ML22136A175

Prop Written Exam

ML22136A175
Person / Time
Site:	Point Beach
Issue date:	07/02/2019
From:	NRC/RGN-III/DRS/OLB
To:	Point Beach
Baker R
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ML17215A929	List: ML22136A152 ML22136A161 ML22136A166 ML22136A175 ML22136A184 ML22136A194 ML22136A199 NRC 2019-0015, Letter Submitting Proposed Exam
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Download: ML22136A175 (335)

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QUESTIONS REPORT for 2019 NRC Exam Master

1. 2019 NRC 001/EPE/007EK1.03/3.7/2-DR/RO/BANK/BG-EOP-0/031.02.LP0405.005 Given the following:

Unit 1 is at Rated Thermal Power Both Steam Generator Feed Pumps trip A manual reactor trip is initiated The crew enters EOP-0, Reactor Trip or Safety Injection The Main Turbine FAILS to trip when the Emergency Turbine Trip pushbutton is pressed The next action directed by EOP-0 is to ___(1)___ and the reason for this action is to ___(2)___.

A. (1) shut BOTH Main Steam Isolation Valves (2) prevent steaming the SGs dry B. (1) shut BOTH Main Steam Isolation Valves (2) prevent an excessive RCS cooldown C. (1) manually run back turbine at maximum rate (2) prevent steaming the SGs dry D. (1) manually run back turbine at maximum rate (2) prevent an excessive RCS cooldown Tuesday, April 30, 2019 11:03:44 AM 1

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2015-03 Comanche Peak Question 39 K/A:

007EK1.03 Reactor Trip Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Reasons for closing the main turbine governor valve and the main turbine stop valve after a reactor trip (Imp 3.7/4.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the reason for verifying the main turbine is tripped after a reactor trip Cognitive Level:

Comprehension 2-RI: Requires the operator to recall the actions necessary to trip the turbine when the pushbutton fails to work, and the basis for taking those actions.

10 CFR Part 55 Content:

55.41 8, 10 55.43 Tuesday, April 30, 2019 11:03:44 AM 2

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-0, Reactor Trip or Safety Injection Rev 66 BG-EOP-0, Background Reactor Trip or Safety Injection Rev 44 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 sequence of events:

Reactor power = 100%

Running MFW Pumps trip The reactor tripped The crew enters EOP-0.0A, REACTOR TRIP OR SAFETY INJECTION The Balance of Plant Operator attempts to trip the Main Turbine using the Turbine Trip pushbutton The Main Turbine fails to trip Based on the above plant conditions, the next action directed by EOP-0.0A is to

____(1)____ and the reason for this action is to ____(2)____.

A. (1) close ALL Main Steam Isolation Valves (2) prevent steaming the SGs dry B. (1) close ALL Main Steam Isolation Valves (2) prevent an excessive RCS cooldown C. (1) pull-out ALL EHC fluid pumps (2) prevent steaming the SGs dry D. (1) pull-out ALL EHC fluid pumps (2) prevent an excessive RCS cooldown Proposed answer: B Tuesday, April 30, 2019 11:03:44 AM 3

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per EOP-0 step 3 RNO a. Manually trip the turbine. If the turbine will not trip, then shut MSIVs". The turbine is tripped to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require.

A INCORRECT: The first part is correct. The second part is incorrect, plausible because analysis during an ATWS assumes a turbine trip within 30 seconds for the conservation of SG inventory.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, plausible because this is the RNO action taken for failure of the turbine to trip in CSP-S.1. The second part is incorrect, plausible because analysis during an ATWS assumes a turbine trip within 30 seconds for the conservation of SG inventory.

D INCORRECT: The first part is incorrect, plausible because this is the RNO action taken for failure of the turbine to trip in CSP-S.1. The second part is correct.

Learning Objective:

Identify the basis for the steps in the Emergency Operating Procedures.

(031.02.LP0405.005)

State the immediate action steps for EOP-0, and given access to appropriate site specific simulator indications, verity that their intent is satisfied.

(031.02.LP0405.006)

Tuesday, April 30, 2019 11:03:44 AM 4

QUESTIONS REPORT for 2019 NRC Exam Master

2. 2019 NRC 002/APE/008AK2.01/2.7*/3-SPR/RO/BANK/NUC-GFP-HXF-004/NUC-GRP-HXF-004.023 Given the following:

The unit is at Rated Thermal Power One Pressurizer PORV is leaking past its seat Pressurizer pressure is 2235 psig Pressurizer Steam Space temperature is 653°F PRT pressure is 15 psig Which of the following is the approximate expected temperature downstream of the leaking Pressurizer PORV?

A. 653°F B. 650°F C. 300°F D. 250°F RO Tier 1 Group 1 Source: Bank Question History:

2015 McGuire Question 40 K/A:

008AK2.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Valves (Imp 2.7*/2.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand the relationship between valves, and the correct indications of tailpipe temperature, and that it is less than PZR vapor space.

Cognitive Level:

Comprehension 3-SPR: Requires the operator to use the steam tables, and knowledge to determine what actual tailpipe temperature is if a leak is actually occurring.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:44 AM 5

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

NUC-GFP-HXF-004, Thermodynamic Processes, Rev 0 Proposed reference to be provided to the applicants during examination:

Steam Tables Original Question:

Given the following conditions on Unit 2:

The unit is operating at 100% RTP One PZR PORV is leaking past its seat Pressurizer pressure is 2235 PSIG Pressurizer Steam Space temperature is 653°F PRT pressure is 15 PSIG Which ONE (1) of the following is the approximate expected temperature downstream of the leaking PZR PORV?

REFERENCE PROVIDED A. 220°F B. 240°F C. 250°F D. 300°F Proposed answer: C Tuesday, April 30, 2019 11:03:44 AM 6

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the PORV is throttling, the pressurizer is at 2235 psig (2250 psia), and the PRT is 15 psig (or 30 psia), a wet vapor is entering the PRT. Therefore the temperature of the wet vapor entering the PRT will be the Saturation temperature for the given pressure (30 psia) or 250°F.

A INCORRECT: Plausible if the operator disregards the throttling effect of the PORV and the change to PSIA, and determines the saturation temperature giving operating pressure of 2250 psia.

B INCORRECT: Plausible if the operator disregards the throttling effect of the PORV but remembers to change units to PSIA, and determines the saturation temperature giving operating pressure of 2220 psia.

C INCORRECT: Plausible if the operator uses the Mollier Diagram and correctly located the intersection of constant enthalpy line and the 30 psia constant pressure line but then incorrectly goes straight up to the saturation line.

D CORRECT: See above.

Learning Objective:

Given an h-s diagram and / or the steam tables and the upstream conditions, derive downstream conditions from a throttling process (NUC-GRP-HXF-004.023)

Tuesday, April 30, 2019 11:03:44 AM 7

QUESTIONS REPORT for 2019 NRC Exam Master

3. 2019 NRC 003/EPE/009EK2.03/3.0/3-SPK/RO/BANK/BG-EOP-1.2/031.02.LP0435.014 Given the following:

Unit 1 was at Rated Thermal Power when an automatic reactor trip and safety injection occurred Twenty minutes after the initial transient:

No RCPs are running Core exit thermocouples read 580°F Pressurizer level is off-scale low RCS pressure is 1310 psig The crew begins drawing more steam from the SGs and raising the AFW flow rate to maintain SG levels How and why will ECCS flow be affected as a result of these operator actions?

ECCS flowrate will . . .

A. remain the same. Cooling down the RCS will raise subcooling, but will not affect RCS pressure.

B. remain the same. The ECCS pumps will maintain a constant pressure for a specific size of RCS leak.

C. rise. As the RCS cools down, its pressure lowers, and the ECCS pumps operate at a lower discharge pressure.

D. rise. Cooldown of RCS will lower pressurizer spray temperature, lowering pressurizer pressure, allowing it to refill.

Tuesday, April 30, 2019 11:03:44 AM 8

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2014 Byron Question 2 K/A:

009EK2.03 Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following: S/Gs (Imp 3.0/3.3*)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand the relationship between AFW flow effect on the steam generators, and what effect that will in turn have on the RCS during a small break loss of coolant accident.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to understand the relationship of flow of AFW to the steam generators, what affect that will have on RCS pressure and temperature and how that will effect ECCS flowrates.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:44 AM 9

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

BG-EOP-1.2, Background Post LOCA Cooldown and Depressurization Rev 28 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 had an automatic reactor trip and safety injection occur from full power.

Twenty minutes after the initial transient:

No RCP's are running.

Core exit TC's read 580°F.

PZR level is off-scale low.

RCS pressure is 1310 psig.

The operators begin withdrawing more steam from the S/G's and raise the AFW flow rate to maintain S/G levels.

How and why will ECCS flow change as a result of these operator actions?

ECCS flowrate will...

A. remain the same. The ECCS pumps will maintain a constant pressure for a specific size of RCS leak.

B. remain the same. Cooling down the RCS will raise subcooling, but will not affect RCS pressure.

C. increase. Cooldown of RCS will lower Pzr spray temperature, lowering pressurizer pressure, allowing it to refill.

D. increase. As the RCS cools down, its pressure lowers, and the ECCS pumps operate against a lower head.

Proposed answer: D Tuesday, April 30, 2019 11:03:44 AM 10

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

EOP-1.2 steps 4, 6 7, and 8, establish a heat sink volume in the steam generators and then cooldown/depressurize the RCS. The rise in steam flow from the steam generators, and well as the rise in AFW flow to the steam generators, will cause the RCS to cooldown. The pressurizer being off-scale low, means that the pressurizer will not be able to maintain the pressure of the RCS. Therefore as the RCS is cooled down, the pressure will also lower, this lowering in pressure will allow the SI pumps to provide more flow.

A INCORRECT: Plausible if the operator doesnt take into consideration that the pressurizer is empty and therefore cannot control pressure.

B INCORRECT: Plausible if the operator doesnt properly recall the pump and system flow laws where flow rate will be affected by both leak size and cooldown/pressure changes.

C CORRECT: See above.

D INCORRECT: Plausible if the operator disregards the RCPs not running, as the flow rate will rise, but since the RCPs are not running, there will be no effect on spray temperature.

Learning Objective:

Describe the following:

a. Preferred method of reestablishing Core Cooling for a Small Break LOCA where the steam generator plays a vital role as a heat sink

b. Methods of increasing safety injection flow for a lager break LOCA where Safety Injection flow is vital to replace mass lost from the system.

c. Role of refluxing in specified transients (031.02.LP0435.014)

Tuesday, April 30, 2019 11:03:44 AM 11

QUESTIONS REPORT for 2019 NRC Exam Master

4. 2019 NRC 004/EPE/011EA1.06/4.2/1-F/RO/NEW/AOP-22/54.02.LP0133.005 Given the following:

Unit 1 was at Rated Thermal Power A Large Break LOCA coincident with a loss of off-site power occurred Due to plant configuration, G01, Emergency Diesel Generator is loaded to 2650 kW What is the LONGEST amount of time G01 can run at this load without exceeding design limits?

(Do not take into consideration normal required preventative maintenance)

A. Limited to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year at this rate B. Limited to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per year at this rate C. Limited to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per year at this rate D. Continuously RO Tier 1 Group 1 Source: New Question History:

None K/A:

011EA1.06 Large Break LOCA Ability to operate and monitor the following as they apply to a Large Break LOCA: D/Gs (Imp 4.2/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the emergency diesel loading limits which would apply during all conditions including during a LOCA.

Cognitive Level:

Knowledge 1-F: Requires the operator to recall the diesel loading limitations.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:44 AM 12

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-22, EDG Load Management Unit 1 Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The diesel load limits for G01 and G02 are 2850 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> 2963 kW for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> 3000 kW for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> With G01 Emergency Diesel Generator loaded at 2650 kW, the limit would be continous based on being less than 2850 kW A INCORRECT: Plausible if the operator does not correctly recall the diesel loading limits for G01 and G03.

B INCORRECT: Plausible if the operator does not correctly recall the diesel loading limits for G01 and G03.

C INCORRECT: Plausible if the operator does not correctly recall the diesel loading limits for G01 and G03.

D CORRECT: See above.

Learning Objective:

DESCRIBE the procedures which govern operation of the Diesel Generator System. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed or Auxiliary Operators.

(054.02.LP0133.005)

Tuesday, April 30, 2019 11:03:44 AM 13

QUESTIONS REPORT for 2019 NRC Exam Master

5. 2019 NRC 005/APE/015/017AA2.09/3.4/3-SPK/RO/NEW/AOP-1B/055.03.LP2438.004 Given the following:

Unit 2 is at Rated Thermal Power A seismic event has occurred The following was noted on Unit 2 after the event The crew has entered AOP-1B, Reactor Coolant Pump Malfunction Based on the above information, if the trends did not change, which RCP will be required to be tripped first, and why?

A. RCP A, due to stator temperature B. RCP A, due to hottest motor bearing temperature C. RCP B, due to stator temperature D. RCP B, due to hottest motor bearing temperature Tuesday, April 30, 2019 11:03:44 AM 14

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: New Question History:

None K/A:

015/017AA2.09 Reactor Coolant Pump (RCP) Malfunctions Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on high stator temperatures (Imp 3.4/3.5)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the rate of rise and the recall the limits of the RCPs, and determine that an RCP needs to be tripped based on exceeding the stator temperature limits.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to determine the rate temperature trends and the recall the limits of the RCPs, and determine which RCP needs to be tripped based on which limit that was exceeded 10 CFR Part 55 Content:

55.41 55.43 5

Reference:

AOP-1B, Reactor Coolant Pump Malfunction Unit 2 Rev 26 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:44 AM 15

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per the foldout page of AOP-1B, RCP stator temperature limits are RCP A >140°C and RCP B >120°C, with RCP cubicle smoke detector alarm or with abnormal amperage The operator must determine the trend for the hottest motor brg and stator for both reactor coolant pumps RCP A RCP B Hottest Stator Amps Hottest Stator Amps Motor Brg Motor Brg Trend ~2°C/10min ~2°C/10min 20 Amps 2°C/10min 8°C/10min Stable Then stable And then determine when a limit will be exceeded. B RCP will be the first RCP requiring a trip per AOP-1B, based on exceeding 120°C with abnormal amps A INCORRECT: Plausible as this value will exceed the limits of 140°C for the stator, but the operator must also recall that it must be combined with abnormal amps.

B INCORRECT: Plausible as the operator will need to recall the correct temperature for both the stator and motor bearing, and this will exceed the limit for motor bearing before the B RCP will.

C CORRECT: See above.

D INCORRECT: Plausible if the operator confused the requirements of motor bearing temperature with that of the stator and requires it to have the abnormal amps also.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following events:

a. Reactor Coolant System leakage

b. Reactor Coolant Pump malfunctions Steam Generator Tube leak (055.03.LP2438.004)

Tuesday, April 30, 2019 11:03:44 AM 16

QUESTIONS REPORT for 2019 NRC Exam Master

6. 2019 NRC 006/APE/025AG2.4.08/3.4/1-B/RO/NEW/SEP-3/031.03.LP2437.008 Given the following:

Unit 1 is preparing to enter MODE 4 from MODE 5 RHR train A is running in DHR mode, and RHR train B is lined up for ECCS injection A break on the suction of 1P-10A, Residual Heat Removal pump causes fluctuating flow and discharge pressure indications.

Coincident with this a station blackout occurs Considering Unit 1 ONLY, as a result of the loss of inventory, RHR cooling and AC power, the crew will . . .

A. enter the AOP network.

B. enter the ECA network.

C. enter the EOP network.

D. enter the SEP network.

RO Tier 1 Group 1 Source: New Question History:

None K/A:

025AG2.4.8 Loss of Residual Heat Removal System (RHRS)

Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

(Imp 3.8/4.5)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine which procedure set need to be entered to take appropriate actions to mitigate the events Cognitive Level:

Knowledge 1-B: Requires the operator to determine, based on initial conditions what procedure set should be entered.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:44 AM 17

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

SEP-1, Degraded RHR System Capability, Unit 1 Rev 17 SEP-3, Loss of All AC Power while on Shutdown Cooling, Unit 1 Rev 38 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The operator needs to make assumptions for plant lineup and conditions in MODE 5, including RCS temperature, accumulator isolation valve position etc as well as the prioritization of which event to mitigate first, the loss of cooling, loss of inventory, or loss of power.

The entry conditions for SEP network are, SEP-1, RHR pump discharge pressure unstable, pump cavitation, loss of power to the operating RHR pump, and inadequate flow for RCS cooling SEP-2, Loss of inventory, drop in pressure SEP-3, Loss of all AC power and RHR is aligned for cooling A INCORRECT: Plausible as several AOPs deal with the various entry conditions, the loss of power to safeguards buses, loss of RHR flow, loss of inventory and ability to charge all have AOPs to deal with those issues.

B INCORRECT: Plausible as SEP-3 directs transition to the ECA network in certain situations and ECA-0.0 entry conditions are the plant in MODE 1, 2, 3, or 4, and transition was just accomplished.

C INCORRECT: Plausible as SEP-1 will direct entry into EOP-0 if the SI accumulators are not isolated at step 1 of the SEP.

D CORRECT: See above.

Learning Objective:

Given appropriate conditions or access to the simulator, implement SEP-3.0 to restore AC power to at least one safeguards bus (031.03.LP2437.008)

Tuesday, April 30, 2019 11:03:44 AM 18

QUESTIONS REPORT for 2019 NRC Exam Master

7. 2019 NRC 007/APE/026AA2.02/2.9/2-DR/RO/BANK/LP0084/051.06.LP0084.004 Given the following:

Unit 1 is at Rated Thermal Power Component Cooling pumps are aligned as shown below:

A manual Reactor Trip was initiated due to a large feedwater leak in the Turbine Building During performance of Immediate Actions of EOP-0, Reactor Trip or Safety Injection an AUTO Safety Injection occurred After completing all immediate actions, a lockout on 1X-04, Low Voltage Station Transformer occurs All automatic actions occur and the safeguards buses are re-energized from the Emergency Diesel Generators Which of the following indicates the status of the Component Cooling Water Pumps after the buses are re-energized?

A. 1P-11A is running, 2P-11B is in standby B. 1P-11A is tripped, 2P-11B is in running C. Both CCW Pumps are running D. Neither CCW Pump is running Tuesday, April 30, 2019 11:03:45 AM 19

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2003 PBNP Question 8 K/A:

026AA2.02 Loss of Component Cooling Water (CCW)

Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The cause of possible CCW loss (Imp 2.9/3.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the status of the component cooling water pumps given a series of events.

Cognitive Level:

Comprehension 2-DI: Requires the operator to determine understand the initial conditions, and then determine what effect the events will have on the status of the component cooling water pumps and what they will indicate.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:45 AM 20

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

LP0084 Lesson Plan, Component Cooling Water System Rev 19 883D195 Sh 9, Safeguards Sequence Logic Drawing Rev 19 AOP-18A, Train A Equipment Operation Unit 1 Rev 17 AOP-18B, Train B Equipment Operation Unit 1 Rev 17 STPT 8.1, Auxiliary Coolant System Setpoints: General Instrumentation Rev 9 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following sequence of events:

- Unit 2 is operating at 100% power.

- 2P-11A, Component Cooling Water Pump, is running.

- 2P-11B, Component Cooling Water Pump, is in standby.

- A manual Reactor Trip was initiated due to a large feedwater leak in the Turbine Building.

- During performance of Immediate Actions of EOP-0, a manual Safety Injection is initiated due low Pressurizer Level.

- After completing all immediate actions, a lockout on Low Voltage Station Transformer 1X-04 occurs.

- All automatic actions occur and the safeguards buses are re-energized from the Emergency Diesel Generators.

Which of the following indicates the status of the Component Cooling Water Pumps after the buses are re-energized?

A. 2P-11A is running, 2P-11B is in standby.

B. Neither CCW Pump is running.

C. 2P-11A is tripped, 2P-11B is running.

D. Both CCW Pumps are running.

Proposed answer: B Tuesday, April 30, 2019 11:03:45 AM 21

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

In the initial conditions A pump is the running pump and pump B is in standby.

With a safety injection signal, the component cooling water pumps will continue to run in the configuration they are in (A running, and B standby). When the lockout on 1X-04 occurs, the SI signal will still be present, this will cause the pumps to be stripped off the bus, and the control switch will need to be taken to the STOP position to reset the pump lockout.

A INCORRECT: Plausible if the operator has the misconception that the safety injection will not affect the pumps, at this is what would happen with just an SI B INCORRECT: Plausible if the operator has the misconception that the running pump breaker will trip open on the bus lockout, as most other breaker do, with the standby pump picking up on low pressure.

C INCORRECT: Plausible if the operator disregarded the effected of the safety injection and only took in consideration of the effect of the lockout on 1X-04.

D CORRECT: See above.

Learning Objective:

DESCRIBE the interlocks associated with the Component Cooling Water System and its major components:

a. Vent valve to atmosphere (CC-17)

b. CC-769

c. RCP Pump CCW valves

d. RADWASTE CCW supply and return valves

e. Standby CCW Pump start (051.06.LP0084.004)

Tuesday, April 30, 2019 11:03:45 AM 22

QUESTIONS REPORT for 2019 NRC Exam Master

8. 2019 NRC 008/APE/027AA1.04/3.9*/1-I/RO/BANK/AOP-18A/055.03.LP2440.002 Given the following:

Unit 1 is responding to a loss of offsite power Both G01 and G03, Emergency Diesel Generators have started and loaded onto their respective buses Safety Injection did NOT actuate Pressurizer level is 25%

What must be done to energize 1T-1C, Backup Group C Heaters?

A. Turn the 1T-1C control switch to OFF.

Then turn the 1T-1C control switch to ON.

B. Reset the 1B-03 Non Safeguards Equipment lockout.

Leave the 1T-1C control switch in AUTO.

C. Restore power to 1B-01, 480V Non-Safeguards bus.

Then turn the 1T-1C control switch to ON.

D. Place 1HC-431K, Pressurizer Pressure Controller, in MANUAL and raise the controller output.

Leave the 1T-1C control switch in AUTO.

Tuesday, April 30, 2019 11:03:45 AM 23

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2012 PBNP Question 4 K/A:

027AA1.04 Pressurizer Pressure Control System (PZR PCS) Malfunction Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: Pressure recovery, using emergency-only heaters.

(Imp 3.9*/3.6*)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the actions necessary to energize an emergency set of pressurizer heaters.

Cognitive Level:

Knowledge 1-I: Requires the operator to recall the interlocks and actions necessary to energize the emergency pressurizer heaters.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 24

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-18A, Train A Equipment Operation Unit 1 Rev 17 883A195 Sh 9, Safeguards Sequence Logic Drawing Rev 19 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following conditions:

Unit 1 is responding to a loss of offsite power Both Emergency Diesel Generators G-01 and G-03 have started and loaded onto their respective buses Safety Injection did NOT actuate Pressurizer level is 25%

What must be done to energize 1T-1C, Backup Group C Heaters?

A. Restore power to 1B-01.

Then turn the 1T-1C control switch to ON.

B. Reset the 1B-03 Non Safeguards Equipment lockout.

Leave the 1T-1C control switch in AUTO.

C. Turn the 1T-1C control switch to OFF.

Then turn the 1T-1C control switch to ON.

D. Place 1HC-431K, Pressurizer Pressure Controller, in MANUAL and raise the controller output.

Leave the 1T-1C control switch in AUTO.

Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 25

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Power is available for the C bank of heaters. This breaker is stripped on under voltage and can be reset by taking the associated control switch to off and back to on.

A CORRECT: See above.

B INCORRECT: The first part is incorrect, 1B-03 has power restored via EDGs.

The Equipment lockout does strip loads but is actuated with an SI, which has not occurred. Resetting this lockout is a common task in the EOP network. Plausible if the operator has the misconception that the effect of the under voltage is the same as an SI. The second part is incorrect, this action will not reset the heaters, plausible if this operator has the misconception that the heaters ride the bus similar to other equipment which does that.

C INCORRECT: The first part is incorrect, power to 1B-01 is not restored via EDGs, this bus is still de-energized with the loss of offsite power, plausible if the operator incorrectly recalls the power supply to the pressurizer heaters. The second part is incorrect, to reset heaters the control switch must be taken to off first.

Plausible if the operator has a misconception of breaker operation and does not recall this breaker needs to be reset.

D INCORRECT: The first part is incorrect, plausible because in normal no loss of power situation, taking HC-431K to manual is a way to turn on pressurizer heaters and raise or maintain pressure. The second part is incorrect, this action will not reset the heaters, plausible if this operator has the misconception that the heaters ride the bus similar to other equipment which does that.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following conditions:

a. Turbine Generator Voltage Regulator failure

b. Loss of Main Generator Hydrogen pressure

c. Total collapse of 345 KV system frequency Loss of electrical buses (055.03.LP2440.002)

Tuesday, April 30, 2019 11:03:45 AM 26

QUESTIONS REPORT for 2019 NRC Exam Master

9. 2019 NRC 009/EPE/029EK3.10/4.1/1-B/RO/BANK/BG-CSP-S.1/043.03.LP1996.013 Given the following Unit 1 is at Rated Thermal Power The basis for manually inserting control rods during an ATWS event is to reduce reactor power to:

A. prevent the fuel from exceeding localized linear heat rate limits B. prevent rapid heatup of the RCS to limit thermally induced stresses across the reactor vessel C. ensure there is sufficient steam dump capacity to prevent opening the steam generator code safeties D. ensure the reactor is shutdown, so the only heat transferred to the RCS is from core decay heat and RCP heat RO Tier 1 Group 1 Source: Bank Question History:

2009-301 Catawba SRO 2008 Retake Question 8 K/A:

029EK3.10 Anticipated Transient Without Scram (ATWS)

Knowledge of the reasons for the following responses as the apply to the ATWS: Manual rod insertion (Imp 4.1/4.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the reason for manual rod insertion during an ATWS Cognitive Level:

Knowledge 1-I: Requires the operator to recall the reason/basis for manual rod insertion during an ATWS 10 CFR Part 55 Content:

55.41 5 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 27

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

BG-CSP-S.1, Background Response to Nuclear Power Generator / ATWS Rev 28 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 is operating at 100% power. The basis for manually inserting control rods during an ATWS event is to reduce reactor power to:

A. prevent exceeding rated thermal power limits defined in Technical Specifications B. prevent rapid heatup of the NC system and potential overfill of the pressurizer C. ensure the only heat added to the NC system is from core decay heat and NC pump heat D. ensure there is sufficient steam dump capacity to prevent opening the steam line code safeties Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 28

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per the background document for CSP-S.1 The safeguards systems that protect the plant during accidents are designed assuming that only decay heat and pump heat are being added to the RCS., so the reactor needs to be shutdown, so that the only heat input is decay heat and RCP heat.

A INCORRECT: Plausible, as the RCP may become solid if the pressurizer PORV lifts due to an overpressure condition and with temperature rising due to the reactor being critical, and the turbine being tripped, this will cause a rise in temperature of the reactor vessel. Control of pressure/temperature of Rx vessel is not the purpose of the manual rod insertion, a controlled cooldown/steam dumps are the effective method to accomplish this.

B INCORRECT: Plausible, if the operator confuses overheating with fuel limts, this is not the basis for manually inserting rods, but will have a desired effect of lowering power as the rods are inserted, lowering heat rates in the fuel.

C INCORRECT: Plausible if the operator has the misconception that manual rod insertion is necessary to ensure the steam dumps system does not exceed its operating capacity.

D CORRECT: See above.

Learning Objective:

List the major actions accomplished by the Subcritically Critical Safety Function Procedures (043.03.LP1996.013)

Tuesday, April 30, 2019 11:03:45 AM 29

QUESTIONS REPORT for 2019 NRC Exam Master

10. 2019 NRC 010/APE/040AK2.02/2.6*/1-I/RO/BANK/STPT 2.2/052.02.LP0153.003 Given the following:

The unit was at Rated Thermal Power A Steam Line Break occurred down-stream of the MSIVs Both SG pressures were initially LOWERING SLOWLY Both SG steam flows reached 4.0 E6 lb/hr Both SG pressures lowered to 615 psig and are LOWERING Pressurizer pressure is 1720 psig and LOWERING TAVG is 539°F and LOWERING Which of the following describes the expected ESF actuations?

Safety Injection Main Steam Isolation A. Low Steamline Pressure High Steamline Flow B. Low Steamline Pressure High-High Steamline Flow C. Low Pressurizer Pressure High Steamline Flow D. Low Pressurizer Pressure High-High Steamline Flow RO Tier 1 Group 1 Source: Bank Question History:

2014 Ginna Question 8 K/A:

040AK2.02 Steam Line Rupture Knowledge of the interrelations between the Steam Line Rupture and the following: Sensors and detectors (Imp 2.6*/2.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to evaluate plant conditions, and determine which parameter setpoint has been exceeded and casued the isolation of the steam line rupture Cognitive Level:

Knowledge 1-I: Requires the operator to evaluate plant conditions, recall the parameters and setpoints for MISV actuation and Safety Injection.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 30

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

STPT 2.1, Safety Injection Setpoint Document, Rev 5 STPT 2.2, Steam Line Isolation Setpoint Document, Rev 7 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following plant conditions:

The plant was at 100% power A Steam Line Break occurred down-stream of the MSIVs Both S/G pressures were initially lowering Both S/G steam flows reached 4.0 x 106 lbm/hr Both S/G pressures lowered to 595 psig and have stabilized Pressurizer pressure is 1720 psig and lowering Tavg is 540°F and lowering WHICH ONE of the following describes the expected ESF actuations?

Safety Injection Main Steam Isolation A. Low Steamline Pressure High Steamline Flow B. Low Steamline Pressure High-High Steamline Flow C. Low Pressurizer Pressure High Steamline Flow D. Low Pressurizer Pressure High-High Steamline Flow Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 31

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With RCS pressure at 1720 psig, this is less than the setpoint of 1735 psig for low pressurizer pressure and will cause the initial SI signal. And with the SI signal present, and steam flow greater than 0.52 E6 lb/hr, and TAVE less than 543°F, High steam flow will cause the closure both MSIVs.

A INCORRECT: The first part is incorrect, plausible if the operator believes that the steam line pressure is less than the SI setpoint of 545 psig.

The second part is correct.

B INCORRECT: The first part is incorrect, plausible if the operator believes that the steam line pressure is less than the SI setpoint of 545 psig.

The second part is incorrect, but plausible as this flow is greater than the rated thermal power normal steam flow, but not higher than the set point of 4.85 E6 lb/hr.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, but plausible as this flow is greater than the rated thermal power normal steam flow, but not higher than the set point of 4.85 E6 lb/hr.

Learning Objective:

DESCRIBE the interlocks associated with the Main Steam System and components.

(052.02.LP0153.003)

Tuesday, April 30, 2019 11:03:45 AM 32

QUESTIONS REPORT for 2019 NRC Exam Master

11. 2019 NRC 011/APE/054AK1.02/3.6/1-I/RO/BANK/BG-CSP-H.1/043.03.LP1998.006 Given the following:

The crew is performing CSP-H.5, Response to Steam Generator Low Level

'A' Steam Generator wide range level is 100 inches and LOWERING SLOWLY Containment Pressure is 6 psig Which answers the following:

(1) What is the highest level which the 'A' Steam Generator is considered "DRY"?

AND (2) Feed flow is NOT established to a "DRY" Steam Generator because . . .

A. (1) LESS THAN 85 inches; (2) significant thermal stresses could be induced on steam generator components when the relatively cold feedwater flow is reinitiated B. (1) LESS THAN 85 inches; (2) feedwater introduction could result in an uncontrolled RCS cooldown and reduction in shutdown margin C. (1) LESS THAN 50 inches; (2) significant thermal stresses could be induced on steam generator components when the relatively cold feedwater flow is reinitiated D. (1) LESS THAN 50 inches; (2) feedwater introduction could result in an uncontrolled RCS cooldown and reduction in shutdown margin Tuesday, April 30, 2019 11:03:45 AM 33

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2012 Ginna Retake Question 11 K/A:

054AK1.02 Loss of Main Feedwater (MFW)

Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G.

(Imp 3.6/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall both the definition of a dry steam generator and the effects when feeding a dry steam.

Cognitive Level:

Knowledge 1-I: Requires the operator to recall both the definition of a dry steam generator based on adverse containment and the effects when feeding a dry steam.

10 CFR Part 55 Content:

55.41 8, 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 34

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

BG-CSP-H.5, Background Response to Steam Generator Low Level Rev 12 CSP-H.5, Response to Steam Generator Low Level Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

Plant conditions:

The crew is performing FR-H.5 RESPONSE TO STEAM GENERATOR LOW LEVEL, for 'A' S/G

'A' S/G wide range level is 130 inches and lowering slowly Containment Pressure is 6 psig (1) At what level will 'A' S/G be considered "dry"?

(2) Feed flow is NOT established to a "dry" S/G because ______'

A. (1) LESS THAN 100 inches; (2) significant thermal stresses could be caused on SIG components when the relatively cold feedwater flow is reinitiated B. (1) LESS THAN 100 inches; (2) feedwater introduction could result in an uncontrolled RCS cooldown and reduction in shutdown margin C. (1) LESS THAN 50 inches; (2) significant thermal stresses could be caused on S/G components when the relatively cold feedwater flow is reinitiated D. (1) LESS THAN 50 inches; (2) feedwater introduction could result in an uncontrolled RCS cooldown and reduction in shutdown margin Proposed answer: a Tuesday, April 30, 2019 11:03:45 AM 35

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The setpoint for a dry steam generator is 85 (adverse containment which is > 5 psig) and per the background document if feed flow to a S/G is isolated and the S/G is allowed to dry out, subsequent reinitiation of feed flow to the S/G could create significant thermal stress condition on S/G components.

A CORRECT: See above B INCORRECT: The first part is correct. The second part is incorrect, but plausible as this the basis statement for restoring feed to a faulted steam generator but would not be the correct basis in this case.

C INCORRECT: The first part is incorrect, plausible because this is the correct setpoint for a non-adverse containment. The second part is correct.

D INCORRECT: The first part is incorrect, plausible because this is the correct setpoint for a non-adverse containment. The second part is incorrect, but plausible as this the basis statement for restoring feed to a faulted steam generator but would not be the correct basis in this case.

Learning Objective:

State the major actions accomplished by each of the following Critical Safety procedures.

CSP-H.1 CSP-H.2 CSP-H.3 CSP-H.4 CSP-H.5 (043.03.LP1998.006)

Tuesday, April 30, 2019 11:03:45 AM 36

QUESTIONS REPORT for 2019 NRC Exam Master

12. 2019 NRC 012/EPE/055EK1.02/4.1/1-P/RO/NEW/BG-ECA-0.0/031.02.LP0462.006 Given the following:

The crew has entered ECA-0.0, Loss of All AC Power, due to a loss of all onsite and offsite power The containment is NOT adverse The crew is depressurizing the intact steam generators to 320 psig Why is the depressurization stopped when meeting this condition?

A. To prevent the loss of pressurizer level, followed by voiding in the reactor vessel upper head region.

B. To prevent injection of safety injection accumulator nitrogen into the RCS which would impact natural circulation.

C. To ensure secondary inventory is maintained prior to the loss of TDAFW pump flow due to low steam generator pressures.

D. To ensure enough secondary inventory is available to prevent the transfer of the pressurizer bubble to the reactor vessel head.

RO Tier 1 Group 1 Source: New Question History:

None K/A:

055EK1.02 Loss of Offsite and Onsite Power (Station Blackout)

Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Natural circulation cooling.

(Imp 4.1/4.4)

Justification for K/A Match:

Matches the K/A by having the operator recall what the reason behind the stopping point for depressurization during a loss of offsite and onsite power, which impacts natural circulation cooling.

Cognitive Level:

Knowledge 1-P: Requires the operator to recall the reason for stopping the depressurization.

10 CFR Part 55 Content:

55.41 8, 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 37

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

BG-ECA-0.0, Background Loss of All AC Power Rev 38 ECA-0.0, Loss of All AC Power Unit 1 Rev 70 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The procedural stopping point for the depressurization is either 320 psig or 32%

narrow range indication. The 320 psig is done to keep the steam generator above 280 psig to prevent a significant injection of nitrogen which would impede natural circulation flow.

C INCORRECT: Plausible as the loss of inventory will be caused by cooling down the plant, and voiding in the upper head region / possible transfer of the bubble from the pressurizer to the vessel can be caused by a lowering of primary pressure, this condition is anticipated and a procedural note directs the depressurization should not be stopped to prevent these occurrences.

D CORRECT: See above B INCORRECT: Plausible because as steam generator pressure lowers, steam supply pressure to the TDAFW pump lowers. Maintaining steam generator inventory is not a concern for maintaining the TDAFW pump, but maintaining cooling for the RCS.

A INCORRECT: Plausible as voiding in the upper head region and possible transfer of the bubble from the pressurizer to the vessel, this condition is anticipated and should not interfere with operator action in the depressurization steps.

Learning Objective:

Given access to the site specific simulator or specific plant conditions, diagnose and respond to a loss of AC power per ECA-0.0 to restore power to at least one Safeguards Bus.

(031.02.LP0462.006)

Tuesday, April 30, 2019 11:03:45 AM 38

QUESTIONS REPORT for 2019 NRC Exam Master

13. 2019 NRC 013/APE/057AK3.01/4.1/2-RI/RO/NEW/AOP-0.2/055.03.LP3456.002 Given the following:

Unit 1 is performing a plant startup per OP 1B, Reactor Startup Reactor power is 1%

1Y-03, White 120V Vital Instrument Panel, de-energized when 1Y-3-M1, Power to 1Y-03, trips open causing the loss of power to:

1N-42, Power Range NI 1N-36, Intermediate Range NI 1N-32, Source Range NI What actions will be required to be taken?

A. Enter EOP-0, Reactor Trip and Safety Injection, verify reactor trip, stabilize the plant, then restore 1Y-03 B. While continuing with the reactor startup, establish minimum charging, and establish excess letdown, then restore 1Y-03 C. Stabilize and maintain reactor power less than P-6, then remove 1Y-03 from service, then continue reactor and plant startup D. Perform Attachment H, Delayed Restart Attempt, manually insert control banks, open reactor trip breakers and verify all rods on are fully inserted, then restore 1Y-03 Tuesday, April 30, 2019 11:03:45 AM 39

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: New Question History:

None K/A:

057AK3.01 Loss of Vital AC Electrical Instrument Bus Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus.

(Imp 4.1/4.4)

Justification for K/A Match:

Matches the K/A by having the operator determine what affect the loss of vital AC electrical instrument bus and what actions are necessary to mitigate per the EOP network and AOP network.

Cognitive Level:

Comprehension 2-RI: Requires the operator to determine the affect the loss of vital AC electrical instrument bus on the plant, and what actions are necessary to mitigate per the EOP network and AOP network.

10 CFR Part 55 Content:

55.41 5, 10 55.43

Reference:

AOP-0.2, Loss of Safety Related Instrument Buses, Attachment, Rev 6 883D195 Sh 11, Nuclear Instrument Trip Signals Rev 7 1-SOP-Y-Y03, 1Y-03, White 120V Vital Instrument Panel, Rev 12 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:45 AM 40

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With 1Y-03 losing power, you will meet the entry conditions for AOP-0.2, with a loss of the white bus. Per AOP-0.2, this will not normally cause a reactor trip, at RTP, but current plant conditions put the plant at < 10% reactor power. With the loss of the bus, will be a loss of power to the intermediate range NI, causing a reactor trip. Even though you will meet the entry conditions to AOP-0.2, the trip must be verified first, and then actions can be done to restore power to the instrument bus.

A CORRECT: See above B INCORRECT: The first part is incorrect, plausible because the loss of the white instrument bus will not cause a reactor trip when greater than 10%. The second part is correct, as these are actions taken in AOP-0.2 because this loss will effect both charging and letdown.

C INCORRECT: The first part is incorrect, plausible because the loss of the white instrument bus will not cause a reactor trip when greater than 10%. The second part is correct, as 1Y-03 will be removed from service.

D INCORRECT: The first part is incorrect, plausible because the loss of the white instrument bus will not cause a reactor trip when greater than 10%. The second part is incorrect, plausible as this is an attachment entered when there will be a delay in the startup, and if the reactor startup has been halted due loss of NI instruments, it will be placed in a stable conditions, (ie, tripped) per OP 1B, until the restart can be commenced.

Learning Objective:

Given access to the site specific simulator or specific plant conditions, RESPOND to the following:

Loss of a DC bus Loss of an Instrumentation Bus (055.03.LP3456.002)

Tuesday, April 30, 2019 11:03:45 AM 41

QUESTIONS REPORT for 2019 NRC Exam Master

14. 2019 NRC 014/APE/058AA1.03/3.1/2-RI/RO/BANK/VENDOR/054.02.LP0123.005 Given the following:

Both Units are at Rated Thermal Power Fuses for D72-26-1, Power to DY-0A Red Alternate Inverter, blow Which of the following would be the expected position or indication on DY-0A following this loss of power?

DC Input Frequency AC Output Bypass Source Breaker Meter Breaker Supplying Load Light A. Open 60 Hz Closed Off B. Closed 60 Hz Closed On C. Closed 0 Hz Open On D. Open 0 Hz Open Off RO Tier 1 Group 1 Source: Bank Question History:

2005 PBNP NRC Question 15 K/A:

058AA1.03 Loss of DC Power Ability to operate and / or monitor the following as they apply to the Loss of DC Power: Vital and battery bus components (Imp 3.1/3.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the indications of a loss of power effects the DC system, if monitoring locally.

Cognitive Level:

Comprehension 2-RI: Requires the operator to understand the initial conditions, normal line up, determine what affect the loss of power will have, then determine what the indications are if locally monitoring the bus.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 42

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

SCI Inverter Tech Manual drawings Proposed reference to be provided to the applicants during examination:

None Original Question:

Breaker 1 on D-26 panel, power to DY0A, Red Swing Inverter, opens. Which of the following would be the expected position or indication on DY0A following this loss of power?

DC Input Breaker Frequency Meter AC Output Breaker A. Open 60 Hz Closed B. Closed 60 Hz Closed C. Closed 0 Hz Open D. Open 0 Hz Open Proposed Answer: B Tuesday, April 30, 2019 11:03:45 AM 43

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With this loss, the input and output breakers will stay shut, and the frequency will go to 60 Hz, as it is now coming from the non-safeguards 120 VAC backup power and the bypass source supplying load light would be lit.

A INCORRECT: The first part is incorrect, there is no input breaker trip on the DC breakers for the Red and Blue train, plausible if the operator confuses with other busses. The second and third parts are correct. The fourth part is incorrect, plausible if the operator has a misconception of how and where power is supplied from.

B CORRECT: See above C INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator does not correctly recall that backup AC power will supply the inverter. The third part is incorrect, plausible if the operator incorrectly determines, that with the loss of input power to the inverter, the output breaker would have also opened. The fourth part is correct.

D INCORRECT: The first part is incorrect, there is no input breaker trip on the DC breakers for the Red and Blue train, plausible if the operator confuses with other busses. The second part is incorrect, plausible if the operator does not correctly recall that backup AC power will supply the inverter. The third part is incorrect, plausible if the operator incorrectly determines, that with the loss of input power to the inverter, the output breaker would have also opened. The fourth part is incorrect, plausible if the operator has a misconception of how and where power is supplied from.

