Text

Draft Written Exam (Folder 2)
	ML101720560
Person / Time
Site:	Ginna
Issue date:	05/05/2010
From:	; Constellation Energy Nuclear Group
To:	Todd Fish; Operations Branch I
	HANSELL S
Shared Package
ML092470059	List: IR 05000244/2010301; ML100400205; ML101680575; ML101720347; ML101720553; ML101720560; ML101720564; ML102320373; ML102320386; ML102320393; ML102320408; ML102520045; ML102520054; ML102710006;
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Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2

.... ~-

Group#: 1

-~ ....

KJA# 022 K1.01 Importance Rating 3.5 Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: SWS/cooling system Proposed Question: RO Question # 1 Plant conditions:

The reactor has tripped.

RCS pressure is 1820 psig and lowering.

Containment pressure is 4 psig and rising.

SG pressures are 1000 psig and stable.

The crew is performing actions of E-O, Reactor Trip or Safety Injection.

Which ONE of the following describes the Containment Cooling alignment for these conditions?

Service Water Outlet Valve Service Water Bypass Valve FCV-4561 FCV-4562 Throttled Throttled A.

Full Open Throttled B.'

Throttled Closed C.

Full Open Full Open D.

Proposed Answer: D Explanation (Optional):

Incorrect. Plant conditions represent safety injection actuated due to high containment pressure. When SI actuates, Service Water to Containment Coolers goes to full open.

A.

Plausible because this is the normal lineup Incorrect. The first part is correct, making it plausible, and also logical that the bypass B. valve could remain throttled.

1 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KlA# 008 K1.02 Importance Rating 3.3 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Loads cooled by CONS Proposed Question: RO Question # 2 Which ONE of the following describes the component(s) supplied by Component Cooling Water that is (are) isolated by automatic closure of Containment Isolation Motor Operated Valves (MOVs)?

Reactor Support Cooling Pad ONLY.

A.

Excess Letdown Heat Exchanger ONLY.

B.

Non-Regenerative Heat Exchanger ONLY.

C.

Reactor Support Cooling Pad, Non-Regenerative Heat Exchanger AND Excess D. Letdown Heat Exchanger.

Proposed Answer: A Explanation (Optional):

Correct. MOV-813 and MOV-814 receive an automatic isolation signal (T). This A. isolates Reactor Support Cooling pad only Incorrect. Plausible because the Excess letdown HX has an isolation boundary from B. the same line as Reactor Support (MOV-817)

Incorrect. Plausible because the Non-Regenerative HX is isolated but not located in C. Containment, and the second part of the option is correct Incorrect. Plausible because it identifies all 3 lines in Containment (with exception of RCPs) and it is logical that a Containment Isolation valve would isolate the loads in D.

CTMT Technical Reference(s): 33013-1246, sheet 2 (Attach if not previously provided) 3 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None R28091C, Obj 1.04 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 D~sign, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

4 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 064 K2.02 Importance Rating 2.8 Knowledge of bus power supplies to the following: Fuel oil pumps Proposed Question: RO Question # 3 Which ONE of the following describes the power supply to 'A' and 'B' EDG Fuel Oil Transfer Pumps?

MCC C and MCC D A.

MCC K and MCC P B.

MCC Hand MCC J C.

MCC Land MCC M D.

Proposed Answer: C Explanation (Optional):

Incorrect. Plausible because the MCCs are paired in same way as correct MCCs and also because the MCCs are located in Aux Building, where safety related equipment is A.

located Plausible because the MCCs are paired in same way as correct MCCs and also because the MCCs are located in Aux Building, where safety related equipment is B.

located Correct.

C.

Plausible because the MCCs are paired in same way as correct MCCs and also because the MCCs are located in Aux Building, where safety related equipment is D.

located Technical Reference(s): P-12 (Attach if not previously provided) 5 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None R0801 C, Obj 1.05 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

6 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 078 K2.02 Importance Rating 3.3

~.--" ...-

Knowledge of bus power supplies to the foliowin~1 Emergency air compressor Proposed Question: RO Question # 4 Plant conditions:

The plant is at 100% power

'C' Instrument Air compressor is running

All major control systems in AUTO

Service and Instrument Air are cross-connected per T-1 C, Instrument AirlService Air Cross Connect

The following alarms are received in the control room:

J-29, 480V TRANSFORMER BREAKER TRIP

L-22, BUS 15 UNDERVOLTAGE NON-SAFEGUARD Assuming no action taken by the crew, which ONE one of the following describes the Air Compressor that will be running?

'A' Instrument Air Compressor A.

'B' Instrument Air Compressor B.

'C' Instrument Air Compressor C.

Service Air Compressor D.

Proposed Answer: A Explanation (Optional):

Correct. Instrument Air Compressor 'A' is powered from 480VAC Bus 13. It will start on A. 105 psig air header pressure decreasing Incorrect. Plausible because air compressor is powered from same series of busses but B. is powered from 480VAC Bus 15.

7 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because air compressor is powered from same series of busses but C. is powered from 480VAC Bus 15.

Incorrect. Plausible because air compressor is powered from same series of busses but is powered from 480VAC Bus 15. The air compressor will not automatically start on D.

loss of another air compressor.

Technical Reference(s): R4701C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R4701 C Obj 5.01 Learning Objective: (As available)

Question Source: Bank # WTSI66526 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Ginna 2006 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or A.nalysis x 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

Editorial modified 2006 question by changing failure to allow for interpretation of alarms 8

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 026 K3.01 Importance Rating 3.9 Knowledge of the effect that a loss or malfunction of the CSS will have on the following: CCS Proposed Question: RO Question # 5 Plant conditions:

The plant is operating at 100% power.

'A' Containment Spray pump has been declared INOPERABLE due to an oil leak.

'A' and 'D' Containment Recirc Fan Coolers are INOPERABLE and isolated due to service water leaks.

All other ECCS equipment is OPERABLE.

Which ONE of the following describes whether or not Containment pressure can be maintained within design limits following a LOCA, and the reason why?

NO, two (2) Containment Spray pumps must be OPERABLE to meet the design basis A for post accident containment cooling and pressure control.

NO, if one Containment Spray pump is out of service, four (4) Recirc Fans are required to be OPERABLE to meet the design basis for post accident containment cooling and B.

pressure control.

YES, only one (1) OPERABLE Containment Spray pump is required to maintain Containment pressure within limits. Recirc Fan Coolers are not considered for post C.

accident Containment cooling and pressure control.

YES, a single OPERABLE Containment Spray pump and two (2) Recirc Fan Coolers OPERABLE meets the deSign basis for post accident Containment cooling and D.

pressure control.

Proposed Answer: D Explanation (Optional):

Incorrect. Only 1 train is required based upon single failure criteria. Plausible because A the applicant may assume that only spray provides post accident pressure control 9

517/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible for same reason as option A. If applicant believes that alignment requirements change based upon equipment out of service, then this option may be B.

picked Incorrect. Plausible because pressure control is normally associated with containment C. spray, while containment coolers are associated with temperature control Correct. One full train of each is required to meet design requirements for post accident D. cooling and pressure control Technical Reference(s): ITS Basis 3.6.6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R2201 C, Obj 1.06a, b Learning Objective: (As available)

Question Source: Bank # x Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity. and functions of emergency systems.

Comments:

Alot of wording changes but left as bank item because conditions are essentially the same 10 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KJA# 076 K3.07 Importance Rating 3.7 Knowledge of the effect that a loss or malfunction of the SWS will have on the following: ESF loads Proposed Question: RO Question # 6 Plant conditions:

The plant is at 100% power.

'8' and 'D' Service Water Pumps are in service.

The following alarms are received in the Control Room:

PPCS-P2160, SERVICE WATER PUMPS A cSt 8 HEADER

C-10, CONTAINMENT RECIRC CLRS WATER OUTLET LO FLOW 1050 GPM

The CO reports that '8' Service Water Pump has tripped.

'A' and 'C' Service Water Pumps tripped upon starting.

The CRS enters AP-SW.2, Loss of Service Water

'D' Service Water Pump is running.

Service Water Header pressure is 55 psig and stable.

Which ONE of the following describes the action required in accordance with AP-SW.2?

Trip the reactor; Align Alternate Cooling Water to 'A' EDG during performance of E-O, A. Reactor Trip or Safety Injection.

Trip the reactor; Align Alternate Cooling Water to '8' EDG during performance of E-O,

8. Reactor Trip or Safety Injection.

Reactor Trip is NOT required; Align Alternate Cooling Water to 'A' EDG and isolate C. Service Water to Non-Essential loads.

Reactor Trip is NOT required; Align Alternate Cooling Water to '8' EDG and isolate D. Service Water to Non-Essential loads.

Proposed Answer: C 11 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Reactor trip is plausible because it will be performed if there are no SW Pumps operating. In this case, there is one operating. Additionally, Alternate Cooling A.

Water will be supplied to 'A' EDG Incorrect. As in Option A, reactor trip is not required at this time. Alt cooling will be B. aligned to 'A' EDG, not 'B' EDG Correct. See AP-SW.2 C.

Incorrect. Plausible because Alternate Cooling is aligned, but wrong because this D, option identifies the wrong EDG

. AP-SW.2 Rev 8 Technical Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP33C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamt~ntal Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

12 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 039 K4.06 Importance Rating 3.3 Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following:

Prevent reverse steam flow on steam line break Proposed Question: RO Question # 7 Plant conditions:

The plant is in Mode 3.

RCS temperature is 547°F.

The Main Condenser is in service.

A Steam Line Break occurs on 'A' SG inside Containment.

Which ONE of the following describes the plant response to this event?

'A' Main Steam Line Pressure will drop at ,a higher rate than 'B' Main Steam Line pressure; MSIVs will receive a CLOSE signal on a Containment pressure setpoint of 4 A.

psig.

'A' Main Steam Line Pressure will drop at a higher rate than 'B' Main Steam Line pressure; MSIVs will receive a CLOSE signal on a Containment pressure setpoint of 18 B.

psig.

BOTH Main Steam Line pressures will lower at approximately the same rate; MSIVs will C. receive a CLOSE signal on a Containment pressure setpoint of 4 psig.

BOTH Main Steam Line pressures will lower at approximately the same rate; MSIVs will D. receive a CLOSE signal on a Containment pressure setpoint of 18 psig.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because the steam pressure response is correct and because there are 2 other MSLI signals that are dependent on SI being generated, which is at 4 psig.

A.

These signals are coincident with Low Tavg B. Correct. Containment pressure at 18 pSig (HIGH-2) will initiate MSIV closure. One SG 13 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination stays constant because there is a check valve downstream of the MSIV that will close on slight pressure differences between steam lines, and in this case, with the Main Condenser in service, the steam lines are connected Incorrect. Plausible because if only the MSIV closure were taken into consideration and not the check valve, and the MSIV closure signals associated with SI, this option would C.

be chosen.

Incorrect. Plausible because the setpoint is correct and because the applicant may not D. consider that there is a check valve prior to MSLI

. R4001 C Rev 23 Technical Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R4001 C, Obj 1.04, 1.07 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic ;and manual features.

Comments:

14 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 013 K4.08

~ ........ --~

Importance Rating 3.1 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following Redundancy Proposed Question: RO Question # 8 Which ONE of the following describes an Engineered Safety Feature as defined by Technical Specifications, and the logic required for actuation of the feature?

Containment Pressure High; 2 out of 3 A.

Containment Pressure High; 2 out of 4 B.

Pressurizer Pressure High; 2 out of 3 C.

Pressurizer Pressure High; 2 out of 4 D.

Proposed Answer: A Explanation (Optional):

Correct. Containment Pressure High is an ESFAS instrument that will initiate Safety A. Injection and Containment Isolation.

Incorrect. Plausible because the parameter is correct, Logic is plausible because 2 of 4 B. is a standard logic.

Incorrect. PRZR High pressure is am RPS instrument, but plausible because PRZR low C. pressure is RPS as well as ESFAS. Logic is correct.

Incorrect. Same description as C, but logic is incorrect and plausible as in option B D.

TS 3.3.2 Technical Reference(s): AR-A-28 (Attach if not previously provided) 15 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None R3501C, Obj 1.13 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

16 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 004 K5.44 Importance Rating 3.2

- _..- ~.-" ....-

Knowledge of the operational implications of the following concepts as they apply to the CVCS:

Pressure response in PZR during in-and-out surge Proposed Question: RO Question # 9 Plant conditions:

The plant is at 100% power.

Control Rods are in MANUAL.

An EH System malfunction caused a load rejection of approximately 50 MWe.

Which ONE of the following describes the immediate effect on pressurizer pressure and Charging flow?

Pressurizer Pressure Charging Flow RISES RISES A.

RISES LOWERS B.

LOWERS RISES C.

LOWERS LOWERS D.

Proposed Answer: B Explanation (Optional):

Incorrect. Pressurizer pressure rises because RCS temperature rises during the load rejection. RCS water is forced into the pressurizer (lNSURGE), causing the bubble to A. be squeezed. The water forced into the pressurizer causes level to rise above program.

Charging flow is automatically reduced to bring level back to program Correct. As described in Option A above B.

Incorrect. Opposite of actual effect. After the immediate response of pressurizer C. pressure, the colder RCS water will tend to depressurize the saturated pressurizer.

This is why heaters are turned on for a high level deviation. If applicant considers this 17 5n1201 0 Rev Draft

Ginna 2010 NRC Written Examination phenomenon, they will choose this option Incorrect. Charging flow does immediately lower, but pressure doesn't lower until later D. in the event.

. ROC02C Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None ROC02C Obj Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

18 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 006 K5.01 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to ECCS:

Effects of temperatures on water level indications Proposed Question: RO Question # 10 Plant conditions:

A LOCA has occurred.

The crew is performing ES-1.2, Post LOCA Cooldown and Depressurization.

The CRS is evaluating whether to stop SI pumps during the RCS cooldown.

Adverse Containment values are in effect.

Which ONE of the following describes the parameters affected by adverse containment values when performing the SI flow reduction sequence in ES-1.2?

Pressurizer Level ONLY A.

RCS Subcooling ONLY B.

RCS Pressure AND Pressurizer Level c.

Pressurizer Level AND RCS Subcooling D.

Proposed Answer: D Explanation (Optional):

Incorrect. Answer is correct but not complete. RCS Subcooling is also considered A.

Incorrect. Answer is correct but not complete. Pressurizer Level is also considered B.

Incorrect. RCS pressure is plausible because it has an adverse value for stopping RHR

c. pumps earlier in the procedure Correct. Both parameters are higher for adverse values than normal values. This is D. based on either radiation levels or pressure in containment, with pressure causing high atmospheric temperatures that cause inaccurate instrument readings 19 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

I Reference ()

Techmca s : ES-1.2, Rev 33, Step 12 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RES12C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic ;and manual features.

Comments:

20 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 012 K6.03 Importance Rating 3.1 Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Trip logic circuits Proposed Question: RO Question # 11 The plant is at 100% power.

PRZR Pressure Channel 429 has failed and has been properly defeated in accordance with ER-INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure.

Which ONE of the following identifies the Reactor Trip and Safety Injection actuation logic required (from the remaining in-service channels) on Low PRZR Pressure?

Reactor Trip Safety Injection 1/3 1/2 A.

2/3 1/2 B.

1/3 2/3 C.

2/3 2/3 D.

Proposed Answer: A Explanation (Optional):

Correct. Reactor trip is normally 2 of 4 logic. The failed channel will be placed in trip and one additional channel in trip wifl initiate a reactor trip. Normally SI 2 of 3 on Low A.

PRZR Pressure, now 1 of 3 Incorrect. Plausible because P-11 and PRZR level reactor trips employ normal 2 of 3 B. logic. Also because SI logic is correct.

Incorrect. Plausible because applicant may not understand that ESF functions are C. placed in trip condition, they may consider the channel bypassed, not tripped.

21 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. If channels were placed in bypass instead of trip, this would be correct for RPS. For SI, it is incorrect but plausible because normal logic is 2 of 4. SI is 2 of 3 on D.

low PRZR pressure Technical Reference(s): P-1) Rx Control and Protection (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3501 C, Obj 1.06 Learning Objective: (As available)

Question Source: Bank # WTSI 59465 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: McGuire 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

22 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 003 K6.14 Importance Rating 2.6 Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

Starting requirements Proposed Question: RO Question # 12 Plant conditions:

The plant is in Mode 4.

RCS temperature is 325°F.

RCS pressure is 440 pSig.

Pressurizer level is 26%.

SG levels are 38% Narrow Range.

SG pressures are 100 psig.

The crew is preparing to start 'A' RCP with the following conditions:

#1 Seal DP is 280 psid.

VCT pressure is 24 psig.

Seal Injection is currently NOT being supplied to 'A' RCP.

CCW flow to 'A' RCP Thermal Barrier Heat Exchanger is 20 GPM.

CCW inlet temperature to 'A' RCP is 92°F.

Which ONE of the following describes the action required in order to start 'A' RCP in accordance with S-2.1, Reactor Coolant Pump Operation?

1 Seal DP must be raised to greater than 320 psid.

A.

Pressurizer level must be raised to greater than 38%.

B.

VCT pressure must be raised to greater than 25 psig.

C.

CCW flow to the Thermal Barrier Heat Exchanger must be raised to greater than 25 D. GPM.

Proposed Answer: D Explanation (Optional):

23 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because there is a lower limit on #1 seal DP, but it is 220 psid, not A. 320 psid.

Incorrect. Because RCS temperature is lE~ss than 330, Pressurizer level must be less than 38%, not greater than or equal to 38%, or SG temperature must be no greater than B.

RCS temperature (100 psig is a few degrees higher than RCS temperature).

Incorrect. VCT pressure is within limits, but plausible because there is a lower limit for C. RCP operation, but is 15 psig.

Correct. Without seal injection, CCW flow to TBHX must be greater than 25 GPM at an D. inlet temperature of less than 100.

Technical Reference(s): S-2.1, Rev 45 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1301C, Obj 1.10 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

24 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KJA# 007 A1.02 Importance Rating 2.7 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank pressure Proposed Question: RO Question # 13 Plant conditions:

The plant is at 100% power.

The following annunciator is received:

F-9, PRT HI PRESS 5 PSI

PRT pressure is 6 psig and RISING SLOWLY.

PRT level is 64% and STABLE.

If alloWed to continue, which ONE of the followin~l is the potential impact of this event and the actions required to restore PRT pressure?

(1) The PRT rupture disc will discharge to containment when pressure rises to_ __

(2) to prevent PRT rupture disc operation.

(1) 50 psig A. (2) Vent the PRT.

(1) 50 psig B. (2) Drain the PRT to the RCDT to reduce level and pressure.

(1)100 psig C. (2) Drain the PRT to the RCDT to reduce level and pressure.

(1 )100 psig D. (2) Vent the PRT.

Proposed Answer: D 25 5n /2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Plausible because action is correct but setpoint for PRT rupture disc is A. incorrect.

Incorrect. Plausible because action would be taken lAW the AR, but level is within the B. low end of the normal band, so that action would not be required. (Low level is 60.8%)

Incorrect. Rupture disc pressure is correct but action is incorrect as in option B C.

Correct. Rupture disc blows at 100 psid, and for high pressure without high level, D. venting the PRT using AOV-527 is the correct action AR-F-9 Technical Reference(s): P-2, RCS P & L (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1401C, Obj 1.07 Learning Objective: (As available)

Question Source: Bank # WTSI56080 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Wolf Creek 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic "md manual features.

Comments:

26 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 063 A1.01 Importance Rating 2.5 Ability to predict and/or monitor changes in parameters associated with operating the dc electrical system controls including: Battery capacity as it is affected by discharge rate Proposed Question: RO Question # 14 Plant conditions:

A loss of all AC power has occurred.

The crew is performing actions of ECA-O.O, Loss of All AC Power.

The crew is evaluating load shed of the DC Batteries Which ONE of the following describes the reason for requirement to shed non-essential DC loads in accordance with ECA-O.O, Loss of All AC Power?

Battery discharge rate is reduced to ensure the station meets the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months technical A. specification design basis requirement for battery capacity following a loss of AC power.

Battery discharge rate is reduced to ensure the station meets the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months technical B. specification design basis requirement for battery capacity following a loss of AC power.

Battery discharge rate is reduced to ensure the station meets the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months coping C. requirement for loss of all AC power Battery discharge rate is reduced to ensure the station meets the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months coping D. requirement for loss of all AC power Proposed Answer: 0 Explanation (Optional):

Incorrect. Plausible since there is a design basis assumption contained in TS, but it is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , not 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is plausible bE~cause it is the allowed TS action time for loss A.

of DC.

Incorrect. Plausible since the TS design basis is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , but load shedding is not B. required to achieve design basis operation of the battery 27 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. In accordance with the Station Blackout Program Plan, the coping requirement is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 2 Hours is plausible because of the TS action time in section C.

3,8 for battery or DC bus inoperability Correct. Load is shed to ensure that a loss of AC power lasting up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months will not D, fully discharge the batteries ECA-O.O step 17 and background document Technical Reference(s): Station Blackout Program Plan (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RECOOC, Obj 1.03 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI19213 (Note changes or attach parent)

New Question History: Last NRC Exam: Harris 2005 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

28 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2

~" ..... --~.--

Group # 1 KIA # 005 A2.01 Importance Rating 2.7 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation Proposed Question: RO Question # 15 Plant conditions:

The plant is in Mode 5.

