ML13310A615

2013-09-Draft Written Examination

ML13310A615
Person / Time
Site:	South Texas
Issue date:	08/07/2015
From:	Vincent Gaddy Operations Branch IV
To:	South Texas
References
Download: ML13310A615 (382)

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2212 Last used on an NRC exam: Never RO Sequence Number: 1 In accordance with 0POP01-ZA-0024, Enhanced Off Normal Operating Procedure Users Guide, which of the following is a responsibility of the Reactor Operators during the performance of an Off Normal Procedure?

A. Ensuring briefings are performed at appropriate transitions or pauses.

B. Predetermining manual actuation setpoints to be used when slowly degrading parameters are unrecoverable.

C. Performance of all immediate actions from memory.

D. Monitoring Conditional Information Pages for possible required actions.

Answer: C Performance of all immediate actions from memory.

Page 1 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2212 K/A Catalog Number: G2.4.11 Tier: 3 Group/Category: 4 RO Importance: 4.0 10CFR

Reference:

55.41(b)(10)

Knowledge of abnormal condition procedures.

STP Lesson: LOT 505.02 Objective Number: 92114 GIVEN a job position, STATE the responsibilities associated with that position as stated in 0POP01-ZA-0024, Enhanced Off Normal Operating Procedure Users Guide.

Reference:

POP01-ZA-0024, section 2.0 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because it is listed in the procedure as a responsibility for the SM and US.

B: INCORRECT: Credible because it is listed in the procedure as a responsibility for the SM and US.

C: CORRECT: This is a listed responsibility for the RO.

D: INCORRECT: Credible because it is a listed responsibility for the US.

Question Level: F Question Difficulty 3 Justification:

The applicant must have a knowledge of the Reactor Operators responsibilities durng Off Normal procedure performance.

Page 2 of 150

0POP01-ZA-0024 Rev. 2 Page 2 of 5 Enhanced Off Normal Operating Procedure Users Guide 1.0 Purpose The purpose of this procedure is to provide guidance for consistent implementation of STPEGS Enhanced Off-Normal Procedures (ONPs). Enhanced Off-Normal Procedures are identified by a statement in the Cover Page heading identifying the procedure as an Enhanced Off-Normal Procedure.

Enhanced ONP usage is intended to be standardized with 0POP01-ZA-0018, Emergency Operating Procedure Users Guide, dual column format rules of usage. The two users guides are to be combined in the future, therefore the user is referred to 0POP01-ZA-0018 when appropriate rather than duplicate information.

2.0 Responsibilities 2.1 Shift Manager (SM) is responsible for:

• The overall implementation of the Enhanced ONPs.

• Directing all plant personnel actions per the Enhanced ONPs.

• Maintaining a broad perspective of events during the implementation of the Enhanced ONPs.

• Predetermining manual actuation setpoints to be used when slowly degrading parameters are unrecoverable AND determining when to abandon attempts to recover degrading parameters and manually initiate safety systems.

• Overview and general direction of recovery actions in accordance with station procedures.

• Implementing the Emergency Plan as required by plant conditions.

• Ensuring briefings are performed at appropriate transitions or pauses to review plant status, Emergency Plan classification, etc.

0POP01-ZA-0024 Rev. 2 Page 3 of 5 Enhanced Off Normal Operating Procedure Users Guide 2.2 Unit Supervisor (US) is responsible for:

• Entering the Enhanced ONPs as warranted by plant conditions.

• Directing all plant personnel actions per the Enhanced ONPs.

• Monitoring Conditional Information Pages for possible required actions.

• Predetermining manual actuation setpoints to be used when slowly degrading parameters are unrecoverable AND determining when to abandon attempts to recover degrading parameters and manually initiate safety systems.

• Ensuring briefings are performed at appropriate transitions or pauses to review plant status, Emergency Plan classification, etc.

2.3 Reactor Operators (RO) are responsible for:

• The performance of all Enhanced ONP immediate actions from memory.

• The performance of all other Control Room actions as directed by the US.

• Communicating the completion or outcome of actions to the US.

• Communicating parameter trends to the US.

• Directing Reactor Plant Operators (RPO) actions per the Enhanced ONPs.

2.4 Reactor Plant Operators (RPO) are responsible for:

• The performance of all local actions as directed by the US or ROs.

• Communicating the completion or outcome of actions to the US or ROs.

• Communicating any abnormal local plant conditions to the US.

3.0 Procedure Hierarchy 3.1 Procedure use follows the normal convention for plant operating procedures, with the highest priority procedure being the primary procedure in use. The procedure priority, with the exception when stated otherwise, is:

3.1.1 Emergency Operating Procedures (EOP) (0POP05s) 3.1.2 Off Normal Operating Procedures (ONP) (0POP04s) 3.1.3 Annunciator Response Procedures (ARP) (0POP09s) 3.1.4 Normal Operating Procedures (OP) (0POP02s, 0POP03s)

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2207 Last used on an NRC exam: Never RO Sequence Number: 2 An audible alarm is received on the RM-11 system for RT-8035, FHB Exhaust.

Which of the following describes the correct response by the Reactor Operator to this condition?

A. Silence the alarm at the RM-11 console and then contact Health Physics to determine subsequent actions.

B Silence the alarm at the RM-11 console and then use POP04-RA-0001, Radiation Monitoring System Alarm Response, to determine subsequent actions.

C. Silence the alarm at CP-023 (RM-23) and then contact Health Physics to determine subsequent actions.

D. Silence the alarm at CP-023 (RM-23) and then use POP04-RA-0001, Radiation Monitoring System Alarm Response, to determine subsequent actions.

Answer: B Silence the alarm at the RM-11 console and then use POP04-RA-0001, Radiation Monitoring System Alarm Response, to determine subsequent actions.

Page 3 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2207 K/A Catalog Number: G2.3.13 Tier: 3 Group/Category: 3 RO Importance: 3.4 10CFR

Reference:

55.41(b)(12)

Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc STP Lesson: LOT 505.01 Objective Number: 92105 STATE the purpose of, and DESCRIBE the scope of the referenced procedure.

Reference:

0POP04-RA-0001, page 3 (Purpose)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because Health Physics are the experts for radiation related matters, however the operators are expected to use the off-normal procedure for system related responses.

B: CORRECT: The audible alarm is silenced with the SYSTEM ACK key on the RM-11 keyboard. A radiation monitor alarm on RM-11 is an entry condition for POP04-RA-0001 which then provides diagnostics/actions for the operator.

C: INCORRECT: Credible because CP-23 does have various controls and indications for this monitor, but no way to silence the alarm. Health Physics are the experts for radiation related matters, however the operators are expected to use the off-normal procedure for system related responses.

D: INCORRECT: Credible because CP-23 does have various controls and indications for this monitor, but no way to silence the alarm.

Question Level: F Question Difficulty 3 Justification:

The applicant requires a knowledge of the purpose of the off-normal procedure for radiation monitoring and an understanding of how the RM-11 system functions.

Page 4 of 150

0POP04-RA-0001 Radiation Monitoring System Alarm Rev. 29 Page 3 of 132

Response

STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED PURPOSE This procedure provides instructions and guidance to Control Room personnel to determine the cause of an ALERT or HIGH Area Process and Effluent Radiation Monitoring System alarm, evaluate the operability of the monitor or perform the initial required actions when the alarm is from the following panels and locations:

x RM-11, or RM-23A (when RM-11 is out of service) radiation monitoring panels.

x N16 Primary to Secondary Leak Monitoring System Alarms on the Plant Computer.

This procedure is applicable to all radiation monitors to provide the initial response to an alarm. However procedure 0PGP03-ZA-0078, Administration of the Radiation Monitoring System, provides the assignment of primary user and responsibility for the individual monitors.

The RM-11 system essentially provides the equivalent of an annunciator alarm display for the radiation monitors. The user may apply the rules for responding to an annunciator alarm when responding to alarms received on the RM-11 system.

SYMPTOMS OR ENTRY CONDITIONS

1. Alert, High or Status alarm on any of the following Radiation Monitors:

x RT-8010A and RT-8010B, Unit Vent Stack x RT-8011, RCB Atmosphere x RT-8012 and RT-8013, RCB Purge Exhaust x RT-8014, RT-8015, RT-8016, RT-8017, RT-8018, RT-8029 and RT-8030, MAB Ventilation x RT-8022, RT-8023, RT-8024 and RT-8025, Steam Generator Blowdown x RT-8027, Condenser Air Removal System x RT-8031, GWPS Inlet x RT-8032, GWPS Outlet x RT-8033 and RT-8034, EAB Air Intake x RT-8035 and RT-8036, FHB Exhaust Step 1. continued on next page

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2208 Last used on an NRC exam: Never RO Sequence Number: 3 Prior to a containment entry at power, Health Physics has requested the Containment Carbon Filter Units be placed in service.

Which of the following describes the purpose for this action?

A. The High Efficiency Particulate Air (HEPA) Filter contained within the unit will remove airborne radioiodine which is mainly an internal dose hazard.

B. The Charcoal Filter contained within the unit will remove airborne radioiodine which is mainly an internal dose hazard.

C. The High Efficiency Particulate Air (HEPA) Filter contained within the unit will remove airborne radioiodine which is mainly an external dose hazard.

D. The Charcoal Filter contained within the unit will remove airborne radioiodine which is mainly an external dose hazard.

Answer: B The charcoal filter contained within the unit will remove airborne radioiodine which is mainly an internal dose hazard.

Page 5 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2208 K/A Catalog Number: G2.3.12 Tier: 3 Group/Category: 3 RO Importance: 3.2 10CFR

Reference:

55.41(b)(12)

Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

STP Lesson: LOT 202.33 Objective Number: 92035 DESCRIBE the flowpath and STATE the functions for each of the following RCB-HVAC subsystems:

B. Containment Carbon Units

Reference:

LOT202.33 handout section 3.2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because the unit has a HEPA filter which is designed to remove radioactive particles, so the applicant must know the design basis for each.

B: CORRECT: The charcaol filter is there to remove iodine which is an internal hazard since it concentrates in the thyroid.

C: INCORRECT: Credible because the unit has a HEPA filter which is designed to remove radioactive particles, so the applicant must know the design basis for each. The radiation hazard is also credible since some radioactive isotopes are more an external hazard than internal.

D: INCORRECT: Credible because some radioactive isotopes are more an external hazard than internal.

Question Level: F Question Difficulty 3 Justification:

The applicant must have an understanding of the purpose of a charcoal filter and the hazards of radioactive iodine.

Page 6 of 150

LOT202.33.HO.01 Page 7 of 7 2.Containment Carbon Units Subsystem A.The Containment Carbon Unit Subsystem consists of two 50 percent capacity units. Each unit consists of the following components:

a.Prefilters b.HEPA filters (two banks - one upstream and one downstream of the carbon filters) c.Carbon filters (rechargeable type) d.Fans (two 100 percent capacity)

B.Prefilters The prefilters are provided upstream of the HEPA filters to protect them from coarse particles, and are designed for 85 percent efficiency.

C.HEPA Filters HEPA filters are provided to remove fine particulate matter from the airstream. HEPA filters are provided downstream of the carbon filters to collect any carbon fines which may be carried into the airstream from carbon filters.

D.Carbon Filter The carbon filters are provided to remove the airborne radioiodine from the airstream. These can be exhausted by solvent vapors and other halogenated materials.

E.Circulating Fans (2)

The circulating fans are centrifugal type with direct-drive, single-speed motors. Fans have totally enclosed, air-cooled motors, and are statically and dynamically balanced.

F.Design Criteria The Containment Carbon Units System shall reduce the airborne radioactivity levels in the containment atmosphere prior to allowing personnel access during normal plant operation or in advance of a scheduled reactor shutdown.

It is not required for safe shutdown.

Rated flow through each carbon unit is 10,000 cfm.

N/R6 08/04/13

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 19 Last used on an NRC exam: 2005 RO Sequence Number: 4 Unit 1 was at 100% power when a Large Break LOCA occurred.

Essential Cooling Water (ECW) Pump 1B was in Auto and started but the ECW Pump 1B discharge valve stopped at 80% open due to mechanical binding.

Which of the following describes the final condition of ECW Pump 1B and the reason why?

ECW Pump 1B A. remains running to continue cooling ESF support systems.

B. trips to protect pump casing from over pressure.

C. remains running to continue flow to the Screen Wash Booster Pump.

D. trips to protect pump shaft bearings from overheating.

Answer: A remains running to continue cooling ESF support systems.

Page 7 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 19 K/A Catalog Number: APE 062 AK3.02 Tier: 1 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the reason for the following responses as they apply to the Loss of Nuclear Service Water:

The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS.

STP Lesson: LOT 201.13 Objective Number: 91193 LIST all automatic functions, switch locations, switch positions, annunciators, local/remote functions, interlocks and permissive for the following: A) Traveling Screens, B) Screen Wash Booster Pump, C)

Screen Wash Valve, D) Strainers, E) Pumps and Motors, F) Discharge Valve, G) Sump, H) Blowdown Valve, I) Sump Pump and Motor

Reference:

9E-EW01-01 Rev 19, 9E-EW04-02 Rev 12 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT - An SI actuation will block the trip of an ECW pump from discharge valve position even though a partial loss of the system may occur.

B: INCORRECT - Credible because a partially closed discharge valve would result in a higher discharge pressure which would raise the stress on the pump casing.

C: INCORRECT - Credible because the ECW pump supplies water to the booster pump, however that is not the reason why the ECW pump trip is blocked during an SI.

D: INCORRECT - Credible because the pump bearings are cooled and lubricated by the discharge flow of the pump and a partially closed valve will reduce the discharge flow.

Question Level: F Question Difficulty 3 Justification:

Applicant must have fundamental knowledge of the reasons for ECW Pump logics.

Page 8 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 52 Last used on an NRC exam: 2005 RO Sequence Number: 5 Given the following Unit 2 conditions:

A Small Break LOCA has occurred SI has been reset Operators have just completed step 1 of 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant The Shift Technical Advisor reports the following:

RCS pressure:.........................1830 psig o

RCS Subcooling:....................60 F 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 Adverse containment conditions do NOT exist Which of the following actions should the operators perform?

A. Manually actuate SI and transition to 0POP05-EO-EO00, Reactor Trip or Safety Injection B. Transition to 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink C. Transition to 0POP05-EO-FRI2, Response to Low Pressurizer Level D. Transition to 0POP05-EO-ES11, SI Termination Answer: D Transition to 0POP05-EO-ES11, SI Termination Page 9 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 52 K/A Catalog Number: EPE W/E02 EA2.2 Tier: 1 Group/Category: 2 RO Importance: 3.5 10CFR

Reference:

55.41(b)(10)

Ability to determine and interpret the following as they apply to the (SI Termination): Adherance to appropriate procedures and operation within the the limitations in the facility's license and amendments.

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 21 Conditional Information Page Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible since SI reinitiation criteria would be met if adverse containment conditions existed.

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.

Page 10 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2219 Last used on an NRC exam: Never RO Sequence Number: 6 The following Unit 1 conditions exist:

MODE 5, cooling down to Mode 6 RHR trains A and B are is service in full cooling mode (RHR heat exchanger outlet valve is full open).

RHR train C is in standby.

A loss of offsite power subsequently occurs.

Without operator action, which of the following describes the final status of RHR?

A. No RHR Pumps are running.

B. Only RHR trains A and B are in service in the full cooling mode.

C. RHR trains A and B are in service in the full cooling mode and RHR pump C is running in recirculation mode.

D. All three RHR trains are in service in the full cooling mode.

Answer: A No RHR Pumps are running.

Page 11 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2219 K/A Catalog Number: APE 025 AK1.01 Tier: 1 Group/Category: 1 RO Importance: 3.9 10CFR

Reference:

55.41(b)(7)

Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation STP Lesson: LOT 201.41 Objective Number: 45253 List the equipment that starts on a Mode I, II, and III ASF Load Sequence signal.

Reference:

LOT201.41, ESF Load Sequencer, handout pages 18 and 21 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: The RHR pumps will strip off because of the LOOP and will not receive an auto start signal from the sequencer.

B: INCORRECT: Credible because safety related equipment does not have undervoltage trip relays, so once the bus is re-energized it might be reasonable to expect the pumps to restart (the pumps do receive a strip signal from the sequencer though).

C: INCORRECT: Credible because the sequencer does send a signal to the RHR pumps (a strip signal), however if the applicant believed it was a start signal, then all 3 pumps could be running with two cooling (as they were prior to the event) and one on recirc.

D: INCORRECT: Credible because the sequencer does send a signal to the RHR pumps (a strip signal), however if the applicant believed it was a start signal, then like other safety related equipment it might be running in its safety configuration (which for RHR would be full cooling mode).

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze the given conditions and using knowledge of the RHR and sequencer systems, determine the final status of the RHR pumps.

Page 12 of 150

LOT201.41.HO.01, Rev. 7 Page 18 of 31 Train B Bus Strip Signals Train C Bus Strip Signals

LOT201.41.HO.01, Rev. 7 Page 21 of 31 Train B Sequence Times in Seconds Function Mode I Mode II Mode III Incoming Breaker 480V Bus E1B1 N/A 1 1 Incoming Breaker 480V Bus E1B2 N/A 1 1 High Head Safety Injection Pump 1B 6 N/A 6 Low Head Safety Injection Pump 1B 10 N/A 10 Containment Spray Pump 1B 15 N/A 15 Reactor Containment Fan Cooler 11B 15 15 15 Reactor Containment Fan Cooler 12B 15 15 15 Component Cooling Water Pump 1B 20 20 20 Essential Cooling Water Pump 1B 25 25 25 Auxiliary Feedwater Pump 12 30 30 30 Control Room Makeup Fan 11B 35 N/A 35 CR/EAB Emergency HVAC 35 35 35 Standby Ess. Chiller and CHW pump 11B 35 35 35 Containment Spray Pump 1B 40 N/A 40 Essential Chiller 12B 240 240 240 Sequence Complete 280 280 280 Containment Spray Pump 1B Permissive 15 N/A 15 Containment Spray Pump 1B Timer 62 17 N/A 17 Containment Spray Pump 1B Permissive 40 N/A 40

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 381 Last used on an NRC exam: Never RO Sequence Number: 7 Which of the following is the BASIS for depressurizing intact Steam Generators to 355 psig at the maximum controllable rate during performance of 0POP05-EO-EC00, Loss of All AC Power?

A. To ensure that a heat sink is maintained due to loss of control of the SG PORVs.

B. To maximize Operator control of secondary pressure.

C. To minimize RCS inventory loss through the RCP seals.

D. To prevent challenging the pressurizer safety valves.

Answer: C To minimize RCS inventory loss through the RCP seals Page 13 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 381 K/A Catalog Number: EPE 055 EK3.02 Tier: 1 Group/Category: 1 RO Importance: 4.3 10CFR

Reference:

55.41(b)(5)

Knowledge of the reasons for the following as they apply to the Station Blackout: Actions contained in EOP for loss of offsite and onsite power STP Lesson: LOT 504.22 Objective Number: 82073 Given a copy of a step caution or note from 0POP05-EO-EC00, STATE/IDENTIFY its basis, its purpose and the result of a failure to comply with its requirements.

Reference:

WOG ERG EC-0.0, R2, page 118 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: Incorrect - Credible because although the hydraulic pump needs AC power and will not be working, control from the control room is still possible until accumulator pressure is used up.

B: Incorrect - Credible since maximizing operator control of secondary pressure is desirable in a loss of all AC power, however it is the basis for maintaining a faulted or ruptured SG isolated.

C: Correct - To minimize RCS inventory loss through the RCP seals D: Incorrect - Credible if the PZR PORVs relied on AC power to operate, however the PZR PORVs will still be available to prevent challenging the pressurizer safeties since they are pressure actuated and DC powered.

Question Level: F Question Difficulty 3 Justification:

The applicant must recall the basis for the rapid depressurization of the RCS during a loss of all AC.

Page 14 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 524 Last used on an NRC exam: 1995 RO Sequence Number: 8 A reactor trip has occurred on Unit 2.

During the performance of the Immediate Operator actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection, the Reactor Operator notes that the Reactor Trip Breakers are both open, but 4 Control Rods are indicating 18 steps withdrawn.

The crew transitions to 0POP05-EO-ES01, Reactor Trip Response.

Which of the following describes the minimum amount of Boric Acid in gallons that must be added to the RCS and the BASIS for the Boric Acid addition?

Emergency Borate A. 3760 gallons or until RCS boron concentration is determined to be greater than 2800 ppm to lower upper range flux to less than 5%.

B. 3760 gallons or until RCS boron concentration is determined to be greater than 2800 ppm to account for the reactivity worth of the stuck rods.

C. 14400 gallons or until RCS boron concentration is determined to be greater than 2800 ppm to lower upper range flux to less than 5%.

D. 14400 gallons or until RCS boron concentration is determined to be greater than 2800 ppm to account for the reactivity worth of the stuck rods.

Answer: B 3760 gallons or until RCS boron concentration is determined to be greater than 2800 ppm to account for the reactivity worth of the stuck rods.

Page 15 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 524 K/A Catalog Number: APE 024 AK3.01 Tier: 1 Group/Category: 2 RO Importance: 4.1 10CFR

Reference:

55.41(b)()

Knowledge of the reasons for the following responses as they apply to the Emergency Boration: When emergency boration is required STP Lesson: LOT 504.06 Objective Number: 81674 Given a step, note, or caution from 0POP05-EO-ES01, STATE/IDENTIFY the basis for the step, note or caution and the basis for the action to include the action itself, its purpose and result

Reference:

0POP05-EO-ES01, step 4 Attached Reference

Attachment:

0POP05-ES-ES01rev.26, page 5 NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because adding acid would lower power, however if power was actually

>5%, a different procedure would be in use.

B: CORRECT - 940 gallons per rod of emergency boration is required for each rod stuck at or below 18 steps (940x4=3760). The basis is to account for stuck rod reactivity worth and ensure adequate shutdown margin.

C: INCORRECT - Credible since this is the amount of boration that would be required if the applicant did not read the procedure closely enough and used the per rod value of boric acid when the rod is more than 18 steps withdrawn (18 steps is the transition point). The basis is credible because adding acid would lower power, however if power was actually >5%, a different procedure would be in use.

D: INCORRECT - Credible since this is the amount of boration that would be required if the applicant did not read the procedure closely enough and used the per rod value of boric acid when the rod is more than 18 steps withdrawn (18 steps is the transition point).

Question Level: H Question Difficulty 3 Justification:

The applicant must determine the correct amount of boric acid to add from the given conditions and recall the basis for the action.

Page 16 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2160 Last used on an NRC exam: Never RO Sequence Number: 9 Chemistry has reported a rise in reactor Coolant activity and the Unit Supervisor has entered 0POP04-RC-0001, High Reactor Coolant System Activity.

Which of the following correctly describes an action that should be taken and the reason for the action?

A. Start a second Centrifugal Charging Pump and place all letdown orifices in service to maximize purification flow through the in service Mixed Bed Demineralizer.

B. Start a second Centrifugal Charging Pump and place all letdown orifices in service to maximize filtration by the reactor coolant filter.

C. Place a Cation Bed Demineralizer in service to remove lithium which reduces RCS pH and minimizes the chances of an RCS crud burst.

D. Place a Cation Bed Demineralizer in service to maximize effective purification.

Answer: D Place a Cation Bed Demineralizer in service to maximize effective purification.

Page 17 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2160 K/A Catalog Number: APE 076 AK3.06 Tier: 1 Group/Category: 2 RO Importance: 3.2 10CFR

Reference:

55.41(b)(5)

Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity:

Actions contained in EOP for high reactor coolant activity.

STP Lesson: LOT 505.01 Objective Number: 92110 Given a precaution, note, or step(s) and the context in which it is used from the referenced procedure, DESCRIBE its basis and any applicable limits.

Reference:

POP04-RC-0001 step 7 basis Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because raising letdown flow is a procedural action and doing so will raise deminerilizer flow and help lower activity. However, starting and running a second charging pump would be required with all orifices in service and would be a procedure violation.

B: INCORRECT: Credible because raising letdown flow is a procedural action and doing so will raise flow through the filter and help lower activity. However, starting and running a second charging pump would be required with all orifices in service and would be a procedure violation.

C: INCORRECT: Credible because the cation demineralizer is used during normal operation to lower lithium concentration in the RCS, however a procedural caution warns that doing so may initiate a crud burst (not minimize).

D: CORRECT: Placing a cation bed in service will aid in removing fission products.

Question Level: F Question Difficulty 3 Justification:

The applicant musy have knowledge of procedural requirements and basis.

Page 18 of 150

0POP04-RC-0001 High Reactor Coolant System Rev. 10 Page 18 of 28 Activity Addendum 1 Basis Basis Page 10 of 19 STEP DESCRIPTION FOR 0POP04-RC-0001 STEP 7.0 STEP: CHECK Cation Demineralizers In Service PURPOSE: To inform the operator to place cation demineralizers in service to facilitate reducing RCS activity if not already in service in Step 4.0.

BASIS: Adequate resin bed capacity will ensure maximum effective purification.

ACTIONS: Place cation demineralizers in service.

INSTRUMENTATION: N/A CONTROL/EQUIPMENT: N/A KNOWLEDGE: N/A This Procedure is Applicable in Modes 1-5

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1296 Last used on an NRC exam: 2005 RO Sequence Number: 10 Unit 2 is operating in Mode 1 with the following Component Cooling Water (CCW) Pump lineup:

CCW Pump 2A - running CCW Pump 2B - standby CCW Pump 2C - tagged out for maintenance Subsequently:

A failure of CCW Pump 2A discharge valve causes it to drift partially closed.

CCW Header pressure drops to 85 psig.

Normal letdown has remained in service.

Which of the following describes the system response an operator should expect to observe?

A. Letdown flow diverts to the Recycle Holdup Tank (RHUT)

B. Initial rise then return to normal in letdown temperature downstream of the Letdown Heat Exchanger C. Initial rise then return to normal in seal water return temperature downstream of the Seal Water Heat Exchanger D. Letdown flow diverts to the Reactor Coolant Drain Tank (RCDT)

Answer: B Initial rise then return to normal in letdown temperature downstream of the Letdown Heat Exchanger Page 19 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1296 K/A Catalog Number: APE 026 AA1.06 Tier: 1 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(7)

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

Control of flow rates to components cooled by the CCWS STP Lesson: LOT 201.12 Objective Number: 5213 Given a plant or system condition, PREDICT the operation of the Component Cooling Water System.

Reference:

LOT201.06.HO.01, pages 18 and 29; LOT201.12.HO.01, page 15 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because letdown flow will divert to the RHUT but only on high VCT level not high letdown temperature caused by reduced CCW flow and pressure.

B: CORRECT: Lowering CCW flow will cause letdown temp to rise, then as letdown heat exchanget TCV-4494 opens, the temperature will return to normal. NOTE: CCW Pumps would not change status because a header pressure of 85 psig would not be low enough to start a standby CCW pump.

C: INCORRECT: This distractor is credible because the Seal Water Heat Exchanger is cooled by CCW but a TCV does not control CCW flow to the Seal Water Heat Exchanger so in this situation, temperature will rise and remain at the higher temperature.

D: INCORRECT: This distractor is credible because CVCS letdown can be diverted to the RCDT but only if Excess Letdown is in service.

Question Level: H Question Difficulty 3 Justification:

Requires ability to determine the effect the reduced system flow will have on heat transfer in the CVCS letdown heat exchanger and the response of the downstream temperature control valve.

Page 20 of 150

LOT201.06 Rev. 14 Page 18 of 51 to fail open on loss of air or electrical power; however, if it were to fail shut, a manual bypass valve is provided in order to continue letdown operation. TCV-381A, the return from the BTRS Reheat Heat Exchanger, fails closed on loss of air or electrical power.

The reactor coolant purification pump is used to circulate water from the RHR system during cold or refueling shutdown for purification. The purification pump capacity of 450 gpm provides maximum purification during shutdown low pressure operations. The purification pump is a centrifugal pump designed for 600 psig and 400F.

During this mode of purification, both mixed beds and the RHR return path downstream of the reactor coolant filter is used. The purification pump has "STOP, NORMAL, START" control switch on CP-004 with a P-T-L position.

Over pressure protection is provided by an alarm which actuates on a R.C. purification pump discharge pressure of 540 psig and by a relief valve which relieves to the VCT at 600 psig. The purification pump is located on the 10 foot elevation of the Mechanical Auxiliary Building (MAB).

The R.C. Purification Pump Discharge Valve (HCV-133) is used to control the pump flow rate. The hand controller is located on CP-004.

The letdown heat exchanger (Figure 5) is U-tube type heat exchanger with letdown through the tubes being cooled by component cooling water on the shell side. The component cooling water control valve automatically controls the temperature of the letdown flow from the letdown heat exchanger at 115F. If the letdown outlet temperature increases above 120F, a letdown high temperature alarm will sound on the main control board. The letdown heat exchanger is located on the 10 foot elevation of the MAB.

Low Pressure Letdown Control Valve (PCV-135) maintains the pressure downstream of the letdown orifices automatically at ~350 psig. Maintaining this backpressure prevents the letdown flow from "flashing to steam" as it passes through the orifice valves.

Flashing would cause excessive erosion of the orifices. During solid plant operations,,

the low pressure letdown control valve's function is to maintain the pressure of the RCS.

PCV-135 fails open on a loss of control air to the valve, or a loss of control voltage to the air controller. A manually operated bypass valve is installed around PCV-135 to allow continued operation with normal letdown flow if the valve should stick or fail. A letdown high pressure alarm at 500 psig is provided to warn the operator. The controller for PCV-135 is located on CP-004.

Low pressure Letdown Line Relief Valve will relieve to the VCT when the pressure in the low pressure letdown line increases to 300 psig.

Letdown Flow is read on FI-132 on CP-004. Letdown flow is sensed by FT-132 downstream of PCV-135. This instrument actuates a letdown Hi/Lo flow alarm at 260 gpm/70 gpm respectively.

Letdown flow is continuously sampled by a liquid Radiation Monitor (RIT-8039) downstream of PCV-135. The differential pressure across the orifice used to measure

LOT201.06 Rev. 14 Page 29 of 51 The Seal Water Leakoff Header Relief Valve is located inside the containment. It is set to relieve to the pressurizer relief tank (PRT) at a pressure of 150 psig. Seal water return header containment isolation valves (MOV-077 and MOV-079) close on a Phase A isolation signal and can be operated from CP-004.

The Seal Water Return Filter is a 25 micron disposable filter and has a maximum designed flow of 250 gpm. The filter is replaced if the pressure across it increases to 20 psid or if it has a radiation level of 5 R/Hour on contact. A manual bypass valve is provided to allow continued operation during filter replacement.

The Seal Water Heat Exchanger (see Figure 12) receives flow from the RCP #1 seal leakoff (~12 gpm), the excess letdown heat exchanger (~20 gpm) and the centrifugal charging pumps recirculation flow (~60 gpm). The seal water heat exchanger is cooled by component cooling water. A relief valve on the inlet to the heat exchanger is set to relieve to the VCT at a pressure of 150 psig. The outlet of the seal water heat exchanger is normally directed to the suction of the charging pumps, but can also be directed back to the VCT through a spray nozzle. The seal water return flow would be directed to the VCT if it became necessary to maintain the hydrogen concentration of the RCS when the normal letdown path is not available. A manual bypass valve and line is provided around the heat exchanger for use during maintenance or leak conditions.

Excess Letdown System (Figure 11)

The Excess Letdown System is used if the normal letdown path is inoperable, to maintain the flow balance between the letdown and charging systems or for additional letdown when necessary. Excess letdown is taken from the reactor coolant loop 4 intermediate leg upstream of the RCP and flows to the excess letdown heat exchanger.

Excess Letdown Isolation Valves (MOV-082 and MOV-083) are motor operated and controlled from CP-004 by "CLOSE, NORMAL, OPEN" spring return to "NORMAL" switches. Valve position status lamps are located above the switch. These valves fail "as is" on loss of electrical power. On a SI or Phase A signal these valves must be closed immediately by the operator to prevent damage to the CCW side of the Excess Letdown Heat Exchanger and loss of reactor coolant through the Seal Water Leakoff Header Relief Valve since there are no automatic closure signals.

The Excess Letdown Heat Exchanger is a stainless steel, tube and shell heat exchanger, cooled by Component Cooling System. It reduces the letdown water temperature to approximately 160F. A high temperature alarm sound on CP-004 if the temperature rises to 175F. Letdown flows through the tube side and component cooling water flows through the shell side. Excess letdown heat exchanger outlet pressure and temperature can be read on CP-004. The excess letdown heat exchanger is located inside containment in the northeast section of the 52 foot elevation.

The Excess Letdown Flow Control Valve (HCV-227) is used to control letdown flow to a maximum of approximately 20 gpm. HCV-227 is a motor operated valve and fails as is.

Caution should be observed when establishing excess letdown flow because of the effect it can have on #1 seal water leakoff backpressure. The seal return flow shares a common line. Increased backpressure could cause a change in the #1 seal water flow.

LOT201.12.HO.01 Rev. 13 PAGE 15 OF 40 The pump net positive suction head requirement for all modes of plant operation is satisfied by locating the pump lower than the surge tank.

The surge tank centerline is at elevation 68' 6" while the pump centerline is at elevation 13' 9-1/4". The CCW pump motors are powered from the 4.16 KV Class 1E power distribution system.

4.4.1 For normal operation, the CCW trains are selected to be off, running or in standby for automatic start, along with the respective ECW train, by CCW/ECW Mode Selector Switches located on CP-002. Normally one train is in "OFF", one is in "Standby" and one is in "RUN".

A. When selected for "Standby", low pressure in the CCW common header (76 psig) or a low pressure in the other two ECW loops (30 psig) initiates automatic startup of the pump after a 15 second delay (if LOOP or SI does not exist). The delay in auto starting allows for switching the operating pumps during normal operations. Automatic startup of the corresponding ECW pump will also occur simultaneously.

B. When selected to "Run", the CCW and ECW pump for that train will start unless a LOOP or SI signal is present.

C. If two of the three selector switches are in "Off", a "Standby Train Not Selected" alarm will annunciate.

4.4.2 Transfer of control and indication for the CCW pumps from the transfer switch panels to the main control room is provided by "Local/Remote" switches located on the transfer switch panels in the switchgear rooms.

A. When in "local", a BYPASS INOP alarm is sounded and all automatic functions are disabled.

4.4.3 Individual STOP/AUTO/START control switches are provided on CP-002 and the transfer switch panels along with running (red) and stopped (green) pump status lights.

The control room switch also has a Pull-To-Lock (PTL) feature that will stop the pump under all conditions.

A. The status lights indicate only at the location selected by the Local/Remote switch.

4.4.4 Auto starts

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1577 Last used on an NRC exam: 2010 RO Sequence Number: 11 An operator action of 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS, is to Ensure 480V LC 1K1 (2K1) and 1L1 (2L1) feeder breakers open.

This step will de-energize power to the....

A. Rod Drive MG Set motors. Opening only one of the breakers will cause a reactor trip.

B. Rod Drive MG Set motors. Both breakers must be opened to cause a reactor trip.

C. Reactor Trip Breaker shunt trip coils. Opening only one of the breakers will cause a reactor trip.

D. Reactor Trip Breaker shunt trip coils. Both breakers must be opened to cause a reactor trip.

Answer: B Rod Drive MG Set motors. Both breakers must be opened to cause a reactor trip.

Page 21 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1577 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 interrrelations between components following an ATWS: Breakers, relays, and disconnects STP Lesson: LOT 201.18 Objective Number: 3069 IDENTIFY major components, system interfaces, interlocks and relative location of components and instrumentation by drawing and labeling a block diagram of the Rod Control System.

Reference:

LOT 201.18 PowerPoint slide 19 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because only one trip breaker must be open to trip the reactor, but both MG sets must be de-energized to initiate a reactor trip.

B: CORRECT: The Load Centers supply power to the MG motors. Either MG set providing power to the Rod Control System will be sufficient to power the rod drives (design redundancy) so both MG must be de-energized as stated.

C: INCORRECT: Credible because the trip breakers will open if the shunt trip coil is de-energized, however it is powered from a different source. Part two is credible because only one trip breaker must be open to trip the reactor, but both MG sets must be de-energized to initiate a reactor trip.

D: INCORRECT: Credible because the trip breakers will open if the shunt trip coil is de-energized, however it is powered from a different source.

Question Level: F Question Difficulty 3 Justification:

Must know the distribution for rod drive power including the design redundancy that must be accounted for to perform a reactor trip.

Page 22 of 150

1. 1 480 VAC LC 1K1 480 VAC LC 1L1

1. 2 RTS RTR BYS BYR 19

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1651 Last used on an NRC exam: 2007 RO Sequence Number: 12 A liquid release is in progress from a Waste Monitor Tank (WMT).

Which of the following correctly describes operation of the system if the Liquid Release Rad Monitor, RT-8038 reaches its HIGH alarm setpoint?

Liquid Waste Discharge Valve, FV-4077, should A. CLOSE to stop flow from the WMT. If the valve fails to close, the Control Room operator must manually close the valve from the RM-11 console.

B. RE-POSITION to recirc the contents of the WMT. If the valve fails to re-position, the Control Room operator must manually re-position the valve to recirc from the RM-11 console.

C. CLOSE to stop flow from the WMT. If the valve fails to close, a Plant Operator will have to be dispatched to close the valve using handswitch on Rad Waste Controlroom panel.

D. RE-POSITION to recirc the contents of the WMT. If the valve fails to re-position, a Plant Operator will have to be dispatched to re-position the valve using handswitch on Rad Waste Controlroom panel.

Answer: D Liquid Waste Discharge Valve, FV4077, will re-position to recirc the contents of the WMT. If the valve fails to re-position, a Plant Operator will have to be dispatched to re-position the valve using handswitch on Rad Waste Controlroom panel.

Page 23 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1651 K/A Catalog Number: APE 059 AA1.01 Tier: 1 Group/Category: 2 RO Importance: 3.5 10CFR

Reference:

55.41(b)(13)

Ability to operate and/or monitor the following as they apply to the Accidental Liquid Radwaste Release:

Radioactive-liquid monitor STP Lesson: LOT 202.41 Objective Number: 92122 LIST the initiating condition and resultant automatic action for the PERMS radiation monitors associated with the following systems: A. Boron Recycle System, B. Gaseous Waste Processing System, C.

Liquid Waste, Processing System, D. Turbine Generator Building Sump and Drain System, E.

Condensate Polishing System, F. Steam Generator Blowdown System, G. Containment Building, H.

Electrical Auxiliary Building and Control Room Envelope HVAC, I. Fuel Handling Building Ventilation System

Reference:

LOT202.41 student handout page 26 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Valve position is credible because the desire would be to stop the discharge (which closing would). Valve control is credible since the monitor is displayed on the RM-11 console in the CR and has some control functions but only for the monitor, the valve cannot be operated from the panel.

B: INCORRECT - Credible because the monitor is displayed on the RM-11 console in the CR and has some control functions but only for the monitor, the valve cannot be operated from the panel.

C: INCORRECT - Credible because the desire would be to stop the discharge (which closing would).

D: CORRECT- FV4077 repositions to recirc tank if it fails to actuate a plant operater will have to go to local hand switch.

Question Level: H Question Difficulty 3 Justification:

The applicant requires a knowledge of the automatic actions associated with the affected radiation monitor and where the valve can be operated from.

Page 24 of 150

LOT202.41.HO01.REV15 PAGE 26 OF 45 Turbine Generator Building (TGB) Sump 1 monitor Failed Fuel Monitor (CVCS)

Condensate Polishing System Monitor Gaseous Waste Processing System (GWPS) Inlet Monitor MAB HVAC (7)

GWPS Discharge Monitor Main Steam Line Monitors (4) (Class 1E)

SG Blowdown Monitors (4) (Class 1E)

PERMS CONTROL FUNCTIONS The PERMS monitor control functions are also outlined in Table 1 for those monitors possessing control functions. Some examples of typical control functions from these monitors can be shown as follows:

Liquid Waste Processing System (LWPS) - RT-8038 In the event of high radiation in this system, or a monitor failure condition a diversion valve will send the liquid effluent from the system back to the waste monitor tanks. WL-FV-4077 Boron Recycle System - RT-8037 Monitor serves to divert flow back to the BRS Evaporator Feed Demineralizers on a high radiation signal or monitor failure signal. BR-RCV-4204 Gaseous Waste Processing System (GWPS)- RT-8032 High radiation as measured at the GWPS discharge or a monitor failure condition results in the shutdown of the GWPS. The High Rad or Monitor Failure sends a signal to the GWPS shutdown circuitry to close the discharge valve, the inlet valve , the BRS vent and secure the Bellows Compressor.

Turbine Generator Building Sump and Drain System - RT-8041 High radiation at the sump pump discharge or a monitor failure condition will stop the sump pump.

Condensate Polishing System (CPS) - RT-8042 High radiation at the discharge of the CPS to the neutralization basin or a monitor failure condition will close this discharge valve. CP-FV-5804 Steam Generator Blowdown (SGBD) System RT-8043 High radiation in the steam generator blowdown liquid or a monitor failure condition closes the SGBD discharge to the neutralization basin. SB-FV-5019 closes.

Containment Building Ventilation System RT-8012 & 8013 -High radiation in the RCB Purge System Exhaust sends a signal to the Solid State Protection System (SSPS) for Containment Ventilation Isolation (CVI). (Normal and supplementary purge)

Electrical Auxiliary Building and Control Room Envelope (HVAC) - RT-8033 & 8034 High radiation level at the EAB air intake initiates Control Room/EAB emergency ventilation.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1733 Last used on an NRC exam: 2007 RO Sequence Number: 13 Unit 1 is at 15% reactor power during a plant shutdown.

The Main Generator is at 210 MWe.

A large transient on the grid causes switchyard frequency to lower to 56.5 Hz.

Assuming no operator action, which of the following correctly identifies the status of the reactor coolant pump breakers and reactor trip breakers following the grid transient?

Reactor Coolant Pump Reactor Trip Breakers Breakers A. Open Open B. Open Closed C. Closed Closed D. Closed Open Answer: A Open, Open Page 25 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1733 K/A Catalog Number: EPE 007 EK2.02 Tier: 1 Group/Category: 1 RO Importance: 2.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the interrelations between a reactor trip and the following: Breakers, relays and disconnects.

STP Lesson: LOT 201.20 Objective Number: 26026 Given a description of plant conditions DETERMINE if an automatic reactor trip signal would be generated.

Reference:

LOT201.20, Handout page 41 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT - The reactor trip setpoint for underfrequency is 57.2 Hz. If reactor power is greater than P-7, then the underfrequency trip will open the reactor trip breakers and the reactor coolant pump breakers. The same signal causes both sets of breakers to trip.

B: INCORRECT - This distractor is credible because the RX Trip breakers would not open if Reactor is less than 10% power.

C: INCORRECT - This distractor is credible if the applicant believes the frequency did not lower enough to cause a trip of the reactor or RCPs.

D: INCORRECT - This distractor is credible if the applicant is aware of the underfrequency reactor trip, but did not realize the RCPs also receive a trip signal (they do not trip on undervoltage, althought the reactor still trips).

Question Level: H Question Difficulty 3 Justification:

From the given conditions, the applicant must use their knowledge of reactor trip setpoints (underfrequency), permissives (P-7) and interlocks (tripping of all RCPs) to determine the correct response.

Page 26 of 150

LOT201.20.HO.01 Rev. 17 PAGE 41 OF 69

4. Time delays are incorporated to prevent spurious trips from momentary electrical transients

a. Delay is set at < .4 Seconds from trip of two or more RCP Bus circuit breakers to signal reaching reactor trip breakers.

Tech Specs require < 1.2 seconds.

5. Alarms and annunciators

a. Located on CP-005

b. RCP Bus Undervolt E. Reactor coolant pump (TP .42)

1. Provide reactor core protection against DNB as a result of underfrequency in more than one RCP

a. Loss of forced reactor coolant flow

2. < 57.2 HZ; 2/4 UF sensors and not blocked (i.e. 2/4 Pumps having UF condition)

3. Automatically blocked below P-7

4. Will also trip open all RCP breakers to allow for coastdown

5. Time delays are incorporated to prevent spurious trips from momentary electrical transients

a. Delay is set at < 0.6 Seconds from time underfrequency trip setpoint is reached to signal reaching reactor trip breakers

6. The signal to trip open all RCP breakers is blocked if the RCP Undervoltage relay is actuated.

7. Alarms and annunciators

a. Located on CP-005

b. RCP Bus Underfrequency RX Trip

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1820 Last used on an NRC exam: 2011 RO Sequence Number: 14 Given the following:

Unit 1 is operating at full power.

480V LC E1A2 TRBL alarm occurs.

480V Load Center E1A2 Bus Volts = 0 volts Annunciator 125 VDC SYSTEM E1A11 TRBL alarms.

Channel 1 BATT CUR indicates 30 amps discharge.

Assuming the plant responds as designed and without operator action, A. DP001 is now powered from its Voltage Regulating Transformer.

B. DP1201 is now powered from its Voltage Regulating Transformer.

C. Bus E1A11 is being powered from its respective ESF Battery.

D. Bus E1A11 is being powered from its Standby Battery Charger.

Answer: C Bus E1A11 is being powered from its respective ESF Battery.

Page 27 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1820 K/A Catalog Number: APE 058 AA1.03 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 DC Power: Vital and battery bus components.

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

Reference:

LOT 201.37 PowerPoint slide 14, LOT201.38 PowerPoint slide 93 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because the 480v inputs to the 120v DP panel systems are split between the two load centers and the applicant may believe the loss of the 480v input will cause a loss of the normal supply, causing it to swap to the voltage regulating transformer. However, as long as battery voltage is available, the swap to the voltage regulating transformer will not occur.

B: INCORRECT: Credible because the 480v inputs to the 120v DP panel systems are split between the two load centers and the applicant may believe the loss of the 480v input will cause a loss of the normal supply, causing it to swap to the voltage regulating transformer. However, as long as battery voltage is available, the swap to the voltage regulating transformer will not occur.

C: CORRECT: The Class 1E 125 VDC Bus E1A11 would not lose power with the given conditions. The symptoms indicate that a loss of the in-service Battery Charger has occurred. This will not result in a loss of power to Class 1E 120 volt vital DP 1201 and 001. Class 1E Battery E1A11 will automatically supply power to DP 1201 and 001 through their respective inverters.

D: INCORRECT: Credible because each battery has two chargers, each powered from a different load center, but only one is normally in service.

Question Level: H Question Difficulty 3 Justification:

Must be able to determine whether a loss of Vital DC power has occurred from the symptoms given.

Then, based on what was lost, determine how the 125 VDC System will respond.

Page 28 of 150

X 480VAC 129VDC Battery Bus D 120VAC I S

T R

I 480VAC B Rectifier Inverter U 3Ø T 132VDC 120VAC I output single Ø Static Manual O output Transfer Bypass N Switch Switch P

Normal Normal A

N E

L

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1863 Last used on an NRC exam: 2009 RO Sequence Number: 15 Given the following conditions:

Unit 1 is raising Reactor Power and is currently at 40% power.

Subsequently:

Instrument Air pressure began to lower and is currently at 95 psig and trending down slowly.

The Control Room Crew is working through 0POP04-IA-0001, Loss of Instrument Air.

Based on current given conditions, which of the following describes the next appropriate crew response in accordance with 0POP04-IA-0001, Loss of Instrument Air?

A. Isolate CVCS Charging and Letdown flow.

B. Trip the Main Turbine and Isolate Main Steam.

C. Trip the Reactor and Ensure the Main Turbine is tripped.

D. Verify that the Instrument Air to Service Air Isolation Valve has closed.

Answer: D Verify that the Instrument Air to Service Air Isolation Valve has closed Page 29 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1863 K/A Catalog Number: APE 065 G2.4.11 Tier: 1 Group/Category: 1 RO Importance: 4.0 10CFR

Reference:

55.41(b)(10)

Loss of Instrument Air: Knowledge of abnormal condition procedures.

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 step 6 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: This distractor is credible because the procedure states in a note at the beginning that CVCS letdown isolation valve, FV-0011, will start to drift closed at 80 psig. If the crew lost letdown flow then they would also isolate charging flow.

B: INCORRECT: This distractor is credible because in the basis of the procedure it states that MSIVs will begin to drift closed at 56 psig. With the Reactor at 40% power and less than P-9, if the crew closed the MSIVs then they could trip the main turbine and not the Reactor.

C: INCORRECT: This distractor is credible because on the CIP of the procedure it states that if IA pressure lowers to 60 psig then Trip the Reactor. If the crew tripped the reactor they would have to ensure the Main Turbine tripped as well.

D: CORRECT: This question is answered correctly by having knowledge of the different IA pressures that effect IA valves and require specific actions. In this case with IA pressure at 95 psig and lowering the next appropriate action is to verify IA to SA Isolation valve is closed. The IA to SA isloation valve set point for closing is given in a note at the begining of the procedure and the valve is verified closed in step 6 in the body of the procedure when IA pressure is below 100 psig.

Question Level: F Question Difficulty 3 Justification:

Must know the Off Normal procedure requirements`

Page 30 of 150

0POP04-IA-0001 Loss Of Instrument Air Rev. 16 Page 4 of 152 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 5.0 CHECK IA Pressure - LESS THAN PERFORM the following:

100 PSIG a. IF IA pressure lowers to less than 100 psig, THEN GO TO Step 6.0.

b. GO TO Step 14.0.

_____ 6.0 CHECK IA Pressure - LESS THAN PERFORM the following:

90 PSIG a. VERIFY Service Air Flow Secured on ICS Display IA-001 SA Flow {CFM}

1) IF Service Air Flow is NOT Isolated, THEN DISPATCH an Operator to perform the following:

a) VERIFY Service Air Isolation Valve N1(2)IA-PV-9785 Closed.

