perry nrc exam 2015 · perry nrc exam 2015 . question ro 1 . the plant is operating at 72% reactor...

Perry NRC Exam 2015 QUESTION RO 1 The plant is operating at 72% Reactor Power with 64 Mlbs/hr Core Flow, when the following occurs: • Reactor Recirculation B Pump tripped • ONI-C51 Unplanned Change in Reactivity or Power has been entered • Reactor Power initially lowered to 40%, and is varying between 34% and 46% The oscillations in Reactor Power are a result of ____.

A. thermal hydraulic instability resulting from High Rod Line and Low Core Flow

B. core flow variation resulting from unstable vortex formation in the Recirculation Loop A Risers

C. jet pump cavitation resulting from the increased Reactor Recirculation Loop A Flow and reduced Feedwater Flow

D. the combined effects of lowering RPV Water Level and Feedwater Temperature following a sudden power reduction

1

Perry NRC Exam 2015

QUESTION RO 1

Examination Outline Cross-Reference

Level: RO SRO Tier # 3 Group # K/A# Generic 2.1.7 Importance Rating 4.4

K&A: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Generic Explanation: Answer A – Oscillations > 10% in Reactor Power are indicative of Thermal Hydraulic Instability

B – Incorrect - plausible, <10% peak to peak variation in APRM Power, at rated power, is associated with Bistable Core Flow. At reduced power conditions, APRM Power variation due to Bistable Core Flow is NOT expected.

C – Incorrect - plausible, When B Recirculation Pump tripped, BOTH Feedwater Flow lowered AND A Recirculation Loop Flow increased. Cavitation Downshift will protect Recirculation Pumps at <22% Feedwater Flow for 15 seconds. This value is not reached by power lowering to 34%.

D – Incorrect - plausible, reduction in RPV Level (from steam programming) and a reduction in Feedwater Temperature (from lower Turbine Load) occur as a result of a downpower, but this will not result in a 12% variation in Reactor Power.

Technical Reference(s): ONI-C51 Chart Rev & PDB-A6 Rev 14

Reference Attached: ONI-C51 Chart (partial) & PDB-A6 p 3

Proposed references to be provided to applicants during examination: None

Learning Objective (As available): OT-Combined-C51AP_OPRM M

Question Source: Bank # Perry 2009 # RO-01 Modified Bank # New

Question History: Previous NRC Exam

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x

10 CFR Part 55 Content: 55.41 x 55.43

Comments: Level of Difficulty = x

2

Perry NRC Exam 2015 QUESTION RO 2 Refer to the figure below to answer this question. Per NOP-OP-1002, Conduct of Operations, who may authorize non-shift personnel to enter the Control Panel Restricted Access Area?

A. Any member of the on-duty control room staff

B. The RO-ATC or RO-BOP only

C. The Unit Supervisor only

D. The Shift Manager only

3

Perry NRC Exam 2015

QUESTION RO 2



K&A: Knowledge of facility requirements for controlling vital/controlled access.

Generic Explanation: Answer B – Per NOP-OP-1002, either the ATC or BOP can authorize entry into the Control Panel Restricted Access Area.

A – Incorrect – This is true for the Controls Area.

C & D – Incorrect – the ATC or BOP will authorize entry.

Technical Reference(s): NOP-OP-1002 Rev 9 Reference Attached: NOP-OP-1002 pp 41 & 96


Learning Objective (As available): OT-3039-01-G

Question Source: Bank # Grand Gulf 2010 # RO-03 Modified Bank # New


Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis



4

Perry NRC Exam 2015 QUESTION RO 3 RHR Loop ‘B’ is being placed in the Shutdown Cooling mode in accordance with IOI-11, Shutdown From Outside the Control Room. Which of the following describes the operator action required to position RHR B HX’S OUTLET VALVE, 1E12 F003B, for this evolution? RHR B HX’S OUTLET VALVE, 1E12 F003B, is manipulated using its control switch located at ____.

A. MCC disconnect EF1D07-D without requiring the use of a Transfer Switch on the Division 2 Remote Shutdown Panel, 1C61-P002

B. MCC disconnect EF1D07-D only after a Transfer Switch is placed in EMERG on the Division 2 Remote Shutdown Panel, 1C61-P002

C. the Division 2 Remote Shutdown Panel without requiring the use of a Transfer and Control Switch on the Division 2 Remote Shutdown Panel, 1C61-P002

D. the Division 2 Remote Shutdown Panel only after a Transfer and Control Switch is placed in EMERG on the Division 2 Remote Shutdown Panel, 1C61-P002

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Perry NRC Exam 2015

QUESTION RO 3



K&A: Ability to locate and operate components, including local controls.

Generic Explanation: Answer A – 1E12 F003B is controlled from MCC EF1D07 D. The Transfer Switch is also located at the MCC not the RSD panel.

B – Incorrect – Remote S/D operation of 1E12-F003B does not require manipulation of any switches on the Div 2 RSP.

C & D – Incorrect - 1E12 F003B is not controlled from the Div 2 RSP and Remote S/D operation of 1E12 F003B does not require manipulation of any switches on the Div 2 RSP.

Technical Reference(s): IOI-11 Rev 31 Reference Attached: IOI-11 pp 3-4, 10, 75-76


Learning Objective (As available): OT-Combined-C61-F.2

Question Source: Bank # Perry 2001 # 71 Modified Bank # New





6

Perry NRC Exam 2015 QUESTION RO 4 Technical Specification Rounds perform surveillances that require CHANNEL CHECKS. Which of the following is the definition of a CHANNEL CHECK, in accordance with Technical Specifications?

A. A test of all required components of a logic circuit, from as close to the sensor as practicable up to the actuated equipment, to verify operability.

B. A comparison of a channel's indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

C. The adjustment of a channel's output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors.

D. The inputting of a simulated or actual signal as close to a channel's sensor as practicable to verify operability, including alarms, interlocks, displays, and trip functions.

7

Perry NRC Exam 2015

QUESTION RO 4



K&A: Knowledge of surveillance procedures.

Generic Explanation: Answer B – Per TS 1.1, Definitions, page 1.0-1: A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A – Incorrect – This is the definition for a Logic System Functional Test.

C – Incorrect – This is the definition for a Channel Calibration

D – Incorrect – This is the definition for a Channel Functional Test.

Technical Reference(s): TS 1.1 Reference Attached: TS 1.1 p 1.01-.02, & .0-4


Learning Objective (As available): OT-3037-07-A.2

Question Source: Bank # INL-0956 Modified Bank # New





8

Perry NRC Exam 2015 QUESTION RO 5 The plant was operating at 50% rated power when a transient resulted in a reduction of both reactor power and reactor pressure. Which of the following combinations of reactor power and reactor pressure indicate that a Safety Limit violation occurred?

Reactor Power Reactor Pressure

A. 10% 750 psig

B. 20% 770 psig

C. 30% 775 psig

D. 35% 810 psig

9

Perry NRC Exam 2015

QUESTION RO 5



K&A: Knowledge of conditions and limitations in the facility license.

Generic Explanation: Answer C – Per TS 2.1.1.1: With the Reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: Thermal Power shall be </= 23.8% RTP.

A – Incorrect – This combination does not violate SL 2.1.1.1.

B – Incorrect – This combination does not violate SL 2.1.1.1.

D – Incorrect – With power ≥ 10% rated flow and pressure ≥ 785 psig, MCPR is the limiting Safety Limit.

Technical Reference(s): TS 2.2.1 Reference Attached: TS 2.2.1 p 2.0-1


Learning Objective (As available): x

Question Source: Bank # INL-2108 Modified Bank # New





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Perry NRC Exam 2015 QUESTION RO 6 Which of the following is an acceptable method to alert the Operator of Control Room annunciators that have been removed from service?

A. Minor Deficiency Monitoring (MDM) Tag

B. Temporary Modification Tag

C. Information Tag

D. Danger Tag

11

Perry NRC Exam 2015

QUESTION RO 6



K&A: Knowledge of the process used to track inoperable alarms.

Generic Explanation: Answer C – IAW PAP-1404, Info tags or Caution tags are to be used to identify Control Room annunciators that are removed from service.

A – Incorrect – The MDM Process is for the management of maintenance deficiencies whose significance is so minor that it would not be prudent to remove the equipment from service to repair. Not tracking of annunciators. Plausible if operator confuses these tags with Repair Tags.

B – Incorrect – Although the Temp Mod procedure controls annunciators removed from service, TM tags are not used. Additionally, Not-in-Service stickers are no longer allowed to be used to identify OOS annunciators in the Control Room. Plausible if operator not very familiar with TM procedure.

D – Incorrect – Although the Caution Tags can be used to track annunciators removed from service, Danger Tags are not used. Plausible if operator not very familiar with tagging procedure.

Technical Reference(s): PAP-1404 Rev 7 & NOP-OP-1010 Rev 3

Reference Attached: PAP-1404 p 4 & NOP-OP-1010 p 25



Question Source: Bank # Modified Bank # New x





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Perry NRC Exam 2015 QUESTION RO 7 What is the lowest radiation exposure that would allow the Shift Manager to waive the Independent Verification / Concurrent Verification of a component?

A. 9 mrem

B. 11 mrem

C. 16 mrem

D. 21 mrem

13

Perry NRC Exam 2015

QUESTION RO 7



K&A: Knowledge of radiation exposure limits under normal or emergency conditions.

Generic Explanation: Answer B – Per NOP-OP-1002, the SM can waive IV/CV requirements if dose exceeds 10 mrem.

A – Incorrect – This dose is below the 10 mrem for a waiver of CV/IV

C & D – Incorrect – These are not the lowest doses for waiving CV/IV.

Technical Reference(s): NOP-OP-1002 Rev 9 Reference Attached: NOP-OP-1002 p 82



Question Source: Bank # Perry 2007-2 # RO-71 Modified Bank # New





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Perry NRC Exam 2015 QUESTION RO 8 The following conditions exist: • The plant is in a refueling outage with core alterations in progress. • Troubleshooting is in progress on the Traversing In-core Probe (TIP) Drive Indexer Mechanism ‘D’

because it is stuck. The TIP Indexer is then unstuck and the TIP detector is being inserted and withdrawn. Which of the following hazards is created by this action?

A. High airborne contamination level in Containment due to TIP purging

B. High radiation exposure to personnel on the 360 degree platform

C. Damage to the TIP mechanism if fuel movement is in progress

D. High radiation exposure to personnel in Containment 599’

15

Perry NRC Exam 2015

QUESTION RO 8



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

Generic Explanation: Answer D – Per FTI-A001, operation of the TIP system may cause a very high radiation area in the vicinity of the TIP drive boxes, which are in containment on the 599’ elevation.

A – Incorrect – The indexer enclosure is an air-tight unit. No air escapes into containment to cause high airborne contamination.

B – Incorrect – Personnel on the 360° platform are shielded by the upper pool.

C – Incorrect – The TIP probes are located within the LPRM tubes. No damage will occur due to fuel movement.

Technical Reference(s): FTI-A001 Rev 11, Survey Map –Containment 599’, SDM C51 (TIP) Rev 7

Reference Attached: FTI-A001 pp 3 & 5, Survey Map –Containment 599’, SDM C51 (TIP) pp 5 & 29


Learning Objective (As available): OT-Combined-C51-TIPS-D.1 & -D.2

Question Source: Bank # Nine Mile-2 2009 # RO-71 Modified Bank # New





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Perry NRC Exam 2015 QUESTION RO 9 The plant was at rated power when a fire was detected in Reactor Recirculation Pump A. At 13:58 the CO2 system automatically initiated. The CNTMT CO2 SUPPLY OUTBOARD ISOL VALVE, 1P54-F340, was opened by the Control Room Operator at 14:40. Which of the following describes the current status of the CO2 system? CO2 for the Reactor Recirculation Pump fire was ____.

A. released into the Drywell and was discharged for the required amount of time

B. released into the Drywell and was not discharged for the required amount of time

C. not released into the Drywell; therefore, a Drywell entry will be required to suppress the fire

D. not released into the Drywell; therefore, the CO2 System will need to be manually initiated

17

Perry NRC Exam 2015

QUESTION RO 9



K&A: Knowledge of “fire in the plant” procedures.

Generic Explanation: Answer D – Since P54-F340 was not opened for 42 minutes after CO2 automatically initiated, no discharge occurred. The system will need to be manually initiated.

A – Incorrect – The discharge timer is set for 40 minutes. The containment isolation valve was not opened for 42 minutes. Therefore, no CO2 was released into the drywell.

B – Incorrect – No CO2 was released into the drywell. Discharge timer timed out prior to opening the containment isolation valve.

C – Incorrect – No drywell entry is required to suppress fire. The CO2 can be manually initiated to suppress the fire.

Technical Reference(s): ONI-P54 Rev 18 Reference Attached: ONI-P54 pp4-6


Learning Objective (As available): OT-Combined-P54-CO2-J.1






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Perry NRC Exam 2015 QUESTION RO 10 You, the Field Supervisor are currently performing an Observation of NLO’s in Control Complex 574’. Then, the Shift Manager entered the Emergency Plan and Site accountability was been initiated in accordance with EPI-B5, Personnel Accountability/Site Evacuation. Which of the following describes your responsibility to meet site accountability requirements?

A. Report to the Unit 1 or 2 Control Room and use your key card in the Accountability Card Reader.

B. Remain inside the plant and use your key card in the nearest Plant Card Reader.

C. Verbally report your location directly to the Secondary Alarm Station (SAS).

D. Verbally report your location directly to the ATC-RO.

19

Perry NRC Exam 2015

QUESTION RO 10



K&A: Knowledge of RO responsibilities in emergency plan implementation.

Generic Explanation: Answer A – IAW EPI-B5, Control Room personnel must report to the unit 1 or unit 2 control rooms and use the accountability card reader.

B – Incorrect – EPI-B5 does not have a provision for using Plant Card Readers (NOP-LP-1207, 4.7) other than in Unit 1or 2 control rooms for control room staff.

C – Incorrect – Although the SAS is instrumental in coordinating Site Accountability, EPI-B5 directs the use of the accountability card readers in unit 1 or 2 control rooms.

C – Incorrect – The Shift Manager initiates accountability, but EPI-B5 directs the use of the accountability card readers in unit 1 or 2 control rooms. Locations should not be reported to the ATC-RO.

Technical Reference(s):EPI-B05 Rev 16 Reference Attached: EPI-B05 pp 6 & 15


Learning Objective (As available): EPL-0804-01

Question Source: Bank # Perry 2002 # SRO-83 Modified Bank # New





20

Perry NRC Exam 2015 QUESTION RO 11 The plant was operating at rated power when Reactor Recirculation Pump ‘B’ tripped. The current conditions are as follows: • Reactor power is 63% • Core flow is 41 Mlbm/hr What action is required? Reference Provided: PDB-A06 pp 3-7

A. Insert control rods in reverse order using the pull sheets

B. Raise core flow using the ‘A’ Flow control valve

C. Insert a manual scram

D. Insert Cram Rods

21

Perry NRC Exam 2015

QUESTION RO 11


Level: RO SRO Tier # 1 Group # 1 K/A# 295001 2.4.11 Importance Rating 4.0

K&A: Knowledge of abnormal condition procedures.

Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 Explanation: Answer D – With OPRM’s operable, using the correct P/F map would require the use of Cram Rods to exit the Controlled Entry/Immediate Exit Region per ONI-C51 Immediate Actions.

A – Incorrect – Inserting control rods in reverse order would require stopping at Insert Limit for each step.

B – Incorrect – Raising core flow is acceptable method to exit the Controlled Entry/Immediate Exit Region only if both Recirc Pumps are operating.

C – Incorrect – This would be correct action if OPRM’s were inoperable. The stem does not identify the OPRM’s as inoperable.

Technical Reference(s): PDB-A06 Rev 14, ONI-C51 Chart Rev J

Reference Attached: PDB-A06 pp 3-7 & ONI-C51 Chart (partial)

Proposed references to be provided to applicants during examination: PDB-A06 pp 3-7

Learning Objective (As available): OT-Combined-C51-AP-OPRM-J.2






22

Perry NRC Exam 2015 QUESTION RO 12 A Station Blackout (SBO) occurred ten minutes ago and the following plant conditions now exist: • Reactor pressure 830 psig and stable • RCIC automatically started and is injecting to maintain reactor level stable at 185” • All rods at position 00 Which of the following should be performed?

A. Within 20 minutes, cycle SRVs while maintaining a cooldown rate of <100°F/hr

B. Within 20 minutes, throttle Main Turbine Bypass Valves while maintaining a cooldown rate of <100°F/hr

C. Immediately open at least 5 SRVs to rapidly lower reactor pressure to 0 psig, irrespective of cooldown rate

D. Immediately open Main Turbine Bypass Valves to rapidly lower reactor pressure to 0 psig, irrespective of cooldown rate

23

Perry NRC Exam 2015

QUESTION RO 12


Level: RO SRO Tier # 1 Group # 1 K/A# 295003 AK1.06 Importance Rating 3.8

K&A: Knowledge of the operational implications of the following concepts as they apply to Partial Or Complete Loss Of A.C. Power: Station blackout: Plant-Specific

Partial or Complete Loss of AC / 6 Explanation: Answer A – Since RCIC auto started, the plant would enter EOP-1 on Level 2. A new Time Critical Operator Action requires operators to commence a cooldown near 100°F/hr within 30 minutes of when MSIV’s are closed. A station Blackout would result in MSIV closure.

B – Incorrect – During a Station Blackout, the bypass valves would not be available.

C – Incorrect – No conditions were given that would require an Emergency Depressurization.

D – Incorrect - No conditions were given that would require anticipating Emergency Depressurization and during a Station Blackout, the bypass valves would not be available

Technical Reference(s): EOP-1 Bases Rev 4 Reference Attached: EOP-1 Bases pp 81-82


Learning Objective (As available): OT-3402-04-AF






24

Perry NRC Exam 2015 QUESTION RO 13 In the event of a loss of the battery chargers, the Unit 1, Division 1 and 2 Batteries are sized to maintain voltage to required loads at >105 VDC for how long? Division 1 battery is sized for a minimum time of (1) hours. Division 2 battery is sized for a minimum time of (2) hours.

(1) (2)

A. 2 2

B. 2 4

C. 4 2

D. 4 4

25

Perry NRC Exam 2015

QUESTION RO 13



K&A: Knowledge of the interrelations between Partial Or Complete Loss Of D.C. Power and the following: Battery charger

Partial or Total Loss of DC Pwr / 6 Explanation: Answer A – Both Div 1 & 2 Batteries are sized to maintain > 80% nameplate rating (105 VDC) for two hours on the loss of a battery charger.

B – Incorrect – The greater time for Div 2 is plausible if they recall the rating for the non-divisional ‘B’ battery.

C – Incorrect – The greater time for Div 1 is plausible since the Div 1 battery has more load (RCIC) and more cells.

D – Incorrect – Both Divisional batteries are sized for 2 hours. Plausible if they mistake this for the rating for the non-divisional ‘B’ battery.

Technical Reference(s): SDM OT-Combined-R42 Rev 9 and Lesson Plan OT-Combined-R42 Rev 2

Reference Attached: SDM OT-Combined-R42 p 11 and Lesson Plan OT-Combined-R42 p 72


Learning Objective (As available): OT-Combined-R42-D






26

Perry NRC Exam 2015 QUESTION RO 14 Plant startup following a refuel outage is in progress with the following conditions. • Reactor power is 25% • Generator is synchronized to the grid • RFPT A is feeding the RPV • Control rod withdrawal to raise reactor power is planned to recommence in 5 minutes Then, annunciator H13-P870-009-G2, EHC STBY PUMP START-HEADER PRESSURE LOW alarms. EHC pressure lowered to 1050 psig prior to standby EHC pump starting. Based on this information, the main generator is (1) synchronized to the grid and feedwater inlet temperature is (2) .

(1) (2)

A. still stable

B. still lowering

C. no longer stable

D. no longer lowering

27

Perry NRC Exam 2015

QUESTION RO 14



K&A: Knowledge of the reasons for the following responses as they apply to Main Turbine Generator Trip: Feedwater temperature decrease

Main Turbine Generator Trip / 3 Explanation: Answer D – When EHC pressure drops < 1100 psig, the turbine trips and trips the main generator. When the turbine trips, extraction steam to all feedwater heaters is isolated. Therefore, final feedwater temperature will lower.

A & B – Incorrect – First part – The generator is no longer synchronized to the grid as the turbine has tripped.

A & C – Incorrect – Second part – Feedwater temperature will lower as extraction steam is isolated to all feedwater heaters.

Technical Reference(s): ARI-H13-P870-09-G2 Rev 10, ONI-N32 Rev 11, ONI-N36 Rev 18

Reference Attached: ARI-H13-P870-09-G2 p 29, ONI-N32 pp 3-5 & 12, ONI-N36 p 4


Learning Objective (As available): OT-3035-09(LP)-B.2 & OT-3035-13(LP)-A.2






28

Perry NRC Exam 2015 QUESTION RO 15 The plant automatically scrammed from rated power one minute ago. The following conditions currently exist: • HPCS and RCIC automatically started • Reactor level is 168 inches and rising • Reactor power is 3% and lowering • Reactor Recirculation Pump A is OFF • Reactor Recirculation Pump B is running in FAST What is your required action?