Reduced Inventory Operations are in progress for work on Reactor Coolant Pumps.

'A' RHR pump is operating.

RHR flow was initially at 450 GPM.

RHR Discharge flow is currently oscillating between 50 and 800 GPM.

Pump motor amps are oscillating, and discharge pressure is fluctuating.

RCS temperature is 110°F and slowly rising.

The CRS enters AP-RHR.2, Loss of RHR While Operating at Reduced Inventory Conditions.

Which ONE of the following is occurring, and which of the following actions is required for operation of RHR Pumps in accordance with AP-RHR.2?

Air entrainment in RHR suction; Reduce RHR flow with RHR Pump 'A' running until A conditions stabilize.

Air entrainment in RHR suction; Stop 'A' RHR Pump; After restoring RCS conditions to B. support operation of an RHR Pump, restart one RHR Pump.

RHR Pump 'A' has reached Runout conditions; Reduce RHR flow with RHR Pump 'A' C. running until conditions stabilize.

RHR Pump 'A' has reached Runout conditions; Stop RHR Pump 'A'; After restoring RCS D. conditions to support operation of an RHR Pump, restart one RHR Pump.

Proposed Answer: B 29 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Action would be correct for normal cavitation, but at reduced RCS level, just A. reducing flow will not alleviate air entrainment Correct. Discharge pressure and flow fluc:tuations indicate that NPSH is lost. This is caused by low level and high flow for that level. Action is correct in accordance with B.

philosophy of AP-RHR.2. Air is vented from the system prior to restart of RHR pumps Incorrect. Plausible because actions are reasonable for the condition and prescribed by the procedure for flow being 2 high for existing level, and because high flow is C.

consistent with Runout (Fluctuations to 800 GPM)

Incorrect. Plausible because actions are correct for the actual condition presented, and D. because high flow is consistent with Runout (Fluctuations to 800 GPM)

AP-RHR.2 Westinghouse ARG-1 Technical Reference(s): Background, Loss of RHR at Mid- (Attach if not previously provided)

Loop conditions Proposed References to be provided to applicants during examination: None RAP25C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

30 517/2010 Rev Draft

Ginna 2010 NRC Written Examination 31 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 010 A2.01 Importance Rating 3.3 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures Proposed Question: RO Question # 16 Plant conditions:

The plant is at 100% power when a plant transient occurred.

After the transient, power is 96% and stable.

The following alarms are received in the control room:

F-6, PRESSURIZER HEATER BREAKER TRIP

F-10, PRESSURIZER LO PRESS 2205 PSI

PRZR pressure indicates 2200 psig and slowly lowering.

The HCO reports that PRZR Backup Heater breaker has tripped.

The CRS enters AP-PRZR.1, Abnormal Pressurizer Pressure.

Which ONE of the following describes the impact, if any, on Technical Specification LCOs, and the action required to restore Pressurizer Backup Heaters?

Technical Specification action is ...

REQUIRED; reset the breaker, place to ON, and verify load increase on Bus 14.

A.

REQUIRED; reset the breaker, place to ON, and verify load increase on Bus 16.

B.

NOT REQUIRED; reset the breaker, place to ON, and verify load increase on Bus 14.

c.

NOT REQUIRED; reset the breaker, place to ON, and verify load increase on Bus 16.

D.

Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible because everything is correct with the exception of bus 14 loading 32 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. LCO for RCS pressure must be entered (DNBR, 3.4.1) Action is correct, and B. pressurizer backup heaters are fed from Bus 16 Incorrect. Plausible because the applicant may only consider TS 3.4.9 for pressurizer heater capacity. In this case, the DNBR TS LCO must be addressed. Additionally, Bus C.

14 is wrong Incorrect. Plausible because the applicant may only consider TS 3.4.9 for pressurizer D. heater capacity. In this case, the DNBR TS LCO must be addressed.

TS 3.4.1 COLR Technical Reference(s): AP-PRZR.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP11 C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

33 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group '# 1 KIA # 059 A3.06 Importance Rating 3.2 Ability to monitor automatic operation of the MFW, including: Feedwater isolation Proposed Question: RO Question # 17 Plant conditions:

A plant shutdown is in progress.

Reactor power is at approximately 60%.

A failure of ADFCS for 'B' Steam Generator causes the 'B' Feedwater Regulating Valve to remain in its current position.

Assuming NO action by the crew, which ONE of the following describes the effect on the plant as turbine load is reduced?

'B' SG level will rise and a Feedwater Isolation signal will be sent to 'B' SG ONLY at A. 85% narrow range level

'B' SG level will rise and a Feedwater Isolation signal will be sent to BOTH SGs when 'B' B. SG reaches 85% narrow range level

'B' SG leve! will lower and a reactor trip will occur when 'B' SG !evel reaches 20%

C.

'B' SG level will lower and a reactor trip will occur when 'B' SG level reaches 17%

D.

Proposed Answer: A Explanation (Optional):

Correct. If feedwater requirements are lowering with the feedwater reg valve in its A. current position, overfeeding will occur at some point during the shutdown.

Incorrect. Feedwater isolation occurs only for the affected SG. Results in closing B. Feedwater Reg Valves and Feedwater Reg Bypass Valves Incorrect. Plausible because SG level will change but it will go in the opposite direction.

C.

The concept of a valve failing in position is easily confused. This setpoint is incorrect, 34 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination but plausible because it is significantly below normal level, and some APs require a trip at this setpoint Incorrect. Plausible because SG level will change but it will go in the opposite direction.

D. The concept of a valve failing in position is easily confused. Setpoint is correct however Technical Reference(s): 33013-1352, Sheet 9 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R4401 C, Obj 1.06 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

35 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 103 A3.01 Importance Rating 3.9 Ability to monitor automatic operation of the contain-ment system, including: Containment isolation Proposed Question: RO Question # 18 Plant conditions:

An RCS leak resulted in the following conditions:

TIME EVENT 0812 Manual Reactor Trip.

0826 Pressurizer Pressure 1800 psig and lowering.

0828 Manual Safety Injection.

0907 Containment Pressure 4 psig and rising.

0941 Containment Pressure 29 psig and rising.

1003 RCS Pressure 220 psig and stable.

Assuming NO additional actions were taken, which ONE of the following choices describes the EARLIEST time an AUTOMATIC Containment Isolation signal was generated?

0826 A.

0828 B.

0907 c.

0941 D.

Proposed Answer: C Explanation (Optional):

Incorrect. PRZR pressure is not quite below the SI setpoint yet. Plausible because it is A. close to setpoint.

36 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Manual SI does not initiate Containment isolation. Plausible because CI is B, associated with SI actuation, and does actuate on any other auto SI Correct. Containment pressure of 4 psig will initiate Si and CI will be actuated based on C, that signal Incorrect. Plausible because setpoint for Containment Spray is 28 psig, which is D. associated with containment pressure HIGH-3 signal Technical Reference(s): P-1, Rx Control and Protection (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3501 C, Obj 1.07 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 6E3192 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamf~ntal Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components. and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Modified times and shifted events to arrive at diff~3rent answer from Ginna 2007 37 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 062 A4.04 Importance Rating 2.6 Ability to manually operate and/or monitor in the control room: Local operation of breakers Proposed Question: RO Question # 19 Plant conditions:

The plant is in Mode 5 with the 'A' Residual Heat Removal Pump in service.

During a plant heatup the control power fuse for the 'A' Residual Heat Removal Pump blows.

Which ONE of the following describes how the Main Control Board Residual Heat Removal Pump indication and local breaker control is affected by the loss of control power?

Main Control Board red I green running indications will be lost.

Local OPEN 1 CLOSE light indication is available, and local breaker control will be lost A.

until control power is restored.

Main Control Board red I green running indications will be lost.

Local OPEN I CLOSE mechanical indication is available, and local breaker control is B.

possible without the control power.

Main Control Board red I green running indications will be available.

Local OPEN I CLOSE light indication is available, and local breaker control is C.

possible without the control power.

Main Control Board red 1 green running indications will be available.

Local OPEN I CLOSE mechanical indication is available, and local breaker control will D.

be lost until control power is restored.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because Main Control Board red I green running indications will be A. lost, however, local breaker control is possible without control power. The breaker may be mechanically operated locally 38 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. With a loss of control power, Main Control Board red / green running B. indications will be lost. Local breaker control is stili possible.

Incorrect. Plausible because local breaker control is possible without control power, C however, local OPEN I CLOSE light indication is NOT available.

Incorrect. Plausible because local OPEN/CLOSE indication is available, however, Main D. control board indications are lost and local control is available.

. LP RGF11C Technical Reference{s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RGF11C,Obj Learning Objective: (As available)

Question Source: Bank # WTSI63033 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Comanche Peak 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

39 5n/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 073 A4.02 Importance Rating 3.7 Ability to manually operate and/or monitor in the control room Radiation monitoring system control panel Proposed Question: RO Question # 20 Plant conditions:

R-5 indicates a RANGE Alarm

Detector Display indicates 0.00

Monitor Bar Graph display is extinguished Which ONE of the following describes the reason for this indication, and the condition required to reset the alarm?

Detector power loss; alarm must be manually reset once the condition is clear A

Detector power loss; alarm will automatically reset if the condition clears B.

Radiation field is below the instrument range; alarm will automatically reset once the C. condition is clear Radiation field is above the instrument range; alarm must be manually reset once the D. condition is clear Proposed Answer: C Explanation (Optional):

Incorrect. If the detector had a power loss, the display would read low. However, this A condition would cause a FAIL alarm.

Incorrect. Power loss will also cause a fail alarm, but it is correct that auto reset occurs B.

Correct. A RANGE alarm with display of all zeroes indicates that the instrument is C. below minimum range Incorrect. Plausible because there is an E-Value for over-range, (EEEEE) but if the D.

detector was over-ranged, it would automatically reset once the detector returned to 40 517/2010 Rev Draft

Ginna 2010 NRC Written Examination scale Technical Reference(s): R3901C, Rev 23 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3901 C, Obj 1.07 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamemtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 PurpQse and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

41 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2

- -.... ~ _ .... _

Group # 1 KJA# 061 2.4.34 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Proposed Question: RO Question # 21 Plant conditions:

The reactor was manually tripped in accordance with AP-CR.1, Control Room Inaccessibility.

The Control Room has been evacuated due to a toxic atmosphere.

There is NO fire in progress.

Which ONE of the following describes the preferred method for local operation of Auxiliary Feedwater?

MDAFW Pumps throttled to maintain approximately 52% Narrow Range SG level A.

MDAFW Pumps throttled to maintain approximately 350 inches Wide Range SG level B.

SAFW Pump throttled to maintain approximately 52% Narrow Range SG level C.

SAFW Pump throttled to maintain approximately 350 inches Wide Range SG level D.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because this is the normal level maintained in SGs. Also plausible A. because MDAFW is preferred Correct. See procedure step 6 performed by HCO B.

Incorrect. SAFW pump is plausible, because it is an alternate means of providing AFW and the pump also has local controls to start if required. NR level plausible because it is C.

the normal level maintained Incorrect. Same reason as C and also because WR level is correct D.

42 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Technical Reference(s): AP-CR.1, Rev. 24 (Attach if not previously provided)

R5101C Proposed References to be provided to applicants during examination: None RAP04C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

43 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 076 2.1.27 Importance Rating 3.9

~- ..... -

Conduct of Operations: Knowledge of system purpose and I or function.

Proposed Question: RO Question # 22 Plant conditions:

The plant is at 100% power.

'A' and '0' Service Water Pumps are running.

'B' and 'C' Service Water Pumps are selected.

Safety Injection occurs.

Immediately following the Safety Injection actuation, a loss of off-site power occurs.

Which ONE of the following describes (1) the number of Service Water Pumps that will be running in each train, and (2) a safety related load that will be automatically isolated?

(1) ONE A. (2) AFW Pump thrust bearing and oil coolers (1) ONE B. (2) CCW Heat Exchangers (1) TWO C. (2) AFW Pump thrust bearing and oil cooll~rs (1) TWO D. (2) CCW Heat Exchangers Proposed Answer: B Explanation (Optional):

Incorrect. One pump is selected in each train and will be running upon completion of A. the sequencing, because the running (but not selected) pump is stripped (load shed).

Correct. One pump will start (selected) in each train. Non-critical load is also correct B.

C. Incorrect. Number of pumps is incorrect but plausible because if this was SI only, both 44 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination pumps would be running Incorrect. Plausible because the load is correct, and if off-site power was available, D. both pumps would be running

. R5101C Rev 27 Technical Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R5101C, Obj 1.02 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

Meets KA because knowledge of the purpose of the system during ESFAS actuation is required to answer the question 45 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group '# 1 KlA# 013 K3.01 Importance Rating 4.4 Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Fuel Proposed Question: RO Question # 23 Plant conditions:

A LOCA has occurred.

RCS pressure is 100 psig.

SI Pumps 'A' and 'B' are tripped.

'C' SI Pump is running.

RHR Pump 'N and 'B' are tripped.

RVLlS indicates Rx Vessel level is 20%.

If this condition continues, which ONE of the following describes the effect on the fuel assemblies?

Fuel failure is NOT likely to occur. Minimum safety function requirements are met.

A.

Fuel failure is NOT likely to occur. SI Accumulator injection will maintain core cooling.

B.

Fuel failure is likely to occur. Minimum safety function train requirements are met, but C. RVLlS indication is too low to sustain core cooling.

Fuel failure is likely to occur. Minimum safety function train requirements are NOT met D. and RVLlS indication is too low to sustain core cooling.

Proposed Answer: D Explanation (Optional):

Incorrect. Only one SI pump running with RHR pump flow significantly below what it A. should be, at least one train of ECCS is not available or running properly.

Incorrect. Accumulators should have already injected and RVLlS indication shows that B. Reactor Vessel level is low.

C. Incorrect. Plausible because 'C' SI Pump is running, but wrong because it is not 46 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination providing enough flow to sustain core cooling since no other equipment is running, and it is true that RVLlS is low.

Correct. Minimum safety function requirements of at least one full train of ECCs is NOT D, met at this time, and fuel damage is likely to occur Technical Reference(s): UFSAR 15.6.4.1.3.2 (Att h'f t . I 'd d)

Westinghouse Setpoint Document ac I no prevIous y provi e Proposed References to be provided to applicants during examination: None N/A Learning Objective: (As available)

Question Source: Bank# WTSI52628 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: VC Summer 2006 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

47 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2

-'-~-- - -....- -

Group '/l! 1 KIA # 010 2.4.20 Importance Rating 3.8 Emergency Procedures I Plan: Knowledge of operational implications of EOP warnings, cautions, and notes.

Proposed Question: RO Question # 24 Plant conditions:

A Steam Generator Tube Rupture has occurr,ed.

Off-Site power was lost shortly after the reactor was tripped.

'A' Steam Generator is isolated.

RCS cooldown is in progress.

The crew is preparing to depressurize the ReS to refill the pressurizer.

Which ONE of the following is a concern associated with the depressurization for these conditions, in accordance with E-3, Steam Generator Tube Rupture?

Thermal shock to the pressurizer spray nozzle A.

Loss of RCS pressure control when the PRZR goes solid B.

Voiding in the upper head region of the RCS C.

Voiding in the RCS side of the tube bundlE~ region of the ruptured SG D.

Proposed Answer: C Explanation (Optional):

Incorrect. Since off-site power is lost, RCPs will be tripped, causing normal spray to be lost. PORVs are the second choice to depressurize. Plausible because this may be a A.

concern if aux spray had to be used Incorrect. Using the PORVs will cause PRZR level to rapidly rise, and the depressurization may need to be terminated in High PRZR level, but the reason for that B.

would be voiding under the head. pressure control would not be lost C. Correct. See caution prior to step 22. With no RCPs running, there is no circulation in 48 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination the head area. Depressurization may result in voiding Incorrect. Voiding should not be occurring in tube region because the SG is isolated D. and tube bundle is covered with water Technical Reference(s): E-3, caution for step 22 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REP03C, Obj 1.03 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

49 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group "1/: 1

- - -..... ~ - _..... _ -

KJA# 007 A3.01 Importance Rating 2.7 Ability to monitor automatic operation of the PRTS, including: Components which discharge to the PRT Proposed Question: RO Question # 25 Plant conditions:

The plant is at 100% power.

All control systems are operating normally in AUTO.

Letdown Pressure transmitter PT-135 fails lov~.

Which ONE of the following describes how Letdown Pressure Control Valve, PCV-135, responds and which tank level will increase as a result of this failure?

PCV-135 closes; A. RCDT level will increase.

PCV-135 closes; B. PRT level will increase.

PCV-135 opens; C. RCDT level will increase.

PCV-135 opens; D. PRT level will increase.

Proposed Answer: B Explanation (Optional):

Plausible because the RCDT is the only other tank in Ctmt that could receive water from A. the primary system.

Correct. If pressure input fails low, the valve will close in an attempt to maintain pressure. When the valve closes, Letdown is essentially isolated, and the LP relief valve B.

to the PRT (203) will lift.

50 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because the valve is reverse acting, which provides a common misconception on valve fail position. VCT and LRW are the 2 flow paths of water that C. are possible if the valve did fail open. VCT normally and Waste Holdup Tank if VCT level exceeded the high setpoint for divert.

Plausible because the valve is reverse acting, which provides a common misconception on valve fail position. VCT and LRW are the 2 flow paths of water that D. are possible if the valve did fail open. VCT normally and Waste Holdup Tank if VCT level exceeded the high setpoint for divert.

R1601C, Rev 24 Technical Reference(s): CPI-PRESS-135 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1601 C, Obj 1.07g, 1.1 OJ Learning Objective: (As available)

Question Source: Bank # WTSI63653 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Harris 2007 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

51 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 039 A2.05 Importance Rating 3.3 Ability to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on predictions, use procedures to correlct, control, or mitigate the consequences of those malfunctions or operations: Increasing steam demand, its relationship to increases in reactor power Proposed Question: RO Question # 26 Plant conditions:

Reactor Startup is in progress following a mid-cycle outage.

The plant is at 5% power when a SG ARV fails open.

RCS temperature decreases and stabilizes at 538°F.

Which of the following predicts the plant response and the operator actions required in accordance with plant procedures and Technical Specifications?

Reactor power rises; Restore Tave above the Technical Specification minimum or the A. unit must be subcritical in Mode 2 in a maximum time of 15 minutes Reactor power rises; Restore Tave above the Technical Specification minimum or the B. unit must be subcritical in Mode 2 in a maximum time of 30 minutes The reactor becomes subcritical; withdraw control rods to raise Tavg above the Technical Specification minimum within a maximum of 15 minutes or initiate a shutdown C.

to Mode 3 The reactor becomes subcritical; withdraw control rods to raise Tavg above the Technical Specification minimum within a maximum of 30 minutes or initiate a shutdown D.

to Mode 3 Proposed Answer: B Explanation (Optional):

Incorrect. Plausible since reactor power will increase, but temperature is not to be A. restored within 15 minutes. 30 minutes is allowed in accordance with TS 3.4.2.

52 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. Temperature may then be recovisred by using control rods in a slow and controlled manner. Temperature has to be restored to greater than 540°F within 30 B.

minutes due to the requirements of TS 3.4.2 Incorrect. Plausible since the 15 minute time limit is similar to the time associated with C. restoration, but the reactor does not become subcritical.

Incorrect. Plausible since the time is correct, but the reactor does not become D. subcritical.

Technical Reference(s): TS 3.4.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank # WTSI 19193 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Harris 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

53 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 063 A4.01

.... --~---- ....

Importance Rating 2.8 Ability to manually operate and/or monitor in the control room: Major breakers and control power fuses Proposed Question: RO Question # 27 During operation at power with the Reactor Trip Breakers (RTBs) closed, a loss of 125 VDC control power to one of the RTBs occurs.

Which ONE of the following describes the effect on the RTB?

RTB opens due to loss of power to the undervoltage trip coil.

A.

RTB opens due to loss of power to the shunt trip coil.

B.

RTB remains closed, and the undervoltag~3 trip coil will not function on a reactor trip

c. signal from the Reactor Protection System.

RTB remains closed, and the shunt trip coil will not open on a reactor trip signal from D. the Reactor Protection System.

Proposed Answer: A Explanation (Optional):

Correct. Indication, UV coil, and shunt trip coil receive power from DC bus. Loss of DC results in a loss of power to the UV coil, causing it to drop out and causing the breaker A.

to open.

Incorrect. The breaker does not trip on loss of power to shunt coil because the shunt coil requires control power to operate. The undervoltage coil losing power would cause B.

a reactor trip breaker to open.

Incorrect. Indication is lost and power is lost to UV coil, but plausible because the breaker shunt coil uses control power and it will not be capable of tripping on a shunt C.

trip.

Incorrect. Indication and shunt trip capability lost. Plausibility is as described in options D.

above.

54 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

. R3501 C Rev 28 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3501C, Obj 1.10 Learning Objective: (As available)

Question Source: Bank # WTSI4B308 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Callaway 2005 Question Cognitive Level: Memory or Fundaml9ntal Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

55 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KJA# 003 K5.02 Importance Rating 2.8

~--

Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP coastdown on RCS parameters Proposed Question: RO Question # 28 The plant is at 100% power:

RCP 'B' trips due to a failure of the RCP breaker.

