RO 2 Group# 1 KIA# Importance Rating National Grid.A Incorrect-A control rod withdrawal block is active due to SRM A being

ES-401 Written Examination Question W,o.:;.:....::rk.::..:s:..:..h:..::e..:.e.:..t ___ _::F....::o:...:.r.:..:m.;....;;;;E..:.S_-4:....::0....::1---=--5

Examination Outline Cross-Reference:

AC Electrical Distribution

Level Tier# Group# KIA# Importance Rating

RO 2 1 262001 K1.03 3.4

Knowledge of the physical connections and/or cause-effect relationships between A. C. ELECTRICAL DISTRIBUTION and the following: Off-site power sources

Proposed Question: #1

The plant is operating at 100% power with the following:

• MOD-10017, NORTH-SOUTH 115KV BUS DISC SW, is open due to a request from National Grid.

• The rest of the electrical distribution system is in a normal alignment for 100% power. • Then, Line 3, FITZPATRICK- LIGHTHOUSE HILL, de-energizes.

Which one of the following describes the impact of the Line 3 loss on the electrical distribution system?

A. Bus 10300 de-energizes.

B. Bus 10400 de-energizes.

C. Bus 10300 remains energized, but loses its alternate power source.

D. Bus 10400 remains energized, but loses its alternate power source.

Proposed Answer: D

Explanation: With MOD-1 0017 open, the primary windings of T2 receive power from Line 3 alone. Therefore on a loss of Line 3, T2 will de-energize. This will cause a loss of the reserve supply to Buses 10200 and 10400. However, at 100% power, these buses are powered from house service transformer T4. Therefore Bus 10400 will rE~main energized.

A. Incorrect- With MOD-1 0017 open, Bus 10300 is affected by off-site power Line 4, not Line 3.

B. Incorrect- A normal alignment at 100% power has Bus 10400 powered from house service transformer T 4, not T2.

C. Incorrect- With MOD-1 0017 open, Bus 10300 is affected by off-site power Line 4, not Line 3.

Technical Reference(s): SDLP-710 Figure 1, OP-44, OP-46A

Proposed references to be provided to applicants during examination: None

Learning Objective: SDLP-71 D 1.09.a

Question Source: New

Question History:

Question Cognitive Level: Comprehension or Analysis

10 CFR Part 55 Content: 55.41 (4)

Comments:

ES-401 Written Examination Question Wo::..:r:...:..k:.::s~h=e.::.et=--------=-F-=o:..:..r::..:.m:....:E::.S.::.-....;4:....:0...:.1....:.-5


PCIS/Nuclear Steam Supply Shutoff


RO 2 1 223002 K1.19 2.7

Knowledge of the physical connections and/or cause-effE~ct relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following: Component cooling water systems


Given the following primary containment isolation valves in th1e RBCLC system:

• 15AOV-132A(B), RBCLC to A(B) Recirc Pump and Motor Coolers Inlet Isolation Valve • 15AOV-133A(B), RBCLC to A(B) Recirc Pump and Motor Coolers Outlet Isolation Valve

Which one of the following describes the operation of thesE~ valves in the event of a high Drywell pressure condition?

If Drywell pressure exceeds 2.7 psig, these valves ...

A. automatically close by venting off instrument air.

B. automatically close by venting off instrument nitrogen.

C. remain open, unless manually closed by venting off instrument air.

D. remain open, unless manually closed by venting off instrument nitrogen.

Proposed Answer: D

Explanation: The nine primary containment isolation valves in the RBCLC system are all pneumatically operated by instrument nitrogen. These valves are provided to allow isolation of the RBCLC lines that penetrate the primary containment in th,e event of a LOCA and concurrent RBCLC leak in the primary containment. However, these valves do not automatically close on a high Drywell pressure isolation signal (2. 7 psig).

Note: The question meets the KIA by testing knowledge of how component cooling water primary containment isolation valves respond to an automatic primary containment isolation signal (ie cause and effect between these two systems).

A. Incorrect -These valves are operated with instrument nitrogen, not instrument air. These valves do not automatically close on a high Drywell pressure isolation signal (2. 7 psig).

B. Incorrect- These valves do not automatically close on a high Drywell pressure isolation signal (2. 7 psig).

C. Incorrect- These valves are operated with instrument nitrogen, not instrument air.

Technical Reference(s): SDLP-15, AOP-15, OP-40


Learning Objective: SDLP-151.05.a.6


Question History:

Question Cognitive Level: Memory or Fundamental Knowledge

1 0 CFR Part 55 Content: 55.41 (9)

Comments: TRH 3/6/14- Added KIA statement to explanation, based on NRC comment.

ES-401 Written Examination Question Worksheet

Examination Outline Cross-Reference: RO 2 1

Form ES-401-5


203000 K2.02 2.5

RHRILPCI: Injection Mode

Knowledge of electrical power supplies to the following: Valves



• MCC-155 is de-energized due to an electrical fault. • MCC-161 is de-energized due to a malfunction of the supply breaker.

Which one of the following describes the ability to align RHR loop A for LPCI and Drywell spray from the Control Room?

1\ i.C ~\_-t.~ +-"~c.\-,~"~ ~UW'I\t 'f\•C..'t'.'SSIVf

RHR loop A... ~ -"! _.. 'i 1~ "3 /''f A can NOT be aligned from the Control Room for either LPCI or Drywell spray. t)l 'f(l-.~'1

B. can be aligned from the Control Room for Drywell spray, but NOT LPCI.

C. can be aligned from the Control Room for LPCI, but NOT Drywell spray.

D. can be aligned from the Control Room for both LPCI and Drywell spray.

Proposed Answer: B

Explanation: MCC-155 provides power for the RHR loop A LPCI injection valves. Therefore with MCC-155 de-energized, the LPCI inboard injection valve~ (normally closed) cannot be opened from the Control Room to align LPCI. MCC-161 provides power for the RHR loop B Drywell spray valve, but does not affect RHR loop A Neither of the given faults prevent use of RHR loop A for Drywell spray.

A Incorrect- RHR loop A can be aligned from the Control Room for Drywell spray. C. Incorrect- RHR loop A cannot be aligned from the Control Room for LPCI. RHR loop A can

be aligned from the Control Room for Drywell spray. D. Incorrect- RHR loop A cannot be aligned from the Control Room for LPCI.

Technical Reference(s): OP-13, SDLP-10


Learning Objective: SDLP-1 0 1.04.a/c


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5


RPS

------------------------Level Tier# Group# KIA# Importance Rating

RO 2 1 212000 K2.01 3.2

Knowledge of electrical power supplies to the following: RPS motor-generator sets


Which one of the following identifies the power supply to 71 RP-1 B, RPS MG Set B?

A. MCC-251

B. MCC-252

C. MCC-261

D. MCC-262

Proposed Answer: c

Explanation: RPS MG Set B is supplied power from MCC-261.

A Incorrect- This is the power supply for RPS MG Set A, not B. B. Incorrect- This is the power supply for the alternate supply transformer to RPS System A D. Incorrect- This is the power supply for the alternate supply transformer to RPS System B.

Technical Reference(s): OP-18


Learning Objective: SDLP-05 1.03.b


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 --~~~-----------------


Source Range Monitor

Level Tier# Group# KIA# Importance Ri3ting

RO 2 1 215004 K3.02 3.4

Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR {SRM) SYSTEM will have on following: Reactor manual C4)ntrol: Plant-Specific


A plant startup is in progress with the following:

• The Reactor Mode Switch is in START & HOT STBY. • AIIIRMs are on Range 2. • All SRMs are partially withdrawn. • While an RO is re-positioning SRMs, a malfunction of the SRM drive control circuitry

results in the following indications:

o SRM A: 60 cps o SRM 8: 7 x 1 03 cps o SRM C: 300 cps o SRM D: 5 x 104 cps

Which one of the following describes the status of the SRM rod block?

The SRM rod block is ...

A. NOT received.

B. received due to an SRM Hi Flux signal.

C. received due to an SRM Downscale signal.

D. received due to an SRM Detector Not Full In signal.

Proposed Answer: D

Explanation: The SRM Detector Not Full In rod block occurs when an SRM is not fully inserted with< 100 cps, associated IRMs on range 1 or 2, and the Mode Switch is not in RUN. This prevents control rod withdrawal with RMCS under the given eonditions due to SRM A

A Incorrect- A control rod withdrawal block is active due to SRM A being <1 00 cps and partially withdrawn.

B. Incorrect- SRM Hi Flux rod block occurs with SRM counts above 105 cps. All SRMs are less than 1 05 cps.

C. Incorrect- SRM Downscale rod block occurs with SRM counts at 3 cps or less. All SRMs are above 3 cps.

Technical Reference(s): OP-16, SDLP-07B


Learning Objective: SDLP-07B 1.05.c.1

Question Source: Modified Bank- 2010 NRC #3Ei

Question History:



Comments: TRH 3/6/14- Changed SRM counts from 130 to 300 in ste!m, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 ·--------------------------Examination Outline Cross-Reference:

Reactor Water Level Control


RO 2 1 259002 K3.05 2.8

Knowledge of the effect that a loss or malfunction of the REACTOR WATER LEVEL CONTROL SYSTEM will have on following: Recirculation flow control system



• Both Recirc pumps are operating at approximately 50% speed. • Feedwater pump A is operating. • Feedwater pump B is shut down.

Then, a malfunction in the Feedwater Level Control system relsults in the following:

• Feedwater pump A speed begins to oscillate. • Feedwater flow oscillates between 4.8 X 1 as lbm/hr and 5.6 X 1 as Ibm/hr. • Reactor water level oscillates between 195" and 205".

Which one of the following describes the response of Recirc pump speed demands?

Recirc pump speed demands ...

A. run back to Speed Demand Limiter #1 (30%).

B. runback to Speed Demand Limiter #2 (44%).

C. oscillates in response to the Feedwater pump oscillation.

D. remain the same because all parameters remain abov,e runback setpoints.

Proposed Answer: B

Explanation: Speed Demand Limiter #2 runs Recirculation pump speed demand back to 44% because Reactor water level goes less than 196.5" with only one Feedwater pump operating.

A. Incorrect- Speed Demand Limiter #1 runs speed demand to 30% if Feedwater flow lowers below 20% of rated (approximately 2 x 106 lbm/hr) or the Recirculation pump discharge valve leaves the full open position. Neither of these conditions are met in this question.

C. Incorrect- RWR demand signal does not \will not oscillate with Feedwater flow oscillaitions. D. Incorrect- Reactor water level lowers below the 196.5" setpoint for Speed Demand Limiter

#2.



Learning Objective: SDLP-021 1.05.c.2


Question History:



Comments:



SGTS


RO 2 1 261000 K4.01 3.7

Form ES-401-5

Knowledge of STANDBY GAS TREATMENT SYSTEM des;ign feature(s) and/or interlocks which provide for the following: Automatic system initia1tion



• Annunciator 09-3-2-40, RX BLDG VENTRAD MON HI-HI, is received. • Reactor Building Ventilation exhaust radiation monitOI"S indicate:

o 17RM-452A: 3 x 104 cpm o 17RM-452B: 5 x 103 cpm

Which one of the following describes the expected status of the Standby Gas Treatment (SGT) system?

A. Both trains of SGT remain in standby.

B. SGT train A auto-initiates. SGT train B remains in standby.

C. SGT train B auto-initiates. SGT train A remains in standby.

D. Both trains of SGT auto-initiate.

Proposed Answer: 8

Explanation: 17RM-452A is above the hi-hi setpoint of 1 x 104 cpm and 17RM-4528 is below the hi-hi setpoint. Each of these radiation monitors causes auto-initiation of their respective SGT train when above the hi-hi setpoint. Therefore, only SGT train A auto-initiates and SGT train 8 remains in standby.

A. Incorrect- SGT train A auto-initiates due to hi-hi radiation. C. Incorrect- SGT train 8 remains in standby. SGT train A auto-initiates due to hi-hi radiation. D. Incorrect- SGT train 8 remains in standby.

Technical Reference(s): ARP 09-3-2-40, OP-:20, AOP-15, SDLP-01 8


Learning Objective: SDLP-01 8 1.05.c.1


Question History:



Comments:



EDGs


RO 2 1 264000 K4.02 4.0

Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Emergency generator trips (emergency/LOCA)


Which one of the following lists two parameters that each pmvide an Emergency Diesel Generator (EDG) trip signal following an EDG start due to a sustained high Drywall pressure?

A. Engine overspeed and loss of generator field

B. Loss of generator field and low lube oil pressure

C. High jacket water temperature and engine overs peed

D. Low lube oil pressure and high jacket water temperature

Proposed Answer: A

Explanation: Engine overs peed and loss of generator field provide EDG trip signals even with a sealed in high Drywell pressure (LOCA start) signal. Both low lube oil pressure and high jacket water temperature provide EDG trip signals on a non-LOCA start, but are bypassed with high Drywell pressure present.

B. Incorrect- Low lube oil pressure provides an EDG trip si~1nal on a non-LOCA start, but is bypassed with high Drywell pressure present.

C. Incorrect- High jacket water temperature provides an EDG trip signal on a non-LOCA start, but is bypassed with high Drywell pressure present.

D. Incorrect- Low lube oil pressure and high jacket water temperature provide EDG trip signals on a non-LOCA start, but are bypassed with high Drywell pressure present.

Technical Reference(s): ARP-09-8-2-9, OP-2~~. SDLP-93


Learning Objective: SDLP-93 1.05.c.1.g


Question History:

Question Cognitive Level: Memory or Fundamental Knowll3dge


Comments: TRH 3/10/14- Revised to get rid of high crankcase pressure as an answer choice, based on NRC comment.



DC Electrical Distribution


RO 2 1 263000 K5.01 2.6

Form ES-401-5

Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION: Hydrogen generation during battery charging.



• An equalizing charge is in progress on Station Battery A. • Battery Room Ventilation is lost. • Battery Room temperature is 80°F.

Which one of the following describes how to control the equalizing charge and the associated reason, in accordance with OP-43A, 125 VDC Power System?

The equalizing charge ...

A. must be secured to prevent excessive temperature rise.

B. must be secured to prevent excessive hydrogen gas buildup.

C. may continue as long as Battery Room temperature remains below 100°F.

D. may continue as long as portable air samples indicate less than 0.6% hydrogen.

Proposed Answer: 8

Explanation: OP-43A precaution C.2.1 states, "Equalizing charges shall only be applied to batteries while battery room ventilation is operating to prevent hydrogen buildup." Therefore, with Battery Room Ventilation out of service, the equalizing c:harge must NOT continue to be applied.

A. Incorrect- The equalizing charge must be secured, but tl1e concern is hydrogen generation, not temperature rise.

C. Incorrect- The equalizing charge must be secured due to concern over hydrogen generation. 1 00°F is based on the upper limit for Batte~ry room air temperature in OP-43A section E.

D. Incorrect- The equalizing charge must be secured due to concern over hydrogen generation. 0.6% is based a threshold hydrogen concentration used in EOP-4.

Technical Reference(s): OP-43A


Learning Objective: SDLP-718 1.13.a


Question History:


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

Comments:


Examination Outline Cross-Reference: Level Tier# Group# KIA#

RO 2

Importance Rating

ADS

1 218000 K5.01 3.8

Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic operation

Proposed Question: #1 0

The plant was operating at 1 00% power when a loss of coolant accident resulted in the following:

Time

10 seconds

40 seconds

144 seconds

155 seconds

Condition(s)/Event(s)

• • • •

• •

•

•

Reactor water level is 59 inches and lowering . A failure prevented the auto sta rt of ALL ECCS pumps. Annunciator 09-4-1-28, ADS Tl MERS ACTUATED, is received. ADS Normal/Override switches on Panel 09-4 are in the NORMAL position.

Reactor water level is 30 inches and slowly lowering . Core Spray pump A is manually started and is running on minimum

>f 130 psig. flow with a discharge pressure c

Reactor water level is 40 inches. and slowly rising .

Reactor water level is 60 inches and slowly rising .

Which one of the following describes the status of the ADS valves at time 190 seconds?

At time 190 seconds, the ADS valves are ...

A. closed because the ADS timers reset before the valves opened.

B. closed because the low pressure ECCS pump running logic is not satisfied.

C. open but will automatically close if Reactor water level rises above 177 inches.

D. open and will remain open until manual action is tak:en to reset the ADS signal.

Proposed Answer: D

Explanation: At 10 seconds, the ADS timers actuate due to Reactor water level being below 59.5 inches (with confirmatory level being below 177 inches). The timers count down for 134 seconds. By time 144 seconds, the timers will have timed out since Reactor water level is still below 59.5 inches. When the timers time out, with at least one Core Spray pump running and discharge pressure above 1 00 psig, the ADS valves open. The ADS valves remain open even if Reactor water level rises above the ADS setpoint. ADS must then be manually reset for the valves to close.

A. Incorrect- The ADS timers finished timing at approximately 144 seconds and Reactor water level was not above the ADS setpoint by this time.

B. Incorrect -A single Core Spray pump running with discharge pressure above 100 psig satisfies the low pressure ECCS pump running logic and allows the ADS valves to open at approximately time 144 seconds.

C. Incorrect- Once the ADS timers finish timing, Reactor water level rising above the ADS setpoint will not close the ADS valves. Manual action must be taken to reset ADS in this situation.

Technical Reference(s): SDLP-02J


Learning Objective: SDLP-02J 1.05.c.1

Question Source: Bank- 2010 NRC #38

Question History:



Comments: TRH 3/6/14- Added "loss of' before "coolant accident" in first sentence, based on NRC comment.



Form ES-401-5


400000 K6.05 2.8

Component Cooling Water

Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Motors



• RBCLC pumps A and B are running. • RBCLC pump C is tagged out for maintenance.

Then, RBCLC pump A trips on motor overcurrent. RBCLC pump B trips on motor overcurrent one minute later.

Which one of the following describes the response of the Emergency Service Water (ESW) system, if any, and the need for a manual Reactor scram, in accordance with AOP-11, Loss of Reactor Building Closed Loop Cooling?

A. ESW remains in standby. A manual Reactor scram is required.

B. ESW remains in standby. A manual Reactor scram is NOT required.

C. ESW automatically starts and injects into the RBCLC system. A manual Reactor scram is required.

D. ESW automatically starts and injects into the RBCLC system. A manual Reactor scram is NOT required.

Proposed Answer: c

Explanation: With RBCLC pump C tagged out and both RBCLC pump A and 8 tripped, RBCLC pressure will lower below 40 psig. This causes ESW to automatically inject into RBCLC. A manual scram is required by AOP-11 due to trip of all RBCLC pumps.

A. Incorrect- With no RBCLC pumps running, RBCLC pressure lowers below 40 psig. This causes ESW to automatically inject into RBCLC.

B. Incorrect- With no RBCLC pumps running, RBCLC pressure lowers below 40 psig. This causes ESW to automatically inject into RBCLC. A me~nual scram is required by AOP-11 due to trip of all RBCLC pumps.

D. Incorrect- A manual scram is required by AOP-11 due to trip of all RBCLC pumps.

Technical Reference(s): AOP-11


Learning Objective: SDLP-15 1.05.c.5 and 1.15.c:t


Question History:


10 CFR Part 55 Content: 55.41(4)

Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~--------~--------


SLC


RO 2 1 211000 K6.03 3.2

Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM: A.C. power


An A TWS has occurred with the following:

• The 10500 bus has de-energized due to a fault. • MCC-162 has de-energized due to a fault. • Boron injection is required.

Which one of the following describes the action required to inject boron into the Reactor?

A Start Standby Liquid Control pump A

B. Start Standby Liquid Control pump B.

C. Use EP-4, Boron Injection Using CRD System, to inject with Control Rod Drive pump A

D. Use EP-4, Boron Injection Using CRD System, to inject with Control Rod Drive pump B.

Proposed Answer: D

Explanation: With the 10500 bus de-energized due to a fault, L-15 and MCC-152 are also deenergized. SLC pump A is unavailable due to the loss of MCC-152. Additionally, SLC pump 8 is unavailable due to the loss of MCC-162. With no SLC pump available and boron injection required, EOP-3 directs use of EP-4 to inject boron using a CRD pump. CRD pump A is unavailable due to the loss of L-15. Therefore, CRD pump 8 must be used to inject boron.

A. Incorrect- SLC pump A is unavailable due to the loss of IMCC-152. B. Incorrect- SLC pump 8 is unavailable due to the loss of IMCC-162. C. Incorrect- CRD pump A is unavailable due to the loss of L-15.

Technical Reference(s): OP-17, OP-25, EOP-3, I=P-4


Learning Objective: SDLP-11 1.04.c


Question History:



Comments:

ES-401 Written Examination Question W....::o..:...;rk:..:.:s:..:.h::..:e....;:e..:..t ___ ....:..F....::o..:...;rm:..:..:....:E::..:S::....-4....:....:....01..:..-...:..5


UPS (AC/DC)


RO 2 1 262002 K4.01 3.1

Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.G./D.C.) design feature(s) and/or interlocks which provide for the following: Transfer from preferred power to alternate power supplies



• Breaker 12602, L26 600V FOR BKR, trips. • The UPS (71 UPS-1) DC to AC Inverter output fails to zero.

Which one of the following describes the response of the 7'1ACUPS Distribution Bus?

The 71ACUPS Distribution bus ...

A. de-energizes and remains de-energized.

B. is energized without interruption from MCC 252.

C. is energized without interruption from MCC 262.

D. is energized without interruption from Battery 71 BCB-~~A.

Proposed Answer: B

Explanation: Bus 12600 (L26) feeds MCC 262 which is the normal feed to the UPS. With breaker 12602 tripped, Bus 12600 and MCC 262 de-energize. With the normal feed unavailable, the UPS transfers to battery supply and uses the Inverter section. With a fault in the Inverter, the alternate feed (MCC 252) will supply the UPS bus. The UPS has a static electronic switch that seamlessly transfers all power sources to the 71ACUPS Distribution bus.

A. Incorrect- The alternate feed from MCC 252 maintains the bus energized. C. Incorrect- MCC 262 normally powers the UPS, however it is de-energized due to the trip of

the supply breaker to bus 12600. D. Incorrect- The battery would normally maintain power, however the inverter fault prevents

this normal function.

Technical Reference(s): OP-46B


Learning Objective: SDLP-71 F 1.05

Question Source: Bank- March 2012 NRC #4~7

Question History: March 2012 NRC #47



Comments: TRH 3/28/14- Edited first bullet, added UPS designator, and changed out original distractor D, based on NRC comment.



Form ES-401-5


239002 A 1. 09 3.1

SRVs

Ability to predict and/or monitor changes in parameters clssociated with operating the RELIEF/SAFETY VALVES controls including: Indicated vs. actual steam flow: PlantSpecific


The plant has experienced a scram from 100% power with tht~ following:

• MSIV are closed.

• Reactor pressure is 1000 psig and slowly rising.

• The CRS has directed pressure control using SRVs.

• A Reactor Operator places the SRV A control switch in OPEN.

Which one of the following describes the expected response of the Main Steam Line flow indicators, 06FI-88A, (8), (C) and (D) on Panel 09-5?

A. All four flow indicators remain at zero.

B. All four flow indicators rise equally.

C. Two flow indicators rise, while the other two remain at zero.

D. One flow indicator rises, while the other three remain at zero.

Proposed Answer: A

Explanation: The SRVs tap off the Main Steam Lines upstream of the flow detectors; therefore flow through the SRV is NOT included in the Main Steam Line flow indications on Panel 09-5.

B. Incorrect- SRV flow is not included in Main Steam Line flow indication. This would be true if the SRV tapped off far enough downstream to draw flow through the cross-connected Main Steam Line, such as TBVs do.

C. Incorrect- SRV flow is not included in Main Steam Line flow indication. This would be true if the SRV tapped off downstream of flow detector and pairs of Main Steam Lines were cross-connected.

D. Incorrect- SRV flow is not included in Main Steam Line flow indication. This would be true if the SRV tapped off just downstream of the flow detector.

