ML16138A090

2016-05 Final Written Exam

ML16138A090
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Site:	South Texas
Issue date:	05/04/2016
From:	Vincent Gaddy Operations Branch IV
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Download: ML16138A090 (214)

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Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 96 Last used on an NRC exam: 2013 RO Sequence Number: 1 Which of the following describes the BACKUP power supply to Instrument Air Compressor #14 upon loss of power to LC 1U?

A. TSC diesel generator which energizes LC 1U B. TSC diesel generator which energizes MCC 1G5 C. BOP diesel generator which energizes LC 1U D. BOP diesel generator which energizes MCC 1G5 Answer: D BOP diesel generator which energizes MCC 1G5 Page 1 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 96 K/A Catalog Number: 078 K2.02 Tier: 2 Group/Category: 1 RO Importance: 3.3 10CFR

Reference:

55.41(b)(4)

Knowledge of the bus power supplies to the following:

Emergency air compressor.

STP Lesson: LOT 202.26 Objective Number: 25610 LIST the systems that interface with the Instrument Air and Service Air systems and STATE the function of each interface.

Reference:

LOT201.35 PowerPoint Rev 9, Slide 23 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - LC is credible because LC 1U normally energizes MCC 1G5 (which is the supply to IA #14).  TSC DG is credible because it is another non-class DG which (despite its name) supplies power to plant components such as the CVCS PDP, transformer cooling fans, select plant lighting, etc.

B: INCORRECT - TSC DG is credible because it is another non-class DG which (despite its name) supplies power to plant components such as the CVCS PDP, transformer cooling fans, select plant lighting, etc.

C: INCORRECT - Credible because the BOP DG supplies IA compressor #14, but it does so by energizing MCC 1G5 which is normally powered through LC 1U.

D: CORRECT - The BOP DG powers up MCC 1G5 (which supplies IA compressor 14) upon a loss of power to the MCC.

Question Level: F Question Difficulty 2 Justification:

The applicant must combine the knowledge of the instrument air system and the backup diesel generators to eliminate the incorrect distractors and select the correct answer.

Page 2 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 486 Last used on an NRC exam: 2007 RO Sequence Number: 2 The power supply for the Turbine Driven Auxiliary Feedwater Pump Steam Inlet Valve, MS-MOV-0143, is ESF 125 VDC Bus __________.

A. E1A11 B. E1B11 C. E1C11 D. E1D11 Answer: D E1D11 Page 3 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 486 K/A Catalog Number: 061 K2.01 Tier: 2 Group/Category: 1 RO Importance: 3.2 10CFR

Reference:

55.41(b)(8)

Knowledge of bus power supplies to the following:

AFW system MOVs STP Lesson: LOT 202.28 Objective Number: 92049 List the typical loads on the Class 1E 125 VDC System.

Reference:

LOT 202.28, lesson plan Rev 10 slide #96 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: All distractors are plausible because they are ESF 125 VDC Busses.

B: INCORRECT: All distractors are plausible because they are ESF 125 VDC Busses.

C: INCORRECT: All distractors are plausible because they are ESF 125 VDC Busses.

D: CORRECT: Power supply to AFW steam supply valve is Class 1E 125 VDC E1D11.

Question Level: F Question Difficulty 2 Justification:

Requires the fundamental knowledge of power supplies for AFW equipment.

Page 4 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1101 Last used on an NRC exam: 2009 RO Sequence Number: 3 In accordance with 0POP08-FH-0009, Core Refueling, which of the following is an administrative task for a Licensed Control Room Operator during refueling operations?

A. Inform the Core Load Supervisor of the next core location to have a fuel assembly loaded.

B. Operate the remote television monitoring equipment used to observe refueling activities.

C. Monitor the Core Monitoring NI channels during and following insertion of each fuel assembly.

D. Evaluate Inverse Count Rate Ratio (ICRR) data on loaded fuel assemblies.

Answer: C Monitor the Core Monitoring NI channels during and following insertion of each fuel assembly.

Page 5 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1101 K/A Catalog Number: G2.1.40 Tier: 3 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(10)

Knowledge of refueling administrative reqirements.

STP Lesson: LOT 201.43 Objective Number: 66407 DESCRIBE the procedural requirements of the fuel handling equipment operating procedure(s) to include purpose, scope, precautions and limitations

Reference:

0POP08-FH-0009, Rev 44, Core Refueling, Page 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because move sheets are available to and monitored by the RO during fuel movement, but not an assigned responsibility.

B: INCORRECT: Credible because the equipment is available to the operator, but not an assigned responsibility C: CORRECT: per 0POP08-FH-0009, Core Refueling D: INCORRECT: Credible because ICRR data is collected. Incorrect because per 0POP08-FH-0009, step 6.2.10.4, it is collected by Reactor Engineering. NOTE: Reactor Operator is required to evaluate/perform an ICRR per the Task Analysis of LOT Training during their reactor startup certification..

Question Level: F Question Difficulty 2 Justification:

The student must have knowledge of the administrative responsibilities for Licensed Operators during refueling.

Page 6 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2508 Last used on an NRC exam: Never RO Sequence Number: 4 Turbine Impulse Pressure is giving a Temperature Reference (Tref) of 590°F to the Rod Control System.

What would this reading equate to in percent Reactor Power?

A. 98%

B. 92%

C. 88%

D. 82%

Answer: B 92%

Page 7 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2508 K/A Catalog Number: 045 K4.01 Tier: 2 Group/Category: 2 RO Importance: 2.7 10CFR

Reference:

55.41(b)(7)

Knowledge of MT/G system design feature(s) and/or interlock(s) which provide for the following:

Programmed controller for relationship between steam pressure at T/G inlet (impulse, first stage) and plant power level.

STP Lesson: LOT 201.18 Objective Number: 86061 DESCRIBE the instrumentation and controls available to monitor and operate the Rod Control System.

Reference:

LOT 201.18 Lesson Plan on Rod Control.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.

B: CORRECT: Tref spans from 567 to 592 degrees F with a corresponding percent power of 0% to 100%. This calculates out to 1/4 Degree F/% power. A Tref of 590 degrees F would then calculate to 92% power.

C: INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.

D: INCORRECT: All distractors are plausible because the student has to know the Tref vs percent power graph to calculate what percent power would be.

Question Level: H Question Difficulty 2 Justification:

The student must have knowledge of the percent power vs Tref and be able to calculate the two for different power levels. NOTE: Turbine Impulse pressure can be read in an actual pressure, however, an ICS computer point calculates the pressure to a Temperature Reference. For operational validity the question uses Tref.

Page 8 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2489 Last used on an NRC exam: Never RO Sequence Number: 5 The Unit is cooling down after a SGTR per 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill, with the following conditions:

RHR Pump A and C are in service.

RCS temperature is 310ºF and slowly lowering.

RCS pressure is being maintained between 325 and 400 psig.

Subsequently:

ESF 4.16KV C loses power due to an overcurrent lockout on the BUS.

Which of the following would be the correct action for the crew to perform?

A. Suspend 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and perform 0POP09-AN-01M2, C-8, RHR PUMP C TRIP, ONLY.

B. Suspend 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and perform 0POP09-AN-01M2, C-8, RHR PUMP C TRIP, AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus.

C. Continue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, RHR PUMP C TRIP, ONLY.

D. Continue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, RHR PUMP C TRIP, AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus.

Answer: D Continue with 0POP05-EO-ES31, POST-SGTR Cooldown Using Backfill and, if manpower is available, perform 0POP09-AN-01M2, C-8, RHR PUMP C TRIP, AND 0POP04-AE-0001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus.

Page 9 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2489 K/A Catalog Number: APE 025 G2.4.8 Tier: 1 Group/Category: 1 RO Importance: 3.8 10CFR

Reference:

55.41(b)(10)

Loss of RHR System:

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

STP Lesson: LOT 504.04 Objective Number: 92283 GIVEN a set of conditions and the occurrence of a Red, Orange, or Yellow path CSF, STATE the action required per 0POP01-ZA-0018, EOP Users Guide.

Reference:

0POP01-ZA-0018, Rev 21, Emergency Operating Procedure User's Guide (step 4.26.4)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the student has to remember that Emergency Operating Procedures have precedence over Off-Normal/Annunciator Response Procedures. Also, the student would need to realize that 0POP04-AE-0001 is equally important to 0POP09-AN-1M02 because without addressing the issue ESF DG on Train C would be running without cooling water.

B: INCORRECT: Plausible because the student has to remember that Emergency Operating Procedures have precedence over Off-Normal/Annunciator Response Procedures.

C: INCORRECT: Plausible because the student would need to realize that 0POP04-AE-0001 is equally important because without addressing the issue ESF DG on Train C would be running without cooling water.

D: CORRECT: Per 0POP01-ZA-0018, Emergency Operating Procedure User's Guide, Off-Normal and Annunciator Response procedures can be performed if they do not conflict with the EOP and adequate resources are available. In this case if resources available then both 0POP09-AN-1M02 and 0POP04-AE-0001 should be performed to restart RHR pump B and secure ESF DG on Train C.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given conditions and have knowledge of the rules of usage for off-normal and emergency procedures to determine the correct answer.

Page 10 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2493 Last used on an NRC exam: Never RO Sequence Number: 6 A Reactor Trip and Safety Injection have occurred with the following conditions:

SI has been reset.

Due to some confusion with the event, the Unit Supervisor has entered 0POP05-EO-ES00, Rediagnosis.

Subsequently a Reactor Operator observes the following:

RCS Pressure: ..............................1710 psig and slowly lowering RCS Subcooling:..........................60oF and slowly lowering Pressurizer Level:.........................20% and slowly lowering All SG Pressures: .........................1175 psig and slowly lowering SG A NR Level: ...........................8% and slowly rising SG B NR Level: ...........................10% and slowly rising SG C NR Level: ...........................17% and slowly rising SG D NR Level: ...........................19% and slowly rising Total AFW Flow: .........................400 gpm Containment Pressure:..3 psig and slowly rising Containment Radiation Level:......4 R/HR and stable Which procedure should the crew transition to?

A. 0POP05-EO-FRZ1, Response to High Containment Pressure B. 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink C. 0POP05-EO-EO20, Faulted Steam Generator Isolation D. 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Answer: D 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Page 11 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2493 K/A Catalog Number: W/E01 EA2.1 Tier: 1 Group/Category: 2 RO Importance: 3.2 10CFR

Reference:

55.41(b)(10)

Ability to determine and interpret the following as they apply to the (Reactor Trip and Safety Injection Rediagnosis):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

STP Lesson: LOT 504.09 Objective Number: 81187 DISCUSS the indications available to determine plant status during a loss of primary or secondary coolant accident.

Reference:

0POP05-EO-EO10 Rev 22, Loss of Reactor or Secondary Coolant and 0POP05-EO-ES00, Rediagnosis Rev 8 page 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Modified Modified from 52 Distractor Justification A: INCORRECT: Credible because this transition may be done with a higher containment pressure (9.5 psig).

B: INCORRECT: Credible because this transition is required if adverse containment conditions existed or if all SG levels were less than 14%.

C: INCORRECT: Credible because this transition is required if a SG level and pressure are lowering in an uncontrolled manner. Pressure is not given but can assume to be controlled if levels are as indicated.

D: CORRECT: With PZR level and pressure slowly lowering and given the other parameters and trends, 0POP05-EO-EO10 would be the correct procedure to enter.

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze the given conditions and apply their knowledge of Rediagnosis procedure.

Page 12 of 150

Exam Bank No.: 52 A Reactor Trip and Safety Injection have occurred with the following conditions:

SI has been reset Operators have just completed step 1 of 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Due to some confusion with the event, the Unit Supervisor also enters 0POP05-EO-ES00, Rediagnosis.

Subsequently a Reactor Operator observes the following:

RCS Pressure: ..............................1830 psig RCS Subcooling:..........................60oF Pressurizer Level:.........................20%

SG A NR Level: ...........................8%

SG B NR Level: ...........................10%

SG C NR Level: ...........................17%

SG D NR Level: ...........................19%

Total AFW Flow: .........................400 gpm Containment Pressure:..4 psig Containment Radiation Level:......5 R/HR Which procedure should the crew transition to?

A. 0POP05-EO-FRZ1, Response to High Containment Pressure.

B. 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink.

C. 0POP05-EO-FRI2, Response to Low Pressurizer Level.

D. 0POP05-EO-ES11, SI Termination.

Answer: D Transition to 0POP05-EO-ES11, SI Termination

Exam Bank No.: 52 K/A Catalog Number: W/E01 EA2.1 Tier: 1 Group/Category: 2 RO Importance: 3.2 10CFR

Reference:

55.41(b)(10)

Ability to determine and interpret the following as they apply to the (Reactor Trip and Safety Injection Rediagnosis):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

STP Lesson: LOT 504.09 Objective Number: 81187 DISCUSS the indications available to determine plant status during a loss of primary or secondary coolant accident.

Reference:

0POP05-EO-EO10, Conditional Information Page Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because this transition may be done with a higher containment pressure (9.5 psig).

B: INCORRECT: Credible because this transition is required if adverse containment conditions existed or if all SG levels were less than 14%.

C: INCORRECT: Credible because transition may be done with a lower pressurizer level (17%).

D: CORRECT: The given conditions would allow transition to ES11 which would be the expected action for the Crew.

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze the given conditions and apply their knowledge of SI termination and reinitiation requirements and the loss of heat sink and integrity transitions in order to eliminate the incorrect responses and choose the correct response.

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 16 Last used on an NRC exam: 1999 RO Sequence Number: 7 A Large Break LOCA has occurred with the following plant conditions:

Containment pressure is 10.2 psig Containment temperature is 240 F Containment radiation is 3.0E + 03R/HR The Crew has entered 0POP05-EO-FRZ3, Response to High Containment Radiation Level.

Which of the following actions is required in accordance with 0POP05-EO-FRZ3, Response to High Containment Radiation Level?

A. Verify Containment Phase A Isolation has occurred.

B. Ensure all Reactor Containment Fan Coolers running.

C. Verify Containment Ventilation Isolation has occurred.

D. Ensure all Containment Spray Pumps running.

Answer: C Verify Containment Ventilation Isolation has occurred.

Page 13 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 16 K/A Catalog Number: W/E16 EA1.3 Tier: 1 Group/Category: 2 RO Importance: 2.9 10CFR

Reference:

55.41(b)(10)

Ability to operate and/or monitor the following as they apply to the High Containment Radiation: Desired operating results during abnormal and emergency situations.

STP Lesson: LOT 504.42 Objective Number: T50442 Without using reference material unless provided, the student will be able to use 0POP05-EO-FRZ3 to correctly respond to a high containment radiation level.

Reference:

0POP05-EO-FRZ3, Rev 2 Step 1 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because containment phase A isolation should have occurred but other procedures verify phase A and this is not an action directed by 0POP05-EO-FRZ3.

B: INCORRECT: Plausible because it is all RCFCs should be in operation but this action is not directed by 0POP05-EO-FRZ3.

C: CORRECT: CVI should be ensured to prevent release.

D: INCORRECT: Plausible because it is desirable to have CS Pumps running however in this condition only 2 would be running and this action is not directed by 0POP05-EO-FRZ3.

Question Level: H Question Difficulty 3 Justification:

Student must analyze the given conditions and apply the mitigating strategies for a high radiation condition in containment to the given conditions and determine whether any operating limits have been exceeded for components contained within the distractors.

Page 14 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 18 Last used on an NRC exam: 2010 RO Sequence Number: 8 Per 0POP05-EO-EC11, Loss of Emergency Coolant Recirculation, which of the following is the correct SEQUENCE to stop backflow from the RWST to the Containment Sump?

Stop the LHSI, HHSI and CS Pumps, then ....

A. 1. Open the RWST to SI Suction Header Valves

2. Open the SI Pump Mini Flow Valves

3. Close the Containment Sump Suction Valves

4. Start the LHSI, HHSI and CS Pumps as necessary B. 1. Close the Containment Sump Suction Valves

2. Open the RWST to SI Suction Header Valves

3. Open the SI Pump Mini Flow Valves

4. Start the LHSI, HHSI and CS Pumps as necessary C. 1. Open the RWST to SI Suction Header Valves

2. Close the Containment Sump Suction Valves

3. Open the SI Pump Mini Flow Valves

4. Start the LHSI, HHSI and CS Pumps as necessary D. 1. Open the SI Pump Mini Flow Valves

2. Close the Containment Sump Suction Valves

3. Open the RWST to SI Suction Header Valves

4. Start the LHSI, HHSI and CS Pumps as necessary Answer: B 1. Close the Containment Sump Suction Valves; 2. Open the RWST to SI Suction Header Valves; 3. Open the SI Pump Mini Flow Valves; 4. Start the LHSI, HHSI and CS Pumps as necessary Page 15 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 18 K/A Catalog Number: W/E11 EA1.2 Tier: 1 Group/Category: 1 RO Importance: 3.5 10CFR

Reference:

55.41(b)(7)

Ability to operate and / or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation): Operating behavior characteristics of the facility.

STP Lesson: LOT 504.27 Objective Number: 82598 DESCRIBE the indications and anticipated readings used to determine that there is no backflow from the RWST to the emergency sump.

Reference:

0POP05-EO-EC11 Rev 19 page 11 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.

B: CORRECT: The containment sump valves must be closed before the RWST and the SI Pump mini flow valves can be opened C: INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.

D: INCORRECT: All distractors are plausible because the end alignment would be the same but it would not work without understanding the operating characteristic of the system that the containment sump valve must be closed before the RWST and the SI Pump mini flow valves can be opened.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the data given and have knowledge of systems interlocks to determine the proper sequence for this valve re-alignment.

Page 16 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 27 Last used on an NRC exam: 2009 RO Sequence Number: 9 Following a Loss of Offsite Power (LOOP), which ONE of the following statements describes the Pressurizer heater groups that will be available to maintain Pressurizer pressure?

A. Backup heater groups A and B ONLY B. Backup heater groups D and E ONLY C. All Backup heater groups EXCEPT control heater group C D. All Backup heater groups Answer: A Backup heater groups A and B ONLY Page 17 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 27 K/A Catalog Number: 011 K2.02 Tier: 2 Group/Category: 2 RO Importance: 3.1 10CFR

Reference:

55.41(b)(7)

Knowledge of bus power supplies to the following:

PZR heaters STP Lesson: LOT 201.14 Objective Number: 8860 List the power supplies to the pressurizer heaters.

Reference:

LOT201.14 Lesson Plan on PZR Level and Pressure Control Rev 14 Slide 9 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: Backup heater groups A and B are supplied power from Class 1E Load Centers E1A1 and E1C1 respectively. These loads are backed by their respective ESF D/Gs. All other heaters are non-safety related and do NOT have any backup power.

B: INCORRECT: Plausible because students have to remember which to PZR heater groups are backed by an ESF D/G.

C: INCORRECT: Plausible because of the importance of having PZR heaters available and the student believing that all heaters would have ESF Backup power except for the control group.

D: INCORRECT: Plausible because of the importance of having PZR heaters available and the student thinking that all heaters would be backedup by ESF D/Gs.

Question Level: F Question Difficulty 3 Justification:

The candidate must know which heater groups are safety related (i..e ESF diesel powered).

Page 18 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 594 Last used on an NRC exam: Never RO Sequence Number: 10 Unit 1 was stable at 100 % power when a piping failure in the Generator Hydrogen Supply Header caused Main Generator hydrogen pressure to lower.

After the leak was isolated, the following conditions exist:

Generator hydrogen pressure - 45 psig Generator output - 1300 MWe Generator reactive load - 300 MVARs OUT Based on these conditions, and referring to Figure 7.1 (attached), which of the following represents the HIGHEST ALLOWABLE amount the generator load should be reduced to?

A. 1225 MWe and 240 MVARs OUT.

B. 1225 MWe and 100 MVARs OUT.

C. 1175 MWe and 340 MVARs OUT.

D. 1175 MWe and 200 MVARs OUT.

Answer: D 1175 MWe and 200 MVARs OUT.

Page 19 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 594 K/A Catalog Number: G2.1.25 Tier: 3 Group/Category: 1 RO Importance: 3.9 10CFR

Reference:

55.41(b)(6)

Ability to interpret reference materials, such as graphs, curves, tables, etc.

STP Lesson: LOT 202.17 Objective Number: 3872 DISCUSS the relationship between Main Generator load and Generator Hydrogen gas pressure. Include in the discussion how generator capacity varies with reduced Hydorgen gas pressure and what gas pressure would require a generator shutdown.

Reference:

LOT 202.17 Lesson Plan Rev 10 on the Main Generator and Main Generator Capability Curve- Unit 1 Plant Curve Book Figure 7.1 Attached Reference

Attachment:

Main Generator Capability Curve- Unit 1 Plant Curve Book Figure 7.1 NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.

B: INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.

C: INCORRECT: All distractors are plausible because the student must be able to use the Main Generator Capability Curve.

D: CORRECT: 1175 MWE and 200 MVARS OUT would represent the only loading that would meet the restrictions of the Main Generator Capability curve with the conditions given.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the conditions given and be able to use the Main Generator Capability Curve.

Page 20 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 715 Last used on an NRC exam: Never RO Sequence Number: 11 The Unit is in Mode 6 with the following condition:

Audio Count Rate is selected to the SR N31 position.

Subsequently:

Source Range N31 fails due to a loss of power.

Alarm(s) (1) will annunciate.

AND Audio count rate will (2) to Source Range Channel N32.

A. (1) SR HI VOLT TRBL ONLY (2) automatically swap B. (1) SR HI VOLT TRBL ONLY (2) have to be manually aligned C. (1) SR HI VOLT TRBL AND SR/IR TRIP BYPASS (2) automatically swap D. (1) SR HI VOLT TRBL AND SR/IR TRIP BYPASS (2) have to be manually aligned Answer: B (1) SR HI VOLT TRBL ONLY (2) have to be manually aligned Page 21 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 715 K/A Catalog Number: APE 032 AK2.01 Tier: 1 Group/Category: 2 RO Importance: 2.7 10CFR

Reference:

55.41(b)(7)

Knowledge of the interrelations between the Loss of Source Range Nuclear Instrumentation and the following: Power supplies, including proper switch positions.

STP Lesson: LOT 201.16 Objective Number: 4886 DESCRIBE all the interlocks, trips, permissives, alarms and/or indications associated with the Nuclear Instrument System, including setoints and coincidences.

Reference:

LOT 201.16, Rev 13, Lesson Plan on Excore Nis Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because it would be reasonable to believe that audio count rate would automatically swap to an operable channel.

B: CORRECT: With the condition given only the SR HI VOLT TRBL would alarm and the audio count rate has to be amnually swapped.

C: INCORRECT: Plausible because SR/IR TRIP BYPASS alarm is associated with the source range instruments. Incorrect because the alarm only comes in if the channel is physically placed in BYPASS. Plausible because it would be reasonable to believe that audio count rate would automatically swap to an operable channel.

D: INCORRECT: Plausible because SR/IR TRIP BYPASS alarm is associated with the source range instruments. Incorrect because the alarm only comes in if the channel is physically placed in BYPASS.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition and determine the action necessary to restore the Control Room Audio Count Rate.

Page 22 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2580 K/A Catalog Number: 001 K1.04 Tier: 2 Group/Category: 2 RO Importance: 3.2 10CFR

Reference:

55.41(b)(7)

Knowledge of the physical connections and/or cause-efect relationships between the CRDS and the following systems:

RCS STP Lesson: LOT 201.18 Objective Number: 86061 Describe the instrumentation and controls available to monitor and operate the Rod Control System.

Reference:

LOT 201.18 Rev 15 slide 66 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: With a 3 degree Tave and Tref deviation rods will step in at 6 steps per minute and continue until the T error is 1 degree difference and then stop.

B: INCORRECT: Plausible because it is desired to have Tave and Tref matched with a T error of 0 degrees. Incorrect because even though with a 3 degree Tave and Tref deviation rods will step in at 6 steps per minute, they stop moving at a Terror of 1 degree.

