ML16280A435

2016-09 Final Written Exam

ML16280A435
Person / Time
Site:	River Bend
Issue date:	09/30/2016
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML16280A435 (200)

Text

Examination Outline Cross-Reference Level RO 203000 RHR/LPCI: Injection Mode Knowledge of electrical power supplies to the following: Pumps Tier #

2 Group #

1 K/A #

K2.01 Rating 3.5 Question 1 What is the electrical power supply to RHR Pump C?

A. ENS-SWG3A B. ENS-SWG3B C. ENS-SWG1A D. ENS-SWG1B Answer: D Explanation:

A is wrong because the power supply for the C RHR pump is ENS-SWG1B. Plausible if applicant confuses with recirc pump power supplies.

B is wrong because see A C is wrong because the C RHR pump is powered from the ENS-SWG1B. Plausible if applicant believes the C pump is powered from the Division 1 emergency bus.

D is correct because the correct power supply for RHR Pump C is ENS-SWG1B.

Technical

References:

R-STM-0204.012, Residual Heat Removal System (RHR)

R-STM-0300, Rev 028, AC Distribution References to be provided to applicants during exam: None Learning Objective: RLP-STM-0204, Obj 11.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No

Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(8)

Examination Outline Cross-Reference Level RO 203000 RHR/LPCI: Injection Mode Knowledge of the effect that a loss or malfunction of the RHR/LPCI: injection mode (plant specific) will have on following: K3.04 Adequate core cooling Tier #

2 Group #

1 K/A #

K3.04 Rating 4.6 Revision 1

Question 2 During a LOCA, the BOP gives the following ECCS status update to CRS:

• E21-MOVF005, LPCS Injection Valve, red and green lights are off.

• RHR PUMP A IN MAN OVERRIDE Annunciator lit

• RHR B has an amber light lit for its associated injection valve

• RHR C has a red light and a white light lit for its associated pump control switch

• E22-S002 bus fault No other equipment is available for use and LOCA signals are still present.

After Emergency Depressurization, without operator action, adequate core cooling is assured by _____.

A. RHR A, and RHR B only B. RHR B and RHR C only C. RHR C only D. HPCS and LPCS only Answer: C Explanation:

A is wrong because RHR A pump has a manual override in. The pump would need to be reset by pushing the reset pushbutton and the auto initiation signal would need to be clear.

B is wrong because the amber light indicates the valve has been manually overridden similar to RHR A.

C is correct because RHR C has proper indications for a pump in an ED situation and is ready to refill the core which would lead to core submergence and therefore adequate core cooling.

D is wrong because HPCS bus has a fault making it inoperable. E22-MOVF004 (HPCS INJECTION VALVE) is powered from E22-S002. E21-MOVF005, LPCS Injection Valve, red and green lights are off, indicating the valve has no power to operate.

Technical

References:

R-STM-0203, High Pressure Core Spray, Rev 8 R-STM-0204, Residual Heat Removal System. Rev 12 R-STM-0205, Low Pressure Core Spray, Rev 6 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0204, Obj 6 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(8)

Examination Outline Cross-Reference Level RO 205000 Shutdown Cooling Knowledge of shutdown cooling system (RHR shutdown cooling mode) design feature(s) and/or interlocks which provide for the following: Adequate pump NPSH Tier #

2 Group #

1 K/A #

K4.04 Rating 2.6 Question 3 The A RHR pump is running in Shutdown cooling mode.

One design feature of the RHR system while in this mode that protects this pump from low Net Positive Suction Head is an interlock that trips the pump if _____.

A. Valve E12-F010, RHR SDC MAN ISOL VLV, is NOT FULL OPEN.

B. Valve SFC-V109, DRYER STORAGE POOL OUTAGE PURIFICATION SYS, is OPEN.

C. Valve E12-F008, RHR SHUTDOWN COOLING OUTBD ISOL VALVE, is NOT FULL OPEN.

D. Valve E12-F066A, RHR A FUEL POOL COOLING SUCTION, is NOT FULL CLOSED.

Answer: C Explanation:

A is wrong because this valve is a manual valve and does not have an interlock associated with the RHR pump for NPSH/low suction pressure.

B is wrong because this valve has a caution in SOP-0031 to ensure that it is not open to prevent low NPSH but there is no interlock for it with the pump for NPSH/low suction pressure.

C is correct per the Figure 10 of the STM-204 and the SOP-0031 guidance. This valve must be FULL OPEN or the K19A relay energizes and trips the pump due to low NPSH concerns.

D is wrong because the interlock is on the NOT FULL OPEN logic not the NOT FULL CLOSED logic.

Technical

References:

STM-0204, rev 12, page 70, Figure 10, and SOP-0031, revision 327.

References to be provided to applicants during exam: None Learning Objective: STM-204 Obj 8 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 209001 LPCS Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM: K5.04 Heat removal (transfer) mechanisms Tier #

2 Group #

1 K/A #

K5.04 Rating 2.8 Revision 1

Question 4 While in the EOPs, the following conditions exist:

• RPV level is -195 inches and stable

• RPV pressure is 0 psig

• LPCS flow rate is 5010 gpm What is the PRIMARY method of core cooling for these conditions?

A. Core Submergence B. Spray Cooling C. Steam Cooling with Injection D. Steam Cooling without Injection Answer: B Explanation:

A is wrong because according to EOP-1 core submergence is the primary method above -

162 inches B is correct because according to EOP-1 spray cooling is assured with LPCS flow greater than 5000 gpm and RPV level above -211 inches C is wrong because steam cooling with injection is assured greater than -187 inches D is wrong because steam cooling without injection is assured greater than -200 inches.

Technical

References:

EPSTG-02, EOP Bases References to be provided to applicants during exam: None

Learning Objective: RLP-OPS-HLO-511, Obj 4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental H2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(14)

Examination Outline Cross-Reference Level RO 209002 HPCS Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS): Condensate Storage Tank Water Level Tier #

2 Group #

1 K/A #

K6.02 Rating 3.4 Revision 1

Question 5 A stroke test of E22-F001, HPCS Pump CST Suction Valve, was in progress when the valve became mechanically bound on its closed seat. While the valve is bound in the closed position a large break LOCA occurs that automatically initiates HPCS.

With NO operator action the HPCS system ___(1)___ provide cooling water flow to the RPV ___(2)___.

A. (1) will (2) because the E22-F015, HPCS Pump Suppression Pool Suction Valve, automatically opens at < 2.4 feet CST level.

B. (1) will not (2) because CST level will not reach the actuation set point for E22-F015, HPCS Pump Suppression Pool Suction Valve, to open with E22-F001, HPCS Pump CST Suction Valve, closed.

C. (1) will (2) because E22-F015, HPCS Suppression Pool Suction Valve, automatically opens due to high suppression pool level at 19 6.

D. (1) will not (2) because annunciator HPCS PUMP SUCTION PRESSURE HI/LOW will be lit and prevents auto start of the HPCS pump.

Answer: B Explanation:

Plausibility Issue: STP-203-6305 step 7.1.1 says to close the 1 valve and record the closing time; it also has a caution stating that Closing the 1 valve will isolate the suction source for the HPCS line fill pumpdo not prolong the time that E22-MOVF001 is closed. The next step after recording the closing time is to open the 15 valve.

A. Incorrect while 2.4 feet in the CST is a signal to open MOV015 the level will not be reached due to no flow through the normal suction valve from the CST B. Correct C. Incorrect, High suppression pool level is an automatic open set point for MOVF015 it could be reached will a large break loss of coolant accident occurring if the right conditions were met but the setpoint is wrong for auto open of the 15 valve (it is set to open when Supp. Pool Level is > 20'4".

D. Incorrect, The alarm would not immediately come in and even if it did it does not prevent an auto start of the pump.

Technical

References:

R-STM-0203, Revision 8, page 33.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0203 Obj D Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 211000 Standby Liquid Control System Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: SBLC system lineup Tier #

2 Group #

1 K/A #

A1.09 Rating 4.0 Question 6 The plant is operating at 100% power, and the operators are preparing to run SLC Pump A with a suction from the test tank.

C41-F031, TEST TK OUTLET VLV, has been opened, but the pump has yet to be started.

Subsequently, a reactor SCRAM occurs with an ATWS.

The CRS has directed injection with Standby Liquid Control using SLC Pump B.

The operator takes the pump switch for SLC Pump B to RUN.

What will be the position of C41-F001B, SLC PUMP B SUCT VLV, and C41-F004B, SQUIB CONTINUITY VALVE?

C41-F001B C41-F004B A. OPEN OPEN B. OPEN CLOSED C. CLOSED OPEN D. CLOSED CLOSED Answer: C Explanation:

A. INCORRECT. C41-F001B is interlocked to C41-F031, and will not open unless C41-F031 is closed. C41-F004B will open when the pump switch is taken to run.

B. INCORRECT. C41-F001B is interlocked to C41-F031, and will not open unless C41-F031 is closed. C41-F004B will open when the pump switch is taken to run.

C. CORRECT. C41-F001B is interlocked to C41-F031, and will stay closed since C41-F031 is open. C41-F004B will open when the pump switch is taken to run.

D. INCORRECT. C41-F001B is interlocked to C41-F031, and will stay closed since C41-F031 is open. C41-F004B will open when the pump switch is taken to run.

Technical

References:

STM-0201, Standby Liquid Control, Revision References to be provided to applicants during exam: None Learning Objective: RLP-STM-0201, Objective F (a)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41.5

Examination Outline Cross-Reference Level RO 212000 RPS Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Changing mode switch position Tier #

2 Group #

1 K/A #

A2.16 Rating 4.0 Question 7 A reactor startup is in progress, using procedure GOP-0001 Reactor power is 13% and stable Reactor System Mode Switch is in the RUN position Intermediate Range Monitors (IRMs) are still inserted, with all on Range 9 with 110/125 of scale

1. If an operator took the Reactor System Mode Switch to the START & HOT STBY position, what effect would it have on the Reactor Protection System?

2. Based on these conditions, what procedural entry is necessary?

A. 1)

Initiates a reactor scram

2)

AOP-0001, Reactor Scram B. 1)

Initiates a reactor scram

2)

GOP-0002, Power Decrease/Plant Shutdown C. 1)

Bypasses Reactor Water Level - High signal

2)

AOP-0001, Reactor Scram D. 1)

Bypasses Reactor Water Level - High signal

2)

GOP-0002, Power Decrease/Plant Shutdown Answer: D

Explanation:

A is wrong. This is plausible if the applicant believes that any of the reactor scram set points in effect have been exceeded.

B is wrong. There is no reactor scram. Even if one was initiated, after AOP-0001 is entered and completed, it directs entry afterwards into GOP-0003, Scram Recovery.

C is wrong. The Reactor Water Level - High signal is bypassed. However, nothing indicated in plant conditions would cause a reactor scram, and there isnt a condition that would cause an operator to initiate a reactor scram.

D is correct. Per R-STM-0508, when the Reactor System Mode Switch is in the START &

HOT STBY position, the MSIV Closure and Reactor Water Level - High scram signals are bypassed. Also, the APRM Neutron Flux - High scram set point is reduced to 15% power.

The IRMs are still inserted, so their scram set point is 120/125 of scale on its ranges (R-STM-0503). Based on these conditions, there is no signal that would cause a reactor scram, and there isnt a condition that should cause an operator to initiate a reactor scram.

The purpose of GOP-0002 is To provide guidelines to reduce power from any point to any standby or shutdown mode. Therefore, this procedure would be entered to address the situation.

Technical

References:

R-STM-0500, Rod Control and Information System, Revision 4 R-STM-0503, Neutron Monitor Instruments, Revision 9 R-STM-0508, Reactor Protection System, Revision 6 GOP-0001, Plant Startup, Revision 86 GOP-0003, Scram Recovery, Revision 27 GOP-0002, Power Decrease/Plant Shutdown, Revision 73 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0508, Obj 5, 7.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 215003 IRM Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM)

SYSTEM including: A3.02 Annunciator and alarm signals Tier #

2 Group #

1 K/A #

A3.02 Rating 3.3 Revision 1

Question 8

• The plant is conducting a reactor startup, with the Reactor Mode Selector Switch in the Startup position

• Reactor power is being monitored in the intermediate range

• Power is available to all neutron monitoring instrumentation and drive control

• Alarm Window P680-07-C01, CONTROL ROD WITHDRAWAL BLOCK, is lit Which of the following indications would be consistent with the above conditions?

Digital Recorder C51-R603B, red pen reads:

A. 36 on range 5 B. 122 on range 4 C. 40 on range 6 D. 1 on range 1 Answer: A Explanation:

A is correct because the rod block is >108/125 of scale which is >34.56 on the 0 - 40 scale.

The student must calculate what the rod block setpoint is for the odd ranges.

B is wrong because this condition would cause both a rod withdrawal block and a reactor scram signal. The rod block and scram set points are 108/125 and 120/125, respectively.

See Pages 34 and 36 of R-STM-0503.

C is wrong because this would cause nothing, but plausible because the student must know which scale is used on range 6. An indication of 40 on an odd range will cause a RPS scram signal.

D is wrong because IRMs on Range 1 bypasses the IRM Downscale rod withdrawal block (

5/125 scale). See Page 34 of R-STM-503. Therefore, this condition would not support the existence of the rod block indication provided in the questions initial conditions.

Technical

References:

System Training Manual R-STM-0503, Neutron Monitoring Instruments System, Revision 9 Procedure ARP-680-07, P608-07 Alarm Response, Revision 35 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0503, Obj 13.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 215004 Source Range Monitor Ability to manually operate and/or monitor in the control room: A4.07 Verification of proper functioning/ operability Tier #

2 Group #

1 K/A #

A4.07 Rating 3.4 Revision 2

Question 9 The reactor has been shut down for an extended refueling outage.

SRM indications are as follows:

• SRM A reads 2X100 cps

• SRM B reads 4X100 cps

• SRM C reads 5X100 cps

• SRM D reads 2X105 cps The reason for a rod withdraw block is due to which of the following?

A. Source Range Downscale and Source Range Monitor High Flux Only B. Source Range Downscale Only C. Source Range Monitor High Flux Only D. Source Range Monitor High Flux and SRM Inoperable Only Answer: A Explanation:

A. Correct B. Incorrect because Channel SRM channel greater than 1X105 cps will also cause a rod withdrawal block C. Incorrect because low source range monitor of < 3 cps will also generate a rod withdrawal block D. Incorrect, because low source range monitor of < 3cps will also generate a rod withdrawal block

Technical

References:

R-STM-0503, Rev 9 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0503, Obj 4.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.41(b) 2

Examination Outline Cross-Reference Level RO 215005 APRM / LPRM 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Tier #

2 Group #

1 K/A #

2.4.34 Rating 4.2 Revision 1

Question 10 The Main Control Room is evacuated due to a fire.

• The reactor is manually scrammed from 100% power prior to evacuation.

Control rod positions are unknown.

• The ATC proceeds to the Division 1 Remote Shutdown Panel (RSP) to implement AOP-0031 Attachment 12, Shutdown Outside Main Control Room:

ATC Operator Actions.

After 7 minutes, the ATC observes 1 SRV remains open on the RSP with reactor pressure constant.

1) What APRM-equivalent value of thermal power does this indicate?

2) What is the source of this power?

A. 1) 3%

2) decay heat only B. 1) 7%

2) decay heat only C. 1) 3%

2) decay heat plus fission heat D. 1) 7%

2) decay heat plus fission heat Answer: D Explanation:

There is no APRM/LPRM indication of power at the Div I Remote Shutdown Panel. Reactor power must be estimated by the number of SRVs open.

Each SRV full open is 7% power. Decay heat 6 minutes after a SCRAM is approximately 3%

thermal power which is indicated by one SRV cycling.

IF Reactor power is greater than 3%, 6 minutes after the SCRAM, THEN enter EOP-0001, Emergency Operating Procedure - RPV Control.

To answer this question, the applicant must understand how to approximate reactor power from SRV position at the RSP, know the expected value of thermal power post-scram, and recognize that a positive deviation from this expected value of thermal power must be an indication of continued fission above the point of adding heat. This is knowledge required for an RO to effectively perform his functions at the RSP.

A is wrong because 3% power would be indicated by one cycling SRV, and the RSP indicates that there is continuing fission above POAH in addition to decay heat.

B is wrong because one SRV full open 6 minutes post-SCRAM indicates that there is continuing fission above POAH in addition to decay heat.

C is wrong because 3% power would be indicated by one cycling SRV.

D is correct because one SRV full open is equivalent to 7% power, and since decay heat is expected to be 3% thermal power post-SCRAM, the increase is due to fission above POAH.

Technical

References:

AOP-0031, Shutdown Outside Main Control Room, Attachment 12, Revision 323.

References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross-Reference Level RO 217000 RCIC Knowledge of the physical connections and/or cause/effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: K1.04 Main condenser Tier #

2 Group #

1 K/A #

K1.04 Rating 2.6 Question 11 The RCIC system piping has drain pots located on the RCIC turbine steam supply header that provide protection for the turbine and piping from water hammer.

These drain pots are normally ___(1)___ and drain to the ___(2)___.

A. 1) open

2) suppression pool B. 1) closed

2) suppression pool C. 1) open

2) main condenser D. 1) closed

2) main condenser Answer: C Explanation:

A is wrong due to the second part being incorrect. They are normally open but they drain to the main condenser not the suppression pool.

B is wrong for both parts. They are normally open (not closed) and they drain to the main condenser (not supp pool).

C is correct - they are normally open and drain to the main condenser D is wrong because the first part is incorrect - they are normally open, not closed. The second part is correct.

Technical

References:

STM-0209, revision 12, page 17.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0209 N10 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 218000 ADS Knowledge of ADS design features and/or interlocks which provide for the following:

Logic control Tier #

2 Group #

1 K/A #

K4.03 Rating 3.8 Question 12 The following conditions exist:

• A leak inside the drywell has occurred.

• Only one RHR pump is running.

• Automatic Depressurization System automatically actuated at 105 seconds and all ADS SRVs are open.

• RPV water level is now steady at -150 inches.

• Drywell pressure peaked at 1.5 psid and is now lowering.

• RPV pressure is 200 psig.

If both Div 1 and Div 2 ADS TIMER/LEVEL 3 SEAL IN RESET buttons are depressed and then released, which of the following describes the result on the Automatic Depressurization System (ADS)?

The ADS SRVs will _____.

A. close and then reopen after 5 minutes plus 105 seconds.

B. close and then reopen after 105 seconds.

C. close and remain closed.

D. remain open.

Answer: A Explanation:

A. Correct-they will close when the reset pushbuttons (ADS TIMER/LEVEL 3 SEAL-IN RESET pushbuttons, S13A(B)) are operated, which resets the 105-second timers and

because the low water level signal is still in (-143 inches) the 5 minute timer also starts. Unless level goes above -143 inches they will reopen once the timer is out.

B. Incorrect-see A - credible because it you dont understand the logic or forget about the 5 minute timer you might pick this distracter.