A. Transfer Reactor Recirculation Pump B to SLOW.

B. Start Reactor Recirculation Pump A in SLOW.

C. Start Reactor Recirculation Pump A in FAST.

D. Trip Reactor Recirculation Pump B to OFF.

29

Perry NRC Exam 2015

QUESTION RO 15


Level: RO SRO Tier # 1 Group # 1 K/A# 295006 AA1.04 Importance Rating 3.1

K&A: Ability to operate and/or monitor the following as they apply to Scram: Recirculation system

SCRAM / 1 Explanation: Answer D – With RPV water level lowering to L2, HPCS & RCIC automatically initiate. Also at L2, Reactor Recirc pumps should trip to OFF. Since B pump is still running in FAST, the correct action per NOP-OP-1002 is to trip B pump to OFF since it did not trip automatically.

A – Incorrect – This action is directed in the scram Hardcard. However, the corrective action for the auto action failing to occur is higher priority.

B – Incorrect – This would be correct for a loss of SDC if no Recirc pumps were running.

C – Incorrect – Recirc pumps are always started using the FAST speed breakers, then transferred to SLOW.

Technical Reference(s): ONI-C71-1 Rev 18, SDM-C22 Rev 6, ARI-H13-P601-16 Rev 17, ARI-H13-P601-21 Rev 15, NOP-OP-1002 Rev 9

Reference Attached: ONI-C71-1 p 5, SDM-C22 p 2, ARI-H13-P601-16 p 35, ARI-H13-P601-21 p 13, NOP-OP-1002 p 61


Learning Objective (As available): OT-COMBINED-C22-I1 & OT-COMBINED-B33-E4






30

Perry NRC Exam 2015 QUESTION RO 16 The plant was operating at rated power when control room was evacuated due to a fire.

All ONI-C61, Evacuation Of The Control Room, Immediate Actions were complete.

Cooldown is in progress IAW IOI-11, Shutdown From Outside The Control Room.

SPDS is not available.

The following data was recorded for IOI-11, Attachment 7, Reactor Cooldown Log.

Time RPV Pressure (psig) 15:15 827 15:45 775 16:15 671 16:45 527 17:15 437 17:45 359 18:15 281 18:45 125

What was the highest cooldown rate achieved in any one hour period? Reference Provided: Steam Tables

A. ~59.3°F/hr

B. ~63.2°F/hr

C. ~85.2°F/hr

D. ~103.8°F/hr

31

Perry NRC Exam 2015

QUESTION RO 16



K&A: Ability to determine and/or interpret the following as they apply to Control Room Abandonment: Cooldown rate

Control Room Abandonment / 7 Explanation: Answer C – By calculation – cooldown from 17:45 to 18:45 was ~438.1°F – ~352.9°F = 85.2°F/hr.

A – Incorrect – This is the rate from 16:45 to 18:15.

B – Incorrect – This is the rate from 18:15 to 18:45.

D – Incorrect – This is the rate from 17:15 to 18:45

Technical Reference(s): Steam Tables, IOI-11 Rev 31 Reference Attached: Worksheet from steam tables, IOI-11 pp 11, 13, 59


Learning Objective (As available): OT-3037-08-B






32

Perry NRC Exam 2015 QUESTION RO 17 The following conditions exist: • It is day 20 of 1R14 outage • Fuel shuffle is complete • Head Tensioning is in progress • Upper Pool gates are installed • Lower Pool gates are removed • Div 2 electrical outage in progress • Spent Fuel Pool temperature is 90°F NCC A pump then trips on a ground fault. NCC C pump cannot be started. Based on the above conditions, determine time to boil in the Spent Fuel Pool? Reference Provided: PBD-A016 & PDB-A017

A. 19 hrs

B. 25 hrs

C. 50 hrs

D. 170 hrs

33

Perry NRC Exam 2015

QUESTION RO 17



K&A: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Partial or Total Loss of CCW / 8 Explanation: Answer C – FPCC provides cooling to the Spent Fuel Pools and is cooled by the NCC system. The decay heat load on day 20 in the Fuel Pool is ~10.25 MBTU/hr based on curves for decay after adding 248 bundles. With SFP water temp initially at 90°F, the time to boil is approximately 50 hours. With Div 2 outage in progress, B NCC pump is not available. With head tensioning in progress, there is no communication between the upper and lower.

A – Incorrect – Plausible if use decay heat curve for 748 bundles discharge

B – Incorrect – Plausible if use time to get to 150°F and correct decay heat curve

D – Incorrect – Plausible if use correct time to boil and decay heat curve prior to adding discharged bundles.

Technical Reference(s): PBD-A016 Rev 13 & PDB-A017 Rev 11

Reference Attached: PBD-A016 p 7 & PDB-A017 p 19

Proposed references to be provided to applicants during examination: PBD-A016 & PDB-A017

Learning Objective (As available): OT-3035-11(LP)-A.1

Question Source: Bank # Modified Bank # Perry 2010 # RO-18 New





34

Perry NRC Exam 2015 QUESTION RO 18 An Instrument air leak in the Aux Building necessitated isolating Instrument Air in the Aux Building. What is the impact of this Instrument Air loss on the Residual Heat Removal System? The capability to ____ is lost.

A. sample RHR A heat exchangers

B. remotely place RHR A in Alternate Keep-fill

C. place RHR B in Fuel Pool Cooling Assist mode

D. pump down the suppression pool with RHR B to Radwaste

35

Perry NRC Exam 2015

QUESTION RO 18



K&A: Knowledge of the interrelations between Partial Or Complete Loss Of Instrument Air and the following: RHR/LPCI: Plant-Specific

Partial or Total Loss of Inst. Air / 8 Explanation: Answer B – Valve 1E12-F300A is opened to place RHR A in Alt Keep-fill and is an AOV. Remote operation of F300 requires instrument air.

A – Incorrect – The RHR A HX sample valves E12-F060A and E12-F075A are solenoid operated valves – no air required.

C – Incorrect – Fuel pool cooling assist requires operation of MOV’s and manually operated valves only.

D – Incorrect – Pumping down the SP to Radwaste only requires operation of MOV’s and manually operated valves in the Aux Building.

Technical Reference(s): SOI-E12 Rev 63, DWGs 302-642 Rev JJ, 302-243 Rev FF, 816-706 Rev N, & 208-013 Sh 9 Rev V

Reference Attached: SOI-E12 p 52, DWGs 302-642, 302-243, 816-706, & 208-013 Sh 9


Learning Objective (As available): OT-COMBINED-P51_52-J.4






36

Perry NRC Exam 2015 QUESTION RO 19 The plant is shutdown for a forced outage following a 325 day run. Reactor coolant temperature is 180°F. The following events then occur: • A valve misalignment results in RPV water level lowering to 126 inches. • Two minutes later bus EH11 experiences a ground fault. Which of the following methods of decay heat removal is available?

A. Start a Reactor Recirculation Pump

B. Place RHR B loop in Shutdown Cooling

C. Inject with RHR C and open SRV’s to feed and bleed the RPV

D. Reject heat through the RWCU non-regenerative heat exchangers

37

Perry NRC Exam 2015

QUESTION RO 19



K&A: Knowledge of the reasons for the following responses as they apply to Loss Of Shutdown Cooling: Feeding and bleeding reactor vessel

Loss of Shutdown Cooling / 4 Explanation: Answer C – With the listed conditions, feed and bleed is the only method available to remove decay heat. Since power is lost to the EH11 Bus, using RHR C is the preferred method.

A – Incorrect – Starting a Recirc pump is an action listed in ONI-E12-2, but it is for circulating water to obtain a true RPV water temperature. The Recirc pump adds heat, it does not remove heat.

B – Incorrect – RHR B can not be placed in SDC since the RPV level is < L3.

C – Incorrect – Rejecting heat through the RWCU NRHX’s will not work as RWCU will isolate on RPV L2 (130”)

Technical Reference(s): ONI-E12-2 Rev, ARI-H13-P601-19 Rev 17, ARI-H13-P680-05 Rev 15

Reference Attached: ONI-E12-2 pp 21-22,24-30, ARI-H13-P601-19 p39, ARI-H13-P680-05 p 27



Question Source: Bank # Nine Mile 2 2009 - # RO-47 Modified Bank # New





38

Perry NRC Exam 2015 QUESTION RO 20 Core alterations are in progress with M14 Containment Vessel and Drywell Purge System operating in the Refuel Mode. The following then occurs: • A HIGH alarm on the GAS channel for 1D17-K686, Containment Atmosphere Radiation Monitor

is received • Annunciator CNTMT VENT EXH RAD HI (H13-P680-07-B3) alarms • 1D17-K609A through D are reading 2.0 mr/hr The Refueling Supervisor informs the Control Room that bubbles are observed while withdrawing a fuel bundle from the core. Entry into (1) is required and the operator will (2) .

(1) (2)

A. only ONI-J11-2, Fuel Handling Accidents

confirm that M14 Containment Vessel and Drywell Purge has isolated

B. ONI-J11-2, Fuel Handling Accidents and ONI-D17, High Radiation Levels Within Plant

confirm that M14 Containment Vessel and Drywell Purge has isolated

C. only ONI-J11-2, Fuel Handling Accidents

need to manually activate the Containment Evacuation Alarm

D. ONI-J11-2, Fuel Handling Accidents and ONI-D17, High Radiation Levels Within Plant

need to manually activate the Containment Evacuation Alarm

39

Perry NRC Exam 2015

QUESTION RO 20



K&A: Ability to operate and/or monitor the following as they apply to Refueling Accidents: †Containment building ventilation: Mark-III

Refueling Acc / 8 Explanation: Answer B – For these alarms, both ONI-D17 and ONI-J11-2 need to be entered. Also, radiation levels > 1.2 mr/hr will automatically isolate the M14 system.

A – Incorrect – ONI-D17 also needs to be entered.

C – Incorrect – ONI-D17 also needs to be entered and the containment evacuation alarm will automatically initiate on a HIGH alarm on D17-K686.

D – Incorrect – The containment evacuation alarm will automatically initiate on a HIGH alarm on D17-K686.

Technical Reference(s): ONI-D17 Rev 18, ONI-J11-2 Rev 15, ARI-H13-P680-07 Rev 24, & ARI-H13-P804 Rev 3, Dwgs 208-050 sh 5 Rev J, & 806-007 Rev L

Reference Attached: ONI-D17 p 3, ONI-J11-2 p 3, ARI-H13-P680-07 p 27, & ARI-H13-P804 pp 29-30, Dwgs 208-050 sh 5, & 806-007


Learning Objective (As available): OT-3035-14(LP)-A.4 & A.5






40

Perry NRC Exam 2015 QUESTION RO 21 The plant was operating at rated power when an Instrument Air leak in the Drywell caused Drywell pressure to increase to 2.5 psig. Additionally, the Parallel Instrument Air header pressure lowered to 45 psig. Based on this information, radiological conditions in the (1) will (2) .

(1) (2)

A. Turbine Building remain the same

B. Turbine Building degrade

C. Containment remain the same

D. Containment degrade

41

Perry NRC Exam 2015

QUESTION RO 21


Level: RO SRO Tier # 1 Group # 1 K/A# 295024 EA2.07 Importance Rating 3.4

K&A: Ability to determine and/or interpret the following as they apply to High Drywell Pressure: Containment radiation levels: Mark-III

High Drywell Pressure / 5 Explanation: Answer D – Drywell pressure > 1.68 psig will require entry into EOP-2 and initiate an automatic scram. When Instrument air pressure lowers to < 48 psig, the inboard MSIV’s will shut. This will require pressure control on the SRV’s. Radiation levels in the containment will increase due to SRV operation.

A – Incorrect – Turbine Building radiation levels will actually lower following an MSIV isolation event. When the plant is shutdown, the Turbine Building is not a posted as a locked high rad area.

B – Incorrect – Turbine Building radiation levels will actually lower following an MSIV isolation event. When the plant is shutdown, the Turbine Building is not a posted as a locked high rad area.

C – Incorrect – The containment radiation levels will increase due to SRV operation.

Technical Reference(s): ONI-P52 Rev 17, EOP-2 Bases Rev 2, & VSDS Radiation surveys, SOI-B21 Rev 17

Reference Attached: ONI-P52 p 28, EOP-2 Bases p 8, VSDS Radiation surveys, SOI-B21 p 4








42

Perry NRC Exam 2015 QUESTION RO 22 The unit was operating at 90% rated power when the following occurred: • Both SB&PR pressure regulators have failed • Annunciator RPS RX PRESS HI (H13-P680-05-A8) alarmed • Annunciator RX PRESS HI (H13-P680-07-D1) alarmed • Reactor pressure is 1065 psig and rising • No automatic actions have occurred Which operator action is required to lower reactor pressure?

A. Lower Load Set

B. Lower the Max Combined Flow Limiter

C. Place the Mode Switch in SHUTDOWN

D. Close Reactor recirculation FCV’s until core flow is approximately 58 Mlbm/Hr

43

Perry NRC Exam 2015

QUESTION RO 22



K&A: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

High Reactor Pressure / 3 Explanation: Answer C – The Immediate operator action for pressure regulator failure is to place mode switch in shutdown.

A – Incorrect – Lowering on Load Set will not lower Rx pressure.

B – Incorrect – Lowering on the Max Combined Flow Limiter is an action if the pressure regulator fails open vs. closed, and the automatic scram setpoint for Rx pressure has been exceeded. Therefore, taking manual action is required.

D – Incorrect – Lowering Recirc flow to 58 Mlbm/hr is an Immediate Action for other ONI’s, but not this situation.

Technical Reference(s): ONI-C85 Rev 1, ARI-H13-P680-05 Rev 15, ARI-H13-P680-07 Rev 24

Reference Attached: ONI-C85 p 3-5, ARI-H13-P680-05 p 19, ARI-H13-P680-07 p 91



Question Source: Bank # Modified Bank # Limerick 2002 # RO-73 New





44

Perry NRC Exam 2015 QUESTION RO 23 Compared to its normal heat capacity, the ability of the Suppression Pool to condense steam from RCIC during an extended Station Blackout is reduced by which of the following initial conditions?

A. Reactor pressure of 900 psig

B. Suppression Pool level of 18.6’

C. Suppression Pool Temperature of 96°F

D. A stuck closed Containment Vacuum Breaker

45

Perry NRC Exam 2015

QUESTION RO 23


Level: RO SRO Tier # 1 Group # 1 K/A# 295026 EK1.02 Importance Rating 3.5

K&A: Knowledge of the operational implications of the following concepts as they apply to Suppression Pool High Water Temperature: Steam condensation

Suppression Pool High Water Temp. / 5 Explanation: Answer C – Elevated initial Suppression Pool temperature of 96°F is > normal value and decreases the temperature difference to SP saturation temperature. This will decrease the margin to the Heat Capacity Limit and inhibit the ability sooner to condense steam.

A – Incorrect – Rx press of 900 psig is lower than normal value and increases margin to HCL.

B – Incorrect – SP level of 18.6’ is greater than normal and increases margin to HCL.

D – Incorrect – This could decrease margin to PSP not HCL.

Technical Reference(s): EOP-2 Rev 2 Reference Attached: EOP-2 p 60


Learning Objective (As available): OT-3402-06-C.1

Question Source: Bank # Quad Cities 2009 # RO-20 Modified Bank # New





46

Perry NRC Exam 2015 QUESTION RO 24 Which of the following lists the order of preference (from most preferred to least preferred) for indications to be used when determining Containment temperature in accordance with the Emergency Operating Procedures?

A. 1) Validated SPDS 2) Highest reading functional instrument 3) Post Accident recorders

B. 1) Post Accident recorders 2) Highest reading functional instrument 3) Validated SPDS

C. 1) Highest reading functional instrument 2) Validated SPDS 3) Post Accident recorders

D. 1) Validated SPDS 2) Post Accident recorders 3) Highest reading functional instrument

47

Perry NRC Exam 2015

QUESTION RO 24



K&A: Knowledge of the interrelations between High Containment Temperature (Mark III Containment Only) and the following: SPDS/ERIS/CRIDS/GDS

High Containment Temperature / 5 Explanation: Answer D – IAW EOP Bases, the priority for determining bulk containment temperature is validated ICS, PAMS recorders, then the highest reading functional instrument. ICS is synonymous with SPDS

A, B, & C – Not the correct priorities.

Technical Reference(s): EOP Bases Rev 4 Reference Attached: EOP Bases p 30








48

Perry NRC Exam 2015 QUESTION RO 25 Power was removed from the Drywell Temperature recorders D23-R090A and D23-R090B on panel H13-P803 for maintenance. TS Rounds data was gathered using only the M13-R110 recorder on H13-P883. Using the attached Tech Spec Rounds sheet determine if there are sufficient DW temperature instruments available and if the LCO for DW Average Temperature should be entered. There (1) sufficient instruments to determine Drywell Average Air Temperature. The LCO for Drywell Average Air Temperature (2) be entered. See attached sheet for information. PRI-TSR page 28 (with data)

(1) (2)

A. are should

B. are should not

C. are not should

D. are not should not

49

Perry NRC Exam 2015

QUESTION RO 25



K&A: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

High Drywell Temperature / 5 Explanation: Answer A – Per Tech Spec Rounds, only one instrument is needed from each elevation to determine Average DW Temp. Additionally, the average is calculated for each elevation then using each elevations average, the DW average is calculated. Using the proper method of calculating temperature, TS LCO 3.6.5.5 is exceeded (145°F).

B & D – Incorrect – Second part - DW Average Temp LCO is exceeded

C & D – Incorrect – First Part – There are sufficient channels for calculating the average DW Temp.

Technical Reference(s): PRI-TSR Rev 32, TS 3.6.5.5 Reference Attached: PRI-TSR pp 28 & 47, TS 3.6.5.5 p 3.6-70, Calculation sheet

Proposed references to be provided to applicants during examination: PRI-TSR page 28 (with data)

Learning Objective (As available): OT-3037-10.A






50

Perry NRC Exam 2015 QUESTION RO 26 The plant is at rated power. Suppression Pool level begins to lower. Per EOP-2 Bases, what is the consequence if Suppression Pool Level lowers to 14.0 feet?

A. The ECCS pumps will lose suction

B. Emergency Depressurization will challenge Drywell Integrity

C. Emergency Depressurization will challenge Containment Integrity

D. The elimination of radioactive gas scrubbing will result in elevated radiation release

51

Perry NRC Exam 2015

QUESTION RO 26



K&A: Knowledge of the operational implications of the following concepts as they apply to Low Suppression Pool Water Level: Heat capacity

Low Suppression Pool Wtr Lvl / 5 Explanation: Answer C – The reason for ED prior to SP level lowering to 14.25’ (2’ above the horizontal vents) is to prevent exceeding the HCL of the Suppression Pool. If a blowdown occurs after exceeding HCL, Containment Integrity is challenged.

A – Incorrect – The ECCS pumps will not start losing suction until level is lowered to 7.25 feet.

B – Incorrect – Drywell integrity is challenged by a hydrogen explosion if HDOL is exceeded.

D – Incorrect – While SP level is > 12.25’, (elevation of 1st row of horizontal vents) scrubbing of radioactive gasses will still occur. Scrubbing is a concern when containment venting is required.

Technical Reference(s): EOP Bases Rev 4, EOP-2 Bases Rev 2.

Reference Attached: EOP Bases p 79, EOP-2 Bases p 41



Question Source: Bank # Clinton 2008 # RO-16 Modified Bank # New





52

Perry NRC Exam 2015 QUESTION RO 27 Which of the following describes the RHR Pump start sequence on a Low Reactor Water level initiation signal concurrent with a Loss of Offsite Power? After the respective Emergency Diesel Generator Output Breaker closes, ____.

A. each RHR pump starts immediately

B. each RHR pump starts 5 seconds later

C. RHR Pumps A and B start immediately and RHR Pump C starts after a 5 second time delay

D. RHR Pumps A and B start after a 5 second time delay and RHR Pump C starts immediately

53

Perry NRC Exam 2015

QUESTION RO 27



K&A: Knowledge of the interrelations between Reactor Low Water Level and the following: Emergency generators

Reactor Low Water Level / 2 Explanation: Answer D – On a LOCA concurrent with a LOOP, loads are sequenced onto the diesel generators to prevent overloading. When the DG breaker closes to reenergize the EH Bus, RHR C pump starts immediately, followed by RHR A & B pumps 5 seconds later.

A – Incorrect – RHR A & B each has a 5 second time delay.

B – Incorrect – RHR C pump does not have a time delay.

C – Incorrect – This is opposite to the correct sequencing.

Technical Reference(s): SDM-OT-COMBINED-E12 Rev 3

Reference Attached: SDM-OT-COMBINED-E12 p 37



Question Source: Bank # Clinton 2004 #RO-50 Modified Bank # New





54

Perry NRC Exam 2015 QUESTION RO 28 The plant was operating at rated power when a LOCA occurred. The current plant conditions are as follows: • Reactor water level 100 inches • Reactor pressure is 700 psig • Injection sources available RCIC only • Offsite release ALERT declared • Steam leak in RCIC room • RCIC room temperature 140° F • Suppression pool temperature 130° F Which of the following identifies the required action regarding RCIC in accordance with EOP-3, Radioactivity Release Control? RCIC should ____.