C: INCORRECT: Plausible because rods do step at 33 steps a minute but only if the Tave and Tref deviation is 4 degrees.

D: INCORRECT: Plausible because rods do step at 33 steps a minute but only if the Tave and Tref deviation is 4 degrees. Also plausible because it is desired to have Tave and Tref matched with a T error of 0 degrees.

Question Level: H Question Difficulty 3 Justification:

The student must be able to anlayze the given conditions and have fundamental knowledge of how the RCS is affected by Control Rod movement.

Page 24 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 816 Last used on an NRC exam: Never RO Sequence Number: 13 Which of the following is the indication used in 0POP05-EO-EC12, LOCA Outside Containment, that ensures actions taken to isolate a leak inside the FHB has been successful and that the procedure can be exited?

A. FHB Area Radiation alarms clearing.

B. RCS pressure rising.

C. FHB SI/CS Pump Sump level alarms clearing.

D. RCS Hot Leg temperatures lowering.

Answer: B RCS pressure rising.

Page 25 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 816 K/A Catalog Number: W/E04 EK1.2 Tier: 1 Group/Category: 1 RO Importance: 3.5 10CFR

Reference:

55.41(b)(10)

LOCA Outside Containment Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment):

Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).

STP Lesson: LOT 504.46 Objective Number: 82657 From Memory STATE/IDENTIFY indications and trends used to determine that the break is isolated in accordance with POP05-EO-EC12.

Reference:

0POP05-EO-EC12, Rev. 9, Steps 4j. and 6a.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because the presence of FHB Area Radiation alarms are used to indicate that a LOCA has ocurred in the FHB but they are not used in reverse to allow exiting the procedure.

B: CORRECT: Once the RCS leak is isolated, RCS pressure should begin to rise and the procedure can be exited.

C: INCORRECT: Plausible because the presence of FHB Sump alarms are used to indicate that a LOCA has ocurred in the FHB but they are not used in reverse to allow exiting the procedure.

D: INCORRECT: Plausible because Hot leg temperatures lowering would be desirable and indicative of restoring RCS conditions but it is not an indication used in 0POP05-EO-EC12 that the LOCA outside containment has been isolated.

Question Level: F Question Difficulty 3 Justification:

Student must have fundamental knowledge of the steps in 0POP05-EO-EC12, LOCA Outside Containment.

Page 26 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2584 Last used on an NRC exam: Never RO Sequence Number: 14 Which of the following would cause contamination of Component Cooling Water system at 100% power while in a normal lineup?

A leak in the heat exchanger.

A. Seal Water B. RCP Thermal Barrier C. Excess Letdown D. Charging Pump Lube Oil Answer: B RCP Thermal Barrier Page 27 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2584 K/A Catalog Number: G2.3.14 Tier: 3 Group/Category: 3 RO Importance: 3.4 10CFR

Reference:

55.41(b)(12)

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

STP Lesson: LOT 204.01 Objective Number: 20401 Given plant or system conditions, predict the response of the plant and/or systems

Reference:

LOT 201.12 Rev 14 lesson plan slide 44 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: The seal water heat exchanger is at a lower pressure than CCW so it would not leak into CCW system. Plausible because seal water HX is cooled by CCW.

B: CORRECT: RCP thermal barrier heat exchanger is at a higher pressure than CCW. Therefore, it would leak into the CCW system causing a contamination/radiation hazard.

C: INCORRECT: Because it is not inservice normally at 100% power. Plausible, because it is at a higher pressure than CCW so IF it was in service then it would leak into CCW system.

D: INCORRECT: The charging pump lube oil heat exchanger is at a lower pressure than CCW so it would not leak into CCW system. Plausible because charging pump lube oil HX is cooled by CCW.

Question Level: F Question Difficulty 3 Justification:

The applicant must be able to determine which action will control (minimize) the radiation release with the given conditions.

Page 28 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2582 Last used on an NRC exam: Never RO Sequence Number: 15 An area radiation monitor in the MAB detects (1) radiation, and its reading can be obtained on (2) radiation display panel in the control room.

A. (1) alpha (2) RM-11 B. (1) gamma (2) RM-11 C. (1) alpha (2) RM-23 D. (1) gamma (2) RM-23 Answer: B (1) gamma (2) RM-11 Page 29 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2582 K/A Catalog Number: APE 061 G2.4.31 Tier: 1 Group/Category: 2 RO Importance: 4.2 10CFR

Reference:

55.41(b)(6)

ARM System Alarms:

Knowledge of annunciator alarms, indications, or response procedures.

STP Lesson: LOT 202.41 Objective Number: 92124 List the types of alarms associated with the area radiation monitoring system (ARMS).

Reference:

LOT 202.41 Rev 15 slide 11 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Area rad monitors detect beta and gamma radiation. Plausible because alpha radiation is a type of radiation that can be detected. However, ARMs is no tdesigned to accurately detect alpha radiaiton.

B: CORRECT: Area rad monitors detect beta and gamma radiation of various places in the plant. Their indications are monitored on RM-11 in the control room.

C: INCORRECT: Area rad monitors detect beta and gamma radiation of various places in the plant.

Their indications are monitored on RM-11 in the control room. Plausible because alpha radiation is a type of radiation that can be detected. However, ARMs is no tdesigned to accurately detect alpha radiaiton. Also, plausible because RCB area rad monitors RT-8050 and RT-8051 are also read on RM-23.

D: INCORRECT: The area rad monitors in the MAB are read on RM-11 in the control room. Plausible because RCB area rad monitors RT-8050 and RT-8051 are also read on RM-23.

Question Level: F Question Difficulty 2 Justification:

Student must have fundamental knowledge of ARMs and their indications.

Page 30 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1053 Last used on an NRC exam: 2009 RO Sequence Number: 16 The Unit was in Mode 1 when a loss of offsite power occurred with the following condition:

ESF DG #11 failed to start.

Before operator actions are taken, 125 VDC Bus E1A11 will A. continue to supply its loads from its Battery Bank for a minimum of 12 hrs.

B. continue to supply its loads automatically from its Standby Battery Charger.

C. not be supplied by a Battery Charger until power is restored to Train A 4160 v Bus.

D. not be supplied by a Battery Charger until power is restored to Train A 4160 v Bus AND a Battery Charger is manually realigned.

Answer: C not be supplied by a Battery Charger until power is restored to Train A 4160 v Bus.

Page 31 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1053 K/A Catalog Number: APE 058 AK1.01 Tier: 1 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(8)

Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation STP Lesson: LOT 201.37 Objective Number: 63901 GIVEN a loss of power, PREDICT the operation of the class 1E 125 VDC Electrical Distribution System to include automatic actions and interlocks.

Reference:

LOT 201.37, Class 1E 125 VDC System Rev 9 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because the batteries would continue to supply the bus but are rated for 2 hrs. minimum, not 12.

B: INCORRECT: Plausible because there is a second charger but it is also supplied via 4KV bus "A" and must be placed in service manually when it is used.

C: CORRECT: The battery charger that was in service before the LOOP will remain aligned and be returned to service when power is restored to the Train A 4160 V Bus.

D: INCORRECT: Plausible if the student believes a charger that was previously in service would need to be realigned to the DC Bus after power is restored.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition to determine how the Vital DC distribution system will react.

Page 32 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1189 Last used on an NRC exam: 2005 RO Sequence Number: 17 Which of the following conditions concerning the Personnel Air Lock would cause a loss of CONTAINMENT INTEGRITY?

A. The outer and inner doors are opened simultaneously for a normal transit entry into containment while in MODE 4.

B. One air lock door fails acceptance test criteria while the plant is in MODE 6.

C. Welding cables are laid through both airlock doors while the plant is in MODE 5.

D. The outer door is opened for a normal transit entry into containment while in MODE 3.

Answer: A The outer and inner doors are opened simultaneously for a normal transit entry into containment while in MODE 4.

Page 33 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1189 K/A Catalog Number: 103 K3.02 Tier: 2 Group/Category: 1 RO Importance: 3.8 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect that a loss or malfunction of the containment system will have on the following:

Loss of containment integrity under normal operations STP Lesson: LOT 503.01 Objective Number: 92101 From memory, DEFINE terms used in the Technical Specifications and the Technical Requirements Manual (TRM).

Reference:

LOT 503.01 Lesson Plan for Tech Specs and TRM, TS 3.6.1.3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: CONTAINMENT INTEGRITY is required in Modes 1 to 4 and if both Air Lock doors are open at the same time then CONTAINMENT INTEGRITY would be affected.

B: INCORRECT: Plausible because a failed surveillance would constitute a loss of CONTAINMENT INTEGRITY if the plant was in MODE 1 to 4.

C: INCORRECT: Plausible because this would constitute a loss of CONTAINMENT CLOSURE if in Mode 6 but not CONTAINMENT INTEGRITY.

D: INCORRECT: Plausible because it would be reasonalble to believe that with just one door open that CONTAINMENT INTEGRITY could be affected while in the higher Mode.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of the definition of CONTAINMENT INTEGRITY and what modes of operation it applies.

Page 34 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1355 Last used on an NRC exam: Never RO Sequence Number: 18 A Reactor Trip and Safety Injection have occurred from 100% power with the following conditions:

SG A NR level lowered to 22% but now is slowly rising due to a SGTR.

SG B, C & D NR levels lowered to 10% and are now slowly rising due to AFW flow.

The Secondary Operator is directed to CLOSE A SG AFW OCIV but the valve will not stay closed.

Which of the following need to be reset for A SG AFW OCIV to operate properly?

1. AFW AUTO FLOW CONT

2. SG Lo-Lo Level Actuation

3. Safety Injection A. 1 ONLY B. 2 and 3 ONLY C. 3 ONLY D. 1 and 3 ONLY Answer: B 2 and 3 ONLY Page 35 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1355 K/A Catalog Number: 013 K1.07 Tier: 2 Group/Category: 1 RO Importance: 4.1 10CFR

Reference:

55.41(b)(7)

Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems:

AFW System STP Lesson: LOT 202.28 Objective Number: 43805 DESCRIBE the AFW system controls and indications the the MCR.

Reference:

Lesson Plan LOT 202.28 Auxiliary Feedwater Rev 10 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because the Auto Flow Control has to be reset for the AFW reg valve to be operated manually but not for the AFW OCIV.

B: CORRECT: The condition states that an SI has actuated and a SG LO-LO level actuatiion would have also occurred on a trip from 100% power. It takes a reset of Safety Injection and SG LO-LO Level Actuation to be able to operate the AFW OCIV.

C: INCORRECT: Plausible because the condition states that SG A level did not go below the SG LO-LO actuation setpoint of 20% but the other SGs would have so it takes a reset of both Safety Injection and SG LO-LO Level Actuation.

D: INCORRECT: Plausible because the Auto Flow Control has to be reset for the AFW reg valve to be operated manually but not for the AFW OCIV. and the condition states that a Safety Injection has occurred.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given conditions to determine the correct ESFAS signal to reset.

Page 36 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2597 Last used on an NRC exam: Never RO Sequence Number: 19 The smoke detectors in the 35 EAB Relay Room have lost power.

A trouble alarm with a (1) light will be received on the Control Room Fire Detection Computer initiated from the (2) Fire Protection System used in the Relay Room.

A. (1) YELLOW (2) Automatic Wet Pipe Sprinkler B. (1) YELLOW (2) Halon C. (1) RED (2) Automatic Wet Pipe Sprinkler D. (1) RED (2) Halon Answer: B (1) YELLOW - (2) Halon Page 37 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2597 K/A Catalog Number: 086 K6.04 Tier: 2 Group/Category: 2 RO Importance: 2.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect of a loss or malfunction of the following on the Fire Protection system:

Fire, smoke, and heat detectors.

STP Lesson: LOT 201.29 Objective Number: 53554 DESCRIBE the Fire Detection System response to a generic alarm condition to include local panel indications and control room indications.

Reference:

LOT 201.29 Rev 7 LP and 0POP02-FA-0001 Rev 15 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because all other rooms on the 35' of the EAB use a Automatic Wet Pipe Fire Protection System. Incorrect because the Relay Room uses a Halon Fire Protection System.

B: CORRECT: A loss of power will cause a trouble alarm that is indicated by a yellow light. The Relay Room uses a Halon Fire Protection System.

C: INCORRECT: Plausible because a red light is the color of other alarms on the control room fire detection computer. Also plausible because all other rooms on the 35' of the EAB use a Automatic Wet Pipe Fire Protection System. Incorrect because a loss of power creates a yellow trouble alarm.

Incorrect because the Relay Room uses a Halon Fire Protection System.

D: INCORRECT: Plausible because a red light is the color of other alarms on the control room fire detection computer. Incorrect because a loss of power creates a yellow trouble alarm.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of the different types of fire protection for different systems and knowledge of basic detector failures with response of the Fire Protection system.

Page 38 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1579 Last used on an NRC exam: Never RO Sequence Number: 20 Which of the following describes the logic to actuate AMSAC?

When turbine impulse pressure is greater than 30%, SG NR level in...

A. one SG lowering to <15% for 25 seconds will actuate Main Turbine relays that will automatically trip the Main Turbine.

B. three SGs lowering to <15% for 25 seconds will actuate Main Turbine relays that will automatically trip the Main Turbine.

C. one SG lowering to <15% for 25 seconds AND low feedwater flow in 3 feedwater lines will actuate relays that will actuate Auxiliary Feedwater.

D. three SGs lowering to <15% for 25 seconds AND low feedwater flow in 3 feedwater lines will actuate relays that will actuate Auxiliary Feedwater.

Answer: B three SGs lowering to <15% for 25 seconds will actuate Main Turbine relays that will automatically trip the Main Turbine.

Page 39 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1579 K/A Catalog Number: EPE 029 EK2.06 Tier: 1 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Knowledge of the interrelations between ATWS and the following: Breakers, Relays, and Disconnects STP Lesson: LOT 201.40 Objective Number: 91840 STATE how the control room operator is aware of AMSAC condition.

Reference:

LOT 201.40 Rev 9 slide 27 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because a low SG level less than 15%. Incorrect because it has to be in 3 out of 4 SGs to actuate AMSAC.

B: CORRECT: AMSAC is actuated when greater than 30% Turbine impulse pressure (2/2) and when SG NR levels (3/4) <15% for >25 sec. AMSAC automatically trips the Main Turbine and actuates AFW (also, secures SG blowbown and samples) to minimize secondary inventory loss and ensure a secondary heat sink is maintained.

C: INCORRECT: Plausible because low feedwater flow could be indicative of a loss of Main Feedwater but it is NOT the signal to actuate AMSAC. Plausible because a low SG level less than 15%.

Incorrect because it has to be in 3 out of 4 SGs to actuate AMSAC.

D: INCORRECT: Plausible because low feedwater flow could be indicative of a loss of Main Feedwater but it is NOT the signal to actuate AMSAC.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of the actuation logic of AMSAC.

Page 40 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1584 Last used on an NRC exam: Never RO Sequence Number: 21 The Unit is at 100% with the following conditions:

Train A CCW is running (Train Selector Switch in RUN).

Train B CCW is secured (Train Selector Switch in STANDBY).

Train C CCW is secured (Train Selector Switch in OFF).

Subsequently:

CCW HX A OUTL PRESS LO and CCW HX A OUTL FLOW HI/LO alarms.

CCW discharge pressure is 80 psig.

CCW HX A flow is 16,000 gpm.

What is the correct action the operator will take NEXT per 0POP09-AN-02M3?

A. Ensure Train B CCW pump running.

B. Secure Train A CCW pump.

C. Isolate CCW non-essential header.

D. Start Train B or C CCW pump.

Answer: D Start Train B or C CCW pump.

Page 41 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1584 K/A Catalog Number: 008 G2.4.50 Tier: 2 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(7)

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

STP Lesson: LOT 201.12 Objective Number: 5213 GIVEN a plant or system condition, PREDICT the operation of the Component Cooling Water System.

Reference:

0POP09-AN-02M3 Rev 29 page 10 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because the student may think 'B' train auto started on low discharge pressure.

B: INCORRECT: Plausible the student may think 'B' train auto started. The standby train auto starts at 76 psig and then the 0POP09 would direct the operator to secure the 'A' pump.

C: INCORRECT: Plausible because with only one pump in operation and HX flow high, reducing loads would help. However, isolating the non-essential header is not desired at 100% power. The 0POP09 has you isolate it only if there has been a LOOP.

D: CORRECT: The standby train CCW will NOT auto start unitl 76 psig. Therefore, only one train is in operation. Also HX flow is high (greater than 15,500 gpm). Per the 0POP09 the correct action to take is to start an additional CCW pump.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition to determine the correct system response and the required annunciator response actions.

Page 42 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1689 Last used on an NRC exam: Never RO Sequence Number: 22 The greatest likelihood of causing an observable Pressurizer Relief Tank level change while forming a steam bubble in the Pressurizer would be with the RCS in a .

A. solid plant condition because of the potential to lift a Pressurizer PORV.

B. vacuum-filled condition because of the potential for Pressurizer PORV seat leakage.

C. solid plant condition because of the potential to lift a Pressurizer Safety Valve.

D. vacuum-filled condition because of the potential for Reactor Vessel Head Vent Valve seat leakage.

Answer: A solid plant condition because of the potential to lift a Pressurizer PORV.

Page 43 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1689 K/A Catalog Number: 007 K5.02 Tier: 2 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(5)

Knowledge of the operational implications of the following concepts as they apply to the PRTS:

Method of forming a steam bubble in the Pressurizer STP Lesson: LOT 201.04 Objective Number: 91039 DESCRIBE the procedure for formation of a pressurizer bubble.

Reference:

0POP03-RC-0100, RCS Vacuum Fill Rev 41; 0POP03-ZG-0001, Plant Heatup Rev 68 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: The greatest likelihood of causing a PRT level change would be from the PZR PORV during solid plant conditions due to the lift setting for the PZR PORV being lower due to COMS.

B: INCORRECT: This distractor is credible because the PZR PORV and Safety Valves could leak by under a vacuum but the likelihood is small. In addition, the flow would have to be from the PRT, which is low in the RCB, to the top of the PZR. It may cause a slight change in PRT level but not an observable change.

C: INCORRECT: This distractor is credible because PZR Safety Valves could lift, however, during solid plant conditions, the Pressurizer PORVs have a lower setpoint through the COMS control, thus they will lift at a significantly lower pressure. The Pressurizer Safety valves will lift at much higher pressure so the potential for them lifting is not as great as that for the PORVs.

D: INCORRECT: This distractor is credible because the head vent valves could leak by under a vacuum but the likelihood is small. In addition, the flow would have to be from the PRT, which is lower in the RCB, to the Rx Vessel. It may cause a slight change in PRT level but not an observable change.

Question Level: H Question Difficulty 3 Justification:

The student must be able to evaluate the different senarios given for forming a PZR Bubble and determine which method and case given would be the most likely to cause an observable level change in the PRT.

Page 44 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2583 Last used on an NRC exam: Never RO Sequence Number: 23 The Unit is raising reactor power with current reactor power at 8% when the following occurs:

A pressurizer spray valve fails open.

The spray valve cannot be manually closed.

The crew should trip RCP A and D per (1) to prevent a (2) from occurring.

A. (1) 0POP04-RP-0001, Loss of Automatic Pressurizer Pressure Control (2) Reactor Trip B. (1) 0POP04-RP-0001, Loss of Automatic Pressurizer Pressure Control (2) Safety Injection C. (1) 0POP03-ZG-0005, Plant Startup to 100%

(2) Reactor Trip D. (1) 0POP03-ZG-0005, Plant Startup to 100%

(2) Safety Injection Answer: B (1) 0POP04-RP-0001, Loss of Automatic Pressurizer Pressure Control (2) Safety Injection Page 45 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2583 K/A Catalog Number: 013 A2.03 Tier: 2 Group/Category: 1 RO Importance: 4.4 10CFR

Reference:

55.41(b)(10)

Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Rapid depressurization.

STP Lesson: LOT 201.20 Objective Number: 507227 Given a description of plant conditions, ANALYZE the conditions and PREDICT how the Solid State Protection System will respond.

Reference:

0PO04-RP-0001 Rev 15 CIP Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because with the reactor at low power tripping two RCPs would not automatically trip the reactor. Incorrect because the reactor would have to be manually tripped without all 4 RCPs running.

B: CORRECT: If a spray valve fails open that can not be manually closed, the CIP of 0POP04-RP-0001 has the crew trip the reactor/turbine, then secure both RCP A and D before performing the actions of 0POP05-EO-EO00. A rapid depressurization of the RCS occurs when a PZR spray valve fails open and if the RCPs are secured quickly enough a Safety Injection may be avoided.

C: INCORRECT: Plausible because with the reactor at low power tripping two RCPs would not automatically trip the reactor. Incorrect because the reactor would have to be manually tripped without all 4 RCPs running. Since the crew would be using 0POP03-ZG-0005 to raise reactor power it is plausible for the student to believe that 0POP03-ZG-0005 would have steps to perform tripping RCP A and D if a spray valve were to fail open.

D: INCORRECT: Since the crew would be using 0POP03-ZG-0005 to raise reactor power it is plausible for the student to believe that 0POP03-ZG-0005 would have steps to perform tripping RCP A and D if a spray valve were to fail open.

Question Level: H Question Difficulty 3 Justification:

Student must analyze given conditions and predict plant response and select appropriate procedure to mitigate the event.

Page 46 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1755 Last used on an NRC exam: 2009 RO Sequence Number: 24 The Unit is operating at 100% power with the following conditions:

ECW/CCW Train Mode Selector Switches are aligned as follows:

o A in RUN; B in STBY; C in OFF All ECW Pumps are running Subsequently:

The 13.8 KV feeder breaker to ESF 4.16 KV Bus Train C opens causing a loss of power to the Bus.

ECW Pump 1C will be stripped and..

A. must be started manually. Once started, it will automatically supply cooling water flow to

1. 13 ESF D/G and Train C CCW HX ONLY.

B. then sequenced on, automatically supplying cooling water flow to #13 ESF D/G and Train C CCW HX ONLY.

C. must be started manually. Once started, it will automatically supply cooling water flow to

1. 13 ESF D/G, Train C CCW Hx, Train C Essential Chiller and Train C CCW Pump Supplementary Cooler.

D. then sequenced on, automatically supplying cooling water flow to #13 ESF D/G, Train C CCW Hx, Train C Essential Chiller, and Train C CCW Pump Supplementary Cooler.

Answer: D sequenced on, automatically supplying cooling water flow to #13 ESF D/G, Train 'C' CCW Hx, Train 'C' Essential Chiller, and Train 'C' CCW Pump Supplementary Cooler.

Page 47 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1755 K/A Catalog Number: 076 A3.02 Tier: 2 Group/Category: 1 RO Importance: 3.7 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the SWS, including: Emergency heat loads.

STP Lesson: LOT 201.13 Objective Number: 91201 GIVEN a plant or system condition, PREDICT the operation of the Essential Cooling Water System.

Reference:

LOT201.13 ECW System, Rev 7, LOT201.41 ESF Sequencers Rev 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because the student must remember that the ECW/CCW Mode Selector Switch does not affect ECW Pump starts from the sequencer. ECW Pump 1C will be automatically sequenced on. Manual start of the pump is not required. When ECW Pump 1C is running it supplys cooling water flow to #13 ESF D/G, Train C CCW Hx, Train C Essential Chiller, and Train C CCW Supplemnetary Cooler.

B: INCORRECT: Plausible because the student must remember which cooling loads are supplied from the ECW Sytem. ECW Pump 1C is automatically sequenced on, however when it is running it will supply cooling water flow to #13 ESF D/G, Train C CCW Hx, Train C Essential Chiller, and Train C CCW Supplemnetary Cooler.

C: INCORRECT: Plausible because the student must remember that the ECW/CCW Mode Selector Switch does not affect ECW Pump starts from the sequencer. ECW Pump 1C will be automatically sequenced on. Manual start of the pump is not required.

