ML21075A088

DC-2021-01 Final Written Exam

ML21075A088
Person / Time
Site:	Diablo Canyon
Issue date:	02/12/2021
From:	Greg Werner Operations Branch IV
To:	Pacific Gas & Electric Co
References
50-275/21-01, 50-323/21-01 50-275/OL-21, 50-323/OL-21
Download: ML21075A088 (184)

Text

U.S. Nuclear Regulatory Commission Diablo Canyon RO Written Examination Applicant Information Name: ANSWER KEY Date: 12 February 2021 Facility/Unit: Diablo Canyon Region:

I II III IV Reactor Type: W CE BW GE Start Time:

Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets.

To pass the examination, you must achieve a final grade of at least 80 percent. Examination papers will be collected 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the examination begins Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results Examination Value

__________ Points Applicants Score

__________ Points Applicants Grade

__________ Percent

U.S. Nuclear Regulatory Commission Diablo Canyon SRO Written Examination Applicant Information Name: KEY Date: 12 February 2021 Facility/Unit: Diablo Canyon Region:

I II III IV Reactor Type: W CE BW GE Start Time:

Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets.

To pass the examination, you must achieve a final grade of at least 80 percent overall, with 70 percent or better on the SRO-only items. You have 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to complete the combined examination.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values

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Points Applicants Scores

/

Points Applicants Grade

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Percent

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DCPP L191 NRC Exam 12 February 2021 i

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NAME: _RO & SRO KEY______

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DCPP L191 NRC Exam 12 February 2021 ii Multiple Choice (Fill In Your Choice)

NAME: __ RO & SRO KEY ___

If you change your answer, write your selection in the blank and initial.

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DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 001 A1.06 Control Rod Drive - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CRD system controls including:

Reactor Power Tier #

2 Group #

2 K/A #

001 A1.06 Rating 4.1 Question 01

1) Based on the NR-45 recorder indications, coincidence for C-2, Power Range Rod Stop, is

2) C-2 inhibits ______________ outward rod motion.

A.

1) met 2) only automatic B.

1) met 2) automatic and manual C.

1) NOT met 2) only automatic D.

1) NOT met

DCPP L191 Exam Rev 4 2) automatic and manual Proposed Answer: B.

1) met 2) automatic and manual Explanation:

Candidate must predict if the power range indications signify that rod stop C-2 is active. The coincidence for C-2 is 1/4 channels greater than 103%. If so, auto and manual rod motion is inhibited.

A. Incorrect. The coincidence is met, however, both auto and manual rod withdrawal is prevented. This is plausible because rods stops such as C-5, (turbine low power rod stop) stops only auto outward rod motion.

B. Correct. Coincidence for C-2 is 1 of 4. One channel greater than 103% will inhibit both auto and manual rod motion.

C. Incorrect. Both parts incorrect. First part plausible as most coincidences, such as C-3, or C-4 are 2 of 4. Second part incorrect, both auto and manual rod motion inhibited.

D. Incorrect. First part is incorrect. Second part is correct.

Technical

References:

OIM B-6-3, LA-6A (Rod Control)

References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Rod Control System. (40754)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 002 A4.07 Reactor Coolant - Ability to manually operate and/or monitor in the control room: Flow path linking the RWST through the RHR system to the RCS hot legs for gravity refilling of the refueling cavity Tier #

2 Group #

2 K/A #

002 A4.07 Rating 2.8 Question 02 The crew is preparing to gravity fill the refueling cavity from the RWST in accordance with OP B-2:II, RHR - Filling the Refueling Cavity.

The crew needs to OPEN RHR-1-8703, RHR Sys Rtn to RCS Hot Legs Loops 1 and 2 and CLOSE SI-1-8809A, RHR HX 1-1 Out to Loops 1&2.

NOTE: Breakers for both valves are closed.

1) To open RHR-1-8703 from the Control Room:

2) To close SI-1-8809A from the Control Room:

A. 1) Take control switch to OPEN, only

2) Take control switch to CLOSE, only B. 1) Take control switch to OPEN, only

2) CUT IN series contactor and then take control switch to CLOSE C. 1) CUT IN series contactor and then take control switch to OPEN

2) Take control switch to CLOSE, only D. 1) CUT IN series contactor and then take control switch to OPEN

2) CUT IN series contactor and then take control switch to CLOSE Proposed Answer: B.

1)

Take control switch to OPEN, only 2)

CUT IN series contactor and then take control switch to CLOSE Explanation:

Procedure opens 8703 and closes 8809A and B. Then the outlet from the RWST, 8980 is opened to establish gravity flow to fill the cavity.

A. Incorrect. First part is correct. There is no interlock or series contactor for 8703. Second part is incorrect, there is a series contactor for 8809A. Plausible if the interlocks are not known.

B. Correct. Both parts correct. There is not an interlock for 8703 but there is a series contactor interlock for 8809A (and B).

C. Incorrect. Both parts are incorrect.

D. Incorrect. First part is incorrect, there is not an interlock associated with 8703. Plausible because many of the RHR valves have interlocks (valve and/or series contactor). 8703 does not. Second part is correct Technical

References:

OP B-2:II, section 6.4, LB-2 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the RHR System. (35317)

Question Source:

Bank #

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 011 K6.05 Knowledge of the effect of a loss or malfunction of the following will have on the PZR LCS: Function of PZR level gauges as post-accident monitors Tier #

2 Group #

2 K/A #

011 K6.05 Rating 3.1 Question 03

1) All Pressurizer Hot Cal level indications on VB2, ___________ designated as Post Accident Monitoring instrumentation.

2) During an accident, if adverse containment conditions exist, the required Pressurizer level for SI termination, is __________________ the non-adverse containment value.

A. 1) are

2) higher than B. 1) are

2) unchanged from C. 1) are NOT

2) higher than D. 1) are NOT

2) unchanged from Proposed Answer: A. 1) are 2) higher than Explanation:

KA is for post accident monitoring, hence the question deals with accident conditions.

The candidate must identify that the pressurizer level channels are PAM (function of the gauges) and the postulated effect of adverse containment on the instruments, (malfunction of the instruments).

A. Correct. The pressurizer level channels are PAM instrumentation. Additionally, during an accident, if there is adverse containment, higher pressurizer level is required for actions such as SI termination, starting RCPs etc.

B. Incorrect. First part is correct. Second part is incorrect but plausible as other instruments used for EOP actions, such as SI termination, ie pressurizer pressure, do not have different adverse containment values.

C. Incorrect. First part is incorrect, but plausible as there are instruments, such as pressurizer pressure, which are not PAM instrumentation. Second part is correct.

D. Incorrect. Both parts incorrect.

Technical

References:

EOP E-1 References to be provided to applicants during exam: None Learning Objective: Describe PAMS components. (40462)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 014 K4.06 Knowledge of RPIS design feature(s) and interlock(s) which provide for the following: Individual and group misalignment Tier #

2 Group #

2 K/A #

014 K4.06 Rating 3.4 Question 04 A control rod is considered misaligned if, as a minimum, its individual indicated rod position is greater than ___________ steps misaligned from its _______________.

A. 1) 6

2) PPC indication B. 1) 6

2) Bank Demand Position Indication C. 1) 12

2) PPC indication D. 1) 12

2) Bank Demand Position Indication Proposed Answer: D.

1) 12 2) Bank Demand Position Indication Explanation:

A. Incorrect. Both parts incorrect. 6 steps is the accuracy of DRPI not a misaligned rod. A misaligned rod is greater than 12 steps from its Bank Demand Position Indication (BDPI).

PPC is plausible because input 1252 to AR PK03-25 is PPC Rod Pos Deviation or Rod Bank Sequence. Could be thought that the PPC is used to determine rod misalignment.

B. Incorrect. PPC is used for many parameters when taking logs, however, a misaligned rod is 12 steps from its BDPI. Second part correct.

C. Incorrect. First part is correct, a misaligned rod is greater than 12 steps, however, its 12 steps from BDPI not PPC.

D. Correct. A misaligned rod is greater than 12 steps from its BDPI.

Technical

References:

LA-3A, LCO 3.1.4, AR PK03-25 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the Rod Control System.

(9903)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 015 NIS: G2.1.20 Ability to interpret and execute procedure steps Tier #

2 Group #

2 K/A #

015 G2.1.20 Rating 4.6 Question 05 The crew is performing a reactor startup in accordance with section 6.1 of OP L-2, Hot Standby to Startup Mode, and performing step 6.1.18, to withdraw control rods to criticality.

Current ECP data:

RIL =

C @55

• ECP - 100 steps =

D@50 ECP =

D@150

• ECP + 100 steps =

D@228 After completing the rod pull of Control Bank D from 50 steps to 66 steps, the operator reports Source Range counts on both channels are 800 counts and rising with a positive, sustained startup rate (SUR).

Based on rising source range counts and positive SUR and in accordance with the guidance of OP L-2 step 6.1.18, what action should be taken?

A. Trip the reactor.

B. Fully insert all Control Bank rods only.

C. Fully insert the Control Bank rods and initiate emergency boration.

D. Establish a positive stable 0.75 DPM SUR and raise power to 10-8 amps on the Intermediate Range channels.

Proposed Answer: D. Establish a positive stable 0.75 DPM SUR and raise power to 10-8 amps on the Intermediate Range channels.

Explanation:

A. Incorrect. Plausible that the reactor should be tripped if there is a problem during a reactor startup. There is no problem, the reactor is critical below the ECP but above the RIL and the ECP-100 step point. The proper action is to raise power to 10-8.

B. Incorrect. This is the proper action if critical below the ECP-100 or +100 steps C. Incorrect. This is proper action if critical below the RIL.

D. Correct. The proper step execution is to raise power. The reactor is critical below the ECP but within the allowable guidance of the step.

Technical

References:

OP L-2, step 6.1 References to be provided to applicants during exam: None

DCPP L191 Exam Rev 4 Learning Objective: Explain the basis of significant steps associated with OP "L" procedures.

(7921)

• Specifically, as they apply to OP L-2, Hot Standby To Startup Mode.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.2 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 016 K1.03 Knowledge of the physical connections and/or cause-effect relationship between the NNIS and the following systems:

SDS (steam dump system)

Tier #

2 Group #

2 K/A #

016 K1.03 Rating 3.2 Question 06 On a Load Rejection, _______1)________ turbine power will arm the steam dumps and

______2)__________ RCS Tave will cause the trip open bistable(s) to actuate.

A. 1) auctioneered high

2) auctioneered high B. 1) auctioneered high

2) median signal select C. 1) median signal select

2) auctioneered high D. 1) median signal select

2) median signal select Proposed Answer: C. 1) median signal select 2) auctioneered high.

Explanation:

A. Incorrect. Auctioneered high is only used for RCS Tave to generate the trip open bistable signal. Median signal select is used for the turbine load rejection. Second part is correct.

B. Incorrect. This is the reverse of what signals are used for arming and the trip open bistables.

C. Correct. C-7A and B, Load Rejection are a result of the output of the median signal of PT-505A, 506A and PT-8. Auctioneered high is used to compare Tave to Tref and generate a signal to trip open the steam dumps if the temperature error exceeds setpoint.

D. Incorrect. First part is correct. Second part incorrect, auctioneered high Tave is used.

Technical

References:

LC-2B, Steam Dumps, OIM C-2-6 References to be provided to applicants during exam: None Learning Objective: Describe the operation of the Steam Dump System. (9993)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.7 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 028 K5.01 Knowledge of the operational implications of the following concepts as they apply to the HRPS: Explosive hydrogen concentration Tier #

2 Group #

2 K/A #

028 K5.01 Rating 3.4 Question 07 In accordance with the FSAR, to prevent ignition or detonation of containment hydrogen, a minimum of ___________1)________ is/are required to control containment hydrogen concentration and ensure it will not exceed ___2)_____.

A. 1) one hydrogen recombiner

2) 2.0%

B. 1) one hydrogen recombiner and one containment spray train

2) 2.0%

C. 1) one hydrogen recombiner

2) 4.0%

D. 1) one hydrogen recombiner and one containment spray train

2) 4.0%

Proposed Answer: C. 1) one hydrogen recombiner 2) 4.0%

Explanation:

A. Incorrect. FSAR states a maximum of 4% is allowed. Plausible, ECG 24.2, has a 2% limit OXYGEN limit to prevent exceeding explosive gas mixture (when hydrogen is above or assumed to, above 4%). First part is correct.

B. Incorrect. Per the discussion, the maximum allowed hydrogen concentration is 4% due to maintain the hydrogen concentration less than the flammability limit. Plausible due to the ECG reason stated above. Second part is incorrect. Containment spray is used to reduce pressure and temperature and is an aid in removal of iodine but not used for hydrogen removal (but is mentioned, along with CFCUs as mixing containment atmosphere).

C. Correct. The discussion of section 6.2 of the FSAR states The licensing limit of 4.0 percent by volume is assured by operating procedures that direct operators to initiate recombiner operation at hydrogen concentrations as low as 0.5 percent by volume. Thus, neither hydrogen burning nor detonation will occur.

D. Incorrect. 4.0% is correct. First part is incorrect.

Technical

References:

FSAR 6.2, ECG 24.2 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.1

DCPP L191 Exam Rev 4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.8

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 033 K3.03 Knowledge of the effect that a loss or malfunction of the Spent Fuel Pool Cooling System will have on the following:

Spent fuel temperature Tier #

2 Group #

2 K/A #

033 K3.03 Rating 3.0 Question 08 Unit 1 is at 100% power. Spent Fuel Pool (SFP) pump 1-2 is in service and spent fuel pool temperature is stable at 80°F.

Safety Injection occurs. 4 kV vital buses transfer to startup.

SFP temperature will be ______1)_________ because _______2_________.

A. 1) rising

2) CCW cooling is not aligned to the SFP heat exchanger B. 1) rising

2) there are no SFP pumps running C. 1) stable

2) SFP pump 1-1 starts and CCW cooling remains aligned to the SFP heat exchanger D. 1) stable

2) SFP pump 1-2 restarts and CCW cooling remains aligned to the SFP heat exchanger Proposed Answer: B.

1) rising 2) there are no SFP pumps running Explanation:

A. Incorrect. There would be cooling to the heat exchanger - phase A does not close the nonvital CCW header, however, the 1-2 pump is stopped by the phase A (does not restart) and the 1-1 SFP is not automatically started.

B. Correct. On phase A, the 1-2 pump is de-energized and does not restart, the 1-1 pump, unlike other pumps, such as SI or charging or AFW pumps are not started on SI. Loss of flow will cause SFP temperature to rise.

C. Incorrect. There would be cooling, however, there is no swap or auto start of the 1-1 pump when the 1-2 pump is tripped (as for other pumps, such as CCW). If the 1-1 pump had been the inservice pump, this would have been correct.

D. Incorrect. The 1-1 pump would have kept running, it is the 1-2 pump that is tripped, it will not be running.

Technical

References:

OIM B-6-7, LB-7 References to be provided to applicants during exam: None Learning Objective: Describe the operation of the Spent Fuel Pool Cooling System. (35694)

Question Source:

Bank #32 DCPP NRC L161 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 035 A3.02 Ability to monitor automatic operation of the S/G including: MAD valves (per ADAMS search, MAD is acronym for 'MAD' Manual/Automatic Depressurization At DCPP this would be the 10% steam dumps.

Tier #

2 Group #

2 K/A #

035 A3.02 Rating 3.7 Question 09 Unit 1 is at 100% power.

A complete loss of offsite power (230 kV and 500 kV) occurs and the reactor trips.

When the reactor trips, what steam dumps will open?

A. Groups I and II only B. Groups I, II and IV C. Groups III only D. Group IV only Proposed Answer: D. Group IV only Explanation:

A. Incorrect. Due to the loss of power, the condenser steam dumps (Groups I and II) will not be available.

B. Incorrect. This would be the normal response to a reactor trip with offsite power.

C. Incorrect. Group III is blocked from opening due to the opening of the reactor trip breakers.

Plausible if its thought the steam generator valves are the Group III valves or its known the Group IV valves are not blocked by the reactor trip and its thought that blocks all operation of the Group IV valves (which open on high steam generator pressure)

D. Correct. Group IV (10% steam dumps) will be open on the steam generator pressure controller.

Technical

References:

OIM C-2-3, C-2-4, C-2-5 References to be provided to applicants during exam: None Learning Objective: 37810 Describe controls, indications, and alarms associated with the Steam Dump System.

Question Source:

Bank #DCPP Bank A-0089 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.2 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 068 A2.04 Ability to a) predict the impacts of the following malfunctions or operations on the LRS, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of automatic isolation Tier #

2 Group #

2 K/A #

068 A2.04 Rating 3.3 Question 10 A liquid radwaste discharge exceeds the Liquid Radwaste radiation monitor, RE-18 isolation setpoint.

No automatic action(s) associated with RE-18 occur(s).

In accordance with AR PK11-21, an operator will be dispatched to ensure:

1. 0-RCV-18, Liquid Waste to Overboard - CLOSED

2. 0-FCV-477, Filters 04 and 05 outlet to EDRs - OPEN

3. HCV-647, Filter 0-4 to ASW Overboard or EDRs, set to zero demand A. 1 only B. 1 and 2 C. 3 only D. 2 and 3 Proposed Answer: B. 1 and 2 Explanation:

AR PK11-21 states: 2.1.c - IF RE-18 alarms, THEN contact Aux Watch to perform the following:

1. Ensure CLOSED RCV-18, Liquid Waste to Circulating Water Overboard

2. Ensure OPEN FCV-477, Filters 0-4 or 0-5 Outlet to EDR's.

A. Incorrect. This is an automatic action but also, 0-FCV-477 should open.

B. Correct. Both actions should occur automatically.

C. Incorrect. This is an action listed to be performed but it does not occur automatically.

D. Incorrect. HCV-647 does not go to 0 demand. However, 0-FCV-477 opening is an automatic action for RE-18.

Technical

References:

AR PK11-21 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #24 L161 10/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 4 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.11

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO EPE 007 EK3.01 Knowledge of the reasons for the following responses as they apply to a reactor trip: Actions contained in EOP for reactor trip Tier #

1 Group #

1 K/A #

EPE 007 EK3.01 Rating 4.0 Question 11 EOP E-0, Reactor Trip or Safety Injection, requires the operator ensure that there is a minimum AFW flow to the steam generators.

In accordance with the background document for EOP E-0, this is the minimum AFW required to:

A. remove decay heat.

B. prevent dryout of steam generator U-tubes.

C. maintain SG water level in the narrow range.

D. make up for the initial shrink in SG water level.

Proposed Answer: A. remove decay heat.

Explanation:

A. Correct. The basis document states this minimum flow is the minimum for heat removal.The design basis for the AFW minimum flow on a loss of feedwater is to prevent overpressurization of the primary system due to a loss of secondary heat sink if there is not sufficient flow.

B. Incorrect. Plausible because this is the reason for establishing 25 gpm to each steam generator if all steam generators are faulted.

C. Incorrect. Once steam generator level is in the narrow range and above 15%, then AFW flow can be less than 435 gpm because there is an adequate heat sink to remove decay heat.

D. Incorrect. SG level may shrink out of the narrow range but this is not the reason for the minimum AFW flow rate. AFW is typically much greater to return level to the narrow range.

The minimum flow is to ensure decay heat removal..

Technical

References:

Background E-0, LPE-0 References to be provided to applicants during exam: None Learning Objective: 7920A Explain basis of emergency procedure steps (E-0, E-0.1) including:

Bases for TCOAs with operator actions of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less.

Question Source:

Bank #39 DCPP L171 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2019 Yes Question History:

Last Two NRC Exams Yes Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 008 AK2.02 Knowledge of the interrelations between Pressurizer vapor space accident and the following: sensors and detectors Tier #

1 Group #

1 K/A #

APE 008 AK2.02 Rating 2.7 Question 12 GIVEN:

Unit 1 is at 100% power Pressurizer pressure is 2200 psig and lowering slowly Pressurizer level 52% and rising slowly RCS Tave is 570°F and stable Containment pressure is 0.7 psig and rising slowly Containment radiation is rising but below alarm setpoints PORV tailpipe temperature is 110°F and stable PRT pressure is 3 psig and stable Based on the indications, there is a:

A. main steamline leak B. pressurizer vapor space leak C. partially open pressurizer PORV D. small break on one of the RCS loops Proposed Answer: B. pressurizer vapor space leak Explanation:

Based on several detectors, the candidate must determine what is causing the indications.

A. Incorrect. Steam leak would not cause containment radiation to rise and would cause RCS Tave to lower.

B. Correct. Indications are consistent with RCS leakage, with rising pressurizer level, the leak is on the pressurizer vapor space.

C. Incorrect. An open PORV would cause PRT parameters to change.

D. Incorrect. A leak on the loops would cause pressurizer level to lower, not rise. Additionally, while level is rising, as it would if charging rose and was greater than RCS leakage, the indication of falling RCS pressure indicates charging is not causing level to rise.

Technical

References:

Steam Tables, TMI References to be provided to applicants during exam: Steam Tables Learning Objective: Describe the plant response to a loss of reactor coolant including: (41697) -

Vapor Space LOCAs Question Source:

Bank #17 Callaway X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam Callaway 2017 Yes

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO EPE 009 EA1.02 Ability to operate and monitor as they apply to a SBLOCA: RB (containment) sump level Tier #

1 Group #

1 K/A #

EPE 009 EA1.02 Rating 3.8 Question 13 The crew initiates Safety Injection due to a small break LOCA on Unit 1.

In accordance with EOP E-0, Reactor Trip or Safety Injection, at step 11, CHECK RCS -

INTACT, the crew will check Containment Recirc Sump level, LI-940/941 A. only B. and Containment Structure Sump level recorders, LR-60 & 61 C. and Narrow Range Reactor Cavity sump level recorder, LR-62 D. and Wide Range Cavity Sump level recorders, LR-942A/943A Proposed Answer: D. and Wide Range Cavity Sump level recorders, LR-942A/943A Explanation:

A. Incorrect. In addition to LI-940/941, EOP E-0 also instructs the operator to check wide range level recorders, 942A and 943A.