Technical Reference(s): FM-29A


Learning Objective: SDLP-29 1.11.a.2


Question History:



Comments:



LPCS

Level Tier# Group# KIA# Importance Rat~ng

RO 2 1 209001 A2.08 3.1

Ability to (a} predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal condition:s or operations: Valve openings


A loss of coolant accident has resulted in the following:

• Reactor water level is -40" and stable. • Reactor pressure is 90 psig and slowly lowering. • Emergency Depressurization has been performed. • Torus water temperature is 1 05°F and slowly rising1. • Core Spray is the only system available for injection into the Reactor. • Core Spray pump A is running. • Core Spray pump B has tripped. • Core Spray pump A flow is 4700 gpm. • Core Spray system A valves are aligned as shown in the picture on the next page.

Which one of the following describes the status of Core Spray system A valves and, if necessary, the appropriate operator action per OP-14, Core Spray System?

A. Core Spray system A is operating as designed.

B. The full flow test valve is closed, but should be open. Manually open the Core Spray system A full flow test valve.

C. The minimum flow valve is open, but should be closed. Manually close the Core Spray system A minimum flow valve.

D. The inboard injection valve is fully open, but should bel partially closed. Throttle closed the Core spray system A inboard injection valve.

Proposed Answer: c

Explanation: Since Reactor water level is less than 59.5" and Reactor pressure is less than 410 psig, Core Spray A should be automatically aligned for injection. The picture shows the injection valves full open, the full flow test valve closed, and tl1e min flow valve open. With flow greater than 980 gpm, the minimum flow valve should be closed. This is diverting flow from the Reactor. With Reactor water level at -40 inches and stable with limited injection sources, this diverted water is significantly impacting the ability to provide adequate core cooling (as evidenced by Core Spray flow less than 4725 gpm, as requimd by EOP-2 with water level between -19 and -44 inches). Per OP-14, the min flow valve should be closed.

A. Incorrect- The min flow valve should be closed. B. Incorrect- The full flow test valve indicates closed, as it should be for plant conditions. D. Incorrect- Plant conditions require maximizing injection sources and Core Spray NPSH is

not of concern with Torus water temperature at 105°F. Therefore the inboard injection valve should remain full open.





Question History:



Comments:



HPCI


RO 2 1 206000 A2.17 3.9

Ability to (a} predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions~, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: HPCI inadvertent initiation: BWR-2,3,4


The plant has experienced a scram with the following:

• Reactor water level is 205 inches and stable with F1eedwater injecting. • Then, HPCI automatically starts due to an erroneous low Reactor water level initiation

signal. • The initiation signal seals in. • An Operator attempts to secure HPCI by depressing the TURB TRIP 23A-S19

pushbutton, but this fails to trip HPCI.

Which one of the following describes an alternate operation that will secure HPCI flow to the Reactor in accordance with OP-15, High Pressure Coolant Injection?

A. Place 23FIC-108-1, HPCI Flow Control, in MAN and ICiwer HPCI turbine speed.

B. Place 23MOV-14, HPCI Turbine Steam Supply lsol Vadve, control switch in CLOSE.

C. Place 23MOV-16, HPCI Turbine Steam Supply Outbd lsol Valve, control switch in CLOSE.

D. Place 23MOV-19, HPCI Pump Disch to Reactor lnbd lsol Valve, control switch in CLOSE.

Proposed Answer: A

Explanation: With a sealed in initiation signal, 23MOV-14, 2:3MOV-16, and 23MOV-19 all have sealed-in open signals and CANNOT be closed. OP-15 section G.2.2 provides direction for placing 23FIC-108-1 in MAN and lowering Turbine speed to secure injection in this situation.

B. Incorrect- With a sealed in initiation signal, 23MOV-14 has a sealed-in open signal and CANNOT be closed.

C. Incorrect- With a sealed in initiation signal, 23MOV-115 has a sealed-in open signal and CANNOT be closed.

D. Incorrect- With a sealed in initiation signal, 23MOV-1 !3 has a sealed-in open signal and CANNOT be closed.





Question History:



Comments: TRH 3/6/14- Added "in accordance with OP-15 ... " to end of question, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~----~~~~~~


Shutdown Cooling


RO 2 1 205000 A3.03 3.5

Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including: Lights and alarms


The plant is shutdown with the following:

• Reactor pressure is 65 psig and very slowly lowering. • RHR loop B is operating in Shutdown Cooling mode. • RHR loop A is in standby. • Then, a transient results in the following valid annunciators:

o 09-5-1-28, RX WTR LVL ALARM HI OR LOW o 09-5-1-31, RPS RX VESSEL LO LVL TRIP

Which one of the following describes the expected response of 1 OMOV-25B, LPCI INBD INJ VLV, and 10MOV-27B, LPCI OUTBD INJ VLV?

A Both valves close.

B. Both valves remain open.

C. 1 OMOV-25B closes and 1 OMOV-27B remains open.

D. 1 OMOV-25B remains open and 1 OMOV-27B closes.

Proposed Answer: c

Explanation: The given annunciators indicate that Reactor water level has lowered below 177 inches. This results in a PCIS group 2 isolation. With Reactor pressure less than 75 psig and 10MOV-17 and 10MOV-18 not full closed (as evidenced by RHR loop 8 in SOC), 10MOV-25B's logic is altered such that the valve will close as part of a PC IS group 2 isolation. No such logic alteration occurs for 10MOV-278. Therefore, 10MOV-25B closes and 10MOV-278 remains open.

A. Incorrect- 1 OMOV-278 remains open. B. Incorrect- 10MOV-25B closes. D. Incorrect- 1 OMOV-258 closes. 1 OMOV-278 remains open.

Technical Reference(s): AOP-15, SDLP-10


Learning Objective: SDLP-1 0 1.05.a.4.a


Question History:



Comments:



------------------------RO 2 1

Importance Rating 215005 A3.02 3.5

APRM I LPRM

Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: Full core display



• Annunciator 09-5-2-43, LPRM UPSCALE, alarms. • LPRM 20-29A spikes upscale for ten seconds and then returns to the pre-transient

value.

Which one of the following describes the expected indicating light response on the full core display for LPRM 20-29A?

The upscale indicating light. ..

A. will be lit during the spike and remain lit until operator action is taken to clear the alarm condition.

B. will be lit during the spike and then extinguish when the LPRM reading goes below the upscale reset value.

C. will only light during the spike if an adjacent control rod is selected and then it remains lit until operator action is taken to clear the alarm condition.

D. will only light during the spike if an adjacent control rod is selected and then it extinguishes when the LPRM reading goes below the upscale reset value.

Proposed Answer: B

Explanation: The amber LPRM lights on the full-core display illuminate if the associated LPRM experiences an upscale condition (> 120/125). These lights operate independently from control rod selection, unlike the LPRM flux meters just below the full core display. These lights automatically extinguish when the LPRM reading returns to at normal value. Therefore, LPRM 20-29A amber light will be lit during the spike, and then automatically extinguish.

A Incorrect- LPRM lights reset automatically, unlike the control rod drift indications. C. Incorrect- The LPRM upscale lights on the full core display operate independently from

control rod selection, unlike the LPRM flux meters just below the full core display. LPRM lights reset automatically, unlike the control rod drift indications.

D. Incorrect- The LPRM upscale lights on the full core display operate independently from control rod selection, unlike the LPRM flux meters just below the full core display.

Technical Reference(s): SDLP-07C


Learning Objective: SDLP-07C 1.12.d


Question History:

Question Cognitive Level: Memory or Fundamental Knowl1edge


Comments: TRH 3/6/14- Combined the two sentences in choices C and D, based on NRC comment.



Instrument Air

--------------------------Level Tier# Group# KIA# Importance Rating

RO 2 1 300000 A4.01 2.6

Ability to manually operate and/or monitor in the control room: Pressure gauges



• Air Compressor A is operating as the lead compressor. • Air Compressor B is aligned as the 1st standby compressor. • Air Compressor C is aligned as the 2nd standby compressor. • Then, an air leak develops in the plant. • Control Room air pressure indications on Panel 09-6 are as shown in the following

picture, and have been slowly lowering since the leak developed:

Which one of the following describes the expected status of Air Compressor C and 39FCV-11 0, Service Air Header Auto Isolation Valve?

Air Compressor C is ...

A. running and 39FCV-110 is open.

B. running and 39FCV-110 is closed.

C. in standby and 39FCV-110 is open.

D. in standby and 39FCV-110 is closed .

Proposed Answer: c

Explanation: The 1st and 2nd standby air compressors automatically start on air pressures of 107 psig and 104 psig, respectively. 39FCV-110 automatically closes on air pressure of 95 psig. The indicated air pressures are approximately 106-107 psig. Therefore Air Compressor C will still be in standby and 39FCV-110 will still be open.

A Incorrect- Since air pressure is above 104 psig, Air Compressor Cis still in standby. B. Incorrect- Since air pressure is above 104 psig, Air Conn pressor C is still in standby. Since

air pressure is still above 95 psig, 39FCV-110 is still open. D. Incorrect- Since air pressure is still above 95 psig, 39FCV-110 is still open.



Learning Objective: SDLP-39 1.08.a and 1.08.c


Question History:


1 0 CFR Part 55 Content: 55.41(7)

Comments:



Form ES-401-5


215003 A4.06 3.0

IRM

Ability to manually operate and/or monitor in the control room: Detector drives


Which one of the following describes the requirement for withdrawing IRM detectors during a plant startup, in accordance with OP-65, Startup and Shutdown Procedure?

IRMs are withdrawn ...

A. as soon as all APRMs indicate~ 2.5%.

B. once the Reactor Mode Switch is placed in RUN.

C. as needed to maintain indication below 75 on the 0 to 125 scale.

D. as needed to maintain indication below 100 on the 0 to 125 scale.

Proposed Answer: 8

Explanation: OP-65 step 0.20.4 directs withdrawing IRMs. This is done just after step 0.20.2, which places the Reactor Mode Switch in RUN.

A. Incorrect- APRMs indicating above 2.5% (downscale) is a prerequisite for placing the Reactor Mode Switch in RUN and withdrawing IRMs, however IRMs are not withdrawn immediately once all APRMs indicate above 2.5%.

C. Incorrect -IRM range switches are manipulated to maintain indication below 75 on the 0 to 125 scale. IRMs remain fully inserted until the Reactor Mode Switch is in RUN.

D. Incorrect- IRM range switches are manipulated to maintain indication below 100 on the 0 to 125 scale. IRMs remain fully inserted until the Reactor Mode Switch is in RUN.



Learning Objective: SDLP-078 1.05.a.1.d


Question History:


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

Comments: TRH 3/6/14 - Replaced original distractor C, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~------~~~~~~


RCIC


RO 2 1 217000 2.4.20 3.8

Emergency Procedures I Plan: Knowledge of operational implications of EOP warnings, cautions, and notes.


EOP-2, RPV Control, provides the following note:

,.. RCJC I [I Defeat interlocks if necessary (E~

Which one of the following describes a parameter value that would cause a RCIC isolation if the isolation was NOT bypassed?

A. Reactor pressure of 150 psig.

B. RCIC steam flow to turbine of 110%.

C. RCIC equipment room temperature of 110°F.

D. RCIC turbine exhaust diaphragm pressure of 10 psig.

Proposed Answer: D

Explanation: High RCIC turbine exhaust diaphragm pressure causes a RCIC isolation at 5 psig.

Note: The question matches the KIA by testing knowledge of when operators would be required to take action based on an EOP note related to RCIC.

A. Incorrect- RCIC does not isolate on low Reactor pressure until< 93 psig. 150 psig is based on the RCIC operability requirement.

B. Incorrect- RCIC does not isolate on high steam flow until 300%. 110% is based on the approximate RCIC overspeed trip setpoint.

C. Incorrect- RCIC does not isolate on high equipment room temperature until -144°F. 11 0°F is based on being higher than the associated Max Normal temperature in EOP-5.

Technical Reference(s): OP-19, TS 3.3.6.1, SDLP-13


Learning Objective: SDLP-13 1.07.b, c, f, and g


Question History:

Question Cognitive Level: Memory or Fundamental Know,ledge


Comments:

ES-401 Written Examination Question W..::.o.:....:rk~s:..:..h:..:e..::.e.:...t ___ ...:..F..::.o.:..:rm..:..:.....:E::...:S:.....-4_:.0.:...1.:....-..::.5


Shutdown Cooling


RO 2 1 205000 2.1. 7 4.4

Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.


A plant cooldown is in progress with the following:

• RHR loop B is operating in Shutdown Cooling mode. • The Shift Manager has directed a target cooldown ratt3 of 90°F/hr. • The following Reactor coolant temperature data has been logged for the last half hour:

• Time 0000: 302°F • Time 0015: 278°F • Time 0030: 254°F

• 1 OMOV-668, HX B BYP VLV, is being throttled to adjust cooldown rate.

Which one of the following describes the correct operation of 1 OMOV-668 to establish the target cooldown rate?

1 OMOV-668 should be throttled further ...

A. open to lower the cooldown rate.

B. open to raise the cooldown rate.

C. closed to lower the cooldown rate.

D. closed to raise the cooldown rate.

Proposed Answer: A

Explanation: The data gives a cooldown rate of 96°F/hr. This is above the 90°F/hr direction from the Shift Manager. To lower the cooldown rate, 10MOV-66B should be further opened, to allow more water to bypass the heat exchanger.

B. Incorrect- The cooldown rate is above the target and should be lowered. C. Incorrect- 1 OMOV-668 should be further opened, to allow more water to bypass the heat

exchanger and lower the cooldown rate. D. Incorrect- The cooldown rate is above the target and should be lowered. 1 OMOV-668

should be further opened, to allow more water to bypass the heat exchanger.

Technical Reference(s): OP-130, FM-20


Learning Objective: SDLP-1 0 1.05.a.4.e


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ----~------------~----


DC Electrical Distribution


RO 2 1 263000 A3.01 3.2

Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights


The plant is operating at 100% power when the following annunciators alarm:

• 09-8-1-22, 125VDC BATT CHGR B AC SUPP TROUBLE • 09-8-1-23, 125VDC BATT B VOLT LO

125 VDC BATT BUS B meter on Panel 09-8 indicates 0 VDC.

Which of the following automatic operations occur, if any, as a result of these conditions?

(1) ADS Logic B transfers to DC Power System A (2) HPCI Logic B transfers to DC Power System A

A. Neither automatic operation occurs

B. (1) only

C. (2) only

D. (1) and (2)

Proposed Answer: B

Explanation: ADS and HPCI receive power to their Logic B from DC Power System B. The given annunciators and alarms indicate a low voltage condition on DC Power System B. Only ADS logic automatically transfers to DC Power System A.

A Incorrect- ADS logic automatically transfers. C. Incorrect -ADS logic automatically transfers. HPCI logic power does not automatically

transfer. D. Incorrect- HPCIIogic power does not automatically transfer.



Learning Objective: SDLP-71 B 1.09.a.1, 2, and 3


Question History:



Comments:



HPCI


RO 2 1 206000 K1.01 3.8

Form ES-401-5

Knowledge of the physical connections and/or cause- effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the follo,wing: Reactor vessel: BWR-2,3,4


The plant has experienced a failure to scram with the following:

• Initial Reactor power after the failure to scram was 17%. • Reactor water level has been intentionally lowered and is now being controlled in a band

of 0 to 11 0 inches.

Which one of the following describes where HPCI injects into the Reactor vessel and if HPCI is allowed to be used for injection in this situation, in accordance with EOP-3, Failure to Scram?

HPCI injects into the Reactor vessel through ...

A. a dedicated HPCI sparger and is allowed to be used for injection.

B. a dedicated HPCI sparger and is NOT allowed to be used for injection.

C. one of the Feedwater lines and is allowed to be used for injection.

D. one of the Feedwater lines and is NOT allowed to be used for injection.

Proposed Answer: c

Explanation: HPCI injects through Feedwater line B, which injects in the downcomer region outside of the core shroud. HPCI is a Group 1 Water Level Control System allowed to be used for Reactor injection in this situation.

A. Incorrect- HPCI does not have a dedicated sparger. I-I PCI injects through Feedwater line B.

B. Incorrect- HPCI does not have a dedicated sparger. HPCI injects through Feedwater line B. HPCI is allowed to be used for injection.

D. Incorrect- HPCI is allowed to be used for injection.

Technical Reference(s): OP-15, EOP-3, FM-25A, FM-34A


Learning Objective: SDLP-23 1.06.a, MIT-301.11 D "1.07


Question History:

Question Cognitive Level: Memory or Fundamental Knowl«~dge


Comments:



ADS


RO 2 1 218000 A2.04 4.1

Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS failure to initiate


A loss of coolant accident results in the following:

• Reactor water level has lowered per the following graph:

250 .-----------------------------

iii ~ 200 -+------------------------· u

~ ~ 150 +---+-----------------------~ ... Ql .... ~ 100 +---~--------------------... 0 ~ ~ 50 +---------~~------------a::

Q +o .... -.rrrrTT""-~1'1'1"1 oioioiTITITOi ill I I I I I I I I I I I I 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

rl N M ~ ~ ~ ~ 00 m 0 rl N M ~ ~ ~ ~ 00 m 0 rl rl rl rl rl rl rl rl rl rl N

Time (seconds)

• Current time is 200 seconds. • Annunciator 09-4-1-28, ADS TIMERS ACTIVATED, is in alarm. • All SRVs are closed. • Reactor pressure is 500 psig and slowly lowering. • All RHR and Core Spray pumps are operating per design. • No EOP actions have been completed yet.

Which one of the following describes the response of the ADS system and the proper operation of the ADS system in accordance with EOP-2, RPV Control?

ADS has ...

A. operated per design. Override ADS per EOP-2.

B. operated per design. Do NOT override ADS per EOP-2.

C. NOT operated per design. Override ADS per EOP-2.

D. NOT operated per design. Do NOT override ADS per EOP-2.

Proposed Answer: c

Explanation: The ADS logic started timing at approximately t~O seconds (Reactor water level <59.5 inches). After -120 seconds with Reactor water level still <59.5 inches and low pressure ECCS running, ADS should have opened the seven ADS SRVs. Therefore, after time = -170 seconds, ADS has not operated per design since no SRVs are open. EOP-2 requires overriding ADS if the timers initiate and also upon entry into the alternate level control leg. EOP-2 does not direct opening all seven ADS valves until after a determination has been made that Reactor water level cannot be restored and maintained above -19".

A. Incorrect- ADS should have opened seven SRVs at approximately time 170 seconds. B. Incorrect -ADS should have opened seven SRVs at approximately time 170 seconds.

Even though ADS should have opened seven SRVs, EOP-2 requires overriding ADS and delaying emergency depressurization.

D. Incorrect- Even though ADS should have opened seven SRVs, EOP-2 requires overriding ADS and delaying emergency depressurization.

Technical Reference(s): ARP 09-4-1-28, EOP·-2


Learning Objective: MIT-301.11c EO 1.07


Question History:



Comments: TRH 3/6/14- Revised explanation wording for clarity, based on NRC comment.



RPS


RO 2 1 212000 A4.09 3.9

Ability to manually operate and/or monitor in the control room: SCRAM instrument volume level


The plant has experienced a small loss of coolant accident and failure to scram with the following:

• AOP-1, Reactor Scram, immediate actions were taken 5 minutes ago. • Drywell pressure is 4 psig and rising slowly. • Reactor water level is 190 inches and stable. • Reactor pressure is 920 psig and stable. • Reactor power is downscale on APRMs. • Control rod insertion by repeated manual scrams is be~ing performed in accordance with

EP-3, Backup Control Rod Insertion. • The following actions are taken in the sequence listed:

o ARI is reset. o RPS logic jumpers are installed. o The SDIV HI LVL TRIP switch is placed in BYPASS. o The RX SCRAM RESET switch is placed in GHOUP 2&3, then to GROUP 1 &4.

Which one of the following describes when the Scram Discharge Instrument Volume (SDIV) begins to drain during performance of EP-3?

The SDIV begins to drain when ...

A. ARI is reset.

B. the RPS logic jumpers are installed.

C. the SDIV high level trip is bypassed.

D. the Reactor scram is reset.

Proposed Answer: D

Explanation: The SDIV begins to drain when the Reactor sc:ram is reset. This is accomplished by placing the RX SCRAM RESET switch in GROUP 2&~~. then to GROUP 1 &4. Before this is accomplished, CRD charging water is filling the SDIV throug1h all HCU scram valves and the SDIV vents/drains are closed. Once the scram is reset, the scram valves close, which stops water from entering the SDIV. The vents/drains open, which removes water from the SDIV.

A Incorrect- ARI must be reset to allow draining of the SDIV, however the SDIV does not begin to drain until the Reactor scram is also reset.

B. Incorrect- The RPS logic jumpers must be installed to allow resetting the Reactor scram with a sealed in scram signal (high Drywell pressure). However, the SDIV does not begin to drain until the Reactor scram is reset, and that does not happen until the RX SCRAM RESET switch is cycled.

C. Incorrect- The SDIV does not begin to drain until the Re,actor scram is reset. Placing the SDIV HI LVL TRIP switch in BYPASS allows resetting tht3 Reactor scram with high level in the SDIV, but does not starting draining of the SDIV.

Technical Reference(s): EP-3, OP-25, FM-27


Learning Objective: SDLP-03C 1.15.c


Question History:



Comments: TRH 3/25/14- Enhanced distractor wording to equalize lengths, based on NRC comment.



Reactor Feedwater


RO 2 2 259001 K1.06 2.9

Knowledge of the physical connections and/or cause- effect relationships between REACTOR FEEDWATER SYSTEM and the following: Plant air systems



• An Instrument Air leak occurs. • Instrument Air header pressure slowly and continuously lowers.

Which one of the following describes an effect of this condition on the Feedwater system?

Feedwater pump ...

A. minimum flow valves fail open.

B. minimum flow valves fail closed.

C. steam supply control valves fail open.

D. steam supply control valves fail closed.

Proposed Answer: A

Explanation: The Feedwater pump minimum flow valves are air operated and fail open on low air pressure.

B. Incorrect- The Feedwater pump minimum flow valves fail open on low air pressure, not closed.

C. Incorrect- The Feedwater pump steam supply control valves are hydraulically operated using oil pressure from the Main Shaft Oil pump when at full power. These valves are not directly affected by loss of Instrument Air.

D. Incorrect- The Feedwater pump steam supply control valves are hydraulically operated using oil pressure from the Main Shaft Oil pump when at full power. These valves are not directly affected by loss of Instrument Air.

Technical Reference(s): OP-2A, AOP-12


Learning Objective: SDLP-33 1.1 O.a


Question History:

Question Cognitive Level: Memory or Fundamental Knowl1edge


Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 .~~~~------~~~~~~


CRD Hydraulic


RO 2 2 201001 K2.03 3.5

Knowledge of electrical power supplies to the following: Backup SCRAM valve solenoids


Which one of the following describes the electrical power supply and function of the backup scram valve solenoids?

A. 120 VAC, energize to cause a scram

B. 120 VAC, de-energize to cause a scram

C. 125 VDC, energize to cause a scram

D. 125 VDC, de-energize to cause a scram

Proposed Answer: c

Explanation: The backup scram valve solenoids are powered from 125 VDC sources. The solenoids are normally de-energized. The solenoids are energized when required to cause a scram.

A. Incorrect- The power supply is 125 VDC. The normal seram solenoids are AC powered. B. Incorrect- The power supply is 125 VDC. The solenoids are energized to cause a scram.

The normal scram solenoids are AC powered. D. Incorrect- The solenoids are energized to cause a scram.





Question History: 2010 NRC #54



Comments: TRH 3/25/14- Added to plausibility statements, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 .~~~~----~~--~~~


Primary Containment and Auxiliaries


RO 2 2 223001 K3.03 3.4

Knowledge of the effect that a loss or malfunction of the !PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: Containment/drywell pressure: PlantSpecific


Which one of the following describes the effect that a failed open Torus to Drywell vacuum breaker would have on Primary Containment response to a design-basis loss of coolant accident?

A. Both peak Torus and Drywell pressures would be greater than expected.

B. Both peak Torus and Drywell pressures would be lowelr than expected.

C. Peak Torus pressure would be higher than expected atnd peak Drywell pressure would be lower than expected.

D. Peak Torus pressure would be lower than expected and peak Drywell pressure would be higher than expected.

Proposed Answer: A

Explanation: With a failed open Torus to Drywell vacuum breaker, the Torus and Drywell airspaces are directly connected and the pressure-suppression feature of the Primary Containment is bypassed. In the event of a LOCA, this means steam will not be forced under the water in the Torus and condensed. Therefore, pressure will be higher in both the Torus and Drywell than expected for a design-basis LOCA.

