ML110310434

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Operator Licensing Retake Exam - Draft Written Exam (Folder 2)
ML110310434
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 01/13/2011
From:
Public Service Enterprise Group
To: Todd Fish
Operations Branch I
Hansell S
Shared Package
ML103200290 List:
References
TAC U01840
Download: ML110310434 (51)
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Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295037 EA2.03 Importance Rating 4.4 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWI\I : SBLC tank level Proposed Question: SRO Question # 1 While operating at rated power, an MSIV (NS4) isolation occurred and the reactor failed to scram; all rods remain at their pre-trip conditions.

The operators are in the process of deliberately lowering RPV water level.

Given:

  • RX Power 10%.
  • RPV Pressure 900 psig and stable being controlled with SRVs.
  • RPV Level (-49) inches and slowly lowering.
  • Suppression Pool Level 79 inches and rising.
  • Suppression Pool Temp 160°F and slowly rising @ < 1°F/10 minutes.
  • Drywell Pressure 2.7 psig and lowering.
  • SLC is Injecting.
  • SLC Tank Level is 1400 gallons and lowering.

What actions are required?

A. Enter EOP-202 and Emergency Depressurize.

B. Stop lowering RPV level lAW EOP-101A and begin a cool down EOP-101.

C. Continue lowering RPV level until (-50") is reached. Then begin a cool down lAW EOP-101A.

D. Stop lowering RPV level when reactor power < 4% and continue with SLC injection lAW EOP-1 01 A. A cool down is NOT permitted at this time.

Proposed Answer: D

Explanation (Optional):

A. Incorrect - No criteria is met for ED (the HCTL limit at this RPV Pressure is about 190°F suppression pool temperature)

B. Incorrect - a cool down cannot begin until Cold shutdown boron weight has been injected or all rods (except one) are full in.

C. Incorrect - with reactor power below 4%, EOPs direct that lowering of level be stopped D. Correct - lAW EOP-101A step LP-14 the lowering of level must be terminated when reactor power is < 4%. With the reactor not shutdown, SLC injection must continue.

Technical Reference( s): EOP-101A (Attach if not previously provided)

EOP-101A, RC/Q6 LP7 thru RC/Q18 Proposed References to be provided to applicants during examination:

LP15, EOP-102 SPIT7 thru SPIT10 E0101AE006, Given any step of the procedure, explain the reason for performance of that step and/or evaluate Learning Objective: (As available) the expected system response to control manipulations prescribed by that step.

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KfA# 295038 EA2.03 Importance Rating 4.3

-=~-=-=::---

Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:

Radiation levels: Plant-Specific Proposed Question: SRO Question # 2 Given:

  • The Unit was operating at 80% power.
  • The MSIV's failed to completely isolate the leak.
  • The reactor automatically scrammed.
  • Reactor level dropped to -45 inches, and level was restored to +30 inches with HPCI & RCIC.
  • Personnel reported a large steam plume in the turbine building upstream of the turbine stop valves
  • The reactor has been depressurizing through the steam line rupture for the last 75 minutes.

Current conditions (75 minutes after the event began):

  • Reactor pressure is 200 psig and dropping slowly.
  • The steam line leak is still in progress.
  • Reactor level is +30 inches with condensate injecting for level control.
  • HPCI & RCIC have been secured to slow the depressurization rate, and both systems were realigned for automatic operations.
  • Radiation Protection has just provided a Dose Assessment projection based on Plant effluent sample analysis:
  • 3.4E-01 mRem TEDE 4-day dose at the MEA
  • 1.6E-01 mRem Thyroid-CDE Which one of the following describes the HIGHEST event classification and the EOP(s) that were or are required to be entered to mitigate the event?

A General Emergency. EOP-101 & EOP-103/4.

B. Site Area Emergency. EOP-01 01 ONLY.

C. Alert. EOP-101 ONLY.

D. Unusual Event. EOP-101 & EOP-103/4.

Proposed Answer: B Explanation (Optional):

A. Incorrect - The threshold was not reached for GE classification, and EOP-104 entry was not required.

B. Correct: SAE declaration required due to a Main Steam Line break outside containment with a failure of the MSIV's to isolate leak with a downstream pathway to the environment. ECG Section 3 Barriers Table, 3.2.3.b 4 points and 3.3.4.a 2 points 6 point SAE. EOP-0101 was required to be entered on low reactor level; however, EOP-0104 entry was/is NOT required since the dose projection has only reached the Unusual Event level based on TEDE. EOP-0104 entry is required when there is a gaseous radioactive release at the ALERT threshold. The SAE declaration was based on the loss of the RCS and containment barriers NOT the radioactive release, so EOP-104 entry was/is not required.

C. Incorrect: SAE threshold was reached.

D. Incorrect: SAE threshold was reached & EOP-104 entry was not required. Only would get an Unusual Event from ECG 6.1.1.a Technical Reference(s): ECG, EOP-101, EOP-103/4 (Attach if not previously provided)

Barrier Table and Proposed References to be provided to applicants durin~1 examination:

Sect. 3 & 6 of ECG Learning Objective: ABCNT4E001 (As available)

Question Source: Bank # NRC 2007 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KIA # 295031 EA2.04 Importance Rating 4.8 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL:

Adequate core cooling Proposed Question: SRO Question # 3 Given:

  • A large break LOCA is in progress.
  • All EOP Table 1 Preferred Injection Systems sources are unavailable.
  • All EOP Table 2 Injection Systems sources are unavailable.
  • Reactor level is (-11 Off) and lowering at 10"/minute.
  • Reactor pressure is 900 psig and lowering at 40 psig/minute.
  • SLC pumps cannot be started.
  • Condensate Transfer is unavailable.
  • Operators have just began to perform OP-EO.ZZ-310, "Alternate Injection Using Fire Water",

and a pump will be aligned for injection in a minimum of 60 minutes.

Which of the following actions are required to maintain adequate core cooling?

Inhibit ADS when level drops to (-129") and _ _.

A. Immediately enter EOP-202, Emergency Depressurization, and open 5 ADS valves.

B. When RPV level drops to (-185"), enter EOP-202, Emergency Depressurization, and open 5 ADS valves.

C. When RPV level drops to (-200"), enter EOP-202, Emergency Depressurization and open 5 ADS valves.

D. Enter EOP-101 Steam Cooling leg, when "Alternate Injection Using Fire Water", (when injection takes place) fails to restore and maintain RPV level> (-200").

Proposed Answer: C

Explanation (Optional):

A Incorrect - with no EOP Table 1 and 2 injection sources available, EOP-101 action requires entry to the steam cooling leg and when level reaches -200 inches, then an ED is performed B. Incorrect - An emergency blowdown is allowed by EOP-1 01 before -185 inches is reached if sufficient injection sources are available to restore level above -129 inches or restore and maintain it above -185 inches following the blowdown. In the scenario given, there is zero injection available.