B. Incorrect. These level recorders are on PAM1 but not the recorders EOP E-0 checks.

C. Incorrect. Recorder is on PAM1 and monitors cavity sump, but is the narrow range, which is not used for this step.

D. Correct. EOP E-0 has the narrow range channels on VB1, LI-940/941 and the wide range channels, LR-942A/943A on PAM1 is checked.

Technical

References:

LI-1, EOP E-0 References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the Containment Structure. (37589)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO EPE 011 EK1.01 Knowledge of the operational implications of the following concepts as they apply to Large break LOCA:

Natural circulation and cooling, including reflux cooling Tier #

1 Group #

1 K/A #

EPE 011 EK1.01 Rating 4.1 Question 14 For a design basis large break LOCA, RCS decay heat removal is ECCS injection:

A. only.

B. and reflux cooling only.

C. and natural circulation only.

D. and natural circulation, followed by a transition to reflux cooling.

Proposed Answer: A. only Explanation:

A. Correct. With RCS pressure less than steam generator pressure, there is no heat removal from the steam generators.

B. Incorrect. Reflux cooling is not effective without RCS pressure above steam generator pressure to provide the necessary driving head for heat removal. Plausible because RVLIS level indication is at a point that reflux cooling in the hot legs would be occurring for small breaks.

C. Incorrect. Natural circulation is not effective without RCS pressure above steam generator pressure to provide the necessary driving head for heat removal. Plausible because it is effective for small breaks.

D. Incorrect. This would be true for a smaller break.

Technical

References:

LMCDFRC References to be provided to applicants during exam: None Learning Objective: Explain how core cooling is provided during a loss of reactor coolant including the role of the following: (41698)

a. Steam Generators as a Heat Sink

b. Break Flow versus ECCS Flow

c. Natural Circulation

d. Reflux Cooling Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 015 AA2.09 Ability to determine and interpret the following as they apply to RCP malfunctions: When to secure RCPs on high stator temperatures Tier #

1 Group #

1 K/A #

APE 015 AA2.09 Rating 3.4 Question 15 In accordance with the Foldout page in OP AP-28, Reactor Coolant Pump Malfunction, the operator will manually trip the reactor and the affected RCP, if:

A. Seal injection flow is lost.

B. Motor bearing temperature is 175°F.

C. Motor stator temperature is 310°F.

D. CCW flow to the thermal barrier heat exchanger is lost.

Proposed Answer: C.

Motor stator temperature is 310°F.

Explanation:

A. Incorrect. Not required - CCW cooling to thermal barrier exists.

B. Incorrect. A trip is required if motor bearing temperature is greater than 200 degrees.

C. Correct. In accordance with AR PK05-01, Section 2.7, RCP 1-1 High Temperature PPC, requires that if the temperature is at or above 300°F then trip the RCP following manual trip of the reactor.

D. Incorrect. Not required - there is seal injection Technical

References:

AR PK05-01, OP AP-28 References to be provided to applicants during exam: None Learning Objective: 7927 - Given initial conditions and assumptions, determine if a reactor trip or safety injection is required.

Question Source:

Bank #42 L111 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 11/2012 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 022 AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup:

adjustment of the RCP seal back pressure regulator valve to obtain normal flow Tier #

1 Group #

1 K/A #

APE 022 AK3.01 Rating 2.7 Question 16 Unit 1 is at 100% power.

A plant perturbation causes charging to change and results in the following indications.

The operator will turn potentiometer for HCV-142, ____________________________________.

A. clockwise to raise seal injection flow B. clockwise to lower seal DP C. counter clockwise to raise seal injection flow D. counter clockwise to lower seal DP Proposed Answer: C. counter clockwise to raise seal injection flow Explanation:

A. Incorrect. Opening the valve LOWERS seal injection flow. Plausible if the location/operation of the valve is not understood and to increase seal injection flow requires opening the valve.

B. Incorrect. Opening HCV-142 creates less backpressure (lower DP) and raises charging but lowers seal injection. Plausible if the operation/location of the valve in the system is not understood.

C. Correct. HCV-142 is used by the operator to create backpressure and ensure adequate seal injection flow, controlled by the operator at the Center Console. Opening the valve increases charging and lowers seal injection flow, closing down (counter clockwise) on the valve does the opposite. Seal Injection is supplied by the CVCS System at 8 to 13 gpm per RCP. The

DCPP L191 Exam Rev 4 flow rate is normally adjusted by throttling HCV-142 to divert charging flow to the seals.

The purpose of HCV-142 is to create sufficient backpressure in the charging line to ensure that adequate flow is maintained through the RCP seal water injection line upstream of valve HCV-142. Seal injection flow is low, normal flow is 8 to 13 gpm.

D. Incorrect. Turning HCV-142 counter clockwise will raise seal DP not lower DP.

Technical

References:

LB-1A, OIM B-1-1, OP AP-17 References to be provided to applicants during exam: None Learning Objective: State the purpose of CVCS components.

• RCP Seal Flow Control Valve HCV-142 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 025 AK2.02 Knowledge of the interrelations between the loss of RHRS and the following: LPI or Decay Heat Removal/RHR Pumps Tier #

1 Group #

1 K/A #

APE 025 AK2.02 Rating 3.2 Question 17 The plant is in MODE 5 and being drained to mid-loop.

What are the indications that the RHR pumps are vortexing?

____1)_____ pump amps swings and a report that there is ______2)______ noticeable increase in noise level.

A. 1) Large (8 to 10 amps)

2) a B. 1) Large (8 to 10 amps)

2) no C. 1) Small (2 to 3 amps)

2) a D. 1) Small (2 to 3 amps)

2) no Proposed Answer: D. 1) Small (2 to 3 amps) 2) no Explanation:

A. Incorrect. Cavitation of the RHR pumps will result in large pump amp swings and a noticeable noise increase as steam bubbles collapse in the pump impeller.

B. Incorrect. First part is indication of cavitation. Second part is the indication of vortexing (with small pump amp swings)

C. Incorrect. First part correct. Vortexing is indicated by small pump amp swings and no noticeable increase in noise level (second part incorrect).

D. Correct. Both parts correct. Vortexing is indicated by small oscillations in pump amps with no noticeable increase in pump noise.

Technical

References:

OE -DCPP event - RHR pump vortexing 1987, LB-2 References to be provided to applicants during exam: None Learning Objective: Discuss Operating Experience associated with the RHR System.

• Abnormal Conditions Question Source:

Bank #44 DCPP NRC L061C X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 02/2009 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0

DCPP L191 Exam Rev 4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 027 AA1.04 Ability to operate and monitor as they apply to a PZR Pressure Control System Malfunction: Pressure recovery using emergency only heaters Tier #

1 Group #

1 K/A #

APE 027 AA1.04 Rating 3.9 Question 18 GIVEN:

Unit 1 is at 100% power A problem with the normal supply breakers has caused the crew to transfer Pressurizer backup heater groups 12 and 13 to their vital 480 VAC bus Both backup groups are ON for boron equalization A pressurizer pressure control malfunction causes HC-455K output to spike and results in a low RCS pressure trip and letdown to isolate on low pressurizer level. RCS pressure is currently 1935 psig and stable.

What is the current status of the pressurizer backup heater groups 12 and 13?

A. Both pressurizer backup heater groups are energized.

B. Both pressurizer backup heater groups are de-energized.

C. Only pressurizer backup heater group 12 is energized.

D. Only pressurizer backup heater group 13 is energized.

Proposed Answer: A.

Both pressurizer backup heater groups are energized.

Explanation:

BU Heaters 2 and 3 have a vital power supply from G and H respectively. When on vital power, there are no automatic on or off signals when they are controlled by bus G or H. The 17% trip does not function.

A. Correct. Low pressure will cause the heaters to energize, despite pressurizer level of less than 17%.

B. Incorrect. This is the normal response if pressurizer level is less than 17%, the heaters are blocked from energizing OR if safety injection had occurred (strips the heaters).

C. Incorrect. Both will be energized. Plausible that one group would be energized and one would not be based something like diesel loading calculations. The Spent Fuel Pool pumps operate in this manner, only one pump has the capability to be restarted on a loss of power, or temperature interlocks for RHR suction valves opening logic, where one is affected by RCS temperature and one is affected by both RCS temperature and pressure.

D. Incorrect. Both will be energized.

Technical

References:

LA-4A, OIM A-4-5 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Pressurizer, Pressure & Level Control System.

• Backup Heaters Question Source:

Bank #

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO EPE 029 G2.1.8 ATWS - Ability to coordinate personnel activities outside the control room Tier #

1 Group #

1 K/A #

EPE 029 G2.1.8 Rating 3.4 Question 19 The crew has entered EOP FR-S.1, Response to Nuclear Power Generation/ATWS, when the reactor could not be tripped from the Control Room.

In accordance with the immediate actions of EOP FR-S.1, the crew will dispatch an operator to the _____1)______ foot elevation of the Auxiliary Building to locally trip the reactor trip breakers ________2)___________.

A. 1) 100

2) only B. 1) 100

2) and the Rod Drive MG sets C. 1) 115

2) only D. 1) 115

2) and the Rod Drive MG sets Proposed Answer: C.

1) 115 2) only Explanation:

A. Incorrect. The reactor trip and MG sets are on the 115 elevation of the Auxiliary Building.

Second part is correct. Buses 13D and E (Rod Drive MG set power supply) are on the 100 foot elevation.

B. Incorrect. Entry into EOP FR-S.1 from EOP E-0 is made if the reactor trip breakers AND both rod drive MG sets fail to cause the rods to trip and make the reactor subcritical.

Plausible that because the MG sets must have also failed to de-energize from the CR,and its thought they should be stopped as well as tripping the RTBs. Buses 13D and E (Rod Drive MG set power supply) are on the 100 foot elevation.

C. Correct. The immediate action calls for tripping the reactor trip breakers only. The trip breakers are located in the rod drive MG set rooms on the 115 of the Aux Building..

D. Incorrect. First part is correct. Second part incorrect, only the breakers are opened per the procedure.

Technical

References:

EOP E-0, EOP FR-S.1, Lesson Guide A-3A References to be provided to applicants during exam: None Learning Objective: Identify the location of components associated with the Rod Control System.

• Reactor Trip Breakers Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

DCPP L191 Exam Rev 4 Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.6

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 054 AK3.05 Knowledge of the reason for the following responses as they apply to loss of main feedwater: HPI/PORV cycling upon total feedwater loss Tier #

1 Group #

1 K/A #

APE 054 AK3.05 Rating 4.6 Question 20 GIVEN:

The crew is performing step 9, TRY To Establish Feedflow From Condensate System, of EOP FR-H.1, Response to Loss of Secondary Heat Sink In accordance with step 9, the following actions have been done:

o RCS pressure has been reduced to 1500 psig using a Pressurizer PORV o

an operator has been tasked with controlling RCS pressure between 1500 and 1865 psig using a pressurizer PORV According to the EOP FR-H.1 background document, the reason for maintaining RCS pressure less than 1865 psig is to:

A. prevent unblocking SI actuation circuits B. maximize ECCS injection C. minimize the challenge to RCS integrity D. prevent bubble formation in the reactor vessel head Proposed Answer: A.

prevent unblocking SI actuation circuits Explanation:

A. Correct. Caution in EOP FR-H.1, states, SI actuation circuits will automatically unblock if RCS Pressure rises above 1915 PSIG.

B. Incorrect. Plausible because lower pressure would result in more ECCS injection. However, the SI signals have been blocked.

C. Incorrect. Maintaining a lower pressure is plausible to prevent a challenge to RCS integrity as repressurization is a key factor in PTS events.

D. Incorrect. Bubble formation is linked to RCS pressure, however, higher pressure would prevent bubble formation.

Technical

References:

EOP FR-H.1 and background References to be provided to applicants during exam: None Learning Objective: Explain basis of emergency procedure steps (FR-Hs) including: (7920N)

• Bases for TCOAs with operator actions of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or les Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.3

DCPP L191 Exam Rev 4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO EPE 055 EK1.02 Knowledge of the operational implications of the following concepts as they apply to SBO: Natural circulation cooling Tier #

1 Group #

1 K/A #

EPE 055 EK1.02 Rating 4.1 Question 21 The crew is performing EOP ECA-0.0, Loss of All Vital AC Power, step 18, DEPRESSURIZE Intact Steam Generators To Reduce RCS Pressure To Inject Accumulators.

In accordance with EOP ECA-0.0, what condition, monitored during the depressurization, will cause the crew to stop the depressurization in order to ensure natural circulation is not interrupted?

A. Pressurizer level goes offscale low.

B. RCS Loop 1 hot leg temperature stabilizes.

C. RCS cold leg temperatures lower to less than 310°F.

D. Steam generator pressures lower to less than 300 psig.

Proposed Answer: D.

Steam generator pressures lower to less than 300 psig.

Explanation:

A. Incorrect. Note in the step states: PZR Level may be lost and Reactor Vessel Upper Head VOIDING may occur due to depressurization of S/Gs. Depressurization SHOULD NOT be stopped to prevent these occurrences B. Incorrect. Indication of a stagnated/inactive loop is not in the criteria for terminating the cooldown in ECA-0.0. However, it is plausible because guidance in EOP E-0.2, Natural Circulation Cooldown, calls for reducing the cooldown rate by half if this occurs C. Incorrect. RCS temperature limit ensures a challenge to RCS integrity does not occur D. Correct. Caution prior to the step states: Accumulator Nitrogen injection into the RCS may occur if S/Gs are Depressurized to LESS THAN 200 PSIG. From the bases: Steam generators should be depressurized to maximize delivery (into the RCS) of the water contained in the SI accumulators while minimizing delivery of nitrogen. Maintaining steam generator pressures above a value that prevents introduction of a significant volume of nitrogen into the RCS ensures that accumulator nitrogen will not impede natural circulation. Depressurization is stopped at 300 psig prevents this from occurring.

Technical

References:

ECA-0.0, ECA-0.0 background for step 17, EOP E-0.2 References to be provided to applicants during exam: None Learning Objective: 7920G - Explain basis of emergency procedure steps (ECA-0 series)

Question Source:

Bank # 51 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 4 Question Cognitive Level:

Memory/Fundamental 3.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 057 AA2.19 Ability to determine and interpret the following as they apply to Loss of Vital AC Instrument Bus and the following: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Tier #

1 Group #

1 K/A #

APE 057 AA2.19 Rating 4.0 Question 22 Unit 1 is operating at 6% power.

Panel PY-12 de-energizes.

1) _____________________________ excore nuclear instrumentation channel(s) has/have been lost.

2) An automatic reactor trip ____________ occur as a result of the loss of power to instrumentation powered from PY-12.

A. 1) A power range channel only

2) will B. 1) A power range channel only

2) will NOT C. 1) An intermediate range and a power range channel

2) will D. 1) An intermediate range and a power range channel

2) will NOT Proposed Answer: C.

1)

An intermediate range and a power range channel 2) will Explanation:

A. Incorrect. An Intermediate Range channel will also de-energize, N36 and will cause the reactor trip. The PR channel will trip its associated bistable, but PR trips are 2 of 4 coincidence.

B. Incorrect. An Intermediate Range channel, N36 is lost as well. This would be true for PY13 or PY14.

C. Correct. Both Power Range channels N42 and Intermediate Range channel N36 lose power and their bistables trip. The IR high flux bistable will cause a reactor trip because reactor power is below 10% and the trip is not yet blocked.

D. Incorrect. A reactor trip will occur due to IR high flux, which as a coincidence of 1 of 2.

Plausible because the channel indication fails low and it could be thought that therefore, it will not cause the high flux bistable to trip.

Technical

References:

LB-4 References to be provided to applicants during exam: None Learning Objective: State the power supplies to Excore Nuclear Instrumentation System components. (40940)

Question Source:

Bank #

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #30 NRC L161 10/2016 X

New Past NRC Exam DCPP 10/2016 (modified)

Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 058 AK1.01 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation Tier #

1 Group #

1 K/A #

APE 058 AK1.01 Rating 2.8 Question 23 Unit 1 is at 100% power.

Initial and current conditions for Battery Charger 1-1 and Battery 1-1 are indicated below:

Initial Conditions Current Conditions What caused the current conditions?

A. opening the input breaker to DC bus 1-1

DCPP L191 Exam Rev 4 B. a loss of 480 VAC bus H C. placing the battery on equalizing charge D. opening the battery charger DC output breaker to bus 1-1 Proposed Answer: A.

opening the input breaker to DC bus 1-1 Explanation:

A. Correct. Loss of a DC bus indicated by battery charger amps dropping to amps that match battery amps.

B. Incorrect. The normal supply to Battery Charger 1-1 is bus F. loss of Bus H would not impact DC bus 1-1. Plausible as EDG 1-1 supplies bus H.

C. Incorrect. Equalizing charge would have higher battery voltage and amps indicated on the charger would be higher.

D. Incorrect. Amps on the charger would indicate 0 and amps for the battery would go negative as the battery carries the bus as it would if the DC output breaker opened.

Technical

References:

OIM J-1-2, LJ-9 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the DC Power System.

(5193)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #43 L181 03/2020 X

New Past NRC Exam DCPP 03/2020 (modified)

Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 062 AA1.01 Ability to operate and/or monitor the following as they apply to Loss of Nuclear Service Water (SWS): Nuclear service water temperature indications Tier #

1 Group #

1 K/A #

APE 062 AA1.01 Rating 3.1 Question 24 Unit 1 is at 100% power.

1) To determine Ultimate Heat Sink Temperature per STP I-1A, Routine Shift Checks Required by Licenses, the operator will use Circulating Water Pump discharge temperature instrumentation located ______

2) In OP AP-10, Loss of Auxiliary Salt Water, the temperature of ________ at the outlet of the CCW heat exchanger is checked to determine required actions.

A. 1) locally

2) ASW B. 1) locally

2) CCW C. 1) on VB4

2) ASW D. 1) on VB4

2) CCW Proposed Answer: D. 1) on VB4 2) CCW Explanation:

ASW/CCW act as one system, as ASW (ocean) cools CCW, which in turn, cools essential heat loads. KA met by testing the Ultimate Heat Sink temperature (which provides ASW) monitoring location and CCW temperature implications are used in OP AP-10, Loss of ASW.

A. Incorrect. First part is incorrect, CWP temperature is checked on VB4. Second part incorrect. While both go through the heat exchanger, rising CCW temperature is the indication used to determine what actions are required in the AP.

B. Incorrect. IAW The STP, UHS temperature is checked by using Circ Water temperature on VB4. Second part is correct. In OP AP-10, CCW temperature at the outlet of the heat exchanger is checked to determine if actions are required.

C. Incorrect. First part is correct, CW temperature is on VB4. Second part incorrect.

D. Correct. Both parts correct, the indication for CW is on VB4, CCW is the temperature checked in OP AP-10.

Technical

References:

OP AP-10, STP I-1A References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the ASW System. (5330)

Question Source:

Bank #

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 065 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 #

APE 065 AA2.08 Rating 2.9 Question 25 GIVEN:

Unit 1 is in MODE 5 Both trains of RHR are in service Total RHR flow is 3000 gpm (both RHR pumps running)

Instrument air pressure to HCV-637, 1-2 RHR Heat Exchanger outlet valve, has just been lost.

RHR flow to loop 3 and 4 cold legs will:

A. lower to zero.

B. rise to runout conditions.

C. remains the same, the valve fails as-is.

D. rise to a maximum flow limited by a mechanical stop.

Proposed Answer: D. rise to a maximum flow limited by a mechanical stop.

Explanation:

A. Incorrect. HCV-637 will fail open raising flow. Plausible because it may be thought that the valve will fail closed on loss of air as there is not a mechanical stop on closing.

B. Incorrect. HCV-637 will fail open raising flow. However, it is limited by a stop to prevent runout. If this is not known, runout is a logical answer.

C. Incorrect. Plausible because the valve could fail as is and therefore, flow would not change.

D. Correct. HCV-637 will fail open raising flow. Mechanical stops limit the flow rise.

Technical

References:

LB-2 Residual Heat Removal System Rev 18 page 20; OIM B References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the RHR system (20950)

Question Source:

Bank #04 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E04 EK2.2 Knowledge of the interrelations (between the LOCA outside containment) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Tier #

1 Group #

1 K/A #

E04 EK2.2 Rating 3.8 Question 26 The crew is performing EOP ECA-1.2, LOCA Outside Containment.

1) According to the background document for EOP ECA-1.2, what is the major concern with a LOCA outside containment?

2) RHR is the system isolated because it _____________________________________.

A. 1) Elevated radiation readings in the Auxiliary Building.

2) is a low pressure system connected to a high pressure system and the most logical location for the leak.

B. 1) Elevated radiation readings in the Auxiliary Building.

2) provides inventory to the suction to the ECCS pumps when aligned for hot or cold leg recirculation.

C. 1) Loss of recirculation capability.

2) is a low pressure system connected to a high pressure system and the most logical location for the leak.

D. 1) Loss of recirculation capability.

2) provides inventory to the suction to the ECCS pumps when aligned for hot or cold leg recirculation.

Proposed Answer: C. 1) Loss of recirculation capability. 2) is a low pressure system connected to a high pressure system and the most logical location for the leak.

Explanation:

A. Incorrect. Elevated radiation is an entry condition for ECA-1.2, however, the goal of the procedure is to isolate the leak due to losing inventory outside the sump and therefore cannot be used for recirculation. Second part is correct.

B. Incorrect. First part is incorrect. Second part is incorrect. RHR does provide the water to the suction of the ECCS pumps during recirculation.

C. Correct. First part is correct, the goal is attempt to isolate the leak and prevent the loss of inventory outside the containment sump. Second part is correct, the procedure isolates RHR because it is a low pressure system connected to the RCS and shown its failure potential is high enough to require a procedure.

D. Incorrect. First part is correct, second part is not.

Technical

References:

Background ECA-1.2 References to be provided to applicants during exam: None

DCPP L191 Exam Rev 4 Learning Objective: Explain basis of emergency procedure steps for ECA-1.2. (42461, 7920H).