B. Incorrect- This malfunction does effectively expand the size of the Drywell airspace, which would lead to lower pressures for very small LOCAs. However for a design-basis LOCA, the loss of pressure-suppression function would lead to higher peak pressures.

C. Incorrect - Peak Drywell pressure would also be higher during a design-basis LOCA. D. Incorrect- Peak Torus pressure would also be higher during a design-basis LOCA.

Technical Reference(s): Technical Specification Bases 3.6.1.7 Action B.1


Learning Objective: SDLP-16A 1.09.e

Question Source: Bank- NMP1 SYSID 53211

Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ------------------------Examination Outline Cross-Reference:

Traversing In-core Probe


RO 2 2 215001 K4.01 3.4

Knowledge of TRAVERSING IN-CORE PROBE design feature(s) and/or interlocks which provide for the following: Primary containment isolation: Mark-1&11 (Not-BWR1)



• Traversing In-core Probe (TIP) calibration of LPRMs is in progress. • When the TIP detector is inserting and almost full in-core, the following occurs:

• The Reactor scrams due to a Turbine trip. • Reactor water level lowers to a low of 11 0 inches. • Operator action is taken to restore and maintain Reactor water level 180-220 inches.

Which one of the following describes the response of the TIP system?

The TIP detector ...

A. retracts to the lead shield, and then the ball valve c,loses.

B. retracts to the lead shield, but the ball valve remains open.

C. does not retract to the lead shield and the ball valve remains open.

D. does not retract, but the shear valve fires and isolates the guide tube penetration.

Proposed Answer: A

Explanation: Reactor water level <177 inches causes a PCIS group 2 isolation, which includes isolation of the TIPs. With a TIP in the core, the detector is automatically given a retract signal. Once the detector clears limit switches near the lead shield, the ball valve automatically closes.

B. Incorrect- This situation not only interrupts the TIP operation by retracting the detector, but also isolates the penetration by closing the ball valve.

C. Incorrect- Reactor water level <177 inches causes the normal operation of the TIP to be interrupted.

D. Incorrect- Reactor water level <177 inches does require TIP penetration isolation, however this is accomplished by retracting and closing the ball valve, not by automatically firing the shear valve.

Technical Reference(s): RAP-7.3.14


Learning Objective: SDLP-07F 1.05.c

Question Source: Modified- March 2012 NRC #58




Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 -------------------------Examination Outline Cross-Reference: Level

Tier# Group# KIA# Importance Rating

Main Turbine Generator and Auxiliary Systems

RO 2 2 245000 K5.06 2.5

Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: Turbine shaft sealing


The plant is operating at 100% power when the running Steam Packing Exhauster fan trips.

Which one of the following describes the impact of this failure!?

A. The Steam Seal pressure control valve fully closes.

B. The standby Steam Packing Exhauster fan automatically starts.

C. Steam will leak from the Main Turbine seals into the Turbine Building.

D. Main Condenser vacuum will lower due to air in-leaka!~e across the Main Turbine seals.

Proposed Answer: c

Explanation: A running Steam Packing Exhauster fan is required to maintain a vacuum in the Steam Packing Exhauster to draw in steam from the Main Turbine seals. Without this vacuum, some of the steam supplied by Seal Steam will leak past the Main Turbine seals into the Turbine Building.

A. Incorrect- With lowering Steam Packing Exhauster vacuum, seal steam backpressure may rise, causing the Seal Steam pressure control valve to throttle closed slightly to maintain -3 psig seal steam pressure. However, this will not preve·nt the negative consequence of steam leaking through the seals into the Turbine Building ..

B. Incorrect- The Steam Packing Exhauster fans do not have an auto-start feature. D. Incorrect- Seal steam is still supplied at- 3 psig to the Main Turbine seals and will prevent

air in-leakage. This is the consequence for loss of seal steam.

Technical Reference(s): OP-240, SDLP-94A


Learning Objective: SDLP-94A 1.1 O.d


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 .~~~~----~~~~~~


MSIV Leakage Control


RO 2 2 239003 K6.02 2.8

Knowledge of the effect that a loss or malfunction of the Jollowing will have on the MSIV LEAKAGE CONTROL SYSTEM: Standby gas treatment system: BWR-4,5,6(P-Spec)


The plant has experienced a loss of coolant accident with the following:

• All MSIVs are closed. • Standby Gas Treatment (SGT) train A is tagged out for maintenance. • Drywell pressure is 10 psig and slowly rising. • MCC-162 is de-energized due to an electrical fault.

Which one of the following describes the effect of these conditions on the ability of the Main Steam Leakage Control System (MSLCS) to perform its desig1n function as required by AOP-39, Loss of Coolant?

A. Either train of MSLCS is able to perform its design function.

B. MSLCS train A is available. MSLCS train 8 is NOT available.

C. MSLCS train 8 is available. MSLCS train A is NOT available.

D. Neither train of MSLCS is able to perform its design function.

Proposed Answer: D

Explanation: MCC-162 is the power supply to SGT fan B. With MCC-162 de-energized and SGT fan A out of service, no SGT train is available. At least one train of SGT is required for MSLCS to perform its design function. Therefore neither train of MSLCS is available to perform its design function. Note that the MSLCS valves are powered from MCC-152 and MCC-164. Therefore the MSLCS valves are still functional.

A. Incorrect- With no SGT trains in service, neither train of MSLCS is available to perform its design function.

B. Incorrect- With no SGT trains in service, neither train of MSLCS is available to perform its design function.

C. Incorrect -With no SGT trains in service, neither train of MSLCS is available to perform its design function.

Technical Reference(s): AOP-39, OP-1, OP-20




Question History:



Comments:

ES-401 Written Examination Question w,orksheet Form ES-401-5 .~~~~------~~--~~~


RHR/LPCI: Containment Spray Mode


RO 2 2 226001 A1.10 3.0

Ability to predict and/or monitor changes in parameters atssociated with operating the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE controls including: Emergency generator loading


A steam leak in the Drywell has resulted in the following:

• A manual Reactor scram was inserted. • Following the scram, Lines 3 and 4 de-energized. • Only Emergency Diesel Generators (EDG) A and B have started. • Torus spray is placed in service with RHR pump A l'unning. • EDG A load is 2490 KW. • Then, an operator starts RHRSW pumps A and C. • EDG A load rises to 3100 KW.

Which one of the following describes EDG A operation with respect to load limitations during this evolution?

EDG A was operating ...

A. below the continuous load rating before RHRSW pumps were started and is now operating above the 30 minute load rating.

B. below the continuous load rating before RHRSW pumps were started and is now operating below the 30 minute load rating.

C. above the continuous load rating before RHRSW pumps were started and is now operating above the 30 minute load rating.

D. above the continuous load rating before RHRSW pumps were started and is now operating below the 30 minute load rating.

Proposed Answer: A

Explanation: Per OP-22 Precaution and Limitation C.2.11, eatch EDG has the following load ratings:

• :s; 2600 KW- Continuous • > 2600 KW, :s; 2850 KW -· 2000 hours • > 2850 KW, :s; 2950 KW -· 1 E>O hours • > 2950 KW, :s; 3050 KW -· 30 minutes

Before the RHRSW pumps were running, EDG A load was 2490 KW and below the continuous load rating. After the RHRSW pumps were started, EDG A load rose to 3100 KW, which is above the 30 minute load rating.

B. Incorrect- After RHRSW pumps are started, EDG A is above the 30 minute load rating. C. Incorrect- Before RHRSW pumps are started, EDG A is below the continuous load rating. D. Incorrect- Before RHRSW pumps are started, EDG A is below the continuous load rating.

After RHRSW pumps are started, EDG A is above the :30 minute load rating.



Learning Objective: SDLP-1 0 1.05.a.2.a


Question History:



Comments: TRH 3/6/14 -Included OP-22 P&L in explanation, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~----~~~~~~


RWM


RO 2 2 201006 A2.03 3.0

Ability to (a) predict the impacts of the following on the R~OD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of thos•:t abnormal conditions or operations: Rod drift: P-Spec (Not-BWR6)



• Reactor power is 18%. • The Rod Worth Minimizer (RWM) NORMAL BYPASS keylock switch is in NORMAL. • A control rod in the current rod group spuriously drifts from position 24 to position 48. • The current rod group insert limit is 12. • The current rod group withdraw limit is 24.

Which one of the following describes the response of the RWM to this rod drift and the required operator action for the rod drift, in accordance with AOP-27, Control Rod Drift?

The RWM ...

A. enforces a rod block. Place the Reactor Mode Switch in SHUTDOWN.

B. enforces a rod block. Drive the drifted rod full in using RMCS.

C. does NOT enforce a rod block. Place the Reactor Mode Switch in SHUTDOWN.

D. does NOT enforce a rod block. Drive the drifted rod full in using RMCS.

Proposed Answer: D

Explanation: Since Reactor power is above the RWM Low Power Setpoint (LPSP) (1 0% Tech Spec, 16% nominal), the RWM will NOT enforce a rod block due to a rod withdraw error. AOP-27 requires driving the drifted control rod full-in, NOT inserting! a Reactor scram.

A. Incorrect- The RWM does NOT enforce a rod block because Reactor power is above the LPSP.

B. Incorrect- The RWM does NOT enforce a rod block because Reactor power is above the LPSP.

C. Incorrect- AOP-27 requires driving the drifted control rod full-in, NOT inserting a Reactor scram.

Technical Reference(s): OP-64, AOP-27




Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ·--------------------------Examination Outline Cross-Reference:

RWCU


RO 2 2 204000 A3.05 2.8

Ability to monitor automatic operations of the REACTOR WATER CLEANUP SYSTEM including: Reactor water temperature


Given the following Reactor Water Cleanup (RWCU) valve names:

• 12MOV-15, RWCU Supply Inboard Isolation Valve, • 12MOV-18, RWCU Supply Outboard Isolation Valve

Which one of the following describes the setpoint for the RWCU isolation on high nonregenerative heat exchanger (NRHX) outlet temperature and the response of these two valves to this isolation signal?

When NRHX outlet temperature reaches the setpoint of ...

A. 120°F, 12MOV-15 remains open and 12MOV-18 closes.

B. 140°F, 12MOV-15 remains open and 12MOV-18 closes.

C. 120°F, 12MOV-15 and 12MOV-18 both close.

D. 140°F, 12MOV-15 and 12MOV-18 both close.

Proposed Answer: B

Explanation: The high NRHX outlet temperature isolation setpoint is 140°F. This isolation closes 12MOV-18, but not 12MOV-15.

A. Incorrect- The high NRHX outlet temperature isolation setpoint is 140°F. 120°F is the nominal temperature that NRHX outlet temperature is to be maintained below.

C. Incorrect- The high NRHX outlet temperature isolation setpoint is 140°F. 120°F is the nominal temperature that NRHX outlet temperature is to be maintained below. 12MOV-15 closes on most RWCU isolation signals, but not on the isolation due to high NRHX outlet temperature.

D. Incorrect -12MOV-15 closes on most RWCU isolation signals, but not on the isolation due to high NRHX outlet temperature.



Learning Objective: SDLP-12 1.05.c.1 and 1.05.c.4


Question History:



Comments: TRH 3/6/14- Revised to remove reference to 12MOV-69 from all the answer choices, based on NRC comment.



Fuel Pool Cooling/Cleanup


RO 2 2 233000 A4.11 2.5

Form ES-401-5

Ability to manually operate and/or monitor in the control room: Closed cooling water temperature


A refueling outage is in progress with the following:

• A full core offload is in progress. • Both Fuel Pool Cooling (FPC) recirc pumps and heat ~exchangers are in service. • Spent Fuel Pool (SFP) temperature is 95°F and rising. • RBCLC supply temperature is currently 85°F. • 15TCV-1 01, RBCLC Discharge Header Temperature Control Valve, is full closed.

Which one of the following is the expected trend in RBCLG system temperature and the appropriate operator action?

RBCLC system temperature is expected to ...

A. rise. Direct an Operator in the field to raise RBCLC flow through the in-service RBCLC heat exchanger(s).

B. rise. Direct an Operator in the field to raise Service Water flow through the in-service RBCLC heat exchanger(s).

C. remain stable at approximately 85°F. Direct an Operator in the field to raise RBCLC flow to the FPC heat exchangers to not exceed the SFP temperature limit of 1 00°F.

D. remain stable at approximately 85°F. Direct an Operaltor in the field to raise RBCLC flow to the FPC heat exchangers to not exceed the SFP temperature limit of 125°F.

Proposed Answer: 8

Explanation: With rising SFP temperature, R8CLC heat load is rising. This will raise R8CLC temperature since the only automatic temperature control mechanism, 15TCV-101, is already in the maximum cooling position. Further control of R8CLC temperature is performed by throttling further open the R8CLC heat exchanger Service Water outlet valves to raise Service Water flow.

A. Incorrect- R8CLC temperature control is augmented by throttling the Service Water flow through the heat exchanger, not the R8CLC flow.

C. Incorrect- Since 15TCV-1 01 is already in the maximum cooling position and heat load is rising, R8CLC temperature will rise. 1 00°F is the SFP heat exchanger outlet temperature high alarm.

D. Incorrect- Since 15TCV-101 is already in the maximum cooling position and heat load is rising, R8CLC temperature will rise. 125°F is the upper limit on SFP temperature per OP-40.

Technical Reference(s): OP-40, FM-15A, OP-30




Question History:



Comments: TRH 3/8/14- Revised 2nd half of original choices A and 8 to raise plausibility, based on NRC comment. Re-ordered choices based on length.

ES-401 Written Examination Question Wc>rksheet Form ES-401-5 --------------~---------


Nuclear Boiler Instrumentation


RO 2 2 216000 2.1.32 3.8

Conduct of Operations: Ability to explain and apply system limits and precautions


A plant cooldown is in progress with the following:

• Reactor pressure is 200 psig and slowly lowering. • Narrow range Reactor water level instruments are indicating 200 inches and stable. • Wide range Reactor water level instruments are indicating 210 inches and stable. • The deviation between narrow and wide range Reactor water level instruments has been

getting larger as the cooldown has progressed.

Which one of the following explains which instruments are expected to be more accurate under these plant conditions and the reason why?

The more accurate instruments under these plant conditions are the ...

A. wide range instruments due to density compensation.

B. narrow range instruments due to density compensation.

C. wide range instruments due to differences in calibration conditions.

D. narrow range instruments due to differences in calibration conditions.

Proposed Answer: 8

Explanation: The wide and narrow range Reactor water level instruments are calibrated at the same conditions (1000 psig, 546°F, 135°F OW temperature). However, only the narrow range Reactor water level instruments are density compensated. This density compensation makes the narrow range instruments more accurate as Reactor water temperature and pressure lower. The wide range instruments tend to indicate higher than actual Reactor water level as Reactor water temperature and pressure lower.

Note: The question matches the KIA by testing the ability to apply knowledge of Nuclear Boiler Instrumentation (Reactor water level instrumentation) limits (based on calibration conditions) to interpret the accuracy of these instruments. EOP-11, EOP & SAOG Graphs, contains a caution regarding the affects of Reactor pressure on Reactor water level indications. A specific application of this caution is tested in this question.

A Incorrect -The narrow range instruments are more accurate under these conditions due to density compensation.

C. Incorrect- The wide and narrow range Reactor water level instruments are calibrated at the same conditions. The narrow range instruments are mort::! accurate under these conditions due to density compensation.

D. Incorrect- The wide and narrow range Reactor water lev,el instruments are calibrated at the same conditions.

Technical Reference(s): EOP-11, SDLP-028


Learning Objective: SDLP-028 1.1 O.g


Question History:

Question Cognitive Level: Memory or Fundamental Knowl,edge


Comments:

ES-401 Written Examination Question Wtlrksheet Form ES-401-5 ------~~-----------------


Control Rod and Drive Mechanism


RO 2 2 201003 A2.06 3.0

Ability to {a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM; and {b) based on those predictions, use pmcedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of CRD cooling water flow



• Reactor pressure is 700 psig. • The in-service Control Rod Drive (CRD) flow control valve (FCV) has failed closed.

Which one of the following describes a consequence of this failure and the required action(s)?

A. The Reactor scram function is not assured. Scram the Reactor, then place the standby CRD FCV in service.

B. The Reactor scram function is not assured. Place the standby CRD FCV in service. A Reactor scram is NOT required.

C. CRD mechanism temperatures will rise. Scram the Reactor, then place the standby CRD FCV in service.

D. CRD mechanism temperatures will rise. Place the standby CRD FCV in service. A Reactor scram is NOT required.

Proposed Answer: D

Explanation: Closure of the in-service CRD FCV blocks flow to the drive water and cooling water headers, but not the charging water header. CRDM mHchanism temperatures will rise due to loss of cooling, but charging water header pressure is still available to ensure the scram function will work. Based on the low initial Reactor pressure, if charging water header pressure was not maintained, the scram function may not be assured and a Reactor scram would be required. With charging water header pressure maintained, there is not requirement for a Reactor scram. OP-25 must be used to place the standby CRD FCV in service.

A. Incorrect- The scram function is still assured. No Reactor scram is required. B. Incorrect- The scram function is still assured. C. Incorrect- No Reactor scram is required.

Technical Reference(s): FM-27, OP-25, AOP-69


Learning Objective: SDLP-03A 1.10.a


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 --~~~------~---------


Plant Fire On Site


RO 1 1 600000 AK 1. 02 2.9

Knowledge of the operational implications of the following concepts as they apply to Plant Fire On Site: Fire Fighting



• Maintenance personnel report a fire has developed in the East Electric Bay. • A high temperature alarm has been received for this zone on the Fire Protection Panel. • No other adverse effects from the fire have been observed.

Which one of the following describes the operational responsE~ required for these conditions in accordance with EAP-3, Fire Fighting, and AOP-28, Operation During Plant Fires?

A. Dispatch the full Fire Brigade. Inserting a manual Reactor scram is required now.

B. Dispatch the full Fire Brigade. Inserting a manual Reactor scram is NOT yet required.

C. Dispatch the Fire Brigade Leader to assess the need for full Fire Brigade activation. Inserting a manual Reactor scram is required now.

D. Dispatch the Fire Brigade Leader to assess the need for full Fire Brigade activation. Inserting a manual Reactor scram is NOT yet required.

Proposed Answer: B

Explanation: EAP-3 contains guidance for Control Room Operator response to a report from personnel of a fire in the plant. This guidance requires dispatching the full Fire Brigade, NOT just the Fire Brigade Leader. AOP-28 contains criteria for when a Reactor scam is required based on a fire. All four of the following need to be met for a Reactor scram to be required based on this fire:

• Serious fire in progress (E-Pian Alert or worse) • Reactor in mode 1 or 2 • Ionization alarm at FPP, actuation of fire suppression system, and/or verbal report of fire • Unexplained EPIC or annunciator alarm, unexplained loss of equipment, and/or

unexplained actuation of equipment

The given conditions only satisfy the 2nd and 3rd bullets, therefore a Reactor scram is NOT yet required.

A Incorrect- Since no unexplained alarms, equipment loss, or equipment actuation have occurred, a Reactor scram is NOT required per AOP-28 yet.

C. Incorrect- EAP-3 requires full Fire Brigade activation du1e to the verbal report of a fire. Since no unexplained alarms, equipment losses, or equipment actuations have occurred, a Reactor scram is NOT required per AOP-28 yet.

D. Incorrect- EAP-3 requires full Fire Brigade activation due to the verbal report of a fire.

Technical Reference(s): EAP-3, AOP-28


Learning Objective: LP-AOP 1.03.a


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 --~~--------~---------


SCRAM


RO 1 1 295006 AK 1. 02 3.4

Knowledge of the operational implications of the followi111g concepts as they apply to SCRAM: Shutdown margin


The plant was operating at 100% power when a manual Reactor scram was inserted.

Which of the following sets of indications, by itself, provides sufficient information to determine that the Reactor will stay shutdown under all conditions without boron?

(1) All control rods are inserted to position 00 except one. (2) Reactor power is below 2.5% on all APRMs. (3) AIIIRMs are indicating on Range 5.

A. (1) only

B. (2) only

C. (3) only

D. (1) and (3)

Proposed Answer: A

Explanation: The Technical Specification bases for shutdown margin ensure that the Reactor will stay shutdown under all conditions without boron and the highest worth control rod fully withdrawn as long as all other control rods are fully inserted to position 00. Reactor power being below 2.5% on APRMs does not guarantee the Reactor will stay shutdown under all conditions without boron. AIIIRMs indicating on Range 5 means the Reactor is currently shutdown, but does not guarantee shutdown under all conditions.

B. Incorrect- Reactor power being below 2.5% on APRMs does not guarantee the Reactor will stay shutdown under all conditions without boron.

C. Incorrect- AIIIRMs indicating on Range 5 means the Reactor is currently shutdown, but does not guarantee shutdown under all conditions.

D. Incorrect- AIIIRMs indicating on Range 5 means the Reactor is currently shutdown, but does not guarantee shutdown under all conditions.

Technical Reference(s): EOP-2, MIT-301.11B, MIT-301.11C, Tech Spec Bases 3.1.1


Learning Objective: MIT-301.11 1.07


Question History:

Question Cognitive Level: Memory or Fundamental KnowiE~dge

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

Comments: TRH 3/6/14- Changed "or" to "and" in choice D, based on NRC comment.

ES-401 Written Examination Question Wt:>rksheet Form ES-401-5 -------------------------Examination Outline Cross-Reference: Level


RO 1 1 295037 EK1.02 4.1

SCRAM Condition Present and Reactor Power Above API~M Downscale or Unknown

Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER A.BOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level effects on reactor power


A failure to scram has resulted in the following:

• Multiple control rods are still withdrawn and CANNOT be inserted. • Initial Reactor power level after the failure to scram wats 15%. • Reactor water level was intentionally lowered to below 110 inches. • Reactor power is now downscale on all APRMs and mid-scale on IRM range 6. • Reactor water level is 105 inches and slowly rising with Feedwater injecting. • Reactor pressure is 920 psig and slowly lowering. • Standby Liquid Control (SLC) is injecting. • Initial SLC tank level was 80%. • Current SLC tank level is 60%. • Torus water temperature is 1 05°F and stable.

Which one of the following describes the required Reactor water level control strategy, in accordance with EOP-3, Failure to Scram?

A. Maintain Reactor water level below 110 inches. Restoring Reactor water level between 177 inches and 222.5 inches must wait until the proper amount of SLC is injected.

B. Maintain Reactor water level below 110 inches. Restoring Reactor water level between 177 inches and 222.5 inches must wait until the proper IRM indication is reached.

C. Restore and maintain Reactor water level between 17~r inches and 222.5 inches, based on the current IRM indications.

D. Restore and maintain Reactor water level between 177 inches and 222.5 inches, based on the current amount of SLC injected.

Proposed Answer: A

Explanation: With an initial power level of >2.5%, Reactor water level was intentionally lowered to less than 110 inches per EOP-3. Reactor water level must be maintained below this level until either a rod pattern is achieved that ensures the Reactor will remain shutdown under all conditions without boron or hot shutdown boron weight is injected (26% of tank level). Currently, multiple control rods are still withdrawn and only 20% of SLC tank level has been injected, so the conditions for restoring Reactor water level above 110 inches are not met. Therefore Feedwater injection must be throttled to prevent level from exceeding 110 inches.

B. Incorrect -IRM indications can be used to determine the Heactor is currently shutdown and begin an RPV cooldown, but they cannot be used to raise Reactor water level above 110 inches.

C. Incorrect- Since the rod pattern does not ensures the Reactor will remain shutdown under all conditions without boron and hot shutdown boron weight has not been injected, Reactor water level must be maintained below 110 inches.

D. Incorrect- Since the rod pattern does not ensures the Reactor will remain shutdown under all conditions without boron and hot shutdown boron weight has not been injected, Reactor water level must be maintained below 110 inches.

Technical Reference(s): EOP-3, MIT-301.11 D


Learning Objective: MIT-301.11 D 1.07


Question History:



Comments:


Control Room Abandonment


RO 1 1 295016 AK2.02 4.0

Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Local control stations: Plant-Specific


The plant was operating at 100% with the following:

• A fire developed in the Control Room and required evi3cuation. • AOP-43, Plant Shutdown From Outside the Controi Room, is being performed. • The Shift Manager has directed starting RHRSW pump Band RHR pump D.

