ML15127A507

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2015-04-FINAL Written Exam
ML15127A507
Person / Time
Site: Cooper Entergy icon.png
Issue date: 04/21/2015
From: Vincent Gaddy
Operations Branch IV
To:
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Download: ML15127A507 (119)
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Text

ANSWER KEY

1. A 26. A 51. C 76. D
2. A 27. C 52. A 77. A
3. C 28. C 53. C 78. A
4. B 29. B 54. A 79. D
5. A 30. C 55. D 80. A
6. D 31. D 56. A 81. C
7. C 32. C 57. C 82. B
8. B 33. A 58. C 83. B
9. A 34. B 59. B 84. C
10. C 35. C 60. B 85. D
11. B 36. B 61. A 86. A
12. B 37. D 62. D 87. B
13. D 38. B 63. C (Deleted) 88. B
14. A 39. C 64. C 89. D
15. C 40. D 65. C 90. A
16. A 41. D 66. A 91. A
17. C 42. A 67. D 92. D
18. D 43. A 68. B 93. A
19. A 44. C 69. D 94. B
20. C 45. B 70. A 95. C
21. D 46. A 71. B 96. C
22. A 47. A or C 72. D 97. B
23. A 48. A 73. B 98. C
24. C 49. D 74. B 99. B
25. B 50. B 75. D 100. B

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name:

Date: 04/21/2015 Facility/Unit: Cooper Nuclear Station Region: I II III IV Reactor Type: W CE BW GE Start Time: Finish Time:

Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.

Applicant Certification All work done on this examination is my own. I have neither given nor received aid.

Applicants Signature Results RO/SRO-Only/Total Examination Values 75 / 25 / 100 Points Applicants Score / / Points Applicants Grade / / Percent

Question: 1 With the Reactor initially operating at 72% power the following conditions exist:

  • Reactor Recirculation MG A trips on motor generator high air temperature.
  • PMIS shows the plant is operating in the Stability Exclusion Region (red region) of the Power to Flow map.
  • SRM period swings with a fluctuation time of <3 seconds.

Why does Procedure 2.4RR (Reactor Recirculation Abnormal) require a scram?

A. To prevent fuel damage challenges due to high power density.

B. To stop RPV water level oscillations due to reactor power swings.

C. To provide the fastest way to exit the area of operation while in single loop.

D. To ensure the reactor is shut down before exceeding the RPV pressure limit.

Question: 2 With the plant operating at 100% power, a loss of offsite power occurs.

  • Both Diesel Generators fail to start and CANNOT be started.
  • The Supplemental Diesel Generator is NOT available.
  • HPCI and RCIC recover RPV water level to +35 inches (Narrow Range).

What operational restriction applies to the continued use of HPCI in response to this event, until onsite or offsite electrical power is restored IAW 5.3SBO (Station Blackout)?

HPCI must be secured and remain off...

A. after one cycle of operation.

B. after two cycles of operation.

C. just prior to the division's battery being exhausted.

D. no later than 15 minutes from the time injection flow was reduced.

Question: 3 The plant is operating at power with the following conditions:

  • Breaker 1FE red indicating light is illuminated.
  • Breaker 1GE red indicating light is illuminated.
  • 4160V buses A, C, and E indicating lights are off.

What is causing the above conditions?

A. Panel BB1 has a blown fuse.

B. Panel BB3 has a blown fuse.

C. Panel AA1 has a blown fuse.

D. Panel AA3 has a blown fuse.

Question: 4 Reactor power is 35% during a startup.

Main Turbine bearing vibrations are as follows:

  • Bearing 5 vibration rises rapidly to 15 mils and steadies out.
  • Vibration on bearings 4 and 6 are 7 mils and rising.

What action(s) is/are required IAW 2.4TURB (Main Turbine Abnormal)?

A. Trip the Main Turbine ONLY.

B. Scram the Reactor AND trip the Main Turbine.

C. Lower reactor power until bearing vibration lowers to <14 mils.

D. Suspend the startup to allow raising the turbine casing temperature.

Question: 5 The plant is operating at 100% power on the 221st day of continuous operation when all outboard MSIVs go closed.

  • At time T=0, the reactor automatically scrams.
  • At time T=3 seconds, reactor pressure spikes to 1100 psig.
  • At time T=10 minutes, reactor pressure is in its expected band.

What is the status of the SRVs at T= 10 minutes?

A. 1 SRV is cycling.

B. No SRVs are open or cycling.

C. 1 SRV is open and another SRV is cycling.

D. 2 SRVs are open and another SRV is cycling.

Question: 6 Following a toxic gas event requiring the control room to be abandoned, the following conditions exist:

  • BOTH Low-Low Set (LLS) valves are cycling.

The ADS ISOLATION switch in the Alternate Shutdown (ASD) Room is now placed in ISOLATE.

Which LLS valve is able to be controlled from the ASD room?

Which LLS valve continues to cycle automatically?

Controlled from LLS Valve that ASD Room Continues to Cycle Automatically A. RV-71D RV-71F B RV-71F RV-71F C. RV-71D RV-71D D. RV-71F RV-71D

Question: 7 The plant is operating at rated power.

  • A loss of all REC pumps occurs.
  • 5.2REC, LOSS of REC is entered.
  • All attempts to restore REC have failed.
  • The reactor is manually scrammed.

Why is the running CRD pump required to be secured IAW 5.2REC (Loss of REC)?

A. Prevent reactor vessel stratification.

B. Prevent reactor water level overfill.

C. Prevent pump bearing overheating.

D. Prevent pump motor bearing overheating.

Question: 8 The plant is operating near rated power.

  • A station operator has reported that a leak has developed in the Augmented Radwaste Building basement air system.
  • Instrument air header pressure has lowered to 75 psig and has stabilized.
  • A reactor scram has been inserted.

What action is required IAW Procedure 5.2AIR (Loss of Instrument Air)?

A. Close IA-SOV-21, Drywell Instrument Air Supply Valve.

B. Close IA-MO-80, Non Critical Instrument Air Isolation Valve.

C. Open SA-MO-81, Service Air to Instrument Air Crosstie Valve.

D. Ensure SA-AO-PCV-609, Service Air System Isolation Valve is open.

Question: 9 The plant is shutdown for refueling with the following conditions:

  • The reactor has been shutdown for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
  • RPV water level is being maintained 50 to 60 inches.
  • RHR HX inlet temperature indicates 100ºF.

What is MINIMUM time for the reactor coolant to reach 212ºF if shutdown cooling is lost?

A. 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> B. 2.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> C. 4.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> D. 13.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

Question: 10 The plant is in day 12 of a refueling outage.

  • Refueling operations are in progress.

What is the required MINIMUM water level above the top of the RPV flange IAW TS 3.9.6 (Refueling Operations)?

A. 6 feet B. 12 feet C. 21 feet D. 37 feet

Question: 11 During an ATWS, a LOCA occurs with the following conditions present:

  • EOP 6A and 7A have been entered.
  • Average drywell temperature is 210°F and steady.
  • Drywell pressure is 19 psig and rising.
  • Average suppression pool temperature is 205°F and rising.
  • Torus water level is 14 feet and rising slowly.
  • Reactor pressure is 1000 psig and steady.

Why is an Emergency Depressurization required at this time IAW EOP 3A (Primary Containment Control)?

To prevent A. chugging, and possible loss of the pressure suppression function.

B. raising torus pressure above the PCPL A which may result in failure of containment.

C. excessive torus to drywell vacuum breaker operation and possible vacuum breaker failure.

D. the torus water level rise which may cause loss of containment on SRV actuation due to a water column in the system discharge piping.

Question: 12 The plant is operating at power with HPCI inoperable due to maintenance on its Auxiliary Lube Oil Pump.

