ML083640084

Draft - RO & SRO Written Exam (Folder 2)

ML083640084
Person / Time
Site:	Beaver Valley
Issue date:	10/23/2008
From:	FirstEnergy Nuclear Operating Co
To:	David Silk Operations Branch I
Hansell S
Shared Package
ML081060562	List: IR 05000412/2008301 ML083370145 ML083400431 ML083640082 ML083640084 ML083640089 ML090120828 ML090120832 ML090120846 ML090120853 ML090230109 ML090260673 ML091760057
References
TAC U01628
Download: ML083640084 (100)

Text

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

1.

The plant is operating at 100% power with all systems in NSA EXCEPT:

Power Range Channel N44 has been declared inoperable.

Power Range Channel N44 has been removed from service IAW AOP-2.2.1C, Power Rarge Channel Malfunction.

Power Range Channel N43 NOW fails HIGH.

All systems function as designed.

No Operator Actions have been taken Which, of the below listed First Out Annunciators (ANN. A5), will alarm in the FIRST 45 seconds AFTER N43 fails High?

(1) A5-1 D 213 Loops Overtemp AT Reactor Trip (2) A5-2A Reactor Protection System Train A Trouble (3) A5-5G Reactor Trip Due To Turbine Trip (4) A5-66 Turbine Anti-Motoring Turbine Trip (5) A5-6D Turbine Trip Due To Reactor Trip (6) A5-7D Generator Trip Due To Turbine Trip A.

1,3,5 & 6 ONLY

6.

2, 4, & 6 ONLY C.

3,5&6ONLY 1.2.3 & 4 ONLY Answer C

ExplanationlJustification:

A.

6.

Incorrect. N-44 does NOT input into GTAT trip setpoint calculation, therefore this alarm will NOT be energized.

incorrect. Candidate may confuse rod control urgent alarm with protection system trouble. Rod control urgent wiii energize on the trip Anti-motoring would alarm if the output breakers did not open. However, stem of the question states that all systems functioned as designed.

3 and 5 will both be alarmed.

Correct. IAW 2OM-1.4.ABB. 1.4.AA1, 1.4.AAD 26.4.AAF. 26.4.AAI and 35.4.AAF Incorrect N-44 does NOT inpul inlo OTAT trip setpoint calculation, therefore this alarm will NOT be energized. Candidate may confuse rod control urgent alarm witn protection system trouble. Rod control urgent will energize on the trip, Antt-motoring would aiarm if the output breakers dld not open. However. stem of the question states that all systems functioned as designed. 5 and 6 will both be alarmed.

C.

0.

~-

KIA Sys #

KIA System KIA Category KIAStatement 000007 Reactor Trip KIA#

EK2.03 KIA Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

1.4.AAD. 26.4.AAi and 35 4.AAF Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 41.7 145.7)

Knowledge of the interrelations between a reactor trip and the following:

Reactor trip Status panel

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

2.

The plant is operating at 100% power with all systems in NSA.

A PRZR vapor space accident occurs.

PRZR pressure drops to 1200 psig, The Highest Steam Generator pressure is 1000 psig.

HHSI flow is 800 gpm and stable.

All systems functioned as designed.

NO Orange or Red path conditions exist.

The crew is performing the actions of E-I, Loss of Reactor or Secondary Coolant.

At Step 2, Check if RCPs should be stopped, the crew is directed to Stop ALL RCPs WHY MUST the RCPs be stopped at this time?

The RCPs are tripped to:

prevent possible pump damage by running the RCPs under highly voided conditions in order to sa\\,e the pumps for potential future use.

prevent excessive depletion of RCS water inventory which might lead to severe core uncovery if th.:

RCPs were tripped later in the event.

ensure RCS liquid inventory has depleted to the point where tripping the RCPs will cause the break to immediately uncover.

ensure the RCP seal package is not damaged by the excessive temperature or steam voiding associated with this event.

A.

B.

C.

D.

Answer B

planationIJustification:

fi.

E.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement incorrect. This IS the reason they are stopped in FR-C.2 Correct IAW with E-1 step 2 basis and RCP trip generic issue.

Incomct. This is what the RCP trip criteria is attempting to prevent, not ensure.

incorrect. This IS the consequence of losing both seal injection and RCP thermal barrier cooling.

000008 KIA U AK3.04 KIA Imporlance 4.2 Exam Level RO Level Of Difficulty: (1 -5)

Pressurizer Vapor Space Accident Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident:

RCP lripping requirements Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

E-1 step 2 bases; 20M-53B.5.Gi-6 pag?' 6 Objective #:

Task ID#:

10CFR Part55Content:

(CFR41.5,41.10/45.6/45.13 2M paragraph

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

3.

The plant is operating at 100% power with all systems in NSA.

A small break LOCA occurs coincident with a loss of offsite power.

All systems function as designed EXCEPT EDG #2 fails to start and CANNOT be started.

10 minutes after the event began; the crew is performing recovery actions IAW ES-1.2, Pcst COCA Cooldown and Depressurization.

IAW ES-1.2 step 1 Reset SI, the RO depresses the Safety Injection Signal Train A AND Ti-ain B reset pushbuttons.

AFTER the Safety Injection Signal Train A AND Train B reset pushbuttons have been depressed, ihe following plant conditions exist:

PRZR pressure is 1350 psig and slowly rising.

RCS Subcooling is 95°F and stable.

4KV bus 2DF is de-energized.

Annunciator A12-IC Auto Safety Injection Blocked is flashing (white then dark).

Annunciator A12-ID Safety Injection Signal is flashing (white then dark).

Based on these conditions:

What is the current status of the automatic Safety Injection Actuation system AND what is the significance of annunciators A12-IC and A12-1 D flashing?

A.

ONLY one Train of Safety Injection has reset; the flashing annunciators indicate a status difference between the two trains of automatic Safety Injection actuation.

ONLY one Train of Safety Injection has reset; the flashing annunciators indicate the Purple Train 0' electrical power will not respond to an automatic Safety Injection actuation signal.

BOTH Trains of Safety Injection have reset; the flashing annunciators indicate pressurizer pressure is still below the low pressure automatic safety injection setpoint.

BOTH Trains of Safety Injection have reset; the flashing annunciators indicate automatic safety injection actuation will not occur until the reactor trip breakers are re-closed.

B.

L.

D.

Answer A

ExplanationlJustification:

A.

B.

C.

D.

KIA sys #

KIA system KIA Category KIA Statement 000009 Small Break LOCA Ability to operate and monitor the foilowing as they apply ESFAS to a small break LOCA:

KIA #

EA1 13 KIA Importance 4.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

ES-1.2 step 1 background Page 7 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 41.7 / 45.5 / 45.6)

Correct. IAW ES-1.2 step 1 background document page 7 Incorrect. Right status of SI actuation system: Wrong significance of flashing alarms. inoperable electrical trains are indicated by the BlSl sWem NOT the flashing of annunciators Al2-1C and 1D.

Incorrect. Only one train of SI has reset. SI will reset even though an SI signal is still present due to the retentive memory circuit and the P 4 contact development.

Incorrect Only one train of SI has reset. Closing the reactor trip breakers and re-arming automatic SI is indicted when A12-IC goes DARK

~~

~

Page 3 Of '00

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

4.

The plant is operating at 100% power with all systems in NSA.

Annunciator A2-5C, Reactor Coolant Pump Vibration AlerVDanger Alarms "B" RCP shaft vibration is 16 mils and stable "6" RCP frame vibration is 1 mil and stable The crew enters AOP-2.6.8, Abnormal RCP Operation.

While performing the actions of AOP-2.6.8, Abnormal RCP Operation, the following additional alarns and indications are received:

A2-5D, Reactor Coolant Pump Seal Vent Pot Level High/Low (RCP 216 Seal Pot L,,d High, computer address point L0508D)

A2-4D, Reactor Coolant Pump Sea Trouble (RCP 218 Seal Lk Off CHS-FT155B Lt.w, computer address point F0128D)

RCP 21 B Seal Lk Off CHS-FT155B is.80 gpm and stable Based on these alarms and indications, which "6" RCP seal has failed?

A.

1. I seal B.

1. 2 seal C.

1. 3seal D.

Low pressure seal Answer B

ExplanationlJustification:

A.

Incorrect if #1 seal lad faiied seal leak-off flow would be high NOT low.

Correcl. IAW 20M-7 4.AAH. 6.4.AAE and AOP-2.6.8 Incorrect. If #3 seal had failed the seal vent pot level lowwould be indicated NOT high.

Incorrect. The low pressure seal is not functional when the motor is coupled to the pump J.

D.

KIA Sys U KIA System KIA Category KIA Statement 000015/17 RCP Malfunctions Knowledge of the interrelations between the Reactor RCP seals

~

Coolant Pump Malfunctions (Loss of RC Flow) and the following:

KIA U AK2.07 WAlmportance 2.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

2OM-7.4.AAH, 6.4.AAE Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 41.7 145.7)

Page 4 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

5.

The plant is in Mode 6.

Preparations to flood the refueling cavity are underway.

RCS water level is ONE (1) foot below the top of the reactor vessel flange and stable All RCS loop isolation valves are CLOSED.

RCS temperature is 100°F and stable.

RCS is vented to atmosphere.

It has been 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> since the reactor was shutdown.

RHR Pump 2RHS*P21A is operating and RHR Pump 2RHS*P21B is in Standby.

RHR Pump 2RHS*P21A TRIPS and RHR Pump 2RHS*P21B WILL NOT start.

The crew enters AOP-2.10.1, Residual Heat Removal System Loss.

At step 11 of AOP-2.10.1, Residual Heat Removal System Loss, the crew is directed to estimate the time to RCS saturation.

Using the attached AOP-2.10.1 figures and attachments, ESTIMATE the time to RCS saturation.

The estimated time to RCS saturation is A.

16 minutes B.

25 minutes C.

37.5 minutes D.

38.6 minutes

' iswer B

cxplanationlJustification:

A.

B.

C.

D.

KIASysU KIASystem KIACategory KIA Statement GOO025 Loss of RHR System NIA Knowledge of abnormal condition procf lures.

KiA U 24.11 KlAlmportance 4.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Application References provided to Candidate Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 143.5 i 45.13) incorrect. This is the number for 140°F Starting temperature.

Correct IAW figure l C and attachment 1 Incorrect. If candidate uses figure 18 instead of figure 1C they will calculate this value.

incorrect. if candidate uses figure 28 instead of figure 2C they will calculate this value.

~

AOP-2.10.1 figures 1A. 16, lC, 2A. 28, 2C. 3, 8 att. 1 Technical

References:

AOP-2.10.1 figure IC and Attachment

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

6.

The plant is operating at 100% power with all systems in NSA.

Primary Component Cooling Water Pump 2CCP*P21 C is on clearance and unavailable Primary Component Cooling Water Pump 2CCP*P21A is running.

Primary Component Cooling Water Pump 2CCP*P21B is in Standby.

2CCP*P21A TRIPS and cannot be re-started.

2CCP*P21 B FAILS to automatically start and cannot be manually started.

The crew enters AOP-2.15.1, Loss of Primary Component Cooling Water AND is instructed to TRIP the reactor and enter E-0, Reactor Trip Or Safety Injection.

Based on these conditions, how will the loss of CCP cooling to the RCPs NOW be addressed?

Immediately Trip ALL RCPs:

THEN complete the immediate operator actions of E-0 AFTER completing the immediate operator actions of E-0.

THEN manually trip the reactor and enter E-0 AFTER the transition is made out of E-0 to ES-0.1, Rx Trip Response.

A.

B.

C.

D.

Answer B

ExplanationIJustification:

A.

6.

7.

J Incorrect. The directions given in AOP-2.15.1 specifically instruct the operators to complete the IOAs of E-0 before tripping the RCPs.

Correct. iAW AOP-2.15.1 step 2 RNO. AT BVPS the topic of AOP use in conjunction with EOP use has been addressed by providing the a.:tions to be completed within the AOP. This is done by providing WHEN statements within the AOP. (2.6.8.2.15.1, 2.6.7)

Incorrect. The directions given in AOP-2.15.1 specifically instruct the operators to complete the IOAs of E-0 before tripping the RCPs. Man ally tripping the reactor would not be necessary since RCP breaker trip would cause an automatic reactor trip.

Incorrect. The dtrecbons given in AOP-2.15.1 specifically instruct the operators to complete the lOAs of E-fl before tripping the RCPs. Canrildate may believe that the E-fl EOP have a higher priority and must be completed before laking additional non-EOP actions.

KIA Sys #

KIA System KIA Category KIA Statement 000026 Lass of Component NIA Knowledge of how abnormal operating Cooling Water procedures are used in Conjunction With EOPs.

KIA #

2.4 8 KIA Importance 3.8 Exam Level RO Level Of Difficulty: (1.5)

Question Source:

New Question Cognitive Level:

Lower Fundamentai References provided to Candidate None Technical

References:

AOP-2.15.1 step 2 RNO Objective #:

Task ID#

10CFRPart55Content:

(CFR: 41.lOl43.5I45.13)

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

7.

A Plant startup is in progress. All systems are in normal alignment for this power level.

All four Power Range channels are indicating 4% and stable.

Both Intermediate Range channels are indicating 2.1 X amps and stable.

Both Intermediate Range SUR channels are indicating Zero DPM and stable.

PRZR pressure begins to drop rapidly, and the Unit Supervisor directs you to manually trip the rextor.

You attempt to trip the reactor from all available control room reactor trip switches HOWEVER; bo'h reactor trip breakers remain CLOSED. An operator is dispatched to locally trip the reactor trip bre: kers Assuming no other operator actions are taken, what will be the status of the Nuclear instrumentati' ~n system one minute AFTER the reactor trip breakers are locally opened?

Power Range indication will be (I) 1 Intermediate Range indication will be (2)

Intermediate Range SUR indication will be (3)

A.

B.

C.

D.

(1) 4% and slowly dropping; (2) 1.7 X amps and slowly dropping; (3) -.I DPM and stable (1) 2% and slowly dropping; (2) 1.7 X 10~5 amps and slowly dropping; (3) -.I DPM and stable.

(I) 0% and stable; (2) 1.0 X 10~6 amps and slowly dropping; (3) -.33 DPM and stable (1) 0% and stable; (2) 1.0 X 10. amps and slowly dropping; (3) -.33 DPM and stable Answer C

ExplanationlJustificatlon:

A.

C.

D.

Incorrect. PR and IR power are too high. SUR is not low enough. Borating and driving rods inward could produce these indications but this would NOT be indicative of the reactor being tripped.

Incorrect. PR and IR power are too high. SUR is not low enough. Borating and driving rods inward could produce these indications but this flould NOT be indicative of the reactor being tripped, Correct Opening the trip breakers from 4% power will result in PR indication going to zero, IR power will drop - 1 decade and SUR will be stable at 4 3 DPM due to rod absorption of the prompt neutrons.

incorrect. PR indication is correct, iR power is too iow. Candidate may believe that trips from low power will stabilize ai 1.O X 10'O amps wt ch is below the point of adding heat and the setpoint for blocking SR during startup. SUR is correct KIASys #

KIASystem KIA Category KIA Statement 000029 ATWS Ability to determine or interpret the following as they apply Reactor nuclear instrumentation KIA #

EA2.01 KIA Importance 4.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

LP GO-GPF-R3 slide 150 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 43.5 145.13) to a A W S :

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

8.

The plant is operating at 100% power with all systems in NSA.

A Steam Generator Tube Rupture occurs.

The crew enters the EOP network.

The crew is currently implementing E-3, Steam Generator Tube Rupture.

The RCS has been cooled to 500°F in preparation for equalizing RCS pressure with the ruptured SG pressure.

The Unit Supervisor directs you to depressurize the RCS AND while maintaining a minimum of 20 'F of Subcooling.

At the current RCS temperature, what is the lowest RCS pressure can be without violating the 20" of Subcooling requirement?

A.

-666 psig B.

-695 psig C.

-798 psig D. -827 psig Answer C

ExplanationlJustification:

A.

6.

C.

D.

Incorrect. Piausible This would be the saturation pressure for 500°F. (680.86 pSia - 14.7psi)

Incorrect Plausible, if candidate attempts to determine pressure for 500'F and mistakenly adds 14.7 psi to 680.86 psla.

Correct. Saturation pressure for 52OOF is 812.53 minus 14.7 psi yields 797.83.

Incorrect. Plausible If candidate mistakenly adds 14.7 psi to the Saturation pressure.

I A Sys #

KIASystem KIA Category KIA Statement 000038 Steam Gen. Tube Rupture Knowledge of the operational implications of the following concepts as they apply to the SGTR.

Use of steam tables KIA #

EK1.O1 KIAlmporlance 3.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate Steam tables Technical

References:

E-3, Steam tables Objective #:

Task ID#:

10CFR Part55 Content:

(CFR41.8I41.10I45.3)

Paoe 8 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

9.

The plant is operating at 100% power with all systems in NSA.

A Steamline break outside containment occurs.

The MSIVs fail to close and they cannot be manually closed All other systems functioned as designed.

All 3 SGs depressurize to atmospheric pressure.

All RCS cold leg temperatures stabilize at 220°F It has been 30 minutes since the steam break occurred.

RCS Subcooling is 200°F.

The operating crew has entered FR-P.1, Response to Imminent Pressurized Thermal Shock Condition due to the excessive cooldown rate and all RCS cold leg temperatures being below the Reference Transition Nil Ductility Temperature (RT,,,)

of 245°F.

Which One (1) of the below listed actions will limit the overall stress on the Reactor Vessel?

A.

Deoressurize the RCS B.

Commence an RCS cooldown C.

Maximize safety injection flow D.

Stop all running RCPs Answer A

ExplanationiJustification:

A.

i.

Correct. IAW FR-P.l bases page 4 one of the major actions to limit the RPV stress is to depressurize the RCS.

Incorrect. IAW FR-P 1 bases page 4 one of the major actions is to stop any cooldown and allow temperature to soak befure re-commenci $9. Must of the EOP Strategies include cooidowns to get on RHR and achieve Mode 5 status. The situation presented by exceeding RTNar is an ex, eption to mast EOP strategies.

Incorrect. IAW FR-P.l bases page 4 one of the major actions is to terminate SI when the criteria are met. Terminating is done to minimize the cold water effects on the vessel downcomer region. Large break. LOCA strategies include maximizing SI flow. The situation presented by eXcL eding RTNoT is an exception to Large break LOCA strategies.

incorrect Stopping RCPs would potentially increase overali vessel stress by allowing the cold SI water contact Vle vessel downcomer r e g m without any mixing. Therefore, RCPs are left running in FR-P.l until support conditions are no ionger available, and then they are securec C.

D.

.~

W A Sys #

WA System W A Category WAStatement 000040 Steam Line Rupture Knowledge of the operational implications of the following Nil ductility temperature KIA #

AK1.04 WA Importance 3.2 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

FR-P 1 bases page 4 Objective #:

Task ID#:

10 CFRPart55Content:

(CFR41.8I41.1OI45.3) concepts as they apply to Steam Line Rupture:

Page 9 I 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

IO.

The plant is operating at 100% power with all systems in NSA.

An inadvertent Reactor trip occurs WITH a coincidental loss of all 4KV AC power.

All other systems operate as designed.

Twenty minutes after the trip, which ONE (1) of the following sets of parameters indicate that natu:al circulation of the RCS has been established?

SG Pressures Core Exit TC's Tcold A.

1060 psig and rising 590 "F and rising 558 "F and dropping B.

1060 psig and stable 577 "F and stable 558 "F and stable C.

1035 psig and dropping 590 "F and rising 550 "F and stable D.

1035 psig and dropping 577 "F and stable 550 "F and droppii-g Answer D

ExplanationlJustification:

A.

6.

C.

Incorrect. CETs rising.

D.

lncoriect. SG Press and CETs are rising and Tcold above Tsat of SG (Tsat for 1060 psig = 553F) lncoriect. Tcold above Tsat of SG. (Tsat for 1060 psig = 553F)

Correct. All parameters stable or dropping and Tcold at Tsat of SG.

NOTE:

See SRO auestion #SO emlanation as to why this auestion has been evaluated to be different enouah from SRO auestion #EO tc be used on the same exam. '

~

'CIA Sys #

KIA System KIA Category KIA Statement a0055 Station Blackout Ability to determine or interpret the following as they apply to a Station Blackout:

RCS core cooling through natural circilation cooling to SIG cooling KIA #

EA2.02 WAlmportance 4.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate Steam Tables Technical

References:

Steam Tables; EOP Attachment A-1 :'

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 43 5 145.13)

Page l O O f 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

11.

The plant is operating at 100% power with all systems in NSA.

An inadvertent Reactor trip occurs WITH a coincidental loss of offsite power.

All other systems operate as designed.

Both Emergency 4KV Buses are being powered by their respective diesel generators.

The crew performs the actions of E-0, Reactor Trip or Safety Injection, and transitions into ES-0.2, Natural Circulation Cooldown.

IAW ES-0.2, Natural Circulation Cooldown, what is the MINIMUM required steam generator water level that must be maintained to provide a stable heat sink during the natural circulation cooldown?

A.

B.

C.

D.

WR level of at least 14%

NR level of at least 12%

NR level of least 35%

NR level of at least 50%

Answer C

ExplanationlJustification:

A.

E.

C.

D.

WA Sys #

KIA System KIA Category KIA Statement 000056 Loss of Off-site Power incorrect. This is the minimum water level for loss of heal Sink in FR-H.1.

lncarrect This is the minimum water level for maintaining the thermal blanket during SGTR recovery Correcl. IAW step 5 of ES-0.2 and step 5 bases.

Incorrect. This is the Maximum water level for naturai Circulation cooidown in ES-0.2.

=c Ability to determine and interpret the following as they apply to the Loss of Offsite Power:

Necessary SIG water level for naturai circulation

, A #

AA2.88 KIA Importance 4.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

ES-0.2 step 5 and step 5 bases, Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

12 A.

B.

C.

D.

The plant is operating at 100% power with all systems in NSA.

A loss of Vital Bus 2 has occurred as a result of a failure in the inverter.

The static switch has FAILED to automatically transfer to the backup power supply (MCC:?-EOG).

The Unit Supervisor has directed you to restore power to Vital Bus 2 using the Manual Bypass Sw tch. In order to accomplish this manual transfer, the Manual Bypass Switch must be placed in the ___

position?

Normal Operation Alternate Source To Load Bypass (Standby)

Bypass (Isolate)

Answer E

ExplanationlJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA Category 000057 KIA#

AA1 01 KIA Importance 3.7 Exam Level RO Question Source:

New Question Cognitive Level:

Lower Fundamental qeferences provided to Candidate None Technical

References:

AOP-2.38.16 step 6e RNO Ijective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 41.7 145.5 145.6)

Incorrect. Normal Operation position is the position it was in when the static switch iailed to make the transfer.

Correct. iAW AOP~2.38.1B step 6e RNO.

Incorrect. This IS the correct manual transfer switch position for Unit 1 NOT Unit 2.

Incorrect. This IS the incorrect manual transfer switch position for Unit 1 KIA Statement Manual inverter swapping Level Of Difficulty: (1-5)

Loss of Vital AC lnst. Bus Ability to operate w d I or monitor the following as they apply to the Loss of Vital AC Instrument Bus:

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

13.

The plant is operating at 100% power with all systems in NSA.

. A loss of 125VDC Bus 2-1 has occurred.

Step 2 of AOP-2.39.1A. Loss Of 125VDC Bus 2-1 instructs the operating crew to control RCS temperature and pressure using the Steam Generator Atmospheric Steam Dump Control valves

[2SVS*PCVIOlA(B)(C)] OR the Residual Heat Release Valve [2SVS*HCV104].

Under these conditions, WHY are THESE valves used to control RCS temperature?

Because:

The condenser will NOT be available due to loss of all cooling tower pumps.

The condenser will NOT be available due to closure of all steam generator MSlVs Rod control will NOT be available due to an URGENT failure alarm Rod control will NOT be available due to a NON-URGENT failure alarm.

A.

B.

C.

D.

Answer E

ExplanationiJustification:

A.

E.

C.

D.

Incorrect. Loss of DC control power to 4KV breakers will not cause the breaker to trip; it wiil render the automatic trip Circuit inoperable Correcl. IAW AOP 2.39 1A step 2 and Automatic actions listed on page 1.

Incorrect. Urgent failure alarms will block all automatic and manual rod mofion. However, rod control power is not powered by this DC bus 4ND the reactor will be tripped by the closing of the MSlVs.