D: CORRECT: ECW Pump 1C will be automatically sequenced on. ESF D/G #13, Train C CCW HX, Train C Essential Chiller, and Train C CCW Pump Supplementary Cooler are all the loads that will be supplied. These loads are always alligned with manual valves to allow cooling water flow.

Therefore when the ECW Pump is running cooling water flow is automatically supplied to the loads.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given conditions and have an understanding of how a loss of the standby bus affects ECW operation along with a knowledge of the loads supplied by the ECW trains and that all loads are supplied when the ECW Pump in that respective train is running.

Page 48 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1833 Last used on an NRC exam: 2009 RO Sequence Number: 25 The Unit 1 is at 100% power with the following conditions:

A compressor malfunction caused one Starting Air Receiver on ESF DG #12 to completely depressurize.

The second Starting Air Receiver is unaffected and at normal operating pressure.

Subsequently:

A Unit 1 Standby Transformer lockout occurs.

Which of the following correctly describes the effect of the depressurized air receiver on this event?

ESF DG #12 will...

A. NOT receive a start signal, but IS capable of starting if needed.

B. NOT receive a start signal and is NOT capable of starting.

C. receive a start signal and WILL start and run.

D. receive a start signal, but is NOT capable of starting.

Answer: C receive a start signal and WILL start and run.

Page 49 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1833 K/A Catalog Number: 064 K6.07 Tier: 2 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect of a loss or malfunction of the following will have on the ED/G System: Air Receivers STP Lesson: LOT 201.39 Objective Number: 98476 Given a plant condition and/or various diesel modes of operation, PREDICT the response of the emergency diesels.

Reference:

LOT201.39, ESF Diesel Generator, PowerPoint presentation. Rev 16 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because in a normal Unit 1 lineup if the affected ESF D/G was 11 or 13 then they would NOT receive a start signal. With ONLY one receiver out of service the ESF DG is capable of starting.

B: INCORRECT: Plausible because in a normal Unit 1 lineup if the affected ESF D/G was 11 or 13 then they would NOT receive a start signal. With ONLY one receiver out of service the ESF DG is capable of starting.

C: CORRECT: In a normal electrical lineup 13.8 KV Standby BUS 1G feeds ESF 4.16 KV BUS Train B. If the Unit 1 Standby Transformer is lost then ESF D/G #12 will receive a start signal and start as long as at least one Starting Air Receiver is available.

D: INCORRECT: Plausible if the student believes that both Starting Air Recivers are needed to start the ESF D/G.

Question Level: H Question Difficulty 3 Justification:

Candidate must analyze the effect of the loss on the transformer on the diesel and then determine the effect of the depressurized receiver on the start capability.

Page 50 of 150

Print Date 4/18/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2502 Last used on an NRC exam: Never RO Sequence Number: 26 Per the WOG basis for a condition where all Feedwater (Main and Auxiliary) has been lost, why are the Reactor Coolant Pumps (RCPs) secured?

To prevent A. uncovering the Reactor Core.

B. seal damage to the RCPs.

C. heat input from the RCPs.

D. over pressurizing the Reactor Coolant System.

Answer: C heat input from the RCPs Page 51 of 150

Print Date 4/18/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2502 K/A Catalog Number: W/E05 EK3.1 Tier: 1 Group/Category: 1 RO Importance: 3.4 10CFR

Reference:

55.41(b)()

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.

STP Lesson: LOT 504.33 Objective Number: 83013 GIVEN a step, note or caution from 0POP05-EO-FRH1, STATE its basis.

Reference:

LOT 504.33 Lesson Plan on 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because uncovering the core is the basis for a step in 0POP05-EO-FRH1.

Performing the RCS Feed and Bleed.

B: INCORRECT: Plausible because preventing seal damage to the RCPs is the basis for steps in other EOPs that have the operators trip Reactor Coolant Pumps.

C: CORRECT: Securing the RCPs while performing 0POP05-EO-FRH1 removes the heat input form the RCPs and lengthens the time to dry out the SGs when feedwater flow is unavailable.

D: INCORRECT: Plausible because over pressurizing the RCS would be a concern while performing 0POP05-EO-FRH1.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of the reasons for performing emergency actions.

Page 52 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2579 Last used on an NRC exam: Never RO Sequence Number: 27 Which of the following is the entry condition listed in 0POP04-AE-0005, Offsite Power System Degraded Voltage?

(1) Switchyard Bus voltage indicating (2) when in a normal electrical lineup.

A. (1) North OR South (2) < 339 KV B. (1) North OR South (2) < 356 KV C. (1) North AND South (2) < 339 KV D. (1) North AND South (2) < 356 KV Answer: C (1) North AND South - (2) < 339 KV Page 53 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2579 K/A Catalog Number: APE 077 AA2.05 Tier: 1 Group/Category: 1 RO Importance: 3.2 10CFR

Reference:

55.41(b)(10)

Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Operational status of offsite circuit.

STP Lesson: LOT 201.30 Objective Number: 91662 Given control room indications associated with the Offsite Electrical Distribution system, EVALUATE plant conditions.

Reference:

0POP04-AE-0005 Rev 11 page 2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the correct voltage is listed but it applies to both the North and South Bus.

B: INCORRECT: Plausible because the correct voltage is listed for an alternate line up but it applies to both the North and South Bus.

C: CORRECT: The entry conditions for this procedure while in a normal electrical lineup is both North and South switchyard bus voltage less than 339 KV.

D: INCORRECT: Plausible because it would be correct voltage if in an alternate electrical lineup.

Question Level: F Question Difficulty 3 Justification:

student must know entry conditions for offnormal procedure.

Page 54 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2600 Last used on an NRC exam: Never RO Sequence Number: 28 The Core Cooling Critical Safety Function (CSF) uses the _________ from QDPS to determine the status of the CSF.

A. average of the highest CET indication from each core quadrant B. average of the 5th highest CET indication from each core quadrant C. highest of all CET indications D. 5th highest of all CET indications Answer: D 5th highest of all CET indications Page 55 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2600 K/A Catalog Number: EPE 074 EK2.08 Tier: 1 Group/Category: 2 RO Importance: 2.5 10CFR

Reference:

55.41(b)(2)

Knowledge of the interrelationships between the Inadequate Core Cooling and the following:

Sensors and Detectors STP Lesson: LOT 202.44 Objective Number: 7667 Given a change on plant or system condition, EXPLAIN the operation and indications of the QDPS System.

Reference:

LOT 202.44, Rev 12, Lesson Plan and Powerpoint slide 99 and 102.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because MAX QUAD CET average is used by QDPS. Incorrect because MAX QUAD CET average is used in the subcooling calculation.

B: INCORRECT: Plausible because the 5th highest CET average could be viewed as a correct answer by the student. Incorrect because the CSFs uses the 5th highest CET indication.

C: INCORRECT: Plausible because the highest CET indication could be viewed as a correct answer by the student. Incorrect because the CSFs uses the 5th highest CET indication.

D: CORRECT: For Core Cooling CSF, QDPS uses the 5th highest overall CET indication.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of QDPS indications to determine the correct answer.

Page 56 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2599 Last used on an NRC exam: Never RO Sequence Number: 29 Unit 1 tripped from 100% power with the following condition:

RCP 1A breaker tripped open.

The LOOP A FLOW FI-0417A for RCP 1A indication would lower to A. 0% and stabilize.

B. 50% and stabilize.

C. 0% and then reverse back up to 25% and stabilize.

D. 0% and then reverse back up to 50% and stabilize.

Answer: C 0% and then reverse back up to 25% and stabilize.

Page 57 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2599 K/A Catalog Number: 003 K5.02 Tier: 2 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(5)

Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP Coastdown on RCS Parameters.

STP Lesson: LOT 201.02 Objective Number: 86369 Describe the effects on the plant due to tripping a RCP.

Reference:

LOT 201.05 Rev 17 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the indication does go down to 0%. Incorrect because it does not stabilize at 0%.

B: INCORRECT: Plausible if student did not know there was reverse flow. Incorrect because it goes to 0 and then experiences reverse flow.

C: CORRECT: When a RCP trips the flow indication goes to 0% in approximately 30 seconds and then rises and stabilizes at approximately 25% due to reverse flow through the loop.

D: INCORRECT: Plausible because there is some reverse flow indicated on the loop. Incorrect becaue it is approximately 25%.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given condition and determine the effect on the RCS parameters.

Page 58 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2096 Last used on an NRC exam: 2011 RO Sequence Number: 30 The Unit tripped from 100% power with the following conditions:

RCS Tave is 567 oF and stable.

Pressurizer pressure is 1737 psig and lowering.

Pressurizer level is 45% and rising.

Containment pressure is 0.1 psig and stable.

PRT pressure is 20 psig and rising.

Which of the following events has likely occurred?

A. Steam Generator feedwater line break outside Containment.

B. Charging flow control valve, FCV-0205, failed open.

C. A Pressurizer PORV has failed open.

D. Steam Generator steamline break outside Containment.

Answer: C A Pressurizer PORV has failed open.

Page 59 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2096 K/A Catalog Number: APE 008 AK2.01 Tier: 1 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(5)

Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following:

Valves.

STP Lesson: LOT 501.21 Objective Number: 501215 Given a set of conditions or event description, be able to PREDICT the sequence of events and trends of plant parameters for a transient or accident involving a decrease in Reactor Coolant Inventory.

Reference:

LOT 501.21 Rev 5 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: a SG feedline break outside of containment would be an overcooling type of event, however RCS temperature is stable, not going down. Additionally, Pzr. level should be lowering if an overcoolig were occurring. Instead, it's going up.

B: INCORRECT: If the charging flow control valve failed open it would normally cause a rise in Pressurizer level. However, based on the given conditions, a Safety Injection has occurred and the charging line has been isolated by Phase 'A' Isolation.

C: CORRECT: Based on RCS pressure lowering with RCS temperature stable, the basic event going on is a loss of coolant and not an overcooling. With a Pressurizer PORV open, a low pressure area exists in the top of the Pzr causing RCS water to expand into the Pzr. raising Pzr. Level. There is no Containment pressure response because the PORV discharges to the PRT.

D: INCORRECT: a SG steamline break outside of containment would be an overcooling type of event, however RCS temperature is stable, not going down. Additionally, Pzr. level should be lowering if an overcooling were occurring. Instead, it's going up.

Question Level: H Question Difficulty 3 Justification:

Student must be able to determine the event that has occurred based on the given plant conditions.

Page 60 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2097 Last used on an NRC exam: 2011 RO Sequence Number: 31 The Unit is operating at 100% power with the following conditions:

A leak in the CCW system develops causing CCW Surge Tank level to lower.

CCW Surge Tank level is currently at 63%.

Based on the given conditions, CCW flow _____(1)_____ been isolated to the RCPs.

Maintaining CCW flow to the RCPs is important to prevent damage to the _____(2)_____ of any operating RCP.

(1) (2)

A. HAS Thermal barrier B. has NOT Motor bearings C. has NOT Thermal Barrier D. HAS Motor bearings Answer: B has NOT; Motor bearings Page 61 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2097 K/A Catalog Number: APE 015/017AK3.01 Tier: 1 Group/Category: 1 RO Importance: 2.5 10CFR

Reference:

55.41(b)(7)

Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

Potential damage from high winding and/or bearing temperatures.

STP Lesson: LOT 201.05 Objective Number: 97119 Given plant conditions, ANALYZE the conditions and accurately PREDICT Reactor Coolant Pump response.

Reference:

LOT 201.05 Rev 17, LOT 201.12 Rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: CCW flow has NOT yet been isolated to the RCP's. Additionally, the RCP Thermal Barriers are not at risk for damage because their CCW cooling is only important if normal seal injection flow is lost.

B: CORRECT: A Surge Tank level of 63% is below the 'first level isolation' so CCW flow has been isolated to some components, but not the RCP's. If CCW Surge Tank level continued to lower, CCW flow to the RCP's would be isolated at a level of 61.5%. If it is isolated and the RCP continues to run, the motor bearings will damaged due to the loss of cooling.

C: INCORRECT: CCW flow has not been isolated to the RCP's, as stated. However, the RCP Thermal Barriers are not at risk for damage because their CCW cooling is only important if normal seal injection flow is lost.

D: INCORRECT: CCW flow has NOT yet been isolated to the RCP's. If it is isolated and the RCP continues to run, the motor bearings will damaged due to the loss of cooling.

Question Level: H Question Difficulty 3 Justification:

Students must be able to determine how the CCW system has been affected based on the current plant conditions and have knowledge of the RCP components that could be potentially damaged if CCW is lost.

Page 62 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2368 Last used on an NRC exam: 2014 RO Sequence Number: 32 An ECO was hung at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on an ECW Pump for maintenance that is expected to last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

As a result of this, which of the following is required in accordance with 0POP01-ZQ-0022, Plant Operations Shift Routines?

A. - Make an entry in the Control Room Logbook OR Operability Assessment System log.

- Update the completed Safety Function Checklist performed at the beginning of shift.

B. - Make an entry in the Control Room Logbook AND Operability Assessment System log.

- Update the completed Safety Function Checklist performed at the beginning of shift.

C. - Make an entry in the Control Room Logbook OR Operability Assessment System log.

- Add the LCO to the Shift Turnover Checklist.

D. - Make an entry in the Control Room Logbook AND Operability Assessment System log.

- Add the LCO to the Shift Turnover Checklist.

Answer: D -Make an entry in the Control Room Logbook AND Operability Assessment System log;

- Add the LCO to the Shift Turnover Checklist Page 63 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2368 K/A Catalog Number: G2.2.36 Tier: 3 Group/Category: 2 RO Importance: 3.1 10CFR

Reference:

55.41(b)(10)

Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations STP Lesson: LOT 507.01 Objective Number: 92184 Given the title of an administrative procedure, identify the actions that are performed by the control room operator.

Reference:

0POP01-ZQ-0022 Rev 75 steps 3.4.1.7 and 6.4.2.2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because both methods are used for tracking and are somewhat redundant, however both are required and the Safety Function Checklist is used to verify the appropriate equipment is available, but it is only performed once per shift (at the beginning) and not updated.

B: INCORRECT: Plausible because the Safety Function Checklist is used to verify the appropriate equipment is available, but it is only performed once per shift (at the beginning) and not updated.

C: INCORRECT: Plausible because both methods are used for tracking and are somewhat redundant, however both are required.

D: CORRECT: Per the referenced procedure, in this situation the CR log, OAS log (if greater than one shift) and turnover checklist must be updated to alllow tracking the activity.

Question Level: H Question Difficulty 3 Justification:

The student must apply procedural requirements to the given situation.

Page 64 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2442 Last used on an NRC exam: Never RO Sequence Number: 33 An on shift crew was performing 0PSP03-RC-0013, Reactor Makeup Water to PZR Relief Tank Check Valve Test, when it was noticed that the surveillance failed to meet the Acceptance Criteria.

The _____(1)_____ will determine Operability status and LCO Action entry requirements and the _____(2)_____ will ensure that a condition report is initiated.

A. (1) Engineering Manager (2) Test Coordinator B. (1) Shift Manager (2) Test Coordinator C. (1) Engineering Manager (2) Plant Surveillance Coordinator D. (1) Shift Manager (2) Plant Surveillance Coordinator Answer: B (1) Shift Manager (2) Test Coordinator Page 65 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2442 K/A Catalog Number: APE 022 G2.2.12 Tier: 1 Group/Category: 1 RO Importance: 3.7 10CFR

Reference:

55.41(b)(10)

Loss of Reactor Coolant Makeup:

Knowledge of Surveillance Procedures.

STP Lesson: LOT 507.01 Objective Number: 92183 Given the title of an administrative procedure, IDENTIFY the individuals (by job title) with specific responsibilities in the procedure.

Reference:

0PGP03-ZE-0004, Plant Surveillance Program Rev 27 page 28 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the Engineering Manager does determine Maintenance Rule issues but not Operability and LCO Action issues.

B: CORRECT: The Shift Manager determines Operability and LCO requirements and the Test Coordinator would ensure a condition report is written.

C: INCORRECT: Plausible because the Engineering Manager does determine Maintenance Rule issues but not Operability and LCO Action issues and the Plant Surveillance Coordinator does implement the Surveillance Program but would not necessarily be present during a surveillance test.

D: INCORRECT: Plausible because the Plant Surveillance Coordinator does implement the Surveillance Program but would not necessarily be present during a surveillance test.

Question Level: F Question Difficulty 3 Justification:

Student must have fundamental knowledge of the Plant Surveillance Procedure.

Page 66 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2446 Last used on an NRC exam: Never RO Sequence Number: 34 Per 0ERP01-ZV-IN03, Emergency Response Organization Notifications, which of the following is the LOWEST Emergency Classification Level at which the ENS Communicator must activate the Emergency Notification Response System (ENRS)?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: A Unusual Event Page 67 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2446 K/A Catalog Number: G2.4.29 Tier: 3 Group/Category: 4 RO Importance: 3.1 10CFR

Reference:

55.41(b)(10)

Knowledge of the emergency plan.

STP Lesson: LOT 507.01 Objective Number: 68900 Maintain required Mode 1 logs, records, charts, printouts and status boards in accordance with 0POP01-ZQ-0022.

Reference:

0ERP01-ZV-SH01, Rev 30, Shift Manager, 0ERP01-ZV-IN03, Rev 18, Emergency Response Organization Notification Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: The ENRS is activated at an Unusual Event even though it is the lowest activation level.

The ENS Communicator, normally a Reactor Operator, is responsible for this E-Plan Duty.

B: INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.

C: INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.

D: INCORRECT: All distractors are plausible because this is a recent duty added for the ENS Communicator which is noramlly a Reactor Operator. It used to be performed by Security. Plus, at the different Emergency Classification Levels, many specific activations of the Emergency Plan are made.

Question Level: F Question Difficulty 2 Justification:

The student must have knowledge of ENRS and when it needs to be activated.

Page 68 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2466 Last used on an NRC exam: Never RO Sequence Number: 35 Per the WOG basis for 0POP05-EO-EO30, Steam Generator Tube Rupture, the PRIMARY reason RCS pressure is equalized with the ruptured SG pressure is to A. minimize inventory loss from the RCS.

B. minimize dose to the public if a relief opens.

C. prevent the Pressurizer from going water solid.

D. prevent voids from forming in the Reactor Vessel head.

Answer: A minimize inventory loss from the RCS.

Page 69 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2466 K/A Catalog Number: EPE 038 EK3.01 Tier: 1 Group/Category: 1 RO Importance: 4.1 10CFR

Reference:

55.41(b)(5)

Knowledge of the reasons for the following responses as they apply to the SGTR:

Equalizing pressure on the primary and secondary sides of ruptured S/G.

STP Lesson: LOT 504.15 Objective Number: 92408 GIVEN a copy of a step from 0POP05-EO-EO30, STATE/IDENTIFY how the action is performed and the basis for the action to include the action itself, its purpose and the result.

Reference:

WOG Background document for SGTR Rev 2 page 144 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: Per the WOG background document the reason for equalizing pressure in the ruptured SG is to reduce primary to secondary leakage.

B: INCORRECT: Plausible because this is a secondary reason for equalizing pressure and is the reason for adjusting the PORV setpoints. Incorrect because it is not the PRIMARY reason.

C: INCORRECT:Plausible because refilling the PZR is part of depressurizing the RCS but overfiling the PZR is a reason for stopping the depressurization early.

D: INCORRECT: Plausible because it is desired to not have voids in the Reactor Vessel Head which could come from depressurizing the RCS but depending on plant conditions this could be an unavoidable consequence. A separate EOP would be used for this condition.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of the reasons for performing steps in EOPs.

Page 70 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2488 Last used on an NRC exam: Never RO Sequence Number: 37 The crew is responding to a break on the CVCS charging line inside containment. The CVCS charging OCIV MOV-0025 did not fully close from the control room and must be closed locally.

Using the provided RWP and Survey Map which of the following is correct?

The charging line OCIV MOV-0025 is located in a (1) and entry is (2) per the RWP.

A. (1) High Radiation Area ONLY (2) NOT allowed B. (1) High Radiation Area and High Contaminated Area (2) NOT allowed C. (1) Radiation Area ONLY (2) allowed D. (1) Radiation Area and Contaminated Area (2) allowed Answer: A (1) High Radiation Area ONLY (2) NOT allowed Page 73 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2488 K/A Catalog Number: G2.3.7 Tier: 3 Group/Category: 3 RO Importance: 3.5 10CFR

Reference:

55.41(b)(12)

Ability to comply with radiation work permit requirements during normal or abnormal conditions.

STP Lesson: LOT 507.01 Objective Number: 92186 Given the title of an administrative procedure, DISCUSS the requirements associated with the referenced procedure.

Reference:

LOT 507.01 Lesson on Administrative Procedures - 0PGP03-ZR-0051, Radiological Access Controls/Standards Attached Reference

Attachment:

LOT 20.1 RO Q2488 RWP and survey map NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: Using the survey map provided, CV-MOV-0025 is located in a high rad area therefore entry with the given RWP is not allowed.

B: INCORRECT: All distractors are plausible because a valv eis located in each area on the provided survey map. The student must know where MOV-0025 is located to assess the conditions and determine entry conditions per the RWP.

C: INCORRECT: All distractors are plausible because a valv eis located in each area on the provided survey map. The student must know where MOV-0025 is located to assess the conditions and determine entry conditions per the RWP.

D: INCORRECT: All distractors are plausible because a valv eis located in each area on the provided survey map. The student must know where MOV-0025 is located to assess the conditions and determine entry conditions per the RWP.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given conditions and survey map to determine RWP requirements.

Page 74 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2601 Last used on an NRC exam: Never RO Sequence Number: 38 Which of the following procedures takes precedence over all Emergency Operating Procedures?

A. 0POP05-EO-FRS1, Response to Nuclear Generation - ATWS B. 0POP05-EO-EC00, Loss of All AC Power C. 0POP05-EO-EO00, Reactor Trip or Safety Injection D. 0POP05-EO-ES13, Transfer To Cold Leg Recirculation Answer: B 0POP05-EO-EC00, Loss of All AC Power Page 75 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2601 K/A Catalog Number: G2.4.6 Tier: 3 Group/Category: 4 RO Importance: 3.7 10CFR

Reference:

55.41(b)(10)

Knowledge of EOP Mitigation Strategies STP Lesson: LOT 504.04 Objective Number: 92283 Given a set of conditions and the occurrence of a red, orange, or yellow path CSF, state the action required per 0POP01-ZA-0018, EOP Users Guide.

Reference:

0POP01-ZA-0018 Rev 21 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because it is the highest priority CSF. Incorrect because 0POP05-EO-EC00 takes priority over FRPs.

B: CORRECT: 0POP05-EO-EC00 takes precedence over the FRPs and all other EOPs due to specific initiating events per 0POP01-ZA-0018.

C: INCORRECT: Plausible because it is the first procedure you would enter aftera reactor trip if there was not a complete loss of all AC power. Incorrect because 0POP05-EO-EC00 takes priority D: INCORRECT: Plausible because it takes priority over all other EOPs if swap over conditions exist.

Incorrect, because entry into 0POP05-EO-EC00 SHALL be entered during the performance of ANY other EOP.

Question Level: F Question Difficulty 3 Justification:

The student needs to have fundamental knowledge of Emergency procedure Mitigation Strategies.

Page 76 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2252 Last used on an NRC exam: 2014 RO Sequence Number: 39 When heating up the Secondary Plant and performing a Plant Startup, the operating procedures have cautions about avoiding Hydraulic Transients (Water Hammer) when operating the Main Steam and Reheat Steam Systems.

Which of the following describes a cause and definition of a Hydraulic Transient?

CAUSE DEFINITION A. During a manual Cold Start of MSRs, An increase in steam demand with a rapidly initiating Main Steam through the resultant pressure reduction.

Reheat Control Valves.

B. When at NOP/NOT, opening a Main Steam An increase in steam demand with a Isolation Valve with downstream pressure resultant pressure reduction.

60 psig lower.

