technical references: OI-14A, Circulating Water System, Section 6.1.B page 14.

None Page 140 of 162

2012 NRC RO EXAM MASTER KEY A reactor trip has occurred from full power on Unit-2. The following conditions exist:

• Reactivity Control is complete.

• Pressurizer level has stabilized at 1210".

• No automatic ESFAS actuations hav~3 occurred.

• RCS pressure is 1710 PSIA and slowly decreasing.

• Both SG levels are -150" and decreasing.

• SG pressures are 785 PSIA

• T COLD is 516 of and lowering Which ONE of the following sets of operator actions is required?

A. Manually initiate SIAS, trip 2 RCPs, and shut the MSIVs.

B. Manually initiate SIAS, SGIS, and trip all RCPs.

C. Manually initiate SIAS, CIS, and AFAS.

D. Block SIAS, throttle AFW flow, and shut the MSIVs.

Answer: A Answer Explanation:

A. Correct - SIAS should have been initiated by 1725 PSIA, per EOP-O, 2 RCPs are tripped after verifying SIAS.

B. Incorrect - RCS pressure is high enough to support 2 RCPs running per Attachment 1 and SGIS is not required to initiate above a S/G pressure 685 of PSIA.

C. Incorrect - AFAS setpoints are not challenged and there is no information to support initiating CIS.

D. Incorrect - SIAS should not be blocked in EOP-O; although, not stated, it is it is inferred that the conditions are shortly after the trip. If in an Optimal Recovery procedure, there are steps to block SIAS prior to actuation. Also, with S/G levels dropping throttling AFW should not be accomplished at this point.

Page 141 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Question**65 (Q25Q69)

Topic: Operator actions for SLB Tier/Group: /2 035 - Steam Generator

• K3 - Knowledge of the effect that a loss or malfunction of KIA Info:

the S/Gs will have on the following:

• K3.01 - ReS RO Importance: 4.4 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.41 (b)(5)

Question source:

Cognitive level: D Memory/Fundamental l8J Comprehension/Analysis Last NRC Exam used on: No record of use on any NRC exam Exam Bank History: 2010 LOR Session 2 quiz Technical references: EOP-O, Post Trip Immediate Actions EOP-4, Excess Steam Demand Event Comments: None Page 142 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following is required if relief for a brief period is necessary when performing the duties of a "Dedicated Operator" in the Control Room assigned by the Shift Manager (SM)?

A. Any licensed operator on watch in Control Room may relieve following a verbal brief by the "Dedicated Operator" on status of the evolution in progress and any special conditions that may require attention or action during "Dedicated Operators" absence.

B. A licensed operator standing a recertification watch who attended the pre-job brief may relieve with SM permission after being verbally briefed on any special conditions that may require attention or action during the "Dedicated Operators" absence.

C. The Dedicated SRO who attended the pre-job brief for evolution in progress, received a verbal brief by the "Dedicated Operator" on status of evolution in progress, and requires no "hands-on" operations during the "Dedicated Operators" absence.

D. Relieving individual attended the pre-job brief and has permission from the SM/CRS to relieve the "Dedicated Operator", received a verbal brief on the status of the evolution in progress and any special conditions that may require attention or action during absence, and have no concurrent duties.

Answer: D Answer Explanation:

A. Incorrect - One of the requirements is that the relieving individual must have NO concurrent duties wh~n relieving the Dedicated Operator for a brief period.

B. Incorrect - A licensed operator standing recertification watch has an inactive license until signed off by GS-SO and may never assume the role of "Dedicated Operator" and is not allowed to manipulate controls on boards independently; second part of statement is right as permission is needed by relieving individual from SM/CRS and verbal brief on any special conditions that may require attention or action during absence is part of requirement.

C. Incorrect - As stated before Dedicated SRO may not have any concurrent duties and may be required to perform "hands-on" manipulations as needed during "Dedicated Operators" brief absence.