Incorrect. Non-Urgent failure alarms will NOT block rod motion However, Non-urgent alarms are generated from a loss of any 24VDC PO\\ er but the DC power is not provided by this DC bus AND the reactor will be tripped by the closing of the MSlVs.

'</A Sys #

KIA System KIA Category WAStatement 10058 Loss of DC Power Knowledge of the reasons for the following responses as they apply to the Loss of DC Power:

Actions contained in EOP for loss of C :

power KIA 37 AK3.02 KIA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Memory References provided to Candidate None Objective It:

Task ID#:

10CFR Part55 Content (CFR41.5.41.1Oi45.6I45.l)

Technical

References:

AOP 2.39.1A step 2 and Auto actions <In page

1.

Page 13 01 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

14.

The plant is operating at 100% power with all systems in NSA.

A Service WaterlNormal Intake Structure Loss has occurred.

Step 2 of AOP-2.30.1, Service Water/Normal Intake Structure Loss instructs the operating crew tc secure any liquid waste discharges IF service water header pressure cannot be restored above 3r! psig Under these conditions, WHY are liquid waste discharges secured?

Because:

The required liquid waste discharge dilution water flow cannot be assured The liquid waste discharge radiation monitor will be inoperable.

The Liquid Waste Effluent High Radiation Isolation Valve [2SGC-HCV100] will fail shut The Steam Generator Blowdown Test Tanks Pumps [2SGC-P26A, 2681 will trip A.

B.

C.

D.

Answer A

ExplanationIJustification:

A.

6.

C.

D.

Correct. IAW OM Fig. 31-1 and 25-4 Dilution water for liquid waste discharges is provided by the sewice water system.

Incorrect. The liquid waste discharge radiation monitor is not cooled by river water and will remain operable during loss of service water.

Incorrect. 2SGC-HCV100 does fail shut on loss of air. However, domestic water is manualiy aligned to cool the station air compressors. tt ?refore NO loss of air will occur.

lncnrrect. Steam Generator Blowdown Test Tanks Pumps do provide the driving force for the liquid waste discharges. However. these put IPS are not cooled by Service water and will remain operable.

~

WASys#

KIASystem KIA Category KIA Statement 100062 Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water:

JA #

AK3.03 WAlmportance 4.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

OM Fig. 31-1 and 25-4 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 41.4, 41.8 145.7 )

Lass nf Nuclear Service Water Guidance actions contained in EOP fc Loss of nuclear sewice water

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

15.

The plant is in Mode 3 with all systems in normal ;alignment for this mode

. The reactor trip breakers are OPEN.

A Loss of station instrument air occurs.

Station instrument air header pressure is 0 psig.

What impact will this loss of station instrument air have on charging and letdown?

Charging will (1) and letdown will (2)

A.

(1) isolate (2) isolate B.

(1) isolate (2) remain in service C.

(1) remain in service (2) isolate D.

(1) remain in service (2) remain in service Answer C

ExplanationlJustification:

A.

6.

C.

0.

incoriecl. Letdown will isolate WASys#

WASystem KIA Category WAStatement 000065 Loss of instrument Air

"!A #

AA2.08 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5) rleferences provided to Candidate None Technical

References:

AOP-2.34.1 attachment 2.34.1-1 Ch 7 <ail Objective zk Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Incorrect. Charging remains In Sewice.

Incorrect. Charging remains in service and letdown will isolate Correct IAW AOP-2 34.1 attachmenl 2.34.1-1 Ch 7 fail positions.

~

Ability to determine and interpret the following as they apply to the Loss of instrument Air:

Failure modes of air-operated equipme 11 iestion Source:

New Question Cognitive Level:

Higher Comprehension positions.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

16.

Following a reactor trip and safety injection. the crew is performing actions of E-0, Reactor Trip 01 Safety Injection.

The following conditions exist:

All SG pressures are 1000 psig and stable.

All SG NR levels are approximately 35% and stable.

AFW flow is 380 gpm and stable.

RCS pressure is -1000 psig, lowering slowly.

RCS temperature is 545"F, stable.

Auxiliary Building - 710 Area Radiation Monitor [2RMP-RQ203] is in HIGH alarm.

Auxiliary Building - 735 Area Radiation Monitor [2RMP-RQ204] is in HIGH alarm.

Auxiliary Building - 735 Area Radiation Monitor [2RMP-RQ205] is in HIGH alarm.

Containment pressure is 13.45 psia and stable.

PRT conditions are NORMAL.

CNMT sump level and radiation are NORMAL.

Which ONE (1) of the following procedures MUST be entered to mitigate this event?

A.

ES-1.1, SI Termination

9.

ECA-1.2, LOCA Outside Containment.

i.

D.

E-I, Loss Of Reactor Or Secondary Coolant ES-1.2, Post-LOCA Cooldown And Depressurization Answer E

ExplanationlJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement WlE04 LOCA Outside Incorrect. RCS pressure is dropping.

Correct. Per E-0 step 20.

Incorrect. All CNMT parameters are normal.

incorrect. Entry would be from E-1, which would not be used

-=_

Ability to determine and interpret the following as they apply to the (LOCA Outside Confainment)

KIA #

EA2.1 KIA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New References provided to Candidate None Technical

References:

EOP E-0 diagnostic steps Objective #:

Task I D #

10 CFR Part 55 Content:

(CFR: 43.5 i 45.13)

Facility conditions and selection of appwpriate procedures during abnormal and emerg?ncy operations.

Containment Question Cognitive Level:

Higher Comprehension Page 16 Of 100

Beaver Valley Unit 2 NRC Written Exam (2LOT6)

17.

The plant is operating at 100% power with all systems in NSA.

A large break LOCA occurs.

Reactor trip and safety injection actuation occur.

The crew is performing actions of E-I, Loss Of Reactor Or Secondary Coolant.

Cold leg recirculation capability cannot be verified and the crew transitions to ECA-1.1, Lcss Of Emergency Coolant Recirculation.

At step 18 of ECA-1.I.

the crew is instructed to stopktart charging pumps to establish MINIMUM SI flow to remove decay heat.

What is the reason for establishing MINIMUM SI flow in this procedurs step?

Prevent a potential ORANGE path for RCS integrity Prevent PRZR overfill and subsequent RCS overpressurization.

Delay SI accumulator injection and subsequent isolation Delay Refueling Water Storage Tank (RWST) depletion.

A.

B.

C.

D.

Answer D

ExplanationlJustification:

A.

6.

Incorrect. Potential Orange or Red paths on RCS integrity are prevented by limiting the RCS cooldown to 100 'Flhr in this procedure.

Incorrect. PRZR overfill and subsequent RCS overpressurization are the reasons for securing SI flow in ES-1.2 which would be appropriat ' for a SMALL break LOCA but NOT a concern far LARGE break LOCAs. For large break LOCAs the PRZR wlli not overtill and RVLlS is used lo water inventory indications.

Incorrect. A major objective of ECA-1.1 is to CID and depressurize to get the accumulators to inject their inventory.

Correct iAW ECA-1.1 step 18 bases.

C.

D.

-~

'4 Sys #

KIA System WA Category KIA Statement E l 1 Loss of Emergency Knowledge of the reasons for the following responses as they apply to the (Loss of Emergency Coolant Recirculation) emergency situations.

Manipulation of controls required to ob:ain desired operating results during abnorrm and Coolant Recirc.

WA #

EK3.3 KIA Importance 3.8 Exam Level RO Level Of Difficulty: (1-5)

Question Cognitive Level:

Higher Comprehension Question Source:

New References provided to Candidate None Technical

References:

ECA-1.1 step 18 bases Objective #:

Task ID#:

10CFRPartSSContent:

(CFR: 41.5141.10.45.6.45.13)

Page 17 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

18.

The plant is operating at 100% power with all systems in NSA.

A small break LOCA occurs inside containment.

The reactor trips and safety injection actuates.

4KV Emergency Bus 2DF is de-energized.

Quench Spray Pump 2QSS*P21A TRIPPED cannot be started.

Containment pressure is 20 psig and slowly rising.

AFW flow is 100 gpm to each SG.

SG NR levels are 25%.

" A

& " B SG NR levels are slowly dropping.

"C" SG NR level BEGINS rising in an uncontrolled manner.

The crew is performing E-I, Loss Of Reactor Or Secondary Coolant step 15 Verify Cold leg Recirculation Capability.

Cold leg recirculation capability CANNOT be verified.

Based on these plant conditions, what procedural transition is Required?

Transition into:

ECA-1.1, Loss Of Emergency Coolant Recirculation FR-H.I. Response To Loss Of Secondary Heat Sink E-3, Steam Generator Tube Rupture FR-Z.l, Response To High Containment Pressure A.

B.

C.

D.

Answer B

(planationlJustification:

A.

B.

C.

D.

KIA Sys #

K/A System KIA Category KIA Statement WIE05 Incorrect. Although the conditions have been met for ECA-1.1 entry, FR-H.1, FR-Z.1 and E-3 entry conditions are also present and have i higher priority.

Correct. FR-H.l red path entry conditions are present since AFW flow is only 300 gpm and NR levels in all SGs is less than the required '31%

adverse CNMT level.

Incorrect. Although the conditions have been met for E-3 entry based on LHP criteria. FR-H.1. and FR-Z.l entry conditions are also pres nt and have a higher priority.

Incorrect. Although the conditions have been met for FR-Z.l entry, FR-H.l entry conditions are also present and have a higher priority.

Loss of Secondary Heat Sink Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink)

KIA #

EA2.1 KIA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 i 45.13)

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Technical

References:

F-0.3 status tree and EOP users guide page 9 paragraph B.1 Page 18 of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

19.

The Plant is operating at 50% power BOL with all systems in NSA.

Control Bank D is at 175 steps.

Control Bank D Demand step counters are at 175 steps.

Control Rod Group Selector Switch is in the "MAN" position Turbine control is in "First Stage Out".

The following VALID control room alarms are received:

A4-4F NIS Power Range Comparator Deviation A4-4G NIS Power Range Neutron Flux Rate High A4-9F Rod At Bottom A4-3C Tavg Deviation from Tref A4-1 D Pressurizer Control Pressure Highilow A4-1 E Pressurizer Control Press Deviation High/low No operator actions have been taken.

Based on these conditions, what will be the status of the following parameters 5 minutes after the :?vent began?

RCS Tavg will be (1) than 558°F.

PRZR Pressure will be (2) than 2235 psig.

PRZR Backup Heaters will be Reactor power will be (4) 50%.

(3)

A.

1. lower

2. lower

3. energized

4. lower than

1. lower

2. lower

3. energized

4. equal to C.

1. higher

2. higher

3. de-energized

4. lower than D.

1. higher

2. higher

3. de-energized

4. equal to Answer A

ExplanationlJustification:

A.

Correct. All of the alarms iisted will alarm for either high or low conditions EXCEPT A4-9F Rod At Bottom. Since A4-9F has energized and IS valid, the candidate will need to identify there is a dropped rod event in progress. This also will eliminate higher Tavg and PRZR pressure from consideration. With the turbine In "first stage our reactor power will be lower since the governor valves will not reposition to adjust for the lower steam pressures. Reactor power would be the same if the turbine was in "First stage in" BOL was selected to provide the most definite ctlmges to the listed parameters.

Incorrect. With the turbine in "first stage our reactor power will be lower since the governor valves will not reposition to adjust for the iowcr steam pressures Reactor power would be the same if the turbine was in "First stage In".

Incorrect. Tavg and PRZR pressure will be lower. The PRZR BIU heaters will be energized.

incorrect. Tavg and PRZR pressure will be lower. The PRZR BiU heaters will be energized. Reactor power will be lower.

6.

C.

D.

WA Sys #

WA System WACategory KIAStatement 000005 InoperableISluck Control NIA Ability to verify that the alarms are con,;istent Rod with the plant conditlons.

WA #

2.4.46 WA Importance 4.2 Exam Level R 3 Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

AOP-2.1.8 symploms (ran on Unit 2 simulator to confirm all parameters)

Objective #:

Task ID#:

10CFR Part55 Content:

(CFR: 41.10143.5l45.3i45.12)

~

~

Page 19 Of io0

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

20.

The plant is in Mode 2 with a reactor startup is in progress. All systems are in normal alignment fo, this condition.

The reactor trip breakers are closed with the shutdown banks withdrawn.

Control rod withdrawal is in progress.

Two control bank A rods fail to move when required, and become misaligned by 15 steps BOTH Source Range detectors SIMULTANEOUSLY become inoperable.

Reactor power is 1 X I O4 CPS and stable.

What are ALL of the IMMEDIATELY Required Technical Specification actions?

1. Suspend operations involving positive reactivity additions.

2. Open the reactor trip breakers.

3. Initiate action to restore one source range neutron flux monitor to OPERABLE status.

4. Verify SDM is within the limits specified in the COLR.

A.

1 and4 B.

1 and2 C.

2and3 D.

3and4 Answer B

ExplanationlJustification:

A.

B.

J.

Incorrect. #1 IS correct. #4 IS the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirement forthe misaligned rods; it IS NOT an immediate requirement.

Correct. IAW TS 3.3.1 conditions H and I.

Incorrect. #2 is correct. #3 IS the immediate requirement for both source ranges inoperable in Mode 6.

Incorrect. #3 is the immediate requirement for both source ranges inoperable in Mode 6 and #4 is the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirement for the misalign',d rods; 11 1s NOT an immediate requirement.

KIA Sys #

KIA System WA Category KIA Statement 000032 Loss of Source Range NI NIA Anon ecge 01 ess : n m 21 iq..a io OI. n:...r Tecnri cd S w c f cat on aci 0 7 siaiem nis lor s,siems W A #

2.2 39 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

TS 3.3.1 Table 3.3.1-1 and condition: Hand I.

Objective #:

Task ID#:

10 CFRPart55Content:

(CFR: 41.7141.10143.2145.13)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

21.

The plant is operating at 100% power with all systems in NSA.

A 50 gpm Steam Generator Tube leak develops.

The crew enters AOP 2.6.4, Steam Generator Tube Leakage.

A controlled shutdown to Mode 3 has been completed.

It has been determined that the leaking Steam Generator shall be cooled and depressurized using the "Backfill" method.

Which of the below listed attributes are advantages to using the "Backfill" method over other rnetho'ds?

1. Facilitates processing of contaminated primary coolant,

2. Minimizes Radiological releases.

3. Fastest means to cool the leaking Steam Generator.

4. Minimizes boron dilution of the primary coolant.

A.

1 and2 B.

1 and4 C.

2and3 D.

3and4 Answer A

ExplanatianlJustification:

A.

Q.

0.

KIA Sys #

KIA System KIA Category KIA Statement 000037 Steam Generator Tube Correct. IAW E-3 step 43 background (AOP does not have a background document). The AOP used the EOP background to develop attar iments to address cooling of the leaking SG).

Incorrect Boron dilution wiil NOT be limited by this method.

Incorrect. The fastest means to cooldown the SG is the steam dump method.

Incorrect. These are advantages of the biowdown method,

~

Knowledge of the reasons for the followiog responses as they apply to the Steam Generator Tube Leak:

Use of "feed and "bleed" process.

Leak K/A #

AK3.04 WAlmportance 2.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

E-3 step 43 background Objective #:

Task ID#:

10CFR Part 55Content:

(CFR41.5.41.1OI45.6I45.13) page21 O f 1 0 0

22 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

Which ONE (1) of the following constitutes a loss 'of an OPERABLE containment?

While in...

MODE 5, it is discovered that the Phase 'B' isolation valve for CCP to the RCPs will NOT CLOSE.

MODE 4, a review of Integrated Leak Rate test results show that leakage is NOT WITHIN LIMITS.

MODE 3, it is discovered that Containment Atmosphere Purge Makeup valve will NOT OPEN.

MODE 6, it is discovered that one of the Emergency Airlock (EAL) doors will NOT CLOSE.

Answer B

ExplanationlJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 000069 Loss of CTMT Integrity Ability to determine and interpret the foilowing as they (WIE14) apply to the Loss of Containment Integrity:

KIA #

AA2.01 KIA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

TS 3.6.1 and bases.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Incorrect. An OPERABLE Containment is not required in Mode 5.

Correct. IAW Technical Specification 3.6.1 and its bases. (Operable coritainment equates to CNMT integrity with the new ITS)

Incorrect. Purge Makeup Valve should not normally be open in Mode 3 and wouid not be a loss of OPERABLE containment if it does not (,PEN Incorrect, In Mode 6. 1 airlock door may remain open.

d p

Loss of containment integrity Page 22 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

23.

The plant is operating at 100% power with all systems in NSA.

A small break LOCA occurs inside containment.

The crew performs the actions of E-0, Reactor Trip Or Safety Injection and transitions intc, E-1, Loss Of Reactor Or Secondary Coolant.

Consider the below listed criteria:

What are ALL of the criteria that MUST be met before a transition to ES-1.1, SI Termination can bi?

made?

1. One emergency diesel must be operating

2. One train of CIA must exist

3. Cold leg recirculation capability must exist

4. The RCS must be subcooled

5. A secondary heat sink must exist

6. RCS pressure must be stable or rising

7. PRZR level must be indicating on span

8. A RCP must be operating A.

8.

C.

1, 3, 4, 5. & 8 ONLY 1, 2, 4, 6, & 8 ONLY 2, 3, 5, & 7 ONLY 4, 5, 6, & 7 ONLY Answer D

ExplanationIJuslification:

A.

Incorrect. IAW the background document for ES-1.I the only 4 criteha mat must be met are items 4, 5. 6, and 7. The other 4 items are n i c ~

to have during recoveiy effolts, but they are not required for SI termination. Additionally, cold ieg recirculation capability isn't even checked b.!fore a transition into ES-1.1 is allowed. The 4 correct criteria (RCS Subcooling. heat sink, RCS pressure rising, and indicated PRZR level) combi e to indicate that the RCS is in a safe state with adequate core cooling.

Incorrect. IAW the background document for ES-1.1 the only 4 criteria that must he met are items 4, 5. 6. and 7. The other 4 items are nice to have during recovery effoits, but they are not required for SI termination. Additionally, cold leg recirculation capability isn't even checked b-:fore a transition into ES-1.1 is allowed. The 4 correct criteria (RCS Subcooling. heat sink, RCS pressure rising, and indicated PRZR level) cornbi ie lo indicate that the RCS IS in a safe state with adequate core cooling.

lncoriect IAW the background document for ES-1.1 the only 4 criteria that must be met are items 4, 5, 6. and 7. The other 4 items are nict to have during recovery efforts but they are not required lor SI termination. Additionally, cold leg recirculation capability isn't even checked b fore a transition into ES-1.1 is allowed, The 4 correct criteria (RCS Subcooling heat sink, RCS pressure rising, and indicated PRZR level) comhi e to indicate that the RCS is in a safe state with adequate core cooling.

Correct. IAW the background document for ES-1.1 the only 4 criteria that must be met are items 4, 5. 6. and 7. These 4 Criteria (RCS Sub, 2oling.

heat sink, RCS pressure rising. and indicated PRZR level) combine to indicate that the RCS is in a safe state with adequate core cooling.

B.

C.

D.

KIA Sys #

KIA System KIA Category WAStatement WIE02 SI Termination Knowledge 01 the interrelations between the (SI Termination) and the following:

Facility's heat removal sysiems. includ ng primary cooiant, emergency coolant, the decay heat removal systems, and reia ons between the proper operation of these systems to the operation of the facility.

WA U EK2.2 KIA Importance 3 5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New References provided to Candidate None Technical

References:

ES-1 1 background page 1 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.7)

Question Cognitive Level:

Higher Comprehension Page 23 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

24.

The plant has been operating at 100% power with all systems in NSA for the past 100 days.

An inadvertent Turbine trip occurs coincident with a loss of offsite power.

IAW the Plant Technical Specifications, which of the following components MUST operate to prevvnt Steam Generator Overpressure (> 110% of design pressure)?

1. Atm Stm Dump Control Valves

2. Steam Generator Safety Valves

3. Residual Heat Release Valve

4. Diesel Driven Air Compressor A.

1 &2ONLY

0.

2ONLY C.

3&4ONLY D.

4ONLY Answer E3 ExplanationlJustification:

A.

Incorrect. Aim Stm Dump Control Valves are for decay heat removal. NOT overpressure protection. They will limit SG pressure by virtue c removing decay heat, but they are NOT required by UFSAR. Since the Atm Stm Dump Control Valves remove decay heat arid thus limit p essure there is a Common misconception that they are required to prevent SG overpressure.

Correct IAW Tech Spec bases 3.7.1 page 1 1" paragraph and page 2 lit paragraph.

incorrect. The residual heat release valve is designed for decay heat only and no credit is taken for limiting SG overpressure. The Diesel I riven Air Compressor IS not required, but nice to have for plant control. Many inain steam system valves are kept open by air, but none are reqt red to prevent SG overpressure.

Incorrect The Diesel Driven Air Compressor is not required, but nice to have for plant control. Many main steam system valves are kept o!,en by air, but none are required to prevent SG overpressure.

6.

C.

0.

~

KIA Sys #

KIA System KIA Category KIA Statement WiE13 Steam Generator Over-Components and functions of control e>?d pressure Generator Overpressure) and the following:

safety systems. including instiumenta1.m signals, interlocks. failure modes, and automatic and manual features.

Knowledge of the interrelations between the (Steam KIA #

EK2.1 KIA Importance 3.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New References orovided to Candidate None Objective #:

Task ID#:

Question Cognitive Level:

Lower Fundamental Technical

References:

Tech SDec bases 3.7.1 oaoe 1 1 " ~ a r i XaDh

,~

and page 2 I" paragraph.

10 CFR Part 55 Content:

(CFR: 41.7 i 45.7)

Page 24 of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

25.

The plant is operating at 100% power with all systems in NSA.

A large break LOCA occurs inside containment.

The crew is implementing the actions of E:-l, Loss Of Reactor Or Secondary Coolant.

The STA then reports the following CSF status:

YELLOW-Core Cooling-FR-C.3. Response To Saturated Core Cooling (Based on core ex t temperatures less than 729°F and RVLIS greater than 40% full range).

ORANGE-Containment-FR-Z.2, Response to Containment Flooding (Based on a contain,nent sump level of 189 inches).

YELLOW-Containment-FR-Z.3, Response To High Containment Radiation Level (Based on a containment radiation level of 76 Rihr).

The crew transitions to FR-Z.2, Response to Containment Flooding and completes all of the actions of this procedure. The STA then reports THE SAME CSF status that was reported earlier.

What procedural transition, if any, is now Required?

A.

B.

C.

D.

Return to Step 1 of FR-Z.2, Response to Containment Flooding.

Return to step in effect of E-I, Loss Of Reactor Or Secondary Coolant Go to FR-C.3, Response To Saturated Core Cooling Go to FR-Z.3, Response To High Containment Radiation Level.

nswer B

ExplanationlJUsiification:

A.

6.

C.

D.

incorrect. Plausible since normally the EOP usage rules do not allow a transition out of a red or orange path procedure until the symptoms bave been corrected. However, FR-Z.2 is an exception and the crew is directed to return to step and procedure In effect.

Correct. IAW F0.5 bases for step 4 page 7 knowledge paragraph.

Incorrect. Plausible. Core cooling E a higher priority status tree terminus than either containment radiation or flooding. However, the termlr JS IS only yellow. and transition to this procedure is optionai not required. The question specifically asks for required transition.

incorrect. Plausible. Containment radiation is a higher priority status tree terminus than returning to E-I. However. the terminus is only yell,.w. and transition to this procedure is optional not required. The question specifically asks for required transition.

~

KIA Sys #

KIA System KIA Category KIA Statement WIE15 Containment Flooding Knowledge of the operational implications of the following concepts as they apply to the (Containment Flooding).

Normal, abnormal and emergency ope ating procedures associated with (Containment Flooding).

KIA #

EKl.2 KIA Importance 2 7 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

F0.5 bases for step 4 page 7 knowledp?

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.8 i 41 10, 45.3) paragraph.

Page 25 1 io0

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

26.

The plant is operating at 100% power with all systems in NSA.

A small break LOCA occurs inside containment.

All systems function as designed.

The crew is implementing the actions of ES-I 2, Post LOCA Cooldown and Depressurizat:on.

All RCPs have been secured.

Both trains of RVLIS are 00s.

SI, CIA, and CIB have all been reset.

While depressurizing the RCS to minimize subcooling in step 24 of ES-1.2. the following I: ant conditions are observed:

+

PRZR level is 45% and rapidly rising.

RWST level is 395 inches and slowly dropping.