C. Incorrect-if you dont recognize the initiation signal is still active for RPV level then you might think this is correct.

D. Incorrect - if you dont understand the logic of the reset seal-in buttons you might think that you have to use the manual buttons on the SRVs to close them once the reset is selected, similar to reset on ECCS equipment, but this is not true for this logic circuit.

Technical

References:

STM-0202 Rev 2, pages 7 and 13-15, and Figures 2 and 4.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0202 Obj H3 and L2 Question Source:

Bank #

NRC 2003 (Q14)

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental H3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b) 8

Examination Outline Cross-Reference Level RO 223002 PCIS/Nuclear Steam Supply Shutoff Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

Turbine building radiation Tier #

2 Group #

1 K/A #

K3.06 Rating 2.8 Question 13 The cause of turbine building radiation levels increasing while LOCA signals are present is ____ isolation logic failing to fully actuate.

A. Reactor Water Clean-Up B. Main Steam Isolation Valve C. Residual Heat Removal D. Balance-of-Plant Answer: B Explanation:

A is wrong because while RWCU logic is actuated from high drywell pressure the failure would result in Auxiliary Building radiation levels to rise. See R-STM-58 p. 40. Also, main steam tunnel temp high is one of the isolation signals for RWCU but this is not a LOCA signal.

B is correct because a failure to isolate the MSIVs and Main Steam Line drains, which protect against a steam line break, would result in Turbine Building radiation levels increasing with a break in that area. See R-STM-0058, Page 41.

C is wrong because LOCA signals feed into the isolation logic through the LPCI initiation logic the valves that would be isolated (RHR test return, condensate flush line to suppression pool) would not affect turbine building radiation levels. See STM-58, pp. 29 and 42.

D is wrong because while LOCA signals feed into the isolation logic the lines affected either arent radioactive or dont penetrate the turbine building. See STM-58, pp 17 and 42.

Technical

References:

R-STM-0058, Containment and Reactor Vessel Isolation Control System (CRVICS), Rev 10 References to be provided to applicants during exam: None

Learning Objective: 5.5.

Effects of loss or malfunction of CRVICS.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) (7)

Examination Outline Cross-Reference Level RO 239002 SRVs Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: Insures that only one or two SRVs reopen following the initial portion of a reactor isolation event (LLS logic).

Tier #

2 Group #

1 K/A #

K4.01 Rating 3.9 Rev 1 Question 14 The design feature of Low-Low Set mode for the Safety Relief Valves is that the designated SRVs will ___(1)___ in order to reduce cyclical stresses on the ___(2)___.

A. (1) open earlier and stay open longer (2) RPV B. (1) open earlier and stay open longer (2) containment C. (1) open later and stay open shorter (2) RPV D. (1) open later and stay open shorter (2) containment Answer: B Explanation:

A is wrong because the second part is wrong, it is to reduce cyclical stresses on containment not the RPV.

B is correct, to open earlier and stay open longer to reduce stresses on containment C is wrong because both aspects are wrong.

D is wrong because the first part is wrong.

Technical

References:

R-STM-0109, Rev. 15, page 10.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0109, Obj 4 and 5

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 239002 SRVs Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES: Discharge line quencher operation Tier #

2 Group #

1 K/A #

K5.05 Rating 2.6 Question 15 Per EOP basis, which one of the following is the highest suppression pool level at which opening an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, quencher, or quencher supports (SRV Tail Pipe Level Limit)?

A. 15 feet 5 inches B. 19 feet 6 inches C. 20 feet D. 21 feet 3 inches Answer: D Explanation:

A is wrong because when suppression pool level decreases to two feet above the top of the Mark III horizontal vents, any further drop in water level could result in direct exposure of the drywell atmosphere to the containment airspace thus compromising the pressure suppression function of the containment. Suppression pool level should therefore be maintained above this elevation. Per EOP-2 Step SPL-9, this level is 15 feet 5 inches.

B is wrong because per Tech Spec 3.6.2, 19 feet 6 inches is the minimum operating suppression pool level in modes 1, 2, and 3.

C is wrong because per Tech Spec 3.6.2, 20 feet is the maximum operating suppression pool level in modes 1, 2, and 3.

D is correct because per EOP basis, the STPLL is 21 feet 3 inches. Since SRV operation with suppression npool level above the STPLL could lead to containment failure, the RPV is not permitted to remain at pressure if suppression pool level will exceed this level.

The SRV Tail Pipe Level Limit (STPLL) is the lesser of:

• The Maximum Pressure Suppression Primary Containment Water Level.

• The highest suppression pool level at which opening an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, quencher, or quencher supports.

Technical

References:

EPSTG*0002, Page B-8-25 Rev 17 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0109, Objective 3 f.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 259002 Reactor Water Level Control Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: K6.02 A.C. power Tier #

2 Group #

1 K/A #

K6.02 Rating 3.3 Question 16 The reactor is operating at 55 % rated thermal power.

Three Reactor Feed Pumps are operating The 480 volt breaker NHS-MCC1C1 for FWL-P1A Reactor Feed Pump Main Oil Pump trips on overcurrent.

Based on the provided information what is the expected response of the Reactor Vessel Level?

A. RPV level will initially lower and then be restored to programmed level B. RPV Level will initially lower and then stabilize at a new lower level C. RPV Level will remain constant D. RPV Level will lower to the point of an automatic reactor scram Answer: C Explanation:

A. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating B. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating C. Correct D. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating

Technical

References:

R-STM-0107 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0107, Obj 15 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 259002 Reactor Water Level Control Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor power Tier #

2 Group #

1 K/A #

A1.03 Rating 3.8 Revision 2

Question 17 The reactor is operating at 90% power with the Feedwater Reg Valve Master Flow controller in Manual.

(1) What happens when RPV water level drops to Level 4 without any operator action?

(2) What happens when RPV level drops to Level 3 without any operator action (after the SCRAM and Recirc Pumps downshift to slow speed)?

A. (1) A Recirc FCV Runback will occur.

(2) Master Flow Controller Setpoint Setdown is initiated.

B. (1) A Recirc FCV Runback will not occur.

(2) Master Flow Controller Setpoint Setdown is not initiated.

C. (1) A Recirc FCV Runback will occur.

(2) Master Flow Controller Setpoint Setdown is not initiated.

D. (1) A Recirc FCV Runback will not occur.

(2) Master Flow Controller Setpoint Setdown is initiated.

Answer: B Explanation:

A is wrong because 3 feed pumps are running, so the Recirc FCV Runback will not occur.

Also since the Master Flow Controller is in manual, Setpoint Setdown is not initiated.

B is correct.

C is wrong because 3 feed pumps are running, so the Recirc FCV Runback will not occur.

D is wrong because since the Master Flow Controller is in manual, Setpoint Setdown is not initiated.

Technical

References:

R-STM-0107, Rev 27, REACTOR FEEDWATER AND LEVEL CONTROL SYSTEMS References to be provided to applicants during exam: None Learning Objective: RLP-STM-0107 Obj # B3, H5, L4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 261000 SGTS Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low system flow Tier #

2 Group #

1 K/A #

A2.01 Rating 2.9 Rev 1 Question 18 Following a LOCA inside containment, the SGTS system started automatically.

While implementing emergency operating procedures, the B train of SGTS was placed in standby with Inlet Damper, GTS-AOD1B closed.

Later, with LOCA conditions still present, flow in the A train of SGTS lowers to 650 cfm.

What is the impact of this condition?

The B train _____.

A. will not auto start due to flow not being low enough for the auto start signal.

B. will auto start and then reposition the inlet damper automatically.

C. will auto start after the inlet damper fully opens.

D. will only start manually after the inlet damper has been fully opened.

Answer: C Explanation:

According to the system training the GTS will only start if the inlet damper is fully open.

There doesnt appear to be any auto opening feature for the inlet dampers. The setpoint for the redundant train to auto start is less than 656 cfm.

A is wrong because the flow is low enough to auto start but the inlet damper is shut B is wrong because there doesnt appear to be an auto open signal for the inlet damper C is correct because see above D is wrong because the GTS can be auto started also.

Other potential distractors:

Will auto start after inlet damper begins to move from its closed seat.

Can only be started manually and the inlet damper will open automatically.

Can only be started manually after the inlet damper begins to move from its closed seat.

Technical

References:

R-STM-257, Standby Gas Treatment System, Rev 5 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0257, Obj 5 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 261000 SGTS Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: System Temperature Tier #

2 Group #

1 K/A #

A3.04 Rating 3.0 Revision 1

Question 19 A plant event occurs that causes the automatic start of both trains of Standby Gas Treatment System (GTS).

A control room operator can monitor the charcoal bed inlet temperatures at __(1)__.

Charcoal bed temperature exceeding __(2)__ would cause the annunciator SGT FILTER TRAIN FLT1A(B) CHARCOAL TEMPERATURE HIGH to alarm.

A. 1) panel H13-P863

2) 233.5F B. 1) panel H13-P863

2) 195F C. 1) process computer

2) 233.5F D. 1) process computer

2) 195F Answer: A Explanation:

A is correct. Per R-STM-0257, Page 10 of 28, GTS-RTD27A(B) provides temperature indication of the charcoal bed inlet temperature on panel H13-P863. Table 4 of R-STM-0257 indicates that the selected annunciator comes in when charcoal bed temperature exceeds 233.50F.

B is wrong. Part 1) is correct, but 2) is incorrect. The GTS filters are provided with a charcoal bed fire detector (which alarms in the control room at 195F and 225F). (R-STM-257 Rev 5, Page 6).

C is wrong. GTS-RTD14A(B), which does provide indication on the process computer, is an indication of charcoal bed outlet temperature.

D is wrong. See the Explanation for answers B and C.

Technical

References:

R-STM-0257, Standby Gas Treatment System, Revision 5 ARP-863-73, P863-73 Alarm Response, Revision 8 SOP-0043, Standby Gas Treatment System (SYS #257), Revision 17 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0257, Obj 3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 262001 AC Electrical Distribution Ability to manually operate and/or monitor in the control room: Local operation of breakers Tier #

2 Group #

1 K/A #

A4.03 Rating 3.2 Rev 4

Question 20

1) Breaker 20620, RSS #1 North 230KV Bus Feeder Breaker, can be operated from _____.

2) Breaker 20650, Port Hudson North 230KV Bus Feeder Breaker, can be operated from _____.

A. 1) Control Room

2) Control Room B. 1) Control Room

2) Southwest Transmission Operations Center Dispatcher C. 1) Southwest Transmission Operations Center Dispatcher

2) Control Room D. 1) Southwest Transmission Operations Center Dispatcher

2) Southwest Transmission Operations Center Dispatcher Answer: B Explanation:

A is wrong. See B.

B is correct because breaker 20620 is remotely operated by River Bend and breaker 20650 is remotely operated by Dispatcher.

C is wrong because see B D is wrong because see B Technical

References:

OSP-48, Attachment 2, Revision 31 R-STM-0300, page 18 & 36, Revision 028.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-300 Revision 9, Objective 5 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 262002 UPS (AC/DC) 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Tier #

2 Group #

1 K/A #

2.4.31 Rating 4.2 Revision 1

Question 21 Which of the following conditions will cause the receipt of annunciator H13-P808/87A/A08, ENB-INV01B / 01B1 VITAL BUS INV TROUBLE?

A. Bypass Source Overvoltage B. High DC Battery Voltage C. Inverter Output Current High D. Single Fan Failure Answer: D Explanation:

A is wrong because Bypass source UNDER voltage is a cause.

B is wrong because LOW DC voltage from battery is a cause.

C is wrong because Inverter Output current LOW is an alarm indication on Non SCI inverters.

(R-STM-0300, p. 31)

D is correct because a single fan failure does bring in this alarm Technical

References:

R-STM-0300, REV 028 - AC Distribution ARP-808-87, P808-87 ALARM RESPONSE, Rev 24 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0300, Obj 12.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F4 Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 263000 DC Electrical Distribution Knowledge of the physical connections and/or cause/effect relationships between D.C.

ELECTRICAL DISTRIBUTION and the following: A.C. electrical distribution Tier #

2 Group #

1 K/A #

K1.01 Rating 3.3 Revision 1

Question 22 BYS-CHGR1D (BACKUP BATTERY CHARGER) is being placed IN service.

(1) What is the normal power supply for this battery charger?

(2) Which breaker should be closed first (AC or DC), when placing battery chargers in service?

A. (1) EJS-SWG2B (2) AC Breaker B. (1) NJS-SWG1U (2) DC Breaker C. (1) EJS-SWG2B (2) DC Breaker D. (1) NJS-SWG1U (2) AC Breaker Answer: B Explanation:

A is wrong because EJS-SWG2B is the power supply for battery charger IHS-CHGR1D. AC breaker is also incorrect. When placing a battery charger in service the DC breaker must be closed first, then the AC.

B is correct.

C is wrong because EJS-SWG2B is the wrong power supply. The DC breaker is correct.

D is wrong because closing the AC breaker first is incorrect.

Technical

References:

R-STM-0305, Rev. 7, DC Distribution References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0305, Obj 4, 12 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 264000 EDGs Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Emergency core cooling systems Tier #

2 Group #

1 K/A #

K3.01 Rating 4.2 Rev 1 Question 23 The station has experienced a Loss of Coolant Accident concurrent with a Loss of Off-site Power.

In this situation, a ___(1)___ on the Division 1 Emergency Diesel Generator would cause the loss of ___(2)___ needed for the LOCA conditions.

A. (1) Generator Differential (2) only the LPCS and RHR A pumps B. (1) Generator Differential (2) only the RHR A and RHR C pumps C. (1) Sustained lube oil pressure of 20 psig (2) only the LPCS and RHR A pumps D. (1) Sustained lube oil pressure of 20 psig (2) only the RHR A and RHR C pumps Answer: A Explanation:

A is correct because LO pressure trips are disabled for LOCA events while a generator differential fault is not. The second part is knowing what the division 1 EDG powers and the answer is LPCS and RHR A pump only.

B is incorrect because the second part is wrong. RHR pump C is off of the division 2 bus/EDG. The first part is correct.

C is incorrect because LO pressure trips are disabled for LOCA events and the second part is correct.

D is incorrect because LO pressure trips are disabled for LOCA events and the pumps are incorrect because RHR pump C is off of the division 2 bus/EDG.

Technical

References:

STM-309S, revision 13.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-309S, 5A, 6D.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 300000 Instrument Air System Knowledge of electrical power supplies to the following: Instrument air compressor Tier #

2 Group #

1 K/A #

K2.01 Rating 2.8 Rev 1 Question 24 Which of the following provides power to instrument air compressor IAS-C2B?

A. NHS-MCC1E B. NHS-MCC1G C. NJS-SWG1G D. NJS-SWG1H Answer: D Explanation:

A. INCORRECT. This is a non-safety MCC power supply to non-safety equipment, but could be picked because it is on 68 ft level of turbine bldg., where other IAS equipment is located B. INCORRECT. This is a non-safety MCC power supply to non-safety equipment, but could be picked because it is on 68 ft level of turbine bldg., where other IAS equipment is located C. INCORRECT. This is the power supply to instrument air compressor IAS-C2A.

D. CORRECT. This is the power supply to instrument air compressor IAS-C2B.

Technical

References:

EE-001AC, Startup Electrical Distribution Chart, Rev 47 References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0121, Objective E Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 300000 Instrument Air Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Air compressors Tier #

2 Group #

1 K/A #

K5.01 Rating 2.5 Question 25 The instrument air compressors are all available. Air compressor IAS-C2C is currently the only one operating to maintain system pressure and the local SEQUENCE CONTROL switch is in Position 3 (C-A-B).

A relay failure in the control circuit for compressor IAS-C2C causes it to shutdown.

With no change in Instrument Air System usage, which of the following describes the effect of the compressor shutdown on the Instrument Air System?

A. IAS-C2A operates alone, maintaining a lower header pressure.

B. IAS-C2A operates alone, maintaining the same header pressure.

C. Both IAS-C2A and C2B are operating, maintaining the same header pressure.

D. Both IAS-C2A and C2B are operating, maintaining a lower header pressure.

Answer: A Explanation:

A is Correct-IAS-C2A is the second compressor to sequence on in Position 3, so it will start as pressure lowers to the mid range setpoint (115.5 psig) which is lower than the pressure setpoint for the start of IAS-C2C (118.5 psig).

B is wrong because it will not maintain the same header pressure.

C is wrong because system demand is unchanged, so there is no need for two compressors to run. Additionally, system pressure will not be maintained the same as before the shutdown (118.5 psig) because IAS-C2A is being controlled by the mid range pressure switches (115.5 psig) in Position 3.

D is wrong because system demand unchanged, so there is no need for two compressors to run.

Technical

References:

R-STM-0121 Rev 16 Pg 14 of 69 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0121 Obj 3a Question Source:

Bank #

NRC 2012 (Q47)

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)4

Examination Outline Cross-Reference Level RO 400000 Component Cooling Water Knowledge of electrical power supplies to the flowing: CCW pumps Tier #

2 Group #

1 K/A #

K2.01 Rating 2.9 Question 26 What are the power supplies to the RPCCW (CCP) pumps?

A. CCP-P1A NJS-SWG1B CCP-P1C NJS-SWG1D B. CCP-P1A NJS-SWG1A CCP-P1C NJS-SWG1B C. CCP-P1A NJS-SWG1A CCP-P1C NJS-SWG1C D. CCP-P1A NJS-SWG1A CCP-P1C NJS-SWG1A Answer: D Explanation:

A is wrong -these are credible as they are 480 MCC but not correct (see D)

B is wrong - these are credible as they are 480 MCC but not correct (see D)

C is wrong - these are credible as they are 480 MCC but not correct (see D)

D is correct -Both pumps are off of the NJS switchgear and on both are on the A bus.

Technical

References:

R-STM-0115, Rev 6, Reactor Plant Component Cooling System (CCP)

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0115 Obj E5 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 201001 CRD Hydraulic Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Tier #

2 Group #

2 K/A #

2.2.44 Rating 4.2 Revision 2

Question 27 While performing friction testing, the control rod that is selected moves an additional half-step than intended with --(double blank) being indicated as the control rod position when the insert pushbutton is released.

The Control Rod Drift annunciator would be ___(1)___ and the control rod would be moved ___(2)___.

1 2

A.

lit; in to 00 B.

lit; to its original position C.

extinguished; in to 00 D.

extinguished; to its original position Answer: D Explanation:

A is wrong because the annunciator would not be lit but if it was it would be required to be inserted to 00.

B is wrong because the annunciator would not be lit.

C is wrong because the operator would not be required to move the rod to 00 since the rod is not drifting.

D is correct because with a normal insert command the control rod drift annunciator would not illuminate. Due to the rod not going in more than one additional step the operator would be allowed to pull the rod back out to the original step per procedure guidance.

