ML22083A063

Wat 2022-02 Draft Written Exam

ML22083A063
Person / Time
Site:	Waterford
Issue date:	03/09/2022
From:	Heather Gepford NRC/RGN-IV/DORS/OB
To:	Entergy Operations
References
Download: ML22083A063 (199)

Text

Examination Outline Cross-Reference Level RO 003 Reactor Coolant Pump System (K6.04) Knowledge of the of the effect of a loss or malfunction on the following will have on the REACTOR COOLANT PUMP SYSTEM: Containment isolation valves affecting RCP operation Tier #

2 Group #

1 K/A #

K6.04 Rating 2.8 Question 1 With CC-641, CCW TO CONTAINMENT OUTSIDE CNTMT ISOLATION, failed shut, component cooling water flow is lost to ________ and must be restored within ________ or the affected equipment must be secured.

A. 1A/1B Reactor Coolant Pumps only, 10 minutes B. CEDM Fan Coolers, 3 minutes C. 2A/2B Reactor Coolant Pumps only, 10 minutes D. All Reactor Coolant Pumps, 3 minutes Answer:

D Explanation:

A is wrong because CC-641 failing shut secures CCW flow to all RCPs and requires flow to be restored within 3 minutes or the RCPs must be secured.

B is wrong because CC-641 failing shut secures CCW flow to CEDM fan coolers but does not require flow restoration within a certain time.

C is wrong because CC-641 failing shut secures CCW flow to all RCPs and requires flow to be restored within 3 minutes or the RCPs must be secured.

D is correct because CC-641 failing shut secures CCW flow to all RCPs and requires flow to be restored within 3 minutes or the RCPs must be secured.

Technical

References:

OP-002-003, Component Cooling Water, Revision 320, Attachment 11.1 OP-901-130 Reactor Coolant Pump Malfunction, Revision 012, E1 Seal Failure, Step 7 WLP-OPS-CC00, Component Cooling Water, Revision 43, slide 154 WLP-OPS-RCP00, Reactor Coolant Pumps, Revision 28, slide 106 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-RCP00, EO2 Identify physical connections and STATE interrelations with: CCW.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 004 (CVCS) Chemical and Volume Control System (G2.4.18) CHEMICAL AND VOLUME CONTROL SYSTEM: Knowledge of the specific bases for EOPs.

Tier #

2 Group #

1 K/A #

G2.4.18 Rating 3.3 Question 2 Following receipt of a Safety Injection Actuation Signal (SIAS), the basis for CVC-183, VCT outlet valve, being __________ is __________.

A. Open, to allow boric acid makeup tank discharge to the charging pump suction B. Shut, to allow boric acid makeup tank discharge to the charging pump suction C. Open, due to VCT level reaching 5.5% following the SIAS.

D. Shut, due to VCT level reaching 7.8% following the SIAS.

Answer:

B Explanation:

A is wrong because CVC-183 is shut following a SIAS.

B is correct because CVC-183 is shut following a SIAS to allow boric acid makeup tank discharge to be directed to the charging pump suction.

C is wrong because CVC-183 shuts when VCT level reaches 5.5% and also shuts on a SIAS.

D is wrong because CVC-183 resets to open when VCT level reaches 7.8%.

Technical

References:

SD-CVC, Chemical and Volume Control, Revision 19 pages 21, 30, 31 WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure, Revision 22 slides 1,17 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure, Revision 22 EO19 State the criteria required AND the basis for each of the following operations in the LOCA procedure, OP-902-002: Verify SIAS Actuation 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:

55.41(b)10 55.43

Examination Outline Cross-Reference Level RO 005 (RHRS) Residual Heat Removal System (K3.06) Knowledge of the effect that a loss or malfunction of the RHRS will have on the following (CFR: 41.7 / 45.6): CSS Tier #

2 Group #

1 K/A #

K3.06 Rating 3.1 Question 3 Following receipt of a Containment Spray Actuation Signal, while operating in Mode 1, CC-963A, Shutdown Cooling Heat Exchanger Outlet Valve, failing in its setpoint (SET PT) position would result in component cooling water flow through the Shutdown Cooling Heat Exchanger A being limited to _______.

A. 3100 gpm B. 3000 gpm C. 2100 gpm D. 100 gpm Answer:

C Explanation:

A is wrong because 3100 gpm is the component cooling water flow with CC-963A in the OPEN position, which includes 100 gpm of bypass flow.

B is wrong because 3000 gpm is the component cooling water flow with CC-963A in the OPEN position with no bypass flow present.

C is correct because the setpoint (SET PT) position is where the valve is open to pass 2000 gpm of component cooling water in addition to the already present 100 gpm of bypass flow.

D is wrong because 100 gpm is the bypass flow with CC-963A in the shut position.

Technical

References:

SD-SDC, Shutdown Cooling, Revision 9, pages 10-11 WLP-OPS-SDC00, Shutdown Cooling, Revision 30 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-SDC00, Shutdown Cooling, Revision 30, EO2, Identify the physical connections and state the interrelationships with the: Containment Spray System (CS)

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 006 Emergency Core Cooling (K2.04) Knowledge of the bus power supplies to the following ECCS: ESFAS-operated valves Tier #

2 Group #

1 K/A #

K2.04 Rating 3.6 Question 4 What is the power supply to SI-225A, HPSI HDR A to RC LOOP 1A FLOW CONTROL VLV?

A. Bus 213A B. Bus 311A C. Bus 313A D. Bus 317A Answer:

B Explanation:

A is wrong because Bus 213A does not power SI-225A, it is a non-safety related bus located in the same switchgear room as Bus 311A.

B is correct because Bus 311A is the power supply for SI-225A.

C is wrong because Bus 313A does not power SI-225A but is located in the same switchgear room as Bus 311A.

D is wrong because Bus 317A does not power SI-225A but is located in the same switchgear room as Bus 311A.

Technical

References:

SD-SI, Safety Injection, Revision 19, Page 63 SD-480, 480 VAC Distribution, Revision 11, Page 48 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-SI00, Safety Injection, Revision 32, EO3, Identify the physical connections and state the interrelationships with the: Power Supplies.

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 007 (PRTS) Pressurizer Relief / Quench Tank System (K5.02) Knowledge of the operational implications of the following concepts as they apply to the PRTS/QUENCH TANK SYSTEM:

Method of forming a steam bubble in the PZR Tier #

2 Group #

1 K/A #

K5.02 Rating 3.1 Question 5 In accordance with procedure OP-001-001, Reactor Coolant System Fill and Vent, a bubble is being drawn in the pressurizer for startup. In section 6.6 of this procedure, it states to use all available heaters in auto and when pressurizer water temperature reaches saturation temperature for RCS pressure, then adjust letdown flow to adjust Pressurizer level using

____1_____.

In step 6.6.7 of this procedure it states that a bubble is formed when _____2_____.

A. 1. RC-ILC-0110, Pressurizer Level Controller

2. level is raised with no corresponding increase in pressure B. 1. RC-ILC-0110, Pressurizer Level Controller

2. level is lowered with no corresponding drop in pressure C. 1. CVC-IPIC-0201, Backpressure Regulator Controller

2. level is raised with no corresponding increase in pressure D. 1. CVC-IPIC-0201, Backpressure Regulator Controller

2. level is lowered with no corresponding drop in pressure Answer: D Explanation:

Per section 6.6 the backpressure regulator is used once pressurizer water temp reaches saturation temperature for RCS pressure. It is used to raise letdown flow which lowers pressurizer level and when level drop is no longer causing a pressure drop then the bubble is formed per the NOTE above step 6.6.7 A is wrong because both part 1 and part 2 are incorrect (see above). Plausible because pressurizer level controller does control level during normal operations but no for this case.

Part 2 is plausible because this could be true for saturated conditions in the pressurizer but is not the correct method.

B is wrong because part 2 is incorrect (see above)

C is wrong because although part 1 is the correct controller the part 2 is incorrect D is correct because this is the correct controller and the correct method to draw a bubble.

Technical

References:

OP-001-001, Revision 41, step 6.6.7, page 34.

WLP-OPS-PLC-00, revision 23, slide 181.

WLP-OPS-CVC-00, revision 29 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CVC-00, LO3 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:

55.41(b)5

Examination Outline Cross-Reference Level RO 008 (CCW) Component Cooling Water System (A3.08) Ability to monitor automatic operations of the CCW SYSTEM including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal Tier #

2 Group #

1 K/A #

A3.08 Rating 3.6 Question 6 The following plant conditions exist:

• Plant has tripped from 100% power.

• RCS pressure is 1680 PSIA and slowly lowering.

• Containment pressure is 17.5 PSIA and slowly rising.

• NO operator actions have been taken.

A. CC-114A, CCW PUMP A TO AB SUCTION CROSS CONNECT, received a CLOSE signal B. CC-200A, CCW HEADER A TO AB SUPPLY ISOLATION, received a CLOSE signal C. CC-563, CCW HEADER AB TO B RETURN ISOLATION, received an OPEN signal D. CC-727, CCW HEADER AB TO A RETURN ISOLATION, received a CLOSE signal.

Answer:

A Explanation:

A is correct because CC-114A receives a CLOSE signal upon SIAS as part of splitting the A and B CCW headers, based on given conditions, SIAS signal is present (> 17.1 PSIA, <1684 PSIA) but CSAS is not (>/= 17.7 PSIA).

B is wrong because CC-200A remains OPEN during a SIAS and receives a CLOSE signal upon CSAS.

C is wrong because CC-563 receives a CLOSE signal upon SIAS to isolate the B CCW header from the AB header and the non-safety related CCW header.

D is wrong because CC-727 remains OPEN during a SIAS and receives a CLOSE signal upon CSAS.

Technical

References:

SD-CC, Component Cooling Water, Revision 25, Pages 65-66 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CC00, Component Cooling Water, Revision 43, EO-5, Explain the interrelations with the Engineered Safety Features Actuation System (ESFAS):

Safety Injection Actuation Signal (SIAS) Containment Spray Actuation Signal (CSAS)

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:

55.41(b)7 55.43

Examination Outline Cross-Reference Level RO 010 (PZR PCS) Pressurizer Pressure Control System (A4.02) PZR PCS - Ability to manually operate and/or monitor in the control room: PZR heaters Tier #

2 Group #

1 K/A #

A4.02 Rating 3.6 Question 7 At full power and normal lineup for the pressurizer pressure and level control systems, the LO LEVEL HEATER CUTOFF switch is monitored from ___1___ and is normally positioned to

____2_____.

A. 1. CP-2

2. X B. 1. CP-4

2. X C. 1. CP-2

2. BOTH D. 1. CP-4

2. BOTH Answer: C Explanation:

Per training slides the switch is selected to BOTH for normal operations and is selected away from a bad channel if it occurs. The switch is located on CP-2. CP-4 is credible because there are several other pressurizer level control switches and controllers on CP-4. X is credible because it is selected for most other items as the primary controlling instrument, however this is the exception and BOTH is selected with no failures of the X or Y channels.

A is wrong because part 2 is wrong (see above)...

B is wrong because part 1 and part 2 are wrong (see above)

C is correct because the switch is selected to BOTH for normal operations and is selected away from a bad channel of it occurs. The switch is located on CP-2 D is wrong because part 1 is wrong. Part 2 is correct (see above)

Technical

References:

WLP-OPS-PLC00, revision 23, slides 42 and 55 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PLC00 Objective EO1 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 012 (RPS) Reactor Protection System (A1.01) Ability to predict and/or monitor changes in parameters associated with operating the RPS controls including: Trip setpoint adjustment Tier #

2 Group #

1 K/A #

A1.01 Rating 2.9 Question 8 The following plant conditions exist:

The plant is operation at 100% with Normal Operating Pressure and Normal Operating Temperature OP-903-107, Plant Protection System Channel A Functional Test, is in progress.

It is reported that the NPO inadvertently depressed the LO SG Press Setpoint Reset pushbutton.

Based on this condition which of the following pressures (PSIA) is the approximate value at which a Steam Generator Low Pressure trip signal would first be generated, and which steam generator(s) are affected.

A. 750 PSIA, Steam Generator #1 only B. 646 PSIA, Steam Generator #1 and #2 C. 750 PSIA, Steam Generator #1 and #2 D. 646 PSIA, Steam Generator #1 only Answer:

B Explanation:

A is wrong because 750 PSIA is the normal setpoint value for the pretrip alarm, the LO SG Press Setpoint Reset pushbutton lowers the setpoint for both steam generators.

B is correct because the LO SG Press Setpoint Reset pushbutton lowers the setpoint by 184 PSI below the steam generator pressure (normally approximately 830 PSIA with given plant conditions) and it lowers the setpoint for both steam generators.

C is wrong because 750 PSIA is the normal setpoint value for the pretrip alarm, the LO SG Press Setpoint Reset pushbutton lowers the setpoint for both steam generators.

D is wrong because the LO SG Press Setpoint Reset, the pushbutton lowers the setpoint by 184 PSI below the steam generator pressure (normally approximately 830 PSIA with given plant conditions) but it lowers the setpoint for both steam generators.

Technical

References:

WLP-OPS-PPS, Plant Protection System, Revision 17, Slides 160-163 References to be provided to applicants during exam: None.

Learning Objective:

Summarize the operation and function of PPS (EO-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 3

10CFR Part 55 Content:

55.41(b)5 55.43

Examination Outline Cross-Reference Level RO 004 (CVCS) Chemical and Volume Control System (K6.05) Knowledge of the of the effect of a loss or malfunction on the following will have on the CVCS: Sensors and detectors Tier #

2 Group #

1 K/A #

K6.05 Rating 2.5 Question 9 The plant is operating at 90% power and Volume Control Tank level transmitter, CVC-ILT-0227, develops a leak in the reference leg. What initial effect would this malfunction have on the chemical and volume control system?

A. CVC-169, Volume Control Tank Inlet / Bypass To Holdup Tanks, will align to the Boron Management System position B. A low level alarm on Panel CP-4 only.

C. CVC-507, Refueling Water Storage Pool To Charging Pumps Suction Isolation, will align to the OPEN position D. CVC-183, Volume Control Tank Discharge Valve, will align to the CLOSED position Answer:

A Explanation:

A is correct because a reference leg leak causes indicated VCT level to read High which results in an automatic diversion of letdown to the boron management system.

B is wrong because of the effect described in A (high level).

C is wrong because CVC-507 opens on low VCT level.

D is wrong because CVC-183 closes on low VCT level.

Technical

References:

SD-CVC, Chemical and Volume Control, Revision 19, Pages 19 and 55.

References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CVC00, Chemical and Volume Control, Revision 29, EO6, Summarize the effect of a degradation of the CVC System on the: Boration flowpaths 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:

55.41(b)7 55.43

Examination Outline Cross-Reference Level RO 013 (ESFAS) Engineered Safety Features Actuation System (K6.01) Knowledge of the of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors Tier #

2 Group #

1 K/A #

K6.01 Rating 2.7 Question 10 A low failure of which of the below instruments would result in a Reactor Protection System actuation but not an Engineering Safety Features Actuation System actuation?

A. Steam Generator 1 Pressure B. Wide Range Pressurizer Pressure C. Steam Generator 2 Level D. Reactor Coolant Flow Answer:

D Explanation:

A is wrong because it causes both a RPS actuation and a ESFAS actuation (MSIS).

B is wrong because it causes both a RPS actuation and a ESFAS actuation (CIAS).

C is wrong because it causes both a RPS actuation and a ESFAS actuation (EFAS).

D is correct because it only causes an RPS actuation, low RCS flow is not an ESFAS actuation signal.

Technical

References:

SD-PPS, Plant Protection System Description, Revision 22, Pages 83,85 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPS, Plant Protection System, Revision 17, EO-3, Summarize the operation and function of PPS: Sensors and Detectors 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:

55.41(b)7 55.43

Examination Outline Cross-Reference Level RO 000007 Reactor Trip, Stabilization, Recovery (000007EK2.03) Knowledge of the interrelations between REACTOR TRIP, STABILIZATION, RECOVERY and the following: Reactor trip status panel Tier #

1 Group #

1 K/A #

EK2.03 Rating 3.5 Question 11 Given:

100% power Channel A Narrow Range Pressurizer Pressure indicates 2340 PSIA Channel B Narrow Range Pressurizer Pressure indicates 2275 PSIA At the Reactor Trip Status Panel, the operator would expect to see the ____1______.

After the operator returns from the reactor trip status panel to check pressurizer pressure indications a second time, the following is seen:

Channel A Narrow Range Pressurizer Pressure indicates 2355 PSIA Channel B Narrow Range Pressurizer Pressure indicating 2352 PSIA At the Reactor Trip Status Panel, the operator would now expect to see the ____2______.

A. 1. pre-trip light illuminated for Channel A High Pressurizer Pressure only

2. trip light illuminated for Channel A and B High Pressurizer Pressure, Relays K1-K4 lights extinguished, TCB 1 - TCB 8 lights Red B. 1. pre-trip light illuminated for Channel A and Channel B High Pressurizer Pressure

2. bistable relays K101-K104 on the Channel A 29 card drops out subsequently resulting in a high pressurizer pressure reactor trip C. 1. trip light illuminated for Channel A and B High Pressurizer Pressure, Relays K1-K4 lights extinguished, TCB 1 - TCB 8 lights Red

2. trip light illuminated for Channel A and B High Pressurizer Pressure, Relays K1-K4 lights extinguished, TCB 1 - TCB 8 lights Red D. 1. pre-trip light illuminated for Channel A High Pressurizer Pressure only

2. trip light illuminated for Channel A High Pressurizer Pressure only, Relay K1 lights extinguished, TCB 1 and TCB 5 lights Red Answer: A

Explanation:

A is correct because the pressurizer pressure high pretrip is above 2308 PSIA so only Channel A would generate the signal. The pressurizer high pressure trip signal is 2350 PSIA so both Channel A and B would result reactor trip on high pressurizer pressure.

B is wrong because only the A channel pressure is high enough for the pretrip signal and both A and B channel would produce the high pressurizer pressure reactor trip signal.

C is wrong because the initial pressures are not high enough on either channel to generate the reactor trip signal.

D is wrong because for the second part both the A and B channel pressures are high enough to generate the reactor trip signal.

Technical

References:

SD-PPS, Plant Protection System Description, Revision 22 Pages 1,26,27,106-108,123 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPS00, Plant Protection System, Revision 17 EO-16, Discuss and/or apply methods that provide for the means of monitoring automatic/manual operation.

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000009 (EPE 9) Small Break LOCA (000009EA1.09) Ability to operate and / or monitor the following as they apply to SMALL BREAK LOCA: RCP Tier #

1 Group #

1 K/A #

EA1.09 Rating 3.6 Question 12 While Operating at 100% power:

A Small Break LOCA occurred After SPTAs were completed the crew entered procedure OP-902-002, LOCA Recovery and the following conditions were noted:

o Containment Pressure is 3 psia and slowly rising o

RCS T cold is 379 degrees F and slowly lowering o

RCS Pressure is 1750 psia and slowly lowering What action is required and why?

A. Secure only TWO RCPs to reduce heat input into the RCS B. Secure only TWO RCPs to prevent core lift C. Secure ALL RCPs to prevent core lift D. Secure ALL RCPs to reduce heat input into the RCS Answer: B Explanation:

A is wrong because the reason to secure only two RCPs is wrong. In this case the RCS pressure is above the required trip two leave two RCPs running for RCS pressure (not below 1621 psig).....

B is correct because two are tripped and two are left running to prevent core lift C is wrong because all four RCPs are not secured for these conditions. The reason is correct on why they are secured.

D is wrong because all four RCPs are not secured for these conditions. The reason is also incorrect (core lift is reason).

Technical

References:

WLP-OPS-PP02, Revision 22, slide 87 and 88.

OP-902-002, LOCA Recovery, Revision 21, page 9.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PP02 Objective EO-19

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:

55.41(b)10

Examination Outline Cross-Reference Level RO 000022 (APE 22) Loss of Reactor Coolant Makeup (000022AA2.03) Ability to determine and interpret the following as they apply to the LOSS OF REACTOR COOLANT MAKEUP:

Failures of flow control valve or controller Tier #

1 Group #

1 K/A #

AA2.03 Rating 3.1 Question 13 The following plant conditions exist:

The reactor is operating at 95% power and stable Pressurizer Level is 53%

Charging Pump B is running with A & AB in STANDBY Letdown flow is 40 GPM with the NORMAL Letdown Flow Control (FCV) valve and Backpressure Regulating (BPRV) valve in service The Letdown Low Flow alarm was just received. Letdown flow indicates ZERO (0).

Letdown FCV controller output is at 100%.

NORMAL Letdown Flow Control (FCV) valve indicates closed Backpressure Regulating (BPRV) valve controller output is at 0%.

NORMAL Backpressure Regulating (BPRV) valve indicates closed.

Crew has entered OP-901-112, Charging or Letdown Malfunction procedure.

Which ONE of the following actions will be required for these conditions?

A. Place Flow Control Valve Manual/Auto station to MANUAL and raise the flow B. Place ALTERNATE Letdown BPRV in service C. Place ALTERNATE Letdown FCV in service D. Secure both Charging and Letdown

.Answer: C Explanation:

A is wrong because cannot operate FCV M/A in manual if valve is failed closed, plausible because applicant may not understand that a failed shut FCV can be opened manually from the M/A station.

B is wrong because BPRV is responding properly to the loss of Letdown flow, plausible because applicant may misdiagnose the initial conditions as a failed BPRV.

C is correct because, OP-901-112, directs placing the ALT FCV in service.

D is wrong because this would be the action for a loss of ability to control pressurizer level, plausible because applicant may misdiagnose initial conditions as loss of pressurizer level control.

Technical

References:

OP-901-112, Charging or Letdown Malfunction, Revision 9, Pages 15-16 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CVC00, Chemical and Volume Control, Revision 29, LO-3, Summarize the operation and function of Major Components in the Letdown, Charging and Boron Management Subsystems: Letdown Flow Control Valve, CVC-113A(B)

Question Source:

Bank #

2009 NRC Q57 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2009 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)7 55.43

Examination Outline Cross-Reference Level RO 000026 (APE 26) Loss of Component Cooling Water (000026AK3.01) Knowledge of the reasons for the following responses as they apply to the LOSS OF COMPONENT COOLING WATER:

The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCWS coolers Tier #

1 Group #

1 K/A #

AK3.01 Rating 3.2 Question 14 Given:

The site has experienced a Loss of Offsite Power A large break Loss of Coolant Accident has occurred A Safety Injection Actuation Signal has occurred Once their associated 4kV bus is repowered, the ACC-110A(B), ACC Pump Discharge Line Isolation Valves, begin to _____1_____ in order to _____2_____.

