ML21075A225

DC-2021-01 Draft Written Exam

ML21075A225
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: ML21075A225 (181)

Text

DCPP L191 Exam Rev 0 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

2) automatic and manual

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 #

(note changes; attach parent)

Modified Bank #

DCPP L191 Exam Rev 0 New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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) Pressurizer level indications on VB2, LI-459, 460 and 461, ___________ identified 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 Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 Learning Objective: Explain the basis of significant steps associated with OP "L" procedures.

DCPP L191 Exam Rev 0 (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 X

10CFR Part 55 Content:

55.41.10 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5 Difficulty: 2.7

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.8 Difficulty: 2.1

DCPP L191 Exam Rev 0 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 Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 Question Cognitive Level:

Memory/Fundamental

DCPP L191 Exam Rev 0 Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.11 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 to ensure 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. maintain SG water level in the narrow range.

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

D. ensure a sufficient heat sink to initiate natural circulation.

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. 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.

C. 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..

D. Incorrect. Raising flow does enhance natural circulation, however, natural circulation is not addressed in E-0 and there is not a flow rate associated with adequate natural circulation.

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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 The plant is operating at 100% power. PRT pressure is normal.

A Pressurizer PORV begins to leak by. RCS pressure has lowered to 2100 psig and PRT pressure is 15 psig and rising.

PORV tailpipe temperature instrument, TI-463, indication will __________________________.

A. rise until offscale high B. rise to a temperature based on PRT pressure saturation temperature C. lower to a temperature based on RCS pressure saturation temperature D. lower to a temperature based difference between RCS and PRT pressure saturation temperature Proposed Answer: B. rise to a temperature based on PRT pressure saturation temperature.

Explanation:

A. Incorrect. Temperature indication follows PRT pressure. Plausible that its known top of scale is 400°F and the temperature is based on RCS pressure.

B. Correct. As long as the conditions remain inside the saturation dome, as PRT pressure rose, Tsat for the PRT rose as well, and the temperature of the tailpipe will reflect the rising tailpipe temperature.

C. Incorrect. Plausible if its thought the temperature response is based on RCS pressure, which has lowered.

D. Incorrect. Plausible if its thought the temperature response is based on RCS to PRT pressure difference, which has lowered.

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 #40 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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.5

DCPP L191 Exam Rev 0 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.

LR-60 and LR-61, Containment Structure Sump Level indication on PAM 1 will fail

_____1)________ due to _______________2)__________________.

A. 1) low

2) isolation of instrument air to containment B. 1) low

2) loss of power to the transmitters C. 1) high

2) isolation of instrument air to containment D. 1) high

2) loss of power to the transmitters Proposed Answer: A. 1) low 2) isolation of instrument air to containment Explanation:

Level transmitter (LT-60) is a bubbler instrument supplied by instrument air. On containment isolation, e.g. safety injection, instrument air is isolated and indicated sump level fails low. LT-60 also provides structure sump 1-1 level recording on PAM1, which is rendered unreliable following containment isolation.

A. Correct. Instrument Air to containment is isolated by Phase A. When pressure is lost to the bubblers, they will fail low.

B. Incorrect. First part is correct. Plausible because most instruments, such as pressurizer or pressurizer level, fail when power is lost to them. Power is not lost to the level instruments, its is the loss of air that causes the indication to go low.

C. Incorrect. Second part is correct. Loss of air causes level to fail but low not high.

D. Incorrect. Both parts are incorrect.

Technical

References:

LI-1 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 X

Comprehensive/Analysis

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.7 Difficulty: 2.4

DCPP L191 Exam Rev 0 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 Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 RCP number 1 seal supply flow indication (same on all RCPs).

The operator will turn potentiometer for HCV-142, RCP Seal Flow Control valve,

______1)________ because current RCP seal injection is ____2)_____.

A. 1) clockwise

2) low B. 1) clockwise

2) high C. 1) counter clockwise

2) low D. 1) counter clockwise

2) high Proposed Answer: C.

1) counter clockwise 2) low 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, Second part is correct, normal seal injection is 8 to 13 gpm.

B. Incorrect. Plausible if its thought that 6 gpm is higher than normal (because normally only

DCPP L191 Exam Rev 0 about 3 gpm is returned to the VCT) and 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 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. First part correct. Second part incorrect.

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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.1

DCPP L191 Exam Rev 0 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, mid-loop.

The crew enters OP AP SD-0, Loss of, or Inadequate Decay Heat Removal RHR, when RHR flow begins to oscillate. The operator reports that based on pump amps and reports from the field that the RHR pumps are cavitating.

What are the indications that the RHR pumps are cavitating?

____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: A.

1) Large (8 to 10 amps) 2) a Explanation:

A. Correct. 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 correct. Second part incorrect. This is the indication of vortexing (with small pump amp swings)

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

D. Incorrect. Both parts incorrect.

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

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 Unit 1 is at 100% power. Pressurizer backup heater group 12 is energized by its vital 480 VAC bus.

The reactor trips and pressurizer level drops to 14%. The crew is performing EOP E-0, Reactor Trip or Safety Injection.

1) What is the vital 480 VAC bus to Pressurizer backup heater group 12?

2) What is the current status of pressurizer backup heater group 12?

A. 1) Bus F

2) energized B. 1) Bus F

2) de-energized C. 1) Bus G

2) energized D. 1) Bus G

2) de-energized Proposed Answer: C.

1)

Bus G 2) energized Explanation:

Proportional group 1 - 13D Backup group 2 - 13D/1G*

Backup group 3 - 13E13E/1H*

Backup group 4 - 13E 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 as they control the main breaker at 13D/E and not the vital breaker at bus G or H. The 17% trip does not function.