CNMT pressure is 15 psig and slowly dropping What procedural transition, if any, is now Required?

A.

B.

C.

D.

Continue with step 24 of ES-1.2, Post LOCA Cooldown and Depressurization Go to ES-0.4, Natural Circulation Cooldown With Steam Void In Vessel (Without RVLIS).

Go to ES-1.3. Transfer To Cold Leg Recirculation Go to FR-Z.l, Response To High Containment Pressure Answer C

ExpianationlJustification:

A.

Incorrect. Plausible if the candidate does not recognize the need to transition to ES-1.3 based on RWST level below 400 inches.

incorrect. Plausible since RCPs are off and a natural circulation cooidown is in progress AND PRZR level rapidly rising is indicative of bubt e formation in the upper head region. However, the transition to ES-0 4 can only be made from ES-0.2 where there is no other accident in prc gress.

Correct IAW ES-1.2 LHP item 4. ES-1.2 bases page 1 4Ih paragraph.

Incorrect Plausible since CNMT pressure is above 11 psig However, this IS incorrect because both QS pumps are operating and thls is on i a yellow path procedure.

C.

D.

=_

KIASys#

KIASystem KIA Category KIA Statement WE03 LOCA Cooldown -

Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization)

Facility conditions and selection of appI,>priate procedures during abnormal and emerpency operations.

Depress.

KIA #

EA2 1 KIA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Anaiysis References provided to Candidate None Objective 8:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Technical

References:

ES-1.2 IHP item 4. ES-1.2 bases page I 4Ih paragraph Page 26 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

27.

The plant has been operating at 100% power with all systems in NSA for the past 100 days.

An inadvertent reactor trip occurs coincident with a loss of offsite power.

All systems function as designed.

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

RCS temperature is 350°F and stable.

RCS Subcooling is 200°F and stable.

RCS Pressure 1200 psig and stable.

RCS cooldown rate is 20"F/hr and stable.

Alarm A I 1-5G CRDM Shroud Fan Auto-StarVAuto-Stop is received. ALL CRDM shroud fans have tripped and cannot be restarted.

What impact will the loss of these CRDM Shroud Fans have on the continued performance of ES4.2, Natural Circulation Cooldown?

Further RCS cooldown (below 350°F) cannot continue UNTIL a suitable RX vessel head soak has 3een performed.

Further RCS depressurization (below 1200 psig) cannot continue UNTIL a suitable RX vessel heac! soak has been performed.

Immediately INCREASE RCS pressure 100 psig to RAISE RCS subcooling Immediately DECREASE RCS pressure 100 psig to LOWER RCS subcooling A.

B.

C.

D.

Answer B

ExpianationlJustification:

IncorieCt. The restriction to perform a head soak only applies dnen the RCS 'is below 350'F. However. cooldowns below 350-F are Stlll alirlwed when CRDM fans are unavailable.

Correct. IAW ES-0.2 step 13 and background.

Incorrect. Raising pressure 100 psig is a technique employed by ES-0.4 natural circulation procedure when the head void growth becomes too large.

Incorrect. Minimizing Subcooling is a technique employed when RX vessel stresses are the concern but NOT when RX vessel head voids : re the concern. Decreasing pressure may actually cause a void to form

6.

C.

0.

~_=

KIA Sys #

KIA System KIA Category KIA Statement WIE09 Natural Circ Knowledge of the operationai implications of the following Annunciators and conditions indicating concepts as they apply to the (Natural Circulatlon Operations) the (Natural Circulation Operations).

signals, and remedial actions assoc!atetJ with KIA #

EKl 3 KIA Importance 3.3 Exam Level 30 Level Of Difficulty: (1-5)

Question Source:

New References provided to Candidate None Technical

References:

ES-0.2 step 13 and background.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.8 / 41.10.45.3)

Question Cognitive Level:

Higher comprehension Page 27 01 130

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

28 A.

B.

C.

D.

The plant is operating at 40% power with all systems in normal alignment for this power level.

. "B" RCP breaker OPENS due to a mechanical failure.

What impact will this OPEN breaker have on the reactor protection system (RPS)?

A reactor trip signal will...

NOT be generated. At this power level it takes 213 RCP breakers open to generate a reactor trip si(jnal.

NOT be generated. At this power level it takes 2/3 RCS Loops Low Flow generate a reactor trip sicnal.

Be generated by the single open RCP breaker.

Be generated by the single RCS loop flow low.

Answer D

ExplanationlJustification:

A.

E.

C.

D.

KIA Sys #

KIA System WA Category KIA Statement 003 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following:

KIA #

K3.04 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

UFSAR logic Figure 7.3-10 Incorrect. It is true that it takes 213 breakers open to generate a trip signal at this power level. However, the single loop flow low wdl genen'e a trip signal.

incorrect. It only takes a single locp flaw low to generate a trip signal.

Incorrect. It IS true that a trip signal will be generated. However, it is not generated from the RCP breaker opening.

Correct IAW UFSAR logic Figure 7.3-10

~

Reactor Coolant Pump RPS

'>jective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.6)

Page 28 01 130

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

29.

A.

B.

C.

D.

In the CVCS.

1. Which ONE (1) of the below listed components is designed to prevent flashing at the downstream side of the letdown orifices?

2. How is this accomplished?

1. Letdown Orifice lsol Vlvs [2CHS*AOV200A(B)(C)]

2. Close on high temperature downstream of the orifices

1. Non-Regen HX Disch Press Control Vlv [2CHS*PCV145]

2. Maintains pressure downstream of the orifices above saturation pressure

1. Non-Regen HX Disch Diverting Vlv [2CHS*TCV143]

2. Diverts letdown flow to the degasifiers on high temperature

1. Non-Regen HX Temp Control Vlv [2CCP*TCV144]

2. Maintains letdown temperature downstream of the orifices below saturation temperature.

Answer B

ExplanationlJustification:

A.

6.

C.

D.

Incorrect. These valves do not have a high temperature isolation signal, although this would prevent flashing. They will isolate on CIA.

Correct. IAW 2QM-7.1.C page 8 1" paragraph Incorrect. This valve is downstream of the Non-regen HX and diverts water away from the demineralizers to protect them from high tempel jture It does NOT divert water to the degasifiers.

incorrect. This valve does cool the letdown water. BUT it 15 too far downstream to prevent flashing at the letdown orifices.

'ASysU WASystem WACategory WA Statement J4 Chemical and Volume Knowledae of CVCS d e s m featurelsl andlor interiock(sl TemDeraturelDreSSure Control in letdohn line:

Control which provide for the following:

KIA It K4.11 KIA Importance 3.1 Exam Level RO Question Source:

New References provided to Candidate None Objective #:

Task ID#:

prevent boiling, lifling reliefs. hydraulic ';hock, piping damage, and burst Level Of Difficulty: (1-5)

Question Cognitive Level:

Lower Fundamental Technical

References:

10 CFR Part 55 Content:

20M-7.1.C page 8 1" paragraph (CFR: 41.7)

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

30.

The plant is in Mode 4 with RCS temperature at 210°F.

0....

RHR Pump 2RHS*P21A is on clearance.

Train " 6 of RHS is in service and being used for an RCS cooldown at 25"F/hr.

All Train "8" RHS components are arranged in their normal alignment for plant cooldown.

RHR HX Flow Control Valve [2RHS*HCV758B] is 30% OPEN.

RHR HX Bypass Valve [2RHS*FCV605B] is 50% OPEN.

As a result of poor Foreign Material Exclusion (FME) practices, a rubber Anti-C boot becomes lodc:ed in the tube side of the "6" RHS Heat Exchanger. The boot BLOCKS 25% of the tubes in the heat exchanger.

IF the RCS cooldown is to CONTINUE at 25"F/hr, the reactor operator will be required to A.

Manually CLOSE RHR HX Bypass Valve [2RHS*FCV605B] and allow RHR HX Flow Control Valve (2RHS*HCV758B] to automatically throttle OPEN to maintain total RHS system flow.

Manually OPEN RHR HX Bypass Valve [2RHS*FCV605B] and allow RHR HX Flow Control Valve

[2RHS'HCV758B] to automatically throttle CLOSED to maintain total RHS system flow.

Manually CLOSE RHR HX Flow Control Valve [2RHS*HCV758B] and allow RHR HX Bypass Valve

[2RHS*FCV605B] to automatically throttle OPEN to maintain total RHS system flow.

Manually OPEN RHR HX Flow Control Valve [2RHS*HCV758B] and allow RHR HX Bypass Valve

[2RHS*FCV605B] to automatically throttle CLOSED to maintain total RHS system flow.

B.

C.

D.

Answer D

planationlJustification:

A.

8.

C.

D.

incorrect Manualiy closing 2RHSFCV6058 will force more water through the HX. However, ZRHS'FVC758B does not have automatic ilo:?

control and will NOT automatically throttle open to maintain RHS system flow.

incorrect. Manually opening 2RHSFCV605B will divert water away from the HX which would cause a heatup. Also. ZRHS'FVC7588 does,701 have automatic flow control and wiii NOT automatically throttle open to maintain RHS system flow.

Incorrect. These actions will slow the RCS cooldown.

Correct. IAW OM figure 10-1. ZRHS'FCV605B has the automatic flaw control. and ZRHS'FCV758B is a manually adjusted valve to Control flow through the RHS HX.

~~

KIA Sys #

KIA System KIA Category KIA Statement 005 Knowledge of tne effect oia loss or malfunction On the following wiii have on the RHRS:

KIA #

K6.03 KIA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

OM iigure 10.1 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.7)

Residual Heat Removal RHR heat exchanger PaOe 30 01 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

31.

The plant is operating at 100% power with all systems in NSA.

. Low Head SI Pump 2SIS"P21A becomes inoperable due to a bearing failure on the pump In the event of a Large break LOCA, how will this failure impact Train " A ECCS performance?

BEFORE transfer to cold leg recirculation there will be (1)

AND AFTER transfer to cold leg recirculation there will be (2)

(Assume all other systems function as designed during the Large break LOCA)

A.

1. NO Low Head SI flow.

2. Low Head SI flow available via Recirc spray pump 2RSS*P21C AND High Head SI flow will be available.

B.

1. Low Head SI flow available via Recirc spray pump 2RSS*P21C

2. NO Low Head SI flow BUT High Head SI flow will be available.

1. NO Law Head SI flaw.

2. NO Low Head SI flow AND NO High Head SI flow.

1. Low Head SI flow available via Recirc spray pump 2RSS*P21C.

2. Low Head SI flow will be available via Recirc spray pump 2RSS*P21C AND High Head SI flow H 'I1 be C.

D.

available.

.iswer A

cxplanationIJustification:

A.

B.

C.

D.

WASys #

WASystem WA Category KIA Statement 006 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

KIA #

K6.03 KIA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

EOP Attachment A-0.7 and ES-1.3 anc VOND Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.7)

Correct. IAW EOP Attachment A-0.7 and ES-1.3 and VOND 11-1 & 13-1 incorrect. Before transfer to cold leg Recirc there will be NO LHSl flow; after transfer to cold Recirc there will be Row via 2RSSP21C.

incorrect After transfer to cold leg Recirc there will be LHSl flow via 2RSSP21C and it will provide suction to the HHSl pump.

Incorrect Before transfer to cold leg Recirc there will be NO LHSi Row.

Emergency Core Cooling Safety Injection Pumps 11-1 & 13-1

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

32.

The plant is operating at 100% power with all systems in NSA.

Charging/High Head Safety lnj Pump 2CHS*P21A is running.

ChargingMgh Head Safety lnj Pump 2CHS*P21B is in standby.

CharginglHigh Head Safety Inj Pump 2CHS*P21C is NOT racked onto any bus How will CharginglHigh Head Safety lnj Pump 2CHS*P21A SUCTION and DISCHARGE be impacted by the receipt of a SI signal?

Suction will realign to the (1)

Discharge will realign for (2)

AND A.

1. VCT

2. Cold leg injection B.

1. RWST

2. Cold leg injection C.

1. VCT

2. Hot leg injection D. 1. RWST

2. Hot leg injection Answer B

planationlJustification:

A.

B.

C.

D.

KIA Sys #

KIA System KIA Category KIAStatement 006 KIA tt A4 05 KIA Importance 3.9 Exam Level RO Level Of Difficulty. (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

20M-7.1.D page 33, 35; 2OM-11.1.D p:ye 4 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.5 to 45.8)

Incorrect. Suction will re-align to the RWST Correct IAW 20M-7.1.D page 33, 35; 20M-11.1.D page 4 incorrect. Suction wiil re-align to the RWST and discharge will re-align to the cold leg injection flowpath Incorrect. Discharge will re-align to the cold ley injection flowpath.

~

Emergency Core Cooling Ability to manually operate andlor monitor in the control room:

recirculation Transfer of ECCS flowpaths prior to Page 32 of IC0

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

33. The plant is operating at 100% power with all systems in NSA.

. The Pressurizer Relief Tank (PRT) level transmitter malfunctions and the operators inadvertently overfill the PRT to 95% level.

How will containment parameters be affected by this overfill condition IF a PRZR PORV opens an(.

continuously discharges to the PRT?

Containment temperatures will ___ ( 1 )

will (2) and containment radiation A.

1. Remain constant

2. Remain constant B.

1. Remain constant

2. Rise C.

1. Rise

2. Remain constant D.

1. Rise

2. Rise Answer D

ExplanationIJustification:

A.

incoriecl. The higher water level is not enough to keep a continuous discharge from blowing the PRT rupture disc and discharging RCS to Containment. This will result in higher containment temperatures. Although the higher water level is capable of scrubbing some iodine from 'he RCS, it wiil not prevent containment radiation level: from increasing.

Incorrect. The higher water level is not enough to keep a continuous discharge from blowing the PRT rupture disc and discharging RCS to containment. This will result in higher containment temperatures. Containment radiation will rise.

Incorrect The higher water level is not enough to keep a continuous discharge from blowing the PRT rupture disc and discharging RCS to Containment. This will result in higher containment temperatures. Although the higherwater level IS capable of scrubbing some iodine from he RCS, it 'wll not prevent containment radiation level; from increasing.

Correct. IAW 20M-6.1 C page 34 3d paragraph. The PRT is not designed for continuous discharge. Candidate must understand that even vith the higher water level. the continuous discharge of a PORV will be a LOCA as far as the containment parameters are concerned.

C.

D.

~ -

KIA Sys #

KIA System WA Category WASiatement 007 Pressurizer RelieflQuench Knowledge of the effect that a loss or malfunction of the Containment PRTS will have on the following:

KIA ff K3.01 WA Importance 3.3 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

20M-6.1.C page 34 3" paragraph Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.6)

Tank

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

34.

The plant is operating at 75% power BOL with all :systems in normal alignment for this power level Rod control is in Automatic All RCS T,,,

indications are Matched with T,,.

VCT level is 45% and stable.

With NO INITIAL change in turbine load, control rods begin to slowly step INWARD Which ONE (1) of the below listed failures will cause this inward rod motion?

A.

B.

C.

D.

Answer D

ExplanationlJustification:

A.

B.

C.

The Loop " A Tc transmitter slowly failing HIGH A Loop " A TH transmitter slowly failing HIGH.

Boric Acid to Boric Acid Blender. [2CHS*FCV113A] failing OPEN, Non-Regen HX Disch Temp Control Vlv [2CCP*TCV144] failing OPEN lncorrecl. If the Tc transmitter fails high, lhen Tavg for thal loop will rise. However, this will NOT cause rods to move Since the Tavg signal '0 rod control IS median selected.

Incorrect. If a Th transmitter fails high, tnen Tavg for that loop will rise. However, this wiii NOT cause rods to move since the Tavg slgnai ti rod control IS median selected.

Incorrect. This would appear to be a potential boratlon path. However, with blender setup in NSA. FCVl13B and FCV114B are both CLO! 3 isolating any potential flowpath. Additionaliy, at Unit 2 the boric acid transfer pump is NOT running untii a Makeup signal is generated. Wit VCT level at 45% there would be NO makeup signal available.

Correct. IAW VOND 15-5 grid F-4 and LP GPF.C4 page 30. Colder water will allow the demineralizers to absorb more boron.

D.

~-

'A Sys #

KIA System KIA Category KIA Statement

,'I8 Component Cooling Water NIA Ability to diagnose and recognize treni's in an accurate and timely manner utilizing tt ?

appropriate control room reference mc erial.

KIA #

2.4.47 KIA Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Objective U Task ID#:

10 CFR Part 55 Content:

(CFR: 41 10 143.5 145.12)

Technical

References:

VOND 15-5 grid F-4 and LP GPF.C4 I: ige 30.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

35.

The plant is in Mode 2 preparing for a turbine startup all systems in normal alignment for this mode.

+

Reactor power is 3% and stable.

The PRZR Master Pressure Controller output is at 42% demand signal.

PRZR pressure is 2235 psig and stable.

Both PRZR spray valve controllers are in AUTOMATIC.

PRZR Spray Valve 2RCS*PCV455A is 2G% OPEN.

PRZR Spray Valve 2RCS*PCV455B is CLOSED.

The PRZR control heater is in AUTO (Red Target).

The Steam Dump Control Mode Selector switch is in the Stm Press position.

The Main Stm Manifold Press Controller [2MSS-PK464] is in AUTOMATIC with a setpoint of 1000 psig.

RCS temperature is 547°F and stable.

Main Stm Manifold Stm Press [2MSS-PT464] transmitter THEN fails LOW.

How will the PRZR Pressure control system INITIALLY respond to this failure?

(Assume NO operator actions)

PRZR Master Pressure Controller output will (1) and cause PRZR Spray Valve 2RCSPCV455A to (2)

A.

(1) RISE above 42%

(2) OPEN more than 20%

6. (1) RISE above 42%

(2) fully CLOSE

i. (1) DROP below 42%

(2) OPEN more than 20%

D.

(1) DROP below 42%

(2) fully CLOSE Answer A

ExplanationIJustification:

A.

Correct IAW 20M-6.4.IFattachment 2 and 20M-21.5.A.12. A low failure of 2MSS-PT464 will cause the steam dumps to close as the contrder is trying to maintain 1000 psig and the input is now zero. Closing the steam dumps will cause an RCS heatup which will raise RCS pressure. With rising RCS pressure the PRZR master controller output will rise and spray valve 455A will OPEN farther to drop pressure.

incorrect. Spray valve 455A will OPEN.

Incorrect. Master controller output will rise.

Incorrect. Master controller output will rise and Spray valve 455A will OPEN.

8.

C.

0.

KIA Sys #

KIA System KIA Category KIAStatement 010

~

~

Pressurizer Pressure Control Ability to predict andlor monitor changes in parameters [to prevent exceeding design limits) associated with operating the PZR PCS controls including:

KIA #

A1.06 KIA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

20M-6.4.IFattachment 2 and 20M-21.:.A.12.

Objective #:

Task ID#

10 CFR Part 55 Content:

(CFR: 41.5 145.5)

RCS heatup and cooldown effect on pl :Ssure

36 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

The plant is operating at 100% power with all systems in NSA.

. PRZR Channel 1 Press 2RCS-PT455 has failed HIGH.

The control room crew has tripped all associated bistables IAW 20M-6.4.IF, Instrument Fcilure.

PRZR Control Pressure [2RCS-PT445] THEN fails HIGH.

What will be the INITIAL plant response to this additional failure?

PRZR Spray Valve 2RCS*PCV455A & 2RCS*PCV455B will OPEN.

PRZR PORV 2RCS-PCV455C WIII OPEN.

PRZR PORVs 2RCS-PCV455D & 2RCS-PCV456 will OPEN.

High PRZR Pressure Reactor Trip will ACTUATE.

Answer C

ExplanationlJustification:

A.

B.

C.

D.

KIASys #

KIASystem KIA Category KIA Statement 010 Pressurizer Pressure Ability to monitor automatic operation of the PZR PCS.

PZR pressure KIA #

A3 02 KIA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher comprehension eferences provided to Candidate None Technical

References:

20M-6.4.IF attachment 2.

Incorrect. This would be the INITIAL response if 2RCS-PT444 failed High.

Incorrect. This would be the next response if 2RCS-PT444 failed High.

Correct. IAW 20M-6.4.1F attachment 2.

Incorrect. Failures are one control channel and one protection channel. therefore NO reactor trip Control mcluding:

ojective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.1 145.5)

Page 36 01 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

37 B.

C.

D.

Which ONE (1) of the following power changes will change the "RCS Loop Low Flow" Automatic Reactor Trip Logic from a 213 coincidence to a 113 coincidence?

Raising power on 214 Power range channels from 8% to 12%.

Raising power on 214 Power range channels from 28% to 32%.

Lowering power on 214 Power range channels from 12% to 8%.

Lowering power on 214 Power range channels from 32% to 28%.

Answer B

ExplanationIJuStification:

A.

B.

C.

0.

KIA Sys II KIA System KIA Category KIAStatement 012 Reactor Protection Knowledge of RPS design feature(s) and/or in!erlock(s) which provide for !he following:

KIA II K4.06 KIA Importance 3.2 Exam Level RO Level Of Difficulty: (1.5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

UFSAR Figs 7.3-9 and 7.3-10.

Objective U:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7)

Incorrect. This power change would Arm the trip. Below P-10 the low flow trip is NOT active.

Correct. IAW UFSAR Figs 7.3-9 and 7.3-10.

Incorrect. This wouid remove all RCS low Row trips.

Incorrect This changes the logic from 113 to 213. ________

~~

Automatic or manual enabieldisabie of ?PS trips Page 37 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

38.

The plant is operating in Mode 1 with all systems in NSA.

A secondary calorimetric has just been completed. The calorimetric indicates that reactor power is 99.6%. All power range channels are OPERABLE.

Power Range indications are as follows:

9 N41 -99.8%

N42-99.5%

N43 - 99.0%

. N44 - 100.0%

What Power Range gain adjustments are Required?

Lower ONLY N44 indicated power to (99.6%

Lower N41 AND N44 indicated power to 599.6%

Raise N42 AND N43 indicated power to 299.6%

Raise N41, N42 AND N43 indicated power to 100%

A.

B.

C.

D.

Answer C

ExplanationlJustification:

A.

6.

C.

D.

VIA Sys #

KIA System KIA Category KIA Statement incorrect. Adiustments not required for Nis with indicated power above actual power.

Incorrect. Adjustments not required for Nis with indicated power above actual power.

Correct IAW 20M-54.4.Cl page 14 2"' bullet.

Incorrect. Although this would be conservative. it is NOT required to raise N41.

2 Reactor Protection Ability to predict andlor monitor Changes in parameters Trip setpoint adjustment (to prevent exoeeding design limits) associated with operating the RPS controls including:

KIA #

Al.01 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1.5)

Question Source:

New Question Cognitive Level:

Lower References provided to Candidate None Technical

References:

20M-54.4.Cl page 14 2" bullet.

Objective #:

Task ID#:

10 CFR Pari 55 Content:

(CFR: 41.5 145.5)

Fundamental PaQe 38 0, '"0

39 B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

Which ONE (1) of the following IS the power supply to the Train "B" Solid State Protection System (SSPS) slave relays?

Vital Bus 1 Vital Bus 2 125VDC Bus 1 125VDC Bus 2 Answer B

ExplanationlJustification:

A.

B.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 013 Engineered Safety Knowledge of bus power suppiies to the following:

ESFASisafeguards equipment control KIA #

K2.01 KIA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)

Incorrect. This is the power supply to the Train a siave relays Correct IAW AOP-2.38.lB page 21 item 7.

incorrect. Siave relay power is provided by AC Vital bus 1 Incorrect. Slave relay power IS provided by AC Vital bus 1 Features Actuation Question Source:

New References provided to Candidate None Objective #:

Task ID#:

Question Cognitive Level:

Lower Memory Technical Referewas.

10 CFR Pari 55 Content:

&OOP-2.3B.16 page 2? item 7 (CFR: 41.7)

P a m 39 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

40.

The plant is operating at 100% power with all systems in NSA, In ADDITION to bus undervoltage and motor electrical protection, which of the below listed signals will directly TRIP Containment Air Recirculation Fan 2HVR-FNZOIA?

1. 2HVR-FN201A hi-hi vibration signal.

2. CIA signal.

3. CIB signal.

4. Safety Injection signal.

5. Containment Sump water level high signal A.

1,3, & 4 ONLY B.

1, 4, & 5 ONLY C.

2, &5ONLY, D.

2&3ONLY Answer B

ExplanationlJustification:

A.

B.

C.

D.

Incorrect. Will Not trip on CIB and it will trip on sump level high.