Which one of the following describes the panellocation(s) where these actions are performed?

A Both actions are performed at 25RSP (Reactor Building 300' North).

B. RHRSW pump B is started at 25RSP (Reactor Building 300' North) and RHR pump D is started at 25ASP-2 (East Crescent Stairway).

C. RHRSW pump B is started at 25ASP-2 (East Crescent Stairway) and RHR pump D is started at 25RSP (Reactor Building 300' North).

D. Both actions are performed at 25ASP-2 (East Crescent Stairway).

Proposed Answer: A

Explanation: 25RSP, located on Reactor Building 300' North, contains the controls to both start RHRSW pump B and RHR pump D. 25ASP-2, located at the East Crescent Stairway, contains multiple RHR valve controls, but no controls for pumps.

B. Incorrect- Both controls are at 25RSP. C. Incorrect- Both controls are at 25RSP. D. Incorrect - Both controls are at 25RSP.



Learning Objective: SDLP-1 0 1.11.b.3


Question History:

Question Cognitive Level: Memory or Fundamental Knowh3dge


Comments:



Loss of Shutdown Cooling


RO 1 1 295021 AK2.03 3.6

Knowledge of the interrelations between LOSS OF SHLJT[)OWN COOLING and the following: RHR/shutdown cooling


The plant is shutdown during a refueling outage with the following:

• New fuel is being loaded into the Reactor core. • RHR pump A is running for Shutdown Cooling. • All other RHR pumps are available and in standby. • Then, a leak develops from a through-wall crack between 1 OMOV-18, RHR Shutdown

Cooling Outboard Isolation Valve, and 10MOV-17, RHR Shutdown Cooling Inboard Isolation Valve.

• The Shift Manager has directed securing RHR pump A and isolating the leak by closing 10MOV-18 and 10MOV-17.

Which one of the following describes the impact of this evolution on Shutdown Cooling?

A. Shutdown Cooling will be secured temporarily during RHR pump A shutdown, but can be re-established with any of the other three RHR pumps once the leak is isolated.

B. Shutdown Cooling will be secured temporarily during RHR pump A shutdown, but can be re-established with either RHR pump B or D once the leak is isolated. RHR pump C will be unavailable for Shutdown Cooling.

C. The RHR Shutdown Cooling mode will be unavailable as long as 1 OMOV-18 and 10MOV-17 are closed. Alternate decay heat removal using RHR in the Fuel Pool Cooling Assist mode may be manually aligned.

D. The RHR Shutdown Cooling mode will be unavailable as long as 1 OMOV-18 and 1 OMOV-17 are closed. Alternate decay heat removal using RHR in the Fuel Pool Cooling Assist mode is NOT available to be manually aligned.

Proposed Answer: c

Explanation: 10MOV-18 and 10MOV-17 are required to be open to align any RHR pump for Shutdown Cooling. Therefore, with these valves closed, the~ RHR Shutdown Cooling mode will be unavailable. However, RHR may still be placed in the Fuel Pool Cooling Assist mode, which requires 10MOV-18 and 10MOV-17 to be closed. With the :SFP gates removed, as evidenced by loading of new fuel into the Reactor core, this provides alternate decay heat removal.

A Incorrect -10MOV-18 and 10MOV-17 are required to be open to align any RHR pump for Shutdown Cooling.

B. Incorrect -10MOV-18 and 10MOV-17 are required to be open to align any RHR pump for Shutdown Cooling.

D. Incorrect- The Fuel Pool Cooling Assist mode is available for decay heat removal with 10MOV-18 and 10MOV-17 closed.

Technical Reference(s): OP-130, AOP-30, OP-·13F, FM-20


Learning Objective: SDLP-10 1.06.e and 1.06.h


Question History:



Comments:



Reactor Low Water Level


RO 1 1 295031 EK2.15 3.2

Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following: A.C. distribution: Plant-Specific



• The Feedwater header ruptures in the Turbine Building. • Reactor water level quickly lowers to 140 inches.

Which one of the following describes the electrical distribution system response to this transient?

A The 10100, 10200, 10300, 10400, and 10700 buses transfer to Transformers T2 and T3. The 10500 and 10600 buses transfer to the Emergency Diesel Generators.

B. The 10100, 10200, 10300, 10400, and 10700 buses transfer to Transformers T2 and T3. The 10500 and 10600 buses remain energized from the 10300 and 10400 buses, respectively.

C. The 10100, 10200, 10300, and 10400 buses transfer to Transformers T2 and T3. The 10700 bus de-energizes. The 10500 and 10600 buses transfer to the Emergency Diesel Generators.

D. The 10100, 10200, 10300, and 10400 buses transfer to Transformers T2 and T3. The 10700 bus de-energizes. The 10500 and 10600 buses remain energized from the 10300 and 10400 buses, respectively.

Proposed Answer: D

Explanation: The low Reactor water level condition causes a Reactor scram (177"). Due to the Reactor scram, the 10100, 10200, 10300, and 10400 buses fast-transfer from Transformer T4 to Transformers T2 and T3. Also due to the scram and loss of power to Transformer T4, the 10700 bus de-energizes. The 10500 and 10600 buses remain energized from the 10300 and 1 0400 buses.

A. Incorrect- 10700 de-energizes and 10500/10600 do NOT transfer to the EDGs. B. Incorrect- 10700 de-energizes. C. Incorrect- 10500/10600 do NOT transfer to the EDGs.

Technical Reference(s): OP-46A, OP-22, SDLP-'71 E, SDLP-93


Learning Objective: SDLP-71 E 1.05.c.1, SDLP-93 1.07.b.2


Question History:



Comments:



High Drywell Temperature


RO 1 1 295028 EK3.01 3.6

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: Emergency depressurization


Which one of the following describes the requirement in EOP-·4, Primary Containment Control, for performing an emergency depressurization due to high D~rwell temperature and the basis of this requirement?

EOP-4 requires emergency depressurization if Drywell temperature cannot be restored and maintained below ...

A. 260°F. This requirement is based on applicable component qualification or structural design temperature limits.

B. 260°F. This requirement is based on the upper limit for Primary Containment temperature indication.

C. 309°F. This requirement is based on applicable component qualification or structural design temperature limits.

D. 309°F. This requirement is based on the upper limit for Primary Containment temperature indication.

Proposed Answer: c

Explanation: EOP-4 contains the following step:

II' .......

IF THEN

Drywell temperature cannot be EMERGENCY RPV restored and maintained below 309° F DEPRESSURIZATION IS

REQUIRED. enter EOP-2, "RPV Control," and execute it concurrently with this procedure.

\... ~

The EOP bases generically state that this temperature is based on "applicable component qualification or structural design limits". 309°F is specifically the design temperature of the Drywell.

A. Incorrect- The requirement is 309°F, not 260°F (the threshold for isolating cooling water to Drywell coolers).

B. Incorrect- The requirement is 309°F, not 260°F (the threshold for isolating cooling water to Drywell coolers). The temperature is based on component qualification or structural design limits, not the upper limit on Drywell temperature indica1tion (such as the horizontal portion of the RPV Saturation Temperature curve).

D. Incorrect- The temperature is based on component qualification or structural design limits, not the upper limit on Drywell temperature indication (such as the horizontal portion of the RPV Saturation Temperature curve).

Technical Reference(s): EOP-4, MIT-301.11 E, SDLP-16A


Learning Objective: MIT-301.11 E 4.05


Question History:


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

Comments:



Partial or Complete Loss of AC


RO 1 1 295003 AK3.03 3.5

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Load shedding


Which one of the following describes the reason for load shedding during a Station Blackout, in accordance with AOP-49, Station Blackout?

A. Reduce heat generation to ensure Battery Room temperatures remain below the limit of 120°F.

B. Reduce heat generation to ensure Battery Room temperatures remain below the limit of 200°F.

C. Preserve capacity of station batteries to ensure meeting the coping time requirement of 4 hours.

D. Preserve capacity of station batteries to ensure meeting the coping time requirement of 8 hours.

Proposed Answer: c

Explanation: AOP-49 requires load shedding in attachment 21 to be completed within 30 minutes of the start of the station blackout. This load shedding is required to reduce loading on station batteries to preserve enough capacity to meet the required station coping time of 4 hours.

A. Incorrect- While load shedding will limit temperature rise in the associated areas, this is not the basis for load shedding. 120°F is the limit in the UFSAR Station Blackout discussion related to Control and Relay Room temperature.

B. Incorrect- While load shedding will limit temperature rise in the associated areas, this is not the basis for load shedding. 200°F is the maximum expeeted temperature in the Primary Containment in the UFSAR Station Blackout discussion.

D. Incorrect -The coping time requirement is 4 hours, not 8. 8 hours is the expected station battery discharge time at the lowest listed discharge rate.

Technical Reference(s): AOP-49, UFSAR section 8.11.3


Learning Objective: SDLP-71 E 1.14.f


Question History:


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

Comments: TRH 3/6/14- Changed "Reactor Building" to "Battery Room" in choices A and B, based on NRC comment.



High Drywell Pressure


RO 1 1 295024 EK3.01 3.6

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Drywell spray operation: Mark-1&11


Which one of the following describes the consequences of spraying the Drywell if conditions are in the unacceptable region of the Drywell Spray Initiation Limilt Curve in EOP-11, EOP & SAOG Graphs?

A. The steam produced by spraying cold water into a superheated atmosphere may over pressurize the primary containment.

B. The cold spray water may put excessive thermal stress on the Drywell, which may lead to structural failure of the primary containment.

C. Convective cooling may result in an immediate, rapid, and large reduction in Drywell pressure which could cause a loss of primary containment integrity due to Drywell to atmosphere pressure being negative.

D. Evaporative cooling may result in an immediate, rapid, and large reduction in Drywell pressure which could cause a loss of primary containment integrity due to Drywell to atmosphere pressure being negative.

Proposed Answer: D

Explanation: With Drywell pressure and temperature in the unacceptable region of the Drywell Spray Initiation Limit, the initiation of Drywell sprays may result in a large drop in primary containment pressure due to evaporative cooling. This drop in pressure can occur faster than can be compensated for by the vacuum relief system, and could result in challenging primary containment integrity.

A. Incorrect- The concern is excessive negative pressure!, not positive pressure. B. Incorrect- Cold water from spray would cause some thermal stress to spray nozzles, but

this is not the concern before the Drywell Spray Initiation Limit. C. Incorrect- The pressure drop of concern is caused by rapid evaporative cooling, not slower

convective cooling.

Technical Reference(s): MIT-301.11 E



Question Source: Bank - 2010 NRC #11


Question Cognitive Level: Memory or Fundamental KnowiE~dge

10 CFR Part 55 Content: 55.41 (1 0)

Comments: TRH 3/10/14- Resampled KIA and replaced question, based on NRC comment.



Form ES-401-5


295030 EA 1. 05 3.5

Low Suppression Pool Water Level

Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI


The plant was operating at 100% power when the following occurred:

• A seismic event caused a Reactor scram. • HPCI and RCIC are the only injection sources currently injecting into the Reactor. • No RHR, Core Spray or Condensate pumps are available for injection. • Reactor water level is -5 inches and lowering. • Reactor pressure is 700 psig and lowering. • Torus water level is 9.5 feet and lowering.

Which one of the following describes how HPCI and RCIC are required to be operated, in accordance with the Emergency Operating Procedures?

A. Both HPCI and RCIC must be tripped.

B. Both HPCI and RCIC must be left running.

C. HPCI must be tripped, RCIC must be left running.

D. HPCI must be left running, RCIC must be tripped.

Proposed Answer: c

Explanation: EOP-4 requires tripping HPCI if Torus water level cannot be maintained above 10.75 feet to prevent over-pressurizing the Containment by allowing HPCI turbine exhaust to enter directly into the Torus airspace. RCIC does not have a similar requirement due to its smaller size. With Reactor water level -5 inches and lowerin!~ at 700 psig, EOP-2 requires continued use of RCIC injection to preserve adequate core cooling as long as possible.

A Incorrect- RCIC does not need to be tripped even though its exhaust line may be passing steam directly into the Torus airspace.

B. Incorrect- HPCI must be tripped to prevent over-pressurizing the Containment, even though adequate core cooling is challenged.

D. Incorrect- HPCI must be tripped to prevent over-pressunizing the Containment, even though adequate core cooling is challenged. RCIC does not need to be tripped even though its exhaust line may be passing steam directly into the Torus airspace.

Technical Reference(s): EOP-4, EOP-2



Question Source: Bank- March 2012 NRC #7!~



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

Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ----~~~--------~~~~~~


High Off-Site Release Rate


RO 1 1 295038 EA 1.06 3.5

Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Plant ventilation



• A steam leak has developed from the RCIC supply piping into the Reactor Building. • Reactor Building ventilation exhaust radiation is 3 x 104 cpm and slowly rising. • RCIC room area temperature is 150°F and slowly rising. • Reactor Building differential pressure is +0.1" H20 and stable. • The Shift Manager has declared that the plant is in an Unusual Event due to offsite

release rate from the Reactor Building ventilation exhaust.

Given the following portion of EOP-5, Secondary Containment Control:

REACTOR BUILDING AREA TEMPERATURIES

AREA INSTRUMENT MAXIMUM MAXIMUM AREA INSTRUMENT MAXIMUM MAXIMUM NORMAL SAFE NORMAL SAFE

Ractor UuUdin(C. fk'at:tor BuiklinA r 2 ft :\("Jftckntion

HU F ll.!"f ckvalkm 11outhcast

IU-t'f 1'\;\"1' 6(JRTJ).10(J OOtl-1tJ6.Panc:JOI)..-'\ .!:\RTf).U.!:C 1!,\-.!0.-i.-\.P:uad Ct'J-9" (K'JRTI).UIH bbTI-lt1K.PiandiJ9-... '\ 1::\RTD-O.!J) 2.\-.!0-18.P..Jfk'llt9AX•

<.Jut!UOC '.-\' (JJ(.I lfPCil>rrwcU Eauran~:c 8;~.Uc:T}" Em .. inrllltC IIJ4'f 11.\"f I -\R1'1>-IO.!( 1:\-.!0J!C.Pand I~ uo·F l'\l'f

filJiitTU-11'\ EPIC Out,· 1:\RTl>-Wll) l\-.!02U.P.andt~

Ht'k"'· &dud Fluor RCI<: (lf}~'(:U Etttr.ml:c

ExhaWlt IU4"f ll.i"f 1:\RTO.Wl.A 1.\-JUlA. P;md IKJ...f,n lltff ll8"f ()C,Iff0-10'\ ()('1'1-10'\.P'.an~l 09-...,.'\ 1:\RTO.IO""D 1\-.!0'"'B.Pafl(ll~

OUbkk 'B' LPCl lk'...:wr 8uikJl8._, z-z ft Haucrr F .. ndusurc IU.f"f 11,\'F ckv;~hon "'Ml\'C'It

10.-'f ltK•"f ()(lR.ll>-116 EPICUnt,.· .Z:\Rll)..4HC .!.\-!02:.-\. Panel tJQ..'i"i

.!-\Rl1).4UI) 2:\-.!0..!8. P".md ~~ SL<: Pump Area • 10-f'F IH'F '.\' RIIR Ucat

bClRll>-ll·i EPICOnt,· fy~·rRonm

1.\0'f l.fl"f Fuel ~~,} <:uolb~A. .!:\Rll).(HA l.\-.!OIA.P.ulc:ll)l)..'}"l

Pump Room ,., .. -p L\;\'F .!.\R'flMHU l;\-..!018. P".tnd I)()..IJ(t

(,{tRll>-11:\ EPI<:ontr Turns Runm • S.uud•

Rcac.:tor UuiWinM :\'1)1) ft lti'UStcamlkk lltrf .lHU"F

dc,·.auon nonhcut • IO.f"f 1'\M"f I :\RTO-l(r<: I J-..!0-'l:. Pmc:l1)1}.9"\

('bKTl>-112 EPJ<: ·~·~·

l_iR'TD-101> I \-l(f"l>. P.utd ~~

R"'Cl T fkat Ext.:hanflt:t Tt..-m Rn111m • S.uudt"'C:'st RCJC ~1..--amlinc

Room 11'\'f ltJ.i''f I .\-2(r A. P:moCI 09-Qi

llO'f l*)"f llTE-11-E P.mdt)l)..ll 1:\RTO.Hi-.\

LZTE-II ... F P.mdOQ-..!1 1:\RTil-IU.ZR l:\-.!02:8.P.IIldltQ.JJ6

'8' n·nr Pump Room Eafii:Cr~cnt JIJ..f'f I i .... F tirE-ti-c P..md09-.ll ll'i"f ll"i' f t"tRTD-HttJ8 MTI-11190. P.uk:l CI'J. .,..,

llT£.11-u P.mdtl9-ll lfPCIIloom

'A' W'<:l1 Puf111) Room ..!:\R'f[).Q.4A l.\-.l9iA.P..mcllt'J.9'\

llTE-.Iil""A P:iru.i()Q-.2:1 ll'\'f ll'\'f .HRT().AN8 2 :\-.!9-lft.. P.utd •t9-96 tn .. ·p 1-\--F

IZTE-11 .. 8 P.md09-2:1 .!:\RlU-11'"".-\ l.\-.!1-.\. Pmd t)tW"i HRTD-11"'11 l_\-.!l""B.Pancltt').Q{t

R~.u.:tur HutldinM :\UU ft l!'k,·.atkm !i(M•h~~M 0 IU-4~F 1-1-F ROC Rnom

bllKTI>-111 EPICUnl)· liR1l).W.J.\ l.\-UW.\.P:utdf.N.91 10-&'F 11--F IIRTI""'R l :\-.!tWB. PancJ IJ9.')6

'8' RffR Heat Ex,:h.a.llAL-r Roum W'ol CrcM:cnt

.!-\R.TI>--0.!.\ 2.\-lO.iA. Panel 09-9i 1,\fl f l-l . .!'f liRil>-""bA 1.\-l""bA. P.1ucl 09-9i IOt'f I i ... "F

l:\&TI>-0.!8 .!,\-20.\8. P.1nd lt9-9C• IIRID-oR t .\-rOK. Po~nclct9-Qtl

Which one of the following identifies the number of entry conditions met for EOP-5, Secondary Containment Control, and EOP-6, Radioactivity Release Control?

EOP-5 EOP-6

A. 3 1

B. 3 0

C. 2 1

D. 2 0

Proposed Answer: B

Explanation: EOP-5 entry is required due to Reactor Building ventilation exhaust radiation (>1x103 cpm), RCIC room area temperature (>104°F), and Reactor Building differential pressure (at or above 0" H20). EOP-6 entry is not required, because although Reactor Building exhaust radiation is elevated, it only correlates with the Unusual Event level. EOP-6 entry is only warranted if offsite release rate exceeds the higher Alert level.

Note: This question matches the KIA by requiring the candidate to monitor Reactor Building ventilation parameters (D/P, exhaust radiation level) relate!d to a radioactivity release outside of Secondary Containment to make EOP entry decisions.

A. Incorrect- No EOP-6 entry condition is met by the give'n conditions. C. Incorrect- Three EOP-5 entry conditions are met and no EOP-6 entry condition is met by

the given conditions. D. Incorrect- Three EOP-5 entry conditions are met by the given conditions.

Technical Reference(s): EOP-5, EOP-6, ARP-09-·3-2-40, ARP-09-3-3-2(12), EALs


Learning Objective: MIT-301.11 F 1.04, MIT-301.11 G 6.05


Question History:



Comments:



RO 1

Importance Rating

1 295025 EA1.07 4.1

High Reactor Pressure

Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: ARI/RPT/ATWS: Plant-Specific



• A malfunction occurs with the Main Turbine pressure regulator. • Reactor pressure rises. • An Operator places the Reactor Mode Switch in SHUTDOWN. • All control rods insert. • Reactor pressure rises to a high value of 1170 psi!~ and then lowers to 920 psig on

Turbine Bypass Valves. • Reactor pressure was above 1 080 psig for two seconds during the transient.

Which one of the following describes the expected status of the Alternate Rod Insertion (ARI) solenoids and the Reactor Water Recirculation (RWR) pumps after this transient?

ARI Solenoids RWR Pumps

A. Energized Tripped

B. Energized Running

C. De-energized Tripped

D. De-energized Running

Proposed Answer: A

Explanation: Reactor pressure rising above 1153 psig actuates both the ARI and A lWS-RPT logic. The ARI logic energizes the ARI solenoids to vent the scram air header and cause a alternate Reactor scram method. The A lWS-RPT logic trips the RWR pumps. There is no time delay on the high pressure actuation of ARI and A lWS-RPT logic.

B. Incorrect- RWR pumps are expected to be tripped due to A lWS-RPT logic actuation. C. Incorrect- ARI solenoids are expected to be energized due to ARI logic actuation. D. Incorrect- ARI solenoids are expected to be energized due to ARI logic actuation. RWR

pumps are expected to be tripped due to A lWS-RPT logiG actuation.

Technical Reference(s): OP-25, OP-27


Learning Objective: SDLP-03C 1.05.c.8, SDLP-02H 1.05.c.2


Question History:



Comments:



Main Turbine Generator Trip


RO 1 1 295005 AA2.07 3.5

Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Reactor water level



• Feedwater pump A is operating in automatic. • Reactor water level is 200 inches and stable.

Then, a Main Turbine trip occurs.

Which one of the following describes the effect on indicated narrow range Reactor water level IMMEDIATELY after the Main Turbine trip?

IMMEDIATELY after the Main Turbine trip, indicated narrow range Reactor water level ...

A. rises due to the effect of control rod insertion on voids.

B. lowers due to the effect of control rod insertion on voids.

C. rises due to the effect of the pressure transient on voids.

D. lowers due to the effect of the pressure transient on voids.

Proposed Answer: D

Explanation: A Main Turbine trip is accomplished by rapid closing of the Turbine Stop Valves. This quickly lowers steam flow from the Reactor vessel and raises Reactor pressure. Turbine Bypass Valves then open to limit the pressure rise, but not be~fore Reactor pressure initially rises. Rising Reactor pressure collapses voids located in the core region inside the shroud. When these voids collapse, water from the annulus region outside the shroud lowers as it moves to fill the space previously occupied by voids inside th1e shroud. Narrow range Reactor water level measures level in the annulus region. Therefore, narrow range Reactor water level initially lowers due to these pressure effects.

A. Incorrect- Reactor water level initially lowers. Control rods do not insert because initial Reactor power was below 29%.

B. Incorrect- Control rods do not insert because initial Reactor power was below 29%. C. Incorrect - Reactor water level initially lowers.

Technical Reference(s): AOP-2, UFSAR Chapter 14 Section 5


Learning Objective: SDLP-94A 1.09.b

Question Source: Modified Bank- NMP1 2010 NRC #41

Question History:



Comments:



Form ES-401-5


295019 AA2.01 3.5

Partial or Complete Loss of Instrument Air

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Instrument air system pressure



• A large leak develops in the Instrument Air header. • AOP-12, Loss of Instrument Air, is being executed. • Operators in the field are attempting to isolate the leal<. • Air pressure is 70 psig and lowering. • Annunciator 09-5-2-3, ROD DRIFT, is in alarm. • One control rod has begun drifting into the core.

Which one of the following describes the requirement for a manual Reactor scram per AOP-12?

A Reactor scram ...

A. is required at this time.

B. is NOT yet required, but will be if another control rod begins to drift.

C. is NOT yet required, but will be if air pressure lowers below 65 psig.

D. is NOT yet required, but will be if the operators report the leak is un-isolable.

Proposed Answer: A

Explanation: AOP-12 contains an override that requires a manual Reactor scram given either of two conditions: (1) Air pressure continues to trend downward and corrective action cannot be immediately accomplished or (2) Any control rod drift alarm oGcurs. The given conditions have a control rod drifting. This meets the second requirement for a manual Reactor scram.

B. Incorrect- A manual scram is already required. A second drifting control rod is part of the scram requirement in AOP-27.

C. Incorrect- A manual scram is already required. 65 psig is the setpoint for the low CRD air pressure alarm, which is expected in this situation.

D. Incorrect- A manual scram is already required. Determination that corrective action cannot be immediately accomplished is part of the scram override in AOP-12.



Learning Objective: SDLP39 1.15.a


Question History:



Comments: TRH 3/25/14- Added bullet regarding ROD DRIFT annunciator in alarm, based on NRC comment.