Discuss why having a solid understanding of plant design, engineering principles, and sciences is a necessary operator fundamental. (56220)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.8 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E11 G2.1.7 - Loss of Emergency Coolant Recirc: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrumentation interpretation Tier #

1 Group #

1 K/A #

E11 G2.1.7 Rating 4.4 Question 27 GIVEN:

At 0900 the Reactor Trips due to an earthquake At 0920 a LOCA occurs At 0950 the crew transitioned to ECA-1.1, Loss of Emergency Recirculation, due to the failure of both RHR pumps At 1030 the crew is performing EOP ECA-1.1 Step 16 RNO to establish the minimum required ECCS flow to remove decay heat.

What is the minimum flow rate that would satisfy the EOP ECA-1.1 Step 16 RNO?

A. 300 gpm B. 330 gpm C. 350 gpm D. 420 gpm Proposed Answer: B. 330 gpm Explanation:

A. Incorrect. Scale begins at 10 minutes, if it is not noticed, time used would be at the 100 minute mark B. Correct. This is the time of the trip, 90 minutes C. Incorrect. This is time of 70 minutes (time from small LOCA)

D. Incorrect. This is time of 40 minutes (time from entry into ECA-1.1)

Technical

References:

ECA-1.1 References to be provided to applicants during exam: ECA-1.1 appendix G Learning Objective: 42460 - Explain basis of emergency steps of ECA-1.1 Question Source:

Bank #18 L091 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 08/2011 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.0 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E05 G2.4.1 Loss of Secondary Heat Sink: Knowledge of EOP entry conditions and immediate action steps Tier #

1 Group #

1 K/A #

E05 G2.4.1 Rating 4.6 Question 28 GIVEN:

Unit 1 trips from 100% power due to a loss of offsite power The crew is exiting EOP E-0, Reactor Trip or Safety Injection Containment pressure is 0.1 psig and stable AFW flow is 0 gpm

1) In accordance with EOP F-0, attachment 3, F-0.3, Heat Sink, what is the minimum number of steam generator narrow range levels that must be less than 15% to require entry into EOP FR-H.1, Response to Loss of Secondary Heat?

2) In accordance with EOP FR-H.1, what is the minimum number of steam generator wide range levels that must be less than 18% to require the crew to immediately take actions to establish Bleed and Feed?

A. 1) 2

2) 3 B. 1) 3

2) 2 C. 1) 3

2) 4 D. 1) 4

2) 3 Proposed Answer: D.

1) 4 2) 3 Explanation:

A. Incorrect. All 4 steam generators must be less than 15% to enter EOP FR-H.1. 2 is plausible as this is the coincidence for reactor trips, such as low-low steam generator level. Second part is correct. Second part is correct, 3 wide range levels are the coincidence for immediately performing the actions to initiate Bleed and Feed.

B. Incorrect. Both parts incorrect. All narrow range levels low is the entry condition for EOP FR-H.1 - 3 is the number of wide range levels required for Bleed and Feed. 3 wide range levels are required to initiate Bleed and Feed.

C. Incorrect. This is the reverse of the respective levels required.

D. Correct. All steam generator narrow range levels less than 15%, with less than 435 gpm of AFW flow results in a RED terminus for the Heat Sink CSF. Once 3 wide range levels are below 18%, Bleed and Feed is initiated per the Foldout Page.

Technical

References:

EOP FR-H.1

DCPP L191 Exam Rev 4 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.3 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 003 K5.01 Knowledge of the operational implications of the following concepts as they apply to RCPS: relationship between RCPS flow rate and the nuclear power core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop T-hot pressure)

Tier #

2 Group #

1 K/A #

003 K5.01 Rating 3.3 Question 29 Unit 1 is at 100% power.

The breaker for RCP 1-1 trips open.

1) PK04-11, Reactor Trip Initiate will alarm due to input from ____________________.

2) The reason for the reactor trip is to provide protection against __________.

A. 1) Loop Low Flow only

2) DNB B. 1) Loop Low Flow only

2) exceeding allowable heat generation (kw/ft)

C. 1) Loop Low Flow and RCP Breaker open

2) DNB D. 1) Loop Low Flow and RCP Breaker open

2) exceeding allowable heat generation (kw/ft)

Proposed Answer: A.

1)

Loop Low Flow only 2)

DNB Explanation:

A. Correct. Above P-8 only low flow will initiate a reactor trip in one loop. The low flow trip provides DNB protection.

B. Incorrect. First part correct. Second part incorrect. Kw/foot protection is provided by trips such as OPdT. Plausible to think the low flow would cause a reduction in temperature and reactor power to rise.

C. Incorrect. First part incorrect. Only loop low flow will cause the reactor trip alarm. Second part correct.

D. Incorrect. Both parts incorrect.

Technical

References:

LB-6A, OIM B-6-4a References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 2.7 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 004 K2.01 Knowledge of bus power supplies to the following:

Boric acid makeup pumps Tier #

2 Group #

1 K/A #

004 K2.01 Rating 2.9 Question 30 GIVEN:

A loss of offsite power occurs on both Unit 1 and Unit 2 Emergency Diesel Generator 1-2 has tripped Emergency Diesel Generator 2-1 has tripped How many Boric Acid Transfer pumps are available on Unit 1 and Unit 2?

Available on Available on Unit 1 Unit 2 A.

1 1

B.

1 2

C.

2 1

D.

2 2

Proposed Answer:

A. 1 1 Explanation:

Note: Unit Difference Unit 1 EDG Unit 2 EDG 1-1 = Bus H 2-1 = Bus G 1-2 = Bus G 2-2 = Bus H 1-3 = Bus F 2-3 = Bus F A. Correct. Boric Acid Transfer pumps are powered from bus F and. bus G, for both Units. The diesels lost both power bus G and therefore, each unit has lost one BA pump and still has one pump available.

B. Incorrect. One of each pump is lost. Plausible if 2-1 EDG powered F and its thought the BA pumps were on G and H.

C. Incorrect. One of each pump is lost. Plausible if its thought the Boric Acid pumps are powered from F and H.

D. Incorrect. This would be correct if the diesels powered bus H.

Technical

References:

LB-1B, OIM J-1-1 References to be provided to applicants during exam: None Learning Objective: State the power supplies to Reactor Makeup Control System components.

(36476)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

DCPP L191 Exam Rev 4 New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 005 K5.09 Knowledge of the operational implications of the following concepts as they apply to the RHRS: dilution and boration considerations Tier #

2 Group #

1 K/A #

005 K5.09 Rating 3.2 Question 31 GIVEN:

RCS temperature is 275°F RCS boron concentration is 1800 ppm A cooldown for a refueling outage is in progress All RCPs and RHR pumps have been temporarily secured Auto makeup to the VCT initiates, however, the TARGET BLEND PPM value is set to 1600 ppm.

What is the operational concern(s) with the TARGET BLEND PPM value set at 1600 ppm?

1. Reduction in SHUTDOWN MARGIN

2. Unplanned heatup to MODE 3

3. Unplanned cooldown to MODE 5 A. 1 only B. 2 only C. 3 only D. 1 and 2 Proposed Answer: A. 1 only Explanation:

A. Correct. The accident of dilution is analyzed as a reduction of SDM and a return to criticality concern.

B. Incorrect. Unlike normal operation, no change of RCS temperature will occur. At power, lowering boron would cause an addition of positive reactivity and temperature would rise.

Plausible if its thought MTC would cause a temperature change.

C. Incorrect. Plausible if believed there could be a positive MTC and cause temperature to be reduced D. Incorrect.. Temperature will not change and cause a heatup.

Technical

References:

LPA-33 References to be provided to applicants during exam: None Learning Objective: Discuss operator behaviors and practices related to the operator fundamental of closely monitoring plant indications and conditions. (56218)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 006 A2.11 Ability to a) predict the impacts of the following malfunctions or operations on the ECCS, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture of ECCS header Tier #

2 Group #

1 K/A #

006 A2.11 Rating 4.0 Question 32 30 minutes after a large break LOCA, the crew completes EOP E-1.3, Transfer to Cold Leg Recirculation and returns to EOP E-1, Loss of Reactor or Secondary Coolant.

The crew has just entered OP AP-32, ECCS Train Isolation During Post Accident Recirculation due to a suspected leak on Train 1 of ECCS.

1) According to the entry conditions for OP AP-32, the procedure is entered based on

2) In accordance with OP AP-32, the crew will isolate Train 1 by stopping RHR pump 1-1 and both __________ pumps.

A. 1) the Foldout page of EOP E-1

2) Charging B. 1) the Foldout page of EOP E-1

2) Safety Injection C. 1) RM-13, RHR Exhaust Duct radiation monitor indication

2) Charging D. 1) RM-13, RHR Exhaust Duct radiation monitor indication

2) Safety Injection Proposed Answer: C. 1) RM-13, RHR Exhaust Duct radiation monitor indication 2)

Charging Explanation:

KA is based on accident conditions.

A. Incorrect. First part incorrect. Plausible the EOP would check the integrity system following the alignment. Second part correct.

B. Incorrect. Both parts incorrect. Second part plausible as the SI pumps are train 2 and would be stopped if RHR pump 1-2 is shutdown.

C. Correct. The procedure is performed on when a leak has been identified in the Aux Building and must be isolated, such as RM-13 readings and RHR sump pumps running. The Charging pumps are on train 1.

D. Incorrect. First part is correct. SI pumps are train 2.

Technical

References:

OP AP-32, AR PK02-17 References to be provided to applicants during exam: None Learning Objective: Discuss why having a solid understanding of plant design, engineering principles, and sciences is a necessary operator fundamental. (56220)

DCPP L191 Exam Rev 4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.5 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 007 A3.01 Ability to monitor automatic operation of the PRTS, including: components which discharge to the PRT Tier #

2 Group #

1 K/A #

007 A3.01 Rating 2.7 Question 33 Unit 1 is at 100% power. PRT level and pressure are normal.

A pressure transient causes letdown pressure to begin to rise.

1) Letdown Relief valve, RV-8117 will lift at a setpoint of __________ psig.

2) Assume when RV-8117 lifts, PRT pressure begins to rise at 5 psig/minute. The PRT rupture disks would rupture in approximately _________ minutes.

A. 1) 450

2) 3 B. 1) 450

2) 19 C. 1) 600

2) 3 D. 1) 600

2) 19 Proposed Answer: D.

1) 600 2) 19 Explanation:

The auto actions to meet the KA are the lifting of the relief valve at 600 psig, which discharges to the PRT and the pressure at which the rupture disc fails of 100 psig. Normal level in the PRT is approximately 85%. Normal pressure is approximately 3 psig.

A. Incorrect. 450 psig is the RHR relief to PRT pressure. In 3 minutes pressure is only about 20 psig. Plausible if the 20 is added the 85 (level).

B. Incorrect. First part incorrect. Second part is correct, in approximately 19 minutes PRT pressure will be at the rupture disk rupture pressure of 100 psig.

C. Incorrect. First part is correct. The setpoint for the letdown relief is 600 psig. Second part is incorrect.

D. Correct. Closing of 8152 will cause relief valve 8117 at 600 psig. The rupture disc setpoint is 100 psig. Normal pressure is approximately 3 psig, so it would take about 19 minutes for pressure to rise to the setpoint Technical

References:

LA-4B, LB-2, LB-1A, OPP B-1A:XII References to be provided to applicants during exam: None Learning Objective: 40573 Describe PRT system components Question Source:

Bank #6 L162 NRC Exam X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 008 G2.1.28-CCW - Knowledge of the purpose and function of major system components and controls Tier #

2 Group #

1 K/A #

008 G2.1.28 Rating 4.1 Question 34

1) The purpose of the CCW Surge Tank pressurization system is to provide assurance that boiling in the _______________________ will not occur as a result of a design basis accident.

2) If the normal supply for the CCW Surge Tank pressurization system is lost, ______________

will automatically supply the pressurization system.

A. 1) CCW pumps

2) nitrogen bottles B. 1) CCW pumps

2) instrument air C. 1) CFCUs

2) nitrogen bottles D. 1) CFCUs

2) instrument air Proposed Answer: C. 1) CFCUs 2) nitrogen bottles Explanation:

A. Incorrect. Plausible because CCW pumps are the motive force for cooling accident loads on CCW. Second part is correct.

B. Incorrect. Plausible because CCW pumps are the motive force for cooling accident loads on CCW. Second part is incorrect. Instrument air is the source if the backup, nitrogen bottles are lost.

C. Correct. During analysis of CCW System heat loading, it was discovered that a postulated scenario could result in flashing in the CCW piping.

• In the event of a Large Break LOCA coincident with a loss of off-site power, (or degraded 230KV power that would cause double sequencing) the CFCUs would coast down. At some point during the coast down the CFCUs would sequence on before the CCW pumps would start.

• The resulting heat up of the CCW fluid, due to the high temperature inside Containment, would cause flashing of the fluid and subsequent water hammer when the CCW pumps sequenced on. *Bottled nitrogen provides the normal backup and automatically supplies the system pressure drops.

D. Incorrect. First part is correct. Second part is incorrect. Instrument air can be valved in if nitrogen is lost to the system.

Technical

References:

LF-2 References to be provided to applicants during exam: None Learning Objective: Describe CCW System components.

• Surge Tank Pressurization System Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

DCPP L191 Exam Rev 4 New X

Past NRC Exam Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 3.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 010 K4.03 Knowledge of the PZR PCS design feature(s) and/or interlocks which provide for the following: Overpressure control Tier #

2 Group #

1 K/A #

010 K4.03 Rating 3.8 Question 35 GIVEN:

Unit 1 is in MODE 4 LTOP is in service A loss of decay heat removal causes RCS temperature to rise to 285°F on all loops and RCS pressure indication to rise to 450 psig on PI-403A and 425 psig on PI-405A.

What is the status of the Pressurizer PORVs?

A. None of the PORVs are open B. One PORV is open C. Two PORVs are open D. All three PORVs are open Proposed Answer: A.

None of the PORVs are open Explanation:

A. Correct. Normally, rising pressure above 435 psig causes one or both PORVs to open.

However, LTOP is automatically disabled if temperature rises above 280°F and none of the PORVs will open.

B. Incorrect. This would be the response if temperature was below the enable setpoint and one channel (PI-403A rose above 435 psig). The PORVs respond to the channel they are associated with, one responds to PI-403A and one opens if PI-405A rises above 435 psig.

However, because temperature is above 280°F, no PORV will open.

C. Incorrect. Plausible that rising pressure causes both to open, however, due to high temperature, none will open.

D. Incorrect. At power, all PORVs will open at the same pressure. At low pressure, only the Class I PORVs open. However, for the current conditions, none will open.

Technical

References:

OIM A-4-7 References to be provided to applicants during exam: None Learning Objective: 36923 - Analyze automatic features and interlocks associated with the Pzr, Pzr Pressure and Level Control System Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 012 K1.05 Knowledge of the physical connections and/or cause-effect relationships between the RPS and the following: ESFAS Tier #

2 Group #

1 K/A #

012 K1.05 Rating 3.8 Question 36 GIVEN:

Reactor power is 4% when a steam break inside containment occurs Containment pressure peaks at 8 psig The reactor trips The MSIVs close What automatic action(s) could have caused:

1) the reactor to trip?

2) the MSIVs to close?

A. 1) Safety Injection only

2) Low Steam Line pressure only B. 1) Safety Injection only

2) High Containment pressure or Low Steam Line pressure C. 1) RCS low pressure trip or Safety Injection

2) Low Steam Line pressure only D. 1) RCS low pressure trip or Safety Injection

2) High Containment pressure or Low Steam Line pressure Proposed Answer: A.

1) Safety Injection only 2) Low Steam Line pressure only Explanation:

KA is relationship to RPS and ESFAS, question tests the connection, which occurs when an accident occurs.

A. Correct. SI will cause a reactor trip. The MSIVs are closed by the low steam line pressure signal.

B. Incorrect. First part is correct. High containment pressure will isolate containment when SI causes Phase A to actuate. However it does not cause the MSIVs to close. The MSIVs close on: low steamline pressure (above P-11) and Phase B (high - high containment pressure)

C. Incorrect. Power is below P-7, low pressurizer pressure will not cause a reactor trip (blocked until above 10%). Second part is correct.

D. Incorrect. Both parts are incorrect.

Technical

References:

OIM B-6-4b, B-6-10, B-6-8 References to be provided to applicants during exam: None Learning Objective: 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source:

Bank #49 L161 10/2016 X

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.7 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 012 K6.03 Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Trip logic circuits Tier #

2 Group #

1 K/A #

012 K6.03 Rating 3.1 Question 37 GIVEN:

Unit 1 is at 100% power Pressurizer Level Channel LT-459 is in BYPASS While LT-459 is in BYPASS, Pressurizer Level channel LT-460 fails HIGH.

Which is the plant response?

A. The reactor trips due to satisfying the 2 of 3 High Pressurizer Level trip coincidence.

B. The reactor trips due to satisfying the 2 of 4 High Pressurizer Level trip coincidence.

C. The reactor does not trip due to NOT satisfying the 2 of 3 High Pressurizer Level trip coincidence.

D. The reactor does not trip due to NOT satisfying the 2 of 4 High Pressurizer Level trip coincidence.

Proposed Answer: C. The reactor does not trip due to NOT satisfying the 2 of 3 High Pressurizer Level trip coincidence.

coincidence.

Explanation:

For Pressurizer high level, the logic is 2 of 3 channels to trip, (unlike trips, such Pressurizer High pressure 2 of 4 channels) or if they include cold cal channel. BYPASS does not trip a channel and the channel does not input to the trip logic.

A. Incorrect. The reactor will not trip. While in Bypass, the channel will not trip and the matrix is both of the remaining channels (2) to trip. Logic is 2 of 3. Plausible if its thought the channel is tripped.

B. Incorrect The reactor will not trip. While in Bypass, the channel not cause a trip and the matrix is both of the remaining channels (2) to trip. Plausible - Logic is 2 of 3, not 2 of 4 as it is for trips such as high or low pressurizer pressure.

C. Correct. The logic is still 2 channels to trip and only channel has tripped (LT-460)

D. Incorrect The logic is still 2 channels to trip and only channel has tripped (LT-460).

Plausible - The trip logic is 2 of 3 not 2 of 4, as it is for trips such as high or low pressurizer pressure.

Technical

References:

OIM B-6-4b References to be provided to applicants during exam: None Learning Objective: 37051 - Discuss abnormal conditions associated with the RPS Question Source:

Bank #11 L161 10/2016 X

(note changes; attach parent)

Modified Bank #

New

DCPP L191 Exam Rev 4 Past NRC Exam DCPP NRC 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 013 K1.18 Knowledge of the physical connections and/or cause-effect relationships between the ESFAS and the following:

Premature reset of ESF actuation Tier #

2 Group #

1 K/A #

013 K1.18 Rating 3.7 Question 38 Unit 1 is at 100% power.

A spurious Safety Injection occurs. About 30 seconds later, the operator presses the SI reset pushbuttons on VB1.

PK 08-22, Auto SI Blocked ________________________________________.

A. will not alarm B. immediately alarms and remains lit C. immediately alarms but clears when the operator releases the pushbuttons D. will alarm and remain lit about 30 seconds after the operator presses the reset pushbuttons Proposed Answer: A.

will not alarm Explanation:

KA must occur during SI actuation in order to be met.

A. Correct. There is a 65 second timer that prevents reset of SI. Any attempts to reset before the timer is done will not result in the alarm as there would not be any output from the AND block for SI Reset/Block.

B. Incorrect. Plausible if the time delay is not known or the duration is not known.

C. Incorrect. Plausible that the signal could be reset as long as the reset pushbuttons are in reset but would clear when the buttons are releasedl D. Incorrect. The timer blocking SI reset is approximately 65 seconds. If its thought the reset locks in, then this would be plausible.

Technical

References:

OIM B-6-5 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Reactor Protection System.

• Safety Injection Actuation Signal Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 013 A3.02 Ability to monitor automatic operation of the ESFAS, including the following: operation of actuated equipment Tier #

2 Group #

1 K/A #

013 A3.02 Rating 4.1 Question 39 GIVEN:

Unit 1 is performing a heatup in accordance with OP L-1, Plant Heatup From Hot Shutdown to Hot Standby Electrical power is aligned to startup RCS temperature is 525°F RCS pressure is 1900 psig PK08 indicates as shown below: (red outlined annunciators are lit)

A steam break, outside containment, upstream of the MSIV occurs on Steam Generator 11.

1) SI will _____________ actuate.

2) Once SI is automatically or manually actuated, what AFW pumps will get a start signal?

A. 1) automatically

2) only the motor driven AFW pumps B. 1) automatically

2) motor and turbine driven AFW pumps C. 1) NOT automatically

2) only the motor driven AFW pumps D. 1) NOT automatically

2) motor and turbine driven AFW pumps

DCPP L191 Exam Rev 4 Proposed Answer: C.

1) NOT automatically 2) only the motor driven AFW pumps Explanation:

Post trip conditions required to meet KA.

Tests what the operator would see in the Control Room (operational validity) when RCS pressure is less than 1915 psig and an accident occurs (steam break). Additionally, the response of the AFW pumps when the SI (ESFAS actuation) occurs.

A. Incorrect.SI on low RCS pressure is blocked below P-11 (1915 psig) (PK08-06 LIT).

Plausible if P-11 is thought to only affect RCS pressure SI. Also, SI would automatically actuate if the break was inside containment and pressure rises to greater than 3 psig. Second part correct, SI starts the motor driven pumps.

B. Incorrect. Both parts incorrect. The TDAFW pump does not get a start signal from the SI.

C. Correct. SI will not actuate automatically and have to manually actuated. The SI signal will start the MDAFW pumps.

D. Incorrect. First part correct, second part incorrect.

Technical

References:

OIM B-6-2, B-6-6, D-1-2 References to be provided to applicants during exam: None Learning Objective: 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source:

Bank #

(note changes; attach parent)

Modified Bank #12 L162 01/2018 X

New Past NRC Exam DCPP NRC 01/2018 (modified)

Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 022 K4.03 Knowledge of the Containment Cooling design feature(s) and/or interlocks which provide for the following:

Automatic containment isolation Tier #

2 Group #

1 K/A #

022 K4.03 Rating 3.6 Question 40 What radiation monitor(s) will/would generate a Containment Ventilation Isolation signal to close any open containment purge valves?