A. 1. CLOSE

2. prevent pump runout of the associated Auxiliary Component Cooling Water pump B. 1. OPEN

2. initiate cooling flow for post-LOCA required heat loads C. 1. CLOSE

2. prevent a water hammer event due to potential leakage past ACC-108A(B), ACCW Pump Discharge Check Valve D. 1. OPEN

2. prevent essential chillers from tripping on low cooling water flow Answer: C Explanation:

A is wrong because the valve(s) closes to prevent a water hammer event not pump runout.

B is wrong because the valve(s) close not open and this is not the reason for their repositioning either.

C is correct because the valve(s) close to prevent a water hammer event.

D is wrong because the valve(s) close not open and part 2 is not correct for these valves.

Technical

References:

SD-CC, Component Cooling Water, Revision 25, Pages 1,47-49 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CC00, Component Cooling Water, Revision 43, EO-3, Summarize the operation and functions of the following major equipment. Include interlocks: ACCW flow control.

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:

55.41(b)14

Examination Outline Cross-Reference Level RO 000029 (EPE 29) Anticipated Transient Without Scram (000029EK1.01) Knowledge of the operational implications of the following concepts as they apply to ANTICIPATED TRANSIENT WITHOUT SCRAM: Reactor nucleonics and thermo-hydraulics behavior Tier #

1 Group #

1 K/A #

EK1.01 Rating 2.8 Question 15 The Diverse Reactor Trip System uses ____1______ as its actuation parameter because it will

_____2_____ during an Anticipated Transient Without Scram event.

A. 1. steam generator pressure

2. rise rapidly due to RCS heat-up B. 1. pressurizer pressure

2. rise rapidly due to RCS heat-up C. 1. steam generator pressure

2. lower rapidly due actuation of the Diverse Emergency Feedwater Actuation System D. 1. pressurizer pressure

2. lower rapidly due actuation of the Diverse Emergency Feedwater Actuation System Answer: B Explanation:

A is wrong because DRTS uses pressurizer pressure at its ATWS indicator.

B is correct because DRTS uses presurizer pressure as its ATWS indicator because pressurizer pressure rises rapidly during an ATWS event due to RCS heatup.

C is wrong because DRTS uses pressurizer pressure at its ATWS indicator.

D is wrong because pressurizer pressure rises rapidly during an ATWS due to RCS heatup.

Technical

References:

SD-ATS, Anticipated Transient System DRTS & DEFAS, Revision 5, Pages 6-7 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-ATS00, Anticipated Transient System, Revision 13, EO1 - Summarize the design features which provide for ATWS protection.

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:

55.41(b)8 and 10 55.43

Examination Outline Cross-Reference Level RO 000040 (APE 40) Steam Line Rupture -

Excessive Heat Transfer (000040AK1.04) Knowledge of the operational implications of the following concepts as they apply to STEAM LINE RUPTURE -

EXCESSIVE HEAT TRANSFER: Nil ductility temperature Tier #

1 Group #

1 K/A #

AK1.04 Rating 3.2 Question 16 For an Excess Steam Demand event, as the life of the plant progresses the potential severity of pressurized thermal shock on the Reactor Coolant Pressure Boundary becomes

_____1_____ because _____2_____.

A. 1. greater

2. of the increase in the reference nil ductility temperature caused by irradiation of the reactor coolant pressure boundary B. 1. less

2. of the increase in the reference nil ductility temperature caused by irradiation of the reactor coolant pressure boundary C. 1. greater

2. of the decrease in the reference nil ductility temperature caused by irradiation of the reactor coolant pressure boundary D. 1. less

2. of the decrease in the reference nil ductility temperature caused by irradiation of the reactor coolant pressure boundary Answer: A Explanation:

A is correct because the increase in the reference nil ductility temperature causes the chance of a pressurized thermal shock event causing damage to the reactor coolant pressure boundary to increase (become greater).

B is wrong because the increase in the reference nil ductility temperature causes the chance of a pressurized thermal shock event causing damage to the reactor coolant pressure boundary to increase (become greater) not become less.

C is wrong because the reference nil ductility temperature becomes greater.

D is wrong because the reference nil ductility temperature becomes greater.

Technical

References:

UFSAR, Revision 311, Section 5.3.2.3, Page 5.3-16

References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE04, Excess Steam Demand Recovery Procedure with Transient and Accident Analysis, Revision 18, EO2, Explain the following theoretical concepts as they apply to an Excess Steam Demand Event: Nil-ductility temperature 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:

55.41(b)8 and 10 55.43

Examination Outline Cross-Reference Level RO 000056 (APE 56) Loss of Offsite Power (G2.4.18) Loss of Offsite Power: Knowledge of the specific bases for EOPs.

Tier #

1 Group #

1 K/A #

G2.4.18 Rating 3.3 Question 17 Given:

Loss of offsite power has occurred Crew has entered OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery The sequencers for both Emergency Diesel Generators are timed out The CRS has directed the BOP to Verify Containment Cooling The BOP will ____(1)____. Per the Tech Guide for OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery, the reason for this action is to ____(2)____.

A. 1. secure one containment fan cooler

2. conserve emergency diesel generator fuel oil B. 1. align containment fan coolers to slow speed

2. conserve emergency diesel generator fuel oil C. 1. secure one containment fan cooler

2. prevent damage to containment cooling ductwork D. 1. align containment fan coolers to slow speed

2. prevent damage to containment cooling ductwork Answer: C Explanation:

A is wrong because Part 2 is wrong. Securing one CFC will help EDG fuel oil consumption, but with no SIAS, fuel oil consumption is not a concern.

B is wrong because the only time four CFCs should be running together in slow speed is following a SIAS or surveillance testing. Neither of these events/evolutions are in progress, so going to slow speed is not required. Aligning the CFCs to slow speed will help EDG fuel oil consumption, but with no SIAS, fuel oil consumption is not a concern.

C is correct. Following a LOOP without a SIAS, all CFCs will start in Fast speed. If all four CFCs start in Fast there is a concern for damaging duct work.

D is wrong because the only time four CFCs should be running together in slow speed is following a SIAS or surveillance testing. Neither of these events/evolutions or in progress, so going to slow speed is not required. Part 2 is correct.

Technical

References:

OP-902-003 Rev. 11, step 13, page 12 TG-OP-902-003 Rev. 307, step 13 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE05, LOOP - Loss of Forced Circulation and Station Blackout, Revision 35, EO6 - Explain the procedure mitigation strategy for Loss of Offsite Power/Loss of Forced Circulation and Station Blackout Recovery Question Source:

Bank #

2017 NRC (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 000058 (APE 58) Loss of DC Power (AA1.01) Ability to operate and / or monitor the following as they apply to the LOSS OF DC POWER: Cross-tie of the affected dc bus with the alternate supply Tier #

1 Group #

1 K/A #

AA1.01 Rating 3.4 Question 18 Following a Loss of 125 VDC Bus TGG-DC, off normal procedure OP-901-313, Loss of a 125 Volt DC Bus, directs the operator to ____1______ and subsequently ____2______.

A. 1. start EDG A only

2. remotely transfer Bus 1A and 1B supply from the Unit Auxiliary Transformer A to Startup Transformer A and locally transfer Bus 2A and 2B supply from the Unit Auxiliary Transformer B to Startup Transformer B B. 1. start EDG A and EDG B

2. locally transfer Bus 1A and 1B supply from the Unit Auxiliary Transformer A to Startup Transformer A and remotely transfer Bus 2A and 2B supply from the Unit Auxiliary Transformer B to Startup Transformer B C. 1. start EDG B only

2. remotely transfer Bus 1A and 1B supply from the Unit Auxiliary Transformer A to Startup Transformer A and locally transfer Bus 2A and 2B supply from the Unit Auxiliary Transformer B to Startup Transformer B D. 1. start EDG A and B

2. remotely transfer Bus 1A and 1B supply from the Unit Auxiliary Transformer A to Startup Transformer A and locally transfer Bus 2A and 2B supply from the Unit Auxiliary Transformer B to Startup Transformer B Answer: D Explanation:

A is wrong because OP-901-313 directs starting both EDG A and EDG B, part 2 is correct B is wrong because for part 2, 1A/1B are transferred remotely and 2A/2B are transferred locally due to the loss of control power to the 2A/2B supply breakers C is wrong because OP-901-313 directs starting both EDG A and EDG B, part 2 is correct.

D is correct because OP-901-313 directs starting both EDG A and EDG B and 1A/1B are transferred remotely and 2A/2B are transferred locally due to the loss of control power to the 2A/2B supply breakers Technical

References:

OP-901-313, Loss of a 125 Volt DC Bus, Revision 307, Subsection E4 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-DC00, DC Distribution, Revision 31, EO9 - Given OP-901-313, Summarize the effects of a loss of the different buses: 125VDC Bus TGB-DC 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 012 (RPS) Reactor Protection System (A4.01) RPS Ability to manually operate and/or monitor in the control room: Manual trip button Tier #

2 Group #

1 K/A #

A4.01 Rating 4.5 Question 19 The RPS manual trip pushbuttons are located on main control room panels _____1_____

and trip the reactor by ____2______.

A. 1. CP-7 and CP-10

2. de-energizing the Reactor Trip Breaker shunt trip coils B. 1. CP-2 and CP-8

2. energizing the Reactor Trip Breaker shunt trip coils C. 1. CP-7 and CP-10

2. opening the Control Element Drive Mechanism MG sets output load contactors D. 1. CP-2 and CP-8

2. opening the Control Element Drive Mechanism MG sets output load contactors Answer: B Explanation:

A is wrong because the RPS manual trip pushbuttons are located on CP-2 and CP-8, part 2 is correct.

B is correct because RPS manual trip pushbuttons are located on CP-2 and CP-8 and they trip the reactor by energizing the Reactor Trip Breaker shunt trip coil and deenergizing the undervoltage trip coil C is wrong because the RPS manual trip pushbuttons are located on CP-2 and CP-8, part 2 is correct.

D is wrong because they trip the reactor by energizing the Reactor Trip Breaker shunt trip coil and deenergizing the undervoltage trip coil, part 1 is correct.

Technical

References:

SD-PPS, Plant Protection System Description, Revision 22 Page 14 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPS00, Plant Protection System, Revision 17 EO3 - Summarize the operation and function of PPS: Trip Actuation Devices 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:

55.41(b)7 55.43

Examination Outline Cross-Reference Level RO 022 (CCS) Containment Cooling System (G2.4.34) CONTAINMENT COOLING SYST Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Tier #

2 Group #

1 K/A #

G2.4.34 Rating 4.2 Question 20 Given the following:

The plant is operating in MODE 1.

A Loss of Coolant Accident has occurred.

The following conditions are noted:

o RCS pressure = 1750 PSIA.

o Containment Pressure = 17.3 PSIA.

o Containment Fan Cooler A has failed The operators should override and close Containment Fan Cooler A Component Cooling Water valves __________.

A. From Panel CP-18, Main Control Room B. From Panel LCP-43, +21 Reactor Auxiliary Building C. From Aux Panel 2, +35 Relay Room D. From Aux Panel 1, +35 Relay Room Answer: D Explanation:

A is wrong because the valves are overridden closed from Aux Panel 1, plausible because CP-18 is the MCR panel where some CCW controls/indications are located.

B is wrong because the valves are overridden closed from Aux Panel 1, plausible because LCP-43 has CCW pump controls but not valve controls.

C is wrong because the valves are overridden closed from Aux Panel 1, plausible because the B containment fan cooler CCW valves are overridden closed from Aux Panel 2.

D is correct because the valves are overridden closed if containment pressure is above the CIAS setpoint to ensure containment isolation.

Technical

References:

OP-902-009, Standard Appendices, Revision 319, Pages 139-140 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE08, Functional Recover Procedure, Revision 16, EO Given plant and/or equipment conditions, explain the bases for procedural steps and requirements, including notes and cautions, for the Functional Recovery Procedure.

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:

55.41(b)10

Examination Outline Cross-Reference Level RO 026 (CSS) Containment Spray System (A2.08) Ability to (a) predict the impacts of the following on the CSS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Safe securing of containment spray when it can be done Tier #

2 Group #

1 K/A #

A2.08 Rating 3.2 Question 21 Given:

A Loss of Coolant Accident has occurred in containment, a containment spray actuation signal was received, containment fan cooler A failed to start Containment pressure is 16.2 PSIA and slowly lowering Containment High Range Monitors are not in ALARM Containment Particulate/Iodine/Gaseous radiation monitor low flow is in ALARM Containment Temperature is 120 deg F and lowering Based on these conditions, OP-902-002, Loss of Coolant Accident Recovery, and its associated technical guideline, would direct the operators to __________.

A. terminate containment spray because the containment vacuum relief setpoint has been reached B. terminate containment spray because containment pressure is below the spray initiation setpoint and stable C. keep containment spray in service because of the containment temperature reading and fan cooler failure D. keep containment spray in service because it has not reached the setpoint where it will not reinitiate Answer: B Explanation:

A is wrong because this value is 15 psia and it is not reached.

B is correct because all conditions are met for containment spray termination and OP-902-009 Attachment 5-E is used to reset the CSAS.

C is wrong because all conditions are met for containment spray termination, containment temperature is lowering.

D is wrong because all conditions are met for containment spray termination, containment pressure is less than 16.7 PSIA and stable or lowering. There is no setpoint where it will not reinitiate.

Technical

References:

OP-902-002, Loss of Coolant Accident Recovery, Revision 021, Page 18 TG-OP-902-002, Technical Guide for Loss of Coolant Accident Recovery, Revision 020, Pages 46-47 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure with Transient and Accident Analysis, Revision 22, EO Given plant and/or equipment conditions, Interpret content, location, and sequencing of procedural steps applicable to the LOCA procedure, OP-902-002.

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:

55.41(b)5

Examination Outline Cross-Reference Level RO 039 (MSS) Main and Reheat Steam System (A1.06) Ability to predict and/or monitor changes in parameters associated with operating the MSS controls including: Main steam pressure Tier #

2 Group #

1 K/A #

A1.06 Rating 3.0 Question 22 A pressure surge resulting from a rapid change in fluid momentum in the Main Steam System (MSS) can be prevented by which ONE of the following:

A. Ensuring < 250 psid across MS-124A, Main Steam Isolation Valve, prior to opening B. Locally throttling open MS-1246A, MSIV 1 Bypass Line Downstream Isolation valves, after MSIVs are opened.

C. Cycling MS-119A, MSIV 1 Upstream Drip Pot Startup Drain, during startup of the MSS.

D. Verifying SBCS control switches are in manual with setpoint adjusted to zero prior to pressurizing the MSS.

Answer: C Explanation:

A is wrong because. OP-005-004 requires <100 psid prior to opening MSIV.

B is wrong because bypass line drains are opened prior to MSIVS being opened.

C is correct because the upstream drip pot startup drain is cycled during startup to remove condensate from main steam system.

D is wrong because SBCS are verified to be in AUTO prior to pressurizing main steam system.

Technical

References:

OP-005-004, Main Steam, Revision 36 WLP-OPS-MS00, revision 34 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-MS00, Objective 8 Question Source:

Bank #

R06042 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2006 RO Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)5

Examination Outline Cross-Reference Level RO 059 (MFW) Main Feedwater System (A4.11) MFW SYSTEM Ability to manually operate and/or monitor in the control room:

Recovery from automatic feedwater isolation Tier #

2 Group #

1 K/A #

A4.11 Rating 3.1 Question 23 Waterford was at full power when an inadvertent Main Steam Isolation Signal (MSIS) occurs on all channels. During recovery from this event, the MSIS initiation relay indicators are monitored from ___1_____ and are verified to be reset when _____2_____.

A. 1. CP-10

2. appropriate trip path indicators are out B. 1. CP-10

2. appropriate trip path indicators are illuminated C. 1. Cabinet K

2. appropriate trip path indicators are out D. 1. Cabinet K

2. appropriate trip path indicators are illuminated Answer: B Explanation:

An Inadvertent MSIS isolates main feed and main steam. IAW the off-normal procedure OP-901-504, Inadvertent ESFAS Actuation, section E4 (MSIS actuation), step 14.2, states to Verify the following Trip Path indicators Illuminated on the ENGINEERED SAFETY FEATURES SYSTEM mimic on CP-10.

A is wrong because part 2 is wrong. The trip path indicators are out before reset but illuminated after reset. The part 1 information (panel) is correct. Plausible if applicant confuses trip path indicators with initiation indicators or alarm reset indicators on cabinet K.

B is correct because CP-10 is the correct panel and the trip path indicators are illuminated after reset per reference above.

C is wrong because CP-10 is correct panel for these indicators, not Cabinet K. Plausible because Cabinet K is where the verification is done for MSIS logic initiated alarms are clear after the reset is complete at CP-10 D is wrong because location is wrong but the part 2 is correct (see above)

Technical

References:

OP-901-504, Inadvertent ESFAS Actuation, Rev 010, page 28.

WLP-OPS-PPS00, Rev 17.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPS00-EO16.

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 061 (AFW) Auxiliary / Emergency Feedwater System (K4.06) Knowledge of AUXILIARY /

EMERGENCY FEEDWATER SYSTEM design feature(s) and or interlock(s) which provide for the following: AFW startup permissives Tier #

2 Group #

1 K/A #

K4.06 Rating 4.5 Question 24 Given the following:

Diverse Reactor Trip and Diverse Emergency Feedwater switches in ENABLE Pressurizer pressure reading 2245 psia on all channels Steam Generator 1 pressure is 900 psia on all channels Steam Generator 2 pressure is 770 psia on all channels Steam Generator 1 NR level is 33% and WR level is 54% on all channels Steam Generator 2 NR level is 28% and WR level is 48% on all channels All emergency feedwater pumps are running due to A. EFAS 1 initiation ONLY.

B. EFAS 2 initiation ONLY.

C. EFAS 1 and EFAS 2 initiation.

D. DEFAS initiation.

Answer: D Explanation:

A is wrong because see D.

B is wrong because see D.

C is wrong because see D.

D is correct because all initiation signals have been met. Greater than 123 psid exists between steam generators but both steam generators are greater than 27.4 % NR so neither EFAS 1 or 2 will actuate. If someone thinks EFAS 1 or 2 has actuated than D cant be correct making the others plausible.

Technical

References:

SD-ATS, Anticipated Transient System, Rev 5, p. 8 References to be provided to applicants during exam: None.

Learning Objective: SUMMARIZE the operation and function of DRTS and DEFAS. [EO-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 2

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 062 AC Electrical Distribution System (K3.03) Knowledge of the effect that a loss or malfunction of the AC ELECTRICAL DISTRIBUTION SYSTEM will have on the following: DC System Tier #

2 Group #

1 K/A #

K3.03 Rating 3.7 Question 25 When all AC power is lost, the DC system will give a maximum (1) hours of power provided that loads are stripped in a maximum of (2) minutes.

A. (1) four (2) thirty B. (1) four (2) sixty C. (1) eight (2) thirty D. (1) eight (1) sixty Answer: A Explanation:

A is correct.

B is incorrect. Part 1 is correct. Part 2 is incorrect but plausible because batteries A (B) are rated for 1116 (1167) amp-hours for one hour.

C is incorrect. Part 1 is plausible because batteries are rated for 2320 amp-hours for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Part 2 is correct.

D is incorrect. Part 1 is plausible for the reason stated in distractor C. Part 2 is plausible for reason stated in distractor B Technical

References:

OP-902-005, Station Blackout Recovery, Revision 022, Pages 6 and 18 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-DC00, Objective 2 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 064 (EDG) Emergency Diesel Generator System (A3.13) Ability to monitor automatic operations of the EDG SYSTEM including: Rpm controller/megawatt load control (breaker-open/ breaker-closed effects)

Tier #

2 Group #

1 K/A #

A3.13 Rating 3.0 Question 26 Given:

EDG B is being synchronized to the grid in accordance with OP-009-002, Emergency Diesel Generator, for a retest following maintenance The synchroscope is slowly rotating in the counter-clockwise (Slow) direction.

EDG voltage is approximately 10 volts lower than Bus voltage IF the EDG output breaker is closed under these conditions, the potential effect of an (a)

____(1)____ trip exists.

Taking action to raise EDG _____(2)_____ PRIOR to closing the EDG output breaker will prevent the effect.

A. (1) overcurrent (2) speed (frequency)

B. (1) overcurrent (2) voltage C. (1) reverse power (2) speed (frequency)

D. (1) reverse power (2) voltage Answer: C Explanation:

A is wrong because overcurrent is not a concern with EDG voltage 10 volts below bus voltage. Plausible because EDG voltage should be match or slightly higher. Second part is correct B is wrong because overcurrent is not a concern with EDG voltage 10 volts below bus voltage. Plausible because EDG voltage should be match or slightly higher. Second part -

raising voltage will not correct the synchroscope rotation.

C is correct because synchroscope rotating in the counter-clockwise (Slow) direction has the potential of tripping the EDG output breaker on reverse power. Raising EDG speed (frequency) will correct the rotation.

D is wrong because the second part is wrong - raising voltage will not correct the synchroscope rotation. First part is correct.

Technical

References:

OP-009-002, Revision 357, Precaution 3.1.7 on page 6.

WLP-OPS-EDG-00, Revision 42.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-EDG00 obj. 2 & 8.

Question Source:

Bank #

2015 NRC (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 073 (PRM) Process radiation Monitoring System (K5.01) Knowledge of the operational implications of the following concepts as they apply to the PRM SYSTEM: Radiation theory, including sources, types, units, and effects Tier #

2 Group #

1 K/A #

K5.01 Rating 2.5 Question 27 The Main Steam Line N-16 Monitors, IRM-RE-5501A/B, consist of _____(1)______

detectors.

These monitors will first cause an alarm when detected activity is equivalent to a primary-secondary leak rate of ___(2)___.

(1)

(2)

A.