C. Correct - With zero injection, EOP-101 directs that the operators wait until the Minimum Zero Injection Water Level is reached (MZIRWL, - 200 inches). This allows steam cooling to be used for as long as possible before an emergency depressurization is performed. Once MZIRWL is reached, then adequate core cooling is not assured and an emergency depressurization is required even without an injection source available.

D. Incorrect - In the Steam Cooling leg, at -200 inches, an ED is required regardless if level is recovering, Technical Reference(s}: EOP-101, Steps AUC-19 and SC-3 (Attach if not previously provided)

EOP-101 ,ALC-2 thru Proposed References to be provided to applicants during examination:

ALC-21 E01 01 PE008, Given any step of the procedure, describe the reason for Learning Objective: performance of that step and/or expected (As available) system response to control manipulations prescribed by that step.

Question Source: Bank# WTS Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KJA# 295025 2.4.11 Importance Rating 4.2 Emergency Procedures I Plan: Knowledge of abnormal condition procedures. (High reactor pressure)

Proposed Question: SRO Question # 4 Given:

  • The plant is operating at 25% power.
  • OHA C5-C2 TCV FAST CLOSURE & MSV TRIP BYP is ILLUMINATED.

Then, the Main Turbine Generator trips:

  • OHA B3-E5 RPV PRESSURE HI is ILLUMINATED.
  • OHA C5-A2 & B2 for TCV FAST CLOSURE and MAIN STOP VALVE CLOSURE are ILLUMINATED.
  • OHAs C3-A2, C3-A3, C3-A4 & C3-A5 for REACTOR SCRAM TRIP LOGIC A1/A2/B1/B2 are EXTINGUISHED.

Which one of the following actions is required lAW with Technical Specifications and a plant abnormal procedure?

lAW TS, _(1)_ and lAW plant abnormal procedure, _.(2)_.

A. (1) reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) Reduce reactor pressure below the alarm setpoint immediately lAW AB.RPV-005.

B. (1) reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) Lock the Reactor Mode Switch in Shutdown immediately lAW AB.BOP-002.

C. (1) reduce the pressure to less than 1020 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Reduce reactor pressure below the alarm setpoint immediately lAW AB.RPV-005.

D. (1) reduce the pressure to less than 1020 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Lock the Reactor Mode Switch in Shutdown immediately lAW AB.BOP-002.

Proposed Answer: B Explanation (Optional):

A. Incorrect - Part (2) would be correct with reactor power <26%

B. Correct lAW with TS 3.4.6.2 and AB-BOP-0002 C. Incorrect Part (1) is incorrect because TS does not mention Cold Shutdown and the action is within 15 minutes. Part (2) would be correct with reactor power <26%

D. Incorrect - Part (1) is incorrect because TS does not mention Cold Shutdown and the action is within 15 minutes AB-BOP-0002 and AB.RPV-0005, Technical Reference(s}: (Attach if not previously provided)

TS 3.4.6.2 Proposed References to be provided to applicants durin~1 examination: none TECSPECPCE010 - Given specific plant operating conditions and a copy of HCGS Learning Objective: (As available)

TS, evaluate planUsystem operability and plant actions to be taken Question Source: Bank # NRC 2005 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 1 Group # 1 KIA # 295026 2.2.12 Importance Rating 4.1

~ ~-- =-~:---

Equipment Control: Knowledge of surveillance procedures. (Suppression Pool High Water Temp)

Proposed Question: SRO Question # 5 The plant is operating at rated power.

  • HPCI is scheduled to be placed in service for its quarterly surveillance test, OP-IS.BJ-0001, (HPCI MAIN AND BOOSTER PUMP SET - OP204 AND OP217 - INSERVICE TEST).

Which one of the following describes; (1) Technical Specification Suppression Chamber Average Temperature requirements during the HPCI test AND (2) The surveillance test requirement for placing Suppression Pool Cooling in service prior to OR during the test.

A. (1) With the suppression chamber average water temperature greater than 95°F, restore the average temperature to less than or equal to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Suppression Pool Cooling must be placed in service prior to placing HPCI in service.

B. (1) With the suppression chamber average water temperature greater than 95°F, restore the average temperature to less than or equal to 95°F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) Suppression Pool Cooling must be placed in service when the EOP-102 entry condition is met.

C. (1) With the suppression chamber average water temperature greater than 105°F, restore the average temperature to less than or equal to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(2) Suppression Pool Cooling must be placed in service prior to placing HPCI in service.

D. (1) With the suppression chamber average water temperature greater than 105°F, restore the average temperature to less than or equal to 95°F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(2) Suppression Pool Cooling must be placed in service when the EOP-102 entry condition is met.

Proposed Answer: C Explanation (Optional):

A. Incorrect - lAW TS, the limit is 105° F. during testing which adds heat to the suppression chamber (HPCI)

B. Incorrect - lAW TS, the limit is 1050 F. during testing which adds heat to the suppression chamber {HPCI}, Suppression Pool Cooling must be in service prior to running HPCI C. Correct - lAW TS 3.6.2.1.a.2.a) and HPCI 1ST BJ-0001, step 5.6 "ENSURE RHR is in Suppression Pool Cooling lAW HC.OP-SO.BC-0001 (Q), RHR System Operation, OR, HC.OP IS.BC-0001(0002)(Q), A(B) Residual Heat Removal Pump -A(B)P202 -In-service Test."

D. Incorrect - The TS requirement is to restore the average temperature to less than or equal to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TS 3.6.2.1.a.2.a)

Technical Reference(s}: (Attach if not previously provided)

HPCI 1ST-BJ-0001 Proposed References to be provided to applicants during examination: none PRICONE009 Given a Scenario of applicable operating conditions and access to Technical Specifications:

Learning Objective: a. Select those sections that are (As available) applicable to the Primary Containment Structure lAW HCGS Technical Specifications.

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group #

KJA# 295006 2.2.39 Importance Rating 4.5

~--~---- -----_.-------.

Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems. (Scram)

Proposed Question: SRO Question # 6 Hope Creek is at 20% power when the plant scrams.

OP-AS.ZZ-OOOO, Reactor Scram, has been entered:

Plant conditions are now:

  • RPV Level is +33" and stable
  • RPV Pressure is 1000 psig and stable
  • Mode Switch is Locked in Shutdown
  • NI surveillances were complete as of 5 days ago As the Control Room Supervisor, you have reached step "S-8":

"IF Conditions permit THEN RESET the Scram AND INSERT a Half Scram (if required)

Which of the following conditions would REQUIRE ordering the RO to INSERT a Half Scram within one (1) hour?