An inadvertent PCIS Group 1 isolation occurs.

What system is capable of maintaining adequate core cooling IAW the EOPs as LLS actuates?

A. SLC B. RCIC C. MC/RF D. Core Spray

Question: 13 The plant is experiencing an ATWS with the following conditions:

  • Reactor Power is 15% and steady.
  • Suppression Pool temperature is 96°F and rising.

Why is boron injection required to be initiated before average suppression pool water temperature reaches 140° F IAW EOP 6A {Reactor Power (Failure to Scram)}?

A. Prevents exceeding the 25% peak-to-peak periodic neutron flux oscillations.

B. Prevents violating Technical Specification Limit for Suppression Pool Temperature.

C. Ensures the reactor will be shutdown under all conditions before the suppression pool is heated beyond its design limits.

D. Ensures the reactor will be shutdown under hot-standby conditions before the suppression pool reaches the Heat Capacity Temperature Limit.

Question: 14 Drywell spray has been placed in service during a LOCA due to high Drywell temperature.

  • Drywell temperature and pressure are lowering.

When is Drywell spray required to be stopped IAW EOP 3A (Primary Containment Control)?

Drywell spray is REQUIRED to be stopped before A. drywell pressure lowers to zero psig.

B. the suppression chamber to drywell vacuum breakers open.

C. the reactor building to suppression chamber vacuum breakers open.

D. drywell temperature and pressure lower to the UNSAFE (Red) region of the Drywell Spray Initiation Limit (DWISL) curve.

Question: 15 Following a LOCA, the following conditions are present:

  • RPV water level above TAF and slowly rising
  • Torus pressure is 4.5 psig (stable).
  • Only RHR pump C is injecting at 6500 gpm.
  • Average torus water temperature is 135°F and rising.

What Torus average water temperature is RHR pump C flow FIRST required to be reduced to ensure compliance with NPSH limits?

A. 140°F B. 165°F C. 185°F D. 210°F

Question: 16 RPV water level lowers to the value requiring entry into EOP-1A (RPV Control) and remains steady at that level.

No other EOP entry conditions are satisfied.

What group isolation(s) automatically occur(s)?

A. Group 2 only.

B. Group 3 only.

C. Groups 3 and 6 only.

D. Groups 2, 3, and 6.

Question: 17 The plant is operating at full power when the following occurs:

  • All MSIVs close resulting in a Reactor Scram with 27 rods failing to fully insert.
  • RPV water level is intentionally lowered due to level/power conditions being met.

When is a normal reactor cooldown FIRST allowed to commence IAW EOP 6A {RPV Pressure (Failure to Scram)}?

A. When all APRMs indicate downscale.

B. When Hot Shutdown Boron Weight is injected.

C. When Cold Shutdown Boron Weight is injected.

D. When one control rod is at 48 and all other control rods are at position 02.

Question: 18 The plant is operating at rated power when the following annunciator is received:

MAIN STM LINE PANEL/WINDOW:

HIGH RAD 9-4-1/A-5 1 minute later the following annunciator is received:

OFFGAS PANEL/WINDOW:

HIGH RAD 9-4-1/C-5 What automatic action minimizes off-site release rates if the ERP, MSL and OG radiation monitors continue rising?

What procedure(s) is/are required to be entered due to these alarms?

A. MSIV closure 2.4OG (Off-Gas Abnormal) ONLY.

B. MSIV closure 2.4OG (Off-Gas Abnormal) AND 5.2FUEL (Fuel Failure).

C. Offgas isolation 2.4OG (Off-Gas Abnormal) ONLY.

D. Offgas isolation 2.4OG (Off-Gas Abnormal) AND 5.2FUEL (Fuel Failure).

Question: 19 The plant is operating at 100% when the Shift Manager directs a Control Room evacuation due to a fire in the control room.

Why are all the AC powered Reactor Feedwater Pump lube oil pump control switches placed in PULL-TO-LOCK IAW Procedure 5.4FIRE-S/D (Fire Induced Shutdown from Outside Control Room)?

To ensure A. a reactor water overfill event is prevented.

B. reactor water level is intentionally lowered to aid in FW preheating.

C. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump operation.

D. automatic start of DC lube oil pumps to maintain RFP bearing lubrication during pump coast down.

Question: 20 With the plant operating at 70% power the following conditions exist:

  • NPPD System frequency is 59.8 hertz and steady.

Is the cause of the disturbance due to CNS or the Grid?

What is the required position of the GEN VOLTAGE REGULATOR switch under these conditions IAW Procedure 5.3GRID (Degraded Grid Voltage)?

A. The Grid OFF B. The Grid ON C. CNS OFF D. CNS ON

Question: 21 The Plant is operating at full power with the following conditions:

  • Alarm B-1/B-3 TG LOW VACUUM PRE TRIP sounds.
  • Main Generator load is 805 MWe.

(1) Where is the alarm setpoint validated?

(2) What action is directed by B-1/B-3 TG LOW VACUUM PRE TRIP, when vacuum CANNOT be maintained 23" Hg?

A. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.

(2) Perform a rapid power reduction.

B. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.

(2) Scram the reactor and trip the turbine.

C. (1) CONDENSER PRESSURE TRIP GRAPH on RVLC HMI.

(2) Perform a rapid power reduction.

D. (1) CONDENSER PRESSURE TRIP GRAPH on DEH HMI.

(2) Scram the reactor and trip the turbine.

Question: 22 The plant is operating at 100% power.

  • The operator takes manual control of RPV pressure using the SRVs.

Why are the SRVs alternated while controlling pressure IAW Procedure 2.2.1 (Nuclear Pressure Relief System)?

To prevent A. localized torus overheating.

B. SRV failure due to excessive cycling.

C. SRV failure due to excessive valve temperature.

D. inaccurate average torus temperature indication.

Question: 23 A plant startup is in progress with pressure set at 926 psig and reactor power at 8%.

  • CRD Pump A trips and cannot be restarted.
  • All attempts by the crew to start CRD Pump B are unsuccessful.

What is the consequence to the Control Rod Drive Mechanism (CRDM) if operation continues with these conditions?

A. Reduced drive piston seal life.

B. Inability to scram control rods.

C. Cooling water orifice plugging.

D. Unlatching of the collet fingers.

Question: 24 The plant is operating at 100% power when an un-isolable RCIC steam line failure results in high radiation in the secondary containment.

What is the reason for inserting a reactor scram before secondary containment radiation levels reach the Max Safe value IAW EOP-5A (Secondary Containment Control)?

A. To preclude flooding and subsequent unmonitored release from the NE Quadrant.

B. To prevent ANY component environmental qualification (EQ) limit from being exceeded.

C. To reduce the driving head of the primary system discharging into secondary containment.

D. To ensure the reactor is shut down before radiation release rates exceed the values for a General Emergency.

Question: 25 The following Reactor Building Ventilation parameters are observed while operating at power:

  • Exhaust Rad Channel A 11 mr/hr
  • Exhaust Rad Channel B 6 mr/hr
  • Exhaust Rad Channel C 5 mr/hr
  • Exhaust Rad Channel D 10 mr/hr
  • Reactor Building DP -0.20 wg NO plant system responses occur.

Which of the following is the initial required Control Room Operator action for Secondary Containment Ventilation systems IAW Procedure 2.1.22 (Recovering from a Group Isolation)?

A. Start ONE SGT train ONLY.

B. Start BOTH SGT trains.

C. Start the Standby RB supply fan.

D. Start the Standby RB exhaust fan.

Question: 26 The plant is conducting refueling operations involving the movement of recently irradiated fuel assemblies in the secondary containment.