A. Incorrect. The power supply is Bus G. Plausible because it could be thought the power supplies are Bus F and G and therefore, group 12 would be from Bus F (and 13 then from bus G). Second part correct, when on backup vital power, the heater group remains on when pressurizer level is below 17%.

B. Incorrect. Power supply is Bus G. Second part would be correct if SI had actuated.. However, they will trip on vital BU power if there is an SI.

C. Correct. Power supply is correct. The auto turn off when pressurizer level is below 17% is defeated when on the emergency power supply.

D. Incorrect. Power supply is bus G however, the heaters will remain energized.

Technical

References:

LA-4A References to be provided to applicants during exam: None

DCPP L191 Exam Rev 0 Learning Objective: Analyze automatic features and interlocks associated with the Pressurizer, Pressure & Level Control System.

• Backup Heaters 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.6

DCPP L191 Exam Rev 0 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.

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.

Plausilbe 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.

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.Incorrect.

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

Past NRC Exam No

DCPP L191 Exam Rev 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.6 Difficulty: 2.3

DCPP L191 Exam Rev 0 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

• Low Pressurizer pressure and main steamline pressure SI have been blocked

• RCS pressure has been reduced to 1500 psig using a Pressurizer PORV In accordance with step 9 in EOP FR-H.1, 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 RCS subcooling 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. Raising RCS pressure would raise subcooling and its plausible that its thought that subcooling should be minimized as it is in EOPs such as ECA-3.1.

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 X

DCPP L191 Exam Rev 0 Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.3

DCPP L191 Exam Rev 0 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. Temperature limit ensures a challenge to RCS integrity does not occur 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 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.5 Difficulty: 3.5

DCPP L191 Exam Rev 0 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 #

(note changes; attach parent)

Modified Bank #30 NRC L161 10/2016 X

DCPP L191 Exam Rev 0 New Past NRC Exam DCPP 10/2016 (modified)

Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.6

DCPP L191 Exam Rev 0 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 Conditions Current Conditions Current conditions are consistent with:

A. opening the input breaker to DC bus 1-1 B. a loss of 480 VAC bus H Battery Charger Battery Charger

DCPP L191 Exam Rev 0 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 the amps to the battery.

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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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.

A loss of ASW pumps has occurred. The crew entered OP AP-10, Loss of Auxiliary Saltwater and then transitioned to OP AP-11, Malfunction of Component Cooling Water.

In accordance with OP AP-11, what condition(s) require the crew to trip the reactor and stop all RCPs?

1. ASW cannot be restored

2. CCW temperature exceeds 120°F

3. ASW temperature exceeds the Technical Specification limit A. 1 only B. 1 and 2 C. 3 only D. 2 and 3 Proposed Answer: B.

1 and 2 Explanation:

A. Incorrect. Inability to restore ASW requires tripping the reactor but also high CCW temperature does as well.

B. Correct. IAW OP AP-11, both conditions require a reactor trip (and stopping RCPs). OP AP-10 does not have trip criteria - it sends you to OP AP-11.

C. Incorrect. LCO 3.7.8 has a limit of 64F however high ASW temperature is not a criteria for tripping the reactor. Plausible that high temperature for a system that supplies CCW could require a reactor trip.

D. Incorrect. CCW temperature requires a reactor trip. Plausible that both systems would require a reactor trip.

Technical

References:

OP AP-10, AP-11 References to be provided to applicants during exam: None Learning Objective: Given initial conditions and assumptions, determine if a reactor trip or safety injection actuation is required. (7927)

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 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 for 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. lower to a minimum flow limited by a mechanical stop.

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 it may be thought that both Heat Exchangers supply all loops and HCV-637 fails closed, therefore, flow would be reduced. If thought there is a mechanical stop to ensure some minimum flow (like there is for opening), then this is a logical answer.

D. Correct. HCV-637 will fail open raising flow. Backup air is supplied thru the cut-in switch and then controlled using the open/close switch on VB3.

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 X

10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 0 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 Learning Objective: Explain basis of emergency procedure steps for ECA-1.2. (42461, 7920H).

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.8

DCPP L191 Exam Rev 0 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)

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 X

10CFR Part 55 Content:

55.41.10 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 References to be provided to applicants during exam: None

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 Comprehensive/Analysis X

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.5 Difficulty: 2.7

DCPP L191 Exam Rev 0 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 A loss of all offsite power occurs on Unit 2.

Emergency Diesel Generator 2-1 trips.

How many Boric Acid Transfer pumps and Charging pumps are currently available?

Boric Acid Transfer Charging Pumps Pumps Available Available A.

1 1

B.

2 2

C.

2 1

D.

1 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. EDG powers bus G for Unit 2. Two charging pumps are on Bus G (one on Bus F) and Boric Acid Transfer pumps are powered from bus F and. bus G. Therefore one Boric Acid and two charging pumps are de-energized and one of each remains 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 EDG 2-1 powered Bus F. Plausible the 2-1 EDG powers F (and 2-2 and 2-3 power G and H respectively).

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 #

New X

Past NRC Exam No Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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:

• Unit 1 is in MODE 4

• RCS temperature is 275°F

• RCS boron concentration is 1800 ppm

• A cooldown to MODE 5 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 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.10 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 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)

Question Source:

Bank #

DCPP L191 Exam Rev 0 (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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 3.5

DCPP L191 Exam Rev 0 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. Incorrect. 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 0 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 L191 Exam Rev 0 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

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

will automatically supply the pressurization system.

A. 1) maintain chemistry within limits

2) nitrogen bottles B. 1) maintain chemistry within limits

2) instrument air C. 1) prevent water hammer during a postulated accident scenario

2) nitrogen bottles D. 1) prevent water hammer during a postulated accident scenario

2) instrument air Proposed Answer: C.