C. During a manual Cold Start of MSRs, The shock imposed on piping from rapidly initiating Main Steam through the initiating steam flow through pipes Reheat Control Valves. containing liquid condensate.

D. When at NOP/NOT, opening a Main Steam The shock imposed on piping from Isolation Valve with downstream pressure initiating steam flow through pipes 60 psig lower. containing liquid condensate.

Answer: C During a manual Cold Start of MSRs, rapidly initiating Main Steam through the Reheat Control Valves. - The shock imposed on piping from initiating steam flow through pipes containing liquid condensate.

Page 77 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2252 K/A Catalog Number: 039 K5.01 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(5)

Knowledge of the operational implications of the following concepts as they apply to the MRSS:

Definition and causes of steam/water hammer.

STP Lesson: LOT 102.57 Objective Number: N99862 Explain operational implications of water (fluid) hammer.

Reference:

LOT 102.57 Lesson Plan Handout Page 46 and Procedures 0POP03-ZG-0003 Rev 37, Secondary Plant Startup, and 0POP02-MS-0001 Rev 54, Main Steam System.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because it describes the definition of SWELL which is related to thermodynamic processes.

B: INCORRECT: This is a credible distractor because it decribes a condition in 0POP03-ZG-0003, where opening the MSIV with a differential pressure of greater than 50 psig can cause the phenomenom of SWELL in the corresponding SG but at the given differential of 60 psig it would not cause water hammer. Also, it describes the definition of SWELL which is related to thermodynamic processes.

C: CORRECT: Manually rapidly opening an MSR Reheat Control Valve during a Cold Start (described in 0POP02-MS-0001) can cause water hammer and thereby system damage. Procedure requires valves to be throttled to raise temperature no greater than 100 degrees F per hour. The definition of water hammer includes the shock imposed on piping from initiating steam flow through pipes containing liquid condensate.

D: INCORRECT: This is a credible distractor because it decribes a condition in 0POP03-ZG-0003, where opening the MSIV with a differential pressure of greater than 50 psig can cause the phenomenom of SWELL in the corresponding SG but at the given differential of 60 psig it would not cause water hammer.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of thermodynamics and procedures covering Secondary Plant Startup from cold conditions to 100% power.

Page 78 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2305 Last used on an NRC exam: Never RO Sequence Number: 40 The Unit was at 100% power when the following occurred:

A Steam Line break on SG 1A in the IVC up stream of the Main Steam Isolation Valve.

One (1) Steam Pressure Transmitter on SG 1A is damaged causing it to read off-scale HIGH on Control Board and QDPS indications.

The damaged Steam Pressure Transmitter was selected for control with the associated Steam Flow Transmitter for SG Water Level Control.

(1) The Controlling Channel of Steam Flow will indicate _____(1)_____ than actual steam flow?

AND (2) Will a Main Steam Isolation (MSI) automatically occur if required?

(1) (2)

A. HIGHER NO B. LOWER YES C. LOWER NO D. HIGHER YES Answer: D Indicated Steam Flow will read HIGHER than actual Steam Flow. - If needed, an MSI will AUTOMATICALLY actuate.

Page 79 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2305 K/A Catalog Number: APE 040 AK2.02 Tier: 1 Group/Category: 1 RO Importance: 2.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the interrelations between the Steam Line Rupture and the following:

Sensors and detectors.

STP Lesson: LOT 202.02 Objective Number: 12768 Given a plant or system condition, PREDICT the operation of the Main Steam System.

Reference:

LOT 202.02 Lesson Plan Rev 11 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because the steam pressure channel did fail high but when the other two steam pressure channels lower to the actuation setpoint then an MSI will automatically actuate.

B: INCORRECT: This distractor is credible because the failed high steam pressure channel does affect the associated steam flow channel but the pressure compensation causes the steam flow channel to read higher than actual steam flow.

C: INCORRECT: This distractor is credible because the failed high steam pressure channel does affect the associated steam flow channel but the pressure compensation causes the steam flow channel to read higher than actual steam flow. Also, the steam pressure channel did fail high but when the other two steam pressure channels lower to the actuation setpoint then an MSI will automatically actuate.

D: CORRECT: With a high failure of a controlling steam pressure channel, the controlling steam flow channel will indicate higher than actual steam flow due to the pressure compensation. One of the three channels of Steam Pressure feeding the MSI actuation will not affect the ability for an automatic actuation on 2 of the remaining 3 Steam Pressure Channels.

Question Level: H Question Difficulty 3 Justification:

The student must be able to evaluate the given conditions to determine the affects on indicated steam flow and MSI actuation.

Page 80 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2321 Last used on an NRC exam: Never RO Sequence Number: 41 Which of the following is:

(1) The design feature that will prevent the Pressurizer (PZR) Heaters from being uncovered during an instantaneous 10% addition of turbine load?

AND (2) The PZR level at which the PZR Heaters will FIRST de-energize?

(1) (2)

A. Design Water and Steam Volume of 17% PZR Level the PZR.

B. Design Water and Steam Volume of 8% PZR Level the PZR.

C. Design Control Response of Charging 17% PZR Level Flow Control Valve CV-FV-0205.

D. Design Control Response of Charging 8% PZR Level Flow Control Valve CV-FV-0205.

Answer: A Design Water and Steam Volume of the PZR. - 17% PZR Level Page 81 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2321 K/A Catalog Number: 010 K4.02 Tier: 2 Group/Category: 1 RO Importance: 3.0 10CFR

Reference:

55.41(b)(7)

Knowledge of the PZR PCS design feature(s) and/or interlock(s) which provide for the following:

Prevention of uncovering PZR heaters.

STP Lesson: LOT 201.04 Objective Number: 91011 DESCRIBE the basis for the sizing of the Pressurizer and the PRT.

Reference:

LOT 201.04 Rev 8 Lesson Plan on PZR, PRT and RCDT. Slide #28 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: The design water and steam volume of the PZR will prevent uncovering the PZR Heaters on a 10% load increase and the PZR Heaters will de-energize at 17% PZR level.

B: INCORRECT: This distractor is credible because the PZR level given (8%) is used in many of the emergency procedures to direct the operator to restart SI pumps and/or initiate SI.

C: INCORRECT: This distractor is credible because the charging flow control valve will open on a lowering PZR level but it is not sized to prevent uncovering the PZR Heaters on a step load increase of 10%.

D: INCORRECT: This distractor is credible because the charging flow control valve will open on a lowering PZR level but it is not sized to prevent uncovering the PZR Heaters on a step load increase of 10%. Also, the PZR level given (8%) is used in many of the emergency procedures to direct the operator to restart SI pumps and/or initiate SI.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of PZR Pressure and Level Control system design features and control interlocks.

Page 82 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2323 Last used on an NRC exam: Never RO Sequence Number: 42 The Unit is at 100% power with the following conditions:

The Component Cooling Water (CCW) MODE SEL switches are in the following position:

o Train A is in RUN o Train B is in AUTO o Train C is in OFF CCW Pumps Control Room handswitches are as follows:

o CCW Pump 1A is in AUTO and is running.

o CCW Pump 1B and 1C are in AUTO and are NOT running.

Reactor Containment Fan Cooler (RCFC) Control Room handswitches are as follows:

o RCFCs Trains A and B are in AUTO and running with RCB Chill Water aligned.

o RCFC Train C is in AUTO and NOT running with RCB Chill Water aligned.

Subsequently:

A Loss of Offsite Power occurs.

Which of the following manual actions will the Control Room Operator have to perform in order to establish all three trains of cooling to Containment?

Manually A. START CCW Pump 1C.

B. START RCFCs on Train C C. CLOSE the RCB Chill Water to RCFC Motor Operated isolation valves.

D. OPEN the CCW to RCFC Motor Operated isolation valves.

Answer: D OPEN the CCW to RCFC Motor Operated isolation valves.

Page 83 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2323 K/A Catalog Number: 022 A4.04 Tier: 2 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(7)

Ability to manually operate and/or monitor in the control room:

Valves in the CCS.

STP Lesson: LOT 201.12 Objective Number: 5213 GIVEN a plant or system condition, PREDICT the operation of the Component Cooling Water System.

Reference:

LOT 201.12 Lesson Plan PPT Rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because even though the MODE SEL switch on Train C is in the 'OFF' position the CCW Pump 1C will still get a start signal.

B: INCORRECT: This distractor is credible because even though the MODE SEL switch on Train C is in the 'OFF' position the RCFCs on Train C will still get a start signal.

C: INCORRECT: This distractor is credible because even though the CCW to RCFCs Motor Operated isolation valves do not get a signal to auto open, the RCB Chill Water to RCFCs Motor Operated isolation valves do get an Auto close signal.

D: CORRECT: The CCW to RCFC Motor Operated isolation valves do not get a signal to auto open and must be manually opend to establish containment cooling on all 3 trains.

Question Level: H Question Difficulty 3 Justification:

The student must be able to evaluate the given conditions to determine which manual action needs to be performed.

Page 84 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2494 Last used on an NRC exam: Never RO Sequence Number: 43 Plant procedures put a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit on loading over 6050kw when running an ESF Diesel Generator.

Running the Diesel Generator overloaded could result in a Generator _____(1)_____ condition causing the Diesel Generator to trip during _____(2)_____ mode of operation.

A. (1) DIFFERENTIAL (2) NORMAL OR EMERGENCY B. (1) OVERCURRENT (2) ONLY NORMAL C. (1) DIFFERENTIAL (2) ONLY NORMAL D. (1) OVERCURRENT (2) NORMAL OR EMERGENCY Answer: B (1) OVERCURRENT (2) ONLY NORMAL Page 85 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2494 K/A Catalog Number: 062 A1.01 Tier: 2 Group/Category: 1 RO Importance: 3.4 10CFR

Reference:

55.41(b)(7)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

Significance of D/G load limits STP Lesson: LOT 201.39 Objective Number: 45057 STATE the Emergency Diesel Generator trips in the emergency mode and in the test mode

Reference:

LOT 201.39 Lesson Plan and 0POP09-AN-0102, B-8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the generator differential is similar to the generator overcurrent in that it trips the DG and opens the output breaker but the generator differential would be indicative of generator damage not necessarily caused by overloading the generator. Also the generator overcurrent trip is only active in the normal mode.

B: CORRECT: Overloading the DG would cause a generator overcurrent trip and is only active in the normal mode. In addition this condition would likely only happen while the DG is paralled to offsite power. When paralled to offsite power the DG is running in the normal mode.

C: INCORRECT: Plausible because the generator differential is similar to the generator overcurrent in that it trips the DG and opens the output breaker but the generator differential would be indicative of generator damage not necessarily caused by overloading the generator.

D: INCORRECT: Plausible, however, the generator overcurrent trip is only active in the normal mode.

Question Level: F Question Difficulty 3 Justification:

The applicant must have knowledge of the trips availables during the different modes of ESF DG operation and now the difference between a Generator Overcurrent and a Generator Differential Trip.

Page 86 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2332 Last used on an NRC exam: Never RO Sequence Number: 44 The Unit is at 100% power with the following conditions:

A fuel handling accident has occurred in the FHB involving a spent fuel assembly.

FHB radiation levels are rising.

A high alarm has been received on RT-8035 and RT-8036 resulting in the required FHB HVAC actuation.

(1) Which of the following identifies the expected trend for RT-8035 and RT-8036?

AND (2) What action must the operators perform in accordance with 0POP04-FH-0001, Fuel Handling Accident?

(1) (2)

Check only one train of FHB Exhaust Fans A. Continue to rise.

(Main and Booster) in operation.

Place Control Room Envelope HVAC in B. Continue to rise.

Emergency Mode.

Check only one train of FHB Exhaust Fans C. Begin to lower.

(Main and Booster) in operation.

Place Control Room Envelope HVAC in D. Begin to lower.

Emergency Mode.

Answer: B Continue to rise; Place Control Room Envelope HVAC in Emergency Mode Page 87 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2332 K/A Catalog Number: 073 A1.01 Tier: 2 Group/Category: 1 RO Importance: 3.2 10CFR

Reference:

55.41(b)(11)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: Radiation levels STP Lesson: LOT 202.38 Objective Number: 2207 LIST all the systems that interface with the Fuel Handling HVAC System and state the function of each interface

Reference:

0POP04-FH-0001 Rev 19 steps 19 and 24 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because the POP04 has the operator check only 1 filter train in service (and 2 sets of fans).

B: CORRECT: Since the rad monitors are upstream of the filters, rad levels will continue to rise. The POP04 requires operators to place CRE HVAC in emergency mode.

C: INCORRECT: Credible because the high rad alarm will put filters in service which will lower the rad levels in the HVAC exhaust, so the applicant must understand system design to determine indicated rad will still rise. Action is credible because the POP04 has the operator check only 1 filter train in service (and 2 sets of fans).

D: INCORRECT: Credible because the high rad alarm will put filters in service which will lower the rad levels in the HVAC exhaust, so the applicant must understand system design to determine indicated rad will still rise.

Question Level: H Question Difficulty 3 Justification:

The applicant must evaluate the given information and use knowledge of system design to determine the proper trend. Knowledge of procedural requirements is also required.

Page 88 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2333 Last used on an NRC exam: Never RO Sequence Number: 45 The Unit is at 100% power with the following condition:

ESF D/G #11 is paralleled to the ESF 4.16 KV BUS E1A to support 0PSP03-DG-0001, Standby Diesel 11(21) Operability Test.

Subsequently:

A Loss of Offsite Power (LOOP) occurred.

ESF DG #13 failed to start and cannot be manually started.

If needed, which component(s) listed below would NOT have auto started but could be manually started?

A. BOTH Centrifugal Charging Pumps 1A and 1B B. ONLY Centrifugal Charging Pump 1A C. BOTH HHSI Pumps 1A and 1B D. ONLY HHSI Pump 1B Answer: C BOTH HHSI Pumps 1A and 1B Page 89 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2333 K/A Catalog Number: 064 K3.03 Tier: 2 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:

ED/G (manual loads)

STP Lesson: LOT 201.41 Objective Number: 98035 Given a plant or system condition, predict the operation of the ESF Load Sequencer.

Reference:

LOT 201.41 Lesson Plan PPT on ESF Load Sequencers Rev 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because Centrifugal Charging pumps do not Auto start on a Mode II signal but CCP 1A is powered from Train C and will not be able to be manually started either.

B: INCORRECT: This distractor is credible because Centrifugal Charging pumps do not Auto start on a Mode II signal but CCP 1A is powered from Train C and will not be able to be manually started either.

C: CORRECT: A LOOP will cause the ESF DG Sequencers to start on a Mode II. However, a Mode II signal will not Auto start the HHSI Pumps. Therefor HHSI Pump 1A and 1B would have to be manualy started if needed.

D: INCORRECT: This distractor is credible because other surviellances use SI signals to emergency start the ESF DG for testing and if the student believed that a Mode I (SI) was present on Train A when the Mode II (LOOP) occurred then only HHSI Pump 1B would be manually started if needed.

Question Level: H Question Difficulty 3 Justification:

The student must be able to evaluate the given conditions to determine which components, if needed, did not auto start and could be manually started.

Page 90 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2439 Last used on an NRC exam: Never RO Sequence Number: 46 After a Reactor Trip from 100% power, indicated Reactor power will lower to _____(1)_____

power in 2 to 3 seconds and then lower to a subcritical equilibrium level at a rate of

_____(2)_____ decades per minute.

A. (1) 6%

(2) - 0.3 B. (1) 10%

(2) - 0.1 C. (1) 6%

(2) - 0.1 D. (1) 10%

(2) - 0.3 Answer: A (1) 6%

(2) - 0.3 Page 91 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2439 K/A Catalog Number: EPE 007 EK1.04 Tier: 1 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(1)

Knowledge of the operational implications of the following concepts as they apply to the reactor trip:

Decrease in reactor power following reactor trip (prompt drop and subsequent decay).

STP Lesson: LOT 101.25 Objective Number: N99751 Explain the shape of the curve of reactor power versus time after a reactor trip.

Reference:

Fundamental Lesson Plan LOT 101.25 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: After a reactor trip power makes a prompt drop to about 6% at which time delayed neutrons slow the decrease at a calculated rate of minus 0.3 decades per minute.

B: INCORRECT: Plausible because 10% reactor power is a basis for other reactor limits such as when certain reactor trips need to be blocked/activated. (6% power is close to the 5% basis for the size of Aux Feedwater.) Also, a minus 0.1 decades per minute lowering of reactor power would be indicative of a failed intermediate range instrument which would require manually energizing source range instruments. (A similar event happened on a recent reactor trip in Unit 1 CR-16-1227)

C: INCORRECT: Plausible because a minus 0.1 decades per minute lowering of reactor power would be indicative of a failed intermediate range instrument which would require manually energizing source range instruments. (A similar event happened on a recent reactor trip in Unit 1 CR-16-1227)

D: INCORRECT: Plausible because 10% reactor power is a basis for other reactor limits such as when certain reactor trips need to be blocked/activated. (6% power is close to the 5% basis for the size of Aux Feedwater.)

Question Level: H Question Difficulty 2 Justification:

Stundent must analyze the condition given and have the fundamental concept of how reactor power reacts after a reactor trip.

Page 92 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2440 Last used on an NRC exam: Never RO Sequence Number: 47 A Reactor Trip and Safety Injection have occurred due to a Small Break LOCA with the following conditions:

All Reactor Coolant Pumps (RCPs) are running.

HHSI Pumps A and B are running. HHSI Pump C failed to start due to an overcurrent.

All LHSI Pumps are secured and in AUTO.

Subsequently:

HHSI Pump B trips on overcurrent.

RCS Pressure lowers to 390 psig and stabilizes.

Per 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, RCPs should be

______(1)______ and LHSI Pumps should be _____(2)______.

A. (1) running (2) secured in AUTO B. (1) secured (2) secured in AUTO C. (1) running (2) running D. (1) secured (2) running Answer: D secured - running Page 93 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2440 K/A Catalog Number: EPE 011 EA2.01 Tier: 1 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(10)

Ability to determine or interpret the following as they apply to a Large Break LOCA:

Actions to be taken, based on RCS temperature and pressure - saturated and superheated STP Lesson: LOT 504.09 Objective Number: 81103 From memory, STATE/IDENTIFY the criteria on the conditional information page of POP05-EO-EO10 to include operator response, initiating parameter(s) and values.

Reference:

CIP of 0POP05-EO-EO10 Rev 22 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because if no HHSI Pumps are running then the RCPs would be required to stay running and even though RCS pressure has stabilized at about LHSI Pump shut off head the procedure requires the pumps to be started if pressure drops below 415 psig.

B: INCORRECT: Plausible because even though RCS pressure has stabilized at about LHSI Pump shut off head the procedure requires the pumps to be started if pressure drops below 415 psig.

C: INCORRECT: Plausible because if no HHSI Pumps are running then the RCPs would be required to stay running.

D: CORRECT: With the given conditions the RCPs would be required to be manually secured and the LHSI Pumps would be manually started.

Question Level: H Question Difficulty 3 Justification:

Student must be able to analyze the given conditions to determine the correct answer.

Page 94 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2441 Last used on an NRC exam: Never RO Sequence Number: 48 During a SBLOCA the following parameters are given:

RCS pressure is indicating 1650 psig.

PZR pressure is 1700 psig.

RCS Hot Leg temperatures are indicating 590ºF.

Max Quad TC Avg temperature is indicating 600ºF.

Which of the following would be the correct indication for subcooling from QDPS?

A. 10ºF B. 14ºF C. 20ºF D. 24ºF Answer: A 10 degrees F Page 95 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2441 K/A Catalog Number: EPE 009 EA1.16 Tier: 1 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(7)

Ability to operate and monitor the following as they apply to a small break LOCA:

Subcooling margin monitors STP Lesson: LOT 202.44 Objective Number: 91674 List the plant systems/components controlled by QDPS.

Reference:

QDPS Lesson Plant LOT 202.44 Rev 12 slide 93 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: QDPS uses Max Quad TC average and RCS pressure to calculate subcooling.

B: INCORRECT: Plausible because answer reflects using PZR pressure in the calculation instead of RCS pressure.

C: INCORRECT: Plausible because answer reflects using Hot Leg temperatures in the calculation instead of Max Quad TC Avg.

D: INCORRECT: Plausible because answer reflects using Hot Leg temperatures in the calculation instead of Max Quad TC Avg and reflects using PZR pressure in the calculation instead of RCS pressure.

Question Level: H Question Difficulty 3 Justification:

Student must be able to know how subcooling is calcualted using QDPS.

Page 96 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2444 Last used on an NRC exam: Never RO Sequence Number: 49 The Unit was at 100% power when a SI occurred due to a LOCA in Containment with the following condition:

Containment pressure is at 7 psig and rising.

Subsequently:

The normal feeder breaker to Train B 4.16 KV ESF bus tripped open causing a loss of power on the bus.

AT THE SAME TIME that the Train B Diesel Generator output breaker closed in, re-energizing the bus, a Containment Spray (CS) Actuation Signal was generated from containment pressure.

(1) Which of the following states the times at which Train B CS Pump started?

AND (2) Which of the following states when the Train B CS Pump discharge valve started to open?

(1) (2)

A. Immediately 1 second B. Immediately 15 seconds C. 15 seconds 15 seconds D. 15 seconds 1 second Answer: D 15 seconds, 1 second Page 97 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2444 K/A Catalog Number: 026 A3.01 Tier: 2 Group/Category: 1 RO Importance: 4.3 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic action of the CSS, including:

Pump starts and correct MOV positioning.

STP Lesson: LOT 201.11 Objective Number: 2009 GIVEN a plant or system condition, PREDICT the operation of the Containment Spray System.

Reference:

LOT 201.11 Lesson Plan Containment Spray Rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the CS Actuation Signal does start the CS Pumps but in this case since the ESF D/G just energized the BUS there is a 15 second time delay before the CS Pump gets a start signal.

B: INCORRECT: Plausible because the CS Actuation Signal does start the CS Pumps but in this case since the ESF D/G just energized the BUS there is a 15 second time delay before the CS Pump gets a start signal. Also many pumps in the plant have their discharge valves start to open when a pump starts but in this case the CS Pump discharge valves start to open on the CS Actuation signal.

C: INCORRECT: Plausible because many pumps in the plant have their discharge valves start to open when a pump starts but in this case the CS Pump discharge valves start to open on the CS Actuation signal.

D: CORRECT: Per the given conditions the CS Pump would start 15 seconds after the ESF BUS is energized but the CS Discharge valve would start to open on the CS Actuation signal. (1 second is for load center breakers to close after the DG output breaker closes)

Question Level: H Question Difficulty 3 Justification:

The candidate must have a knowledge of the conditions required to actuate both the spray pumps and discharge valves. This knowledge must then be applied to the conditions given to determine the correct response.

Page 98 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2588 Last used on an NRC exam: Never RO Sequence Number: 50 The CCW surge tank normal makeup valve LV-4501 can be operated (1) .

AND The CCW SURGE TK LVL LO annunciator will be (2) IF the level is at 65%.

A. (1) locally OR from the control room (2) extinguished B. (1) ONLY locally (2) extinguished C. (1) locally OR from the control room (2) lit D. (1) ONLY locally (2) lit Answer: C (1) locally OR from the control room - (2) lit Page 99 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2588 K/A Catalog Number: APE 026 AA1.05 Tier: 1 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(7)

Ability to operate and/or monitor the following as they apply to the Loss of Component Cooling Water:

The CCWS surge tank, including level control and level alarms, and radiation alarm.

STP Lesson: LOT 201.12 Objective Number: 80198 DESCRIBE the Instrumentation and Controls available to monitor and operate the CCW System.

Reference:

LOT 201.12 Rev 14 Slide 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the first low level isolation does not occur until 64.6%. Incorrect because the alarm comes in at 66.7%

B: INCORRECT: Plausible because the first low level isolation does not occur until 64.6%. Also plausible because the surge tank alternate makeup valve is operated locally ONLY. Also there are several tanks with makeup valves that can not be operated from the control room (i.e. AFW storage tank)

C: CORRECT: The makeup valve can be operated from both the control room and locally. The low surge tank level alarm comes in at 66.7%.