Geiger-Mueller 5 GPD B.

Geiger-Mueller 30 GPD C.

Scintillation 5 GPD D.

Scintillation 30 GPD Answer: D Explanation:

There are 2 primary detectors which monitor for primary-secondary tube leakage: main steam line N-16 monitors (Na-I scintillation) and the Air Evacuation Discharge Radiation Monitor (RE0004).

N-16 Monitors: If Primary-to-secondary leakage reaches 30 GPD or Primary-to-Secondary leakage rate of change reaches 30 GPD/HR, an alarm is generated at the RM-11 and on CP-

36. The detectors and monitor are non-safety related. It is technically an Area Radiation Monitor which monitors a process.

AE Discharge Monitor: The PMC alarm setpoint is initially set at an activity value equivalent to a 5 gpd leak and the high alarm setpoint is set for an equivalent 30 gpd leak. The detector is a PIG detector, but on the gaseous portion is used. It is a noble gas detector which uses a beta scintillator.

A is wrong because the N-16 monitors are Scintillation detectors, and first alarm at 30 GPD.

Plausible because the N-16 monitors are the only ARMs that are not Geiger-Mueller or Ion Chamber detectors, and 5 GPD is the first alarm setpoint of the AE Discharge monitor.

B is wrong because the N-16 monitors are Scintillation detectors. Part 2 correct. Plausible as described in A.

C is wrong because the N-16 monitors first alarm at 30 GPD. Part 1 correct. Plausible because 5 GPD is the first alarm setpoint of the AE Discharge monitor.

D is correct.

Technical

References:

SD-RMS Rev 15, Radiation Monitoring System WLP-OPS-RMS00 Rev 31, Radiation Monitoring System References to be provided to applicants during exam: None.

Learning Objective: SD-RMS LO-1: Summarize the operation and function of: Process Radiation Monitors, Area Radiation Monitors.

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:

55.41(b)5 55.43

Examination Outline Cross-Reference Level RO 076 (SW) Service Water System (K1.19) Knowledge of the physical connections and/or cause-effect relationships between the SWS SYSTEM and the following:

SWS emergency heat loads Tier #

2 Group #

1 K/A #

K1.19 Rating 3.6 Question 28 The CCW COOLING MODE selector switches are in AUTO for essential chiller operations.

As A and B component cooling water heat exchanger outlet temperatures lower from 107F to 94F, the ____(1)____. During this change in mode of operation, the ACCW inlet valves (ACC-112A/B) will start to close ____(2)____ the ACCW outlet valves (ACC-139A/B).

(1)

(2)

A ACCW valves close BEFORE the CCW valves open At the same time as B

CCW valves will open BEFORE the ACCW valves close Before C

ACCW valves close BEFORE the CCW valves open Before D

CCW valves will open BEFORE the ACCW valves close At the same time as Answer: A Explanation:

For Waterford the ACCW is one type of emergency service water (dry cooling towers and wet cooling towers the other two) and a heat load for it is the CCW heat exchangers (and therefore a KA match).

A is correct because with the mode selector switch in AUTO, the essential chillers should have switched to WET cooling mode at 105F. As CCW HX outlet temperature lowers to

<95F, the system will automatically realign to DRY cooling mode. This is accomplished by closing the ACCW valves first then opening CCW valves. Additionally, the ACCW outlet and inlet valves start closing at the same time.

B is wrong because the system will automatically align to DRY cooling mode by closing the ACCW valves first then opening the CCW valves. Applicant may incorrectly choose this since it intuitively maintains flow thru the essential chillers as it is switching the heat sink.

Additionally, the ACCW outlet and inlet valves are only staggered in operating when the system is automatically switching from DRY to WET and NOT from WET to DRY.

C is wrong because the ACCW outlet and inlet valves are only staggered in operating when the system is automatically switching from DRY to WET and NOT from WET to DRY.

D is wrong because the system will automatically align to DRY cooling mode by closing the ACCW valves first then opening the CCW valves. Applicant may incorrectly choose this since it intuitively maintains flow thru the essential chillers as it is switching the heat sink.

Technical

References:

SD-CC, Component Cooling Water Rev 25, page 29 WLP-OPS-CC00, Rev 43, slide 100 References to be provided to applicants during exam: None.

Learning Objective: EO-3 Summarize the operation and functions of the following major equipment. Include interlocks: ACCW supply CHW.

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 078 (IAS) Instrument Air System (K1.05) Knowledge of the physical connections and/or cause-effect relationships between IAS and the following: Service air Tier #

2 Group #

1 K/A #

K1.02 Rating 2.7 Question 29 In the event of an Instrument Air (IA) malfunction resulting in lowering IA header pressure, the Station Air (SA) to Instrument Air Cross-Connect valve, SA-125, is expected to be

__(1)__ opened upon a lowering IA header pressure of __(2)__ psig?

__(1)__ __(2)__

A. Manually 88 B. Automatically 88 C. Manually 105 D. Automatically 105 Answer: D SA-125 connects the SA system to the IA system. This valve is normally closed and will automatically open when the IA system pressure falls below an adjustable setpoint (normally set to about 105 psig). This allows the SA system to supply the IA system to prevent loss of instrument air to vital instrumentals, controls and valve operations.

Explanation:

A is wrong because both part 1 and part 2 of the question are incorrect. SA-125 is an automatically operated valve upon a lowering pressure setpoint, set to 105 psig. The distractor is plausible as 1) as the applicant may confused the SA containment isolation valve (SA-908) is a manually operated valve, and the SA to IA Cross-Connect valve, and 2) when IA pressure drops below 88 psig, valves with nitrogen backup are automatically supplied with nitrogen to ensure valves continue uninterrupted operation.

B is wrong because part 2 of the question is incorrect. SA-125 automatically opens valve upon a lowering pressure setpoint, set to 105 psig. Part 2 of the distractor is plausible as IA pressure drops below 88 psig, valves with nitrogen backup are automatically supplied with nitrogen to ensure valves continue uninterrupted operation.

C is wrong because part 1 of the question is incorrect. SA-125 is an automatically operated valve upon a lowering pressure setpoint, set to 105 psig. The distractor is plausible as the applicant may confused the SA containment isolation valve (SA-908) is a manually operated valve, and the SA to IA Cross-Connect valve.

D is correct. See answer explanation.

Technical

References:

Lesson Plan SD-AIR, Revision 14, page 19 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 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 103 (CNT) Containment System (K4.01) Knowledge of CNT SYSTEM design feature(s) and or interlock(s) which provide for the following: Vacuum breaker protection Tier #

2 Group #

1 K/A #

K4.01 Rating 3.0 Question 30 When Train A high containment to annulus differential pressure occurs, the containment vacuum relief system automatically A. closes CV-101, containment vacuum relief valve when pressure drops less than 8.5 INWD B. opens CV-201, containment vacuum relief valve when pressure rises greater than 8.5 INWD C. starts the Train B annulus negative pressure fan when pressure rises to greater than 5.5 INWD D. stops the Train B annulus negative pressure fan when pressure drops to less than 5.5 INWD Answer: B Explanation:

For train A the correct relief is CVR-201 and it opens when pressure exceeds 8.5 INWD per OP-008-005, rev 306, page 10. There is not an interlock for the fan (manually controlled) so C and D are incorrect based on stem stating automatically. A is incorrect because closing the relief would not be correct for either train. Values are plausible based on values in the procedure.

A is wrong because it is wrong valve for Train B and it opens not closes...

B is correct (see above)

C is wrong because annulus fans are not an auto feature for this system.

D is wrong because annulus fans are not an auto feature for this system.

Technical

References:

OP-008-005, rev 306, page 10.

WLP-OPS-CB00, Rev 24RM, slide 5 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CB00, 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000008 (APE 8) Pressurizer Vapor Space Accident (000008AK3.02) Knowledge of the reasons for the following responses as they apply to the PRESSURIZER VAPOR SPACE ACCIDENT:

Why PORV or code safety exit temperature is below RCS or PZR temperature Tier #

1 Group #

1 K/A #

AK3.02 Rating 3.6 Question 31 The plant was at 100% power when RC-317A, Pressurizer Safety Valve, failed open. The crew tripped the reactor and initiated Safety Injection Actuation Signal and Containment Isolation Actuation Signal.

The following conditions exist:

RCS pressure - 1255 PSIA RCS Thot temperature - 555 °F Quench Tank pressure - 55 PSIG What is the expected temperature indication of the Pressurizer Relief Line?

A. 287 degrees F B. 303 degrees F C. 555 degrees F D. 573 degrees F Answer: B Explanation:

A is wrong because the value is equal to the given pressure if the candidate does not convert to psia from psig which is the normal units for Quench Tank Pressure indication on CP-2....

B is correct because the value corresponds to the value that can be obtained from a Molliere diagram for 70 psia, assuming that the throttling process across the open relief valve is isenthalpic.

C is wrong because the value is equal to the Hot Leg temperature given, which could be selected if the candidate does not realize that the throttling process is isenthalpic and assumes that an in-surge from the Hot Leg has lowered PZR Temperature to Hot Leg Temperature.

D is wrong because value is equal to the saturation temperature for the RCS pressure given and could be selected if the candidate realizes the RCS is still subcooled, and therefore no significant in-surge has occurred to the PZR, but does not realize that the throttling process is isenthalpic.

Technical

References:

Steam Tables - Properties of Saturated and Superheated Steam References to be provided to applicants during exam: Steam Tables - Properties of Saturated and Superheated Steam Learning Objective:

WLP-OPS-RCS00, Reactor Coolant System, Revision 12-12, EO-4, Summarize the operation and function of code safety valves.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)5 and 10

Examination Outline Cross-Reference Level RO 000011 (EPE 11) Large Break LOCA (000011EA2.06) Ability to determine and interpret the following as they apply to LARGE BREAK LOCA: That fan is in slow speed and dampers are in accident mode during LOCA Tier #

1 Group #

1 K/A #

EA2.06 Rating 3.7 Question 32 Given:

Plant is operating at 100% power A LOCA occurs Containment pressure is 17.3 PSIA and rising Pressurizer pressure is 1700 PSIA and lowering The containment fan coolers will be operating in ____(1)______ speed. Containment fan cooler discharge will be to the containment cooling _____(2)______.

(1)

(2)

A. slow system ring header B. slow safety discharge dampers C. fast safety discharge dampers D. fast system ring header Answer: B Explanation:

A is wrong because Part 1 is correct. Discharge of the containment fan coolers will swap from CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS.

B is correct because SIAS is present. PZR pressure does not meet the setpoint for a SIAS but containment pressure does. On a SIAS, all containment fan coolers will swap to slow speed. Discharge of the containment fan coolers will swap from the CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS.

C is wrong because on a SIAS, all containment fan coolers will swap to slow speed. Part 2 is correct.

D is wrong because SIAS is present. PZR pressure does not meet the setpoint for a SIAS but containment pressure does. On a SIAS, all containment fan coolers will swap to slow speed. Discharge of the containment fan coolers will swap from the CCS ring header to CCS-102A and CCS-102B. These dampers fail open on a SIAS.

Technical

References:

SD-PPS Rev. 18, Table 3 OP-008-003 Rev. 302, page 10 References to be provided to applicants during exam: None.

Learning Objective: Lesson WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure Enabling Objective 19:

STATE the criteria required AND the basis for each of the following operations in the LOCA procedure, OP-902-002: [EO-19]

Verify Containment Isolation and Cooling Question Source:

Bank 2018 NRC Q41 (note changes; attach parent)

Modified New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 3

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 000015 (APE 15) Reactor Coolant Pump Malfunctions (000015 G2.4.111) Reactor Coolant Pump Malfunctions: Knowledge of abnormal condition procedures Tier #

1 Group #

1 K/A #

G2.4.11 Rating 4.0 Question 33 The plant is at full power when a lower seal failure is diagnosed due to rising controlled bleed-off temperatures. In accordance with OP-901-130, Reactor Coolant Pump Malfunction, you will lower component cooling water temperature by A. starting an additional wet cooling tower fan only.

B. starting an additional reactor cavity cooling fan.

C. starting an auxiliary component cooling water pump and raising the setpoint of ACC-126A(B).

D. starting dry cooling tower fans.

Answer: D Explanation:

A is wrong because the procedure has you start an ACCW pump and associated WCT fan.

B is wrong because not in the procedure but plausible if someone thinks the system will provide cooling to the RCPs but doesnt read the question well enough.

C is wrong because the procedure has you lower the setpoint of ACC-126A(B). Plausible if someone doesnt understand how the system works.

D is correct because it is one of the steps in the procedure.

Technical

References:

OP-901-130, Reactor Coolant Pump Malfunction, Rev. 12, p. 10 References to be provided to applicants during exam: None.

Learning Objective: Explain the procedural steps and requirements for OP-901-130 [EO-10]

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:

55.41(b)10

Examination Outline Cross-Reference Level RO 000025 (APE 25) Loss of Residual Heat Removal System (AK3.01) Knowledge of the reasons for the following responses as they apply to the LOSS OF RESIDUAL HEAT REMOVAL SYSTEM:

Shift to alternate flowpath Tier #

1 Group #

1 K/A #

AK3.01 Rating 3.1 Question 34 The unit is in Mode 3 with shutdown cooling Train A in service. The LPSI pump starts to cavitate and the CRS enters procedure OP-901-131, Shutdown Cooling Malfunction. The LPSI pump is tripped, and RCS makeup is required. What pump is started for makeup to the RCS (1) and why (2)?

A. (1) HPSI B pump (2) reduces the risk of gas intrusion using the non-affected train B. (1) HPSI A pump (2) this is the piping design for hot and cold leg injection C. (1) LPSI B pump (2) reduces the risk of gas intrusion using the non-affected train D. (1) CCP A pump (2) with a vent open this will clear the gas intrusion event Answer: A Explanation:

This KA involves a shift to the alternate flowpath, in this case the B Train because of a cavitating A LPSI pump. The opposite train HPSI pump is used for makeup due to risk of a gas intrusion event on the A train (ie this is the unaffected train). Doesnt matter if you are in the section for loss of cooling flow or for RCS leakage (Section E1 or E2). The makeup pump choice is the same in both cases.

A is correct (see above)

B is incorrect but plausible. Part 1 is plausible if applicant confuses piping design for hot and cold leg injection. Part 2 is plausible if applicant thinks to clear the gas event you need to fill that trains loop from the same train.

C is incorrect but plausible. Part 1 is incorrect but plausible if applicant confuses the train needed for SDC (the B train) and the pump needed for makeup. Part 2 is plausible for the reason stated in distractor B.

D is incorrect but plausible because of the CCP discharge pressure it could be used to clear a gas event but it is not true for makeup for this case. Part 2 is incorrect but plausible for same reason.

Technical

References:

OP-901-131, Revision308, page 11 WLP-OPS-RQE21, revision 21, slide 65.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-REQ21, LO6 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000027 (APE 27) Pressurizer Pressure Control System Malfunction (AA2.11) Ability to determine and interpret the following as they apply to the PRESSURIZER PRESSURE CONTROL SYSTEM MALFUNCTION: RCS pressure Tier #

1 Group #

1 K/A #

AA2.11 Rating 4.0 Question 35 Given:

The plant is at 100% power Control Systems are in normal 100% power alignments The Pressurizer Pressure Controller, RC-IPIC-0100, output fails HIGH Assuming no action by the crew, a reactor trip will occur first due to (1) Pressurizer Pressure generated by (2).

A. (1) High (2) PPS B. (1) High (2) CPCs C. (1) Low (2) PPS D. (1) Low (2) CPCs Answer:

Explanation: D A is wrong because the failure of the PZR controller high will cause the actual pressure to lower because the spray valves open. There is a high pressure trip associated with PPS B is wrong because the failure of the PZR controller high will cause the actual pressure to lower because the spray valves open. There is a high pressure trip associated with CPCs C is wrong because the low pressure trip is correct but the trip will first occur with the CPS.

There is a low pressure trip associated with the PPS D is correct.

Technical

References:

OP-901-120, section E2, revision 303 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPO10, revision 20, enabling objective 4 and WLPOPS-CPC00, revision 24, page 81 Question Source:

Bank #

2015 NRC (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b) 5 and 7

Examination Outline Cross-Reference Level RO 000038 (EPE 38) Steam Generator Tube Rupture (EA1.19) Ability to operate and / or monitor the following as they apply to STEAM GENERATOR TUBE RUPTURE: MFW System status indicator Tier #

1 Group #

1 K/A #

EA1.19 Rating 3.4 Question 36 The plant is operating at 100% power.

A steam generator tube leak has increased just beyond the capacity of the charging system to make up to the RCS.

Pressurizer and VCT levels are lowering slowly.

Steam Generator Water Level Control is in automatic.

With the reactor still at power, what indications will be seen on the affected Steam Generator?

A. Steam Flow rising Feed Flow lowering Steam Generator level maintained steady B. Steam Flow steady Feed Flow lowering Steam Generator level maintained steady C. Steam Flow rising Feed Flow steady Steam Generator level rising D. Steam Flow steady Feed Flow steady Steam Generator level rising Answer: B Explanation:

Because the leak is greater than the capacity of the charging system, it is categorized as a tube rupture. With the reactor still at power and the magnitude of the rupture just above charging capacity, the Steam Generator Water Level Control System will lower feedwater flow to maintain SG level at program. There is no change in load and therefore no change in steam flow.

A is wrong because steam flow will not rise.

B is correct as described above.

C is wrong because steam flow will not rise, feed flow will lower, which will cause SG level to be maintained at program and not rise.

D is wrong because feed flow will lower, and SG level will be maintained steady on program.

Technical

References:

WLP-OPS-PPE07 Rev 22, SGTR Recovery Procedure (OP-902-007) and Transient and Accident Analysis References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE07 Rev 22 EO-1: INTERPRET and ANALYZE plant conditions to: Diagnose Steam Generator Tube Rupture 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:

55.41(b)5 55.43

Examination Outline Cross-Reference Level RO 000054 (APE 54 CE E06) Loss of Main Feedwater (AA1.02) Ability to operate and / or monitor the following as they apply to the LOSS OF MAIN FEEDWATER: Manual startup of electric and steam-driven AFW pumps Tier #

1 Group #

1 K/A #

AA1.02 Rating 4.4 Question 37 While operating at 100 percent power:

a main feedline break occurred the reactor tripped followed by a loss of offsite power the A diesel generator failed to start EFW pump AB was started locally with OP-902-009 Standard Appendices, Appendix 36, Local Manual Control of EFW Pump AB Turbine The plant operator reports that MS-416, EFW Pump AB Turbine Stop Valve, is full open.

Emergency feedwater system deliverable flow is ______ gpm.

A. 350 B. 600 C. 700 D. 1050 Answer: D Explanation:

A is wrong because this is the flowrate for a single motor driven EFW pump and does not take into account the steam driven EFW pump (stop valve full open is full flow not zero flow);

is plausible because there is only one motor driven pump operating.

B is wrong because it is the flowrate of two motor driven EFW pumps operating but the A pump is not operating due to the LOOP and failure of the A EDG to start; is plausible because it is the system flowrate for operation with two motor driven EFW pumps.

C is wrong because it is the capacity of only the steam driven EFW pump and there is still power to the B EFW pump; is plausible because there is a LOOP and the failure of an EDG to start.

D is correct because it is the combined deliverable flows of the steam driven EFW pump (700 gpm) and a single motor driven EFW pump (350 gpm).

Technical

References:

WLP-OPS-EFW00, pages 45-48 OP-902-009 Standard Appendices, Appendix 36 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 2

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 000057 (APE 57) Loss of Vital AC Instrument Bus (G2.4.4) Loss of Vital AC Instrument Bus:

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Tier #

1 Group #

1 K/A #

G2.4.4 Rating 4.5 Question 38 The plant is at 100% full power conditions when and numerous alarms come in.

The following are two major equipment indicators for this event:

Reactor trip path breakers 1,2, 5, and 6 open Control room indications for EDG A are lost The crew enters AOP-901-312, Loss of A Vital Instrument Bus for loss of which vital bus?

A. PDP MA B. SUPS 014AB C. SUPS AB D. PDP MC Answer: D Explanation:

Answer is PDP MC causes both of these indications in the stem. Other choices are plausible because they are other vital buses and have impacts on the plant but not these indications.

A is wrong because it is PDP MC that causes these RTB to open and EDG A CR indications to be lost.

B is wrong (see above)

C is wrong (see above)

D is correct because this bus loss causes RTB 1, 2, 5, and 6 to open and also causes loss of EDG A control room indications.

Technical

References:

OP-901-312, Loss of Vital Instrument Bus, Revision 319 WLP-OPS-PPO30, Revision 17 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPO30. EO1

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000065 (APE 65) Loss of Instrument Air (AA2.08) Ability to determine and interpret the following as they apply to the LOSS OF INSTRUMENT AIR: Failure modes of air-operated equipment Tier #

1 Group #

1 K/A #

AA2.08 Rating 2.9 Question 39 The following plant conditions exist:

Instrument air pressure is 75 psig and decreasing The crew has entered OP-901-511, INSTRUMENT AIR MALFUNCTION The following annunciators are in alarm:

• SG1 FW ISOL VLV ACCU PRESS LO

• SG1 FW ISOL VLV AIR RSVR PRESS LO

• SG2 FW ISOL VLV ACCU PRESS LO

• SG2 FW ISOL VLV AIR RSVR PRESS LO According to OP-901-511, main feedwater isolation valves FW-184A and FW-184B have

___1___ and they will fail ___2___ for these conditions.

A. 1) N2 accumulators

2) as is B. 1) N2 Accumulators

2) closed C. 1) Air accumulators

2) as is D. 1) Air accumulators

2) closed Answer: C Explanation:

A is wrong because part 1 is wrong. They are air accumulators, and they fail as is.

Plausible because many of these valves have N2 accumulators, however, these two valves use Air.

B is wrong because part 1 and part 2 are wrong.

C is correct because they are air accumulators, and they fail as is.

D is wrong because part 2 is wrong (see discussion above).

Technical

References:

OP-901-511, Step 13 WLP-OPS-AIR00 Rev 26, slide 176

References to be provided to applicants during exam: None.