Correct. IAW 20M-44C.l.D page 3 2" paragraph.

Incorrect Does not trip on CIA.

Incorrect. Does not trip on CIA or CIE.

-~

ASysU KIASystem KIA Category KIA Statement 022 Containment Cooling Ability to manually operate andlor monitor in the control CCS fans KIA #

A4.01 KIA Importance 3.6 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided lo Candidate None Technical

References:

20M-44C.l.D page 3 2"' paragraph Objective It:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.5 to 45.8) room:

Page 40 of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

41.

The plant is operating at 100% power with all systems in NSA.

A large break LOCA occurs inside CNMT.

When the main generator tripped, the 2A Normal 4KV bus FAILED to transfer to the off site power source (SSST).

CNMT pressure reaches 20 psig and is stable.

RWST level reaches 360 inches and is slowly dropping.

All ESF equipment operates as designed.

Based on these conditions, how many HHSI/Charging and Recirculation Spray pumps will be DISCHARGING DIRECTLY into the reactor vessel?

HHSllCharqinq Recirculation Spray A.

E C

D 1

0 1

1 n

2 2

L 4

Answer C

ExplanationlJustification:

A.

B.

C.

lncofrect. EOG functioned thereinre both trains of emergency power are available. If candidates belleves only one train avaiiable and do N IT recognize that RWST level is below the transfer to Recirc setpoint. then they will choose this answer.

incorrect. EDG functioned therefore both trains of emergency power are available. If a candidate believes only one train available. then the will choose this answer.

Correct. IAW 20M-13.1.B page 3 3"paragraph.

lncnrrect. HHSi pumps are correct. However. two of the 4 Recim spray pumps are re-aligned to inject into the core. All 4 pumps wiii be run1 ing, but only 2 are injecting into the core. The other 2 continue to inject into the CNMT spray header.

KIA Sys #

KIA System KIA Cateoory KIA Statement 026 Containment Spray Knowledge of the physical connections andlor cause-ECCS effect relationships between the CSS and the following systems:

KIA #

K1.O1 KIA Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

20M-13.1.B page 3 3 paragraph.

Objective 11:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.2 to 41.9 145.7 to 45.8)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

42 A.

B.

C.

D.

The plant is operating at 100% power with all systems in NSA.

. Recirculation Spray Cooler 2RSS*E21A becomes INOPERABLE.

What Technical Specification actions are REQUIRED?

Entry into Technical Specification LCO:

3.6.7 Condition C ONLY 3.6.7 Condition C AND 3.6.8 Condition A ONLY 3.6.6 Condition A AND 3.6.7 Condition C ONLY 3.6.6, 3.6.8 Condition A, AND 3.6.7 Condition C.

Answer A

ExplanationIJustification:

A.

8.

C.

D.

Correct IAW Tech Spec 3.6.7 condition C Incorrect. Candidates who believe chemical addition is through the Recirc spray system will select this answer since the chem. Add subsy! :em would also be inoperable. However. the chem. Add system injects into the QS system NOT the Recirc spray system.

Incorrect. Candidate may believe that loss ofthis heat exchanger also impact one train of OS. However. the QS dispersion rinq is a seDarz e header and is NOT impacted by a loss of the RS heat exchanger.

Incorrect. Candidates who believe chemical addition is through the Recirc spray system will select this answer since the chem. Add subsy! 'em would also be inoperable. However, the chem. Add system injects into the QS system NOT the Recirc spray system. Candidate mayALS(

believe that loss of this heat exchanger will impact one train of QS. However. the QS dispersion ring is a separate header and is NOT imp: 3ed by a loss of the RS heat exchanger.

~

KIA Sys U KIA System KIA Category KIA Statement

'"6 Containment Spray NIA Ability to apply Technical specification:^ for a

.dAU 2.2.40 KIA Importance 3.4 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Application References provided to Candidate Tech Spec Section 3.6 Technical

References:

Tech Spec 3.6.7 condition C.

Objective #:

Task ID#:

10CFR Pari55Content:

(CFR:41.lOI43.2I435/45.3) system.

Page 42 01 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

43.

The plant is operating at 40% power with all systems in normal alignment for this mode.

0 Rod Control is in MANUAL.

A FULL load rejection occurs.

The reactor trip breakers remain CLOSED.

Tavg - Tref deviation indicates 6°F.

All systems function as designed.

Which ONE (1) of the following describes how the Steam Dump system will be operating for these conditions?

Steam Dump Bank 1 will be PARTIALLY OPEN. All other Steam Dump Banks will be CLOSED.

Bank 1 will be FULL OPEN. Bank 2 will be PARTIALLY OPEN. Banks 3 & 4 will be CLOSED.

Banks 1 & 2 will be FULL OPEN. Bank 3 will be PARTIALLY OPEN. Bank 4 will be CLOSED.

Banks 1, 2, & 3 will be FULL OPEN. Bank 4 will be PARTIALLY OPEN A.

B.

C.

D.

Answer A

ExplanaiionIJustification:

A.

8.

C.

D.

Correct. IAW 2OM-21.5.A.12 and 13 Incorrect Banks 3 & 4 response is correct but Bank 1 wiil oniy be partially open and bank 2 will be ciosed. The temperature error IS not lar( ?

enough to fully open bank 1 and partially open bank 2.

Incorrect. Bank 4 response is correct but Bank 1 will only be partially opfn and bank 2 will be closed. The temperature error IS not large en, ugh to fully open bank 1 and partially open banks 2 and 3.

Incorrect. Bank 1 wiii only be partially open and banks 2 8 3 will be closed. The temperature error is not large enough to fully open bank I ;;nd paifiaiiy open banks 2, 3, 8 4.

~~

. JA Sys U KIA System WA Category KIAStatement 039 KIA U A4.07 KIA Importance 2.8 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided io Candidate None Technical

References:

20M-21.5.A.12 and 13.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.5 to 45.8)

Main and Reheat Steam Ability to manually operate andlor monitor in the control room:

Steam dump valves

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

44 A.

8.

C.

D.

The plant is operating at 100% power with all systems in NSA Which ONE (1) of the below listed failures will cause the associated Main Feed Regulating valve ' 3 INITIALLY throttle CLOSED.

(Assume NO operator Action)

An associated level transmitter fails HIGH.

The selected steam flow transmitter fails HIGH.

The selected feed flow transmitter fails LOW.

The associated steam pressure transmitter fails LOW Answer D

ExplanationlJustification:

A.

6.

C.

D.

incorrecl. Level 1s median selected. Therefore, a single failure either way will not impact MFRV operation. IF level were NOT median selec this failure would cause the MFRV to throttle closed.

Incorrect. This will cause the valve to throttle OPEN.

lncorrecl. This will cause the valve to throttle OPEN.

Correct. IAW 2OM-24.4.1F attachment 4 page 31 2"' NOTE. Steam pressure is used to compensate steam flow indication for density. and as the same effect as steam flow. Therefore, a pressure transmitter failing low will cause the SGWLC system to see low steam flow with respect I I feed flow This anticipatory signal will drive the MFRV closed in an attempt to match feed flow to the Steam flow. This will be the initial response Since the SGWLC system is level dominant, when level drops as a result of this initial response, the MFRV will be driven open again in an attem t to restore level back to program value.

KIA Sys U KIA System KIA Category KIA Statement

'59 Main Feedwater Knowledge of MFW design feature(s) andlor interlock(s)

Feedwater regulatory valve operation (on which provide for the following:

basis of steam flow, feed flow mismatch)

.JA U K4.08 KIA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

2OM-24.4.1F attachment 4 page 31 2" VOTE Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

45.

The plant is operating at 70% power with all systems in normal alignment for this power level Both Main feed pumps are in operation.

The "B" Main Feed pump THEN trips and cannot be re-started, (1) What impact will this failure have on SG Feed Pump 21B Recirculation Valve 2FWR-(2) IAW AOP-2.24.1, Loss of Main Feedwater, what actions will be required in response to FCV150B operation?

this failure?

A.

(1) 2FWR-FCV150B will OPEN.

(2) Trip the Reactor and Go to E-0, Reactor Tip or Safety Injection (1) 2FWR-FCV150B will REMAIN CLOSED.

(2) Start the SG Startup feed pump [2FWS-P24]

(1) 2FWR-FCV150B will REMAIN CLOSED.

(2) Trip the Reactor and Go to E-0, Reactor Tip or Safety Injection B.

C.

D.

(1) 2FWR-FCV150B will OPEN.

(2) Start the SG Startup feed pump [2FWS-P24]

Answer B

ExpianationlJustification:

A.

Incorrect 2FWR-FCV15OB will remain CLOSED. The number of pumps running does impact the control system for the feed pump recircul. lion valves. However, the feed at 70% power is large enough (-16,000 gpm) to keep the valve closed. The reactor IS not required lo be tripped One main feed pump and the startup feed pump are adequate to supply the necessary feed flow. Therefore, AOP-2.24.1 requires the crew to st Irt 2FWS-P24.

Correct. IAW AOP-2.24.1 step 2b RNO and 20M-24.1.0 pages 11 and 12.

Incorrect. The reactor is not required to be tripped. One main feed pump and the startup feed pump are adequate to supply the necessary ?ed flow. Therefore, AOP-2.24.1 requires the crew to Start 2FWS-P24.

Incorrect 2FWR-FCVISOB will remain CLOSED. The number of pumps running does impact the control System for the feed pump recirculi:tion valves However, the feed at 70% power IS large enough (-16,000 gpm) to keep the valve closed.

D.

~

KIA Sys 11 KIA System KIACategory KIA Statement 059 Main Feedwater Ability to (a) predict the impacts of the following malfunctions or operations on the MFW: and (b) based on those predictions, use procedures to correct, controi. or mitigate the consequences of those malfunctions or operations:

Failure of feedwater control system KIA #

A2 11 KIA Importance 3.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

AOP-2.24.1 step 2b RNO and 20M-24. '.D Objective #:

Task iD#

10CFRPart55Content:

(CFR: 41.5/43.5l45.3/45.13) pages 11 and 12.

Page 45 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

46.

The plant is operating at 100% power with all systems in NSA.

EDG #I is on clearance for a lube oil change-out AND maintenance has just removed all lube oil from the crankcase.

An inadvertent reactor trip occurs COINCIDENT with a loss of offsite power.

All SG levels "Shrink" to 10% NR as a result of the trip.

All systems function as designed.

Based on these conditions:

Which motor driven auxiliary feed pump will be running AND which electrical bus will be providing f i e power to the pump?

" A AFW pump powered from 480V Bus 8N.

" A AFW pump powered from 4KV Bus 2AE.

"6" AFW pump powered from 480V Bus 9P, "6" AFW pump powered from 4KV Bus 2DF.

A.

6.

C.

D.

Answer D

ExplanationlJustification:

A.

6.

C.

D.

ulA Sys #

KIA System KIA Category KIA Statement Incorrect. Entire A Train will be de-energized and AFW pump motors are 4KV not 480V.

lncoriect. Entire ATrain will be de-energized.

Incorrect. AFW pump motors are 4KV not 480V.

Correct. iAW 20M-24.1.C page 5 1" paragraph.

3 1

AuxiliarylEmergency Knowledge of bus power supplies to the following:

AFW electric drive pumps KIA #

K2.02 KIA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

20M-24.1.C page 5 1" paragraph.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7)

Feedwater Page 46 of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

47.

The plant is operating at 100% power with all systems in NSA.

A Steam Generator Tube Rupture (SGTR) occurs on the " B Steam Generator.

All systems function as designed EXCEPT the turbine driven AFW pump did NOT start and could NOT be locally started.

The crew has entered E-3, Steam Generator Tube Rupture.

"B" Steam Generator NR level is 35% and rising.

At step 5 of E-3, the crew is attempting to isolate AFW flow to the "B" Steam Generator 218 SG AFW Throttle Valve 2FWE*HCVlOOC will NOT close and CANNOT be closed ?om the control room.

IAW E-3, Steam Generator Tube Rupture step 5 how will AFW be isolated to the "B" Steam Generyitor?

A.

B.

C.

D.

Reset SI THEN secure " A AFW pump.

Reset SI THEN secure "B" AFW pump, Secure "A" ANV pump THEN reset SI Secure "B" AFW pump THEN reset SI.

Answer A

ExplanationlJustification:

A.

B.

C.

q.

Correct IAW E-3 step 5. (predict which controls will need to be operated to prevent overfilling the SG)

Incorrect. Wrong pump. ZFWE'HCVIOOC is an '"A train mechanical vaibe, therefore the "A motor driven AFW pump must be secured to s',p Row to the "E SG.

Incorrect. Resetting SI is done AFTER shutting down the turbine driven AFE pump, but is done BEFORE shutting down the motor driven p Imps.

Incorrect. Wrong pump FWE'HCVIOOC is an " A train mechanical valve, therefore the " A motor driven AFW pump must be secured to sto, flow to the "E SG. Resetting Si is done AFTER shutting down the turbine dnven AFE pump, but is done BEFORE shutting down the motor driv.n pumps.

=

KIA Sys #

KIA System KIA Category KIA Statement 061 AuxiliarylEmergency Ability lo predict and/or monitor changes in parameters (to SG Level Feedwater prevent exceeding design limits) associated with operating the AFW controls including:

KIA #

Al.O1 KIA Importance 3.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Objective #:

Task ID#:

Technical

References:

E-3 step 5 10 CFR Part 55 Content:

(CFR: 41.5 145.5)

Page 47 01 io0

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

48.

The plant is operating in Mode 3.

. All svstems are in normal alianment for this mode EXCEPT Primarv Comoonent Coolin 1 Water Pump 2CCP*P21C is-racked onto the 2AE bus and IS running with'its control switch in AFTER START (Red Target).

Primary Component Coolhg'Water Pump 2CCP*P21A is racked onto the 2AE bus with ts control switch in AFTER STOP (Green Target).

Primary Component Cooling Water Pump 2CCP*P21B is racked onto the 2DF bus with,is control switch in AFTER STOP (Green Target).

A loss of offsite power occurs and all systems function as designed AFTER the EDG have completed sequentially loading all equipment, WHICH Primary Component Cooling Water Pump(s) will be running?

A.

ONLY 2CCP'P21 6.

6.

ONLY 2CCP*P21A AND 2CCP*P21B C.

D.

ONLY 2CCP*P21 C AND 2CCP*P21 B ALL Primary Component Cooling Water Pumps Answer B

ExplanationIJustification:

A.

B.

Incorrect. 2CCP'ZiA will also start.

Correct IAW 20M-15.1 D page 3 1"paragraph and 20M-15.1.D page 6 last paragraph NOTE.

2"a part of the WA has not been addressed. Use procedures to mitigate, correct or control is not applicabie to the situation. The alignment c the standby CCP pump 1s correct and the plant response is correct. Therefore, additional operator actions would NOT be required. Additionally if the situation posed in the question is modified such that the plant response is incorrect. the only procedural guidance that exists to correct wou 1 be to simply start the pump that should have started OR stop the pump that started inappropriately. Based on this information. the more impartar part of the WA has been addressed. it is important for the operators to understand the CCP start logic when the standby pump is racked onto ai emergency bus.

Incorrect. 2CCP21G will NOT start because 2CCP-2iA is NOT in disconnect posihon on the 2AE bus. 2CCP21Awill start.

Incorrect. ZCCP'2iC will NOT start because ZCCP-21A is NOT in disconnect position on the 2AE bus.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 062

=_

AC Electrical Distribution Ability to (a) predict the impacts of Ihe following malfunctions or operations on the ac distribution system:

and ( b ) based on those predictions, use procedures to correct. controi, or mitigate the consequences of those malfunctions or operations:

Aligning standby equipment with correct emergency power source (DIG)

KIA #

A2.11 KIA Importance 3.7 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension.

References provided to Candidate None Technical

References:

20M-15.l.D page 6 last paragraph Objective #:

Task ID#:

10CFR Part55 Content:

(GFR: 4i.5143.5145.3145.13)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

49 A.

0.

C.

D.

The plant is operating at 100% power with all systems in NSA.

Battery Charger *2-1 FAILS and its associated output breaker OPENS All systems function as designed.

Based on these conditions, what will be the status of 125VDC Switchboard 2-I?

ENERGIZED by 120VAC Vital Bus 1 ENERGIZED by station Battery *2-1 DE-ENERGIZED until the spare charger is installed as a replacement.

DE-ENERGIZED until Vital Bus 1 Manual Bypass Switch is placed to "Bypass".

Answer 6

ExplanationlJustification:

A.

6.

C.

D.

Incorrect #1 inverter automaticaily receives DC power on loss of input power; it does NOT output power to the DC SWBD.

Correct. IAW 2OM-39.1.0 page 3 3" paragraph Incorrect. lnstaliing the spare charger will restore the AC power to the SWBD. However, the battely will provide 125VDC power in the Intel n.

Incorrect. The battery will provide 125VOC power. Placing the Vital bus manual bypass switch to Bypass will restore AC power to a Vitai b s that failed to transfer thru its Static switch.

NOTE At Unit 2 there is NO position labeled Bypass. This is a Unit 1 term. Bypass position was used in this question to avoid giving any hints to le candidate on how to answer question #12.

KIA Sys #

KIA System KIA Category KIAStatement 063 DC Electrical Distribution Knowledge of the physical connections andlor cause-AC electrical system effect relationships between the DC electrical system and the following systems:

KIA #

K1.02 KIAlrnportance 2.7 Exam Level RO Level Of Difficulty: (1-5) iestion Source:

New Question Cognitive Level:

Lower Fundamental deferences provided to Candidate None Technical

References:

20M-39.1.B page 3 3' paragraph Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.2 to 41.9 145.7 to 45.8)

Page 49 M 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

50.

The plant is operating at 100% power with all systems in NSA.

An inadvertent reactor trip occurs COINCIDENT with a loss of offsite power.

BOTH EDGs FAIL to start and cannot be started.

The operating crew enters the appropriate Emergency Operating Procedure to address these conditions.

- 30 minutes after the reactor trip, CNMT pressure rises to 5.0 psig,

.. SI and CIA actuate What will be the status of the following CIA components AFTER this CIA actuation?

1. Letdown orifice isolation valves.

2. RCP seal water return CNMT isolation valves.

3. CNMT Instrument Air Compressor suction isolation valves.

4. Non-Regen Heat exchanger Letdown inlet valve.

A.

1. Closed.

2. Closed.

3. Closed.

4. Closed.

B.

1. Closed.

2. Open.

3. Closed.

4. Open.

C.

1. Closed.

2. Open.

3. Open.

4. Closed.

D.

1. Open.

2. Open.

3. Open.

4. Open.

Answer C.

ErplanationlJustification:

A.

EL C.

D.

KIA Sys U KIA System KIA Category KIAStatement 064 Emergency Diesel KIA #

K3.02 WA Importance 4.2 Exam Level RO Level Of Difficulty: (1-5)

Incorrect. This IS the position they should all be in if power is available. However, items 2 and 3 are 480V motor operated valves that woulr have already been OPENING when power was lost. Without power to close the valves, they would remain open.

IncOrreCt. Item 2 is a 480V motor operated valve that would have already been OPEN when power was lost. Without power to close the va re it would remain open. Item 4 is DC powered. and DC power is still available (batteries are designed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) to close the valve.

Correct. IAW EOP attachment A-0.2 pages 6-9.

Incorrect. Items 1 & 4 are DC powered, and DC power is still available (batteries are designed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) to close these valves.

~

Knowledge of the effect that a loss or malfunction of the EDIG system will have on the following:

ESFAS controlled or actuated systems Generator Question Source:

New References provided to Candidate None Objective #:

Task ID#

Question Cognitive Level:

Lower Fundamental Technical

References:

10 CFR Part 55 Content:

EOP attachment A-0.2 pages 6-9 (CFR: 41.7 145.6)

Paqe 50 01 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

51 A.

8.

C.

D.

The plant is in Refueling Mode with all systems aligned for core off-load.

While lowering a spent fuel assembly into the Spent fuel pool, the assembly ruptures ard releases ALL of the gases from ALL of the rods in that assembly ONLY.

NO other fuel assemblies have been damaged.

Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 goes into HIGH alarm.

Based on these conditions, will Fuel Building Vent Radiation Monitor 2RMF-RQ301NB ALSO go into a HIGH alarm condition? Why or Why Not?

NO, Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 is designed to detect gamma radiation (GM tL be)

AND Fuel Building Vent Radiation Monitor 2RMF-RQ301NB is designed to detect beta radiation (scintillation).

NO, The iodine and xenon released from the fuel assembly WILL BE sufficiently scrubbed out by t'ie water above the assembly.

YES, Fuel Pit Bridge Radiation Monitor 2RMF-RQ202 is designed to detect beta radiation (scintillarion)

AND Fuel Building Vent Radiation Monitor 2RMF-RQ301NB is designed to detect gamma radiaticn (GM tube).

YES, The iodine and xenon released from the fuel assembly WILL NOT BE sufficiently scrubbed o'it by the water above the assembly.

Answer D

ExplanationIJustification:

A.

C.

D.

Incorrect. The type of detectors is correct. However, if the gases released are emiding enough gamma radiation to actuate the high aiarm i r 2RMF-RQ-202. then there will be more than enough Xe and iodine to actuate the high alarm on 2RMF-RQ1301 incorrect. Some iodine will be scrubbed by the 23 feet of water, but enough iodine and other gases will be present to actuate the high alar, on 2RMF-RQ1301 Incorrect. Yes the alarm will actuate, but not because of detector types which are not correct.

Correct IAW 20M-43.1.C page 28 AND UFSAR Section 15.7.4.3. The analyzed fuel handling accident in the fuel pool will result in an offsi-5 dose.

The 2RMF-RQ1301 radiation monitor will detect this release and actuate the alarms. AOP-2.49.1 for the fuei handling accident also lists b< th monitors as symptoms of the event.

KIA Sys #

WA System KIA Category KIA Statement 073 Process Radiation Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

KIA #

K5.01 WA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental.

References provided to Candidate None Technical

References:

20M-43.1.C page 28 AND UFSAR sec ion Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.5 145.7)

Radiation theory, including sources, tyi es, units, and effects Monitoring 15.7.4.3

Beaver Valley Unit 2 NRC Written Exam (ZLOTG)

52.

The plant is operating at 100% power with all systems in NSA.

Service Water Pumps 2SWS'PZlA AND B are BOTH in service.

Service Water Pump 2SWS*P21C is on clearance and unavailable.

" A and "6" Primary Plant Component Cooling Water Heat Exchangers are BOTH in sei-vice "A and "5' Secondary Plant Component Cooling Water Heat Exchangers are BOTH in service.

A large Service water leak develops at the inlet to the " A Primary Plant Component Cooling Water Heat Exchanger. The leak causes the following Service water header pressure indications:

Service Water Header Press 2SWS-PI113A is 30 psig and stable.

Service Water Header Press 2SWS-PI1138 is 40 psig and stable.

(1) IF these Service Water Header Pressures are sustained for greater than 1 minute, what will be the impact on Secondary Plant Component Cooling Water Heat Exchanger operations?

(2) IAW AOP-2.30.1, Service WateriNormal Intake Structure Loss, what actions will be REQUIRED IF BOTH Service Water Header Pressures drop below 34 psig and cannot be restored abo\\!e 34 psig?

A.

(1) ONLY the " A Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED (2) Manually trip the reactor and Go to E-0, Reactor Trip or Safety Injection.

(1) ONLY the "A" Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED (2) Perform an emergency shutdown IAW AOP-2.51.I, Emergency Shutdown.

(1) NEITHER Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED.

(2) Manually trip the reactor and Go to E-0, Reactor Trip or Safety Injection.

(1) NEITHER Secondary Plant Component Cooling Water Heat Exchanger will be ISOLATED (2) Perform an emergency shutdown IAW AOP-2.51. I, Emergency Shutdown.

6.

C.

D.

Answer C

ExplanationlJustification:

A.

Incorrect. ESWS'MOV107A will auto close when pressure is less than 34 psig for greater than 45 seconds. However. this only Isolates the K header. The ' ' 0 header will continue to supply BOTH secondary plant component cooling water heat exchangers. Manually trip the reactor s correct Incorrect. 2SWSMOV107A WIII auto ciose when pressure is less than 34 psig for greater than 45 seconds. However, this only isolates the ' K header The "8" header will continue lo sundv BOTH secondarb nlant comnonent coolina water heat exchangers. Manuallv trio the reactor s

6.