Learning Objective:

IDENTIFY and DESCRIBE the local controls, alarms, and indications associated with the Instrument Bus Electrical System, including:

a. Location and function of component and/or system operating controls and control stations

b. Alarming locations and response to major system and component alarms

c. Plant, system, and component conditions or permissives required for local operation Setpoints associated with major system alarms and/or interlocks (054.02.LP0123.005)

Tuesday, April 30, 2019 11:03:45 AM 44

QUESTIONS REPORT for 2019 NRC Exam Master

15. 2019 NRC 015/APE/0656AA2.01/2.9/2-RI/RO/MODIFIED/ARB/051.06.LP0086.011 Given the following:

Units are at Rated Thermal Power P-32A, P-32B,and P-32D Service Water pumps are running The following annunciators are received:

C01 A 1-5, SERVICE WATER STRAINERS DELTA P HIGH C01 A 2-5, NORTH OR SOUTH SERVICE WATER HEADER STRAINERS C01 A 3-5, NORTH OR SOUTH SERVICE WATER HEADER PRESSURE LOW C02 D 3-6, G-01 EMER DIESEL COOLER LOW FLOW C02 F 3-1, G-02 EMER DIESEL COOLER LOW FLOW 2C20 B 3-5, UNIT 2 TURBINE BUILDING SUMP LEVEL HIGH Which of the following indicates the cause of these alarms?

A. There is a leak in the North Service Water Header B. There is a leak in the South Service Water Header C. The Unit 1 Turbine Building Zurn Strainer is clogged D. The South Service Water Main Zurn Strainer is clogged Tuesday, April 30, 2019 11:03:45 AM 45

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Modified Question History:

2007 PBNP NRC Question 15 K/A:

062AA2.01 Loss of Nuclear Service Water Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: Location of a leak in the SWS (Imp 2.9/3.5)

Justification for K/A Match:

Matches the K/A by having the requiring the operator to determine not only the cause of the indications, but also location.

Cognitive Level:

Comprehension 2-RI: Requires the operator to understand the initial conditions, normal line up, determine what affect the loss of power will have, then determine what the indications are if locally monitoring the bus.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:45 AM 46

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ARB C01 A 1-5, Service Water Strainers Delta P High Rev 8 ARB C01 A 2-5, North Or South Service WaterHeader Strainers Rev 4 ARP C01 A 3-5, North Or South Service Water Header Pressure Low Rev 1 Proposed reference to be provided to the applicants during examination:

None Original Question:

Units 1 and 2 are at 100% power. Service Water Pumps P-32A, B and F are running. The following annunciators are then received:

- "Service Water Strainers delta P High"

- "North or South Service Water Header Pressure Low"

- "G01 Emergency Diesel Cooler Low Flow"

- "G02 Emergency Diesel Cooler Low Flow"

- "Unit 1 Turbine Building Sump Level High" Which of the following indicates the cause of these alarms and the appropriate remedial action?

(OI-70 is Service Water System Operation)

(AOP-9A is Service Water System Malfunction)

A. The Unit 1 Turbine Building Zurn Strainer is clogged; use OI-70 to backwash the strainer.

B. The South Service Water Main Zurn Strainer is clogged; use OI-70 to backwash the strainer.

C. There is a leak in the North Service Water Header, use AOP-9A to isolate the leak.

D. There is a leak in the South Service Water Header, use AOP-9A to isolate the leak.

Proposed Answer: D Tuesday, April 30, 2019 11:03:45 AM 47

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given all indication, the leak is in the south header. A leak would cause both G0-1 and G0-2 low flow alarms, as well as the Zurn strainers and header low pressure alarms. The sump alarm is what will indicate the location of the leak.

A CORRECT: See above B INCORRECT: As most indication would be found during a leak, the unit 1 turbine hall sump level would not indicate the north heater, but the south header. Plausible as the operator may confuse the headers with the location in the plant.

C INCORRECT: A Zurn strainer could account for several of the indications, but not all, specifically low flow on both emergency diesels, and low header pressure. Plausible if the operator has a misconception of the indications of issues with the Zurn strainers which is located in the south header. The turbine hall Zurn strainer does not impact service water flow to the EDGs.

D INCORRECT: A Zurn strainer could account for several of the indications, but not all. I would account for one of the low flow on the emergency diesels, but would cause the sump alarm. Plausible if the operator has a misconception of the indications of issues with the Zurn strainers as this is the south header Zurn strainer.

Learning Objective:

IDENTIFY and DESCRIBE the Control Room controls, alarms, and indications associated with the Service Water System, including:

a. Location and function of component and/or system operating controls and control stations

b. Alarming locations and response to major system and component alarms

c. Plant, system, and component conditions or permissives required for Control Room operation

d. Setpoints associated with major system alarms and/or interlocks (051.06.LP0086.011)

Tuesday, April 30, 2019 11:03:45 AM 48

QUESTIONS REPORT for 2019 NRC Exam Master

16. 2019 NRC 016/SYS/077AG2.1.32/3.8/2-DR/RO/NEW/AOP-0.1/055.03.LP2440.002 Given the following:

Both units are at Rated Thermal Power Grid frequency is 58.2 Hz and LOWERING at 0.05 Hz/min Bus Section 2 and 4 are at 351 KV and LOWERING SLOWLY Per AOP-0.1, Declining Frequency on 345KV Distribution System, which of the following correctly states:

(1) The actions to be implemented by PBNP for BOTH units AND (2) Why?

A. (1) Enter AOP-17A, shed 30% load in 10% increments (2) Limit main generator overheating from operating at a lowered frequency B. (1) Enter AOP-17A, shed 30% load in 10% increments (2) Prevent exceeding RCS Pressure Safety Limits from multiple dropped rods caused by operating Rod Drive MGs at a lowered frequency C. (1) Manually trip, go to EOP-0, Reactor Trip or Safety Injection, stabilize the units, trip all RCPs (2) Limit main generator overheating from operating at a lowered frequency D. (1) Manually trip, go to EOP-0, Reactor Trip or Safety Injection, stabilize the units, trip all RCPs (2) Prevent exceeding RCS Pressure Safety Limits from multiple dropped rods caused by operating Rod Drive MGs at a lowered frequency Tuesday, April 30, 2019 11:03:45 AM 49

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: New Question History:

None K/A:

077AG2.1.32 Generator Voltage and Electric Grid Disturbances Ability to explain and apply system limits and precautions.

(Imp 3.8/4.0)

Justification for K/A Match:

Matches the K/A by having the operator determine what actions and reasons for those actions when limits have been reached during a grid disturbance.

Cognitive Level:

Comprehension 2-RI: Requires the operator to understand the initial conditions, apply the procedure to determine what steps are necessary to be taken, and the reason why those steps are necessary.

10 CFR Part 55 Content:

55.41 10 55.43 2

Reference:

AOP-0.1, Declining Frequency on 354KV Distribution System Rev 18 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

With a declining voltage and frequency, AOP-0.1 will be entered. This procedure is divided in to 3 parts prior to a probable grid collapse, which are; greater than 59.35 Hz, between 59.35 and 58.3 Hz, and less than 58.3 Hz. For 58.3 or less, trip the reactor, start diesel generator after divorcing the plant from the gird, tripping reactor coolant pumps, shutting MSIVs, etc. With the initial conditions of frequency still lowering, these actions will be performed. The automatic protection scheme prevents motor damage during a frequency anomaly by removing the power supply prior to the onset of insulation overheating degradation and failure. However, the extent of rapid heatup of the main generators at PBNP is unknown. This potential for damage is avoided by manual separation of the main generators from the grid if frequency remains greater than the 86/TG01-02 setpoint of 56.07 Hz.

A INCORRECT: The first part is incorrect, plausible, because there are 3 Tuesday, April 30, 2019 11:03:45 AM 50

QUESTIONS REPORT for 2019 NRC Exam Master interlocked 10% load reductions which will occur automatically during a grid disturbance where frequency keeps lowering.

These load reductions will be performed by ATC, not PBNP, and the operator may have the misconception that the 10%

load reductions will be performed by PBNP. The second part is correct, as this is a concern during a declining frequency/grid disturbance.

B INCORRECT: The first part is incorrect, plausible, because there are 3 interlocked 10% load reductions which will occur automatically during a grid disturbance where frequency keeps lowering.

These load reductions will be performed by ATC, not PBNP, and the operator may have the misconception that the 10%

load reductions will be performed by PBNP. The second part is incorrect, plausible as multiple rods may drop with a lowering of frequency, but this should cause pressure to lower not rise.

The operator may have the misconception that the pressure safety limit is linked to the reactor core SL, which is a concern during lowering frequency due to lowering flow, not pressure.

C CORRECT: See above D INCORRECT: The first part is correct. The second part is incorrect, plausible as multiple rods may drop with a lowering of frequency, but this should cause pressure to lower not rise. The operator may have the misconception that the pressure safety limit is linked to the reactor core SL, which is a concern during lowering frequency due to lowering flow, not pressure.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following conditions:

Turbine Generator Voltage Regulator failure Loss of Main Generator Hydrogen pressure Total collapse of 345 KV system frequency Loss of electrical buses (055.03.LP2440.002)

Tuesday, April 30, 2019 11:03:45 AM 51

QUESTIONS REPORT for 2019 NRC Exam Master

17. 2019 NRC 017/EPE/E04G2.4.35/3.8/1-P/RO/NEW/ECA-1.2/031.02.LP0465.008 Given the following:

Unit 1 heating up from a refueling outage OP-1A, Cold Shutdown to Hot Standby has just been completed A LOCA outside containment occurs The crew is now in ECA-1.2, LOCA Outside Containment based on indications Which of the following correctly identifies actions to be taken in CONTAINMENT per ECA-1.2 for . . .

(1) A valve that will require local operator action to locally shut if manual operations from control is unsuccessful?

AND (2) The indication used to determine if this valve operation successfully isolated the LOCA outside containment?

A. (1) 1RH-720, RHR Return to RC MOV (2) Rising RCS pressure B. (1) 1RH-720, RHR Return to RC MOV (2) Rising RHR Pump Discharge Pressure C. (1) 1CV-313A, 1P-1A/B RCP #1 Seal Water Return Isolation (2) Rising RCS pressure D. (1) 1CV-313A, 1P-1A/B RCP #1 Seal Water Return Isolation (2) 1P-1A RCP Labyrinth Seal P zero or negative Tuesday, April 30, 2019 11:03:45 AM 52

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: New Question History:

None K/A:

E04G2.4.35 LOCA Outside Containment Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

(Imp 3.8/4.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall which valves are directed to be locally operated by the procedure, and what indications are used to determine whether or not the leak has been isolated.

Cognitive Level:

Comprehension 1-P: Requires the operator to recall which valves are directed to be operated, and what indication will be used to determine if the LOCA outside the containment has been isolated.

10 CFR Part 55 Content:

55.41 10 55.43 5

Reference:

ECA-1.2, LOCA Outside Containment Unit 1 Rev 25 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:45 AM 53

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The procedure directs the verification of a proper valve lineup for control, and if valves cannot be manually shut, to locally shut the valves. 1RH-720, position can be verified. ECA-1.2 used rising RCS pressure to verify the leak outside containment has been isolated.

A CORRECT: See above B INCORRECT: The first part is correct. The second part is incorrect, but plausible since the operator may conclude that the leak is isolated from RHR pump discharge pressure, which would be incorrect indication for leak isolation determination.

C INCORRECT: The first part is incorrect, plausible based on this valve being checked during a containment isolation actuation, with actions to shut the valve if found open. The second part is correct.

D INCORRECT: The first part is incorrect, plausible based on this valve being checked during a containment isolation actuation, with actions to shut the valve if found open. The second part is incorrect, plausible as the labyrinth seal P would go to zero if the leak was on this line and isolated by shutting this valve, but would not be used to verify the leak is isolated per procedure.

Learning Objective:

Given appropriate conditions/parameters and access to the site specific Simulator, IMPLEMENT the following procedures for the specified conditions:

a. ECA-1.1 to respond to a loss of Containment Sump Recirculation

b. ECA-1.2 to respond to an intersystem LOCA

c. ECA-1.3 to respond to containment sump blockage

d. ECA-2.1 to respond to both Steam Generators being faulted (031.02.LP0465.008)

Tuesday, April 30, 2019 11:03:45 AM 54

QUESTIONS REPORT for 2019 NRC Exam Master

18. 2019 NRC 018/EPE/E05EK3.1/3.4/2-DR/RO/BANK/BG-CSP-H.1/043.03.LP1998.007 Given the following:

A Feedwater transient resulted in a reactor trip The crew entered CSP-H.1 Response to Loss of Secondary Heat Sink, when all AFW flow was lost 1LT-460, SG A WR level is 150 inches and LOWERING SLOWLY 1LT-470, SG B WR level is 75 inches and LOWERING SLOWLY The crew has just secured all Reactor Coolant pumps The OATC notes RCS system pressure is RISING Which answers the following:

(1) Why have Reactor Coolant pumps been secured?

AND (2) Why is RCS system pressure rising?

A. (1) To begin RCS feed and bleed (2) Due to RCS temperature rise B. (1) To minimize heat input (2) Due to letdown being secured C. (1) To begin RCS feed and bleed (2) Due to letdown being secured D. (1) To minimize heat input (2) Due to RCS temperature rise Tuesday, April 30, 2019 11:03:45 AM 55

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 1 Source: Bank Question History:

2008 Catawba Question 16 K/A:

E05EK3.1 Loss of Secondary Heat Sink Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink): Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

(Imp 3.4/3.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine why RCP were secured, and the reason for the effects on pressure seen after this transient on the plant.

Cognitive Level:

Comprehension 2-DR: Requires the operator to understand the initial conditions, and where in the procedure the crew is currently, to determine why RCP were secured, and the reason for the effects on pressure after the RCPs are secured.

10 CFR Part 55 Content:

55.41 5 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 56

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

CSP-H.1, Response to Loss of Secondary Heat Sink Unit 1 Rev 44 BG-CSP-H.1, Background Response to Loss of Secondary Heat Sink Unit 1 Rev 30 Proposed reference to be provided to the applicants during examination:

None Original Question:

A Feedwater transient resulted in a reactor trip and the operating crew entered EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink) when all Auxiliary Feedwater flow was lost. Given the following:

S/G 1A wide range level - 31%

S/G 1B wide range level - 20%

S/G 1C wide range level - 23%

S/G 1D wide range level - 28%

The BOP has just secured all the NC pumps The OATC notes NC system pressure is increasing

1. Why have NC pump been secured?

2. Why is NCS pressure increasing?

A. 1. To begin NCS bleed and feed

2. Due to NC temperature increase B. 1. To minimize heat input

2. Due to letdown being secured C. 1. To begin NCS bleed and feed

2. Due to letdown being secured D. 1. To minimize heat input

2. Due to NC temperature increase Proposed answer: D Tuesday, April 30, 2019 11:03:45 AM 57

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

CSP-H.1 has two times where the RCPs are secured. The first is the CAS step to start feed and bleed when the conditions are met, and the second it when SG level is above the feed and bleed criteria, and you are now attempting to utilize the feed and condensate system to feed the steam generators. Based on SG level, this would be the second time and it is done to minimize heat up and to extnd effectiveness of the water in the SG.

When the RCPs are secured, this will cause an interim plant transient on RCS pressure and temperature as natural circulation flow conditions are established in the RCS.

A INCORRECT: The first part in incorrect, plausible as this is one of the two reasons for securing the RCPs in CSP-H.1. The second part is correct.

B INCORRECT: The first part is correct. The second part is incorrect, plausible as securing letdown maybe seen as the cause of the transient on the RCP causing the pressure rise.

C INCORRECT: The first part in incorrect, plausible as this is one of the two reasons for securing the RCPs in CSP-H.1. The second part is incorrect, plausible as securing letdown maybe seen as the cause of the transient on the RCP causing the pressure rise.

D CORRECT: See above.

Learning Objective:

APPRAISE and PRIORITIZE each operator-initiated recovery technique in its ability to restore the Heat Sink Critical Safety Function.

(043.03.LP1998.007)

Tuesday, April 30, 2019 11:03:45 AM 58

QUESTIONS REPORT for 2019 NRC Exam Master

19. 2019 NRC 019/APE/003AG2.4.45/4.1/3-SPK/RO/NEW/AOP-6A/SD86.4.2.4.45 Given the following:

Unit 1 is at Rated Thermal Power Control Rod Bank Selector switch is in AUTO One Rod Bottom light is LIT The following annunciators are LIT:

1C04 1A 1-5, ROD BOTTOM ROD DROP 1C04 1A 1-7, AUTOMATIC ROD MOTION 1C04 1A 3-3, POWER RANGE CHANNEL DEVIATION 1C04 1A 3-5, POWER RANGE OVERPOWER ROD STOP 1C04 1A 4-2, POWER RANGE HIGH SETPOINT CHANNEL ALERT 1C04 1A 4-5, POWER RANGE ROD DROP 1C04 1A 4-8, OVERTEMP T AUTO TURBINE RUNBACK 1C04 1C 1-2, PRESSURIZER PRESSURE HIGH OR LOW Control Rod Bank D Demand Step Counter reads 220 Which of the following procedures will be entered to address these conditions?

A. EOP-0, Reactor Trip or Safety Injection B. AOP-6A, Dropped Rod C. AOP-6B, Stuck or Misaligned Control Rod D. AOP-6C, Uncontrolled Motion of RCCA(S)

Tuesday, April 30, 2019 11:03:45 AM 59

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 2 Source: New Question History:

None K/A:

003AG2.4.45 Dropped Control Rod Ability to prioritize and interpret the significance of each annunciator or alarm.

(Imp 4.1/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to note the plant conditions and annunciators to assess the plant conditions to determine what has occurred and what procedure will be entered to address the plant conditions.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to understand what events have happened based on plant conditions and annunciators received, determine what procedure will be entered to address those conditions.

10 CFR Part 55 Content:

55.41 10 55.43 5

Reference:

ARB 1C04 1A 1-5, Rod Bottom Rod Drop Rev 4 AOP-6A, Dropped Rod Rev 20 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:45 AM 60

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the plant conditions, a single dropped rod has occurred. Annunciators 1C04 1A 3-5, Power Range Overpower Rod Stop and 1C04 1A 4-2, Power Range High Setpoint Channel Alert indicate that a rod closer to the outside of the core than the center has dropped. Annunciators 1C04 1A 3-5, power range overpower rod stop and 1C04 1A 4-2, Power Range High Setpoint Channel Alert indicate that rod motion will not occur. Given these indications, only one rod has dropped, and the procedure to be entered would be AOP-6A, Dropped Rod.

A INCORRECT: Plausible, because if two rods where to have dropped, then actions would direct entry into to EOP-0.

B CORRECT: See above.

C INCORRECT: Plausible which a rod drop at the center of the core, this would be the correct procedure to enter, then a transition would be done to AOP-6A.

D INCORRECT: Plausible as the drop rod could be considered misaligned, and the auto turbine runback alarm indicates that a power change is happening, but the other annunciators indicated that rod motion has been blocked.

Learning Objective:

Ability to prioritize and interpret the significance of each annunciator or alarm.

(SD86.4.2.4.45)

Tuesday, April 30, 2019 11:03:45 AM 61

QUESTIONS REPORT for 2019 NRC Exam Master

20. 2019 NRC 020/APE/005AA2.03/3.5/2-DR/RO/BANK/EOP-0.1/032.02.LP0405.003 Given the following:

An inadvertent reactor trip has just occurred Both reactor trip breakers indicate open Control rod K-7 IRPI reads 220 steps, its rod bottom light is NOT lit Control rod L-6 IRPI reads 30 steps, its rod bottom light is NOT lit Control rod G-3 IRPI reads 15 steps, its rod bottom light is LIT The current procedure in effect is EOP-0.1, Reactor Trip Response Which of the following describes the amount of boration required for these conditions?

A. No boration is required B. 2825 gal C. 5650 gal D. 8475 gal Tuesday, April 30, 2019 11:03:45 AM 62

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 2 Source: Bank Question History:

2003 PBNP Question 20 K/A:

005AA2.03 Inoperable/Stuck Control Rod Ability to determine and interpret the following as they apply to the Inoperable /

Stuck Control Rod: Required actions if more than one rod is stuck or Inoperable (Imp 3.5/4.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what determines a stuck/inoperable rod, and what needs to be accomplished procedurally when there is more than one stuck/inoperable rod.

Cognitive Level:

Comprehension 2-DR: Requires the operator to understand the initial conditions, determine which rods are considered not fully inserted, then recall what actions must be performed given that determination.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:45 AM 63

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-0.1, Reactor Trip Response Rev 46 BG-EOP-0.1, Background Reactor Trip Response Rev 32 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following Unit 1 plant conditions:

- An inadvertent reactor trip has just occurred.

- Both reactor trip breakers indicate open.

- Control rod K-7 IRPI reads 220 steps, its rod bottom light is NOT lit.

- Control rod L-6 IRPI reads 35 steps, its rod bottom light is NOT lit.

- Control rod G-3 IRPI reads 10 steps, its rod bottom light is lit.

The current procedure in effect is EOP-0.1, "Reactor Trip Response".

Which of the following describes the amount of boration required for these conditions?

A. No boration is required since the reactor trip breakers are open.

B. A 1200 gallon boration is required since only one control rod is considered not fully inserted.

C. A 2400 gallon boration is required since only two control rods are considered not fully inserted.

D. A 3600 gallon boration is required since three control rods are considered not fully inserted.

Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 64

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per BG-EOP-0.1, control rods are considered inserted when IRPI indicates less than 20 steps with reactor trip breaker open and the rod bottom light lit." In the initial conditions, there are 2 rods not fully inserted, K-7 and L-6, these do not meet the requirement to be called fully inserted per the background .

Per EOP-0.1, an emergency boration in needed if two or more rods are not fully inserted, and a boration of 2825 gal should be done for each rod not fully inserted. A total boration of 5650 gal A INCORRECT: Plausible, if the operator has a misconception of the definition of a fully inserted rods, and only determines one rod is not fully inserted.

B INCORRECT: Plausible if the operator has a misconception or incorrectly recalls the boration amount required. The step reads Borate 2825 gallons for each control rod not fully inserted. and thinks there is only one rod requiring boration.

C CORRECT: See above.

D INCORRECT: Plausible if the operator has a misconception of the amount required. The step reads Borate 2825 gallons for each control rod not fully inserted. and thinks there are three rods requiring boration.

Learning Objective:

Describe the major actions accomplished by each of the following procdures:

a. EOP-0, Reactor Trip or Safety Injection

b. EOP-0.0, Rediagnosis

c. EOP-0.1, Reactor Trip Response

d. EOP-1.1. SI Termination (031.02.LP0405.003)

Tuesday, April 30, 2019 11:03:45 AM 65

QUESTIONS REPORT for 2019 NRC Exam Master

21. 2019 NRC 021/APE/033AK1.01/2.7*/1-I/RO/BANK/PBN LP2416/053.03.LP2416.009 Given the following:

A reactor shutdown is in progress in accordance with OP-3B, Reactor Shutdown Compensating voltage is lost on one of the intermediate range detectors How does this failure affect the operation of the Nuclear Instrumentation System?

A. The intermediate range detector will be under compensated, which will cause the source range detectors to energize early.

B. The intermediate range detector will be under compensated, which will prevent the automatic energizing of the source range detectors.

C. The intermediate range detector will be over compensated, which will cause the source range detectors to energize early.

D. The intermediate range detector will be over compensated, which will prevent the automatic energizing of the source range detectors.

RO Tier 1 Group 2 Source: Bank Question History:

None K/A:

033AK1.01Loss of Intermediate Range Nuclear Instrumentation Knowledge of the operational implications of the following concepts as they apply to Loss of Intermediate Range Nuclear Instrumentation: Effects of voltage changes on performance (Imp 2.7/3.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what the effect is when a change in voltage happens to the intermediate range NR.

Cognitive Level:

Knowledge 1-I: Requires the operator to recall the effect of the loss of voltage on the IR NI and what this interlock is for SR NIs.

10 CFR Part 55 Content:

55.41 8 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 66

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

PBN LP2416 Nuclear Instruments Rev 6 883D195 Sh 11, NIS Trip Signal Logic Drawing Rev 7 883D195 Sh 12, NIS Permissives and Block Logic Drawing Rev 10 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A reactor shutdown is in progress in accordance with OP-3B, Reactor Shutdown Compensating voltage is lost on one of the intermediate range detectors How does this failure affect the operation of the Nuclear Instrumentation System?

A. The intermediate range detector will be under compensated, which will cause the source range detectors to energize early.

B. The intermediate range detector will be under compensated, which will prevent the automatic energizing of the source range detectors.

C. The intermediate range detector will be over compensated, which will cause the source range detectors to energize early.

D. The intermediate range detector will be over compensated, which will prevent the automatic energizing of the source range detectors.

Proposed answer: B Tuesday, April 30, 2019 11:03:45 AM 67

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

A loss of voltage will cause the IR NI to become undercompensated, which in turn will cause it to read higher than indicated power level, due to the gamma radiation no longer being removed or compensated for. This in turn will prevent the automatic energizing of the SR NIs, but they can be manually energized.

A INCORRECT: The first part is correct. The second part it incorrect, plausible if the operator has the misconception that being undercompensated will cause the detector to read lower than actual power level.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, plausible if the operator has the misconception that being a loss of voltage will cause the meter to be overcompensated. The second part is incorrect, plausible if the operator has the misconception that being overcompensated by a loss of voltage will cause the detector to read lower than actual power level.

D INCORRECT: The first part is incorrect, plausible if the operator has the misconception that being a loss of voltage will cause the meter to be overcompensated. The second part is correct.

Learning Objective:

DESCRIBE the Nuclear Instrumentation System response to the following events/malfunctions:

a. Core Voiding

b. Overcompensation

c. Undercompensation

d. Loss of Instrument Power

e. Loss of Control Power

f. Loss of Compensating Power (053.03.LP2416.009)

Tuesday, April 30, 2019 11:03:45 AM 68

QUESTIONS REPORT for 2019 NRC Exam Master

22. 2019 NRC 022/APE/036AK2.01/2.9/1-P/RO/BANK/OI 53/112.01.LP0260.002 Given the following:

A fuel assembly is being transferred from the Spent Fuel Pool to Containment The Fuel Transfer Carriage stops in the transfer tube The CABLE OVERLOAD light illuminates and stays lit Which of the following is the initial method used for Carriage retrieval?

A. Use the Frame down switch to lower the upender frame to ensure interlocks are met.

B. Use the Manual upender handwheel to lower the upender frame to ensure interlocks are met.

C. Perform Emergency Retrieval using the Polar Crane and Emergency Pull cable to Containment side.

D. Release the brake, use the Manual Carriage drive handwheel to position the carriage at the pool side.

RO Tier 1 Group 2 Source: Bank Question History:

None K/A:

036AK2.01 Fuel Handling Incidents Knowledge of the interrelations between the Fuel Handling Incidents and the following: Fuel handling equipment (Imp 2.9./3.5)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the method used utilizing fuel handling equipment during a fuel handling incident.

Cognitive Level:

Knowledge 1-P: Requires the operator to recall the initial method used to for carriage retrieval during a fuel handling incident.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 69

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

OI 53, Positioning of the Fuel Transfer Cart Rev 25 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A fuel assembly is being transferred from the Spent Fuel Pool to Containment The Fuel Transfer Carriage stops in the transfer tube The CABLE OVERLOAD light illuminates and stays lit Which of the following is the initial method used for Carriage retrieval?

A. Use the Frame down switch to lower the upender frame to ensure interlocks are met.

B. Use the Manual upender handwheel to lower the upender frame to ensure interlocks are met.

C. Perform Emergency Retrieval using the Polar Crane and Emergency Pull cable to Containment side.

D. Release the brake, use the Manual Carriage drive handwheel to position the carriage at the pool side.

Proposed answer: D Tuesday, April 30, 2019 11:03:45 AM 70

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With the cable overload light lit, manual handwheel operation is the preferred method. This method consists of releasing the brake, removing the winch drive motor cover, installing the cart handwheel on the winch drive motor and hand cranking the carriage to the desired position.

A INCORRECT: This is not the initial method, plausible if the operator has the misconception that the cable overload light is a protection for the upender, as this is in the procedure to perform if the upender interlocks are not met, but the cable overload interlock is not an upender interlock.

B INCORRECT: This is not the initial method, plausible if the operator has the misconception that the cable overload light is a protection for the upender, as this is in the procedure to perform if the upender interlocks are not met and is a backup in the automatic feature is not working, as would be the case if the overload light was an upender protection, but the cable overload interlock is not an upender interlock.

C INCORRECT: This is not the initial method, as the PAB crane is used in an emergency pull cable method to move the carriage. Plausible if the operator has the misconception that since the carriage is going to the containment, the polar crane would be used in the emergency pull cable method.

D CORRECT: See above.

Learning Objective:

IDENTIFY and DISCUSS the purpose, operation and utilization of the following IAW plant procedures:

a. Fuel Transfer Conveyor

b. Upender

c. New and Spent Fuel Handling Tools

d. Burnable Poison Rod Handling Tool Portable RCCA Change Tool (112.01.LP0260.002)

Tuesday, April 30, 2019 11:03:45 AM 71

QUESTIONS REPORT for 2019 NRC Exam Master

23. 2019 NRC 023/APE/037AK3.08/4.1/3-SPK/RO/BANK/AOP-3/055.03.LP2438.004 What is the reason for maintaining SG level above the top of the U Tubes in the affected steam generator during a Steam Generator Tube Leak?

A. To minimize erosion of the break from steam flow.

B. To prevent thermal shock to the tubes during RCS cooldown.

C. To prevent a rapid depressurization of the affected SG during RCS cooldown.

D. To ensure that the pressure and temperature limits of the SG shell are maintained.

RO Tier 1 Group 2 Source: Bank Question History:

None K/A:

037AK3.07 Steam Generator (S/G) Tube Leak Knowledge of the reasons for the following responses as they apply to the Steam Generator Tube Leak: Actions contained in EOP for S/G tube leak.

(Imp 4.2/4.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the required actions contained the EOP network for a depressurization during a SGTL when combined with a loss of subcooling.

Cognitive Level:

Knowledge 1-P: Recall the reason for the action of maintaining SG level in the SG with a Steam Generator Tube Leak.

10 CFR Part 55 Content:

55.41 5 10 55.43 Tuesday, April 30, 2019 11:03:45 AM 72

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-3, Steam Generator Tube Leak Unit 1 Rev 13 BG AOP-3, Background Steam Generator Tube Leak Rev 11 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 2 was tripped due to a SG tube rupture of approximately 150 GPM on SG 'B' The operating crew has entered EOP-3, Steam Generator Tube Rupture Operators are addressing the step concerning isolation of feed flow to SG

'B' What is the reason for maintaining SG 'B' level above the top of the U Tubes?

A. To provide partitioning of Iodine from the RCS B. To prevent thermal shock to the tubes during RCS cooldown.

C. To prevent a rapid depressurization of the 'B' SG during RCS cooldown.

D. To ensure that the pressure and temperature limits of the SG shell are maintained.

Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 73

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Affected steam generator level is checked for two reasons; To reduce feed flow to the affected steam generators to minimize the potential for steam generator overfill, and to establish and maintain a water level in the affected steam generators above the top of the U-tubes in order to promote thermal stratification to prevent affected steam generator depressurization.

A INCORRECT: Plausible because flow may cause the break area to erode and become worse, but is not the reason for maintaining level, but depressurizing the RCS.

B INCORRECT: Plausible because prevention of thermal shock is a valid concern of not wanting to exacerbate the tube leak, but is not the reason for maintaining water level.

C CORRECT: See above.

D INCORRECT: Plausible because maintaining pressure / temperature limits is a valid concern of not wanting to exacerbate the tube leak, but is not the reason for maintaining water level.

Learning Objective:

DESCRIBE the plant and operator(s) response to the following conditions:

a. Failure of Pressurizer Pressure and/or Level Control system

b. Reactor Coolant System leak

c. Reactor Coolant Pump malfunction Steam Generator Tube leak (055.03.LP2438.001)

Tuesday, April 30, 2019 11:03:45 AM 74

QUESTIONS REPORT for 2019 NRC Exam Master

24. 2019 NRC 024/APE/067AA1.05/3.0/1-I/RO/NEW/TS 78/052.01.LP0003.008 Given the following:

Both units are at Rated Thermal Power W-13A1, Cable Spreading Room Recirc fan is RUNNING W-13A2, Cable Spreading Room Recirc fan is in AUTO A fire causes a Halon Activation in Zone 1, Cable Spreading Room Which of the following describes the indications observed in the control room for W-13A1 and W-13A2, Cable Spreading Room Recirc fans after the Halon Activation?

(Assume no operator action)

A. ONLY W-13A1 will be running B. W-13A1 AND W-13A2 will be secured C. BOTH W-13A1 and W-13A2 will start on low flow after a 10 minute time delay D. W-13A1 will be secured AND W-13A2 will start on low flow after a 10 minute time delay RO Tier 1 Group 2 Source: New Question History:

None K/A:

067AA1.05 Plant fire on site Ability to operate and / or monitor the following as they apply to the Plant Fire on Site: Plant and control room ventilation systems (Imp 3.0/3.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to monitor the indications in the control room after a plant fire on site, by identifying the ventilation system fans running.

Cognitive Level:

Knowledge 1-I: Requires the operator to recall the interlock for the cable spreading room ventilation during a plant fire on site.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 75

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 78, Semi-Annual Halon 1301 Fire Suppression System Surveillance Test Rev 34 TS 9, Control Room Heating and Ventilation System Monthly Checks Rev 48 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The operator will have to recall the normal lineup for fan operation. On an actuation of the Halon system for Zone 1, Cable Spreading Room, both of the recirc will stop, and neither fan will start. With no fire present, if one fan stopped, the other would start on a low flow condition.

A INCORRECT: Plausible if the operator has a misconception of the operation of the system and assumes that W-13A1 was running and would stay running to ensure halon is distributed throughout the area since the recirc fans only recirc the air in the cable spreading room, and one fan running would ensure distribution of the halon.

B CORRECT: See above.

C INCORRECT: Plausible as there is a low flow interlock, if the fan wasnt secured by the fire detection system, and the 10 minute delay is plausible as no entry into the room is allowed for 10 minutes to ensure halon effectiveness procedurally.

D INCORRECT: Plausible as there is a low flow interlock that would start the stby fan, if the fan wasnt secured by the fire detection system, and the 10 minute delay is plausible as no entry into the room is allowed for 10 minutes to ensure halon effectiveness procedurally.

Learning Objective:

IDENTIFY and DESCRIBE the Control Room controls, alarms, and indications associated with the Fire Protection System, including:

a. Location and function of component and/or system operating controls and control stations

b. Alarming locations and response to major system and component alarms

c. Plant, system, and component conditions or permissives required for Control Room operation Setpoints associated with major system alarms and/or interlocks (052.01.LP0003.008)

Tuesday, April 30, 2019 11:03:45 AM 76

QUESTIONS REPORT for 2019 NRC Exam Master

25. 2019 NRC 025/APE/068AA2.07/4.1/2-DR/RO/NEW/AOP-40A/055.03.LP1275.001 Given the following:

Both units are at Rated Thermal Power The crew transitioned to AOP-40A, Control Room Abandonment Due to Fire All control room actions were completed prior to evacuation Which answers the following:

1) Where will PZR level be monitored?

AND

2) What is the guidance regarding maintaining PZR level and RCS subcooling?

A. 1) C-45, Alternate Shutdown Control Panel

2) Control charging to maintain greater than 20% pressurizer level and operate pressurizer heaters to maintain subcooling greater than 35°F subcooling B. 1) C-45, Alternate Shutdown Control Panel

2) Establish minimum charging and excess letdown, to maintain pressurizer level and after AFW flow is established to the steam generators, maintain subcooling using the steam dumps C. 1) 1N-11/2N04, Charging Pump/PZR Heater Local Control Station

2) Control charging to maintain greater than 20% pressurizer level and operate pressurizer heaters to maintain subcooling greater than 35°F subcooling D. 1) 1N-11/2N04, Charging Pump/PZR Heater Local Control Station

2) Establish minimum charging and excess letdown, to maintain pressurizer level and after AFW flow is established to the steam generators, maintain subcooling using the steam dumps Tuesday, April 30, 2019 11:03:45 AM 77

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 2 Source: New Question History:

None K/A:

068AA2.07 Control Room Evacuation Ability to determine and interpret the following as they apply to the Control Room Evacuation: PZR level (Imp 4.1/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine PZR level once the control room has be evacuated and recalling actions to maintain subcooling and PZR level.

Cognitive Level:

Comprehension 2-DR: Requires the operator to recall where PZR level is determined, and and the relationship between subcooling and PZR level to determine the necessary actions to take.

10 CFR Part 55 Content:

55.41 55.43 5

Reference:

AOP-40A, Control Room Abandonment Due to Fire Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:45 AM 78

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

AOP-40A, differs from AOP-10, the difference being AOP-40A is control room evacuation due to a fire, while AOP-10 is due to other than fire. The both procedures are written to maintain the plant in a safe condition, AOP-10 has a goal of hot shutdown, and AOP-40A is maintaining the plant in a safe condition while mitigating the consequences of the fire. In AOP-40A maintaining pressurizer level greater than 20%is the goal, and if greater than 12% then maintain subcooling through the use of pressurizer heaters.

A INCORRECT: The first part is incorrect, plausible as this is used in several places in the procedure, for charging power, SW pumps etc.

The second part is incorrect, plausible because both methods would be successful, the establishment of minimum charging coupled with excess letdown, would control pressurizer level, and with AFW flow established, the SG inventory would be sufficient to use the steam dumps. Both of which are accomplished in AOPs.

B INCORRECT: The first part is incorrect, plausible as this is used in several places in the procedure, for charging power, SW pumps etc.

The second part is correct.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible because both methods would be successful, the establishment of minimum charging coupled with excess letdown, would control pressurizer level, and with AFW flow established, the SG inventory would be sufficient to use the steam dumps. Both of which are accomplished in AOPs.

Learning Objective:

DESCRIBE the major actions of AOP-10 and the AOP-40 series procedures to include:

a. Actions taken prior to abandoning the control room

b. Controlling equipment locally to maintain both units in a safe shutdown condition

c. Opening and/or closing MCC load breakers

d. Opening and/or closing 480V or 4160V breakers

e. Transferring breaker control power (055.03.LP1275.001)

Tuesday, April 30, 2019 11:03:45 AM 79

QUESTIONS REPORT for 2019 NRC Exam Master

26. 2019 NRC 026/APE/069AG2.2.44/4.2/2-DR/RO/NEW/TS 3.6.2/057.02.LP3342.002 Given the following:

Unit 1 is in MODE 4 recovering from a refueling outage The following annunciators are NOT lit:

1C20 A 1-1, UNIT 1 CONTAINMENT UPPER HATCH OUTER DOOR OPEN 1C20 A 2-1, UNIT 1 CONTAINMENT UPPER HATCH INNER DOOR OPEN The Designated Airlock Operator is stationed RP is entering the containment for survey purposes The Designated Airlock Operator contacted the control room and has been granted permission for RP to enter containment 1C20 A 1-1, is now LIT Which of the following identifies the status of LCO 3.6.2., Containment Air Locks and required actions?

A. NOT met Verify bulkhead door and equalizing valves are closed on the operable bulkhead in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. NOT met Initiate action to evaluate overall containment leakage rate per LO 3.6.1 IMMEDIATELY C. NOT met Verify bulkhead door and equalizing valves are closed on both doors in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND initiate action to evaluate overall containment leakage rate per LO 3.6.1 IMMEDIATELY D. MET Annunciator is expected and a qualified Containment Airlock Operator stationed as required Tuesday, April 30, 2019 11:03:45 AM 80

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 2 Source: New Question History:

None K/A:

069 AG2.2.44 Loss of Containment Integrity Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(Imp 4.2/4.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to interpret control room indication, to determine the status of a system and then determine if that status meets the licensing directives and if actions are necessary.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to determine the status of the annunciators, determine the status of containment, and then what actions are necessary during an entry.

10 CFR Part 55 Content:

55.41 5 55.43 5

Reference:

1-SOP-CONT-001, Containment Airlock Operation and Entry, Rev 21, Section 3.5 TS 3.6.2., Containment Systems - Containment Air Locks, Rev 1 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:45 AM 81

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per 1-SOP-CONT-001, the annunciators have a blue dot because they are expected and will be nuisance alarms due to the operation of the hatch. If the annunciators are in a cleared condition (not lit), the first time they are activated they will light. With permission being granted to enter the containment, and operation of the outer door, the TSAC will no longer be met, because this will not be entering to perform airlock repairs, therefore the LCO is not met, and condition A will be entered A CORRECT: See above.

B INCORRECT: The first part is correct. The second part is plausible if the operator has a misconception for the actions needed when LCO 3.6.2 is not met, this is the first action for one or more containment air locks inoperable for reasons other than A or B.

C INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has a misconception for the actions needed when LCO 3.6.2 is not met, these are the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less action for one or more containment air locks inoperable for reasons other than A or B.

D INCORRECT: The first part is incorrect, but plausible if the operator has the misconception that with annunciators and containment hatch watch as procedurally required, the LCO is met, The second part is correct, these actions are procedurally required.

Learning Objective:

Given specific plant conditions, ASSESS and APPLY Technical Specification requirements as appropriate.

(057.02.LP3342.002)

Tuesday, April 30, 2019 11:03:45 AM 82

QUESTIONS REPORT for 2019 NRC Exam Master

27. 2019 NRC 027/EPE/E16EK2.1/3.0/1-P/RO/BANK/OM 3.7/031.02.LP0405.010 Given the following:

Unit 1 has experienced a Large Break LOCA Containment Pressure, Temperature, Humidity, and Radiation are all reading abnormally high due to the LOCA conditions The Reactor Operator has made the announcement the Unit 1 containment is adverse Which of the following describes the proper use of Adverse Containment setpoints?

Once the use of Adverse Containment setpoint has been implemented due to . . .

A. pressure, adverse values must be used for the duration of the event.

B. temperature, adverse values can be used when temperature lowers to a normal value.

C. humidity, adverse values can be used when humidity lowers to normal to a normal value.

D. radiation, if integrated dose is exceeded, adverse values must be used for the duration of the event.

Tuesday, April 30, 2019 11:03:45 AM 83

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 1 Group 2 Source: Bank Question History:

2009 Callaway Question 27 K/A:

E16EK2.1 High Containment Radiation Knowledge of the interrelations between the (High Containment Radiation) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

(Imp 3.0/3.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall how high containment radiation effect adverse containment (instrument readings), and how long that effect will last.

Cognitive Level:

Knowledge 1-P: Requires the operator to recall the initial method used to for carriage retrieval during a fuel handling incident.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:45 AM 84

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

OM 3.7, AOP and EOP Procedure Usage for Response to Plant Transients Rev 31 Section 6.13.5 Proposed reference to be provided to the applicants during examination:

None Original Question:

The Callaway Plant has experienced a large Loss of Coolant Accident (LOCA).

Containment Pressure, Temperature, Humidity, and Radiation are all reading abnormally high due to the LOCA conditions. The Reactor Operator has made the announcement the plant is now in Adverse Containment Which ONE of the following describes the proper use of Adverse Containment?

Once in Adverse Containment . . . .

A. Due to pressure, adverse values must be used for the duration of the event.

B. Due to temperature, adverse values can be used when temperature lowers to a normal value.

C. Due to humidity, adverse values can be used when humidity lowers to normal to a normal value.

D. Due to radiation, adverse values must be used for the duration of the event.

Proposed answer: D Tuesday, April 30, 2019 11:03:45 AM 85

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given an abnormally high pressure and radiation, adverse containment condition can be entered either due to pressure or radiation. Because one of the possibilities is radiation, then adverse containment will remain in effect for the duration of the event, or until it can be determined that the integrated dose is less than the setpoint, which would be determined by the TSC.

A INCORRECT: Plausible if the operator applies the requirement for relaxing adverse containment due to radiation to pressure. Adverse containment entered due to pressure will be exited after pressure reduces below the setpoint.

B INCORRECT: Plausible if the operator has the misconception that temperature determines adverse containment.

C INCORRECT: Plausible if the operator has the misconception that humidity determines adverse containment.

D CORRECT: See above.

Learning Objective:

RECOGNIZE the conditions requiring use of adverse Containment values and when the operator can return to using normal values.