1. RM-12, Containment Rad Gas

2. RM-14/14R, Plant Vent

3. RM-44A/B, Containment Exhaust A. 1 only B. 3 only C. 1 and 2 D. 2 and 3 Proposed Answer: B. 3 only Explanation:

A. Incorrect. While RM-12 is a containment radiation monitor, it does not generate a CVI.

B. Correct. Either RM-44A and B monitor the containment exhaust during a purge and generator a CVI to close all the valves when high radiation is sensed.

C. Incorrect. RM-14/14R could be thought to act as a backup to generate a CVI.

D. Incorrect. Because the exhaust is is directed to the plant vent, it makes sense to think RM-14/14R would also generate a CVI.

Technical

References:

OIM B-6-9a, G-3-1 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Reactor Protection System.

• Containment Ventilation Isolation Actuation Signal Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.2 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 026 K1.01 Knowledge of the physical connections and/or cause-effect relationships between the CSS and the following: ECCS Tier #

2 Group #

1 K/A #

026 K1.01 Rating 4.2 Question 41 Which of the following would prevent opening the residual heat removal (RHR) to containment spray rings isolation valve (9003A)?

A. RHR cold leg injection isolation valve (8809A) - OPEN B. RHR to containment spray rings isolation valve (9003B) - OPEN C. Containment spray pump discharge valve (9001A) - CLOSED D. Containment recirculation sump suction valve (8982A) - CLOSED Proposed Answer: D.

Containment recirculation sump suction valve (8982A) - CLOSED Explanation:

Accident conditions required to meet KA.

A. Incorrect. There is no interlock between 9003 and 8809 valves. Plausible because it may be thought that an interlock would be required to prevent one RHR pump from supplying both sprays and injection.

B. Incorrect. There is no interlock between the 9003 valves. Plausible because one train is aligned by the operators to supply sprays, and it may be thought that an interlock would be necessary to prevent both trains from supplying the spray header while on recirculation mode.

C. Incorrect. There is no interlock between 9003 and 9001 valves. Plausible because it may be thought that an interlock would be necessary to prevent a CS pump and RHR pump from supplying the spray header simultaneously.

D. Correct. The associated 8982A/B must be open in order to open 9003A or B (this ensures the ECCS system is in the recirculation mode and not drawing suction from the RWST).

Technical

References:

EOP E-1.3, LI-2 References to be provided to applicants during exam: None Learning Objective: Describe the operation of the Containment Spray System. (40805)

Question Source:

Bank # 15 L161, 10/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP NRC 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 026 A4.05 Containment Spray: Ability to manually operate and/or monitor in the control room: containment spray reset switches Tier #

2 Group #

1 K/A #

026 A4.05 Rating 3.5 Question 42 GIVEN:

A LOCA occurs on Unit 1 SI is not reset PK01-18, CONTMT SPRAY ACTUATION is in alarm Phase B Red lights, on Monitor Light Box D, are lit The operator presses both Containment Spray reset pushbuttons.

When the operator presses the reset pushbuttons, PK01-18:

A. remains lit because SI is NOT reset.

B. remains lit because Phase B is NOT reset.

C. clears because the Containment Spray reset has "retentive memory".

D. clears but reflashes because the Containment Spray reset does NOT have "retentive memory".

Proposed Answer: C. clears because the Containment Spray reset has "retentive memory".

Explanation:

Accident conditions required to meet KA A. Incorrect. If SI is reset, containment spray will not ACTUATE, however, it does not block reset.

B. Incorrect. Containment spray reset is a latch, which will reset the spray alarm and clear the alarm. Phase B has the same setpoint as Containment Spray but is not affected by resetting spray.

C. Correct. Spray will reset and alarm will reset. Because the reset is "retentive" it can be reset any time after actuation.

D. Incorrect. Spray will reset and alarm will reset. Plausible because for many signals, such as FWI with high containment pressure, the signal must be clear to allow reset.

Technical

References:

PK01-18, LB-6A OIM B-6-8, B-6-12 References to be provided to applicants during exam: None Learning Objective: 37578 - Describe controls, indications, and alarms associated with the CSS Question Source:

Bank ##15 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 3.2 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 039 A1.05 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits associated with operating the MRSS (Main and reheat steam) including: RCS T-ave Tier #

2 Group #

1 K/A #

039 A1.05 Rating 3.2 Question 43 GIVEN:

Unit 1 is at 3% power, EOL A plant startup is in progress in accordance with OP L-3, Secondary Plant Startup Steam Dumps are maintaining RCS Tave in AUTOMATIC in Steam Pressure mode If HC-507, 40% Stm Dump Vlvs Press Cont, setpoint is changed from 83.8% to 82.5%.

RCS Tave will ______1)________ and reactor power will ________2)_________.

A. 1) lower

2) lower B. 1) lower

2) rise C. 1) remain the same

2) lower D. 1) remain the same

2) rise Proposed Answer: B. 1) lower 2) rise Explanation:

Candidate must understand how the system and feedbacks respond when low in power Raising the setpoint will cause Tave to rise to match the new steam demand. Lowering the setpoint will cause the opposite to occur.

A. Incorrect. Correct that Tave lowers but the effect is to raise power with rising steam flow not lower.

B. Correct. Tave will lower, the rise in steam flow (to maintain the lower temperature) will cause power to rise..

C. Incorrect. Tave lowers. Normal plant response of steam dumps is to maintain constant Tave, also plausible if the effect of the setpoint change was thought to close dumps, power could be thought to lower.

D. Incorrect. Tave lowers. Its thought the function is to maintain constant Tave, then raising steam flow to maintain Tave would raise power.

Technical

References:

LC-2B References to be provided to applicants during exam: None Learning Objective: Desribe system interrelationships between the Steam Dump System andother plant systems. (8042)

DCPP L191 Exam Rev 4 Question Source:

Bank #34 L162 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.2 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 059 K3.03 Knowledge of the effect that a loss or malfunction of the MFW will have on the following: S/Gs Tier #

2 Group #

1 K/A #

059 K3.03 Rating 3.5 Question 44 Unit 1 is at 65% power.

The operator notes the following indications over the last 50 seconds:

1) At the current rate and level, a reactor trip setpoint will be reached in approximately

________ seconds

2) At the Digital Feedwater Control Station, the operator will observe that a A. 1) 30

2) main feedwater reg valve has failed closed B. 1) 30

2) steam generator narrow range level channel has failed low C. 1) 50

2) main feedwater reg valve has failed closed D. 1) 50

2) steam generator narrow range level channel has failed low Proposed Answer: C. 1) 50 2) main feedwater reg valve has failed closed

DCPP L191 Exam Rev 4 Explanation:

Level dropped 25% in 50 seconds or 0.5% per second. To reach the trip setpoint of 15%, level will lower another 25% or another 50 seconds.

A. Incorrect. 25% is the adverse containment number. At power, Containment is not adverse. To reach 25%, level would need to lower another 15% at 0.5%/second, this is 30 seconds.

Second part is correct. The closing a MFP reg valve would cause level lower as feed to the steam generator ceases.

B. Incorrect. 30 seconds is plausible if the adverse containment value is used. Level channel failure could cause the indication to lower however, the control circuit would remove the channel from the circuit and feed flow would be unaffected.

C. Correct. Level will reach 15% is 50 seconds. The closing a MFP reg valve would cause level lower as feed to the steam generator ceases.

D. Incorrect. First part is correct. Feed flow channel failure could cause the indication to go low but the channel would be de-selected by the control circuit and level would be maintained on program.

Technical

References:

LC-8B, OIM References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the DFWCS. (37642)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 061 A1.05 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits associated with operating the AFW controls including: AFW flow/amps Tier #

2 Group #

1 K/A #

061 A1.05 Rating 3.6 Question 45 GIVEN:

Unit 1 is in MODE 3 1-2 and 1-3 Motor Driven AFW pumps are running AFW flow is approximately 100 gpm to each steam generator All AFW LCVs are in AUTO.

PT-433, AFW Pp 1-2 discharge pressure transmitter, is reading zero. There are no PCS alarms.

AFW Flow meters on VB3 will indicate:

A. zero flow to two steam generators.

B. zero flow to all steam generators.

C. greater than 400 gpm flow to two steam generators.

D. greater than 400 gpm flow to all steam generators.

Proposed Answer:

A.

zero flow to two steam generators Explanation:

A. Correct. Pressure transmitters PT-433 and 434 control the LCVs for their respective motor driven AFW pump and the steam generators are not cross connected when both AFW pumps are running. PT-433 failing low causes the 1-2 AFW pump LCVs to close as part of the runout protection for the pump.

B. Incorrect. The PT only controls the LCVs for the 1-2 pump. Plausible if its thought the steam generators are cross tied or its believed that either PT causes the LCVs to close and cease all flow to all steam generators.

C. Incorrect. The failure low causes the LCVs to close, not open. This would be correct if the failure caused the valves to open.

D. Incorrect. As stated above, the steam generators are not cross tied and the PT only affects the LCVs for the steam generators being fed by the motor driven pump, not all LCVs.

Technical

References:

LD-1 References to be provided to applicants during exam: None Learning Objective: Discuss operator behaviors and practices related to the operator fundamental of closely monitoring plant indications and conditions. (56218)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 4 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 062 A4.01 Ability to manually operate and/or monitor in the control room: All breakers (including available switchyard)

Tier #

2 Group #

1 K/A #

062 A4.01 Rating 3.3 Question 46 Unit 1 is preparing to perform OP J-2:V, Offsite Power Sources - Backfeeding the Unit from the 500kV System.

In accordance with OP J-2:V, the operator will check that the Exciter Field Breaker is open on

______1)_____ and close Main Gen Output CKT Bkrs CB-532 and CB-632 at _____2)_____.

A. 1) CC3 2)

CC3 B. 1) CC3 2)

VB5 C. 1) VB5 2)

CC3 D. 1) VB5 2)

VB5 Proposed Answer: A. 1)

CC3 2) CC3 Explanation:

A. Correct. The Exciter Field breaker must be open. The lights (Red/Green) are located on CC3.

CB-532 and CB-632 are closed from CC3.

B. Incorrect. Both are on or operated from CC3. VB5 is plausible. It has multiple breakers to operate, such as 12 kV cross tie breaker, 52-VU-11, a mimic for the 230 kV and 500 kV connection to the unit, as well as meters and indications for the startup and aux transformers and switchyard alarms.

C. Incorrect. The Exciter Field breaker is on CC3. Second part correct.

D. Incorrect. With the multitude of meters, breakers and indications for offsite breakers and transformers, its plausible both are there.

Technical

References:

J-2:V References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the Switchyard and Offsite Power System.

Question Source:

(note changes; attach parent)

X No Question History:

No Question Cognitive Level:

2.5 Bank #

Modified Bank #

New Past NRC Exam Last Two NRC Exams Memory/Fundamental Comprehensive/Analysis

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 063 K2.01 Knowledge of bus power supplies to the following:

Major DC loads Tier #

2 Group #

1 K/A #

063 K2.01 Rating 2.9 Question 47 Which of the following receive control power from Vital DC Bus 1-1?

A. Loads on 4 kV Bus F and EDG 1-1 B. Loads on 4 kV Bus F and EDG 1-3 C. Loads on 4 kV Bus H and EDG 1-1 D. Loads on 4 kV Bus H and EDG 1-3 Proposed Answer: B.

Loads on 4 kV Bus F and EDG 1-3 Explanation:

A. Incorrect. DC Bus 1-1 supplies control power to Bus F. Plausible because the EDG is 1-1 and could imply that its bus F. Bus F is correct.

B. Correct. Both receive control power from from DC Bus 1-1.

C. Incorrect. Neither part is correct. Plausible that DC bus 1-1 supplies EDG 1-1, which supplies power to Bus H.

D. Incorrect. DC bus 1-1 supplies control power to EDG 1-3 (correct). Plausible that bus H would be aligned to EDG 1-3.

Technical

References:

OP AP-23 References to be provided to applicants during exam: None Learning Objective: Explain the consequences of loss of DC vital bus. (7116)

Question Source:

Bank #20 L171 01/2019 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2019 Yes Question History:

Last Two NRC Exams Yes Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.8

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 063 A2.01 Ability to a) predict the impacts of the following malfunctions or operations on the DC Electrical system, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Grounds Tier #

2 Group #

1 K/A #

063 A2.01 Rating 2.5 Question 48 Unit 1 is at 100% power.

A ground causes the DC breaker supplying control power to 12 kV bus D to open.

1) Indications in the Control Room for 12 kV bus D will be a loss of indicating lights

2) The crew will ______________________________________________.

A. 1) and the breakers for the affected loads open

2) investigate the ground, a reactor trip is not required at this time B. 1) and the breakers for the affected loads open

2) ensure the reactor tripped and go to EOP E-0, Reactor Trip or Safety Injection C. 1) however, the breakers for the affected loads remain closed

2) investigate the ground, a reactor trip is not required at this time D. 1) however, the breakers for the affected loads remain closed

2) ensure the reactor tripped and go to EOP E-0, Reactor Trip or Safety Injection Proposed Answer: C.

1) however, the breakers for the affected loads remain closed 2) investigate the ground, a reactor trip is not required at this time Explanation:

First part of KA is knowing that the impact is the loss of indication however, the breakers do not open (and cannot be opened from the control room). Second part is determining if the conditions warrant a reactor trip or continued ground investigation.

NOTE: the opening of the DC control power breaker causes a loss of indication for RCPs 1-1, 1-3 and CWP 1-2. It also causes a loss of indication for the redundant breakers for the other two RCPs which does not affect the Control Room indication. The RCP alarms for the other two RCPs is received, however, there is no loss of indication in the Control Room. They are left out of the information given as it is not pertinent to the action to be taken.

A. Incorrect. A condition, such as a ground has caused a loss of control power, which causes the lights to go out, however, the breakers remain closed. Second part is correct.

B. Incorrect. Both parts incorrect. The breakers remain closed. Per PK05-01 or 05-03, the action is to dispatch an operator to check the DC control power breaker, 72-1233, "12KV SWGR BUS E & SVD6R, SVD7R" (the lights for the RCP redundant breakers). A reactor trip is not required, the pumps continue to run. Plausible that its thought the pumps trip or the reactor must be tripped due to inability to trip the loads from the control room is necessary.

C. Correct. The breakers remain closed and the pumps continue to run. There is no reason to trip the loads, the action is to investigate the ground.

DCPP L191 Exam Rev 4 D. Incorrect. First part is correct. There is no reason to trip the reactor. Reactor trip is plausible as it could be thought the action is to trip if control of the breakers from the control room is lost or it could be thought the components need to be tripped. If the RCPs were to be tripped, the reactor is tripped first Technical

References:

AR PK05-01, 05-03, drawings 445075, 477848 References to be provided to applicants during exam: None Learning Objective: 37793 -Describe controls, indications, and alarms associated with the DC Power System Question Source:

Bank #22 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 064 K4.02 Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Trips for ED/G while operating (normal or emergency)

Tier #

2 Group #

1 K/A #

064 K4.02 Rating 3.9 Question 49 Unit 1 is at 100% power.

An emergency diesel generator has been started from the Control Room and paralleled with offsite power for a normal surveillance.

For the current plant conditions, what condition(s) could cause the emergency diesel generator to trip?

1) High Jacket Water temperature

2) Differential

3) Overspeed A. 1 only B. 2 only C.

1 and 3 D. 2 and 3 Proposed Answer: D. 2 and 3 Explanation:

A. Incorrect. High Jacket Water temperature is only active when control is in LOCAL. If the EDG was started in the Control Room, Local/Remote Control Selector switch is in Remote and the trip is not active.

B. Incorrect. Answer is incomplete, Differential is one trip that is active, but overspeed is as well.

C. Incorrect. Overspeed will trip the diesel, however, the jacket water temperature trip is not active at this time.

D. Correct. Both trips will cause the diesel to trip.

Technical

References:

LJ-6B References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Diesel Generator System. (37725)

Question Source:

Bank #23 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 064 K6.08 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks Tier #

2 Group #

1 K/A #

064 K6.08 Rating 3.2 Question 50 If there is less than the required level in the diesel fuel oil storage tanks, the emergency diesel generators may not operate for the required Engineered Safeguards MINIMUM assumed time of:

A. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. 7 days C. 14 days D. 30 days Proposed Answer: B. 7 days Explanation:

A. Incorrect. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a frequently used time in technical specifications.

B. Correct. The design is for the diesels to be able to run for 7 days.

C. Incorrect. Could be thought as the time required to get FLEX equipment to become available or frequently its a allowable outage time.

D. Incorrect. A month is a plausible time that the diesels could be required due to a stranded plant condition.

Technical

References:

LJ-6B References to be provided to applicants during exam: None Learning Objective: Explain significant Diesel Generator System design features and theimportance to nuclear safety.

Question Source:

Bank #27 L111 11/2012 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 11/2012 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.8

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 073 A1.01 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits associated with operating the PRM system controls including: Radiation levels Tier #

2 Group #

1 K/A #

073 A1.01 Rating 3.2 Question 51 Unit 2 is at 100% power.

A small steam generator tube leak is causing steam line radiation monitor RM-73 to read 1000 cpm.

If power is reduced to 50%, indication on RM-73 will lower due to ___________________.

A. the burnout of Xenon B. less steam flow past RM-73 C. a decrease in I-131 production D. a decrease in N-16 production Proposed Answer: D.

a decrease in N-16 production.

Explanation:

A. Incorrect. On a down power, Xe would rise for a period of time and also do not affect indication on the steamline radiation monitors. Plausible because Xe is present and levels are proportional to power.

B. Incorrect. A lower flowrate of steam could affect air ejector reading but the steamline rad monitors are not in the flow stream but outside the piping and measure N-16.

C. Incorrect. Iodine production will decrease but the detector senses N-16 not iodine (or xenon).

D. Correct. The steam line radiation monitors detect N-16 from the tube leakage, as power is lowered, N-16 production lowers and the reading on RM-73 lowers. Once the unit is shutdown, N-16 production ceases and the indication will decrease.

Technical

References:

LG4A - Radiation Monitoring, SOE-93-001 References to be provided to applicants during exam: None Learning Objective: LG4A - Radiation Monitoring, SOE-93-001 Question Source:

Bank #25 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.5 10CFR Part 55 Content:

55.41.11

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 076 K2.01 Knowledge of bus power supplies to the following:

Service water pumps (Aux Saltwater Pumps - DCPP)

Tier #

2 Group #

1 K/A #

076 K2.01 Rating 2.7 Question 52 The power supply for ASW pump 1-1 is bus __1)___, and for ASW pump 1-2 is bus __2)___.

A. 1) F

2) G B. 1) F

2) H C. 1) G

2) H D. 1) H

2) F Proposed Answer: A. 1)

F 2) G Explanation:

A. Correct. ASW pumps are powered from bus F (1-1) and G (1-2)

B. Incorrect. This would be correct for SI pumps.

C. Incorrect. This would be correct for Containment Spray Pumps D. Incorrect. This would be correct for AFW pumps.

Technical

References:

OIM J-1-1 References to be provided to applicants during exam: None Learning Objective: 5339 -State the power supplies to ASW system components Question Source:

Bank #25 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 076 A4.02 Ability to manually operate and/or monitor in the control room: SWS valves Tier #

2 Group #

1 K/A #

076 A4.02 Rating 2.6 Question 53 Unit 1 is at 100% power.

ASW and CCW are in their normal full power alignment.

In accordance with STP I-D, Routine Monthly Checks Required by Licenses, as a minimum, what valves must be open?

NOTE:

1-FCV-496, ASW pump 1-1 discharge cross tie 1-FCV-495, ASW pump 1-2 discharge cross tie 1-FCV-602, ASW inlet to CCW heat exchanger 1-1 1-FCV-603, ASW inlet to CCW heat exchanger 1-2 A. Either FCV-495 or 496 and either FCV-602 or 603 B. Either FCV-495 or 496 and both FCV-602 and 603 C. Both FCV-495 and 496 and either FCV-602 or 603 D. Both FCV-495 and 496 and both FCV-602 and 603 Proposed Answer: C. Both FCV-495 and 496 and either FCV-602 or 603 Explanation:

A. Incorrect. A full power alignment is both cross ties open and one CCW heat exchanger in service. This is plausible as it would give the normal amount of flow required for power operation.

B. Incorrect. This is the opposite of whats required. This lineup is used in OP AP-10. Both cross ties are normally open and one CCW heat exchanger valve open.

C. Correct. Normal power lineup is both trains of ASW cross tied and one CCW heat exchanger in service.

D. Incorrect. This would be the alignment if there is abnormal conditions, such as high ASW or CCW temperature.

Technical

References:

STP I-1D References to be provided to applicants during exam: None Learning Objective: Describe the ASW System operation. (3896)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 4 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 078 K3.02 Knowledge of the effect that a loss or malfunction of the IAS system will have on the following: Systems having pneumatic valves and controls Tier #

2 Group #

1 K/A #

078 K3.02 Rating 3.4 Question 54 Unit 1 is at 100% power, at the middle of core life.

Instrument air is lost to TCV-130, Letdown Heat Exchanger CCW outlet valve.

Which of the following describes the effect on letdown system temperature and the potential reactivity effect?

Letdown temperature ___1)____ and adds ____2)____ reactivity.

A. 1) lowers

2) negative B. 1) lowers

2) positive C. 1) rises

2) negative D. 1) rises

2) positive Proposed Answer: B.

1) lowers 2) positive Explanation:

A. Incorrect. First part is correct. TCV-130 fails open. Second part incorrect. The result will be letdown temperature will lower, removing boron - adding positive reactivity, not negative.

B. Correct. TCV-130 fails open, lowering letdown temperature. Lowering temperature causes a reduction in boron concentration. This reduction in boron has a positive reactivity effect (could cause a slight rise in reactor power).

C. Incorrect. Both parts incorrect. This would be true if the valve failed closed.

D. Incorrect. First part incorrect. Second part is correct.

Technical

References:

LPA 9, LB-1A References to be provided to applicants during exam: None Learning Objective: List the effects that a loss of Instrument Air would have on the plant.

