Text

Farley, Nov/Dec 2007 Exam 05000348-07-301 & 05000364-07-301- Final Combined Ro/Sro Written Exam with Answers, Kas, & References
	ML080280360
Person / Time
Site:	Farley
Issue date:	01/28/2008
From:	; NRC/RGN-II
To:	;
References
	50-348/07-301, 50-364/07-301
	Download: ML080280360 (260)
	v • d • e

Final Submittal (Blue Paper)COMBINED RO/SRO WRITTEN EXAM WITH KAS, ANSWERS, REFERENCES,

QUESTIONS REPORT for 25 SRO Questions 1.006 A2.IO 004 Unit 1 is in Mode 1.Chemistry has provided sample results for boron concentration

of1A and1B Accumulators

with the following results:*1A

boron concentration

is 2350 ppm.*1B Accumulator

boron concentration

is 2198 ppm.Which ONE of the following describes the impact of this condition;

and the action required in accordance

with Technical Specifications

and SOP-8.0, Safety Injection System-Accumulators?

A.*Ability to maintain subcriticality

after an accident is reduced;*Drain and fill the1A Accumulator

to lower boron concentration.

B.*Ability to maintain minimum boron precipitation

time is reduced;*Drain and fill the1A Accumulator

to lower boron concentration.*Ability to maintain subcriticality

after an accident is reduced;*Feed and bleed the1B Accumulator

to raise boron concentration.

D.*Ability to maintain minimum boron precipitation

time is reduced;*Feed and bleed the1B Accumulator

to raise boron concentration.

Monday, January 14, 20082:43:54

PM 1

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)2 requirements

for SRO level question.A.is incorrect;

111 All Accumulator

boron concentration

is within spec.SOP-8.0 does provide guidance that fills and/or drains the accumulators

in a separate section, but not to raise or lower the Boron concentration.

Step 4.1 fills the accumulators

but does not address TS requirements

for level and pressure while filling.Also does not address boron concentration

unless a 12%level change is made, so sampling is not required due to the fill.Step 4.2 provides guidance to lower Accumulator

level but not to lower boron C.Appendix 2 is provided expressly to raise boron C.to>2300 PPM B.incorrect, 111 All Accumulator

boron concentration

is within spec.C.Correct.TS 3.5.4 basis states that boron concentration

of the RWST is designed to ensure subcriticality

is maintained

with uncontrolled

cooldown coincident

with most reactive rod stuck fully out.For a large break LOCA analysis, the minimum water volume limit of 321 ,000 gallons and the lower boron concentration

limit of 2300 ppm are used to compute the post LOCA sump boron concentration

necessary to assu're subcriticality.

The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.Within the same bases the following is found and may cause the applicant to choose the precipitation

idea which is what bounds the upper limit.A water volume of 506,600 gallons and the upper limit on boron concentration

of 2500 ppm are used to determine the maximum allowable time to switch to hot leg recirculation

following a LOCA..The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation

in the core following the accident.SOP-B.O, APP 2, FEED AND BLEED OF ACCUMULATOR

1A (18, 1C)TO RAISE BORON CONCENTRA TION>2300 PPM would be used to raise boron concentration

SOP-8.0 CAUTION: Accumulator

boron concentration

must be maintained

between 2200 and 2500 ppm;the intent of this appendix is to raise accumulator

boron concentration>

2300 ppm.D.Incorrect.

Basis is incorrect Monday, January 14, 2008 2:43:54 PM 2

QUESTIONS REPORT for 25 SRO Questions 006 A2.1 0 Emergency Core Cooling Ability to (a)predict the impacts following malfunctions

or operations

on the and (b)based on those predictions, use procedures

to correct, control, or mitigate the consequences

of those malfunctions

or operations:, Low boron concentration

in SIS.Question Number: Tier 2 Group 1 86 Importance

Rating: 3.9 Technical Reference:

TS 83.5.1/3.5.2

and basis, SOP-8, SOP-2.3, Proposed references

to be provided to applicants

d,uring examination:

None Learning Objective:

OPS521 02801 10 CFR Part 55 Content: 43.2 Comments: fixed per FJE comments MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: MODIFIED LOWER SRO GTO Version:0123456789

Answer: CAD C B DCAAD Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:54 PM 3

QUESTIONS REPORT for 25SRO Questions 2.009"EA2.01

005 Given the following:*A small break LOCA has occurred on Unit 1.*The crew is performing

EEP-1.0, Loss of Reactor or Secondary Coolant, step 7 that checks SI Termination

Criteria.*Containment

pressure is 3.6 psig.*Subcboled Margin Monitor value is 14°F in CETC mode.*RCS pressure is 1100 psig and stable.*Pressurizer

level is 20%and rising slowly.Which ONE of the following correctly describes the procedure flow path when evaluating

step 7, Check SI Termination of EEP-1.0, and the reason?The crew wilL..A':'remain in EEP-1 because RCS subcooling

is too low.B.remain in EEP-1 because RCS pressure is NOT rising.C.remain in EEP-1 because pressurizer

level is too low.D.go to ESP-1.1, SI Termination, because all SI Termination

criteria are met.Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.A:Correct.

Subcooling

does not meet requirements.

Check SUB COOL"ED MARGIN MONITOR indication

-GREATER THAN 16F{45F}SUBCOOLED IN CETC MODE.B: Incorrect.

pressure may be stable or rising.7.3 Check RCS pressure-STABLE OR RISING.C: Incorrect.

level meets requirement.

7.4 Check pressurizer

level-GREATER THAN 13%{43%}.D: Incorrect.

Subcooling

must be raised by cooldown or pressure increase NOTE: For certain break sizes, SI termination

criteria may be met due to injection flow exceeding mass flow out of the break.Step 7.5 is not intended to terminate SI a known I DCA exists 7.5 IF all SI termination

criteria satisfied, THEN go to FNP-1-ESP-1.1, SI TERMINATION.

Monday, January 14, 20082:43:54

PM 4

QUESTIONS REPORT for 25 SRO Questions 009 EA2.01 009 small break LOCA Ability to determine or interpret the following as they

to a small break LOCA: Actions to be taken, based on RCS temperature

and

saturated and superheated

Question Number: 76 Tier 1 Group 1 Importance

Rating: SRO 4.8 Technical Reference:

EEP-1, Step 7 Proposed references

to be provided to applicants

during examinatio'n:

'None Learning Objective:

OPS52301 B09 10 CFR Part 55 Content: 43.5 Scramble Range:A-D NEW illGHER SRO GTO Source:

Level: Job Position: reviewed: Comments: fixed per FJE comments 009 smaH break LOCA AbiHty to determine or interpret the following as they apply to a smaU break LOCA: Actions to be taken, based on ReS temperature

and

saturated and superheated

MCS Time: 1 Points: 1.00 Version:0123456789 Answer: AAAACBCC CA Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:54 PM 5

QUESTIONS REPORT for 25 SRO Questions 3.011 A2.12 001 Given the following:*A reactor trip has occurred and1C RCP is the only operating RCP.*Auxiliary Spray has been placed in service lAW ESP-0.1, Reactor Trip Response, to control and reduce RCS pressure.*The plant is preparing for a cooldown lAW UOP-2.2, Shutdown of Unit from Hot Standby to Cold Shutdown, with the following parameters:

-RCS pressure is 2230 psig.-RCS temperature

is 537°F.-Pressurizer

level is 23%.*The crew is at step 5.2 to begin raising pressurizer

level to 55%.Which ONE of the following correctly describes the limit associated

with cooling down the pressurizer, and while raising pressurizer

level, the method used to prevent thermal stratification

in accordance

with UOP-2.2?A.*The temperature

difference

between the pressurizer

steam space and charging water must not exceed 320°F;*A pressurizer-

insurge must occur during the pressurizer

cooldown.B!'*The temperature

difference

between the pressurizer

steam*space and charging water must not exceed 320°F;*A pressurizer

outsurge must occur during the pressurizer

cooldown.C.*The pressurizer

cooldown rate must not exceed 100°F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period;*A pressurizer

insurge must occur during the pressurizer

cooldown.D.*The pressurizer

cooldown rate must not exceed 100°F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period;*A pressurizer

outsurge must occur during the pressurizer

cooldown.Monday, January 14, 2008 2:43:54 PM 6

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)2 and 5 requirements

for SRO level question.5.4 IFauxiliaryspray

is in operation, THEN on Data Sheet 1 record the time and the differential

temperature

between regenerative

heat exchanger outlet charging TI-123 and pressurizer

vapor space TI-454.Ensure that the differential

temperature

does not exceed 320 o P.A incorrect;

outsurge is required;correct parameters

used to determine differential

temperature.

B correct;step 5.2 begins raising the level to 55%and there is a caution prior to and a note after that step to limit delta T and level<63.5%and an outsurge is to be maintained.

5.35.1 Verify Delta T between pressurizer

and charging<320°F.5.35.1.1 Commence recording delta T in PNP-1-STP-1.0, OPERATIONS

DAILY AND SHIFf SURVEILLAN'CE

REQUIREMENTS, misc.section every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

P&L ofUOP-2.2 3.3.3 Do not exceed a 200 0 PIhr pressurizer

cooldown rate.3.3.4 The temperature

differential

between the pressurizer

and the RCS must not exceed 320 0 P.The pressurizer

liquid, surge line and loop B hot leg temperatures

should be monitored to ensure that a pressurizer

outsurge is taking place whenever the pressurizer

is being cooled or filled.This will prevent thermal stratification

from taking place.A pressurizer

outsurge is indicated by surge line temperature

approximately

equal to pressurizer

liquid temperature

and greater than"B" Hot Leg temperature.

C incorrect;

PRZR cooldown rate is 200°F CAUTION: PRZR cool down rate must be limited to<200°Flhr.P&L 3.3.2 Do not exceed RCS cooldown rate specified in PTLR section 2.0, Operating Limits.The maximum cooldown rate is 100 0 P in anyone hour period.D incorrect;

PRZR cooldown rate limit is 200°F The following flow path could cause entry into UOP-2.2 with aux spray in service.If 1 C RCP is the only running pump after a Rx trip aux spray would be put on service as long as Letdown is in service..ESP-O.1 will send you to UOP-2.3 which could send you to UOP-2.2.In this scenario the P&Ls of UOP-2.2 would apply and the limits of 320°F and 200°F would be applica,.LJ..bu:;Ie::....------------------------

Monday, January 14, 20082:43:54

PM 7

QUESTIONS REPORT.for 25 SRO Questions 011 Pressurizer

control 12 Ability to (a)predict the impacts of the following malfunctions

or operations

on the PZR LCS;and (b)bas on those predictions, use procedures

to correct}control, or mitigate the consequences

of Operation of auxiliary spray Question Number: 91 Tier 2 Group 2 Importance

Rating: 3.3 Technical Reference:

UOP-2.2, TRM B 13.4, TS B 3.4.3, STP-35.0 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52510E06

10 CFR Part 55 Content: 43.5 Comments: Scramble Range:A-D NEW IDGHER SRO GO fixed per FJE comments and added some verbiage to stem to clarify where procedurally

you are at.Otherwise the question does not make sense.MCS Time: 1 Points: 1.00 Version: a123456789 Answer: BDCCBB DCAA Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:43:54

PM 8

QUESTIONS REPORT for 25 SRO Questions 4.016 G2.4.31 001 Unit 1 is at 95%power when the following occurred:*LT-474, 1A SG NR LVL, was declared INOPERABLE

and the channel placed in trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago lAW Tech Specs 3.3.1, Reactor Trip System (RTS)Instrumentation

and 3.3.2, Engineered

Safety Feature Actuation System (ESFAS)Instrumentation.

Subsequently, the card power supply for LT-475, 1A SG NR LVL, failed.*The following MCB annunciators

are in alarm:-EC1, PROC CAB PWR FAILURE-JA 1,1A SG LO L VL-JC1, 1A SG LO-LO LVL ALERT-JD1, 1A SG HI-HI LVL Alert-JF1, 1A SG LVL DEV Which one of the following is the appropriate

procedure(s)

and actions to be taken for this condition?*Enter EEP-O, Reactor Trip and Safety Injection, and then go to ESP-O.1, Reactor Trip Response.*Control AFW flow>395 gpm until at least one SG is>31%NR level.*Maintain SG levels 31%-65°/6 when conditions

permit.B.*Enter EEP-O, Reactor Trip and Safety Injection, and then go to ESP-O.1, Reactor Trip Response.Implement FRP-H.3, Response to Steam Generator High Level, in conjunction

with ESP-O.1.*Verify BOTH SGFPs are tripped and Main Feedwater and AFW is isolated to ALL SGs.*When ALL SGs are<65%NR level, then maintain SG levels 31-65%.C.*Enter AOP-1 00, Instrumentation

Malfunction.

Control1A SG FRV as required to raise1A SG level to 65%.*Return SGWLC to Automatic when conditions

permit.D.*Enter AOP-1 90, Instrumentation

Malfunction.

C*ontrol1A SG FRV as required to lower1A SG level to 650/0.*Reference T.S.3.3.1 and 3.3.2, and notify the Shift Manager.Monday, January 14, 20082:43:54

PM 9

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.A: Correct.With 2 L T$on one SG less than 28%, a Rx trip w,ill be generated and an autostart of MDAFWPs is generated.

The appropriate

path is EEP-O to ESP-0.1.No SI signal is generated.

B: Incorrect.

With2L Ts on one SG less than 28%, a Rx trip will be generated and an autostart of MDAFWPs is generated.

If a candidate thought that a card failure would cause a high level and the card caused a high level condition which is plausible in that JD1 is in alarm, then this would be an appropriate

action to take.Since the failures listed cause a low level alarm and condition, a Rx trip occurs and ESP-0.1 actions taken.C: Incorrect.

L T failure meets entry conditions

for AOP-1 00 and subsequent

required actions, however, is the incorrect procedure to enter based upon ERG entry requirement

D: Incorrect.

L T failure meets entry conditions

for AOP-1 00 and subsequent

required actions, however, is the incorrect procedure to enter based upon ERG entry requirement

With 2 L Ts on one SG less than 280/0, a Rx trip will be generated and an autostart of MDAFWPs is generated.

The appropriate

path is EEP-O to ESP-0.1.Monday, January 14, 2008 2:43:54 PM 10

QUESTIONS REPORT for 25 SRO Questions 016 Non-Nuclear

Instrurnentation

System

2.4.31 Emergency Procedures

I Plan Knowledge of annunciators

alarms and indications, and use of the response instructions.

Question Number: 92 Tier 2 Group 2 Importance

Rating: 3.4 Technical Reference:

EEP-O and AOP-100 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content:.43.2 Scramble Range:A-D FARLEY MODIFIED HIGHER SRO GO Comments: changed to a different bank question and modified it due to many technical issues.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: AC AAB AABBC Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:55 PM 11

QUESTIONS REPORT for 25 SRO Questions 5.022 G2.1.14 002 Given the following:

Unit 2 is at 100%power*The following alarms are received:*HA1, PRZR LVL HI RX TRIP ALERT*HA2, PRZR LVL DEV HI B/U HTRS ON*HB1, PRZR LVL HI*DE1, REGEN HX L TDN FLOW DISCH TEMP HI*EA2, CHG HDR FLOW HI-LO*Actual Pressurizer

level is 46%and trending DOWN.*VCT level is 43%and trending UP.*RCS temperature

and pressure are stable.Which ONE of the following describes the procedure entry required, and a required notification

for the event in progress?A.*Enter AOP-1 00, Instrument

Malfunction;

Notify the Shift Manager to initiatea1 hour report lAW EIP-8.0,Non-Emergency

Notifications.*Enter AOP-1 00, Instrumentation

Malfunction;

Initiate a CR.and notify the Work Week Coordinator.

C.*Enter AOP-1.0, RCS Leakage;*Notify the Shift Manager to initiatea1 hour report lAW EIP-8.0, Non-Emergency

Notifications

..D.*Enter AOP-1.0, RCS Leakage;*Initiate a CR and notify the Work Week Coordinator.

Monday, January 14,20082:43:55

PM 12

QUESTIONS REPORT for 25 8RO Questions Meets 10 CFR 55.43 (b)5 requirements

for 8RO level question.A.Incorrect.

Credible due to correct procedure for a failed LT.This question gives the indicatio'ns

for a failedLT and lAW AOP-1 00 the WWC would be notified and the crew would initiate a CR.This is nota1 hour report.AOP entry is found in EIP-8.0 under 20.0 Additional

Corporate Duty Manager Notifications

20.9 Events requiring entry into the EOPs or AOPs The time.required to notify the CDM is not defined in this section but would be done as soon as reasonably

possible.The point of the above is that a notification

is made lAW EI P-8 and the candidate would have to know that the notification

is nota1 hour notification.

B.Correct.Due to a failed level instrument, AOP-100 would be entered.The following people need to be notified, both the 8M and the WWC.The reason for the 8M in the other distracters

is not correct since the E-plan does not need to be implemented

for this condition.

C.Incorrect.

Credible due to PZR level trend, and this is nota1 hour report.AOP entry is found in EIP-8.0 under 20.0 Additional

Corporate Duty Manager Notifications

20.9 Events requiring entry into the EOPs or AOPs D.Incorrect.

Incorrect pro.cedure

and incorrect person to notify for an AOP-1 00 entry, but correct for an AOP-1 00 entry.LT 459 has failed high, BU heaters will be on, Charging flow will go to a minimum value, and due to letdown still on service with charging at a minimum, DE1 will be in alarm.Due to this failure, Pressurizer

level is trending DOWN and VCT level is trending up due to the charging flow to a minimum.AOP-100 actions 8 Notify the Shift Manager.9 Submit a Condition Report for the failed level channel, and notify the Work Week Coordinator (Maintenance

ATL on backshifts)

of the Condition Report.Monday, January 14,2008 2:43:55 PM 13

QUESTIONS REPORT for 25 SRO Questions 022 G2.1.14 APE 022 Loss of Reactor coolant makeup (this affects charging system)Conduct of Operations:

Knowledge of system status criteria which require the notification

of plant personnel..

Question Number: 77 Tier 1 Group 1 Importance

Rating: SRO 3.3 Technical Reference:

AOP-100 and above ARPs Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5 Scramble Range:A-D NEW HIGHER SRO GTO Comments: changed to meet KA for notification

requirements

and procedural

entry.MCS Time: 1 Points: 1.00 Version: a 123456789 Answer: BAB B CADDAA Source if Banlc Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:55 PM 14

QUESTIONS REPORT for 25 SRO Questions 6.025 AA2.04 004 Given the*following:

Unit 1 is in Mode 5 at 120°F with SG manways open to remove nozzle da-ms after core reload.*'A'Train RHR is in service with'B'Train RHR in standby.*The following alarms are received:*EC5-RCS LVL HI-LO*BE5-BOP PANELS ALARM*LE2-(BOP)1A RHR PUMP RM SUMP LVL HI-HI OR TRBL The operator observes the following indications:

RCS level 123'1"andfalling.

The leak is estimated to be 25 gpm.*Both 1A RHR pump room sump pumps are running.*1A RHR pump flow, amps, and discharge pressure are stable.Which ONE of the following correctly describes the*procedure

required to be entered, what the procedure will accomplish, and how to apply Technical Specification

3.4.13.for the conditions

above?*AOP-12.0, Res'idual Heat Removal System Malfunction, is required to be entered and will id.entify the location of the leak and WILL isolate the leak.*LCO 3.4.13, RCS Operational

LEAKAGE, is NOT applicable

in Mode 5.B.*AOP-12.0, Residual Heat Removal System.Malfunction, is required to be entered and will identify the location of the leak but will NOT isolate the leak.*LCO 3.4.13, RCS Operational

LEAKAGE, is NOT applicable

in Mode 5.C.*AOP-1.0, RCS Leakage, is required to be entered and will identify the location of the leak and WILL isolate the leak.*Enter LCO 3.4.13, RCSOperationalLEAKAGE, for IDENTIFIED

leakage.D.*AOP-1.0, RCS Leakage, is required to be entered and will identify the location of the leak but will NOT isolate the leak.*Enter LCO 3.4.13, RCS Operational

LEAKAGE, for IDENTIFIED

leakage.Monday, January 14, 2008 2:43:55 PM 15

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.A.CorrectlAW EC5, validation

of the low level alarm would send the operator to AOP-12.This procedure applies to mode 4,5 6.

will identify where the leak is and aid in isolting the leak at step 5.EC5 setpoint PROBABLE CAUSE 1.Improper RCS level control 2.Im.proper valve lineup 3.RCS leakage ACTION 2.IF low level condition exists, THEN monitor RHR pump(s)for evidence of cavitation

and if necessary, THEN refer to FNP-1-AOP-12.0, RHR SYSTEM MALFUNCTION.

lAW LE2, the operator could secure the running-pump and then go to SOP-7.0, but this is not an option.The leak is outside ctmt due to the running sump pumps which shows the leak to be in the1A RHR pump room.The TS for RCS leakage is NOT applicable

in modes 5 or 6 but is applicable

in modes 1-4.B.Incorrect-AOP-12

will isolate the leak and this question says AOP-12 will not isolate the leak.C.Incorrect-

wrong procedure Since AOP-1.0 is not applicable

in-this mode.AOP-1 is only applicable

in Mode1-3 The TS for ReS leakage is NOT applicable

in modes 5 or 6 but is applicable

in modes 1-4.If.in mode 1-4 then this would be correct.D.incorrect.

wrong procedure.

AOP-1 only applicable

in Mode1-3 Monday, January 14, 2008 2:43:55 PM 16

QUESTIONS REPORT for 25 SRO Questions 025 Loss of the RH RS 2x04 Ability to deterrnine

and interpret following as they apply to the Loss Residual Heat Removal System: location and isolability

of leaks Question Number: 78 Tier 1 Group 1 Importance

Rating: SRO 3.6 Technical Reference:

AOP-12.0 and AOP-1.0 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 43.5 Scramble Range:A-D BANK IDGHER SRO GO Comments: Fixed per FJE comment to include the location of the leak to meet the KA and then the procedural

guidance to be entered to meet the SRO portion of the question.MCS Time: 1 Points: '1.00 Version: a123456789 Answer: ADABCB DBAD Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level:.Job Position: reviewed: Monday, January 14, 2008 2:43:55 PM 17

QUESTIONS REPORT for 25 SRO Questions 7.029 G2.1.33 002 Given the following:

.*The plant was at 100%power.*At 1 000, Both Reactor Trip Breakers were declared INOPERABLE.

SSPS has been determined

to be operable.*The crew immediately

initiated a plant shutdown.*At 1025, a reactor trip signal was generated.

The Reactor Trip Breakers did NOT open.*1A CRDM MG set breaker did NOT open.At 1030, ALL Reactor Trip and Bypass Breakers were verified open.UOP-2.3, Shutdown of Unit following Reactor Trip, has been entered.Which ONE of the following correctly describes the mode the unit is allowed to remain in or must be placed in lAW Technical Specifications

and the reason?*The plant can remain in Mode 3 indefinitely;

since the RTBs are now open and rod control is no longer capable of rod withdrawal.

B.*The plant must proceed to Mode 4, but can remain in Mode 4 indefinitely;

since the RTBs are now open and rod control is no longer capable of rod withdrawal..

C.*The plant can remain in Mode 3 for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , but must be in Mode 4 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months and Mode 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months ;*sfnce BOTH RTBs are inoperable

and Technical Specification

3.0.3 is in effect.D.*The plant can remain in Mode 3 for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months while trying to repair one RTB, but if one RTB cannot be fixed" the plant must be in Mode 4 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months and Mode 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months ;*since BOTH RTBs are inoperable

and Technical Specification

3.0.3 is in effect.Monday, January 14, 20082:43:55

PM 18

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)2 requirements

for SRO level question due to the application

of 3.0.3 and knowledge that 3.0.3 applies in this case and how it applies, specifically.

A.Correct, 3.0.3 no longer applies since the RTBs are opened.The (a)With RTBs closed and Rod Control System capable of rod withdrawal

for modes 3, 4, 5 show that when the RTBs are open, the spec no longer applies.B.incorrect

-but plausible because 3.0.3 applies until the RTBs are open and has to be evaluated.

C.incorrect.

The plant can remain in Mode 3 indefinitely

so the plant can remain in mode 3 for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , however the plant does not have.to go to mode 4 and 3.0.3 is no longer in effect due to the RTBs being open.D is incorrect-The plant can remain in Mode 3 indefinitely

so the plant can remain in mode 3 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , however the plant does not have to go to mode 4 and 3.0.3 is no longer in effect due to the RTBsbeing open.3.3.1 18.Reactor Trip Breakers (j)1 ,2 2 trains 3 (a),4 (a),5 (a)2 trains*R, V C,V (a)WithRTBsclosed

and Rod Control System capable of rod withdrawal.

v.Two RTS trains inoperable.

V.1 Enter LCO 3.0.3.mediately LCO 3.0.3 When an LCO is not met and the associated

ACTIONS are not met, an associated

ACTION is not provided, or if directed by the associated

ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable.

Action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in: a.MODE 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ;b.MODE 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ;and c.MODE 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .Monday, January 14, 20082:43:55

PM 19

QUESTIONS REPORT for 25 SRO Questions EPE 029AT G2x 1.33 Conduct of Operations:

AbiUty to recognize indications

for system operating parameters

which are entry level conditions

for technical specifications

Question Number: 79 Tier 1 Group 1 Importance

Rating: SRO 4.0 Technical Reference:

TS 3.3.1 and 3.0.3 Proposed references

to be provided to applicants

during examination:

Learning Objective:

10 CFR Part 55 Content: 43.2 no reference Scramble Range:A-D NEW IDGHER SRO GTO Comments: This was rewritten to incorporate

an ATWT into the stem and in such a way as to make TS entry a requirement

to meet.Just entering Mode 3 on an ATWT event with 3.0.3 in effect would require the plant to be in mode 5, 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months after the RTBs were found to be inoperable

as long as they can not be opened.Since they are opened in the stem, and as expected per procedure, then the plant is no longer bound to be in mode 4 or 5 and no time limit applies.MCS Time: 1 Points: 1.00 Version: a123456789 Answer: AADC CAAAB A Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Mon,day, January 14,2008 2:43:55 PM 20

QUESTIONS REPORT for 25 SRO Questions 8.035 A2.06 003 Given the following:*A small break LOCA has occurred on Unit 2.*RCS pressure is 1550 psig and lowering slowly.*SG pressu"res

are 1000 psig and stable.*Total AFW flow is 500gpm.*SG narrow range levels are 5%and rising slowly.*PRZR level is off scale low.*Containment

pressure is 3 psig and stable.*The crew is in EEP-1, Loss of Reactor or Secondary Coolant.Which ONE of the following describes the correct sequence of actiens the crew must use to cool down the RCS in order to place RHR in service?A.*Cooldown in accordance

with EEP-1 until RCS pressure is less than SG pressure;*Go to ESP-1.2, Post LOCA Cooldown and Depressurization, and cooldown to RHR entry conditions.*NO cooldown will be performed in EEP-1;*Go to ESP-1.2, Post LOCA Cooldown and Depressurization, and cooldown to RHR entry conditions.

C.*Cooldown in accordance

with EEP-1 until RCS pressure is less than SG pressure;*Go to ESP-1.2, Post LOCA Cooldown and Depressurization, and cooldowntoHot Standby;*Then go to UOP-2.2, Shutdown of Unit from Hot Standby to Cold Shutdown, and cooldown to RHR entry conditions.

D.*NO cooldown will be performed in EEP-1;-*Go to ESP-1.2, Post LOCA Cooldown and Depressurization, and cooldowntoHot Standby;*Then go to UOP-2.2, Shutdown of Unit from Hot Standby to Cold Shutdown, and cooldown to RHR entry conditions.

Monday, January 14, 2008 2:43:55 PM 21

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.A.Incorrect.

A cooldown is not done in EEP-1.SGWL is maintained

and the procedure transitions

to either ESP-1.2 or ESP-1.3 for a small break LOCA.B.Correct.Entry to ESP-1.2 is required and cooldown to RHR entry is directed in ESP-1.2.C.Incorrect.

A cooldown is not done in EEP-1.SGWL is maintained

and the procedure transitions

to either ESP-1.2 or ESP-1.3 for a small break LOCA.ESP-1.2 does not senq you to UOP-2.2.see below discussion.

D.Incorrect.

ESP-1.2"cools

the plant down to RHR entry conditions, not Hot Standby conditions.

Plausibility-

While UOP-2.2 is not required or directed by EOPs, it could be used in part to recover the plant from this point.It might be directed by the TSC staff to enter at a step that would consider getting the plant into a condition in which the appropriate

procedure would be used.Since UOP-2.2 is cooldown from Hot Stby and the unit will be on RHR with temp<200°F, this UOP would not be appropriate

at certain steps and sections being used and others being N/A ed.However, The TSC would Evaluate long term plant status lAW ESP-1.2 and then look for an appropriate

procedure to use to clean up the plant and get back to operational

status.They could decide many different and/or appropriate

procedures

depending on the plant conditions.

What is entirely incorrrect

with this statement is that ESP-1.2 would not cooldown to Hot Standby, it actually goes all the way down to 200°F before the appropriate

procedure would be addressed.

035 SG system.06 Ability to (a)predict the impactsofthe following mal-functions

or operations

on the S;and (b)base.d on those predictions, use procedures

to correct,

or mitigate the consequences

of those malfunctions

or operations:

Small break LOCA Question Number: Tier 2 Group 2 93 Importance

Rating: 4.6 Technical Reference:

EEP-1 and

Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 43.5 Comments: fixed per FJE comments Monday, January 14, 20082:43:56

PM 22

FARLEY NO Scramble Range:A-D QUESTIONS REPORT for 25 SRO Questions 1.00 Version:0123456789

Answer: BDCCCBB CAA Source if Banle Difficulty:

Plant: Previous 2 NRC exams: NEW HIGHER SRO GTO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:56 PM 23

QUESTIONS REPORT for 25 SRO Questions 9.037 G2.4.11 001 The following Unit 1 conditions*exist

while at 10%power: The Shift Radiochemist

reports the following:*1A SG Primary to Secondary Leakage=148 gpd*1B SG Primary to Secondary Leakage*=185 gpd*1C SG Primary to Secondary Leakage=134 gpd The OATC reports the following:

Pressurizer

PORV-445A is leaking to the PRT at 2.2 gpm.Which ONE of the following correctly describes the procedure that must be entered and the required action and completion

time lAW Technical Specification

LCO 3.4.13, RCS Operational

LEAKAGE?A.*Enter AOP-2.0, Steam Generator Tube Leakage.*Reduce leakage to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .*EnterAOP-2.0, Steam Generator Tube Leakage.*Be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .C.*Enter AOP-1.0, RCS Leakage.*Reduce leakage to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .D.*EnterAOP-1.0, RCS Leakage.*Be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Monday, January 14,20082:43:56

PM 24

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.ANSWER/DISTRACTOR

ANALYSIS A.Incorrect.

All leakage listed in stem is identified

leakage.Plausible because applicant may think that PORV leakage is unidentified

and at 2.2 gpm, this would exceed the limit.AOP-2 is correct but the actions are not correct.B.Correct.150 gpd is the TS limit.bases 3.4.13-3"The ReS operational

primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience

with SG tube degradation

mechanisms

that result in tube leakage.The operational

leakage rate criterion in conjunction

with the implementation

of the Steam Generator Program is an effective measure for minimizing

the frequency of steam generator tube ruptures.AOP-2 is the correct procedure to enter for the above conditions

and the TS is a immediately

go to mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .C.Incorrect.

All leakage listed in stem is identified

leakage.Plausible for same reason as A above and AOP-1 could be entered but not for the reasons given.The action is not correct for a tube leak.D.Incorrect.

Incorrect procedure.

Plausible because it is partially correct in that the action is correct.REFERENCES

1.Technical Specification

3.4.13, Operational

Leakage.2.Technical Specification

3.4.13 Basis.AOP-2 lesson plan The guidance is based on anticipating

a tube rupture and is more restrictive

than the required actions of Technical Specifications.

If the steam generator leak rate is determined

to be greater than 150 gpd in any steam generator and the unit is in MODE 1 or 2, then the unit is to be placed in MODE 3 per UOP-3.1 and UOP-2.1 , within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Monday, January 14, 20082:43:56

PM 25

QUESTIONS REPORT for 25 SRO Questions 037 Steam generator tube leakage G2.4.11 Knowledge of Abnormal operating procedures.

Question Number: original question#82 Tier 1 Group 2 Importance

Rating: 3.6 Technical Reference:

AOP-2 and bases 3.4.13-3 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5B5 wrote new question for new KA approved by FJE.10-30-2007(was

KA 059G2.4.30)

Scramble Range:A-D Version:0123456789

Answer: Be CADDC CDA Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO NEW HIGHER SRO GTO KIA MATCH ANALYSIS This question tests the AOP selection and the TS involved.Since it is an immediate be in mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months then it is required knowldge and reference is not provided.Tech Specs can be considered

a procedure that is used by the operators.

The question tests the knowledge of whether or not a'limit is violated.The applicant must have this knowledge in order to have the ability to execute the Tech Specs.Testing the procedural

entry requirement

makes it SRO-only level.MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:56 PM 26

QUESTIONS REPORT for 25 SRO Questions 10.039 A2.01 002 A Large Break LOCA has occurred on Unit 1 with the following conditions:*B Train is the on service train.*An LOSP has occurred and B Train emergency power is not available.

SG pressures are 680 psig and stable.*Containment

pressure rose to 31 psig and is now 8 psig and slowly lowering.*The crew is at the step to verify SI flow stable in ESP-1.3, Transfer to Cold Leg Recirculation, with*the following conditions:

Containment

Spray is aligned to the RWST.1A RHR pump is running with proper flow and is aligned to the containment

sump.1A Charging pump has tripped on overcurrent.

Which ONE of the following describes the procedure flow path required and the action that would be taken to reduce SG pressure?A.*Transition

to ECP-1.1, Loss of Emergency Coolant Recirculation;

Dump steam from the SGs using the steam dumps and maintain the cooldown rate less than 1 OQoF per hour.B.*Transition

to ECP-1.1, Loss of Emergency Coolant Recirculation;

Dump steam from the SGs using the Atmospheric

Relief Valves and maintain the cooldown rate less than 100°F per hour.C.*Continue in ESP-1.3 and align the CS system for recirculation, then transition

back to EEP-1.0, Loss of Reactor or Secondary Coolant;*Dump steam from the SGs using the steam dumps at the maximum attainable

rate.*Continue in ESP-1.3 and align the CS system for recirculation, then transition

back to EEP-1.0, Loss of Reactor or Secondary Coolant;*Dump steam from the SGs using the Atmospheric

Relief Valves at the maximum attainable

rate.Monday, January 14, 2008 2:43:56 PM 27

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)5 requirementsforSRO level question.A.incorrect-The ste*p in ESP-1.3 (7.27)has the crew verify SI flow is stable.If it is, then continue in ESP-1.3 and place CS on recirculation, then go to procedure and step in effect.Since a LBLOCA has occurred, EEP-1 would be the procedure used to get to ESP-1.3.Since the 1A RHR pump is running with proper flow, SI flow will be stable and the transition

to ECP-1.1 not required.The way the step is written, as shown below, could confuse the candidate in that they could assume all SI flow)s stable when only one is required to be stable.7.27 Verify SI flow-STABLE.ATRN HHSI FLOW[]FI 943 HHSI B TRN RECIRC FLOW[]FI 940 1 A(1 B)RHR HDR FLOW[]FI 605A[]FI 605B 7.27 IF at least one train of flow from the containment

sump to the RCS*can NOT be established

or maintained, THEN go to FNP-1-ECP-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION.

If the applicant decided ECP-1.1 was the proper procedural

flow path, then the two choices of using dumps or ARVs is available and the CDR would be correct for this procedure.

Dumps are not available since CTMT pressure went to 31 psig and the MSIVs are closed.B.incorrect-see above C.incorrect-since the MSIVs would be closed due to the LBLOCA and the LOSP, the dumps would not and could not be used.Our design of the MSIVs do not allow them to be re-opened until'a 50 PSID is reached across the valve.D.Correct-The step in ESP-1.3 (7.27)has the crew verify SI flow is stable.Since it is, then they continue in ESP-1.3andplace CS on recirculation, then go to procedure and step in effect.Since a LBLOCA has occurred, EEP-1 would be the procedure used to get to ESP-1.3 and would be returned to.Then EEP-1 has the crew decide to release pressure from the SGs to decrease the delta P across the tubes.Since the dumps are not available, the ARVs would be used.The max attainable

rate is procedurally

driven by EEP-1 and makes for a great distracter

as well because someone not familiar with the reason for decreasing

pressure at this time would be confused and would

Monday, January 14,20082:43:56

PM 28

QUESTIONS REPORT for 25 SRO Questions 039 A2.01 Main and Reheat Steam Ability t (a)predict the hllpacts of the following mal-functions

or operations

on the Main Reheat Stearn System;and (b)based on predictions, use procedures

to correct,.control, or rnitigate consequences

of those malfunctions

or operations:

Flow paths of steam during a

Question Number: Tier 2 Group 1 87 Importance

Rating: 3.2 Technical Reference:

E-1 and ESP-1.3 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5 Comments: This question was rewritten in entirety.One reason is the original submitted question did not meet the KA.I had to rewrite it to a LB LOCA since 035 A2.06 on this exam tests the procedural

transition

to ESP-1.2.This would have been ideal for this question but was double jeopardy.I did not find a procedural

transition

question to ECP-1.1 or back to EEP-1 on this exam.This question also had to deal with steam flows during a LOCA.There is no steam flow in EEP-1 except at step 18 which depressurizes

the SGs.Since ESP-1.2 has been taken away by a previous question, the logical step was to go to step 18.To get the procedural

transition

piece, I had to place enough in the stem for the applicant to analyze and to decide which procedure would be best.If this is not satisfactory, I will need another suggestion

or a replacement

KA.We ran on simulator to get proper pressures and temperatures

MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER SRO GTO Version:0123456789

Answer: DABBA CADDD I Scramble Range: A-D*Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January*14, 20082:43:56

PM 29

QUESTIONS REPORT for 25 SRO Questions 11.040 AA2.0 1 002 Given the following on Unit1:*Reactor Trip and Safety Injection have'occurred.