{29 ft TGB}

b) IF N1(2)IA-PV-9785 is NOT closed, THEN CLOSE 1(2)-SA-9982 SERVICE AIR SYSTEM PV-9785 INLET ISOLATION VALVE.{29 ft TGB Between IA Dryer 11(21) and Dry Receiver 12(22)}

b. ENSURE all available IA Compressors running OR aligned to Start/Load.

c. IF IA pressure lowers to less than 90 psig, THEN GO TO Step 7.0.

d. GO TO Step 14.0.

This Procedure is Applicable in All Modes

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1864 Last used on an NRC exam: 2009 RO Sequence Number: 16 Given the following:

Unit 1 operators are establishing RCS bleed and feed in accordance with 0POP05-EO-FRH1, Loss of Secondary Heat Sink.

While verifying RCS bleed path per step 13, the Reactor Operator observes that ONE of the Pressurizer PORVs will not open.

Which of the following describes the appropriate response in accordance with 0POP05-EO-FRH1, Loss of Secondary Heat Sink, step 13, and the reason for this response?

A. Open the Reactor Vessel Head Vent valves because the RCS may not depressurize sufficiently to permit adequate SI flow to remove core decay heat.

B. Close the open PORV and continue efforts to restore AFW flow because one PORV will not depressurize the RCS sufficiently to allow SI to maintain RCS inventory.

C. No action is required because the RCS will still depressurize sufficiently with one PORV open to permit adequate SI flow to remove core decay heat.

D. Close the open PORV, then open the Reactor Vessel Head Vent valves to restrict the mass loss sufficiently to ensure RCS inventory can be maintained with SI.

Answer: A Open the Reactor Vessel Head Vent valves because the RCS may not depressurize sufficiently to permit adequate SI flow to remove core decay heat.

Page 31 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1864 K/A Catalog Number: EPE E05 EK2.1 Tier: 1 Group/Category: 1 RO Importance: 3.7 10CFR

Reference:

55.41(b)(5)

Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

STP Lesson: LOT 504.33 Objective Number: 83085 DESCRIBE the indications and anticipated readings used to determine that the Reactor Coolant System bleed path is adequate

Reference:

0POP05-EO-FRH1 step 13 WOG ERG Background FRH.1 (Rev 2)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: Reason and action IAW with the references cited.

B: INCORRECT: This distractor is credible because continuing efforts to restore AFW flow and returning to the beginning of the procedure is an action if adequate RCS FEED path cannot be established.

C: INCORRECT: This distractor is credible because it identifies a lack of knowledge with the function of the PZR PORVs and their interrelation with the Loss of Secondary Heat Sink. One PORV will NOT depressurize the RCS sufficiently to allow enough SI flow to maintain inventory and remove decay heat.

D: INCORRECT: This distractor is credible because it identifies a lack of knowledge with the function of the PZR PORVs and the Reactor Head Vent valves and their interrelation with the Loss of Secondary Heat Sink. Opening the Reactor Vessel Head Vent Valves provides additional depressurization of the RCS but is still not enough to equal 2 PZR PORVs being open even though this same action is performed if no PZR PORVs can be opened. If only one PZR PORV can be opened, then it needs to stay open.

Question Level: H Question Difficulty 3 Justification:

Must be able to determine appropriate procedure response for given plant conditions and understand why those actions are necessary.

Page 32 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2150 Last used on an NRC exam: Never RO Sequence Number: 17 Given the following:

Unit 1 is operating at 100% power All control systems are operating in automatic One Pressurizer Power Operated Relief Valve (PORV) fails full open No Operator actions are taken Considering the leak rate through the open Pressurizer PORV, which of the following describes the plant response to this event?

A. An Over-Temperature Delta-T reactor trip will occur along with a low Pressurizer pressure Safety Injection. RCS pressure will stabilize above the shutoff head for the LHSI Pumps due to injection by the HHSI Pumps.

B. An Over-Temperature Delta-T reactor trip will occur along with a low Pressurizer pressure Safety Injection. RCS pressure will stabilize below the shutoff head of the LHSI Pumps.

C. An Over-Power Delta-T reactor trip will occur along with a low Pressurizer pressure Safety Injection. RCS pressure will stabilize above the shutoff head for the LHSI Pumps due to injection by the HHSI Pumps.

D. An Over-Power Delta-T reactor trip will occur along with a low Pressurizer pressure Safety Injection. RCS pressure will stabilize below the shutoff head of the LHSI Pumps.

Answer: A An Overtemperature Delta-T reactor trip will occur along with a low Pressurizer pressure Safety Injection. RCS pressure will stabilize above the shutoff head for the LHSI Pumps due to injection by the HHSI Pumps.

Page 33 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2150 K/A Catalog Number: APE 008 AA2.25 Tier: 1 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(5)

Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident:

Expected leak rate from open PORV or code safety 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:

LOT501.21, handout page 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT - An OTDT trip will occur first at this power level in response to a lowering pressure.

RCS pressure will continue to drop until it stabilizes based on HHSI pump flow (SBLOCA response)

B: INCORRECT - This distractor is credible because it shows a lack of knowledge of the amount of flow that can come from an open PZR PORV. RCS pressure will lower, but not to the extent of having LHSI pumps injecting based on the leak size.

C: INCORRECT - This distractor is credible because with a lowering pressure at full power, OTDT trip will occur first. Although OPDT has a variable setpoint like OTDT, OPDT is not changed by a lowering PZR Pressure. RCS pressure response is correct.

D: INCORRECT - This distractor is credible because with a lowering pressure at full power, OTDT trip will occur first. Although OPDT has a variable setpoint like OTDT, OPDT is not changed by a lowering PZR Pressure. It also shows a lack of knowledge of the amount of flow that can come from an open PZR PORV. RCS pressure will lower, but not to the extent of having LHSI pumps injecting based on the leak size.

Question Level: H Question Difficulty 3 Justification:

Applicant must have an understanding of the leak rate through a PORV, then apply that leak to the RCS to determine plant response.

Page 34 of 150

LOT501.21.HO.01 Rev. 5 Page 6 of 47 Inadvertent Opening of a Pzr Safety or Relief Valve Event Definition The inadvertent opening of a Pzr safety or relief valve is defined as an accidental depressurization of the RCS caused by the spurious actuation of a Pzr safety or relief valve.

Major Concerns The major concerns associated with the unmitigated inadvertent opening of a Pzr safety or relief valve is possible fuel cladding damage resulting from the decrease in RCS pressure and subsequent violation of the safety analysis limit DNBR value.

Event Hazards/Challenges The hazards and challenges associated with the inadvertent opening of a Pzr safety or relief valve are:

A challenge to the fuel and fuel cladding due to the rapid reduction in the RCS pressure along with the high power and could violate the safety analysis limit DNBR.

A radiological hazard would be created if the reactor coolant water is discharged to the containment through the PRT.

Analysis Objective The objective of the analysis is to prove that the reactor protection system will automatically terminate the event prior to DNB occurring.

NOTE: Following reactor trip, the operator is expected to be able to isolate the open valve (assuming the failure is a PORV) and prevent further adverse reactor conditions. If the valve can not be isolated (such as a safety valve), the event is no longer a RCS depressurization but a small-break LOCA. The long term plant response due to the opening of a valve that can not be isolated is bounded by the limiting small-break LOCA.

Types of Accidents Analyzed The accident analyzed is the inadvertent opening of a Pzr safety valve while at full power operation.

A Pzr safety valve will relieve approximately twice the steam flow rate of a relief valve, and will therefore allow a much more rapid depressurization upon opening.

LOT501.21.HO.01 Rev. 5 Page 8 of 47 Inadvertent Opening of a Pzr Safety or Relief Valve (continued)

Sequence of Events DBD, Module 17, Table 17-4, Figures 17-1 to 17-3 The sequence of events is shown below.

Time Pzr Pressure Core Average Nuclear Power DNBR (sec) Temperature 0-21 Pzr pressure starts Core average As the pressure DNBR is decreasing decreasing temperature slowly decreases, more due to the decreasing immediately after increases with nucleate boiling RCS pressure.

the Pzr safety valve increasing reactor occurs in the core opens. power until reactor causing Tave to Pressure has trip occurs. increase and fuel decrease enough temperature to to cause a decrease.

reactor trip on The positive OTT at 20.8 MTC and seconds. negative Doppler cause power to increase.

21-23 Pzr pressure Core average Nuclear power DNBR reaches its continues decreasing temperature decreases decreases following minimum value at following the rods following the rods the rods dropping 22.9 seconds, above dropping into the dropping into the into the core. the safety analysis core. core. limit value and immediately increases thereafter due to decreasing nuclear power and RCS temperatures.

Results As was stressed earlier, this analysis stops following the reactor trip. It assumes that the valve will be isolated following the reactor trip and the RCS depressurization will be stopped. The objective of the analysis was to show that DNB would not occur prior to the reactor trip.

Radiological Consequences The radiological consequences for the inadvertent opening of a Pzr safety or relief valve are minimal.

Even assuming a direct release to the containment atmosphere, the radiological consequences of such an event are substantially less than that of a LOCA because:

less primary coolant is released, and the activity is lower since no fuel damage is postulated during the event.

Conclusions The results of the analysis show that the OTT reactor trip signal provides adequate protection against the RCS depressurization event.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2151 Last used on an NRC exam: Never RO Sequence Number: 18 Given the following:

Unit 2 is operating at 100% power.

A small Reactor Coolant System leak identified from a Containment Area Rad Monitor develops into a leak exceeding the capacity of a centrifugal charging pump within a two hour period.

Which of the following sets of procedures will be used to directly address this condition?

A. System Operating (POP02) and Emergency (POP05) procedures B. General Operating (POP03) and Enhanced Off-Normal (POP04) procedures C. General Operating (POP03) and Emergency (POP05) procedures D. Enhanced Off-Normal (POP04) and Emergency (POP05) procedures Answer: D Enhanced Off-Normal (POP04) and Emergency (POP05) procedures Page 35 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2151 K/A Catalog Number: EPE 009 G2.2.38 Tier: 1 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(10)

Small Break LOCA: Knowledge of conditions and limitations in the facility license.

STP Lesson: LOT 505.02 Objective Number: 92115 GIVEN a list of Emergency Operating Procedures, Off Normal Operating Procedures, Annunciator Response Procedures, and Operating Procedures, ARRANGE them in order of hierarchy.

Reference:

LOT505.02, lesson plan page 6 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - This distractor is credible because the Emergency Procedures (POP05) do refer to System Operating procedures but would not be directly used to address the RCS leak.

B: INCORRECT - This is a credible distractor because there are two centrifugal charging pumps and it would be reasonable to think that starting the second CCP would provide enough volume to maintain PZR level allowing a normal shutdown of the unit with a General Operating procedure (POP03). But one must remember that a leak greater than the capacity of just one CCP also affects make up capabilty to the VCT and thereby requires entry into the emergency procedures (POP05).

Entering the Enhanced Off-Normal procedure (POP04) is correct.

C: INCORRECT - This is a credible distractor because there are two centrifugal charging pumps and it would be reasonable to think that starting the second CCP would provide enough volume to maintain PZR level allowing a normal shutdown of the unit with a General Operating procedure (POP03). But one must remember that a leak greater than the capacity of just one CCP also affects make up capabilty to the VCT. Entering the Emergency procedure (POP05) is correct.

D: CORRECT - Once indication of RCS leakage identified from an area rad monitor, an Enhanced Off-Normal procedure (POP04) will be entered. A leak larger than the capacity of a charging pump will eventually require a manual SI due to inability to maintain PZR level which in turn will require entry into the Emergency procedures (POP05).

Question Level: H Question Difficulty 3 Justification:

From the given information, the applicant must determine that a manual SI will be required and then with their knowledge of procedure hierarchy determine how the event will be addressed.

Page 36 of 150

LP.NO.:LOT505.02.LP Rev. 0 Page 6 of 10 INSTR. NOTE: Obj. 92115 4.0 PROCEDURE HEIRARCHY 4.1 The highest priority procedure is to be the primary procedure in use. Other procedures may be used in conjunction with the primary procedure provided there is no conflict. The hierarchy is:

4.1.1 Emergency Operating Procedures (EOP) (0POP05s) 4.1.2 Off Normal Operating Procedures (ONP0 (0POP04s) 4.1.3 Annunciator Response Procedures (ARP) (0POP09s) 4.1.4 Normal Operating Procedures (OP) (0POP02s, 0POP03s)

INSTR. NOTE: If the Control Room has been evacuated, then Control Room Evacuation procedure, 0POP04-ZO-0001, takes precedence over the EOPs and 0POP04-ZO-0008 and 0POP04-ZO-0009.

If a fire occurs in Fire Areas 02-78, then 0POP04-ZO-0009, Safe Shutdown Fire Response takes precedence over all EOPs.

INSTR. NOTE: Obj. 92116 5.0 PROCEDURE USE AND ADHERENCE 5.1 Review PURPOSE and SYMPTOMS OR ENTRY CONDITIONS to verify ONP is appropriate. (example: Annunciator alarm determined to be not valid) 5.2 May be entered for any of the following reasons:

A condition is present which is specified in the SYMPTOMS OR ENTRY CONDITIONS SECTION.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2152 Last used on an NRC exam: Never RO Sequence Number: 19 Given the following:

Unit 2 is operating at 100% power 0POP04-RC-0002, Reactor Coolant Pump Off Normal has been entered Plant Computer has been lost ICS annunciator functions are still working Under these conditions, the procedure directs the operators to verify the status of RCP oil reservoir annunciators and if necessary, enter the containment and inspect the RCP.

Which of the following is true concerning the performance of these actions?

Oil Reservoir Annunciator Verification RCP Inspection CAN be performed from within the Control CAN be performed under these plant A.

Room Horseshoe area conditions CANNOT be performed from within the CAN be performed under these plant B.

Control Room Horseshoe area conditions CAN be performed from within the Control CANNOT be performed under these C.

Room Horseshoe area plant conditions CANNOT be performed from within the CANNOT be performed under these D.

Control Room Horseshoe area plant conditions Answer: C CAN be performed from within the Control Room Horseshoe area; CANNOT be performed under these plant conditions Page 37 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2152 K/A Catalog Number: APE 015/017AA1.02 Tier: 1 Group/Category: 1 RO Importance: 2.8 10CFR

Reference:

55.41(b)(3)

Ability to operate and / or monitor the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

RCP oil reservoir level and alarm indicators STP Lesson: LOT 201.05 Objective Number: 4829 DESCRIBE the instrumentation available for the reactor coolant pumps in the control room and locally

Reference:

LOT201.05 handout page 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - Alarm verification is correct. This is a credible distractor because an RCB entry would be allowed at this power level and is required to inspect RCP oil level but the risk of a radiolagical over exposure is too great at a power level above 10% if entering the boisheild where the RCPs are located, so the Unit must be down powered first.

B: INCORRECT - This distractor is credible because there are numerous alarms that are available only on back panels or plant computer, however RCP reservoir annunciators are located within the horseshoe area. This is a credible distractor because an RCB entry would be allowed at this power level and is required to inspect RCP oil level but the risk of a radiolagical over exposure is too great at a power level above 10% if entering the boishield where the RCPs are located, so the Unit must be down powered first.

C: CORRECT - Annunciators are located on Panel 5 within the control room horseshoe area. RCPs are located within the bioshield in containment which is inaccessible at this power level (procedure has the operators reduce power to less than 10% for RCP inspection).

D: INCORRECT - This distractor is credible because there are numerous alarms that are available only on back panels or plant computer, however RCP reservoir annunciators are located within the horseshoe area. RCB entry portion is correct.

Question Level: F Question Difficulty 3 Justification:

Applicant must have knowledge of the annunciator location for RCP oil reservoir levels in the control room and a basic knowledge of containment conditions (radiologically) under these plant conditions.

Page 38 of 150

LOT201.05.HO. Rev. 15 PAGE 8 OF 31 RCP RELATED REACTOR TRIPS UnderVoltage sensed on the Class 1E 15 Kv RCP Cubicle (Dummy breaker). Does NOT trip the RCP, but this signal feeds into SSPS and on a 2 of 4 logic, generates a Reactor Trip.

As previously discussed, Underfrequency - sensed on the Class 1E 15 Kv RCP Cubicle (Dummy breaker). This signal feeds into SSPS and on a 2 of 4 logic, trips ALL RCPS and also generates a Reactor Trip.

INSTRUMENTATION/ALARMS The following instrumentation is provided as part of each Reactor Coolant Pump motor:

A. RCP oil pressure switches: (2 per pump)

RCP oil lift pump No. 11 - PS-699A-1/2 RCP oil lift pump No. 12 - PS-699B-1/2 RCP oil lift pump No. 13 - PS-699C-1/2 RCP oil lift pump No. 14 - PS-699D-1/2 The above RCP oil pressure switches are part of an interlock system that prevents starting of a Reactor Coolant Pump until the Lube Oil Lift System is operating properly. The RCP upper thrust bearing must be supplied with lubricating oil at minimum pressure of 600 psig for starting.

When oil pressure gets above 600 psig a blue permissive light (CP-004) comes on.

Administrative controls require that the RCP not be started for at least two minutes after the oil lift pump pressure light has come on. This time delay helps protect against starting the pump with inadequate oil lift pressure, as might be caused by an obstruction in the oil supply line temporarily blocking flow while maintaining pressure above setpoint.

B. Upper Oil Reservoir Liquid Level A level switch (LS-687A, LS-687B, LS-687C, and LS-687D) is provided in the oil reservoir for the motor upper radial bearing and thrust bearing, with contacts to actuate high and low alarms on CP-005.

The annunciators on CP-005 are as follows:

RCP 1A UPPER OIL RSVR LEVEL HI/LO (1-1) on 1LB005A RCP 1B UPPER OIL RSVR LEVEL HI/LO (1-2) on 1LB005A RCP 1C UPPER OIL RSVR LEVEL HI/LO (1-3) on 1LB005A RCP 1D UPPER OIL RSVR LEVEL HI/LO (1-4) on 1LB005A Low oil level signals may indicate leaks in the oil piping or oil reservoir. High oil level signals may indicate cooling water leakage permitting oil to mix with water, or possibly a faulty oil-indicating device.

C. RCP Lower Oil Reservoir Liquid Level

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2153 Last used on an NRC exam: Never RO Sequence Number: 20 Pressurizer backup heaters have been energized to recover from a Pressurizer Pressure Control malfunction that resulted in Pressurizer pressure lowering 30 psig.

During the recovery, (1) heat is being added to raise the fluid temperature in the Pressurizer and (2) heat is being added to change saturated liquid into a saturated vapor. It takes (3) energy to change 1 (one) pound of saturated fluid to saturated vapor than it does to raise 1 (one) pound of saturated liquid 1ºF.

(1) (2) (3)

A. sensible latent less B. latent sensible less C. latent sensible more D. sensible latent more Answer: D sensible, latent, more Page 39 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2153 K/A Catalog Number: APE 027 AK1.03 Tier: 1 Group/Category: 1 RO Importance: 2.6 10CFR

Reference:

55.41(b)(14)

Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: Latent heat of vaporization/condensation STP Lesson: LOT 102.54 Objective Number: N99793 Define the following terms: Latent Heat of Vaporization

Reference:

LOT102.54 handout page 13 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - This distractor is credible because it shows a lack of fundemental knowledge of the concepts of latent heat of vaprization as it applies to the PZR. This distractor does not have all parts of the statement correctly identified.

B: INCORRECT - This distractor is credible because it shows a lack of fundemental knowledge of the concepts of latent heat of vaprization as it applies to the PZR. This distractor does not have all parts of the statement correctly identified.

C: INCORRECT - This distractor is credible because it shows a lack of fundemental knowledge of the concepts of latent heat of vaprization as it applies to the PZR. This distractor does not have all parts of the statement correctly identified.

D: CORRECT - LATENT heat is added for a phase change and SENSIBLE heat is added for a temperature change. It takes MORE heat energy for a phase change than it does to change temperature.

Question Level: F Question Difficulty 3 Justification:

Applicant must have an understanding of the fundamental thermodynamic processes which occur inside the Pressurizer.

Page 40 of 150

102.54 GPST3.doc The latent heat of fusion is the amount of heat higher than the critical point, the substance is that must be added to change the phase of a solid considered a fluid; something neither gas or to a liquid at constant temperature and pressure. liquid. At pressures lower than the critical The latent heat of fusion can also be thought of pressure (but at higher temps), the substance is as the change in specific enthalpy of the considered a gas substance when changing phase from a solid to a liquid, at a given temperature and pressure. If a No definable phase change occurs above the liquid is frozen, the latent heat of fusion critical point. Two rather arbitrarily drawn lines represents the amount of heat that must be added are extended to the right and upward from the to change it to a liquid. The latent heat of fusion critical point to constitute an area where a gas or for ice is 144 Btu/lbm at a pressure of one liquid state is not readily apparent. Any atmosphere and a temperature of 32°F. substance whose property values cause it to fall within this area is referred to as a fluid. A fluid As heat is added to a liquid at some constant is neither gas nor liquid. A phase transition does pressure below critical pressure (Pc), we move to not occur at the points that define these lines, but the right on the P-T diagram as the temperature they do correspond to an arbitrary definition of of the liquid is elevated, until the Vaporization what is a liquid and what is a gas. The fluid Line is reached. At this point, any addition of region simply resolves the indeterminate area in heat results in a phase change as the liquid between these states.

evaporates into a gas. The latent heat of vaporization is the amount of heat that must be The critical point of water occurs at a pressure of added to cause this phase transition. The change 3,208.2 psia and a temperature of 705.47°F. At in enthalpy per lbm of the substance when the critical point, the latent heat of vaporization changing phase from a liquid to a gas, at a is zero, since steam and water are perceived as constant temperature and pressure, is equal to the one and the same. Most atmospheric gases have latent heat of vaporization. The latent heat of critical temperatures much lower than water. For vaporization for water is 970 Btu/lbm at a example, the critical temperature of helium is pressure of one atmosphere and a temperature of 9.54°R or -450.46°F. Therefore, helium at room 212°F. Less heat is required at higher temperature is at a temperature approximately 55 temperatures and pressures to produce the same times greater than its critical temperature.

phase change. For example, water at a pressure Conversely, the critical temperature of metals is of 1,000 psia requires the addition of only typically much higher than the critical 650.4 Btu to vaporize one lbm of water. The temperature of water.

saturated steam tables contain the values for The single point at which the three phase lines latent heat of vaporization for water.

come together is called the "triple point" for the The points at which a substance can exist as both substance. This single point is unique because all a liquid and gas in equilibrium is represented by three phases (solid, liquid, and gas) can exist in the Vaporization Line. Although the Fusion equilibrium with each other while at this Line has no upper limit, the Vaporization Line pressure and temperature. For example, the terminates at a point defined as the "critical triple point of water is at a temperature of point" for the substance. 32.02°F and a pressure of 0.089 psia. At this state point, ice, water, and water vapor would The highest temperature (critical temperature) exist together.

and pressure (critical pressure) at which a gas and liquid can exist in equilibrium as distinguishable phases is represented by the critical point. At temperatures and pressures PWR / THERMODYNAMICS / CHAPTER 3 13 of 47 © 2011 GENERAL PHYSICS CORPORATION

/ STEAM REV 3 GF@gpworldwide.com www.gpworldwide.com

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2156 Last used on an NRC exam: Never RO Sequence Number: 21 Given the following:

Unit 1 was operating at 100% power.

A Main Steam leak occurred inside containment.

Containment reached a peak pressure of 8 psig.

Containment pressure and temperature are now slowly lowering.

Which of the following describes the reason for the current containment pressure and temperature response?

Heat is being removed from the containment atmosphere by A. both the Containment Spray System and the RCB Chilled Water System flowing through the Reactor Containment Fan Cooler (RCFC) cooling coils.

B. only the RCB Chilled Water System flowing through the Reactor Containment Fan Cooler (RCFC) cooling coils.

C. both the Containment Spray System and the Component Cooling Water System flowing through the Reactor Containment Fan Cooler (RCFC) cooling coils.

D. only the Component Cooling Water System flowing through the Reactor Containment Fan Cooler (RCFC) cooling coils.

Answer: D only the Component Cooling Water System flowing through the Reactor Containment Fan Cooler (RCFC) cooling coils.

Page 41 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2156 K/A Catalog Number: APE 040 AK3.06 Tier: 1 Group/Category: 1 RO Importance: 3.4 10CFR

Reference:

55.41(b)(7)

Knowledge of the reasons for the following responses as they apply to the Steam Line Rupture:

Containment temperature and pressure considerations STP Lesson: LOT 201.12 Objective Number: 57126 DESCRIBE the operation of the Component Cooling Water System and its major components. Include automatic actions, interlocks and trips

Reference:

LOT201.12 handout page 23 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - This distractor is credible because a knowledge of the Containment Spray actuation setpoint of 9.5 psig is required. This distractor is credible because RCB chilled water is normally aligned to the RCFCs but swaps to the CCW system on an SI. (Containment Pressure 3.0 psig)

B: INCORRECT - This distractor is credible because RCB chilled water is normally aligned to the RCFCs but swaps to the CCW system on an SI. (Containment Pressure 3.0 psig)

C: INCORRECT - This distractor is credible because a knowledge of the Containment Spray actuation setpoint of 9.5 psig is required. CCW through the RCFC coils is correct.

D: CORRECT - A Containment Spray actuation has not occurred, but SI has actuated which swaps RCFC cooling to Component Cooling Water.

Question Level: H Question Difficulty 3 Justification:

From the given information, the applicant must determine that a Containment Spray actuation has not occurred (happens at 9.5 psig) and that an SI actuation has occurred (3 psig). The applicant must also know that cooling for the RCFCs swaps from the normal chilled water supply to Component Cooling water upon an SI actuation.

Page 42 of 150

LOT201.12.HO.01 Rev. 13 PAGE 22 OF 40 4.7.19 Post-Accident Sampling System CCW provides cooling water to post-accident sample coolers.

4.7.20 ESF Status Monitoring Monitors safety-related components of the CCWS and provides alarms and status indication for inoperable equipment, failure to actuate and equipment bypassed conditions.

5.0 SAFETY INJECTION Upon receipt of the safety injection signal, the following operations are automatically initiated:

5.1 Starting of all three CCW pumps.

5.2 Opening of the CCW heat exchanger outlet valves MOV-0643, MOV-0645, and MOV-0647 and closing of CCW heat exchanger bypass valves MOV-0642, MOV-0644, and MOV-0646. This prevents the CCW heat exchanger from being bypassed thereby ensuring full heat removal capability.

5.3 The isolation valves MOV-0447, MOV-0032, MOV-0235, MOV-0236, MOV-0297, MOV-0392, MOV-0393, FV-4540 and FV-4541 are closed to isolate the following non-ESF components:

SFP heat exchangers BTRS chiller LWPS evaporator package Letdown heat exchanger Excess letdown heat exchanger BRS evaporator package RCDT heat exchanger Primary sample cooler Post-accident sample cooler Boric acid sample cooler 5.4 The RHR heat exchanger isolation valves FV-4531, FV-4548 and FV-4565 open allowing CCW flow to the RHR heat exchangers.

LOT201.12.HO.01 Rev. 13 PAGE 23 OF 40 5.5 Switchover of the RCFCs cooling water source from chilled water to CCW, by automatic closing of chilled water supply and return valves and automatic opening of CCW supply and return valves.

5.6 All the safety injection automatic functions can be performed remotely (manually) from the control room should the automatic system fail.

5.7 Should a Phase B containment isolation signal exist, the CCW to and from the containment for the RCPs will also be by automatic closing of MOV-0318, MOV-0291, MOV-0404, MOV-0542, MOV-0403 and AOV-4493.

5.8 Recirculation Phase Recirculation mode is the condition after the injection phase. During this mode the CCWS provides cooling to the same components as the injection phase. Two CCW trains are capable of safely cooling down the reactor following a DBA. During the recirculation phase the operator has 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to restore the CCW supply to the SFP heat exchangers.

6.0 LOSS OF OFFSITE POWER (LOOP)

The motor-operated valves remain in the same position before the power failure, until the standby diesel generators automatically start to provide power to all CCW safety-related components and valves. As soon as the emergency power is available, all three CCW trains are sequenced on and the chilled water to and from the RCFCs is isolated by closing MOV-0059, MOV-0070, MOV-0137, MOV-0149, MOV-0199, MOV-0209, FV-0864, FV-0852 and FV-0863. CCW to the RCFCs is to be manually provided by the operator within 30 minutes. Additionally MOV-0392 for isolating the RCDT heat exchanger automatically closes. If instrument air is lost during the LOOP, the following pneumatic valves go to their failed position:

RHR Heat Exchanger Outlet Fails Open.

Auto Makeup (Demin) Fails Closed.

Charging pump header cross connects Fail Closed.

CCW from RCPs, OCIV FV-4493 Fails Closed.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2157 Last used on an NRC exam: Never RO Sequence Number: 22 When starting a Reactor Coolant Pump (RCP) in 0POP05-EO-EC11, Loss of Emergency Coolant Recirculation, a Note in the procedure gives the preferred running order of RCPs as follows:

First - Loop D Second - Loop A Third - Loops B AND C Which of the following is true regarding the Note?

The Note helps to ..

A. minimize the effect of RCP operation on RHR Pump performance.

B. minimize the effect of RCP operation on LHSI Pump performance.

C. ensure normal Pressurizer spray flow is available when needed.

D. ensure adequate mixing in all portions of the Reactor Coolant System.

Answer: C ensure normal Pressurizer spray flow is available when needed.

Page 43 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2157 K/A Catalog Number: EPE E11 G2.4.20 Tier: 1 Group/Category: 1 RO Importance: 3.8 10CFR

Reference:

55.41(b)(10)

Loss of Emergency Coolant Recirc: Knowledge of the operational implications of EOP warnings, cautions, and notes.

STP Lesson: LOT 504.27 Objective Number: 82520 Given a step, note or caution from 0POP05-EO-EC11, STATE its basis.

Reference:

EOPT03.20, page 10 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - Credible because Loop D is the only loop that does not have an RHR pump connected to it and it is listed first, and a running RCP would apply additional back pressure on the discharge of the RHR pump (but not enough to degrade the operation of the RHR pump).

B: INCORRECT - Credible because Loop D is the only loop that does not have an LHSI pump connected to it and it is listed first. However, LHSI pumps injecting and RCPs running would not occur at the same time.

C: CORRECT - The basis for the note is to indicate which combination of RCP(s) will provide normal spray flow.

D: INCORRECT - Credible because each pump produces a different flow distribution within the RCS, however that is not a consideration for the starting order.

Question Level: F Question Difficulty 3 Justification:

The applicant must have knowledge of procedure basis or an understanding of spray flow dynamics within the RCS.

Page 44 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2165 Last used on an NRC exam: Never RO Sequence Number: 23 Given the following:

Unit 2 is at 60% power and raising power to 90% at 10%/hour Generator Voltage Regulator is ON (auto)

Inclement weather has caused a grid disturbance GEN MAX EXCT alarm on annunciator panel 7M01 illuminates Which of the following describes the grid disturbance that has occurred and the required operator action in response to the alarm?

A. Grid voltage has risen causing the generator voltage regulator to raise excitation. Lower excitation using the VOLTAGE ADJUSTER control.

B. Grid voltage has risen causing the generator voltage regulator to raise excitation. Lower excitation using the BASE ADJUSTER control.

C. Grid voltage has lowered causing the generator voltage regulator to raise excitation.

Lower excitation using the VOLTAGE ADJUSTER control.

D. Grid voltage has lowered causing the generator voltage regulator to raise excitation.

Lower excitation using the BASE ADJUSTER control.

Answer: C Grid voltage has lowered causing the generator voltage regulator to raise excitation. Lower excitation using the "VOLTAGE ADJUSTER" control.

Page 45 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2165 K/A Catalog Number: APE 077 AK1.02 Tier: 1 Group/Category: 1 RO Importance: 3.3 10CFR

Reference:

55.41(b)(4)

Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Over-excitation STP Lesson: LOT 202.17 Objective Number: 91963 DESCRIBE manual and auto voltage regulation.

Reference:

0POP09-AN-07M1, window C-4 (page 19)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - This distractor is credible because rising grid voltage could be thought to also raise excitation on the Main Generator but the opposite is true. Correct control is indicated.

B: INCORRECT - This distractor is credible because rising grid voltage could be thought to also raise excitation on the Main Generator but the opposite is true. This distractor is credible because the Base Adjuster can raise or lower excitation on the Main Generator but not when the Voltage regualtor is ON.and in Auto.

C: CORRECT - Lowering grid voltage will cause the voltage regulator to raise excitation and result in the given alarm. The operator must use the voltage adjuster to lower field current and thereby lower Main Generator excitation.

D: INCORRECT - Lowering grid voltage will cause the voltage regulator to raise excitation and result in the given alarm. This distractor is credible because the Base Adjuster can raise or lower excitation on the Main Generator but not when the Voltage regualtor is ON.and in Auto.

Question Level: H Question Difficulty 3 Justification:

From the given conditions, the applicant must determine that Main Generator excitation has risen.

Knowledge of generator control is needed to determine that lowering grid voltage will result in the auto voltage regulator raising excitation. The applicant then must know which controller will manually change excitation under these conditions.

Page 46 of 150

0POP09-AN-07M1 Rev. 16 Page 19 of 35 Annunciator Lampbox 7M01 Response Instructions GEN MAX EXCT Automatic Actions: 1) IF the "VOLT REG CONT" switch is in AUTO, THEN exciter current is limited to 112 amps.

Immediate Actions: None Subsequent Actions: 1) IF the "VOLT REG CONT" switch is in AUTO, THEN PERFORM the following:

a) DECREASE "EXC FLD CUR" until alarm clears, using the "VOLTAGE ADJUSTER" control.

b) ENSURE Main Generator MVARS less than 400 MVAR positive.

CAUTION Extended manual operation of the Voltage Regulator is not recommended because the exciter limiter circuits are not functional (protection circuits remain functional).

NOTE

• IF "VOLT REG NULL" is NOT at zero, THEN sudden MVAR reading change should be expected when turning "VOLT REG CONT" switch from AUTO to OFF

• IF controlling voltage manually, THEN increased operator awareness is required to maintain MVAR loading on Main Generator.

2) IF Voltage Regulator output can NOT be reduced in Auto, THEN PERFORM the following:

a) PLACE "VOLT REG CONT" switch in OFF.

b) DECREASE "EXC FLD CUR" until alarm clears, using the "BASE ADJUSTER" control.

c) ENSURE Main Generator MVARS less than 400 MVAR positive.

Page 1 of 2 7M01-C-4 GEN MAX EXCT

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2168 Last used on an NRC exam: Never RO Sequence Number: 24 The table below lists Intermediate Range NI readings before and after a unit trip.

100% power 10 minutes after trip 30 minutes after trip 4x10 -4 amps 8x10-10 amps 6x10-10 amps NI-35 stable slowly lowering stable

-4 -10 4.5x10 amps 1.5x10 amps 2x10-11 amps NI-36 stable lowering stable Which of the following is true regarding the current status of the Intermediate Range NIs?

A. NI-35 has lost its compensating voltage (under compensated)

B. NI-35 is over compensated C. NI-36 has lost its compensating voltage (under compensated)

D. NI-36 is over compensated Answer: A NI-35 has lost its compensating voltage Page 47 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2168 K/A Catalog Number: APE 033 AA2.11 Tier: 1 Group/Category: 2 RO Importance: 3.1 10CFR

Reference:

55.41(b)(5)

Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Loss of compensating voltage STP Lesson: LOT 201.16 Objective Number: 91250 DEFINE overcompensation and undercompensation and DESCRIBE their effect on intermediate range detector operation.

Reference:

LOT201.16, Excore NIs, Powerpoint slide #8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT - An IR detector will normally lower to about 10-10 amps within 10 to 15 minutes following a trip and finally stabilize just above idling current of 10-11 amps. A detector that is reading higher than it should (due to gamma interactions that are not compensated out) is termed to be undercompensated or has lost its compensating voltage.

B: INCORRECT - This distractor is credible because it could be thought that NI-35 is over compensated because it is reading too high.

C: INCORRECT - This distractor is credible because if it were thought that NI-36 was reading too low and if there was a misconception of how compensation works in the Intermediate Range Detectors then it would be under compensated. However, the readings/trends for NI-36 is indicative of normal response following a trip. This detector is correctly compensated.

D: INCORRECT - This distractor is credible because if it were thought that NI-36 was reading too low, then the detector would be over compensated. However, the readings/trends for NI-36 is indicative of normal response following a trip. This detector is correctly compensated.

Question Level: H Question Difficulty 3 Justification:

The applicant must evaluate the given information and then based on knowledge of normal detector response and definition of terms, determine the correct answer.

Page 48 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2170 Last used on an NRC exam: Never RO Sequence Number: 25 0POP05-EO-FRC1 RESPONSE TO INADEQUATE CORE COOLING Step 13 instructs operators to stop ALL Reactor Coolant Pumps prior to depressurizing ALL Intact SGs to Atmospheric Pressure.

Select the RCP parameter that is of concern.

A. RCP motor stator winding temperature B. RCP lower seal water bearing temperature C. RCP motor lower radial bearing temperature D. RCP number 1 seal differential pressure Answer: D RCP number 1 seal differential pressure Page 49 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2170 K/A Catalog Number: EPE 074 EA1.06 Tier: 1 Group/Category: 2 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Ability to operate and monitor the following as they apply to a Inadequate Core Cooling: RCPs STP Lesson: LOT 504.30 Objective Number: 82939 DESCRIBE the indicators available to determine that the RCPs should be stopped.

Reference:

WOG FR-C1 Background Document. Pg 42 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: Incorrect - This distractor is credible because the student may associate the containment conditions in this accident with the possible effects on the RCP motor winding temperatures and RCP motor winding temperatures are a for the RCPs..

B: Incorrect - This distractor is credible because the student may associate the RCS conditions in this accident with the possible effects on the RCP lower seal water bearing temperature and RCP lower seal water bearing temperatures are a concern for the RCPs.

C: Incorrect - This distractor is credible because the student may associate the containment conditions in this accident with the possible effects on the RCP motor bearing temperatures and RCP motor bearing temperatures are a concern for the RCPs.

D: Correct - AS the SGs are depressurized the RCS pressure will lower rapidly. The number 1 seal surfaces will come in contact with each other. The RCPs are secured to prevent damage to the seals so that they will be avalible for use later if CETs exceed 1200F Question Level: F Question Difficulty 3 Justification:

The applicant must recognize the effect on the RCS and recall the basis for securing RCPs prior to depressing the SGs.

Page 50 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2171 Last used on an NRC exam: Never RO Sequence Number: 26 Given the following:

Unit 1 is in Mode 6 during a rapid refueling outage Core reload has just been completed 2 assemblies were placed in the wrong core location during the reload resulting in Keff being higher than predicted Which of the following is true concerning Shutdown Margin (SDM) as a result of this event?

A. SDM has not changed assuming boron concentration has remained the same.

B. SDM has not changed since all control rods are fully inserted.

C. SDM is larger. The reactor is farther from criticality.

D. SDM is smaller. The reactor is closer to criticality.

Answer: D SDM is smaller. The reactor is closer to criticality.

Page 51 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2171 K/A Catalog Number: APE 036 AK1.02 Tier: 1 Group/Category: 2 RO Importance: 3.4 10CFR

Reference:

55.41(b)(1)

Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Incidents : SDM STP Lesson: LOT 201.43 Objective Number: 92856 DISCUSS reactivity management concerns when the plant is in Mode 5 & 6 to include:

a. fuel assembly movement

Reference:

LOT101.19 handout pages 24 and 29 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: This distractor is credible because the fuel assemblies were put in the wrong place but overall the core still has the same assemblies, however, by placing the assemblies in the incorrect location and it resulted in a higher Keff, the resulting SDM would have to be reduced especially if the Boren concentration did not change even though it would be at about 2800 ppm.

B: INCORRECT: This distractor is credible because the fuel assemblies were put in the wrong place but overall the core still has the same assemblies, however, by placing the assemblies in the incorrect location and it resulted in a higher Keff, the resulting SDM would have to be reduced and in a rapid refueling the control rods would not be inserted but would be locked full out in the Reactor Head.

C: INCORRECT: This distractor is credible because there has been times when students have had misconceptions with how changes in Keff will affect SDM. Placing the assemblies in the incorrect location resulted in Keff higher reducing SDM.

D: CORRECT: If Keff is higher, then SDM would be smaller and the core closer to criticality.

Question Level: F Question Difficulty 3 Justification:

The applicant must recall what paramenters affect SDM and how.

Page 52 of 150

(PR02Sr4_LifeCycle May 2011) LOT10119GPST4.DOC The answer to this problem is a concept called REACTIVITY REACTIVITY.

Reactivity is the measure of the departure of a Thus far the discussion has centered on keff and reactor from criticality. Reactivity is defined as the affects on the factors in the six factor the fractional change in neutron population per formula. Reactor operators, however, do not generation and is indicated by the Greek letter use keff very often in referring to the condition rho (). The fractional change in neutron of the reactor. There are several reasons that population per generation (reactivity) can be have been given for this. keff is a multiplication shown by the equation given below.

factor, made up by multiplying other parameters together. Reactor operators do not have the k e ff 1 luxury of considering only one parameter k eff change at a time. Often, they must consider the affect on the core of many changes that have Equation 2-19 occurred.

The key is that reactivity terms are additive. It Using the multiplicative nature of keff, the is a common scale with which we can quantify operator might have to multiply the keff of the affect on the core due to moderator having moderator temperature going up, with temperature change as well as the change due to the keff associated with rods moving out. This is rod motion. It can quantify the change due to very awkward. What is needed is something boron and xenon concentration changes as well that can be added and subtracted.

as the effects on the core from power level Consider a change in rod position and a changes. It is the common scale that the commensurate change in moderator temperature Reactor Operator needs to control reactor when operating in the power range (why this power.

happens will be discussed in detail in Chapter Before we can start using it in this way, we need 4). The rods move in or out and due to the to understand the units used to quantify change in neutron absorption cause a change in reactivity.

the probability of neutrons surviving to cause fission. This creates an imbalance in the core, keff itself is dimensionless (that is has no units).

causing neutron population to change, resulting Thus, reactivity is also dimensionless. Operator in a change in moderator temperature. As rods use of it in quantifying reactor behavior, move into the core the neutron absorber in the however, leads to a need for some sort of rod capture more neutrons. A reduction in the dimension. The formula itself is used to define neutron population causes reactor power to the natural reactivity unit k/k. Also used are lower and less heat is generated. This results in the units %k/k and pcm as follows:

a lower RCS moderator temperature.

So, how much rod motion is needed to change k eff k eff 1 temperature by a degree Fahrenheit? A common scale is necessary. Temperature k eff 1 k eff k change and rod movement are both changing the k eff k eff k conditions in the core, but we have no common scale on which to measure the two effects. Equation 2-20 PWR / REACTOR THEORY / CHAPTER 2 24 of 39 © 2007 GENERAL PHYSICS CORPORATION

/ NEUTRON LIFE CYCLE REV 4 GF@gpworldwide.com www.gpworldwide.com

(PR02Sr4_LifeCycle May 2011) LOT10119GPST4.DOC NOTE: Most of the discussion of Shut Down SHUTDOWN MARGIN Margin (SDM) is beyond the scope of what has been covered thus far in this course. While a brief discussion of SDM is included here, the Shutdown margin (SDM) is the instantaneous main discussion is found in Chapter 8, Reactor amount of reactivity that the core is, or can be Operational Physics.

made, subcritical from its present condition with the most reactive control rod fully withdrawn from the core at any time during the core cycle.

Technical Specifications require a shutdown margin with the most reactive rod withdrawn from the core. A typical value required is a shutdown margin of 1.3% k/k. These values change depending on the operational mode.

The shutdown margin is calculated using the following equation.

1 k eff SDM k eff Equation 2-25 Note that this equation may look like the reactivity equation, but the equation is different; the terms in the numerator are reversed.

Calculate the shutdown margin of a shutdown reactor with a core reactivity value of -0.0045 k/k.

Example 2-18 PWR / REACTOR THEORY / CHAPTER 2 29 of 39 © 2007 GENERAL PHYSICS CORPORATION

/ NEUTRON LIFE CYCLE REV 4 GF@gpworldwide.com www.gpworldwide.com

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2173 Last used on an NRC exam: Never RO Sequence Number: 27 With the plant operating at 100% power, which of the following would violate the definition of containment integrity?

A loss of containment integrity would occur if ..

A. the Supplementary Containment Purge exhaust OCIV and ICIV are opened to reduce containment pressure.

B. BOTH doors of the Personnel Airlock (PAL) OR Auxiliary Airlock (AAL) are opened for material passage.

C. a normally closed air operated containment isolation valve for RCS sampling is opened for chemistry to grab a sample.

D. an automatic containment isolation valve is closed and de-energized for maintenance on the control circuit.

Answer: B BOTH doors of the Personnel Airlock (PAL) OR Auxiliary Airlock (AAL) are opened for material passage.

Page 53 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2173 K/A Catalog Number: APE 069 AK2.03 Tier: 1 Group/Category: 2 RO Importance: 2.8 10CFR

Reference:

55.41(b)(7)

Knowledge of the interrelations between the Loss of Containment Integrity and the following: Personnel access hatch and emergency access hatch STP Lesson: LOT 503.01 Objective Number: 92101 From memory, DEFINE terms used in the Technical Specifications and the Technical Requirements Manual (TRM).

Reference:

Tech Spec definition 1.7, Containment Integrity (page 1-2)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - This distractor is credible because normal containment purge valves cannot be opened at power but the supplemental purge system valves automatically close when required.

B: CORRECT - Each airlock must be operable (no more than 1 door open at a time)

C: INCORRECT - This distractor is credible because manual containment isolation valves cannot be opened without affecting containment integrity but air operated isolation valves will automatically close when required.

D: INCORRECT - This distractor is credible because working on containment isolation valves can affect containment integrity but if an MOV is in its required position and de-energized then containment integrity would still be satisfied.

Question Level: F Question Difficulty 3 Justification:

The applicant must have a working knowledge of the definition of containment integrity.

Page 54 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2183 Last used on an NRC exam: Never RO Sequence Number: 28 Given the following:

A Loss of Offsite Power has occurred.

ALL 4.16KV ESF busses are powered from their associated Emergency Diesel Generator.

The Unit supervisor directs the operator to ENSURE Pressurizer Heaters are ON Which of the following (1) list the available Pressurizer Heaters and (2) the correct operator action to energize them?

A. (1) Only A, B and C Pressurizer Heater Groups (2) Reset ESF Load Sequencers, then turn the Control Room Handswitch to ON B. (1) Only A and B Pressurizer Heater Groups (2) Reset ESF Load Sequencers, then turn the Control Room Handswitch to ON C. (1) Only A, B and C Pressurizer Heater Groups (2) Reset ESF Load Sequencers, then cycle Control Room Handswitch OFF then ON D. (1) Only A and B Pressurizer Heater Groups (2) Reset ESF Load Sequencers, then cycle Control Room Handswitch OFF then ON Answer: D (1) Only A and B Pressurizer Heater Groups (2) Reset ESF Load Sequencers, then Cycle Contro lroom Handswitch OFF then ON Page 55 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2183 K/A Catalog Number: APE 056 AA2.17 Tier: 1 Group/Category: 1 RO Importance: 3.4 10CFR

Reference:

55.41(b)(7)

Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Operational status of PZR backup heaters.

STP Lesson: LOT 201.14 Objective Number: 80414 STATE the pressurizer pressure and level control system actuation signals, setpoints, logic, coincidence, and interlocks.

Reference:

LOT201.14 PowerPoint slides 9 and 10 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Part 1 is credible since different groups of heaters have different power supplies, so the applicant must have specific knowledge of system design to determine the correct response.

Part 2 is credible since most components will change state by going to ON with their control switch.

It must be known that an additional interlock exists.

B: INCORRECT: Part 2 is credible since most components will change state by going to ON with their control switch. It must be known that an additional interlock exists.

C: INCORRECT: Part 1 is credible since different groups of heaters have different power supplies, so the applicant must have specific knowledge of system design to determine the correct response.

D: CORRECT: The A and B Pressurizer Heaters would have power available and resetting the ESF Load Sequencer and cycling the control room handswitch would enable Pressurizer Heaters A and B.

Question Level: F Question Difficulty 3 Justification:

The Reactor Operator must have knowledge of the logics associated with the Pressurizer Heaters.

Page 56 of 150

Pressurizer Pressure and Level Control System Objective 2: Heater power supplies and KW ratings PZR HTR B/U GRP A LC-E1A1 431 KW PZR HTR B/U GRP B LC-E1C1 431 KW CONTROL GROUP C LC-1N 485 KW PZR HTR B/U GRP D LC-1P 377 KW PZR HTR B/U GRP E LC-1J2 377 KW TOTAL 2101 KW 9

10 Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2210 Last used on an NRC exam: Never RO Sequence Number: 29 With a containment purge in progress in Mode 3, high radiation in containment caused a high alarm on RT-8012, RCB Purge Exhaust.

A note in 0POP04-RA-0001, Radiation Monitoring System Alarm Response, states the following:

A high alarm on RT-8012 or RT-8013 will cause a Containment Ventilation Isolation (CVI). This, in turn, causes RT-8011 sample lines to be isolated and renders RT-8011 radiation monitor inoperable.

Which of the following describes the implications of the note?

With the RT-8011 sample lines isolated, the sample pump A. may run indefinitely on recirculation flow. RT-8011 is inoperable; however Technical Specifications are not affected.

B. must be secured to prevent damage. RT-8011 is inoperable; however Technical Specifications are not affected.

C. may run indefinitely on recirculation flow. RT-8011 is inoperable and Technical Specification entry will be required.

D. must be secured to prevent damage. RT-8011 is inoperable and Technical Specification entry will be required.

Answer: D must be secured to prevent damage. RT-8011 is inoperable and Technical Specification entry will be required.