The reactor trips.

Which ONE of the following would be CORRECT regarding the loop 'B' SG level and RCS Loop Delta T INITIAL responses following the Rep trip?

'B' SG level would shrink lower than 'A' SG immediately after the RCP trips.

A. RCS loop 'B' delta T would be lower than RCS loop 'A' delta T

'B' SG level would swell higher than 'A' SG immediately after the RCP trips.

B. RCS loop 'B' delta T would be lower than RCS loop 'A' delta T

'B' SG level would swell higher than 'A' SG immediately after the RCP trips.

C. RCS loop 'B' delta T would be higher than RCS loop 'A' delta T

'B' SG level would shrink lower than 'A' SG immediately after the RCP trips.

D. RCS loop 'B' delta Twould be higher than RCS loop 'A' delta T Proposed Answer: A Explanation (Optional):

Correct. SG level will shrink immediately following the RCP trip due to less heat input. Loop delta T will lower due to no heat removal through the affected loop, as well A.

as reverse flow since the other loop is still in service Incorrect. Plausible candidate could think SG level would swell due to no heat B. removal from the loop or that feedwater will cause a rise. Delta T response is correct Incorrect. Plausible because candidate could think SG level would swell due to no heat C.

removal from the loop and think delta T could rise due to setting up natural circulation.

56 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination In this case, natural circulation would not occur because the other loop is in service Incorrect. SG level response is correct. Plausible candidate could think no heat removal from the loop could cause a higher delta T, or consider natural circulation D.

setting up

. LP ROC01S Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None ROC01S,Obj Learning Objective: (As available)

Question Source: Bank # WTSI64678 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Vogtle 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pmssure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

57 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 011 K5.12 Importance Rating 2.7 Knowledge of the operational implications of the following concepts as they apply to the PZR LCS Criteria and purpose of PZR level program Proposed Question: RO Question # 29 Which ONE of the following describes the reason for the Pressurizer Level program, and the benefit of using a programmed level?

Maintains a constant mass in the RCS throughout the full range of power, and ensures A. pressurizer heaters will remain covered on a reactor trip.

Maintains a constant mass in the RCS throughout the full range of power, and ensures B. that water will not reach the PRZR PORVs on a turbine trip without reactor trip.

Maintains a constant volume in the RCS throughout the full range of power, and C. ensures pressurizer heaters will remain covered on a reactor trip.

Maintains a constant volume in the RCS throughout the full range of power, and D. ensures that water will not reach the PRZR PORVs on a turbine trip without reactor trip.

Proposed Answer: A Explanation (Optional):

CORRECT. Pressurizer level is programmed from 20%-56% to maintain a constant RCS mass over the automatic control range of Tavg (547-574"F). There is an internal lower limit setpoint in the control circuit related to no load Tavg. The purposes of the A.

program are to ensure heaters remain covered following a trip, and to minimize size requirements of CVCS Incorrect. Plausible because pressurizer level will rise on an ATWS caused by load B. rejection, but this is basis for High PRZR level trip.

Incorrect. Plausible because the difference between between mass and volume can be C. easily confused and second part is correct.

Incorrect. Plausible because the difference between between mass and volume can be D.

easily confused and second part is correct for basis for High PRZR level reactor trip.

58 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Technical Reference(s): LP 1901C, Page 23 of 77 [c.1)] (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1901 C, 1.01 Learning Objective: (As available)

Question Source: Bank# 56138 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pn~ssure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

59 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 015 A4.02 Importance Rating 3.9 Ability to manually operate and/or monitor in the control room: NIS indicators Proposed Question: RO Question # 30 Plant conditions:

The reactor tripped from 100% power due to a reactor coolant pump breaker failure.

The crew is performing ES-0.1, Reactor Trip Response.

Intermediate Range Nuclear Instrumentation Channel N-35 has stabilized at 8 E-9 amps

Intermediate Range Nuclear Instrumentation Channel N-36 is showing a normal post trip response and is approaching 2 E-10 amps Which of the following describes how the Source Range Nuclear Instrumentation Channels (N31 and N32) will reinstate?

N31 and N32 must be manually reinstated by the HCO.

A.

N31 and N32 will automatically reinstate on interlock from N-35.

B.

N31 and N32 will automatically reinstate on interlock from P-10.

C.

N31 will automatically reinstate on interlock from N35; N32 must be manually reinstated D. by the HCO.

Proposed Answer: A Explanation (Optional):

CORRECT. Automatic reinstatement requires both IRNIS channels to be responding A. properly (less than 5 E-11 amps) since the logic for automatic reinstatement is 2/2.

Plausible if applicant believes the reinstatement logic is 1/2; the common logic for 2 B. channel system trips.

60 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible if applicant does not understand or properly apply the unblock reinstatement permissive provided by P-10. P-10 only allows manual or automatic reinstatement of C.

the SRNIS.

Plausible if applicant believes the reinstatement is via "channel matching" (N31/N35 and D. N32/N36).

IR f () R3301C, Page 17 of 86 (Item 5) (Attach if not previously provided)

Technlca e erence s :

Proposed References to be provided to applicants during examination: None R3301C, 1.07 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Ques~ion History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Higher order because the applicant must apply the off-normal trip response to NIS/RPS design and determine that manual action is required to ensure proper NIS indication.

Meets KIA by requiring knowledge that manual action is required to ensure proper NIS indication following a plant trip.

61 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group ~~ 2 KJA# 016 A3.01 Importance Rating 2.9 Ability to monitor automatic operation of the NNIS, including: Automatic selection of NNIS inputs to control systems Proposed Question: RO Question # 31 Plant conditions:

100% power

Multiple alarms have actuated

The plant is in a stable condition

Turbine First Stage Pressure Channel PT-485 is reading 635 PSIG

Turbine First Stage Pressure Channel PT-486 is reading 200 PSIG Which choice fills in the blank in the following statement?

Both feedwater control valves remain in AUTO with _ _ _ providing the signal to the steam generator level control setpoint calculator.

PT-485 A.

Average STEAM FLOW B.

Total FEEDWATER FLOW C.

Median select STEAM PRESSURE D.

Proposed Answer: A Explanation (Optional):

CORRECT. PT-485 is the first choice as the ARBITRATE SELECT SIGNAL and within A. the acceptable range for the stated power.

Plausible because Average Steam Flow provides the ARBITRATE SIGNAL for 1st B. Stage Pressure and is the second choice as the ARBITRATE SELECT SIGNAL.

62 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because Total Feedwater Flow provides the power signal for the selection C. between the HIGH and LOW Power mode of operation in ADFWCS.

Plausible because Avg Steam Pressure is used as an input signal in the ADFWCS and, D. much like 1st Stage Pressure, varies predictably from 0-100% power.

. LP R4401 C Pg. 35 (6)

Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None LP R4401C 1.07.a.2 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Meets KIA by requiring knowledge of the automatic selection of an NNIS input to a control system when an instrument failure occurs.

Identified as MEMORY level but borderline comprehension since applicant must know PT-485 is within the validation range to answer the question correctly.

63 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KJA# 001 K2.05

~ .. -~ ....... ------

Importance Rating 3.1 Knowledge of bus power supplies to the following: MIG sets .....

Proposed Question: RO Question # 32 Plant conditions:

100% power with all components in a normal alignment

A tagging operation is underway for a component fed by Bus 13 Which ONE of the following describes plant response if the AO performing the tagging mistakenly opens the Rod Drive MG Set Supply Breaker on Bus 13?

MG Set "A" trips and the reactor trips.

A MG Set "A" trips and the plant continues to operate at 100% power.

B.

MG Set "B" trips and the reactor trips.

C.

MG Set "B" trips and the plant continues to operate at 100% power.

D.

Proposed Answer: B Explanation (Optional):

Plausible because the first part is correct; Bus 13 feeds MG "An supply breaker.

A However, the MG Sets are in parallel feeding a series circuit so no trip occurs.

CORRECT. Bus 13 feeds MG "A" supply breaker. With MG "B" operating in parallel B. and feeding a series circuit no trip will occur.

Plausible if applicant does not know which bus (13 or 15) feeds which MG Set and does C. not apply the parallel circuit feeding a series circuit design.

Plausible if applicant does not know which bus (13 or 15) feeds which MG Set. The D. second part is correct.

64 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

IR f () LP 3001C, Page 21/65, Item 2.a (Attach if not previously provided)

Techmca e erence s :

Proposed References to be provided to applicants during examination: None R3001 C, 1.05 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # 45562 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

Modified to match GINNA plant design and changed format to four different combinations of two consequences.

Memory because of power supply recall and could answer by applying recall of E-O, REACTOR TRIP OR SAFETY INJECTION, immediate actions (RNO).

Meets KJA by requiring knowledge of power supply to one Rod Drive MG set and, for operational validity, how the circuit is designed.

65 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 002 K6.04 Importance Rating 2.5 Knowledge of the effect or a loss or malfunction on the following RCS components: RCS vent valves Proposed Question: RO Question # 33 Plant conditions:

A loss of secondary heat sink has occurre!d

RCS Bleed and Feed was directed but neither Pressurizer PORV could be opened

As a consequence, FR-H.1 has specified opening the Reactor Head Vent Valves (SOV 590, SOV-591, SOV-592, SOV-593)

Which ONE of the following identifies both the minimum valve operation required to establish at least one Reactor Head Vent flowpath and to where is that flow directed?

Any single valve; Containment A.

Any single valve; Pressurizer Relief Tank B.

SOV-590 and SOV-592 OR SOV-591 and SOV-593; Containment C.

SOV-590 and SOV-592 OR SOV-591 and SOV-593; Pressurizer Relief Tank D.

Proposed Answer: C Explanation (Optional):

Plausible because the second part is COrrE~Ct. Applicant must know the head vent piping design (two sets of valves in series in a parallel flowpath) since the procedure provides A.

no specific guidance.

Plausible because the applicant must know the head vent piping design (two sets of valves in series in a parallel flowpath) since the procedure provides no specific B.

guidance and the PRT is in the normal bleed (PRZR PORV) flowpath.

CORRECT. The head vent piping design (two sets of valves in series in a parallel C.

ftowpath) requires that two valves in the same line must open. The reactor head vents 66 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination terminate in the refueling cavity.

Plausible because the first part is correct ,and the PRT is in the normal bleed (PRZR D. PORV) flowpath.

LP R1001C Technical Reference(s): FR-H.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1001C, 1.08 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Mechanical components and design features of reactor primary system.

Comments:

Meets KIA by requiring knowledge of the minimum system alignment required to perform one of the intended functions of the reactor head vent system.

67 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KJA# 071 K4.01 Importance Rating 2.6 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Pressure capability of the waste gas decay tank Proposed Question: RO Question # 34 Which one of the choices identifies the pressure protection component installed on each Waste Gas Decay Tank discharge line that has the LOWER setpoint and where it discharges to?

Relief Valve; Plant Vent A.

Relief Valve; Vent Header B.

Rupture Disk; Plant Vent C.

Rupture Disk; Vent Header D.

Proposed Answer: D Explanation (Optional):

Plausible as one of two pressure protection components for each Waste Gas Decay Tank and it does relieve to the Plant Vent. However, at 150 PSIG it is set higher than A.

the Rupture Disk (142 PSIG).

Plausible as one of two pressure protection components for each Waste Gas Decay Tank. However, at 150 PSIG it is set higher than the Rupture Disk (142 PSIG) and it is B.

the Rupture Disk that relieves to the Vent Header.

Plausible because the first part is correct: Rupture Disk @ 142 PSIG and Relief Valve C. @ 150 PSIG. However, the relief valve lifts to the Plant Vent not the Rupture Disk.

CORRECT. The Rupture Disk fails @ 142 PSIG and Relief Valve lifts @ 150 PSIG.

The Rupture Disk relieves pressure back to the Vent Header. The Relief Valve relieves D.

pressure back to the Plant Vent.

Technical Reference(s): LP 3801 C, Pages 22 (bottom) and (Attach if not previously provided) 68 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination 23 (top) of 31 Proposed References to be provided to applicants during examination: None R3801C Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

Recall of relative value of setpoints. Stayed awalf from recall of setpoints since all system operations are performed in the field or at a local panel, by AO's.

Meets KIA by requiring knowledge of pressure protection component setpoints and their flowpaths.

69 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KJA# 072 K3.02 Importance Rating 3.1 Knowledge of the effect that a loss or malfunction of the ARM system will have on the following: Fuel handling operations Proposed Question: RO Question # 35 Plant conditions:

A refueling outage has started

Core off-load is in progress

E-24, RMS Area Monitor High Activity, ha:s actuated

Operators have determined that the FAIL ALARM has actuated on Radiation Monitor R 2, Containment Area Monitor

Parts to repair R-2 are estimated to arrive on site in approximately 3 days Which of the choices completes the following statement?

With R-2 failed, core off-load - - - -

must be stopped until R-2 is restored to operability.

A.

must be stopped but may resume when a local monitor is installed on the manipulator B. bridge.

may continue if R-7, In-Core Detectors Area Monitor, is operable.

C.

may continue if either R-29 or R-30, Containment High Range Monitor, is operable.

D.

Proposed Answer: B Explanation (Optional):

Plausible because it is partially correct. However, the procedure allows a local monitor A. on the manipulator bridge to be substituted.

CORRECT. 0-15.1 specifies either R-2 or a local monitor on the manipulator bridge.

B.

70 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because a substitute monitor is permitted. R-7 is a monitor inside C. containment but is NOT the designated substitute.

Plausible because a substitute monitor is permitted. R-29 and R-30 are monitors inside D. containment but are NOT the designated substitute.

Technical Reference(s): 0-15.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3701C,1.09 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Q~estion History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 Radiological safety principles and procedures.

Comments:

71 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 041 A2.02 Importance Rating 3.6 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:

Steam valve stuck open Proposed Question: RO Question # 36 Plant conditions:

Reactor power is stable at 1% power whille reactor engineers gather data during a reactor startup

One previously closed condenser steam dump valve has failed full open

SG pressure is reduced to 965 psig prior to reclosing the valve

RCS temperature indicates 541°F on Loop 'A' and 541.5°F on Loop 'B' Which ONE of the following describes the status of RCS temperature and the action that will be taken to restore plant conditions?

RCS temperature is ...

above the technical specification minimum temperature for criticality; RCS temperature A. will be restored by valve closure and/or manual rod withdrawal.

above the technical specification minimum temperature for criticality; RCS temperature B. will be restored by automatic rod withdrawal.

below the technical specification minimum temperature for criticality; RCS temperature C. will be restored by valve closure and/or manual rod withdrawal.

below the technical specification minimum temperature for criticality; RCS temperature D. will be restored by automatic rod withdrawal.

Proposed Answer: A Explanation (Optional):

A. CORRECT. Rod Control is in MAMUAL at less than 15% power. Also, Minimum 72 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination temperature for criticality is 540 degrees F.

Incorrect. Plausible because the first part is correct. However, automatic rod control is B. not available until reactor power is above 15%.

Incorrect. Plausible because the action is correct but the temperature is incorrect.

Applicant can confuse temperatures because no load is 547 and steam dumps close on C. low temperature with reactor trip breakers open to ensure inadvertent cooldown does not occur.

Incorrect. Plausible as described in Options above.

D.

0-1.2 Technical Reference(s): TS 3.4.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R4501 C, 1.06 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KJA by requiring knowledge of how the rest of the steam dump system and the integrated plant responds to a steam dump valve failure at low power.

Higher order because the applicant must apply ht9at transfer, reactivity balance and steam dump system design concepts to answer the question.

73 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination 74 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 055 K1.06 Importance Rating 2.6 Knowledge of the physical connections and/or cause effect relationships between the CARS and the following systems: PRM system Proposed Question: RO Question # 37 Which ONE of the following radiation monitors' alarm setpoints are set to ensure that Technical Specification limits for steam generator tube leakage are not exceeded?

R-31/R-32, Main Steam Line A.

R-19, Steam Generator Blowdown B.

R-47, Air Ejector/Exhaust C.

R-48, Air Ejector/Exhaust D.

Proposed Answer: C Explanation (Optional):

Incorrect. R31 and R32 are good indicators of which SG is ruptured as a backup to A. other indications of an SGTR Incorrect. R19 activity is helpful in determining trend of leak rate but is not used or B. calibrated for TS leakage limits Correct.

C.

Incorrect. R48 is used as a high range detector that is helpful in establishing EAL D. classifications R3901C Technical Reference(s): P-9 (Attach if not previously provided) 75 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None R4301 C, 1.06.b Learning Objective: R3901 C, 1.0B.a (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

Also meets 10CFR55.41(b) item 4 76 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KJA# 086 A1.04

~-- ....---

Importance Rating 2.7

- -... ~~

Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including: Fire dampers Proposed Question: RO Question # 38 Plant conditions:

The unit is in Mode 5, heating up to Mode 4

An AO on rounds has reported that the linkage from the damper operator to Fire Damper BB-46-P in Battery Room B appe,ars to be de-formed

The damper has been declared INOPERABLE pending further evaluation and Work Control has contacted maintenance for assistance Which ONE of the following meets Technical SpecificationlTRM requirements for this condition?

Fire Damper BB-46-P - - - -

must be placed in the closed position within one hour and then "B" Battery must be A. declared INOPERABLE.

must be placed in the closed position within one hour. The unit must not enter Mode 4 B. until all fire barrier penetration seals are OPERABLE.

can remain open but a continuous fire watch must be established on one side of the C. damper within one hour.

can remain open. Verify all Battery Room "B" fire detectors OPERABLE within one hour D. then perform a once-per-shift fire watch inspection of "B" Battery Room.

Proposed Answer: C Explanation (Optional):

Plausible because it is typical for failed components to be placed in the safety-response position. In this case, if the damper is closed then Battery Ventilation and therefore A.

operability must be evaluated. However, the damper is NOT required to be closed.

77 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because it is typical for failed components to be placed in the safety-response position and, generally, mode changes are not permitted with reliance on action B. statements. However, fire barrier penetration seals are required to be operable in all modes (not just Mode 4 and higher) and the damper is NOT required to be closed.

CORRECT. The damper is NOT required to be placed in the safety response position.

C. One hour action per TRM TR 3.7.5, Condition A.

Plausible because the first part is correct and the second part is partially correct.

D. However, a once-per-hour fire watch inspection is required.

Techmca* I Ref erence ()

s : TRM TR 3.7.5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None LP R5901C, 1.12 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # x (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KIA indirectly. Knowledge of compensatory action(s) required because the position of the fire damper will NOT be changed in order that battery ventilation be maintained. Licensed operators have limited responsibility for recall of specific fire damper operations but are required to recall one hour actions from TSfTRM.

Bank Question (attached) used as a basis for this question in order to be "in the ballpark" of expected operator knowledge for this system.

78 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination 79 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KJA# 009 EK1.02 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables Proposed Question: RO Question # 39 Plant conditions:

A LOCA has occurred.

A manual Rx trip and Safety Injection were initiated

RCS pressure is 1085 psig.

CETs indicate 525°F.

Containment pressure is 7 psig.

Transition to ES-1.1, SI Termination, is being evaluated by the CRS.

The CRS directs the HCO to determine RCS subcooling using Figure 1.0, Figure Min Subcooling.

Which ONE of the following identifies current RCS subcooling in accordance with Figure 1.0, and whether all conditions are met to transition to ES-1.1?

(Reference Provided)

RCS subcooling is OaF; transition will be made A.

RCS subcooling is 10°F; transition will be made B.

RCS subcooling is OaF; transition will NOT be made C.

RCS subcooling is 10°F; transition will NOT be made D.

Proposed Answer: D Explanation (Optional):

Incorrect. Subcooling will indicate higher than zero. The applicant may be off by a gradient on the figure. Transition will not be made with zero subcooling, but transition is A.

plausible if applicant doesn't consider that greater than zero required.

80 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because if applicant uses normal or adverse containment values, B. they will believe transition may be made.

Incorrect. Subcooling will indicate higher than zero. The applicant may be off by a gradient on the figure. Plausibility enhanced because if the applicant arrives at zero and C.

correctly applies the transition, this option would be chosen Correct. Plausible because the value is correct, but if value is greater than zero, the transition would be made based on subcooling alone, but will NOT be made under this D.

condition where RCS pressure is too low.

E-1 Technical Reference(s): Figure 1.0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: Figure 1.0 REPOOC, Obj 2.01 Learning Objective: (As available)

Question Source: Bank # WTSI64256 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Salem 2006 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

81 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group #

KIA # 007 EK1.04 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decrease in reactor power following reactor trip (prompt drop and subsequent decay)

Proposed Question: RO Question # 40 Which ONE of the following describes Nuclear Instrumentation response to a reactor trip from 100% power until the Source Range instruments energize, from the time control rods begin to drop?

Prompt Drop of approximately 3 decades, followed by a -1/3 DPM startup rate for A. approximately 20 minutes.

Prompt Drop of approximately 3 decades, followed by a -1/3 DPM startup rate for

8. approximately 3-4 hours.

Prompt Drop to approximately 5% power, followed by a -1/3 DPM startup rate for C. approximately 3-4 hours.