D. Correct - This is what is required per NO-1-200 page 28 Section 5.2.8.3 Page 143 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY

. Short term relief of Dedicated Operator Tier/Group:

2.1 - Conduct of Operations KIA Info:

• 2.1.3 - Knowledge of shift or short-term relief turnover practices.

RO Importance: 3.7 Proposed references to be None provided to applicant:

Recall the purpose of each of the safety function boxed Learning Objective:

steps of EOP-O.

10 CFR Part 55 Content:

I Cognitive level: k8J Memory/Fundamental o Comprehension/Analysis I Last NRC Exam used on: No record of use on any exam Exam Bank History: N/A NO-1-200 Page 28 Section 5.2.B.3 None i

Page 144 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY The plant tripped from 100% power due to a LOCA. EOP-O actions were taken and the crew transitioned to EOP-5, Loss of Coolant Accident.

The following conditions exist 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after entry into EOP-5:

• SIAS, CIS, and CSAS were verified in EOP-O

• RAS has actuated and been verified

• Containment pressure is 3.0 PSIG and slowly lowering

• RCS pressure is 360 PSIA and slowly lowering

• RCS subcooling is O°F

• ALL RVLMS lights are green on PAMS

• Containment Wide Range Level indicates 50 inches and steady

• Both Containment Spray Pumps are running normally

• HPSI flow is throttled (and balanced) to the minimum allowed per EOP Att. 10, HPSI Flow

• 11 and 1~~ HPSI Pump current and flow are fluctuating Per EOP-5, which ONE of the following actions must be taken to stabilize HPSI flow?

A. Secure both Containment Spray Pumps.

B. Throttle HPSI injection flow.

C. Secure ONLY one Containment Spray Pump.

D. Secure one HPSI Pump and readjust HPSI flow to minimum allowed.

Answer: A Answer Explanation:

A. Correct - Since HPSI flow is at the minimum, EOP-5 Step S.1.j.2 states secure BOTH spray pumps and THEN check for acceptable HPSI pump performance.

B. Incorrect - Throttling HPSI flow even more is NOT allowed as it is at the minimum required cooling flow for time since LOCA.

C. Incorrect - Securing only one pump does provide relief for HPSI pumps; however, the sump level is adequate and NOT the cause of cavitation.

It is the sump screens becoming clogged.

D. Incorrect - Securing a HPSI pump would significantly reduce the flow to the vessel and since spray pumps are still operating, it would be more appropriate to secure both of these pumps before stopping a HPSI pump.

Page 145 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: i Conduct of Operations Tier/Group: 3 2.1 - Conduct of Operations KIA Info:

• 2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

RO Importance: 4.3 Proposed references to None i be provided to applicant:

Given a LOCA in progress, evaluate plant conditions and Learning Objective: perform the required action to prevent HPSI pump cavitation i

10 CFR Part 55 Content: 55.41(b)(10)

Question source:

Cognitive level: cg] Comprehension/Analysis Last NRC Exam used on: NEW Exam Bank History: None Technical references: EOP-5 Block Step S.1.j.2 Comments: Adapted from Millstone 2, 2008 NRC RO exam Page 146 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Which ONE of the following explains the reason for the difference between the required shutdown boron concentration for Mode 3/4 and Mode 5?

A. Less positive reactivity is inserted during a cooldown in Mode 3 or 4 than Mode 5 at EOG.

B. More positive reactivity is inserted during a cooldown in Mode 3 or 4 than Mode 5 at EOG.

G. Less negative reactivity is inserted during a cooldown in Mode 3 or 4 than Mode 5 at EOG.

D. More negative reactivity is inserted during a cooldown in Mode 3 or 4 than Mode 5 at EOG.

Answer: B Answer Explanation:

A. Incorrect - On a cooldown, more NOT less positive reactivity is added in Mode 3 or 4 than Mode 5 at EOG due to cooldown from a steam line break which is most restrictive accident to challenge 8DM. Mode 5 is below 200°F and no cooldown from a steam accident would occur.