D: INCORRECT: Plausible because the surge tank backup makeup valve is operated locally ONLY.

Also there are several tanks with makeup valves that can not be operated from the control room (i.e. AFW storage tank) Incorrect because LV-4501 can be operated from both the control room and locally.

Question Level: F Question Difficulty 3 Justification:

Student must have knowledge of the CCW system alarms and controls.

Page 100 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2467 Last used on an NRC exam: Never RO Sequence Number: 51 The Unit was at 100% power when a control rod dropped into the core to the Rod Bottom position.

The Power Range NIs now read the following:

NI-0041B 98%

NI-0042B 98%

NI-0043B 97%

NI-0044B 98%

The control rod that dropped into the core was at the _____(1)_____ of the core.

AND As the dropped control rod is being recovered its Differential Rod Worth will be _____(2)_____

as it approaches the mid-plane of the core.

A. (1) center (2) less B. (1) edge (2) less C. (1) center (2) more D. (1) edge (2) more Answer: C (1) center - (2) more Page 101 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2467 K/A Catalog Number: APE 003 AK1.19 Tier: 1 Group/Category: 2 RO Importance: 2.8 10CFR

Reference:

55.41(b)(1)

Knowledge of the operational implications of the following concepts as they apply to Dropped Control Rods:

Differential Rod Worth.

STP Lesson: LOT 101.22 Objective Number: N99697 Define control rod worth, differential rod worth and intergral rod worth.

Reference:

LOT 101.22 Lesson Handout on Control Rod Worths Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.

B: INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.

C: CORRECT: With a single control rod dropping into the center of the core the power range NIs would equally lower. As the control rod is being recovered the DRW would be more as it approched the mid-plane of the core.

D: INCORRECT: All distractors are plausible because the student must have an understanding of how the Power Range NIs respond to a control rod dropped into the center of the core vs one dropped into the edge of the core. If the control rod that dropped was at the edge of the core then the Power Range NI closest to the dropped control rod would be significantly lower than the rest of the NIs by as much as 10%. The student must alos have fundamental knowledge of how DRW changes with rod height.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given NI power range readings to determine what part of the core the control rod dropped and have fundamental knowledge of control rod worths for control rods at different locations/heights in the core.

Page 102 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2468 Last used on an NRC exam: Never RO Sequence Number: 52 During a fast load reduction from 100% power the Control Rods were placed in MANUAL and the following DRPI indication was observed:

Control Rod(s) _____(1)_____ is(are) misaligned. The crew will _____(2)_____.

A. (1) D4 and M12 (2) place the Unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. (1) D4 and M12 (2) trip the Reactor C. (1) D4 ONLY (2) place the Unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. (1) D4 ONLY (2) trip the Reactor Answer: A (1) D4 and M12 - (2) place the Unit in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Page 103 of 150

Print Date 4/25/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2468 K/A Catalog Number: APE 005 AA1.05 Tier: 1 Group/Category: 2 RO Importance: 3.4 10CFR

Reference:

55.41(b)(6)

Ability to operate and/or monitor the following as they apply to the Inoperable/Stuck Control Rod:

RPI STP Lesson: LOT 201.19 Objective Number: 93001 Given a plant or system condition, PREDICT the operation of the Rod Position Indication System.

Reference:

LOT 201.19 Lesson Plan, Rev 12, and 0POP04-RS-0001, Rev 15 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: Control Rods D4 and M12 are both misaligned from the other control rods by more than 12 steps. Per 0POP04-RS-0001, the Unit will be placed in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B: INCORRECT: Plausible because for 2 dropped control rods the crew would trip the reactor.

C: INCORRECT: Plausible because the indicated positions of Control Rods D4 and M12 are different and the student must have the knowledge of how many steps different from other control rods would be considered a misalignment.

D: INCORRECT: Plausible because the indicated positions of Control Rods D4 and M12 are different and the student must have the knowledge of how many steps different from other control rods would be considered a misalignment. Also the student must have knowledge of procedural requirements.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition to determine which Control rods are misaligned and the required procedural action to take.

Page 104 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2470 Last used on an NRC exam: Never RO Sequence Number: 53 The Unit is at 100% power with the following conditions:

CVCS Cation Bed 1A was used 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago for RCS lithium control.

Subsequently:

An Alert Alarm is received on Failed Fuel Monitor RT-8039.

The Crew is responding per 0POP04-RC-0001, High Reactor Coolant System Activity.

Chemistry has requested that Cation Bed 1A be placed back in service to help control RCS activity.

Failed Fuel Monitor RT-8039 readings can be validated by _____(1)_____.

AND A consequence of placing Cation Bed 1A in service is that _____(2)_____ would lower.

(1) (2)

A. obtaining a current RCS sample RCS pH B. obtaining a current RCS sample boron concentration C. trending the GWPS inlet Radiation Monitor RCS pH D. trending the GWPS inlet Radiation Monitor boron concentration Answer: A obtaining a current RCS sample - RCS pH Page 105 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2470 K/A Catalog Number: APE 076 G2.2.44 Tier: 1 Group/Category: 2 RO Importance: 4.2 10CFR

Reference:

55.41(b)(7)

High Reactor Coolant Activity:

Ability to interpret control room indications to verify the status of a system and understand how operator actions and directives affect plant and system conditions.

STP Lesson: LOT 505.01 Objective Number: 92109 GIVEN a plant condition, DESCRIBE and/or INTERPRET the reqirements and/or limits of a precaution or step of a referenced procedure.

Reference:

0POP04-RC-0001, High RCS Activity Rev 11 page 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: Per 0POP04-RC-0001 an alternate method for verifing High RCS activity when RT-8039 is trending upward is to get a current RCS sample. A caution in the procedure also warns the operator that placing a CVCS Cation bed in service will lower RCS lithium thus lowering RCS pH and possibly cause a Crud Burst.

B: INCORRECT: Plausible because placing a CVCS Cation Bed in service can lower boron concentration, however, it was stated in the conditions of the question that the CVCS Cation Bed was placed in service 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago which indicates that its boron concentration is equal to the RCS.

C: INCORRECT: Plausible because in 0POP04-RC-0001 GWPS is monitored because VCT purge flow rate is raised but GWPS is not monitored from a stand point of verifing RCS activity.

D: INCORRECT: Plausible because in 0POP04-RC-0001 GWPS is monitored because VCT purge flow rate is raised but GWPS is not monitored from a stand point of verifing RCS activity. Also, placing a CVCS Cation Bed in service can lower boron concentration, however, it was stated in the conditions of the question that the CVCS Cation Bed was placed in service 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago which indicates that its boron concentration is equal to the RCS.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given conditions to determine what affect an action will have on the RCS and have knowledge of alternate ways to verify high RCS activity.

Page 106 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2501 Last used on an NRC exam: Never RO Sequence Number: 54 Which of the following 480 Volt Motor Control Centers (MCC) supplies power to Boric Acid Transfer Pump 1B?

A. MCC E1A4 B. MCC E1C4 C. MCC 1G8 D. MCC 1K3 Answer: A MCC E1A4 Page 107 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2501 K/A Catalog Number: 004 K2.01 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Knowledge of bus power supplies to the following:

Boric acid makeup pumps.

STP Lesson: LOT 201.07 Objective Number: 91054 DESCRIBE the boric acid pumps to include: A. Function and locartion B. Pump flow and pressure capacity C. Electrical power supply D. Pump controls and meaning of associated alarms.

Reference:

LOT 201.07 Lesson Plan on Reactor Makeup. Rev 13 slide 49 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: The Boric Acid Transfer Pumps power supply comes from Class 1E 480 Volt MCCs.

BAT Pump 1B is powered from MCC E1A4. (NOTE: Same Train of power as CCP 1B)

B: INCORRECT: Plausible because the power for the Boric Acid pumps is not Train specific and MCC E1C4 is the power for BAT Pump 1A.

C: INCORRECT: Plausible because some safety significant equipment is powered from NC MCC 1G8 like the Positive Displacement Charging Pump. Plus MCC 1G8 is backed by the TSC DG.

D: INCORRECT: Plausible because some some safety support equipment is powered from NC MCC 1K3 like the RWST Purification Pump. Plus MCC 1K3 is physically close to the Boric Acid Transfer Pumps.

Question Level: F Question Difficulty 2 Justification:

The student has to have fundamental knowledge of power supplies for safety significant and safety related equipment.

Page 108 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2473 Last used on an NRC exam: Never RO Sequence Number: 55 During a plant heatup, which of the following represents the HIGHEST ALLOWABLE flow rate through one CVCS demineralizer?

A. 350 gpm B. 300 gpm C. 250 gpm D. 200 gpm Answer: C 250 gpm Page 109 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2473 K/A Catalog Number: 004 A4.14 Tier: 2 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(7)

Chemical and Volume Control:

Ability to manually operate and/or monitor in the control room:

Ion exchangers and demineralizers.

STP Lesson: LOT 201.06 Objective Number: 70033 STATE the maximum specified purification flow rates through the CVCS System Demineralizers including the reasons for the limit.

Reference:

LOT 201.06 Lesson Plan for CVCS and 0POP03-ZG-0001, Plant Heatup Rev 68 page 13 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.

B: INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.

C: CORRECT: The limit on flow rate through one CVCS demineralizer during a plant heatup is 250 gpm.

D: INCORRECT: All distracters are plausible since the student must have knowledge of the CVCS demineralizer flow limits for one demin during a plant heatup. Flow rates above 250 gpm are allowed if 2 demins of the same type are aligned.

Question Level: F Question Difficulty 2 Justification:

The student must have knowledge of CVCS demin flow requirements.

Page 110 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2503 Last used on an NRC exam: Never RO Sequence Number: 56 The Unit has been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and mid-loop operations have just commenced with the following conditions:

RHR Pumps A and C are in service.

RHR pump B is in standby.

RCS Hot Leg level is +9 inches.

Subsequently:

RHR Pump A receives alarm RHR PUMP A CURRENT LO.

The alarm came in and out 2 times and is now locked in.

RCS Hot Leg level is observed at +4 inches.

(1) With the conditions given, the RHR PUMP A CURRENT LO alarm is indication of RHR Pump _____(1)_____.

AND (2) The Crew should _____(2)_____.

(1) (2)

A. RUNOUT secure RHR Pump A & C and refill RCS using a LHSI Pump B. RUNOUT lower RHR Pump A flow to less than 1000 gpm C. VORTEXING secure RHR Pump A & C and refill RCS using a LHSI Pump D. VORTEXING lower RHR Pump A flow to less than 1000 gpm Answer: C VORTEXING - secure RHR Pump A & C and refill RCS using a LHSI Pump Page 111 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2503 K/A Catalog Number: 005 A2.01 Tier: 2 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(10)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, or mitigate the consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation.

STP Lesson: LOT 201.09 Objective Number: 4245 GIVEN a plant or system condition, PREDICT the operation of the Residual Heat Removal System.

Reference:

LOT 201.09 Lesson Plan on RHR and 0POP04-RH-0001, Loss of Residual Heat Removal Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because pump runout is a condition where pump current is effected but for pump runout current would be high.

B: INCORRECT: Plausible because pump runout is a condition where pump current is effected but for pump runout current would be high. Also plausible because lowering RHR Pump flow would be helpful but lowering flow less than 1000 gpm would be an incorrect action because the pump trips at 925 gpm. The procedure states to lower flow for some conditions to between 1000 and 1500 gpm.

C: CORRECT: With the given conditions the CURRENT LO alarm would indicate vortexing. 0POP04-RH-0001 will have the crew stop all running RHR pumps and refill the RCS.

D: INCORRECT: Plausible because lowering RHR Pump flow would be helpful but lowering flow less than 1000 gpm would be an incorrect action because the pump trips at 925 gpm. The procedure states to lower flow for some conditions to between 1000 and 1500 gpm.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition to determine the correct procedural action and have knowledge of the RHR Pump lcurrent low alarm.

Page 112 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2475 Last used on an NRC exam: Never RO Sequence Number: 57 During a Large Break LOCA, if the Safety Injection Accumulators failed to inject, what would be the effect on water being used to cool the core?

During the recirculation phase, if the Safety Injection Accumulators failed to inject, the ECCS water recirculated through the core would have a _____(1)_____ boron concentration and a

_____(2)_____ pH.

A. (1) higher (2) lower B. (1) lower (2) lower C. (1) higher (2) higher D. (1) lower (2) higher Answer: D (1) lower - (2) higher Page 113 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2475 K/A Catalog Number: 006 K6.02 Tier: 2 Group/Category: 1 RO Importance: 3.4 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

Core flood tanks (accumulators).

STP Lesson: LOT 201.10 Objective Number: 29419 GIVEN a plant condition, PREDICT the operation of the ECCS to include automatic actuations, interlocks and/or trips.

Reference:

LOT 201.10 Lesson Plan on ECCS Rev 19 and LOT 201.11 Lesson Plan on Containment Spray Rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.

B: INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.

C: INCORRECT: All distracters are plausible because the student must have knowledge of the contents of the SI Accumulators and know how chemisry of cooling water would be effected if there was a loss of the accumulators.

D: CORRECT: Without the water from the accumulators injecting into the RCS during a LB LOCA it would make the water being used for cooling the core at a lower boron concentration and a higher PH.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of the contents of the SI Accumulators and how they would effect system operation during a LB LOCA.

Page 114 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1116 Last used on an NRC exam: 2003 RO Sequence Number: 58 At 2335 psig (87.5%) and increasing, the automatic response of the Pressurizer Pressure Master Controller SHOULD be:

A. Spray Valves CLOSED, and PORV CLOSED.

B. Spray Valves FULLY OPEN, and PORV OPEN.

C. Spray Valves PARTIALLY OPEN, and PORV OPEN.

D. Spray Valves PARTIALLY OPEN, and PORV CLOSED.

Answer: B Spray Valves FULLY OPEN, and PORV OPEN.

Page 115 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 1116 K/A Catalog Number: 010 A3.02 Tier: 2 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the PZR PCS, including: PZR pressure STP Lesson: LOT 201.14 Objective Number: 92779 Given plant conditions, determine their effects on the pressurizer prssure and level control system.

Reference:

LOT201.14 Rev 14 slide 31 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: All distractors are plausible because the positions are determined by the demand signal from the pzr pressure master controller in automatic.

B: CORRECT: With pzr pressure master controller in automatic the sprays are fully open at 72% and the PORV opens at 87.5%

C: INCORRECT: All distractors are plausible because the positions are determined by the demand signal from the pzr pressure master controller in automatic.

D: INCORRECT: All distractors are plausible because the positions are determined by the demand signal from the pzr pressure master controller in automatic.

Question Level: F Question Difficulty 3 Justification:

Student must have fundamental knowledge of the pzr pressure master controller and its setpoints.

Page 116 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2477 Last used on an NRC exam: Never RO Sequence Number: 59 The Reactor will trip at a specified Reactor Coolant Pump frequency setpoint.

If Reactor Coolant Pump frequency became an issue and the Reactor trip setpoint had drifted

_____(1)_____ the Reactor would be more likely to exceed _____(2)_____ limits.

A. (1) high (2) DNB B. (1) high (2) Fuel Integrity C. (1) low (2) DNB D. (1) low (2) Fuel Integrity Answer: C (1) low - (2) DNB Page 117 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2477 K/A Catalog Number: 012 A1.01 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including:

Trip setpoint adjustment.

STP Lesson: LOT 202.20 Objective Number: 3832 DESCRIBE the reactor protection system control and permissive interlocks including inputs, setpoints, coincidences, and functions.

Reference:

LOT 201.20 Rev 18 Lesson Plan on SSPS and Technical Specifications bases for the reasons for the different Reactor Trips. Section 2 Basis.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because it is reasonalble to believe the Reactor Trip setpoint could drift high and cause the Reactor to be more likely to exceed DNB limits.

B: INCORRECT: Plausible because it is reasonalble to believe the Reactor Trip setpoint could drift high and plausible because Fuel Integrity is a concern from the reactor but Fule Integrity is protected by the OPDT Reactor Trip.

C: CORRECT: The Reactor will trip on low Reactor Coolant Pump frequency. If the setpoint were to drift low then the Reactor would be more likely to exceed DNB limits.

D: INCORRECT: Plausible because Fuel Integrity is a concern from the reactor but Fule Integrity is protected by the OPDT Reactor Trip.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of Reactor Trip setpoints and the reason for the trips.

Page 118 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2478 Last used on an NRC exam: Never RO Sequence Number: 60 The Unit is at 28% power when the following occurs:

RCP C develops a sheared shaft.

What automatic response will occur?

A. SG C MFRV will open to restore level in SG C.

B. SG C MFRV will close to restore level in SG C.

C. Auxiliary Feedwater starts due to an automatic Reactor Trip signal.

D. Main Feedwater isolates due to an automatic Reactor Trip signal.

Answer: A SG C MFRV will open to restore level in SG C.

Page 119 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2478 K/A Catalog Number: 059 K4.13 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Knowledge of the MFW design feature(s) and/or interlocks which provide for the following:

Feedwater fill of S/G upon loss of RCP.

STP Lesson: LOT 202.13 Objective Number: 20359 GIVEN plant/system conditions, PREDICT the operation of the Feedwater System.

Reference:

LOT 202.13 Lesson Plan on Main Feedwater Rev 17 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: A loss of RCP flow will cause a loss of load in the affected loop. SG level will quickly lower and the SG MFRV will open to restore level. An automatic reactor trip will not immediately occur on the loss of one RCS loop flow because reactor power is below 40% power. (P-8)

B: INCORRECT: Plausible because the student has to remember that on a loss of RCS loop flow without a Reactor trip that SG level in the affected loop will lower. If the student thought the level would rise then they would chose that the MFRV would close.

C: INCORRECT: Plausible because on a Reactor Trip AFW will eventually start if SG levels get below 20% and the student has to remember that the unit will not automatically trip on loss of one RCS loop flow when below 40% power. (P-8)

D: INCORRECT: Plausible because on a Reactor Trip Main Feedwater will isolate but only when RCS temperature lowers below 571 degrees F and the student has to remember that the unit will not automatically trip on loss of one RCS loop flow when below 40% power. (P-8)

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given condition to determine how MFW will respond.

Page 120 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2479 Last used on an NRC exam: Never RO Sequence Number: 61 The overspeed trip set point for the AFW Pump Terry Turbine is (1) of normal speed AND the overspeed trip must be reset (2) .

A. (1) 115% (2) locally B. (1) 115% (2) from the Control Room C. (1) 105% (2) locally D. (1) 105% (2) from the Control Room Answer: A (1) 115% - (2) locally Page 121 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2479 K/A Catalog Number: 061 K4.07 Tier: 2 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(7)

Knowledge of AFW design Feature(s) and/or interlock(s) which provide for the following:

Turbine trip including overspeed.

STP Lesson: LOT 202.28 Objective Number: 43847 DISCUSS the following elements associated wht the AFW turbine driven Pump: A. Signals that will trip the trip and throttle valve B. How to reset the trip and throttle valve C. Lubrication system D. How to determine status E. Exhaust and drain paths F. Reason for warming up steam lines before startup G.

Transfer switch operation to transfer operation to ASP H. Associated Alarms.

Reference:

0POP09-AN-06M4 Rev 34, Page 36 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: Overspeed trip set point for the Terry Turbine is 115% and must be reset locally.

B: INCORRECT: Plausible because starting and stopping the Terry Turbine using AF-MOV-0514 can be done from the control room. Incorrect because if the Terry Turbine is shut down due to overspeed the overspeed trip linkage must be reset locally.

C: INCORRECT: Plausible because the indication listed is more than the nornal speed but the trip set point is 115%.

D: INCORRECT: Plausible because the indication listed is more than the nornal speed but the trip set point is 115%. Plausible because starting and stopping the Terry Turbine using AF-MOV-0514 can be done from the control room. Incorrect because if the Terry Turbine is shut down due to overspeed the overspeed trip linkage must be reset locally.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of the AFWP #14 design feature for overspeed.

Page 122 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2480 Last used on an NRC exam: Never RO Sequence Number: 62 A Large Break LOCA has occurred with the following conditions:

Containment pressure is 17 psig and slowly lowering.

The crew has completed all actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection, and transitioned to 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant.

(1) Which Control Room Panel can the RO operate the Containment Spray (CS) Pumps?

AND (2) How many CS Pumps should be running?

A. (1) CP-001 (2) 3 CS Pumps B. (1) CP-001 (2) 2 CS Pumps C. (1) CP-002 (2) 3 CS Pumps D. (1) CP-002 (2) 2 CS Pumps Answer: D (1) CP-002 - (2) 2 CS Pumps Page 123 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2480 K/A Catalog Number: 103 G2.1.31 Tier: 2 Group/Category: 1 RO Importance: 4.6 10CFR

Reference:

55.41(b)(10)

Containment:

Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

STP Lesson: LOT 201.11 Objective Number: 2009 GIVEN a plant or system condition, PREDICT the operation of the Containment Spray System.

Reference:

0POP05-EO-EO00, Rev. 23, Reactor Trip or Safety Injection, CIP.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because CP-001 also has safety related equipment that would need to be operated under these conditions. Also, with the given containment pressure it would be reasonable to believe that all CS Pumps should remain running.

B: INCORRECT: Plausible because CP-001 also has safety related equipment that would need to be operated under these conditions.

C: INCORRECT: Plausible because with the given containment pressure it would be reasonable to believe that all CS Pumps should remain running.

D: CORRECT: CS Pumps Controls are located on CP-002 and the CIP for 0POP05-EO-EO00, Reactor Trip or Safety Injection, states that for RWST conservation a 3rd CS Pump should be stopped.

Question Level: H Question Difficulty 2 Justification:

The student must analyze the given conditions to determine the correct system lineup and have knowledge of where to operate the CS Pumps.

Page 124 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2482 Last used on an NRC exam: Never RO Sequence Number: 64 The Unit is at 75% Reactor power with the following conditions:

PZR level control is in automatic controlling PZR level on program.

LOOP A Tavg is 586ºF.

LOOP B Tavg is 586ºF.

LOOP C Tavg is 584ºF.

LOOP D Tavg is 585ºF.

Subsequently:

LOOP B Tavg Channel fails high.

Which LOOP Tavg is now determining PZR program level?

A. LOOP A Tavg.

B. LOOP B Tavg.

C. LOOP C Tavg.

D. LOOP D Tavg.

Answer: B LOOP 'B' Tavg.

Page 127 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2482 K/A Catalog Number: 016 A3.01 Tier: 2 Group/Category: 2 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the NNIS, including:

Automatic selection of NNIS inputs to control systems.

STP Lesson: LOT 201.14 Objective Number: 92779 GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control system.

Reference:

LOT 201.14 Lesson Plan on PZR pressure and level control systems. Rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.

B: CORRECT: PZR program level setpoint is automatically set off of the highest channel of RCS LOOP Tavg.

C: INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.

D: INCORRECT: All distractors are plausible because other systems like SSPS (uses lowest 2 Tavg to close steam dumps) and RCS Thot (bypass a failed Thot channel to calculate Thot) use different ways to calculate setpoints. In the question, after LOOP 'B' fails high each channel has a different temperature.

Question Level: H Question Difficulty 2 Justification:

The student must analyze the given condition and have knowledge of the PZR level control system to determine which channel would set PZR program level setpoint.

Page 128 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2589 Last used on an NRC exam: Never RO Sequence Number: 65 The Unit is at 100% power with the following condition:

Loop A (T-0410) Tavg and Delta T channels are in the Tripped condition per Technical Specifications.

Subsequently:

Loop D Tcold (TI-0440B) fails high.

The Steam Dumps arm due to a (1) signal.

AND With no operator action Tavg will lower to (2) .

A. (1) C-8 (2) 563°F B. (1) C-7 (2) 563°F C. (1) C-8 (2) 561°F D. (1) C-7 (2) 561°F Answer: A (1) C (2) 563°F Page 129 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2589 K/A Catalog Number: 041 A1.01 Tier: 2 Group/Category: 2 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SDS controls including : Tave, verification above low/low setpoint.