References provided to Candidate None aojective #:

Task ID#:

Technical

References:

AOP-2.30.1 automatic actions on page

~ &

VOND 30-1 grid G-6 and 7: Part 2 IAW,AOP-2.30.1 step 2 RNO e.

10 CFR Part 55 Content:

(CFR: 41.5 143.5 14513 145113)

P a m 52 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

53.

The plant is operating at 100% power with all systems in NSA.

A Large Break LOCA occurs inside CNMT.

CNMT pressure rises to 35 psig.

All equipment functions as designed.

Which of the below listed components will NOW be cooled by Service Water?

1. CNMT Air Recirc Coolers

2. Charging pump lube oil coolers

3. Primary Plant Component Cooling Water Heat Exchangers

4. Recirculation Spray Heat Exchangers

5. Secondary Plant Component Cooling Water Heat Exchanger

6. Rod Control Area N C Units A.

B.

C.

D.

2, and 4 ONLY 3, 5. and 6 ONLY 1, 3. and 5 ONLY 1, 2, 4, and 6 ONLY Answer D

ExplanationlJustification:

A.

6.

incorrecl. Primary and Secondary plant component cooling water heat exchangers are isolated. Rod Controi area is NOT isolated at the he tder, but the inlet MOVs will only open when temp exceeds 107°F. Even if they did open, the outlet valve is manually closed so there will be no f IW.

Incorrect. CNMT air Recirc coolers are NOT isolated at the header, However, Sewice water is a backup to normal cooling and MUST be mi nually aligned. Primary and Secondary plant component cooling water heal exchangers are isolated incorrect. CNMT air Recirc coolers are NOT isolated at the header. However, sewice water IS a backup to normal cooling and MUST be m: iually aligned. Rod Controi area is NOT isolated at the header, but tne inlet MOVs will only open when temp exceeds 107OF. Even if they did ope, the outlet valve is manually closed so there will be no flow. Charging pump lube 011 coolers and Recirc spray heat exchangers are correct.

Correct. IAW VOND 30-1 grid D-6: 30-2 grid D-1; EOP Attachment A-0.5 D.

KIA Sys #

KIA System KIA Category WAStatement 076 Service Water Ability to manually operate andlor monitor in the control Emergency heat ioads KIA #

A4.04 KIAlmporlance 3.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

VOND 30-1 grid D-6: 30-2 grid D-1; EO;'

Objective #:

Task ID#:

10 CFR Pari 55 Content:

(CFR: 41.7 ! 45.5 to 45.8)

~~

room:

Attachment A-0.5

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

54.

The plant is operating at 100% power with all systems in NSA EXCEPT Station Air Compressor [2:;AS-C21 B] is on clearance and unavailable.

A large leak develops in the station sewice air header.

Station air header pressure begins to drop.

As station air header pressure continues to drop, at what setpoint will each of the below listed aubnatic actions occur:

1. Diesel-Driven Air Compressor 21AS-C21 -AUTOMATIC START

2. Condensate Polishing Air Compressor 2SAS-C22 - AUTOMATIC START

3. SAS Main Header to Service Air Header AOV 2SAS-AOV105 - AUTOMATIC CLOSE A.

1. 82 psig

2. 90 psig

3. a6 psig B.

1. 86 psig

2. 90 psig

3. 82 psig C.

I. a2 psig

2. 86 psig

3. 90 psig D.

1. 90 psig

2. 86 psig

3. a2 psig Answer A

ExpianationlJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 078 Instrument Air Ability to monitor automatic opera!ion of the IAS, Air pressure KIA #

A3.01 KIA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

20M-34.2.6 page 2 pressure setpoints Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 145.5)

Correct IAW 20M-34.2.6 page 2 pressure setpoints. The candidate will need to know the sequence of starting (which one first, Zm and 3m but will NOT need to have these three setpoint memorized. All of these automatic actions are geared towards maintaining Instrument Air available Incorrect. Wrong setpoints for 21ASX21 and 2SAS-AOV105.

Incorrect. Wrong setpoints for 2SAS-CZ2 and 2SAS-AOV105.

Incorrect. All setpoints are wrong.

~

including:

Paae 54 0,100

55.

h B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

Which ONE (1) of the below listed components DIRECTLY receives a CIA signal to CLOSE?

HEPA Filter House No. 1 Outlet Damper 2HVS*MOD211A Pri Comp Clg Wtr Supply Hdr B lsol2CCP*MOV175-1 Control Room ACU Outside Air Intake DMPR 2HVCiMOD201A Regen HX Normal Charging Disch Vlv 2CHS'MOV310 Answer B

ExplanationlJustification:

A.

6.

C.

D.

KIA Sys U KIA System KIA Category KIA Statement 103 Containment Ability to monitor automatic operation of the containment Containment isolation KIA #

A3.01 KIAlmportance 3.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Memory References provided to Candidate None Technical

References:

EOP Attachment A-0.2 page 7 Objective #:

Task ID#:

10 CFR Pari 55 Content:

(CFR: 41.7 145.5)

Incorrect. This damper receives a CIA OPEN signal Correct. IAW EOP Attachment A-0.2 page 7 Incorrect. This damper receives a CIB signal.

Incorrect. This valve receives a SI signal.

system, including:

Page 55 01 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

56.

The plant is operating at 100% power with all systems in NSA

. All Rods are indicating 228 steps on DRPI.

The following alarms and indications are THEN received in the control room:

Annunciator A4-8B, Rod Control System Non-Urgent Alarm - has ALARMED ALL 48 DRPl General Warning (GW) LED lights - "Flashing".

DRPl Rod Deviation 1, R, & 2 LED lights - LIT.

DRPl Data A Failure 1, 2, & 3 LED lights - "Flashing".

DRPl Central Control Failure 1, 2, & 3 LED lights - NOT LIT.

DRPl Urgent Failure 1, 2, & 3 LED lights - NOT LIT.

DRPl Data B Failure 1, 2, & 3 LED lights - NOT LIT.

All DRPl Rod Bottom (RB) lights - NOT LIT.

All Rods are STILL indicating 228 steps on DRPI.

Reactor power REMAINS at 100% and stable.

Based on these conditions:

1. Rod positions will be indicated every steps.

2. IAW ARP A4-8B, Rod Control System Non-Urgent Alarm, the REQUIRED action is to Place the Accuracy Mode selector switch to the position.

A.

1. 6

2. "A + B" B.

1. 6

2. "B Only"

1. 12

2. "A + B" D.

1. 12

2. "B Only" Answer D

ExplanationIJusiification:

A.

Incorrect. Indications given in the stem of the question indicate that Data A has failed. If the WAC power to the "A" coils is lost, these indic 4tions would be present. The presence of the Non-urgent alarm resulk in "haif-accuracy" mode. This means DRPl will indicate every 12 steps in!,!ead of every 6. Placing the switch to A+B is incorrect. The NSA position is A+B Candidate may confuse this A+B switch with the SSPS A+B posi-on where the NSA position is for 1 train to be in A+B and the other train is in A or B only.

Incorrect. Indications given in the stem of the question indicate that Data A has failed. If the WAC power to the "A" coils is lost. these indic.itions would be present. The presence of the Non-urgent alarm results in "half-accuracy" mode. This means DRPl will indicate every 12 steps in! 'ead of every 6. Piacing the switch to 'W ONLY is correct. The NSA position is AtB. Candidate may confuse this A+B switch with the SSPS A+B osition where the NSA position is for 1 train to be in A+B and the other train is in A or B only.

Incorrect Indications given in the stem of the question indicate that Data A has failed, If the 6VAC power to the '"P coils is lost, these indic.itions would be present. The presence of the Nan-urgent alarm results in "half-accuracy"mode. This means DRPl will indicate every 12 steps in every 6. Placing the switch to A+B IS incorrect. The NSA position is A+B. Candidate may confuse this A+B switch with the SSPS A+B pos where the NSA position is for 1 train to be in A+B and the other train is in A or B only.

Correct. IAW ARP A4-8B (20M-1.4.AAK page 3) Ran on simulator io verify all indications for Data Afailure.

6.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 014 Rod Position Indication Ability to (a) predict the impacts of the following maifunctions or operations on the RPIS. and (b) based on Loss of power to the RPIS those on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.02 KIA Importance 3.1 Exam Level RO Level Of Difficulty: (1-5)

KIA U

'uestion Source:

New Question Cognitive Level:

Higher Comprehension

.eferences provided to Candidate None Technical

References:

ARP A4-8B (20M-1.4.AAK page 3)

Objective #:

Task 1011:

10CFR Part55Conteni:

(CFR: 41.5143.5145.3145.13)

Page 56 O ' l O O

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

57.

The plant is in Mode 3 with the RX trip breakers CLOSED and the shutdown banks withdrawn. All systems are aligned normally for this plant condition.

. BOTH Source range channels are indicating 500 CPS and stable.

20ST-2.3, Nuclear Source Range Channel Test MUST be performed, for N32 ONLY before the stiirtup can proceed.

During the performance of this surveillance, what control room actions will be REQUIRED to preve,it the reactor from tripping on Source Range High Flux?

Place the N32 SR drawer:

"High Flux at Shutdown" switch to the BLOCK position.

"Level Trip" switch to the BYPASS position "HV Manual OniOff switch to the HV ON position "Operation Selectoi' switch to the I O 4 CPS position A.

B.

C.

D.

Answer E

ExplanationiJustification:

A.

incorrect. This does NOT block the High Flux trip; rather it enables the High Flux at SID alarm.

6.

Correct. iAW 2OST-2.3 page 37 step 7.

C.

Incorrect. This ensures the HV power to the detector. It does NOT block the trip by holding the power to SSPS relays.

D. Incorrect. This injects a test signal equal to I O ' CPS. It does NOT keep the signal from exceeding I O ' CPS.

",A Sys #

KIA System KIA Category KIA Statement

~

2 Nuclear Instrumentation Ability to manually operate andlor monitor in the control Trip bypasses room:

KIA #

A4.03 KIA Importance 3.8 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

20ST-2.3 page 37 step 7.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.7 i 45.5 to 45.8)

Page 57 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

58.

The plant is operating at 100% power with all systems in NSA.

. The "Main Turb First Stage Press Sensor Select" switch is in the PM 446 position.

IF the "Main Turb First Stage Press Sensor Select' switch is placed in the PM 447 position, what iriipact will this have on plant operations?

The Press 2MSS-PT447 transmitter INSTEAD of 1A First Stage STM Press 2MSS-PT446 transmitter.

will NOW be coming from 1 B First Stage STbl A.

6.

C.

AMSAC "Bypass" permissive D.

T,,

signal to Rod control Tlef signal to Steam Dumps Steam Dump Load Rejection "Arming " signal Answer D

ExplanationlJustification:

A.

8.

C.

D.

Correct IAW 20M-24.4.1F attachment 5 KIA Sys #

KIA System KIA Category KIA Statement 016 Non-nuclear Ability to manually operate andlor monitor in the control NNI channel select controls

'A #

A4.01 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5) ilestion Source:

New Question Cognitive Level:

Higher comprehension References provided to Candidate Nane Technical

References:

20M-24.4.lF attachment 5 Objective #:

Task ID#:

10 CFR Pari 55 Content:

(CFR: 41.7 145.5 to 45.8)

Incorrect. This signai IS not selectable.

Incorrect. This signal is always provided by the 447 transmitter.

Incorrect. AMSAC bypass requires both 446 and 447 input. Not selectable vp Instrumentation room:

Page 58 or 100

59 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

Which ONE (1) of the following is NOT a source of hydrogen inside containment following a Desigi~i Bases Large break LOCA of an RCS cold leg?

Pressurizer Relief Tank gas space.

Zirc - water reaction between the fuel clad and the reactor coolant.

Corrosion of aluminum and zinc by the ECCS water.

Radiolysis of water in the core and CNMT sump.

Answer A

ExplanationlJustification:

A.

E.

C.

D.

KIA Sys #

KIA System KIA Category KIAStatement 028 KIA U K5 03 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

Lesson Plan 2SQS46.1 slide 7 Objective U:

Task IO#.

20 CFR Part 55 Content.

(CFR: 41.5 I 45.7)

Correct PRT gas space gas is nitrogen NOT hydrogen. Also. any gas in the PRT gas space will remain in the PRT during a 1.arge Coid le!

LOCA.

Incorrect. This IS a source of hydrogen in containment.

Incorrect. This is a source of hydrogen in containment.

incorrect. This is a source of hydrogen in Containment. ____

~~

Hydrogen Recombiner and Purge Control Knowledge of the operational implications of the following concepts as they appiy to the HRPS:

Sources of hydrogen within containme1 t

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

60 A.

B.

C.

D.

A large break LOCA has occurred and the following plant conditions exist:

. RVLIS is available.

. All RCPs are STOPPED.

. The RCS is 50°F SUPERHEATED.

Which ONE (1) of the following plant conditions will REQUIRE a RED PATH entry into FR-C.1, Response To Inadequate Core Cooling?

The two hottest core exit TC is 125O"F, ALL the other core exit TCs are 700°F AND RVLIS Full range level is 33%.

The two hottest core exit TC is 125O"F, ALL the other core exit TCs are 700°F AND RVLIS Dynam c range level is 33%.

The three hottest core exit TCs are 750"F, ALL the other core exit TCs are 700°F AND RVLIS Full range level is 33%

The three hottest core exit TCs are 750"F, ALL the other core exit TCs are 700°F AND RVLIS Dyn m i c range level is 33%.

Answer C

ExplanationIJustification:

A.

Incorrect. The three MAX TCs are NOT greater than 1200°F NOR are they greater than 729'F RVLlS level IS low enough to require entry ' TCs are hot enough.

6.

Incorrect. The three MAX TCs are NOT greater than 1200'F NOR are they greater than 72TF. Wrong RVLlS range.

C.

Correct. IAW EOP Status tree F-0.2.

Incorrect. The three MAX TCs are NOT greater than 120OoF. Wrong RVLlS range.

~

KIA Sys #

KIA System KIA Category KIA Statement 017 In-Core Temperature Monitor System (ITM) room:

RCSiRCP operation during inadequate core Ability to manually operate andlor monitor in the control Temperature values used to determine cooiing (!.e., if appiicable. average offi, e highest values)

KIA #

A4.02 KIAlmporlance 3.8 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

EOP status tree F-0.2.

Objective #:

Task ID#

10 CFR Part 55 Content:

(CFR: 41.7 145.5 to 45.8)

Page 60 of 100

61 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

The Unit is in Mode 6. A fuel assembly is being lowered into the core.

IF the fuel assembly "BINDS" against another fuel assembly, downward motion of the hoist will be automatically stopped to prevent fuel assembly damage.

What manipulator crane interlock provides this protection?

Tube Down Underload Overload Bridge-Trolley-Hoist Answer B

ExplanationlJustification:

A.

B.

C.

D.

Incorrect. Tube down interlock will stop hoist downward motion when the hoist is all the way down.

Correct. IAW LP 3505-6.13 slide 49. (2RP-3.3)

Incorrect. Overload will stop UPWARD motion if an assembly IS binding while moving upward.

Incorrect. Bridge-Trolley-Hoist interlock will only allow rnotionlrnovement in one direction at a time.

KIASysU KIASystem KIA Category KIA Statement 034 KIA U K4 01 KIA Importance 2.6 Exam Level RO Level Of Difficulty: (1-5)

Fuel Handling Equipment Knowledge of design feature(s) andlor interlock(s) which provide for the following:

Fuel protection from binding and drop1 ing Question Source:

New

'eferences provided to Candidate None ojective U:

Task ID#

Question Cognitive Level:

Lower Fundamental Technical

References:

10 CFR Part 55 Content:

LP 3SQS-6.13 slide 49. (2RP-3.3)

(CFR: 41.7)

Page61 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

62.

The plant is operating at 100% power with all systems in NSA.

An inadvertent turbine trip occurs.

The "B" reactor trip breaker FAILS to OPEN.

All other systems function as designed.

Without any operator action, where will RCS temperature automatically stabilize?

A.

541°F

0.

547°F C.

550°F D.

554°F Answer C

ExplanationlJustification:

A.

6.

C.

Correct IAW 20M-21.5.A.12.

D.

Incorrect. This is where RCS would stabilize if it were relying on the steam dump lo-lo Tavg interlock to stop a cooldown.

Incorrect. This is where RCS would stabilize if it were being col?trolled by the Rx trip controller. However, with ' " E trip breaker Still closed, t e steam dumps will function on the load rejection controller which has a 3'F deadband before it will open the steam dumps.

Incorrect. This is where RCS would stabilize if it were relying on the SG safeties to control temperature. This would be necessary if the ste. m dumps were NOT armed. However. the ' ' A reactor trip breaker opening will arm the dumps.

~

KIA Sys U KIA System KIA Category KIA Statement 041 Steam Durnpnurbine Knowledge of the Physical connections and/or cause-RCS Bypass Controi effect relationships between the SDS and the foliowing systems:

'A #

K1.05 KIA Importance 3.5 Exam Level RO L w e i Of Difficulty: (1-5)

Jestion Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

2OM-21.5.A.12.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41 2 to 41.9 145.7 to 45.8)

Page 62 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

63 A.

B.

C.

D.

The plant is operating at 100% power with all systems in NSA.

. Condensate Bypass Vlv 2CNM-AOVI 00 inadvertently OPENS What effect will this have on plant operations?

Feedwater inlet Temperature to the Steam Generators will DROP Condenser hotwell level will RISE.

Main feed Pump Suction pressure will DROP.

Turbine Plant Demineralized Water Storage Tank will RISE.

Answer A

ExplanationlJustification:

A.

8.

C.

D.

KIASys #

KIASystem KIA Category WA Statement 056 Condensate Knowledge of the physical connection5 andlor cause-MFW effect relationships between the Condensate System and the following systems:

Correct. IAW VOND 22A-2 grid 6-5 Incorrect. This would be true if the bypass around the normal LCV was failed open. (LCViO3).

Incorrect. This would be true if the condensate pump Recirc valve was failed open. (FCVIOI).

Incorrect. This would be true if the bypass around the normal condensate pump reject MOV was failed open. (LCVIOI)

KIA #

K1.03 KIA Importance 2.6 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided to Candidate None Technical

References:

VOND 22A-2 grid 6-5 Objective #:

Task ID#:

10CFRPart55Content (CFR: 41.2 to41.9l457to45.8)

Page 63 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

64 B.

C.

D.

Which ONE (1) of the below listed set of conditions are the MINIMUM REQUIRED conditions to ac'Jate annunciator A I 2-4C, Condenser Unavailable?

2 out of 2 Condenser Pressure transmitters below 19.5" of Hg vacuum 4 out of 4 Circulating Water Pumps NOT running 2 out of 2 Condenser Pressure transmitters above 19.5 of Hg vacuum 3 out of 4 Circulating Water Pumps NOT running 1 out of 2 Condenser Pressure transmitters below 19.5" of Hg vacuum 4 out of 4 Circulating Water Pumps NOT running 1 out of 2 Condenser Pressure transmitters above 19.5" of Hg vacuum 3 out of 4 Circulating Water Pumps NOT running OR OR OR OR Answer C

ExplanationlJustification:

A.

6.

C.

Correct.ZOM-26.4.ABM page 3 setpoints.

D.

incorrect. This will actuate the alarm, but it is NOT the MINIMUM required conditions. It only takes K transmitters.

Incorrect. It only takes X transmitters. vacuum is below NOT above. Also MUST have 414 circ pumps OFF.

Incorrect Vacuum is below NOT above, Also MUST have 414 circ pumps OFF.

~

KIA Sys #

KIA System KIA Category KIAStatement 175 Circulating Water Knowledge of circulating water system design feature@)

Heat sink

.dA #

K4.01 KIA Importance 2.5 Exam Level RO Level Of Difficulty: (1-5) and interlock(s) which provide for the following:

Question Source:

New References provided to Candidate None Objective #:

Task ID#:

Question Cognitive Level:

Lower Memory Technical

References:

10 CFR Part 55 Content:

20M-26.4.ABM page 3 setpoints (CFR: 41.7)

65 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

The plant is operating at 100% power with all systems in NSA.

10 Ton CO, Storage Tank 2FPD-TK22 MUST be removed from service for maintenancf:

10 Ton CO, Storage Tank 2FPD-TK23 is able to supply CO, to the System 2 Zones.

24 Ton CO, Storage Tank 2FPD-TK24 is able to supply GO2 to the System 2 Zones.

IAW 20M-33.4.G. CO, Fire Protection System Staeup And Storage Tank Fill, how will the C02 sys!em be re-aligned to maintain operability of the system?

Align the -(I) for service, then place the Smoke Detection Panel 2FPS-PNL-XL3 MAIN/RESERVE switch to (2)

, AND isolate (3)

(1) 24 Ton COz Storage Tank 2FPD-TK24 (2) MAIN (3) ONLY 10 Ton CO, Storage Tank 2FPD-TK22 (1) 24 Ton C02 Storage Tank 2FPD-TK24 (2) RESERVE (3) BOTH 10 Ton CO, Storage Tanks 2FPD-TK22 & 23 (1) 10 Ton CO? Storaae Tank 2FPD-TK23 i2j MAIN (3) ONLY 10 Ton C02 Storage Tank 2FPD-TK22 (1) 10 Ton COz Storage Tank 2FPD-TK23 (2) RESERVE (3) BOTH CO, Storage Tanks 2FPD-TK22 & 24 iswer B

ExplanationiJustification:

A.

8.

C.

D.

KIA Sys #

KIA System KIACategory KIAStatement 086 Fire Protection Ability to predict andlor monitor changes in parameters (to FPS lineups Incorrect Switch must be placed to reserve for 24 ton tank master valve to function. All other items are correct Correct. iAW 20M-33.4.G page 25 step 5. This 'is NOT minutia. rather it tests the candidates ability to predict what line-up changes are net i to prevent operating the COz outside of the required aiignment.

incorrect. 24 ton unit must be aligned for service. Smoke Detection Panel 2FPS-PNL-XL3 MAINIRESERVE switch must be placed to reser e, and BOTH 10 ton units must be isolated.

Incorrect. 24 ton unit must be aligned for service, items 2 and 3 are correct prevent exceeding design limits) associated with Fire Protection System operating the controls including:

KIA#

A I.05 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

20M-33.4.G page 25step 5 Rev. 10.

Objective #:

Task ID#:

10 CFR Part 55 Content (CFR: 41.5 145.5)

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

66 A.

B.

C.

D.

The plant is in Mode 5 preparing to enter Mode 4.

Valve alignments are being performed on a Safety-Related system.

The REQUIRED N5A position of a manually operated globe valve is 2 Turns OPEN.

The valve must be in this position PRIOR to Mode 4 entry.

The valve has MINIMAL safety significance.

The valve list REQUIRES Concurrent verification for this valve.

The second verifier will receive 5 mR performing the Concurrent verification.

The valve has NO remote valve indication The valve CANNOT be verified in the correct position by the performance of a functional test.

IAW the guidance provided in 1/20M-48.3.D, Administrative Control Of Valves And Equipment, hoi / will the Concurrent verification for this valve be addressed?

The Shift Manager shall waive the Concurrent verification for this valve based on MINIMAL safety significance and HIGH radiation exposure to the second verifier.

The First verifier places the valve in the required position; WITHIN 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the second verifier verifii?s the valve in the required position.

The First verifier places the valve in the required position; the second verifier remains OUTSIDE thi~, line of sight of the first verifier THEN verifies the valve in the required position.

The First verifier places the valve in the required position WHILE the second verifier observes the f rst verifier placing the valve in the required position.

Answer D

Cxplanation/Justification:

Incorrect. The shift manager may waive the independent verification of a safety related valve if it has minimal safety significance and will re iult In lOmR exposure to the second verifier. This valve only has 5 mR exposure. Also since this valve requires a number of turns. the only indivi' ual that can waive the concurrent verification is the operations manager.

Incorrect. These are the requirements for independent verification of Tech Spec related actions that support current plant conditions. Sincr this valve is required for Mode 4 entry it is NOT required for the current plant Mode.

Incorrect. These are the requirements for independent verification NOT concurrent verification. Additionally, this valve must be concurrentl verified since independent verification would negate the original condition.

Correct. IAW 1120M-48.3 D lll.C and VI.A.9.a. The valve requires concurrent verification and it cannot be waived by the shift manager.

Concurrent verification is Specifically defined for valves that require a number of turns. This definition specifically states that the second ve fier will observe the original manipulation.

6.

C.

D.

E

-~

KIA Sys #

WA System KIA Category WAStatement NIA Generic Conduct Of Operations Knowledge of how to conduct system 1; ieups.