B. Correct - On a cooldown, more positive reactivity is added in Mode 3 or 4 than Mode 5 at EOG due to cooldown from a steam line break which is most restrictive accident to challenge 8DM. Mode 5 is below 200 OF and no cooldown from a steam accident would occur.

G. Incorrect - The most limiting M8LB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line outside containment, initiated at the end of core life. Positive NOT negative reactivity is added at EOG.

D. Incorrect - The most limiting M8LB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line outside containment, initiated at the end of core life. Positive NOT negative reactivity is added at EOG.

Page 147 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Generic - Conduct of Operations Tier/Group: 3 2.1 - Conduct of Operations

• 2.1.43 - Ability to use procedures to determine the KIA Info: effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

RO Importance: 4.1 Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content:

Cognitive level: ~ Memory/Fundamental Last NRC Exam used on: No record of use on any exam Exam Bank History:

None Page 148 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY Per EOP-O, which of the following sets of actions is performed if any Unit-1 MSR 2 nd Stage Source MOV or Unit-2 MSR 2 nd StagE~ Control valve fails to shut after the immediate actions have been performed? Assume NO loss of power has occurred.

A. For Unit-1: shut BOTH MSIVs; For Unit-2: shut BOTH MSIVs B. For Unit-1: place the MSR 2nd Stg Stm Source MOVs handswitch, 1-HS-4025 in the closed position; For Unit-2: depress the RESET button on the MSR control panel.

C. For Unit-1, close the MSR 2 nd Stage High Load MOVs and verify the MSR 2nd Stage Bypass Control valve panel loaders in manual with panel loader output at zero; For Unit-2, shut the Main Steam Supply to the MSR 2nd Stage isolation valve.

D. For Unit-1, shut the appropriate Main Steam Supply to MSR 2nd Stage manual isolation valve; For Unit-2, verify the MSR 2nd Stage bypass control valve panel loaders in manual with panel loader output at zero.

Answer: C Answer Explanation:

A. Incorrect - These actions are performed for loss of power conditions, turbine speed not lowering, MTSV fails to close(U-1) and TV fails to close (U-2)

B. Incorrect - These are the immediate actions for each unit which have been performed as stated in the stem.

C. Correct - Per EOP-O, these are the correct actions to do per alternate actions for turbine trip.

D. Incorrect - This is action for Unit-2 not Unit-1; these are actions for Unit-1 not Unit 2. Both of these actions are a part of alternate actions response.

Page 149 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Generic 2.2 - Equipment Control Tier/Group: 3 2.2 - Equipment Control

• 2.2.4 - (multi-unit license) Ability to explain the KJA Info: variations in control board/control room layouts, systems, instrumentation and procedural actions between units at a facility.

RO Importance: 3.6 Proposed references to be None provided to applicant:

Learning Objective: Recall how a Unit 1 and Unit 2 turbine trip are verified.

10 CFR Part 55 Content: 55.41 (b)( 10)

Question source:

Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: NEW Exam Bank History: None EOP-O Unit 1 and Unit 2 Ensure Turbine Trip step 8.3 None Page 150 of 162 Rev. 3

2 NRC RO EXAM MASTER KEY Unit-1 is in Mode 1 and the latest leakage reports are:

• 8.3 GPM - Pressurizer safety valve leakage

• 1.8 GPM - leakage past check valves from the RCS to the SI system

• 0.2 GPM - 12 Steam Generator primary-to-secondary leakage

• 10.9 GPM - total leakage Which ONE of the following pairs of Technical Specification RCS leakage limits is exceeded?

A. Primary to Secondary leakage and Identified leakage.

B. Primary to Secondary leakage and Pressure Boundary leakage.

C. Identified leakage and Unidentified leakage.

O. Pressure Boundary leakage and Identified leakage.

Answer: A Answer Explanation:

A. Correct - 12 S/G Primary to secondary leakage (0.2 GPM x 60 x 24 = 288 GPO) exceeds the T.S. limit of 100 GPO. Identified leakage is 10.3 GPM which is greater than the T.S. limit of 10 GPM.