Document learning objective: Given Plant Status, IDENTIFY and APPLY procedural steps and requirements, including notes and cautions for the following procedures:

[EO-5]

- OP-901-511, Instrument Air Malfunction 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000077 (APE 77) Generator Voltage and Electric Grid Disturbances (AK3.02) Knowledge of the reasons for the following responses as they apply to the GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Actions contained in abnormal operating procedure for voltage and grid disturbances.

Tier #

1 Group #

1 K/A #

AK3.02 Rating 3.6 Question 40 Given:

Reactor power is 100%

The CRS has entered procedure OP-901-314, Degraded Grid Conditions due to grid instability issues.

A Transmission Loading Relief (TLR) Request was made to the Waterford 3 control room and the BOP operator has been directed to raise main generator voltage to 200 MVAR (out) to address the grid instability issues.

The main concern with this adjustment is ________.

A. excessive overexcitation could cause main generator high rotor field winding temperature B. excessive overexcitation could cause main generator pole slippage C. excessive reactive load could cause main generator high stator temperatures D. excessive reactive load could cause main generator high end iron temperatures Answer: A Explanation:

Per WLP-OPS-GEN-00, revision 15, slide 27, the main concern when operating over-excited (the top part of the generator operation curve) is high rotor temperatures (lagging or MVAR out). Pole slippage and high end iron temperatures are a concern when operating under-excited (leading or MVAR in). In this case raising voltage will raise excitation and will produce a lagging condition, which makes A the correct answer.

A is correct because (see above discussion).

B is wrong because this is correct for lowering voltage and creating an under-excited generator.

C is wrong because this is correct for a unity power factor only.

D is wrong because this is correct for lowering voltage and creating an under-excited generator.

Technical

References:

WLP-OPS-GEN-00, revision 15, slide 27

OP-901-314, Revision 5, section E0, steps 2 and 7.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-GEN00 Obj. 6.

Question Source:

Bank #

2012 NRC (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 000001 Continuous Rod Withdrawal Knowledge of the interrelations between the (APE 1) CONTINUOUS ROD WITHDRAWAL and the following: Rod motion lights Tier #

1 Group #

2 K/A #

AK2.05 Rating 2.9 Question 41 Given the following:

Reactor Power is 12% with power ascension in progress Reg Group 6 is being withdrawn manually in MS mode to raise power to 15%

Reg Group 6 CEAs have a 0.0 deviation from the Group No deviation exists between Pulse Counters and reed switch position transmitters for Group 6 The ATC releases the IN-HOLD-OUT switch at 135 and Reg Group 6 continues out due to failure of the OUT contacts in the switch The crew enters OP-901-102, CEA or CEDMCS Malfunction Assuming the reactor does NOT trip and no operator actions are taken yet,

1) what is the status of Group 6 position on CP-2?

2) what is the status of their associated rod motion indicator lights on CEDMCS control panel on CP-2?

A. (1) Group 6 pulse counters read approximately 150 (2) Red indicators are extinguished B. (1) Group 6 pulse counters on CP-2 read approximately 145.5 (2) Red indicators are extinguished C. (1) Group 6 pulse counters read approximately 150 (2) Red indicators are illuminated D. (1) Group 6 pulse counters on CP-2 read approximately 145.5 (2) Red indicators are illuminated Answer: B Explanation:

A is wrong because part 1 is wrong. Upper Group Stop would stop the group 6 rods at 145.5, not the Upper Electrical Limit of 150. This would occur at 145.5 on the pulse counters on CP-2. The red lights would be extinguished because they are actuated by the Upper Electrical Limit for the individual rod, not the Upper Group Stop.

B is correct because Upper Group Stop would occur at 145.5 on the pulse counters as stated. The red lights would be extinguished as stated because they are actuated by the Upper Electrical Limit for the individual rod, not the Upper Group Stop.

C is wrong because both parts are wrong. Upper Group Stop would occur as stated; however, this would occur at 145.5 on the pulse counters, not 150. The red lights would be extinguished not illuminated.

D is wrong because part 2 is incorrect. Upper Group Stop would occur as stated. This would occur at 145.5 on the pulse counters as stated. However, the red lights would be extinguished not illuminated.

Technical

References:

SD-CED, Control Element Drive, Revision 12 Pages 26-27 OP-901-102, revision 306 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CED00, Control Element Drive, Revision 24, EO-1, Summarize the design features which provide for: Circuitry and Principle of Operation of Reed Switches Question Source:

Bank #

2010 NRC Q57 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 000003 Dropped Control Rod Knowledge of the reasons for the following responses as they apply to the DROPPED CONTROL ROD: RIL and PDIL Tier #

1 Group #

2 K/A #

AK3.10 Rating 3.2 Question 42 While operating, Mode 1, 100% power, with All Rods Out, CEA 22 (Group 6) control element assembly rod bottom light is illuminated, a Power Dependent Insertion Limit alarm is received. TS 3.1.3.1, Moveable Control Assemblies, requires action to verify _____1_____

and if these limits cannot be maintained to _____2_____.

(1)

(2)

A.

radial xenon redistribution effects are maintained within limits emergency borate B.

radial xenon redistribution effects are maintained within limits restore the CEA to within limits within two hours C.

minimum shutdown margin requirements are maintained restore the CEA to within limits within two hours D.

minimum shutdown margin requirements are maintained emergency borate Answer: D Explanation:

Ro knowledge required to know 1 HR TS action statements.

A is wrong because radial xenon redistribution effects are maintained within limits is the basis for maintaining CEAs above the long term and short term insertion limits. Part 2 is correct.

B is wrong because radial xenon redistribution effects are maintained within limits is the basis for maintaining CEAs above the long term and short term insertion limits. The crew is allowed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore alignment.

C is wrong because the crew is allowed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore alignment. Part 1 is correct.

D is correct because TS 3.1.3.1 and the PDIL annunciator response procedure requires SDM verification and if it cannot be maintained per TS 3.1.1.1 then to emergency borate.

Technical

References:

W3 Technical Specifications 3.1.1.1, Shutdown Margin Any CEA Withdrawn, Amendment 249 and 3.1.3.1 Moveable Control Assemblies CEA Position, Amendment 182 OP-500-008, Control Room Cabinet H, Revision 44, Attachment 4.78, Power Dependent Insertion Limit References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CED00, Control Element Drive, Revision 24, EO-3, Given the Technical Specifications and plant status, identify and evaluate the operability, actions, and bases for the following: TS 3.1.3.1, CEA Operability.

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:

55.41(b)5, 10

Examination Outline Cross-Reference Level RO 000028 Pzr Level Control Malfunction Knowledge of the operational implications of the following concepts as they apply to (APE

28) PRESSURIZER (PZR) LEVEL CONTROL MALFUNCTION: PZR reference leak abnormalities Tier #

1 Group #

2 K/A #

AK1.01 Rating 2.8 Question 43 While operating at 100%, with Pressurizer Level Controller, RC-ILIC-0110, selected to Pressurizer Level Transducer, RC-ILT-0110X, a leak develops in the reference leg of RC-ILT-0110X. The operator would expect pressurizer level indicator RC-ILI-0110X to _____1_____

and in response would _____2_____ utilizing OP-901-110, Pressurizer Level Control Malfunction.

A. 1. Rise

2. Place RC-ILIC-0110 in MANUAL and adjust the OUTPUT to restore Pressurizer level to program B. 1. Lower

2. Place RC-ILIC-0110 in MANUAL and adjust the OUTPUT to restore Pressurizer level to program C. 1. Rise

2. Transfer Pressurizer Level Control CHANNEL SELECT switch to the non-faulted channel D. 1. Lower

2. Transfer Pressurizer Level Control CHANNEL SELECT switch to the non-faulted channel Answer: A Explanation:

A is correct because a reference leg leak on RC-ILT-0110X would cause the indicated level to rise, OP-901-110 directs the operator to place the controller in manual for a pressurizer level control malfunction.

B is wrong because a reference leg leak on RC-ILT-0110X would cause the indicated level to rise not lower. Part 2 is correct.

C is wrong because OP-901-110 directs the operator to place the controller in manual for a pressurizer level control malfunction, it contains a note that selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level. Part 1 is correct.

D is wrong because a reference leg leak on RC-ILT-0110X would cause the indicated level to rise not lower. OP-901-110 directs the operator to place the controller in manual for a pressurizer level control malfunction, it contains a note that selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level.

Technical

References:

SD-PLC, Pressurizer Level and Pressure Control, Revision 11, Pages 22 and 57 OP-901-110, Pressurizer Level Control Malfunction, Revision 11, Page 9 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PLC00, Pressurizer Level Control, Revision 23, EO-2, Explain the theoretical concepts as they apply to the PLC/PPC system: Level Detection 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:

55.41(b)8, 10

Examination Outline Cross-Reference Level RO 000033 Loss of Intermediate Range Nuclear Instrumentation Ability to operate and / or monitor the following as they apply to the (APE 33) LOSS OF INTERMEDIATE RANGE NUCLEAR INSTRUMENTATION: Level trip bypass Tier #

1 Group #

2 K/A #

AA1.02 Rating 3.0 Question 44 Given:

Plant is at approximately 5% power during a normal reactor shutdown.

Channel A Excore Nuclear Instrument middle detector fails low OP-500-009, Control Room Cabinet K has been entered Which of the following is true for Channel A:

A. High Log Power Trip Operating Bypass - Removed B. DNBR/LPD Trip Operating Bypass - Initiated C. Low SG Flow Trip Operating Bypass - Initiated D. High SG Level Trip Bypass - Removed Answer: A Explanation:

A is correct because High Log Power Trip Operating Bypass is removed automatically when reactor power decreases below 10^-4% power.

B is wrong because the DNBR/LPD Trip Operating Bypass is manually initiated below 10^-

4% power.

C is wrong because Low SG Flow Trip Operating Bypass is manually initiated when reactor power is below 8.5 x 10^-5% power.

D is wrong because High SG Level Trip Bypass is Removed manually by keyswitches (initiated manually when below 20% reactor power)

Technical

References:

SD-PPS, Plant Protection System, Revision 22 Page 84 SD-NI, Nuclear Instrumentation, Revision 8, Page 22 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPS00, Plant Protection System, Revision 17, EO4 - Identify the interlocks which provide: Manual enable and automatic disable of Operating Bypasses for certain RPS and ESFAS trips 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 000037 Steam Generator Tube Leak Knowledge of the reasons for the following responses as they apply to the (APE 37)

STEAM GENERATOR TUBE LEAK: Criteria for securing RCP Tier #

1 Group #

2 K/A #

AK3.08 Rating 4.1 Question 45 OP-902-007, Steam Generator Tube Rupture Recovery, Step 8, RCP Trip Strategy, establishes an RCP trip strategy if Pressurizer pressure is less than 1621 psi and SIAS is actuated with adequate RCP net positive suction head available. At this step

1) Which RCPs are secured?

2) ) What is the reason for this action for the given stem conditions (in accordance with TG-902-007)?

A. 1. All RCPs

2. This step reduces the potential for RCS inventory loss should a loss of coolant accident diagnosis be confirmed B. 1. All RCPs

2. This step ensures that the correct number of pumps are secured when RCS heat removal capability may be challenged by excessive leak rate if forced circulation is continued.

C. 1. One RCP per loop

2. It reduces the potential for RCS inventory loss should a loss of coolant accident diagnosis be confirmed D. 1. One RCP per loop

2. This step ensures that the correct number of pumps are secured when RCS heat removal capability may be challenged by excessive leak rate if forced circulation is continued.

Answer: C Explanation:

A is wrong because 1 RCP per loop is secured based on given conditions, part 2 is correct.

Tripping all RCPs is plausible because it is the action required in the same step if NPSH cannot be maintained.

B is wrong because 1 RCP per loop is secured based on given conditions. Tripping all RCPs is plausible because it is the action required in the same step if NPSH cannot be maintained (this is given in the stem that NPSH is met). Part 2 is plausible because it is the rationale for tripping all RCPs if NPSH cannot be maintained and is to ensure that all four RCPs are tripped as stated in the TG-OP-902-007 to ensures that all four RCPs are tripped for LOCAs in which the RCS leak rate may challenge RCS heat removal capability if forced circulation is

continued. This would be a correct reason to trip all four RCPs if part 1 were correct, but tripping four is not desired for the stem conditions.

C is correct because based on given conditions 1 RCP per loop would be secured for the listed reason.

D is wrong because Part 2 is incorrect but plausible because it is the rationale for tripping all RCPs if NPSH cannot be maintained. Part 1 is correct.

Technical

References:

OP-902-007, Steam Generator Tube Recovery Procedure, Revision 18, Page 8 TG-OP-902-007, Technical Guide Steam Generator Tube Recovery Procedure, Revision 309, Pages 26-27 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE07, SGTR Recovery Procedure (OP-902-007) and Transient and Accident Analysis, Revision 22, EO3 - Summarize the bases or reasons for: Starting or stopping Reactor Coolant Pumps 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:

55.41(b)5, 10

Examination Outline Cross-Reference Level RO 000067 Plant Fire On Site Ability to operate and / or monitor the following as they apply to the (APE 67) PLANT FIRE ON SITE: Fire alarm reset panel Tier #

1 Group #

2 K/A #

AA1.07 Rating 4.1 Question 46 Given the following:

Reactor is in Mode 1 at 100%

Control room receives an alarm on the Fire Detection Main Control Panel (FDMCP) o Detector RCB-04 is alarming (inside containment on EL +21)

1) What is the expected response of this panel in the control room?

2) Which ONE of the following actions will be completed first IAW FP-001-020, Fire Emergency/ Fire Report?

A. 1. strobe light turns on and large audible sounds

2. sound the fire alarm B. 1. only large audible sounds in main control room

2. sound the fire alarm C. 1. strobe light turns on and large audible sounds

2. perform a soft reset by pressing the Reset System soft key D. 1. only large audible sounds in main control room

2. perform a soft reset by pressing the Reset System soft key Answer: C Explanation:

A is wrong because this step is only performed if the fire alarm remains locked in after attempting a reset from the control panel.

B is wrong because this type of reset is only performed with the assistance of an I&C technician since this would be a system hard reset. Applicant may incorrectly conclude this is the answer since the procedure does require a system reset to check if the alarm remains locked in.

C is correct because FP-001-020, step 6.4.2.9.2 requires a reset on the control panel to check if the fire alarm clears.

D is wrong because this step in the procedure is completed well after attempting an alarm reset.

Technical

References:

FP-001-020, page 14-16 WLP-OPS-FP00, revision 23, slides 225, 228 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-FP00, LO-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:

55.41(b)10

Examination Outline Cross-Reference Level RO 000069 Loss of Containment Integrity (000069 Loss of Containment Integrity:

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 #

1 Group #

2 K/A #

G2.4.21 Rating 4.0 Question 47 Given:

An excess steam demand event has occurred with a diagnosed tube rupture on the A steam generator. The CRS has entered the functional recovery procedure OP-902-008 to evaluate the safety function for loss of containment integrity due to containment pressure exceeding the safety function requirement. With only containment fan coolers available (no containment spray pumps), containment pressure cannot exceed what value to meet this safety function status check?

A. 10.7 psia with coolers in normal mode B. 29.7 psia with coolers in normal mode C. 17.7 psia with coolers in emergency mode D. 50 psia with coolers in emergency mode Answer: C Explanation:

The value from the functional recovery procedure is 17.7 psia in emergency mode or 16.4 psia in normal mode and no CS pumps available.

A is wrong (see above) but plausible if the applicant confuses the values.

B is wrong (see above) but plausible if applicant believes it is less than the 50 psia for pressure with CS pumps available C is correct (see above)

D is wrong (see above) but plausible if confuse the pressure for both CS pumps available Technical

References:

OP-902-008, rev 30, page 35.

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 3

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 000074 Inadequate Core Cooling Ability to determine and interpret the following as they apply to (EPE 74) INADEQUATE CORE COOLING: Relationship between RCS temperature and main steam pressure Tier #

1 Group #

2 K/A #

EA2.04 Rating 3.7 Question 48 With both SG pressures at 650 PSIA, which ONE of the following conditions indicate core uncovery during a SBLOCA?

A. Representative CET temperature 500°F; RCS pressure 700 PSIA B. Representative CET temperature 510°F; RCS pressure 750 PSIA C. Representative CET temperature 520°F; RCS pressure 800 PSIA D. Representative CET temperature 530°F; RCS pressure 900 PSIA Answer: C Explanation:

Note: the SG pressures are at equilibrium with main steam pressures. Per the Technical Guide to this EOP, Core uncovery and, therefore, superheat on the CETs, indicate an advanced phase in the approach to inadequate core cooling.

A is wrong because the RCS is not under super-heated conditions B is wrong because the RCS is not under super-heated conditions C is correct because the RCS is under super-heated conditions.

D is wrong because the RCS is not under super-heated conditions Technical

References:

Steam tables TG-OP-902-002, Revision 20, page 198.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE02, revision 22, enabling Objective 14.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2008 NRC Q62 New Question History:

Last NRC Exam No

Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 000074 Natural Circulation (CE A13) Natural Circulation: Ability to interpret and execute procedure steps.

Tier #

1 Group #

2 K/A #

G2.1.20 Rating 4.6 Question 49 The following plant conditions exist:

A grid disturbance has caused a loss of offsite power The crew has completed the applicable actions of OP-902-001, Reactor Trip Recovery Procedure The crew is currently performing actions in accordance with OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery In accordance with the procedural steps of OP-902-003, which of the following is ONE indication that there is single phase natural circulation in at least one loop?

A. RCS subcooling of 25°F based on representative CET temperature B. Loop T of 60°F C. Hot and Cold Leg temperature constant D. Thot and representative CET temperature T of 15°F Answer: C Explanation:

A is wrong because subcooling must be at least 28°F; plausible as it is a positive value close to the setpoint B is wrong because loop delta T needs to be less than 58°F; plausible because positive loop delta T could be considered indicative of NC flow C is correct D is wrong because delta T needs to be less than 10°F. Plausible as 15 is close to the 10 setpoint Technical

References:

OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery, Rev. 11, Page 15 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPEO5, EO-2 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:

55.41(b) 14

Examination Outline Cross-Reference Level RO 062 AC Electrical Distribution AC ELECTRICAL DISTRIBUTION SYSTEM: Knowledge of electrical power supplies to the following: Major system loads Tier #

2 Group #

1 K/A #

K2.01 Rating 3.3 Question 50 Heater Drain Pump C is powered from which bus?

A. 2A B. 2B C. 4A D. 4B Answer: A Explanation:

A is correct.

B is incorrect but plausible because bus 2B is a non-safety 4160 volt bus.

B is incorrect but plausible because bus 4A is a non-safety 4160 volt bus.

B is incorrect but plausible because bus 4B is a non-safety 4160 volt bus.

Technical

References:

OP-003-034, Feed Heater Vents and Drains, Revision 025, Page 136 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-ED00, Objective 3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 063 DC Electrical Distribution DC ELECTRICAL DISTRIBUTION SYSTEM Knowledge of electrical power supplies to the following: Major DC loads Tier #

2 Group #

1 K/A #

K2.01 Rating 2.9 Question 51 Which of the following pumps are powered by 125VDC buses?

1. Diesel Generator A Standby Booster Fuel Oil Pump

2. Diesel Generator A Circulating Jacket Water Pump

3. Generator Air Side Seal Oil Back Up Pump

4. Generator H2 Side Seal Oil Pump

5. FWPT Emergency Lube Oil Pump A

6. Main Turbine Emergency Bearing Oil Pump A. 2, 4, 5, 6 B. 1, 3, 4, 5 C. 1, 3, 5, 6 D. 2, 3, 4, 6 Answer: C Explanation:

1, 3, 5, and 6 are DC motors. 2 and 4 are AC motors, but are plausible because they are in systems that have a complicated balance between AC and DC powered auxiliary support systems.

A is wrong because 2 and 4 are AC components. 5 and 6 are DC components.

B is wrong because 4 is an AC component. 1, 3 and 5 are DC components.

C is correct.

D is wrong because 2 and 4 are AC components. 3 and 6 are DC components Technical

References:

SD-DC, 125V DC Electrical Distribution, Rev 12 References to be provided to applicants during exam: None.

Learning Objective: [Could not locate relevant learning objective]

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:

55.41(b)7

Examination Outline Cross-Reference Level RO 073 Process Radiation Monitoring System Knowledge of the effect that a loss or malfunction of the PROCESS RADIATION MONITORING SYSTEM will have on the following: Radioactive effluent releases Tier #

2 Group #

1 K/A #

K3.01 Rating 3.6 Question 52 If power supply PDP 345AB fails to 0V, which of the following process radiation monitors will automatically isolate discharge flow?

A. Gaseous Waste Management Extended Range Monitor B. Fuel Handling Building Normal Exhaust Monitor C. Component Cooling Water System Liquid Monitors D. Industrial Waste Sump Turbine Building Liquid Monitor Answer: A Explanation:

A is correct because this monitor is powered by PDP 345AB and will isolate on a loss of power supply to the monitor. Two other monitors are powered from the same power supply and will auto isolate discharge flow. Liquid Waste Management Liquid Monitor and Boron Management System Liquid Monitor.

B is wrong because it is not powered by the power supply and has no auto isolation on loss of power or high activity.

C is wrong because it is not powered from the power supply and has no auto isolation on loss of power or high activity.

D is wrong because it is not powered from the power supply and has no auto isolation on loss of power. This monitor does auto isolate on high activity until an operator manually completes the flow path to the Reactor Auxiliary Building Waste Tanks.

Technical

References:

SD-RMS, Radiation Monitoring System, Rev 15, p. 14 References to be provided to applicants during exam: None.

Learning Objective: STATE vital and/or non-vital power supplies to process/area radiation monitors and RMS console. [EO-12]

Summarize the effect of a degradation of the RMS on radioactive effluent releases. [EO-4]

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No

Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 076 Service Water System Ability to (a) predict the impacts of the following on the SERVICE WATER SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Service water header pressure Tier #

2 Group #

1 K/A #

A2.02 Rating 2.7 Question 53 The following conditions exist:

the plant had a LOCA heat loads exceeded the DCT capacity ACCW is being prepared to cool the CCW heat exchangers The ACCW jockey pump was not running Subsequently, the following alarm annunciates:

AUX CCW SYS A PRESS LO/JOCKEY PUMP TRIP/TROUBLE (K-9)

What is the response of the system and what procedure is used to verify the expected system response?