1) prevent water hammer during a postulated accident scenario. 2) nitrogen bottles Explanation:

A. Incorrect. The backup is bottles of nitrogen. First part is incorrect. Plausible because higher pressure in the system is a method of keeping gas in solution.

B. Incorrect. Both parts incorrect. Instrument air is available but is only used if nitrogen is not available and must be valved in. Plausible because air provides primary and backup motive force to many plant components.

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.

• In this scenario the integrity of the CCW system could be challenged. To preclude flashing and subsequent water hammer, the CCW surge tank is pressurized to increase the static head on the system. Bottled nitrogen provides the normal backup and automatically supplies the system pressure drops.

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

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 0 New X

Past NRC Exam Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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

• RCS temperature is 270°F

• RCS pressure is 390 psig Which of following would cause a Pressurizer PORV to open?

1. RCS wide range pressure rising to 450 psig

2. RCS cold leg temperature lowering to 265°F

3. RCS cold leg temperature rising to 285°F A. 1 only B. 2 only C. 3 only D. 1 and 3 Proposed Answer: A.

1 only Explanation:

A. Correct. Only rising pressure causes the PORVs to open. If pressure exceeds 435 psig, the two safety related PORVs will open.

B. Incorrect. Lowering temperature arms LTOP but does not cause the PORV to open.

C. Incorrect. Rising temperature will disable LTOP but not cause the PORVs to open.

D. Incorrect. Only pressure rising will cause the Class 1 PORVs to 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 #56 L121 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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

(note changes; attach parent)

Modified Bank #

DCPP L191 Exam Rev 0 New Past NRC Exam DCPP 10/2016 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.7

DCPP L191 Exam Rev 0 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 the channel 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 Past NRC Exam DCPP NRC 10/2016 Yes

DCPP L191 Exam Rev 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 0 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.0

DCPP L191 Exam Rev 0 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-11, Containment Air Particulate

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-11 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 The crew is performing the actions of EOP E-1.3, Transfer to Cold Leg Recirculation.

Containment Spray is in service.

When EOP E-1.3 is complete, ___1)____ train(s) of RHR will be aligned

__________________2)________________________.

A. 1) one

2) to the suction of the containment spray pumps B. 1) one

2) downstream of 9001A/B, Containment Spray Pump Discharge valves C. 1) both

2) to the suction of the containment spray pumps D. 1) both

2) downstream of 9001A/B, Containment Spray Pump Discharge valves Proposed Answer: B. 1) one 2) downstream of 9001A/B, Containment Spray Pump Discharge valves Explanation:

Accident conditions required to meet KA.

A. Incorrect. First part is correct. Second part incorrect. The RHR train is aligned to supply the containment spray header, downstream of the spray pump discharge valves, 9001 A and B, which are closed and 9003A and B are opened.

B. Correct. One train of RHR is aligned to the spray header, downstream of the spray pump discharge valves, 9001 A and B.

C. Incorrect. Both parts incorrect.

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

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 #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam DCPP NRC No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 Comprehensive/Analysis X

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.7 Difficulty: 3.2

DCPP L191 Exam Rev 0 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)

Question Source:

Bank #34 L162 X

DCPP L191 Exam Rev 0 (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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 100% power.

The operator notes the following indications over the last 7 minutes:

1) At the Digital Feedwater Control Station, the operator will observe that

2) Based on the rate of level rise from the time the event began, P-14 will actuate in approximately ________ minutes.

A. 1) a main feedwater reg valve has failed open

2) one to two B. 1) a main feedwater reg valve has failed open

2) four to five C. 1) speed on one or both main feedwater pumps has risen

2) one to two D. 1) speed on one or both main feedwater pumps has risen

2) four to five

DCPP L191 Exam Rev 0 Proposed Answer: B. 1) a main feedwater reg valve has failed open 2) four to five Explanation:

A. Incorrect. First part is correct. Level on only one steam generator has risen. Second part incorrect. Plausible that the time was incorrectly used to calculate the rise or the setpoint is not know.

B. Correct. A main feedwater reg valve (on steam generator 1-1) has failed open. Level has risen for 7 minutes from 65% (normal operating level) to 80%. 15%/7 = ~2%/minute.

Setpoint for P-14 is 90%. Therefore, in approximately five minutes, steam generator level will reach the P-14 setpoint.

C. Incorrect. Both parts incorrect. If speed had risen on a MFP(s), level would be higher on all steam generators. Time to P-14 is incorrect.

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

Technical

References:

LTAA 6 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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 Unit 1 is in MODE 3. All steam generator narrow range levels are 65%.

The 1-1 AFW pump starts spuriously.

AFW Flow meters on VB3 will indicate:

A. no flow because design discharge pressure of the 1-1 AFW pump is less than Main Feedwater pump discharge pressure.

B. no flow because the level control valves will immediately close because the steam generator levels are on program.

C. maximum flow, however, flow will slowly lower due to the modulating action of the 1-1 AFW pump level control valves.

D. maximum flow until operator action is taken to close the 1-1 AFW pump level control valves or the pump is shutdown.

Proposed Answer:

D. maximum flow until operator action is taken to close the 1-1 AFW pump level control valves or the pump is shutdown Explanation:

A. Incorrect. Rated discharge pressure is higher than main feedwater pressure. The pump will feed forward.

B. Incorrect. The 1-1 AFW pump LCVs do not modulate, they must be operated by the operator.

Plausible because the LCVs for the motor driven AFW pumps would operate this way.

C. Incorrect. This would be correct for the motor driven AFW pumps..

D. Correct. The 1-1 LCVs do not modulate, they are controlled by the operator. Because the discharge pressure of the AFW pump is higher than MFW pressure, it will feed the steam generators and continue to do so until action is taken by the crew.