RCS Pressure is 2010 psig and DECREASING.

Pressurizer

level is 22%and rising.*LOOPA-Tcold on TR0410, RCS COLD LEG, is 510°F and DECREASING.

Containment

Pressure is 16 psig and INCREASING.*1A SG Pressure is 520 psig and DECREASING.*1 Band1C SG pressures are 840 psig and STABLE.*Sub Cooled Margin Monitor isreading130°F.

Which ONE of the following describes the location of the break and the next procedure the crew will perform after transition

from EEP-O.O, Reactor Trip or Safety Injection?

A.Downstream

of 1A MSIV;ESP-1.1, SI Termination.

B.Downstream

of1A MSIV;EEP-2.0, Faulted Steam Generator Isolation.

C.Upstream of1A MSIV;ESP-1.1, SI Termination.Upstream of1A MSIV;EEP-2.0, Faulted Steam Generator Isolation.

Meets 10 CFR 55.43 (b)5 requirements

for SRO level question.A.incorrect because the break is in the wrong place, as indicated by containment

pressure.B.incorrect because the break is in the wrong place, as indicated by containment

pressure.C.incorrect due to incorrect procedure, but credible because the break is in the correct place and the procedure would be correct for a downstream

break.D.correct.Indications

of a Faulted SG upstream of MSIV due to containment

pressure.EEP-2 will be addressed because the SG will eventually

depressurize.

Monday, January 14,2008 2:43:56 PM 30

QUESTIONS REPORT for 25 SRO Questions 040 Steam line rupture AA2.01 Ability to determine and interpret the following as they apply to the Steam Rupture: Occurrence

and location of a steam line rupture from pressure and flow indications

Question Number: Tier 1 Group 1 Importance

Rating: 80 SRO 4.7 Technical Reference:

EEP-O.O Proposed references

to be provided to applicants*during

examination:

None Learning Objective:10 CFR Part 55 Content: 43.5 Comments: The parameters

were picked based on running this event on the simulator and picking hypothetical

values this could happen depending on the reaction time of the crew and a small.steam break.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER SRO GTO Version: a123456789 Answer:DBBDCBCBBB

Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:56 PM 31

QUESTIONS REPORT for 25 SRO Questions 12.061 AA2.03 001 Given the following:

Unit 1 is at 1 00%power.*A fuel shuffle is in progress in the SFP room.*The following alarm is received:*FH1, RMS HI-RAD The OA TC reports the following:

R-5, Spent Fuel Pool Area Monitor, Red HIGH alarm light is LIT.*R-25A and R-25B, SFP VENT, radiation monitor-Amber ALERT light is LIT.-Red HIGH alarm light is NOT LIT.Which ONE of the following describes the current status of Spent Fuel Pool Supply and Exhaust Fans, and the actions that will be required lAW FH1, RMS HI-RAD?A.*Spent Pool Supply and Exhaust fans are running;*Implement EIP-9, Emergency Actions, determine if Automated Rapid Dose Assessment (ARDA)has actuated, and verify both trains of PRF running.B.*Spent Fuel Pool Supply and Exhaust fans are tripped;*Implement EIP-9, Emergency Actions, determine if Automated Rapid Dose Assessment (ARDA)has actuated, and verify both trains of PRF running.C.*Spent Fuel Pool Supply and Exhaust fans are tripped;*Enter AOP-30.0, Refueling Accident, isolate the Control Room and place Control Room Emergency Filtration

system (CREFs)in service.*Spent Fuel Pool Supply and Exhaust fans are running;*Enter AOP-30.0, Refueling Accident, isolate the Control Room and place the Control Room Emergency Filtration

system (CREFs)in service.Monday, January 14, 2008 2:43:56 PM 32

Setpoint 4.44e-4 IJc/ml 1.20e-6 IJc/ml 4.00e-5 IJc/ml.027 RIhr.038 RIhr.038 R/hr Setpoint 13000 (VI)11571 (V2)CPM 1800 (VI)4280 (U2)CPM 156 (VI)143 (V2)CPM QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)4 and 5 requirements

for SRO level question.A.incorrect.

because the actions are inco'rrect

for the event taking place.Plausi'ble

because it is an action that would be performed for different conditions.

ARDA will activate when R-25A/B go red, and then EIP-9 would be referred to.Both trains of PRF would actuate for R-25A/B high alarm SFP exhaust goes to the AB exhaust plenum which feeds the plant vent stack.The plant vent stack is monitored by R-14, 29 and 22 which would activate ARDA.EIP-9.1 ARDA will automatically

start when any of the following monitors go into alarm for two consecutive

system polls one minute apart on the applicable

unit and use the latest 15 minute average monitor value to perform the calculations:

Monitor Plant Vent Stack R29 (SPING)Noble Gas Iodine Particulate

Steam Jet air Ejector R15C TDAFW Exhaust R60D Steam Generator AlBIC R60AIBIC ARDA will also automatically

start when any of the following monitors go into alarm for two consecutive

system polls one minute apart on the applicable

unit.The ARDA system will use the plant Vent stack SPING latest 15 minute average monitor value to perform the calculations

when these monitors activate the system: Monitor Plant Vent stack Monitors Gas monitor R 14 Particulate

monitor R 21 Gas monitor R 22 B.incorrect.

R-25A or B ,RED alarm light realigns FHB ventilation

NOT the AMBER alert light.C is incorrect.

because FHB fans are running.Credible because R-25A/B RED alarm setpoint has not been reached, so applicant may think FHB ventilation

has realigned.

AOP-30 directs Control Room isolation and starting CREFs.D.Correct.because FHB fans are running.Credible bec,ause R-25A/B RED alarm setpoint has not been reached,.AOP-30 directed Control Room isolation and starting CREFs Monday, January 14, 2008 2:43:56 PM 33

QUESTIONS REPORT for 25 SRO Questions 061 Area Rad Monitoring

alarms: AA2.03 Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM)System Alarms: Setpoints for alert and high alarms.Question Number: Tier 1 Group 2 83 3.3 ARP FH5.and FH1 and EIP-9.1;U258400;AOP-30;A181015;43.5/6 and 55.43(b)(4 and 5)Importance

Rating: Technical Reference:

OPS-52106D

Proposed references

to be provided to applicants

during examination:

Learning Objective:

10 CFR Part 55 Content: Comments: None Rewrote to FJE comments.MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: NEW HIGHER SRO GTO Version:0123456789

Answer: DCA C D DADBD Scramble Range:A-D Source if Bank: Difficulty:

Plant: 'FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:56 PM 34

QUESTIONS REPORT for 25 SRO Questions 13.062 G2.1.33 002 Given the following:

Unit 1 is in Mode 1.*WD2,1B INV FAULT, comes into alarm.*The ROVER reports the following indications

on the1B Inverter panel:-The BYPASS SOURCE POWERING LOAD light is LIT.The'INVERTER POWERING LOAD light is NOT LIT.-The battery input breaker has tripped open.-The BYPASS SOURCE AVAILABLE light is LIT.Which ONE of the following statements

describes the Technical Specification

ACTION statement(s)

that MUST be entered?Art*Enter the TS LCO action statement for 3.8.7, Inverters-Operating.

LCO 3.8.9, Distribution

Systems-Operating, entry is NOT required.B.*Enter the TS LCO action statement for 3.8.7, Inverters-Operating.

Enter the TS LCO action statement for 3.8.9, Distribution

Systems-Operating.

C.*LCO 3.8.7, Inverters-Operating, entry is NOT required.*Enter the TS LCO action statement for LCO 3.8.9, Distribution

SystemsOperating.

D.*LCO 3.8.7, Inverters-Operating, entry is NOT required.*LCO 3.8.9, Distribution

Systems-Operating,entry

is NOT required.Monday, January 14, 2008 2:43:56 PM 35

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)2 requirements

for SRO level question.This is SRO only in that a candidate would have to look at bases to determine whether 3.8.7 and 3.8.*9 apply.The definitions

of operability

are found in bases..A.Correct.Since the1B Inverter has lost the DC source, the indications

above show that the inverter swapped to the bypass source.Since this is true, the1B vital panel is still energized.

T8 3.8.7 applies and Condition A has the SRO look atapplicabilityof

3.8.9.If the inverter did not swap per design and the vital panel was de-energized, then 3.8.9..would be required to be entered also.The reason the inverter is INOPERABLE

is blc of the DC source is required to be the primary source of power to the inverter.Ac is just the backup.I*The vital panel is operable since it is powered from an inverter.3.8.9 says the vital panel can be powered from an inverter that is powered from either AC or DC source..3.8.7 Operable inverters require the associated

vital bus to be powered by the inverter with output voltage and frequency within tolerances, and power input to the inverter from a 125 VDC station battery.

E:nter applicable

Conditions

and Required

of LCO 3.8.9, IIDistribution

-Operating" with any vital bus deenergized.

bases of 3.8.7.With a required inverter inoperable, its associated vital bus becomes inoperable

until it is re-energized

from its Class 1 E: CVT.For this reason a has been included in Conditionrequiring the entry into the Conditions

and Required

of LCO 3.8.9,"Distribution

Systems-Operating." This ensures that the vital bus is re-energized

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .The'associated

static transfer switch normally provides a bumpless transfer of power to the alternate source (Class1E CVT).3.8.9

vital bus electrical

power distribution

subsystems

require the associated

buses to be energized to their proper voltage from the associated

inverter via inverted DC voltage or Class 1 E: constant voltage transformer.

B.Incorrect.

3.8.7 is entered and 3.8.9 is NOT.C.Incorrect.

3.8.7 is entered and 3.8.9 is NOT.D.Incorrect.

3.8.7 is entered and 3.8.9 is NOT.Monday, January 14, 2008 2:43:57 PM 36

QUESTIONS REPORT for 25 SRO Questions 062 AC electrical

distribution

-G221.33 Conduct of operations:

Ability to recognize indications

for systerTI operating parameters

which are entry conditions

for technical specifications

Question Number: Tier 2 Group 1 88 Importance

Rating: 4.0 Technical Reference:

TS 3.8.7,

and bases Proposed references

to be provided to applicants

during examination:

Learning Objective:

10 CFR Part 55 Content: 43.2 None-Comments: This KA tests the recognition

of entry conditions

to TSs and isSRO in that bases knowledge has to be understood

for the TS referenced

and also detailed knowledgeofthe note in LCO 3.8.7 that sends the SRO to 3.8.9 and why.MCS Time: Source: Cognitive Level: Job

reviewed: Points: 1.00 NEW illGHER SRO GTO Version: a123456789 Answer: ADCCBDC BCD Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:57 PM 37

QUESTIONS REPORT for 25 SRO Questions 14.073 A2.01 001 Given the following:

Unit 1 is at 100%power.*The following alarm is received:*FH2, RMS CH FAILURE R-23S, SGSD TO DI*LUTION, radiation monitor is indicating

downscale with all indicating

lights extinguished.

Which ONE of the following describes the effect of the failure and the associated

ODeM requirement?

A.*FGV-1152, SGSD Heat Exchanger Discharge valve, will close;*SGSD releases to the environment

can NOT continue.B.*FCV-1152, SGSD Heat Exchanger Discharge valve, will close;*SGSD releases can continue provided chemistry analyzes grab samples.*RCV-23S, SGSD Dilution Discharge valve, will close;*SGSD releases can continue provided chemistry analyzes grab samples.D.*RCV-23S, SGBD Dilution Discharge valve, will close;*SGSD releases to the environment

can NOT continue.Meets 10 CFR 55.43 (b)2 requirements

for SRO level question in the realm of the ODCM requirements

when a radiation monitor fails.A.incorrect.

FCV-1152, SGSD Heat Exchanger Discharge valve will not close for this rad monitor.credible due to R-23A will close 1152.The release can be continued.

Purification

Outlet Radiation Monitor (RE-23A)The purification

outlet radiation monitor determines

the activity level of the fluid entering the surge tank.When the demineralizer

train is bypassed, this instrument

indicates the activity of the untreated blowdown fluid.If the blowdown is being processed, this instrument

will indicate a radioactive

breakthrough

across the demineralizers.

In any case, a high activity signal from RE-23A closes FCV-1152, which stops the blowdown.Indication

and a high alarm are on the radiation monitoring

system (RMS)panel in the main control room.B.Incorrect-FC'v'-1152, SGSD Heat Exchanger Discharge valve will not close for this rad monitor.The action listed is correct.c.correct.since the rad monitor gives a high signal upon a loss of power, the actions for the high alarm will occur.This will close RCV-23B.Monday, January 14,20082:43:57

PM 38

QUESTIONS REPORT for 25 SRO Questions FH2 AUTOMATIC ACTION 1.The radiation monitors fail to a"High Radiation" condition on loss of instrument

and/or control power that-will result in actuation of associated

automatic functions.

Refer to annunciator

FH1 for automatic actions.ODeM requirements

Instrument

Minimum Channels OPERABLE ACTION.Gross Radioactivity

Monitors Providing Automatic Termination

of Release a.Liquid Radwaste Effluent Line (RE-18)b.Steam Generator Slowdown Effluent Line (RE-23S)1 1 28 29 ODCM page 2-4 ACTION 29-With the num.ber of channels OPERABLE less than required by the Minimum'Channels OPERABLE requirement, effluent releases via this pathway may continue, provided grab samples are analyzed for gross radioactivity (beta or gamma)at a MINIMUM DETECTABLE

CONCENTRATION

no higher than1x 10-7 micro Ci/mL.Discharge Radiation Monitor (RE-23Bl RE-23B monitors the activity level of the fluid"leaving

the SGBD.A high activity signal from this instrument

closes RCV-23B, which prevents the discharge of high activity fluid to the environment.

Indication

and a high alarm are located on the RMS panel in the main control room.D is incorrect since the release can continue.FH1 guidance R-23A SG Slowdown Surge Tank Liquid Inlet (AS 130 1)R-23B SG Blowdown Surge Tank Liquid ODCM Discharge (AB 130 1)Monday, January 14,20082:43:57

PM Scint.(W)Scint.(W)Closes Perform Step FCV-1152 4.23 Closes'Perform Step RCV-23S 4.23 39

QUESTIONS REPORT for 25 SRO Questions*

073 Process radiation monitoring

A2.01 Ability to (a)predict the impacts of the following malfunctions

or operations

on the PRM system;and (b)based on those predictions, use procedures

to

control, or mitigate the consequences

of those malfunctions

or operations:

Erratic or failed power supply Question Number: Tier 2 Group 1 89 Importance

Rating: 2.9 Technical Reference:

ARP-1.6, FH2, FH1 SGSD lesson plan and aDCM page 2-4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

.10 CFR Part 55 Content: 43.4 Comments: This was rewritten to incorporate

comments from FJE and made the ODCM applicable

to make it SRO level MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER SRO GTO Version:0123456789

Answer:CCA CABCCAA Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:57 PM 40

SRO QUESTION 076G2.1.2 Distractor

"8" is also a correct answer.See examination

report 05000348/2007301

and 05000364/2007301

Enclosure 2.

QUESTIONS REPORT for 25 SRO Questions 15.076 G2.1.2 001 Given the following:

Unit 2 is at 100%power withllA II Train on service.*At 1200 on 11/7/2007, 2E Service Water pump tripped and IIBII Train SW was declared INOPERABLE.

Which ONE of the following describes the Technical Specification

REQUIRED ACTION lAW 3.7.8, Service Water System, and the action required to make IIB II Train Service Water OPERABLE?Art*Immediately

declare the DG supported by Train IIB II Service Water INOPERABLE.

Place IIB II Train of SW on service and align 2C SW pump to auto start for 2E SW pump lAW SOP-24.0, Service Water System.B.*Immediately

declare the DG supported by Train IIB II Service Water INOPERABLE.

Align 2C SW pump to auto start for 2E SW pump lAW AOP-1 0.0, Loss of Service Water.C.*Declare the DG supported by Train IIBII Service Water INOPERABLE

no later than 1600 on 11/7/2007 (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months later).*Align 2C SW pump to auto start for 2E SW pump lAW AOP-1 0.0, Loss of Service Water.D.*Declare the DG supported by Train IIB II Service Water INOPERABLE

no later than 1600 on 11/7/2007 (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months later).*Place B Train of SW on service and align 2C SW pump to auto start for 2E SW pump lAW SOP-24.0, Service Water System.Meets 10 CFR 55.43 (b)2 and 5 requirements

for SRO level question A.Correct.This TS im.mediately

entered from 3.7.8 and the DG is declared INOP.Then the 2C SW pump is aligned to auto start for 2E.The trains are swapped to do this.This will allow both trains to be operable.*AOP-10 CAUTION: Based on plant needs, shifting electrical

trains in FNP-1-S0P-24.0, SERVICE WATER SYSTEMS, may be delayed.Subsequent

shifting of electrical

trains is required for train separation.

19 IF affected train NOT leaking, THEN evaluate aligning1C SW pump to affected train using FNP-2-S0P-24.0, SERVICE WATER SYSTEM.Bases 3.7.8 LCO Two SWS trains are required to be OPERABLE to provide the required redundancy

to ensure that the system functions to remove post accident heat loads, assuming that the worst Monday, January 14, 2008 2:43:57 PM 41

QUESTIONS REPORT for 25 SRO Questions case single active failure occurs coincident

with the loss of offsite power.An SWS train is considered

OPERABLE during MODES 1, 2, 3, and 4 when: a.Two pumps are OPERABLE;and b.The associated

piping, valves, and instrumentation

and controls required to perform the safety related function are OPERABLE.Notefrom A.1 The first Note indicates that the applicable

Conditions

and Required Actions of LCO 3.8.1,"AC Sources-Operating," should be entered if an inoperable

SWS train results in an inoperable

emergency diesel generator.

FSD 181001 3.1.5.1 The Service Water pumps shall be automatically

started by a signal from the LOSP or ESS sequencer.

The Service Water swing pump shall be automatically

started by a signal from the LOSP or ESS sequencer when inservicereplacing

one of the train oriented pumps.(References

6.7.039 and 6.1.009)SOP-24 P&L 3.3 Service Water pump Ie may be selected for auto-start

from the ESS or the LOSP sequencers, instead of an A Train or B Train pump, by using key-interlocked

selector switches at the SW local control panels.Normal position of both the A Train and B Train selector switches will be the lC position and lC SW pump will not autostart.

B.Incorrect.

This TS is immediately

entered from 3.7.8 and the DG is declared INOP.The second part is in part correct but B Train would be however AOP-1 0 sends the operator to SOP-24 to select the 2C SWP to autostart and if this was done wlo swapping trains it would be in an incorrect alignment.

This has to be done in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (3 days later)lAW TS 3.7.8.NOT 7 days..c.incorrect.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would NOT be allowed to declare inop if DG was OOS.The second part is in NOT correct.See above.D.incorrect.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would NOT be allowed to declare inop if DG was OOS.Second part of this is correct.Monday, January 14, 2008 2:43:57 PM 42

QUESTIONS REPORT for 25 SRO Questions 076 Service Water System G2.1.2 Conduct of Operations:

Knowledge of operator responsibilities

during all modes of plant operation.

Question Number: Tier 2 Group 1 90 Importance

Rating: 4.0 Technical Reference:

TS 3.7.8,3.8.1, AOP-10, SOP-24 Proposed references

to be provided to applicants

during examination:

NoneLearningObjective:

10 CFR Part 55 Content: 43.2 Comments: fixed per FJE comments and added how to restore B train to operable status.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW ffiGHER SRO GO Version:0123456789

Answer: ADDDACC BCD Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:57 PM 43

QUESTIONS REPORT for 25 SRO Questions 16.E03 02.4.4 006 A spurious SI has occurred on Unit 1 with the following conditions:

All systems functioned

as required.*ALL but one charging pump has been secured lAW EEP-O, Reactor Trip or Safety Injection.

After establishing

normal charging in EEP-O, RCS pressure started to decrease and PRZR level started trending down from 35%and is now 14%.Which ONE of the following describes the actions and procedural

transition

the SRO must direct at this point?A.Reinitiate

a manual Safety Injection and transition

back to step 1 of

B.Return to the diagnostic

steps (13 through 15)of EEP-O, and then transition

to EEP-1, Loss of Reactor or Secondary Coolant.C.Transition

to ESP-1.1, SI Termination, step 6, and apply the foldout page of ESP-1.1 to re-establish

HHSI flow.Re-establish

HHSI flow per EEP-O, and then transition

to E8P-1.2, Post LOCA Cooldown and Depressurization.

Meets 10 CFR 55.43 (b)5 requirements

for SRO level questionA-Incorrect;

If PZR level can not be maintained, the flow path must be reestablished

and a transition

to ESP-1.2 is warranted.

There is no need to manually 81 and transition

back to step 1 of EEP-O.B-Incorrect;

returning to diagnostics

once they have been completed and additional

actions taken may seem plausible to get to EEP-1, but this is not allowed per sop-O.8 procedural

use guidelines

procedure.

Also the RNO column of EEP-O directs the correct actions for this condition.C-Incorrect;

the very next step of EEP-O sends the user to ESP-1.1.While this would eventually

lead the crew to the right place, ESP-1.1 foldout page has the operator go to EEP-1 after the HHSI was re-established.

This may seem to be a method to use but is'not procedurally

correct.D-Correct;From EEP-O, step 19 and 21, RNO, says that if PZR level or pressure cann'ot be maintained, the procedural

requirement

is to go to E8P-1.2.Monday, January 14, 2008 2:43:57 PM 44

QUESTIONS REPORT for 25 SRO Questions E03'lOCA cooldown and depressurization

G2.4.4 Emergency Procedures

/Plan Ability to recognize abnormal indications

for system operating parameters

which are entry-level

conditions

for emergency and abnormal operating procedures.

Question Number: B4 Tier 1 Group 2 Importance

Rating: 4.3 Technical Reference:

EEP-O and SOP-O.B Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5 Scramble Range:A-D NEW mGHER SRO GTO Comments: This question was written to address the double jeopardy transition

issue and not giving away answers by other questions with 035 A2.06.Instead of using EEP-1 to transition

to ESP-1.2, EEP-O is being used to give a similar procedural

transition.

MCS Time: 1 Points: 1.00 Version:0123456789

Answer: DCAB CDBAAA Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:57 PM 45

QUESTIONS REPORT for 25 SRO Questions 17.E04 G2.2.22 001 Given the following:

Reactor trip and safety injection have occurred on Unit 1.*ECP-1.2, LOCA Outside Containment;

has been completed.

ESP-1.1, SI Termination, has been completed.

UOP-2.1, Shutdown of Unit from Minimum Load to Hot Standby, is in progress.*PRZR level is 75%and increasing

slowly.*1B RCP is operating.*1A and1C RCPs are secured.*RCS Tavg is 526°F and increasing

slowly.Which ONE of the following describes theTechnicalSpecification

LCO action statement in effect for the given conditions

and the basis for the LCO?*3.4.9, Pressurizer;

To maintain pressure control to minimize the consequences

of potential overpressure

transients.

B.*3.4.9, Pressurizer;

To maintain RCS subcooling

during natural circulation

conditions.

C.*3.4.5 RCS Loops-Mode 3;*To ensure adequate decay removal from the core in the event of an inadvertent

control rod withdrawal.

D.*3.4.5 RCS Loops-Mode 3;*To ensure adequate decay heat removal from the core and proper boron mixing throughout

the RCS.Monday, January 14, 2008 2:43:57 PM 46

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)2 requirements

for SRO level question A is correct.Since the PZR must be kept below 63.5%in mode 3.Adequate volume for a steam bubble for pressure control is the reason for the 63.5%max pzr level in Mode 1-3.The pressurizer

shall be OPERABLE with: a.Pressurizer

water level<or equal to 63.5%indicated;

Bases for LeO 3.4.9 page 83.4.9-2 The LCO requirement

for the pressurizer

to be OPERABLE with a water volume=868 cubic feet, which is equivalent

to 63.5%indicated, ensures that a steam bubble exists.'Limiting the LCO maximum operating water level preserves the steam space for pressure control.The LCO has been established

to ensure the capability

to establish and maintain pressure control for steady state operation and to minimize the consequences

of potential overpressure

transients.

Requiring the presence of a steam bubble is also consistent

with analytical

assumptions.

B.is incorrect.

Basis for PZR heaters.C.is incorrect.

3.4.5 is not in effect since the RTBs are open and there is one RCP running.Two RCS loops shall be OPERABLE, and either: a.Two RCS loops shall be in operation when the Rod Control System is capable of rod withdrawal;

or b.One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

b'ut basis for that spec would be correct in Mode 2 with RTBs closed D.is incorrect.

Incorrect spec (see above), but correct basis for operability

requirements

with 1 RCP, RTBs open.Monday, January 14, 2008 2:43:57 PM 47

QUESTIONS REPORT for 25 SRO Questions E04 lOCA outside ctmt G2.2.22 Equipment Control: Knowledge Limiting Conditions

for Operations

and safety limits Question Number: Tier 1 Group 1 81 Importance

Rating: SRO 4.1 Technical Reference:

TS 3.4.9&3.4.5 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 43.2 Comments: Changed per FJE comments note from es-401-9 below: Exal'Diner

Note: Question meets first half of KIA (LOCA outside of containment)

because the event is necessary to provide a credible context for the given plant conditions.

MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER SRO GTO Version:0123456789

Answer: ADBAABD BCD Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:58 PM 48

QUESTIONS REPORT for 25 SRO Questions 18.E15 EA2.1 005 Given the following:*A Large Break LOCA has occurred on Unit1.*'RWST level is 14.8 feet and slowly lowering.The crew is at step 15 in EEP-1.0, Loss of Reactor or Secondary Coolant, to check LHSI flow in progress, when the following containment

indications

are reported by the OATC:*FI-958A, CS flow, reads a gpm.*FI-958B, CS flow reads 1850 gpm.*Ctmt Pressure is 29.5 psig and rising slowly.*Ctmt Sump Level 8.0 feet and rising slowly.*Ctmt Radiation Level is 3.6 Rem/Hr on both High Range instruments.

Which ONE of the following'describes

the next action to take for these conditions?

A.Implement FRP-Z.1, Response to High Containment

Pressure.Implement FRP-Z.2, Response to Containment

Flooding.C.Implement FRP-Z.3, Response to High Containment

Radiation.

D.Transition

to ESP-1.3, Transfer to Cold Leg Recirculation.

Meets 10 CFR 55.43 (b)2 requirements

for SRO level question A.Incorrect.

27 psig is ORANGE Path on pressure.Plausible, because if flow for CS dropped below 1000 gpm this would be the correct procedure on an orange path.As is it would be a yellow path IF CTMT sump level was less than 7.6 feet.B.Cor'rect.Since CS flow is>1000 gpm, and Ctmt sump level>7.6 feet, this is an orange path on Z.2.C.Incorrect.

This is a yellow path in the same CSF network and the c,andidate

has to know it is a yellow path and it is below or less critical than Ctmt sump level.This would be a correct choice if ctmt sump level is<7.6 feet.D.lncorrect.

Once ESP-1.3 is entered, no CSF applies.In this case, the crew is holding at step 160f level

transition

to ESP-1.3.Due to the RWST level at 14.8 feet the crew would not wait on the transition

but would enter FRP-z.2 until ESP-1.3 was required.Monday, January 14, 2008 2:43:58 PM 49

E15 CTMT flooding QUESTIONS REPORT for 25 SRO Questions EA2.1 Ability to determine and interpret the following as they apply to (Containment

Flooding)Facility conditions

and selection of appropriate

procedures

during abnormal and emergency operations.

Question Number: Tier 1 Group 2 85 Importance

Rating: 3.2 Technical Reference:

CSF-O.5, CSFSTs Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS 52533M02 10 CFR Part 55 Content: 43.5 Comments: Rewrote the question to specifically

address the KA.With the conditions

given the candidate has to evaluate the CSFs and determine the appropriate

procedure to go to which is tied to the KA.Scramble Range:A-D FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 MODIFIED HIGHER SRO GO Version: a123456789 Answer: B ACBCBBB A C Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:58 PM 50

19.G2.1.25 004 QUESTIONS REPORT for 25 SRO Questions Unit 1 is in a refueling outage.The following conditions

exist:*Nozzle dams are installed on ALL Steam Generators.

Both trains of RHR are in service.*RCS level is 123 1 3 11 and stable.*RCS temperature

is 120°F and stable.*Secondary side of all SGs is>85%wide range.*A charging pump is in service;Band C are tagged out.*The equipment hatch is closed.*The time to saturation

is 35 minutes.The OATC reports that both RHR pumps have just tripped due to breaker problems.Using UOP-4.0 Appendix1, SHUTDOWN SAFETY ASSESSMENT, which one of the following is the correct procedure to go to and proper condition based on the events in progress?.References

Provided C.Go to AOP-45, SHUTDOWN INVENTORY, under an Orange condition.

D.Go to AOP-45, SHUTDOWN INVENTORY, under a Red condition.

Monday, January 14,20082:43:58

PM 51

References:

QUESTIONS REPORT for 25 SRO Questions UOP-4.0 Appendix 1, figure 1a page 6 of 16 Version 28 Meets 10 CFR 55.43 (b)requirements

for SRO level question in that this is a task only performed by an SRO at FNP.A.Incorrect-

This is a red condition not orange.plausible since time to saturation

is a 1 and if the SGs were evaluated improperly, then an orange condition would be selected.B.Correct-RED 25 min to saturation

based on Table B for 100°F.The SG tubes are not filled and vented which gives them a 0 and#5 is not met as well.With the loss of the RHR system, AOP-42 would be

Even if a 1 was entered for#5, the condition would still be an unexpected

RED.CORE COOLING 1.2 SOs Avail with loops filled (Ref step 2.7)2.Cavity level=142'1" wI Upper Internals Removed 3.RHR Subsystems

Available (0, 1 or 2)4.RCS level=126'6" 5.Time to saturation>

30 minutes OR RCS press>325 psig with at least one RCP available for operation and at least one SO available Core Cooling Subtotal Subtotal Condition__0_0-1 RED_--",-0_2-3 ORANGE__0_4 YELLOW 0__ GREEN1_(GREEN if Defueled)__1_AOP 42 42 C.Incorrect-

yellow is correct evaluated condition for Inventory, not orange and thenifit was yellow the AOP would not be addressed.

D-Incorrect-

yellow is correct.Plausible b/cthe HHSI flow path is not identified

and not in use and could be considered

not available, especially

in a shutdown mode.The RCS is intact since the nozzle dams are installed.

INVENTORY 1.Refueling Cavity23 Feet (142'I")Above Fuel 2.,LHSI PumplFlowpath

Available 3.HIlSI PumplFlowpath

Available 4.RCS is Intact below the Reactor Vessel Flange Inventory Subtotal Monday, January 14, 20082:43:58

PM Subtotal Condition°'RED 1 ORANGE 2 YELLOW 3-4 GREEN (GREEN if Defueled), AOP 45 45 52

QUESTIONS REPORT for 25 SRO Questions G2.1.25 Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain perforrnance

data.Question Number: 94 Tier 3 Group 1 Importance

Rating: 3.1 Technical Reference:

UOP-4.0 Proposed references

to be provided to applicants

during examination:

UOP-4.0 Appendix 1, figure1a page 6 of 16 Version 28 Learning Objective:

10 CFR Part 55 Content: Scramble Range:A-D FARLEY Version:0123456789

Answer: B DB ACBB DAB Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO MODIFIED HIGHER SRO GO Comments: This question was changed out to get a better match to the KA.This is an SRO task during an outage done daily.MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:58 PM 53

FARLEY NOVEMBER/DECEMBER

2007 FINAL EXAMINATION

NOS.05000348/2007301

&05000364/2007301

THIS PAGE CONTAINING

SROQUESTIONG2.1.6

OMITTED FROM DISTRIBUTION

TO THE PUBLIC

QUESTIONS REPORT for 25 SRO Questions 21.G2.2.17 008 Given the following:

During a power reduction to 19%power, an SO reports a flange leak of 5 drops per minute on a Main Steam Line flange upstream of the MSIVs in the MSVR that can not be isolated.*The FIN team leader wants his team to tighten the flange and stop the leak as TOOLPOUCH MAINTENANCE.

Which ONE*of the following correctly states whether the work ca'n be performed as TOOLPOUCH MAINTENANCE

per Guidelines

for Scheduling

of On-Line Maintenance, and the reason?A.May be performed as TOOLPOUCH work because the work will not interrupt the flow of process fluid.B.May be performed as TOOLPOUCH work because the flange is not part of a safety related system.May NOT be performed as TOOLPOUCH work because the system pressure and temperature

are too high.D.May NOT be performed as TOOLPOUCH work because the'work will require entry into a technical specification

LCO.Monday, January 14, 20082:43:58

PM 57

QUESTIONS REPORT for 25 SRO Questions Meets 10 CFR 55.43 (b)requirements

for SRO level question due to being a supervisory

knowledge of work control procedures.

This is also an IR of 2.3 for an RO.A.Incorrect-

This is not acceptable

because the flange is part of a high temperature/high

pressure system.Even though the work will not i'nterupt flow, the TPM can not be perform'ed.

B.Incorrect-

This is not acceptable

because the flange is part of a high temperature/high

pressure system and because of the high energy of the system.C.Correct-NOT acceptable

because TOOLPOUCH WORK is not allowed on systems where the pressure is greater than 1000 psig or temperature

is greater than 200 degrees F.ACP-52.1 section 2.0 o Tightening

of un-isolatable

fittings with process fluids<1000 psig or<200 degrees F can be done as tool pouch work.If the system pressure is>1000 psig or temperature

is>200 degrees F, then a work order is required with Team Leader or above approval.(AI#2004202241)

D.Incorrect-itis not acceptable

but the reason given is incorrect.

TS LCO entry would not be required to tighten the flange.The stem placed the flange upstream of the MSIVs to give TS entry credibility.

TOOLPOUCH WORK is defined as work that can be conducted without detailed written Instructions

and without overall plant scheduling.

section 3.0 table Flanges Tighten to stop leakage (within maximum torque limits)Monday, January 14,20082:43:58

PM 58

QUESTIONS REPORT for 25 SRO Questions G2.2.17 Knowledge of the process for managing m.aintenance

activities

during power operations.

Question Number: 97 Tier 3 Group 2 Importance

Rating: 3.5 Technical Reference:

ACP-52.1, Appendix 3;

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5 Items Not Scrambled MODIFIED HIGHER SRO GTO Comments: Fixed per FJE comments and removed ctmt from stem.each distracter

has a valid reason why it could or could not be correct for the conditions

given.MCS Time: 1 Points: 1.00 Version:0123456789

Answer:CCCCCC C.CCC Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:43:58

PM 59

QUESTIONS REPORT for 25 SRO Questions 22.02.2.25 005 Technical Specification

3.4.16, RCS Specific Activity, states: The'specific activity of the reactor-coolant

shall not exceed 100/E bar microCi/gm

of gross activity.If this limit is not satisfie*d, the reactor shall be shut down and cooled to 500°F or less within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after detection.

Which one of the following is the basis for reducing Tavg to less than 500°F if the specific activity of the reactor coolant is not within the limits of LCO 3.4.16?A.Minimize the release of radioactivity

in the event of a LOCA outside containment.Prevent venting a ruptured steam generator to the environment.

C.Ensure that the,1-hour

dose at the SITE BOUNDARY will not exceed a small fraction of the10 CFR Part 1 00 dose guideline limits in the event of a SGTR.D.Ensure that the 1-hour dose at the SITE BOUNDARY willnotexceed a small fraction of the10 CFR Part 20 dose guideline limits in the event of a LOCA.Meets 10 CFR 55.43 (b)2 requirements

for SRO level question.A.Incorrect, LOCA dose is not the bases.B.Correct-Minimize the release of radioactivity

should a steam generator tube rupture occur.APPLICABLE:

The LCO limits on the specific activity of the reactor coolant ensures SAFETY ANALYSES that the resulting doses will not exceed an appropriate

fraction of the 10 CFR 100 dose guideline limits following a SGTR accident.The SGTR safety analysis (Ref.2)assumes the specific activity of the reactor coolant at 0.5 ocCi/gm, a conservatively

high letdown flow of 145 gpm, and a bounding reactor coolant steam generator (SG)tube leakage of 1 gpm total for three SGs.The MSLB analysis assumes a steam generator tube leakage of 500 gpd in the faulted loop and 470 gpd in each of the intact loops for a total leakage of 1440 gpd.Condition 8.1 The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and RCS average temperature

<500°F lowers the saturation

pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment

in an SGTR event.The allowed Completion

Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500°F from full power conditions

in an orderly manner and without challenging

plant systems.C.Incorrect, this isa2 hour dose and the reason for the specific activity limit, not the temperature.

D.Incorrect,LOCA

is not the concern and 10 CFR part 20 is not correct.TS 3.4.16'Basis Monday, January 14, 2008 2:43:58 PM 60

QUESTIONS REPORT for 25 SRO Questions G2.2.25 Knowledge of bases in technical specifications

for limiting conditions

for operations

and safety

Question Number: 96 Tier 3 Group 2 Importance

Rating: 3.7 Technical Reference:

TS 3.4.16 Basis Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 43.2 Scramble Range:A-D FARLEY MODIFIED LOWER SRO GO Comments: Changed out the question since the other question did not meet the KA.Per our telephone discussion

this KA can test bases in technical specifications

for limiting conditions

for operations

and bases in technical specifications

for limiting conditions

for safety limits (which are RO as well as SRO knowledge level questions.)

MCS Time: 1, Points: 1.00 Version:0123456789

Answer: BDDDC CDDAD Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:43:59 PM 61

QUESTIONS REPORT for 25 SRO Questions 23.02.3.8 002 Given the following:*A Waste Gas release of Waste Gas Decay Tank#5 was started on November 8, 2007 at 1500.Which ONE of the following describes a condition that would require termination

of the release once initiated, and the person who is required to be notified lAW with SOP-51.1, Waste Gas System Gas Decay Tank Release?A.*R-29A, PLANT VENT STACK, is declared INOPERABLE.