(1) APRM channels "An and "C" INOP (2) IRM channels "E" and !IF" INOP (3) 1 Reactor Vessel Steam Dome Pressure High Transmitter INOP A. (1) ONLY S. (2) ONLY C. (1) AND (2) ONLY D. (1), (2) AND (3)

Proposed Answer: A Explanation (Optional):

A. Correct - Per TS 3.3.1.a. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least 1 trip system in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1. For the APRM's in OPCON 3 - Minimum OPERABLE Channels per Trip System is 2, therefore if 2 APRM were INOP you would only have 1 in that Trip System OPERABLE and would need to insert a Half-scram B. Incorrect - Per TS 3.3.1-1 in OPCON 3 you are only required to have 2 IRM's OPERABLE per trip system, since you have 3 available having 1 INOP still leaves 2 that are OPERABLE and so you would NOT have to insert a Half-Scram C. Incorrect - see B above D. Incorrect - see B above, also Reactor Steam Dome Pressure High transmitter is only required in OPCON 1 or 2, since you are in OPCON 3 this would be N/A Technical Reference(s): TS 3.3.1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: none APRMOOE008, Given a scenario of applicable conditions and access to Technical Specifications:

Learning Objective: (As available)

a. Select those sections which are applicable to the APRMS lAW Technical Specifications.

Question Source: Bank # WTS 10509 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 KJA# 295018 2.4,30 Importance Rating 4,1 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, Proposed Question: SRO Question # 7 Given:

  • The reactor has been shutdown for two (2) days following power operation,

Then, a grassing event causes a loss of Service Water flow to the SACS HXs.

  • HC,OP-AB,COOL-0001, Station Service Water has just been entered with NO positive results.
  • HC.OP-AB.RPV-0009, Shutdown Cooling has just been entered and any remaining actions can NOT be completed for sixty (60) minutes.
  • RWCU bottom head drain temperature is now 180°F and rising at 5° F/minute.

Based on given information, which one of the following is the highest reporting requirement?

A. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report B. 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report C. Alert D. Site Area Emergency Proposed Answer: C Explanation (Optional):

A. Incorrect - Based on 11.1.1.c but does not apply B. Incorrect - Based on 11.2.2.b but not the highest reporting requirement C. Correct - HC ECG 8.1.2 D. Incorrect. Based on a.i.3.b. but not in Op Con 1,2, or 3 Technical Reference(s): ECG (Attach if not previously provided)

Proposed References to be provided to applicants during examination: ECG Sect. a & 11 ABCOL 1E007, Given plant conditions and plant procedures, determine required Learning Objective: actions of the retainment override(s) and (As available) subsequent operator actions in accordance with Station Senfice Water.

Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 1 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295036 EA2.03 Importance Rating 3.8

=-:-:-:-:::-:-:-

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SLIMP/AREA WATER LEVEL: Cause of the high water level Proposed Question: SRO Question #8 The plant was operating at 30% power when a seismic event occurred.

Given:

  • A LOP occurred and ALL Vital Buses are being powered with their respective EDGs.
  • HPCI is running and controlling RPV level at +30 inches.
  • RPV pressure is 900 psig and stable.
  • A break has occurred in the cooling water supply line to the HPCI and RCIC room coolers.
  • The HPCI room water level has been above its Max Safe Op Floor Level Limit for 45 minutes.
  • The RCIC room water level has been above its Max Safe Op Floor Level Limit for 45 minutes.

Which of the following describes the source of the leak and actions required?

A. SACS is the source of the leak. The criteria has been met for performing an Emergency Depressurization lAW EOP-103/4, Reactor Building & Rad Release Control.

B. Chilled Water is the source of the leak. The criteria has been met for performing an Emergency Depressurization lAW EOP-103/4, Reactor Building & Rad Release Control.

C. SACS is the source of the leak. An Emergency Depressurization is NOT required. A cool down can be commenced lAW EOP-101, RPV Control.

D. Chilled Water is the source of the leak. An Emergency Depressurization is NOT required. A cool down can be commenced lAW EOP-101, RPV Control.

Proposed Answer: C Explanation (Optional):

A. Incorrect - SACS is not a primary system [reactor coolant] and therefore lAW EOP-103/4, an ED is not required.

B. Incorrect - Chill water does not supply the HPCI & RCIC Room Coolers and is not a primary system [reactor coolant] and therefore lAW EOP-103/4, an ED is not required.

C. Correct - SACs supplies the HPCI & RCIC Room Coolers and is NOT a primary system [reactor coolant] therefore lAW EOP 103/4 (RB-15), a shutdown is required (RB-22), NOT an ED (RB 19). The cool down is directed in EOP-101 step RC/P-7 D. Incorrect - Chill water does not supply the HPCI & RCIC Room Coolers. The cool down is directed in EOP-101 step RC/P-7 EOPs 101, 103/4 Technical Reference(s): (Attach if not previously provided)

NOH04STACSOC-06, Pages 14, 15 EOP-103/4 steps RB14 thru RB 20 Learning Objective: NOH04STACSOC-O, OBJ. 2 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 KIA # 295014 2.1.9 Importance Rating 4.5 Ability to direct personnel activities inside the control room. (Inadvertent reactivity addition)

Proposed Question: SRO Question #9 Given:

  • The Reactor is shutdown with refueling activities in progress.
  • Mode Switch is in Refuel.
  • No significant changes in SRM counts were noted.

Then, fifteen (15) seconds later:

  • OHA C3-D1, SRM PERIOD alarms.
  • OHA C6-C1, SRM UPSCALE OR INOPERATIVE alarms.
  • No Radiation Monitors are in alarm at this time.

Which one of the following isfare the Abnormal procedure(s) entered based on the above alarms and what action(s) isfare required?

A. Enter AB.RPV-0001. Reactor Power ONLY and direct Chemistry to sample and analyze reactor water.

B. Enter AB.IC-0003, Reactor Protection System ONLY and direct resetting the Y2 Scram.

C. Enter AB.RPV-0001, Reactor Power AND AB.IC-0003, Reactor Protection System and ONLY direct resetting the }2 Scram.

D. Enter AB.lC-0003, Reactor Protection System AND AB.RPV-0001, Reactor Power and ONLY direct Chemistry to sample and analyze reactor water.

Proposed Answer: A Explanation (Optional):

A. Correct - Condition E of AB.RPV-0001 for a dropped rod directs sampling reactor water.

B. Incorrect - AB.IC-0003, is not required to be entered because RPS has not actuated. The SRM UpscalellNOP will not cause a half/full scram in this condition, as the shorting links are installed.

C. Incorrect - AB.lC-0003, is not required to be entered because RPS has not actuated. The SRM Upscale/INOP will not cause a half/full scram in this condition, as the shorting links are installed.

D. Incorrect - AB.lC-0003, is not required to be entered because RPS has not actuated. The SRM UpscalellNOP will not cause a half/full scram in this condition, as the shorting links are installed.

AB-RPV-0001 Technical Reference(s): (Attach if not previously provided)

AB-IC-0003 Proposed References to be provided to applicants during examination: none ABRPV1 E001 - Recognize abnormal conditions/alarms and/or procedural Learning Objective: (As available) requirements for implementing reactor power Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2,5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 295034 EA2.01 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Ventilation radiation levels Proposed Question: SRO Question #10 The Unit is in OPCON 1 with spent fuel moves in progress on the refueling floor. Ventilation systems are lined up in the normal lineup to support the plant conditions.

Then:

  • A spent fuel bundle is dropped from the full up position.
  • The bundle impacts other fuel bundles in the fuel pool.
  • Bubbles are observed rising from the dropped bundle.
  • Refuel Floor Exhaust radiation rises to 1.8 x 10-3 /lCi/cc.
  • Reactor Building Exhaust radiation is normal.