Technical

References:

AOP-61, Control Rod(s) Mispositioned/Malfunction, Rev 8 STP-052-0102, Partially Withdrawn Control Rod Insertion Operability Check, Rev 9 References to be provided to applicants during exam: None Learning Objective: RLP-OPS-HLO-0711, Obj 2, 4; RLP-STM-0052, Obj 9 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross-Reference Level RO 201005 RCIS Knowledge of the operational implications of the following concepts as they apply to ROD CONTROL AND INFORMATION SYSTEM (RCIS): (CFR: 41.5 / 45.3)

K5.09 High power setpoints BWR-6 Tier #

2 Group #

2 K/A #

K5.09 Rating 3.5 Question 28 Which of the following describes the operational implications of maintaining control rods within designated withdrawal distances during control rod movements above the high power setpoint?

A. Establish a 2 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

B. Establish a 4 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

C. Establish a 2 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

D. Establish a 4 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

Answer: C Explanation:

During an up-power maneuver, once the LPSP is reached, the constraints of the RPCS are no longer in effect. At this point, the Rod Withdrawal Limiter (RWL) comes into effect to limit the allowable number of notches of rod withdrawal. When power level is between the LPSP (27.5%) and the High Power Setpoint (HPSP) (67.9%), the RWL limits allowable notches of withdrawal for any control rod to 4 notches. Once power level passes above the HPSP (again sensed by first stage turbine pressure) the RWL limits allowable notches of withdrawal for any control rod to 2 notches. These notch withdrawal limitations are based upon reactor physics considerations relative to reactor power levels. The limits are imposed on each withdrawal operation of each rod. That is, once the RWL imposes its limits on a particular rod, the rod can be de-selected, re-selected, and withdrawn again.

A. incorrect - the rod pattern controller of RC&IS is designed to mitigate the consequences of a control rod drop accident B. incorrect - the rod pattern controller of RC&IS is designed to mitigate the consequences of a control rod drop accident C. Correct D. incorrect - the 4-notch limit is enforced between the LPSP and the HPSP Technical

References:

R-STM-0500, revision 4, page 16.

References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(6)

Examination Outline Cross-Reference Level RO 202002 Recirculation Flow Control Knowledge of the effect that a loss or malfunction of the RECIRCULATION FLOW CONTROL SYSTEM will have on following:

Recirculation pump speed: Plant-Specific Tier #

2 Group #

2 K/A #

K3.05 Rating 3.2 Revision 1

Question 29 The plant is starting up, coming out of an outage where maintenance was conducted on the reactor recirculation system controls.

Per procedure GOP-0001, Attachment 2, Section 4.1, a Control Board Lineup using A of SOP-0003 has been completed.

A control room operator is in the process of starting Reactor Recirculation pump A using SOP-0003, Section 4.4.11.

When the control room operator attempts to depress the RELEASE pushbutton on the STOP/PUSH TO LOCK control switch, the pushbutton will not depress.

With this malfunction in place, what effect does it have on Reactor Recirculation pump speed control logic?

A. Prevents recirculation pump slow starts ONLY B. Trips the recirculation pump after starting in slow speed ONLY C. Prevents recirculation pump slow AND fast starts D. Trips the recirculation pump after starting in slow or fast speeds ONLY Answer: C Explanation:

A is wrong. This malfunction would prevent recirculation pump slow and fast starts. See citations in the Explanation for Answer C.

B is wrong. See C.

C is correct. With the Reactor Recirculation system control board recently lined up according to SOP-0003, Attachment 4A, the PUSH TO LOCK pushbutton on the pumps STOP/PUSH TO LOCK control switch has been depressed. This prevents closing the CB-5 breaker, which limits the speed control functions that can used with the associated recirculation pump [see Section D.3.b)(1), Page 8 of 76, in R-STM-0053]. Three limitations are in place. Of these, answer C is one of the limitations stated.

D is wrong. It will trip the pump if it was running in fast speed only (not both fast and slow)

Technical

References:

R-STM-0053, Reactor Recirculation System, Revision 14 SOP-0003, Reactor Recirculation System (SYS #053), Revision 313 GOP-0001, Plant Startup, Revision 86 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0053, Obj 2, 6 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 204000 Reactor Water Cleanup System Knowledge of Reactor Water Cleanup System design feature(s) and/or interlocks which provide for the following: Over temperature protection for system components.

Tier #

2 Group #

2 K/A #

K4.03 Rating 2.9 Question 30 Temperature element G33-N007, Non-Regenerative Heat Exchanger Outlet temperature, is reading 142°F.

Which of the following automatic actions has occurred due to this temperature?

A. G33-F044, RWCU F/D Bypass Valve, automatically opened B. G33-F104, RWCU HX/DEMIN Bypass Valve, automatically opened C. G33-F004, RWCU Pump Outboard Suction Valve, automatically closed D. G33-F054, RWCU Pump Outboard Discharge Valve, automatically closed Answer: C Explanation:

A. INCORRECT. G33-F044, RWCU F/D Bypass Valve, is used to throttle flow around the filter demineralizers, and has no automatic function.

B. INCORRECT. G33-F104, RWCU HX/Demin Bypass Valve, is provided to allow for throttling flow around both the RHX and NRHX.

C. CORRECT. NRHX Outlet Temperature, or F/D Inlet Temperature, is interlocked with G33-F004, RWCU Pump Outboard Suction Valve, such that if temperature reaches 140°F, G33-F004 isolates.

D. INCORRECT. G33-F054, RWCU Pump Outboard Discharge Valve, RWCU Pump Room ambient tempratures.

Technical

References:

STM-0601, Reactor Water Cleanup (RWCU) System, Revision 8 References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0601, Objective D Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

GG 2010-06 X

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 216000 Nuclear Boiler Instrumentation Knowledge of the physical connections and/or cause/effect relationships between Nuclear Boiler Instrumentation and the following: Low pressure core spray Tier #

2 Group #

2 K/A #

K1.06 Rating 3.9 Question 31 Which of the following trip units are associated with the initiation of Low Pressure Core Spray?

A. Level 1 trip B. Level 2 trip C. Level 3 trip D. Level 4 trip Answer: A Explanation:

A. CORRECT. Low Pressure Core Spray is associated with a Level 1 trip.

B. INCORRECT. RCIC y is associated with a Level 2 trip.

C. INCORRECT. SCRAM is associated with a Level 3 trip.

D. INCORRECT. Feed pump is associated with a Level 4 trip.

Technical

References:

STM-0051, Nuclear Boiler Instrumentation, Revision 6 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0051, Objective M Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode Knowledge of electrical power supplies to the following: Valves Tier #

2 Group #

2 K/A #

K2.01 Rating 2.5 Revision 1

Question 32 Which of the following is the correct power supply for the RHR B & D Heat Exchanger Shell Bypass Valve E12-MOVF048B?

A. EHS-MCC2E B. EHS-MCC2F C. EHS-MCC2G D. EHS-MCC2H Answer: B Explanation:

A. INCORRECT. EHS-MCC2E is the power supply for E12-MOV048A.

B. CORRECT. EHS-MCC2F is the power supply for E12-MOV048B.

C. INCORRECT. EHS-MCC2G is the power supply for Division I RHR valves.

D. INCORRECT. This is not the correct power supply (see B).

Valve is important because after recovering reactor water level, it may be necessary to place at least one of the RHR loops in the suppression pool cooling mode of operation.

Ten minutes after the initiation signal is received, the open signal to the heat exchanger bypass valves MOVF048A(B) is removed, allowing them to be throttled back or closed as required.

Technical

References:

STM-0204, Residual Heat Removal System (RHR), Revision 12 References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0204, NLO Objective F (f)

Question Source:

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

Examination Outline Cross-Reference Level RO 223001 Primary CTMT and Aux.

Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES: Containment cooling: Mark-III Tier #

2 Group #

2 K/A #

K6.02 Rating 3.5 Question 33 During surveillance testing, I&C personnel inadvertently isolated all containment chilled water supply and return isolation valves.

What is the MAXIMUM containment or drywell temperature that would require BOTH the mode switch to be placed in shutdown and an emergency depressurization?

A. 110°F B. 145°F C. 165°F D. 185°F Answer: D Explanation:

185°F containment temperature is the correct answer.

110°F is the suppression pool temperature requiring the mode switch to be placed in shutdown but would not require an ED.

145°F drywell temperature would require a shutdown but not an ED.

165°F is the maximum post LOCA long term bulk air temp. This is more of a design temperature/limit for containment. 330°F drywell temperature would also require a shutdown and an ED.

Technical

References:

R-STM-0057, PRIMARY CONTAINMENT AND AUXILIARIES, Revision 5, p. 6-12 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0057, REV. 7, ATTACHMENT 3, Objective D

Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(9)

Examination Outline Cross-Reference Level RO 241000 Reactor/Turbine Pressure Regulator Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

High reactor pressure Tier #

2 Group #

2 K/A #

A2.02 Rating 3.7 Revision 3

Question 34 The plant is operating at 90% power.

Bypass EHC HPU pump D002A is isolated for repairs.

A reactor transient causes reactor pressure to increase to 1090 psig and steady.

Main condenser pressure is 25 in. Hg vacuum and steady.

Based on the plant conditions, the main turbine bypass valve(s) are ___(1)___.

Alarm response procedures that apply due to the expected alarm(s) direct the operator to ___(2)___.

A. (1) Closed (2) stop all control rod withdrawal B. (1) Open (2) stop all control rod withdrawal C. (1) Closed (2) enter the EOPs D. (1) Open (2) enter the EOPs Answer: B

Explanation:

A is wrong. For (1), it is plausible that either the secured bypass EHC HCU pump or vacuum could cause the bypass valves to stay closed. Citing the reference in answer Bs explanation, this is incorrect. (2) is correct.

B is correct. (1) is correct because R-STM-0509, Section F.2.g)(1) (Page 57 of 81) says that for the EHC HPUs, one operating pump will provide adequate EHC fluid pressure for continued operation. Condenser vacuum is lowering, but it is not at the pressure that would preclude bypass valves from opening (8.5 Hg). (2) is correct because the TURBINE BYPASS VALVE OPEN alarm would be lit (ARP-680-07, A07). The Operator Action is to stop all control rod withdrawal.

C is wrong. See Explanations for answers A and D.

D is wrong. (1) is correct, but (2) is incorrect. It is an action for the CONDENSER LO VACUUM alarm (ARP-680-07, D08). Condenser vacuum isnt low enough to get this alarm, and it isnt low enough to cause a MSIV isolation (8.5 Hg for both). Also OSP-53, Attach. 36 Trigger point for condenser vacuum, states to insert a scram by 23 vacuum.

Technical

References:

System Training Manual R-STM-0509, Electro-Hydraulic Control (EHC), Revision 14 ARP-680-07, P680-07 Alarm Response, Revision 35 OSP-53, Attachment 36 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0509, Obj 7 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 268000 Radwaste Ability to predict and/or monitor changes in parameters associated with operating the RADWASTE controls including: Radiation level Tier #

2 Group #

2 K/A #

A1.01 Rating 2.7 Question 35

• A liquid effluent discharge of a Recovery Sample Tank is in progress.

• An alarming condition caused by a High Alarm as measured by RMS-RE107, Liquid Radwaste Radiation Monitor, comes in at the Auxiliary Control Room.

In the Main Control Room, this condition can be confirmed by viewing a channel display of the associated radiation monitor with the color ___(1)___ on panel ___(2)___.

A. 1) Red

2) DSPL230 B. 1) Red

2) P878 C. 1) Yellow

2) P878 D. 1) Yellow

2) DSPL230 Answer: A Explanation:

A is correct. a) When a liquid effluent discharge is prepared for, Alert and Alarm levels are established for RMS-RE107. These values are input into the Digital Radiation Monitoring System (DRMS). When the channel display for the radiation monitor achieves its Alarm level, it is Red in color. See R-STM-0511, Section C.3.a). CSP-0110, Definitions 3.27 and 3.28, confirm that Alert Alarm and High Alarm set points are calculated. To match the DRMS scheme, these would correspond to its Alarm and Alert points.

For b), this radiation monitor, a non safety-related monitor, only has monitoring available via the DRMS in the Main Control Room. This is displayed in panel DSPL230. This interface is referred to in SOP-0113 and SOP-0086.

B is wrong. The channel display color is correct. However, panel P878 provides MCR safety-related radiation monitor indication.

C is wrong. The Alert color of Yellow is assigned to a different value than that assigned to the ACR alarm and the Alarm level set in DRMS. The panel doesnt provide indication for the radiation monitor in question.

D is wrong. The panel is correct, but the display color is incorrect.

Technical

References:

R-STM-0603, Radwaste Systems, Revision 7 R-STM-0511, Radiation Monitoring, Revision 15 SOP-0113, Liquid Radwaste Processing/Recovery Sample Tank System, Rev 24 SOP-0086, Digital Radiation Monitoring System (SYS #511), Revision 16 ADM-0054, Radioactive Liquid Effluent Batch Discharge, Revision 6A ARP-RMS-DSPL230, DRMS Computer CRT (RMS-DSPL230) Alarm Response, Rev 9 CSP-0110, Radioactive Liquid Effluent Batch Discharge, Revision 19 ARP-LWS-PNL187-4, LWS-PNL18704 Alarm Response, Revision 303 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0511, Obj 5, 7 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 271000 Offgas Ability to monitor automatic operations of the OFFGAS SYSTEM including: Process radiation monitoring system indications Tier #

2 Group #

2 K/A #

A3.07 Rating 3.4 Rev 1 Question 36 The major source of radiation monitored by the OFF-GAS radiation monitors is primarily gamma radiation ___(1)___. When the OFF-GAS POST TRT HIGH RADIATION alarm is received the OFF-GAS system will automatically ___(2)___.

A. (1) from O-19 and N-16.

(2) place the charcoal beds on line for removal of radioactive particulates.

B. (1) from O-19 and N-16.

(2) shut OFF-GAS DISCH TO VENT VLV, 1N64-F060, to isolate the off-gas stream from discharging to the stack exhaust ventilation.

C. (1) emitted by radioactive particles.

(2) place the charcoal beds on line for removal of radioactive particulates.

D. (1) emitted by radioactive particles.

(2) shut OFF-GAS DISCH TO VENT VLV, 1N64-F060, to isolate the off-gas stream from discharging to the stack exhaust ventilation Answer: C Explanation:

A. Is wrong because the sample is extracted from the line at a point where the shorter lived radioactive nuclei, principally O-19 and N-16, have had sufficient time to decay, and contribute only minimally to the gross radioactivity.

The charcoal beds would automatically be placed in service on receipt of this alarm.

B. Is wrong because the sample is extracted from the line at a point where the shorter lived radioactive nuclei, principally O-19 and N-16, have had sufficient time to decay, and contribute only minimally to the gross radioactivity The OFF-GAS DISCH TO VENT VLV, 1N64-F060 is isolated if An isolation of the Off-gas System will occur if both channels receive HI-HI-HI and/or INOP signals from the trip units C. Correct

D. Is wrong because OFF-GAS DISCH TO VENT VLV, 1N64-F060 is isolated if An isolation of the Off-gas System will occur if both channels receive HI-HI-HI and/or INOP signals from the trip units Technical

References:

System Training Manual, R-STM-0511 Rev 15, Radiation Monitoring Systems References to be provided to applicants during exam: None Learning Objective: RLP-STM-0511 Obj 6; RLP-STM-0606, Obj 1, 4, 7 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) 7

Examination Outline Cross-Reference Level RO 290001 Secondary CTMT Ability to manually operate and/or monitor in the control room: System status lights and alarms: Plant-Specific Tier #

2 Group #

2 K/A #

A4.09 Rating 3.2 Revision 1

Question 37 A LOCA is in progress and primary and secondary containment isolation signals are present.

The following dampers are analyzed in the Main Control Room to ensure correct Auxiliary Building isolation.

1) HVR-AOD164(143), Aux Bldg Inlet Isolation Dampers

2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers

3) HVR-AOD 249, Aux Bldg Outlet Isolation Damper

4) HVR-FN6A(B), Aux Bldg Supply Fans Which one of the following energized status lights is correct for the above dampers and fans?

HVR-AOD164(143)

HVR-AOD18A(B)

HVR-AOD 249 HVR-FN6A(B)

A. Green Green Green Red B. Green Red Green Green C. Green Green Green Green D. Green Red Green Red Answer: B Explanation:

A is wrong because (2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers are open and (4) supply fans are tripped.

B is correct C is wrong because (2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers are open.

D is wrong because (4) supply fans are tripped.

Technical

References:

R-STM-0409, Rev. 6, Auxiliary Building HVAC References to be provided to applicants during exam: None Learning Objective: RLP-STM-0406, Obj 2, 11 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental H3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(9)

Examination Outline Cross-Reference Level RO 290002 Reactor Vessel Internals Ability to explain and apply system limits and precautions.

Tier #

2 Group #

2 K/A #

2.1.32 Rating 3.8 Revision 2

Question 38 The RCS Pressure Safety Limit is set at ___(1)___ as measured at the ___(2)___.

A. (1) 1325 psig (2) steam dome B. (1) 1325 psig (2) vessel head flange C. (1) 1250 psig (2) steam dome D. (1) 1250 psig (2) vessel head flange Answer: A Explanation:

Per Tech Spec bases, the maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping, 1650 psig for discharge piping between the pump and the discharge valve, and 1550 psig beyond the discharge valve. The most limiting of these allowances is the 110% of the suction piping design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

Per R-STM-0051, Safety Limit 2.1.2, Reactor Coolant System Pressure. Reactor steam dome pressure shall be 1325 psig. Based upon allowing a maximum transient to 110% of design pressure as allowed by ASME Boiler and Pressure Vessel Code, in which 1325 psig yields 1250 +.10(1250 psig) = 1375 at the lowest point in the vessel.

A. CORRECT. See above.

B. INCORRECT. First part is correct. Second part is plausible because that is where the vessel temps are taken and is a weak point for high pressure with the flange being located there.

C. INCORRECT. First part is plausible because it is the design pressure of the suction piping.

D. INCORRECT. See above explanations.

Technical

References:

Tech Spec Bases 2.1.2, Page B 2.0-7 Per R-STM-0051, Revision 7, Page 29 of 47 References to be provided to applicants during exam: None Learning Objective: RLP-OPS-HLO-401, Obj 2, 6 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

Examination Outline Cross-Reference Level RO 295001 Partial or Complete Loss of Forced Core Flow Circulation Knowledge of the reasons for the following responses as they apply to partial or complete loss of forced core flow circulation: Reactor power response Tier #

1 Group #

1 K/A #

AK3.02 Rating 3.7 Question 39 According to AOP-0062, Jet Pump Failures, for a displaced jet pump mixer, core thermal power will ___(1)___ and flow in the other jet pump on the same riser will

___(2)___.

A. (1) decrease; (2) increase.

B. (1) remain constant; (2) decrease.

C. (1) remain constant; (2) increase.

D. (1) decrease; (2) decrease.

Answer: D Explanation:

A is wrong because flow in the other jet pump lowers. Plausible if applicant believes other jet pump flow will increase to compensate.

B is wrong because reactor power will decrease due to less core flow. Plausible if applicant believes that overall core flow remains unchanged.

C is wrong because see A & B.