(3541)

Question Source:

Bank #07 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 103 A2.04 Ability to a) predict the impacts of the following malfunctions or operations on the Containment, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Containment evacuation (including recognition of the alarm)

Tier #

2 Group #

1 K/A #

103 A2.04 Rating 3.5 Question 55

1) The containment evacuation alarm is activated by an operator in ____________________.

2) The containment evacuation alarm is ________________________________________.

A. 1) the Control Room only

2) a constant, high-pitched tone B. 1) the Control Room only

2) an electronic warbler that sounds a series of falling tones C. 1) containment or by an operator in the Control Room

2) a constant, high-pitched tone D. 1) containment or by an operator in the Control Room

2) an electronic warbler that sounds a series of falling tones Proposed Answer: B.

1) the Control Room only 2) an electronic warbler that sounds a series of falling tones Explanation:

A. Incorrect. First part is correct. Second part is incorrect. This is the Fire alarm.

B. Correct. The operator activate the containment evacuation alarm. Second part is correct, the signal is as described in the answer.

C. Incorrect. the alarm is not activated from containment. Plausible as it is an alarm to evacuate containment. Second part is incorrect, this the fire alarm.

D. Incorrect. First part incorrect, the alarm is not activated from containment. Plausible as it is an alarm to evacuate containment. Second part is correct.

Technical

References:

LI-1, OP AP-21, GPAASite References to be provided to applicants during exam: None Learning Objective: State the purpose of Containment Structure components.

• Containment Evacuation Alarm Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 3.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.11

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO 103 G2.4.46 Containment - Ability to verify that alarms are consistent with plant conditions Tier #

2 Group #

1 K/A #

103 G2.4.46 Rating 4.2 Question 56 Unit 1 is at 100% power.

Proposed Answer: B. 3 only Explanation:

A. Incorrect. FCV-357 closes on Phase B.

B. Correct. This occurs as part of the Phase A and would confirm at least a partial Phase A has occurred.

C. Incorrect. FCV-357 closes on Phase B and the SG Blowdown IC valves close on MSI, the SGBD OC valves close on Phase A.

D. Incorrect. Only 8112 closes, the SG Blowdown inside isolation valves close on MSI, the OUTSIDE valve close on Phase A.

Technical

References:

OIM B-6-7, D-2-2 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.2 10CFR Part 55 Content:

55.41.7 A. 1 only B. 3 only C. 1 and 2 D. 2 and 3 PK02-01, CONTMT ISOLATION PHASE A/B (RED), alarms. No other alarms are received in the Control Room. The operator checks Monitor Light Box B, Containment Isolation Phase A and reports that there are indications of a partial Phase A actuation.

What indication(s), on Monitor Light Box B, could the operator have used to confirm that a partial Phase A has occurred?

1. RCP Thermal Barrier CCW Return valve FCV-357 - has closed

2. SG Blowdown Inside Containment Isolation valve FCV-760 - has closed

3. RCP Seal Return valve 8112 - has closed

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 003 AK2.05 Knowledge of the interrelations between the Dropped Control Rod and the following: control rod drive power supplies and logic circuits Tier #

1 Group #

2 K/A #

APE 003 AK2.05 Rating 2.5 Question 57 A Unit 1, Control Bank D, Group 1, rod drops into the core during power operation.

The crew is recovering the dropped rod in accordance with OP AP-12C, Dropped Control Rod.

In accordance with OP AP-12C:

1) the relay disconnect(s) for _________ will be opened prior to recovering the dropped rod.

2) _________________ should be expected when the crew begins to withdraw the dropped rod.

A. 1) only the affected rod

2) PK 03-17, ROD CONT URGENT FAILURE B. 1) only the affected rod

2) PK 03-18, ROD CONT NON-URGENT FAILURE C. 1) all unaffected rods in the bank

2) PK 03-17, ROD CONT URGENT FAILURE D. 1) all unaffected rods in the bank

2) PK 03-18, ROD CONT NON-URGENT FAILURE Proposed Answer: C.

1) all unaffected rods in the bank 2)

PK 03-17, ROD CONT URGENT FAILURE Explanation:

A. Incorrect. First part is incorrect, the dropped rod is left connected. The remaining rods in the bank are disconnected Second part is correct.

B. Incorrect. Both parts incorrect. All the unaffected rods are disconnected. The expected alarm is the Urgent failure alarm. PK03-17.

C. Correct. Both parts correct. All the rods, except the dropped rod are disconnected. When no rods in group 2 move, PK 03-17 will alarm.

D. Incorrect. First part is correct. Second part is incorrect.

Technical

References:

OP AP-12C, AR PK03-17, AR PK 03-18 References to be provided to applicants during exam: None Learning Objective: Given an abnormal condition, summarize the major actions of OP AP-12A, 12B, 12C, & 12D to mitigate an event in progress. (3477M)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 024 G2.1.30 Emergency Boration - Ability to locate and operate components, including local controls Tier #

1 Group #

2 K/A #

APE 024 G2.1.30 Rating 4.4 Question 58 In accordance with OP AP-6, Emergency Boration, the crew is aligning emergency boration using Emergency Boration valve, CVCS-8104.

1) In accordance with OP AP-6, CVCS-8104 will be opened __________________.

2) Once the valve is opened, emergency boration flow of greater than 50 gpm can be read locally at XFIT-113 located _________________________________________.

A. 1) locally

2) in the Cable Spreading Room B. 1) locally

2) at the Dedicated Shutdown Panel C. 1) from VB2

2) in the Cable Spreading Room D. 1) from VB2

2) at the Dedicated Shutdown Panel Proposed Answer: C.

1) from VB2 2) in the Cable Spreading Room Explanation:

A. Incorrect. CVCS-8104, Emergency Boration valve is operated from VB2. Manual Emergency Boration valve CVCS-8471 is opened locally. Second part is correct.

B. Incorrect. The valve is opened from VB2. DSDP is plausible as there are indications there, such as RCS cold leg temperature and steam generator pressure, but not boration flow.

C. Correct. 8104 is opened from VB2. XFIT-113 is located in the Cable Spreading Room.

D. Incorrect. First part correct. Second part is incorrect, there is no boration flow indication on the DSDP.

Technical

References:

OP AP-6, LA-8 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the Reactor Makeup Control System. (40581)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.3 Comprehensive/Analysis

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 028 AK1.01 Knowledge of the operational implications of the following concepts as they apply to PZR Level control malfunction: PZR reference leak abnormalities Tier #

1 Group #

2 K/A #

APE 028 AK1.01 Rating 2.8 Question 59 Unit 1 is at 100% power.

A leak has developed on a reference leg for a hot cal Pressurizer level channel.

The leak will cause indicated pressurizer level for the channel to be __1)__ than actual level because the DP has _____2)_____.

A. 1) lower

2) risen B. 1) lower

2) lowered C. 1) higher

2) risen D. 1) higher

2) lowered Proposed Answer: D.

1) higher 2) lowered Explanation:

A reference leg leak will result in a higher than actual pzr level indication due to a lower DP across the level transmitter - Pressure on variable leg (pressurizer) remains constant while pressure on reference leg is lowered.

A. Incorrect. Indicated level will be higher than actual level. DP lowers as the leak occurs.

B. Incorrect. First part incorrect. Second part correct.

C. Incorrect. First part is correct. Second part is incorrect.

D. Correct. The leaking reference leg causes DP to decrease, causing dp to lower which causes indicated level to increase.

Technical

References:

sensors and detectors References to be provided to applicants during exam: None Learning Objective: Given a potential failure mode for a differential pressure cell used for level indication, describe how indicated level will be affected. a. Reference leg leak or break Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 051 AA2.02 Ability to determine and interpret the following as they apply to loss of condenser vacuum: Conditions requiring reactor and/or turbine trip Tier #

1 Group #

2 K/A #

APE 051 AA2.02 Rating 3.9 Question 60 Unit 1 is at 53% power, 610 MWe.

Condenser pressure is 8.5 inches Hg absolute and rising slowly. The crew enters OP AP-7, Degraded Condenser.

1) In accordance with OP AP-7, Attachment 2, Turbine Operating Limitations, condenser pressure ______ in the Trip Turbine Immediately region.

2) If required, the operator will trip the ______________.

A. 1) is

2) reactor B. 1) is

2) turbine C. 1) is NOT

2) reactor D. 1) is NOT

2) turbine Proposed Answer: A. 1) is 2) reactor Explanation:

A. Correct. The setpoint ramps from 7.4 inches to 10.2 at full load. At 610 MWe, the setpoint is less than 8.5 inches. Despite the turbine being in the trip turbine immediately, because power is above 50% (above P-9), a reactor trip is procedurally required.

B. Incorrect. First part is correct. Second part plausible as the graph region of unacceptable operation is labelled trip turbine immediately, (procedure directs a reactor trip if above 50% power).

C. Incorrect. First part incorrect. Vacuum is above the allowable setpoint. Plausible as it is less than the full power value and if its not known the setpoint ramps.

D. Incorrect. Both parts incorrect.

Technical

References:

OP AP-7, attachment 2, OIM page B-6-2 References to be provided to applicants during exam: Attachment 2 Learning Objective: 3477G Given an abnormal condition, summarize the major actions of OP AP-7 to mitigate an event in progress.

Question Source:

Bank #60 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.2 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 061 AK2.01 Knowledge of the interrelations between the Area Radiation Monitoring System alarms and the following:

detectors at each ARM location Tier #

1 Group #

2 K/A #

APE 061 AK2.01 Rating 2.5 Question 61 Units 1 and 2 are at 100% power. Control Room Vent Train Mode Selector Switches are in Mode 1.

Control Room Vent Intake radiation monitors, 1-RE-25 and 1-RE-26, go into high alarm.

Control Room Press Intake radiation monitors, 1-RE-51 and 52, are NOT in alarm.

What is the current status of the Control Room Ventilation System?

A. Mode 3 with Unit 1 Pressurization Fan S99 running B. Mode 3 with Unit 2 Pressurization Fan S97 running C. Mode 4 with Unit 1 Pressurization Fan S99 running D. Mode 4 with Unit 2 Pressurization Fan S97 running Proposed Answer: D. Mode 4 with Unit 2 Pressurization Fan S97 running Explanation:

Knowledge of where the detectors are located, and which fan is started based on the detectors in alarm meets the KA A. Incorrect. Mode 3 is a full recirc mode of operation, originally designed for chlorine.

However, its plausible to think a full recirc mode would be preferred if there is high radiation.

B. Incorrect. Mode 3 is not used.

C. Incorrect. The opposite unit pressurization fan starts.

D. Correct. The opposite unit pressurization fan starts.

Technical

References:

LH-5 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Control Room Ventilation System - Mode 4 Operation Question Source:

Bank #37 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.11

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E06 EA1.3 Ability to operate and monitor the following as they apply to Degraded Core Cooling: Desired operating results during abnormal and emergency operations Tier #

1 Group #

2 K/A #

E06 EA1.3 Rating 3.7 Question 62 GIVEN:

The crew is performing EOP FR-C.2, Response to Degraded Core Cooling The crew is performing step 3, VERIFY ECCS Flow RCS pressure is 1100 psig and stable

1) At EOP FR-C.2, step 3, VERIFY ECCS Flow, the operator checks for a minimum flow of

_____ on FI-917, Charging Injection flow.

2) Based on the indications shown below, there should _______ greater than the minimum flow indicated on FI-917.

A. 1) 25 gpm

2) be B. 1) 25 gpm

2) NOT be C. 1) 100 gpm

2) be D. 1) 100 gpm

2) NOT be Proposed Answer: D. 1) 100 gpm 2) NOT be Explanation:

A. Incorrect. 25 gpm is used in EOP ECA-2.1 as the amount of AFW flow to each faulted steam generator. Second part is incorrect. The suction of the ECCS CCPs is through 8805A and B.

With both closed and VCT outlet vavles 112B and C closed, there will be no flow from the

DCPP L191 Exam Rev 4 RWST to either charging pump and, therefore, no charging injection. The minimum flow check will not be met. Plausible as the remainder of the required valves are open and it could be thought these valves by themselves will not hinder CCP flow.

B. Incorrect. First part incorrect, minimum flow is 100 gpm. Second part is correct.

C. Incorrect. First part is correct. Second part incorrect, for the current alignment, there will not be flow from either charging pump.

D. Correct. The step looks flow of greater than 100gpm. With the valves closed, there will be no flow to the RCS cold legs through the charging injection line and the flow indicator for the pumps will be reading 0 gpm.

Technical

References:

EOP FR-C.2, LB-3, sim reference References to be provided to applicants during exam: None Learning Objective: Describe Emergency Core Cooling System components.

• Safety Injection Pumps Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO APE 076 AK3.06 Knowledge of the reasons for the following responses as they apply to High Reactor Coolant Activity: actions contained in the EOP for high reactor coolant activity Tier #

1 Group #

2 K/A #

APE 076 AK3.06 Rating 3.2 Question 63 GIVEN:

A plant shutdown is in progress because RCS activity levels are greater than allowed by Technical Specifications, when a small break LOCA occurs The crew has transitioned to EOP E-1.2, Post LOCA Cooldown and Depressurization The crew is now preparing to establish RCP seal return flow CCW valves to the RCP have remained open In accordance with EOP E-1.2, prior to opening 8100 and 8112, RCP Seal Water Return Stop Valves, an evaluation should be performed to assess the consequences of which of the following?

A. Inter-system LOCA B. Thermal shock to the RCP seals C. Flashing in the seal water heat exchanger D. Increased radiation levels in the auxiliary building Proposed Answer: D.

Increased radiation levels in the auxiliary building Explanation:

A. Incorrect. Plausible as it is a high pressure system going to VCT pressure.

B. Incorrect. This is true if all seal injection has been lost. However, if CCW not been isolated, cooling has been to the seals has been maintained.

C. Incorrect. Plausible that the hot RCS going through the seal water heat exchanger and the step is to restore RCP seal return flow.

D. Correct. Caution at step 28 states: If excess activity levels in the RCS are suspected, then an evaluation of the consequences of re-establishing seal return flow should be made prior to placing RCP seal return flow in service.

Technical

References:

EOP E-1.2 References to be provided to applicants during exam: None Learning Objective: Explain basis of emergency procedure steps (E-1.1, E-1.2). (7920S)

Question Source:

Bank #72 L091C 03/2012 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 03/2012 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.41.12

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E02 EK3.2 Knowledge of the reasons for the following responses as they apply to SI termination: Normal, abnormal and emergency operating procedures associated with (SI Termination)

Tier #

1 Group #

2 K/A #

E02 EK3.2 Rating 3.3 Question 64 The crew is performing step 9, VERIFY ECCS Flow Not Required, of EOP E-1.1, SI Termination.

In accordance with EOP E-1.1, what parameter should the Shift Foreman have the operator check first and why is that parameter checked first?

A. Subcooling - it is the most direct indication that there is adequate core cooling.

B. Subcooling - it is the most direct indication that there are no voids in the upper head of the reactor vessel.

C. Pressurizer level - it is the most direct indication that there is adequate core cooling.

D. Pressurizer level - it is the most direct indication that there are no voids in the upper head of the reactor vessel.

Proposed Answer: A. Subcooling - it is the most direct indication that there is adequate core cooling Explanation:

A. Correct. Subcooling is the most direct check of adequate core heat removal.

B. Incorrect. While voids in the RCS are not desirable, and RVLIS is the most direct indication of voids,this is not the reason subcooling is checked first.

C. Incorrect. Pressurizer level is not checked until after subcooling and secondary heat sink.

D. Incorrect. Both parts are incorrect.

Technical

References:

E-1.1, Westinghouse Executive Volume - SI Termination References to be provided to applicants during exam: None Learning Objective: 7920S Explain basis of emergency procedure steps (E-1.1, E-1.2)

Question Source:

Bank #63 L091C 03/2012 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 03/2012 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO E13 EA1.3 Ability to operate and/or monitor the following as they apply to Steam Generator Overpressure: Desired operating results during abnormal and emergency conditions Tier #

1 Group #

2 K/A #

E13 EA1.3 Rating 3.1 Question 65 GIVEN:

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

The ruptured S/G has been identified.

1) In accordance with EOP E-3, the operator will raise the setpoint of the 10% steam dump valve on the ruptured steam generator to a setting corresponding to __________.

2) In accordance with the background document for EOP E-3, the basis for the pressure setpoint is to ensure _______________.

A. 1) 1040 psig

2) the 10% steam dump opens prior to lifting a code safety valve B. 1) 1040 psig

2) there is adequate subcooling when the RCS cooldown is complete C. 1) 1065 psig

2) the 10% steam dump opens prior to lifting a code safety valve D. 1) 1065 psig

2) there is adequate subcooling when the RCS cooldown is complete Proposed Answer: A. 1) 1040 psig 2) the 10% steam dump opens prior to lifting a code safety valve Explanation:

For a ruptured steam generator, pressure will rise once it is isolated. This is an overpressure condition which is mitigated by setting the setpoint to 1040 psig which prevents further pressure rise to the safety valve setpoint (the desired result)

A. Correct. The 10% steam dump is set to 1040 psig. Per the background document this is minimize atmospheric release but corresponds to approximately 25 psig below the lowest setpoint for the steam generator code safety valves.

B. Incorrect. First part correct. Second part incorrect. Plausible because subcooling is checked when the RCS cooldown is complete.

C. Incorrect. First part incorrect. This is the lowest setting for the steam generator code safety valve. Second part correct.

D. Incorrect. Both parts incorrect.

Technical

References:

EOP E-3, EOP E-3 background References to be provided to applicants during exam: None Learning Objective: Explain basis of emergency procedure steps Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

DCPP L191 Exam Rev 4 Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.41.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.1.3 Knowledge of Shift or short-term relief turnover practices Tier #

3 Group #

1 K/A #

G2.1.3 Rating 3.7 Question 66 According to OP1.DC31, Dissemination of Operations Information, how is a shift order that discusses a recent event transmitted to an operator?

The shift order shall be ______________

1. covered by the Shift Manager at shift brief

2. given to each operator for review and signature

3. placed in the Shift Foreman Shift Turnover notes for review at shift turnover A. 1 only B. 3 only C.

1 and 2 D. 2 and 3 Proposed Answer: A.

1 only Explanation:

Shift order book shall contain two types of information: standing orders, and shift orders. The purpose of the incident summary report is to transmit to the shift operators a concise review of any incident and its cause that the operations manager may deem important. The incident summary shall be reviewed with the crew at the shift briefing.

A. Correct. The incident shall be covered at shift brief.

B. Incorrect because it is only required to be covered by the SM at the shift briefing. Plausible because other information is placed for the SFM to review.

C. Incorrect because it is only required to be covered by the Shift Manager (SM) at the shift briefing. It is forwarded to the CRA when review is complete, and not kept in the Shift Order book. Plausible because it may seem reasonable for the SM and SFM to both have for turnover to the crews.

D. Incorrect because it is only required to be covered by the SM at the shift briefing. Plausible because other information is required to be signed for by each operator.

Technical

References:

OP1.DC31 References to be provided to applicants during exam: None Learning Objective: Discuss Operating Experience associated with Operations Department Policies and Administrative Procedures. (46416, 46639)

Question Source:

Bank #66 L161 10/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc Tier #

3 Group #

1 K/A #

G2.1.4 Rating 3.3 Question 67 In accordance with OP1.DC10, Conduct of Operations, regarding maintaining a license, the NRC shall be notified of a change to a Reactor operators license if the operator:

A. sprains their wrist and assigned light duty B. is placed on high blood pressure medication C. begins license class to obtain an SRO license D. is quarantined for 14 days due to COVID close contact Proposed Answer: B. is placed on high blood pressure medication Explanation:

A. Incorrect. While this is an injury, it is not a permanent condition or restriction B. Correct. In accordance with OP1.DC10, the following requires notifying the NRC if there are changes to:

Legal name Address Type of license (e.g., a downgrade from an SRO)

Permanent medical condition or restriction.

o New diagnosis/condition o

New prescription o

New medical equipment (CPAP, pacemaker) o Any other condition that may cause impaired judgement C. Incorrect. Upgrade to SRO will result in a new license but beginning class is not a condition to be reported to the NRC.

D. Incorrect. While the operator would not be able to stand watch, like getting the flu or other sickness it is not a condition requiring notification to the NRC.

Technical

References:

OP1.DC10 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management Tier #

3 Group #

1 K/A #

G2.1.37 Rating 4.3 Question 68 Unit 1 is at 100% power.

In accordance with OP1.ID3, Reactivity Management Program, what is the definition of a transient?

A. Any unexpected rod motion B. A load change of any magnitude C. Automatic rod motion of greater than 3 steps D. A power change in excess of the administrative limit of 5 MW/min Proposed Answer: D. A power change in excess of the administrative limit of 5 MW/min Explanation:

A. Incorrect. Rod motion requires investigation and notification of the SFM but not a transient IAW OP1.ID3.

B. Incorrect. This must be logged but not a transient unless greater than 5 mw/min.

C. Incorrect. 3 step pull and wait is the method for moving rods but greater than 3 steps is not a transient.

D. Correct. Per OP1.ID3, a transient is a planned or unplanned power change of greater than 5 MW/min (step 5.3.2)

Technical

References:

OP1.ID3 References to be provided to applicants during exam: None Learning Objective: 67250 - Describe reactivity management requirements and expectations:

Question Source:

Bank # DCPP bank P-51952 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.12

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.2.3 Knowledge of the design, procedural, and operational differences between units Tier #

3 Group #

2 K/A #

G2.2.3 Rating 3.8 Question 69 What are the maximum allowable cooldown rates for Unit 1 and Unit 2 in EOP E-0.2, Natural Circulation Cooldown?

A. The rates for Unit 1 and Unit 2 are the same, (25 °F/hour).

B. The rates for Unit 1 and Unit 2 are the same, (50 °F/hour).

C. The Unit 1 rate is half of the Unit 2 rate, (U1 - 25 °F/hour / U2 - 50 °F/hour).

D. The Unit 1 rate is twice of the Unit 2 rate, (U1 - 50 °F/hour / U2 - 25 °F/hour).

Proposed Answer: C. The Unit 1 rate is half of the Unit 2 rate, (U1 - 25 °F/hour / U2 - 50

°F/hour).