Shift Supervisor

B.*R-29A, PLANT VENT STACK, is declared INOPERABLE.

Health Physics Foreman C.*Waste Gas Decay Tank#4 pressure decreases during the release.*Health Physics Foreman*Waste Gas Decay Tank#4 pressure decreases during the release.*Shift Supervisor

Monday, January 14, 2008 2:43:59 PM 62

QUESTIONS REPORT for 25 SRO Questions A is incorrect.

R-29A becoming INOPERABLE

will not cause the release to be stopped, but R-14 would (see below).R-29A is a backup to R-29B in the event that R-29B fails and would be used to comply with ODCM to take grab samples.3.2 IF R-14 becomes inoperable

while discharging

gaseous waste to the vent stack, THEN discharge shall be stopped immediately

and the Shift Supervisor

notified.B is incorrect.

first not correct (see above)..Second part not correct, see C below..C is incorrect.

first part is correct, see below.second part NOT correct.The HP foreman is in the approval chain for the release, but is not required to be notified of termination

per SOP-51.1 D is correct.If another tank pressure drops, stop the release and Notify the Shift Supervisor.

SOP-51.1 step 4.1.15 Monitor all gas decay tank pressures during the release.Ensure that only the tank which is being released exhibits a pressure decrease and no other tank pressure increases.

Stop the release and notify the Shift Supervisor

if one of the above occurs.Monday, January 14, 2008 2:43:59 PM 63

QUESTIONS REPORT for 25 SRO Questions G2.3.8 Knowledge of the process for performing

a planned gaseous radioactive

release.Question Number: Tier 3 Group 3 98 Importance

Rating: 3.2 Technical Reference:

FNP-1-80P-51.1, FNP-O-CC*P-213.0

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.4 Comments: This fits the KA in that it is the 88 job function to know the process for the release and what is required should a particular

P&L not be met or if an instrument

should fail such as R-14 or R-22.MCS Time: 3 Source: Cognitive Level: Job Position: reviewed: Points: 1.00 MODIFIED LOWER SRO GTO Version:0123456789

Answer:DDCCAAADB D Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:59 PM 64

QUESTIONS REPORT for 25 SRO Questions 24.G2.4.27 002 Given the following:

Unit 1 and 2 are operating at 100%Power.*The Outside System Operator reports a fire in the Liquid H2 storage tank vent stack.Which ONE of the following describes the INITIAL response the SRO should direct lAW AOP-29.0, Plant Fire;and contains only notifications

REQUIRED, lAW EIP-8.0, Non-Emergency

Notifications?

A.*Assemble the fire brigade and direct the Fire Brigade Leader to extinguish

the fire by spraying water directly on the hydrogen vent stack.*FNP Duty Manager and the Air Products company.B.*Assemble the fire brigade and direct the Fire Brigade Leader to extinguish

the fire by spraying water directly on the hydrogen vent stack.*Corporate Duty Managerandthe Nuclear Regulatory

Commission

Operations

Center (NRCOC).*Direct the Outside SO to use SOP-34.0 to extinguish

the fire by establishing

a.helium purge and isolating the leak.*FNP Duty Manager and the Air Products company.D.*Direct the Outside SO to use SOP-34.0 to extinguish

the fire by establishing

a helium purge and isolating the leak.*Corporate Duty Manager and the Nuclear Regulatory

Commission

Operations

Center (NRCOC).Meets 10 CFR 55.43 (b)due to EIP notifications

are the responsibility

of the SRO position.A.incorrect.

The first part is not correct.It is not directed since it would not secure the source of the hydrogen and may not put the fire out.In the case of a hydrogen vent stack fire, AOP-29 has the following on the symptoms and entry page: I.IF the fire is in the Liquid H2 storage tank vent stack, THEN go to FNP-0-SOP-34.0, section 4.10, HYDROGEN-OXYGEN SYSTEM.IF the actions of FNP-0-SOP-34.0

are unsuccessful,THEN

the Shift Supervisor

should enter FNP-O-AOP-

29.0 and at his/her discretion, assemble the fire brigade to respond.SOP-34 P&L 3.11 In the event of a fire at exit of vent stack do not spray water on the vent stack or safety relief valves.Allow fire to continue to burn at top of vent stack until hydrogen source is located, THEN extinguish

per section 4.10.Monday, January 14, 2008 2:43:59 PM 65

QUESTIONS REPORT for 25 SRO Questions The second part is correct lAW EIP-8.SOP-34 and EIP-13.Notification

of the FNP Duty Manager and the Air Products company is required (if the emergency involves a liquid hydrogen tank, a liquid oxygen tank, or associated

use systems.)by EIP-13.0 Fire Emergencies

and EIP-8.0 Non-Emergency

Notifications.

B.incorrect.

first part and second part is NOT correct.The Nuclear Regulatory

Commission

Operations

Center (NRCOC)is NOT required to be notified as delieniated

in EIP-8.0 P&Ls unless the fire is an emergency classification

per step'6.2.3 EIP-8.0.Some fires are Emergency classifications, but a vent stack fire is not.Corporate Duty Manager is correct for this event.There is no requirement

to file a non-emergency

report per EIP-8 (which would require notifying the NRCOC per figure 1)nor is there a requirement

to notify the NRCOC in SOP-34.C.Correct.This is the correct action and notifications

lAW AOP-29, SOP-34 would be used to extinguish

the fire per the below steps of SOP-34 4.10.1 Note tank pressure and attempt to quickly determine probable source of H2 leakage by frost.on lines from relief valves or purge lines.The most probable source of H2 leakage is PCV-3 which is set at 130 psig and is isolated by valve NSP14V757 (V-27).4.10.2 Open isolation valve on installed helium bottle, THEN establish helium purge of vent stack.4.10.3 Isolate or attempt to isolate leaking valve.EIP-8.0 6.0 Notification

for EIP-13,"Fire Emergencies" NOTE: Notifications

are required for all plant fires including.small fires and hydrogen vent stack fires.EXCEPTION:

Notifications

are not required for intentionally

set fires at th\e Fire Training Facility.6.1 The Shift Manager shall ensure the following are notified: 6.1.5 The FNP Duty Manager.6.1.8 Air Products, if the emergency involves a liquid hydrogen tank, a liquid oxygen tank, or associated

use systems.6.2 The ED/FNP Duty Manager shall notify: 6.2.2 Corporate Duty Manager.D.incorrect-first part is correct, notifications

is incorrect.

Monday, January 14,20082:43:59

PM 66

QUESTIONS REPORT for 25 SRO Questions G2.4.27 Knowledge of fire in the plant

Question Number: Tier 3 Group 4 99 Importance

Rating: 3.5 Technical Reference:

AOP-29.0 and SOP-34 and EIP-8.0 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.5 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER SRO GO Version: a123456789 Answer:CCBDB CADDD Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:43:59

PM 67

SRO QUESTION G2.4.44 The correct answer for this question is"c" and not"B." The distractor

analysis for"B" should read as follows: B.Incorrect:

First part incorrect (see first part of A,)second part incorrect.

For initial notifications

the Form"Guideline

1 11 states: (Unaffected

Unit(s)Status Not Required for Initial Notifications)

The answer analysis for"c" should read as follows: C: Correct: Notification

of Protective

Action Recommendations

is required to be completed for the Initial Notification

of a General Emergency.(Not required for any other classification

including Site Area Emergency).

Announcement

with evacuation

instructions

required per step II.A.2.of Guideline 2, EIP-9.0.

QUESTIONS REPORT..for 25 SRO Questions 25.G2.4.44 045 A Site Area Emergency was declared 35 minutes ago.Subsequently, conditions

have'degraded and a General Emergency classification

needs to be declared.When upgrading to the General Emergency classification, which one of the following contains ONLY required actions lAW FNP-O-EIP-9.0, Emergency Actions?A.*Sounding of the plant emergency alarm.*Announce needed evacuation

instructions

to plant personnel.*Sounding of the plant emergency alarm.*Notify Alabama and Georgia of the status of the unaffected

Unit.C.*Notify Alabama and Georgia of Protective

Action Recommendations.

Announce needed evacuation

instructions

to plant personnel.

D.*,Notify Alabama and Georgia of Protective

Action Recommendations.

Notify Alabama and Georgia of the status of the unaffected

Unit.Meets 10 CFR 55.43 (b)requirements

for SRO level question since the IR is a 2.1 for an RO and not required knowledge or an action for an RO.EIP-9.0 A: Incorrect:

This action is required by the General Emergency Guideline Procedure only when not already previously

performed.

The SRO must know that it was required, and was already sounded, for the SAE.Second part correct II.Emergency Director Actions NOTE: THE SHIFT MANAGER SHALL PERFORM THE DUTIES OF THE EMERGENCY DIRECTOR UNTIL HIS ARRIVAL AND ASSUMPTION

OF DUTIES..Initials A.Notify personnel on site 1.If the Plant Emergency alarm has not already been activated, then announce over the public address system"All Plant Personnel Report to Designated

Assembly Area," activate the PEA[Plant Emergency alarm]for 30 seconds and repeat the announcement.

2.Announce the classification, and the condition, request setup of the TSC and OSC and give needed evacuation

instructions

over plant public address system.Monday, January 14, 2008 2:43:59 PM 68

QUESTIONS REPORT for 25 SRO Questions

f:$,crma:i

ci!'enSfrMi,fm})mIt::=.s

i

..

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25.m:t:

Tf:f..:fE..O!\T......E

....3.

NUCLEA1Rm07HER._".6.

t2.,!Ufiii7

STATtJ!1t..:B: tti

i.Jt*Tt:me.

Date._{-:t_:*_

lfWt

f:*

!d'Tmm,e.-i f__}_'_

renart:s;D i'feD'3

QF:{.*B.Correct: Notification

of Protective

Action Recommendations

is required to be completed for the Initial'Notification

of a General Emergency.(Not required for any other classification

including Site Area Emergency).

Announcement

with evacuation

instructions

required per step II.A.2.of Guideline 2, EIP-9.0.C: Incorrect:

First part correct, second part incorrect.

For initial notifications

the Form IIGuideline

1 11 states: (Unaffected

Unit(s)Status Not Required for Initial Notifications)

D: Incorrect:

First part correct, second part incorrect.

G2.4.44 Knowledge of emergency plan protective

action recommendations.

Question Number: 100 Tier 3 Group 4 Importance

Rating: 4.0 Technical Reference:

EIP-9.0 Proposed references

to be provided to applicants

during examination:

NO Learning Objective:

10 CFR Part 55 Content: 43.5 Comments: Replaced the question with one that does not require reference material and is not a direct lookup.It is more closely related to what an SRO duty is in the emergency plan and required knowledge for an SRO.MCS Time: Points: 1.00 Version: a123456789 Answer:BB DAB B DDC C Scramble Range:A-D Monday, January 14, 2008 2:43:59 PM 69

Source: Cognitive Level: Job Position: reviewed: BANK LOWER SRO GO QUESTIONS REPORT for 25 SRO Questions Source if Bank: FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:43:59 PM 70

QUESTIONS REPORT for 75 RO Questions 1.001 AK2.01 001 Initial conditions (Time=1000)with Rod control in AUTO:*Tavg-Tref deviation is O°F and stable.*Pressurizer

level is 45%and stable.*Reactor Power is approximately

75%and stable.*Control Bank D step counters are at 144 steps.Current conditions (Time=1 002)with no load change in progress:*Tavg-Tref deviation is approximately

+2°F and rising.*Pressurizer

level 46%and slowly rising.*Pressurizer

spray valves have throttled open.*Reactor Power is approximately

76%and slowly rising.*Control Bank D step counters are at 150 steps and rising at 8 steps per minute.Which ONE of the following describes the event in progress;and then the FIRST action that must be performed lAW AOP-19.0, Malfunction

of Rod Control System?A.*Inadvertent

RCS dilution;*Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.

B.*Inadvertent

RCS dilution;*Place the rod control mode selector switch to MANUAL and match Tavg with Tref by inserting rods.C.*Uncontrolled

Continuous

Rod Withdrawal;

Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.*Uncontrolled

Continuous

Rod Withdrawal;

Place the rod control mode selector switch to MANUAL and verify that rod motion stops.Monday, January 14, 20082:42:14

PM 1

QUESTIONS REPORT for 75 RO Questions A is incorrect;

if an inadvertent

dilution were taking place, the rods would go in not OUT To trip the reactor at this point would be incorrect.

t B is incorrect;

See above for the first part.Second part is corre'ct lAW AOP-19 for a continuous

rod withdrawal.

C is incorrect;

is the correct accident, however, the action stated is the RNO'if rods do not cease moving once they have been placed in manual lAW AOP-19.D.is correct for the stated situation.

A CRW is taking place due to temperature

shows rods should actually be moving in due to high temperature

and the action is to place rods in Manual if they are stepping while in AUTO 001 AK2.01 Continuous

Rod Withdrawal

Knowledge of the interrelations

between the Continuous

Rod Withdrawal

and the following:

Rod bank step counters Question Number: 57 Tier 1 Group 2 Importance

Rating: 2.8 Technical Reference:

OPS 52201 E, AOP-19.0 Proposed references

to be provided to applicants

during examlnation:

None Learning Objective:

OPS52520S07

10 CFR Part 55 Content: 41.5 Comments: Fixed per FJE comments 10/4/2007 MCS Time: 1 Points: 1.00 Version:0123456789

Answer: DAB DABCBAB Scramble Range:A-D Source: BANK Source if Banle WBN BANK Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant: FARLEY reviewed: GO Previous 2 NRC exams: NO Monday, January 14, 20082:42:14

PM 2

QUESTIONS REPORT for 75 RO Questions 2.001 K5.36 001 During a power DECREASE, the change in power defect will add (1)reactivity

to the core.Assuming the operator does NOT borate or dilute, control-rod (2)will initially be required to maintain Tavg on program.A.(1)negative (2)insertion B.(1)negative (2)withdrawal(1)positive (2)insertion D.(1)positive (2)withdrawal

A.incorrect because power defect adds positive reactivity

on a power decrease.B.incorrect because power defect adds positive reactivity

on a power decrease.C.correct.Power defect adds positive reactivity

for a negative change in load.Curve 27 shows for 6000 MWO/MTU power defect will go from-1269 to-658.Positive reactivity

will cause Tavg to rise.Withnoboration, rods must be inserted.D.incorrect because rods must be inserted to maintain Tavg on program.Monday, January 14, 20082:42:14

PM 3

QUESTIONS REPORT for 75 RO Questions 001 K5.36 Control Rod Drive Systems Knowledge of the following operational

implications

as they apply to the CRDS: Significance

of sign.(always minus)of a calculated

power defect Question Number: Tier 2 Group 2 29 Importance

Rating: 3.1 Technical Reference:

T&AA, CORE PHYSICS CURVEs pcb-1-voI1-crv27, 34&60 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52510F04

10 CFR Part 55 Content: 41.1 Comments: Fixed per FJE comments 10/5/2007 MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GO Version:0123456789

Answer: C DBAAABDDC Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:14 PM 4

5 QUESTIONS REPORT for 75 RO Questions 3.002 K3.02 001.Which ONE of the following correctly describes the reason that, in the event of a design basis Large Break LOCA, the plant is realigned from Cold Leg Recirculation

to Simultaneous

Cold and Hot Leg Recirculation?

Art To prevent fuel temperatures

from i'ncreasing

due to boron precipitation

at the TOP of the core.B.To prevent fuel temperatures

from increasing

due to boron precipitation

at the BOTTOM of the core.C.To prevent a reduction in Shutdown Margin due to boron precipitation

at the TOP of the core.D.To prevent a reduction in Shutdown Margi'n due to boron precipitation

at the BOTTOM of the core.A.Correct.Hot Leg Recirc is aligned to Ibackflush

l the core due to boron precipitation

that occurs due to boil-off.B.Incorrect.

Concern is top of the core, not the bottom which will be covered with water and have continuos flow.C.Incorrect.

Shutdown Margin may be ultimately

affected, but core cooling and blockage ofchannels is the concern that Hot Leg Recirc addresses D.Incorrect.

Shutdown Margin may be ultimately

affected, but core cooling and blockage of flow channels is the concern that Hot Leg Recirc addresses, and the cO'ncern is at the top of the core Executive volume Rev 2 of ERG guidelines

The operators should continue with the guideline and transfer to cold leg recirculation (ES-1.3)when the RWST level reaches the switchover

setpoint.The plant engineering

staff may also recommend hot leg recirculation (ES-1.4, TRANSFER TO,.HOT LEG RECIRCULATION

), at a later time, if a boron precipitation

concern is possible.CONCERN(s)

Should the SI system be aligned for hot-leg recirculation

in order to prevent boron precipitation

in core?Boron precipitate

can plate out on the fuel cladding surface, thereby reducing heat transfer from the fuel to the coolant..This requirement

is conservative

in that for all cases except the design-basis

LOCA, the actual rate of boron concentration

within the core will be less than that assumed in the FSAR design-basis

calculation

of the time at which switchover

is required from cold-to hot leg recirculation.

This is due to the core boiling rate being less than that assumed in the calculation.

Additionally, for any LOCA smaller than the design-basis

LOCA, the saturation

temperature

will be higher than that assumed in the calculation, resulting in a boron precipitation

limit that is higher than assumed.These factors substantially

lengthen the Monday, January 14, 20082:42:14

PM

QUESTIONS REPORT for 75 RO Questions time to the onset of boron precipitation

within the core and the time before switchover

from cold-to hot-leg recirculation

is required.Conservative

analysis has shown that, following a large cold-leg break in the RCS, the boric acid concentration

limit established

by the NRC (the boric acid solubility

limit of 27.53%minus 4%for conservatism)

would be exceeded if cold leg recirculation

is maintained

for an extended period.The analysis considers the increase in boric acid concentration

in the reactor vessel during the long-term cooling phase of a LOCA assuming a conservatively

small effective vessel volume including only the free volumes of the reactor core and the upper plenum below the bottom*of the hot leg nozzles.This assumption

conservatively

neglects the mixing of boric acid solution with directly connected volumes,such

as the reactor vessel lower plenum.Effects of Break Location Cold Leg Break The calculation

of boric acid concentration

in the reactor vessel considers a cold I*eg break of the reactor coolant system in which steam is generated in the core from decay heat while the boron associated

with the boric acid solution is completely

separated from the steam and remains in the effective vessel volume.The cold leg safety injection flow is not effective in counteracting

this boiloff from the core since for larger breaks the downcomer level is low and the injection flow is primarily refilling the downcomer as opposed to the core, and no flushing of the core occurs.If the plant is transferred

from cold leg to hot leg recirculation

prior to the time the boric acid concentration

limit is reached in the reactor vessel, the hot leg safety injection flow will dilute the vessel boron concentration

by passing relatively

dilute boron solution from the hot leg through the vessel to the cold leg break location and will terminate boiloff from the core.This will prevent boron precipitation

in the core along with any resultant plateout on the fuel cladding which could reduce heat transfer from.the fuel to the reactor coolant..Lesson text ESP-1.3 OPS-52531 G Approximately

7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following the loss of coolant accident (LOCA), the cold leg recirculation

phase will be terminated

and the simultaneous

cold and hot leg recirculation

phase is initiated.

Switching to a hot leg recirculation

path will wash out the boron that may have plated out on the fuel rods at the top of the core.Maintaining

a leg recirculation

path provides a normal flow path through the core.If the boron were allowed to build up in the top of the core, it could reduce flow through the core*and degrade the heat transfer

of the fuel.This would also result in a depletion of the boron concentration

in the recirculated

fluid from the sump.Monday, January 14, 2008 2:42:14 PM 6

QUESTIONS REPORT for 75 RO Questions 002 K3.02 Reactor Coolant System Knowledge of the effect that a loss or malfunction

of the ReS will have on the following:

Fuel Question Number: Tier 2 Group 2 30-Importance

Rating: 4.2 Technical Reference:

OPS-521 028 Lesson text ESP-1.3 OPS-52531G, Executive volume Rev 2 of ERG guidelines

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS521 0280310 CFR Part 55 Content: 41.7 Comments: Fixed per FJE comments 10/5/2007 MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GO Version:0123456789

Answer:ACBB ABC B C A Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:15

PM 7

QUESTIONS REPORT for 75 RO Questions 4.003 A1.04 001 Given the following:

Unit 1 is in Mode 4.*1A RCP has just been started.*A CCW leak is occurring in the tube section of the upper bearing oil cooler of the 1A RCP.Which ONE of the following correctly describes the effect on the1A RCP Oil Reservoir level and the MINIMUM motor bearing temperature

that requires tripping the RCP?Art B.c.D.Oil Reservoir Level INCREASES INCREASES DECREASES DECREASES MINIMUM Temperature

195 0 F A.correct.CCW would leak into the bearing oil reservoir because it is at a higher pressure.The correct temperature

to trip the RCP is HG1 2.IF any 1A Rep motor bearing temperature

exceeds 195°F, THEN perform the following actions: a)Trip the reactor, AND go to FNP-I-EEP-O.O, REACTOR TRIP OR SAFETY INJECTION.

b)Stop IB RCP.c)Perform the actions required by FNP-I-AOP-4.0, LOSS OF REACTOR COOLANT FLOW.d)Manually close pressurizer

spray valve, PK 444C B.incorrect due to temperature

setpoint.KK5 PHASE 1 alarm setpoint 275°P MFG max safe operating temp.302°P c.incorrect because the reservoir level will be high, not low.D.incorrect because the reservoir level will be high, not low.195°F correct per ARP, 302°F is Plausible because per ARP-1.10, KK5, Max temperature

for Rep Motors is 302°F.On a complete Loss of CCW Flow to RCP Motor Bearing Oil Coolers, the bearing temperatures

will exceed 195°F in approximately

2 minutes.Monday, January 14, 20082:42:15

PM 8

QUESTIONS REPORT for 75 RO Questions 003 A1,,04 Reactor Coolant Pump System (RepS)Ability to predict and/or monitor changes'in parameters (to prevent exceeding design limits)associated

with operating the RepS controls including:

Rep oil reservoir le,vels Question Number: Tier 2 Group 1 Importance

Rating: 2.6 Technical Reference:

HG1&HH1 ARP-1.8, AOP-4.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52520107

10 CFR Part 55 Content: 41.5 Comments: Fixed per FJE comments 10/5/2007 MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER RO GO Version:0123456789

Answer:ABBD DABBDA Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:15

PM 9

QUESTIONS REPORT for 75 RO Questions 5.003 K5.03 001 Unit 1 is at 6%reactor power when1B Rep trips.Which ONE of the following describes the INITIAL response of Tavg in the 1 BLoop and the reason for that response with no operator action?A.INCREASE because Thot in the unaffected

loops INCREASES.

B.INCREASE due to the reverse flow of primary coolant in the 1 Bloop.C.DECREASE because Tcold in the unaffected

loo.ps DECREASES.DECREASE due to the reverse flow of primary coolant in the 1 Bloop.Monday, January 14,20082:42:15

PM 10

QUESTIONS REPORT for 75 RO Questions A.Incorrect;

1 Bloop Tave will not increase even though the unaffected

loops Tavg will increase.B.Incorrect;

because BLoop Tavg will not increase.Plausible because the applicant may misunderstand

Thot and Tcold values for reverse flow in a loop.C.Incorrect;

1 Bloop Tavg will decrease but the unaffected

loops Tc will increase.o.Correct;Tavg will initially decrease due to reverse flow, which occurs when the 1 B Loop RCP is tripped.OPS-525200 If the reactor is less than 30%power and there is a loss of coolant flow in one loop (two or more loops if below 10%power), the operator must respond in anefficientmanner

in order to minimize the effects on primary and secondary systems.In the loop that has lost coolant flow, temperatures

will stabilize at approximately

the cold leg temperature.(TC)of the unaffected

loop(s).This will drop the saturation

temperature

and pressure of the affected loop's steam generator (SG), causing SG level to drop (shrink), and will also reduce the amount of steaming and power output from the affected SG to a minimum.The loop flow indications

observed by the operators would be as follows: For the affected loop, flow would slowly decrease to 0 and then return to approximately

10%;for the unaffected

loops, the flow should increase to approximately

105%(each loop).The flow indication

in the idle loop occurs as flow stops and then begins again in the reverse direction.

Since flow rates in the ReS loops are derived from the differential

pressure felt in an elbow in each loop, any flow at all will be indicated, regardless

of the direction.

The indication

observed in the two loops with the running pumps is due simply to the pumps in those loops picking up'a small portion of the flow lost in the idle loop.Monday, January 14, 2008 2:42:15 PM 11

QUESTIONS REPORT for 75 RO Questions 003 K5.03 Reactor Coolant Pump System (RepS)Knowledge of the operational

implications

of the following concepts as they apply to the RepS: Effects of Rep shutdown on T-ave., including the reason for the unreliability

of T-ave.in the shutdown loop Question Number: Tier 2 Group 1 2 Importance

Rating: 3.1 Technical Reference:

OPS 52520D Proposed references

to be provided to applicants*

during examination:

None Learning Objective:

OPS40301 AOa10 CFR Part 55 Content: 41.7 Comments: Fixed per FJE comments 10/5/2007 meets the KA in that this addresses the operational

implications

of the loss of a RCP on Tavg and the reason the temperature

is reading below the other 2 loops (ie., not reliable or different from the operating loops)Scramble Range:A-D FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK LOWER RO GO Version: a123456789 Answer: DB CDDDDBCC Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:15

PM 12

QUESTIONS REPORT for 75 RO Questions 6.004 A4.05 001 Given the following:

Unit 1 is in Mode 5.*Solid plant operations

are in progress.*HIK-142, RHR TO LTDN HX, has been adjusted to the full open position.*FK-122, CHG FLOW, is in MANUAL.*PK-145, LTDN PRESS, is in MANUAL.The OATC lowers the demand on PK-145.Which ONE of the following describes the effect on PCV-145, Letdown PCV, and RCS pressure?PCV-145 throttles__(_1)__, RCS pressure (2.-..)_A.(1)OPEN B.(1)CLOSED(1)OPEN D.(1)CLOSED (2)INCREASES (2)INCREASES (2)DECREASES (2)DECREASES A.incorrect-

PCV-145 WILL open to decrease pressure when the controller

is taken to the the lower position.Due to the location of the valve in the system and with HCV-142 fully open when PCV-145 is opened RCS pressure will drop with no change in charging flow.B.incorrect-The valve will open, not close.Distractor

is credible because changing demand does change valve position, and it is easy to associate lowering demand with valve closure C.correct.Reducing the demand on PK-145 in manual will cause the valve to open, reducing backpressure

on the letdown line, therefore reducing RCS pressure upstream.D.incorrect-The valve will open, not close.Distracter

is credible because changing demand does change valve position, and it is easy to associate lowering demand with valve closure 0%demand on the controller

=lower system pressure and the valve will open 100%demand on the controller

=higher system pressure and the valve will close Monday, January 14, 20082:42:15

PM 13

QUESTIONS REPORT for 75 RO Questions 004 A4.05 Chemical and Volume Control System Ability to manually operate and/or monitor in the control room: Letdown pressure and temperature

control valves Question Number: Tier 2 Group 1 4 Importance

Rating: 3.6 Technical Reference:

CVCS LP OPS-521 01 F Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS40301 FOa 10 CFR Part 55 Content: 41.5 Comments: MCS Time: Source:.Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER RO GO Version: a123456789 Answer:CDBCCCBCBA

Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42: 15 PM 14

QUESTIONS REPORT for 75 RO Questions 7.004 K2.0l 001 Which ONE of the following states the power supply to1A Boric Acid Transfer Pump?600 Volt MCC----- B.1B C.1D D.1E A is correct.per the load list for unit 1, page F-92,1A Boric Acid Transfer pump comes off FAC4, which comes from ED10 and from DF03.B is incorrect.

Plausible because it supplies power to1B BAT pump.(FBB4 on MCC1B which comes from EE10 and DG03)C is incorrect.

Plausible because it supplies power to CVCS components:1A Charging pump Aux Lube Oil pump HDL5.This MCC is on the rad side aux bldg and supplies many rad side AB loads.D is incorrect.

Plausible because this MCC is on the non-rad side aux bldg but supplies some rad side AB loads such as the Boric acid batching tank cond return unit and power to CVCS components:

1 Band1C Charging pump Aux Lube Oil pumps from HEK2 and K3.Monday, January 14, 20082:42:15

PM 15

QUESTIONS REPORT for 75 RO Questions 004 K2.01 Chemical and Volume Control System Knowledge of bus power supplies to the following:

Boric acid makeup pumps Question Number: Tier 2 Group 1 3 Importance

Rating: 2.9 Technical Reference:

OPS 521011, F,&G, FNP-Unit 1 Load List A-506250 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS40301104

10 CFR Part 55 Content: 41.5 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GTO Version:0123456789

Answer:AC CADBCCCC Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:15

PM 16

QUESTIONS REPORT for 75 RO Questions 8.005 K6.03 001 Which one of the following would prevent the 1 A(B)RHR heat exchangers

from performing

their design function?A.A loss of air to Heat Exchanger discharge valves HCV-603A and HCV-603B.Br Closing the component cooling water outlets from the RHR heat exchangers

during Mode 3 operation.

C.Closing the manual valve to HCV-142, RHR Discharge to CVCS Letdown Line, during Mode 5 solid plant operation.

D.A loss of air to Heat Exchanger bypass valves FCV-605A and FCV-605B.Distractor

analysis: A: Incorrect-Loss of air to the HCV-603 I s, HXs discharge valves, will not prevent the HX from performing

their design function, since these valves fail open the HXs are still available.

B: Correct-CCW system must be able to provide flow through the RHR HXs in order for them to perform their design function of removing RCS heat to facilitate

cooldown from 350 to 140 within 16 hrs.C: Incorrect-Manually closing this valve during solid plant operation will result in a loss of UD and may cause a pressure increase in the RCS, however, this is not the design function of the RHR HX.D: Incorrect-Loss of air to the FCV-605 I s, HXs bypass valves, will not prevent the HX from performing

their design function, since these valves fail closed the HXs are still available.

REFERENCES:

.1.FNP-1-S0P-7.0, RESIDUAL HEAT REMOVAL SYSTEM 2.

COMPONENT COOLING WATER Also found in bases bases 3.4.6 In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG)secondary side coolant or the component cooling water via the residual heat removal (RHR)heat exchangers.

The seconda ry f1lnption ofihe reactor

soluble n'eutron poison, boric acid.bases 3.4.7 In MODE 5 with the ReS loops filled, the primary function of the Monday, January 14, 20082:42:15

PM 17

QUESTIONS REPORT for 75 RO Questions reactor coolant is the removal of decay heat and transfer this heat either to the steam generator (SG)secondary side coolant via natural circulation (Ref.1)or the component cooling water via the residual heat removal (RHR)heat exchangers.

While the principal means for decay heat removal is via the RHR System, the SGs via natural circulation (Ref.1)are specified as a backup means for redundancy

bases 3.7.7 In-MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger.

bases 3.9.4 The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant and to.prevent boron stratification (Ref.1).Heat is removed from the RCS by circulating

reactor coolant through the RHR heat exchanger(s), where the heat is transferred

to the Component Cooling Water System 005 K6.03 Knowledge of the effect of a loss or malfunction

on the following will have on the RHRS: RHR heat exchanger Question Number: 5 Tier 2 Group 1 Importance

Rating: 2.5 Technical Reference:

1.FNP-1-S0P-7.0, RESIDUAL HEAT REMOVAL SYSTEM 2.OPS-52102G-40204A

COMPONENT COOLING WATER Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS40301 K06 10 CFR Part 55 Content: 41.7 BANK LOWER RO GO Source: Cognitive Level: Job Position: reviewed: Comments: This exact question has been used on 3 NRC exams, 2002 surry exam, 2006 summer exam and 2006 FNP exam for the same KA.It tests the knowledge of the loss of the RHR ht exchanger (ie.cooling function)has on the design basis for the ht exchanger.

MCS Time: 1 Points: 1.00 Version: a123456789________________....=-DI-::J---------

Source if Banlc FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: YES , Monday, January 14, 20082:42:15

PM 18

QUESTIONS REPORT for 75 RO Questions 9.006 A4.01 011 Given the following:

Unit 1 was operating at 100%power.*B Train is on service with1B charging pump running.*An SI/LOSP has just occurred.*At 22 seconds after the SI/LOSP actuation annunciator

EB1, CHG PUMP OVERLOAD TRIP, comes into alarm.*The operator notices the amber light on the handswitch

for the1C Chg pump.Which ONE of the following is correct concerning1B Chg Pump?1B Chg Pump_A.must be manually started.B.will start from the LOSP sequencer.

C!'will start due to1C Chg Pump tripping on overload.D.will remain running throughout

the event per design.Monday, January 14, 2008 2:42: 16 PM 19

QUESTIONS REPORT for 75 RD Questions A.incorrect;

On the LOSP, the1B chg pump will load shed.1B charging pump does not need to be manually started since when the1C CHG pump trips, the1B pump should automatically

start due to an overload trip.B.Incorrect.

Plausible because the sequencer will only start the1B charging pump if the1C charging pump breaker is racked out or has tripped on overload.After 22 seconds have passed the sequencer will be at about step 2 of returning equipment to service.Once a step is complete, the sequencer signal is no longer available to start any other component on a previous step.Charging pumps come off step 1 and this will occur about 17 seconds into the event.C.Correct.The sequencer sequences on the1C chg pump (unless it is racked out, then it would sequence on the 1 B)'after about 17 seconds (approx.12

secs for DG to start and tie on, no more than 5 secs for sequencer to start load.).Then, an overload trip of1C will cause1B chg pump (when aligned to same train)to auto start.In this case B Train is on service so1B chg pump is aligned to the B train with1C Chg pump.P&L of SDP-2.1 3.30 If the on-service

charging pump trips on overload, the off-service

charging pump for the particular

train which has two operable charging pumps will automatically

start.3.31 If lA (IC)Charging Pump trips on overload or is racked out, IB Charging Pump will automatically

start upon safety injection or loss of offsite power.D.Incorrect.

because if there was an SI with no LOSP, the SI Sequencer would leave1B chg pump running, and would not load shed1B Chg pump and not start 1 C.006 A4.0.1 Chemical and Volume Control System Ability to manually operate and/or monitor in the control room: Pumps Question Number: 7 Tier 2 Group 1 Importa'nce

Rating: 4.1 Technical Reference:

CVCS LP OPS-521 01 F FNP-1-ARP-1.5

EB1 SOP-2.1, CHEMICAL AND VOLUME CONTROL SYSTEM PLANT STARTUP AND OPERATION version 84 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS521 01 FOe 10 GFR Part 55

Comments: Monday, January 14, 2008 2:42: 16 PM 20

FARLEY NO Scramble Range:A-D FARLEY QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer: C BADCDBCBA Source if Banle Difficulty:

Plant: Previous 2 NRC exams: BANK HIGHER RO GO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:16

PM 21

QUESTIONS REPORT for 75 RO Questions 10.006 K3.01 001 Given the following:

A.small break LOCA has occurred on Unit 1 and EEP-O, Reactor Trip or Safety Injection, is in progress.*Sub Cooled Margin Monitor is reading 38°F.*Containment

pressure is 9 psig.*ALL RCPs are running.*FI-943, A TRN HHSI FLOW, indicates°GPM.Which ONE of the following describes how the RCPs must be operated lAW EEP-O and the reason?A':'RCPs must remain operating to provide core cooling.B.RCPs must remain operating to simplify RCS temperature

and pressure control during plant recovery.C.RCPs must be tripped to prevent damage to the RCPs seals due to the loss of seal injection flow.D.RCPs must be tripped to prevent excessive loss of RCS water inventory and to keep the core covered.Monday, January 14, 20082:42:16

PM 22

QUESTIONS REPORT for 75 RO Questions A.Correct.RCPs may not be tripped because there is no HHSI flow per EEP-O Fold out page.RCP Trip Criteria RCP trip criteria have been developed and incorporated

into the ERPs to provide for RCP trip when required for Small Break LOCAs and to minimize the probability

of RCP trip when not required.The RCP trip criteria consist of two fundamental

parts:.Successful

operation of the SI system AND*Subcooling

less than 16°F{45°F}In the ERPs, the RCPs are not tripped unless this two-part criterion" is satisfied.

The following summary is provided from the RCP TRIP/RESTART

document: If RCPs continue to operate during a small break LOCA, the_forced circulation

provides core cooling, but also results in greater loss of coolant inventory due to continued discharge of saturated liquid (rather than steam)from the break.Continuous

operation of the RCPs during a LOCA cannot be guaranteed

since tripping of the RCPs would occur upon a loss of offsite power or other essential support conditions

which could occur at any time.The reason for purposely tripping the RCPs during an accident (when the RCP trip criterion is met)is to prevent excessive loss of RCS water inventory through a small break which might lead to severe core uncovery if the RCPs were tripped for some reason later in the accident.B.incorrect.

although RCPs do remain running the"reason is not to make it easier to control temperature

and pressure with no charging flow.C.Incorrect.

Would be correctifHHSI flow was being indicated.

D.incorrect, RCPs would be tripped if HHSI flow was being indicated.

A caution in E-O says the following:.CAUTION: RCP seal degradation

may occur if seal injection flow is not maintained

to all RCPs.This could be used to trip the RCPs if Seal injection were lost, however, core cooling is.more important at this time due to the loss of HHSI flow.Monday, January 14, 20082:42:16

PM 23

QUESTIONS REPORT for 75 RO Questions 006 K3.01 Emergency Core Coolant System Knowledge of the effect that a loss or malfunction

of the ECCS will have on the following:

ReS Question Number: Tier 2 Group 1 6 Importance

Rating: 4.1 Technical Reference:

EEP-O.O foldout page Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52530A0310 CFR Part 55 Content: 41.1 0 Comments: This tests the ability to determine what to do with RCPs running and a loss of subcooling

with a S8 LOCA in EEP-O, and the effects on the RCS of that decision in relation to a loss of HHSI,.which meets the KA above.MCS Time: Source:

Level: Job Position: reviewed: Points: 1.00 NEW HIGHER RO GO Version: a123456789 Answer:AABBB CADDA Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:16

PM 24

QUESTIONS REPORT for 75 RO Questions 11.007 K5.02 001 The crew is forming a pressurizer

steam space (drawing a bubble)per UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby.The vacuum refill procedure will NOT be performed.