D is correct because reactor power will decrease due to less core flow and the second jet pump flow will lower due to preferential flow through the failed jet pump.

Technical

References:

AOP-0062, Jet Pump Failures, Rev 002 References to be provided to applicants during exam: None Learning Objective: RLP-OPS-AOP-0062, Obj 4

Question Source:

(note changes; attach parent)

Bank #

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 295003 Partial or Complete Loss of AC Ability to operate and/or monitor the following as they apply to partial or complete loss of A.C. power: D.C. electrical distribution system Tier #

1 Group #

1 K/A #

AA1.04 Rating 3.6 Question 40 The reactor is operating at 100% power.

A station blackout occurs.

AC power restoration is expected to take 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The Station Blackout Diesel Generator should be aligned to 125 VDC bus ___(1)___,

via BYS-CHGR1D, 125 VDC BACKUP BATT CHGR, in accordance with AOP-50, Station Blackout.

The associated battery is designed to supply all SBO loads and still have enough capacity to perform switching operation to restore AC and DC power at the end of

___(2)___ hours.

A. (1) ENB-SWG01A (2) 4 B. (1) ENB-SWG01A (2) 8 C. (1) ENB-SWG01B (2) 4 D. (1) ENB-SWG01B (2) 8 Answer: A Explanation:

Per STM0305, (p. 6) Division I and II batteries have a rating of 2100 amp-hour. They are designed to supply all SBO loads or all loads not automatically tripped under LOCA conditions and still have enough capacity to perform switching operation to restore AC ad DC power at the end of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The Division III (HPCS) battery has a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rating of 825 amp-hours and is capable of continuously supplying the required Div. III DC loads for at least two hours.

A is correct see above explanation.

B is wrong. Wrong coping time C is wrong because AOP-50 directs the backup battery charger to be aligned to Div I ENB-SWG01A, not Div II.

D is wrong because AOP-50 directs the backup battery charger to be aligned to Div I ENB-SWG01A, not Div II.

Technical

References:

AOP-50 Rev 55, Station Blackout, Revision 55.

R-STM-0305 Rev 7, DC Distribution.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0541, Obj 4; RLP-STM-0305, Obj 3 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b) 8

Examination Outline Cross-Reference Level RO 295004 Partial or Total Loss of DC Pwr Ability to determine and/or interpret the following as they apply to partial or complete loss of D.C. power: Cause of partial or complete loss of D.C. power Tier #

1 Group #

1 K/A #

AA2.01 Rating 3.2 Revision 2

Question 41 Initially the ENB bus is being supplied by the backup charger.

A ___(1)___ causes a partial or complete loss of DC, because it automatically trips the

___(2)___.

A. LOP battery breaker B. LOCA battery breaker C. LOCA backup charger feeder breaker D. LOP backup charger feeder breaker Answer: C Explanation:

Per R-STM-305, Rev 7, Page 9 of 37, the backup charger feeder breakers to the standby DC buses automatically trip on a LOCA.

A is incorrect but plausible because of explanation above.

B is wrong but plausible because the battery breaker could be confused with the backup battery charger breaker.

C is correct. See explanation above.

D is wrong but plausible (LOCA not LOP is correct)

Technical

References:

R-STM-0305, DC Distribution, Revision 7 AOP-0014, Loss of 125VDC, Revision 24 SOP-0049, 125 VDC System (SYS #305), Revision 35 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0305, Obj 9.

Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross-Reference Level RO 295005 Main Turbine Generator Trip 2.1.19 Ability to use plant computers to evaluate system or component status.

Tier #

1 Group #

1 K/A #

2.1.19 Rating 3.9 Revision 1

Question 42 The plant is operating at 100% power when a turbine trip occurs. After 15 seconds from the time of the turbine trip, the following data is observed on the boards and Plant Process Computer (PPC):

• Condenser vacuum reads 26.5Hg

• The Exciter Field Breaker is closed

• Main Generator Gross MWe indicates 0 MWe (1) Immediately after the turbine trip, the turbine is _______ (motoring/generating).

(2) According to AOP-2, MAIN TURBINE AND GENERATOR TRIPS, 15 seconds after the turbine trip, the main generator output breakers should be

______(open/closed).

(1)

(2)

A. generating open B. generating closed C. motoring closed D. motoring open Answer: D. motoring, open Explanation:

A is wrong because it is motoring at this point. Open for part 2 is correct.

B is wrong because it is motoring at this point.. Also part 2 is incorrect because the breakers should be open not closed.

C is wrong because of part 2, closed is incorrect.

D is correct because the anti-motoring relay actuates with a 5 sec time delay, so 30 seconds into this is it motoring. The O/P breakers should be open at this point.

Technical

References:

AOP-0002, MAIN TURBINE AND GENERATOR TRIPS, Rev 27 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0521, Obj 3, 9; RLP-STM-0310, Obj 7.

Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(4)

Examination Outline Cross-Reference Level RO 295006 SCRAM Knowledge of the operational implications of the following concepts as they apply to SCRAM: AK1.01 Decay heat generation and removal Tier #

1 Group #

1 K/A #

AK1.01 Rating 3.7 Revision 1

Question 43 The plant has been operating at full power for 9 months when an electrical transient causes all of the turbine control valves to fully open.

The reactor automatically Scrams and no operator action is taken.

Five minutes following the Scram, decay heat is being removed by SRVs in___(1)____

Mode and __(2)___ are also available to assist in removing decay heat.

A. 1) Low-Low Set

2) RWCU System Components B. 1) Relief

2) RWCU System Components C. 1) Low-Low Set

2) Main Turbine Bypass Valves D. 1) Relief

2) Main Turbine Bypass Valves Answer: A Explanation:

A is correct. On a full open of all control valves the reactor scrams on low steam pressure and the MSIVs go closed. The bypass valves would not be available to control decay heat but RWCU is available to help with removing decay heat after MSIV closure.The SRVs (in auto) would initially control in relief mode but would transition to low low set mode, which makes A correct.

B is incorrect because SRVs would control in low-low set mode.

C is wrong because bypass valves are not available once MSIVs go closed.

D is correct because SRVs control in low-low set mode and bypass valves are not available Technical

References:

R-STM-0109, revision 15.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0109, Obj E Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 295016 Control Room Abandonment Knowledge of the interrelations between control room abandonment and the following:

Local control stations: Plant-Specific Tier #

1 Group #

1 K/A #

AK2.02 Rating 4.0 Question 44 A fire in the main control room is in progress and the CRS has ordered it to be abandoned.

What is required to be performed by the Reactor Building Operator in less than 10 minutes of scramming the reactor?

A. Locally initiate a full NSSS Isolation at the RPS MG sets and EPA Breakers.

B. Locally isolate all 10 Condensate Demineralizers.

C. Align for remote operation plant electrical and HVAC systems for plant cooldown D. Transfer power at E51-MOVF063 from Div. II to Div. I Alternate Power.

Answer: D Explanation:

A is wrong because this is what is required of the Unit Operator (UO) and to be accomplished within five minutes.

B is wrong because this is what is required of the Aux Control Room Operator and to be accomplished within three minutes.

C is wrong because this is what is required of the RB Operator after the first ten minutes..

D is correct.

Technical

References:

AOP-0031, Rev. 323, Shutdown From Outside the Main Control Room References to be provided to applicants during exam: None Learning Objective: RLP-OPS-AOP-0031, Obj 4.

Question Source:

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 295018 Partial or Total Loss of CCW Knowledge of the reasons for the following responses as they apply to partial or complete loss of component cooling water: Starting standby pump Tier #

1 Group #

1 K/A #

AK3.04 Rating 3.3 Revision 1

Question 45 CCP pressure is 94 psig and lowering due to a system malfunction.

Based on the above plant conditions the standby CCP pump will start to prevent CCP pressure from lowering to ___(1)___ where CCP Extreme Low Pressure isolation(s) for

___(2)___ will occur.

A. (1) 65 psig; (2) Div I and Div II ONLY B. (1) 56 psig; (2) Div I and Div II ONLY C. (1) 65 psig; (2) Div I ONLY D. (1) 56 psig; (2) Div I ONLY Answer: B Explanation:

A. Incorrect, CCW extreme low pressure is actuated by 56 psig CCW pressure B. Correct C. Incorrect, CCW extreme low pressure is actuated by 56 psig CCW pressure and both divisions will actuate at this pressure D. Incorrect, Both division of CCW extreme low pressure will actuate at the given pressure Technical

References:

AOP-0011 References to be provided to applicants during exam: None

Learning Objective: Objective 5 of RLP-OPS-AOP11.

Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295019 Partial or Total Loss of Inst. Air Ability to operate and/or monitor the following as they apply to partial or complete loss of instrument air: Backup air supply Tier #

1 Group #

1 K/A #

AA1.01 Rating 3.5 Question 46 The plant is at 100% power when pressure in the instrument air system begins to lower.

What is the pressure set-point where SAS-AOV134, Instrument Air Header Cross-Tie Valve, starts to open?

A. 120 psig B. 117 psig C. 114 psig D. 113 psig Answer: D Explanation: 113 psig is the pressure that opens the cross connect valve. All other values are associated with starting of system air compressors.

Technical

References:

R-STM-0121, Plant Air Systems, Rev. 16, p. 26 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0517, Obj 4; RLP-STM-0121, Obj 3 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295023 Refueling Accidents Ability to determine and/or interpret the following as they apply to refueling accidents:

Fuel Pool Level Tier #

1 Group #

1 K/A #

AA2.02 Rating 3.4 Question 47 During movement of irradiated fuel in the Spent Fuel Storage Pool, Storage Pool level starts to drop.

Which of the following is the Minimum Technical Specification level that is required to be maintained over the irradiated fuel assemblies in the Spent Fuel Storage Pool?

A. 21 B. 22 C. 23 D. 24 Answer: C Explanation:

A. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

B. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

C. CORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

D. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

Technical

References:

Technical Specification 3.7.6, Amendment 81 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0416, Objective 2

Question Source:

Bank # Columbia 2011-04 Exam, Question 10 X (2010 NRC)

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

Examination Outline Cross-Reference Level RO 295024 High Drywell Pressure Ability to interpret control room indications to verify status and operation of a system, and understand how operator actions and directives affect plant and system conditions Tier #

1 Group #

1 K/A #

2.2.44 Rating 4.2 Revision 1

Question 48 The plant is at 100% power and RPS channel B is in bypass for testing.

• CV71-PT-N050A (Drywell Pressure Transmitter) fails HIGH

• Annunciator P680-06A, NSSSS INT DRYWELL HIGH PRESSURE is in alarm What is the status of the RPS system?

What is the basis for the high drywell pressure reactor scram?

A. RPS has initiated a Half Scram; to ensure the Pressure Suppression function of the containment is maintained in the event Emergency Depressurization is required B. RPS has initiated a Full Scram; to ensure the Pressure Suppression function of the containment is maintained in the event Emergency Depressurization is required.

C. RPS has initiated a Half Scram; to minimize the possibility of fuel damage due to a reactor coolant pressure boundary leak by reducing the amount of energy being added to the coolant.

D. RPS has initiated a Full Scram; to minimize the possibility of fuel damage due to a reactor coolant pressure boundary leak by reducing the amount of energy being added to the coolant.

Answer: C Explanation:

A. Incorrect but plausible because a 1/2 Scram will occur however the pressure suppression function is based on RB pressure not drywell pressure.

B. Incorrect but plausible because the applicant could see that both trains of RPS have things affecting them but only a 1/2 scram will occur C. Correct

D. Incorrect but plausible because the reason for the high drywell scram is correct. A high drywell pressure condition results due to a leak of the primary system. Due to the loss of coolant, an inability to cool the fuel may result. A reactor scram occurs to minimize the energy being produced in the RPV. However in the stated conditions only a 1/2 scram will occur Technical

References:

R-STM-0508, Reactor Protection System References to be provided to applicants during exam: None Learning Objective: RLP-STM-0508 Obj. 2.

Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental H3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295025 High Reactor Pressure Knowledge of the operational implications of the following concepts as they apply to high reactor pressure: Pressure effects on reactor power Tier #

1 Group #

1 K/A #

EK1.01 Rating 3.9 Question 49 During an ATWS the following conditions exist:

• Reactor power is stable at 12%

• Reactor pressure is being controlled by the Main Turbine Bypass Valves at 920 psig.

• Reactor water level is being maintained at -100 inches How will reactor power be affected if the Main Turbine Bypass Valves failed shut?

Assume no operator actions taken.

A. Reactor power will rise due to a rise in coolant temperature caused by the rise in pressure.

B. Reactor power will be unchanged as the steam line drains control reactor pressure.

C. Reactor power will lower due to the negative reactivity effect of coolant temperature rising.

D. Reactor power will rise due to a reduction in voids caused by rising reactor pressure.

Answer: D Explanation:

A. Incorrect - A rise in coolant temperature would cause a reduction in power.

B. Incorrect - The stem indicates no action has been taken. Operation of the steam line drains requires manual action.

C. Incorrect - The negative reactivity due to voids has a larger affect the moderator temperature.

D. Correct - The closing of the BPVs will cause pressure to reduction in voids in the coolant, resulting in more moderation and higher power.

Technical

References:

R-STM-0509 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0509 Obj. 15a.

Question Source:

Bank #

2010 NRC (Q11)

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) 5

Examination Outline Cross-Reference Level RO 295026 Supp. Pool High Water Temp Knowledge of the interrelations between suppression pool high water temperature and the following: Containment pressure: Mark-III Tier #

1 Group #

1 K/A #

EK2.05 Rating 3.0 Question 50 During an accident, with RPV pressure at 800 psig, the Average Suppression Pool Temperature reaches the Heat Capacity Temperature Limit (HCTL) and is still increasing.

At this point the EOPs direct operators to Emergency Depressurize.

An ED at this point will prevent exceeding CERTAIN CONTAINMENT LIMITS.

Which of the following is the MOST LIMITING containment parameter for these conditions?

A. The Primary Containment Pressure Limit.

B. The Pressure Suppression Pressure Limit.

C. The Drywell Design Temperature Limit.

D. The Primary Containment Temperature Limit.

Answer: B Explanation:

A. Incorrect. The primary containment pressure limit is around 30 psig and the PSP limit is 3.2 to 5 psig and would be reached first for these conditions B. Correct - the PSP limit (variable between 3.2 to 5.5 psig SAFE region) is reached before any of the other variables and is the main concern for these conditions to be able to reject the energy from the RPV during an ED.

C. Incorrect DW Temperature is a concern but not the first concern reached for these conditions.

D. Incorrect, Containment temp is a concern but not the major concern for these conditions.

Technical

References:

EOP-2, Primary Containment Control, Revision 16 EOP bases document, Step CP-5, page B-8-14.

References to be provided to applicants during exam: None Learning Objective: HLO-0514 Obj 5 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b) 9

Examination Outline Cross-Reference Level RO 295027 High Containment Temperature Knowledge of the reasons for the following responses as they apply to high containment temperature (Mark III containment only):

Emergency depressurization: Mark-III Tier #

1 Group #

1 K/A #

EK3.01 Rating 3.7 Revision 1

Question 51 In accordance with EOP-2 and its associated bases for step CT-6, a containment temperature which cannot be maintained less than the design set-point requires Emergency Depressurization ______.

A. in order to extend equipment operability within containment for as long as possible during an event B. while the rate of energy transfer from the RPV to containment is less than the capacity of the containment vents during an event C. to reduce the driving head and flow of primary systems that are discharging into secondary containment D. to reduce the severity of an offsite radioactivity release Answer: A Explanation:

A is correct in accordance with EOP-2 and its bases, step CT-6, an ED is required at 185F to terminate, or reduce as much as possible, any continued containment temperature increase and thereby maintain equipment operability for as long as possible.

B is incorrect because these words are part of the basis for step SPT-6 to ED when HCTL is exceeded.

C is incorrect because these words are part of the basis for step SC-16 in EOP-3 D is incorrect because these words are part of the basis for step RR-6 in EOP-3.

Technical

References:

EOP-2 and EOP-2 bases, step CT-6, page B-8-9, Revision 17 References to be provided to applicants during exam: None Learning Objective: HLO-514 Obj. 5

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

X (2008 audit exam)

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)9

Examination Outline Cross-Reference Level RO 295028 High Drywell Temperature Ability to operate and/or monitor the following as they apply to high drywell temperature:

Drywell ventilation system Tier #

1 Group #

1 K/A #

EA1.02 Rating 3.9 Rev 1 Question 52

• The plant is operating at 100% power

• The control switches for drywell unit coolers DRS-UC1A through -UC1D are in RUN

• Temperature controllers DRS-H/A52A through -H/A52D are in AUTO

• Plant Engineering and responsible System Engineers have been briefed on current plant configuration

• On H13-P808, CMS-TR41A and -TR41B, DRYWELL ATMOS TEMP RECORDERs read 110F and have been stable for days (1) Per SOP-0060, Drywell Cooling, what are the minimum permissible number of drywell unit coolers that can be running for these conditions?

A.

2 B.

3 C.

4 D.

5 Answer: C Explanation:

A is wrong (see C)

B is wrong (see C)

C is correct. Procedure SOP-0060, Drywell Cooling (SYS #404), Section 4.1 contains a Caution and a Note which provides stipulations on how many drywell cooling units can be run in different situations. Per the Notes, it is permissible to run four drywell unit coolers, provided that the average drywell temperature is less than or equal to 1450F, and Plant Engineering and the System Engineer are notified. Therefore, 4 is correct. 1450F.

D is wrong (see C)

Technical

References:

SOP-0060, Drywell Cooling (SYS #404), Revision 10 EOP-2, Primary Containment Control, Revision 16 System Training Manual R-STM-0057, Primary Containment and Auxiliaries, Revision 5 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0057, Obj 24; RLP-STM-0403, Obj 9 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis F2 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295030 Low Suppression Pool Wtr Lvl Ability to determine and/or interpret the following as they apply to low suppression pool water level: EA2.03 Reactor pressure Tier #

1 Group #

1 K/A #

EA2.03 Rating 3.7 Question 53 An ATWS has occurred following an inadvertent MSIV isolation.

Which Reactor pressure band is appropriate for the given suppression pool levels and temperatures?

Reactor Pressure Band Supp Pool Level Supp Pool Temp A.

500 - 700 psig 15 4 130F B.

800 - 1090 psig 16 11 128F C.

800 - 1090 psig 19 5 140F D.

500 - 700 psig 21 3 150F Answer: A Explanation:

A is correct because it is the only choice that allows full use of the pressure band without exceeding the HCTL curve, therefore it is correct.

B is wrong because its band falls within the unsafe region (exceeds the HCTL curve)

C is wrong because its band falls within the unsafe region (exceeds the HCTL curve)

D is correct because its band falls within the unsafe region (exceeds the HCTL curve)

Technical

References:

EOP-0001 Figure 2 HCTL curve, revision 27.