Form ES-401-5


295001 AA2.03 3.3

Partial or Complete Loss of Forced Core Flow Circulation

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Actual core flow



• Both Reactor Water Recirculation (RWR) pumps in operation. • A malfunction occurs in the RWR controller system.

The plant stabilizes at the following conditions:

• Reactor power is 45% and steady. • RWR flow is 35% of rated core flow and steady.

Note: The expanded two loop power - flow map is provided on the next page.

Which one of the following actions is required per AOP-8, Loss or Reduction of Reactor Coolant Flow?

A. Lower RWR pump speed.

B. Manually scram the Reactor.

C. Raise RWR pump speed or insert rods.

D. Commence a Reactor shutdown per OP-65, Startup and Shutdown.

95

90

85

80

75

i 70 0

0..

; 65

a: • 60

55

50

45

40

35

EXPANDED POWER - FLOW MAP Two Loop Operation

20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 (IU) (23.1) (30.8) (38.5) (46.2) (53.8) ($l6) ($$.3) (17.0)

• Rated Core Flow ~)

Proposed Answer: c

Explanation: The given power and flow indications place the plant in the Buffer Zone of the Two Loop Power-Flow Map. AOP-8 requires either inserting control rods or raising RWR pump speed to exit the Buffer Zone.

A. Incorrect- While lowering RWR pump speed would lower power, it simultaneously lowers core flow. This does not lead to exiting the Buffer Zone atnd is not the action directed by AOP-8.

B. Incorrect- The plant conditions are NOT in the red area of the Power-Flow Map, NOR are there indications of THI.

D. Incorrect- AOP-8 does NOT require a complete shutdown, just exiting the Buffer Zone.

Technical Reference(s): AOP-8, Power-Flow l\llap


Learning Objective: SDLP-02H 1.09.b, 1.15.a


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~----~~---------

Examination Outline Cross-Reference: Level Tier# Group# KIA# Importance Rating

Generator Voltage and Electric Grid Disturbances

Equipment Control: Knowledge of surveillance procedums.



RO 1 1 700000 2.2.12 3.7

• Preparations are underway to perform ST-988, EDG 13 and D Full Load Test and ESW Pump Operability Test.

• EDG 8 has been declared inoperable, but has NOT yet been started. • Then, grid voltage begins to degrade. • AOP-72, 115 KV Grid Loss, Instability, or Degradation, is entered. • Power Control informs the Control Room that the low voltage post contingency alarm

has been received. • Grid voltage has stabilized, but is still at a degraded level.

Which one of the following describes how to control the Emergency Diesel Generators (EDGs)?

A. Do NOT continue with ST-988. Restore EDG 8 operability and maintain all EDGs in standby.

B. Do NOT continue with ST-988. Start and load all EDGs per OP-22, Diesel Generator Emergency Power.

C. Do NOT continue with ST-988. Start and load either E:DGs A and Cor EDGs 8 and D per OP-22, Diesel Generator Emergency Power. Maintain the other pair of EDGs in standby.

D. Continue with performance of ST-988 because AOP-72 requires EDG operability to be verified in response to the low voltage post contingenc:y alarm.

Proposed Answer: A

Explanation: During performance of ST-988, EDGs are declared inoperable. AOP-72 requires minimizing testing and inoperability of critical components of the electrical distribution system during degraded grid conditions. Therefore ST-988 should not be performed while in AOP-72.

B. Incorrect- AOP-72 requires returning inoperable EDGs to operable status as soon as possible, but does not require starting and loading EDGs, such as is required in AOP-13 in response to a hurricane warning.

C. Incorrect- AOP-72 requires returning inoperable EDGs Ito operable status as soon as possible, but does not require starting and loading EDGs, such as is required in AOP-13 in response to a hurricane warning.

D. Incorrect- AOP-72 requires minimizing testing and inop~3rability of critical components of the electrical distribution system during degraded grid conditions.

Technical Reference(s): ST -988, AOP-72


Learning Objective: SDLP-71 D 1.14.c


Question History:


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

Comments: TRH 3/8/14 - Revised stem conditions and choices to raise plausibility of distractors, based on NRC comment.

ES-401 Written Examination Question W()rksheet Form ES-401-5 ~~~------~~~~~~


Partial or Complete Loss of Component Cooling Water

RO 1 1 295018 2.1.32 3.8

Conduct of Operations: Ability to explain and apply systom limits and precautions.



• A loss of coolant accident occurs. • RBCLC and ESW flow to Drywell components is lost.

Which one of the following describes the required control of RBCLC and ESW flow to Drywell components and the reason for this response, in accordance with AOP-11, Loss of Reactor Building Closed Loop Cooling, and OP-40, Reactor Building Closed Loop Cooling?

A. Re-establish RBCLC and/or ESW flow to Drywell components to ensure Drywell cooling capacity meets design assumptions.

B. Re-establish RBCLC and/or ESW flow to Drywell components to prevent further degradation of Recirculation pump seals.

C. Do NOT re-establish RBCLC and/or ESW flow to Drywell components due to the potential for damage from severe water hammer.

D. Do NOT re-establish RBCLC and/or ESW flow to Drywell components due to the potential for damage from severe thermal stresses.

Proposed Answer: c

Explanation: Both a caution in AOP-11 and OP-40 P&L C.:2.2 direct NOT re-establishing RBCLC/ESW flow to the Drywell if lost during LOCA conditions. The reason for this restriction is to avoid damage from water hammer that could jeopardize primary containment integrity.

A. Incorrect- RBCLC/ESW flow is NOT allowed to be re-established. B. Incorrect- RBCLC/ESW flow is NOT allowed to be re-established. D. Incorrect- The concern is water hammer, not thermal stress.

Technical Reference(s): OP-40, AOP-11


Learning Objective: SDLP-151.13


Question History:




ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~------~------~~


Refueling Accidents


RO 1 1 295023 2.2.44 4.2

Equipment Control: Ability to interpret control room indic:ations to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.


You are on-watch in the Control Room with refueling activities in progress. The following conditions exist:

• The Reactor Mode Switch is in REFUEL. • All control rods are fully inserted. • The Refuel Bridge main hoist is loaded with a fuel bundle and over the core. • Annunciator 09-5-2-02, ROD BLOCK, is NOT in alarm. • SRM A indicates 100 cps with oo period. • SRM B indicates 110 cps with oo period. • SRM C indicates 90 cps with oo period. • SRM D indicates 115 cps with oo period.

Which one of the following describes the impact of the rod block and SRM indications on continued fuel movement?

A. These indications are appropriate for continued fuel movement.

B. You must direct the Refuel Bridge to stop fuel movem1ent due to improper operation of the REFUELING INTERLOCKS rod block. SRM indications are appropriate for continued fuel movement.

C. You must direct the Refuel Bridge to stop fuel movement due to improper SRM indication. REFUELING INTERLOCKS rod block operation is appropriate for continued fuel movement.

D. You must direct the Refuel Bridge to stop fuel movem1ent due to improper operation of the REFUELING INTERLOCKS rod block and improp1er SRM indications.

Proposed Answer: B

Explanation: With the Refuel Bridge hoist loaded with an irradiated fuel bundle and located over the core, the REFUELING INTERLOCKS rod block should cause annunciator 09-5-2-02 to be alarming. Fuel movement must be stopped until this malfunction is fixed.

A. Incorrect- With the Refuel Bridge hoist loaded with an irradiated fuel bundle and located over the core, the REFUELING INTERLOCKS rod block should cause annunciator 09-5-2-02 to be alarming. Fuel movement must be stopped until this malfunction is fixed.

C. Incorrect- SRMs are indicating appropriately for fuel movements (> 3 cps and stable). D. Incorrect- SRMs are indicating appropriately for fuel movements(> 3 cps and stable).

Technical Reference(s): ST-20F, OSP-66.001


Learning Objective: SDLP-03F 1.05.b.1.j


Question History:



Comments: TRH 3/25/14- Deleted extra words in choice A, based on NRC comment.



------------------------RO 1 1

Importance Rating 295026 EA2.02 3.8

Suppression Pool High Water Temperature

Ability to determine and/or interpret the following as the~' apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool level


The plant is operating at 100% power when a Main Turbine trip and high-power A TWS result in the following :

• Reactor pressure is 850 psig and slowly rising. • The current Reactor pressure control band is 800 to 1000 psi g. • Torus water temperature is 176°F. • Torus water level is 15.0 feet.

Which one of the following describes the impact of these parameters on the required Reactor pressure control strategy, in accordance with EOP-3, Failure to Scram?

Heat Capacity Temperature Limit

220

lfl 200 ~ = r: - .._. s 1111: ~ Torus 1-

~ ~ 1-0

...... ~ ~ -- Level [ft]: 1-......... 180 ..... :""" ::: li 1-

(U .....

!! =s 16 1-I.. ..... ..... :I ...

1-..... 14 ~ s ~ !i 1-nl I.. ...... 17.85 1-(U 160 1-c.. - r- 12 E ..... 1-

~ - 1-- 1--VI 140 - 1-2 ...... 1-0 9.58 1-

I-

120

0 200 400 600 BOO 1000 1200

RPV Pressure [psig]

A. Emergency depressurization is required .

B. Reactor pressure must be maintained at or below approximately 900 psig.

C. Reactor pressure must be maintained at or above approximately 900 psig.

D. Continued use of the entire current Reactor pressure band is acceptable.

Proposed Answer: 8

Explanation: With Torus water level at 15.0 feet, the 14 feet HCTL curve must be used. With Torus water temperature at 176°F, this HCTL curve requires Reactor pressure to be below approximately 900 psig.

A. Incorrect- With current Reactor pressure at 850 psig, op,eration is still in the "good" region of the HCTL curve, therefore emergency depressurization is not required.

C. Incorrect- Reactor pressure must be maintained below SIOO psig, not above, to satisfy HCTL.

D. Incorrect- Reactor pressure may not go into the upper half of the current pressure band without violating HCTL. This would be the correct answer if the 16 feet HCTL curve was erroneously used.

Technical Reference(s): EOP-4


Learning Objective: MIT-301.11 E 4.05 and 4.07

Question Source: Bank- 2009 NMP1 NRC #15

Question History:



Comments:

ES-401 Written Examination Question W..::.o.:...:rk~s:..:.h:..::e..::.e..:...t ___ ..:..F..::.o.:...:rm~E::...:S:....-4....:.0..:...1.:....-...:..5


Partial or Complete Loss of D.C. Power


RO 1 1 295004 AK 1. 05 3.3

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Loss of breaker protection


The plant is operating at 1 00% power when an electrical fault causes a loss of DC power system B.

Which one of the following describes the impact of this failum on Bus 10600 breakers?

All Bus 10600 breakers ...

A. fail open.

B. remain in their initial position, CANNOT be opened or closed from the Control Room, and lose all automatic trip capabilities.

C. remain in their initial position, CANNOT be opened or closed from the Control Room, and lose all automatic trip capabilities except for the overcurrent trip.

D. continue to operate normally with control power supplied from DC power system A.

Proposed Answer: B

Explanation: Control power for Bus 10600 breakers is supplied by DC system B, with no automatic transfer to DC system A With no control power to Bus 10600 breakers, they remain in their initial position, lose the ability to be opened/closed from the Control Room, and lose automatic trip capabilities.

A Incorrect- Although the breakers lose control power, they remain in their initial positions. C. Incorrect- The breakers lose all automatic trip capabilities. D. Incorrect- Bus 10600 breakers are supplied control power from DC system B, with no

automatic transfer to DC system A



Learning Objective: SDLP-71E 1.10.c


Question History:



Comments: TRH 3/6/14 - Edited B and C to make C more plausible, based on NRC comment.



High Drywell Pressure

------------------------Level Tier# Group# KIA# Importance Ratin!~

RO 1 2 295010 AK1.03 3.2

Knowledge of the operational implications of the followin!g concepts as they apply to HIGH DRYWELL PRESSURE: Temperature increases



• Changes in Drywell parameters have been observed over the last 12 hours as follows:

12 Hours A o Now 130°F

2.2 psig 100 cQ_m 0.2 gpm 0.8 gpm

• Both Drywell cooling assemblies are running. • Each Drywell cooling assembly has three fans runnin!l

Which one of the following describes a malfunction that is consistent with these indications and an appropriate operator action in response to the malfunction?

Malfunction Operator Action

A. Reactor coolant leak Run four fans on each Drywell cooling into the Drywell assembly to lower Drywell temperature

B. Reactor coolant leak Vent the Containment to maintain Drywell into the Drywell pressure below limits

C. Degraded Drywell cooling Run four fans on each Drywell cooling assembly performance assembly to lower Drywell temperature

D. Degraded Drywell cooling Vent the Containment to maintain Drywell assembly performance pressure below limits

Proposed Answer: D

Explanation: Degraded Drywell cooling assembly performance causes Drywell temperature to rise. Since the Drywell is a closed space, this temperature rise causes a pressure rise. Rising Drywell pressure erodes margin to the high Drywell pressure scram setpoint, therefore Containment venting is appropriate to lower Drywell pressur~e. The daily surveillance checks require Drywell pressure s 1.95 psig.

A. Incorrect- If there was a Reactor coolant leak into the Drywell, the CAM and Drywell floor drain leakage would rise.

B. Incorrect- If there was a Reactor coolant leak into the Drywell, the CAM and Drywell floor drain leakage would rise.

C. Incorrect- Running four fans on each Drywell cooling assembly is prohibited except for a short period of time during fan swaps. Prolonged operation with four fans risks overloading fan motors.

Technical Reference(s): OP-53, OP-37, SDLP-168, ARP 09-5-1-34




Question History:



Comments:

ES-401 Written Examination Question Wc•rksheet Form ES-401-5 --~~------------------


High Containment Hydrogen Concentration

RO 1 2 500000 EK2. 03 3.3

Knowledge of the interrelations between HIGH CONTAINMENT HYDROGEN CONCENTRATIONS the following: Containment Atmosphere Control System



• EOP-4A, Primary Containment Gas Control, is being e~xecuted due to slightly elevated hydrogen concentrations in both the Torus and the Dr)IWell.

• Hydrogen and oxygen concentrations in the Torus and Drywell are the same. • Offsite release rate is expected to stay below LCO limits during any venting operation. • T crus water level is 14 feet and stable. • The CRS directs venting and purging the Containment per the preferred strategy in EP-

6, Post-Accident Containment Venting and Gas Control.

Which one of the following describes the preferred venting and purging strategy, in accordance with EP-6?

A. Vent from the Torus and purge to the Torus.

B. Vent from the Torus and purge to the Drywell.

C. Vent from the Drywell and purge to the Torus.

D. Vent from the Drywell and purge to the Drywell.

Proposed Answer: B

Explanation: In EP-6, the preferred method of venting is from the Torus (with purge to the Drywell) to maintain pressure suppression function, scrub gases before release, and ensure effective vent and purge of both the Drywell and Torus. With Torus water level below 19.5 feet, T crus venting is available.

A. Incorrect- In EP-6, the preferred method of venting is from the Torus (with purge to the Drywell) to maintain pressure suppression function, scrub gases before release, and ensure effective vent and purge of both the Drywell and Torus.

C. Incorrect -In EP-6, the preferred method of venting is from the Torus (with purge to the Drywell) to maintain pressure suppression function, scrub gases before release, and ensure effective vent and purge of both the Drywell and Torus.

D. Incorrect- In EP-6, the preferred method of venting is from the Torus (with purge to the Drywell) to maintain pressure suppression function, scrub gases before release, and ensure effective vent and purge of both the Drywell and Torus.

Technical Reference(s): EOP-4A, EP-6


Learning Objective: MIT-301.11 E


Question History:


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

Comments: TRH 3/6/14- Fixed typo ("from" vs. "to") in choice D, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 .~~~~----~~~~~~


Loss of Main Condenser Vacuum


RO 1 2 295002 AK3.01 3.7

Knowledge of the reasons for the following responses as they apply to LOSS OF MAIN CONDENSER VACUUM: Reactor SCRAM: Plant-Specific


The plant is operating at 32% power with the following :

• Circulating Water pumps A and B have tripped. • Main Condenser vacuum has lowered and is now stable at 24.7 inches Hg. • Main Generator output is 240 MWe. • Given the following portion of AOP-31 , Loss of Condenser Vacuum:

ATTACHMENT 1 Page 1 of 1

CONDENSER VACUUM VS. TURBINE LOAD

27.00·~-------------------------------------------------------,

co X

:§. 25.50

~

5 25.00 ~ ffi 24.50 VI z w ~ 24.00 8

23.50

; TURBINE LOAD (Mwe)

4.0

Which one of the following describes the need for a manual Reactor scram, in accordance with AOP-31?

A manual Reactor scram .. .

A. is NOT required due to the current Reactor power level.

B. is NOT required due to Main Condenser vacuum being stable above 24.0".

C. is required due to proximity to the Reactor feed pump low vacuum trip setpoint.

D. is required due to Main Turbine operation restrictions with low vacuum at low load.

Proposed Answer: D

Explanation: AOP-31 requires a manual Reactor scram if Reactor power is above 29% in either of the following situations:

(1) Turbine load is less than 267 MWe and vacuum is less then 25.5 inches Hg (2) Condenser vacuum lower to or is rapidly approaching the Main Turbine trip setpoint

(22.5 inches Hg).

In the given conditions, Condenser vacuum is below 25.5 inches Hg with Turbine load less than 267 MWe (low vacuum at low load) and Reactor power above 29%, therefore a manual Reactor scram is required.

A. Incorrect- A manual Reactor scram is required due to Main Turbine load less than 267 MWe with Condenser vacuum below 25.5 inches Hg and Reactor power above 29%.

B. Incorrect- A manual Reactor scram is required due to Main Turbine load less than 267 MWe with Condenser vacuum below 25.5 inches Hg and Reactor power above 29%. 24.0" is the normal benchmark for ordering a manual scram based on lowering Main Condenser vacuum.

C. Incorrect- Condenser vacuum has stabilized with over 4 inches Hg of margin above the Reactor feed pump trip setpoint (20 inches Hg), therefore a manual Reactor scram is not required based on approaching the Reactor feed pump trip.



Learning Objective: SDLP-94A 1.14.e


Question History:



Comments:



Form ES-401-5

Level Tier# Group# KIA# Importance Ratin!~

295007 AA 1. 04 3.9

High Reactor Pressure

Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Safety/relief valve operation: Plant-Specific


The plant is operating at 100% power when MSIVs spuriously close and the Reactor fails to scram.

Which one of the following describes the Reactor pressure at which the first SRVs will begin to open and the Reactor pressure at which the last SRVs will begin to open?

The first SRVs will begin to open at. ..

A. 1080 psig and the last SRVs will begin to open at 1145 psig.

B. 1080 psig and the last SRVs will begin to open at 1250 psig.

C. 1135 psig and the last SRVs will begin to open at 1145 psig.

D. 1135 psig and the last SRVs will begin to open at 1250 psig.

Proposed Answer: c

Explanation: SRV electric lift setpoints are as follows:

Valve 02RV-71 K, L 02RV-71 D, E 02RV-71 A, B, C, F, G, H, J

Setpoint 1135 psi.g 1140 psig 1145 psig

Therefore the first SRVs begin to lift at 1135 psig and the last SRVs begin to lift at 1145 psig.

A. Incorrect- The first SRVs begin to open at 1135 psig. 1080 psig is associated with the Reactor scram setpoint.

B. Incorrect- The first SRVs begin to open at 1135 psig. 1080 psig is associated with the Reactor scram setpoint. The last SRVs being to open at 1145 psig. 1250 psig is the RPV design pressure.

D. Incorrect- The last SRVs being to open at 1145 psig. 1.250 psig is the RPV design pressure.

Technical Reference(s): SDLP-020

Proposed references to be provided to applicants during exc:tmination: None

Learning Objective: SDLP-02J 1.05.a.1


Question History:



Comments:

ES-401 Written Examination Question Wmksheet Form ES-401-5 ------------------------Examination Outline Cross-Reference: Level


Secondary Containment High Sump/Area Water Level

RO 1 2 295036 EA2.01 3.0

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Operatbility of components within the affected area



• The Mode Switch is in STARTUP/HOT STANDBY. • Reactor pressure is 100 psig and stable. • A Reactor Building floor drain high level alarm is received. • EOP-5, Secondary Containment Control, has been entered. • Investigation reveals a fire header is leaking in the East Crescent. • Area water level is 18 inches and rising.

Which one of the following describes equipment which is required to be operable under these plant conditions and whose operability may be challenged by the location of this leak?

A. RHR Pump D and HPCI

B. Core Spray Pump B and RCIC

C. RHR Pump A and Core Spray Pump A

D. RHR Pump B and Core Spray Pump B

Proposed Answer: D

Explanation: High area water level in the Reactor Building is of concern due to the potential effects on operability of equipment in the area. A high area water level in the East Crescent is of concern to HPCI, Core Spray Pump B, and RHR Pumps Band D, but not RCIC, Core Spray Pump A, or RHR Pumps A and C. In Mode 2 below 150 psig, neither HPCI nor RCIC are required to be operable. Therefore only Core Spray Pump B and RHR Pumps B and D are both required to be operable and affected by this leak location.

A. Incorrect- HPCI is affected by the leak, but not required to be operable with Reactor pressure less than 150 psig.

B. Incorrect- RCIC is in the West Crescent, not the East Crescent. C. Incorrect- RHR Pump A and Core Spray Pump A are in the West Crescent, not the East

Crescent.

Technical Reference(s): SDLP-1 0, 13, 14, 23; Technical Specifications 3.5.1 and 3.5.3


Learning Objective: SDLP-23 1.16


Question History:


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

Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ------------------------Examination Outline Cross-Reference: Level


High Secondary Containment Area Temperature

RO 1 2 295032 2.4.45 4.1

Emergency Procedures I Plan: Ability to prioritize and iint4upret the significance of each annunciator or alarm.



• Annunciator 09-3-3-02, DIV I AMBIENT TEMP HI, alarms. • EPIC point 1275, HPCI ROOM, is alarming above the trip setpoint of 133°F. • EPIC point 1273, HPCI ROOM, is also alarming above the trip setpoint of 133°F. • No other annunciators or EPIC points are in alarm.

Which one of the following describes the significance of these alarm conditions?

HPCI. ..

A. remains un-isolated.

B. isolates due to closure of the inboard isolation valve, only.

C. isolates due to closure of the outboard isolation valve, only.

D. isolates due to closure of both the inboard and outboard isolation valves.

Proposed Answer: B

Explanation: Multiple temperature detectors input to annunciator 09-3-3-02. The two given EPIC points indicate that two detectors are giving an isolation signal (on Division I) to HPCI. The HPCIIogic is setup such that the inboard isolation valve closes on these signals alone. Since annunciator 09-3-3-12 is NOT also in alarm, the outboard isolation valve does NOT close.

A. Incorrect- Even though annunciator 09-3-3-12 is not also in alarm, the HPCI inboard isolation valve does close, isolating HPCI.

C. Incorrect- Since annunciator 09-3-3-12 is not in alarm, the HPCI outboard isolation valve does not isolate.

D. Incorrect- Since annunciator 09-3-3-12 is not in alarm, tlhe HPCI outboard isolation valve does not isolate.

Technical Reference(s): ARP-09-3-3-02


Learning Objective: SDLP-23 1.09.d


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~---------------------


Secondary Containment Ventilation High Radiation

RO 1 2 295034 EK3.02 4.1

Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Starting SBGT/FRVS: Plant-Specific


Which one of the following describes the basis for having the Standby Gas Treatment system automatically start on high ventilation radiation levels?

A. Ensures offsite radiation doses are maintained below 10 CFR 100 limits.

B. Ensures Control Room personnel radiation doses are maintained below 10 CFR 50 limits.

C. Ensures equipment operability in the Reactor Buildin!;J is maintained to support safe shutdown requirements.

D. Ensures personnel access to the Reactor Building is maintained to support safe shutdown requirements.

Proposed Answer: A

Explanation: Standby Gas Treatment (SGT) automatically initiates on high Reactor Building ventilation exhaust radiation to provide a filtered, elevated release of fission products during accident conditions to ensure 10 CFR 100 limits for offsite radiation dose are not exceeded.