(031.02.LP0405.010)

Tuesday, April 30, 2019 11:03:45 AM 86

QUESTIONS REPORT for 2019 NRC Exam Master

28. 2019 NRC 028/SYS/003K1.01/2.6*/2-DR/RO/BANK/AOP-1B/055.03.LP2438.004 Given the following:

Unit 1 is operating at Rated Thermal Power 1C04 1C 3-11, 1P-1A OR B RCP UPPER OR LOWER SUMP OIL LEVEL HIGH OR LOW is lit 1C04 1C 3-10, 1TR-2001 TEMPERATURE MONITOR is lit Unit 1 Component Cooling Water Surge Tank level is 49% and LOWERING The follow are 1P-1A, RCP indications:

Upper Thrust Bearing is 92°C and RISING Lower Thrust Bearing is 92°C and RISING Seal injection flow is 8 gpm and STABLE No. 1 seal leakoff is 2.6 gpm and STABLE Which of the following describes the required action and the reason for the action that would explain all of the above abnormal conditions?

A. The Seal Return Heat Exchanger must be bypassed because CCW is diluting the RCS through a leak in the heat exchanger.

B. Unit 1 Reactor must be tripped and 1P-1A, RCP stopped because its oil has been emulsified with CCW and has affected bearing lubrication.

C. Unit 1 Reactor must be tripped and 1P-1A, RCP stopped because an oil leak has caused low reservoir level, resulting in poor bearing lubrication.

D. The position of 1CC-761A, 1P-1A RCP Thermal Barrier Outlet AOV, must be checked to ensure it is shut because the thermal barrier is leaking.

Tuesday, April 30, 2019 11:03:45 AM 87

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2003 PBNP NRC Question 5 K/A:

003K1.01 Reactor Coolant Pump System (RCPS)

Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems: RCP lube oil (Imp 2.6/2.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand the physical connections between RCP lube oil, and CCW, and the effect of a CCW leak on the RCP lube oil system cause on the RCP and what actions would be necessary.

Cognitive Level:

Comprehension 2-RI: Requires the operator understand the physical connections between RCP lube oil, and CCW, understand the initial conditions determine what happened and what actions will be needed.

10 CFR Part 55 Content:

55.41 2 to 9 55.43 Tuesday, April 30, 2019 11:03:45 AM 88

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-1B, Reactor Coolant Pump Malfunction Unit 1 Rev 29 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 is operating at 100% power. The following indications are noted:

- Annunciator "1P-1A or B RCP Upper or Lower Sump Oil Level High or Low" is lit.

- Points 2 and 4 on recorder 1TR-2001 (1P-1A RCP Thrust Bearing Upper and Lower Shoe temperatures) are in alarm and are currently reading 92 ºC and rising.

- Unit 1 Component Cooling Water Surge Tank level is 49% and lowering.

- 1P-1A RCP seal injection flow is 6 gpm.

- 1P-1A RCP No. 1 seal leakoff is 1.2 gpm.

Which of the following describes the required action and the reason for the action that would explain all of the above abnormal conditions?

A. Unit 1 Reactor must be tripped and 1P-1A RCP stopped because oil has leaked out of the pump resulting in poor bearing lubrication.

B. The position of 1CC-761A, Thermal Barrier Outlet AOV, must be checked to ensure it is shut because the thermal barrier is leaking.

C. Unit 1 Reactor must be tripped and 1P-1A RCP stopped because its oil has been emulsified with CCW and this has affected bearing lubrication.

D. The Seal Return Heat Exchanger must be bypassed because CCW is diluting the RCS through a leak in the heat exchanger.

Proposed answer: C Tuesday, April 30, 2019 11:03:45 AM 89

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The given indication of oil sump high or low, combined with thrust bearing temperatures and a surge tank level lower, leads to the conclusion that, a CCW leak has occurred and is affecting the RCP lube oil system. CCW system water in the RCP lube oil system will cause an emulsification of the oil, which in turn will cause a rise in thrust bearing temperatures. The pump will be required to be tripped as it meets the RCP trip criteria for bearing temperature, and the reactor will need to be tripped prior to the pump being tripped based on power level.

A INCORRECT: Plausible as there is a leak out of the CCW system, and if the leak was located in this heat exchanger it would cause a dilution condition in the RCS.

B CORRECT: See above.

C INCORRECT: Plausible, as the actions are correct and the system affected is the RCP lube oil system, but this does not account for all of the indications in the question stem, such as the surge tank lowering.

D INCORRECT: Plausible if the operator determines this valve should have shut based on the indications, but leakage is causing the leak.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following events:

a. Reactor Coolant System leakage

b. Reactor Coolant Pump malfunctions

c. Steam Generator Tube leak (055.03.LP2438.004)

Tuesday, April 30, 2019 11:03:45 AM 90

QUESTIONS REPORT for 2019 NRC Exam Master

29. 2019 NRC 029/APE/003A4.05/3.1/3-SPR/RO/BANK/AOP-1B/055.03.LP2438.001 Given the following:

Unit 1 is at Rated Thermal Power The following annunciators on 1C03 1D are LIT 1-1, 1P-1A OR B RCP LABYR SEAL WATER INLET OR BEARING TEMP HIGH 1-2, 1P-1A RCP STAND PIPE LEVEL HIGH 2-1, 1P-1A OR B RCP LABYR SEAL P LOW 3-1, 1P-1A OR B RCP NO. 1 SEAL WATER OUTLET TEMPERATURE HIGH 3-2, 1P-1A RCP NO. 1 SEAL WATER FLOW HIGH OR LOW RDCT level is RISING approximately 10% per minute RCP indication on 1C04 are as follows:

Question continued on next page Question continued from previous page Tuesday, April 30, 2019 11:03:46 AM 91

QUESTIONS REPORT for 2019 NRC Exam Master

29. 2019 NRC 029/APE/003A4.05/3.1/3-SPR/RO/BANK/AOP-1B/055.03.LP2438.001 What is the status of the A RCP seal package?

A. #1 seal is failing only B. #2 seal is failing only C. #3 seal is failing only D. #1 and #2 seals are failing RO Tier 2 Group 1 Source: Bank Question History:

2017 PBNP Question 5 Previous 2 NRC Exams K/A:

003A4.05 Reactor Coolant Pump System (RCPS)

Ability to manually operate and/or monitor in the control room: RCP seal leakage detection instrumentation (Imp 3.1/3.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to use control room instrumentation, including seal leakage detection instrumentation to diagnose which seal(s) has failed on the reactor coolant pump.

Cognitive Level:

Comprehension 3-SPR: The operator must diagnose the seal package failure based on the plant indications.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

AOP-1B Unit 1, Reactor Coolant Pump Malfunction, Rev 29, Step Steps 4, 9, and 11, and Foldout Page Item 3 BG AOP-1D, Reactor Coolant Pump Malfunction Background Document, Rev 17, Steps 4, 9, and 11, and Foldout Page Item 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:46 AM 92

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The #1 and #2 seals have failed. The #1 seal is indicated by an increase in temperature on the RCP seal inlet and bearing temperature as well as the RCP No 1 seal outlet temp and the decrease in lab seal P. The temperatures would not increase if only #2 seal failed, nor would the lab seal P decrease. Also the annunciators 1-1, 2-1, 3-1, and 3-2 are indicative of a #1 seal failure, while 1-2 and 3-2 are indicative of a #2 seal failure. The low seal leakage is also a symptom of a #2 seal failure. So combining the symptoms, both seals #1 and

1. 2 have failed.

A INCORRECT: Plausible as the indication of an increase in temperatures, decrease in lab seal P are symptoms of a #1 seal failure, it does not explain the additional annunciators nor the low leakoff values.

B INCORRECT: Plausible as the indication of a lowering seal leakage, as well as annunciators 1-2 and 3-2 are symptoms of a #2 seal failure, it does not explain the additional annunciators nor the increased temperatures or decreased P.

C INCORRECT: Plausible as the stand pipe and leakoff symptoms could be confused with this failure as it is neither Seal #1 only or Seal #2 only.

D CORRECT: See above.

Learning Objective:

DESCRIBE the plant and operator(s) response to the following conditions:

a. Failure of Pressurizer Pressure and/or Level Control system

b. Reactor Coolant System leak

c. Reactor Coolant Pump malfunction

d. Steam Generator Tube leak (055.03.LP2438.001)

Tuesday, April 30, 2019 11:03:46 AM 93

QUESTIONS REPORT for 2019 NRC Exam Master

30. 2019 NRC 030/SYS/004K4.12/3.1/3-PEO/RO/BANK/LP0079/051.02.LP0079.004 Given the following:

Unit 1 is at Rated Thermal Power Two Letdown Orifices are in service Reactor Makeup settings are:

1HC-111, Rx Makeup Water Flow Controller set to 40 1HC-110, Boric Acid Flow Controller set to 5 1LT-141, VCT Level fails to 100%

Which of the following describes the expected impact to the VCT?

With NO operator action, the ACTUAL VCT level will lower . . .

A. due to Letdown diverting. Auto makeup will NOT start and eventually the VCT level will lower to the RWST swap over setpoint. 1LT-141 will actuate a switchover to RWST Suction on Low-Low VCT level.

B. slowly due to normal RCS losses. Auto makeup will attempt to control VCT.

Eventually VCT level will lower to the RWST swap over setpoint. 1LT-112 will actuate a swap over to RWST Suction on Low-Low VCT level.

C. due to Letdown diverting. Auto makeup will attempt to control VCT level.

Eventually VCT level will lower to the RWST swap over setpoint. With 1LT-141 failed high, the coincidence for swap over will not be made up and level will continue lowering to 0%.

D. slowly due to normal RCS losses. Auto makeup will NOT start and eventually the VCT level will lower to the RWST swap over setpoint. With 1LT-141 failed high, the coincidence for swap over will not be made up and level will continue lowering to 0%.

Tuesday, April 30, 2019 11:03:46 AM 94

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2012 DC Cook Quesiton 4 K/A:

004K4.12 Chemical and Volume Control System (CVCS)

Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following: Minimum level of VCT (Imp 3.1/3.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to use knowledge of the required logic for the low-low VCT level and the response of the system to a level channel failure.

Cognitive Level:

Comprehension 3-PEO: The operator must understand the initial conditions, determine what affect the instrument failure will have, and then determine what will occur based on the intial conditions if no operator actions are taken.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:46 AM 95

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

LP0079, Chemical and Volume Control System, Rev 25 STPT 7.1, Setpoint Document Chemical and Volume Control System Setpoint:

General Instrumentation, Rev 17 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following plant conditions:

Unit 2 is operating at 100%.

120 GPM letdown is in service.

Volume Control Tank (VCT) blender controls are in automatic.

VCT level transmitter QLC-452 fails to 100%.

Which ONE of the following describes the expected impact to the VCT?

With NO operator action, the VCT level will lower ...

A. slowly due to normal RCS losses. Auto VCT makeup will NOT start and eventually the VCT level will lower to the RWST switchover setpoint. With QLC-452 failed high, the coincidence for switchover will not be made up and level will continue lowering to 0%.

B. due to Letdown diverting. Auto VCT makeup will NOT start and eventually the VCT level will lower to the RWST switchover setpoint. QLC-451 will actuate a switchover to RWST Suction on Lo-Lo VCT level.

C. due to Letdown diverting. Auto VCT makeup will attempt to control VCT level.

Eventually VCT level will lower to the RWST switchover setpoint. With QLC-452 failed high, the coincidence for switchover will not be made up and level will continue lowering to 0%.

D. slowly due to normal RCS losses. Auto VCT makeup will attempt to control VCT. Eventually VCT level will lower to the RWST switchover setpoint. QLC-451 will actuate a switchover to RWST Suction on Lo-Lo VCT level.

Proposed answer: C Tuesday, April 30, 2019 11:03:46 AM 96

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With letdown at 80 gpm, and a failure of 1LT-141 to 100% this will cause the following. The divert valve will modulate to full open, actual VCT level will start to decrease. Auto makeup will occur from 17-28%, but with letdown at 80 gpm, and makeup at 40 gpm, there will be a net loss of level, until the VCT empties.

There will be no auto swap over to the RWST as that requires both instruments (1LT-112 and 1LT-141) to be at the lo-low setpoint.

A INCORRECT: VCT level will be lowering during to diverting, but makeup will occur and swap over to the RWST will not occur. Plausible if the operator has a misconception of the logic controls for VCT level given no operator actions.

B INCORRECT: Level will be lowering due to diverting, and auto swap over to the RWST will not occur. Plausible if the operator has a misconception of the logic controls for VCT level given no operator actions.

C CORRECT: See above.

D INCORRECT: Level will be lowering due to diverting, and auto makeup will occur. Plausible if the operator has a misconception of the logic controls for VCT level given no operator actions.

Learning Objective:

DESCRIBE the interlock associated with the Chemical Volume and Control (include administrative limitations)

a. Orifice Isolation valves

b. Letdown isolation valves

c. Divert valve Volume Control Tank outlet valve and Refueling water makeup supply valves Containment Isolation Valves Charging pumps (051.02.LP0079.004)

DESCRIBE the interlock associated with the CVC Boration and Dilution Control Systems (include administrative limitations)

a. Makeup Mode Selector Switch

b. Boric Acid and Makeup water flow controllers

c. Flow Control Vale, VCT Inlet valve , and VCT outlet valve (051.02.LP0082.005)

Tuesday, April 30, 2019 11:03:46 AM 97

QUESTIONS REPORT for 2019 NRC Exam Master

31. 2019 NRC 031/SYS/005K5.05/2.7*/3-PEO/RO/BANK/110E018-1/051.03.LP0069.010 Given the following:

Unit 1 is shutdown and in MODE 4 at 300°F and 300 psig RHR cooling is established and a cooldown rate of 25°F/hr has been initiated using 1HX-11A, Residual Heat Removal Heat Exchanger 1P-1B, Reactor Coolant pump is running The RCS is solid with PCV-135, Letdown Line Backpressure CV maintaining RCS pressure in MANUAL If 1RH-624, HX-11A RHR HX Outlet FCV, were to have its instrument air supply line rupture, what alarm would the operator be expected to address for this malfunction?

(Assume NO operator action)

A. 1C03 1D 4-1, 1P-1B RCP No. 1 SEAL P LOW B. 1C03 1D 4-4, RHR LOOP LOW FLOW C. 1C04 1C 3-5, LOW TEMPERATURE OVERPRESSURE D. 1C04 1C 4-7, 1HX-3A & B NONREGEN HX LETDOWN OUTLET TEMPERATURE HIGH Tuesday, April 30, 2019 11:03:46 AM 98

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2011 PBNP Question 7 K/A:

005K5.05 Residual Heat Removal System (RHRS)

Knowledge of the operational implications of the following concepts as they apply the RHRS: Plant response during "solid plant": pressure change due to the relative incompressibility of water.

(Imp 2.7*/3.1*)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine how this failure will affect the plant when it is solid plant and automatic features are in manual.

Cognitive Level:

Comprehension 3-PEO: The must operator understand the initial conditions, determine what affect the failure will have on the valve and how that will affect the system.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:46 AM 99

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ARB 1C03 1D 4-1, 1P-1B RCP No. 1 Seal P Low Rev 6 110E018-1, Auxiliary Coolant System (RHR) Rev 71 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 is shutdown and currently in MODE 4 at 300°F and 300 psig:

RHR cooling is established and a cooldown rate of 25°F/Hr has been initiated using 1HX-11A, Residual Heat Removal Heat Exchanger.

The 1P-1B RCP is running.

The RCS is solid with PCV-135, Letdown Line Backpressure CV maintaining RCS pressure in MANUAL.

If 1RH-624, HX-11A RHR HX Outlet FCV, were to have its instrument air supply line rupture, what alarm would the operator be expected to address first for this malfunction?

(Assume NO operator action)

A. RHR Low Flow B. NRHX Outlet Temperature High C. 1P-1B RCP No. 1 Seal P Low D. Low Temperature Overpressure Proposed answer: C Tuesday, April 30, 2019 11:03:46 AM 100

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

1RH-624 fails open which causes max cooling of the RCS. At solid plant conditions the cooldown will decrease RCS pressure which will lower #1 Seal DP to a point where the RCP may need to be tripped. Therefore this alarm needs to be addressed.

A CORRECT: See above.

B INCORRECT: Plausible, if the operator has a misconception that the valve will limit or stop flow based on the failure. The low flow alarm will not come in because flow control valve 1RH-626 should maintain flow control.

C INCORRECT: Plausible if the operator has the misconception that the valve will reduce the cooldown causing the RCS pressure to rise.

Low Temperature Overpressure will not alarm below 420 psig.

D INCORRECT: Plausible if the operator has the misconception that letdown would increase due to the failure. The NRHX outlet temperature is controlled by TCV-130, and it is sized such that any additional letdown flow could be adequately cooled.

Learning Objective:

Predict the effect of a RHR system malfunction on the RHR and RCS systems during RHR cooldown.

(051.03.LP0069.010)

Tuesday, April 30, 2019 11:03:46 AM 101

QUESTIONS REPORT for 2019 NRC Exam Master

32. 2019 NRC 032/SYS/006K4.09/3.8/2-RI/RO/BANK/883D195 SH 9/051.03.LP0066.013 Given the following:

Unit 1 is at Rated Thermal Power An automatic Safety Injection occurs Concurrent with the Safety Injection, the following happens:

1A-06, 4160 VAC Safeguards bus lockout occurred 1X-04, Low Voltage Station Auxiliary Transformer lockout Which of the following describes the status of 1SI-841A(B) SI Accumulator Outlet Valves and 1SI-852A(B) Low Head SI Core Deluge Isolations 2 minutes after the actuation of Safety Injection Signal?

1SI-841A 1SI-841B 1SI-852A 1SI-852B A. OPEN OPEN OPEN OPEN B. OPEN CLOSED OPEN CLOSED C. OPEN OPEN OPEN CLOSED D. CLOSED OPEN CLOSED OPEN RO Tier 2 Group 1 Source: Bank Question History:

None K/A:

006K4.09 Emergency Core Cooling System (ECCS)

Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Valve positioning on safety injection signal (Imp 3.9/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what valves are interlocked to reposition.

Cognitive Level:

Comprehension 2-RI: The must operator understand the initial conditions, determine what affect the failures will have on the valves which will reposition on an SI signal.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:46 AM 102

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

883D195 Sh 9, Safeguard Sequence Logic Rev 19 PBE-7033, Simplified Electrical Power Distribution Diagram Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 1 is at Rated Thermal Power An Auto Safety Injection occurs Concurrently the following happens:

1A06 4.16 KV Bus Lockout occurred 1X04 Low Voltage Station Transformer lockout Which of the following describes the status of 1SI-841A(B) SI Accumulator Outlet Valves and 1SI-852A(B) Low Head SI Core Deluge Isolations?

1SI-841A 1SI-841B 1SI-852A 1SI-852B A. OPEN OPEN OPEN OPEN B. OPEN CLOSED OPEN CLOSED C. OPEN OPEN OPEN CLOSED D. CLOSED OPEN CLOSED OPEN Proposed answer: C Tuesday, April 30, 2019 11:03:46 AM 103

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With an automatic actuation of Safety Injection, all the valves in question have a time 0 open actuation signal. The operator must recall that the 1SI-841A(B), SI Accumulator Outlet valves have a normal at power position of open with the breaker power off. And that the 1SI-852A(B), Low Head SI Core Deluge Isolation are powered from 1B-32, 480V Safeguards MCC PAB Safeguards bus and 1B-42, 480V Safeguards MCC PAB Safeguards bus respectively, and with the 1A06, 4160 VAC Safeguards bus lockout, 1SI-852B will have no power.

A INCORRECT: The first part is correct. The second part is incorrect as 1SI-852B will have no power, plausible if the student has a misconception of power supplies or type of valve (AOV vice MOV), as AOVs in the SI system have a fail/loss of power position which lines them up for system actuation.

B INCORRECT: The first part is incorrect, plausible if the student has a misconception on that the valves need to reposition and 1A06 powers the required bus for 1SI-841B. The second part is correct.

C CORRECT: See above.

D INCORRECT: The first part is incorrect, plausible if the student has a misconception on that the valves need to reposition and 1A06 powers the required bus for 1SI-841A. The second part is incorrect, plausible if the student has a misconception on that the valves need to reposition and 1A06 powers the required panel for 1SI-852A.

Learning Objective:

STATE, from memory, the Safety Injection sequence.

(051.03.LP0066.013)

STATE the power supply for the Safety Injection System components:

a. Safety Injection Pumps Motor Operated Valves (051.03.LP0066.003)

DESCRIBE the interlocks associated with the Safety Injection System and its major components:

a. Safety Injection Pump Recirculation and Sump Suction valves

b. Containment Sump isolation and Safety Injection Pump Recirculation valves

c. RHR Pump to Spray Pump suction valves

d. RHR to SI suction cross connect and RHR pump discharge pressure.

RHR to SI suction cross connect and SI pump suction valve.

(051.03.LP0066.004)

Tuesday, April 30, 2019 11:03:46 AM 104

QUESTIONS REPORT for 2019 NRC Exam Master

33. 2019 NRC 033/SYS/007K1.01/2.9/2-RI/RO/MODIFIED/FSAR/051.01.LP0078.002 Given the following:

A small break LOCA has occurred Pressurizer PORVs are being used to reduce RCS pressure per EOP-1.2, Post LOCA Cooldown and Depressurization Containment pressure is 12 psig Which of the following is the minimum pressure in PSI, inside the Pressurizer Relief Tank that will cause the PRT rupture disc to rupture?

A. 88 psi B. 100 psi C. 112 psi D. 127 psi RO Tier 2 Group 1 Source: Modified Question History:

2005 Summer Question 11 K/A:

007K1.01 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the physical connections and/or cause-effect relationships between the PRTS and the following systems: Containment system (Imp 2.9/3.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what affect the containment system pressure will have on the PRT/Quench tank system.

Cognitive Level:

Comprehension 2-RI: The operator must determine how the initial conditions, will affect the function of the PRT/Quench tank rupture disk.

10 CFR Part 55 Content:

55.41 2 to 9 55.43 Tuesday, April 30, 2019 11:03:46 AM 105

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

FSAR Table 4.1-3 541F091 Sh2, Reactor Coolant System P&ID Rev 41 Proposed reference to be provided to the applicants during examination:

None Original Question:

A small break LOCA has occurred. Pressurizer PORVs are being used to reduce RCS pressure per EOP-2.1, "Post-LOCA Cooldown and Depressurization."

- Containment pressure is 14 psig.

Which ONE of the following represents the maximum pressure that could be reached inside the Pressurizer Relief Tank (PRT) before the PRT rupture disc ruptures?

A. 90 psig B. 100 psig C. 114 psig D. 128 psig Proposed answer: C Tuesday, April 30, 2019 11:03:46 AM 106

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With the containment at 12 psig, this will affect the PRT rupture disc, causing it to rupture at higher pressure than the design 100 psig (or a 100 psid), since the containment side of the rupture disc is seeing 10 psig, the rupture disc will rupture at 112 psig (100+12 psig=112psig).

A INCORRECT: Plausible if the operator has the misconception that the pressure in the containment will affect the rupture disc in the opposite manner.

B INCORRECT: Plausible as this is the design pressure when the rupture disc should rupture if the containment was at or near 0 psig.

C CORRECT: See above.

D INCORRECT: Plausible if the operator converts to psia and takes in to account the 12 psig.

Learning Objective:

DRAW and DISCUSS a one line diagram of the Pressurizer, Level Control, Pressure Control, and Relief System. Discussion of this drawing should include system flowpaths, major components, and interfaces with other major systems:

Pressurizer Vessel and Instrumentation taps Power Operated Relief (PORV) and Block Valves Safety Valves and associated discharge line Loop Seals Normal and Auxiliary Spray Valves and Spray Valve Bypass Pressurizer Relief Tank (PRT) and penetrations Discharge to Waste Gas Nitrogen supply Reactor Makeup Water Supply Discharge to Reactor Coolant Drain Tank (RCDT)

Rupture Disc (051.01.LP0078.002)

Tuesday, April 30, 2019 11:03:46 AM 107

QUESTIONS REPORT for 2019 NRC Exam Master

34. 2019 NRC 034/SYS/008K3.01/3.4/2-RI/RO/BANK/AOP 5B/051.02.LP0079.001 Given the following:

Unit 2 is at Rated Thermal Power 2TCV-130, Component Cooling Water Return from the Non-Regenerative Heat Exchanger Temperature Control valve, fails due to a broken air line.

Which of the following describes the effect of this failure on the plant?

(Assume no operator action)

A. Letdown temperature goes up; the rise in letdown temperature causes the letdown demineralizers to remove less boron, resulting in a boration.

B. Letdown temperature goes up; the rise in letdown temperature causes the letdown demineralizers to remove more boron, resulting in a dilution.

C. Letdown temperature goes down; the decrease in letdown temperature causes the letdown demineralizers to remove less boron, resulting in a boration.

D. Letdown temperature goes down; the decrease in letdown temperature causes the letdown demineralizers to remove more boron, resulting in a dilution.

RO Tier 2 Group 1 Source: Bank Question History:

2009 PBNP Question 33 K/A:

008K3.01 Component Cooling Water System (CCW)

Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: Loads cooled by CCWS.

(Imp 3.4/3.5)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine after a fault, the effect on the loads being cooled by CCWS.

Cognitive Level:

Comprehension 2-RI: Requires the operator to determine the effect of the loss of air on the valve, then determine how that will affect CCWS and ultimate the loads cooled.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:46 AM 108

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP 5B Loss of Instrument Air Attachment H Rev 49 191007, INPO Lesson Plan Demineralizers and Ion Exchangers Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

TCV-130, "Component Cooling Water Return from the Non-Regenerative Heat Exchanger Temperature Control Valve", fails due to a broken air line.

Assuming no action by the crew, which of the following describes the effect of this failure on the plant?

A. Letdown temperature goes down; the decrease in letdown temperature causes the letdown demineralizers to remove more boron, resulting in a minor dilution.

B. Letdown temperature goes up; the rise in letdown temperature causes the letdown demineralizers to remove more boron, resulting in a minor dilution.

C. Letdown temperature goes up; the rise in letdown temperature causes the letdown demineralizers to remove less boron, resulting in a minor boration.

D. Letdown temperature goes down; the decrease in letdown temperature causes the letdown demineralizers to remove less boron, resulting in a minor boration.

Proposed answer: A Tuesday, April 30, 2019 11:03:46 AM 109

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The loss of air will cause the valve to fail open. The valve failure will cause more flow and therefore more cooling, this in turn will cause more boron to be removed.

A INCORRECT: The first part is incorrect, plausible if the operator has the misconception that the valve is a failed close valve, which most air operated valves are. The second part is incorrect, plausible as this would occur if the temperature went up.

B INCORRECT: The first part is incorrect, plausible if the operator has the misconception that the valve is a failed close valve, which most air operated valves are. The second part is correct.

C INCORRECT: The first part is correct. The second part is incorrect, if the operator has the misconception that a reduction in temperature will cause a reduction in boron.

D CORRECT: See above.

Learning Objective:

DESCRIBE the function and/or purpose, design bases, and operating characteristics of CVCS. Description should include design operation of the following specific components:

a. Letdown orifices and isolation valves

b. Delay loop section of the letdown line

c. Demineralizers

d. Volume Control Tank

e. Regenerative and non-regenerative heat exchangers

f. Charging Pumps and charging control valve

g. Pressure and temperature control valves

h. Seal injection filters and valves

i. Auxiliary spray and auxiliary charging (051.02.LP0079.001)

Tuesday, April 30, 2019 11:03:46 AM 110

QUESTIONS REPORT for 2019 NRC Exam Master

35. 2019 NRC 035/SYS/008G2.4.21/4.0/2-RI/RO/BANK/1-SOP-CC-001/051.03.LP0069.010 Given the following:

Unit 1 is in MODE 5 The RCS is solid on RHR RCS pressure is 350 psig 1HC-135, Letdown Line Pressure Controller, is in MANUAL to control RCS pressure Charging pumps are in MANUAL Letdown is aligned to RHR All RCS and RHR conditions are STABLE The PAB operator performs Attachment C, CC HX SW Blowdown, of 1-SOP-CC-001, Component Cooling System, for 1HX-12A, Component Cooling Water Heat Exchanger If no other operator actions are taken, RCS pressure will . . .

A. lower because RHR temperature will lower.

B. rise because more RHR flow will bypass the RHR heat exchangers.

C. rise because Service Water blowdown flow will bypass the CCW heat exchanger tubes.

D. lower because the Non-Regenerative heat exchanger Letdown outlet temperature will rise.

Tuesday, April 30, 2019 11:03:46 AM 111

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2003 PBNP Question 52 K/A:

008G2.4.21 Component Cooling Water System (CCW)

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.

(Imp 4.0/4.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what affect the blowdown will have on core cooling and primary pressure given the initial conditions.

Cognitive Level:

Comprehension 2-RI: Requires the operator to determine the effect of the blowdown on the heat exchangers, then determine what that will cause given the current lineup of the plant..

10 CFR Part 55 Content:

55.41 7 55.43 5 Tuesday, April 30, 2019 11:03:46 AM 112

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

1-SOP-CC-001, Component Cooling Water System, Attachment D Rev 32 Proposed reference to be provided to the applicants during examination:

None Original Question:

- Unit 1 is in Mode 5.

- The RCS is solid at 300 psig and on RHR.

- 1HC-135, Letdown Line Pressure Controller, is in Manual to control RCS pressure.

- All RCS and RHR conditions are stable.

The PAB operator then performs a blowdown of the Service Water side of 1HX-12A and HX-12B, Component Cooling Water Heat Exchangers.

If no other operator actions are taken, RCS pressure will:

A. lower because RHR temperature will lower.

B. rise because more RHR flow will bypass the RHR Heat Exchangers.

C. rise because Service Water blowdown flow will bypass the CCW Heat Exchanger tubes.

D. lower because the Non-Regenerative Heat Exchanger letdown outlet temperature will rise.

Proposed answer: A Tuesday, April 30, 2019 11:03:46 AM 113

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Service water flow will rise from the cleaning effect of the blowdown on the heat exchanger. This will cause a lower CCW temperature leaving the CCW HX.

This in turn cools RHR more as it passes through the RHR HX, causing temperature and pressure to lower in the RCS due to the controllers being in manual and not automatic.

A CORRECT: See above.

B INCORRECT: The temperature of RHR is manually adjusted. The 624/625 valves are manually set to pass a certain amount of flow through the HX and 626 changes position to maintain RHR FLOW, vice temperature, plausible if the operator has the misconception these are in automatic.

C INCORRECT: SW is being blown down from the bottom of the HX bells, which will cause SW flow to rise, thus lowering temperature, plausible if the operator has the misconception that due to the rise in SW flow, the more RHR will be bypass the heat exchanger and this will cause the rise in pressure.

D INCORRECT: NRHX temp will not rise, since there is more cooling flow to CCW, NRHX outlet temp would be expected to lower, plausible if the operator has the misconception that manual on the controller will control the setpoint not valve position.

Learning Objective:

PREDICT the effects and RECOGNIZE the associated corrective actions for malfunctions of the Residual Heat Removal System.

(051.03.LP0069.010 Tuesday, April 30, 2019 11:03:46 AM 114

QUESTIONS REPORT for 2019 NRC Exam Master

36. 2019 NRC 036/SYS/010A3.01/3.0/2-RI/RO/BANK/ARB 1C04 1.1/051.01.LP0078.002 Given the following:

Unit 2 is in MODE 1 Testing of 2RC-515, 2RC-431C PZR PORV Isolation MOV per IT 25, Reactor Coolant Valves (Quarterly) is being performed The CO opens 2RC-515 2RC-431C, PZR PORV seat leakage causes the following annunciators on 2C04 2C to alarm 1-11, 2T-2 PRT PRESS HIGH TEMP HIGH LEVEL HI OR LO 1-4, PRESSURIZER SAFETY OR RELIEF LINE TEMPERATURE HIGH 2RC-515 will NOT close.

How will the PRT temperature and pressure trends initially respond after these annunciators are received?

(Assume no further operator action)

A. PRT temperature and pressure will both start to lower.

B. PRT temperature will rise, but pressure will lower.

C. PRT temperature will lower, but pressure will rise.

D. PRT temperature will rise, and pressure will rise.

Tuesday, April 30, 2019 11:03:46 AM 115

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2009 Millstone Unit 3 Question 38 K/A:

010A3.01 Pressurizer Pressure Control System (PZR PCS)

Ability to monitor automatic operation of the PZR PCS, including: PRT temperature and pressure during PORV testing.

(Imp 3.0/3.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what happening to the temperature and pressure trends for the PRT during PZR PORV block valve testing with PZR PORV seat leakage.

Cognitive Level:

Comprehension 2-RI: The operator must determine how the pressure and temperature trends will be effected in the PRT given a set of initial conditions..

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:46 AM 116

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ARB 2C04 2C 1-11, 2T2 PRT Pressure High Temp Hi Level Hi or Lo Rev 5 541F091 Sh2, Reactor Coolant System P&ID Rev 41 Proposed reference to be provided to the applicants during examination:

None Original Question:

The plant is in MODE 3 with PORV Testing in progress, and the following sequence of events occurs:

1. The RO opens the "A" PORV, and the PORV remains open.

2. PRT temperature and pressure start to increase.

3. The PZR REL TK PRESSURE HI annunciator comes in on MB4.

4. All automatic system responses occur, as designed.

How will the PRT temperature and pressure trends initially respond after this annunciator is received?

A. PRT temperature will continue to increase, but pressure will start to decrease.

B. PRT temperature will continue to increase, and pressure will start to increase at a faster rate.

C. PRT temperature and pressure will both continue to increase at the same rate.

D. PRT temperature and pressure will both start to decrease.

Proposed Answer: B Tuesday, April 30, 2019 11:03:46 AM 117

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the Unit is in MODE 1, at NOP/NOT, and the fact that the block valve will not shut with the PORV exhibiting signs of leakage, the pressure and temperature of the PRT will rise, and continue to rise until either the leakage is isolated or the rupture disk ruptures.

A INCORRECT: Plausible, because this would occur when the PRT rupture disc ruptured, the sudden loss of pressure would cause the lowering of temperature. Without the low pressurizer pressure annunciator, the leak is small, therefore would not reach the setpoint for the rupture disc.

B INCORRECT: Plausible, if the operator has the misconception the vent opens at the 7 psig interlock vice closes as this would limit the pressure rise seen.

C INCORRECT: Plausible if the operator has the misconception that the sparging effect in the PRT will have an overall cooling effect on the temperature.

D CORRECT: See above.

Learning Objective:

DRAW and DISCUSS a one line diagram of the Pressurizer, Level Control, Pressure Control, and Relief System. Discussion of this drawing should include system flowpaths, major components, and interfaces with other major systems:

Pressurizer Vessel and Instrumentation taps Power Operated Relief (PORV) and Block Valves Safety Valves and associated discharge line Loop Seals Normal and Auxiliary Spray Valves and Spray Valve Bypass Pressurizer Relief Tank (PRT) and penetrations Discharge to Waste Gas Nitrogen supply Reactor Makeup Water Supply Discharge to Reactor Coolant Drain Tank (RCDT)

Rupture Disc (051.01.LP0078.002)

Tuesday, April 30, 2019 11:03:46 AM 118

QUESTIONS REPORT for 2019 NRC Exam Master

37. 2019 NRC 037/SYS/010K5.01/3.5/3-SPK/RO/BANK/STEAM TABLES/055.02.LP0162.010 Given the following:

Unit 1 is in MODE 4 Pressurizer pressure is 350 psig Pressurizer temperature is 436°F Pressurizer level is 65%

Which answers the following:

(1) Is a bubble drawn in the pressurizer?

AND (2) What is being used to maintain pressurizer pressure?

Bubble in PZR Pressure Control A. YES PZR Spray B. YES Letdown PCV C. NO PZR Spray D. NO Letdown PCV Tuesday, April 30, 2019 11:03:46 AM 119

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2016 Prairie Island Question 36 K/A:

010K5.01 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the operational implications of the following concepts as they apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables.

(Imp 3.5/4.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to use steam tables to determine the condition of the fluid in the pressurizer.

Cognitive Level:

Comprehension 2-SPR: The operator must determine use steam tables to determine the condition of the fluid in the pressurizer, and then recall how pressure will be controlled when the fluid is in that condition.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:46 AM 120

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

Steam Tables OP 1A, Cold Shutdown to Hot Standby Unit 1 Rev 17 Step 5.30.12 Proposed reference to be provided to the applicants during examination:

Steam Tables Original Question:

- Unit 1 is in Mode 4.

- Pressurizer pressure is 335 psig.

- Pressurizer temperature is 432°F.

- Pressurizer level is 65%.

(1) Is a bubble drawn in the pressurizer, and, (2) What is being used to maintain pressurizer pressure?

Bubble in PRZR Pressure Control A. YES Letdown PCV B. YES PRZR Heaters C. NO Letdown PCV D. NO PRZR Heaters Proposed answer : B Tuesday, April 30, 2019 11:03:46 AM 121

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With pressure at 350 psig (365 psia) and temperature at 436°F, the fluid in the pressurizer is saturated, indicating there is steam and bubble in the pressurizer.

Since there is a steam bubble in the pressurizer, the letdown pressure control valve will have little to no effect on pressurizer pressure. Per OP 1A, the operator is directed to maintain RCS pressure with spray valves.

A CORRECT: See above.

B INCORRECT: The first part is correct. The second part is incorrect, plausible as a bubble is drawn in the pressurizer and if the operator incorrectly believes the letdown control valve is used to maintain pressurizer pressure with a bubble in the pressurizer, Letdown PCV will be used to control flow through the letdown system as pressure rises.

C INCORRECT: The first part is incorrect, plausible if the operator incorrectly converts psig to psia when using the steam tables to determine if the pressurizer fluid is saturated. This could result in the examinee incorrectly believing the fluid in the pressurizer is subcooled. The second part is correct, plausible if the examinee incorrectly believes pressurizer sprays are used to control pressurizer pressure during all modes, because all pressurizer heaters are energized while drawing a bubble.

D INCORRECT: The first part is incorrect, plausible if the operator incorrectly converts psig to psia when using the steam tables to determine if the pressurizer fluid is saturated. This could result in the examinee incorrectly believing the fluid in the pressurizer is subcooled. The second part is correct for the condition.

Learning Objective:

Given access to the Site Specific Simulator or a set of plant conditions, DEMONSTRATE the ability to draw a bubble in the Pressurizer and RECOGNIZE the indications of bubble formation.

(055.02.LP0162.010)

Tuesday, April 30, 2019 11:03:46 AM 122

QUESTIONS REPORT for 2019 NRC Exam Master

38. 2019 NRC 038/SYS/012A4.06/4.3/2-RI/RO/NEW/883D195 SH 5A/053.02.LP0273.007 Given the following:

Unit 1 is at Rated Thermal Power Performing 1ICP 02.003A, Reactor Protection System Logic Train A 31 Day Surveillance Test, resulted in the current reactor trip breaker configuration shown below:

The OS directs a MANUAL trip of the reactor based on degrading condtions When the CO manually tripped the reactor using both pushbuttons, AND the TRAIN B Reactor Trip Pushbutton FAILED to function Which of the following shows the position of the reactor trip and bypass breakers 5 seconds after the manual trip?

(Assume NO automatic Reactor trip signal is present AND the second set of Reactor Trip pushbuttons was NOT used)

A. B.

C. D.

Tuesday, April 30, 2019 11:03:47 AM 123

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

012A4.06 Reactor Protection System Ability to manually operate and/or monitor in the control room: Reactor trip breakers (Imp 4.3/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to monitor the reactor trip breaker operation given a fault during actuation of the system.

Cognitive Level:

Comprehension 2-RI: The operator must understand the initial conditions, determine the effect of the fault on both the reactor trip and bypass breakers, and also the RPS.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

883D195 Sh 2, Reactor Trip Signal Logic Diagram Rev 9 883D195 Sh 15, RC Trip Signal Logic Diagram Rev 9 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:47 AM 124

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With the RTA and RTB reactor trip breakers racked in and closed and additionally BYA bypass breaker racked in and closed, given a manual trip, with no automatic trip signal present, this will cause a trip signal to be sent to the Train A breakers (RTA reactor trip breaker and BYB bypass breaker). Only the RTA reactor trip breaker will open, and furthermore the logics for the turbine trip (20AST and 20/ET) and feedwater isolation will not be made up, so the reactor will not trip.

A CORRECT: See above.

B INCORRECT: Plausible if the operator reverses how the actuation failure will affect the reactor trip and bypass breakers without taking into consideration the signal from BYB bypass breaker which is racked out.

C INCORRECT: Plausible if the operator has the misconception that the Train A pushbuttons affect the A components, (i.e., RTA reactor trip breaker and BYA bypass breaker).

D INCORRECT: Plausible as this would be the result if Train B functioned and Train A did not, taking into consideration the BYB bypass breaker signal caused by it being tripped. When RTB reactor trip breaker with BYB bypass breaker already being tripped, this will make up the logic that will cause Feedwater isolation and the turbine trip. This in turn will turbine trip to reactor trip due to meeting the P-7 and P9 permissives.

Learning Objective:

DESCRIBE the Reactor Trips, automatic functions and interlocks associated with the Reactor Protection System and its major components. Description should include actuation setpoints, actuation logic, logic acceptability, requirements to enable actuation, and protection afforded by each of the following:

a. Reactor Trips

b. Reactor Trip Permissives

c. Rod Stops

d. Turbine Runbacks (053.02.LP0273.007)

Tuesday, April 30, 2019 11:03:47 AM 125

QUESTIONS REPORT for 2019 NRC Exam Master

39. 2019 NRC 039/SYS/012K1.02/3.4/1-P/RO/BANK/AOP-0.0/05503.LP3456.001 According to AOP 0.0, Vital DC System Malfunction, a loss of which of the following would cause a DUAL UNIT TRIP?

A. D-01, 125 VDC Distribution Panel B. D-13, 125 VDC Distribution Panel C. D-18, 125 VDC Distribution Panel D. D-21, 125 VDC Distribution Panel RO Tier 2 Group 1 Source: Bank Question History:

2017 PBNP Question 69 Previous 2 NRC Exams K/A:

012K1.02 Reactor Protection System Knowledge of the physical connections and/or cause effect relationships between the RPS and the following systems: 125V dc system (Imp 3.4/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand the electrical connections to the 125 VDC system the effect of a loss on the units.

Cognitive Level:

Knowledge 1-P: The operator must recall the loss of which will cause a dual unit trip.

10 CFR Part 55 Content:

55.41 2 to 9 55.43 Tuesday, April 30, 2019 11:03:47 AM 126

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-0.0, vital DC system Malfunction, Attachment E, Rev 36 Proposed reference to be provided to the applicants during examination:

None Original Question According to AOP 0.0, Vital DC System Malfunction, a loss of which of the following would cause a DUAL UNIT TRIP?

A. D01 B. D13 C. D18 D. D21 Proposed answer: A Justification:

A loss of D01 or D02 will cause a dual unit trip.

A CORRECT: See above.

B INCORRECT: Plausible a loss of D13 will cause a Unit 2 trip.

C INCORRECT: Plausible a loss of D18 will cause a Unit 2 trip.

D INCORRECT: Plausible, a loss of D21 will cause a Unit 1 trip.

Learning Objective:

Given access to the appropriate equipment indication, DIAGNOSE and DESCRIBE the plant and operator(s) response to the following condition:

a. Loss of a DC Bus (055.03.LP3456.001)

Tuesday, April 30, 2019 11:03:47 AM 127

QUESTIONS REPORT for 2019 NRC Exam Master

40. 2019 NRC 040/SYS/013K6.01/2.7*/2-DR/RO/BANK/883D195 SH7/053.06.LP0486.017 Given the following:

A Containment Pressure transmitter that feeds both SI and Containment Spray fails LOW Action has been taken in accordance with Technical Specifications to place the failed channel in TRIP Which of the following identifies the correct ESF actuation logic for the remaining Containment pressure channels?