References to be provided to applicants during exam: EOP-0001 Figure 2 Learning Objective: RLP-HLO-517 Obj 2 Question Source:

Bank #

X (2010 NRC)

(note changes; attach parent)

Modified Bank #

New

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 295031 Reactor Low Water Level Ability to determine operability and/or availability of safety related equipment.

Tier #

1 Group #

1 K/A #

2.2.37 Rating 3.6 Question 54

• An electrical fault occurs which requires removing power to power supply EHS-MCC2J for repairs.

• A LOCA occurs during repair activities, requiring implementation of EOP-1, RPV Control Which of the Preferred Injection Systems, listed in Table L-1, would not be available?

A. RHR B Train in LPCI Mode B. LPCS C. HPCS D. RHR C Train in LPCI Mode Answer: B Explanation:

A is wrong because the source bus for RHR pump B and its related MCC loads is ENS-SWG1B.

B is correct because power supply EHS-MCC2J (source bus ENS-SWG1A) powers E21-MOVF005, LPCS pump discharge valve. This valve is normally closed, so electrical power is needed to open it to provide injection when if reactor vessel level reaches Level 1 (-143).

C is wrong because the source bus for HPCS and its MCC loads is source bus SWGR E22-S004.

D is wrong because the source bus for RHR pump C and its related MCC loads is ENS-SWG1B.

Technical

References:

R-STM-0205, Low Pressure Core Spray System, Revision 6 R-STM-0204, Residual Heat Removal System, Revision 12 R-STM-0203, High Pressure Core Spray System (HPCS), Revision 8 R-STM-0300, AC Distribution, Revision 28

EOP-1, RPV Control, River Bend Station, Revision 27 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0205, Obj 17.

Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Knowledge of the operational implications of the following concepts as they apply to SCRAM condition present and reactor power above APRM downscale or unknown:

Shutdown margin Tier #

1 Group #

1 K/A #

EK1.07 Rating 3.4 Question 55 The unit was at full power when an event occurred that required a SCRAM and it did not occur.

As the ATC you placed the mode switch in shutdown, pushed the reactor SCRAM pushbuttons, and initiated ARI.

Current plant conditions are:

• Reactor power is 10%

• Several rods are not fully inserted The operational implication of this event because of the lack of adequate shutdown margin is that the ___(1)___ must be controlled differently than EOP-1, RPV Control, strategies with respect to shutdown margin in order to prevent damage to the

___(2)___.

A. (1) Power and Level ONLY (2) core and the RPV B. (1) Power and Level ONLY (2) core only C. (1) Power, Pressure, and Level (2) core and the RPV D. (1) Power, Pressure, and Level (2) core only Answer: C

Explanation:

A. Incorrect because all three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage to the RPV and core.

B. Incorrect because all three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage. Also the second part of this distracter does not include the RPV and is can be potentially damaged due to this event if not managed in accordance with EOP-1A.

C. Correct. All three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage to the RPV and core.

D. Incorrect First part is correct but the second part of this distracter does not include the RPV and is can be potentially damaged due to this event if not managed in accordance with EOP-1A.

Technical

References:

EPSTG-0002, page B-6-5 and B-6-6, B-5-14, revision 17.

References to be provided to applicants during exam: None Learning Objective: HLO-0513, Objective 4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)8 Comments:

Examination Outline Cross-Reference Level RO 295038 High Off-site Release Rate Knowledge of the interrelations between high off-site release rate and the following: Site emergency plan Tier #

1 Group #

1 K/A #

EK2.05 Rating 3.7 Revision 1

Question 56 The MINIMUM required EOP-3 entry category for the emergency plan for a high off-site release rate is ____(1)____ and the associated instrument used for this determination is ____(2)_____.

A. 1) ALERT

2) RMS-RE125, MAIN PLANT EXHAUST B. 1) ALERT

2) RMS-RE16, CONTAINMENT POST ACCIDENT MONITOR C. 1) SITE AREA EMERGENCY

2) RMS-RE125, MAIN PLANT EXHAUST D. 1) SITE AREA EMERGENCY

2) RMS-RE16, CONTAINMENT POST ACCIDENT MONITOR Answer: A Explanation:

A is correct. EOP-3 entry is for minimum ALERT, and the instrument used is RMS-125.

B is wrong because the second part is wrong (wrong instrument)

C is wrong because the first part is wrong, it is at Alert level, not SAE.

D is correct because both parts are wrong.

Technical

References:

EOP-3, revision 17, EIP-2-001, revision 26.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0515 Obj 2 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F4 Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)11

Examination Outline Cross-Reference Level RO 600000 Plant Fire On Site Knowledge of the reasons for the following responses as they apply to plant fire on site:

Actions contained in the abnormal procedure for plant fire on site Tier #

1 Group #

1 K/A #

AK3.04 Rating 2.8 Revision 2

Question 57 There is a fire in the D Tunnel.

The CRS has directed the crew to enter AOP-0052, FIRE OUTSIDE THE MAIN CONTROL ROOM IN AREAS CONTAINING SAFETY RELATED EQUIPMENT.

Section 5 of this procedure then directs AOP-020 be used to place RHR B in alternate shutdown cooling mode.

Per AOP-0052, Step 5.7.1, Division ____(1)___ SRVs are to be used in this situation.

The reason this division of SRVs is used is ___(2)___.

A. (1) I (2) to ensure that at least eight SRVs are available when control air may be lost for long-term operations B. (1) II (2) because they are the most reliable when considering the equipment design aspects for a fire outside the control room C. (1) I (2) because they are the most reliable when considering the equipment design aspects for a fire outside the control room D. (1) II (2) to ensure that at least eight SRVs are available when control air may be lost for long-term operations Answer: D

Explanation:

A is wrong. (a) The wrong division of SRVs is provided. See AOP-0052, Section 5.6, sixth bullet. (b) The answer for this part is correct.

B is wrong. (a) The correct division of SRVs is provided. (b) This is incorrect but plausible because the design includes SRVs for RSP operation (this is Div 1 equipment though).

Someone who confuses the design aspects of RSP or forgets which division would pick this distracter.

C is wrong because both answers are incorrect. See Explanations for answer A and B.

D is correct. (a) Per AOP-0052, Section 5.6, sixth bullet, in cases where RHR will be used in alternate shutdown cooling mode, it itemizes which SRVs are to be used. The list itemizes each division. For use of RHR B, Division II SRVs are listed. They are available for RPV pressure control, as confirmed on Attachment 1 of AOP-0052. (b) The Note prior to the sixth bullet explains why this action is taken. Fires in various areas can affect the long-term SRV control air supply (compressed air supply MOV spurious closures). This ensures that there are 8 SRVs available in this condition.

Technical

References:

AOP-0052, Fire Outside the Main Control Room, Revision 25 OSP-0019, Electrical Bus Outages, Revision 308 System Training Manual R-STM-0109, Main Steam System, Revision 15 SEP-FPP-RBS-002, River Bend Station Fire Fighting Procedure, Revision 2 EPSTG*0002, Revision 17 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0544, Obj 2, 3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) (5)

Examination Outline Cross-Reference Level RO 700000 Generator Voltage and Electric Grid Disturbances Ability to operate and/or monitor the following as they apply to generator voltage and electric grid disturbances: Turbine/generator controls Tier #

1 Group #

1 K/A #

AA1.02 Rating 3.8 Rev 3 Question 58 The Reactor is operating at 100% power and all Turbine Generator controls are in automatic.

P808-86A-H01, GRID TROUBLE is in alarm and AOP-0064, Degraded Grid has been entered.

The Shift Manager receives a report from the SOC that Grid voltage is low at 223 KV and requests that we attempt to maintain grid voltage between 224.25 KV and 242 KV.

The CRS directs you to attempt to adjust voltage by adjusting ___(1)___ on the Main Generator; the emergency limit that you are trying to stay within is the emergency

___(2)___.

A. 1) MVARs

2) VARS limit of - 60 MVARS to +359 MVARS B. 1) MVARs

2) VARS limit of 0 MVARS to +230 MVARS C. 1) Voltage

2) Voltage limit of 224 KV to 245 KV D. 1) Voltage

2) Voltage limit of 225.86 KV to 242 KV Answer: A A. Correct, With the TG VR in auto, MVARS are what you adjust to attempt to control voltage per step 5.4 of the AOP. The emergency limit on MVARS is -60 to +359 MVARS per the note directly above this step in the procedure.

B. Incorrect - first part is correct but the second part is the normal limit per SOP-0060 which is not the emergency limit and is what is required for these conditions in the stem of the question and what is asked for ini the stem of the question (ie the emergency limit).

C. Incorrect, voltage is not adjusted on the TG for these conditions unless in manual, which is not given in the stem. Also the second part is wrong as well. This is a plausible voltage range based on values given in the AOP.

D. Incorrect, voltage is not adjusted on the TG for these conditions unless in manual, which is not given in the stem. Also the second part is wrong as well. This is a plausible voltage range based on values given in the AOP.

Technical

References:

AOP-0064, Degraded Grid References to be provided to applicants during exam: None Learning Objective: RLP-OPS-AOP-0064, Obj 5 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) 10

Examination Outline Cross-Reference Level RO 295002 Loss of Main Condenser Vac Knowledge of the reasons for the following responses as they apply to loss of main condenser vacuum: Main steam isolation valve: Plant-Specific Tier #

1 Group #

2 K/A #

AK3.05 Rating 3.4 Question 59 The reactor plant is operating at 100% power. Changes in the plant result in the following indications:

• Main steam line pressure is 1085 psig

• Reactor vessel level is 30

• Main condenser vacuum is 8 Hg

• Main steam line tunnel temperature is 160F Based on these conditions, the Main Steam Isolation Valves (MSIVs) automatically close. In this case, what is the reason for this protective action?

A. To isolate a leak in the main steam lines B. To prevent a rapid depressurization of the reactor pressure vessel C. To isolate or prevent a leak in the main turbine D. To isolate or prevent a leak in the main condenser Answer: D Explanation:

A is wrong because this is the reason for an automatic closure of MSIVs for main steam line tunnel temperature. The set point for this is 1730F.

B is wrong because this is one of the reasons for an automatic MSIV closure for main steam line low pressure. The set point for this is 849 psig.

C is wrong but plausible because the main turbine has a low vacuum trip (22.3 Hg) to protect the main turbine. The MSIVs automatically close to protect the main condenser.

D is correct because main condenser vacuum 8.5 Hg automatically closes the MSIVs. The reason for this automatic capability is to protect against leaks in the main condenser.

Technical

References:

R-STM-0109, Main Steam System, Revision 15 R-STM-0051, Nuclear Boiler Instrumentation, Revision 6 R-STM-0058, Containment and Reactor Vessel Isolation Control System (CRVICS), Rev 10 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0109, Obj 22 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 295007 High Reactor Pressure Ability to operate and/or monitor the following as they apply to high reactor pressure:

AA1.04 Safety/relief valve operation: Plant-Specific Tier #

1 Group #

2 K/A #

AA1.04 Rating 3.9 Question 60 The reactor is at 100% power when reactor pressure begins to rise to 1140 psig.

Following the initial opening of SRV F051D you would expect to observe F051D cycling between ___(1)___ psig and ___(2)___ psig.

A. (1) 956 (2) 1063 B. (1) 956 (2) 1103 C. (1) 966 (2) 1063 D. (1) 966 (2) 1103 Answer: A Explanation:

After the initial opening of the SRV the low-low set mode of operation would be active.

F051D is the first SRV to open and, due to low-low set, would cycle between 956 psig and 1063 psig. The 1140 psig in the stem is lower than the setpoint for the second SRV (F051C) to open. The other values (966 and 1103) are the low-low set values for F051C.

A is correct - see above explanation B is incorrect because of part 2s pressure is the closed sp for F051C not F051D C is incorrect because both setpoints are incorrect D is incorrect because both setpoints are incorrect Technical

References:

R-STM-0109, Main Steam System, Rev 15 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0109, Obj 4.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(7)

Examination Outline Cross-Reference Level RO 295012 High Drywell Temperature Ability to determine and/or interpret the following as they apply to high drywell temperature: AA2.02 Drywell pressure Tier #

1 Group #

2 K/A #

AA2.02 Rating 3.9 Revision 1

Question 61 The following conditions exist:

• Drywell cooling has been lost

• Drywell temperature is rising

• Efforts are being made to restore Drywell cooling (1) Drywell pressure will ___(1)___;

(2) EOPs would be entered first on ___(2)___.

A. (1) rise (2) 1.68 psid Drywell pressure B. (1) rise (2) 145F Drywell temperature C. (1) remain the same until Drywell temperature reaches 212F (2) 145F Drywell temperature D. (1) remain the same until Drywell temperature reaches 212F (2) 1.68 psid Drywell pressure Answer: B Explanation:

A is wrong because.... see B. Part 2 is plausible because during a leak in the drywell, the EOPs will be entered on drywell pressure before drywell temperature.

B is correct because as drywell temperature is rising drywell pressure will also slowly rise.

Due to the size of the drywell and the smaller margin to maximum drywell temperature, the

EOPs will be entered on Drywell Temperature first. Normal operating drywell temperature is approximately 120F.

C is wrong because of B and the drywell pressure will slowly rise before temperature reaches 212F.

D is correct because see B and C.

Technical

References:

EOP-2 Basis for step PCC-1; References to be provided to applicants during exam: None Learning Objective: RLP-STM-0403, Obj 16 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(8)

Examination Outline Cross-Reference Level RO 295015 Incomplete SCRAM / 1 Ability to identify post-accident instrumentation.

Tier #

1 Group #

2 K/A #

2.4.3 Rating 3.7 Revision 1

Question 62 The plant has experienced an ATWS condition are you are asked to verify reactor water level with a Post-Accident Monitoring (PAM) instrument by the CRS prior to lowering level for power control.

Which of the following reactor water level instruments is appropriate for this request and how can you tell if it is a PAM instrument?

A. 1) Wide range recorder B21 LR/PR-R623B on panel P601

2) red label with Post-Accident Monitor letters in white or black B. 1) Fuel zone instrument B21 LI-R610 on panel P680

2) red label with Post-Accident Monitor letters in white or black C. 1) Fuel zone instrument B21 LI-R610 on panel P680

2) tan label with Post-Accident Monitor letters in white or black D. 1) Wide range recorder B21 LR/PR-R623B on panel P601

2) tan label with Post-Accident Monitor letters in white or black Answer: D Explanation:

A is wrong because the label is tan not red. The second part (the instrument is correct)

B is wrong because of both aspects. The label is tan not red and this fuel zone instrument is not PAM and is also not located on the P680 panel.

C is wrong because this fuel zone instrument is not PAM and is also not located on the P680 panel.

D is correct because it is tan and this is the correct instrument and location.

Technical

References:

ADM-0037, rev 17, page 12, R-STM-0051 Rev 6, page 39.

References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0051 Obj B3-B6.

Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(2)

Examination Outline Cross-Reference Level RO 295017 High Off-site Release Rate Knowledge of the operational implications of the following concepts as they apply to high off-site release rate: Protection of the general public.

Tier #

1 Group #

2 K/A #

AK1.02 Rating 3.8 Question 63 Due to a steam leak, the Main Steam Line Tunnel area temperatures are all between 160°F and 170°F.

All automatic isolations have occurred as designed.

Because of the leak location and isolation actions, NO LOCA signal occurred from high drywell pressure or RPV low level.

An ALERT has been declared based on offsite release rates.

Which one of the following will reduce the UNMONITORED release rate?

A. Shutdown the Turbine Building Ventilation System if operating.

B. Shutdown the Radwaste Building Ventilation System if operating.

C. Start the Turbine Building Ventilation System if NOT operating.

D. Start the Fuel Building Charcoal Filtration trains if NOT operating.

Answer: C Explanation:

A is wrong per RR-1 of EOP-3 for high dose rate associated with ALERT or higher, the turbine building vent system must be restarted if not running, so C is correct and the others are wrong.

B is wrong (see A above)

C is correct because this is the directed ventilation system and it is directed to restart it if not running. This ties to the KA because reducing the rate reduces the dose to the public, which protects them.

D is wrong (see A above).

Technical

References:

EOP-3, RR-2, revision 17 EPSTG-2, Page B-10-4, revision 17 References to be provided to applicants during exam: None Learning Objective: HLO-515 OBJ-6 Question Source:

Bank #

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2003 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)13

Examination Outline Cross-Reference Level RO 295029 High Suppression Pool Wtr Lvl Knowledge of the interrelations between high suppression pool water level and the following:

RCIC: Plant-Specific Tier #

1 Group #

2 K/A #

EK2.09 Rating 3.1 Rev 2 Question 64 RCIC has been initiated after a loss of feed transient. It is currently maintaining level between -10 and +51 inches. The following indications are observed in the control room:

• Suppression Pool Level High Alarm is in on P808

• Suppression Pool level indicates 20 1 on ERIS When the Suppression Pool Level reaches ___(1)___ the RCIC system will ___(2)___.

A. (1) 20 3.5 (2) trip the RCIC turbine on High Suction Pressure B. (1) 21 3 (2) automatically align RCIC suction source to the Suppression Pool C. (1) 21 3 (2) trip the RCIC turbine on High Suction Pressure D. (1) 20 3.5 (2) automatically align RCIC suction source to the Suppression Pool Answer: D Explanation:

A is wrong because a low suction pressure trips the turbine, not a high suction pressure; a high pressure only gives an alarm..

B is wrong because operator action is required at 21 3 to perform an Emergency Depressurization which would trip the RCIC turbine (at 60 psig RPV pressure).

C is wrong because operator action is required at 21 3 to perform an Emergency Depressurization which would trip the RCIC turbine (at 60 psig RPV pressure) and there is not high suction pressure trip for RCIC turbine.

D is correct because 20 3.5 is the correct suppression pool level that is not tied to the ED (not given in stem that ED occurred or is needed) where auto actions occur as given and for this level RCIC automatically re-aligns its suction source to the suppression pool.

Technical

References:

ARP-P808-84A-G02, and P601-21A-C05 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0209, Obj 5.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(5)

Examination Outline Cross-Reference Level RO 295032 High Secondary Containment Area Temperature Knowledge of the reasons for the following responses as they apply to high secondary containment area temperature: Isolating affected systems Tier #

1 Group #

2 K/A #

EK3.03 Rating 3.8 Revision 1

Question 65 You are performing a RCIC surveillance when an alarm indicating a room high temperature is received.

Temperature readings are as follows:

• RCIC Equipment Room Ambient Temperature is 184°F rising

• RHR/RCIC Area Temperature is 124 degrees and rising

• RHR Equipment Room Ambient Temperature is 110°F and rising

• RHR area Temperature is 108 degrees and rising Based on the indications provided what group isolation signals have been generated and why?