Which of the following describes additional AOP or EOP procedure entry requirements and secondary containment ventilation status?

Enter AB.CONT -0005, Irradiated Fuel Damage and __.

A. EOP-103/4, Reactor Building and Rad Release Control. The Reactor Building ventilation supply and exhaust fans are running, FRVS vent fans are NOT running, and FRVS recirculation fans are NOT running.

B. AB.CONT-0003, Reactor Building. The Reactor Building ventilation supply and exhaust fans are NOT running, ONE FRVS vent fan is running, and SIX FRVS recirculation fans are running.

C. EOP-103/4, Reactor Building and Rad Release Control. The Reactor Building ventilation supply and exhaust fans are NOT running, ONE FRVS vent fan is running, and FOUR FRVS recirculation fans are running.

D. AB.CONT-0003, Reactor Building. The Reactor Building ventilation supply and exhaust fans are NOT running, TWO FRVS vent fans are running, and SIX FRVS recirculation fans are running.

Proposed Answer: A Explanation (Optional):

A. Correct - The entry condition for EOP 103/4 has been met (1 x 10-3 /-lCi/cc). No other procedural entry conditions were met. The Refuel Floor Exhaust radiation level did not meet the auto initiation setpoint for swapping over to FRVS (2 x 10-3 /-lCi/cc).

B. Incorrect - No entry condition exists for AB.CONT -0003, Reactor Building. Also, The RBVS fans would be running, the FRVS would not be running C. Incorrect - The RBVS fans would be running, the FRVS would not be running.

D. Incorrect - No entry condition exists for AB.CONT-0003, Reactor Building. Also, The RBVS fans would be running, the FRVS would not be running.

NOH01 SECCONC-02 page 10 Technical Reference(s): (Attach if not previously provided)

EOP-103/4 Proposed References to be provided to applicants during examination: EOP-103/4 EOP1 03E002 - Given a set of plant Learning Objective: conditions, analyze and determine if entry (As available) conditions into HC.OP-EO.ZZ-103 exist.

NRC 2007 - slightly Question Source: Bank #

modified Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 239002 A2.04 Importance Rating 4.2 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations: ADS actuation Proposed Question: SRO Question # 11 The plant was operating at rated power when a LOCA occurred.

Given:

  • Reactor Pressure is 500 psig and lowering
  • RPV Level reached (-129") two (2) minutes ago
  • SLC is injecting
  • Preferred Injection Systems are in service and injecting
  • Drywell Pressure is 13 psig and rising slowly
  • Drywell Temperature is 240°F and rising slowly
  • Drywell Spray is in service
  • ADS has NOT been inhibited Which one of the following describes the status of the ADS SRVs and the EOP being implemented to mitigate the given plant conditions?

A. The ADS SRVs are OPEN. ADS should have previously been inhibited lAW EOP-101, RPV Control.

B. The ADS SRVs are OPEN. ADS should have previously been inhibited lAW EOP-101A, ATWS RPV Control.

C. The ADS SRVs are CLOSED. ADS must be inhibited lAW EOP-101, RPV Control.

D. The ADS SRVs are CLOSED. ADS must be inhibited lAW EOP-101A, ATWS-RPV Control, Proposed Answer: B

Explanation (Optional):

A. Incorrect - EOP-101A is being implemented due to the ATWS, NOT EOP-101. Step LP-5 directs inhibiting ADS, which will keep the valves closed, even after the 105 second timer has timed out.

B. Correct The ADS initiating logic has been met and the ADS valves are OPEN.

EOP-101A is being implemented due to the ATWS. Step LP-5 directs inhibiting ADS, which will keep the valves closed, even after the 105 second timer has timed out.

C. Incorrect - The ADS initiating logic has been met and the ADS valves are OPEN. EOP-101A is being implemented due to the ATWS, NOT EOP-*101.

D. Incorrect - The ADS initiating logic has been met and the ADS valves are OPEN.

EOP-101A Technical Reference(s}: NOH01ADSSYSC-04, Page 9 and 10 (Attach if not previously provided)

Proposed References to be provided to applicants durin!J examination: none Learning Objective: NOH01ADSSYSC-04, Obj. R2 (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 215005 A2.08 Importance Rating 3.4

- - - _......- =-=:-=-=--=-:::-::-:-

Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Faulty or erratic operation of detectors/systems.

KA Match Justification. The OPRMs get their inputs from APRM/LPRM inputs Proposed Question: SRO Question # 12 The plant is operating at rated power.

When:

  • A partial loss of core flow occurs.
  • The plant is currently operating at 53% power with a rated core flow of 40%.

lAW HC Technical Specifications, the minimum number of operable OPRM channels is _ _(1 AND_{2)_.

A. (1) 2 (2) if an OPRM alarm annunciates on an operable channel an orderly shutdown is required lAW 10.ZZ-0004, Shutdown From Rated Power B.

(1) 2 (2) if you do NOT have the minimum number of operable channels, a reactor scram is required immediately lAW AB.RPV-0003, Recirculation System / Reactor Power Oscillations C.

(1) 4 (2) if an OPRM alarm annunciates on an operable channel an orderly shutdown is required lAW 10.ZZ-0004, Shutdown From Rated Power D.

(1) 4 (2) if you do NOT have the minimum number of operable channels, a reactor scram is required immediately lAW AB.RPV-0003, Recirculation System 1 Reactor Power Oscillations

Proposed Answer: 0 Explanation (Optional):

A. Incorrect - TS 3.3.11 requires 4 operable channels. HC.OP-AB.RPV-0003 requires the Mode Switch placed in SO if in Region 1 and OPRMs inoperable.

B. Incorrect - TS 3.3.11 requires 4 operable channels.

C. Incorrect - HC.OP-AB.RPV-0003 requires the Mode Switch placed in SO if in Region 1 and OPRMs inoperable.

O. Correct - TS 3.3.11 requires 4 operable channels. HC.OP-AB.RPV-0003 requires the Mode Switch placed in SO if in Region 1 and OPRMs inoperable lAW TS.

AB-RPV-0003 Technical Reference(s): TS 3.3.11 (Attach if not previously provided)

NOH040PRMOOC-01 page 12 PF Map OPRMS Proposed References to be provided to applicants during examination: Operable and Inoperable.

ABRPV3E007, Given plant conditions and plant procedures, determine required actions of the retainment override(s) and Learning Objective: (As available) subsequent operator actions in accordance with the Recirculation System/Power Oscillations.

Question Source: Bank # WTS 10556 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 2 Group #

KIA # 264000 2.4.8 Importance Rating 4.5 Emergency Procedures 1 Plan: Knowledge of how abnormal operating nrr.I'ortl are used in conjunction with EOPs. (EDGs)

Proposed Question: SRO Question # 13 The plant was operating at 30% power. A monthly surveillance test on the "C" EDG was in progress. The output breaker is closed to synchronize the EDG to the grid and within a second, a Station Blackout occurred (loss of all onsite and offsite power). The reactor scrammed.