  • A fuel bundle is dropped on the top of the core.
  • A Group 6 isolation occurs due to bundle damage.
  • Calm winds are indicated on the SPDS Weather Display.
  • HV-MO-262 MG SET-1A INLET and HV-AO-263 MG SET-1A INLET fail to fully isolate.
  • Reactor Building Average DP indicates -0.04 wg and stable.

Which of the following identifies the actual Off-Site radiological release rate response?

A. ERP release rate rises.

B. ERP release rate lowers then rises.

C. Reactor Building monitored release rate rises.

D. Reactor Building unmonitored release rate rises.

Question: 27 Which of the following completes the statement below?

EOP 5.8.21 (PC Venting and Hydrogen Control) has a CAUTION alerting the operator(s) that a combination of 6% H2 and (1) will produce (2) .

A. (1) 5% air (2) a combustible mixture B. (1) 5% air (2) an explosive mixture C. (1) 5% O2 (2) a combustible mixture D. (1) 5% O2 (2) an explosive mixture

Question: 28 With the plant operating at rated power a LOCA occurs.

The following conditions exist:

  • Reactor pressure is 600 psig and lowering.
  • Drywell pressure is 5.2 psig and rising slowly.
  • Torus pressure is 4.0 psig and rising slowly.

When does LPCI injection flow into the RPV FIRST occur for the listed reactor pressures below?

A. 550 psig B. 435 psig C. 220 psig D. 105 psig

Question: 29 Residual Heat Removal (RHR) loop A is aligned for shutdown cooling mode of operation with RHR pump A operating. The following conditions exist:

  • The vessel head is tensioned.
  • SDC system flow is 5000 gpm.
  • Temperatures are logged as follows:

RHR-TR-131, CH 9 NBI-TR-89, CH 9 (reactor NBI-TR-89, CH6 (reactor TIME (RHR HX inlet temp) vessel metal temp) vessel flange temp) 0100 210ºF 222ºF 332ºF 0115 188ºF 196ºF 331ºF 0130 166ºF 173ºF 331ºF 0145 143ºF 147ºF 330ºF What RHR action is required and why IAW Procedure 2.2.69.2 (RHR System Shutdown Operations)?

A. Throttle CLOSED RHR-MO-27A (Inboard Injection Valve) to reduce the cooldown rate.

B. Throttle OPEN RHR-MO-66 (RHR Heat Exchanger Bypass Valve) to reduce the cooldown rate.

C. Throttle CLOSED RHR-MO-27A (Inboard Injection Valve) to ensure accurate temperature indication at the RHR HX Inlet.

D. Throttle OPEN RHR-MO-66 (RHR Heat Exchanger Bypass Valve) to ensure accurate temperature indication at the RHR HX Inlet.

Question: 30 The plant is operating at 10% of rated power.

HPCI-MO-15, STM SUPP INBD ISOL VLV is found closed and cannot be opened with its control switch.

HPCI is declared Inoperable.

What Technical Specification ACTION is required?

A. Enter Technical Specification 3.0.3 immediately.

B. Verify all ADS SRVs are Operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Verify RCIC system is Operable by administrative means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Isolate HPCI by deactivating HPCI-MO-15 and HPCI-MO-16 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Question: 31 What is the alternate suction source for Core Spray Pump A and when is it aligned to this source?

A. ECST, to raise torus level in Modes 1/2/3.

B. CST A, to raise torus level in Modes 1/2/3.

C. ECST, for core reflood capability when the torus is drained in Modes 4/5.

D. CST A, for core reflood capability when the torus is drained in Modes 4/5.

Question: 32 The plant is operating at near rated power when a loss of both RVLC/RFPT CORE switches causes RFPT control to transfer to MDEM.

How are the RFPs affected?

A. Speed lowers to idle speed.

B. Speed rises to upper automatic clamp.

C. Speed is held constant at current speed.

D. Speed lowers to minimum governor speed.

Question: 33 What is the power supply that is used to fire the A Standby Liquid Control (SLC) squib valve?

A. MCC K B. MCC S C. MCC M D. CPP

Question: 34 The reactor is operating at 100% rated power.

The RPS MG Set A motor supply fused disconnect fuse blows.

(1) What is the status of the RPS A Electrical Protection Assemblies (EPA)?

(2) What is the status of RPS?

A. (1) Two (2) RPS A EPA Breakers opened.

(2) Full reactor SCRAM.

B. (1) Two (2) RPS A EPA Breakers opened.

(2) 1/2 scram ONLY on RPSPP1A.

C. (1) Four (4) RPS A EPA Breakers opened.

(2) Full reactor SCRAM.

D. (1) Four (4) RPS A EPA Breakers opened.

(2) 1/2 scram ONLY on RPSPP1A.

Question: 35 IRM A detector is installed with twice the argon fill pressure of the other IRM detectors.

How does this affect the operation of the IRM A versus the others when subjected to the same neutron field?

IRM A High High trip is A. less conservative and the downscale rod block is less conservative.

B. less conservative and the downscale rod block is more conservative.

C. more conservative and the downscale rod block is less conservative.

D. more conservative and the downscale rod block is more conservative.

Question: 36 The plant is performing a reactor startup requiring the SRM Shorting Link Switches to be placed in the OPEN position for testing.

All Shorting Link Switches have been placed in OPEN, however while placing switches for the B Channel to OPEN, the contacts remain CLOSED (switch position changes while switch contacts do not change state).

What is the impact of this RPS Logic malfunction if SRM A reaches the Hi Hi setpoint?

A. A full Reactor Scram occurs.

B. Only a half scram from "A" RPS trip system occurs.

C. Only a half scram from "B" RPS trip system occurs.

D. A full Reactor Scram occurs only if a second SRM trip signal is received.

Question: 37 The reactor is operating at 15% rated power when the following occurs:

  • IRM "B" fails upscale.

Then, before any operator action is taken:

  • APRMs "B" and "E" both fail downscale.

What is/are the minimum action(s) that will allow all rod block and/or scram signals to be cleared?

A. Bypass IRM "B" only.

B. Bypass APRM "E" only.

C. Bypass APRM "B" and IRM "B".

D. Bypass APRM "B" and APRM "E".

Question: 38 Given the following conditions:

  • RCIC is injecting to the reactor at 400 gpm.
  • RCIC suction is from the ECST.
  • Torus temperature is 125°F and steady.
  • ECST level is 25 inches and lowering.
  • Torus level is 12 9 and rising.

(1) If these conditions persist, how is RCIC impacted?

(2) What action is required IAW 9-4-1/F-2 (RCIC Suction Transfer)?

(1) (2)

RCIC Suction aligns to torus Action required A. at 23 ECST Level. Secure RCIC B. at 23 ECST Level. Makeup to the ECSTs C. at 13 1 Torus Level . Secure RCIC D. at 13 1 Torus Level . Makeup to the ECSTs

Question: 39 Which of the following completes the statement below with the plant operating at power?

The Safety Relief Valves associated with the ADS system are normally aligned with ____(1)____ to open and the accumulators associated with them are sized to allow a MINIMUM of ____(2)____ valve cycles at 70% of design drywell pressure.

A. (1) Nitrogen (2) 5 B. (1) Instrument Air (2) 5 C. (1) Nitrogen (2) 2 D. (1) Instrument Air (2) 2

Question: 40 A manual reactor Scram is inserted due to a transient while operating at 50% power.

The following conditions occur AFTER the Mitigating Task Scram Actions are complete:

  • RPV pressure lowered to 600 psig and is currently rising slowly.
  • RPV water level lowered to -24 inches and is currently rising slowly.
  • Drywell pressure rises to 1.2 psig and stabilizes.

The following valves currently indicate OPEN:

  • RHR-MO-920 {(Div 1), AOG STEAM SUPPLY VLV}

Which valve(s) is/are required to be CLOSED to ensure PCIS initiation is complete?