A. Safety Injection - 1 of 2; Containment Spray - 1 of 2 plus 2 of 3 B. Safety Injection - 1 of 2; Containment Spray - 2 of 3 plus 2 of 3 C. Safety Injection - 1 of 3; Containment Spray - 1 of 2 plus 1 of 3 D. Safety Injection - 1 of 3; Containment Spray - 2 of 3 plus 1 of 3 RO Tier 2 Group 1 Source: Bank Question History:

2011 Ginna Retake Question 11 K/A:

013K6.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors (Imp 2.7*/3.1*)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand how a failure of a detector will effect on the functioning of ESFAS.

Cognitive Level:

Comprehensive 2-DR: The operator must understand the initial conditions, and the ramifications of the event. Then the operator must determine what effect this event will have on the remaining ESF actuation logic 10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:47 AM 128

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

883D195 SH 7, Safeguards Actuation Signals Logic Diagram Rev 25 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A Containment Pressure transmitter that feeds both SI and Containment Spray fails LOW.

Action has been taken in accordance with Technical Specifications to place the failed channel in TRIP.

Which ONE of the following identifies the correct ESF actuation logic for the remaining Containment pressure channels?

A. Safety Injection -1/2; Containment Spray -1/2 plus 2/3 B. Safety Injection -1/2; Containment Spray -213 plus 2/3 C. Safety Injection -1/3; Containment Spray -1/2 plus 1/3 D. Safety Injection -1/3; Containment Spray -2/3 plus 1/3 Proposed answer: A Tuesday, April 30, 2019 11:03:47 AM 129

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Safety Injection is normally 2 of 3 logic for Containment Pressure and Containment Spray is 2 of 3 taken twice. When a protection channel that feeds both actuations is removed from service, bistables are tripped in all cases.

AUTO SI will occur if either of the two remaining bistables trip, and an AUTO CS normally needs 2 of 3 for 2 sets of inputs, since one is already in a tripped condition, only 1 additional is needed to meet the 2 of 3 requirement for the first set of inputs, and for the other, it will need and additional 2 of 3 channels to trip.

A CORRECT: See above.

B INCORRECT: The first part is correct. The second part is incorrect, plausible as it would be correct if Containment Spray was placed in BYPASS rather than trip.

C INCORRECT: The first part is incorrect, SI is 1 of 2, plausible because PZR Pressure SI actuation is 2 of 3 logic, and the operator can confuse the logic with the 2 of 4 logic provided by other RPS actuations. The second half is correct.

D INCORRECT: The first part is incorrect, SI is 1 of 2, plausible because PZR Pressure SI actuation is 2 of 3 logic, and the operator can confuse the logic with the 2 of 4 logic provided by other RPS actuations. The second part is incorrect, plausible as it would be correct if Containment Spray was placed in BYPASS rather than trip.

Learning Objective:

Describe the automatic functions associated with the Engineered Safety Features Actuation System (ESFAS) and its major components. Description should include actuation setpoints, actuation logic, logic acceptability, requirements to enable actuation, and protection afforded by the system.

(053.06.LP0486.017)

Tuesday, April 30, 2019 11:03:47 AM 130

QUESTIONS REPORT for 2019 NRC Exam Master

41. 2019 NRC 041/SYS/022A2.04/2.9*/2-DR/RO/NEW/FSAR/031.02.LP0405.015 Given the following:

Unit 1 is at Rated Thermal Power at MOL 1P-14B, Containment Spray pump is Out of Service A design basis Steam Line Break inside containment occurs 1SW-2907 AND 2908, Containment Ventilation Cooler Emergency Flow Control Valves remain SHUT Containment Accident Recirc Fans indicate as shown below:

Which answers the following:

(1) With no operator actions, will the design pressure limits of the containment be exceeded?

AND (2) What, if any actions are necessary?

A. (1) No (2) No action needed, the required amount flow is received with SW-2907 and SW 2908 closed B. (1) No (2) BOTH SW-2907 and SW 2908 are opened to ensure margin to prevent excess pressure in the containment C. (1) Yes (2) EITHER SW-2907 or SW 2908 must be opened for required flow D. (1) Yes (2) BOTH SW-2907 and SW 2908 must be opened for required flow Tuesday, April 30, 2019 11:03:47 AM 131

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

022A2.04 Containment Cooling System (CCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of service water (Imp 2.9*/3.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to understand how a failure of the SW system will affect the containment and what actions are necessary to mitigate the loss.

Cognitive Level:

Comprehensive 2-DR: The operator must understand determine the effect of the loss based on the initial conditions and talking no operator actions, then recall what operator actions the are necessary to mitigate the loss of SW.

10 CFR Part 55 Content:

55.41 5 55.43 5

Reference:

FSAR, Final Safety Analysis Report Rev 11/12/18, Section 14.2.5 Containment Response Calculation Section Page 14.2.5-10 0f 30 EOP-0, Reactor Trip of Safety Injection Rev 66 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:47 AM 132

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The two accidents that can challenge the containment structure are LOCA and SLB inside containment. Containment structure protection is shown for a DBLOCA inside containment if only one train of containment cooling is available (i.e., one spray pump and 2 accident fans), but containment structure protection is shown during a MSLB with both trains of containment cooling functioning.

This is due to other assumptions made during a MSLB as the FIV failure. Per EOP-0, only one valve (SW-2907 or 2908) needs to be open to satisfy the necessary cooling flow.

A INCORRECT: The first part is incorrect, protection is shown if both trains of containment cooling function correctly, and SW flow will be low, plausible as only one train of containment cooling is needed for a LOCA, and all components are running. The second part is incorrect, plausible due to the first part of the answer.

B INCORRECT: The first part is incorrect, protection is shown if both trains of containment cooling function correctly, and SW flow will be low, plausible as only one train of containment cooling is needed for a LOCA, and all components are running. The second part is incorrect, plausible if the operator recalls the step in EOP-0 in a conservative manner.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator recalls the step in EOP-0 in a conservative manner.

Learning Objective:

Describe worst case criteria analyzed for various steam line break accidents.

(043.02.LP02464.007)

Given access to the Site Specific Simulator, RESPOND to accident conditions in accordance with the plant's Emergency Procedures.

(031.02.LP0405.015)

Tuesday, April 30, 2019 11:03:47 AM 133

QUESTIONS REPORT for 2019 NRC Exam Master

42. 2019 NRC 042/SYS/026A1.01/3.9/1-P/SRO/BANK/883D195 SH 7/051.03.LP0064.010 Given the following:

A Large Break LOCA is occurring on Unit 1 Containment pressure is 5.5 psig and RISING Which of the following identifies (1) The operation of the Containment Spray system as containment pressure rises throughout the event.

AND (2) Once actuation has occurred, the operation of Containment Spray as containment pressure is lowering?

Containment Spray pumps will automatically start . . .

A. (1) immediately on a Containment HI-HI Pressure signal (2) Containment Spray flow will be reduced to one train running prior to transition out of EOP-0.

B. (1) 10 seconds after a Containment HI-HI Pressure signal is present (2) Containment Spray flow will be reduced to one train running prior to transition out of EOP-0.

C. (1) immediately on a Containment HI-HI Pressure signal (2) Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 20 psig prior to transitioning out of EOP-1.

D. (1) 10 seconds after a Containment HI-HI Pressure signal is present (2) Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 20 psig prior to transitioning out of EOP-1.

Tuesday, April 30, 2019 11:03:47 AM 134

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2013 Turkey Point Question 41 K/A:

026A1.01 Containment Spray System (CSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment pressure (Imp 3.9/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator monitor the operation of the containment spray system and determine how it is required to be operated during an event.

Cognitive Level:

Knowledge 1-P: The operator must recall how the containment spray system auto actuation happens, and how the EOP network requires the system tobe operated.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:47 AM 135

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-0, Reactor Trip or Safety Injection Rev 66 883D195 SH 7, Safeguards Actuation Signals Logic Diagram Rev 25 883D195 SH 8, Safeguard Sequence Logic Diagram Rev 19 883D195 SH 9, Safeguards Sequence Logic Rev 19 Proposed reference to be provided to the applicants during examination:

None Original Question:

Plant conditions:

A LOCA is occurring on Unit 3.

Containment pressure is 5.5 psig and rising.

All equipment is functioning as designed.

Which ONE of the following identifies the operation of the Containment Spray system as containment pressure rises throughout the event, and once actuation has occurred, the operation of Containment Spray as containment pressure is lowering?

Containment Spray pumps will automatically start...

A. directly on a Containment Pressure HI-HI signal; Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 20 psig.

B. directly on a Containment Pressure HI-HI signal; Containment Spray flow will be reduced to one train running after transitioning from E-1, Loss of Reactor or Secondary Coolant.

C. on a Sequencer signal if a Containment Pressure HI-HI signal is present; Both Containment Spray pumps will be stopped as soon as Containment Pressure is below 20 psig.

D. on a Sequencer signal if a Containment Pressure HI-HI signal is present; Containment Spray flow will be reduced to one train running after transitioning from E-1, Loss of Reactor or Secondary Coolant.

Proposed answer: D Tuesday, April 30, 2019 11:03:47 AM 136

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Automatic Safety Injection will be actuated at 4.8 psig in the containment, and since this is a large break LOCA, the pressure will continue to rise until MSIV Closure setpoint is reached at 15 psig, and Containment Spray actuation at 25 psig. As show by 883D195 Sh 7, the containment spray actuation signal is separate from the SI signal, and it goes to a sequencer for actuation using timing relays. When EOP-0 is entered and an SI has happened, Attachment A is completed. Attachment A will secure and place in pullout one of the trains of Containment Spray pumps regardless of the event. A DBLOCA only requires one train of containment cooling (or one spray pump) to show protection.

A INCORRECT: The first part is incorrect. The spray pumps have a sequencer and timing circuits that ensure the valve operation and pump starting timing is correct, plausible if the operator has the misconception of how the spray system is actuated. The second part is correct B CORRECT: See above.

C INCORRECT: The first part is incorrect. The spray pumps have a sequencer and timing circuits that ensure the valve operation and pump starting timing is correct, plausible if the operator has the misconception of how the spray system is actuated. The second part is incorrect, with a large break LOCA, EOP-1.3 will be in effect, and the spray system will be operated as directed by EOP-1.3. Plausible because the spray pump operation is directed by EOP-1, if EOP-1.3 is not in effect, and if pressure is less than 20 psig, then the pumps are secured.

D INCORRECT: The first part is correct. The second part is incorrect, with a large break LOCA, EOP-1.3 will be in effect, and the spray system will be operated as directed by EOP-1.3. Plausible because the spray pump operation is directed by EOP-1, if EOP-1.3 is not in effect, and if pressure is less than 20 psig, then the pumps are secured.

Learning Objective:

STATE actuation setpoints and EXPLAIN effects of automatic actuations for the following components:

a. Containment Spray Pumps

b. Spray additive valves Containment Spray valves (051.03.LP0064.010)

Given access to the Site Specific Simulator, RESPOND to accident conditions in accordance with the plant's Emergency Procedures.

(031.02.LP0405.015)

Tuesday, April 30, 2019 11:03:47 AM 137

QUESTIONS REPORT for 2019 NRC Exam Master

43. 2019 NRC 043/SYS/039A1.10/2.9*/2-RDR/RO/BANK/NEED HELP/055.03.LP32438.004 Given the following:

Unit 1 is at Rated Thermal Power While monitoring the trend for 1RE-215 and 1RE-232 on PPCS, a 15 gpm tube leak develops in the B Steam Generator Initially, the crew would expect to see elevated radiation level readings to appear first on ___(1)___. If a reactor trip is required, the crew would expect the reading on 1RE-215 to ___(2)___ immediately after the trip.

1RE-215, Condenser Air Ejector Gas Monitor 1RE-232, Steam Line 1B Monitor A. (1) 1RE-215 and then on 1RE-232 (2) slowly lower B. (1) 1RE-215 and then on 1RE-232 (2) drop sharply C. (1) 1RE-232 and then on 1RE215 (2) slowly lower D. (1) 1RE-232 and then on 1RE215 (2) drop sharply Tuesday, April 30, 2019 11:03:47 AM 138

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2005 Summer Question 33 K/A:

039A1.10 Main and Reheat Steam System (MRSS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: Air ejector PRM (Imp 2.9*/3.0*)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the change in reading on the air ejector rad monitor at the start of a steam generator tube leak, and after a required trip.

Cognitive Level:

Comprehension 2-DR: The operator must determine how a small steam generator tube leak will be indicated on the RMS system, and then determine what effect tripping the reactor will cause on the RMS system.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:47 AM 139

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ML12065A095, Steam Generator Management Program: PWR Primary-to-Secondary Leak Guidelines Rev 4 Proposed reference to be provided to the applicants during examination:

None Original Question:

The Unit is operating at 100% power.

As an A SG tube leak slowly develops (from 0 gpm to a few gallons per minute), initially, the crew would expect to see elevated radiation level readings to appear first on ______________

If a reactor trip is required, the crew would expect the reading on RM-A9 to

_______________ immediately after the trip.

A. RM-G19A and then on RM-A9; slowly decrease and the reading on RM-G19A to drop sharply B. RM-A9 and then on RM-G19A; slowly decrease and the reading on RM-Gl9Ato drop sharply C. RM-A9 and then on RM-G19A; drop sharply and the reading on RM-G19A to slowly decrease D. RM-G19A and then on RM-A9; drop sharply and the reading on RM-G19A to slowly decrease Proposed answer: B Tuesday, April 30, 2019 11:03:47 AM 140

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The first indication of a SGTL will be 1RE-232, as it will take time to build up in the condenser. Post trip, the reading on 1RE-215 will slowly lower instead of drop sharply immediately due to short-lived N-16 not being a mjor contributer to the what is being sensed, and that steam usage drops significantly after the trip.

A INCORRECT: The first part is incorrect. there will be a slow build up in the condenser, plausible if the operator has a misconception on sensitivity of the detectors. The second part is correct.

B INCORRECT: The first part is incorrect. there will be a slow build up in the condenser, plausible if the operator has a misconception on sensitivity of the detectors. The second part is incorrect, plausible if the student has the misconception that the majority of what the detector is sensing is N-16 due to the steam system acting like a delay loop.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible if the student has the misconception that the majority of what the detector is sensing is N-16 due to the steam system acting like a delay loop.

Learning Objective:

Given access to the site Specific Simulator or specific plnt coditons, resons to the following events:

a. Reactor Coolant System leakage

b. Reactor Coolant Pump malfunction

c. Steam Generator Tube leak (055.03.LP2438.004)

Tuesday, April 30, 2019 11:03:47 AM 141

QUESTIONS REPORT for 2019 NRC Exam Master

44. 2019 NRC 044/SYS/039A3.02/3.1/1-I/RO/BANK/883D195 SH 3/052.03.LP0021.003 Given the following:

Unit 1 is operating at Rated Thermal Power 1MS-2017, 'B' Main Steam Isolation Valve, failed CLOSED Which of the following is the expected response of the Turbine and Turbine Trip solenoids 20ET and 20AST to the MSIV closing?

A. Both the 20ET solenoid and the 20AST solenoid will energize to trip the turbine.

B. Neither the 20ET solenoid nor the 20AST solenoid will energize and the turbine will not trip.

C. 20ET solenoid will not generate a turbine trip signal and the 20AST solenoid will energize to trip the turbine.

D. 20ET solenoid will energize to trip the turbine and the 20AST solenoid will not generate a turbine trip signal.

RO Tier 2 Group 1 Source: Bank Question History:

None K/A:

039A3.02 Main and Reheat Steam System (MRSS)

Ability to monitor automatic operation of the MRSS, including: Isolation of the MRSS (Imp 3.1/3.5*)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what automatic operation causing isolation of the MRSS system will happen.

Cognitive Level:

Knowledge 1-I: The operator must recall the interlock which cause the turbine to trip.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:47 AM 142

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

883D195 Sh 3, Turbine Trip Signals Logic Diagram Rev 35 499B466 Sh 824, Elemtary Wiring Diagram Steam Isol and Atmos Dump Indication Rev 8 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 1 is operating at Rated Thermal Power 1MS-2017, 'B' Main Steam Isolation Valve, failed CLOSED.

Which of the following is the expected response of the Turbine Trip solenoids 20ET and 20AST to the MSIV closing?

A. Both the 20ET solenoid and the 20AST solenoid will energize to trip the turbine.

B. Neither the 20ET solenoid nor the 20AST solenoid will energize and the turbine will not trip.

C. 20ET solenoid will not generate a turbine trip signal and the 20AST solenoid will energize to trip the turbine.

D. 20ET solenoid will energize to trip the turbine and the 20AST solenoid will not generate a turbine trip signal.

Proposed answer: A Tuesday, April 30, 2019 11:03:47 AM 143

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The Main Steam Isolation Valves input to both the 20ET and 20AST relays.

A CORRECT: See above.

B INCORRECT: Plausible if the operator has a misconception of how the logics function.

C INCORRECT: Plausible if the operator has a misconception of how the logics function.

D INCORRECT: Plausible if the operator has a misconception of how the logics function.

Learning Objective:

DESCRIBE the interlocks, actuation setpoints, and permissives associated with major components and operations associated with the Turbine Protection Trip System.

(052.03.LP0021.003)

Tuesday, April 30, 2019 11:03:47 AM 144

QUESTIONS REPORT for 2019 NRC Exam Master

45. 2019 NRC 045/SYS/059A2.05/3.1*/3-SPK/RO/BANK/883D195 SH 10/031.02.LP0441.016 Given the following:

Unit 1 is at 70% of Rated Thermal Power A rupture of the main feedwater system occurs inside containment UPSTREAM of the SG Feedwater Check valve BOTH Steam Generator Feed Water Pumps will be tripped by a ___(1)___.

Entry conditions for ___(2)___ are met.

A. low-low steam generator level reactor trip EOP-0, Reactor Trip or Safety Injection with a transition to EOP-0.1, Reactor Tip Response, B. low-low steam generator level reactor trip EOP-0, Reactor Trip or Safety Injection with a transition to EOP-1.1, SI Termination, C. high containment pressure safeguards actuation EOP-0, Reactor Trip or Safety Injection with a transition to EOP-1.1, SI Termination, D. low steam generator level and high steam flow/feed flow mismatch AOP-2B, Feedwater System Malfunction with a branch to AOP-17A, Rapid Power Reduction, Tuesday, April 30, 2019 11:03:47 AM 145

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2012 Prairie Island Question 44 K/A:

059A2.05 Main Feedwater (MFW) System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Rupture in MFW suction or discharge line (Imp 3.1*/3.4*)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what impacts a rupture in the Main Feedwater system will have and what procedures will be used to mitigate it.

Cognitive Level:

Comprehension 3-SPK: The operator must understand the initial conditions, determine what automatic action will happen based on the initial conditions, and then recall the procedures that will be utilized to mitigate the event.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:47 AM 146

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

883D195 Sh 7, Safegurds Actuation Signals Logic Diagram Rev 25 883D195 Sh 10, Logic Diagram Feewater Control and Isolation Rev 17 EOP-0, Reactor Trip or Safety Injection Unit 1 Rev 66 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following conditions:

Unit 1 is at 70% power.

A rupture of the main feedwater system occurs inside containment UPSTREAM of the check valve.

BOTH Main Feed Water Pumps will be tripped by a ______.

Entry conditions for ______ are met.

A. low-low steam generator level reactor trip 1E-0 followed by 1ES-0.1, Reactor Tip Recovery, B. high containment pressure safeguards actuation 1E-0 followed by 1ES-0.2, SI Termination C. low-low steam generator level reactor trip 1E-0 followed by 1ES-0.2, SI Termination D. SGWLC system high steam flow/feed flow mismatch 1C1.4AOP 1, Rapid Power Reduction Unit 1, Proposed answer: B Tuesday, April 30, 2019 11:03:47 AM 147

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Feedwater temperature entering the steam generators at approximately 70% of Rated Thermal Power will be 425°F. This will quickly turn to steam in the containment, pressurizing the containment to the Safety Injection actuation pressure of 4.8 psig. With the break in the Main Feedwater system being in the containment and UPSTREAM of the check valve in the feedline, the steam generators can still be fed by the AFW system, and the leak will be effectively isolated once the steam generator feed pumps are tripped by the actuation of SI.

This means that the procedure path for event mitigation will be EOP-0, to EOP-1.1, as there will be no reason for the SI pumps to be running once the leak is isolated (steam generator feed pumps tripped).

A INCORRECT: The first part is incorrect, plausible, if the operator has the misconception this could cause the reactor to trip, and trip the steam generator feed pumps. The second part is incorrect, but plausible as is would be a correct procedure path based on the first part of the answer if an SI was not needed to trip the steam generator feed pumps.

B INCORRECT: The first part is incorrect, plausible, if the operator has the misconception this could cause the reactor to trip, and trip the steam generator feed pumps. The second part is correct given an SI.

C CORRECT: See above.

D INCORRECT: The first part is incorrect, plausible if the operator has a misconception that a mismatch when compbined with a SG low level will generate a reactor trip. The second part is incorrect, plausible as AOP-2B,and AOP-17A are called out when a Feedpump trip occurs and power is < 75%.

Learning Objective:

Given access to Site Specific Simulator, IMPLEMENT the EOPs to respond to a faulted or ruptured Steam Generator.

(031.02.LP0441.016)

Tuesday, April 30, 2019 11:03:47 AM 148

QUESTIONS REPORT for 2019 NRC Exam Master

46. 2019 NRC 046/SYS/061K2.02/3.7*/2-RI/RO/NEW/CSP-H1/051.05.LP0169.003 Given the following:

Unit 1 is in MODE 1, Unit 2 is in MODE 5 EDGs are aligned to their respective busses A Loss of All Off Site Power to BOTH units occurs G03, Emergency Diesel Generator fails to start, and cannot be started 2A-06, 4160 VAC Safeguards Bus is locked out When Unit 1 transitions to EOP-0.1, Reactor Trip Response the following occurs:

1P-29, Turbine Driven AFW pump fails catastrophically Unit 1 enters CSP-H.1, Response to Loss of Secondary Heat Sink Which of the following would be the FIRST available pump that the crew will use to restore feed to one or both steam generators?

(Assume no change in electric plant status)

A. P-38A, Standby SG Feed pump B. P-38B, Standby SG Feed pump C. 1P-53, Motor Driven AFW pump D. 2P-53, Motor Driven AFW pump Tuesday, April 30, 2019 11:03:47 AM 149

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

061K2.02 Auxiliary / Emergency Feedwater (AFW) System Knowledge of bus power supplies to the following: AFW electric drive pumps (Imp 3.7*/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine which pumps still have power based a LOOP complicated by additional losses.

Cognitive Level:

Comprehension 2-RI: The operator must understand the initial conditions, recall the major procedure flow path and determine which pump will be the first procedurally which will still have power to feed the steam generators given a LOOP and additional faults.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

PBE-7033, Simplified Electrical Power Distribution Diagram Rev 13 MDB 3.2.3 Panel 1B03, 480 V AC Unit 1 Rev 19 MDB 3.2.4 Panel 2B04, 480 V AC Unit 2 Rev 14 CSP-H.1, Response to Loss of Secondary Heat Sink Unit 1 Rev 44 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:47 AM 150

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The power supplies for the pumps are:

P-38A, Standby SG Feed pump 1B03 P-38B, Standby SG Feed pump 2B04 1P-53, Motor Driven AFW pump 1A06 2P-53, Motor Driven AFW pump 2A05 CSP-H.1 will attempt to restore feed in the following order 1P-29, Turbine Driven AFW pump (catastrophically failed) 1P-53, Motor Driven AFW pump (no power, G-03, EDG failure)

P-38A, Standby SG Feed pump is the correct answer P-38B, Standby SG Feed pump (no power, 2A06 bus lockout) 2P-53, Motor Driven AFW pump (has power, but P-38A would be utilized first)

A CORRECT: See above.

B INCORRECT: Plausible if the operator has the misconception of the power supply of this pump, thinking it is still powered.

C INCORRECT: Plausible if the operator has the misconception of the power supply of this pump, thinking it is still powered.

D INCORRECT: Plausible if the operator has the misconception of major actions of CSP-H.1, as it still has power, and Unit 2 is in MODE 5 and should not need it.

Learning Objective:

STATE the power supply for the following Auxiliary Feedwater System Components.

a. Electric-Driven Auxiliary Feedwater Pumps

b. Motor Operated Isolation Valves (052.05.LP0169.003)

Tuesday, April 30, 2019 11:03:47 AM 151

QUESTIONS REPORT for 2019 NRC Exam Master

47. 2019 NRC 047/SYS/062A1.03/2.5/2-DI/RO/BANK/TS 81/182.01.LP2519.001 Given the following:

The Crew is performing TS-81, Emergency Diesel Generator G01 Monthly Test While paralleling G01 to 1A-05, the following conditions are noted:

Sync Selector Switch for 1A52-60, G01 to 1A-05 breaker, is ON Synchroscope is rotating 10 RPM in the SLOW direction Running and Incoming Voltmeter read as shown below:

What must the operator do to match voltage and make the Synchroscope turn 2 to 5 RPM in the FAST direction?

Go to ___(1)___ on the G01 Diesel Generator Voltage Regulator to equal or slightly exceed bus voltage.

Go to ___(2)___ on the G01 Diesel Generator Governor control switch to make Synchroscope rotate properly.

(1) (2)

A. Raise Raise B. Lower Raise C. Raise Lower D. Lower Lower Tuesday, April 30, 2019 11:03:47 AM 152

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2017 PBNP Question 47 Previous 2 NRC Exams K/A:

062A1.03 A.C. Electrical Distribution Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies (Imp 2.5/2.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine how to manipulate controls while switching power supplies.

Cognitive Level:

Comprehension 2-DR: The operator must recall understand the initial conditions, what adjusting the governor control and voltage regulator will do the EDG, then determine which way and what to adjust to get the desired results which will meet the limitation and requirements to allow the source to be shifted.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:47 AM 153

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 81, Emergency Diesel Generator G01 Monthly Test, Rev 92, Step 5.29 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

The Crew is performing TS-81, Emergency Diesel Generator G01 Monthly Test While paralleling G01 to 1A-05, the following conditions are noted:

Sync Selector Switch for 1A52-60, G01 to 1A-05 breaker, is ON Synchroscope is rotating 10 RPM in the SLOW direction Running and Incoming Voltmeter read as shown below:

What must the operator do to match voltage and make the Synchroscope turn 2 to 5 RPM in the FAST direction?

Go to ___(1)___ on the G01 Diesel Generator Voltage Regulator to equal or slightly exceed bus voltage.

Go to ___(2)___ on the G01 Diesel Generator Governor control switch to make Synchroscope rotate properly.

(1) (2)

A. Raise Raise B. Lower Raise C. Raise Lower D Lower Lower Proposed answer: B Tuesday, April 30, 2019 11:03:47 AM 154

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Running Voltage is the voltage of the bus, in this case. Incoming Voltage is the EDG voltage. EDG voltage needs to be lowered to match bus, thus Lower voltage on G01 make Incoming match Running. If the synch scope is going FAST in the SLOW direction, the EDG is not turning fast enough, the governor adjust will need to be raised to make the scope turn slow in the fast direction.

A INCORRECT: Plausible if the student has the misconception of which is running and which is incoming or the correct action to be taken.

B CORRECT: See above.

C INCORRECT: Plausible if the student has the misconception of which is running and which is incoming or the correct action to be taken and on what the actions are necessary to have the sync scope rotate in the correct direction.

D INCORRECT: Plausible this is the correct action to be taken to adjust EDG voltage regulator and if the student has a misconception on the actions necessary to have the sync scope rotate in the correct direction.

Learning Objective:

Given access to the PBNP Site Specific Simulator, the trainee should be able to OPERATE the major components of the specified system in accordance with acceptable practices and procedures.

(182.01.LP2519.001)

Tuesday, April 30, 2019 11:03:47 AM 155

QUESTIONS REPORT for 2019 NRC Exam Master

48. 2019 NRC 048/SYS/063A2.01/2.5/3-SPK/RO/NEW/ARB 12C20 A 1-3/055.03.LP3456.002 Given the following:

Both units are at Rated Thermal Power Annunciator 2C20 A 1-3, D-107 BATTERY CHARGER TROUBLE alarms Annunciator 2C20 A 2-2, D-01/D-03 125V DC BUS UNDER/OVER VOLTAGE alarms The AO reports the GROUND light is lit on D-107 Below is the panel indication from 2C20:

Which of the following answers (1) What protective function is the cause for the indications?

AND (2) What actions are required per the alarm response procedure?

A. (1) D-107 Battery Charger DC Breaker has tripped open (2) Line up D-109, Battery Charger to D-105, Battery B. (1) D72-03-03, Feed to D-03 from D-105 Station Battery fuses blown (2) Line up D-109, Battery Charger to D-105, Battery C. (1) D-107 Battery Charger DC Breaker has tripped open (2) Line up D-107, Battery Charger to D-305, Swing Battery D. (1) D72-03-03, Feed to D-03 from D-105 Station Battery fuses blown (2) Line up D-107, Battery Charger to D-305, Swing Battery Tuesday, April 30, 2019 11:03:48 AM 156

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

063A2.01 D.C. Electrical Distribution Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Grounds (Imp 2.5/3.2*)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what automatic actions were caused by the ground based on given indications, and what procedural actions will be needed to mitigate the ground.

Cognitive Level:

Comprehension 3-SPK: The operator must understand the initial conditions and determine was automatic action has been caused, and how to restore the system.

10 CFR Part 55 Content:

55.41 5 55.43 5

Reference:

ARP 2C20 A 1-3, D-107 Battery Charger Trouble Rev 6 ARP 2C20 A 2-2, D-01/D-03 125V DC Bus Under/Over Voltage Rev 6 0-TS-EP-001, Weekly Power Availability Verification Rev 23 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:48 AM 157

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With a possible ground indicated on D-107 charger, as indicated by annunciator 2C20 A 1-3, and confirmed by the AOs report, followed by the next annunciator 2C20 A 2-2 and panel readings, indicates that the discharge breaker for charger D-107 has tripped open. Actions required to restore or line up a charger to the D03, would be to line up D-109 Battery charger to supply the bus and battery.

A CORRECT: See above.

B INCORRECT: The first part in incorrect, plausible if the operator has the misconception of the charger powering the bus directly, and this breaker will cause one of the two annunciators (undervoltage),

but will not cause the other (charger trouble). The second part is correct.

C INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has a misconception the the ground is located on the battery, this is a physically possible lineup which would line up a battery to D-03.

D INCORRECT: The first part in incorrect, plausible if the operator has the misconception of the charger powering the bus directly, and this breaker will cause one of the two annunciators (undervoltage),

but will not cause the other (charger trouble).The second part is incorrect, plausible if the operator has a misconception the the ground is located on the battery, this is a physically possible lineup which would line up a battery to D-03.

Learning Objective:

Given access to the site specific simulator or specific plant conditions, RESPOND to the following:

Loss of a DC bus Loss of an Instrumentation Bus (055.03.LP3456.002)

Tuesday, April 30, 2019 11:03:48 AM 158

QUESTIONS REPORT for 2019 NRC Exam Master

49. 2019 NRC 049/SYS/064K6.07/2.7/1-I/RO/BANK/FSAR/054.02.LP0133.002 Given the following:

Unit 1 is at Rated Thermal Power A relief valve has failed open on T-170A, G03 EDG Starting Air Receiver The leakage exceeds the capacity of the G03 EDG Starting Air Compressor The AO is responding to the associated alarm Which of the following describes the response of G03, Emergency Diesel Generator if a start signal occurs before any operator action is taken?

A. It will start in the normal time from Starting Air Receivers T-170C and T-170D via all four starting air motors.

B. It will take longer to start or may NOT start because the starting air system will be depressurizing/depressurized.

C. It will take longer to start or may NOT start because the fuel rack fails to the zero fuel position and may not reset.

D. It will start in the normal time via the two starting air motors associated with the Starting Air Receiver T-170C and T-170D.

RO Tier 2 Group 1 Source: Bank Question History:

2012-12 Diablo Canyon Question 19 K/A:

064K6.07 Emergency Diesel Generators (ED/G)

Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers (Imp 2.7/2.9)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what impact the loss of the air receivers will have on the EDG system.

Cognitive Level:

Knowledge 1-I: The operator must recall the design and function of the emergency diesel generator air starting system.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 159

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

FSAR, Final Safety Analysis Report Rev USFAR 2018 Section 8.8, Page 8.8-3 of 11 Proposed reference to be provided to the applicants during examination:

None Original Question:

GIVEN:

Unit 1 is at 100% power.

A relief valve has failed open on Starting Air Receiver A for Diesel Generator 1-1.

The leakage exceeds the capacity of the starting air compressor.

The NO is responding to the associated alarm.

Which of the following describes the response of Diesel Generator 1-1 if a start signal occurs before any operator action is taken?

A. It will start in the normal time from Starting Air Receiver B via all four starting air solenoids.

B. It will start in the normal time via the two starting air solenoids associated with the Starting Air Receiver B.

C. It will take longer to start or may NOT start because the starting air system will be depressurizing/depressurized.

D. It will take longer to start or may NOT start because the fuel rack fails to no fuel position and may not reset.

Proposed answer: B Tuesday, April 30, 2019 11:03:48 AM 160

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The Train B EDGs (G03 and G04) are automatically started by two pairs of air motors. Each engine has its own independent starting system, including two banks of two air storage tanks. Each bank of air receivers has sufficient storage to crank the engine five times for the normal cranking duration. The starting air systems are completely redundant for each diesel generator A INCORRECT: Both air start systems are independent, plausible if the operator has a misconception of the design of the air start system and thinks both air banks feed all 4 air start motors.

B INCORRECT: The air start systems are able to be cross-connected, but are not normal in that configuration, plausible if the operator incorrectly thinks the air systems are cross-connected.

C INCORRECT: The air system has no effect on the fuel system, plausible if the operator has a misconception of the starting sequence and function of the air start system.

D CORRECT: See above.

Learning Objective:

DISCUSS the Diesel Generator and support systems associated with the Diesel Generators. Discussion of this drawing should include system flowpaths, major components and their physical location, and interfaces with other major systems:

a. Safeguards AC Electrical Distributions system

b. DC Electrical Distribution system

c. Fuel Oil system

d. Lubricating Oil systems

e. Air systems Engine Cooling (054.02.LP0133.002)

Tuesday, April 30, 2019 11:03:48 AM 161

QUESTIONS REPORT for 2019 NRC Exam Master

50. 2019 NRC 050/SYS/073K4.01/4.0/1-I/RO/BANK/RMSASRB CI RE-214/051.04.LP0052.004 What system condition will automatically shut WG-0014, Gas Decay Tank Radiation Control Valve, terminating a Gas Decay Tank discharge?

A. Gas Decay Tank reaches the low pressure setpoint.

B. High Radiation Signal on Auxiliary Building Ventilation Stack.

C. High Volumetric Air Flow at Auxiliary Building Ventilation Stack.

D. Air Flow through Auxiliary Building Ventilation Stack reverses direction.

RO Tier 2 Group 1 Source: Bank Question History:

None K/A:

073K4.01 Process Radiation Monitoring (PRM) System Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: Release termination when radiation exceeds setpoint (Imp 4.0/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what relationship between a high radiation signal and the termination of a release.

Cognitive Level:

Knowledge 1-I: The operator must recall the interlocks associated with the waste gas and radiation monitoring systems.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 162

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

RMSASRB CI RE-214, Auxiliary Building Vent Exhaust Gas Monitor Rev 9 Proposed reference to be provided to the applicants during examination:

None Original Question:

What system condition will automatically shut WG-0014, Gas Decay Tank Radiation Control Valve during a Gas Decay Tank discharge?

A. Gas Decay Tank reaches the low pressure setpoint.

B. High Radiation Signal on Auxiliary Building Ventilation Stack.

C. High Volumetric Air Flow at Auxiliary Building Ventilation Stack.

D. Air Flow through Auxiliary Building Ventilation Stack reverses direction.

Proposed answer: B Justification:

A high radiation signal will cause the vent gas release valve to shut if open.

A INCORRECT: This will not cause the valve to shut, plausible if the operator has a misconception of the how the system functions as assumes an auto shutoff feature based on tank pressure.

B CORRECT: See above.

C INCORRECT: The system will shift to the carbon filter, but a high volumetric flow will not cause the valve to shut, plausible if the operator has a misconception of the system function.

D INCORRECT: The system will shift to the carbon filter, but a high volumetric flow will not cause the valve to shut, plausible if the operator has a misconception of the system function.

Learning Objective:

DESCRIBE the interlocks associated with the Waste Gas System (051.04.LP0052.004)

Tuesday, April 30, 2019 11:03:48 AM 163

QUESTIONS REPORT for 2019 NRC Exam Master

51. 2019 NRC 051/SYS/073G2.4.47/4.2/2-DR/RO/NEW/OI 140B/051.04.LP0063.005 Given the following:

Waste Distillate Tank A is being discharged overboard via RE-223, per OI 140B, Standard Radioactive Batch Liquid Release - Waste Distillate Tanks Circ water is NOT lined up for ice melt PPCS indications for following are attached: (4 pages)

RE-223, Waste Distillate Overboard Liquid monitor 1RE-229, Unit 1 Service Water Discharge monitor 2RE-229, Unit 2 Service Water Discharge monitor Which of the following actions should be taken in response to these indications?

A. Shut WL-18, Waste Condensate Overboard Discharge to SW Header Control.

B. Shut BE-LW-15, Waste Distillate Overboard Discharge Flow Control and secure the discharge line up per OI 140B.

C. BE-LW-15, Waste Distillate Overboard Discharge Flow Control will automatically close, verify discharge flow has terminated.

D. Throttle Waste Distillate flow using BE-LW-15, Waste Distillate Overboard Discharge Discharge Flow Control, to clear the RE-229 alarm.

Tuesday, April 30, 2019 11:03:48 AM 164

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

073G2.4.47 Process Radiation Monitoring (PRM) System Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(Imp 4.2/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to use trends to diagnose an issue with a waste distillate tank and what actions are necessary based on that diagnosis.

Cognitive Level:

Comprehension 2-RI: The examinee must understand the given indications and trends, that actions need to be taken, and recall what action must be taken.

10 CFR Part 55 Content:

55.41 10 55.43 5

Reference:

OI 140B, Standard Radioactive Batch Liquid Release - Waste Distillate Tanks Rev 12 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:48 AM 165

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Based on the trend for 2RE-229, and the individual reading, the radiation monitor has reached an alert level, and per OI-140B, the discharge line must be secured starting with shutting the point of discharge, which in this case is BE-LW-15, Waste Distillate Overboard Discharge flow control valve.

A INCORRECT: Plausible because this would be a correct action if the discharge path was via RE-218, and it has an auto close feature based on RE-218 reading.

B CORRECT: See above.

C INCORRECT: Plausible, has a misconception on which radiation monitor will cause an automatic operation, because RE-223 has an auto close feature based on RE-233 readings.

D INCORRECT: Procedural directed action is to terminate the discharge, plausible because reducing discharge flow may have the desired effect of clearing the RMS alarm and the examinee may believe this is a reasonable first action to address the high RMS alarm.

Learning Objective:

IDENTIFY and DESCRIBE the controls, alarms, and indications associated with the Liquid Waste Disposal System, including:

a. Location and function of component and/or system operating controls and control stations

b. Alarming locations and response to major system and component alarms

c. Plant, system, and component conditions or permissives required for operation

d. Setpoints associated with major system alarms and/or interlocks (051.04.LP0063.005)

Tuesday, April 30, 2019 11:03:48 AM 166

QUESTIONS REPORT for 2019 NRC Exam Master

52. 2019 NRC 052/SYS/076K2.08/3.1*/2-DR/RO/BANK/MDB 3.2.6 2B32/051.06.LP0086.004 Given the following:

Both units are in MODE 1 A Unit 2 reactor trip occurs and safety injection and containment isolation are MANUALLY actuated 2X-04, LV Station Auxiliary Transformer locked out The crew is responding per EOP-0, Reactor Trip or Safety Injection G04, emergency Diesel Generator failed to start EOP-0 Attachment A, Automatic Action Verification has just been completed Considering Service Water ONLY; which of the following describes the plant response?

(2SW-2907 and 2908, Containment Ventilation Cooler Emergency Flow Control Valves)

A. 2SW-2907 and 2SW-2908 BOTH remained shut.

B. 2SW-2907 and 2SW-2908 BOTH automatically opened.

C. 2SW-2907 automatically opened and 2SW-2908 remains shut.

D. 2SW-2907 remains shut and 2SW-2908 automatically opened.

Tuesday, April 30, 2019 11:03:48 AM 167

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

2012 PBNP NRC Question 53 K/A:

076K2.08 Service Water System (SWS)

Knowledge of bus power supplies to the following: ESF-actuated MOVs (Imp 3.1*/3.3*)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the relationship between ESF actuated MOVs and the power supplies, with a lockout and manual operation of the SI pushbuttons.

Cognitive Level:

Comprehension 2-DR: The operator must recall the relationship between 2X04, G04, and the ESF operated SW valves, given a 2X04 lockout and diesel failure with manual SI actuation.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 168

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

MDB 3.2.6 2B32, 480 V AC Motor Control Centers Rev 23 MDB 3.2.6 2B42, 480 V AC Motor Control Centers Rev 22 883D195 Sh 9 Safeguard Sequence Logic Rev 19 Proposed reference to be provided to the applicants during examination:

None Original Question:

Consider the following conditions:

- Both units are in MODE 1

- A Unit 2 reactor trip occurs and safety injection is MANUALLY actuated

- 2X04 LV Station Auxiliary Transformer locked out

- The crew is responding per EOP-0, 'Reactor Trip or Safety Injection'

- G-04 EDG failed to start

- EOP-0 Attachment 'A', Automatic Action Verification has just been completed Considering Service Water ONLY; which of the following describes the plant response?

(2SW-2907 and 2908, Containment Ventilation Cooler Emergency Flow Control Valves)

A. 2SW-2907 and 2SW-2908 BOTH automatically opened.

B. 2SW-2907 and 2SW-2908 BOTH remained shut.

C. 2SW-2907 automatically opened and 2SW-2908 remains shut.

D. 2SW-2907 remains shut and 2SW-2908 automatically opened.

Proposed answer: C Tuesday, April 30, 2019 11:03:48 AM 169

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

A high radiation signal will cause the vent gas release valve to shut if open.

A INCORRECT: One of the valves will remain shut, plausible if the operator has a misconception of the MOV power supplies and determines that due to the loss of 2X04 and the diesel that neither will have power.

B INCORRECT: This is the normal response when both buses have power, plausible if the operator has a misconception of the MOV power supplies and that this is the normal operation of the system when both buses have power.

C CORRECT: See above.

D INCORRECT: This the opposite of what will happen, plausible if the operator reverses the power supplies for the valves.

Learning Objective:

STATE the power supply for the Service Water Pumps and Motor Operated Valves IAW the Master Data Book.

(051.06.LP0086.004)

Tuesday, April 30, 2019 11:03:48 AM 170

QUESTIONS REPORT for 2019 NRC Exam Master

53. 2019 NRC 053/SYS/076A3.02/3.7/2-DR/RO/BANK/OR-01/051.06.LP0086.005 Given the following:

Both Units are in MODE1 with a normal electric plant alignment P-32A, B, and D Service Water Pumps are running The following sequence of events occurs:

2X-04, Low Voltage Station Auxiliary Transformer LOCKS OUT Diesel generators FAIL to automatically restore power to safeguards buses Operators manually energize Unit 2 'B' train safeguards buses using the associated standby emergency power source Which of the following describes the expected Service Water Pumps running before and after the operators close the diesel output breaker?

Before After A. A A, B, D & E B. A & B A, B, E & F C. A & B A, B, C, D & E D. A, B & D A, B, C, D & F Tuesday, April 30, 2019 11:03:48 AM 171

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: Bank Question History:

None K/A:

076A3.02 Service Water System (SWS)

Ability to monitor automatic operation of the SWS, including: Emergency heat loads (Imp 3.7/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the automatic operation of the service water pumps, to ensure sufficient cooling is available for emergency heat loads before and after the closure of the diesel output breakers.