A. 1) When ACC header pressure reaches 5 psig, ACC Pump A will auto start

2) OP-901-510, Component Cooling Water System Malfunction B. 1) When ACC header pressure reaches 10 psig, ACC Pump A will auto start

2) OP-901-510, Component Cooling Water System Malfunction C. 1) When ACC header pressure reaches 5 psig, ACC Pump A will auto start

2) OP-500-002, Control Room Cabinet B D. 1) When ACC header pressure reaches 10 psig, ACC Pump A will auto start

2) OP-500-002, Control Room Cabinet B Answer: C Explanation:

A is wrong because part 2 is wrong. Part 1 is correct per alarm response procedure OP-500-002, for the reported alarm, window K-9. The alarm comes in at 10 psig and if pressure lowers to 5 psig on ACC header pressure then the A ACC pump will auto start.

B is wrong because part 1 and part 2 are wrong (see above). Plausible because this is the alarm setpoint not pump start setpoint and could believe the off-normal has something for this in it, but it does not.

C is correct (see above discussion)

D is wrong because part 1 is incorrect.

Technical

References:

OP-500-002, revision 38, window K-9.

WLP-OPS-CC00, revision 43, page 192 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CC00 Obj E08 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:

55.41(b)5

Examination Outline Cross-Reference Level RO 078 Instrument Air System Ability to monitor automatic operations of the INSTRUMENT AIR SYSTEM including: Air pressure Tier #

2 Group #

1 K/A #

A3.01 Rating 3.1 Question 54 Given the following conditions:

100 percent power Instrument Air Header pressure is 120 psig An air leak occurs, causing Instrument Air Header pressure to drop to 100 psig and stabilizes.

What is one way the control room can verify proper response of instrument air components to this condition?

A. Verify SA-125 Station Air Backup is OPEN by verifying that the associated alarm Inst Air Press Back-up Vlv Open is illuminated on CP-36 B. Verify SA-125 Station Air Backup is OPEN by verifying that the associated alarm Inst Air Press Back-up Vlv Open is illuminated on CP-1 C. Verify SA-123 Air Dryer Bypass is OPEN by verifying that the associated alarm Inst Air Dryers By-Passed is illuminated on CP-36 D. Verify SA-123 Air Dryer Bypass is OPEN by verifying that the associated alarm Inst Air Dryers By-Passed is illuminated on CP-1 Answer: A Explanation:

A is correct because SA-125 opens at 105 psig and this pressure has been exceeded at 100 psig. The location for the alarm is correct on CP-36.

B is wrong because the alarm location is wrong (it is on CP-36)-Plausible because some alarms are on CP-1 for IA and the instrument air header pressure instrument is located on CP-1 as well.

C is wrong because SA-123 opens at 95 psig and this pressure has not been reached. The alarm for this valve being open is on CP-36. Plausibility is same as distracter B.

D is wrong because SA-123 opens at 95 psig and this pressure has not been reached. CP-1 is correct.

Technical

References:

SD-Air, Revision 14, page 50 OP-003-016, Revision 28, pg 27 WLP-OPS-AIR00, revision 26

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-AIR00 obj. 01.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)4

Examination Outline Cross-Reference Level RO 103 Containment System Ability to monitor automatic operations of the CONTAINMENT SYSTEM including: Containment isolation Tier #

2 Group #

1 K/A #

A3.01 Rating 3.9 Question 55 Initial Conditions Reactor is tripped.

RCS Pressure is 1680 psia Containment Pressure is 18 psia Refueling Water Storage Pool level is at 10%

SI-120A, SI RECIRC HEADER A TO RWSP ISOLATION is SHUT CS-125B, CS HEADER ISOLATION VALVE is OPEN SI-602A, SAFETY INJECTION SUMP ISOLATION is OPEN SI-343, SI TANKS DRAIN is OPEN Which valve requires operation action given plant conditions?

A. SI-120A B. CS-125B C. SI-602A D. SI-343 Answer: D Explanation:

A is wrong because a Recirculation Actuation Signal (RAS) is generated when RWSP level is at 10%. Once automatic actions from a RAS is verified, one of the valves the operator must shut is SI-120A B is wrong because CS-125B will automatically open on a Containment Spray Actuation Signal (CSAS). A CSAS signal is generated when a Safety Injection Actuation Signal (SIAS)

AND a High-High Containment Pressure signal are coincident. SIAS is satisfied because pressure is greater than 17.1 psia AND RCS pressure is less than 1684 psia. High High Containment Pressure is met since containment pressure is greater than 17.7 psia.

C is wrong because when a RAS signal is generated, SI-602A is automatically opened.

D is correct because a Containment Isolation Actuation Signal is generated when pressure is 1684 psig or lower OR containment pressure is 17.1 psia or higher. This valve should have automatically shut and is out of position given plant conditions.

Technical

References:

SD-CB, Containment Building and Containment Isolation, Rev. 11, pg 18.

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 #

X (Q#27, 2008 Exam)

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.1.19) Ability to use plant computers to evaluate system or component status.

Tier #

3 Group #

K/A #

G2.1.19 Rating 3.9 Question 56 The Core Operating Limit Supervisory System Reactor Power analog instrument is suspect on CP-2 and operators need to verify it with the plant computer. To check this, they would type _____as the first letter, A.

G B.

K C.

A D.

C Answer: D Explanation:

A. Incorrect, plausible because it is for analog outputs to PMC driven meters.

B. Incorrect, plausible because it is used for COLSS constants.

C. Incorrect, plausible because it is for analog.

D. Correct.

Technical

References:

SD-PMC, Plant Monitoring Computer, Revision 12 Page 49 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PMC00, Plant Monitoring Computer, Revision 13, EO1 - Summarize the design features that provide for PMC Data Point Processing and Site Specific Programs.

Question Source:

Bank #

2010 RO 67 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2010 RO 67 Question Cognitive Level:

Memory/Fundamental 3

Comprehensive/Analysis

10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.1.21) Ability to verify the controlled procedure copy.

Tier #

3 Group #

K/A #

G2.1.21 Rating 3.5 Question 57 EN-AD-103, Document Control and Records Management Programs, allows for work to be performed without verifying the controlled procedure copy current revision when:

A. Hard copies located in an emergency facility are used during an emergency B. Hard copies located in the field are used during an emergency C. A site procedure has outstanding changes that were verified in Asset Suite (AS)

D. A site procedure has outstanding changes that were verified in Echelon Electronic Book Reference Library (ECH eB REFLIB)

Answer: A Explanation:

A is correct because EN-AD-103 allows for in an emergency or simulated emergency, only hard copy controlled documents or Disaster Recovery (DR) computers located in the Emergency Facilities may be used without verifying the current revision in EDMS/AS.

B is wrong because EN-AD-103 only allows for use for procedures are in a emergency facilities not the field. Plausible because the applicant may focus on the emergency nature of the use not the location of the procedure.

C is wrong because EN-AD-103 directs the use of Asset Suite for verification of outstanding changes to controlled documents, plausible because the applicant may not understand the difference between revision verification and verification of outstanding changes.

D is wrong because EN-AD-103 directs use of ECH eB REFLIB for verification of Nuclear Management Manual procedures not site procedures, plausible because the applicant may not understand the difference between revision verification and verification of outstanding changes.

Technical

References:

EN-AD-103, Document Control and Records Management Programs, Revision 24 References to be provided to applicants during exam: None.

Learning Objective:

WPPT-OPS-PPA00, Administrative Procedures, Revision 15, EO-1, Given plant and/or equipment conditions, identify applicable administrative procedures.

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:

55.41(b)10

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.1.8) Ability to coordinate personnel activities outside the control room.

Tier #

3 Group #

K/A #

G2.1.8 Rating 3.4 Question 58 The unit is in Mode 5 following a refueling outage; startup preparations are underway. You are the at-the-controls (ATC) reactor operator. An auxiliary operator (AO) calls the control room and informs you that while conducting a valve lineup, a manually operated valve was found to be stiff and requires the use of a torque amplifying device.

Per EN-OP-115-05, Operation of Components, what coordination/direction do you give the AO?

A. Torque amplifying devices are not allowed without an engineering change B. As a qualified AO, the use of a torque amplifying device is at their discretion C. You, as ATC, may grant permission to use the torque amplifying device D. SM/CRS approval is required to use a torque amplifying device Answer: D Explanation:

A is wrong because EN-115-05 has a CAUTION which states the use of torque amplifying devices may be acceptable in certain situations; is plausible because the use of such devices on power operated components are not allowed without an engineering change.

B is wrong because step 5.5.2 states that SM/CRS approval is required; is plausible because section 4.0 of EN-115-05 states that Non-Licensed Operators are responsible for positioning of components as authorized by approved procedures, clearance requests or at SM/CRS direction.

C is wrong because step 5.5.2 states that SM/CRS approval is required; is plausible because section 4.0 of EN-115-05 states that Reactor Operators are responsible for positioning of components as authorized by approved procedures, clearance requests or at SM/CRS direction.

D is correct because EN-115-05 has a CAUTION which states the use of torque amplifying devices may be acceptable in certain situations, and section 5.5.2 states that if such a device is required then operators must obtain SM/CRS approval.

Technical

References:

EN-115-05, Operation of Components, rev 3, page 14 References to be provided to applicants during exam: None.

Learning Objective: WPPT-OPS-PPA00, Administrative Procedures Enabling Objective 3: From memory, DESCRIBE and APPLY procedural steps, cautions, and notes of Informational Use procedures as applicable to assigned operational duties. [EO-3]

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 001 Control Rod Drive System (001A2.13) Ability to (a) predict the impacts of the following on the (SF1 CRDS) CONTROL ROD DRIVE SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: ATWS Tier #

2 Group #

2 K/A #

A2.13 Rating 4.4 Question 59 In response to an anticipated transient without scram, operators depress both Reactor Trip push buttons on CP-2. Based on this action, Reactor Trip Circuit Breakers _____1_____ should open. If this action is unsuccessful, the operators depress the Diverse Reactor Trip System push button which will open the control element drive mechanism control system motor generator set _____2_____.

A. 1. 1,4,5,8

2. load contactors B. 1. 2,3,6,7

2. load contactors C. 1. 1,4,5,8

2. supply breakers D. 1. 2,3,6,7

2. supply breakers Answer: A Explanation:

A is correct because the CP-2 pushbuttons open TCB 1,4,5,8. The DRTS operates to open the CEDM control system MG set load contactors.

B is wrong because the CP-2 pushbuttons actuate TCB 1,4,5,8. Plausible because the CP-8 push buttons open 2,5,6,7. Part 2 is correct.

C is wrong because the DRTS operates to open the CEDM control system MG set load contactors. Plausible because opening the supply breakers is a step in OP-902-008. Part 1 is correct.

D is wrong because the CP-2 pushbuttons open TCB 1,4,5,8. The DRTS operates to open the CEDM control system MG set load contactors. Plausible because the CP-8 push buttons open 2,5,6,7 and opening the supply breakers is a step in OP-902-008.

Technical

References:

SD-CED, Control Element Drive, Revision 12, page 18

SD-ATS, Anticipated Transient System, Revision 5, page 7 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CED00, Control Element Drive, revision 24, EO4 - Summarize the operation and function of : Reactor Trip Breakers, including Controls, Location and Operation of CEDM MG Sets and Control Panel, including trips Question Source:

Bank #

NRC 2009 Q9 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2009 Question Cognitive Level:

Memory/Fundamental 3

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)5

Examination Outline Cross-Reference Level RO 002 Reactor Coolant System (002K5.12) Knowledge of the operational implications of the following concepts as they apply to the REACTOR COOLANT SYSTEM:

Relationship of temperature average and loop differential temperature to loop hot-let and cold-leg temperature indications Tier #

2 Group #

2 K/A #

K5.12 Rating 3.7 Question 60 While operating at 100% power with the T Cold Loop 1 switch selected to the NORM position and the Tave Loop Selector Switch selected to the Both 1 and 2 position, RC-ITE-0115, 1-B cold leg, fails low. Operators would expect RC-ITI-0111Y, Dual Cold Leg Temperature Indicator, to indicate _____1_____ and the Tave calculated by the Reactor Regulating System to indicate _____2_____.

A. 1. correctly

2. correctly B. 1. low

2. correctly C. 1 correctly

2. low D. 1. low

2. low Answer: A Explanation:

A is correct because with the T Cold Loop 1 switch in the NORM position the 1-A cold leg temperature is provided to the RC-ITI-011Y, Dual Cold Leg Temperature Indicator, for display so it would display correctly. It also provides the 1-A cold leg temperature to the RRS so the Tave calculated by the RRS would be correct.

B is wrong because the T Cold Loop 1 switch in the NORM position the 1-A cold leg temperature is provided to the RC-ITI-011Y, Dual Cold Leg Temperature Indicator, for display so it would display correctly. Plausible if the applicant does not understand the functions of the T cold loop selector switch in the NORM and ALT positions. Part 2 is correct.

C is wrong because the T Cold Loop 1 switch in the NORM position the 1-A cold leg temperature is provided to the RRS so the Tave calculated by the RRS would be correct.

Plausible if the applicant does not understand the functions of the T cold loop selector switch in the NORM and ALT positions. Part 1 is correct.

D is wrong because the T Cold Loop 1 switch in the NORM position the 1-A cold leg temperature is provided to the RC-ITI-011Y, Dual Cold Leg Temperature Indicator, for display so it would display correctly. It also provides the 1-A cold leg temperature to the RRS so the Tave calculated by the RRS would be correct. Both are plausible if the applicant does not understand the functions of the T cold loop selector switch in the NORM and ALT positions.

Technical

References:

SD-RR, Reactor Regulating, Revision 7, pages 7-9 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-RR00, Reactor Regulating System, Revision 12, EO2 - Summarize the operation and function of: Tavg calculator and Temperature Input Selector Switches 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:

55.41(b)5

Examination Outline Cross-Reference Level RO 011 Pressurizer Level Control System (011A4.03) PRESSURIZER LEVEL CONTROL SYSTEM Ability to manually operate and/or monitor in the control room:

PZR heaters Tier #

2 Group #

2 K/A #

A4.03 Rating 3.3 Question 61 Given:

• Plant is at 100% power

• The Pressurizer Level Control Channel Select Switch is in the Channel Y position

• The ATC reports that the Pressurizer Level Hi/Lo and Pressurizer Level Lo/Lo annunciators are in alarm

• One Charging Pump is running

• Pressurizer level as read on RC-ILI-110X indicates 0%

Normal control of pressurizer heaters can be restored by:

A. 1. Taking the Pressurizer Lo Level Cutoff Switch to the CHANNEL Y position

2. Manually resetting the Proportional Heaters by taking the control switch to ON B. 1. Taking the Pressurizer Lo Level Cutoff Switch to the CHANNEL X position

2. Manually resetting the Proportional Heaters by taking the control switch to ON C. 1. Taking the Pressurizer Lo Level Cutoff Switch to the CHANNEL Y position

2. Manually resetting the Proportional Heaters by taking the control switch to AUTO D. 1. Taking the Pressurizer Lo Level Cutoff Switch to the CHANNEL X position

2. Manually resetting the Proportional Heaters by taking the control switch to AUTO Answer: A Explanation:

A is correct because once it is determined that the non-selected level channel has failed low, the operator need only position the Pressurizer heater Lo Level Cutout switch to the channel Y position to regain normal operation of the Backup Pressurizer heaters.

The Proportional heaters will have to be reset manually by taking the control switches to ON.

B is wrong because this would select the faulted channel, plausible if applicant believes this switch cutoff the selected channel. Part 2 is correct.

C is wrong because Proportional Heaters are reset by taking the switch to ON, plausible because AUTO is a position for the Backup Heater control switch. Part 1 is correct.

D is wrong because this would select the faulted channel, plausible if applicant believes this switch cutoff the selected channel. Part 2 is wrong because Proportional Heaters are reset by taking the switch to ON, plausible because AUTO is a position for the Backup Heater control switch.

Technical

References:

SD-PLC, Pressurizer Level and Pressure Control, Revision 11 Pages 23-24 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PLC00, Pressurizer Level Control, Revision 23, EO5 - Summarize the operation and functions of: Pressurizer Heaters Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2017 RO 21 New Question History:

Last NRC Exam 2017 RO 21 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO 014 Rod Position Indication System (014K5.01) Knowledge of the operational implications of the following concepts as they apply to the (SF1 RPI) ROD POSITION INDICATION SYSTEM: Reasons for differences between RPIS and step counter Tier #

2 Group #

2 K/A #

K5.01 Rating 2.7 Question 62 Given:

Plant is performing a startup and is at 80% power Group P CEAs are being withdrawn for ASI control Group Select switch is in the P position CEA 25 in Group P becomes mechanically bound With the CEA Manual Shim switch in WITHDRAW, PMC pulse counter indication for CEA 25 is ____(1)____. CEAC position indication for CEA 25 is ______(2)____.

(1)

(2)

A.

rising stationary B.

stationary stationary C.

rising rising D.

stationary rising Answer: A Explanation:

A is correct because the PMC pulse counters will be rising as long as the shim switch is taken to withdraw. The CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

B is wrong because the PMC pulse counters will be rising as long as the shim switch is taken to withdraw. Part 2 is correct, the CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

C is wrong because the CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs. Part 1 is correct, the PMC pulse counters will be rising as long as the shim switch is taken to withdraw.

D is wrong, because the PMC pulse counters will be rising as long as the shim switch is taken to withdraw. The CEA position used in the CEACs work on reed switch position indicators and will see actual CEA position. Therefore, the CEACs will show a stationary position for CEAs.

Technical

References:

SD-PMC, Plant Monitoring Computer, Revision 12, Pages 37-38 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CED00, Control Element Drive, Revision 24, EO17 - Summarize the operation and function of: Purpose and Operation of Sensors feeding into the CEDMCS,Location and Operation of CEA position indication system Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2012 RO58 New Question History:

Last NRC Exam 2012 RO58 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)5

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.2.12) Knowledge of surveillance procedures.

Tier #

3 Group #

K/A #

G2.2.12 Rating 3.7 Question 63 EN-WM-110, Surveillance Program, requires certain notifications be made if a surveillance test cannot be completed within its specified frequency or applicable mode. Which of the following personnel DOES NOT have to be notified?

A. Shift Manager/Control Room Supervisor B. Surveillance Coordinator C. Discipline Supervisor D. Operations Manager Answer: D Explanation:

A is wrong because EN-WM-110 requires notification of the Discipline Supervisor, Surveillance Coordinator, and SM/CRS. Plausible because applicant may believe that the Operations Manager must also be notified.

B is wrong because EN-WM-110 requires notification of the Discipline Supervisor, Surveillance Coordinator, and SM/CRS. Plausible because applicant may believe that the Operations Manager must also be notified.

C is wrong because EN-WM-110 requires notification of the Discipline Supervisor, Surveillance Coordinator, and SM/CRS. Plausible because applicant may believe that the Operations Manager must also be notified.

D is correct because EN-WM-110 requires notification of the Discipline Supervisor, Surveillance Coordinator, and SM/CRS.

Technical

References:

EN-WM-110, Surveillance Program, Revision 2, Page 16 References to be provided to applicants during exam: None.

Learning Objective:

WPPT-OPS-PPA00, Administrative Procedures, Revision 15, EO2 - From memory, identify Operations responsibilities.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.2.38) Knowledge of conditions and limitations in the facility license.

Tier #

3 Group #

K/A #

G2.2.38 Rating 3.6 Question 64 Per the Waterford 3 Operating License, Entergy Operations Inc. is authorized to operate the Waterford 3 reactor core at power levels not to exceed:

A. 3716 MWt B. 3739 MWt C. 3761 MWt D. 3793 MWt Answer: A Explanation:

A is correct because this is the value listed in the current oper. license post power uprate.

B. is wrong because this is the value listed in the current Operating License post power uprate + the most recent evaluation of RCP heat contribution. of 23 MW.

C. is wrong because the last two numbers are transposed with the answer A D. is wrong because the last two numbers are transposed from the value of current operating License post power uprate + the most recent evaluation of RCP heat contribution.of 23 MW...

Technical

References:

Waterford Facility License, Amendment 258, page 4.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-TS00 Obj. 2.

Question Source:

Bank #

2010 NRC Q69 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)2

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.2.43) Knowledge of the process used to track inoperable alarms.

Tier #

3 Group #

K/A #

G2.2.43 Rating 3.0 Question 65 In accordance with EN-OP-0115-08, an inoperable alarm is removed from service (disabled) due to a malfunction in the circuit. This action is required to be tracked by making an entry into

___1____ and placing a ___2___ on the alarm panel.

A. (1) annunciator logbook (2) red dot B. (1) annunciator logbook (2) remove from service marker C. (1) watch station deficiency database (2) red dot D. (1) watch station deficiency database (2) remove from service marker Answer: B Explanation:

The procedure was recently changed from an OI procedure to the new EN-OP-115-08 procedure. The old procedure required an entry into the watch station deficiency database with a red dot over the window while the new procedure requires entry into the annunciator logbook and a remove from service marker placed on the alarm window.

A is wrong because part 2 is wrong. Part 1 is correct.

B is correct because (see above explanation).

C is wrong because both parts are wrong, and this was the previous method to perform this activity with the old procedure.

D is wrong because Part 1 is wrong. Part 2 is correct.

Technical

References:

EN-OP-115-08, revision 6, page 13 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 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 027 Containment Iodine Removal System CONTAINMENT IODINE REMOVAL SYSTEM Ability to manually operate and/or monitor in the control room: CIRS fans Tier #

2 Group #

2 K/A #

A4.03 Rating 3.3 Question 66 Loss of 480 VAC MCC 313A(B) would result in the loss of _____1_____ which is controlled from _____2_____.

A. 1. Containment Fan Cooler A(B)

2. CP-35 B. 1. Airborne Radioactivity Removal System Exhaust Fans A(B)

2. CP-18 C. 1. Containment Fan Cooler A(B)

2. CP-33 D. 1. Airborne Radioactivity Removal System Exhaust Fans A(B)

2. CP-33 Answer: B Explanation:

A is wrong because Containment Fan Coolers are powered from 480 VAC MCC 317A(B)-S, plausible because of being 480 VAC and part of the containment cooling system similar to the ARRS Exhaust Fans. CFC A(B) is controlled at CP-18, not CP-35, plausible because the power lost annunciator for CFC (B) is at CP-35.