Unit 1 is in Mode 5 maintaining

325-375 psig.*1B RCP is running.*A Train RHR is on service with low pressure letdown aligned.*RCS is in solid plant pressure control with pressurizer

temperature

at 178°F.*All PRZR heaters have been energized.

Which ONE of the following correctly describes the condition that will indicate when the pressurizer

is at saturation

conditions (ie.a bubble is ready to be formed)lAW UOP-1.1;and the effect on PRT level during this evolution?

A.*Letdown flow decreases;

PRT level will remain constant.*RCS Pressure will increase;*PRT level will remain constant.C.*RCS Pressure will increase;*PRT level will rise.D.*Letdown flow decreases;

PRT level will rise.Monday, January 14, 20082:42:16

PM 25

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

Plausible, because RCS pressure will start to rise and letdown flow will increase as pressure starts to rise.The candidate may not know what to expect from the letdown flow as they may not know the position of PCV-145, LETDOWN PCV, FCV-122, CHG FLOW REG, and HCV-142, RHR TO LETDOWN LINE.The PRT parameters

will remain constant.B.Correct.lAW step 5.10, letdown flow will increase as RCS pressure increases.

The PRT parameters

will remain constant since the liquid from the pzr is diverted to the RHTs.Uop-1.1 step 5.10 WHEN pressurizer

temperature

increases to the saturation

temperature

for 375 psig (approximately

442°F)as indicated by increasing

RCS pressure or letdown flow, THEN establish a steam space in the pressurizer

as follows: UOP-1.1 shows that the liquid from the pressurizer

will go to the RHTs.There will be no level increase or liquid that will go to the PRT.5.10.5 WHEN VCT level increases to 81%, THEN verify VCT HI LVL IDIVERT'VLV

Q1 E21 LCV115A in the fully diverted position.C Incorrect.

First part is correct.second part is NOT correct.see above.D.Incorrect.

both first and second part are not correct.007 K5.02 Pressurizer

Relief Tank Knowledge, of the operational

implications

of the following concepts as the apply to PRTS: Method of forming a steam bubble in the PZR Question Number: 8 Tier 2 Group 1 Importance

Rating: 3.1 Technical Reference:

UOP 1.1, RHR FSD A-181002 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS40301 F0610 CFR Part 55 Content: 41.1 0.Comments: In order to meet the KA, the PRT had to be used to in some fashion.Since the PZR liquid is directed to the RHTs as the level is being decreased, the PRT level, temp and pressure will be unaffected.

To meet the KA;the method of forming the bubble is addressed by the indications

that will be available when the steam space justbegins

to be formed.Operational

'implications

of the PRT are none so level, pressure and temp will remain constant.MCS Time: Points: 1.00 Version:0123456789

Answer: B AAACBBBBA Scramble Range:A-D Monday, January 14, 20082:42:16

PM 26

Source: Cognitive Level: Job Position: reviewed: NEW mGHER RO GO QUESTIONS REPORT for 75 RO Questions Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14,20082:42:16

PM 27

QUESTIONS REPORT for 75 RO Questions 12.008 A2.08 001 Given the following:

Unit 1 is at 1 00%power.*The temperature

input to TCV-3083 (also called TCV-144), L TDN HX CCW DISCH TCV, fails low.*The OATC reports that TI-144, CCW L TDN HX,

indicator, is reading at the bottom of the scale (50°F)due to the temperature

input failure.The consequences

of the failure is a small RCS1;and the action required by the OATC would be to use MCB TK-144, L TDN HX OUTLET TEMP, in Manual Control and 2 CCW flow.A.1.dilution 2.increase B.1.dilution 2.decrease1.boration 2.increase D.1.boration 2.decrease Monday, January 14, 20082:42:16

PM 28

QUESTIONS REPORT for 75 RO Questions From SOP-23.0: CAUTION: CCW temperature

should be maintained

as stable as possible due to the effects on reactivity

due to changes in letdown temperature.

Also, changing CCW temperature

could affect RCP oil levels which could cause level annunciators

to come in.From SOP-2.1 rev 84 CAUTION: Changes in letdown temperature

can have a significant

effect on reactor power.Care should be taken to closely coordinate

changes in CCW flow between personnel at L TDN HX CCW TEMP CONT, QIP17TV3083, and Control Room personnel at L TDN HX OUTLET TEMP TK144.Letdown Temperature

controller, TK-144, failed low.The controller

senses a lower temperature

and sends a signal to the CCW valve toclosedown to provide less cooling to raise the temperature

of Letdown.When Letdown temperature

goes up, the demineralizers

have less affinity for boron, and some of the boron in the demineralizers

is released.This is a boration effect.ARP'S DF5&DF1 both direct taking manual control of TK-144 when needed to control temperature.

A.incorrect;

a dilution will not occur, increasing

is correct.B.incorrect;

a dilution will not occur.Decreasing

CCW flow is not correct even though it would be done using TK-144 IN MANUAL.It is plausible because the TCV failure does cause a temperature

change, just opposite from the change that will cause a dilution.C.Correct.a boration will occur and increasing

CCW.flow to the Ht exchanger using TK-144 IN MANUAL is the correct answer.As letdown water heats up, boron will be released in ion exchangers, resulting in a small boration.D.incorrect;

a boration will occur, Decreasing

CCW flow is not correct.Monday, January 14, 20082:42:16

PM 29

QUESTIONS REPORT for 75 RD Questions 008 A2.08 Component Cooling Water System Ability to (a)predict the impacts of the following malfunctions

or operations

on the CCWS, and (b)b.ased on those predictions, use procedures

to correct, control, or mitigate the consequences

of those malfunctions

or operations:

Effects of shutting (automatically

or otherwise)

the isolation valves of the letdown cooler Question Number: Tier 2 Group 1 9 Importance

Rating: 2.5 Technical Reference:

CVCS LP, SOP-2.1, Sec 4.18&4.19 cautions Proposed references

to be provided to applicants

during examination:

'None Learning Objective:

OPS521 01 F0210 CFR Part 55 Content: 41.7 Comments: This questionmeetsthe KA in that it is a failure closed of the CCW isolation valve to a cooler and has operational

procedures

and actions to combat the event.Scramble Range:A-D FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 MODIFIED HIGHER RO GO Version:0123456789

Answer: CBAB CADCAD Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:16

PM 30

QUESTIONS REPORT for 75 RO Questions 13.008 AK2.01 005 Given the following plant conditions:*A reactor trip and safety injection have occurred.*RCS pressure is1050 psig and lowering.*Tavg is 550°F and lowering.*Pressurizer

level is 65%and rising rapidly.*Containment

pressure is 2 psig and rising.Which ONE of the following describes the cause of this event?A.Letdown line break.B.Small Break LOCA on an RCS cold leg.Stuck open pressurizer

PORV.D.Stuck open pressurizer

spray valve.A&B.Incorrect.

PZR level would be lowering or off-scale low if either of these events occurred.C.Correct.A vapor space LOCA isoccurring,due

to RCS pressure lowering and Containment

pressure rising with PZR level rising.D.Incorrect because spray valve failure would not result in containment

pressure rising.008 AK2.01 Pressurizer

Vapor Space Accident-Knowledge of the interrelations

between the Pressurizer

Vapor Space Accident and the following:

Valves Question Number: 39 Tier 1 Group 1 Importance

Rating: R02.7 Technical Reference:

AOP-100, HC1 ARP-1.8 HE3 and 4 and 5 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52201 H1210 CFR Part 55 Content: Comments: This was originally

written as a spray valve failure This is not a vapor space

not meet the KA.Rewritten to meet the KA for vapor space accident and a valve issue.Subsequent

comments from FJE made question unsat, swapped for question on 2004 Robinson NRC exam and also on VC Summer 2007 NRC exam.Monday, January 14, 2008 2:42: 16 PM 31

FARLEY Scramble Range:A-D FARLEY QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer: CB CADAAAB A Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: BANK HIGHER RO GO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42: 16 PM 32

QUESTIONS REPORT for 75 RO Questions 14.o09EAl.17001

Given the following:*A reactor trip and safety injection have occurred.*RCS pressure is 1450 psig.*Containment

pressure is 7.5 psig.*SG pressures are 1000 psig.*All equipment has operated as designed.Which ONE of the following would have a rising level due to the RCP#1 seal return flow?A.VCTPRT C.RCDT D.Containment

Sump A is incorrect.

credible because it is the normal#1 seal flowpath.B is correct.Containment

isolation will isolate seal return flow, and the seal return relief valve will lift and direct the flow to the PRT.C is incorrect.

credible because it is the#2 seal flowpath.D is incorrect.

credible because#3 seal flow path is directed to the Ctmt sump.Monday, January 14, 20082:42:17

PM 33

QUESTIONS REPORT for 75 RO Questions 009 EA1.17 Small Break LOCA-Ability to operate and monitor the following as they apply to a small break LOCA: PRT Question Number: Tier.1 Group 1 Importance

Rating: 40 R03.4 Technical Reference:

OPS-521 01 F EEP-O attachment

3 figure 1 drawing 0-175039 sheet 1 and 0-175037 sheet 2 0-7 Proposed references

to be provided to applicants

during examination:

Learning Objective:'

OPS40301 F0510 CFR Part 55 Content: Comments: I changed the stem from describes where the the RCP seal return flow is being directed to would have a rising level due to the Rep seal return flow to meet the ability to monitor the PRT parameters

piece of the KA.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK LOWER RO GTO Version: a123456789 Answer:BCCCBDCBBB

Scramble Range:A-D Source if Bank: SONGS 2005 Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14,20082:42:17

PM 34

QUESTIONS REPORT for 75 RO Questions 15.010 K3.0l 002 Given the following:

Unit 1 is at 100%power.*All control systems are in their normal alignments.

PT-445, Pressurizer

Pressure Channel fails HIGH.Which ONE of the following describes the initial effect on RCS pressure and the reason for that effect?RCS pressure_A.rises due to ONLY Variable heaters energizing.

B.rises due to ALL Backup and Variable heaters energizing.lowers due to one PRZR PORV opening.D."lowers due to ALL Backup and Variable heaters de-energizing, and both spray valves and one PORV opening.A incorrect;

because pressure will lower initially.

Credible because it is consistent

with a controller

failure, which could be confused with an input failure from PT-444.Also the heaters will come on when pressure starts dropping from PK-444A.The pressure will not initially drop and will not rise until the PORV closes (cycles at 2000#)B incorrect;

see above, only all heaters are involved and could be confused between a level control failure, PT 444 failure and this failure.C Correct;When PT-445 fails high, PORV 445A opens and will drop pressure to 2000 psig.The valve will close per design at 2000 psig which comes from PT-455, 456 and 457 on 2/3<2000 psig.The PORV will cycle at 2000 psig if a rx trip and SI did not occur.D incorrect.

Would be correct for PT-444 failing high.The lesson plan says that all heaters will turn on when the RCS pressure drops, sprays will close and PORV 444B will remain closed.Monday, January 14, 20082:42:17

PM 35

QUESTIONS REPORT for 75 RO Questions 010 K3.01 Pressurizer

Pressure Control System Knowledge of the effect that a loss or malfunction

of the PZR PC,S will have on the following:

ReS Question Number: Tier 2 Group 1 Importance

Rating: Technical Reference:

10 3.8 AOP-100, OPS-52201 H Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52201 H1710 CFR Part 55 Content: 41.7 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW IDGHER RO GO Version:0123456789

Answer:CDB DCACCCB Scramble Range:A-D Source if Barne Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO'Monday, January 14, 20082:42:17

PM 36

16..011 A3.03 002 QUESTIONS REPORT for 75 RO Questions Given the following conditions:

The plant is stable at 90%power.*Charging, Letdown, and Pressurizer

Level Control systems are in automatic.

The Pressurizer

Level Selector Switch is in the 1/11 Position.*L T-459, Pressurizer

level Transmitter, has failed low.Which one of the following describes the system response?No operator action is taken Charging flow will

__and letdown flow will__-..,;;;;;;(2;."...)

_A.(1)increase (2)remain the same(1)increase (2)decrease C.(1)decrease (2)remain the same D.(1)decrease (2)decrease Monday, January 14, 2008 2:42: 17 PM 37

QUESTIONS REPORT for 75 ROQuestions

A.Incorrect.

charging flow increases through FCV-122, but letdown will isolate and flow will drop to zero.B.correct-Charging flow will increase and letdown will isolate and flow will drop to zero.C.Incorrect.

Charging flow will increase due to indicated PRZR level low and letdown will isolate and flow will drop to zero.D.Incorrect.

Charging flow will increase due to lower indicated PRZR level and letdown will isolate and flow will drop to zero.Reference:

CFR: 41.7/45.5 OPS-52201 H (in part 459 (I)LowLow level alarm LCV-459 closes Orifice isolation valves close All pressurizer

heaters turn off Charging flow increases to maximum Actual pressurizer

level increases because of secured letdown and maximum charging flow High level alarm from channel 460 (III)Reactor trip on high pressurizer

level if no

action is taken 011 A3.03 Pressurizer

Level Control System Ability to monitor automatic operation of the PZR LeS, including:

Charging and letdown Question Number: 31 Tier 2 Group 2 Importance

Rating: 3.2 Technical Reference:

PZR level/Press

LP OPS-521 01 E&H, UOP-3.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52201 H15 10 CFR Part 55 Content: 41.7 Comments: changed out questior riJtcitwas

not correct.This meets the KA in that the questions has the candidate monitor letdown and charging flows and the affects on one other system during this failure.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: BCD ABC DAB C Scramble Range:A-D Monday, January 14, 20082:42:17

PM 38

Source: Cognitive Level: Job Position: reviewed: MODIFIED HIGHER RO GO QUESTIONS REPORT for 75 RO Questions Source if Banle FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:17 PM 39

QUESTIONS REPORT for 75 RO Questions 17.012 A2.02 001 At 1 0:00 plant conditions

were as follows:*Unit 1 was at 41%power, ramping down due to RCS leakage greater than Tech Spec limit.*120V AC vital panel1A was been de-energized

2 hrs ago due to damage to the breaker panel.*DF01,1A S/U transformer

to 1F 4160V bus, tripped open.At 10:10, a Large Break LOCA occurred.Which ONE of the following describes the (1)the status of the Reactor Trip Breakers at 10:05;and (2)the action(s)required lAW EEP-O, Reactor Trip or Safety Injection, concerning

ESF components?

At 10:05 the Reactor Trip Breakers will be (1)After the LBLOCA, the operator is required to manually align (2)A':'(1)open (2)"A" Train ESF components

ONLY.B.(1)open (2)BOTH trains of ESF components.

C.(1)closed (2)"A" Train ESF components

ONLY.D.(1)closed (2)BOTH trains of ESF components.

A.Correct, The RTBs will open due to the loss of power from the solas to the RCP Single loop loss of flow (SLLOF)to 2 RCPs on A Train.At 41%power, this will result in a Rx trip.Then due to the loss of the vital panel, the"A" Train ESF components

will not actuate since the II A" train output relay cabinet slave relays will not actuate due to the loss of power.FSD A-181 007 2.Reactor Coolant Pump Breaker Trip Opening of one or two reactor coolanLpump-breakers

upon power level), which is indicative

of an imminent loss of coolant flow in the loop or loops, will cause a reactor trip.If two of three pump breakers trip with plant power>P-7 10%RTP, a reactor trip will occur.Below P-7 (10%RTP), the trip is Monday, January 14, 20082:42:17

PM 40

QUESTIONS REPORT for 75 RO Questions automatically

blocked.Also if 1/3 pump breakers are opened with plant power>P-8 (30%RTP), a reactor trip will occur..(References

6.1.003, 6.4.007, 6.7.012)0p8-52201 RPS Low Flow or Rep Breaker Open Trip There are three low flow protection

bistable status lights for each loop (total of nine lights)on TSLB-2.There is also a protection

bistable status light for each reactor coolant pump (RCP)breaker on TSLB-2.A low flow condition in any loop as detected by the open RCP breaker or 2/3 low flow signals will energize the respective

loop low flow partial reactor trip alarm, A(B, C)RCS LOOP FLOW LO OR A(B, C)RCP BKR OPEN.If reactor power is greater than the P-8 setpoint (30 percent), the low flow condition will cause a reactor trip.This is indicated by the ONE LOOP LO FLOW OR RCP BKR OPEN RX TRIP alarm.If reactor power is less than the P-8 setpoint but greater than the P-7 setpoint, a low flow condition in 2/3 loops will cause a reactor trip.This is indicated by the TWO LOOP LO FLOW OR RCP BKRS OPEN RX TRIP alarm.B.Incorrect-first part is correct.second part in not correct in that the master relay is not the relay that causes this issue and B Train ESF components

will actuate from B Train.c.Incorrect, RTBs will open second part is correct.D.Incorrect, both parts incorrect.

see above for explanation.

FSD A-181 007 Figure F-l is a block diagram, illustrating

the Reactor Protection

System FSD boundaries.

The equipment shown depicts the Reactor Protection

System and its interfaces

as follows: 1.Analog protection

system cabinets (W 7300 System Racks)containing

the bistables which input to the Reactor Protection.

Although the process instruments

which interface with the analog protection

system are considered

part of the reactor protection

system as defined by IEEE 279, they are also considered

as part of their respective

fluid systems.For completeness, the process sensors are included in this FSD.The functional

requirements

associated

with the.interfacing

process input components

to the RPS are those that are applicable

to these type devices on a generic basis.2.NIS Racks (the bistables which input to the RPS and the sensors contained within).3.Control board switches 4.Field contacts (RCP breakers, turbine stopvalves,etc.)

5.Solid State Protection

System (SSPS)initiates reactor trip or ESF actuation in accordance

with defined logic that is based on the bistable outputs from the process racks

T-8 depicts tile system interfaces

associated

with the SSPS Output Cabinet providing Reactor Trip and ESF actuation functions.

6.Reactor trip switchgear, normal and bypass breakers 7.Computer and control board demultiplexers (demux)Monday, January 14, 2008 2:42:17 PM 41

QUESTIONS REPORT for 75 RO Questions 8.AMSAC (Anticipated

Transient Without Trip (ATWT)Mitigation

System Actuation Circuitry)

is not shown nor considered

part of the RPS FSD but is being mentioned here for completeness.

The W 7300 and NIS Racks provide the signal conditioning, setpoint comparison, process analog signal actuation, control board/control

room!miscellaneous

indications

and compatible

electrical

signal output to the protection

devices.The bistable outputs pertaining

to these systems which provide this input to the RPS have been included as part of this FSD and are listed in Section 7.The process instrumentswhichinput

into the W 7300 and NIS Racks have been included as part of this FSD and are listed in Table T-7.(References

6.1.001,6.1.002,6.7.001,6.7.002,6.7.004, 6.7.010, 6.7.057)Monday, January 14, 2008 2:42: 17 PM 42

QUESTIONS REPORT for 75 RO Questions 012 Reactor Protection

System A2..02 Ability to (a)predict the impacts of the following malfunctions

or operations

on the RPS;and (b)based on those predictions, use procedures

to correct, control, or mitigate the consequences

of those malfunctions

or operations:

Loss of Instrument

Power Question Number: Tier 2 Group 1 Importance

Rating: Technical Reference:

25 3.6 FSD A-181 007 ,Figure 12 of ops-52201 RPS lesson plan Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS52201 D12 10 CFR Part 55 Content: 41.1 0 Comments: This question tests the loss of power to both sides of SSPS, the affects from the loss of power thru the vital panel and the affects from the loss of power from the solas, and the action required by the ERGs due to that loss.The second half of the KA is met by knowing from the question what the effects are to the plant to the failures listed, and then using EEP-O, know what the operator will have to do based on the failure to correct the malfunctions

that are identified.

A loss of power to solid state affects the ESF components

as well as RPS components

equally since power is lost to both protection

and control as well as RPS.Logic Cabinet (Figures 4 and 11)The logic cabinet is to the right of the input relay cabinet.It houses the circuitry to make logic decisions.

The logic circuitry receives signals from the input relay cabinet and if appropriate

signals are received, it will initiate a reactor trip or actuate the ESF systems.In addition to logic decision making, information

is collected, stored, and transmitted (via multiplexing

techniques)

to the computer and control board (via demultiplexing

techniques).

Both signals come from the logic cabinets.The logic circuits look for coincidence

between protection

channels.If the logic requirements

are met for a reactor trip, the circuit sends a signal to the UV driver card.The UV driver card output drops from 48V DC to zero and de-energizes

its associated

reactor trip and bypass breaker UV coils.This action trips open the breakers and de-energizes

the control rod drive mechanisms.

This releases the control rod assemblies

into the core.The train A UV driver card sends its trip signal to reactor trip breaker RTA, and to bypass breaker BYB.If an unsafe condition calls for safeguards

actuation, the logic circuits will send a signal to thesafeguardsdriver

card.The card's output will increase from zero to 48V DC and will energize the required master relays for the specific safeguards

actuation.

The master relays energize their slave relays using 120V AC,which supply either AC or DC control power to ESF loads as appropriate.

MCS Time: Points: 1.00 Version: a123456789 Answer: ACCB AC BDBD Scramble Range:A-D Monday, January 14, 20082:42:17

PM 43

Source: Cognitive Level: Job Position: reviewed: MODIFIED HIGHER RO GO QUESTIONS REPORT for 75 RO Questions Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: FARLEY FARLEY Monday, January 14, 20082:42:17

PM 44

QUESTIONS REPORT for 75 RO Questions 18.012 A4.01 001 Given the following:

Unit 1 is operating'

at 100%power when a PRZR PORV spuriously

opens.*The control room operators attempt to close the PORV but are unsuccessful.

The UO closes the block valve for the open PORV.The following conditions

exist:*Tavg is 575°F.*PRZR level is 63%.*PRZR pressure is 1845 psig and rising slowly.*Reactor power is 98%.Which ONE of the following actions are required?A.Restore RCS pressure to>2209 psig within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .B.Commence plant shutdown and be in hot standby per UOP-3.1, Power Operation.

Cr Manually trip the reactor, initiate SI, and enter EEP-O, Reactor Trip or Safety Injection.

.D.Maintain the PORV block valve closed with power available.

A.lncorrect

-TS 3.4.1 requires restoration

w/i 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in mode 2 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .This would be true if a reactor trip were not required at this time.B.'Incorrect

-This would be an action due to rising PRZR level lAW 3.4.9 when pzr level is greater than 63.5%.C.Correct-this action is necessary since PRZR pressure is below the reactor trip and SI setpoint.D.Incorrect-This is the TS action if the plant would remain at power.Monday, January 14, 20082:42:17

PM 45

QUESTIONS REPORT for 75 RO Questions 012 A4.01 Reactor Protection

System Ability to manually operate and/or monitor in the control room: Manual trip button Question Number: Tier 2 Group 1 11 Importance

Rating: 4.5 Technical Reference:

AOP-100;TS 3.4.11 Proposed references

to be provided to applicantsduringexamination:

None Learning Objective:

10 CFR Part 55 Content: 41.1 0 Scramble Range:A-D FARLEY Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK LOWER RO GO Version:0123456789

Answer: CBABBB AADB Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42: 18 PM 46

QUESTIONS REPORT for 75 RO Questions 19.013 A4.02 002 Given the following:*A LOCA has occurred.*RCS pressure is 500 psig and stable.*Containment

pressure is 29 psig and lowering slowly.*All equipment is operating as designed.*The crew is performing

actions contained in ESP-1.2, Post LOCA Cooldown and Depressurization, preparing to reset ESF Actuation signals.Which ONE of the following describes the conditions

required to be met, if any, to reset Containment

Isolation Phase A and B?*Phase A may be reset without additional

conditions.

Phase B may be reset without additional

conditions.

B.*Phase A may be reset without additional

conditions.

Containment

Spray must be reset prior to resetting Phase B.C.,.Phase A may be reset without additional

conditions.

Phase A must be reset before resetting Phase B.D.*Safety Injection must be reset before resetting Phase A.*Containment

pressure must be less than the actuation setpoint before resetting Phase B.A is correct.Manual resets for Phase A and Phase B may be performed even with actuating signal present.B is incorrect.

Credible because CTMT spray and Phase B have the same actuating signal.C8 actuation and Phase B can be reset independently

of each other and at any time.C is incorrect.

Credible because procedure directs Phase B reset after Phase A.Phase B can be reset at any time.D is incorrect.

Credible because 81 is the automatic initiation

signal for Phase A.Phase A can be reset prior to SI reset.Monday, January 14,20082:42:18

PM 47

QUESTIONS REPORT for 75 RO Questions 013 Engineered

Safety Features Actuation System A4..02 Ability to manually operate and/or monitor in the control room: Reset of ESFAS channels Question Number: 13 Tier 2 Group 1 Importance

Rating: 4.3 Technical Reference:

EEP-1.0, Reactor Protection

System FSD, A181 007, Figure F-2 sheet8&Table T-4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.7/45.5 to 45.8 Scramble Range:A-D Version:0123456789

Answer: ABADBABDDA

Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO 1.00 BANK HIGHER RO GO Points: Comments: meets the KA in that the question asks for the conditions

needed to reset ESFAS channels in the CR MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:18 PM 48

QUESTIONS REPORT for 75 RO Questions 20.013 G2.4.31 008 Plant conditions

at 09:00 were as follows:*Unit 1 was at 100%power.*SSPS train"B II surveillance

testing was in progress.*ED4, SSPS B TRN TRBL, was in alarm due to the SSPS testing.*The"B" Reactor Trip Bypass Breaker were closed and the associated

annunciator

alarms have been acknowledged.

At 09:30 the following occurs:*EC4, SSPS A TRN TRBL, has come into alarm.*All remaining annunciators

are unchanged.

Theplant operator reports that the SSPS train"A" Output Relay Mode Selector switch was inadvertently

placed in the TEST position.Which ONE of the followi*ng

actions are required?A.Apply Technical Specification

3.0.3 and initiate actions within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to shut down to Hot Standby.B.Stop the testing, check the"B" Reactor Trip Breaker closed, then open the"B" Reactor Trip Bypass Breaker.Initiate a manual reactor trip;if unsuccessful, enter FRP-*S.1, Response to Nuclear Power Generation/ATWT.

D.Immediately

place the SSPS train"A" Output Relay Mode Selector switch in the OPERATE position;then verify EC4, SSPS A TRN TRBL, annunciator

has cleared.Monday, January 14, 2008 2:42:18 PM 49

QUESTIONS REPORT for75 RO Questions For this condition while testing is on-going in the SSPS B Train cabinets, with the bypass breaker closed, the general warning light will be lit for B Train.This will cause ED4 to be in alarm.Then when the SSPS train"A" Output Relay Mode Selector switch is placed in the TEST position the other trains GW light will be LIT and 2 GW lights should cause a reactor trip.some P&Ls of STP-33.0 follow: 4.2 Ensure that the GENERAL WARNING lamps on SOLID-STATE

PROTECTION

TRAIN-A and B LOGIC CABINETS are OFF prior to commencing

this test.4.6 IF a failure occurs during testing, THEN hold at the point the failure occurs and contact Maintenance

for troubleshooting

and repair.4.1 The GENERAL WARNING lamps on SOLID-STATE

PROTECTION

TRAIN-B and B LOGIC CABINETS will be ON during this test.A.incorrect.

Both trains of Solid State are inoperable, and a common TS which applies when both trains of safety related equipment is inoperable

is 3.0.3, but a reactor trip is.required for this condition.

B.incorrect.

IIBacking ouf'of a test when the other train becomes inoperable

is normally done, but due to the GW on both trains a reactor trip is required.C.correct.Rea'ctor should have tripped with 2 trains in test due to both trains have a general warning in which is the input to the automatic trip coincidence

of 2/2 GW alarms in.ARP EC4 AUTOMATIC ACTION 1.IF both Train A AND B Solid State Protection

System Trouble alarms are actuated, THEN a reactor trip will occur.D.incorrect.

If SSPS train A had not been momentarily

inoperable, no further a"ction would be required.The coincidence

for a reactor trip would no longer be met.Quickly restoring the output switch to it's original position would not change the fact that the SSPS should have initiated an automatic trip, and met coincidence

for one, but it did not occur.A reactor trip is necessary.

Monday, January 14, 20082:42:18

PM 50

QUESTIONS REPORT for 75 RO Questions 013 G2.4..31 Engineered

Safety Features Actuation System Emergency Procedures

I Plan: Knowledge of annunciators

alarms and indications, and use of the response instructions

..Question Nu.mber: Tier 2 Group 1 12 Importance

Rating: 3.3 Technical Reference:

ARP EC4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK HIGHER RO GTO Version: a12345678"9 Answer:CC BCDBBACB.Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: Monday, January 14, 20082:42:18

PM 51

QUESTIONS REPORT for 75 RO Questions 21.015 AK3.01 002 Given the following:

Unit 2 is in Mode 4 with two RCPs running.The crew is at a step in UOP-1.1, Startup of Unit from Cold Shutdown to Hot Standby, to start a third RCP.Which ONE of the following correctly describes a RCP failure mechanism that will still allow the remaining RCP to be started, the damage that would occur and the reason?Art*The anti-reverse

rotation device pawls are not engaged in the ratchet plate.*RCP motor winding damage due to high starting currents.B.*The anti-reverse

rotation device pawls are not engaged in the ratchet plate.*RCP radial bearing damage due to reverse flow through the RCP._C.*The oil lift pump does not develop 600 psig oil lift pressure.*RCP radial bearing damage due to high starting torque.D.*The oil lift pump does not develop 600 psig oil lift pressure.*RCP motor winding damage due high starting torque.A.Correct-the anti rotation device pawls not being engaged in the ratchet plate would cause the high motor winding temps and possible damage to the windings due to high starting currents.A flywheel and an anti-reverse

rotation device have been mounted at the top of the motor.Stopping one or more Reps while other pumps are running will cause a reverse flow through the inactive loops.Reverse flow will turn the de-energized

pump backwards.

Although no mechanical

damage would result from reverse rotation, an attempt to start a pump in this condition would cause starting currents to exist for an

length of time, resulting in overheating

of the motor.To prevent reverse rotation, each pump has been equipped with an anti-reverse

rotation device.B.Incorrect-first part is correct.second part is not due to the fact that no damage will result to a RCP due to reverse flow thru the pump.C.Incorrect-The oil lift pump does not develop 600 psig or greater oil lift pressure wouldnotallow the RCP to be run.If it were to be started with the pressure not being>600 psig, then the second part is in fact correct for the oil lift pump but not for the radial bearing.It would actually potentially

damage the windings._______

attempt to start a Rep unl essitS-oi 1 lift

the upper thrust shoes for at least two minutes.Observe the oil lift pumps indicating

lights to verify correct oil pump motor operation and oil pressure.The oil lift pumps should run at least 1 minute after the Rep's are started.An interlock will prevent starting a Rep until 600 psig oil pressure is established.

Monday, January 14, 20082:42:18

PM 52

QUESTIONS REPORT for 75 RO Questions system before starting the motor.The oil"lifts" the thrust shoes away from the thrust runner.D.Incorrect-see above for the first part.RCP would not start in thi-s condition due to an interlock.

second part deals with the lower bearing.information

below.RCP motor winding damage could result due to the upper thrust shoes.Rep lesson plan The lower thrust bearing takes the weight of the rotating parts when the reactor coolant loop is at low pressure.As the loop pressure increases, the unbalanced

force on the number one seal causes the shaft to lift and transfer the thrust to the upper thrust bearing.By the time loop pressure is sufficient

to allow pump operation (350 psig), all thrust acts on the upper bearing, which is the normal operating condition.

The lower thrust bearing functions only when the motor runs uncoupled from the pump.In this condition, the weight of the motor rotating parts acts downward.In order to reduce starting torque, the thrust bearing shoes receive oil from the oil lift system before starting the motor.The oil"lifts" the thrust shoes away from the thrust runner.The lower thrust bearing takes the weight of the rotating parts when the reactor coolant loop is at low pressure.As the loop pressure increases, the unbalanced

force on the number one seal causes the shaft to lift and transfer the thrust to the upper thrust bearing.By the time loop pressure is sufficient

to allow pump operation (350 psig), all thrust acts on the upper bearing, which is the normal operating condition.

The lower thrust bearing functions only when the motor runs uncoupled from the pump.In this condition, the weight of the motor rotating parts acts downward.015 Reactor Coolant Pump Malfunction

-AK3.01 Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Potential damage from high winding"and/or bearing temperatures

Question Number: Tier 1 Group 1 Importance

Rating: 41 R02.5 Technical Reference:

RCP LP OPS-521 01 D, UOP-1.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 Comments: had to replace this question since it did not meet the KA, ie the reason portion of Hie KA, and it is an exact question from 003 A1.04 meets the KA in that there is a potential failure identified

that would cause loss of RC flow form the RCP and the reason that malfunction

would cause that problem, specifically

motor windings.I did not test the bearing side of the KA due to the KA already on the test 003A 1.04.Monday, January 14, 20082:42:18

PM 53

FARLEY NO Scramble Range:A-D FARLEY QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer: ABCBCCAABC

Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: MODIFIED HIGHER

RO GO Points: MCS Time:.Source: Cognitive Level: Job Position: reviewed: Monday, January 14,20082:42:18

PM 54

QUESTIONS REPORT for 75 RO Questions 22.015 K6.02 002 Given the following:

Unit 2 was initially at 95%power.*The reactor has tripped.*Compensating

voltage on N-35, Intermediate

Range NI, is set too HIGH.Which ONE of the following correctly describes the response of Intermediate

Range N-35 following the trip AND the effect of this response on the Source Range (SR)HI FLUX TRIP?N-35 will indicate (1)than actual power.(2)The SR HI FLUX TRIP will reinstate

A.(1)LOWER (2)as soon as N-35 reaches the P-6 s"etpoint.(1)LOWER (2)only when N-36 reaches the P-6 setpoint.C.(1)HIGHER (2)only when N-35 reaches the P-6 setpoint.D.(1)HIGHER (2)as soon as N-36 reaches the P-6 setpoint.A Incorrect;

If one channel indicates low, it will satisfy 1/2 of the required P-6 reset logic.Therefore, if the logic was 1 of 2, A would be correct.B Correct;An overcompensated

channel means that compensating

voltage is too high for the channel, cancelling

out part of the actual signal, resulting in a lower indication.

The P-6 permissive

is satisfied when 2 out of2IR channels are below the setpoint.C Incorrect;

the channel would indicate lower per above discussion

and Would be correct if the channel was undercompensated

with the actual logic (2 out of 2 logic for going below P-6)D.Incorrect;

If the logic was 1 out of 2 versus 2 out of 2, and the IR indicated high (undercompensated)

Monday, January 14, 20082:42:18

PM 55

QUESTIONS REPORT for 75 RO Questions 015 Nuclear Instrumentation

System K6.02 Knowledge of the effect of a loss or malfunction

on the following will have on the NIS: Discriminator/compensation

circuits Question Number: Tier 2 Group 2 32 Importance

Rating: 2.6 Technical Reference:

UOP-2.3, ESP-O.1, step 11 and NI LP OPS-52201 D Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 Comments: good match Scramble Range:A-D VCS MCS Time: 4 Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK HIGHER RO GO Version: a123456789 Answer:BC CADAB B BA Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:18

PM 56

QUESTIONS REPORT for 75 RO Questions 23.017 K1.02 001 The Subcooled Margin Monitor is in the RTD mode.Which ONE of the following correctly describes the temperature

instruments

used in the Subcooling

calculations

for this mode of operation?

.The highest reading RTD of the 3 Wide Range RCS Hot leg and 3 Wide Range RCS Cold Legs is used.B.The highest reading of Core Exit and Upper Head Thermocouples

is used.C.The fifth hottest of all Core Exit.Thermocouples (excluding

the Upper Head thermocouples)

is used.D.The fifth hottest RTD of all Wide Range RCS Hot legs and Cold Legs is used.A.Correct.This is the method used in the RTD m.ode, but has a greater time delay in indication

of a loss of subcooling

due to loop transit time and instrument

RTD response time than the CETC mode.B.Incorrect.

This is correct for the Individual

Value display mode.C.Incorrect.

This incorrect for the Individual

Value display mode.The upper headTCs are not excluded from the calculation

in the individual

value mode, and also the fifth hottest is not used in the calculation

even though the fifth hottest is used for diagnostic

purposes throughout

the ERG procedure network.D.Incorrect.

This is incorrect.