Group 2:

E51 F031 (RCIC PUMP SUP PL SUCTION VALVE)

E51 F064 (RCIC STEAM SUPPLY OUTBD ISOL VALVE)

Group 5:

E12 F008 (RHR SHUTDOWN COOLING OUTBD ISOL VALVE)

E12 F053A (RHR PUMP A SDC INJECTION VALVE)

Group 14:

E12 F037A (RHR A TO UPPER POOL FPC ASSIST)

E12 F075A (RHR A HX DN STREAM SAMPLE VLV)

Group 17:

RHS-AOV63 (SPC SUCTION VALVE)

RHS-AOV64 (SPC DISCHARGE VALVE)

A. Group 2 valves only; solely to reduce the potential of off-site release B. Groups 2 valves only; because this allows selected valves in those systems essential to mitigate the effects of an event to remain open or move to their open position

C. Groups 5, 14 and 17 only; because this allows selected valves in those systems essential to mitigate the effects of an event to remain open or move to their open position D. Groups 5, 14 and 17 only; solely to reduce the potential of off-site release Answer: B Explanation:

A. Incorrect, because high RCIC equipment room temp of 182 degrees will result in group 2 valves isolating only, the first part is correct, however, the second part is incorrect. the distractor is plausible because the high temperature and the reduction in the potential for off-site release is a function of containment isolation signals but it is not the sole reason.

B. Correct C. Incorrect, because only the RCIC room ambient temperature has exceeded the actuation set point of 182 degrees which will isolate group 2 valves only. The distractor is plausible because if RHR Area Temperature was > 130.9 degrees you would get Groups 5,14 and 17 valve isolations D. Incorrect, because only the RCIC room ambient temperature has exceeded the actuation set point for Group 2 isolations if the RHR/RCIC Area high temperature had exceeded 130.9 or the RHR Equipment Room Ambient temperature exceeded 117 degrees groups 2, 5, 14 anbd17 valves would have received an isolation signal.

Technical

References:

AOP-0003 Rev 34 page References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0522, AOP 0003, Automatic Isolations, Objective C Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b) 7

Examination Outline Cross-Reference Level RO Knowledge of shift or short-term relief turnover practices.

Tier #

3 Group #

K/A #

2.1.3 Rating 3.7 Revision 1

Question 66 You are the oncoming ATC. Which of the following MUST be completed PRIOR to assuming the shift?

A. Review LCO/Tracking LCO Logs B. Review Night/Standing Orders C. Back Panel Walkdown D. Annunciator Status Review Answer: B Explanation:

A. INCORRECT. This is to be completed as early in the shift as possible.

B. CORRECT. According to Attachment 3, this must be completed prior to assuming the shift.

C. INCORRECT. This is to be completed as early in the shift as possible.

D. INCORRECT. This is to be completed as early in the shift as possible.

Technical

References:

OSP-002, Shift Relief and Turnover, Rev 47 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0206, Objective D Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No

Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO Conduct of Operations 2.1.39 Knowledge of conservative decision making practices. (CFR: 41.10 / 43.5 / 45.12)

Tier #

3 Group #

K/A #

2.1.39 Rating 3.6 Revision 1

Question 67 In accordance with EN-OP-115, Conduct of Operations, which of the following attributes are required to be followed when the control room is faced with a time-critical decision?

1.

Suspend the use of peer checking

2.

Take time to validate available information

3.

Contact off-site personnel, if necessary

4.

Do not allow uncertainty to prevent timely action A. 2 only B. 1 and 4 only C. 2 and 3 only D. 1, 3 and 4 only Answer: C Explanation:

Based on EN-OP-115 Conduct of Operations which describes the thought process utilized for making conservative decisions. The procedure covers a list of attributes that are to be followed when faced with a time critical decision.

A. Is wrong because contacting offsite personnel is also included B. Is wrong because peer checking is not suspended and action should not be taken in the face of uncertainty C. Correct - validating information and contacting offsite help if necessary are listed D. Is wrong because peer checking is not suspended and action should not be taken in the face of uncertainty Technical

References:

EN-OP-115, Rev 17, page 22.

References to be provided to applicants during exam: None Learning Objective: HLO-0206 Obj U.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41(b)(10)

Examination Outline Cross-Reference Level RO Ability to identify and interpret diverse indications to validate the response of another indication Tier #

3 Group #

K/A #

2.1.45 Rating 4.1 Revision 1

Question 68 The ATC needs to verify a board indication for a given parameter with ERIS.

The ERIS parameter is bordered in CYAN (light blue).

This parameter should be considered _____.

A. bad data.

B. validated.

C. at the ALERT setpoint.

D. at the ALARM setpoint.

Answer: B Explanation:

A. Incorrect, plausible because magenta - In all cases, this color is used to indicate bad data (i.e., not measured, out of range, DAS failure, etc.). (Page 11)

B. Correct. Cyan - This color is used in conjunction with parameter validation for the control parameters on SPDS. When a control parameter is fully validated, its digital value appears in a window bordered in cyan. (Page 11)

C. Incorrect. RTAD displays messages in a border. RPV Alert and CONTMT Alert border colors are Yellow. (Page 10)

D. Incorrect. RTAD displays messages in a border. RPV Alarm and CONTMT Alarm border colors are Red. (Page 11)

Technical

References:

STM-0514, rev 11, page 11 References to be provided to applicants during exam: None

Learning Objective: RLP-STM-0514, Rev 1 Objective 3 and 10 Question Source:

Bank #

X (RBS OPS)

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO Equipment Control Ability to manipulate console controls as requird to operate the facility between shutdown and power levels Tier #

3 Group #

K/A #

2.2.2 Rating 4.6 Question 69 When operating a motor-operated throttle valve from the control room, the hand switch should be held in the ___(1)___ because ___(2)___.

A. (1) closed position for at least 5 seconds after the red light extinguishes (2) the red light is set to extinguish before the valve reaches its closed seat B. (1) open position for at least 3 seconds after the green light extinguishes (2) the green light is set to extinguish before the valve starts to come off its open seat C. (1) closed position for at least 3 seconds after the red light extinguishes (2) the red light is set to extinguish before the valve reaches its closed seat D. (1) open position for at least 5 seconds after the green light extinguishes (2) the green light is set to extinguish before the valve starts to come off its open seat Answer: A Explanation:

A is correct per OSP-0022, page 11.

B is incorrect because the time is wrong (it is 5 sec not 3 sec) and also it is for the red light and its closed seat not the green light and the open seat.

C is incorrect because the time is wrong (it is 5 sec not 3 sec) the second part is correct.

D is incorrect because of the second part, it is set to the red light and the closed seat but the time of 5 sec is correct.

Technical

References:

OSP-0022, rev 86, page 11.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0211, Obj 2 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO Equipment Control Knowledge of the process used to track inoperable alarms Tier #

3 Group #

K/A #

2.2.43 Rating 3.0 Revision 1

Question 70 An annunciator has been frequently alarming in the Main Control Room. The cause was discovered to be that the alarm is coming in due to an out of calibration instrument.

Based on the guidance in OSP-0015, Problem Annunciator Resolution Program, the Shift Manager has determined that the annunciator should be flagged with ______.

A. an orange dot, indicating a Disabled Annunciator - Partial.

B. a yellow dot, indicating a Nuisance Alarm.

C. a green dot, indicating an Out of Service Annunciator.

D. a pink dot, indicating a Disabled Annunciator - Full.

Answer: C Explanation:

A is incorrect - A Disabled Annunciator - Partial is used for partially disabling an input by lifting a single lead or bypassing a point B is Incorrect - There is not a yellow dot used in the control room at RBS. Alslo a nuisance alarm is one that is distracting from safe operation of the plant, but has not been disabled yet C Correct - An annunciator that is not functioning properly due to (1) an alarm without a valid condition, (2) An alarm that fails to alarm with a valid condition, or (3) an alarm condition due to an out of cal instrument D is incorrect-A Disabled Annunciator - Full is used for fully disabling all inputs by pulling the annunciator card Technical

References:

OSP-0015, Rev 307, section 4.7

References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible Question Source:

Bank #

2008 Audit Q71 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO Knowledge of the process for controlling equipment configuration or status.

Tier #

3 Group #

K/A #

2.2.14 Rating 3.9 Revision 1

Question 71 A pump in the plant has developed a minor seal leak found during operator rounds. The operator secured the pump and then started the standby pump as directed.

Which of the following describes the appropriate control method for the status of the secured pump?

A. Place a Test and Maintenance Tag on the switch to allow maintenance activities.

B. Place a Lockout Device on the switch to prevent use of this equipment.

C. Place a Danger Tag on the switch to prevent use of this equipment.

D. Place a Caution Tag on the switch that documents only use under the direction of the OSM/CRS.

Answer: D Explanation:

A. INCORRECT. This answer is plausible because a T&M tag would be used for a component that would need to be manipulated during maintenance activities, such as determining the leak location, but is incorrect because a T&M tag permits operation of the equipment by only authorized persons signed on to the tagout.

B. INCORRECT. This answer is plausible because during maintenance activities a lockout device would be used however the device would be installed on the pump breaker.

C. INCORRECT. This answer is plausible because the Danger hold tag would be used to place the equipment in a safe condition if maintenance were to be performed. When maintenance is to be performed, the Caution tag would be replaced by a Danger tag.

D. CORRECT. Caution tags provide precautions or special instructions that relate to unusual or out-of-normal conditions.

Technical

References:

EN-OP-102, Revision 18, page 5 and 60 References to be provided to applicants during exam: None

Learning Objective: RLP-HLO-0216 Objective G Question Source:

Bank # 2015 NRC X

(note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam Yes Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis F3 10CFR Part 55 Content:

55.41.10

Examination Outline Cross-Reference Level RO Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions.

Tier #

3 Group #

K/A #

2.3.4 Rating 3.2 Revision 1

Question 72 Which of the following exposure limits are Entergy's Routine Annual Administrative Guidelines?

TEDE: Total Effective Dose Equivalent SDE: Shallow Dose Equivalent WB: Whole Body LDE: Lens Done Equivalent TEDE (mrem per year)

SDE, WB (rem)

LDE (rem)

A.

2000 40 12 B.

5000 40 12 C.

5000 50 15 D.

2000 50 15 Answer: A Explanation:

A. is correct, the limits are Entergy Routine Annual Administrative Guidelines B. Is incorrect, Maximum Annual Administrative guidelines for Entergy C. is incorrect, they are Annual Regulatory limits D. is incorrect, they are a mix of different limits Technical

References:

EN-RP-201 References to be provided to applicants during exam: None Learning Objective: Rad Worker Objectives.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(12)

Examination Outline Cross-Reference Level RO Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Tier #

3 Group #

K/A #

2.3.12 Rating 3.2 Question 73 In accordance with TS 5.7, High Radiation Area:

The lowest area radiation level where locked or continuously guarded doors are required to prevent unauthorized entry is _____.

Entry into a locked high radiation area is permitted under an approved RWP that shall specify the dose rate levels in the immediate work areas and _____

individuals in those areas.

A. 1000 mrem/hr; periodic RP surveillance of B. 500 mrem/hr; periodic RP surveillance of C. 500 mrem/hr; maximum allowable stay times for D. 1000 mrem/hr; maximum allowable stay times for Answer: D Explanation:

A is wrong because periodic surveillance is incorrect. See B.

B is wrong because 1000 mrem/hr is the lowest radiation level that requires locked/guarded doors. Plausible if applicant confuses 500 rad/hr limit for VHRA. Continuous surveillance is required in lieu of max stay times. Plausible if applicant considers periodic surveillance to be adequate.

C is wrong because 500 mrem/hr is incorrect. See B.

D is correct because 1000 mrem is the lowest level requiring. Max stay times designated in the RWP meets the requirement.

Technical

References:

TS 5.7.2 References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)(12)

Examination Outline Cross-Reference Level RO Emergency Procedures / Plan Knowledge of general operating crew responsibilities during emergency operations.

Tier #

3 Group #

K/A #

2.4.12 Rating 4.0 Revision 1

Question 74 In accordance with OSP-0053, Emergency and Transient Response, which of the following EOP procedure steps is considered explicit enough that the skill of the operator alone is all that is required to complete it without hard cards or other procedure guidance?

A. Injecting with Standby Liquid Control during ATWS.

B. Vent the Scram air header C. Inject with ECCS systems which are running but not injecting D. Emergency vent containment Answer: C Explanation:

A is incorrect. This is not listed in the OSP and has a hard card for it.

B is incorrect. This is not listed in the OSP and has an enclosure for it (Encl 11).

C is correct. This is the only item specifically listed in OSP-0053 that does not require other procedure guidance to complete.

D is incorrect. This is not listed in the OSP and has an enclosure (Encl. 21) for it.

Technical

References:

OSP-0053, revision 23, page 6.

References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Tier #

3 Group #

K/A #

2.4.21 Rating 4.0 Question 75 EOP-1 contains Graph 5, the Hydrogen Deflagration Overpressure Limit (HDOL) curve.

For this curve, the maximum allowed hydrogen concentration in percent (%) decreases as containment pressure rises.

Which one of the following is the reason for this relationship?

A. This ensures that a hydrogen deflagration at the limit combined with current pressure will not exceed containment overpressure failure limits.

B. As containment pressure rises, the capability of the hydrogen recombiners to remove hydrogen is reduced.

C. As containment pressure rises, the containment hydrogen analyzer system response time is adversely affected.

D. The deflagration pressure of hydrogen drops with increasing containment pressure and therefore requires a lower concentration of allowed hydrogen Answer: A Explanation:

A. CORRECT. According to the table in EPSTG-2, the combined pressure of containment and hydrogen deflagration pressures could exceed the design pressure of containment.

B. INCORRECT. This is not correct and is not the reason for the EOP curve for H2 C. INCORRECT. The system response is independent of H2 concentration or containment pressure D. INCORRECT. This is not correct. If deflagration pressure goes up with increasing H2 concentration then that might be correct but is not the reason for the curve-EPSTG-2.

Technical

References:

EOP-1 revision 27, EPSTG-2, revision 17, Page A-15-17.

References to be provided to applicants during exam: None.

Learning Objective: HLO-514, Objective 5 Question Source:

Bank # Q77 2003 NRC Exam (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.41.7

Examination Outline Cross-Reference Level SRO 215005 APRM / LPRM Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Recirculation flow channels flow mismatch Tier #

2 Group #

1 K/A #

A2.07 Rating 3.4 Revision 1

Question 76 While operating at 100 percent power, the following conditions develop in sequence:

• Reactor and main generator power rapidly drop to 90 percent of rated

• Recirculation loop A flow is now 10 percent higher than loop B with no change in flow control valve positions

• RPV level rapidly rises to Level 8, followed by a reactor scram and turbine trip Based on this, the SRO should enter _____.

A. AOP-0006, Condensate and Feedwater Failures B. AOP-0009, Loss of Normal Service Water C. AOP-0042, Loss of Instrument Bus D. AOP-0062, Jet Pump Failures Answer: D Explanation:

A. Incorrect: water level may rise, but not trend in power and recirc flow differential B. Incorrect: Loss of all service water can cause a scram, but not due to a level 8 trip C. Incorrect: Loss of an instrument bus may cause some of the indications, but not all D. Correct: Indications are those of a Jet Pump Failure

Technical

References:

References to be provided to applicants during exam: None Learning Objective: RLP-OPS-AOP062 Objective: 2 Question Source:

Bank #

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam June 2007 NRC Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis C3 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 239002 SRVs Knowledge of the bases in Tech Specs for LCOs and Safety Limits Tier #

2 Group #

1 K/A #

2.2.25 Rating 4.2 Question 77 According to Technical Specification bases, the safety mode of the Safety/Relief Valves is designed to protect against which event?

A. MSIVs close with a high flux scram B. Main generator load reject/generator trip C. Rod ejection accident at low power D. Main turbine stop valve closure with the failure of bypass valves Answer: A Explanation:

A. Correct - Per Technical Specification Bases Section 3.4.4, the most severe pressure transient which the safety function (mode) of the SRV is designed to protect against is closure of all the MSIVs with the reactor scram signal coming from high flux rather than the MSIV valve closure signal.

B. Incorrect - This answer is incorrect because the generator load reject is not the most severe transient. This event assumes the reactor scrams on stop valve closure initiation. This answer is plausible because the event does cause a pressure and power transient.

C. Incorrect - this is not the reason for the safety mode but could be picked because it is an accident analyzed in the USAR.

D. Incorrect - This answer is incorrect because this is not the most severe transient which the SRV safety function is designed to preclude. The stop valve closure will cause a reactor scram and the pressure transient is not the most severe. This answer is plausible because a pressure perturbation will occur as a result of the transient.

Technical

References:

TS Bases, 3.4.4, page B-3.4-18, rev 0.

References to be provided to applicants during exam: None Learning Objective: RLP-0109, Obj H16, L7 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 262002 UPS (AC/DC)

Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Over voltage Tier #

2 Group #

1 K/A #

A2.02 Rating 2.5 Question 78 During the performance of STP-302-0102, Power Distribution System Operability Check, the Unit Operator has reported a voltage reading of 123.2 volts AC on VBS-PNL01A.

A report from the field confirms 123.2 volts AC at the output of ENB-INV01A.

(1) What is the impact of this condition; and (2) What actions should the CRS direct?

A. (1) ENB-INV01A is inoperable and the associated vital bus distribution system is inoperable; (2) SOP-0048 Section 5.5, Transfer from ENB-INV01A to ENB-INV01A1 supplying VBS-PNL01A.

B. (1) Only ENB-INV01A is inoperable.

(2) SOP-0048 Section 5.5, Transfer from ENB-INV01A to ENB-INV01A1 supplying VBS-PNL01A.

C. (1) Only ENB-INV01A is inoperable.

(2) SOP-0048 Section 5.2 Transferring an ENB Inverter from Normal Operation to Maintenance Bypass.

D. (1) ENB-INV01A is inoperable and the associated vital bus distribution system is inoperable.

(2) SOP-0048 Section 5.2, Transferring an ENB Inverter from Normal Operation to Maintenance Bypass.

Answer: A

Explanation:

A. Correct - The surveillance contains acceptance criteria for both the inverter and the vital bus. Both are outside of the given acceptance criteria therefore the surveillance requirement is not met so both components must be declared inop (TS 3.8.7 & 3.8.9).

Transferring to ENB-INV01A will satisfy TS 3.8.7 and will correct the voltage condition which will also allow exit of TS 3.8.9).

B. The vital bus distribution system is also inop due to voltage being outside the acceptance criteria. Part 2 is correct.

C. Part 1 is incorrect, see B. Operating an inverter in Manual Bypass does not satisfy TS 3.8.7. The correct action is to transfer to the alternate inverter.

D. Part 1 is correct, but operating an inverter in Manual Bypass does not satisfy TS 3.8.7.

The correct action is to transfer to the alternate inverter.

Technical

References:

STP-302-0102, TS 3.8.7, TS 3.8.9 References to be provided to applicants: TS 3.8.7, TS 3.8.9, STP-302-0102 Learning Objective: RLP-STM-0300 Obj 10 Question Source:

Bank #

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2014-3 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 264000 EDGs Knowledge of the operational implications of EOP warnings, cautions, and notes.

Tier #

2 Group #

1 K/A #

2.4.20 Rating 4.3 Rev 1 Question 79 The plant is operating at 100% power with the Division II Standby Diesel Generator running in parallel with offsite power when a loss of offsite power occurs.