Technical

References:

LD-1, DCPP simulator 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 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.10

DCPP L191 Exam Rev 0 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 at 100% power.

PK16-21, 4KV Bus H Aux or SU Breakers, alarms due to input 294, "4KV Bus-H Aux Bkr 52-HH-13 OC Trip".

What is the expected Vital 4 kV Bus H indication on VB4?

NOTE: Outlined lights are LIT.

A.

B.

C.

D.

DCPP L191 Exam Rev 0 Proposed Answer: B. Picture 2 Explanation:

A. Incorrect. This would be the indication for a normal breaker opening and transfer to S/U.

B. Correct. The blue light is lit for overcurrent trips and transfer to SU or diesel is blocked. All breakers would be open.

C. Incorrect. The blue light would be lit however, the SU and diesel breakers would be open.

Transfer is blocked by overcurrent trip.

D. Incorrect. All the breakers would be open, but the blue light would be lit.

Technical

References:

OIM J-5-1d and 1e, AR PK16-21 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Electrical Power Transfer System. (4295)

Question Source:

Bank #19 L171 01/2019 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam Yes Question History:

Last Two NRC Exams DCPP 01/2019 Yes Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.7 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.8 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 A ground causes the DC breaker supplying control power to the 11 RCP, 13 RCP and 12 CWP to open.

All indicating lights for 11 RCP, 13 RCP and 12 CWP are out.

In accordance with AR PK05-01, 11 RCP, or AR PK05-03, 13 RCP, what action(s) should be taken by the crew in the Control Room?

1) Dispatch an operator to check a possible ground and the status of the DC control power breaker to the affected components on bus D

2) Dispatch an operator to check a possible ground and the status of the DC control power breaker to the affected components on bus E

3) From the Control Room trip the reactor, 11 RCP, 13 RCP and 12 CWP A. 1 only B. 2 only C. 1 and 3 D. 2 and 3 Proposed Answer: B. 2 only Explanation:

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 loads are on Bus E not bus D.

B. Correct. A condition, such as a ground has caused a loss of control power, which causes the lights to go out. The three loads listed are on 12 kV bus E. 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).

C. Incorrect. Loads are on Bus D. There is no reason to trip the loads and due to the loss of control power and they cannot be tripped from the control room. Plausible because there is no indication and it could be thought the components need to be tripped. If the RCPs were to be tripped, the reactor is tripped first.

D. Incorrect. There is no reason to trip the loads and due to the loss of control power and they cannot be tripped from the control room. Plausible because there is no indication and it could

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.10 Difficulty: 2.2

DCPP L191 Exam Rev 0 Examination Outline Cross-Reference Level RO 064 K4.10 Knowledge of ED/G system design feature(s) and/or interlock(s) which provide for the following: Automatic load sequencer: Blackout Tier #

2 Group #

1 K/A #

064 K4.10 Rating 3.5 Question 49 The crew has completed the actions of EOP E-1.3, Transfer to Cold Leg Recirculation.

A loss of startup occurs and the emergency diesel generators start and begin to energize their respective vital 4 kV bus.

What ECCS pumps, will be automatically started?

1. CCP

2. SI

3. RHR A. 1 only B. 2 only C. 1 and 3 D. 2 and 3 Proposed Answer: A. 1 only Explanation:

A. Correct. Before the loss of power, all the pumps will be running. With SI reset, the RHR and SI pumps will not restart. The charging pumps get a start signal with or without an SI signal.

B. Incorrect. While the RHR pumps supply the suction to the charging and SI pumps, they do not receive a start signal without an SI signal present.

C. Incorrect. The charging pumps start, the SI pumps do not.

D. Incorrect. Neither of these pumps start.

Technical

References:

OIM J-6-1 References to be provided to applicants during exam: None Learning Objective: Analyze automatic features and interlocks associated with the Electrical Power Transfer System. (4295)

Question Source:

Bank #22 L121 08/2014 X

(note changes; attach parent)

Modified Bank #

New Past NRC Exam 08/2014 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 L191 Exam Rev 0 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 #22 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.8 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.11 Difficulty: 3.5

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 and Unit 2 are at 100% power.

ASW and CCW are in their normal full power alignment.

ASW to Unit 2 is lost. In accordance with OP AP-10, Loss of Auxiliary Salt Water, the crew is aligning ASW pump 1-2 to supply Unit 2.

In accordance with OP AP-10, the crew will have to open FCV-601 ______________________.

NOTE:

• 1-FCV-496, ASW pump 1-1 discharge cross tie

• 1-FCV-495, ASW pump 1-2 discharge cross tie

• 0-FCV-601, Unit 1 and 2 ASW Cross-tie A. only B. and open 1-FCV-495 C. and close 1-FCV-495 D. and close 1-FCV-496 Proposed Answer: D.

and close 1-FCV-496 Explanation:

A. Incorrect. 0-FCV-601 must be opened but also the cross tie valve from the train NOT supplying Unit 2 must be closed. If its not known what the lineup is, then its plausible that only opening 601 is the only valve manipulation necessary. This would cross tie both units but would be cross tieing both trains of both units.

B. Incorrect. This is plausible if its thought the cross ties are normally closed (they are normally open).

C. Incorrect. Plausible if its thought the cross tie valve is only for Unit 1 (closed to split Unit 1 trains) and closing it realigns the flow path to Unit 2.

D. Correct. The cross tie valves are both normally open. To supply the other unit the cross tie valve for the pump NOT supplying ASW to the opposite unit must be closed.

Technical

References:

OP AP-10, LE-5 References to be provided to applicants during exam: None Learning Objective: Discuss abnormal conditions associated with the ASW System. (5360)

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

Past NRC Exam No

DCPP L191 Exam Rev 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.10 Difficulty: 2.6

DCPP L191 Exam Rev 0 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 Comprehensive/Analysis X

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.5 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 Unit 1 is in MODE 6.