A. remain running since RCIC isolations were overridden per EOP-2

B. remain running because it is required to be operated by other EOP’s

C. be manually isolated as it should have isolated on high room temperature

D. be manually isolated to prevent damage to the RCIC pump due to high suppression pool temperature

55

Perry NRC Exam 2015

QUESTION RO 28



K&A: Knowledge of the reasons for the following responses as they apply to High Off-Site Release Rate: System isolations

High Off-site Release Rate / 9 Explanation: Answer B – IAW EOP-3 systems discharging outside of primary containment should be isolated unless needed by other EOP’s. Per EOP-1, since RCIC is the only injection system available, RCIC is needed for adequate core cooling.

A – Incorrect – If needed, RCIC isolations are overridden in EOP-1 & EOP-1A, not EOP-2.

C – Incorrect – The RCIC room temperature is less than the isolation setpoint (143°F). RCIC would not have isolated on high temperature yet.

D – Incorrect – The SP temperature that equipment damage is possible is 140°F, not 130°F.

Technical Reference(s): EOP-3 Bases Rev 4 Reference Attached: EOP-3 Bases p 32.



Question Source: Bank # Brunswick 2008 # RO-60 Modified Bank # New





56

Perry NRC Exam 2015 QUESTION RO 29 The plant is operating at 100% power with Control Room HVAC Train A in normal and Control Room HVAC Train B in standby. The following then occurs: • Annunciator CONT RM EMERG RCIRC A CHAR FLTR TEMP HIGH (H13-P904-01-A4

has alarmed • The Secondary Alarm Station (SAS) reports smoke detected in duct of Control Room

HVAC Train A • M26-R032A indicates 260°F and increasing Based on this information, the operator will ____.

A. manually initiate deluge by opening the local deluge supply isolation valve

B. confirm auto initiation of charcoal deluge system on smoke in HVAC Train A

C. confirm auto initiation of charcoal deluge system on high charcoal temperature

D. manually initiate deluge by arming and depressing the CONT RM EMG RCIRC A CHAR FLTR DELUGE switch on H13-P904.

57

Perry NRC Exam 2015

QUESTION RO 29



K&A: Ability to operate and / or monitor the following as they apply to Plant Fire On Site: Plant and control room ventilation systems

Plant Fire On Site / 8 Explanation: Answer A – When indications of a fire in the HVAC charcoal filters exist, the deluge system must be manually lined up locally. All automatic features have been defeated.

B & C – Incorrect – Automatic initiation of charcoal deluge was eliminated. Initiation must be done manually.

D – Incorrect – The arm & depress switch will only trip the running fans and prevent a fan start. It will no longer initiate deluge.

Technical Reference(s): ARI-H13-P904-01 Rev 10, SOI-P54 (WTR) Rev 19, SDM-M25/26 Rev 6

Reference Attached: ARI-H13-P904-01 p 11, SOI-P54 (WTR) pp 49 & 108, SDM-M25/26 p 26


Learning Objective (As available): OT-COMBINED-M25_26-C






58

Perry NRC Exam 2015 QUESTION RO 30 Perry is operating with the following conditions: • Main Generator Terminal Voltage 21.9 KV • Main Generator Megawatts 1280 Mwe • Main Generator VARs 100 MVARs lagging • Main Generator Hydrogen Pressure 72 psig A grid disturbance results in the following: • Steadily increasing grid voltage • The Main Generator voltage regulator shifts to MANUAL With no operator action, this transient could result in ____. Reference Provided: PDB-C002, Generator Capability Curve

A. exceeding the Generator Underexcited Reactive Amp Limit

B. a Generator Lockout due to field over-excitation relay trip

C. a Generator Lockout due to reverse power relay trip

D. overheating the Main Generator stator windings

59

Perry NRC Exam 2015

QUESTION RO 30



K&A: Ability to determine and/or interpret the following as they apply to Generator Voltage And Electric Grid Disturbances: Operating point on the generator capability curve

Generator Voltage and Electric Grid Disturbances / 6 Explanation: Answer A – If grid voltage rises, gen VARs will lower. If voltage mismatch is big enough, VARs can lower enough to cause the gen to operate in the LEAD area of the Generator Capability Curve, and possibly to the point at which the Under Excitation Limit (UEL) is exceeded.

B – incorrect – Field over-excitation results from field current too high, which can be caused by the voltage regulator raising generator output voltage (VAR too high)

C – incorrect – Reverse power trip occurs when real load (MW) is reduced to the point where the grid supplies the generator. The given conditions would not result in lowering MW

D – incorrect – Could happen if gen voltage was higher than grid voltage to point where picking up additional VARS would result in exceeding the capability curve (B-C) and hydrogen pressure of 60 psig.

Technical Reference(s): PDB-C002 Rev 6, SOI-N32/41 Rev 28, & LP OT-COMBINED-N41_51 Rev 2

Reference Attached: PDB-C002 p 4, SOI-N32/41 p 7, & LP OT-COMBINED-N41_51 p29

Proposed references to be provided to applicants during examination: PDB-C002, Generator Capability Curve

Learning Objective (As available): OT-COMBINED-N41_N51-H.3 & OT-3035-14(LP)-A

Question Source: Bank # Modified Bank # Peach Bottom 2008 # RO-58 New





60

Perry NRC Exam 2015 QUESTION RO 31 The plant is operating at 25% rated power with Circ Water Pump ‘A’ & ‘B’ tagged out for cleaning the intake screens. Then a trip of Circ Water Pump ‘C’ occurs. With no operator intervention, what automatic protective actions will occur?

A. Only Reactor Feedwater Pump Turbine trip and Main Turbine Bypass Valve closure

B. Only MSIV closure, Reactor Feedwater Pump Turbine trip, and Main Turbine Bypass Valve closure

C. Only Main Turbine trip, Reactor Feedwater Pump Turbine trip, and Main Turbine Bypass Valve closure

D. MSIV closure, Main Turbine trip, Reactor Feedwater Pump Turbine trip, and Main Turbine Bypass Valve closure

61

Perry NRC Exam 2015

QUESTION RO 31



K&A: Knowledge of the operational implications of the following concepts as they apply to Loss Of Main Condenser Vacuum: Loss of heat sink

Loss of Main Condenser Vacuum / 3 Explanation: Answer D – Since all circ water is lost, vacuum will degrade. As vacuum degrades, the Main Turbine trips @ 8.1”Hg. The RFPT’s trip @ 11.5” Hg in the Aux Condensers, The Bypass close at 20” Hg and the MSIV’s close at 21.5” Hg.

A, B, & C – Incorrect – These actions occur. However, these are not the only actions that occur. All plausible if they do not recall each action.

Technical Reference(s): ONI-N62 Rev 9 Reference Attached: ONI-N62 p 4


Learning Objective (As available): OT-COMBINED-N62-J.1






62

Perry NRC Exam 2015 QUESTION RO 32 The plant is at rated power when a malfunction in the Feedwater Level Control System occurs causing RPV level to increase to 225 inches. Which of the following describes the potential consequence that is avoided by inserting a scram?

A. Main Turbine blade damage

B. Excessive Jet pump vibration

C. Failure of SRV’s to fully reseat after opening

D. Failure of MSIV’s to fully close when needed

63

Perry NRC Exam 2015

QUESTION RO 32



K&A: Knowledge of the interrelations between High Reactor Water Level and the following: Main turbine: Plant-Specific

High Reactor Water Level / 2 Explanation: Answer A – Per the B21 System Description Manual, the bases for the L8 (219”) trip is to prevent excessive moisture carryover that could result in main turbine blade damage.

B – Incorrect – Excessive jet pump vibration is caused by imbalances between Recirc loop flows.

C – Incorrect – The feedwater system also stops feeding at L8. Therefore, level will not increase to the level of the main steam lines causing SRV’s to not reseat.

D – Incorrect – The feedwater system also stops feeding at L8. Therefore, level will not increase to the level of the main steam lines causing MSIV’s to not close when needed.

Technical Reference(s): SDM-B21(NBPI) Rev 9 & SDM-B33 Rev 10

Reference Attached: SDM-B21(NBPI) pp 5-6 & SDM-B33 p 46


Learning Objective (As available): OT-COMBINED-N32_C85-C.3






64

Perry NRC Exam 2015 QUESTION RO 33 The Control Room Operators entered EOP-2, Primary Containment Control due to high Containment temperature. EOP-2 directs the Control Room Operators to “operate all available Containment cooling”. Current plant conditions are as follows: • CVCW Chiller ‘A’ is operating • CVCW Chill Water Pump ‘A’ is operating • Containment Vessel Cooling Fans ‘A’, ‘C’, ‘D’, and ‘F’ are operating What action can be taken to “operate all available Containment cooling”?

A. Start CVCW Chiller ‘C’.

B. Start CVCW Chill Water Pump ‘C’.

C. Start Containment Vessel Cooling Fans ‘B’ and ‘E’.

D. Manually close the CVCW three-way temperature control valve to isolate chill water bypass flow around the Containment Vessel Cooling Air Handling Units.

65

Perry NRC Exam 2015

QUESTION RO 33



K&A: Knowledge of the reasons for the following responses as they apply to High Containment Temperature (Mark III Containment Only): Increased containment cooling

High Containment Temp. / 5 Explanation: Answer C – Per EOP-2 Bases, this action includes starting all available cooling fans.

A – Incorrect – Since CVCW chiller and chill water pump are given as running, no L2 BOP isolation has occurred. Only one CVCW chill water pump can be operation at a time. Starting an additional CVCW chill water pump would cause CVCW chill water pump A to trip.

B – Incorrect – A CVCW chiller needs its respective chill water pump running prior to starting the chiller. Starting an additional CW pump would trip the running CW pump.

D – Incorrect – The TCV is an automatically controlled valve that would already be in full flow to the AHU due to the elevated containment temperature.

Technical Reference(s): EOP-2 Bases Rev 2 Reference Attached: EOP-2 Bases p 73








66

Perry NRC Exam 2015 QUESTION RO 34 During an ATWS, which of the following set of conditions/indications will result in a Redundant Reactivity Control System (RRCS) ‘APRM NOT Downscale’ permissive signal?

A. APRM power > 4%, APRM bypassed, Loss of +20VDC in APRM cabinet

B. Loss of +20VDC in APRM cabinet, APRM INOP, APRM power > 4%

C. APRM bypassed, Loss of +20VDC in APRM cabinet, APRM INOP

D. APRM INOP, APRM power > 4%, APRM bypassed

67

Perry NRC Exam 2015

QUESTION RO 34



K&A: Ability to operate and/or monitor the following as they apply to Incomplete Scram: Neutron monitoring system

Incomplete SCRAM / 1 Explanation: Answer B – APRM INOP, APRM > 4%, or loss of 20 VDC to the ARPM cabinet will result in an APRM NOT DOWNSCALE permissive signal.

A, C, & D – Incorrect – Bypassing an APRM will remove the APRM NOT DOWNSCALE permissive signal even though any other APRM NOT DOWNSCALE condition exists.

Technical Reference(s): SDM-C22 Rev 6 Reference Attached: SDM-C22 pp11-12


Learning Objective (As available): OT-COMBINED-C22-C






68

Perry NRC Exam 2015 QUESTION RO 35 The plant was operating at 100% power with Annulus Exhaust Gas Treatment System Fan B in operation. The following conditions are present: • A steam leak in the annulus • ALERT and HIGH alarms on appropriate PLANT VENT GAS Radiation Monitor • The Shift Manager has declared an Unusual Event (HU-1) based on radiation release to the

environment Entry into (1) is required. Monitor the (2) Plant Vent Radiation Monitor to track release rate.

(1) (2)

A. ONI-D17, High Radiation Levels Within The Plant only

Unit 1

B. ONI-D17, High Radiation Levels Within The Plant only

Unit 2

C. ONI-D17, High Radiation Levels Within The Plant and EOP-05, Radioactivity Release Control,

Unit 1

D. ONI-D17, High Radiation Levels Within The Plant and EOP-05, Radioactivity Release Control,

Unit 2

69

Perry NRC Exam 2015

QUESTION RO 35



K&A: Ability to determine and/or interpret the following as they apply to High Off-Site Release Rate: †Source of off-site release

High Off-Site Release Rate / 9 Explanation: Answer B– The receipt of an Alert or High alarm on the Plant Vent rad monitors requires entry into ONI-D17. At the Unusual Event level, there is no entry requirement for EOP-5. With AGET fan B operating, the Unit 2 Plant Vent is the correct release point.

A & C – Incorrect – AEGT fan B discharges through the Unit 2 Plant Vent.

C & D – Incorrect – No entry conditions for EOP-5 are met.

Technical Reference(s): ODCM Rev 20, ONI-D17 Rev 18,ARI-H13-P680-07 Rev 24, & EOP-3 Bases Rev 4

Reference Attached: ODCM p 33, ONI-D17 pp 3-4,ARI-H13- P680-07 pp 11, 12, & 15, & EOP-3 Bases p 72


Learning Objective (As available): OT-COMBINED-D17-O






70

Perry NRC Exam 2015 QUESTION RO 36 Plant startup is in progress. The following conditions exist: • Mode Switch in STARTUP • Feed water shift in progress from RFBPs on Low flow controller to the Motor Feed Pump • CRD ‘A’ pump is out of service for bearing replacement • CRD ‘B’ pump trips on over current • Alarm window H13-P680-5A-E8, CRD HCU LEVEL HI/PRESS LO is illuminated Per ONI-C11-1, Inability To Move Control Rods, the following is correct with respect to current plant conditions? The Mode Switch must be placed in SHUTDOWN 1 with 2 with associated accumulators inoperable.

(1) (2)

A. immediately 2 control rods @ position 00

B. within 20 minutes 2 control rods @ position 00

C. immediately 2 control rods @ position 48

D. within 20 minutes 2 control rods @ position 48

71

Perry NRC Exam 2015

QUESTION RO 36



K&A: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Loss of CRD Pumps / 1 Explanation: Answer C – This is the Required Action per TS 3.1.5 and ONI-C11-1 for reactor pressure < 600 psig. RFBP transfer to MFP shift occurs @ ~250 psig.

A – Incorrect – There is no requirement to shutdown the reactor with inoperable accumulators as long as the associated control rods are not withdrawn.

B – Incorrect – There is no requirement to shutdown the reactor with inoperable accumulators as long as the associated control rods are not withdrawn.

D – Incorrect – A 20 minute delay is allowed if the Rx is in Mode 1 or 2 if Rx pressure > 600 psig.

Technical Reference(s): TS 3.1.5, ONI-C11-1 Rev 14 & ONI-C11-1 Rev 15

Reference Attached: TS 3.1.5 p 3.1-16-17, ONI-C11-1 p 5


Learning Objective (As available): OT-COMBINED-C11_CRDH-J.2, OT-3037-05.D






72

Perry NRC Exam 2015 QUESTION RO 37 The following conditions exist: • The Fuel Handling Building Ventilation System is in operation in accordance with SOI-M40 • Movement of irradiated fuel in the FHB Then an irradiated fuel bundle is dropped. Shortly thereafter, a HIGH radiation alarm is received on the FHB Ventilation Exhaust GAS and IODINE modules. EOP-3, Secondary Containment Control, is entered. Which of the following describes the current status of the FHB Ventilation System based on these plant conditions?

A. Only two Exhaust Fans are running due to the HIGH alarm on the iodine channel.

B. Only two Exhaust Fans are running due to the HIGH alarm on the noble gas channel.

C. Two Exhaust Fans and one Supply Fan are running due to the HIGH alarm on the noble gas channel.

D. Two Exhaust Fans and one Supply Fan are running due to the HIGH alarm on the iodine channel.

73

Perry NRC Exam 2015

QUESTION RO 37



K&A: Knowledge of the operational implications of the following concepts as they apply to Secondary Containment Ventilation High Radiation: †Radiation releases

Secondary Containment Ventilation High Radiation / 9 Explanation: Answer B – A HIGH on the FHB Gas channel will cause the running FHB supply fan to trip. Since the FHB ventilation was running IAW SOI-M40, 2 exhaust fans would be running.

A – Incorrect - The High on the Iodine channel does not cause any automatic trips.

C & D – Incorrect – This is the Normal mode of operation for the FHB ventilation system

Technical Reference(s): SOI-M40 Rev 9, ONI-D17 Rev 18, SDM-M40 Rev 7

Reference Attached: SOI-M40 p 16, ONI-D17 p 18, SDM-M40 p 4


Learning Objective (As available): OT-COMBINED-M40-B.4, OT-3035-01(LP)-A.3






74

Perry NRC Exam 2015 QUESTION RO 38 The following conditions exist: • The reactor was scrammed from 25% rated power • Emergency Depressurization was required • RHR A is providing adequate core cooling • Suppression Pool Makeup has initiated • Suppression Pool level is lowering 1 inch per minute • Current suppression pool level is 15 feet 2 inches Per EOP Bases, the earliest that continued operation of RHR A pump will be threatened due to possible RHR equipment damage is in ____ minutes.

A. 11

B. 35

C. 95

D. 113

75

Perry NRC Exam 2015

QUESTION RO 38


Level: RO SRO Tier # 2 Group # 1 K/A# 203000 A1.05 Importance Rating 3.8

K&A: Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: Injection Mode (Plant Specific) controls including: Suppression pool level

RHR/LPCI; Injection Mode Explanation: Answer D – With suppression pool level lowering at 1”/minute, it will take 113 minutes to reach 5.75’. At this level, pump damage from air entrainment becomes a concern.

A – Incorrect – This is the time to uncover the SRV tailpipes

B – Incorrect – This is the time to uncover the horizontal vents

C – Incorrect – This is the time when damage to RCIC may occur

Technical Reference(s): EOP Bases Rev 4 Reference Attached: EOP Bases p 60


Learning Objective (As available): OT-3402-01-B.2

Question Source: Bank # Modified Bank # Perry 2010 #RO-25 New





76

Perry NRC Exam 2015 QUESTION RO 39 The plant is shutdown for a forced outage with the following conditions: • RHR A loop is operating in Shutdown Cooling • RHR B loop is in standby • All electrical buses and panels are on their normal power supplies Then, a lockout on Bus L11 occurs. What is the effect on Shutdown Cooling and what ONI(s) will be entered? RHR A pump will be (1) and entry into (2) is required.

(1) (2)

A. tripped only ONI-E12-2, Loss of Decay Heat Removal

B. tripped ONI-E12-2, Loss of Decay Heat Removal and ONI-C71-2, Loss of One RPS Bus

C. running only ONI-R22, Loss of A Non-Essential 13.8KV Or 4.16KV Bus

D. running only ONI-R22, Loss of A Non-Essential 13.8KV Or 4.16KV Bus and ONI-C71-2 Loss of One RPS Bus

77

Perry NRC Exam 2015

QUESTION RO 39



K&A: Ability to (a) predict the impacts of the following on the Shutdown Cooling System (RHR Shutdown Cooling Mode); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: SDC/RHR pump trips

Shutdown Cooling Explanation: Answer B – When Bus L11 is lost, the B RPS bus loses power which causes a SDC isolation and RHR pump trip. Both ONI-C71-2 and ONI-E12-2 need to be entered.

A – Incorrect – ONI-C71-2 and ONI-R22 also need to be entered.

C & D – Incorrect – First part - The RHR pump will trip when the E12-F008 & F009 valves close.

C – Incorrect – ONI-E12-2 and ONI-C71-2 also need to be entered.

D – Incorrect – ONI-E12-2 also needs to be entered

Technical Reference(s): ONI-E12-2 Rev 32, ONI-C71-2 Rev 9, ONI-R22 Rev 9, SDM-C71 Rev 12, & SDM-E12 Rev 3

Reference Attached: ONI-E12-2 pp 3-4, ONI-C71-2 9, 12-13, ONI-R22 pp 3, 8, SDM-C71 p 68, & SDM-E12 p 51


Learning Objective (As available): OT-COMBINED-E12-I.1 & OT-3035-11(LP)-A.1






78

Perry NRC Exam 2015 QUESTION RO 40 RHR Loop A has just been placed into the Shutdown Cooling Mode of operation using the Normal Return Path. The cooldown rate is excessive. In accordance with SOI-E12, Residual Heat Removal System, which of the following actions will reduce the cooldown rate?

A. Throttle close the RHR A HX’S BYPASS VALVE, E12-F048A, and throttle open the RHR A HX’S OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 7000-7100 gpm.

B. Throttle open the RHR A HX’S BYPASS VALVE, E12-F048A, and throttle close the RHR A HX’S OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 7000-7100 gpm.

C. Throttle open the RHR A HX’S BYPASS VALVE, E12-F048A, and throttle close the RHR A HX’S OUTLET VALVE, E12-F003A, while maintaining a system flowrate of 2575-7100 gpm.

D. Throttle ESW flow through the RHR Heat Exchanger using RHR A HX’S ESW OUTLET VALVE, P45-F068A.

79

Perry NRC Exam 2015

QUESTION RO 40


Level: RO SRO Tier # 2 Group # 1 K/A# 205000 K5.02 Importance Rating 2.8

K&A: Knowledge of the operational implications of the following concepts as they apply to Shutdown Cooling System (RHR Shutdown Cooling Mode): Valve operation

Shutdown Cooling Explanation: Answer B – This is the approved method for adjusting the cooldown rate while in Normal Mode SDC.

A – Incorrect – This action will raise the cooldown rate.

C – Incorrect – This is the required flow rate band when using the alternate return path via E12-F042A.