Current plant conditions are:

  • Drywell temperature 339°F rising slowly
  • Drywell pressure 6.7 psig rising slowly
  • Torus level 100" and steady
  • RPV pressure 320 psig dropping slowly
  • Reactor power all rods fully inserted
  • Reactor level (-70") lowering slowly
  • RCIC CIT and disassembled for emergent maintenance work
  • HPCI tripped on over-speed and will NOT restart
  • "A"EDG tripped - maintenance investigating
  • "B"EDG tripped on Bus differential over-current - cause unknown
  • "C"EDG running unloaded output breaker failed open on anti pump circuitry
  • "D" EDG start failure, low starting air pressure -20 psig What is required to either restore power to a vital bus or control containment parameters?

A. lAW HC.OP-SO.KJ-0001, Emergency Diesel Generator Operation, direct the NEO to reset the Bus Differential Over-current on the "B" EDG and restart the "B" EDG. Then, start a plant cool down at <1 OO°F/hour lAW EOP-101.

B. lAW HC.OP-AB.ZZ-0135, Station Blackout, Loss of Offsite Power, Diesel Generator malfunction, direct the NEO to reset the Bus Differential Over-current on the "B" EDG and restart the "B" EDG. Then, start a plant cool down at <100°F/hour lAW

EOP-101.

C. lAW HC.OP-EO.ZZ-0202, Emergency Depressurization is required based on high Drywell temperature. Then, lAW HC.OP-AB.ZZ-0135, Station Blackout, Loss of Offsite Power, Diesel Generator malfunction, direct the RO to depress the TRIP pushbutton on the "C" EDG output breaker and verify output breaker closes.

D. lAW HC.OP-EO.zZ-0202, Emergency Depressurization is required before Reactor Water Level decreases to -129". Then, lAW HC.OP-AB.zZ-0135, Station Blackout, Loss of Offsite Power, Diesel Generator malfunction, direct the RO to depress the TRIP pushbutton on the "C" EDG output breaker and verify output breaker closes.

Proposed Answer: C Explanation (Optional):

A. Incorrect - bus differential current should NOT be reset without electrical maintenance determining and correcting the cause. An ED must be performed first.

B. Incorrect - This action is not directed from the abnormal. An ED must be performed first.

C. Correct - With Drywell temperature approaching 340 F, an ED is performed lAW EOP -102 step 0

Dwrr-8. lAW HC.OP.AB.ZZ-0135, Station Blackout p. 22 step 5.16 - The Anti-pump circuitry on the DIG output breaker could cause the output breaker to fail open. To load the DIG under this condition the operator must depress the TRIP push-button (even though the breaker is already tripped) to reset the logic. When the TRIP push-button is released, then the breaker will close and the DIG will load.

D. Incorrect - Emergency Depressurization procedure should NOT be entered until level is less than

-129" but before level decreases to -185" EOP-102 Technical Reference(s): HC.OP.AB.ZZ-0135, Station Blackout (Attach if not previously provided)

p. 22 step 5.16 EOP-101 RC/L 1 thru Proposed References to be provided to applicants during examination: ALC1and EOP-102 DWrr1 thru DWrr9 E01 01 PE008 Given any step in the procedure, describe the reason for Learning Objective: performance of that step and/or expected (As available) system response to control manipulations prescribed by that step.

Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)

New

Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KiA # 263000 2.2.40 Importance Rating 4.7 Ability to apply Technical Specifications for a system. (DC electrical)

Proposed Question: SRO Question # 14 The plant is at rated power:

Given:

  • 4.16Kv breaker 52-40101 ,Bus 10A401 Alt Feed 1BX501 , breaker is tagged in DISCONNECT for Maintenance work
  • All other systems are in a normal lineup Then:

Station Service Transformer BX501 suffers a phase to ground fault.

Assuming NO operator actions, what is the resulting impact to the operability of 125 VDC Battery 1AD411 and why?

Battery 1AD411 is_ _.

A. Inoperable; because 1OA401, 4.16Kv bus is energized by the running "A" EDG.

B. Inoperable; because both battery chargers are de-energized.

C. Operable; because 1OA401, 4.16Kv bus is energized by normal power.

D. Operable; because both battery chargers are energized by the running "A" EDG.

Proposed Answer: C Explanation (Optional):

A. Incorrect - 10A401 would still be energized. Normally closed breaker 52-40108 would remain

closed and powered from offsite power.

B. Incorrect. Both chargers would be energized.

C. C. Correct Normally closed breaker 52-40108 would remain closed and powered from offsite power. The battery charger needs to be energized from the 1E bus to be operable.

D. Incorrect - A EDG would not be running on a loss of only one infeed.

TS 3.8.3.1 Technical Reference(s): DC electrical distribution lesson plan (Attach if not previously provided) page 24 Proposed References to be provided to applicants during examination: none TECSPECPCE010 Given specific plant operating conditions and a copy of HCGS Learning Objective: (As available)

TS, evaluate plant/system operability and plant actions to be taken Question Source: Bank # Audit 2007 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 211000 A2.05 Importance Rating 3.4

- - -..... _ - -~-- ....- -

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 Proposed Question: SRO Question # 1 5 The plant is at rated power.

Then:

  • SLCS LINE A TEMP locally reads 69° F and lowering.
  • SLCS LINE B TEMP locally reads 80° F and steady.
  • SLC TANK TEMP locally reads 71° F and lowering.
  • SLC solution TEMP locally reads 71 ° F and lowering.

Followed by an investigation that indicates:

  • SLC Tank mixing and operating heaters have lost power.

Which one of the following describes the status of the SLC system and action(s) required?

A. The TS LCO for SLC does NOT apply because SLC solution temperature is NOT below the TS requirement. However, actions to restore the power supply to the mixing heater, the operating heater and heat trace must be taken.

B. The TS LCO for SLC must be entered because the SLCS line "A" temperature is less than the TS requirement. Actions to restore the power supply to the operating heater and heat trace must be taken.

C. The TS LCO for SLC does NOT apply because ONLY the SLCS Line "An temperature is less than the TS requirement. BOTH SLC lines must be affected similarly before SLC is considered inoperable. Actions to restore the power supply to the operating heater and heat trace must be

taken.

D. The TS LCO for SLC must be entered because the SLC solution AND SLCS line "A" temperature is less than the TS requirement. Actions to restore the power supply to the operating heater and heat trace must be taken.

Proposed Answer: B Explanation (Optional):

A. Incorrect - The TS limit is <::70°F for the tank OR the heat tracing circuit as indicated by suction line temperature <::70°F. Therefore, the TS LCO applies.

B. Correct - lAW TS SR 4.1.5 (3), The heat trace circuit and the other heaters are powered from NON-IE power lAW NOH01 SLCSYSC-02 page 35.

C. Incorrect - SLC is inoperable lAW SR 4.1.5 (3) and heaters are powered from NON-IE supplies D. Incorrect - The tank solution temperature is not < the TS requirement of 70 degrees. The heaters are powered from NON-IE supplies.