A. MSIVs B. RR-AO-741 C. RWCU-MO-15 D. RHR-MO-920

Question: 41 While performing Surveillance 6.ADS.201, ADS Manual Valve Actuation (IST) during startup following a refuel outage, the BOP Operator reports that SRV RV-71 E Control Switch has been placed to OPEN.

When conditions stabilize, which of the following indications validate that SRV RV-71 E is full open IAW 6.ADS.201?

A. Main Generator output lowers.

B. Total indicated steam flow rises.

C. PMIS temperatures within MAX T.

D. Bypass valves throttle in the closed direction.

Question: 42 What information must the Rod Worth Minimizer (RWM) receive from the Reactor Vessel Level Control System (RVLCS) in order to determine its mode of operation?

A. Total main steam flow and feedwater flow rates.

B. Total main steam flow and feedwater flow rate mismatch.

C. Time that total main steam flow and feedwater flow are above limits.

D. Reactor power level calculated from total main steam flow and feedwater flow rates.

Question: 43 An accident occurred that resulted in fuel failure and a breach of the reactor coolant system boundary.

SGT A train is in service to support Primary containment venting when the following alarms:

SGT A PANEL/WINDOW:

HIGH MOISTURE K-1/A-2 What SGT system component is primarily affected and how is the offsite release rate affected?

A. The Charcoal Filter and release rate rises primarily due to iodine activity.

B. The Charcoal Filter and release rate rises primarily due to particulate activity.

C. The High Efficiency Inlet Filter and the release rate rises primarily due to iodine activity.

D. The High Efficiency Inlet Filter and the release rate rises primarily due to particulate activity.

Question: 44 What is the normal power source to the Emergency Service Station Transformer?

A. The OPPD line via the T6 transformer.

B. The Auburn line via the T7 transformer.

C. The 161 KV substation via the T6 transformer.

D. The 161 KV substation via the T7 transformer.

Question: 45 The plant is in a normal full power electrical lineup. The following alarm is received:

NO BREAK PANEL/WINDOW:

INVERTER 1A C-4/E-7 VOLT FAILURE What is the current source of power to the NBPP?

A. MCC-L via a step down transformer and the inverter cabinet static switch.

B. MCC-R via a step down transformer and the inverter cabinet static switch.

C. MCC-L via a step down transformer and then directly to the power panel.

D. MCC-R via a step down transformer and then directly to the power panel.

Question: 46 The plant is operating at 100% power with the following conditions:

  • Battery charge in progress following the replacement of a cell in the Division II 250 VDC station battery.
  • A complete loss of Battery room ventilation occurs.

What is the PRIMARY operational concern for the present conditions?

A. Hydrogen buildup in the battery room is a fire hazard.

B. Hydrogen buildup in the battery room displaces oxygen.

C. Battery room temperature rise causes cell overheating and loss of electrolyte level.

D. Battery room temperature rise leads to cell reversal conditions in the new replacement battery cell.

Question: 47 DG1 is manually started for post maintenance testing:

  • DIESEL GEN 1 BKR EG1 is open.
  • Engine driven lube oil pump shaft completely shears.

What condition FIRST trips the Diesel Generator?

A. Low Lube Oil Pressure B. High Lube Oil Temperature C. Low Turbocharger Oil Pressure D. High Thrust Bearing Oil Temperature

Question: 48 The air compressors are operating with the Compressor Control Module (CCM) in LOCAL with the following lineup:

  • Air Compressor 1A is in Lead and running.
  • Air Compressor 1B is First Backup and is idle.
  • Air Compressor 1C is Second Backup and is idle.

Reactor Equipment Cooling (REC) to the air compressors is lost due to an REC pipe rupture.

What condition automatically trips Air Compressor 1A?

What automatically occurs following the trip of Air Compressor 1A?

A. High air temperature.

Air Compressor 1B starts and continuously supplies air loads.

B. High air temperature.

Air Compressor 1B starts but trips soon after it starts.

C. Low cooling water pressure.

Air Compressor 1B starts and continuously supplies air loads.

D. Low cooling water pressure.

Air Compressor 1B starts but trips soon after it starts.

Question: 49 The Unit is operating at 50% power with REC Pump A breaker tagged OPEN.

REC Heat Exchanger B and 3 REC pumps are in service.

  • One REC pump trips and is unavailable.
  • M-1/A-1, REC SYSTEM LOW PRESSURE alarms.

(1) What is the system response with no operator action?

(2) What Immediate Action(s) is/are required upon receipt of this alarm with system pressure at 60 psig IAW Procedure 5.2REC (Loss of REC)?

A. (1) Non-critical loads isolate after a 20 second time delay.

(2) Isolate Augmented Radwaste cooling ONLY.

B. (1) Non-critical loads isolate after a 20 second time delay.

(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.

C. (1) Non-critical loads isolate after a 40 second time delay.

(2) Isolate Augmented Radwaste cooling ONLY.

D. (1) Non-critical loads isolate after a 40 second time delay.

(2) Isolate Augmented Radwaste cooling and RWCU Non-Regen HX Inlet.

Question: 50 The plant operating at 100% rated power with SLC pump 1A out of service when an ATWS occurs:

  • Average torus temperature is just below BIIT and rising slowly.
  • RPV level remains at +35 inches.

SLC pump 1B is placed to START and the following conditions are present:

  • Reactor power is 40% and slowly lowering.
  • Boron is injecting with an initial tank level of 80%.

Assuming the SLC pump is operating at its design flow rate and RPV level remains steady, what is the approximate SLC tank level and status of the Average Power Range Monitor (APRM) downscales after 25 minutes?

SLC Tank Level APRM Downscale Alarm (9-5-1/C-8)

A. 54% OFF B. 54% ON C. 26% OFF D. 26% ON

Question: 51 A Plant startup is in progress with reactor power at or near the point of adding heat.

Reactor period is near infinity with the Reactor Mode Select switch in STARTUP / HOT STANDBY position.

IRM C is on Range 7 and indicates 35 on the Panel 9-5 Recorder, when its range switch is taken to Range 6.

What is the new indication for IRM C and what automatic action(s) occur(s)?

Recorder Indication Automatic Action A. 97 Rod Block Only.

B. 97 Rod Block and Half Scram.

C. 110 Rod Block Only.

D. 110 Rod Block and Half Scram.

Question: 52 Which one of the following is a Post Accident Monitoring (PAM) instrument?

A. Source Range Monitor (SRM)

B. Traversing In-core Probe (TIP)

C. Condensate Storage Tank Level Indicator D. Reactor Building Ventilation Exhaust Plenum Radiation Monitors

Question: 53 Following a plant transient the following conditions are present:

  • A steam leak develops in the RCIC Room.
  • RCIC Room area temperature is 195°F and rising.

What is the effect on RCIC?

A. ONLY RCIC-MO-16 (RCIC STEAM SUPPLY OUTBOARD ISOLATION VALVE) closes and the RCIC turbine trips.

B. ONLY RCIC-MO-15 (RCIC STEAM SUPPLY INBOARD ISOLATION VALVE) closes and the RCIC turbine coasts down (no trip).

C. BOTH RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine trips.

D. BOTH RCIC-MO-15 and RCIC-MO-16 (RCIC STEAM SUPPLY INBOARD and OUTBOARD ISOLATION VALVE) close and the RCIC turbine coasts down (no trip).

Question: 54 The Plant is performing a Reactor Startup.

What is the flow indication on Panel 9-5 CRD-FI-305 (Drive Water Flow) while the rod is withdrawing?