D – Incorrect – Not an approved method to control the cooldown rate per SOI-E12.

Technical Reference(s): SOI-E12 Rev 63 Reference Attached: SOI-E12 pp 100-101


Learning Objective (As available): OT-COMBINED-E12-B.4

Question Source: Bank # Perry 2010 # RO 39 Modified Bank # New

Question History: Previous NRC Exam Perry 2010




80

Perry NRC Exam 2015 QUESTION RO 41 The following plant conditions exist: • Mode 3, forced cooldown in progress • Reactor Pressure 400 psig • Reactor Level 185” An inadvertent initiation of Low Pressure Core Spray (LPCS) occurs. Based on this information, LPCS injection (1) and in accordance with ONI-E12-1, Inadvertent Initiation of ECCS/RCIC you would override LPCS by (2) .

(1) (2)

A. occurred stopping the LPCS Pump

B. occurred shutting the LPCS Injection Valve

C. did not occur stopping the LPCS Pump

D. did not occur shutting the LPCS Injection Valve

81

Perry NRC Exam 2015

QUESTION RO 41



K&A: Ability to predict and/or monitor changes in parameters associated with operating the Low Pressure Core Spray System controls including: Core spray flow

LPCS Explanation: Answer A – LPCS will begin injecting at ~450 psig. Therefore, injection occurred. IAW ONI-E12-1, the proper method for overriding LPCS is to stop the pump.

B & D – Incorrect – Second part - Shutting the LPCS injection valve is a supplemental action following stopping the pump.

D & D – First part – Injection occurred – the shutoff head for LPCS is ~450 psig.

Technical Reference(s): ONI-E12-1 Rev 9 Reference Attached: ONI-E12-1 pp 3, 5


Learning Objective (As available): OT-3035-15(LP) A.1






82

Perry NRC Exam 2015 QUESTION RO 42 The plant was operating at rated power when a LOCA occurred coincident with a LOOP. All ECCS systems started automatically. Then Emergency Closed Cooling Pump ‘A’ tripped. This will result in rising temperatures in the ____.

A. ECC Pump area

B. LPCS Pump room

C. HPCS Pump room

D. RHR B Pump room

83

Perry NRC Exam 2015

QUESTION RO 42



K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Low Pressure Core Spray System: ECCS room cooler(s)

LPCS Explanation: Answer B – ECC ‘A’ loop supplies cooling water to the LPCS Pump Room Cooler. With a LOOP, Aux Building ventilation will not be running to provide cooling to the ECCS rooms. With no cooling water flow to the room cooler and the LPCS pump running, room temperature will increase.

A – Incorrect – Only one ECC Pump Area Cooler is required and cooling is provided by CCCW (P47).

C – Incorrect – The HPCS pump room cooler is supplied by HPCS-ESW and is designed to provide 100% of the cooling requirements for the HPCS Pump room.

D – Incorrect – RHR B room cooler is supplied by ECC ‘B’ loop and is designed to provide 100% of the cooling requirements for the RHR B Pump room.

Technical Reference(s): SOI-M28 Rev 7, ARI-H13-P800-01 Rev 7, SDM-M39 Rev 4, SDM-P42 Rev 12.

Reference Attached: SOI-M28 p 3, ARI-H13-P800-01 p 31, SDM-M39 pp 2, 3, & 5, SDM-P42 p 2


Learning Objective (As available): OT-COMBINED-M39-L.1






84

Perry NRC Exam 2015 QUESTION RO 43 High Pressure Core Spray (HPCS) is in Standby with its suction on the Condensate Storage Tank (CST). The following events occur in sequence: 1) A LOCA is in progress 2) CST Level Instrument 1E22-N654C, COND STG TK LVL – LOW fails low HPCS Suction source ____ be shifted to the CST with the control switches on H13-P601 IAW EOP-SPI 6.4 HPCS Injection.

A. will shift to the Suppression Pool, and can

B. will shift to the Suppression Pool and can not

C. will not shift to the Suppression Pool, but can

D. will not shift to the Suppression Pool, and can not

85

Perry NRC Exam 2015

QUESTION RO 43



K&A: Ability to (a) predict the impacts of the following on the High Pressure Core Spray System (HPCS); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings

HPCS Explanation: Answer A – HPCS will shift to the Suppression Pool on a low CST level signal. Using steps in EOP-SPI 6.4, HPCS suction can be returned to the CST. (EOP-SPI 6.4 only defeats the hi SP level)

B & D – Incorrect – Second Part - HPCS can be overridden to the CST.

C & D – Incorrect – First part – HPCS will auto transfer to the Suppression pool on a low CST level signal.

Technical Reference(s): EOP-SPI 6.4 Rev 1, PDB-I005 Rev 9

Reference Attached: EOP-SPI 6.4 pp 4 & 5, PDB-I005 p 12


Learning Objective (As available): OT-COMBINED-E22-F.3






86

Perry NRC Exam 2015 QUESTION RO 44 The following plant conditions exist: • An ATWS is in progress and EOP-1A, Level Power Control, has been entered • Both SLC Pump keylock switches are in the ON position • Both SLC Pump Suction Valves, 1C41-F001A & B, indicate open • Annunciator SLC STORAGE TANK LEVEL HI/LO has alarmed • SLC Storage Tank Level is 4000 gallons • The C41-F004A white SQUIB CONTINUITY indicating light is OFF • The C41-F004B white SQUIB CONTINUITY indicating light is ON Which of the following describes the current status of the SLC System?

A. Only SLC Pump A has started and is injecting into the RPV.

B. Both SLC Pumps have started and are injecting into the RPV.

C. Only SLC Pump A has started; however, it is not injecting into the RPV.

D. Both SLC Pumps have started; however, only SLC Pump A is injecting into the RPV.

87

Perry NRC Exam 2015

QUESTION RO 44



K&A: Ability to monitor automatic operations of the Standby Liquid Control System including: Lights and alarms: Plant-Specific

SLC Explanation: Answer B – SLC piping is configured with the squib valves downstream of the pump discharge cross-connect piping. Pump start is independent of squib valve firing. Therefore, if one squib valve fails to fire, SLC flow is unaffected.

A – Incorrect – Based on the indications provided, both pumps have started.

C – Incorrect – Based on the indications provided, both pumps have started and both are injecting.

D – Incorrect – Since the cross-connect piping is upstream of the squib valves, both pumps are injecting.

Technical Reference(s): SOI-C41 Rev 19, ARI-H13-P601-19 Rev 17, SDM-C41 Rev 9

Reference Attached: SOI-C41 p 7, ARI-H13-P601-19 p 95, SDM-C41 pp 9 & 29


Learning Objective (As available): OT-COMBINED-C41-F.1 & F.2






88

Perry NRC Exam 2015 QUESTION RO 45 The following conditions exist. • The reactor has scrammed and the Mode Switch is in SHUTDOWN • The cause of the scram was identified and corrected • Annunciator RPS INST VOL HI is sealed in Then: • All INST VOL LEVEL HI SCRAM BYPASS keylock switches were placed in BYPASS • All RPS SCRAM RESET pushbuttons were depressed • 30 seconds later the Mode Switch was placed in STARTUP Based on the above information: Reset of RPS (1) successful. And, what effect, if any, resulted from placing the Mode Switch in STARTUP (2) ?

(1) (2)

A. was a full scram occurred

B. was only a half scram occurred

C. was no effect

D. was not no effect

89

Perry NRC Exam 2015

QUESTION RO 45



K&A: Ability to manually operate and/or monitor in the control room: Scram instrument volume level

RPS Explanation: Answer A – The INST VOL LEVEL HI SCRAM BYPASS switches bypass the hi SDV scram signal when the Mode Switch is in Shutdown or Refuel and allows RPS to be reset. Since it takes several minutes for the SDV to drain after the scram is reset, placing the Mode Switch to Startup 30 seconds after placing the switches in bypass would generate scram signal.

B – Incorrect – Since a FULL scram would occur, ONLY a half scram is wrong.

C – Incorrect – A full scram signal would be generated.

D – Incorrect – The scram can be reset and another scram signal would be generated.

Technical Reference(s): SOI-C71 Rev 20,ARI-H13-P680-05 Rev 15, Dwgs 208-040 sh 5 Rev BB sh 9 Rev T

Reference Attached: SOI-C71 p 64,ARI-H13-P680-05 p 17, Dwgs 208-040 sh 5 & sh 9 (partials)


Learning Objective (As available): OT-COMBINED-C71-F

Question Source: Bank # Hope Creek 2010 # RO-20 Modified Bank # New





90

Perry NRC Exam 2015 QUESTION RO 46 A plant startup following a refuel outage is in progress with the following: • The mode switch is in STARTUP • The Unit Supervisor has declared IRM’s B, D, F & H INOPERABLE Which of the following describes an associated Technical Specification requirement?

A. Immediately restore RPS trip capability for IRM’s.

B. Within one hour place Mode Switch in SHUTDOWN.

C. Within one hour, restore RPS trip capability for IRM’s.

D. Immediately place RPS channel B in tripped condition.

91

Perry NRC Exam 2015

QUESTION RO 46



K&A: Knowledge of less than or equal to one hour Technical Specification action statements for systems.

IRM Explanation: Answer C – With all IRM’s on RPS B being INOP, trip capability for RPS channel B is not maintained. The TS Action is to restore RPS trip capability within 1 hour.

A – Incorrect – This is a 1 hour action, not an immediate action

B – Incorrect – If trip capability not restored in one hour then be in Mode 3 in 12 hours – no requirement to place Mode Switch in S/D in 1 hour.

D – Incorrect – This is a 6 hour action. Placing RPS B in trip condition could also fulfill the requirement to restore trip capability. However that is a 1 hour requirement, not an immediate action.

Technical Reference(s): TS 3.3.1.1, SDM-C51(IRM) Rev 8

Reference Attached: TS 3.3.1.1 pp 3.3-1, 2, & 7, and SDM-C51(IRM) p 63


Learning Objective (As available):OT-3037-H






92

Perry NRC Exam 2015 QUESTION RO 47 Plant conditions are as follows: • Reactor Mode Switch is in STARTUP/STANDBY • All IRMs are on Range 3 • Source Range Monitor (SRM) A is reading 0.5 cps • SRMs B and C are reading 8.3 x 10E4 • SRM D drawer mode switch is in STANDBY • A rod block signal has been generated Which of the following has caused the rod block?

A. SRM Upscale

B. SRM Inoperable

C. SRM Downscale

D. SRM Detector Wrong Position

93

Perry NRC Exam 2015

QUESTION RO 47



K&A: Knowledge of the physical connections and/or cause-effect relationships between Source Range Monitor (Srm) System and the following: Rod control and information system: Plant-Specific

Source Range Monitor Explanation: Answer B – With Reactor Mode Switch not in RUN and SRM detector channel switch out of OPERATE, a Rod Block on SRM INOP signal will be generated .

A – Incorrect – An Upscale SRM signal does not come in until 1E5.

C – Incorrect – The SRM downscale signal is bypassed when associated IRM’s are on range 3

D – Incorrect - Detector Wrong Position will not generate a Rod Block, with IRMs on or above range 3

Technical Reference(s): SDM-C11(RC&IS) Rev 9 Reference Attached: SDM-C11(RC&IS) pp55-56


Learning Objective (As available): OT-COMBINED-C51_SRM-F

Question Source: Bank # Hope Creek 2010 # RO-01 Modified Bank # New





94

Perry NRC Exam 2015 QUESTION RO 48 The plant was operating at rated power with APRM A INOPERABLE and in BYPASS. Subsequently, an electrical fault resulted in a loss of power to ATWS Dist Panel EV-1-B Div 2. Based on the above information, ____ signal is generated.

A. a full scram

B. only a half scram

C. a Division 2 RRCS ARI

D. only a Recirc Flow Control Valve closure

95

Perry NRC Exam 2015

QUESTION RO 48



K&A: Knowledge of electrical power supplies to the following: APRM channels

APRM / LPRM / OPRM Explanation: Answer B – When power is lost to EV-1-B, APRM’s B, D, F, & H lose power which causes a ½ scram signal.

A – Incorrect – Since APRM A is bypassed, no full scram signal will result.

C – Incorrect – Div 2 RRCS is not powered from EV-1-B. Plausible since name of panel is ATWS.

D – Incorrect – A FCV closure signal is generated in addition to a ½ if power was lost to EV-1-A.

Technical Reference(s): ARI-H13-P680-06 Rev 9, ELI-R14/15 Rev 7, Dwg 206-65 Rev S

Reference Attached: ARI-H13-P680-06 p 39-40, ELI-R14/15 p 7, Dwg 206-65


Learning Objective (As available): OT-COMBINED-C51-AP_OPRM-L.1 & OT-COMBINED-C22-F






96

Perry NRC Exam 2015 QUESTION RO 49 The following conditions exist: • The reactor scrammed due to a loss of the feedwater system • High Pressure Core Spray pump tripped due to a seized shaft • Reactor Core Isolation Cooling is maintaining RPV level at 30 inches • Reactor pressure is 870 psig and lowering slowly The receipt of which annunciator would indicate an immediate threat to RCIC’s ability to maintain RPV level?

A. RCIC ISOL DIAPHRAGM RUPTURED, (H13-P601-0021-B1)

B. STEAM TUNNEL LD AMB TEMP P632, (H13-P601-0019-G4)

C. RCIC TURBINE OIL COOLER OUT TEMP HIGH, (H13-P601-0021-C4)

D. RCIC SUPR POOL SUCT VLV OPEN SUPR PL LVL HI, (H13-P601-0021-G5)

97

Perry NRC Exam 2015

QUESTION RO 49



K&A: Knowledge of the effect that a loss or malfunction of the Reactor Core Isolation Cooling System (RCIC) will have on following: Reactor water level

RCIC Explanation: Answer A – An exhaust diaphragm rupture will cause an immediate RCIC turbine trip.

B – Incorrect – The steam tunnel high temperature has a 29 minute time delay to isolate.

C – Incorrect – High RCIC lube oil temp does not cause an immediate loss of the turbine.

D – Incorrect – A high suppression level will cause a suction shift, but the suppression pool valve opens fully before the CST valve closes – in this case it doesn’t matter which source is lined up.

Technical Reference(s): ARI-H13-P601-021 Rev 15, ARI-H13-P601-19 Rev 17, & SOI-E31 Rev 8

Reference Attached: ARI-H13-P601-021 pp 19, 39, & 89, ARI-H13-P601-19 p 111, & SOI-E31 p19


Learning Objective (As available): OT-COMBINED-E51-F.2






98

Perry NRC Exam 2015 QUESTION RO 50 The following conditions exist: • Reactor Core Isolation Cooling (RCIC) is operating in CST to CST Mode • 1E51-R600, RCIC Pump Flow Controller is in AUTO set at 700 gpm • RCIC turbine speed is 3500 rpm If RCIC FIRST TEST VALVE TO CST, 1E51-F022 valve is throttled in the CLOSED direction, which of the following describes the expected stable values of RCIC turbine speed and RCIC Pump flow?

A. RCIC Turbine Speed 3000 rpm; RCIC Pump Flow 600 gpm

B. RCIC Turbine Speed 3000 rpm; RCIC Pump Flow 700 gpm

C. RCIC Turbine Speed 4000 rpm; RCIC Pump Flow 600 gpm

D. RCIC Turbine Speed 4000 rpm; RCIC Pump Flow 700 gpm

99

Perry NRC Exam 2015

QUESTION RO 50



K&A: Ability to predict and/or monitor changes in parameters associated with operating the Reactor Core Isolation Cooling System (RCIC) controls including: RCIC flow

RCIC Explanation: Answer D – When throttling RCIC FIRST TEST VALVE TO CST, 1E51-F022 in the CLOSED direction, flow will immediately lower until the Flow Controller increases Turbine speed to re-establish flow at the 700 gpm setpoint.

A – Incorrect – This would be the response if the Flow Controller were lowered to 600 gpm

B – Incorrect – This would be the response if the 1E51-F022 valve were throttled in the OPEN direction.

C – Incorrect – This would be the response if the Flow Controller were lowered to 600 gpm and 1E51-F022 were throttled in the CLOSED direction

Technical Reference(s): SDM-E51 Rev 12 Reference Attached: SDM-E51 p 22



Question Source: Bank # Fermi 2001 # RO-70 Modified Bank # New





100

Perry NRC Exam 2015 QUESTION RO 51 The following occurred: • An ATWS occurred • The Unit Supervisor directed ADS to be inhibited • The RO placed ADS A and B Inhibit switches in INHIBIT • ADS A INHIBIT light failed to illuminate • Annunciator ADS A TIME DELAY LOGIC TIMER RUNNING is sealed in Based on this information, (1) . To delay an undesired ADS actuation the Operator must depress (2) .

(1) (2)

A. RPV Level 3 and RPV Level 1 have been reached and RHR A or LPCS is running

both ADS A and B Logic Seal In Reset pushbutton

B. RPV Level 3 and RPV Level 1 have been reached

both ADS A and B Logic Seal In Reset pushbutton

C. RPV Level 3 and RPV Level 1 have been reached and RHR A or LPCS is running

only ADS A Logic Seal In Reset pushbutton

D. RPV Level 3 and RPV Level 1 have been reached

only ADS A Logic Seal In Reset pushbutton

101

Perry NRC Exam 2015

QUESTION RO 51



K&A: Knowledge of Automatic Depressurization System design feature(s) and/or interlocks which provide for the following: Prevent inadvertent initiation of ADS logic

ADS Explanation: Answer D – Annunciator ADS A TIME DELAY LOGIC TIMER RUNNING indicates that both L3 & L1 have been reached and ADS will initiate after a 105 second time delay. Since the B inhibit was successful, only the ADS A Logic Seal In Reset pushbutton needs to be depressed.

A & C – Incorrect – First part – A low pressure ECCS pump is not required to be running for the timer to start and the annunciator to alarm.

A & B – Incorrect – Second part – Since the B inhibit was successful, only the ADS A Logic Seal In Reset pushbutton needs to be depressed.

Technical Reference(s): ARI-H13-P601-19 Rev 17 and ONI-E12-2 Rev 11

Reference Attached: ARI-H13-P601-19 p 71 and ONI-E12-2 p 6


Learning Objective (As available): OT-COMBINED-B21C-F & L.1






102

Perry NRC Exam 2015 QUESTION RO 52 The plant is in Mode 4 with RHR B loop in Shutdown Cooling. SVI-B21-T0034-B, RPV Level 3 and Level 8 RPS/RHR Shutdown Isolation Channel B Functional for 1B21-N680B, LVL 3 & 8, is in progress. During performance of this SVI an RPV Level 3 trip signal is input. Concurrently, 1B21-N680C, LVL 3 & 8 trip unit fails low. Based on this information, ____.

A. only 1E12-F009, Shutdown Cooling INBD SUCT ISOL will close

B. only 1E12-F008, Shutdown Cooling OTBD SUCT ISOL will close

C. 1E12-F009, Shutdown Cooling INBD SUCT ISOL and 1E12-F053B, Shutdown Cooling B To FDW Shutoff will close

D. 1E12-F008, Shutdown Cooling OTBD SUCT ISOL and 1E12-F053B, Shutdown Cooling B To FDW Shutoff will close

103

Perry NRC Exam 2015

QUESTION RO 52



K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off: Nuclear boiler instrumentation

PCIS/Nuclear Steam Supply Shutoff Explanation: Answer C – Both 1E12-F009 & 1E12-F053B are receive an isolation signal when 1B21-N680B is dialed low and 1B21-N680C fails low.

A – Incorrect – 1E12-F0053B and 1E12-F037B also will close.

B – Incorrect – PCIS isolation logic is Inboard/Outboard. E12-F008 is not an inboard valve. Plausible since most ‘C’ instruments are Div 1 instruments and E12-F008 is a Div 1 powered valve.

D – Incorrect – 1E12-F008 will not close on this combination of trip unit signals.

Technical Reference(s): PDB-I5 Rev 9, SVI-B21-T0034B Rev 12

Reference Attached: PDB-I5 p 35



Question Source: Bank # Modified Bank # Perry 2007-1 # RO-39 New





104

Perry NRC Exam 2015 QUESTION RO 53 The plant is operating at rated power when the following indications are observed: • Core Thermal Power initially decreases and then stabilizes at a higher value • Main Generator electrical output decreases • No operator actions have been taken Based on these indications, entry into ____ is required.

A. ONI-B21-1 due to an open SRV

B. ONI-C11-3 due to a single control rod scram.

C. ONI-C51 due to a Reactor Recirculation FCV drifting open.

D. ONI-N36 due to a feedwater heater extraction steam valve closing.

105

Perry NRC Exam 2015

QUESTION RO 53



K&A: Ability to predict and/or monitor changes in parameters associated with operating the Relief/Safety Valves controls including: Reactor power

SRVs Explanation: Answer A – When an SRV opens, Rx pressure decreases. As pressure decreases, Rx power decreases due to the pressure coefficient of reactivity. With the SRV remaining open, power will return to previous levels then increase slightly. This was verified in the simulator on 7/16/14.

B – Incorrect – A single rod scram would decrease power, but it would remain at a lower level.

C – Incorrect – a FCV drifting open would cause power to increase.

D – Incorrect – A FW heater isolation would reduce FW temperature and cause an increase in power with no initial decrease in power.