Cognitive Level:

Comprehension 2-DR: The operator must understand the initial conditions and electrical lineup, determine which pumps will be running due to the event, and then determine which pumps will be running after operator actions are taken.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 172

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

MDB 3.2.3 Panel 1B03, 480 V AC Unit 1 Rev 19 MDB 3.2.3 Panel 1B04, 480 V AC Unit 1 Rev 15 MDB 3.2.4 Panel 2B03, 480 V AC Unit 2 Rev 14 MDB 3.2.4 Panel 2B04, 480 V AC Unit 2 Rev 14 883D195 Sh 8 Safeguard Sequence Logic Rev 19 110E163 Sh 12A, Schematic Diagram SI Logic ESF Systems Train A Reactor Safeguards Rev 21 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Both Units are operating in MODE1 with a normal electric plant alignment P-32A, B, and D Service Water Pumps are running The following sequence of events occurs:

2X-04, LOCKS OUT Diesel generators FAIL to automatically restore power to safeguards buses Operators manually energize Unit 2 'B' train safeguards buses using the associated standby emergency power source Which of the following describes the expected Service Water Pumps running before and after the operators close the diesel output breaker?

Before After A. A A, B, D & E B. A & B A, B, E & F C. A & B A, B, C, D & E D. A, B & D A, B, C, D & E Proposed answer: C Justification:

Service Water Pump Normal power supplies:

P32A - 1B03 P32B - 1B03 P32C - 1B04 P32D - 2B04 P32E - 2B04 P32F - 2B03 Service Water Pump Alternate power supplies:

Tuesday, April 30, 2019 11:03:48 AM 173

QUESTIONS REPORT for 2019 NRC Exam Master P32B - B08 P32C - B09 P32E - B09 P32F - B08 Auto start signals G01 or G02 EDG breaker closure - A, B, F Pumps G03 or G04 EDG breaker closure - C, D, E Pumps SI- Pumps A & C start @ 15.5 sec, B & D @ 20.5 sec, E & F@25.75 sec With the initial conditions, pump P-32A and B will still have power and after breaker closure which will cause the auto start of P-32C, D, and E, which will all have power still supplied. P-32F will have lost power and not received an auto start signal.

A INCORRECT: The first part is incorrect, P-32B will remain running, plausible if the operator has a misconception of the power supply. The second part is incorrect, P-32C will start when G04 output breaker is shut and is powered from 1B04, plausible if the operator has a misconception about the power supplies, and/or auto start features of the SW pumps.

B INCORRECT: The first part is incorrect, P-32F will remain not be run, plausible if the operator has a misconception of the power supply. The second part is incorrect, P-32C will start when G04 output breaker is shut and is powered from 1B04, P-32D also has power and will receive an auto start signal, plausible if the operator has a misconception about the power supplies, and/or auto start features of the SW pumps C CORRECT: See above.

D INCORRECT: The first part is incorrect, P-32D will lose power and not be running, plausible if the operator has a misconception on the power supply. The second part is correct.

Learning Objective:

DESCRIBE the interlocks associated with the Service Water System and its major components:

a. Service Water Pumps

b. Service Water Screens

c. Service Water Zurn Strainers (051.06.LP0086.005)

Tuesday, April 30, 2019 11:03:48 AM 174

QUESTIONS REPORT for 2019 NRC Exam Master

54. 2019 NRC 054/SYS/078K3.03/3.0/2-DR/RO/NEW/AOP-5B/055.03.LP2349.005 Given the following:

Unit 1 is in MODE 4, cooling down to MODE 5 1P-1A, RCP, is running 1P-10A, RHR pump, is operating in decay heat removal mode Unit 2 is at Rated Thermal Power Subsequently:

A Loss of Off Site power occurs The crew is unable to restart any Instrument or Service air compressors The crew is implementing AOP-5B, Loss of Instrument Air in addition to other procedures Unit 1 RCS temperature is 280°F and LOWERING SLOWLY Unit 2 Reactor tripped In accordance with AOP-5B, which of the following identifies the action(s) to take for controlling Unit 1 and Unit 2?

UNIT 1 UNIT 2 A. Start and stop the RHR pump Take MANUAL control of Main Feed Regulating valves and restore SG levels to program B. Start and stop the RHR pump Ensure 2P-29, TDAFW pump is running and control flow with 2AF-4000 and 2AF-4001, AFW Pump Discharge valves C. Locally adjust 1RH-716A, Take MANUAL control of Main Feed RH-624 Flow Control Outlet valve Regulating valves and restore SG levels to program D. Locally adjust 1RH-716A, Ensure 2P-29, TDAFW pump is running RH-624 Flow Control Outlet valve and control flow with 2AF-4000 and 2AF-4001, AFW Pump Discharge valves Tuesday, April 30, 2019 11:03:48 AM 175

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 1 Source: New Question History:

None K/A:

078K3.03 Instrument Air System (IAS)

Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Cross-tied units.

(Imp 3.0/3.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the actions necessary during a loss of instrument air for cross-tied units.

Cognitive Level:

Comprehension 2-DR: The operator must understand the initial conditions and determine the impact of the loss of instrument air, and determine what actions are necessary during a loss of instrument air given differing plant conditions.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

AOP-5B, Loss of Instrument Air Rev 49 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:48 AM 176

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The loss of off-site power will cause Unit 2 to trip. With the inability to restart any instrument or service air compressors, the instrument and service air headers will depressurize. With the depressurization, control of the Unit 1 cooldown will be accomplished by local operation of 1RH-716A, due to the 1RH-642 failing open. Control of the AFW flow to Unit 2 will be accomplished by ensuring the TDAFW pump is running and controlling flow using the discharge valves. 2P-53, MDAFW pump will not be utilized due to the discharge valves are air operated and fail open, so there would be no way to control the feed rate to the steam generators.

A INCORRECT: The first part is incorrect, plausible as this method will work, but is not procedurally allowed. The second part is incorrect, with the main feed regulating valves losing air, they will fail shut, taking manual control will not prevent this. Plausible if the operator has a misconception of the valve operation.

B INCORRECT: The first part is incorrect, plausible as this method will work, but is not procedurally allowed. The second part is correct.

C INCORRECT: The first part is correct. The second part is incorrect, with the main feed regulating valves losing air, they will fail shut, taking manual control will not prevent this. Plausible if the operator has a misconception of the valve operation.

D CORRECT: See above.

Learning Objective:

Given access to the Site Specific Simulator, APPLY the appropriate guidance provided in the applicable AOPs for various system/component malfunctions.

(055.03.LP2439.005)

Tuesday, April 30, 2019 11:03:48 AM 177

QUESTIONS REPORT for 2019 NRC Exam Master

55. 2019 NRC 055/SYS/103K3.03/3.7/1-P/RO/BANK/CL 1E/051.05.LP0099.004 How is containment closure capability ensured during a refueling outage when RCS temperature is <200°F and movement of irradiated fuel assemblies for core reload is in progress?

A. Containment closure capability is not required below 200°F as long >23 feet of water above the reactor vessel flange is maintained.

B. CL 1E, Containment Closure Checklist, is maintained by the control room tracking all open penetrations.

C. Personnel and equipment hatch interlocks mechanisms are maintained in an operable status.

D. The Containment Isolation system and components are maintained operable.

RO Tier 2 Group 1 Source: Bank Question History:

2011 PBNP NRC Question 55 K/A:

103K3.03 Containment System Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under refueling operations.

(Imp 3.7/4.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the method which containment integrity is maintained during refueling.

Cognitive Level:

Knowledge 1-P: The operator must recall the method to maintain containment integrity during refueling.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 178

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TRM 3.9.3 Containment Penetrations Rev 2 CL 1E, Containment Closure Checklist Unit 1 Rev 28 Proposed reference to be provided to the applicants during examination:

None Original Question:

How is containment closure capability ensured during a refueling outage when RCS temperature is <200°F and movement of irradiated fuel assemblies for core reload is in progress?

A. Containment closure capability is not required below 200°F as long as we maintain >23 feet of water above the vessel flange.

B. CL-1E, Containment Closure Checklist is maintained by the control room tracking all open penetrations.

C. Personnel and equipment hatch interlocks mechanisms are maintained in an operable status.

D. The Containment Isolation system is operable.

Proposed answer: B Tuesday, April 30, 2019 11:03:48 AM 179

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

CL 1E is established and maintained by the Control Room as a list of containment penetrations, including containment airlocks and the fuel transfer tube, which must be isolated in the event of a fuel handling accident or other fission product release with RCS temperature is less than 200°F and containment integrity is not maintained.

A INCORRECT: Plausible as this is a Tech Spec requirement to be maintained, for refueling, and containment integrity is not always necessary, but the conditions listed in the stem do requirement containment closure.

B CORRECT: See above.

C INCORRECT: Plausible as hatch interlock will be maintained to ensure containment closure capability, but not less than 200°F.

D INCORRECT: Plausible as Containment Isolation is required when in modes 1-4, but not during refueling, and if it was operable, it would shut most closures to the containment.

Learning Objective:

DESCRIBE the procedures which govern operation of the Containment Structure. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non Licensed Operators.

(051.05.LP0099.004)

Tuesday, April 30, 2019 11:03:48 AM 180

QUESTIONS REPORT for 2019 NRC Exam Master

56. 2019 NRC 056/SYS/001K2.02/3.6/1-F/RO/NEW/617F354 SH 5/053.02.LP0315.02 Which of the following correctly describes components in the power flow path to the Reactor Trip Breakers?

52/RTA and 52/RTB Reactor Trip Breakers B-01 and B-02, 480V Non-Safeguards bus B-03 and B-04, 480V Safeguards bus G-06 and G-07, Rod Control Motor Generator Set RDC-1AC/RDC-1BD/RDC-2AC, Rod Drive Control Power Cabinets RDC-Logic, Rod Drive Control Logic Circuit Cabinet A. B-01/B-02 Motor Generator Sets Rx Trip Breakers B. B-03/B-04 Motor Generator Sets Power/Logic Cabinets Rx Trip Breakers C. B-01/B-02 Motor Generator Sets Power/Logic Cabinets Rx Trip Breakers D. B-03/B-04 Motor Generator Sets Rx Trip Breakers RO Tier 2 Group 2 Source: New Question History:

None K/A:

001K2.02 Control Rod Drive System Knowledge of bus power supplies to the following: One-line diagram of power supply to trip breakers.

(Imp 3.6/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the one-line diagram of for the power supply to the reactor trip breakers.

Cognitive Level:

Knowledge 1-F: The operator must recall the one-line diagram for the power supply to the reactor trip breakers.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 181

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

617F354 Sh5A, Reactor Protection System Reactor Trip Breaker Switchgear Train A Unit 1 Schematic Diagram, Rev 5 617F354-2 Sh5A, Reactor Protection System Reactor Trip Breaker Switchgear Train A Unit 2 Schematic Diagram, Rev 5 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

Per 617F354 Sh 5, the 480 V bus B-01 and B-02 supply the CRDM MG sets then the Reactor trip breakers.

A CORRECT: See above.

B INCORRECT: The first part is incorrect, plausible as these are also safety related switchgear. The second part is correct. The third part is incorrect, plausible if the operator has the misconception that power goes to the power/logic cabinets prior to the reactor trip breakers.

C INCORRECT: The first part is correct. The second part is correct. The third part is incorrect, plausible if the operator has the misconception that power goes to the power/logic cabinets prior to the reactor trip breakers.

D INCORRECT: The first part is incorrect, plausible as these are also safety related switchgear. The second part is correct.

Learning Objective:

DESCRIBE the system operation to include electrical flowpaths, major components, signal generation, and Reactor Trip Breaker/Trip Bypass Breaker operation.

(053.02.LP0315.002)

Tuesday, April 30, 2019 11:03:48 AM 182

QUESTIONS REPORT for 2019 NRC Exam Master

57. 2019 NRC 057/SYS/011K4.06/3.3/1-F/RO/NEW/LP0078/051.01.LP0078.004 On a lowering Pressurizer level event at what pressurizer level will letdown automatically isolate and why does it isolate at this level?

A. 12% to prevent uncovering pressurizer heaters B. 12% to prevent pressurizer level transmitter reference leg flashing C. 20% to prevent uncovering pressurizer heaters D. 20% to prevent pressurizer level transmitter reference leg flashing RO Tier 2 Group 2 Source: New Question History:

None K/A:

011K4.06 Pressurizer Level Control System (PZR LCS)

Knowledge of PZR LCS design feature(s) and/or interlock(s) which provide for the following: Letdown isolation (Imp 3.3/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the interlock level where automatic letdown isolation occurs and what the function is of the automatic isolation.

Cognitive Level:

Knowledge 1-F: The operator must recall the interlock level and function.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

LP0078, Pressurizer Pressure and Level Control, Rev 15 DBD-25 Section B, NSSS Control Systems - Pressurizer Level Control (Section B) Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:48 AM 183

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

At 12% pressurizer level the interlock for automatic isolation for the letdown system is met, this is done to prevent uncovering the pressurizer heaters.

A CORRECT: See above.

B INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has the misconception that the reference leg will flash due to level being too low to maintain the required head/water level necessary to prevent flashing.

C INCORRECT: The first part is incorrect, plausible if the operator has a misconception of how the automatic isolation setpoint is developed, the 12% value for the isolation is derived from being 20% of gage span and also 20% is the low limit of the normally maintained pressurizer level from no-load to full-load conditions. The second part is correct.

D INCORRECT: The first part is incorrect, plausible if the operator has a misconception of how the automatic isolation setpoint is developed, the 12% value for the isolation is derived from being 20% of gage span and also 20% is the low limit of the normally maintained pressurizer level from no-load to full-load conditions. The second part is incorrect, plausible if the operator has the misconception that the reference leg will flash due to level being too low to maintain the required head/water level necessary to prevent flashing.

Learning Objective:

DESCRIBE the automatic functions and interlocks associated with the Pressurizer, Level Control, Pressure Control, and Relief System and its major components:

a. Pressurizer Pressure

b. Pressurizer level Low Temperature Over Pressure Protection (051.01.LP0078.004)

Tuesday, April 30, 2019 11:03:48 AM 184

QUESTIONS REPORT for 2019 NRC Exam Master

58. 2019 NRC 058/SYS/015K6.04/3.1/3-SPK/RO/BANK/LP2416/053.03.LP2416.002 Given the following:

A Unit 2 reactor startup is in progress Source range counts are at 2000 cps N-32, Source Range NI Instrument Power fuses blew Based on these indications, what is the status of the reactor core with no operator actions?

A. All Rod bottom lights should be lit B. No Rod position change should occur C. Control Bank Rod bottom lights only should be lit D. Shutdown bank Rod bottom lights only should be lit RO Tier 2 Group 2 Source: Bank Question History:

2014 DC Cook Question 58 K/A:

015K6.04 Nuclear Instrumentation System Knowledge of the effect of a loss or malfunction on the following will have on the NIS: Bistables and logic circuits (Imp 3.1/3.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what effect a loss of power to the bistables will have one the NI, and plant.

Cognitive Level:

Comprehension 3-SPK: The operator must recall what effect the loss of power will have on the instrument and bistables, and given the plant conditions, what effect it will have..

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 185

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

LP2416, Nuclear Instrumentation System Rev 6 Slide 52 VTM 00193-1 BOOK 2, Nuclear Instrumentation System - Book 1, Rev 42 Page 151 Proposed reference to be provided to the applicants during examination:

None Original Question:

A Unit 2 reactor startup is in progress with source range counts at 2000 cps when the following alarms come in on Annunciator #210 Response: Flux Rod:

Annunciator 210 Drop 3, "SOURCE RANGE HI FLUX AT SHUTDOWN" The reactor operator verifies that Source Range Channel N32 level indicates off-scale HIGH.

Based on these indications, what is the status of the reactor core with no operator actions?

a. All Rod bottom lights should be lit.

b. No Rod position change should occur.

c. Control Bank Rod bottom lights only should be lit.

d. Shutdown bank Rod bottom lights only should be lit.

Proposed answer: A Tuesday, April 30, 2019 11:03:48 AM 186

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Pulling INSTRUMENT POWER fuses will cause a loss of +20Vdc SCR gate voltage and will cause a trip signal to be generated unless level trip switch is in bypass or SR trips are blocked, which does not happen until power is greater than P-6.

A CORRECT: See above.

B INCORRECT: P-6 has not been met, so the SR trips cannot be blocked, plausible if the operator has a misconception about either P-6 or the coincidence for the SR high flux trip, or how a loss of power will affect the NI.

C INCORRECT: All rods will receive a trip signal, plausible if the operator has a misconception on the type of trip caused by the SR NIs.

D INCORRECT: All rods will receive a trip signal, plausible if the operator has a misconception on the type of trip caused by the SR NIs.

Learning Objective:

DRAW and DISCUSS a block diagram of the Source Range Nuclear Instrumentation System. Discussion should include electrical flow paths and major components.

a. Detector

b. Preamplifier and Power Supply

c. Pulse Shaper and Driver

d. Log Pulse Integrator

e. Level Amplifier

f. High Flux at Shutdown Bistable

g. Source Range High Flux Reactor Trip Bistables (053.03.LP2416.002)

Tuesday, April 30, 2019 11:03:48 AM 187

QUESTIONS REPORT for 2019 NRC Exam Master

59. 2019 NRC 059/SYS/016A2.01/3.0*/3-PEO/RO/BANK/883D195 SH 7/055.03.LP3455.001 Given the following:

The unit is at Rated Thermal Power PT-482, SG A Pressure (Blue) has failed high for an unknown reason and has been removed from service IAW 0-SOP-IC-001 Blue, Routine Maintenance Procedure Removal of Safeguards or Protection Sensor from Service - Blue While I&C was collecting comparison data from PT-468, SG A Pressure (Red) an inadvertent short caused PT-468 output to fail low Which of the following describes the expected crew response to this failure?

A. Place FIC-466, SG A MFRV Controller, HC-480, SG A Feed Reg Bypass controller and HC-468, SG A Atmos Steam Dump controller, in MANUAL per AOP-24, Response to Instrument Malfunctions. Remove PT-468 from service IAW 0-SOP-IC-001 Red.

B. Reduce load per AOP-17A, Rapid Power Reduction, to prevent overpower from excessive steam and feed flow, place HC-468, SG A Atmos Steam Dump controller, in MANUAL.

C. Enter EOP-0, Reactor Trip or Safety Injection, and subsequently transition to EOP-0.1, Reactor Trip Response.

D. Enter EOP-0, Reactor Trip or Safety Injection, and subsequently transition to EOP-1.1, SI Termination.

Tuesday, April 30, 2019 11:03:48 AM 188

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 2 Source: Bank Question History:

2005 PBNP Question 59 K/A:

016A2.01 Non-Nuclear Instrumentation System (NNIS)

Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure.

(Imp 3.0*/3.1*)

Justification for K/A Match:

Matches the K/A by requiring the operator to diagnose the result of a detector failure, and which procedures would mitigate the consequences of the malfunction.

Cognitive Level:

Comprehension 3-PEO: The operator must understand the initial conditions and determine what effect the second detector failure will affect the plant, and what procedures are required to mitigate the effect.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:48 AM 189

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-0, Reactor Trip of Safety Injection Rev 66 0-SOP-IC-001 Blue, Routine Maintenance Procedure Removal of Safeguards or Protection Sensor from Service - Blue Rev 16 883D195 Sh 7, Safegurds Actuation Signals Logic Diagram Rev 25 Proposed reference to be provided to the applicants during examination:

None Original Question:

Consider the following Unit 1 conditions:

- PT-482, Steam Generator Pressure Loop A (Blue) channel has failed for an unknown reason and has been removed from service IAW 0-SOP-IC-001 Blue, Removing Safeguards or Protection Sensor from Service.

- While I&C was collecting comparison data from PT-468, Steam Generator Pressure Loop A (Red) channel, an inadvertent short caused PT-468 output to fail low.

Which of the following describes the expected crew response to this failure?

A. Place HC-466, SG 'A' MFRV Controller, HC-480, SG 'A' MFRV Bypass controller and HC-468, SG 'A' Atmospheric Controller, in manual per AOP-24, Response to Instrument Malfunctions. Remove PT-468 from service IAW 0-SOP-IC-001 Red.

B. Enter EOP-0, Reactor Trip or Safety Injection, and subsequently transition to EOP-0.1, Reactor Trip Response.

C. Enter EOP-0, Reactor Trip or Safety Injection, and subsequently transition to EOP-1.1, SI Termination.

D. Reduce load per AOP-17A, Rapid Power Reduction, to prevent overpower from excessive steam and feed flow.

Proposed answer: C Tuesday, April 30, 2019 11:03:48 AM 190

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With PT-482 removed from service, the failure of a second SG press instrument, this will cause an SI, therefore entry to EOP-0, then transition to EOP-1.1.

A INCORRECT: Plausible because these actions would be correct if PT-482 was in service at the time of the inadvertent short.

B INCORRECT: Plausible, because these actions would be correct for PT-482, which was already taken out of service.

C INCORRECT: Plausible because the instrument failure will cause a trip, but also an SI, so transition to trip response would be incorrect.

D CORRECT: See above.

Learning Objective:

DESCRIBE the plant and operator(s) response to the following conditions:

a. Loss of PPCS

b. Loss of Instrumentation

c. Inadvertent Safety Injection Actuation Inadvertent Containment Isolation Actuation (055.03.LP3455.001)

Tuesday, April 30, 2019 11:03:48 AM 191

QUESTIONS REPORT for 2019 NRC Exam Master

60. 2019 NRC 060/SYS/034A4.02/3.5/1-P/RO/MODIFIED/TS 3.9.2/SD85.1 2.1.40 Given the following:

The unit is in REFUELING MODE Fuel movement in Containment is in progress Concerning the Source Range nuclear instrumentation, as a MINIMUM, the Control Operator must verify and monitor which of the following per Technical Specifications AND RP 1C, Refueling ?

___(1)___ operating, visually monitored in the Control Room, and has audible indication in the ___(2)___.

A. (1) ONE channel is (2) Containment ONLY B. (1) ONE channel is (2) Control Room AND Containment C. (1) TWO channels are (2) Containment ONLY D. (1) TWO channels are (2)Control Room AND Containment Tuesday, April 30, 2019 11:03:48 AM 192

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 2 Source: Modified Question History:

None K/A:

034A4.02 Fuel Handling Equipment System (FHES)

Ability to manually operate and/or monitor in the control room: Neutron levels.

(Imp 3.5/3.9)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what is required to be monitored in the control room during fueling handling concerning neutron levels.

Cognitive Level:

Knowledge 1-P: The operator must recall the requirements for source range nuclear instrument during a refueling condition.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 193

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 3.9.2, Refueling Operation - Nuclear Instrumentation Rev 2 RP 1C, Refueling Rev 81, Section 4.5, 4.6, 5.14.1, and Attachment C Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

The plant is in REFUELING Mode Fuel movement in Containment is in progress What is the responsibility of the Control Operator concerning the Source Range nuclear instrumentation during this operation?

As a MINIMUM, the operator must verify . . .

A. ONE channel is operating, and it is visually monitored in the Control Room.

B. ONE channel is operating, it is visually monitored in the Control Room and has audible indication in the Control Room and Containment.

C. TWO channels are operating, each channel is visually monitored in the Control Room, and ONE channel has audible indication in Containment.

D. TWO channels are operating, each channel is visually monitored in the Control Room and ONE channel has audible indication in the Control Room.

Proposed answer: D Tuesday, April 30, 2019 11:03:48 AM 194

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The LCO states: Two source range neutron flux monitors shall be operable, and one source range audible count rate circuit shall be operable in the control room. RP-1C states: An audible signal output from N31 or N32 is OPERABLE and being heard both in the Control Room and Containment.

A INCORRECT: Plausible if the operator has a misconception of the source range requirements for refueling.

B INCORRECT: Plausible if the operator has a misconception of the source range requirements for refueling.

C INCORRECT: Plausible if the operator has a misconception of the source range requirements for refueling.

D CORRECT: See above.

Learning Objective:

Knowledge of refueling administrative requirements.

(SD86.1 2.1.40)

Tuesday, April 30, 2019 11:03:48 AM 195

QUESTIONS REPORT for 2019 NRC Exam Master

61. 2019 NRC 061/SYS/041G2.1.7/4.4/2-DR/RO/MODIFIED/STPT 5.2/052.02.LP0035.005 Given the following:

Unit 1 is at Rated Thermal Power when an inadvertent turbine trip causes a reactor trip.

CO3 is monitoring plant response to the trip and observes the following:

TAVG is 551°F and LOWERING Condenser Steam Dump Valves 1MS-2053 and 2057 indicate OPEN Annunciator 1C03 1E2 4-2, TAVG STEAM DUMP CHANNEL ALERT is LIT 1HFC-484, Cond Steam Dump Controller indicates approximately 25%

controller output Based on these observed indications the response of the condenser steam dumps (CSD) system is ___(1)___,

AND during performance of the TEMPERATURE CONTROL STEP of EOP-0, the operators should ___(2)___.

(1) (2)

A. normal continue monitoring for proper operation no additional action needed B. normal place Steam Dump Mode Selector switch in MAN C. abnormal place Steam Dump Mode Selector switch in MAN AND 1HFC-484 in MAN to close CSDs D. abnormal direct field operators to locally isolate CSDs by shutting manual isolation valves Tuesday, April 30, 2019 11:03:48 AM 196

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 2 Source: Modified Question History:

None K/A:

041G2.1.7 Steam Dump System (SDS) and Turbine Bypass Control Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(Imp 4.4/4.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to evaluate plant performance and system response using instrument interpretation and determine if operations are normal and what actions are necessary.

Cognitive Level:

Comprehension 2-DR: The student must understand the initial conditions, recall the interlocks of the SDS, then based on indications validate how the system is functioning and what actions are necessary.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:48 AM 197

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

STPT 5.2, Major Control Systems Setpoints - Steam Dump Control Rev 7 883D195 Sh 17, Steam Dump Control Logic Diagram Rev 20 EOP-0, Reactor Trip or Safety Injection, Rev 66 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 1 is operating at Rated Thermal Power when an inadvertent turbine trip causes a reactor trip.

CO3 is monitoring plant response to the trip and observes the following:

TAVG is 551°F and LOWERING Condenser Steam Dump Valves 1MS-2053 and 2057 indicate OPEN Annunciator 1C03 1E2 4-2, TAVG STEAM DUMP CHANNEL ALERT is LIT 1HFC-484, Cond Steam Dump Controller indicates approximately 25% controller output Based on these observed indications the response of the condenser steam dumps (CSD) system is ___(1)___, and operators should ___(2)___.

(1) (2)

A. normal continue monitoring for proper operation B. abnormal place Steam Dump Mode selector switch in MAN to close CSDs C. abnormal place Steam Dump Mode selector switch in MAN AND 1HFC-484 in MAN to close CSDs D. abnormal direct field operators to locally isolate CSDs by shutting manual isolation valves Proposed answer: A Tuesday, April 30, 2019 11:03:48 AM 198

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator must determine the TAVG Steam Dump Channel Alert annunciator being lit is not an abnormal indication, this annunciator will energize when the control switch is in auto, turbine tripped, and exceeded the blow open setpoint of the turbine bistables. After the completion of immediate actions, the operator will be directed to take the steam dump mode selector switch to manual based on the fact that the steam dumps are currently available.

A INCORRECT: The first part is correct. The second part is incorrect, but plausible if the operator has a misconception of the action taken in EOP-0 or does not recall that even with normal indication of a functional steam dump system, the control should be taken to manual based on no longer wanting to control the plant using TAVE.

B CORRECT: See above C INCORRECT: Plausible if the operator diagnoses a controller malfunction taking the Mode selector switch and CSD controller to MANUAL would be a possible course of action to address the CSD malfunction.

D INCORRECT: Plausible if the operator diagnoses a controller malfunction directing local operator actions would be a possible course of action to address the CSD malfunction.

Learning Objective:

RECOGNIZE and ASSESS the response of the Steam Dump System to:

a. Turbine Trip

b. Load Rejection Loss of Main Condenser Availability (052.02.LP0035.005)

Tuesday, April 30, 2019 11:03:48 AM 199

QUESTIONS REPORT for 2019 NRC Exam Master

62. 2019 NRC 062/SYS/045A3.07/3.5/2-DR/RO/BANK/STPT 15.1/052.03.LP0021.001 Given the following:

A Unit 1 Turbine Generator startup is in progress A failure in the electro-hydraulic control system (EHC) causes turbine speed to rise The EHC Auxiliary Governor actuates The Independent Overspeed Protection System (IOPS) actuates Turbine speed reaches 1895 rpm Which of the following lists the expected positions of the Turbine Stop and Governor valves one (1) minute after the Turbine Protection system has actuated?

STOP Valves GOVERNOR Valves A. OPEN OPEN B. OPEN SHUT C. SHUT OPEN D. SHUT SHUT Tuesday, April 30, 2019 11:03:48 AM 200

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 2 Source: Bank Question History:

None K/A:

45A3.07 Main Turbine Generator (MT/G)

Ability to monitor automatic operation of the MT/G system, including: Turbine stop/governor valve closure on turbine trip.

(Imp 3.5/3.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine a turbine trip has occurred, and then to determine what the position of the stop and governor valves after a turbine protection system actuation.

Cognitive Level:

Comprehension 2-DR: The operator must understand the initial conditions determine how the turbine protection system will respond, and then determine what the end state position of the stop and governor valves.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 201

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

STPT 15.1, Setpoint Document - Turbine Trips, Alarms, and AMSAC Rev 19, Section 1.1 LP0021, Turbine Protection Trip System Rev 14 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A Unit 1 Turbine Generator startup is in progress A failure in the electro-hydraulic control system (EHC) causes turbine speed to increase to 1872 rpm The EHC Auxiliary Governor actuates The IOPS (Independent Overspeed Protection System) actuates Which of the following lists the expected positions of the Turbine Stop and Governor valves while the EHC Auxiliary Governor and IOPS are actuated?

STOP Valves GOVERNOR Valves A. OPEN OPEN B. OPEN SHUT C. SHUT OPEN D. SHUT SHUT Proposed answer: D Tuesday, April 30, 2019 11:03:48 AM 202

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator must determine the mechanical overspeed limit has been reached as well as the auxiliary governor and independent overspeed protection system actuations. Given the RPM has reached the mechanical overspeed limit, that the stop and governor valves will go and stay closed.

A INCORRECT: Plausible if the operator has the misconception that the mechanical overspeed protection limit was not reached, and the other systems have actuated and cleared, allowing the stop and governor valves to reopen.

B INCORRECT: Plausible if the operator has the misconception that only the governor valves will be affected by the auxiliary governor system actuation, since it actuated first.

C INCORRECT: Plausible if the operator has a misconception that the auxiliary governor system will affect only the stop valves and not the governor valves.

D CORRECT: See above.

Learning Objective:

DESCRIBE the function and/or purpose, design basis, and operating characteristics of the Turbine Protection Trip system and major components.

(052.03.LP0021.001)

Tuesday, April 30, 2019 11:03:48 AM 203

QUESTIONS REPORT for 2019 NRC Exam Master

63. 2019 NRC 063/SYS/071A1.06/2.5/1-I/RO/BANK/RMSASRB CI RE-214/051.05.LP2711.004 Given the following:

Both units are at Rated Thermal Power PAB Ventilation is aligned as follows:

W-35, PAB Supply fan, is running W-30A, PAB Filter fan, is running, W-30B is secured W-21A, PAB Stack fan, is running, W-21B is secured PAB Ventilation filters are in the normal alignment Subsequently, a Waste Gas decay tank begins to leak RE-214, Aux Building Vent Exhaust monitor, goes into HIGH ALARM Which of the following correctly describes the FINAL PAB Ventilation system alignment assuming no operator actions?

W-21/W-30 PAB W-35, PAB F-23 PAB F-29 PAB Exhaust Fans Supply Fan Charcoal Filter HEPA Filter A. As-is Running Aligned Secured B. As-is Running Secured Aligned C. All Running Off Aligned Secured D. All Running Running Aligned Aligned Tuesday, April 30, 2019 11:03:48 AM 204

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 2 Group 2 Source: Bank Question History:

2007 PBNP Question 23 K/A:

071A1.06 Waste Gas Disposal System (WGDS)

Ability to predict and/or monitor changes in parameters(to prevent exceeding design limits) associated with Waste Gas Disposal System operating the controls including: Ventilation system.

(Imp 2.5/2.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to predict the changes in the ventilation systems based on an exhaust rad monitor going into high alarm.

Cognitive Level:

Knowledge 1-I: The operator recall the effect of the high alarm PAB ventilation system.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:48 AM 205

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

RMSARB CI RE-214, Auxiliary Building Vent Exhaust Gas Monitor Rev 9 M-215 Sh 3, Heating and Ventilation P&ID Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

Consider the following plant conditions:

- Both units are at 100% power.

- PAB Ventilation is aligned as follows:

- W-35, PAB Supply fan, is running.

- W-30A, PAB Filter fan, is running, W-30B is secured.

- W-21A, PAB Stack fan, is running, W-21B is secured.

- PAB Ventilation filters are in the normal alignment.

- Subsequently, a Waste Gas decay tank begins to leak.

- RE-214, Aux Building Vent Exhaust monitor, goes into HIGH ALARM.

Which of the following correctly describes the FINAL PAB Ventilation system alignment assuming no operator actions?

W-21/W-30 PAB W-35, PAB F-23 PAB F-29 PAB Exhaust Fans Supply Fan Charcoal Filter HEPA Filter As-is Running Aligned Secured As-is Running Secured Aligned All Running Off Aligned Secured All Running Running Aligned Aligned Proposed answer: A Tuesday, April 30, 2019 11:03:48 AM 206

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

On a RE-214 high alarm, the PAB charcoal filter is automatically aligned and the HEPA filter secured along with no change in the fan alignment.

A CORRECT: See above.

B INCORRECT: Plausible if the operator has the misconception that the HEPA filter will stay aligned and the charcoal filter will remain secured.

C INCORRECT: Plausible if the operator has the misconception that the supply fan will shut off, but the exhaust fan will remain running and charcoal filter will remain in service to keep a negative pressure on the PAB.

D INCORRECT: Plausible if the operator has a misconception that both filter will be in service and all supply and exhaust fans will operate to ensure maximum filtration of air from the PAB.

Learning Objective:

DESCRIBE the interlocks associated with the Auxiliary and South Service Building Ventilation Systems and its major components:

(051.05.LP2711.004)

Tuesday, April 30, 2019 11:03:48 AM 207

QUESTIONS REPORT for 2019 NRC Exam Master

64. 2019 NRC 064/SYS/072K5.01/2.7/1-F/RO/BANK/FSAR/053.05.LP0286.001 Which type of radiation is RE-101, Control Room Monitor designed to detect?

A. Alpha radiation B. Beta radiation C. Gamma radiation D. Neutron radiation RO Tier 2 Group 2 Source: Bank Question History:

2011 Millstone Unit 3 Question 62 K/A:

072K5.01 Area Radiation Monitoring (ARM) System Knowledge of the operational implications of the following concepts as they apply to the ARM system: Radiation theory, including sources, types, units, and effects.

(Imp 2.7/3.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall what type of radiation detector is utilized for the control room and what type of radiation it monitors for.

Cognitive Level:

Knowledge 1-F: The operator must recall the type of radiation being monitored by the control room radiation monitor.

10 CFR Part 55 Content:

55.41 5 55.43 Tuesday, April 30, 2019 11:03:48 AM 208

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

FSAR, Final Safety Analysis Report Rev 11/12/18, Section 11.5 page 11.5-5 of 27 Proposed reference to be provided to the applicants during examination:

None Original Question:

Which type of radiation is Millstone 3 Radiation Monitor 3RMS22-1, "Control Room" designed to detect?

A. Alpha radiation B. Beta radiation C. Gamma radiation D. Neutron radiation Proposed answer: C Tuesday, April 30, 2019 11:03:48 AM 209

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The Control Room Monitor is part of the area monitoring radiation system, and the purpose of the system is to monitor radiation levels in various areas which are subject to changing radiological conditions. The area radiation monitoring system utilizes low range fixed position gamma sensitive G-M tube detector assemblies.

A INCORRECT: Plausible because this is a type of ionizing radiation of concern at a nuclear plant or if the operator confuses this detector with the process radiation monitors in the control room.

B INCORRECT: Plausible because this is a type of ionizing radiation of concern at a nuclear plant or if the operator confuses this detector with the process radiation monitors in the control room.

C CORRECT: See above.

D INCORRECT: Plausible because this is a type of ionizing radiation of concern at a nuclear plant or if the operator confuses this detector with the process radiation monitors in the control room.

Learning Objective:

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Radiation Monitoring System. Description should include:

a. Major components and their physical location

b. Tasks and Control Functions performed by a DAM and a SPING

c. Operation of an AMAU and AMBU Type of detector utilized and its operation for each RMS channel (053.05.LP0286.001)

Tuesday, April 30, 2019 11:03:48 AM 210

QUESTIONS REPORT for 2019 NRC Exam Master

65. 2019 NRC 065/SYS/075K3.07/3.4*/1-F/RO/NEW/TRM 3.7.7/057.02.LP3410.003 Given the following:

Both Units are at Rated Thermal Power A large fish intrusion occurs Circ Water Pump Bay levels for BOTH units stabilize at -13 feet Which of the following components will cause TLCO 3.7.7, Service Water (SW)

System to no longer be met with Circ Water Pump Bay level at -13 Feet?

A. PAB Battery Room Vent Coolers B. 66' EL Containment Accident Fan Coolers C. Component Cooling Water Heat Exchangers D. G01 and G02, Emergency Diesel Generators RO Tier 2 Group 2 Source: New Question History:

None K/A:

075K3.07 Circulating Water System Knowledge of the effect that a loss or malfunctions of the circulating water system will have on the following: ESFAS.

(Imp 3.4*/3.5*)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall how a loss of Circ Water as indicated by a Circ Water Pump Bay level effect ESFAS by identifying which component is required to be declared inoperable.

Cognitive Level:

Knowledge 1-F: The operator must recall which component must be immediately declared inoperable based on Circ Water Pump Bay level.

10 CFR Part 55 Content:

55.41 7 55.43 Tuesday, April 30, 2019 11:03:48 AM 211

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TRM 3.7.7, Service Water (SW) System Rev 17 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

With Circ Water Pump Bay level being at -13 feet means that 66' EL containment accident fan cooler units need to be declared inoperable immediately (TRMAC 3.7.7.B), which is required to be done at a level of less than -11.5 feet.

A INCORRECT: Plausible, because PAB Battery Room Vent Cooler will be declared inoperable if the forebay temperature is greater than 85°F.

B CORRECT: See above.

C INCORRECT: Plausible, because Component Cooling Water Heat Exchanges are declared inoperable if the forebay temperature is greater than 85°F.

D INCORRECT: Plausible, because G01 and G02 will be declared inoperable if the Circ Water Pump Bay level is less -15 feet.

Learning Objective:

Given specific plant conditions, ASSESS and APPLY Technical Specification Technical Requirements Manual requirements as appropriate.

(057.02.LP3410.003)

Tuesday, April 30, 2019 11:03:48 AM 212

QUESTIONS REPORT for 2019 NRC Exam Master

66. 2019 NRC 066/SG/G2.1.18/3.6/1-P/RO/NEW/OM 3.31/SD86.2 2.01.18 Given the following:

Both Units are at Rated Thermal Power Due to multiple nuisance alarms for C01B 4-11, AUX BLDG STACK OR FILTER LOW FLOW the following actions have been taken:

C01B 4-11 has been placed in NOT AUTO CLEAR Action Request/Work Request (AR/WR) has been written Compensatory actions of verifying fan operation twice per shift are in place Per OM 3.31, Removal and Restoration of Alarm, is further tracking of the C01B 4-11 annunciator required and why?

A. No further tracking, due to the initiation of the AR/WR B. No further tracking due a backup Plant Process Computer System (PPCS) alarm being available C. Yes, PBF-2126, Single Shift Alarm OOS Log, requires tracking the accomplishment of the twice per shift checks D. Yes, PBF 2121, Alarm OOS Log, is completed and retained until normal function of the alarm is restored to track annunciator status Tuesday, April 30, 2019 11:03:48 AM 213

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 3 Source: New Question History:

None K/A:

G2.1.18 Conduct of Operations Ability to make accurate, clear, and concise logs, records, status boards, and reports.

(Imp 3.6/3.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall if a log entry is required, and what is required for that entry.

Cognitive Level:

Knowledge 1-P: The operator must recall the log requirements for removing an annunciator from service.

10 CFR Part 55 Content:

55.41 10 55.43

Reference:

OM 3.31, Removal and Restoration of Alarm, Rev 11 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:48 AM 214

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per Om 3.31, placing an annunciator in NOT AUTO CLEAR, a alarm log entry needs to be completed on PBF-2126 or 2121. The 2121 form would be filled out and maintained active until the restoration of the alarm A INCORRECT: Plausible, because if the alarm is taken out of service due per the fleet level procedure a WO or AR should be written to track the issue, but per the local or station procedure, additional actions must occur.

B INCORRECT: Plausible, because if the alarm is taken out of service by a procedure, or is caused by a procedural action, (e.g., operation of a system in this manner can cause multiple alarms) no further action is needed for logging, and this is one of the possible compensatory actions of alternate means for monitoring the affected parameter.

C INCORRECT: Plausible, because this form is one of the possible forms to use, but the action of recording the completion of the compensatory actions, is incorrect, as only what the compensatory action is needs to be recorded on this log.

D CORRECT: See above.

Learning Objective:

Ability to make accurate, clear, and concise logs, records, status boards, and reports.

(SD86.2 2.01.18)

Tuesday, April 30, 2019 11:03:49 AM 215

QUESTIONS REPORT for 2019 NRC Exam Master

67. 2019 NRC 067/GEN/G2.1.25/3.9/3-SPR/RO/MODIFIED/ROD 1.3/055.01.LP0244.004 Given the following:

Unit 2 has commenced a power ascension from 80% to 85% per OP 1C, Startup to Power Operation Unit 2 Burnup is 8500 MWD/MTU Using ROD 1.3, Reactivity Plan for Power Changes, CALCULATE the expected amount of dilution to offset the power change without the use of the control rods.

(The effect of Xenon is NOT to be considered)

A. 395 gal B. 495 gal C. 598 gal D. 701 gal RO Tier 3 Source: Modified Question History:

None K/A:

G2.1.25 Conduct of Operations Ability to interpret reference materials, such as graphs, curves, tables, etc.

(Imp 3.9/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to use both a table and graph to determine a dilution.

Cognitive Level:

Comprehensive 3-SPR: The operator must use provided references to determine the required dilution.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 216

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ROD 1.3, Reactivity Plant for Power Changes - U2C37, Rev 16 Proposed reference to be provided to the applicants during examination:

ROD 1.3, Reactivity Plant for Power Changes - U2C37 Burnup Range (8,000-9,000) MWD/MTU Original Question:

Given the following:

Unit 2 has commenced a power ascension from 80% to 85% per OP 1C, Startup to Power Operation Unit 2 Burnup is 3000 MWD/MTU Bank D rods are at 164 steps CALCULATE the expected amount of dilution in order to maintain TAVG without the use of the control rods. (the effect of Xenon is NOT to be considered)

(in Gallons of Water)

A. 240 B. 363 C. 426 D. 500 Proposed answer: B Tuesday, April 30, 2019 11:03:49 AM 217

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Answer using ROD 1.3 (Table 3)

Power Coefficient: -20.1 pcm/1% power = -20.1 x 5% = -100.5pcm HFP Differential Boron Worth: -6.4 pcm/1 ppm = -100.5pcm ÷ -6.4 = 15.7 ppm Dilution required for 1 ppm change: 38.1 gal/1 ppm = 38.1 x 15.7 = 598 gal A INCORRECT: Plausible, if the operator has a misconception in the calculation of a dilution, and a misapplication of Table 2 of ROD 1.3.