B is correct because 480 VAC MCC 313 A(B) powers ARRS A(B) and they are controlled at CP-18.

C is wrong because Containment Fan Coolers are powered from 480 VAC MCC 317A(B)-S, plausible because of being 480 VAC and part of the containment cooling system similar to the ARRS Exhaust Fans. CFC A(B) is controlled at CP-18, not CP-33, plausible because the power lost annunciator for CFC (A) is at CP-33.

D is wrong because CFC A(B) is controlled at CP-18, not CP-33, plausible because the unit trouble annunciator for CFC (A) is at CP-33.

Technical

References:

SD-CCS, Containment Cooling System, Revision 8, pages 21-23 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CCS00, Containment Cooling System, Revision 20, EO1 - Summarize the operation and functions of: Airborne Radioactivity Removal System 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:

55.41(b)7

Examination Outline Cross-Reference Level RO 033 Spent Fuel Pool Cooling System Ability to predict and/or monitor changes in parameters associated with operating the SPENT FUEL POOL COOLING SYSTEM controls including: Radiation monitoring systems Tier #

2 Group #

2 K/A #

A1.02 Rating 2.8 Question 67 Which of the following radiation monitors will generate a Fuel Handling Accident Signal upon detecting activity in excess of the setpoint?

1. Fuel Handling Building Area Radiation Monitors (RE0300.1S/.2S/.3S/.4S)

2. Fuel Handling Building Normal Exhaust Monitors (RE5107A/B)

3. Fuel Handling Building Emergency Exhaust Monitor (RE3032)

A. 1 ONLY B. 2 ONLY C. 1 and 3 ONLY D. 2 and 3 ONLY Answer: A Explanation:

The Fuel Handling Ventilation System provides for cooling of the SFP Cooling/Purification equipment. Should a fuel handling accident occur, radiation monitors would produce a fuel handling accident signal. This will isolate the Fuel Handling Building and start the safety related emergency filtration exhaust fans.

Only the Fuel Handling Building ARMs (Isolation Monitors) generate a Fuel Handling Accident signal upon a HI-HI radiation alarm. This sends signals to start the associated train of emergency filtration for the FHB Ventilation system while shutting the isolation dampers and securing the normal ventilation system.

The Fuel Handling Building Normal Exhaust Monitor (RE5107A and B) are PIGs and give operating personnel indications of Fuel Pool Ventilation system activity levels. Each of these will generate alarms the Control Room at CP-36 if setpoints are reached. These alarm conditions would indicate a low level leakage problem and alert operators to conduct additional surveys to locate the source of the leakage. They DO NOT generate a Fuel Handling Accident Signal.

FHB Emergency Exhaust Monitor (RE3032) will generate an alarm in the Control Room at CP-36 when a release rate of 4.00 x 106 Ci/sec is detected. It DOES NOT generate a Fuel Handling Accident Signal.

A is correct.

B is wrong because The Fuel Handling Building Normal Exhaust Monitor (RE5107A and B) generate an alarm, but do not generate a fuel handling accident signal.

C is wrong because Fuel Handling Building Emergency Exhaust Monitor (RE3032) generates an alarm, but do not generate a fuel handling accident signal.

D is wrong because both The Fuel Handling Building Normal Exhaust Monitor (RE5107A/B) and Fuel Handling Building Emergency Exhaust Monitor (RE3032) generate an alarm, but do not generate a fuel handling accident signal.

Technical

References:

SD-RMS Rev 15, Radiation Monitoring System SD-FS Rev 13, Fuel Pool Cooling and Purification SD-HVF Rev 11, Fuel Handling Building HVAC References to be provided to applicants during exam: None.

Learning Objective: Fuel pool cooling EO-4: SUMMARIZE the effects of a degradation of the Fuel Pool Cooling and Purification system on the: HVF System, RMS System 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:

55.41(b)11

Examination Outline Cross-Reference Level RO 035 Steam Generator System Knowledge of the physical connections and/or cause-effect relationships between STEAM GENERATOR SYSTEM and the following:

MRSS Tier #

2 Group #

2 K/A #

K1.02 Rating 3.2 Question 68 If main steam pressure is 1120 psia, a MAXIMUM of (1) steam generator code safety valves should be open, per steam generator.

A. 2 B. 3 C. 4 D. 5 Answer: C Explanation:

There are six code safety valves per steam generator, and their lift setpoints are 1070, 1085, 1100, 1115, 1125 and 1135 psia. Four is the correct answer since the pressure is above 1115 psia but below 1125 psia. The other distractors are plausible since the question is essentially knowledge of the safety valve setpoints.

Technical

References:

Technical Specification 3.7.1, Table 3.7-1, Amendment 142 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-MS00, Revision 34, Objective 2 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.3.13) Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Tier #

3 Group #

K/A #

G2.3.13 Rating 3.4 Question 69 Which of the following is classified as a Very High Radiation Area with no access allowed in MODE 1?

A.

Area outside the Regen HX Cubicle with Letdown in service B.

-4ft MSL area around the Reactor Drain Tank C.

Reactor Vessel Annulus D.

Pressurizer cubicle Answer: C Explanation:

A is wrong because the dose rates in this area would be high with Letdown in service but will not be a VHRA.

B is wrong because the background radiation in this area is also high but would not be a VHRA.

C is correct because HP-001-213 step 5.2.2 states that the following areas have been designated as Very High Radiation Areas. Entries into these areas are forbidden when the reactor is in MODE 1: (Hot and Cold Leg D ring wall penetrations, Reactor Vessel Annulus, and the Reactor Cavity.)

D is wrong because the dose rates in the Pressurizer cubicle will be high but will not be a VHRA.

Technical

References:

HP-001-213, Control of Reactor Containment Building Power Entries, Revision 306, Page 9 References to be provided to applicants during exam: None.

Learning Objective:

WPPT-OPS-PPA00, Administrative Procedures, Revision 15, EO-3, From memory, describe and apply procedural steps, cautions, and notes of Informational Use procedures as applicable to assigned operational duties.

Question Source:

Bank #

2012 RO73 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2012 RO73 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.41(b)12

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.3.7) Ability to comply with radiation work permit requirements during normal or abnormal conditions.

Tier #

3 Group #

K/A #

G2.3.7 Rating 3.5 Question 70 With the plant in Mode 1, which ONE of the following conditions is a requirement to enter an area with only a Self Brief and a General Radiation Work Permit (vice requiring a Specific Radiation Work Permit and RP brief)?

A. An area assessed as Low Risk Work B. An area with airborne rates less than 5 DAC C. An area with dose rate fields less than 25 mrem/hour D. An area with contamination fields less than 100,000 dpm/100 cm2 Answer: C Explanation:

A is wrong because low risk work must range between 25mrem/hr and 1000 mrem/hr and cannot be self-briefed unless it is less than 25 mrem/hr.

B is wrong because DAC must be below 1 DAC for medium risk work and would not allow either aspects (self brief or GA work permit)

C is correct because self brief low risk work requires dose rate fields less than 25 mrem/hr.

D is wrong because contamination fields must be less than 10,000 dpm/cm2, not 100,000 dpm/cm2.

Technical

References:

EN-RP-110-04, revision 8, pg 3, 8, and 9 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPA00 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 2

10CFR Part 55 Content:

55.41(b)12

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.4.17) Knowledge of EOP terms and definitions.

Tier #

3 Group #

K/A #

G2.4.17 Rating 3.9 Question 71 When BOTH Steam Generators (SG) have a leak (either a fault or a rupture), OI-038-000, EOP Operations Expectations/Guidance Procedure, defines MOST AFFECTED SG as A. The SG with the largest tube rupture B. The SG with the largest ESD C. The SG with the largest effective leak rate D. The SG that does not have an EFW pump and/or ADV Answer: C Explanation:

Most affected SG is defined as the SG with the largest effective leak rate when both SGs are believed to have a SGTR, both have an ESDE, or one has a SGTR and one has an ESDE.

A is wrong because (see above definition). Plausible if conclude the release to environment is more important and therefore is most affected.

B is wrong because (see above definition). Plausible if conclude the steam pipe failure is more important due to uncontrolled cooldown and usually larger than a tube rupture and is therefore the most affected.

C is correct (see above)

D is wrong because although this is important and will impact cooldown and is a big decision for the definition, this is not correct.

Technical

References:

OI-038-000, Emergency Operating Procedures Operations Expectations / Guidance, Revision 19, Page 4 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE01, Emergency Operating Procedures, Revision 17, EO2 - Define specific terms used in Function Based EOPs.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO No System (Tier 3)

(G2.4.30) Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Tier #

3 Group #

K/A #

G2.4.30 Rating 2.7 Question 72 Per Emergency Plan Procedure EP-002-010, as the ENS communicator you are normally required to maintain continuous communications with the NRC at A. A NOUE or higher B. An Alert or higher C. A Site Area Emergency or higher D. A General Emergency Answer: B Explanation:

A is wrong because it is at An Alert or higher....

B is correct per the reference section 5.2.4, page 20.

C is wrong because it is at An Alert or higher.

D is correct because it is at An Alert or higher.

Technical

References:

EPP-002-010, revision 316, page 20 WLP-OPS-EP00 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-EP00 Obj. 12.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 041 Steam Dump/Turbine Bypass Control System Knowledge of the effect that a loss or malfunction of the STEAM DUMP/TURBINE BYPASS CONTROL SYSTEM will have on the following: Reactor power Tier #

2 Group #

2 K/A #

K3.04 Rating 3.5 Question 73 The following plant conditions exist:

• The plant is at 75% power

• A loss of Computer SUPS caused a spurious operation of the SBCS which resulted in MS-320A opening and remaining OPEN. MS-319A was out of service for maintenance at the time of the event.

Disregarding the effects of Moderator Temperature and Doppler coefficients, which ONE of the following reactor power levels will be the approximate resulting plant output?

A. 65%

B. 75%

C. 85%

D. 95%

Answer: C Explanation:

Modified by changing stem to only one valve going open and this also changes answer.

A. Incorrect: One valve is rated for 9.88% power output in addition to 75% by MT. This is 85%. Might believe that it takes 10% power away from reactor so 65% is plausible.

B. Incorrect: One valve is rated for 9.88% plus 75% by MT or 85%. Plausible because might think the turbine cuts back and power stabilizes back where it was before at 75%.

C. Correct: One valve is rated for 9.88% each for power output in addition to 75% by MT.

This is 85% and is the correct answer.

D. Correct: One valve is rated for 9.88% plus power output added to 75% by MT is 85%.

Plausible if though it was 20% per valve or memorized original question where 95%

was original answer.

Technical

References:

SD-SBC, Steam Bypass Control, Revision 10 OP-901-312, Loss of Vital Instrument Bus, Revision 319 References to be provided to applicants during exam: None.

Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2009 (Q61)

New Question History:

Last NRC Exam Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)5

Examination Outline Cross-Reference Level RO 056 Condensate System CONDENSATE SYSTEM: Ability to perform specific and integrated plant procedures during all modes of operation.

Tier #

2 Group #

2 K/A #

G2.1.23 Rating 4.3 Question 74 Given the following:

A plant startup is in progress.

Condensate Pump A was started from a cold condition, run for 20 minutes at rated speed, then manually tripped 3 minutes ago.

Investigation revealed an erroneous indication, and the pump may be run as required.

In accordance with OP-003-003, Condensate System, which of the following describes the time requirement for restart of Condensate Pump A?

A. The pump may be started ONE time immediately.

B. An additional MINIMUM of 2 minutes must pass prior to attempting restart.

C. An additional MINIMUM of 27 minutes must pass prior to attempting restart.

D. An additional MINIMUM of 57 minutes must pass prior to attempting restart.

Answer: D Explanation:

A is wrong but plausible if confused with cold start requirements B is wrong but plausible if thought 5 minute wait time required C is wrong but plausible if thought a 30 minute wait time required or confused with no wait time required if the pump was running at rated speed for 30 minutes.

D is correct because if pump is run at rated speed for less than 30 minutes, a 60 minute wait time is required Technical

References:

OP-003-003 Revision 312, Precaution 3.1 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CD00, EO-3 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

X

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.41(b)10

Examination Outline Cross-Reference Level RO 075 Circulating Water System Knowledge of CIRCULATING WATER SYSTEM design feature(s) and or interlock(s) which provide for the following: Heat sink Tier #

2 Group #

2 K/A #

K4.01 Rating 2.5 Question 75 Initially there are no Circulating Water (CW) Pumps running. An operator is directed to start the A and B CW Pumps. Once the A CW pump is running with its discharge valve (CW-103A) fully open, the operator places the B CW switch to the START position. The operators notes that B CW pump has not started. Why?

A. A 15-minute delay is required to prevent overloading the pump on startup B. B CW pump discharge valve must be 90% open before the pump starts C. B CW pump discharge valve must be 20% open before the pump starts.

D. The operator needs to wait at least 100 seconds after CW-103A is fully open Answer: D Explanation:

A is wrong because the 15 min delay only occurs with the start of the first CW pump.

Additionally, the 15 min delay is an administrative hold and does not provide an actual interlock.

B is wrong because the 90% discharge valve position does trip the associated pump off.

However, this is for when the pump control switch is taken to the STOP position, not the START position.

C is wrong because the 20% valve position is when the pump would start. However, the valve would not move since the 100 sec interlock has not been satisfied.

D is correct because there is a time delay of 100 seconds that prevents the starting of a second pump until the previous pumps discharge valve has been fully opened.

Technical

References:

SD-CW, Circulating Water, page 16, rev. 18 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CW00 PowerPoint Presentation, Identify the interlocks which provide for CW Pump and Discharge Valve Operations [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 3

10CFR Part 55 Content:

55.41(b)7

Examination Outline Cross-Reference Level SRO 006 Emergency Core Cooling (006A2.02) Ability to (a) predict the impacts of the following on the EMERGENCY CORE COOLING SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of flow path Tier #

2 Group #

1 K/A #

A2.02 Rating 4.3 Question 76 The following plant conditions exist:

LOCA occurred 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago Crew is performing OP-902-002, Loss of Coolant Accident Recovery Recirculation Actuation Signal has occurred, and required actions have been taken RCS is 15°F subcooled; CET temperatures are rising very slowly QSPDS RVLMS levels 1-6 indicates void BOP notes that HPSI A and B pumps cold leg flow has dropped off slowly with discharge pressure rising and no oscillations HPSI Cold Leg flow criteria is met Containment Spray Pumps A and B are running with flow ~ 2000 GPM each train and stable The CRS should ______(1)______. The conditions given indicate _____(2)_____.

A. 1. remain in OP-902-002 and align hot and cold leg injection

2. core flow blockage is starting to occur due to boron precipitation B. 1. secure one HPSI pump and go to OP-902-008, Functional Recovery

2. core flow blockage is starting to occur due to boron precipitation C. 1. remain in OP-902-002 and align hot and cold leg injection

2. the SI Sump Strainers are starting to clog from debris collection D. 1. secure one HPSI pump and go to OP-902-008, Functional Recovery

2. the SI Sump Strainers are starting to clog from debris collection Answer: A Explanation:

A is correct because the indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps. At 2-3 hours post LOCA the crew should align hot and cold leg injection to avoid concentrating boric acid in the reactor vessel and possible boron precipitation in the core which could restrict coolant flow through the core...

B is wrong because Core flow blockage indications are given making Part 2 correct. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

C is wrong because The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers).

Step 53 of OP-902-002 directs performance of Hot and Cold Leg Injection 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> post-LOCA. Either condition listed in Part 2 is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

D is wrong because Both parts are incorrect. The indications given indicate a problem developing on the discharge side of the High Pressure Safety Injection Pumps, not the suction side (SI Sump Strainers). Step 47 of OP-902-002 directs going down to one HPSI pump operating to prevent heat buildup and subsequent pump seal damage in low flow conditions. This condition is addressed in OP-902-002, and therefore, OP-902-008 entry is not required.

Technical

References:

OP-902-002, Loss of Coolant Accident Recovery, Revision 021, Page 47 TGOP-902-002, Technical Guide for Loss of Coolant Accident Recovery, Revision 020, Pages 118-119 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure with Transient and Accident Analysis, Revision 22, EO19 - Objective 19, State the criteria required and the basis for each of the following operations in the LOCA procedure, OP-902-002: Hot and Cold Leg Injection.

Question Source:

Bank #

2012 NRC Q11 (note changes; attach parent)

Modified Bank #

New 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 007 pressurizer Relief/Quench Tank System (007A2.02) Ability to (a) predict the impacts of the following on the PRESSURIZER RELIEF/QUENCH TANK SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Abnormal pressure in the PRT Tier #

2 Group #

1 K/A #

A2.02 Rating 3.2 Question 77 The following plant conditions exist:

The plant is at 100% power.

Pressurizer Safety Valve, RC-317B, is leaking at 1.5 gpm.

Quench Tank temperature and pressure are both rising.

OP-901-111, Reactor Coolant System Leak would direct Operators to utilize procedure

_____1_____ to _____2_____.

A. 1. OP-003-019, Nitrogen System

2. Vent the Quench Tank nitrogen blanket to reduce pressure B. 1. OP-002-011, Primary Makeup System

2. Fill and drain the Quench Tank to reduce temperature C. 1. OP-007-001, Boron Management

2. Vent the Quench Tank nitrogen blanket to reduce pressure D. 1. OP-007-001, Boron Management

2. Fill and drain the Quench Tank to reduce temperature Answer: D Explanation:

A is wrong because OP-901-111, directs if leakage into Quench Tank is indicated by rising level OR temperature, then fill and drain Quench Tank to reduce temperature in accordance with OP-007-001, Boron Management. Plausible because initial nitrogen blanket pressurizes quench tank to 3 psig and venting it could reduce pressure but is not directed in OP-901-111.

B is wrong because OP-901-111, directs if leakage into Quench Tank is indicated by rising level OR temperature, then fill and drain Quench Tank to reduce temperature in accordance with OP-007-001, Boron Management. Plausible because the PMU system is used in the the fill and drain procedure but OP-007-001 contains the procedure steps. Part 2 is correct.

C is wrong because OP-901-111, directs if leakage into Quench Tank is indicated by rising level OR temperature, then fill and drain Quench Tank to reduce temperature in accordance with OP-007-001, Boron Management. Plausible because initial nitrogen blanket pressurizes

quench tank to 3 psig and venting it could reduce pressure but is not directed in OP-901-111.

Part 1 is correct.

D is correct because OP-901-111, directs if leakage into Quench Tank is indicated by rising level OR temperature, then fill and drain Quench Tank to reduce temperature in accordance with OP-007-001, Boron Management.

Technical

References:

OP-901-111, Reactor Coolant System Leak, Revision 305, Page 11 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPO10, Off-Normal Procedures Primary Systems, Revision 20, EO4 - interpret content, location and sequencing of procedural steps applicable to the diagnosed event.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

Similar stem to 2006 RO87 New Question History:

Last NRC Exam Similar stem to 2006 RO87 Question Cognitive Level:

Memory/Fundamental 3

Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 010 Pressurizer Pressure Control System PRESSURIZER PRESSURE CONTROL SYSTEM): Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and derivatives affect plant and system conditions.

Tier #

2 Group #

1 K/A #

G2.2.44 Rating 4.4 Question 78 A turbine control system malfunction has caused the turbine to runback from 100% power.

The following indications are present two minutes after the turbine stops running back:

REACTOR COOLANT Tave-Tref HI alarm in SELECTED COLD LEG 1 TEMPERATURE HI alarm in PRESSURIZER PRESSURE HI/LO alarm in Tcold is 534°F PZR Pressure is 2277 psia All PZR Heaters have de-energized automatically Pressurizer pressure controller output is 0%

Pressurizer spray valve controller output is 0%

The CRS will address pressurizer pressure control by entering OP-901-120, Pressurizer Pressure Control Malfunction, subsection ____(1)____ and enter the Tech Spec action(s) for

_____(2)_____.

____(1)____ _______(2)________

A. E2, Pressurizer Pressure pressurizer pressure (TS 3.2.8) only Controller Malfunction B. E3, Pressurizer Spray pressurizer pressure (TS 3.2.8) only Valve Malfunction C. E2, Pressurizer Pressure pressurizer pressure (TS 3.2.8) and Controller Malfunction reactor coolant Tcold temp (TS 3.2.6)

D. E3, Pressurizer Spray pressurizer pressure (TS 3.2.8) and Valve Malfunction reactor coolant Tcold temp (TS 3.2.6)

Answer: C Explanation:

A is wrong because part 2 is wrong. PZR pressure controller output should be high (i.e.

>75%) for the given conditions. The applicant must also recognize that a 0% output from the pressure controller will input into the spray valve controller and prevent the spray controller

from responding in automatic. Tcold temperature is not within the TS LCO band of 536-549°F therefore part 2 is incorrect (both pressure and temperature LOCs apply)

B is wrong because part 1 and part 2 are wrong. The spray valve controller will not respond in automatic because it is receiving a 0% signal from the master controller. The CRS will enter E2 which direct actions to place both the PZR pressure and spray valve controllers in manual and control pressure manually. Part 2 is wrong as stated in A above.

C is correct because part 1 is the correct section (it is a controller malfunction as stated above) and part 2 is correct because Tcold temperature is not within the TS LCO band of 536-549°F and pressurizer pressure is also exceeding the TS LCO band.

D is wrong because part 1 is wrong. The spray valve controller will not respond in automatic because it is receiving a 0% signal from the master controller. The CRS will enter E2 which direct actions to place both the PZR pressure and spray valve controllers in manual and control pressure manually. Part 2 is correct (see above discussion).

Technical

References:

OP-901-120, Pressurizer Pressure Control Malfunction, Revision 303.

TS 3.2.6 and TS 3.2.8, Amendment 249 SD-PLC, Revision 11.