B. Incorrect - 12 S/G Primary to secondary leakage (0.2 GPM x 60 x 24 = 288 GPO) exceeds the T.S. limit of 100 GPO; however no pressure boundary leakage exists. Tech Specs define Pressure Boundary leakage as "LEAKAGE (except primary to secondary LEAKAGE) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall".

C. Incorrect - Identified leakage of 10.3 GPM is greater than the T.S. limit of 10 GPM. Total leakage of 10.9 GPM minus Identified leakage of 10.3 GPM =

0.6 GPM unidentified leakage which does not exceed the T.S. limit of 1 GPM unidentified leakage.

O. Incorrect - Tech Specs define Pressure Boundary leakage as "LEAKAGE (except primary to secondary LEAKAGE) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall". No Pressure Boundary leakage exists. Identified leakage of 10.3 GPM is greater than the T.S. limit of 10 GPM.

Page 151 of 162 Rev. 3

2012 NRC RO EX,AM MASTER KEY Topic: Equipment Control- Tech Spec entry conditions Tier/Group: 3 2.2 - Equipment Control

• KiA Info:

• 2.2.42 - Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

RO Importance: 3.9 Proposed references to None be provided to applicant:

Given RCS leakage values, determine the leakage limits Learning Objective:

exceeded per tech spec LCO 3.4.13 10 CFR Part 55 Content: 55.41 (b)(10)

Question source:

Cognitive level: [8J Comprehension/Analysis Last NRC Exam used on:

'Exam Bank History: LOR 11-6C Biennial Written exam (11/11)

Technical references: Unit-1, Tech Spec 3.4.13 and leakage definitions Comments: Modified from Q92906 Page 152 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Per CCNPP procedures, which ONE of the following would be the first threshold TEDE dose limit requiring an extension and required approval?

A. TEDE annual dose limit to exceed 1,~~50 but not greater than 4,000 millirem/yr; your department Manager and GS.

B. TEDE annual dose limit to exceed 2,000 but not greater than 3,000 millirem/yr; GS-RP, your department Manager and GS.

C. TEDE annual dose limit to exceed 3,000 but not greater than 4,000 millirem/yr; GS-RP, your department Manager and GS.

D. TEDE annual dose limit to exceed 4,000 but not greater than 5,000 millirem/yr; GS-RP, your department Manager and GS, PGM, and VP-CCNPP.

Answer: B Answer Explanation:

A. Incorrect - This value is still below the first threshold of 2,000 mRem/yr to requiring an extension and approval.

B. Correct - Per Table 2 of RP-1-100, the first dose extension and approval is required when exceeding 2,000 mRem/yr C. Incorrect - This would be the next threshold requiring an extension per Table 2; also approval of PGM is required. However, this includes dose from ALL sources (this applies for contractors and permanent personnel who worked at other nuclear sites).

D. Incorrect - This is the next threshold per Table 2 and it requires approval from VP-CCNPP in addition to the approvals to exceed 3,000 mRem/yr without exceeding the federal limit of 5,000 millirem/yr.

Page 153 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY i Radiation Control - Exposure Limits Tier/Group: 3 2.3 - Radiation Control KIA Info:

• 2.3.4 - Knowledge of radiation exposure limits under I normal or emergency conditions.

I RO Importance: 3.2

, Proposed references to None I be provided to applicant:

State whose approval is required to exceed CCNPP I Learning Objective:

administrative dose limits.

10 CFR Part 55 Content: .55.41(b)(12)

Question source:

Cognitive level: *[g] Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use on any exam Exam Bank History: None

.Technical references: RP-1-100, Radiation Protection Table 2 Comments: None Page 154 of 162 Rev. 3

12 NRC RO EXAM MASTER KEY In accordance with CNG-OP-1.01-2003, Alarm Response and Control, if one or more inputs to a multiple input alarm is out of service, the alarm will be designated with a ...