TS 3.1.5 SR 4.1.5 Technical Reference(s): (Attach if not previously provided)

NOH01 SLCSYSC-02 Page 35 Proposed References to be provided to applicants during examination: TIS 3.1.5 TECSPECPCE010 - Given specific plant operating conditions and a copy of HCGS Learning Objective: (As available)

TS, evaluate plant/system operability and plant actions to be taken Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 1 KIA # 271000 A2.03 Importance Rating 3.8

.~ ..................- -

Ability to (a) predict the impacts of the following on the OFFGAS SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Main steam line high radiation Proposed Question: SRO Question # 16 The plant is at 75% power with power ascension in progress.

Then:

  • OHA C6-A3, MN STM LINE RADIATION HI, alarms
  • OHA C6-C1, RADIATION MONITORING ALARMITRBL, alarms Which one of the following describes the status of the Offgas System and action(s) required?

A. The Offgas System would be isolated. lAW AB-RPV-0008, Reactor Coolant Activity, reduce reactor power as necessary to clear the MN STM LINE RADIATION HI alarm.

B. The Offgas System would NOT be isolated. lAW AB-RPV-0008, Reactor Coolant Activity, reduce reactor power as necessary to clear the MI\I STM LINE RADIATION HI alarm.

C. The Offgas System would be isolated. lAW AB.CONT-0004, Radioactivity Release, reduce reactor power as necessary to clear the RADIATION MONITORING ALARMITRBL alarm.

D. The Offgas System would NOT be isolated. lAW AB.CONT-0004, Radioactivity Release, reduce reactor power as necessary to clear the RADIATION MONITORING ALARMITRBL alarm.

Proposed Answer: B Explanation (Optional):

A. Incorrect -There is no isolation of the offgas system due to high Main Steam Line radiation levels. There are offgas isolations based on other offgas parameters exceeding their setpoints.

B. Correct The offgas system would remain in service. lAW AB-RPV-0008, Reactor Coolant Activity, Step B.1, reduce reactor power as necessary to clear the MN STM LINE RADIATION HI alarm.

C. Incorrect There is no isolation of the offgas system due to high Main Steam Line radiation levels. There are offgas isolations based on other offgas parameters exceeding their setpoints.

There is NO direction in AB.CONT-0004, Radioactivity Release, to reduce reactor power as necessary to clear the RADIATION MONITORING ALARMrrRBL alarm.

D. Incorrect - There is NO direction in AB.CONT-0004, Radioactivity Release, to reduce reactor power as necessary to clear the RADIATION MONITORING ALARMrrRBL alarm.

AB-RPV-0008, Step B.1.

Technical Reference(s): (Attach if not previously provided)

AB. CO NT-0004 Proposed References to be provided to applicants durin~, examination: None ABRPV8E001 - recognize abnormal indications/alarms and/or procedural Learning Objective: (As available) requirements for implementing Reactor Coolant Activity Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New x Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier# 2 Group # 2 KIA # 226001 2.4.6 Importance Rating 4.7

c--

Emergency Procedures / Plan: Knowledge of EOP mitigation strategies (RHRlLPCI: Containment Spray System Mode)

Proposed Question: SRO Question # 17 Due to a large reactor coolant leak:

  • Drywell sprays have been initiated
  • Drywell pressure is 10 psig and lowering
  • Drywell temperature is 310 OF and lowering If lowering drywell pressure and temperature lowering results in entering the UNSAFE region of the Drywell Spray Initiation Limit (DWT-P), what action is required?

A. lAW EOP-102 Primary Containment Control, Drywell Temperature leg, Secure all drywelJ sprays when in the UNSAFE region of the DrywelJ Spray Initiation Limit curve.

B. lAW EOP-102 Primary Containment Control, DrywelJ Pressure leg, Secure drywell sprays at 9.5 psig drywell pressure and initiate suppression chamber sprays.

C. lAW EOP-202, Emergency Depressurize the reactor.

D. lAW EOP-102, Primary Containment Control, Continue drywell sprays until drywell pressure approaches 0 psig.

Proposed Answer: D Explanation (Optional):

A. Incorrect - Curve is based on spray initiation, not securing spray B. Incorrect - DW sprays remain in service until 0 psig. There is no indication in the stem that the RHR pump is needed for adequate core cooling. No direction to secure drywell sprays and initiate suppression chamber sprays.

C. Incorrect With OW pressure lowering, an EO is not warranted.

O. Correct - this is a Retainment Override action, step PCC-1 Technical Reference{s): EOP-102 and EOP-202 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: none EOP1 02E01 0 - Given plant conditions and access to the following curves, determine Learning Objective: the region of acceptable operation and (As available) explain the bases for the curve lAW Primary Containment Control Question Source: Bank # NRC 2007 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility: Hope Creek Vendor: GE Exam Date: 2/2011 Exam Type: S Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 KIA # 290001 2.2.37 Importance Rating 4.6 Equipment Control: Ability to determine operability and lor availability of safety related equipment.

(Secondary Containment)

Proposed Question: SRO Question # 18 Given:

  • Refueling and OPDRV activities are in progress.

Then:

  • Steam Vent Blowout Panel 1AS224 indicates open on the RM-11.
  • Visual observation confirms that 1AS224 is NOT fully closed.

Which one of the following describes whether Secondary Containment is operable and the operational impact of this equipment problem?

Secondary Containment is _(1 With the Blowout Panel not fully closed, operationally _(2)_

A. (1) Operable until Reactor Building vacuum is <0.25 inches of vacuum water gauge.

(2) Reactor BUilding Llp will degrade and continue to degrade. The Main Steam Tunnel Coolers will be unaffected.

B. (1) Operable until Reactor Building vacuum is <0.25 inches of vacuum water gauge.

(2) Reactor Building Llp will degrade and continue to degrade and the Main Steam Tunnel Coolers will have degraded cooling capability.

C. (1) NOT Operable.

(2) ALL fuel movement must be suspended. OPDRVs may continue.

D. (1) NOT Operable.

(2) ALL fuel movement must be suspended AND operations with a potential for draining the reactor vessel must be suspended.

Proposed Answer: D Explanation (Optional):

A. Incorrect - Secondary Containment is Inoperable RB t.p will initially degrade when the panel is open but will then return to its normal setpoint.

B.

Incorrect - Secondary Containment is Inoperable. The steam tunnel coolers are in a different location.

C. Incorrect - ONLY movement of recently irradiated fuel must be suspended. For example, moving new fuel in the spent fuel pool would be permitted.

D. Correct. The Steam Vent Blowout Panel is part of secondary containment and is verified in place using the RM-11 and surveillance HC.OP-ST.ZZ-0003. The failure of the blowout panel results in a TS LCO entry 3.6.5.1.B. for secondary containment integrity. With it inoperable, In Operational Condition *, suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.