During core offload, a damaged fuel assembly causes rising radiation levels and containment ventilation isolation. The crew enters OP AP-21, Irradiated Fuel Damage.

1) The containment evacuation alarm will be activated ____________________.

2) The containment evacuation alarm is ________________________________________.

A. 1) by the operator in the Control Room only

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

2) an electronic warbler that sounds a series of falling tones C. 1) both automatically by the CVI and by the operator in the Control Room

2) a constant, high-pitched tone D. 1) both automatically by the CVI and by the operator in the Control Room

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

1) by the operator in 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. OP AP-21 has the operator activate the containment evacuation alarm. Second part is correct, the signal is as described in the answer.

C. Incorrect. Both parts incorrect. Plausible if they believe the alarm actuates like the Fuel Handling Building rad alarm and operates automatically on high radiation or can be manually actuated.

D. Incorrect. First part incorrect, not an auto action of CVI. 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

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.11 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 GIVEN:

• Safety Injection has actuated

• Containment pressure has risen to the values shown below:

PK02-01, CONTMT ISOLATION PHASE A/B (RED), is in alarm.

What valve(s) should be closed as a direct result of the actuation signal(s) that caused PK02-01 to alarm?

1. CVCS-8112, RCPs #1 Seal Water Return

2. FCV-363, RCP Lube Oil Cooler Return

3. CVCS-8166, Excess Letdown Isolation A. 1 only B. 2 only C. 1 and 3 D. 2 and 3 Proposed Answer: A.

1 only Explanation:

Only Phase A has actuated. Phase B setpoint is 22 psig.

A. Correct. Phase A Containment Isolation closes CVCS-8100 and CVCS-8112, RCPs #1 Seal Outlet valve.

DCPP L191 Exam Rev 0 B. Incorrect. FCV-363 closes on Phase B. Plausible if its thought its closed by Phase A.

C. Incorrect. CVCS-8166 are isolation valves but do not get an automatic close signal from either Phase A or Phase B.

D. Incorrect. FCV-363 is closed by Phase B and CVCS-8166 does not get an automatic close signal.

Technical

References:

LB-1A, LF-2 References to be provided to applicants during exam: None Learning Objective: Describe controls, indications, and alarms associated with the Radiation Monitoring System. (37875)

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 X

10CFR Part 55 Content:

55.41.7 Difficulty: 3.2

DCPP L191 Exam Rev 0 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 DCPP 06/2008 No Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.2 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 in the _________________________________________.

A. 1) locally

2) Cable Spreading Room B. 1) locally

2) Dedicated Shutdown Panel C. 1) from VB2

2) Cable Spreading Room D. 1) from VB2

2) Dedicated Shutdown Panel Proposed Answer: C.

1) from VB2 2) 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.7

DCPP L191 Exam Rev 0 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 A main steam line break has occurred inside Unit 1 Containment.

Elevated containment temperature causes indicated pressurizer level for LT-459 to be __1)__

than actual level because the density of the fluid in the reference leg 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:

High temperature can affect instrumentation in Containment is by heating the reference leg of a level instrument. This would cause the density of the water in the reference leg to decrease, lowering the sensed D/P. This would make the indicated level appear to be higher than actual level.

A. Incorrect. Indicated level will be higher than actual level. Density lowers as temperature rises.

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

C. Incorrect. First part is correct. Density lowers.

D. Correct. Heating the reference leg causes density to decrease, causing dp to lower which causes indicated level to increase.

Technical

References:

LMCDIAR References to be provided to applicants during exam: None Learning Objective: Explain the effects adverse Containment conditions have on instrumentation. (7952)

Question Source:

Bank #DCPP bank A-0018 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 X

DCPP L191 Exam Rev 0 10CFR Part 55 Content:

55.41.5 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.10 Difficulty: 3.2

DCPP L191 Exam Rev 0 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

1) On a loss of power to RM-58, Fuel Handling Building monitor, the Fuel Handling Building Evacuation alarm ___________ actuate.

2) The Fuel Handling Building Evacuation alarm will actuate if the radiation reading on RM-58 rises, as a minimum, to the ________________ alarm setpoint.

A. 1) will

2) Alert (amber light)

B. 1) will

2) High (red light)

C. 1) will NOT

2) Alert (amber light)

D. 1) will NOT

2) High (red light)

Proposed Answer: B. 1) will 2)

High (red light)

Explanation:

A. Incorrect. First part is correct, loss of power causes the RM to go into alarm and one of the actions that occurs is the FHB Evacuation (local) alarm. Second part is incorrect. The alarm occurs at high, not alert level.

B. Correct. Loss of power causes the alarm. Additionally, High alarm causes the auto actions, which includes the FHB Evacuation alarm.

C. Incorrect. Both parts are incorrect. Loss of power or high alarm causes the auto actions to occur.

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

Technical

References:

LG-4A References to be provided to applicants during exam: None Learning Objective: State the purpose of Radiation Monitoring System components.

RM-58 and RM-59 Fuel Handling Building Monitors 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 X

10CFR Part 55 Content:

55.41.11 Difficulty: 2.5

DCPP L191 Exam Rev 0 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, 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 RWST to either charging pump and, therefore, no charging injection. The minimum flow

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.12

DCPP L191 Exam Rev 0 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.5 Difficulty: 2.5

DCPP L191 Exam Rev 0 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

Past NRC Exam No

DCPP L191 Exam Rev 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.41.5 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 Question History:

Last Two NRC Exams No

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.4

DCPP L191 Exam Rev 0 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, regarding maintaining a license, the NRC shall be notified of a change to a Reactor operators license if the operator:

A. is audited by the IRS.

B. receives a traffic citation for speeding.

C. is placed on high blood pressure medication.

D. begins license class to obtain an SRO license.

Proposed Answer: C. is placed on high pressure medication.