The fifth hottest of all RTDs is not used 017 K1.02 In-Core Temperature

Monitor System Knowledge of the physical connections

and/or cause effect re'lationships

between the ITM system and the following systems: ReS Question Number: 33 Tier 2 Group 2 Importance

Rating: 3.3 Technical Reference:

ICCMS LP OPS-52202E, SOP-68.0 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5 Comments: Monday, January 14, 20082:42:18

PM 57

FARLEY NO Scramble Range:A-D QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer: ACBB BAB AB D Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: MODIFIED LOWER RO GO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:18

PM 58

QUESTIONS REPORT for 75 RO Questions 24.022 AA2.04 002 Given following:

Unit 1 is at 60%power.*Pressurizer

levelon program.*All charging flow has been lost and NO charging pump is running.*Letdown has been secured.*PRZR level is lowering at a rate of 1%every five (5)minutes.*1 00%Tref=573°F Approximately

how much time will pass before all pressurizer

heaters will automatically

secure assuming no operator action?A.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B.1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months s2 hours D.2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Monday, January 14, 20082:42:19

PM 59

QUESTIONS REPORT for 75 RO Questions A.Incorrect;

1/2 the value of C.B.Incorrect;

60%X 28.8 will result in approx.17.8%.Level lowering at1%every 5 minutes yields approx.89 minutes, about 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .C.Correct;At 60%power, program level is approximately

38.7%.This is calculated

by taking 573-547=26 which=2.6 X 10%change in power.60%power or 6 x 2..6=15.6+547=562.6°F.Level at this temperature

is 38.7%based on 50.2-21.4=28.8 or 2.88 per10 6 X 2.88=17.28+21.4=38.68 (level at 50%+35.8 or 1/2 of 28.8+21-.4)Letdown isolates at 15%.38.7-15=23.68%level decrease and since 1%per 5 minutes is the level decrease, the time would be 5 X 23.68=118.4 or 2 minutes less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months D.Incorrect;

Using the 18.7%and subtracting

from 50.2%level yields approx.34.6%x 5=157.5 minutes or 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months calculation:

at 100%power level is 50.2%and at 0%power level is 21.4%.This gives a level'change of approx..288%.per%power.Letdown isolates at 15%.-A508617 pzr level setpoint document shows pzr level to be 21.4%at 547°F and 54.9%at 577.2°F.The lesson plan shows level to be 21.4%at*547°F and 50.2 at 573°F.figure 6 shows program level to be 21.4 to 50.2 from 547 to 573°F Tavg Options are plausible

they are symmetrical

and not significantly

different from actual time.A math error or misunderstanding

of program level at this power could direct an applicant to any option.Monday, January 14, 20082:42:19

PM 60

QUESTIONS REPORT for 75 RO Questions 022 Loss of Reactor Coolant Makeup-AA2.04 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Pump Makeup: How long PZR level can be maintained

within limits Question Number: 42 Tier 1 Group 1 Importance

Rating: R02.9 Technical Reference:

OPS 52201 H FIG 6 and AOP-16 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 FARLEY Scramble Range:A-D WTSI BANK HIGHER

RO GO Comments: Tref is given due to the fact that both units are different and we teach Tref at 573°F and explain that this value changes most every outage and is a moving target.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: C BABCC CACD Source if Banle Difficulty:

Plant:

Previous 2 NRC exams: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:19

PM 61

QUESTIONS REPORT for 75 RO Questions 25.022 Kl.Ol 001 Given the following initial conditions

with Unit 1 at 1 00%power: There is a smclilleak

in the 1A containment

cooler.The cooler has been isolated lAW*SOP-12.1, Containment

Cooling System.The following valves are closed for leak isolation:

-MC)V-3019A, SW TO1A CTMT CLR AND CTMT FPS-MC)V-3024A, EMERG SW FROM 1A CTMT CLR-MC)V-3441A, SW FROM 1A CTMT CLR A Large Break LOCA occurs at this time.Which ONE of the following correctly describes the:Service Water flow rate (if any)through the1A containment

cooler with no operator action?SW flow will bE3-----A.secured B.approximately

600 gpm C.approximately

800 gpmapproximately

2000 gpm Monday, January 14, 20082:42:19

PM 62

QUESTIONS REPORT for 75 RO Questions Cooling water normally discharges

through a 10-inch line including MOV-3441 A, B, C, and D, which are located inside containment, then through a 6-inch line and MOV-3023A, B, C, and D.On an IISII signal, water also discharges

through a 10-inch discharge line through MOV-3024A, B, C, and D, thus increasing

the flow through the coolers.MOV-3024A, B, C, and D Containment

Cooler Emergency Service Water Discharge Valves (Figure 19)Each motor-operated

valve is controlled

by a three-position

MCB handswitch (CLOSE/AUTO/OPEN, spring return to AUTO).In the AUTO position, the valve automatically'

opens upon receiving an S-signal.Valve position indication

lights are above each switch.MOV-3441A, B, C, and D Containment

Cooler Service Water Discharge: Isolation Valves The operation of these MCB motor-operated

valves

19)is identical to the emergency service water discharge valves (3024A, B, C, and D).MOV-3019A,B,C, and D Containment

Cooler Service Water Inlet Isolation Valves The operation of these MCB motor-operated

valves (Figure 19)is identical to the emergency service water discharge valves (3024A, B, C, and D).A.Incorrect-due to the SI signal MOV-3019A, MOV-3441A and MOV-3024A open to provide emergency SW to the coolers.MOV-3023A is normally open and is not closed for the leak isolation lAW SOP-12.1.This would be correct if the candidate did not know what valves rolled open then the effect on containment

temperature

due to the failure.B.Incorrect-This could be confused with the TS bases requirement

to have 600 GPM flow from one Ctmt cooler to meet post LOCA conditio,ns.

C.Incorrect-This could be confused with not knowing the correct valve line u'p and this is the normal flow rate thru the coolers which if the emergency SW to the valve was not taken into account forthisevent, this would be the correct answer.D.Correct-Since due to the SI signal MOV-3019A, MOV-3441A and MOV-3024A open to provide emergency SW to the coolers flow rate would increase to 2000 gpm thru all coolers with two SW pumps running whether the fan is running or not.Cooling at 2000 gpm will provide the cooling necessary in a LBLOCA to cool containment.

Monday, January 14, 20082:42:19

PM 63

QUESTIONS REPORT for 75 RO Questions 022 Containment

Coo,ling System (CCS)K1.01 Knowledge of the physical connections

and/or cause-effect

relationships

between the CCS and the following systems: SWSI cooling system Question Number: 14 Tier 2 Group 1 Importance

Rating: 3.5 Technical Reference:

FSD A 181001 Service Water System OPS-52102F, SW lesson plan and TS bases 3.6.6 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55.Content: 41.5 Scramble Range:A-D FARLEY Version:0123456789

Answer: DC CDC CDDCA Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO MODIFIED HIGHER RO GO Source: Cognitive Level: Job Position: reviewed: Comments: This KA was changed to 022K1.01 due to not being able to meet the selected KA.This was approved by FJE and recorded on ES-401-4 record of rejected KAs.This meets the KA in that it asks for the cause effect of only having one sw flow path to one operating ctmt cooler.MCS Time: 1, Points: 1.00 Monday, January 14, 20082:42:19

PM 64

QUESTIONS REPORT for 75 RO Questions 26.025 AK2.03 002 The following conditions

exist:*The plant is in Mode 1.*1A CCW pump is tagged out to have the motor rebuilt.*1 Band1C CCW pumps are both in operation.*8 Train is the liOn Service" train.The1C CCW pump has just tripped.Which one of the following correctly describes ONLY components

that have lost ALL CCW flow due to the1C CCW pump trip?A.18 RHR Hx, 18 RHR Pump Seal Cooler, 18 Spent Fuel Pool Hx 8.18 RHR Hx, 18 RHR Pump Seal Cooler, 1A Spent Fuel Pool Hx C.1A RHR Hx, 1A RHR Pump Seal Cooler, 1A Spent Fuel Pool Hx D¥1A RHR Hx, 1A RHR Pump Seal Cooler, 18 Spent Fuel Pool Hx Monday, January 14, 20082:42:19

PM 65

QUESTIONS REPORT for 75 RO Questions C CCW pump is the A train pump-The B CCW pump is the onservice train and iscarryingthe

misc.header and on B Train.A.Incorrect-RHR components

are not correct, SFP is correct.B.Incorrect-All are B Train ESF loads supplied by B CCW pump.C.Incorrect-RHR components

are correct, SFP is not correct.D.correct-All are A Train ESF loads supplied by C CCW pump OPS-52102G

The ESS loads consist of the following:

1.Charging pum.ps 2.Spent fuel pool heat exchangers

3.RHR heat exchangers

4.RHR pumps The secondary heat exchanger loads consist of the following:

1.RCP oil coolers and thermal barrier heat exchangers

2.Reactor coolant drain tank (RCDT)heat exchanger 3.Excess letdown heat exchanger 4.Seal water heat exchanger 5.Letdown heat exchanger 9.Waste gas compressors

10.Sample system heat exchangers

11.Gross failed fuel detector The CCW system provides cooling for the Train A and Train 8 emergency core cooling system components.

The C CCW pump and the C heat exchanger are designated

as Train A.The A CCW pump and the A CCW heat exchanger are designated

as Train B.The 8 CCW pump and the 8 CCW heat exchanger can be aligned to either train.CCW is normally lined up so that one CCW pump and one CCW heat exchanger is in operation supplying the on-service

train.The on-service

train is'the one that supplies the secondary heat exchangers.

The swing pump and heat exchanger (18 CCW pump and 18 HX)is normally aligned in standby to the on-service

train with the heat exchanger outlet valve shut.The remaining pump and heat exchanger is valved into a closed loop with the redundant safety train.This train is idle and is designated

as the off-service

train.The off-service

train CCW pump must be running before starting the off-service

train charging pump or RHR pump.Monday, January 14, 20082:42:19

PM 66

QUESTIONS REPORT for 75 RO Questions.

25 Loss of Residual Heat Removal System AK2..03 Knowledge of the interrelations

between the Loss of Residual Heat Removal System and the following:

Service water or Closed cooling water pumps Question Number: 43 Tier 1 Group 1 Importance

Rating: R02.7 Technical Reference:

ARP-1.1 AD5 and AE4, UOP-1.1, SOP-7.0 and SOP-23 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55Content:41.7

/45.7 Scramble Range:A-D FARLEY MODIFIED LOWER RO GO Comments: meets the KA in that the interrealtionships

between RHR and CCW and a CCW pump trip has to be evaluated to arrive at the correct answer.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: DABDDDB CDC Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14,20082:42:19

PM 67

QUESTIONS REPORT for 75 RO Questions 27.026 A3.01 004 Given the following:

Unit 1 was in Mode 5;Unit 2 was at 100%power.*A

Loss of Offsite Power has occurred.*All DGs have started and tied onto the 4160V ESF buses.*Vital load sequencing

has been completed.

ESP-0.1, Reactor Trip Response, has been entered on Unit 2.*An inadvertent

'A'Train CS actuation signal is received while the crew is responding

to the reactor trip.Which ONE of the following correctly describes the status of the Unit 2 Train IA 1 Containment

Spray (CS)system?2A CS Pump A.StoppedStopped C.Running D.Running Monday, January 14, 20082:42:19

PM 2A CS Pump Discharge Valve Closed Open Closed Open 68

QUESTIONS REPORT for 75 RO Questions A.Incorrect-Neither CS pump will start, due to the LOSP with no SI actuation (pump controls logic diagram), but the Train A inadvertent

actuation for containment

spray would cause the A train pump discharge MOV to open.B.Correct-See FSD narrativebelow.ESS loading sequencer requires an SI signal to be present before sequencing

on ESS loads.An SI signal has not occurred for this transient so the pump will not start but the valve will open.C.Incorrect-The inadvertent

Train A containment

spray actuation would not cause the 2A spray pump to start for reason given in A above.The A train MOV would open for reasons given in A above.The B Tr?in would be unaffected.

D.Incorrect-See 2A pump will not start per the FSD narrative below.OPS-52102C

Containment

Spray Pumps (Figure 6)A three-position (STOP/AUTO/START, spring return to AUTO)handswitch

controls each pump.Placing the switch in theST ART position will start the pump.Placing the'switch in the STOP position will stop the pump and reset the 86 relay.In the AUTO position, the pump will automatically

start upon receipt of a containment

spray actuation signal ("P" signal)if an LOSP has not occurred.If an LOSP has occurred with the IIp ll signal present, a safety injection signal to the ESF sequencer must also be present, or the ESF sequencer must be in test, to start the pump.T Containment

Spray System FSD A 181008 3.1.5.2 With offsite power available, the"P" signal shall start both CSS pumps.Without offsite power available*, the CSS pumps shall start by the diesel generator ESS loading sequencer.

Starting will occur at step two of the sequence if the"P" signal is present at that time.If the"P" signal occurs between the completion

of step two and step six of the ESS sequence, then starting will occur at the completion

of step six of the loading sequence.If the"P" signal occurs after the completion

of step six, starting will take pi_ace immediately.

Automatic starting of the CSS pumps shall not occur unless the pump control switch on the main control board is in the"AUTO" position (References

6.4.001 , 6.4.006, 6.4.007, 6.4.008).3.6.1.1 These active valves shall open automatically

upon receipt of a containment

spray actuation signal ("P" signal)from the ESFAS and remain open for the containment

spraying function (Reference

6.2.001).Monday, January 14, 20082:42:19

PM 69

'QUESTIONS

REPORT for 75 RO Questions 026 Containment

Spray System'A3.01 Ability to monitor automatic operation of the CSS, including:

Pump starts and correct MOV positioning

Question Number: Tier 2 Group 1 15 Importance

Rating: 4.3 Technical Reference:

CS&Cool OPS-52102C

drawings-0207195020764502076530207646

Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK HIGHER RO GO Version:0123456789

Answer: BADCAC DAD C Scramble Range:A-D Source if Banle FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:19

PM 70

QUESTIONS REPORT for 75 RO Questions 28.026 AAl.02 005 The following conditions

exist:*Unit 2 is in Mode 3 preparing for a reactor startup.*1B CCW pump is tagged out.*The on service CCW pump trips due to over-current.

The other CCW pump wiU not start from the MCB.*The crew just tripped the RCPs lAW AOP-9.0, Loss of Component Cooling Water.Which ONE of the following correctly describes operation of the charging pumps while performing

AOP-9.0, Attachment1, Establishing

Firewater Cooling to a Charging Pump?A.*Stop all charging pumps until CCW or fire water is established

to at least one charging pump.*Maximum allowable Charging Pump lube oil temperature

is 160°F.B.*Stop all charging pumps until CCW or fire water is established

to at least one charging pump.*Maximum allowable Charging Pump lube oil temperature

is 140°F.*Swap operating charging pumps until fire water is established

to one charging pump.*Maximum allowable Charging Pump lube oil temperature

is 160°F.D.*Swap operating charging pumps until fire water is established

to one charging pump.*Maximum allowable Charging Pump lube oil temperature

is 140°F.Monday, January 14, 20082:42:19

PM 71

QUESTIONS REPORT for 75 RO Questions AOP-9.0, Version 18 A.Incorrect-

It is not correct to stop all charging pumps sincethe RCP seals could be damaged.The temperature

is correct.B.Incorrect-It is not correct to stop all charging pumps since the RCP seals could be damaged.The temperature

is NOT correct.C.Correct-This is the correct way to operate the chg pumps and the correct temperature.

D.Incorrect-This is the correct way to operate the chg pumps and NOT the correct temperature.

AOP-9 attachment

1 Note: Until alternate cooling is established, swapping the operating CHG PUMP may lengthen the time that RCP seal injection is maintained.

EA3 Dispatch operator to determine the affected pump and the actual temperature

as indicated on the local temperature

indicators.

IF local temperature

indication

is: 1.1 Between 140°F and 155°F, THEN operation may continue during subsequent

troubleshooting.

1.2 Between 155°F and 160°F, THEN consider shutdown of pump.1.3>160°F, THEN immediately

shutdown the affected charging pump.2.IF a loss of CCW has occurred, THEN perform the actions required by FNP-1-AOP-9.0

LOSS OF COMPONENT COOLING WATER.026 Loss Component Cooling Water AA1.02 Ability to operate and/or monitor the following as they apply to the Loss of Component Cooling Water: Loads on the CCWS in the control room Question Number: 44 Tier 1 Group 1 Importance

Rating: R03.2 Technical Reference:

CCW LP OPS-52102G

and AOP-9 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

l11-f--10I-le----t-F

.......

Comments: Monday, January 14, 2008 2:42:20 PM 72

FARLEY NO Scramble Range:A-D FARLEY QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789*Answer: CDACABBBBB Source if Bank: Difficulty:

Plant:

Previous 2 NRC exams: BANK illGHER RO GO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14,2008 2:42:20 PM 73

QUESTIONS REPORT for 75 RO Questions 29.026 G2.1.2 002 Given the following conditions

on Unit 2:-The plant was at 100%power when the 2A SG Main Steam line ruptured inside containment.-All systems actuated as per design.-Containment

pressure spiked to 33 psig and is now continuing

to decrease slowly.-The crew has entered ESP-1.1, SI Termination.

Which one of the following correctly describes the MAXIMUM containment

pressure and MINIMUM recirculation

time that will allow the OATC to secure the CS pumps per ESP-1.1?Containment

Pressure A.15 psig15 psig C.18 psig D.18 psig Monday, January 14, 20082:42:20

PM Time Aligned for Recirculation

7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 10 hours 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 10 hours 74

QUESTIONS REPORT for 75 RO QuestionsA-Incorrect,'The 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of operation'applies to HHSI/LHSI transferring

from Cold Leg recirc to Simultaneous

hoVcold leg recirc, but is not long enough to meet the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months minimum for operation of CS on recirc prior to securing CS.EEP-1[CAl WHEN 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months have passed since the start of the event, THEN go to FNP-1-ESP-1.4, TRANSFER TO SIMULTANEOUS

COLD AND HOT LEG RECIRCULATION.B-Correct, CS has been aligned for recirculation

flow for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months and containment

pressure is15 psig.ESP-1.3, does provide guidance;Containment

pressure is<16#and the time on recirc is>8 hours.C-Incorrect, containment

pressure has to be<16 psig and it is not.The RWST, at 4.5 feet and decreasing

would be a time to align the CS pumps for recirc, and if this could not be done then they would be secured.D-Incorrect, 18 psig is too high a pressure and CS has not been aligned for recirc for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .ESP-1.3.step 10.3 WHEN containment

spray recirculation

flow has been established

for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , AND containment

pressure islessthan 16 psig, THEN stop both CS PUMPs.ESP-1.1 step 18.3 18.3[CAl WHEN containment

spray recirculation

flow has been aligned for at least8hours, AND containment

pressure is less than 16 psig, THEN stop both CS pUMPs.026 G2.1.2 Containment

Spray System Conduct of Operations:

Knowledge of operator responsibilities

during all modes of plant operation.

Question Number: Tier 2 Group 1 16 Importance

Rating: 3.0 Technical Reference:

ECP-1.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: replaced to match the KA.original question did not have an operator responsibility.

This question does have an operator responsibility

blc this is a Continuing

action step and would___

period of time later in which they would be responsible

for identifying

and securing this system properly.MCS Time: 1 Points: 1.00 Version:0123456789

Answer:BBACBBCBBA

Scramble Range:A-D Monday, January 14, 20082:42:20

PM 75

Source: Cognitive Level: Job Position: reviewed: MODIFIED LOWER RO GO QUESTIONS REPORT for 75 RO Questions Source if Bank: FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:20 PM 76

QUESTIONS REPORT for 75 RO Questions 30.027 AA2.16 001 Given the following:

Unit 2 is at 100%power.*PT-444, PRZR PRESS, fails LOW.Which ONE of the following lists an action contained in AOP-1 00, Instrumentation

Malfunction, that will terminate the pressure transient and stabilize ReS pressure?Ar:'Place all pressurizer

heaters to the OFF position and cycle the heaters as required.B.Fully open both PRZR Spray valves and cycle the heaters as required.C.Take manual control of PK-444A, PRZR PRESS REFERENCE, and reduce demand to 0%.D.Take manual control of PK-444A, PRZR PRESS REFERENCE, and increase demand to 100%.Monday, January 14, 2008 2:42:20 PM 77

QUESTIONS REPORT for 75 RO Questions A correct.PT-444 inputs to the master controller.

If it fails low, heaters will energize in an attempt to raise pressure.Eventually

pressure will rise to a point where 1 PORV (PORV 445A)will open unless heaters are secured manually.B incorrect.

opening sprays partially would stop the pressure rise and stabilize pressure, but opening them FULLY will cause pressure to drop and continue dropping.C incorrect.

Lowering demand to 0%will secure heaters, but additionally

open spray valves and one PORV.This would cause pressure to continue to drop.D incorrect.

PT-444 failing low does not cause the controller

to fail low.The controller

output demand is actually high at 100%due to the failure.PT 444 Fails Low Backup heaters tum on Variable heaters tum on to maximum Spray valves close (if open)Actual pressurizer

pressure increases to PORV PCV-445A open setpoint causing PORV to open Actual pressurizer

pressure eventually

decreases to the PORV PCV-445A close setp'oint, causing PORV to close Plant pressure cycles around PORV open/close

setpoints AOP-IOO step 4 actions: IF an alarm was caused by a CONTROL instrument (PT*444/445)

OR component failure, THEN perform the following as required to restore RCS pressure to desired value.Take manual control of the following as required:*Pressurizer

Heaters[]IA PRZR HTR GROUP BACKUP[]IB PRZR HTR GROUP BACKUP[]IC PRZR HTR GROUP VARIABLE[]ID PRZR HTR GROUP BACKUP[]IE PRZR HTR GROUP BACKUP Monday, January 14, 20082:42:20

PM 78

QUESTIONS REPORT for 75 RO Questions 027 Pressurizer

Pressure Control Malfunction

-AA2.16 Ability to determine and interpret the following as they apply to the Pressurizer

Pressure Control Malfunctions:

Actions to be taken if PZR

instrument

fails low Question Number: 45 Tier 1 Group 1 Importance

Rating: RO 3.6 Technical Reference:

AOP-100 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 Scramble Range:A-D NEw illGHER RO GO Comments: This is the action to be taken on a PT-444 failure low lAW AOP-100 and matches the KA.Revised all choices per FJEs comments.MCS Time: 1 Points: 1.00 Version: a123456789 Answer: ADB B DAB DAB Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14,20082:42:20

PM 79

QUESTIONS REPORT for 75 RO Questions 31.028 K2.01 002 Unit 1 has just lost power to 600V Motor Control Center 1 B.Which ONE of the following components

will not have power?1 B Post LOCA Hydrogen Recombiner.

B.1 B Containment

Cooler Fan-High speed.C.HHSI TO RCS CL Isolation Valve, MOV-8803B.

D.1 B Accumulator

Discharge Isolation Valve, MOV-8808B.

A.Correct-1 B Post LOCA Hydrogen Recombiner.

OPS-52102D

B.The recombiners

receive power from separate vital electrical

power trains.Recombiners

A and B are powered from 600V Mee A and B, respectively.

B.Incorrect-LCC B is the power supply.C.Incorrect-M/CC V;valve is outside CTMT.D.Incorrect-MCC V;valve is inside CTMT.028K2.01 Hydrogen Recombiner

and Purge Control System Knowledge of bus power supplies to the following:

Hydrogen Recombiners

Question Number: 34 Tier 2 Group 2 Importance

Rating: 2.5 Technical Reference:

Post LOCA Atm Control, OPS-LP 521 02D Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5 Comments: different way to ask the power supply to a component to make it different from 004 K2.01.MCS Time: Points: 1.00 Version:0123456789

Answer: AB CAAB CAB C Scramble Range:A-D Monday, January 14, 2008 2:42:20 PM 80

Source: Cognitive Level: Job Position: reviewed: BANK LOWER RO GTO QUESTIONS REPORT for 75 RO Questions Source if Bank: FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:20 PM 81

QUESTIONS REPORT for 75 RO Questions 32.029 EKl.Ol 001 Given the following:

An ATWT has occurred on Unit 2 during coastdown prior to entering a refueling outage.*The crew is performing

actions of FRP-S.1, Respon$e to Nuclear Power Generation/ATWT.

An operator has been dispatched

to trip the reactor locally.*Attempts to establish Emergency Boration have been unsuccessful.

Reactor power indicates 6%.*Intermediate

Range Startup rate is slightly positive.*The RCS temperature

is slowly rising.Which ONE of the following describes the actions required lAW FRP-S.1?A':'Allow the RCS to heat up, and continue attempts to place the reactor in a subcritical

condition.

B.Allow the RCS to heat up, and open one PORV as necessary to maintain pressurizer

pressure less than 2135 psig to increase charging flow.C.Stop the RCS heatup by increasing

AFW flow to greater than 700 gpm" and verify dilution paths isolated.D.Stop the RCS heatup by dumping steam to the main condenser, and continue attempts to place the reactor in a subcritical

condition.

FRP-S.1 version 25 17 Continue emergency boration.17 Perform the following.

17.1 Determine if moderator temperature

coefficient

positive or negative.[]Core Physics Curve 5 17.2 IF moderator temperature

coefficient

negative, THEN allow RCS to heat up.A.correct.During coastdown at EOl, MTC is negative under all conditions.

Do not leave FRP-S.1 until power below 5%.IF power was to be>%5 or a positive SUR on the IR, then in addition to continuing

the emergency boration, if the MTC is negative, then the RCS would be allowed to'HU to add positive reactivity

to the core and help shut it down.B.incorrect The RCS is allowed to heatup, but the PORvs are not cycled to maintain pressure less than 2135 psig unless pressure is>2335 psig.

'-t-t-i nt-r.Cl:-+Q

rl-HreG-t--Qe-GaYSe-£-.--:1--Gtoe-s-AQt-have--the

A FlAt f-lew to be

but does have an RNO step to increase AFW flow to 700 gpm if SGWls are not>31%D.incorrect RCS temperature

is not stabilized, it is

to rise: , Monday, January 14, 2008 2:42:20 PM 82

QUESTIONS REPORT for 75 RO Questions 029 Anticipated

Transient Without Scram (ATWS)EK1.01 Knowledgeofthe operational

implications

of the following concepts as they apply to the ATWS: Reactor nucleonics

and thermo-hydraulics

behavior Question Number: 46 Tier 1 Group 1 Importance

Rating: R02.8 Technical*Reference:

FRP-S.1 Proposed references

to be provided to applicants

during E!xamination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Scramble Range:A-D NEW IDGHER RO GO Source: Cognitive Level: Job Position: reviewed: Comments: This meets the KAin that this tests the operational

implic8ltions

during an ATWT and the effects that we would take if the reactor was still critical after emergency boration and rods going in what would happen temperature

wise, ie.thHrmo-hydralic

behavior.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: ADB DCDC B BD Source if Bank: Difficulty:

Plant: FARLEY Previous 2 exams: NO Monday, January 14, 20082:42:20

PM 83

QUESTIONS REPORT for 75 RO Questions 33.033 AK3.01 057 Given the following:

Unit 1 is in Mode 2, a reactor startup is in progress.*Intermediate

Range Instrument

N-35 indicates 1 X 10-8 Amps.*FB3, NI 36 LOSS OF COMPENSATING

VOLTAGE, has come in to alarm.*N-36 indicates 1 X 10-11 Amps and has failed LOW.*Repairs to N-36 will take 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .Which one of the following will satisfy the requirements

of FB3 and Technical Specifications, and the reason for those actions?At:'Shutdown the reactor since two IR Nls are required to remain at the current power level for the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .B.Remain at 1 X 10-8 Amps since two IR Nls are required to raise power to the POAH.C.Remain at 1 X 10-8 Amps since positive reactivity

additions must be suspended at this power level with one IR NI failed.D.Increase power to>5%since the plant can remain at>5%power indefinitely

with one IR NI failed.Monday, January 14, 2008 2:42:20 PM 84

QUESTIONS REPORT for 75 RO Questions A.Correct-Due to TS 3.3.1 below with IR power above P-6 and below P-1 0, tWQ IR Nls are required or power has to be decreased below P-6 in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or>1 0%power in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .Since no answer allows to go to>10°/0 power, this is the only option with one IR NI broke.B.incorrect-

power can not remain at 10-8 Amps and 2 IR range Nls are not required to go to the POAH.This iswherecritical

data is taken and power is leveled off during a startup.C.Incorrect, with power in the IR and loss of 1 channel, TS 3.3.1 requires power to be>P-10 where the PR instruments

are operable, or<P-6 where SR will be operable to provide protection

against uncontrolled

rod withdrawl (TS Basis).D.incorrect-

if the plant could get to>10%power then the plant could remain in this mode indefinitely.

Since the reactor power is in the IR there is not time to get 100/0 power due tot he hold at 85 AND due to placing a SGFP onservice and meeting all m'ode 1 entry requirements.

Also the 8%power is not high enough to stay there and the UOP-1.3 has the plant stabilize at 8%.F THERMAL POWER>P-6 F.1 Reduce THERMAL and<P-10, one POWER to<P-6.Intermediate

Range Neutron Flux channel OR inoperable.

F.2 Increase THERMAL POWER to>P-10.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours With thermal power>P-6 and<P-IO, one intermediate

range neutron flux channel inoperable

requires that thermal power be either reduced to<P-6 or raised to>P-IO within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .A failure of both intermediate

range detectors, with thermal power between P-6 and P-IO, requires suspension

of operations

involving positive reactivity

additions immediately

and reduction of thermal power to<P-6 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .If thermal power is<P-6 and either one'or both of the intermediate

range detectors become inoperable, actions must be taken to restore channel(s)

to operable status prior to increasing

thermal power to>P-6.Monday, January 14,20082:42:21

PM 85

QUESTIONS REPORT for 75 RO Questions 033 AK3.01 Loss of Intermediate

Range Nuclear Instrumentation

Knowledgeofthe reasons forthe following responses as they apply to the Loss of Intermediate

Range Nuclear Instrumentation:

Terminati*on

of startup following loss of intermediate-

range instrumentation

Question Number: 58 Tier 1 Group 2 Importance

Rating: 3.2 Technical Reference:

TS 3.3.1 Function 4;A-181007 T5-1'Proposed references

to be provided to applicants

during examination:

none Learning Objective:

10 CFR Part 55 Content: 41.10/43.2

FARLEY NO Scramble Range:A-D FARLEY MODIFIED HIGHER RO GO Comments:.replaced question as the original question required SRO knowledge.

MCS Time: 1 Points: 1.00 Version:0123456789

Answer: AC CDAABDAB Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:21

PM 86

QUESTIONS REPORT for 75 RO Questions 34.035 K4.06 001 Which ONE of the following describes the design requirement

of the Steam Generator Safety Valves?Limits SG pressure to no greater than 1100/0 of design_A.assuming a 100%loss of load ,with no credit taken for reactor'trip

on High PRZR pressure.assuminga1 00%loss of load with no credit taken for automatic steam dump or rod control operation.

C.assuming a limiting ATWT initiated by a loss of feedwater with no credit taken for operation of primary PORVs or secondary ARVs.D.assuming a limiting ATWT initiated by a loss of feedwater with no credit taken for operation of primary PORVs or automatic steam dump operation.

A is incorrect but credible because credit is not taken for the ionitiating

event, although the turbine trip could ultimately

give a high pressure trip-It is credited.B is correct.C is incorrect because the initiating

event is wrong.Credible because turbine trip and PORV operation are tied to

valves and their capacity, and the events not credited in these options are similar to actual conditions

for the safety analysis.D is incorrect because the initiating

event is wrong and the PORV operations

is wrong.035 Steam Generator System K4.06 Knowledge of S/GS design feature(s)

and/or interlock(s)

which provide for the following:

S/G pressure Question Number: 35 Tier 2 Group 2 Importance

Rating: 3.1 Technical Reference:

TS basis, 3.7.1 FSAR, 15.2 Proposed references

toprovided to applicants

during examination:

None Learning Objective:

___--"--"10 CFR Part

41

_Comments: this is written at an S'RO level due to the references, but it is a base level of knowledge about the design of the SG safety valves.Monday, January 14, 2008 2:42:21 PM

MCS Time: Source: Cognitive Level: Job Position: reviewed: 1 Points: NEW LOWER RO GO QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer:BCB BDB CADD Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: Scramble Range:A-D FARLEY NO Monday, January 14, 2008 2:42:21 PM 88

QUESTIONS REPORT for 75 RO Questions 35.038 EKl.02 001 Given the following:*A Steam Generator Tube Rupture has occurred on Unit 1.*RCS cooldown and depressurization

are complete.*The creW is maintaining

the plant stable while preparing to transition

to ESP-3.1, Post SGTR Cooldown using Backfill.*The ruptured SG narrow range level is 73%and slowly decreasing.

PRZR level is approximately

38%and slowly increasing.

The OATC turns on PRZR heaters lAW the guidance in EEP-3, Steam Generator Tube Rupture.Which ONE of the following describes the reason for this action?To maintain pressurizer

saturation

temperature

_corresponding

to ruptured SG pressure to minimize SG leakage into the RCS.B.above the intact SG pressure to maintain adequate secondary heat sink with

SGs.C.above the corresponding

ruptured SG pressure to ensure RCS Subcooling

is maintained.

D.corresponding

to intact SG pressure to ensure RCS Subcooling

is maintained.

A is correct.Attempting

to maintain an inventory balance between RCS and ruptured SG prior to ruptured SG cooldown.B is incorrect because if intact SG pressure was higher than RCS pressure, EEP-3.0 would not be the governing procedure, ECP-3.1 would.C.is incorrect.

Coold'own is to ensure subcooling.

RCS and ruptured SG will act like 2 pressurizers.

Subcooling

is not the issue.D.is incorrect.

Cooldown is to ensure subcooling.

RCS and ruptured SG will act like 2 pressurizers.

Subcooling

is not the issue.Background

document for EEP-3 page 83 of 119 Purpose: To control RCS pressure and charging flow to maintain an indicated pressurizer

level while minimizing

primary-to-secondary

leakage.Basis: In order to explain the basis for the guidance provided in this step, consider again equilibrium

conditions

between leakage through the failed SG tube and charging flow, as shown in Figure 30.For primary system pressures greater than the ruptured steam generator pressure (PSG), primary-to-secondary

leakage will occur so that excess charging flow, i.e., greater than letdown and coolant shrinkage, is necessary to maintain pressurizer

inventory.

Conversely, for letdown flows greater than charging flow, the equilibrium

RCS pressure is less than the Monday, January 14,20082:42:21

PM 89

QUESTIONS REPORT for 75 RO Questions ruptured steam generator pressure and secondary-to-primary

leakage will occur.The ideal conditions, shown by Point B, occur when charging flow exactly compensates

for letdown and coolant shrinkage so that RCS pressure and the ruptured steam generator pressure equalize.For these conditions

both the pressurizer

and ruptured steam generator inventories

will remain constant.Obviously fluctuations

about these ideal conditions

will occur due to variations

in ruptured steam generator pressure, cooldown rates, and letdown flows.Consequently, the operator must continuously

adjust RCS pressure and charging flow to control pressurizer

and ruptured steam generator inventories.

This step provides guidance for performing

these actions in the form of a table.Figure 30 can be divided into four different regions which are characterized

by pressurizer

and ruptured steam generator level behavior.For primary pressures greater than the ruptured steam generator pressure, leakage into the steam generator will increase steam generator water level (LSG).Alternatively, water level will decrease forRCS pressures less than the ruptured steam generator pressure.Similarly, pressurizer

level (LPRZR)will increase for RCS pressures less than equilibrium.

This leads to the four regions illustrated

in Figure 30.The steps one performs to stabilize the plant at the ideal, equilibrium

conditions

depend on the pressurizer

inventory and ruptured, steam generator water level behavior.For example, if pressurizer

level is low, region II or region ill must be entered to increase pressurizer

level.This requires one to increase charging flow or decrease RCS pressure, as shown in Figure 30.The further into these regions, the more rapidly pressurizer

level will increase.Of course, if pressurizer

level is high, the opposite response would be necessary.

However, the ruptured steam generator water level must also be considered.

STEP DESCRIPTION

TABLE FOR E-3Step29 If the steam generator water level is increasing, RCS pressure must be reduced to stop primary_to_secondary

leakage.If the steam generator water level is decreasing, primary pressure should be increased by energizing

pressurizer

heaters to minimize leakage into the ReS.Note that in some cases, actions which address pressurizer

level conflict with those which address steam generator*

level.For example, if steam generator level is increasing

one must decrease RCS pressure.Since this will also increase pressurizer

level, the pressurizer

could fill with water if level is initially high.However, by reducing charging flow, pressurizer

level will decrease.Since this will also decrease RCS pressure if heaters are not energized, steam generator level will also stabilize.

Hence, for this situation the preferred action is to reduce charging flow.038 Steam Generator Tube Rupture EK1..02 Knowledge of the operational

implications

of the following concepts as they apply to the SGTR: Leak rate vs.pressure drop Question Number: 47 Tier 1 Group 1 Importance

Rating: R03.2 Technical Reference:

EEP-3 and background

documents page 83 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

1 OCFR Part 55 Content: 41.5 Scramble Range:A-D Version: a 123456789 Answer: ADBBBB DDC C Comments: this meets the KA in that the question tests the concept of why PRZR pressure and temp affect the leak rate into or out of the Przr MCS Time: 1 Points: 1.00 Monday, January 14, 20082:42:21

PM 90

Source: Cognitive Level: Job Position: reviewed: BANK LOWER RO GTO QUESTIONS REPORT for 75 RO Questions Source if Banle MCGUIRE 2003 Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:21

PM 91

QUESTIONS REPORT for 75 RO Questions 36.039 Al.lO 001 Given the following:

Unit 1 is at 100%power.*A Steam Generator Tube Leak has developed.

R-15A, SJAE EXH, is approximately

mid-scale.

Which ONE of the following describes:

(1)the R-15A indication

that will be observed if SJAE Filtration

is placed on service and (2)the R-15A indication

that will be observed if a reactor trip occurs?R-15A.........(_1.-.-)when SJAE Filtration

is placed on service.R-15A (2)if the reactor trips.A.(1)Trends down (2)Remains stable B.(1)Trends down (2)Trends down C.(1)Remains stable (2)Remains stable(1)Remains stable (2)Trends down Monday, January 14,20082:42:21

PM 92

QUESTIONS REPORT for 75 RO'Questions A.Incorrect (1)is incorrect, Plausible because it is true for R-158&R-.15C which are both downstream

of the SJAE Filtration

system (which is normally bypassed)(2)is incorrect.

but plausible because the reactor trip does not isolate or stop the SG Tube leak, and may not consider the dip across the tube is proportional

to leak flow.8.Incorrect (1)is incorrect, Plausible because it is true for R-158&R-15C which are both downstream

of the SJAE Filtration

system (which is normally bypassed)(2)is correct.C.Incorrect (1)is correct.(2)is incorrect, but plausible because, the reactor trip does not isolate or stop the SG Tube leak, and may not consider the dip across the tube is proportional

to leak flow.D.Correct.When the Steam Jet air ejector Filtration

system is placed on service there is no change in the reading since the SJAE is upstream of the Filtration

system.When the reactor trips, steam flow is decreased, steam pressure goes up,&dip across SG U-tubes goes down.This causes tube leakage rate to decrease which causes R-15 indication

to decrease.039 A1.10 Main and Reheat Steam System (MRSS)Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits)associated

with operating the MRSS controls including:

Air ejector PRM Question Number: 17 Tier 2 Group 1 Importance

Rating: 2.9 Technical Reference:

RAD MONITORING

LP OPS 52106D Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5/45.5.Comments: Points: 1.00 MCS Time: Version: a123456789 Answer: DCACCBBABD Scramble Range:A-D Source: NEW Source if Banle Cognitive Level: HIGHER Difficulty:

Job Position: RO Plant:.FARLEY--------1'"reviewed-o--.

t-G-.rt-Q-l---------J,;1-Prro-'evious

2 NRC exaUlS.NO Monday, January 14,20082:42:21

PM 93

QUESTIONS REPORT for 75 RO Questions 37.041 A1.02 016 Given the following:

Unit 1 reactor power is steady at 14%.*Tavg is at 551 0 F.*Rod control is in Manual.*Turbine power is steady at 7%.*Steam Dumps are open in the STM PRESS mode.*PK-464, STM HDR PRESS, controller

is in AUTO and set to control temperature

at 551Which ONE of the following is the correct response of the steam dump system if PT-464, STM HDR PRESS, fails HIGH under these conditions?