STP Lesson: LOT 202.09 Objective Number: 93002 Given plant conditions, DETERMINE their effects on the Steam Dump System.

Reference:

LOT 202.09 Rev 13 LP on Steam Dump System Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: With the conditions given C-8 would arm the steam dumps and with no operator action Tavg will lower to the P12 set point of 593 degrees F at which point the Steam Dumps will close and cycle at that temperature. P-12 closes the steam dumps at 563 degrees F to prevent an uncontrolled cooldown which could cause a reactivity addition that could restart the reactor.

B: INCORRECT: Plausible because C-7 is an input to the Steam Dump System that arms the Steam Dumps. Incorrect because with the condition given the reactor would have tripped and C-8 would be the signal that arms the Steam Dumps.

C: INCORRECT: Plausible because 561 degrees F is the TS low limit for Tavg when in Modes 1 and

2. Incorrect because the Steam Dump P-12 setpoint is 563 degrees F.

D: INCORRECT: Plausible because C-7 is an input to the Steam Dump System that arms the Steam Dumps. Incorrect because with the condition given the reactor would have tripped and C-8 would be the signal that arms the Steam Dumps. Plausible because 561 degrees F is the TS low limit for Tavg when in Modes 1 and 2. Incorrect because the Steam Dump P-12 setpoint is 563 degrees F.

Question Level: H Question Difficulty 3 Justification:

Student must analyze given conditions and predict the operation of the steam dump system.

Page 130 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2484 Last used on an NRC exam: Never RO Sequence Number: 66 The Unit is at 100% power with the following condition:

A recent Chemistry sample of RCB Atmosphere Noble Gas is reading 6.50E-4 µCi/cc resulting in a purge permit notification level of 9.75E-4 µCi/cc.

(1) What procedural actions should be performed prior to starting the purge?

AND (2) If a purge was started WITHOUT performing this action what would be the consequences?

A. (1) Raise the isolation setpoint on RT-8012 and RT-8013.

(2) An ALERT alarm ONLY will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.

B. (1) Raise the isolation setpoint on RT-8012 and RT-8013.

(2) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.

C. (1) Verify Containment Ventilation Actuation is BLOCKED.

(2) An ALERT alarm ONLY will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.

D. (1) Verify Containment Ventilation Actuation is BLOCKED.

(2) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.

Answer: B (1) Raise the isolation setpoint on RT-8012 and RT-8013.

(2) An ALERT and HIGH alarm will come in on RT-8012 and RT-8013 RCB Purge Exhaust Monitors.

Page 131 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2484 K/A Catalog Number: 029 A2.04 Tier: 2 Group/Category: 2 RO Importance: 2.5 10CFR

Reference:

55.41(b)(10)

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, or mitigate the consequences of those malfunctions or operations:

Health physics sampling of containment atmosphere.

STP Lesson: LOT 202.33 Objective Number: 97097 Discuss 0POP02-HC-0003 including: A. Purpose and scope B. Precautions C. Notes.

Reference:

0POP02-HC-0003, Supplemental Containment Purge Rev 25 page 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the student has to be aware of the isolation alarm (High Alarm) set point that would cause an ESF Containment Ventalation Isolation (CVI).

B: CORRECT: RT-8011 reading is above the level which will cause an isolation alarm (High Alarm) on RT-8012 and RT-8013 (5.00E-4). The procedural action is to raise the RT-8012 and RT-8013 isolation alarm set point (High Alarm) to prevent an ESF actuation prior to performing the purge.

C: INCORRECT: Plausible because the student has to be aware of the isolation alarm (High Alarm) set point that would cause an ESF Containment Ventalation Isolation (CVI). Also, you can block CVI but the procedural action to verify that CVI is blocked is for Mode 5 and 6 and Defueled only.

D: INCORRECT: Plausible because you can block CVI but the procedural action to verify that CVI is blocked is for Mode 5 and 6 and Defueled only.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition to determine the correct procedural actions and have knowledge of the ESF actuation setpoint for CVI. NOTE: Chemistry performs samples of contaiment at STP. NOT Health Physics.

Page 132 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2486 Last used on an NRC exam: Never RO Sequence Number: 67 The Unit is at 100% power when the following alarm annunciates:

RWST LEVEL HI/LO on CP-001 ACC TK 1A(2A) PRESS HI/LO on CP-001 CNTMT PRESS HI/LO on CP-002 PZR DNBR PRESS LOW on CP-004 Considering each one separately, which alarm based on its setpoints would require a one hour entry into Technical Specifications?

A. RWST LEVEL HI/LO on CP-001 B. ACC TK 1A(2A) PRESS HI/LO on CP-001 C. CNTMT PRESS HI/LO on CP-002 D. PZR DNBR PRESS LOW on CP-004 Answer: C CNTMT PRESS HI/LO on CP-002 Page 133 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2486 K/A Catalog Number: G2.2.39 Tier: 3 Group/Category: 2 RO Importance: 3.9 10CFR

Reference:

55.41(b)(10)

Knowledge of less than or equal to one hour Technical Specification action statements for systems.

STP Lesson: LOT 503.01 Objective Number: 80056 GIVEN a system condition, DETERMINE the applicable Technical Specification and/or the Technical Requirements Manual (TRM) and APPLY the specification(s).

Reference:

LOT 503.01 Lesson Plan On TSs, especially those that require the unit to take action wihtin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. TS 3.6.1.4 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because the alarm is yellow indicating a higher priority, is related to safety systems and TSs apply but not based just on alarm setpoints.

B: INCORRECT: Plausible because the alarm is yellow indicating a higher priority, is related to safety systems and TSs apply but not based just on alarm setpoints.

C: CORRECT: Containment Pressure HI/LO alarm is the only one listed that has setpoints at the TS limits. (for hi or lo limit) With this alarm in containment pressure has to be restored to within limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or a shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is required. With the other alarms the setpoints for the alarms are within the limits of the TS for that parameter.

D: INCORRECT: Plausible because the DNBR alarm does require a TS entry but it is a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> TS.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of alarms that would cause a direct entry in to Technical Specifications.

Page 134 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2487 Last used on an NRC exam: Never RO Sequence Number: 68 In regards to 480 V Load Center 12L; (1) Which Unit(s) supplies power to the Load Center?

AND (2) Which Unit(s) has a trouble alarm for the Load Center?

(1) (2)

A. Unit 1 ONLY Both Unit 1 and Unit 2 B. Both Unit 1 and Unit 2 Both Unit 1 and Unit 2 C. Unit 1 ONLY Unit 1 ONLY D. Both Unit 1 and Unit 2 Unit 1 ONLY Answer: A Unit 1 ONLY - Both Unit 1 and Unit 2 Page 135 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2487 K/A Catalog Number: G2.2.3 Tier: 3 Group/Category: 2 RO Importance: 3.8 10CFR

Reference:

55.41(b)(7)

Knowledge of the design, procedural, and operational differences between units.

STP Lesson: LOT 203.21 Objective Number: N0056 STATE the design differences between Unit 1 and Unit 2.

Reference:

LOT 203.21, Rev 4, Lesson Plan on Unit Differences.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: For LC 12L only Unit 1 supplies power to this single ended load center but both units have alarms for the bus.

B: INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.

C: INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.

D: INCORRECT: All distractors are credible because the student must know which sytems in the plant are supplied by one or both units.

Question Level: F Question Difficulty 2 Justification:

the student must have fundamental knowledge of Unit differences.

Page 136 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2490 Last used on an NRC exam: Never RO Sequence Number: 69 The Unit is at 100% power.

To lower RCB iodine levels prior to a RCB entry, the crew should start (1) from (2) .

A. (1) Normal Containment Purge (2) CP-022 B. (1) Normal Containment Purge (2) CP-002 C. (1) Containment Carbon Filter Unit (2) CP-022 D. (1) Containment Carbon Filter Unit (2) CP-002 Answer: C (1) Containment Carbon Filter Unit (2) CP-022 Page 137 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2490 K/A Catalog Number: 027 A4.03 Tier: 2 Group/Category: 2 RO Importance: 3.3 10CFR

Reference:

55.41(b)(7)

Ability to manually operate and/or monitor in the control room:

CIRS fans.

STP Lesson: LOT 202.33 Objective Number: 80242 DESCRIBE the instrumentation and controls available to monitor and operate the RCB-HVAC systems.

Reference:

0POP02-HC-0001, Containment HVAC Rev 27 page 20 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because a purge can be used to lower iodine levels in containment prior to a containment entry but in Modes 1 to 4 Supplemental Purge must be used not Normal purge. The system is operated on CP-002 in the control room.

B: INCORRECT: Plausible because a purge can be used to lower iodine levels in containment prior to a containment entry but in Modes 1 to 4 Supplemental Purge must be used not Normal purge. Also plausible because the RCFCs are operated from panel 2.

C: CORRECT: Containment Carbon Unit Filters can be used when requested by Health Physics to lower iodine levels in containment prior to a containment entry. The system is operated on CP-002 in the control room.

D: INCORRECT: Plausible because the RCFCs are operated from panel 2.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of the procedural requirements for starting the Containment Carbon Units and when they would be needed.

Page 138 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2586 Last used on an NRC exam: Never RO Sequence Number: 70 The Reactor Operator can monitor Instrument and Service Air header pressures on (1) and the normal operating pressure is (2) .

A. (1) CP-007 (2) 110 psig B. (1) CP-002 (2) 110 psig C. (1) CP-007 (2) 125 psig D. (1) CP-002 (2) 125 psig Answer: C (1) CP-007 - (2) 125 psig Page 139 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2586 K/A Catalog Number: 079 G2.1.31 Tier: 2 Group/Category: 2 RO Importance: 4.6 10CFR

Reference:

55.41(b)(7)

Service Air:

Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

STP Lesson: LOT 202.26 Objective Number: 80556 DEScRIBE the instrumentation and controls available to monitor and operate the Instrument Air and Service Air systems.

Reference:

0POP04-IA-0001, Rev 17, Page 2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Plausible because 110 psig is above the pressure for parts of IA/SA to start isolating.

Incorrect because at this pressure all of the IA/SA Compressors would be running and loaded.

B: INCORRECT: Plausible because IA OCIV is operated on CP-002 and this is the only control for IA/SA on the control panels. Incorrect because the IA/SA header pressure findication os on CP-007.

Plausible because 110 psig is above the pressure for parts of IA/SA to start isolating. Incorrect because at this pressure all of the IA/SA Compressors would be running and loaded.

C: CORRECT: IA Header and SA Header Indications can be found on CP-007 and the normal operating pressure for both is about 120 to 125 psig.

D: INCORRECT: Plausible because IA OCIV is operated on CP-002 and this is the only control for IA/SA on the control panels. Incorrect because the IA/SA header pressure findication os on CP-007.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of location of control room indication and IA/SA System pressures.

Page 140 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2116 Last used on an NRC exam: 2011 RO Sequence Number: 71 In accordance with 0POP04-IA-0001, Loss of Instrument Air, which of the following is the FIRST Instrument Air pressure reached that would require a manual Reactor Trip?

A. 90 psig B. 80 psig C. 70 psig D. 60 psig Answer: D 60 psig Page 141 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2116 K/A Catalog Number: APE 065 AA2.06 Tier: 1 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

When to trip reactor if instrument air pressure is decreasing.

STP Lesson: LOT 505.01 Objective Number: 92108 Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.

Reference:

0POP04-IA-0001, Rev.17 CIP Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Plausible because 90 psig is a setpoint but its when the yard isolation valve closes.

B: INCORRECT: Plausible because 80 psig is a setpoint but its when the dryer bypass valve opens.

C: INCORRECT: Plausible because 70 psig is the approximate setpoint when the main feed reg valves begin to close.

D: CORRECT: If air pressure goes below 60 psig, a manual Reactor trip is required by 0POP04-IA-0001. 55 psig is the FIRST pressure indicating less than 60 psig.

Question Level: F Question Difficulty 3 Justification:

Student must have fundamental knowledge of the procedural requirement to perform a manual Reactor Trip on lowering Inst. Air pressure.

Page 142 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2197 Last used on an NRC exam: 2013 RO Sequence Number: 72 Step 1 of 1POP09-AN-03M2 A-1, 125VDC System E1A11 TRBL, requires a verification that E1A11 battery discharge current is less than 200 amps.

Per the annunciator response, the battery discharge current for this battery should be read using the A. BATT CUR indicator on CP-003.

B. BATT CUR indicator on CP-010.

C. QDPS plasma display.

D. local battery current indication.

Answer: A 'BATT CUR' indicator on CP-003.

Page 143 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2197 K/A Catalog Number: 063 A4.03 Tier: 2 Group/Category: 1 RO Importance: 3.0 10CFR

Reference:

55.41(b)(7)

Ability to manually operate and/or monitor in the control room: Battery discharge rate STP Lesson: LOT 201.37 Objective Number: 92986 DESCRIBE the local and MCR instrumentation available to monitor the Class 1E 125 VDC System

Reference:

1POP09AN03M2 Rev 30 window A-1 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: This information is available on CP-003 B: INCORRECT: Credible because CP-010 is also an electrical panel, however this indication is not located on it.

C: INCORRECT: Credible because the QDPS computer provides safety related system information, but not this. NOTE: This information is available on the ICS computer system (in the control room).

D: INCORRECT: Credible because many plant parameters are only available through local indication.

The applicant must be familiar with what indication is on the control panels to correctly respond.

Question Level: F Question Difficulty 2 Justification:

The applicant must have knowledge of the indications available in the control room for the batteries.

Page 144 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 62 Last used on an NRC exam: 2001 RO Sequence Number: 73 Which of the following radiation monitor automatic actions control radioactive releases to the environment?

A high activity alarm on A. RT-8038 LWPS rad monitor will cause Waste Monitor Tank Pumps to stop.

B. RT-8041 TGB Drain rad monitor will cause TGB Sump #1 Sump Pumps to stop.

C. RT-8042 Condensate Polisher rad monitor will cause TDS Waste Discharge to Neutralization Basin Valve to shift to RECIRC.

D. RT-8043 SG Blowdown rad monitor will cause SG Blowdown Isolation to Neutralization Basin Valve to shift to RECIRC.

Answer: B RT-8041 TGB Drain rad monitor will cause TGB Sump #1 Sump Pumps to stop.

Page 145 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 62 K/A Catalog Number: G2.3.11 Tier: 3 Group/Category: 3 RO Importance: 3.8 10CFR

Reference:

55.41(b)(11)

Ability to control radiation releases.

STP Lesson: LOT 505.01 Objective Number: 92107 DISCUSS automatic actions expected to occur on entry conditions for the reference procedure.

Reference:

0POP04-RA-0001, Page 43 (Rev 32)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: Incorrect: Waste Monitor Tank Pump does not stop; LWPS Discharge Valve WL-FV-4077 shifts to RECIRC B: Correct: TGB Sump Number 1 sump pumps will stop on RT-8041 high activity C: Incorrect:1(2)-FV-5804 TDS Waste Discharge to Neutralization Basin Valve CLOSES on RT-8042 high activity.

D: Incorrect: 1(2)-SB-FV-5019 SG Blowdown Isolation to Neutralization Basin Valve CLOSES on RT-8043 high activity Question Level: F Question Difficulty 3 Justification:

The candidate must comprehend the systems physical arrangement, operation during discharges and interlocking scheme.

Page 146 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2581 Last used on an NRC exam: Never RO Sequence Number: 74 The Unit is at 100% power when the following occurs:

Annunciator 3M02-A-5, 120V AC CH1 DIST PNL 1201 TRBL, alarms.

DP-1201 has lost power.

The failure will result in the loss of (1) .

AND The crew will FIRST enter (2) .

A. (1) NI-41 AND NI-35 (2) 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution B. (1) NI-41 AND NI-35 (2) 0POP04-NI-0001, Nuclear Instrument Malfunction C. (1) N-41 ONLY (2) 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution D. (1) N-41 ONLY (2) 0POP04-NI-0001, Nuclear Instrument Malfunction Answer: A (1) NI-41 AND NI-35 (2) 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution Page 147 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2581 K/A Catalog Number: APE 057 G2.2.44 Tier: 1 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(7)

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.

STP Lesson: LOT 505.01 Objective Number: 92108 Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.

Reference:

0POP04-VA-0001 Rev 32 Addendum 4 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: A loss of DP-1201 will casue a loss of PR NI-41 and IR NI-35. The first procedure to enter will address the loss of DP-1201.

B: INCORRECT: Plausible because if the NI failures were the only problem then 0POP04-NI-0001 would be entered first. Incorrect because other failurse have to be addressed with the loss of DP-1201 and 0POP04-VA-0001 is the correct procedure to enter first.

C: INCORRECT: Plausible because NI-35 is affected by the loss of DP-1201. Incorrect because NI-41 is also affected.

D: INCORRECT: Plausible because NI-35 is affected by the loss of DP-1201. Incorrect because NI-41 is also affected. Plausible because if the NI failures were the only problem then 0POP04-NI-0001 would be entered first. Incorrect because other failurse have to be addressed with the loss of DP-1201 and 0POP04-VA-0001 is the correct procedure to enter first.

Question Level: H Question Difficulty 3 Justification:

Student must analyze given conditions and predict the plant response to given failure and know procedure entry criteria.

Page 148 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2471 Last used on an NRC exam: Never RO Sequence Number: 75 Reactor Coolant Pumps are secured if an inadvertent containment isolation (1) occurs.

They are secured due to a loss of (2) to the RCPs.

A. (1) Phase A (2) seal injection B. (1) Phase A (2) CCW cooling C. (1) Phase B (2) seal injection D. (1) Phase B (2) CCW cooling Answer: D (1) Phase B (2) CCW cooling Page 149 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC RO EXAM Exam Bank No.: 2471 K/A Catalog Number: 003 K1.08 Tier: 2 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(7)

Knowledge of the physical connections and/or cause-effect relationships between the RCPs and the following systems:

Containment isolation.

STP Lesson: LOT 201.05 Objective Number: 50785 DESCRIBE the physical relationship between the reactor collant pumpsand the following: A. RCP Seals B. Chemical Volume and Control System C. Containment Isolation System D. Reactor Coolant System E.

Component Cooling Water.

Reference:

LOT 201.05 Lesson Plan for RCPs Rev 17 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Phase A does not require RCPs to be secured and seal injection is not lost solely on a Phase A. Plausible because Containment Isolation Phase A isolates many systems to containment but not any that requires tripping the RCPs. Also, plausible because seal injection is isolated on phase A along with a low charging discharge pressure.

B: INCORRECT: Phase A does not require RCPs to be secured. Plausible because Containment Isolation Phase A isolates many systems to containment but not any that requires tripping the RCPs C: INCORRECT: Seal injection is not lost to RCPs on phase B isolation. Plausible because seal injection is isolated on other signals.

D: CORRECT: If a Containemnt Isolation Phase B were to occur then CCW to the RCPs would be lost and the RCPs would have to be tripped.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of the effects of a containment isolation has on RCPs.

Page 150 of 150

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2461 Last used on an NRC exam: Never SRO Sequence Number: 76 The Unit performed a rapid load reduction from 100% power to 80% power over a 40 minute time period.

Per Technical Specifications, an RCS sample to perform a/an (1) will be required because thermal power changed more than (2) in one hour.

A. (1) Isotopic Analysis for Iodine (2) 15%

B. (1) Radiochemical Analysis for determination (2) 15%

C. (1) Isotopic Analysis for Iodine (2) 10%

D. (1) Radiochemical Analysis for determination (2) 10%

Answer: A (1) Isotopic Analysis for Iodine (2) 15%

Page 1 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2461 K/A Catalog Number: G2.2.42 Tier: 3 Group/Category: 2 SRO Importance: 4.6 10CFR Reference or SRO Objective: 55.43(b)(2)

Ability to recognize system parameters that are entry level conditions for Technical Specifications STP Lesson: LOT 503.01 Objective Number: SRO92102 Given the topic or title of a specification included in the Technical Specifications, or the Technical Requirements Manual (TRM), DESCRIBE the general requirements of the specification to include components or administrative requirements affected, limitations, major time frames involved, major surveillance in order to comply, and the bases for the specification.

Reference:

TS 3/4.4.8 surveillance requirements table 4.4-4 action 4.b Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: An analysis for Iodine is the only one required after a power change of 15% or more in a one hour period.

B: INCORRECT: Radiochemical for E bar is required in the same TS surveillance requirement but no because power changed greater than 15% in one hour.

C: INCORRECT: Plausible because this is the correct analysis to be performed. However, it is not required until power has changed at least 15%.

D: INCORRECT: Radiochemical for E bar is required in the same TS surveillance requirement but not because power changed greater than 15% in one hour.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of Technical Specification requirements.

Page 2 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2577 Last used on an NRC exam: Never SRO Sequence Number: 77 The crew is performing the steps of 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, due to a Small Break LOCA.

A depressurization of all SGs to 1000 psig has just been performed.

Subsequently:

All SG pressures are 950 psig and lowering rapidly.

All SG flows 0.3 106 lbm/hr and rising.

Containment Pressure is 3.5 psig and rising.

RWST level is 125,000 gallons and lowering.

All MSIVs are open and will not close from the Control Room.

The Unit Supervisor will A. transition to 0POP05-EO-EC21, Uncontrolled Depressurization of all Steam Generators, and perform Addendum 1, to locally close MSIVs.

B. transition to 0POP05-EO-EO20, Faulted Steam Generator Isolation, and perform Addendum 1, to locally close MSIVs.

C. transition to 0POP05-EO-ES13, Transfer to Cold Leg Recirculation, and reset SI.

D. remain in 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, and reset SI.

Answer: B transition to 0POP05-EO-EO20, Faulted Steam Generator Isolation, and perform Addendum 1, to locally close MSIVs.

Page 3 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2577 K/A Catalog Number: APE 040 AA2.01 Tier: 1 Group/Category: 1 SRO Importance: 4.7 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine and interpret the following as they apply to the Steam Line Rupture:

Occurrence and location of a steam line rupture from pressure and flow indications.

STP Lesson: LOT 504.13 Objective Number: 81261 STATE/IDENTIFY the indications and anticipated readings used to determine that a faulted Steam Generator exists and which Steam Generator(s) is/are faulted.

Reference:

0POP05-EO-EO10, Rev 22, Step 3, Page 8 and 0POP05-EO-EO20, Rev 11, Step 1, Page 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because all SG pressures are lowering rapidly. Incorrect because a transition to 0POP05-EO-EC21 is not made directly from any procedure execept 0POP05-EO-EO20 B: CORRECT: With the given conditions it would be determined that a faulted SG has occurred in containment. This knowledge is assumed to be known by the SRO. The US will transition to 0POP05-EO-EO20 and with the MSIVs open will use Addendum 1 to close them locally.

C: INCORRECT: Plausible because with the original SBLOCA RWST level would be lowering and if this procedure were entered it would take priority. Incorrect because RWST level is still not low enough to transition to 0POP05-EO-ES13.

D: INCORRECT: Plausible because if the US thought that this subsequent condition was from a larger RCS break then staying in 0POP05-EO-EO10 would be justified. Incorrect because with the presence of SG flow going up then this indicates that the issue is a faulted SG.

Question Level: H Question Difficulty 3 Justification:

The student must anlayze the given conditions to determine the correct procedure transition.

Page 4 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 661 Last used on an NRC exam: 2014 SRO Sequence Number: 78 A LOCA with core damage has occurred with the following conditions:

An SAE has been declared.

The TSC and EOF are activated.

A MAB entry is required at the 41 containment penetration area.

Projected dose rate in the area is 1.16E+5 mR/hr.

Duration of the exposure is expected to be 3 minutes.

Who must authorize this exposure?

A. Plant General Manager B. STPNOC Vice President C. Radiological Director D. Emergency Director Answer: D Emergency Director Page 5 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 661 K/A Catalog Number: G2.3.4 Tier: 3 Group/Category: 3 SRO Importance: 3.7 10CFR Reference or SRO Objective: 55.43(b)(4)

Knowledge of radiation exposure limits under normal or emergency conditions.