79 x 5 = 395 B INCORRECT: Plausible, if the operator misapplies the peak xenon worth

(-20.1 x 5) + (3.5 x 5) = 83 83 ÷ 6.4 = 13 38.1 x 13 = 495 C CORRECT: See above.

D INCORRECT: Plausible, if the operator applies the peak xenon worth Plausible, if the operator misapplies the peak xenon worth

(-20.1 x 5) - (3.5 x 5) = 118 118 ÷ 6.4 = 18.4 38.1 x 18.4 = 701 Learning Objective:

Given access to the site Specific Simulator, DEMONSTRATE the ability to maintain TAVG while operating the control rods in manual and automatic, and DISCUSS the required actions if TAVG cannot be maintained.

(055.01.LP0244.004)

Tuesday, April 30, 2019 11:03:49 AM 218

QUESTIONS REPORT for 2019 NRC Exam Master

68. 2019 NRC 068/GEN/G2.1.43/4.1/2-DR/RO/BANK/THERM 007/PBN LOI GFE THERM007 Given the following:

The unit is operating at Rated Thermal Power The feedwater temperature input to the LEFM thermal power calculation was incorrectly calibrated to 7°F higher than actual feedwater temperature Calibration of the Power Range NIs is being performed How will LEFM power compare to actual thermal power and how will adjustment of the NIs be affected using the calculated value of LEFM?

Calculated thermal power per TS-RE-001, Power Level Determination, will be

___(1)___ than actual power.

NI adjustment per TS-RE-002, Power Range Detector Power Level Adjustment, will be ___(2)___ conservative.

A. (1) lower (2) less B. (1) higher (2) less C. (1) lower (2) more D. (1) higher (2) more Tuesday, April 30, 2019 11:03:49 AM 219

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 3 Source: Bank Question History:

2008 DC Cook Question 67 K/A:

G2.1.43 Conduct of Operations Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

(Imp 4.1/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine the effects on reactivity of differences in secondary plant temperatures based on knowledge of the calorimetric procedure.

Cognitive Level:

Comprehensive 2-DR: The operator determine the effect of the calibration error on the calorimetric and how that will affect the adjustment of the Power Range NIs.

10 CFR Part 55 Content:

55.41 10 55.43 6 Tuesday, April 30, 2019 11:03:49 AM 220

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

193007, Heat Transfer Instructor Guide Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following plant conditions on Unit 1:

The unit is operating at 100%.

The feedwater temperature input to the LEFM thermal power calculation was incorrectly calibrated to 7°F higher than actual feedwater temperature.

Calibration of the power range nuclear instruments (NIs) is being performed.

How will LEFM power compare to actual thermal power and how will adjustment of the NIs be affected using the calculated value of LEFM?

A. Calculated thermal power is lower than actual power.

NI adjustment will be less conservative.

B. Calculated thermal power is higher than actual power.

NI adjustment will be less conservative.

C. Calculated thermal power is lower than actual power.

NI adjustment will be more conservative.

D. Calculated thermal power is higher than actual power.

NI adjustment will be more conservative.

Proposed answer: A Tuesday, April 30, 2019 11:03:49 AM 221

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Due to actual Feedwater temperature being 7°F lower than indicated/calibrated feedwater temperature, calculated thermal power will indicate lower than actual thermal power. This will cause NI calibration to result in setpoints that are less conservative (i.e. farther from the trip setpoint).

A CORRECT: See above.

B INCORRECT: The first part is incorrect, plausible if the operator has a misconception or reverses the effect of the Feedwater temperature calibration error will have on the calorimetric. The second part is correct.

C INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has a misconception or reverses the effect of the PR NI calibration.

D INCORRECT: The first part is incorrect, plausible if the operator has a misconception or reverses the effect of the Feedwater temperature calibration error will have on the calorimetric. The second part is incorrect, plausible if the operator has a misconception or reverses the effect of the PR NI calibration.

Learning Objective:

Explain methods of calculating core thermal power (PBN LOI GFE THERM 007 K1.06)

Tuesday, April 30, 2019 11:03:49 AM 222

QUESTIONS REPORT for 2019 NRC Exam Master

69. 2019 NRC 069/GEN/2.2.14/3.9/1-P/RO/BANK/OM 3.41/SD86.2 2.02.14 Given the following:

A piece of equipment is placed in an OFF-NORMAL position and is NOT procedurally controlled This equipment does not affect any LCOs or TLCOs An action request(AR)/work request(WR) to document the condition has been written In accordance with OM 3.41, System Status Control, what additional actions are required to track the position of the shut gauge isolation?

A. Enter the component on the Abnormal Alignment List AND attach Information Tag if directed by Shift Management.

B. Enter the component information in the Station Log AND ensure an item is included on the appropriate Turnover Notes.

C. No additional actions are required; the AR/WR is sufficient to ensure the component is tracked via the work control process.

D. Annotate the system checklist for the component out of the normal position AND maintain that checklist in the Procedure-in-Progress file.

Tuesday, April 30, 2019 11:03:49 AM 223

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 3 Source: Bank Question History:

2017 PBNP Question 69 Previous 2 NRC Exams K/A:

G2.2.14 Equipment Control Knowledge of the process for controlling equipment configuration or status.

(Imp 3.9/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the process for maintaining the status of a component which is put in an off normal position/condition.

Cognitive Level:

Knowledge 1-P: The operator must recall the requirements for maintaining the status of components placed in an off normal position/condition.

10 CFR Part 55 Content:

55.41 10 55.43 3 Tuesday, April 30, 2019 11:03:49 AM 224

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

OM 3.41, System Status Control Rev 11, Attachment B Section 4.2.3.

Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A piece of equipment is placed in an OFF-NORMAL position and is NOT procedurally controlled This equipment does not affect any LCOs or TLCOs An action request(AR)/work request(WR) to document the condition has been written In accordance with OM 3.41, System Status Control, what additional actions are required to track the position of the shut gauge isolation?

A. Enter the component on the Abnormal Alignment List AND attach an Information Tag as soon as practicable.

B. Enter the component information in the Station Log AND ensure an item is included on the appropriate Turnover Notes.

C. No additional actions are required; the AR/WR is sufficient to ensure the component is tracked via the work control process.

D. Annotate the system checklist for the component out of the normal position AND maintain that checklist in the Procedure-in-Progress file.

Proposed answer: A Tuesday, April 30, 2019 11:03:49 AM 225

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

From OM-3.41, Attachment B, 4.2.3, When equipment is placed in an Off-normal positon and is NOT procedurally controlled, ENSURE the following for any equipment and any affected instrumentation, switches, etc.:

An action request (AR) with a 90 day due date, and a work request (WR) are initiated to track and correct the condition as needed.

The following entered into the Abnormal Alignment List:

Reason for alignment AR number Original abnormal alignment date EC number (if applicable) 10CFR50.59 tracking number (AR number) is placed in the 50.59 tracking AR column of the Abnormal Alignment List.

All applicable information fields are filled in on the Information Tag Information Tag is affixed to the identified equipment as soon as practicable.

Purple Abnormal Alignment magnetic arrows should be used on the Main Control Boards in lieu of information tags.

A CORRECT: See above.

B INCORRECT: Plausible because these are reasonable actions to take for this situation. Station log entries are a commonly required action.

C INCORRECT: Plausible if the examinee believes that the work control process is appropriate to maintain component status.

D INCORRECT: Plausible because maintaining procedures in the procedure-in

-progress file is an appropriate status control action for when a procedure is suspended.

Learning Objective:

Knowledge of the process for controlling equipment configuration or status.

(SD 86.2 2.2.14)

Tuesday, April 30, 2019 11:03:49 AM 226

QUESTIONS REPORT for 2019 NRC Exam Master

70. 2019 NRC 070/GEN/G2.2.42/3.9/1-F/RO/BANK/TS 3.4.13/057.02.LP3339.002 Given the following:

Both units are at Rated Thermal Power Which of the following exceeds the limits of Tech Spec 3.4.13, Reactor Coolant System, RCS Operation LEAKAGE?

A. 6 gpm identified leakage B. 0.5 gpm unidentified leakage C. 90 GPD tube leakage 1HX-1B, Steam Generator D. 120 GPD tube leakage 2HX-1A, Steam Generator RO Tier 3 Source: Bank Question History:

2010 McGuire Question 69 K/A:

G2.2.42 Equipment Control Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

(Imp 3.9/4.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the limits for RCS Operational Leakage which constitutes entry condition for Tech Specs Cognitive Level:

Knowledge 1-F: The operator must recall the limits for RCS Operational Leakage.

10 CFR Part 55 Content:

55.41 7, 10 55.43 2, 3 Tuesday, April 30, 2019 11:03:49 AM 227

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

Tech Spec 3.4.13, Reactor Coolant System, RCS Operation LEAKAGE Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

With Unit I operating at 100% RTP, which ONE (1) of the following exceeds the limits of Tech Spec 3.4.13 (RCS Operational Leakage)?

A. 6 GPM identified leakage B. 0.5 GPM unidentified leakage C. 140 GPD tube leakage in 1C SC D. 356 GPD total primary-to-secondary leakage through all SGs Proposed answer: C Justification:

Operational leakage limits are:

NO pressure boundary leakage 1 gpm unidentified leakage 10 gpm identified leakage 72 gpd (Unit 1) and 150 gpd (unit 2) primary to secondary leakage through any one steam generator A value of 90 gpd tube leakage for unit 1 will exceeds the limit of 72 gpd.

A INCORRECT: Plausible if the operator confuses identified leakage limits with either unidentified or pressure boundary leakage because this exceeds the allowable limit for pressure boundary leakage and unidentified leakage.

B INCORRECT: Plausible if the operator confuses unidentified leakage limits with pressure boundary leakage because this exceeds the allowable limit for pressure boundary leakage.

C CORRECT: See above D INCORRECT: Plausible if the operator confuses the limits for the units because this exceeds the limit for unit 1 tube leakage.

Learning Objective:

Given specific plant conditions, ASSESS and APPLY Technical Specification requirements as appropriate.

(057.02.LP3339.002)

Tuesday, April 30, 2019 11:03:49 AM 228

QUESTIONS REPORT for 2019 NRC Exam Master

71. 2019 NRC 071/GEN/G2.3.4/3.3/1-P/RO/MODIFIED/NP 4.2.14/SD86.2 2.03.04 Given the following:

A 20 year old PBNP employee has 500 mrem of TEDE exposure for 2019 Which of the following describes:

(1) The MAXIMUM amount of additional TEDE exposure that may be received without additional authorization (i.e., Site VP, Plant Manager, Radiation Protection Manager, department manager approval) at PBNP?

AND (2) What is the MAXIMUM amount of additional TEDE exposure that may be received prior to exceeding 10CFR20 (NRC) exposure limits?

(1) (2)

A. 500 mrem 4500 mrem B. 1500 mrem 4500 mrem C. 500 mrem 5000 mrem D. 1500 mrem 5000 mrem RO Tier 3 Source: Modified Question History:

2013 Comanche Peak Question 70 K/A:

G2.3.4 Radiation Control Knowledge of radiation exposure limits under normal or emergency conditions.

(Imp 3.2/3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the exposure limits under normal and emergency conditions.

Cognitive Level:

Knowledge 1-P: The operator must recall the exposure limits under normal and emergency conditions.

10 CFR Part 55 Content:

55.41 12 55.43 4 Tuesday, April 30, 2019 11:03:49 AM 229

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

NP 4.2.14, Administrative Dose Levels/Dose Level Extension Procedure Rev 11, Table 1 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following condition:

A 20 year old CPNPP employee has 1000 mrem of TEDE exposure for 2013.

Which of the following describes the MAXIMUM amount of additional TEDE exposure that may be received without additional CPNPP authorization (i.e.,

Plant Manager approval, Radiation and Industrial Safety Manager, employee supervisor, etc.), and what is the MAXIMUM amount of additional TEDE exposure that may be received prior to exceeding 10CFR20 (NRC) exposure limits?

A. 1000 mrem; 4000 mrem B. 3000 mrem; 4000 mrem C. 1000 mrem; 5000 mrem D. 3000 mrem; 5000 mrem Proposed Answer: A Tuesday, April 30, 2019 11:03:49 AM 230

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per NP 4.2.14, you are allowed 2000 mrem TEDE administratively a year. The question asks about the PBNP extensions, so 1500 mrem TEDE is all that is allowed without extension for a total of 2000 mrem TEDE. The 10CFR20 limit is 5000 mrem TEDE, therefore only 4500 mrem TEDE can be added to 500 mrem TEDE to be below that value.

A INCORRECT: The first part is incorrect, plausible because the limit was changed from 1000 mrem to the current limit. The second part is correct.

B CORRECT: See above C INCORRECT: The first part is incorrect, plausible because the limit was changed from 1000 mrem to the current limit. The second part is incorrect, plausible if the operator has a misconception of the application of the legal limit, because the legal limit is 5000 mrem.

D INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has a misconception of the application of the legal limit, because the legal limit is 5000 mrem.

Learning Objective:

Knowledge of radiation exposure limits under normal or emergency conditions.

(SD86.3 2.03.04)

Tuesday, April 30, 2019 11:03:49 AM 231

QUESTIONS REPORT for 2019 NRC Exam Master

72. 2019 NRC 072/GEN/G2.3.11/3.8/1-P/RO/BANK/EOP-3/031.02.LP0441.002 Given the following:

The B SG is ruptured The crew is performing the actions of EOP-3, Steam Generator Tube Rupture Which of the following describes how 'B' Atmospheric Dump Valve (ADV) controller will be aligned to minimize steam dumped from the 'B' SG?

Controller placed in . . .

A. AUTO, set at 1005 psig.

B. AUTO, set at 1050 psig.

C. MANUAL, maintaining 'B' SG pressure at 1005 psig.

D. MANUAL, maintaining 'B' SG pressure at 1050 psig.

RO Tier 3 Source: Bank Question History:

2011 Ginna Retake Question 71 K/A:

G2.3.11 Radiation Control Ability to control radiation releases.

(Imp 3.8/4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the atmospheric dump valve control setting for a steam generator with a tube rupture.

Cognitive Level:

Knowledge 1-P: The operator must recall the controller setpoint for ruptured SG ADV.

10 CFR Part 55 Content:

55.41 11 55.43 4 Tuesday, April 30, 2019 11:03:49 AM 232

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-3 U1, Steam Generator Tube Rupture Unit 1, Rev 55, Step 3 a or b Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

'B' SG is ruptured.

The crew is performing the actions of E-3, Steam Generator Tube Rupture.

Which ONE of the following describes how 'B' ARV Controller will be aligned to minimize steam dumped from the 'B' Steam Generator?

Controller placed in...

A. AUTO, set at 1005 PSIG.

B. AUTO, set at 1050 PSIG.

C. MANUAL, maintaining 'B' SG pressure at 1005 psig.

D. MANUAL, maintaining 'B' SG pressure at 1050 PSIG.

Proposed Answer: B Tuesday, April 30, 2019 11:03:49 AM 233

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Step 3.b of EOP-3 will Ensure 1HC-478, B ADV Controller - Set to 1050 psig.

EOP Step 3 of the background states the purpose is to isolate flow from the ruptured steam generator to minimize radiological releases. And The atmospheric dump valve on the ruptured steam generator should remain available to limit steam generator pressure unless it fails open. This will minimize any challenges to the code safety valve.

A INCORRECT: Plausible because this is the normal settings for the controller.

B CORRECT: See above C INCORRECT: Plausible because the action of taking the control to manual would ensure it would not open in the event of a controller failure which would minimize the release potential, but the procedure does not direct this, and 1005 is the normal setting for the ADV.

D INCORRECT: Plausible because the action of taking the control to manual would ensure it would not open in the event of a controller failure which would minimize the release potential, but the procedure does not direct this, and 1050 is the correct setting for the ADV with a tube rupture.

Learning Objective:

LIST the major actions accomplished by the following procedures:

a. EOP-2

b. EOP-3

c. EOP-3.1

d. EOP-3.2

e. EOP-3.3 (031.02.LP0441.002)

Tuesday, April 30, 2019 11:03:49 AM 234

QUESTIONS REPORT for 2019 NRC Exam Master

73. 2019 NRC 073/GEN/G2.3.14/3.4/1-F/RO/BANK/FSAR/SD86.3 2.3.14 Given the following:

Refueling is in progress on Unit 1 A spent fuel assembly is being moved from the Upender to the Spent Fuel Pool The spent fuel assembly contacts the divider wall Gas bubbles are observed rising from the damaged spent fuel assembly to the surface of the Spent Fuel Pool Which of the following correctly describes the primary radiation hazard to personnel in the immediate vicinity AND the related source products released from the damaged spent fuel assembly?

A. Alpha radiation from activated hydrogen gas.

B. Gamma radiation from fission product gases.

C. Alpha radiation from Reactor Coolant particulate fission products.

D. Gamma radiation from particulate fission products and corrosion products.

RO Tier 3 Source: Bank Question History:

2011 South Texas Projects Question 45 K/A:

G2.3.14 Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(Imp 3.4/3.8)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall potential hazards associated with spent fuel assemblies.

Cognitive Level:

Knowledge 1-F: The operator must recall the potential hazards associated with spent fuel assemblies.

10 CFR Part 55 Content:

55.41 12 55.43 4 Tuesday, April 30, 2019 11:03:49 AM 235

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

FSAR, Section 14.2.1 and Table 14.2.-2 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Refueling is in progress on Unit 1.

A spent fuel assembly is being moved from the reactor to the Upender.

The spent fuel assembly is dropped to the bottom of the Transfer Canal.

Gas bubbles are observed rising from the damaged spent fuel assembly to the surface of the Transfer Canal.

Which one of the following correctly describes the primary radiation hazard to personnel in the immediate vicinity AND the related source products released from the damaged spent fuel assembly?

A. Alpha radiation from activated hydrogen gas.

B. Gamma radiation from fission product gases Xenon and Krypton.

C. Alpha radiation from Reactor Coolant particulate fission products.

D. Gamma radiation from particulate fission products and corrosion products.

Proposed answer: B Tuesday, April 30, 2019 11:03:49 AM 236

QUESTIONS REPORT for 2019 NRC Exam Master Spent fuel assemblies have some amount of fission gases within the fuel rods depending on the degree of use of the fuel assembly. Fission yields for Xenon and Krypton are higher than most other elements therefore these gases will be present in larger amounts than other gases. Both of these gases decay by gamma emission thereby creating the radiation hazard if released from the clad.

Because they are gases, they will rise to the surface of the Spent Fuel Pool where they will enter the surrounding atmosphere and create a radiation hazard due to their decay.

A INCORRECT: Plausible if the operator has a misconception of the hazards associated with spent fuel. Hydrogen gas within a fuel rod acts to counter balance external pressure on the rod when the fuel is in the reactor at operating pressure. 'Activated Hydrogen' is Tritium which decays by beta emission. It can be easily shielded and so will not present a significant radiation hazard unless it is ingested B CORRECT: See above C INCORRECT: Plausible if the operator has a misconception of the hazards associated with spent fuel. Fission products predominantly undergo beta, gamma decay. Only the heavier isotopes in the fuel undergo Alpha decay and these are typically not fission products but fuel isotopes or other heavy isotopes created as a result of neutron activation D INCORRECT: Plausible if the operator has a misconception of the hazards associated with spent fuel. Gamma radiation from particulate fission products and corrosion products can be a significant radiation source, however since the damage to the spent fuel assembly occurred at the bottom of the Transfer Canal, these products will remain in the water and not be released to the surrounding environment as gases will be.

Learning Objective:

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(SD86.3 2.3.14)

Tuesday, April 30, 2019 11:03:49 AM 237

QUESTIONS REPORT for 2019 NRC Exam Master

74. 2019 NRC 074/GEN/2.4.3/3.7/1-P/RO/BANK/TS 3.3.3/SD86.2 2.04.03 Which of the following is a Post Accident Monitoring Instrument, AND controlled by Tech Spec 3.3.3, Instrumentation - Post Accident Monitoring (PAM)

Instrumentation?

A. LT-460, SG A Wide Range level B. FI-128, Charging Line Flow C. RE-109, Failed Fuel Monitor D. RE-211, Containment Air Particulate Monitor RO Tier 3 Source: Bank Question History:

2013 Catawba Question 75 K/A:

G2.4.3 emergency Procedures / Plan Ability to identify post-accident instrumentation.

(Imp 3.7/3.9)

Justification for K/A Match:

Matches the K/A by requiring the operator to identify post-accident instrumentation from a list of instruments controlled by Tech Specs.

Cognitive Level:

Knowledge 1-P: The operator must identify post-accident instrumentation from a list of instruments controlled by Tech Specs.

10 CFR Part 55 Content:

55.41 6 55.43 Tuesday, April 30, 2019 11:03:49 AM 238

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 3.3.3, Instrumentation - Post Accident Monitoring (PAM) Instrumentation Rev 3, Table 3.3.3-1 Proposed reference to be provided to the applicants during examination:

None Original Question:

Which ONE of the following is a Post Accident Monitoring Instrument, AND is controlled by Tech. Spec. 3.3.3, [Post-Accident Monitoring (PAM)

Instrumentation]?

A. Containment Humidity B. SG Water Level Wide Range C. I EMF-36L, Unit Vent Gas Monitor D. I EMF-38L, Containment Monitor - Particulate Proposed answer: B Justification:

This instrument is identified as a Post-Accident Monitoring Instrument per LCO 3.3.3 item 15, Steam Generator Water Level (Wide Range).

A CORRECT: See above.

B INCORRECT: Plausible because this instrument is used in several places in both the AOP and EOP network.

C INCORRECT: Plausible because this instrument is required by the Emergency Plan Implementing procedures and used post-accident to diagnose failed fuel/cladding damage.

D INCORRECT: Plausible because this radiation monitor is controlled by Tech Specs, under the LCO 3.4.15 RCS Leak Detection Instrument.

Learning Objective:

Ability to identify post-accident instrumentation.

(SD86.4 2.04.03)

Tuesday, April 30, 2019 11:03:49 AM 239

QUESTIONS REPORT for 2019 NRC Exam Master

75. 2019 NRC 075/GEN/G2.4.46/4.2/1-P/RO/BANK/AOP-6A/053.01.LP1547.005 Given the following:

Unit 1 is at 68% Reactor Power One Shutdown Bank A rod has dropped into the core The crew is recovering the dropped rod Annunciator 1C04 1A 1-6, ROD CONTROL SYSTEM URGENT FAILURE is received when rod withdrawal begins.

Which of the following describes the Rod Control System Urgent Failure alarm and the plant response?

The alarm is . . .

A. unexpected. Rod withdrawal will not occur until the alarm is reset at the Logic Cabinet.

B. expected. The alarm will have to be reset to allow rod recovery to continue.

C. unexpected. Rod withdrawal will not occur until the alarm is reset at the Power Cabinet.

D. expected. Rod withdrawal is unaffected and recovery may continue.

Tuesday, April 30, 2019 11:03:49 AM 240

QUESTIONS REPORT for 2019 NRC Exam Master RO Tier 3 Source: Bank Question History:

2012 Diablo Canyon Question 75 K/A:

G2.4.46 emergency Procedures / Plan Ability to verify that the alarms are consistent with the plant conditions.

(Imp 4.2/4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to identify if an alarm is consistent with plant conditions.

Cognitive Level:

Knowledge 1-P: The operator must recall a procedure note about the urgent failure alarm and the effect it will have on the plant.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 241

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-6A Unit 1, Dropped Rod Rev 20, Step 26 BG AOP-6A, Background Dropped Rod Rev 17, Step 26 ARP 1C04 1A 1-6, Rod Control System Urgent Failure Rev 0 Proposed reference to be provided to the applicants during examination:

None Original Question:

GIVEN:

68% reactor power.

One Shutdown Bank A Rod has dropped into the core.

The crew is recovering the dropped rod.

PK03-17, ROD CONT URGENT FAILURE is received when rod withdrawal begins Which of the following describes the Rod Control System Urgent Failure alarm and the plant response?

The alarm is:

A. unexpected. Rod withdrawal will not occur until the alarm is reset at the Logic Cabinet.

B. expected. The alarm will have to be reset to allow rod recovery to continue.

C. unexpected. Rod withdrawal will not occur until the alarm is reset at the Power Cabinet.

D. expected. Rod withdrawal is unaffected and recovery may continue.

Proposed Answer: D.

Tuesday, April 30, 2019 11:03:49 AM 242

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per AOP-6A note prior to step 26, Retrieving any RCCA in shutdown Bank A OR control Banks A or C will cause an urgent failure alarm. This alarm should not stop rod motion. A logic cabinet urgent failure occurs when a pulse is sent but is not received by any rod in an individual group. This occurs when all the lift coil disconnects are open on all the rods of a group. This will not prevent rod motion in the with the Control Rod Bank Selector switch in the bank position, but will prevent rod motion in the manual or automatic position.

A INCORRECT: Plausible because this would be correct if the rod was on a bank different than shutdown Bank A OR control Banks A or C.

B INCORRECT: Plausible because the alarm is expected, and this reset would be necessary is the rod was on a bank shutdown different than Bank A OR control Banks A or C C INCORRECT: Plausible because this alarm is on the power cabinet and would occur if the rod was on a bank different than shutdown Bank A or control Banks A or C.

D CORRECT: See above.

Learning Objective:

DESCRIBE the procedures which govern operation of the Rod Control System.

Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed or Auxiliary Operators.

(053.01.LP1547.005)

Tuesday, April 30, 2019 11:03:49 AM 243

QUESTIONS REPORT for 2019 NRC Exam Master

76. 2019 NRC 076/EPE/007EG2.1.30/4.0/3-SPK/SRO/NEW/BG-EOP-0.1/031.02.LP0405.015 Given the following:

Unit 1 experienced a complete loss of Charging flow with rising Seal Outlet temperatures The crew tripped the reactor and entered EOP-0, Reactor Trip or Safety Injection Concurrent with the reactor trip, a loss of Component Cooling Water occurred During the restoration of Charging flow, the OS1 directed the local closure of the Seal Water Injection Throttle valves The crew transitioned to EOP-0.1, Reactor Trip Response Component Cooling Water flow AND Charging flow were restored The crew has reached step 10, Check B RCP - Running Concerning the RCPs, answer the following:

(1) Where did the auxiliary operator go to shut the Seal Water Injection Throttle valves AND (2) Should the RCPs be started and why?

A. (1) 8/PAB/Pipeway #1 (2) Yes, once the conditions for starting an RCP have been established in Attachment C, Starting an RCP, starting the RCP will no longer result in degradation to the RCP bearings or seals.

B. (1) 8/PAB/Pipeway #1 (2) No, the RCP should only be started after a status evaluation or when an RCP start is required to address a CSF challenge.

C. (1) 8/PAB/Pipeway #2 (2) Yes, once the conditions for starting an RCP have been established in Attachment C, Starting an RCP, starting the RCP will no longer result in degradation to the RCP bearings or seals.

D. (1) 8/PAB/Pipeway #2 (2) No, the RCP should only be started after a status evaluation or when an RCP start is required to address a CSF challenge.

Tuesday, April 30, 2019 11:03:49 AM 244

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: New Question History:

None K/A:

007EG2.1.30 Reactor Trip Ability to locate and operate components, including local controls.

(Imp 4.0)

Justification for K/A Match:

Matches the K/A by requiring identification of the location for local operation of the valves necessary to stop seal water injection, during the stabilization in EOP-0.1, Reactor Trip Response.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to use information contained in notes and background documents to base the decision.

Cognitive Level:

Comprehension 3-SPK: The operator must understand and apply information contained in procedure notes and background documentation.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

EOP-0.1, Reactor Trip Response Unit 1, Rev 46 BG-EOP-0.1, Background Reactor Trip Response, Rev 32 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:49 AM 245

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The location for valve operation is 8/PAB/Pipeway #1 EOP-0.1 step 10, has a caution which states, If RCP seal cooling had previously been lost, affected RCPs should not be started prior to a status evaluation The background document further states that, if excessive temperatures develop, the affected RCP should not be restarted prior to a complete RCP evaluation. There is not delineation of time or temperature in the question stem, and seal water injection valves were closed, so the pump should not be started unless needed during a CSP Under a CSF challenge, potential RCP damage is an acceptable consequence if RCP start is required to address a CSF challenge.

A INCORRECT: The first part is correct. The second part is incorrect, but plausible, if the operator forgets the note and/or background information. If Attachment C conditions can be met, then the pump will be started.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, plausible because this is the location of the Component Cooling Water Throttle valves . The second part is incorrect, but plausible, if the operator forgets the note and/or background information. If Attachment C conditions can be met, then the pump will be started.

D INCORRECT: The first part is incorrect, plausible because this is the location of the Component Cooling Water Throttle valves. The second part is correct Learning Objective:

Given access to the Site Specific Simulator, respond to accident conditions in accordance with the plants Emergency Procedures.

(031.02.LP0405.015)

Tuesday, April 30, 2019 11:03:49 AM 246

QUESTIONS REPORT for 2019 NRC Exam Master

77. 2019 NRC 077/APE/022AA2.01/3.8/2-DR/SRO/MODIFIED/TRM 3.5.1/051.02.LP3410.003 Given the following:

Unit 1 is at Rated Thermal Power Letdown flow is 80 gpm on FI-134 LP Letdown Line Flow The following indications are noted:

1FR-177, RCP A&B Seal Leakage High Range Flow is 1.8 gpm and STABLE 1FI-128, Charging Line Flow 65 gpm and LOWERING 1TI-127, 1HX-2 Regen HX Letdown Outlet Temperature is 320°F and RISING VCT Level is 55% and LOWERING Pressurizer level is 45% and LOWERING SLOWLY TAVG is 575.9°F and STABLE 1C20 A 3-4, AUX BLDG SOUTH SUMP LEVEL HIGH is LIT Which of the following:

(1) Describes a correct leak location AND (2) The implication on Technical Specifications based on isolation of this leakage location.

(See following page for leak location)

REFERENCE PROVIDED TRM 3.5.1 (2 pages)

TS 3.6.3 (4 pages)

A. (1) letdown line between 1CV-371A and 1CV-371, Letdown Line Containment Isolation valves (2) TS 3.6.3 is met B. (1) letdown line between 1CV-371A and 1CV-371, Letdown Line Containment Isolation valves (2) declare LCO is NOT met and enter TSAC 3.6.3.A C. (1) piping upstream of 1CV-323A, RC Loop B Cold Leg Auxiliary Charging Line Isolation valve (2) declare TLCO 3.5.1 NOT met and enter TRMAC 3.5.1.C only D. (1) piping upstream of 1CV-323A, RC Loop B Cold Leg Auxiliary Charging Line Isolation valve (2) declare TLCO 3.5.1 NOT met and enter TRMAC 3.5.1.C AND TRMAC 3.5.1.E Tuesday, April 30, 2019 11:03:49 AM 247

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: Modified Question History: 2016 Callaway Question 76 SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to assess plant conditions utilize interpret the leakage location and determine the correct TSAC to en ter.

K/A:

022AA2.01 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: Whether charging line leak exists (Imp 3.8)

Justification for K/A Match:

Matches the K/A by requiring identification of a charging leak, then interpret the impact on tech specs.

Cognitive Level:

Comprehension 2-DR: The operator must understand initial conditions, and apply the relationship between charging, letdown to determine the location of the leak, and then determine the required action conditions per Tech Specs.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 248

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TRM 3.5.1, Chemical and Volume Control System, Rev 11 684J741 Sh 2, Chemical and Volume Control P&ID, Rev 82 684J741 Sh 3, Chemical and Volume Control P&ID, Rev 17 Proposed reference to be provided to the applicants during examination:

TRM 3.5.1, Chemical and Volume Control System pages 1 and 2 TS 3.6.3, Containment Isolation Valves, pages 1-4 Original Question:

Reactor Power is 100%.

All control systems are in normal alignment.

Letdown flow is 75 gpm on BG FI-132, LTDN HX OUTLET FLOW.

The following parameters are now noted on the CVCS system:

Seal Return Flows are 3 gpm per Reactor Coolant Pump.

Charging flow is 94 gpm and rising.

BG TI-130, LTDN HX OUTLET TEMP, has risen 25°F from its steady state value.

Volume Control Tank level is 55% and lowering.

Pressurizer level is 53% and lowering slowly.

Reactor Coolant System (RCS) average temperature is 585°F and stable.

Which of the following describes the correct leak location AND associated leakage monitoring requirements?

The Leakage is from the ........

A. letdown line between the letdown isolation valves and the orifices valves. This leakage must be monitored as post accident recirculation flowpath leakage, as required by TS 5.5.2, Primary Coolant Sources Outside Containment.

B. charging line between the flow indicator and Containment. This leakage is required to be monitored as a radiological effluent, as required by TS 5.5.4, Radioactive Effluent Controls Program.

C. letdown line between the letdown isolation valves and the orifices valves. This leakage is required to be monitored as a radiological effluent, as required by TS 5.5.4, Radioactive Effluent Controls Program.

D. charging line between the flow indicator and Containment. This leakage must be monitored as post accident recirculation flowpath leakage, as required by TS 5.5.2, Primary Coolant Sources Outside Containment.

Proposed answer: D Tuesday, April 30, 2019 11:03:49 AM 249

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the indications of the leak, a leak in the charging line will cause the VCT level to lower, PZR level to lower. Charging flow will rise based on the lowering of PZR level. And the NRHX temperature will rise. Given there is leakage in the charging line, the impact on TS will be that TRM 3.5.1 will not be met, and TRMAC 3.5.1.C One required boron injection flow path non-functional in MODE 1, 2, 3, or 4. This is based only of the three main paths from charging to the primary, Charging via RCS Loop A cold leg is now isolated due to the leak. The other current path that still exists is the seal injection flow path from the charging system.

A INCORRECT: The first part is incorrect, plausible if the operator has a misconception of the system flow and layout. A leak between the letdown isolation valves and the orifice valves would not cause a rise in the NRHX temperature. The second part is incorrect, plausible because it would be a correct condition for TS 3.6., given this leak condition.

B INCORRECT: The first part is incorrect, plausible if the operator has a misconception of the system flow and layout. A leak between the letdown isolation valves and the orifice valves would not cause a rise in the NRHX temperature. The second part is incorrect, plausible if the operator has the misconception that a leak in the penetration flow path with two containment isolation valves would be inoperable due to the leak location.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible is misconception that with the loss of charging all required charging paths are lost given no path from the charging pumps to the primary.

Learning Objective:

Given specific plant conditions, ASSESS and APPLY Technical Specification Technical Requirements Manual requirements as appropriate.

(057.02.LP3410.003)

Tuesday, April 30, 2019 11:03:49 AM 250

QUESTIONS REPORT for 2019 NRC Exam Master

78. 2019 NRC 078/EPE/038EG2.4.4/4.7/2-DR/SRO/MODIFIED/AOP-3A/055.03.LP2438.004 Given the following:

Unit 1 was operating at Rated Thermal Power A 50 gpd Tube Leak developed on the 'B' SG with a rate of change LESS THAN 30 gpd/hr Unit 1 was shutdown in response to this condition Plant cooldown is in progress with RCS temperature at 495°F

'B' SG tube leakage has just risen to 275 gpm Which of the following describes:

(1) the procedure in effect for cooldown PRIOR to the elevated tube leakage?

AND (2) the appropriate actions to implement due to the elevated leakage?

A. (1) Cooldown is being performed per AOP-3, Steam Generator Tube Leak.

(2) Start Safety Injection pumps and go to EOP-3, Steam Generator Tube Rupture.

B. (1) Cooldown is being performed per AOP-3, Steam Generator Tube Leak.

(2) Initiate Safety Injection and Containment Isolation and go to EOP-0, Reactor Trip or Safety Injection.

C. (1) Cooldown is being performed per OP 3A, Power Operations to Hot Standby.

(2) Start Safety Injection pumps and go to EOP-3, Steam Generator Tube Rupture.

D. (1) Cooldown is being performed per OP 3A, Power Operations to Hot Standby.

(2) Initiate Safety Injection and Containment Isolation and go to EOP-0, Reactor Trip or Safety Injection.

Tuesday, April 30, 2019 11:03:49 AM 251

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: Modified Question History:

2013 Indian Point Unit 3 Question 79 K/A:

038EG2.4.4 Steam Generator Tube Rupture (SGTR)

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

(Imp 4.7)

Justification for K/A Match:

Matches the K/A by requiring identification of the entry procedure required for mitigation of a SGTL, as this would not require an AOP entry on unit 2.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure will be entered initially, procedure flow path required for plant cooldown, and the course of mitigation for a SGTR, based on the procedure path required.

Cognitive Level:

Comprehension 2-DR: The operator must understand initial conditions, which procedure is entered, what procedure controls cooldown, and what the procedural requirements are for mitigation of the SGTR.

10 CFR Part 55 Content:

55.41 10 55.43 2 Tuesday, April 30, 2019 11:03:49 AM 252

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-3, Steam Generator Tube Leak Unit 1 Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

The following conditions exist at Unit 3:

Plant was operating at 100% when a 5 gpm SGTL developed on 34 SG.

The plant was shutdown in response to this condition.

Plant cooldown is in progress with RCS temperature at 499°F.

34 SG tube leakage has just increased to 275 gpm.

Which of the following describes the correct procedure in place at this time and the appropriate actions?

A. Cooldown is being performed per AOP-SG-1, Steam Generator Tube Leak.

Initiate SI and go to E-O, Reactor Trip or Safety Injection.

B. Cooldown is being performed per AOP-SG-1, Steam Generator Tube Leak.

Start SI Pumps and go to E-3, Steam Generator Tube Rupture.

C. Cooldown is being performed per POP-3.3, Plant Cooldown -Hot to Cold Shutdown. Initiate SI and go to E-O, Reactor Trip or Safety Injection.

D. Cooldown is being performed per POP-3.3, Plant Cooldown -Hot to Cold Shutdown. Start SI Pumps and go to E-3, Steam Generator Tube Rupture.

Answer: A Tuesday, April 30, 2019 11:03:49 AM 253

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator must recall that 50 gpd leakage on unit 1 requires a shutdown per AOP-3. AOP-3 will required a shutdown of the unit per AOP-17A or OP 3A depending on the speed of the shutdown required. Given the initial condition of 50 gpd, a shutdown to MODE 3 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is required, so OP 3A will be utilized.

AOP-3 is the procedure in effect, and also controls the cooldown of the plant after shutdown. When pressurizer level cannot be maintained within 7% of programmed level, which will happen given the rise of the SGTL to 275 gpm, the procedure in effect requires a manual SI, CI and transition to EOP-0.

A INCORRECT: The first part is correct, see above. The second part is incorrect but plausible because if these actions were taken, this would work to mitigate the SGTR.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, but plausible as the allowable SGTL amount for unit 2 is different and at 50 gpd, AOP-3 would not be the procedure in effect, and the shutdown would be performed per OP 3A. The second part is correct, see above.

D INCORRECT: The first part is incorrect, but plausible as the allowable SGTL amount for unit 2 is different and at 50 gpd, AOP-3 would not be the procedure in effect, and the shutdown would be performed per OP 3A. The second part is incorrect but plausible because if these actions were taken, this would work to mitigate the SGTR.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following events:

a. Reactor Coolant System leakage

b. Reactor Coolant Pump malfunctions

c. Steam Generator Tube leak (055.03.LP2438.004)

Tuesday, April 30, 2019 11:03:49 AM 254

QUESTIONS REPORT for 2019 NRC Exam Master

79. 2019 NRC 079/APE/056AA2.44/4.5/3-SPR/SRO/NEW/TS 3.8.1/057.02.LP3344.002 Given the following:

Both units are at Rated Thermal Power The electric plant is in a normal lineup An electrical perturbation causes the indication pictured on the next page Operators have restored D-07, DC Station Battery Charger, to service Considering UNIT 1 ONLY, which of the following:

(1) Lists the required actions for LCO 3.8.1 Electrical Power Systems, AC Sources - Operating?

AND (2) The operability status of 1P-15A, SI pump?

(Assume no additional operator action)

REFERENCE PROVIDED TS 3.8.1 (5 pages)

A. (1) Enter TSAC 3.8.1.B, 3.8.1 E and 3.8.1F (2) Inoperable, due to the loss of offsite power B. (1) Enter TSACs 3.8.1.B, 3.8.1.C, 3.8.1.D, 3.8.1 E and 3.8.1F (2) Inoperable, due to the loss of offsite power C. (1) Enter TSAC 3.8.1.B, 3.8.1 E and 3.8.1F (2) Operable, 1P-15A will be declared inoperable after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. (1) Enter TSACs 3.8.1.B, 3.8.1.C, 3.8.1.D, 3.8.1 E and 3.8.1F (2) Operable, 1P-15A will be declared inoperable after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Tuesday, April 30, 2019 11:03:49 AM 255

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: New Question History:

None K/A:

056AA2.44 Loss of Offsite Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Indications of loss of offsite power.

(Imp 4.5)

Justification for K/A Match:

Matches the K/A by requiring identification of the loss of offsite power given control room indications.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to assess plant conditions utilize TS bases for redundant power supplies in the application of required actions (TS Section 3) and apply completion times for equipment, declaring it inoperable.

Cognitive Level:

Comprehension 3-SPR: The operator must understand the plant conditions and final lineup of the cause the by perturbation, and apply tech specs to determine which action(s) are required to be entered and the status (operable or inoperable) for 1P-15A.

10 CFR Part 55 Content:

55.41 10 55.43 2 Tuesday, April 30, 2019 11:03:49 AM 256

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 3.8.1, Electrical Power Systems, AC Sources - Operating Rev 4 TS B 3.8.1 Basis - Electrical Power Systems, AC Sources - Operating Rev 20 Proposed reference to be provided to the applicants during examination:

TS 3.8.1, Electrical Power Systems, AC Sources - Operating Rev 4 PAGES 1-5 Original Question:

None Justification:

The operator must diagnose the loss of offsite power to the safeguard busses, complicated by G03, EDG not starting. Given the assumption that no further actions have occurred after the restoration of D-07, the operator must determine that Unit 1 is the only unit affected by the loss of the G03 and that TSAC-3.8.1.B/C/D need to be entered based on the initial conditions. The operator must also determine that 1P-15A is operable but must complete required action D.1 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> form discovery.

A INCORRECT: The first part is incorrect, plausible as this is one of the TSACs needed to be entered, but not a completed list. The second part is correct based on the loss of offsite power with loss of G03.

B INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has the misconception that the loss of offsite power causes the pump to immediately inoperable C INCORRECT: The first part is incorrect, plausible as this is one of the TSACs needed to be entered, but not a completed list. The second part is incorrect, plausible if the operator determine that after the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that 1P-15A is no longer operable.

D CORRECT: See above.

Learning Objective:

Given specific plant conditions, ASSESS and APPLY Improved Technical Specification requirements as appropriate.

(057.02.LP3344.002)

Tuesday, April 30, 2019 11:03:49 AM 257

QUESTIONS REPORT for 2019 NRC Exam Master

80. 2019 NRC 080/APE/065AG2.1.31/4.3/3-PEO/SRO/NEW/AOP-5B/055.03.LP2439.004 Given the following:

Unit 1 is at Rated Thermal Power Steam Generator A NR is 60% and LOWERING Steam Generator B NR is 60% and LOWERING South Instrument Air Header pressure is 58 psig and LOWERING SLOWLY The following annunciators are LIT:

1C03 1D 1-9, 1MS-2018 SG A MAIN STM STOP VALVE AIR PRESSURE LOW 1C03 1D 2-9, 1MS-2017 SG B MAIN STM STOP VALVE AIR PRESSURE LOW 1C03 1D 4-9, BLEEDER TRIP VALVES AIR PRESSURE LOW 1C03 1E2 2-1, 1P-28A SG FEED PUMP SUCTION PRESSURE LOW 1C03 1E2 2-5, 1P-28B SG FEED PUMP SUCTION PRESSURE LOW 1CO3 1F 3-2, MSR A, B, C, OR D HOTWELL LEVEL HIGH C01A 1-8, AIR SYSTEMS COMMON TROUBLE C01A 1-9, INSTRUMENT AIR HEADER PRESSURE LOW Which of the following correctly completes the statement below?