Assume no operator action is taken.*All steam dumps will open and then close at P-12.*PK-464 will shift to MANUAL.B.*All steam dumps will open and then close at P-12.*PK-464 will remain in AUTO.c.*All steam dumps will open and then cycle at P-12.*PK-464 will remain in AUTO.D.*All steam dumps will open and then cycle at P-12.*PK-464 will shift to MANUAL.Monday, January 14, 2008 2:42:21 PM 94

QUESTIONS REPORT*for 75 RO Questions A.Correct-The low-low T AVG (P-12)block actuates when 2/3 T AVG instruments

indicate below 543°F.It should be noted that the AUTO feature of PK-464 can be selected only under certain conditions.

First, if the mode selector switch is in the T AVG mode, PK-464 shifts to'manual control.Secondly, if the low-low T AVG signal (P-12)exists and th.e BYP INTLK position on both A and B Train STEAM DUMP INTERLOCK SWITCHES has not been selected, PK-464 will shift to manual control.Byshiftingto

manual control, the output of the P+I portion of the controller

is set to zero and thus prevents small pressure errors from being integrated

into large controller

output signals.B.Incorrect-

they do go closed at 543°F, and the block does reset at 545°F, however, the controller

shifts to minimum and manucil and the dumps do not cycle.C.Incorrect-

the steam dumps will go

and shift to minimum and manual.They will not cycle.D.Incorrect-

the dumps do open and PK 4164 will go to manual but the dumps will remain closed and.not cycle to control

041 Main Turbine Generator System A1.02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits)associated

with operating the 80S controls including:

Steam pressure Question Number: 36 Tier 2 Group 2 Importance

Rating: 3.1 Technical Reference:

SD LP OPS-52201 G;AOP-100 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5/45.5 Scramble Range:A-D FARLEY BANK HIGHER RO GO Source: Cognitive Level: Job Position: reviewed: Comments: this meets the KA in that the failure of the stm pressure transmitter

affects the dumps and RCS temp and the operator has to predict and monitor the RCS for these changes.MCS Time: 1 Points: 1.00 Version: a123456789 Answer: AC BADBDCB C Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:21 PM 95

QUESTIONS REPORT for 75 RO Questions 38.054 AA2.05 001 The following plant conditions

exist:*Unit 1 is operating at 28%power.*A Feed Water control malfunction

has caused1A SG NR level to reach 84%.As'suming the plant re'sponds AS DESIGNED, with NO operator action, which ONE of the following describes the current valve alignment?

Assume all valves were open prior to the event Main Feed SGFP Feedwater Reg Valves Discharge Valves Isolation Valves FCV-478,488, 498 MOV503A, B MOV-3232A,B,C

A.All open All open All shut B.All shut All open AII'shut C.All open All shut All openAll shut All shut All shut A.Incorrect-The SGFP trip will cause the discharge valves to go closed.The FWI signal causes the FRV and bypasses to close.B.Incorrect-SGFP discharge valves go shut due to the SGFP trip at 82%level.C.Incorrect-FRVs close on FWIS and FWI valves close on SGFP trip.D.Correct.See FSD, Student text,&SOP-21.0.The FWI signal will cause the SGFPs to trip and the main turbine to trip.The SGFP trip will cause the discharge valves to go closed and the FW isolation.valves

to go closed.The FWI signal causes the FRV and bypasses to close.FNP Units 1&2 REACTOR PROTECTION

SYSTEM A-181 007[FSD]: 2-26 Rev.10 2.7.1 4.Main Feedwater Isolation and Turbine Trip The Main Feed Line Isolation is initiated to prevent excessive cooldown of the reactor or to lessen the severity of the transient overall.The following signals are utilized to initiate the Main Feed line Isolation:

a.Safety injection b.High-high steam generator water level (P-14)set at=820/0 of narrow range steam generator span on 2/3 coincidence

c.Low Tavg;=554°F in coincidence

with reactor trip P-4 Monday, January 14,20082:42:21

PM 96

QUESTIONS REPORT for 75 RO Questions manual reset to clear.(References

6.1.022,6.4.007,6.4.015,6.7.012,6.4.21)

7.High-High Steam Generator Water Level If water level in a steam generator increases to 82%on 2/3 narrow range level instruments, the main turbine trips, the main feedwater pumps trip and main feedwater isolation signals are initiated.

Tripping the turbine is a protective

measure to ensure no damage occurs from moisture carry-over.

Main feedwater is isolated so that no further water is added to the steam generator with thehighhigh level to protect the primary side from excessive cooldown when safety injection is actuated.(References

6.1.022, 6.4.007,6.4.014,6.7.012)

SOP-21.0, CONDENSATE

AND FEEDWATER SYSTEM 4.6.3 At approximately

2450 RPM, trip the SGFP.4.6.4 Verify that the SGFP high and low pressure stop valves are closed.NOTE: In the following step, annunciator

KC3 should clear after approximately

three minutes.4.6.5 Verify that annunciator

KC3, lA OR IB SGFP TRIPPED comes in.4.6.6 Verify closed the lA(lB)SGFP DISCH VALVE, NIN21V503A(B).

4.6.7 At the lA(lB)SGFP Oil Test Station, depress the LOW LEVEL TEST push-button

until the low level alarm light is illuminated.

CONDENSATE

&FEEDWATER STUDENT TEXT, OPS-52104C, OPS-40201B

Main Feedwater Stop Valves (3232A, B, and C)(Figure 14)A three-position

handswitch (CLOSE/AUTO/OPEN, spring return to AUTO)on the MCB controls each motor-operated

isolation valve.In AUTO, the valve automatically

closes on a SGFP trip signal from both pumps.Valve position lights indicate above each switch.APE 054 Loss of Main Feedwater-AA2.05 Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW): Status of MFW pumps, regulating

and stop valves Question Number: Tier 1 Group 1 Importance

Rating: Technical Reference:

48 R03.5 FSD, Student text,&SOP-21.0.Proposed references

to be provided to applicants

during examination:

None 1 a CFR Part 55 Content: 41.1 0 Comments: original question did not meet KA.Replaced.This question meets the KA in that when the SGWL reached 82%, the SGFPs tripped.The operator has to determine the status of the Monday, January 14, 20082:42:21

PM 97

FARLEY NO Scramble Range:A-D FARLEY QUESTIONS REPORT for 75 RO Questions 1.00 Version:0123456789

Answer: DADBDAABDA

Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: BANK HIGHER RO GTO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:22

PM 98

QUESTIONS REPORT for 75 RO Questions 39.055 EAl.02 001 Given the following:*A LOSS OF ALL AC POWER has occurred on Unit 1.*VA2, 1B DG GEN FAULT TRIP, annunciator

has come into alarm.*The crew is at the step in ECP-O.O, Loss of All AC Power, to verify breakers for'major, loads OPEN.*A Safety Injection occurs on Unit 1 at this time.Which ONE of the following describes how the 2C DG will be started and the events that will take place or need to take place to energize the ESF equipment?

A.*Start 2C DG from EPB in Mode 2 using the start pushbutton.

The LOSP sequencer will run to start all ESF loads.B.*Start 2C DG from EPB in Mode 2 using the start pushbutton.

ALL ESF loads will have to be manually aligned.C.*Start 2C DG from EPB in Mode 1 using the start pushbutton.

The LOSP sequencer will run to start all ESF loads.*Start 2C DG from EPB in Mode 1 using the start pushbutton.

ALL ESF loads will have to be manually aligned.Monday, January 14, 20082:42:22

PM 99

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

not mode2-ECP-O.O step 5 directs starting 2C DG in MODE 1.the LOSP sequencer will not run due to the SI signal present.Using MODE 2 is plausible since all other DGs are started in Mode 2 in this condition lAW ECP-O.2C DG is the only DG that is started in Mode 1.The sequencer is plausible since it would run if the SI signal was not present.B.Incorrect.

not mode 2.second part is correct.C.Incorrect.

first part is correct.Second part is NOT correct.D.Correct.ECP-O.O step 5 directs starting 2C DG in MODE 1.the LOSP sequencer will not run due to the SI signal present.see below.The LOSP sequencer will not run per the note below in ECP-O with an SI signal present.ECP-O note at step 5.2.1.5 RNO NOTE: The LOSP sequencer should run when output breaker closes, if no SI signal is present.If an SI signal is present, neither sequencer will run and Slloads must be started manually.5.2 Perform the following:

5.2.1 Perform 2C DG SSO start as follows.5.2.1.1 Verify 2C DG MODE SELECTOR switch in MODE 1.5.2.1.3 WHEN load shed verified,THEN

depress 2C DG DIESEL START pushbutton.

055 EA 1.02 Station Blackout Ability to operate and monitor the f.ollowing

as they apply to a Station Blackout: Manual ED/G start Question Num-ber: 49 Tier 1 Group 1 Importance

Rating: RO 4.3 Technical Reference:

FNP-1-ECP-O.O

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41..10 Comments: replaced this question since the original did not meet the KA.This demonstrates

the ability to start and monitor a manual start of the 2C DG and the subsequent

actions to energize equipment In that train after the start.MCS Time: Points: 1.00 Version:0123456789

Answer: DADCDCAAAC

Scramble Range:A-D Monday, January 14, 20082:42:22

PM 100

Source: Cognitive Level: Job Position: reviewed: NEW HIGHER RO GTO QUESTIONS REPORT for 75 RO Questions Source if Banle Difficul ty: Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14,2008 2:42:22 PM.101

QUESTIONS REPORT for 75 RO Questions 40.056 A2.04 010 Given the following:

Unit 2 is at 80%power ramping to 100%.*2C Condensate

Pump has tripped.*Annunciator

KB4, SGFP SUCT PRESS LO, came into alarm 35 seconds ago.*PR-4039, SGFP Suction PRESS recorder, indicates SGFP pressure is 280 psig and is decreasing.

Which ONE of the following is the expected result of this condition and action required per AOP-13.0, Condensate

and Feedwater Malfunction?

A.*The standby condensate

pump should have AUTO started;*Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.*The standby condensate

pump should have AUTO started;*Verify the standby condensate

pump started and if suction pressure is still falling, then reduce load rapidly lAW AOP-17, Rapid Load Reduction.

C.*The standby condensate

pump should NOT have AUTO started;*Trip the reactor and enter EEP-O, Reactor Trip or Safety Injection.

D.*The standby condensate

pump should NOT have AUTO started;*Verify the standby condensate

pump started and if suction pressure is still falling, then reduce load rapidly lAW AOP-17, Rapid Load Reduction.

Monday, January 14, 2008 2:42:22 PM 102

QUESTIONS REPORT for 75 RO Questions The condensate

pumps have 2 auto starts, one on the trip of the other pump and one for suction pressure<275 psig for>10 sec.In this question the auto start for suction pressure is not met but the, other one is.This has been a high miss question since most do not think of the autostart for the tripped pump.A-Incorrect;

The stby condensate

pump should start immediately

when the other condensate

pump trips (if the stby pump is in AUTO, which is the normal alignment at 80%power).Plausible because the low suction pressure auto start of<275 psig for>10 seconds has not yet been met.Tripping the reactor is not required unless approaching

trip criteria or if 80TH SGFPs are tripped and this has not happened.8-Correct;The standby condensate

pump SHOULD auto start immediately

when the other condensate

pump trips.With suction pressure dropping, AOP-13 directs verifying stby pump started prior to 275 psig decreasing, If pressure continues to drop, rapidly ramp down lAW AOP-17, Rapid Load Reduction.C-Incorrect;

first part is NOT true: condensate

pump should have auto started on the trip of the other pump.second part is not correct for this situation.

It is plausible in that if the candidate thought both SGFPs tripped due to the alarm being in for>275 psig, then this would be correct.D-Incorrect;

first part is NOT true: condensate

pump should have auto started on the trip of the other pump.Second part is incorrect for this condition also, but plausible because the ramp at:s.5 MW/MIN is incorrect for a condensate

pump but plausible since it is correct for a HDT pump trip also covered by AOP-13.At 275 psig falling the standby condensate

pump will start after 10 sec.IF suction pressure is NOT greater than 275 psig within 30 sec, THEN the SGFP's will trip.This could result in a reactor trip.Monday, January 14, 20082:42:22

PM 103

QUESTIONS REPORT for 75 RO Questions 056 A2.04 Condensate

System Ability to (a)predict the impacts of the following malfunctions

or operations

on the Condensate

System;and (b)based on those predictions, use procedures

to correct, control, or mitigate the consequences

of those mal-functions

or operations:

Loss of condensate

pumps Question Number: 37 Tier 2 Group 2 Importance

Rating: 2.6 Technical Reference:

AOP-13&ARP 1.10 KB4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.1 0 Scramble Range:A-D Version:0123456789

Answer: B DCADADDB B Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO MODIFIED HIGHER RO GO Comments: significantly

modified the distracters

to make sure this does not answer 059 A 1.03 and this is also a different part of AOP-13.MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:22 PM 104

QUESTIONS REPORT for 75 RO Questions 41.057 AK3.01 001 Given the following:

Unit 1 is at 14%power and ramping up in preparation

for rolling the main turbine.*The bypass feed regulating

valves are in service, and SG level is being'maintained

at 65%.*1G.4160 V Bus has been de-energized

due to DG15,.1B.S/U XFMR to 4160V Bus 1 G, tripping open.*The1B DG has re-energized

the'BI Train emergency buses.Which ONE of the following describes the correct operator actions lAW AOP-5.0, Loss of A or B Train Electrical

power?.A.Trip the reactor and restore the1G 4160 V bus to the grid.B.Shut down the reactor and place the unit in Mode 3.Stabilize the plant and restore the1G 4160 V'bus to the grid.D.Verify the reactor tripped and stabilize the unit in Mode 3.AOP-5.0 Version 24 A.incorrect-AOP-5.0 no longer directs tripping the reactor.B.Incorrect-

Ramping off'Iine is not required in AOP-5.C.correct-lAW AOP-5, step 14 the long term status is to maintain the reactor stable and then restore the grid, then continue with whatever procedure the operatorwas

in at the time of the problem.This was changed a few years ago to prevent an unnecessary

reactor trip for this condition.

D.Incorrect-an automatic reactor trip will not occur.Monday, January 14, 2008 2:42:22 PM 105

QUESTIONS REPORT for 75 RO Questions.APE 057 Loss of Vital AC Instrument

Bus-AK3.01 Loss of Vital AC Instrument

Bus, Knowledge of the reasons for the following responses as they apply to the (Loss of Vital AC Instrument

Bus): Actions contained in the EOP for loss of vital AC electrical

instrument

bus.Question Number: 56 Tier 1 Group 1 Importance

Rating: 4.1 Technical Reference:

AOP-5.0 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.10 Scramble Range:A-D FARLEY o 123456789 CCCCBAACBA

Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO MODIFIED HIGHER RO GO Comments: this KA is for a loss ofa vital bus which is an emergency bus and/or panel at FNP.This meets the KA in that there is a loss of the vital bus and the procedure guidance of AOP-5 is followed.The reason for those actions is implied and agreed upon as part of the actions that would be done..FNP does not have a loss of a vital panel procedure and that issue is vaguely discussed in an ARP in which it says if a vital panel has been lost, then recover from it when the event is over by doing the following that apply.MCS Time: 1 Points: 1.00 Version: Answer: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:22 PM 106

QUESTIONS REPORT for 75 RO Questions 42.058 G2.1.32 002 Given the following on Unit1:*A loss of"A" Train Auxiliary Building Battery bus has occurred.*The crew is performing

AOP-29.1, Plant.Stabilization

in Hot Standby and Cooldown Without"A" Train AC or DC Power.*The RCS temperature

is 547°F and pressure is being maintained

2220 psig.*Seal injection flow has been lost to all three Reactor Coolant Pumps (RCPs).*LB3, RCP THRM BARR ISO HV-3184 AIR PRESS LO, annunciator

has come into alarm.*All three RCPs have been tripped.Which ONE of the following correctly describes one of the actions required by AOP-29.1 for the conditions

given above, and the reason for performing

that action?A':'Isolate seal injection flow to all RCPs to prevent potential RCP seal damage.B.Isolate seal injection flow to all RCPs to prevent a potential radioactive

release to the Auxilia'ry

Building from occurring.

C.Isolate seal return flow to all RCPs to prevent potential RCP seal damage.D.Isolate seal return flow to all RCPs to prevent potentialbarrierheat

exchanger damage.A is correct.CAUTION (for step 5): CAUTION: To prevent potential'

seal damage, neither seal injection flow nor CCW flow to the thermal barrier shall be re-established

to an RCP which has lost both seal injection and CCW cooling.the background

documents for a loss of all ac describe what happens on a loss of chg and CCW to a seal.The seal injection is isolated to prevent potential seal damage.B is incorrect.

This is a correct action but the reason is for the seal return flow.C.is incorrect.

see explanation

below.D is incorrect.

seal return is isolated but not to prevent seal damage.It is isolated for several reasons: Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT)(via seal return relief valve outside containment)

and subsequent

transfer to other auxiliary building holdup tanks (via VCT relief valve)with the potential for radioactive

release within the auxiliary building.Such a release, without auxiliary building ventilation

available, could----.---1-1-'1

imit-p-er-so-rm-et-aceess--f-or-toeatoperations.

Isolating the s.eal return line also enables pressure in the number 1 sealleakoff

line to increase up to the relief valve setpoint of 150 psig.Maintaining

a backpressure

in the sealleakoff

line.of at least 150 psig enables development

of high pressure in the number 1 seal leakoff cavity with a steady-state

seal leakage rate established

due to the self-limiting

leakage characteristic

of Monday, January 14, 2008 2:42:22 PM 1 07

QUESTIONS REPORT for 75 RO Questions the number 1 seal.Under these conditions, with the number 1 seal functioning

as expected and the number 2 seal remaining closed, the expected leakage flow rate is 21.1 gpm/pump.This is consistent

with the steady statepressuredistribution

and seal leakage determined

in the WCAP-10541

analysis and used in the latest RCP seal leakage PRA model in WCAP-15603

3.0 IF ONE of the following conditions

occur at anytime during the event, AND cannot be readily restored.[]A total loss of RCP seal cooling as indicated by loss of seal injection and loss of CCW to the Thermal Barrier Heat Exchanger.

OR[]A total loss of the operating train of charging without the ability to quickly restore the redundant train, and RCP sealleakoff

before the loss was less than 2.5 gpm per pump.bac,kground

document for ECP-O Purpose: To isolate the RCP seals Basis: This step groups three'actions, with different purposes, aimed at isolating the RCP seals.The actions are grouped since all require an auxiliary operator, dispatched

from the control room, to locally close containment

isolation valves (the reference plant utilizes motor operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal injection lines).This grouping assumes that the subject valves are located in the same penetration

room area and that they are accessible.

Concurrent

with dispatching

the auxiliary operator, the control room operator should place the valve switches for the motor operated valves in the closed position so that the valves remain closed when ac power is restored..Isolating the RCP seal injection lines prepares the plant for recovery while protecting

the RCPs from seal and shaft damage that may occur when a charging/51

pump is started as part of the recovery.With the RCP seal STEP DESCRIPTION

TABLE FOR ECA-O.OStep

8 injection lines isolated, a charging/51

pump can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs.Seal injection can subsequently

by established

to the RCP consistent

with appropriate

plant specific procedures.

Isolating the RCP thermal barrier CCW return outside containment

isolation valve prepares the plant for recovery while protecting

the CCW system from steam formation due to RCP thermal barrier heating.Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal area.Monday, January 14, 20082:42:22

PM 108

QUESTIONS REPORT for 75 RO Questions APE 58 G2..1.32 Loss of DC Power Conduct of Operations:

Ability to explain and apply all system limits and precautions.

Question Number: 50 Tier 1 Group 1 Importance

Rating: 3.4 Technical Reference:

AOP-29.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7/1 0 Items Not Scrambled NEW HIGHER RO GO Comments: This meets the KA in that under the conditions

given, a loss of DC power, a system caution to isolate CCW and Seal injection to the RCP seal to prevent seal damage upon re-initiation

of either Seal injection or CCW flow which could cause a SBLOCA from the RCP seal of 300 gpm.This question asks the operator to explain why these actions are taken.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: AAAAAAAAAA

Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:22

PM 109

QUESTIONS REPORT for 75 RO Questions 43.059 Al.03 001 Given th'e following:

Unit 1 is at 100%power.*1A SGFP has tripped.*The crew has entered AOP-13.0, Condensate

and Feedwater Malfunction.

Emergency Boration is in progress.Which ONE of the following describes the subsequent

restrictions

on operation of the unit in accordance

with AOP-13.0?(1)Load must be reduced to_(2)The reactor must be tripped immediately

if_A.(1)730 MWe;(2)SG NR levels cannot be maintained

above the minimum value specified.

B.(1)730 MWe;(2)FE1, CONT ROD BANK POSITION LO, annunciator

comes into alarm with Tavg at 577°F.(1)540 MWe;(2)SG NR levels cannot be maintained

above the minimum value specified.

D.(1)540 MWe;(2)FE1, CONT ROD BANK POSITION LO, annunciator

comes into alarm with Tavg at 577°F.Monday, January 14, 2008 2:42:22 PM 110

QUESTIONS REPORT for 75 RO Questions A.is incorrect.

Due to load setpoint, although plausible because this is the value at which the decrease load button is released when DEH manual load reduction is used per the RNO column.B is incorrect.

Due to load setpoint, although plausible because this is the value at which the decrease load button is released when DEH manual load reduction is used per the RNO column.T AVG 541°F-580°F 577 F is still within the band to not trip.AOP-13 has a step to evaluate the plant per the below: 1.14 IF the Team is NOT confident that a parameter is being restored, THEN trip the reactor and go to FNP-1-EEP-O, REACTOR TRIP OR SAFETY INJECTION.

one of the parameters

checked is FE2, not FE1.Since the team is emergency borating per the procedure, it is unlikely this alarm would come in, but if it did the reactor is not required to be immediately

tripped.If FE2 came into alarm, then action here would be to place rods in manual and with the emergency boration in progress, evaluate the plant, not immediately

trip the reactor.C.Correct.AOP-13.0 directs load to be reduced to 540 MWe, and SG NR levels must remain above 28%.AOP-13 step 1.8 IF SG narrow range levels NOT maintained

greater than 28%, THEN trip the reactor and go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION.

D is incorrect.

see B above for second part.TAVG 541°F-580°F 577 F is still within the band to not trip.OPERATOR ACTION for FE!1.Check indications

and determine that actual control bank rod position is at low insertion limit.1.1 Click on Rod Supervision

button on Applications

Menu.1.2 Click on Rod Insertion Limits button.1.3 Determine if low insertion limit exceeded.2.IF reactor coolant system dilution is in progress, THEN stop dilution.3.IF a plant transient is in progress, THEN place the turbine load on"HOLD".4.Refer to FNP-1-UOP-3.1, POWER OPERATIONS.

5.Borate the Control Bank"OUT" as necessary using the Boron Addition Nomographs.

{CMT 0008900}6.Refer to the Technical Specifications

section on Reactivity

.-C-'-F\ontror+--l.----------------------------------

Monday, January 14,20082:42:22

PM 111

QUESTIONS REPORT for 75 RO Questions 059 A1.03 Main Feedwater System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits)associated

with operating the MFW controls including:

Power level restrictions

for operation of MFW pumps and valves.Question Number: Tier 2 Group 1 18 Importance

Rating: 2.7 Technical Reference:

AOP-13.0, FE2 and KB4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.7 Comments: good match for KA since it tests power level restrictions

with 1 SGFP and monitoring

levels to prevent exceeding design limits and the actions required.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER RO GO Version:0123456789

Answer:CCCACCDACC

Scramble Range:A-D Source if Barue Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:23 PM 112

QUESTIONS REPORT for 75 RO Questions 44.061 K1.07 001 Which ONE of the following describes the Service Water Train normally aligned to the TDAFW Pump for Emergency Makeup, and if the TDAFW Pump is running and aligned to the CST, the MINIMUM time available to swap to the emergency supply when JD4 and JE4, CST LVL La-La A and B TRN, alarms are received?(1)

Water Train normally aligned to_(2)The minimum time available to swap to the emergency supply after receiving the CST La-La level alarm is_Atf (1)A Train (2)20 minutes B.(1)A Train (2)2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C.(1)B Train (2)20 minutes D.(1)B Train (2)2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months A is correct.The CST La-La level alarm received means that at least 20 minutes of normal supply remains.65,300 Gallons (5 1 3")is when the La-La comes in.B is incorrect.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is the time that 150,000 gallons of CST will maintain Hot Standby..C is incorrect.

Wrong train, plausible because B train can be aligned'and would be if A train power or SW was unavailable.

D is incorrect.

Wrong train and time.SOP-22, AFW section 4.7 4.7 Aligning Service Water to the AFW System.This shows how to swap SW to B train and how to align it to A Train.The initial valve line up per the sOP also shows it is aligned to A Train and D-l 75007 shows the normal line uR is to A Train.FSAR 6.55 instrumentation

after recieving low level alarm setpoint Monday, January 14, 20082:42:23

PM 113

QUESTIONS REPORT for 75 RO Questions 061 K1..07 Auxiliary Emergency Feedwater (AFW)System Knowledge of the physical connections

and/or cause-effect

relationships

between the AFW and the following systems: Emergency water source.Question Number: Tier 2 Group 1 Importance

Rating: Technical Reference:

19 3.6 LP 52102H AFW D-175007 and SOP-22 AFW Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GTO Version:0123456789

Answer:AAAA DCA D B C Scramble Range: A-" D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14,2008 2:42:23 PM 114

QUESTIONS REPORT for 75 RO Questions 45.062 G2.1.23 001 Given the following:

Unit 1 is operating at 38%power with IA 1 Train on service.*The1C Service Water Pump is tagged out for motor replacement.

A fire is reported on the1K 4160 volt bus and the plant operators de-energize1K 4160 volt bus.The following conditions

exist:*CCW FROM CCW HX TEMP, TI-3042C, is elevated slightly and rising slowly.*PI-3001 A, A Train SW TO CCW HX HDR PRESS, reads 42 psig.*PI-3001 B, B Train SW TO CCW HX HDR PRESS, reads 75 psig.Which one of the following actions are required lAW AOP-1 0.0, Loss of Service Water?A.Trip the reactor and perform EEP-O, Reactor Trip or Safety Injection.

B.Reduce power to less than 35%, then trip the Main Turbine andreferto AOP-3.0, Turbine Trip below the P-9 Setpoint.Start1A CCW pump and1C charging pump, secure1A charging pump, and swap on service trains of CCW.D.Start1C CCW pump and1A charging pump, secure1C charging pump, and swap on service trains of CCW.Monday, January 14,2008 2:42:23 PM 115

QUESTIONS REPORT for 75 RO Questions I*A.Incorrect-a rx trip is not required at this time.AOP-1 0.0 Step 4.2.2.2 RNO and step 6.3 has the crew trip the reactor if one train of SW does not have at least 60 psig.Since one train i.s available and operating, a reactor trip is not called for.B.Incorrect-removing the main turbine from service would be an option if both trains of SW were less than 60 psig per the RNO step 6 of AOP-10 if power was less than 35%.Being so close to 35%in the stem and temperatures

are elevated slightly makes this plausible since it would make sense to ramp to below 35%and remove the turbine from service vs trip the rx.C.Correct-Start a CCW pump and charging pump in the nonaffected

train, secure affected train charging pump, and swap on service trains of CCW.This is the correct response because enough time is allowed before RCP temps increase to 195°F to mitigate the loss of SW and prevent the need to trip the reactor.It will take time to heat up the On service train of CCW, and AOP-1 0.0 takes that into account.D.Incorrect-Swapping on-service

trains is partially correct, but the unit is not ramped to 35%power and the main turbine tripped.APE 062 Loss of Nuclear Service Water-G2.1.23 Conduct of Operations:

Ability to perform specific system and integrated

plant procedures

during all modes of plant operation.

Question Number: 51 Tier 1 Group 1 Importance

Rating: 3.9 Technical Reference:

AOP-10 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: this meets the KA in that the ability to perform AOP-1 0 is demonstrated

during mode 1 at a low power level.Points: 1.00 MCS Time: Version: a123456789 Answer:CA CAB BCDAA Scramble Range:A-D

Cognitive Level: IDGHER Difficulty:

Job Position: R.O Plant: FARLEY reviewed: GO Previous 2 NRC exams: NO Monday, January 14, 20082:42:23

PM.116

QUESTIONS REPORT for 75 RO Questions 46.062 K2.01 001 Given the following conditions

on Unit 2:*IA I train is the liOn Service ll train.*2A Charging Pump breaker has been racked out for maintenance.

2G 4.160 V Bus has been de-energized

due to a fault.Which one of the following states the ECCS pumps that will have power based on current conditions?

A':'2B Charging Pump, 2A RHR Pump.B.2B Charging Pump, 2B RHR Pump.C.2C Charging Pump, 2A RHR Pump.D.2C Charging Pump, 28 RHR Pump.A.correct, both powered from 2F B.Incorrect, 28 chg powered from 2F.28 RHR powered from 2G C.Incorrect, 2C chg powered from 2G.2A RHR powered from 2F D.Incorrect, both powered from 2G 062 K2.01 A.C..Electrical

Distribution

Knowledge of bus power supplies to the following:

Major system loads.QuestionNumber:

20 Tier 2 Group 1 Importance

Rating: 3.3 Technical Reference OPS 521038, Unit 1 Equipment Load List A506250 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.5 Comments: KA match in that these are major loads on the emergency busses.MCS'rIme: POInts: lJJO VerSIon: O-r2-r456789 Answer: ADDCDB CAB D Scramble Range:A-D Monday, January 14, 2008 2:42:23 PM 117

Source: Cognitive Level: Job Position: reviewed: MODIFIED LOWER RO GTO QUESTIONS REPORT for 75 RO Questions Source if Bank: FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:23 PM 118

QUESTIONS REPORT for 75 RO Questions 47.063 A3.0l 001 Unit 1 is at 1000/0 power with the following conditions:*1A Battery Charger is on service.*EM personnel are doing preventative

maintenance

on the1A battery..The following indications

are received:*The UNIT 1 AUX BLDG DC BUS-A TRN GROUND DET white light comes on momentarily

and then goes OFF..*Then the following alarms are received:*WC2,1A 125V DC BUS UV OR GND*WC3, 1A 125V DC BUS BATT BKR 72-LA05 TRIPPED*THEN WC2 clears.Which ONE of the following describes the status of the indications

on the EPB for the 1A DC BUS and the1A and1B Inverters?

1A DC BUS VOLTAGE reads approximately

(1)1A and1B INVERTER AMPERES are reading approximately

(2)A.(1)0 DC VOLTS.(2)25 amps and being powered from the bypass source.B.(1)0 DC VOLTS.(2)0 amps and being powered from the normal source.C.(1)125 DC VOLTS.(2)0 amps and being powered from the bypass source.(1)125 DC VOLTS.(2)25 amps and being powered from the normal source.Monday, January 14, 20082:42:23

PM 119

QUESTIONS REPORT for 75 RO Questions explanation

When the Battery output breaker is opened, LA-05, WC3 will come into alarm due to the b contact from breaker LA05.WC2 shows either a low voltage condition or a ground.In this case it would be a ground.The battery output breaker has opened due to a ground on the battery and when it opens WC2 clears.The annunciators

provide indication

that the breaker opened and the white light provides indication

of the ground.For this set of circumstances, the battery'is

no longer aligned to the bus and the battery charger is carrying the load.The indications

will remain normal and the inverters will have normal indications.

The inverters will not swap to the bypass source and will still be powered from the BC.A.Incorrect.

0 DC volts on the1A DC bus indicates the bus is de-energized.

The bus still has power from the Batt.chger.The inverters will be powered from the BC or the normal supply and will indicate 25 amps.If it were to swap to the bypass source, it would still have amp readings, but if the manual bypass switch were to be placed in the bypass position, then the amps would be 0 amps.B.Incorrect.

0 is not correct for both.Normal is correct.C.Incorrect-125 is correct.0 is not correct and bypass is not correct.D.Correct.125 is correct and 25 is correct from the normal source.063 D.C.Electrical

Distribution

A3.01 Ability to monitor automatic operation of the de electrical

system, including:

Meters, annunciators, dials, recorders, and indicating

lights Question Number: 21 Tier 2 Group 1 Importance

Rating: 2.7 Technical Reference:

ARP WC2, WC3 and 0177082 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

lO CFR Part 55 Content: 41.7 Comments: This was replaced to fully meet the KA.It meets the KA in that it tests the ability to determine the proper readings on the EPB for an abnormal condition based on the indications

and alarms received (white light and annunciators).

The automatic portion of the KA is the breaker opening on an overcurrent

con-rntlnn.

MCS Time: Points: 1.00 Version:0123456789

Answer:DCBDD DCABA Scramble Range:A-D Monday, January 14, 20082:42:23

PM 120

Source: Cognitive Level: Job Position: reviewed: NEW LOWER RO GO QUESTIONS REPORT for 75 RO Questions Source if Banlc Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:23

PM 121

QUESTIONS REPORT for 75 RO Questions 48.063'K4.04 001 Given the following:

Unit 2 is at 100%power.Reactor Trip Breakers A (RT A)and B (RTB)are closed, Reactor Trip Bypass Breakers are open.*125V DC distribution

panel breaker 2B-16, IIAII Reactor Trip switchgear

control power to Bypass breaker and Reactor Trip breaker, has tripped open.Which one of the following statements

correctly describes how a loss of DC to the A Train reactor trip switchgear

would effect the operation of Reactor Trip Breaker A?A.RT A will immediately

trip open.RTA will remai'n closed and will still open from either a manual or automatic signal.C.RT A will remain closed and will not open from either a manual or automatic signal.D.RTA will remain closed and will not open from a manual reactor trip signal;an automatic trip will still open the breaker.Monday, January 14, 2008 2:42:23 PM 122

QUESTIONS REPORT for 75 RO Questions A.Incorrect-This will not cause a direct Rx trip due to SSPS still powered up and the only loss is DC to the STC Rx Trip.B.Correct-RTA would still open from either a manual or automatic signal.125 V DC bus A allows the RT A shunt trip coil to energize on a Rx trip.With no DC available, the Rx trip Brkers will still open on a signal from SSPS A train by de-energizing

the UV coil.With the loss of the DC, no RT brker will open immediately

because SSPS is still energized and the breaker does not have a trip signal.C.Incorrect-It will open on both.D.Incorrect-manual Rx trips cause the STC to be energized and the UV coil to be de-energized.

A Rx trip will still operate on a loss of DC due to the UV coil.according to Table 6 of OPS-521 03C, 125V DC distribution

panel feeds to Rx trip swgr#1.According to the load list'page F-51/52 LA-13 feeds 125V DC A Rx trip swgr control power to Byp brker&Rx trip bker.(2B-16)FSD A-181007 2.2.18 The RPS shall be designed for fail safe operation.

Loss of power to the protection

logic or rod control system shall trip the reactor.The only exception to fail safe criteria shall be containment

spray (H1-3)and shunt trip attachment

for reactor trips.(References

6.1.022, 6.7.031)page 3-10 The Shunt Trip Attachment

coil-shall operate on 125 Vdc and function as a backup for the undervoltage

trip device.From the load list.this is the breaker designation

and the nomenclature

for this panel 125V DC distribution

panel breaker 2B-16, IIA II Reactor Trip switchgear

control power to Bypass breaker and Reactor Trip breaker, Monday, January 14, 20082:42:23

PM 123

QUESTIONS REPORT for 75 RO Questions 063 K4..04 D"C..Electrical

Distribution

Knowledge of DC electrical

system design feature(s)

and/or interlock(s)

which provide for the following:

Trips Question Number: Tier 2 Group 1 22 Importance

Rating: 2.6 Technical Reference:

Electrical

Dist FSD, DC LP;SEQ LP OPS-52103F

reactor protection

lesson plan and FSD A-18100?Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.7 Comments: replaced this question to meet the KA.This is a backward way to meet the KA in that DC power is provided to theRx trip breakers to cause a rx trip via the shunt trip coil.A loss of the DC will not allow DC to provide for a rx trip so one will not occur from the STC but will occur from SSPS via the UV coil.This requiresknowledgeof

the DC bus supply to the RT*breakers and what the loss means to the operator.Scramble Range:A-D FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK LOWER RO GO Version:0123456789

Answer: B CDC CAACDB Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:23 PM 124

QUESTIONS REPORT for 75 RO Questions 49.064 K6.08 001 Given the following:

Hurricane force winds have caused damage in the HVSYD and a dual unit LOSP.*Long term Emergency Diesel Generator operation is anticipated.

It is not known how soon the on-site Diesel supplies can be replenished

The Fuel Oil Storage Tank (FOST)for 1-2A DG is empty.*The Fuel Oil Day Tank for 1-2A DG is full.Which ONE of the following describes the maximum time available for 1-2A DG to cO.ntinue to run at full load?A.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s4 hours c.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Monday, January 14, 20082:42:24

'PM 125

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

The day tank is sized to supply 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at full load, 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months at minimum TS level.B.Correct.