KIA #

2.1.29 KIAlmportance 4.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

IAW 1120M-48.3.D i1l.C and VI.A.9.a.

Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 145.1 145.12) such as valves. breakers, switches. etr Page 66 oi 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

67.

The plant is operating in Mode 6 with all systems in normal alignment for this Mode, Core Re-loading activities are in progress.

There are 100 fuel assemblies in the core.

Source Range Channel N32 fails low.

Source Range Channel N31 remains OPERABLE.

Which ONE (1) of the below listed evolutions can STILL be performed WITHOUT violating the Tec'inical Specification required actions for Source Range Instrumentation?

A.

Removing a SPENT fuel assembly from its fully lowered core position and placing it into the fuel tri isfer cart.

Moving an underwater camera from one core l o c a t i o n to another to verify proper seating of fuel assemblies.

Removing a temporary secondary source device that was installed in the center core location to as-.ist in plotting 1/M data.

Moving a fuel assembly from a temporary core location into the final core location that is adjacent ti source range channel N31 B.

C.

D.

Answer B

ExpianationlJustification:

A.

6.

Incorrect. Even though this wouid iessen the overall reactivity of the core, it would violate the TS action for one inoperable source range ch :nnei.

This is a Core AlleraOon since it is fuel movement within the vessel with fuel in the vessel.

Correct. Loss of HV power supply will render N32 inoperable, TS action is to IMMEDIATELY suspend core alterations. Core alterations are defined as movement of fuel. sources, or reactivity control components within the vessel WITH fuel in the vessel. Underwater cameras are #lone of these therefore. this evolution would be permitted.

Incorrect Even though this is removing a source device, it would violate the TS action for one inoperable source range channel. This is a C Ire Alteration since it is movement as a source within the vessei with fuel in the vessei.

Incorrect. Even thought he assembly is already in the core in its temporary location, it would violate the TS action for one inoperable SOurCt range channei This is a Core Alteration since it is fuel movement within the vessel with fuel in the vessel.

KIA Sys #

KIA System KIA Category KIAStatement NIA Generic Conduct Of Operations Knowledge of procedures and limitatior 5 KIA #

2.1 36 KIA Importance 3.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

Tech Spec 3.9.2; Tech Spec Definition if Objective ff:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 143.6 145.7)

D.

~

involved in core alterations.

Core Alteration.

Page 67 Of 'DO

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

68.

The plant is operating at 90% power with all systems in NSA Control Bank D is at 229 steps.

Core Burnup is 3000 MWDIMTU.

RCS Boron Concentration is 1250 ppm.

Equilibrium Xenon concentration conditions exist.

Tavg is equal to Tref.

How many gallons of dilution water will be needed to raise power to 95% and keep Tavg equal to 7 ref?

Assume the Boron Correction factor is 1.O and disregard any changes in Xenon concentration.

A.

20 gals.

B.

420gals C.

520gals D.

720 gals Answer C

ExplanationlJustification:

A.

6.

C.

Incorrect. If the candidate makes a math error and stops after determining the change in boron concentration, they will choose this answel Incorrect. If the candidate does all of the calculations correctly and but sloppily uses the correct nomograph for dilution, they will choose th 5 answer. Sloppy use of the nomograph means to inaccurately apply the straight edge to the nomograph.

Correct CE-28 3000 mdirntu equals boron worth of -6.0 pcmipom. CB-21 1250 ppm power defect for 90.95% is 90 pcm. 90pcm1-6pcm1p~ n equals -1 5 ppm. Must reduce RCS boron 15 ppm to Compensate for power defect. Using CE-33 nomograph detemine volume of water nt ?ded

-520 gals.

NOTE: Alternate method of using CB-33 formula 8069FTI.02264FT3/lbm/8.33 X In(CilCfj= 516 gals. Candidate will NOT have the table that specifies the volume of the RCS and will therefore need to use the nomograph to determine the answer.

Incorrect. If the candidate does all of the calculations correctly but DOES NOT realize the correct nomograph for dilution is NOT linear, thf I will choose this answer.

0.

KIA Sys #

KIA System KIA Category KIA Statement NIA Generic Conduct Of Operations Ability to use procedures to determine the effects on reactivity of pian1 changes,,;mh as reactor coolant system temperature.

secondary plant, fuel depletion, etc.

KIA #

2 1.43 KIAlmportance 4.1 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 143.6 145.6)

Curve book curves CB-28, 21, &

33 Technical

References:

Curve book curves CB-28.21. 8 33

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

69.

What is the Technical Specification basis for the Reactor Core Safety Limit?

There must be a least a 95% probability at a 95% confidence level that the :

Hot fuel rod in the core does not experience DNB or centerline fuel melting.

Integrity of the Reactor Coolant System will be protected against overpressurization.

Core will be protected against rapid increases in neutron flux Maximum clad oxidation does not exceed 17% of clad thickness A.

B.

C.

D.

Answer A

ExplanationlJustification:

A.

6.

C.

D.

KIA Sys 11 KIA System KIA Category KIA Statement NIA Generic Equipment Control Knowledge of the bases in Technical Correct IAW Tech Spec bases 2.1.1 page B 2.1.1-2 Incorrect. This is the bases for the other Tech spec Safety limit.

incorrect. This is the Tech Spec bases for the high positive rate trip. Setpoint.

Incorrect. This IS an ECCS acceptance criteria NOT the Tech Spec bases for the core safety limit.

~

Specifications for limiting conditions for operations and safety limits.

KIA #

2.2.25 KIA Importance 3.2 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Memory References provided to Candidate None Technical

References:

Tech Spec bases 2.1.1 page B 2.1.1-2 Objective #:

Task ID#:

10CFRPart55Content:

(CFR: 41.5141.7143.2)

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 1

70.

Refer to the drawing of a typical valve control circuit for a 480 VAC motor-operated valve (see figure below).

With NO initiating condition present, the valve is currently OPEN. If the SI pushbutton is depressec, the valve will and when the SI pushbutton is subsequently released the valve will

-1 25 V D C CLOSED BY INITIATING CON0 ITION si

1. 3 CONTACT VAL\\

-1 25 VDC

RGIZE TO OPEN VALVE; UERGIZE To CLOSE IE I TYPICAL VALVE CONTROL CIRCUIT I

A.

remain open; remain open B.

close; remain closed C.

remain open; close D.

close; open Answer B

ExplanationlJustification:

A.

B.

C.

D.

KIASys#

KIASystem WACategory KIAStatement NIA Generic Equipment Control Ability lo obtain and interpret station elxtrical and mechanical drawings.

KIA #

2.2.41 KIA Importance 3.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

BVPS Bank Question 13933 Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

Print reading skills Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41 10145.12 145.13) incorrect. Wrong initial response: wrong subsequent response Correct Right initial response; right subsequent response.

Incorrect. Wrong initial response: right subsequent response.

Incorrect. Right initial response: wrong subsequent response.

Page 70 0' 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

71.

You have been assigned the task of venting a radioactive system that is located in a Locked High Radiation Area (LHRA).

When you open the vent valve you receive an UNEXPECTED dose rate alarm on your electronic alarming dosimeter (EAD).

IAW NOP-WM-7025, High Radiation Area Program, what are your Required actions for these conditions?

A.

B.

C.

D.

Immediately notify Radiation Protection (RP) and stay in the area to await further instructions.

Close the vent valve and report the alarm to the control room supervisor and Radiation Protection I QP)

Immediately exit the area and perform whole body frisk.

Close the vent valve and immediately exit the area Answer D

ExplanationlJustification:

A.

8.

C.

D.

KIASys#

KIASystem WACategory NIA Generic Radiation Control Knowledge of radiological safety princ ples incorrect. These are the correct actions personnel contamination, Incorrect. These would be appropriate actions for an alarming air monitor Incorrect Frisking IS required before exiting the RCA but not necessarily required as pari of LHRA exit Correct. IAW NOP-WM-7025 step 4.2.12 on page 6 and 7.

~~

KIA Statement pertaining to iicensed operator duties such as containment entry requirements, fuel 'landling responsibilities, access to locked higP radiation areas, aligning filters. etc.

KIA #

2.3.12 KIA Importance 3.2 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided lo Candidate None Objective #:

Task ID#:

i 0 CFRPart55Content:

(CFR: 41.12145.9/45.10)

Technical

References:

NOP-WM-7025 step 4.2 12 on page t and 7.

Page 71 Of 10"

72 A.

B.

C.

D.

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

What type of radiation detector is used in the In-Containment High Range Area Radiation Monitors 2RMR*RQ206 and 207?

Proportional.

Geiger-Mueller.

Ion Chamber.

Beta/Gamma Scintillator.

Answer C

ExplanationlJustification:

A.

6.

C.

D.

KIA Statement KIA Sys #

KIA System KIA Category NIA Generic Radiation Control Knowledge of radiation monitoring systems.

such as fixed radiation monitors and al3rms.

portable survey instruments. personnc monitoring equipment, etc.

KIA #

2.3.15 KIA Importance 2.9 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Memory References provided to Candidate None Technical

References:

20M-43.1.C page 51 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.12 143.4 145.9) incorrect. This is the type of detector used in the source range instrument incorrect. This IS the type of detector used in most area monitors, however in order to meet the Reg. guide 1.97 criteria for post accident n 2nitors.

Ion chambers are needed to avoid saturating the detector from the extremely high radiation fields that the monitors are designed to detect Correct. IAW 20M-43.1.C page 51. In order to meet the Reg guide 1.97 criteria for post accident monitors. Ion chambers are needed to a oid saturating the detector from the extremely high radiation fieids that the monitors are designed to detect Incorrect. These are the type detectors used in the process radiation monitoring system.

Page 72 of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

73.

The plant is in Mode 4 with a plant shutdown in progress.

RHS has just been placed in service.

RHR Pump 2RHS*P21A is in service.

RHR Pump 2RHS*P21B is out of service.

PRZR level is 15% and stable.

21C RCP is in service.

Annunciator AI-5H Residual Heat Removal System Trouble (RHR TRN A FLW RHS'FT605A LOW:

computer address point F0600D) -Alarms The following control room indications NOW exist:

RHR Train A Flow [2RHS-F1605A] is oscillating between 0 and 1400 gpm 21A RHR Pump Amps [2RHS-I121A] are erratically oscillating.

21A RHR HX Bypass Vlv [2RHS*FCV605A] is erratically oscillating.

21C RCP Amp [2RCS-I121C] indicates 688 amps and stable.

PRZR level remains at 15% and stable.

In order to address these conditions, what procedure are you Required to enter?

A.

AOP-2.6.5, Shutdown LOCA B.

AOP-2.6.8, Abnormal RCP Operation C.

q.

AOP-2.10.1, Residual Heat Removal System Loss AOP-2.10.2, Loss of RHS While At Reduced Inventory/Midloop Conditions C

ExplanationlJustification:

A.

E.

C.

D.

Incorrect. This procedure entry would be appropriate if the RHR system was displaying these symptoms due to a loss of inventory. There i'.e no indications that a loss of inventory is progress. PRZR level is 15 and stable.

incorrect. Entry into the procedure would he appropriate if the RCP was displaying the erratic amps and flow. Since the RHP pump IS displ wing the erratic amps and flow and RCP amps are stable entry into this procedure is NOT appropriate or required.

Correct. IAW symptoms listed in AOP-2.10.1.

Incorrect Although all of the symptoms listed in the stem are symptoms in this AOP also. you must also be at reduced inventory or midloor, before entry is required. With PRZR level stable at 15% the plant is NOT at reduced inventory OR midioop.

KIA Sys #

KIA System KIA Category KIA Statement NIA Generic Emergency ProcedureslPlan Ability to recognize abnormal indicaliohs for system operating parameters that are I ?try-level conditions for emergency and abr 3rrnal operating procedures.

KIA #

2.4.4 KIA Importance 4.5 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

Symptoms listed in AOP-2.10.1 Objective #:

Task I D #

10CFRPart55Content:

(CFR:41.10/43.2/45.6)

Page 73 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

74.

Which of the below listed Abnormal Operating Procedures contain Immediate Operator Actions?

(1)

(2)

(3)

AOP-2.1.8, Rod lnoperability (4)

(5)

(6)

(7)

(8)

AOP-2.1.3, RCCA Control Bank Inappropriate Continuous Movement AOP-2.1.7, Rod Position Indication Malfunction AOP-2.6.4, Steam Generator Tube Leakage AOP-2.24.1, Loss of Main Feedwater AOP-2.26.1, Turbine and Generator Trip AOP-2.36.1, Loss of All AC Power When Shutdown AOP-2.36.2, Loss of 4KV Emergency Bus A.

0.

C.

D.

Answer C

ExplanationlJustification:

A.

B.

c.

Correct.iAWAOPs2.1.3,2.1.8.2.26.1.82.36.2.

D.

KIA Sys U KIA System KIA Category KIA Statement WA Generic Emergency ProcedureslPian Knowledge of abnormal condition procr.dures.

4 #

2.4.11 KIA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)

-estion Source:

New Question Cognitive Level:

Lower Memory References provided to Candidate None Technical

References:

AOPs 2.1.3. 2.1.8, 2.26.1, 8 2.36.2.

Objective #:

Task ID#:

10CFR Part55 Content:

(CFR: 41.10143.5145.13) 1, 2, 6, 7, & 8 ONLY 2, 3. 4, 5, & 7 ONLY

1. 3, 6, & 8 ONLY 3,4, 5, & 7 ONLY Incorrect. 2 and 7 do NOT have iMAs: 3 does Incorrect. 3 is the only one with IMAs.

Incorrect. 4, 5,8 7 do NOT have IMAs; 1, 6. 8 8 have IMAs

~ _ _

Page 74 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

75.

The plant is operating at 25% power with all systems in normal alignment for this power level A Steam Generator Tube Rupture (750 gpm) occurs.

RCS pressure slowly drops to the Low PRZR pressure reactor trip setpoint.

Ruptured SG NR level is 20% and slowly rising.

The BOP operator wishes to pre-emptively isolate feed flow to the ruptured SG.

IAW the guidance provided in 1120M-538.2, User's Guide, how will this pre-emptive action be accomplished?

The BOP operator is REQUIRED to:

Complete the Immediate actions of E-0, Reactor or Safety Injection, THEN obtain concurrence fron the SM/US. THEN isolate feed flow to the ruptured SG.

Complete the Immediate actions of E-0, Reactor or Safety Injection, THEN isolate feed flow to the ruptured SG, THEN at the first crew brief inform the SM/US of the preemptive actions taken.

Isolate feed flow to the ruptured SG, THEN complete the Immediate actions of E-0, Reactor or Safety Injection, THEN at the first crew brief inform the SMiUS of the preemptive actions taken.

Obtain concurrence from the SMIUS, THEN isolate feed flow to the ruptured SG, THEN complete tile Immediate actions of E-0, Reactor or Safety Injection.

A.

B.

C.

D.

Answer A

ExplanationlJustification:

A.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement NIA Generic Emergency ProcedureslPlan Knowledge of crew roles and responsil ilities KIA #

2.4.13 KIA Importance 4.0 Exam Level RO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

1120M-53.B.2 item 10 on page 7 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41 10 145.12)

Correct. IAW 1120M-53.B.2 item 10 on page 7. Preemptive actions can oniy be performed after completing the IMAs and after obtaining S' IIUS concurrence Incorrect. Must obtain permission first. This is the requirements for any automatic action that failed to occur.

incorrect. This wouid be the appropriate response to completing an automatic action that failed to occur EXCEPT the actions were comple ?d out of order Incorrect Completing the IMAs MUST is accomplished first

~

during EOP usage.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY The plant is operating at 100% power with all systems in NSA A LOCA occurs coincident with a loss of offsite power.

All systems respond as designed.

The crew has entered procedure ECA-1.1, Loss Of Emergency Coolant Recirculation due tm-i the inability to verify cold leg recirculation capability.

At step 13 the crew is attempting to perform an RCS cooldown to Mode 5 at IOO"F/hr.

IF the RCS cooldown cannot be manually established from the control room:

(1) What local actions will be REQUIRED to perform the cooldown?

(2) What would be the consequences of NOT performing these actions?

Direct Local operators to:

(1) Open SG Atm steam Dump Valves [2SVS*PCVlOlA(B)(C)] IAW EOP Attachment A-1.1 1, Man:ial (2) RCS depressurization will NOT be permitted and the time to RWST depletion will be shortened (1) Open SG Atm steam Dump Valves [2SVS*PCVlOlA(B)(C)] IAW EOP Attachment A-1.1 1, Man ial (2) RCS bleed and feed will be immediately required to maintain core cooling.

(1) Perform EOP Attachment A-1.18, ERFS Diesel Generator Startup THEN start the Station and LNMT (2) RCS depressurization will NOT be permitted and the time to RWST depletion will be shortened (1) Perform EOP Attachment A-1.18, ERFS Diesel Generator Startup THEN start the Station and CNMT (2) RCS bleed and feed will be immediately required to maintain core cooling.

AND Handpump Operations Of Hydraulically Actuated Valves.

Handpump Operations Of Hydraulically Actuated Valves.

air compressors.

air compressors.

76 A.

B.

C.

D.

Answer A

ExplanatlonlJustification:

A.

6.

C.

D.

Correct. Action directed by ECA-1.1 step 13 and ECA-1.1 bases page 3 item 3, the cooldown is being done to allow RCS depressurizatio to limit breakfiow and prolong the time to RWST depletion.

Incorrect. Correct action but incorrect consequence for not completing the action. Feed and bleed is not required as long as AFW flow is functloning AFW would be in service if systems responded as designed and stated in the stem.

Incorrect. These would be correct actions if the SG Atm steam Dump Valves were air operated valves like Unit 1. Correct consequence.

Incorrect. These would be correct actions if the SG Atm steam Dump Valves were air operated valves like Unit 1. Incorrect consequence -or not completing the action. Feed and bleed is not required as long as AFW flow IS functioning. AFW would be in Service if systems responded.Is designed and stated in the stem.

~ _ _

WASys +i WASystem WA Category WAStatement 00001 1 Large Break LOCA NIA Knowledge of local auxiliary operator asks during an emergency and the resultallt operationai effects.

WA #

2.4.35 WA Importance 4.0 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

ECA-1.1 bases, EOP Att. A-1.11 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 143.5 145.13)

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

77.

The plant is operating at 26% power with all systems in normal alignment for this power level Annunciator A2-4D Reactor Coolant Pump Seal Trouble is in alarm (computer address RCF 218 SEAL LK OFF HIGH).

RCP21B No. 1 seal leakoff flow indicates >6 gpm (off-scale high).

RCP21B No. 2 seal leakoff flow is less than 0.1 GPM.

VCT pressure is 26 psig.

RCP21 B seal injection flow is 9.5 gpm.

IAW AOP 2.6.8, Abnormal RCP Operation, which ONE (1) of the following actions and sequence o!

actions are you REQUIRED to direct the crew to perform?

A.

Stop RCP 218, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV3038] within 3 - 5 minutes of securing the pump. THEN initiate an Emergency Shutdown to Hot Standby in accordance with AOF 2.51. I, Emergency Shutdown.

Trip the reactor and go to E-0, Reactor Trip or Safety Injection. Complete the immediate actions of ::-O THEN Stop RCP 21 B, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV303B]

within 3 - 5 minutes of securing the pump.

Monitor seal return temperature, THEN maintain seal injection flow to RCP 218 greater than 9 gpm THEN initiate an Emergency Shutdown to Hot Standby in accordance with AOP 2.51.1, Emergency Shutdown.

Trip the reactor. THEN Stop RCP 218 and go to E-0, Reactor Trip or Safety Injection. Complete the immediate actions of E-0, THEN shut the Seal Water Leakoff Vlv [2CHS*MOV303B]

within 3 - 5 minutes of securing the pump.

5.

C.

Answer E

ExplanationIJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA category KIAStatement 000015/17 RCP Malfunctions Incorrect. Wrong Sequence and wrong procedural guidance.

Correct. IAW AOP~2.6.8 step 2.9 RNO Incorrect. Wrong actions and wrong procedural guidance.

Incorrect. Wrong sequence of correct actions.

Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

When to secure RCPs on loss of coolir 1 or seal injection KIA if AA2.10 KIAlmportance 3.7 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

BVPS Unit 1 Bank (1LOT7 Audit Exam)

Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

AOP-2.6.8 step 2.9 RNO 2OM-7.4.W -I Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR 43.5 145.13)

Page 77 Of 100

78 A.

B.

D.

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY The plant is operating at 100% power with all systems in NSA EXCEPT:

. The control switch for PZR PORV 2RCS*PCV455C is in the CLOSE position and its associi ted block valve is closed and de-energized.

The Technical Specification surveillance for RCP seal injection flow has just been satisfactorily completed.

Immediately after completing the RCP seal injection flow surveillance, Pressurizer pressure transmitter 2RCS*PT444 fails HIGH.

NO operator actions have been taken.

How will this failure impact RCP seal injection flow?

(Assume NO reactor trips have occurred)

RCP seal injection flow will ~

(1)

IF an accident were to NOW occur, the amount of ECCS flow that WOULD BE diverted from the EOCS injection path will be -(2) the range assumed in the safety analysis.

(1) increase (2) outside of (1) increase (2) within (1) decrease (2) outside of (1 ) decrease (2) within Answer B

ExplanationlJustification:

A.

B.

Incorrect. Correct impact on seal injection flow. Incorrect impact on assumed ECCS flow.

Correct. RCS pressure will drop, which allows seal injection flow to increase, However the RCP seal injection SuNeillanCe Sets the flow lo Mhin the limits assumed in the safety anaiysis. Even if pressure is dropping, the manual throttle valves have been adjusted to ensure ECCS injection flow is within the values assumed in the accident analysis.

Incorrect. Incorrect impact on seal injection flow. Incorrect impact on assumed ECCS flow.

Incorrect. Incorrect impact on seal injection Row. Correct impact on assumed ECCS flow.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 000027 Pressurizer Pressure Ability to determine and interpret the following as they RCP injection flow Control System Malfunction apply to the Pressurizer Pressure Control Malfunctions:

WA #

FA2.14 KIA Importance 2.9 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

Technical Specification 3.5.5 and bas :s:

20M-6.4.IF Attachment 2 Objective #:

Task ID#

10 CFR Part 55 Content:

(CFR: 43.5 / 45.13)

Page 78 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

79.

The plant is operating at 100% power with all systems in NSA.

A Steam Generator Tube Rupture occurs on the " B Steam Generator.

All systems function as designed.

The crew has just entered E-3, Steam Generator Tube Rupture.

" A and " C Steam Generators are intact.

The following plant conditions exist:

Total AFW flow is 900 gpm and stable CNMT pressure is 1 psig and stable.

" A and "C" NR SG level are 0%.

" B NR SG level is 5% and rising.

The reactor operator requests permission to perform "pre-emptive" actions and isolate all AFW flov. to the "B" Steam Generator.

IAW the guidance contained in E-3, Steam Generator Tube Rupture, what direction are you REQUIRED to give the reactor operator AND what is the bases for this direction?

A.

Isolate feed flow to the "B" Steam Generator, the required heat sink will be maintained by " A and "C" Steam Generators.

Isolate feed flow to the "6" Steam Generator, "B" Steam Generator overfill must be avoided to limit he radiological consequences.

Continue feeding the "B" Steam Generator until NR level is >12%, "6" Steam Generator is required for a heat sink.

B.

D.

Continue feeding the " B Steam Generator until NR level is >12%, "B" Steam Generator tubes mus.

remain covered to avoid SG depressurization.

Answer D

ExplanationlJustification:

A.

6.

C.

D.

KIASys#

WASystem U A Category 000038 Steam Generator Tube NIA Knowledge of the specific bases for E!)Ps.

Incorrect. Feed flow IS NOT to be isolated UNLESS NR of >12% has been reached (pre-emptive requirement of EOP users guide page 7 ilem 10 41h bullet) Heat Sink requirement is correct.

Incorrect.. Feed flow IS NOT to be isolated UNLESS NR of >12% has been reached (pre-emptive requirement of EOP users guide page 7 item 10 41h bullet), SG overfill is a concern but not at the expense of allowing a ruptured SG to depressurize.

Incorrect. Correct direction, however "0' SG wiil not be needed as a heat sink with A and C intact.

Correct. IAW E-3 step 5 bases page 65 2"' bullet.

KIA Statement Rupture (SGTR)

KIA #

2.4.18 KlA Importance 4.0 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

E-3 step 5 bases page 62 2"' bullet.