Explanation:

A. Incorrect. Unit difference, max rate for U2 is 50 F/hr B. Incorrect. Unit difference, max for U1 is 25 F/hr, U2 is 50 F/hr C. Correct. Correct rates for Unit1 and Unit 2 D. Incorrect. Correct rates but for wrong units.

Technical

References:

EOP E-0.2 - U1 & U2 References to be provided to applicants during exam: None Learning Objective: 7920C - Explain basis of emergency procedure steps (E-0.2 series)

Question Source:

Bank #69 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.2.12 Knowledge of surveillance procedures Tier #

3 Group #

2 K/A #

G2.2.12 Rating 3.7 Question 70

1) In accordance with STP I-1A, Routine Shift Checks Required by Licenses, Shift checks shall be performed on a nominal _____ hour interval.

2) In STP I-1A, what is the significance of bracketed limits, i.e. [120°F]?

A. 1) 8

2) These are the Technical Specification or Equipment Control Guideline limits.

B. 1) 8

2) These are the limits if using Control Board meters instead of the PPC.

C. 1) 12

2) These are the Technical Specification or Equipment Control Guideline limits.

D. 1) 12

2) These are the limits if using Control Board meters instead of the PPC.

Proposed Answer: C. 1) 12 2) These are the Technical Specification or Equipment Control Guideline limits.

Explanation:

A. Incorrect. A shift could be thought to be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This would be correct.

B. Incorrect. The checks are based on 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. There are times Unit 2 values, such as in OP AP-32, or PPC values are given, but in (), brackets in the surveillance are the Technical Specification or Equipment Control Guideline limits.

C. Correct. The frequency is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and brackets are the Technical Specification or Equipment Control Guideline limits.

D. Incorrect. First part is correct. Second part incorrect..

Technical

References:

STP I-1A References to be provided to applicants during exam: None Learning Objective:.

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam #

No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.4 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.2.43 Knowledge of the process used to track inoperable alarms Tier #

3 Group #

2 K/A #

G2.2.43 Rating 3.0 Question 71 In accordance with OP1.DC24, Control of Annunciator System Problems,

1) the Control Operator is required to review the open main annunciator problem evaluation sheets/defeat logs ____________________.

2) the main annunciator problem evaluation sheets/defeat logs are maintained in a/an A. 1) daily

2) binder B. 1) daily

2) electronic spreadsheet C. 1) at the beginning of each shift

2) binder D. 1) at the beginning of each shift

2) electronic spreadsheet Proposed Answer: C. 1) At the beginning of each shift 2) binder Explanation:

A. Incorrect. First part is incorrect, the review is to be done at the beginning of each shift.

Second part is correct, the review consists of reviewing the annunciators on attachment 2 of the procedure.

B. Incorrect. Some logs are taken (and then reviewed) on a daily basis, however, the annunciator review is the beginning of each shift. Second part is incorrect, while logs, trends etc are electronic, the log is in a binder.

C. Correct. The review is performed by the operator at the beginning of each shift. The sheets are maintained in a binder.

D. Incorrect. First part is correct, second part is incorrect.

Technical

References:

OP1.DC24 steps 5.4 and 5.5 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #69 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

Tier #

3 Group #

3 K/A #

G2.3.4 Rating 3.2 Question 72 Which of the exposures listed below, would be the highest exposure a male operator could receive in a year without requiring an extension above the DCPP Administrative Guideline?

A. 490 mrem B. 1450 mrem C. 1950 mrem D. 3480 mrem Proposed Answer: C.

1950 mrem Explanation:

A. Incorrect. 500 mrem is the limit for declared pregnant worker.

B. Incorrect. If thought the admin guideline limit was 1500 mrem. Is also below the 2000 mrem Admin Limit but not the highest exposure below the limit.

C. Correct. Admin guideline is 2000 mrem. To exceed, need an extension.

D. Incorrect. Would be correct if its thought the admin guideline is the Admin limit of 4000 mrem.

Technical

References:

RP1.ID6 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #72 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.3 10CFR Part 55 Content:

55.41.12

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

Tier #

3 Group #

3 K/A #

G2.3.12 Rating 3.2 Question 73 Unit 1 is at 100% power.

An operator is preparing to enter Unit 1 containment for a non-emergency entry.

In accordance with RCP D-230, Radiological Control for Containment Entry, as part of exposure control for the operator, the __________________________ shall maintain possession of the MIDS keys during the containment entry?

A. Operator making the entry B. Unit 1 Shift Foreman (or designee)

C. Work Control Shift Foreman (or designee)

D. Radiation Protection Foreman (or designee)

Proposed Answer: D. Radiation Protection Foreman (or designee)

Explanation:

A. Incorrect. While it may seem that having the operator control the key would be a positive control, the keys shall be in the possession of the RP Foreman (or designee)

B. Incorrect. The SFM authorizes entry and controls keys, but not the MIDS keys.

C. Incorrect. The WCSFM is responsible for authorizing work packages, but does not control the MIDS keys D. Correct. The RP foreman shall be in possession of the keys (or designee).

Technical

References:

RCP-230 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #71 L181 03/2020 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 03/2020 Yes Question History:

Last Two NRC Exams Yes Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.12

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.4.3 Ability to identify post-accident instrumentation Tier #

3 Group #

4 K/A #

G2.4.3 Rating 3.7 Question 74 Which of the following can be monitored on PAM 1?

A. Gammametrics B. Pressurizer level C. Steamline pressure D. Auxiliary Feedwater Flow Proposed Answer: A. Gammametrics Explanation:

A. Correct. Gammametrics is monitored by PAM 1 B. Incorrect. PAM instrument on VB2, not monitored on PAM 1 C. Incorrect. PAM on VB3, not monitored on PAM 1 D. Incorrect. PAM on VB3, not on PAM 1 Technical

References:

LB-10, Post-Accident Monitoring System References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the PAMS.

- PAMS Panel 1 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #75 L141 04/2016 X

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level RO G2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions Tier #

3 Group #

4 K/A #

G2.4.5 Rating 3.7 Question 75 The Emergency Operating Procedure (EOP) network is entered directly by entering EOP E-0, Reactor Trip or Safety Injection _____________________________.

A. only B. or EOP ECA-0.0, Loss of All Vital AC Power C. or EOP E-0.1, Reactor Trip Response D. or EOP FR-S.1, Response to Nuclear Power Generation/ATWS Proposed Answer: B. or EOP ECA-0.0, Loss of All Vital AC Power only Explanation:

Technical

References:

EOP E-0, ECA-0.0, FR-S.1 References to be provided to applicants during exam: None Learning Objective: Describe when procedure transitions are made while using the EOP set.

(7988)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 A. Incorrect. EOP E-0 is direct entry into the EOP network, but EOP ECA-0.0 can also be entered directly as well.

B. Correct Both EOP E-0 and EOP ECA-0.0 are direct entries into the EOP network..

C. Incorrect. EOP E-0.1 is always entered from EOP E-0. Plausible that if there is no SI actuation, entry into EOP E-0 is not required D. Incorrect. EOP FR-S.1 is entered from EOP E-0 or from the CSF status tree Red or Magenta, which are not monitored until exit from EOP E-0 or directed to by the procedure in effect.

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 026 G2.2.25 Loss of Component Cooling Water - Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Tier #

1 Group #

1 K/A #

APE 026 G2.2.25 Rating 4.2 Question 76 GIVEN:

Unit 1 reactor is at 100% power A 200 gpm CCW leak occurs CCW makeup is NOT available In accordance with the bases for LCO 3.7.7, Vital Component Cooling Water (CCW) System,

1) what is the minimum time the surge tank is designed to provide system make-up?

2) what is the minimum time based on?

A.

1) 5 minutes

2) time until there is system impairment due to the water loss.

B.

1) 5 minutes

2) time until there is damage to RCP seals due to the loss of cooling C.

1) 20 minutes

2) time until there is system impairment due to the water loss D.

1) 20 minutes

2) time until there is damage to RCP seals due to the loss of cooling Proposed Answer:

C. 1) 20 minutes 2) time until there is system impairment due to the water loss Explanation:

The seal package is a safety-related component and is often applicable when discussing cooling water flow to the seal package and its potential failure (creates a seal LOCA) and this lends credibility to B2 and D2. The time of 5 minutes is credible because this is often associated with the time without {all} cooling to the RCP seal package before damage to the seals occurs.

A. Incorrect. First part is incorrect, Tech spec bases states that the surge tank volume provides a minimum of 20 minutes before the system becomes impaired. OP AP-11, Appendix C provides instruction to make cooling available to a charging pump and that with no seal injection or thermal barrier cooling, the integrity could be lost in as little as 5 minutes.

However, this would not occur until after the system is impaired. Second part is correct.

B. Incorrect. First part is incorrect. Second part is incorrect.

C. Correct. Tech Spec Bases 3.7.7 states that 20 minutes based on a non-mechanistic leakage rate of 200 gpm, for operators to locate and isolate the leak or realign the CCW system into two separate vital loops before the system becomes impaired due to water loss.

D. Incorrect. First part is correct. Second part is incorrect.

Technical

References:

Tech Spec bases B3.7.7, OP AP-11 section C and Appendix C References to be provided to applicants during exam: None Learning Objective: 9694G - Apply TS 3.7 Technical Specification bases

DCPP L191 Exam Rev 4 Question Source:

Bank #77 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO EPE 038 G2.4.21 SGTR - 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 #

1 K/A #

EPE 038 G2.4.21 Rating 4.6 Question 77 GIVEN:

A steam generator tube rupture has occurred in Steam Generator 1-3 RCPs have been tripped The crew is commencing an RCS cooldown in accordance with EOP E-3, Steam Generator Tube Rupture, using the 10% steam dumps on the intact steam generators During the cooldown, Tcold in loop 1-3 drops rapidly. The STA reports the RCS Integrity Critical Safety Function status tree is MAGENTA. There are no other RED or MAGENTA paths.

What action should be taken by the Shift Foreman?

A. Immediately go to EOP FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

B. Immediately go to EOP FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition.

C. Remain in EOP E-3, status trees are monitored for information only during the cooldown.

D. Remain in EOP E-3, the indication is false and should be disregarded.

Proposed Answer:

D.

Remain in EOP E-3, the indication is false and should be disregarded.

Explanation:

A. Incorrect. This would be correct in most instances but EOP E-3 states the indication is false and should be disregarded until after SI is terminated (step 35).

B. Incorrect. Plausible that the Magenta would send the crew to P.2 but P.2 and P.1 utilize the same FR, FR-P.1.

C. Incorrect. There are procedures the CSFs are monitored for info only, ECA-0.0, E-1.3.

However, E-3 is not one of those procedures.

D. Correct. Caution in E-3 states: If RCPs are not running, the following steps may cause a false F-0.4, Integrity Status Tree indication for the ruptured loop. Disregard the ruptured loop TCOLD indication until after performing step 35 (page 26).

Technical

References:

EOP E-3 References to be provided to applicants during exam: None Learning Objective: Explain basis of emergency procedure steps (E-3 series) including:

• Bases for TCOAs with operator action of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less. (7920F)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

DCPP L191 Exam Rev 4 New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO E12 EA2.2 Uncontrolled Depressurization of all Steam Generators: Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments Tier #

1 Group #

1 K/A #

E12 EA2.2 Rating 3.9 Question 78 The crew entered EOP ECA-2.1, Uncontrolled Depressurization of All Steam Generators due to an inability to close any MSIV. SI termination is complete in accordance with EOP ECA-2.1, steps 14 to 25.

At step 26, Check RCS Hot Leg Temperatures - STABLE OR LOWERING, the operator reports MSIV, FCV-43, is closed and pressure in Steam Generator 1-3 is rising.

What action should be taken by the Shift Foreman?

A. Continue in ECA-2.1 until ALL MSIVs are closed.

B. Continue in ECA-2.1 until at least one additional MSIV is closed.

C. Stop performance of ECA-2.1 and go to EOP E-1.1, SI Termination.

D. Stop performance of ECA-2.1 and go to EOP E-2, Faulted Steam Generator Isolation.

Proposed Answer:

D.

Stop performance of ECA-2.1 and go to EOP E-2, Faulted Steam Generator Isolation.

Explanation:

A. Incorrect. Plausible that the procedure is performed until all steam generators are isolated as it could be thought that ECA-2.1 is performed in lieu of E-2.

B. Incorrect. Because there are still 3 faulted steam generators, plausible to continue on in the procedure until at least one more is isolated.

C. Incorrect. Although SI termination has been performed, a transition to E-2 is required to verify the isolation of the steam generator prior to going to E-1.1.

D. Correct. If pressure begins to rise in any steam generator, a transition is made to E-2.

Technical

References:

EOP ECA-2.1 References to be provided to applicants during exam: None Learning Objective: 5433 - Identify exit conditions for the EOPs Question Source:

Bank #80 L171 01/2019 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam Yes Question History:

Last Two NRC Exams Yes Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 056 G2.2.37 Loss of Offsite Power - Ability to determine operability and/or availability of safety-related equipment Tier #

1 Group #

1 K/A #

APE 056 G2.2.37 Rating 4.6 Question 79 GIVEN:

A loss of all AC power had occurred on Unit 1 Estimated time for offsite power restoration is 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> ASW pump 1-1 is out of service Diesel Generator 1-1 has been started and aligned to its vital 4 kV bus in accordance with ECA-0.3, Restore 4 kV Buses Diesel generator 1-1 load is 2.30 MWe Unit 2 is at full power Unit 2 ASW pump 2-1 is running Unit 2 ASW pump 2-2 is in standby ASW pump 1-2 has max demand of 375 KW.

What actions should be taken by the Shift Foreman to restore ASW to Unit 1?

A. Obtain permission from Emergency Director, perform Appendix X, Crosstie of Vital Bus, then start ASW pump 1-2.

B. Obtain permission from Site Emergency Coordinator, perform Appendix X, Crosstie of Vital Bus, then start ASW pump 1-2 C. Refer to OP AP-10, Loss of Auxiliary Salt Water, to cross-connect Unit 1 and Unit 2 ASW systems through cross-tie valve, FCV-601.

D. Refer to OP AP-10, Loss of Auxiliary Salt Water, to crosstie ASW and Circ Water Bays through ASW pump 1 bay valve, FCV-432, and Demusseling valve, FCV-604.

Proposed Answer: C. Refer to OP AP-10, Loss of Auxiliary Salt Water to cross-connect Unit 1 and Unit 2 ASW systems through cross-tie valve, FCV-601.

Explanation:

The candidate must determine if the emergency diesel is available and then upon determining it is NOT available, determine the method of restoring ASW to Unit 1.

Load limit is 2.6 MW. 2.300 +.375 = 2.675 MW. Site Emerg Coordinator permission required.

The SM or SFM would obtain this approval and direct the cross-tie of the buses. Due to the length of time to potential power restoration, the ASW pump will have to be on the diesel essentially continuously and the 2 hr/24 hour limit of 2860 cannot be assumed to apply but the continuous rating of 2600 would be the limit that applies (also exceeds 2000 hr/year rating of 2750)

A. Incorrect.The Site Emergency Coordinator (not ED) authorization is required however, the limit would be exceeded and the pump would not be started. Performing Appendix X would not result in restoration of ASW to Unit 1.

B. Incorrect. The Site Emergency Coordinator authorization is required however, the limit

DCPP L191 Exam Rev 4 would be exceeded and the pump would not be started. Performing Appendix X would not result in restoration of ASW to Unit 1.

C. Correct. Note (***) in appendix Q that If an ASW Pp cannot be energized, Then refer to OP AP-10 to cross-connect units through FCV-601 D. Incorrect. This is a step in OP AP-10 if amps on a running ASW pump is not steady and the pump is cavitating (AP-10, step 2 RNO)

Technical

References:

ECA-0.3 appendix Q and X, OP AP-10 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #90 L141, 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2106 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 077 AA2.09 Ability to determine and interpret the following as they apply to Generator Voltage and Grid Disturbances:

Operational status of the emergency diesel generators Tier #

1 Group #

1 K/A #

APE 077 AA2.09 Rating 4.3 Question 80 Unit 1 is at 100% power.

Diesel Generator 1-1 is paralleled to the Auxiliary Transformer in accordance with OP J-6B:IV, Diesel Generators - Manual Operation of DG 1-1.

1) While the diesel is paralleled to the Auxiliary Transformer, the diesel is considered

2) If a grid disturbance causes the Main Unit Transformer to trip, the operator will ensure the Auxiliary Feeder breaker opens and place the diesel control switch in _____________.

A. 1) inoperable

2) AUTO B. 1) inoperable

2) MANUAL C. 1) OPERABLE

2) AUTO D. 1) OPERABLE

2) MANUAL Proposed Answer: A. 1) inoperable 2) AUTO Explanation:

A. Correct. To perform the parallel, the diesel is placed in MANUAL (DROOP) control. This makes the diesel inoperable. OE - all Diesels were running to due a loss of startup and SRO directed the operator to place all the control switches in MANUAL and shutdown the diesels.

The crew failed to recognize the diesels were inoperable with the switches in MANUAL and resulted in entry into LCO 3.0.3. If a unit trip occurs, the feeder breaker opens automatically (procedure states ENSURE the breaker opens). Additionally, the control switch for the diesel is returned to AUTO.

B. Incorrect. First part is correct. Second part is incorrect. The control switch is in MANUAL when paralleled and must be returned to AUTO. MANUAL is plausible as this is a normal response for system upsets, such as a controller malfunction and its plausible to believe having the diesel in MANUAL is required during the resulting system upset of losing the Aux Transformer.

C. Incorrect. First part is incorrect. The diesel is inoperable when operating in MANUAL. It must be in manual (droop) to be paralleled with the grid (through the Aux transformer).

Second part is correct.

D. Incorrect. Both parts are incorrect..

Technical

References:

OP J-6B:IV, OE SAPN 50570582

DCPP L191 Exam Rev 4 References to be provided to applicants during exam: None Learning Objective: 6408 - Describe significant precautions and limitations associated with the Diesel Generator System Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 058 AA2.03 - Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems Tier #

1 Group #

1 K/A #

APE 058 AA2.03 Rating 3.9 Question 81 The crews for both units have entered EOP ECA-0.0, Loss of All Vital AC Power.

If the loss of all vital AC power is determined to last for at least ___1)_______, then the crews should ____2)_______ FSG 04, DC Bus Load Shed and Management to extend the availability of the vital batteries.

A. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) GO TO B. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) IMPLEMENT C. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) GO TO D. 1) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

2) IMPLEMENT Proposed Answer: B. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2)

IMPLEMENT Explanation:

Use of the FSG is SRO knowledge, per the FSG "With careful evaluation, directed by the SM/SEC, this guideline can be modified to suit the plant conditions existing at the time of its use. Additionally, whether the procedure is performed or implemented is not required RO knowledge (that there is a procedure to address DC load shed, would be RO knowledge, not the type of adherence required). The knowledge of how quickly to implement FSG 04 will aid the operating crew in the ability to operate and monitor the minimum key systems and indications associated with DC on a loss of all power event..

A. Incorrect. An ELAP is a loss of power for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The FSGs are implemented.

ECA-0.0 is not left.

B. Correct. The FSG is IMPLEMENTED if power is lost for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Incorrect. The procedure is implemented after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the design basis for the batteries.

D. Incorrect. The procedure is implemented, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is the design for the batteries to maintain voltage on the DC buses.

Technical

References:

ECA-0.0, FSG 04 References to be provided to applicants during exam: None Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source:

Bank #80 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New

DCPP L191 Exam Rev 4 Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 037 G2.4.8 - Steam Generator Tube Leak Knowledge of how Abnormal Operating Procedures are used in conjunction with EOPs Tier #

1 Group #

2 K/A #

APE 037 G2.4.8 Rating 4.5 Question 82 GIVEN:

RM-15R indicates a leak rate of approximately 200 gpd A plant shutdown to MODE 3 is in progress in accordance with OP AP-3, Steam Generator Tube Failure The plant trips due to loss of 12 kV bus D. All narrow range steam generator levels are 35% and rising slowly. RCS pressure is 2240 psig and rising slowly.

1) Prior to the reactor trip, in accordance with LCO 3.4.13, RCS Operational Leakage, how long did the crew have to reach MODE 3.

2) After the reactor trip, the crew will go to EOP E-0, Reactor Trip or Safety Injection the Shift Foreman should; A. 1) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

2) direct the operator to initiate Safety Injection and subsequently transition to EOP E-3, Steam Generator Tube Rupture.

B. 1) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

2) transition to EOP E-0.1, Reactor Trip Response and when complete, then complete the actions of OP AP-3 or perform in parallel if resources allow.

C. 1) 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />

2) direct the operator to initiate Safety Injection and subsequently transition to EOP E-3, Steam Generator Tube Rupture.

D. 1) 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />

2) transition to EOP E-0.1, Reactor Trip Response and when complete, then complete the actions of OP AP-3 or perform in parallel if resources allow.

Proposed Answer: B. 1) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 2) transition to EOP E-0.1, Reactor Trip Response and when complete, then complete the actions of OP AP-3 or perform in parallel if resources allow..

Explanation:

A. Incorrect. First part is correct. Second part is incorrect. Initiation of SI is not required in this case. The trip was due to a loss of the bus. However, SI and reactor trip are initiated in OP AP-3 if there is a leak of such magnitude that normal charging cannot maintain pressurizer level. Also since a tube leak was in progress and the plant tripped, its plausible that the proper procedure is now EOP E-3.

B. Correct. LCO 3.4.13, states the completion time for primary to secondary leakage is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 3. OP AP-3 must be completed, per the note, when the EOPs are complete. However, OP AP-3 also states it may be done is parallel if resources allow.

C. Incorrect. First part is incorrect. 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> would be the time if REQUIRED ACTION A.1 (reduce leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) and REQUIRED ACTION B.1 (MODE 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />)

DCPP L191 Exam Rev 4 applied. This would be correct for pressure boundary leakge - which if plausible if its believed the leaking U-tubes are pressure boundary leakage. Second part incorrect, SI and entry into EOP E-3 is not required.