Explanation:

A. Incorrect. An audit does not require notification to the NRC.

B. Incorrect. A felony could impact a license, a speeding citation does not.

C. 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.

D. Incorrect. Upgrade to SRO will result in a new license but beginning class is not a condition to be reported 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.12 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.2 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.4

DCPP L191 Exam Rev 0 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, how often is the Control Operator required to review the open main annunciator problem evaluation sheets/defeat logs?

A. At the beginning of each shift B. Daily C. Weekly D. Monthly Proposed Answer: A.

At the beginning of each shift Explanation:

A. Correct. The review is to be done at the beginning of each shift.

B. Incorrect. Some logs are taken (and then reviewed) on a daily basis.

C. Incorrect. An audit of the log is performed on a weekly basis.

D. Incorrect. Many surveillances are on a monthly basis and because the number of alarms in defeat is typically a small number, the review could be thought to be be on a longer cycle.

Technical

References:

OP1.DC24 step 5.5.1.a.

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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.1

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.41.12 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 in MODE 6. Core offload is in progress in accordance with OP B-8DS1, Core Unloading.

While an assembly is being moved into the Spent Fuel Pool, there are reports of high radiation in the Fuel Handling Building.

In accordance with OP B-8DS1, what is the responsibility of the Reactor Operator in the Control Room?

1. Ensure FHBVS in iodine removal mode using an available exhaust fan

2. Ensure the fuel assembly has been placed in a safe location

3. Make a PA announcement to evacuate the FHB A. 1 and 2 B. 1 and 3 C. 2 only D. 3 only Proposed Answer: B. 1 and 3 Explanation:

Question deals with the safety principles of limiting exposure to personnel and fuel handling responsibilities.

A. Incorrect. 1 is correct. 2 is incorrect, this may seem prudent but not a responsibility of the operator in the control room, would be done by the refueling SRO.

B. Correct. Per OP B-8DS1, in the event of a FHB Evacuation alarm, the Control Room shall:

• place FHB ventilation in iodine removal mode using an available exhaust fan

• Make a PA announcement to evacuate the FHB

• and refer to AR PK11-10 and TS 3.3.8 for more information on this alarm.

C. Incorrect. Per OP B8DS1, not the responsibility of the control room operator.

D. Incorrect. This is an action for the control room to take but ensuring FHB ventilation is in Iodine Removal is another responsibility.

Technical

References:

OP B-8DS1 References to be provided to applicants during exam: None Learning Objective:

Question Source:

Bank #

(note changes; attach parent)

Modified Bank #

New X

DCPP L191 Exam Rev 0 Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.12 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 PAM1?

A. Pressurizer level B. Steamline pressure C. Auxiliary Feedwater Flow D. Reactor Cavity Sump Level Proposed Answer: D.

Reactor Cavity Sump Level Explanation:

A. Incorrect. PAM instrument on VB2, not monitored on PAM1 B. Incorrect. PAM on VB3, not monitored on PAMS 1 C. Incorrect. PAM on VB3, not on PAMS1 D. Correct. (WR) Cavity 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 #75 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.

Difficulty: 2.0

DCPP L191 Exam Rev 0 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 only C. or EOP FR-S.1, Response to Nuclear Power Generation/ATWS only D. EOP ECA-0.0, Loss of All Vital Power 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:

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 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.

D. Incorrect. EOP ECA-0.0 is correct. EOP FR-S.1 is not, it is entered from EOP E-0 or when monitoring the CSF status trees and the status tree is Red or Magenta.

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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.41.10 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 According to the bases of Technical Specification 3.7.7, Component Cooling Water, what is the minimum ASW and CCW pump(s) necessary to support a Design Bases Accident?

A. one ASW pump and one CCW pump B. one ASW pump and two CCW pumps C. two ASW pumps and one CCW pump D. two ASW pumps and two CCW pumps Proposed Answer:

B.

one ASW pump and two CCW pumps Explanation:

A. Incorrect. Per Tech Spec 3.7.7, The CCW system is designed to provide sufficient heat removal for normal and post accident ESF heat loads without overheating. The CCW system and ASW system are essentially considered a single heat removal system for the purpose of assessing the ability to sustain either a single active or passive failure and still perform design basis heat removal. Only one ASW pump and one CCW heat exchanger is required, as assumed in the safety analysis, to provide sufficient heat removal from containment to mitigate a DBA. However, to ensure maximum heat removal capability, operators are instructed to place the second CCW heat exchanger in service early in the emergency operating procedures. (However, for CCW, 2 pumps are required for a vital loop.)

B. Correct. The Tech Spec states: In the event of a DBA, one vital CCW loop is required to provide the minimum heat removal capability assumed in the safety analysis for the systems to which it supplies cooling water. A vital CCW loop is considered OPERABLE when:

a. Two CCW pumps, one CCW heat exchanger, one vital CCW header and the surge tank are OPERABLE; and

b. The associated piping, valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

C. Incorrect. Two CCW pumps required.

D. Incorrect. This is more than the minimum required.

Technical

References:

B3.7.7 References to be provided to applicants during exam: None Learning Objective: 9694G - Apply TS 3.7 Technical Specification bases Question Source:

Bank #77 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

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.5 Difficulty: 2.0

DCPP L191 Exam Rev 0 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 0 New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.5 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.43.5 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 0 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 X

10CFR Part 55 Content:

55.43.5 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 trip of the Main Unit Transformer occurs, 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 0 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 X

10CFR Part 55 Content:

55.43.2 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 0 Past NRC Exam DCPP 01/2018 Yes Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.5 Difficulty: 2.5

DCPP L191 Exam Rev 0 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.