Prompt Drop to approximately 5% power, followed by a -1/3 DPM startup rate for D. approximately 20 minutes.

Proposed Answer: D Explanation (Optional):

Incorrect. Plausible because the applicant may confuse the magnitude of the prompt drop (3 decades would make power approximately 0.1 %, which nearly 0%, similar to A.

5% being near 0%) Additionally, the -1/3 DPM SUR for 20 minutes is correct.

Incorrect. Plausible same reason as option A, and also because power does show a B. noticeable decrease for 3-4 hours, but the rate will be lower after the first 20 minutes.

Incorrect. Plausible because the prompt drop is correct, and also for same reason as C. option B for time that SUR is -1/3 DPM.

Correct. Decay heat is approximately 5-7%, and when the reactor trips, that will be the D. immediate indication. Short Lived delayed neutron precursors will cause the Startup Rate to be -1/3 DPM for approximately 20 minutes until they have decayed sufficiently, 82 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination the SUR lowers for about 3-4 hours until power is stable in Source Range.

. RRT04C Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank# WTSI 52812 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Harris 2004 NRC Question Cognitive Level: Memory or FundamEmtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 1 55.43 Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Comments:

83 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group '/l! 1 KJA# E05 EK1.1 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) Components, capacity, and function of emergency systems.

Proposed Question: RO Question # 41 Plant conditions:

A loss of Feedwater has resulted in a reactor trip.

The crew is performing actions of FR-H.1, Loss of Secondary Heat Sink.

Which ONE of the following describes the operation of the PRZR PORVs during this event?

May be operated to depressurize the RCS so that Low Steam Pressure and Low PRZR Pressure SI signals can be blocked; this prevents unwanted SI actuation when A.

depressurizing SGs prior to performing action to establish Main Feedwater flow.

May be operated to depressurize the RCS so that Low Steam Pressure and Low PRZR Pressure SI signals can be blocked; this prevents unwanted SI actuation when B.

depressurizing SGs prior to performing action to establish Condensate flow.

Remain closed with block valves closed to preserve RCS inventory unless conditions C. exist that require initiation of Bleed and Feed cooling of the RCS.

Remain in automatic and allowed to operate with block valves open for PRZR pressure control. One PORV is opened if Bleed and Feed cooling is required, with the other D.

PORV remaining available for automatic control.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because the phrasing is similar, but no depressurization is required A. in order to initiate Feedwater flow Correct. Aux Spray is the preferred method here, but one PORV may be used if Aux B. Spray is not available (Step 10) 84 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because opening a PORV can possible lead to loss of inventory.

C. The only way block valves will be closed is if a PORV is failed or leaking.

Incorrect. Plausible because this is simila,r to the use of PORVs in FR-C.1 and FR-C.2.

D. The second part of this option is correct.

Technical Reference(s): FR-H.1, Rev 38, Step 10 RNO (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRH1C, Obj 1.04 Learning Objective: (As available)

Question Source: Bank #

Mod ified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

85 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group '# 1 KJA# E11 EK2.2 Importance Rating 3.9 Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, and the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Proposed Question: RO Question # 42 Which ONE of the following describes the reason and requirements for the INITIAL depressurization of the RCS per ECA-1.1, Loss of Emergency Coolant Recirculation?

To ensure SI Accumulator injection, depressurize the RCS maintaining minimum RCS A. subcooling, then terminate SI pump flow.

To minimize RCS leakage, depressurize the RCS maintaining minimum RCS B. subcooling, then reduce SI pump flow.

To minimize RCS leakage, depressurize the RCS maintaining maximum RCS subcooling, then stabilize RCS temperature while attempting to restore makeup C.

sources.

To ensure SI Accumulator injection, depressurize the RCS maintaining maximum RCS subcooling, then stabilize RCS temperature while attempting to restore makeup D.

sources.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because setup for accumulator injection is performed later in procedure, and depressurizing will ensure injection at some point, but SI is not secured, A.

only minimized Correct. The depressurization is performed to decrease leakage, therefore B. decreasing RCS makeup requirements.

Incorrect. Maximizing subcooling will delay depressurization, and will not stabilize RCS C. temperature. Plausible because normally RCS subcooling is maximized 86 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because setup for accumulator injection is performed later in the D. procedure after SG depressurization, but the crew will not stabilize RCS temperature.

ECA-1.1, ECA-1.1 Background Technical Reference(s): Doc (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REC11C, Obj 1.01 Learning Objective: (As available)

Question Source: Bank # WTSI64936 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Wolf Creek 2009 NRC Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

87 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group~! 1 KJA# E04 EK2.1 Importance Rating 3.5 _.

Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Proposed Question: RO Question # 43 Plant conditions:

The plant was initially at 100% power.

A Reactor Trip and Safety Injection have occurred.

RCS pressure is 1200 psig.

R-14, Plant Effluent Gas Monitor, is in alarm.

Area Radiation Monitor Readings in the Auxiliary Building are increasing.

The Crew has transitioned from E-O to ECA-1.2, LOCA Outside Containment.

Why does ECA-1.2 direct the closing RHR Discharge to Reactor Vessel Deluge valve, MOV 852A?

To assist in establishing a Hot Leg Injection flow path A

To determine if the leak is in the RHR discharge piping B.

To prevent RWST inventory from being lost to the environment if the break is in the C. RHR discharge piping To maintain RHR discharge piping isolated when RHR pumps are stopped because D. they are not required due to RCS pressure Proposed Answer: B Explanation (Optional):

Incorrect. ECA-1.2 does not align hot leg injection. MOV-852A would be closed for hot leg injection but not in this procedure. Plausible because it isolates cold leg injection A

from RHR B. Correct. Purpose is to locate and isolate the leak 88 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because there is a leak and RHR takes suction from RWST, but in C. this condition, RCS is leaking, not RWST.

Incorrect. RHR would only remain isolated if closing the valve stops the leak. Plausible because RCS pressure is above RHR pump shutoff head, and they will be shut down in D.

E-1, but the valve is not closed due to securing RHR pumps Technical Reference(s}: ECA-1.2, Step 3, Background (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REC12C, Obj 1.02 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 19024 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Robinson 2002 NRC, modified by adding condition to stem so that distracter options could be made more plausible 89 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 077 AK2.07 Importance Rating 3.6 Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Turbine 1 generator control Proposed Question: RO Question # 44 Plant conditions:

The plant is operating at 100% power.

Severe weather is in the area.

Grid instabilities are being reported.

Circuit 911 is lost due to a lightning strike.

RG&E ECC is evaluating conditions for potential power restrictions on the grid.

Which ONE of the following describes the action that will be required in accordance with 0-6.9, Ginna Station Operating Limits for 13A Transmission?

Refer to the applicable attachment to determine operability of off-site power sources A. and adjust Main Generator Excitation if required.

Start and load both EDGs due to inoperability of 480 volt Safeguards Buses.

B.

Reduce Main Generator Output in accordance with RG&E instructions if ECC C. determines that net generation is too high.

Trip the reactor; enter E-O, Reactor Trip or Safety Injection.

D.

Proposed Answer: C Explanation (Optional):

Incorrect. Plausible because this would be performed if a Low Voltage Contingency A. alarm was received, and is covered by the same procedure Incorrect. Plausible because this would bE~ performed if it was determined that off-site B. power was inoperable and affected the safeguards bus operability.

C. Correct. RG&E ECC may call to require a load reduction when an off-site line is lost 90 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination because net generation may be too high for degraded grid capacity Incorrect. The reactor would only be tripped if the facility cannot reduce net generation D. within the time period committed to.

Technical Reference(s): 0-6.9 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R4101C, 1.11 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

91 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group t~ 1 KIA # 022 AK3.03 Importance Rating 3.1

-~- ..... -- ~ ....--

Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Pump Makeup: Performance of lineup to establish excess letdown after determining need Proposed Question: RO Question # 45 Plant conditions:

The plant is at 100% power when a small RCS leak develops.

The crew takes the necessary actions to restore and stabilize PRZR level.

Subsequently, the crew attempts to determine the leakage source in accordance with the appropriate AP.

After isolating letdown, the crew observes the! following parameters/indications:

Letdown flow 0 gpm

i Containment pressure approximately 0.8 psigl and lowering

RCS pressure approximately 2235 psig and stable

PRZR level approximately 72% and rising

Charging flow 40 gpm

Containment radiation monitors trending downward slowly Which ONE of the following describes the required actions in accordance with AP-CVCS.1, CVCS Leak?

Immediately trip the reactor and enter E-O, Reactor Trip or Safety Injection because A. PRZR level exceeds the maximum limit.

Place Excess Letdown in service and adjLlst seal injection flow to maintain PRZR level B. within limits.

Initiate Shutdown per 0-2.1, Normal Shutdown to Hot Shutdown, due to RCS leakage C. exceeding limits.

Re-establish normal Letdown and dispatch leak search team to identify leakage D. source.

Proposed Answer: B 92 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Applicant may mistake 72% PRZR level for the reactor trip value; however, A. the initial conditions do not require an immediate reactor trip.

Correct. The RCS leak was determined to be on normal letdown based on containment pressure decreasing and radiation monitor readings stabilized. Therefore to maintain pressurizer level and RCS chemistry (chemical feed and boration), excess letdown is required.

B.

The applicant must understand from the initial conditions the RCS leak was on the letdown line and has been isolated. In order to maintain pressurizer level and chemistry excess letdown is required.

Incorrect. Since pressurizer level is stabilized and is controllable by placing Excess Letdown in service, the other initial conditions indicated the RCS leak has been isolated C.

a shutdown is not required.

Incorrect. The leak has been isolated to the normal letdown line. Therefore, a leak D. search team is unnecessary.

Technical Reference(s): AP-CVCS.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP05C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank # WTSI64349 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Sequoyah 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 10 93 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

94 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 026 AK3.03 Importance Rating 4.0 Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: Guidance actions contained in EOP for Loss of CCW/nuciear service water Proposed Question: RO Question # 46 Plant conditions:

The plant is in Mode 4.

A loss of CCW has occurred.

The crew is performing AP-CCW.3, Loss of CCW- Plant Shutdown.

The HCO is ensuring that seal injection flow is maintained to RCPs.

A cooldown is being initiated as the crew attempts to restore CCW.

Which ONE of the following describes the requirement for maintaining seal injection in this procedure, and the reason for the requirement?

Maintain seal injection to RCPs until RCS temperature reaches ...

150°F; To ensure a high pressure source of water to the seal area to keep CRUD from A. the RCS from fouling the seal package.

150°F; To ensure RCP seal cooling is maintained.

B.

250°F; To ensure a high pressure source of water to the seal area to keep CRUD from C. the RCS from fouling the seal package.

250°F; To ensure RCP seal cooling is maintained.

D.

Proposed Answer: B Explanation (Optional):

Incorrect. This is true for an idle RCP in cold shutdown or hot standby but for a loss of CCW, the concern for fouling a seal is subordinate to seal cooling. Plausible because it A.

is an RCP action and the temperature is correct.

B. Correct. Procedure recommends ensuringl seal injection is maintained until RCS is 95 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination below 150.

Incorrect. This is true for an idle RCP in cold shutdown or hot standby but for a loss of CCW, the concern for fouling a seal is subordinate to seal cooling. Plausible because it C.

is an RCP action.

Incorrect. Westinghouse recommends that some seal cooling be maintained above D. 150°F. If seal injection was terminated at 250, then the RCP could be damaged.

AP-CCW.2 Note and Background Technical Reference(s): Information (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP03C 1.03 Learning Objective: (As available)

Question Source: Bank # WTSI66509 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Ginna 2006 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or A.nalysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

96 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 040 AK3.02 Importance Rating 4.4

~-- .. ~- - -....... --

Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture: ESFAS initiation Proposed Question: RO Question # 47 Which ONE of the following describes the reason for automatic Safety Injection actuation during a Main Steam Line Break (MSLB)?

Maintains RCS pressure to prevent loss of subcooling and reactor vessel head voiding A. during the RCS cooldown.

Minimizes the probability of a Pressurized Thermal Shock event by limiting the pressure B. transient on the RCS.

Ensures that a Containment Isolation signal will be generated in the event of a MSLB C. inside Containment.

Ensures borated water is added to the ReS to offset the positive reactivity added during D. the RCS cooldown.

Proposed Answer: D Explanation (Optional):

Incorrect. Plausible because head voiding is a concern during the performance of some EOPs. In this case, subcooling will be high as the RCS cools down and the RCS will A.

shrink. But, because it a MSLB, inventory is not lost and head voiding is unlikely.

Incorrect. SI increases chance of PTS because once the faulted SG is fully depressurized, pressure will be high, and heatup will squeeze the RCS causing B.

pressure to rise further with SI inventory already in the RCS Incorrect. Plausible because it is a true statement by itself, but is not the reason that C. Safety Injection is actuated during a Main Steam Line Break Correct. SI actuates to ensure boron counteracts the positive reactivity from a severe D. cooldown of the ReS 97 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

I Reference ()

Tech!llca s : ITS 3.5.2 Basis (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R2701C 1.13 Learning Objective: (As available)

Question Source: Bank # WTSI19018 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Robinson 2002 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

MOdified distracters for plausibility 98 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KJA# 056 AA1.24 Importance Rating 2.9

--'~'--

Ability to operate and I or monitor the following as they apply to the Loss of Offsite Power:

Plant computer, to call up in-core temperature monitoring group Proposed Question: RO Question # 48 Plant conditions:

Reactor trip followed by a loss of off-site power.

The crew is performing the actions of ES-0.2, Natural Circulation Cooldown.

The HCO is monitoring the RCS cooldown.

Which ONE of the following describes the use of the PPCS to monitor the RCS cooldown in accordance with A-503.1, Emergency and Abnormal Operating Procedures User's Guide?

PPCS is the required means of plotting thle cooldown using the Core Exit A. Thermocouples page PPCS is the primary means of plotting the cooldown using Th and T cold trending using B. the RCS display MCB indication is the primary source of information during EOP use, but PPCS may be C. used to monitor Core Exit Thermocouples to monitor RCS subcooling as desired MCB indication is the primary source of information during EOP use. Information provided by PPCS may NOT be acted upon UNLESS MCB indicators are consistent D.

with that information Proposed Answer: C Explanation (Optional):

Incorrect. Not required, only used as backup indication, but PPCS may be used, it is A. just not required to be used Incorrect. RCS display doesn't trend, and PPCS not primary indicator in accordance B. with A503.1 99 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Correct.

C.

Incorrect. There is SPDS info that can only be gotten from PPCS, and is acted upon D. during performance of EOPs

I Ref erence()

Technlca s : A-503.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3201 C, Obj 1.04 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signa;ls, interlocks, failure modes, and automatic and manual features.

Comments:

100 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group ~! 1 KJA# 015 AA1.21 Importance Rating 4.4 Ability to operate and 1 or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Development of natural circulation flow Proposed Question: RO Question # 49 Plant conditions:

The plant was operating at 100% power when a reactor trip and subsequent loss of off-site power occurred.

Which ONE of the following describes the effect on RCS temperature immediately after the RCPs trip as natural circulation is developing?

That rises until Natural Circulation develops, then lowers as decay heat load lowers.

A. Loop Delta T remains constant throughout the event.

Tcold lowers until Natural Circulation develops, then stabilizes as decay heat load B. lowers. Loop Delta T remains constant throughout the event.

That rises until Natural Circulation develops, then lowers as decay heat load lowers.

C. Loop Delta T lowers as decay heat load lowers.

Tcold lowers until Natural Circulation develops, then stabilizes as decay heat load D. lowers. Loop Delta T lowers as decay heat load lowers.

Proposed Answer: C Explanation (Optional):

Incorrect. First part is correct. Second part is plausible if the applicant misunderstands A. T cold behavior. T cold will track SG pressure Incorrect. Tcold actually rises immediately as the SG pressure rises due to Turbine stop valves closing and reduced steam demand against the initial reactor decay heat.

B.

Description of Loop Delta T same as in option A Correct. When the reactor trips and RCPs are lost, there is less flow through the core to C.

remove decay heat. As natural circulation sets up, That becomes hotter because the 101 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination water spends more time in the core with a low flow rate. Natural circulation develops due to density difference between Thot and Tcold, and as decay heat lowers and natural circulation flow rises, Thot lowers because there is less decay heat generated and the water spends less time in the corEl.

Incorrect. Same description as Option B, except that loop Delta T description is correct D.

Westinghouse Executive Volume Technical Reference(s): Natural Circulation (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 14 55.43 Principles of heat transfer, thermodynamics and fluid mechanics.

Comments:

102 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 KIA # 055 EA1.01 Importance Rating 3.7 Ability to operate and monitor the following as they apply to a Station Blackout: In-core thermocouple temperatures Proposed Question: RO Question # 50 Plant conditions:

A loss of off-site power has occurred.

Both EDGs tripped upon starting.

The crew is performing actions of ECA-O.O, Loss of All AC Power.

RCS Tavg is 555°F and rising.

RCS pressure is 2120 psig and lowering slowly.

Core Exit Thermocouple temperatures are 612°F and rising.

SG Narrow Range levels indicate 10%.

TDAFW flow is 220 GPM.

Which ONE of the following describes the action required to stabilize or reverse the trend of Core Exit Thermocouple temperature?

Start available Control Rod Shroud Fans for cooling of the vessel head.

A.

Energize pressurizer heaters to raise RCS subcooling.

B.

Lower Atmospheric Relief Valve controller setpoint to raise SG steaming rate.

C.

Lower the Main Condenser Steam Dump pressure setpoint to raise SG steaming rate.

D.

Proposed Answer: C Explanation (Optional):

Incorrect. Plausible because this action is on the list of actions to enhance Natural A. Circulation on Att 13. However, there is no power available to start the fan Incorrect. Plausible because this action is on the list of actions to enhance Natural B. Circulation on Att 13. However, there is no power available to energize heaters 103 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct.

C.

Incorrect. With Circ Water Pumps tripped, Main Condenser will not be available. The crew will have to use SG ARVs to steam the SGs. Plausible because this is the normal D.

action for enhancing Natural Circulation

I Ref erence ()

Techmca s : ECA-O.O (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RECOOC, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

104 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1

~~"".----- ~--------

Group # 1 KIA # 057 AA2.17 Importance Rating 3.1 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: System and component status, using local or remote controls Proposed Question: RO Question # 51 Plant conditions:

The plant is operating at 100% power.

The following alarms are received in the control room:

E-3, INVERTER TROUBLE

E-6, LOSS A INSTR BUS

Instrument Bus 'A' voltage is zero volts.

Which ONE of the following describes the effect on the plant?

All RED (Channel 1) bistable lights on the Status Panel are LIT; Rod Control remains A. available in AUTO All RED (Channel 1) bistable lights on the Status Panel are LIT; Rod Control must be B. placed in MANUAL All RED (Channel 1) and WHITE (Channel 2) bistable lights on the Status Panel are C. OFF; Rod Control remains available in AUTO All RED (Channel 1) and WHITE (Channel 2) bistable lights on the Status Panel are D. OFF; Rod Control must be placed in MANUAL Proposed Answer: B Explanation (Optional):

Incorrect. Rod Control must be placed in manual because the +/- 4°F auto rod stop will A. be actuated. One set of indications supplied by Instrument Bus A will be lost.

105 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. See P-12, section 2.5 B.

Incorrect. Indication is for a loss of Instrument Bus B, which supplies power to the C. bistable lights. Also, manual rod control is required as described in Option A Incorrect. Indication is for a loss of Instrument Bus B, which supplies power to the D. bistable lights.

P-12 Technical Reference(s): ER-INST.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R0901C, Obj 1.11 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

106 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 027 AA2.12 Importance Rating 3.7 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: PZR level Proposed Question: RO Question # 52 Plant conditions:

The plant is operating at 100% power.

The following alarm is received in the control room:

0

F-19, PRZR PORV OUTLET HI TEMP 145

The HCO determines that PORV PCV-431C indicates open.

PRZR pressure indicates 2000 psig and lowering.

The HCO trips the reactor.

The PORV is isolated when PRZR pressure has lowered to 825 psig.

Safety Injection failed to automatically actuatE~, and was manually actuated at the same time that the PORV was closed.

Which ONE of the following describes indicated PRZR level at the time the PORV is isolated, and whether PRZR level is an accurate indication of RCS inventory?

Off-Scale low or lowering rapidly; if level is on scale, it is considered an accurate A. indication of RCS inventory.

Off-Scale low or lowering rapidly; if level is on scale, it is NOT considered an accurate

8. indication of RCS inventory.

Off-Scale high or rising rapidly; if level is on scale, it is considered an accurate C. indication of RCS inventory.

Off-Scale high or rising rapidly; if level on scale, it is NOT considered an accurate D. indication of RCS inventory.

Proposed Answer: 0 107' 5n12010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Pressure has decreased to a value below saturation pressure. Therefore, with RCPs tripped, a bubble will form under the head. Water will be expelled through A. the pressurizer opening, resulting in high indicated level. Water level only lowers prior to reaching saturation in the RCS.

Incorrect. Same discussion as in option A, but second half of the statement is correct...

B.

Incorrect. Level indication is correct, but it will not be an accurate indication, particularly C. if SI was not actuated when required.