WLP-OPS-PLC00, Revision 20 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PLC00 Objectives 4 and 5 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2015 SRO Q13 New Question History:

Last NRC Exam 2015 (SRO Q13)

Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 013 Engineered Safety Features Actuation System (013 G2.2.38) Engineered Safety Features Actuation System: Knowledge of conditions and limits in the facility license Tier #

2 Group #

1 K/A #

G2.2.38 Rating 4.5 Question 79 Per the Waterford 3 Operating License, Entergy Operations, Inc. is required to have an operable Engineered Safety Features Actuation System to assist the Reactor Protection System in preventing the plant from exceeding the below Specified Acceptable Fuel Design Limit of A. fuel centerline temperature shall be maintained less than 5080ºF B. departure from nucleate boiling ratio shall be maintained greater than 1.14 C. planar heat rate shall be maintained less than 23 kW/ft D. integrated radial peaking factor low shall be maintained greater than 1.08 Answer: A Explanation:

A is correct because it is one of the two SAFDLs and the temperature of 5080ºF is correct.

B is wrong because the nucleate boiling ration must be maintained above 1.24; is plausible because it is one of the SAFDLs that PPS prevents and 1.14 may be confused with 1.24.

C is wrong because planar heat rate is not one of the SAFDLs that PPS (RPS and ESF) prevents; is plausible because linear heat rate is limited to 21 kW/ft (close to 23 kW/ft).

D is wrong because it is not one of the SAFDLs that PPS prevents; is plausible because the CPCs DNBR low algorithm does have an integrated radial peaking factor limit of 1.28.

Technical

References:

TS bases, page B2-1, B2-2, B2-2a, B2-6, References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPS00, rev 17, Plant Protection System Enabling Objective 7: SUMMARIZE the effect of a degradation of the RPS or ESFAS on the following: (EO-7)

Fuel Reactor Coolant System (RCS)

Containment

Control Element Drive Mechanism System 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)1

Examination Outline Cross-Reference Level SRO 026 Containment Spray System (026A2.07) Ability to (a) predict the impacts of the following on the CONTAINMENT SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff (interlock) limit Tier #

2 Group #

1 K/A #

A2.07 Rating 3.9 Question 80 The following plant conditions exist:

LOCA occurred 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago Crew is performing OP-902-002, Loss of Coolant Accident Recovery Recirculation Actuation Signal has occurred, and required actions have been taken RCS is 15°F subcooled; CET temperatures are rising very slowly Containment Pressure is 28psia and slowly lowering BOP notes that HPSI A and B pumps are running with flow ~ 400 GPM each train and stable Containment Spray Pump A is running with flow and amp oscillations and varying suction and discharge pressure Containment Spray Pump B is running with flow ~ 1200 GPM and stable The CRS is evaluating step 49 for Monitor for Loss of ECCS Pump Suction.

The CRS should ____________.

A. Secure both CS pumps only and go to OP-902-008, Functional Recovery B. Secure 1 HPSI pump and CS pump A only and remain in OP-902-002, LOCA Recovery C. Secure both CS pumps only and remain in OP-902-002, LOCA Recovery D. Secure 1 HPSI pump and both CS pumps and go to OP-902-008, Functional Recovery Answer: D Explanation:

KA involves cavitation and loss of suction (clogged sump) while in recirculation mode.

Step 49a states to secure 1 HPSI and 1 CS pump due to erratic or low flow and then check results. In this case, the erratic CS pump is pump A so it would be secured first and then one HPSI pump secured. However, the second CS pump is be required to be secured IAW step 49c. With both CS pumps secured, step 49f states Check at least ONE CS pump operating

acceptably by stable parameters and flow greater than 1750 gpm. With no CS pumps operating the functional procedure is required.

A is wrong (see above discussion).

B is wrong because you secure both CS pumps and 1 HPSI pump for these conditions and do not stay in the LOCA procedure. (see above).

C is wrong because you secure both CS pumps and 1 HPSI pump and do not remain in the LOCA procedure but transition to the FRP D is correct because you secure both CS pumps and 1 HPSI pump and go to the FRP.

Technical

References:

OP-902-002, Rev 21, pages 44-45 TG-902-002, rev 20, step 49 WLP-OPS-PPE02, rev 22 WLP-OPS-PPE08, rev 16 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE08, Obj E01 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)5

Examination Outline Cross-Reference Level SRO 000009 (EPE 9) Small Break LOCA (000009EA2.01) Ability to determine and interpret the following as they apply to (EPE 9)

SMALL BREAK LOCA: Actions to be taken, based on RCS temp and press, saturated and superheated Tier #

1 Group #

1 K/A #

EA2.01 Rating 4.8 Question 81 Given:

• A Small Break LOCA has occurred

• SIAS, CIAS, and MSIS are present

• The crew is performing actions of OP-902-002, Loss of Coolant Accident Recovery Procedure After the LOCA is isolated, the ATC informs the CRS that the 200°F subcooled line of the RCS Pressure and Temperature Curve (PT curve) is being exceeded.

To restore the RCS to within the limits of the PT curve, the CRS will direct the crew to take actions to lower RCS ____1____.

After the RCS is restored to within the limits of the PT curve, the crew will maintain the RCS within the curve by re-energizing the 32A and 32B busses in accordance with OP-902-009, Standard Appendices, Appendix ______2______.

A. 1. Temperature

2. 12, Electrical Restoration B. 1. Pressure

2. 12, Electrical Restoration C. 1. Pressure

2. 25, Restore Pressurizer Heater Control D. 1. Temperature

2. 25, Restore Pressurizer Heater Control Answer: C Explanation:

A is wrong because applicant is required to know exceeding the upper limit of the PT curve means that the RCS is over sub-cooled. OP-902-002 step 30 requires the crew to stop the

cooldown and lower pressure. Appendix 12 has various sections for restoring electrical power, but guidance to close in the 32 feeder breakers with a SIAS is located in Appendix 25 B is wrong because Appendix 12 has various sections for restoring electrical power, but guidance to close in the 32 feeder breakers with a SIAS is located in Appendix 25. Part 1 is correct.

C is correct because if the upper limit of the PT curve is exceeded, the concern is PTS. To restore the RCS to within limits, the crew will need to lower RCS pressure. OP-902-002 step 30 directs the crew to Appendix 25, Restore Pressurizer Heater Control guidance for energizing the 32A(B) busses and restoring pressurizer heaters.

D is wrong because the applicant is required to know exceeding the upper limit of the PT curve means that the RCS is over sub-cooled. OP-902-002 step 30 requires the crew to stop the cooldown and lower RCS pressure. Part 2 is correct.

Technical

References:

OP-902-002, Loss of Coolant Accident Recovery, Revision 21, page 27 TG-OP-902-002, Technical Guide for Loss of Coolant Accident Recovery Procedure, Revision 20, pages 71-72 OP-902-009, Standard Appendices, Revision 319, Attachment 25, all pages References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE02, Loss of Coolant Accident Recovery Procedure with Transient and Accident Analysis, EO17 - Given plant and/or equipment conditions, identify and apply procedural steps, cautions, and notes applicable to the LOCA procedure or to plant and/or equipment conditions which may require entry into the LOCA procedure, OP-902-002.

Question Source:

Bank #

2017 NRC SRO 76 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2017 NRC SRO 76 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 000025 Loss of Residual Heat Removal System (000025 (APE 25) Loss of Residual Heat Removal System: (G2.2.40) Ability to apply Tech Specs for a system.

Tier #

1 Group #

1 K/A #

G2.4.40 Rating 4.7 Question 82 In Mode 5 with the reactor coolant loops NOT filled, with A and B low pressure safety injection pumps running and both shutdown cooling loops in operation, A low pressure safety injection pump fails. TS LCO 3.4.1.5 would require action _____1_____.

The basis for this LCO in these plant conditions is _____2_____.

A. 1. immediately

2. to ensure proper boron mixing, prevent stratification of the RCS, and produce gradual reactivity changes during boron concentration reductions.

B. 1. immediately

2. based on single failure considerations and the unavailability of the steam generators for heat removal C. 1. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2. to ensure proper boron mixing, prevent stratification of the RCS, and produce gradual reactivity changes during boron concentration reductions.

D. 1. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2. based on single failure considerations and the unavailability of the steam generators for heat removal Answer: B Explanation:

A is wrong because based on the given conditions, the basis for TS LCO 3.4.1.5 states that single failure considerations and the unavailability of steam generators for heat removal require two shutdown cooling trains to be operable. Plausible because the basis for TS LCO 3.4.1.5 also states that one shutdown cooling loop needs to be in operation for ensuring mixing, preventing stratification of the RCS, and producing gradual reactivity changes during boron concentration reductions. Part 1 is correct.

B is correct because based on the given conditions, TS LCO 3.4.1.5 requires immediate corrective action to restore the required loops to operable status. The basis for TS LCO 3.4.1.5 states that single failure considerations and the unavailability of steam generators for heat removal require two shutdown cooling trains to be operable.

C is wrong because based on the given conditions, TS LCO 3.4.1.5 requires immediate corrective action to restore the required loops to operable status. Also, the basis for TS LCO 3.4.1.5 states that single failure considerations and the unavailability of steam generators for

heat removal require two shutdown cooling trains to be operable. Part 1 is plausible because TS LCO 3.4.1.5 allows a shutdown cooling (LPSI) pump to be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when other specific conditions are met (these conditions are not met based on the given information). Part 2 is plausible because the basis for TS LCO 3.4.1.5 also states that one shutdown cooling loop needs to be in operation for ensuring mixing, preventing stratification of the RCS, and producing gradual reactivity changes during boron concentration reductions.

D is wrong because based on the given conditions, TS LCO 3.4.1.5 requires immediate corrective action to restore the required loops to operable status. Plausible because TS LCO 3.4.1.5 allows a shutdown cooling (LPSI) pump to be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when other specific conditions are met (these conditions are not met based on the given information). Part 2 is correct.

Technical

References:

Waterford 3 Technical Specifications, RCS Cold Shutdown - Loops Not Filled, LCO 3.4.1.5, Page 3/4 4-6, Amendment 249 Waterford 3 Technical Specification Bases, 3/4/4/1 Reactor Coolant Loops and Coolant Circulation, Page B 3/4 4-1, Change 34 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-SDC, Shutdown Cooling, Revision 30, EO-6, Given plant status, identify and evaluate the operability, actions and bases for the following Technical Specifications as they relate to the Shutdown Cooling System: 3/4/4/15.

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)2

Examination Outline Cross-Reference Level SRO 000025 Pressurizer Pressure Control System Malfunction (000027 (APE 27) Pressurizer Pressure Control System Malfunction: (G2.2.36) 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 #

G2.2.36 Rating 4.2 Question 83 Given:

Unit at full power Proportional heater bank 1 has one heater out of service that will be replaced in the next outage Maintenance was attempting to troubleshoot the issue and tripped the heater breakers for all six banks of backup heaters.

Maintenance wont be able to get the breakers closed for the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> now.

The crew took timely actions to take manual control of spray valves and all available heaters Pressurizer pressure lowered to 2220 psia and stabilized at this pressure.

Based on these conditions and with respect to Technical Specifications ONLY, the CRS is required to take action to A. reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to LCO 3.2.8 for pressurizer pressure B. restore pressurizer pressure to within the limits specified with the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to LCO 3.2.8 for pressurizer pressure C. restore at least one bank of backup heaters to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> due to LCO 3.4.3.1 for pressurizer D. be in at least HOT STANDBY within the next six hours with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Answer: C Explanation:

LCO 3.2.8 requires pressurizer pressure to go below 2125 psia and it did not so this TS LCO is met. It is the value where the abnormal procedure OP-901-120 is entered for a pressurizer pressure low event. The value in the stem is above this value in the TS so A and B distracters are not correct. For LCO 3.4.3.1, it requires at least two banks that have 150kw EACH, and with only 2 banks on Prop 1 heaters (which is 100kw of power) it is considered inoperable and so a backup heater set is needed with 150kw of power to meet this TS. Since it is not, then C is the correct answer. D is credible but also not correct because you have 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to meet LCO condition A of 3.4.3.1 and distracter D is condition D of this LCO.

A is wrong because this LCO is not applicable for this pressure...

B is wrong because this LCO is not applicable for this pressure C is correct because (see above)

D is wrong because (see above discussion)

Technical

References:

TS 3/4-4.3, amendment 249, page 3/4-4.9 TS 3.2.8, amendment 249, page 3/4-2.13 WLP-OPS-PLC00, revision 23, slides 110-111 WLP-OPS-PPO10, revision 20, slides 115-116 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPO10, Objective E08.

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 000038 Steam Generator Tube Rupture (000038EA2.08) Ability to determine and interpret the following as they apply to (EPE

38) STEAM GENERATOR TUBE RUPTURE /

3 (CFR: 43.5 / 45.13): Viable alternatives for placing plant in safe condition when condenser is not available Tier #

1 Group #

1 K/A #

EA2.08 Rating 4.4 Question 84 The Plant was operating at 100% when a reactor trip is initiated due to a Steam Generator Tube Rupture in Steam Generator 1 OP-902-007, Steam Generator Tube Rupture Recovery was entered When the Control Room Supervisor reaches the step to implement the Placekeeper, a Loss of Offsite Power occurs Emergency Diesel Generator B trips on Overspeed A rapid cooldown to Thot of 520°F will be performed in accordance with (1), using the Atmospheric Dump Valve(s) on (2) Steam Generator(s).

A. (1) OP-902-008, Functional Recovery (2) both B. (1) OP-902-008, Functional Recovery (2) the unaffected C. (1) OP-902-007, Steam Generator Tube Rupture Recovery (2) both D. (1) OP-902-007, Steam Generator Tube Rupture Recovery (2) the unaffected Answer: C Explanation:

A is incorrect. Part 1 is plausible if one believes that with the loss of offsite power, the Maintenance of Vital Auxiliaries safety function would not be satisfied, and the Functional Recovery procedure would be entered. Part 2 is correct.

B is incorrect. Part 1 is plausible for the reason stated in distractor A. Part 2 is plausible if one believes you do not cooldown using a ruptured steam generator.

C is correct.

D is incorrect. Part 1 is correct. Part 2 is plausible for the reason stated in distractor c.

Technical

References:

OP-902-007, Steam Generator Tube Rupture Recovery, Revision 018, Step 11.1, Page 12 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source:

Bank #

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam 2011, Q77 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 000055 Station Blackout (000055 (EPE 55) Station Blackout) (G2.1.7)

Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 43.5)

Tier #

1 Group #

1 K/A #

G2.1.7 Rating 4.7 Question 85 Given the following:

0800 - EDG A is tagged out for maintenance 0900 - All offsite power is lost 0900 - EDG B starts and trips on overspeed 0910 - CRS enters OP-902-005, Station Blackout RCS pressure is 2175 psia and rising Pressurizer level is 33% and rising All other equipment operated as designed Offsite power recovery time is not known The Temporary Diesel would not start due to water in the fuel

1) What is the required time to strip battery loads to meet the coping time at Waterford?

2) What procedure directs this requirement?

A. 1. At time 0930

2. TGOP-902-005 B. 1. At time 0930

2. OI-038-000 C. 1. At time 0940

2. TGOP-902-005 D. 1. At time 0940

2. OI-038-000 Answer: B Explanation:

25 minutes is a TCA for getting a temporary diesel if power cannot be restored but 30 minutes from start of the event is the required time (or 0900 plus 30 is 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />). Also, since it is from the start of the event (not when you enter the procedure) it is from 0900 not 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br /> when SPTAs are estimated to be complete (or 0900 plus 30 is 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />). This requirement is in procedure OI-038-000. Page 19 of the TGOP-902-005 states that the note is not in the EPG but was added to the EOP as a deviation at Waterford.

A is wrong because part 2 is incorrect (see above).

B is correct (see above)

C is wrong because both parts are wrong (see above)

D is wrong because part 1 is wrong. Part 2 is correct (see above).

Technical

References:

OI-038-000, Revision 19, page 37.

TGOP-902-005 SBO TG, revision 311, page 19.

WLP-OPS-PPE05, revision 34, slide 88 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE05 E03, E07.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 000062 Loss of Nuclear Service Water (000062AA2.03) Ability to determine and interpret the following as they apply to the (APE 62) LOSS OF NUCLEAR SERVICE WATER: The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition Tier #

1 Group #

1 K/A #

AA2.03 Rating 2.9 Question 86 Per the Waterford 3 FSAR, Section 9.2.5.3, Ultimate Heat Sink, Safety Evaluation:

1) In the event of a tornado strike to the Dry Cooling Towers, damage to the cooling coils will cause the Dry Cooling Tower to be automatically isolated and bypassed based on

2) With the DCTs bypassed, Wet Cooling Tower basins have sufficient water storage capacity to remove the heat from the vital services for approximately A. 1) CCW Surge Tank level lowering

2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. 1) CCW Surge Tank level lowering

2) 30 days C. 1) CCW pressure lowering

2) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> D. 1) CCW pressure lowering

2) 30 days Answer: A Explanation:

FSAR section 9.2.5.3, in discussing design basis tornado, states that, In the unlikely event of a tornado strike to the dry cooling tower, damage by the tornado missiles to the dry cooling tower coils is automatically detected by a decrease in the CCW surge tank water level and the dry cooling towers are automatically bypassed. Wet cooling tower basins have sufficient water storage capacity to remove the heat from the vital services for approximately two hours. This time period allows the operator to isolate the damaged dry cooling tower cells and put the operable dry cooling tower cells back into service. DCTs bypass and isolate at 20% CCW Surge Tank level, lowering.

Separately, FSAR section 9.2.5.3 in discussing the design basis of the Ultimate Heat Sink (which is the DCTs, WCTs, and WCT Basin), states: Capacity: The ultimate heat sink is designed, and sufficient water is stored in two wet cooling tower basins accounting for fuel pool cooling, to provide sufficient cooling for 30 days or longer and to permit safe

shutdown and cooldown of the plant and maintain it in a safe shutdown condition without additional make-up water. To conserve water, the system is so designed that the dry cooling towers are operating at full capacity post accident.

A is correct. See above.

B is wrong because with the DCTs bypassed, the WCT basins can only provide design basis cooling for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 30 days is plausible because that is the design basis cooling timeframe with DCTs available.

C is wrong because DCTs bypass and isolate based on surge tank level, not CCW pressure.

Plausible because the related ACCW system pressure low causes automatic actions, such as pump start.

D is wrong because with the DCTs bypassed, the WCT basins can only provide design basis cooling for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 30 days is plausible because that is the design basis cooling timeframe with DCTs available. Also, DCTs bypass and isolate based on surge tank level, not CCW pressure. Plausible because the related ACCW system pressure low causes automatic actions, such as pump start.

Technical

References:

Waterford 3 FSAR Rev 312, Section 9.2.5.3.3, Site Phenomena.

WLP-OPS-CC00 Rev 43, Component Cooling Water References to be provided to applicants during exam: None.

Learning Objective: SUMMARIZE the operation and functions of the following major equipment. INCLUDE interlocks: Dry Cooling Towers A and B [EO-3]

Given plant status, PREDICT the response of the CCW System to leakage anywhere in the system. [EO-7]

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)1

Examination Outline Cross-Reference Level SRO 000024 Emergency Boration Emergency Boration: (G2.2.37) Ability to determine operability and/or availability of safety related equipment Tier #

1 Group #

2 K/A #

G2.2.37 Rating 4.6 Question 87 Given:

Plant is performing a cooldown for a refueling outage RCS temperature is 195 ºF RCS pressure is 350 psia EDG A is tagged out Offsite power train B is declared inoperable because of a voltage problem.

The CRS will verify ________ is aligned as the operable boration flow path per OP-903-002, Boration Flow Path Valve Lineup Verification.

A. the RWSP and HPSI pump A B. the RWSP and HPSI pump B C. BAM pump A and Charging pump A D. BAM pump B and Charging pump B Answer: B Explanation:

A is wrong. Even though HPSI pump A is operable and available to be used as a boration flow path, offsite train A cannot be credited as the operable power source...

B is correct because OP-903-002 is the surveillance procedure used to verify SR 4.1.2.1. TS 3.1.2.1 bases specifies that the operable power source must be an EDG. HPSI pump B is the only component that could be powered from EDG B. TS bases knowledge is required to determine the correct boration flow path.

C is wrong because BAM pump A and Charging pump A are operable and available but offsite train A cannot be credited as the operable power source.

D is wrong because BAM pump B is powered from safety bus A and therefore EDG B cannot be credited as the operable power source.

Technical

References:

OP-903-002 Boration Flow Path Valve Lineup Verification, Revision 304, Page 18 TS 3.1.2.1 & bases, Amendment 249, Page 3/4 1-6 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-CVC00, Chemical and Volume Control System, Revision 1-7, LO8 - From memory, identify and evaluate the operability, actions and bases for the following Technical Specifications, which require action in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less 3.1.2.1, Boration Systems, Flow Paths -Shutdown Question Source:

Bank #

2015 NRC Q8 (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2

10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 000036 Fuel Handling Accidents Ability to determine and interpret the following as they apply to the Fuel handling accidents:

Magnitude of potential radioactive release Tier #

1 Group #

2 K/A #

AA2.03 Rating 4.2 Question 88 During refueling operations, a spent fuel bundle is dropped to the Refueling Cavity floor, resulting in numerous area and effluent radiation alarms.

What action contained in OP-901-405, Fuel Handling Incident, has the HIGHEST priority to minimize an uncontrolled release?

A. Start Airborne Radioactivity Removal System B. Ensure Fuel Handling Building exterior and Cargo Train Bay doors closed C. Close the Personnel and Escape Air Locks D. Close and secure the Equipment Hatch Answer: D Explanation:

A is wrong because this is referenced late in OP-901-405 and is less effective in limiting the release of radioactivity to the environment than securing Containment.

B is wrong because this is referenced early in OP-901-405, but is ineffective for a dropped assembly in the Containment.

C is wrong because this is part of the Containment Closure Checklist referenced in OP-901-405 but directed after closing the equipment hatch.

D is correct because this action is directed in OP-901-405 prior to directing operators to complete the Containment Closure Checklist per OP-901-131.

Technical

References:

OP-901-405, Fuel Handling Incident, Revision 8, Page 11 WLP-OPS-PPO40, Off Normal Operating Procedures Radiation Monitoring, Revision 11, Slide 35 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPO40, Off Normal Operating Procedures Radiation Monitoring, Revision 11, EO3 - Apply procedural steps, cautions, and notes applicable to the diagnosed event.