Technical Reference(s): Reactor Theory (GFE) Lesson Plan Rev 4

Reference Attached: Reactor Theory (GFE) Lesson Plan p 49


Learning Objective (As available): OT-COMBINED-B21_N11-O & OT-3035-07(LP)-A.1






106

Perry NRC Exam 2015 QUESTION RO 54 The plant is operating at 70% rated power with the following conditions: • RFPT A & B are on DFWCS in 3-element (3E) Level Control • Motor Feed Pump is in Standby with MFP Auto Xfer feature ARMED

Then the following occurs: • DFWCS RPV Level Channel A fails upscale • DFWCS RPV Level Channel B fails downscale The failure of these channels will cause the DFWCS to shift (1) . The Reactor Operator will (2) .

(1) (2)

A. to the manual speed control dial shift DFWCS to 1-element (1E) Level Control per SOI-C34, Feedwater Control System

B. to the manual speed control dial maintain RPV level 192” to 200” per ONI-C34, Feedwater Flow Control Malfunction

C. the feed pump flow controllers to manual

shift DFWCS to 1-element (1E) Level Control per SOI-C34, Feedwater Control System

D. the feed pump flow controllers to manual

maintain RPV level 192” to 200” per ONI-C34, Feedwater Flow Control Malfunction

107

Perry NRC Exam 2015

QUESTION RO 54



K&A: Ability to (a) predict the impacts of the following on the Reactor Water Level Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor water level input

Reactor Water Level Control Explanation: Answer D – The loss of 2 of 3 level inputs causes DFWCS flow controllers to shift to manual. RPV level is maintained IAW ONI-C34.

A & B – Incorrect – First part - DFWCS no longer auto shifts to the manual speed control dials.

A & C – Incorrect – Second Part – There is no requirement to shift DFWCS to 1E control. Plausible since a loss of a steam flow signal, feed pump suction flow signal, feed pump recirc flow signal causes a shift from 3 Element to 1 Element control.

Technical Reference(s): SOI-C34 Rev 34 & ONI-C34 Rev 9

Reference Attached: SOI-C34 p 6 & ONI-C34 pp 5-6


Learning Objective (As available): OT-COMBINED-C34-M






108

Perry NRC Exam 2015 QUESTION RO 55 The following conditions exist: • The plant is at 100% power • Digital Feedwater Operator Rx Level Setpoint was set at 200 inches A manual scram is now inserted. Which of the following describes the response of the DFWCS following the scram?

A. Upon receipt of the scram signal, the level demand signal will be 200 inches for 10 seconds and then lower to 178 inches.

B. Upon receipt of the scram signal, the level demand signal will be 196 inches for 10 seconds and then lower to 178 inches.

C. When level reaches 178 inches, the level demand signal will be 200 inches for 10 seconds and then lower to 178 inches.

D. When level reaches 178 inches, the level demand signal will be 196 inches for 10 seconds and then lower to 178 inches.

109

Perry NRC Exam 2015

QUESTION RO 55



K&A: Ability to monitor automatic operations of the Reactor Water Level Control System including: Changes in reactor power

Reactor Water Level Control Explanation: Answer C – With the Operator Rx Level Setpoint set at 200”, when RPV level drops below L3, Setpoint Setdown logic demands the Operator Rx Level Setpoint setting for 10 seconds then lowers to 178”.

A & B – Incorrect – The scram signal does not initiate Setpoint Setdown logic.

B & D – Incorrect – the Operator Rx Level Setpoint was set at 200”. Therefore the Setpoint Setdown logic demands 200” not 196”.

Technical Reference(s): Lesson Plan OT-COMBINED-C34 Rev 5

Reference Attached: Lesson Plan OT-COMBINED-C34 p 69


Learning Objective (As available): OT-COMBINED-C34-G






110

Perry NRC Exam 2015 QUESTION RO 56 The plant is operating at rated power with the following conditions: • Annulus Exhaust Gas Treatment System ‘A’ is operating • Annulus Exhaust Gas Treatment System ‘B’ is in Standby An initiation signal then causes AEGT train B to start. The start of the AEGT train B will cause ____.

A. 1M15-F070A, AEGT EXH DAMPER A to open more

B. 1M15-F070B, AEGT EXH DAMPER B to open more

C. 1M15-F080A, AEGT RECIRC DAMPER A start closing

D. 1M15-F080B, AEGT RECIRC DAMPER B start opening

111

Perry NRC Exam 2015

QUESTION RO 56



K&A: Ability to monitor automatic operations of the Standby Gas Treatment System including: Valve operation

SGTS Explanation: Answer D – With an AEGT train in standby, the Exhaust damper is full open and the Recirc damper is full closed. When the idle fan starts, the resulting increase in annulus to Containment DP causes the B Recirc damper to start opening and both Exhaust dampers to start closing.

A – Incorrect – With greater vacuum in the annulus, the A Exhaust damper will close to restore DP.

B – Incorrect – In standby, the B Exhaust damper is full open. It cannot open any further upon fan start.

C – Incorrect – The AEGT A Recirc damper will respond to the higher DP by starting to open, not close.

Technical Reference(s): SDM-M15 Rev 7 Reference Attached: SDM-M15 p 16


Learning Objective (As available): OT-COMBINED-M15-F






112

Perry NRC Exam 2015 QUESTION RO 57 Bus EH13 is being transferred from the Division 3 Emergency Diesel Generator (EDG) back to the grid. The SYNC SEL SWITCH is in the TH1 position. The following indications are observed on panel H13-P601-22: • BUS EH13 VOLTS RUNNING, 1R22-R022C 4200 VAC • BUS EH13 VOLTS INCOMING, 1R22R021C 4100 VAC • Synchroscope is rotating slowly is the SLOW direction

Before the Preferred Source Breaker, EH1303 can be closed, you must: Adjust the DIESEL GEN VOLTAGE REGTR in the (1) direction to match BUS EH13 RUNNING and INCOMING VOLTS. And Adjust the DIESEL GEN GOVERNOR in the (2) direction until the Synchroscope is moving slowly in the FAST direction.

(1) (2)

A. LOWER RAISE

B. LOWER LOWER

C. RAISE RAISE

D. RAISE LOWER

113

Perry NRC Exam 2015

QUESTION RO 57



K&A: Ability to manually operate and/or monitor in the control room: Synchroscope, including understanding of running and incoming voltages

AC Electrical Distribution Explanation: Answer C – Need to raise Incoming Volts to match Running Volts and need to raise governor to cause Synchroscope to start rotating in the FAST direction.

A & B – Incorrect – First part - Lowering on the voltage regulator will cause voltages to diverge not match.

B & D – Incorrect – Second part – Lowering on the governor will cause the syschroscope to rotate faster in the SLOW direction.

Technical Reference(s): SOI-E22B Rev 30 Reference Attached: SOI-E22B pp 52-53


Learning Objective (As available): OT-COMBINED-E22B-G.1

Question Source: Bank # Modified Bank # Perry 2007-1 # RO-56 New





114

Perry NRC Exam 2015 QUESTION RO 58 The plant was operating at rated power when a loss of off-site power occurred. Subsequently, the Unit Supervisor entered EOP-1 due to lowering reactor water level. Current RPV pressure is 60 psig. RPV level cannot be maintained >130” using available systems. What method can be used to provide additional water into the RPV?

A. ADHR Alternate Injection, per ONI-SPI E-4

B. FPCC Header Alternate Injection per EOP-SPI 4.3

C. Fast Fire Water Alternate Injection per EOP-SPI 4.6

D. Condensate Transfer Alternate Injection per EOP-SPI 4.4

115

Perry NRC Exam 2015

QUESTION RO 58



K&A: Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

AC Electrical Distribution Explanation: Answer C – With the conditions stated, Fast Fire Water using the Diesel Fire Pump is the only method available to add water to the RPV.

A – Incorrect – ADHR Alternate Injection uses the ADHR pump which is powered from a non-class 1E bus that lost power during the LOOP.

B – Incorrect – FPCC Header Alternate Injection uses the SPCU pump which is powered from a non-class 1E bus that lost power during the LOOP.

C – Incorrect – CT Alternate Injection uses the Condensate Transfer Pumps which are powered from a non-class 1E bus that lost power during the LOOP.

Technical Reference(s): EOP-SPI 4.6 Rev 1, ONI-SPI E4 Rev 0, EOP-SPI 4.3 Rev 2, EOP-SPI 4.4 Rev 1

Reference Attached: EOP-SPI 4.6 p 2, ONI-SPI E4 pp 2, 7, EOP-SPI 4.3 p 2, EOP-SPI 4.4 pp 2, 6


Learning Objective (As available): OT-3402-02.F






116

Perry NRC Exam 2015 QUESTION RO 59 A reactor startup is in progress with the following conditions: • Reactor power approximately 16% • Turbine generator ready to synchronize to the grid Vital inverter DB-1-A experienced a failure. Additionally, the static transfer switch failed to shift to the Alternate Source resulting in a loss of power to Bus V-1-A. Based on these conditions, which ONI will be entered?

A. ONI-C11-1, Inability To Move Control Rods due to loss of RC&IS

B. ONI-C34, Feedwater Flow Malfunction due to Motor Feed Pump Flow Controller failing to 100% open

C. ONI-C85, Pressure Regulator Failure due to loss of one channel of Pressure Regulator

D. ONI-N32, Turbine And/Or Generator Trip due to Main Turbine trip caused by loss of Both Speed Signals

117

Perry NRC Exam 2015

QUESTION RO 59



K&A: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

UPS (AC/DC) Explanation: Answer A – With a loss of both the inverter and failure of the UPS to transfer to the regulating transformer, RC&IS is lost and ONI-C11-1 is entered.

B – Incorrect – The Recirc Controller Fails open, not the Pump Flow Controller.

C – Incorrect – Plausible since V-1-A supplies one source to an ABT that supplies part of the SB&PR system – no loss.

D – Incorrect – Plausible since V-1-A supplies power to various main turbine indications. No turbine trips would result from a loss of V-1-A.

Technical Reference(s): PDB-H0044 Rev 7 & ONI-C11-1 Rev 14

Reference Attached: PDB-H0044 pp 16 & 17 and ONI-C11-1 pp 3 & 9


Learning Objective (As available): OT-COMBINED-R14_15-J.1






118

Perry NRC Exam 2015 QUESTION RO 60 The Unit 1 Division 1 battery voltage is reading 130 VDC locally and in the Control Room. The Control Room Operator directs a PO to adjust the FLOAT potentiometer for the on-service battery charger to obtain 135 VDC in accordance with SOI-R42 (Div 1), Section 7.5. Following the adjustment, with ED1A bus loading remaining within the capacity of the Normal Charger, the DIV 1 BATT AMPS meter on H13-P877 will indicate a value in the 1 region with the ability to restore battery voltage to 135 VDC 2 on the number of DC loads.

(1) (2)

A. CHARGE dependent

B. CHARGE independent

C. DISCHARGE dependent

D. DISCHARGE independent

119

Perry NRC Exam 2015

QUESTION RO 60



K&A: Knowledge of the physical connections and/or cause- effect relationships between D.C. Electrical Distribution and the following: Battery charger and battery

DC Electrical Distribution Explanation: Answer B – When the Float voltage is increased, the battery ammeter will indicate in the Charge Region. The charger has the capacity to maintain the battery voltage constant with stable load.

A – Incorrect – The charging rate is determined by the difference in voltage between the battery terminal voltage and the charger output voltage. It is not affected by the loading on the bus provided that the bus loading remains within the capacity of the charger.

C & D – Incorrect – Raising the setting on the FLOAT potentiometer will increase the charger output voltage. Since the battery was initially producing 130 VDC, this action will cause a charge indication, not a discharge indication.

Technical Reference(s): SOI-R42 Rev 18, SDM-R42 Rev 9.

Reference Attached: SOI-R42 pp 34 & 60, SDM-R42 p 11


Learning Objective (As available): OT-COMBINED-R42






120

Perry NRC Exam 2015 QUESTION RO 61 The following conditions exist: • A Loss of Off-Site Power (LOOP) occurred • Divisional diesel generators are supplying their respective buses • DIESEL GENERATOR CONTROL TRANSFER switch is in LOCAL on 1H51-P055A • DIESEL GENERATOR CONTROL TRANSFER switch is in CONT RM on 1H51-P055B The following alarms are then received on H13-P877: Division 1 • DIV 1 DIESEL GENERATOR TROUBLE, (H13-P877-0001-D5) • DG TRIP* LUBE OIL TEMP HIGH, (H13-P877-0001-C3) Division 2 • DIV 2 DIESEL GENERATOR TROUBLE, (H13-P877-0002-D5) • DG TRIP DIFF RELAY LOCKOUT, (H13-P877-0002-E4) What will be the status of Buses EH11 and EH12?

EH11 EH12

A. Energized Energized

B. Energized De-energized

C. De-energized Energized

D. De-energized De-energized

121

Perry NRC Exam 2015

QUESTION RO 61



K&A: Knowledge of the effect that a loss or malfunction of the Emergency Generators (Diesel/Jet) will have on following: A.C. electrical distribution

EDGs Explanation: Answer D – With Div 1 DG in LOCAL control, the LOOP/LOCA trips are not bypassed. The high lube oil temperature will cause the Div 1 DG to trip and de-energize Bus EH11. Also, the Differential Relay Lockout Trip is always active. Therefore, div 2 DG will trip and de-energize Bus EH12.

A – Incorrect – Both Buses will be de-energized. Plausible since some trips are bypassed in a LOOP condition.

B – Incorrect – Bus EH 11 is also de-energized.

C – Incorrect – Bus EH 12 is also de-energized.

Technical Reference(s): SDM-R43 Rev 13, SOI-R43 Rev 43, ARI-H13-P877-001 Rev 11, & ARI-H13-P877-002 Rev 12

Reference Attached: SDM-R43 pp 7, 12, SOI-R43 p 6, ARI-H13-P877-001 pp 31, 43-44, & ARI-H13-P877-002 pp 43-44, 53


Learning Objective (As available): OT-COMBINED-R43_48-F.9






122

Perry NRC Exam 2015 QUESTION RO 62 The plant is operating at rated power with the Safety Related Instrument Air Compressor out of service for corrective maintenance. Air pressures as indicated on ADS AIR STRG PRESS, 1P57-R026A and 1P57-R026B, are 115 psig and slowly lowering. If ADS air receiver pressure continues to lower, the (1) MSIV’s will be affected. Restore ADS air pressure to normal using (2) .

(1) (2)

A. inboard a temporary air compressor

B. inboard Instrument Air System

C. outboard a temporary air compressor

D. outboard Instrument Air System

123

Perry NRC Exam 2015

QUESTION RO 62



K&A: Knowledge of Instrument Air System design feature(s) and or interlocks which provide for the following: Cross-over to other air systems

Instrument Air Explanation: Answer C – IAW SOI-P57 P&L 2.3, if ADS B air pressure lowers to 90 psig, TS 3.6.1.3 needs to be entered. At <45 psig, the MSIV’s may not be leak tight following a DBA LOCA. The ADS Air System is designed to be pressurized with Instrument Air up to ~ 125 psig only. However, to restore air pressure to normal (165 psig), a temporary air compressor is used.

A & B – Incorrect – First part - Inboard MSIV’s are not affected by lowering ADS pressure.

B & D – Incorrect – Second part – The Instrument Air System is used recharge the ADS air system to only 125 psig.

Technical Reference(s): SOI-P57 Rev 17, ARI-H13-P601-019 Rev 17

Reference Attached: SOI-P57 pp3, 15, 17, ARI-H13-P601-019 pp 129-130


Learning Objective (As available): OT-COMBINED-P57-M






124

Perry NRC Exam 2015 QUESTION RO 63 On a loss of off-site power, the TBCC HXs SW TCV, 1P41-F003 will fail (1) and the TBCC HX SW TCV BYP, 1P41-F390 will fail (2) .

(1) (2)

A. open open

B. open close

C. close open

D. close close

125

Perry NRC Exam 2015

QUESTION RO 63



K&A: Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Valves

Component Cooling Water Explanation: Answer D – On a LOOP, the TBCC HXs SW TCV, 1P41-F003 and the TBCC HX SW TCB BYP, 1P41-F290 are designed to fail in the closed position to isolate service water from the TBCC heat exchangers.

A, B, C – Incorrect – Both valves fail in the close position.

Technical Reference(s): SDM-P41 Rev 10 & SOI-P40/41 Rev 11

Reference Attached: SDM-P41 pp 17-18 & SOI-P40/41 p 22


Learning Objective (As available): OT-COMBINED-P41-F






126

Perry NRC Exam 2015 QUESTION RO 64 The plant was operating at rated power with the following conditions: • Reactor Pressure was 1025 psig • Control Rod 30-31 accumulator pressure was 1600 psig Then a malfunction in the Reactor pressure regulating system caused reactor pressure to lower. What effect would a lowering Reactor Pressure have on Control Rod 30-31 scram time? Control Rod 30-31 scram time would _____.

A. become shorter as Reactor Pressure lowered

B. become longer as Reactor Pressure lowered

C. initially become longer and then get shorter once Reactor Pressure lowered <950psig

D. initially become shorter and then get longer once Reactor Pressure lowered <600 psig

127

Perry NRC Exam 2015

QUESTION RO 64



K&A: Knowledge of the effect that a loss or malfunction of the following will have on the Control Rod And Drive Mechanism: Reactor pressure

Control Rod and Drive Mechanism Explanation: Answer A – The accumulator remains operable with accumulator pressure >1520 psig, as reactor pressure lowers scram time will become shorter.

B – Incorrect – opposite of A, misconception of Ball Check valve operation.

C – Incorrect – If the accumulator was inoperable the control rod would be declared slow, (longer scram time) and 950 psig is the lower pressure that the control rod scram times are determined.

D – Incorrect – This is true for an inoperable accumulator.

Technical Reference(s): Tech Spec 3.1.4 and 3.1.5 Reference Attached: Tech Spec 3.1.4 pp 3.1.12 & 14, TS 3.1.5 p 3.1.15-17


Learning Objective (As available): OT-OT-C11_CRDM-F, OT-3037-05-H






128

Perry NRC Exam 2015 QUESTION RO 65 Which of the following describes the operational implications for the Rod Control and Information System (RCIS) when Turbine 1st Stage pressure decreases below the Low Power Alarm Point (LPAP)?

A. Control Rod withdrawal is automatically limited to two notches to prevent any Control Rod from having an excessive rod worth.

B. Control Rod withdrawal is automatically limited to four notches to prevent any Control Rod from having an excessive rod worth.

C. The control rod pattern should be adjusted to match the selected Control Rod sequence during a power reduction prior to reaching the Low Power Setpoint (LPSP).

D. The control rod pattern should be adjusted to match the selected Control Rod sequence during a power increase prior to reaching the High Power Setpoint (HPSP).

129

Perry NRC Exam 2015

QUESTION RO 65



K&A: Ability to predict and/or monitor changes in parameters associated with operating the Rod Control And Information System (RCIS) controls including: First stage shell pressure/turbine load: BWR-6

RC&IS Explanation: Answer C – If power is reduced below the LPSP, the constraints of the RPC will prevent any further control rod movement unless the control rods are in pattern for the selected sequence.

A – Incorrect – This describes operation above the High Power Setpoint.

B – Incorrect – Rod withdrawal is limited to four notches but not to limit rod worth (this is not the significance of the transition zone).

D – Incorrect – The transition zone is activated at 38% power decreasing on a power reduction and allows control pattern adjustment before the LPSP is activated.

Technical Reference(s): SDM-C11(RC&IS) Rev 9 & IOI-3 Rev 52

Reference Attached: SDM-C11(RC&IS) pp 29, 48-49 & IOI-3 pp 85-86


Learning Objective (As available): OT-COMBINED-C11(RC&IS)-D & L.1

Question Source: Bank # Perry Audit 2003 # RO-88 Modified Bank # New





130

Perry NRC Exam 2015 QUESTION RO 66 The plant was operating at rated power when the following annunciators alarmed: • RECIRC A INNER SEAL FLOW HI/LO • RECIRC A OUTER SEAL LEAKAGE HI • RECIRC A/B TEMP HI B33-R601 indicates Recirc A Seal Cavity Temp is 215°F Leakage from the Reactor Recirculation Pump seal is considered (1) leakage. The action required to address this condition is: (2) .

(1) (2)

A. identified trip Recirc Pump A

B. identified transfer Recirc Pump A to slow speed

C. unidentified trip Recirc Pump A

D. unidentified transfer Recirc Pump A to slow speed

131

Perry NRC Exam 2015

QUESTION RO 66



K&A: Ability to (a) predict the impacts of the following on the Recirculation System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation system leak

Recirculation Explanation: Answer A – Leakage from the Recirc Pump seal is considered ‘identified’ leakage. When Recirc pump seal temperature reaches 210°F, the operator is required to trip the pump.

B & D – Incorrect – Second part – This is the action to consider when seal temperature starts to rise.

C & D – Incorrect – First part – Leakage from the Recirc Pump seals is routed to the Equipment Drain Sump and is considered ‘identified’ leakage.