D. Incorrect. First part is incorrect, second part is correct.

Technical

References:

OP AP-3, LCO 3.4.13 References to be provided to applicants during exam: None Learning Objective: 3794 Given initial conditions, assumptions, and symptoms, predict the operational implications for any size S/G tube leak.

Question Source:

Bank #78 L091C X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 03/2012 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 067 G2.4.41 Plant Fire On Site - Knowledge of the emergency action level thresholds and classifications Tier #

1 Group #

2 K/A #

APE 067 G2.4.41 Rating 4.6 Question 83 Unit 1 is at 100% power.

At 1620, fire alarms for Containment are received in the Control Room. At 1630 the alarm is confirmed and the fire is continuing.

The Shift Manager declared an Unusual Event per HU4.1 at 1636 hours0.0189 days <br />0.454 hours <br />0.00271 weeks <br />6.22498e-4 months <br />.

The declaration of HU4.1 was:

A. accurate and timely.

B. accurate but not timely.

C. timely but not accurate.

D. not accurate. The fire does not meet declaration criteria at this time.

Proposed Answer: B. accurate but not timely.

Explanation:

DCPP L191 Exam Rev 4 A. Incorrect. This time is in accordance with the guidance in the bases document. Time based EALs should be evaluated upon first indication of the conditions. If someone is working to mitigate the condition in less than the time required, the declaration can wait to see if they are successful within the time constraints. If there is indication that the threshold will be exceeded for the time period, the declaration should immediately be declared, regardless of the time remaining..Note 1 of the EALs states: The SM/SEC/ED should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

For EAL HU4.1 assessment purposes, the emergency declaration clock starts at the time that multiple alarms or indications are received, the report was received, or the time that a single alarm is confirmed by subsequent verification action. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. In this case, the clock started at 1620 not the confirmation time of 1630 and the declaration should have been made by 1635. Had this been a single alarm this would have been correct.

B. Correct. The declaration of HU4.1 is correct. However, it is outside the 15 minute time frame, which was until 1635. Candidate must know that Containment is on Table H-1 (not given as reference).

C. Incorrect. If its thought the declaration should have been HU4.2 based on single alarm then the declaration was within the 30 minute window of when the confirmation was made or 4.3 if its not known that Containment is on table H-1.

D. Incorrect. If its not known that containment is on H-1 then none of the classifications would apply at this time.

Technical

References:

EAL charts ALL, EAL Bases References to be provided to applicants during exam: None Learning Objective: As described in EP G-1, explain the time limits for emergency classifications. (42282)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.1

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO APE 069 AA2.01 Ability to determine and interpret the following as they apply to the loss of containment integrity: Loss of containment Integrity Tier #

1 Group #

2 K/A #

APE 069 AA2.01 Rating 4.3 Question 84 EOP ECA-1.1, Loss of Emergency Coolant Recirculation, is in effect.

Containment pressure rises the STA reports the Containment Critical Safety Function status tree is MAGENTA.

The Shift Foreman should:

A. remain in EOP ECA-1.1 and direct the operators to verify all available CFCUs are running.

B. remain in EOP ECA-1.1, however, direct the operators to start and operate the containment spray pumps in accordance with EOP FR-Z.1.

C. go to EOP FR-Z.1 and direct the operators to start and operate the containment spray pumps in accordance with EOP FR-Z.1.

D. go to EOP FR-Z.1, however, direct the operators to start and operate the containment spray pumps in accordance with EOP ECA-1.1.

Proposed Answer: D. go to EOP FR-Z.1, however, direct the operators to start and operate the containment spray pumps in accordance with EOP ECA-1.1 Explanation:

A. Incorrect. There is an action in EOP ECA-1.1 checking the status of the CFCUs and its plausible to think the spray pumps are not started to conserve RWST inventory.

B. Incorrect. This is the reverse of the correct action. Plausible to think the FR procedure is used as a reference to guide the operation of the spray pumps and that EOP ECA-1.1 is the priority procedure to be performed.

C. Incorrect. This is the normal response if EOP FR-Z.1 were entered from other EOPs.

D. Correct. EOP FR-Z.1 must be entered to address the severe challenge to Containment Integrity, however because a goal of EOP ECA-1.1 is to conserve RWST inventory, the action is to operate the spray pumps using the guidance of EOP ECA-1.1, not EOP FR-Z.1.

Technical

References:

EOP FR-Z.1, EOP ECA-1.1 References to be provided to applicants during exam: None Learning Objective: Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event. (3552)

Question Source:

Bank #80 L161 10/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 2.5 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO E15 EA2.1 Ability to determine and interpret the following as they apply to the Containment Flooding: Facility conditions and selection of appropriate procedures during abnormal and emergency operations Tier #

1 Group #

2 K/A #

E15 EA2.1 Rating 3.2 Question 85 GIVEN:

The crew is performing EOP FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, due to a valid MAGENTA path for RCS Integrity The CSF for RCS Integrity is currently YELLOW.

The STA reports that Containment Sump level is 99 feet and rising slowly.

What action should be taken by the Shift Foreman?

A. Continue in EOP FR-P.1 because it must be completed prior to addressing the MAGENTA Containment Integrity CSF.

B. Continue in EOP FR-P.1 because Containment Integrity CSF is YELLOW and a lower priority than RCS Integrity.

C. Go to EOP FR-Z.2, Response to Containment Flooding because Containment Integrity CSF is MAGENTA and now a higher priority the RCS Integrity.

D. Go to EOP FR-Z.2, Response to Containment Flooding because Containment Integrity CSF is RED and now a higher priority the RCS Integrity.

Proposed Answer: A. Continue in EOP FR-P.1 because it must be completed prior to addressing the MAGENTA Containment Integrity CSF.

Explanation:

A. Correct. Per the rules of usage, the Containment Integrity CSF is Magenta. The RCS Integrity CSF, which was entered on a Magenta path, once entered must be completed unless a higher CSF challenge (Red or higher priority Magenta) occurs. Containment Integrity is Magenta but a lower priority. Therefore the proper action is to complete EOP FR-P.1 prior to addressing containment flooding.

B. Incorrect. While the action is to stay in EOP FR-P.1, the reason is incorrect. Containment Integrity is Magenta due to containment flooding.

C. Incorrect. Per the rules of usage, despite the clearing of the Magenta for RCS Integrity, it must be performed until directed to exit.

D. Incorrect. This would be correct if containment flooding was a Red path.However, the only Red challenge for Containment Integrity is containment pressure.

Technical

References:

EOP F-0 References to be provided to applicants during exam: None Learning Objective: Apply the Rules of Usage in EOPs for the CSFSTs and FRGs, including:

• the six status trees

• the priority of use of the status trees

• the priority of use of the color of each CSF

DCPP L191 Exam Rev 4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 005 A2.02 Ability to a) predict the impacts of the following malfunctions or operations on the RHRS, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Pressure transient protection during cold shutdown Tier #

1 Group #

2 K/A #

005 A2.02 Rating 3.7 Question 86 Unit 1 is solid in MODE 5.

A pressure transient causes RHR Suction Relief valve, RV-8707 to lift and it fails to reseat. PRT level is rising.

The Shift Foreman should go to:

A. OP AP-16, Malfunction of the RHR System B. OP AP-24, Shutdown LOCA C. OP AP SD-2, Loss of RCS Inventory D. OP AP SD-5, Loss of Residual Heat Removal Proposed Answer: C. OP AP SD-2, Loss of RCS Inventory Explanation:

SRO level - entry into abnormal procedures that are not major abnormal procedures. SRO must decide on procedure based on mode of applicability and which SD is applicable.

A. Incorrect. This procedure is in Mode 4 and deals with a loss of RHR flow.

B. Incorrect. There is a LOCA however, OP AP-24 is used in MODE 3 after isolating accumulators and in MODE 4.

C. Correct. OP AP SD-2 is used in MODEs 5 and 6 for a loss of RCS inventory. The leak will be isolated at step 6 when RHR integrity is checked.

D. Incorrect. OP AP SD-5 is used in MODEs 5 and 6 but for a loss of RHR flow. It does not deal with a break. Specifically states that it is not used if there is an inventory problem.

Technical

References:

LB-2, OP AP-16, OP AP-24, OP AP SD-2, OP AP SD-5 References to be provided to applicants during exam: None Learning Objective: Given initial conditions, assumptions, and symptoms, determine the correct abnormal operating procedure to be used to mitigate an operational event. (3478)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 006 G2.2.40 ECCS - Ability to apply Technical Specifications for a system Tier #

2 Group #

1 K/A #

006 G2.2.40 Rating 4.7 Question 87 Unit 1 is at 100% power. The crew is verifying ECCS inspection points have detectable water levels by performing STP M-89, ECCS System Venting.

It is determined that ECCS inspection point RHR-1-998, 73 HX Rm (RHR HX 1-1 hi pt), is NOT FULL.

1) What is the basis for verifying the ECCS inspection points have detectable water levels?

2) What Technical Specification LCO should be entered as a result of the RHR-1-998, NOT FULL, condition?

A. 1) Prevent water hammer of ECCS piping.

2) LCO 3.0.3 B. 1) Prevent water hammer of ECCS piping.

2) LCO 3.5.2, ECCS - Operating C. 1) Prevent gas binding of the running CCP.

2) LCO 3.0.3 D. 1) Prevent gas binding of the running CCP.

2) LCO 3.5.2, ECCS - Operating Proposed Answer: B. 1) Prevent water hammer of ECCS piping. 2) LCO 3.5.2, ECCS -

Operating Explanation:

STP M-89 checks for minimum water levels at all ECCS vent (inspection) locations. These levels represent the levels which can challenge the ECCS and RCS components by voiding, water hammer, pump cavitation, gas binding or pumping of non-condensable gases into the reactor vessel following an SI signal or during shutdown cooling.

Candidate must determine the bases - (both are part of the SR Bases) and also the vent point is such that only one train is inoperable, therefore, only LCO 3.5.2 Action A is applicable.

A. Incorrect. Only the train is affected, (supported by note in STP M-89 table, "Train AOT").

Reason is correct.

B. Correct. The impact is water hammer (per the bases) and only the train is impacted, and LCO 3.5.2, Action A applies C. Incorrect. The impact is only on one train. Per the table in STP M-89, for charging pumps, the impact is "void migration" which means the impact is a void being transferred to another location.

D. Incorrect. Does not impact the CCP. Gas binding is discussed in bases but the charging pump is not impacted.

Technical

References:

STP M-89 Attachment 9.1 and LCO 3.5.2 and bases References to be provided to applicants during exam: None

DCPP L191 Exam Rev 4 Learning Objective: 9694E - Apply TS 3.5 Technical Specification bases Question Source:

Bank #78 L171 01/2019 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2019 Yes Question History:

Last Two NRC Exams Yes Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.3

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 008 A2.03 Ability to a) predict the impacts of the following malfunctions or operations on the CCW, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High/Low CCW Temperature Tier #

2 Group #

1 K/A #

008 A2.03 Rating 3.2 Question 88 Unit 1 is at 100% power.

The operator reports:

Ocean temperature has been slowly rising and is now 65°F CCW heat exchanger outlet temperature is 80°F and rising slowly The Shift Foreman should direct the crew to:

A. commence a plant shutdown to comply with LCO 3.0.3 because there are no OPERABLE CCW loops due to the high ocean temperature.

B. trip the reactor and all RCPs in accordance with OP AP-11, Loss of Component Cooling Water System, to prevent damage to the RCP thermal barrier.

C. trip the reactor and all RCPs in accordance with OP AP-11, Loss of Component Cooling Water System, to prevent possibly exceeding CCW maximum design temperature during a design basis accident.

D. place a second CCW heat exchanger in service, in accordance with OP E-5:II, Auxiliary Saltwater System Two CCW Heat Exchanger Operation, to prevent possibly exceeding CCW maximum design temperature during a design basis accident.

Proposed Answer: D. place a second CCW heat exchanger in service, in accordance with OP E-5:II, Auxiliary Saltwater System Two CCW Heat Exchanger Operation, to prevent possibly exceeding CCW maximum design temperature during a design basis accident.

Explanation:

A. Incorrect. ASW cools CCW using the ocean (the ultimate heat sink). The Ultimate heat sink is inoperable due to high temperature, but it does not result in cascading to the CCW LCO and making CCW inoperable. If it did, then both trains would be inoperable and there is no action in LCO 3.7.7 for two inoperable trains and entry into 3.0.3 would be required.

B. Incorrect. 80F is elevated CCW temperature (usually in the 60s), coupled with rising ASW temperature, this abnormal situation could indicate a need to trip. If there was no ASW flow or CCW temperature reached 120F, then by procedure the reactor and RCPs would be tripped. Because the RCP thermal barriers are cooled by CCW and tripped in the procedure, the reason is credible.

C. Incorrect. Correct reason, but a reactor trip is not required at this time.

D. Correct. The action of LCO 3.7.9 is to place a second heat exchanger in service, this is done, IAW OP E-5:II, to keep CCW temperature below design if a DBA occurred (and is effective as long as ocean temperature remains below 70°F) Technical

References:

OP AP-11, LCO 3.7.7, 3.7.9 Technical

References:

OP AP-11, LCO 3.7.7, 3.7.9, Bases 3.7.9, OP E-5:II

DCPP L191 Exam Rev 4 References to be provided to applicants during exam: None Learning Objective: 9694G Apply TS 3.7 Technical Specification bases Question Source:

Bank #79 L121 08/2014 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 08/2014 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 061 G2.2.25 AFW - Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits Tier #

2 Group #

1 K/A #

061 G2.2.25 Rating 4.2 Question 89 Both Unit 1 and Unit 2 are at 100% power.

1) According to the bases for LCO 3.7.6, the limiting event for the minimum required CST volume is ____________________________.

2) If CST level is less than required, the Shift Foreman will enter LCO 3.7.6_______________.

A. 1) natural circulation cooldown to RHR

2) only B. 1) natural circulation cooldown to RHR

2) and LCO 3.7.5, Auxiliary Feedwater (AFW) System C. 1) large feedwater line break coincident with a loss of offsite power

2) only D. 1) large feedwater line break coincident with a loss of offsite power

2) and LCO 3.7.5, Auxiliary Feedwater (AFW) System Proposed Answer: A. 1) natural circulation cooldown to RHR 2) only Explanation:

A. Correct. Per the bases for LCO 3.7.6, The limiting event for AFW supply, i.e., CST minimum tank volume, is based on a natural circulation cooldown due to a loss of offsite power using seismically-qualified water sources.

Per OP1.DC38, 5.10 Common Support Systems, 5.10.2 - CST - The AFW system will not be able to perform its design function without a supply of water for RCS decay heat removal via the SGs. The Required Actions for inoperability of the CST is more restrictive than for the case if all three AFW trains are inoperable. The appropriate action is to follow the TS Required Actions for an inoperable CST and not to enter the Required Actions for an inoperable AFW system. The LOSF evaluation will conclude that although there is a degradation for maintaining an AFW heat sink there is not a loss of safety function as long as there is useable inventory. The CST Required Actions are bounding for this case.

B. Incorrect. First part is correct. Second part incorrect, only the action of LCO 3.7.6 is entered.

C. Incorrect. First part incorrect. Plausible because it is discussed in the bases for 3.7.6 and because most a large part of Technical Specifications is to ensure equipment required to meet design basis accidents is OPERABLE. Second part correct.

D. Incorrect. Both parts are incorrect.

Technical

References:

B3.7.6, References to be provided to applicants during exam: None Learning Objective: 9694G Apply TS 3.7 Technical Specification bases Question Source:

Bank #

DCPP L191 Exam Rev 4 (note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 3.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 062 G2.2.22 AC Distribution: Knowledge of limiting conditions for operations and safety limits Tier #

2 Group #

1 K/A #

062 G2.2.22 Rating 4.7 Question 90 Unit 1 is at 100% power.

In accordance with OP J-2:VIII, Guidelines for Reliable Transmission Service for DCPP, what condition(s) is/are required for 500 kV to be OPERABLE?

1. At least one Mesa transformer in service

2. At least one 500 KV line connected to the grid with minimum voltage

3. Los Padres area load greater than or equal to the limit for Normal Transfer Level A. 1 only B. 2 only C. 1 and 3 D. 2 and 3 Proposed Answer: B. 2 only Explanation:

A. Incorrect. Mesa transformers are checked when determining 230 kV OPERABILITY when a line is out of service B. Correct. At least one line connected to the grid (and voltage greater than 512 kV) is required for 500 kV to be OPERABLE.

C. Incorrect. Los Padres area load is checked for 230 kV OPERABILITY.

D. Incorrect. At least one line connected is a requirement, Los Padres load is not (it is a requirement for 230 kV OPERABILITY).

Technical

References:

B3.8.1, J-2:VIII References to be provided to applicants during exam: None Learning Objective: 9697H Apply TS 3.8 Technical Specification LCOs Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 017 A2.02 Ability to a) predict the impacts of the following malfunctions or operations on the ITM, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Core Damage Tier #

2 Group #

2 K/A #

017 A2.02 Rating 4.1 Question 91 The crew is performing EOP ECA-0.0, Loss of All Vital AC Power.

In EOP ECA-0.0, at step 25, CHECK Core Exit TCs - Less Than 1200°F, all core exit thermocouples are greater than 1200°F and rising.

The Shift Foreman should:

A. continue in EOP ECA-0.0.

B. IMPLEMENT FSG 01, Long Term RCS Inventory Control.

C. GO TO EOP FR-C.1, Response to Inadequate Core Cooling.

D. GO TO SAG 01, Control Room Initial Response.

Proposed Answer: D. GO TO SAG 01, Control Room Initial Response Explanation:

A. Incorrect. 1200°F will cause the Critical Safety Function (CSF) status tree to be RED but in EOP ECA-0.0 CSF status trees are monitored for information only. If its thought the RED path is not addressed, or if its thought the procedure has steps to address the condition (such as increasing AFW flow to avoid a loss of heat sink, which is what the procedure does) then continuing in the procedure is plausible.

B. Incorrect. If its known that FR-C.1 is not correct but knows something is done, this FSG seems like a plausible choice.

C. Incorrect. A transition is normally required, however, in EOP ECA-0.0 the FRPs are not used because there is no power for equipment.

D. Correct. For the current conditions, a transition to SAG01 is required based CETCs greater than 1200°F and rising. According to the background document 1200°F is an indication of inadequate core cooling which leads to core damage.

Technical

References:

EOP ECA0.0, background document for ECA-0.0 References to be provided to applicants during exam: None Learning Objective: 3552 - Given initial conditions, assumptions, and symptoms, determine the correct Emergency Operating Procedure to be used to mitigate an operational event Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0

DCPP L191 Exam Rev 4 10CFR Part 55 Content:

55.43.

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 034 K4.01 Knowledge of FHS design feature(s) and interlock(s) which provide for the following: Fuel protection from binding and dropping Tier #

2 Group #

2 K/A #

034 K4.01 Rating 3.4 Question 92 While a bowed fuel assembly is being loaded into the core, the Refueling SRO observes the SLACK CABLE light energize.

What would be an indication that the fuel assembly was properly loaded onto the core plate and not hung up on an adjacent assembly?

A. Underload light is lit.

B. Z-Z tape indicates full down.

C. Minimal load indicated on the fuel cell.

D. Verification that the assembly has been lowered on index.

Proposed Answer: B. Z-Z tape indicates full down.

Explanation:

A. Incorrect. underload would be encountered for both situations.

B. Correct. per OP B-8DS2,6.9.13 indications are:

• Minimal load is indicated on the load cell

• SLACK CABLE light is ON

• TUBE DOWN light is ON

• Z Z tape indicates full down

• Expected Gemco position for down on core plate C. Incorrect. this indication could indicate either condition.

D. Incorrect. if the assembly is bowed, it may have to be loaded off-index (as stated in precaution 5.4.11.a:If a fuel assembly being lowered is bowed or out of plumb such that the bottom nozzle is off the core location when the crane is indexed, or if the top of an adjacent assembly is violating the space for an assembly being lowered, it may be necessary to move the crane off index to permit entry into the core.) Additionally, just a verification that the assembly was on index does not mean it is on the core plate.

Technical

References:

OP B-8DS2 References to be provided to applicants during exam: None Learning Objective: 36964 - Describe controls, indications, and alarms associated with the Fuel Handling system Question Source:

Bank #93 L141 04/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 04/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO 071 A2.09 Ability to a) predict the impacts of the following malfunctions or operations on the WGS, and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck open relief valve Tier #

2 Group #

2 K/A #

071 A2.09 Rating 3.5 Question 93 GIVEN:

VCT pressure lowers rapidly to 0 psig Letdown flow is 75 gpm PK11-25, PLANT VENT RADIATION, is in alarm PK11-21, HIGH RADIATION, is in alarm The operator reports VCT relief valve, RV-8120, is open.

1) The Shift Foreman should enter ___________.

2) Per EP RB-2, DCPP Emergency Exposure Guidelines, if Emergency Exposure authorization is required to isolate the valve and the TSC and EOF are not activated, it should be approved by the _______________________.

A. 1) OP AP-14, Tank Ruptures

2) Shift Manager B. 1) OP AP-14, Tank Ruptures

2) Station Director C. 1) OP AP-17, Loss of Charging

2) Shift Manager D. 1) OP AP-17, Loss of Charging

2) Station Director Proposed Answer: A. 1)

OP AP-14, Tank Ruptures 2) Shift Manager Explanation:

A. Correct. Low VCT pressure and level are indicative of a VCT rupture. The procedure to address it is OP AP-14. In AP-14, step 4 is to implement radiological emergency procedures.

First bullet is RB-2. RB-2 states the SM or the SEC or ED approves emergency exposures.

B. Incorrect. The procedure is AP-14, however, the SM not the Station Director approves emergency exposures. Plausible as the SD is in charge of operations.

C. Incorrect. PK11-21 could be in alarm for either loss of charging or tank rupture. Also, VCT level would lower for a large charging line leak (but pressure would not fall to 0 psig). Also, RV-8120 could be confused with RV-8117 which is the letdown relief valve. The SEC approves emergency exposure.