Correct. Saturation has been reached under the vessel head. PRZR level will rise as a bubble forms under the head. It is not an accurate indication of RCS inventory because D.

RVLlS will show that there is voiding. If pressure was raised, PRZR level would fall

I Ref erence()

Techmca s : UFSAR 15.6.1.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

108 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 038 EA2.01 Importance Rating 4.1 Ability to determine or interpret the following as they apply to a SGTR: When to isolate one or more S/Gs Proposed Question: RO Question # 53 Plant conditions:

A Steam Generator Tube Rupture is in progress.

The crew is performing actions of E-3, Steam Generator Tube Rupture.

SG 'A' has been identified as ruptured.

SG A Narrow Range level - 7%

SG B Narrow Range level - 3%

In accordance with E-3, which ONE of the following describes the EARLIEST time that AFW flow may be isolated to SG 'A'?

Immediately.

A.

ONLY when both SG Narrow Range levels are above 7%.

B.

Not until 'A' SG narrow range level is above 25%.

C.

ONLY when both SG narrow range levels are above 25%.

D.

Proposed Answer: A Explanation (Optional):

Correct. >7% in the ruptured SG is required prior to isolation of that SG, to ensure SG A. tubes are covered and a thermal layer may be established above the break.

Incorrect. The ruptured SG level must be > 7% to ensure an adequate thermal layer exists prior to isolation of Aux Feed. This will insulate the steam space in the SG from B. the cooler RCS water as the RCS is cooled down to allow depressurization of the RCS to stop the leakage to the SG without losing subcooling, if the SG were to depressurize while trying to equalize RCS to SG pressure. 7% is !\lOT required in a steam generator 109 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination that is unaffected.

Incorrect. Plausible because this value is listed in the procedure as the value that would C. be used if adverse containment conditions existed.

Incorrect. Plausible because both SGs are desired fro symmetrical heat removal on a reactor trip, but to isolate a ruptured SG, only the affected SG must be >7%. The applicant would choose this if they believe that >25% in the unaffected SG is required to D.

maintain minimum heat removal capability. 25% will be familiar, because it is the value required if containment conditions were adverse.

. E-3 Rev 46 Step 5 Techrucal Reference(s): ' , (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REP03C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank # WTSI6Ei819 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

11 CI 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group "# 1 KIA # 011 2.1.23 Importance Rating 4.3 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Proposed Question: RO Question # 54 Plant conditions:

A LOCA has occurred.

The crew is performing actions of E-O, Reactor Trip or Safety Injection.

RCS pressure is 200 psig.

The crew has just verified immediate actions.

ECCS is operating as designed.

Containment pressure is 31 psig and rising.

Annunciator A-27, CNMT SPRAY, is LIT.

BOTH Containment Spray Pumps GREEN breaker indicating lights are LIT.

The HCO reports there is NO Containment Spray flow on either train.

Which ONE of the following describes (1) the MINIMUM action required to manually actuate Containment Spray, and (2) the action required in accordance with E-O for operation of RCPs?

(1) Depress either 1 of 2 Containment Spray actuation pushbuttons A. (2) Leave RCPs running until evaluating status at the appropriate step in E-O (1) Depress either 1 of 2 Containment Spray actuation pushbuttons B. (2) IMMEDIATELY trip both RCPs based on E-O Foldout page requirements (1) Depress BOTH Containment Spray actuation pushbuttons C. (2) Leave RCPs running until evaluating status at the appropriate step in E-O (1) Depress BOTH Containment Spray actuation push buttons D. (2) IMMEDIATELY trip both RCPs based on E-O Foldout page requirements Proposed Answer: D Explanation (Optional):

A. Incorrect. Both pushbuttons must be depressed at the same time. This minimizes the 111 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination probability of an inadvertent spray actuation. Second part is correct.

Incorrect. Both pushbuttons must be depressed at the same time. This minimizes the probability of an inadvertent spray actuatkm. RCP operation is plausible because it is B.

common to trip RCPs as part of spray actuation step due to loss of CCW.

Incorrect. RCP evaluation is a continuous action step.

C.

Correct.

D.

E-O, Rev 43, Steps 5 Technical Reference(s): E-O Foldout Page (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REPOOC, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

112 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 054 2.4.34 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Proposed Question: RO Question # 55 Plant conditions:

The reactor was tripped due to loss of all CCW to 'B' RCP.

CCW CANNOT be restored to 'B' RCP.

Subsequently, a loss of Feedwater occurred.

The crew is performing FR-H.1, Loss of Secondary Heat Sink.

MDAFW Pumps are tripped and CANNOT be restarted.

TDAFW Pump steam supply valves MOV-3504A and 3505A are OPEN.

V-3652, Turb Drvn AFW Pump Governor Vlv, indicates GREEN light lit, RED light extinguished.

, TDAFW flow is 0 gpm

. TDAFW discharge pressure is 0 psig

SG levels are 45 inches in BOTH SGs.

Bleed and Feed has been initiated.

RCS temperature is lowering slowly.

For the plant condition, which ONE of the following describes the local operator actions that are required in accordance with FR-H.1, and if the actions are unsuccessful, the NEXT action that will be required to restore secondary heat sink?

Locally reset TDAFW Pump governor valve ...

ONL Y; if AFW flow cannot be restored, attempt to initiate Main Feedwater flow A.

ONLY; if AFW flow cannot be restored, attempt to initiate Standby Aux Feedwater flow B.

AND close 'B' RCP seal injection needle valve; if AFW flow cannot be restored, attempt C. to initiate Main Feedwater flow AND close 'B' RCP seal injection needle valve; if AFW flow cannot be restored, attempt D. to initiate Standby Aux Feedwater flow 113 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed Answer: D Explanation (Optional):

Incorrect. Plausible because the governor valve will be reset and also because Main A. Feedwater flow is the next action required Incorrect. Plausible because the governor valve will be reset. Standby AFW is not B. initiated at this point in the event. The next action after AFW is Main feedwater Incorrect. Plausible because all actions are correct except initiation of Main Feedwater C. is not the next step.

Incorrect. Plausible because all actions are correct with the exception of initiating SAFW flow next.

D.

Correct.

. FR-H.1 Rev 38 Techntcal Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRH1C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

114 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination 115 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 008 2.1.32 Importance Rating 3.8 Conduct of Operations: Ability to explain and apply all system limits and precautions.

Proposed Question: RO Question # 56 Plant conditions:

Reactor Trip and Safety Injection have actuated due to a LOCA.

The crew is performing actions of ES-1.2, Post LOCA Cooldown and Depressurization.

PRZR level is off-scale high.

RCS subcooling indicates 5°F superheat.

RCS cooldown is in progress.

During the cooldown, an ORANGE condition is indicated on the INTEGRITY Critical Safety Function Status Tree.

The CRS determines that the crew will remain in ES-1.2 until RHR injection is attained, before addressing the Orange condition.

Which ONE of the following describes the reason for the decision to remain in ES-1.2?

Since RCS subcooling is negative, Pressurized Thermal Shock is NOT an immediate A. concern, even if RCS cooldown rates are high.

Once RHR injection is attained, the Orange condition on the Integrity Critical Safety B. Function Status Tree will clear, and the actions of the FR procedure will not be required.

Challenges to the Integrity Critical Safety Function are expected when performing the C. actions of ES-1.2 for a pressurizer vapor space LOCA.

Once an operator controlled cooldown is underway, the likelihood of Pressurized D. Thermal Shock is reasonably remote.

Proposed Answer: A Explanation (Optional):

Correct. Caution prior to step 7 says not to go to FR-P.1 until the cooldown is complete A. to the point of RHR injection 116 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because it sounds reasonable that RHR injection will mean no PTS. However, the CSF Status Tree looks at cooldown rate and if the max rate is B.

exceeded, it is possible to get an orange or red condition Incorrect. RCS cooldown rates are expected to be below the rate required to receive a C. challenge to the Integrity CSFST Incorrect. Plausible because it sounds similar to reason for RCP trip criteria not D. applying once a cooldown is started in E-3 ES-1.2 Background, caution prior Technical Reference(s): to step 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RES12C, Obj. 1.02 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

117 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KJA# 024 AK1.01 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to Emergency Soration: Relationship between boron addition and change in T-ave Proposed Question: RO Question # 57 Plant conditions:

A component failure while RPS testing was in progress resulted in an automatic trip signal on High Nuclear Power

The reactor failed to trip and the crew has. entered FR-S.1, RESPONSE TO REACTOR RESTART/ATWS

The Main Turbine is tripped and both motor-driven AFW Pumps are running

Emergency Soration has been initiated

Tavg is 590 of Assuming NO automatic or manual control rod motion, which ONE of the following describes the trend of Tavg due to the RCS boration?

Tavg will ....

lower continuously until Emergency Soration is stopped.

A.

lower continuously until steam dumps gain control of T avg and then stabilize as steam S. dumps modulate closed.

remain the same until reactor power is less than AFW heat removal capability and then

c. begin to lower.

remain the same until control rods drop, then lower rapidly and stabilize when Steam D. Dump valves are closed.

Proposed Answer: S Explanation (Optional):

A. Plausible because Tavg will begin lowering as the Emergency Soration reaches the 11B 517/201 0 Rev Draft

Ginna 2010 NRC Written Examination reactor. However, it has no effect on Tav9 once the reactor is sub-critical.

CORRECT. When the Emergency Soration reaches the reactor the effect is similar to but not as rapid as the negative reactivity insertion of control rod motion. With Steam Dumps wide open, Tavg will lower until they gain control. At that point the steam dumps S.

will modulate closed as Tavg is attempting to lower due to boration. With steam dumps modulating closed, Tavg will stabilize until they are fully closed Plausible from a heat balance standpoint if the applicant believes that only power is changing. However, Tavg will be 10werin~J continuously from the negative reactivity C.

insertion.

Plausible because this is an instantaneous capsule response on normal plant trip.

D. However, Tavg will be lowering continuously from the negative reactivity insertion.

RRT04C Technical Reference(s): RRT07C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # 60219 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 1 55.43 Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

Comments:

Meets KJA by requiring knowledge of Tavg response to an emergency boration at power. FR S.1 is the only time emergency boration is initiated at power at GINNA.

11B 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Considered higher order because the applicant must apply reactor theory and HT-FF in a dynamic situation.

1201 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KJA# 074 EK2.08 Importance Rating 2.5 Knowledge of the interrelations between the following Inadequate Core Cooling: Sensors and detectors Proposed Question: RO Question # 58 Plant conditions:

A small break LOCA resulted in a reactor trip and safety injection

The crew was performing E-1. LOSS OF REACTOR OR SECONDARY COOLANT.

when multiple ECCS failures resulted in transition to FR-C.1. RESPONSE TO INADEQUATE CORE COOLING

The crew is initiating depressurization of SGs to facilitate SI Accumulator injection Which ONE of the following describes the parameters monitored to determine if the depressurization will be stopped?

SG pressure at 160 psig; Thot below 390'F A.

SG pressure at 160 psig; Core Exit Thermocouples below 700'F B.

RCS pressure at 160 psig; Thot below 390'F C.

RCS pressure at 160 psig; Core Exit Thermocouples below 700'F D.

Proposed Answer: A Explanation (Optional):

Correct. WR Thot is evaluated at less than 390' F so that the depressurization can be terminated until ECCS Accumulators are isolated. SG pressure at 160 psig when A

depressurization is stopped Incorrect. SG pressure is correct. However, CETs are evaluated throughout FR-C.1 and also as entry conditions to FR-C.1. but not for purposes of RCS depressurization B.

for accumulator injection.

121 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because the pressure value is correct and Thot is also correct, but the pressure is required for SGs, not RCS. The applicant may consider 160 psig in C,

RCS because RCS pressure must be low for accumulator injection.

Incorrect. Plausible for same reasons as Band C.

0,

. FR-C.1 Step 16 Technical Reference{s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRC1C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KIA by requiring identification of the correct detector and the associated value requiring action during a high-level evolution in FR-C.1. While it appears to be a setpoint memorization question, it can be answered by knowing the bases associated with the step.

122 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KIA # 061 AK3.02 Importance Rating 3.4 Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system Proposed Question: RO Question # 59 Plant conditions:

Reactor at 98% power

NIS Channel N-44 is out of service

Reactor Engineering is running an incore flux map to determine QPTR

Annunciator E-24, RMS AREA MONITOR HIGH ACTIVITY, activates

Monitor R-7, Incore Detection Area, is in alarm

No other RMS alarm conditions exist Which ONE of the following describes the correct action to be taken?

Notify RP and request a containment air sample.

A Run an RCS leak rate surveillance to determine if an RCS leak has initiated.

B.

Verify that R-7 is functioning properly; this is an expected alarm for these conditions.

C.

Direct Reactor Engineering to terminate the flux map while the cause is investigated.

D.

Proposed Answer: C Explanation (Optional):

Plausible because this would be a logical action if the flux map was not in progress.

A Plausible if applicant assumes every containment RMS alarm should be addressed as B. potential indication of RCS leakage into containment.

CORRECT. The sensitivity and setpoint are such that it does alarm when the fission C. chambers are not in the storage location or reactor.

D. Plausible because RMS alarm response procedures often direct termination of activities 123 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination that would interfere with analysis of the situation or may have caused the alarm.

AR-RMS-7 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R3901 C, 4,01 Learning Objective: (As available)

Question Source: Bank # C072,0024 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55,4 1 11 55,43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

124 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # E02 EA1.3 Importance Rating 3.8 Ability to operate and 1 or monitor the following as they apply to the (SI Termination) Desired operating results during abnormal and emergency situations.

Proposed Question: RO Question # 60 Plant conditions:

An RCS leak resulted in a MANUAL SI and CI.

The crew is now performing ES-1.1, SI TERMINATION, and is restoring letdown.

The ES-1.1 FOLDOUT requires monitoring subcooling for SI REINITIATION requirements.

Which ONE of the following identifies the required method for monitoring subcooling and the action required if subcooling goes below the minimum value?

RCS Subcooling Margin Monitor; Start SI Pumps as necessary A.

RCS Subcooling Margin Monitor; Initiate a MANUAL SI and CI B.

FIG 1.0 - FIGURE MIN SUBCOOLlNG; Start SI Pumps as necessary C.

FIG 1.0 - FIGURE MIN SUBCOOLlNG; Initiate a MANUAL SI and CI D.

Proposed Answer: C Explanation (Optional):

Plausible because the RCS Subcooling Margin Monitor is available for monitoring but there is no alarm associated with the reinitiation setpoint and the FOLDOUT specifically A.

directs the use of FIG 1.0. The second palrt is correct.

Plausible because the RCS Subcooling Margin Monitor is available for monitoring but there is no alarm associated with the reinitiation setpoint and the FOLDOUT specifically B.

directs the use of FIG 1.0.

Correct.

C.

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because the foldout directs use of Figure 1.0 but SI initiation is not required.

D.

Techmca* IR f () ES-1.1, FOLDOUT Page (Attach if not previously provided) e erence s :

Proposed References to be provided to applicants during examination: None RES11, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KIA by requiring knowledge of indication(s} monitored for subcooling and action required if minimum value is NOT met.

126 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # E13 EA2.2 Importance Rating 3.0 Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure) Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Proposed Question: RO Question # 61 Plant conditions:

The plant has tripped.

The crew has entered FR-H.2, Response to Steam Generator Overpressure, based upon a YELLOW condition on the Heat Sink CSF Status Tree.

SG 'A' pressure is 1150 psig.

SG 'B' pressure is 1050 psig.

SG 'A' Narrow Range level is 55%.

Instrument Air header pressure has been lost.

Which ONE of the following actions will mitigate the steam generator overpressure condition in accordance with FR-H.2?

Raise SG Blowdown Flow rate.

A Open MSIV Bypass Valves and/or steam supply to TDAFW pump.

B.

Transition to FR-H.3 to reduce pressure by reducing SG level.

C.

Place Steam Dump Controller in MANUAL in the STEAM PRESSURE mode, and D. increase demand.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because raising SG blowdown flow would lower level, but this is not A in accordance with FR-H.2 Correct. Even without instrument air, the MSIV Bypass Valves can be opened to B.

reduce pressure 127 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. SG level would have to be >90% to perform this action but it is plausible C. because it is a step in the procedure.

Incorrect. Instrument Air is not available, so steam dumps would also not be available.

Plausible because if IA was available, this could possibly be used if MSIS had not D.

occurred.

. FR-H.2 Rev 7 Technical Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRH2C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Called new but developed for a Watts Bar exam in 2007 that was not administered 128 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 2 KJA# 069 2.2.39 Importance Rating 3.9 Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems.

Proposed Question: RO Question # 62 Plant conditions:

The unit is returning to service from a refueling outage

Currently in Mode 4, heating up to Mode 3

An AO in the final group exiting containml3nt reports that the air lock inner door seal appears to be damaged

The Shift Manager has made the conservative decision to call the affected air lock inoperable until further evaluation is completed Which ONE of the following actions must be performed within one hour?

Terminate the heatup and return to Mode 5.

A.

Verify the affected air lock outer door is closed.

B.

Verify the affected air lock door interlock mechanism is operable.

C.

Run the leakage surveillance test on the unaffected air lock to verify operability.

D.

Proposed Answer: B Explanation (Optional):

Plausible because all containment operability requirements are effective in Mode 4.

However with the unit already in Mode 4, compliance with the REQUIRED ACTION A.

allows the heatup to continue.

CORRECT. ITS 3.6.2, REQUIRED ACTION A.1.

B.

Plausible because the interlock affects air lock operability. However, the air lock is C.

already inoperable and the interlock is designed to prevent opening both doors 129 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination simultaneously; not to compensate for a bad seal.

Plausible as similar to the requirements for dual train components (such as EDG).

However, the inner and outer doors in each air lock provide their own backup protection, D. not the other air lock. In fact, one door could be inoperable in each air lock as long as the REQUIRED ACTION for each one is met.

ITS 3.6.2, REQUIRED ACTION Technical Reference(s): A.1. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 9 55.43 Shielding, isolation, and containment design features, including access limitations.

Comments:

Meets KIA by requiring knowledge of a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months REQUIRED ACTION for a containment integrity issue.

130 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 076 AK3.05 Importance Rating 2.9 Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity: Corrective actions as a result of high fission-product radioactivity level in the RCS Proposed Question: RO Question # 63 Plant conditions:

50% power

8 days ago, after a rapid power reduction from 100% power, RCS activity level began rising slowly and has continued to slowly rise

Radiation Monitor R-9 has gone into high alarm and AR-RM-9, R-9 LETDOWN LINE MONITOR, has directed entry into AP-RCS-3, HIGH REACTOR COOLANT ACTIVITY

The CRS determines that a plant shutdown and reduction in RCS Tavg is required.

Which ONE of the following describes the reason that a reduction in RCS temperature is required in accordance with AP-RCS.3 and/or Technical Specifications?

The RCS must be placed in an operatiom:ll mode where the TS limit on RCS activity A. does not apply.

RCS temperature must be lowered to comply with accident analysis assumptions for a B. Main Steam Line Break with TS maximum primary to secondary leak rate.

RCS temperature must be below the saturation temperature for the MSSV lift setpoints C. in case a steam generator tube rupture occurs.

Reducing RCS temperature raises the decontamination factor of Letdown System ion exchangers, providing reasonable assurance that 10CFR100 dose limits will not be D.

exceeded in the event of a LOCA.

Proposed Answer: C Explanation (Optional):

A. Incorrect. The statement by itself is true but is an incorrect answer for the question 131 5n/2010 Rev Draft

Ginna 2010 NRC Written Examination proposed.

Incorrect. Plausible because it is similar to other TS bases related to SG tube leakage B.

Correct. In the event of a required shutdown, Tavg must be lowered to 500'F to C. minimize the chance that an MSSV will open during a SGTR.

Incorrect. Plausible because it is similar to discussions about reactivity effects of raising D. and lowering temperature throl1gh an ion exchanger.

Technical Reference(s): TS 3.4.16 Basis (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP17C, 1.03 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KIA by requiring knowledge of both an operator action in a procedure with only one "hands-on" operation and the basis for a required shutdown and cooldown.

132 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KJA# E10 2.2.37 Importance Rating 3.6 Equipment Control: Ability to determine operability and 1 or availability of safety related equipment.

Proposed Question: RO Question # 64 Plant conditions:

A reactor trip occurred with a subsequent loss of off-site power

Bus 16 tripped on protective relay actuation

Charging Pump 1A is running

AOV-427, LETDOWN ISOLATION VALVE, is failed closed

Excess Letdown is NOT in service

The crew is performing ES-0.3, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL

The crew has completed the cooldown to 500"F and are preparing to "Depressurize ReS To 1500 PSIG" Which ONE of the following describes the component(s) available to depressurize the RCS to 1500 PSIG in accordance with ES-0.3?

Auxiliary Spray Valve, AOV-296 A.

PRZR PORV-430 ONLY B.

PRZR PORV-431C ONLY C.

PRZR PORV-430 and PRZR PORV-431C D.

Proposed Answer: B Explanation (Optional):

Plausible because this is the preferred path but is not implemented with Letdown OOS.

A.

CORRECT. A procedure NOTE specifies using a PORV with an operable block valve.

B.

Power is available to MeC "c" which feeds MOV-516, the block valve for PORV-430.