The Feed Regulating Valve controller DEMAND signal indication is ___(1)___,

and the OS1 should direct entry into ___(2)___

A. (1) Going to 100% OPEN (2) AOP-2B, Feedwater System Malfunction, reduce power per AOP-17A, Rapid Power Reduction B. (1) Going to 100% CLOSED (2) AOP-2B, Feedwater System Malfunction, reduce power per AOP-17A, Rapid Power Reduction C. (1) Going to 100% OPEN (2) AOP-5B, Loss of Instrument Air, direct the reactor be tripped D. (1) Going to 100% CLOSED (2), AOP-5B, Loss of Instrument Air, direct the reactor be tripped Tuesday, April 30, 2019 11:03:49 AM 258

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: New Question History:

None K/A:

065AG2.1.31 Loss of Instrument Air Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(Imp 4.3)

Justification for K/A Match:

Matches the K/A by requiring identification of the correct control room indication given the plant conditions.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure implementation will mitigate consequences of an event and actions which are necessary.

Cognitive Level:

Comprehension 3-PEO: The operator must analyze the initial conditions, determine the event and use knowledge to determine what course of actions are needed to mitigate the situation given the initial conditions.

10 CFR Part 55 Content:

55.41 10 55.43 Tuesday, April 30, 2019 11:03:49 AM 259

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-5B, Loss of Instrument Air Rev 49 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The operator must diagnose the loss of instrument air based on the conditions.

The loss of instrument air will cause the failure of the main feed regulating valves to have a demand of at or going to 100% open, while actual valve position will be going shut, depending on the actual air pressure in the supply line to the valve. With instrument air pressure that low, procedure direction is to trip the reactor.

A INCORRECT: The first part is correct. The second part is incorrect, as this is the incorrect procedure to enter based on plant indications and annunciators, but plausible, as there are several places in the procedure would require a reduction in power.

B INCORRECT: The first part is incorrect but plausible if the operator has the incorrect concept of how the controller functions during a loss of instrument air pressure. The second part is incorrect, as this is the incorrect procedure to enter based on plant indications and annunciators, but plausible, as there are several places in the procedure would require a reduction in power.

C CORRECT: See above.

D INCORRECT: The first part is incorrect but plausible if the operator has the incorrect concept of how the controller functions during a loss of instrument air pressure. The second part is correct.

Learning Objective:

Given access to the Site Specific Simulator, EVALUATE and RESPOND to plant indications associated with the following events:

a. Secondary Coolant System leak

b. Feedwater system malfunction

c. Loss of Condenser Vacuum Loss of Instrument Air (055.03.LP2439.004)

Tuesday, April 30, 2019 11:03:49 AM 260

QUESTIONS REPORT for 2019 NRC Exam Master

81. 2019 NRC 081/EPE/E11EA2.1/4.2/3-SPR/SRO/BANK/ECA-1.2/031.02.LP0465.008 Given the following:

A Reactor Coolant Leak has been identified in the Primary Auxiliary Building on Unit 1 Efforts to enter the area to isolate the leak have been unsuccessful The Crew has worked their way through the EOP network and are about transition from ECA-1.2, LOCA Outside Containment RWST level is 50% and LOWERING SLOWLY Reactor Coolant System pressure continues to LOWER SLOWLY Which of the following is the correct procedure to transition to and why?

A. EOP-1, Loss of Reactor or Secondary Coolant, to continue actions to address the LOCA.

B. ECA-1.1, Loss of Containment Sump Recirculation, to identify the location of the LOCA and isolate it.

C. ECA-1.1, Loss of Containment Sump Recirculation, to address the loss of inventory available for core cooling.

D. EOP-1, Loss of Reactor or Secondary Coolant, to initiate actions to lower RCS pressure to minimize break flow to slow the loss of inventory.

Tuesday, April 30, 2019 11:03:49 AM 261

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 1 Source: Bank Question History:

2008 Ginna Question 80 K/A:

E11EA2.1 Loss of Emergency Coolant Recirculation Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(Imp 4.2)

Justification for K/A Match:

Matches the K/A by requiring selection of the appropriate procedure for a given set of initial conditions during emergency operations.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure will need to be transitioned to and to procedure background as to why that transition is necessary.

Cognitive Level:

Comprehension 3-SPR: The operator must understand the plant conditions and where they are procedurally; make a decision on where to transition to and the reason for this choice.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 262

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

ECA-1.2, LOCA Outside Containment Unit 1 Rev 25 BG-ECA-1.2, LOCA Outside Containment Rev 15 Proposed reference to be provided to the applicants during examination:

None Original Question:

Plant conditions occurred as follows:

A Reactor Coolant Leak has been identified in the Auxiliary Building.

Efforts to enter the area to isolate the leak have been unsuccessful.

The Control Room team has worked their way through the EOPs and are about to leave ECA-1.2, LOCA Outside Containment.

RWST level is 55% and lowering slowly.

Reactor Coolant System pressure and Pressurizer water level continue to lower slowly.

Which procedure will be transitioned to and why?

a. E-1, Loss of Reactor or Secondary Coolant, to continue actions to address the LOCA.

b. ECA-1.1, Loss of Emergency Coolant Recirculation, to identify the location of the LOCA and isolate it.

c. E-1, Loss of Reactor or Secondary Coolant, to initiate actions to lower RCS pressure to minimize break flow to slow the loss of inventory.

d. ECA-1.1, Loss of Emergency Coolant Recirculation, to address the loss of inventory available for core cooling Proposed answer: D Tuesday, April 30, 2019 11:03:49 AM 263

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator must determine where they are procedurally and determine since the leak is not isolated, the loss of makeup inventory for core cooling requires actions and transition to ECA-1.1 is warranted.

A INCORRECT: The first part is incorrect, plausible because this is the transition for continuing actions in the EOP network if the leak is isolated.

The second part is not correct, plausible as actions in EOP-1 will address a LOCA.

B INCORRECT: The first part is correct. The second part is incorrect, due to the LOCA being outside of containment, but plausible as EOP-1 does have steps that address LOCAs.

C CORRECT: See above.

D INCORRECT: The first part is incorrect, plausible because this is the transition for continuing actions in the EOP network if the leak is isolated.

The second part is not correct, plausible as this action would in fact slow the loss of inventory, but are not prudent at this time.

Learning Objective:

Given appropriate conditions/parameters or access to the site specific simulator, implement the following procedures for the specified conditions:

a. ECA-1.1 to respond to a loss of Containment Sump Recirculation

b. ECA-1.2 to respond to an intersystem LOCA

c. ECS-1.3 to respond to containment sump blockage

d. ECA-2.1 to respond to both Steam Generators being faulted (031.02.LP0465.008)

Tuesday, April 30, 2019 11:03:49 AM 264

QUESTIONS REPORT for 2019 NRC Exam Master

82. 2019 NRC 082/APE/024AG2.1.27/4.0/1-B/SRO/NEW/TS 3.1.1/057.02.LP3338.001 TS 3.1.1, Shutdown Margin (SDM) utilizes the Control Rod System and Boration System to ensure the LCO is met. Per the basis of TS 3.1.1, answer the following:

(1) In MODE 2 with Keff (<1.0) the function of the Control Rod System is to show protection against what most limiting accident?

AND (2) In MODE 5 the function of the Boration System is to show protection against what most limiting accident?

A. (1) Main Steam Line Break (2) Inadvertent Boron Dilution B. (1) Main Steam Line Break (2) Uncontrolled Cooldown C. (1) Control Rod Ejection (2) Inadvertent Boron Dilution D. (1) Control Rod Ejection (2) Uncontrolled Cooldown Tuesday, April 30, 2019 11:03:49 AM 265

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 2 Source: New Question History:

None K/A:

024AG2.1.27 Emergency Boration Knowledge of system purpose and/or function.

(Imp 4.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the function of the system in the form of the accident protected against.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to have knowledge of the TS bases and terminology.

Cognitive Level:

Knowledge 1-B: The operator must recall the most limiting accident protected against.

10 CFR Part 55 Content:

55.41 7 55.43

Reference:

TS B 3.1.1, Basis - Shutdown Margin (SDM) Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:49 AM 266

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per the basis for TS 3.1.1, The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions. The most limiting accident for the SDM requirements is based on a main steam line break (MSLB) For MODE 5, the primary safety analysis that relies on the SDM limit is the boron dilution analysis.

A CORRECT: See above.

B INCORRECT: The first part is correct. The second part is incorrect, but plausible as an uncontrolled cooldown is part of the MSLB which is the most limiting accident while operating.

C INCORRECT: The first part is incorrect, but plausible as the control rod system must show protection against a rod ejection. The second part is correct.

D INCORRECT: The first part is incorrect, but plausible as the control rod system must show protection against a rod ejection. The second part is incorrect, but plausible as an uncontrolled cooldown is part of the MSLB which is the most limiting accident while operating.

Learning Objective:

IDENTIFY and DISCUSS the Technical Specifications (TS) associated with the following Reactivity Control Systems components, parameters, and operation including Limiting Conditions for Operation (LCO),LCO Applicability, Action Conditions, Required Actions and Surveillance Requirements as they pertain to the following requirements:

a. Shutdown Margin

b. Core Reactivity

c. Moderator Temperature Coefficient

d. Rod Group Alignment Limits

e. Shutdown Bank Insertion Limits

f. Control Bank Insertion Limits

g. Rod Position Indication

h. Physics Tests Exceptions-Mode 2 (057.02.LP3338.001)

Tuesday, April 30, 2019 11:03:49 AM 267

QUESTIONS REPORT for 2019 NRC Exam Master

83. 2019 NRC 083/APE/060AG2.4.30/4.1/3-SPR/SRO/BANK/EPIP 1.2.1/SD86.4.2.4.30 Given the following:

Both Units are at Rated Thermal Power Chemistry is in the process of sampling T-20A, Gas Decay Tank when the following occurs:

2204 Annunciator 1C20 B 2-9, COMMON PROCESS RADIATION MONITOR HIGH alarms 2205 RE-317 AB Exhaust Mid Range Gas Monitor, is reading 4.9E-2 Ci/cc AND RP reports the release is two times ODCM limit 2208 The Auxiliary Operator reports the diaphragm on WG-1607A, T-20A GDT Master Inlet Isolation is cracked 2209 RP is notified to survey at WG-1607A 2210 Maintenance notified to determine time required to repair/terminate leak 2212 RE-317, peaks at 1.7E+0 Ci/cc 2217 RE-317, is reading 1.1+0 Ci/cc and is LOWERING 2219 RE-317, reading is 6.2E-4 Ci/cc and STABLE 2223 RP reports the release rate from WG-1607A is three times ODCM limit 2245 Maintenance reports it will take 30 minutes to isolate/terminate the leak Which of the choices completes the following statement?

The highest required emergency classification is ___(1)___ and the NRC is required to be notified within ___(2)___.

REFERENCE PROVIDED EAL Charts (2 pages)

A. (1) UNUSUAL EVENT (2) 15 minutes B. (1) UNUSUAL EVENT (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) ALERT (2) 15 minutes D. (1) ALERT (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tuesday, April 30, 2019 11:03:49 AM 268

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 2 Source: Bank Question History:

2016 North Anna Question 83 K/A:

060AG2.4.30 Accidental Gaseous Radwaste Release Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

(Imp 4.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to have the knowledge of events that must be reported to internal or external agencies, such as the NRC.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant parameters and determine selection of a notification based on that diagnosis. The operator must have knowledge of the EAL chart and have knowledge of the required EP procedures for proper time of notifications.

Cognitive Level:

Comprehension 3-SPR: The operator must assess plant parameters and then determine the selection of a notification based on that assessment. The operator must have knowledge of the EAL chart and have knowledge of the required EP procedures for proper time of notifications.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 269

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

Draft EAL Wallboards EPIP 2.1, Notifications - ERO, State and Counties and NRC Rev 57 Proposed reference to be provided to the applicants during examination:

EAL Wall boards Draft 2 pages (Hot and ALL MODE)

Original Question:

Both Units are at 100% power. Chemistry is in the process of sampling "A" WGDT (1-GW-TK-1A) when the following occurs:

2204 Unit 2 receives annunciator 2B-B5, PROCESS VENT VNT STACK A&B HI HI RADIATION.

2205 1-VG-RI-180-2, Vent Stack B, is reading 4E+5 uCi/sec 2208 The Auxiliary Building operator reports the diaphragm on 1-GW-TK-1A sample isolation valve is cracked and cannot be isolated.

2209 HP is notified to survey at 1-GW-TK-1A 2210 Mechanics notified to determine time required to repair/terminate leak.

2215 1-VG-RI-180-2, Vent Stack B, peaks at 4.06E+6 uCi/sec 2217 1-VG-RI-180-2, Vent Stack B, is reading 3E+6 uCi/sec and is decreasing 2221 1-VG-RI-180-2, Vent Stack B, reading is steady at 2E+5 uCi/sec 2223 HP confirms the release rate from 1-GW-TK-1A is three times ODCM limit 2245 Mechanics report it will take them 30 minutes to isolate the leak.

Which of the choices below completes the following statement?

The highest required emergency classification is ___(1)___ and the NRC is required to be notified within ___(2)___.

REFERENCE PROVIDED A. (1) NOUE (2) 15 minutes B. (1) NOUE (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) ALERT (2) 15 minutes D. (1) ALERT (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> None Proposed answer is B Tuesday, April 30, 2019 11:03:49 AM 270

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator will declare an UNUSUAL EVENT based on RU1.2. This will be based on the confirmation of 3 times the ODCM rate combined with the report from the Mechanics of 30 minutes to isolate. The total time of the leak/release will have been 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 11 minutes from the start of the leak. The notification will be made per EPIP 2.1, which contains the Precaution and Limitation of The NRC shall be notified immediately following the state and county notifications, not to exceed on hour from declaration of a classified emergency."

A INCORRECT: The first part is correct. The second part is incorrect, plausible because 15 minutes is the required time for the notification of the State and local government agencies.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, but plausible RE-317 reading peaks at a value greater than the Alert level, as the operator could select RA1.1 based on the reading being greater than the alert level, but it was not sustained for the required 15 minutes. The second part is correct.

D INCORRECT: The first part is incorrect, but plausible RE-317 reading peaks at a value greater than the Alert level, as the operator could select RA1.1 based on the reading being greater than the alert level, but it was not sustained for the required 15 minutes. The second part is incorrect, plausible because 15 minutes is the required time for the notification of the State and local government agencies.

Learning Objective:

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

(SD86.4.2.4.30)

Tuesday, April 30, 2019 11:03:49 AM 271

QUESTIONS REPORT for 2019 NRC Exam Master

84. 2019 NRC 084/EPE/E07EA2.1/4.0/3-SPR/SRO/BANK/ECA-3.2/031.02.LP0473.008 Given the following:

A Steam Generator Tube Rupture has occurred Due to equipment failures and conditions in the facility, the crew transitioned to and is performing the steps of ECA-3.2, SGTR with Loss of Reactor Coolant -

Saturated Recovery Desired The STA informs you that all CSF Status Trees are GREEN with the exception of the following:

Core Cooling -YELLOW path for CSP-C.3, Response to Saturated Core Cooling Inventory -YELLOW path for CSP-I.2, Response to Low Pressurizer Level Which of the following describes the required procedural action for this event, and the reason?

A. Transition to CSP-I.2 to restore the Inventory to a GREEN condition B. Transition to CSP-C.3 to restore the Core Cooling to a GREEN condition C. Remain in ECA-3.2. The actions contained in CSP-C.3 and CSP-I.2 conflict with ECA-3.2 actions D. Remain in ECA-3.2. CSPs shall not be implemented prior to the completion of Emergency Contingency procedures.

Tuesday, April 30, 2019 11:03:49 AM 272

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 2 Source: Bank Question History:

2011 Ginna Retake Question 84 K/A:

E07EA2.1 Saturated Core Cooling Ability to determine and interpret the following as they apply to the (Saturated Core Cooling): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(Imp 4.0)

Justification for K/A Match:

Matches the K/A by requiring the operator to select the appropriate course of action based on initial conditions of a transition to SGTR with Loss of Reactor Coolant - Saturated Recovery Desired.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of mitigation strategy and requirements within procedures to determine the correct course of action.

Cognitive Level:

Comprehension 3-SPR: The operator must assess the mitigative strategy of the procedure in effect and determine what course of action(s) are necessary and why given watchstander report on plant conditions.

10 CFR Part 55 Content:

55.41 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 273

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

CSP-C.3, Response to Saturated Core Cooling Unit 1 Rev 13 CSP-I.2, Response to Low Pressurizer Level Unit 1 Rev 13 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

A Steam Generator Tube Rupture has occurred.

Due to equipment failures, the crew is performing actions contained in ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT -SATURATED RECOVERY DESIRED.

The STA informs you that all CSF Status Trees are GREEN with the exception of the following:

Core Cooling -YELLOW path for FR-C.3, RESPONSE TO SATURATED CORE CONDITIONS Inventory -YELLOW path for FR-I.2, RESPONSE TO LOW PRESSURIZER LEVEL Which ONE of the following describes the required implementation of procedures for this event, and the reason?

A. Transition from ECA-3.2 to FR-I.2 to restore the Inventory CSF to a green condition B. Transition from ECA-3.2 to FR-C.3 to restore the Core Cooling CSF to a green condition C. Remain in ECA-3.2. The actions contained in FR-C.3 and FR-I.2 conflict with ECA-3.2 actions D. Remain in ECA-3.2. Implementation of Yellow Path procedures is not allowed when using Emergency Contingency procedures.

Proposed Answer: C Tuesday, April 30, 2019 11:03:49 AM 274

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The operator should remain in ECA-3.2, both CSP-C.3 and CSP-I.2 contains conflicting actions from the goal of ECA-3.2. Both CSP-C.3 and CSP-I.2 has a caution to not perform this procedure when ECA-3.2 is the procedure in effect.

A INCORRECT: The first part is incorrect, CSP-I.2 contains a caution which states not to perform this procedure is ECA-3.2 is the procedure in effect. Plausible as there is no caution or note in ECA-3.2 concerning the implementation of CSPs and yellow path CSPs are at the discretion of the OS. The second part is incorrect, but is plausible as it is the correct answer if the transition was made.

B INCORRECT: The first part is incorrect, CSP-C.3 contains a caution which states not to perform this procedure is ECA-3.2 is the procedure in effect. Plausible as there is no caution or note in ECA-3.2 concerning the implementation of CSPs and yellow path CSPs are at the discretion of the OS. The second part is incorrect, but is plausible as it is the correct answer if the transition was made.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, there is no caution or note in ECA-3.2 which precludes the performance of CSPs, plausible as there are other procedures within the EOP network which contain cautions similar to this.

Learning Objective:

Given access to the Site Specific Simulator or appropriate plant/system conditions, implement the ECAs in response to each of the following event:

a. SGTR in faulted Steam Generator

b. SGTR with a LOCA

c. SGR with a loss of Pressurizer Pressure Control (031.02.LP0473.008)

Tuesday, April 30, 2019 11:03:49 AM 275

QUESTIONS REPORT for 2019 NRC Exam Master

85. 2019 NRC 085/EPE/E15EA2.2/3.3/3-SPK/SRO/BANK/EOP-1.3/043.03.LP1995.013 Given the following:

A reactor trip and Safety Injection occurred as a result of a large break LOCA about 90 minutes ago EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection is in progress Containment Spray has been aligned for recirculation and the crew is aligning charging pump suction to the VCT The following plant conditions are noted:

Reactor Coolant System Pressure 15 psig STABLE Core Exit Thermocouples 186°F LOWERING SLOWLY RCS Subcooling 60°F RISING SLOWLY Containment pressure 8 psig LOWERING SLOWLY Steam Generator A Level 55% STABLE Steam Generator B Level 58% STABLE Containment Sump 'A' level 10 inches STABLE Containment Sump 'B' level 90 inches RISING SLOWLY RWST Level 17% STABLE Containment Radiation Monitor 31 R/hr RISING SLOWLY Reactor Vessel Narrow Range Level 17 ft STABLE Reactor Vessel Wide Range Level 45 ft STABLE Service Water Header -South 35 psig STABLE Service Water Header - North 55 psig STABLE Reactor Coolant Pumps are secured Which of the following actions should be taken in response to these indications?

A. Stay in EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection, CSPs may not be implemented while this procedure is in progress.

B. Enter ECA-1.3, Containment Sump Blockage, based on high Containment Sump B level.

C. Enter CSP-P.1, Response to Imminent Pressurized Thermal Shock, and take actions including a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> RCS temperature soak.

D. Enter CSP-Z.2, Response to Containment Flooding, to address containment flooding and direct U2 OS to implement AOP-9A, Service Water Malfunction in parallel.

Tuesday, April 30, 2019 11:03:49 AM 276

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 1 Group 2 Source: Bank Question History:

2017 PBNP Question 90 Previous 2 NRC Exams K/A:

E15EA2.2 Containment Flooding Ability to determine and interpret the following as they apply to the (Containment Flooding): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

(Imp 3.3)

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Justification for K/A Match:

Matches the K/A by requiring the operator to assess plant conditions and based on that select the appropriate course of action and implement procedures to accomplish this.

Cognitive Level:

Comprehension 3-SPK: The operator must understand the initial condition, which one of the conditions must be calculated by the operator, and apply those conditions to the critical safety function status Trees, and interpret how the information impacts the EOP Network and determine the course of action to take next.

10 CFR Part 55 Content:

55.41 55.43 5

Reference:

CSP-ST.0, Critical Safety Function Status Trees Unit 1 Rev 10 EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection Unit 1 Rev 58 Proposed reference to be provided to the applicants during examination:

Tuesday, April 30, 2019 11:03:49 AM 277

QUESTIONS REPORT for 2019 NRC Exam Master None Original Question:

Given the following:

A reactor trip and Safety Injection occurred as a result of a large break LOCA about 90 minutes ago EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection is in progress Containment Spray has been aligned for recirculation and the crew is aligning charging pump suction to the VCT The following plant conditions are noted:

Reactor Coolant System Pressure 15 psig STABLE Core Exit Thermocouples 186°F LOWERING SLOWLY RCS Subcooling 60°F RISING SLOWLY Containment pressure 8 psig LOWERING SLOWLY Steam Generator A Level 55% STABLE Steam Generator B Level 58% STABLE Containment Sump 'B' level 90 inches RISING SLOWLY RWST Level 17% STABLE Containment Radiation Monitor 31 R/hr RISING SLOWLY Reactor Vessel Narrow Range Level 17 ft STABLE Reactor Vessel Wide Range Level 45 ft STABLE Service Water Header -South 35 psig STABLE Service Water Header - North 55 psig STABLE Reactor Coolant Pumps are secured Which of the following actions should be taken in response to these indications?

A. Stay in EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection, CSPs may not be implemented while this procedure is in progress.

B. Enter ECA-1.3, Containment Sump Blockage, based on high Containment Sump B level.

C. Enter CSP-P.1, Response to Imminent Pressurized Thermal Shock, and take actions including a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> RCS temperature soak.

D. Enter CSP-Z.2, Response to Containment Flooding, to address containment flooding and direct U2 OS to implement AOP-9A, Service Water Malfunction in parallel.

Proposed answer: D Tuesday, April 30, 2019 11:03:49 AM 278

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the initial conditions, the crew has aligned containment spray for recirculation, the operator must determine where in EOP-1.3 the crew currently is, and then based on that, determine if a transition out of EOP-1.3 to a CSP is allowed. This is based on the note at Step 1 Steps 1 through 32 should be performed without delay. CSPs should not be implemented prior to completion of these steps. The crew is currently on step 35. The AOP-9A actions are necessary to mitigate the high sump level to protect critical plant components necessary for plant recovery. Service Water indication provides further basis for addressing containment flooding.

A INCORRECT: The step in effect in EOP-1.3 is past the note which allows implementation of CSPs. Plausible because EOP-1.3 suspends implementation of CSPs for most of the procedure.

B INCORRECT: ECA-1.3 is incorrect because the indication provided do not result from sump blockage, and lowering CETs provide positive indication that sump recirculation is effective. This response is plausible if the examinee has a misconception that sump blockage could result in rising containment level.

C INCORRECT: Plausible as the temperature soak is normally the method utilized to mitigate the excessive cooldown. This will not be performed because RCS pressure is 15 psig, and since the crew is in EOP-1.3, RHR flow is greater than 550 gpm, so CSP-P.1 will be exited at step 1, and the crew will return to the procedure in effect; not taking the soak actions.

D CORRECT: See above.

Learning Objective:

IMPLEMENT the Critical Safety Function Status Tree and Critical Safety Procedure rules of usage.

(043.03.LP1995.013)

Tuesday, April 30, 2019 11:03:49 AM 279

QUESTIONS REPORT for 2019 NRC Exam Master

86. 2019 NRC 086/SYS/004A2.10/4.2/3-SPK/SRO/BANK/AOP-6F/055.03.LP3718.004 Given the following:

Unit 1 is entering a refueling outage RCS temperature is 130°F and STABLE RCPs are secured Both trains of RHR are aligned for decay heat removal with 'A' train providing cooling Draindown to 40% Pressurizer level per OP 4D Part 1, Draining the Reactor Coolant System, was just completed The Control Operator then reports the following information to you:

Pressurizer level is RISING SLOWLY Source range counts on both N-31 and N-32 are RISING 1C04 1C 2-8 BA FLOW DEVIATION OR POTENTIAL DILUTION IN PROGRESS alarm is LIT As the Operations Supervisor of that unit what actions are you going to take?

A. Use OP 4A, Filling and Venting Reactor Coolant System, Attachment D, Blending Operation for Unit 1 and Unit 2, to borate the RCS from the blender using an available Charging Pump.

B. Enter AOP-6F, Low Concentration Water Pockets in RCS, and borate the RCS from the RWST using a Safety Injection Pump.

C. Enter SEP-1, Degraded RHR System Capability, in an effort to raise RHR flow to 3000 gpm ensuring prompt mixing of the RCS.

D. Enter CSP-S.1, Response to Nuclear Power Generation/ATWS, and emergency borate the RCS from the RWST using all available Charging Pumps.

Tuesday, April 30, 2019 11:03:49 AM 280

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 1 Source: Bank Question History:

2015 PBNP Question 82 Previous 2 NRC Exams K/A :

004A2.10 Chemical and Volume Control System (CVCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent boration/dilution (Imp 4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to assess plant conditions where the CVCS system is being used during cooldown operations, and based on that select the appropriate course of action and implement procedures to mitigate the inadvertent dilution.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure implementation will mitigate consequences of an event and actions that are necessary.

Cognitive Level:

Comprehension 3-SPK: The operator must understand the initial condition, which one of the conditions must be calculated by the operator, and apply those conditions to the critical safety function status Trees, and interpret how the information impacts the EOP Network and determine the course of action to take next.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 281

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-6F, Low Concentration Water Pockets in RCS Rev 3 ARB 1C04 1C 2-8 BA Flow Deviation or Potential Dilution in Progress Rev 9 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 1 is entering a refueling outage RCS temperature is 130°F and stable RCPs are secured Both trains of RHR are aligned for decay heat removal with 'A' train providing cooling Draindown to 40% Pressurizer level per OP 4D Part 1, Draining the Reactor Coolant System, was just completed The Control Operator then reports the following information to you:

Pressurizer level is slowly rising Source range counts are rising on both N-31 and N-32 1C04 1C 2-8 BA FLOW DEVIATION OR POTENTIAL DILUTION IN PROGRESS alarm is LIT As the Operations Supervisor of that unit what actions are you going to take?

A. Use OP 4A, Filling and Venting Reactor Coolant System, Attachment D, Blending Operation for Unit 1 and Unit 2, to borate the RCS from the blender using an available Charging Pump.

B. Enter AOP-6F, Low Concentration Water Pockets in RCS, and borate the RCS from the RWST using a Safety Injection Pump.

C. Enter SEP-1, Degraded RHR System Capability, in an effort to raise RHR flow to 3000 gpm ensuring prompt mixing of the RCS.

D. Enter CSP-S.1, Response to Nuclear Power Generation/ATWS, and emergency borate the RCS from the RWST using all available Charging Pumps.

Proposed answer: B Tuesday, April 30, 2019 11:03:49 AM 282

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Given the plant indications, there are multiple entry conditions, and the AOP is referenced in the ARB, the actions are found in the procedure.

A INCORRECT: This is the proper procedure used to borate the RCS for recovery from refueling. Plausible as this is a procedure used during refueling.

B CORRECT: See above.

C INCORRECT: This procedure would potentially result in a faster mixing of the boron, also the requirements to enter this procedure are not met. Plausible as this procedure does provide a timely mitigation strategy.

D CORRECT: This procedure is used for boration, but when the plant is above 350°F. Plausible as this procedure is the one the operators usually use for boration when in the EOP Network.

Learning Objective:

Given access to the Site specific Simulator or plant conditions, DIAGNOSE and RESPOND to CVCS malfunction in accordance with the appropriate procedures(s).

(055.03.LP3718.004)

Tuesday, April 30, 2019 11:03:49 AM 283

QUESTIONS REPORT for 2019 NRC Exam Master

87. 2019 NRC 087/SYS/005G2.1.23/4.4/3-SPK/SRO/NEW/OP 3C/055.01.LP0272.001 Given the following:

Unit 1 is in MODE 3 The crew entered OP 3C, Hot Standby to Cold Shutdown to cool down the plant for a refueling outage During the course of the shutdown:

Both RCS loops will remain filled Level in both Steam Generators will be 40% narrow range Which of the following identifies the HIGHEST MODE the following activities can occur WITHOUT violating procedural requirements, or UNNECESSARILY entering a TSAC?

(1) The OS directing the parallel performance of OP 7A, Placing Residual Heat Removal System in Operation, to place RHR in service AND (2) After placing both trains of RHR in service, the removal of one train of RHR for extended maintenance of more than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

A. (1) MODE 4 (2) MODE 4 B. (1) MODE 4 (2) MODE 5 C. (1) MODE 5 (2) MODE 5 D. (1) MODE 5 (2) MODE 6 Tuesday, April 30, 2019 11:03:49 AM 284

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 1 Source: New Question History:

None K/A:

005G2.1.23 Residual Heat Removal System (RHRS)

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(Imp 4.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine integrated plant procedure coordination to establish RHR and remove a train of RHR during a shutdown.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of when to coordinate parallel procedure implementation to maneuver the plant.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to assess plant conditions and use knowledge of when to coordinate parallel procedure implementation to maneuver the plant, and determine what conditions are needed to repair the RHR pump, and when these can first be accomplished, by comparing the requirements of varying Tech Specs.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:49 AM 285

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

OP 3C, Hot Standby to Cold Shutdown Unit 1 Rev 17 TS 3.4.6, Reactor Coolant System (RCS) RCS Loops - MODE 4, Rev 3 TS 3.4.7 Reactor Coolant System (RCS) RCS Loops - MODE 5 Loops Filled Rev 3 TS 3.5.3, Emergency Core Cooling Systems (ECCS) ECCS - Shutdown Rev 2 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Tuesday, April 30, 2019 11:03:50 AM 286

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

OP 3C directs the performance of OP 7A to align the 'A' RHR train for decay heat removal between the temperatures of 330°F and 340°F which corresponds to MODE 4.

TS 3.4.6 requires two loops of the combination of RHR/RCS, but is modified by a note which allows an RHR pump to be shutdown for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

TS 3.4.7, requires one RHR loop operable and in operation and either and additional RHR loop operable or > 35% narrow range in 1 SG in MODE 5 TS 3.5.3, requires one loop of ECCS be operational in MODE 4, but an RHR train may be considered operable during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation.

Even though TS 3.5.3 only requires one loop in operation in MODE 4 it must able capable to be manually realigned, TS 3.4.6, will only allow this alignment for an hour, and 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to complete the maintenance is needed. TS 3.4.7 allows this in MODE 5.

A INCORRECT: The first part is correct. The second part is incorrect, but is plausible if the operator only apply TS 3.5.3 to this situation which would allow only needing one RHR loop aligned for decay heat removal to be met.

B CORRECT: See above.

C INCORRECT: The first part is incorrect, but plausible if the operator incorrectly recalls the shutdown sequence of the plant and applying that the SGs are still effective until >200°F. The second half is correct.

D INCORRECT: The first part is incorrect, but plausible if the operator incorrectly recalls the shutdown sequence of the plant and applying that the SGs are still effective until >200°F. The second half is incorrect, plausible in this MODE.

Learning Objective:

DESCRIBE the procedures which govern the operations associated with taking the plant from Hot Standby Operation to Cold Shutdown Operation. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed Operators.

(055.01.LP0272.001)

Tuesday, April 30, 2019 11:03:50 AM 287

QUESTIONS REPORT for 2019 NRC Exam Master

88. 2019 NRC 088/SYS/022A2.06/3.2*/3-SPK/SRO/NEW/CSP-Z.1/043.03.LP2000.007 Given the following:

Unit 1 was at Rated Thermal Power when a Steam Generator Fault AND LOCA occurred in containment 1B-04, 480V Safeguards bus normal supply breaker tripped open concurrent with the reactor trip The crew has transitioned to EOP-2, Faulted Steam Generator Isolation The STA notes the indications pictured on PPCS screen 2220 on the following page Which of the following is . . .

(1) The correct course of action to take for these indications is to transition to CSP-Z.1, and . . .

AND (2) The reason for this course of action?

EOP-0, Reactor Trip or Safety Injection CSP-Z.1, Response to High Containment Pressure A. (1) complete and return to procedure and step in effect.

(2) 1P-14A, Containment Spray pump tripped B. (1) remain in CSP-Z.1 until pressure is less than 25 psig.

(2) 1P-14A, Containment Spray pump tripped C. (1) complete and return to procedure and step in effect (2) 1P-14A, Containment Spray pump secured per EOP-0, Attachment A D. (1) remain in CSP-Z.1 until pressure is less than 25 psig (2) 1P-14A, Containment Spray pump secured per EOP-0, Attachment A Tuesday, April 30, 2019 11:03:50 AM 288

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 1 Source: New Question History:

None K/A:

022A2.06 Containment Cooling System (CCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of CCS pump (Imp 3.2*)

Justification for K/A Match:

Matches the K/A by requiring the operator to diagnose the cause of the containment pressure rise (the loss of the second containment spray pump), and select the appropriate actions to mitigate the rise in containment pressure.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of procedure implementation to mitigate consequences of an even.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to assess plant conditions with the use of the PPCS display, and determine the cause of the pressure rise in the containment, and then determine the actions for mitigation.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 289

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

CSP-Z.1, Response to High Containment Pressure Unit 1 Rev 29 BG-CSP-Z.1, Background Response to High Containment Pressure Unit 1 Rev 20 OM 3.7, AOP and EOP Procedure Usage for Response to Plant Transients Rev 31 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The operator must diagnose the following from indications of the PPCS show that there is no flow from either containment spray pump. 1P-14B was lost with the lockout of bus 1B-04 during the start of the event. 1P-14A has tripped, with the cause being unknown. Entry into CSP-Z.1 is required due to containment pressure being greater than 25 psig with less than one train of containment cooling. CSP-Z.1 contains a return to procedure and step in effect step, and the background contains a statement if the Containment function is not restored to green "If this is the case, this procedure does not need to be implemented again since all necessary actions have already been performed." OM 3.7, Section 6.18.11 contains a similar statement.

A CORRECT: See above.

B INCORRECT: The first part is correct, The second part is incorrect, but plausible because several CSP procedures will have a loop in the procedure if actions are unsuccessful in mitigating an even C INCORRECT: The first part is incorrect plausible, as Attachment A of EOP-0 secures a containment spray pump to conserve RWST water during a large break LOCA. The second part is correct D INCORRECT: The first part is incorrect plausible, as Attachment A of EOP-0 secures a containment spray pump to conserve RWST water during a large break LOCA. The second part is incorrect, but plausible because several CSP procedures will have a loop in the procedure is actions are unsuccessful in mitigating an event.

Learning Objective:

Implement the CSPs to respond to plant conditions where the Containment Status Tree is not satisfied.

(043.03.LP2000.007)

Tuesday, April 30, 2019 11:03:50 AM 290

QUESTIONS REPORT for 2019 NRC Exam Master

89. 2019 NRC 089/SYS/062A2.06/3.9/3-SPK/SRO/MODIFIED/AOP-18A/055.03.LP2440.002 Given the following:

Unit 1 is in MODE 1 Unit 2 is in MODE 6 for a refueling outage All required Auxiliary Feedwater pumps are OPERABLE All six Service Water pumps are OPERABLE All required offsite power sources to Units 1 & 2 are OPERABLE G02, Emergency Diesel Generator is OUT OF SERVICE with G01, Emergency Diesel Generator aligned to both 4160v Safeguards Buses.

Given these conditions, which statements correctly . . .

(1) describes the requirements associated with cross-tying 2B-03 to 2B-04, Unit 2 480v AC Safeguards buses? The buses may be cross tied for . . .

AND (2) In an emergency, which procedure specifically verifies 2B-03 and 2B-04 are no longer cross-tied by tie breaker position during a loss of power event?

A. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided 1 RHR loop is OPERABLE OR the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange.

(2) AOP-18A, Train A Equipment Operation OR AOP-18B, Train B Equipment Operation B. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided 1 RHR loop is OPERABLE OR the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange.

(2) AOP-18, Electrical System Malfunction C. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided 1 RHR loop is OPERABLE AND the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange.

(2) AOP-18A, Train A Equipment Operation OR AOP-18B, Train B Equipment Operation D. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided 1 RHR loop is OPERABLE AND the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange.

(2) AOP-18, Electrical System Malfunction Tuesday, April 30, 2019 11:03:50 AM 291

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 1 Source: Modified Question History:

2011 PBNP Question 90 K/A:

062A2.06 AC Electrical Distribution System Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Keeping the safeguards buses electrically separate (Imp 3.9)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine what condition are necessary to enter in to a cross-tied safeguards bus condition, and what procedure will ensure the cross-tied buses have been electrically separated in a loss of power event.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure has the specific actions needed to be taken to mitigate consequences of an event.

Cognitive Level:

Comprehension 3-SPK: Requires the operator to assess plant conditions determine what additional requirements are necessary to enter a safeguards bus cross-tied condition, and then during a loss of power event, determine which procedure will specifically validate the tie breakers for the safeguards buses are open.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 292

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 3.8.9, Electrical Power System - Distribution Systems - Operating Rev 1 TS 3.8.10, Electrical Power System - Distribution Systems - Shutdown Rev 1 AOP-18A, Train A Equipment Operation Rev 17 AOP-18B, Train B Equipment Operation Rev 15 Proposed reference to be provided to the applicants during examination:

None Original Question:

Unit 1 is in MODE 1.

Unit 2 is in MODE 6 for a refueling outage.

All Auxiliary Feedwater pumps are OPERABLE All six Service Water pumps are OPERABLE All required offsite power sources to Units 1 & 2 are OPERABLE G02, Emergency Diesel Generator is OUT OF SERVICE with G01, Emergency Diesel Generator aligned to both 4160v Safeguards Buses.

Given these conditions, which of the following statements correctly describes the requirements associated with cross-tying 2B-03 to 2B-04, Unit 2 480v AC Safeguards buses?

The buses may be cross tied for ___(1)___ provided ___(2)___.

A. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (2) 1 RHR loop is OPERABLE OR the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange B. (1) up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (2) 1 RHR loop is OPERABLE AND the Unit 2 Reactor Cavity water level is maintained > 23 feet above the Reactor Vessel flange C. (1) up to 7 days (2) loads on the Unit 2 480v Safeguards Buses are limited to avoid overloading the standby Emergency Diesel Generator D (1) up to 7 days (2) G02, Emergency Diesel Generator is restored to an OPERABLE status and re-aligned to its associated 4160v Safeguards Bus Proposed answer: B Tuesday, April 30, 2019 11:03:50 AM 293

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per TS 3.8.10, one of two conditions must be met to cross-tie the safeguards buses being (1) Two residual heat removal loops are operable when the unit is in MODE5 or MODE 6 with reactor cavity water level < 23 Ft or (2) One residual heat removal loop is operable with the unit is in MODE 6 with reactor cavity water level > 23 Ft. During a loss of power event, AOP-18A or AOP-18B, will specifically check 2B52-40C, 2B-03 to 2B-04 Bus Tie Break - OPEN during the restoration of the of the cross-tied bus (2B-03 or 2B-04).

A INCORRECT: The first part is incorrect as 1 RHR loop is needed if the if cavity water level is kept at > 23 Ft, but the OR makes this conditions, so 2 RHR loops are needed, but plausible since is cavity water level is > 23 Ft only one loop of RHR is needed. The second part is correct.

B INCORRECT: The first part is incorrect as 1 RHR loop is needed if the if cavity water level is kept at > 23 Ft, but the OR makes this conditions, so 2 RHR loops are needed, but plausible since is cavity water level is > 23 Ft only one loop of RHR is needed. The second part is incorrect, plausible as this procedure will be used after normal power to safeguards busses is restored and the EDG are unloaded and stopped.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible as this procedure will be used after normal power to safeguards busses is restored and the EDG are unloaded and stopped.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following conditions:

a. Turbine Generator Voltage Regulator failure

b. Loss of Main Generator Hydrogen pressure

c. Total collapse of 345 KV system frequency

d. Loss of electrical buses (055.03.LP2440.002)

Tuesday, April 30, 2019 11:03:50 AM 294

QUESTIONS REPORT for 2019 NRC Exam Master

90. 2019 NRC 090/SYS/063G2.4.49/4.4/3-PEO/SRO/NEW/AOP-0.0/055.03.LP2440.002 Given the following:

Unit 1 is operating at Rated Thermal Power An electrical perturbation causes the loss of D11, 125 VDC Distribution Panel Which of the following actions are necessary?

ECA-0.0, Loss of All AC Power EOP-0, Reactor Trip or Safety Injection EOP-0.1, Reactor Trip Response AOP-0.0, Vital DC System Malfunction AOP-18, Electrical System Malfunction AOP-18B, Train B Equipment Operation AOP-19A, Train A Safeguards Bus Restoration OI 35A, Standby Emergency Power Alignment A. An automatic reactor trip should have occurred.

Enter ECA-0.0, to restore power to the safeguard busses, and then perform AOP-18 to restore power to other de-energized buses.

B. An automatic reactor trip should have occurred Enter EOP-0, perform immediate actions, and then AOP-0.0 should be performed in parallel with EOP-0.1 to restore power to de-energized busses C. An automatic reactor trip should NOT have occurred.

Enter AOP-19A, to restore power to de-energized buses, and perform AOP-18B, in parallel as a secondary propriety to AOP-19A, until completed.

D. An automatic reactor trip should NOT have occurred Enter AOP-0.0, to align G02, Emergency Diesel Generator to 1A-05, 4160 VAC Safeguards Bus per OI 35A, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> due to the loss of control power to G01 Tuesday, April 30, 2019 11:03:50 AM 295

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 1 Source: New Question History:

None K/A:

063G2.4.49 DC Electrical Distribution System Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(Imp 4.4)

Justification for K/A Match:

Matches the K/A by requiring the operator to determine if a reactor trip occurred and perform immediate actions which are done without a procedure until the verification portion.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of which procedure implementation will mitigate consequences of an event and how to coordinate procedure usage.

Cognitive Level:

Comprehension 3-PEO: Requires the operator to assess plant conditions determine what the effect of the perturbation was and based on those effects, take the appropriate actions and coordinate procedures to mitigate the event.