B. Incorrect -This is the basis for Control Room ventilation and the Control Room habitability program.

C. Incorrect- This is part of the reason behind max safe radiation levels and the associated actions in EOP-5.

D. Incorrect- This is part of the reason behind max safe radiation levels and the associated actions in EOP-5.

Technical Reference(s): UFSAR section 5.3.4


Learning Objective: SDLP-01 B 1.02


Question History:



Comments: TRH 3/10/14- Replaced original distractor A, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 -----------------------Examination Outline Cross-Reference: Level

Tier# Group# KIA# Importance Ratin!~

RO 3

2.1.14 3.1

Knowledge of criteria or conditions that require plant-widte announcements, such as pump starts, reactor trips, mode changes, etc.


Given the following events:

(1) Commencing radiography activities. (2) Fire requiring fire brigade response. (3) Personnel injury requiring first aid team response.

Which one of the following identifies which of these events require a plant-wide announcement AND sounding the Station Alarm, in accordance with OP-63, llntra-Piant Communications System?

A. (2) only

B. (3) only

C. (2) and (3) only

D. (1 }, (2), and (3)

Proposed Answer: B

Explanation: OP-63 posted attachment 10 contains plant-wide announcements for radiography activities, fires, and personnel injury. Radiography activities do not require any plant alarm. A fire requires fire alarm activation, but not station alarm activation. Personnel injury requires station alarm activation.

A. Incorrect- A fire requires fire alarm activation, but not station alarm activation. Personnel injury requires station alarm activation.

C. Incorrect- A fire requires fire alarm activation, but not station alarm activation. D. Incorrect- Radiography activities do not require any plant alarm. A fire requires fire alarm

activation, but not station alarm activation.

Technical Reference(s): OP-63 posted attachment 1 0


Learning Objective: EP-12.5.4.2 E0-1.05


Question History:



Comments: TRH 3/25/14- Enhanced emphasis in question, based on NIRC comment.



RO 3

2.1.5 2.9

Form ES-401-5

Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.



• The shift is at minimum manning. • Then, an NPO becomes injured while in a contaminated area. • The NPO is transported via ambulance to Oswego Hospital. • The RP Tech accompanies the NPO to the hospitaL • The ambulance leaves the site at 0300. • The on-coming shift is expected to arrive at 0600.

Which one of the following describes the requirement to repla1ce the NPO and the RP Tech in accordance with EN-OP-115, Conduct of Operations, and Technical Specifications?

A. Both the NPO and RP Tech must be replaced earlier than the expected arrival of the oncoming shift.

B. The NPO must be replaced earlier than the expected arrival of the on-coming shift. The RP Tech does NOT need to be replaced earlier than the expected arrival of the oncoming shift.

C. The RP Tech must be replaced earlier than the expected arrival of the on-coming shift. The NPO does NOT need to be replaced earlier than the expected arrival of the oncoming shift.

D. NEITHER the NPO NOR the RP Tech must be replaGed earlier than the expected arrival of the on-coming shift.

Proposed Answer: A

Explanation: Both NPOs and an RP Tech are part of required for shift manning. Since the shift started at minimum manning, loss of both the NPO and the RP Tech takes the shift below minimum manning in both of these positions. Technical Specifications and EN-OP-115 Attachment 9.5 Section 1 allow up to 2 hours to return shift manning to at least the minimum requirements. Since the on-coming shift is not expected to arrive for another 3 hours, both the NPO and the RP Tech must be replaced earlier than arrival of the on-coming shift.

B. Incorrect- The RP Tech is part of minimum shift mannin~1 and must be replaced within 2 hours (by 0500).

C. Incorrect- The NPO is part of minimum shift manning and must be replaced within 2 hours (by 0500).

D. Incorrect- Both the NPO and the RP Tech are part of minimum shift manning and must be replaced within 2 hours (by 0500).

Technical Reference(s): EN-OP-115, Technical Specifications


Learning Objective: LP-AP 46.01 F


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~------~---------


RO 3

2.2.2 4.6

Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.



• All SRMs are indicating five (5) count rate doublings over the initial count rate. • The Reactor is NOT yet critical. • The next control rod to be withdrawn is at position 12 and has a target position of 48. • The control rod is NOT identified as fast on the list of CRD deficiencies.

Which one of the following describes the ability to withdraw the control rod using continuous rod withdrawal, in accordance with OP-26, Control Rod Drive Manual Control, and OP-65, Startup and Shutdown?

A. The control rod may be withdrawn continuously from position 12 to 48.

B. The control rod must NOT be withdrawn continuously at any position from 12 to 48.

C. The control rod must be notch withdrawn from position 12 to 30. The control rod may be withdrawn continuously from position 30 to 48.

D. The control rod may be withdrawn continuously from position 12 to 16 and from position 30 to 48. The control rod must be notch withdrawn from position 16 to 30.

Proposed Answer: B

Explanation: OP-65 requires only notch control rod withdrawal during the approach to criticality once SRMs have achieved four (4) count rate doublings over the initial count rate. Since SRMs have exceeded this limitation and the Reactor is not yet critical, the control rod may not be withdrawn continuously.

A. Incorrect- The Reactor is sub-critical and SRMs are above four count rate doublings, therefore continuous rod withdrawal is not allowed. Th~s answer would be correct if SRMs were below three count rate doublings.

C. Incorrect- The Reactor is sub-critical and SRMs are above four count rate doublings, therefore continuous rod withdrawal is not allowed. This answer would be correct if SRMs were between three and four count rate doublings.

D. Incorrect- The Reactor is sub-critical and SRMs are above four count rate doublings, therefore continuous rod withdrawal is not allowed. This answer is based on concept in OP-65 step D.14.2.i.

Technical Reference(s): OP-26, OP-65


Learning Objective: SDLP-03F 1.13. b.2


Question History:

Question Cognitive Level: Memory or Fundamental Knowlledge

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

Comments:



RO 3

2.2.1 4.5

Form ES-401-5

Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.



• The Reactor Mode Switch is in STARTUP. • Control rods are being withdrawn. • The Rod Worth Minimizer (RWM) has just failed with 11 control rods withdrawn.

Which one of the following describes the requirements to continue the startup, in accordance with OP-64, Rod Worth Minimizer?

A. Bypass the RWM and then station a second individual with no concurrent duties to independently verify control rod movements. This individual must be a licensed operator. There is NO restriction on the last time such a startup has been performed.

B. Bypass the RWM and then station a second individual with no concurrent duties to independently verify control rod movements. This individual may be a licensed operator or a Shift Technical Advisor. There is NO restriction on the last time such a startup has been performed.

C. Verify that startup with RWM inoperable has NOT be,en performed in the same calendar year, bypass the RWM, and then station a second individual with no concurrent duties to independently verify control rod movements. This individual must be a licensed operator.

D. Verify that startup with RWM inoperable has NOT been performed in the same calendar year, bypass the RWM, and then station a second individual with no concurrent duties to independently verify control rod movements. This individual may be a licensed operator or a Shift Technical Advisor.

Proposed Answer: D

Explanation: With less than 12 rods withdrawn during a startup, OP-64 and TS 3.3.2.1 allow continuing startup only if startup with inoperable RWM has not been performed this calendar year and a second qualified individual is stationed to verify rod movements (licensed operator or STA).

A. Incorrect- The second individual may also be an STA. A startup with an inoperable RWM must NOT have been performed in the same calendar year.

B. Incorrect- A startup with an inoperable RWM must NOT have been performed in the same calendar year.

C. Incorrect- The second individual may also be an STA

Technical Reference(s): Tech Spec 3.3.2.1, OP-154 section E.1


Learning Objective: SDLP-030 1.15.A





Comments:



RO 3

2.3.12 3.2

Form ES-401-5

Knowledge of radiological safety principles pertaining tc1 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked highradiation areas, aligning filters, etc.


Personnel are preparing to enter the Drywell at power to inve~stigate a problem.

Which one of the following describes the requirements ofAP-12.02, Drywell Entries During Primary Containment?

The upper limit for Reactor power is (1) and the lower limit for oxygen concentration is (2)

(1) (2)

A. 15% 15%

B. 15% 19.5%

C. 25% 15%

D. 25% 19.5%

Proposed Answer: B

Explanation: AP-12.02 requires Reactor power to be less than 15% and Drywell oxygen concentration to be between 19.5% and 23.5% to allow personnel entry to the Drywell when Primary Containment is required.

A. Incorrect- The Drywell oxygen concentration limit is 19.5'Yo, not 15%. C. Incorrect- Reactor power must be less than or equal to 15%, not 25%. The Drywell oxygen

concentration limit is 19.5%, not 15%. D. Incorrect- Reactor power must be less than or equal to 15%, not 25%.

Technical Reference(s): AP-12.02


Learning Objective: LP-AP E0-45.04 and E0-45.05





Comments: TRH 3/25/14 - Edited 2nd column distractors from 23.5% to 15%, based on NRC comment.

ES-401 Written Examination Question Worksheet Form ES-401-5 --------------------------Examination Outline Cross-Reference: Level

Tier# Group# KIA#

RO 3

Importance Rating 2.3.14 3.2

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.


The Control Room has been evacuated due to a fire with the following:

• Operators are required to enter a radiologically posted area in order to manually close Primary Containment Isolation Valves.

• The highest general area dose rate in the area is 2500 mRem/hr.

Which one of the following describes the radiological requirements for the area and the entry?

This area is required to be posted as a (1) . While the operators are working in the area, the entrance (2) required to be locked or continuously guarded.

(1) (2)

A. Very High Radiation Area is

B. Very High Radiation Area is NOT

C. Locked High Radiation Area is

D. Locked High Radiation Area is NOT

Proposed Answer: c

Explanation: This is required to be posted as a Locked High Radiation Area because the dose rate is above 1000 mRem/hr. This is NOT required to be posted as a Very High Radiation Area because rad levels are below 500 rads/hour. Either of these areas are required to be locked except when continuously guarded or during periods of personnel/equipment entry/exit.

A. Incorrect- This is not required to be posted as a Very High Radiation Area (500 rad/hour). B. Incorrect- This is not required to be posted as a Very High Radiation Area (500 rad/hour).

Even when inside the area, the entrance needs to be either guarded or closed and locked to prevent unauthorized entry.

D. Incorrect- Even when inside the area, the entrance needs to be either guarded or closed and locked to prevent unauthorized entry.

Technical Reference(s): EN-RP-108


Learning Objective: LP-AP EO 31.02 and 31.03

Question Source: Bank- NMP1 2009 NRC #72

Question History:



Comments:

See Separate Attachment for Question #72

Due to Sensitive Security Information


Examination Outline Cross-Reference: Level RO Tier# 3 Group# KIA# 2.4.1 Importance Rating 4.6

Knowledge of EOP entry conditions and immediate actio,n steps.


A plant transient has resulted in the following:

• Reactor water level is 190 inches and rising slowly. • Reactor pressure is 1090 psig and steady. • Reactor power is downscale on APRMs. • Drywell pressure is 2 psig and rising slowly. • Drywell average temperature is 145°F and rising slowly. • Torus water level is 14.2 feet and rising slowly. • Torus temperature is 82°F and rising slowly. • Annunciator 09-4-1-29, RX BLDG EQUIP SUMP A LVL HI, is in alarm. • An operator reports Reactor Building Equipment Surnp A water level is one (1) foot

below floor level.

Which one of the following lists the Emergency Operating Procedures that are required to be entered based on current conditions?

A. EOP-2, RPV Control, and EOP-4, Primary ContainmE~nt Control, only

B. EOP-2, RPV Control, and EOP-5, Secondary Containment Control, only

C. EOP-4, Primary Containment Control, and EOP-5, SE~condary Containment Control, only

D. EOP-2, RPV Control, EOP-4, Primary Containment Control, and EOP-5, Secondary Containment Control

Proposed Answer: A

Explanation: EOP-2 entry is required due to Reactor pressure above 1080 psig. EOP-4 entry is required due to Torus water level above 14.0 feet and Drywall temperature above 135°F. No EOP-5 entry conditions are met. High Reactor Building fle>or drain sump level is an entry condition, but high Reactor Building equipment drain sump le~vel is NOT.

B. Incorrect- EOP-4 entry is required due to Torus water level above 14.0 feet and Drywall temperature above 135°F. No EOP-5 entry conditions are met. High Reactor Building floor drain sump level is an entry condition, but high Reactor Building equipment drain sump level is NOT.

C. Incorrect- EOP-2 entry is required due to Reactor pressure above 1080 psig. No EOP-5 entry conditions are met. High Reactor Building floor drain sump level is an entry condition, but high Reactor Building equipment drain sump level is 1\JOT.

D. Incorrect- No EOP-5 entry conditions are met. High Reactor Building floor drain sump level is an entry condition, but high Reactor Building equipment drain sump level is NOT.

Technical Reference(s): EOP-2, EOP-4, EOP-5

Proposed references to be provided to applicants during e~xamination: None

Learning Objective: MIT-301.11C 1.02, MIT-301.11E4.02, MIT-301.11F 1.04


Question History:



Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 -------------------------Examination Outline Cross-Reference: Level


RO 3

2.2.22 4.0

Knowledge of limiting conditions for operations and safe,ty limits.


The plant was operating at 100% power when the following occurred:

• The Main Turbine tripped. • Turbine Bypass Valves failed to actuate. • Reactor pressure peaked at 1345 psig. • Reactor water level control systems initially malfunctioned. • Reactor water level lowered to 20 inches before being restored.

Which one of the following describes the status of Safety Limits during this transient, in accordance with Technical Specifications?

A. No Safety Limits were exceeded.

B. The Safety Limit for Reactor pressure was exceeded, only.

C. The Safety Limit for Reactor water level was exceeded, only.

D. Both the Safety Limits for Reactor pressure and Reactor water level were exceeded.

Proposed Answer: B

Explanation: The Reactor pressure Safety Limit of 1325 psig was exceeded. The Reactor water level Safety Limit of 0" (top of active fuel) was not excet~ded.

A. Incorrect- Although the Reactor water level Safety Limit was NOT exceeded, the Reactor pressure Safety Limit was exceeded.

C. Incorrect- The Reactor water level Safety Limit was NOT exceeded, however the Reactor pressure Safety Limit was exceeded.

D. Incorrect- The Reactor water level Safety Limit was approached, but NOT exceeded.

Technical Reference(s): Technical Specification 2.1


Learning Objective: SDLP-02A 1.02





Comments:

ES-401 Written Examination Question Worksheet Form ES-401-5 ~~~~--------------------


RO 3

2.3.4 3.2

Knowledge of radiation exposure limits under normal or emergency conditions.


The plant is shutdown with the following:

• A 31 year old Operator is entering the Drywell for a jCib. • Expected dose for transiting to and from the job location is a total of 20 mRem. • Expected dose rate for the job location is 3 Rem/hr. • It will take 45 minutes to complete the job. • The Operator's previous TEDE for the year is 1710 rnRem. • No dose extension has been previously obtained for lthis Operator. • No emergency is in progress.

Which one of the following describes the Operator's expected dose, in accordance with EN-RP-201, Dosimetry Administration?

The Operator's expected dose ...

A. will stay within the normal annual administrative limit.

B. will exceed the normal annual administrative limit, but stay within the federal annual dose limit.

C. will exceed the normal annual administrative and federal dose limits. If it is still necessary to use this Operator, a Planned Special Exposure must be processed.

D. will exceed the normal annual administrative and federal dose limits. Under no circumstance can this Operator exceed the federal dose limits for this job because no emergency is in progress.

Proposed Answer: B

Explanation: The given job information will result in the Operator's annual dose rising to 3980 mRem (1710 mRem + 20 mRem + .75 hours* 3000 mRem/hour). This is above the normal annual administrative dose limit of 2000 mRem, but below the~ federal annual dose limit of 5000 mRem.

A. Incorrect- The expected total annual dose of 3980 mRem will exceed the 2000 mRem normal annual administrative dose limit.

C. Incorrect- The expected total annual dose of 3980 mRem will NOT exceed the 5000 mRem federal annual dose limit.

D. Incorrect- The expected total annual dose of 3980 mRem will NOT exceed the 5000 mRem federal annual dose limit. The federal annual dose limit can be exceed outside of an emergency using a Planned Special Exposure.

Technical Reference(s): EN-RP-201


Learning Objective: LP-AP 30.03.f


Question History:



Comments:



Low Suppression Pool Water Level


SRO 1 1 295030 EA2.03 3.9

Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Reactor pressure


The plant is operating at 100% power when a seismic event results in the following:

• The Reactor has scrammed. • Reactor pressure is 910 psig and slowly rising. • Two control rods remain withdrawn at position 36. • All other control rods are fully inserted. • All APRMs indicate downscale. • AIIIRMs indicate downscale on range 2. • No boron has been injected into the Reactor. • Torus water level is 10.75 feet and slowly lowering. • Operators are attempting to locate the Torus leak. • Operators are attempting to align Torus makeup.

Which one of the following describes the appropriate Reactor pressure control strategy, in accordance with the Emergency Operating Procedures?

A. Implement EOP-3. Reactor pressure must be stabilized near the current value. Reactor depressurization may NOT be commenced.

B. Implement EOP-3. Reactor pressure may be reduced, but the cooldown rate must be kept less than 1 00°F/hr.

C. Implement EOP-2. Reactor pressure may be reduced irrespective of the cooldown rate using Turbine Bypass Valves.

D. Implement EOP-2. Emergency RPV Depressurization is required using SRVs.

Proposed Answer: B

Explanation: Torus water level at 10.75 feet and lowering indicates that Emergency RPV Depressurization may be needed in the future if a determination is made that Torus water level cannot be maintained above 9.58 feet. Since Torus water level cannot be maintained above 10.75 feet, EOP-2 entry is required. The failure of two control rods at position 36 requires exit of EOP-2 and entry into EOP-3. Since EOP-2 is not in use, the pressure control override regarding anticipating Emergency Depressurization may not be used to violate cooldown rate. EOP-3 allows cooldown because the IRM indications show that the Reactor is currently shutdown and no boron has been injected into the Reactor. EOP-3 restricts this cooldown to less than 1 00°F/hr.

A. Incorrect- EOP-3 allows cooldown since IRM indications show that the Reactor is currently shutdown and no boron has been injected into the Reactor.

C. Incorrect- Since EOP-2 is not in use, the pressure contml override regarding anticipating Emergency Depressurization with Turbine Bypass Valves may not be used to violate cooldown rate. EOP-3 restricts the cooldown to less than 1 00°F/hr.

D. Incorrect- There is not enough information to determine that Torus water level cannot be maintained above 9.58 feet, therefore Emergency RPV Depressurization with SRVs is not yet required.

Technical Reference(s): EOP-4, EOP-2, EOP·-3


Learning Objective: MIT-301.11 C, MIT-301.11 D, MIT-301.11 E


Question History:



Comments: TRH 3/8/14- Added "Implement EOP-2/3" to each choice, based on NRC comment.

ES-401 Written Examination Question Wo.:..:r...:..:k:::s.:..:h:::..ee.::..t=-------=F~o:...:r...:..:m:..:....::E:.:S:.--4...:..0.:..1.:...-...::.5


Partial or Complete Loss of Instrument Air

SRO 1 1 295019 AA2.02 3.7

Ability to determine and/or interpret the following as the~· apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safet~-related instrument air system loads (see AK2.1 - AK2.19)


A failure to scram has occurred with the following:

• RCIC is out of service. • A complete loss of Instrument Air pressure occurred 45 minutes ago. • EOP-3, Failure to Scram, is being executed. • Reactor water level is being intentionally lowered. • Reactor water level is 90 inches and slowly lowering. • Reactor pressure is 950 psig and slowly lowering with two SRVs open. • Reactor power is 10% and slowly lowering. • Torus temperature is 140°F and slowly rising. • Drywell pressure is 1.5 psig and stable.

Which one of the following describes the Reactor water level control strategy, in accordance with EOP-3?

A. Re-commence injection with Feedwater now and maintain Reactor water level between -19 inches and 110 inches.

B. Re-commence injection with HPCI now and maintain Reactor water level between -19 inches and 11 0 inches.

C. Continue to lower Reactor water level until Reactor power drops below 2.5%, Reactor water level lowers to 0 inches, or all SRVs remain closed. Then re-commence injection with Feedwater.

D. Continue to lower Reactor water level until Reactor power drops below 2.5%, Reactor water level lowers to 0 inches, or all SRVs remain closed. Then re-commence injection with HPCI.

Proposed Answer: D

Explanation: EOP-3 contains the following steps:

IF

Rc.IA.:tor pLmtot•r i~ ~Lhd\"\.~ 1. i'' , or t:.ann,n ~ dctt.·rmincd

AND RIJ'\' w.ater 14..'\"d i~ ,;m..n:e IIU in.

Rc;u:tor pm.n:r i~ J~n"C !. i".,._ nr t:.mnot be

dctcmtin<:d

AND

AND

l'orus tt.•RJpt•r.Jtun: itt .ahtn-c the Huron lniet:don lnitiJtion 1't'11lpcrdlurt•

AND

An SR\' j:'i ••pen ~Jr d11-.·t.·11 pn .. ·~,.un~ i10 ;Lht.)\'oC

!.- p•ill

ntEN

Lower RP\' wat-.·r len.i to bt.~k•"· I I <I in. b)· tcrmiiLitio'p; and pn:wntinl! J.U iL1i<'ll:ti•Mt CX~CJll ~~L<:. R<:l<: and {:Rn <tl'·iJ.

lrn:spc:t:th:c oflrt~w..·tl)f p.m·cr eN' RPV water It\"\:' .I ~~~·JlLati.uns. JoY~o'l!r RP\' \\"att-r lc'\·c.ll')' lcnninJtinfl .tnd pn.1o·cntit1p; .1U i11lit:"'-"ti(U\ cx'-·cpt su:. RCIC Jnd (:RU !EI'·'iJ Lllllil either:

:- Rc.a..:tor puwcr tJirdp~ bchm: .!. ;,•1,,_

OR

C)R

:- All SR\'s f<t'll1.1in -t.:lo~.'d Jnd llif1""-'Cll

pn."!'i~Urt• n:m.dn"" llc! low .! . - J"tSiJ&.

The first step is satisfied with Reactor water level at 90 inches and would allow re-injection. However, since Reactor power is above 2.5%, Reactor water level is above 0 inches, Torus temperature is above the BIIT (-115°F at 10% power), and SRVs are open, the second step requires further lowering of Reactor water level. Once injection is allowed, Feedwater is unavailable due to the complete loss of instrument air, which causes outboard MSIVs to close and loss of motive steam to the Feedwater pumps. At 950 psig with RCIC out of service and Feedwater unavailable, only HPCI is available for injection. HPCI is unaffected by the loss of instrument air.

A. Incorrect- The given conditions require Reactor water level to be lowered further. Feedwater is unavailable due to the prolonged, complete: loss of instrument air pressure.

B. Incorrect- The given conditions require Reactor water level to be lowered further. C. Incorrect- Feedwater is unavailable due to the prolonged, complete loss of instrument air

pressure.

Technical Reference(s): EOP-3, EOP-11, AOP-12

Proposed references to be provided to applicants during examination: BIIT Curve



Question History:



Comments: TRH 3/8/14- Deleted BIIT curve from question and added as provided reference, based on NRC comment.



Main Turbine Generator Trip

------------------------Level Tier# Group# KIA# Importance Rating

SRO 1 1 295005 AA2. 06 2.7

Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Feedwater temperature


The plant is operating at 20% power during a startup with the following:

• Ten minutes ago, the Main Generator was synchroniz.ed to the grid and loaded to 60 MWe per OP-11A, Main Generator, Transformers, and Isolated Bus Phase Cooling.

• No subsequent steps in the startup have been performed yet. • Then, a Main Turbine trip occurs. • The Reactor does NOT scram. • Reactor power begins to slowly rise.

Which one of the following describes whether this is the expected Reactor response to the Main Turbine trip and a required action to be directed, in accordance with the AOPs and EOPs?

This is ...

A. the expected Reactor response. Enter AOP-2, Main Turbine Trip Without Scram, and direct a manual Reactor scram.

B. the expected Reactor response. Enter AOP-2, Main Turbine Trip Without Scram, and direct a Reactor power reduction. A manual Reacto1r- scram is NOT required.

C. NOT the expected Reactor response. Enter AOP-32, Unplanned Power Change, to determine the cause of the power rise. A manual Reactor scram is NOT required.