A. Black Dot B. Blue Dot C. Red Dot D. Yellow Dot Answer: 0 Answer Explanation:

A. Incorrect - Per CNG-OP-1.01-2003, Alarm Response and Control, a Black dot placed on an annunciator window is used to signify one of the following:

• A maintenance activity in the station that causes an alarm on a repeated basis.

• For identification of a locked in alarm that is caused by a current station configuration due to maintenance in the field or an Operations' lineup.

• For placement on alarm windows of nuisance alarms with the approval of the Control Room Senior Reactor Operator.

B. Incorrect - Per CNG-OP-1.01-2003, Alarm Response and Control, a Blue dot placed on an annunciator window is used to signify the associated annunciator window has been taken out of service.

C. Incorrect - Per CNG-OP-1.01-2003, Alarm Response and Control, a Red dot placed on an annunciator window is used to signify the associated component or annunciator window is part of a tagout.

D. Correct - Per CNG-OP-1.01-2003, Alarm Response and Control, a Yellow dot placed on an annunciator window is used to signify that one or more inputs to a multiple input annunciator are out of service.

Page 155 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY What color dot indicates an input to a multiple input Topic:

annunciator window is OOS

. Tier/Group: 3 2.2 - Equipment Control KIA Info:

• 2.2.43 - Knowledge of the process used to track inoperable alarms.

RO Importance: 3.0 i

• Proposed references to None be provided to applicant:

Learning Objective:

10 CFR Part 55 Content: 55.41(b)(10)

Question source:

Cognitive level: ~ Memory/Fundamental Last NRC Exam used on: No previous use Exam Bank History: LOI-2006 Audit Exam Technical references: CNG-OP-1.01-2003, Alarm Response and Control Page 156 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY As a licensed operator you have assumed the watch as the ABO. You are signed in on RWP-2, Operations Activities, including Fuel Shuffle, and Non-High radiation areas.

An emergency situation requires you to enter a locked high radiation area. No EAL classification thresholds have been met.

Which ONE of the following choices describes the requirements to gain access to the area?

A. Sign in under an Emergency Work Permit (EWP) and obtain RP coverage.

B. Enter the area under your current Radiological Work Permit (RWP) without RP coverage.

C. Obtain RP coverage and enter the area under your current RWP.

D. Sign in under the applicable EWP, RP coverage is not required if another operator is available.

Answer: C Answer Explanation:

A Incorrect - Emergency Work Permits are only used when EAL of Alert or higher is declared. They are used for plant equipment, lifesaving, and protecting large populations. EWPs are not used under routine operations.

B. Incorrect - RWP has the following contingency:

• EMERGENCY CONTINGENCY: In the event of an emergency, responders may enter any areas using this activity. Continuous RP coverage is required.

Following closure of the emergency, responders may not enter the RCA without approval of RP Supervision.

C. Correct - RWP has the following contingency:

• EMERGENCY CONTINGENCY: In the event of an emergency, responders may enter any areas using this activity. Continuous RP coverage is required.

Following closure of the emergency, responders may not enter the RCA without approval of RP Supervision.

D. Incorrect - Emergency Work Permits are only used when EAL of Alert or higher is declared. They are used for plant equipment, lifesaving, and protecting large populations. EWPs are not used under routine operations.

Page 157 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Topic: Generic 2.3 - Radiation Control Tier/Group: 3 2.3 - Radiation Control

• 2.3.12 - Knowledge of radiological safety principles KIA Info: pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

RO Importance: 3.2 Proposed references to None be provided to applicant:

Apply the requirements of RP-1-1 00 for Locked High Learning Objective:

Radiation Access.