Upon the loss of offsite power, the Division II EDG trips on a spurious high jacket water temperature.

The CRS should direct ___(1)__. Ensure that jacket water temperatures do not exceed a maximum of ___ (2) ___ to prevent damage to the diesel.

A. (1) implementation of OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, to emergency start the Div II diesel generator.

(2) 155°F B. (1) implementation of OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, to emergency start the Div II diesel generator.

(2) 200°F C. (1) operators to place the High Temperature Trip Bypass switch to BYPASS and transition to SOP-0053, Standby Diesel Generator and Auxiliaries, to start the Div II EDG.

(2) 155°F D. (1) operators to place the High Temperature Trip Bypass switch to BYPASS and transition to SOP-0053, Standby Diesel Generator and Auxiliaries, to start the Div II EDG.

(2) 200°F Answer: D Explanation:

A is wrong because the high temp bypass switch must be taken to bypass. Plausible if applicant believes that the diesel does not trip on high jacket temp when started in emergency mode.

B is wrong because see A.

C is wrong because see D. Part 2 plausible if applicant believes that operation above 155F is prohibited. The jacket water heater will de-energize when control switch is in AUTO and standpipe water temperature is greater than 155F. (R-STM-309S Rev 13, Page 26)

D is correct because AOP-0004, Loss of Offsite Power, directs that the high temp trip bypass must be taken to bypass and then the diesel restarted IAW SOP-0053. AOP-0004 note further states that jacket water may be allowed to exceed the alarm setpoint of 186F, but temperatures above 200F could cause damage to the diesel. SRO level question due to selection of appropriate procedures.

Technical

References:

AOP-0004, Loss of Offsite Power, Rev 52 SOP-0053, Standby Diesel Generator and Auxiliaries, Rev 332 OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, Rev 23 R-STM-309S Rev 13 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0309S, Rev.6, Enabling OBJ E.a Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 212000 RPS 2.2.25 Knowledge of the bases in Tech Specs for LCOs and Safety limits Tier #

2 Group #

1 K/A #

G2.2.25 Rating 2.5 Rev 1 Question 80 One of the ways that the Reactor Protection System provides protection from neutronic/thermal instabilities is that the Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function allowable values are specified in the COLR and are limited by LCO 3.2.4 by restricting the Fraction of Core Boiling Boundary (FCBB).

The normal and setup values are selected by operator manipulation of a Setup button on each flow control trip reference card.

The normal value provides protection when ___(1)___ the restricted region and with the FCBB limit___(2)___.

A. (1) inside (2) required to be met.

B. (1) outside (2) not required to be met.

C. (1) outside (2) required to be met.

D. (1) inside (2) not required to be met.

Answer: B Explanation:

A. Incorrect because normal (ie setup) value is for protection when outside the restricted region not inside it and with FCBB not required to be met per the bases document.

Both aspects are wrong for this distracter.

B. Correct - because normal (ie setup) value is for protection when outside the restricted region and with FCBB NOT required to be met per the bases document.

C. Incorrect because normal (ie setup) value is for protection when outside the restricted region (first part is correct) and with FCBB not required to be met, so the second part of this distracter is wrong per the bases document.

D. Incorrect because normal (ie setup) value is for protection when outside the restricted region not inside it. Second part of this distracter is correct.

Technical

References:

TS Bases pages3.3-8a and 3.3-8b, Revision 4-8.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0404, Obj 2, 3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 201002 RMCS Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Tier #

2 Group #

2 K/A #

2.2.42 Rating 4.6 Rev 1 Question 81 Consider the Following Plant Conditions:

• Reactor Power 12%

• Power escalation is in progress

• One Rod fails to move when attempting to withdraw from notch 20 to 24 Based on the above conditions alone, what Technical Specification should be entered and what is its basis?

A. TS 3.1.3 Control Rod Operability, because the capability of inserting the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.

B. TS 3.1.6 Control Rod Pattern, because the capability of inserting the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.

C. TS 3.1.3 Control Rod Operability, because this assures that the control rod patterns are consistent with the assumptions of the CRDA analyses.

D. TS 3.1.6 Control Rod Pattern, because this assures that the control rod patterns are consistent with the assumptions of the CRDA analyses.

Answer: A Explanation:

A. Correct B. is wrong because TS 3.1.6 is not applicable with thermal power greater than 10%

plausible because this is the correct basis C. is wrong because this is not the reason in the basis for TS 3.1.3 but plausible because this is the basis for TS 3.1.6 D. is correct because TS 3.1.6 is not applicable with the unit greater than 10% thermal power but plausible because this is the correct basis for TS.3.1.6

Technical

References:

TS 3.1.3 and TS 3.1.6 and the Basis for each References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0402, Obj 1 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 234000 Fuel Handling Equipment Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock Failure Tier #

2 Group #

2 K/A #

A2.01 Rating 3.7 Question 82 The Refuel Platform is lowering a fuel bundle into the center of the reactor vessel.

The Safety Travel interlock inadvertently actuates.

(1) What is the impact of this condition on the Refuel Platform?

(2) What action is required to complete the fuel move?

A. (1) Only Bridge movement is prevented. Trolley and Main Hoist movement may continue.

(2) Obtain Refuel Floor SRO permission and use the "Travel Override" button.

B. (1) Only Bridge movement is prevented. Trolley and Main Hoist movement may continue.

(2) Obtain Refuel Floor SRO permission and use the "Hoist Override" button.

C. (1) All Bridge, Trolley, and Main Hoist movements are prevented.

(2) Obtain Refuel Floor SRO permission and use the "Hoist Override" button.

D. (1) All Bridge, Trolley, and Main Hoist movements are prevented.

(2) Obtain Refuel Floor SRO permission and use the "Travel Override" button.

Answer: D

Explanation:

A. Incorrect because all Bridge, Trolley and Main Hoist movements are prevented.

B. Incorrect because all Bridge, Trolley and Main Hoist movements are prevented and the Hoist Override button is only used to raise the main hoist up past normal up.

C. Incorrect because the Travel Override button must be used D. Correct Technical

References:

FHP-0003,Refuel Platform Operation, (Rev 36)

Section 6.10.3 and 6.11.5.2 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0055 Obj 11b Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)7

Examination Outline Cross-Reference Level SRO 290003 Control Room HVAC 2.2.22 Knowledge of limiting conditions for operations and safety limits.

Tier #

2 Group #

2 K/A #

2.2.22 Rating 4.7 Rev 1 Question 83 During the rotation from Division 1 HVK to Division 2 HVK, Division 2 failed to start due to the failure of the oncoming service water pump.

The attempt to rotate back to Division 1 HVK was also unsuccessful because the oncoming chiller did not start.

As the CRS, you enter Technical Specification 3.7.3 for two inoperable control room AC subsystems inoperable.

It requires you to verify control room area temperature less than or equal to 104°F.

(1) Temperature is verified as being met by using a _____.

(2) The action required within 30 minutes for this condition is to prop open the doors at the _____.

A. 1) lollipop temperature indicator.

2) 116 ft. elevation for the battery and inverter rooms.

B. 1) hand held pyrometer.

2) 116 ft. elevation for the battery and inverter rooms.

C. 1) lollipop temperature indicator.

2) 98 ft. elevation for the switchgear rooms.

D. 1) hand held pyrometer.

2) 98 ft. elevation for the switchgear rooms.

Answer: B

Explanation:

A. Incorrect-Although there are lollipop indicators in some of the other rooms, the main control room does not have these and a handheld pyrometer must be used. The second part is correct per AOP-0060, step 5.1.6. The battery and inverter room doors must be propped open within 30 minutes of this event per this step.

B. Correct - this contains the hand held pyrometer, which per Att 1 of AOP-0060, is what is used to measure MCR temperature, and the battery room doors are propped open per step 5.1.6 of the AOP.

C. Incorrect - Although there are lollipop indicators in some of the other rooms, the main control room does not have these and a handheld pyrometer must be used. The second part is also incorrect per AOP-0060, step 5.1.6. The switchgear room doors must be propped open within 2 hrs of this event per this step.5.1.7.

D. Incorrect - The first part is correct, the hand held pyrometer, which per Att 1 of AOP-0060, is what is used to measure MCR temperature The second part is incorrect per AOP-0060, step 5.1.6. The switchgear room doors must be propped open within 2 hrs of this event per this step.5.1.7.

Technical

References:

AOP-0060, rev 10, page 6 and Attachment 1, page 1.

References to be provided to applicants during exam: None Learning Objective: RLP-OPS-710 Obj. 1 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 295026 Suppression Pool High Water Temp Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature: Suppression pool water temperature Tier #

1 Group #

1 K/A #

EA2.01 Rating 4.2 Revision 2

Question 84 The Plant is operating at 90% power and RCIC is being run in CST to CST mode for post maintenance testing.

The Suppression Pool Average Temperature was recorded as follows:

0800 94.0 °F 0805 96.2 °F 0815 100.5 °F 0818 102.3 °F RCIC turbine tripped 0820 105.3 °F 0821 106.0 °F CR initiated for failed testing Based on the above data, the CRS would enter Tech Spec 3.6.2.1______.

A. Condition A for SP Temp exceeding 100°F at time 0815.

B. Condition C for SP Temp exceeding 105°F at time 0820.

C. Condition A for SP Temp exceeding 100°F at time 0818.

D. Condition C for SP Temp exceeding 105°F at time 0821.

Answer: C Explanation:

While testing 105°F is limit, when testing secured 100°F is limit. Therefore condition C never applies for this event.

A. Incorrect - At 0815 testing is still being performed, so Condition A does not apply.

Condition C would apply but the higher limit of 105°F is not reached so this doesnt apply.

B. Incorrect - At 0820 testing is secured so condition C does not apply.

C. Correct - At time 0818 as soon as the pump is secured the LCO Condition A TS limit for 100°F with no testing is immediately met and must be entered.

D. Incorrect - At 0820 testing is secured so condition C does not apply.

Because the above the line information is required to be known from memory for RO and SRO applicants, the only concept this question is really testing is knowing that condition C only applies when testing is going on (adds heat to the Supp pool so higher temp is used during this time). Once you throw out condition C, it only leaves condition A and the 100°F limit is met after testing is secured (immediately).

Technical

References:

TS 3.6.2.1, Amendment 81, page 3-6-33.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0656 OBj H12, L7 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 295004 Partial or Total Loss of DC Pwr 2.2.40 Ability to apply Technical Specifications for a system.

Tier #

1 Group #

1 K/A #

2.2.40 Rating 4.7 Question 85 Plant is operating in MODE 3 with BYS-TRS4, XFR SW FOR SBO POWER CONNECTION is out of service for repairs, the following alarms are received on H13-P808 Insert 87, including:

• 125VDC BAT CHGR ENB-CHGR1B TROUBLE

• DIV II 125VDC CHGR BRKR ENB-ACB580- OPEN Based on these indications, which of the following conditions should the CRS enter?

A. LCO 3.8.7 (Inverters - Operating) Condition A, One inverter on Division I or II inoperable B. LCO 3.8.4 (DC Sources - Operating) Condition A, One required battery charger on Division I or II inoperable C. LCO 3.8.4 (DC Sources - Operating) Condition B, Division I or II DC electrical power subsystem inoperable for reasons other than Condition A D. LCO 3.8.5 (DC Sources - Shutdown) Condition A, One or more required DC electrical power subsystems inoperable.

Answer: B.

Explanation:

A., Incorrect T.S. 3.8.7 Condition A is not correct because B. Correct T.S. 3.8.4 Condition A basis requires the Non safety related Backup charger and SBO Diesel Generator to be available and with BYS-TRS4 out of service the SBO Diesel Generator is NOT available to supply the back up charger, Condition A is applicable.

C. Incorrect because this LCO is for conditions other than Condition A which is not applicable due to the SBO diesel generator not being available.

D. Incorrect LCO 3.8.5 is only applicable in Modes 4 and 5. Plausible if operator assumes mode 3 is in applicable.

Technical

References:

TS 3.8.4, RLP-STM-0305 References to be provided to applicants during exam: None Learning Objective: RLP-STM-0305 Obj 7 Question Source:

Bank #

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2014 -12 NRC Exam (Q76)

Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)2 2

Examination Outline Cross-Reference Level SRO 295005 Main Turbine Generator Trip Ability to determine and/or interpret the following as they apply to main turbine generator trip: Reactor Power Tier #

1 Group #

1 K/A #

AA2.05 Rating 3.9 Question 86 According to the basis for TS 3.3.1.1, the Turbine Stop Valve Closure function provides the primary reactor scram signal for a ___(1)___ event, and ensures that the ___(2)___

Safety Limit is not exceeded.

A. (1) Turbine Trip; (2) Reactor Coolant Pressure B. (1) Generator Load Rejection; (2) Reactor Coolant Pressure C. (1) Generator Load Rejection; (2) Minimum Critical Power Ratio D. (1) Turbine Trip; (2) Minimum Critical Power Ratio Answer: D Explanation:

A is wrong because... the Turbine Stop Valve Closure function protects against violating the MCPR safety limit. Plausible due to the potential pressure increase following a turbine trip.

B is wrong because the Turbine Stop Valve Closure function provides the primary scram signal for the turbine trip event. Plausible because the Turbine Control Valve Fast Closure function provides protection against a Generator Load Rejection event. The Turbine Stop Valve Closure function protects against violating the MCPR safety limit.

Plausible due to the potential pressure increase following a turbine trip.

C is wrong because the Turbine Stop Valve Closure function provides the primary scram signal for the turbine trip event. Plausible because the Turbine Control Valve Fast Closure function provides protection against a Generator Load Rejection event.

D is correct because per the TS basis, the Turbine Stop Valve Closure function is the primary scram signal for a turbine trip event and protects against exceeding MCPR.

Technical

References:

TS 3.3.1.1 Bases References to be provided to applicants during exam: None Learning Objective: RLP-STM-0508, Rev. 03, Objective #3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level RO 295016 Control Room Abandonment Knowledge of the emergency plan Tier #

1 Group #

1 K/A #

2.4.29 Rating 4.4 Rev 1

Question 87 At 1004, the control room is evacuated due to a fire.

The ATC mans the Div I Remote Shutdown Panel.

At 1022, the ATC reports that RCIC initiation has just commenced, and that RPV level is being maintained above -160 inches, and RPV pressure is being maintained less than 1094.7 psig.

The SRO should declare a(n) _____.

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: C Explanation:

A. INCORRECT.

B. INCORRECT.

C. CORRECT. AOP-0031 requires initiating RCIC for RPV Pressure and Level control within 15 minutes. EIP-2-001 classifies a site area emergency, HS3, for Control room evacuation having been initiated, and control of the plant cannot be established in accordance with AOP-0031, Shutdown from Outside the Main Control Room, within 15 minutes. An Alert is also applicable, HA4, for the fire, but the higher classification of the Site Area Emergency bounds.

D. INCORRECT.

Technical

References:

AOP-0031, Shutdown from Outside the Main Control Room, Revision 323; EIP-2-001, Classification of Emergencies, Revision 026 References to be provided to applicants during exam: EAL table only.

Learning Objective: RCBT-EP-SRORMED-Mod 1, Obj 6, 7 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295021 Loss of Shutdown Cooling Ability to determine and/or interpret the following as they apply to loss of shutdown cooling: RHR/shutdown cooling system flow Tier #

1 Group #

1 K/A #

AA2.02 Rating 3.4 Rev 1 Question 88 The Reactor has been shut down for a refueling outage for 30 days fuel shuffle/reload has been completed.

• The B Train of RHR is in operation removing decay heat.

• Fuel Transfer canal has been drained

• A surveillance is to be performed on the A Train of RHR that will render it inoperable.

• The A train is expected to be inoperable for a total of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(1) Is the removal of one of the required Trains of RHR allowed?

(2) What is the Basis for allowing or not allowing this surveillance to occur?

A. Yes, Removal of one required train is allowed per Technical Specifications.

The basis for this allowance is that RCS pressures and Temperatures are being closely monitored as required by LCO 3.4.11, RCS Pressure and Temperature Limits.

B. No, Removal of one required train is not allowed per Technical Specifications.

The basis for not allowing this is because decay heat removal is an important safety function that must be accomplished or core damage could result.

C. Yes, Removal of one required train is allowed per Technical Specifications.

The basis for this allowance is because the core heat generation can be low enough and the heatup rate slow enough to allow for loss of redundancy in the RHR system.

D. No, Removal of one required train is not allowed per Technical Specifications.

The basis for this that two RHR subsystems must remain operable to allow for accurate average coolant temperature monitoring and management of gas voids.

Answer: C Explanation:

A. Is incorrect because this is the basis for securing both trains to preform inservice leak testing and hydrostatic testing.

B. Is incorrect because This is allowed per the note in LCO 3.4.10 C. Correct D. Is incorrect because it is allowed per note in LCO3.4.10 Technical

References:

TS 3.4.10 and Basis References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0405, Obj 1.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 3

Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295024 High Drywell Pressure Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Tier #

1 Group #

1 K/A #

2.2.36 Rating 4.2 Revision 2

Question 89 SR 3.3.3.1.3 (CHANNEL CALIBRATION) was mis-performed, causing PAM Drywell Pressure Instrument ILMS-PT-2A to peg high.

30 days have elapsed and the condition has not yet been corrected.

(1) What Technical Specifications are required to be entered?

(2) Per Technical Specification bases, what is the reason for allowing 30 days completion time?

A. (1) TS 3.3.3.1 only (2) Because of the diversity of sensors available to provide trip signals and the redundancy of the system design B. (1) TS 3.3.3.1, TS 3.3.1.1 (RPS Instrumentation), TS 3.3.5.1 (ECCS Instrumentation), and others.

(2) Because of the diversity of sensors available to provide trip signals and the redundancy of the system design C. (1) TS 3.3.3.1 only (2) Because of the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments).

D. (1) TS 3.3.3.1, TS 3.3.1.1 (RPS Instrumentation), TS 3.3.5.1 (ECCS Instrumentation), and others.

(2) Because of the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments).

Answer: C Explanation:

Part (1) tests on the applicants knowledge of whether the Drywell Pressure PAM instrument uses the same transmitter as RPS and ECCS. RPS and ECCS use 4 transmitters, C71-PT-N050A-D, while PAM uses 2 separate transmitters, ILMS-PT-2A/B. An inoperable PAM drywell pressure transmitter does not affect the RPS/ECCS tech specs.

The tech specs that would be affected by an inoperable RPS-associated drywell pressure transmitter are: TS 3.3.1.1 (RPS Instrumentation), 3.3.5.1 (ECCS Instrumentation), 3.3.6.1 (Primary Containment and Drywell Isolation Instrumentation), 3.3.6.2 (Secondary Containment and Fuel Building Isolation Instrumentation), 3.3.6.3 (Containment Unit Cooler System Instrumentation), 3.3.7.1 (Control Room Fresh Air System Instrumentation). For simplicity, the distractors only name the RPS and ECCS instrumentation tech specs, but acknowledge that others also apply.