D. NOT the expected Reactor response. Enter EOP-2, RPV Control, and direct a manual Reactor scram. If the Reactor still does NOT scram, transition to EOP-3, Failure to Scram.

Proposed Answer: 8

Explanation: This is the proper Reactor response. The Reactor does NOT scram on the Main Turbine trip because Reactor power is below 29%. Reactor power rises due to the loss of Feedwater heating on the Main Turbine trip. AOP-2 entry is required. The subsequent actions of AOP-2 require directing a Reactor power reduction due to the expected rise in power due to loss of Feedwater heating.

A. Incorrect- While AOP-2 does require responding to the l~eactor power rise that results from the loss of Feedwater heating, it requires a power reducti1on, but NOT a Reactor scram. A Reactor scram would only become required if appropriate action was not taken and Reactor power reached 29%.

C. Incorrect- This is the expected response. AOP-32 entry is required, although the cause of the power rise is already known to be loss of Feedwater heating.

D. Incorrect- This is the expected response. EOP-2 entry would only be required if the Reactor was supposed to scram.

Technical Reference(s): AOP-2, AOP-32, EOP-:2

Proposed references to be provided to applicants during Elxamination: None

Learning Objective: LP-AOP 1.02


Question History:



Comments: TRH 3/10/14- Revised question to better meet SRO only guidelines, based on NRC comment.


Examination Outline Cross-Reference: SRO 1 1

Form ES-401-5

Level Tier# Group# KIA# Importance Ratin~J

295021 2.4.41 4.6

Loss of Shutdown Cooling

Emergency Procedures I Plan: Knowledge of the emergency action level thresholds and classifications.


A plant shutdown and cooldown is in progress with the following:

• RHR pump 8 is running in a Shutdown Cooling lineup. • Reactor water temperature is 21 0°F and slowly lowering. • Reactor pressure is 0 psig. • Then, Shutdown Cooling isolates due to an electrical fault (Time= 0 minutes). • At Time = 40 minutes, Shutdown Cooling (SOC) is restored. • The following conditions occurred during this time period:

Reactor Water R ea ctor Pressure Time (minutes) Temperature (°F) (psig)

0 210

(SOC isolates) 0

5 213 0.3 10 216 1.2 15 220 2.5 20 224 3.8 25 228 5.3 30 232 6.9 35 236 8.5 40

240 (SOC restored) 10.3

45 236 8.5

Which one of the following describes the highest required emergency action level (EAL) that is met during this time period, if any?

A. No EAL is met.

B. An Unusual Event EAL is met.

C. An Alert EAL is met.

D. A Site Area Emergency EAL is met.

Proposed Answer: c

Explanation: Alert CA3.1 is met at time 40 minutes due to "unplanned RPV pressure increase > 10 psig due to loss of RCS cooling". The Cold EAL chart must be used because Reactor water temperature was < 212°F at the beginning of the transient. The Cold EAL chart must continue to be used even once Reactor water temperature rises above 212°F.

A. Incorrect- Both UE and Alert EALs are met using the Cold EAL chart. However, if the candidate uses the wrong EAL chart, no EALs are met on the Hot EAL chart.

B. Incorrect- UE CU3.1 is met due to "unplanned event results in RCS temperature > 212°F due to loss of decay heat removal capability". However, this is not the highest EAL met due to Alert CA3.1.

D. Incorrect- No Site Area Emergency EAL is met.

Technical Reference(s): IAP-2

Proposed references to be provided to applicants during examination:

Learning Objective:


Question History:



Comments:

Hot and Cold EAL charts (with SU5. 1 and Table F-1 Row D blocked out)



Form ES-401-5


295023 2.4.18 4.0

Refueling Accidents

Emergency Procedures I Plan: Knowledge of the specific bases for EOPs.


The plant is in a refueling outage with the following:

• Reactor coolant temperature is 85°F and stable. • Core shuffle is in progress. • A new fuel assembly has been dropped in the Spent Fuel Pool. • The Spent Fuel Pool area radiation monitor, 18RIA-OS1-12, has exceeded the Maximum

Normal reading but is below the Maximum Safe reading.

Which one of the following describes the entry requirement for EOP-5, Secondary Containment Control, and AOP-44, Dropped Fuel Assembly?

A. Both EOP-5 and AOP-44 entry is required.

B. EOP-5 entry is NOT required due to plant mode. AOP-44 entry is required.

C. EOP-5 entry is required. AOP-44 entry is NOT required since the dropped fuel assembly was NOT previously irradiated.

D. EOP-5 entry is NOT required due to plant mode. AOP-44 entry is NOT required since the dropped fuel assembly was NOT previously irradiated.

Proposed Answer: B

Explanation: The given area radiation monitor condition meets an EOP-5 entry condition. However, EOP bases specifically do NOT require the EOPs to be entered if Reactor coolant temperature is less than 212°F and a Reactor startup or shutdown is NOT in progress. Since the plant is in Mode 5, Reactor coolant temperature is less than 212°F and a Reactor startup or shutdown is NOT in progress. Therefore EOP-5 entry is NOT required. AOP-44 entry is required, as the procedure entry conditions do NOT differentiate between new and irradiated fuel bundles.

A. Incorrect- EOP-5 entry is NOT required even though an entry condition is met since EOP bases do NOT require entry given the current plant mode.

C. Incorrect- EOP-5 entry is NOT required even though an entry condition is met since EOP bases do NOT require entry given the current plant mode. AOP-44 entry is required, as the procedure entry conditions do NOT differentiate between new and irradiated fuel bundles.

D. Incorrect- AOP-44 entry is required, as the procedure entry conditions do NOT differentiate between new and irradiated fuel bundles.

Technical Reference(s): EOP-5, MIT-301.11 F, EP-1, AOP-44


Learning Objective: MIT-301.11 F 1.04, LP-AOP 1.09


Question History:



Comments: TRH 3/10/14 - Revised question to more fully meet SRO only guidelines, based on NRC comment. TRH 3/17/14- Revised first bullet to not directly state Mode 5, based on NRC comment.



Form ES-401-5


295038 2.2.37 4.6

High Off-site Release Rate

Equipment Control: Ability to determine operability and lor availability of safety related equipment.



• Waste Sample Tank A is being discharged to the canal. • Then, 17RM-350, Radwaste Effluent Monitor, fails downscale. • The discharge is temporarily suspended to allow investigation.

Which one of the following describes the ability to continue the discharge?

A. The discharge must be stopped until 17RM-350 is fixed.

B. The discharge may continue if the appropriate grab samples are collected and analyzed at least once per 12 hours.

C. The discharge may continue with no further actions since the minimum required number of operable radiation monitor channels is still met.

D. The discharge may continue if two independent samples are analyzed and two technically qualified individuals verify the discharge valve lineup.

Proposed Answer: D

Explanation: 17RM-350 is the only radiation monitor installed to monitor discharges from Radwaste to the canal. With this monitor failed downscale, the Minimum Channels Operable of ODCM Table 3.1-1 is NOT met. The associated Action (a) st:ates, "With the number of operable channels less than the required minimum number, effluent releases may continue provided that prior to initiating a release:

1. Two independent samples are analyzed; 2. Two technically qualified members of the facility staff verify the discharge line valving;

Otherwise, suspend release of radioactive effluents via this pathway."

A. Incorrect- ODCM Table 2.1-1 Action (a) provides alternate requirements that will allow continued discharge.

B. Incorrect- This is the requirement for Service water system effluent line with less than the required Minimum Channels Operable in ODCM Table 2.'1-1. The canal discharge relies on a separate pathway and radiation monitor.

C. Incorrect- 17RM-350 is the only radiation monitor installe~d to monitor discharges from Radwaste to the canal.

Technical Reference(s): ODCM 2.1, OP-49, FM-170

Proposed references to be provided to applicants during examination: ODCM 2.1



Question History:



Comments:



Reactor Low Water Level


SRO 1 1 295031 2.4.35 4.0

Emergency Procedures I Plan: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.



• A fire in the Control Room requires execution of AOP-43, Plant Shutdown From Outside the Control Room.

• Operators have completed all required field actions olf AOP-43. • Then, a small loss of coolant accident develops. • Reactor water level is lowering. • All ability to monitor Reactor water level at 25RSP is lost.

Which one of the following describes the ability of RCIC to automatically inject to the Reactor and required direction in response to the loss of Reactor wat,er level indication?

A. RCIC will automatically inject. Direct flooding of the Reactor per EOP-7, RPV Flooding.

B. RCIC will NOT automatically inject. Direct flooding of the Reactor per EOP-7, RPV Flooding.

C. RCIC will automatically inject. Direct monitoring of R1:!actor water level from Rack 25-51 per AOP-43.

D. RCIC will NOT automatically inject. Direct monitorin~1 of Reactor water level from Rack 25-51 per AOP-43.

Proposed Answer: D

Explanation: AOP-43 actions result in closing of the RCIC trip throttle valve, which prevents RCIC from automatically injecting to the Reactor given low water level. AOP-43 requires local monitoring of Reactor water level from Rack 25-51 if level i1ndication is unavailable from 25RSP. With Reactor water level indication still available from Rack 25-51, EOP-7 entry is not required.

Note: The question meets SRO only question guidance by requiring in-depth knowledge of the strategy and actions taken in AOP-43 and assessing how these procedure actions affect subsequent plant operation.

A. Incorrect- RCIC will not automatically inject because operator action has been taken per AOP-43 to close the trip throttle valve. Entry into EOP-7 is not required because alternate Reactor water level indication is available from Rack 25-51.

B. Incorrect- Entry into EOP-7 is not required because alternate Reactor water level indication is available from Rack 25-51.

C. Incorrect- RCIC will not automatically inject because operator action has been taken per AOP-43 to close the trip throttle valve.



Learning Objective: LP-AOP 1.08


Question History:



Comments: TRH 3/8/14- Added note regarding SRO only level of question, based on NRC comment.



Inadvertent Containment Isolation

Level Tier# Group# KIA# Importance Ratin!~

SRO 1 2 295020 AA2.05 3.6

Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor water level


The plant is shutdown for a refueling outage with the followin£1:

• RHR loop B is in a Shutdown Cooling Lineup. • RWR pump A is running at 30% speed. • RWR pump B is shutdown. • Reactor water level is being controlled in a band of 200-220 inches. • Then, an error during I&C testing results in an inadvertent PCIS Group 2 isolation. • The isolation CANNOT be immediately restored.

Which one of the following describes the required direction to be given for control of Reactor water level and the validity of Reactor coolant temperature indications, in accordance with AOP-30, Loss of Shutdown Cooling?

Reactor water level. ..

A. may continue to be controlled at the current level. Reactor coolant temperature indications remain valid.

B. may continue to be controlled at the current level. The validity of Reactor coolant temperature indications is NOT assured.

C. control must be raised to a higher band. Reactor coolant temperature indications remain valid.

D. control must be raised to a higher band. The validity of Reactor coolant temperature indications is NOT assured.

Proposed Answer: A

Explanation: The PCIS Group 2 isolation causes a SOC isolation. RWR pump A is unaffected and remains in service providing forced core circulation. With no SDC in service, but with an RWR pump still running, AOP-30 Attachment 2 requires Reactor water level controlled in a band of 200-270 inches. This allows Reactor water level to continue to be controlled in the current band. AOP-30 Attachment 3 cautions that Reactor coolant temperature indication validity is threatened if all RWR and SOC pumps are off with Reactor water level less than 234.5 inches. Since RWR pump A is still in service, Reactor coolant temperature indications remain valid, even with Reactor water level in the band of 200-220 inches.

B. Incorrect- With RWR pump A in service, Reactor coolant temperature indications remain valid.

C. Incorrect- With RWR pump A in service, Reactor water level may remain controlled in the 200-220 inch band.

D. Incorrect- With RWR pump A in service, Reactor water level may remain controlled in the 200-220 inch band and Reactor coolant temperature indiGations remain valid.



Learning Objective: LP-AOP 1.03.a


Question History:



Comments:



Loss of CRD Pumps


SRO 1 2 295022 2.4.45 4.3

Emergency Procedures I Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.



• Following a trip of CRD pump A, CRD pump 8 is started per AOP-69, Control Rod Drive Pump Trouble.

• Annunciator 09-5-1-43, CRD ACCUM PRESS LO OR LVL HI, remains in alarm. • The yellow ACCUM lights are lit for control rods 02-2·?, 10-31, and 30-39. • An operator in the field reports the following CRD HCU accumulator pressures:

o Control rod 02-27: 900 psig o Control rod 1 0-31 : 960 psig o Control rod 30-39: 825 psig

• Each of these control rods is at position 48. • The scram times for these control rods during the last surveillance were:

Control Rod 02-27 10-31 30-39

Scram Times when Reactor Stea m[ otc Notch Position 46 Notch Position 36 N

0.32 1.05 0.41 1.10 0.29 0.80

>ome Pressure ~ 800 l?~ig) :h Position 26 Notch Position 06

1.79 3.40 1.92 3.32 1.50 2.72

• All other control rods have been verified to meet surveillance scram times.

Which one of the following describes the current operability of these control rod scram accumulators, and the required action if these low accumulator pressures CANNOT be restored, in accordance with Technical Specifications?

A. All three control rod scram accumulators are currently inoperable. All three control rods must be declared either "slow" or inoperable. None of the three control rods is required to be inserted.

B. All three control rod scram accumulators are currently inoperable. Control rod 30-39 must be declared either "slow" or inoperable. Control rods 02-27 and 10-31 must be declared inoperable, fully inserted within 3 hours, ancl disarmed within 4 hours.

C. Control rod scram accumulators 02-27 and 30-39 are currently inoperable. Control rod scram accumulator 10-31 is currently operable. Control rods 02-27 and 30-39 must be declared either "slow" or inoperable. Neither of these· two control rods is required to be inserted.

D. Control rod scram accumulators 02-27 and 30-39 are currently inoperable. Control rod scram accumulator 10-31 is currently operable. Control rod 30-39 must be declared either "slow" or inoperable. Control rod 02-27 must be declared inoperable, fully inserted within 3 hours, and disarmed within 4 hours.

Proposed Answer: D

Explanation: Annunciator 09-5-1-43 is likely to be received while in AOP-69, but should normally clear once a CRD pump is restarted. The annunciator alone does not conclusively determine control rod operability. Further analysis of accumulator pressure and Technical Specifications is required. Control rod scram accumulators are inoperable if their pressure is < 940 psig. Therefore, control rod scram accumulators 02-27 and 30-39 are inoperable, while control rod scram accumulator 10-31 is operable. The scram times for control rod 02-27 do NOT meet the requirements of TS table 3.1.4-1, while the scram times for control rod 30-39 do meet these requirements. Therefore, TS 3.1.5 allows declaring control rod 30-39 either "slow" or inoperable. Declaring this control rod "slow" avoids the need to insert the control rod. Since control rod 02-27 had slow scram times, TS 3.1.5 requires de!Ciaring the control rod inoperable. TS 3.1.3 then requires inserting the control rod within 3 hours, and disarming the rod within 4 hours.

A. Incorrect- Control rod scram accumulator 10-31 is operable because its pressure is above 940 psig.

B. Incorrect- Control rod scram accumulator 10-31 is operable because its pressure is above 940 psig.

C. Incorrect- Control rod 02-27 does NOT meet the requirement sof TS table 3.1.4-1, therefore it must be declared inoperable, inserted within 3 hours, and disarmed with 4 hours.

Technical Reference(s): AOP-69, ARP 09-5-1-4~~, Technical Specifications 3.1.3, 3.1.4, and 3.1.5

Proposed references to be provided to applicants during examination: Technical Specifications 3.1.3, 3.1.4, and 3.1.5

Learning Objective: SDLP-03C 1.16 and 1.18


Question History:



Comments:



295015 Incomplete SCRAM


SRO 1 2 295015 AA2.02 4.2

AA2.02 - Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM: Control rod position



• All control rod position indication is lost. • Then, a Main Turbine trip occurs. • APRMs indicate downscale. • Reactor water level lowers to 150 inches and then rises with Feedwater injecting.

Which one of the following describes the proper Emergency Operating Procedure (EOP) execution for this transient?

Enter EOP-2, RPV Control, and ...

A. remain in EOP-2 until all EOP-2 entry conditions clear.

B. then exit to EOP-3, Failure to Scram. EOP-3 may NOT be exited until control rod positions are known.

C. then exit to EOP-3, Failure to Scram. EOP-3 may be exited once all IRMs are fully inserted and indicate less than range 7.

D. then exit to EOP-3, Failure to Scram. EOP-3 may be exited if Standby Liquid Control is used to inject the Cold Shutdown Boron Weight.

Proposed Answer: B

Explanation: EOP-2 entry is required because Reactor water level lowered below 177". With no control rod position information both of the following steps must be answered "No/Unknown":

No/Unknown ~hts pmcedure and enter ~:)·:\."Failure to Scram.'"

Therefore EOP-2 must be exited and EOP-3 must be entered. EOP-3 is only exited once one of the following conditions is satisfied:

IF

All t:untoH nH.k .IR' in~rtt:tl tu Uf" oc,·und

l"'»iiCiounOl

OR It b.L"' been do&:~t.•rntincd tl\4.' rc.u.·tur "'-'ill "m~in ~tuldtmm under .Ill ...-.. mdil~un~ wLtmn11 burun

N

THEN

L IF .......... hnrun ls llt."inJt inft'-.1c.-d lu

~hw duwn dw R:'A"tor.

THEN ... tt'mtln:tU: htwHn lnit.'1.'lkm.

!.. Exil thk pluc:t"ttun• .tnd cntc.·r £t •P-l. *RP\' <:untmt"

Neither of these two determinations can be made until control rod positions are known. Therefore EOP-3 cannot be exited until control rod positions are known.

A. Incorrect- With control rod position unknown, EOP-2 must be exited and EOP-3 must be entered. APRMs being downscale is not enough to avoid EOP-3 entry.

C. Incorrect- EOP-3 cannot be exited until control rod positions are known. IRMs less than range 7 only allows RPV cooldown while still in EOP-3.

D. Incorrect- EOP-3 cannot be exited until control rod positions are known. Injection of Cold Shutdown Boron Weight only allows RPV cooldown and restoration of normal water level while still in EOP-3.

Technical Reference(s): EOP-2, EOP-3, MIT-30"1.11 C, MIT-301.11 D


Learning Objective: MIT-301.11C 1.03 and 1.07, MIT-301.110 1.02 and 1.03


Question History:



Comments:



PCIS/Nuclear Steam Supply Shutoff


SRO 2 1 223002 A2.11 3.9

Form ES-401-5

Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Standby liquid initiation



• I&C and Engineering have determined that a wiring problem will prevent the input from the Standby Liquid Control (SLC) pump control switch to the Reactor Water Cleanup Isolation circuitry from re-positioning upon SLC pump start.

• I&C and Engineering have determined that all other SLC pump control switch wiring is connected properly.

Which one of the following describes the immediate impac:t of this wiring problem on Technical Specifications (TS)?

A. TS 3.1. 7 Conditions A and B must be entered.

B. TS 3.3.6.1 Conditions A and B must be entered.

C. TS 3.1.7 Condition A must be entered. TS Condition B does NOT need to be entered.

D. TS 3.3.6.1 Condition A must be entered. TS Conditic1n B does NOT need to be entered.

Proposed Answer: B

Explanation: With the input from the Standby Liquid Control (SLC) pump control switch to the Reactor Water Cleanup Isolation circuitry not re-positioning upon SLC pump start, TS table 3.3.6.1-1 Function 5.d is NOT met and this isolation function is NOT maintained, requiring TS 3.3.6.1 Conditions A and B to be entered.

A. Incorrect- TS 3.1.7 entry is not required unless the required completion time of TS 3.3.6.1 Condition A orB is not met, in which case TS 3.3.6.1 Condition I may require declaring a SLC subsystem inoperable.

C. Incorrect- TS 3.1.7 entry is not required unless the required completion time of TS 3.3.6.1 Condition A or B is not met, in which case TS 3.3.6.1 Condition I may require declaring a SLC subsystem inoperable.

D. Incorrect- Since all input from the SLC switch to the RWCU isolation logic is not working, TS table 3.3.6.1-1 Function 5.d is not capable of occurrin~). This results in the need to enter TS 3.3.6.1 Condition B.

Technical Reference(s): Technical Specifications 3.1.7 and 3.3.6.1, ESK-6RE, 1.70-110


Learning Objective: SDLP-111.18


Question History:



Comments: TRH 3/8/14- Added "immediate" to question, based on NRC comment.

Technical Specifications 3.1. 7 and 3.3.6.1 (allowable value column blocked out)



Form ES-401-5

Level Tier# Group# KIA# Importance Ratin~)

239002 A2.02 3.2

SRVs

Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, ccmtrol, or mitigate the consequences of those abnormal conditions or operations: Leaky SRV



• Annunciator 09-4-1-16, SRV LEAKING, alarms. • SRV B tailpipe temperature is 275°F and rising approximately 1 °F/hr. • All other SRV tailpipe temperatures indicate approximately 160°F and stable. • Main Generator output is unchanged. • Reactor pressure is unchanged. • Torus water temperature is 95°F and rising approximately 0.25°F/hr.

Based on the current indications, which one of the following describes the required response?

A. Direct a Reactor scram per EOP-4, Primary Containment Control.

B. Direct a power reduction to 85% per AOP-36, Stuck Open Relief Valve(s).

C. Direct commencing a Reactor shutdown per OP-65, Startup and Shutdown Procedure.

D. Direct raising the SRV B tailpipe temperature alarm setpoint per OP-1, Main Steam System.

Proposed Answer: D

Explanation: With Annunciator 09-4-1-16 in alarm, Torus water temperature slowly rising, but Main Generator output and Reactor pressure unchanged, SFtV B is leaking, but NOT open. ARP 09-4-1-16 directs raising the SRV B tailpipe temperature alarm setpoint to 345°F per OP-1 Section G.9 while other actions are taken to assess need for repairs to the SRV.

A. Incorrect- EOP-4 entry is required based on Torus water temperature, however the procedure does not require a Reactor scram until Torus water temperature approaches the Boron Injection Initiation Temperature (11 0°F at this Reactor power). Based on the current trend in Torus water temperature, reaching 11 0°F is not a concern for many hours.

B. Incorrect- While some symptoms listed in AOP-36 are present, AOP-36 does NOT require a power reduction.

C. Incorrect- While a normal plant shutdown may be performed in order to fix a leaking SRV, there is no procedural requirement to execute OP-65 based on the given conditions.

Technical Reference(s): ARP 09-4-1-16, OP-1, EOP-4, AOP-36


Learning Objective: SDLP-02J 1.14.d


Question History:






ADS

Level Tier# Group# KIA# Importance Ratin!g

SRO 2 1 218000 2.4.3 3.9

Form ES-401-5

Emergency Procedures I Plan: Ability to identify post-accident instrumentation.


The plant is operating at 100% power when power is lost to the SRV acoustic valve monitoring system (VMS).

Which one of the following describes a required action, if any, per Technical Specifications and/or the Technical Requirements Manual?

A. No actions are required by Technical Specifications or the Technical Requirements Manual.

B. Restore acoustic monitors to operable status within a maximum of 7 days.

C. Restore acoustic monitors to operable within a maximum of 12 hours or be in Mode 3 within a maximum of 12 hours.

D. Verify SRV thermocouples are operable immediately i3nd if any are inoperable, restore either the thermocouple or acoustic monitor for the affected SRV(s) within a maximum of 48 hours.

Proposed Answer: D

Explanation: TRM 3.3.C, Post-Accident Monitoring (PAM) Instrumentation, Table T3.3.C-1 requires SRV acoustic monitors to be operable. With loss of power to the entire SRV acoustic valve monitoring system, both channels of acoustic monitors are inoperable for all SRVs. Therefore TRM 3.3.C Condition F applies, which requires verifying the associated thermocouples are operable immediately. If any thermocouplle is inoperable, then TRM 3.3.C Condition G applies for the associated SRV and requires restoring at least one acoustic monitor or the thermocouple to operable status within 48 hours.

A. Incorrect- TRM 3.3.C Condition F applies and requires actions. This is plausible if the candidate is unable to identify SRV acoustic monitors as PAM instrumention.

B. Incorrect- This is the required action for two channels of inoperable PAM Instrumentation in TS 3.3.3.1, however SRV acoustic monitors are covered by the TRM 3.3.C, not TS 3.3.3.1.