C.Incorrect.

Plausible because it is close to the amount of time and symmetrical

as a distractor

D.Incorrect.

Plausible due to amount of time and also name of tank (day tank)Lesson plan 52101 i Diesel Fuel Oil Storage System Refer to Figure 14.Each diesel generator is connected to a shared fuel oil storage and transfer system, which consists of five storage tanks, two fuel oil transfer pumps per storage tank, a day tank for each diesel generator, and interconnecting

piping and valves.The diesel fuel oil storage system has a total of five 40,000-gallon

storage tanks, two 1000-gallon

day tanks (for the little diesels), three 1325-gallon

day tanks (for the big diesels),*and two redundant capacity fuel oil transfer pumps per storage tank.The storage tanks are designed with sufficient

fuel oil storage capacity to supply the minimum number of diesels required for seven days of operation with ten percent excess for testing using the deliverable

capacity of four of the five storage tanks.The electrical

distribution

system supplies Class 1 E, 120V AC power to each storage tank level transm itter.Two fuel oil transfer pumps are mounted on each storage tank.The pumps are motordriven, vertical, submersible, wet pit-type pumps.The capacity of each pump is in excess of the amount required to simultaneously

supply the diesel generator full load fuel requirements

and fill the associated

day tank.One pump automatically

maintains the required day tank level, and the other is strictly manual.When full, the day tanks provide sufficient

storage for four hours of full load*operation.

Tech Spec required minimum day tank volume ensures'sufficient

fuel oil is available to allow 3hoursof full load operation of the respective

diesel generator.

FSD 3.6.3.2 A-181005 Functional

Requirements

Each diesel engine has an individual

fuel oil day tank.When full, the fuel oil day tanks have sufficient

fuel oil volume to supply their respective

diesel for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at continuous

rated load.However, the Technical Specifications

state that the minimum amount of fuel oil is 900 gallons each in the day tanks of the 1-2A, 1 Band 2B diesel engines and 700 gallons each in the day tanks of the 1 Cand 2C diesel engines.This capacity is sufficient

to ensure that each day tank can supply its corresponding

diesel with at least three hours of fuel at continuous

rated load (References

6.1.006, 6.1.014, 6.3.012, 6.5.014,6.7.009

and 6.7.030).Monday, January 14, 20082:42:24

PM 126

QUESTIONS REPORT for 75 RO Questions 064 KG.OS Emergency Diesel Generators

-Knowledge of the effect of a loss or malfunction

of the following will have on the ED/G system: Fuel oil storage tanks Question Number: Tier 2 Group 1 23 Importance

Rating: 3.2 Technical Reference:

TS 3.8.3.a, f;DG LP OPS-521 021 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10/43.2

Comments: good question that tests length of time a DG can run with the day tank full andother fuel available.

This is the loss of portion of the KA and the affect as it relates to the DGs.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GTO Version: a123456789 Answer: B BCDCCC CDD Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:24

PM 127

QUESTIONS REPORT for 75 RO Questions 50.065 AK3.03 003 Given the following:

Unit 2 is at 100%power.*N2P19HV3825, Instrument

Air to Penetration

Room valve, has closed and cannot be opened.Which ONE of the following will occur with no operator action taken?A.Pressurizer

pressure and level will remain stable.B.Pressurizer

pressure will increase until the PORVs lift.C!'Pressurizer

pressure and level will increase until a reactor trip occurs.D.Pressurizer

level will decrease until letdown isolates and backup heaters turn off, then increase until a reactor trip occurs.HV 3825 supplies air to AB loads.FCV-122 will go full open, letdown will secure sprays, PORVs will go closed and stay closed.A.Incorrect-

Due to the loss of air pre'ssure Przr pressure will be rising and level will be rising due to FCV-122 and loss of letdown.B.Incorrect-

there isnoair to the PORVs C.Correct-due to no air, charging will be at a max rate and letdown will secure.Pressure will also be rising and a Rx trip on high pressure will occur.D.Incorrect-

Level will actually rise.APE 065 Loss of Instrument

Air AK3.03 Knowledgeofthe reasons for the following responses as they apply to the Loss of Instrument

Air: Knowing effects on plant operation of isolating certain equipment from instrument

air Question Number: 52 Tier 1 Group 1 Importance

Rating: 2.9 Technical Reference:

AOP-6.0____----"-P--=-r-=-toPQ-s-fid---Le1e

re

Learning Objective:10 CFR Part 55 Content: 41.7 Comments: changed out the question completely

to meet the KA.Monday, January 14, 20082:42:24

PM 128

Scramble Range:A-D FARLEY FARLEY NO QUESTIONS REPORT for 75 RO Questions 1.00 Version: a123456789 Answer: CDCBDCBACD

Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: MODIFIED LOWER RO GO Points: MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:24 PM 129

QUESTIONS REPORT for 75 RO Questions 51.068 G2.1.20 002 Given the following:*Both Units are operating at 100%power.*Toxic gas has made the Control Room inaccessible.

AOP-28.0, Control Room Inaccessibility, has been implemented.

Which ONE of the following are the minimum and complete actions required lAW AOP-28.0 before leaving the control room?A.Trip the reactor an'd trip the main turbine ONLY.B.Trip the reactor, trip the main turbine, trip both SGFPs, and sound the plant emergency alarm.Trip the reactor, trip the main turbine, verify at least one train of 4160 V ESF buses are energized, and sound the plant emergency alarm.D.Trip the reactor, trip the main turbine, verify at least one train of 4160 V ESF buses are energized and a,ctuate a safety injection.

Monday, January 14, 2008 2:42:24 PM 130

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

These are the Immediate actions of FRP-S.1, not lAW AOP-28.These actions are not complete lAW AOP-28.B.Incorrect.

Both SGFPs are not required to be tripped at this point in the procedure.

It is done locally at step 15.Sounding the plant emergency alarm is the only place this can be done and is in the first four steps..C.Correct.These are the strategies

addressed by the steps in the procedure.

AOP-28.0 rev 11 actions: 1.0 Verify reactor tripped.2.0 Verify the turbine tripped.3.0 Verify at least o,ne train of 4160 V ESF buses energized.

4.0 Perform the following.

4.1 Direct Operation's

personnel to man the Hot Shutdown Panels.4.2 Actuate the plant emergency alarm.4.3 Announce"Main control room evacuation.

Report to your designated

assembly areas." 4.4 Verify control room and C.A.S.evacuated.

4.5 Notify appropriate

support groups to report to the Hot Shutdown??Panels.4.6 Direct Security to station personnel at each co'ntrol room door to prevent entry.D.Incorrect.

These are almost the first 4 steps of E-O, an SI is actuated if one is required.however, if no signal is calling for an SI, it would not be conservative

to actuate an SI, and might be considered

a good idea, but not lAW AOP-28.Checking the SI actuated is not required.Monday, January 14, 2008 2:42:24 PM 131

QUESTIONS REPORT for 75 RO Questions APE068 Control Room Evacuation

G2.1..20 Conduct of Operations:

Ability to execute procedure steps.Question Number: 65 Tier 1 Group 2 Importance

Rating: 4.3 Technical Reference:

AOP-28 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS 52533M02 10 CFR Part 55 Content: 41.1 0 Scramble Range:A-D FARLEY a123456789 CADDDABBDB

Source if Banle Difficulty:

Plant: FARLEY..Previous 2 NRC exams: YES MODIFIED LOWER RO GO Source: Cognitive Level: Job Position: reviewed: Comments: This is a new KA replaced per FJE.This question asks the operator the actions required that an RO should know to evacuate the control room.These should be committed to memory and if not properly executed would cause operational

concerns.This demonstrates

the ability to execute procedural

steps in AOP-28.0.MCS.Time: 1 Points: 1.00 Version: Answer: Monday, January 14, 2008 2:42:24 PM 132

QUESTIONS REPORT for 75 RO Questions 52.069 AA2.01 007 Which ONE of the following conditions

represents

a loss of containment

integrity and would cause entry into Technical Specification

3.6.1, Containment?

A..Mode 3 and one of the Personnel Airlock doors will not close.B.rtI Mode 4 and Integrated

Leak Rate test determines

that leakage is not within limits.C.Mode 5 and it is discovered

that the Phase IBI isolation valve for CCW to the RCPs, will not close.D.Mode 6 and the Equipment Hatch is held in place by 4 bolts.Containment

integrity A is incorrect.

Both doors inop would be a loss of Containment

Integrity, this is just an inop of one of the doors in the Personnel Airlock Plausible becasue one of two series valves makes containment

integrity LCO not met.B is correct.Surveillance

requires ILRT to be within limits for Containment

Integrity to be set.C is incorrect.

because Containment

Integrity is not required in Mode 5, plausible because the valve is part of a containment

penetration

that would affect integrity in modes 1-4.D is incorrect.

4 bolts meets the minimum requirement

for Containment

Closure in Mode 6, but not containment

integrity in the modes that containment

integrity is required.Monday, January 14, 20082:42:24

PM 133

QUESTIONS REPORT'for 75 RO Questions APE 069 AA2..01 Loss of Containment

Integrity-Ability to determine and

the following as they apply to the Loss of Containment

Integrity:

Loss of containment

integrity Question Number: 59 Tier 1 Group 2 Importance

Rating: 3.7 Technical Reference:

TS section 3.6 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS-52102A-1

10 CFR Part 55 Content: 43.2/41.10

Comments: meets the KA in that it tests the ability to determine IF Ctmt integrity is met in different modes lAW Tech Specs.Items Not Scrambled HARRIS BANK LOWER RO GTO Mode applicability

(1-4)&one hour or less tech specs (one or more air locks with one door inoperable)

are RO Knowledge and this question meets KIA for ROs.MCS Time: 1 Points: 1.00 Version:0123456789

Answer:BBBBBBBBBB

Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:24 PM 134

QUESTIONS REPORT for 75 RO Questio'ns

53.072 G2.1.2 001 Which ONE of the following Area Radiation Monitors requires entry to a Technical Specification

Action Statement if it is declared INOPERABLE?

A.R-1 A, Control Room Area Radiation (Unit 1)B.R-2, Containment

Area Radiation C.R-4, Charging Pump Area RadiationR-27A, Containment

Area Radiation (High Range)A.Incorrect.

Plausible because this is an important radiation monitor indicating

the habitability

of the Control Room, but it is not in TS.B.'Incorrect.

Plausible because this is an important radiation monitor indicating

the abnormally

high radiation level in containment.

This is used in the emergency procedures

for diagnosis of a LOCA, but it is not in TS.C.Incorrect.

Plausible because this is an important radiation monitor indicating

radiation levels in the Charging Room area.This is used in the emergency procedures

for diagnosis of aLOCAoutside

Containment, but it is not in TS.D is correct.This radiation monitor is in TS in 3.3.3 table, and monitored on STP-1.0 every shift to ensure operable.19.Containment

Area Radiation (High Range)072 G2.1.2 Area Radiation Monitor Conduct of ,Operations:

Knowledge of operator responsibilities

during all modes of plant operation.

Question Number: 38 Tier 2 Group 2 Importance

Rating: 3.0 Technical Reference:

FNP-1-ARP-1.6

FH1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10/43.2

Comments: This meetsttieKAmt11at1f1Sfne

operator responsfolltty

to know what area rad monitors are entry conditions

to TSs.Monday, January 14, 2008 2:42:24 PM 135'

MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: NEW LOWER RO GTO QUESTIONS REPORT for 75 RO Questions 1.00 Version:0123456789

Answer: DB C CDADB AA Source if Bank: Difficulty:

Plant: Previous 2 NRC exams: Scramble Range:A-D FARLEY NO Monday, January 14, 20082:42:24

PM 136

QUESTIONS REPORT for 75 RO Questions 54.073 A4.02 001 Given the following:

R-19, SGBD SAMPLE, radiation monitor is in alarm and stable above the-alarm setpoint.*The Shift Chemist requests to sample the Steam Generators.

Which ONE of the following correctly describes the actions that will allow the shift chemist to obtain a sample of the SGs lAW SOP-45.0, Radiation Monitoring

System?A.Manually open the sample valves one at a time.B.Pull the INSTRUMENT

power fuses for R-19 to allow opening the sample valves.C.Pull the DC power fuses to each sample valve solenoid to fail the valve open.Place R-19 Operations

Selector Switch to the RESET position, then open the sample valves.A.Incorrect.

These valves can not be manually opened.The SGBD sample valves do not have manual jacks as they have solenoids powered from DC power and fail closed.B.Incorrect.

This is the procedure directed action-for

a monitor in saturation, but not to clear a valid alarm.c.Incorrect.

these solenoid valves are DC powered and fail closed.Even though the designator

is HV3328 and there is no manual operator on the valve.D.Correct.SOP-45.0, Section 4.4 directs this.QIPI5HV3328

IA SteamGeneratorBlowdown

sample valve QIPI5HV3329

IB Steam Generator Blowdown sample valve QIPI5HV3330

Ie SteamGeneratorBlowdown

sample valve Monday, January 14, 20082:42:24

PM 137

QUESTIONS REPORT for 75 RO Questions 073 Process*Radiation

Monitoring

System A4.02 Ability tomanuallyoperate

and/or monitor in the control room: Radiation monitoring

system

panel Question Number: Tier 2 Group 1 24 Importance

Rating: 3.7 Technical Reference:

OPS 521060;SOP-45.0 section 4.4 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.11 Comments: MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GO Version: a123456789 Answer: DABBDAACAA Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:24 PM 138

QUESTIONS REPORT for 75 RO Questions 55.'076 K3.0l 001 Given the following:

Unit 2 is at 39%power.*IA I Train is On-Service.

1-2A DG is running at full load lAW STP-80.1, Diesel Generator 1-2A Operability

Test, and tied to Unit 2.*A Loss of IA I Train Service Water is occurring due to SW Pump failure.*The crew is performing

actions of AOP-1 0.0, Loss of Train A or B Service Water.*IB I Train SW header pressure is 72 psig.Which ONE of the following describes a potential effect on the unit and the actions required in accordance

with AOP-1 O.O?A.*RCP Motor Air Coolers will lose cooling water flow;*Trip the reactor and any RCP if its motor stator temperature

exceeds the temperature

limit.*RCP bearing temperatures

will rise;*Trip the reactor and any RCP if its bearing temperature

exceeds the temperature

limit.C.*1-2A DG will lose cooling water flow;*Isolate Service Water to the Turbine Building and trip the reactor.D.*Main Generator bearing and Hydrogen temperatures

will rise;*Isolate Service Water to the Turbine'Building and trip the reactor.Monday, January 14, 2008 2:42:25 PM 139

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

plausible because Service Water does supply cooling to the motor air coolers, but Train B does, not Train A.B.Correct.CCW temperature

will rise, and as it does, RCP bearing temperatures

will rise.This action is done in both AOP-10 and AOP-9 which AOP-10 sends the user to to accomplish

in conjunction

with AOP-1 0 at step 11.C.Incorrect.

plausible because Service Water does supply cooling to this DG and is in fact aligned to both units SW.Therefore a loss of Unit 2 does not cause a loss of cooling water to the DG.The candidate may believe that for a unit 2 STP, SW would be secured from unit 1 or flow is affected since STP-80.1 has the following note.If service water from either unit is secured, a partial surveillance

may be performed for the non affected unit.Full surveillance

credit for this STP may be taken once service water is returned to service and verified operable by rerunning this STP.The actions would be performed if the DG was required.It is NOT required.The RNO of step 4.2 says to isolate SW to the TB for the affected train and trip the reactor if BOTH trains were isolated.In this case the DG would be secured or left running if SW flow was sufficient

from unit1.D.Incorrect.

First part is correct.second part incorrect since at step 6, SW pressure is checked to be>60 psig.For this event, the SW pressure is 72 psig.(If it was less than 60 psig, then actions would be performed to isolate Service water to the Turbine Bldg.If this were done, then a reactor trip would be required.)

Since it is not required to go to the RNO column, then it is not correct to do these actions.076 K3.01 Service Water System Knowledge of the effect that a loss or malfunction

of the SWS will have on the following:

Closed cooling water Question Number: 26 Tier 2 Group 1 Importance

Rating: 3.4 Technical Reference:

FNP-2-AOP-10.0

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: This meets the KA in that there is a loss of SW to a train and requires knowledge of hOV\l_th----Cis=----_

affects the equipment that receives CCW and is cooled by the SWS.This also tests the actions required for this event.MCS Time: Points: 1.00 Version:0123456789

Answer: B ADABCB CDA Scramble Range:A-D Monday, January 14, 2008 2:42:25 PM 140

Source: Cognitive Level: Job Position: reviewed: NEW HIGHER RO GO QUESTI"ONS

REPORT for 75 RD Questions Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:25 PM 141

QUESTIONS REPORT for 75 RO Questions 56.078 G2.1.32 001 Which ONE of the following describes the reason why one Air Compressor

should be aligned with the air compressor

panel key switch in'LOCAL OR in the MCB position'with the AUTOMATIC OPERATION LED (green light)LIT?A-I To prevent a complete loss of Instrument

Air pressure due to a single failure of the sequencer panel pressure transducer.

B.To allow one air compressor

to be started from the MCB and operate from the selected sequencer for any complete loss of Instrument

Air pressure situation.

C.To allow one air compressor

to start after an LOSP and load and unload based on its Internal Mode" pressure setting.D.To prevent all three air compressors

from running at the same time to prevent a complete loss of Instrument

Air pressure in the event that Service Water islostto the Turbine Building.Monday, January 14, 2008 2:42:25 PM 142

QUESTIONS REPORT for 75 RO Questions A.Correct.see P&L below 3.19 Failure of the sequencer panel pressure transducer

could unload all air compressors

selected (integrated)

and result in complete loss of air pressure.To prevent loss of air from a single failure, at least one air compressor

should be aligned with the air compressor

panel key switch in LOCAL OR in MCB with the AUTOMATIC OPERATION LED lit (green).B is incorrect.

The air compressor

in LOCAL will start from the MCB if OFF is selected first, then AUTO, but will not operate on the sequencer;

but will start by its internal mode pressure , switch.3.8 Any air c9mpressor

with the panel key switch in the MCB position will (1)stop if the MCB handswitch

is selected to OFF and (2)start and load if the MCB handswitch

is taken to the START/RUN position and returned to AUTO position, based on the Internal Mode pressure settings on the air compressor.

The AUTOMATIC OPERATION LED on the air compressor

panel will be lit (green)when the MCB handswitch

has been taken to the START/RUN position and returned to AUTO.The lit LED indicates the air compressor

will load and unload based on its Internal Mode pressure settings.IF the MCB handswitch

is taken from START/RUN to OFF, THEN the air compressor

panel AUTOMATIC OPERATION LED will not be lit AND the LED will remain off if the MCB handswitch

is taken from OFF back to AUTO without going to START/RUN AND the air compressor

will not load and unload based on its Internal Mode pressure settings.C is incorrect.

There is a P&L applicable

to1C air compressor

and the compressor

will cycle on the sequencer after the load shed and LOSP is complete.This is not necessarily

true for any air compressor

operation after an LOSP.The Air compressor

will also sequence b,ack on***and run on the sequencer, not the Internal Mode pressure setting.3.11 During an LOSP or SI/LOSP the emergency section of Load Center1A will automatically

align to Load Center 1 D, and1C air compressor

supply breaker EA-15 will automatically

close.If the air compressor

was operating prior to the LOSP, the compressor

will resume operation after the LOSP if (1)the MCB handswitch

is in AUTO (returned from START/RUN and not been take.n to OFF)and the1C panel key switch is in MCBposition

OR (2)the1C panel key switch is in the SEQ position and1C is selected (integrated)

to the sequencer OR (3)the1C panel key switch is in the LOCAL position.D.incorrect.

If air pressure drops with the switches in the above configuration, then all 3 air compressors

will be running.Thisdoesnot prevent 3 a/cs from running.The normal system line-up is three air compressors

in AUTO on the MCB, two air compressor

selected (integrated)

on the sequencer, and one air compressor

de-selected (isolated)

from the sequencer.

This will allow the sequencer to control two air compressors

based on header pressure and allow the de-selected (isolated)

air compressor

to auto start based on its receiver pressure.Monday, January 14, 2008 2:42:25 PM 143

QUESTIONS REPORT for 75 RO Questions 078 G2.1.32 Instrument

Air System Conduct of Operations:

Ability to explain and apply all system limits and precautions.

Question Number: Tier 2 Group 1 27 Importance

Rating: 3.4 Technical Reference:

FNP 1-S0P-31.0

Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: This meets the KA in that a precaution

is required to be known and applied for a failure of a pressure switch for the instrument

air compressors

..MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW LOWER RO GO Version:0123456789

Answer: A CADAB AACA Scramble Range:A-D Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO---------------------'---------------------------

Monday, January 14, 20082:42:25

PM 144

QUESTIONS REPORT for 75 RO Questions 57.103 K4.06 001 Given the following conditions:*A LOCA has occurred.*Containment

pressure is currently 19 psig and rising.*All automatic actions have occurred as required.*No manual actions have been taken.Which ONE of the following describes the ESF actuations

that have taken place?A.Safety Injection ONLY.B.Safety Injection and Containment

Isolation Phase A ONLY.Safety Injection, Containment

Isolation Phase A, and Main Steam Line Isolation ONLY.D.Safety Injection, Containment

Isolation Phase A, Main Steam Line Isolation,"and Containment

Isolation Phase B.A is incorrect.

because if SI actuates, Phase A will also be actuated.B is incorrect.

because Phase A is actuated, but MSLI is also actuated.C is correct.Containment

Isolation Phase A, and Main Steam Line Isolation ONLY, due to containment

pressure.D is incorrect.

because containment

pressure is not high enough for phase B.Monday, January 14, 20082:42:25

PM 145

QUESTIONS REPORT for 75 RO Questions 103 K4.06 Containment

System Knowledge of containment

system design feature(s)

and/or interlock(s)

which provide for the following:Containmentisolation

system Question Number: 28 Tier 2 Group 1 Importance

Rating: 3.1 Technical Reference:

E-O Attachment

3 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

1 0 CFR Part 55 Content: Scramble Range:A-D WOLFCREEK Version:0123456789

Answer: CDCDDABCBB Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO BANK LOWER RO GTO Comments: This tests the KA appropriately

in that these are design features that provide for ctmt isolation at an RO level of knowledge.

MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:25 PM 146

QUESTIONS REPORT for 75 RO Questions 58.E02 EAl.3 001 Given the following:

Unit 1 was operating at 10%Reactor power when a Loss of Off-Site Power caused'a loss of ALL RCP*s and a spurious safety injection.

Due to the LOSP, no RCP can be started.*The crew has just adjusted the atmospheric

relief valves.*The crew is performing

actions of ESP-1.1, SI Termination, and are at the step to determine if adequate natural circulation

exists.Which ONE of the following correctly lists indications

that are consistent

with adequate natural circulation

lAW ESP-1.1?1-RCS hot leg temperature

---stable or decreasing2-RCS hot leg temperature

---increasing3-SG pressure---stable or decreasing

4'-SG pressure---increasing5-RCS hot leg temperature

---at saturation

for SG pressure6-RCS cold leg temperature

---at saturation

for SG pressure A.2,3, and 5 B.2,4, and 6 C.1,4, and 5 D!'1, 3, and 6 Monday, January 14, 2008'2:42:25 PM 147

Scramble Range:A-D QUESTIONS REPORT for 75 RO Questions A.incorrect.

RCS HL temps would not be increasing, and not RCS HL at SG saturation

temperature.

B.incorrect.

RCS HL temps would not be increasing, SG pressure would not be increasing

C.incorrect.

SG pressure would not be increasing

D.Correct.1, 3, and6-ESP-1.1 step 21.4 RNO lists NC flow requirements:

Verify adequate natural circulation.

a)Check SG pressure stable or falling.#3 b)Check SUB COOLED MARGIN MO-NITOR indication

greater than 16°F subcooled in CETC mode.c)Check RCS hot leg temperatures

stable or falling.#1 d)Check core exit TICs stable or falling.e)Check RCS cold leg temperatures

at saturation

temperature

for SG pressure.#6 W/E02 81 Termination

-EA1.3 Ability to operate and I or monitor the following as they apply to the (81 Termination):

Desired operating results during abnormal and emergency situations.

Question Number: 60 Tier 1 Group 2 Importance

Rating: 3.8 Technical Reference:

ESP-1.1 steps'2-18 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: This question tests the ability to monitor for desired operating parameters

lAW ESP-1.1, SI termination, during natural circ flow conditions.012345

Answer: DAAC CADCBB Monday, January 14, 2008 2:42:25 PM 148

Source: Cognitive Level:.Job Position: reviewed: MODIFIED LOWER RO GO QUESTIONS REPORT for 75 RO Questions Source if Banle FARLEY Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:25

PM 149

QUESTIONS REPORT for 75 RO Questions 59.E04 EKl.3 002 Given the following:*A reactor trip and an SI have occurred.*Containment

pressure is reading 2 psig.*RCSpressure

is reading 1755 psig.*All systems have operated as required." At the step in EEP-1.0, Loss of Reactor or Secondary Coolant, the following indications

are observed by the Unit Operator:*The following BOP annunciator

is in alarm:-NE2,1B RHR PUMP RM SUMP LVL HI-HI OR TRBL*1A and1B RHR pump discharge pressures are reading 750 psig.*MK4, LIQ OR GAS PROC PNL ALARM, has just come into alarm.Which ONE of the following describes 1)the operator actions;AND 2)the operational

implications

of those actions performed lAW ECP-1.2, LOCA Outside Containment, in an attempt to mitigate this leak?Art 1)Isolate the discharge to ONE train of RHR and check RCS pressure rising;2)Loss of one train of LHSI for injection and recirculation.

B.1)Isolate the RWST suction to ONE train of RHR and check RWST level stable;2)Loss of one train of LHSI for injection 9NLY.'C.1)Isolate the discharge to BOTH trains of RHR and check RCS pressure rising;2)Loss of,BOTH trains of LHSI for inje'ction

and recirculation.

D.1)Isolate the RWST suction to BOTH trains of RHR and check RWST level stable;2)Loss of BOTH trains of LHSI for injection ONLY.A.Correct.At FNP, the most credible source of an ISLOCA is from the RCS to the LHSI pump suction piping which is a low pressure system.The high level action steps of ECP-1.2 are to verify proper valve alignment, attempt to isolate the break, check if the break is isolated.The first system to be isolated is the RHR system which in this case would cause a loss of one train of LHSI due to the isolation of RHR valves.MOVs 8888A and 8887 A are closed and the leak checked, then the other train.Only one train is checked at a time and isolated so BOTH trains are not affected.At FNP, the most credible mechanism for initiation

of RHR suction ISLOCA during power operation is the catastrophic

rupture of the closed MOVs isolating the RHR pump suction from the RCS.Following the RCS to RHR pressure boundary failure, the RHR system will be able to withstand RCS pressure if the hoop stress imposed on the RHR system by exposure to RCS pressures is below the yield stress.However, the RHR pump sealsinboth trains are expected to rupture.This rupture of pump seals'is assumed to result in failure of both RHR pumps Monday, January 14, 2008'2:42:25

PM.150'

QUESTIONS REPORT for 75 RO Questions (i.e., motors short due to water spray).B.

correct location, incorrect strategy.incorrect operational

implication.

The RHR system is not secured to both trains at the same time.ECP-1.1 isolates one train first and thenrestoresand

isolates the other train if the problem is not corrected.

ECP-1.2 never isolates the RWST to any component.

However, ECP-1.1 does.The loss of the train would be for injection and recirc for one train.C.Incorrect.

Incorrect location, incorrect strategy.incorrect operational

implication.

D.incorrect.

incorrect location.incorrect strategy.incorrect operational

implication.

Location-Components

between the RCS and the low pressure LHSI Pump hot leg injection piping include three check valves.A LOCA through the upstream hot leg injection piping is less likely than through the cold leg piping due to the addition of an additional

in-series check valve and because the upstream isolation valves are normally closed.Also the Background

states that the piping is able to withstand RCS pressure if the hoop stress imposed on the RHR system by exposure to RCS pressures is below the yield stress.The RWST will not be lost due to the isolation of the sy.stem, it will be saved due to this action.ECP-1.2 never isolates the RWST to any component.

However, ECP-1.1 does.Further background-

Purpose: To ensure that normally closed valves are closed Basis: This step instructs the operator, to verify that all normally closed valves in low pressure lines and other plant specific lines that penetrate containment

are closed.The valving connecting

the RHR System to the RCSis of particular

interest in this step since the RHR System is a low pressure system (600 psig)connected to the highpressurereactor

coolant system (2500 psig).Therefore, a rupture or break outside containment

is most probable to occur in the low pressure RHR System piping.ERG StepText: Check If Break Is Isolated Purpose: To determine if the LOCA outside containment

has been isolated from previous actions Basis: This step instructs the operator to check RCS pressure to determine if the break has been isolated by previous actions.If the break is isolated in Step 2, a significant

RCS pressure increase.will occur due to the SI flow filling up the RCS with break flow stopped.The operator transfers to E-1, LOSS OF REACTOR OR SECONDARY COOLANT, if the break has been isolated, for further recovery actions.If the break has not been isolated, the operator is sent to ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, for further recovery actions since there will be no inventory in the sump.ERP StepText: Identify source of leak.ERG StepText: Try To Identify And Isolate Break Purpose: To attempt to identify and isolate a LOCA outside containment

Basis: This step instructs the operator to sequentially

close and open all normally opened valves in paths that penetrate containment

to identify and isolate the break outside containment.

Again, as in Step 1, the valving connecting

the low pressure

high pressure (2500 psig)RCS is of primary interest, since the probability

of a break occurring outside containment

is most probable to occur in the low pressure RHR System piping.Knowledge:

The potential exists for RWST inventory to be lost to the auxiliary building for a LOCA that occurs outside containment

in the RHR system piping.The RWST could be drained to the auxiliary building if the RCS pressure is reduced to below the static head pressure in the Monday, January 14, 2008 2:42:26 PM 151

QUESTIONS REPORT for 75 RO Questions RWST.If this condition occurs, actions should be taken to isolate this potential leakage path and loss of inventory from the RWST.W/E04 LOCA Outside Containment

EK1.3 Knowledge of the operational

implications

of the following concepts as they apply to the (LOCA Outside Containment):

Annunciators

and conditions

indicating

signals, and remedial actions associated

with the (LOCA Outside Containment).

Question Number: Tier 1 Group 1 53 Importance

Rating: 3.5 Technical Reference:

ECP-1.2,"FNP--O-ECB-1.2

specific background

document for ECP-1.2, lesson plan for ECP-1.2, OPS-52532E, WOG background

document Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: This question gives the indications

of an ISLOCA and then theoperationalimplications

and remedial actions or high level actions.The procedure would have the operator isolate one train at a time so a complete loss of injection is not done.The implications

of this action if the system is isolated lAW ECP-1.2 is to lose one train of LHSI.The RWST is a credible distracter

in that ECP-1.1 which is where this procedure could send the operator to does isolate and turn off pumps due to low RWST level.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 NEW HIGHER RO GTO Version:0123456789

Answer:ACDACBBBBA

Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:26

PM 152

QUESTIONS REPORT for 75 RO Questions 60.E05 02.1.27003

A Reactor Trip and Safety Injection have occurred on Unit 1: The crew has entered FRP-H.1, Response to Loss of Secondary Heat Sink, from EEP-1, Loss of Primary or Secondary Coolant, with the following conditions:

RCS Pressure is 175 psig and decreasing.

Intact SG pressures are 475 psig and trending down.Which ONE of the following describes the status of the Steam Generators

and the associated

procedural

requirement

for the conditions

given above?The Steam Generators

are-----------

A.*Available to provide secondary heat sink.*Remain in FRP-H.1.*NOT Available to provide secondary heat sink.*Return to EEP-1.C.*Available to provide secondary heat sink.*Return to EEP-1.D.*NOT Available to provide secondary heat sink.*Remain in FRP-H.1.A-Incorrect.

Secondary heat sink is not required if SGs are at a higher pressure than the RCS.They act as a heat source.Plausible because the conditions

for FRP-H.1 entry are met except for the first step of FRP-H.1.RNO sends to procedure and step in effect.B-Correct.If SGs are NOT required for heat sink, the crew will return to EEP-1.C-tncorrect.

SGs are NOT required, because RCS pressure is below SG pressure.D-Incorrect.

LBLOCA, RCS less than SG pressure, return to EEP-1.Monday, January 14, 20082:42:26

PM 153*

QUESTIONS REPORT for 75 RO Questions W/E05 loss Secondary Heat Sink G2.1..27 Conduct of Operations:

Knowledge of system purpose and or function.Question Number: 54 Tier 1 Group 1 Importance

Rating: 2.8 Technical Reference:'

FRP-H.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: '41.10 Scramble Range:A-D WTSI BANK HIGHER RO GO Comments: meets the KA in that the operator has to know the purpose oftheSGs and the function/role

they play in removing/adding

heat on a LOCA..MCS Time: 1 Points: 1.00 Version:0123456789

Answer:BB DADB ACB B Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:26

PM 154

QUESTIONS REPORT for 75 RO Questions 61.E06 EAl.2 064 Given the following:*A LOCA has occurred.*Due to ECCS failures, the crew is performing

FRP-C.2, Response to Degraded Core Cooling.*The crew is depressurizing

ALL Steam Generators

to 100 psig.*The STA reports a RED condition on the Integrity CSF Status Tree.*The Shift Supervisor

continues in FRP-C.2.Which ONE of the following describes the reason that the SS remains in FRP-C.2?A.Actions cannot be taken for a RED condition on the Integrity CSF because there is inadequate

ECCS equipment available to mitigate the degraded core cooling condition.

B.The RED condition on the Integrity CSF is not valid because of the dumping of steam to the condenser at a maximum rate lAW FRP-C.2.C.FRP-C.2 has a higher priority than any lower level CSF and no other procedural

actions are allowed to be implemented

until a transition

is directed lAW FRP-C.2.The RED condition on the Integrity CSF is expected and is based upon the accumulators

injecting.

Monday, January 14, 2008 2:42:26 PM 155

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

Actions taken would actually be to reduce ECCS flow, so unavailability

of SI for a PTS issue would not be a priority.-B.Incorrect.

The red condition is valid.The dumps are NOT opened to dump steam at a maximum rate, the limit of 60°F/hour is the limit in C.2.In FRP-C.1, Steam is dumped at a maximum rate.C.Incorrect.

Core Cooling is high priority, but FRP-C.-2 is entered on an orange condition, so a red condition on another CSF tree would take priority.However, in the specific case of the integrity CSF, the dumping of the accumulators

is expected and subsequent

entry into this FRP would only cause core temperatures

to rise and C.1 entry could be required.D.Correct.Due to the dumping of the steam above, accumulator

injection isthegoal and the dumping of steam is done at a rate of 60°FI hr.FRP-C.2 Caution prior to step 12 just before depressurizing

all intact SGs to 100 psig.CAUTION: Performance

of step 12 will cause accumulator

injection which may result in a red path on the INTEGRITY st,atus tree.This procedure should be completed before transition

to FNP-1-FRP-P.1

, RESPONSE TO-IMMINENT

PRESSURIZED

THERMAL SHOCK CONDITIONS.

background

documentation

ERP Step Text: Performance

of step 12 will cause accumulator

injection which may result in a red path on the INTEGRITY status tree.This procedure should be completed before transition

to FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED

THERMAL SHOCK CONDITIONS.

ERG StepText: The following step will cause accumulator

injection which may cause a red path condition in F-O.4, INTEGRITY Status Tree.This guideline should be completed before transition

to FRP.1, RESPONSE TO IMMINENT PRESSURIZED

THERMAL SHOCK.Purpose: To alert the operator to complete entire guideline FR-C.2 even if a red path occurs in the Integrity Status Tree, F-0.4.Basis: Once the RCS is cooled/depressurized

in step 10 to the point at which the accumulators

inject, the RCS cold leg temperature

could be reduced such that a transition

to FR-P.1, Response to Imminent Pressurized

Thermal Shock Condition, is required via the red path of Status Tree F-0.4.The operator would stop the cooldown after entering FR-P.1.While the operator is allowing the thermal shock to soak out, the core will continue to boil away-the injected accumulator

water and begin to uncover once again.Eventually, core exit temperatures

and/or RVLIS level values could existwhich

would require the operator to transfer to FR-C.1, Response to Inadequate

Core Cooling, via one of the red paths on Status Tree F-0.2.Thus, by going from FR-C.2 to FR-P.1 and stopping the cooldown and soaking, a degraded core cooling condition could be allowed to deteriorate-to an inadequate

core cooling condition.

Therefore, this caution will require the operator to complete guideline FRC.

0.4.'Monday, January 14,20082:42:26

PM 156

QUESTIONS REPORT for 75 RO Questions W/E06 Degraded Core Cooling EA1.2 Knowledge of the reasons for the following responses as they apply to the (Degraded Core Cooling)Operating behavior characteristics

of the facility.Question Number: Tier 1 Group 2 Importance

Rating: Technical Reference:

61 3.5 1-FRP-C.2 and FNP-O-FRB-C.2

background

document Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.7 Comments: This question tests the operating behavior characteristic

in that when this condition is entered and the SGs are depressurized, the resulting accumulator

injection is expected to cause FRP-P.1 conditions

due to the rapid cooldown.This is a high level action of FRP-C.2 that an RO is expected to know.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 MODIFIED LOWER RO GTO Version:0123456789

Answer:DD DCADBDDA Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:26

PM 157

QUESTIONS REPORT for 75 RO Questions 62.E08 EA2.1 001 Given the following:

While operating at 100%power, Unit 1 experienced

a steam break accident inside CTMT.*After the transition

from EEP-O, Reactor Trip or Safety Injection, RCS cold.leg temperatures

had dropped to 240°F in 20 minutes.*Reactor power is currently<1%and 80th Intermediate

Range SUR meters are reading-.03 dpm.*AFW flow is reading 300 gpm.*SG narrow range water levels are reading: 1A SG=49%18 SG=29%1C SG=30%*Sub Cooled.Margin

Monitor is reading 34°F in the CETC mode.*Containment

pressure is currently 28 psig and slowly decreasing.