Page 57 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2210 K/A Catalog Number: EPE 016 G2.4.20 Tier: 1 Group/Category: 2 RO Importance: 3.8 10CFR

Reference:

55.41(b)(10)

High Containment Radiation: Knowledge of the operational implications of EOP warnings, cautions, and notes.

STP Lesson: LOT 505.01 Objective Number: 92110 Given a precaution, note, or step(s) and the context in which it is used from the referenced procedure, DESCRIBE its basis and any applicable limits.

Reference:

POP04-RA-0001, page 20 and 105 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Sample pump status is credible because most pumps have a recirculation flowpath, but this sample pump has no recirculation flowpath. Tech Spec is credible because a component either is or is not contained in Tech Specs, specific knowledge is needed to make this determination.

B: INCORRECT: Tech Spec is credible because a component either is or is not contained in Tech Specs, specific knowledge is needed to make this determination.

C: INCORRECT: Sample pump status is credible because most pumps have a recirculation flowpath, but this sample pump has no recirculation flowpath.

D: CORRECT: Must ensure the pump is secured to prevent damage. The radiation monitor is required by the RCS Leakage Detection Tech Spec.

Question Level: F Question Difficulty 3 Justification:

The applicant must have a knowledge of the operation of the radiation monitor and Tech Spec entry conditions.

Page 58 of 150

0POP04-RA-0001 Radiation Monitoring System Alarm Rev. 29 Page 20 of 132

Response

Addendum 3 RT-8012 And RT-8013 RCB Purge Exhaust Addendum 3 Page 3 of 5 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE A high alarm on RT-8012 or RT-8013 will cause a Containment Ventilation Isolation (CVI). This, in turn, causes RT-8011 sample lines to be isolated and renders RT-8011 radiation monitor inoperable.

_____ 3.0 ENSURE RT-8011 RCB Atmosphere Radiation Monitor Sample Pump -

STOPPED

_____ 4.0 REFER TO Technical Specifications (TS) 3.3.2 And 3.4.6.1 For Further Actions

_____ 5.0 CHECK For Increased Readings On The Following Radiation Monitors:

x RT-8010A RT-8010B x RT-8050 RT-8051 RE-8052 x RE-8053 RE-8054 RE-8055 x RE-8056 RE-8099

_____ 6.0 NOTIFY Chemistry And Health Physics To Obtain Grab Samples As Needed To Confirm Increased Readings On Radiation Monitor RT-8012 Or RT-8013

0POP04-RA-0001 Radiation Monitoring System Alarm Rev. 29 Page 105 of 132

Response

Addendum 30 Basis Basis Page 12 of 38 STEP DESCRIPTION FOR 0POP04-RA-0001 Addendum 2 ADDENDUM: RT-8011 RCB Atmosphere PURPOSE: To provide the operator with response steps specific to the radiation monitor(s).

BASIS: Due to the different applications provided by the radiation monitors, i.e., area (ARMS), process and liquid effluent (PERMS), monitors required by technical specifications, TRM, ODCM, automatic actuations etc., response steps will vary. Depending on which monitor is in alarm the operator performs the steps required for that particular monitor.

ACTIONS: The operator performs the appropriate Addendum steps.

INSTRUMENTATION: N/A CONTROL/EQUIPMENT: An RM-11 computer console is located in the Control Room, health physics, and the TSC. The RM-23 is located in the Control Room. RT-8011 sample isolation valves (MOVs-001, 003, 004, and 006) are located on CP002.

KNOWLEDGE: The RCB Atmospheric Monitor is a Non-Class 1E process and effluent monitor. The monitor has three detectors, one for particulate, one for iodine, and one for noble gas. A high alarm on RT-8012 or RT-8013, RCB purge exhaust radiation monitors will cause a Containment Ventilation Isolation (CVI). The CVI causes the sample valves for RT-8011 to shut and render the monitor inoperable. RT-8011 sample pump should automatically shut down after approximately one minute on a loss of sample flow, however the operator should ensure the sample pump is secured to prevent damage to the sample pump. The Unit Supervisor/Shift Manager may elect not to secure containment purge when it is in progress and the alarm condition does not cause the purge permit requirements to be exceeded. If RT-8011 is not operable then Health Physics will take Grab Samples. Chemistry should also be notified because RT-8011 is used for the RCB Purge Permits.

RT-8011 Particulate monitor is only Required in Modes 1 through 4. (Reference 1.a)

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2213 Last used on an NRC exam: Never RO Sequence Number: 30 Given the following:

A Loss of Coolant Accident (LOCA) has occurred Operators are performing the steps of POP05-EO-EO10, Loss of Reactor or Secondary Coolant SG A, B, C and D pressures are 800, 810, 790 and 450 psig respectively and slowly lowering LOOP A, B, C, and D Tcold are 450, 435, 330 and 440 °F respectively and slowly lowering Reactor Vessel Plenum level indicates 20%

CETs are approximately 375 °F and slowly lowering Which of the following indicates the status of natural circulation cooling and the expected action for these conditions?

Natural Circulation Action Adequate natural circulation A. Reduce number of running ECCS pumps cooling exists Adequate natural circulation B. Maintain current steam flow from SGs cooling exists Adequate natural circulation C. Maintain operation of ECCS pumps cooling does NOT exist Adequate natural circulation D. Open SG PORVs to raise steam flow cooling does NOT exist Answer: C Adequate natural circulation does NOT exist; Maintain operation of ECCS pumps Page 59 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2213 K/A Catalog Number: EPE 011 EA2.09 Tier: 1 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(5)

Ability to determine or interpret the following as they apply to a Large Break LOCA: Existence of adequate natural circulation STP Lesson: LOT 504.25 Objective Number: 92230 STATE/IDENTIFY the indications available to determine Reactor Coolant System Natural Circulation cooldown rate.

Reference:

LOT102.59 student handout page 20 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: NC status is credible because some of the classic indications exist (CETs lowering, loop temperature lowering), however natural circulation is not present with no fluid in the loops (plenum level 20%). Action is credible since E10 does look at stopping LHSI pumps (but based on RCS pressure).

B: INCORRECT: NC status is credible because some of the classic indications exist (CETs lowering, loop temperature lowering), however natural circulation is not present with no fluid in the loops (plenum level 20%). Action is credible because the indicated action would maintain adequate natural circulation (from part one of the distractor)

C: CORRECT: Without fluid in the loops (based on plenum level), natural circulation cannot exist, therefore the correct action is to maintain cooling via ECCS injection.

D: INCORRECT: Action is credible because that is the action that would be taken if NC is not adequate, but only if loops are full.

Question Level: H Question Difficulty 3 Justification:

The applicant must evaluate the given conditions and determine the state of natural circulation. Then based on the determination, select the correct action.

Page 60 of 150

LOT10259GPST4.doc NATURAL CIRCULATION Conditions Required For Natural Circulation Natural circulation is a basic thermal hydraulic phenomenon that occurs during a loss of the Natural circulation will only occur if the correct reactor coolant pumps. Heating and cooling of conditions exist. Even after natural circulation water changes the density of the coolant. As the begins, removal of any one of the required density decreases, a given volume of water has conditions will stop the natural circulation. The less mass. The heated water tends to rise, while following conditions must exist for natural the cooled water tends to fall. This is similar to circulation:

the principles of operation of a hot air balloon.

To rise, heat is added to the gas volume of the A density difference. In all practical hot air balloon. As the hot air cools, the hot air systems, this density difference is produced balloon sinks. Natural circulation is the by a temperature difference. The warmer mechanism by which the coolant is transferred fluid is less dense.

out of the reactor vessel to the steam generators during accident conditions when force A height difference. The cooler, denser circulation is not available which act as a heat fluid must be at a higher elevation than the sink. warmer, less dense fluid.

There are two methods used to cause a fluid Fluids in physical contact with each other.

flow: forced circulation and natural circulation.

Forced circulation requires a pump. Natural There must be two bodies of fluid at different circulation requires no mechanical work and no temperatures. This could also be one body of moving parts. fluid with areas of different temperatures. The difference in temperature is necessary to cause a The driving force for natural circulation flow is density difference in the fluid. The density the difference in density between two adjacent difference is the driving force for natural masses of fluid. circulation flow.

The difference in temperature must be maintained for the natural circulation to continue. Addition of heat by a heat source must exist at the high temperature area. Continuous removal of heat by a heat sink must exist at the low temperature area. Otherwise, the temperatures would equalize and no further circulation would occur.

The warm area must be at a lower elevation than the cool area. A warmer fluid is less dense and will tend to rise, and a cooler fluid is denser and will tend to sink. The greater the elevation differences between the warm and the cool fluid masses, the greater the natural circulation flow rate. This is referred to as the heat sink being above the heat source.

PWR / THERMODYNAMICS / CHAPTER 8 20 of 33 © 2007 GENERAL PHYSICS CORPORATION

/ THERMAL HYDRAULICS REV 4 GF@gpworldwide.com www.gpworldwide.com

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 164 Last used on an NRC exam: Never RO Sequence Number: 31 An EOP mitigation strategy to ensure a secondary heat sink is to maintain a minimum AFW flow of 576 gpm.

In accordance with 0POP05-EO-EO00, which of the following cases allows AFW flow to be reduced to Less Than 576 gpm?

SG NR Level Ctmt Press. Ctmt Rad.

A-30% C-24%

A. 7 psig 1 R/hr B-29% D-33%

A-3% C-17%

B. 3 psig 10 R/hr B-2% D-6%

A-17% C-18%

C. 5 psig 1 x 106 R/hr B-14% D-21%

A-12% C-0%

D. 3 psig 1 x 104 R/hr B-13% D-6%

Answer: B A-3%, C-17%, B-2%, D-6%; 3 psig; 10 R/hr Page 61 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 164 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 504.05 Objective Number: 80399 From memory, STATE/IDENTIFY how total AFW flow is verified to be sufficient in the event of a Safety Injection and/or Reactor Trip.

Reference:

0POP05-EO-EO00 pages 3 and 40 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: Incorrect: Credible because the criteria consists a specified level in one or more SGs which is dependant upon wether or not adverse containment conditions exist (based on pressure and radiation levels). Knowledge of the level criteria and entry conditions for adverse containment plus the ability to apply these criteria are needed to correctly respond. In this case, flow cannot be reduced because adverse containment conditions exist and a SG level is not at least 34%.

B: Correct: Not in adverse containment (pressure less than 5 psig and radiation levels less than 1E5 R/HR) with at least 1 SG greater than 14% NR so total AFW flow can be reduced to less than 576 gpm C: Incorrect: Credible because the criteria consists a specified level in one or more SGs which is dependant upon wether or not adverse containment conditions exist (based on pressure and radiation levels). Knowledge of the level criteria and entry conditions for adverse containment plus the ability to apply these criteria are needed to correctly respond. In this case, flow cannot be reduced because adverse containment conditions exist and a SG level is not at least 34%.

D: Incorrect: Credible because the criteria consists a specified level in one or more SGs which is dependant upon wether or not adverse containment conditions exist (based on pressure and radiation levels). Knowledge of the level criteria and entry conditions for adverse containment plus the ability to apply these criteria are needed to correctly respond. In this case, flow cannot be reduced because no SG level is at least 14% (adverse containment conditions do not exist).

Question Level: F Question Difficulty 3 Justification:

Student must know the values and parameters used for adverse containment and review information to select correct condition satisfied.

Page 62 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 278 Last used on an NRC exam: 2007 RO Sequence Number: 32 A licensed individual has worked the following daytime schedule:

Primary Operator - 11/6 Primary Operator - 11/7 OFF - 11/8 Training - 11/9 Training - 11/10 Training - 11/11 Which of the following correctly identifies the logbook entries the individual is REQUIRED to review per 0POP01-ZQ-0022, Plant Operations Shift Routines, during shift turnover as Primary Operator on 11/12?

Review of Control Room Logbook entries is required A. for only the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. for only the previous 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

C. for only the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. since the individuals last on-shift duty.

Answer: C for only the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Page 63 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 278 K/A Catalog Number: G2.1.3 Tier: 3 Group/Category: 1 RO Importance: 3.7 10CFR

Reference:

55.41(b)(10)

Knowledge of shift or short-term relief turnover practices.

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

Reference:

0POP01-ZQ-0022, R68, step 3.3.4 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because the requiremnet consists of a specified time period. While the indicated time period may seem logical, it is not correct.

B: INCORRECT: Credible because the requiremnet consists of a specified time period. While the indicated time period may seem logical, it is not correct.

C: CORRECT - Procedure requires that on-coming personnel review pertinent information (e.g. special instructions and watchstation logbooks) that have been generated since their last on-shift duty or in the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, whichever is less.

D: INCORRECT: Credible because the requiremnet consists of a specified time period. While the indicated time period may seem logical, it is not correct.

Question Level: H Question Difficulty 3 Justification:

The applicant must evaluate the given data and make a determination regarding required review of logs.

Page 64 of 150

0POP01-ZQ-0022 Rev. 68 Page 6 of 99 Plant Operations Shift Routines 3.3 Shift Turnover 3.3.1 Off-going Watchstanders SHALL complete applicable portions of the Shift Turnover Checklist Form 4, 5, 6, 7, 8 or 19 for the respective watchstation.

3.3.2 Shift turnover SHALL take place at the normal shift watchstations or their designated locations.

3.3.3 Routine business SHOULD NOT be conducted in the control room during the shift turnover process.

3.3.4 On-coming personnel SHALL review pertinent information (e.g. special instructions and watchstation logbooks) that have been generated since their last on-shift duty or in the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, whichever is less.

3.3.5 On-coming and off-going Control Room Watchstanders SHALL walkdown the Control Boards and discuss Shift Turnover Checklist items. This discussion should include, but is NOT limited to:

x Plant Operational Mode x Status of operating systems and components x Abnormal equipment alignments x Inoperable equipment x Equipment under clearance x Abnormal annunciator status x Surveillance or equipment work in progress x Any events occurring during the shift x Evolutions in progress 3.3.6 During the control board walkdown, on-coming Control Room Watchstanders SHALL scan the panels to ensure normally lit indications are illuminated (Reference 9.7).

3.3.7 Off-going watchstanders SHALL remain on watch until one of the following conditions is satisfied: (Reference 9.49) x Their watchstation relief is fully aware of plant conditions.

x IF there is NOT an on-coming relief, THEN watch station plant conditions should be turned over to peer watchstation personnel. (e.g. IF the CP Watch is being secured, THEN ENSURE the TGB Watch is fully aware of the CP watchstation conditions.)

3.3.8 On-coming POs SHALL notify Control Room Personnel that they have assumed the watch as soon as Shift Turnover is complete.

3.3.9 The on-coming Fire Brigade Leader SHALL CONTACT the Off-going Fire Brigade Leader for information that will impact STPEGS as related to fire fighting. (Reference Form 19)

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 854 Last used on an NRC exam: 2010 RO Sequence Number: 33 Given the following:

Assume that today is January 15 of the current year.

A Staff RO, maintaining an active license, has performed the functions of an RO during one 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift since January 1.

Which of the following actions will maintain the ROs license in an active status in accordance with 0POP01-ZA-0014, Licensed Operator License Maintenance?

A. One more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift performing RO functions during January.

Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during February.

B. Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during February.

Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during March.

C. Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during March.

Four more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during April.

D. Four more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during April.

Four more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift performing RO functions during May.

Answer: B Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during February, Two more 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts performing RO functions during March.

Page 65 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 854 K/A Catalog Number: G2.1.4 Tier: 3 Group/Category: 1 RO Importance: 3.3 10CFR

Reference:

55.41(b)(10)

Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license, 10CFR55, etc.

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-ZA-0014, Rev 25, Step 4.3.1 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because an active license is maintained by performing a specified number of watches within a specified time period. Knowledge of the specific requirements and the ability to apply the requirements are needed to correctly respond.

B: CORRECT: Performing this action will result in five watches during the first quarter (meets the requirement).

C: INCORRECT: Credible because an active license is maintained by performing a specified number of watches within a specified time period. Knowledge of the specific requirements and the ability to apply the requirements are needed to correctly respond.

D: INCORRECT: Credible because an active license is maintained by performing a specified number of watches within a specified time period. Knowledge of the specific requirements and the ability to apply the requirements are needed to correctly respond.

Question Level: H Question Difficulty 3 Justification:

The applicant must have a knowledge of the license maintenance requirements and be able to apply those requirements to the given conditions in order to determine the correct response.

Page 66 of 150

0POP01-ZA-0014 Rev. 25 Page 9 of 19 Licensed Operator License Maintenance 4.3 Active License Maintenance NOTE The seven 8-hour or five 12-hour shifts SHALL be within the same calendar quarter AND DO NOT have to be on consecutive shift cycle days. The 8 or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a shift must be consecutive hours and according to current on-shift scheduled working hours. The shift hours must be in the position of US or SM for SRO Watchstations and Primary/Secondary RO for RO Watchstation.

4.3.1 The requirements for maintaining an active license for the next calendar quarter are met when:

4.3.1.1 An individual has completed seven 8-hour shifts OR five 12-hour shifts within the same calendar quarter.

(10CFR55.53 and NUREG 1262) 4.3.1.2 Individual or Supervisor has verified their License and Respirator Physicals are current and valid.

4.3.1.3 An individual is current in Licensed Operator Requalification Training requirements. Licensed Operator Upgrade Training is not a substitute for this requirement.

4.3.2 Only Qual King or the Training Qual Matrix SHALL be used to determine if individuals with active licenses or current STA qualifications may assume the watch.

4.3.3 Operations Administrative Technician, Training Department, or other authorized individual SHALL update the TRDS database as necessary to ensure the database is maintained current.

4.3.3.1 CERT 821 (RO Watchstation Activation/Maintenance) 4.3.3.2 CERT 822 (SRO Watchstation Activation/Maintenance).

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 32 Last used on an NRC exam: 1999 RO Sequence Number: 34 Given the following:

A Reactor Trip occurs due to a Loss of Offsite Power The ESF Diesel Generators have all started and restored power to their ESF buses The Control Room crew has just completed the Immediate Actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection Which of the following correctly identifies the status of Containment Cooling?

A. The RCFCs are running and CCW is flowing through the cooling coils.

B. The RCFCs are running and there is NO flow through the cooling coils.

C. The RCFCs are NOT running and CCW is flowing through the cooling coils.

D. The RCFCs are NOT running and there is NO flow through the cooling coils.

Answer: B The RCFCs are running and there is NO flow through the cooling coils.

Page 67 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 32 K/A Catalog Number: 022 A1.04 Tier: 2 Group/Category: 1 RO Importance: 3.2 10CFR

Reference:

55.41(b)(7)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Cooling water flow.

STP Lesson: LOT 202.33 Objective Number: 4967 State the sources of cooling water to the RCFCs and when each is used.

Reference:

Logics 9-Z-42041 Rev 7, 9-Z-42042 Rev 7, 9-Z-41630 Rev 9 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because this is the normal safety configuration following an ESF actuation.

B: CORRECT - RCFCs are started by the sequencer following a LOOP, and cooling flow to the RCFCs is isolated by the sequencer following a LOOP.

C: INCORRECT - Fan status is credible because a component will either start or not start following a LOOP, and RCB pressure/temperature would not be a factor for a period of time, so it would be reasonable to conclude the fans would not be needed immediately following a LOOP. Flow status is credible because the CCW pumps are started (and needed) following a LOOP. Knowledge of system design is required to choose the correct response.

D: INCORRECT - Fan status is credible because a component will either start or not start following a LOOP, and RCB pressure/temperature would not be a factor for a period of time, so it would be reasonable to conclude the fans would not be needed immediately following a LOOP. Knowledge of system design is required to choose the correct response.

Question Level: H Question Difficulty 4 Justification:

The applicant must analyze the given conditions and: 1) Determine that upon a LOOP power is lost to the RCB chill water system which normally supplies cooling to the RCFCs. 2) Determine that the chill water supply valves to the RCFCs are closed and recognize that the CCW supply valves to the RCFCs are closed because an SI signal is not present. Thus, there is no flow through the RCFC cooling coils.

3) Recognize that the RCFCs are sequenced on to their respective ESF buses on a LOOP signal.

Page 68 of 150

Obj. 7 RCFC CONTROL LOGIC ZLP-002 HS-9666 MT MO SR MO PULL TO STOP AUTO START SEQUENCER LOCK CP-002 BUS STRIP SIGNAL SEQUENCER ZLP 700 Train A MODELS I, II OR III ZLP 701 Train B THIS DWG D D THIS DWG ZLP 709 Train CC F THIS DWG E THIS DWG ZLP-XXX ZLP-780 HS-9666 HS-9666B TYPE 3 TYPE 4 (NOTE 2)

MT (NOTE 2) MT MO SR MO STOP START RCFC CR LOCAL STOP NORMAL START SUPPLY FAN #11A (2-VM-1-FN001)

STOPPED RUNNING D THIS DWG C THIS DWG LOT202.33.TP.23 E THIS DWG DWG 07/25/00 F THIS DWG

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 82 Last used on an NRC exam: 1999 RO Sequence Number: 35 Unit 1 Train A AND Train C 4.16 KV ESF Busses were de-energized with an expected duration of 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Train A 4.16 KV ESF Bus has been re-energized using the Train B ESF Diesel Generator per 0POP04-AE-0004, Loss of Power To One or More 4.16 KV ESF Bus.

Which of the following describes the BASIS for this action?

A. To enable the start of CCP A for RCS Inventory Control.

B. To enable the start of CCW Pump A for Spent Fuel Pool cooling.

C. To extend the use of the Plant Computer System for accident monitoring.

D. To maintain accident monitoring instruments energized by getting a Train A charger in service.

Answer: D To maintain accident monitoring instruments energized by getting a Train A charger in service.

Page 69 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 82 K/A Catalog Number: 062 K1.03 Tier: 2 Group/Category: 1 RO Importance: 3.5 10CFR

Reference:

55.41(b)(10)

Knowledge of the physical connections and/or cause-effect relationships between the AC distribution system and DC distribution.

STP Lesson: LOT 201.37 Objective Number: 92047 State how the Class 1E 125 VDC System interfaces with other systems.

Reference:

0POP04-AE-0004 Rev 13, pages 77 and 88 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible since this would be desireable, however CCP A is powered from E1C (Train C 4.16 KV)

B: INCORRECT - Credible since maintaining SFP cooling is desireable, however any CCW pump can supply SFP cooling and it is given that Train B diesel is available.

C: INCORRECT - Credible since the basis is to maintain accident monitoring instrumentation.

However the Plant Computer does not support accident monitors instrumentation nor does it have a Class 1E power supply (QDPS computer system would have been correct).

D: CORRECT - The basis for energizing the Train A bus with the Train B DG is to extend the Train A battery life for accident monitoring instrumentation.

Question Level: F Question Difficulty 3 Justification:

Requires a knowledge of the basis for the procedure steps in 0POP04-AE-0004 Page 70 of 150

0POP04-AE-0004 Loss Of Power To One Or More Rev. 13 Page 77 of 95 4.16 KV ESF Bus Addendum 14 Basis Basis Page 1 of 18 PROCEDURE PURPOSE The purpose of this procedure is to restore power to any ESF bus which is not energized. In the case where only one ESF bus is energized by a DG, and another one cannot be energized by the associated DG or offsite power, then steps are taken to operate breakers and disconnects to use the one running DG to supply key loads on another bus.

MAJOR ACTION CATEGORIES x Tie the operating DG to another bus via the emergency switchgear bus 1L(2L).

x Energize at least one ESF bus from the Emergency Transformer.

x Control and load essential equipment on to the available ESF buses.

DISCUSSION:

STP has committed under specific conditions related to loss of offsite and onsite power to energize at least two ESF buses from a running DG in order to energize specific loads needed to extend station battery life or provide availability of ESF equipment that is electrically powered from one of two specific ESF buses.

This Procedure is Applicable in all Modes

0POP04-AE-0004 Loss Of Power To One Or More Rev. 13 Page 88 of 95 4.16 KV ESF Bus Addendum 14 Basis Basis Page 12 of 18 STEP DESCRIPTION FOR 0POP04-AE-0004 Addendum 4 AND Addendum 6 STEP: Various PURPOSE: To provide the actions necessary to energize an ESF bus from an already energized bus via the Emergency Bus 1L(2L).

BASIS: These two Addendum provide the switching instructions necessary to connect either A or C ESF bus to B ESF bus when it is energized from its associated DG. Included are steps for reloading the energized bus and shedding of loads in the event that A or C bus can not be energized. STP is committed in ST-AE-HL-94678 to having a method to energize any ESF bus from an operating DG. At Step 31.0 of Addendum 4, actions are taken to maintain accident monitoring instruments energized by getting a battery charger in service or by shedding non accident monitoring loads. Steps 41.0 and 42.0 of Addendum 4 to isolate air to MSIVs may be needed because of the MSIV Energize to Actuate DCP (Unit 1: DCP 00-01937-90, Unit 2:

DCP 00-01937-91). The value of 105.5 VDC to open the battery output breaker comes from Calculation EC-5008, Rev. 13, sheet 252. The Caution uses 105 volts as the minimum and opening at 105.5 provides some margin to equipment damage.

ACTIONS: Energize either A or C ESF bus from B ESF bus via the Emergency Switchgear Bus 1L(2L).

INSTRUMENTATION: N/A CONTROL/EQUIPMENT: Various breakers and disconnects are operated to accomplish this action.

KNOWLEDGE: Due to the planned ESF Bus outages in conjunction with the DG 22 EAOT in 2RE10 outage, steps are included to ensure SFP pump in service for modes 5, 6 and core offloaded to SFP. Having steps to place SFP cooling in service on a loss of power event help the risk profile for 2RE10.

This Procedure is Applicable in all Modes

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 432 Last used on an NRC exam: 1997 RO Sequence Number: 36 A tube leak has occurred in 1D Steam Generator. The Unit is currently performing a rapid plant shutdown for repair.

Under these conditions, failure of which of the following Process and Effluent radiation monitors would allow a release to the environment that would otherwise be automatically prevented?

A. Condenser Air Removal System Monitor (RT-8027)

B. Steam Generator 1D Blowdown Monitor (RT-8025)

C. Turbine Generator Building Drain Monitor (RT-8041)

D. Main Steam Line D Monitor (RT-8049)

Answer: C Turbine Generator Building Drain Monitor (RT-8041)

Page 71 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 432 K/A Catalog Number: 073 K3.01 Tier: 2 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Knowledge of effect that a loss or malfunction of the PRM will have on the following: radioactive effluent releases.

STP Lesson: LOT 202.41 Objective Number: 80695 STATE the basis for the automatic actions provided by the process radiation monitors.

Reference:

0POP04-RA-0001 page 43 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because this monitor will read upscale during the event, however there is no automatic function associated with this monitor.

B: INCORRECT: Credible because this monitor will read upscale during the event, however there is no automatic function associated with this monitor. There is an automatic function associated with the Steam Generator Blowdown System (RT-8043) that would prevent a release to the environment.

C: CORRECT: Prevents potential release; automatic function stops the TGB Sump #1 pump from discharging to the Oily Waste Storage Tank (outside in the yard).

D: INCORRECT: Credible because this monitor will read upscale during the event, however there is no automatic function associated with this monitor.

Question Level: F Question Difficulty 3 Justification:

Applicant must determine from the available distracters which monitor performs an automatic function to prevent a release to the enviroment.

Page 72 of 150

0POP04-RA-0001 Radiation Monitoring System Alarm Rev. 29 Page 43 of 132

Response

Addendum 15 RT-8041, TGB Drain Monitor Addendum 15 Page 1 of 1 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 1.0 CHECK HIGH Alarm Exists On Radiation GO TO Step 3.0 of this Addendum.

Monitor RT-8041

_____ 2.0 ENSURE TGB Sump Number 1 Sump Pumps 1A(2A), 1B(2B), 1C(2C), And 1D(2D) - STOPPED {TGB 29 west}

_____ 3.0 PERFORM The Following:

_____ a. NOTIFY Chemistry Of The Alarm Condition

_____ b. REQUEST Chemistry Sample The Monitor For Radioactivity

_____ 4.0 NOTIFY Health Physics Of The Alarm Condition

_____ 5.0 GO TO Procedure And Step In Effect

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 478 Last used on an NRC exam: 1995 RO Sequence Number: 37 Conditions have occurred while responding to a Reactor Trip which requires Emergency Boration be initiated. The operator attempts to emergency borate using the Normal Emergency Boration Flowpath, but is unsuccessful at starting a Boric Acid pump.

Which of the following contains two flowpaths, each of which would independently meet the Emergency Boration requirements of 0POP04-CV-0004, Emergency Boration, for the given condition?

A. Boration through Normal Boration Flowpath OR Emergency Boration through "1(2)-CV-0221 MANUAL ALTERNATE IMMEDIATE BORATE" B. Emergency Boration from RWST OR Emergency Boration via Gravity Feed C. Emergency Boration through "1(2)-CV-0221 MANUAL ALTERNATE IMMEDIATE BORATE" OR Emergency Boration via Gravity Feed D. Emergency Boration via Gravity Feed OR Boration through Normal Boration Flowpath Answer: B Emergency Boration From RWST OR Emergency Boration Via Gravity Feed Page 73 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 478 K/A Catalog Number: 004 K6.17 Tier: 2 Group/Category: 1 RO Importance: 4.4 10CFR

Reference:

55.41(b)(6)

Knowledge of the effect of a loss or malfunction on the following CVCS Components: Flow paths for emergency boration STP Lesson: LOT 201.07 Objective Number: 91060 DESCRIBE the steps necessary to commence an emergency boration in accordance with 0POP04-CV-0003, Emergency Boration.

Reference:

0POP04-CV-0003, Rev 12, Emergency Boration (Pgs 3 & 5)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because both are valid flowpaths for getting boric acid to the RCS, however Normal Boration Flow Path will not meet the flow requirements without a Boric Acid pump running and Emergency Boration through 1(2) CV-0221Manual Alternate Immediate Borate Valve is not a method listed in the procedure.

B: CORRECT - The methods identified meet the flow requirments and are identified in the procedure.

C: INCORRECT - Credible because Emergency Boration through 1(2) CV-0221Manual Alternate Immediate Borate Valve will get acid to the RCS, but is not a method listed in the procedure. The second method is correct.

D: INCORRECT - First method is correct. The second method is credible because it will get acid to the RCS under normal conditions, but requires a boric acid pump.

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze that the inability to start a boric acid pump will mean some of the available flow paths will not provide adequate flow and recall the minimum required flow and alternate flow path identified in the procedure.

Page 74 of 150

0POP04-CV-0003 Emergency Boration Rev. 12 Page 3 of 39 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

____ 1.0 CHECK Boric Acid Storage Tank - GO TO Addendum 1, Emergency Boration OPERABLE OR AVAILABLE From the RWST.

_____ 2.0 ENSURE CVCS Charging Line OCIV DISPATCH an operator to open MOV-0025 - OPEN "1(2)-CV-MOV-0025" "CVCS CHG LINE OCIV." {29 ft MAB RM 108C}

_____ 3.0 ENSURE One Of The Following Valves Are Open:

• Normal Charging LOOP A ISOL MOV-0003.

OR

• Alternate Charging LOOP C ISOL MOV-0006.

____ 4.0 CHECK Charging Pump - RUNNING PERFORM the following:

a. CLOSE all seal injection OCIVs:

• RCP 1A(2A) SEAL INJ ISOL MOV-0033A

• RCP 1B(2B) SEAL INJ ISOL MOV-0033B

• RCP 1C(2C) SEAL INJ ISOL MOV-0033C

• RCP 1D(2D) SEAL INJ ISOL MOV-0033D

b. CLOSE the discharge valve for the Centrifugal Charging Pump to be started:

• CCP 1A(2A) DISCH ISOL MOV-8377A

• CCP 1B(2B) DISCH ISOL MOV-8377B Step 4.0 continued on next page This Procedure is Applicable in Modes 1-5

0POP04-CV-0003 Emergency Boration Rev. 12 Page 5 of 39 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 5.0 PERFORM Both Of The Following: PERFORM all of the following:

• START a BA Transfer Pump {CP004} a. COMMENCE boration using the normal boration flowpath.

• OPEN "ALT BORATION ISOL b. DISPATCH an operator to perform the MOV-0218" {CP004} following:

1) IF Boric Acid Tank A is available, THEN DISPATCH an operator to perform the following:

• OPEN "1(2)-CV-0333 BORIC ACID TANK 1A(2A) CHARGING PUMP SUCTION ISOLATION VALVE." {19 ft MAB Room 076}

• UNLOCK AND OPEN "1(2)-CV-0226 BORIC ACID TANK TO CHARGING PUMP SUCTION BORATION VALVE."

{19 ft MAB Room 079}

Step 5.0 continued on next page This Procedure is Applicable in Modes 1-5

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 490 Last used on an NRC exam: 1995 RO Sequence Number: 38 While operating in Mode 4, annunciator window 5M02-B-7, RCS COLD OVERPRESS ALERT-TRN B, illuminates. The operator notes that COMS has NOT actuated.

Which of the following instrument failures could be the cause of the annunciator?

A. Loop C WIDE range cold leg temperature failed High (TE-434)

B. Loop C WIDE range cold leg temperature failed Low (TE-434)

C. Loop C NARROW range cold leg temperature failed High (TE-430)

D. Loop C NARROW range cold leg temperature failed Low (TE-430)

Answer: B Loop 'C' WR cold leg temperature failed low (TE-434)

Page 75 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 490 K/A Catalog Number: 002 K4.10 Tier: 2 Group/Category: 2 RO Importance: 4.2 10CFR

Reference:

55.41(b)(7)

Knowledge of RCS design feature(s) and/or interlock(s) which provide for the following: Overpressure protection STP Lesson: LOT 201.14 Objective Number: 80414 STATE the pressurizer pressure and level control system actuation signals, setpoints, logic, coincidence, and interlocks.

Reference:

0POP04-RP-0005, Rev 13, (Pg 6); LOT 201.14, Rev 14, (Pg 6)

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because temperature is an input. The applicant must understand COMS design/operation to correctly respond.

B: CORRECT - A lowering temperature will lower the COMS lift setpoint and cause the alarm to illuminate (alarm comes in when pressure is within 20 psi of lift setpoint).

C: INCORRECT - Credible because temperature is an input. The applicant must understand COMS design/operation to correctly respond.

D: INCORRECT - Credible because temperature is an input. The applicant must understand COMS design/operation to correctly respond.

Question Level: H Question Difficulty 3 Justification:

The applicant has to recall that COMS Train B receives input from auctioneered Lo WR Tcold and that only the Alert alarm is actuated if COMS is not ARMED.

Page 76 of 150

0POP04-RP-0005 COMS Actuation Or Failure Rev. 13 Page 6 of 20 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 4.0 CHECK COMS Actuation Due To Instrument Malfunction

_____ a. VERIFY following channels - a. PERFORM the following:

OPERABLE: 1) PLACE the Cold "OVERPRESSURE

• PT-403, Loop B Wide range pressure MIT" Switch for PCV-0655A in (QDPS) BLOCK.

• TE-413, Loop A W/R Hot Leg (QDPS) 2) IF PCV-0655A is closed, THEN ENSURE PCV-0655A isolation valve

• TR-413 (CP 018) open.

• TE-423, Loop B W/R Hot Leg (QDPS) 3) ENSURE PCV-0655A handswitch in AUTO.

• TR-423 (CP 018)

• TE-433, Loop C W/R Hot Leg (QDPS)

• TR-433 (CP 018)

• TE-443, Loop D W/R Hot Leg (QDPS)

• TR-443 (CP 018)

_____ b. VERIFY following channels - b. PERFORM the following:

OPERABLE: 1) PLACE the Cold "OVERPRESSURE

• PT-404, Loop C Wide range pressure MIT" Switch for PCV-0656A in (QDPS) BLOCK.

• TE-414, Loop A W/R Cold Leg (QDPS) 2) IF PCV-0656A is closed, THEN ENSURE PCV-0656A isolation valve

• TR-413 (CP 018) open.

• TE-424, Loop B W/R Cold Leg (QDPS) 3) ENSURE PCV-0656A handswitch in AUTO.

• TR-423 (CP 018)

• TE-434, Loop C W/R Cold Leg (QDPS)

• TR-433 (CP 018)

• TE-444, Loop D W/R Cold Leg (QDPS)

• TR-433 (CP 018)

This Procedure is Applicable in Modes 3, 4, 5, And 6 With The Head On The Reactor Vessel

0POP04-RP-0005 COMS Actuation Or Failure Rev. 13 Page 9 of 20 Addendum 1 Cold Overpressure Limits Addendum 1 Page 1 of 2 LOW PORV, PCV-0656A This Procedure is Applicable in Modes 3, 4, 5, And 6 With The Head On The Reactor Vessel

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 492 Last used on an NRC exam: 1995 RO Sequence Number: 39 What is the minimum configuration of ECCS equipment assumed by the FSAR to inject into the reactor vessel to assure adequate core cooling in the event of the design basis LOCA?

A. Two HHSI pumps, two LHSI pumps, two Accumulators B. Two HHSI pumps, two LHSI pumps, one Accumulator C. One HHSI pump, one LHSI pump, two Accumulators D. One HHSI pump, one LHSI pump, one Accumulator Answer: C One HHSI pump, one LHSI pump, two Accumulators Page 77 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 492 K/A Catalog Number: 006 K1.03 Tier: 2 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(8)

Knowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: RCS STP Lesson: LOT 201.10 Objective Number: 4123 State the function of the ECCS and each of it major components.

Reference:

LOT201.10,HO.01 handout page 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT- Credible because some combination of HHSI, LHSI and acumulators is required.

The applicant must have knowledge of these requirements to choose the correct response.

B: INCORRECT- Credible because some combination of HHSI, LHSI and acumulators is required.

The applicant must have knowledge of these requirements to choose the correct response.

C: CORRECT - Minimum ECCS flow assumes one train of ECCS fails to start, one train dumps to containment through initiating break, and one train reaches core. Accumulators assume one accumulator dumps to containment through initiating break, and two accumulators reach core.

D: INCORRECT- Credible because some combination of HHSI, LHSI and acumulators is required.

The applicant must have knowledge of these requirements to choose the correct response.

Question Level: F Question Difficulty 3 Justification:

The applicant must recall the minimum ECCS equipment for the design basis LOCA to ensure core cooling.

Page 78 of 150

LOT201.10.HO.01 Rev. 18 PAGE 3 OF 30 In the event of a break which maintains RCS pressure > LHSI pump shut-off head, flow provided from one HHSI pump and two Accumulators is sufficient to meet minimum ESF performance criteria.

In Summary, A minimum of 2 Accumulators delivering to two unaffected loops and one HHSI and one LHSI pump delivering to an unaffected loop will assure adequate core cooling for the design basis LOCA.

Safe Shutdown Assessment Contained in "STP Fire Hazard Analysis Report" (FHAR) and used to show compliance with the requirements of 10CFR50 Appendix R, section III.G Fire Protection of Safe Shutdown Capability, and III.L, Alternative and dedicated Shutdown Capability.

Safe shutdown analysis assumes pressure control capability during a cooldown/depressurization.

During depressurization of the RCS, the SI Accumulators need to be isolated or depressurized to prevent injection and allow the RCS to be depressurized. The Accumulator isolation valves and Accumulator nitrogen venting valve are required for this capability.

If closing their outlet valves cannot isolate the Accumulators; the gas pressure will be vented to the RCB atmosphere. Either way, injection of the contents would be prevented.

FLOWPATHS The injection phase is defined as period in which borated water is delivered from the RWST and Accumulators to the RCS cold legs.

The ECCS minimizes and prevents core damage by rapidly refilling the reactor vessel and reflooding the core providing short term core cooling and terminates reactivity increases.

During the injection phase, the RWST provides borated water to 3 CS/SIS suctions connecting off the main header.

The HHSI pumps inject when RCS pressure is less than 1600 psig and are normally lined up for cold leg injection. The pumps are provided with miniflow return lines to the RWST to protect running against shut-off head.

The Accumulators inject stored borated water when RCS pressure is 590-670 psig.

Accumulator injection pressure is established by nitrogen and they can only inject into the cold legs.

LHSI pumps inject when RCS pressure is less than 300 psig and are normally lined up through the RHR heat exchanger for cold leg injection. The pumps are provided with miniflow returns lines to RWST to protect running against shutoff head.

The cold leg recirculation phase is defined as that period in which borated water is recirculated from containment sumps to the RCS cold legs via LHSI/HHSI pumps.

The cold leg recirculation phase terminates core boiling and is initiated automatically by the Auto-Recirc signal:

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2220 Last used on an NRC exam: Never RO Sequence Number: 40 Emergency Diesel Generator Trip Solenoids have Class 1E and Non-Class control power.

Which of the following states the source of the Trip Solenoids control power?

Class 1E Non-Class Control Power Control Power A. Class 1E 120 Volt Vital AC Non-Class 125 Volt DC B. Class 1E 125 Volt DC Non-Class 125 Volt DC C. Class 1E 120 Volt Vital AC Non-Class 120 Volt Vital AC D. Class 1E 125 Volt DC Non-Class 120 Volt Vital AC Answer: B Class 1E 125 Volt DC - Non-Class 125 Volt DC Page 79 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2220 K/A Catalog Number: 064 K2.03 Tier: 2 Group/Category: 1 RO Importance: 3.2 10CFR

Reference:

55.41(b)(7)

Emergency Diesel Generators (ED/G)

Knowledge of bus power supplies to the following:

Control Power STP Lesson: LOT 201.39 Objective Number: 44288 STATE the normal source of power for the Emergency Diesel Generator system, sub systems and components.

Reference:

LOT 201.39 Powerpoint slide 191 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because solenoids can be either AC or DC powered. The applicant must have knowledge of system design to correctly respond.

B: CORRECT: Emergency Diesel Generator Emergency Trip Solenoids are powered from Class 1E 125 VDC and the Non-Emergency Trip Solenoids are powered from Non-Class 125 VDC.

C: INCORRECT: Credible because solenoids can be either AC or DC powered. The applicant must have knowledge of system design to correctly respond.

D: INCORRECT: Credible because solenoids can be either AC or DC powered. The applicant must have knowledge of system design to correctly respond.

Question Level: F Question Difficulty 3 Justification:

The Reactor Operator must have knowledge of power supplies to different components of the Emergency Diesel Generator system.

Page 80 of 150

EMER 1E 1E Non 1E Mode 191

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 922 Last used on an NRC exam: Never RO Sequence Number: 41 Upon receipt of a Safety Injection signal, Pressurizer heaters that are supplied by ESF busses are de-energized.

Which of the following describes the Pressurizer heaters that will be de-energized?

A. Proportional Heater Group C and Backup Heaters Group A and B.

B. Proportional Heater Group C and Backup Heaters Group D and E.

C. Only Backup Heaters Group A and B.

D. Only Backup Heaters Group D and E.

Answer: C Only Backup Heaters Group A and B.

Page 81 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 922 K/A Catalog Number: 010 K2.01 Tier: 2 Group/Category: 1 RO Importance: 3.0 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 power supplies for pressurizer heaters.

Reference:

LOT201.14 handout #1, page 4 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - This distractor is credible because it shows a lack of knowledge with PZR Heater power supplies. Only the ESF electrical powered heaters are shed. The control heaters and backup heaters D and E are nonclass power and are only controlled from pressurizer pressure control system and low pressurizer level.

B: INCORRECT - This distractor is credible because it shows a lack of knowledge with PZR Heater power supplies. Only the ESF electrical powered heaters are shed. The control heaters and backup heaters D and E are nonclass power and are only controlled from pressurizer pressure control system and low pressurizer level.

C: CORRECT - The A and B backup heaters are powered from class 1E powered 480V LC E1A1 and E1C1 respectively.

D: INCORRECT - This distractor is credible because it shows a lack of knowledge with PZR Heater power supplies. Only the ESF electrical powered heaters are shed. The control heaters and backup heaters D and E are nonclass power and are only controlled from pressurizer pressure control system and low pressurizer level.

Question Level: F Question Difficulty 3 Justification:

The applicant is required to recall that the ESF sequencer strips only backup heaters A and B that powered from ESF electrical busses.

Page 82 of 150

LOT201.14.01. Rev. 14 Page 4 of 38 NOTE: Slide 9 HEATER POWER SUPPLIES AND KW PZR HTR B/U GRP A LC-E1A1 431 KW PZR HTR B/U GRP B LC-E1C1 431 KW CONTROL GROUP C LC-1N 485 KW PZR HTR B/U GRP D LC-1P 377 KW PZR HTR B/U GRP E LC-1J2 377 KW TOTAL 2101 KW NOTE: Slide 10 The ESF powered Group A and B heaters will be de-energized on an ESF actuation. After the actuation has been reset, the heater hand switch on CP-004 must be taken to OFF to clear the seal-in signal which turned the theaters off. Then the heaters can be energized (providing level in the pressurizer is >17%).

This feature only applies to the ESF powered heaters. This is to prevent the heaters from cycling on when the ESF bus may be powered from the ESF diesel.

NOTE: Slides 11-14 COLD OVERPRESSURE MITIGATION SYSTEM (COMS)

Provides protection against RCS over-pressurization when temperature is below RTNDT.

The system uses the power operated relief valves (PORVs). Inputs to the COMS are:

1. Train A COMS (PCV-655A) which receives signals from auctioneered low RCS wide range TH and wide range RCS pressure channel 403.

2. Train B COMS (PCV-656A) which receives signals from auctioneered low RCS wide range TC and wide range RCS pressure channel 404.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 940 Last used on an NRC exam: Never RO Sequence Number: 42 The Reactor tripped and Auxiliary Feedwater has actuated.

Which of the following completes the statement concerning the heat transfer relationship between the RCS and SGs?

The heat transfer rate between the RCS and the SGs will A. lower as RCS temperature rises and AFW flow rises.

B. lower as AFW temperature lowers and AFW flow rises.

C. rise as AFW temperature rises and RCS flow lowers.

D. rise as RCS temperature rises and AFW flow rises.

Answer: D rise as RCS temperature rises and AFW flow rises.

Page 83 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 940 K/A Catalog Number: 061 K1.04 Tier: 2 Group/Category: 1 RO Importance: 3.9 10CFR

Reference:

55.41(b)(14)

Knowledge of the physical connections and/or cause-effect relationships between the AFW and the following systems: RCS STP Lesson: LOT 102.58 Objective Number: N99867 Describe three mechanisms of heat transfer.

Reference:

LOT102.58 GP instructor guide page 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because temperature and/or flow changes in either the hot or cold fluids will affect heat transfer rate. The applicant must closely analyze the given conditions to determine if the indicated changes in the two systems will cause the transfer rate to change as indicated.

B: INCORRECT - Credible because temperature and/or flow changes in either the hot or cold fluids will affect heat transfer rate. The applicant must closely analyze the given conditions to determine if the indicated changes in the two systems will cause the transfer rate to change as indicated.

C: INCORRECT - Credible because temperature and/or flow changes in either the hot or cold fluids will affect heat transfer rate. The applicant must closely analyze the given conditions to determine if the indicated changes in the two systems will cause the transfer rate to change as indicated.

D: CORRECT - as the RCS heats up and AFW flow rises, a large delta T is created raising the heat transfer rate.

Question Level: H Question Difficulty 3 Justification:

The applicant must recall the cause and effect relationship of differential temprature and heat transfer rate and apply the concept to the given situation.

Page 84 of 150

102.58 GPIG3.doc KEY POINTS, AIDS, INSTRUCTOR GUIDE QUESTIONS/ANSWERS Q = total heat added (Btu)

Cp = heat capacity (Btu/°F)

T = change in temperature (°F) m = mass (lbm) cp = specific heat (Btu/lbm °F)

B. The British thermal unit (Btu) is unit of heat energy and is defined in terms of these relationships

1. One Btu is defined as amount of heat required to raise temperature of one pound-mass of water at standard atmospheric pressure by one degree Fahrenheit C. In most power plant applications, heat is added to flowing fluids rather than stagnant bodies

1. For these applications, it is convenient to rewrite these relationships in terms Equation 7-2 of heat addition or removal rate ( Q) m Q c p T and mass flow rate ( m )

Where:

Q = heat addition or removal rate (Btu/hr) m = mass flow rate (lbm/hr) cp = specific heat (Btu/lbm °F)

T = change in temperature (°F)

Example 7-1 Calculate rate of heat addition for a heat exchanger operating with these conditions:

Coolant temperature in = 535°F Coolant temperature out = 551°F Coolant flow rate = 7 107 lbm/hr PWR / THERMODYNAMICS / CHAPTER 7 8 of 73 © 2011 GENERAL PHYSICS CORPORATION

/ HEAT TRANSFER REV 3 GF@gpworldwide.com www.gpworldwide.com

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1330 Last used on an NRC exam: 2011 RO Sequence Number: 43 Which one of the following correctly describes the SEQUENCE of events as Instrument Air pressure lowers from the normal operating value?

A. -Air Compressor 14 (24) starts/loads.

-Service Air Isolation Valve PV-9785 closes.

-Instrument Air to Yard Valve PV-8568 closes.

-Instrument Air Dryer Bypass Valve PV-9983 opens.

B. -Instrument Air to Yard Valve PV-8568 closes.

-Air Compressor 14 (24) starts/loads.

-Instrument Air Dryer Bypass Valve PV-9983 opens.

-Service Air Isolation Valve PV-9785 closes.

C. -Air Compressor 14 (24) starts/loads.

-Instrument Air Dryer Bypass Valve PV-9983 opens.

-Service Air Isolation Valve PV-9785 closes.

-Instrument Air to Yard Valve PV-8568 closes.

D. -Service Air Isolation Valve PV-9785 closes.

-Air Compressor 14 (24) starts/loads.

-Instrument Air Dryer Bypass Valve PV-9983 opens.