STP Lesson: LOT 803.14 Objective Number: SRO-65180 Given a description of responsibilities related to an Emergency Response Organization position that interfaces with the Emergency Director, DETERMINE the responsible individual by title.

Reference:

0ERP01-ZV-IN06 Rev 7 page 6 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT: Plausible because the Plant General Manager is responsible for authorizing exposures in excess of 2 Rem at STP or 3 Rem Total during NORMAL operating conditions.

B: INCORRECT: Plausible because the STPNOC Vice President is responsible for authorizing exposures in excess of 2 Rem at STP or 4 Rem Total during NORMAL operating conditions.

C: INCORRECT: Plausible because the Radiological Director is responsible for authorizing exposures above 2 Rem but less than 5 Rem when responding to EMERGENCY conditions.

D: CORRECT: Emergency Director is responsible for authorizing exposures above 5 Rem when responding to EMERGENCY conditions. With a projected dose rate of 1.16E+5 mR/hr the total dose to respond to this emergency condition is 5.8 Rem (1.16E+5 mR/hr / 60minutes x 3 minutes =

5.8 R)

Question Level: H Question Difficulty 3 Justification:

The student must accurately determine the projected dose and then use that knowledge to determine the approval authority based on the applicable exposure limits for each position.

Page 6 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2505 Last used on an NRC exam: Never SRO Sequence Number: 79 The Unit is at 100% power with the following condition:

Pressurizer Heater Group A is out of service.

Subsequently:

At 08:00 an issue occurs on Pressurizer Heater Group B requiring trouble shooting.

At 10:00 maintenance determines heater group B has a capacity of 150KW.

The Unit Supervisor should be in at least HOT STANDBY by (2) .

A. 14:00 B. 16:00 C. 20:00 D. 22:00 Answer: A 14:00 Page 7 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2505 K/A Catalog Number: 010 A2.01 Tier: 2 Group/Category: 1 SRO Importance: 3.6 10CFR Reference or SRO Objective: 55.43(b)(2)

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures STP Lesson: LOT 507.01 Objective Number: 92106 Given plant conditions/symptoms, EVALUATE the conditions/symptoms and STATE whether or not the referenced procedure is to be used.

Reference:

TS 3.4.3 action b Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: PZR Heater Groups A and B are the 2 groups of heaters that are supplied by ESF power and covered by TS. TS requires each group of the heaters to have 175 KW capacity. Two heater group is inoperable and TS 3.4.3 action b applies - be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock starts when the malfunction occurs. Therefore the unit must be in hot standby by 14:00.

B: INCORRECT: Plausible because PZR Heater Groups A and B are the 2 groups of heaters that are supplied by ESF power and covered by TS. TS requires each group of the heaters to have 175 KW capacity. Two heater group is inoperable and TS 3.4.3 action b applies - be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Incorrect because student must know that the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to trouble shoot is used and is no longer allowed to be used to get to HOT STANDBY.

C: INCORRECT: Plausible because 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a common TS action. For this TS 3.4.3 b 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the requirment to be in shotdown. Incorrect because it is a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> requirement for hot standby.

D: INCORRECT: Plausible because 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a common TS action. For this TS 3.4.3 b 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the requirment to be in shotdown. Incorrect because it is a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> requirement for hot standby. Also incorrect because the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to trouble shoot is used and is no longer allowed to be used.

Question Level: H Question Difficulty 3 Justification:

The student must determine the operability requirments of the pressurizer and know which TS actions to apply.

Page 8 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 968 Last used on an NRC exam: Never SRO Sequence Number: 80 The crew is performing actions of 0POP05-EO-ES01, Reactor Trip Response, when the following occurs:

Containment pressure is 11 psig and rising.

STA validates an ORANGE path on Containment.

The Unit Supervisor will A. immediately re-enter 0POP05-EO-EO00, Reactor Trip or Safety Injection, at step 1.

B. complete the current step of 0POP05-EO-ES01, then re-enter 0POP05-EO-EO00, Reactor Trip or Safety Injection at step 1.

C. immediately transition to 0POP05-EO-FRZ1, Response to High Containment Pressure.

D. complete the current step of 0POP05-EO-ES01, then transition to 0POP05-EO-FRZ1, Response to High Containment Pressure.

Answer: A immediately re-enter 0POP05-EO-EO00, Reactor Trip or Safety Injection, at step 1.

Page 9 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 968 K/A Catalog Number: G2.4.16 Tier: 3 Group/Category: 4 SRO Importance: 4.4 10CFR Reference or SRO Objective: 55.43(b)(5)

Knowledge of EOP implementation hierarchy and cordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

STP Lesson: LOT 504.04 Objective Number: 92284 State the conditions, in accordance with 0POP01-ZA-0018, that the CSFs would be monitored but the FRPs not implemented.

Reference:

0POP01-ZA-0018 Rev. 21 page 25 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: CORRECT: If 0POP05-EO-EO00 is exited at step 4 to go to 0POP05-ES01 and subsequently a SI occurs, then 0POP05-EO-EO00 shall be re-entered at step 1 and discontinue CSF monitoring.

B: INCORRECT: Plausible because student may think he has to complete the current step of 0POP05-EO-ES01 because containment CSF is an orange path. Other times when orange path exists you complete the current step in effect before transitioning.

C: INCORRECT: Plausible to think you would go immediately to FRZ1 because you can immediately transition from ES01 to an FRP. However, since an SI occurred you must go back to EO00 first.

D: INCORRECT: Plausible because student may think he has to complete the current step of 0POP05-EO-ES01 because containment CSF is an orange path. Other times when orange path exists you complete the current step in effect before transitioning. However, in this case EO00 should be re-entered before addressing the orange path in containment.

Question Level: H Question Difficulty 3 Justification:

Student must have fundamental knowledge of prioritiztion of EOPs and CSFs per EOP user's guide.

Page 10 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2598 Last used on an NRC exam: Never SRO Sequence Number: 81 The Unit is performing a reactor startup with the following conditions:

The Reactor is critical with both Intermediate Range Nuclear Instruments indicating 1E-8 AMPS.

Intermediate Range SUR is 0 DPM.

Subsequently:

Intermediate Range Channel NI-35 indication fails low.

Which of the following correctly describes the action required by the Unit Supervisor?

A. Stop the reactor startup and be 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 TS 3.3.1 Action 1.

B. Continue with the reactor startup provided NI-35 is restored to operable status prior to raising power above the P-10 setpoint per TS 3.3.1 Action 3b.

C. Stop the reactor startup and suspend operations from positive changes in reactivity per TS 3.3.1 Action 4.

D. Continue with the reactor startup provided NI-35 is placed in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per TS 3.3.1 Action 5a.

Answer: B Continue with the reactor startup provided NI-35 is restored to operable status prior to raising power above the P-10 setpoint per TS 3.3.1 Action 3b.

Page 11 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2598 K/A Catalog Number: 015 A2.01 Tier: 2 Group/Category: 2 SRO Importance: 3.9 10CFR Reference or SRO Objective: 55.43(b)(5)

Nuclear Instrumentation System: Ability to a) predict the impacts of the following malfunctions or operations on the NIS; and b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Power supply loss or eratic operation.

STP Lesson: LOT 201.16 Objective Number: 4884 DESCRIBE the basic principle of operation of the various NIS detectors.

Reference:

TS 3.3.1.5 Action 3b.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because this is a partical action for reactor trip instrumentation from TS 3.3.1 Action 1. Incorrect because IR NIs do not have a specific shutdown action unless both channels are inoperable and then it would be from TS 3.0.3.

B: CORRECT: With reactor power above the P-6 setpoint, one intermediate range NI failure would cause entry into TS 3.3.1.5 Action 3b, restore NI-35 to operable status prior to raising power above P-10 (10% power).

C: INCORRECT: Plausible because the action list is for nuclear instrumentation failures. Incorrect because the action listed is actually for a Source Range instrument failure.

D: INCORRECT: Plausible because the action list is a common action for reactor trip instumentation.

Incorrect because there are only 2 channels of Intermediate Range NIs.

Question Level: H Question Difficulty 3 Justification:

The student has to analyze the given conditon to determine the correct TS action to take.

Page 12 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2590 Last used on an NRC exam: Never SRO Sequence Number: 82 Per the WOG Background Documents, why is INTEGRITY the fourth priority in the Critical Safety Function hierarchy?

INTEGRITY protects the A. Fuel Cladding ONLY.

B. RCS ONLY.

C. Fuel Cladding and the RCS.

D. Containment and the RCS.

Answer: B RCS ONLY Page 13 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2590 K/A Catalog Number: G2.4.22 Tier: 3 Group/Category: 4 SRO Importance: 4.4 10CFR Reference or SRO Objective: 55.43(b)(4)

Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

STP Lesson: LOT 504.04 Objective Number: 80488 STATE the basis for monitoring the Critical Safety Function Status Trees.

Reference:

WOG F-0 Background Document Page 14 and 15 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because if the RCS were to break, a loss of inventory would occur and inventory is tied to the Core Cooling critical safety function which protects the Fuel Cladding.

Incorrect because per the WOG the Integrity only protects the RCS.

B: CORRECT: Per the WOG INTEGRITY only potects the RCS.

C: INCORRECT: Plausible because if the RCS were to break it is reasonable that the student would then think that the Fuel Cladding would also need protecting. Incorrect because per the WOG the Integrity only protects the RCS.

D: INCORRECT: Plausible because if the RCS were to break it is reasonable that the sudent would then think that the Containment was also need to be protected due to the release of energy to containment from the RCS. Incorrect because per the WOG the Integrity only protects the RCS.

Question Level: F Question Difficulty 3 Justification:

The student has to have fundamental knowledge of the basis for prioritizing Critical Safety Functions.

Page 14 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2445 Last used on an NRC exam: Never SRO Sequence Number: 83 The Unit is at 100% power with the following conditions:

PZR pressure channel PT-458 is in BYPASS for maintenance on a card.

PZR pressure channel PT-457 is the CONTROLLING channel.

Subsequently:

PZR pressure channel PT-457 fails high.

Per Technical Specifications the Unit Supervisor will A. within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore at least one inoperable channel to operable or be 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 />.

B. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to shutdown the Unit and be 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 />.

C. place PT-457 in BYPASS condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. place PT-457 in TRIPPED condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Answer: B within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> take action to shutdown the Unit and be 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 />.

Page 15 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2445 K/A Catalog Number: APE 027 AA2.15 Tier: 1 Group/Category: 1 SRO Importance: 4.0 10CFR Reference or SRO Objective: 55.43(b)(2)

Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Actions to be taken if PZR pressure instrument fails high.

STP Lesson: LOT 201.14 Objective Number: 92779 GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control systems.

Reference:

LOT 201.14, Rev 14, Lesson on PZR pressure and level control systems. TS 3.0.3.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because the TS action described is an action taken in the same table of TSs. Incorrect because the action describes one for Reactor Trip Breakers.

B: CORRECT: It is assumed that the SRO knows that with another channel failed, PT-457, that there are now two channels inoperable. The correct action is to then enter TS 3.0.3 - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to take action to shutdown unit and be in hit standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Since the correct answer is apply TS 3.0.3 it is not a one hour TS entry and is considered SRO knowledge.

C: INCORRECT: Plausible because the TS action described is an action taken. Incorrect because putting a channel in BYPASS is only done tempararily to take the channel out of the logic.

D: INCORRECT: Plausible because the TS action described is an action taken for one channel being inoperable. Incorrect because this makes two channals inoperable.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given PZR control system conditions to determine the correct answer.

Page 16 of 50

Print Date 4/28/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2447 Last used on an NRC exam: Never SRO Sequence Number: 84 Unit 2 is at 100% power with a normal electrical lineup.

Subsequently:

x A lock-out occurs on the switch yard SOUTH Bus.

The Unit 2 Unit Supervisor will enter Technical Specification _____(1)_____ and direct the actions of _____(2)_____ to restore lost electrical power once the switchyard SOUTH Bus is restored.

A. (1) 3.8.1.1.a due to loss of ONE required offsite circuit (2) 0POP04-AE-0002, Loss of One or More 13.8 KV Auxiliary or Non-Class 4.16 KV Bus D B. (1) 3.8.1.1.a due to loss of ONE required offsite circuit (2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby Bus C. (1) 3.8.1.1.e due to loss of TWO required offsite circuitV (2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby Bus D. (1) 3.8.1.1.e due to loss of TWO required offsite circuitV (2) 0POP04-AE-0002, Loss of One or More 13.8 KV Auxiliary or Non-Class 4.16 KV Bus D Answer: C (1) 3.8.1.1.e due to loss of TWO required offsite circuitV (2) 0POP04-AE-0003, Loss of Power to One or More 13.8 KV Standby Bus Page 17 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2447 K/A Catalog Number: APE 056 G2.4.11 Tier: 1 Group/Category: 1 SRO Importance: 4.2 10CFR Reference or SRO Objective: 55.43(b)(2)

Loss of Off-Site Power:

Knowledge of abnormal condition procedures.

STP Lesson: LOT 201.31 Objective Number: 62351 GIVEN a plant or system condition, PREDICT the operation of the Non-Class 1E 13.8 to 4.16 volt AC distribution system.

Reference:

LOT 201.31, Rev 14, Lesson Plan on Non-Class 13.8 and 4.16 KV power and TS 3.8.1.1.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because losing one swityard bus constitues losing one offsite circuit. It is reasonable for a student to forget that the Stanby Bus G would also be effected. Also plausible because for this condition the Unit Supervisor would start directing actions per 0POP04-AE-0001 and then have to make a determination of which procedure to use next based on the extent of the loss of power.

B: INCORRECT: Plausible because losing one swityard bus constitues losing one offsite circuit. It is reasonable for a student to forget that the Stanby Bus G would also be effected.

C: CORRECT: Under a normal electrical lineup, if the South Bus was lost, it would constitue losing TWO required offsite sources because 13.8 KV Standby Bus 2G would also be effected. 0POP04-AE-0001 would first be entered and then 0POP04-AE-0003 would be used to restore power to Unit 2 13.8 KV Standby Bus 2G once the South switch yard Bus was restored.

D: INCORRECT: Plausible because for this condition the Unit Supervisor would start directing actions per 0POP04-AE-0001 and then have to make a determination of which procedure to use next based on the extent of the loss of power.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the condition to determine the extent of the loss of power and have knowledge of normal electrical lineups and of off-normal electrical procedures.

Page 18 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2448 Last used on an NRC exam: Never SRO Sequence Number: 85 The Unit is at 100% power with the following conditions:

PZR level control is selected to 465/466 for control Subsequently:

Vital AC Bus DP-1201 loses power PZR level will begin to _____(1)_____.

The Unit Supervisor will enter _____(2)_____.

(1) (2)

A. lower 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permits B. lower 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control. The US will also enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, if manpower permits C. rise 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permits D. rise 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control. The US will also enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, if manpower permits Answer: C rise until the RO takes manual control - 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution. The US will also enter 0POP04-RP-0002, Loss of Automatic Pressurizer Level Control, if manpower permits Page 19 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2448 K/A Catalog Number: APE 057 AA2.12 Tier: 1 Group/Category: 1 SRO Importance: 3.7 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

PZR level controller, instrumentation, and heater operation.

STP Lesson: LOT 201.14 Objective Number: 92779 GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control systems.

Reference:

LOT 201.14, Rev 14, Lesson on PZR pressure and level control systems. 0POP04-VA-0001 Rev 32 page 12 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because the student has to know how the failure affects the PZR level control system.

B: INCORRECT: Plausible because the student has to know how the failure affects the PZR level control system. Also it would be plausible to enter 0POP04-RP-0002 first because the failure would directly affect PZR level control.

C: CORRECT: A loss of power to DP-1201 would cause the controlling PZR level channel, LT-465, to fail low. This would cause actual level to rise as CV-FV-0205 would begin to open until manual control was made by the RO. The US would enter 0POP04-VA-0001 and only enter 0POP04-RP-0002 if manpower was available.

D: INCORRECT: It would be plausible to enter 0POP04-RP-0002 first because the failure would directly affect PZR level control.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given PZR control system conditions to determine the correct answer.

Page 20 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2463 Last used on an NRC exam: Never SRO Sequence Number: 86 The Unit is at 100% power with the following conditions:

Accumulator A boron concentration is 2688 ppm Accumulator B boron concentration is 2806 ppm Accumulator C boron concentration is 2912 ppm The Unit Supervisor will restore the boron concentration of (1) accumulator(s) to within the required technical specification limits within (2) .

A. (1) A and B ONLY (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. (1) A and B ONLY (2) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C. (1) A ONLY (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> D. (1) A ONLY (2) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Answer: D (1) A ONLY (2) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Page 21 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2463 K/A Catalog Number: 006 A2.10 Tier: 2 Group/Category: 1 SRO Importance: 3.9 10CFR Reference or SRO Objective: 55.43(b)(2)

Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Low boron concentration in SIS STP Lesson: LOT 201.10 Objective Number: 92102 Given the topic or title of a specification included in the TS, or TRM, describe the general requirements of the specification to include components or administrative requirements affected, limitations, major time frames involved, major surveillance in order to comply, and the basis for the specification.

Reference:

TS 3.5.1 action c Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Incorrect because with the given conditions only one accumulator is outside the TS limits for boron concentration. Plausible that the student may think two accumulators are inoperable with the given conditions.

B: INCORRECT: Incorrect because with the given conditions only one accumulator is outside the TS limits for boron concentration. Plausible that the student may think two accumulators are inoperable with the given conditions. Also, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action is for only one accumulator being inoperabe but plausible because this is an action within the same TS.

C: INCORRECT: It is correct that only one accumulator is inoperable per the surveillance requirements of TS. However, the one hour action is for more than one accumulator being inoperable. Plausible because this is an action within the same TS.

D: CORRECT: Student must know the surveillance requirement of TS to determine operability of the accumulators based on boron concentration. Accumulator boron concentration needs to be within 2700 to 3000 ppm. With the given conditions, one accumulator is inoperable due to low boron concentration. The correct action to take is restore it to wihtin the TS limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Question Level: F Question Difficulty 3 Justification:

Student has knowledge of TS surveillance requirements for operability of accuulators.

Page 22 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2450 Last used on an NRC exam: Never SRO Sequence Number: 87 The Unit experienced an ATWS and the crew is performing 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS, with the following condition:

Emergency Boration has been aligned to the RCS per Step 4 The crew is in the process of completing Step 5, ENSURE Containment Ventilation Isolation.

Subsequently Operators make the following observations:

Core Exit TCs are 680ºF and lowering Extended Range NIs are 4% and lowering Extended Range NIs Startup Rate is -0.1 What procedural steps should the Unit Supervisor perform NEXT?

A. TRANSITION TO 0POP05-EO-EO00, Reactor Trip or Safety Injection.

B. TRANSITION TO 0POP05-EO-FRC2, Response to Degraded Core Cooling.

C. GO TO Step 18 and then SECURE Emergency Boration.

D. GO TO Step 6 and CHECK SI Status.

Answer: C GO TO step 18 and then secure emergency boration.

Page 23 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2450 K/A Catalog Number: EPE 029 G2.1.20 Tier: 1 Group/Category: 1 SRO Importance: 4.6 10CFR Reference or SRO Objective: 55.43(b)(5)

ATWS:

Ability to interpret and execute procedure steps.

STP Lesson: LOT 504.28 Objective Number: 83681 DESCRIBE the indications and anticipated readings used to determine that all dilution paths are isolated.

Reference:

LOT 504.28, Rev 8, Lesson on Response to Nuclear Power Generation - ATWS and 0POP05-EO-FRS1 Rev 18 CIP Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because transitioning to 0POP05-EO-EO00 would eventually be performed but it is not the NEXT step. Emergency Boration must be secured first.

B: INCORRECT: Plausible because a high CET readings. However they are not elevated enough for entry into FRC2.

C: CORRECT: The conditions given would indicate that the reactor is being shutdown and that emergency boration can be secured. Going to step 18 is a CIP step.

D: INCORRECT: Plausible because it would be reasonable to believe that conditions warrant to continue with the procedure to the next step.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given conditions to determine what the next required procedure step needs to be taken.

Page 24 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2353 Last used on an NRC exam: Never SRO Sequence Number: 88 The Unit is at 100% power when I/C Maintenance has determined that a Shutdown Bank Control Rod is stuck full out and is untrippable.

What is the MAXIMUM amount of time that can be taken to enter HOT STANDBY?

A. 1 Hour B. 4 Hours C. 6 Hours D. 12 Hours Answer: C 6 Hours Page 25 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2353 K/A Catalog Number: APE 005 G2.2.40 Tier: 1 Group/Category: 2 SRO Importance: 4.7 10CFR Reference or SRO Objective: 55.43(b)(2)

Inoperable/Stuck Control Rod:

Ability to apply Technical Specifications for a system.

STP Lesson: LOT 503.01 Objective Number: 80056 Given a system scenario, DETERMINE the applicable Technical Specification and/or the Technical Requirements Manual (TRM) for the system and APPLY the specification(s).

Reference:

TS 3.1.3.1 Action a.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT: This distractor is credible because 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a time limit listed in this TS for other situation such as initially verifying shutdown margin.

B: INCORRECT: This distractor is credible because because 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is a time limit listed in this TS for other situations such as changing the High Neutron Flux Trip Setpoint.

C: CORRECT: T.S. 3.1.3.1 acition A - With only one stuck and untrippable control rod the unit must by in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D: INCORRECT: This distractor is credible because because 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a time limit listed in the TS for other situations such as determining shutdown margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of TS for the Control Rods.

Page 26 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2452 Last used on an NRC exam: Never SRO Sequence Number: 89 The Unit is at 100% power with the following conditions:

Steam A B C D Generators Steam Line 1.7E-2 uCi/cc 1.5E-2 uCi/cc 1.4E-2 uCi/cc 3.9E-1 uCi/cc Radiation Blowdown 2.7E-4 uCi/cc 2.4E-4 uCi/cc 2.3E-4 uCi/cc 4.6E-2 uCi/cc Radiation N-16 9.0 gpd 0.2 gpd 0.1 gpd 77.0 gpd Monitors Chemistry reports total current primary to secondary leak rate is 79 gpd.

The rate of increase is about 1 to 2 gpd/hr.

(1) Which Steam Generator(s) have tube leak(s)?

AND (2) What action will the Unit Supervisor take per 0POP04-RC-0004, Steam Generator Tube Leakage?

A. (1) ONLY Steam Generator D has a tube leak.

(2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. (1) ONLY Steam Generator D has a tube leak.

(2) Transition to 0POP04-TM-0005, Fast Load Reduction, and reduce power to <50% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C. (1) Steam Generators A and D have tube leaks.

(2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. (1) Steam Generators A and D have tube leaks.

(2) Transition to 0POP04-TM-0005, Fast Load Reduction, and reduce power to <50% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Answer: A (1) ONLY Steam Generator D has a tube leak.

(2) Transition to 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, and shutdown to Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Page 27 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2452 K/A Catalog Number: APE 037 AA2.01 Tier: 1 Group/Category: 2 SRO Importance: 3.4 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:

Unusual readings of the monitors; steps needed to verify readings.

STP Lesson: LOT 505.01 Objective Number: 92108 Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.

Reference:

LOT 505.01, Rev 7, Lesson Plans for Off-Normal Procedures. 0POP04-RC-0004, Rev 31, Steam Generator Tube Leakage add. 6 page 1 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: With the given conditions only SG D would have a tube leak and the correct procedure would be 0POP04-RC-0004.

B: INCORRECT: Plausible because if the believed to be large enough leak then tripping the reactor would be warranted however total leakage would have to be close to or greater than the capacity of a charging pump. About 200gpm.

C: INCORRECT: Plausible because of the elevated N-16 reading for SG A. However, the N-16 reading would be normal on SG A with a leak on SG D because SG A N-16 monitor would pick up some indications from SG D.