133 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible because the valve numbers and power supplies are somewhat inverted in that C. MCC "C" feeds MOV-516 and MCC "0" feeds MOV-515.

Plausible if applicant does not know that only one PORV is opened if Auxiliary Spray is O. unavailable.

ES-0.3, NOTE prior to Step 5 and Technical Reference(s): Step 5 (Attach if not previously provided)

P-12 Proposed References to be provided to applicants during examination: None RES03C, 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

Meets KIA by having applicant identify the depressurization flowpath that meets the operability requirements of the procedure step in ES-0.3.

Higher order because the applicant must carefully evaluate plant conditions and apply the procedure NOTE to correctly answer the question.

134 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: level RO SRO Tier # 1 Group # 2 KJA# 003 AA1.01 Importance Rating 2.9 Ability to operate and I or monitor the following as they apply to the Dropped Control Rod:

Demand position counter and pulse/analog converter Proposed Question: RO Question # 65 The operating crew is preparing to retrieve a dropped Bank D control rod in accordance with ER-RCC.1, RETRIEVAL OF A DROPPED RCC Which ONE of the following describes the operation of the associated Group Step Counters and Pulse to Analogue (PIA) converter prior to the rod retrieval?

Set only the associated group step counters to ZERO; select the appropriate bank on A. the PIA converter and reset the digital indication to ZERO.

Set only the associated group step counte!rs to ZERO; record the reading on the PIA B. converter but do NOT change the settings.

Set both group step counters in the associated bank to ZERO; select the appropriate C. bank on the PIA converter and reset the digital indication to ZERO.

Set both group step counters in the associated bank to ZERO; record the reading on the D. PIA converter but do NOT change the settings.

Proposed Answer: A Explanation (Optional):

CORRECT. The PIA Converter feeds only rod control components associated with Control Bank rods and it is manually set to ZERO after the bank position is recorded A.

and before rod withdrawal begins.

Plausible because the first part is correct and the setting must be correct. However, it is B. set to ZERO so that it counts up to the bank position as if the bank were moving.

Plausible in that the applicant must understand the function of the PIA converter. Since C. there is no overlap or insertion alarm assclciated with the Shutdown Bank rods, the PIA converter is unnecessary. The second part is correct.

135 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Plausible in that the applicant must understand the function of the PIA converter. Since there is no overlap or insertion alarm associated with the Shutdown Bank rods, the PIA D.

converter is unnecessary. The second part balances the choices.

Techmca ~

I Relerence ()

s : ER-RCC.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RER11C, 4.0 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

Meets KIA by requiring knowledge of the design application and operation of the PIA Converter during a dropped rod retrieval. The PIA Converter is in the field at GINNA so the control room team is directing someone to set it.

136 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 1 KJA# G1 2.1.5 Importance Rating 2.9

- - - -...... ~ - -......-

Conduct of Operations: Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Proposed Question: RO Question # 66 Plant conditions:

The date is 6/21/2010.

The time is 1615.

You are scheduled to relieve the shift at 1830 this evening.

You have worked the 1900-0700 shift for the past 5 nights.

You receive a call from the on-shift CRS to provide emergency relief for the HCO, who was transported to the hospital at 1545 this afternoon.

Which ONE of the following describes (1) the LATEST time that a relief must be on-shift, and (2) whether you may arrive PRIOR to your scheduled shift and COMPLETE your shift WITHOUT violating overtime limitations?

(1) 1645 (2) You may relieve the shift early and work your entire shift without violating overtime A.

limitations (1) 1645 (2) You may NOT relieve the shift early and work your entire shift without violating B.

overtime limitations (1) 1745 (2) You may relieve the shift early and work your entire shift without violating overtime C.

limitations (1) 1745 (2) You may NOT relieve the shift early and work your entire shift without violating D.

overtime limitations Proposed Answer: D 137 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. Plausible because only one overtime limit will be violated, and applicant may A. not consider it. Additionally, one hour is reasonable to provide shift relief Incorrect. Plausible because the second half is correct and one hour is reasonable to B. provide shift relief Incorrect. The time is correct for providing relief but the applicant may not relieve the shift and continue through entire shift because they would violate 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in 7 day C.

requirements Correct. Would violate the rule for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in 7 days D.

Technical Reference(s): ~~g*-~~-1.01-1002 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

138 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KJA# G1 2.1.8 Importance Rating 3.4 Conduct of Operations: Ability to coordinate personnel activities outside the control room.

Proposed Question: RO Question # 67 Plant conditions:

A station blackout has occurred.

The crew is performing actions of ECA-O.O, Loss of All AC Power.

Off-Site power has NOT been restored.

BOTH EDGs are tripped and have NOT been restarted.

SGs depressurization is in progress.

Which ONE of the following actions may be taken to provide power to a safeguards bus, and the procedure use following the power restoration?

Direct an AO to align the TSC Diesel to ...

Bus 14; the bus is considered restored when power is aligned and the crew may A. transition to the appropriate recovery procedure when directed.

Bus 14; the bus is NOT considered restored when power is aligned and the crew must B. remain in ECA-O.O until another power source is aligned to a safeguards bus.

Bus 16; the bus is considered restored when power is aligned and the crew may C. transition to the appropriate recovery procedure when directed.

Bus 16; the bus is NOT considered restored when power is aligned and the crew must D. remain in ECA-O.O until another power source is aligned to a safeguards bus.

Proposed Answer: D Explanation (Optional):

Incorrect. TSC diesel is aligned to opposite safeguards bus. Although the bus would A. be energized, it is NOT considered restored lAW ECA-O.O B. Incorrect. second part is correct but bus 16 is restored, not bus 14 139 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect Correct bus identified, but bus is not considered restored, and crew must C. remain in ECA-O.O until the bus is restored for an emergency DG Correct. Note prior to ECA-O.O, Step 27 D.

IR f () ECA-O.O, note prior to step 27 Techmca e erence s : (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R0801C, Obj 1.10 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

140 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 KIA # G2 2.2.37 Importance Rating 3.6 Equipment Control: Ability to determine operability and I or availability of safety related equipment.

Proposed Question: RO Question # 68 Plant conditions:

The plant is operating at 100% power.

A loss of DC Bus 'A' occurs.

Which ONE of the following describes equipment that is made INOPERABLE by this failure?

All Steam Dump Valves A.

Pressurizer PORV 431 C B.

Train 'A' SI Actuation Instrumentation C.

Turbine Driven AFW Pump D.

Proposed Answer: C Explanation (Optional):

Incorrect. Losing DC Bus 'B' would make this true. Losing DC Bus 'A' will fail several A. valves, and the remainder will only go full open or closed Incorrect. PORV 430 is inoperable, but PORV 431 C is powered by Train B B.

Correct. DC Power required for SI actuation circuitry C.

Incorrect. If DC Bus B was lost, the TDAFW Discharge MOV would fail as is, but Train D. A loss does not make TDAFW inoperable Technical Reference(s): ER-ELEC.2, Attachment 1 (Attach if not previously provided) 141 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None R0901C, Obj 1.06 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamt:mtal Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 D~si9n, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

Comprehension because it requires understanding not only of power supplies, but the effect it has on the components that receive power 142 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 KIA # G2 2.2.35 Importance Rating 3.6 Equipment Control: Ability to determine Technical Specification Mode of Operation.

Proposed Question: RO Question # 69 Plant conditions:

The plant is shutdown, preparing for refueling.

All systems are normally aligned for the current plant conditions.

At 0600, the following conditions are noted:

Cold Leg WR Temperature indicators are 17E.oF.

WR RCS Pressure indicators are 100 psig.

Maintenance is preparing to de-tension reactor vessel head closure bolts.

RHR is lost and ALL Cold Leg WR Temperature indicators begin rising at 2°F/minute.

Based on the above indications and assuming the above conditions and trends continue, which ONE of the following correctly identifies the plant MODE at 0600 and MODE at 0700?

MODE at 0600 MODE at 0700 MODE 5 MODE 3 A.

MODE 5 MODE4 B.

MODE 6 MODE 4 C.

MODE 6 MODE 5 D.

Proposed Answer: B Explanation (Optional):

Incorrect. At 0600, plant is initially in Mode 5 (RCS temp < or equal to 200°F, Keff <

0.99). At 0700, RCS temperature will be 295°F, which is Mode 4 (RCS temp < 350°F.

A.

Plausible because the applicant may eithe!r calculate incorrectly or confuse modes 143 5n/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. Plant is in Mode 5 initially. Plant is in Mode 4 starting at approximately 0613 B.

Incorrect. Plant is in Mode 5 initially. Plant would be in Mode 6 initially if one or more reactor vessel head bolts was less than fully tensioned and will enter Mode 4 at 0613 if C.

current trends continue.

Incorrect. Plant is in Mode 5 initially. Plant would be in Mode 6 initially if one or more D. reactor vessel head bolts was less than fully tensioned.

Technical Specification table 1.1 Technical Reference(s): 1, Definitions (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 6S657 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Question meets KA. Question requires examinee ability to determine Mode of Operation based on plant indications and procedures.

Modified from Braidwood 2007 NRC exam. Modified heatup rate from 3 to 2 to change correct answer 144 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 3 KIA # G3 2.3.14 Importance Rating 3.4 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Proposed Question: RO Question # 70 Plant conditions:

A Steam Generator Tube Rupture has occurred on SG "B".

The crew has performed all actions of E-3, Steam Generator Tube Rupture, up to the step to commence depressurization of the RCS.

All equipment is fUnctioning as designed.

Which ONE of the following describes the status of "B" SG Atmospheric Relief Valve, and the reason for the status?

CLOSED with controller in Manual; prevent radioactive release to atmosphere.

A.

CLOSED with controller in Manual; ensures minimum RCS subcooling will be B. maintained when RCS depressurization is initiated.

Set at 1050 psig with controller in AUTO; prevent uncontrolled radioactive release due C. to SG safety valve lifting.

Set at 1050 psig with controller in AUTO; ensures minimum RCS subcooling will be D. maintained when RCS depressurization is initiated.

Proposed Answer: C Explanation (Optional):

Incorrect. Plausible because it is logical to maintain valve closed but controller will not A. be in manual. Reason is correct Incorrect. Same reason as option A, and additionally, reason is plausible because if the ARV stuck open on a ruptured SG, the depressurization would also cause B.

depressurization of the RCS. This would result in loss of RCS subcooling.

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. Controller is in AUTO which allows valve to open as required if pressure rises.

This prevents safety valves from lifting and potentially becoming stuck open, causing

c. radioactive release.

Incorrect. Correct for status of valve, but reason is incorrect. Plausible because valve would be placed in manual and closed if it stuck open below 1050 psig, but this is not D. the reason that the valve is placed in AUTO. The remainder of the steps for SG isolation are correct for this reason.

E-3, Rev 46, step 3 Technical Reference{s): Westinghouse Setpoint Document (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REP03C, Obj 1.03, 2.01 Learning Objective: (As available)

Question Source: Bank # WTSI64973 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Wolf Creek 2009 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

146 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 3 KIA # G3 2.3.13 Importance Rating 3.4 Radiation Control: Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Proposed Question: RO Question # 71 Plant conditions:

The unit is at 90% power.

All systems are in their normal alignment.

Auxiliary Building (AB) Ventilation IMS Switch is in " FILTER IN".

"C" Gas Decay Tank (GDT) is being released.

1. What is the effect if R-13, Auxiliary Building Particulate Monitor, or R-14, Auxiliary Building Gas Monitor, goes into high alarm?

and

2. What are the operator actions for that plant response?

1. If R-13, goes into hjgh alarm, "C" Gas Decay Tank release is automatically secured.

2. Verify the GOT release AOV-RCV-014 to the plant vent closes.

A. Ensure 1F AB Exhaust fan is no longer running.

Ensure 1A, 1 Band 1C Intermediate Building Exhaust fans are no longer running

1. If R-13, goes into high alarm, "C" Gas Decay Tank release is automatically secured.

2. Verify the GDT release AOV-RCV-014 to the plant vent closes.

B. Ensure 1A, 1B, 1C and 1F AB Exhaust fans are no longer running.

Ensure 1A, 1Band 1C Intermediate Building Exhaust fans are no longer running.

1. If R-14, goes into high alarm, "c" Gas Decay Tank release is automatically secured.

2. Verify the GDT release AOV-RCV-0141:0 the plant vent closes.

C. Ensure 1F AB Exhaust fan is no longer running.

Ensure 1A, 1Band 1C Intermediate Building Exhaust fans are no longer running.

1. If R-14, goes into high alarm, "C" Gas Decay Tank release is automatically secured.

2. Verify the GOT release AOV-RCV-014 to the plant vent closes.

D.

Ensure 1A, 1B, 1C and 1F AB Exhaust fans are no longer running.

Ensure 1A, 1Band 1C Intermediate Building Exhaust fans are no longer running.

147' 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed Answer: C Explanation (Optional):

R-13 will cause this realignment of the AB exhaust system but it will not stop the Gas A. Decay Tank release so RCV-14 (the GDT release AOV to the plant vent) will not close R-13 will cause a realignment of the AB exhaust system but it will not stop the Gas Decay Tank release so RCV-14 (the GDT release AOV to the plant vent) will not close.

B. This is correct if the Auxiliary Building (AB) filters are "OUT", but in the stem the Auxiliary Building (AB) filters are "IN", so the additional fans are not affected With a high alarm on R-14 and the AB filters are "IN" the automatic actions that should occur are: RCV-14 (the GDT release AOV to the plant vent) closes. 1F AB Exhaust fan C.

receives a trip signal. 1A, 1Band 1C Intermediate Building fans receive trip signals With a high alarm on R-14, RCV-14 (the GDT release AOV to the plant vent) closes.

This is correct if the Auxiliary Building (AB) filters are "OUT", but in the stem the D.

Auxiliary Building (AB) filters are "IN", so the additional fans are not affected LP R3901 C, Radiation Monitoring System pg. 18 Technical Reference(s): LP R3801 C, Waste Disposal (Attach if not previously provided)

System pg.25 Proposed References to be provided to applicants during examination: None R3901C Obj 1.03 Learning Objective: (As available)

Question Source: Bank # WTSI65980 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2008 Ginna Question Cognitive Level: Memory or Fundamental Knowledge 148 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

149 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 4 KJA# G4 2.4.30 Importance Rating 2.7 Emergency Procedures 1 Plan; Knowledge of eVE!nts related to system operation I status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

Proposed Question: RO Question # 72 Plant conditions:

A reactor trip occurred from 100% power.

The crew is performing ES-O.1, Reactor Trip Response.

Due to equipment failures following the trip, the Shift Manager declares an Unusual Event based on his conservative judgment and opinion.

There is NO radiological concern or release in progress of anticipated.

Which ONE of the following describes whether or not the local and state authorities must be notified, and the reason why?

State and Local authorities ...

must be notified because an unplanned re,actor trip has occurred.

A.

must be notified because the event resulted in declaration of an Emergency B. Classification.

are NOT required to be notified because there is no radioactive release in progress or C. anticipated.

are NOT required to be notified because local and state notifications are only required D. at the ALERT level or above.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because the NRC would be notified within 4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months depending A. on the event and level of equipment malfunction, if there was no classification 150 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. State and Local notifications are required within 15 minutes for emergency B. classifications Incorrect. Plausible because with no release in progress, the applicant may believe that since evacuation or shelter are not required, that local authorities do not require C.

notification.

Incorrect. Plausible because the entire ERO is not required to be manned unless the D. classification is alert or above, but state and local must always be notified

. EPIP 1-1 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

151 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3

---~" ....-

Group # 4 KIA # G4 2.4.39 Importance Rating 3.9 Emergency Procedures I Plan: Knowledge of the RO's responsibilities in emergency plan implementation.

Proposed Question: RO Question # 73 Plant conditions:

The plant is in a General Emergency.

The Ginna Station Nuclear Emergency Response Plan is being implemented.

From the list below, which one of the following correctly identifies the responsibilities of the HCO/CO during the General Emergency in accordance with EPIP 5-7, Emergency Organization?

1. Report all unusual observations or communications to the Shift Manager.

2. Sound the alarm and make announcements as necessary.

3. Peer check the Shift Manager in the diagnosis of unusual event conditions and above.

4. Check that the Control Room ventilation system is in re-circulation mode.

5. Assist as directed and inform Shift Manager of all Control Room changes.

6. Ensure search and rescue is initiated, if necessary, per EPIP 1-8.

1,3,5 A.

2,4,6 B.

1,3,6 C.

2,4,5 D.

Proposed Answer: D 152 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. # 1 is the responsibility of the C/R Communicator. # 3 is the responsibility of A. the STA per EPIP 5-7, Emergency Organization Incorrect. # 6 is the responsibility of the Emergency Coordinator per EPIP 5-7, B. Emergency Organization Incorrect. # 1 is the responsibility of the CIR Communicator. # 3 is the responsibility of the STA per EPIP 5-7, Emergency Organization .. # 6 is the responsibility of the C.

Emergency Coordinator.

Correct Per EPIP 5-7, Emergency Organization D.

EPIP 5-7, Emergency Technical Reference(s): Organization (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None N/A Learning Objective: (As available)

Question Source: Bank# WTSI65965 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Ginna 2008 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or A.nalysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

153 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group ~~ 2 KJA# G2 2.2.13 Importance Rating 4.1 Equipment Control: Knowledge of tagging and clearance procedures.

Proposed Question: RO Question # 74 Which ONE of the following jobs would REQUIRE an Isolated Work Area (a TAGOUT), due to its meeting the definition of a "Hazardous Energy" in accordance with CNG-OP-1. 0 1-1007, Clearance and Safety Tagging? (Assume NO instrument calibrations will be performed.)

Work on a(n) ...

AC circuit where the maximum voltage is 45 volts AC.

A.

DC circuit where the maximum voltage is 40 volts DC.

B.

Hydraulic system which has a maximum pressure of 100 psig and a maximum C. temperature of 75°F.

Hydraulic system which has a maximum pressure of 40 psig and a maximum D. temperature of 110°F.

Proposed Answer: C Explanation (Optional):

Incorrect. Per procedure a hold is required for> 50 volts (AC OR DC).

A.

Incorrect. Per procedure a hold is required for> 50 volts (AC OR DC).

B.

Correct. Per procedure a hold is required for> 50 psig hydraulic pressure and/or>

C. 120°F temperature Incorrect. Per procedure a hold is required for> 50 psig hydraulic pressure and/or>

D. 120°F temperature

. CNG-OP-1.01-1007 Technical Reference(s): (Attach if not previously provided) 154 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None NIA Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI 65954 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

I didn't have this procedure, just changed the question based upon the information in the existing question Modified from a 2008 Ginna NRC exam question 155 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 1 KIA # G1 2.1.3 Importance Rating 3.7 Conduct of Operations: Knowledge of shift or sholrt-term relief turnover practices.

Proposed Question: RO Question # 75 Given the following:

The time and date is 0630, June 25, 2010.

You are the on-coming HCO performing shift turnover.

The last time you were on shift was June 21, 2010.

In accordance with OPS-SHIFT-TURNOVER, prior to assuming the shift, you must review shift logs (official records) back to AT LEAST 0630 on ...

June 24 A.

June 23 B.

June 22 C.

June 21 D.

Proposed Answer: B Explanation (Optional):

Incorrect. June 24 would represent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and the MINIMUM review is 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or back to the last time you were on shift, whichever is less. Plausible because it is A.

reasonable that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months could be the standard Correct. 2 days is 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , which is less time than the last time you were on shift B,

Incorrect. Time represents 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , which is plausible because it falls directly between C. the 2 times identified in the procedure Incorrect. Plausible because it is the last time you were on shift, which is one of the 2 D. times identified in the procedure.

156 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

. CNG-OP-1.01-1002 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

157 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group:#: 1 KJA# 056 AA2.23 Importance Rating 3.9 Ability to determine and interpret the following as they apply to the Loss of Offsite Power:

Turbine trip-reactor button and indicator Proposed Question: SRO Question # 76 Plant conditions:

The plant is at 100% power.

A reactor trip occurred

Subsequent to the trip, a loss of off-site powelc occurred.

Turbine stop valve position cannot be verified.

The CO manually depresses the Turbine Trip pushbutton.

He reports that he still cannot determine whether the turbine has tripped.

Which ONE of the following describes the action that will be required next, and if conditions require the plant to be placed in Mode 5, which procedure will be used to perform RCS cooldown to RHR entry conditions?

Shut MSIVs; ES-0.2, Natural Circulation Cooldown.

A.

Shut MSIVs: 0-2.2, Plant Shutdown From Hot Shutdown to Cold Conditions.

B.

Manually run back the turbine or trip EHC Pumps; ES-0.2, Natural Circulation C. Cooldown.

Manually run back the turbine or trip EHC Pumps; 0-2.2, Plant Shutdown From Hot D. Shutdown to Cold Conditions.

Proposed Answer: A Explanation (Optional):

Correct. The crew will go from E-O to ES-0.1, then transition to ES-0.2 until RHR entry A. conditions. 0-2.2 can be used once Mode! 5 is reached Incorrect. Closing MSIVs is correct but 0-2.2 will not be used until after exit from ES B.

0.2 158 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because these methods will also result in a turbine shutdown.