10 CFR Part 55 Content:

55.41 10 55.43 2 Tuesday, April 30, 2019 11:03:50 AM 296

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

AOP-0.0, Vital DC System Malfunction Rev 36 OM 3.7, AOP and EOP Procedure Usage for Response to Plant Transients Rev 31 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

The loss of D11 will cause a single unit trip, the operator can either enter EOP-0 directly based on the reactor trip, or enter AOP-0.0, which will check for a reactor trip, and if one has occurred, direct the operator to stabilize plant using EOPs while continuing in this procedure.

A INCORRECT: The first part is correct. The second part is incorrect, but plausible, as this procedure can be used to address de-energized busses and normalizing the plant after power restoration to the safeguard bussese.

B CORRECT: See above.

C INCORRECT: The first part is incorrect a unit 1 reactor trip should have occurred. The second part is incorrect but plausible as there was a loss of breaker indications the A safeguards train.

D INCORRECT: The first part is incorrect a unit 1 reactor trip should have occurred. The second part is incorrect but plausible if the operator confuses D11 with D01, as this action is necessary if D01 has lost power and D01 supplies D11.

Learning Objective:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following conditions:

a. Turbine Generator Voltage Regulator failure

b. Loss of Main Generator Hydrogen pressure

c. Total collapse of 345 KV system frequency

d. Loss of electrical buses (055.03.LP2440.002)

Tuesday, April 30, 2019 11:03:50 AM 297

QUESTIONS REPORT for 2019 NRC Exam Master

91. 2019 NRC 091/SYS/029A2.01/3.6/1-F/SRO/NEW/OP 9C/051.05LP0057.005 Given the following:

Unit 1 is in MODE 6 1200 03/27/19 - Containment air sample for air particulate, noble gas iodine, and tritium was taken 2000 03/27/19 - Containment Purge Permit was authorized 2100 03/27/19 - All required steady state and source checks of 1RE-211/212, Containment Air Particulate and Containment Noble Gas Monitors were completed 2200 03/27/19 - Containment purge per OP 9C, Containment Venting and Purging was commenced 0400 03/28/19 - Maintenance required the containment purge to be interrupted for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> then restarted Which of the following actions (if any) are needed to authorize the RESTART of the containment purge per OP-9C, Containment Venting and Purging?

A. An air sample and new Containment Purge Permit must be completed B. No additional actions are necessary containment purge can be restarted C. Notify RP and start the Clean Up fan prior to restarting containment purge D. The steady state and source checks for 1RE-211/212 must be performed and within tolerance to previous data Tuesday, April 30, 2019 11:03:50 AM 298

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 2 Source: New Question History:

None SRO:

10CFR55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing RO knowledge of radiological safety principles (e.g., radiation work permit requirements, stay time, and DAC hours); AND requires the operator to recall the actions necessary during an interruption of a containment purge to authorize the restart of that purge.

K/A:

029A2.01 Containment Purge System (CPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Maintenance or other activity taking place inside containment (Imp 3.6)

Justification for K/A Match:

Matches the K/A by requiring the operator to recall the radiation monitoring system requirement imposed on the containment purge after an interruption caused by maintenance to authorize the restart of that purge.

Cognitive Level:

Knowledge 1-F: Requires the operator to recall the radiation monitoring system requirements.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 299

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

OP-9C, Containment Venting and Purging Unit 1 Rev 16 Proposed reference to be provided to the applicants during examination:

None Original Question:

None Justification:

An air sample has to be performed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to the start of a containment purge, given that the original purge started within those 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, but due the interruption due to maintenance, which is procedurally allowed for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without need for an air sample, a new sample needs to be taken. If a new air sample is required, then a new permit is also needed.

A CORRECT: See above.

B INCORRECT: Plausible if the operator has a misconception of the allowable time permit is valid or confuses it with the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> requirement for air sampling prior to starting a containment purge or vent.

C INCORRECT: Plausible because of the procedural note to contact RP about starting the cleanup fan prior to the required source checks of RE-211/212.

D CORRECT: Plausible because this is a necessary action if the purge is interrupted more than 20 minutes but less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Learning Objective:

DESCRIBE the procedures which govern operation of the Containment Ventilation System. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non Licensed Operators.

(051.05.LP0057.005)

Tuesday, April 30, 2019 11:03:50 AM 300

QUESTIONS REPORT for 2019 NRC Exam Master

92. 2019 NRC 092/SYS/033G2.4.41/4.6/3-SPR/SRO/BANK/EPIP 1.2.1/SD86.4.2.4.41 Given the following:

Unit 1 is in MODE 2 performing physics testing following a refueling outage Unit 2 is at Rated Thermal Power The Security Shift Supervisor reports that a HOSTILE ACTION is occurring in the PROTECTED AREA The perpetrator is contained at the Spent Fuel area of the PAB Annunciator C01 C 4-10, SPENT FUEL POOL TEMP HIGH LEVEL HIGH OR LO is received Spent fuel pool temperature is 115°F RISING Spent fuel pool level is 43 LOWERING RE-105, SFP Area Low Range Radiation Monitor is 3.8 R/hr RISING RAPIDLY The highest required emergency classification is . . .

REFERENCE PROVIDED EAL Charts (3 pages)

RU2 (2 pages)

RA2 (2 pages)

RS2 (1 page)

HU1 (2 pages)

HA1 (2 pages)

HS1 (2 pages)

HG1 (2 pages)

A. UNUSUAL EVENT B. ALERT C. SITE AREA EMERGENCY D. GENERAL EMERGENCY Tuesday, April 30, 2019 11:03:50 AM 301

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 2 Source: Bank Question History:

2014 DC Cook Question 92 K/A:

033G2.4.41 Spent Fuel Pool Cooling System (SFPCS)

Knowledge of the emergency action level thresholds and classifications.

(Imp 4.6)

Justification for K/A Match:

Matches the K/A by requiring the operator the EAL classification based on an event which affects the Spent Fuel Pool Cooling SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant parameters and determine selection of a notification based on that diagnosis. The operator must have knowledge of the EAL chart and bases.

Cognitive Level:

Comprehension 3-SPR: The operator must assess plant parameters and then determine the selection of a notification based on that assessment and the use of the EAL charts and pertinent bases documents.

10 CFR Part 55 Content:

55.41 10 55.43 5

Reference:

Draft EAL Wallboards Draft Emergency Action Levels Bases Document Proposed reference to be provided to the applicants during examination:

EAL ALL MODES Chart Rev Draft 2 pages Hot and ALL MODE (proveded for an earlier question)

Bases Documents for the following EAL levels:

RU2 (2 pages)

RA2 (2 pages)

Tuesday, April 30, 2019 11:03:50 AM 302

QUESTIONS REPORT for 2019 NRC Exam Master HU1 (2 pages)

HA1 (2 pages)

HS1 (2 pages)

HG1 (2 pages)

Original Question:

Given:

Unit 1 is in MODE 1 performing physics testing following a refueling outage Security has informed the control room that the facility has been breached and the perpetrator is contained in the Fuel Handling building.

The following alarms occur:

Annunciator #134 Response: Spent Fuel Pit, Drop 2, SFP Water Level Low Annunciator #134 Response: Spent Fuel Pit, Drop 3, SFP Water Temperature High Annunciator #134 Response: Spent Fuel Pit, Drop 6, North SFP Pump Failure Annunciator #105 Response: Containment Spray, Drop 28, Spent Fuel Pit Temp High Annunciator #105 Response: Containment Spray, Drop 26, Spent Fuel Pit Subpanel Alarm Spent fuel pit temperature is rising Spent fuel pit level is 623 and lowering R-5, Spent Fuel Pit Area radiation monitor is reading 50 mR and rising rapidly All other plant systems are indicating normal Using PMP-2080-EPP-101, Initiating Conditions Table, what is the correct classification of the event?

a. General Emergency

b. Site Area Emergency

c. Alert

d. Unusual Event Proposed answer A Tuesday, April 30, 2019 11:03:50 AM 303

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Based on the deteriorating condition in the Spent Fuel Pool, and that a hostile action is happening in the protected area, the correct classification would a GENERAL EMERGENCY base on HG1.1.

A INCORRECT: Plausible if the operator classifies based on an unplanned rise in the RE-105 reading, or an unplanned water level drop in the refueling pathway.

B INCORRECT: Plausible if the operator classifies based on the rapidly rising radiation reading on RE-105 or the Spent Fuel Pool level being below 49.

C INCORRECT: Plausible if the operator classifies based on a hostile action in the protected area or the lowering Spent Fuel Pool level, which will be going less than the required level based on trend.

D CORRECT: See above.

Learning Objective:

Knowledge of the emergency action level thresholds and classifications.

(SD86.4.2.4.41)

Tuesday, April 30, 2019 11:03:50 AM 304

QUESTIONS REPORT for 2019 NRC Exam Master

93. 2019 NRC 093/SYS/056A2.05/2.5*/3-SPK/SRO/BANK/NP 3.2.3/055.03.LP2439.005 Given the following:

Unit 1 is at Rated Thermal Power Annunciator 1C20 A 1-9, SECONDARY SYSTEM SAMPLE PANEL UNIT 1 is lit OI-38A, SF-6 Condenser Tube Leak Test has identified tube leaks in WB #4 in the 'A' condenser Chemistry has validated Steam Generator Blowdown Sodium concentration is 1500 ppb Which of the following is performed to mitigate the situation?

A. Trip the reactor and enter EOP-0, Reactor Trip or Safety Injection Maximize SG Blowdown flow B. Trip the reactor and enter EOP-0, Reactor Trip or Safety Injection Secure and isolate 1P-25A, Condensate pump C. Shutdown the plant per AOP-17A, Rapid Power Reduction or OP-3A, Power Operation to Hot Standby Maximize SG Blowdown flow D. Shutdown the plant per AOP-17A, Rapid Power Reduction or OP-3A, Power Operation to Hot Standby Secure and isolate 1P-25A, Condensate pump Tuesday, April 30, 2019 11:03:50 AM 305

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 2 Group 2 Source: Bank Question History:

None K/A:

056A2.05 Condensate System Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Condenser tube leakage (Imp 2.5*)

Justification for K/A Match:

Matches the K/A by requiring the operator determine the required actions to mitigate the effects of condenser tube leakage.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditons and determine actions per facility operating limitation and determine actions necessary to direct to mitigate the situation.

Cognitive Level:

Comprehension 3-SPK: The operator must assess plant parameters and then determine the course of actions needed to mitigate the situation.

10 CFR Part 55 Content:

55.41 5 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 306

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

NP 3.2.3, Secondary Water Chemistry Monitor Program Rev 29 ARB 1C20 A 1-9, Secondary System Sample Panel Unit 1 Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following conditions:

Unit 1 is at 100% power.

Annunciator 1-ALB-7B, Window 3.5 - SEC SMPL PNL TRBL, is in alarm.

ABN-304, Main Condenser and Circulating Water System Malfunction, Section 4.0, Main or Auxiliary Condenser Tube Leak, is in progress.

Chemistry has verified Steam Generator Blowdown Sodium concentration at 1500 ppb.

Which of the following is performed to mitigate the situation?

A. Trip the Reactor and enter EOP-0.0A, Reactor Trip or Safety Injection.

Remove all Condensate Polishers from service.

B. Trip the Reactor and enter EOP-0.0A, Reactor Trip or Safety Injection.

Maximize Steam Generator Blowdown flow.

C. Initiate a power reduction per IPO-003A, Power Operation to be less than or equal to 50% power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Maximize Steam Generator Blowdown flow.

D. Initiate a power reduction per IPO-003A, Power Operation to be less than or equal to 50% power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Remove all Condensate Polishers from service.

Proposed Answer: B Tuesday, April 30, 2019 11:03:50 AM 307

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Based on the results of steam generator sodium being above the Action level 3 condition limits, a shutdown to condition as quickly as safe plant operation permits and clean up by feed and bleed or drain and refill as appropriate A INCORRECT: The first part is incorrect, plausible as actions level 3 directs to shut down as quickly as safe plant operation permits. The second part is correct.

B INCORRECT: The first part is incorrect, plausible as actions level 3 directs to shut down as quickly as safe plant operation permits. The second part is incorrect, plausible if the operator has the misconception that this will isolate the 'A' condenser from the 'B' condensate pump and condensate system.

C CORRECT: See above.

D INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator has the misconception that this will isolate the 'A' condenser from the 'B' condensate pump and condensate system.

Learning Objective:

Given access to the Site Specific Simulator, APPLY and RESPOND using appropriate guidance provided in the applicable AOPs for various system/component malfunctions.

(055.03.LP2439.005)

Tuesday, April 30, 2019 11:03:50 AM 308

QUESTIONS REPORT for 2019 NRC Exam Master

94. 2019 NRC 094/GEN/G2.1.5/3.9/1-P/SRO/BANK/OM 3.1/SD86.4.2.1.5 Given the following:

Both Units are at Rated Thermal Power The time is 0015 AO Staffing is as follows:

Lead AO - Fully Qualified AO (FBL)

Unit 1 Turbine Hall (TH) - Fully Qualified AO (EP1)

Unit 2 Turbine Hall (TH) - Fully Qualified AO (EP2)

Primary Auxiliary Building (PAB) - Fully Qualified AO (FB1)

AO Trainee - Qualified Fire Brigade, EP Dose Assessor, and EP Responder and currently standing Unit 1 TH under instruction AO - Fully Qualified (FB2)

Fire Tech - Qualified Fire Brigade (FB3)

Fire Tech - Qualified Fire Brigade (FB4)

The Unit 1 TH Operator must leave immediately for a family emergency.

Which of the following describes the actions required to be taken?

A. The Unit 2 TH operator will assume the Unit 1 TH responsibilities; AO Trainee will assume EP Responder duties. Staffing may be maintained in this configuration up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. Notify duty and call personnel while attempting to replace Unit 1 TH watch. STA must remain in the Unit 1 TH watchstation until suitable replacement has reported.

C. Third or Fourth License may be utilized to cover the Unit 1 TH watchstation for the remainder of the shift; AO Trainee will assume EP Responder duties.

D. Immediately begin callout to replace the Unit 1 TH operator, to ensure that staffing can be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Tuesday, April 30, 2019 11:03:50 AM 309

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Bank Question History:

2005 PBNP Question 95 K/A:

G2.1.5 Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

(Imp 3.9)

Justification for K/A Match:

Matches the K/A by requiring the operator recall the minimum crew staffing and actions needed to be taken.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to have knowledge of the TS organization requirements terminology.

Cognitive Level:

Knowledge 1-P: The operator must recall the minimum staffing requirements and actions to necessary to meet it.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 310

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

TS 5.2, Administrative Controls - Organization Rev 3 OM 3.1, Operations Shift Staffing Requirements Rev 18 Proposed reference to be provided to the applicants during examination:

None Original Question:

Both Units are at 100% reactor power. The crew is working regular 12-hour shifts. The time is 0015.

AO Staffing is as follows:

- Unit 1 Turbine Hall (TH) - Fully Qualified AO.

- Unit 2 Turbine Hall (TH) - Fully Qualified AO.

- PAB - Fully Qualified AO.

- Water Treatment (WT) - AO Qualified WT and Fire Brigade only.

- AO Trainee - Fire Brigade qualified only, standing WT Under Instruction.

The Unit 1 TH Operator must leave immediately for a family emergency.

Which of the following correctly describes the actions that must be taken?

A. The Unit 2 TH operator will assume the Unit 1 TH responsibilities, AO Trainee will assume fire brigade duties. Staffing may be maintained in this configuration up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. Immediately begin callout to replace the Unit 1 TH operator, the operator must be replaced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. Third or Fourth License may be utilized to cover the Unit 1 TH watchstation for the remainder of the shift, with AO Trainee assuming Fire Brigade Duties.

D. Notify duty and call personnel while attempting to replace Unit 1 TH watch.

STA must remain in the Unit 1 TH watchstation until suitable replacement has reported.

Proposed answer: B Tuesday, April 30, 2019 11:03:50 AM 311

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

Per Tech Spec section 5.2.2.b states: Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew OM 3.1 section 5.2.7., 5.2.8, and 5.2.9 state the requirements for watches. With the Unit 1 TH AO leaving, there is a shortage of 1 AO, therefore, section 5.4.1 must be performed, which states "Call out required replacement personnel shall be implemented immediately... to ensure that proper shift staffing can be restored within two hours."

A INCORRECT: Both parts are wrong, the Unit 2 TH operator is not allowed to assume the Unit 1 TH operator duties, the AO trainee can assume the EP Responder duties, but it does not make up for the loss of the Unit 1 TH operator. Plausible as the Unit 2 TH operator is fully qualified, and the AO Trainee can assume the EP Responder.

B INCORRECT: The STA is not required to remain in the Unit 1 TH operator watchstation. Plausible as the duty and call personnel will be notified, and the STA if a licensed SRO, is by virtue of position qualified the duties of the auxiliary operators.

C INCORRECT: The third or fourth license is required to be manned in the control room. Plausible as the third or fourth license are also by virtue of position qualified the duties of the auxiliary operators and the AO trainee can assume the EP Responder duties.

D CORRECT: See above.

Learning Objective:

Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

(SD86.4.2.1.5)

Tuesday, April 30, 2019 11:03:50 AM 312

QUESTIONS REPORT for 2019 NRC Exam Master

95. 2019 NRC 095/GEN/G2.1.45/4.3/3-SPK/SRO/MODIFIED/BG-EOP-0/SD86.4.2.1.45 Given the following:

A LOCA has occurred The containment is adverse SI alignment has been verified Current plant instrumentation readings are as follows:

Train A RCS Subcooling indicates 41°F Train B RCS Subcooling indicates 0°F Pressurizer pressure indicates 1700 psig RCS Wide Range pressure indicates 1550 psig Hottest Channel Core Exit thermocouple indicates 566°F Which of the following answers (1) What is actual RCS subcooling?

AND (2) What action should be taken based on these indications?

A. (1) 35°F (2) All Reactor Coolant pumps are secured following a Small Break LOCA to prevent damage to the RCPs to maintain availability.

B. (1) 35°F (2) All Reactor Coolant pumps are secured following a Small Break LOCA to prevent loss of excessive inventory if the RCPs trip later in the event.

C. (1) 48°F (2) All Reactor Coolant pumps should remain running if possible to provide normal Pressurizer Spray flow and forced RCS flow.

D. (1) 48°F (2) All Reactor Coolant pumps should remain running if possible because it is desirable to minimize operator actions such as tripping the RCPs then restarting them later.

Tuesday, April 30, 2019 11:03:50 AM 313

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Modified Question History:

2009 Seabrook Question 95 K/A:

G2.1.45 Conduct of Operations Ability to identify and interpret diverse indications to validate the response of another indication.

(Imp 4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator determine which indications are necessary to diagnose the situation, use those indication to determine subcooling.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant parameters and determine the correct action and basis for that action.

Cognitive Level:

Comprehension 3-SPK: The operator must determine which indications are necessary to determine subcooling, and then calculate subcooling, determine the impact of subcooling on the RCPs, what actions should be taken and the basis for that action.

10 CFR Part 55 Content:

55.41 7 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 314

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

EOP-0, Reactor Trip or Safety Injection Unit 1 Rev 66 BG-EOP-0, Background Reactor Trip or Safety Injection Rev 44 Proposed reference to be provided to the applicants during examination:

None Original Question:

The following conditions exist:

A LOCA has occurred.

SI alignment has been verified.

Current plant instrumentation readings are as follows:

Train "A" RCS Subcooling indicates 35 degs F.

Train "B" RCS Subcooling indicates 0 degs F.

Pressurizer pressure indicates 1600 psig.

RCS Wide Range pressure indicates 1470 psig.

Hot Channel Core Exit thermocouple indicates 560 degs F.

What is actual RCS subcooling and what action should be taken based on these indications?

A. Actual RCS subcooling is 34 degs F. For Small Break LOCA conditions the primary concern is protection of the RCP for later use B. Actual RCS subcooling is 34 degs F. All Reactor Coolant pumps are secured following a small break LOCA to prevent loss of excessive inventory when the RCSs are tripped.

C. Actual RCS subcooling is 46 degs F. All Reactor Coolant pumps should remain running if possible to provide normal Pressurizer Spray flow and forced RCS flow.

D. Actual RCS subcooling is 46 degs F. All Reactor Coolant pumps should remain running if possible because it is desirable to minimize operator actions such as tripping the RCPs then restarting them later.

Proposed Answer: B Tuesday, April 30, 2019 11:03:50 AM 315

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

RCS Subcooling is 35 based on the 566°F CET temperature and wide rage RCS pressure of 1550 psig, thus the subcooling monitor board indication is not functioning properly. The background document states that RCPs should be secured following a small break LOCA to prevent the excessive loss of inventory and core uncovery when the RCPs trip later in the event.

A INCORRECT: The first part is correct. The second part is incorrect, but plausible as this action will protect the RCP, but that is not the primary concern of the performance of securing the RCPs during a small break LOCA, which is what the RCS pressure is indicating.

B CORRECT: See above.

C INCORRECT: The first part is incorrect 48°F is based on the pressurizer pressure indication which has a low peg at 1700 psig, this is plausible if the operator determines the pressurizer indication is more accurate than the wide range. The second part is correct, the WOG generic issue volume states that RCPs should not be tripped early so they provide normal pressurizer spray flow and forced RCS flow.

D INCORRECT: The first part is incorrect 48°F is based on the pressurizer pressure indication which has a low peg at 1700 psig, this is plausible if the operator determines the pressurizer indication is more accurate than the wide range. The second part is correct, the WOG generic issue volume states that RCPs should remain running if possible because it is desirable to minimize operator action such as tripping the RCPs then restarting them later Learning Objective:

Ability to identify and interpret diverse indications to validate the response of another indication.

(SD86.4.2.1.45)

Tuesday, April 30, 2019 11:03:50 AM 316

QUESTIONS REPORT for 2019 NRC Exam Master

96. 2019 NRC 096/GEN/2.2.22/4.7/3-SPK/SRO/BANK/TS 3.6.6/SD86.4.2.2.22 Given the following:

Unit 1 is at 1% power, with preparations being made for entry into MODE 1 The following sequence of events occurs:

Maintenance reports that they have added the wrong type of oil to 1P-14A, Containment Spray pump.

The crew enters TSAC 3.6.6.A, Containment Spray and Cooling Systems, One containment spray train inoperable, Restore containment spray train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the applicable TSAC for 3.6.7 Spray Additive System The STA reviews the surveillance history for Containment Spray, and discovers the monthly surveillance valve lineup, 1-TS-ECCS-001, Safeguard Systems Valve and Lock Checklist (Monthly) Unit 1, scheduled for 20 days ago, was incomplete for ONLY B Train Containment Spray valves The last time the B Train Containment Spray lineup was completed was 50 days ago Work Control estimates the B Train valve lineup will be completed in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Per Technical Specifications 3.6.6 Containment Spray and Cooling Systems ONLY, what action is the crew required/allowed to take?

A. Within one hour, take action to place the unit in MODE 3, within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, per LCO 3.0.3, since failure to perform the surveillance within the specified interval shall be failure to meet the LCO, per Surveillance Requirement SR 3.0.1.

B. Take action to place the unit in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, per LCO 3.0.3.

The one hour allowance of LCO 3.0.3 cannot be utilized, since the estimated time to complete the valve lineup in excess of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Remain in the current ACTION CONDITION for A Train, since B Train surveillance time has not exceeded its maximum allowable extension per Surveillance Requirement SR 3.0.2.

D. Remain in the current ACTION CONDITION for A Train, since the crew has 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete the B Train surveillance per Surveillance Requirement SR 3.0.3.

Tuesday, April 30, 2019 11:03:50 AM 317

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Bank Question History:

2009 Millstone Unit 3 Question 96 K/A:

G2.2.22 Equipment Control Knowledge of limiting conditions for operations and safety limits.

(Imp 4.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to use knowledge of the generic LCO requirements and apply it to limiting conditions for operations.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to have knowledge of application of generic LCO requirements (LCO 3.0.1 through LCO 3.0.7 and SR 3.0.1 through SR 3.0.4)

Cognitive Level:

Comprehension 3-SPK: The operator must determine how the generic LCO conditions apply to the initial condition and what the outcome of that application should be.

10 CFR Part 55 Content:

55.41 5 55.43 2

Reference:

TS 3.6.6, Containment Systems - Containment Spray and Cooling System, Rev 4 TS 3.0, LCO Applicability Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

Initial Conditions:

The plant is at 1% power, with preparations being made for entry into MODE 1, when the following sequence of events occurs:

Tuesday, April 30, 2019 11:03:50 AM 318

QUESTIONS REPORT for 2019 NRC Exam Master

1. Maintenance reports that they have added the wrong type of oil to the "A" QSS Pump.

2. The crew enters LCO 3.6.2.1 "Containment Quench Spray System,"

ACTION with one QSS subsystem INOPERABLE, to restore the pump to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3. The STA looks at the surveillance history of the "B" Train of QSS, and discovers its monthly surveillance valve lineup, scheduled for 20 days ago, was inadvertently missed.

4. The last time the "B" Train QSS lineup was completed was 50 days go.

5. Work Control estimates the "B" Train valve lineup will be completed in1 hour.

In accordance with section 3/4.0 of Technical Specifications, what ACTION is the crew required/allowed to take?

A. Within one hour, take action to place the unit in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, per LCO 3.0.3, since failure to perform a surveillance within the specified interval shall be failure to meet the LCO, per surveillance requirement 4.0.1.

B. Take action to place the unit in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, per LCO 3.0.3. The one hour allowance of LCO 3.0.3 cannot be utilized, since the estimated time to complete the valve lineup in excess of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Remain in the current ACTION for Train "A", since the "B" train surveillance time has not exceeded its maximum allowable extension per surveillance requirement 4.0.2.

D. Remain in the current ACTION for Train "A", since the crew has 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to complete the "B" train surveillance per surveillance requirement 4.0.3.

Proposed Answer: D Tuesday, April 30, 2019 11:03:50 AM 319

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

SR 3.0.3 states If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. So a delay of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed, prior to declaring the B Train of Containment Spray inoperable.

A INCORRECT: Plausible since SR 3.0.1 states: Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillances, shall be failure to meet the LCO. Not meeting this LCO would cause both trains of Containment Spray to be inoperable, and invoke 3.0.3 B INCORRECT: Plausible since SR 3.0.1 states: Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillances, shall be failure to meet the LCO. Not meeting this LCO would cause both trains of Containment Spray to be inoperable, and invoke 3.0.3, taking into consideration that the valve lineup would be completed during the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to place the unit of 3.0.3.

C INCORRECT: Plausible if the operator incorrectly applies SR 3.0.2 which allow an extension of 1.25 times the interval specified in the Frequency. A proper extension would be 38.75 days (1.25 v 31 days).

D CORRECT: See above.

Learning Objective:

Knowledge of limiting conditions for operations and safety limits.

(SD86.4.2.2.22)

Tuesday, April 30, 2019 11:03:50 AM 320

QUESTIONS REPORT for 2019 NRC Exam Master

97. 2019 NRC 097/GEN/G2.3.13/3.8/3-SPK/SRO/BANK/RP 1C/112.01.LP0285.002 Given the following:

Unit 2 Refueling operations are in progress The Containment Equipment hatch is removed Irradiated fuel is being moved in the manipulator from the core to the containment upender for transfer to the Spent Fuel Pool A leak is reported from the refueling cavity drain line cleanout flange Refueling Cavity water level is LOWERING SLOWLY Identify the direction the Core Load Supervisor should provide during this event.

A. Store the irradiated fuel assembly in any available core location, THEN shut the Transfer Tube Gate Valve.

B. Store the irradiated fuel assembly in the Spent Fuel Pool upender Frame Down, THEN shut the Spent Fuel Pool Transfer Canal Doors.

C. Store the irradiated fuel assembly in any available core location, THEN have the Control Room secure Unit 2 Purge Supply and Exhaust Fans.

D. Store the irradiated fuel assembly in the containment upender Frame Down, THEN shut the Transfer Tube Gate Valve AND the Spent Fuel Pool Transfer Canal Doors.

Tuesday, April 30, 2019 11:03:50 AM 321

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Bank Question History:

2011 PBNP Question 98 K/A:

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

(Imp 3.7)

Justification for K/A Match:

Matches the K/A by requiring the operator to use give take action during a fuel handling event to minimize the impact due to radiation, and refueling water level loss.

SRO:

10CFR55.43(b)(7) Fuel handling facilities and procedures.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it requires refuel floor direction from the Core Load Supervisor, which is an SRO position, for actions taken and safe storage locations for fuel during a refueling cavity water level loss event.

Cognitive Level:

Comprehension 3-SPK: The operator must understand the initial conditions, and utilizing this information, determine the correct location to place the fuel assembly, and then was further actions are needed.

10 CFR Part 55 Content:

55.41 12 55.43 4 Tuesday, April 30, 2019 11:03:50 AM 322

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

RP 1C, Refueling, Rev 82, Section 3.3.5 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following on Unit 2:

Refueling operations are in progress It has been 7 days since shutdown The Containment Equipment hatch is removed Irradiated fuel is being moved in the manipulator from the core to the containment upender for transfer to the Spent Fuel Pool A leak is reported from the refueling cavity drain line cleanout flange Refueling Cavity water level is slowly lowering Identify the direction the Core Load Supervisor should provide during this event.

A. Store the irradiated fuel assembly in the containment upender Frame Down, THEN shut the Transfer Tube Gate Valve AND the Spent Fuel Pool Transfer Canal Doors.

B. Store the irradiated fuel assembly in the Spent Fuel Pool upender Frame Down, THEN shut the Spent Fuel Pool Transfer Canal Doors.

C. Store the irradiated fuel assembly in any available core location, THEN have the Control Room secure Unit 2 Purge Supply and Exhaust Fans.

D. Store the irradiated fuel assembly in any available core location, THEN shut the Transfer Tube Gate Valve.

Proposed Answer: D Tuesday, April 30, 2019 11:03:50 AM 323

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

With a refueling water leak at an elevation below the reactor refueling seal ledge (which would be the refueling cavity drain line cleanout flange), the correct locations for the placement of an irradiated fuel assembly is the reactor or spent fuel racks. The transfer tube valve should be left open until this is completed, this is done to minimize the amount of level lost in the refueling cavity.

A CORRECT: See above.

B INCORRECT: Wrong action. Plausible, as the storage location is correct, but without shutting the transfer tube valve, the level in the spent fuel pool will also be lost.

C INCORRECT: Wrong action. Plausible as this is a correct storage location, but the action is incorrect based on the initial conditions, with the equipment hatch open, purge supply and exhaust balances air flow.

D INCORRECT: Wrong location, and shutting the spent fuel pool transfer canal doors is not necessary. Plausible as this is one of the safe storage locations utilized if the leak is at or above the refueling seal ledge level.

Learning Objective:

DESCRIBE the procedures which govern Fuel Handling Operations. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non Licensed Operators.

(112.01.LP0285.002)

Tuesday, April 30, 2019 11:03:50 AM 324

QUESTIONS REPORT for 2019 NRC Exam Master

98. 2019 NRC 098/GEN/G2.3.15/3.1/2-DR/SRO/BANK/TS 3.37.9/SD86.3 2.3.15 Given the following:

Unit 1 is in MODE 5 Unit 2 is preparing to begin core reload RE-101, Control Room Monitor has been removed from service for maintenance Control Room Emergency Filtration System (CREFS) is in Mode 1 What is the status of LCO 3.7.9 Control Room Emergency Filtration System and impact on fuel handling?

A. CREFS is OPERABLE. Fuel handling may begin.

B. CREFS is INOPERABLE. Fuel handling may not begin without the performance of a risk assessment.

C. CREFS is INOPERABLE. Fuel handling may begin on Unit 2 because CREFS is not required in MODE 5 or 6.

D. CREFS is INOPERABLE. Fuel handling may begin as the REQUIRED ACTIONS do not suspend movement of irradiated fuel assemblies.

Tuesday, April 30, 2019 11:03:50 AM 325

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Bank Question History:

None K/A:

G2.3.15 Radiation Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(Imp 3.1)

Justification for K/A Match:

Matches the K/A by requiring the operator to use knowledge of fixed radiation monitors and their impact on both Tech Specs and refueling.

SRO:

10CFR55.43(b)(2) Facility operating limitations in the technical specifications and their bases.

10CFR55.43(b)(7) Fuel handling facilities and procedures.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing < 1-hour TS/TRM Actions, the LCO/TRM information listed above the line, or TS safety limits; AND requires the operator to have knowledge of application of generic LCO requirements (LCO 3.0.1 through LCO 3.0.7 and SR 3.0.1 through SR 3.0.4) and refueling requirements.

Cognitive Level:

Comprehension 2-RI: The operator understand the initial conditions and the impact the rad monitor will have on the CREFS system, and refueling as well as changing plant modes.

10 CFR Part 55 Content:

55.41 12 55.43 4 Tuesday, April 30, 2019 11:03:50 AM 326

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

0-SOP-FH-001, Fuel-Insert-Component Movement In The Spent Fuel Pool Or New Fuel Vault Rev 32 TS 3.7.9, Plant Systems - Control Room Emergency Filtration System, Rev 5, LCO, Applicability, and Actions Sections TS B 3.7.9, Basis Plant Systems - Control Room emergency Filtration System, Rev 6 TS 3.3.5, Instrumentation - Control Room Emergency Filtration System (CREFS) Actuation Instrumentation, Rev 2 TS B 3.3.5, Basis Instrumentation - Control Room Emergency Filtration System (CREFS) Actuation Instrumentation, Rev 4 RP 1C, Refueling, Rev 82 TS 3.0, Limiting Condition for Operation (LCO) Applicability, Rev 3 Proposed reference to be provided to the applicants during examination:

None Original Question:

Given the following:

Unit 1 is in MODE 5 Unit 2 is preparing to begin core off load W-14B, Control Room Charcoal Filter Fan, is OOS for minor maintenance What is the status of LCO 3.7.9 Control Room Emergency Filtration System and impact on fuel handling?

A. CREFS is OPERABLE. Fuel handling may begin.

B. CREFS is INOPERABLE. Fuel handling may not begin without the performance of a risk assessment.

C. CREFS is INOPERABLE. Fuel handling may begin on Unit 2 because CREFS is not required in MODE 5.

D. CREFS is INOPERABLE. Fuel handling may begin as the REQUIRED ACTIONS do not suspend movement of irradiated fuel assemblies.

Proposed answer: B Tuesday, April 30, 2019 11:03:50 AM 327

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The CREF system is not operable, based on RE-101 being OOS causing TS 3.3.5 to be not operable. With Unit 2 preparing to move fuel, the CREFS and CREFS Actuation system are not applicable, but unit 2 cannot commence moving fuel, as this would put the unit in the mode applicability and TS 3.7.9 and 3.3.5. are not met. Also RP-1C checklist requires TS 3.3.5 and 3.7.9 to be met or compliance with TS LOC 3.0.4 is required prior to entering the MODE/Condition of applicability A INCORRECT: The first part is incorrect, plausible if the operator does not take into consideration TS 3.3.5. The second part is incorrect, plausible because fueling handling my begin, but after the performance of a risk assessment.

B CORRECT: See above.

C INCORRECT: The first part is correct. The second part is incorrect, MODE/condition of applicability is mode 1-4 AND when moving irradiated fuel. Plausible if the operator recalls in error the mode of applicability for CREFS and the requirements of fuel handling D INCORRECT: The first part is correct. The second part is incorrect, plausible if the operator mis-applied the TS 3.0 rules as the REQUIRED ACTIONS do not require suspension of irradiated fuel movement.

Learning Objective:

Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(SD86.3 2.3.15)

Tuesday, April 30, 2019 11:03:50 AM 328

QUESTIONS REPORT for 2019 NRC Exam Master

99. 2019 NRC 099/GEN/G2.4.9/4.2/3-SPK/SRO/NEW/EPIP 1.2.1/SD86.4.2.4.9 Given the following:

Unit 1 RCS is solid, making preparations to transition to MODE 4 per OP 1A, Cold Shutdown to Hot Standby Containment Closure has been established The crew has just begun to stabilizing RCS temperature in preparation for Safety Injection Pump testing RCS temperature is 195°F and VERY SLOWLY RISING RCS pressure is 310 psig and STABLE A RHR pump is in service Offsite power to Unit 1 is lost A 15 GPM RCS leak occurs coincident with the loss of offsite power G01, Emergency Diesel Generator fails to start, and CANNOT be started Current Conditions 12 minutes later The crew adjusts Charging flow to compensate for the RCS leak The crew has located the leak on a cracked fitting, and it will take 5 minutes to isolate RCS temperature is 205°F and RISING SLOWLY RCS pressure is 340 psig and RISING SLOWLY The CURRENT highest emergency classification is . . .

REFERENCE PROVIDED CU3 (2 pages)

CA3 (2 pages)

SU5 (2 pages)

A. UNUSUAL EVENT based on CU3.1 B. UNUSUAL EVENT based on SU5.1 C. ALERT based on CA3.1 D. ALERT based on CA3.2 Tuesday, April 30, 2019 11:03:50 AM 329

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: New Question History:

None K/A:

G2.4.9 Emergency Procedures/Plan Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(Imp 4.2)

Justification for K/A Match:

Matches the K/A by requiring the operator to use knowledge of low power accidents, determine the choice of the proper classification for the notification of government agencies so the proper actions may occur.

SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant parameters and determine selection of a notification based on that diagnosis. The operator must have knowledge of the EAL chart utilization.

Cognitive Level:

Comprehension 3-SPK: The operator must analyze the initial conditions, use the EAL charts and determine the classification and basis for that classification.

10 CFR Part 55 Content:

55.41 10 55.43 5 Tuesday, April 30, 2019 11:03:50 AM 330

QUESTIONS REPORT for 2019 NRC Exam Master

Reference:

Draft EAL Wallboards Draft Emergency Action Levels Bases Document Proposed reference to be provided to the applicants during examination:

EAL ALL MODES Chart Rev Draft 2 pages Cold and ALL MODE (ALL MODE proveded for an earlier question)

Bases Documents for the following EAL levels:

CU3 (2 pages)

CA3 (2 pages)

SU5 (2 pages)

Original Question:

None Justification:

UNUSUAL EVENT based an unplanned rise in RCS temperature greater than 200°F, due to initial condition of the crew stabilizing temperature at 195°F in preparation for entering MODE 4, not actually entering MODE 4.

A CORRECT: See above.

B INCORRECT: The level is correct but the basis is incorrect, plausible because if the plant was in MODE 4 intentionally (RCS >200°F) this would be a correct basis for classification.

C INCORRECT: The level and basis are incorrect but plausible if the operator does not take into consideration that RHR what running until the LOOP occurred. The crew had not fully stabilized the heat up, therefore, it has continued for more than the 12 minutes with the leak lasting 17 minutes.

D INCORRECT: The level and basis are incorrect but plausible as it would be the correct classification the plant was not in solid plant operations.

Learning Objective:

Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(SD86.4.2.4.9)

Tuesday, April 30, 2019 11:03:50 AM 331

QUESTIONS REPORT for 2019 NRC Exam Master 100. 2019 NRC 100/SG/2.4.20/4.3/3-PEO/SRO/BANK/ECA-3.1/031.02.LP0473.006 Given the following:

EOP-3, Steam Generator Tube Rupture is in progress The cooldown to establish subcooling has been completed RC-430, Pressurizer PORV was opened to minimize break flow and refill the Pressurizer When the criteria to close RC-430 was met, both RC-430 and RC-516, RC-430 PZR PORV Isolation valve would not close Following the required procedure transition, the OS eventually reached the following CAUTION:

Which of the following identifies:

(1) The procedure that the OS transitioned to, AND (2) The basis for the caution?

A. (1) ECA-3.1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired (2) Minimize the potential for thermal shock of the steam generator tubes B. (1) ECA-3.1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired (2) Prevent exacerbating the RCS cooldown by feeding a faulted steam generator C. (1) ECA-3.3, SGTR Without Pressurizer Pressure Control (2) Minimize the potential for thermal shock of the steam generator tubes D. (1) ECA-3.3, SGTR Without Pressurizer Pressure Control (2) Prevent exacerbating the RCS cooldown by feeding a faulted steam generator Tuesday, April 30, 2019 11:03:50 AM 332

QUESTIONS REPORT for 2019 NRC Exam Master SRO Tier 3 Source: Bank Question History:

None SRO:

10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Justification for SRO-ONLY Question:

This is an SRO-ONLY question based on it cannot be answered solely by knowing system knowledge, immediate operator actions, knowing entry condition for AOPs or plant parameters which direct entry into major EOPs, the purpose, overall sequence of events, or overall mitigative strategy of a procedure; AND requires the operator to assess plant conditions and use knowledge of mitigation strategy and requirements within procedures to determine the correct course of action and must recall the basis of procedure steps.

K/A:

G2.4.20 Emergency Procedures/Plan Knowledge of the operational implications of EOP warnings, cautions, and notes.

(Imp 4.3)

Justification for K/A Match:

Matches the K/A by requiring the operator to use knowledge of the notes in the EOP network to implement a mitigation strategy for the situation.

Cognitive Level:

Comprehension 3-PEO: The operator must analyze the initial conditions, use knowledge to determine what course of actions are needed to mitigate the situation given the initial conditions and recall the reason for an applicable note.

10 CFR Part 55 Content:

55.41 10 55.43 5

Reference:

EOP-3.0, Steam Generator Tube Rupture Rev 38 ECA-3.1 Unit 1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired Rev 42 BG-ECA-3.1, Background Steam Generator Tube Rupture Rev 33 Proposed reference to be provided to the applicants during examination:

None Tuesday, April 30, 2019 11:03:50 AM 333

QUESTIONS REPORT for 2019 NRC Exam Master Original Question:

Given the following:

EOP-3, Steam Generator Tube Rupture is in progress The cooldown to establish subcooling has been completed RC-430, Pressurizer PORV was opened to minimize break flow and refill the Pressurizer When the criteria to close RC-430 was met, both RC-430 and RC-516, RC-430 PZR PORV Isolation valve would not close Following the required procedure transition, the OS eventually reached the following CAUTION:

Which of the following identifies:

(1) The procedure that the OS transitioned to, AND (2) The basis for the caution?

A. (1) ECA-3.1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired (2) Minimize the potential for thermal shock of the S/G tubes B. (1) ECA-3.1, SGTR With Loss Of Reactor Coolant -Subcooled Recovery Desired (2) Prevent exacerbating the RCS cooldown by feeding a faulted steam generator C. (1) ECA-3.3, SGTR Without Pressurizer Pressure Control (2) Minimize the potential for thermal shock of the S/G tubes D. (1) ECA-3.3, SGTR Without Pressurizer Pressure Control (2) Prevent exacerbating the RCS cooldown by feeding a faulted steam generator Proposed answer: B Tuesday, April 30, 2019 11:03:50 AM 334

QUESTIONS REPORT for 2019 NRC Exam Master Justification:

The required transition is based on EOP-3, step 18 requires a transition to ECA-3.1 for a condition where the PORV was used to minimize break flow and refill the Pressurizer and RCS pressure continues to lower, which is the case in the initial conditions. The basis of the note is to communicate that feeding a ruptured faulted steam generator may aggravate an uncontrolled cooldown of the RCS and may rise the possibility of steam generator overfill.

A INCORRECT: The first part is correct. The second part is incorrect, plausible, if the operator has a misconception of the reason for not feeding a faulted steam generator, not a faulted/ruptured steam generator.

B CORRECT: See above C INCORRECT: The first part is incorrect, plausible if the operator confuses no pressure control (to lower pressure), with loss of pressure control, (to lower pressure when minimizing break flow).The second part is incorrect, plausible, if the operator has a misconception of the reason for not feeding a faulted steam generator, not a faulted/ruptured steam generator.

D INCORRECT: The first part is incorrect, plausible if the operator confuses no pressure control (to lower pressure), with loss of pressure control, (to lower pressure when minimizing break flow). The second part is correct.

Learning Objective:

Given access to the Site Specific Simulator and appropriate plant/system conditions, diagnose a loss of RCS or Pressurizer Pressure control.

(031.02.LP0473.006)

Tuesday, April 30, 2019 11:03:50 AM 335