A. 2 gpm B. 4 gpm C. 6 gpm D. 8 gpm

Question: 55 The Plant is conducting a Reactor startup with the following conditions present:

  • Group 3 Rod Movement Sheet Insert Limit: 04
  • Group 3 Rod Movement Sheet Withdraw Limit: 08
  • When Control Rod 26-27 is withdrawn from 06 to 08, it double notches and settles with the Four Rod position indication being BLANK.

Which RWM IDT targets turn RED?

A. WITHDRAW BLOCK only.

B. WITHDRAW BLOCK and OUT OF SEQUENCE only.

C. WITHDRAW BLOCK, OUT OF SEQUENCE and SELECT ERROR only.

D. WITHDRAW BLOCK, INSERT BLOCK, OUT OF SEQUENCE and SELECT ERROR only.

Question: 56 Reactor power is steady at 26%.

What is the HIGHEST Core Flow which is a TS Safety Limit Violation?

CORE FLOW A. 9%

B. 10%

C. 11%

D. 13%

Question: 57 Where does the Reactor Water Cleanup System (RWCU) return water to the reactor?

Via the A. suction side of A reactor recirculation pump.

B. suction side of B reactor recirculation pump.

C. A feedwater line downstream of the outboard check valve.

D. B feedwater line downstream of the outboard check valve.

Question: 58 What supplies power to the switch contact logic of Wide Range Level Indicating Switch, NBI-LIS-57A (PCIS initiation logic)?

A. AA-1 B. CDP-1A C. RPSPP1A D. 24V Power Panel DC-A

Question: 59 The Plant is operating at power when a steam leak occurs in the Drywell.

  • RHR is spraying the torus and drywell.
  • Torus pressure is 11 psig lowering 1 psig/minute.
  • Drywell temperature is 245°F lowering slowly.

ALL redundant High Drywell pressure inputs (2 psig signal) fail to 1.5 psig.

What is the effect on drywell temperature?

Drywell temperature A. rises due to the loss of Drywell FCUs.

B. rises due to the Drywell Spray valves going full closed.

C. lowers due to the start of Drywell FCUs.

D. lowers due to the Drywell Spray valves going full open.

Question: 60 The plant is operating at 100% power when a small break LOCA occurs. The following conditions are present:

  • RHR Loop A is in Torus Spray and Suppression Pool Cooling.
  • Torus pressure is 2.3 psig and lowering slowly.

What is the HIGHEST flow which automatically opens RHR-MO-16A (LOOP A MIN FLOW BYP VLV) as RHR Loop A flow is being lowered?

A. 2500 gpm B. 2000 gpm C. 1500 gpm D. 1000 gpm

Question: 61 Which of the following identifies how main turbine equalizing header pressure is impacted as turbine load is raised from no load to full turbine load?

Main turbine equalizing header pressure will A. rise but less than reactor steam dome pressure.

B. remain constant at the main turbine pressure setpoint.

C. rise by the same amount as reactor steam dome pressure.

D. lower slightly as the main turbine governor valves throttle open.

Question: 62 The plant is operating at 95% power on the 100% rod line when the Non-Return Isolation Check Valve for Feedwater Heater 1-A-5 goes full closed due to a short in its motor operator.

(1) How does feed water temperature respond?

(2) Where does operation stabilize on the power-to-flow map?

(1) (2)

Feedwater temperature 100%Rod Line A. Lowers and returns to near the previous temperature. Below B. Stabilizes at a lower temperature ONLY. Below C. Lowers and returns to near the previous temperature. Above D. Stabilizes at a lower temperature ONLY. Above

Question: 63 The plant is operating at 90% power.

What is the impact on Main Steam steady state temperature if a control room operator opens a Safety Relief Valve (SRV)?

Main Steam steady state temperature will A. rise as the result of the turbine governor valves throttling closed thereby raising steam pressure and temperature.

B. rise as the result of reactor power rising.

C. lower as the result of the turbine pressure control system offset lowering steam pressure and temperature.

D. remain constant because the turbine pressure control system offset will compensate for the open SRV.

Question: 64 Maintenance is being performed on RMP-RM-150B (OFFGAS RAD MONITOR B) while operating at rated power with the following annunciator in alarm:

OFFGAS RAD PANEL/WINDOW:

MON DOWNSCALE 9-4-1/D-4 OR INOP (1) What is the impact on the Off-Gas system if radiation levels rise causing RMP-RM-150A (OFF GAS RAD MONITOR A) to reach the Hi Hi Trip Setpoint?

(2) What action is required following the Off-Gas System isolation IAW Procedure 2.4OG (Off-Gas Abnormal)?

A. (1) Off-Gas isolates IMMEDIATELY.

(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).

B. (1) Off-Gas isolates IMMEDIATELY.

(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).

C. (1) Off-Gas isolates after a 15 minute time delay.

(2) SCRAM and enter Procedure 2.1.5 (Reactor Scram).

D. (1) Off-Gas isolates after a 15 minute time delay.

(2) Lower reactor power per Procedure 2.1.10 (Station Power Changes).

Question: 65 What automatically starts the Essential Control Building Ventilation system?

A. Battery room exhaust fan trip.

B. High Computer room temperature.

C. High Critical Switchgear room temperature.

D. Non-essential Control Building supply fan trip.

Question: 66 With the plant at power, a manual valve is closed in order to maintain Primary Containment OPERABLE IAW TS LCO 3.6.1.3 (PCIVs).

The valve is required to be opened per CRS direction with a NLO stationed at the valve.

Which of the following completes the statements below regarding the actions required to maintain Administrative Control of this PCIV while open IAW Procedure 2.0.1 (Plant Operations Policy)?

Direct the NLO to establish ____(1)____ communication with the Control Room and to close the valve in event of an accident condition.

The instructions provided to the NLO ____(2)____ required to be documented in the Control Room log.

A. (1) continuous (2) are B. (1) continuous (2) are NOT C. (1) intermittent (2) are D. (1) intermittent (2) are NOT

Question: 67 The Reactor has just been declared CRITICAL during plant startup. In addition to:

  • Date/Time
  • Sequence
  • Reactor Pressure What additional criticality information is required to be logged in the Control Room Operators Log IAW Procedure 2.1.1 (Startup Procedure)?

A. SRM Counts AND IRM Overlap.

B. SRM Counts AND Moderator Temperature.

C. Reactor Period AND IRM Overlap.

D. Reactor Period AND Moderator Temperature.

Question: 68 Which of the following identifies the indication and location that confirms the reactor is in Hot Shutdown when the Mode Switch is in the Shutdown position during a reactor cooldown?

A. Feedwater Temperature indicates 200°F on Panel A.

B. RR Suction Temperature indicates 240°F on Panel 9-4.

C. RHR HX Inlet Temperature indicates 200°F on Panel 9-3.

D. Vessel Head Adjacent to Flange temperature indicates 240°F on Panel 9-21.

Question: 69 Which of the following identifies the FIRST two tagging order steps for placing a system pump under clearance, and the reason for this order IAW Procedure 0.9 (Tagout)?

OPEN the pump breaker and then close the A. suction valve to minimize draining time.

B. discharge valve to minimize draining time.

C. suction valve to protect low pressure components.

D. discharge valve to protect low pressure components.

Question: 70 An Operator is preparing to write a Clearance Order on a Solenoid Valve.

(1) Where can the Operator obtain Controlled Copies of the electrical and mechanical drawings needed to prepare the Clearance?

(2) When interpreting the electrical drawing, what is the status of Contact 3-4 when the switch is in the CLOSE position?

(1) (2)

A. Control Room Open B. Operations Support Center Open C. Control Room Closed D. Operations Support Center Closed

Question: 71 Which of the following is the FIRST accumulated dose value ABOVE WHICH a tour member is required to have an NRC Form 5 or equivalent issued IAW 9.ALARA.1 (Personnel Dosimetry and Occupational Radiation Exposure Program)?