C. Correct procedure Incorrect. Plausible same reason as option C but wrong procedure in this option D.

E-O, step 2 RNO Technical Reference(s): E-O, step 20 RNO (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None REPOO, Obj 2.01 Learning Opjective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

159 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 055 EA2.04

-~--

Importance Rating 4.1 Ability to determine or interpret the following as they apply to a Station Blackout: Instruments and controls operable with only dc battery power available Proposed Question: SRO Question # 77 Plant conditions:

The plant was at 100% power.

A loss of all off site power occurred approximately 20 minutes ago.

Both "A" and "B" Diesel Generators failed to start and cannot be started.

Which ONE of the following Control Room controls or indications will remain usable to control the initial response and the impact on reporting to off-site authorities?

Microprocessor Rod Position Indication (MRPI)

First report to off-site authorities is required within 15 minutes of initial emergency A.

classification Steam Generator ARV Controllers First report to off-site authorities is required within 15 minutes of initial emergency B.

classification Microprocessor Rod Position Indication (MRPI)

First report to off-site authorities is required within 60 minutes of initial emergency C.

classification Steam Generator ARV Controllers First report to off-site authorities is required within 60 minutes of initial emergency D.

classification Proposed Answer: B Explanation (Optional):

Incorrect - MCC-1 K supplies power to MRPI and will be de-energized, and state and A. local authorities must be notified within 15 minutes 160 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct - ARVs are available for use as stated in ECA-O.O, and 15 minutes is the B. required first call Incorrect - MCC-1 K supplies power to MRPI and will be de-energized, and 60 minutes is C. plausible because the NRC is required to be notified in that time Incorrect. ARVs are available, and also plausible on time for same reason as option C D.

ECA-O.O Technical Reference(s): EPIP 1-0 (Attach if not previously provided)

EPIP 1-5 Proposed References to be provided to applicants during examination: None R3101C, Obj 1.02 Learning Objective: RECOOC, Obj 2.01 (As available)

Question Source: Bank # WTSI62905 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Callaway 2009 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

changed distracter options but stem essentially left intact.

161 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1

.... -~

Group # 1 KIA # 026 AA2.06

--.~.---.--

Importal1ce Rating 3.1 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The length of time after the loss of CCW flow to a component before that component may be damaged Proposed Question: SRO Question # 78 Initial conditions:

The plant is operating at 100% power.

Component Cooling Water temperatures have been rising, and the CRS is addressing TS 3.7.7 for operability of the Component Cooling Water System.

To comply with the action statement of TS 3.7.7, the CRS determines that a plant shutdown to Mode 3 is required.

Current conditions:

A loss of Component Cooling Water has occurred and the crew is performing actions of AP CCW.2, Loss of CCW During Power Operation.

The CCW System has been without flow for 2 minutes.

RCP motor bearing temperatures are 192°F and rising.

Which ONE of the following describes the action required, and assuming operability cannot be restored to CCW, the effect on compliance with TS 3.7. 7?

Trip the reactor, trip the RCPs, and perform E-O, Reactor Trip or Safety Injection; All A requirements of TS 3.7.7 are met when the plant is in Mode 3.

Remain in AP-CCW.2 and continue attempts to restore CCW; All requirements of TS B. 3.7.7 are met when the plant is in Mode 3.

Trip the reactor, trip the RCPs, and perform E-O, Reactor Trip or Safety Injection; All requirements ofTS 3.7.7 will NOT be met until the plant is in Mode 4.

C.

Remain in AP-CCW.2 and continue attempts to restore CCW; All requirements of TS 3.7.7 will NOT be met until the plant is in Mode 4. Plausible because initial action is to D.

go to Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed Answer: C Explanation (Optional):

Incorrect. TS 3.7.7 requires the unit to be in Mode 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if operability A. cannot be restored.

Incorrect. Plausible because RCP temperatures have not reached the trip setpoint yet.

B. Action is correct for operability Correct. TS 3.7.7 requires continuing to Mode 4. RCPs have been without CCW for C. over 2 minutes Incorrect. Plausible same as Option B D.

AP-CCW.2 Technical Reference(s): (Attach if not previously provided)

TS 3.7.7 Proposed References to be provided to applicants during examination: None RAP02C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamemtal Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

163, 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group '# 1 KIA # 062 2.4.34 Importance Rating 4.1 Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Proposed Question: SRO Question # 79 Plant conditions:

The plant is at 100% power.

A Service Water leak is in progress.

The crew is performing AP-SW.1, Service Water Leak.

Leak location has NOT been identified.

All Service Water Pumps are operating.

Temperatures of components cooled by Service Water are stable.

Service Water Header pressure is 48 psig on both loops.

Which ONE of the following describes the local actions that will be performed in accordance with AP-SW.1, and the action required based on current plant conditions?

Refer to AP-SW.2, Loss of Service Water, to reduce unnecessary Service Water loads to raise Service Water header pressure as required.

A.

Trip the reactor, enter E-O, Reactor Trip or Safety Injection Dispatch an operator to split Service Water Headers A and B in an attempt to isolate the leak in accordance with Attachment 2.2.

B.

Trip the reactor, Enter E-O, Reactor Trip or Safety Injection Refer to AP-SW.2, Loss of Service Water, to reduce unnecessary Service Water loads to raise Service Water header pressure as required.

C.

Initiate a controlled plant shutdown using AP-TURB.5, Rapid Load Reduction.

Dispatch an operator to split Service Water Headers A and B in an attempt to isolate the leak in accordance with Attachment 2.2.

D.

Initiate a controlled plant shutdown using AP-TURB.5, Rapid Load Reduction.

164 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed Answer: D Explanation (Optional):

Incorrect. Reactor trip not required, but plausible because it would be required on loss A. of all Service Water Pumps. They will continue in AP-SW.1 Incorrect Reactor trip not required as described in Option A B.

Incorrect. AP-SW.1 does not direct shedcling loads, but is plausible because doing that C. would raise system pressure by reducing demand Correct. AP-SW.1 step 9 identifies action D.

. AP-SW.1 Step 9 Technical Reference(s): ' (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP19C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Meets KA because the headers are split locally 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # E12 2.2.22 Importance Rating 4.7 Equipment Control: Knowledge of limiting conditions for operations and safety limits.

Proposed Question: SRO Question # 80 The crew is performing actions of ECA-2.1, Uncontrolled Depressurization of Both Steam Generators.

Which ONE of the following describes the Technical Specification implications of the event?

RCS cooldown rates above 100°F per hour may potentially result in brittle fracture of the A. reactor vessel.

Pressurizer cooldown rates above 100°F per hour may potentially result in pressurized B. thermal shock to the pressurizer.

Loss of SG inventory may ultimately result in the RCS Pressure Safety Limit being C. exceeded.

Loss of SG inventory may ultimately result in the Reactor Core Safety Limit being D. exceeded.

Proposed Answer: A Explanation (Optional):

Correct. Concern about non-ductile failure (brittle fracture) of reactor vessel is basis for A. maximum cooldown rates.

Incorrect. PTS requires neutron embrittlement and pressurizer does not receive embrittlement. Plausible because PTS is a concern related to cooldown rates, and the B. pressurizer does have administrative restrictions on cooldown rates, just not as limiting as the RCS.

Incorrect. Plausible because ECA-2.1 is associated with loss of SG inventory because there is an uncontrolled loss of SG mass. However, ECA-2.1 accounts for this by

c. directing action for AFW. Also, if SG inventory was lost. the RCS pressure safety limit could be challenged due to lack of heat removal.

16E, 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because the applicant must know the factors affecting the RCS Core Safety Limit, and RCS temperature is a factor. A loss of SG inventory will result in D. a rise in RCS temperature. However, this safety limit is associated with plant operation in Mode 1.

Technical Reference(s): TS 3.4.3 (Attach if not previously provided)

PTLR Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 025 2.4.3 Importance Rating 3.9 Emergency Procedures I Plan: Ability to identify post-accident instrumentation.

Proposed Question: SRO Question # 81 Plant conditions:

The plant is in Mode 5.

RCS temperature is 185°F.

A loss of RHR has occurred due to trip of BOTH RHR Pumps.

Pressurizer Level is 14% and lowering at approximately 1% per minute.

RCS temperature is rising at approximately 2"F per minute.

The Shift Manager is evaluating EPIP-1.0 to determine the Emergency Classification.

Which ONE of the following instruments will be used to determine the appropriate emergency classification for this event?

(Reference Provided)

Core Exit Thermocouples; event currently meets the criteria for EAL classification.

A.

Core Exit Thermocouples; event does NOT currently require classification but will meet B. the criteria for EAL classification in less th':1n 10 minutes.

Pressurizer Level; event currently meets the criteria for EAL classification.

C.

Pressurizer Level; event does NOT currently require classification but event will meet D. the criteria for EAL classification in less than 10 minutes.

Proposed Answer: B Explanation (Optional):

Incorrect. Classification will be required when RCS temperature reaches 200. CETs A. are post-accident instrumentation Correct.

B.

168 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. PRZR level helps diagnose an event, but RVLlS is used for classification on C. RCS inventory Incorrect. PRZR level helps diagnose an event, but RVLlS is used for classification on D. RCS inventory Technical Reference(s): EPIP-1.0, EAL 7.2.4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

169 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group ;# 2 KJA# 036 AA2.03 Importance Rating 4.2 Ability to determine and interpret the following as they apply to the Fuel Handling Incidents:

Magnitude of potential radioactive release Proposed Question: SRO Question # 82 In accordance with Technical Specifications, which ONE of the following describes ONLY plant parameters with LCOs that are designed to limit iodine release and off-site dose following a postulated Fuel Handling Accident?

Refueling cavity level ONLY A.

RCS boron concentration ONLY B.

RCS temperature and Refueling Cavity level C.

RCS temperature and RCS boron concentration D.

Proposed Answer: A Explanation (Optional):

Correct.

A.

Incorrect. Plausible because neutrons are absorbed by boron, but boron concentration B. concern is inadvertent criticality, not rad release from fuel handling accident Incorrect. Plausible because there is a temperature limitation in Mode 6 C.

Incorrect. Same as C D.

Technical Reference(s): TS 3.9.6 basis (Attach if not previously provided) 170 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 7 Fuel handling facilities and procedures.

Comments:

171 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1

-~-~

Group # 2 KIA # 033 AA2.12 Importance Rating 3.1 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Maximum allowable channel disagreement Proposed Question: SRO Question # 83 Plant conditions:

Plant startup in progress.

N31 and N32 are indicating 2X104 CPS

N35 indicates 1.5XlO-10 amps

N36 indicates 7.0X1 0- 9 amps

N41, 42, 43, 44 indicate 0%

Which ONE of the following describes the operability of the Nls, and the actions, if any, that are necessary?

Source Range and Intermediate Range Nls are operable, since Power Range indicates

<8%. NO Technical Specification action is required. A CR must be initiated for the A.

deviation between Intermediate Range Instruments.

Intermediate Range NI deviation is outside of tolerance but will be deenergized when Power Range Nls indicate >8%. NO Technical Specification action is required, but a CR B.

must be initiated.

N35 is reading too low for the current Source Range level. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reactor power must be raised OR lowered to a value where Intermediate Range instruments are C.

not required.

N36 is reading too high for the current Source Range level. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reactor power must be raised OR lowered to a value where Intermediate Range instruments are D.

not required.

Proposed Answer: D Explanation (Optional):

172 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because a CR will be written, but IR is required below P-10.

A.

Incorrect. Plausible because SR de-energizes >P-6, but IR does not. Operability B. requirements change> 10% power.

Incorrect. For the current SR indication, IR should be significantly less than N36.

C.

NI-4 is approximately 1 decade high for the current source range level. Maximum D. disagreement for IR detectors is 5% of scale TS-3.3.1 Technical Reference(s): R3301C, T43C-006A (Attach if not previously provided) 0-6.13 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTSI58467 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamemtal Knowledge Comprehension or A.nalysis x 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

173 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KJA# E03 2.4.30 Importance Rating 4.1 Emergency Procedures / Plan; Knowledge of events related to system operation / status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

Proposed Question: SRO Question # 84 Plant conditions:

A LOCA is in progress.

The crew is initiating an RCS cooldown in accordance with ES-1.2, Post LOCA Cooldown and Depressurization.

The Shift Manager has completed all required local, state, and NRC notifications for an initial Emergency Classification.

Which ONE of the following describes the additional reporting required to the NRC, if any, beyond the reports already completed?

(Reference Provided)

An additional ONE hour report is required ONLY.

A.

An additional FOUR hour report is required ONLY.

B.

Additional ONE hour AND FOUR hour reports are required.

C.

NO additional NRC reports are required because an emergency classification is in effect D.

Proposed Answer: B Explanation (Optional):

Incorrect. Reportability procedure requires one hour for classification, but not for ECCS A. discharge or reactor trip Correct.

B.

Incorrect. Plausible because a four hour report is required, but the one hour report has

c. been completed for EP 174 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible because a report has been made, but there are 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reports that D. also must be made

. CNG-NL-1.01-1004 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: CNG-NL-1.01-1004 N/A Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

I Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

175 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 1 Group '# 2 KIA # E14 2.4.6 Importance Rating 4.7 Emergency Procedures I Plan: Knowledge of EOP mitigation strategies.

Proposed Question: SRO Question # 85 Plant conditions:

A LOCA has occurred.

Component Cooling Water has been lost.

The crew was required to perform ECA-1.1, Loss of Emergency Coolant Recirculation.

The Crew is now entering FR-Z.1, Response to High Containment Pressure.

Containment pressure is 31 psig and STABLE.

BOTH Containment Spray Pumps are OFF.

Which ONE of the following describes the strategy for reducing Containment Pressure?

OPERATE Containment Spray Pumps in accordance with the guidance in ECA- 1.1, as A. directed by FR-Z.1. Continue in FR-Z.1 until exit criteria is met.

START both Containment Spray Pumps in accordance with FR-Z.1. RED CSF B. conditions take precedence over ECA actions.

Perform ONLY the FR-Z.1 actions that do NOT conflict with or undo the action taken in ECA-1.1. Two Containment Recirc Coolers will provide adequate depressurization to C.

meet the Containment Safety Function requirements.

Do NOT perform actions of FR-Z.1 until the RWST LO-LO level alarm is clear and Containment Spray Pumps may be restarted. Ensure all other automatic actions related D. to containment isolation have occurred as required.

Proposed Answer: A Explanation (Optional):

Correct. If Containment Spray Pumps have been shut down in ECA-1.1, the action in A. FR-Z.1 is subordinate to ECA-1.1 actions due to limited inventory for spray and ECCS B. Incorrect. Plausible because red conditions typically do take precedence over ECA 176 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination actions. This is an exception Incorrect. Plausible because Containment Cooling is adequate with 2 coolers, but not C. containment depressurization under emergency conditions.

Incorrect. Plausible because with RWST LO-LO level alarm, there may not be enough D, NPSH for spray pumps to take a suction on the tank.

Technical Reference(s): FR-Z.1, note prior to step 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRZ01C, Obj 2.01 Learning Opjective: (As available)

Question Source: Bank # WTSI55223 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Callaway 2005 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

177' 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 026 A2.04 Importance Rating 4.2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of spray pump Proposed Question: SRO Question # 86 Plant conditions:

Reactor was manually tripped during a loss of Service Water.

Service Water CANNOT be recovered at this time.

The crew transitioned to ES-0.1, Reactor Trip Response

Subsequently, a LOCA occurred.

After transition back to E-O, Reactor Trip or Safety Injection, Containment Spray actuates.

'A' Containment Spray Pump control switch has a TRIP indication.

Train 'A' Containment Spray flow is 0 gpm.

Containment pressure indicates 29 psig and slowly rising.

Which ONE of the following describes the action that will be required?

Enter FR-Z.1, Response to High Containment Pressure, due to a RED Path.

A.

Enter FR-Z.1, Response to High Containment Pressure, due to an ORANGE Path.

B.

Remain in E-O, Reactor Trip or Safety Injection, and use the Continuous Action step to C. attempt to start 'A' Containment Spray Pump.

Continue in E-O and operate Containment Spray in accordance with E-1, Reactor or D. Secondary Coolant when that procedure is in effect.

Proposed Answer: B Explanation (Optional):

Incorrect. Procedure is correct but Containment is in an ORANGE, not RED, condition A.

Correct. Have already left E-O to go to ES-0.1, so CSF Status Tree Monitoring is in B.

effect.

178 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. The step to initiate spray is a CA step, but because CSF STs are in effect, C. the crew will go to FR-Z.1 Incorrect. If already in E-1, the 'crew would follow that guidance as directed in the note D. prior to step 1 of FR-Z.1 Technical Reference(s): A-503.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRZ1C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 059 A2.04 Importance Rating 3.4 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a dry S/G Proposed Question: SRO Question # 87 Plant conditions:

A loss of Feedwater has occurred.

The crew is performing actions of FR-H.1, Response to Loss of Secondary Heat Sink.

SG Wide Range levels are 45 inches and lowering.

Bleed and Feed has been initiated, but only one train of SI is operating.

Core Exit thermocouples are approximately 590°F and slowly lowering.

The crew has been able to restore one Main Feedwater Pump.

Attachment 22, Attachment Restoring Feed Flow, is being performed.

Which ONE of the following describes how feedwater will be initiated in accordance with 2, and which procedure transition will be made when heat sink is restored and Bleed and Feed is terminated?

Initiate flow at less than or equal to 100 gpm; Transition to ES-1.1, SI Termination.

A.

Initiate flow at less than or equal to 100 gpm; Transition to E-O, Reactor Trip or Safety B. Injection.

Initiate flow at maximum rate: Transition to ES-1.1, SI Termination.

C.

Initiate flow at maximum rate: Transition to E-1, Loss of Reactor or Secondary Coolant.

D.

Proposed Answer: A Explanation (Optional):

Correct. With no further complications, the crew will ensure PORVs are closed and SI A. is shut down, then go to ES-1.1 at end of FR-H.1 B. Incorrect. Flow is right but procedure is wrong. Will go to E-1 if can't isolate any bleed 180 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination path. E-O would be correct if directed to transition to procedure and step in effect, and original transition was directly from E-O Incorrect. Maximum flow would be initiatE!d to only ONE SG if RCS temperatures were C. rising. Correct procedure Incorrect. Maximum flow would be initiated to only ONE SG if RCS temperatures were D. rising.

Technical Reference(s): ATT-22.0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRH1C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Part of this was drawn from bank questions, but the transition is added. parameters and conditions new and specific to facility 181 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KJA# 010 2.2.38 Importance Rating 4.5 Equipment Control: Knowledge of conditions and limitations in the facility license.

Proposed Question: SRO Question # 88 Plant conditions:

The plant is at 100% power.

Engineering reported that BOTH Pressurizer Safety Valves were set incorrectly during the last Refueling outage.

Each valve is set to Relieve at 2572 psig.

Which ONE of the following describes the MAXIMUM amount of time allowed to place the Unit in Mode 3 in accordance with Technical Specifications, and the basis for the TS LCO?

Six (6) hours; operability of the PRZR Safety valves ensures that RCS pressure is limited to less than 120% of the Safety Limit for RCS pressure for ALL anticipated A.

transients with the EXCEPTION of an RCP locked rotor.

Six (6) hours; operability of the PRZR SafE~ty valves ensures that RCS pressure is limited to less than 110% of design pressure for ALL anticipated transients with the B.

EXCEPTION of an RCP locked rotor.

Fifteen (15) minutes; operability of the PRZR Safety valves ensures that ReS pressure is limited to less than 120% of the Safety Limit for ReS pressure for ALL anticipated

c. transients with the EXCEPTION of an RCP locked rotor.

Fifteen (15) minutes; operability of the PRZR Safety valves ensures that ReS pressure is limited to less than 110% of desigr'[ pressure for ALL anticipated transients with the D.

EXCEPTION of an Rep locked rotor.

Proposed Answer: B Explanation (Optional):

Incorrect. Plausible because the wording of the basis is close to the actual wording.

A. Locked rotor events may cause pressure to go to 120% of design pressure. Safety limits are 110% of design pressure 182 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months because both valves are inoperable. If one valve was inoperable, 15 B. minutes would be available to correct the problem prior to shutting down Incorrect. 15 minutes is the required TS action time for one safety valve inoperable.

Additionally second half wrong because a locked rotor will reach 120% of design, not C

safety limit Incorrect. Time is incorrect because both valves are affected. Second half is correct D.

Technical Reference{s): TS 3.4.10 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R1901C, Obj 1.13 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

183~

517/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group t~ 1 KIA # 004 2.4.9 Importance Rating 4.2 Emergency Procedures 1 Plan: Knowledge of low power 1 shutdown implications in accident (e.g., loss of coolant accident or loss of residuall1eat removal) mitigation strategies.

Proposed Question: SRO Question # 89 Initial conditions:

The plant is in Mode 4.

RCS temperature is 210°.

The pressurizer is solid.

PRZR pressure is approximately 350 psig.

Charging flow is approximately 75 gpm.

Charging and Letdown are in MANUAL control.

Current conditions:

RCS pressure is lowering rapidly.

RCS temperature is stable.

Charging and Letdown flow are stable in Manual control.

PRZR level remains off-scale high.

Which ONE of the following describes the procedure entry and initial action that will be required to mitigate this event?