Part (2) tests on the applicants knowledge of basis for waiting 30 days for Tech spec entry.

From TS 3.3.1.1 Bases: RPS Instrumentation A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (Ref. 9) to permit restoration of any inoperable channel to OPERABLE status.

From TS 3.3.3.1 PAM Instrumentation A.1 When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel(s) (or in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments),

and the low probability of an event requiring PAM instrumentation during this interval.

A is wrong but plausible because of the bases for RPS completion time.

B is wrong because of the bases for RPS completion time. Also, only the PAM TS 3.3.3.1 is affected by the inoperable PT, not the RPS and ECCS tech specs.

C is correct.

D is wrong because only the PAM TS 3.3.3.1 is affected by the inoperable PT, not the RPS and ECCS tech specs.

Technical

References:

R-STM-0057, Rev. 4 (Primary Containment and Auxiliaries)

R-STM-0508, Rev. 6 (RPS)

References to be provided to applicants during exam: TS 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.

Learning Objective: RLP-HLO-0404, Obj 1 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 295028 High Drywell Temperature Ability to determine and/or interpret the following as they apply to high drywell temperature: Drywell temperature Tier #

1 Group #

1 K/A #

A2.01 Rating 4.1 Rev 1 Question 90 The plant is operating at 100% power with all drywell coolers running.

Shortly thereafter the following occurs:

• Half scram on Channel B with a Division II NSSSS isolation

• Standby Gas Treatment B and CMS H2 Analyzer B auto start

• Fuel and Control Building Charcoal Ventilation Treatment B starts

• Control Building HVAC B re-aligns to the Charcoal Ventilation Treatment Train A few minutes later the following annunciator is received on H13-P601:

• AIR TEMP MON R608 DRYWELL AMBIENT HIGH TEMP

• DW ambient temperature indicates 142F and is stable What procedure should the CRS use first for these events?

A. AOP-0010, Loss of One RPS Bus B. EOP-0002 and Enclosure 20 C. SOP-0060, Drywell Cooling D. AOP-0003, Automatic Isolations Answer: A Explanation:

A. Correct - these conditions indicate a loss of RPS bus B which also causes a loss of two of four unit drywell coolers (UC1B and UC1D) and the corresponding high drywell temperature. This procedure has guidance to allow restoring DW coolers without defeating interlocks.

B. Incorrect - EOP-0002 requires entry at 145F and the alarm comes in at approx.

142F. Also, to install enclosure 20 and defeat the interlocks it must be determined that DW temp cannot be maintained below 145F.

C. Incorrect - Although SOP-0060 has directions for starting unit coolers, UC1F cant be started because it has no power so this is not a good choice for DW cooling and does not address loss of the RPS bus either.

D. Incorrect - Resetting isolations per AOP-0003 would be ineffective without power restored to RPS Bus B.

Technical

References:

AOP-0010, rev 21, EOP-0002, rev 16, SOP-0060, rev 10, and AOP-0003, rev 34.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0403 Obj. 10; RLP-STM-0057 Obj. 4a Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 295020 Inadvertent Cont. Isolation Ability to interpret and execute procedure steps.

Tier #

1 Group #

2 K/A #

2.1.20 Rating 4.6 Question 91 A plant transient has resulted in numerous alarms and valve actuations in the Main Control Room, including the following:

• All 4 Division 1 RPS solenoid lights on H13-P680 are EXTINGUISHED

• All 4 Division 2 RPS solenoid lights on H13-P680 are ILLUMINATED

• All Division 1 RWCU containment isolation valves are CLOSED

• All Division 2 RWCU containment isolation valves are OPEN

• All Division 1 CCP containment isolation valves are CLOSED

• All Division 2 CCP containment isolation valves are OPEN

• Both the UP STREAM B33-AOVF019, DIV2 and the DN STREAM B33-AOVF020, DIV 1 Reactor Water Sample Line isolation valves closed

1) Which of the following procedures should the CRS direct for these conditions?

2) What initial subsequent actions should the CRS direct for these conditions?

A. 1) AOP-0003, Automatic Isolations

2) Reset the isolation signal using isolation reset pushbuttons B. 1) AOP-0003, Automatic Isolations

2) Verify all isolations completed as designed C. 1) AOP-0010, Loss of One RPS Bus

2) Reset the isolation signal using isolation reset pushbuttons D. 1) AOP-0010, Loss of One RPS Bus

2) Verify all isolations completed as designed Answer: C

Explanation:

C is correct-these indications are for loss of RPS Bus A and the required immediate steps to combat this event are to transfer RPS to the alternate bus, which is contained only in AOP-0010, not AOP-0003. The second part of the question deals with subsequent action steps which would be contained in the attachment for AOP-10 where the first step is to reset the isolation.

A is incorrect-wrong procedure and the second part is also wrong the initial subsequent steps are different than what is stated in the answer. Operators eventually reset the isolation in step 5.9 using the isolation reset pushbuttons B is incorrect because wrong procedure and wrong actions for part 2. Verification of isolation happens at step 5.7.

D is incorrect because the second part is wrong. Verification of isolation does not happen in the procedure until later in the restoration using the attachment.

Verify the isolation is complete Restore all systems to their normal lineup Reset the isolation signal Restore power to RPS A Restore power to RPS B Technical

References:

AOP-0010, Loss of One RPS Bus, revision 21.

AOP-0003, Automatic Isolations, Revision 34 References to be provided to applicants during exam: None Learning Objective: RLP-OPS-AOP0010 Obj 3 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295033 High Secondary Containment Area Radiation Levels Ability to determine and/or interpret the following as they apply to high secondary containment area radiation levels: Cause of high area radiation Tier #

1 Group #

2 K/A #

EA2.03 Rating 4.2 Rev 1 Question 92 Due to a report from security of various steam leaks in ECCS pump rooms you have tasked operators to search for the exact locations of the steam leaks. Before the operators report the leak locations the following radiation alarms come in on the DRMS computer. All indications are currently rising.

• HPCS Area 7.2 E+02 mR/hr

• HPCS Penetration Area 8.7 E+01 mR/hr

• LPCS Equipment Room 1.0 E+02 mR/hr

• RHR A Equipment Room 1.2 E+03 mR/hr

• RHR B Equipment Room 3.1 E+02 mR/hr

• RHR C Equipment Room 9.3 E+03 mR/hr (1) Based on the conditions what procedure is required to be entered?

(2) Why is the reason for entering the procedure?

A. (1) EOP-0001, EMERGENCY OPERATING PROCEDURE - RPV CONTROL (2) Because 2 areas are above their max safe value B. (1) EOP-0001, EMERGENCY OPERATING PROCEDURE - RPV CONTROL (2) Because any area is above its max safe value and a primary system is discharging into secondary containment C. (1) GOP-0002, POWER DECREASE/PLANT SHUTDOWN (2) Because 2 areas are above their max safe value D. (1) GOP-0002, POWER DECREASE/PLANT SHUTDOWN (2) Because any area is above its max safe value and a primary system is discharging into secondary containment Answer: B

Explanation:

A is wrong because while the first part is correct the second is not because there is a primary system discharging into secondary containment due to the increased rad levels.

B is correct because according to the EOP bases for step SC-8 any radiation level above its max normal operating value is an indication of a primary system leak. SC-14 is a before step and without knowing if the operators will find the leak before the setpoint is reached than SC-15 applies to enter EOP-1.

C is wrong because of the radiation levels not exceeding max safe in 2 areas. Had this question used temperatures or area water levels then GOP-2 could have applied if 2 areas exceeded their max safe D is wrong because of the incorrect procedure.

Technical

References:

References to be provided to applicants: EOP-3 Table SC-2 (Area Rad Levels ONLY)

Learning Objective: 4. Given an EOP step, identify the basis for the action taken.

Question Source:

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 500000 High CTMT Hydrogen Conc.

Knowledge of EOP mitigation strategies.

Tier #

1 Group #

2 K/A #

2.4.6 Rating 4.7 Rev 1 Question 93 A LOCA occurred, which for a time uncovered fuel in the core until adequate core cooling could finally be established.

EOP-0001, RPV Control and EOP-0002, Primary Containment Control have been entered and the following containment conditions exist:

• Pre-LOCA Containment Temperature 90°F

• POST-LOCA Containment Pressure 2 psig

• POST-LOCA Containment Temperature 165°F

• POST-LOCA Containment Hydrogen 3.6%

When the hydrogen igniters were placed in service, several failed to start.

Based on these conditions, the Control Room Supervisor should direct the crew to use:

A. SAP-2 to ensure all H2 removal equipment is running B. Figure 5 HDOL curve in EOP-0001 to determine if containment should be vented C. Enclosure 21, Emergency Containment Venting, to vent containment D. Enclosure 31, H2 Control Systems, to ensure all H2 removal equipment is running Answer: A Explanation:

A. Correct - SAP-2 is required to be entered immediately once 3.5% H2 is reached, as given in the stem. Exit from the EOPs and entry to the SAPs are provided at the hydrogen concentration limit of 3.5% to reinforce the SAP entry required in EOP-1/1A and EOP-4/4A

B. Incorrect - Although the curve is on EOP-0002 and SAP-2, it is not on EOP-0001 and would not be used prior to entering the SAP-2.

C. Incorrect - Although this could be used it is not correct at this point to implement this enclosure D. Incorrect - Although this could be used it is not correct at this point to implement this enclosure.

Technical

References:

EPSTG*0002, page B-8-33, rev 17, SAP-2 rev 6.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0512, Obj 7 Question Source:

Bank #

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level RO Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Tier #

3 Group #

K/A #

2.1.5 Rating 3.9 Question 94 Per EN-OP-115, Conduct of Operations, a minimum of ___(1)___ members of the site fire brigade is required to fulfill minimum shift staffing requirements.

The STA ___(2)___ be a member of the fire brigade A. (1) 4 (2) may B. (1) 5 (2) may C. (1) 4 (2) may not D. (1) 5 (2) may not Answer: D Explanation:

A. INCORRECT. Part 1 is incorrect but plausible if one believes that the fire brigade can go below minimum staffing for two hours. Part 2 is correct.

B. INCORRECT. According to EN-OP-115, Part 1 is correct as the minimum complement of the fire brigade is five. Part 2 is incorrect but plausible if one believes since the STA is not a licensed position, he can serve on the fire brigade.

C. INCORRECT. Part 1 is incorrect but plausible if one believes that the fire brigade can go below minimum staffing for two hours. Part 2 is incorrect but plausible if one believes since the STA is not a licensed position, he can serve on the fire brigade.

D. CORRECT. According to EN-OP-115, Part 1 is correct as the minimum complement of the fire brigade is five, and part 2 is correct as the STA may not be on the fire brigade.

Technical

References:

EN-OP-115, Conduct of Operations, Revision 17 References to be provided to applicants during exam: None Learning Objective: RLP-HLO-0206, Objective F Question Source:

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.5

Examination Outline Cross-Reference Level SRO Conduct of Operations Knowledge of new and spent fuel movement procedures.

Tier #

3 Group #

K/A #

2.1.42 Rating 3.4 Question 95 River Bend is in a refueling outage and in the process of placing a fuel bundle in a peripheral location.

When approaching the location to place the fuel, the bridge should be operated in the

___(1)___ mode because approaching a peripheral location in the other mode results in

___(2)___ which could potentially cause the bundle to contact the wall.

A. (1) Manual (2) Bundle Swing B. (1) Automatic (2) Bundle Swing C. (1) Manual (2) Location Overshoot D. (1) Automatic (2) Location Overshoot Answer: A Explanation:

A. Correct B. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall C. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall

D. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall Technical

References:

FHP-0003, REFUEL PLATFORM OPERATION. Rev 35 References to be provided to applicants during exam: None Learning Objective: RLP-RF-PROC, Obj 20.

Question Source:

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(6)

Examination Outline Cross-Reference Level SRO Equipment Control Knowledge of conditions and limitations in the facility license.

Tier #

3 Group #

K/A #

2.2.38 Rating 4.5 Question 96 Which of the following Limiting Conditions for Operation (LCOs) listed below provide guidance concerning a supported system LCO not being met because the support system LCO is not met?

A. LCO 3.0.4 B. LCO 3.0.5 C. LCO 3.0.6 D. LCO 3.0.7 Answer: C Explanation:

A. Incorrect - this is for mode changes and LCOs B. Incorrect - this is for equipment returned to service administratively for testing C. Correct - this is for support system LCOs D Incorrect - this is for special operations LCOs Technical

References:

TS Section 3, pages 3.02 and 3.03, Amendment number 173.

References to be provided to applicants during exam: None Learning Objective: RLP-HLO-416 Obj 15 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis F3 10CFR Part 55 Content:

55.43(b)1

Examination Outline Cross-Reference Level SRO Radiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Tier #

3 Group #

K/A #

2.3.5 Rating 2.9 Question 97 RMS-RE125, MAIN PLANT EXHAUST WRGM has been declared inoperable.

Which of the following is correct regarding meeting the requirements of TRM 3.3.11.3, Radioactive Gaseous Effluent Monitoring Instrumentation?

RMS-RE126, PARTICULATE AND GAS MONITOR ________.

A. satisfies the requirements of TRM 3.3.11.3. No further action is required.

B. can NOT meet the requirements of TRM 3.3.11.3. Periodic sampling by Chemistry is required.

C. satisfies the requirements of TRM 3.3.11.3 if an auxiliary sample holder for particulate and iodine grab samples is placed in line on the particulate and gas skid.

D. can NOT meet the requirements of TRM 3.3.11.3. Suspend the release of gaseous effluents from this pathway.

Answer: C Explanation:

A. As designed RMS-RE126 does not provide/meet Function 1b and 1c of TRM 3.3.11.3.

B. RMS-RE126 can be used to meet TRM 3.3.11.3. See C.

C. Correct - Per SOP-0086 P&L 2.5, RMS-RE126 can satisfy TRM 3.3.11.3 if a particulate and iodine sampler is utilized on the skid.

D. See C.

Technical

References:

SOP-0086 revision 16, P&L 2.5, TRM 3.3.11.3 rev 5.

References to be provided to applicants during exam: TRM 3.3.11.3 Learning Objective: RLP-STM-0511 Obj. 12, 13 Question Source:

Bank #

2010S Audit Q97 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental H4 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(4)

Examination Outline Cross-Reference Level SRO Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Tier #

3 Group #

K/A #

2.3.15 Rating 3.1 Rev 1

Question 98 Which of the following is required to discharge an LWS tank to the Mississippi River if RMS-RE107 is INOPERABLE?

A. Analyze two independent samples from the tank B. Analyze one sample from the tank and one sample from the effluent monitor line C. Analyze one sample from the tank and one sample from the recovery sample line D. Use LWS-FE197 to monitor the discharge and ensure no additional liquid is added to the tanks during the discharge Answer: A Explanation:

A. Correct - ADM-0054 requires that two indep. Samples be taken to perform the release. Other choices are plausible because these are streams that could be used to determine if the amount exceeds the ODCM limits. LWS-FE197 is also required to be operable nut is not a backup for RE107.

B. Incorrect - two samples are required.

C. Incorrect - two samples are required D. Incorrect - two samples are required Technical

References:

TRM 3.3.11.2 and ADM-0054 Rev 6a, section 5.5, page 9.

References to be provided to applicants during exam: None Learning Objective: RLP-STM-0603 Obj 8c

Question Source:

Bank #

2012 NRC Q98 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)4

Examination Outline Cross-Reference Level SRO Emergency Procedures / Plan Knowledge of operator response to loss of all annunciators.

Tier #

3 Group #

K/A #

2.4.32 Rating 4.0 Revision 1

Question 99 A reactor start-up is in progress and the ATC is raising power with recirc flow control valves from 85% to 90% following a planned sequence exchange when the following annunciators come in:

• P680-07A-C07, ANNUNCIATOR LOSS OF DC POWER P630

• P680-08A-A04, ANNUNCIATOR LOSS OF DC POWER P850 The CRS will _________________.

A. declare an ALERT within 15 minutes B. declare a NOUE within 15 minutes C. direct the ATC to insert a SCRAM D. direct the UO to start both EDGs Answer: B Explanation:

A. Incorrect - an ALERT would be declared if compensatory instruments were also lost during this event, and there is no other info given that this occurs in the stem.

B. Correct - NOUE is correct based on AOP-0055 and EIP-2-001 for loss of all annunciators without a corresponding loss of comp instruments with no significant transient in progress.

C. Incorrect - this is in direct conflict with AOP-0055 D. Incorrect - this is in direct conflict with AOP-0055 Technical

References:

EIP-2-0001, rev 26, page 108, AOP-0055, rev 21.

References to be provided to applicants during exam: EAL Chart Learning Objective: RLP_OPS-0547, Obj 2, 3 Question Source:

Bank #

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Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO Emergency Procedures / Plan Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Tier #

3 Group #

K/A #

2.4.38 Rating 4.4 Revision 2

Question 100 The plant is operating at 100% power when the control room receives the following reports:

• 0800: Field operator makes a report of a fire in the fuel building.

• 0810: Fire Brigade Leader reports that additional outside assistance will be required to control the fire.

• 0815: Fire Brigade Leader reports that the fire continues to burn, however no safety-related equipment has been affected.

• 0823: The Fire Brigade Leader reports damage to Fuel Pool Cooling Pump, P1A.

At the same time, the Shift Communicator reports to the Shift Manager that the initial notification form is ready to be transmitted.

Assuming that it takes the Shift Communicator 5 minutes to prepare or update the notification form, the Shift Manager/Emergency Director should take the following actions:

A. Declare a NOUE at time 0815. Declare an Alert at 0823.

Transmit the NOUE notification by 0830 and send an Alert notification by time 0838.

B. Declare a NOUE at time 0815. Declare an Alert at 0823.

Direct the Shift Communicator to update the notification form with the Alert declaration and transmit it before 0830.

C. Declare a NOUE at time 0810. Declare an Alert at 0823.

Direct the Shift Communicator to cancel the NOUE notification.

Send an Alert notification by 0838.

D. Declare a NOUE at time 0810.

Direct the Shift Communicator to submit the NOUE notification by 0825.

Declare an Alert at 0823 and send an Alert notification by 0838.

Answer: D Explanation:

A is wrong because...the UE should be declared at 0810 when it is known that the fire will not be stopped within 15 minutes. The second half is also incorrect, as there would be sufficient time to update the initial notification to an Alert level prior to sending it at 0830.

B is wrong because the UE should be declared at 0810 when it is known that the fire will not be stopped within 15 minutes. The second half would be correct if 0815 was the correct UE declaration time.

C is wrong because the UE notification should not be cancelled. Since there is not sufficient time to update the UE notification to an Alert, the UE notification should be transmitted and an Alert notification should be transmitted by 0838.

D is correct because the UE declaration time is correct and the notifications are correct.

Technical

References:

EIP-2-001 CLASSIFICATION OF EMERGENCIES EIP-2-002 CLASSIFICATION ACTIONS References to be provided to applicants during exam: EAL Charts Learning Objective: Document learning objective if possible.

Question Source:

Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)