Op Con * - When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

TS Surveillance 4.6.5.1.b.1 and LCO Technical Reference(s): 3.6.5.1.a (Attach if not previously provided)

HC.OP-ST.ZZ-0003 Proposed References to be provided to applicants durin!~ examination: none TECSPECPCE010 Given specific plant operating conditions and a copy of HCGS Learning Objective: (As available)

TS, evaluate planUsystem operability and plant actions to be taken NRC 2010 slightly Question Source: Bank #

modified Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2010 slightly modified Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier# 3 Group # 1 KIA # G1 2.1.35 Importance Rating 3.9 Conduct of Operations: Knowledge of the fuel-handling responsibilities of SRO's.

Proposed Question: SRO Question # 19 Given:

  • Core Alterations are in progress.
  • A fuel bundle move to core location 21-38 has been completed.
  • As the refueling platform is returning to the fuel pool the spotter determines that the fuel assembly was supposed to have been placed in core location 17-38.

lAW NF-AA-310, Special Nuclear Material And Core Component Movement, what is the proper sequence for the required actions?

(1) CONTACT the Reactor Engineer / SNMC for further instructions.

(2) DOCUMENT the condition in the Corrective Action Program.

(3) REPORT the condition to Control Room Supervision.

(4) STOP all fuel movement.

A. (3), (1), (4), (2)

B. (2), (4), (1), (3)

C. (1), (2), (3), (4)

D. (4), (3), (1), (2)

Proposed Answer: D Explanation (Optional):

A.

Incorrect - Sequence is not correct

B.

Incorrect - Sequence is not correct C.

Incorrect - Sequence is not correct D. Correct - Per NF-AA-310, Special Nuclear Material And Core Component Movement, step 4.4.7, the required actions are:

1. STOP all fuel movement,
2. REPORT the condition to Control Room Supervision,
3. CONTACT the Reactor Engineer I SNMC for further instructions,
4. DOCUMENT the condition in the Corrective Action Program.

Technical Reference(s): NF-AA-310, Pg 12 Sect 4.4.7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None ADMPROE003 - Describe the conditions under which performance of a procedure Learning Objective: (As available) should be stopped and the necessary subsequent actions Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier# 3 Group #

KJA# G2 2.2.18 Importance Rating 3.9 Equipment Control: Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Proposed Question: SRO Question # 20 The plant is in OPCON 3, restoring "S" RHR Pump from an outage.

  • An experienced NEO is releasing a HPCI Work Clearance Document (WCD), installing fuses, in the Lower Relay Room.
  • A newly qualified NEO has been assigned to verify the release.
  • As the experienced NEO releases the WCD, he shows the newly qualified NEO the component locations.

Concerning this WCD release verification, which of the following actions will be required lAW HU-M 101, Human Performance Tools and Verification Practices, prior to restoring "S" RHR to operability?

A. The NEO that observes the release can use documented Peer Checks to verify the release of all components, signing for each component as he observes its restoration.

B. The NEO that observed the release must perform an Independent Verification AFTER the NEO who performed the WCD release is finished.

C. The NEO that observed the release can use Concurrent Verification to satisfy the verification requirements as the release is being performed.

D. The NEO that observed the release CANNOT perform an Independent Verification. An Independent Verification of the WCD must be performed by a different person.

Proposed Answer: D

Explanation (Optional):

A Incorrect An IV or CV is required, NOT a documented peer check.

B. Incorrect The second NEO is no longer independent.

C. Incorrect - A CV cannot be led and must independently identify the component.

D. Correct -lAW HU-AA-101 4.3, Independent Verification (IV) or Concurrent Verification (CV) shall be performed for safety related equipment when returning these systems to their final configuration prior to considering them operable. The definition of IV states " ... a second qualified individual who has not witnessed the activity that established the desired condition performs the Independent Verification." Since the second NEO observed the restoration, he cannot perform an IV. Step 4.3.3.3.2 for Concurrent Verification states "The performer and verifier independently identify the component. .. " Since the first NEO led the second NEO, the independent identification of the component had been compromised.

Technical Reference(s): HU-AA-101 4.3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None ADMPR01 03CE002 - Determine how to Learning Objective: (As available) perform an independent verification Question Source: Bank# ID: Q59785 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier # 3 Group #

KIA # G3 2.3.6 Importance Rating Radiation Control: Ability to approve release permits.

Proposed Question: SRO Question # 21 Given:

  • The plant is in Operational Condition 3 - Hot Shutdown, proceeding to Cold Shutdown.
  • The forced outage is due excessive unidentified RCS leakage.
  • Reactor pressure is 920 psig.
  • Drywell O 2 concentration is 2.5%.

Which of the following requirements must be met prior to purging the Primary Containment?

A. Reactor power is required to be ,:::15% per TIS. 3.6.1.1 Primary Containment Integrity.

B. The plant must be in Operational Condition 4 Cold Shutdown per TIS, 3.6.1.1 Primary Containment Integrity.

C. A Drywell walk down must be completed lAW HC.OP-10.ZZ-0004, Shutdown from Rated Power to Cold Shutdown.

D. A Valve Open Time permit must be prepared lAW OP-HC-103-105, Administrative Control of Containment Atmosphere Control Valve Open Time.

Proposed Answer: 0 Explanation (Optional):

A. Incorrect There is no power requirement to be met prior to venting, there is a TIS limit on The total time these valves are open for Purging or Venting.

B. Incorrect There is no requirement to be in Shutdown prior to venting, The TIS states that without Primary Containment Integrity, restore Primary Containment Integrity within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In this case with the valves operable Primary Containment is still met.

C. Incorrect There is no requirement to walk down the Drywell prior to venting.

D. Correct - Radiation Protection will determine if a Gaseous Effluent Permit is required for the pending valve manipulation. The Gaseous permit must include a valve open time permit to track

  1. of hours that purge valves are open lAW OP-HC-103-105. The total time these valves are open for Purging or Venting, shall not exceed 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> for any rolling 365 day period. This is to maintain compliance with TIS 3.6.1.8, TIS 4.6.1.8.1 HC.oP-SO-GS.0001 (Q)

Technical Reference(s): (Attach if not previously provided)

OP-HC-103-105 Proposed References to be provided to applicants during examination: None Learning Objective: INERTOE015 (As available)

Question Source: Bank # ID: Q55956 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 KIA # G4 2.4.40 Importance Rating 4.5 Emergency Procedures / Plan: Knowledge of the SRO's responsibilities in emergency plan implementation.

Proposed Question: SRO Question # 22 Both Hope Creek and the Salem Units are operating at full power when:

  • Security reports an attack on the Hope Creek million gallon diesel fuel oil tank.
  • Armed intruders have entered the Hope Creek Protected Area.
  • You are the Shift Manager.

In accordance with NC.EP-EP .ZZ-01 02, Emergency Coordinator Response which one of the following prompt actions is required?

A. Immediately scram the plant, then direct the CRS to contact the Salem plant and determine their plant status.

B. Contact the Salem Shift Manager and determine their security situation, if they are under attack, ONLY Hope Creek will be scrammed.

C. Contact the Salem Shift Manager and decide on the units Trip/Scram sequence. Notify the CRS of Trip/Scram sequence. Scram Hope Creek first.