-Instrument Air to Yard Valve, PV-8568 closes.

Answer: A -Air Compressor 14 (24) starts/loads.

-Service Air Isolation Valve PV-9785 closes.

-Instrument Air to Yard Valve PV-8568 closes.

-Instrument Air Dryer Bypass Valve PV-9983 opens.

Page 85 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1330 K/A Catalog Number: 078 K1.02 Tier: 2 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(4)

Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: Service Air STP Lesson: LOT 202.26 Objective Number: 92995 Given a scenario in which Instrument Air pressure is decreasing, PREDICT Instrument and Service Air system component automatic actions that will occur as pressure decreases.

Reference:

0POP04-IA-0001, Loss Of Instrument Air, Rev. 16 pg 2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT: This is the correct sequence IAW the system off-normal procedure.

B: INCORRECT: Credible because all the actions occur. The applicant must have knowledge of system design/operation to respond correctly.

C: INCORRECT: Credible because all the actions occur. The applicant must have knowledge of system design/operation to respond correctly.

D: INCORRECT: Credible because all the actions occur. The applicant must have knowledge of system design/operation to respond correctly.

Question Level: F Question Difficulty 3 Justification:

Applicant must know the automatic actions and setpoints for the IA and SA systems.

Page 86 of 150

0POP04-IA-0001 Loss Of Instrument Air Rev. 16 Page 2 of 152 PURPOSE This procedure provides the necessary operator actions for responding to a significant degradation or loss of Instrument Air (IA) capacity.

Instrument Air Pressure Automatic Actuation (Decreasing)

IA Compressor 11(21) Starts/Loads in Local 122 psig Control IA Compressor 12(22) Starts/Loads in Local 119 psig Control IA Compressor 13(23) Starts/Loads in Local 116 psig Control IA Compressor 14(24) (air cooled and BOP DG 113 psig powered) Starts/Loads Service Air Isolation Valve N1(2)IA-PV-9785 100 psig Closes Instrument Air to Yard Valve 90 psig N1(2)IA-PV-8568 Closes Instrument Air Dryer Bypass 80 psig N1(2)IA-PV-9983 Opens SYMPTOMS OR ENTRY CONDITIONS

1. The following Control Room annunciator alarms:

x SAS ISOL VLV CLOSE Lampbox 08M3, Window F-3 x SAS HDR PRESS LO Lampbox 08M3, Window E-3 x IAS HDR PRESS LO Lampbox 08M3, Window D-3

2. All operable IA compressors running continuously.

3. No IA compressors running.

4. Various air operated valves observed to be drifting to failure positions.

This Procedure is Applicable in All Modes

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1401 Last used on an NRC exam: 2009 RO Sequence Number: 44 Given the following conditions:

Unit 1 is in Mode 5 RHR Train A is in service providing shutdown cooling.

FV-8565, IA OCIV, subsequently fails closed.

Which of the following correctly describes the effect of the valve failure?

RHR Train A is.....

A. AVAILABLE to provide shutdown cooling since instrument air accumulator tanks in containment will continue to supply the necessary air and allow normal operation.

B. AVAILABLE to provide shutdown cooling since the RHR Heat Exchanger Outlet valve fails open and the RHR Heat Exchanger Bypass valve fails closed providing full cooling flow.

C. NOT available to provide shutdown cooling since the RHR Heat Exchanger Outlet valve fails closed and the RHR Heat Exchanger Bypass valve fails open providing no cooling flow.

D. NOT available to provide shutdown cooling since the RHR Pump Recirculation valve fails open which would not allow adequate cooling water flow to reach the RCS.

Answer: B AVAILABLE to provide shutdown cooling since the RHR Heat Exchanger Outlet valve fails open and the RHR Heat Exchanger Bypass valve fails closed providing full cooling flow.

Page 87 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1401 K/A Catalog Number: G2.2.37 Tier: 3 Group/Category: 2 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Ability to determine operability and/or availability of safety related equipment.

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

Reference:

LOT201.09, RHR System, PowerPoint slides 14 and 15 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible since some plant air operated valves have air accumulators for increased reliability during loss of air scenarios.

B: CORRECT - The valves fail as indicated, thus providing full cooling flow to the RCS.

C: INCORRECT - Credible since air operated valves can fail in either direction, depending on design.

The applicant must understand design/safety considerations for the system.

D: INCORRECT - Credible because the pumps have a large recirc valve. But unlike most pumps which have an air operated recir that fails open to protect the pump, the RHR pump recirc is motor operated (which is not affected by loss of air.

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze the given conditions to determine the affect on the system and its availability.

Page 88 of 150

RHR Heat Exchanger Flow Control Valve (9-Z-42183 Rev. 8)

RHR RHR HEAT HEAT EXCHANGER EXCHANGER FLOW FLOW CONTROL CONTROL ZCP-001 ZCP-001 HC HC 864 864 CONTROL CONTROLXFERRED XFERRED TO TOAUX.

AUX. SHUTDOWN SHUTDOWN PANEL PANEL ZLP-100 ZLP-100 AA HK HK 864 ERF ERF 864 DI DI RHHD0864 RHHD0864 TT AA FY FY 3860 3860 II PP f(x) f(x) 9-Z-41811 9-Z-41811 AA BYPASS STATUS BYPASS STATUS 125 125 VDC VDC INOPERATIVE INOPERATIVE LOGIC LOGIC TRAIN TRAINAA SS ERF ERF DI DI HCV HCV EE EE RHZD0864 RHZD0864 864 864 AA ZLP-659 ZLP-659 ZSC ZSC ZSO ZSO RR GG 0864 0864 0864 0864 ZCP-001 ZCP-001 ZCP-001 ZCP-001 0%

0% 100%

100%

F.O.

F.O.

RHR RHR HEAT HEAT EXCHANGER EXCHANGER LOT 201.09 SLIDE 13 FLOW FLOW CONTROL CONTROLVALVEVALVE

RHR RHRHX HXBYPASS BYPASS FLOW FLOWCONTROL CONTROL FT FT 851 851 FC FC 851 851

++

KK FK FK ZCP-100 ZCP-100 CONTROL 851 851 CONTROLXFERRED XFERRED TO TOAUX.

AUX.SHUTDOWN SHUTDOWN PANEL PANEL TT FK FK AA ZLP-100 851 ZLP-100 851 AA DI DI SIHD0851 ERF ERF NOTES NOTES1&4 1&4 RHR Heat SIHD0851 FY FY TT AA Exchanger 3857 3857 II PP Bypass Control f(x) f(x) FY Valves (9-Z-42030 FY 3857 3857 AA Rev. 8)

SS NOTE NOTE33 CC 125 125VDC VDC TRAIN TRAINAA FCV FCV 851 851 AA ZSC ZSC ZSO ZSO RR GG 0851 0851 0851 0851 ZLP-001 ZLP-001 ZCP-001 ZCP-001 0%

0% 100%

100%

RR GG ZLP-100 ZLP-100 F.C.

F.C. ZCP-100 ZCP-100 RHR RHRHEAT HEATEXCHANGER EXCHANGER EE ZLP-659 FLOW FLOW CONTROLVALVE CONTROL VALVE EE ZLP-659 NOTE NOTE11 ERF 9-Z-41811 DI ERF 9-Z-41811 DI BYPASS BYPASSSTATUS STATUS SIZD0851 SIZD0851 INOPERATIVE INOPERATIVELOGIC LOGIC LOT 201.09 SLIDE 14

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1559 Last used on an NRC exam: Never RO Sequence Number: 45 Given the following:

A reactor startup is in progress on Unit 1.

SR Channel N-31 indicates 5 x 104 cps.

4 SR Channel N-32 indicates 7 x 10 cps.

IR Channel N-35 indicates 2 x 10-8 amps IR Channel N-36 indicates 2 x 10-10 amps Which of the following describes the NIS response indicated by these readings?

A. All SR and IR Channels are functioning correctly; P-6 permissive is enabled.

B. SR Channel N-32 is reading abnormally high for existing conditions; P-6 permissive is NOT enabled.

C. IR Channel N-35 is reading abnormally high for existing conditions; P-6 permissive is enabled.

D. IR Channel N-36 is reading abnormally low for existing conditions; P-6 permissive is NOT enabled.

Answer: C IR Channel N-35 is reading abnormally high for existing conditions; P-6 permissive is enabled.

Page 89 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1559 K/A Catalog Number: 015 A3.03 Tier: 2 Group/Category: 2 RO Importance: 3.9 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the NIS, including: Verification of proper functioning/operability STP Lesson: LOT 201.16 Objective Number: 57121 STATE the ranges and regions of overlap for the excore NIS.

Reference:

LOT201.16 PowerPoint slide 3 and 16 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because numerically, the pairs of SR and IR indications are relatively close together. The applicant must have an understanding of the meaning and significance of the indications given.

B: INCORRECT: Credible because N-32 is reading higher than N-31. The applicant must understand the significance of the the indication (numerically, the SRs are reading further apart than the IRs). P-6 is credible because 2E-8 could be construed as not greater than 1E-10 (8 is less than 10)

C: CORRECT: IR channels come on scale with SR's slightly less than 1E4 cps. IR reading of 2E-8 amps is indicative of power operation. P6 is enabled with both IR's greater than 1E-10 amps.

D: INCORRECT: Credible because N-36 is indicating less than N-35. N-36 is reading about where it should for the given SR indications. P6 is enabled with both IR's greater than 1E-10 amps. P-6 is credible because 2E-8 could be construed as not greater than 1E-10 (8 is less than 10)

Question Level: H Question Difficulty 3 Justification:

The applicant must apply knowledge of NI overlap and interlocks to the given condition to determine which reading is erroneous and P-6 status.

Page 90 of 150

INTERMEDIATE RANGE INSTRUMENT LOSS DET. VOLTAGE A 100 VOLTSBELOWNORMAL 300 TO1500V DC CIC LOSS COMPENSATINGVOLT.

A <50%OF NORMAL 100 VOLT DC LOG TEST SIGNALS CURRENT AMP 10-11 IDLINGCURRENT PROVIDES AMPLIFIED OUTPUT PROPORTIONAL TOLOGOF INPUT I

A MCB METER SUR NR-45 PLANT COMPUTER

• BISTABLES CANBE BYPASSEDFOR MAINTENANCE OR TESTING.

1/2, 10-10AMPS

• P-6, SOURCE RANGE PERMISSIVE A

1/2, 20%CURRENT EQUIV.

• C-1 ROD STOP LEVEL TRIP IR BLOCK P-10 BYPASS 1/2, 25%CURRENT EQUIV.

• Rx TRIP 10-11 TO10-3AMPS LOT201.16.TP.14 DWG- 11/28/02

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1665 Last used on an NRC exam: 2007 RO Sequence Number: 46 Given the following:

Reactor power is 100%

Control rods are in AUTO Channel II of Pressurizer Pressure is being calibrated and associated bistables have been TRIPPED Channel IV T-hot output from QDPS fails high Which of the following describes the effect of these conditions on the Rod Control System?

A. Control rods drive in due to auctioneered Tave failed high.

B. Control rods drive in due to auctioneered T failed high.

C. Reactor trip breakers open due to two channels of OPT bistables tripped.

D. Reactor trip breakers open due to two channels of OTT bistables tripped.

Answer: D Reactor trip breakers open due to two channels of OTT bistables tripped.

Page 91 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1665 K/A Catalog Number: 012 K3.01 Tier: 2 Group/Category: 1 RO Importance: 3.9 10CFR

Reference:

55.41(b)(6)

Knowledge of the effect that a loss or malfunction of the RPS will have on the following: CRDS STP Lesson: LOT 201.15 Objective Number: 92495 Given a description of plant conditions, PREDICT the indications received in the control room.

Reference:

LOT201.15, Temperature Monitoring Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT - Credible because this would be correct if the PZR pressure work was not on-going.

B: INCORRECT - Credible because RCS temperature does input into rod control (Tave rather than delta-T).

C: INCORRECT - Credible because OPT channel IV will trip (due to the failed T--hot), so the applicant must have knowledge of the other inputs to determine this is not correct (due to PZR pressure calibration).

D: CORRECT - Channel II OTT will be tripped due to the PZR Pressure channel out t of service and the Channel IV OTT will trip when thT-hot fails.

Question Level: H Question Difficulty 3 Justification:

From the given conditions, the applicant must determine the effect on the reactor protection system and the rod control system and with that knowledge determine the effect on the plant.

Page 92 of 150

OT T Formula

• T < To[K1 - K2(T - T) + K3(P - P) - f1( I)]

Taverage Measured delta T by Pressurizer pressure RCS instrumentation Trip stpt only reduced Indicated delta T at when delta I exceeds RATED THERMAL

+8%

POWER 28

OP T Formula

• T < To[K4 - K5(T) - K6(T - T) - f2( I)]

Taverage Set to 0 Indicated delta T at RATED THERMAL POWER Measured delta T by RCS instrumentation 30

+7 Rod Block Rod Block Obj. 4 a,b,d,e,f,h 1

1.5% < stpt 1.5% < stpt 5 & 5 a,b,c,d,f 39

• see COLR

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1675 Last used on an NRC exam: 2007 RO Sequence Number: 47 Given the following:

Unit 1 is in Mode 3 Steam Dumps are in Steam Pressure Mode controlling at 1185 psig.

All RCPs are running All Steam Dump Valves fail open causing the RCS to cool at >100 ºF/hr.

Which of the following correctly describes the MINIMUM operator action/s that would ensure all Steam Dumps are closed and the reason cooldown limits are established?

A. Place EITHER Steam Dump Interlock Selector Switch to OFF/RESET. Excessive cooldown can result in non-ductile failure of the Reactor Vessel.

B. Place BOTH Steam Dump Interlock Selector Switches to OFF/RESET. Excessive cooldown can result in ductile failure of the Reactor Vessel.

C. Place EITHER Steam Dump Interlock Selector Switch to BYPASS INTERLOCK.

Excessive cooldown can result in non-ductile failure of the Reactor Vessel.

D. Place BOTH Steam Dump Interlock Selector Switches to BYPASS INTERLOCK.

Excessive cooldown can result in ductile failure of the Reactor Vessel.

Answer: A Place EITHER Steam Dump Interlock Selector Switch to OFF/RESET.

Excessive cooldown can result in non-ductile failure of the Reactor Vessel.

Page 93 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1675 K/A Catalog Number: 039 K5.05 Tier: 2 Group/Category: 1 RO Importance: 2.7 10CFR

Reference:

55.41(b)(2)

Knowledge of the operational implications of the following concepts as they apply to the MRSS: Bases for RCS cooldown limits STP Lesson: LOT 102.61 Objective Number: N99926 Describe and differentiate between the stresses induced in a reactor vessel wall during heatup and cooldown.

Reference:

LOT 202.09 PowerPoint slides 101 and 103; LOT102.61 handout page 25 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: CORRECT - the Steam Dump Interlock Selector Switches are redundant control devices such that either one being positioned to 'Off/Reset' will remove the control signal from all steam dumps, closing them. Cooldown limits are based on non-ductile failure (brittle).

B: INCORRECT: Switch operation is credible because some operations (i.e. CS manual actuation, blocking SI) requires operation of both train switches, however in this case both are not needed.

Failure mechanism is credible because ductile failure in metals can occr, so knowledge of the difference is required.

C: INCORRECT - Switch operation is credible because 'Bypass Interlock' is another position on the Steam Dump Selector Switches, so knowledge of system operation is required.

D: INCORRECT -Switch operation is credible because 'Bypass Interlock' is another position on the Steam Dump Selector Switches, so knowledge of system operation is required. Failure mechanism is credible because ductile failure in metals can occr, so knowledge of the difference is required.

Question Level: F Question Difficulty 3 Justification:

A knowledge of Steam Dump Controls and material properties is required.

Page 94 of 150

Interlock Selector Switches INTLK SEL (Train A and B)

OFF/RESET All dump valves are ON F BY blocked closed.

IN PA OF TE S After bypassing the RL S ET low-low Tavg interlock OC ES for cooldown, the K /R switch must be returned to the OFF/RESET to reactivate the interlock.

Interlock Selector Switches INTLK SEL (Train A and B)

BYPASS INTERLOCK Spring returns to ON ON F BY position.

IN PA OF TE S Allows bank No. 1 RL S ET dump valves to OC ES continue dumping K /R operations below 563°F.

Brittle Fracture and Vessel Thermal Stress LOT102.61GPST3 Student Handout Page 25 of 34 For a specific irradiated nuclear pressure vessel, the embrittlement depends on the neutron doses, neutron spectrum, irradiation temperature, steel material, and the amount of trace impurities, copper and phosphorous. For trace impurities, shifts in transition temperature by as much as 100°F above the shift otherwise predicted have been observed for increases in copper content from 0.1% to 0.2% at a flux of 2 1019 n/cm2. Thus the change in transition temperature is quite sensitive to impurity level.

REACTOR VESSEL STRESSES To prevent brittle fracture, a vessel must not be stressed too heavily while it is cool. As its temperature increases, it can withstand higher pressures since the metal becomes more ductile. Engineers can calculate the minimum metal temperature required to prevent brittle fracture, given the stress conditions at a certain pressure and heatup rate, as shown in Figure 10-8.

It is possible to construct curves showing the minimum required wall temperature to prevent failure at a given reactor pressure and heatup rate. This would not be of much use to control room operators, however, as they have no direct indication of vessel wall temperatures.

To circumvent this problem, the engineers took into account the fact that the metal temperature would lag the coolant temperature as the coolant temperature rose or fell. Once the temperature differences had been calculated, the minimum reactor temperatures required to safely support the existing pressures were computed. This information is useful to operators because they have direct indication of coolant pressures and temperatures. These parameters are used to establish and control vessel heatup and cooldown rate.

Figure 10-8 Minimum Vessel Temperature vs. Vessel Pressure The reactor vessel and associated piping are normally pressurized from atmospheric pressure up to greater than 2,000 psig. The stress from the pressure is called tensile stress. Both the inner and outer walls of the vessel are subject to tensile stress, with the inner wall experiencing the greatest stress.

PWR / THERMODYNAMICS / CHAPTER 10 © 2011 GENERAL PHYSICS CORPORATION

/ BRITTLE FRACTURE & THERMAL STRESS REV 3 GF@gpworldwide.com www.gpworldwide.com

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1700 Last used on an NRC exam: Never RO Sequence Number: 48 Which of the following correctly describes a condition that could cause the DRPI and the step counter for a particular control rod group to disagree AND result in a Tech Spec entry?

A. Rod Control Logic Cabinet Urgent Alarm.

B. Rod Control Logic Cabinet Non-Urgent Alarm C. DRPI Urgent Alarm D. DRPI Non-Urgent Alarm Answer: C DRPI Urgent Alarm Page 95 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1700 K/A Catalog Number: 014 K5.01 Tier: 2 Group/Category: 2 RO Importance: 2.7 10CFR

Reference:

55.41(b)(5)

Knowledge of the operational implications of the following concepts as they apply to the RPIS: Reasons for differences between RPIS and step counters.

STP Lesson: LOT 201.19 Objective Number: 98055 GIVEN a plant or system condition, PREDICT the operation of the Rod Position Indication System.

Reference:

LOT 201.19 PowerPoint slide 27 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because a logic cabinet urgent failure will prevent rods from moving in manual or auto and result in a Tech Spec entry, however DRPI and step counters will still agree.

B: INCORRECT: Credible because it is a logic cabinet alarm, and without in-depth knowledge of the system could be construed as a possible cause.

C: CORRECT: a DRPI urgent failure will cause a loss of DRPI indication for that rod, thus there will be disagreement with the step counter position for that rod. Would enter the TS for an inoperable DRPI channel.

D: INCORRECT: Credible because it is a DRPI alarm and could cause a difference in the board indication (but not Tech Spec entry).

Question Level: H Question Difficulty 3 Justification:

The applicant must have a knowledge of the urgent and non-urgent alarms involved and apply it to the circumstances described in the question. A basic knowledge of Tech Spec requirements is also needed.

Page 96 of 150

ALARMS AND INDICATIONS DISPLAY UNIT ALARMS URGENT NON-URGENT DEVIATION URGENT ALARM Data A and B Failure Detector Failure Control Unit Failure LOT201.19.TP.027

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1726 Last used on an NRC exam: Never RO Sequence Number: 49 If the fire detectors in the EAB 35 Relay Room do not function, then the __________ system in the Relay Room ________________.

A. Carbon Dioxide (CO2); can still be actuated manually B. Halon; can still be actuated manually C. Halon; will not actuate manually or automatically D. Carbon Dioxide (CO2); will not actuate manually or automatically Answer: B Halon; can still be actuated manually Page 97 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1726 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 will have on the Fire Protection System:

Fire, Smoke, and Heat Detectors.

STP Lesson: LOT 201.29 Objective Number: 91394 LIST the types of fire detectors used and DESCRIBE their basic principle of operation.

Reference:

LOT201.29, lesson on Fire Protection Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because CO2 extinguishers are available in the area, but the relay room uses Halon for fire protection.

B: CORRECT: The relay room uses Halon for fire protection. Manual actuation is always available.

C: INCORRECT: Credible because the use of detectors and auto or manual actions would prevent inadvertent operation of the system.

D: INCORRECT: System type is credible because CO2 extinguishers are available in the area, but the relay room uses Halon for fire protection. Action is credible because the use of detectors and auto or manual actions would prevent inadvertent operation of the system.

Question Level: F Question Difficulty 2 Justification:

The applicant requires a basic knowledge of the fire protection system located in the Relay Room.

Page 98 of 150

LOT201.29.HO.01 Rev. 7 PAGE 12 OF 35 G. Foam-Water Sprinkler Systems (3)

This system is installed to protect the Standby Diesel Generator Fuel Oil Storage Tanks (FOST) located above the engine rooms. Each of the Foam-Water Sprinkler Systems is dedicated to protect only one diesel fuel tank. The system operation is basically the same as described above for the AFOST with the exception that instead of sending foam solution through a nozzle, it passes through a sprinkler designed to mix air with the solution to produce a foam spray.

The water supply to the foam-water system is supplied from the Ring Main by deluge valves in separate valve houses for each diesel generator located adjacent to the north wall of the DGB. The foam concentrate for each of the three systems is stored in a separate 100 gallon capacity tank located next to the respective FOST room.

H. Halon 1301 System Halon systems consist of a stored pressurized gas that, when released, stops the combustion reaction of fire by chemical interaction. Halon is the trade name for halogenated agent bromotrifluoromethane and is much safer for use in fire suppression than other gases such as Carbon Dioxide because it is breathable in the range of 7 to 7.5% concentration for which the system is designed.

The process computer room (including the raised floor),

computer battery room, and relay room on the 35 foot elevation and the Technical Support Center (TSC) on the 70 foot elevation of the EAB are protected by halon systems. The halon system supplying the rooms on the 35 foot elevation is located on the 10 foot elevation of the EAB inside a room adjacent to the corridor on the west end. The TSC halon system is separate and located on the 70 foot elevation.

The Halon System that protects the EAB computer and relay rooms consists of 20 storage bottles for the main bank and 20 in a reserve bank. The pressurized storage bottles and a Halon fire control panel is located in the room on the 10 foot elevation of the EAB previously mentioned. This system uses halon bottles pressurized to 600 psig and is designed to flood the computer or relay room with a concentration of 7 to 7.5%

within 10 seconds of receiving the fire alarm and actuation signal.

LOT201.29.HO.01 Rev. 7 PAGE 13 OF 35 Automatic actuation of the system is initiated by the Halon fire control panel when 1 of 2 ionization smoke detectors in each of two redundant (cross) zones of detection sense a fire in that area. The use of cross zoning prevents an inadvertent halon actuation should a detector fail. When the halon system is actuated, the affected area ventilation dampers are automatically closed, and solenoid valves open on the selected halon cylinders and main supply line to flood the area with halon. A pre-discharge evacuation alarm is sounded in the affected area to warn personnel. 20 bottles are released for a relay room fire and 5 are used for a computer room fire.

Manual fire alarm switches are provided at the entrance to each room to actuate the system. The system can also be actuated by mechanical dump at the cylinders. A selector switch is located in the halon system room to allow selection of the reserve bank after the main bank has been used. A pressure switch is located in the halon system discharge piping to warn operators that a halon system actuation has occurred.

I. Portable Fire Extinguishers Portable Extinguishers are available throughout the plant.

Carbon Dioxide, pressurized water, and dry chemical extinguishers are provided in selected locations based on their anticipated needs.

1.3 General Design Criteria 1.3.1 The Fire Protection System is designed in accordance with 10 CFR 50 and other industrial requirements. The following are those applicable to this lesson:

- Safety-related structures that have exposed steel are protected with spray-applied fire proofing material that has a fire rating of at least three hours.

- Materials of low combustible and/or low fuel contribution, flame spread, and smoke development are used within safety-related structures.

- An automatic sprinkler system is provided for the one case where open cable trays are routed above a suspended ceiling in the Health Physics Office area in the MAB.

- Floor drains which may collect water from a radioactive area are routed to the LWPS.

- Floor drains are designed to prevent fire spread from one drainage area to another.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2167 Last used on an NRC exam: Never RO Sequence Number: 50 You are performing the actions of Addendum 5, Verification of SI Equipment Operation of 0POP05-EO-EO00, Reactor Trip or Safety Injection.

At Step 6, VERIFY Containment Isolation Phase A, on the ESF Status Panel, you note the following on the CONTAINMENT ISOLATION PHASE A/B status monitoring panel:

Train PHASE A ISOL BYP INOP F/ACT red light white light white light A ON OFF ON B ON ON OFF C ON OFF OFF Which of the following correctly describes the status of Phase A Isolation and any actions that may be required?

A. At least one Train A valve is open; manually close the valve(s).

B. At least one Train B valve is open; manually close the valve(s).

C. At least one Train A valve is open; manually actuate Phase A isolation.

D. At least one Train B valve is open; manually actuate Phase A isolation.

Answer: A At least one Train A valve is open; manually close the valve(s).

Page 99 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2167 K/A Catalog Number: 103 A3.01 Tier: 2 Group/Category: 1 RO Importance: 3.9 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the containment system, including: Containment isolation STP Lesson: LOT 504.05 Objective Number: 80483 Given a copy of a subsequent step or from memory an immediate action step from POP05-EO-EO00, STATE/IDENTIFY how the action is performed and the basis for the action to include the action itself, its purpose and result.

Reference:

0POP05-EO-EO00 Rev 22 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT - The red lights tell the operator all trains received an actuation signal. The F/ACT light tells the operator the status of the actuation. If the light is lit, then the actuation is not fully complete, as is the case for train 'A'.

B: INCORRECT - Credible because the BYP INOP light is lit which tells the operator that component conditions could prevent an actuation from occuring, but the fact that the F/ACT light is off for the train indicates the actuation is complete.

C: INCORRECT - Credible because if an actuation has not occurred, then a manual action would be required. In this case, the actuation signal is present, but all components did not actuate so a manual action will not help.

D: INCORRECT - Status is credible because the BYP INOP light is lit which tells the operator that component conditions could prevent an actuation from occuring, but the fact that the F/ACT light is off for the train indicates the actuation is complete. Action is credible because if an actuation has not occurred, then a manual action would be required. In this case, the actuation signal is present, but all components did not actuate so a manual action will not help.

Question Level: H Question Difficulty 3 Justification:

The applicant must first determine the meaning of the light indication given in the stem. Then using their knowledge of sytem design, determine the correct course of action for the condition given.

Page 100 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2175 Last used on an NRC exam: Never RO Sequence Number: 51 Unit 1 is in MODE 4 with a plant heat up in progress.

A Train CCW is out for maintenance.

Train B and C RHR are in service.

RCS Temperature is 345ºF.

Subsequently one of the suction valves closes on Train B RHR.

Which of the following describes (1) the impact of this malfunction to Train B RHR and (2) the actions that should be taken to prevent an inadvertent entry in to MODE 3?

A. (1) Train B RHR trips on low flow.

(2) Use SG PORVs to steam SGs per 0POP03-ZG-0001, Plant Heatup.

B. (1) Train B RHR trips on low flow.

(2) Start Train A RHR per 0POP02-RH-0001, Residual Heat Removal System Operation.

C. (1) Train B RHR miniflow valve automatically opens due to low flow.

(2) Secure Train B RHR and Start Train A RHR per 0POP02-RH-0001, Residual Heat Removal System Operation.

D. (1) Train B RHR miniflow valve automatically opens due to low flow.

(2) Secure Train B RHR and use SG PORVs to steam SGs per 0POP03-ZG-0001, Plant Heatup.

Answer: A (1) Train B RHR trips on low flow.

(2) Use SG PORVs to steam SGs per 0POP03-ZG-0001, Plant Heatup.

Page 101 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2175 K/A Catalog Number: 05 A2.04 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(3)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS and (b) based on those predictions use procedures to correct, control or mitigate those malfunctions: RHR valve malfunction STP Lesson: LOT 201.09 Objective Number: 4245 Given a plant or system condition, predict the operation to the Residual Heat Removal System.

Reference:

0POP02-RH-0001, RHR System Operation, discusses the RHR low flow trip and 0POP03-ZG-0001, Plant Heatup, discusses use of SG PORVs prior to and when transitioning form MODE 4 to MODE 3.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: RHR Pumps will trip on low flow and SGs can be used to control RCS temperature when at the transition point for going from MODE 4 to MODE 5.

B: INCORRECT: Credible since under normal conditions, this could be correct. However it is given that CCW 'A' is OOS, so RHR 'A' would not be placed in service.

C: INCORRECT: Impact is credible because most recirc valves will automatically maintain minimum flow for the pump, however the RHR pump recircs have no auto function. Action is credible since under normal conditions, this could be correct. However it is given that CCW 'A' is OOS, so RHR

'A' would not be placed in service..

D: INCORRECT: Impact is credible because most recirc valves will automatically maintain minimum flow for the pump, however the RHR pump recircs have no auto function.

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator must evaluate the given conditions to determine the effects of the malfunction and the actions to take.

Page 102 of 150

0POP03-ZG-0001 Rev. 58 Page 51 of 136 Plant Heatup Initials 7.0 Mode 3 Heat-Up CAUTION Cold Shutdown (68qF, Xenon-free) RCS boron concentration SHALL be maintained while SI Actuation below P-11(1985 psig) is Blocked. (Reference 2.5.35 & 2.5.43) 7.1 IF Core Exit Thermocouple/Resistance Temperature Detector Cross Calibration is to be performed, THEN REVIEW 0PSP10-RC-0002 Core Exit Thermocouple/Resistance Temperature Detector Cross Calibration Section 5.3 for plant parameters necessary for the TEST. _____

7.2 PLACE a minimum of two CRDM vent fans and one reactor cavity vent supply and exhaust train in operation per 0POP02-HC-0001, Containment HVAC. _____

NOTE WHEN RCS temperature is greater than 340qF, THEN the S/U SGFP 14(24) is the preferred method to feed the Steam Generators.

7.3 WHEN RCS temperature is greater than 340qF, THEN PERFORM the following to feed the SGs.

7.3.1 IF S/U SGFP 14(24) is NOT available AND Feedwater Booster Pump desired for feed, THEN GO TO Step 7.3.3.1. _____

7.3.2 IF S/U SGFP 14(24) is NOT available, THEN CONTINUE AFW operation. _____

7.3.3 IF S/U SGFP 14(24) is available, THEN PERFORM the following:

7.3.3.1 ESTABLISH feedwater flow to the SGs using S/U SGFP 14(24) OR Feedwater Booster Pump per 0POP03-ZG-0003, Secondary Plant Startup. _____

7.3.3.2 SECURE all AFW pumps per 0POP02-AF-0001, Auxiliary Feedwater. _____

7.3.3.3 ENSURE the ESF Standby Readiness Lineup Section 16.0 has been performed per 0POP02-AF-0001, Auxiliary Feedwater. _____

This procedure, when completed, SHALL be retained for the life of the plant.

0POP03-ZG-0001 Rev. 58 Page 52 of 136 Plant Heatup Initials CAUTION The Main Steam lines upstream of the MSIVs may require periodic blowdown for moisture control. This can be accomplished by performing Addendum 5. (CR 03-3694) 7.3.3.4 INITIATE warmup of the Main Steam lines per 0POP03-ZG-0003, Secondary Plant Startup, or 0POP03-ZG-0011, Secondary Plant Cold Startup, per the Unit Operations Manager. _____

NOTE x Either steaming method mentioned in Steps 7.4 and 7.5 may be used.

x IF SG PORVs are required to be controlled in manual to maintain RCS or Secondary Side temperatures, THEN an OAS entry is required to ENSURE compliance with TS 3.3.5.1 and TS 3.7.1.6.

x IF RHR was REMOVED from service and RCS Pressure is between 550 psig and 600 psig in from Step 6.22, THEN Steps 7.7, 7.8 and 7.9 are NA.

x IF the (Alternate) steam dumps method is used in Step 7.4, THEN monitor hotwell levels to prevent overfill.

7.4 IF SGs are being fed with AFW, THEN ESTABLISH RCS cooling by steaming SGs utilizing the (Preferred) SG PORVs or (Alternate) steam dumps. {CP006} _____

7.5 IF SGs are being fed with Main Feedwater, THEN ESTABLISH RCS cooling by steaming SGs utilizing steam dumps or SG PORVs. {CP007} _____

7.6 MAINTAIN SG Narrow Range levels between 55 and 75%. _____

7.7 IF RHR is in service, THEN REDUCE cooling by the RHR system as heat removal is established by means of the SGs. _____

7.8 ENSURE RHR system REMOVED from operation per 0POP02-RH-0001, Residual Heat Removal System Operation. _____

7.9 WHEN RHR has been removed from service, THEN RAISE RCS Pressure to between 550 psig and 600 psig. _____

7.10 ESTABLISH required ECCS lineup per 0POP02-RH-0001, Residual Heat Removal System Operation. _____

This procedure, when completed, SHALL be retained for the life of the plant.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2176 Last used on an NRC exam: Never RO Sequence Number: 52 Unit 2 is in MODE 3 with all 4 Reactor Coolant Pumps (RCP) operating, when a breaker fault causes RCP 2A to trip.

Which of the following describes the response of the LOOP A FLOW indicator?

Indication A. instantly drops to 0% and stabilizes B. instantly drops to 0%, then rises to ~20%

C. drops to 0% over a period of time and stabilizes D. drops to 0% over a period of time, then rises to ~20%

Answer: D drops to 0% over a period of time, then rises to ~20%

Page 103 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2176 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 201.05 Objective Number: 86369 DESCRIBE the effects on the plant due to tripping a Reactor Coolant Pump.

Reference:

LOT201.05 handout page 7 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because pumps without flywheels (most pumps) will exhibit an immediate drop in flow when secured. Also other systems with multiple pumps do not have reverse flow in the idle pumps, so it is not expected to have indicated flow on an idle pump.

B: INCORRECT: Credible because pumps without flywheels (most pumps) will exhibit an immediate drop in flow when secured.

C: INCORRECT: Credible because other systems with multiple pumps do not have reverse flow in the idle pumps, so it is not expected to have indicated flow on an idle pump.

D: CORRECT: The flywheel is designed to maintain flow for a short period of time following a trip of the pump. Once the pump stops, the core DP generaterd by the 3 running pumps will force a small amount of flow backwards through the idle loop and cause the flow indicator to rise again.

Question Level: H Question Difficulty 3 Justification:

The applicant must understand the design features of the flywheel and then must realize that with the other 3 pumps running there will be reverse flow through the idle loop due to the reactor vessel DP.

Page 104 of 150

LOT201.05.HO. Rev. 15 PAGE 7 OF 31 The flywheel stores inertial energy to keep the pump rotating for a short but critical period following station blackout. The flywheel provides approximately one minute of flow coastdown to provide DNB protection during the early stages of station blackout when the decay heat level is high. The flywheel is keyed to the motor shaft above the upper bearing assembly.

In a multi-loop plant, de-energization of one or more reactor coolant pumps while another pump or pumps are running causes a reverse flow through the inactive loops. This reverse flow tends to turn the de-energized pumps backwards. Although no mechanical damage would result from such reverse rotation, if an attempt were made to start a pump in this condition, excessive starting currents would be drawn for an excessive time, resulting in over-heating of the motor.

To prevent this reverse rotation, each pump is equipped with an anti-reverse rotation device. The anti-reverse mechanism consists of five pawls mounted on the outside diameter of the flywheel, a serrated ratchet plate mounted on the motor frame, a spring return for the ratchet plate, and two shock absorbers. After the motor has come to a stop, one pawl will engage the ratchet plate and, as the motor starts to rotate in the opposite direction, the ratchet plate will also rotate slightly until stopped by the shock absorbers. The rotor will remain in this position until the motor is energized again. After the motor has started to rotate, the ratchet plate will be returned to its original position by the spring return. When the motor is started, the pawls will drag over the ratchet plate until the motor reaches approximately 70 rpm. After this time, centrifugal force will keep the pawls in an elevated position.

STATOR The stator core laminations are made from high-silicon electrical sheets coated with alkoplus for insulation. Stacks of laminations periodically separated by air vent spacers are held in place by studs and clamped by steel end plates.

The stator windings are made of insulated copper wire fitted into slots in the core. The ends of the windings, which extend beyond the slots, are braced by an insulated support ring and are separated by non-woven polyester felt packing to withstand the mechanical forces associated with full-voltage starts.

The entire stator core and windings are insulated and are moisture resistant. A solid epoxy resin strengthens the bracing system of the windings. This resin is susceptible to radiation damage at levels of 100 rads/hr. Normal radiation levels are less than 50 rads/hr; therefore, no deterioration is expected.

RCP ELECTRICAL TRIPS Trips for the RCPs are:

Undervoltage - sensed on the 13.8 Kv Aux Bus cub 13 Reverse Phase - sensed on the 13.8 Kv RCP breaker (Aux bus cubicle 11)

Overcurrent - sensed on the 13.8 Kv RCP breaker (Aux bus cubicle 11)

Underfrequency - sensed on the Class 1E 15 Kv RCP Cubicle (Dummy breaker). This signal feeds into SSPS and on a 2 of 4 logic, trips ALL RCPS and also generates a Reactor Trip.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2179 Last used on an NRC exam: Never RO Sequence Number: 53 Given the following:

Pressurizer pressure is 2235 psig PRT PRESS HI annunciator is in PRT TEMP HI annunciator is in PRT pressure on CP-04 is indicating 10 psig and slowly rising PRT temperature on CP-04 is indicating 110 F and slowly rising Which of the following is the likely cause for this condition?

A. Pressurizer PORV seat leakage.

B. Reactor Makeup Water to the PRT (FV-3650) is open.

C. Letdown Stop valve leakoff flow high.

D. RCP #1 seal leakoff flow high.

Answer: A Pressurizer PORV seat leakage Page 105 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2179 K/A Catalog Number: 007 A4.10 Tier: 2 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(3)

Ability to manually operate and/or monitor in the control room: Recognition of leaking PORV/code safety STP Lesson: LOT 201.04 Objective Number: 80883 DESCRIBE the indications available to determine that a Pressurizer power operated relief valve is leaking.

Reference:

LOT201.04 PZR, PRT and RCDT Power Point Presentation Slide #58 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: The discharge of the PZR PORV is directed to the PRT this would cause temperature and pressure to rise in the PRT.

B: INCORRECT: Credible because this flowpath would cause PRT pressure to rise, but temperature should not.

C: INCORRECT: Credible because these indications would occur in the RCDT (another collection tank located in containment). The applicant must have knowledge of the different system flowpaths.

D: INCORRECT: Credible because this source would give these indications in the PRT following an SI actuation, but not during normal operations. There is no indication in the stem that an SI has occurred.

Question Level: H Question Difficulty 3 Justification:

The candidate must recognize from indications that a high energy source is going into PRT and recall which influents are directed to PRT and which are high enough energy to give the indications provided.

Page 106 of 150

PRT FEED AND BLEED LOT201.04.TP.58

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2186 Last used on an NRC exam: Never RO Sequence Number: 54 Given the following:

Unit 1 is at 100% power A Train ECW/CCW systems are running B Train ECW is running B ECW/CCW is the Train selected for STBY.

A CCW HX outlet flow indicates 12,700 gpm Subsequently, the following occurs:

CCW HX 1A OUTL FLOW HI/LO alarms CCW HX 1A OUTL PRESS LO alarms CCW HX 1A OUTL PRESS PI-4513 indicates 74 psig Select the malfunction that could have caused the given indications and the automatic action that should occur.

MALFUNCTION AUTOMATIC ACTION A. A loss of power to RHR 1A CCW Outlet The 1B CCW Pump will auto Valve caused the valve to fail open. start after a time delay.

B. A loss of power to RHR 1A CCW Outlet The 1B CCW Pump will start Valve caused the valve to fail open. immediately.

C. A loss of power to CCW HX 1A Outlet TCV The 1B CCW Pump will auto caused the valve to fail open. start after a time delay.

D. A loss of power to CCW HX 1A Outlet TCV The 1B CCW Pump will start caused the valve to fail open. immediately.

Answer: A A loss of power to RHR 1A CCW Outlet Valve caused the valve to fail open The 1B CCW Pump will auto start after a time delay.

Page 107 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2186 K/A Catalog Number: 008 A3.03 Tier: 2 Group/Category: 1 RO Importance: 3.0 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the CCWS including:

All flow rate indications and the ability to evaluate the performance of this closed cycle cooling system.

STP Lesson: LOT 201.12 Objective Number: 57126 Describe the operation of the Component Cooling Water System and its major components. Include automatic actions, interlocks and trips.

Reference:

LOT 201.12 Power Point Presentation Slide #8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: The RHR HX CCW Oulet Valve fails open on loss of power or loss of air. The train flow will increase by approximately 5000 gpm, Stby CCW pump auto starts on low common header pressure of 76 psig after a time delay.

B: INCORRECT: Pump start is credible because that is original plant design and the way most systems function.

C: INCORRECT: Valve failure is credible because open failure of a safety related TCV would be reasonable and desired, however these valves are MOVs and will fail as is if power is lost to the valve.

D: INCORRECT: Valve failure is credible because open failure of a safety related TCV would be reasonable and desired, however these valves are MOVs and will fail as is if power is lost to the valve. Pump start is credible because that is original plant design and the way most systems function.

Question Level: H Question Difficulty 3 Justification:

The candidate must recognize that the RHR HX outlet valve has failed open given lost valve indication, high flow alarm and low system press. And that the stby train will auto start after a time delay.

Page 108 of 150

RHR HX 1A CCW Outlet Valve is an air operated Temperature Control Valve LOT 19 RO Question

1. 54 CCW HX 1A Outlet TCV is a Motor Operated Valve

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2189 Last used on an NRC exam: Never RO Sequence Number: 55 Given the following:

Unit 1 is at 100% Power.

Train A and C ECW/CCW pumps are running to support Unit operations Train B ECW/CCW is in standby.

Subsequently the following occur simultaneously:

A large Steam Line Break in Containment MCC E1B3 loses power.

Which of the following describes an impact due to these events and the actions that will mitigate the consequences?

IMPACT ACTIONS Place CS Pump 1B in PTL to prevent running the CS Pump 1B discharge pump at shutoff head.

A. valve is CLOSED and will not OPEN. Start two additional RCFCs to provide additional containment cooling.

Place CS Pump 1B in PTL to prevent runout CS Pump 1B discharge conditions.

B. valve is OPEN and will not CLOSE. Start two additional RCFCs to provide additional containment cooling.

ECW Pump 1B discharge Place ECW Pump 1B in PTL to prevent running the C. valve is CLOSED and pump at shutoff head.

will not OPEN. Place EDG #12 in PTL to prevent overheating.

ECW Pump 1B discharge Place ECW Pump 1B in PTL to prevent runout D. valve is OPEN and will conditions.

not CLOSE. Place EDG #12 in PTL to prevent overheating.

Answer: C ECW Pump 1B discharge valve is CLOSED and will not OPEN; Place ECW Pump 1B in PTL to prevent running the pump at shutoff head.;

Place EDG #12 in PTL to prevent overheating.

Page 109 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2189 K/A Catalog Number: 076 A2.01 Tier: 2 Group/Category: 1 RO Importance: 3.5 10CFR

Reference:

55.41(b)(7)

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Loss of SWS STP Lesson: LOT 201.13 Objective Number: 91201 GIVEN a plant or system condition, PREDICT the operation of the Essential Cooling Water System.

Reference:

0POP05-EO-EO00 Addendum 5 step 7 RNO Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Impact is credible if the stated MCC supplied this valve (powered from MCC E1B2).

Action is credible because placing CS Pump 1B in PTL would be the correct action if the discharge valve did not open. Since both CS and RCFCs are designed to provide containment cooling, starting RCFCs to replace a CS pump would be appropriate, however in the given situation all RCFCs would already be running.

B: INCORRECT: Impact is credible because some pumps (i.e. AFW Pumps) start with discharge valves open and they must automatically throttle to prevent pump runout, however CS Pump discharge valves are not normally open while in a standby condition. Action is credible because placing CS Pump 1B in PTL would be the correct action if the pump was running in a runout condition. Since both CS and RCFCs are designed to provide containment cooling, starting RCFCs to replace a CS pump would be appropriate, however in the given situation all RCFCs would already be running.

C: CORRECT: MCC E1B3 provides power exclusively to Train B ECW components. The valve would have been closed when it lost power and would not open on pump start. Pump damage could occur while running with discharge closed (no recirc and no cooling). Stopping the DG is also a required action in this case since it is not being provided any cooling water flow.

D: INCORRECT: Impact is credible because some pumps (i.e. AFW Pumps) start with discharge valves open and they must automatically throttle to prevent pump runout, however ECW Pump discharge valves are not normally open while in a standby condition.

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator has to evaluate the given condition to determine the impact and the procedure steps to mitigate the consequences.

Page 110 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2190 Last used on an NRC exam: Never RO Sequence Number: 56 Unit 1 has experienced a Reactor Trip, Safety Injection and LOOP on Class 1E 4.16 KV Bus Train A. Diesel Generator #11 is supplying power to Class 1E 4.16 KV Bus Train A.

Subsequently AFWP #14 tripped on overspeed. Mechanical Maintenance corrected a mechanical issue with the trip linkage and says that AFWP #14 overspeed trip can now be reset.

Which of the following describes and explains a precaution the Operations Crew should take prior to resetting the overspeed trip for AFWP #14?

A. Reset ESF Load Sequencer Train A and SG LO-LO Actuations to ensure AF-MOV-0514, TURB TRIP/THROT remains closed during and after the reset.

B. Reset Safety Injection and SG LO- LO Actuations to ensure AF-MOV-0514, TURB TRIP/THROT remains closed during and after the reset.

C. Reset Safety Injection and SG LO- LO Actuations to ensure MS-MOV-0143, MN STM ISOL remains closed during and after the reset.

D. Reset ESF Load Sequencer Train A and SG LO-LO Actuations to ensure MS-MOV-0143, MN STM ISOL remains closed during and after the reset.

Answer: B Reset Safety Injection Actuations and SG LO-LO Actuations to ensure AF-MOV-0514, TURB TRIP/THROT remains closed during and after the reset.

Page 111 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2190 K/A Catalog Number: 013 G2.1.32 Tier: 2 Group/Category: 1 RO Importance: 3.8 10CFR

Reference:

55.41(b)(10)

Engineered Safety Features Actuation System: Ability to explain and apply system limits and precautions.

STP Lesson: LOT 202.28 Objective Number: 43847 DISCUSS the following elements associated with the AFW turbine driven pump: B. How to reset the trip and throttle valve.

Reference:

LOT 202.28 and 0POP02-AF-0002, Resetting Auxiliary Feedwater Pump 14(24) Mechanical Overspeed Trip Device Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because the ESF Load Sequencers provide a start signal to all the other AFW Pumps, but not AFWP #14.

B: CORRECT: AFWP #14 gets a start signal from SI and LO-LO SG level. These two actuations (specifically Train A) must be reset prior to resetting the overspeed trip device to ensure the trip throttle valve, MOV-0514, does not open while resetting.

C: INCORRECT: Credible because MS-MOV-0143, MS Isolation to AFWP #14 does get an open signal upon an AFW actuation, but this valve would already be open (and is normally open).

D: INCORRECT: Reset is credible because the ESF Load Sequencers provide a start signal to all the other AFW Pumps, but not AFWP #14. Explanation is credible because MS-MOV-0143, MS Isolation to AFWP #14 does get an open signal upon an AFW actuation, but this valve would already be open (and is normally open).

Question Level: H Question Difficulty 3 Justification:

The Operator must evaluate the given condition and then apply knowledge of the precautions associated with AFWP #14.

Page 112 of 150

0POP02-AF-0002 Rev. 7 Page 4 of 13 Resetting Auxiliary Feedwater Pump 14(24) Mechanical Overspeed Trip Device 4.0 Notes and Precautions 4.1 WHEN stopping turbine driven AFW Pump 14(24), THEN AFW Pump 14(24) SHOULD be tripped to avoid any possibility of operation less than 1900 rpm. (Reference 2.5) 4.2 Components on and around AFW Pump 14(24) MAY be HOT. Caution SHOULD be used when working on or around AFW Pump 14(24).

4.3 Resetting TRAIN A SAFETY INJECTION and TRAIN A SG LO-LO LEVEL actuation signals prior to resetting the overspeed trip device should keep MOV-0514 from automatically opening as soon as the operator resets the overspeed trip device.

5.0 Resetting Auxiliary Feedwater Pump 14(24) Mechanical Overspeed Trip Device NOTE All component locations are IVC 10 AFW Pump 14(24) Pump Cubicle unless otherwise noted.

5.1 VERIFY the AFW Pump 14(24) TURB TRIP/THROT MOV-0514 is CLOSED (CP 006). _____

5.2 VERIFY the AFW Pump 14(24) MOV-0514 Turbine Mechanical Overspeed Trip Linkage is TRIPPED by observing the Latch Up Lever NOT engaged with the Latch Trip Hook.