Question Source:

Bank #

Similar to 2006 SRO91 (note changes; attach parent)

Modified Bank #

New 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 CE/A11 RCS Overcooling Ability to determine and interpret the following as they apply to the (RCS Overcooling)

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Tier #

1 Group #

2 K/A #

AA2.2 Rating 3.4 Question 89 Given:

Plant was operating at 100% power with EDG A danger tagged An Excess Steam Demand occurred SIAS, CIAS, and MSIS have been initiated Pressurizer level is 0%

After entering OP-902-004, Excess Steam Demand Recovery, the following conditions change:

Representative CET temperature and RCS pressure start to rise.

A Loss of Off-Site Power occurs All components respond as designed to the event.

Based on these conditions, the CRS should:

A. Stay in OP-902-004, Excess Steam Demand Recovery, only B. Exit OP-902-004, Excess Steam Demand Recovery, and Enter OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery Procedure C. Exit OP-902-004, Excess Steam Demand Recovery, and Enter OP-902-008, Functional Recovery Procedure D. Stay in OP-902-004, Excess Steam Demand Recovery, and Use Standard Attachment 13 to Stabilize RCS Temperature Answer: A Explanation:

A is correct because exit conditions for OP-902-004 do not exist and with one safety bus still having power this procedure still has priority.

B is wrong because although loss of offsite power has occurred this is not an exit criteria from OP-902-004.

C is wrong because although loss of offsite power has occurred this is not an exit criteria from OP-902-004 nor complicated enough to meet entry criteria for the Function Recovery procedure OP-902-008.

D is wrong because the Attachment 13 is not needed here-stabilize RCS Pressure would be next and the steps are contained within the OP-902-004 procedure. This appendix is not

mentioned in OP-902-004 but is plausible because many of the standard appendices are used in this procedure but this is not one of them.

Technical

References:

OP-902-004 Rev 17, page 4 References to be provided to applicants during exam: None Learning Objective: WLP-OPS-PPE04 obj. 7 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

2014 NRC Q86 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 CE A16 Excess RCS leakage CE A16 Excess RCS leakage: (G2.2.25)

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Tier #

1 Group #

2 K/A #

G2.2.25 Rating 4.2 Question 90 The technical bases for the surveillance requirements for RCS pressure isolation valves within Technical Specification 3.4.5.2 is _____1______ and the operational leakage from these valves is considered ______2______.

A. 1. Reduce the probability of an Intersystem LOCA

2. UNIDENTIFIED LEAKAGE B. 1. Reduce the probability of a Large Break LOCA

2. UNIDENTIFIED LEAKAGE C. 1. Reduce the probability of an Intersystem LOCA

2. IDENTIFIED LEAKAGE D. 1. Reduce the probability of a Large Break LOCA

2. IDENTIFIED LEAKAGE Answer: C Explanation:

Per TS bases page B 3/4 4-4e, Change 86, the SR requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from these valves is considered IDENTIFIED LEAKAGE.

A is wrong because part 2 is wrong. Plausible if confuse what type of leakage it is characterized as within the TS.

B is wrong because part 1 and part 2 are wrong. Part 1 plausible as discussed above. Part 2 plausible for LB LOCA because this is the biggest event for design considerations.

C is correct (see above).

D is wrong because part 1 is incorrect (see above for plausibility)

Technical

References:

Technical Specification Basis for 3.4.5.2, Change 86, page B 3/4 4-4e References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-RCS00 EO-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 3

10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO 015 Nuclear System Instrumentation System (015A2.05) Ability to (a) predict the impacts of the following on the NUCLEAR INSTRUMENTATION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Core void formation Tier #

2 Group #

2 K/A #

A2.05 Rating 3.8 Question 91 Given the following conditions:

The unit had a large break LOCA event several hours ago conditions have escalated in the RCS with inadequate core cooling CET temperatures are currently reading approximately 1100 degrees F.

The CRS is currently in the Functional Recovery Procedure (FRP) OP-902-008.

1) What are the impacts to nuclear instruments as read at the PMC?

2) What procedure should the CRS use to continue with this event (assuming the CRS is still the decision maker for procedures)?

A. 1. NIs will indicate negative values and BAD quality

2. Stay in OP-902-008 B. 1. NIs will indicate large positive values

2. Stay in OP-902-008 C. 1. NIs will indicate negative values and BAD quality

2. Exit FRP and enter SAMG-SAG-01 D. 1. NIs will indicate large positive values

2. Exit FRP and enter SAMG-SAG-01 Answer: A Explanation:

A is correct. Part 1 is correct per SD-NI, page 52-53, thermionic emissions become dominant over 1000 degrees F and polarity will shift to negative values and have BAD quality in the PMC. Part 2 is correct because you dont exit the EOP framework and enter SAMGs until CETs are greater than 1200 degrees F. You stay in the FRP. Large positive values for the NIs occur when fuel melting starts to occur, around 2400 degrees F B is wrong because part 1 is wrong (see above)

C is wrong because part 2 is wrong. Part 1 is correct (see above)

D is wrong because both parts are wrong. See above.

Technical

References:

SD-NI, Revision 8, pages 52-53 OP-902-008, Revision 30, page 121.

WLP-OPS-PPE08, Revision 16 References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-PPE08. Obj E05 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)5

Examination Outline Cross-Reference Level SRO 071 Waste gas Disposal System WASTE GAS DISPOSAL SYSTEM (G2.1.32)

Ability to explain and apply system limits and precautions.

Tier #

2 Group #

2 K/A #

G2.1.32 Rating 4.0 Question 92 Given:

Plant is operating at 100%

The crew needs to discharge at least one Gas Decay Tank (GDT)

The crew is using procedure OP-007-003, Gaseous Waste Management Per the precaution and limitations sections of this procedure, the normal methodology when writing the release permit is to:

A. discharge all three tanks simultaneously due to GDT inlet valve leakage B. discharge one tank at a time due to GDT inlet valve leakage C. discharge all three tanks simultaneously due to GDT outlet valve leakage D. discharge one tank at a time due to GDT relief valve leakage Answer: A Explanation:

A is correct per precaution and limitations step 3.2.4...

B is wrong because it is the inlet valve leakage that is of concern, but it is done simultaneously (all three) because of this issue. Plausible because you might think that it is more controlled with leakage to only to discharge one tank at a time.

C is wrong because it is not the outlet valves that are the concern, the inlet valves. Plausible if not familiar with the P and L and the CRs assigned to this procedure for leakage of these valves.

D is wrong because it is the inlet valves not the relief valve. Plausible if not familiar with the P and L and the CRs assigned to this procedure for leakage of these valves.

Technical

References:

OP-007-003, Gaseous Waste Management, Revision 310, step 3.2.4.

WLP-OPS-GWM00, revision 16.

References to be provided to applicants during exam: None.

Learning Objective: WLP-OPS-GWM00 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 2

10CFR Part 55 Content:

55.43(b)4

Examination Outline Cross-Reference Level SRO 079 Station Air System (079A2.01) Ability to (a) predict the impacts of the following on the STATION AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Cross-connection with IAS Tier #

2 Group #

2 K/A #

A2.01 Rating 3.2 Question 93 Operators noted instrument air system pressure is steady at 98 psig for approximately 1.5 hrs.

The CRS will direct the crew to _____1_____ in accordance with the guidance in _____2_____.

A. 1. cross connect backup air stations

2. OP-003-016, Instrument Air B. 1. open, IA-969, Instrument Air High Pressure Supply Station 1-2 Low Pressure Cross Connect

2. OP-003-016, Instrument Air C. 1. verify IA-123, Instrument Air Dryers Bypass Solenoid Valve, is open

2. OP-901-511, Instrument Air Malfunction D. 1. verify SA-125, Station Air to Instrument Air Cross-Connect, is open

2. OP-901-511, Instrument Air Malfunction Answer: D Explanation:

A is wrong because OP-501-511 directs verifying SA-125 is open or raising the setpoint to force it open. Plausible because OP-901-511 also directs action to cross connect backup air stations using OP-003-016 if backup air supply pressure is lowering due to a leak.

B is wrong because OP-501-511 directs verifying SA-125 is open or raising the setpoint to force it open. Plausible because applicant may not understand that you cannot use essential air banks as a supply to instrument air.

C is wrong because OP-501-511 directs verifying SA-125 is open or raising the setpoint to force it open. Plausible because IA-123 should open at 95 psig which is not the case here.

D is correct because OP-501-511 directs verifying SA-125 is open or raising the setpoint to force it open.

Technical

References:

OP-901-511, Instrument Air Malfunction, Revision 17 page 7 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-AIR00, Plant Air Systems, Revision 26, EO5, Given Plant Status, identify and apply procedural steps and requirements, including notes and cautions for the following procedures: OP-901-511, Instrument Air Malfunction 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)5

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.1.34) Knowledge of primary and secondary plant chemistry limits.

Tier #

3 Group #

K/A #

G2.1.34 Rating 3.5 Question 94 While operating in Mode 1, primary Hydrogen concentration is normally maintained

_____1_____ to _____2_____.

A. 1. 5 - 15 cc/kg

2. reduce the formation of nitric acid in the RCS B. 1. 5 - 15 cc/kg

2. control oxygen concentration in the RCS C. 1. 25 - 50 cc/kg

2. control oxygen concentration in the RCS D. 1. 25-50 cc/kg

2. reduce the formation of nitric acid in the RCS Answer: C Explanation:

A is wrong because the normal band is 25-50 cc/kg to control oxygen concentration.

Plausible because 5 - 15 cc/kg are the Action Level 2 and 3 concentrations and because reduction of nitric acid formation is a benefit of hydrogen concentration in the RCS.

B is wrong because the normal band is 25-50 cc/kg to control oxygen concentration.

Plausible because 15 - 25 cc/kg are the Action Level 2 and low end of Action Level 1 concentrations and part 2 is correct to control oxygen in the RCS.

C is correct because the normal band is 25-50 cc/kg to control oxygen concentration.

D is wrong because wrong because part 2 is wrong. The normal band is 25-50 cc/kg to control oxygen concentration. Plausible because the reduction of nitric acid formation is a benefit of hydrogen concentration in the RCS.

Technical

References:

CE-002-006, Maintaining Reactor Coolant Chemistry, Revision 320, Page 12 WLP-OPS-CHM03, Primary Chemistry, Revision 19, Slide 30 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-CHM03, Primary Chemistry, Revision 19, EO5 - State the effects, limits, and concerns of H2 in the Primary System.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental 2

Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.1.35) Knowledge of the fuel-handling responsibilities of SROs.

Tier #

3 Group #

K/A #

G2.1.35 Rating 3.9 Question 95 Given the following:

Reactor is in Mode 6 with refueling in progress Currently 3 new fuel assemblies (non-irradiated) are seated in the core A new fuel assembly (non-irradiated) is being moved into position in the core Reactor cavity level is greater than 23 ft above the flange LPSI A has been secured but will be operating and running in 30 minutes.

LPSI B pump breaker is racked out.

Source range is audible in all locations.

Source range meters are inoperable in the control room.

Technical Specification ________ is not in compliance. Fuel movement is ______.

A. 3.9.8.1 for one operable SDC train; allowed B. 3.9.2 for two operable and operating source range flux monitors; not allowed C. 3.9.8.2 for two operable SDC trains with one in service; allowed D. 3.9.10.2 for minimum water level above the fuel; not allowed Answer: B Explanation:

A is wrong because only one shutdown cooling train (LPSI A) is required to be operable and in service. This TS also allows a cooling loop to be removed from operation for up to an hour per 8-hour period. Also, the applicant may incorrectly choose this since the actions for non-compliance state to stop all activities that increase decay heat. New fuel should not have any decay heat generating.

B is correct because this requires audible indications in both the control room and containment. Also, requires continuous visual indication in the control room, which is absent in this question. Failure to meet this requires suspension of all core alterations.

C is wrong because this TS is not applicable since water level is greater than 23 ft above seated fuel. If the applicant does not recognize applicability, then they may incorrectly conclude that 2 trains of SDC are required. Additionally, none of the actions require suspension of fuel movement.

D is wrong because this TS is not applicable since all seated CEAs are not irradiated. Also, water level is greater than 23 ft above the fuel since the flange is at a higher elevation than the top of the fuel. Non-compliance with this does require suspension of all CEA movement within the pressure vessel.

Technical

References:

Technical Specification, Amendment 249, page 3/4 9-2 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 Comprehensive/Analysis 2

10CFR Part 55 Content:

55.43(b)2

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.2.11) Knowledge of the process for controlling temporary design changes.

Tier #

3 Group #

K/A #

G2.2.11 Rating 3.3 Question 96 EN-DC-136, Temporary Modifications, specifies for Emergency Temporary Modification Implementation that concurrence be received by the _____1_____ and a _____2_____ be completed prior to implementation.

A. 1. Plant Support Engineering Manager or designee

2. 10 CFR 50.59 screening B. 1. Plant Support Engineering Manager or designee

2. EN-DC-136, Attachment 9.2 Temporary Modification Screening and Exclusions C. 1. Engineering Director or designee

2. 10 CFR 50.59 screening D. 1. Engineering Director or designee

2. EN-DC-136, Attachment 9.2 Temporary Modification Screening and Exclusion Answer: C Explanation:

A is wrong because EN-DC-136 specifies the Shift Manager, with the concurrence of the Engineering Director, or designee, and a completed 50.59 screening prior to implementation.

Part 1 is plausible because the Plant Support Engineering Manager must be notified of the emergency temporary modification. Part 2 is correct.

B is wrong because EN-DC-136 specifies the Shift Manager, with the concurrence of the Engineering Director, or designee, and a completed 50.59 screening prior to implementation.

Part 1 is plausible because the Plant Support Engineering Manager must be notified of the emergency temporary modification. Part 2 is plausible because Attachment 9.2 must be completed for non-emergency temporary modifications.

C is correct because EN-DC-136 specifies the Shift Manager, with the concurrence of the Engineering Director, or designee, and a completed 50.59 screening prior to implementation.

D is wrong because EN-DC-136 specifies the Shift Manager, with the concurrence of the Engineering Director, or designee, and a completed 50.59 screening prior to implementation.

Part 2 is plausible because Attachment 9.2 must be completed for non-emergency temporary modifications. Part 1 is correct.

Technical

References:

EN-DC-136, Temporary Modifications, Revision 21, page 17

References to be provided to applicants during exam: None.

Learning Objective:

WPPT-OPS-PPA00, Administrative Procedures, Revision 15, EO2 - From memory identify Operations responsibilities.

Question Source:

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Question History:

Last NRC Exam No Question Cognitive Level:

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

55.43(b)3

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.2.21) Knowledge of pre-and post-maintenance operability requirements.

Tier #

3 Group #

K/A #

G2.2.21 Rating 4.1 Question 97 You just completed a surveillance test on a safety system You are directed by the SM to restore the equipment back to its normal alignment with the exception of one valve, which needs to be placed in the CLOSE position This valves normal alignment position is OPEN There is no other guidance in the surveillance procedure realignment package Which procedure provides the direction to complete this task?

A. EN-OP-115-05, Operation of Components B. EN-OP-104, Operability Determinations C. OP-100-14, Technical Specification and Technical Requirements Compliance D. EN-OP-115-07, Configuration Control Answer: D Explanation:

A is wrong because this requires a deviation process, which is contained in EN-OP-115-07. It is credible because of the title for operating components.

B is wrong (see A above). It is plausible because having a valve out of normal position might cause an applicant to think that they need to enter this procedure, but this is specifically covered in the deviation process in EN-OP-115-07.

C is wrong (see A above). It is plausible for same reason as B.

D is correct because section 5.3 of EN-OP-115-07 is the Deviations Process and this is specifically defined as a deviation Technical

References:

EN-OP-115-07, Configuration Control, Revision 4, page 7.

WPPT-OPS-PPA00, revision 15, slide 142.

References to be provided to applicants during exam: None.

Learning Objective: WPPT-OPS-PPA00 Obj E03 Question Source:

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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 No System (Tier 3)

(G2.3.11) Ability to control radiation releases.

Tier #

3 Group #

K/A #

G2.3.11 Rating 4.3 Question 98 Which of the following are required actions per TRM 3.3.3.10, Radioactive Liquid Effluent, to release a Boric Acid or Waste Condensate Tank with its associated discharge Radiation Monitor Inoperable?

A. Ensure release rate calculations have been verified by two technically qualified personnel prior to the discharge and the discharge valve lineup is independently verified.

B. Ensure release rate calculations have been verified by two technically qualified personnel prior to the discharge and obtain grab samples every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during discharge.

C. Calculate release flow rate every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge and the discharge valve lineup is independently verified.

D. Calculate release flow rate every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge and obtain grab samples every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during discharge.

Answer: A Explanation:

A is correct because these actions are required per TRM 3.3.3.10 action 1b for an inoperable BM or LWM Radiation Monitor.

B is wrong because independent samples are required prior to the discharge, not during the discharge (action 1a)

C is wrong because Part 1 is for an inoperable flow instrument (action 2). Part 2 is correct.

D is wrong because Part 1 is for an inoperable flow instrument (action 2). Independent samples are required prior to the discharge but not every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the discharge (action 1a).

Technical

References:

Technical Requirements Manual, 3/4.3.3.10 Radioactive Liquid Effluent, Page 3/4 3-23, Amendment 51 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-BM00, Boron Management System, Revision 21, EO9 - Given the Technical Requirements Manual and plant status, identify and evaluate the operability, actions, and bases for the following: TRM 3.3.3.10

Question Source:

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2017 SRO 23 (note changes; attach parent)

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New Question History:

Last NRC Exam 2017 SRO 23 Question Cognitive Level:

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

55.43(b)4

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.4.23) Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

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3 Group #

K/A #

G2.4.23 Rating 4.4 Question 99 Following a loss of offsite power and reactor trip, operators completed OP-902-000, Standard Post Trip Actions, and using Appendix 1 Diagnostic Flowchart, operators diagnosed a Loss of Coolant Accident had also occurred. Emergency Operating Procedure _____1_____ should then be entered because _____2_____.

A. 1. OP-902-002, Loss of Coolant Accident Recovery

2. Offsite power is not required to adequately mitigate the event B. 1. OP-902-008, Functional Recovery

2. Simultaneous performance of more than one optimal recovery procedure is prohibited C. 1. OP-902-003, Loss of Offsite Power/Loss of Forced Circulation Recovery

2. Vital Auxiliaries is the higher priority safety function D. 1. OP-902-002, Loss of Coolant Accident Recovery

2. RCS Inventory Control is the higher priority safety function Answer: A Explanation:

A is correct because OP-100-017, Emergency Operating Procedure Implementation Guide, prohibits the use of more than one optimal recovery procedure, directing the use of the functional recovery procedure instead. OP-100-017 also states that since offsite power is not required to mitigate a LOCA if a LOOP occur concurrently then the optimal recovery procedure is entered.

B is wrong because OP-100-017, Emergency Operating Procedure Implementation Guide, prohibits the use of more than one optimal recovery procedure, directing the use of the functional recovery procedure instead. OP-100-017 also states that since offsite power is not required to mitigate a LOCA if a LOOP occur concurrently then the optimal recovery procedure is entered. Plausible because of the guidance contained in OP-100-017 for multiple events.

C is wrong because OP-100-017, Emergency Operating Procedure Implementation Guide, prohibits the use of more than one optimal recovery procedure, directing the use of the functional recovery procedure instead. OP-100-017 also states that since offsite power is not required to mitigate a LOCA if a LOOP occur concurrently then the optimal recovery procedure is entered. Plausible because Vital Auxiliaries is a higher priority safety function vs inventory control.

D is wrong because OP-100-017, Emergency Operating Procedure Implementation Guide, prohibits the use of more than one optimal recovery procedure, directing the use of the functional recovery procedure instead. OP-100-017 also states that since offsite power is not required to mitigate a LOCA if a LOOP occur concurrently then the optimal recovery procedure is entered. Plausible if applicant does not know the safety function hierarchy.

Part 1 is correct.

Technical

References:

OP-100-017, Emergency Operating Procedure Implementation Guide, Revision 5, Pages 18-19 References to be provided to applicants during exam: None.

Learning Objective:

WLP-OPS-PPE01, Emergency Operating Procedures, Revision 17, EO4 - DIscuss EOP Procedure Usage as described in OP-100-017, Emergency Operating Procedure Implementing Guide.

Question Source:

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Question History:

Last NRC Exam No Question Cognitive Level:

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

55.43(b)5

Examination Outline Cross-Reference Level SRO No System (Tier 3)

(G2.4.28) Knowledge of procedures relating to a security event (non-safeguards information).

Tier #

3 Group #

K/A #

G2.4.28 Rating 4.1 Question 100 Given the following:

An explosion was reported in the B Emergency Diesel Generator Room.

Security reports hostile action in the B EDG room.

Per OP-901-523, Security Events, the NRC must be notified within 15 minutes (1).

The correct EAL classification in this event would be ___(2)___.

A. (1) of discovery of the event (2) HS1 B. (1) after initial EAL classification (2) HA4 C. (1) of discovery of the event (2) HA4 D. (1) initial EAL classification (2) HS1 Answer: A Explanation:

A is correct because OP-901-523, page 43, states that NRC must be notified within 15 minutes of the discovery of the event. The correct EAL would be HS1 since this is for hostile action within the protected area (the EDG room).

B is wrong because OP-901-523, page 43 states that NRC must be notified within 15 minutes of the discovery of the event. Applicant may incorrectly conclude that EAL classification is the priority over notifying the NRC of the event. Also, the correct answer is HS1. An EXPLOSION in a room containing a plant safety system is correctly classified as HA4. However, the hostile action in the protected area would take precedence since its a higher classification of HS1.

C is wrong because An EXPLOSION in a room containing a plant safety system is correctly classified as HA4. However, the hostile action in the protected area would take precedence since its a higher classification of HS1.

D is wrong because OP-901-523, page 43 states that NRC must be notified within 15 minutes of the discovery of the event. Applicant may incorrectly conclude that EAL classification is the priority over notifying the NRC of the event.

Technical

References:

OP-901-523 NRC Bulletin 2005-02 EP-001-001

References to be provided to applicants during exam: EAL Chart.

Learning Objective: Document learning objective if possible.

Question Source:

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Modified Bank #

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Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3

10CFR Part 55 Content:

55.43(b)5