10 CFR Part 55 Content: 55.41(b)(12)

Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No record of use Exam Bank History: LOI-2008 Admin Comp (06/10)

Technical references: RWP-2

,Comments: None Page 158 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Upon entry into an emergency operating procedure it becomes necessary to perform actions that are not contained within the controlling technical procedure and that are not parallel actions.

Which ONE of the following describes the minimum approval required to deviate from the emergency operating procedure?

A. At least 2 Senior Reactor Operators.

B. The Shift Manager AND the Control Room Supervisor.

C. The Shift Technical Advisor.

D. The Shift Manager.

Answer: D Answer Explanation:

A. Incorrect - NO-1-201, Calvert Cliffs Operating Manual, specifies "Deviations shall be approved by the 8M or the CR8 in the absence of the 8M".

B. Incorrect - NO-1-201, Calvert Cliffs Operating Manual, specifies "Deviations shall be approved by the 8M or the CR8 in the absence of the 8M".

C. Incorrect - NO-1-201, Calvert Cliffs Operating Manual, specifies "Deviations shall be approved by the 8M or the CR8 in the absence of the 8M" but not the 8TA.

D. Correct - NO-1-201, Calvert Cliffs Operating Manual, specifies "Deviations shall be approved by the SM or the CRS in the absence of the SM". "For deviations approved by the CRS, the CRS shall inform the SM as soon as practical".

2012 NRC RO EXAM MASTER KEY Topic: 2.4 - Emergency Procedures Tier/Group: 3 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.14 - Knowledge of general guidelines for EOP usage.

RO Importance: 3.8 Proposed references to None be provided to applicant:

Apply the requirements of NO-1-201, Calvert Cliffs Operating Learning Objective:

Manual, for deviation from an approved procedure.

10 CFR Part 55 Content: 55.41(b)(10)

Question source: ~ Bank D Modified DNew Cognitive level: ~ Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: No previous NRC Exam use Exam Bank History: LOI-2010 Panel Comp (06/11)

Technical references: NO-1-201, Calvert Cliffs Operating Manual Comments: None Page 160 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY You are attending LOR training with your Ops crew when an Alert is declared by the Operating Crew.

The Shift Manager (SM) makes an announcement over the plant page system, directing all ERO members to report to their designated assembly areas.

At which ONE of the following locations should you assemble?

A. Assemble in the South Service Building Cafeteria.

B. Assemble in the Control Room behind the electrical panels.

C. Assemble outside the GS-Ops Training Office on 2 nd floor of OTF.

D. Assemble in the pre-designated area in the OTF/NOF first floor hallway.

Answer: A Answer Explanation:

A. Correct - Per ERPIP-317, this is where Operators in training will assemble, for an Alert declaration or higher, for accountability and assignment of tasks when directed by Control Room.

B. Incorrect - This is where "On-Shift" Operators would assemble, for an Alert declaration or higher, if not involved in actions to address emergency event in progress.

C. Incorrect - This is where Ops Training personnel assemble, for an Alert declaration or higher, if they do not have an assigned position in the ERO.

D. Incorrect - This would only be appropriate for non ERO personnel who have a regular work location within the protected area. Operators are considered part of the ERO when on site.

Page 161 of 162 Rev. 3

2012 NRC RO EXAM MASTER KEY Training Crew Assembly Area for ERPIP declaration I Tier/Group: 3 2.4 - Emergency Procedures / Plan KIA Info:

• 2.4.29 - Knowledge of the emergency plan RO Importance: 3.1 Proposed references to None be provided to applicant:

Determine your assembly area when an Alert or higher EAL Learning Objective:

is declared.

10 CFR Part 55 Content: 55.41 (b}(10)

Question source:

Cognitive level: [g1 Memory/Fundamental D Comprehension/Analysis Last NRC Exam used on: Millstone 2,2008 RO exam Exam Bank History: No record of use on any exam Technical references: ERPIP-317, Operations Team (OSC)

Question stem modified from Millstone 2,2008 RO exam to Comments:

reflect Calvert Cliffs emergency plan.

Page 162 of 162 Rev. 3