D. Immediately direct the CRS to scram the plant and implement the appropriate abnormal procedure, then contact the Salem plant and decide on the overall security status.

Proposed Answer: C Explanation (Optional):

A. Incorrect - The SM must contact the other station's SM and decide on units Trip/Scram sequence.

B. Incorrect - Both plants will be scrammed even if Salem was NOT under attack. Also if either Salem Unit was under attack they would require scramming the Units.

C. Correct - In accordance with NC.EP-EP.lZ-0102, Emergency Coordinator Response, IF in the Shift Managers judgment an actual attack to OCNPA property is occurring then CONTACT other station's SM and decide on units Trip/Scram sequence, INFORM CRS of Trip/Scram sequence and unless there is a Salem specific attack in progress, Hope Creek should be the first unit ScrammedlTripped.

D. Incorrect - The SM must contact the other station's SM and decide on units Trip/Scram sequence.

Technical Reference{s): NC.EP-EP.ZZ-0102 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: EP-EP.lZ-102 Learning Objective: SECAB-0001-01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier # 3 Group #

KIA # G1 2.1.23 Importance Rating 4.4 Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Proposed Question: SRO Question # 23 Given:

  • The plant is operating at 100% power.
  • OHA E6-C5 "RBVS & WING AREA HVAC PNL 10C382" alarms.
  • The Reactor Operator reports Reactor Building Differential Pressure is negative at 0.25 inches water gauge.

Which one of the following actions is required?

A. Place FRVS in service lAW HC.OP-SO.GU-0001 FRVS Operation.

B. Start another Reactor Building Supply fan lAW HC.OP-AB.ZZ-0001 Transient Plant Conditions.

C. Isolate RBVS Isolation Dampers lAW HC.OP-AB.CONT-0003 Reactor Building.

D. Start another Reactor Building Supply fan lAW HC.OP-SO.GR-0001 Reactor Building Ventilation.

Proposed Answer: A Explanation (Optional):

A. Correct - Place FRVS in service lAW HC.OP-SO.GU-0001 FRVS Operation. FRVS is required by AB-CONT-0003 because RB L:lP is less than the required 0.30" WC.

B. Incorrect - Start another Reactor Building Supply fan HC.OP-AB.zZ-0001 Transient Conditions.

AB-ZZ- 0001 does not provide direction for starting RBVS.

C. Incorrect - Isolate RBVS Isolation Dampers lAW HC.OP-AB.CONT-0003 Reactor Building, abnormal states to verify OPEN dampers not closed D. Incorrect - Start another Reactor Building Supply fan lAW HC.OP-SO.GR-0001 Reactor Building Ventilation. Starting a supply fan would aggravate the problem.

Technical Reference(s): HC.OP-AB.CONT-0003 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: ABCNT3E004 (As available)

Question Source: Bank # 80659 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 KIA # G2 2.2.7 Importance Rating 3.6 Equipment Control: Knowledge of the process for conducting special or infrequent tests.

Proposed Question: SRO Question # 24 The plant is planning to install a new flow control system for HPCI during the next outage.

In accordance with LS-AA-104, 50.59 Review Process, a 10CFR 50.59 Evaluation would determine if the new flow control system requires _(1 )_. The administrative process to perform initial acceptance testing of HPCI with the new flow control system is controlled by _(2)_.

A. (1 ) ONLY NRC notification prior to implementation (2) OP-AA-1 03-1 03, OPERATION OF PLANT EQUIPMENT B. (1) ONLY NRC notification prior to implementation (2) OP-AA-108-110, EVALUATION OF SPECIAL TESTS OR EVOLUTIONS C. (1) NRC approval prior to implementation (2) OP-AA-103-103, OPERATION OF PLANT EQUIPMENT D. (1) NRC approval prior to implementation (2) OP-AA-108-110, EVALUATION OF SPECIAL TESTS OR EVOLUTIONS Proposed Answer: D Explanation (Optional):

A. Incorrect - NRC approval is the purpose, not notification. The purpose of the OP-AA-103-103 procedure is to provide clear poliCies regarding who is authorized to manipulate or operate plant equipment.

B. Incorrect NRC approval is the purpose, not notification C. Incorrect - The purpose of the OP-AA-103-103 procedure is to provide clear policies regarding who is authorized to manipulate or operate plant equipment.

D. Correct lAW LS-AA-104 - Purpose - "This procedure establishes the requirements for preparing, reviewing, approving, and documenting evaluations performed pursuant to the requirements of 10 CFR 50.59 "Changes, tests, and experiments," for determining if a facility or procedure change, test, or experiment requires NRC approval prior to implementation.

Special tests or evolutions are defined in AU 1 of OP-AA-108-110. Specifically examples of special evolutions include: Appropriate portions of plant startup after an outage that involves significant change to plant systems, equipment or procedure related to the core, reactivity control, or reactor protection.

Additionally, evolutions that require the use of special tests in conjunction with existing procedures may also be classified as special evolutions, including Complex Modification Function Testing Effects equipment essential to ECCS Operability:

Any activity that may prevent the actuation of, or ability of the Core Spray system, Low Pressure Coolant Injection system, High Pressure Coolant Injection System, or the Automatic Depressurization System from performing their ECCS function.

Any activity that does not fit the above specific criteria, but involves significant deviation from normal operations.

Additional justification for identifying this as a special test or evolution are found in Att 2 of OP AA-1 08-11 O.

Technical Reference(s): OP-AA-1 08-11 O. AU 1 & 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None NOH04ADA 104C State the purpose Learning Objective: (As available) of the 50.59 review process Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 3 Comments:

Facility Hope Creek Vendor GE Exam Date 2/2011 Exam Type S Examination Outline Cross-reference: Level RO SRO Tier # 3 Group #

KIA # G3 2.3.4 Importance Rating 3.7 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Proposed Question: SRO Question # 25 Given:

  • An equipment operator with specific skills must enter a locked door high radiation area to isolate a leaking valve.
  • Their current dose for the year is 1,725 mRem.
  • It will be necessary to raise the operators' administrative dose control level to 3,000 mRem.
  • The operator's exposure records are complete and up to date.

Who must approve this dose limit extension?

A. Shift Manager ONLY.

B. Radiation Protection Manager ONLY.

C. Radiation Protection Manager and the Shift Manager.

D. Radiation Protection Manager and the Plant Manager.

Proposed Answer: C Explanation (Optional):

A. Incorrect - Approval is required by the Radiation Protection Manager and the work group supervisor.

B. Incorrect - Approval is required by the Radiation Protection Manager and the work group supervisor.

C. Correct - To raise the ADCL to 3000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manager and the work group supervisor.

Per RP-AA-203 4.2.5. "To raise the ADCL to 3000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manager and the work group supervisor."

D. Incorrect - Approval is required by the Radiation Protection Manager and the work group supervisor.

Technical Reference(s): RP-AA-203, Step 4.2.5, pg 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: NOH04ADM024C-04, (As available)

Question Source: Bank#

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:

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