A. 50 mrem B. 100 mrem C. 500 mrem D. 1000 mrem

Question: 72 What is/are the MINIMUM personnel monitoring requirement(s) for exiting a contaminated area in the Reactor Building and dressing in street clothing IAW 9.EN-RP-100 (Radiation Worker Expectations)?

A. Perform a whole body frisk with a frisker ONLY.

B. Perform a hand and foot frisk with a frisker ONLY.

C. Perform a whole body frisk with a frisker then a whole body contamination monitor scan using a PCM.

D. Perform a hand and foot frisk with a frisker then a whole body contamination monitor scan using a PCM.

Question: 73 An Operator has to enter a room to close a valve to stop a small water leak.

  • The affected work area general radiation level is 1600 mrem/hour.

What type of entry permit is required?

Is continuous RP coverage required during the entry?

Entry Permit Continuous RP Coverage A. SWP NOT Required B. SWP Required C. RWP Required D. RWP NOT Required

Question: 74 The plant is at power when a transient occurs.

  • Multiple Black and Yellow outlined annunciators are in alarm.

Which colored alarm takes precedence and why IAW Procedure 2.3.1 (General Alarm Procedure)?

A. BLACK; an EOP entry is required ONLY.

B. YELLOW; a plant shutdown condition may be present ONLY.

C. BLACK; BOTH an EOP entry is required AND a plant shutdown condition may be present.

D. YELLOW; BOTH an EOP entry is required AND a plant shutdown condition may be present.

Question: 75 There are operational circumstances when operators must perform immediate operator actions without reference to procedures.

Which statement represents one of those circumstances?

A. When an EOP directs performing the action.

B. When Technical Specifications direct a scram.

C. When an Alarm procedure directs performing the action.

D. When an Abnormal procedure directs performing the action.

Question: 76 The Main Turbine trips while operating at 31% Power.

Two minutes later, ALL Mitigating Task Scram Actions per 2.1.5 (Reactor Scram) are completed with the following indications reported:

  • Reactor Pressure is steady at 985 psig.

(1) What is the current reactor power based upon Main Turbine Bypass valve status?

(2) What is the HIGHEST Emergency Action Level (EAL) required to be declared IAW Procedure 5.7.1 (Emergency Classification)?

A. (1) Between 10% and 15%.

(2) Alert B. (1) Between 20% and 25%.

(2) Alert C. (1) Between 10% and 15%.

(2) Site Area Emergency D. (1) Between 20% and 25%.

(2) Site Area Emergency

Question: 77 The plant is operating at 100% power when the following conditions occur:

  • SW Pressure on both Divisions has risen but is still in the green band.
  • REC system pressure is steady and is in the green band.
  • REC Surge Tank Level High alarms.
  • RR MG Set Oil Temperatures are rising.
  • Drywell temperature and pressure are rising.
  • RWCU F/D Inlet Temp High alarms.

Which of the following identifies the cause of these conditions and the required action to correct the problem IAW Procedure 2.0.1.2 (Operations Procedure Policy)?

The REC Heat Exchanger A. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).

B. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2REC (Loss of REC).

C. SW Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).

D. REC Outlet valve has closed, shift REC heat exchangers IAW 5.2SW (Service Water Casualties).

Question: 78 The plant is operating at near rated power when a fire in the Control Building Basement causes a loss of all air compressors. The fire is extinguished but the air compressors are unable to be restarted.

The damage in the basement is limited to air system components.

How will reactor level and pressure have to be controlled?

A. HPCI and RCIC for level control.

HPCI/RCIC and/or SRVs for pressure control.

B. HPCI and RCIC for level control.

Manual control of Main Turbine Bypass valves for pressure control.

C. Condensate and Feed with RFP minimum flow valves manually closed for level control.

Manual control of Main Turbine Bypass valves for pressure control.

D. Condensate and Feed with RFP minimum flow valves manually closed for level control.

HPCI/RCIC and/or SRVs for pressure control.

Question: 79 The plant is in MODE 5 with refueling in progress when an irradiated fuel bundle is dropped over the core.

The Refueling Floor ARM (RA-1) is 5.5 x 104 mR/hr and rising.

The ERP Kaman is indicating 1.80E +06 µCi/sec and rising.

(1) What is the current MINIMUM required Emergency Classification?

(2) What is the LOWEST ERP Kaman reading requiring escalation to the next higher classification?

A. (1) Unusual Event (2) 3.70E+06 Ci/sec B. (1) Unusual Event (2) 3.70E+07 Ci/sec C. (1) Alert (2) 3.70E+06 Ci/sec D. (1) Alert (2) 3.70E+07 Ci/sec

Question: 80 While operating at 80% power, the following indications are observed:

  • Drywell Temperature is 148°F and rising 1°F/5 min.
  • Drywell Pressure is 0.35 psig and rising slowly (0.05 psig/min).

What action is required to mitigate these conditions?

Ensure all available DW FCU control switches are in A. RUN IAW 2.4PC (Primary Containment Control).

B. RUN IAW H-1/A-2 (Drywell Zone 1 High Temp).

C. OVERRIDE IAW H-1/A-2 (Drywell Zone 1 High Temp).

D. OVERRIDE IAW 2.4PC (Primary Containment Control).

Question: 81 The plant is operating at rated power with the Startup Transformer out of service for maintenance.

The Main Turbine trips due to loss of vacuum.

(1) What power source automatically reenergizes the Critical Busses?

(2) What action is required IAW 5.3EMPWR (Emergency Power During MODES 1, 2, or 3)?

A. (1) Diesel Generators (2) Direct DCC to perform the CNS-Black Plant Procedure.

B. (1) Diesel Generators (2) Coordinate with DCC to backfeed through the Normal Transformer.

C. (1) Emergency Transformer (2) Direct DCC to perform the CNS-Black Plant Procedure.

D. (1) Emergency Transformer (2) Coordinate with DCC to backfeed through the Normal Transformer.

Question: 82 While the Fire Brigade is combatting a Temporary Trailer fire (south of the Training Building) with fire hoses, the following indications are observed on the Main Control Room Fire Protection Panel:

G R G R G R ELECTRIC FIRE DIESEL FIRE ELECTRIC FIRE PUMP C PUMP D PUMP E AUTO AUTO AUTO PULL TO PULL TO LOCK LOCK The operator then places the control switch for Fire Pump E to START and the pump starts.

(1) What is the known MINIMUM pressure reached in the fire protection header prior to the operator taking the above action?

(2) What is the operability status of the Fire Suppression Water System?

A. (1) 68 psig (2) Operable B. (1) 68 psig (2) Inoperable C. (1) 141 psig (2) Operable D. (1) 141 psig (2) Inoperable

Question: 83 (1) Which of the following is the LOWEST Reactor Steam Dome Pressure requiring entry into TS LCO 3.4.10, Reactor Steam Dome Pressure?

(2) What is the TS Bases for the Steam Dome Pressure of LCO 3.4.10?

A. (1) 1025 psig (2) To ensure ASME Code limits for the reactor coolant pressure boundary are not exceeded.

B. (1) 1025 psig (2) To ensure the assumptions of the overpressure protection analysis are conserved.

C. (1) 1050 psig (2) To ensure ASME Code limits for the reactor coolant pressure boundary are not exceeded.

D. (1) 1050 psig (2) To ensure the assumptions of the overpressure protection analysis are conserved.

Question: 84 A loss of Division 1 RPSPP1A power occurs while operating at power. Troubleshooting the problem is about to commence.

What is the response of the RWCU PCIV(s)?

Which procedure directs how to control the configuration of the RWCU PCIV(s) during troubleshooting?

A. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.

(2) 0.31(Equipment Status Control).

B. (1) ONLY RWCU-MO-15 (INBOARD ISOLATION VALVE) closes.

(2) 0-CNS-WM-102 (Work Implementation and Closeout).

C. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.

(2) 0.31(Equipment Status Control).

D. (1) BOTH RWCU-MO-15 and RWCU-MO-18 (OUTBOARD ISOLATION VALVE) close.

(2) 0-CNS-WM-102 (Work Implementation and Closeout).

Question: 85 The plant is in Mode 1 and annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD alarms due to a valid signal.

(1) What is the LOWEST radiation level which causes this alarm to actuate?

(2) What is the TS Bases for the allowable value of this instrument setpoint?

A. (1) 5 mR/hr (2) Detect a steam leak in Secondary Containment.

B. (1) 5 mR/hr (2) Detect gross fuel cladding failure.

C. (1) 10 mR/hr (2) Detect a steam leak in Secondary Containment.

D. (1) 10 mR/hr (2) Detect gross fuel cladding failure.

Question: 86 The following conditions exist during a large break LOCA from rated power:

  • No Off-Site is power available.
  • DG1 is unavailable.
  • RHR Pump D is unavailable.
  • Reactor Building is inaccessible.
  • RPV Pressure is 35 psig and steady.

(1) Which RHR Loop is available for LPCI injection?

(2) What action is procedurally required if additional injection is required to assure adequate core cooling?

A. (1) Loop A (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.

B. (1) Loop A (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.

C. (1) Loop B (2) Enter 5.3ALT-STRATEGY (Alternate Core Cooling Mitigating Strategies) and inject using Fire Protection to RHR.

D. (1) Loop B (2) Enter 5.3EMPWR (Emergency Power During MODES 1, 2, OR 3) and use the Supplemental Diesel Generator (SDG) to re-energize 4160V 1F Bus.

Question: 87 The plant is in Mode 5 with all OPERABLE Control Rods (Core Cell contains fuel) fully inserted.

Procedure 4.5 (Reactor Protection/Alternate Rod Insertion Systems), Section 7 (Bypass Reactor Mode Switch - Shutdown Position Scram and transfer MODE switch), directs the following Step:

7.3.2.2 Inform Shift Manager that Reactor Mode Switch - Shutdown Position Scram (LCO 3.3.1.1, Function 10) is inoperable due to being bypassed.

a. Enter applicable Conditions and Required Actions for following LCOs as required:
1. LCO 3.3.1.1.
2. LCO 3.10.3.
3. LCO 3.10.4.

What is/are the MINIMUM Required Action(s) for this step IAW TS LCO 3.3.1.1 (RPS Instrumentation) if an OPERABLE Control Rod is fully withdrawn with the DIV 1 Reactor Mode Switch Scram bypass jumper installed?

Place the bypassed channel in trip...

A. within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ONLY.

B. OR Insert the Control Rod within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. OR Insert the Control Rod within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. OR Insert the Control Rod within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Question: 88 A reactor startup is in progress with the following conditions present:

  • Reactor power is below the point of adding heat.

The following annunciator alarms and clears multiple times within 30 seconds and repeats for 3 minutes due to IRM A spiking.

IRM RPS CH A PANEL/WINDOW:

UPSCALE TRIP 9-5-1/D-7 OR INOP (1) What is the result of IRM A spiking?

(2) What is the MOST Restrictive required Technical Specification action?

A. (1) Rod block ONLY (2) Place Channel or Associated Trip system in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. (1) Rod block and 1/2 scram (2) Place Channel or Associated Trip system in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. (1) Rod block ONLY (2) Place Channel in one trip system OR one trip system in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. (1) Rod block and 1/2 scram (2) Place Channel in one trip system OR one trip system in trip within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Question: 89 The Plant is operating at rated power on March 10th.

At 1300 the RO observes the following:

  • RFC-LI-94A indicates 36 inches and is slowly rising.
  • RFC-LI-94B indicates 32 inches and is slowly lowering.
  • RFC-LI-94C indicates 37 inches and is slowly rising.

If these conditions continue, what is the LATEST time to exit the Mode of Applicability IAW TS 3.3.2.2, Feedwater and Main Turbine High Water Level Trip?

A. March 10th at 1500 B. March 10th at 1900 C. March 17th at 1300 D. March 17th at 1700

Question: 90 The plant is in Mode 3.

An event requires IMMEDIATELY de-energizing 4160V Bus 1A.

(1) What is the impact of IMMEDIATELY de-energizing 4160V Bus 1A?

(2) How many Off Site Power Circuits are Inoperable IMMEDIATELY following power transfer?

A. (1) A momentary loss of Bus 1F occurs.

(2) One B. (1) A momentary loss of Bus 1F occurs.

(2) Two C. (1) A momentary loss of Bus 1G occurs.

(2) One D. (1) A momentary loss of Bus 1G occurs.

(2) Two

Question: 91 6.RWM.301 (Rod Worth Minimizer Functional Test For Startup) is being performed.

Which of the following completes the statements below regarding how the Select Error function of the RWM is tested IAW 6.RWM.301 AND the significance of maintaining the RWM operable during startup IAW TS Bases?

The operator is required to verify the SELECT ERROR indicator turns red while selecting ____(1)____

from each RWM group (except Group 1).

The RWM enforces compliance with BPWS which ensures that the initial conditions of the analysis for a

____(2)____ is NOT violated.

A. (1) a single rod (2) control rod drop accident B. (1) a single rod (2) single control rod withdrawal error C. (1) ALL individual rods (2) control rod drop accident D. (1) ALL individual rods (2) single control rod withdrawal error

Question: 92 The plant is operating at 97% power when RFP A trips.

(1) How is the Recirculation Flow Control System affected if RPV level lowers to 25 inches on NR Level instruments?

(2) If RRMG B speed does not change and CANNOT be reduced, what Attachment provides direction to trip RRMG B with RPV level stable at 20 inches on NR Level instruments IAW 2.4RR (Reactor Recirculation Abnormal)?

A. (1) Both Recirculation Pumps run back to 22% speed.

(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.

B. (1) Both Recirculation Pumps run back to 22% speed.

(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).

C. (1) Both Recirculation Pumps run back towards 45% speed.

(2) Attachment 1 {Trip of Reactor Recirculation Pump(s)}.

D. (1) Both Recirculation Pumps run back towards 45% speed.

(2) Attachment 4 (Reactor Recirculation Flow Control Failure/RRMG Scoop Tube Lockout).

Question: 93 The following conditions exist following an earthquake:

  • SRVs cannot to be opened for pressure control.
  • No means of RPV injection are available.
  • RPV water level is -150 inches corrected fuel zone and steady.
  • One MSL remains open.
  • RPV pressure is 950 psig.
  • Torus water level is 9.5 feet and lowering 0.2 feet/minute.

(1) What is the required CRS response?

(2) What CAUTION is of concern when carrying out this response?

A. (1) Transition to EOP 2A and direct Emergency Depressurization.

(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.

B. (1) Transition to EOP 2A and direct Emergency Depressurization.

(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.

C. (1) Transition to EOP 2A and direct Steam Cooling.

(2) When placing SJAEs in service, caution of sending personnel through potentially high radiation areas must be addressed.

D. (1) Transition to EOP 2A and direct Steam Cooling.

(2) When placing SJAEs in service, caution of forcing the shutter slide on the breaker could cause damage to the breaker preventing closure.

Question: 94 Which of the following completes the statements below regarding SRO License requirements per CNS procedures?

A MINIMUM of ____(1)____ 12-hour shifts under instruction is required to allow taking the watch after a four month absence from watchstanding.

A Special Prescription Respirator Glasses Verification is required IAW ____(2)____?