(See Addendum 1, Picture 1 for component location and identification) _____

5.3 ENSURE 1(2)-AF-ZSC-7537A, TERRY TURBINE MECHANICAL OVERSPEED TRIP SWITCH Limit Switch roller is on the MOV-0514 side of the Head Lever (Above the Turbine Pump Casing).

(See Addendum 1 for component location and identification) _____

NOTE Spring tension pulls the linkage away from the Trip and Throttle Valve.

5.4 RESET the Mechanical Overspeed Trip Linkage by PUSHING/PULLING the Linkage Connecting Rod towards 1(2)-MS-0514 MAIN STM TO TERRY TURBINE THROTTLE MOV. _____

5.5 ENSURE that the Trip Hook and Latch Up Lever are fully engaged.

(See Addendum 1, Pictures 2 and 3 for a proper comparison) _____

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2191 Last used on an NRC exam: Never RO Sequence Number: 57 Unit 1 Control Room is notified by workers excavating in the yard east of the unit that a buried pipe has been damaged.

A Control Room operator observes IA press PI- 8563 lowering below 95 psig.

Annunciator IAS HDR PRESS LO 8M03-D3 is received.

The IA header press stops lowering and begins to rise.

Which of the following correctly identifies what has occurred?

At 90 psig A. IA Yard Isolation Valve 1-IA-PV-8568 auto closed to isolate an Instrument Air pipe leak.

B. IA Dryer Emergency Bypass Valve opened to provide sufficient air volume to overcome the leak.

C. SA Isolation Valve 1-IA-PV-9785 auto closed to isolate a Service Air pipe leak.

D. #14 IA Compressor auto started and is supplying sufficient volume to overcome the leak.

Answer: A IA Yard Isolation Valve 1-IA-PV-8568 auto closed to isolate an Instrument Air pipe leak.

Page 113 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2191 K/A Catalog Number: 078 A3.01 Tier: 2 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(4)

Ability to monitor automatic operation of the IAS, including: Air pressure STP Lesson: LOT 202.26 Objective Number: 92995 Given a scenario in which Instrument Air pressure is decreasing, PREDICT Instrument and Service Air system component automatic actions that will occur as pressure decreases.

Reference:

0POP04-IA-0001, Loss of Instrument Air Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: 90 psig is the set point for IA HDR PRESS LO alarm and closes IA yard Isol valve B: INCORRECT: Credible because the IA Emergency Bypass receives an open signal on low pressure (80 psig). IA pressure has not reached the 80 psig setpoint yet.

C: INCORRECT: Credible because the Service Air Isolation Valve receives a closed signal on low pressure (100 psig). If this would have isolated the leak then IA pressure would have started to go back up before it reached 95 psig.

D: INCORRECT: Credible because the 14 IA Compressor receives an auto start signal on low pressure (113 psig). If this would have helped the leak then IA pressure would have started to go back up before it reached 95 psig.

Question Level: F Question Difficulty 3 Justification:

The candidate must correlate reported field information and check IA header, recall alarm setpoint and which component will actuate to isolate leak.

Page 114 of 150

0POP04-IA-0001 Loss Of Instrument Air Rev. 16 Page 2 of 152 PURPOSE This procedure provides the necessary operator actions for responding to a significant degradation or loss of Instrument Air (IA) capacity.

Instrument Air Pressure Automatic Actuation (Decreasing)

IA Compressor 11(21) Starts/Loads in Local 122 psig Control IA Compressor 12(22) Starts/Loads in Local 119 psig Control IA Compressor 13(23) Starts/Loads in Local 116 psig Control IA Compressor 14(24) (air cooled and BOP DG 113 psig powered) Starts/Loads Service Air Isolation Valve N1(2)IA-PV-9785 100 psig Closes Instrument Air to Yard Valve 90 psig N1(2)IA-PV-8568 Closes Instrument Air Dryer Bypass 80 psig N1(2)IA-PV-9983 Opens SYMPTOMS OR ENTRY CONDITIONS

1. The following Control Room annunciator alarms:

x SAS ISOL VLV CLOSE Lampbox 08M3, Window F-3 x SAS HDR PRESS LO Lampbox 08M3, Window E-3 x IAS HDR PRESS LO Lampbox 08M3, Window D-3

2. All operable IA compressors running continuously.

3. No IA compressors running.

4. Various air operated valves observed to be drifting to failure positions.

This Procedure is Applicable in All Modes

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2192 Last used on an NRC exam: Never RO Sequence Number: 58 Given the Following:

Unit 2 Control Room is performing 0POP05-EO-EC12 LOCA OUTSIDE CONTAINMENT.

FHB -4 ft el Area Radiation Monitor is in High Alarm on RM-11 A Train SI/CS Hi/Hi sump alarm on QDPS Which of the following describes the impact of this condition and the actions the crew should take to mitigate the consequences?

IMPACT ACTIONS RWST inventory will not be Place A Train SI/CS pumps in A. available for recirculation phase. PTL.

Sump tanks in the FHB will have to Align the A Train SI/CS sump to B. be transferred to the RCB for discharge to the Containment recirculation phase. Emergency Sump.

Align the A Train SI/CS sump to RWST inventory will not be C. discharge to the Containment available for recirculation phase.

Emergency Sump.

Sump tanks in the FHB will have to Place A Train SI/CS pumps D. be transferred to the RCB for in PTL.

recirculation phase.

Answer: A RWST inventory will not be available for recirculation phase. Place 'A' Train SI/CS pumps in PTL.

Page 115 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2192 K/A Catalog Number: 006 A2.11 Tier: 2 Group/Category: 1 RO Importance: 4.0 10CFR

Reference:

55.41(b)(10)

Emergency Core Cooling System:

Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture of ECCS Header STP Lesson: LOT 201.10 Objective Number: 17259 DESCRIBE the flowpath for the ECCS to include major components and valves.

Reference:

LOT 201.10 Lesson on ECCS and LOT 504.46 Lesson on 0POP05-EO-EC12 LOCA OUTSIDE CONTAINMENT step 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: RWST inventory leaked into FHB would not be available for injection into RCS. The FHB radiation monitor alarm and SI/CS pump room sump are indications used to identify a potential leak and location, then the train is secured to try and isolate the leak.

B: INCORRECT: Credible because both given options would allow recovery of lost inventory, however plant design does not support.

C: INCORRECT: The action is credible because it provides a possible method of recovering lost inventory but is not supported by plant design.

D: INCORRECT: Impact is credible because it provides a reasonable action that could recover lost RWST inventory, but is not supported by plant design.

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator must be able to predict the impact of the malfunction and then identify the correct mitigation based on the given information.

Page 116 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2193 Last used on an NRC exam: Never RO Sequence Number: 59 Unit 1 has experienced a Reactor Trip from 100% power. The AFW System has actuated but no Operator actions have been performed yet.

Which of the following describes the effect on the AFW System if a QDPS #2 APC were to lose power?

A. All AFW Pump Flow Regulating Valves would fail in the AS IS POSITION.

B. Only the AFW Pump Flow Regulating Valve associated with the failed #2 APC would fail in the AS IS POSITION.

C. All AFW Pump Outside Containment Isolation Valves would fail in the AS IS POSITION.

D. Only the AFW Pump Outside Containment Isolation Valve associated with the failed #2 APC would fail in the AS IS POSITION.

Answer: B Only the AFW Pump Flow Regulating Valve associated with the failed #2 APC would fail in the AS IS POSITION.

Page 117 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2193 K/A Catalog Number: 016 K3.06 Tier: 2 Group/Category: 2 RO Importance: 3.5 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: AFW System STP Lesson: LOT 202.44 Objective Number: 7667 Given a change in plant or system condition EXPLAIN the operation and indications of the QDPS System.

Reference:

LOT 202.44 QDPS and LOT 202.28 AFW System Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because each reg valve receives a control signal, but not all from the same QDPS APC.

B: CORRECT: QDPS #2 APCs control their associated train of AFW Reg Valve.

C: INCORRECT: Credible because the AFW Outside Containment Isolation Valves automatically receive an open signal upon an actuation, but it comes from a source other than the QDPS APC (SSPS).

D: INCORRECT: Credible because the AFW Outside Containment Isolation Valves automatically receive an open signal upon an actuation, but it comes from a source other than the QDPS APC (SSPS) and the APCs are train related.

Question Level: F Question Difficulty 3 Justification:

The Reactor Operator must have fundemental knowledge of two systems; QDPS and AFW.

Page 118 of 150

The Upper Section of the # 2 APCs provides the Valve Control

• Automatic and Manual control of the SG PORVs.

• Automatic control of the AFW throttle valves.

• Manual control of the Reactor Head Vent Valves.

LOT202.44 slide 44

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2194 Last used on an NRC exam: Never RO Sequence Number: 60 Following a Loss of Coolant Accident (LOCA), the pH of the fluid used by the Containment Spray pumps is controlled to reduce corrosion and maintain iodine in solution.

Which of the following describes how the pH of the Containment Spray fluid is controlled?

A. Minimum boric acid concentration requirements for the Refueling Water Storage Tank (RWST) ensure the proper pH is maintained in the emergency sumps.

B. Sodium Hydroxide (NaOH) from the Spray Additive Tanks mixes with the Containment Spray Pump discharge.

C. Powdered trisodium phosphate stored in six baskets located on the -11 foot elevation of the RCB dissolves into the fluid which flows into the emergency sumps.

D. Lithium Hydroxide (LiOH) is injected into the emergency sumps.

Answer: C Powdered trisodium phosphate stored in six baskets located on the -11 foot elevation of the RCB dissolves into the fluid which flows into the emergency sumps.

Page 119 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2194 K/A Catalog Number: 026 K4.02 Tier: 2 Group/Category: 1 RO Importance: 3.1 10CFR

Reference:

55.41(b)(7)

Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following: Neutralized boric acid to reduce corrosion and remove inorganic fission product iodine from steam (NAOH) in containment spray STP Lesson: LOT 201.11 Objective Number: 29767 STATE the name and the function of the chemical used in the Recirculation Fluid pH Control System

Reference:

LOT201.11 handout page 7 & 8 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - Credible because boric acid will change pH (lower it), however the additive used raises pH.

B: INCORRECT - Credible because spray additive tanks were part of original plant design (and are still physically in the plant), but system was abandoned in place and replaced by the trisodium phosphate baskets.

C: CORRECT - baskets of trisodium phosphate dissolve into the fluid on the containment floor to raise the pH of the fluid.

D: INCORRECT - Credible because LiOH is used during normal operation to control the pH of the RCS.

Question Level: F Question Difficulty 3 Justification:

The applicant must have knowledge of the design and function of the recirculation fluid pH control system.

Page 120 of 150

LOT201.11.HO.01 Rev. 14 PAGE 7 OF 13 Automatic Actions: None Immediate Actions: None Probable Causes: Electrical Fault Breaker opened locally Setpoint: Breaker open and no bus strip signal and control switch not in STOP or PTL

2. CS Pump Discharge Flow Lo Automatic Actions: None Immediate Actions: None Probable Causes: Improper valve lineup Loss of suction Instrument failure Setpoint: 500 gpm

3. ESF bypass/INOP

a. Pull-To-Lock

b. Breaker not racked in

c. Loss of control power

4. Fail To Actuate - CS Actuation Signal and Pump not running RECIRCULATION FLUID PH CONTROL SUB-SYSTEM Six Stainless Steel Baskets (approximately 4' X 8') are filled with powdered Tri-Sodium Phosphate (TSP). When dissolved by the break flow of a LOCA or major Main Steam Line Break (MSLB) in addition to Containment Spray Flow and SI Flow, the PH of the fluid collected in the three (3) Containment Emergency Sumps is increased to 7.0 -

9.5.

LOT201.11.HO.01 Rev. 14 PAGE 8 OF 13 The Boric Acid in the RWST and RCS can cause the PH of the Fluid in the Sumps to be as low as 4.5. Tech Specs requires that between 11,500 lbs to 15,100 lbs of TSP be available for mixing with the SI Injection Fluid and Break Fluid. This guarantees that a Fluid PH of 7.0 to 9.5 is achieved in the three (3) Containment Emergency Sumps.

Controlling PH 7.0 ensures that entrapped iodine remains in solution. Maintaining PH 9.5 minimizes Chloride Induced Stress Corrosion Cracking of Austenitic Stainless Steel.

The six (6) TSP baskets are located at the following azimuths on the 11' elevation in the RCB:

Inside the Biological Shield - at 45, 90and 330 Outside the Biological Shield - at 140, 227 and 270 (NOTE: 0 is at Plant East.)

To ensure that each basket is filled to the proper level, an "Indicator" mark is checked during each refueling outage. This satisfies the Tech.

Spec. Surveillance.

SPRAY HEADERS & NOZZLES The CS Pump Discharge Headers penetrate Containment and join a common Spray Header climbing up the Containment Interior Wall to the Spray Rings. The Common Spray Header supplies 4 concentric rings and associated nozzles. The nozzles are drilled with 3/8 diameter orifices allowing 1/3 particle passage. The nozzle locations and orientation angles ensure a minimum of 90 percent Containment coverage. The Spray Ring is located as high as possible in the RCB without incurring Spray Pattern Interference.

Ring 1 is a 4 diameter ring with a 13 radius, 225 elevation and 21 nozzles.

Ring 2 is a 6 diameter ring with a 26 radius, 221 elevation and 49 nozzles.

Ring 3 is a 6 diameter ring with a 45 radius, 210 elevation and 59 nozzles.

Ring 4 is an 8 diameter ring with a 64 radius, 188 elevation and 119 nozzles.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2195 Last used on an NRC exam: Never RO Sequence Number: 61 During the response to a Unit 2 Reactor Trip, all Auxiliary Feedwater was lost. The crew is performing 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink, Addendum #1, Establishing Main Feedwater Flow.

The following conditions exist:

RCS Feed and Bleed has been established.

RCS Wide Range TH is 600ºF and rising.

All SG Wide Range levels are 11% and lowering.

The crew is ready to feed Main Feedwater to the Steam Generators.

Which of the following describes the IMPACT of feeding Main Feedwater to dry Steam Generators and the REQUIRED ACTION from 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink, which should minimize the IMPACT?

IMPACT REQUIRED ACTION A Steam Generator integrity will be If CETs are rising establish maximum impacted due to thermal shock. flow rate to ALL SGs.

B Steam Generator integrity will be If CETs are rising establish maximum impacted due to high pressure. flow rate to ONLY ONE SG.

C Steam Generator integrity will be If CETs are rising establish maximum impacted due to thermal shock. flow rate to ONLY ONE SG.

D Steam Generator integrity will be If CETs are rising establish maximum impacted due to high pressure. flow rate to ALL SGs.

Answer: C Steam Generator integrity will be impacted due to thermal shock. If CETs are rising establish maximum flow rate to ONLY ONE SG.

Page 121 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2195 K/A Catalog Number: 059 A2.04 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(5)

Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a Dry Steam Generator 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 on 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink.

See CIP Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because the given RCS temperature is high for post trip conditions (and rising), so the first instinct is to lower temperature as quickly as possible.

B: INCORRECT: Credible because adding water to a hot vessel will quickly produce steam and raise pressure, however the SG safeties are still available and designed to protect the SG from a high pressure, loss of integrity event.

C: CORRECT: The impact of feeding a dry SG is thermal shock. Because of this feedwater is established to only one SG at a time. The heat removal capability from one SG is greater than decay heat.

D: INCORRECT: Impact is credible because adding water to a hot vessel will quickly produce steam and raise pressure, however the SG safeties are still available and designed to protect the SG from a high pressure, loss of integrity event. Action is credible because the given RCS temperature is high for post trip conditions (and rising), so the first instinct is to lower temperature as quickly as possible.

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator needs to have fundamental knowledge of the basis for cautions in emergency procedures and, when given information, determine the required actions from the procedure to take.

Page 122 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2196 Last used on an NRC exam: Never RO Sequence Number: 62 Given the following:

A Unit 1 reactor trip has occurred.

An AFW actuation occurred as expected.

The pump coupling on AFW Pump #13 broke causing the pump shaft to separate from the motor shaft.

Based on these conditions, which of the following is true?

A. AFWP 13 TRIP alarm will annunciate due to motor overspeed after the coupling breaks.

B. AFWP 13 TRIP alarm will annunciate when pump flow drops to <90 gpm after the coupling breaks.

C. AFWP 13 DISCH PRESS LO alarm will annunciate, the motor should be stopped by placing the hand switch in the STOP position.

D. AFWP 13 DISCH PRESS LO alarm will annunciate, the motor should be stopped by placing the hand switch in the PTL position.

Answer: D AFWP 13 DISCH PRESS LO alarm will annunciate, the motor should be stopped by placing the hand switch in the PTL position.

Page 123 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2196 K/A Catalog Number: 061 K6.02 Tier: 2 Group/Category: 1 RO Importance: 2.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Pumps STP Lesson: LOT 202.28 Objective Number: 43805 DESCRIBE the AFW system controls and instrumentation in the MCR.

Reference:

POP09-AN-06M4 pages 15 and 28 & Logic 9Z40131 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT - Credible because turbine driven and engine driven pumps have overspeed trips and motor speed will temporarily rise when the coupling breaks.

B: INCORRECT - Credible because 90 gpm is the point where the recirc valve begins to open and low flow is often indicative of the need for a pump trip.

C: INCORRECT - Credible because normally, placing a switch in "stop" will secure the pump, but since an AFW actuation is present (given) the switch must be taken to PTL.

D: CORRECT - A low discharge pressure condition will occur when the coupling breaks and the pump shaft stops. The motor can only be stopped in this condition by placing the handswitch in PTL due to the AFW actuation signal present.

Question Level: F Question Difficulty 3 Justification:

The applicant must have a fundamental knowledge of the effects of a shaft shear on pump operation and how it relates to the indications available for the AFW pumps.

Page 124 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2197 Last used on an NRC exam: Never RO Sequence Number: 63 A lockout has occurred on 4.16 KV bus E2A. The Unit Supervisor has directed you to determine the E2A11 battery discharge current.

Which of the following correctly describes how you should obtain this information for the Unit Supervisor?

This information can A. be obtained from CP-003 using the BATT CUR indicator.

B. be obtained from CP-010 using the BATT CUR indicator.

C. only be obtained from one of the QDPS plasma displays.

D. only be obtained locally. A Plant Operator must be dispatched.

Answer: A be obtained from CP-003 using the BATT CUR indicator.

Page 125 of 150

Print Date 8/7/2013 STP LOT-19 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:

LOT201.37 PowerPoint slide 55 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New 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. 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.

Tha applicant must be familiar with what indication is on the control panels to correctly respond.

Question Level: F Question Difficulty 3 Justification:

The applicant must have knowledge of the indications available in the control room for the batteries.

Page 126 of 150

Control Room Alarms/Indications Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2198 Last used on an NRC exam: Never RO Sequence Number: 64 Given the following:

Unit 1 is in Mode 3.

All Shutdown Rods are fully withdrawn preparing for a Reactor Startup.

Pressurizer Backup Heaters D and E are energized.

Subsequently:

D RCP Trips Which of the following describes the correct INITIAL response of Pressurizer Pressure Control and Pressurizer Temperature?

Pressurizer Pressure Master Pressurizer Controller Output Temperature A. Rise Rise B. Rise Lower C. Lower Lower D. Lower Rise Answer: A Rise, Rise Page 127 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2198 K/A Catalog Number: 004 A3.15 Tier: 2 Group/Category: 1 RO Importance: 3.5 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the CVCS, including:

PZR pressure and temperature STP Lesson: LOT 201.14 Objective Number: 92779 GIVEN plant conditions, DETERMINE their effects on the Pressurizer pressure and level control system.

Reference:

LOT 201.02, rev 10 & LOT 201.14 rev 14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: PZR spray line connects to the 'D' RCS loop. With the 'D' RCP tripped, PZR spray flow wil lower causing PZR pressure to rise. Master Controller output will rise which will open spray valves to replace lost spray flow. PZR temperature will rise with pressurizer pressure.

B: INCORRECT: This distractor is credible because with a RCP trip, there is a loss of heat input which could be thought to lower PZR pressure and temperature and it shows lack of knowledge of the PZR Pressure Control system response to a drop in pressure. If PZR pressure and temperature go down then controller output will lower NOT rise. The loss of spray flow will have more affect on the PZR Pressure control than the loss of RCP heat input.

C: INCORRECT: This distrator is credible because the master controller output will change and the applicant must understand system operation to determine how it will change. With a RCP trip there is a loss of heat input which could be thought to lower PZR pressure and temperature due to the outsurge, but the loss of spray flow will have more affect on the PZR Pressure control than the loss of heat input to the RCS.

D: INCORRECT: This distrator is credible because the master controller output will change and the applicant must understand system operation to determine how it will change. If PZR pressure and temperature go up then controller output will rise not lower.

Question Level: H Question Difficulty 3 Justification:

The applicant must analyze the given condition and recall that pressurizer spray is on the D RCS loop and there would be lower flow to the spray nozzles and then analyze the effect on pressurizer pressure control system in auto.

Page 128 of 150

34 Reactor Reactor Coolant Coolant System System LOT201.02. page.19

Simplified Simplified Drawing Drawing Prressurizer Prressurizer LOT201.02. page.29

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2199 Last used on an NRC exam: Never RO Sequence Number: 65 Unit 1 is at 75% power. The crew is currently raising power at 10%/Hour.

The Primary Reactor Operator over the last 10 minutes has noticed RCP Motor Upper and Lower Radial Bearing Temperatures trending up. The bearing temperatures were all at 140ºF and are now stable at the following temperatures:

Motor Upper Radial Motor Lower Radial Bearing Temperature Bearing Temperature RCP 1A 200ºF 200ºF RCP 1B 200ºF 190ºF RCP 1C 190ºF 190ºF RCP 1D 190ºF 180ºF At a minimum, which of the following should the Unit Supervisor have the Reactor Operator perform in accordance with 0POP04-RC-0002, Reactor Coolant Pump Off Normal?

Trip the Reactor, Ensure Main Turbine tripped and stop A. RCP 1A only B. RCP 1A and 1B only C. RCP 1A, 1B and 1C only D. RCP 1A, 1B, 1C and 1D Answer: B RCP 1A and 1B only Page 129 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2199 K/A Catalog Number: 003 A1.02 Tier: 2 Group/Category: 1 RO Importance: 2.9 10CFR

Reference:

55.41(b)(5)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: RCP pump and motor 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 Lesson on RCPs and LOT 501.01 Lesson on 0POP04-RC-0002, Reactor Coolant Pump Off Normal, CIP Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because all given temperatures are above the normal operating range (~140 degrees), so the applicant must have knowledge of the limit to correctly respond.

B: CORRECT: With the given information RCPs 1A and 1B will have to be tripped. If either RCP Motor Upper OR Lower Radial Bearing temperature is above 195 degrees F then the RCP is required to be tripped.

C: INCORRECT: Credible because all given temperatures are above the normal operating range (~140 degrees), so the applicant must have knowledge of the limit to correctly respond.

D: INCORRECT: Credible because all given temperatures are above the normal operating range (~140 degrees), so the applicant must have knowledge of the limit to correctly respond.

Question Level: F Question Difficulty 3 Justification:

The Reactor Operator needs to have fundamental knowledge of those parameters that require RCPs to be tripped.

Page 130 of 150

0POP04-RC-0002 Reactor Coolant Pump Off Normal Rev. 29 Page 99 of 99 Conditional Information Page RCP TRIP CRITERIA IF ANY VALID condition listed below occurs, THEN PERFORM the following:

1. IF the Reactor is critical, THEN PERFORM the following:

a. TRIP the Reactor.

b. ENSURE Main Turbine tripped.

2. STOP affected RCP(s)

3. CONTINUE at Step 1.0 of procedure x Motor Upper or Lower Radial Bearing Temp - GREATER THAN OR EQUAL TO 195qF x Lower Seal Water Bearing Temp - GREATER THAN OR EQUAL TO 230qF x Seal 1 Water Inlet Temp - GREATER THAN OR EQUAL TO 230qF x Motor Stator Winding Temp - GREATER THAN OR EQUAL TO 310qF x Number 1 Seal DP - LESS THAN 220 PSID x Case Vibration - a. GREATER THAN OR EQUAL TO 5 MILS Mtr_Accel-Vert b. GREATER THAN OR EQUAL TO 3 MILS AND RATE Mtr_Accel-Horiz OF VIBRATION INCREASE IS GREATER THAN OR EQUAL TO 0.2 MIL PER HOUR x Shaft Vibration - a. GREATER THAN OR EQUAL TO 20 MILS Brg2-Vert b. GREATER THAN OR EQUAL TO 15 MILS AND RATE Brg2-Horiz OF VIBRATION INCREASE IS GREATER THAN OR EQUAL TO 1.0 MIL PER HOUR RCP MOTOR THRUST BEARING TEMPERATURE HIGH IF Motor Upper or Lower Thrust Bearing Temp - GREATER THAN OR EQUAL TO 195qF, THEN, PERFORM Step 2 of this procedure.

RCP TRIP CRITERIA FOR LOSS OF SEAL INJECTION AND LOSS OF THERMAL BARRIER CCW IF an RCP experiences a simultaneous loss of seal water injection flow AND loss of CCW flow to thermal barrier, THEN STOP affected RCP within 1 minute.

RCP TRIP CRITERIA FOR HIGH NUMBER 1 SEAL LEAKOFF FLOW IF RCP Number 1 Seal leakoff flow increases to GREATER THAN 6 gpm OR pegged high, THEN PERFORM the following:

1. IF the Reactor is critical, THEN PERFORM the following:

a. Trip the Reactor.

b. ENSURE Main Turbine tripped.

2. STOP the affected RCP.

3. PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection.

4. CONTINUE actions of this procedure as resources permit.

5. CLOSE affected RCP Number 1 Seal leakoff isolation valve between 3 to 5 minutes after stopping RCP.

x RCP 1A(2A) "SEAL NO 1 LKF ISOL FV-3154" x RCP 1B(2B) "SEAL NO 1 LKF ISOL FV-3155" x RCP 1C(2C) "SEAL NO 1 LKF ISOL FV-3156" x RCP 1D(2D) "SEAL NO 1 LKF ISOL FV-3157"

6. MONITOR CCW flow - ADEQUATE.

This Procedure is Applicable in ALL Modes

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2200 Last used on an NRC exam: Never RO Sequence Number: 66 Unit 2 is performing a Plant Startup with Reactor Power currently at 8%.

Power Range Channel N41 INSTRUMENT and CONTROL power fuses are removed while maintenance is being performed on the detector. All protective bistables associated with N41 are in the TRIPPED condition.

Subsequently the INSTRUMENT power is lost to Power Range Channel N42.

Which of the following describes the IMPACT of this malfunction and the ACTION taken to mitigate the consequences?

IMPACT ACTION All protective bistables associated Manually control LPFRVs to respond to the A.

with N42 will BYPASS. Steam Generator level transient.

Perform immediate actions of 0POP05-EO-All protective bistables associated B. EO00, Reactor Trip or Safety Injection, to with N42 will TRIP.

respond to the Reactor Trip.

Perform immediate actions of 0POP05-EO-All protective bistables associated C. EO00, Reactor Trip or Safety Injection, to with N42 will BYPASS.

respond to the Reactor Trip.

All protective bistables associated Manually control LPFRVs to respond to the D.

with N42 will TRIP. Steam Generator level transient.

Answer: B All protective bistables associated with N42 will TRIP. Perform the immediate actions of 0POP05-EO-EO00, Reactor Trip or Safety Injection, to respond to the Reactor Trip.

Page 131 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2200 K/A Catalog Number: 012 A2.02 Tier: 2 Group/Category: 1 RO Importance: 3.6 10CFR

Reference:

55.41(b)(7)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument power.

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:

LOT 201.20 lesson on the Solid State Protection System and LOT 201.16 lesson on Excore NIS Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Impact is credible because the Nis have a bypass feature (however it is manually initiated) and some actuations have been changed from de-energize to actuate to energize to actuate (main steam and feedwater isolations). Action is credible because Power Range NIs do have an input to LPFRV automatic control (and are taken to manual in the off-normal for NIs).

B: CORRECT: The bistables for NIs actuate to the tripped condition when instrument or control power is lost. With N41 bistables in the tripped condition the Reactor will trip.

C: INCORRECT: Impact is credible because the Nis have a bypass feature (however it is manually initiated) and some actuations have been changed from de-energize to actuate to energize to actuate (main steam and feedwater isolations).

D: INCORRECT: Action is credible because Power Range NIs do have an input to LPFRV automatic control (and are taken to manual in the off-normal for NIs) and several other reactor trips are bypassed when less than 10% (i.e. pressurizer pressure low, pressurizer level high, RCP underfrequency).

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator must have knowledge of how instrument power for the NIs can affect the SSPS and be able to evaluate the given condition to determine the correct action to implement.

Page 132 of 150

POWERRANGE INSTRUMENT MAIN CONTROL AMPS BOARD I/A I/A NR-45 DETECTOR RPS CURRENT COMPUTER COMPARATOR I/A DT TRIPS

%POWER COMPUTER OVER I/A POWER POWER RECORDER SUPPLY I

AMP CHANNEL CP-018 COMPARATOR B/S I/A ROD CONTROL B/S OVERPOWER DETECTOR RODSTOP CURRENT COMPUTER COMPARATOR HIGH B/S LOWERPOWER VOLTAGE REACTORTRIPS FAILURE I/A I/A B/S HIGHPOWER B/S P-8 PERMISSIVE AMPS DIFF POSITIVE RATE B/S B/S P-9 AMP PERMISSIVE LOT201.16.TP.16 DWG - 11/28/02 B/S P-10 PERMISSIVE

POWERRANGE INSTRUMENT 100 V < NORM LOSS DET VOLT.

UPPER UIC UIC LOWER A B POWER SUPPLY A A AMMETER I

SAME I TOOT T, 0P T A

I RECORDER I

LOWERCURRENT, RECORDER A

SUMMING PLANT COMPUTER I METER LEVEL AMP DET. CURRENT COMPARATOR 2%DEVIATION (60 SECTIME DELAY) 25%

PR LORANGE TRIP 2/4 O-120%MCB ABOVE P-10

• I NR 45 1/4 A A PLANT COMPUTER OTHER3 CHANNELS P-10 PERMITS BLOCK OF PR 10%2/4 LORANGE TRIP AND INPUTS TOP-7 P

• I A

AUCTIONEER POWER PNUC PTURB MISMATCH DEFEAT SW.

PRHI RANGE 109%2/4 FLUX TRIP OVERPOWER A

• I A

RECORDER 0-200%

PR CHANNEL 1/4 COMPARATOR 4%DEV.

2/4 40% (5 SEC TIME DELAY)

P-8, 3 LOOP FLOWPERMISSIVE C-2 1/4 103%

• RODSTOP (2 SECTIME DELAY) A RODSTOP BYP. SW. 2/4 50%

P-9, TURBINE TRIP PERMISSIVE HI FLUX COMPARATOR

• BISTABLES BYPASSEDFOR CANBE RATE +5%IN 2 SEC MAINTENANCE ANDTESTING.

2/4 COMPARES 1/4 A O-120%NIS PANEL

• POWERLEVEL NOWWITH 2 SECONDS AGO LOT201.16.TP.17 DWG- 11/28/02

LOT201.16 TP.29a

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2201 Last used on an NRC exam: Never RO Sequence Number: 67 Given the following on Unit 1:

A LOOP has occurred and all Emergency Diesel Generators (EDG) are running.

A fire in the relay room has caused a Control Room evacuation and 0POP04-ZO-0001, Control Room Evacuation, is being performed.

As secondary RO, you have performed the initial actions at ZLP-653 and ZLP-700 in Train A Switchgear Room per 0POP04-ZO-0001 to transfer equipment control to the local panels.

Which of the following describes actions associated with Essential Cooling Water (ECW) performed at ZLP-653 and 700 and the resultant affects on ESF Diesel #11 operation?

Action(s) Affects ECW Pump A remains running and the Transferred controls for ECW Pump A A. discharge valve remains open with no and discharge valve affect on the operation of ESF DG #11 Transferred control of ECW Pump A ECW Pump A remains running with no B.

only affect on the operation of ESF DG #11 ECW Pump A trips and the discharge Transferred controls for ECW Pump A C. valve remains open requiring the pump to and discharge valve be restarted to restore cooling to the DG.

ECW Pump A trips requiring the pump Transferred control of ECW Pump A D. to be restarted to restore cooling to the only DG.

Answer: B Transferred control of ECW Pump A only; ECW Pump A remains running with no affect on the operation of ESF DG #11 Page 133 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2201 K/A Catalog Number: 064 G2.4.34 Tier: 2 Group/Category: 1 RO Importance: 4.2 10CFR

Reference:

55.41(b)(10)

Emergency Diesel Generator: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

STP Lesson: LOT 201.13 Objective Number: 91193 LIST all automatic functions, switch locations, switch positions, annunciators (and where indicated),

local/remote functions, interlocks and permissive for the following:

A. ECW Traveling Screens B. ECW Screen Wash Booster Pumps C. ECW Screen Wash Valves D. ECW Strainers E. ECW Pumps and Motors F. ECW Discharge Valves G. ECW Sump H. ECW Blowdown Valve I. ECW Sump Pump and Motor

Reference:

LOT201.13 PowerPoint slides 40 & 44 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Action is credible because it would be reasonable to think the valve controls would be located with the pump controls (such as with the CCW pumps).

B: CORRECT: The secondary RO only transfers control of the ECW pump (valve control is transferred at the ECW structure). When control is transferred, equipment status is not affected.

C: INCORRECT: Action is credible because it would be reasonable to think the valve controls would be located with the pump controls (such as with the CCW pumps). Affect is credible because the transfer operation physically moves control of the equipment from one location to another which could result in the loss of the run signal and stop the pump.

D: INCORRECT: Affect is credible because the transfer operation physically moves control of the equipment from one location to another which could result in the loss of the run signal and stop the pump.

Question Level: F Question Difficulty 3 Justification:

Applicant requires a knowledge of local controls for ECW and the operational consequences of their use.

Page 134 of 150

ECW PUMPS Transfer switches located in the EAB Switchgear Rooms on ZLP-653, 654, 655; Trains A,B,C Remote: allows auto operation from the ECW/CCW Train selector switches and sequencer Auto start:

SI starts the pump after 25 sec. sequencer delay Loss of power (LOOP) or SI coincident with LOOP starts the pump 25 sec. after the DG output is connected to ESF bus Low CCW header pressure, at 76 psig dec.

Low pressure on the other two trains of ECW, at 30 psig dec. (15 sec TD added for CC/ECW Lo pressure auto start)

DISCHARGE VALVE Transfer switch at the MCC "Local" control at the MCC prevents ECW Pump control from CR "Remote" control:

Pump start: valve opens after 10 seconds Pump stop: valve closes after 2 minutes Status Monitoring alarm if: Valve fails to open Transfer switch in "LOCAL" Loss of control power Thermal overload computer alarm with indication on the MCC

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2202 Last used on an NRC exam: Never RO Sequence Number: 68 Which of the following describes the physical location of Core Exit Thermocouples and their relationship to the Reactor Coolant System?

A. Thermocouples are positioned just above the top of EACH Fuel Assembly and are used to determine Reactor Coolant System Subcooling.

B. Thermocouples are positioned just above the top of EACH Fuel Assembly and are used to determine Control Rod Insertion Limits.

C. Thermocouples are positioned just above the top of SELECTED Fuel Assemblies and are used to determine Control Rod Insertion Limits.

D. Thermocouples are positioned just above the top of SELECTED Fuel Assemblies and are used to determine Reactor Coolant System Subcooling.

Answer: D Thermocouples are positioned just above the top of SELECTED Fuel Assemblies and are used to determine Reactor Coolant System Subcooling.

Page 135 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2202 K/A Catalog Number: 017 K1.02 Tier: 2 Group/Category: 2 RO Importance: 3.3 10CFR

Reference:

55.41(b)(3)

Knowledge of the physical connections and/or cause-effect relationship between the ITM system and the following systems: RCS STP Lesson: LOT 201.17 Objective Number: 91337 DESCRIBE the operation of the Incore Thermocouples.

Reference:

LOT 201.17 lesson on Incore Nuclear Instrumentation.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because it would be reasonable to believe temperature is monitored at the exit of all assemblies.

B: INCORRECT: Number is credible because it would be reasonable to believe temperature is monitored at the exit of all assemblies. Useage is credible because the calculation for Control Rod Insertion Limits has an RCS temperature input, but it is core delta-T.

C: INCORRECT: Useage is credible because the calculation for Control Rod Insertion Limits has an RCS temperature input, but it is core delta-T.

D: CORRECT: There are 50 CETs placed just above selected Fuel Assemblies. They measure exit temperature of the core and feed into the calculation for RCS subcooling.

Question Level: F Question Difficulty 3 Justification:

The Reactor Operator needs to have fundamental knowledge of the operation of the core exit thermocouples.

Page 136 of 150

LOT201.17.HO.01 Rev. 3 Page 12 of 19 LEAK DETECTION SYSTEM The leak detection system consists of a drain header connecting the 10-path transfer devices, a pressure switch, drainage solenoid valve, an alarm light and reset pushbutton mounted on the distribution panel in the control room. Liquid collecting in a 10-path transfer due to a leak will cause the water level to rise in the drain header and thus actuate the pressure switch. The switch will then energize the audible alarm and the alarm light on the distribution panel while energizing the drainage solenoid to dump the water to the plant drain. The alarm is acknowledged by pressing the lighted reset pushbutton. This silences the audible alarm. When the water level in the drain header decreases below the pressure switch setting the alarm light goes out, the solenoid valve closes and the system returns to normal.

If a reactor coolant leak should develop in any of the incore thimbles, it may be detected by the leak detection system or by indicated abnormal radiation levels within the plant containment. Also difficulty of detector insertion may indicate a leaky thimble. Once such a leak has been detected, it should be possible to determine which thimble is faulty by visual observation at the path indicator switches (after removal of the covers). If it is a small leak, it may be possible to determine which thimble is affected by lightly touching the tubes between the seal plate and the movable frame to find the one having the highest temperature.

Proof of a small leak can best be done by inserting a dummy cable manually into the suspected thimble. If water is collected, it can be chemically analyzed to determine if it is reactor coolant. After identification of the leaking thimble, a small leak can be isolated during either a hot or cold shutdown by capping the appropriate thimble. A large leak would probably require cold shutdown for access to the thimble.

INCORE THERMOCOUPLE SYSTEM Thermocouple The Thermocouple System utilizes 50 thermocouples, positioned to measure fuel assembly coolant outlet temperature at preselected core locations. The thermocouples are the chromel-alumel type and have an accuracy of 2F. The thermocouple system is class IE, is divided into two redundant, independent and separate groups, Train A and Train C. (CET - 18 in Unit 1 is abandoned in place per DCP 06-677)

Thermocouple Routing and Seal Assemblies Each thermocouple is 1/8-inch (nominal) diameter, stainless steel sheathed, aluminum oxide insulated, with the trailing end terminated in a male thermocouple connector. The thermoelectric characteristics conform to the K calibration curve within 2 from zero to 530F and within 3/8 percent of point from 530F to 700F. Each thermocouple is supplied to the specific length required for its assigned location.

The sheaths, which are removable, are routed in guide tubes which position the thermocouple end at the selected core location. The guide tubes extend the entire distance from the core location to the seal assemblies.

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2203 Last used on an NRC exam: Never RO Sequence Number: 69 Unit 1 has just tripped from 100% Power.

The crew tripped the Reactor when it was determined that Control Rods were stepping out in an uncontrolled manner.

Given this condition, Deaerator (DA) level is expected to be A. trending up. Open DA high level dump bypass valves to prevent DA level from trending above 80%.

B. trending up. Close Condensate to DA inlet valves, CD-MOV-0574 & 0575 to prevent DA level from trending above 80%.

C. trending down. Start available Condensate Pumps to prevent DA level from trending below 30%.

D. trending down. Stop all Feedwater Booster Pumps to prevent DA level from trending below 30%.

Answer: A trending up. Open DA high level dump bypass valves to prevent DA level from trending above 80%.

Page 137 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2203 K/A Catalog Number: 045 A1.06 Tier: 2 Group/Category: 2 RO Importance: 3.3 10CFR

Reference:

55.41(b)(5)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with the MT/G system controls including: Expected response of secondary plant parameters following T/G trip.

STP Lesson: LOT 504.06 Objective Number: 81674 Given a step, note, or caution from 0POP05-EO-ES01, STATE/IDENTIFY the basis for the step, note or caution and the basis for the action to include the action itself, its purpose and result.

Reference:

LOT 504.06 lesson on 0POP05-EO-ES01, Reactor Trip Response Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT: DA level would trend up after a main turbine/generator trip with out complications on the secondary. Due to water entering the DA after the trip from sources other than the condensate inlet (e.g. condensate vent condenser) opening the DA high level dump bypass valves will be effective in controlling the high water level expected in the DA. Keeping DA level below 80% will prevent adverse effects on secondary piping.

B: INCORRECT: Credible because it is reasonable to think that closing the inlet MOVs would prevent water from entering the DA, however there are sources that bypass these valves.

C: INCORRECT: Trend is credible since several sources of water (MSDT discharge, FWH #11 drains) are no longer present. Action is credible because condensate pumps supply water to the DA and starting additional pumps would add additional water.

D: INCORRECT: Trend is credible since several sources of water (MSDT discharge, FWH #11 drains) are no longer present. Action is credible because the booster pumps remove water from the DA and the action would stop the removal.

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator needs to have knowledge of the response of the secondary plant after a main turbine/generator trip and be able to evaluate that the given condition would not complicate the secondary response to the main turbine/generator trip.

Page 138 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2204 Last used on an NRC exam: Never RO Sequence Number: 70 Unit 1 is at 100% Power.

A Condensate Pump trips and the Standby Condensate Pump will not start leaving just one Condensate Pump running.

Which of the following components is affected to the extent that Operator Action is required to maintain proper operation?

Seal Water to the A. Low Pressure Heater Drip Pumps B. Feedwater Booster Pumps C. Condensate Pumps D. Turbine Driven Steam Generator Feed Pumps Answer: D Turbine Driven Steam Generator Feed Pumps Page 139 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2204 K/A Catalog Number: 056 K1.03 Tier: 2 Group/Category: 2 RO Importance: 2.6 10CFR

Reference:

55.41(b)(7)

Knowledge of the physical connections and/or cause-effect relationship between the Condensate system and the following systems: Main Feedwater STP Lesson: LOT 202.10 Objective Number: 40110 LIST all the systems that interface with the Condensate System and state the function of each interface.

Reference:

LOT 202.10 lesson on the Condensate System and 0POP04-CD-0001, Loss of Condensate Flow Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because condensate supplies seal water to these pumps, but at a lower pressure.

B: INCORRECT: Credible because condensate supplies seal water to these pumps, but at a lower pressure.

C: INCORRECT: Credible because condensate supplies seal water to these pumps, but at a lower pressure.

D: CORRECT: Under the given condition, condensate pressure would not be high enough to supply proper seal water flow to the SGFPTs and operator action would be required to raise condensate pressure (manually reduce flow to the DA).

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator must be able to evaluate the given condition to determine the effect on other interfacing systems.

Page 140 of 150

0POP04-CD-0001 Loss of Condensate Flow Rev. 14 Page 6 of 53 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED Step 3.0 continued from previous page

e. CHECK for other valves in the Condensate to Deaerator flowpath that may be isolated. Refer to Addendum 4, Condensate Flowpath To the DA .

f. IF flowpath can NOT be re-established, THEN COMMENCE a turbine load reduction per Addendum 3, Turbine Load Reduction.

_____ 4.0 CHECK Any Of The Following GO TO Step 6.0 Annunciators Lit:

x "SGFPT 11(21) SEAL WTR DP LO" Lampbox 6M03, Window E-5 x "SGFPT 12(22) SEAL WTR DP LO" Lampbox 6M04, Window E-1 x "SGFPT 13(23) SEAL WTR DP LO" Lampbox 6M04, Window E-5

_____ 5.0 TAKE Manual Control Of The DEAER Level Control Valves AND THROTTLE Closed Until Condensate Header Pressure Is High Enough To Clear All "SGFPT SEAL WTR DP LO" Alarms {CP009}

(Approximately 400 To 450 PSIG)

This Procedure is Applicable in Modes 1, 2, and 3

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2205 Last used on an NRC exam: Never RO Sequence Number: 71 A fault in the steam pressure transmitter for the controlling Steam Flow channel on SG 1A causes the pressure indication to drop 50 psig.

Which one of the following describes the Steam Generator Water Level Control System response to this fault?

Assume program p for SGFPT speed control does NOT change.

SG 1A Main Feedwater Regulation Valve will initially A. open to match feedwater flow with indicated steam flow. When stabilized, SG 1A level will be controlling on program.

B. open to match feedwater flow with indicated steam flow. When stabilized, SG 1A level will be controlling slightly higher than program.

C. close to match feedwater flow with indicated steam flow. When stabilized, SG 1A level will be controlling on program.

D. close to match feedwater flow with indicated steam flow. When stabilized, SG 1A level will be controlling slightly lower than program.

Answer: C close to match feedwater flow with indicated steam flow. When stabilized, SG 1A level will be controlling on program.

Page 141 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2205 K/A Catalog Number: 035 A3.01 Tier: 2 Group/Category: 2 RO Importance: 4.0 10CFR

Reference:

55.41(b)(7)

Ability to monitor automatic operation of the S/G including: S/G water level control STP Lesson: LOT 202.15 Objective Number: 21005 IDENTIFY the level controller, the manual/auto station, all input signals to and all output signals from the SGWLCS. STATE how a change in each input signal will affect the position of the Main Feed Regulating Valves.

Reference:

LOT202.15 handout page 8 & 9 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because steam flow is pressure compensated and if the applicant incorrectly believes the failure will cause indicated steam flow to rise, then the valve response would be correct.

B: INCORRECT: Valve response is credible because steam flow is pressure compensated and if the applicant incorrectly believes the failure will cause indicated steam flow to rise, then the valve response would be correct. Long term response is credible because the indicated response would be correct if the controller did not have an integral function.

C: CORRECT: A drop in indicated steam pressure will cause a drop in indicated steam flow which will result in the MFRV closing to reduce feed flow to match indicated steam flow. Since the controller has an integral function, level will ultimately be restored to program.

D: INCORRECT: Long term response is credible because the indicated response would be correct if the controller did not have an integral function.

Question Level: H Question Difficulty 3 Justification:

The applicant must first determine how the pressure change will affect steam flow, then apply the change in steam flow to the level control system to determine initial response. A knowledge of the type of controller (PI) is then needed to determine long term response of the affected SG level.

Page 142 of 150

LOT202.15.HO. Rev. 9 PAGE 8 OF 11 The large gain of 2.5 to 1 in the P+I level controller ensures that the control system is level dominate and will attempt to always return indicated level to setpoint. However, when a level instrument fails, the indicated level does not show actual level and the control system will respond to the failed instrument, driving actual level away from setpoint regardless of the magnitude of the instrument failure.

Anytime actual feed flow is changed, the feed-to-steam DP input to the SGFP Master Speed Controller changes. If a controlling level channel fails high or low, the actual feed flow and the feed/steam DP is decreased or increased accordingly.

This will make the SGFPs speed up or slow down to recover the system DP to the programmed setpoint and may help in slowing the actual SG level change; however, the constant level error from the failed channel will continually attempt to close or open the MFRVs.

6.2 STEAM FLOW FAILURE 6.2.1 Steam Flow Channel Failing LOW A controlling steam flow channel failing low from 100% power will initially result in a summed feed/steam flow error of 100%. This large flow error will rapidly decrease feedwater flow because the MFRV receives a signal to go closed to about 40% demand (a 0.6%

proportional gain between flow error and level error output signal). The MFRV will slowly continue to close further as long as there is a difference between flow error and level error.

As the MFRV closes, feedwater flow rapidly decreases. Actual SG level drops and now a level error is developed. Again, a proportional gain of 2.5 is applied on this level error to make this a level dominant system. When the level error output signal is large enough to overcome the flow error, the MFRV will start to come open and increase feed flow in an attempt to restore actual SG level to program.

However, with the Steam Flow channel still failed low, as soon as feed flow increases, the flow error gets larger and starts to counteract the level error output signal. From this point, its a race to the SG LO LO level trip setpoint.

At the time of the steam flow channel failing low the total steam flow signal decreases resulting in a lower DP setpoint for the SGFP Master Controller. The SGFPs will eventually control to a lower DP setpoint.

As the flow control valve closes, feedwater flow rapidly decreases, SG level drops and a level error is developed. A proportional gain of 2.5 is applied on this level error to make this a level dominate system.

An additional item to consider is the SGFPT speed control system. At the time of the steam flow channel failing low, the total steam flow signal decreases resulting in a lower DP setpoint for the SGFP Master Controller. The SGFPs will eventually control to a lower DP setpoint.

LOT202.15.HO. Rev. 9 PAGE 9 OF 11 For example, with the plant at 100% power and a Steam Flow channel failed low, a 100% flow error is produced. As the MFRV closes, the actual feedwater flow decreases and the flow error rapidly decreases from 100%.

Once the flow error signal reaches <50%, for example, then it would take a level error signal of 20% to overcome the flow error (20% x 2.5 = 50% so, 50.7% actual SG level). However, once actual feed flow starts to increase, the flow error would get larger, and once again, close the MFRV. With the plant at 100% power, it would take a 40% level error signal to restore 100% feed flow (40% x 2.5 = 100%

so, 30.7% actual SG level) With the SG LO LO level setpoint of 20 %,

this means that SG level could theoretically be restored with no operator action.