AP-RCS.1, Reactor Coolant Leak; Isolate Letdown and control Charging in attempt to A. maintain PRZR pressure constant.

AP-RCS.1, Reactor Coolant Leak; Isolate Letdown and start one SI Pump to prevent a B. loss of RCS subcooling and maintain RCS inventory.

AP-RCS.4, Shutdown LOCA; Isolate Letdown and control Charging in attempt to C. maintain PRZR pressure constant.

AP-RCS.4, Shutdown LOCA; Isolate Letdown and start one SI Pump to prevent a loss D. of RCS subcooling and maintain RCS invEmtory.

Proposed Answer: C 184 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Incorrect. This procedure would be entered from higher modes. Plausible because a A. LOCA is in progress and actions are correct.

Incorrect. This procedure would be entered from higher modes. Plausible because a B. LOCA is in progress Correct. The crew would perform these actions in accordance with the guidance in AP C, RCS.4 Incorrect. SI pump operation would not be required as an initial action. Alot more 0, inventory would be lost prior to that action Technical Reference(s): AP-RCS.4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RAP18C, Obj 1.02, 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

185 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 006 A2.11 Importance Rating 4.4 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture of ECCS header Proposed Question: SRO Question # 90 Plant conditions:

The unit has tripped LOCA

RCS pressure - 1450 psig and stable.

RVLlS Water Level - 65% and stable.

Containment pressure - 8.5 psig and slowly rising.

CETs - 810°F and slowly rising.

RWST level - 70% and lowering.

. Containment sump 78 inches.

. No RCPs are running.

Containment radiation monitors are in alarm.

. All required ECCS systems are running.

, E~O, Reactor Trip or Safety Injection, is being performed.

1. I Which ONE of the following describes the impact on the SI system if a rupture were to occur on the inlet weld to MOV-878A, SI PUMP 'N DISCHARGE ISOL MOV TO LOOP 8 HOT LEG?

AND

2. Which procedure flowpath will be implemented upon transition from E-O?

1. SI Line to RCS Loop "8" - SI flow on FI-924 and SI pressure on PI-922 would remain stable.

A

2. E-1 to FR-C.2

1. SI Line to RCS Loop "8" - SI flow on FI-924 would rise and SI pressure on PI-922 8.

2. FR-C.2 to E-O

1. SI Line to RCS Loop "8" - SI flow on FI-924 and SI pressure on PI-922 would C. remain stable.

186 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination

2. FR-C.2 to E-O

1. SI Line to RCS Loop "B" - SI flow on FI-924 would rise and SI pressure on PI-922 would lower.

D.

2. E-1 to FR-C.2 Proposed Answer: D Explanation (Optional):

1. MOV-878A is a normally closed valve that gets SI flow from the "Afl SI pump and delivers it to the fiB" Hot leg. With the leak on the upstream side where MOV-878B also taps in will result in high flows and lowering pressure on the SI line to the "B" RCS loop.

2. Transition out of E-O to E-1 prior to start of CSFST monitoring. Once in E-1 CSFST A.

monitoring starts, then go to FR-C.2 then back to E-1, ES-1.2 and ES-1.3. ECA-1.1 will be entered if the leak can't be stopped. Any transition to ECA-1.1 will only happen after ES-1.3 is entered

1. This would be correct if MOV-878A was on a common line that went to the flA" RCS loop but it does not, it is on a common line that goes to the "B" RCS loop.

B.

2. Not monitoring CSFST in E-O yet, so transition would only happen when in E-1.

1. MOV-878A is a normally closed valve tihat gets SI flow from the "A" SI pump and delivers it to the "B" Hot leg. With the leak on the upstream side where MOV-8788 also taps in will result in high flows and lowering pressure on the SI line to the "B" RCS loop.

C. 2. Not monitoring CSFST in E-O yet, so transition would only happen when in E-1.

1. MOV-878A is a normally closed valve that gets SI flow from the "A" SI pump and delivers it to the "8" Hot leg. With the leak on the upstream side where MOV-8788 also taps in will result in high flows and lowering pressure on the SI line to the "8" RCS loop.

D.

2. Transition out of E-O to E-1 prior to start of CSFST monitoring. Once in E-1 CSFST monitoring starts, then go to FR-C.2 A-503.1 Technical Reference(s): 33013, Sheet 1 and 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None R2701C Obj 1.07 Learning Objective: (As available) 187 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Question Source: Bank # WTS16E5045 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Ginna 2008 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

188:

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KJA# 068 2.1.32 Importance Rating 2.6 Conduct of Operations: Ability to explain and apply all system limits and precautions.

Proposed Question: SRO Question # 91 Plant conditions:

A Liquid Waste Release has been in progress for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Reactor power has been reduced from 60% to 49% in the last 60 minutes due to a Circulating Water Pump vibration problem.

"At! Circulating Water Pump is being removed from service in accordance with T-8A, Startup and Shutdown of Circulatillg Water Pumps A and B.

Based upon these conditions, which ONE of the following describes the action(s) required?

Notify Chemistry to update release rate calculations or stop the release A.

Notify Chemistry to sample the RCS for Iodine and Gross Activity B.

Notify Chemistry to sample the RCS for Iodine and Gross Activity AND notify RP to C. update release rate calculations or stop the release Direct that the liquid Waste release flow rate be throttled to within the capacity of 1 D. Circulating Water Pump, and refer to the ODCM Proposed Answer: A Explanation (Optional):

Correct. If Circ Water Flow Rate is chang l9d, RP must recalculate release rate A.

Incorrect. Power changes >15% in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months require sample B.

Incorrect. Power changes >15% in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months require sample C.

Incorrect. Flow rate will be terminated, not throttled D.

189 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination T-8A Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTSI66157 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Ginna NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

190 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 011 2.4.21 Importalnce Rating 4.6 Emergency Procedures 1 Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Proposed Question: SRO Question # 92 Plant conditions:

A Steam Generator Tube Rupture is in progress.

Complications were observed while performing E-3, Steam Generator Tube Rupture.

The crew is performing ECA-3.2, SGTR with Loss of Reactor Coolant - Saturated Recovery Desired.

SI Pumps 'A' and 'C' are running

RCS subcooling is O°F.

RCP 'B' is running.

RCS fluid fraction is 55%.

PRZR level is off-scale low.

Which ONE of the following describes the condition of the INVENTORY CSF Status Tree, and the action that will be required?

INVENTORY CSF Status Tree is ...

GREEN because Safety Injection Pumps are operating. If the Status Tree turns yellow, A. the CRS would remain in ECA-3.2.

YELLOW because Pressurizer level is low. The CRS may perform FR-1.2, Response to B. Low Pressurizer Level, at his discretion.

YELLOW due to voiding in Reactor Vessel; the CRS may perform FR-1.3, Response to C. Voids in the Reactor Vessel, at his discretion.

YELLOW due to voiding in the Reactor Vessel; the CRS will NOT perform actions D. contained in FR-1.3 because they conflict with the ECA procedure in use.

Proposed Answer: A 19'1 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Explanation (Optional):

Correct. Yellow Path for Inventory will not exist if SI is in service. If Yellow path existed, A. then it would not be implemented due to conflict with ECA-3.2 Incorrect. Yellow Path does not exist with SI pumps running, and action would be B. normal for yellow conditions, but not for current conditions Incorrect. Plausible because there is a low water density, but with RCP running, FR-1.1 C. cannot be reached.

Incorrect. Plausible because there is a low water density, but with RCP running, FR-1.1 D. cannot be reached. However, actions are correct for this condition Technical Reference(s): FR-1.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RFRI3C Obj 2.01, 1.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

192:

5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KJA# 001 A2.19 Importance Rating 4.0 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS* and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Axial flux distribution Proposed Question: SRO Question # 93 Plant conditions:

The plant is operating at 86% power following a down power due to a Main Condenser leak.

Control Rod insertion was primarily used.

Bank D control rods are currently just above the Rod Insertion Limit.

Which ONE of the following describes (1) the direction that Axial Flux Difference (AFD) will trend, and (2) the action required if AFD exceeds the limit specified in the Core Operating Limits Report (COLR)?

AFD will trend in the ...

POSITIVE direction; restore AFD within 30 minutes or be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A.

POSITIVE direction: reduce reactor power to <50% within 30 minutes of exceeding the B. limit.

NEGATIVE direction; restore AFD within 30 minutes or be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C.

NEGATIVE direction; reduce reactor power to <50% within 30 minutes of exceeding the D. limit.

Proposed Answer: D Explanation (Optional):

Incorrect. AFD will trend in the negative direction because flux will trend toward the bottom of the core with rod insertion. Action is plausible because 30 minutes is part of A. the action statement, and the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months represents normal shutdown to Mode 3 action statements 193 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Plausible as in A. but action to restore is correct. Applicant may confuse B. positive versus negative if they misunderstand the calculation for AFD Incorrect. Right direction but wrong action. Plausible as described in Option A C.

Correct.

D.

Technical Reference(s): TS 3.2.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Needed Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Facility operating limitations in the technical specifications and their bases.

Comments:

Using a procedure number was redundant so I eliminated. This is SRO based upon the TS action 194 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 KIA # G1 2.1.25 Importance Rating 4.2 Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables, etc.

Proposed Question: SRO Question # 94 Plant conditions:

The plant was operating at 100% power when a reactor trip occurred on low pressurizer pressure.

"B" S/G Tube Rupture was diagnosed, and E-3, Steam Generator Tube Rupture, was entered.

RCS Cooldown and Depressurization is complete.

Plant conditions:

SG "S" level is 32% and rising.

SG "A" level is 52% and stable.

PRZR level is 37% and lowering.

Which ONE of the following describes the required operator action lAW E-3, and which ONE of the following procedures will subsequently be used for the ruptured SG Cooldown if radioactive release and contamination must be minimized?

(Reference Provided)

Depressurize RCS; ES-3.1, Post SGTR Cooldown Using Backfill A.

Depressurize RCS; ES-3.2, Post SGTR Cooldown Using Slowdown S.

Energize PRZR Heaters; ES-3.2, Post SGTR Cooldown Using Slowdown C.

Energize PRZR Heaters; ES-3.1, Post SGTR Cooldown Using Sackfill D.

Proposed Answer: A Explanation (Optional):

195 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Correct. Table shows with PRZR level between 20% and 50% and SG level rising, A. RCS must be depressurized.

Incorrect. Plausible because action is correct, but preferred procedure for conditions B. indicated is ES-3.1 Incorrect. Plausible because this action is in the column next to the required action. If SG level were lowering, this action would be taken. ES-3.2 would be used if there was C.

a reason NOT to use ES-3.1 for SG cooldown under these conditions.

Incorrect. Plausible because the correct procedure is indicated. Action would be D. correct if SG level was lowering.

. E-3 Rev 46 Step 36 Techmcal Reference(s): ' , (Attach if not previously provided)

Proposed References to be provided to applicants during examination: E-3, Step 36 REP03C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI66155 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamemtal Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

Modified from 2007 Ginna NRC exam. Changed levels and direction. Replaced first part of distracter A and B and made conditions so that there is a new correct answer.

196 5nl2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 2 KIA # G2 2.2.44 Importance Rating 4.4 Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Proposed Question: SRO Question # 95 Plant conditions:

A reactor trip and safety injection have occurred.

All equipment is operating as designed.

The crew is performing diagnostic actions of E-O, Reactor Trip Or Safety Injection.

Containment pressure is 19 psig and LOWERING.

RCS pressure is 1250 psig and STABLE.

RCS subcooling margin is 46°F and STABLE.

RWST level is 77% and dropping slowly

PRZR level is 4% and RISING.

All AFW pumps are running with 400 gpm total flow.

BOTH RCP's have been secured.

The Shift Manager intends to restart one RCP as soon as is practical.

Which ONE of the following procedures will be in use when the crew starts a reactor coolant pump?

ES-1.1, SI Termination A.

ES-1.2, Post-LOCA Cooldown And Depressurization B.

ES-1.3, Transfer to Cold Leg Recirculation C.

E-1, Loss Of Reactor Or Secondary Coolant D.

Proposed Answer: B Explanation (Optional):

Incorrect. Conditions do not exist to transition to ES-1.1. RCS pressure must be A.

greater than 1650 psig. Plausible because conditions are stable and parameters are 197 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination recovering.

Correct. From E-O, crew will transition to 1::-1. In E-1, transition will be made to ES-1.2, B. where an RCP may be started for RCS cooldown.

Incorrect. Plausible because a LOCA is in progress and RWST level is lowering.

However, RWST is not at a level where transition to ES-1.3 is imminent, so this C.

procedure will not be the next one entered.

Incorrect. Plausible because E-1 is the next procedure entered. However, E-1 stabilizes conditions and verifies equipment operation prior to transition to another D. procedure. Actions such as starting RCPs are performed in the subsequent procedure; in this case, ES-1.2 Technical Reference(s): ES-1.2, Rev 33, Step 11 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RES12C, Obj 2.01 Learning Objective: (As available)

Question Source: Bank # WTSI55134 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Beaver Valley Unit 2 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

198 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group '# 3 KJA# G3 2.3.5 Importance Rating 2.9 Radiation Control: Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO Question # 96 Plant conditions:

The plant is in Mode 6.

Core alterations are in progress.

A Containment Purge is being prepared in accordance with S-23.2.2, Containment Purge Procedure.

Which ONE of the following describes (1) an acceptable ventilation alignment, and (2) the radiation monitors required to be operable for these conditions?

(1 ) 2 Purge Supply Fans running; 1 Purge Exhaust Fan running; Containment A. (2) R-11 and R-12 MUST be operable for the release to proceed (1 ) 2 Purge Supply Fans running; 1 Purge Exhaust Fan running; 1 B. (2) R-12A MAY be used to satisfy the requirements for the release (1 ) 1 Purge Supply Fan running; 2 Purge Exhaust Fans running; Containment C. (2) R-11 and R-12 MUST be operable for the release to proceed (1 ) 1 Purge Supply Fan running; 2 Purge Exhaust Fans running; 1 D. (2) R-12A MAY be used to satisfy the requirements for the release Proposed Answer: C Explanation (Optional):

Incorrect. Not acceptable to have 2 supply and 1 exhaust fan. Negative pressure in containment is required. R-11 and R-12 are not required in Mode 6, the applicability is A.

for modes 1-4 and procedure allows use of R-12A for this evolution in Mode 5 or 6 199 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Not acceptable to have 2 supp1y and 1 exhaust fan. Negative pressure in B, containment is required. Correct applicabon of radiation monitor use Correct. Because refueling is in progress, R-11 and R-12 must be operable C

Incorrect. R-12A cannot satisfy requirements for release with refueling in progress D,

Technical Reference(s): S-23.2.2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI66188 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 4 Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

Modified by changing 2nd part 200 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier # 3 Group#: 4 KJA# G4 2.4.41 Importance Rating 4.6 Emergency Procedures I Plan: Knowledge of the emergency action level thresholds and classifications.

Proposed Question: SRO Question # 97 Initial Conditions:

A controlled shutdown was being performed due to a 200 GPD primary-to-secondary leak on "B" steam generator.

During the shutdown, SG tube leakage increased to approximately 150 GPM.

The reactor did NOT automatically trip when required.

Manual reactor trip from the MCB was successful.

The CO then reported all steam generator pre~ssures at 850 psig and lowering rapidly.

The crew manually initiated Safety Injection and Main Steam Isolation signals.

All ESF systems functioned properly.

Current Conditions:

Containment pressure is 0.6 psig and stable.

"8" steam generator pressure continues to lower rapidly.

Which ONE of the following describes the HIGHEST emergency classification for this event?

(Reference Provided)

Alert (1.1.1)

A.

Alert (3.1.2)

B.

Site Area Emergency (1.1.2)

C.

Site Area Emergency (3.2.2)

D.

Proposed Answer: D Explanation (Optional):

201 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect because the classification is not high enough, although conditions A. are met for this call because the reactor did not automatically trip Incorrect because the classification is not high enough, although conditions are met for this call. Leakage is higher and there is an uncontrolled depressurization of B.

SG secondary side Incorrect because the conditions are not me for this classification. There is no C. Subcriticality red path for this event Correct D.

Technical Reference(s): EPIP-1.0, EALs (Attach if not previously provided)

EPIP-1.0, EAL Proposed References to be provided to applicants during examination:

Tables Needed Learning Objective: (As available)

Question Source: Bank# WTSI6~~633 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Harris 2007 Question Cognitive Level: iVlemory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

202 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 3 KJA# G3 2.3.4

~-- ....- - - - - _...._

Importance Rating 3.7 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: SRO Question # 98 Plant conditions:

A General Emergency is in progress.

A worker is critically injured and unconscious in the RHR pit.

Extremely high radiation levels exist in the arE~a.

The Duty RP Tech is determining expected dose for two volunteers that will be assigned Search and Rescue responsibilities in accordance with EPIP 1.8, Search and Rescue.

Which ONE of the following describes the maximum dose guideline and highest approval required to allow the dose in accordance with EPIP 2-8, Voluntary Acceptance of Emergency Radiation Exposure?

25 Rem under all circumstances A.

>25 Rem only if the volunteers are fully aware of the risks involved B.

10 Rem under all circumstances C.

>10 Rem but less than or equal to 25 Rem only jf the volunteers are fully aware of the D. risks involved Proposed Answer: B Explanation (Optional):

Incorrect 25 Rem is the guideline but that dose may be exceeded by volunteers that A. have been briefed on the risks involved.

Correct.

B.

Incorrect, This value is for saving plant equipment/protecting valuable property C.

203 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Same as C but the remainder ()f this option is logical and consistent with B D.

Technical Reference(s): EPIP-2.8 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None RSC02C 17.00 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # WTSI66461 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or FundamEmtal Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 4 Radiation hazards that may arise during normal cmd abnormal situations, including maintenance activities and various contamination conditions.

Comments:

204 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 -

Group 'Ii! 4 KJA# G4 2.4.38 Importance Rating 4.4 Emergency Procedures 1 Plan: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Proposed Question: SRO Question # 99 Which ONE of the following Emergency Coordinator responsibilities may be delegated to another individual?

Classifying and declaring emergencies and requesting the formation of emergency A. teams.

Directing operations of emergency response organizations and providing Off-Site B. Protective Action Recommendations.

Initiating the implementation of on-site protective actions and directing operations of C. emergency response organizations.

Contacting Off-Site authorities by telephone in accordance with EPIP-1.5.

D.

Proposed Answer: D Explanation (Optional):

Incorrect. Classification is a non-delegable responsibility of the EC. Plausible because A. once the ERO is staffed, the TSC provides a significant amount of input to the EC.

Incorrect. PARs are responsibility of the EC. Plausible because once the ERO is B. staffed, the TSC provides a significant amount of input to the EC.

Incorrect. On-site actions are directed by EC. Plausible because on-site organizations such as the OSC are directed by their own coordinators, but protective actions are C.

responsibility of the EC.

Correct.

D.

Technical Reference(s): EPIP-1.5 (Attach if not previously provided) 517/2010 Rev Draft

Ginna 2010 NRC Written Examination Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTSI55219 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: Callaway 2005 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 1 Conditions and limitations in the facility license Comments:

206 5/7/201 0 Rev Draft

Ginna 2010 NRC Written Examination Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 1 KJA# G1 2.1.42 Importance Rating 3.4 Conduct of Operations: Knowledge of new and spent fuel movement procedures.

Proposed Question: SRO Question # 100 Plant conditions:

The plant is in Mode 6.

Refueling is in progress.

RF-301, Refueling Operations (Offload, Shuffle, Reload) is being performed for a core offload.

A fuel assembly is in transit from its position in the core to the Upender.

Refueling Cavity level begins to lower rapidly.

Sump A level is rising rapidly.

The Refueling SRO evacuates Containment and notifies the Control Room.

Which ONE of the following describes additional action that will be required in accordance with RF-601, Fuel Handling Accident Instructions?

Return the fuel assembly to its position in the core; return the Fuel Transfer Cart to the A. Spent Fuel Pit.

Return the fuel assembly to its position in the core; Ensure the transfer cart is in the B. Refueling Cavity.

Place the fuel assembly in the emergency location in the fuel transfer slot and lower to C" the bottom of the slot area; Ensure the transfer cart is in the Refueling Cavity.

Place the fuel assembly in the emergency location in the fuel transfer slot and lower to D. the bottom of the slot area; return the Fuel Transfer Cart to the Spent Fuel Pit.

Proposed Answer: D Explanation (Optional):

Incorrect. Would place fuel assembly here if it was over its core position, but not in A. transit to the upender.

207 5/7/2010 Rev Draft

Ginna 2010 NRC Written Examination Incorrect. Would place fuel assembly here if it was over its core position, but not in transit to the upender. Also, Transfer cart belongs in SFP so that the transfer tube gate B.

valve can be closed.

Incorrect. Correct location for fuel assembly but Transfer cart belongs in SFP so that C. the transfer tube gate valve can be closed.

Correct.

D.

Technical Reference(s): RF-601, Rev 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None NIA Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 7 Fuel handling facilities and procedures.

Comments:

208 5/7/2010 Rev Draft

v t e Letter Sequence Draft Other
TAC:U01766, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML102320373, ML102320386, ML102320393, ML102320408, ML102520045, ML102520054, ML102710006
MONTHYEAR2010-05-05 [Table View]

ML101720560

Text

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