Objective #:

Task I D #

10CFRPart55Conient:

(CFR:41.10I43.1 145.13)

Page 79,f 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

80.

The plant is operating at 100% power with all systems in NSA

. A Reactor trip occurs coincident with a loss of offsite power.

. The steam driven AFW pump failed to start and cannot be started.

All other plant equipment responded as designed.

30 minutes AFTER the reactor trip the following plant conditions exist:

. All SG pressures are 1005 psig and stable.

. RCS Subcooling is 35°F and stable.

. Loop ATs are indicating upscale and stable.

. All SG NR levels are 45% and slowly dropping.

. Total AFW flow is 100 gpm and stable.

(1) When natural circulation has been established, what will be the status of TColdand Tho,?

(2) IAW EOP Attachment A-I.7, Natural Circulation Verification, what directions are you REQUIRED to give the crew in order to enhance natural circulation?

A.

(1) Tcoldwill be at 547°F and Thot will be stable or rising (2) Raise SG NR levels by increasing AFW flow.

(1) Tcoldwill be at 512°F and That will be stable or dropping (2) Raise the rate at which steam is being dumped.

(1) Tcaldwill be at 512°F and Thol will be stable or rising (2) Raise SG NR levels by increasing AFW flow.

(1) TColdwill be at 547°F and Thol will be stable or dropping (2) Raise the rate at which steam is being dumped.

B.

C.

D.

Answer D

ExplanationlJustification:

A.

6.

C.

D.

Incorrect. Right Tc response; Wrong Th response and wrong enhancement directions. TC will be at Saturation temperature for SG pressurr Th will be rising as natural circulation is being developed BUT it will be stable or dropping once it has been developed. If Th IS stiil rising then atural circulation has not been developed. Raising SG levels may seem plausible however, NR SG Ieveis are within the band of 35~50%

The dir ction for enhancing IS to raise the steam dump rate.

Incorrect. Wrong Tc response; Right Th response and Right enhancement directions. Tc will be at saturation temperature for SG pressure.vhich is 1005 psig (547~F).

Th will be rising as natural circulation is being developed BUT it will be stable or dropping once it has been develope. if Th is still rising then natural circulation has not been deveioped. 512°F corresponds to 35'F below Tsat. A non-discriminating candidate may lliink this 1s where Tc will be for these conditions.

Incorrect. Wrong TC response: Wrong Th response and wrong enhancement directions. Tc will be at saturation temperature for SG pressu e. Th will be rising as natural circulation is being developed BUT it will be stabie or dropping once it has been developed. If Th is Still rising then atural circulation has not been developed. Raising SG ieveis may seem plausible however, NR SG levels are within the band of 35.50%. The dir ction for enhancing is to raise the steam dump rate. 512'F corresponds to 35°F below Tsat. A non-discriminating candidate may think this is wh. re Tc will be for these conditions.

Correct. TC will be at saturation temperature for SG pressure which is 1005 psig (547°F). Th will be rising as natural circulation IS being de eioDed BUT it will be stable or dropping once it has been developed if Th is still rising then natural circulation has not been developed. IAW EOP Attachment A-1.7 Natural Circulation Verification the direction for enhancing natural Circulation is to raise the rate of dumping steam.

NOTE:

This question is NOT too similar to RO question # I O. RO question # I O requires candidate to know only trends and the concept of Tc being equal to or less than saturation pressure in the SGs. This SRO question requires the SRO candidate to calculate Tc and addresses Th which is NOT addressed in the RO question. This question also requires the SRO candidate to determine what acti3ns will be directed to enhance natural circulation.

KIA Sys #

KIA System KIA Category KIA Statement 000056 Loss of Offkite Power KIA #

AA2.19 KIAlmportance 4.2 Exam Level SRO Level Of Difficulty: (1-5) h e s t i o n Source:

New Question Cognitive Level:

Higher Comprehension Ability to determine and interpret the following as they apply to the Loss of Offsite Power:

T-cold and T-hot indicators (wide rang?)

aferences provided to Candidate Steam Tables Technical

References:

EOP Attachment A-1.7 Natural Circulz'ion Verification Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Page 80 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

81.

The plant is operating at 100% power with all systems in NSA A Reactor trip and Safety injection occur.

All systems function as designed.

The crew has implemented E-0, Reactor Trip or Safety Injection and has begun the diagnostic steps of E-0 with the following plant conditions:

All SG pressures are 800 psig and stable.

All SG NR levels are 35% and slowly rising.

All Secondary radiation monitors are consistent with pre-event values.

CNMT Pressure is -1.O psig and stable.

CNMT sump level is consistent with pre-event values.

CNMT radiation is consistent with pre-event values.

RCS Subcooling is 40°F and slowly dropping.

AFW flow is 700 gpm and stable.

RCS Pressure is 1125 psig and slowly dropping.

PRZR level is 12% and slowly dropping.

Auxiliary Building Radiation levels are rising.

Auxiliary Building sump levels are rising.

Based on these conditions:

What procedural entry is REQUIRED?

A.

ECA-1.2, LOCA Outside Containment J.

ES-1.I, SI Termination.

C.

D.

E-2, Faulted Steam Generator Isolation.

E-3, Steam Generator Tube Rupture Answer A

ExplanationiJustification:

A.

6.

C.

D.

Correct. iAW E-0 step 20 Auxiliary Building radiation ievels rising and evidence of a LOCA outside CNMT requires entry into ECA-1.2.

Incorrect. RCS pressure must be stable or rising and PRZR level must be > 17 'A for this to be the right procedural entry.

incorrect. At least one SG pressure must be dropping in an uncontrolled manner for this to be the right procedural entry.

Incorrect. At least one SG level must be rising in an uncontrolled manner or secondary radiation must be inconsistent with pie-event ValUl'S for this to be the right procedural entry.

~.

KIA Sys #

KIA System KIA Category KIA Statement WIE04 LOCA Outside Ability to determine and interpret the following as they apply to the (LOCA Outside Con!ainment)

Facility conditions and selection of ap.iropriate procedures during abnormal and erne-gency operations.

Level Of Difficulty: (1-5)

Containment KIA 11 EA2.1 KIA lmoortance 4.3 Exam Level SRO

~~

..~~

~

Question Source:

New References provided to Candidate None Objective #:

Task ID#:

Question Cognitive Level:

Higher Analysis Technical

References:

10 CFR Part 55 Content:

E-0 steps 14-20: E-0 Step 20 bases.

(CFR: 43.5 145.13)

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

82.

The Plant is operating at 100% power with all systems in NSA.

Control Bank D is at 229 steps.

Control Bank D Demand step counters are at 229 steps.

Control Rod Group Selector Switch is in the "MAN" position The following control room alarms are received:

A4-9F Rod At Bottom A4-3C Tavg Deviation from Tref A4-4F NIS Power Range Comparator Deviation A4-4G NIS Power Range Neutron Flux Rate High Plant Parameters are NOW as follows:

Tavg is 575°F and slowly dropping.

RCS Pressure is 2230 psig and slowly dropping.

Reactor power has dropped to 96% and is slowly rising.

PR N-41 Negative Rate Trip bistable is LIT All other PR Negative Rate Trip bistables are NOT LIT Control Bank D Demand step counters remain at 229 steps Based on these conditions:

What procedure contains the REQUIRED guidance to address these plant conditions?

A.

C.

D.

AOP 2.1.8, Rod lnoperability E-0, Reactor Trip or Safety Injection.

AOP 2.1.3, RCCA Control Bank Inappropriate Continuous Movement.

AOP 2.1.7, Rod Position Indication Malfunction.

Answer D

ExplanationIJustification:

A.

6.

C.

Incorrect. No entry conditions for E-0 have been met. The PR rate coincidence is 214 and oniy one rate bistable has been actuated.

Incorrect. The entry conditions for this procedure would require rods stepping out in conjunction with A4-3C Tavg Deviation from Tref in a1:irm The rods are not stepping in the conditions of the question.

Incorrect. Entry into this procedure is required ONLY if there IS no evidence of a plant response to the alarms. In this question, the plant hzij responded to a dropped rod with corresponding templpressurelpower change. This procedure may be entered as part of the initial diagno:itics.

however entry into this procedure is not REQUIRED.

Correct. IAW symptoms listed for AOP 2.1.8. the alarms and plant response are consistent with a dropped rod. AOP 2.1.5 Dropped rod ha.; been deleted, and AOP 2.1.8 now addresses a dropped rod in Part A. SRO candidate must evaluate the given conditions and those that are NC'T present to determine that a rod has dropped, and is in fact at zero steps NOTE: The stem IS worded usina the word contains due to the ailowance for entrf into AOP 2.1.7 which would then diaunose AOP 2.1.8 i the D.

correct procedure to address these conditions. If the stem asked what procedure entry is required, then there wouid be 2 possible answer! Only AOP 2.1.8 "Contains" the appropriate guidance.

KIA Sys #

KIA System KIA Category KIA Statement 000003 Droooed Control Rod Abiiitv to determine and intermet the followina as thev Rod position indication to actual rod D( jition Level Of Difficulty: (1-5) applito the Dropped Control' Rod

~

KIA #

AA2 01 KIA Importance 3 9 Exam Level SRO Question Source:

New References provided to Candidate None iective #:

Task ID#:

Question Cognitive Level:

Higher Analysis Technical

References:

AOPs 2.1.3. 2.1.7. 8 2.1.8 10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Page 82 oi 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

83. The Plant is operating at 100% power with all systems in NSA.

RCS activity is high, right at the Technical Specification Limits, due to leaking fuel element i.

At T = 0, RCP 21 C Thermal barrier heat exchanger develops a leak AND 21C RCP Therm 31 Barrier Outlet lsol Vlv [2CCP*AOV107C] FAILS to isolate and CANNOT be closed.

Thermal barrier outlet flow is 60 gpm and stable.

At T = 1 minute, The following alarms and indications are received:

. Annunciator A4-5C Radiation Monitoring L.evel High -Alarms

. Component Cooling Heat Exchanger Radiation Monitor 2SWS-RQII 01 AND Component Cooling Service Water Radiation Monitor 2SWS-RQI102 are BOTH in - HIGH Alarm

. Radiation Monitor 2SWS-RQI101 is reading 9.0 X 10.' pCilml.

. Radiation Monitor 2SWS-RQ1102 is reading 9.0 X pCilml No Reactor Trip or SI signals have been actuated.

No Reactor Trip or SI signals are required.

If all of these conditions continue until T = 20 minutes, What is the highest Emergency Plan Classification REQUIRED, if any, at T = 20 minutes?

(Assume NO Dose projections will be available until T = 50 minutes).

A.

B.

Unusual Event Alert D.

Site Area Emergency.

No Emergency Plan Classification is required.

Answer C

ExplanationlJuStificatio":

A.

B.

C.

Incorrect. Candidate could choose this based on RCS identified leakage being less than 25 gpm (Tab 2.6) based on thermal barrier outlet ow rising from a nominal 45 gpm to 60 gpm.

incorrect. Candidate could choose this based on [2SWS-RQllOZJ being greater than 2 times the ODCM setpoint (Tab 7.2). however this M'JST be for a period of greater than 60 minutes to be a UE.

Correct. IAW Tab 7.2 and the bases for Tab 7.2. [2SWS-RQIlOl] is 200 times the ODCM setpoint and this has been for greater than 15 m 'lutes.

The candidate must refer to the EAL Tab 7.2 and the corresponding table 7-1 and apply the given data to the EAL matrix. The keys to the question are to recognize that the given radiation monitors are indicators used to determine if an EAL criterion has been exceeded AND to recognize that the numbers in table 7-1 have been exceeded. After analyzing and applying this informalion, the candidate may still choose JE since this value has been exceeded, but NOT for Incorrect. Candidates could choose this if they incorrectly declare both the Fuel barrier and RCS barrier to be potential losses.

60 minutes. Making the correct EAL determination demonstrates the SRO ability.

D.

WASysU WASystem WA Category WA Statement 000059 Accidental Liquid NIA Ability to diagnose and recognize trend!; in an RadWaste Rel.

accurate and tlrnely manner utilizing th.~,

appropriate controi room reference ma ?rial.

KIA 11 2.4.47 WA Importance 4.2 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Appiication References provided to Candidate EALs.

Technical

References:

EAL Tab 7.2 and table 7-1 Objective #:

Task ID#:

10CFRPart55Content:

(CFR:41.10/43.5l4512)

Page 83 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

84.

The Plant is operating at 100% power with all systems in NSA.

A major fire has started in the control room.

The fire is NOW out of control and the fire brigade has not been able to extinguish the fire.

As the Shift manager, you direct the implementation of 20M-56C, Alternate Safe Shutdowmi From Outside The Control Room.

Which ONE (1) of the following methodsilocations will be used to bring the unit to cold shutdown?

Direct the crew to conduct a :

A.

B.

C.

D.

Natural circulation cooldown from the Alternate Shutdown Panel (ASP).

Natural circulation cooldown from the Emergency Shutdown Panel (ESP),

Forced circulation cooldown from the Alternate Shutdown Panel (ASP).

Forced circulation cooldown from the Emergency Shutdown Panel (ESP)

Answer A

ExplanationiJustification:

A.

E.

C.

D.

"A Sys #

KIA System KIA Category KIASiatement

,0067 Piant Fire On-site NIA Knowledge of "fire in the planr proced res Correct IAW 20M-56C.l.B page 2 3 paragraph and 20M-56C.4.A page 2 I" paragraph and 20M-56C.4.B page 3 2"'item Incorrect Implementation of 20M-56C. Alternate Safe Shutdown From Outside The Control Room requires the ASP to be activated. The E 3P is activated for small fires. toxic fumes, etc.

Incorrect. RCPs are tripped before leaving the control room.

Incorrect. RCPs are tripped before leaving the control room KIA #

2.4.27 KIAlrnportance 3.9 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Objective #:

Task ID#:

Technical

References:

20M-56C.l.B page 2 3" paragraph an I ZOM-56C.4.A page 2 1" paragraph and 2Ob -

56C.4.B page 3 2"' item.

10CFR Part55Content:

(CFR: 41.10i43.5l45.13)

Page 84 01 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

85.

The Plant IS operating at 100% power with all systems in NSA.

The control room crew is performing 20ST-43.6, Containment High Range Area Monitor Channel Test.

During the surveillance, the HIGH alarm setpoint for In-Containment High Range Area motiitor

[2RMR*RQ206] is found to be set at 2.6 X I O 4 R/hr.

Background radiation is 100 mr/hr.

Based on these conditions, what is the MINIMUM Technical Specification/LRM action, if any, that 5 REQUIRED?

A.

B.

C.

D.

No Technical SpecificationlLRM action is required.

Restore the required alarm channel to OPERABLE status within 30 days.

Adjust the alarm setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Declare the radiation monitor alarm inoperable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Answer C

Explanation/Justification:

A.

6.

C.

D.

-'A Sys #

KIA System KIA Category KIA Statement Incorrect. The alarm setpoint is out of range, and must be adjusted.

Incorrect. This would be the required action if the MONITOR was inoperable.

Correct In this case, the LRM provides specific actions for the alarm setpoint being out of range. Therefore, the minimum required action i to adjust the setpoint as specified in LRM 3.3.15 Condition A.l Incorrect. This action in an option BUT it must be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> NOT 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

~~

ii061 ARM System Alarms Ability to determine and interpret the following as they appiy to the Area Radiation Monitoring (ARM) System Alarms:

Required actions if alarm channel is 01 of service KIA U AA2.06 WA Importance 4.1 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher References provided to Candidate 20ST-43.6. LRM 3.3.15, TS 3.3.3 Technical

References:

LRM 3.3.15 Condition A.1 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 43.5 145.13)

Application Page 85 01 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

86.

A plant heatupistartup is in progress with RCS average temperature at 325°F.

Other plant conditions are as follows:

Recirculation Pump 2RSS*P21 D is INOPERABLE.

ChargingiHHSI Pump 2CHS*P21 C i s on clearance for maintenance A risk assessment for this condition has NOT YET been performed.

ChargingiHHSl Pump 2CHS*P21 B becomes INOPERABLE Which ONE (1) of the following describes the Technical Specification REQUIRED Actions?

A.

Restore the 2RSS'P21D recirculation pump and 2CHS*P21B ChargingiHHSl Pump to OPERABLE status BEFORE exceeding 350°F.

Restore ONLY the ChargingiHHSI Pump 2CHS*P21B to OPERABLE status BEFORE exceeding 350°F.

Restore the 2RSS*P21D recirculation pump and 2CHS*P21B Charging/HHSI Pump to OPERABLE status BEFORE exceeding 375°F.

Restore ONLY the ChargingiHHSI Pump 2CHS*P21B to OPERABLE status BEFORE exceeding 375°F.

6.

C.

D.

Answer A

ExplanationlJustification:

1.

d.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement 006 Emergency Core Cooling NiA Knowledge of conditions and limiiatior 5 in the KIA #

2.2.38 KIA Importance 4.5 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

BVPS Bank 56320 Question Cognitive Level:

Higher Application References provided to Candidate LCO 3.5.2 and LCO 3.5.3 Technical

References:

LCO 3.5.2 and Bases; LCO 3.5.3 and 3ases; Objective #:

Task ID#:

10CFR Part55 Content:

(CFR:41.7141.10i43.1 145.13)

Correct. iAW LCO 3 5.2 both pumps must be operable before transitioning above 350°F. At Unit 2 RSS' P21C and D provide the LHSl fur :tion during recirculation phase.

incorrect. Bath pumps are required before exceeding 350°F.

Incorrect. The 25 degree allowance in Note 2 of TS 3.5.2 is only applicable to the charging pump. The Recirc spray pump is required befo 13 exceeding 350°F.

Incorrect. Both pumps are required before exceeding 350°F.

facility license.

and LCO 3.0.4 Page 86 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY 87 The Plant IS operating at 100% power with all systems in NSA The control switches for all PORV Motor Operated Is01 Vlvs [2RCS*MOV535, 536, 5371 an in AUTO An inadvertent reactor trip occurs 2 Minutes later Pressurizer Spray Valve 2RCSPCV455A fails OPEN and is stuck OPEN The PORV Motor Operated lsol Vlvs [2RCS*MOV535, 536, 5371 will AUTOMATICALLY close when (1)

Pumps (2)

E-0, Reactor Trip or Safety Injection REQUIRES you to direct the crew to stop Reactor Coolant A.

(1) 2/3 PZR Protection channels decrease to less than 2000 psig (2) 21A and 21C (1) 2/3 PZR Protection channels decrease to less than 2185 psig (2) 21A and 21 C (1) 2/3 PZR Protection channels decrease to less than 2000 psig (2)216and 21C (1) 2/3 PZR Protection channels decrease to less than 2185 psig (2) 218 and 21C B.

C.

D.

swer B

ExplanationIJustification:

A.

8.

C.

0.

KIA Sys #

KIA System KIA Category KIA Statement nin Pressurizer Pressure Ability to la) Dredict the imDacts of the followinq Sprav valve failures InCorrect. Wrong setpoint for auto closure. 2000 psig is the P-11 permiss ve. At Unit 1 the P-'1 interlock performs this function. At Unit 2 tht function is performed at 2185 psig.

Correct. iAW 20M-6.4.lF auto close feature is 213 protection channels below 2185 psig. IAW E-0 step 12b RNO secure the 21A and C RCF's Incorrect. Wrong setpoint for auto closure. Wrong pumps for the PCV455A failure.

Incorrect. Correct setpoint for auto ciosure Wrong pumps for the PCV45SA failure.

~

Corlt'o ma f A CIOW 0, 3:eia:onS on me I'ZR PCS a& !>,

oasw m m s e prec CI om cuolio

?I m t gate !ne conseq.tri::es 01 t'lcse OIL! f x c l i n s c r opemi LI s

~ x e f l. i e s IO c o m :I K I A #

A2 C L KIA lmporlancc 3 i Exam Level S t W Level Of Difficulty. 11-51 Question Source:

\\c n Question Cognitive Level.

n gne(

Convreiitns on References provided l o Canaidale huns Technical References' E-0 slep 123 Rho. 2GM-i 4 F Oolective D:

Task IOU:

10CFRPan55Content.

lCFR 41 5 131. 4 5 ' ~ 2513,

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

88.

The Plant is operating at 100% power with all systems in NSA.

(1) What will be the status of the SSPS Rx trip relay for N42 Overpower Trip High Range?

(2) What are ALL of the applicable Reactor Trip System (RTS) Instrumentation Functions that,vi11 An Instrument Power fuse for Power Range NIS Channel 2 (N42) Blows REQUIRE Technical Specification action? (Choose from the list below)

a. Power range neutron flux -High

b. Power range neutron flux -Low

c. Power range neutron flux High positive rate

d. Overtemperature AT

e. P8 Power range neutron flux interlock

f.

P9 Power range neutron flux interlock

g. P I 0 Power range neutron flux interlock A.

(1) Tripped (2) a, c, d, e, f, and g (2) a, b, c, d, e, f, and g (2) a, c, d, e, f, and 9 (2) a, b, c, d. e, f, and 9 B.

(1) Tripped C.

(1) NOTTripped D.

(1) NOT Tripped

.iswer A

ExplanationIJustification:

A.

Correct. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a loss of contro power will cause the bistable to trip since control power is what powers the drawer. However, the bistable relay driver will input a trip for loss of eil?er power supply. The six TS actions are applicable.

Incorrect. Action 2b is not applicable since reactor power is above the P-IO interlock.

Incorrect. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a ioss of contr'il power will cause the bistable to trip since control power is what powers the drawer. However, the bistable relay driver will input a trip for l o s of either power supply. The SIX TS actions are appiicable.

incorrect. Loss of control power OR instrument power will cause the bistable to trip. It is a common misconception that only a loss of conti )I power will cause the bistable to trip since control power is what powers the drawer However. the bistable relay driver will input a trip for lo:-s of either power supply. Action 2b is not applicable since reactor power is above the P-IO interlock.

6.

C.

D.

KIA Statement Loss of instrument Power KIA Sys #

KIA System KIA Category 012 Reactor Protection System Abiljty 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:

KIA #

A2.02 KIA Importance 3.9 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

BVPS Bank #61405 Modified to make closed book Question Cognitive Level:

Higher Comprehension References provided to Candidate None Objective #:

Task ID#

Technical

References:

10 CFR Pari 55 Content:

TS Section 3.3.1 and AOP 2.2.1C Syn82tom

1. 5, LP 3SQS-2.1 slide 59 (CFR: 41.5 143.5 145.3 1 45.5)

Page 88 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

89.

The Plant is operating at 100% power with all systems in NSA.

...... No automatic actions occur.

Battery *2-2 is being charged, per maintenance request.

Emergency Switchgear Exhaust Fan 2HVZ*FN262A is running.

Emergency Switchgear Exhaust Fan ZHVZ'FN262B is in Auto.

The running Battery Room Exhaust Fan 2HVZ*FN216A TRIPS.

Annunciator A I 0-7H Battery Room Exhaust Fan Auto-Start/Auto-Stop is received.

Based on these conditions:

(1) What impact will this have on the battery rooms?

(2) IAW ARP Al0-7H, Battery Room Exhaust Fan Auto-StarUAuto-Stop, what actions are you REQUIRED to direct the crew to perform in order to address this alarm condition?

A.

(1) Oxygen concentrations will buildup.

(2) Start Emergency Switchgear Exhaust Fan 2HVZ*FN262B, (1) Hydrogen concentrations will buildup.

(2) Start Emergency Switchgear Exhaust Fan 2HVZ*FN262B (1) Oxygen concentrations will buildup.

(2) Start Battery Room Exhaust Fan 2HVZ*FN216B (1) Hydrogen concentrations will buildup.

(2) Start Battery Room Exhaust Fan 2HVZ*FN216B B.

C.

D.

4nswer D

.rplanationlJustification:

A.

Incorrect. Battery charging generates hydrogen gas NOT Oxygen gas. The excess hydrogen gas could buildup to explosive leve!S if the exhaust system (s not functioning. Starting the Emergency Switchgear Exhaust Fan ZHVZ"FN262B may seem iike a viable Soiution since the batter.

rooms are located In emergency switchgear. However. the Emergency Switchgear supply and Exhaust Fans provide fresh cool air to the emergency switchgear area and the battery room exhaust fans will pull this air into the battery room and exhaust it to outside. The procedu,ai guidance is !O start the redundant battery room exhaust fan NOT Start the redundant Emergency Switchgear Exhaust Fan.