Technical Reference(s): ARI-H13-P680-04 Rev 22, ARI-H13-P601-18 Rev 15

Reference Attached: ARI-H13-P680-04 p 5-6, 41-42, & 109, ARI-H13-P601-18 p 3


Learning Objective (As available): OT-COMBINED-B33-C1 & N.6, and OT-COMBINED-E31-I.1






132

Perry NRC Exam 2015 QUESTION RO 67 The following conditions exist for Reactor Recirculation Hydraulic Power Unit A: • Subloop 1 – OPERATIONAL, LEAD, READY, PRESSURIZED • Subloop 2 – READY Which one of the following conditions will initiate annunciator FCV A MOTION INHIBITED and cause Reactor Recirculation FCV A to hydraulically lockup?

A. Oil reservoir level decreases to 60 gal.

B. Oil reservoir temperature increases to 145°F.

C. Subloop 1 discharge pressure decreases to 1650 psig.

D. Subloop 1 discharge filter differential pressure increases to 35 psid.

133

Perry NRC Exam 2015

QUESTION RO 67



K&A: Ability to monitor automatic operations of the Recirculation Flow Control System including: Flow control valve operation: BWR-5,6

Recirculation Flow Control Explanation: Answer A – If HPU reservoir tank level lowers to ≤ 60 gallons, the HPU will shift to Maintenance Mode and ‘lockup’.

B – Incorrect – Oil reservoir temperature has to be ≥ 150°F to lockup HPU.

C – Incorrect – This condition will cause a transfer from Subloop 1 to Subloop 2 but will not hydraulically lockup the FCV.

D – Incorrect - This condition will cause a FCV A HPU NEEDS MAINT alarm but will not hydraulically lockup the FCV.

Technical Reference(s): ARI-H13-P680-04 Rev 22 Reference Attached: ARI-H13-P680-04 p 15


Learning Objective (As available): OT-COMBINED-B33-C






134

Perry NRC Exam 2015 QUESTION RO 68 RHR A and B loops were placed in Suppression Pool Cooling following a scram with MSIV closed. Then, an automatic initiation of LPCI occurred. Which of the following conditions must be met in order to lineup RHR Loop A to cool the Suppression Pool in accordance with SOI-E12, Residual Heat Removal System?

A. The LPCI initiation logic must be bypassed on panel P629 in order to open RHR A TEST VALVE TO SUPR POOL, 1E12-F024A.

B. The LPCI initiation logic must be reset on panel P601 in order to open RHR A TEST VALVE TO SUPR POOL, 1E12-F024A.

C. 110 seconds must elapse in order to close RHR A HX’S BYPASS VALVE, 1E12-F048A.

D. 10 minutes must elapse in order to close RHR A HX’S BYPASS VALVE, 1E12-F048A.

135

Perry NRC Exam 2015

QUESTION RO 68



K&A: Ability to manually operate and/or monitor in the control room: Valve lineup

RHR/LPCI: Torus/Pool Cooling Mode Explanation: Answer D – Upon a LPCI initiation signal, RHR A/B will shift from Suppression Pool Cooling Mode to LPCI injection mode. This will cause the HX bypass valve to open. The HX bypass valve cannot be closed for 10 minutes following a LPCI initiation.

A – Incorrect – The LPCI initiation logic does not need to be bypassed. The 1E12-F024A/B has override capability.

B – Incorrect – The LPCI initiation logic does not need to be reset. The 1E12-F024A/B has override capability.

C – Incorrect – The 110 second time delay is to allow the 1E12-F024A to close before the 1E12-F048A/B opens upon an LPCI initiation.

Technical Reference(s): SOI-E12 Rev 63 & SDM-E12 Rev 3

Reference Attached: SOI-E12 pp 21-22, 29-30 & SDM-E12 p 44


Learning Objective (As available): OT-COMBINED-E12-F






136

Perry NRC Exam 2015 QUESTION RO 69 A LOCA occurred in containment. Hydrogen Analyzers were directed to be started. Twenty minutes after the Hydrogen Analyzers were started, the BOP Operator reports that blue PERMISSIVE lights are ON above each H2 Analyzer Channel Select Switch. The Hydrogen Analyzers are controlled from (1) . And, the blue PERMISSIVE light being illuminated indicates that the 2 .

(1) (2)

A. H13-P800, Heating Ventilation & Air Conditioning Control Panel (Unit 1)

selected area solenoid valve is open

B. H13-P800, Heating Ventilation & Air Conditioning Control Panel (Unit 1)

sample readings are representative for the area selected

C. H13-P904, Common Heating, Ventilation, & Air Conditioning Control Panel

selected area solenoid valve is open

D. H13-P904, Common Heating, Ventilation, & Air Conditioning Control Panel

sample readings are representative for the area selected

137

Perry NRC Exam 2015

QUESTION RO 69



K&A: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

Primary CTMT and Aux Explanation: Answer B – The controls for the Hydrogen Analyzers are on H13-P800. The blue Permissive light indicates that 3.5 minutes have elapsed since the channel was selected and indicates that the sample reading is representative. The sample should be valid after about 15 minutes.

A & C – Incorrect – The selected area solenoid valve will open upon operation of the Channel Select switch after a short time delay.

C & D – Incorrect - The controls for the Hydrogen Analyzers are on H13-P800. Plausible since the panel names are very similar.

Technical Reference(s): SDM-M51 Rev 9 & SOI-M51/56 Rev 26

Reference Attached: SDM-M51 p 13 & SOI-M51/56 pp 6 & 43


Learning Objective (As available): OT-COMBINED-M51_56-N






138

Perry NRC Exam 2015 QUESTION RO 70 The plant was operating at rated power when the following occurred: • At 09:45 an inadvertent RCIC initiation occurred and all ONI actions were subsequently performed • At 09:55 ECC pump A was overridden off due to high vibrations • At 09:58 a steam leak developed in the Drywell • At 10:03 high Drywell pressure caused an automatic scram • At 10:04 a leak in the scram discharge volume caused Containment pressure to start increasing At 10:12 the following conditions exist: • Containment pressure is 7.8 psig • Drywell pressure is 7.6 psig • RPV level is 10 inches Based on the above information, a Containment Spray automatic initiation signal (1) been generated and ECC pump A (2) be running.

(1) (2)

A. has will

B. has will not

C. has not will

D. has not will not

139

Perry NRC Exam 2015

QUESTION RO 70



K&A: Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI: Containment Spray System Mode and the following: Component cooling water systems

RHR/LPCI: CTMT Spray Mode Explanation: Answer C – Containment Spray will not automatically initiate for 10.85 minutes following a LPCI initiation signal to allow for adequate core cooling. After 10 minutes if DW pressure is > 1.68 psig and Containment pressure is > 7.72 psig, then Containment Spray will initiate. Also, a new initiation (LPC) signal will restart a previously overridden ECC pump.

A & B – Incorrect – First part – Containment Spray has not initiated as 10 minutes have not elapsed since the LPCI initiation signal.

A & C – Incorrect – Second Part – ECC A pump will not restart on a Containment Spray initiation signal. Since it started on a LPCI initiation signal, then overridden off, only LOOP or RCIC initiation signals will auto start ECC A pump.

Technical Reference(s): SDM-P42 Rev 12, SDM-E12 Rev 3, & SOI-E12 Rev 63

Reference Attached: SDM-P42 pp 9-10, SDM-E12 p 33, & SOI-E12 p 27


Learning Objective (As available): OT-COMBINED-P42-F & OT-COMBINED-E12-F






140

Perry NRC Exam 2015 QUESTION RO 71 The plant is operating at 50% rated power with the following conditions: • Hotwell Pumps A and C are in operation • Condensate Booster Pumps B and C are in operation If a lockout then occurs on Bus H11, which condensate system pumps will remain in operation?

A. Hotwell Pump A and Condensate Booster Pump B.

B. Hotwell Pump A and Condensate Booster Pump C.

C. Hotwell Pump C and Condensate Booster Pump B.

D. Hotwell Pump C and Condensate Booster Pump C.

141

Perry NRC Exam 2015

QUESTION RO 71



K&A: Knowledge of electrical power supplies to the following: System pumps

Reactor Condensate Explanation: Answer C – A lockout on Bus H11 prevents bus H11 from transferring to its alternate source and will de-energize. The pumps fed from Bus H11 will not remain in operation. HW Pump C and CBP B will remain running.

A – Incorrect – HW Pump A will lose power.

B – Incorrect – HW Pump A & CBP C will lose power.

D – Incorrect – CBP C will lose power.

Technical Reference(s): SDM-R10 Rev 10 & PDB-H006 Rev 0

Reference Attached: SDM-R10 pp 15, 31 & PDB-H006 pp 3-4


Learning Objective (As available): OT-COMBINED-N21_N63-C.6 & -C.6 and OT-COMBINED-R10-C.7 & -F.18






142

Perry NRC Exam 2015 QUESTION RO 72 The plant was operating at 90% rated power when 1N27-F135, HEATER 6A & 6B FDW BYPASS VLV began to open due to an electrical problem. Initially, reactor power indication on APRM’s will (1) due to increasing (2) .

(1) (2)

A. rise core flow

B. rise core inlet subcooling

C. lower core flow

D. lower core inlet subcooling

143

Perry NRC Exam 2015

QUESTION RO 72



K&A: Knowledge of the effect that a loss or malfunction of the Reactor Feedwater System will have on following: Core inlet subcooling

Reactor Feedwater Explanation: Answer B – When 1N27-F135 starts to open, Feedwater not heated by the 6A & 6B Feedwater Heaters is admitted to the reactor, thus increasing subcooling. The cooler Feedwater will cause power on the APRM’s to rise. The effect is the same as a loss of extraction steam to the 6 feedwater heaters. This was demonstrated in the simulator on 7/22/14 – RJT.

A & C – Incorrect – Second part – The Feedwater level control system maintains the same amount of Feedwater going to the vessel. And, Recirc FCV position has not changed. Therefore, there is no increase in core flow.

C & D – Incorrect – First part – Reactor power will go up not down.

Technical Reference(s): OT-3302-08(LP) Rev 4, OT-3302-09(LP) Rev 4, OT-3301-08(LP) Rev 5, ONI-N36 Rev 18

Reference Attached: OT-3302-08(LP) p 25, OT-3302-09(LP) p 36, OT-3301-08(LP) p 58, ONI-N36 p 4


Learning Objective (As available): OT-3035-13(LP)-A.2 & OT-3302-08






144

Perry NRC Exam 2015 QUESTION RO 73 The Mechanical Vacuum Pumps are designed to operate up to (1) . And, at rated power, the in service Offgas Catalytic Recombiner outlet temperature will be approximately (2) .

(1) (2)

A. 5% reactor power 350°F

B. 5% reactor power 610°F

C. 5% hydrogen concentration 350°F

D. 5% hydrogen concentration 610°F

145

Perry NRC Exam 2015

QUESTION RO 73



K&A: Knowledge of Offgas System design feature(s) and/or interlocks which provide for the following: The prevention of hydrogen explosions and/or fires

Offgas Explanation: Answer B – The mechanical vacuum pumps are limited to operation up to 5% reactor power. Above 5% reactor power, hydrogen concentration can increase above the flammable limit of 4%. Additionally, the on service recombiner will be ~610°F due to the recombination process. (610°F was validated in Plant on 8/19/14 at 100% power)

A & C – Incorrect – Second part – 350°F is the standby temperature for the off-service recombiner and the temperature where recombination starts.

C & D – Incorrect – First part – The mechanical vacuum pumps are limited to 5% reactor power.

Technical Reference(s): SOI-N64/62 Rev 35, Lesson Plan N62 Rev 4, Lesson Plan N64 Rev 1, & SDM-N64 Rev 0

Reference Attached: SOI-N64/62 p 5, Lesson Plan N62 p 7, Lesson Plan N64 p 24, & SDM-N64 pp 2, 12


Learning Objective (As available): OT-COMBINED-N64-C & H






146

Perry NRC Exam 2015 QUESTION RO 74 A fire has been detected in the Technical Support Center. In order to automatically initiate the installed Halon system, 1 detector in (1) must go into alarm. Halon extinguishes a fire by (2) .

(1) (2)

A. 2 zones chemically inhibiting the combustion reaction

B. 2 zones displacing all the oxygen needed to support combustion

C. 1 zone chemically inhibiting the combustion reaction

D. 1 zone displacing all the oxygen needed to support combustion

147

Perry NRC Exam 2015

QUESTION RO 74



K&A: Knowledge of the operational implications of the following concepts as they apply to Fire Protection System: Effect of Halon on fires: Plant-Specific

Fire Protection Explanation: Answer A – The installed Halon extinguishing systems require 1 detector in 2 zones to go into alarm (cross zone). The first detector will only give an alarm. Also, the effect that has on a fire is to interrupt the chemical reaction.

B – Incorrect – Second part – This is how CO2 extinguishes a fire, not Halon.

C & D – First Part – Multiple detectors in a single zone will not cause a Halon discharge, only an alarm.

Technical Reference(s): SOI-P54(GAS) Rev 7 & SDM-P54_Halon Rev 1.

Reference Attached: SOI-P54(GAS) p 17 & SDM-P54_Halon pp 2-3, 8 12-13


Learning Objective (As available): OT-COMBINED-P54_Halon-C & -F






148

Perry NRC Exam 2015 QUESTION RO 75 A LOCA occurred. The ADS INHIBIT switches failed and all ADS valves opened. Which of the following set of conditions maintain adequate core cooling? Reference Provided: EOP-SPI Supplement (partial)

A. RPV level stable at 20” on Fuel Zone RPV Pressure is 15 psig CRDH A Pump injecting at rated flow Drywell Temperature is 260°F

B. RPV level stable at 20” on Fuel Zone RPV Pressure is 20 psig LPCI A & B Pumps injecting at rated flow Drywell Temperature is 280°F

C. RPV level stable at -50” on Fuel Zone RPV Pressure is 20 psig LPCI C Pump injecting at rated flow Drywell Temperature is 200°F

D. RPV level stable at -40” on Fuel Zone RPV Pressure is 200 psig No injection into the RPV Drywell Temperature is 240°F

149

Perry NRC Exam 2015

QUESTION RO 75



K&A: Knowledge of effect that a loss of malfunction of the following will have on the Reactor Vessel Internals: Heat transfer mechanisms

Reactor Vessel Internals Explanation: Answer D – Steam cooling has been established. Adequate core cooling exists with no injection into the RPV as long as water level is > -42.5” below TAF.

A – Incorrect – With Drywell temperature > RPV saturation temperature (obtained from steam tables), no level indication is considered accurate (RPV Level Caution) and RPV Flooding must be performed.

B – Incorrect – With Drywell temperature > RPV saturation temperature (obtained from steam tables), no level indication is considered accurate (RPV Level Caution) and RPV Flooding must be performed.

C – Incorrect – RPV level is below the steam cooling RPV level (-25) with injection.

Technical Reference(s): EOP-1 Bases Rev 4 Reference Attached: EOP-1 Bases 55

Proposed references to be provided to applicants during examination: EOP-SPI Supplement (partial)

Learning Objective (As available): OT-3402-02-F & OT-3402-01-C.1

Question Source: Bank # Modified Bank # RQL-0021 New





150

Perry NRC Exam 2015 QUESTION SRO 1 Which of the following must be directly supervised by a Senior Reactor Operator?

A. Raising and lowering of irradiated fuel bundles in the fuel prep machine during MODE 5.

B. Movement of individual fuel pins during in pool inspection of irradiated fuel assemblies during MODE 1.

C. Movement of irradiated fuel assemblies from one location in the fuel pool to a different location in the fuel pool during MODE 4.

D. Movement of irradiated fuel assemblies from one location in the reactor vessel to a different location in the reactor vessel during MODE 5.

151

Perry NRC Exam 2015

QUESTION SRO 1



K&A: Knowledge of procedures and limitations involved in core alterations.

Generic Explanation: Answer D – The movement described is a Core Alteration by Tech Specs. As such, it is required to be directly supervised by a Licensed Senior Reactor Operator per PAP-0802.

A – Incorrect - Movement of irradiated fuel during MODE 5 is required to be directly supervised by a Fuel Handling Supervisor with the exception of Core Alterations.

B – Incorrect - The Fuel Handling Supervisor can supervise the movement in MODE 1. Fuel Handling Supervisors are not required to be SROs.

C – Incorrect - This is not defined as a Core Alteration and may be supervised by a Fuel Handling Supervisor.

Technical Reference(s): Tech Specs 1.0, IOI-9 Rev 32, NOP-OP-1002 Rev 9, NOP-NF-3001 Rev 8, & SOI-F15 Rev 18

Reference Attached: Tech Specs p 1.0-2, IOI-9 p 7, NOP-OP-1002 pp 19-20, NOP-NF-3001 p 14, & SOI-F15 p 124


Learning Objective (As available): OT-3602-01-B.4

Question Source: Bank # Columbia 2011 # SRO-05 Modified Bank # New



10 CFR Part 55 Content: 55.41 55.43 b(6)

Comments: SRO Justification – Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)] • Administrative requirements associated with refueling activities, such as approvals required to amend core

loading sheets or administrative controls of potential dilution paths and/or activities.

152

Perry NRC Exam 2015 QUESTION SRO 2 The plant is in a refueling outage with the following conditions: • A full core off-load was performed due to work on the vessel bottom head. • Core reload is scheduled to commence in one hour. • Control Rods 22-59 and 58-35 are withdrawn and uncoupled for CRDM replacement and are not

scheduled to be completed for 4 more hours. • All other control rods are inserted. Based on this information, core reload ____

A. can commence as scheduled using the Spiral Reload method.

B. cannot commence until control rods 22-59 and 58-35 are reinserted.

C. cannot commence until leads are lifted and jumpers are installed to defeat RCIS interlocks.

D. can commence as scheduled as long as no fuel is loaded around control rods 22-59 and 58-35 until they are reinserted.

153

Perry NRC Exam 2015

QUESTION SRO 2



K&A: Knowledge of the refueling process.

Generic Explanation: Answer B – Core reload is not allowed with any control rod withdrawn.

A – Incorrect – Plausible since this is mentioned in TS 3.10.6. However, the spiral reload with control rods withdrawn has not been analyzed for Perry.

C – Incorrect – Plausible since this is done during “At All Times” to allow bridge movement over core.

D – Incorrect – Plausible since maintaining adequate shutdown margin is a concern during core reload

Technical Reference(s): FTI-D06 Rev 17 & IOI-9 Rev 32 Reference Attached: FTI-D06 p 3 & IOI-9 pp 55 & 116


Learning Objective (As available): OT-3035-12(LP)-E, OT-3037-13-A, & OT-3602-01-D.3

Question Source: Bank # Modified Bank # New X




Comments: SRO justification - Fuel handling facilities and procedures. [10 CFR 55.43(b)(7)]

• Assessment of surveillance requirements for the refueling mode.

154

Perry NRC Exam 2015 QUESTION SRO 3 A proposed plant modification must always have prior approval from the NRC if it ____.

A. requires a 50.59 evaluation

B. involves a system described in the UFSAR

C. involves a system included in the Technical Specifications

D. results in a design basis limit for Primary Containment being altered

155

Perry NRC Exam 2015

QUESTION SRO 3



K&A: Knowledge of the process for making design or operating changes to the facility.

Generic Explanation: Answer D – Design bases alterations requires a license amendment prior to implementation IAW 10CFR50.59.

A – Incorrect - This evaluation will determine if NRC approval is required.

B – Incorrect – This must have a 50.59 evaluation but not necessarily NRC approval.

C – Incorrect – This must have a 50.59 evaluation but not necessarily NRC approval.

Technical Reference(s): 10CFR50.59, NOBP-LP-4003A Rev 7, Form NOP-LP-4003-3

Reference Attached: 10CFR50.59 pp 1-2, NOBP-LP-4003A pp 9 & 54, and Form NOP-LP-4003-3


Learning Objective (As available): OT-3037-15D

Question Source: Bank # Perry 2009 SRO # 21 Modified Bank # New




Comments: SRO Justification - Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

• 10 CFR 50.59 screening and evaluation processes.

156

Perry NRC Exam 2015 QUESTION SRO 4 In accordance with Technical Specification Bases, the RPV water level high (Level 8) trip function, of Reactor Protection System, ensures that ____.

A. fuel clad and reactor coolant pressure boundary challenge is minimized

B. the Minimum Critical Power Ratio (MCPR) does not exceed the MCPR Safety Limit

C. along with safety relief valves, limits the peak reactor pressure to less than the ASME code limits

D. the possibility of fuel damage is minimized and to reduce the amount of energy being added to the coolant

157

Perry NRC Exam 2015

QUESTION SRO 4



K&A: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Generic Explanation: Answer B – > per Tech Spec Bases for RPS Instrumentation on high Rx Water level.

A – incorrect – TS Bases for Steam Dome Pressure High

C – incorrect – TS Bases for APRM Fixed Neutron Flux High

D – incorrect – TS bases for DW Pressure High

Technical Reference(s): TS 3.3.1.1 Bases Rev 7 Reference Attached: TS 3.3.1.1 Bases p B 3.3-13


Learning Objective (As available): OT-3037-07.G

Question Source: Bank # Perry NRC Exam 2009 # SRO-08 Modified Bank # New

Question History: Previous NRC Exam Perry NRC Exam 2009



Comments: SRO Justification - B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

• Knowledge of TS bases that are required to analyze TS required actions and terminology.