D. Incorrect. The appropriate procedure is AP-14 not AP-17 and the SM approves emergency exposures.

Technical

References:

OP AP-14, OP AP-17, EP RB-2

DCPP L191 Exam Rev 4 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #86 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 2.6 10CFR Part 55 Content:

55.43.4

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.1.34 Knowledge of primary and secondary plant chemistry limits Tier #

3 Group #

1 K/A #

G2.1.34 Rating 3.5 Question 94 Unit 2 is at 3% and raising power in accordance with OP L-3, Secondary Plant Startup.

Chemistry reports the following Reactor Coolant System sample results:

Chloride concentration is 0.66 ppm Fluoride concentration is 1.13 ppm Oxygen concentration is 1.10 ppm The Shift Foreman will enter Condition A:

A. only B. and Condition C only C. and Condition E only D. Condition C and Condition E Proposed Answer: B. and Condition C only Explanation:

The Surveillance Requirements must be evaluated. The limits are: steady state 0.01 for dissolved oxygen and 0.15 for chloride and fluoride. The transient limits are 10x higher. Based on the results, all are above the steady state limit and oxygen is above the transient limit.

However, only oxygen is above is steady state limit of 1.0 ppm. With all above the steady state limit, the unit is in ECG 7.4, Condition A. Additionally, because oxygen is above the transient limit, ECG 7.4 Condition C also applies. Because fluoride and chloride are below their transient limits, Condition E does not apply.

A. Incorrect. This is plausible if the limit of oxygen is misapplied (read as 1.5 vice the 1.0 ppm)

B. Correct. Because oxygen is above 1.0 ppm, Condition C applies.

C. Incorrect. Condition E applies if chloride and/or fluoride is above the transient limit, they are not.

D. Incorrect. This would be true if either chloride and/or fluoride were above the transient limit.

Technical

References:

ECG 7.4 References to be provided to applicants during exam: ECG 7.4 Learning Objective: 66040 -Apply the requirements of System 7 ECGs Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 4 Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.2

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.1.42 Knowledge of new and spent fuel movement procedures Tier #

3 Group #

1 K/A #

G2.1.42 Rating 3.4 Question 95 In accordance with OP B-8DS1, Core Unloading, the Refueling SRO has the responsibility for which of the following?

1. Stopping fuel movement if Source Range counts double on one channel

2. Supervision of core alteration activities

3. Providing technical guidance and trending of source range count rates.

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

Per OP OP B-8DS1 (Core Unload), the following are the responsibility of the refueling SRO:

• Direct supervision of CORE ALTERATION activities with no concurrent duties.

• All fuel handling operations.

• Safe and orderly evacuation of the refueling crew in the event of a high radiation alarm at a refueling station.

• Determining the cause of high radiation alarms.

• Determining that fuel handling personnel are properly qualified for their duty stations.

A. Incorrect. Fuel unloading would be stopped by the Refueling SRO if counts increase by a factor of 3 on a single channel or 2 on all channels.This is not the case, only one channel has doubled.

B. Correct. This is the only one listed that is the responsibility of the refueling SRO C. Incorrect. #1 is a duty of the operator in the control room, #3 is the duty of the reactor engineer. Plausible as the refueling SRO provides direct oversight and these could be thought of as falling under that definition.

D. Incorrect. 3 is the responsibility of the reactor engineer Technical

References:

OP B-8DS1 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #94 L161 10/2016 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5

DCPP L191 Exam Rev 4 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.7

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.2.13 Knowledge of tagging and clearance procedures Tier #

3 Group #

2 K/A #

G2.2.13 Rating 4.3 Question 96

1) In accordance with OP2.ID2, Tagging Requirements, what should be done if a control board caution (CBC) tag is for a component to be opened but whose switch position cannot be maintained, (e.g. spring return to NEUTRAL)?

2) In accordance with OP2.ID2, in an emergency, the Shift Foreman ________ have the authority to authorize operation of a CBC tagged component.

A. 1) Mark the CBC tag OPEN-THEN-AUTO

2) does B. 1) Mark the CBC tag OPEN-THEN-AUTO

2) does NOT C. 1) Place a pink tag on the control switch

2) does D. 1) Place a pink tag on the control switch

2) does NOT Proposed Answer: A. 1)

Mark the CBC tag OPEN-THEN-AUTO 2) does Explanation:

A. Correct. The procedure states the SFM may authorize operation of the caution tagged component. The CBC tag is marked as noted in the answer.

B. Incorrect. First part is correct. The SFM has the authority. Plausible that its thought that responsibility lies with the SM.

C. Incorrect. Pink Tags for status control prior to issuing a clearance or in cases where procedures to not maintain status control. Second part is correct, the SFM has the authority.

D. Incorrect. Both parts are incorrect.

Technical

References:

OP2.ID2, OP1.DC10 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 3.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.2.20 Knowledge of the process for managing troubleshooting activities Tier #

3 Group #

2 K/A #

G2.2.20 Rating 3.8 Question 97 In accordance with MA1.ID26, Troubleshooting:

1) Operator diagnostics such as alarm response and instrumentation validation __________

require entry into MA1.ID26.

2) Level A troubleshooting, (troubleshooting activities with high risk) shall be approved by the A. 1) do

2) Shift Foreman B. 1) do

2) Shift Manager C. 1) do NOT

2) Shift Foreman D. 1) do NOT

2) Shift Manager Proposed Answer: D. 1) do NOT 2) Shift Manager Explanation:

A. Incorrect. Normal operator diagnotics do not require entry. The SM shall approve level A troubleshooting. Plausible the SFM, who approves most work, would approve troubleshooting on the unit.

B. Incorrect. First part incorrect. Second part correct.

C. Incorrect. First part is correct. Second part incorrect. SM approval required.

D. Correct. Per MA1.ID26, Normal operator diagnostics such as alarm response and instrumentation validation do not require entry into this procedure. Direct component manipulation or operational maneuvering using approved plant procedures is allowed outside of this procedure to establish initial condition assessment. SM approval is required for Level A (high risk troubleshooting).

Technical

References:

MA1.ID26 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 4 Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.10

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.3.6 Ability to approve release permits Tier #

3 Group #

3 K/A #

G2.3.6 Rating 3.8 Question 98 A gas decay tank discharge is being setup in accordance with OP G-2:V, Gaseous Radwaste -

Gas Decay Tank Discharge.

In accordance with Form 69-21595, Gas Decay Tank Discharge Authorization, ____1)______ is responsible for preparing and the _____2)______ is responsible for approving a Unit 1 gaseous radwaste discharge permit.

A. 1) Chemistry

2) Shift Foreman B. 1) Chemistry

2) Shift Manager C. 1) Radiation Protection

2) Shift Foreman D. 1) Radiation Protection

2) Shift Manager Proposed Answer: A. 1) Chemistry 2) Shift Foreman Explanation:

A. Correct. The permit is prepared by chemistry and approved by the Shift Foreman.

B. Incorrect. The shift manager has overall control of the plant, but the SFM approves work or discharges on their unit.

C. Incorrect. While offsite dose (RP) is part of the permit, the calculation and preparation of the permit is done by chemistry.

D. Incorrect. Prepared by chemistry, approved by the SFM..

Technical

References:

OP G-2:V, Form 69-21595 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #98 L162 01/2018 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 2.5 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.4

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.4.13 Knowledge of crew roles and responsibilities during EOP usage Tier #

3 Group #

4 K/A #

G2.4.13 Rating 4.6 Question 99

1) In accordance with OP1.DC10, Conduct of Operations, if an update is used during EOP procedure transitions, at a minimum, the SFM shall cover the Critical Safety Function Tree status and _________________________________.

2) If the crew was required to transition to EOP E-3, Steam Generator Tube Rupture, the SFM will perform ________________.

A. 1) current EAL classification and upgrade criteria

2) an update B. 1) plant direction and major actions to take

2) an update C. 1) current EAL classification and upgrade criteria

2) a procedure transition brief D. 1) plant direction and major actions to take

2) a procedure transition brief Proposed Answer: A. 1) current EAL classification and upgrade criteria 2) an update Explanation:

Candidate must know the generic concept of update vs transition brief usage for EOPs and that that Time Critical Operator Actions in an EOP will determine whether an update or transition brief is required. Additionally, the SRO must know what must be covered in an update vs a transition brief.

A. Correct. First part is correct. EAL classification and upgrade criteria shall be covered Second part correct. For procedures with TCOAs, an update is used. EOP E-3 is one of those procedures.

B. Incorrect. First part is incorrect. Major actions are part of a transition brief. This is plausible to think the SFM would give the crew a road map of what the milestones are in the procedure to be entered and EALs determined by the SM with assistance from the STA, so plausible that as this is SM/STA function, it is not covered prior to performing an EOP with time critical actions. Second part correct, update is used per OP1.DC10 for EOPs with time critical actions.

C. Incorrect. First part is correct. EAL classification and upgrade criteria are required.Second part incorrect. EOP E-3 is an EOP with time critical actions, a transition brief is done for procedures w/o time critical actions D. Incorrect. Both parts correct.

Technical

References:

OP1.DC10, EOP E-3 References to be provided to applicants during exam: None Learning Objective:

DCPP L191 Exam Rev 4 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental 3.0 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.5

DCPP L191 Exam Rev 4 Examination Outline Cross-Reference Level SRO G2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations Tier #

3 Group #

4 K/A #

G2.4.22 Rating 4.4 Question 100 A Functional Restoration Guideline, which is used for either a RED or MAGENTA condition, is being performed by the crew due to the MAGENTA condition being met.

The challenged CSF Status Tree turns to RED.

The Shift Foreman should:

A. return to step 1 because the previous steps should be reperformed to address degrading plant conditions and the actions taken may now be different.

B. return to step 1 because this is the proper action based on the rules of usage in accordance with EOP F-0, Critical Safety Function Status Trees.

C. continue in the guideline in effect from the current step because the corrective actions for the RED and MAGENTA conditions are the same.

D. continue in the guideline in effect from the current step because the MAGENTA path will have a defined transition to the RED path.

Proposed Answer: C. continue in the guideline in effect from the current step because the corrective actions for the RED and MAGENTA conditions are the same.

Explanation:

The question tests the actions and reason taken when the RED and MAGENTA CSF status trees refer to the same procedure. FRGs such as EOP FR-S.1, or FR-P.1 is used for RED and MAGENTA challenges.

A. Incorrect. Plausible that because the CSF going RED would indicate degrading plant conditions and going back to the beginning would be a seemingly logical step.

B. Incorrect. This is the action if the procedures are not the same, such as for Core Cooling and then the higher priority RED challenge would be addressed.

C. Correct. Per EOP F-0, IF a Functional Restoration Guideline is in progress due to a severe challenge (MAGENTA PATH) AND the CSF Status Tree goes to an extreme challenge (RED PATH) AND references the SAME guideline, THEN the operator should continue in the guideline from the current step since the corrective actions are the same regardless of the severity of the challenge.

D. Incorrect. This is true if the challenge is not a higher challenge.

Technical

References:

EOP F-0 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No

DCPP L191 Exam Rev 4 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis 3.0 10CFR Part 55 Content:

55.43.5

Cover sheet for Handouts for Written Exam

Loss of Emergency Coolant Recirculation EOP ECA-1.1 R27E Page 25 of 32 UNIT 1 EOP_ECA-1!1u1r27.DOC 1227.1508 APPENDIX G Minimum ECCS Flow Rate After Trip 0

100 200 300 400 500 600 700 800 10 100 1000 10000 TIM E AFTER T R IP (M IN U TES)

MINIMUM REQUIRED FLOW RATE (GPM)

Appendix G, Page 1 of 1

Turbine Operating Limitations OP AP-7 R51 Page 25 of 31 U1&2 Attachment 2: Page 1 of 1 OP_AP-7u3r51.DOC 0929.2228 TURBINE OPERATING LIMITATIONS (Breaker CLOSED) 0 1

2 3

4 5

6 7

8 9

10 0

120 240 360 480 600 720 840 960 1080 1200 TURBINE LOAD, MWe C

O N

D P

R E

S S

U R

E, IN H

G A

B S

TRIP TURBINE IMMEDIATELY ACCEPTABLE OPERATING REGION 11 600 1182 7.2 10.2 TURBINE TRIP SETPOINT TURBINE OPERATING LIMITATIONS (Breaker OPEN) 0 1

2 3

4 5

6 7

8 9

10 0

180 360 540 720 900 1080 1260 1440 1620 1800 TURBINE SPEED, RPM C

O N

D P

R E

S S

U R

E, IN H

G A

B S

TRIP TURBINE IMMEDIATELY ACCEPTABLE OPERATING REGION 11 1260 1620 7.2 10.2 TURBINE TRIP SETPOINT

Chemistry ECG 7.4 DIABLO CANYON - UNITS 1 & 2 Effective Date:____________________

Rev. 3 ECG_7!4u3r03.DOC 1019.1129 Page 1 of 4 7.0 REACTOR COOLANT SYSTEM 7.4 Chemistry ECG 7.4 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 7.4-1.

APPLICABILITY: At all times.

ACTIONS In MODES 1, 2, 3, and 4:

CONDITION REQUIRED ACTION COMPLETION TIME A.

Any one or more chemistry parameter in excess of its Steady-State Limit BUT Within its Transient Limit A.1 Restore the parameter to within its Steady-State Limit 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B.

Action A.1 not done within the required Completion Time B.1 Be in at least MODE 3 AND 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.2 Be in at least MODE 5 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.

Any one or more chemistry parameter in excess of its Transient Limit C.1 Perform Action B.1 AND As specified in Action B.1 C.2 Perform Action B.2 As specified in Action B.2 (continued) 11/10/09

Chemistry ECG 7.4 DIABLO CANYON - UNITS 1 & 2 Rev. 3 ECG_7!4u3r03.DOC 1019.1129 Page 2 of 4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D.1 Reduce the pressurizer pressure to less than or equal to 500 psig, if applicable AND Immediately D.2 Perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine if the Reactor Coolant System remains acceptable for continued operation Prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to Mode 4 E.

Concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Transient Limit E.1 Perform Action D.1 AND As specified in Action D.1 E.2 Perform Action D.2 As specified in Action D.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 7.4 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters in Table 7.4-1.

At the frequencies specified in Table 7.4-1

Chemistry ECG 7.4 DIABLO CANYON - UNITS 1 & 2 Rev. 3 ECG_7!4u3r03.DOC 1019.1129 Page 3 of 4 Table 7.4-1 Reactor Coolant System CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS PARAMETER STEADY-STATE LIMIT TRANSIENT LIMIT SAMPLE AND ANALYSIS FREQUENCY At least once per:

Dissolved Oxygen*

0.10 ppm 1.00 ppm 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Chloride 0.15 ppm 1.50 ppm 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Fluoride 0.15 ppm 1.50 ppm 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

• Not applicable with Tavg 250°F.

Chemistry ECG 7.4 DIABLO CANYON - UNITS 1 & 2 Rev. 3 ECG_7!4u3r03.DOC 1019.1129 Page 4 of 4 BASES The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

When the RCS is drained to a reduced inventory configuration with all fuel removed and all loops drained, RCS sampling is not possible. This is because level is so low that sample flow through RCS loop sampling taps or instrument taps is not possible. Therefore, compliance with this ECG is achieved by following the action statement requirements and performing an analysis to show the RCS did not exceed the ECG requirements by evaluating the Reactor Coolant before and after the "loops drained" operating condition. The ECG surveillance requirements shall be followed once reactor vessel level is returned to a level that allows RCS sampling.

REFERENCES

1. License Amendment Request 94-07, "Relocation of Selected 3/4.4 Technical Specifications in Accordance with NRC Final Policy Statement and NUREG-1431"

2. License Amendment 98 (Unit 1) and 97 (Unit 2), dated March 9, 1995

3. PSRC Interpretation 90-04

Parent questions for Modified Questions Q22, 23, 39, and 74

DCPP L161 Exam Rev 1 Examination Outline Cross-Reference Level RO 015 K2.01 - Knowledge of bus power supplies to NIS channels, components, and interconnections.

Tier #

2 Group #

2 K/A #

015 K2.01 Rating 3.3 Question 30 Unit 1 is operating at 6% power Panel PY-13 is de-energized.

Which of the following states the excore instrumentation channels that have been lost and if an automatic reactor trip occurs as a result of the loss of the PY?

A. A power range channel only, reactor trip WILL NOT occur B. An intermediate range and a power range channel; reactor trip WILL NOT occur C. A power range channel only, reactor trip WILL occur D. An intermediate range and a power range channel; reactor trip WILL occur Proposed Answer:

A. A power range channel only; reactor trip WILL NOT occur Explanation:

A. Correct. PY-13 powers the 3rd column of the NIS cabinet in the control room, with PR N43. Reactor power is below P10, and 25% (IR and PR trip setpoint), however, the PR logic is 2 of 4 and no IR has been lost.

B. Incorrect because the IR channels are powered from PY-11 and 12. Plausible because it could be the IR channels are not all on the same channels as the SR channels. If an IR lost power, the reactor would trip.

C. Incorrect because there is no reactor trip. Plausible as only a PR is lost.

D. Incorrect because there is no reactor trip and only a PR is lost. Plausible as the power supplies must be known and if IR had lost power, this would be the correct answer.

Technical

References:

LB4 References to be provided to applicants during exam: None Learning Objective: State the power supplies to Excore Nuclear Instrumentation System components. (40940)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank # S-32224 X

New Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

Rev 2 Examination Outline Cross-Reference Level RO APE 058 AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation Tier #

1 Group #

1 K/A #

APE 058 AK1.01 Rating 2.8 Question 43 Unit 1 is at 100% power.

Initial Indications Current Indications The current indications are consistent with:

A. the loss of DC Bus 1-1.

B. a loss of 480 VAC bus H.

Rev 2 C. placing of the battery on equalizing charge.

D. opening the battery charger output breaker.

Proposed Answer: D. opening the battery charger output breaker.

Explanation:

A. Incorrect. While DC amps of the charger are at 0, negative amps on the battery indicate the battery is carrying the bus, not a loss of the bus.

B. Incorrect. The normal supply to Battery Charger 1-1 is bus F. loss of Bus H would not impact DC bus 1-1. Plausible as EDG 1-1 supplies bus H.

C. Incorrect. Equalizing charge would have higher battery voltage and there would still be amps indicated on the charger.

D. Correct. Opening the charger output breaker would result in the battery supplying the bus.

Indications would be 0 amps from the charger and negative amps from the battery, as it is now carrying load..

Technical

References:

OIM J-1-1 and J-1-2 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the DC Power System.

(5193)

Question Source:

Bank #49 DCPP L091 07/2011 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam #49 DCPP NRC 07/2011 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L162 Exam Rev 0 Examination Outline Cross-Reference Level RO 013 G2.4.9 - ESFAS: Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Tier #

2 Group #

1 K/A #

013 G2.4.9 Rating 3.8 Question 12 GIVEN:

Unit 1 is performing a heatup in accordance with OP L-1, Plant Heatup From Hot Shutdown to Hot Standby Electrical power is aligned to startup RCS temperature is 525°F RCS pressure is 1900 psig PK08 indicates as shown below: (red outlined annunciators are lit)

A steam break, outside containment, upstream of the MSIV occurs on the 11 Steam Generator.

1) SI will _____________ actuate.

2) Once SI is automatically or manually actuated, the running AFW pumps will A.

1) automatically

2) remain running without interruption B.

1) NOT automatically

2) remain running without interruption C.

1) automatically

2) stop and restart when sequenced on to their respective Emergency Diesel Generator D.

1) NOT automatically

2) stop and restart when sequenced on to their respective Emergency Diesel Generator L162 Answer Key

DCPP L162 Exam Rev 0 Proposed Answer: B. 1) NOT automatically

2) remain running without interruption Explanation:

Tests what the operator would see in the Control Room (operational validity) when RCS pressure is less than 1915 psig and an accident occurs (steam break). Additionally, the response of the AFW pumps when the SI occurs (do they or dont they stop and sequence on when SI occurs).

A. Incorrect.SI on low RCS pressure is blocked below P-11 (1915 psig) (PK08-06 LIT).

Plausible if P-11 is thought to only affect RCS pressure SI. Also, SI would automatically actuate if the break was inside containment and pressure rises to greater than 3 psig.

B. Correct. SI will not actuate automatically, however, the AFW are not stripped and then restarted. They would be if there was also a transfer to diesel, but if startup is available, no load stripping occurs.

C. Incorrect. This would be the response at power and the AFW pumps were not running and startup was not available.

D. Incorrect, P-11 blocks Low Pressurizer pressure AND low steamline pressure SI (PK08-16 and 17 LIT). The AFW are not stripped. Plausible because for most SI actuations which cause a reactor trip and bus transfer from 500 kV to either startup or diesel, the AFW pumps would be sequenced. In this case, already on startup, no bus stripping occurs and the pumps will remain running.

Technical

References:

OIM B-6-2, B-6-5 and J-6-1 References to be provided to applicants during exam: None Learning Objective: 41313 - Analyze automatic features and interlocks associated with the Eagle-21/SSPS Question Source:

Bank #

(note changes; attach parent)

Modified Bank #11 DCPP NRC L091C 03/12 X

New Past NRC Exam Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.3 L162 Answer Key

DCPP L141 Exam Rev 2 Examination Outline Cross-Reference Level RO G2.4.3 - Ability to identify post accident instrumentation.

Tier #

3 Group #

4 K/A #

2.4.3 Rating 3.7 Question 75 Which of the following is monitored on PAM1?

A. Pressurizer Level B. Auxiliary Feedwater Flow C. Steamline Pressure D. Reactor Cavity Sump Level Proposed Answer:

D. Reactor Cavity Sump Level Explanation:

A. Incorrect. Not monitored on PAM1 B. Incorrect. Not monitored on PAMS 1 C. Incorrect. Not on PAMS1 (vertical board)

D. Correct. (WR) Containment Sump level is monitored by PAMS 1 Technical

References:

LB-10, Post-Accident Monitoring System References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the PAMS.

- PAMS Panel 1 Question Source:

Bank # 74 NRC L081 (1/2010)

X (note changes; attach parent)

Modified Bank #

New Question History:

Last NRC Exam No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7