A. (1) four (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)

B. (1) four (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)

C. (1) five (2) NTP8.1 (Administration of Licensed Operator Medical Examination Program)

D. (1) five (2) 2.0.7 (Licensed Operator Active/Reactivation/Medical Status Maintenance Program)

Question: 95 Which activity REQUIRES Refuel Floor Supervisor permission during refueling operations in Mode 5?

A. Allowing under vessel access.

B. Allowing access to the refuel floor.

C. Re-commencing fuel handling operations.

D. Using greater than 50 gallons of demineralized water on the refuel floor.

Question: 96 The plant is operating at power in Mode 1.

What is the MAXIMUM time a Temporary Alteration In Support of Maintenance (TASM) can be installed on plant equipment WITHOUT performing a 10CFR50.59 Review IAW Procedure 3.4.4 (Temporary Configuration Change)?

A. 30 days B. 60 days C. 90 days D. 120 days

Question: 97 Which of the following identifies an item required to be addressed by a SRO performing a Work Order Standard Plant Impact Statement IAW 0-CNS-WM-102 (Work Implementation and Closeout)?

A. Verifying the Work Instructions are accurate.

B. Identification of single valve isolation requirements.

C. Verification that correct parts have been procured and are available.

D. Identifying required walkdowns by Shop/Responsible Work Center.

Question: 98 Which of the following completes the statement below regarding access to Locked High Radiation Areas (LHRA) IAW 9.EN-RP-101 (Access Control For Radiologically Controlled Areas)?

The ____(1)____ is required to give the final approval and the ____(2)____ shall perform the pre-job brief for personnel entering a LHRA with general area dose rates > 1.5 Rem/hr in the actual work area.

A. (1) Operations Shift Manager (2) Radiation Protection Supervisor B. (1) Operations Shift Manager (2) Dose Assessor C. (1) Radiation Protection Manager (2) Radiation Protection Supervisor D. (1) Radiation Protection Manager (2) Dose Assessor

Question: 99 Given the following:

  • At 1200 the threshold for a NOUE is exceeded.
  • At 1210 the Emergency Director declared classification of a NOUE.
  • At 1215 the threshold for an ALERT is exceeded.
  • At 1220 the Emergency Director declared classification of an ALERT.

What is the LATEST time that the State/Local agency notifications of the NOUE classification is required to be performed IAW EPIP 5.7.6 (Notification)?

A. 1215 B. 1225 C. 1230 D. 1235

Question 100 and associated references redacted due to SUNSI considerations

NOTE 3 Torus overpressure is sum of torus pressure and hydrostatic head above suction strainer Torus pressure (psig) ______________

Hydrostatic head (psig)

PC water level (ft.) ____________

Strainer level (ft.) -4 0.43 x +

____________ = _____________

Torus overpressure (psig) ______________

ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY Figure 1 - TIME TO BOILING - WATER LEVEL AT HIGH LEVEL TRIP PROCEDURE 2.4SDC REVISION 14 PAGE 15 OF 24

ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY Figure 2 - TIME TO BOILING - WATER LEVEL AT FLANGE PROCEDURE 2.4SDC REVISION 14 PAGE 16 OF 24

ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY Figure 3 - TIME TO BOILING - WATER TO LEVEL FLOODED TO 1001' PROCEDURE 2.4SDC REVISION 14 PAGE 17 OF 24

ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY Figure 4 - TIME TO CORE UNCOVERY PROCEDURE 2.4SDC REVISION 14 PAGE 18 OF 24

ATTACHMENT 5 TIME TO CORE BOILING/TIME TO CORE UNCOVERY Figure 5 - TIME TO CORE UNCOVERY PROCEDURE 2.4SDC REVISION 14 PAGE 19 OF 24

INFORMATION ONLY RP$ Instrumentation

- 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable. trip.

OR A.2 Place associated trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system in trip.

B. One or more Functions B.l Place channel in one 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with one or more trip system in trip.

required channels inoperable in both OR trip systems.

B.2 Place one trip system 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in trip.

c. One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability.

capability not maintained.

{continued)

Cooper 3.3-1 Amendment No. 178 INFORMATION ONLY

INFORMATION ONLY RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D. 1 Enter the Condition Immediately associated Completion referenced in Table Time of Condition A, B, 3.3.1.1-1 for the channel.

or C not met.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in Table core cells containing one or 3.3.1.1-1. more fuel assemblies.

Amendment 231 3.3-2 07/16/08 INFORMATION ONLY

INFORMATION ONLY RPS Instrumentation 3.3.1.1 Table.3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEillANCE AU. OWABLE FUNCTION CONDITIONS SYSTEM ACTIOND.1 REQUIREMENTS VALUE

7. Scram Discharge Volume Water Level- High
a. Level Transmitter 1,2 2 G SR 3.3.1 1.4 _:: 90 inches SR 3.3.1 . 1.9 SR 3.3.1 .1.12(a,b)

SR 3 3.1 .1.13 SR 3.3. 1.1.15 s<cl 2 H SR33.1.1 .4 .::90 inches SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.15

b. level Switch 1,2 2 G SR 3.3.1.1.4  ;:: 90 inches SR 3.3.1 .1.9 SR 3.3.1.1. 12 SR 3.3.1.1.13 SR 3.3.11.15 2 H SR 3.3.11.4  :::90 inches SR 3.3.1.1.9 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3.3.1.1.15
8. Turbine Stop ~29.5% RTP 2 E SR 3.3.1.1.4 ~ 10% closed Valve - Closure SR 3.3.1.1 .9 SR 3.3.1.1.12 SR 3.3.1 .1.13 SR 3.3.1.1.14 SR 3.3.1.1.15
9. Turbine Control Valve ~29.5% RTP 2 E SR 3.3.1.1.4 ~ 1018 psig Fast Closure, DEH SR 3.3.1.1.9 ( b)

Trip Oil SR 3.3.1.1.12 a, Pressure - Low SR 3.3.1 ,1.13 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode 1,2 G SR 3.3.1.1 .11 NA Switch - Shutdown SR 3.3.1.1.13 Position s<cl H SR 3.3.1.1.11 NA SR 3.3.1.1.13 11 Manual Scram 1,2 G SR 3.3.1.1 .9 NA SR 3.3.1.1.13 5(c} H SR 3.3.1.1.9 NA SR 3.3.1 .1.13 (a) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before retuming the channel to seJVice.

(b} The instrument channel setpoint shall be reset to a value that Is with in lhe as-left tolerance around the limiting Trip Seipoint (lTSP} at the completion of the suJVeillance; otlleiWise. the channel shall be declared inoperable. Setpoints more conservative than the LTSP are acceptable provided that the as-found a11d as-left tolerances apply to the actual setpoint implemented In the Surveillance procedures (Nominal Trip Selpo1nt) to confirm channel performance. The Limiting Trip Setpoint and the methodologies used to determine the as-found and the as-left tolerances are specified in the Technical Requirements Manual.

(c) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Cooper 3.3-8 Amendment No. 242 INFORMATION ONLY

INFORMATION ONLY Feedwater and Main Turbine High Water Level Tr p Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation LCO 3.3.2.2 Three channels of feedwater and main turbine high water level trip instrumentation shall be OPERABLE.

APPLICABILITY: THERMAL POWER ~ 25% RTP.

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One feedwater and main A.l Place channe 1 in 7 days turbine high water trip.

level trip channel inoperable.

B. Two or more feedwater B.l Restore feedwater and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and main turbine high main turbine high water level trip water level trip channels inoperable. capability.

c. Required Action and C.l Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to < 25% RTP.

Time not met.

Cooper 3.3-20 Amendment No. 178 INFORMATION ONLY

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