C. Incorrect- This is based on the required action for two or more ADS valves inoperable per TS 3.5.1 Condition G. However, inoperable acoustic monitors are not cascaded to make the associated ADS valves inoperable.

Technical Reference(s): TS 3.3.3.1, TRM 3.3.C, TS 3.4.3, TS 3.5.1

Proposed references to be provided to applicants during examination: TS 3.3.3.1, TRM 3.3.C, TS 3.4.3, TS 3.5.1

Learning Objective: SDLP-02J 1.18


Question History:



Comments: TRH 3/8/14- Reworded choice C to connect time requirement with first half of sentence, based on NRC comment.



HPCI


SRO 2 1 206000 2.4.4 4.7

Form ES-401-5

Emergency Procedures I Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.



• Reactor water level is -5 inches and stable. • HPCI is the only available injection system and flow is maximized. • A Reactor Operator reports that HPCI Room temperature indications are between 115°F

and 125°F and rising. • An Operator in the field reports a steam leak in the HPCI Room. • Given the following excerpt from EOP-5, Secondary Containment Control, Reactor

Building Area Temperatures table:

AREA INSTRUMENT MAXIMUM MAXIMUM NORMAL SAFE

HPCI Room 23RTD-94A 23-294A. Panel (J<J-9"; 23RTD-948 23-2948. Panel 09·96 l().i"f 137"f 23RTIJ..ll7 A 23-2 l7 A. Panel 09-9"; 23RTD-1178 23-2 I 78, Panel 09-96

Which one of the following describes the requirements of EOP-5, Secondary Containment Control, based on the HPCI Room temperature indications?

EOP-5 entry is ...

A. NOT yet required because HPCI is in operation, but will be if HPCI Room temperatures rise above the Maximum Safe value.

B. required. HPCI must be isolated.

C. required. HPCI does NOT need to be isolated because it is required to operate for another EOP.

D. required. HPCI does NOT need to be isolated unless HPCI Room temperatures rise above the Maximum Safe values.

Proposed Answer: c

Explanation: The given HPCI Room temperatures are above the Max Normal value of 1 04°F but below the Max Safe value of 137°F. This meets an EOP·-5 entry condition. It would also normally require isolation of HPCI, however isolation is not re~quired if the system is needed for other EOP actions. In these circumstances, adequate core cooling is challenged and only HPCI is available to continue injection to the Reactor. Therefore, HPCI should NOT be isolated until another injection system can be made available.

A. Incorrect- EOP-5 entry is required because the given HPCI Room temperature indications are above the Max Normal value of 1 04°F. Even with HPCI in operation, the Max Normal values, not the Max Safe values, are used to determine EOP-5 entry.

B. Incorrect- HPCI would normally be required to be isolateld, but not under these circumstances because it is required to maintain adequate core cooling.

D. Incorrect- HPCI Room temperatures would already normally require HPCI isolation even though they are below the Max Safe value.

Technical Reference(s): EOP-5, EOP-2

Proposed references to be provided to applicants during e~xamination: None

Learning Objective: MIT-301.11 F 1.04 and 1.07


Question History:



Comments:



IRM


SRO 2 1 215003 2.1.31 4.3

Conduct of Operations: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.



• Control rod withdrawals are in progress. • Annunciator 09-5-2-32, IRM DOWNSCALE, is received. • IRM indications are as shown on the next two pages.

Which one of the following describes the impact of these conditions on Technical Specification (TS) 3.3.1.1, Reactor Protection System (RPS) Instrumentation?

A. All TS requirements for IRMs are still met.

B. TS 3.3.1.1 Condition A must be entered. TS 3.3.1.1 Conditions B and C do NOT need to be entered.

C. TS 3.3.1.1 Conditions A and B must be entered. TS ~3.3.1.1 Condition C does NOT need to be entered.

D. TS 3.3.1.1 Conditions A, B, and C must be entered.

IRM BYP

flANGE SWITCH

----- ~ -- - -

IRH A !RM C

IRM BYP

Proposed Answer: B

Explanation: The given indications show IRMs A, C, and D failed downscale. This means these IRMs are inoperable for their upscale scram function. TS Tablle 3.3.1.1-1 requires 3 operable channels per trip system for the IRM upscale scram function. Trip system A only has IRMs E and G operable, therefore it has one required channel inoperable. Trip system B has IRMs B, F, and H operable, therefore it has all three required operable channels. Therefore, TS 3.3.1.1 Condition A must be entered, but Conditions Band C do NOT need to be entered.

A. Incorrect- Trip system A has one of the three required channels inoperable, therefore TS 3.3.1.1 is NOT fully met.

C. Incorrect- Trip system B has all three required operable channels, therefore TS 3.3.1.1 Condition B does NOT need to be entered.

D. Incorrect- Trip system A still has RPS trip capability on IRM upscale (1 out of 4 taken twice), therefore TS 3.3.1.1 Condition C does NOT need to be entered.

Technical Reference(s): TS 3.3.1.1, SDLP-05




Question History:



Comments:

TS 3.3.1.1 (allowable value column blocked out)



Secondary Containment


SRO 2 2 290001 A2.04 3.7

Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High airborne radiation



• Irradiated fuel is being moved in the Spent Fuel Pool in preparation for loading new fuel for an upcoming outage.

• An Operator on the Refueling Bridge reports an irradiated fuel assembly has been dropped.

Which one of the following describes the required actions, in accordance with AOP-44, Dropped Fuel Assembly?

A. Evacuate personnel from the entire Reactor Building. Isolate Control Room and Relay Room ventilation.

B. Evacuate personnel from the entire Reactor Building. Maintain Control Room and Relay Room ventilation in the normal alignment.

C. Evacuate personnel from the Refuel Floor, only. Isolate Control Room and Relay Room ventilation.

D. Evacuate personnel from the Refuel Floor, only. Maintain Control Room and Relay Room ventilation in the normal alignment.

Proposed Answer: A

Explanation: Drop of an irradiated fuel assembly is an analyzed accident expected to release fission product gases from the fuel into the Secondary Containment and cause high airborne radiation/contamination levels. AOP-44 is designed to mitiga1te this release of radiation/contamination. AOP-44 directs evacuating all personnel from the entire Reactor Building, not just the Refuel Floor. AOP-44 also directs isolating Control Room and Relay Room ventilation to protect Control Room habitability.

B. Incorrect- Control Room and Relay Room ventilation are isolated to protect Control Room habitability.

C. Incorrect- AOP-44 directs evacuating all personnel from the entire Reactor Building, not just the Refuel Floor.

D. Incorrect- AOP-44 directs evacuating all personnel from the entire Reactor Building, not just the Refuel Floor. Control Room and Relay Room ventilation are isolated to protect Control Room habitability.



Learning Objective: LP-AOP

Question Source: Modified Bank- March 2012 NRC #72




Comments:



Recirculation


SRO 2 2 202001 2.4.6 4.7

Emergency Procedures I Plan: Knowledge of EOP mitigation strategies.


The plant has experienced a failure to scram with the following:

• The Reactor Mode Switch is in SHUTDOWN. • ARI has been initiated. • Reactor power is approximately 50% and stable. • Reactor water level is 197 inches and stable with Feedwater injecting.

Form ES-401-5

• Reactor pressure is 1025 psig and stable with the Main Generator still online.

Which one of the following describes the required strategy for controlling Reactor Recirculation flow, in accordance with EOP-3, Failure to Scram?

A. Maintain Recirculation flow at the current value as lon!~ as Reactor parameters remain stable.

B. Ensure Recirculation flow is runback to minimum. Then, if Reactor power is above 2.5%, ensure the Recirculation pumps are tripped.

C. Ensure Recirculation flow is runback to minimum. Th1~n. maintain Recirculation pumps operating as long as the Main Generator is still online.

D. Immediately trip the Recirculation pumps.

Proposed Answer: B

Explanation: Since the Main Generator and Feedwater are still operating, Recirculation flow must be run back to minimum before tripping Recirculation pumps. This action lowers Reactor power while attempting to prevent high level trip of the Main Turbine or Feedwater pumps. Once Recirculation flow is run back to minimum, diagnostic steps lead to tripping both Recirculation pumps if Reactor power remains above 2.5%.

Note: The question meets SRO only guidance by requiring the candidate to assess conditions and use detailed knowledge of the steps and logic in the EOP-3 power leg to determine the correct answer.

A. Incorrect- The Reactor power leg of EOP-3 requires ensuring Recirculation flow is runback to minimum in this situation because the Main Generator and Feedwater are still operating.

C. Incorrect- Operating state of the Main Generator is used in the determination to initially runback Recirculation flow. Then, the determination to trip Recirculation pumps is based solely on Reactor power, not operating state of the Main Generator.

D. Incorrect- Since the Main Generator and Feedwater are operating, it is required to ensure Recirculation flow is runback to minimum before tripping Recirculation pumps, to attempt to avoid tripping the Main Generator and Feedwater pumps on high level.

Technical Reference(s): EOP-3




Question History:



Comments: TRH 3/8/14- Added note regarding how the question meets SRO only guidance, based on NRC comment.


Control Rod and Drive Mechanism


SRO 2 2 201003 2.4.8 4.5

Emergency Procedures I Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.


The plant was operating at 100% power with the following:

• AOP-24, Stuck Control Rod, was being performed du1e to a control rod stuck at position 02.

• Then, a Main Turbine trip resulted in a Reactor scram. • A loss of Feedwater resulted in Reactor water level lowering to 150 inches. • The control rod is still stuck at position 02.

Which one of the following describes the procedure use requirements for this situation?

A Enter EOP-2, RPV Control. Continue performing AOP-24. In the event of a conflict between the procedures, EOP-2 is the overriding document.

B. Enter EOP-2, RPV Control. Continue performing AOP-24. In the event of a conflict between the procedures, AOP-24 is the overriding document.

C. Exit AOP-24 and enter EOP-2, RPV Control. AOP-24 is re-entered at the step in progress after exiting EOP-2.

D. Exit AOP-24 and enter EOP-2, RPV Control. AOP--24 entry conditions are re-evaluated after exiting EOP-2.

Proposed Answer: A

Explanation: EOPs and AOPs may be used in conjunction with each other. However, in the event of a conflict between an EOP and an AOP, the EOP is the higher tiered document and takes precedence. Therefore, EOP-2 must be entered due to presence of the low Reactor water level condition, but AOP-24 should continue to be performed until the stuck control rod is resolved.

B. Incorrect- EOPs are higher tiered documents and take precedence in the event of a conflict with a lower tiered document, such as an AOP.

C. Incorrect- The AOP may still be used in conjunction with the EOP and does not need to be exited.

D. Incorrect- The AOP may still be used in conjunction with the EOP and does not need to be exited.

Technical Reference(s): EP-1


Learning Objective:

Question Source: Modified Bank- 2008 NMP1 NI~C #97

Question History: 2008 NMP1 NRC #97



Comments:



Knowledge of the fuel-handling responsibilities of SROs.


SRO 3

2.1.35 3.9

The plant is in Mode 5 conducting refueling operations with the following:

Form ES-401-5

• You are the on-coming Refuel Bridge SRO preparing Ito take the shift and continue fuel movements.

• The time is 1600 on October 2nd. • Reactor water temperature is 65°F. • All control rods are inserted. • OSP-66.001, Management of Refueling Activities, Attachment 9, Refueling Checklist,

was last performed at 0300 on October 151.

Which one of the following describes your ability to continue fuel movements, in accordance with OSP-66.001?

Fuel movements ...

A. may continue with these conditions.

B. must be suspended because the last completion time for Attachment 9 is unsatisfactory. Reactor water temperature is satisfactory to support fuel movements.

C. must be suspended because Reactor water temperature is unsatisfactory. The last completion of Attachment 9 is satisfactory to support fuel movements.

D. must be suspended because both Reactor water temperature and the last completion time for Attachment 9 are unsatisfactory.

Proposed Answer: D

Explanation: OSP-66.001 contains requirements that must be met to allow fuel movements. One requirement is that Attachment 9, Refueling Checklist, be performed once each day. This requirement is NOT met, since Attachment 9 was last completed > 24 hours ago. A second requirement is for Reactor water temperature to be 68°F to 1 :35°F. Since Reactor water temperature is below this range, continued fuel movements are not allowed.

A. Incorrect- Fuel movements must not continue. B. Incorrect- Reactor water temperature is unsatisfactory. C. Incorrect- The last completion time for Attachment 9 is unsatisfactory.

Technical Reference(s): OSP-66.001


Learning Objective: LP-AP 73.03

Question Source: Modified - 2010 NRC #94




Comments:



SRO 3

2.2.23 4.6

Ability to track Technical Specification limiting conditions for operations.

Form ES-401-5



• Today is 5/2/2014. • Reactor power has been approximately 100% for the last 21 days. • Technical Specification Surveillance Requirement (SH) 3.2.1.1 was last completed at

1230 on 5/1/2014. • Problems with 30 Monicore are challenging the ability to verify APLHGR. • See the following for a portion of Technical Specificatlion 3.2.1.

ACTIONS

CONDITION

A. Any APLHGR not within A.l limits.

B. Required Action and 8.1 associated Completion Time not met.

SURVEILLANCE REQUIREMENTS

SURVEILLANCE

REQUIRED ACTION

·) to Restore APLHGR<s within limits.

Reduce THERMAL F to< 25% RTP.

•OWER

SR 3.2.1.1 Verify all APLHGRs are less than or e to the l1m1ts specified in the COLR.

qual

COMPLETION TIME

2 hOurs

4 hours

FREQUENCY

Once within 12 hours after :t 25% RTP

AND

24 hours thereafter

Which one of the following describes the latest time allowed to reduce thermal power less than 25% RTP if APLHGR can NOT be verified?

A. 0930 on 5/2/2014

B. 1230 on 5/2/2014

C. 0030 on 5/3/2014

D. 0630 on 5/3/2014

Proposed Answer: c

Explanation: Since Reactor power has been> 25% RTP for at least 21 days, SR 3.2.1.1 frequency is 24 hours. SR 3.0.2 allows completion of the surveillance to be extended to 30 hours (1.25*24 hours) if necessary without declaring the surveillance not met. Once the surveillance is not met, LCO 3.2.1 Condition A is entered. If the surveillance is still not performed 2 hours later, LCO 3.2.1 Condition B is entered. If the surveillance is still not performed 4 hours later, power must be less than 25% RTP. This would need to be completed by 1830 on 5/2/2014, which is 36 hours after the last performance of the surveillance (1.25*24+2+4=36).

A. Incorrect- This is 21 hours after last performance of the surveillance, which could be arrived at if a 12 hourfrequencywas used (1.25*12+2+4=21).

B. Incorrect- This is 24 hours after last performance of the surveillance, which could be arrived at if only the 24 hour frequency was allowed.

D. Incorrect- This is 42 hours after last performance of the surveillance, which could be arrived at if a 1.5 multiplier was used instead of a 1.25 multiplier for the frequency extension (1.5*24+2+4=42).

Technical Reference(s): TS 1.3, 1.4, SR 3.02, SR 3.0.3 and LCO 3.2.1


Learning Objective: JLP-OPS-ITS01 1.01.f

Question Source: Modified Bank- 2005 Peach Bottom SRO #21

Question History: 2005 Peach Bottom SRO #21



Comments:

Peach Bottom February 2005 Initial Senior Reactor Operator NRC Examination

21. Unit 2 is in MODE 1 at full power.

• An applicable Tech Spec Surveillance with a 24 hour frequency was last perfonned satisfactorily at 1230 on 1114105.

• The Limiting Condition for Operation (LCO) required actions direct that the equipment be restored to OPERABLE status in 4 hours, or be in MODE 3 in 12 hours and MODE 4 in 36 hours.

If a plant priority on Unit 3 prevents the surveillance from be perfonned, track the LCO to determine the EARLIEST time Unit 2 is rcq·llired to be in MODE 4.

A. By 0430 on 1117/2005.

B. By 0630 on 1/1712005.

C. By 1030on 111712005.

D. By 2230 on 1/1712005.



SRO 3

2.3.5 2.9

Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.


An unintentional radioactivity release to the environment is in progress.

Which one of the following describes the types of radiation de~tectors used to determine the need for entry into EOP-6, Radioactivity Release Control, and the types of radiation detectors used to determine the need for an Emergency RPV Depressurization in EOP-6?

A.

B.

C.

D.

Used to Determine Need for Entry

Installed process radiation monitors, only

Portable survey instruments, only

Installed process radiation monitors and portable survey instruments


Used to Determine Need for Emergency RPV Depressurization

Installed process radiation monitors, only


Portable survey instruments, only

Installed process radiation monitors and portatble survey instruments

Proposed Answer: D

Explanation: EOP-6 entry is determined by "Offsite radioactivity release rate above the Emergency Plan "Alert" level". This includes conditions that would classify as a Site Area Emergency or General Emergency. Both installed process radiation monitors and field instruments may be used to classify conditions at one of these levels. The need for Emergency RPV Depressurization is determined by the step "Before offsite radioactivity release rate reaches the offsite release rate which requires a General Emergency". Stack, TB exhaust, and RB exhaust rad monitors can all be used to classify a General Emergency. Additionally, field instruments may be used to survey offsite conditions to classify a General Emergency. Therefore, both installed process radiation monitors and field instruments may be used to determine the need for an Emergency RPV Depressurization.

A. Incorrect- Portable survey instruments may also be used to determine EOP-6 entry and the need for Emergency RPV Depressurization.

B. Incorrect- Installed process rad monitors may also be used to determine EOP-6 entry. C. Incorrect -Installed process rad monitors may also be us~ed to determine the need for

Emergency RPV Depressurization.

Technical Reference(s): EOP-6, IAP-2


Learning Objective: MIT-301.11G 6.03 and 6.05


Question History:


1 0 CFR Part 55 Content: 55.43(4 & 5)

Comments: TRH 3/8/14- Changed both columns of choice A from "portable survey instruments, only" to "installed process radiation monitors, only", based on NRC comment.



Knowledge of the emergency plan.



SRO 3

2.4.29 4.4

Form ES-401-5

Which one of the following describes when initial Protective Action Recommendations (PARs) are made to the County and State?

Initial PARs are required to be made to the County and State within a time limit of ...

A. 15 minutes after declaration of a General Emergency.

B. 30 minutes after declaration of a General Emergency.

C. 15 minutes after the offsite release rate exceeds the Technical Specification limit.

D. 30 minutes after the offsite release rate exceeds the Technical Specification limit.

Proposed Answer: A

Explanation: Initial PARs are required to be made within a maximum of 15 minutes after declaration of a General Emergency.

B. Incorrect- Initial PARs are required to be made within a maximum of 15 minutes after declaration of a General Emergency. 30 minutes is based on the time for update notifications to the County and State.

C. Incorrect- Initial PARs are required to be made within a maximum of 15 minutes after declaration of a General Emergency, not 15 minutes after the offsite release rate exceeds the Technical Specification limit.

D. Incorrect- Initial PARs are required to be made within a maximum of 15 minutes after declaration of a General Emergency, not 15 minutes after the offsite release rate exceeds the Technical Specification limit. 30 minutes is based on the time for update notifications to the County and State.

Technical Reference(s): IAP-1, EAP-1.1


Learning Objective:


Question History: NMP1 2013 NRC #100



Comments: TRH 3/8/14- Added "initial" to stem and question and revised second half of choices C and D, based on NRC comment. TRH 3/17/14- Revised 2nd half of choices C and D, based on NRC comment.



Knowledge of conservative decision making practices.



SRO 3

2.1.39 4.3

Form ES-401-5

• Drywell Floor Drain leakage has risen over the last 24 hours, but remains below all action thresholds in plant technical procedures and Technical Specifications.

• Based on current trending, Drywell Floor Drain leakage will exceed the Technical Specification limit in 48 hours.

Which one of the following processes would be appropriate to initiate in response to this condition?

A. AP-19.01, Surveillance Testing Program, Attachment 2, Surveillance Program Intervals.

B. EN-OP-115, Conduct of Operations, Attachment 9.5, Equipment Status Control Flow Chart.

C. EN-DC-205, Maintenance Rule Monitoring, Attachment 9.1, Functional Failure Determination Form.

D. EN-OP-111, Operation Decision-Making Issue (ODMI) Process, Attachment 9.3, Emergent Issue Checklist.

Proposed Answer: D

Explanation: EN-OP-111 provides a framework for dealing with significant plant issues that do not warrant immediate operations action (such as that required in an AOP, EOP, or Technical Specifications) but present a challenge to plant operation or safety. Attachment 9.3, Emergent Issue Checklist, is provided for Operations Shift Management to communicate such an emergent issue and start the conservative decision making process.

A. Incorrect- AP-19.01 Attachment 2 provides information regarding implementation of surveillance testing frequency, but does not provide any guidance on increasing surveillance testing frequency in response to degraded conditions. Increases monitoring frequency is likely to be initiated in response to this problem through the process in EN-OP-111.

B. Incorrect- EN-OP-115 Attachment 9.5 is used to ensure configuration control is maintained when manipulating plant equipment, but is NOT used to ensure proper decision making given degraded plant equipment.

C. Incorrect- EN-DC-205 Attachment 9.1 is applicable for degraded systems, structures, and components that fall under the Maintenance Rule program, however an elevated Drywell leak rate does not directly relate to a Maintenance Rule system.

Technical Reference(s): EN-OP-111


Learning Objective: LP-AP


Question History:



Comments: TRH 3/8/14- Revised choice A, based on NRC comment.



SRO 3

2.3.15 3.1

Form ES-401-5

Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.


Which one of the following identifies a radiation monitor and associated threshold reading used to define Loss of the Fuel Clad Barrier, in accordance with IAP-2, Classification of Emergency Conditions, Table F-1, Fission Product Barrier Matrix?

A. Drywell radiation; 450 R/hr

B. Drywell radiation; 3000 R/hr

C. Offgas radiation; hi alarm setpoint

D. Offgas radiation; hi-hi alarm setpoint

Proposed Answer: B

Explanation: IAP-2 Table F-1, Fission Product Barrier Matrix, identifies Drywall radiation > 3000 R/hr as a Loss of the Fuel Clad Barrier.

Note: The provided reference for Question #79 is to be edited to avoid making this question a direct lookup.

A. Incorrect- 3000 R/hr is the threshold value, NOT 450 IR/hr. 450 R/hr is the hi-hi alarm setpoint for the Drywall rad monitors.

C. Incorrect- Offgas radiation monitoring is used to define Unusual Event SU5.1. This indicates fuel clad degradation, but does not define loss of the barrier.

D. Incorrect- Offgas radiation monitoring is used to define Unusual Event SU5.1. This indicates fuel clad degradation, but does not define loss of the barrier.

Technical Reference(s): IAP-2


Learning Objective:


Question History:



Comments:



SRO 3

2.2.18 3.9

Form ES-401-5

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

Proposed Question: #1 00

An outage risk assessment has determined that one Key Safety Function risk assessment will be Yellow for a short period of time.

Which one of the following describes the status of this Key Safety Function while its risk assessment is Yellow and the need for a contingency plan during the time period when the risk assessment is Yellow, in accordance with AP-1 0.09, Outage Risk Assessment?

While the risk assessment is Yellow, the minimum Technical Specification requirements for this Key Safety Function are ...

A. met, but a contingency plan is needed.

B. met and NO contingency plan is needed.

C. NOT met and a contingency plan is needed. The highest level of approval for the contingency plan is the Operations Manager.

D. NOT met and a contingency plan is needed. The highest level of approval for the contingency plan is the General Manager Plant Operations.

Proposed Answer: A

Explanation: A Yellow risk assessment for a Key Safety Function means the function is degraded below what is desired, but still meets Technical Spl3Cification requirements. A Yellow risk assessment requires development of a contingency plan.

B. Incorrect- A contingency plan is required. C. Incorrect- Minimum Technical Specification requirements are met. D. Incorrect- Minimum Technical Specification requirements are met. The contingency plan

for Yellow requires Operations Manager approval, but NOT GMPO approval. Contingency plans for Orange require GMPO approval.

Technical Reference(s): AP-10.09


Learning Objective: LP-AP 41.02.e


Question History:



Comments: TRH 3/8/14- Revised C and D to raise plausibility of D, based on NRC comment.

RO 2 Group# 1 KIA# Importance Rating National Grid.A Incorrect-A control rod withdrawal block is active due to SRM A being

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