Incorrect Right impact: wrong procedural actions, Starting the Emergency Switchgear Exhaust Fan ZHVZ'FN262B may seem like a viable solution since the battery rooms are located in emergency switchgear. However, the Emergency Switchgear supply and Exhaust Fans pro\\ de fresh cool air to the emeraencv switchaear area and the batter$ room exhaust fans will ~ u l l this air into the batter$ room and exhaust it to oi tside.

6.

The procedural guidance is to start the"redundant battery room exhaust fan NOT stat3 the redundant Emergency'switchgear Exhaust Fan Incorrect. Wrong impact: correct actions.

C.

D.

KIASys#

KIASystem KIACategory KIAStatement 063 Correct Correct impact; correct actions.

DC Electrical Distribution Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems:

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operafions:

Loss of ventilation during battery chargiiig KIA #

A2.02 KIA Importance 3.1 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided lo Candidate None Technical

References:

2OM-44F.AAH: 20M-44F.l.B page 3 Of4 Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.5 143.5 145.3 i 45.13)

Page 89 01 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

90.

The Plant is operating at 15% power with all systems in normal alignment for this power level.

SGWLC is being maintained Automatically by the SG Feedwater Bypass Control Vlvs

[2FWS*FCV479(489)(499)].

Annunciator A6-3C Station Instrument Air Receiver Tank Trouble is received.

Station Instrument Air Header Pressure is 80 psig and slowly dropping.

A local operator reports that the station instrument air dryers have malfunctioned, and bott dryers are venting.

(1) IAW AOP 2.34.1, Loss Of Station Instrument Air, what directions are you REQUIRED to give the local operator to address the degrading Station Instrument Air Header Pressure?

(2) IF Station Instrument Air Header Pressure continues to drop below 30 psig, how will the SG Feedwater Bypass Control Vlvs FAIL?

A.

(1) Place the Instrument Air Bypass filters in service THEN isolate the Instrument Air dryers (2) OPEN.

(1) Place the Instrument Air Bypass filters in service THEN isolate the Instrument Air dryers (2) CLOSED.

(1) Supply Station Instrument Air with Containment Instrument Air bv OPENING CNMT instrument Air B.

C.

backup supply Valve [21AC-MOV131] and CNMT instrument Airsupply lsol Valve [21AC-MOV1: 01 (2) OPEN.

D. (1) Supply Station Instrument Air with Containment Instrument Air by OPENING CNMT instrument >Air backup supply Valve [21AC-MOV131] and CNMT instrument Air supply lsol Valve [21AC-MOVI: 01.

(2) CLOSED.

Answer B

ExplanationIJustification:

A.

E.

C.

D.

Incorrect. Correct actions; wrong failure mode for the SG Feedwater Bypass Control Vlvs Correct. IAW AOP-2.34.1 place bypass filters in Senice and SG Feedwater Bypass Control Vlvs fail closed.

Incorrect. These are the actions for loss of containment instrument air. Containment instrument air can be supplied by station instrument ai. by opening these valves and these are the directions given in AOP 2.34.2; wrong failure mode for the SG Feedwater Bypass Control Vlvs Incorrect These are the actions for toss of containment instrument air. Containment instrument air can be supplied by station instrument ai. by opening these valves and these are the directions given in AOP 2.34.2; correct failure mode for the SG Feedwater Bypass Control Vlvs

~

KIA Sys #

KIA System KIA Category KIA Statement 078 Instrument Air Ability to (a) predict the impacts of the following malfunctions or operations on the IAS: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Air dryer and filter maifunctions KIA #

A2.01 KIA Importance 2.9 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Objective #:

Task ID#:

10 CFR Part %Content:

(CFR:41.5143.5145.3145.13)

Technical

References:

AOP 2.34.1 step 3 and NOTE prior to r,'ep 7 Page 90 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

91.

The Plant is operating at 100% power with all systems in NSA.

A Reactor Trip coincident with a loss of offsite power occurs.

Both trains of RVLIS are NOT functioning.

All other systems functioned as designed.

RCS Hot leg temperatures are 450°F and stable.

The crew is performing a Natural Circulation Cooldown IAW ES-0.4, Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS).

Throughout this procedure, the plant is depressurized in several discrete phases. AFTER each depressurization phase there is a check of pressurizer level to ensure it is less than 90%.

During this check of pressurizer level, IF Pressurizer level is greater than 90% :

(1) What directions are you REQUIRED to give the crew to address the pressurizer level situati m?

(2) What is the basis for this action?

A.

(1) Maximize letdown flow.

(2) Prevent a water solid PRZR and the resultant loss of pressure control (1) Raise RCS pressure 100 psig using PRZR Heaters.

(2) Prevent a water solid PRZR and the resultant loss of pressure control.

B.

C.

(1) Maximize letdown flow.

(2) Partially or wholly collapse the Rx vessel void (1) Raise RCS pressure 100 psig using PRZR Heaters, (2) Partially or wholly collapse the Rx vessel void.

Answer D

ExplanationlJustification:

A.

Incorrect. This is the required action and basis for high pressurizer levei while in ES-0.3 where the technique employed for the RCS c0oldo:m is dramatically different. In ES-0.3 charging and letdown are controlled throughout the cooldown to keep PRZR levei below 90%. In ES-0.4 ct83rging and letdown are set PRIOR to the cooldown and thereafter NOT adjusted. PRZR ievel rise is then used to monitor void growth.

Incorrect Right action; wrong basis.

Incorrect. Wrong action: right basis.

Correct. IAW ES-0 4 step 9 and basis.

6.

C.

D.

WA Sys #

KIA System KIA Category KIA Statement 011 Pressurizer Level Control NIA Ability to perform specific system and -

integrated plant procedures during aii r odes of plant operation.

KIA #

2.1 2 3 KIA Importance 4.4 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Technical

References:

ES-0.4 step 9 and basis; Objective #:

Task ID#:

10CFRPart55Content:

(CFR: 41.10143.5145.2145.6)

Page 91 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

92.

The Plant is operating at 100% power with all systems in NSA.

Gaseous Waste Storage Tanks [2GWS-TK25A-G] pressures are 10 psig and stable Gaseous Waste Surge Tank [ZGWS-TKZI] pressure is 62 psig and stable.

The Waste Gas Storage Tanks Radiation Monitor [2GWS-RQ104] is out of service.

Oxygen Analyzer [2GWS-OAlOOA] is out of service.

RCS Coolant activity is 25 lCi/ml.

It is desired to fill the Gaseous Waste Storage Tanks IAW 20M-19.4.G, Filling Unit 2 Gaseous Wa:.re Storage Tanks From Unit 2 Surge Tank.

While filling the Gaseous Waste Storage Tanks, under these conditions, what LRMlODCM compensatory actions are REQUIRED?

At least once per:

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; take grab samples and analyze for BOTH Oxygen concentration and radioactive content 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; take grab samples and analyze for Oxygen concentration ONLY.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; take grab samples and analyze for Oxygen concentration and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for radioactive con:ent.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; take grab samples and analyze for Oxygen concentration ONLY.

A.

B.

C.

D.

Answer A

ExplanationlJustification:

A.

Correct. IAW LRM 3.3.12 condition B.l and OOCM attachment 0 surveillance 4.1 1.2.5.1.

Incorrect At Unit 2 Both Oxygen and radioactive content must be sampled and analyzed. If the candidate does NOT correctly apply the O K M surveillance. then this distractor would appear plausible. Unit 1 does NOT have to perform this Surveillance if RCS activity is below 100 pCi'ml.

Incorrect. This oxygen sample time is the time limit if BOTH oxygen analyzers were 00s. Right actions for radioactive content.

incorrect. This oxygen sample time is the time limit if BOTH oxygen analyzers were 00s. If the candidate does NOT correctly apply the OClCM Surveillance, then this distractor would appear plausible. Unit 1 does NOT have to perform this surveillance if RCS activity is below 100 VCi ml.

2.

D.

~-

KIA Sys #

KIA System KIA Category KIA Statement 071 Waste Gas Disposal NIA Ability to interpret and execute procedure steps.

KIA #

2.1.20 KIA Importance 4.6 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Appiication References provided to Candidate Objective #:

Task ID#:

10CFR Part 55Content:

(CFR: 41.10143.5145.12)

% ODCM section 2.02 ~ LRM 3.3.12 Technical

References:

LRM 3.3.12 condition 0.1 and %-ODC-:l 03 attachment 0 surveillance 4.1 1.2.5.1 Page 92 Of 100

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

93.

The Plant is in Mode 5.

Train A is the declared protected Train.

System Station Transformer 2A is supplying the 2A and 2AE 4KV Buses.

The Deluge valve for System Station Transformer 2A inadvertently actuates and sprays thi transformer.

Operators locally isolate the Deluge valve.

System Station Transformer 2A remains in service.

IAW %-ADM-1900, Fire Protection Program, what actions, if any, are REQUIRED for the isolated Deluge valve?

A.

No actions required.

B.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish an hourly fire watch patrol with backup fire suppression capability and estal: ish controls to prohibit transient combustibles in the affected area.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish an hourly fire watch patrol in the affected area with backup fire suppression capability, and to check for proper cooling, no oil leakage, or any abnormal conditions.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, establish an hourly fire watch patrol in the affected area to check for proper cooling, no oil leakage, or any abnormal conditions.

C.

D.

Answer D

ExplanationlJustification:

A.

B.

Incorrect. Candidate may think that since the 2A transform IS not safety related that no actions are required.

Incorrect These are the required actions for safety related equipment that is protected by the C02 system.

Incorrect. This is a combination of CO21water and safetylnon-safety related actions.

Correct. IAW Att B of %-ADM-lSflO item 3b for non-safety related equipment that IS required to be operable. Candidate must realize that t t, ? 2A transformer is non-safety related and is required to be operable in Mode 5 with Train A protected.

_I

~

KIA Sys #

KIA System KIA Category KIA Statement 086 Fire Protection NIA Ability to evaluate plant performance ar 1 make operational judgments based on operating characteristics. reactor behai or, and instrument interpretation.

KIA #

2.1.7 KIA Importance 4.7 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Application References provided to Candidate

%-ADM-lgOfl Technical

References:

!&AOM-l9fl0 Attachment B item 3b Objective #:

Task I D #

10 CFR Pari 55 Content:

(CFR: 41.51 43.5 145.12 145.13)

Page 93 Of 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

94.

You are a Licensed Senior Reactor Operator at Beaver Valley.

You have been Licensed for Five and one-half years.

Your License renewal medical examination (NRC Form 396, Certification Of Medical Examination By Facility Licensee) is due to the NRC Regional Administrator in 6 months.

Your License is "Active" and you are currently assigned as the Unit 2 Control Room Shift Manager.

Your License contains NO medical restrictions.

You have been experiencing some difficulties with your "distant" vision.

On your first relief day, youf personal physician (a licensed optometrist) determines that your "dista.it" vision has permanently degraded and you will NOW be required to wear corrective lenses at all tim?s IAW 10CFR 50.74, Notification of Change In Operator or Senior Operator Status, when are you arc REQUIRED to notify the NRC Regional Administrator of this change in your medical status?

A.

Immediately B.

C.

D.

Answer C

ExplanationlJustification:

A...

D.

KIASys#

KIASystem KIA Category KIA Statement NIA Generic Conduct Of Operations Knowledge of individual licensed operator Prior to assuming your next shift.

Within 30 days of the diagnosis.

Within 60 days of the diagnosis, incorrect. You should immediately begin wearing the corrective lenses, but not required to reportfor 30 days.

Incorrect. You should begin wearing the corrective lenses prior to your next shift. but not required to report for 30 days Correct. IAW IOCFR 50 74, 55.25, 55.23, NRC form 396. (Beaver Valley specific OE CR 07-2231 1)

Incorrect. Must be within 30 days.

responsibilities related to shift staffing. iuch as medical requirements "no-solo" opeaation, maintenance of active license status, 10CFR55, etc.

KIA U 2.1.4 KIA Importance 3.8 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental References provided l o Candidate None Technical

References:

IOCFR 50.74, 55.25. 55.23, NRC form I96 Objective #:

Task ID#:

10 CFR Part %Content:

(CFR: 41.10l43.2)

Page 94 01 '00

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

95.

The Plant is operating at 100% steady state power with all systems in NSA.

The shift chemist reports the following STABLE Primary and Secondary plant chemistry conditions-Secondary Specific Activity is 0.025 lCi/gm DOSE EQUIVALENT 1-131 RCS Specific Activity is 25.0 vCi/gm DOSE EQUIVALENT 1-131 RCS Dissolved Oxygen is 0.15 ppm RCS Chlorides are 0.10 ppm RCS Fluorides are 0.10 ppm Based on these Chemistry conditions, what Technical SpecificationlLRM actions are REQUIRED?

Within:

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be in Mode 3 and within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> be in Mode 5 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> be in Mode 3 with Tavg less than 500°F.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore DOSE EQUIVALENT 1-131 to within its limit.

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> restore RCS Dissolved Oxygen to within the steady state limit A.

B.

C.

D.

Answer B

ExplanationlJ ustification:

A.

Incorrect. These are the required actions if Oxygen, chloride, or fluoride, are outside their transient limits. Oxygen is outside the steady statr! limit but within the transient limit.

B.

Correct. IAW TS 3.4.16 condition C 1 since DOSE EQUIVALENT 1-131 IS in the unacceptable region of TS iigure 3.4.16-1.

C.

Incorrect Time limit is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> NOT 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

incorrect. Time limit is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> NOT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

~~~

A Sys #

WA System KIA Category KIA Statement chemistly iimits.

NIA Generic Conduct Of Operations Knowledge of pnmary and secondary piant KIA #

2 1.34 KIA Importance 3.5 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Application References provided to Candidate TS 3 4.16 condition C.l and Objective #:

Task ID#:

10 CFR Part 55 Content:

(CFR: 41.10 i 43.5 i 45.12)

TS 3.4 16, 3.7.13; LRM 3.4.2 Technical

References:

TS figure 3.4.16-1 Page 95 Of 00

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY

96.

The Plant is operating at 100% power with all systems in NSA All PORVs and associated block valves are OPERABLE.

. Technical Specification LCO 3.4.1 1 requires each PORV and associated block valve to be OPERABLE.

. Every 92 days, Surveillance 20ST-6.6. PORV Isolation Valve Test and Position Check, is performed to meet this requirement.

While performing 20ST-6.6:

2RCSMOV535 PORV Motor Operated lsol Vlv CLOSES but WILL NOT OPEN.

Maintenance finds a bad power supply breaker to the MOV, and replaces the entire breaker assembly at the MCC.

ALL of their required work package instructions have been completed.

The tagout has been lifted, 2RCS*MOV535 is ENERGIZED and CLOSED.

2RCSMOV535 is ready for operations post-maintenance testing.

For these conditions:

What MINIMUM post-maintenance testing will be REQUIRED to verify compliance with Technical Specification LCO 3.4.1 I ?

(For each of the below actions, assume all valve stroke times and indications are within acceptable limits)

Open 2RCS*MOV535, no other actions required Open 2RCS*MOV535; then Close; then re-open Cycle the associated PORV through one complete cycle, then open 2RCS*MOV535.

Cycle the associated PORV through one complete cycle, then open 2RCS*MOV535; then Close; then re-open.

A.

B.

D.

Answer B

ExplanationiJustification:

A.

Incorrect. The surveillance requirement is for a complete cycle. Opening the valve would ONLY meet half of a cycle. if the candidate beiievcd that the other half was satisfactorily performed earlier, then the candidate would select this choice. Since maintenance was on the breaker the v.!lve must be again cycled through a complete cycle (open and closed)

Correct. The surveillance requirement is for a complete cycle. Since maintenance was on the breaker the valve must be again cycied throu!:h a complete cycle (open and closed)

Incorrect. Since maintenance was on the breaker the valve must be again cycled through a complete cycle (open and closed). The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any Surveillance activities for the PORV.

Incorrect. Right actions for the block valve. Wrong actions for the PORV. The LCO addresses both the PORV and the block valve, but maintenance was only performed on the block valve. NO requirement to perform any surveiiiance activities for the PORV.

B.

C.

D.

~

KIA Sys tt KIA System KIA Category KIA Statement NIA Generic Equipment Control Knowledge of pre-and post-rnaintenani 3 operability requirements.

KIA It 2.2.21 KIA Importance 4.1 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Analysis References provided to Candidate None Objective #:

Task ID#:

Technical

References:

Technical Specification 3.4.11 SR 3.4.1 1.1; 20ST-6.6 Acceptance Criteria page 5 a Id 6 10 CFR Part 55 Content:

(CFR: 41.10 i 43.2)

P a m 96 0, 30

97.

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY The Plant is operating at 100% power with all systems in NSA.

A turbine runback occurs.

All systems respond as designed.

The crew is stabilizing the plant in accordance with the appropriate procedure.

Control Bank "D" Group Counters are at 180 steps.

On DRPI, one Control Bank "D" rod indicates 196 steps; all other rods indicate 182 steps.

The affected rod has a blown movable gripper fuse and has been determined to be trippable Power stabilizes at 85%

Which ONE (1) of the following describes the Technical Specification implications of this event?

A.

The rod is INOPERABLE AND NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained The rod is INOPERABLE AND NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

The rod is OPERABLE, BUT NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure acceptable power distribution limits are maintained.

The rod is OPERABLE, BUT NOT within alignment limits; Realign the rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to ensure Shutdown Margin is maintained.

B.

C.

D.

Answer C

planationlJustification:

incorrect. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1s required by T.S. 3.1.4 Condition A, but rod is not inoperable if it is trippable. If tne rod was untrippable, then SDM would be affected Power distribution limits are the correct reason. Common misconception is that a rod is INOPERABLE if it is misaligned. This misconception stems for the OLD Technical Specincations where misaligned rods WERE INOPERABLE.

incorrect. Would be true if the rod was untrippable Correct. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required by T.S. 3.1.4 Condition B. Misalignment iimits are based on impact on power distribution limits.

Incorrect. Correct call on operabiiity, but the concern for the situation presented is not shutdown margin E.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement NiA Generic Equipment Control Ability to determine operability andior KIA #

2.2.37 KIA importance 4.6 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

Bank 1LOT7 NRC Exam Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

TS 3.1.4, Condition B. and basis Objective #:

Task ID#

10 CFR Part55Content:

(CFR: 41.7i43.5l45.12)

~~

availability of safety related equipment.

Page 97 01 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY

98.

The Plant is operating at 100% power with all systems in NSA.

A Large break LOCA occurs inside containment coincident with a loss of offsite power.

Rx Trip, SI, CIA and CIB all actuate as designed.

2-2 Emergency Generator TRIPS and ALL Train B AC Equipment de-energizes.

As Shift manager, you are assessing the EAL Fission Product Barrier Matrix Which of the below listed Radiation Monitors will be the ONLY available monitors for this assessment?

1. Control Room Area [2RMC*RQ201]

2. Control Room Area [2RMC*RQ202]

3. In-Containment High Range Area [2RMR*RQ206]

4. In-Containment High Range Area [2RMR*RQ207]

5. Recirc Spray Heat Exchanger [2SWS*RQ1100A]

6. Recirc Spray Heat Exchanger [2SWS*RQIlOOB]

7. Recirc Spray Heat Exchanger [2SWS*RQIlOOC]

8. Recirc Spray Heat Exchanger [2SWS*RQI100D]

9. Containment Purge [2HVR*RQ104A]

IO. Containment Purge [2HVR*RQ104B]

A.

1, 3, 5, 6, & 9 B.

2, 4, 6, 8, & 10

c.

2, 4, 5, & 7

2.

1, 3, 5, & 7 Answer D

ExplanationlJustification:

A.

B.

C.

D.

WA Sys #

KIA System KIA Category KIA Statement NIA Generic Radiation Control Knowledge of radiation monitoring sys ?ms, such as fixed radiation monitors and al.irms, portable sutvey instruments. personne monitoring equipment, etc.

incorrect. 2SWS'RQllOOB has no power and 2HVR'RQ104A is not part of the fission product harrier matrix.

Incorrect. 2RMC'RQ202,2RMR'RQ207, 2SWS'RQ1100B & ZSWS*RQIlOOD have no power and 2HVR'RQ104A is not part of the fission iroduct harrier matrix.

Incorrect. ZRMC'RQ202 and ZRMR'RQ207 have no power.

Correct All of these have power and are a part of the fission product barrier matrix.

WA #

2.3.15 KIAlmporlance 3.1 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Higher Comprehension References provided to Candidate None Technical

References:

EAL Fission product harrier matrix; 20'A-Objective #:

Task ID#

10 CFR Part 55 Content:

(CFR: 41.12 i 43.4 i 45.9) 43.3.C page 3

Beaver Valley Unit 2 NRC Written Exam (ZLOT~)

SRO ONLY 99 Throughout the EOP network there are numerous steps that address SI reset PRIOR to these steps IS a "Caution" that reads as follows:

"If offsite power is lost after SI reset, manual action may be required to restart safeguards equipme-it" What is the basis of this "Caution"?

It is a reminder that normal sequencing of safeguards loads onto the emergency bus after diesel generator startup -(I) occur AND a blackout sequencer (bus undervoltage) actuation possible.

A.

(1) will NOT (2) is B.

(1) will NOT (2) is NOT C.

(1) will (2) is D.

(1) will (2) is NOT Answer A

planationlJustification:

Correct. IAW ES-1.1 step 1 caution 1 basis. In order to correctly answer this question, the candidate must understand the workings of the I 'G sequencer Once SI has been reset, any of the equipment that was started due to the SI that is then secured, will not receive a sequencer,.tart signal UNLESS the equipment is also on the "blackour sequencer.

Incorrect. Right normal response; wrong blackout response.

Incorrect. Wrong normal response right blackout response.

Incorrect. Wrong normal response: wrong blackout response.

B.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement NIA Generic Emergency ProcedureslPlan Knowledge of the operational implicatic 7s of KIA #

2.4 20 KIA Importance 4.3 Exam Level SRO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower Fundamental.

References provided to Candidate None Technical

References:

ES-1.1 step 1 caution 1 basis.

Objective #:

Task I D #

10 CFRPart 55Content:

(CFR: 41.10143.5145.13)

EOP warnings, cautions, and notes.

Page 99 or 100

Beaver Valley Unit 2 NRC Written Exam ( 2 ~ 0 ~ 6 )

SRO ONLY 100.

A large Steam break accident inside containment has occurred.

Containment pressure peaked at 20 psig.

All Equipment functioned as designed EXCEPT all seal injection flow has been lost.

SI, CIA, and CIB have all been reset.

SWS has been restored to the CCP heat exchangers.

CCP flow has been restored.

While performing EOP Attachment A-I 2, Establishing RCP CCP Cooling and Seal Injectim, the Reactor Operator is unable to "OPEN" 21A RCP Thermal Barrier Outlet lsol Vlv

[2CCP*AOV107A], using the benchboard control switch.

In order to "OPEN" 21A RCP Thermal Barrier Outlet Is01 Vlv [2CCP*AOV107A] it will be necessary :o defeat the "CLOSE" signal to 21A RCP Thermal Barrier Outlet lsol Vlv [2CCP*AOV107A].

IAW EOP Attachment A-1.2, Establishing RCP CCP Cooling and Seal Injection:

What directions are you REQUIRED to give the local operator to defeat the "CLOSE" signal to 21A RCP Thermal Barrier Outlet Is01 Vlv [2CCP*AOV107A]?

A.

B.

C.

D.

Install jumpers across the opening contacts of the valve's control circuit Remove the valve's associated secondary process rack power supply card Remove the valve's associated control circuit power supply fuse Install jumpers across the contacts of the high discharge flow transmitter.

swer B

ExplanationIJustification:

A.

6.

C.

D.

KIA Sys #

KIA System KIA Category KIA Statement NIA Generic Emergency ProceduresiPlan Knowledge of local auxilialy operator t..sks Incorrect. Although this may open the valve, it is NOT IAW EOP attachment A-1.2.

Correct iAW EOP attachment A-1.2 step 4.a.3.

Incorrect. Tnis action will fail the valve closed.

incorrect. This action will only defeat the high flow signal 8UT NOT the high pressure and it is NOT IAW EOP attachment A-1.2 during an emergency and the resultant operational effects.

KlA #

2 4.35 KIA Importance 4.0 Exam Level SKO Level Of Difficulty: (1-5)

Question Source:

New Question Cognitive Level:

Lower References provided to Candidate None Technical

References:

EOP attachment A-1.2 step 4.b.3.

Objective #:

Task ID#:

10CFR Part55 Content:

(CFR: 41.10l43.5i45.13)

Fundamental