This will happen on a controlling steam flow channel failing low when

>50% power. On steam flow channel low failures when <50% power, the response is not as drastic and the resulting level error output signal can overcome the flow error and will respond to return the SG level to setpoint. As long as the level deviates from setpoint, the P+I level controller will generate a level error output signal on an 1800 second reset time constant, allowing level to slowing return to setpoint with minimal overshoot.

6.2.2 Steam Flow Channel Failing HIGH The feedwater and steam flow transmitters are calibrated for 0 to 118% flow. A controlling steam flow channel failing high from 100%

results in an 18% flow error that will initially open the MFRV an additional 11% demand and then slowly continue to open the valve until the level error output signal becomes strong enough to start closing the valve. When the steam flow channel failed high the total steam flow signal increases and results in a higher DP setpoint for the SGFP Master Speed Controller. The SGFPs will eventually be controlled at a higher DP setpoint.

As actual feedwater flow increases, the SG level increases and a level error output signal is developed. When actual level increases > 7.2%

(7.2% x 2.5 = 18), a level error output signal will be generated that is strong enough to overcome the 18% flow error. Additionally, with feedwater flow increasing, the flow error is lower and the level may not have to increase as far above setpoint before the generated level error output signal will start to close the MFRV. Because the flow error is < the level controllers level error output signal, it can compensate to bring feedwater flow back to 100% and slowly return level to setpoint.

If a controlling steam flow channel fails high at < 68% power (a >

50% flow error), the level controllers level error output signal

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2209 Last used on an NRC exam: Never RO Sequence Number: 72 Given the following:

During core off load it is reported by the refueling crew that SFP and Rx cavity level is lowering.

Plant personnel report water coming out of an open Steam Generator manway (nozzle dam failure)

Control Room operators enter 0POP04-FC-0002, Refueling LOCA.

A Plant Operator is dispatched to close the FUEL TRANSFER TUBE GATE VALVE per 0POP04-FC-0002.

Which of the following describes how the plant operator will close the valve and the resultant effect of the valve closure?

Valve Closure Effect Closed by manually turning local SFP level stabilized, ICSA and Rx A. hand wheel Cavity continue to lower Closed by manually turning local SFP and ICSA level stabilized, Rx B. hand wheel Cavity level continues to lower Closed by placing local Handswitch SFP level stabilized, ICSA and Rx C. to CLOSE Cavity continue to lower Closed by placing local Handswitch SFP and ICSA level stabilized, Rx D. to CLOSE Cavity level continues to lower Answer: A closed by manually turning hand wheel; SFP level stabilized, ICSA and Rx Cavity continue to lower Page 143 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2209 K/A Catalog Number: 033 G2.4.35 Tier: 2 Group/Category: 2 RO Importance: 3.8 10CFR

Reference:

55.41(b)(10)

Spent Fuel Pool Cooling: Knowledge of local auxilary operator task during an emergency and resultant operational effects.

STP Lesson: LOT 505.01 Objective Number: 38635 Given an abnormal operating event, PREDICT the symptoms expected to occur in accordance with the appropriate off normal operating procedure.

Reference:

0POP04-FC-0002 rev 14 and 0POP08-FH-0003 rev 32 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: CORRECT - Fuel transfer canal gate valve is a manually operated valve which isolates the SFP from the ICSA and the Rx Cavity. With a leak inside containment, the ICSA and Rx cavity will continue to lower.

B: INCORRECT - Credible because there is also an ICSA gate, which if closed in this situation would produce the given indication.

C: INCORRECT - Credible because many large valves are motor operated (this one is not).

D: INCORRECT - Valve closure is credible because many large valves are motor operated (this one is not). Effect is credible because there is also an ICSA gate, which if closed in this situation would produce the given indication.

Question Level: H Question Difficulty 3 Justification:

The applicant must have knowledge of and recall that the valve is manually operated and determine leak location based on the given information which will lead to a conclusion for overall plant status.

Page 144 of 150

0POP04-FC-0002 Refueling LOCA Rev. 14 Page 4 of 71 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 6.0 SUSPEND Movement Of Loads Over The SFP

_____ 7.0 CHECK Fuel Transfer Cart - AT SFP PLACE fuel transfer cart on the SFP side with SIDE WITH UPENDER DOWN the upender down.

_____ 8.0 CHECK 1(2)-FH-0001, FUEL CLOSE 1(2)-FH-0001, FUEL TRANSFER TRANSFER TUBE GATE VALVE" - TUBE GATE VALVE CLOSED

_____ 9.0 CHECK Reactor Internals Movement - IN GO TO Step 11.0.

PROGRESS

_____ 10.0 SECURE Reactor Internals In The Desired Storage Area

_____ 11.0 PERFORM The Following:

_____ a. NOTIFY Health Physics of the current plant conditions

_____ b. INSTRUCT Health Physics to perform a radiation survey of the SFP and RCB areas

_____ c. ENSURE Appropriate actions are taken based on the results of the survey This Procedure is Applicable Anytime the Reactor Head is Off the Vessel

0POP08-FH-0003 Rev. 32 Page 34 of 47 Fuel Transfer System 8.4.10 OPEN the following power supply breakers for the RCB Control Console: [35 ft EAB Pen Space MCC 1K1(2K1)]

8.4.10.1 "1K1(2K1)/A3R RX SIDE FUEL HANDLING CONT PNL ZLP-111" _____

8.4.10.2 "1K1(2K1)/D4L B/U BKR-RX SIDE FUEL HANDLING CONT PNL ZLP-111" _____

CAUTION The Operator closing Fuel Transfer Tube Gate Valve SHALL be aware it takes approximately 109 turns to close gate valve from full open position.

Closing 1(2)-FH-0001 FUEL TRANSFER SYSTEM TRANSFER TUBE ISOLATION VALVE more than 109 turns may result in damage to the valve.

IF 1(2)-FH-0001 FUEL TRANSFER SYSTEM TRANSFER TUBE ISOLATION VALVE is turned approximately 109 turns in closed direction AND position indicator does NOT indicate closed, THEN valve SHALL NOT be closed any further until difference is resolved. (Reference 2.9) 8.4.11 CLOSE "1(2)-FH-0001 FUEL TRANSFER SYSTEM TRANSFER TUBE ISOLATION VALVE". (68 ft FHB N End of Transfer Canal) _____

8.4.12 NOTIFY Maintenance to close Fuel Transfer Quick Opening Hatch. _____

8.4.13 NOTIFY Health Physics that Fuel Transfer Operations through transfer canal have been secured. _____

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2206 Last used on an NRC exam: Never RO Sequence Number: 73 Unit 2 tripped from 100% Power due to a LOCKOUT on the Unit Aux Transformer.

Currently the Crew is performing 0POP05-EO-ES01, Reactor Trip Response, and is checking to see if Reactor Coolant Pumps can be started.

Main Steam has been isolated.

The following temperatures and pressures are reported by the Reactor Operator:

PZR Pressure 1925 psig and lowering CET Temperature 605ºF and rising S/G Pressures 1200 psig and lowering RCS Hot Leg Temperatures 600 ºF and rising RCS Cold Leg Temperatures 595 ºF and rising RCS subcooling 27 ºF and lowering Which of the following describes the action (and basis) the Crew should perform next?

A. Raise Steam Dumping Rate to aid in establishing natural circulation.

B. Raise Aux Feedwater Flow to aid in establishing natural circulation.

C. Initiate Safety Injection due to loss of subcooling.

D. Initiate Safety Injection to aid in identifying a faulted S/G.

Answer: C Initiate Safety Injection due to loss of subcooling.

Page 145 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2206 K/A Catalog Number: G2.1.7 Tier: 3 Group/Category: 1 RO Importance: 4.4 10CFR

Reference:

55.41(b)(5)

Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.

STP Lesson: LOT 504.06 Objective Number: 81674 Given a step, note, or caution from 0POP05-EO-ES01, STATE/IDENTIFY the basis for the step, note or caution and the basis for the action to include the action itself, its purpose and result.

Reference:

LOT 504.06 lesson on 0POP05-EO-ES01, Reactor Trip Response (CIP) and Steam Tables Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because the given conditions indicate natural circulation is not occurring and raising steaming rate would be appropriate to help establish natural circulation.

B: INCORRECT: Credible because the given conditions indicate natural circulation is not occurring and raising Aux Feedwater flow could help establish natural circulation.

C: CORRECT: With RCS Subcooling at 27 degrees F, initiation of SI is required because it is below the required limit of 35 degrees F. (See procedure CIP)

D: INCORRECT: Credible because an SI is required, however it would not assist in identifying a faulted generator (a main steam isolation would) and there are no given indications of a faulted generator (pressures are lowering but it is due to loss of natural circulation).

Question Level: H Question Difficulty 3 Justification:

The Reactor Operator needs to be able to evaluate the given condition and determine the appropriate action.

Page 146 of 150

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Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2071 Last used on an NRC exam: Never RO Sequence Number: 74 Given the following:

New Fuel receipt is in progress in the Fuel Handling Building.

A New Fuel Assembly is dropped resulting in a breach of the Fuel Cladding.

The resulting radiation hazard is primarily A. EXTERNAL exposure due to the presence of neutron radiation.

B. EXTERNAL exposure due to the presence of alpha radiation.

C. INTERNAL exposure due to the presence of neutron radiation.

D. INTERNAL exposure due to the presence of alpha radiation.

Answer: D INTERNAL exposure due to the presence of alpha radiation.

Page 147 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 2071 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 103.02 Objective Number: N91217 Contrast the effects of external and internal exposure.

Reference:

LOT 103.04 student handout page 23 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from Distractor Justification A: INCORRECT: Credible because neutron radiation is generally associated with nuclear fuel, more specifically it should be spent nuclear fuel. The hazard is credible because if there were neutrons, the hazard would be primarily external as stated.

B: INCORRECT: Credible because all other forms of radiation have sufficient penetrating power to be an external hazard, however Alpha particles travel only a short distance due to their ionizing potential and can easily be shielded by a person's skin.

C: INCORRECT: Credible because neutron radiation is generally associated with nuclear fuel, more specifically it should be spent nuclear fuel.

D: CORRECT: Alpha radiation is the primary hazard for new fuel due to the decay of the uranimum.

Alpha particles travel only a short distance due to their ionizing potential and can easily be shielded by a person's skin.

Question Level: F Question Difficulty 3 Justification:

Applicant must know the radiation/contamination hazard for new fuel and that alpha particles are primarily an internal radiation hazard.

Page 148 of 150

LOT103.04.HO.01 Rev. 7 Page 23 of 26 Concrete 0.089 cm-1 Rule of Thumb: Neutron dose rate in water.

The tenth-thickness concept can also be applied to neutron dose rate attenuation. The rule of thumb is: 10 inches of water or polyethylene is necessary to reduce the neutron dose rate by a factor of 10.

Example:

Calculate the tenth value layer of water for fast neutrons given r(H20) = 0.103 cm-1.

Solution:

I I o e r x 1

I Io 10 1

I o I o e r x1/10 10 1 1 e (0.103cm )x1/10 10 1

ln (0.103cm 1 )x1/10 10 1

ln x1/10 10 0.103cm 1 x1/10 22.4cm 8.8inches Alpha and Beta Attenuation:

Since alpha and beta radiation have relatively low penetrating power (because of their high ionizing potential) they are primarily an internal radiation hazard.

Alpha particles lose energy rapidly in any medium because of high specific ionization (large size and charge). Alpha particles normally produced by fission are contained within the fuel elements and furthermore can be stopped by a sheet of paper. The outer layer of skin will absorb alpha particles up to 7.5 MeV.

Betas are usually absorbed by material containing the radioactive source or by any shielding employed to reduce gamma levels. Beta particles, have a lower specific ionization, therefore their penetration into any absorber will be much greater that of an alpha particle. Beta is considered a slight external hazard, since a 70 KeV beta will penetrate the skin. It is primarily a hazard to the lens of the eyes.

Date: 08/05/13

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1830 Last used on an NRC exam: 2009 RO Sequence Number: 75 To satisfy the Reactor Coolant System Pressure SAFETY LIMIT, Reactor Coolant System pressure cannot exceed _______ psig.

A. 2380 B. 2485 C. 2735 D. 3110 Answer: C 2735 Page 149 of 150

Print Date 8/7/2013 STP LOT-19 NRC RO EXAM Exam Bank No.: 1830 K/A Catalog Number: G2.2.22 Tier: 3 Group/Category: 2 RO Importance: 4.0 10CFR

Reference:

55.41(b)(5)

Knowledge of limiting conditions for operations and safety limits.

STP Lesson: LOT 201.02 Objective Number: 92102 (RCS) 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 adminstrative requirements affected, limitations, major time frames involved, major surveillance in order to comply, and the bases for the specification.

Reference:

Safety Limit 2.1.2 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified from Distractor Justification A: INCORRECT: Credible because 2380 is an RCS setpoint (high pressure trip setpoint).

B: INCORRECT: Credible because 2485 is an RCS setpoint (Pzr safety lift setpoint).

C: CORRECT: Per Safety Limit 2.1.2, RCS pressure must not exceed 2735 psig.

D: INCORRECT: Credible because 3110 is an RCS setpoint (hydro criteria for the RCS).

Question Level: F Question Difficulty 3 Justification:

The applicant must have a knowledge of the Safety Limits.

Page 150 of 150

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2181 Last used on an NRC exam: Never SRO Sequence Number: 76 The following events occur simultaneously:

A Fire in the Relay Room Requiring Fire Brigade Response A Large Main Steam Line Break in Containment Which of the following procedures takes precedence?

A. 0POP05-EO-EO00, Reactor Trip or Safety Injection B. 0POP05-EO-FRZ1, Response to High Containment Pressure C. 0POP04-ZO-0008, Fire/Explosion D. 0POP04-ZO-0001, Control Room Evacuation Answer: D 0POP04-ZO-0001, Control Room Evacuation Page 1 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2181 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 coordination 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: 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:

LOT 504.04 - 0POP01-ZA-0018, Emergency Operating Procedure User's Guide Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because this is the procedure normally entered first in the event of a steam break.

B: INCORRECT: Credible because a large steam break in containment will require performing this procedure.

C: INCORRECT: Credible because this procedure will be entered during a fire, but does not take precedence.

D: CORRECT: 0POP04-ZO-0001, Control Room Evacuation, takes precedence over all other EOPs and 0POP04-ZO-0008, Fire/Explosion or 0POP04-ZO-0009, Safe Shutdown Fire Response.

Question Level: F Question Difficulty 3 Justification:

The Unit Supervisor needs to have knowledge of the priority of procedure usage for a given condition.

Page 2 of 50

0POP01-ZA-0018 Rev. 21 Page 29 of 48 Emergency Operating Procedure User's Guide 7.0 EOP Network Usage 7.1 Entry into the EOPs is limited to the following conditions:

7.1.1 WHEN the Reactor is in Mode 1, 2, OR 3 with RCS pressure GREATER THAN 1000 PSIG AND any reactor trip or safety injection occurs OR is required (this includes a manual reactor trip and / or safety injection in response to approaching a reactor trip / safety injection setpoint such that an automatic action is imminent), THEN 0POP05-EO-EO00, Reactor Trip Or Safety Injection, SHALL be entered, unless the Control Room has been evacuated OR a complete loss of all AC ESF busses has occurred.

7.1.2 WHEN the Reactor is in Mode 1, 2, 3, OR 4 AND a complete loss of power on all AC ESF busses occurs, THEN 0POP05-EO-EC00, Loss Of All AC Power, SHALL be entered, unless the Control Room has been evacuated. This entry condition also applies during the performance of ANY other EOP.

7.1.3 IF the Control Room has been evacuated, THEN 0POP04-ZO-0001, Control Room Evacuation, SHALL take precedence over all EOPs AND 0POP04-ZO-0008 and 0POP04-ZO-0009.

7.1.4 IF a fire occurs in Fire Areas 02 - 78, THEN 0POP04-ZO-0009, Safe Shutdown Fire Response, SHALL take precedence over all EOPs.

• IF a reactor trip occurs or is directed during the performance of 0POP04-ZO-0009, THEN the actions of 0POP04-ZO-0009 SHALL be performed as directed by the CIP page or procedure step.

• Operator actions of this procedure section 4.0 that DO NOT conflict with the actions of 0POP04-ZO-0009 may be taken during 0POP04-ZO-0009 following a reactor trip to limit the effects of the reactor trip on the plant.

The actions to establish and maintain a heat sink, limit RCS cooldown and establish RCS pressure and inventory control are examples of (but not limited to) prudent actions that should be taken.

• Actions may be taken per EOPs, Off Normal Operating Procedures and Annunciator Response Procedures that DO NOT conflict with the actions of 0POP04-ZO-0009 if adequate resources are available. The EOP, Off Normal Operating Procedure or Annunciator Response Procedure should be entered and procedure steps followed. (e.g., IF during the performance of the 0POP04-ZO-0008/9 there are indications of abnormal RCP conditions, THEN the RCP Off Normal Operating Procedure SHOULD be entered.)

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2178 Last used on an NRC exam: Never SRO Sequence Number: 77 Authorization has been given to allow work on the packing of a motor operated valve (MOV) using the valve backseat as the boundary.

In accordance with 0PGP03-ZO-ECO1A, Equipment Clearance Order Instructions, who was responsible for giving authorization to perform the work and how shall the authorization be documented?

AUTHORIZATION DOCUMENTATION A Plant Manager and Engineering In the ECO Notes section of the ECO Division Manager Form B Plant Manager and Engineering In the General Information of the Shift Division Manager Manager Shift Turnover Checklist C Unit Operations Manager and In the General Information of the Shift Maintenance Division Manager Manager Shift Turnover Checklist D Unit Operations Manager and In the ECO Notes section of the ECO Maintenance Division Manager Form Answer: D Unit Operations Manager and Maintenance Division Manager. In the ECO Notes section of the ECO Form.

Page 3 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2178 K/A Catalog Number: G2.2.13 Tier: 3 Group/Category: 2 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(3)

Knowledge of tagging and clearance procedures.

STP Lesson: LOT 802.31 Objective Number: SRO-1172 STATE the MOV Manual Seating Guidelines

Reference:

LOT 507.01 - 0PGP03-ZO-ECO1A, Equipment Clearance Order Instructions, addendum 3 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because at one time the Plant Manager was the approval authority for this and engineering is considered the equipment experts.

B: INCORRECT: Credible because at one time the Plant Manager was the approval authority for this and engineering is considered the equipment experts. Documentation is credible because the authorization could be listed on the Shift Manager Turnover Checklist, but there is no retention requirements for use of the checklist whereas there is retention requirements for an ECO.

C: INCORRECT: Documentation is credible because the authorization could be listed on the Shift Manager Turnover Checklist, but there is no retention requirements for use of the checklist whereas there is retention requirements for an ECO.

D: CORRECT: Lists the correct authorization and documentation requirements.

Question Level: F Question Difficulty 3 Justification:

SRO must have knowledge of authorization and administrative requirements of the ECO program.

Page 4 of 50

0PGP03-ZO-ECO1A Rev. 19 Page 77 of 107 Equipment Clearance Order Instructions Addendum 3 MOV Manual Seating Requirements Page 1 of 2 CAUTION

• MOVs should be manually seated (handwheel) with minimum applied force to prevent exceeding the compensating spring pack deflection.

• Valve wrenches SHALL NOT be used on MOVs for manual seating. (Reference 2.2.24)

• Manually seated valves should be returned to normal position prior to a thermal cycle on the valve.

• For Technical Specification required cooldown, manually seated MOVs SHALL be returned to normal position as soon as possible.

• Any safety-related MOV that is manually seated, SHALL be evaluated for operability by the Shift Manager/Unit Supervisor for entry into OAS.

• Declutching and manually closing an MOV is NOT the normal method of isolating equipment AND requires notification of the Field OR Unit Supervisor. This action will allow some MOVs to OPEN due to system differential pressure AND requires the valve to be manually unseated prior to energizing the MOV motor upon restoration.

• Whenever possible, the MOV should be de-energized prior to manual operation.

• Concurrent manual and remote operation of an MOV SHALL NEVER be attempted.

1. Motor Operated Valves (MOV) Manipulation:

1.1 MOVs should normally be opened or closed using the motor handswitch. IF the MOV has a handswitch in the Control Room, THEN the Control Room handswitch should be used for valve positioning.

1.2 Any time an MOV must be declutched and manually closed for ECOs (to stop system leakage),

the Field or Unit Supervisor will be informed and the applicable ECO noted. This is to ensure the MOV is manually unseated on ECO restoration prior to energizing the motor. Also see Chapter 9 of Conduct of Operations Manual for MOV manipulation.

2. The backseat of a valve may be used as a BOUNDARY for maintenance provided that:

2.1 The applicable WORK DOCUMENT allows for maintenance on the valve on its backseat.

2.2 The applicable Unit Operations Manager and the applicable Maintenance Division Manager or General Maintenance Supervisor have authorized performance of maintenance on the valve on the backseat. This authorization MAY be delivered verbally to the Shift Manager.

2.3 This authorization, including the date, time and method (phone, email, etc.), SHALL be documented in the ECO Notes.

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1627 Last used on an NRC exam: 2007 SRO Sequence Number: 78 Unit 2 is in Mode 6 when the following occurs:

CNTMT NORMAL SUMP LVL HI-HI alarm actuates CNTMT SEC NORM SUMP LVL HI-HI alarm actuates Personnel in the Fuel Handling Building report lowering level in the Spent Fuel Pool Based on this information, which of the following correctly identifies the location of the leak AND the procedure to be entered by the Unit Supervisor?

A. Spent Fuel Pool Cooling leak in the Fuel Handling Building; 0POP04-RC-0007, Mode 5 Or Mode 6 LOCA With The Reactor Vessel Head On B. Spent Fuel Pool Cooling leak in the Fuel Handling Building; 0POP04-FC-0002, Refueling LOCA C. Residual Heat Removal System leak in containment; 0POP04-RC-0007, Mode 5 Or Mode 6 LOCA With The Reactor Vessel Head On D. Residual Heat Removal System leak in containment; 0POP04-FC-0002, Refueling LOCA Answer: D Residual Heat Removal System leak in containment; POP04-FC-0002, Refueling LOCA Page 5 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1627 K/A Catalog Number: APE 025 AA2.03 Tier: 1 Group/Category: 1 SRO Importance: 3.8 10CFR Reference or SRO Objective: 55.43(b)(5)

Loss of RHR System: Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Increasing reactor building sump level 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:

LOT 505.01 - Instruction on 0POP04-FC-0002, Refueling LOCA Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT - Credible leak location because SFP level is lowering. Credible procedure because 0POP04-RC-0007 is used for a LOCA in Mode 6 (but only with the RX Head on). With level lowering in the SFP and levels rising in the RCB Sumps this would indicate that the RX Head is off and the SFP is tied to the RX Cavity through the Fuel Transfer Canal making 0POP04-FC-0002 the correct procedure.

B: INCORRECT - Credible leak location because SFP level is lowering.

C: INCORRECT - Credible because 0POP04-RC-0007 is used for a LOCA in Mode 6 but only with the RX Head on. With level lowering in the SFP and levels rising in the RCB Sumps this would indicate that the RX Head is off.

D: CORRECT - Sump alarms given are in the containment, SFP level is lowering because the transfer tube is open and the cavity flooded with the reactor head off. The correct procedure to use during a LOCA with the head off is POP04-FC-0002 Question Level: H Question Difficulty 3 Justification:

The applicant must determine that the leak is from RHR in containment (sump alarms) even though SFP level is lowering (which it will with an RHR leak while the transfer tube is open). The applicant must also have a knowledge of the entry conditions (Refueling LOCA vs Mode 5 or 6 LOCA) of the referenced procedures.

Page 6 of 50

0POP04-FC-0002 Refueling LOCA Rev. 13 Page 2 of 71 PURPOSE This procedure provides guidelines for protection of the Reactor Core, Fuel Assemblies, and Personnel in the event of a LOCA with the Reactor Vessel Head off of the vessel.

SYMPTOMS OR ENTRY CONDITIONS

1. Any of the following annunciator alarms are possible symptoms:

• "CNTMT NORM SUMP LVL HI-HI" Lampbox 5M03, Window D-8

• "CNTMT SEC NORM SUMP LVL HI-HI" Lampbox 5M03, Window E-8

• "SFP WATER LVL HI/LO" Lampbox 22M02, Window F-5

• "SFP TROUBLE" Lampbox 22M02, Window F-6

2. Any of the following local indications are entry conditions:

• SG Nozzle Dam failure.

• Rising Reactor Containment Building (RCB) radiation levels (RM-11, RM-23).

• Containment Ventilation Isolation (CVI) due to high radiation levels.

• Rising Spent Fuel Pool (SFP) area radiation monitor levels (RT-8090 or RT-8091).

• Fuel Handling Building Ventilation system shifted to Emergency Mode of operation due to high radiation levels.

• Reactor Cavity or SFP water level lowering in an uncontrolled manner.

This Procedure is Applicable Anytime the Reactor Head is Off the Vessel

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2154 Last used on an NRC exam: Never SRO Sequence Number: 79 Unit 1 was operating at 100% power when an event occurred that tripped the reactor and initiated a Safety Injection.

The crew is performing 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Based on the following conditions of the Steam Generators and Containment; Steam A B C D Generators Pressure 1095 psig 1085 psig 1090 psig 1010 psig Slowly Lowering Slowly Lowering Slowly Lowering Slowly Lowering Level 20% NR 19% NR 29% NR 31% NR Slowly Rising Slowly Rising Stable Slowly Lowering AFW Flow 150 gpm 150 gpm 50 gpm 50 gpm Containment Pressure 3.2 psig - Rising Temperature 130ºF - Rising Humidity 110ºF-dew point - Rising Which of the following procedures should the Unit Supervisor perform next?

A. 0POP05-EO-EO20, Faulted Steam Generator Isolation B. 0POP05-EO-EO30, Steam Generator Tube Rupture C. 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant D. 0POP05-EO-FRZ1, Response to High Containment Pressure Answer: C 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Page 7 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2154 K/A Catalog Number: EPE 009 EA2.11 Tier: 1 Group/Category: 1 SRO Importance: 4.1 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine or interpret the following as they apply to a Small Break LOCA: Containment temperature, pressure and humidity STP Lesson: LOT 504.05 Objective Number: 80474 From memory STATE/IDENTIFY how the RCS is checked to be intact per POP05-EO-EO00, Reactor Trip or Safety Injection.

Reference:

LOT 504.05 - Procedure training on 0POP05-EO-EO00, Reactor Trip or Safety Injection Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because SG D pressure is lower than the other three with its level is also lowering and the given containment conditions could be caused by a steam leak inside the RCB.

However this is normal for the given conditions due to the turbine driven aux feedwater pump that is connected to this steam generator.

B: INCORRECT: Credible because there are SG levels that are rising. However closer analysis of the given conditions reveals that the SGs with rising level are also being supplied with more AFW flow, which with no other indications should rule out tube leakage.

C: CORRECT: Containment conditions given are indicative of a SBLOCA or small steam break, however secondary indications do not support the steam break.

D: INCORRECT: Credible because containment pressure is elevated. However, it has not yet reached the level required for 0POP05-EO-FRZ1 entry.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given conditions to determine the event that has occurred and then determine which procedure is appropriate to transition to based on the event in progress.

Page 8 of 50

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Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2155 Last used on an NRC exam: Never SRO Sequence Number: 80 A reactor trip has occurred from 2% reactor power.

The crew is performing Step 1 of 0POP05-EO-ES01, Reactor Trip Response Source Range Detectors have not energized yet.

Subsequently a Reactor Operator reports the following; Extended Range Startup Rate is 0.1 DPM.

DRPI shows three control rods at 24 steps.

Which indication should the Unit Supervisor interpret as a priority condition and which procedure should be entered?

A. Extended Range Startup Rate is 0.1 DPM - Enter 0POP04-CV-0003, Emergency Boration B. Extended Range Startup Rate is 0.1 DPM - Enter 0POP05-FO-FRS1, Response to Nuclear Power Generation - ATWS C. DRPI indicates three control rods at 24 steps - Enter 0POP04-CV-0003, Emergency Boration D. DRPI indicates three control rods at 24 steps - Enter 0POP05-FO-FRS1, Response to Nuclear Power Generation - ATWS Answer: B Extended Range Startup Rate is 0.1 DPM - Enter 0POP05-FO-FRS1, Response to Nuclear Power Generation - ATWS Page 9 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2155 K/A Catalog Number: EPE 029 G2.1.7 Tier: 1 Group/Category: 1 SRO Importance: 4.7 10CFR Reference or SRO Objective: 55.43(b)(5)

ATWS: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavor and instrument interpretation.

STP Lesson: LOT 504.28 Objective Number: 84506 STATE the basis for monitoring the conditions listed in the subcriticality safety function status tree associated with 0POP05-EO-FRS1.

Reference:

LOT 504.28 Instruction on 0POP05-FO-FRS1, Response to Nuclear Power Generation -

ATWS Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because the POP04 procedure will add boron to the RCS, however within the EOP network, using the FRP is the correct response.

B: CORRECT: A positive start up rate would be a priority condition because it is a true indication of core reactivity and entering the critical safety function procedure would be the correct path to take.

0POP05-EO-FRS1 would be entered with RX power greater than 5E-6% (Source Range NIs not energized yet = RX power of at least 1E-5% or greater) and start up rate greater than 0.

C: INCORRECT: Credible because DRPI is indication of rod position which directly affects core reactivity. The indicated procedure is credible because the POP04 procedure will add boron to the RCS, however within the EOP network, using the FRP is the correct response.

D: INCORRECT: Credible because DRPI is indication of rod position which directly affects core reactivity.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given instrument response and enter the appropriate procedure for the condition.

Page 10 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2158 Last used on an NRC exam: Never SRO Sequence Number: 81 Unit 1 is at 100% Power and stable with all systems in a normal lineup.

An event occurs and an RO reports the following for Steam Generator levels:

Steam Generator A - 73% and rising.

Steam Generator B - 73% and rising.

Steam Generator C - 70% and stable.

Steam Generator D - 70% and stable.

Which of the following (1) describes a failure that could cause this event and (2) the procedure that has the appropriate actions to control Steam Generator levels?

A. (1) Loss of power to Feedwater Pump Discharge Pressure Transmitter PT-0558 (2) Enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, to manually control SGFP Master Speed Controller.

B. (1) Loss of power to Feedwater Pump Discharge Pressure Transmitter PT-0558 (2) Enter 0POP04-FW-0002, Steam Generator Feed Pump Trip, to manually control SGFP Master Speed Controller.

C. (1) Loss of power to Class 1E 120 VAC DP-1201 (2) Enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, to deselect affected channels controlling Steam Generator water level.

D. (1) Loss of power to Class 1E 120 VAC DP-1201 (2) Enter 0POP04-FW-0002, Steam Generator Feed Pump Trip, to deselect affected channels controlling Steam Generator water level.

Answer: C (1) Loss of power to Class 1E 120 VAC DP-1201 (2) Enter 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution, to deselect affected channels controlling Steam Generator water level.

Page 11 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2158 K/A Catalog Number: APE 057 G2.4.49 Tier: 1 Group/Category: 1 SRO Importance: 4.4 10CFR Reference or SRO Objective: 55.43(b)(5)

Loss of Vital AC Instrument Bus: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

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:

LOT 505.01 lesson on off normal procedure 0POP04-VA-0001, Loss of 120 VAC Class Vital Distribution.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible beacause a loss of power to PT-0558 would cause SG levels to rise, however all would rise. The procedure is credible because it contains steps for controlling SG levels, but would not be entered on a loss of power to PT-0558 (a non-class instrument).

B: INCORRECT: Credible beacause a loss of power to PT-0558 would cause SG levels to rise, however all would rise. The procedure is credible because it contains steps for responding to a malfunction of PT-0558.

C: CORRECT: With a normal lineup of SG level controls, a loss of DP-1201 would cause level to rise on SG A and B. 0POP04-VA-0001 has the required actions to conrol SG level in more than one SG and specifically has the extra action to deselect the affected channels.

D: INCORRECT: The procedure is credible because it contains steps for controlling SG levels under abnormal conditions (i.e. loss of PT-0557 or 0558, failure of master speed controller), but not this condition.

Question Level: H Question Difficulty 3 Justification:

The Unit Suprvisor has to evaluate the conditions given to determine a possible cause of the event and the correct procedure to use.

Page 12 of 50

0POP04-VA-0001 Loss Of 120 VAC Class Vital Rev. 28 Page 4 of 198 Distribution STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE x Steps 2.0 through 8.0 may be performed in any order.

x Foldout CIP should be opened.

_____ 1.0 CHECK Plant Mode - IN MODE 1, 2, OR 3 GO TO Step 4.0.

_____ 2.0 CHECK SG Levels - ALL LEVELS BEING IF using main feedwater to supply SGs, CONTROLLED ON PROGRAM LEVEL THEN PERFORM the following:

IN AUTOMATIC a. PLACE any SG feedwater regulating valve(s) or low power feedwater regulating valve(s) NOT properly responding in MANUAL to match feedwater flow to steam flow:

x SG 1A(2A) NORM FCV-0551 x SG 1B(2B) NORM FCV-0552 x SG 1C(2C) NORM FCV-0553 x SG 1D(2D) NORM FCV-0554 x SG 1A(2A) LOW PWR FV-7151 x SG 1B(2B) LOW PWR FV-7152 x SG 1C(2C) LOW PWR FV-7153 x SG 1D(2D) LOW PWR FV-7154

b. ADJUST affected SG feedwater regulating valve(s) as necessary to restore affected SG NR level(s) to between 68% and 74%.

c. IF Feedwater Flow Transmitter FT-0510, FT-0520, FT-0530 or FT-0540 indicates channel failure, THEN PERFORM the following:

1) ENSURE DA Storage Tank Level Control LK-7406 in MANUAL.

2) MAINTAIN DA Storage Tank level between 65% and 80.

Step 2.0 continued on next page This Procedure Is Applicable At All Times

0POP04-VA-0001 Loss Of 120 VAC Class Vital Rev. 28 Page 5 of 198 Distribution STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED Step 2.0 continued from previous page

d. ENSURE an operable channel selected for affected feedwater regulating valve(s):

x feed flow x steam flow x SG level

e. PLACE affected SG feedwater regulating valve(s) in AUTO AND VERIFY proper operation.

f. IF SG NR level can NOT be maintained between 20% and 87.5%, THEN PERFORM the following:

1) TRIP the Reactor.

2) PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection.

_____ 3.0 CHECK SGFP Master Speed Controller - IF using main feedwater to supply SGs, RESPONDING PROPERLY IN THEN PERFORM the following:

AUTOMATIC a. PLACE SGFP master speed controller in MANUAL AND ADJUST as necessary to maintain Feedwater/Steam DP per Addendum 1.

b. IF additional feed flow is necessary, THEN:

1) ENSURE SU SGFP is running.

2) START a Standby FW Booster Pump.

c. IF SG NR level can NOT be maintained between 20% and 87.5%, THEN PERFORM the following:

1) TRIP the Reactor.

2) PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection.

This Procedure Is Applicable At All Times

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2159 Last used on an NRC exam: Never SRO Sequence Number: 82 Unit 1 is operating at 100% power when the CCW SURGE TK LVL LO annunciator alarms.

Given the following conditions:

The Crew has entered 0POP04-CC-0001, Component Cooling Water System Leak, due to the lowering Component Cooling Water Surge Tank level.

Plant Operators in the field have NOT identified the source of the leak.

Component Cooling Water Non-Vital Supply Valves closed as required at 64.6% level.

Component Cooling Water Surge Tank continued to lower to 60% and is now stable.

Which of the following (1) identifies a possible leak location and (2) the correct action the Unit Supervisor should take based on the EXISTING plant conditions?

A. (1) A leak in the Letdown Heat Exchanger.

(2) Isolate letdown and enter 0POP04-CV-0004, Loss of Normal Letdown.

B. (1) A leak in RCP C Motor Air Cooler.

(2) Trip the Reactor, Turbine, and Reactor Coolant Pumps, then enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

C. (1) A leak in RCP C Motor Air Cooler.

(2) Enter 0POP04-TM-0005, Rapid Load Reduction and reduce power to less than P8 (40%) to allow tripping RCP C D. (1) A leak in the Letdown Heat Exchanger.

(2) Isolate letdown and enter 0POP04-RC-0008, Boron Dilution Event.

Answer: B (1) A leak in RCP C Motor Air Cooler.

(2) Trip the Reactor, Turbine, and Reactor Coolant Pumps, then enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Page 13 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2159 K/A Catalog Number: APE 026 AA2.01 Tier: 1 Group/Category: 1 SRO Importance: 3.5 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water:

Location of leak in the CCWS STP Lesson: LOT 500.01 Objective Number: 92108 Given a plant condition, STATE the actions required to be performed per the applicable Off-Normal procedure.

Reference:

0POP04-CC-0001, Loss of Component Cooling Water, pages 4, 9, and 10 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Leak location is credible because CCW supplies cooling to the letdown heat exchanger, but not correct since the surge tank level continued to lower after the first isolation.

Action is credible since a loss of cooling to the heat exchanger would necessitate removing letdown from service.

B: CORRECT: A leak on an RCP would still be occurring after the first CCW isolation (64.6%), but stop after the second isolation (61.5%). The second isolation will isolate CCW to all RCPs so manual trip would be required.

C: INCORRECT: Action is credible because P8 allows for three RCP operation less than 40% power (although plant procedures do not). Also, action is not correct because plant conditions affects all RCPs and not just RCP C.

D: INCORRECT: Leak location is credible because CCW supplies cooling to the letdown heat exchanger, but not correct since the surge tank level continued to lower after the first isolation.

Action is credible because a tube leak in other CVCS heat exchangers (i.e. seal water heat exchanger) would cause a dilution of the RCS, however this procedure would not be used in Mode 1 Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given conditions and then determine which procedure would be appropriate to perform.

Page 14 of 50

0POP04-CC-0001 Component Cooling Water System Rev. 14 Page 4 of 47 Leak STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 2.0 M ONIT OR For CCW Surge Tank Low GO TO Step 8.0.

Level Non-Vital Supply Valves Isolation:

• CCW Surge Tank Level - LESS THAN 64.6%

OR

• CCW Surge Tank Level - HAS BEEN LESS THAN 64.6%

Plant Computer Points

• Train A - CCLE4504

• Train B - CCLE4506

• Train C - CCLE4508

_____ a. CHECK the following valves are closed: a. Manually CLOSE the valves.

• "NNS LOADS INL ISOL MOV-0235" (CCW to common supply header isolation)

• "NNS LOADS INL ISOL MOV-0236" (CCW to common supply header isolation)

• "BRANCH ISOL MOV-0297" (CCW to RCDT Hx and Excess LD Hx isolation)

• "RCDT HX 1A(2A) INL MOV-0392" (CCW to RCDT Hx isolation)

• "EXCESS LETDOWN HX 1A(2A)

INL MOV-0393" (CCW to Excess LD isol)

This Procedure is Applicable anytime a CCW Pump is Operating.

0POP04-CC-0001 Component Cooling Water System Rev. 14 Page 8 of 47 Leak STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 10.0 M ONIT OR For CCW Surge Tank Low GO TO Step 24.0.

Level Commom Header Valves Isolation:

• CCW Surge Tank Level - LESS THAN 61.5%

OR

• CCW Surge Tank Level - HAS BEEN LESS THAN 61.5%

Plant Computer Points

• Train A - CCLE4504

• Train B - CCLE4506

• Train C - CCLE4508 Step 10.0 continued on next page This Procedure is Applicable anytime a CCW Pump is Operating.

0POP04-CC-0001 Component Cooling Water System Rev. 14 Page 9 of 47 Leak STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED Step 10.0 continued from previous page

_____ a. CHECK the following valves are Manually CLOSE the valves.

closed:

Charging System Components

• "SUPPLY ISOL MOV-0768"(CCW Train A isol to charging pump header)

• "RET ISOL MOV-0772"(CCW return header isolation from charging pumps to Train A)

• "SUPPLY X-CONN FV-4656"(CCW to charging pump A crosstie)

• "RETURN X-CONN FV-4657"(CCW return from charging pump A crosstie)

CCW Header Isolations Train A

• "CCW SPLY HDR ISOL" "MOV-0316"

• "CCW RET HDR ISOL" "MOV-0052" Train B

• "CCW SPLY HDR ISOL" "MOV-0314"

• "CCW RET HDR ISOL" "MOV-0132" Train C

• "CCW SPLY HDR ISOL" "MOV-0312"

• "CCW RET HDR ISOL" "MOV-0192" This Procedure is Applicable anytime a CCW Pump is Operating.

0POP04-CC-0001 Component Cooling Water System Rev. 14 Page 10 of 47 Leak STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 11.0 CHECK Plant MODE - MODE 1 or 2 ENSURE RCPs secured.

_____ a. TRIP the Reactor

_____ b. SECURE RCPs

_____ c. PERFORM 0POP05-EO-EO00, Reactor Trip Or Safety Injection

_____ d. CONTINUE Actions Of This Procedure As Resources Permit CAUTION

• A loss of CCW to CCP Supplemental Cooler may cause respective CCP motor failure in as little as 4 minutes.

• A loss of CCW to CCP Lube Oil Cooler may cause pump failure in as little as 8 minutes.

_____ 12.0 CHECK Charging System Status - ANY GO TO Step 17.0.

PUMP RUNNING

_____ 13.0 CHECK Centrifugal Charging GO TO Step 15.0.

Pump 1A(2A) Status - RUNNING This Procedure is Applicable anytime a CCW Pump is Operating.

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2218 Last used on an NRC exam: Never SRO Sequence Number: 83 Given the following:

Unit 1 is in Mode 5 RCS Boron Concentration is currently at 2810 ppm Control Rods are locked in the refueling position Preparations are being made for refueling operations Reactor Trip breakers are closed for maintenance and testing Rod Drive MG Sets are running for maintenance and testing Subsequently:

A fire breaks out in the Train C ESF Switchgear room.

The Fire Brigade Leader reports a concern that the fire appears to be spreading toward the Rod Drive MG Set Room which could threaten the ability to move the Control Rods.

Based on the given information, which of the following describes the actions the Unit Supervisor should perform?

A. Direct I&C to restore Control Rod operation per 0PMP07-DM-0003, Rapid Refueling Rod Holdout Operation, and then ensure all Control Rods are fully inserted.

B. Enter 0POP04-RS-0001, Control Rod Malfunction, open the Reactor Trip Breakers and then ensure all Control Rods are fully inserted.

C. Direct a Plant Operator to secure the Rod Drive MG Sets per 0PMP07-DM-0003, Rapid Refueling Rod Holdout Operation, and then ensure all Control Rods are fully inserted.

D. Enter 0POP04-RS-0001, Control Rod Malfunction, place all Control Rod Lift Coil Disconnect Switches in the DISCONNECT position and then ensure all Control Rods are fully inserted.

Answer: A Direct I&C to restore Control Rod operation per 0PMP07-DM-0003, Rapid Refueling Rod Holdout Operation, and then ensure all Control Rods are fully inserted.

Page 15 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2218 K/A Catalog Number: APE 067 AA2.16 Tier: 1 Group/Category: 2 SRO Importance: 4.0 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to determine and interpret the following as they apply to the Plant Fire on Site:

Vital equipment and control systems to be maintained and operated during a fire.

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 Instruction on 0POP04-ZO-0008, Fire/Explosion Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: 0PMP07-DM-0003 is the only procedure used that will lock or unlock the control rods for rapid refueling. Restoring the operation of the control rods from a locked out rapid refueling position, would allow the Unit Suprvisor to direct the control rods be fully inserted or dropped into the reactor core.

B: INCORRECT: Credible because POP04-RS-0001 directs recovery from control rod malfunctions.

Under normal circumstances, opening the trip breakers would cause control rods to fall, but not with the given conditions.

C: INCORRECT: Credible because 0PMP07-DM-0003 will provide instructions on how to make rods capable of insertion under these conditions, however securing the MG sets is not one of them and will not work with the conditions given.

D: INCORRECT: Credible because POP04-RS-0001 directs recovery from control rod malfunctions and provides instruction for operation of the lift coil disconnect switches, but for other reasons (recovery of a dropped rod).

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate plant conditions and then select the appropriate procedure. Also, there must be a knowledge of how the Control Rods operate normally and when placed in the Rapid Refueling position.

Page 16 of 50

0POP04-ZO-0008 Fire/Explosion Rev. 21 Page 9 of 77 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 16.0 PERFORM 0POP04-ZO-0004, Personnel Emergencies, Concurrently With This Procedure

_____ 17.0 MONITOR With Fire Brigade Leader GO TO Step 19.0.

For Smoke Affecting Adjacent Buildings

_____ 18.0 DETERMINE If Any Of The Following Actions Are Required For Smoke Affecting Adjacent Buildings:

x Evacuate affected building(s) x SECURE OR REALIGN HVAC in affected building(s) x VENTILATE affected building(s)

_____ 19.0 CHECK The Following Conditions exist: GO TO Step 21.0.

x Control Rods are locked in the rapid refueling position x The Reactor Vessel head is - ON x The fire may degrade the ability to move the control rods This Procedure is Applicable in All Plant Conditions

0POP04-ZO-0008 Fire/Explosion Rev. 21 Page 10 of 77 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

___ 20.0 PERFORM The Following:

_____ a. DIRECT I&C to restore control rod operation per 0PMP07-DM-0003, Rapid Refueling Rod Holdout Operation

_____ b. Fully INSERT all Control Rods

_____ 21.0 CHECK The Fire Alarm Indicates A GO TO Step 23.0.

Possible RCB Cable Tray Fire

_____ 22.0 PERFORM, Addendum 1, Response To RCB Cable Tray Alarms

_____ 23.0 MONITOR A Valid Hi Temperature CONTINUE at Step 25.0 while monitoring Alarm Received For Any Of The filter unit temperature alarms AND Following Filter Units: PERFORM Step 24.0 when any valid filter unit temperature alarm is received.

x Containment Carbon Units x FHB Exhaust Filter Units x EAB Filter Units x MAB Filter Units This Procedure is Applicable in All Plant Conditions

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2161 Last used on an NRC exam: Never SRO Sequence Number: 84 Unit 1 is operating at 100% Power.

The crew has implemented 0POP04-RC-0004, Steam Generator Tube Leakage, due to the following current Radiation Monitor Readings given to the Unit Supervisor for the Steam Generators.

Steam A B C D Generators Steam Line 1.8E-2 uCi/cc 1.5E-2 uCi/cc 1.4E-2 uCi/cc 3.9E-1 uCi/cc Radiation Blowdown 3.1E-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 75 gpd.

Leakage rate is rising 4 gpd/hr.

Which Steam Generator(s) have tube leaks and what action will the Unit Supervisor perform?

A. Only Steam Generator D has a tube leak - Place the Unit in MODE 3 in less than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using 0POP03-ZG-0006, Plant Shutdown From 100%.

B. Only Steam Generator D has a tube leak - Reduce power until the turbine is tripped using 0POP04-TM-0005, Fast Load Reduction.

C. Steam Generators A and D have tube leaks - Reduce power until the turbine is tripped using 0POP04-TM-0005, Fast Load Reduction.

D. Steam Generators A and D have tube leaks - Place the Unit in MODE 3 in less than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using 0POP03-ZG-0006, Plant Shutdown From 100%.

Answer: A Only Steam Generator D has a tube leak; Place the Unit in MODE 3 in less than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using 0POP03-ZG-0006, Plant Shutdown From 100%.

Page 17 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2161 K/A Catalog Number: APE 037 AA2.02 Tier: 1 Group/Category: 2 SRO Importance: 3.9 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:

Agreement/disagreement among redundant radiation monitors 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-RC-0004, Steam Generator Tube Leak.

Attached Reference

Attachment:

0POP04-RC-0004, Steam Generator Tube Leak, Page 5 (Step

8) and Page 34 (Add 6)

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: The elevated leak rate from the N-16 monitor for SG 'A' is due to it being close to SG

'D' steam line. The N-16 monitors are in the TGB where there is no concrete to shield in between the steam lines. With the Steam Line and Blowdown Rad monitors shielded in the IVC and not being elevated on all but SG 'D', the comparison would indicate that only SG 'D' had a tube leak.

With one SG leaking greater than 75 gpd with a rate of change less than 30 gpd/hr, the correct action is to use 0POP03-ZG-0006 and be in MODE 3 in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B: INCORRECT: Credible because this action would be correct if the leakrate was escalating faster

(>= 30 gpd/hr).

C: INCORRECT: Credible because SG A N-16 reading is considerable larger than B or C, but this is due to shine from the SG D steam line. Credible action because this action would be correct if the leakrate was escalating faster (>= 30 gpd/hr)

D: INCORRECT: Credible because SG A N-16 reading is considerable larger than B or C, but this is due to shine from the SG D steam line.The D steam line is on the outside 'A' is only one adjacent to

'D'.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given conditions and then select the appropriate procedure.

Page 18 of 50

0POP04-RC-0004 Steam Generator Tube Leakage Rev. 29 Page 5 of 116 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 6.0 MAINTAIN VCT Level - GREATER PERFORM the following:

THAN 15% WITH CHARGING PUMP a. TRIP the Reactor.

SUCTION ALIGNED TO VCT (CP004)

b. INITIATE Safety Injection.

x Auto makeup

c. GO TO 0POP05-EO-EO00, Reactor Trip x Manual makeup Or Safety Injection.

_____ 7.0 CHECK Main Turbine In Service GO TO Step 13.0.

_____ 8.0 CHECK For One Of The Following: PERFORM the following:

x Leakage From Any One SG Is a. COMPARE SG Tube Leak Rates to values GREATER THAN OR EQUAL TO 75 listed in Addendum 6.

gpd AND Continues To Increase At b. IF Mode 3 is REQUIRED by GREATER THAN OR EQUAL TO 30 Addendum 6, THEN PERFORM the gpd/hr following:

OR 1) COMMENCE plant shut down per x Leakage From Any One SG Is 0POP03-ZG-0006, Plant Shutdown GREATER THAN OR EQUAL TO 75 From 100% To Hot Standby, per gpd AND Loss of Continuous response time requirements of Radiation Monitoring Addendum 6.