158

Perry NRC Exam 2015 QUESTION SRO 5 The Plant is in a refueling outage. A contract worker performing an inspection of a twice burnt fuel bundle in the fuel inspection machine received an unexpected exposure of 5105 mRem TEDE when the upper stop failed on the fuel inspection machine. Not including Events Of Potential Public Interest, what notification to the NRC is required? Reference Provided: NOP-OP-1015 & NOBP-OP-1015

A. An Immediate Notification

B. A 4 Hour Notification

C. A 24 Hour Notification

D. A Thirty Day Notification

Due to resolution of a post-examination comment, this question has been deleted from the exam. See Perry Examination Report (05000440/2015301) for an explanation.

159

Perry NRC Exam 2015

QUESTION SRO 5



K&A: 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.

Generic Explanation: Answer C – Per 10CFR20.2202(b), Each licensee shall, within 24 hours of discovery of the event, report any event involving loss of control of licensed material possessed by the licensee that may have caused, or threatens to cause, any of the following conditions: (1) An individual to receive, in a period of 24 hours (i) A total effective dose equivalent exceeding 5 rems

A – Incorrect – Plausible since a radiation exposure >25 REM is immediately reportable.

B – Incorrect – Plausible as a 4-hour notification is required for an incident involving the spent fuel (72.75(b)(1)).

D – Incorrect – Plausible since this is a requirement for lost or stolen SNM.

Technical Reference(s): NOP-OP-1015 Rev 0 & NOBP-OP-1015 Rev 0

Reference Attached: NOP-OP-1015 p 34 & NOBP-OP-1015 pp 111-112

Proposed references to be provided to applicants during examination: NOP-OP-1015 & NOBP-OP-1015

Learning Objective (As available): OT-EVENTREPORT_PY-01-A





Comments: SRO Justification - Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.[10 CFR 55.43(b)(4)]

• The SRO is responsible for making NRC Notifications.

160

Perry NRC Exam 2015 QUESTION SRO 6 (This exam question contained security material and has been redacted from the written exam.)

A. 1, 2, 3

B. 2, 3, 4

C. 1, 3, 4

D. 1, 2, 4

161

Perry NRC Exam 2015

QUESTION SRO 6



K&A: Knowledge of procedures relating to a security event (non-safeguards information).

Generic (Exam material has been redacted due to security purposes.)

Technical Reference(s): ONI-P56-2 Rev 21 Reference Attached: ONI-P56-2 pp 7-8 & 12-14







Comments: SRO Justification - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

162

Perry NRC Exam 2015 QUESTION SRO 7 The following conditions exist: • The Shift Manager acting as the Emergency Coordinator has declared a General Emergency. • None of the Emergency Response Facilities are activated yet. Duties that can be delegated to another emergency team member by the Shift Manager include (1) and (2) .

(1) (2)

A. Conduct site accountability actions. Reclassify the event if necessary

B. Direct notifications to the state of Ohio Provide information and assistance to the Public Information Organization

C. Direct activation of the designated Emergency Response Facilities

Request mobilization of the Corporate Assistance Center (CAC)

D. Recommend protective actions for the general public to state and counties

Coordinate and direct the actions necessary to terminate or mitigate the effects of the emergency

163

Perry NRC Exam 2015

QUESTION SRO 7



K&A: Knowledge of SRO responsibilities in emergency plan implementation.

Generic Explanation: Answer C – IAW EPI-A2, the three non-delegable duties of the Shift Manager acting as the Emergency Coordinator are; Determine the emergency classification including reclassification and termination; direct notification of offsite agencies and organizations; and recommend protective actions for the general public to state and local county officials. These 2 actions can be delegated.

A – Incorrect – Reclassification of the Event is not delegable. This is plausible as the operator may only remember that initial Classification is not delegable.

B – Incorrect - Directing notifications to offsite agencies is not delegable. This is plausible since the performance of the notifications is done by a communicator.

D – Incorrect – Recommendations for PARs are not delegable. Plausible since the Shift Lead chemistry Tech develops the PARs prior to the TSC/EOF becoming operational. However, ‘recommendation’ of the PARs is the Shift Managers responsibility.

Technical Reference(s): EPI-A2 Rev 18 Reference Attached: EPI-A2 pp 2-4


Learning Objective (As available): EPL-0801-01-3




10 CFR Part 55 Content: 55.41 55.43 x

Comments: SRO Justification - Plant Specific Exemptions – See page 10 of Clarification Guidance for SRO-only Questions. The following Tasks are uniquely linked to the Shift Manager (Emergency Coordinator) position: 344-502-05-03 - Determine Protective Action Recommendations (PARs) and Ensure that the State and Counties are Notified 344-019-05-03 - Classify/Reclassify Emergency Events Requiring Emergency Plan Implementation 344-020-05-03 - Ensure that Required Notifications of On-Site and Off-Site Personnel of Emergency Events are made The following Objective is uniquely linked to the emergency coordinator position: EPL-0801-01-3: Describe EOF Emergency Coordinator actions and responsibilities during an emergency.

164

Perry NRC Exam 2015 QUESTION SRO 8 Thirty minutes ago the plant was operating at rated power when breaker 1CB6, BUS ED-1-C MAIN BREAKER tripped open due to a battery fault. The following annunciators have alarmed on 1H13-P601: • DIV 3 BATTERY DC SYSTEM TROUBLE • DC BUS ED-1-C UNDERVOLTAGE The current status of High Pressure Core Spray is (1) . If bus ED-1-C is energized from bus ED-2-C via the Unit 2 battery 2E22-S005 and Unit 2 Normal Charger EFD-2-C, then bus ED-1-C will be (2) .

(1) (2)

A. Inoperable Operable

B. Inoperable Inoperable

C. Operable Inoperable

D. Operable Operable

165

Perry NRC Exam 2015

QUESTION SRO 8



K&A: Knowledge of annunciator alarms, indications, or response procedures.

Partial or Total Loss of DC Pwr / 6 Explanation: Answer B – With CB-6 opening, Div 3 DC Bus de-energizes. Per TS 3.8.7 Condition D, HPCS must be declared INOP immediately. Additionally, per TS 3.8.4 Bases, the Unit 2 Normal changer is not listed for OPERABILITY of Div 3 DC subsystem (not powered from a Class 1E source).

A – Incorrect – With ED-1-C powered from the Unit 2 charger, the bus is INOP.

C – Incorrect – When Div 3 DC becomes INOP; HPCS must be declared INOP immediately. Plausible since this may be confused with the 1-hour to verify RCIC/HPCS Operable with HPCS/RCIC INOP.

D – Incorrect - When Div 3 DC becomes INOP; HPCS must be declared INOP immediately. Plausible since this may be confused with the 1-hour to verify RCIC/HPCS Operable with HPCS/RCIC INOP. With ED-1-C powered from the Unit 2 charger, the bus is INOP.

Technical Reference(s): ARI-H13-P601-16 Rev 17, TS 3.8.4 Bases, Rev 7, TS 3.8.7, TS Bases 3.8.7 Rev 1

Reference Attached: ARI-H13-P601-16 pp 87 & 89, TS 3.8.7 pp 3.8-36 & 37, TS Bases 3.8.4 p B3.8-53, & TS 3.8.7 Bases p B3.8-73


Learning Objective (As available): OT-COMBINED-R10-K.6





Comments: SRO Justification - Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]


166

Perry NRC Exam 2015 QUESTION SRO 9 The following conditions exist: • The plant is in MODE 4. • RHR Loop ‘A’ is operating in the Shutdown Cooling mode. • Reactor Recirculation Pump ‘A’ is operating in slow speed. • SHUTDOWN COOLING OTBD SUCT ISOL VLV (E12-F008) closes due to a failed relay in the

Division 1 NS4 RHR Isolation logic. • The isolation signal cannot be reset. • Reactor water temperature is increasing but still within the specified temperature band. Which of the following describes the status of the RHR loop(s); including an alternate method of decay heat removal the Unit Supervisor could establish per ONI-E12-2, Loss of Decay Heat Removal?

A. Only RHR Loop ‘A’ is unavailable for shutdown cooling; RHR Loop ‘B’ should be placed in the Shutdown Cooling Mode using the LPCI injection return flow path.

B. Only RHR Loop ‘A’ is unavailable for shutdown cooling; RHR Loop ‘B’ should be placed in the Shutdown Cooling Mode using the normal shutdown cooling return path.

C. Both RHR loops are unavailable for shutdown cooling; RWCU should be started as an alternate shutdown cooling method.

D. Both RHR loops are unavailable for shutdown cooling; A second Reactor Recirculation Pump should be started as an alternate shutdown cooling method.

167

Perry NRC Exam 2015

QUESTION SRO 9



K&A: Ability to determine and/or interpret the following as they apply to Loss Of Shutdown Cooling: RHR/shutdown cooling system flow

Loss of Shutdown Cooling / 4 Explanation: Answer C – With E12-F008 closed, both loops of SDC are unavailable since E12-F008 is a common suction isolation valve. ONI-E12-2 (Loss of Decay Heat Removal) Attachment 2 Step 4 gives a list of alternate DHR systems to start prior to temperature going out of band. RWCU is on this list.

A & B – Incorrect – Neither loop of RHR would be available since the 1E12-F008 is a common suction line isolation valve.

D – Incorrect - Starting a second recirc pump in this situation is not required and is not an alternate decay heat removal method. It would add pump heat to the RCS.

Technical Reference(s): ONI-E12-2 Rev 33 & Drawing 302-642 Rev JJ.

Reference Attached: ONI-E12-2 pp 20, 22, 24 & Drawing 302-642 (Partial)


Learning Objective (As available): OT-COMBINED-E12-I.1






The applicant’s knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure’s content is required to correctly answer the written test item.

168

Perry NRC Exam 2015 QUESTION SRO 10 The plant was operating at rated power when an MSIV isolation occurred. Five minutes after the MSIV isolation, the following conditions exist: • Not all control rods are inserted • Pressure control is on SRV’s with one SRV open and a second SRV cycling on setpoint. • RPV level has been stabilized at 178” using Feedwater • Suppression Pool temperature is 100°F and rising • All appropriate EOP’s have been entered Based on the above conditions, which of the following describes the action(s) currently required?

A. Maintain RPV water level 178 to 219 inches

B. Maintain RPV water level 150 to 219 inches

C. Terminate and Prevent ECCS injection and maintain RPV water level -25 to 219 inches

D. Terminate and Prevent ECCS injection and Feedwater injection and maintain RPV water level -25 to 100 inches

169

Perry NRC Exam 2015

QUESTION SRO 10



K&A: Ability to determine and/or interpret the following as they apply to High Reactor Pressure: Reactor power

High Reactor Pressure / 3 Explanation: Answer D – With an MSIV isolation, a reactor high pressure condition would cause a reactor scram. Based on not all control rods inserted and 1 SRV open and 1 SRV cycling, Rx power is > 5% (APRM down scale is 4%). Therefore, the US would transition to the EOP-1A flowchart. The conditions of RPV level > 100” and APRM’s not downscale are met. The US should direct T&P of ECCS and FW and lower level to -25 to 100 inches (nominal band is 50 to 100”).

A & B – Incorrect – These are the level bands directed from EOP-1 (Nominal & SRV cycling)

C – Incorrect – This is the level band directed if in an ATWS and power is <4%.

Technical Reference(s): PYBP-POS-30 Rev 2, EOP-1 Chart Rev E, & EOP-1A Chart Rev E

Reference Attached: PYBP-POS-30 p 8, EOP-1 Chart Partial, & EOP-1A Partial


Learning Objective (As available): OT-3402-03






• Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

170

Perry NRC Exam 2015 QUESTION SRO 11 The plant is operating at rated power when the following alarms are received: • H13-P601-0020-E3 - DRYWELL PRESS A HIGH • H13-P601-0017-D5 - DRYWELL PRESS B HIGH • H13-P601-0020-F4 - CONTAINMENT TEMP A HIGH • H13-P601-0017-D2 - CONTAINMENT TEMP B HIGH • H13-P601-0018-A1 - DRYWELL IDENTIFIED LEAK RATE HIGH • H13-P601-0018-A2 - CNTMT IDENTIFIED LEAK RATE HIGH Drywell pressure per validated SPDS is 0.5 psig and rising slowly Drywell temperature per validated SPDS is 142°F and rising slowly Containment temperature per validated SPDS is 96°F and rising Containment pressure per validated SPDS is 0.3 psig and rising Based on the above conditions, which action has the highest priority?

A. Shutdown the reactor and enter EOP-1, RPV Control

B. Blowdown the RPV per EOP-4-2, Emergency Depressurization

C. Operate available Containment cooling per EOP-2, Containment Control

D. Initiate Containment Spray per EOP-SPI 3.1, Containment Spray Operation

171

Perry NRC Exam 2015

QUESTION SRO 11



K&A: Ability to prioritize and interpret the significance of each annunciator or alarm.

High Containment Temperature / 5 Explanation: Answer C – Based on the indications, containment temperature is above the EOP-2 entry criteria. To control containment temperature containment cooling is maximized. Although the DW & Containment Identified leak alarms are indicative of possible leakage, these are not a priority.

A – Incorrect – This would be correct after containment cooling is maximized if containment temperature continued to go up and approach 185°F.

B – Incorrect – This would be correct if containment temperature would exceed 185°F or containment pressure could not be restored and maintained below PSP

D – Incorrect – Containment Spray is required prior to reaching 185°F. However, at 0.3 psig, the Containment Spray Initiation Limit is not yet reached.

Technical Reference(s): EOP-2 Chart rev C, ARI-H13-P601-0017 Rev 13, ARI-H13-P601-0018 Rev 15, & ARI-H13-P601-0020 Rev 17

Reference Attached: EOP-2 Chart partial, ARI-H13-P601-0017 pp 47 & 53, ARI-H13-P601-0018 pp 3, 4, 7, & 8, & ARI-H13-P601-0020 pp 69 & 83


Learning Objective (As available): OT-3402-07-C

Question Source: Bank # Clinton 2008 # SRO-10 Modified Bank # New






172

Perry NRC Exam 2015 QUESTION SRO 12 The plant is operating at rated power. While the Unit Supervisor was reviewing Technical Specification Rounds he discovered the following: • Containment Average Air Temp 78°F • Suppression Pool validated level 17.2 feet • Half of the suppression pool temperature detectors on H13P868 and H13-P869 read 77.5°F • Half of the suppression pool temperature detectors on H13P868 and H13-P869 read 85°F Based on the readings from Tech Spec Rounds, ____.

A. All Suppression Pool temperature instruments are OPERABLE

B. The Suppression Pool temperature instruments reading 77.5°F are OPERABLE

C. The Suppression Pool temperature instruments reading 85°F are OPERABLE

D. None of the Suppression Pool temperature instruments are OPERABLE

173

Perry NRC Exam 2015

QUESTION SRO 12



K&A: Ability to determine and/or interpret the following as they apply to Low Suppression Pool Water Level: Suppression pool temperature

Low Suppression Pool Wtr Lvl / 5 Explanation: Answer C – Eight of sixteen SP level instruments are uncovered and exposed to the containment atmosphere when SP level drops below 17.33’. In order for the SP level instruments to perform their intended function (Operable) they must be submerged.

A – incorrect – the instruments not submerged are not operable

B – incorrect – these instruments are reading containment air temperature and are not operable

D – incorrect – the instruments still submerged are operable – plausible is candidate is confused about which sets of instruments are reading accurately

Technical Reference(s): TS 1.0, TS 3.3.3.1 TS Bases 3.3 Rev 1, & Dwg 240-082 Rev M

Reference Attached: TS 1.0 p 1.0-5, TS 3.3.3.1 pp 20-23 TS Bases 3.3 p B 3.3-54 & Dwg 240-082


Learning Objective (As available): OT-COMBINED-D23-L.2







174

Perry NRC Exam 2015 QUESTION SRO 13 With the plant operating at rated power, per Tech Spec 3.6.1.3, Primary Containment Isolation Valves, opening the 42 inch inboard Containment Purge Isolation Damper is .

A. prohibited because it may not be capable of closing during a LOCA

B. only allowed if a dedicated operator is standing-by to isolate the damper

C. prohibited because it cannot be leak rate tested with the reactor at power

D. only allowed when operating Containment Vessel and Drywell Purge system in Intermittent Mode

175

Perry NRC Exam 2015

QUESTION SRO 13



K&A: Ability to explain and apply system limits and precautions.

High Off-site Release Rate / 9 Explanation: Answer A – SOI-M14 P&L 2.4 prohibits operating M14 in Refuel Mode in Modes 1, 2, & 3. The 42” INBD valve is only opened in Refuel Mode. Per TS Bases, the INBD valve may not close in the event of a LOCA. This may contribute to high off-site release rates. The OTBD 42” valve is opened during Intermittent Mode, which can be operated in Mode 1.

B – Incorrect – The 42” INBD valve is never allowed to be opened in Modes 1, 2, & 3. Plausible because other penetrations are allowed to be opened under Admin Controls with a Dedicated operator.

C – Incorrect - The 42” INBD cannot be leak tested in Modes 1, 2, & 3. However, this is not the TS Bases reason for not opening the valve.

D – Incorrect - The 42” INBD valve is opened when operating M14 (CVDWP) in Intermittent Mode.

Technical Reference(s): SOI-M14 Rev 23, TS 3.6.1.3, & TS Bases 3.6.1.3 Revs 1, 4, & 7

Reference Attached: SOI-M14 p 4, TS 3.6.1.3 p 3.6-9, & TS Bases 3.6.1.3 pp 17-20


Learning Objective (As available): OT-COMBINED-M14-H & -K.2







176

Perry NRC Exam 2015 QUESTION SRO 14 Which of the following conditions requires a Continuous Fire Watch Patrol? Reference Provided: PAP-1910 Fire Protection Program Body & Attachment #3

A. RCIC Pump Room Wet-Pipe Sprinkler will not deliver water

B. Heat Detection for Reactor Recirculation Pump B is out of service

C. Unit 2 Division 1 Cable Spreading Pre-Action Spray System will not deliver water

D. General area smoke detectors in Containment are functional but the detection system will not transmit an alarm to SAS.

177

Perry NRC Exam 2015

QUESTION SRO 14



K&A: Ability to determine and interpret the following as they apply to Plant Fire On Site: Requirements for establishing a fire watch

Plant Fire On Site / 8 Explanation: Answer A – This requires a continuous fire watch with backup fire suppression equipment.

B – Incorrect - Fire watch is not required, hourly remote monitoring is required.

C – Incorrect – The UNIT 2 requires hourly watch; UNIT 1 requires a continuous fire watch.

D – Incorrect – This requires an hourly fire watch.

Technical Reference(s): PAP-1910 Rev 31 Reference Attached: PAP-1910 pp 59-61, 65-66, 78-80, & 82-84

Proposed references to be provided to applicants during examination: PAP-1910 Body & Attachment #3

Learning Objective (As available): OT-3039- Admin-03-J

Question Source: Bank # Modified Bank # Perry 2009 # SRO-04 New





• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

178

Perry NRC Exam 2015 QUESTION SRO 15 Plant startup is in progress following a forced outage with the following conditions: • The Unit Supervisor is operating in IOI-3, Power Changes • The main generator was synchronized to the grid 5 minutes ago. • The MFP is feeding the RPV. • RFPT A is @ 1100 RPM • Main condenser pressure is 5.1”HgA and degrading slowly • Aux condenser pressure is 5.0” HgA and degrading slowly • Offgas flow is 33 SCFM and rising slowly • RPV level is 208 inches and rising slowly Based on this information, the Unit Supervisors highest priority is to direct the actions contained in (1) and (2) IOI-3, Power Changes.

(1) (2)

A. ONI-N62, Loss of Main Condenser Vacuum

continue in

B. ONI-N62, Loss of Main Condenser Vacuum

exit from

C. ONI-C71, Reactor Scram continue in

D. ONI-C71, Reactor Scram exit from

179

Perry NRC Exam 2015

QUESTION SRO 15



K&A: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Loss of Main Condenser Vac / 3 Explanation: Answer A – Per ONI-N62, Subsequent Actions direct tripping of the main turbine if condenser pressure degrades to > 5” HgA with generator load < 375 MWE. Since the generator was just synchronized to the grid, load would only be ~ 200 MWE. The operator would remain in IOI-3 as a turbine trip at this low power level would not cause a Rx scram.

B & D – Incorrect – Second part - Operator remains in IOI-3. Under these conditions, there is no scram. So no need to exit IOI-3.

C & D – Incorrect – First part – with Rx level rising and vacuum degrading with Offgas flow rising, the Operator may think scramming is appropriate. However, vacuum is already > 5” HgA. Therefore the trip of the main turbine has highest priority.

Technical Reference(s): ONI-N62 Rev 9 & IOI-3 Rev 52 Reference Attached: ONI-N62 p 5 & IOI-3 pp 3-4


Learning Objective (As available): OT-3043-03(DAY5) & OT-3035-10(LP)-A.1






Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

180

Perry NRC Exam 2015 QUESTION SRO 16 The plant was preparing to shift Reactor Recirculation pumps to fast speed with the following conditions: • RFPT ‘A’ was feeding the RPV in AUTO • RFPT ‘B’ was tripped due to high vibrations during warm-up • The Motor Feed Pump was tagged out due to a lube oil leak Then RFPT A tripped on thrust bearing wear. The following conditions are now present: • All control rods are inserted • RPV pressure is 940 psig and stable • RPV level 160 inches and trending down The procedure that contains specific actions to stop the level decrease is (1) . The Allowable Value for the level 3 scram setpoint is chosen to (2) .