D: INCORRECT: Plausible because if the believed to be large enough leak then tripping the reactor would be warranted however total leakage would have to be close to or greater than the capacity of a charging pump. About 200gpm. Also, because of the elevated N-16 reading for SG A. However, the N-16 reading would be normal on SG A with a leak on SG D because SG A N-16 monitor would pick up some indications from SG D.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given readings for each SG and determine which SG(s) have tube leaks or tube ruptures and apply the correct procedure.

Page 28 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2454 Last used on an NRC exam: Never SRO Sequence Number: 91 The Unit is in Mode 6 with fuel movement in progress.

Subsequently:

Containment High Range Monitor RT-8050 brings in a Dark Blue condition on RM-11.

Containment High Range Monitor RT-8051 brings in a Magenta condition on RM-11.

(1) Which of the following is correct about the condition of RT-8050 and RT-8051?

AND (2) What action will the Unit Supervisor take?

A. (1) Both RT-8050 and RT-8051 are Non-Functional.

(2) Suspend fuel movement.

B. (1) Both RT-8050 and RT-8051 are Non-Functional.

(2) Suspend Polar Crane operations with loads over the Reactor Cavity.

C. (1) ONLY RT-8051 is Non-Functional.

(2) Suspend fuel movement.

D. (1) ONLY RT-8051 is Non-Functional.

(2) Suspend Polar Crane operations with loads over the Reactor Cavity.

Answer: A (1) Both RT-8050 and RT-8051 are Non-Functional.

(2) Suspend fuel movement.

Page 31 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2454 K/A Catalog Number: APE 061 G2.4.45 Tier: 1 Group/Category: 2 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(2)

ARM Systems Alarms:

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

STP Lesson: LOT 202.41 Objective Number: 68793 DESCRIBE the meanings of the colors on the RM-11 display.

Reference:

LOT 202.41 Lesson Plan on Radiation Monitors Rev 15 slide 51, and TRM 3.9.15.2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: If both RT-8050 and RT-8051 are non-functional then fuel movement must be suspended or an alternate method for monitoring a radioactive release must be available.

B: INCORRECT: Plausible because with some Refueling requirements not met an additonal action is to secure crane movement with loads over fuel. This is not the case, however, with containment high range area radiation monitors.

C: INCORRECT: Plausible because the magenta condition is the highest priority condition but both magenta and dark blue cause the radiation monitor to be non-functional.

D: INCORRECT: Plausible because the magenta condition is the highest priority condition but both magenta and dark blue cause the radiation monitor to be non-functional. Also, with some Refueling requirements not met an additonal action is to secure crane movement with loads over fuel. This is not the case, however, with containment high range area radiation monitors.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given refueling conditon to determine the correct conditon of the monitors and action to take.

Page 32 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2455 Last used on an NRC exam: Never SRO Sequence Number: 92 How many Reactor Coolant Pumps (RCPs) must be in operation in Mode 2 AND the action required if the condition is not met?

_____(1)_____ RCPs must be in operation.

_____(2)_____ is the action to be taken.

A. (1) Two (2) Place the required RCPs in operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. (1) Two (2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. (1) All (2) Place the required RCPs in operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> D. (1) All (2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Answer: D (1) All (2) Be in at least HOT STANBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Page 33 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2455 K/A Catalog Number: 003 G2.2.22 Tier: 2 Group/Category: 1 SRO Importance: 4.7 10CFR Reference or SRO Objective: 55.43(b)(2)

Reactor Coolant Pumps:

Knowledge of the limiting conditions for operations and safety limits.

STP Lesson: LOT 503.01 Objective Number: 80056 GIVEN a system scenario, DETERMINE the applicable Technical Specifications and/or Technical Requirements Manual (TRM) for the system and APPLY the specification(s).

Reference:

TS 3.4.1.1 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because two RCP in operation are required in Mode 3 with Reactor Trip Breakers closed. Also, plausible because the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is associated with the operablity of the loop only. Not if the loop is in operation. (RCP running) and is associated with Mode 3 Requiremeints.

B: INCORRECT: Plausible because two RCP in operation are required in Mode 3 with Reactor Trip Breakers closed.

C: INCORRECT: Plausible because the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is associated with the operablity of the loop only. Not if the loop is in operation. (RCP running) and is associated with Mode 3 Requiremeints.

D: CORRECT: Mode 2 requires all RCPs in operation or be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of the TS for Reactor Coolant Loops in Mode 2.

Page 34 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2449 Last used on an NRC exam: Never SRO Sequence Number: 93 A Unit trip has occurred from 100% power with the following conditions:

No AFW Pumps can be started All SG levels are 10% narrow range and lowering Based on these conditions which have been observed for 15 minutes, what emergency action level would the Emergency Director declare?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: C Site Area Emergency Page 35 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2449 K/A Catalog Number: W/E05 G2.4.41 Tier: 1 Group/Category: 1 SRO Importance: 4.6 10CFR Reference or SRO Objective: 55.43(b)(5)

Loss of Secondary Heat Sink:

Knowledge of emergency action level thresholds and classifications.

STP Lesson: LOT 803.14 Objective Number: SRO-74026 GIVEN an emergency condition and a copy of the emergency classification tables from 0ERP01-ZV-IN01, Emergency Classification, CLASSIFY the emergency condition.

Reference:

0ERP01-ZV-IN01, Rev 10, Emergency Classification, Addendum 4 Attached Reference

Attachment:

0ERP01-ZV-IN01, Rev 10, Emergency Classification, Addendum 4

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.

B: INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.

C: CORRECT: The student must be able to recognize that the given condition represents a loss of heat sink RED PATH. This causes a potential loss of both the RCS and Fuel Cladding. Therefore the correct classification is a Site Area Emergency.

D: INCORRECT: Any other classification is plausible because the student must be able to analyze the given conditions and apply the emergency classification tables and procedure.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given conditions and apply the emergency classification tables and procedure.

Page 36 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2457 Last used on an NRC exam: Never SRO Sequence Number: 94 The Unit has experienced a Small Break LOCA with the following conditions:

The operators performed a Reactor Trip and Safety Injection.

The Containment Normal Sump is pumping to the Waste Holdup Tank.

RCP A #1 Seal Leakoff Flow is 3 gpm and stable.

A CCW SURGE TK LVL LO annunciator is lit.

A TURB L.O. RSVR LVL LO-LO annunciator is lit.

Which condition should the Unit Supervisor address FIRST?

A. Ensure the Containment Sump penetration is isolated per Addendum 5 of 0POP05-EO-EO00, Reactor Trip or Safety Injection.

B. Trip RCP A per 0POP04-RC-0002, Reactor Coolant Pump Off Normal.

C. Restore the CCW Surge Tank per 0POP04-CC-0001, Component Cooling Water System Leak.

D. Restore Turbine Lube Oil Reservoir level per 0POP09-AN-07M3/D-5, TURB L.O.

RSVR LVL LO-LO.

Answer: A Ensure the Containment Sump penetration is isolated per Addendum 5 of 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Page 37 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2457 K/A Catalog Number: 013 G2.4.8 Tier: 2 Group/Category: 1 SRO Importance: 4.5 10CFR Reference or SRO Objective: 55.43(b)(5)

Engineered Safety Features Actuation:

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

STP Lesson: LOT 504.04 Objective Number: 92283 GIVEN a set of conditions and the occurrence of a Red, Orange or Yellow path CSF, STATE the action required per 0POP01-ZA-0018, EOP User's Guide.

Reference:

LOT 504.04, Rev 4, Lesson Plan and 0POP01-ZA-0018, EOP User's Guide Rev 21 page 18 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: Containment Isolation is an Engineered Safety Features system. Off-normal procedures can only be performed in parrallel with EOPs if they don't preclude taking action of an EOP. The failure of the Containment Penetration to isolate would have to be addressed first.

B: INCORRECT: Plausible because tripping a RCP when needed is very important.

C: INCORRECT: Plausible because CCW supports the cooling of the ECCS but is still addressed with an off-normal procedure.

D: INCORRECT: Plausible because ensuring the main turbine can coast down without damage is very important and is a Lessons Learned at STPNOC.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of how off-normal procedures work with EOPs.

Page 38 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2458 Last used on an NRC exam: Never SRO Sequence Number: 95 The Unit is at 100% power with the following conditions:

Train A and Train B RCFCs are running.

Train A CCW Pump is running.

Subsequently:

13.8KV Standby BUS G loses power.

The Unit Supervisor will direct actions of A. Addendum 1, Sequencer Loading Verification of 0POP04-AE-00001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus, and align B Train CCW to the RCFCs.

B. Addendum 3, Equipment Placed in Pull-To-Lock of 0POP04-AE-00001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus, and place B Train equipment in PTL.

C. 0POP02-HC-0002, Normal Containment Purge, Section 5 to start the Normal Containment Purge System.

D. 0POP02-HC-0001, Containment HVAC, Section 5 to re-start B train RCFCs.

Answer: A .Addendum 1, Sequencer Loading Verification of 0POP04-AE-00001, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus, and align B Train CCW to the RCFCs.

Page 39 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2458 K/A Catalog Number: 022 A2.04 Tier: 2 Group/Category: 1 SRO Importance: 3.2 10CFR Reference or SRO Objective: 55.43(b)(5)

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.

STP Lesson: LOT 202.33 Objective Number: 51319 STATE the power supplies for the RCB-HVAC systems.

Reference:

0POP04-AE-0001 Rev 44 Addendum 1 page 2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: With the given conditions, a loss of 13.8 KV Standby Bus 1G will cause the 4.16 KV ESF Bus 'B' to lose power. This in turn causes a loss of cooling water (chill water) flow to Train 'A' and 'B' RCFCs. ESF DG #12 will start and load the Bus with a Mode II signal. The Mode II signal starts CCW Pump 1B but does not restore any cooling water (CCW or chill water) flow to Train 'A' or 'B' RCFCs. Thus containment pressure will rise. The US will enter 0POP04-AE-0001 for the loss of B train 13.8 KV bus. Step 5 will have US perform Addendum 1 which will re-align CCW to RCFCs.

B: INCORRECT: Plausible because Addendum in the procedure that is performed. However, it is only performed if the sequencer is not functioning.

C: INCORRECT: Plausible but 0POP02-HC-0002, Normal Containment Purge, is only used in Modes 5 and 6 when the Unit is shutdown.

D: INCORRECT: Plausible because section 5 of 0POP02-HC-0001 does start the RCFC. However, with the given conditions the sequencer would have started the RCFCs.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given conditions to determine the impact of lossing cooling flow to containment and decide which procedure to use based on the Unit staying on line in Mode 1.

Page 40 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2459 Last used on an NRC exam: Never SRO Sequence Number: 96 The crew is performing 0POP05-EO-ES01, Reactor Trip Response, after a reactor trip from 100% power when the following conditions occur:

A loss of offsite power.

All ESF DGs started and loaded normally.

All SG Pressures 1200 psig and stable.

Containment pressure is 1.1 psig and slowly rising.

RCS pressure is 1925 psig and stable.

PZR level is 20% and stable.

The impact to the RCS due to these conditions is that there will be less cooling to the

_____(1)_____.

AND The Unit Supervisor will transition ______(2)_____ to mitigate the event.

A. (1) Reactor Vessel Head (2) back to 0POP05-EO-EO00, Reactor Trip or Safety Injection, B. (1) Reactor Vessel Head.

(2) to 0POP05-EO-ES02, Natural Circulation Cooldown, C. (1) Reactor Coolant Pump seals.

(2) back to 0POP05-EO-EO00, Reactor Trip or Safety Injection, D. (1) Reactor Coolant Pump seals.

(2) to 0POP05-EO-ES02, Natural Circulation Cooldown, Answer: B (1) Reactor Vessel Head.

(2) to 0POP05-EO-ES02, Natural Circulation Cooldown.

Page 41 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2459 K/A Catalog Number: 002 A2.03 Tier: 2 Group/Category: 2 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of forced circulation.

STP Lesson: LOT 504.25 Objective Number: 92228 STATE the basis for maximum cooldown rate associated with natural circulation cooldown.

Reference:

0POP05-EO-ES01, Reactor Trip Response, Rev 28, Page 25 of 26 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because rules of procedure usage would require a transition back to EO00 if Safety Injection conditions warranted. Incorrect because the conditions for safety injection are not satisfied. (PZR pressure, Contaianment pressure and SG pressure.

B: CORRECT: Loss of forced circulation causes less cooling to the Reactor Vessel Head which can cause a bubble to form in the head during cooldown. The conditions listed would have the US transition to 0POP05-EO-ES02.

C: INCORRECT: Plausible because cooling to the RCP seals is a concern to the RCS because the seals are a part of the RCS pressure boundary. If this were a loss of ALL AC power seal cooling would be lost and a LOCA through the seals is likely. Also, at the beginning of the procedure for a natural circulation cooldown it does have the operator check RCP seal cooling but it is only for a restart of the RCP if it were to become available. Also plausible because rules of procedure usage would require a transition back to EO00 is conditions warranted.

D: INCORRECT: Plausible because cooling to the RCP seals is a concern to the RCS because the seals are a part of the RCS pressure boundary. If this were a loss of ALL AC power seal cooling would be lost and a LOCA through the seals is likely. Also, at the beginning of the procedure for a natural circulation cooldown it does have the operator check RCP seal cooling but it is only for a restart of the RCP if it were to become available.

Question Level: H Question Difficulty 3 Justification:

The student must analyze the given conditions and have fundamental knowledge of the affect of losing forced circulation and to determine the correct procedure transition.

Page 42 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2460 Last used on an NRC exam: Never SRO Sequence Number: 97 A Loss of Coolant Accident has occurred with the following:

The crew is performing 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant.

RCS pressure is 550 psig and slowly lowering.

Pressurizer level is 90% and slowly rising.

RWST level is 70,000 gallons and slowly lowering.

Containment pressure is 9.6 psig and slowly rising.

All LHSI pumps are running.

The Unit Supervisor can validate the LHSI Control Board indications observed by determining the LHSI Pumps _____(1)_____ and the Unit Supervisor should transition to _____(2)_____.

(1) (2)

A. ARE providing flow into the RCS 0POP05-EO-ES13, Transfer to Cold Leg because PZR level is rising Recirculation B. are NOT providing flow into the RCS 0POP05-EO-ES13, Transfer to Cold Leg because RCS pressure is too high Recirculation C. ARE providing flow in to the RCS 0POP05-EO-FRZ1, Response to High because PZR level is rising Containment Pressure D. are NOT providing flow into the RCS 0POP05-EO-FRZ1, Response to High because RCS pressure is too high Containment Pressure Answer: B are NOT providing flow into the RCS because RCS pressure is too high -

0POP05-EO-ES13, Transfer to Cold Leg Recirculation Page 43 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2460 K/A Catalog Number: G2.1.45 Tier: 3 Group/Category: 1 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to identify and interpret diverse indications to validate the response of another indication.

STP Lesson: LOT 504.09 Objective Number: 81103 Fom memory STATE/IDENTIFY the criteria on the conditional information page of 0POP05-EO-EO10 to include operator response, initiating parameter(s) and values.

Reference:

0POP01-ZA-0018, EOP User's Guide Rev 21 page 31 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because PZR level rising can be an indication of added inventory but it can also be an indication of a PZR Vapor space leak.

B: CORRECT: With the given conditions RCS pressure is too high for LHSI flow to the RCS. When conditions exist for transfer to cold leg recirculation then that procedure is the priority.

C: INCORRECT: Plausible because PZR level rising can be an indication of added inventory but it can also be an indication of a PZR Vapor space leak. Also, containment pressure is high enough for a transition to the containment integrity safety function but the transfer to cold leg recirculation takes priority.

D: INCORRECT: Plausible because containment pressure is high enough for a transition to the containment integrity safety function but the transfer to cold leg recirculation takes priority.

Question Level: H Question Difficulty 3 Justification:

The student must be able to analyze the given condition and determine the response of the LHSI Pumps and the procedure that should be entered.

Page 44 of 50

Print Date 4/25/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2293 Last used on an NRC exam: 2014 SRO Sequence Number: 98 Due to an issue with pressure fluctuations, a local pressure recorder has been installed on ECW Pump 1B discharge line.

The pressure recorder is to remain installed for the next SIX weeks.

Which of the following SHALL be used to obtain authority and/or control the installation of the pressure recorder?

A. An entry in the Operator Aid Log in accordance with 0PGP03-ZO-0039, Operations Configuration Management.

B. Written instructions prepared in accordance with 0PGP03-ZA-0010, Performing and Verifying Station Activities.

C. A daily entry in the US Shift Turnover Checklist in accordance with 0POP01-ZQ-0022, Plant Operations Shift Routines.

D. Completion of a Temporary Modification Package in accordance with 0PGP03-ZO-0003, Temporary Modifications.

Answer: D Completion of a Temporary Modification Package in accordance with 0PGP03-ZO-0003, Temporary Modifications.

Page 45 of 50

Print Date 4/25/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2293 K/A Catalog Number: G2.2.14 Tier: 3 Group/Category: 2 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(3)

Knowledge of the process for controlling equipment configuration or status.

STP Lesson: LOT 802.10 Objective Number: SRO-10224 Given the description of a change to installed plant equipment, DETERMINE if the change consitutes a temporary modification in accordance with 0PGP03-ZO-0003.

Reference:

LOT 802.10 Lesson Plan and 0PGP03-ZO-0003, Temporary Modifications Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT: This distractor is credible because 0PGP03-ZO-0039 does track some types of configuration changes but not the installation of a pressure recorder.

B: INCORRECT: This distractor is credible because 0PGP03-ZA-0010 does authorize control of equipment where no procedure instructions exist but not the change to equipment.

C: INCORRECT: This distractor is credible because 0POP01-ZQ-0022 does give on shift personel a way to track issues in the plant but it does not give authority to make the actual changes.

D: CORRECT: The authority to install a pressure recorder to permanent plant equipment consitutes a temporary modification is controlled by 0PGP03-ZO-0003.

Question Level: F Question Difficulty 3 Justification:

The student must have knowledge of procedures that give authority to make changes to facility equipment.

Page 46 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2464 Last used on an NRC exam: Never SRO Sequence Number: 99 A fire breaks out behind CP-005 in the Unit 1 Control Room causing a Reactor Trip.

The Unit 1 Unit Supervisors priority would be to use _____(1)_____ to respond.

AND Using this procedure the Unit 1 Unit Supervisor will _____(2)_____.

(1) (2)

A. 0POP04-ZO-0001, Control notify Unit 2 to announce the Unit 1 Control Room Evacuation Room Evacuation B. 0POP04-ZO-0008, notify Unit 2 to announce the Unit 1 Fire/Explosion Fire/Explosion C. 0POP04-ZO-0001, Control direct the Unit 1 Secondary RO to announce the Room Evacuation Control Room Evacuation D. 0POP04-ZO-0008, direct the Unit 1 Secondary RO to announce the Fire/Explosion Fire/Explosion Answer: C 0POP04-ZO-0001, Control Room Evacuation - direct the Unit 1 Secondary RO to announce the Control Room Evacuation Page 47 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2464 K/A Catalog Number: G2.1.14 Tier: 3 Group/Category: 1 SRO Importance: 3.1 10CFR Reference or SRO Objective: 55.43(b)(5)

Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

STP Lesson: LOT 505.01 Objective Number: 92106 GIVEN plant conditions/symptoms, EVALUATE the conditions/symptoms and STATE whether or not the referenced procedure is to be used.

Reference:

0POP04-ZO-0001, Rev. 39, Page 3 of 223 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because for a fire where the control room is evacuated the unaffected Unit does NOT make an announcement of the evacuation but they do make the announcement of the fire/explosion because they would implement 0POP04-ZO-0008 per 0POP04-ZO-0001, step 15.

B: INCORRECT: Plausible because 0POP04-ZO-0008 is entered by the unaffected unit per 0POP04-ZO-0001, step 15.

C: CORRECT: For a fire that requires a control room evacuation the affected Unit's Secondary RO will announce the control room evacuation. The affected Unit's Unit Supervisor will contact the unaffected Unit to enter 0POP04-ZO-0008. 0POP01-ZA-0018, EOP Users Guide, states that if the control room is evacuated then 0POP04-ZO-0001 SHALL take precedence over all EOPs and 0POP04-ZO-0008. Also, for a fire in the control room that has caused unexpected equipment actuation then the Unit Supervisor is required to enter 0POP04-ZO-0001 first.

D: INCORRECT: Plausible because 0POP04-ZO-0008 is entered by the unaffected unit per 0POP04-ZO-0001, step 15. Also, in this case, 0POP04-ZO-0001 has the Unit Supervisor direct the Unit 1 secondary RO to announce the Control Room evacuation NOT the fire.

Question Level: F Question Difficulty 2 Justification:

The student must have fundamental knowledge of who makes announcements during a Fire/Explosion and Control Room Evacuation.

Page 48 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2491 Last used on an NRC exam: Never SRO Sequence Number: 100 In accordance with the STPNOC Operating License as listed in the UFSAR, conditions allowing release of liquid radwaste require a/an _____(1)_____ to alarm on high effluent radioactivity and automatically _____(2)_____ to stop the liquid radwaste release.

A. (1) Area Radiation Monitor (2) secure the effluent release pump B. (1) Area Radiation Monitor (2) divert the release to recirculation C. (1) Process Radiation Monitor (2) secure the effluent release pump D. (1) Process Radiation Monitor (2) divert the release to recirculation Answer: D (1) Process Radiation Monitor - (2) divert the release to recirculation Page 49 of 50

Print Date 4/13/2016 STP LOT-20.1 NRC SRO EXAM Exam Bank No.: 2491 K/A Catalog Number: 068 G2.2.38 Tier: 2 Group/Category: 2 SRO Importance: 4.5 10CFR Reference or SRO Objective: 55.43(b)(1)

Liquid Radwaste:

Knowledge of conditions and limitations in the facility license.

STP Lesson: LOT 203.11 Objective Number: 92083 State the purpose of the LWPS Effluent Radiation Monitor (RT-8038) and Discharge Divert Valve (FV-4077).

Reference:

USFAR Section 11.2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Plausible because are used to detect high radiation in system piping such as for Main Steam they are not used in any of th eprocess efflent release systems. Also, stopping the pump is plausible because this method is used to stop a release from the TGB Sump #1.

B: INCORRECT: Plausible because are used to detect high radiation in system piping such as for Main Steam they are not used in any of th eprocess efflent release systems.

C: INCORRECT: Stopping the pump is plausible because this method is used to stop a release from the TGB Sump #1.

D: CORRECT: Description in the UFSAR matches the conditions - Process Radiation Monitor that diverts the release to recirculation.

Question Level: F Question Difficulty 3 Justification:

The student must have fundamental knowledge of conditions and limitations of the facility license.

Page 50 of 50

Hand-outs for STP-2016-05 Written Retake Examination

Key for Hand-outs for STP-2016-05 Written Retake Examination (Note this was not given to applicants during exam but is included for the Adams package)

Attachment Reference for RO Question #10 The point of intesection (1175 MWe and 200 MVAR out) is within the 45 psig curve. All other options for question are outside of the 45 psig curve.

Attachment for RO Question #37 3 Pages

CV-MOV-0025 is located at 'G' making this a High Radiation Area and entry is not allowed on this RWP.

1. dpm/100 cm2 Bg Alpha Comment 1 < 1K N/A Floor 2 < 1K N/A Floor 3 < 1K N/A Pipe 4 < 1K N/A Floor 5 6K N/A Floor 6 15K N/A Pipe and Floor 7 15K N/A Floor 8 110K N/A Pipe and Floor 9 15K N/A Pipe and Floor

0ERP01-ZV-IN01 Rev. 10 Page 176 of 179 Emergency Classification Addendum 4 Fission Product Barrier Chart Page 1 of 1 KEY The conditions given indicate a Red Path on Heat Sink which gives a potential loss to fuel clad and a potential loss of the RCS. The potential loss of two barriers gives an SAE.