2) GO TO Step 10.0.

c. IF Shutdown NOT required, THEN RETURN to procedure and step in effect.

_____ 9.0 PERFORM The Following:

_____ a. COMMENCE Plant Shutdown Per 0POP04-TM-0005, Fast Load Reduction Per The Response Time Requirements Of Addendum 6

_____ b. CONTINUE performance of 0POP04-TM-0005, Fast Load Reduction until TURBINE TRIPPED This Procedure is Applicable in Modes 1, 2, 3 and 4

0POP04-RC-0004 Steam Generator Tube Leakage Rev. 29 Page 34 of 116 Addendum 6 Recommended Response Times Addendum 6 Page 1 of 1 CAUTION Post Reactor Shutdown conditions in the primary to secondary leakage (RCS temperature and pressure decreasing and SG pressure increasing) may reduce the SG Tube Leakage Rate. Plant Shutdown and Cooldown Rates should be based on the initial or increased leakage and NOT reduced leakage due to the Post Shutdown conditions.

Action Leak Rate Increasing Response Times (1)

Level Leak Rate Reduce Rx PWR to

> 75 gpd Rate of inc > 30 gpd/hr < 50% in 1 hr 3 (4) AND Mode 3 in the next 2 hr

> 75 gpd Reduce Rx PWR to AND < 50% in 1 hr Loss of Continuous Radiation NA AND 3 Monitoring Mode 3 in the next 2 hr (3) 3 > 150 gpd NA Mode 3 < 6 hr

> 75 gpd 2 Mode 3 < 24 hr (4) Rate of inc < 30 gpd/hr

< 75 gpd AND Loss of Continuous Radiation Continued Operations N/A 1 Monitoring (2)

(3)

Continued Operations 1 > 30 gpd N/A (2)

Increased Continued Operations

> 5 gpd N/A Monitoring (2)

(1) Response times are the maximum times allowed. Power Reduction and Mode Change(s) may be completed in less time.

(2) Continued Operations per Plant Management direction. Refer to 0PGP03-ZO-0041, Action For Monitoring Primary to Secondary Leakage.

(3) Loss of Continuous Radiation Monitoring as defined in 0PGP03-ZO-0041, Action for Monitoring Primary to Secondary Leakage.

(4) With a continued increase in leakage rate over 30 minute time interval per the next column.

This Procedure is Applicable in Modes 1, 2, 3 and 4

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2162 Last used on an NRC exam: Never SRO Sequence Number: 85 Unit 2 is operating at 60% Power.

Subsequently:

A Reactor Operator reports that Condenser Vacuum is 23 HG and stable and Main Generator Megawatt output is slowly lowering.

Further investigation reveals that the MAIN COND VACUUM LO alarm is EXTINGUISHED and the C9 COND AVAILABLE FOR STEAM DUMP light is ILLUMINATED.

Based on the given plant conditions, which of the following describes actions that the Unit Supervisor should perform?

A. (1) Trip the Reactor, Ensure the Main Turbine tripped and enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

(2) Write a Condition Report on the MAIN COND VACUUM LO alarm which is NOT responding properly.

B. (1) Trip the Reactor, Ensure the Main Turbine tripped and enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

(2) Write a Condition Report on the MAIN COND VACUUM LO alarm and the C9 COND AVAILABLE FOR STEAM DUMP light. Both are NOT responding properly.

C. (1) Enter 0POP04-CR-0001, Loss of Condenser Vacuum, and begin a Turbine Load Reduction.

(2) Write a Condition Report on the MAIN COND VACUUM LO alarm and the C9 COND AVAILABLE FOR STEAM DUMP light. Both are NOT responding properly.

D. (1) Enter 0POP04-CR-0001, Loss of Condenser Vacuum, and begin a Turbine Load Reduction.

(2) Write a Condition Report on the MAIN COND VACUUM LO alarm which is NOT responding properly.

Answer: D (1) Enter 0POP04-CR-0001, Loss of Condenser Vacuum, and begin a Turbine Load Reduction.

(2) Write a Condition Report on the 'MAIN COND VACUUM LO' alarm which is NOT responding properly.

Page 19 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2162 K/A Catalog Number: APE 051 G2.4.46 Tier: 1 Group/Category: 2 SRO Importance: 4.2 10CFR Reference or SRO Objective: 55.43(b)(5)

Loss of Condenser Vacuum: Ability to verify that the alarms are consistent with the plant conditions.

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-CR-0001, Loss of Condenser Vacuum Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because conditions can exist where a manual trip is required, but not until 21 inches of vacuum.

B: INCORRECT: Credible because conditions can exist where a manual trip is required, but not until 21 inches of vacuum. Part 2 is credible because both the alarm and the light are dependent on vacuum, however the setpoint for the light is 22 inches.

C: INCORRECT: Part 2 is credible because both the alarm and the light are dependent on vacuum, however the setpoint for the light is 22 inches.

D: CORRECT: Condenser vacuum is above 21" HG but below 26" HG. A Turbine Load reduction is required. The "MAIN COND VACUUM LO" alarm did not annunciate at 26" HG when it was supposed to.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given conditions and then determine the correct action to take.

Page 20 of 50

0POP04-CR-0001 Loss of Condenser Vacuum Rev. 17 Page 2 of 30 PURPOSE This procedure provides guidelines for operator response to a lowering or reduced Main Condenser vacuum.

The procedure has two levels of response. The first response is a diagnostic process to identify and correct the cause of the loss of vacuum. The second is to monitor condenser vacuum and ensure automatic actuations occur at the required setpoint.

Condenser Vacuum Automatic Actuation (in. Hg)

(UNIT 1 ONLY) Condenser Vacuum Low alarm

> 80% Power 24 (UNIT 1 ONLY)

< 80% Power Condenser Vacuum Low alarm 26 (UNIT 2 ONLY)

Condenser Vacuum Low alarm 26 26 Standby Condenser Air Removal Pump starts 22 C-9, Steam Dump Block Permissive 21 Main Turbine Trip SYMPTOMS OR ENTRY CONDITIONS

1. The following Control Room annunciator alarm:

x "MAIN COND VACUUM LO" Lampbox 07M3 Window E-7 x "CWP TRIP/FAIL START" Lampbox 09M1 Window A-3

2. Automatic starting of the standby Condenser Air Removal (CARs) pump.

3. Unexplained continuing lowering in Main Turbine Load.

4. Lowering Main Condenser vacuum.

This Procedure is Applicable anytime a vacuum exists in the Main Condenser

0POP04-CR-0001 Loss of Condenser Vacuum Rev. 17 Page 12 of 30 Addendum 2 Main Turbine Exhaust Pressure Limitations Addendum 2 Page 1 of 2 Unit 1 Exhaust Pressure Limitations Maximum Permissible Condensing Pressure This Procedure is Applicable anytime a vacuum exists in the Main Condenser

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2216 Last used on an NRC exam: Never SRO Sequence Number: 86 Instrument air has been lost to CV-FV-0011.

Which of the following correctly describes the action to be taken by the control room staff and why?

The Unit Supervisor should enter A. 0POP04-RP-0002, Loss Of Automatic Pressurizer Level Control, and close FCV-0205 to minimize thermal stress on the charging nozzle at the RCS pipe.

B. 0POP04-CV-0004, Loss of Normal Letdown, and close FCV-0205 to minimize thermal stress on the charging nozzle at the RCS pipe.

C. 0POP04-RP-0002, Loss Of Automatic Pressurizer Level Control, and close FCV-0205 to maintain VCT level and prevent a loss of suction to the charging pump.

D. 0POP04-CV-0004, Loss of Normal Letdown, and close FCV-0205 to maintain VCT level and prevent a loss of suction to the charging pump.

Answer: B 0POP04-CV-0004, Loss of Normal Letdown, and close FCV-0205 to minimize thermal stress on the charging nozzle at the RCS pipe.

Page 21 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2216 K/A Catalog Number: APE 065 G2.1.32 Tier: 1 Group/Category: 1 SRO Importance: 4.0 10CFR Reference or SRO Objective: 55.43(b)(5)

Loss of Instrument Air: Ability to explain and apply system limits and precautions.

STP Lesson: LOT 201.06 Objective Number: 48669 In regard to POP02-CV-0004, POP02-CV-0005, AND PSP03-CV-0011, DESCRIBE the following: 1.

Purpose and Scope, 2. Precautions, and 3. Notes and Cautions

Reference:

POP04-CV-0004, step 2.0 basis, POP02-CV-0004, step 4.14 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Procedure is credible because pressurizer level will rise during this event as it will during a low failure of the controlling level channel which POP04-RP-0002 is designed to address; therefore it would not be a correct entry in this instance.

B: CORRECT: Loss of IA to the valve will cause it to close and result in letdown flow going to 0.

POP04-CV-0004 will address this condition. If letdown flow is lost, then charging should be isolated to minimize thermal stress on the charging nozzle since preheating is no longer occurring.

C: INCORRECT: Procedure is credible because pressurizer level will rise during this event as it will during a low failure of the controlling level channel which POP04-RP-0002 is designed to address; therefore it would not be a correct entry in this instance. Reason is credible because this failure will cause VCT level to lower and closing FCV-0205 will minimize that, however loss of suction to the charging pump is not a concern due to auto makeup and low level swapover to the RWST.

D: INCORRECT: Reason is credible because this failure will cause VCT level to lower and closing FCV-0205 will minimize that, however loss of suction to the charging pump is not a concern due to auto makeup and low level swapover to the RWST.

Question Level: H Question Difficulty 3 Justification:

The applicant must determine the effect the given failure will have and select the proper procedure. A knowledge of system design/limits is needed to deterrmine the reason for the action taken.

Page 22 of 50

0POP04-CV-0004 Loss of Normal Letdown Rev. 11 Page 2 of 59 PURPOSE This procedure provides guidance for Operator response due to a loss of Normal Letdown.

This procedure provides guidance for recovery actions including response due to an instrument failure, letdown line failure or a High Energy Line Break (HELB). The following is a list of valves in the letdown path and possible causes of closure.

Loss of signal Pzr Level Containment High CVCS Loss of from Letdown

< 17% Isol Phase A Room Temp Power/Air Pressure PT-0135 LCV-0465 LCV-0468 FV-0011 FV-0012 FV-0013 MOV-0014 MOV-0023 MOV-0024 PCV-0135 SYMPTOMS OR ENTRY CONDITIONS

1. Any of the following Control Room annunciator alarms:

• "LETDN HX OUTL PRESS HI" Lampbox 4M08, Window C-4

• "LETDN HX OUTL FLOW HI/LO" Lampbox 4M08, Window D-4

2. The following symptoms are indicative of a loss of Normal Letdown

• Low Letdown flow as indicated on FI-0132

• HELB Actuation as indicated by closure of MOV-0023 or MOV-0024, letdown containment isolation valves

3. This procedure is entered any time Normal Letdown is isolated unexpectedly. This procedure is NOT entered if loss of letdown is due to a Safety Injection.

This Procedure is Applicable in Modes 1-4

0POP04-CV-0004 Loss of Normal Letdown Rev. 11 Page 29 of 59 Addendum 7 Basis Basis Page 2 of 31 STEP DESCRIPTION FOR 0POP04-CV-0004 STEP 2.0 STEP: VERIFY Letdown Containment Isolation Valves - OPEN

• ICIV MOV-0023

• OCIV MOV-0024 PURPOSE: To determine if either of these valves have closed and take additional action if either valve has closed.

BASIS: These valves close to isolate a high energy line break in the MAB resulting from letdown line breaks. If either of these valves have closed additional upstream valves must be closed manually to prevent a loss of coolant through PSV-3100, which is between the letdown orifices and the containment isolation valves. (Ref. P&ID 9F05005)

ACTIONS: If either valve has closed, manually close FV-0011, which ensures that letdown flow is isolated. The charging flow control valve (FCV-0205) is closed, the miniflow recirc valve is opened for the operating charging pump and RCP seal injection valve (HCV-0218) is adjusted to maintain RCP seal injection within the specified band. The orifice valves are closed, followed by LCV-0465 and 0468. It is necessary to close FV-0011 following the closure of MOV-0023 and MOV-0024 to minimize the amount of time the letdown relief valve is challenged.

INSTRUMENTATION: N/A CONTROL/EQUIPMENT: Switches for FV-0011, MOV-0014, FV-0013, FV-0012, LCV-465, LCV-468, FCV-0205, HCV-0218 on CP004.

KNOWLEDGE: Part of isolating letdown is also to isolate charging to minimize thermal shock at the charging nozzle at the RCS pipe (Ref. PLS 5Z010ZS1101, page 47). Before normal letdown is restored, an evaluation should be made to verify it is feasible, i.e., any leakage points isolated do not prevent letdown from being restored.

This Procedure is Applicable in Modes 1-4

0POP04-CV-0004 Loss of Normal Letdown Rev. 11 Page 33 of 59 Addendum 7 Basis Basis Page 6 of 31 STEP DESCRIPTION FOR 0POP04-CV-0004 STEP 6.0 STEP: VERIFY Letdown Orifice Header Isolation Valve LETDN ORIF HDR ISOL FV-0011 - OPEN PURPOSE: Determine if a malfunction has occurred with valve FV-0011.

BASIS: Valve FV-0011 closes on low pressurizer level or a Containment Phase A isolation signal (SSPS Train C). Both of these conditions have been checked in previous steps. The other possible causes for FV-0011 to close is a loss of instrument air or loss of electrical power. FV-0011 may also close on a temporary loss of power to LT-0465 or LT-0468, which may not close LCV-0465 or LCV-0468. This step determines if FV-0011 is open or closed and investigates for possible causes if the valve has closed.

ACTIONS: The determination is made if FV-0011 is open or closed. If the valve is closed instrument air is checked and the power supply to the valve is checked. If the source of the problem is identified and corrected, the letdown orifices are closed, and then letdown can be restored per Addendum 4, Placing Normal Letdown In Service.

INSTRUMENTATION: N/A CONTROL/EQUIPMENT: Switches for MOV-0014, FV-0013, FV-0012, PK-0135, FV-0011 on CP004.

KNOWLEDGE: FV-0011 fails closed on loss of air or loss of power to the solenoid.

This Procedure is Applicable in Modes 1-4

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2182 Last used on an NRC exam: Never SRO Sequence Number: 87 Unit 1 was at 100% power when an event occurred that required a fast load reduction to 80%

power. During the power reduction, 2 rods in Control Bank D failed to move while the remainder of the rods in the bank inserted 18 steps.

Which of the following describes the ACTION the Unit Supervisor should now take and the EFFECTS this condition may have on Unit 1?

ACTION EFFECTS Attempt to re-align the Control Rods per Fuel Integrity may be A. 0POP04-RS-0001, Control Rod Malfunction, challenged by Power Peaking Addendum 2, Recovery of Misaligned Rods. Factor limits being exceeded.

Place Unit 1 in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per Fuel Integrity may be B. 0POP03-ZG-0006, Plant Shutdown from challenged by DNBR limits 100%. being exceeded.

Attempt to re-align the Control Rods per Fuel Integrity may be C. 0POP04-RS-0001, Control Rod Malfunction, challenged by DNBR limits Addendum 2, Recovery of Misaligned Rods. being exceeded.

Place Unit 1 in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per Fuel Integrity may be D. 0POP03-ZG-0006, Plant Shutdown from challenged by Power Peaking 100%. Factor limits being exceeded.

Answer: D Place Unit 1 in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per 0POP03-ZG-0006, Plant Shutdown from 100%; Fuel Integrity may be challenged by Power Peaking Factor limits being exceeded.

Page 23 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2182 K/A Catalog Number: APE 005 G2.1.43 Tier: 1 Group/Category: 2 SRO Importance: 4.6 10CFR Reference or SRO Objective: 55.43(b)(5)

Inoperable/Stuck Control Rod: Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant temperature, secondary plant, fuel depletion, etc.

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-RS-0001, Control Rod Malfunction and TS 3.1.3.1 Action d.

Attached Reference

Attachment:

0POP04-RS-0001, pages 1-10 NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Action is credible because the procedure contains actions to align rods, but only if there is only one rod affected.

B: INCORRECT: Effect is credible because DNBR limits are often a concern when power is affected, however the limits are not expected to be exceeded with misaligned control rods.

C: INCORRECT: Action is credible because the procedure contains actions to align rods, but only if there is only one rod affected. Effect is credible because DNBR limits are often a concern when power is affected, however the limits are not expected to be exceeded with misaligned control rods.

D: CORRECT: With 2 rods misaligned a plant shutdown is required. Fuel Integrity is challenged by Power Peaking Factors possibly being exceeded.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor must evaluate the given condition and choose the correct procedure and identify the effect on the Unit caused by the condition.

Page 24 of 50

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 5 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 4.0 CHECK For Misaligned Rods:

_____ a. CHECK All Rods - ANY RODS a. GO TO Step 5.0.

MISALIGNED

_____ b. CHECK All Rods - ONLY ONE ROD b. IF two or more rods are NOT aligned in MISALIGNED MODES 1 OR 2, THEN:

1) REFER TO Technical Specification 3.1.3.1 Action d for appropriate action.

2) COMMENCE load reduction per 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, to place the unit in Mode 3 within six hours of the time of misalignment.

3) GO TO Step 5.0.

_____ c. GO TO Addendum 2, Recovery of Misaligned Rods

_____ 5.0 CHECK Reactor Trip Breakers - CLOSED IF all rods have NOT fully inserted following a Reactor Trip or shutdown, THEN GO TO Addendum 3, Insertion of Rods Which Fail to Fully Insert Following Reactor Trip or Shutdown.

This Procedure is Applicable in Modes 1, 2, and 3

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2211 Last used on an NRC exam: Never SRO Sequence Number: 88 A large leak of radioactive water is occurring in the Mechanical Auxiliary Building.

The following sumps are experiencing a rise in level due to the leak.

1. Component Cooling Water Sump

2. Mechanical Auxiliary Building Floor Drain Sump #2

3. Essential Cooling Water Sump

4. Mechanical Auxiliary Building Elevator Sump Under these conditions securing the sump pumps for which of the above sumps would prevent an unmonitored release directly to the environment?

A. 1 B. 2 C. 3 D. 4 Answer: C 3 Page 25 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2211 K/A Catalog Number: G2.3.11 Tier: 3 Group/Category: 3 SRO Importance: 4.3 10CFR Reference or SRO Objective: 55.43(b)(4)

Ability to control radiation releases.

STP Lesson: LOT 203.10 Objective Number: 98076 Given a set of conditions, PREDICT the effect(s) and/or response(s) on the Equipment and Floor Drains system.

Reference:

LOT 203.10 lesson on Equipment and Floor Drains Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: This is a credible distractor because Component Cooling Water Sump could collect radioactive effluent but the sump is pumped to the Condensate Polisher Regeneration Waste Collection Tank (CPRWCT) and then to a WHT where the water is processed and monitored prior to release.

B: INCORRECT: This is a credible distractor because MAB Sump #2 could collect radioactive effluent but is pumped to the Floor Drain Tank and then to a WHT where the water is processed and monitored prior to release.

C: CORRECT: Essential Cooling Water Sump pumps directly to the enviroment through the Circ Water system without being monitored.

D: INCORRECT: This is a credible distractor because MAB Elevator Sump could collect radioactive effluent but is pumped to MAB Sump #3, then to the Floor Drain Tank and then to a WHT where the water is processed and monitored prior to release.

Question Level: F Question Difficulty 3 Justification:

In order to control a potential release, the Unit Supervisor needs fundemental knowledge of where MAB sumps discharge.

Page 26 of 50

ESSENTIAL COOLING WATER SUMP

• Location - 10MAB

• Input - various non-behind Essential radioactive drains in Chillers MEAB (CCW Pump

• Not potentially Area drains, Essential contaminated, reduces Chillers ECW Relief's, amount to be processed EAB HVAC Sump)

• Raised ledge around

• Output - pumped to ECW Sump and Drain Circ Water Discharge Hubs piping 104

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2180 Last used on an NRC exam: Never SRO Sequence Number: 89 Given the following:

A Waste Monitor Tank release needs to be performed RT-8038, LWPS Monitor #1, was declared inoperable 3 days ago In accordance with 0PSP07-WL-LDP2, Liquid Effluent Permit with RT-8038 Inoperable, who approves the discharge and what are the additional requirements for the release per the Offsite Dose Calculation Manual (ODCM)?

Approval ODCM requirement Continuous surveys of the discharge piping A. Shift Manager during the release At least two independent samples of the B. Shift Manager monitor tank are analyzed prior to the release Continuous surveys of the discharge piping C. Health Physics Manager during the release At least two independent samples of the D. Health Physics Manager monitor tank are analyzed prior to the release Answer: B Shift Manager; At least two independent samples of the monitor tank are analyzed prior to the release Page 27 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2180 K/A Catalog Number: G2.3.6 Tier: 3 Group/Category: 3 SRO Importance: 3.8 10CFR Reference or SRO Objective: 55.43(b)(2)

Ability to approve release permits.

STP Lesson: Objective Number: SRO-13400 AUTHORIZE a release of liquid waste in accordance with 0PSP07-WL-LDP1, Liquid Effluent Permit.

Reference:

0PSP07-WL-LDP2 step 5.1.23, ODCM Table 3.3-12 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because continuous surveys could give an indication of a release of excess activity, however it does not match the requirments B: CORRECT: Approval and requirements correct per the given references C: INCORRECT: Credible because Health Physics is involved with all things concerned with radiation, but they are not the approval authority for a discharge. Although continuous surveys could give an indication of a release of excess activity, it does not match the requirments D: INCORRECT: Credible because Health Physics is involved with all things concerned with radiation, but they are not the approval authority for a discharge.

Question Level: F Question Difficulty 3 Justification:

The SRO applicant must have a knowledge of release and ODCM requirements Page 28 of 50

0PSP07-WL-LDP2 Rev. 10 Page 13 of 18 Liquid Effluent Permit With RT-8038 Inoperable 1

Initials 5.1.20.5 Document jumper installation in the Effluent Release Logbook. _____

5.1.21 IF the discharge will occur on the fourteenth day OR greater from the time the monitor was declared inoperable, AND the discharge will proceed, THEN perform the following:

5.1.21.1 Generate a Condition Report to document making releases with RT-8038 inoperable 14 days. (Or record CR# if a CR was already generated on a previous release) _____

CR #_____________________

5.1.21.2 Obtain Chemistry Manager or Designee approval for release(s) to occur with RT-8038 inoperable 14 days: _____

Approval obtained (Initials):_________Date: ________ Time: ________

5.1.22 Perform an Independent Verification of the Liquid Effluent Permit.

Independent Verification by: (Init.)__________ Date:_________ Time:________

5.1.23 Forward package to Shift Manager for signature for the following: _____

• Review of the Liquid Effluent Permit

• Approval for the discharge Shift Manager Date: Time:

This procedure, when completed, shall be retained for the life of the plant.

TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT ACTION OPERABLE

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release Liquid Waste Processing Discharge Monitor (N1RA-RT-8038 or 1 43 N2RA-RT-8038)

2. Flow Rate Measurement Devices Liquid Waste Processing Discharge Line (N1WL-FT-4078 or N2WL-FT-4078) 1 46 ACTION STATEMENTS ACTION 43 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Control 4.11.1.1.1, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 44 - (Not used)

ACTION 45 - (Not used)

ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in place may be used to estimate flow.

03/01/2011 15 ODCM Rev. 17

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2217 Last used on an NRC exam: Never SRO Sequence Number: 90 Given the following:

Unit 1 is at 100% power.

Train A ECW Pump is inoperable to allow for maintenance.

Subsequently:

Electrical Maintenance requests approval to perform Troubleshooting on Train B ECW Pump Room Fan 11B breaker which showed a hot spot during a thermal imaging scan 2 weeks ago.

In accordance with 0POP01-ZO-0012, Operations Troubleshooting Process, the ____________

Manager would approve troubleshooting activities. In the given scenario Troubleshooting activities ____________ be allowed.

A. Shift, would NOT B. Shift, would C. Maintenance, would D. Maintenance, would NOT Answer: A Shift, would NOT Page 29 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2217 K/A Catalog Number: G2.2.20 Tier: 3 Group/Category: 2 SRO Importance: 3.8 10CFR Reference or SRO Objective: 55.43(b)(3)

Knowledge of process for managing troubleshooting activities.

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

Reference:

0POP01-ZO-0012, Operations Troubleshooting Process, steps 5.1 and 6.1.13 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Modified Modified From 2134 Distractor Justification A: CORRECT: Per POP01-ZO-0012, the shift manager would approve troubleshooting activities. In this case, since it is a cross train related component which could become inoperable, troubleshooting would not be allowed.

B: INCORRECT: Credible because the procedure does allow for cross train troubleshooting provided the system will not be rendered inoperable. In this case, if the vent fan becomes inoperable, the system is inoperable.

C: INCORRECT: Credible because the Maintenance manager must approve certain work activities (i.e. work behind a freeze seal, work on main/backseated valves) and electrical maintenance is involved. The work allowance is credible because the procedure does allow for cross train troubleshooting provided the system will not be rendered inoperable.

D: INCORRECT: Credible because the Maintenance manager must approve certain work activities (i.e. work behind a freeze seal, work on main/backseated valves) and electrical maintenance is involved.

Question Level: F Question Difficulty 3 Justification:

The applicant must have knowledge of the Operations Troubleshooting process.

Page 30 of 50

0POP01-ZO-0012 Rev. 1 Page 3 of 14 Operations Troubleshooting Process 3.1.8 0POP01-ZO-0011, Operability, Functionality, and Reportability Guidance 3.2 Conduct of Operations Manual 3.3 CR 09-19595, Process Shortfall Concerning Operations Troubleshooting Activities 3.4 NRC NOV 95-06-02 (CR 95-6790), Operators Changed Configuration of Safety-Related Equipment Without Written Instructions 4.0 Prerequisites 4.1 OBTAIN Shift Manager approval for entry into this procedure. _____

5.0 Precautions and Notes 5.1 Troubleshooting should only be performed on operable systems which will not be rendered inoperable as determined by the Shift Supervisor, or on systems which have already been declared inoperable.

5.2 During Plant Operations troubleshooting activities, the Shift Manager SHALL maintain communications with the Division Manager and the Operations Manager as appropriate to keep them updated with progress and plant impact.

5.3 During performance of this procedure, a system, structure or component (SSC) may be found to be functioning other than as expected. 0POP01-ZO-0011 (Operability, Functionality, and Reportability Guidance) provides departmental guidance for determining Operability and Functionality status of SSCs that come into question.

5.4 Activities during troubleshooting SHALL NOT be performed using uncontrolled written instructions.

5.5 Whenever component manipulations are required during the troubleshooting process, they should be performed using approved plant procedures or CROEs.

5.6 The use of approved plant procedures or CROEs to assist in troubleshooting SHALL NOT warrant exit from this procedure. This procedure SHALL remain in effect throughout the entire troubleshooting process.

5.7 Once approved, any revisions to the original Troubleshooting Plan as documented in Addendum 2 SHALL require an additional LCR and additional approval signatures, including the Shift Manager.

5.8 Conditions/problems discovered during the troubleshooting process SHALL be reported in accordance with 0PGP03-ZX-0002 (Condition Reporting Process).

STP LOT-18 NRC EXAM Print Date 8/4/2013 Exam Bank No.: 2134 Last used on an NRC exam: 2011 Given the following:

Unit 1 is at 100% power.

Train A ECW Pump is inoperable to allow for maintenance.

Subsequently:

During maintenance on Train A ECW Pump, an unexpected condition arose on Train A ECW Discharge Motor Operated Valve (MOV) and the craft wants to troubleshoot the MOV, however troubleshooting the MOV will render the MOV inoperable.

In accordance with 0POP01-ZO-0012, Operations Troubleshooting Process, troubleshooting activities on the MOV _________ allowed. Trouble shooting plans are approved by the

____________ Manager.

A. ARE, Shift B. are NOT, Shift C. are NOT, Plant D. ARE, Plant Answer: A ARE, Shift Page 1 of 2

STP LOT-18 NRC EXAM Print Date 8/4/2013 Exam Bank No.: 2134 RO Outline Number: SRO Outline Number:

K/A Catalog Number: G2.2.20 Tier: 3 Group/Category: 2 RO/SRO Importance: 2.6 / 3.8 RO-10CFR55.41 #

SRO-10CFR55.43 # 3 or SRO Obj:

Knowledge of process for managing troubleshooting activities.

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

Reference:

0POP01-ZO-0012, Operations Troubleshooting Process, steps 5.1 and 6.1.13 Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified from:

Distractor Justification A: CORRECT: Troubleshooting activities are allowed since the pump (system) are already inoperable.

The Shift Manager is the correct approval authority.

B: INCORRECT: Troubleshooting activities would not be allowed if the train was operable, but are since the pump (system) are already inoperable. The approval authority is correct.

C: INCORRECT: Troubleshooting activities would not be allowed if the train was operable, but are since the pump (system) are already inoperable. The Plant Manager is the approval authority for many activities, however it is the Shift Manager in this case.

D: INCORRECT: Activity analysis is correct. The Plant Manager is the approval authority for many activities, however it is the Shift Manager in this case.

Level F Difficulty 3 Justification:

The applicant must have knowledge of the Operations troubleshooting process.

Page 2 of 2

0POP01-ZO-0012 Rev. 1 Page 8 of 14 Operations Troubleshooting Process NOTE The reviewer for technical adequacy SHALL NOT be the plan developer.

6.1.9 OBTAIN a review of the troubleshooting plan for technical adequacy from a Senior Reactor Operator, along with reviewer signature on Addendum 2. _____

6.1.10 ENSURE that a License Compliance Review (LCR) of the troubleshooting plan has been performed per 0PAP01-ZA-0103, License Compliance Review. _____

6.1.10.1 On the LCR form, RECORD this procedure number and revision. _____

6.1.10.2 In the Procedure Title or Description line of the LCR form, INCLUDE a brief description of the troubleshooting task to be performed. _____

6.1.10.3 ATTACH the LCR to Addendum 2 of this procedure. _____

6.1.10.4 ENSURE that the requirements of 0PGP03-ZA-0010 Section 6.8 have been met. _____

6.1.11 ENSURE that a Security Compliance Review (SCR) has been performed in accordance with 0PAP01-ZA-7358 Security Compliance Review. _____

6.1.11.1 ATTACH the SCR to Addendum 2 of this procedure. _____

6.1.12 ENSURE that a Work Risk Assessment has been performed per 0PGP03-ZA-0090, Work Process Program, and documented in the CR referred to in Step 6.1.2. _____

6.1.13 OBTAIN approval signature for the plan from the Shift Manager on Addendum 2. _____

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2163 Last used on an NRC exam: Never SRO Sequence Number: 91 Unit 1 is at 100% Power.

A Reactor Operator reports the following to the Unit Supervisor:

VCT level is 60% and slowly rising.

A VCT Auto Make-up is NOT in service.

Tavg is 1ºF below Tref and slowly lowering.

VCT HI/LO alarm is illuminated.

VCT LO/LO alarm is illuminated.

Which of the following (1) describes the cause of these indications and (2) the action the Unit Supervisor should take?

A. (1) VCT level transmitter, CV-LT-0113, has failed low causing an inadvertent boration.

(2) Enter 0POP09-AN-04M8, Annunciator Response for the VCT LO/LO alarm, OPEN CV-MOV-0113A, VCT OUTL ISOL, and then de-energize the MOV.

B. (1) VCT level transmitter, CV-LT-0113, has failed low causing an inadvertent boration.

(2) Enter 0POP04-TM-0001, Turbine Load Rejection, and lower Main Turbine load to match Tref with the lower Tavg.

C. (1) VCT level transmitter, CV-LT-0112, has failed low causing an inadvertent boration.

(2) Enter 0POP09-AN-04M8, Annunciator Response for the VCT LO/LO alarm, OPEN CV-MOV-0112B, VCT OUTL ISOL, and then de-energize the MOV.

D. (1) VCT level transmitter, CV-LT-0112, has failed low causing an inadvertent boration.

(2) Enter 0POP04-TM-0001, Turbine Load Rejection, and lower Main Turbine load to match Tref with the lower Tavg.

Answer: A (1) VCT level transmitter, CV-LT-0113, has failed low causing an inadvertent boration.

(2) Enter 0POP09-AN-04M8, Annunciator Response for the VCT LO/LO alarm, OPEN CV-MOV-0113A, VCT OUTL ISOL, and then de-energize the MOV.

Page 31 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2163 K/A Catalog Number: 004 A2.10 Tier: 2 Group/Category: 1 SRO Importance: 4.2 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertant boration/dilution STP Lesson: LOT 201.06 Objective Number: 507226 Given a description of plant conditions, ANALYZE the conditions and PREDICT how the CVCS will respond.

Reference:

LOT 201.06 Lesson Plan for CVCS Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: A low failure of CV-LT-0113 will close CV-MOV-0113A, VCT OUTL ISOL, and open CV-MOV-0113B, RWST TO CHG PUMP SUCTION ISOL effectively borating the RCS. The annunciator response for this failure will give direction to open and de-energize CV-MOV-0113A.

B: INCORRECT: This distractor is credible because a low failure of CV-LT-0113 will close CV-MOV-0113A, VCT OUTL ISOL, and open CV-MOV-0113B, RWST TO CHG PUMP SUCTION ISOL effectively borating the RCS. Entry in to 0POP04-TM-0001, Turbine Load Rejection, is credible because turbine load would start to lower from 100% power with Tavg lowering and the Main Turbine in IMP-IN but this would not be a true load rejection in the sense that Tavg normally goes up with a turbine load rejection.

C: INCORRECT: This distractor is credible because a low failure of CV-LT-0112 will close CV-MOV-0112B, VCT OUTL ISOL, and open CV-MOV-0112C, RWST TO CHG PUMP SUCTION ISOL effectively borating the RCS but CV-LT-0112 controls make-up to the VCT and if it was failed low a continuous auto make-up would occur.

D: INCORRECT: This distractor is credible because a low failure of CV-LT-0112 will close CV-MOV-0112B, VCT OUTL ISOL, and open CV-MOV-0112C, RWST TO CHG PUMP SUCTION ISOL effectively borating the RCS but CV-LT-0112 controls make-up to the VCT and if it was failed low a continuous auto make-up would occur. Entry in to 0POP04-TM-0001, Turbine Load Rejection, is credible because turbine load would start to lower from 100% power with Tavg lowering and the Main Turbine in IMP-IN but this would not be a true load rejection in the sense that Tavg normally goes up with a turbine load rejection.

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to evaluate the given condition and then enter the correct procedure.

Page 32 of 50

0POP09-AN-04M8 Rev. 38 Page 68 of 82 ANNUNCIATOR LAMPBOX 04M8 RESPONSE INSTRUCTIONS VCT LEVEL LO-LO Automatic Actions: 1) IF LT-0112 output drops below setpoint (3%), THEN RWST ISOL MOV-0112C will OPEN AND OUTL ISOL MOV-0112B will CLOSE.

2) IF LT-0113 output drops below setpoint (3%), THEN RWST ISOL MOV-0113B will OPEN AND OUTL ISOL MOV-0113A will CLOSE.

Immediate Actions: None Subsequent Actions: 1) IF actual VCT level is less than 3%, THEN ENSURE Charging pump suction aligns to RWST.

2) CHECK the following Plant Computer points to identify a failed VCT level transmitter:

a) CVLA0112 b) CVLA0113

3) IF actual VCT level is less than 3%, THEN PERFORM the following:

a) ENSURE Reactor Makeup system started in AUTO.

b) IF RCS Makeup can NOT be started in AUTO, THEN Manually INITIATE makeup to the RCS per 0POP02-CV-0001, Makeup to the Reactor Coolant System.

4) IF VCT low level is due to excessive RCS leakage, THEN GO TO the leakage procedure appropriate for plant conditions:

• 0POP04-RC-0003, Excessive RCS Leakage

• 0POP04-RC-0006, Shutdown LOCA

• 0POP04-RC-0007, Mode 5 Or Mode 6 LOCA With The Reactor Vessel Head On Page 1 of 4 04M8-F-2 VCT LEVEL LO-LO

0POP09-AN-04M8 Rev. 38 Page 69 of 82 ANNUNCIATOR LAMPBOX 04M8 RESPONSE INSTRUCTIONS VCT LEVEL LO-LO Subsequent Actions: 5) IF a VCT level transmitter (LT-0112 or LT-0113) is failed less than or (continued) equal to 3%, THEN PERFORM the following:

NOTE Holding the VCT outlet isolation valve handswitch in the OPEN position bypasses an automatic SI function to close the valve.

a) OPEN affected VCT outlet isolation valve by taking the VCT outlet isolation valve handswitch to OPEN AND HOLDING it in the open position until the action of either substep b) or c) is complete. {CP004}

• VCT OUTL ISOL MOV-0112B

• VCT OUTL ISOL MOV-0113A b) DISPATCH an Operator to OPEN affected VCT outlet isolation MCC breaker:

• VCT OUTLET ISOL 1(2)-CV-MOV-0112B; MCC E1C1(E2C1)/G3

• VCT OUTLET ISOL 1(2)-CV-MOV-0113A; MCC E1B2(E2B2)/E1 Page 2 of 4 04M8-F-2 VCT LEVEL LO-LO

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2164 Last used on an NRC exam: Never SRO Sequence Number: 92 In regards to the Pressurizer Pressure Reactor Trip Setpoint, which of the following describes the Technical Specification Bases for the Setpoint?

A. Prevent water relief through the Pressurizer Safety Valves.

B. Protect against Departure from Nucleate Boiling and Reactor Coolant System over pressure.

C. Prevent Fuel Pellet melting and greater than 1% cladding strain.

D. Protect against consequences of a power excursion from all power levels.

Answer: B Protect against Departure from Nucleate Boiling and Reactor Coolant System over pressure.

Page 33 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2164 K/A Catalog Number: 010 G2.2.25 Tier: 2 Group/Category: 1 SRO Importance: 4.2 10CFR Reference or SRO Objective: 55.43(b)(2)

Pressurizer Pressure Control: Knowledge of the basis in Technical Specifications for limiting conditions of operations and safety limits.

STP Lesson: LOT 201.14 Objective Number: 8559 STATE the function of the pressurizer pressure and level control system components, controls and instrumentation.

Reference:

LOT 201.14 Lesson on PZR Pressure and Level Control and TS 2.2.1 Reactor Trip System Instrumentation Setpoints Basis.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible because it is the bases for a different reactor trip (pressurizer level)

B: CORRECT: Preventing DNB is the basis for PZR pressure low Reator Trip setpoint and protecting the RCS from over pressure is the basis for the high Reactor Trip setpoint.

C: INCORRECT: Credible because it is the bases for a different reactor trip (overpower delta-T)

D: INCORRECT: Credible because it is the bases for a different reactor trip (power range high flux)

Question Level: F Question Difficulty 3 Justification:

The Unit Supervisor needs to have knowledge of Technical Specification basis.

Page 34 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2174 Last used on an NRC exam: Never SRO Sequence Number: 93 Unit 1 is at full power operations.

Annunciator 125V DC SYSTEM E1A11 TRBL is received and the crew sends a Plant Operator to the Class 1E 125V DC E1A11 Bus.

The Plant Operator reports the following:

Bus voltage reads 129 VDC Negative voltage reads 25 VDC Positive voltage reads 105 VDC Which of the following describes the malfunction and the procedure the Unit Supervisor should use to mitigate the consequences?

MALFUNCTION PROCEDURE Use 0POP02-EE-0001, ESF (Class1E) DC A. Class 1E 125 VDC Bus has a ground. Distribution System, to de-energize the Bus and apply appropriate Tech Specs.

Use 0POP01-ZO-0009, Ground Isolation, to B. Class 1E 125 VDC Bus has a ground.

attempt to isolate the ground fault.

Use 0POP02-EE-0001, ESF (Class1E) DC Class 1E 125 VDC Bus has high C. Distribution System, to place a second battery current.

charger in service.

Class 1E 125 VDC Bus has high Use 0POP01-ZO-0009, Ground Isolation, to D.

current. de-energize unnecessary loads.

Answer: B Class 1E125 VDC Bus has a ground. Use 0POP01-ZO-0009, Ground Isolation, to attempt to isolate the ground fault.

Page 35 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2174 K/A Catalog Number: 063 A2.01 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 DC electrical systems; and (b) based on those predictions, control or mitgate the consequences of those malfunctions or operations: Grounds STP Lesson: LOT 201.37 Objective Number: 92986 DESCRIBE the local and MCR instrumentation available to monitor the Class 1E 125 VDC System.

Reference:

LOT 201.37 and Annunciator Response, 1POP09-AN-03M2, Window A-1, 125V DC SYSTEM E1A11 TRBL.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: De-energizing the bus is credible because it will alleviate the ground, however this action is not proceduralized or desired in this condition.

B: CORRECT: A difference in negative and positive bus voltage is an indication of a ground. Normal positive bus volts is 65 and negative bus volts is 65. The Unit Supervisor should try to locate the source of the ground using the indicated procedure.

C: INCORRECT: Credible because high current may lower voltage, but will not cause a difference in positive and negative readings. Additional charger capacity could prevent damage to batteries in an overcurrent condition, however these chargers are not designed to operate in parallel.

D: INCORRECT: Credible because high current may lower voltage, but will not cause a difference in positive and negative readings. ZO-09 provides guidance on de-energizing loads, however it would be an incorrect application of the procedure.

Question Level: H Question Difficulty 2 Justification:

The Unit Supervisor has to evaluate the given condition and select the correct malfunction and Procedure.

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1POP09-AN-03M2 Rev. 29 Page 6 of 59 Annunciator Lampbox 1-03M-2 Response Instructions 125V DC SYSTEM E1A11 TRBL Subsequent Actions: 11) IF bus current is greater than or equal to 200 amps OR battery (Continued) charger current is greater than or equal to 300 amps, THEN CONTACT Electrical Maintenance for assistance in determining the cause.

12) IF a ground is indicated, THEN DIRECT an Operator to isolate equipment as directed by the Unit/Shift Supervisor to attempt to isolate the fault. REFER TO 0POP01-ZO-0009, Ground Isolation.

13) TAKE appropriate actions per Technical Specifications 3.8.1.1, 3.8.1.2, 3.8.1.3, 3.8.2.1, 3.8.2.2, 3.8.3.1, 3.8.3.2, TRM 3.8.1, and 3.8.2.

14) INITIATE a CR to document faulted conditions and to repair the faulted condition.

15) WHEN desired to return system to normal after repair, THEN PERFORM applicable section of 0POP02-EE-0001, ESF (Class1E) DC Distribution System OR 0POP02-AE-0004, 120 VAC ESF Vital Distribution Power Supplies.

Probable Causes: 1) Loss of battery charger AC input power

2) Battery charger failure

3) Any E1A11 switchboard breaker tripped

4) E1A11 switchboard trouble

5) Instrument failure

6) System ground Page 4 of 5 03M2-A-1 125V DC SYSTEM E1A11 TRBL

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2177 Last used on an NRC exam: Never SRO Sequence Number: 94 Which of the following requires Core Load Supervisor approval prior to performing?

During an outage and while offloading the Reactor Core A. the FHB Upender Operator uses the PROX SWITCH BYPASS key switch to operate the FHB upender because the UPENDER CLEAR light will not illuminate.

B. the FHB Upender Operator positions the TRAVERSE RX/POOL selector switch to RX position to move the carriage from the FHB to Containment.

C. the Fuel Handling Machine Operator lowers a fuel assembly into a Region 2 Fuel Cell.

D. the Fuel Handling Machine Operator unlatches a fuel assembly after being placed in a SFP Region 1 Fuel Cell.

Answer: A the FHB Upender Operator uses the"PROX SWITCH BYPASS" key switch to operate the FHB upender because the "UPENDER CLEAR" light will not illuminate.

Page 37 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2177 K/A Catalog Number: 034 G2.1.42 Tier: 2 Group/Category: 2 SRO Importance: 3.4 10CFR Reference or SRO Objective: 55.43(b)(7)

Fuel Handling Equipment: Knowledge of new and spent fuel movement procedures.

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:

LOT 201.43 Introduction to Refueling and 0POP08-FH-0002, Fuel Handling Machine, and 0POP08-FH-0003, Fuel Transfer System.

Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: Per procedural guidance the Core Load Supervisor, during refueling operations, has to approve any bypass operation of Fuel Handling equipment.

B: INCORRECT: This is a credible distractor because moving the Fuel Transfer system carriage requires two verifications prior to moving; the Upender Frame Down light must be illuminated and the operator must visually verify that the frame is down but this action does not require Core Load Supervisor permission.

C: INCORRECT: This is a credible distractor because independent verification is required to lower a fuel assemble, but not from the Core Load Supervisor.

D: INCORRECT: This is a credible distractor because Step Verifier agreement is required to unlatch a fuel assemble, but not from the Core Load Supervisor.

Question Level: F Question Difficulty 3 Justification:

The Unit Supervisor needs to have knowledge of Fuel Handling Procedure precautions and notes.

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0POP08-FH-0003 Rev. 31 Page 6 of 42 Fuel Transfer System 4.0 Notes and Precautions 4.1 Precautions specified in applicable Radiation Work Permit SHALL be observed.

4.2 The Fuel Transfer Canal and Refueling Cavity SHALL be governed by Housekeeping Zone 3 guidelines per 0PGP03-ZA-0098, Station Housekeeping.

4.3 The areas above the spent fuel pool, fuel transfer canals, an uncovered new fuel vault, the reactor vessel, and the in-containment storage area (with stored fuel or during movement of fuel in the RCB), SHALL be considered a FMEA Level 1 in accordance with 0PGP03-ZA-0014, Foreign Material Exclusion Program.

4.4 0PEP02-ZM-0009, Spent Fuel Pool Storage and Work, provides requirements and guidelines for all work conducted in the SFP, ICSA, the Reactor Cavity, and the fuel transfer canals when filled with water. These requirements and guidelines apply to all individuals who perform work in the above areas.

4.5 IF moving fuel or other irradiated components, THEN fuel transfer equipment SHALL NOT be operated until Transfer Tube Quick Opening Hatch AND Fuel Transfer Tube Gate Valve are opened.

4.6 Key operated interlock bypass switches are administratively controlled.

4.7 IF moving fuel or other irradiated components, THEN key operated interlock bypass switches SHALL ONLY be used with Core Loading Supervisor's permission.

4.8 Control console heaters are wired independently of main power input. IF control consoles are required to be opened for maintenance activities with the heater circuit energized, THEN personnel must take precautions to avoid electrical shock hazards. All work inside the control consoles will be performed in accordance with an approved work package.

4.9 WHEN fuel transfer activities are in progress, THEN communications SHALL be maintained between the Refueling Machine, Fuel Handling Machine, RCB Control Console, FHB Control Console, and Main Control Room. (TRM 3.9.5 and Reference 2.10) 4.10 Prior to lifting a fuel assembly from either Upender, ensure Upender is in vertical position.

4.11 WHEN irradiated fuel is being transported through Fuel Transfer Tube, THEN high radiation conditions may exist in FHB penetration space in vicinity of Fuel Transfer Tube.

Access to this area SHALL be controlled per Form 3 of 0PRP07-ZR-0009, Performance of High Exposure Work. (Reference 2.8) 4.12 This procedure applies to transfer of irradiated material from RCB to FHB, provided material has been configured or packaged so it can be placed in the Fuel Transfer System Fuel Container. In such cases the words "fuel assembly" in this procedure can be construed to mean, "irradiated material".

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2184 Last used on an NRC exam: Never SRO Sequence Number: 95 Unit 2 is at 100% Power.

The Reactor Trip Switchgear loses DC control power from Class 1E 125VDC E1A11.

Subsequently SSPS receives an automatic Reactor Trip signal.

Which of the following (1) describes how the Reactor Trip Breakers are affected and (2) the direction the Unit Supervisor should give to mitigate the consequences of the malfunction?

A. (1) Only Reactor Trip Breaker R Shunt Trip Coil loses power.

(2) Enter 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS.

B. (1) Reactor Trip Breaker R and S Shunt Trip Coil loses power.

(2) Enter 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS.

C. (1) Reactor Trip Breaker R and S Shunt Trip Coil loses power.

(2) Enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

D. (1) Only Reactor Trip Breaker R Shunt Trip Coil loses power.

(2) Enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Answer: D (1) Only Reactor Trip Breaker R Shunt Trip Coil loses power.

(2) Enter 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Page 39 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2184 K/A Catalog Number: 012 A2.07 Tier: 2 Group/Category: 1 SRO Importance: 3.7 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequeces of those malfunctions or operations: Loss of dc control power 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:

LOT 201.20 Lesson on SSPS and LOT 201.37 Lesson on Class 1E 125VDC Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Part 2 is credible because all other remotely operated breakers will not work if control power is lost, however trip breakers have two independent means of opening and will still open when DC power is lost.

B: INCORRECT: Part 1 is credible because both reactor trip breakers receive control power from Class 1E DC, but not from the same source. Part 2 is credible because all other remotely operated breakers will not work if control power is lost, however trip breakers have two independent means of opening and will still open when DC power is lost.

C: INCORRECT: Part 1 is credible because both reactor trip breakers receive control power from Class 1E DC, but not from the same source.

D: CORRECT: Reactor Trip Breaker 'R' Shunt Trip Coil gets power from Class 1E 125 VDC E1A11.

Entering 0POP05-EO-EO00, Reactor Trip or SI is correct because the 48 volt dc uv coil would still trip both Reator Trip Breakers open.

Question Level: H Question Difficulty 4 Justification:

The Unit Supervisor must evaluate the given condition and determine the correct procedure to implement. Must also have basic knowledge of how the RTBs operate.