What action should be taken by the Shift Foreman concerning the steam generator tube leak?

At step 4 of EOP E-0, Reactor Trip or Safety Injection, the Shift Foreman should:

A. direct the operator to initiate Safety Injection and then transition to EOP E-3, Steam Generator Tube Rupture from EOP E-0.

B. transition to OP AP-3 to complete the recovery. EOP E-0.1, Reactor Trip Response does not need to be completed.

C. 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.

D. transition to EOP E-0.1, Reactor Trip Response only. OP AP-3 does not need to be completed.

Proposed Answer: C.

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. 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.

B. Incorrect. OP AP-3 has steps to cool down the unit following the reactor trip. EOP E-0.1 does need to be completed.

C. Correct. The AP must be completed when the EOPs are complete. However, OP AP-3 states it may be done is parallel if resources allow.

D. Incorrect. OP AP-3 directs that it must be completed after the EOPs unless a transition to EOP E-3 is made, which is not the case here.

Technical

References:

OP AP-3 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.

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.43.5 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 GIVEN:

• At 1100, PK10-10, Fire Detected, alarms on Unit 1

• At 1102, the fire is reported to be in a Charging Pump room

• At 1105, the Shift Manager begins reviewing EAL charts for possible EP classification

• At 1107, report from the fire brigade reports that the fire will take at least 30 minutes to extinguish The Shift Manager should __________________________________________.

A. wait 8 minutes and then make the declaration at 1115 B. wait 15 minutes and then make the declaration at 1122 C. not wait any longer and promptly declare based on the determination made at 1105 D. not wait any longer and promptly declare upon getting the report at 1107 Proposed Answer: D. not wait any longer and promptly declare upon getting the report at 1107 Explanation:

A. Incorrect. 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. This would be correct if no word was received or for EALs w/o a time requirement.

B. Incorrect. This is 15 minutes from the report at 1107.

C. Incorrect. This is based on the time of determining what EAL is applicable.

D. Correct. The EAL declaration should be made once its determined the time will be exceeded.

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 #

DCPP L191 Exam Rev 0 New X

Past NRC Exam No Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.1 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 operator 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.X 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 0 Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.2 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 0 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 X

10CFR Part 55 Content:

55.43.5 Difficulty: 2.6

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.43.5 Difficulty: 3.0

DCPP L191 Exam Rev 0 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. To satisfy a surveillance requirement of LCO 3.5.2, ECCS-Operating, 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 performing STP M-89?

2) What Technical Specification LCO should be entered by the Shift Foreman 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.

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.

Proposed Answer: B. 1) Prevent water hammer of ECCS piping. 2) LCO 3.5.2 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 0 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 X

10CFR Part 55 Content:

55.43.3 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 (43.5)

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:

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. direct the operator to 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. direct the operator to 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. take the action of LCO 3.7.9, Ultimate Heat Sink, to place a second CCW heat exchange in service, to prevent possibly exceeding CCW maximum design temperature during a design basis accident.

Proposed Answer: D. take the action of LCO 3.7.9, Ultimate Heat Sink, to place a second CCW heat exchange in service, 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 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 References to be provided to applicants during exam: None

DCPP L191 Exam Rev 0 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 X

10CFR Part 55 Content:

55.43.2 Difficulty: 3.0

DCPP L191 Exam Rev 0 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 According to the bases for LCO 3.7.5, Auxiliary Feedwater (AFW) System, AFW is designed to remove decay heat for which of the following?

1. Steam Line Break

2. Feed Line Break

3. Large Break LOCA A. 1 only B. 1 and 2 C. 2 and 3 D. 3 only Proposed Answer: B. 1 and 2 Explanation:

Per the bases for LCO 3.7.5, The limiting Design Basis Accidents (DBAs) and transients for the AFW System are as follows:

a. Feedwater (FWLB) or Main Steam Line Break (MSLB); and

b. Loss of normal feedwater both with and without a loss of offsite power.

A. Incorrect. Steam Line break is one of the accidents but feed line break is also DBA for AFW.

B. Correct. Both steam line and feed line breaks are DBAs for AFW.

C. Incorrect. Large LOCA is not a DBA for AFW. However, a small break LOCA considers AFW in its analysis.

D. Incorrect. Large LOCA is not correct. For a large LOCA, RCS pressure is less than steam generator pressure and the steam generators are not used for heat removal.

Technical

References:

B3.7.5 References to be provided to applicants during exam: None Learning Objective: 9694G Apply TS 3.7 Technical Specification bases 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.2 Difficulty: 2.5

DCPP L191 Exam Rev 0 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:

1) If a single 230 kV line is lost, 230 kV _____________ OPERABLE.

2) If a single 500 kV line is lost, 500 kV _____________ OPERABLE.

NOTE: assume 230 kV and 500 kV bus voltage remains unchanged.

A. 1) remains

2) remains B. 1) remains

2) may, depending on area load and bus voltage, remain C. 1) may, depending on area load and bus voltage, remain

2) remains D. 1) may, depending on area load and bus voltage, remain

2) may, depending on area load and bus voltage, remain Proposed Answer: C. 1) may, depending on area load and bus voltage, remain

2) remains Explanation:

A. Incorrect. 230 kV may or may not be OPERABLE, depends on voltage and load. Second part is correct.

B. Incorrect. Both parts incorrect. It is the opposite of the correct answer.

C. Correct. There are two 230 kV lines and three 500 kV lines. According to the bases for LCO 3.8.1, the sources are OPERABLE if one 230 kV (with conditions) and two 500 kV lines satisfy the LCO. A loss of one line of each does not affect operability providing voltage and load for the existing 230 kV lineup is acceptable. For 500 kV a loss of a single line does not make the system inoperable, the 500 kV system remains OPERABLE.