CS flow is reading 950 gpm with one CS pump running.*RCS pressure is currently 1500 psig with all ECCS equipment running.Which ONE of the following is the highest level Functional

Restoration

Procedure (FRP)required to be entered under these conditions?

A.FRP-Z..1, Response to High Containment

Pressure.8.FRP-S.1, Response to Nuclear Power Generation-

ATWT.FRP-P.1, Response to Imminent Pressurized

Thermal Shock Conditions.

D.FRP-H.1, Response to Loss of Secondary Heat Sink.reference provided is the RCS pressure-temperature

graph CSF-O.4 rev 17 so a determination

can be made to which area of the graph applies.All distracters

are plausible in that evaluation

has to be made to determine which FRP is valid and what condition it is in, orange green yellow.Then a determination

of which order these are referenced

in.Answer A is incorrect:

Z.1 is an Orange path but it is lower than P.1 orange path.Answer 8 is incorrect:

FRP S.1 is a green path based on<5%power, IR SUR more negative than-.02.Answer C is correct: P.1 entered on an orange path.Answer.D is incorrect.

H.1 entry conditions

are<395 gpm or all sg nr levels<310/0.if one sgwl NR is>31%no entry is required*.

Monday, January 14, 2008 2:42:26 PM 158

QUESTIONS REPORT for 75 RO Questions E08 EA2.1 Pressurized

Thermal Shock-EPE Ability to operate and I or monitor the following as they'apply to the (Pressurized

Thermal Shock)Facility conditions

and selection of appropriate

procedures

during abnormal and emergency operations.

Question Number: Tier 1 Group 2 Importance

Rating: Technical Reference:

Learning Objective:

10 CFR Part 55 Content: 62 3.4 none OPS 52533K-8 41.10 Comments: matches KA in that the operator has to monitor the correct parameters

and then based on those parameters

select the appropriate

procedure to follow during the emergency event and it deals with a PTS event.All distracters

are plausible in that evaluation

has to be made to determine which FRP is valid and what condition it is in, orange green yellow.Then a determination

of which order these are referenced

in.This does not overlap with G2.4.22 MCS Time: 1 Points: 1.00 Scramble Range:A-D FARLEY Source: Cognitive Level: Job Position: reviewed: MODIFIED HIGHER

RO GO Version: a123456789 Answer: CAB eBB B BCD Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:26

PM 159

QUESTIONS REPORT for 75 RO Questions 63.EIO EKI.I 005 Given the following:

An RCS cooldown is in progress lAW ESP-0.3, Natura'i Circulation

Cooldown with Allowance for Reactor Vessel Head Steam Voiding (WITH RVLIS).*The plant is being depressurized

using auxiliary spray.*Charging and Letdown flows are matched.*As RCS pressure drops through 1300 psig, a rapid rise in pressurizer

level is observed.*Pressurizer

level has increased to 66%.*Reactor vessel level indication

has dropped below the minimum required UPPER PLENUM level of 44%.Which ONE of the following correctly describes the required response lAW ESP-0.3?A.Increase auxiliary spray flow and verify Both CRDM cooling fans running.B.Reduce charging flow, increase letdown flow and stop the cooidown in progress.C.Increase the RCS cooldown rate while maintaining

charging and letdown flows matched.Reduce the auxiliary spray flow and energize additional

pressurizer

heaters.A.Incorrect-Plausible because verifying both CRDM cooling fans running would cool the head and reduce void formation&is required later in procedure.

Spray flow should be reduced, not increased, to establish subcooling.

B.Incorrect-Plausible because ESP-0.3 says to control or reduce charging and increase letdown or continue the cooldown if PRZR level is>90%..Level is only 660/0.the method at this point in the procedure.

C.Incorrect-Plausible because cooling down the RCS will cool the head, but with a time delay.With the head still hot, this will cause the void to increase and PRZR to go solid.D.Correct-This is the correct response lAW ESP-0.3 when Reactor vessel level indication

drops to less than 44%upper plenum raise RCS pressure and this would be done by controlling

pressure with sprays and heaters.Monday, January 14, 2008 2:42:26 PM 160

QUESTIONS REPORT for 75 RO Questions W/E10 EK1.1 Natural Circulation

with Steam Void in Vessel with/withoutKnowledge of the operational

implications

of the following concepts as they apply to the (Natural Circulation

with Steam Void in Vessel with/without

RVLIS)Components, capacity, and function of ef)lergency

systems.Question Number: 63 Tier 1 Group 2 Importance

Rating: 3.3 Technical Reference:

ESP-0.3 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS 52531 C06 1,0 CFR Part 55 Content: 41.1 0 Scramble Range:A-D FARLEY BANK HIGHER RO GO Comments: matches KA in that it tests the knowledge of ESP-0.3 and the implications

of void formation and what to do about it using the components

that the operator os required to control during this evolution.'

.MCS Time: Points: 1.00 Version: a123456789 Answer: DAAAACDBBD Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:26 PM 161

QUESTIONS REPORT for 75 RO Questions 64.Ell EK2.2 004 The following conditions

exist on Unit1:*1B RHR pump is tagged out and the oil is drained from the motor.*The Control room team is responding

to a LOCA.*The reactor was tripped and an SI manually actuated.*RCS pressure is 1000 psig..*1A RHR pump has tripped.The control room team has transitioned

to ECP-1.1, Loss Of Emergency Coolant Recirculation.

Make up has been established

to the RWST.Which one of the following describes the correct actions to take in ECP-1.1 undertheseconditions?*Initiate an RCS cooldown to Cold Shutdown at less than 100°F/hr,*establish only one charging pump running and*reduce RCS pressure to reduce break flow.B.*Initiate an RCS cooldown to Cold Shutdown at the maximum rate possible,*establish only one charging pump running and*reduce RCS pressure to dump the accumulators.

C.*Initiate an RCS cooldown to Cold Shutdown at less than 100°F/hr,*establish two charging pumps running and*reduce RCS pressure to dump the accumulators.

D.*Initiate an RCS cooldown to Cold Shutdown at the maximum rate possible,*establish two charging pumps running and.*reduce RCS pressure to*reduce break flow.Monday, January 14, 2008 2:42:26 PM 162

QUESTIONS REPORT for 75 RO Questions ECP-l.l, LOSS OF EMERGENCY COOLANT RECIRCULATION

OPS-52532D

A.Correct-Start makeup to the RWST, initiate an RCS cooldown, minimize ECCS flow and reduce RCS pressure.The following criteria are the high level actions needed to be successful

in ECP-1.1 Makeup to the RWST is necessary Inventory in the RWST is a concern for recovery from a loss of ECR capability.

Makeup is added to the RWST to extend the time the SI pumps and containment

spray pumps (if operating)

can take suction from the RWST and providecorecooling

to the RCS.Begin Cool Down to Cold Shutdown The purpose is to begin a controlled

RCS cool down to cold shutdown temperature

using a'preferred

oralternatemethod

with a specified maximum cool down rate.Shutdown margin should be monitored during RCS cool down using Curve 61 and/or 61A.The objective is to reduce the overall temperature

of the RCS coolant and metal to reduce the need for supporting

plant systems and equipment required for heat removal.The maximum cool down rate of 1 OO°F/hr will preclude violation of the integrity status tree, thermal shock limits.Stop SI Pumps To reduce flow into the RCS, the low-head injection pumps and all but one high-head pump are stopped.Satisfaction

of conditions

for SI termination

implies that control can be maintained

by the operator without all of the ECCS pumps running.In this step, all but one high-head pump are stopped and placed in standby for future use.Reduce RCS Pressure to Reduce Subcooling

This step is performed to decrease RCS pressure to the lowest pressure possible without losing adequate subcooling.

The RCS pressure reduction is done to decrease RCS break flow.The RCS should be depressurized

until RCS subcooling

indicates between 16°F (45°F)and 26°F (55°F)on the Subcooled Margin Monitor in CETC mode.A second criterion for stopping the pressure reduction is PRZR level greater than 730/0 (50%).B.Incorrect-See above, first and third part incorrect, second part correct C.Incorrect-See above, first part correct, second and third part incorrect D.Incorrect-See above, first and second part incorrect, third part correct Monday, January 14, 2008 2:42:27 PM 163

QUESTIONS REPORT for 75 RO Questions W/E11 Loss of Emergency Coolant Recirculation

-EK2..2 Knowledge of the interrelations

between the (Loss of Emergency Coolant Recirculation)

and the following:

Facility*s

heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.Question Number: Tier 1 Group 1 55 Importance

Rating: 3.9 Technical Reference:

ECP-1.1 and ECP-l.l, LOSS OF EMERGENCY COOLANT RECIRCULATION

OPS-52532D

and FNP-O-ECB-l.l

Proposed references

to be provided to applicantsduringexamination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: This meets the KA in that the operator has to know the strategy of the procedure and the proper operation of the the various heat removal systems to control the casualty in progress.Items Not Scrambled FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK IDGHER RO GO Version:0123456789

Answer: AAAAAAAAAA

Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:27 PM 164

QUESTIONS REPORT for 75 RO Questions 65.E15 EK3.4 003 Which ONE of the following is a potential source of the flooding that is checked for in FRP-Z.2;and the concern if the maximumexpectedpost-accident

containment

water level (design basis containment

flood level)is exceeded?A.*Condensate

Storage Tank..*Thermal shock to the reactor vessel lower head due to quenching.

B.*Condensate

Storage Tank.*Damage to vital system or components

rendering them inoperable.*Service Water system.*Damage to vital system or components

rendering them inoperable.

D.*Service Water system.*Thermal shock to the reactor vessel lower head due to quenching.

Monday, January 14, 2008 2:42:27 PM 165

QUESTIONS REPORT for 75 RO Questions A.Incorrect.

CST is not one of the potential sources of water FRP-Z.2 addresses and the background

does not mention.However, since AFW goes into ctmt and is in use, it is plausible that this large source of water would be checked for and isa concern.The maintenence

sump isisolatedfrom

the bottom of the reactor vessel by a wall with an elevation higher than the vital equipment of concern.Plausible, because a high enough containment

level would allow water to potentially

thermally shock the hot post accident reactor vessel.B.Incorrect.

CST is not one of the potential sources of water FRP-Z.2 addresses and the background

does not mention.The reason is correct.C.Correct.Service water is the most likely source since it does not isolate to the ctmt on a phase A or B signal and is the largest source of water available to ctmt.The purpose of the sump is to collect and divert water in areas that will not affect vital plant equipment.

Flooding may jeopardize

that function.RWST, CST,&RCS are expected to provide their full volumes to the CTMT sump in accident analysis.D.Incorrect.

SW.is correct.The reason is incorrect.

Plausible, because a high enough containment

level would allow water to potentially

thermal shock the hot post accident reactor vessel.FNP-O-FRB-Z.2

specific background

document for FRP-Z.2 for step 1 Basis: This step instructs the operator to try to identify the unexpected

source of the water in the containment

sump.Containment

flooding is a concern since critical plant components

necessary for plant recovery may be damaged and rendered inoperable.

A water level greater than the design basis flood level provides an indication

that water volumes other than those represented

by the emergency stored water sources (e.g., RWST, accumulators, etc.)have been introduced

into the containment

sump.Typical sources which penetrate containment

are service water, component cooling water, primary makeup water and demineralized

water.All possible*plant specific sources which penetrate containment

should be included in this step.These systems provide large water flow rates to components

inside the containment

and a major leak or break in one of these lines could introduce large quantities

of water into the sump.Identification

and isolation of any broken or leaking water line inside containment

is essential to maintaining

the water level below the design basis flood*level.Monday, January 14, 20082:42:27

PM 166

QUESTIONS REPORT for 75 RO Questions E15 EK3.4 Containment

Flooding-EPE Knowledge of the reasons for the following responses as they apply to the (Containment

Flooding)RO or SRO function as a within the control room team as appropriate

to the assigned position, in such a way that procedures

are adhered to and the limitations

in the facilities

license and amendments

are not violated.Question Number: Tier 1 Group 2 64 Importance

Rating: 2.9 Technical Reference:

FRP-Z.2 and FNP-O-FRB-Z.2

specific background

document for FRP-Z.2 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

OPS 52533M0110 CFR Part 55 Content: 41.1 0 Comments: This meets the KA in that it tests the knowledgeofthe operator on where the most likely source of water would come from and then the limitations

that the bkgrd documents speak of.MCS Time: Source: Cognitive Level: Job Position: reviewed:" Points: 1.00 NEW LOWER RO GO Version: a123456789 Answer:.CAAAC C"DCCA Scramble Range:A-D Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:27 PM 167

QUESTIONS REPORT for 75 RO Questions 66.G2.1.10 001 Unit 2 is in MODE 2 and a reactor startup is in prqgress.In accordance

with Technical Specifications 2.1, Safety Limits (SLs), which ONE of the following describes the RCS Pressure Safety Limit, and the MAXIMUM time to take action'if it is exceeded?Limit MAXIMUM time A.2735 psig 5 minutes2735 psig 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C.2750 psig.5 minutes D.2750 psig 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months A.Incorrect.

In Modes 3,4,5, TS allows 5 minutes to restore pressure B.Correct.In Modes 1 or 2, TS allows 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to Mode 3.In Modes 3,4,5, TS aUows 5 minutes to restore pressure.C.Incorrect.

2750 would be correct in psia, but not in psig.D.Incorrect.

2750 would be correct in psia, but not in psig.G2.1.10 Conduct of Operations

Knowledge of conditions

and limitations

in the facility license.Question Number: 67 Tier 3 Group 1 Importance

Rating: 2.7 Technical Reference:

TS 2.2.2 Proposed references

to be pro.vided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.2/41.10

Comments: This meets the KA in that it tests the knowledg,e

of a RCS safety limit and the applicable

tech spec requirements

for that limit for RO knowledge.

MCS Time: Points: 1.00 Version:0123456789"Answer:BCCCDBDBBC

Scramble Range:A-D Monday, January 14, 20082:42:27

PM 168

Source: Cognitive Level: Job Position: reviewed: MODIFIED LOWER RO GTO QUESTIONS REPORT for 75 RO Questions Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:27 PM 169

QUESTIONS REPORT for 75 RO Questions 67.G2.l.l2 001 Unit 1 is at 1000/0 power.All three Auxiliary Feedwater Pumps have just been declared INOPERABLE.

Which ONE of the following actions MUST be taken?A.Be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in Mode 4 in in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .B.Take action to restore at least one AFW pump to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and a second AFW pump within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or be in Mode 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Immediately

take action to restore at least one AFW pump to OPERABLE status.D.Immediately

enter LCO 3.0.3 and take actions to initiate a shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .A.Incorrect.

T.S.3.7.5 with 2 trains INOP or Required Action of A or B not met, then the action is to enter mode 3 in 6 and mode 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .B.Incorrect.

T.S.3.7.5 does not havea1 hour Action for AFW.If the AFW pump was returned in one hour a case can be made it was RTS immediately, but the second pump being INOP is not allowed to be INOP for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months wlo action.Therefore, the distracter

is incorrect since when one AFW pump is RTS, Required Action for C.is to place the unit in mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , not wait an additional

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to fix it.This is a sly way of usinga6 hour LCO from memory in that the unit has to be placed in mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months but the LCO C is entered immediately

once one AFW pump is RTS.C.Correct.This answer reflects the Note contained in action D as discussed above.TS 3.7.5 in Condition D has a NOTE that states: LCO 3.0.3 and all other action statements

requiring a Mode change are suspended until one AFW train is restored to operable status.This prevents placing the plant in a much higher risk condition than required.D.1 says: Initiate action to restore one AFW train to OPERABLE status.D.Incorrect.

note in LCO 3.7.5 LCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.Monday, January 14, 20082:42:27

PM 170

QUESTIONS REPORT for 75 RO Questions G2.1.12 Conduct of Operations

Ability to apply technical specif-ications

for a system.Question Number: 66 Tier 3 Group 1 Importance

Rating: 2.9 Technical Reference:

T8 section 3.7.5 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 43.2/41.10

Scramble Range:A-D FARLEY GTO MODIFIED LOWER Comments: this meets the KA at an RO expected level of knowledge for the AFW system.MCS Time: 1 Points: 1.00 Version: a123456789 Answer: CDCAAADACB

Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:27 PM 171

QUESTIONS REPORT for 75 RO Questions 68.G2.1.8 004 Given the following:

An oil spill has occurred from a non-PCB oil source in the Turbine Building.*The following conditions

exist:*The oil has reached the Turbine Building sump.*The sump pump is*running and releasing water to the environment.

Which ONE of the following actions is required to be performed by the control room team lAW AP-60, Oil Spill Prevention

Control and Countermeasure

Plan, Hazardous Waste Contingency

Plan?A':'*Dispatch*the SSS to the scene.*Direct the TB System Operator stop the sump pump, then close and tag the discharge valve.B.*Dispatch the shift chemist to the scene.*Direct the TB System Operator place the sump on recirc until the sump contents can be analyzed.C.*Dispatch the SSS to the scene.*Direct the TB System Operator place the sump on recirc until the sump contents.can be analyzed.D.*Dispatch the shift chemist to the scene.*Direct the TB System Operator stop the sump pump, then close andtagthe discharge valve..A.Correct-Dispatch the SSS to the scene, have the TB System Operator stop the sump pump, close and tag the discharge valve.AP-60 requires these actions: SSS to the scene, stop the release and close and tag sump discharge valves.B.Incorrect-The shift radiochemist

is not required to be dispatched, though this may be a good idea, however chemistry supervision

is required to be notified.The shift radiochemist

is not necessarily

supervision.

The TB System Operator should not place the sump on recirc.With a release in progress the requirement

is to stop the release.C.Incorrect-

The TB System Operator is required to stop the release immediately, not evaluate how much more water can be released.The release needs to analyzed prior____

s'lpervisor-apptO¥aLto-stactc-l-itb-.

D.Incorrect-The shift radiochemist

is not required to be dispatched, the release would not continue with oil going to it due to the potential for release to the enviro.nment, and the discharge valve would be closed, and also tagged.Monday, January 14, 2008 2:42:27 PM 172

QUESTIONS REPORT for 75 RD Questions G2.1.8 Conduct of Operations

Ability to coordinate

personnel activities

outside the control room.Question Number: Tier 3 Group 1 68 Importance

Rating: 3.8 Technical Reference:

AP-60 Appendix 1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: Comments: Items Not Scrambled.FARLEY MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK LOWER RO GO Version:0123456789

Answer: AAAAAAAAAA

Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: YES Monday, January 14, 2008 2:42:27 PM 173

QUESTIONS REPORT for 75 RO Questions 69.G2.2.22 002 Unit 1 is in a Refueling Outagewithfuel being loaded into the core.Which one of the following describes the MINIMUM temperature

and the MINIMUM borated water volume that must be met to maintain an operable Boric Acid Storage Tank (BAT Ta.nk)?Solution Temperature

Borated Water Volume A.35°F 2,000 gal.B.35°F 11 ,336 gal.65°F 2,000 gal.D.65°F 11 ,336 gal.Reference:

Technical

Manual, TRM 13.1.6.4 and 13.1.6.6 A.Incorrect, Mode 5 and 6.TRS 13.1.6.6 Verify the contained borated water volume in the boric acid storage tank is 2: 2,000 gal., TRS 13.1.6.1 Verify RWST solution temperature

is>or equal to 35°F B.Incorrect, Mode 5 and 6, TRS 13.1.6.1 Verify RWST solution temperature

is>...Q[equal to 35°F.Mode 1,2,3&4, TRS 13.1.7.4 Verify the contained borated water volume in the boric acid storage tank is 2: 11 ,336 gal C.Correct, Plant is in Mode 6.The following TRSs apply.TRS 13.1.6.4 Verify boric acid storage tank solution temperature

is>or equal to 65°F, TRS 13.1.6.6 Verify the contained borated water volume in the boric acid storage tank is 2: 2,000 gal D.Incorrect, Mode 5 and 6, TRS 13.1.6.4 Verify boric acid storage tank solution temperature

is>or equal to 65°F.Mode 1 ,2,3&4, TRS 13.1.7.4 Verify the contained borated water volume in the boric acid storage tank is 2: 11 ,336 gal Monday, January 14, 2008 2:42:27 PM 174

QUESTIONS REPORT for 75 RO Questions G2*.2.22 Equipment Control Knowledge of limiting conditions

for operations

and safety limits.Question Number: 69 Tier 3 Group2.Importance

Rating: Technical Reference:

3.4 TRM 13.1.6.4 and 13.1.6.6 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10/43.2

Comments: replaced this question since the concept is already tested on the SRO portion of the exam, ie., IR instrument

failed at 10-8 amps and what to do and why.Scramble Range:A-D FARLEY Version: a123456789 Answer: CDADBDCADB

Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO 1.00 BANK LOWER RO GTO Points: This is more of an RO question with no overlap on the exam'and meets the LCO requirements

above.MCS Time: Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 20082:42:27

PM 175

QUESTIONS REPORT for 75 RO Questions 70.02.2.34002

Given the following:

Unit 1 is in Mode 3.*Reactor tripped from 1 000/0 RTP.*ECC has been calculated

for astartup12 hours after the trip.*Estimated critical rod position is Control Bank D at 1 00 steps.*Startup is delayed for TWO (2)hours.Which ONE of the following describes the effect on 11M plot data taken during the approach to critical?The 11M plot will predict criticality

at a.....A.LOWER rod height due to Xenon concentration

greater than that assumed in ECC calculation.LOWER rod height due to Xenon concentration

less than that assumed in ECC calculation.

C.HIGHER rod height due to Xenon concentration

less than that assumed in ECC calculation.

D.HIGHER rod height due to Xenon concentration

greater than that assumed in ECC calculation.

Monday, January 14, 20082:42:28

PM 176

QUESTIONS REPORT.for 75 RO Questions A: Incorrect.

Lower rod height is correct.Xenon concentration

greater is incorrect.

Xenon concentration

will be less but will be adding positive reactivity

which will result in a lower rod height for criticality

to be obtained.B: Correct.Delay will affect core reactivity

since Xenon is decaying, reducing the negative reactivity

in the core.Rods will not have to be withdrawn as far to make the reactor critical.C: Incorrect.

Higher rod height is incorrect.

Xenon concentration

will be less but will be adding positive reactivity

which will result in a lower rod height for criticality

to be obtained.D: Incorrect.

Rods will not have to be withdrawn as far to make the reactor critical.Delay will affect core reactivity

since Xenon is decaying, reducing the negative reactivity

in the core.Candidate needs to demonstrate

an understanding

of the time that it takes Xenon to peak from a full power trip, which is typically the square root of the equilibrium

power level..G2.2.34 Equipment Control Knowledge of the process for determining

the internal and external effects on core reactivity.

Question Number: 70 Tier 3 Group 2 Importance

Rating: 2.8 Technical Reference:

Physics curve 60 Proposed references

to be provided to applicants

during examination:

None Learning ObJective:

10 CFR Part 55 Content: 41.1 Comments: This KA tests the operator to evaluate core reactivity

and the effects of xenon after a rx trip for a startup.MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 MODIFIED HIGHER RO GO Version:0123456789

Answer: B D-cADCDAC D Scramble Range:A-D Source if Bank: SEQUOY AH 2004 Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:28 PM 177

QUESTIONS REPORT for 75 RO Questions 71.02.3.10007

What precaution

is required to be taken at the 121 1 Piping Penetration

Room (PPR)prior to lowering RCS level to mid-loop lAW UOP-4.3, Mid Loop Operations?

A.The door to the 121 1 P.PR must be locked.B.Health physics (HP)must survey the 121 1 PPR.A caution sign must be placed at the entrance of the 121 1 PPR.D.All vent valves on systems in the 121 1 PPR penetrating

containment

must be verified closed.UOP-4.3 2.24 Prior to reducing Res level, a caution sign concerning

the establishment

of containment

closure must be placed at the entrance of the following locations.

NOTE: The signs can be obtained from the Shift Support Supervisor

and are normally stored in the CCW Storage Room on Unit 2.2.24.1 139'Electrical

Penetration

Room 2.24.2 121'Piping Penetration

Room 2.24.3 100'Piping Penetration

Room 2.24.4 Main Steam Valve Room 2.24.5 Personnel Access Hatch 2.24.6 Auxiliary Access Hatch A.Incorrect-Door not required to be locked SOP 0.0, 15.3.5 The following doors will be locked closed when unattended

during unit operation in Modes 1 through 4:*.139'Electrical

Penetration

Room Doors 317 A/2317*121'Piping Penetration

Room Doors 214/2214 B.Incorrect-Surveys are not required prior to reducing level.C.Correct-per the above initial condition of UOP-4.3 D.Incorrect-Air to air barrier not required for midloop integrity (ctmt closure in 2 hrs)Monday, January 14, 20082:42:28

PM 178

QUESTIONS REPORT for 75 RO Questions G2,,3..10 Radiation Control Ability to perform procedures

to reduce excessive levels of radiation and guard against personnel exposure.Question Number: 71 Tier 3 Group 3 Importance

Rating: 2.9 Technical Reference:

Health.Physics manual Proposed references

to be provided to applicants

during examination:

None Learning Objective:

Describe how the rad monitoring

system helps to protect the health and safety of plant workers and the public.(ESP521 06008)10 CFR Part 55 Content: 43.4 Scramble Range:A-D FARLEY Version:0123456789

Answer:CB CACAAADA Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO BANK LOWER RO GO Comments: This question tests the basic generic applicability

of the KA at an RO level of knowledge.

MCS Time: 1 Points: 1.00 Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:28 PM 179

QUESTIONS REPORT for 75 RO Questions 72.G2.3.9 001 The following conditions

exist on Unit 2 while at 100%power:*Containment

Main PurgeandMini-Purge

are secured.*CTMT to ATMOS DP is currently 0.3 psid.*All pre-requisites

to perform a batch release of the containment

atmosphere

have been met.Which ONE of the following describes where the containment

purge system discharges

to and how the system is operatedtoreduce containment

pressure lAW SOP-12.2, Containment

Purge and Pre-access

Filtration

System, Appendix 3, Batch Releases of Containment

Atmosphere?

A.*Discharges

directly to the plant vent stack;*Open the Mini-Purge

dampers, then start the Mini-Purge

supply and exhaust fans to initiate the release.B.*Discharges

directly to the exhaust plenum;*Open the Mini-Purge

dampers to initiate the release.When CTMT to ATMOS DP is<0.25 psid, then start the Mini-Purge

s'upply and exhaust fans.C.*Discharges

directly to the plant vent stack;*Open Purge Filter Outlet Valve, V-294, then open the Mini-Purge

dampers and start the Mini-Purge

supply and exhaust fans.Then close V-294.*Discharges

directly to the exhaust plenum;*Open Purge Filter Outlet Valve, V-294, then open the Mini-Purge

dampers.When CTMT to ATMOS DP is<0.25 psid, then close V-294 and start the Mini-Purge

supply and exhaust fans.Monday, January 14, 2008 2:42:28 PM 180

QUESTIONS REPORT for 75 RO Questions DISTRACTOR

ANALYSIS: A Incorrect.

not the correct order, the incorrect release path and v294 is not being used as required.B Incorrect.

Incorrect order and v294 is not being used as required.C Incorrect.

incorrect release path and the pressure has to be.checked<.25 psid to start the fans.D Correct.per the procedure below and the prints REFERENCES:

1.SOP-12.2 Containment

Purge and Pre-access

filtration

system, Rev.34/27 3.2 Open N2P13V294, PURGE FILTER COOLING OUTLET VALVE.3.3 WHEN performing

the following valve manipulations, THEN note the start time for recording purposes: 3.3.1 Place the following CTMT Purge DMPRS hand switches to MINI to initiate CTMT Batch Release: HS-3196 HS-3198 3.3.2 Record start date/time data in Part III of the Batch Gaseous Waste Release Permit.IY 3.4 WHEN CTMT DIFF PRESSURE decreases to=0.25 psid, THEN perform, the following, noting the fan start time and containment-to-atmosphere

delta pressure for recording purposes: 3.4.1 Close N2P13V294, PURGE FILTER COOLING OUTLET VALVE.3.4.2 Start MINI PURGE SUPP/EXH FAN.3.4.3 Record fan start date/time data in Part III of the Batch Gaseous Waste Release Permit.Monday, January 14, 20082:42:28

PM 181

QUESTIONS REPORT for 75 RO Questions G2.3.9 Radiation Control Knowledge of the process for performing

a containment

purge.Question Number: 72 Tier 3 Group 3 2.5 SOP-12.2, step 4.4 version 33.0 Importance

Rating: Technical Reference:

P&lOs: 5.1.1 0-205010, sheets 1 and 2, Containment

Cooling and Purge SystemP&10 5.1.2 0-207783, Elem.Oiag., Containment

Mini-Purge

Fans 5.1.3 0-207204, Elem.Oiag., Containment

Purge Iso.Oampers Train A 5.1.4 0-207199, Elem.Oiag., Containment

Purge Iso.Oampers Train B 5.1.5 0-207236, Elem.Oiag., Containment

Purge Air Handling Unit Fan 5.1.6 0-207237, Elem.Oiag., Containment

Purge Exhaust Fan 5.1.7 0-204654, Conn.Diag., Containment

Purge Starter Panels Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 43.4 Scramble Range:A-D FARLEY BANK LOWER RO GO Comments: This is according to SOP-12.2 appendix 3 for the batch release.This question requires knowledge of process (procedure)

for equalizing

containment

pressure with atmospheric

pressure when initiating

a containment

purge.One'answer choice (distractor)

may result in system damage.MCS Time: 1 Points: 1.00 Version:0123456789

Answer: DBACADCADA

Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Source: Cognitive Level: Job Position: reviewed: Monday, January 14, 2008 2:42:28 PM 182

QUESTIONS REPORT for 75 RO Questions 73.02.4.22 002 Given.the following:

FRP-H.1, Response to Loss of Secondary Heat Sink, is in progress in response to a Red Heat Sink condition.

The crew is still progressing

through FRP-H.1 when critical safety function status tree conditions

are reported as follows:*Subcriticality:

Orange*Core Cooling: Green*Heat Sink: Yellow*Integrity:

Green*Containment:

Red*Inventory:

Yellow Which ONE of the following describes the actions the ,crew should take in response to the conditions

given above?A.Complete FRP-H.1, then transition

to FRP-S.1.Complete FRP-H.1, then transition

to FRP-Z.1.c.Immediately

exit FRP-H.1 and transition

to FRP-S.1.D.Immediately

exit FRP-H.l and transition

to FRP-Z.1.A is incorrect.

S.1 is higher priority CSF, but lower challenge (Orange).B is correct.Entered H.1 on red condition, must complete prior to any other lower procedure.

C is incorrect.

Even though heat sink is no longer red, would not immediately

leave.D same as C, except that current conditions

indicate a transition

to Z.1 is required.Monday, January 14, 2008 2:42:28 PM 183

QUESTIONS REPORT for 75 RO Questions G2..4..22 Emergency Procedures

I Plan Knowledge of the bases for prioritizing

safety functions during abnormal/emergency

operations.

Question Number: Tier 3 Group 4 75 Importance

Rating: 3.0 Technical Reference:

FRP-H.1 and EOP Users Guide Proposed references

to be provided to applicants

during examination:

None Learning Objective:

.10 CFR Part 55 Content: 41.1 0 Comments: Scramble Range:A-D NORTH ANNA MCS Time: Source: Cognitive Level: Job Position: reviewed: Points: 1.00 BANK HIGHER RO GTO Version: a123456789 Answer: B AAC BADCCB Source if Bank: Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:28

PM 184

QUESTIONS REPORT for 75 RO Questions 74.G2.4.34 001 Gi'ven the following:*A fire required evacuation

of the control room.*The crew is performing

actions of AOP-28.2, Fire in the Control Room.*HSO Panel A is manned and functional.

The crew is at the step to adjust HIK-122, CHG FLOW, to maintain pressurizer

level within the required band when the following is reported:*HIK-122 on the HSO panel'is not controlling

FCV-122 properly.*Pressurizer

level is 16%and trending down.Which ONE of the following contains a correct method for controlling

PRZR level and the correct location of the components

to be operated lAW

A.Close LCV-459 or 460, L TON LINE ISO, from the HSO panel.Control charging flow using the bypasses around FCV122 locally in the 1 00*Piping Penetration

Room entrance.B.Close HV-8149A and 8149B or C, L TON ORIF ISO, at the HSO panel.Control charging flow using the bypasses around FCV-122 locally in the 1 00'hallway BIT area.C.Close LCV-459 or 460, L TON LINE ISO, from the HSO panel.Control charging flow using MOV-8803A or B, HHSI TO RCS CL, locally in the 100'Piping Penetration

Room entrance.Close HV-8149A and 8149B or C, L TON ORIF ISO, at the HSO panel.Control charging flow using MOV-8803A or B, HHSI TO RCS CL, locally in the 100'hallway BIT area.Monday, January 14, 2008 2:42:28 PM 185

QUESTIONS REPORT for 75 RD Questions A.incorrect.

LCV-459 or 460 can not isolated at the HSDP and would not be isolated procedurally

per the note below.Control of charging using the bypass valves is an option, actually the first option, (step 14.6)but in this case the first part of the distracter

is not correct and the location is correct.ADP-28.2 NOTE: Isolation of letdown due to low pressurizer

level (15%)will unnecessarily

complicate

plant recovery (LeV 459&460 cannot be re-opened from the HSDP, Reactor head vents must then be used for removing mass from the primary system).Therefore, emphasis should be placed on controlling

charging flow to establish a stable or slowly rising pressurizer

level that compensates

for any effect on level due to cooldown.B.incorrect.

Placing 2 orifices on service is correct at step 25 in the procedure for controlling

level, and bypassing FCV-122 is correct, but the

is not correct., C.incorrect.

LCV-459 or 460 can not isolated at the HSDP and would not be isolated procedurally

per the note below.control charging flow using MOV8803A or B, HHSI TO RCS CL, is correct but the location is not correct.D.Correct.Placing 2 orifices on service is correct step 25 in the procedure for controlling

level, and control charging flow using MOV8803A or B, HHSI TO RCS CL, is correct and the location is correct.G2.4.34 Emergency Procedures

I Plan Knowledge of RO tasks performed outside the main control room during emergency operations

including system geography and system implications.QuestionNumber:

74 Tier 3 Group 4 Importance

Rating: 3.8 Technical Reference:

FNP-1-AOP-28.2

step 14.6 and the note above as well as step 25.2 Proposed references

to be provided to applicants

during examination:

None Learning Objective:10 CFR Part 55 Content: 41.1 0 Comments: This question tests the knowledge of an RO task outside the control room during an emergency and tests geography, plant location as well as procedural

guidance and operational

implications, which include the letdown portion of the question and the note-----eles-ertb-ifl-g-why-teteewft-is-not-iselatee.b---------------------

MCS Time: Points: 1.00 Version:0123456789

Answer: DDCAB AB DB B Scramble Range:A-D Monday, January 14, 20082:42:28

PM'186

Source: Cognitive Level: Job Position: reviewed: NEW LOWER RO GO QUESTIONS REPORT for 75 RO Questions Source if Bailie Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 20082:42:28

PM 187

QUESTIONS REPORT for 75 RO Questions 75.02.4.49 002 Given the following:

The crew is performing

the actions of FRP-H.1, Loss of Secondary Heat Sink, following a reactor trip due to a loss of feedwater.

RCS pressure is 2280 psig.*Containment

pressure is 1 psig.Which one of the following sets of steam generator wide range level parameters

meet the FRP-H.1 foldout page criteria for feed and bleed for the conditions

given?A.1A SG-28%1'B SG-29%1C SG-0%B.1A SG-0%1 B SG-13%1CSG-15%C.1ASG-31%1B SG-29%1C SG-32%1A SG-11%1B SG-11%1C SG-14%Monday, January 14, 2008 2:42:28 PM 188

1.1.1 Stop all RCPs.RCP[]1A[]1B[]1 C 1.1.2 Proceed, to Step 12 QUESTIONS REPORT for 75 RO Questions A.Incorrect.

Plausible, 2 of 3 SG WR levels<31%is the figure for adverse containment

initiation

of feed and bleed.B.Incorrect.

plausible since 2 of 3 SG WR levels<28%was chosen since candidate could confuse with the adverse containment

figure for SG NR level control of 28%.Also this could be chosen if the candidate did not remember 2 of 3 and thought it was 1 of 3 for non adverse numbers.C.Incorrect.

Plausible, 2 of 3 SG WR levels<or equal to 31%is the setpoint for a Dry SG during adverse containment.

Possible candidate may confuse the setpoint.Also this could be chosen if the candidate did not remember 2 of 3 and thought it was 1 of 3 for adverse numbers or did not remember greater than 31%.D.Correct.2 of 3 SG WR levels<12%requires feed and bleed with normal containment

pressure conditions.

FRP-H.1 Foldout page requirements

1 Monitor bleedandfeed criteria.(applicable

steps 1 thru 11 only)1.1 Check at least two SG*wide 1.1 Perform the following.

range levels-GREATER THAN 12%{31%}.G2..4.49 Emergency Procedures

I Plan Ability to pertorm without reference to procedures

those actions that require immediate operation of system components

and controls.Question Number: Tier 3 Group 4 73 Scramble Range:A-D Importance

Rating: 4.0 Technical Reference:

FRP-H.1 Proposed references

to be provided to applicants

during examination:

None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments: meets the KA in that the question tests the FO page of H.1 which are IOAs of that procedure when those conditions

require entry.This is required RO knowledge.

MCS Time: 1 Points: 1.00 Version:0123456789

Answer: DDABBBB DBA VOGTLE Source: Cognitive Level: Job Position: reviewed: BANK LOWER RO GO Source if Banle Difficulty:

Plant: FARLEY Previous 2 NRC exams: NO Monday, January 14, 2008 2:42:28 PM 189

ML080280360

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