Page 40 of 50

A-Train Loads

• Reactor Trip Breaker Control Power

• ESF DG A Field Flash

• PL-139A/39A (DG Control Panels)

• 4160/480 VAC (ESF) breaker control power

• Load Sequencer A

• Inverters 001 and 1201

• Pzr PORV 655A

B-Train Loads

• Reactor Trip Breaker Control Power

• ESF DG B Field Flash

• PL-139B/39B (DG Control Panels)

• 4160/480 VAC (ESF) breaker control power

• Load Sequencer B

• Inverter 1203

• Pzr PORV 656A

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2185 Last used on an NRC exam: Never SRO Sequence Number: 96 Unit 1 is in a Site Area Emergency based on the following:

A Core Cooling Orange Path has been in effect for 20 minutes.

Core Exit Thermocouples are 715ºF and slowly rising.

RCS Plenum level is 20%.

Which of the following parameters would cause an escalation to a General Emergency?

NOTE: Consider each of the following separately.

A. Reactor Coolant System Activity (DEI) is reported at 350µCi/gm.

B. Reactor Coolant Failed Fuel Monitor, RT-8039, is reading 900µCi/ml.

C. Containment Pressure is 25psig.

D. Containment Hatch Monitor is reading 200mR/hr.

Answer: A Reactor Coolant System Activity (DEI) is reported at 350µCi/gm.

Page 41 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2185 K/A Catalog Number: 002 G2.4.41 Tier: 2 Group/Category: 2 SRO Importance: 4.6 10CFR Reference or SRO Objective: Objective SRO 74026 Reactor Coolant: Knowledge of the emergency action level thresholds and classifications.

STP Lesson: LOT 803.14 Objective Number: 74026 Given an emergency condition and a copy of the emergency classification tables from 0ERP01-ZV-IN01, Emergency Classification, CLASSIFY the emergency condition.

Reference:

LOT 803.14 - 0ERP01-ZV-IN01, Emergency Classification Attached Reference

Attachment:

0ERP01-ZV-IN01, Emergency Classification, Addendum 1, Emergency Classification Tables, Page 2 and 3.

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: CORRECT: With the given conditions the Fission Product Barrier Degradation totals 8 points. 3 points for potential loss of fuel clad due to Core Cooling Orange Path or RCS Plenum level at 20%

or CETs at 715 degrees F. 4 points for loss of RCS due to Core Cooling Yellow with subcooling less than 0 degrees F. (Core Cooling Orange and CETs at 715 degrees F would satisfy this) 1 point for potential loss of Containment due to Core Cooling Orange greater than 15 minutes. With RCS DEI at 350 uCi/gm fuel clad would go from potential loss to loss. (3 points to 4 points) This would make the Fission Product Barrier total 9 points and thus raise the E-Plan declaration to a General Emergency.

B: INCORRECT: This is a credible distractor because although the failed fuel monitor reading of 900uCi/ml is greater than the limit of 870uCi/ml this represents on a potential loss of the fuel clad which has already been identified.

C: INCORRECT: This is a credible distractor because athough Containment pressure of 25 psig is considerably high, Containment pressure alone could not cause an escalation to a GE because it would only represent a potential loss of Containment which has already been identified.

D: INCORRECT: This is a credible distractor because although an elevated reading on the Containment Hatch Rad Monitor could cause an escalation in the E-Plan classification the reading would have to go above 222mR/hr.

Question Level: H Question Difficulty 3 Justification:

The SRO has to evaluate the condition given and determine which parameter would cause an escalation of emergency classification level to a GE.

Page 42 of 50

0ERP01-ZV-IN01 Rev. 9 Page 11 of 114 Emergency Classification Addendum 1 Emergency Classification Tables Page 2 of 28 RECOGNITION CATEGORY F FISSION PRODUCT BARRIER DEGRADATION INITIATING CONDITION MATRIX Determine which combination of the three barriers are lost or have a potential loss and use the following matrix to classify the event. Also, an event (or multiple events) could occur which result in the conclusion that the loss or potential loss is IMMINENT (within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this IMMINENT loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT (1-2) ALERT (3-4) SITE AREA EMERGENCY (5-8) GENERAL EMERGENCY (9-10)

FU1 ANY Loss or ANY Potential Loss FA1 ANY Loss or ANY Potential Loss FS1 Loss of BOTH Fuel Clad and FG1 Loss of ANY Two Barriers AND of Containment of Fuel Clad or RCS RCS OR Potential Loss of BOTH Potential Loss or Loss of Third Fuel Clad and RCS Barrier FU2 Fuel Clad Degradation See SU6 OR FU3 RCS Leakage - See SU7 Potential Loss of EITHER Fuel Clad or RCS AND Loss of ANY Additional Barrier Operating Modes 1 through 4 Note: 1. At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from General Emergency.

2. The ability to escalate to higher emergency classes as an event degrades must be maintained. RCS leakage steadily increasing would represent an increasing risk to public health and safety.

Determination of Emergency Classification Level Select values from the top of the columns on the next page, which describe specific Fission Product Barrier degradation. Select the higher value that applies from each barrier. Add the values to arrive at the total challenge to the Fission Product Barriers. The emergency classification is determined from the range of values shown in parentheses in the table above.

0ERP01-ZV-IN01 Rev. 9 Page 12 of 114 Emergency Classification Addendum 1 Emergency Classification Tables Page 3 of 28 RECOGNITION CATEGORY F FISSION PRODUCT BARRIER DEGRADATION INITIATING CONDITION MATRIX FUEL CLAD RCS CONTAINMENT EAL POTENTIAL LOSS (3) LOSS (4) POTENTIAL LOSS (3) LOSS (4) POTENTIAL LOSS (1) LOSS (2)

CSF CSF CSF CSF CSF 1 Core Cooling - Orange Core Cooling - Red RCS Integrity - Red Core Cooling - Yellow Containment - Red OR OR with subcooling < 0 qF OR 2

Heat Sink - Red Heat Sink - Red2 Core Cooling - Orange > 15 min.

RCS Activity RCS Activity RCS Leak Rate RCS Leak Rate Containment Pressure Containment Pressure 2 Failed Fuel Monitor, Dose Equivalent Iodine Unisolable leak exceeding the Leak rate greater than CVCS Greater than 6% hydrogen Initial increase followed by rapid RT-8039, equal to or greater greater than 300 PCi/gm capacity of one centrifugal System's ability to maintain RCS concentration in containment unexplained decrease than 870 PCi/ml charging pump in the normal inventory as indicated by loss of OR OR charging mode. RCS subcooling. Containment pressure greater than Containment pressure or sump 9.5 psig with neither containment level not increasing as expected spray nor RCFC running. with LOCA conditions.

Core Exit Thermocouple Core Exit Thermocouple SG Tube Rupture SG Tube Rupture SG Tube Leak 3 t 708qF 1200qF SG Tube has ruptured and SG Tube is ruptured and has a Primary to secondary leakage the primary to secondary leak non-isolable secondary steam greater than 150 gpd through any rate is greater than the capacity release one steam generator with direct of one centrifugal charging secondary side leakage to pump. atmosphere Reactor Vessel Water Level Containment Bypass Containment Isolation 4 Plenum level less then 20% VALID increase in reading on area Containment isolation signal or ventilation monitors in areas AND adjacent to the containment Valves not closed boundary with a known LOCA AND inside containment. A pathway to the environment exists.

RCB Rad Monitor RCB Rad Monitor RCB Rad Monitor 5 RT-8050 or RT-8051 RT-8050 or RT-8051 RT-8050 or RT-8051 greater than greater than 100 R/hr greater than 100 R/hr 1,000 R/hr OR OR OR Hatch Monitor greater Hatch Monitor greater than 222 Hatch Monitor greater than 2,222 than 222 mR/hr mR/hr mR/hr Note: 1. The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment Barrier. Unusual Event Initiating Conditions (ICs) associated with RCS and Fuel Clad barriers are addressed under SU6 and SU7.

2. CSF indicators must be valid; outside the immediate control of the operator.

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2187 Last used on an NRC exam: Never SRO Sequence Number: 97 A Site Area Emergency has been declared in Unit 1. The TSC, EOF and JIC have not been activated yet.

An Inside Containment Isolation Valve needs to be manually closed to stop a radiological release that is affecting the owner controlled area. It is estimated that it will take a worker 15 minutes to close the valve once in the area. The dose rate in the area of the valve is estimated at 25 REM/Hour.

In accordance with 0ERP01-ZV-IN06, Radiological Exposure Guidelines, who at the minimum should provide approval to perform this task?

A. Emergency Director Only B. Emergency Director and the Worker C. The Worker Only D. Acting Radiological Manager and the Worker Answer: B Emergency Director and the Worker Page 43 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2187 K/A Catalog Number: 103 G2.4.38 Tier: 2 Group/Category: 1 SRO Importance: 4.4 10CFR Reference or SRO Objective: 55.43(b)(4)

Containment: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

STP Lesson: LOT 803.14 Objective Number: 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:

LOT 803.14 - 0ERP01-ZV-SH01, Shift Manager - Responsibilities which include directing control room response to mitigate the emergency condition.

Attached Reference

Attachment:

0ERP01-ZV-IN06, Radiological Exposure Guidelines - 7 Page Procedure NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Credible beacause the Emergency Director must sign, however since 10CFR20 limits are being exceeded, the worker must sign too.

B: CORRECT: Since the exposure to the worker will exceed 5 REM then the Worker and the Emergency Director are required to sign for approval. See Data Sheet 1 in the Procedure.

C: INCORRECT: Credible because the exposure to the worker will exceed 5 REM and the Worker is required to sign but the Emergency Director is always required to sign.

D: INCORRECT: Credible because the Acting Radiological Manager has procedural responsibilities for requesting and tracking approvals, however the procedure does not give them signature authority for the ED.

Question Level: H Question Difficulty 3 Justification:

The SRO has to evaluate the condition and then use the procedure to determine authorization for the exposure approval.

Page 44 of 50

0ERP01-ZV-IN06 Rev. 6 Page 6 of 7 Radiological Exposure Guidelines Data Sheet 1 Emergency Exposure Approval Page 1 of 1 Name: _______________________________ SSN: ______________________________________

Date/Time: ___________________________ Discipline Group: ___________________________

Reason for Dose Extension Request: ______________________________________________________

Current exposure to date: ___________ Rem (TEDE) TLD No.: ____________________________

Exposure Authorized to: ___________________________ Rem (TEDE)

• Emergency Worker Emergency Director

• Required to exceed 10CFR20 TEDE exposure limit (5 rem).

WHEN COM PLETED, THIS RECORD SHALL BE RETAINED IN ACCORDANCE WITH THE DOCUM ENT TYPE LIST (DTL).

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2188 Last used on an NRC exam: Never SRO Sequence Number: 98 Given the following:

A LOCA has occurred in Unit 2 Operators are performing 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant Only one HHSI pump is available and running CETs are 710 °F and rising Core voiding is just beginning to occur Which of the following describes the initial response of the Source Range detectors to the core voiding and the actions the Unit Supervisor should take to control the voiding?

NI Response Actions Enter 0POP05-EO-FRC1, Response to Inadequate Core Indication will A. Cooling and depressurize the RCS using a Pressurizer PORV to RISE 255 psig to allow accumulators to inject.

Indication will Enter 0POP05-EO-FRC2, Response to Degraded Core Cooling B.

RISE and start all available charging pumps to raise RCS inventory.

Enter 0POP05-EO-FRC1, Response to Inadequate Core Indication will C. Cooling and depressurize the RCS using a Pressurizer PORV to LOWER 255 psig to allow accumulators to inject.

Indication will Enter 0POP05-EO-FRC2, Response to Degraded Core Cooling D.

LOWER and start all available charging pumps to raise RCS inventory.

Answer: B Indication will RISE; Enter 0POP05-EO-FRC2, Response to Degraded Core Cooling and start all available charging pumps to raise RCS inventory.

Page 45 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 2188 K/A Catalog Number: 015 A2.05 Tier: 2 Group/Category: 2 SRO Importance: 3.8 10CFR Reference or SRO Objective: 55.43(b)(5)

Ability to 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: Core void formation.

STP Lesson: LOT 502.09 Objective Number: 50384 DESCRIBE the effects on reactor kinetics of coolant voiding in the core region, and relate the Excore Nuclear Instrumentation System response to voiding.

Reference:

LOT502.09 handout page 39, 0POP05-EO-FRC2 step 2 RNO Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: New Modified From Distractor Justification A: INCORRECT: Procedure is credible because FRC1 will be performed if temperature continues to rise. Action is credible because FRC1 does inject accumulators, but by depressurizing SGs.

B: CORRECT: The beginning of core voiding during a LOCA will cause excore NIs to start reading higher due to more neutrons leaking out of the core. With CETs above 708 and not all SI pumps running, FRC2 will be entered and all available charging pumps started to help raise RCS inventory.

C: INCORRECT: NI response is credible because voiding reduces Keff by reducing moderation which would cause NIs to lower (this effect is overcome by increased leakage though). Procedure is credible because FRC1 will be performed if temperature continues to rise. Action is credible because FRC1 does inject accumulators, but by depressurizing SGs.

D: INCORRECT: NI response is credible because voiding reduces Keff by reducing moderation which would cause NIs to lower (this effect is overcome by increased leakage though).

Question Level: H Question Difficulty 3 Justification:

The Unit Supervisor has to have knowledge of the operation and effects on nuclear instrumentation and has to evaluate the given condition to determine the procedure to use to mitigate the accident.

Page 46 of 50

is less attenuation of neutron flux in the downcomer, and the transmission ratio increases.

In fact, as the homogeneous void fraction in the core region increases from 0 to 100 percent, the transmission ratio increases by several orders of magnitude (from four to six orders of magnitude, depending upon the core model and method of calculation used).

The increase in transmission ratio causes the excore detector response to increase with homogeneous voiding.

The results of all excore detector response analyses performed Similarities and following the TMI-2 accident were consistent in the following Variations in respect: For homogeneous void fractions of up to about 80 percent, Analysis Results detector response increases (that is, the count rate goes up) continuously.

For homogeneous void fractions in excess of about 80 percent, the detector response results reported in various analyses are not entirely consistent. Some analyses show detector response leveling off at a void fraction of about 80 percent (fig. 3-1.11). Others show detector response continuously increasing for higher void fractions, all the way up to 100 percent voids (fig. 3-1.12).

In all of the analyses, increases in detector response are attributed to increases in the transmission ratio, which in turn are attributed to decreases in downcomer fluid density. The differences in results for homogeneous void fractions greater than 80 percent voids can be attributed to the following differences in the analyses:

CCalculated effects of voiding on keff CCalculated effects of voiding on core source strength All analyses show keff decreasing as voiding increases.

However, some show keff decreasing more than others, especially at very high void fractions. Similarly, all analyses show core source strength decreasing as voiding increases. However, some show source strength decreasing more than others.

The analyses showing detector response leveling off at very high void fractions are based on comparatively low values for keff and source strength at high void fractions. In these analyses, the reduction in core neutron population that occurs at very high void fractions counterbalances the increased transmission ratio. Although the neutron leakage probability increases with voiding, there are fewer neutrons available to leak. So the detector response levels off.

WOGMCD.3.1-WP

1991 Westinghouse Electric Corp.

Westinghouse Proprietary Class 2

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Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1604 Last used on an NRC exam: Never SRO Sequence Number: 99 Given the following:

The crew is performing 0POP05-EO-FRH1, Response to Loss of Secondary Heat Sink, due to an earlier red path on the Heat Sink Critical Safety Function.

Subsequently:

The STA reports the following current conditions:

o Subcriticality CSF ORANGE o Core Cooling CSF ORANGE o Heat Sink CSF YELLOW o Integrity CSF RED o Containment CSF RED o Inventory CSF YELLOW Which one of the following correctly describes the actions required?

Complete 0POP05-EO-FRH1 and transition to A. 0POP05-EO-FRZ1, Response to High Containment Pressure.

B. 0POP05-EO-FRC2, Response to Degraded Core Cooling.

C. 0POP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition.

D. 0POP05-EO-FRS1, Response to Nuclear Power Generation - ATWS Answer: C 0POP05-EO-FRP1, Response to Imminent Pressurized Thermal Shock Condition.

Page 47 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1604 K/A Catalog Number: G2.4.14 Tier: 3 Group/Category: 4 SRO Importance: 4.5 10CFR Reference or SRO Objective: 55.43(b)(5)

Knowledge of general guidelines for EOP usage.

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, EOP User's Guide Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT: Credible because a red condition is a higher priority than the current yellow condition on Heat Sink, but not correct because it is lower in priority than Integrity.

B: INCORRECT: Credible because Core Cooling is a higher priority CSF than the current Heat Sink, but not correct because red overrides orange and Subcriticality is higher priority than Core cooling.

C: CORRECT: Per ZA-0018, a red or orange priority FRP must be completed unless preempted by a higher priority condition. Once complete, transition is made to the next highest priority (color and CSF) condition.

D: INCORRECT: Credible because Subcriticality is a higher priority CSF than the current Heat Sink, but not correct because red overrides orange.

Question Level: H Question Difficulty 3 Justification:

The applicant must have knowledge of the hierarchy that exists for critical safety functions and their associated colors and be able to evaluate the given conditions and determine the appropriate action IAW 0POP01-ZA-0018 Page 48 of 50

0POP01-ZA-0018 Rev. 21 Page 27 of 48 Emergency Operating Procedure User's Guide NOTE When implementing Addendum 5 Verification of ESF Equipment Operation of 0POP05-EO-EO00, then CSF Status Trees are monitored and NOT implemented even though a transition to an FRP is determined. The transition is delayed in most cases (refer to Step 6.1 for exceptions) until the verification of ESF equipment (per Addendum 5) is complete since this may resolve the CSF Status Tree abnormal indication.

6.5 WHEN monitoring CSF Status Trees, THEN:

6.5.1 Always perform the evaluations in the priority sequence listed in Section 6.3.

6.5.2 Enter at the box marked ENTER located at the left side of the status tree.

6.5.3 Answer the questions based on plant conditions at the time and follow the appropriate branch line to the next question.

6.5.4 WHEN a color-coded terminus is reached, THEN the individual status tree evaluation is complete.

6.6 IF a RED condition is reached, THEN immediately stop any ORP or yellow path FRP actions in progress (i.e., DO NOT complete the step in progress) AND perform the FRP required by the RED condition.

6.7 IF during the performance of a RED condition FRP, a RED condition of higher priority as listed in Section 6.3 arises, THEN the higher priority condition should be addressed first AND the lower priority condition FRP suspended (i.e., complete the step in progress).

6.8 IF an ORANGE condition arises, THEN monitor all of the remaining status trees. IF no RED condition exists, THEN suspend any ORP in progress (i.e., complete the step in progress) and perform the FRP required by the ORANGE condition.

6.9 IF during the performance of an ORANGE condition FRP, a RED condition OR higher priority ORANGE condition arises, THEN the RED or higher priority ORANGE condition is to be addressed first, and the original ORANGE condition FRP suspended (i.e., complete the step in progress). IF a FRP specifically states that a higher priority condition should NOT be addressed, THEN this requirement does NOT apply.

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1732 Last used on an NRC exam: Never SRO Sequence Number: 100 Unit 1 is operating at 100% power when Chemistry notifies the Control Room of the following results:

RCS gross activity of 219 microcuries per gram and a newly calculated (E-bar) of 0.2734.

Based on this information, as a minimum, the Unit Supervisor should....

A. Enter 0PGP03-ZO-0012, Plant Systems Chemistry Control, and shutdown as quickly as safe plant operation allows to comply with Tech Spec 3.4.8, RCS Specific Activity.

B. Enter 0PGP03-ZO-0012, Plant Systems Chemistry Control, and immediately take action to reduce power to 50% or less to comply with Tech Spec 3.4.8, RCS Specific Activity.

C. Enter POP04-RC-0001, High Reactor Coolant System Activity, and raise letdown flow in order to pass more coolant through the Demineralizers.

D. Enter POP04-RC-0001, High Reactor Coolant System Activity, and lower letdown flow in order to allow the coolant to remain in the Demineralizers longer.

Answer: C Enter POP04-RC-0001, High Reactor Coolant System Activity, to raise letdown flow in order to pass more coolant through the Demineralizers.

Page 49 of 50

Print Date 8/3/2013 STP LOT-19 NRC SRO EXAM Exam Bank No.: 1732 K/A Catalog Number: G2.1.34 Tier: 3 Group/Category: 1 SRO Importance: 3.5 10CFR Reference or SRO Objective: 55.43(b)(5)

Knowledge of primary and secondary plant chemistry limits.

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:

TS 3.4.8; POP04-RC-0001, High Reactor Coolant System Activity Attached Reference

Attachment:

NRC Reference Req'd

Attachment:

Source: Bank Modified From Distractor Justification A: INCORRECT - Credible because RCS activity is listed in the chemistry specification procedure (but does not have an action level so ZO-0012 would not be entered). The action specified follows the required action for a level 3 violation in ZO-0012. The Tech Spec inference is credible because there is a Tech Spec limit of 100/E-bar for RCS activity (which is not exceeded by the given conditions).

B: INCORRECT - Credible because RCS activity is listed in the chemistry specification procedure (but does not have an action level so ZO-0012 would not be entered). The action specified follows the required action for a level 2 violation in ZO-0012. The Tech Spec inference is credible because there is a Tech Spec limit of 100/E-bar for RCS activity (which is not exceeded by the given conditions).

C: CORRECT - Reported activity does not exceed the TS limit (calculated limit is 365.76 microcuries) but is an elevated reading so POP04 entry would be required. The POP04 has the crew raise letdown flow to improve RCS cleanup.

D: INCORRECT - Credible beacause lower flow would mean more contact time with the resin in the demineralizer, however raising flow will actually clean up a given volume faster provided channeling does not occur which is the reason for the maximum flow in the POP04.

Question Level: H Question Difficulty 3 Justification:

The applicant must calculate the Tech Spec limit based on the given E-bar to determine if Tech Specs have been exceeded. Then, based on this determination choose the correct procedure response.

Page 50 of 50

0POP04-RC-0001 High Reactor Coolant System Activity Rev. 10 Page 4 of 28 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION Placing a cation bed in service will reduce the lithium concentration in the RCS; thus, reducing the primary coolant pH. Significant reduction in RCS pH, for example from 7.2 to 6.9, may cause a crud burst.

NOTE Raising Letdown flow SHALL be limited to only the flowrate allowed with one CCP in-service.

(CR 11-13341)

_____ 4.0 PERFORM Any Of The Following Per 0POP02-CV-0004, Chemical And Volume Control System As Recommended By Chemistry:

_____ a. RAISE letdown flow by placing additional letdown orifice(s) in service

_____ b. PLACE cation demineralizers in service

_____ c. (Mode 5) PLACE Reactor Coolant Purification Pump (RCPP) In Service Observing All Notes And Cautions

_____ 5.0 CHECK RT-8039, Failed Fuel Monitor - GO TO Step 8.0.

x GREATER THAN ALERT SETPOINT OR x RISING INDICATION This Procedure is Applicable in Modes 1-5

LOT 19 NRC Exam SRO Reference Package

0ERP01-ZV-IN01 Rev. 9 Page 11 of 114 Emergency Classification Addendum 1 Emergency Classification Tables Page 2 of 28 RECOGNITION CATEGORY F FISSION PRODUCT BARRIER DEGRADATION INITIATING CONDITION MATRIX Determine which combination of the three barriers are lost or have a potential loss and use the following matrix to classify the event. Also, an event (or multiple events) could occur which result in the conclusion that the loss or potential loss is IMMINENT (within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). In this IMMINENT loss situation use judgment and classify as if the thresholds are exceeded.

UNUSUAL EVENT (1-2) ALERT (3-4) SITE AREA EMERGENCY (5-8) GENERAL EMERGENCY (9-10)

FU1 ANY Loss or ANY Potential Loss FA1 ANY Loss or ANY Potential Loss FS1 Loss of BOTH Fuel Clad and FG1 Loss of ANY Two Barriers AND of Containment of Fuel Clad or RCS RCS OR Potential Loss of BOTH Potential Loss or Loss of Third Fuel Clad and RCS Barrier FU2 Fuel Clad Degradation See SU6 FU3 RCS Leakage - See SU7 OR Potential Loss of EITHER Fuel Clad or RCS AND Loss of ANY Additional Barrier Operating Modes 1 through 4 Note: 1. At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from General Emergency.

2. The ability to escalate to higher emergency classes as an event degrades must be maintained. RCS leakage steadily increasing would represent an increasing risk to public health and safety.

Determination of Emergency Classification Level Select values from the top of the columns on the next page, which describe specific Fission Product Barrier degradation. Select the higher value that applies from each barrier. Add the values to arrive at the total challenge to the Fission Product Barriers. The emergency classification is determined from the range of values shown in parentheses in the table above.

0ERP01-ZV-IN01 Rev. 9 Page 12 of 114 Emergency Classification Addendum 1 Emergency Classification Tables Page 3 of 28 RECOGNITION CATEGORY F FISSION PRODUCT BARRIER DEGRADATION INITIATING CONDITION MATRIX FUEL CLAD RCS CONTAINMENT EAL POTENTIAL LOSS (3) LOSS (4) POTENTIAL LOSS (3) LOSS (4) POTENTIAL LOSS (1) LOSS (2)

CSF CSF CSF CSF CSF 1 Core Cooling - Orange Core Cooling - Red RCS Integrity - Red Core Cooling - Yellow Containment - Red OR OR with subcooling < 0 F OR 2

Heat Sink - Red Heat Sink - Red2 Core Cooling - Orange > 15 min.

RCS Activity RCS Activity RCS Leak Rate RCS Leak Rate Containment Pressure Containment Pressure 2 Failed Fuel Monitor, Dose Equivalent Iodine Unisolable leak exceeding the Leak rate greater than CVCS Greater than 6% hydrogen Initial increase followed by rapid RT-8039, equal to or greater greater than 300 Ci/gm capacity of one centrifugal System's ability to maintain RCS concentration in containment unexplained decrease than 870 Ci/ml charging pump in the normal inventory as indicated by loss of OR OR charging mode. RCS subcooling. Containment pressure greater than Containment pressure or sump 9.5 psig with neither containment level not increasing as expected spray nor RCFC running. with LOCA conditions.

Core Exit Thermocouple Core Exit Thermocouple SG Tube Rupture SG Tube Rupture SG Tube Leak 3 708F 1200F SG Tube has ruptured and SG Tube is ruptured and has a Primary to secondary leakage the primary to secondary leak non-isolable secondary steam greater than 150 gpd through any rate is greater than the capacity release one steam generator with direct of one centrifugal charging secondary side leakage to pump. atmosphere Reactor Vessel Water Level Containment Bypass Containment Isolation 4 Plenum level less then 20% VALID increase in reading on area Containment isolation signal or ventilation monitors in areas AND adjacent to the containment Valves not closed boundary with a known LOCA AND inside containment. A pathway to the environment exists.

RCB Rad Monitor RCB Rad Monitor RCB Rad Monitor 5 RT-8050 or RT-8051 RT-8050 or RT-8051 RT-8050 or RT-8051 greater than greater than 100 R/hr greater than 100 R/hr 1,000 R/hr OR OR OR Hatch Monitor greater Hatch Monitor greater than 222 Hatch Monitor greater than 2,222 than 222 mR/hr mR/hr mR/hr Note: 1. The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Containment Barrier. Unusual Event Initiating Conditions (ICs) associated with RCS and Fuel Clad barriers are addressed under SU6 and SU7.

2. CSF indicators must be valid; outside the immediate control of the operator.

SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION D0527 STI: 33010739 0ERP01-ZV-IN06 Rev. 6 Page 1 of 7 Radiological Exposure Guidelines Quality Non Safety-Related Usage: N/A Effective Date: 10/26/11 S. Korenek N/A N/A Emergency Response Division PREPARER TECHNICAL USER COGNIZANT ORGANIZATION Table of Contents Page

1.0 Purpose and Scope

.............................................................................................................................2 2.0 Responsibilities ..................................................................................................................................2 3.0 Precautions and Limitations...............................................................................................................2 4.0 References ..........................................................................................................................................3 5.0 Procedure ...........................................................................................................................................3 6.0 Support Documents............................................................................................................................3 Addendum 1, Emergency Cumulative Exposure Limits....................................................................4 Addendum 2, Risks Involved with Exposures Greater Than 25 rem TEDE .....................................5 Data Sheet 1, Emergency Exposure Approval...................................................................................6 Data Sheet 2, Emergency Exposure Approval Log ...........................................................................7

0ERP01-ZV-IN06 Rev. 6 Page 2 of 7 Radiological Exposure Guidelines

1.0 Purpose and Scope

1.1 This procedure provides guidance for authorizing radiological exposures in excess of 10CFR20 limits during response to emergency conditions at an Alert, Site Area Emergency, or General Emergency.

1.2 This procedure implements the requirements of the South Texas Project Electric Generating Station (STPEGS) Emergency Plan specific to radiological exposure guidelines during emergency conditions at an Alert, Site Area Emergency, or General Emergency.

2.0 Responsibilities 2.1 The Emergency Director is responsible for approving ALL requests for exposure extensions above 10CFR20 limits.

2.2 The following individuals are responsible for requesting exposure extensions, and tracking approvals for the following personnel:

2.2.1 Acting Radiological Manager - Onshift personnel until Technical Support Center activation.

2.2.2 Radiological Coordinator - Operations Support Center personnel 2.2.3 Radiological Manager - Technical Support Center, Security and Control Room personnel.

2.2.4 Radiological Director - Emergency Operations Facility and Offsite Field Teams.

3.0 Precautions and Limitations 3.1 Administrative dose limits are not applicable.

3.2 Emergency responders shall be authorized an exposure limit of 5 rem TEDE.

3.3 No individual shall knowingly exceed 10CFR20 exposure limits except when authorized to do so by the Emergency Director.

3.4 Upon Assembly and Accountability completion, ensure all personnel remaining in the Protected Area have Thermoluminescent Dosimetry, Document issue using Form 3, TLD Issuance Log.

0ERP01-ZV-IN06 Rev. 6 Page 3 of 7 Radiological Exposure Guidelines 3.5 Data Sheet 1, Emergency Exposure Approval, Data Sheet 2, Emergency Exposure Approval Log, shall be included in the individual dosimetry files which are forwarded to Records Management for retention. (10CFR20.2106) (ANI 80.1A) 4.0 References 4.1 STPEGS Emergency Plan 4.2 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents 4.3 NUREG-0654 FEMA REP-1, Rev. 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 4.4 0PGP05-ZV-0004, Emergency Plan Implementing Procedure Users Guide 5.0 Procedure 5.1 Verify individual(s) have been issued a Thermoluminescent Dosimeter.

5.2 Determine if alternative actions are available to preclude exposure extension. This should include the use of shielding, assignment of personnel with lower exposures, or rotation of personnel on the job.

5.3 To approve exceeding 10CFR20 exposure limits complete Data Sheet 1, Emergency Exposure Approval, for each individual.

5.4 Log individuals names on Data Sheet 2, Emergency Exposure Approval Log, for all individuals approved to exceed exposure limits.

5.5 Brief individuals expected to receive exposures greater than 25 rem TEDE using Addendum 2, Risks Involved with Exposures Greater Than 25 rem TEDE.

6.0 Support Documents 6.1 Addendum 1, Emergency Cumulative Exposure Limits 6.2 Addendum 2, Risks Involved with Exposures Greater Than 25 rem TEDE 6.3 Data Sheet 1, Emergency Exposure Approval 6.4 Data Sheet 2, Emergency Exposure Approval Log

0ERP01-ZV-IN06 Rev. 6 Page 4 of 7 Radiological Exposure Guidelines Addendum 1 Emergency Cumulative Exposure Limits Page 1 of 1 TEDE Required Approval to Approval Special Considerations Exposure Limit Exceed 10CFR20 Documentation Exposure Limit 5-10 rem Emergency Director Data Sheet 1

• Approval to receive exposure in this range must be based on the protection of valuable property, life saving activities or protecting large populations.

10 - 25 rem Emergency Director Data Sheet 1

• Approval to receive exposure in this range must be based on life saving activities or protecting large populations.

>25 rem Emergency Director Data Sheet 1

• Approval to exceed this limit applies to life saving activities or protecting large populations.

• 25 rem TEDE is planned exposure limit and should not be considered a maximum upper limit to removing personnel from life threatening environments.

• The individual shall be a volunteer and be aware of the risks involved.

0ERP01-ZV-IN06 Rev. 6 Page 5 of 7 Radiological Exposure Guidelines Addendum 2 Risks Involved with Exposures Greater Than 25 rem TEDE Page 1 of 1 You have indicated that you are volunteering to be part of an emergency repair/damage control team. The nature of your task is such that you will probably receive an exposure to radiation that will be at a level above the normal limits. You need to have full awareness of the radiological risks involved. The purpose of this briefing is to make you aware of these risks.

The Emergency Director has authorized you to receive an emergency exposure. As you might recall from your training, our procedures allow such a once in a lifetime exposure. The emergency limits are based upon recommendations by the EPA.

There are two categories of risk associated with this type of radiological exposure that you should be fully aware of. These two risks are the immediate health effect and the delayed health effect.

IMMEDIATE HEALTH EFFECTS:

The immediate health effect of an acute exposure (a large dose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) to radiation will vary with the individual and with the amount of exposure. Generally, as the exposure increases, the immediate health effect is more severe. The exposure you are projected to receive will most likely cause temporary blood changes which may make you temporarily more susceptible to illness. Also, although unlikely, you may experience temporary nausea, vomiting and diarrhea. Your level of exposure is well below the levels that have led to immediate fatalities. As a point of reference, at the 100 rem TEDE whole body dose, without medical treatment, approximately 1% of the exposed population would die. At an exposure of 450 rem TEDE, without medical treatment, 50% would die.

DELAYED HEALTH EFFECTS:

The delayed health effect of an acute exposure to radiation is an increase in the risk of premature death from cancer.

It is hard to estimate the increase in cancer risk; however, it is fair to state that your chance of getting cancer increases with every rem that you receive.

Statistics indicate that if a person 40 to 50 years of age were to receive 25 rem TEDE of exposure, then the risk of premature death due to cancer will increase by approximately 0.5%. Another way of looking at this would be to suppose what would happen if 1000 people were exposed to 25 rem TEDE. These same statistics indicate that we would expect approximately 5 of the 1000 individuals to die prematurely from cancer as a result of the 25 rem TEDE exposure (premature meaning an estimated 15 years of life lost).

In summary, if an individual 45 years old receives a 25 rem TEDE exposure, then it is possible that that person could die prematurely from cancer. The odds are roughly 1 in 200.

0ERP01-ZV-IN06 Rev. 6 Page 6 of 7 Radiological Exposure Guidelines Data Sheet 1 Emergency Exposure Approval Page 1 of 1 Name: _______________________________ SSN: ______________________________________

Date/Time: ___________________________ Discipline Group: ___________________________

Reason for Dose Extension Request: ______________________________________________________

Current exposure to date: ___________ Rem (TEDE) TLD No.: ____________________________

Exposure Authorized to: ___________________________ Rem (TEDE)

• Emergency Worker Emergency Director

• Required to exceed 10CFR20 TEDE exposure limit (5 rem).

WHEN COMPLETED, THIS RECORD SHALL BE RETAINED IN ACCORDANCE WITH THE DOCUMENT TYPE LIST (DTL).

0ERP01-ZV-IN06 Rev. 6 Page 7 of 7 Radiological Exposure Guidelines Data Sheet 2 Emergency Exposure Approval Log Page 1 of 1 Name Approval to Exceed 10CFR20 Exposure limits

>5 rem >10 - <25 rem >25 rem TEDE TEDE TEDE WHEN COMPLETED, THIS RECORD SHALL BE RETAINED IN ACCORDANCE WITH THE DOCUMENT TYPE LIST (DTL).

0POP04-RC-0004 Steam Generator Tube Leakage Rev. 29 Page 5 of 116 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 6.0 MAINTAIN VCT Level - GREATER PERFORM the following:

THAN 15% WITH CHARGING PUMP a. TRIP the Reactor.

SUCTION ALIGNED TO VCT (CP004)

b. INITIATE Safety Injection.

x Auto makeup

c. GO TO 0POP05-EO-EO00, Reactor Trip x Manual makeup Or Safety Injection.

_____ 7.0 CHECK Main Turbine In Service GO TO Step 13.0.

_____ 8.0 CHECK For One Of The Following: PERFORM the following:

x Leakage From Any One SG Is a. COMPARE SG Tube Leak Rates to values GREATER THAN OR EQUAL TO 75 listed in Addendum 6.

gpd AND Continues To Increase At b. IF Mode 3 is REQUIRED by GREATER THAN OR EQUAL TO 30 Addendum 6, THEN PERFORM the gpd/hr following:

OR 1) COMMENCE plant shut down per x Leakage From Any One SG Is 0POP03-ZG-0006, Plant Shutdown GREATER THAN OR EQUAL TO 75 From 100% To Hot Standby, per gpd AND Loss of Continuous response time requirements of Radiation Monitoring Addendum 6.

2) GO TO Step 10.0.

c. IF Shutdown NOT required, THEN RETURN to procedure and step in effect.

_____ 9.0 PERFORM The Following:

_____ a. COMMENCE Plant Shutdown Per 0POP04-TM-0005, Fast Load Reduction Per The Response Time Requirements Of Addendum 6

_____ b. CONTINUE performance of 0POP04-TM-0005, Fast Load Reduction until TURBINE TRIPPED This Procedure is Applicable in Modes 1, 2, 3 and 4

0POP04-RC-0004 Steam Generator Tube Leakage Rev. 29 Page 34 of 116 Addendum 6 Recommended Response Times Addendum 6 Page 1 of 1 CAUTION Post Reactor Shutdown conditions in the primary to secondary leakage (RCS temperature and pressure decreasing and SG pressure increasing) may reduce the SG Tube Leakage Rate. Plant Shutdown and Cooldown Rates should be based on the initial or increased leakage and NOT reduced leakage due to the Post Shutdown conditions.

Action Leak Rate Increasing Response Times (1)

Level Leak Rate Reduce Rx PWR to

> 75 gpd Rate of inc > 30 gpd/hr < 50% in 1 hr 3 (4) AND Mode 3 in the next 2 hr

> 75 gpd Reduce Rx PWR to AND < 50% in 1 hr Loss of Continuous Radiation NA AND 3 Monitoring Mode 3 in the next 2 hr (3) 3 > 150 gpd NA Mode 3 < 6 hr

> 75 gpd 2 Mode 3 < 24 hr (4) Rate of inc < 30 gpd/hr

< 75 gpd AND Loss of Continuous Radiation Continued Operations N/A 1 Monitoring (2)

(3)

Continued Operations 1 > 30 gpd N/A (2)

Increased Continued Operations

> 5 gpd N/A Monitoring (2)

(1) Response times are the maximum times allowed. Power Reduction and Mode Change(s) may be completed in less time.

(2) Continued Operations per Plant Management direction. Refer to 0PGP03-ZO-0041, Action For Monitoring Primary to Secondary Leakage.

(3) Loss of Continuous Radiation Monitoring as defined in 0PGP03-ZO-0041, Action for Monitoring Primary to Secondary Leakage.

(4) With a continued increase in leakage rate over 30 minute time interval per the next column.

This Procedure is Applicable in Modes 1, 2, 3 and 4

D0527 SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION STI 33587455 0POP04-RS-0001 Rev. 35 Page 1 of 145 Control Rod Malfunction Quality Usage: IN HAND Enhanced Off-Normal Procedure Safety-Related CONTROLLING STATION Effective Date: 08/28/2012 D. Rohan P. Travis Crew 2E Generation Support PREPARER TECHNICAL USER COGNIZANT DEPT Table of Contents Purpose, Symptoms or Entry Conditions .................................................................................................. 2 References ................................................................................................................................................... 11 Addendum 1, Recovery of a Dropped Rod .............................................................................................. 13 Addendum 2, Recovery of Misaligned Rods............................................................................................ 26 Addendum 3, Insertion of Rods Which Fail to Fully Insert Following Reactor Trip or Shutdown .. 52 Addendum 4, Technical Specifications .................................................................................................... 57 Addendum 5, Rod Control In Manual While In Rod Realignment Process ........................................ 58 Addendum 6, Basis..................................................................................................................................... 63 Conditional Information Page ................................................................................................................ 145 This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 2 of 145 PURPOSE

1. This procedure provides instructions to stabilize the plant and recover from the following types of rod malfunctions during operation in Mode 1 or 2:

x Dropped or misaligned rod(s) x Immovable rod(s) x Spurious rod movement x Shutdown Rod not fully withdrawn

2. Provide instructions to recover dropped or misaligned rod(s) in Mode 3.

3. Provide instructions to recover from rod(s) which fail to fully insert following a Reactor Trip or plant shutdown.

4. The applicable portions of this procedure may be used to provide guidance in realigning rods that are less than 12 steps from the Group Step Counter Demand position at the discretion of the Shift Manager/Unit Supervisor.

SYMPTOMS OR ENTRY CONDITIONS

1. Dropped or Misaligned Rod

a. A deviation of greater than 12 steps between Digital Rod Position Indication (DRPI) and Group Step Counter Demand position indications for any rod.

b. A deviation between DRPI and Group Step Counter Demand position indications for any rod that requires realignment.

c. Rod Bottom LED on DRPI LIT.

d. Decreasing RCS Tavg.

e. Any of the following annunciators LIT:

x PR UPPER DET FLUX DEV HI/AUTO DEF Lampbox 5M03 Window A-3 x RPI TRBL Lampbox 5M03 Window A-5 x PR LOWER DET FLUX DEV HI/AUTO DEF Lampbox 5M03 Window B-3 x PR CHANNEL DEV Lampbox 5M03 Window C-3 x AXIAL FLUX DIFFERENCE HI Lampbox 5M03 Window D-3 x ROD SUPV MNTR ROD POSITION TRBL Lampbox 5M03 Window D-5 x ROD BOTTOM Lampbox 5M03 Window F-4 This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 3 of 145 SYMPTOMS OR ENTRY CONDITIONS (Continued)

2. Immovable Rod

a. Inability to insert and/or withdraw any control rod.

b. Misaligned rod.

c. ROD CONT URGENT ALARM LIT Lampbox 5M03 Window B-5

3. Spurious Rod Movement

a. Unexpected rod motion with ROD BANK SELECTOR SW in MANUAL or AUTO.

4. Shutdown Rod NOT Fully Withdrawn

a. ROD SUPV MNTR ROD POSITION TRBL LIT Lampbox 5M03 Window D-5

b. Any shutdown rod observed to be NOT withdrawn to the full out rod position.

5. Rod(s) fail to fully insert following a Reactor Trip or shutdown as indicated by absence of rod bottom light(s) lit.

This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 4 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Steps 1.0 through 3.0 are IMMEDIATE ACTION Steps.

_____ 1.0 ENSURE ROD BANK SEL Switch In MANUAL {CP005}

_____ 2.0 VERIFY All Rods - NO ROD MOTION PERFORM the following:

a. TRIP the Reactor.

b. GO TO 0POP05-EO-EO00, Reactor Trip Or Safety Injection.

_____ 3.0 CHECK For Dropped Rods:

_____ a. CHECK All Rods - ANY RODS a. GO TO Step 4.0.

DROPPED

_____ b. CHECK All Rods - ONLY ONE ROD b. IF in Modes 1 OR 2, THEN PERFORM DROPPED the following:

1) TRIP the Reactor.

2) GO TO 0POP05-EO-EO00, Reactor Trip Or Safety Injection.

_____ c. GO TO Addendum 1, Recovery of a Dropped Rod This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 5 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 4.0 CHECK For Misaligned Rods:

_____ a. CHECK All Rods - ANY RODS a. GO TO Step 5.0.

MISALIGNED

_____ b. CHECK All Rods - ONLY ONE ROD b. IF two or more rods are NOT aligned in MISALIGNED MODES 1 OR 2, THEN:

1) REFER TO Technical Specification 3.1.3.1 Action d for appropriate action.

2) COMMENCE load reduction per 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, to place the unit in Mode 3 within six hours of the time of misalignment.

3) GO TO Step 5.0.

_____ c. GO TO Addendum 2, Recovery of Misaligned Rods

_____ 5.0 CHECK Reactor Trip Breakers - CLOSED IF all rods have NOT fully inserted following a Reactor Trip or shutdown, THEN GO TO Addendum 3, Insertion of Rods Which Fail to Fully Insert Following Reactor Trip or Shutdown.

This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 6 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Following a control rod malfunction, I&C troubleshooting and diagnosis will be significantly enhanced if control rods are NOT moved. Therefore rods should NOT be moved unless rod motion is required to maintain Reactor control.

_____ 6.0 CHECK Tavg - WITHIN 1.5qF OF Tref MAINTAIN Tavg within 1.5qF of Tref by any of the following methods:

x ADJUST Turbine load x ADJUST RCS boron concentration x IF Turbine is offline, THEN ADJUST demand on the Steam Generator PORVs OR Steam Dumps.

NOTE x An Urgent Failure in the Logic Cabinet prevents all automatic and manual rod motion in overlap.

x An Urgent Failure in a Power Cabinet prevents all rod motion by the rods powered from the failed cabinet.

_____ 7.0 CHECK ROD CONT URGENT ALARM DISPATCH an Operator to determine alarm Lampbox 5M03 Window B status of rod control logic and power cabinets.

EXTINGUISHED {60 ft EAB Room 323}

_____ 8.0 NOTIFY I&C Personnel To Assist In Troubleshooting And Diagnosis Of The Malfunctioned Rod(s)

_____ 9.0 PERFORM Required Actions And Conditions Of Technical Specification 3.1.3.1 This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 7 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Step 10.0 may be performed concurrently with the rest of this procedure.

_____ 10.0 NOTIFY The Following Of The Rod Malfunction:

_____ x Reactor Engineering Supervisor

_____ x Plant Operations Manager NOTE Quadrant Power Tilt Ratio (QPTR) can be monitored on the Plant Computer by selecting the FT: Radial Flux Tilts display under Nuclear Application Programs.

_____ 11.0 MONITOR The Following At Least Once Every Hour:

x Quadrant Power Tilt Ratio (QPTR) x Axial Flux Difference (AFD)

_____ 12.0 CHECK QPTR And AFD - WITHIN PERFORM the following:

LIMITS OF TECHNICAL a. MONITOR the length of time that the SPECIFICATIONS limits of Technical Specification 3.2.1 or x 3.2.1 3.2.4 are exceeded.

x 3.2.4 b. REFER TO Technical Specifications 3.2.1 and 3.2.4 for appropriate action.

c. IF action statements of Technical Specifications 3.2.1 and 3.2.4 for power reduction can NOT be complied with, THEN PERFORM the following:

1) TRIP the Reactor.

2) GO TO 0POP05-EO-EO00, Reactor Trip Or Safety Injection.

This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 8 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE An immovable rod should NOT be considered untrippable unless I&C personnel have verified that the stationary gripper, movable gripper, and lift coils are being energized with the correct currents and in the proper sequence.

_____ 13.0 CHECK All Rods - TRIPPABLE PERFORM the following:

a. IF two or more rods are determined NOT trippable, THEN:

1) TRIP the Reactor.

2) GO TO 0POP05-EO-EO00, Reactor Trip Or Safety Injection.

b. IF only one rod is determined NOT trippable, THEN:

1) REFER TO Technical Specification 3.1.3.1 Action a for appropriate action.

2) COMMENCE load reduction per 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, to place the unit in Mode 3 within six hours of determining rod is NOT trippable.

_____ 14.0 CHECK DRPI - OPERABLE REFER TO Technical Specification 3.1.3.2 for appropriate action.

This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 9 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 15.0 CHECK All Rods - ALIGNED PERFORM the following:

a. IF two or more rods are NOT aligned in MODES 1 OR 2, THEN:

1) REFER TO Technical Specification 3.1.3.1 Action d for appropriate action.

2) COMMENCE load reduction per 0POP03-ZG-0006, Plant Shutdown From 100% To Hot Standby, to place the unit in Mode 3 within six hours of the time of misalignment.

b. IF only one rod is NOT aligned OR in Mode 3, THEN GO TO Addendum 2, Recovery of Misaligned Rods.

_____ 16.0 CHECK Rod Status - ALL MOVABLE IF any rod(s) is aligned AND NOT movable, THEN PERFORM the following:

a. MAINTAIN compliance with Rod Insertion Limits and overlap specified in the Core Operating Limits Report.

b. MAINTAIN affected bank(s) within 12 steps of the immovable rod(s).

c. IF the immovable rod becomes misaligned, THEN GO TO Addendum 2, Recovery of Misaligned Rods.

d. IF only one rod is immovable, THEN REFER TO Technical Specification 3.1.3.1 Action b for appropriate action.

e. IF two or more rods are immovable, THEN REFER TO Technical Specification 3.1.3.1 Action c for appropriate action.

This Procedure is Applicable in Modes 1, 2, and 3

0POP04-RS-0001 Control Rod Malfunction Rev. 35 Page 10 of 145 STEP ACTIONS/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_____ 17.0 CHECK All Shutdown Rods - FULLY IF two or more Shutdown Rods are NOT fully WITHDRAWN withdrawn, THEN REFER TO Technical Specifications 3.1.3.5 and 3.0.3 for appropriate action.

_____ 18.0 RECORD The Following In The Control Room Log:

_____ x Core location of malfunctioned rod(s)

_____ x Digital Rod Position Indication (DRPI) for malfunctioned rod(s)

_____ x Affected bank(s)

_____ x Group Step Counter Demand position of affected bank(s)

_____ x Type of malfunction (e.g., misaligned, immovable, etc)

_____ x Date and time malfunction occurred

- END -

This Procedure is Applicable in Modes 1, 2, and 3