D. Incorrect. 500 kV remains OPERABLE. First part is correct.

Technical

References:

LCO 3.8.1 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

DCPP L191 Exam Rev 0 Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.2 Difficulty: 2.6

DCPP L191 Exam Rev 0 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 GIVEN:

• The crew is performing EOP E-1, Loss of Reactor or Secondary Coolant, Step 8, CHECK If RHR Pps Should Be Stopped

• RCPs are secured

• MSIVs are closed

• No SI or ECCS CCPs are running

• RCS pressure is 800 psig and stable

• RWST level is 75% and lowering slowly

• Applicable RVLIS level is 75% and lowering slowly The STA reports most incore thermocouples are reading 800°F and rising rapidly.

For the current plant conditions, the Shift Foreman should:

A. continue with EOP E-1 until transitioning to EOP E-1.2, Post LOCA Cooldown and Depressurization.

B. continue with EOP E-1 until the transition to EOP E-1.3, Transfer to Cold Leg Recirculation is required.

C. exit EOP E-1 and go to EOP FR-C.1, Response to Inadequate Core Cooling.

D. exit EOP E-1 and go to EOP FR-C.2, Response to Degraded Core Cooling.

Proposed Answer: D. exit EOP E-1 and go to EOP FR-C.2, Response to Degraded Core Cooling Explanation:

A. Incorrect. Thermocouples (greater than 5) are greater than 720°F. A transition to EOP FR-C.2 is warranted. This would be correct if thermocouples were not elevated.

B. Incorrect. Thermocouples (greater than 5) are greater than 720°F. A transition to EOP FR-C.2 is warranted. This would be correct if thermocouples were not elevated and RCS pressure was less than RHR pressure (300 psig).

C. Incorrect. A transition is required, however, the CSF challenge is MAGENTA, not RED. For a valid RED challenge, greater than 5 thermocouples would have to be greater than 1200°F or RVLIS level would have to be lower than 32%,.

D. Correct. For the current conditions, a transition to EOP FR-C.2 is required based on more than 5 thermocouples greater than 720°F and RVLIS above 32%

Technical

References:

EOP F-0, EOP FR-C.1 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

DCPP L191 Exam Rev 0 Question Source:

Bank #

(note changes; attach parent)

Modified Bank #76 L141 04/2016 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:

55.43.

Difficulty: 3.0

DCPP L191 Exam Rev 0 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 the 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 the 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.7

DCPP L191 Exam Rev 0 Difficulty: 2.2

DCPP L191 Exam Rev 0 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 0 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 X

10CFR Part 55 Content:

55.43.4 Difficulty: 2.6

DCPP L191 Exam Rev 0 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 0 Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.2 Difficulty: 3.0

DCPP L191 Exam Rev 0 Examination Outline Cross-Reference Level SRO G2.1.42 Knowledge of new and spent fuel movement procedures (43.7)

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. Maintaining the record of fuel movement in accordance with the Fuel Movement Tracking Sheet

2. Direct supervision of core alteration activities with no concurrent duties

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. This is a duty of the operator in the control room. Plausible as the refueling SRO provides direct oversight and this could be thought of as falling under that definition.

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 X

DCPP L191 Exam Rev 0 Comprehensive/Analysis 10CFR Part 55 Content:

55.43.7 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 HUTCH interlock on the control switch

2) does D. 1) Place a HUTCH interlock 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 incorrect. The SFM has the authority. Plausible that its thought that responsibility lies with the SM.

C. Incorrect. Second part is correct, the SFM has the authority.

D. Incorrect. Both parts are incorrect.

Technical

References:

OP2.ID2, LPECA-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 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.10 Difficulty: 3.0

DCPP L191 Exam Rev 0 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) Normal operator diagnostics such as alarm response and instrumentation validation

__________ require entry into MA1.ID26.

2) Level A troubleshooting, (troubleshooting activities with very high risk, or 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 0 Question Cognitive Level:

Memory/Fundamental X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.10 Difficulty: 2.3

DCPP L191 Exam Rev 0 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.4 Difficulty: 2.5

DCPP L191 Exam Rev 0 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 An EOP and an AOP, of lesser importance, are to be performed.

In accordance with OP1.DC10, Conduct of Operations, how should the performance of the EOP and AOP be conducted?

The Shift Foreman will direct the performance of the EOP and A. direct the performance of the AOP as time permits.

B. assign the AOP to the SFM on the unaffected unit.

C. assign a board operator to perform the AOP.

D. assign another SRO, preferably the STA, and board operator to perform the AOP.

Proposed Answer: C. assign a board operator to perform the AOP.

Explanation:

A. Incorrect. The SFM remains the procedure of the higher importance and assigns a board operator the role of performing the AOP.

B. Incorrect. Plausible, that the other unit SFM assigned. The other unit SFM can be used during emergencies, such as STA, if required.

C. Correct. Per OP1.DC10, If two procedures are being implemented concurrently:

1. The procedure of highest importance should be implemented in the normal manner (procedure reader/board operators).

2. The procedure of lesser importance should be assigned to a single board operator.

D. Incorrect. This seems plausible because normal procedure performance is a reader and an operator, Technical

References:

OP1.DC10 step 28.17.7.g 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 X

Comprehensive/Analysis 10CFR Part 55 Content:

55.43.5 Difficulty: 2.7

DCPP L191 Exam Rev 0 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 0 Question History:

Last Two NRC Exams No Question Cognitive Level:

Memory/Fundamental Comprehensive/Analysis X

10CFR Part 55 Content:

55.43.5 Difficulty: 3.0

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