(1) (2)

A. OAI-1703, Hardcards ensure that for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS will not be required

B. OAI-1703, Hardcards allow time for the low pressure core flooding systems to activate and provide adequate cooling

C. ONI-C34, Feedwater Flow Malfunction ensure that for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS will not be required

D. ONI-C34, Feedwater Flow Malfunction allow time for the low pressure core flooding systems to activate and provide adequate cooling

181

Perry NRC Exam 2015

QUESTION SRO 16



K&A: Ability to determine and/or interpret the following as they apply to Low Reactor Water Level: Reactor water level

Low Reactor Water Level / 2 Explanation: Answer A – OAI-1703, Hardcards contains the Reactor Scram Hardcard, which gives direction to ’stabilize’ RPV level using Feedwater, RCIC, or HPCS. Tech Spec 3.3.1.1 Bases for L3 scram, states that the Allowable Value is selected to ensure that for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS will not be required.

B & D – Incorrect – (2nd part) This is the Allowable Value Bases for the L1 ECCS Initiation signal.

C & D – Incorrect – (1st part) Although ONI C34 contains general direction to maintain RPV water level 192-200 inches, since all feedwater pumps are tripped or tagged out, this procedure would not be appropriate.

Technical Reference(s): OAI-1703 Rev 18, ONI-C34 Rev 9, TS 3.3.1.1 Bases Rev 3, & TS 3.3.5.1 Bases Rev 9

Reference Attached: OAI-1703 p 35, ONI-C34 p 6, TS 3.3.1.1 Bases p 3.3-12, & TS 3.3.5.1 Bases p 3.3-96


Learning Objective (As available): OT-COMBINED-B21(INST)-K.2 & OT-3037-07-G




10 CFR Part 55 Content: 55.41 55.43 b(2) & b(5)

Comments: SRO Justification - Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] • Knowledge of TS bases that are required to analyze TS required actions and terminology. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)] • Knowledge of when to implement attachments and appendices, including how to coordinate these items

with procedure steps.

182

Perry NRC Exam 2015 QUESTION SRO 17 The plant is operating at rated power, when a partial loss of Drywell Cooling occurs: As a result: • Drywell Average Air Temperature rises and stabilizes at 144.6°F • Drywell-to-Primary Containment d/p rises and stabilizes at +2.1 psid Which of the following describes the required action and the potential consequence of not taking that action? The required action is (1) . The consequence of not taking the action is (2) .

1 2

A restore the Drywell-to-Primary Containment d/p to within its Tech Spec limits

weir wall overflow, should an inadvertent upper pool dump occur

B restore the Drywell-to-Primary Containment d/p to within its Tech Spec limits

direct communication of the blowdown energy contained in the drywell airspace, to the suppression pool inventory, should a LOCA occur.

C restore the Drywell Average Air Temperature to within its Tech Spec Limits

weir wall overflow, should an inadvertent upper pool dump occur

D restore the Drywell Average Air Temperature to within its Tech Spec Limits

direct communication of the blowdown energy contained in the drywell airspace, to the suppression pool inventory, should a LOCA occur.

183

Perry NRC Exam 2015

QUESTION SRO 17



K&A: Ability to determine and/or interpret the following as they apply to High Drywell Pressure: Drywell pressure.

High Drywell Pressure / 5 Explanation: Answer B – With the plant in Mode 1, LCO 3.6.5.4 requires Drywell to Primary Containment Differential Pressure restored ≤ 2.0 psid. This LCO is based on preventing Horizontal Vent clearing at normal Suppression Pool Water Level. This discussion means that too high a drywell-to-CNMT ∆P can cause the vents to be already uncovered ('cleared') at the onset of a DBA LOCA (as a result of the downward force on the annulus water level). If a LOCA, then, were to occur, the RPV blowdown energy would communicate directly into the suppression pool inventory.

A – Incorrect – Second part - Plausible for negative differential pressure.

C & D – Incorrect – First part - Plausible if Drywell Temperature exceeded 145°F.

D – Incorrect – Second part - Plausible for negative differential pressure.

Technical Reference(s): LCO 3.6.5.4 Amendment 69, Technical Specification Bases B3.6.5.4 Rev 1

Reference Attached: TS 3.9.9 & TS 3.9.9 Bases pp B 3.9-30 & 31



Question Source: Bank # Clinton 2005 # SRO-81 Modified Bank # New





Knowledge of TS bases that are required to analyze TS required actions and terminology.

184

Perry NRC Exam 2015 QUESTION SRO 18 The plant is operating at rated power. • Containment pressure is 0.1 psig • Containment temperature is 74°F The NLO Rounds Taker reports that MCC disconnect to the Standby Liquid Control Operating Heater, C41-D002 was inadvertently manipulated to the OFF position. Based on this information, both SLC A and SLC B subsystems are (1) . The procedure used to restore the MCC disconnect to the ON position is (2) .

(1) (2)

A. INOPERABLE ARI for SLC A/B OUT OF SERVICE

B. INOPERABLE ELI-R24, 480 VOLT MCC

C. OPERABLE ARI for SLC A/B OUT OF SERVICE

D. OPERABLE ELI-R24, 480 VOLT MCC

185

Perry NRC Exam 2015

QUESTION SRO 18



K&A: Ability to (a) predict the impacts of the following on the Standby Liquid Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of SBLC tank heaters

SLC Explanation: Answer D – Both SLC subsystems remain operable with the heaters off as long as containment temperature remains ≥70°F (requires knowledge of surveillance requirements). The lineup ELI-R24 is the correct procedure for restoring the MCC for the heaters to the ON position.

A & B – Incorrect – First part – SLC remains operable since the given containment temperature is 74°F. SR 3.1.7.2 only looks at tank temperature.

A & C – Incorrect – Second part – The ARI is plausible since loss of power to other SLC components are annunciated by alarms. The ARI directs the operator to check MCC disconnects for loss of power to pumps and valves.

Technical Reference(s): TS 3.1.7, SOI-C41 Rev 19, ELI-R24 Rev 31, PAP-0205 Rev 20, ARI-H13-P601-19 Rev 17, & ARI-H13-P601-18 Rev 15

Reference Attached: TS 3.1.7 pp 3.1-20 & 21, SOI-C41 p 5, ELI-R24 p 47, PAP-0205 pp 4, 15-16, ARI-H13-P601-19 p 64, & ARI-H13-P601-18 p 14


Learning Objective (As available): OT-COMBINED-C41-K.1






• Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.

186

Perry NRC Exam 2015 QUESTION SRO 19 The plant was operating at rated power when a TLAC occurred. The following conditions now exist: • Reactor pressure is being maintained by manually cycling SRVs • Reactor level is 3” and slowly lowering • All control rods are fully inserted • RCIC initiated and tripped • The operators are in the process of lining up Fast Fire Water • No other injection systems are available Considering only Reactor level, complete the following statement. Based on the current status of injection systems, if Reactor level continues to lower, Emergency Depressurization ____. Reference provided: EOP-1 Chart – (Partial-modified)

A. must be performed before Reactor level reaches –25”

B. must not be performed until Reactor level reaches –42.5”

C. may be performed anytime while Reactor level is between 0” and –25”

D. may be performed anytime while Reactor level is between 0” and –42.5”

187

Perry NRC Exam 2015

QUESTION SRO 19



K&A: Knowledge of EOP mitigation strategies

RCIC Explanation: Answer B – Operating in EOP-1, (Step ALC-7) with RCIC tripped and no other injection systems or injection sub-systems available, the SRO must transition to EOP-4-3, Steam Cooling. While in EOP-4-3, if no injection sources are made available, ED is delayed until level lowers to -42.5”

A – Incorrect – This would be correct if any injection systems/sub-systems were available.

C – Incorrect – ED between 0 and -25” would be correct if any injection subsystem were lined-up for injection.

D – Incorrect – ED between 0 and -42.5” would be correct if any injection source became available and level could not be restored and maintained > -25”

Technical Reference(s): EOP-1 chart Rev E, ONI-R10 Rev 11, EOP-1 Bases Rev 4

Reference Attached: EOP-1 chart (partial), ONI-R10 p 3, EOP-1 Bases p 3, 54-55, 86, 90-91

Proposed references to be provided to applicants during examination: EOP-1 Chart – (Partial- modified - eliminate graphs. Give only Level and Pressure legs)

Learning Objective (As available): OT-3402-02 F




10 CFR Part 55 Content: 55.41 55.43 b.5



188

Perry NRC Exam 2015 QUESTION SRO 20 The following conditions exist: • The plant is at 100% power • I&C is performing surveillance SVI-B21-T0072-C, MAIN STEAM LINE LOW PRESSURE

CHANNEL C CALIBRATION FOR 1B21-N076C • The Unit Supervisor signed for Channel INOPERABILITY at 10:17 May 2nd • All other MSL Low Pressure instruments are OPERABLE • At 12:17 on May 2nd, the technician reports to you that he is unable to adjust the 1B21-N076C

instrument within the allowable value. Based on this information, the isolation function for Main Steam Line Pressure Low (1) maintained. The Required Action(s) is to (2) . Reference Provided: Technical Specification 3.3.6.1 (partial – modified) and Plant Data Book Tab I - partial

(1) (2)

A. is place channel in trip by 10:17 on May 3rd

B. is place channel in trip by 12:17 on May 3rd

C. is not be in Mode 3 by 17:17 on May 2nd

D. is not be in Mode 3 by 19:17 on May 2nd

189

Perry NRC Exam 2015

QUESTION SRO 20



K&A: Ability to (a) predict the impacts of the following on the Primary Containment Isolation System/Nuclear Steam Supply Shut-Off ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal condition or operations: †Surveillance testing

PCIS/Nuclear Steam Supply Shutoff Explanation: Answer B – Per Tech Spec Bases, sufficient channels remain Operable such that a valid signal will isolate the MSL’s. Therefore, the isolation function is maintained. The SR’s are modified by a Note stating that entry into the Conditions and Required Actions may be delayed for up to 6 hrs. Therefore entry into the Spec is 24 hours from time of discovery rather than from the US signing channel inoperability.

A – incorrect – this is 24 hours from the US signing channel inoperability

C & D – incorrect –- Function is not lost – MODE 2 is a RA if function is lost.

Technical Reference(s): PDB-I005 Rev 9, TS 3.3.6.1 & TS Bases B 3.3.6.1 Revs 0, 3 & 4, SVI-B21-T0072-C Rev 5

Reference Attached: PDB-I005 p 17, 20, TS 3.3.6.1 pp 48, & 53-54 & TS Bases B 3.3.6.1 pp B 3.3-142, & -163-164

Proposed references to be provided to applicants during examination: Technical Specification 3.3.6.1 (partial – modified) and Plant Data Book Tab I - partial

Learning Objective (As available): OT-COMBINED-B21(NS4)-l


Question History: Previous NRC Exam: Perry 2010




• Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).


190

Perry NRC Exam 2015 QUESTION SRO 21 The plant is operating in ONI-S11, Hi/Low Voltage, for Degraded Grid condition and EOP-1A, Level/Power Control. The following conditions currently exist: • RPV level is 10” and lowering • HPCS Pump is tagged out for motor replacement • Div 1 DG started and tripped on over-speed • RHR B pump tripped on over current • Buses EH11 and EH12 are powered from the Preferred source • Bus EH13 is powered from the Alternate Preferred Source Predict the impact if the Unit 1 Startup Transformer locks-out and select the procedure that will maximize the available ECCS systems. If Unit 1 Startup Transformer locks-out, (1) . The procedure that will maximize available ECCS systems is (2) .

(1) (2)

A. Bus EH12 will be de-energized with Div 2 DG running

ONI-SPI A-6, Division 1 Diesel Emergency Operation

B. Bus EH12 will be de-energized with Div 2 DG running

ONI-SPI A-12, Supplying Bus EH11 Loads from Division 2 Diesel Generator

C. Bus EH12 will be automatically transferred to Div 2 DG

ONI-SPI A-6, Division 1 Diesel Emergency Operation

D. Bus EH12 will be automatically transferred to Div 2 DG

ONI-SPI A-12, Supplying Bus EH11 Loads from Division 2 Diesel Generator

191

Perry NRC Exam 2015

QUESTION SRO 21



K&A: Ability to (a) predict the impacts of the following on the A.C. Electrical Distribution; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accident

AC Electrical Distribution Explanation: Answer D – With RPV level <16.5”, the Div 1 & 2 DG have started, but are not tied onto the buses. When U1 S/U Xfmr locks-out, the Preferred source for Bus EH12 deenergized and the bus undervoltage actions commence; The Preferred source breaker opens and the Div 2 DG output breaker will close. Since the RHR B pump tripped on over current, Energizing the Div 1 ECCS pumps (RHR A & LPCS) would maximize ECCS system availability. Therefore ONI-SPI A-12 is correct.

A & B – Incorrect – When a bus undervoltage occurs, the Preferred Source breaker will open and the DG output breaker will close. Plausible since the DG starts but doesn’t close in on a LOCA signal.

A & C – Incorrect – ONI-SPI A-6 is only used when DC control power is lost. Plausible if the operator thinks this ONI will get Div 2 DG running.

Technical Reference(s): SDM-OTCOMBINEDR10 Rev 10, ONI-SPI-A12 Rev 3, ONI-SPI-A6 Rev 3

Reference Attached: SDM-OTCOMBINEDR10 p 34, ONI-SPI-A12 p 2, ONI-SPI-A6 p 2


Learning Objective (As available): OT-COMBINED-R10-N







192

Perry NRC Exam 2015 QUESTION SRO 22 An overcurrent condition is sensed on the output of the Division 1 ATWS UPS Inverter. This overload condition will cause the (1) to transfer the loads to the Bypass Source Per TS 3.8.7, the AC Electrical Power Distribution System (2) with Division 1 ATWS UPS System in this lineup,

(1) (2)

A. Static Transfer Switch becomes INOPERABLE

B. Alternate Source Selector Switch becomes INOPERABLE

C. Static Transfer Switch remains OPERABLE

D. Alternate Source Selector Switch remains OPERABLE

193

Perry NRC Exam 2015

QUESTION SRO 22



K&A: Ability to determine operability and/or availability of safety related equipment.

UPS (AC/DC) Explanation: Answer C – An overload condition will cause the Static Transfer Switch to transfer to the Bypass Source. Per TS 3.8.7 Bases, the system remains operable on the Alternate Source.

A & B – Incorrect – Second part – The system remains operable on the bypass source. This info is located in the TS BASES.

B & D – Incorrect – First part – The loads remain energized through the static transfer switch. However, the Alt Source Selector Switch is plausible as the TSC UPS system has this device to transfer loads.

Technical Reference(s): ARI-H13-P680-06 Rev 9, SDM R14_15 Rev 1, TS 3.8.7, TS 3.8.7 Bases Rev 1

Reference Attached: ARI-H13-P680-06 pp 11-12, SDM R14_15 pp 4-6, 24, 28-29, TS 3.8.7 pp 3.8-36 & 37, TS 3.8.7 Bases p B 3.8-71


Learning Objective (As available): OT-COMBINED-R14_15-F.3 &I.1







194

Perry NRC Exam 2015 QUESTION SRO 23 The plant was operating at rated power with RHR A tagged out due to a motor ground. While responding to a plant transient, 1E12-F024B, RHR Test Valve To Suppression Pool stuck closed and blew fuses. Which of the following set of conditions would result in a Site Area Emergency as the highest event classification? Reference Provided: EPI-A1 & EOP-SPI Supplement (partial)

A. Rx Power is 3% RPV level at 5 inches RPV pressure at 825 psig Suppression pool level at 16.4 feet Suppression Pool temperature at 109°F

B. Rx Power is 4% RPV level at 25 inches RPV pressure at 925 psig Suppression pool level at 18.4 feet Suppression Pool temperature at 119°F

C. Rx Power is 5% RPV level at 25 inches RPV pressure at 925 psig Suppression pool level at 15.9 feet Suppression Pool temperature at 106°F

D. Rx Power is 4% RPV level at 25 inches RPV pressure at 925 psig Suppression pool level at 18.6 feet Suppression Pool temperature at 118°F

195

Perry NRC Exam 2015

QUESTION SRO 23



K&A: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

RHR/LPCI: Torus/Pool Cooling Mode Explanation: Answer D – Emergency plan declarations require notification to the NRC within 1 hour. With Rx not shutdown and Suppression pool temperature >110°F and not in UNSAFE region of HCL curve; conditions require a declaration of CS-1.

A – Incorrect – Conditions do not meet criteria for SAE – meets Alert.

B – Incorrect – Conditions meet criteria for a General Emergency.

C – Incorrect – Conditions do not meet criteria for SAE – meets General Emergency

Technical Reference(s): EPI-A1 Rev 26 & EOP-SPI Supplement Rev 4

Reference Attached: EPI-A1 pp 24 & 26 and EOP-SPI Supplement p 8

Proposed references to be provided to applicants during examination: EPI-A1 & EOP-SPI Supplement (partial)








196

Perry NRC Exam 2015 QUESTION SRO 24 The plant is operating at rated power with SVI-E51-T2001, RCIC Pump and Valve Operability Test in progress. The following alarms are received: • SUPR POOL LEVEL A HI/LO • SUPR POOL LEVEL B HI/LO • SUPR POOL TEMP A HIGH • SUPR POOL TEMP B HIGH Validated Suppression Pool level per SPDS is 18.55 feet. Validated Suppression Pool temperature per SPDS is 98°F. Based on the above conditions, the Unit Supervisor will direct Lowering Suppression Pool level per ARI for SUPR POOL LEVEL HI/LO and (1) And Entry into EOP-2 (2) required.

1 2

A. Lowering Suppression Pool temperature per ARI for SUPR POOL TEMP HIGH

is

B. Placing the Mode Switch in SHUTDOWN per Tech Spec Required Actions

is

C. Lowering Suppression Pool temperature per ARI for SUPR POOL TEMP HIGH

is not

D. Placing the Mode Switch in SHUTDOWN per Tech Spec Required Actions

is not

197

Perry NRC Exam 2015

QUESTION SRO 24



K&A: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Primary Containment System and Auxiliaries Explanation: Answer C – Since performance of SVI-E51-T2001 is a planned evolution, entry into EOP-2 is not required. The appropriate action is to return SP parameters to normal.

A – Incorrect – Entry into EOP-2 is not required since this is a preplanned evolution.

B – Incorrect – Placing Mode Switch to Shutdown would be required if SP temp rose to 110°F. Entry into EOP-2 is not required since this is a preplanned evolution.

D – Incorrect - Placing Mode Switch to Shutdown would be required if SP temp rose to 110°F.

Technical Reference(s): EOP-2 Bases Rev 2, ARI-H13-P601-17 Rev 13, SVI-E51-T2001 Rev 37 TS 3.6.2.1

Reference Attached: EOP-2 Bases p 2, ARI-H13-P601-17 pp 58-59, 31-32, SVI-E51-T2001 pp 4-5 TS 3.6.2.1 pp 3.6-36-39


Learning Objective (As available): OT-3402-05-C & 06-C






• Knowledge of administrative implementation, and/or coordination emergency procedures.

198

Perry NRC Exam 2015 QUESTION SRO 25 The plant was operating at 25% rated power following a refuel outage. Annunciator H13-P680-15A-D4, MSOP QUILL SHAFT FAIL alarmed. The reactor operator reports the following: • Shaft Oil Pump pressure on N34-R120 lowered to 165 psig then recovered to normal • Bearing Header oil pressure on 1N34-R121 lowered to 14 psig then recovered to normal The oil pump that should have started under these conditions is the (1) . If the oil pump did not auto start, the procedure that contains the steps to mitigate this casualty is (2) .

(1) (2)

A. Turning Gear Oil Pump ARI-H13-P680-15A-D4, MSOP QUILL SHAFT FAIL

B. Turning Gear Oil Pump ONI-N32, Turbine and/or Generator Trip

C. Emergency Bearing Oil Pump ARI-H13-P680-15A-D4, MSOP QUILL SHAFT FAIL

D. Emergency Bearing Oil Pump ONI- N32, Turbine and/or Generator Trip

199

Perry NRC Exam 2015

QUESTION SRO 25



K&A: Ability to (a) predict the impacts of the following on the Main Turbine Generator And Auxiliary Systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of lube oil

Main Turbine Gen. / Aux Explanation: Answer A – When the Main Shaft Oil pump discharge pressure lowers to 190 psig, the TGOP should auto start. If the bearing header oil pressure lowers to < 12 psig the turbine will trip. Therefore the correct procedure to mitigate this causality is ARI-H13- P680-15A-D4.

B – Incorrect – The turbine would not have tripped yet. The ONI-N32 would not be appropriate.

C – Incorrect – The EBOP will start if the TGOP discharge pressure decreases to < 10 psig.

D – Incorrect – The EBOP will start if the TGOP discharge pressure decreases to < 10 psig. And, the turbine would not have tripped yet. The ONI-N32 would not be appropriate.

Technical Reference(s): ARI-H13-P680-15A-D4 Rev 6 and ONI-N32 Rev 11

Reference Attached: ARI-H13-P680-15A-D4 pp 13-14 and ONI-N32 pp 3, 4, 7, & 12.


Learning Objective (As available): OT-Combined-N34-M






• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

200

perry nrc exam 2015 · perry nrc exam 2015 . question ro 1 . the plant is operating at 72% reactor...

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